Division of Industrial and Medical Nuclear Safety
Office of Nuclear Material Safety and Safeguards
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

Comments On Draft Report

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Any interested party may submit comments on this report for consideration by the NRC staff. Please specify the report number, draft NUREG-1556, Vol. 9, in your comments, and send them by October 30, 1998 to:

Chief, Rules Review and Directive Branch
Office of Administration
Mail Stop T-6 D59
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

Comments may be accompanied by additional relevant information or supporting data. The staff specifically requests comments on (1) whether information on any other technical methods and models for risk-informed, performance-based applications not covered in the report exist and should be reviewed and included, and (2) risk-informed, performance-based applications beyond those discussed in the report that would provide regulatory focus on risk significant issues, and flexibility to licensees in implementing NRC safety objectives.

You may also provide comments via the Comment Form or at the NRC Web Site, http://www.nrc.gov. See the link under "Technical Reports in the NUREG Series" on the "Reference Library" page. Instructions for sending comments electronically are included with the document, NUREG-1556, Vol. 9, at the web site.

For any questions about the material in this report, please contact:

Sally L. Merchant
Mail Stop T-8 F5
U. S. Nuclear Regulatory Commission
Washington, D.C. 20555
Phone: (301) 415-7874
E-mail: slm2@nrc.gov

Abstract

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This draft guide has been developed in parallel with the proposed revision of 10 CFR Part 35, "Medical Use of Byproduct Material." Comments received in response to publication of this draft NUREG will be considered in developing the final guide. Development of the final NUREG will continue to parallel the rulemaking, and result in a guidance document that is consistent with the final rule.

As part of its redesign of the materials licensing process, the United States Nuclear Regulatory Commission (NRC) is consolidating and updating numerous guidance documents into a single comprehensive repository as described in NUREG-1539, "Methodology and Findings of the NRC's Materials Licensing Process Redesign," dated April 1996, and draft NUREG-1541, "Process and Design for Consolidating and Updating Materials Licensing Guidance," dated April 1996. Draft NUREG-1556, Vol. 9, "Consolidated Guidance about Materials Licenses: Program-Specific Guidance About Medical Use Licenses," dated August 1998, is the ninth program-specific guidance document developed for the new process and is intended for use by applicants, licensees, and NRC staff and will also be available for use by Agreement States. This guidance document corresponds with the revision to the Code of Federal Regulations (CFR), Title10, Part 35, concurrently published in draft in August 1998. This document combines and supersedes the guidance previously found in Regulatory Guide (RG) 10.8, Revision 2, "Guide for the Preparation of Applications for Medical Use Programs"; Appendix X to RG 10.8, Revision 2, "Guidance on Complying With New Part 20 Requirements"; Draft RG DG-0009, "Supplement to Regulatory Guide 10.8, Revision 2, Guide for the Preparation of Applications for Medical Use Programs"; Draft RG FC 414-4, "Guide for the Preparation of Applications for Licenses for Medical Teletherapy Programs"; Policy and Guidance Directive (P&GD) FC 87-2, "Standard Review Plan for License Applications for the Medical Use of Byproduct Material"; P&GD FC 86-4, Revision 1, "Information Required for Licensing Remote Afterloading Devices"; Addendum to Revision 1 to P&GD FC 86-4, "Information Required for Licensing Remote Afterloading Devices-Increased Source Possession Limits"; P&GD 3-15, "Standard Review Plan for Review of Quality Management Programs"; RG 8.39, "Release of Patients Administered Radioactive Materials"; RG 8.33, "Quality Management Program"; P&GD 3-17, "Review of Training and Experience Documentation Submitted by Proposed Physician User Applicants"; and RG 8.23, "Radiation Safety Surveys at Medical Institutions, Revision 1."

This draft report, where applicable, provides a more risk-informed, performance-based approach to medical use licensing consistent with the proposed regulations.

Figures

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Figure 2.1 U.S. Map 2-2

Figure 8.1 Location of Use 8-4

Figure 8.2 Decommissioning Records. 8-9

Figure 8.3 Financial Assurance Mechanisms 8-10

Figure 8.4 Gamma Stereotactic Radiosurgery Unit 8-15

Figure 8.5 Beta Radiation Catheter 8-17

Figure 8.6 Stent Implantation. 8-18

Figure 8.7 RSO Responsibilities 8-22

Figure 8.8 Facility Diagram for Nuclear Medicine Suite. 8-31

Figure 8.9 Iodine-131 NaI Administration for the Thyroid Carcinoma Patient 8-32

Figure 8.10 Overhead View of Manual Brachytherapy Patient Treatment Room 8-33

Figure 8.11 Teletherapy and HDR Treatment Room 8-34

Figure 8.12 Annual Dose Limits for Radiation Workers 8-49

Figure 8.13 Proper Security of Licensed Material 8-52

Figure 8.14 Material Receipt and Accountability 8-57

Figure 8.15 Leak Test Sample 8-61

Figure M.1 Diagram of Office and HDR Facility M-3

Foreword

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This draft guide has been developed in parallel with the proposed revision of 10 CFR Part 35, "Medical Use of Byproduct Material," which was published for comment in August 1998. It is distributed for comment to encourage public participation in its development. Comments received in response to publication of this draft NUREG will be considered in finalizing this NUREG. Finalization of the NUREG will continue to parallel the rulemaking, and result in a guidance document that is consistent with the final rule.

The NRC is using Business Process Redesign (BPR) techniques to redesign its materials licensing process. This effort is described in NUREG-1539, "Methodology and Findings of the NRC's Materials Licensing Process Redesign," dated April 1996. A critical element of the new process is consolidating and updating numerous guidance documents into a NUREG-series of reports. Below is a list of volumes currently included in the NUREG-1556 series:

Volume Number	Volume Title	Status
1	Program-Specific Guidance About Portable Gauge Licenses	Final Report
2	Program-Specific Guidance About Industrial Radiography Licenses	Draft for Use and Comment
3	Applications for Sealed Source and Device Evaluation and Registration	Draft for Comment
4	Program-Specific Guidance About Fixed Gauge Licenses	Draft for Comment
5	Program-Specific Guidance About Self- Shielded Irradiator Licenses	Draft for Comment
6	Program-Specific Guidance About 10 CFR Part 36 Irradiator Licenses	Draft for Comment
7	Program-Specific Guidance About Academic, Research and Development, and Other Licenses of Limited Scope	Draft for Comment
8	Program-Specific Guidance About Exempt Distribution Licenses	Final Report

The current document, draft NUREG-1556, Vol. 9, "Consolidated Guidance about Materials Licenses: Program-Specific Guidance About Medical Use Licenses," dated August 1998, is the ninth program-specific guidance document developed for the new process. It is intended for use by applicants, licensees, NRC license reviewers, and other NRC personnel. It combines and updates the guidance for applicants and licensees previously found in RG 10.8, Revision 2, "Guide for the Preparation of Applications for Medical Use Programs," dated August 1987, and the guidance for licensing staff previously found in P&GDs, draft RGs and Standard Review Plans. In addition, this draft report also contains pertinent information found in Information Notices (IN), as listed in Appendix A.

Since this draft report takes a risk-informed, performance-based approach to medical use licensing, it reduces the amount of information needed from an applicant seeking to possess and use relatively safe quantities of byproduct material.

A team composed of NRC and State Department of Health staff drafted this document, drawing on their collective experience in radiation safety in general and as specifically applied to medical use of byproduct material.

This draft report is strictly for public comment and is not for use in preparing or reviewing applications for medical use licenses until it is published in final form. It is being distributed for comment to encourage public participation in its development. NRC is requesting comments on the information provided about medical use of byproduct material, as well as comments on a risk-informed, performance-based approach to licensing. Please submit comments within 90 days of the draft report's publication. Comments received after that time will be considered if practicable.

Address comments to:

Secretary, U. S. Nuclear Regulatory Commission,
Washington, DC 20555
Attention: Rulemakings and Adjudications Staff.

Hand deliver comments to:

One White Flint North,
11555 Rockville Pike,
Rockville, Maryland
between 7:30 a.m. and 4:15 p.m. on Federal workdays.

Copies of comments received may be examined at: NRC Public Document Room, 2120 L Street, NW (Lower Level), Washington, DC.

Draft NUREG-1556, Vol. 9, is not a substitute for NRC regulations. The approaches and methods described in this draft report are provided for information and comment only.

Donald A. Cool, Acting Deputy Office Director, Office of Nuclear Material Safety and Safeguards

Acknowledgments

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The guidance development team thanks the individuals listed below for assisting in the development and review of the draft report. All participants provided valuable insights, observations, and recommendations.

The Participants

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Bolling, Lloyd A.
Cool, Donald A.
DelMedico, Joseph R.
Flack, Diane S.
Haney, Catherine
Hill, Thomas E.
Howard, Marcia
Jones, Samuel Z.
Lohaus, Paul H.
Lieberman, James
Merchant, Sally L.
Paperiello, Carl A.
Rothschild, Marjorie U.
Roe, Mary Louise
Siegel, M.D., Barry
Trottier, Cheryl A.
Treby, Stuart A.
Walter, David

The team also thanks Kay Avery, D. W. Benedict Llewellyn, Grace S. Lee, Steven W. Schawaroch, and Gina G. Thompson, of Computer Sciences Corporation for their assistance in the preparation of this document.

We would also like to recognize the following individuals for their contribution to supporting documents that formed a basis for this draft:

Ayers, Robert
Bhalla, Neelam
Brown, Keith D.
Frazier, Cassandra F.
Fuller, Mike L.
Holahan, Patricia K.
Merchant, Sally L.
Minnick, Sheri A.
Schlueter, Janet R.
Smith, James A.
Taylor, Torre M.

Abbreviations

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AAPM	American Association of Physicists in Medicine
ACMUI	Advisory Committee on the Medical Use of Isotopes
ALARA	as low as is reasonably achievable
ALI	annual limit on intake
AMP	Authorized Medical Physicist
ANP	Authorized Nuclear Pharmacist
ANSI	American National Standards Institute
AU	authorized user
bkg	background
BPR	Business Process Redesign
Bq	Becquerel
CFR	Code of Federal Regulations
Ci	curies
cc	centimeter cubed
cm²	centimeter squared
Co-57	cobalt-57
Co-60	cobalt-60
cpm	counts per minute
Cs-137	cesium-137
DAC	Derived Air Concentration
DIS	decay-in-storage
DOT	United States Department of Transportation
dpm	disintegrations per minute
FDA	United States Food and Drug Administration
ft	foot
G-M	Geiger-Mueller
GPO	Government Printing Office
GSR	gamma stereotactic radiosurgery
HDR	high dose-rate
I-125	iodine-125
I-131	iodine-131
IN	Information Notice
Ir-192	iridium-192
LDR	low dose-rate
mCi	millicurie
ml	milliliter
mR	milliroentgen
mrem	millirem
mSv	millisievert
NaI(Tl)	sodium iodide (thallium doped)
NCRP	National Council on Radiation Protection and Measurements
NIST	National Institute of Standards and Technology
NRC	United States Nuclear Regulatory Commission
NVLAP	National Voluntary Laboratory Accreditation Program
OCFO	Office of the Chief Financial Officer
OCR	optical character reader
OMB	Office of Management and Budget
P-32	phosphorus-32
Pd-103	palladium-103
PDR	pulsed dose-rate
P&GD	Policy and Guidance Directive
QA	Quality Assurance
Ra-226	Radium-226
RG	Regulatory Guide
RSC	Radiation Safety Committee
RSO	Radiation Safety Officer
SDE	shallow-dose equivalent
SI	International System of Units (abbreviated SI from the French Le Systeme Internationale d'Unites)
Sr-90	strontium-90
SSDR	Sealed Source and Device Registration
std	standard
Sv	sievert
Tc-99m	technetium-99m
TEDE	total effective dose equivalent
TLD	thermoluminescent dosimeters
U-235	uranium-235
�Ci	microcurie
WD	Written Directive

1 Purpose of Draft Report

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This document is strictly for public comment and is not for use in preparing or reviewing applications for medical use licenses until this document is published in final form.

This draft report provides guidance to an applicant in preparing a medical use license application as well as NRC criteria for evaluating a medical use license application. It is not intended to address the research and development or the commercial aspects of manufacturing, distribution, and service of medical radionuclides or sources in devices.

Radionuclides are used for a variety of purposes in medicine. Typical uses are:

Diagnostic studies with unsealed radionuclides
Therapeutic administrations with unsealed radionuclides
Diagnostic studies with sealed radionuclides
Manual brachytherapy
Therapeutic administrations with sealed sources in devices [i.e., teletherapy, remote afterloaders, and gamma stereotactic radiosurgery (GSR) units].

This draft report identifies the information needed to complete NRC Form 313 (Appendix B), "Application for Material License," for medical use of radionuclides. The information collection requirements in 10 CFR Parts 30 and 35, and NRC Form 313 have been approved under the Office of Management and Budget (OMB) Clearance Nos. 3150-0017, 3150-0010, and 3150-0120, respectively.

Information in this draft report will also help medical users understand specific regulatory requirements and licensing policies as they apply to medical licenses. Applicants are expected to address all the items on NRC Form 313 and to either follow the specific guidance that will become available in the final version of this draft report or to respond to the items in a manner that assures safe operation and compliance with the regulations that apply.

The format within this document for each item of technical information is as follows:

Regulations -- references the regulations applicable to the item
Criteria -- outlines the criteria used to judge the adequacy of the applicant's response
Discussion -- provides additional information on the topic sufficient to meet the needs of most readers, and
Response from Applicant -- provides suggested response(s) or indicates that no response is needed on that topic during the licensing process.

The regulations require the applicant/licensee to develop and implement procedures that will ensure compliance with the regulations. The appendices describe model radiation protection procedures. Each applicant should carefully read the regulations and model procedures and then decide if the model procedure appropriately addresses specific radiation protection program needs at the applicants facility. Applicants may adopt a model procedure or they may develop their own procedures to comply with the applicable regulation. Written procedures developed by applicants do not need to be submitted as part of the license application. However, the applicant must state that applicable procedures have been developed and implemented in accordance with the regulations.

NRC Form 313 does not have sufficient space for applicants to provide full responses to Items 5 through 11; as indicated on the form, the answers to those items are to be provided on separate sheets of paper and submitted with the completed NRC Form 313.

Appendix C includes sample medical licenses with conditions most often found on these licenses, although not all licenses will have all conditions. Appendix C also contains a checklist to assist the applicant in determining which sections of this document and required procedures apply to the type of medical license requested. Appendices D through Y contain additional information on various radiation protection topics.

In this document, dose or radiation dose means absorbed dose, dose equivalent, effective dose equivalent, committed dose equivalent, committed effective dose equivalent, or total effective dose equivalent (TEDE). These terms are defined in 10 CFR Part 20. Rem, and Sievert (Sv), its SI equivalent (1 rem = 0.01 Sv), are used to describe units of radiation exposure or dose. This is because 10 CFR Part 20 sets dose limits in terms of rem, not rad or roentgen, and the sources commonly used in therapy emit beta and gamma rays, which roughly means that 1 roentgen = 1 rad = 1 rem.

This NUREG not only updates the information and guidance provided in Revision 2 of RG 10.8, "Guide for the Preparation of Applications for Medical Use Programs," but also revises the format in which it is presented to assist with the preparation of a medical use license. Revision 2 was issued in August 1987 to provide guidance for the revised 10 CFR Part 35, which became effective on April 1, 1987. Since then, 10 CFR Part 35 has been revised a number of times. Technology-specific information has been revised and expanded to include technologies that are now more commonly used, for example, computerized remote afterloading brachytherapy [particularly high dose-rate (HDR)] and GSR.

Applicants and licensees should also be aware of two other documents that provide useful information for medical use licensees. The first is the October 1994 Draft RG DG-0005 (second proposed revision to RG 10.5), "Applications for Licenses of Broad Scope," which provides additional licensing guidance on medical use programs of broad scope. The second is the May 1997 NUREG-1516, "Effective Management of Radioactive Materials Safety Programs at Medical Facilities." The guidance in NUREG-1516 emphasizes a team approach to program management as a means to effectively manage radiation protection programs, and provides tools to help licensees manage these programs. Radiation protection team members should include the management of the licensed facility, the Radiation Safety Committee (RSC), if applicable, and the Radiation Safety Officer (RSO). The document also includes discussions on the duties and responsibilities of the RSO and supervised individuals; conduct of required audits; advantages and disadvantages in using consultants or service companies to augment the radiation protection program; resources that may be needed to support the program; and NRC notification and reporting requirements. Specific tools for day-to-day operation of a radiation protection program are provided in several appendices. Additionally, an extensive annotated bibliography lists publications on radiation protection program management at medical facilities.

1.1 Licenses

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NRC regulates the intentional internal or external administration of byproduct material, or the radiation therefrom, to patients or human research subjects for medical use. A specific license of either limited or broad scope is issued to authorize possession and use of licensed material. These licenses are issued pursuant to 10 CFR Parts 30, 33, and 35. NRC issues three types of licenses for the use of byproduct material in medical practices and facilities. These include; general in vitro license, specific license of limited scope, and specific license of broad scope.

NRC usually issues a single byproduct material license to cover an entire radionuclide program -- except for teletherapy, nuclear-powered pacemakers, and GSR. The teletherapy license will also contain the authorization for source material (i.e., depleted uranium) used as shielding in many teletherapy units. Although individual licensees may be issued separate licenses for different medical uses, separate licenses are not usually issued to different departments in a medical facility or to individuals employed by or contracted by the medical facility. Only the facility's management may apply for the license.

An applicant should carefully study this report, related guidance, and all applicable regulations before completing NRC Form 313. When necessary, NRC may ask the applicant for additional information in order to gain reasonable assurance that an adequate radiation protection program has been established.

After a license is issued, the licensee must conduct its program in accordance with the following:

statements, representations, and procedures contained in the application and in correspondence with NRC
terms and conditions of the license
NRC's regulations.

In 10 CFR 30.9, NRC requires that the information in the application be complete and accurate in all material respects. Information is considered to be material if it is likely to change or affect an agency decision on issuing the license.

1.1.1 General in Vitro License

In 10 CFR 31.11, "General license for use of byproduct material" for certain in vitro clinical or laboratory testing, NRC establishes a general license authorizing physicians, veterinarians, clinical laboratories, and hospitals to receive, acquire, possess, or use certain small quantities of byproduct material for in vitro clinical or laboratory tests not involving "medical use" (i.e., not involving administration to humans). Section 31.11 explains the requirements for using materials listed. If the general license alone meets the applicant's needs, only NRC Form 483, "Registration Certificate -- In Vitro Testing With Byproduct Material Under General License," need be filed. Medical use licensees authorized pursuant to 10 CFR Part 35 do not need to file the form.

In 10 CFR 31.11, NRC limits possession to a total of 200 microcuries, at any time, of photon-emitting materials listed in 10 CFR 31.11. The use of materials listed in 10 CFR 31.11 within the inventory limits of that section are subject only to the requirements of that section and not to the requirements of 10 CFR Parts 19, 20, 21, and 35, except as discussed in 10 CFR 31.11.

An applicant needing more than 200 microcuries of these materials must apply for a specific license and may request the increased inventory limit as a separate line item on the NRC Form 313 application. Such applicants generally request an increased limit of 3 millicuries. If requesting an increased inventory limit, the applicant will be subject to the requirements of 10 CFR Parts 19, 20, 21, and 35, including the requirements for waste disposal.

1.1.2 Specific License of Limited Scope

Specific licenses of limited scope are generally issued to physicians in private practice or in a group practice with a limited number of medical disciplines, and not to physicians located within a licensed medical institution. These licensees are not authorized to perform procedures requiring hospitalization of the patient.

Specific licenses of limited scope are also issued to medical institutions. A medical institution is an organization in which several medical disciplines are practiced. These licenses authorize the medical use of byproduct material by physicians named on the institution's license or authorized by the licensee in accordance with 10 CFR Part 35.

A specific license of limited scope may also be issued to a mobile service (10 CFR 35.80, 10 CFR 35.647): Physicians in private practice and medical institutions may apply for authorization to use byproduct material in a mobile service.

1.1.3 Specific License of Broad Scope

Some medical institutions provide patient care and conduct research programs that use radionuclides for in vitro, animal, and medical procedures. In these cases, the NRC may issue a specific license of broad scope as discussed in 10 CFR Part 33. Specific licenses of broad scope for medical use, i.e., licenses authorizing multiple quantities and types of byproduct material for unspecified uses, are issued to institutions that (1) have had previous experience successfully operating under a specific institutional license of limited scope and (2) are engaged in medical research, as well as in routine diagnostic and therapeutic uses of byproduct material. DG-0005 was issued for comment in October 1994 to offer additional guidance to applicants for specific medical use license of broad scope. Both DG-0005 and the final version of this draft NUREG report should be consulted for guidance on applying for a medical use license of broad scope.

1.2 the "As Low as Is Reasonably Achievable (Alara)" Concept

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10 CFR 20.1101 states that "each licensee must develop, document, and implement a radiation protection program commensurate with the scope and extent of licensed activities�" and "the licensee shall use, to the extent practicable, procedures and controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are ALARA." Additionally, this section requires that licensees periodically review the content of the radiation protection program and its implementation.

The success of an ALARA program depends on the cooperation of each person working at the licensed facility. Each individual who is authorized to use byproduct material should appropriately instruct all other individuals who work with or in the vicinity of byproduct material and should ensure that the facility and equipment are adequate for safe use. Each worker should comply with procedures developed to ensure safety and should promptly report incidents and potential problems to the authorized user or RSO.

Management should make a commitment to the ALARA philosophy and implement that commitment with adequate resources. The RSO and management must audit the byproduct material program to ensure the continued safe use of byproduct material. The RSO is responsible for the day-to-day operation of the radiation protection program.

RGs 8.10, "Operating Philosophy for Maintaining Occupational Radiation Exposures ALARA," and 8.18, "Information Relevant to Ensuring That Occupational Radiation Exposures at Medical Institutions Will Be ALARA," provide the NRC staff position on this subject. Several other NRC publications contain background information on the ALARA philosophy and its application in the medical environment. For example, NUREG-0267, "Principles and Practices for Keeping Occupational Radiation Exposures at Medical Institutions ALARA"; NUREG-1134, "Radiation Protection Training for Personnel Employed in Medical Facilities"; and NUREG-1516, "Effective Management of Radioactive Material Safety Programs at Medical Facilities," all contain information, methods, and references useful in establishing radiation protection programs to maintain exposures ALARA in medical institutions. Applicants should consider the ALARA philosophy as detailed in these reports when developing plans for work with licensed radioactive materials.

1.3 Written Directive (WD) Procedures

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10 CFR 35.41 requires medical use licensees to develop, maintain, and implement procedures to provide high confidence that for each administration requiring a WD, the patient's identity is verified prior to the administration and the administration is in accordance with the WD. Information on developing these procedures is found in Appendix S to this document.

1.4 Research Involving Human Subjects

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Effective January 1, 1995, the definition of "medical use" contained in 10 CFR 35.2 was revised to include the administration of byproduct material to human research subjects. Also, 10 CFR 35.6, "Provisions for research involving human subjects," was added to allow limited specific and broad scope medical use licensees to conduct "research involving human subjects," which meets specific criteria. Under 10 CFR 35.6, medical use licensees may conduct such research provided that the research is conducted, funded, supported, or regulated by another Federal agency which has implemented the Federal Policy for the Protection of Human Subjects. Otherwise, the licensee must apply for and receive approval of a specific amendment before conducting such research. Also, whether or not a license amendment is required, licensees must obtain informed consent from human subjects and obtain prior review and approval of the research activities by an Institutional Review Board in accordance with the meaning of those terms under the Federal Policy.

Table 1.1 When to Request NRC Approval for Human Research

Is the licensee one of the 15 Federal Agencies that adopted the Federal Policy?	YES	Research is covered by Federal Policy. No further information is needed.*
No
Is all human research conducted, funded, supported, or regulated by one of the 15 Federal Agencies that adopted the Federal Policy?	YES	Research is covered by Federal Policy. No further information is needed.*
No or not sure
Does the licensee have a valid "Multiple Project Assurance" with the Department of Health and Human Services (or other equivalent assurance with one of the other 14 Federal Agencies that adopted the Federal Policy)?	NO	Research not specifically conducted, funded, supported, or regulated by one of these Federal agencies requires an amendment request.*
Yes
Did the licensee voluntarily state in the "Multiple Project Assurance" (or other equivalent assurance with one of the other 14 Federal Agencies that adopted the Federal Policy) that all human research would be performed in accordance with the assurance?	NO	Research not specifically under the assurance is not covered by Federal Policy. An amendment request is necessary for this research.*
Yes
Research is covered by the Department of Health and Human Services or one of the other 14 agencies that adopted the Federal Policy, and no further information is needed.*
* If the radiation safety program necessitated by or associated with the research goes beyond the applicant's previously described radiation safety program, a revised radiation safety program should be submitted in an amendment request.

2 Agreement States

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Certain states, called Agreement States (see Figure 2.1), have entered into agreements with the NRC that give them the authority to license and inspect byproduct, source, or special nuclear materials used or possessed within their borders. Any applicant other than a Federal agency who wishes to possess or use licensed material in one of these Agreement States needs to contact the responsible officials in that State for guidance on preparing an application. These applications are filed with State officials, not with the NRC.

In the special situation of work at Federally-controlled sites in Agreement States, it is necessary to know the jurisdictional status of the land in order to determine whether NRC or the Agreement State has regulatory authority. NRC has regulatory authority over land determined to be "exclusive Federal jurisdiction," while the Agreement State has jurisdiction over non-exclusive Federal jurisdiction land. Licensees are responsible for finding out, in advance, the jurisdictional status of the specific areas where they plan to conduct licensed operations. NRC recommends that licensees ask their local contact for the Federal agency controlling the site (e.g., contract officer, base environmental health officer, district office staff) to help determine the jurisdictional status of the land and to provide the information in writing, so that licensees can comply with NRC or Agreement State regulatory requirements, as appropriate. Additional guidance on determining jurisdictional status is found in All Agreement States Letter, SP-96-022, dated February 16, 1996, which is available from NRC upon request.

Table 2.1 provides a quick way to check on which agency has regulatory authority.

Table 2.1 Who Regulates the Activity?

Applicant and Proposed Location of Work	Regulatory Agency
Federal agency regardless of location (except that Department of Energy [DOE] and, under most circumstances, its prime contractors are exempt from licensing [10 CFR 30.12])	NRC
Non-Federal entity in non-Agreement State, U.S. territory, or possession	NRC
Non-Federal entity in Agreement State at non-Federally controlled site	Agreement State
Non-Federal entity in Agreement State at Federally-controlled site *not* subject to exclusive Federal jurisdiction	Agreement State
Non-Federal entity in Agreement State at Federally-controlled site subject to exclusive Federal jurisdiction	NRC

Figure 2.1 U.S. Map. Location of NRC Offices and Agreement States. Note: As of October 1, 1998, the Walnut Creek Field Office will close. All communications should be forwarded to Region IV.

Reference: A current list of Agreement States (including names, addresses, and telephone numbers of responsible officials) may be obtained upon request from NRC's Regional or Field Offices. Or visit NRC's Home Page <http://www.nrc.gov>, choose "Nuclear Materials," then "Review of State Radiation Control Program Query Form," and then "Directories."

The All Agreement States Letter, SP-96-022, dated February 16, 1996, is available by contacting NRC's Office of State Programs; call NRC's toll free number (800) 368-5642, and then ask for extension 415-3340. Or visit NRC's Home Page <http://www.nrc.gov>, choose "Nuclear Materials," then choose "Review of State Radiation Control Program Query Form" and follow the directions for submitting a query for "SP-96-022."

3 Management Responsibility

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The NRC recognizes that effective radiation protection program management is vital to achieving safe and compliant operations (see 10 CFR 35.24). NRC believes that consistent compliance with its regulations provides reasonable assurance that licensed activities will be conducted safely.

"Management" refers to the chief executive officer or that person's delegate or delegates who have authority to provide necessary resources to achieve regulatory compliance.

To ensure adequate management involvement, a management representative (i.e., chief executive officer or delegate) must sign the submitted application acknowledging management's commitments and responsibility for the following:

Radiation protection, security and control of radioactive materials, and compliance with regulations
Completeness and accuracy of the radiation protection records and all information provided to NRC (10 CFR 30.9)
Knowledge about the contents of the license and application
Compliance with current NRC and United States Department of Transportation (DOT) regulations and the licensee's operating and emergency procedures
Commitment to provide adequate resources (including space, equipment, personnel, time, and, if needed, contractors) to the radiation protection program to ensure that patients, the public, and workers are protected from radiation hazards
Appointment of a qualified individual to work as the RSO
Approval of qualified individual(s) to serve as Authorized Medical Physicists (AMPs), Authorized Nuclear Pharmacists (ANPs), and authorized users (AUs) for licensed activities.

For information on NRC inspection, investigation, enforcement, and other compliance programs, see "General Statement of Policy and Procedures for NRC Enforcement Actions," NUREG-1600, dated June 1995, "Compilation of NRC Enforcement Policy as of September 10, 1997", and Manual Chapter 87100, Appendix B, "Nuclear Medicine Inspection Field Notes"; see Notice of Availability (on inside front cover of this draft report). NUREG-1600 is also available on the Internet. Visit NRC's Home Page <http://www.nrc.gov>, choose "Nuclear Materials," then "Enforcement," "Enforcement Guidance Documents," and then "Current Enforcement Policy."

4 Applicable Regulations

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It is the applicant's or licensee's responsibility to have up-to-date copies of applicable regulations, read them, and abide by each applicable regulation.

The following Parts of 10 CFR Chapter I contain regulations applicable to medical uses licensees:

10 CFR Part 2, "Rules of Practice for Domestic Licensing Proceedings and Issuance of Orders"
10 CFR Part 19, "Notices, Instructions and Reports to Workers: Inspection and Investigations"
10 CFR Part 20, "Standards for Protection Against Radiation"
10 CFR Part 21, "Reporting of Defects and Noncompliance"
10 CFR Part 30, "Rules of General Applicability to Domestic Licensing of Byproduct Material"
10 CFR Part 32, "Specific Domestic Licenses to Manufacture or Transfer Certain Items Containing Byproduct Material"
10 CFR Part 33, "Specific Domestic Licenses of Broad Scope for Byproduct Material"
10 CFR Part 35, "Medical Use of Byproduct Material"
10 CFR Part 40, "Domestic Licensing of Source Material"
10 CFR Part 71, "Packaging and Transportation of Radioactive Material"

Part 71 requires that licensees or applicants who transport licensed material or who may offer such material to a carrier for transport must comply with the applicable requirements of the DOT that are found in 49 CFR Parts 170 through 189. For ordering information on the regulations, see the Notice of Availability (on inside front cover of this draft report).

10 CFR Part 150, "Exemptions and Continued Regulatory Authority in Agreement States and in Offshore Waters Under Section 274"
10 CFR Part 170, "Fees for Facilities, Materials, Import and Export Licenses, and Other Regulatory Services Under the Atomic Energy Act of 1954, as Amended"
10 CFR Part 171, "Annual Fees for Reactor Operating Licenses, and Fuel Cycle Licenses and Materials Licenses, Including Holders of Certificates of Compliance, Registrations, and Quality Assurance Program Approvals and Government Agencies Licensed by NRC"

In addition to the information provided in the Notice of Availability (on inside front cover of this draft report), to request copies of the above documents, applicants may call the Government Printing Office's (GPO's) order desk in Washington, DC at (202) 512-1800. Order the two-volume bound version of Title 10, Code of Federal Regulations, Parts 0-50 and 51-199 from the GPO, Superintendent of Documents, Post Office Box 371954, Pittsburgh, Pennsylvania 15250-7954. You may also contact the GPO electronically at <http://www.gpo.gov>. Request single copies of the above documents from NRC's Regional or Field Offices (see Figure 2.1 for addresses and telephone numbers). Note that NRC publishes amendments to its regulations in the Federal Register. Title 10 is also available at <http://www.nrc.gov>, choose "Reference Library", and then "Title 10 of The Code of Federal Regulations."

5 How to File

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5.1 Paper Application

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Applicants for a materials license should do the following:

Be sure to use the most recent guidance in preparing an application
Complete NRC Form 313 (Appendix B) Items 1 through 4, 12, and 13 on the form itself
Complete NRC Form 313 Items 5 through 11 on supplementary pages or use Appendix C
Provide sufficient detail for the NRC to determine that equipment, facilities, training and experience, and the radiation safety program are adequate to protect health and safety and minimize danger to life and property
For each separate sheet, other than Appendix C, that is submitted with the application, identify and key it to the item number on the application or the topic to which it refers
Submit all documents, typed, on 8-1/2 x 11 inch paper
Avoid submitting proprietary information unless it is absolutely necessary
Submit an original, signed application and one copy
Retain one copy of the license application for future reference.

As required by 10 CFR 30.32(c) and 10 CFR 35.12(a), applications must be signed by the management of the facility; see section on "Certification."

Using the suggested wording of responses in this draft report will expedite NRC's review.

All license applications will be available for review by the general public in NRC's Public Document Rooms. If it is necessary to submit proprietary information, follow the procedure in 10 CFR 2.790. Failure to follow this procedure could result in disclosure of the proprietary information to the public or substantial delays in processing the application. Employee personal information, i.e., home address, home telephone number, social security number, date of birth, and radiation dose information, should not be submitted unless specifically requested by NRC.

NRC's new licensing process will be faster and more efficient, in part, through acceptance and processing of electronic applications at some future date. NRC will continue to accept paper applications. However, these will be scanned and put through an optical character reader (OCR) to convert them to electronic format. To ensure a smooth transition, applicants are requested to follow these suggestions:

Submit printed or typewritten, not handwritten, text on smooth, crisp paper that will feed easily into the scanner
Choose typeface designs that are sans serif, such as Arial, Helvetica, Futura, Univers; the text of this document is in a serif font called Times New Roman
Choose 12-point or larger font size
Avoid stylized characters such as script, italic, etc.
Be sure the print is clear and sharp
Be sure there is high contrast between the ink and paper (black ink on white paper is best).

5.2 Electronic Application

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As the electronic licensing process develops, it is anticipated that NRC may provide mechanisms for filing applications via diskettes or CD-ROM, and through the Internet. Additional filing instructions will be provided as these new mechanisms become available. The existing paper process will be used until the electronic process is available.

6 Where to File

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Applicants wishing to possess or use licensed material in any State or U. S. territory or possession subject to NRC jurisdiction must file an application with the NRC Regional Office for the locale in which the material will be possessed and/or used. Figure 2.1 shows NRC's four Regional Offices and their respective areas for licensing purposes and identifies Agreement States. The Walnut Creek, California, Field Office, can respond to routine telephone inquiries until September 30, 1998. Effective October 1, 1998, the Walnut Creek, California, Field Office will close and any communications previously involving that Field Office should be addressed to the Region IV Office.

In general, applicants wishing to possess or use licensed material in an Agreement State must file an application with the Agreement State, not NRC. However, if work will be conducted at Federally controlled sites in Agreement States, applicants must first determine the jurisdictional status of the land in order to determine whether NRC or the Agreement State has regulatory authority. See Section 2, "Agreement States," for additional information.

7 License Fees

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Each application for which a fee is specified, including applications for new licenses and license amendments, must be accompanied by the appropriate fee. Refer to 10 CFR 170.31 to determine the amount of the fee. NRC will not issue the new license prior to fee receipt. Once technical review has begun, no fees will be refunded; application fees will be charged regardless of the NRC's disposition of an application or the withdrawal of an application.

Most NRC licensees are also subject to annual fees; refer to 10 CFR 171.16. Consult 10 CFR 171.11 for additional information on exemptions from annual fees and 10 CFR 171.16(c) on reduced annual fees for licensees that qualify as "small entities."

Direct all questions about NRC's fees or completion of Item 12 of NRC Form 313 (Appendix B) to the Office of the Chief Financial Officer (OCFO) at NRC headquarters in Rockville, Maryland, (301) 415-7544. You may also call NRC's toll free number (800) 368-5642 and then ask for extension 415-7544.

Enter the fee category and the amount of the fee enclosed with the application on NRC Form 313.

8 Contents of an Application

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This section explains, item by item, the information requested on NRC Form 313. Items 5 through 11 on the form request specific information about the proposed radiation safety program. To assist the applicant in submitting complete information on these items, the applicable regulations are referenced in the discussion of each item.

Applicants must provide detailed information about the following:

Proposed facilities and equipment
Training and experience of byproduct material users and the RSO
Delegation of authority to RSO
Financial assurance (if applicable)
Mobile use of byproduct material (if applicable).

Additionally, in response to Items 9, 10, and 11, the applicant must provide a commitment to develop and implement various procedures to meet the requirements of the applicable regulation. Table 1 in Appendix C is provided to assist applicants determine which procedures must be developed and implemented for the type of medical use requested. Several appendices in this report present sample procedures that the applicant may use in developing their procedures. If a particular item requires the applicant to develop and implement a procedure, the applicant may use the following wording in each response section on the application:

"We have developed and will implement written procedures for ___________ that meet the requirements of 10 CFR _____."

If an applicant or licensee commits to a section of this report, that commitment will be incorporated as a part of the terms and conditions of the license. The licensee will be inspected against the commitments contained in the referenced section, appendix, or document, just as the applicant/licensee will be inspected against more detailed responses.

If a particular part of a section does not apply simply note "NA" for "not applicable." If a particular section applies, but a procedure does not have to be developed, simply note "N" for "no response required." Short sentence responses, "NA" or "N" responses to Items 5 through 10 should run consecutively on one or more sheets separate from responses provided on NRC Form 313. Lengthy responses should be appended as attachments.

As indicated on NRC Form 313 (Appendix B), responses to Items 5 through 11 should be submitted on separate sheets of paper. Applicants may use Appendix C to assist with completion of the application. Applicants should note that using the suggested wording of responses will expedite NRC's review.

8.1 Item 1: License Action Type

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THIS IS AN APPLICATION FOR (Check appropriate item)

Type of Action	License No.
[ ] A. New License	Not Applicable
[ ] B. Amendment to License No.	XX-XXXXX-XX
[ ] C. Renewal of License No.	XX-XXXXX-XX

Check box A if the application is for a new license

Check box B for an amendment⁽¹⁾ to an existing license, and provide license number.

Check box C for a renewal¹ of an existing license, and provide license number.

8.2 Item 2: Applicant's Name And Mailing Address

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List the legal name of the applicant's corporation or other legal entity with direct control over use of the radioactive material; a division or department within a legal entity may not be a licensee. An individual may be designated as the applicant only if the individual is acting in a private capacity and the use of the radioactive material is not connected with employment in a corporation or other legal entity. Provide the mailing address where correspondence should be sent. A post office box number is an acceptable mailing address.

Note: NRC must be notified before control of the license is transferred or when bankruptcy proceedings have been initiated. See below for more details. NRC IN 97-30, "Control of Licensed Material during Reorganizations, Employee-Management Disagreements, and Financial Crises," dated June 3, 1997, discusses the potential for the security and control of licensed material to be compromised during periods of organizational instability.

Timely Notification of Transferring Control

Regulations: 10 CFR 30.34(b), 10 CFR 35.14(b).

Criteria: Licensees must provide full information and obtain NRC's prior written consent before transferring control of the license, or, as some licensees call it, "transferring the license."

Discussion: Transferring control may be the result of mergers, buyouts, or majority stock transfers. Although it is not NRC's intent to interfere with the business decisions of licensees, it is necessary for licensees to obtain prior NRC written consent before transferring control of the license. This is to ensure the following:

Radioactive materials are possessed, used, or controlled only by persons who have valid NRC licenses
Materials are properly handled and secured
Persons using these materials are competent and committed to implementing appropriate radiological controls
A clear chain of custody is established to identify who is responsible for final disposal of the material
Public health and safety are not compromised by the use of such materials

As provided in 10 CFR 35.14(b), if the licensee's name or mailing address changes, and the name change does not constitute a transfer of control of the license as described in 10 CFR 30.34(b), a licensee must file a written 30 day notification with the NRC instead of obtaining prior NRC written consent.

Response from Applicant: No response is required from an applicant for a new license; Appendix D, excerpted from IN 89-25 (Rev. 1), "Unauthorized Transfer of Ownership or Control of Licensed Activities," dated December 7, 1994, identifies the information to be provided about transferring control.

Reference: See the Notice of Availability on the inside front cover of this draft report to obtain copies of IN 89-25 (Rev. 1), "Unauthorized Transfer of Ownership or Control of Licensed Activities," dated December 7, 1994, and IN 97-30, "Control of Licensed Material during Reorganizations, Employee-Management Disagreements, and Financial Crises," dated June 3, 1997.

Notification of Bankruptcy Proceedings

Regulation: 10 CFR 30.34(h).

Criteria: Immediately following filing of a voluntary or involuntary petition for bankruptcy for or against a licensee, the licensee must notify the appropriate NRC Regional Administrator, in writing, identifying the bankruptcy court in which the petition was filed and the date the of filing.

Response from Applicant: None at time of application for a new license. Licensees must notify NRC immediately (i.e., within 24 hours) of filing a bankruptcy petition.

8.3 Item 3: Address(es) Where Licensed Material Will Be Used or Possessed

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Specify the street address, city, and state or other descriptive address (e.g., on Highway 10, 5 miles east of the intersection of Highway 10 and State Route 234, Anytown, State) for each facility. The descriptive address should be sufficient to allow an NRC inspector to find the facility location. A post office box address is not acceptable (see Figure 8.1). If byproduct material is to be used at more than one location under the license, the specific address (e.g., street and building) must be provided for each facility. If applying for a license for a mobile service as authorized pursuant to 10 CFR 35.18(b), the applicant should refer to Section 8.42 and Appendix V of this report for specific licensing guidance.

Figure 8.1 Location of Use.

Being granted an NRC license does not relieve a licensee from complying with other applicable Federal, State, or local regulations (e.g., local zoning requirements; a local ordinance requiring registration of a radiation-producing device).

Note: As discussed later under "Financial Assurance and Record Keeping for Decommissioning," licensees do need to maintain permanent records on where licensed material was used or stored while the license was in effect. This is important for making future determinations about the release of these locations for unrestricted use (e.g., before the license is terminated). For medical use licensees, acceptable records are sketches or written descriptions of the specific locations where material is used or stored and any information relevant to spills (e.g., where contamination remains after cleanup procedures or when there is reasonable likelihood that contaminants may have spread), damaged devices, or leaking radioactive sources.

8.4 Item 4: Person to Be Contacted About This Application

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Identify the individual who can answer questions about the application and include his or her telephone number. This is typically the proposed RSO, unless the applicant has named a different person as the contact. The NRC will contact this individual if there are questions about the application.

Notify NRC if the contact person or his or her telephone number changes so that NRC can contact the applicant or licensee in the future with questions, concerns, or information. This notice is for "information only" and does not require a license amendment or a fee.

The individual named in Item 4 may or may not be the same individual who signs the application as the "certifying official" on behalf of the licensee and has the authority to make commitments to NRC (see Item 13 on Form 313). Any commitments made by the applicant should be signed by the individual named in Item 13 since only that individual is considered by NRC to have the authority to make commitments on behalf of the applicant. Therefore, NRC will not accept license amendments or renewals signed by the individual identified in Item 4, if this person differs from the one named in Item 13.

NRC recognizes that licensees may use a consultant or consultant group to help prepare the license application and provide support to the radiation protection program. Licensees are reminded that regardless of the role of the consultant in radiation protection program management, the licensee remains ultimately responsible for all aspects of the licensed program, including the services performed by the consultant.

As indicated on NRC Form 313 (Appendix B), Items 5 through 11 should be submitted on separate sheets of paper. Applicants may use Appendix C for guidance and should note that using the suggested wording of responses will expedite NRC's review.

8.5 Item 5: Radioactive Material

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Regulations: 10 CFR 30.32, 10 CFR 30.33, 10 CFR 30.34, 10 CFR 30.35, 10 CFR 35.100, 10 CFR 35.200, 10 CFR 35.300, 10 CFR 35.400, 10 CFR 35.500, 10 CFR 35.600, 10 CFR 35.1000.

Criteria: 10 CFR Part 35 divides byproduct material for medical use into seven types of use (10 CFR 35.100, 35.200, 35.300, 35.400, 35.500, 35.600, and 35.1000).

Discussion: Using the table formats below (see Table 8.1), the applicant should indicate the byproduct material requested. For 35.100, 35.200, and 35.300 material, the chemical/physical form may be "Any." For 35.100 and 35.200 material, the total amount requested may be "As Needed." For 35.300 material, the total amount requested must be specified. For 35.400, 35.500, 35.600, and 35.1000 material, express the radionuclide, the chemical/physical form (e.g., sealed source and manufacturer's name and model number), the total amount in Becquerels (Bq), microcuries (�Ci), millicuries (mCi), or curies (Ci), and maximum number of sources possessed at any one time.

For sealed sources used in devices, an applicant may wish to request two sources, one to be used in the device and one to be stored in its shipping container, to accommodate the total quantity of material in the licensee's possession during replacement of the source in the device. The maximum activity for a single source or source loading may not exceed the activity specified by the manufacturer for the specific device and source combination as stated in the Sealed Source and Device Registration (SSDR) Certificate. However, it is permissible to request a maximum activity, for the source in the shipping container, that exceeds the maximum activity allowed in the device. To request such authorization, applicants should provide certification that the source transport container is approved for the requested activity. A source that is received with a higher activity than permitted in the device must be allowed to decay to, or below, the device source activity limit prior to installation in the device.

If applicable, the applicant should request authorization for possession of depleted uranium [i.e., uranium depleted in uranium-235 (U-235)] in quantities sufficient to include shielding material in both the device(s) and source containers used for source exchange. The applicant should review the manufacturer's specifications for each device specified in the license request to determine (1) whether depleted uranium is used for shielding the source(s) within the device and (2) the total quantity of depleted uranium present in the device (in kilograms). The applicant should also consult the manufacturer's specifications or the source supplier to determine whether depleted uranium is contained within shielding source containers used during source exchange and determine the total quantity of depleted uranium in such containers (in kilograms).

A separate entry should be made for other items that need to be listed (e.g., more byproduct material for in vitro testing than is allowed under 10 CFR 31.11, depleted uranium for linear accelerator shielding, survey meter calibration source, dosimetry system constancy check source, or material for in vitro, animal, or human research studies). Sources that are authorized by 10 CFR 35.65, "Authorization for calibration and references sources" should not be listed. Applicants should number each line entry consecutively following the 10 CFR Part 35 material.

Table 8.1 Sample Format for Byproduct Material

Byproduct Material	Chemical/Physical Form	Maximum Amount
Any byproduct material included in 10 CFR 35.100	Any	As needed
Any byproduct material included in 10 CFR 35.200	Any	As needed
Any byproduct material included in 10 CFR 35.300	Any	300 millicuries
Cesium 137 (i.e., specific brachytherapy radionuclide)	sealed source (Manufacturer Name, Model #XYZ)	2 curies total
Gadolinium 153 (i.e., specific diagnostic sealed source radionuclide)	sealed source (Manufacturer Name, Model #XYZ)	Not to exceed 500 millicuries per source and 1 curie total
Cobalt 60 (i.e., specific teletherapy sealed source radionuclide)	sealed source (Manufacturer Name, Model #XYZ)	Not to exceed 9,000 curies per source and 18,000 curies total
Iridium 192 (i.e., specific afterloader sealed source radionuclide)	sealed source (Manufacturer Name, Model #XYZ)	Not to exceed 10 curies per source and 20 curies total
Cobalt 60 (i.e., specific gamma stereotactic radiosurgery sealed source radionuclide)	sealed source (Manufacturer Name, Model #XYZ)	Not to exceed 36 curies per source and 6,600 curies total
Depleted Uranium	Metal	99 kilograms
Any byproduct material identified in 10 CFR 31.11	Prepackaged Kits	50 millicuries

When determining both individual radionuclide and total quantities, all materials to be possessed at any one time under the license should be included, i.e., materials received awaiting use (new teletherapy or brachytherapy sources for exchange); materials in use or possessed; material used for shielding; and those materials classified as waste awaiting disposal or being held for "decay-in-storage (DIS)."

Response from Applicant: The applicant shall submit the information as described above.

8.6 Item 5: Financial Assurance And Recordkeeping For Decommissioning

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Regulations: 10 CFR 30.34(b), 10 CFR 30.35.

Criteria: Licensees authorized to possess licensed material in excess of the limits specified in 10 CFR 30.35 must provide evidence of financial assurance for decommissioning.

Even if no financial assurance is required, licensees are required to maintain, pursuant to 10 CFR 30.35(g), in an identified location, decommissioning records related to structures and equipment where licensed material is used or stored, spills or spread of contamination, and leaking sealed sources (see Figure 8.2). Licensees must transfer these records important to decommissioning either to the new licensee before licensed activities are transferred or assigned in accordance with 10 CFR 30.34(b) or to the appropriate NRC Regional Office before the license is terminated.

Figure 8.2 Decommissioning Records.

Discussion: The requirements for financial assurance are specific to the types and quantities of byproduct material authorized on a license. Most medical use applicants and licensees do not need to take any action to comply with the financial assurance requirements because either their total inventory of licensed material does not exceed the limits in 10 CFR 30.35 or the half-life of the unsealed byproduct material used does not exceed 120 days. The limits for some sealed sources are shown in Table 8.2. Applicants requesting more than one radionuclide need to use the sum of the ratios method to determine whether financial assurance is needed. See Appendix E for additional information.

Table 8.2 Minimum Sealed Source Inventory Quantity Requiring Financial Assurance

Radionuclide	Activity in GigaBq	Activity in Ci
cesium-137 (Cs-137)	3.7 x 10⁶	100000
cobalt-60 (Co-60)	3.7 x 10⁵	10000
strontium-90 (Sr-90)	3.7 x 10⁴	1000

Applicants and licensees wanting to possess licensed materials exceeding the limits in 10 CFR 30.35 must submit evidence of financial assurance or a decommissioning funding plan (10 CFR 30.35 (b)). Figure 8.3 depicts acceptable methods of providing financial assurance. Regulatory Guide (RG) 3.66, "Standard Format and Content of Financial Assurance Mechanisms Required for Decommissioning Under 10 CFR Parts 30, 40, 70, and 72," dated June 1990, contains approved wording for each mechanism authorized by the regulation to guarantee or secure funds, except for the Statement of Intent for government licensees. See Appendix E for the recommended wording for a Statement of Intent.

Figure 8.3 Financial Assurance Mechanisms.

NRC will authorize possession exceeding the limits shown in Table 8.2 without requiring decommissioning financial assurance, for the purpose of normal sealed source exchange for no more than 30 days.

Licensees using sealed sources as authorized by 10 CFR Part 35, in general, use licensed material in a manner that would preclude releases into the environment, would not cause the activation of adjacent materials, or would not contaminate work areas. Moreover, the licensee's most recent leak test should demonstrate that there has been no leakage from the sealed sources while in the licensee's possession. However, any significant leakage of the sealed source would warrant further review of decommissioning procedures on a case-by-case basis.

Response from Applicants: No response is needed from most applicants. If financial assurance is required, applicants must submit evidence as described in RG 3.66.

Pursuant to 10 CFR 30.35(g), licensees must transfer records important to decommissioning either to the new licensee before licensed activities are transferred or assigned in accordance with 10 CFR 30.34(b) or to the appropriate NRC Regional Office before the license is terminated.

Reference: See the Notice of Availability on the inside front cover of this draft report to obtain copies of Regulatory Guide 3.66, "Standard Format and Content of Financial Assurance Mechanisms Required for Decommissioning Under 10 CFR Parts 30, 40, 70, and 72," dated June 1990, and Policy and Guidance Directive FC 90-2 (Rev. 1), "Standard Review Plan for Evaluating Compliance with Decommissioning Requirements," dated April 30, 1991.

8.7 Item 5: Sealed Sources And Devices

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Regulation: 10 CFR 30.32(g), 10 CFR 30.33(a)(2), 10 CFR 32.210

Criteria: Applicants must provide the manufacturer's name and model number for each requested sealed source and device (except for calibration and reference sources authorized by 10 CFR 35.65). Licensees will be authorized to possess and use only those sealed sources and devices specifically approved or registered by NRC or an Agreement State.

Discussion: NRC or an Agreement State performs a safety evaluation of sealed sources and devices before authorizing a manufacturer to distribute the sources or devices to specific licensees. The safety evaluation is documented in a SSDR Certificate. Applicants must provide the manufacturer's name and model number for each requested sealed source and device so that NRC can verify that they have been evaluated in an SSDR Certificate or specifically approved on a license. If such a review has not been conducted for the specific source/device model(s), licensees should request a copy of NUREG-1556, Vol. 3, "Consolidated Guidance about Materials Licensees: Applications for Sealed Source and Device Evaluation and Registration," dated September 1997, from the NRC Regional Office and submit the information requested therein to NRC for review.

An applicant may consult with the proposed supplier or manufacturer to ensure that requested sources and devices are compatible and conform to the SSDR designations registered with NRC or an Agreement State. Licensees may not make any changes to the sealed source, device, or source/device combination that would alter the description or specifications from those indicated in the respective SSDR Certificates, without obtaining NRC's prior permission in a license amendment. To ensure that sealed sources and devices are used according to the registration certificates, applicants may want to obtain a copy of the certificate and review it or discuss it with the manufacturer.

Response from Applicant: No response is necessary.

Reference: See the Notice of Availability on the inside front cover of this draft report to obtain a copy of NUREG-1556, Vol. 3, "Consolidated Guidance about Materials Licensees: Applications for Sealed Source and Device Evaluation and Registration," dated September 1997.

Note: Information on SSD registration certificates is also available on the Internet at <http://www.hsrd.ornl.gov/nrc/ssdrform.ht m> or contact the Registration Assistant by calling NRC's toll free number (800) 368-5642 and then asking for extension 415-7217.

8.8 Item 6: Purpose(s) for Which Licensed Material Will Be Used

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Regulations: 10 CFR 30.33(a)(1), 10 CFR 35.100, 10 CFR 35.200, 10 CFR 35.300, 10 CFR 35.400, 10 CFR 35.500, 10 CFR 35.600, 10 CFR 35.1000.

Criteria: 10 CFR Part 35 divides byproduct material for medical use into seven types of use as follows:

10 CFR 35.100	Medical Use of Unsealed Byproduct Material for Uptake, Dilution and Excretion Studies for Which a Written Directive is not Required
10 CFR 35.200	Medical Use of Unsealed Byproduct Material for Imaging and Localization Studies for Which a Written Directive is not Required
10 CFR 35.300	Medical Use of Unsealed Byproduct Material for Which a Written Directive is Required
10 CFR 35.400	Medical Use of Sources for Manual Brachytherapy
10 CFR 35.500	Medical Use of Sealed Sources for Diagnosis
10 CFR 35.600	Medical Use of a Sealed Source(s) in a Device for Therapy-Teletherapy Unit
10 CFR 35.600	Medical Use of a Sealed Source(s) in a Device for Therapy-Remote Afterloader Unit
10 CFR 35.600	Medical Use of a Sealed Source(s) in a Device for Therapy-Gamma Stereotactic Radiosurgery Unit
10 CFR 35.1000	Other Medical Uses of Byproduct Material or Radiation from Byproduct Material (Emerging Technology)

Discussion: For 35.100, 35.200 and 35.300 material, the applicant must define the purpose of use by stating the applicable section of 10 CFR Part 35 (e.g., 10 CFR 35.100, 10 CFR 35.200) and the description of the applicable modality (e.g., any uptake dilution and excretion procedure approved in 10 CFR 35.100).

The use of unsealed byproduct material in therapy involves the administration of a radiopharmaceutical, either orally or by injection, to diagnose, treat or palliate a particular disease. The most common form of radiopharmaceutical therapy is the treatment of hyperthyroidism with iodine-131 (I-131) sodium iodide. Less common therapeutic procedures include ablation of thyroid cancer metastases, treatment of malignant effusions, treatment of polycythemia vera and leukemias, palliation of bone pain in cancer patients, and radiation synovectomy for rheumatoid arthritis patients. Table 8.3 contains a summary of several therapeutic radiopharmaceuticals and their use.

Table 8.3 Radiopharmaceuticals Used in Therapy

Agent	Form	Route of Administration	Therapeutic Use
I-131 sodium iodide	solution/capsules	oral	hyperthyroidism thyroid carcinoma total body scan for thyroid metastases (diagnostic)
phosphorus-32 (P-32) chromic phosphate	colloidal suspension	intraperitoneal or intrapleural cavity injection	peritoneal or pleural effusions
P-32 sodium phosphate	solution	oral or IV	polycythemia vera leukemias
strontium-89 chloride	solution	IV	skeletal metastases
samarium-153 EDTMP	solution	IV	skeletal metastases
rhenium-186 HEDP	solution	IV	skeletal metastases
tin-117m DTPA	solution	IV	skeletal metastases
dysprosium-165 FHMA	aggregate in solution	IV	rheumatoid arthritis
yttrium-90 FHMA	aggregate in solution	IV	rheumatoid arthritis

For 35.400 material, the applicant must define the purpose of use by stating the applicable section of 10 CFR Part 35 (i.e., 10 CFR 35.400). For 35.400 material, applicants may need to define the purpose of use by describing the manufacturer's name(s) and model number(s) of devices containing sealed sources (where applicable). The licensee should correlate the sealed sources listed in Item 5 with the devices described in this item.

In manual brachytherapy several types of treatments are available. These may include:

Interstitial Treatment of Cancer. The following sources are routinely used:

Cs-137 and Co-60 as a sealed source in needles and applicator cells
iridium-192 (Ir-192) as seeds encased in nylon ribbon
gold-198, iodine-125 (I-125), and palladium-103 (Pd-103) as a sealed source in seeds

Eye Plaque Implants
The eye plaque consists of a curved soft plastic insert that has a series of grooves molded into the rear convex surface, which are designed to hold the radioactive seeds. After the plastic insert is loaded with the seeds, a solid gold cover, matched in size to the insert, is placed over the convex surface of the insert and cemented in place to seal the seeds into a fixed array within the plaque. The insert is completely surrounded by the gold cover except for the concave surface that is placed against the eye. When used with I-125 and Pd-103 seeds, the gold cover provides considerable shielding of the normal tissues surrounding the eye and limits the external dose rates surrounding the patient. Although not implanted into the tumor, because the plaque is placed in the orbit of the eye over the tumor site and sutured to the sclera of the eye to stabilize its position on the tumor while in the orbit, this is considered interstitial, not topical, treatment.

Intracavitary Treatment of Cancer. In addition, for purposes of NRC's sealed source and device evaluation on radiation safety issues, intraluminal use is considered analogous to intracavitary use. The following sources are routinely used for the intracavitary treatment of cancer:

Cs-137 and Co-60 as a sealed source in needles and applicator cells
Ir-192 and Pd-103 seeds.

Topical (Surface) Applications. The following sources are routinely used for topical applications:

Cs-137 and Co-60 as sealed sources in needles and applicator cells
Sr-90 as a sealed source in an applicator for treatment of superficial eye conditions.

For 35.500 material, the applicant must define the purpose of use by stating the applicable section of 10 CFR Part 35 (i.e., 10 CFR 35.500) and describing the manufacturer's name(s) and model number(s) of devices containing sealed sources (where applicable). The licensee should correlate the sealed sources listed in Item 5 with the devices described in this item. Typically, a licensee should use the following sealed sources according to manufacturer's radiation safety and handling instructions and must use the sources as approved in the SSDR:

I-125, americium-241, or gadolinium-153 as a sealed source in a device for bone mineral analysis; and
I-125 as a sealed source in a portable imaging device

For 35.600 material, the applicant must define the purpose of use by stating the applicable section of 10 CFR Part 35.600 (e.g., teletherapy, remote afterloading, GSR) and describing the manufacturer's name(s) and model number(s) of the device containing a sealed source(s) (e.g., for use in a Leksell Gamma System Model 23016 {a.k.a. Gamma Knife or Cerebral Stereotactic Radiosurgical Unit} radiation therapy unit for the treatment of humans). A schematic of a GSR unit is provided as Figure 8.4. The applicant should correlate the sealed source(s) listed in Item 5 with the device described in this item. If applicable, the applicant should state that depleted uranium is used as shielding for the device and specify that an additional source is requested to be stored in its shipping container incident to source replacement.

Figure 8.4 Gamma Stereotactic Radiosurgery Unit.

For 35.1000 material, the applicant must define the purpose of use by stating the applicable section of 10 CFR Part 35 (e.g., 10 CFR 35.1000) and describing the manufacturer's name(s) and model number(s) of devices containing sealed sources (where applicable). The licensee should correlate the sealed sources listed in Item 5 with the devices described in this item.

One area of an emerging technology is the use of byproduct material to prevent restenosis. Post-operative radiation exposure has recently been shown to reduce the probability of restenosis following balloon angioplasty in patients. Balloon angioplasty can damage the smooth muscle lining the artery, stimulating cell growth that may result in restenosis. Vessel reclosure occurs in approximately 30% to 40% of angioplasty cases within 6 months because the artery wall collapses or scars and fills with tissue. There are two components to restenosis: the first, recoil, is the mechanical collapse of the dilated arteries; and the second, intimal hyperplasia, is the proliferation of smooth muscle cells in response to the vascular injury.

Emerging technologies have been developed using radioactive catheters, pellets, and stents to treat coronary and peripheral vascular problems. These therapy devices contain ionizing radiation in the form of a gas, liquid, or solid that retards recoil and proliferation of smooth muscle cells in the affected vessel wall. The radiation can be ion implanted, plated, or encapsulated in a sealed source device attached to a guide wire used in the angioplasty procedure. The radioactive device can be either permanently implanted or removed via the guide wire following treatment of the affected vessel wall.

Intracoronary radiation therapy is emerging as the primary discipline of the new technology. Because of the trauma and expense of performing repeat coronary procedures, physicians and medical companies have developed devices that appear to inhibit restenosis rates in clinical trials. Three major innovative types of intravascular radiation therapy devices are being clinically investigated following balloon angioplasty:

1.	Intracoronary Beta Radiation Catheter (Figure 8.5) - The catheter is not an implant and the radiation is delivered post balloon angioplasty. The beta radiation catheter is unsheathed in the affected coronary artery where the balloon angioplasty had occurred. The catheter delivers a high dose directly to the arterial wall while having a minimal effect on the whole body. Treatment times average around 5 minutes. The clinician controls the delivery and return of the source train. The catheter is retracted from the artery using the guide wire from the angioplasty procedure.
2.	Intracoronary Beta Radiation Stent (Figure 8.6) - The stent is a permanent implant and the radiation is delivered post balloon angioplasty. The stent is attached to the outside of the balloon catheter. The catheter is inflated and the stent is pressed against the arterial wall which becomes permanently implanted. The balloon is then deflated and retracted from the coronary artery. The stent delivers a high dose directly to the arterial wall while having a minimal effect on the whole body.
3.	Intracoronary Beta Radiation Pellets - Radiation is delivered post balloon angioplasty. The pellets are temporarily implanted at the angioplasty site in the coronary artery for a short treatment time (approximately 30 minutes). The clinician controls the delivery and return of the source train from the coronary artery. The pellets deliver a high dose directly to the arterial wall while having a minimal effect on the whole body.

Figure 8.5 Beta Radiation Catheter.

Figure 8.6 Stent Implantation.

A new clinical procedure uses beta radiation sutures to reduce restenosis in renal dialysis patients. Radiation is thought to produce an antiscarring effect by breaking the strands of DNA in the endothelial cells of the affected vessels.

The United States Food and Drug Administration (FDA) considers the use of radiation in the coronary and/or peripheral vasculature for the prevention of restenosis investigational with the potential for risk to patients. Legal and ethical considerations require U.S. patients to be studied under an investigational device exemption application at the time of this writing.

Benchtesting of the radiation device in an emerging technology includes the following criteria:

Characterization of the radiation source
Biocompatibility of blood-containing materials used in the delivery system (if applicable)
Mechanical integrity of components used in the delivery system (if applicable)
Sterility of the source and delivery system.

Bench testing is conducted by the manufacturer and the clinical investigator(s).

Radionuclides that are being used in the emerging technology include P-32, Sr-90, Ir-192, rhenium-186, rhenium-188, xenon-133, and hydrogen-3. The use of beta radiation and/or low energy photons localized to the affected site has the potential to subsequently reduce or eliminate restenosis.

Applicants must also describe non-medical uses and reference the applicable radioactive material provided in response to Item 5.

Appendix C contains sample licenses that provide guidance on how to respond to Item 6.

Response from Applicant: The applicant shall submit the information as described above.

8.9 Item 7: Individual(s) Responsible For Radiation Safety Program And Their Training And Experience

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Regulations: 10 CFR 30.33(a)(3), 10 CFR Part 35.

Criteria: The RSO, AUs, AMPs, and ANPs must have adequate training and experience.

Discussion: Individuals responsible for the radiation protection program are licensee senior management, the AUs, RSO, RSC (if applicable), AMPs, and ANPs. In 10 CFR 30.33(a)(3), NRC requires that an applicant be qualified by training and experience to use licensed materials for the purposes requested in such a manner as to protect health and minimize danger to life or property. Subparts B, D, E, F, G, H, and J of 10 CFR Part 35 give specific criteria for acceptable training and experience for AUs for medical use, ANPs, the RSO, and AMPs. Applicants should note that a résumé or a curriculum vitae does not usually supply all the information needed to evaluate an individual's training and experience.

NRC holds the licensee responsible for the radiation protection program. Therefore, it is essential that strong management controls and oversight exist to ensure that licensed activities are conducted properly. Management responsibility and liability are sometimes underemphasized or not addressed in applications and are often poorly understood by licensee employees and managers. Senior management should delegate to the RSO and, if applicable, the RSC, in writing, sufficient authority, organizational freedom, and management prerogative, to communicate with and direct personnel regarding NRC regulations and license provisions and to terminate unsafe activities involving byproduct material. The licensee maintains the ultimate responsibility, nevertheless, for the conduct of licensed activities. In addition, licensees with multiple modalities or multiple users must develop, document, and implement administrative procedures for interdepartmental/interdisciplinary coordination of the radiation protection program (10 CFR 35.24). The licensee may use several mechanisms to ensure that all departments are kept informed of all activities regarding the radiation protection program, such as electronic mail or committee meetings, including an RSC.

Licensees may contract for patient services for which they have no in-house expertise. In those instances in which the contracted service is regulated by the NRC, the licensee should be aware that the licensee remains responsible for regulatory compliance and implementation of the radiation protection program. The licensee should not assume that by hiring a consultant to perform certain tasks it has fully satisfied all regulatory requirements or that it has transferred responsibility for the licensed program to a consultant. Licensee management should ensure that adequate mechanisms for oversight are in place to determine that the radiation protection program is effectively implemented by the appropriate individuals.

Response from Applicant: Refer to the subsequent sections specific to the individuals described above.

8.10 Item 7: Radiation Safety Officer (RSO)

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Regulations: 10 CFR 30.33(a)(3), 10 CFR 35.24, 10 CFR 35.50, 10 CFR 35.57, 10 CFR 35.59, 10 CFR 35.900, 10 CFR 35.2024

Criteria: RSOs must have adequate training and experience. The training and experience requirements for the RSO are described in 10 CFR 35.50 and 10 CFR 35.900 and allow for the following three training pathways:

1. Certification by one of the professional boards approved by the NRC

2. Didactic and work experience as described in item b of each section and passing an examination⁽²⁾

3. Identification as an AU, AMP, or ANP on the license and have experience in the types of use for which the individual has RSO responsibilities²

In addition, 10 CFR 35.24 requires that the licensee provide the RSO sufficient authority, organizational freedom, and management prerogative to do the following:

Identify radiation safety problems
Initiate, recommend, or provide corrective actions
Stop unsafe operations involving byproduct material
Verify implementation of corrective actions.

The licensee must also establish in writing the authority, duties, and responsibilities of the RSO.

Discussion: The RSO is responsible for day-to-day oversight of the radiation protection program. The RSO needs independent authority to stop operations that he or she considers unsafe. He or she must be given sufficient time and resources and have sufficient commitment from management to fulfill certain duties and responsibilities to ensure that radioactive materials are used in a safe manner. NRC requires the name of the RSO on the license to ensure that licensee management has always identified a responsible, qualified person and that the named individual knows of his or her designation as RSO. Usually, the RSO is a full-time employee of the licensed facility; however, NRC has authorized individuals that are not employed by the licensee, such as a consultant, to fill the role of RSO or provide support to the facility RSO. Typical RSO duties are illustrated in Figure 8.7 and described in Appendix F. Appendix F also contains a model RSO Delegation of Authority. Appendix G contains forms which can be used to document the RSO's training and experience.

Figure 8.7 RSO Responsibilities. Typical duties and responsibilities of RSOs.

Response from Applicant: Provide the following:

Name of the proposed RSO

AND

Delegation of Authority

AND

Previous license number (if issued by the NRC) or a copy of the license (if issued by an Agreement State) that authorized the uses requested and on which the individual was specifically named as the RSO.

Copy of the certification(s) for the board(s) approved by the NRC and as applicable to the types of use for which he or she has RSO responsibilities.

Description of the training and experience demonstrating that the proposed RSO is qualified by training and experience as applicable to the types of use for which he or she has RSO responsibilities. NRC Forms 313A and 313B may be used for this purpose (see Appendix G).

AND

Documentation of successful completion of the radiation safety exam (e.g., certificate) [see Note above].

AND

Written certification, signed by a preceptor RSO, that the above training and experience has been satisfactorily completed.

Note:

It is a requirement to notify NRC within 30 days if an RSO permanently discontinues his/her duties under the license or has a name change pursuant to 10 CFR 35.14 and to request an amendment to change an RSO pursuant to 10 CFR 35.13.
An AU, AMP, or ANP may be designated as the RSO on the license if the individual has training and experience with the radiation safety aspects of similar types of use of byproduct material for which he or she has RSO responsibilities and has sufficient time to perform the duties.
Descriptions of training and experience will be reviewed using the criteria listed above. This documentation will be reviewed on a case-by-case basis to determine whether the applicable criteria in Subparts B or J are met. If the training and experience do not appear to meet the Subparts B or J criteria, the NRC may request the assistance of its Advisory Committee on the Medical Uses of Isotopes (ACMUI).

8.11 Item 7: Authorized Users

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Regulations: 10 CFR 30.33(a)(3), 10 CFR 35.14, 10 CFR 35.27, 10 CFR 35.57, 10 CFR 35.59, 10 CFR 35.290, 10 CFR 35.292, 10 CFR 35.390, 10 CFR 35.490, 10 CFR 35.590, 10 CFR 35.690, 10 CFR 35.910, 10 CFR 35.920, 10 CFR 35.930, 10 CFR 35.932, 10 CFR 35.934, 10 CFR 35.940, 10 CFR 35.941, 10 CFR 35.950, 10 CFR 35.960, 10 CFR 35.961.

Criteria: AUs must have adequate training and experience. Successful completion of training as described in 10 CFR 35.57, 10 CFR 35.290, 10 CFR 35.292, 10 CFR 35.390, 10 CFR 35.490, 10 CFR 35.590, 10 CFR 35.690, or in Subpart J, as applicable, is evidence of adequate training and experience.

Discussion: AUs involved in medical use have the following special responsibilities:

Radiation safety commensurate with use of byproduct material
Administration of a radiation dose or dosage and how it is prescribed
Direction of individuals under the AU's supervision in the use of byproduct material
Preparation of WDs, if required.

Technologists, therapists, or other personnel may use byproduct material under an AU's supervision when permitted by Federal, State or local laws. Supervision is addressed in 10 CFR 35.27, "Supervision."

For in vitro and animal research, calibration of survey instruments, and other uses that do not involve the intentional exposure of humans, the list of proposed AUs should include those individuals who will actually be responsible for the safe use of the byproduct material for the requested use. An applicant should note which user will be involved with a particular use by referring to Items 5 and 6 of the application and provide their training and experience. Those AUs may direct the use of byproduct material by users under their supervision for the requested use.

Applicants are reminded of the recentness of training requirements described in 10 CFR 35.59. Specifically, physician-AU applicants must have successfully completed the applicable training and experience criteria described in 10 CFR Part 35 within 7 years preceding the date of the application, or additional training may be necessary. This time restriction applies to board certification as well as to other recognized training pathways.

Response from Applicant: Provide the following:

Name of the proposed AU and uses requested.

AND

Previous license number (if issued by the NRC) or a copy of the license (if issued by an Agreement State) on which the physician was specifically named as an AU for the uses requested.

Copy of the certification(s) for the board(s) approved by the NRC and as applicable to the use requested.

Description of the training and experience demonstrating that the proposed AU is qualified by training and experience for the use requested. NRC Forms 313A and 313B may be used for this purpose (see Appendix G).

AND

Documentation of successful completion of the radiation safety exam (e.g., certificate).

AND

Written certification, signed by a preceptor AU physician, that the above training and experience has been satisfactorily completed.

Note:

It is a requirement to notify NRC within 30 days if an AU permanently discontinues his/her duties under the license or has a name change pursuant to 10 CFR 35.14.
Subpart J will be deleted two years after the publication of the final 10 CFR Part 35. Until then, licensees do not have to comply with the training and experience requirements provided in other sections of 10 CFR Part 35 (e.g., radiation safety exam), with the exception of the section titled "Recentness of training".
Descriptions of training and experience will be reviewed using the criteria listed above. This documentation will be reviewed on a case-by-case basis to determine whether the applicable criteria in 10 CFR Part 35 are met. If the training and experience do not appear to meet the 10 CFR Part 35 criteria, the NRC may request the assistance of its ACMUI.
Authorized use for nonmedical use or for uses that do not involve the intentional exposure of humans (e.g., in vitro and animal research, calibration, dosimetry research) will be reviewed on a case-by-case basis.

8.12 Item 7: Authorized Nuclear Pharmacist

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Regulations: 10 CFR 30.33(a)(3), 10 CFR 35.55, 10 CFR 35.57, 10 CFR 35.59, 10 CFR 35.980.

Criteria: ANPs must have adequate training and experience. Successful completion of training as described in 10 CFR 35.55, 10 CFR 35.57, or 10 CFR 35.980, and 10 CFR 35.59 is evidence of adequate training and experience.

Discussion: An ANP is defined in 10 CFR 35.2, "Definitions." Training and experience criteria for an ANP is described in 10 CFR 35.55 and 10 CFR 35.980, both titled "Training for an authorized nuclear pharmacist," and 10 CFR 35.57, "Training for experienced nuclear pharmacists." At many licensed medical facilities, an ANP is directly involved with the preparation and administration of radiopharmaceuticals.

Applicants are reminded of the recentness of training requirements described in 10 CFR 35.59. Specifically, nuclear pharmacist applicants must have successfully completed the applicable training and experience criteria described in 10 CFR Part 35 within 7 years preceding the date of the application, or additional training may be necessary. This time restriction applies to board certification as well as to other recognized training pathways.

Response from Applicant:

Provide the following:

Name of the proposed ANP.

AND

Previous license number (if issued by the NRC) or a copy of the license (if issued by an Agreement State) on which the individual was specifically named as an ANP.

Copy of the certification(s) for the radiopharmacy board(s) approved by the NRC.

Description of the training and experience demonstrating that the proposed ANP is qualified by training and experience. NRC Forms 313A and 313B may be used for this purpose (see Appendix G).

AND

Documentation of successful completion of the radiation safety exam (e.g., certificate).

AND

Written certification, signed by a preceptor ANP, that the above training and experience has been satisfactorily completed and that the individual has achieved a level of competency sufficient to independently operate a nuclear pharmacy.

Note:

It is a requirement to notify NRC within 30 days if an ANP permanently discontinues his/her duties under the license or has a name change, pursuant to 10 CFR 35.14.
Subpart J will be deleted two years after the publication of the final 10 CFR Part 35. Until then, licensees do not have to comply with the training and experience requirements provided in Subpart B of 10 CFR Part 35 (e.g., radiation safety exam), with the exception of the section titled "Recentness of training".
Descriptions of training and experience will be reviewed using the criteria listed above. This documentation will be reviewed on a case-by-case basis to determine whether the applicable criteria in Subparts B or J are met. If the training and experience do not appear to meet the Subparts B or J criteria, the NRC may request the assistance of its ACMUI.

8.13 Item 7: Authorized Medical Physicist

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Regulations: 10 CFR 30.33(a)(3), 10 CFR 35.51, 10 CFR 35.57, 10 CFR 35.59, 10 CFR 35.961.

Criteria: AMPs must have adequate training and experience. Successful completion of training as described in 10 CFR 35.51, 10 CFR 35.57 or 10 CFR 35.961, and 10 CFR 35.59 is evidence of adequate training and experience.

Discussion: An AMP is defined in 10 CFR 35.2, "Definitions." Training and experience criteria for an AMP are described in 10 CFR 35.51, "Training for authorized medical physicist", or 10 CFR 35.961, "Training for teletherapy physicist", and 10 CFR 35.57, "Training for experienced medical physicist." At many licensed medical facilities conducting radiation therapy treatments, an AMP is directly involved with the calculation and administration of the radiation dose. Additionally, the American Association of Physicists in Medicine (AAPM) suggests that a medical physicist limit his or her involvement in radiation therapy to areas for which he or she has established competency.

Applicants are reminded of the recentness of training requirements described in 10 CFR 35.59. Specifically, medical physicist applicants must have successfully completed the applicable training and experience criteria described in 10 CFR Part 35 within 7 years preceding the date of the application, or additional training may be necessary. This time restriction applies to board certification as well as to other recognized training pathways.

Response from Applicant: Provide the following:

Name of the proposed AMP.

AND

Previous license number (if issued by the NRC) or a copy of the license (if issued by an Agreement State) on which the individual was specifically named as an AMP.

Copy of the certification(s) for the board(s) approved by the NRC.

Description of the training and experience demonstrating that the proposed AMP is qualified by training and experience. NRC Forms 313A and 313B may be used for this purpose (see Appendix G).

AND

Documentation of successful completion of the radiation safety exam (e.g., certificate).

AND

Written certification, signed by a preceptor AMP, that the above training and experience has been satisfactorily completed.

Note:

It is a requirement to notify NRC within 30 days if an AMP permanently discontinues his/her duties under the license or has a name change, pursuant to 10 CFR 35.14.
Subpart J will be deleted two years after the publication of the final 10 CFR Part 35. Until then, licensees do not have to comply with the training and experience requirements provided in Subpart B of 10 CFR Part 35 (e.g., radiation safety exam), with the exception of the section titled "Recentness of training".
Descriptions of training and experience will be reviewed using the criteria listed above. This documentation will be reviewed on a case-by-case basis to determine whether the applicable criteria in Subparts B or J are met. If the training and experience do not appear to meet the Subparts B or J criteria, the NRC may request the assistance of its ACMUI.

8.14 Item 8: Training for Individuals Working in or Frequenting Restricted Areas

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Regulations: 10 CFR 19.12, 10 CFR 35.27, 10 CFR 35.310, 10 CFR 35.410, 10 CFR 35.610, 10 CFR 35.2310.

Criteria: Individuals working with, as well as in the vicinity of, licensed material must have adequate training and experience as required by 10 CFR Parts 19 and 35. For those individuals who work in the vicinity of licensed material and, in the course of employment, are likely to receive in a year an occupational dose of radiation over 1 millisievert (mSv) [100 millirem (mrem)], the licensee must provide training as required by 10 CFR 19.12. Also, 10 CFR 35.310, 10 CFR 35.410 and 10 CFR 35.610 describe additional training requirements for individuals involved with therapeutic treatment of patients. 10 CFR 35.27 requires the licensee's AUs and ANPs to provide training to all personnel using byproduct material under their supervision on the licensee's written procedures.

Discussion: AUs, ANPs, AMPs, and RSOs would be most likely to receive doses in excess of 1 mSv (100 mrem) in a year. However, potential radiation doses received by any individual working in or frequenting restricted areas must also be evaluated. All individuals working with or around licensed materials should receive training commensurate with their assigned duties, and if it is likely they could receive doses over 1 mSv (100 mrem) in a year, they must receive instruction as specified by 10 CFR 19.12. For example, a licensee could determine that housekeeping staff, while not likely to receive doses over 1 mSv (100 mrem), need to be informed of the nature of the licensed material and the meaning of the radiation symbol, and need to be instructed not to touch the licensed material and to remain out of the room if the door to the licensed material storage location is open.

Additionally, since nursing staff often have direct involvement with patients undergoing therapy treatments with byproduct material, the licensee should ensure that nursing staff receive adequate training on the level of risks involved with the particular therapy treatment and emergency response procedures.

In accordance with 10 CFR 35.27(a), individuals working with licensed material under the supervision of an AU must be trained regarding the instructions of the supervising AU for medical uses of licensed material, the written radiation protection procedures established by the licensee, and compliance with NRC regulations and license conditions. In accordance with 10 CFR 35.27(b), an ANP or an AU, as allowed by 10 CFR 35.11(c), shall instruct supervised individuals in the preparation of byproduct material for medical use and require the individuals to follow their instructions and the written radiation protection program, the license conditions, and the NRC regulations.

A model training program is provided in Appendix H.

Response from Applicant: No response is necessary.

8.15 Item 9: Facilities and Equipment

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Regulations: 10 CFR 20.1302, 10 CFR 30.33(a)(2), 10 CFR 35.75.

Criteria: Facilities and equipment must be adequate to protect health and minimize danger to life or property.

Discussion: In 10 CFR 30.33(a)(2), NRC states that an application will be approved if, among other things, the applicant's proposed equipment and facilities are adequate to protect health and minimize danger to life or property. Facility and equipment requirements depend on the scope of the applicants operations (e.g., planned use of the material, the types of radioactive emissions, the quantity and form of radioactive materials possessed, etc.). Particular attention should be focused on operations using large quantities of radioactive materials; preparation steps involving liquids, gases, and volatile radioactive materials; and the use of alpha-emitters, high-energy photon-emitters, and high-energy beta-emitters.

Response from Applicant: Refer to the subsequent sections for guidance.

8.16 Item 9: Facility Diagram

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Regulations: 10 CFR 20.1003, 10 CFR 20.1101, 10 CFR 20.1201, 10 CFR 20.1301, 10 CFR 20.1302, 10 CFR 20.2102, 10 CFR 30.33(a)(2), 10 CFR 35.14, 10 CFR 35.75.

Criteria: Facilities and equipment must be adequate to protect health and minimize danger to life or property.

Discussion: Applicants must submit an annotated drawing of the room or rooms and adjacent areas in which byproduct material will be received, used, administered, and stored. Identify it as ATT. 9.1. This includes rooms used to confine patients awaiting nuclear medicine treatments and patients undergoing in-patient therapy procedures, rooms used for administering radiopharmaceuticals or radiation doses, radioactive waste storage areas, and all byproduct material use areas including those used for receipt and storage of the byproduct material. Pursuant to 10 CFR 20.1302, "Compliance with Dose Limits for Individual Members of the Public," licensees must demonstrate compliance with 10 CFR 20.1301, "Dose Limits for Individual Members of the Public," for unrestricted or controlled areas that are adjacent to rooms in which byproduct material will be received, used, administered, and stored. Figure 8.8 depicts a standard nuclear medicine suite.

Figure 8.8 Facility Diagram for Nuclear Medicine Suite.

Regulatory requirements, the principle of ALARA, good medical care, and access control should be considered when determining the location of a therapy patient's room or a therapy treatment room. Use of byproduct material in a room that is not described in the license application requires prior NRC approval through a license amendment, except for areas of use where byproduct material is used in accordance with 10 CFR 35.100 and 10 CFR 35.200. Licensees must notify the NRC, pursuant to 10 CFR 35.14, within 30 days following changes in areas of use for 10 CFR 35.100 and 10 CFR 35.200 byproduct material. Figure 8.9 represents an overhead view of a radioiodine patient isolation room that contains some of the required elements discussed in the response from applicant section. Figure 8.10 represents an overhead view of a manual brachytherapy patient isolation room. Based on an evaluation of shielding and planned use of each area, the applicant must have determined whether each area adjacent to the treatment room will be maintained as a restricted or an unrestricted area, and must demonstrate compliance with NRC regulations. For portable shields, the licensee should assure proper placement of the shield prior to each treatment. Applicants must also submit cross-sectional diagrams to illustrate areas above and below the facilities used for patient therapy treatments (e.g., other patient rooms, stairwells, nursing stations and waiting areas). The radiation dose levels associated with these areas must be in compliance with 10 CFR 20.1302, "Compliance with dose limits for individual members of the public."

Figure 8.9 Iodine-131 NaI Administration for the Thyroid Carcinoma Patient. The patient is required to be isolated in a private room with a private bath. Note: Applicants must also submit cross-sectional diagrams to illustrate areas above and below the patient's room.

Figure 8.10 Overhead View of Manual Brachytherapy Patient Treatment Room.

Note: Applicants must also submit cross-sectional diagrams to illustrate areas above and below the patient's room.

Figure 8.11 represents a combined HDR and teletherapy suite. Based on an evaluation of shielding and the planned use of each area, the applicant must have determined whether each area adjacent to the treatment room used for all therapies involving sealed sources will be maintained as a restricted or an unrestricted area, and must demonstrate compliance with NRC regulations.

Figure 8.11 Teletherapy and HDR Treatment Room.

Note: Applicants must also submit cross-sectional diagrams to illustrate areas above and below the patient's room.

It may be necessary to restrict use of the teletherapy unit's primary beam if the treatment room's walls, ceiling, or floor will not adequately shield adjacent areas from direct or scattered radiation. Electrical, mechanical, or other physical means (rather than administrative controls) must be used to limit movement or rotation of the unit.

The teletherapy unit should be equipped with electrical or mechanical stops that limit use of the primary beam of radiation so as to ensure compliance with Subpart D of 10 CFR Part 20. Some applicants have found it helpful to have a sample response for guidance. The following is an example of an acceptable response on the use of a rotational unit with an integral beam absorber; the angle orientation convention described above applies.

For the primary beam directed toward the integral beam absorber, electrical or mechanical stops are set so that the primary beam must be centered (within plus or minus 2 degrees) on the integral beam absorber and, in that configuration, the attenuated primary beam may be rotated 360 degrees pointing toward the floor, east wall, ceiling, and west wall.
For the primary beam directed away from the integral beam absorber, electrical or mechanical stops permit the unattenuated primary beam to be directed in a 95 degrees arc from 5 degrees toward the west wall to vertically down toward the floor to 90 degrees toward the east wall.

Experience has shown that, given this type of example, many applicants can make changes to accommodate their own situations (e.g., use of a vertical unit, use of a rotational unit without an integral beam absorber).

Response from Applicant: Provide the following on the facility diagrams:

Scale; use the same scale (preferably 1/4 inch = 1 foot) for all drawings
Direction of north
Location, room numbers and principal use of each room or area where byproduct material is used or stored (e.g., patient therapy rooms, radioactive waste storage, nuclear medicine hot lab, manual brachytherapy source storage room)
Location, room numbers and principal use of each adjacent room (e.g., office, file, toilet, closet, hallway) including areas above and below therapy treatment rooms. Note that areas should be described as restricted or unrestricted as defined in 10 CFR 20.1003
Type, thickness, and density of any necessary shielding applicable to the quantities, form, and emissions of the radionuclide that will be used (note: beta emitters should be shielded using a material with a low atomic number to minimize the production of bremsstrahlung). Include a description of any portable shields used (e.g., shielding of proposed patient rooms used for implant therapy including the dimensions of any portable shield, if one is used; source storage safe, etc.).
Nature of and distances to all areas adjacent to therapy patient treatment rooms (including above and below). Note that plans are particularly helpful in showing the relationship of the patient treatment room with the rest of the building
Location of additional radiation safety equipment (e.g., fume hood, L-block, dose calibrator, fixed area monitors, remote handling tools, t-bars, Allen key) within the facility.

In addition to the above, for remote afterloader, teletherapy, and GSR facilities, provide:

Type, thickness, and density of the shielding materials used on all sides of the treatment room, including the walls, floor, and ceiling
Location of doors, windows, conduits, and other penetrations and voids in the shielding materials
Nature of and distances to all areas adjacent to the treatment room (including above and below), with an indication of whether the areas are restricted or unrestricted, as defined in 10 CFR 20.1003. Note that plans and elevation drawings are particularly helpful in showing the relationship among the treatment room, the roof, and the rest of the building

In addition to the above, for teletherapy and GSR facilities, provide:

Location of the treatment unit and source within the treatment room
Directions of primary beam usage for teletherapy units and, in the case of an isocentric unit, the plane of beam rotation
Height of earth against outside walls, as applicable.

In addition to the facility description for remote afterloader units, provide detailed calculations of maximum radiation levels (and dose rates) that will exist in each area, restricted and unrestricted, adjacent to the room(s) where treatment is performed using a remote afterloader device, to demonstrate compliance with 10 CFR 20.1201 and 10 CFR 20.1301, respectively. (This includes areas above and below the treatment room.) The calculations should include the following:

Expected radiation levels for each area adjacent to the room housing the device considering the most adverse source orientations and maximum source activity to be used in the device. These calculations must be sufficient to demonstrate that the expected dose rates in restricted and unrestricted areas adjacent to the treatment room(s), meet the requirements of 10 CFR 20.1201 and 10 CFR 20.1301.
Parameters (equations, constants, assumptions, etc.) used to perform the calculations described above. These parameters must include such factors as distance to each area of concern, the type and thickness of material(s) used in barriers and shields, the transmission factor of the barriers or shields, and the maximum source strength. For HDR and pulsed dose-rate (PDR) remote afterloader devices, the use of portable shielding to meet the dose rate requirements of 10 CFR 20.1201 and 10 CFR 20.1301 is not permitted.
The maximum anticipated workload data, such as device maximum "on time" per hour and per week and occupancy factors used for all adjacent areas.
Calculations to determine the dose received by individuals present in unrestricted areas should reflect continuous occupancy (i.e., occupancy factor of 1), unless the applicant can justify using a lower value.
Demonstrate that the limits specified in 10 CFR 20.1301(a) will not be exceeded. If the calculations demonstrate that these limits cannot be met, provide further steps that will be taken to limit exposure to individual members of the public. The applicant may consider the following options:

adding shielding to the barrier in question, with corresponding modification of the facility description if necessary
requesting prior NRC authorization to operate up to an annual dose limit for an individual member of the public of 5 mSv (0.5 rem) and demonstrating that the requirements of 10 CFR 20.1301(c) will be met. The applicant must demonstrate the need for and the expected duration of operations which will result in an individual dose in excess of the limits specified in 10 CFR 20.1301(a). A program to assess and control dose within the 5 mSv (0.5 rem) annual limit and procedures to be followed to maintain the dose ALARA (10 CFR 20.1101) must be developed.
decreasing exposure times and/or limiting the number of patient treatments or increasing the size of the restricted area surrounding the treatment room

In addition to the facility description for teletherapy and GSR units, provide:

A copy of the manufacturers calculation of source(s) intensity
Calculations of the maximum radiation levels expected in each adjacent area. The calculations should include the following:

maximum anticipated workload data (e.g., maximum number of patients treated per hour and per week; maximum dose and treatment time per patient; maximum on-time per hour and per week)
the value of each parameter used in your calculations. These parameters include such factors as beam orientation, maximum field size, scatter angle, scatter ratio, distance to scatterer, distance to area of concern, type and thickness of materials used in barrier, and transmission factor of barrier
contributions from primary, leakage (with the source in the on position), and scattered radiation (including low-angle scatter that just misses the integral beam absorber for teletherapy)
calculations for each area adjacent to the treatment room, including above and below the room, and a statement as to whether the area will be maintained as a restricted or unrestricted area. Calculations need not be provided for areas that have not been excavated.
worst case situations (e.g., use of maximum beam size; all patients treated in 1 hour using the critical orientation that produces high radiation levels in an adjacent area; if the teletherapy integral beam absorber is not used for all patient treatments, calculations based on use of the unattenuated primary beam where appropriate; situations within the capabilities of the teletherapy unit that are not prohibited by electrical or mechanical stops, regardless of the clinical usefulness of these orientations)
the dose received by individuals present in unrestricted areas. Consider continuous occupancy (i.e., occupancy factor of 1), unless the licensee can make a compelling argument for using a lower value.

For each unrestricted area, a statement of how the requirements of 10 CFR 20.1301 (a)(1) and (2) will be met
Description of the teletherapy unit's mechanical or electrical beam stops that are operational and restrict beam orientation, specify the direction in which the teletherapy head can be moved, and describe the maximum angle (from vertical) of the beam orientation in each direction. Identify the angle orientation convention (e.g., 0 degrees is vertical toward the floor, 90 degrees is horizontal toward the east wall, 180 degrees is vertical toward the ceiling, and 270 degrees is horizontal toward the west wall). If the teletherapy unit has an integral beam absorber (also called a beam catcher), provide similar information for those orientations in which (1) the primary beam is directed toward the integral beam absorber and (2) the primary beam is directed away from the integral beam absorber. Sketches may be used to describe how beam stops limit the use of the primary beam.

National Council on Radiation Protection and Measurements (NCRP) Report 49, "Structural Shielding Design and Evaluation for Medical Use of X-Rays and Gamma Rays of Energies up to 10 MeV" and Report 102, "Medical X-Ray, Electron Beam and Gamma Ray Protection for Energies up to 50 MeV (Equipment Design, Performance and Use)", may be helpful in responding to the items above.

8.17 Item 9: Radiation Monitoring Instruments

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Regulations: 10 CFR 20.1101, 10 CFR 20.1501, 10 CFR 20.2102, 10 CFR 20.2103(a), 10 CFR 30.33(a)(2), 10 CFR 35.61, 10 CFR 35.2061.

Criteria: All licensees shall possess calibrated radiation detection and measuring instruments that will be used for radiation protection, including survey and monitoring instruments and quantitative measuring instruments needed to monitor the adequacy of radioactive materials containment and contamination control.

Discussion: Licensees shall possess instruments used to measure radiation levels, radioactive contamination, and radioactivity, as applicable. Instruments used for quantitative radiation measurements must be calibrated for the radiation measured. The instruments must be available for use at all times when byproduct material is in use. The licensee must possess survey instruments sufficiently sensitive to measure the type and energy of radiation used including, survey instruments used to locate low energy or low activity seeds (e.g., I-125, Pd-103) if they become dislodged in the operating room or patient's room.

Usually it is not necessary for a licensee to possess a survey meter solely for use during sealed source diagnostic procedures since it is not expected that a survey be performed each time a procedure is performed. In these cases it is acceptable for the meter to be available on short notice in the event of an accident or malfunction that could reduce the shielding of the sealed source(s). Surveys may be required to verify source integrity of the diagnostic sealed source and to ensure that dose rates in unrestricted areas and public and occupational doses are within regulatory limits.

Survey meter calibrations must be performed by persons, including licensee personnel, who are specifically authorized by the NRC or an Agreement State to perform calibrations. If a calibration service will be used, the applicant should ensure that the service is licensed to perform these activities by an NRC (or an equivalent Agreement State) license. Applicants seeking authorization to perform survey meter calibrations must submit calibration facility diagrams in accordance with Section 8.16. Appendix I provides guidance regarding appropriate instrumentation and model survey instrument calibration procedures.

Response from Applicant: Provide the following:

Identify the instrument type, sensitivity, and range for each type of radiation detected. Additionally, if applicants possess only one survey instrument to meet the criteria established in 10 CFR Part 35, they should describe what is done when the survey instrument is being calibrated or repaired and either routine or emergency radiation surveys need to be performed.

AND

Provide one of the following:

A statement that: "Radiation monitoring instruments will be calibrated by a person authorized by the NRC or an Agreement State to perform survey meter calibrations."

AND/OR

A statement that: "We have developed and will implement written survey meter calibration procedures in accordance with the requirements in 10 CFR 20.1101 and that meet the requirements of 10 CFR 35.61."

8.18 Item 9: Dose Calibrator And Other Dosage Measuring Equipment

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Regulations: 10 CFR 30.33, 10 CFR 35.60, 10 CFR 35.62, 10 CFR 35.63, 10 CFR 35.2060, 10 CFR 35.2063.

Criteria: In 10 CFR 35.60, 10 CFR 35.62, and 10 CFR 35.63, NRC describes requirements for the use, possession, calibration, and check of dose calibrators and other equipment used to measure patient dosages.

Discussion: As described in 10 CFR 35.63, dosage measurement is required for licensees who prepare patient dosages. If the licensee uses only unit dosages made by a manufacturer or preparer licensed pursuant to 10 CFR 32.72, the licensee is not required to possess an instrument to measure the dosage. However, pursuant to 10 CFR 35.60, if the licensee prepares their own dosages, breaks up unit dosages for patient administration, or decides to measure unit dosages, the licensee is required to possess and calibrate all instruments used for measuring patient dosages. Appendix J provides model dose calibrator calibration procedures. Currently no alpha emitting nuclides are used in unsealed form in medicine. Therefore guidance is not provided in this document on the measurement of these radionuclides. Also, licensees receiving unit dosages of byproduct material and not splitting the dosages may rely on the provider's dose label for the measurement of the dosage.

Equipment used to measure dosages that emit gamma, alpha, or beta radiation must be calibrated for the applicable radionuclide being measured. For other than unit dosages, the activity must be determined by direct measurement or by a combination of measurement and calculation. However, there are inherent technical difficulties to overcome. For beta-emitting radionuclides these difficulties include, dependence on geometry, lack of an industry standard for materials used in the manufacture of both vials and syringes, and lack of a National Institute of Standards and Technology (NIST) traceable standard for all radionuclides used. Licensees must assay patient dosages in the same type of vial and geometry as used to determine the correct dose calibrator settings. The use of different vials or syringes may result in measurement errors, for example, due to the variation of bremsstrahlung created by interaction between beta particles and the differing dosage containers. Licensees are reminded that beta emitters should be shielded using a low-atomic-numbered material to minimize the production of bremsstrahlung, followed by a high-atomic-numbered material thick enough to attenuate the bremsstrahlung intensity.

Calibrations of dosage measuring equipment may be performed by persons authorized by the NRC or an Agreement State to perform such services. If a calibration service will be used, the applicant must ensure that the service is licensed to perform these activities by an NRC (or an equivalent Agreement State) license.

Response from Applicant: If applicable, provide the following:

A statement that: "Dosage measuring equipment will be calibrated by a person authorized by the NRC or an Agreement State to perform dosage measuring equipment calibrations."

AND/OR

A statement that: "We have developed and will implement written dosage measuring equipment calibration procedures that meet the requirements in 10 CFR 35.60 and 10 CFR 35.62 as applicable."

AND

Identify the instrument type, manufacturer, and model number. Additionally, if only one dose calibrator is possessed, applicants shall describe what is done at their facility when the dose calibrator is being calibrated or repaired and patient dosages need to be measured

8.19 Item 9: Dosimetry Equipment - Calibration and Use

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Regulations: 10 CFR 35.432, 10 CFR 35.630, 10 CFR 35.632, 10 CFR 35.633, 10 CFR 35.635, 10 CFR 35. 642, 10 CFR 35.643, 10 CFR 35.644, 10 CFR 35.645, 10 CFR 35.2432, 10 CFR 35.2630, 10 CFR 35.2632, 10 CFR 35.2633, 10 CFR 35.2635, 10 CFR 35.2642, 10 CFR 35.2643, 10 CFR 35.2645.

Criteria: In the above regulations, NRC describes requirements for verification of source activity or output. To perform this measurement, the applicant must possess appropriately calibrated dosimetry equipment.

Discussion: The applicant must possess a calibrated dosimetry system (e.g., Farmer chamber, electrometer, well-type ionization chamber) which will be used to perform calibration measurements of sealed sources. Dosimetry systems and/or sealed sources used to calibrate dosimetry systems must be calibrated by a laboratory accredited by NIST or AAPM. The licensee must maintain records of such calibrations for the duration of the license.

The licensee must develop and implement procedures governing calibration of sealed sources used for therapy. The procedures must be approved by the licensee's AMP. The calibration procedures described in AAPM Task Group Nos. 21, 40 or 56, and Report 54, or any published protocol approved by a nationally recognized body, as applicable, may be used. At a minimum, the calibration procedures must address the following:

Full calibration measurements of the sealed source(s) shall be conducted by the AMP(s).
The method used to determine the exposure rate (or activity) under specific criteria (i.e., distances used for the measurement, whether the measurement is an "in air" measurement or done using a phantom configuration of the chamber with respect to the source(s) and device, scatter factors used to compute the exposure rate, etc.).
A commitment to maintain a record of calibration measurements and associated calculations.
The frequency for calibration measurements must be specified. Full calibrations must be performed: before first medical use; whenever spot-check measurements indicate that the output differs by more than 5% from the output obtained at the last full calibration; following installation of new sources or reinstallation of the unit in a new location not previously described in the license and before patient treatment is resumed; following device repairs involving sealed sources; and at intervals not to exceed 1 year. Manual brachytherapy sources must only be calibrated initially, prior to use.

Response from Applicant:

Provide the following:

A statement that: "We will calibrate dosimetry equipment in accordance with the requirements in 10 CFR 35.630."

AND

A statement that: "We have developed and will implement a written therapy sealed source calibration procedure that meets the requirements in 10 CFR 35.432, 10 CFR 35.632, 10 CFR 35.633, 10 CFR 35.635, 10 CFR 35. 642, 10 CFR 35.643, 10 CFR 35.644, and 10 CFR 35.645 (as applicable to the type of medical use requested)."

AND

Identify the instrument type, manufacturer, and model number.

References: Copies of AAPM Task Group No. 21, "A Protocol for the Determination of Absorbed Dose from High-Energy Photon and Electron Beams," AAPM Task Group No. 40, "Comprehensive QA for Radiation Oncology," AAPM Report No. 54, "Stereotactic Radiosurgery," AAPM Task Group No. 56, "Code of Practice for Brachytherapy Physics," may be obtained from the American Association of Physicists in Medicine, One Physics Ellipse, College Park, MD 20740-3843 or ordered electronically at the following address: <http://www.aapm.org>.

8.20 Item 9: Other Equipment and Facilities

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Regulations: 10 CFR 30.33(a)(2), 10 CFR Part 35.

Criteria: Facilities and equipment must be adequate to protect health and minimize danger to life or property.

Discussion: The applicant must describe other equipment and facilities available for safe use and storage of byproduct material listed in Item 5 of this application (e.g. fume hoods, xenon traps, emergency response equipment, area monitors, remote handling tools, source transport containers, patient viewing and intercom systems, interlock systems, etc.). Identify it as ATT. 9.4.

Describe additional facilities and equipment for the radiopharmaceutical therapy program to safely receive, use, store, and dispose of radioactive material. Particular attention should be focused on facilities you will use for radioactive drug therapy administration and patient accommodations (i.e., private room/private bath). I-131 sodium iodide is the most widely used source of radiopharmaceutical therapy. If the radionuclide is administered in liquid form, it is important to place the patient dosage in a closed environment (i.e., a fume hood) because of the volatility of the radiopharmaceutical. If the patient has exceeded the release limits of 10 CFR 35.75 the patient must be accommodated in a private room with a private bath as described in Item 8.16 of this report. Sources of patient contamination include airborne I-131 and radioactivity in the patient's urine, perspiration, and saliva.

To facilitate decontamination of the patient room, floors, toilet areas, sink areas, countertops, and other permeable surfaces, the licensee should consider covering areas with disposable materials having plastic on one side and absorbent material on the other. In addition, items handled by the patient may be covered with plastic. These articles would include the telephone, faucet and toilet handles, television remote, door handles, and nurse call buttons. P-32 is effectively shielded by a plastic syringe and once the radionuclide has been administered to the patient, there is no external radiation hazard and P-32 does not require that the patient be placed in isolation. However, P-32 administered in colloidal form can result in contaminated bandages and dressings and designated waste containers should be maintained in the patient room.

For teletherapy, GSR and HDR facilities, the licensee shall require any individual entering the treatment room to ensure, through the use of appropriate radiation monitors, that radiation levels have returned to ambient levels. A beam-on radiation monitor permanently mounted in each therapy treatment room that is equipped with an emergency power supply separate from the power supply for the therapy unit meets the requirements of 10 CFR 35.615(c). In addition, the beam-on monitors traditionally installed in therapy treatment rooms are capable of providing a visible indication (e.g., flashing light) of an exposed or partially exposed source.

The applicant shall describe the system, required by 10 CFR 35.615(d), used to view and communicate with the patient continuously while the patient is in the treatment room. If a shielded viewing window will be used, the thickness, density, and type of material used shall be specified. If a closed-circuit television system (or some other electronic system) will be used for viewing the patient, the backup system or procedure used in case the electronic system malfunctions shall be specified or a commitment must be made to suspend all treatments until the electronic system is repaired and functioning again. The communication system must allow the patient to communicate with the unit operator in the event of any medical difficulties. An open microphone system is recommended to allow communication without requiring the patient to move to activate buttons.

The applicant must also provide adequate equipment and controls to maintain exposures of radiation to workers ALARA and within regulatory limits. 10 CFR 35.615(b), in part, requires that each door leading into the treatment room be provided with an interlock to control the on-off mechanism of the therapy unit. The interlock must cause the source to move to the off condition or shield the source(s) if the door to the treatment room is opened when the source is exposed. The interlock system must prevent the operator from initiating a treatment cycle unless the treatment room entrance door is closed. Additionally, the interlock must be wired so that the source(s) cannot be returned to the on condition after interlock interruption until the treatment room door is closed and the system is reset at the control panel.

Due to the unique characteristics of PDR remote afterloaders and lack of constant surveillance of their operation, a more sophisticated alarm system is essential to ensure protection of the patient during treatment. In addition to the above, it is necessary to ensure the following:

The PDR device's control console is not accessible to unauthorized personnel during treatment
A primary care provider check the patient to ensure that the patients device has not been moved, kinked, dislodged, disconnected, etc.
A more sophisticated interlock/warning system is normally installed for PDR devices. This system should perform the following functions or possess the following characteristics:

The signal from the PDR device must indicate that the source is in the "safe" or retracted position connected with the signal indicating the presence of radiation from the room radiation monitor in such a manner that an audible alarm is generated whenever the room monitor indicates the presence of radiation and the device indicates a safe condition.
The alarm circuit must also be wired in such a manner that the audible alarm is also generated for any device internal error condition that could indicate the unintended extension of the source. This would constitute a circuit that generates the audible alarm when either the "source retracted and radiation present" or appropriate internal error condition(s) exist.
The "source safe and radiation present" signal must also be self-testing. This requires that if a "source not safe" input is received without a corresponding "radiation present" signal, the circuit will generate an interlock/warning circuit failure signal that will cause the source to retract. This circuit will require manual resetting before attempting to continue treatment.
The audible alarm must be sufficiently loud to be clearly heard by the facility's responsible device/patient monitoring staff at all times.
No provisions for bypassing this alarm circuit or for permanently silencing the alarm may be made to the circuit as long as the room radiation monitor is indicating the presence of radiation. If any circuitry is provided to mute the audible alarm, such circuitry may not mute the alarm for a period of more than 1 minute. Controls that disable this alarm circuit or provide for silencing the alarm for periods in excess of 1 minute are prohibited.

If the alarm circuit is inoperative for any reason, a commitment must be made to prohibit initiating any patient treatments with the device until the circuit has been repaired and tested. If the alarm circuit fails during the course of a patient treatment, the treatment in progress may continue as long as continuous surveillance of the device is provided during each treatment cycle or fraction.

For patient rooms where low dose-rate (LDR) remote afterloader use is planned, neither a viewing nor an intercom system is required. However, the applicant must describe how the patient and device will be monitored during treatment to ensure that the sources and catheter guide tube are not disturbed during treatment and to provide for prompt detection of any operational problems with the LDR device during treatment.

Response from Applicant: Provide a description of additional facilities and equipment required by 10 CFR Parts 30 and 35. For teletherapy, GSR, and remote afterloader facilities, include a description of the:

Warning systems and restricted area controls (e.g., locks, signs, warning lights and alarms, interlock systems) for each therapy treatment room
Methods for controlling occupancy for each restricted area
Area radiation monitoring equipment
Viewing and intercom systems (except for LDR units)
Steps that will be taken to ensure that no two units can be operated simultaneously, if other radiation-producing equipment (e.g., linear accelerator, x-ray machine) are located in the treatment room.
Methods to ensure that whenever the device is not in use or is unattended, the console keys will be inaccessible to unauthorized persons.

8.21 Item 10: Radiation Protection Program

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Regulations: 10 CFR 20.1101, 10 CFR 20.2102, 10 CFR 30.34(e), 10 CFR 35.24, 10 CFR 35.26, 10 CFR 35.2024, 10 CFR 35.2026.

Criteria: 10 CFR 20.1101 states that each licensee must develop, document, and implement a radiation protection program commensurate with the scope of the licensed activity. The program must assure compliance with established standards, procedures, and provisions of NRC regulations. Additionally, any calculations or measurements used to demonstrate compliance with NRC regulations must be representative of typical quantities in use or maximum patient doses (dosages). The licensee is responsible for the conduct of all licensed activities and the acts and omissions of individuals handling licensed material. 10 CFR 30.34(e) permits NRC to incorporate into licenses additional requirements and conditions that it deems appropriate or necessary to protect health and safety or to minimize danger to life and property. 10 CFR 35.24 describes the licensees management's authorities and responsibilities for the radiation protection program. 10 CFR 35.26 describes when the licensee may revise the radiation protection program without NRC approval. One of the allowances to change the program without approval is when the revision does not reduce radiation safety. Examples of when this might apply include replacement of survey instruments with comparable survey instruments or reassignment of tasks among employees.

Discussion: Applicants/licensees must abide by all applicable regulations, develop and implement procedures, and/or provide requested information about the proposed radiation protection program during the licensing process. The table in Appendix C may be helpful in determining the information that must be provided when requesting a license. The applicant/licensee must consider the following functional areas (as applicable to the type of medical program):

Audit Program
Occupational Dose
Public Dose
Minimization of Contamination
Operating and Emergency Procedures
Material Receipt and Accountability

Ordering and Receiving
Opening Packages
Sealed Source Inventory
Use Records
Leak Tests
Area Surveys
Administrations Requiring a Written Directive
Safe Use of Unsealed Licensed Material
Maintenance of Therapy Devices Containing Sealed Sources
Spill Procedures
Emergency Response for Sealed Sources or Devices Containing Sealed Sources
Patient or Human Research Subject Release
Safety Procedures for Therapy Treatments Where Patients are Hospitalized
Safety and Device Calibration Procedures

Mobile Medical Service
Transportation
Waste Management

Response From Applicant: Respond to subsequent sections of this document regarding Item 10 of the application.

8.22 Item 10: Audit Program

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Regulations: 10 CFR 20.1101, 10 CFR 20.2102.

Criteria: Licensees must annually review the content and implementation of the radiation protection program to ensure the following:

Compliance with NRC and DOT regulations (as applicable), and the terms and conditions of the license
Occupational doses and doses to members of the public ALARA (10 CFR 20.1101)
Records of audits and other reviews of program content are maintained for 3 years.

Discussion: The licensee must develop and implement procedures for the audit program. Appendix K contains model procedures. All areas indicated in Appendix K may not be applicable to every licensee and may not need to be addressed during each audit. For example, licensees do not need to address areas which do not apply to their activities, and activities which have not occurred since the last audit need not be reviewed at the next audit. Generally, audits are conducted at least once every 12 months.

Currently the NRC's emphasis in inspections is to perform actual observations of work in progress. As a part of their audit programs, applicants should consider performing unannounced audits of users to determine if, for example, Operating and Emergency Procedures are available and are being followed.

It is essential that once identified, problems are corrected comprehensively and in a timely manner. IN 96-28, "Suggested Guidance Relating to Development and Implementation of Corrective Action," dated May 1, 1996, provides guidance on this subject, and specifically describes the corrective action process to include three steps:

1.	Conduct a complete and thorough review of the circumstances that led to the violation
2.	Identify the root cause of the violation
3.	Take prompt and comprehensive corrective actions that will address the immediate concerns and prevent recurrence of the violation.

The NRC will review the licensee's audit results and determine if corrective actions are thorough, timely, and sufficient to prevent recurrence. Depending on the significance of a violation, if a violation is identified by the licensee and these steps are taken, the NRC may exercise discretion and may elect not to cite a violation. The NRC's goal is to encourage prompt identification and prompt, comprehensive correction of violations and deficiencies.

For additional information on NRC's use of discretion on issuing a notice of violation, refer to "General Statement of Policy and Procedures for NRC Enforcement Actions," (NUREG 1600, dated June 1995).

Pursuant to 10 CFR 20.2102, licensees must maintain records of audits and other reviews of program content and implementation for 3 years from the date of the record. Audit records should contain the following information: audit findings, noted deficiencies, and corrective actions.

Response from Applicant: The applicant is not required to, and should not, submit its audit program to the NRC for review.

References: See the Notice of Availability on the inside front cover of this draft report to obtain copies of: NUREG-1600, "General Statement of Policy and Procedures on NRC Enforcement Actions," dated June 1995, and IN 96-28, "Suggested Guidance Relating to Development and Implementation of Corrective Action," dated May 1, 1996. NUREG-1600 is also available on the Internet. Visit NRC's Home Page <http://www.nrc.gov>, choose "Nuclear Materials," then "Enforcement," "Enforcement Guidance Documents," and then "Current Enforcement Policy."

8.23 Item 10: Occupational Dose

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Regulations: 10 CFR 20.1101, 10 CFR 20.1201, 10 CFR 20.1202, 10 CFR 20.1204, 10 CFR 20.1207, 10 CFR 20.1208, 10 CFR 20.1501, 10 CFR 20.1502, 10 CFR 20.2102, 10 CFR 20.2106.

Criteria: Applicants must do either of the following:

Maintain, for inspection by NRC, documentation demonstrating that unmonitored individuals are not likely to receive, in 1 year, a radiation dose in excess of 10% of the allowable limits as shown in Figure 8.12.

Monitor external and/or internal occupational radiation exposure, if required by 10 CFR Part 20.

Figure 8.12 Annual Dose Limits for Radiation Workers.

Discussion: The licensee must evaluate the exposure of all occupational workers (e.g., nurses, technologists, etc.) to determine whether monitoring is required to demonstrate compliance with Subpart F of 10 CFR Part 20. Licensees must consider the internal and external dose and the occupational workers assigned duties when evaluating the need to monitor occupational radiation exposure.

When evaluating external dose from xenon gas, the licensee may take credit for the reduction of dose resulting from the use of xenon traps. Additionally, periodic checks of the trap effluent may be used to ensure proper operation of the xenon trap. Licensees may vent xenon gas directly to the atmosphere as long as the effluent concentration is within 10 CFR Part 20 limits.

When evaluating dose from aerosols, licensees may take credit for the reduction of dose resulting from the use of aerosol traps. Licensees may vent aerosols directly to the atmosphere as long as the effluent concentration is within 10 CFR Part 20 limits.

A model procedure for monitoring occupational exposure is provided in Appendix L.

If external dose monitoring is necessary, the applicant should describe the type of personnel dosimetry, i.e., film badges or thermoluminescent dosimeters (TLDs), that will be used by personnel. If occupational workers handle licensed material, the licensee should evaluate the need to provide extremity monitors [i.e., required if likely to receive a dose in excess of 0.05 Sv (5 rem) shallow-dose equivalent (SDE)] in addition to whole body badges. Additionally, applicants should ensure that their personnel dosimetry program contains provisions that personnel monitoring devices should be worn so that the part of the body likely to receive the greatest proportion of its permissible dose equivalent will be monitored.

Some licensees use self-reading dosimeters in lieu of processed dosimetry. This is acceptable if the criteria above are met. See American National Standards Institute (ANSI) N322, "Inspection and Test Specifications for Direct and Indirect Reading Quartz Fiber Pocket Dosimeters" for more information. If pocket dosimeters are used to monitor personnel exposures, applicants should state the useful range of the dosimeters; and procedures and frequency for calibration and maintenance of pocket dosimeters as required by 10 CFR 20.1501(b).

When personnel monitoring is needed, most licensees use either film badges or TLDs that are supplied by a National Voluntary Laboratory Accreditation Program (NVLAP) approved processor. Film badges are usually exchanged monthly due to technical concerns about film fading. TLDs are usually exchanged quarterly. Applicants must verify that the processor is NVLAP-approved. Consult the NVLAP-approved processor for its recommendations for exchange frequency and proper use.

It may be necessary to assess the intake of radioactivity for occupationally exposed individuals in accordance with 10 CFR 20.1204 and 20.1502. If internal dose monitoring is necessary, the applicant must measure the following:

Concentrations of radioactive material in air in work areas
Quantities of radionuclides in the body
Quantities of radionuclides excreted from the body; or
Combinations of these measurements.

For example, for individuals preparing or administering therapeutic dosages of I-131, licensees may need to assess thyroid burden measurements. For those individuals who are occupationally exposed to lesser quantities of I-131, RG 8.20, "Applications of Bioassay for I-125 and I-131, Revision 1", has suggested frequencies of bioassays for individuals, based on quantities handled, type of compounds (volatile/non-volatile), and facilities used.

The applicant should describe in their procedures the criteria used to determine the type of bioassay and the frequencies at which bioassay (both in vivo and in vitro) will be performed to evaluate intakes. The criteria also should describe how tables of investigational levels are derived, including the methodology used by the evaluated internal dose assessments, i.e., the empirical models used to interpret the raw bioassay data. The bioassay procedures should provide for baseline, routine, emergency, and follow-up bioassays. The applicant must describe the equipment and facilities dedicated to the bioassay program. If a commercial bioassay service will be used, ensure that the service is licensed to perform these activities by an NRC (or an equivalent Agreement State) license.RG 8.9, Revision 1, "Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay Program," and NUREG/CR-4884, "Interpretation of Bioassay Measurements," outline acceptable criteria that may be used by applicants in developing their bioassay programs.

10 CFR 20.1202 describes the requirements for summing external and internal doses. Applicants must ensure that their occupational monitoring procedures include criteria for summing external and internal doses.

Response from Applicant: Provide the following:

A description of facilities and equipment used for monitoring occupational exposure.

AND

A statement that: "We have developed and will implement written procedures for monitoring occupational dose in accordance with 10 CFR 20.1101 and that meet the requirements in Subparts C and F of 10 CFR Part 20."

References: National Institute of Standards and Technology (NIST) Publication 810, "National Voluntary Laboratory Accreditation Program Directory," is published annually and is available for purchase from GPO and on the Internet at the following address: <http://ts.nist.gov/ts/htdocs/210/214/do sim.htm>. Copies of ANSI N322 may be obtained from the American National Standards Institute, 1430 Broadway, New York, NY 10018, or ordered electronically at the following address: <http://www.ansi.org>. See the Notice of Availability on the inside front cover of this draft report to obtain copies of RG 8.20, "Applications of Bioassay for I-125 and I-131, Revision 1," RG 8.9, Revision 1, "Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay Program," and NUREG/CR-4884, "Interpretation of Bioassay Measurements."

8.24 Item 10: Public Dose

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Regulations: 10 CFR 20.1301, 10 CFR 20.1302, 10 CFR 20.1801, 10 CFR 20.1802, 10 CFR 20.2107.

Criteria: Licensees must do the following:

Ensure that licensed material will be used, transported, and stored in such a way that members of the public will not receive more than 1 mSv (100 mrem) in 1 year, and the dose in any unrestricted area will not exceed 0.02 mSv (2 mrem) in any one hour, from licensed operations.
Control and maintain constant surveillance of licensed material that is not in storage and secure stored licensed material from unauthorized access, removal, or use.

Discussion: Members of the public include persons who work or may be near locations where licensed material is used or stored and employees whose assigned duties do not include the use of licensed materials and who work in the vicinity where it is used or stored (see Figure 8.13). Public dose is controlled, in part, by ensuring that licensed material is secure (e.g., located in a locked area) to prevent unauthorized access or use. Some medical use devices containing licensed material are usually restricted by controlling access to the keys needed to operate the devices and/or to keys to the locked storage area. Only AUs and personnel using byproduct material under their supervision should have access to these keys.

Figure 8.13 Proper Security of Licensed Material. Licensed Material should be located away from occupied areas and secured to prevent unauthorized use or removal.

Public dose is also affected by the choice of storage and use locations and conditions. Licensed material may present a radiation field, and must be located so that the public dose in an unrestricted area (e.g., an office or the exterior surface of an outside wall) does not exceed 1 mSv (100 mrem) in a year or 0.02 mSv (2 mrem) in any one hour. Use the concepts of time, distance, and shielding when choosing storage and use locations. Decreasing the time, increasing the distance, and using shielding (i.e., brick, concrete, lead, or other solid walls) will reduce the radiation exposure.

Licensees can determine the radiation levels adjacent to licensed material either by calculations or a combination of direct measurements and calculations using some or all of the following: typical known radiation levels provided by the manufacturer, the "inverse square" law to evaluate the effect of distance on radiation levels, occupancy factor to account for the actual presence of the member of the public, and limits on the use of licensed material. See Appendix M for an example demonstrating that individual members of the public will not receive doses exceeding the allowable public dose limits.

If, after making an initial evaluation, a licensee changes the conditions used for the evaluation (e.g., changes the location of licensed material within a designated room, changes the type or frequency of licensed material use, or changes the occupancy of adjacent areas) then the licensee must perform a new evaluation to ensure that the public dose limits are not exceeded and take corrective action, as needed.

Response from Applicant: No response is required from the applicant in a license application, but this matter will be examined during inspection. During NRC inspections, licensees must be able to provide documentation demonstrating, by measurement or calculation, that the total effective dose equivalent to the individual likely to receive the highest dose from the licensed operation does not exceed the annual limit for members of the public. See Appendix M for examples of methods to demonstrate compliance with public dose limits.

8.25 Item 10: Minimization of Contamination

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Regulations: 10 CFR 20.1406

Criteria: Applicants for new licenses must ensure that facility design and procedures for operation will minimize, to the extent practicable, contamination of the facility and the environment, facilitate eventual decommissioning, and minimize, to the extent practicable, the generation of radioactive waste.

Discussion: All applicants for new licenses need to consider the importance of designing and operating their facilities to minimize the amount of radioactive contamination generated at the site during its operating lifetime and to minimize the generation of radioactive waste during decontamination. This is especially important for licensed activities involving unsealed byproduct material. As described in Item 8.37, "Spill Procedures", cleanup procedures should be implemented for any contamination event.

Sealed sources and devices that are approved by NRC or an Agreement State and located and used according to their SSDR Certificates usually pose little risk of contamination. Leak tests performed as specified in the SSDR Certificate should identify defective sources. Leaking sources must be immediately withdrawn from use and decontaminated, repaired, or disposed of according to NRC requirements. These steps minimize the spread of contamination and reduce radioactive waste associated with decontamination efforts. Other efforts to minimize radioactive waste do not apply to programs using only sealed sources and devices that have not leaked.

Response from Applicant: Provide a description of how facility design and procedures for operation will minimize contamination of the facility and the environment, facilitate eventual decommissioning, and minimize the generation of radioactive waste.

8.26 Item 10: Operating And Emergency Procedures

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Regulations: 10 CFR 19.11(a)(3), 10 CFR 20.1101, 10 CFR 20.1801, 10 CFR 20.1802, 10 CFR 20.2102, 10 CFR 20.2201-2203, 10 CFR 21.21, 10 CFR 30.34(e), 10 CFR 30.50, 10 CFR 35.75, 10 CFR 35.310, 10 CFR 35.315, 10 CFR 35.404, 10 CFR 35.406, 10 CFR 35.410, 10 CFR 35.415, 10 CFR 35.610, 10 CFR 35.615.

Criteria: Before using licensed material, licensees must do the following:

Develop, implement, and maintain specific operating and emergency procedures containing the following elements:

Instructions for opening packages containing licensed material, using licensed material, operating therapy treatment devices, and performing routine maintenance on devices containing sealed sources, according to the manufacturer's written recommendations and instructions and in accordance with regulatory requirements
Instructions for conducting area radiation level and contamination surveys
Instructions for administering licensed material in accordance with the WD
Steps to take, and whom to contact (e.g., RSO, local officials), when the following has occurred: (a) leaking or damaged source, (b) device malfunction and/or damage, (c) licensed material spills, (d) theft or loss of licensed material, or (e) any other incidents involving licensed material
Steps for source retrieval and access control of damaged sealed source(s) and/or malfunctioning devices containing sealed source(s)
Steps to ensure that patient release is in accordance with 10 CFR 35.75
Steps to take in case a therapy patient undergoes emergency surgery or dies
Instructions for calibration of survey and dosage measuring instruments
Periodic spot checks of therapy device units, sources, and treatment facilities
Instructions for radioactive waste management.

AND

Make operating and emergency procedures available to all users
Maintain a current copy of the procedures at each location of use (or, if this is not practicable, post a notice describing the procedures and stating where they may be examined).

The licensee should also consider the following when developing their radiation protection program:

Steps to take to keep radiation exposures ALARA
Steps to maintain licensed material accountability

Discussion: Applicants shall develop operating and emergency procedures that minimize radiation safety risks, while keeping radiation exposures ALARA. These procedures must be specific to the type and form of the licensed material used.

Sealed sources and radiopharmaceuticals used for therapy can deliver significant doses in a short period of time. Unauthorized access to licensed material by untrained individuals could lead to a significant radiological hazard. Therefore, operating procedures will also need to address access control. Many licensees achieve access control by permitting only trained individuals to have access to licensed material (e.g. keys, lock combinations, security badges, etc.). Accountability of licensed material may be ensured by conducting physical inventories, controlling receipt and disposal, and maintaining use records.

In the event that a therapy patient should undergo emergency surgery or die, it is necessary to ensure the safety of others attending the patient. As long as the body remains unopened, the radiation received by anyone near it is due almost entirely to gamma rays. The change in emphasis when an operation or autopsy is to be performed is due to the possible exposure of the hands and face to relatively intense beta radiation. Procedures for emergency surgery or autopsy can be found in section 5.3 of NCRP Report No. 37, "Precautions In The Management of Patients Who Have Received Therapeutic Amounts of Radionuclides." Appendix N also provides model procedures for responding to emergency surgery or death of the therapy patient.

Applicants must develop emergency procedures that address a spectrum of incidents (e.g., major spills, leaking source, medical events, interlock failure, stuck source, etc.).

The NRC must be notified when licensed material in excess of 10 times the quantity specified in Appendix C to part 20 is lost or stolen. The RSO must be proactive in evaluating whether NRC notification is required for any incident involving licensed material. Refer to the regulations (10 CFR 20.2201-20.2203, 10 CFR 30.50, and 10 CFR 21.21) for a description of when notifications are required.

Response from Applicant: No response is necessary. Refer to the subsequent sections for guidance.

Reference: Copies of NCRP Report No. 37, "Precautions In The Management of Patients Who Have Received Therapeutic Amounts of Radionuclides," may be obtained from the National Council on Radiation Protection and Measurements, 7910 Woodmont Avenue, Suite 800, Bethesda, MD 20814-3095, or ordered electronically at the following address: <http://www.ncrp.com>.

8.27 Item 10: Material Receipt And Accountability

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Regulations: 10 CFR 20.1801, 10 CFR 20.1802, 10 CFR 20.1906, 10 CFR 20.2201, 10 CFR 30.34(e), 10 CFR 30.35(g)(2), 10 CFR 30.41, 10 CFR 30.51, 10 CFR 35.67.

Criteria: Accountability of licensed material must be ensured at all times to prevent loss, theft, or misuse.

Discussion: As illustrated in Figure 8.14, licensed materials must be tracked from "cradle to grave" in order to ensure accountability, identify when licensed material could be lost, stolen, or misplaced, and ensure that possession limits listed on the license are not exceeded. Licensees exercise control over licensed material accountability by including the following items (as applicable) in their radiation protection program:

Physical inventories of sealed sources at intervals not to exceed 6 months
Ordering and receiving licensed material
Package opening
Use records.

>Figure 8.14 Material Receipt and Accountability. Licensees must maintain records of receipt, transfer, and disposal and conduct semiannual physical inventories of sealed sources.

Response from Applicant: No response is necessary. Refer to the subsequent sections for guidance.

8.28 Item 10: Ordering And Receiving

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Regulations: 10 CFR 20.1801, 10 CFR 20.1802, 10 CFR 20.1906, 10 CFR 30.34(e), 10 CFR 30.51.

Criteria: 10 CFR 20.1906 describes the requirements for receiving packages containing licensed material. Applicants are reminded that security of licensed material, required by 10 CFR 20.1801 and 10 CFR 20.1802, must be considered for all receiving areas. 10 CFR 30.51 requires licensees to maintain records showing the receipt of byproduct material.

Discussion: Licensees must ensure that the type and quantity of licensed material possessed is in accordance with the license. Additionally, licensees must ensure that packages are secured and radiation exposures from packages are minimized.

Appendix O contains model procedures for ordering and receiving licensed material.

Response from Applicant: No response is necessary.

8.29 Item 10: Opening Packages

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Regulations: 10 CFR 20.1906, 10 CFR 20.2103.

Criteria: Licensees must ensure that packages are opened safely and that the requirements of 10 CFR 20.1906 are met. Licensees must retain records of package surveys in accordance with 10 CFR 20.2103.

Discussion: Licensees must develop and implement procedures for opening packages to ensure that the survey requirements of 10 CFR 20.1906 are met and that radiation exposures are minimized. Appendix P contains model procedures. Applicants are reminded that 10 CFR 20.1906(b) requires, in part, that licensees monitor the external surfaces of a labeled package for radioactive contamination within 3 hours of receipt if it arrives during normal working hours, or not later than 3 hours from the beginning of the next working day, if it arrived after working hours.

Response from Applicant: Provide the following:

A statement that "We have developed and will implement written package opening procedures that meets the requirements of 10 CFR 20.1906."

8.30 Item 10: Sealed Source Inventory

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Regulations: 10 CFR 20.1801, 10 CFR 20.1802, 10 CFR 30.51, 10 CFR 35.67, 10 CFR 35.406, 10 CFR 35.2067.

Criteria: NRC requires the licensee in possession of a sealed source or brachytherapy source to conduct a semi-annual physical inventory of all such sources in its possession. Inventory records must be maintained for 3 years.

Discussion: The licensee must conduct a semiannual physical inventory of all sealed sources and brachytherapy sources in your possession, pursuant to 10 CFR 35.67. Individual GSR sources are exempt from the semiannual physical inventory as stated in 10 CFR 35.67(g). However, the licensee must maintain records of GSR sources receipt and disposal, pursuant to 10 CFR 30.51, to indicate the current inventory of sources at the licensee's facility. The licensee shall retain each inventory in accordance with 10 CFR 35.2067. In addition, 10 CFR 35.406 requires the licensee to make a record of brachytherapy source accountability when removing and returning brachytherapy sources from the storage location.

Response from Applicant: No response is necessary.

8.31 Item 10: Use Records

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Regulations: 10 CFR 30.51, 10 CFR 35.2063, 10 CFR 35.2204, 10 CFR 35.2406.

Criteria: Licensees must record the use of licensed material to reflect proper use and accountability. Records of use must be maintained for 3 years.

Discussion: Licensees are required to make and maintain records of dosage activity prior to medical use. Records of each dosage must be made and include:

Radionuclide
Generic name or its abbreviation or trade name
Lot number
Patient's or human research subject's name and identification number if one has been assigned
Prescribed dosage and activity at the time of dosage determination, or a notation that the total activity is less than 1.1 mega-Bq (30 �Ci)
Date and time of dosage determination
Name of the individual who determined the dosage.

Dosage determination for unit dosages may be made either by direct measurement or by a decay correction based on the determination (e.g., measurement) made by the manufacturer or preparer licensed pursuant to 10 CFR 32.72 or equivalent Agreement State requirements.

If molybdenum concentration is measured pursuant to 10 CFR 35.204, records of molybdenum concentration must be made and include:

Ratio of the measurements expressed as �Ci of molybdenum per mCi of technetium-99m
Date and time of the measurement
Name of the individual who made the measurement.

If the licensee uses manual brachytherapy sources, the following records of use must be made:

Each time temporary implant brachytherapy sources are removed from storage, a record will be made and include the number and activity of sources removed, the time and date they were removed from storage, the location of use, and the name of the individual who removed them from storage.
Each time temporary implant brachytherapy sources are returned to storage, a record will be made and include the number and activity of sources returned, the time and date they were returned to storage, and the name of the individual who returned them to storage.
For permanent implants, a record will be made and include the number and activity of sources removed from storage, the date they were removed from storage, the name of the individual who removed them from storage, the number and activity of sources returned to storage, the date they were returned to storage, the name of the individual who returned them to storage, and the number and activity of sources permanently implanted in the patient or human research subject.

Response from Applicant: No response is necessary.

8.32 Item 10: Leak Tests

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Regulations: 10 CFR 20.1501, 10 CFR 20.2103, 10 CFR 30.53, 10 CFR 35.67, 10 CFR 35.2067, 10 CFR 35.3067.

Criteria: NRC requires testing to determine whether there is any radioactive leakage from sealed sources. Records of test results must be maintained for 3 years.

Discussion: Licensees must perform leak testing of any sealed source or brachytherapy source in accordance with 10 CFR 35.67. Appendix Q provides model leak testing procedures. Pursuant to 10 CFR 35.67, licensees are required to perform leak tests at six month intervals or at other intervals approved by the NRC or an Agreement State and specified in the SSDR certificate and before first use unless accompanied by a certificate indicating that the test was performed within the past 6 months. The measurement of the leak test sample is a quantitative analysis requiring that instrumentation used to analyze the sample be capable of detecting 185 Bq (0.005 �Ci) of radioactivity. Leak tests samples should be collected at the most accessible area where contamination would accumulate if the sealed source were leaking (see Figure 8.15)

Figure 8.15 Leak Test Sample.

The leak test may be performed in house or by a contractor who is authorized to perform leak tests as a service to other licensees.

The licensee or contractor does not need to leak test sources if:

Sources contain only byproduct material with a half-life less than 30 days
Sources contain only byproduct material as a gas
Sources contain 3.7 mega-Bq (100 �Ci) or less of beta-emitting or gamma-emitting material or 0.37 mega-Bq (10 �Ci) or less of alpha-emitting material
Sources contain Ir-192 seeds in nylon ribbon
Sources are stored for less than a 10-year period and not being used. The licensee shall, however, test each such source for leakage before any use or transfer unless it has been leak tested within 6 months before the date of use or transfer.

Response from Applicant: No response is necessary.

References: See the Notice of Availability on inside front cover of this draft report to obtain a copy of Draft RG FC 412-4, "Guide for the Preparation of Applications for the Use of Radioactive Materials in Leak-Testing Services," dated June 1985.

8.33 Item 10: Area Surveys

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Regulations: 10 CFR 20.1003, 10 CFR 20.1101, 10 CFR 20.1301, 10 CFR 20.1302, 10 CFR 20.1501, 10 CFR 20.1801, 10 CFR 20.1802, 10 CFR 20.2102, 10 CFR 20.2103, 10 CFR 20.2107, 10 CFR 35.70, 10 CFR 35.2070.

Criteria: Licensees must do the following:

Ensure that licensed material will be used, transported, and stored in such a way that members of the public will not receive more than 1 mSv (100 mrem) in 1 year, and the dose in any unrestricted area will not exceed 0.02 mSv (2 mrem) in any 1 hour, from licensed operations.
Ensure that licensed material will be used, transported, and stored in such a way that occupational doses to individuals will not exceed the limits specified in 10 CFR 20.1201.
Control and maintain constant surveillance over licensed material that is not in storage and secure licensed material from unauthorized access, removal, or use.

Discussion: Licensees must develop and implement area survey procedures. Appendix R contains model procedures with suggested survey frequencies. At a minimum, licensees must perform daily surveys in all areas where radiopharmaceuticals requiring a WD were prepared for use or administered (i.e., diagnostic activities exceeding 30 �Ci of I-131 and all therapy treatments); however, for administrations, requiring a WD, in patients' rooms, the licensee is not required to perform a survey if the patient is not released, but should perform adequate surveys of patients' rooms after patient release and prior to release of the room for unrestricted use.

In addition, therapy sealed sources (including applicators and catheters) may become dislodged during implantation or after surgery, and inadvertently lost or removed. When developing area survey procedures, the licensee should consider:

a survey of the therapy patient's bed linens before removing them from the patient's room, and
a survey of the operating room and the patient's room after source implantation.

Response from Applicant: Provide the following:

A statement that: "We have developed and will implement written procedures for area surveys in accordance with 10 CFR 20.1101 and that meets the requirements of 10 CFR 20.1501 and 10 CFR 35.70."

8.34 Item 10: Procedures for Administrations Requiring a Written Directive

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Regulations: 10 CFR 35.40, 10 CFR 35.41, 10 CFR 35.2040.

Criteria: 10 CFR 35.40 describes the requirements for WDs. 10 CFR 35.41 requires medical use licensees to develop, maintain, and implement written procedures to provide high confidence that licensed material is administered as directed by authorized users.

Discussion: The procedures do not need to be submitted to the NRC. This allows licensees the flexibility to revise the procedures to enhance their effectiveness without obtaining NRC approval. Appendix S provides guidance for developing the procedures.

Response from Applicant: Provide the following:

A statement that: "We have developed and will implement written procedures for administrations requiring a written directive in accordance with 10 CFR 35.41."

8.35 Item 10: Safe Use of Unsealed Licensed Material

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Regulations: 10 CFR 20.1101, 10 CFR 20.1301, 10 CFR 20.1302, 10 CFR 20.2102, 10 CFR 20.2103, 10 CFR 30.34(e), 10 CFR 35.69, 10 CFR 35.70, 10 CFR 35.310.

Criteria: Before using licensed material, the licensee must develop and implement a radiation protection program that includes safe use of unsealed licensed material.

Discussion: Licensees must have adequate equipment and operating controls to ensure that radioactivity, surface contamination, and effluent releases are maintained within regulatory limits.

Users of licensed material must perform surveys required by 10 CFR 20.1302(a) (i.e., surveys of radiation levels and release of effluents to unrestricted and controlled areas). In addition, applicants should establish a program to constrain doses from air emissions in accordance with 10 CFR 20.1101(d). Records of the results of the measurements are required by 10 CFR 20.2103(b)(4).

Applicants must show how releases to the environment will be ALARA. The general guideline is 10% of the limit specified in 10 CFR 20.1301(a)(1). Licensees that possess sufficient quantities of radioactive material to exceed 10 CFR Part 20 air emissions limits should demonstrate a basis for compliance with the applicable requirements. Such basis could include one or more of the following:

Measured concentrations of radionuclides in air effluents are below 10 CFR Part 20, Appendix B, Table 2 concentrations (and external dose < 50 mrem/yr).
Bounding calculations show that air effluents could not exceed 10 CFR Part 20, Appendix B, Table 2 concentrations (and external dose < 50 mrem/yr).
Dose modeling shows that dose equivalent to the individual likely to receive the highest dose does not exceed 10 mrem/yr.

In addition, licensees must develop procedures for implementing protective measures occupational workers should take to maintain their doses ALARA. Protective measures may include:

Use of syringe shields and/or vial shields
Wearing laboratory coats and gloves when handling unsealed byproduct material
Monitoring hands after handling unsealed byproduct material.

Appendix T contains model procedures for safe use of unsealed licensed material.

Response from Applicant: Provide the following:

A statement that: "We have developed and will implement procedures for safe use of unsealed licensed material that meet the requirements of 10 CFR 20.1101, 10 CFR 20.1301 and 10 CFR 35.69."

8.36 Item 10: Maintenance of Therapy Devices Containing Sealed Sources

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Regulations: 10 CFR 20.1101, 10 CFR 30.34(e), 10 CFR 35.605, 10 CFR 35.655, 10 CFR 35.2605, 10 CFR 35.2655.

Criteria: Licensees must ensure that therapy devices containing sealed sources are maintained according to manufacturer's written recommendations and instructions and according to the SSDR and the regulations. In addition, 10 CFR 35.655 requires that teletherapy and GSR units be fully inspected and serviced during source replacement or at intervals not to exceed 5 years, whichever comes first, to ensure proper functioning of the source exposure mechanism. Maintenance is necessary to ensure that the device functions as designed and source integrity is not compromised.

Discussion: Maintenance and repair includes installation, replacement, and relocation or removal of the sealed source(s) or therapy unit that contains a sealed source(s). Maintenance and repair also means any adjustment involving any mechanism on the therapy device, treatment console, or interlocks that could expose the source(s), reduce the shielding around the source(s), or affect the source drive controls.

The NRC requires that maintenance and repair (as defined above) be performed only by persons specifically licensed by the Commission or an Agreement State to perform such services. Most licensee employees do not perform maintenance and repair because they do not have specialized equipment and technical expertise to perform these activities. Applicants requesting authorization to possess and use LDR remote afterloaders should review 10 CFR 35.605 prior to response to this item. 10 CFR 35.605 allows for limited service activities with regard to LDR remote afterloader units.

Response from Applicant: No response is necessary if the licensee contracts with personnel who are licensed by the Commission or an Agreement State to perform maintenance and repair services on the specific therapy device possessed by the licensee. However, if the applicant requests that an employee, who is trained by the manufacturer, be authorized to perform maintenance and repair, the applicant must submit the following:

Name of the proposed employee and types of maintenance and repair requested.

AND

Description of the training and experience demonstrating that the proposed employee is qualified by training and experience for the use requested.

AND

Copy of the manufacturer's training certification and an outline of the training.

Note: The applicant should specify only those maintenance and repair functions described in a certificate or letter from the manufacturer of the device which documents the employee's training in the requested function(s).

8.37 Item 10: Spill Procedures

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Regulations: 10 CFR 19.11(a)(3), 10 CFR 20.1101, 10 CFR 20.1406, 10 CFR 20.2102, 10 CFR 20.2202, 10 CFR 20.2203, 10 CFR 30.34(e), 10 CFR 30.35(g), 10 CFR 30.50, 10 CFR 30.51.

Criteria: Before using licensed material, the licensee must develop and implement a radiation protection program that includes proper response to spills of licensed material.

Discussion: The applicant must develop and implement procedures to be used in the event of spills or other contamination events in order to prevent the spread of radioactive material. Appendix N contains model emergency response procedures, including model spill procedures. Spill procedures should address all types and forms of licensed material used and should be posted in restricted areas where licensed materials are used or stored. The instructions should specifically state the names and telephone numbers of persons to be notified, e.g., RSO, staff, state and local authorities, and the NRC, when applicable. Additionally, the instructions should contain procedures on evacuation of the area, containment of spills and other releases, appropriate methods for reentering, and decontaminating facilities that may have been accidentally contaminated.

Response from Applicant: Provide the following:

A statement that: "We have developed and will implement written procedures for safe response to spills of licensed material in accordance with 10 CFR 20.1101."

8.38 Item 10: Emergency Response for Sealed Sources or Devices Containing Sealed Sources

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Regulations: 10 CFR 19.11(a)(3), 10 CFR 20.1101, 10 CFR 20.2102, 10 CFR 20.2201-2203, 10 CFR 21.21, 10 CFR 21.51, 10 CFR 30.34(e), 10 CFR 30.50, 10 CFR 30.51, 10 CFR 35.410, 10 CFR 35.610.

Criteria: Before handling sealed sources or using devices containing sealed sources, the applicant must develop and implement procedures for emergency response. 10 CFR 35.610 requires, in part, that instructions and telephone numbers of AUs, AMPs and RSO be posted at the therapy unit console. The instructions must inform the operator of procedures to be followed if the operator is unable to turn off the primary beam of radiation with the controls outside of the treatment room, remove the patient from the radiation field, or if any other abnormal operation occurs.

Discussion: The applicant must develop and implement procedures to be used in response to emergencies involving sealed sources or devices containing sealed sources in order to prevent inadvertent release of, or exposure to, licensed material. Appendix N contains model emergency response procedures for teletherapy units. Emergency procedures must address all types of licensed material and devices used and should be posted in restricted areas where sealed sources are used or stored. The instructions must specifically state the names and telephone numbers of persons to be notified, e.g., RSO, staff, state and local authorities, and the NRC, when applicable. Additionally, the instructions must contain procedures on evacuation and security of the involved area(s), source recovery, area reentry, and decontamination of facilities (if applicable). All equipment necessary for complying with emergency procedures shall be immediately accessible in the treatment room or console area; for example, these may include remote handling tools, t-bars, Allen keys, and shielded containers.

The applicant must establish and agree to follow written procedures for emergencies that may occur, e.g., a manual brachytherapy source becomes dislodged, a therapy source fails to retract or return to the shielded position, a GSR couch fails to retract. A copy of the manufacturer's instructions should be given to each individual involved with therapy treatments. Practice drills, using nonradioactive (dummy) sources (when possible), including dry runs of emergency procedures that cover stuck or dislodged sources and applicators (if applicable), and emergency procedures for patient removal must be practiced annually. These procedures, designed to minimize radiation exposure to patients, workers, and the general public, must at a minimum address the following as applicable to the type of medical use:

When the procedures are to be implemented, such as, any circumstance under which the source becomes dislodged, cannot be retracted to a fully shielded position, or the patient cannot be removed from the beam of radiation
The actions specified for emergency source recovery or shielding which primarily consider minimizing exposure to the patient and healthcare personnel while maximizing safety of the patient.
The step-by-step actions for single or multiple failures that specify the individual(s) responsible for implementing the actions. The procedures must clearly specify which steps are to be taken under different scenarios. The procedure must specify situations in which surgical intervention may be necessary and the steps that should be taken in the event that surgical intervention is required.
Identification of the location of emergency source recovery equipment and specification of what equipment may be necessary for the various scenarios described in the procedure. At a minimum, emergency equipment must include shielded storage containers, remote handling tools, and, if appropriate, supplies necessary to surgically remove applicators or sources from the patient.
Giving first consideration to minimizing the exposure to the patient, usually by removing the patient from the room (rather than using tools to attempt to return the source to the off position). Note: If a first step of the emergency procedures for teletherapy units specifies pressing the emergency bar on the teletherapy unit console, the applicant is advised that this action may cause the source to return to the off position but may also cut power to the entire teletherapy unit or to the gantry or the couch.
Instructing the staff to act quickly and calmly, and to avoid the primary beam of radiation.
Specifying who is to be notified.
Requirements to restrict (lock, as necessary) and post the treatment area with appropriate warning signs as soon as the patient and staff are out of the treatment room.

Response from Applicant: Provide the following:

A statement that: "We have developed and will implement written procedures for safe response to emergencies involving sealed sources in accordance with 10 CFR 20.1101 and 10 CFR 35.610 (if applicable)."

8.39 Item 10: Patient or Human Research Subject Release

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Regulations: 10 CFR 35.75, 10 CFR 35.2075.

Criteria: Licensees may release from confinement patients or human research subjects (patients) who have been administered licensed material if the TEDE to any other individual from exposure to the released patient is not likely to exceed 5 mSv (0.5 rem). Licensees must provide radiation safety instructions to patients released in accordance with 10 CFR 35.75 (b).

Discussion: 10 CFR 35.75 requires that the licensee provide the released individual with instructions, including written instructions, on actions recommended to maintain doses to other individuals ALARA if the TEDE to any other individual is likely to exceed 1 mSv (0.1 rem). If the dose to a breast-feeding infant or a child could exceed 1 mSv (0.1 rem), assuming there was no interruption of breast-feeding, the instructions shall also include:

guidance on the interruption or discontinuation of breast-feeding; and
information on the potential consequences of failure to follow the guidance. This implies that the licensee will confirm whether a patient is breast-feeding prior to release of the patient.

In addition, 10 CFR 35.75 (c) requires that the licensee maintain a record of the basis for authorizing the release of an individual, for 3 years after the date of release, if the TEDE is calculated by:

using the retained activity rather than the activity administered;
using an occupancy factor less than 0.25 at 1 meter;
using the biological or effective half-life; or
considering the shielding by tissue.

In 10 CFR 35.75 (d), the licensee is required to maintain a record, for 3 years after the date of release, that instructions were provided to a breast-feeding woman if the radiation dose to the infant or child from continued breast-feeding could result in a TEDE exceeding 5 mSv (0.5 rem).

Appendix U provides guidance to the applicant for determining when:

The licensee may authorize the release of a patient who has been administered radiopharmaceuticals or treated with implants containing radioactive material (Section 1)
Instructions to the patient are required by 10 CFR 35.75(b) (Section 2)
Records are required by 10 CFR 35.75(c) and (d) to be generated and maintained (Section 3).

The appendix lists activities for commonly used radionuclides and their corresponding dose rates with which a patient may be released in compliance with the dose limits in 10 CFR 35.75.

Response from Applicant: Provide the following:

A statement that: "We have developed and will provide written instructions to patients or human research subjects, released pursuant to 10 CFR 35.75, that meet the requirements in 10 CFR 35.75."

8.40 Item 10: Safety Procedures For Treatments Where Patients Are Hospitalized

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Regulations: 10 CFR 20.1501, 10 CFR 20.1801, 10 CFR 20.2103, 10 CFR 35.315, 10 CFR 35.404, 10 CFR 35.415, 10 CFR 35.615, 10 CFR 35.2404.

Criteria: Applicants must develop and implement procedures to ensure that access to therapy treatment rooms and exposure rates from therapy treatments are limited to maintain doses to occupational workers and members of the public ALARA.

Discussion: 10 CFR 35.315, 10 CFR 35.415, and 10 CFR 35.615 (LDR and PDR) requires the licensee to take certain safety precautions regarding radiopharmaceutical therapy, manual brachytherapy, or remote afterloader brachytherapy involving patients hospitalized in accordance with 10 CFR 35.75. The precautions are to ensure compliance with the exposure limits in 10 CFR Part 20.

10 CFR 35.404(b) requires licensees to perform a radiation survey of the patient immediately after removing the last source from the patient. This is done to confirm that all sources have been removed and accounted for. A record of the patient survey must be maintained for 3 years. 10 CFR 35.615(e) requires that licensed activities where sources are placed within the patient's body be limited to treatments which allow for expeditious removal of a decoupled or jammed source.

Applicants must consider the following elements:

Provide a private room with a private sanitary facility for patients treated with a radiopharmaceutical therapy dosage
Post the patient's or the human research subject's door with a "Radioactive Materials" sign and note on the door or in the patient's or human research subject's chart where and how long visitors may stay in the patient's or the human research subject's room
Either monitor material and items removed from the patient's or the human research subject's room (e.g., patient linens, surgical dressings, etc.) to determine that their radioactivity cannot be distinguished from the natural background radiation level or to confirm that they do not contain brachytherapy sources with a radiation detection survey instrument set on its most sensitive scale and with no interposed shielding, or handle them as radioactive waste
Notify the RSO and AU as soon as possible, if the patient or human research subject has a medical emergency and, immediately, if the patient dies.

10 CFR 20.1501 requires licensees to perform adequate surveys to evaluate the extent of radiation levels. Therefore, licensees must evaluate the exposure rates around patients who are hospitalized in accordance with 10 CFR 35.75 following the dosage administration or implant (e.g., measured exposure rates, combination of measured and calculated exposure rates).

10 CFR 20.1801 requires licensees to secure licensed material in storage from unauthorized access or removal. Therefore, licensees must ensure that access to rooms where patients are hospitalized, in accordance with 10 CFR 35.75, is limited to authorized personnel. Access control and appropriate training of authorized personnel may prevent unauthorized removal of licensed material and unnecessary personnel exposures.

In order to control exposures to individuals in accordance with 10 CFR Part 20, the licensee should consider briefing patients on radiation safety procedures for confinement to bed, visitor control, identification of potential problems, notification of medical staff in the event of problems, and other items as applicable and consistent with good medical care.

Response from Applicant: No response is necessary.

8.41 Item 10: Safety And Device Calibration Procedures

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Regulations: 10 CFR 20.1101, 10 CFR 20.1301, 10 CFR 20.2102, 10 CFR 20.2103, 10 CFR 20.2107, 10 CFR 30.34(e), 10 CFR 35.604, 10 CFR 35.610, 10 CFR 35.615, 10 CFR 35.632, 10 CFR 35.633, 10 CFR 35.635, 10 CFR 35.642, 10 CFR 35.643, 10 CFR 35.644, 10 CFR 35.645, 10 CFR 35.652, 10 CFR 35.655, 10 CFR 35.657, 10 CFR 35.2310, 10 CFR 35.2404, 10 CFR 35.2605, 10 CFR 35.2632, 10 CFR 35.2633, 10 CFR 35.2635, 10 CFR 35.2642, 10 CFR 35.2643, 10 CFR 35.2645, 10 CFR 35.2652, 10 CFR 35.2655.

Criteria: Applicants must develop and implement procedures for providing radiation safety for the use of sealed sources in devices. Applicants must also develop and implement procedures to ensure that therapy sources and devices are calibrated and operating correctly.

Procedures should be complete and self-contained. Pertinent information contained in equipment manuals and other publications may be extracted and included in your operating procedures. Applicable AAPM documents may be referenced.

Discussion: Provided below is a separate discussion for each functional area where sealed sources in devices are used. The applicant should review the functional area(s) that apply to the type of medical use requested.

Diagnostic Sealed Sources and Devices

Good health physics practice dictates that the applicant will provide personnel with clear and specific instructions on usage of sealed sources or devices containing sealed sources. These procedures may include but are not limited to:

Service and repair of the device
Routine proper use for sealed sources or devices containing sealed sources in accordance with the NRC license and 10 CFR Parts 19 and 20
Reviews and discussions of the sealed sources "device specific" manufacturer literature and instructions
Description of checks performed on the device to verify its proper operation after it has been moved and before it is used on patients. This should include manufacturer instruction for start-up, warm-up time and phantom analysis for bone mineral analyzers
Safety and security measures

Teletherapy and GSR Sealed Sources and Devices

Good health physics practice dictates that the applicant will provide personnel with operating procedures to give them clear and specific directions in their duties and responsibilities. These duties may include, safety device checks, instrument calibration, periodic spot checks, quality control checks, and leak tests. Operating procedures should not contain information that does not apply specifically to persons to whom they are directed. For example, housekeeping personnel would not follow the same procedures as therapy technologists. Applicants shall develop and implement procedures to ensure that access to therapy treatment rooms and exposure rates from therapy treatments are limited to maintain doses to occupational workers and members of the public ALARA. The applicant must establish and agree to adhere to written procedures governing the operation of the therapy unit. The applicant shall have written operating procedures specifically developed for and given to particular groups of staff members (e.g., therapists) outlining the responsibilities of each group to ensure your facility's compliance with NRC regulations, the terms and conditions of the license, and the commitments made in license applications and correspondence with NRC. The procedures must include:

Use of the therapy unit, including security of the device, the console and the console keys
Safety device checks
Periodic spot check measurements
Inspection and servicing
Full calibration measurements
Relocation of unit
Recordkeeping

The functional areas listed above are described in more detail below.

Use of the Therapy Unit

The operating procedures should specify who may operate the unit, how the unit may be used (i.e., in what orientations, for what purposes), typical treatment times and setups, how the unit is to be operated (i.e., the sequence of steps to be followed to begin treatment), and who must be present during the treatment. For example, the AU and AMP must be present for GSR treatments. The operating procedures shall contain instructions to ensure that the patient is alone in the room when the primary beam is on and may specify certain daily checks of the unit to ensure its proper operation.

Safety Device Checks

Safety devices shall be checked periodically to ensure that they are operating properly. Such devices include timers, mechanical and electrical interlocks, warning lights and alarms, helmet position indicator microswitches, safety switches, door interlocks, beam collimators, and other devices that actively warn of, limit, or prevent radiation exposure to either patients or personnel. The frequencies required by the regulations for each safety device varies from daily to annually. The results shall be recorded. The operating procedures should contain instructions for making the checks, the frequency of such checks, prompt correction of any malfunctions or defects noted, and retention of appropriate records. A simple checklist may be used to complete the task and recordkeeping quickly and efficiently.

When checks of safety devices indicate defects or malfunctions, there may be some delay before the defects or malfunctions can be corrected. The operating procedures should describe the steps that personnel will follow should a delay occur. For example, use of the therapy unit might be prohibited until the problem is corrected.

Documents such as ANSI N449.1-1978, "Procedures for Periodic Inspection of Cobalt-60 and Cesium-137 Teletherapy Equipment", NCRP Report 69, "Dosimetry of X-Ray and Gamma-Ray Beams for Radiation Therapy in the Energy Range 10 keV to 50 MeV," NUREG/CR-6323, "Relative Risk Analysis in Regulating the Use of Radiation-Emitting Medical Devices", NUREG/CR-6324, "Quality Assurance for Gamma Knives" AAPM Report No. 54, "Stereotactic Radiosurgery" provide standards and recommendations for the frequency and procedures for making certain tests. If the standards or recommendations in these documents conflict with NRC regulations or license conditions, the minimum acceptable frequency is that specified in the regulation or license condition.

Relocation of Therapy Unit

10 CFR 35.13 requires that the NRC approve your plans and proposed location before a therapy unit is relocated. The operating procedures should ensure that the necessary amendment to the NRC license is obtained before the therapy unit is relocated.

Inspection and Servicing of the Therapy Unit

10 CFR 35.655 requires that teletherapy and GSR units be fully inspected and serviced during source replacement or at intervals not to exceed 5 years, whichever comes first. This work, to ensure proper functioning of the source exposure mechanism, must be done by a person or firm specifically licensed to do so by the NRC or an Agreement State. Preventive maintenance should also be addressed to ensure that as systems deteriorate from use they are identified and repaired. These items related to the GSR should be included in these sections, such as hydraulic system maintenance; collimator helmet supports, holes, plugs, bushings, and other helmet positioning equipment; and the systems related to the patient couch and the shielding door. Persons holding an Agreement State license are granted a general license to perform the same activities in non-Agreement States, pursuant to the requirements of 10 CFR 150.20. Operating procedures shall be sufficient to ensure compliance with NRC regulations.

Periodic Spot-Check Measurements of Teletherapy Units

10 CFR 35.642 specifies that output spot-check tests must be performed once in each calendar month and 10 CFR 35.630 describes the characteristics of a properly calibrated dosimetry system needed to make the output measurements. 10 CFR 35.642 also describes spot-checks that must be performed at each source installation. The operating procedures should specify when, how, and by whom the spot-check measurements will be made. The measurements required shall be performed in accordance with procedures established by an AMP. The AMP need not actually perform the spot-check measurements; however, the AMP must review the spot-check measurements within 15 days, as required by 10 CFR 35.642(c).

Teletherapy Full Calibration Measurements

10 CFR 35.632 requires that a licensee authorized to use a teletherapy unit for medical use shall perform full calibration measurements on each teletherapy source before the first medical use of the unit and under the conditions listed in 10 CFR 35.632(a)(2). Operating procedures shall be sufficient to ensure compliance with 10 CFR 35.632.

Periodic Spot-Check Measurements of GSR Units

10 CFR 35.645 specifies that output spot-check tests must be performed once in each calendar month, and 10 CFR 35.630 describes the characteristics of a properly calibrated dosimetry system needed to make the output measurements. 10 CFR 35.645 also describes additional spot-checks that must be performed monthly, daily, and after each source exchange. The operating procedures should specify when, how, and by whom the spot-check measurements will be made. The measurements required shall be performed in accordance with procedures established by an AMP. The AMP need not actually perform the spot-check measurements; however the AMP must review the spot-check results within 15 days, as required by 10 CFR 35.645(b).

GSR Full Calibration Measurements

10 CFR 35.635 requires that a licensee authorized to use a GSR unit for medical use shall perform full calibration measurements on each GSR source before the first medical use of the unit and under the conditions listed in 10 CFR 35.635(a)(2). Operating procedures shall be sufficient to ensure compliance with 10 CFR 35.635.

Remote Afterloader

Good health physics practice dictates that the applicant will provide personnel with operating procedures to give them clear and specific directions in their duties and responsibilities. These duties may include, safety device checks, instrument calibration, periodic spot checks, quality control checks, and leak tests. Operating procedures should not contain information that does not apply specifically to persons to whom they are directed. For example, housekeeping personnel would not follow the same procedures as therapy technologists. Applicants shall develop and implement procedures to ensure that access to therapy treatment rooms and exposure rates from therapy treatments are limited to maintain doses to occupational workers and members of the public ALARA. The applicant must establish and agree to adhere to written procedures governing the operation of the therapy unit. The applicant shall develop written operating procedures specifically for and given to particular groups of staff members (e.g., therapists) outlining the responsibilities of each group to ensure your facility's compliance with NRC regulations, the terms and conditions of the license, and the commitments made in license applications and correspondence with NRC. The procedures must include:

Conduct of surveys following source replacement in remote afterloader devices and when the HDR/PDR device location changes significantly from that existing during previous surveys. At a minimum, the survey program must include surveys defined in the SSDR to ensure that the maximum and average radiation levels from the surface of the remote afterloader device safe do not exceed the levels stated in the SSDR with the source(s) in the shielded position.
Conduct of surveys incident to use including patient and device surveys performed immediately after each use of the device, or upon responding to a device alarm, to ensure that the source has been returned to the fully shielded position. The patient must be surveyed, pursuant to 10 CFR 35.604, with a portable radiation detection survey instrument to confirm that the source(s) has been removed and returned to the safe shielded position. The survey instrument should be capable of measuring dose rates of 1 - 1000 mrem per hour at 1 meter.
Steps to ensure that:

The device, console, and treatment or storage room will be secured when unattended.
The patient will be alone in the treatment room, unless contraindicated, during HDR/PDR therapy.
Nursing personnel follow the AU's and RSO's specific instructions regarding care to be provided to a patient during the treatment process. If the treatment is to be conducted over a period of several hours and direct patient care will be required, such instructions should be provided to the nursing staff in writing.
Treatment planning computer systems using removable media to store each patient's treatment parameters for direct transfer to the treatment system should have each card labeled with the corresponding patient's name and identification number. Such media may be reused (and should be relabeled) in accordance with the manufacturer's instructions.
A treatment procedure will not be conducted if a decoupled or jammed source cannot be removed expeditiously from the patient and placed in a shielded container.
During patient treatments, require the AU and the AMP to be present as defined in 10 CFR 35.615.
If the interlock system malfunctions, the device will be locked in the off position and not used, except as may be necessary for repair or replacement of the interlock system, until the interlock system is shown to be functioning properly.
If a radiation monitor used to indicate the source position is found to be either inoperable or intermittently inoperable, the device will be locked in the off position and not used, except as may be necessary for repair or replacement, until the radiation monitor is shown to be functioning properly.

10 CFR 35.643 and 10 CFR 35.644 require that a licensee authorized to use a remote afterloader unit for medical use shall perform periodic spot-checks on each unit and facility. Operating procedures shall be sufficient to ensure compliance with the regulations.
10 CFR 35.633 requires that a licensee authorized to use a remote afterloader unit for medical use shall perform full calibration measurements on each unit before the first medical use of the unit and under the conditions listed in 10 CFR 35.633(a)(2). Operating procedures shall be sufficient to ensure compliance with 10 CFR 35.633.

Recordkeeping

The licensee must maintain certain records to comply with NRC regulations, the conditions of your license, and commitments made in your license application and correspondence with NRC. Operating procedures should identify which individuals within your organization are responsible for maintaining which records. Examples of documents that must be maintained include:

Copies of NRC licenses, your license applications, and correspondence with the NRC in support of a license request
Personnel dosimetry records
Records of survey instrument calibrations
Records of calibration of the dosimetry system used for full calibration measurements
Records of calibration or intercomparison of the dosimetry system used for spot-check measurements
Results of full calibration measurements
Results of spot-check measurements
Results of leak tests
Records of training of new personnel and annual refresher training of personnel
Records of training in emergency procedures
Records of full inspection and servicing of the therapy unit
Records of receipt and disposal of radioactive material

Response from Applicant: Provide the following:

A statement that: "We have developed and will implement written procedures for safe use of sealed sources and calibration of sources in accordance with 10 CFR 20.1101 and that meet the requirements of the applicable section(s) of Subpart H.

References: Copies of ANSI N449.1-1978, "Procedures for Periodic Inspection of Cobalt-60 and Cesium-137 Teletherapy Equipment," may be obtained from the American National Standards Institute, 1430 Broadway, New York, NY 10018, or ordered electronically at the following address: <http://www.ansi.org>. Copies of NCRP Report 69, "Dosimetry of X-Ray and Gamma-Ray Beams for Radiation Therapy in the Energy Range 10 keV to 50 MeV," may be obtained from the National Council on Radiation Protection and Measurements, 7910 Woodmont Avenue, Suite 800, Bethesda, MD 20814-3095, or ordered electronically at the following address: <http://www.ncrp.com>. See the Notice of Availability on the inside front cover of this draft report to obtain copies of NUREG/CR-6323, "Relative Risk Analysis in Regulating the Use of Radiation-Emitting Medical Devices," and NUREG/CR-6324, "Quality Assurance for Gamma Knives." Copies of AAPM Report No. 54, "Stereotactic Radiosurgery," may be obtained from the American Association of Physicists in Medicine, One Physics Ellipse, College Park, MD 20740-3843 or ordered electronically at the following address: <http://www.aapm.org>.

8.42 Item 10: Mobile Medical Service

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Regulations: 10 CFR 20.1101, 10 CFR 30.41, 10 CFR 30.51, 10 CFR 35.80, 10 CFR 35.647, 10 CFR 35.2080, 10 CFR 35.2647, 10 CFR 71.5, 10 CFR 71.12, 10 CFR 71.13, 10 CFR 71.14, 10 CFR 71.37, 10 CFR 71.38, Subpart H of 10 CFR Part 71, 10 CFR 150.20, 49 CFR Parts 171-178.

Criteria: In addition to the requirements in 10 CFR 35.80, mobile medical service licensees must comply with all other applicable regulations.

Discussion: Appendix V describes specific licensing items pertaining to mobile services. The temporary job site at medical care facilities (client's address) is where mobile medical service licensees use byproduct material. Mobile medical service licensees may transport licensed material and equipment from the van into a client's building, or may bring patients into the van. In either case, the van should be located on the client's property that is under the client's control. In-van imaging services may not be considered an NRC licensed activity if services are limited to patient imaging (i.e., byproduct material is not administered), and byproduct material is not possessed or used.

Self-contained mobile service involves a mobile treatment or administration facility that provides ready to deliver mobile services on arrival at a client's site. The facility is entirely self-contained with a shielded treatment or administration area, remote afterloader device (if applicable), and safety equipment (e.g., dose calibrators, patient viewing systems, intercom, etc.).

Transportable mobile service involves transport of the byproduct material for use in a pre-existing shielded treatment or administration facility at the client site. The mobile service licensee may provide the byproduct material, associated equipment, and trained personnel, or the client may choose to provide the trained personnel to use the byproduct material. Patient treatments with remote afterloaders for this type of service require prior installation of the device in an appropriately shielded and permanently located treatment room. Other support equipment, such as viewing systems, area monitors, and intercoms, must have been separately installed and available for use in the treatment room prior to commencing treatment of patients.

Class 1 (byproduct material, trained personnel, and facility) mobile service providers are authorized to provide the device/facility (e.g., in-van use) and the treatment of (or administration to) patients at the client site. Class 1 mobile service providers are responsible for all aspects of byproduct material use and authorized patient treatments (or administrations).

Class 2 (byproduct material and trained personnel) mobile service providers are authorized to provide the transportation to and use of the byproduct material within the client's facility. Class 2 mobile service providers are also responsible for all aspects of byproduct material use and authorized patient treatments (or administrations).

Class 3 (byproduct material only) mobile service providers are authorized to provide the byproduct material to a client site so that the client can perform treatments (or administrations). Under this class of service, the mobile service provider authorization is limited to the possession, limited servicing, and transport of the byproduct material and associated equipment. The client will need a separate authorization (license) to perform patient treatments (or administrations) with the byproduct material and the client will be responsible for all aspects of byproduct material use and patient treatment(s) (or administrations), as applicable, including, but not limited to, dose calibrator measurements, sealed source calibration, remote afterloader device function checks and all safety system checks.

A mobile service provider may apply for one or multiple classes of service. However, a single client site may be authorized for only a single class of service. This restriction on client sites is intended to eliminate possible confusion that may arise over responsibilities for use and control of byproduct materials that could arise at client sites authorized for multiple classes of service.

For Class 1 and Class 2, mobile medical service licensees must ensure that patients treated meet the release criteria in 10 CFR 35.75.

Note: NRC licensees requesting reciprocity for activities conducted in Agreement States are subject to the general license provisions of equivalent Agreement State regulations as described in 10 CFR 150.20. This general license authorizes persons holding a specific license from the NRC to conduct the same activity in Agreement States if the specific license issued by the NRC does not limit the authorized activity to specific locations or installations. NRC licensees who wish to conduct operations at temporary job sites in an Agreement State should contact that State's radiation control program office for information about State regulations. To ensure compliance with Agreement State reciprocity requirements, a licensee shall request authorization well in advance of scheduled work. In addition to the requirements specified in 10 CFR 150.20, applicants performing procedures through a mobile service should contact the applicable State regulatory agency to determine if mobile services are allowed within the State through reciprocity and to clarify requirements associated with the authorization to practice medicine within the State jurisdiction.

Response from Applicant: Refer to Appendix V for the type of additional information to provide.

8.43 Item 10: Transportation

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Regulations: 10 CFR 20.1101, 10 CFR 30.41, 10 CFR 30.51, 10 CFR 71.5, 10 CFR 71.12, 10 CFR 71.13, 10 CFR 71.14, 10 CFR 71.37, 10 CFR 71.38, Subpart H of 10 CFR Part 71, 49 CFR Parts 171-178.

Criteria: Applicants should develop, implement, and maintain safety programs for transport of radioactive material to ensure compliance with NRC and DOT regulations.

Discussion: The general license in 10 CFR 71.12 provides the authorization used by most licensees to transport, or offer for transport, packages of radioactive material and specifies certain conditions. Most packages of licensed material for medical use contain quantities of radioactive material that require use of Type A packages.

Some medical use licensees (e.g., teletherapy or gamma stereotactic radiosurgery) may need to ship licensed material in Type B packages. Before offering a Type B package for shipment, the licensee needs to be registered as a user of the package and have an NRC-approved quality assurance (QA) plan, two of the requirements under the 10 CFR 71.12 general license. For information about QA plans, see Revision 1 of RG 7.10, "Establishing Quality Assurance Programs for Packaging Used in the Transport of Radioactive Material," dated June 1986. For further information about registering as a user of a package or submitting a QA program for review, contact NRC's Spent Fuel Project Office by calling NRC's toll-free number 800-368-5642 and asking for extension 415-8500. For information about associated fees, contact NRC's OCFO by calling NRC's toll-free number 800-368-5642 and asking for extension 415-7544.

Some medical use licensees that ship radioactive material have chosen to transfer possession of radioactive materials to a manufacturer (or service licensee) with an NRC or Agreement State license who then acts as the shipper. The manufacturer (or service licensee), who is subject to the provisions of 10 CFR 71.12 or 10 CFR 71.14, as appropriate, then becomes responsible for proper packaging of the radioactive materials and compliance with NRC and DOT regulations. Licensees who do this must ensure that the manufacturer (or service licensee):

Is authorized to possess the licensed material at temporary job sites (e.g., at the licensee's facilities)
Actually takes possession of the licensed material under its license

Additionally, the licensee should verify and the manufacturer (or service licensee) must:

Use an approved Type B package
Register with the NRC as a user of the Type B package
Possess an NRC-approved QA plan.

For each shipment, it must be clear who possesses the licensed material and is responsible for proper packaging of the radioactive materials and compliance with NRC and DOT regulations.

During an inspection, NRC uses the provisions of 10 CFR 71.5 and a Memorandum of Understanding with DOT on the Transportation of Radioactive Material (signed June 6, 1979) to examine and enforce various DOT requirements applicable to medical use licensees. Appendix W lists major DOT regulations that apply to medical licensees.

Response from Applicant: No response is needed from applicants during the licensing phase. However, before offering a Type B package for shipment, a licensee needs to have registered with NRC as a user of the package and obtained NRC's approval of its QA program. Transportation issues will be reviewed during inspection.

References: "A Review of Department of Transportation Regulations for Transportation of Radioactive Materials (1983 revision)" can be obtained be calling DOT's Office of Hazardous Material Initiatives and Training at (202) 366-4425. See the Notice of Availability on inside front cover of this draft report to obtain a copy of the Memorandum of Understanding with DOT on the Transportation of Radioactive Material, signed June 6, 1979, and Revision 1 of RG 7.10, "Establishing Quality Assurance Programs for Packaging Used in the Transport of Radioactive Material," dated June 1986.

8.44 Item 11: Waste Management

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Regulations: 10 CFR 20.1101, 10 CFR 20.2001-2007, 10 CFR 20.2102, 10 CFR 20.2108, 10 CFR 30.41, 10 CFR 30.51, 10 CFR 35.92, 10 CFR 35.2092, 10 CFR 71.5.

Criteria: Licensed materials must be disposed of in accordance with NRC requirements by: transfer to an authorized recipient; DIS; release to the environment; or treatment or disposal of the licensed material as described in 10 CFR 20.2004. Appropriate records must be maintained.

Discussion: Licensees must develop and implement procedures for waste disposal of licensed material in accordance with 10 CFR 20.1101. Appendix X contains model procedures for DIS and generator or other licensed material return. In 10 CFR 20.2001, NRC requires that licensees dispose of licensed material by specific means. In 10 CFR 20.2006, NRC requires that for licensed material transferred to a land disposal facility, the licensee must comply with the specific requirements in Appendix F to 10 CFR Part 20, i.e., manifest, certification, and control and tracking. In 10 CFR 35.92, NRC specifies the requirements for handling of waste by DIS. In 10 CFR 71.5, NRC requires that licensees who transport licensed material or offer it for transport, comply with the regulations of DOT in 49 CFR Parts 170-189. Applicants shall address the following items (as applicable):

Except for material suitable for DIS and some animal carcasses, solids must be transferred to an authorized recipient licensed to receive such waste in accordance with 10 CFR 20.2006 or in applicable regulations in 10 CFR Parts 30, 40, 60, 61, 70, or 72. Follow the packaging instructions you received from the transfer agent and the burial site operator. For your record of disposal, keep the consignment sheet that the transfer agent gave you.
When setting up a program for DIS, applicants/licensees should consider short-term and long-term storage. Long-term storage should be designed to allow for segregation of wastes with different half-lives (e.g., the use of multiple shielded containers). Containers should have shielded covers to maintain occupational exposure at ALARA levels. Storage areas must be in a secure location.
Waste from in vitro kits that are generally licensed pursuant to � 31.11 is exempt from waste disposal regulations. Radioactive labels should be defaced or removed. There is no need to keep any record of release or make any measurement.
Applicants should describe the monitoring and control mechanisms in place to ensure compliance with the appropriate requirements regarding the release of material into air and water pursuant to 10 CFR 20.1302 and 20.2003, respectively.

Regulations for disposal in the sanitary sewer appear in paragraph 20.2003. Material must be readily soluble or dispersible in the water. There are monthly and annual limits based on the total sanitary sewerage release of your facility. (Excreta from patients undergoing medical diagnosis or therapy is exempt from all the above limitations; see paragraph 20.2003 (b)) Make a record of the date, radionuclide, estimated activity that was released, and of the sink or toilet at which the material was released.
Limits on permissible concentrations in effluents to unrestricted areas are enumerated in Table II of Appendix B to 10 CFR Part 20. These limits apply at the boundary of the restricted area. Make a record of the date, radionuclide, estimated activity that was released and estimated concentration, and of the vent site at which the material was released.
Liquid scintillation - counting media containing 1.85 kilo-Bq (0.05 �Ci) per gram of H-3 or C-14 may be disposed of without regard to its radioactivity (� 20.2005 (a) (1)). Make a record of the date, radionuclide, estimated activity, calculated concentration, and how the material was disposed.

If applicants/licensees propose to treat or dispose of licensed material by incineration, they must receive specific approval from the NRC. Contact your regional NRC office for guidance on treatment or disposal of material by incineration in accordance with 10 CFR 20.2004.
Applicants wishing to use waste volume reduction operations, e.g., compactors, must provide the below detailed description along with their response to Item 8.16 (Facility Diagram):

A description of the compactor to demonstrate that it is designed to safely compact waste generated in your operations. (e.g., manufacturer's specifications, annotated sketches, photographs)
The types, quantities, and concentrations of waste to be compacted
An analysis of the potential for airborne release of radioactive material during compaction activities
The location of the compactors within the waste processing area(s), as well as a description of the ventilation and filtering systems used in conjunction with the compactors and procedures for monitoring filter blockage and exchange
Methods used to monitor worker breathing zones and/or exhaust systems
The types and frequencies of surveys that will be performed for contamination control in the compactor area
The instructions provided to compactor operators, including instructions for protective clothing, checks for proper functioning of equipment, method of handling uncompacted waste, and examining containers for defects.

General Guidance for Waste Disposal

All radioactivity labels must be removed or obliterated from containers and packages prior to disposal in in-house waste. If waste is compacted, all labels that are visible in the compacted mass must be defaced or removed. If waste is to be incinerated, labels may not need to be defaced. Contact your regional NRC office for guidance on this subject.
Remind employees that nonradioactive waste such as leftover reagents, boxes, and packing material should not be mixed with radioactive waste.
Occasionally monitor all procedures to ensure that radioactive waste is not created unnecessarily. Review all new procedures to ensure that waste is handled in a manner consistent with established procedures.
In all cases, consider the impact of various available disposal routes. Consider occupational and public exposure to radiation, other hazards associated with the material and routes of disposal (e.g., toxicity, carcinogenicity, pathogenicity, flammability), and expense.

DIS

For radionuclides of byproduct material with a half-life of less than 120 days, licensees may dispose of waste in ordinary trash as long as they follow the below criteria:

Hold byproduct material for decay until the waste cannot be distinguished from background level with a radiation detection survey meter set on its most sensitive scale and with no interposed shielding
Remove or obliterate all radiation labels
Maintain proper records.

Returning Sources

Because of the nature of the material contained in brachytherapy, teletherapy, and GSR sources, the only option for disposal is transfer to an authorized recipient as specified in 10 CFR 20.2001(a)(1). Authorized recipients are the original manufacturer of the sealed source, a commercial firm licensed by the NRC or an Agreement State to accept radioactive waste from other persons, or another specific licensee authorized to possess the licensed material (i.e., their license specifically authorizes the same radionuclide, form, and use).

Before transferring radioactive material, a licensee must verify that the recipient is properly authorized to receive the material using one of the methods described in 10 CFR 30.41. Additionally, 10 CFR 71.5 requires that licensees who transport licensed material or offer it for transport, comply with the regulations of DOT in 49 CFR Parts 170-189. Records of the transfer must be maintained as required by 10 CFR 30.51.

Licensees should promptly dispose of unused sealed sources to minimize potential problems of access by unauthorized individuals, use for inappropriate purposes, or improper disposal.

Because of the difficulties and costs associated with disposal of sealed sources, applicants should preplan the disposal. Applicants may want to consider contractual arrangements with the source supplier as part of a purchase agreement.

Response from Applicant: Provide the following:

A statement that: "We have developed and will implement written waste disposal procedures for licensed material in accordance with 10 CFR 20.1101, and that meet the requirements of the applicable section of Subpart K to 10 CFR Part 20 and 10 CFR 35.92."

The next two items on NRC Form 313 are to be completed on the form itself.

8.45 Item 12: Fees

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On NRC Form 313, enter the appropriate fee category from 10 CFR 170.31 and the amount of the fee enclosed with the application.

8.46 Item 13: Certification

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Individuals acting in a private capacity are required to date and sign NRC Form 313. Otherwise, representatives of the corporation or legal entity filing the application should date and sign NRC Form 313. These representatives must be authorized to make binding commitments and to sign official documents on behalf of the applicant. An application for licensing a private practice should be signed by a senior partner or the president. An application for licensing a medical institution (e.g., hospital, or medical center) must be signed by its chief executive officer (or delegate). The individual who signs the application should be identified by title of the office held. As discussed previously in "Management Responsibility," signing the application acknowledges management's commitment and responsibilities for the radiation protection program. NRC will return all unsigned applications for proper signature.

Note:

It is a criminal offense to make a willful false statement or representation on applications or correspondence (18 U.S.C. 1001).
When the application references commitments, those items become part of the licensing conditions and regulatory requirements.
It is a violation of 10 CFR 30.9 and 10 CFR 30.10 to make a false statement on an application. Civil sanctions including revocation of the license and/or orders removing individuals from licensed activity may be taken.

9 Amendments and Renewals to a License

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It is the licensee's obligation to keep the license current. If any of the information provided in the original application is to be modified or changed, the licensee must submit an application for a license amendment before the change takes place. Also, to continue the license after its expiration date, the licensee must submit an application for a license renewal at least 30 days before the expiration date (10 CFR 2.109, 10 CFR 30.36(a)).

10 CFR 35.13 describes activities that require a license amendment. These include the following:

Receipt or use of byproduct material for a clinical procedure permitted by Part 35, but not permitted by the license
Permitting anyone to work as an AU, AMP, or ANP, unless the individual meets one of the exceptions listed in 10 CFR 35.13(b)
Changing RSO
Ordering byproduct material in excess of the amount, or radionuclide or form different than currently authorized on the NRC license
Changing an area or address of use identified in the application or on the license.

Applications for license amendment, in addition to the following, must provide the appropriate fee. For renewal and amendment requests applicants must do the following:

Be sure to use the most recent guidance in preparing an amendment or renewal request
Submit in duplicate, either an NRC Form 313 or a letter requesting amendment or renewal
Provide the license number
For renewals, provide a complete and up-to-date application if many outdated documents are referenced or there have been significant changes in regulatory requirements, NRC's guidance, the licensee's organization, or radiation protection program. Alternatively, describe clearly the exact nature of the changes, additions, and deletions.

>Using the suggested wording of responses in this draft report will expedite NRC's review.

10 Termination of Activities

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Regulations: 10 20.1401-1405, 10 CFR 30.34(b), 10 CFR 30.35(g), 10 CFR 30.36(d) and (j), 10 CFR 30.51(f).

Criteria: The licensee must do the following:

Notify NRC, in writing, within 60 days, when its license has expired or a decision has been made to permanently cease licensed activities at the entire site regardless of contamination levels.
Notify NRC, in writing, within 60 days, when principal activities have not been conducted for 24 months or a decision has been made to permanently cease licensed activities in any separate building or outdoor area, if they contain residual radioactivity making them unsuitable for release according to NRC requirements.
Certify the disposition of licensed materials by submission of NRC Form 314, "Certificate of Disposition of Materials" or equivalent information.
Before a license is terminated, send the records important to decommissioning (as required by 10 CFR 30.35(g)) to the appropriate NRC regional office. If licensed activities are transferred or assigned according to 10 CFR 30.34(b), transfer records important to decommissioning to the new licensee.

Discussion: Subpart E to 10 CFR Part 20 describes the radiological criteria for license termination. A licensee's determination that a facility is not contaminated is subject to verification by NRC inspection.

For guidance on the disposition of licensed material, see the section on "Waste Management." For guidance on decommissioning records, see the section on "Financial Assurance and Record Keeping for Decommissioning."

Licensees should promptly dispose of unused licensed material to minimize potential problems of access by unauthorized individuals, use for inappropriate purposes, or improper disposal.

Response from Applicant: The applicant is not required to submit a response to the NRC during the initial application. However, when the license expires or at the time the licensee ceases operations, then the applicant must perform decommissioning activities and submit NRC Form 314 or equivalent information.

Reference: Copies of NRC Form 314, "Certificate of Disposition of Materials," are available upon request from NRC's Regional or Field Offices; see Appendix Y.

Appendix A: List of Documents Considered in Development of this Draft NUREG

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This draft report incorporates and updates the guidance previously found in the Regulatory Guides (RG), Policy and Guidance Directives (P&GD), and Information Notices (IN) listed in Table A.1. When this draft report is issued in final form, the documents in Table A.1 will be considered superseded and should not be used. Other references were also used in this draft report and are listed below.

Document Identification	Title	Date
RG 10.8, Revision 2	Guide for the Preparation of Applications for Medical Use Programs	8/87
Appendix X to RG 10.8, Revision 2	Guidance on Complying With New Part 20 Requirements	6/92
Draft RG DG-0009	Supplement to Regulatory Guide 10.8, Revision 2, Guide for the Preparation of Applications for Medical Use Programs	3/97
Draft RG FC 414-4	Guide for the Preparation of Applications for Licenses for Medical Teletherapy Programs	12/85
P&GD FC 87-2	Standard Review Plan (SRP) for License Applications for the Medical Use of Byproduct Material	12/87
Supplement 1 TO P&GD FC 86-4; Revision 1	Mobile Remote Afterloading Brachytherapy Licensing Module	5/97
P&GD FC 86-4, Revision 1	Information Required for Licensing Remote Afterloading Devices	9/93
Addendum to Revision 1 to P&GD FC 86-4	Information Required for Licensing Remote Afterloading Devices--Increased Source Possession Limits	7/95
P&GD 3-15	Standard Review Plan for Review of Quality Management Programs	6/95
RG 8.39	Release of Patients Administered Radioactive Materials	4/97
RG 8.33	Quality Management Program	10/91
P&GD 3-17 (previously 16)	Review of Training and Experience Documentation Submitted by Proposed Physician User Applicants
RG 8.23	Radiation Safety Surveys at Medical Institutions, Revision 1	1/81

Additionally, the below references were used.

References

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Title 10, Code of Federal Regulations

1.	Part 2	-	Rules of Practice for Domestic Licensing Proceedings and Issuance of Orders
2.	Part 19	-	Notices, Instructions, and Reports to Workers; Inspections
3.	Part 20	-	Standards for Protection Against Radiation
4.	Part 21	-	Reporting of Defects and Noncompliance
5.	Part 30	-	Rules of General Applicability to Domestic Licensing of Byproduct Material
6.	Part 31	-	General Domestic Licenses for Byproduct Material
7.	Part 32	-	Specific Domestic Licenses to Manufacture or Transfer Certain Items Containing Byproduct Material
8.	Part 33	-	Specific Domestic Licenses of Broad Scope for Byproduct Material
9.	Part 35	-	Medical Use of Byproduct Material
10.	Part 40	-	Domestic Licensing of Source Material
11.	Part 70	-	Domestic Licensing of Special Nuclear Material
12.	Part 71	-	Packaging and Transportation of Radioactive Material
13.	Part 150	-	Exemptions and Continued Regulatory Authority in Agreement States and in Offshore Waters Under Section 274
14.	Part 170	-	Fees for Facilities and Materials Licenses and Other Regulatory Services Under the Atomic Energy Act of 1954, As Amended
15.	Part 171	-	Annual Fees for Reactor Operating Licenses, and Fuel Cycle Licenses and Materials Licenses, Including Holders of Certificates of Compliance, Registrations, and Quality Assurance Program Approvals and Government Agencies Licensed by NRC

>Title 49, Code of Federal Regulations

>1.	Part 172	-	Hazardous Materials Table, Special Provisions, Hazardous Materials Communications, Emergency Response Information, and Training Requirements
2.	Part 173	-	Shippers -- General Requirements for Shipments and Packages
3.	Part 177	-	Carriage by Public Highway
4.	Part 178	-	Specifications for Packagings
>
NRC Regulatory Guides (RG)
>
1.	RG 1.86	-	Termination of Operating Licenses for Nuclear Reactors, June 1974
2.	RG 3.66	-	Standard Format and Content of Financial Assurance Mechanisms Required for Decommissioning Under 10 CFR Parts 30, 40, 70, and 72, June 1990
3.	RG 7.10, Revision 1	-	Establishing Quality Assurance Programs for Packaging Used in the Transport of Radioactive Material, June 1986
4.	RG 8.4	-	Direct-Reading and Indirect-Reading Pocket Dosimeters, February 1973
5.	RG 8.7	-	Instructions for Recording and Reporting Occupational Radiation Exposure Data, Revision 1, June 1992
6.	RG 8.9	-	Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay Program, Revision 1, June 1993
7.	RG 8.10	-	Operating Philosophy for Maintaining Occupational Radiation Exposures As Low As Is Reasonably Achievable, Revision 1-B, September 1975
8.	RG 8.13 (Draft)	-	Instruction Concerning Prenatal Radiation Exposure, October 1994
9.	RG 8.18	-	Information Relevant to Ensuring that Occupational Radiation Exposures at Medical Institutions Will Be As Low As Reasonably Achievable, Revision 1, October 1982
10.	RG 8.20	-	Applications of Bioassay for I-125 and I-131, Revision 1, September 1979
11.	RG 8.21	-	Health Physics Surveys for Byproduct Material at NRC-Licensed Processing and Manufacturing Plants
12.	RG 8.23	-	Radiation Safety Surveys at Medical Institutions
13.	RG 8.25	-	Air Sampling in the Workplace, Revision 1, June 1992.
14.	RG 8.29	-	Instruction Concerning Risks from Occupational Radiation Exposure, Revision 1, February 1996
15.	RG 8.34	-	Monitoring Criteria and Methods to Calculate Occupational Radiation Doses, July 1992.
16.	RG 8.36	-	Radiation Dose to the Embryo/Fetus, July 1992
17.	RG 10.2	-	Guidance to Academic Institutions Applying for Specific Byproduct Material Licenses of Limited Scope, Revision 1, December 1976
18.	RG 10.5 (Draft)	-	Applications for Type A Licenses of Broad Scope, October 1994
19.	RG 10.8, 1997	-	Revision (Draft NUREG 1569 - never published), Program-Specific Guidance for Medical Use Licensees
20.	RG FC 412-4 (Draft)	-	Guide for the Preparation of Applications for the Use of Radioactive Materials in Leak-Testing Services, June 1985
21.	RG FC 413-4 (Draft)	-	Guide for the Preparation of Applications for Licenses for the Use of Radioactive Materials in Calibrating Radiation Survey and Monitoring Instruments, June 1985

>NRC Information Notices (IN)
>
1.	IN 89-25, Revision 1	-	Unauthorized Transfer of Ownership or Control of Licensed Activities
2.	IN 94-70	-	Issues Associated with Use of Strontium-89 and Other Beta Emitting Radiopharmaceuticals
3.	IN 96-28	-	Suggested Guidance Relating to Development and Implementation of Corrective Action
4.	IN 97-30	-	Control of Licensed Material during Reorganizations, Employee-Management Disagreements, and Financial Crises

>NRC Policy and Guidance Directives (P&GD)

>1.	P&GD FC 90-2, Revision 1	-	Standard Review Plan for Evaluating Compliance with Decommissioning Requirements, April 1991
>
NRC NUREGs
>
1.	NUREG-0267, Revision 1	-	Principles and Practices for Keeping Occupational Radiation Exposures at Medical Institutions As Low As Reasonably Achievable, October 1982
2.	NUREG-1134	-	Radiation Protection Training for Personnel Employed in Medical Facilities, May 1985
3.	NUREG-1492	-	Regulatory Analysis on Criteria for the Release of Patients Administered Radioactive Material, February 1997
4.	NUREG-1516	-	Effective Management of Radioactive Materials Safety Programs at Medical Facilities, May 1997
5.	NUREG-1539	-	Methodology and Findings of the NRC's Materials Licensing Process Redesign, April 1996
6.	NUREG-1541 (Draft)	-	Process and Design for Consolidating and Updating Materials Licensing Guidance, April 1996
7.	NUREG-1556, Volume 3 (Draft)	-	Consolidated Guidance about Materials Licensees: Applications for Sealed Source and Device Evaluation and Registration, September 1997
8.	NUREG-1600	-	General Statement of Policy and Procedures for NRC Enforcement Actions, June 1995 and Compilation of NRC Enforcement Policy as of September 10, 1997
9.	NUREG/CR-4444	-	Radiation Safety Issues Related to Radiolabeled Antibodies, 1991
10.	NUREG/CR-4884	-	Interpretation of Bioassay Measurement, July 1987
11.	NUREG/CR-6323	-	Relative Risk Analysis in Regulating the Use of Radiation-Emitting Medical Devices: A Preliminary Application, September 1995
12.	NUREG/CR-6324	-	Quality Assurance for Gamma Knives, September 1995

>National Council on Radiation Protection and Measurements (NCRP) Publications

>1.	NCRP Report No. 30	-	Safe Handling of Radioactive Materials, 1989
2.	NCRP Report No. 37	-	Precautions in the Management of Patients Who Have Received Therapeutic Amounts of Radionuclides, 1970
3.	NCRP Report No. 40	-	Protection Against Radiation from Brachytherapy Sources, 1972
4.	NCRP Report No. 49	-	Structural Shielding Design and Evaluation for Medical Use of X Rays and Gamma Rays of Energies up to 10 MeV, 1976
5.	NCRP Report No. 57	-	Instrumentation and Monitoring Methods for Radiation Protection, 1978
6.	NCRP Report No. 58	-	A Handbook of Radioactivity Measurement Procedures, Second Edition, 1985
7	NCRP Report No. 69	-	Dosimetry of X-Ray and Gamma-Ray Beams for Radiation Therapy in the Energy Range 10 keV to 50 MeV, 1981
8.	NCRP Report No. 71	-	Operational Radiation Safety - Training, 1983
9.	NCRP Report No. 87	-	Use of Bioassay Procedures for Assessment of Internal Radionuclide Deposition, February 1987
10.	NCRP Report No. 102	-	Medical X-Ray, Electron Beam and Gamma Ray Protection for Energies up to 50 MeV (Equipment Design, Performance and Use), 1989
11.	NCRP Report No. 105	-	Radiation Protection for Medical and Allied Health Personnel, 1989
12.	NCRP Report No. 107	-	Implementation of the Principle of As Low As Reasonably Achievable (ALARA) for Medical and Dental Personnel, 1990
13.	NCRP Commentary No. 11	-	Dose Limits for Individuals Who Receive Exposure from Radionuclide Therapy Patients, February 1995

>International Commission on Radiological Protection (ICRP) Publications

>1.	ICRP Report No. 26	-	Recommendations of the International Commission on Radiological Protection, 1977
2.	ICRP Report No. 30	-	Limits for Intakes of Radionuclides by Workers, 1978
3.	ICRP Report No. 35	-	General Principles of Monitoring for Radiation Protection of Workers, 1982
4.	ICRP Publication No. 53	-	Radiation Dose to Patients from Radiopharmaceuticals, 1987
5.	ICRP Publication 54	-	Individual Monitoring for Intake of Radionuclides by Workers: Design and Interpretation, 1987

>ANSI Standards

>1.	ANSI N13.4-1971 (R1983)	-	Specification of Portable X- or Gamma Radiation Survey Instruments
2.	ANSI N13.5-1972 (R1989)	-	Performance and Specifications for Direct Reading and Indirect Reading Pocket Dosimeters for X- and Gamma Radiation
3.	ANSI N13.6-1966 (R1989)	-	Practice for Occupational Radiation Exposure Records Systems
4.	ANSI N14.5-1987	-	Leakage Tests on Packages for Shipment of Radioactive Materials
5.	ANSI N42.12-1994	-	Calibration and Usage of Sodium Iodide Detector Systems
6.	ANSI N42.13-1986 (R1993)	-	Calibration and Usage of Dose Calibrator Ionization Chambers for the Assay of Radionuclides
7.	ANSI N42.15-1990	-	Performance Verification of Liquid Scintillation Counting Systems
8.	ANSI N42.17A-1989	-	Performance Specifications for Health Physics Instrumentation-Portable Instrumentation for Use in Normal Environmental Conditions
9.	ANSI N322	-	Inspection and Test Specifications for Direct and Indirect Reading Quartz Fiber Pocket Dosimeters
10.	ANSI N323A-1997	-	Radiation Protection Instrumentation Test and Calibration, Portable Survey Instruments
11.	ANSI N449.1-1978 (R1984)	-	Procedures for Periodic Inspection of Cobalt-60 and Cesium-137 Teletherapy Equipment

>American Association of Physicists in Medicine (AAPM) Reports

>1.	AAPM Task Group No. 21	-	A Protocol for the Determination of Absorbed Dose from High-Energy Photon and Electron Beams, 1984
2.	AAPM Radiation Therapy Committee Task Group No. 40	-	Comprehensive QA for Radiation Oncology, 1994
3.	AAPM Radiation Therapy Committee Task Group No. 56	-	Code of Practice for Brachytherapy Physics, 1998
4.	AAPM Radiation Therapy Committee Task Group No. 56	-	HDR Treatment Delivery Safety, 1997 Draft
5.	AAPM Report No. 41	-	Remote Afterloading Technology, 1993
6.	AAPM Report No. 54	-	Stereotactic Radiosurgery, 1995

>Other Technical Publications

>1.	International Commission on Radiation Units and Measurements (ICRU), "Certification of Standardized Radioactive Sources," Report No. 12, 1968
2.	U.S. Department of Health, Education, and Welfare, "Radiological Health Handbook," 1970
3.	R.C.T. Buchanan and J.M. Brindle, "Radioiodine Therapy to Out-patients - The Contamination Hazard," British Journal of Radiology, Volume 43, 1970
4.	International Atomic Energy Agency (IAEA), "Monitoring of Radioactive Contamination on Surfaces," Technical Report Series No. 120, 1970
5.	IAEA, "Handbook on Calibration of Radiation Protection Monitoring Instruments," Technical Report Series No. 133, 1971
6.	A.P. Jacobson, P.A. Plato, and D. Toeroek, "Contamination of the Home Environment by Patients Treated with Iodine-131," American Journal of Public Health, Volume 68, Number 3, 1978
7.	A. Brodsky, "Resuspension Factors and Probabilities of Intake of Material in Process (or 'Is 10-6 a Magic Number in Health Physics?')," Health Physics, Volume 39, Number 6, 1980
8.	Bureau of Radiological Health, "Radiation Safety in Nuclear Medicine: A Practical Guide," Department of Health and Human Services (HHS) Publication FDA 82-8180, November 1981
9.	Center for Devices and Radiological Health, "Recommendations for Quality Assurance Programs in Nuclear Medicine Facilities," HHS Publication FDA 85-8227, October 1984
10.	S. R. Jones, "Derivation and Validation of a Urinary Excretion Function for Plutonium Applicable over Ten Years Post Intake," Radiation Protection Dosimetry, Volume 11, No. 1, 1985
11.	"Guidelines for Patients Receiving Radioiodine Treatment," Society of Nuclear Medicine, 1987
12.	J. R. Johnson and D. W. Dunford, "GENMOD--A Program for Internal Dosimetry Calculations," AECL-9434, Chalk River Nuclear Laboratories, Chalk River, Ontario, 1987
13.	K.F. Eckerman, A.B. Wolbarst, and A.C.B. Richardson, "Federal Guidance Report No. 11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," Report No. EPA-520/1-88-020, 1988
14.	K. W. Skrable et al., "Intake Retention Functions and Their Applications to Bioassay and the Estimation of Internal Radiation Doses," Health Physics Journal, Volume 55, No. 6, 1988
15.	A.S. Meigooni, S. Sabnis, R. Nath, "Dosimetry of Palladium-103 Brachytherapy Sources for Permanent Implants," Endocurietherapy Hyperthermia Oncology, Volume 6, April 1990
16.	R. Nath, A.S. Meigooni, and J.A. Meli, "Dosimetry on Transverse Axes of 125I and 192Ir Interstitial Brachytherapy Sources," Medical Physics, Volume 17, Number 6, November/December 1990
17.	M.G. Stabin et al., "Radiation Dosimetry for the Adult Female and Fetus from Iodine-131 Administration in Hyperthyroidism," Journal of Nuclear Medicine, Volume 32, Number 5, May 1991
18.	P. Early, D. B. Sodee, "Principles and Practice of Nuclear Medicine," 2nd ed., 1995
19.	M. Stabin, "Internal Dosimetry in Pediatric Nuclear Medicine," Pediatric Nuclear Medicine, 1995
20.	"Intravascular Brachytherapy - Guidance for Data to be Submitted to the Food and Drug Administration In Support of Investigational Device Exemption (IDE) Applications," Draft Version 1.3, 1996
21.	R.O. Dunkelberger, II, "Which Probe Should I Use," Baltimore-Washington Health Physics Society Newsletter

Appendix B: United States Nuclear Regulatory Commission Form 313

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United States Nuclear Regulatory Commission Form 313

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Form 313: Application for Material License

Appendix C: License Application Checklist and Sample Licenses

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The instructions on Table C.1, the Applicability Table, may be followed to determine whether the information must be provided or if "NA" may be the response to each item that follows.

To determine those items to which you must respond, "highlight" the columns under the categories of materials you requested in item 5. If any "Y" beside an item is highlighted, you must provide detailed information in response to the item. If the letters "NA" (not applicable) are highlighted, you may respond "NA" on your application. If any "N" beside an item is highlighted, no information in response is required; however the NRC regulations that apply to the given category apply to your type of license. If any "P" beside an item is highlighted, you must provide a commitment to develop and implement a procedure in response to the item. Note that some modules have additional item numbers that may need to be addressed.

In addition, sample licenses are included that may provide guidance on the particular type of medical use you are requesting.

Table C.1 Applicability Table

Item #	Topic	35.100/200	35.300	35.400	35.500	35.600	35.1000	APP
N/A	Unsealed Byproduct Material - Low Dose	Y
N/A	Unsealed Byproduct Material - High Dose		Y
N/A	Manual Brachytherapy			Y
N/A	Sealed Sources for Diagnosis				Y
N/A	Teletherapy Devices					Y
N/A	Remote Afterloader Devices					Y
N/A	Gamma Stereotactic Radiosurgery Devices					Y
N/A	Emerging Technologies						Y
8.6	Financial Assurance Determination	Y	Y	Y	Y	Y	Y	E
8.7	Sealed Source Registry	N	N	N	N	N	N
8.10	Radiation Safety Officer	Y	Y	Y	Y	Y	Y	F, G
8.11	Authorized User Training and Experience	Y	Y	Y	Y	Y	Y	G
8.12	Authorized Nuclear Pharmacists Training and Experience	Y	Y	N/A	N/A	N/A	Y	G
8.13	Authorized Medical Physicist Training and Experience	N/A	N/A	Y	N/A	Y	Y	G
8.14	Training Program	N	N	N	N	N	N	H
8.16	Facility Diagram and Equipment	Y	Y	Y	Y	Y	Y
8.17	Radiation Monitoring Instrument Calibration	P	P	P	N	P	P	I
8.18	Dose Calibrator Calibration	P	P	N/A	N/A	N/A	P	J
8.19	Dosimetry Equipment and Therapy Sealed Source Calibration	N/A	N/A	P	N/A	P	P
8.20	Other Equipment and Facilities	Y	Y	Y	Y	Y	Y
8.22	Audit Program	N	N	N	N	N	N	K
8.23	Occupational Dose	P	P	P	P	P	P	L
8.24	Public Dose	N	N	N	N	N	N	M
8.25	Minimization of Contamination	Y	Y	Y	Y	Y	Y
8.26	Operating and Emergency Procedures	N	N	N	N	N	N	N
8.28	Ordering and Receiving	N	N	N	N	N	N	O
8.29	Opening Packages	P	P	P	P	P	P	P
8.30	Sealed Source Inventory	N	N	N	N	N	N
8.31	Use Records	N	N	N	N	N	N
8.32	Leak Tests	N	N	N	N	N	N	Q
8.33	Area Surveys	P	P	P	P	P	P	R
8.34	Written Directive Procedures	N/A	P	P	N/A	P	P	S
8.35	Safe Use of Unsealed Licensed Material	P	P	N/A	N/A	N/A	P	T
8.36	Maintenance of Therapy Devices Containing Sealed Sources	N/A	N/A	N/A	N/A	Y	Y
8.37	Spill Procedures	P	P	N/A	N/A	N/A	P	N
8.38	Emergency Response for Sealed Sources or Devices	N/A	N/A	P	P	P	P	N
8.39	Patient or Human Research Subject Release	P	P	P	N/A	N	P	U
8.40	Safety Precautions for Therapy Treatments where Patients are Hospitalized	N/A	N	N	N/A	N/A*	N
8.41	Safety and Device Calibration Procedures	N/A	N/A	N/A	P	P	P
8.42	Mobile Use of Radionuclides	Y	Y	Y	Y	Y	Y	V
8.43	Transportation	N	N	N	N	N	N	W
8.44	Waste Management	P	P	P	P	P	P	X
* N for remote afterloaders

Form 374. Material License Form - Sample

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Material License Form - Sample NRC Form 374(A)
Sample Gamma Knife	Items 1 through 13
	Item 14
Sample Medical Institution Limited	Items 1 through 6
	Items 9 through 11
	Items 12 through 14
	Items 14 through 18
	Item 19
Sample Medical Institution Limited with differing authorizations for two locations of use and 35.75 limitation on 35.300 uses	Items 1 through 9
	Items 10 through 14
	Item 15
Sample Pacemaker License	Items 1 through 16
	Item 17
Sample Medical Private Practice permitting only out-patient therapy	Items 1 through 9
	Items 10 through 13
Sample High Dose Rate Remote Afterloader	Items 1 through 14
	Item 15
Sample Teletherapy	Items 1 through 13
	Items 14 through 16
Sample In-Vitro Testing Laboratory	Items 1 through 13
	Items 13 through 15

Appendix D: Information Needed for Transfer of Control

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Licensees must provide full information and obtain NRC's prior written consent before transferring control of the license (10 CFR 30.34(b)). If NRC finds that the transfer is in accordance with the Act, NRC may, at the NRC's option, authorize the transfer by license amendment. Otherwise, a new license authorizing the transfer may be issued. The licensee must provide the following information concerning changes of control. If any items are not applicable, so state.

1.	The new name of the licensed organization. If there is no change, the licensee should so state.
2.	The new licensee contact and telephone number(s) to facilitate communications.
3.	Any changes in personnel having control over licensed activities (e.g., officers of a corporation) and any changes in personnel named in the license such as RSO, AUs, or any other persons identified in previous license applications as responsible for radiation safety or use of licensed material. The licensee should include information concerning the qualifications, training, and responsibilities of new individuals.
4.	An indication of whether the transferor will continue to exist without an NRC license.
5.	A complete, clear description of the transaction, including any transfer of stocks or assets, mergers, etc., so that legal counsel is able, when necessary, to differentiate between name changes and transferring control.
6.	A complete description of any planned changes in organization, location, facility, equipment, or procedures (i.e., changes in operating or emergency procedures).
7.	A detailed description of any changes in the use, possession, location, or storage of the licensed materials.
8.	An indication of whether all surveillance items and records (e.g., calibrations, leak tests, surveys, inventories, and accountability requirements) will be current at the time of transfer. Provide a description of the status of all surveillance requirements and records.
9.	Confirmation that all records concerning the safe and effective decommissioning of the facility, pursuant to 10 CFR 30.35(g); public dose; and waste disposal by release to sewers, incineration, radioactive material spills, and on-site burials, have been transferred to the new licensee, if licensed activities will continue at the same location, or to the NRC for license terminations.
10.	A description of the status of the facility. Specifically, the presence or absence of contamination should be documented. If contamination is present, will decontamination occur before transfer? If not, does the successor company agree to assume full liability for the decontamination of the facility or site?
11.	A description of any decontamination plans, including financial assurance arrangements of the transferee, as specified in 10 CFR 30.35. Include both information about how the transferee and transferor propose to divide the transferor's assets and responsibility for any cleanup needed at the time of transfer.
12.	Confirmation that the transferee agrees to abide by all commitments and representations previously made to NRC by the transferor. These include, but are not limited to: maintaining decommissioning records required by 10 CFR 30.35(g); implementing decontamination activities and decommissioning of the site; and completing corrective actions for open inspection items and enforcement actions. With regard to contamination of facilities and equipment, the transferee should confirm, in writing, that it accepts full liability for the site and should provide evidence of adequate resources to fund decommissioning, or the transferor should provide a commitment to decontaminate the facility before transferring control. With regard to open inspection items, etc., the transferee should confirm, in writing, either that it accepts full responsibility for open inspection items and/or any resulting enforcement actions, or that the transferee proposes alternative measures for meeting the requirements, or that the transferor provides a commitment to close out all such actions with NRC before license transfer.
13.	Documentation that the transferor and transferee agree to transferring control of the licensed material and activity, including the conditions of transfer, with the transferee made aware of all open inspection items and its responsibility for possible resulting enforcement actions.
14.	A commitment by the transferee to abide by all constraints, conditions, requirements, representations, and commitments identified in the existing license. If the transferee does not make such a commitment, the transferee must provide a description of its program, to ensure compliance with the license and regulations.

Appendix E: Guidance on Financial Assurance Determination

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Determining Need for Financial Assurance for Decommissioning

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The half lives of unsealed byproduct material used by medical licensees have traditionally been less than 120 days. Therefore, most medical use applicants need only to consider licensed material in sealed sources to evaluate the need for financial assurance. Use Table E.1 to determine if financial assurance is required for the sealed sources listed. If requesting sealed sources other than those listed or any other unsealed byproduct material with half lives greater than 120 days, refer to 10 CFR 30.35 and Appendix B to Part 30 for possession limits requiring financial assurance. The sum of the fractions procedure is also depicted in Table E.1 and must be used in determining the need for financial assurance for both sealed and unsealed byproduct material.

Table E.1 Worksheet for Determining Need for Financial Assurance for Sealed Sources

Step Number	Description	Cobalt-60	Cesium-137	Strontium-90
1	Activity possessed, in Curies*
2	Activity requiring financial assurance, in Curies	10,000	100,000	1,000
3	Divide data in Step 1 by data in Step 2 = FRACTION
4	Add the fractions determined in Step 3
* This table uses only conventional units. The conversion to the International System of units (SI) is: 1 Curie = 37 gigabecquerels.

If the sum of the fractions is greater than or equal to 1, the applicant will need to submit financial assurance. RG 3.66,⁽³⁾ "Standard Format and Content of Financial Assurance Mechanisms Required for Decommissioning Under 10 CFR Parts 30, 40, 70, and 72," dated June 1990, provides sample documents for financial mechanisms. The recommended wording for a Statement of Intent for government licensees is shown below, since this mechanism is not described in RG 3.66.

Suggested Wording for a Statement of Intent for a Government Licensee

[DATE]
TO: U. S. NUCLEAR REGULATORY COMMISSION
WASHINGTON, DC 20555 [or appropriate regional address]

STATEMENT OF INTENT

As [Title] of [Licensee Name], I exercise express authority and responsibility to approve funding for decommissioning activities associated with operations authorized by U. S. Nuclear Regulatory Commission Material License No._____. This authority is established by [Name of Document(s) Governing Control of Funds]. Within this authority, I intend to have funds made available when necessary, in an amount up to [Dollar Amount] to decommission [Description of Facilities]. I intend to request and obtain these funds sufficiently in advance of decommissioning to prevent delay of required activities.

A copy of [Name of Documents] is attached as evidence that I am authorized to represent [Licensee Name] in this transaction.

[SIGNATURE]
[NAME]
[TITLE]

Attachment: As stated

Appendix F: Typical Duties and Responsibilities of the Radiation Safety Officer and Sample Delegation of Authority

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Model RSO Duties and Responsibilities

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The RSO's duties and responsibilities include ensuring radiological safety and compliance with NRC and DOT regulations and the conditions of the license (see Figure 8.1). Typically, these duties and responsibilities include ensuring the following:

Activities involving licensed material that the RSO considers unsafe are stopped
Radiation exposures are ALARA
Up-to-date radiation protection procedures in the daily operation of the licensee's byproduct material program are developed, distributed, and implemented
Possession, use, and storage of licensed material is consistent with the limitations in the license, the regulations, the SSDR Certificate(s), and the manufacturer's recommendations and instructions
Individuals installing, relocating, maintaining, or repairing devices containing sealed sources are trained and authorized by an NRC or Agreement State license
Personnel training is conducted and is commensurate with the individual's duties regarding licensed material
Documentation is maintained to demonstrate that individuals are not likely to receive, in one year, a radiation dose in excess of 10% of the allowable limits or that personnel monitoring devices are provided
When necessary, personnel monitoring devices are used and exchanged at the proper intervals, and records of the results of such monitoring are maintained
Licensed material is properly secured
Documentation is maintained to demonstrate, by measurement or calculation, that the total effective dose equivalent to the individual likely to receive the highest dose from the licensed operation does not exceed the annual limit for members of the public
Proper authorities are notified of incidents such as loss or theft of licensed material, damage to or malfunction of sealed sources, or fire
Medical events are investigated and reported to the NRC. Cause(s) and appropriate corrective action(s) are identified, and timely corrective action(s) are taken
Audits of the radiation protection program are performed at least annually and documented
If violations of regulations or license conditions or program weaknesses are identified, effective corrective actions are developed, implemented, and documented
Licensed material is transported in accordance with all applicable DOT requirements
Licensed material is disposed of properly
Appropriate records are maintained
Up-to-date license is maintained and amendment and renewal requests are submitted in a timely manner

Model Delegation of Authority

Memo To:   Radiation Safety Officer
From:   Chief Executive Officer
Subject:   Delegation of Authority

You, have been appointed Radiation Safety Officer and are responsible for ensuring the safe use of radiation. You are responsible for managing the radiation protection program; identifying radiation protection problems; initiating, recommending, or providing corrective actions; verifying implementation of corrective actions; stopping unsafe activities; and ensuring compliance with regulations. You are hereby delegated the authority necessary to meet those responsibilities, including prohibiting the use of byproduct material by employees who do not meet the necessary requirements and shutting down operations where justified by radiation safety. You are required to notify management of situations where staff are not cooperating and not addressing radiation safety issues. In addition, you are free to raise issues with the Nuclear Regulatory Commission at anytime. It is estimated that you will spend _____ hours per week conducting radiation protection activities.

_________________________________
Signature of Management Representative

I accept the above responsibilities,

_________________________________
Signature of Radiation Safety Officer

cc: Affected department heads

Appendix G: Documentation of Training and Experience

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General Guidance

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The required training and experience described in 10 CFR Part 35 must be obtained within the 7 years preceding the date of the application, or the individual must document as having had related continuing education, retraining, and experience since the required training and experience were obtained. Complete retraining is neither practical nor necessary in most cases. Examples of acceptable continuing education and experience are:


1.	Successful completion of didactic review courses that include radiation safety practices relative to the proposed type of authorized medical use.
2.	Practical and laboratory experience with patient procedures using radioactive material, for the same use(s) for which the applicant is requesting authorization.
3.	Practical and laboratory experience under the supervision of an AU at the same or another licensed facility that is authorized for the same use(s) for which the applicant is requesting authorization.
4.	For therapy devices, experience with the therapy unit and/or comparable linear accelerator experience and completion of an inservice to review the emergency procedures relative to the therapy unit to be used by the applicant.

The simplest and most straightforward method of demonstrating acceptable training and experience is through certification by one of the professional boards recognized by the NRC. Equally straightforward is evidence that the applicant is listed as an AU on an NRC or Agreement State license or permit issued by a medical broad scope licensee, provided that the applicant is authorized for the same types of use(s) being requested in the application under review, and that the applicant meets the requirements for recentness of training criteria described in 10 CFR 35.59. For AUs who have been previously authorized under a medical broad scope license, the applicant should submit either verification of previous authorization(s) granted by the broad scope licensee or evidence of acceptable training and experience.

The NRC recognizes supervised work and practical experience, such as that described in 10 CFR 35.292(b), conducted under a preceptor in a licensed material use program. A preceptor is an AU who provides frequent direction, instruction and direct oversight of the student as the student completes the required work and practical experience in the use of byproduct material. Preceptorships may occur at various licensed facilities, from a large teaching university hospital to a small private practice. Upon completion of the supervised work and practical experience, the applicant should either submit the preceptor forms provided as NRC Forms 313A and 313B to NRC Form 313, "Materials License Application" or a letter from the preceptor that indicates that the applicant has obtained all required experience elements.

There is no NRC requirement that an AU must render an interpretation of a diagnostic image or results of a therapeutic procedure. NRC recognizes that the AU may or may not be the physician who interprets such studies. Additionally, NRC regulations do not restrict who can read and interpret diagnostic scans or the results of therapeutic procedures involving the administration of byproduct material to individuals.

NRC Form 313A U.S. Nuclear Regulatory Commission

TRAINING AND EXPERIENCE

Note: Descriptions of training and experience must contain sufficient detail to match the training and experience criteria in the applicable regulations.

1. Name of Individual, Proposed Authorization (e.g., Radiation Safety Officer), and Applicable Training Requirements (e.g., 10 CFR 35.50)

2. For Physicians and Pharmacists, State or Territory Where Licensed

3. Certification

Specialty Board	Category	Month and Year Certified

4. Didactic Training
Description of Training	Location	Clock Hours	Dates of Training
Radiation Physics and Instrumentation
Radiation Protection
Mathematics Pertaining to the Use and Measurement of Radioactivity
Radiation Biology
Chemistry of Byproduct Material for Medical Use
Other

5. Practical Experience with Radiation. (Actual use of radionuclides or equivalent experience)
Description of Experience	Name of Supervising Individual(s)	Location and Corresponding Materials License Number	Clock Hours and Dates	Related Radiation Safety Exam Score

6. Formal Training
Degree, Area of Study	Name of Program and Location with Corresponding Materials License Number	Dates	Name of Organization that Approved the Program (e.g., Accreditation Council for Graduate Medical Education) and the Applicable Regulation (e.g., 10 CFR 35.294)

7. The individual named in item 1 of this form is competent to function independently as an authorized user. [ ] Yes [ ] No *Note:* Response to Item 7 is applicable to proposed authorized users, medical physicists, or radiation safety officer for the type of medical use requested.
8. The training and experience indicated above was obtained under the supervision of:	10. Preceptor's Signature
a. Name of Supervisor
b. Mailing Address	11. Preceptor's Name (Printed Clearly)
c. City
9. Materials License Number	12. Date

NRC Form 313B U.S. Nuclear Regulatory Commission PRECEPTOR STATEMENT *Note:* Descriptions of training and experience must contain sufficient detail to match the training and experience criteria in the applicable regulations. Supplement B must be completed by the individual's preceptor. If more than one is necessary to document experience, obtain a separate preceptor statement from each.
1. Name of Individual, Proposed Authorization (e.g., Authorized User), and Applicable Training Requirements (e.g., 10 CFR 35.490): Name, Proposed Authorization, and Applicable Training Requirements Street Address City State Zip Code
2. Supervised Experience of Above Named Individual
Radionuclide	Type of Use	Number of Cases Involving Personal Participation	Location and Corresponding Materials License Number, Dates, and Clock Hours of Experience











2. Supervised Experience of Above Named Individual (Cont'd)
Radionuclide	Type of Use	Number of Cases Involving Personal Participation	Location and Corresponding Materials License Number, Dates, and Clock Hours of Experience














3. The individual named in item 1 of this form is competent to independently operate a nuclear pharmacy. [ ] Yes [ ] No *Note:* Response to Item 3 is only applicable to proposed authorized nuclear pharmacists.

4. The training and experience indicated above was obtained under the supervision of:	6. Preceptor's Signature
a. Name of Supervisor
b. Mailing Address	7. Preceptor's Name (Printed Clearly)
c. City
5. Materials License Number	8. Date

Appendix H: Model Training Program

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Model Training Program for Medical Uses of Radionuclides, Sealed Sources and Medical Devices Containing Sealed Sources

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Personnel shall be instructed before assuming duties with, or in the vicinity of radioactive materials, during annual refresher training, and whenever there is a significant change in duties, regulations, terms of the license, or type of radioactive material or therapy device used. Records of worker training will be maintained for at least three years. The training records will include the date of the instruction or training and the name(s) of the attendee(s) and instructor(s).

Training for Individuals Involved in the Usage of Byproduct Material

We will instruct the professional staff (e.g., AU, AMP, ANP, RSO, nurse, dosimetrist, technologist, therapist) providing or involved in the care of patients during diagnostic or therapeutic procedures in the following topics, commensurate with his/her duties:

Basic radiation biology, e.g., interaction of ionizing radiation with cells and tissues (10 CFR 19.12)
Basic radiation protection to include concepts of time, distance, and shielding (10 CFR 19.12)
Concept of maintaining exposure ALARA (10 CFR 20.1101)
Risk estimates, including comparison with other health risks (10 CFR 19.12)
Posting requirements (10 CFR 20.1902)
Proper use of personnel dosimetry (when applicable) (10 CFR 20.1201)
Access control procedures (10 CFR 20.1601, 10 CFR 20.1802)
Proper use of radiation shielding, if used (10 CFR 19.12)
Patient release procedures (10 CFR 35.75)
Instruction in procedures for notification of the RSO and AU, when responding to patient emergencies or death, to ensure that radiation protection issues are identified and addressed in a timely manner. The intent of these procedures should in no way interfere with or be in lieu of appropriate patient care. (10 CFR 19.12)
Occupational dose limits and their significance (10 CFR 20.1201)
Dose limits to the embryo/fetus, including instruction on declaration of pregnancy (10 CFR 20.1208)
Workers right to be informed of occupational radiation exposure (10 CFR 19.13)
Each individual's obligation to report unsafe conditions to the RSO (10 CFR 19.12)
Applicable regulations, license conditions, information notices, bulletins, etc. (10 CFR 19.12)
Where copies of the applicable regulations, the NRC license, and its application are posted or made available for examination (10 CFR 19.11)
Proper recordkeeping required by NRC regulations (10 CFR 19.12, 10 CFR 35.27)
Appropriate surveys to be conducted, including surveys of all material leaving radioactive material areas (10 CFR 20.1501)
Proper use of required survey instruments (10 CFR 20.1501)
Emergency procedures (10 CFR 19.12)
Decontamination and release of facilities and equipment (10 CFR 20.1406, 10 CFR 30.36)
Dose to individual members of the public (10 CFR 20.1301)
Licensee's operating procedures (e.g., survey requirements, instrument calibration, waste management, sealed source leak testing) (10 CFR 35.27)

Training for the Staff Directly Involved in Administration to or Care of Patients Administered Therapeutic Quantities of Byproduct Material (Including Greater than 30 microcuries of I-131), or Therapeutic Treatment Planning

In addition to the topics identified above, we will instruct staff involved in the therapy treatment of patients (e.g., nursing, RSO, AMP, AU, and dosimetrist) in the following topics, commensurate with his/her duties:

Leak testing of sealed sources (10 CFR 35.67)
Emergency procedures (including emergency response drills) (10 CFR 35.310, 10 CFR 35.410, 10 CFR 35.610)
Operating instructions (10 CFR 35.27, 10 CFR 35.610)
Computerized treatment planning system (10 CFR 35.657)
Dosimetry protocol (10 CFR 35.610)
Detailed pretreatment quality assurance checks (10 CFR 35.27)
Safe handling (when applicable) of the patient's dishes, linens, excretions (saliva, urine, feces), and surgical dressings that are potentially contaminated or that may contain radioactive sources (10 CFR 35.310, 10 CFR 35.410)
Patient control procedures (10 CFR 35.310, 10 CFR 35.410, 10 CFR 35.610)
Visitor control procedures, such as visitor's stay times and safe lines in radiation control areas (patient's room) (10 CFR 35.310, 10 CFR 35.410, 10 CFR 35.610)
Licensee's WD Procedures, to ensure that each administration is in accordance with the WD, patient identity is verified, and where applicable, attention is paid to correct positioning of sources and applicators to ensure that treatment is to the correct site (or, for GSR; correct positioning of the helmet) (10 CFR 35.41)
Proper use of safety devices and shielding to include safe handling and shielding of dislodged sources (or, in the case of remote afterloaders, disconnected sources) (10 CFR 35.310, 10 CFR 35.410, 10 CFR 35.610)
Size and appearance of different types of sources and applicators (10 CFR 35.410, 10 CFR 35.610)
Previous incidents, events and/or accidents (10 CFR 35.310, 10 CFR 35.410, 10 CFR 35.610)
For remote afterloaders, teletherapy units, and GSR units; initial training provided by the device manufacturer or by individuals certified by the device manufacturer that is device model-specific and includes:

Design, use, and function of the device, including safety systems and interpretation of various error codes/conditions, displays, indicators, and alarms;
Hands-on training in actual operation of the device under the direct supervision of an experienced user including "dry runs" (using dummy sources) of routine patient setup/treatment and implementation of the licensee's emergency procedures; and
A method of determining each trainee's competency to use the device for each type of proposed use, such as practical examinations.

Additional Training for Authorized Medical Physicists

In addition to the training and experience requirements of 10 CFR 35.51 or 10 CFR 35.961, we will verify that the AMP has specific training and experience in performing the measurements and calculations associated with the specific type of therapy treatments that we are requesting (e.g, manual brachytherapy, remote afterloader therapy, teletherapy, GSR therapy) and that the training involved the use of the treatment planning system that will be used for therapy at our facility.

Additional Training for Therapy Authorized Users

In addition to the training and experience requirements of 10 CFR 35.390, 10 CFR 35.490, 10 CFR 35.690 or Subpart J, we will verify that the therapy physician has specific training and experience in performing the specific therapy treatment that we are requesting, including training in the treatment planning system, quality control system, and clinical procedures that will be used at our facility.

Training for Contractors

We will ensure that individuals who work under a contractual arrangement be instructed in the topics described above, equivalent to instruction given to facility employees and commensurate with their duties.

Training for Ancillary Staff

For the purposes of this section, ancillary staff includes personnel engaged in housekeeping, dietary services, laboratory services, security, custodial services, etc.

For individuals whose assigned activities during normal and abnormal situations are likely to result in a dose in excess of 1 mSv (100 mrem), we will provide instruction commensurate with potential radiological health protection problems present in the work place. Alternatively, we may choose to prohibit ancillary personnel from entering restricted or controlled areas unless escorted by trained personnel. Topics of instruction will include the following:

Storage, transfer, or use of radiation and/or radioactive material (10 CFR 19.12)
Health protection problems associated with exposure to radiation and/or radioactive material, in precautions or procedures to minimize exposure, and in the purposes and functions of protective devices employed (e.g., basic radiation protection concepts of time, distance, and shielding) (10 CFR 19.12)
The applicable provisions of NRC regulations and licenses for the protection of personnel from exposure to radiation and/or radioactive material (e.g., posting and labeling of radioactive material) (10 CFR 19.12)
Responsibility to report promptly to the licensee any condition that may lead to or cause a violation of NRC regulations and licenses or unnecessary exposure to radiation and/or radioactive material (e.g., notification of the RSO regarding radiation protection issues) (10 CFR 19.12)
Appropriate response to warnings made in the event of any unusual occurrence or malfunction that may involve exposure to radiation and/or radioactive material (10 CFR 19.12)
Radiation exposure reports that workers may request, as per 10 CFR 19.13 (10 CFR 19.12)

Appendix I: Radiation Monitoring Instrument Specifications and Model Survey Instrument Calibration Program

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Facilities and Equipment

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To reduce doses received by individuals not calibrating instruments, calibrations should be conducted in an isolated area of the facility or at times when no one else is present.
Individuals conducting calibrations will wear assigned dosimetry, if required.

Equipment Selection

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Low energy beta emitters, such as carbon-14 and sulfur-35, are difficult to detect with Geiger-Mueller (GM) probes. Generally the detection efficiency is on the order of 2% for low energy beta emitters. Particular attention to surveying method (e.g., speed and height above surface) is needed to perform adequate surveys. Additionally, wipes should be taken and counted on a liquid scintillation counter to verify potential contamination.
Medium to high energy beta emitters, such as P-32 and Ca-45, can be detected with a pancake GM. The efficiency ranges 15 to 40%, depending on the beta energy.
Low energy gamma emitters, such as I-125, can be detected with a sodium iodide probe. If the probe possesses a thin window and thin crystal, the detection efficiency is approximately 20%.
Medium to high energy gamma emitters, such as Tc-99m, can be detected with either GM or sodium iodide probes, depending on the required sensitivity. The sensitivity of GM probes is much lower than for sodium iodide probes, therefore, sodium iodide probes are the preferred alternative.

Model Procedure for Calibrating Survey Instruments

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(Detailed information about survey instrument calibration may be obtained by referring to ANSI N323A-1997, "Radiation Protection Instrumentation Test and Calibration, Portable Survey Instruments." Copies may be obtained from the American National Standards Institute, 1430 Broadway, New York, NY 10018 or ordered electronically at the following address: <http://www.ansi.org>.

We will implement the following procedure when calibrating survey instruments:

Radiation survey instruments will be calibrated with a radioactive source in accordance with 10 CFR 35.61. Electronic calibrations alone are not acceptable. Survey meters must be calibrated at least annually, before first use, and after servicing. (Battery changes are not considered "servicing.") Instruments used to monitor higher energies are most easily calibrated in known radiation fields produced by sources of gamma rays of approximately the same energies as those to be measured. An ideal calibration source should emit the applicable radiation (e.g., alpha, beta, or gamma) with an energy spectrum similar to that to be measured and have a suitably long half-life.
A radioactive sealed source(s) used for calibrating survey instruments will:

Approximate a point source
Have its apparent source activity or the exposure rate at a given distance traceable by documented measurements to a standard certified to be within � 5% accuracy by NIST
Emit the type of radiation measured
Approximate the same energy (e.g., Cs-137, Co-60) as the environment in which the calibrated device will be employed
Provide a radiation dose rate sufficient to reach the full scale (<1000 mR/hr) of the instrument calibrated.

The inverse square and radioactive decay law must be used to correct changes in exposure rate due to changes in distance or source decay.
A record must be made of each survey meter calibration and retained for 3 years after each record is made (10 CFR 20.2103(a) and 10 CFR 35.2061).
Instrument readings should be within +10% of known radiation values at calibrated points; however, readings within + 20% shall be acceptable if a calibration chart or graph is prepared and made available with the instrument.
The kinds of scales frequently used on radiation survey meters are calibrated as follows:

Linear Readout Instruments must be calibrated at no fewer than two points on each scale. Calibration shall be checked near the ends of each scale (approximately 20% and 80% of each scale).
Logarithmic Readout Instruments must be calibrated at one point (midpoint) on each decade.
Digital Readout Instruments with either manual or automatic scale switching for indicating exposure rates must be calibrated at no fewer than two points on each scale. Calibration shall be checked near the ends of each scale (approximately 20% and 80% of each scale).
Digital Readout Instruments without scale switching for indicating exposure rates must be calibrated at one point (midpoint) on each decade.
Integrating Instruments must be calibrated at two dose rates (approximately 20% and 80% of the dose rate range).

Readings above 1000 mR/hr (250 microcoulombs/kilogram per hour) need not be calibrated; however, such scales may be checked for operation and approximately correct response.
Survey meter calibration reports will indicate the procedure used and the data obtained. The description of the calibration should include:

A description of the instrument, including the manufacturer's name, model number, serial number, and type of detector
A description of the calibration source, including the calibration procedure and the exposure rate at a specified distance on a specified date
For each calibration point, the calculated exposure rate, the indicated exposure rate, the deduced correction factor (the calculated exposure rate divided by the indicated exposure rate), and the scale selected on the instrument
The exposure reading indicated with the instrument in the "battery check" mode (if available on the instrument)
For instruments with external detectors, the angle between the radiation flux field and the detector (i.e., parallel or perpendicular)
For instruments with internal detectors, the angle between the radiation flux field and a specified surface of the instrument
For detectors with removable shielding, an indication whether the shielding was in place or removed during the calibration procedure
The exposure rate from a check source, if used
The name of the person who performed the calibration and the date it was performed

The following information should be attached to the instrument as a calibration sticker or tag:

The source that was used to calibrate the instrument
The proper deflection in the battery check mode (unless this is clearly indicated on the instrument)
Special use conditions (e.g., an indication that a scale or decade was checked only for function but not calibrated)
The date of calibration and the next calibration due date
The apparent exposure rate from the check source, if used.

Calculating the Efficiency of the NaI(Tl) Uptake Probe

The sodium iodide (thallium doped) [NaI(Tl)] uptake probe is commonly used for bioassays of personnel administering I-131 radionuclides in the form of sodium iodide. RG 8.20 gives the details of bioassay requirements for I-131 radionuclides. Appendix B to Part 20 considers the Annual Limits on Intake (ALIs) and Derived Air Concentrations (DACs) of radionuclides for occupational exposure. Converting counts per minute (cpm) to disintegrations per minute (dpm) in determination of accurate bioassay values is of importance in determining thyroid burdens with radioiodine. We will calculate the efficiency of nuclear counting systems on an annual basis, before first use, and/or after repair, using the following procedure:

Check the instrument's counting efficiency using either a standard source of the same radionuclide as the source being tested or one with similar energy characteristics. Accuracy of standards will be within + 5% of the stated value and traceable to a primary radiation standard such as those maintained by NIST.
Calculate efficiency of the instrument.

For example:	[(cpm from std) - (cpm from bkg)] activity of std in microcuries		=	efficiency in cpm/microcurie

	where:	cpm	=	counts per minute
		std	=	standard
		bkg	=	background

The date of the efficiency test will be attached to the instrument as a calibration sticker or tag and the following information should be attached:

The date of the next efficiency due
Results of efficiency calculation(s)

Calculating the Gamma Well Efficiency of Counting Equipment

Gamma well counting equipment is often used for assaying the wipe testing of packages, sealed sources, and areas where unsealed byproduct material is prepared, administered, or stored. Converting cpm to dpm using smear wipes is required when dealing with radiation surveys of sealed and unsealed radioactive materials. We will calculate the efficiency of all instruments used for assaying wipe tests on an annual basis, before first use, and/or after repair, using the following procedure:

Check the instrument's counting efficiency, using either a standard source of the same radionuclide as the source being tested or one with similar energy characteristics. Accuracy of standards will be within + 5% of the stated value and traceable to a primary radiation standard such as those maintained by NIST.
Calculate efficiency of the instrument.

For example:	[(cpm from std) - (cpm from bkg)] activity of std in microcuries		=	efficiency in cpm/microcurie

	>where:	cpm	=	counts per minute
		std	=	standard
		bkg	=	background

The date of the efficiency test will be attached to the instrument as a calibration sticker or tag and the following information should be attached:

The date of the next efficiency due
Results of efficiency calculation(s)

Reference: Draft RG FC 413-4, "Guide for the Preparation of Applications for Licenses for the Use of Radioactive Materials in Calibrating Radiation Survey and Monitoring Instruments," dated June 1985.

Appendix J: Model Procedures for Dose Calibrator Calibration

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Model Procedures for Testing Dose Calibrators Used to Measure Photon-Emitting Radionuclides

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We will test for the following at the indicated frequency. We will consider repair, replacement, or arithmetic correction if the dose calibrator falls outside the suggested tolerances. We will record, for all tests, the name of the individual who performed the test.

Note: A licensee shall repair or replace the dose calibrator if the accuracy or constancy error exceeds 10% and shall mathematically correct dosage readings [for dosages greater than 1.11 MBq (30 �Ci)] if the geometry or linearity error exceeds 10%.

Constancy, at least once each day prior to assay of patient dosages (+10%).
Linearity, at installation and at least annually thereafter (+10%).
Geometry dependence, at installation (+10%).
Accuracy, at installation and at least annually thereafter (+10%).

After repair, adjustment, or relocation to another building of the dose calibrator, we will repeat the above tests before use.

Constancy means reproducibility in measuring a constant source over a long period of time. We will assay at least one relatively long-lived source such as Cs-137, Co-60, cobalt-57 (Co-57)⁽⁴⁾, or radium-226 (Ra-226)² using a reproducible geometry each day before using the calibrator. We will consider the use of two or more sources with different photon energies and activities.

We will use the following procedure:

1.	Assay each reference source using the appropriate dose calibrator setting (e.g., use the Cs-137 setting to assay Cs-137).
2.	Measure background at the same setting, and subtract or confirm the proper operation of the automatic background subtract circuit if it is used.
3.	For each source used, record (e.g., plot, log, etc.) the activity measured, the model and serial number of the instrument, the identity of the radionuclide contained in the check source, and the date of the check.
4.	Using one of the sources, repeat the above procedure for all commonly used radionuclide settings. Record (e.g., plot, log, etc.) the results.
5.	Notify the RSO or the AU if the test results fall outside +10% of the expected results.

Linearity means that the calibrator is able to indicate the correct activity over the range of use of that calibrator. The linearity of a dose calibrator will be ascertained over the range of its use between the maximum activity administered and 1.1 MBq (30 �Ci). This test will be done using a vial or syringe of technetium-99m (Tc-99m) whose activity is at least as large as the maximum activity normally assayed for administration.

Time Decay Method

We will use the following procedure:

1.	Assay the Tc-99m syringe or vial in the dose calibrator, and subtract background to obtain the net activity in millicuries. Record the date, time to the nearest minute, and net activity on the dose calibrator linearity test form.
2.	Repeat the assay at approximately 4 hour intervals during the workday. Continue on subsequent days until the assayed activity is less than 1.1 MBq (30 �curies). For dose calibrators on which you select a range with a switch, select the range you would normally use for the measurement.
3.	Convert the time and date information you recorded to hours elapsed since the first assay.
4.	Record the measured activities, the calculated activities, the time elapsed between measurements, the model number and serial number of the dose calibrator, and the date(s) of the test.
5.	Notify the RSO if the worst deviation is more than +10%.

Shield Method

If we decide to use a set of "sleeves" of various thicknesses to test for linearity, it will first be necessary to calibrate the sleeves.

We will use the following procedure:

Note: The applicant should review the procedure for calibrating sleeves against the manufacturer's instructions. Some sleeve manufacturer's procedures indicate that various sleeves should be stacked to achieve a desired attenuation. The following procedure should be modified to allow for stacking of sleeves.

1.	Begin the linearity test as described in the decay method described above. After making the first assay, the sleeves can be calibrated as follows. Steps 2 through 4 below must be completed within 6 minutes (i.e., approximately 1% of decay of Tc-99m).
2.	Put the base and sleeve 1 in the dose calibrator with the vial. Record the sleeve number and indicated activity.
3.	Remove sleeve 1 and put in sleeve 2. Record the sleeve number and indicated activity
4.	Continue for all sleeves.
5.	Complete the decay method linearity test steps 2 through 5 above.
6.	From the data recorded in step 4 of the decay method, find the decay time associated with the activity indicated with sleeve 1 in place. This is the "equivalent decay time" for sleeve 1. Record that time with the data recorded in step 2.
7.	Find the decay time associated with the activity indicated with sleeve 2 in place. This is the "equivalent decay time" for sleeve 2. Record that time with the data recorded in step 3.
8.	Continue for all sleeves.
9.	The table of sleeve numbers and equivalent decay times constitutes the calibration of the sleeve set.

The sleeve set may now be used to test dose calibrators for linearity.

1.	Assay the Tc-99m syringe or vial in the dose calibrator, and subtract background to obtain the net activity. Record the net activity.
2.	Steps 3 through 5 below must be completed within 6 minutes.
3.	Put the base and sleeve 1 in the dose calibrator with the vial. Record the sleeve number and indicated activity.4. Remove sleeve 1 and put in sleeve 2. Record the sleeve number and indicated activity.
5.	Continue for all sleeves.
6.	Record the measured activities, the calculated activities, the time elapsed between measurements, the model number and serial number of the dose calibrator, and the date(s) of the test.
7.	Notify the RSO if the worst deviation is more than +10%.

Geometry independence means that the indicated activity does not change with volume or configuration. The test for geometry independence will be conducted using syringes and vials that are representative of the entire range of size, shape, and constructions normally used for injections and a vial similar in size, shape, and construction to the generator and radiopharmaceutical kit vials normally used. The following test assumes injections are done with 3-centimeter cubed (cc) plastic syringes and that radiopharmaceutical kits are made in 30-cc glass vials and your predetermined safety margin is +10%.

Note: If you do not use these volumes, change the procedure so that your syringes and vials are tested throughout the range of volumes commonly used.

We will use the following procedure:

1.	In a small beaker or vial, mix 2 cc of a solution of Tc-99m with an activity concentration between 1 and 10 mCi/milliliter (ml). Set out a second small beaker or vial with water.
2.	To test the geometry dependence for a 3-cc syringe, draw 0.5 cc of the Tc-99m solution into the syringe and assay it. Record the volume and activity (e.g., mCi) indicated.
3.	Remove the syringe from the calibrator, draw an additional 0.5 cc of water, and assay again. Record the volume and activity indicated.
4.	Repeat the process until you have assayed a 2.0-cc volume.
5.	Select as a standard the volume closest to that normally used for injections. For all the other volumes, divide the standard activity by the activity indicated for each volume. The quotient is a volume correction factor. Alternately, you may graph the data and draw horizontal 10% error lines above and below the chosen "standard volume."
6.	Record the model and serial number of the dose calibrator, the configuration of the source measured, the activity measured for each volume measured, and the date of the test.
7.	Notify the RSO if any correction factors are greater than 1.1 or less than 0.9, or if any data points lie outside the + 10% error lines.
8.	To test the geometry dependence for a 30-cc glass vial, draw 1.0 cc of the Tc-99m solution into a syringe and then inject it into the vial. Assay the vial. Record the volume and activity indicated.
9.	Remove the vial from the calibrator and, using a clean syringe, inject 2.0 cc of water, and assay again. Record the volume and activity indicated.
10.	Repeat the process until you have assayed a 19.0-cc volume. The entire process must be completed within 10 minutes.
11.	Select as a standard the volume closest to that normally used for mixing radiopharmaceutical kits. For all the other volumes, divide the standard activity by the activity indicated for each volume. The quotient is a volume correction factor. Alternatively, you may graph the data and draw horizontal 10% error lines above and below the chosen "standard volume."
12.	Record the model and serial number of the dose calibrator, the configuration of the source measured, the activity measured for each volume measured, and the date of the test.
13.	Notify the RSO if any correction factors are greater than 1.1 or less than 0.9, or if any data points lie outside the + 10% error lines.

Accuracy means that, for a given calibrated reference source, the indicated activity (e.g., mCi) value is equal to the activity value determined by NIST or by the supplier who has compared that source to a source that was calibrated by NIST. Certified sources are available from the NIST and from many radionuclide suppliers. At least one source with a principal photon energy between 100 keV and 500 keV (e.g., Co-57 or Ba-133) will be used. We will consider using at least one reference source whose activity is within the range of activities normally assayed.

We will use the following procedure:

1.	Assay a calibrated reference source at the appropriate settings (i.e., use the Co-57 setting to assay Co-57), and then remove the source and measure background. Subtract background from the indicated activity to obtain the net activity. Record the net activity.
2.	The measurement should be within +10% of the certified activity of the reference source, mathematically corrected for decay.
3.	Repeat the procedure for any other calibrated reference sources possessed.
4.	Record the model and serial number of the dose calibrator, the model and serial number of each source used, the identity of the radionuclide contained in the source and its activity, the date of the test, and the results of the test.
5.	Notify the RSO if the test results do not agree, within +10%, with the certified value of the reference source(s).
6.	At the same time the accuracy test is done, assay the source that will be used for the daily constancy test (it need not be a certified reference source) on all commonly used radionuclide settings.

We will, through the RSO, ensure that the operation of the dose calibrator is in accordance with approved procedures and regulatory requirements.

Appendix K: Suggested Medical Licensee Audit

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Annual Radiation Protection Medical Licensee Audit

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Note: All areas indicated in audit notes may not be applicable to every license and may not need to be addressed during each audit. For example, licensees do not need to address areas which do not apply to the licensee's activities, and activities that have not occurred since the last audit need not be reviewed at the next audit.

Date of This Audit:	_________________________	Date of Last Audit:	_________________________

>Next Audit Date:	_________________________


>Auditor:	_________________________		Date:	_________________________
	(Signature)

>Management Review:	_________________________		Date:	_________________________
	(Signature)

Audit History

A. Were previous audits conducted annually [20.1101]?

B. Were records of previous audits maintained [20.2102]?

C. Were any deficiencies identified during previous audit?

D. Were corrective actions taken? (Look for repeated deficiencies).

Organization and Scope of Program

A.	Radiation Safety Officer
	1.	If the RSO was changed, was license amended [35.13]?
	2.	Does new RSO meet NRC training requirements [35.50 or 35.900, 35.57, 35.59]?
	3.	Is RSO fulfilling his/her duties [35.24]?
B.	Multiple places of use? If yes, list locations
C.	Are all locations listed on license? [L/C]
D.	Were annual audits performed at each location [20.1101]? If no, explain.
E.	Describe scope of the program (staff size, number of procedures performed, etc.).
F.	Licensed Material
	1.	Isotope, chemical form, quantity and use as authorized [L/C]?
	2.	Does the total amount of radioactive material possessed require financial assurance [30.35(a)]? If so, is financial assurance adequate?
	3.	Calibration and reference sources [35.65]?
		a.	Sealed sources manufactured and distributed by a person licensed pursuant to 10 CFR 32.74 or equivalent Agreement State regulations and sources do not exceed 30 millicuries each [35.65(a)]?
		b.	Any byproduct material with a half-life not longer than 120 days in individual amounts not to exceed 15 millicuries [35.65(b)]?
		c.	Any byproduct material with a half-life longer than 120 days in individual amounts not to exceed 200 microcuries each and does not exceed 10E3 times the quantities in Appendix B of Part 30 [35.65(c)]?
		d.	Technetium-99m in individual amounts as needed [35.65(d)]?
	4.	Unsealed material used under 35.100, 200, and 300 are:
		a.	Obtained from a manufacturer or properly licensed organization[32.72]? AND/OR
		b.	Prepared by a physician user, an authorized nuclear pharmacist, or an individual under the supervision of an authorized nuclear pharmacist or physician user?
G.	Are the sealed sources possessed as described in the Sealed Source and Device Registration (SSDR) Certificate? Are copies of (or access to) SSDR Certificates possessed? Are manufacturers' manuals for operation and maintenance of medical devices possessed? [32.210]?
H.	Are the actual uses of medical devices consistent with the authorized uses listed on the license?
I.	If places of use changed, was the license amended [35.13 (e)]?
J.	If control of license was transferred or bankruptcy filed, was NRC prior consent obtained or notification made, respectively [30.34(b)]?

Radiation Safety Program

A. Minor changes pursuant to [35.26]?

B. Records of changes maintained [35.2026]?

C. Content and implementation reviewed annually by the licensee [20.1101(c)]?

D. Records of reviews maintained [20.2102]?

Use by Authorized Individuals [L/C]

Compliance is established by meeting at least one criterion under each category.

A.	Authorized Nuclear Pharmacist [35.55 or 35.980, 35.57, 35.59] Does Not Apply to Facilities That Are Registered/Licensed by FDA/State Agency as a Drug Manufacturer with Distribution Regulated Under Part 32
	1.	Certified by specialty board
	2.	Identified on NRC or Agreement State license
	3.	Identified on permit issued by broad scope licensee
	4.	Listed on facility license
B.	Authorized User [35.57, 35.59, and 35.290, 35.292, 35.390, 35.490, 35.590, 35.690, or Subpart J]
	1.	Certified by specialty board
	2.	Identified on NRC or Agreement State license
	3.	Identified on permit issued by broad scope licensee
	4.	Listed on facility license
C.	Authorized Medical Physicist [35.51 or 35.961, 35.57, 35.59]
	1.	Certified by specialty board
	2.	Identified on NRC or Agreement State license
	3.	Identified on permit issued by broad scope licensee
	4.	Listed on facility license

Mobile Service

A.	Operates services per [35.80, 35.647]?
B.	Compliance with 20.1301 evaluated and met?
C.	Letter signed by management of each client [35.80(a)]?
D.	Licensed material was not delivered to client's address (unless client was authorized) [35.80(b)]?
E.	Dosage measuring instruments checked for proper function before used at each address of use or on each day of use, if more frequent [35.80(c)]?
F.	Survey instruments checked for proper operation before used at each address of use [35.80(d)]?
G.	Survey of all areas of use prior to leaving each client address [35.80(e)]?
H.	Additional technical requirements for mobile remote afterloaders per [35.647]?

Amendments Since Last Inspection [35.13]

A. Any Amendments since last inspection [35.13]?

Notifications Since Last Inspection [35.14]

A.	Appropriate documentation provided to the NRC for authorized nuclear pharmacists, authorized medical physicist, or authorized user no later than 30 days after the individual starts work [35.14(a)]?
B.	NRC notified within 30 days after authorized user, authorized nuclear pharmacist, authorized medical physicist, or RSO stops work or changes name, licensee's mailing address or name changes without a transfer of control of the license, or licensee has added to or changed an area of use for 35.100 or 35.200 use [35.14(b)]?
C.	Any Notifications since last inspection [35.14]?

Training, Retraining, And Instructions to Workers

A.	Have workers been provided with required instructions [19.12, 35.27]?
B.	Is the individual's understanding of current procedures and regulations adequate?
C.	Training program implemented?
	1.	Operating procedures [35.310, 35.410, 35.610]?
	2.	Emergency procedures [35.310, 35.410, 35.610]?
	3.	Periodic training required and implemented [35.310, 35.410, 35.610]?
	4.	Were all workers who are likely to exceed 1 mSv (100 mrem) in a year instructed and was refresher training provided, as needed [10 CFR 19.12]?
	5.	Was each supervised user instructed in preparation of material, principles and procedures for radiation safety, device usage and administration of written directives, as appropriate [35.27]?
	6.	Are initial and periodic training records maintained for each individual [35.2310]?
	7.	Briefly describe training program:
D.	Additional therapy device instructions/training
	1.	Unit operation, inspection, associated equipment, survey instruments [35.610]?
	2.	License conditions applicable to the use of the unit?
	3.	Emergency drills [35.610]?
E.	Part 20
	Workers cognizant of requirements for:
	1.	Radiation Safety Program [35.24, 35.26, 20.1101]?
	2.	Annual dose limits [20.1301, 20.1302]?
	3.	New forms 4 and 5?
	4.	10% monitoring threshold [20.1502]?
	5.	Dose limits to embryo/fetus and declared pregnant worker [20.1208]?
	6.	Grave Danger Posting [20.1902]?
	7.	Procedures for opening packages [20.1906]?
F.	Supervision of individuals by authorized user in accordance with 10 CFR 35.27?

Manual Brachytherapy And Unsealed Therapy Training

A.	Safety instruction to personnel provided include [10 CFR 35.310, 10 CFR 35.410]:

	1.	Control of patient and visitors?
	2.	Routine visitation to patients in accordance with 10 CFR 20.1301(a)(1) and (3)?
	3.	Contamination control and size/appearance of sources?
	4.	Safe handling and shielding instructions?
	5.	Waste control?
	6.	RSO and AU notification in emergency or death?
	7.	Records retained [35.2310]?

Facilities

A.	Facilities as described in license application?
B.	Therapy device facilities provided with electrical interlock system, viewing and intercom systems, radiation monitor, source retraction mechanism, and source indicator lights [35.615]?
C.	Emergency source recovery equipment available [35.415, 35.615]?
D.	Storage areas:
	1.	Materials secured from unauthorized removal or access [20.1801]?
	2.	Licensee controls and maintains constant surveillance of licensed material not in storage [20.1802]?
E.	Therapy unit operation:
	1.	Device, sources, and console keys controlled adequately [20.1801, 35.610(a)(1)]?
	2.	Restricted to certain source orientations and/or gantry angles [L/C]?
	3.	Ceases to operate in restricted orientation(s) [L/C]?
	4.	Only one radiation device can be placed in operation at a time within the treatment room [L/C]?

Dose or Dosage Measuring Equipment

A.	Possession, use, calibration, and check of instruments to measure activities of photon emitting radionuclides [10 CFR 35.60]:
	1.	Instrumentation possessed and used?
	2.	Constancy and proper operation checked at the beginning of each day of use?
	3.	Accuracy, linearity, and geometry dependence tests performed before initial use and following repair for each instrument?
	4.	Accuracy and linearity tests performed annually?
	5.	Dosage readings mathematically corrected for geometry or linearity errors greater than +10%?
	6.	Instrument repaired or replaced when accuracy or constancy error exceeded +10%?
	7.	Approved procedures followed?
	8.	Records maintained and include required information [10 CFR 35.2060]?
B.	Instrumentation - Alpha- or beta-emitting radionuclides [35.62]?
	1.	List type of equipment used to assay alpha and beta particles:
	2.	Approved procedures for use of instrumentation followed?
	3.	Accuracy, linearity, and geometry dependence tests performed before initial use and following repair?
	4.	Constancy and proper operation checked at the beginning of each day of use?
	5.	Accuracy and linearity tests performed annually?
	6.	Dosage readings mathematically corrected for geometry or linearity errors greater than +10%?
	7.	Instrument repaired or replaced when accuracy or constancy error exceeded +10%?
	8.	Records maintained [10 CFR 35.2060]?
C.	Determination of dosages of unsealed byproduct material [35.63]?
	1.	Each dosage determined and recorded prior to medical use [35.63(a)]?
	2.	Measurement made either by direct measurement or by decay correction [35.63(b)]?
	3.	Measurement for a dosage of photon-emitting radionuclide prepared by the licensee made by direct measurement or by combination of measurement and calculation [35.63(c)]?
	4.	Measurement of an alpha- or beta-emitting radionuclide prepared by the licensee made by direct measurement or by combination of measurements and calculations [35.63(c)]?
D.	Licensee uses generators?
	1.	First eluate after receipt used for radiopharmaceuticals tested for Mo-99 breakthrough [35.204(b)]?
	2.	No radiopharmaceuticals administered with Mo-99 concentrations over 0.15 �Ci per mCi of Tc-99m [35.204(a)]?
	3.	Records maintained [35.2204]?
E.	Dosimetry Equipment [35.630]
	1.	Calibrated system available for use [35.630(a)]?
	2.	Calibrated by NIST or an AAPM-accredited lab within previous two years and after servicing [35.630(a)(1)] OR calibrated by intercomparison per [35.630(a)(2)]?
	3.	Calibrated within the previous four years [35.630(a)(2)]?
	4.	Licensee has available for use a dosimetry system for spot-check measurements [35.630(b)]?
	5.	Record of each calibration, intercomparison, and comparison maintained [35.2630]?

Radiation Protection And Control of Radioactive Material

A.	Use of radiopharmaceuticals:
	1.	Protective clothing worn?
	2.	Personnel routinely monitor their hands?
	3.	No eating/drinking in use/storage areas?
	4.	No food, drink, or personal effects kept in use/storage areas?
	5.	Proper dosimetry worn?
	6.	Radioactive waste disposed in proper receptacles?
	7.	Syringe shields and vial shields used [35.69]?
B.	Leak tests and Inventories:
	1.	Leak test performed on sealed sources and brachytherapy sources [35.67(b)(1)]?
	2.	Inventory of sealed sources and brachytherapy sources performed at intervals not to exceed six months [35.67(g)]?
	3.	Records maintained [35.2067]?

Radiation Survey Instruments

A.	Survey instruments used to show compliance with Part 20 and 30.33(a)(2).
	1.	Appropriate operable survey instruments possessed or available [10 CFR Part 20]?
	2.	Calibrations [35.61(a) and (b)]:
		a.	Before first use, annually & after repairs?
		b.	Within 20% in each scale or decade of interest?
	3.	Records maintained [35.2061]?
B.	Radiation surveys performed [20.1501, 35.70]?
	1.	Daily in all areas where radiopharmaceuticals requiring a written directive are prepared or administered (except patient rooms) [35.70]?
	2.	Weekly in all areas where radiopharmaceuticals or waste is stored?
	3.	Weekly wipes in all areas where radiopharmaceuticals are routinely prepared, administered or stored?
	4.	Trigger levels established?
	5.	Corrective action taken and documented if trigger level exceeded?
	6.	Techniques can detect 0.1 mR/hr, 2000dpm?
	7.	Quarterly in brachytherapy source storage area?
	8.	Surveys as defined in the Sealed Source and Device Registry made to assure that the maximum radiation levels and average radiation levels from the surface of the main source safe with the sources(s) in the shielded position does not exceed the levels stated in the Registry [35.652(a)] and records maintained [35.2652]?
		a.	After new source installation?
		b.	Following repairs to the source(s) shielding, the source(s) driving unit, or other electronic or mechanical mechanism that could expose the source, reduce the shielding around the source(s), or compromise the radiation safety of the device or the source(s)?

Public Dose

A.	Is licensed material used in a manner to keep doses below 1mSv (100 mrem) in a year [10 CFR 20.1301(a)(1)]?
B.	Has a survey or evaluation been performed per 10 CFR 20.1501(a)?
C.	Have there been a ny additions or changes to the storage, security, or use of surrounding areas that would necessitate a new survey or evaluation?
D.	Do unrestricted area radiation levels exceed 0.02 mSv (2 mrem) in any one hour [10 CFR 20.1301(a)(2)]?
E.	Is licensed material used or stored in a manner that would prevent unauthorized access or removal [10 CFR 20.1801]?
F.	Records maintained [10 CFR 20.2103, 10 CFR 20.2107]?

Patient Release

A.	Individuals released when TEDE less than 0.5 rem [35.75(a)]?
B.	Instructions to breast-feeding women include required information [35.75(b)]?
C.	Release records maintained if 35.2075(b) criteria are met [35.2075(a)]?
D.	Records of instructions given to breast-feeding women maintained, if required [35.2075(c)]?

Radiopharmaceutical Therapy

A.	Safety precautions implemented to include patient facilities, posting, stay times, patient safety guidance, release and contamination controls [35.315(a)]?
B.	RSO and AU promptly notified if patient died or had a medical emergency [35.315(b)]?

Brachytherapy

A.	Safety precautions implemented to include posting, stay times, and emergency response equipment [35.415]?
B.	Patients surveyed immediately after implant [35.404(a)]?
C.	Patients surveyed immediately after removing the last temporary implant source [35.404(b)]?
E.	Records maintained [35.2404]?

Radioactive Waste

A.	Disposal
	1.	Decay-in-storage [35.92]
	2.	Procedures followed [35.92]?
	3.	Labels removed or defaced [20.1904, 35.92]?
B.	Special procedures performed as required [L/C]?
C.	Improper/unauthorized disposals [20.2001]?
D.	Records maintained [20.2103(a), 20.2108, 35.2092]?
E.	Effluents
	1.	Release to sanitary sewer [20.2003]?
		a.	Material is readily soluble or readily dispersible [20.2003(a)(1)]?
		b.	Monthly average release concentrations do not exceed 10 CFR Part 20 App. B, Table 2 values?
		c.	No more than 5 Ci of H-3, 1 Ci of C-14 and 1 Ci of all other radionuclides combined released in a year [20.2003]?
		d.	Procedures to ensure representative sampling and analysis implemented [20.1501]?
	2.	Release to septic tanks [20.2003]?
		a.	Within unrestricted limits [10 CFR Part 20 App. B, Table 2, Part 20]?
	3.	Waste incinerated?
		a.	License authorizes [20.2004(a)(3)]?
		b.	Directly monitor exhaust?
		c.	Airborne releases evaluated and controlled [20.1501, 20.1701]?
	4.	Air effluents and ashes controlled [20.1101, 1201, 1301, 1501, 2001, L/C]? {See also IP 87102, RG 8.37}
		a.	Air effluent less than 10 mrem constraint limit [20.1101]
		b.	If no, reported appropriate information to NRC
			i.	Corrective actions implemented and on schedule?
		c.	Description of effluent program
			i.	Monitoring system hardware adequate?
			ii.	Equipment calibrated, as appropriate?
			iii.	Air samples/sampling technique (i.e. charcoal, HEPA, etc.) analyzed with appropriate instrumentation?
F.	Waste storage
	1.	Protection from elements and fire?
	2.	Control of waste maintained [20.1801]?
	3.	Containers properly labeled and area properly posted [20.1902, 20.1904]?
	4.	Package integrity adequately maintained?
G.	Waste disposal
	1.	Sources transferred to authorized individuals [20.301, 20.2001, 30.51]?
	2.	Name of organization:
H.	Records of surveys and material accountability are maintained [20.2103, 20.2108]?

Receipt And Transfer of Radioactive Material

A.	Describe how packages are received and by whom:
B.	Written package opening procedures established and followed [20.1906(e)]?
C.	All incoming packages with a DOT label wiped, unless exempted (gases and special form) [20.1906(b)(1)]?
D.	Incoming packages surveyed [20.1906(b)(2)]?
E.	Monitoring in (C) and (D) performed within time specified [20.1906(c)]?
F.	Transfer(s) performed per [30.41]?
G.	All sources surveyed before shipment and transfer [20.1501(a), 49 CFR 173.475(I)]?
H.	Records of surveys and receipt/transfer maintained [20.2103(a), 30.51]?
I.	Package receipt/distribution activities evaluated for compliance with 20.1301?

Transportation (10 CFR 71.5(a) and 49 CFR 171-189)

A.	Shipments are:
	[ ] delivered to common carriers [ ] transported in own private vehicle [ ] both [ ] no shipments since last audit
B.	Return radiopharmacy doses or sealed sources?
	1.	Licensee assumes shipping responsibility?
	2.	If NO, describe arrangements made between licensee and radiopharmacy for shipping responsibilities:
C.	Packages
	1.	Authorized packages used [173.415, 416]?
	2.	Performance test records on file?
		a.	DOT-7A packages [173.415(a)]
		b.	Special form sources [173.476(a)]
	3.	Two labels (White-I, Yellow-II, Yellow-III) with TI, Nuclide, Activity, and Hazard Class [172.403, 173.441]?
	4.	Properly marked (Shipping Name, UN Number, Package Type, RQ, "This End Up" (liquids), Name and Address of consignee) [172.301, 304, 310, 312, 324]?
	5.	Closed and sealed during transport [173.475(f)]?
D.	Shipping Papers
	1.	Prepared and used [172.200(a)]?
	2.	Proper {Shipping Name, Hazard Class, UN Number, Quantity, Package Type, Nuclide, RQ, Radioactive Material, Physical and Chemical Form, Activity, Category of Label, TI, Shipper's Name, Certification and Signature, Emergency Response Phone Number, "Limited Quantity" (if applicable), "Cargo Aircraft Only" (if applicable)} [172.200-204]?
	3.	Readily accessible during transport [177.817(e)]?

Teletherapy And Gamma Stereotactic Radiosurgery Servicing

A.	Inspection and servicing performed following source replacement or at intervals not to exceed 5 years [35.655(a)]?
B.	Licensee arranged for needed service identified during the inspection?
C.	Service performed by persons specifically authorized to do so [35.655(b)]?

Full Calibration-Therapeutic Medical Devices

A.	Licensee uses one of the proper protocols (TG-21, AAPM 54, TG-56, TG-40) [35.632, 633, 635]?
B.	Performed prior to first patient use [35.632(a)(1), 633(a)(1), 635(a)(1)]?
C.	At intervals not to exceed one year for teletherapy, gamma stereotactic, and LDR remote afterloader; at intervals not exceeding 120 days for HDR and PDR remote afterloaders [35.632(a)(3)], 633(a)(3) and (4), 635(a)(3)]?
D.	Whenever spot-checks indicate output differs from expected by +5% [35.632(a)(2)(i), 633(a)(2)(i), 635(a)(2)(i)]?
E.	After source exchange, relocation, and major repair or modification [35.632(a)(2), 633(a)(2), 635(a)(2)]?
F.	Performed with properly calibrated instrument [35.632(c), 35.633(d), 35.635(c)]?
G.	Includes:
	1.	For teletherapy:
		a.	Output measured within +3% of expected for the range of field sizes, range of distances [35.632(b)(1)]?
		b.	Coincidence of radiation field and field light localizer [35.632(b)(2)]?
		c.	Uniformity of radiation field and beam angle dependence [35.632(b)(3)]?
		d.	Timer constancy and linearity over the range of use [35.632(b)(4)]?
		e.	On-off error [35.632(b)(5)]?
		f.	Accuracy of all measuring and localization devices [35.632(b)(6)]?
	2.	For HDR and PDR remote afterloaders:
		a.	Output measured within +5% of expected [35.633(b)(1)]?
		b.	Source positioning accuracy within +1 millimeter [35.633(b)(2)]?
		c.	Source retraction with backup battery upon power failure [35.633(b)(3)]?
		d.	Electrically assisted treatment room doors with power turned off [35.633(b)(4)]?
		e.	Source guide tubes [35.633(c)(1)]?
		f.	Timer accuracy and linearity over the range of use [35.633(c)(1)]?
		g.	Length of the connectors [35.633(c)(1)]?
		h.	Annual check of source guide tube and connector function [35.633(c)(1)]?
	3.	For LDR remote afterloaders:
		a.	Output measured within +5% of expected [35.633(b)(1)]?
		b.	Source positioning accuracy within +1 millimeter [35.633(b)(2)]?
		c.	Source retraction with backup battery upon power failure [35.633(b)(3)]?
		d.	Autoradiograph of the source(s) to verify source(s) arrangement and inventory [35.633(c)(2)]?
		e.	Quarterly spot check of the absolute timer accuracy [35.633(c)(2)]?
	4.	For gamma stereotactic radiosurgery:
		a.	Output measured within +3% of expected [35.635(b)(1)]?
		b.	Helmet factors [35.635(b)(2)]?
		c.	Isocenter coincidence [35.635(b)(3)]?
		d.	Timer accuracy and linearity over the range of use [35.635(b)(4)]?
		e.	On-off error [35.635(b)(5)]?
		f.	Trunnion centricity [35.635(b)(6)]?
H.	Output corrected mathematically [35.632(e), 35.633(f), 35.635(e)]?
I.	Records maintained [35.2632, 35.2633, 35.2635]?

Periodic Spot Checks For Therapeutic Devices

A.	Performed at required frequency [35.642(a), 643(a), 644(a), 645(a)]?
B.	Procedures established by medical physicist [35.642(b), 35.643(b), 35.644(c), 35.645(b)]?
C.	Procedures followed?
D.	Medical physicist reviews results within 15 days [35.642(c), 35.643(b), 35.644(c), 35.645(b)]?
E.	Performed with properly calibrated instrument [35.642(a)(5), 35.643(c)(2), 35.645(c)(2)(i)]?
F.	Output and safety spot checks include:
	1.	For teletherapy:
		a.	Timer constancy and linearity over the range of use [35.642(a)(1)]?
		b.	On-off error [35.642(a)(2)]?
		c.	Coincidence of radiation field and field light localizer [35.642(a)(3)]?
		d.	Accuracy of all measuring and localization devices [35.642(a)(4)]?
		e.	The output for one typical set of operating conditions [35.642(a)(5)]?
		f.	Difference between measured and expected output [35.642(a)(6)]?
		g.	Interlock systems [35.642(d)(1)]?
		h.	Beam stops and dead-man switches [35.642(d)(2)]?
		i.	Source exposure indicator lights [35.642(d)(3)]?
		j.	Viewing and intercom systems [35.642(d)(4)]?
		k.	Treatment room doors, inside and out [35.642(d)(5)]?
		l.	Electrical treatment doors with power shut off [35.642(d)(6)]?
	2.	For HDR and PDR remote afterloaders:
		a.	Source positioning accuracy [35.643(c)(1)]?
		b.	The output [35.643(c)(2)]?
		c.	Difference between measured and expected output [35.643(c)(3)]?
		d.	Interlock systems [35.643 (d)(1)]?
		e.	Source exposure indicator lights [35.643(d)(2)]?
		f.	Viewing and intercom systems [35.643(d)(3)]?
		g.	Emergency response equipment [35.643(d)(4)]?
		h.	Radiation monitors used to indicate source position [35.643(d)(5)]?
		i.	Timer constancy [35.643(d)(6)]?
		j.	Clock (date and time) in the unit's computer [35.643(d)(7)]?
		k.	Simulated cycle of treatment [35.643(e)]?
	3.	For LDR remote afterloaders:
		a.	Interlock systems [35.644(a)(1)]?
		b.	Source exposure indicator lights [35.644(a)(2)]?
		c.	Viewing and intercom systems [35.644(a)(3)]?
		d.	Emergency response equipment [35.644(a)(4)]?
		e.	Radiation monitors used to indicate source position [35.644(a)(5)]?
		f.	Timer constancy [35.644(a)(6)]?
		g.	Clock (date and time) in the unit's computer [35.644(a)(7)]?
		h.	Simulated cycle of treatment [35.644(b)]?
	4.	For gamma stereotactic radiosurgery:
		a.	Treatment table retraction mechanism [35.645(c)(1)(i)]?
		b.	Helmet microswitches [35.645(c)(1)(ii)]?
		c.	Emergency timing circuits [35.645(c)(1)(iii)]?
		d.	Emergency off buttons [35.645(c)(1)(iv)]?
		e.	Stereotactic frames and localizing devices [35.645(c)(1)(v)]?
		f.	The output for one typical set of operating conditions [35.645(c)(2)(i)]?
		g.	Difference between measured and expected output [35.645(c)(2)(ii)]?
		h.	Source output compared against computer calculation of output [35.645(c)(2)(iii)]?
		i.	Timer accuracy and linearity over the range of use [35.645(c)(2)(iv)]?
		j.	On-off error [35.645(c)(2)(v)]?
		k.	Trunnion centricity [35.645(c)(2)(vi)]?
		l.	Interlock systems [35.645(d)(1)]?
		m.	Source exposure indicator lights [35.645(d)(2)]?
		n.	Viewing and intercom systems [35.645(d)(3)]?
		o.	Timer termination [35.645(d)(4)]?
		p.	Radiation monitors used to indicate room exposures [35.645(d)(5)]?
		q.	Hydraulic cutoff mechanism [35.645(d)(6)]?
G.	Licensee promptly repaired items found to be not operating properly and did not use unit until repaired, if required [35.642(e), 35.643(g), 35.644(d), 35.645(f)]?
H.	Records maintained [35.2642, 35.2643, 35.2645]?

Installation, Maintenance, and Repair of Therapy Devices

A.	Only authorized individuals perform maintenance, repair and inspection [35.605]? Name of organization/individual:
B.	Records of maintenance, inspection and service maintained [35.2605]?

Operating Procedures For Therapy Devices

A.	Instructions on location of procedures and emergency response telephone numbers are posted at the device console [35.610(c)]?
B.	Copy of the entire procedures physically located at the device console [35.610(b)]?
C.	Procedures include:
	1.	Securing the device, the console, and the console keys [35.610(a)(1)]?
	2.	Ensuring that only the patient is in the treatment room before starting a treatment (except for LDR) [35.610(a)(2)]?
	3.	Preventing dual operation of more than one radiation producing device in the treatment room [35.610(a)(3)]?
	4.	Responding to emergencies or abnormal situations [35.610(a)(4)]?
	5.	Radiation survey of patient is performed to ensure source is returned to shielded position [35.604(a)]?
	6.	Records of radiation surveys maintained for three years [35.2404]?
D.	Authorized medical physicist and authorized user:
	1.	Physically present during initiation of patient treatment with remote afterloaders [35.615(f)(1), (2), and (3)]?
	2.	Physically present throughout all patient treatments with a gamma stereotactic radiosurgery device [35.615(f)(4)]?

Personnel Radiation Protection

A.	Exposure evaluation performed [20.1501]?
B.	ALARA program implemented [20.1101(b)]?
C.	External Dosimetry
	1.	Monitors workers per [20.1502(a)]?
	2.	External exposures account for contributions from airborne activity [20.1203]?
	3.	Supplier _________ Frequency _________
	4.	Supplier is NVLAP-approved [20.1501(c)]?
	5.	Dosimeters exchanged at required frequency?
D.	Internal Dosimetry
	1.	Monitors workers per [20.1502]?
	2.	Briefly describe program for monitoring and controlling internal exposures [20.1701, 20.1702]?
	3.	Monitoring/controlling program implemented (includes bioassays) [35.315(a)]?
	4.	Respiratory protection equipment [20.1703]?
E.	Reports
	1.	Reviewed by ____________ Frequency ____________
	2.	Auditor reviewed personnel monitoring records for period __________ to __________
	3.	Prior dose determined for individuals likely to receive doses [20.2104]?
	4.	Maximum exposures TEDE ____________ Other ____________
	5.	Maximum CDEs ____________ Organs ____________
	6.	Maximum CEDE ____________
	7.	Internal and external summed [20.1202]?
	8.	TEDEs and TODEs within limits [20.1201]?
	9.	NRC forms or equivalent [20.2104(d), 2106(c)]?
		a.	NRC-4 Complete:
		b.	NRC-5 Complete:
	10.	Worker declared her pregnancy in writing during inspection period (review records)? If yes, in compliance with [20.1208] and records maintained [20.2106(e)]?
F.	Who performed any PSEs at this facility (number of people involved and doses received) [20.1206, 2104, 2105, 2204]?
G.	Records of exposures, surveys, monitoring, and evaluations maintained [20.2102, 2103, 2106]?

Confirmatory Measurements

Detail location and results of confirmatory measurements

Medical Events

A.	If medical events [criteria in 35.3045(a)] have occurred since the last audit, evaluate the incident(s) and procedures for implementing & administering written directives using the existing guidance.
	1.	Event date	_______________	Information Source	_______________
	2.	Notifications

		>NRC Ops Center Referring Physician In writing/By telephone	Region Patient

		>If notification did not occur, why not?

	>3.	Written Reports [35.3045]
		>a.	Submitted to Region within 15 days?
		b.	Copy to patient within 15 days?

>B.	Records maintained [35.2045]?

Notification and Reports

A.	In compliance with [19.13, 30.50] (reports to individuals, public and occupational, monitored to show compliance with Part 20)?
B.	In compliance with [20.2201, 30.50] (theft or loss)?
C.	In compliance with [20.2202, 30.50] (incidents)?
D.	In compliance with [20.2203, 30.50] (overexposures and high radiation levels)?
E.	Aware of NRC Ops Center phone number?
F.	In compliance with [20.2203] (Constraint on air emissions)?

Posting and Labeling

A.	NRC Form 3, "Notice to Workers" is posted [19.11]?
B.	Parts 19, 20, 21, Section 206 of Energy Reorganization Act, procedures adopted pursuant to Part 21, and license documents are posted, or a notice indicating where documents can be examined is posted [19.11, 21.6]?
C.	Other posting and labeling per [20.1902, 1904] and not exempted by [20.1903, 20.1905]?

Recordkeeping for Decommissioning

A.	Records of information important to the safe and effective decommissioning of the facility maintained in an independent and identifiable location until license termination [30.35(g)]?
B.	Records include all information outlined in 10 CFR 30.35(g)?

Bulletins and Information Notices

A.	Bulletins, Information Notices, NMSS Newsletters, etc., received?
B.	Appropriate action in response to Bulletins, Generic Letters, etc.?

Special License Conditions or Issues

A.	Special license conditions or issues to be reviewed:
B.	Evaluation:

Audits and Findings

A.	Summary of findings:
B.	Corrective and preventive actions:

Appendix L: Model Procedures for an Occupational Dose Program

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Dosimetry is required for individuals likely to receive in 1 year a dose in excess of 10% of the applicable regulatory limits in 10 CFR 20.1201. The Total Effective Dose Equivalent (TEDE) is the sum of the deep-dose equivalent (external exposure) and the committed effective dose equivalent (internal exposure). To demonstrate that dosimetry is not required, the licensee needs to have available, for inspection, an evaluation to demonstrate that the workers are not likely to exceed 10% of the applicable annual limits.

If an individual is likely to receive more than 10% of the annual dose limits, the NRC requires the licensee to monitor the dose, to maintain records of the dose, and, at least on an annual basis, to inform the worker of his/her dose.

The As Low As Reasonably Achievable "ALARA" Program

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10 CFR 20.1101 states that "each licensee must develop, document, and implement a radiation protection program commensurate with the scope and extent of licensed activities�" and "the licensee shall use, to the extent practicable, procedures and controls based upon sound standard protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable (ALARA)." Additionally, 10 CFR 20.1101 requires that licensees periodically review the content of the radiation protection program and its implementation.

External Dose Exposure

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The mechanism by which doses to individuals from exposure to radiation is evaluated is called dosimetry. Dosimetry allows the licensee to ensure that doses are maintained ALARA. Dosimetry also allows the licensee to show compliance with the occupational dose limits required by the NRC.

Providing for the safe use of radioactive materials and radiation is a management responsibility. It is important that management recognize the importance of radiation monitoring in the overall requirements for radiation protection.

There are three dose limits included in 10 CFR 20.1201 that apply to external exposure: deep dose to the whole body (5 rems or 0.05 Sv), shallow dose to the skin or extremities (50 rems or 0.5 Sv), and dose to the lens of the eye (15 rems or 0.15 Sv). According to the definitions in 10 CFR 20.1003, the deep-dose equivalent (DDE) to the whole body is considered to be at a tissue depth of 1 cm (1000 mg/cm2), shallow-dose equivalent to the skin or extremities at 0.007 cm (7 mg/cm2), and eye dose equivalent at 0.3 cm (300 mg/cm2). In evaluating the eye dose equivalent, it is acceptable to take credit for the shielding provided by protective lenses.

Monitoring an individual's external radiation exposure is required by 10 CFR 20.1502(a) if the external occupational dose is likely to exceed 10% of the dose limit appropriate for the individual (i.e., adult, minor, or declared pregnant woman). External radiation monitoring is also required by 10 CFR 20.1502(a)(3) for any individual entering a high or very high radiation area.

The use of individual monitoring devices for external dose is required for the following:

For adults who are likely to receive an annual dose in excess of any of the following (each evaluated separately):

0.5 rem (0.005 Sv) DDE
1.5 rems (0.015 Sv) eye dose equivalent
5 rems (0.05 Sv) shallow-dose equivalent to the skin
5 rems (0.05 Sv) shallow-dose equivalent to any extremity

For minors who are likely to receive an annual dose in excess of any of the following (each evaluated separately):

0.05 rem (0.5 mSv) DDE
0.15 rem (1.5 mSv) eye dose equivalent
0.5 rem (0.005 Sv) shallow-dose equivalent to the skin
0.5 rem (0.005 Sv) shallow-dose equivalent to any extremity.

For declared pregnant women who are likely to receive an annual dose from occupational exposure in excess of 0.05 rem (0.5 mSv) DDE, although the dose limit applies to the entire gestation period.
For individuals entering a high or a very high radiation area.

If the licensee determines that monitoring the occupational exposure of some workers is not necessary, he/she must demonstrate to the NRC that these workers will not exceed 10% of these limits using acceptable criteria. In these cases, the licensee need not provide individual monitoring devices to these workers.

The following are examples of criteria NRC accepts:

the licensee has previous dosimeter monitoring reports for workers in a specific work area that show that the workers are not likely to receive a dose in excess of 10% of the limits;
the licensee has performed appropriate radiation level surveys (using a survey meter or area thermoluminescent dosimeter (TLD)) of the work area, and has determined the number of hours a worker will be present in that work area, and has calculated the dose to workers (including "reasonable" accident scenarios) that shows that the workers are not likely to receive a dose in excess of 10% of the limits;
the licensee performs a reasonable calculation based upon source strength, distance, shielding, and time spent in the work area, that shows that workers are not likely to receive a dose in excess of 10% of the limits.

External dose is determined by using individual monitoring devices, such as film badges or thermoluminescent dosimeters (DDE). These devices must be evaluated by a processor that is National Voluntary Laboratory Accreditation Program (NVLAP) approved, as required by 10 CFR 20.1501. Acceptable exchange frequencies are every 3 months for TLDs and every month for film badges.

The device for monitoring the whole body dose shall be placed near the location expected to receive the highest dose during the year (10 CFR 20.1201( c)). When the whole body is exposed fairly uniformly, the individual monitoring device is typically worn on the front of the upper torso.

If the radiation dose is highly nonuniform, causing a specific part of the whole body (head, trunk, arms above the elbow, or legs above the knees) to receive a substantially higher dose than the rest of the whole body, the individual monitoring device shall be placed near that part of the whole body expected to receive the highest dose. For example, if the dose rate to the head is expected to be higher than the dose rate to the trunk of the body, a monitoring device shall be located on or close to the head.

If, after the exposure is received, the licensee somehow learns that the maximum dose to a part of the whole body was substantially higher than the dose measured by the individual monitoring device, an evaluation shall be conducted to estimate the actual maximum dose.

An acceptable alternative approach for highly nonuniform radiation fields is to use more than one dosimeter to separately track doses to different parts of the whole body. At the end of the year, each of the doses for each location would be summed. The deep-dose equivalent to be recorded would be that of the dosimeter location receiving the highest dose.

If the licensee determines that extremity monitoring is required, it may be appropriate to use an extremity dosimeter for some, but not all, radiation exposure. The licensee could supply an extremity dosimeter when exposure is nonuniform. When exposure is uniform, the shallow-dose equivalent measured by a torso dosimeter would be representative of the shallow-dose equivalent to the extremities, and separate extremity monitoring would not be needed. If protective gloves are used, it is acceptable to place the extremity dosimeter under the gloves.

10 CFR 20.2106 requires that the recording for individual monitoring be done on NRC Form 5 or equivalent. NRC Form 5 is used to record for the calendar year doses received. The monitoring year may be adjusted as necessary to permit a smooth transition from one monitoring year to another. As long as the year begins and ends within the month of January, the change is made at the beginning of the year, and no day is omitted or duplicated in consecutive years.

Because evaluation of dose is an important part of the radiation protection program, it is important that users return dosimeters on time. Licensees shall be vigorous in their effort to recover any missing dosimeters. Delays in processing a dosimeter can result in the loss of the stored information.

If an individual's dosimeter is lost, the licensee needs to perform and document an evaluation of the dose the individual received and add it to the employee's dose record. Sometimes the most reliable method for estimating an individual's dose is to use his/her recent dose history. In other cases, particularly if the individual does non-routine types of work, it may be better to use doses of co-workers as the basis for the dose estimate.

Investigational Levels - External Dose Monitoring

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The NRC has emphasized that the investigational levels in this program are not new dose limits but, as noted in ICRP Report 26, "Recommendations of the International Commission on Radiological Protection," Investigational levels serve as check points above which the results are considered sufficiently important to justify investigations.

When the exposure to a radiation worker exceeds Investigational Level I in Table L.1 of the occupational exposure in a quarter (Action Level I), an investigation should take place by the RSO, and a consideration of actions that might be taken to reduce the probability of recurrence will be reviewed by the RSO. When the exposure exceeds Investigational Level II in Table L.1 of the occupational exposure in a quarter (Action Level II), an investigation should take place by the RSO, with a consideration of actions to be taken to reduce the probability of occurrence, and a report of the actions should be reviewed by management.

Table L.1 Investigational Levels

Part of Body	Investigational Level I (mrems per calendar quarter)	Investigational Level II (mrems per calendar quarter)
whole body; head and trunk; active blood-forming organs; lens of eyes; or gonads	125	375
hands and forearms; feet and ankles	1875	5625
skin of whole body	750	2250

The RSO will review and record on Form NRC-5, "Current Occupational External Radiation Exposures," or an equivalent form (e.g., dosimeter processor's report) results of personnel monitoring. The following actions should be taken at the investigational levels as stated in Table L.1:

Personnel dose less than Investigational Level I.

Except when deemed appropriate by the RSO, no further action will be taken in those cases where an individual's dose is less than Table L.1 values for the Investigational Level I.
Personnel dose equal to or greater than Investigational Level I but less than Investigational Level II.

When the dose of an individual whose quarterly dose equals or exceeds Investigational Level I, a timely investigation should take place by the RSO. A consideration of actions that might be taken to reduce the probability of recurrence should be reviewed by the RSO following the quarter when the dose was recorded. If the dose does not equal or exceed Investigational Level II, no action related specifically to the exposure is required unless deemed appropriate by the RSO. The RSO should, however, review each such dose in comparison with those of others performing similar tasks as an index of ALARA program quality and should make a record of the review.
Personnel dose equal to or greater than Investigational Level II.

The RSO should investigate in a timely manner the causes of all personnel doses equaling or exceeding Investigational Level II. A consideration of actions should be taken by the RSO to reduce the probability of occurrence, and a report of the actions should be reviewed by the licensee's management at its first meeting following completion of the investigation.
Re-establishment of Investigational Level II to a level above that listed in Table L.1.

In cases where a worker's or a group of workers' doses need to exceed an Investigational Level, a new, higher Investigational Level may be established for that individual or group on the basis that it is consistent with good ALARA practices. Justification for new Investigational Levels should be documented.

Declared Pregnancy and Dose to Embryo/Fetus

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10 CFR 20.1208 states that the licensee shall ensure that the dose to an embryo/fetus during the entire pregnancy, due to occupational exposure of a declared pregnant woman, does not exceed 0.5 rem (5 mSv). The licensee shall make efforts to avoid substantial variation above a uniform monthly exposure rate to a declared pregnant woman. The pregnancy is declared in writing to include the worker's estimated date of conception. If the fetal exposures have exceeded 0.5 rem (5 mSv), or are within 0.05 rem (0.5 mSv) of this dose, exposures to the fetus will not exceed 50 mrem (0.5 mSv) during the remainder of the pregnancy. The dose to an embryo/fetus shall be taken as the sum of:

the deep-dose equivalent to the declared pregnant woman; and
the dose to the embryo/fetus from radionuclides in the embryo/fetus and radionuclides in the declared pregnant woman.

Methods for calculating the radiation dose to the embryo/fetus can be found in Regulatory Guide 8.36, "Radiation Dose To The Embryo/Fetus."

Internal Dose Exposure

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With respect to internal exposure, you are required to monitor your occupational intake of radioactive material and assess the resulting dose if it appears likely that you will receive greater than 10% of the ALI from intakes in 1 year. 10 CFR Part 20 provides terms for radionuclide intakes by means of inhalation and ingestion, i.e., DAC and ALI.

Exposure to airborne radioactivity at a level of 1 DAC for 1 year (2,000 hours) would result in either a committed effective dose equivalent of 5 rems (50 mSv) or a committed dose equivalent of 50 rems (0.5 Sv) to any individual organ or tissue, with no consideration for the contribution of external dose.

For each class of each radionuclide, there are two ALIs, one for ingestion and one for inhalation. The ALI is the quantity of radioactive material that, if taken into the body of an adult worker by the corresponding route would result in a committed effective dose of 5 rems (0.05 Sv) or a committed dose equivalent of 50 rems (0.5 Sv) to any individual organ or tissue, again, with no consideration for the contribution of external dose.

The effective dose equivalent concept described above makes it possible to combine both the internal and external doses in assessing the overall risk to the health of an individual. The 10 CFR Part 20 dose methodology evaluates the doses to all major body organs, multiplies these doses by the appropriate organ weighting factors, and then sums the organ-weighted doses to obtain a whole body risk-weighted "effective dose." The ALIs and DACs in 10 CFR Part 20, Appendix B, therefore, reflect the doses to all principal organs that are irradiated. The ALI means the derived limit for the amount of radioactive material taken into the body of an adult worker by inhalation or ingestion in a year. ALI is the smaller value of intake of a given radionuclide in a year by the reference man that would result in a committed effective dose equivalent (CEDE) of 5 rems or a committed dose equivalent (CDE) of 50 rems to any individual organ or tissue.

Assessment of intakes of radioactive materials by medical workers can best be determined by bioassay measurements. Bioassay services shall be available if the types and quantities of radioactive material licensed for use at your facility could, under normal operational occurrences, result in airborne levels in normally occupied areas exceeding DACs. Provisions shall be made for the collection of appropriate samples, analysis of bioassay samples, and evaluation of the results of these analyses to determine intakes.

If a worker receives a dose in excess of any of the annual dose limits, the regulations prohibit any occupational exposure during the remainder of the year in which the limit is exceeded. The licensee is also required to file an overexposure report with the NRC and provide a copy to the individual who received the dose as required by 10 CFR 20.2203 and 20.2205.

Internal Dose Monitoring (Bioassay Program)

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For each patient or human research subject receiving radiopharmaceutical therapy, the licensee should measure the thyroid burden of each individual who prepares or administers a dosage of iodine-131 NaI within 24 to 72 hours after administering the dosage. Determining the appropriate frequency of routine bioassay measurements depends upon the exposure potential and the physical and chemical characteristics of the radioactive material and the route of entry to the body. Elements that shall be considered include: 1) the potential exposure of the individual; 2) the retention and excretion characteristics of the radionuclide; 3) the sensitivity of the measurement technique; and 4) the acceptable uncertainty in the estimate of intake and committed dose equivalent. Bioassay measurements used for demonstrating compliance with the occupational dose limits shall be conducted often enough to identify and quantify potential exposures and resultant intakes that, during any year, are likely to collectively exceed 0.1 times the ALI.

Two separate categories of bioassay measurements further determine the frequency and scope of measurements: routine measurements and special measurements.

Routine Measurements include baseline measurements, periodic measurements, and termination measurements. These measurements shall be conducted to confirm that appropriate control exits and to assess dose.

Baseline measurements shall be conducted prior to initial work activities that involve exposure to radiation or radioactive materials, for which monitoring is required.

Periodic measurements shall be performed on a frequency determined on an a priori basis, considering the likely exposure of the individual. Periodic measurements shall be made when the cumulative exposure to airborne radioactivity, since the most recent bioassay measurement, is > 0.02 ALI (40 DAC hours). Noble gases and airborne particulates with a radioactive half-life of less than 2 hours shall be excluded from the evaluation, since external exposure is generally controlling for these radionuclides. As a minimum, periodic measurements shall be conducted annually.

Termination measurements shall be made when an individual is no longer subject to the bioassay program because of termination of employment or change in employment status. These measurements ensure that any unknown intakes are quantified.

Special Monitoring considers abnormal and inadvertent intakes from situations such as a failed respiratory protective device, inadequate engineering controls, inadvertent ingestion, contamination of a wound, or skin absorption, which shall be evaluated on a case-by-case basis.

Methods are presented for evaluating bioassay data that will result in calculated intakes acceptable to the NRC for evaluating compliance with the occupational dose limits of 10 CFR 20.1202. Examples of specific exposure situations and the physical and biochemical processes considered in the assessment of the exposures are in Appendix A of Regulatory Guide 10.9, "Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay Program."

Estimating Intakes - Evaluation and Investigation Levels

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Licensees shall estimate the intake for any bioassay measurement that indicates internally deposited radioactive material resulting from licensed activity. The scope of the evaluation shall be commensurate with the potential magnitude of the intake. For individual exposures with an estimate of intake less than 0.02 ALI, minimum bioassay measurements are adequate to provide a reasonable approximation of intake. Repeated follow-up measurements or additional exposure data reviews are not necessary, provided a reasonable estimate of the actual intake can be made based on available data.

Evaluation Level

The Evaluation Level is the level at which an intake shall be evaluated beyond the initial bioassay measurement. The Evaluation Level is 0.02 times the annual limit on intake (ALI), which is equivalent to 40 DAC hours. For very small intakes, a single bioassay measurement is adequate to estimate intake. For intakes that represent a significant contribution to dose, other available data shall be evaluated. If initial bioassay measurements indicate that an intake is greater than an Evaluation Level of 0.02 ALI, additional available data, such as airborne measurements or additional bioassay measurements, shall be used to obtain the best estimate of actual intake.

The primary radiopharmaceutical in medical bioassay programs is I-131 sodium iodide in either solution or capsule form. Greater volatility is associated with I-131 NaI in solution. The Evaluation Level (0.02 ALI) for I-131(inhalation) is (1 �Ci). The ALI for class D I-131 is 5E+1 �Ci (from Appendix B to 10 CFR 20.1001-20.2401); therefore, the Evaluation Level is:

0.02 x ALI (50 �Ci) = 1 �Ci (3.7E+4 Bq)

Investigational Level

The Investigational Level is the level at which an intake shall be investigated. The Investigation Level is any intake greater than or equal to 0.1 times the annual limit on intake (ALI). For single intakes that are greater than 10% of the ALI, a thorough investigation of the exposure shall be made; therefore, if a potential intake exceeds an investigation level of 0.1 ALI, multiple bioassay measurements and an evaluation of available workplace monitoring data shall be conducted. If practical, daily measurements shall be made until a pattern of bodily retention and excretion can be established. Such a determination is feasible after as few as three measurements; however, physiologically related variations and uncertainties require that measurements be continued over a longer period of time in some cases. For potential intakes near or exceeding the ALIs, the bioassay data evaluations shall consider any additional data on the physical and chemical characteristics and the exposed individual's physical and biokinetic processes.

The Investigational Level for I-131 is:

0.1(10%) x the ALI (50 �Ci) = 5 �Ci (1.9E+5 Bq)

Table L.2 Bioassay Action Levels for I-131

I-131 Intake Level	Activity Action Level for I-131
Evaluation Level	1 �Ci (2.22 x 10⁶ dpm) 0.02 ALI
Investigation Level	5 �Ci (1.11 x 10⁷ dpm) 0.1 ALI

Actions taken by RSO if radionuclide Intake level is exceeded:

Evaluation Level

The RSO should collect additional available data, such as airborne measurements or additional bioassay measurements to obtain the best estimate of actual intake. A consideration of actions that might be taken to reduce the probability of recurrence should be reviewed by the RSO in a timely manner following the time when the intake was recorded. If the intake does not equal or exceed the Investigational Level, no action related specifically to the intake is required unless deemed appropriate by the RSO. The RSO should, however, review each such intake in comparison with those of others performing similar tasks as an index of ALARA program quality and should make a record of the review.
Investigational Level

The RSO should collect multiple bioassay measurements and should evaluate available workplace monitoring data. If practical, daily measurements shall be made until a pattern of bodily retention and excretion can be established. Such a determination is feasible after as few as three measurements. The RSO should investigate in a timely manner the causes of all personnel intakes exceeding the Investigational Level. A consideration of actions should be taken by the RSO to reduce the probability of occurrence, and a report of the actions should be reviewed by the licensee's management at its first meeting following completion of the investigation.

If a worker receives an intake in excess of ALI, the regulations prohibit any occupational exposure during the remainder of the year in which the limit is exceeded. For I-131, the ALI is 50 �Ci to the thyroid gland. ALI is the smaller value of uptake of a given radionuclide in a year by the reference man that would result in a committed effective dose equivalent of 5 rems (0.05 Sv) to the whole body or a committed dose equivalent of 50 rems (0.5 Sv) to any individual organ or tissue. ALI values for intake by ingestion and by inhalation of selected radionuclides are given in Table 1, Columns 1 and 2, of Appendix B to 10 CFR 20.1001-20.2401.

Types of Measurements

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Characteristics such as mode of intake, uptake, and excretion and mode of radioactive decay shall be considered in selecting the most effective and reliable types of measurements. For example, in vivo lung or total body measurements shortly following exposure generally provide reliable estimates of intakes for most gamma-emitting radionuclides. In vitro measurements shall be used for radionuclides that emit little or no gamma radiation; however, in vitro urine or fecal measurements for the first voiding following exposure, while providing important information for assessing potential significance, do not generally represent equilibrium conditions and thereby shall not be relied upon in evaluating intakes. ICRP Publication 54 and NCRP Report No. 87 provide guidance acceptable to the NRC for determining the types of bioassay measurements that shall be made considering the physical and biological characteristics of the radioactive material.

Recordkeeping

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Records of measurement data, calculations of intakes, and methods for calculating dose must be maintained as required by 10 CFR 20.1204 (c), 20.2103 (b), and 20.2106 (a).

Interpretation of Bioassay Measurements

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The specific scope and depth of the evaluation of bioassay measurements, as discussed in the Estimating Intake - Evaluation and Investigational Levels section, depends on the potential significance of the intake. The methods presented below are acceptable to the NRC staff for correlating bioassay measurements to estimates of intakes for the purpose of demonstrating compliance with the occupational dose limits of 10 CFR 20.1201.

Time of Exposure

Accurate estimation of intake from bioassay measurements is dependent upon knowledge of time of intake. Generally, the time of intake is known considering work activities and other monitoring data, such as air sample data; therefore, the time of intake will be known for all but unusual situations. When the time of intake cannot be determined from monitoring data, it can often be determined from information provided by the individual. When information is insufficient to determine the time of intake, it is acceptable to assume that the intake occurred at the midpoint of the time period since the last bioassay measurement. This initial assumption should be refined by using any available information such as the individual's work schedule, facility operations data, historical air monitoring data, and effective half-life of the radionuclides detected.

Acceptable Biokinetic Models

Determining a worker's intake from bioassay measurements involves comparing the measured bodily retention or excretion to a tabulated value. The models and methods used for evaluating bioassay measurements should provide a reasonable assessment of the worker's exposure. For intakes that are a small fraction of the limit, greater inaccuracy in the estimate of intake can be accepted without significant impact on the overall assessment of a worker's dose; however, for annual exposures for which monitoring is required by 10 CFR 20.1502(b), these methods should not lead to significant underestimation or overestimation of the actual intake.

Variations from predicted retention and excretion for specific individuals can be expected. Excretion of radionuclides may be influenced by the worker's diet, health condition, age, level of physical and metabolic activity, or physiological characteristics. The lung deposition and clearance of the inhaled radionuclide, the particle size distribution, and the time of the excretion also influence the excretion rate of radionuclides.

Important considerations for evaluating bioassay measurements include:

Appropriate measurement technique (in vivo or in vitro) based on radionuclide decay characteristics (i.e., types of radiation emitted) and biokinetic characteristics (i.e., systemic uptake and retention and urine and fecal excretion fractions);
The effects of diuretics or chelation to reduce systemic uptake and to increase excretion or excretion rates;
Representativeness of measurements such as 24-hour or accumulated urine or fecal measurements;
The appropriate lung clearance class (D, W, or Y), if known (see definition of class in 10 CFR 20.1003), or if no information on the biological behavior or chemical form is available, the most restrictive clearance class relevant for the particular element should be assumed (i.e., that class that gives the lowest value of ALI);
Particle size distribution;
Chemical toxicity, as in the case of uranium (see 10 CFR 20.1201(e)).

The metabolic models in ICRP-30 and accompanying addenda and ICRP-54 present acceptable bases for estimating intake from bioassay measurements. Other acceptable models are the tritium model developed by Johnson and Dunford and the plutonium urinary excretion model developed by Jones.

The use of computer codes that apply these models is also acceptable for evaluating bioassay measurements, provided it can be demonstrated through documented testing that the models and methods employed provide results consistent with the acceptable models. There are several commercially available computer codes for interpreting bioassay measurements; these codes may be used as long as the software application is based on acceptable models and provides results that correctly implement the models. No specific computer codes are endorsed by the NRC staff. Licensees are responsible for ensuring that computer codes are appropriate for use in their particular circumstances.

Intake Retention and Excretion Fractions for Calculating Intakes

ICRP-54 presents urinary excretion and fecal excretion equations as a function of time following intake for a number of radionuclides. By differentiating these equations, intake retention functions can be derived. The solution of these equations over a range of times allows the development of tabulated intake retention and excretion fractions. The intake retention fractions⁶ (IRFs) contained in NUREG/CR-4884 were developed in this manner and represent an acceptable basis for correlating bioassay measurements to estimates of intake. To apply the use of IRFs for calculating an individual's radionuclide intake from a single bioassay measurement, divide the total activity in 24-hour urine, 24-hour feces, accumulated urine, or accumulated feces,⁷ or the radionuclide content in the total body, systemic organs, lungs, nasal passages, or GI tract, by the appropriate IRF value in NUREG/CR-4884.

⁶For purposes of this guide and the application of the data from NUREG/CR-4884, the parameter IRF denotes both intake retention fractions and intake excretion fractions.

⁷The term 24-hour urine" means the total urine output collected over a 24-hour period, and the term 24-hour feces" means the total feces output collected over a 24-hour period. Accumulated urine" and accumulated feces" mean the total output since time of exposure.

Equation 1 demonstrates this method:

Equation 1:

where:
	I	=	Estimate of intake with units the same as A(t)
	A(t)	=	Numerical value of the bioassay measurement obtained at time t (decay corrected to time of sampling for in vitro measurements) with appropriate units (�Ci, Bq, or �g)
	IRF(t)	=	Intake retention fraction corresponding to type of measurement for time t after estimated time of intake

Evaluating Spot Samples

If the total urine or feces is not collected for the 24-hour period, the following equations may be used to estimate the total activity excreted or eliminated over the 24-hour period based on less frequent sampling (spot samples).

Equation 2: A_i = C_i E(t_i - t_{i - 1})

Equation 3: A_i = A₁ + A₂ + ... A_i

where:
	A_i	=	Activity or amount of radioactive material in sample I
	i	=	The sequence number of the sample
	C_i	=	The radionuclide concentration in urine (activity/liter) or feces (activity/gram) of sample i, decay corrected to the time of sampling
	E	=	Daily excretion rate (use measured rates when available, or assume values of 1.4 liters/day for urine and 135 grams/day for feces for standard man or 1.0 liter/day for urine and 110 grams/day for feces for standard woman)
	t_i	=	The time (days) after intake that sample i is collected
	A_i	=	Total activity excreted or eliminated up to time

This method is applicable only if spot samples are collected with a frequency consistent with the significance of changes in the excretion rates. In general, spot samples should be collected frequently enough that there is no more than a 30% increase in the IRFs between bioassay measurements. For example, if the IRF for accumulated urine increases at a rate of 30% per day, spot samples should be collected daily. If the rate is 10% per day, collecting spot samples once every 3 days would be adequate. Also, the rapid clearance and excretion of inhaled particles from the N-P region of the lung makes it important that at least one spot sample be collected within the first 24 hours after exposure. Otherwise, the reliability of using accumulated samples and excretion fractions for calculating intakes should be examined; calculations based on spot samples correlated to 24-hour samples may provide better estimates.

For spot samples used to estimate an equivalent 24-hour sample, correcting for abnormal conditions of high or low fluid intake or excessive loss of fluids by perspiration may be warranted. NCRP-87 presents the following method based on a relationship between the specific gravity (sp. gr.) of the sample to the average specific gravity of urine (1.024 g/ml).

Equation 4:

An alternative to this method is a correction based on the expected creatine excretion rate of 1.7 grams/day for men and 1.0 grams/day for women. Refer to NCRP-87 for additional information.

Logarithmic interpolation should be used for interpolating retention and excretion fractions. For example, using the NUREG/CR-4884 data, an IRF value for 2.8 days post intake should be calculated by a logarithmic interpolation between the 2-day and the 3-day IRF values.

Evaluating Multiple Bioassay Measurements

When multiple bioassay measurements are made, a statistical evaluation of the data should be performed. Numerous statistical methods are available for evaluating multiple measurements, but the results will be no better than the reliability of the data set. Measurements that are suspect or known to be inaccurate should be excluded from the analysis. Additional measurements should be used for obtaining an appropriate data set. For the evaluation of multiple measurements, NUREG/CR-4884 recommends the use of unweighted, minimized chi-squared statistics, assuming all variances are the same (i.e., a least squares fit). This method is acceptable to the NRC staff; it is simple and straightforward for evaluating multiple bioassay measurements. The equation is as follows:

Equation 5:

Other statistical analyses of the data may provide a better fit of the data, considering the particulars of the measurements. For example, a minimized chi-squared fit weighted by the inverse of the variance may be used. Several methods are available for estimating the variance of measurements. One approach, applicable to radioactivity measurements, is to assume that the variance is proportional to the value of the measurement itself. Another is the assumption that the variance is proportional to the expected value.

In selecting the statistical method to be used for evaluating multiple measurements, consideration should be given to available information, particularly observed variability of the data and reliability of individual measurements. Other statistical methods are acceptable to the NRC staff, provided it can be demonstrated that the results provide reasonable estimates of intake.

Adjusting Intake Estimates for Multiple and Continuous Intakes

In practice, a worker may receive repeated exposures to the same radionuclide over a period of time. These intakes should be treated as separate acute intakes if measurements collected through the period allow for the individual quantification of each exposure. As a general rule, if intakes are separated in time so that the retained or eliminated fraction from an earlier intake is less than 10% of the retention or excretion fraction for the next intake, each intake may be evaluated separately without regard to any previous intakes.

Continual intakes that are distributed equally in size and time may be approximated using a relationship based on time integration of the IRF. The total intake is estimated by dividing the measured activity by the appropriate time integrated retention or excretion fraction. An example using the IRF values from NUREG/CR-4884 would be to perform a numerical integration over the individual IRF values covering the time period of interest. Any one of a number of standard integration techniques, including numerical and analytical solutions, can be used. For example, using the trapezoidal rule yields the following method:

For bioassay measurements taken during an exposure time interval, the equation is:

Equation 6:

Using the trapezoidal rule to solve Equation 6 yields the following approximation:

Equation 7:

For bioassay measurements taken after an exposure interval, the equation is:

Equation 8:

Likewise, Equation 8 may be approximated using the trapezoidal rule, which yields Equation 9:

Equation 9:

where:
	I	=	Total intake during period T
	A(t)	=	Amount of activity in compartment or whole body at time t following onset of intake
	T	=	Duration of intake (exposure time period)
	t	=	Time from onset of intake to time of measurement
	IRF(u)	=	Intake retention fraction at time u in compartment or whole body for a single intake of a radionuclide
	�	=	Variable time between integration limits
	n	=	number of increments

The number of increments to be used for a numerical integration should be selected to minimize unnecessary errors associated with the particulars of the IRF values over which the integration is being performed. In general, errors associated with the integration technique used should be limited to less than 10%.

Correcting Intake Estimates for Particle Size Differences

The models used for deriving intake retention and excretion fractions, such as those in NUREG/CR-4884, are typically based on 1-micrometer activity median aerodynamic diameter (AMAD) particles. It is acceptable to correct intake estimates for particles of different sizes. These corrections often help explain retention or excretion rates different from those expected, such as would occur for larger particles preferentially deposited in the upper region of the respiratory tract (N-P region) with more rapid clearance times. Guidance for determining AMADs is provided in Regulatory Guide 8.25, Air Sampling in the Workplace."

Equation 10, taken from Appendix B to NUREG/CR-4884, should be used for revising the total body IRFs in NUREG/CR-4884 to particle size distributions between 0.1 to 20 �m AMAD.

Equation 10:

where:
	IRF_AMAD	=	IRF for the activity median aerodynamic diameter (AMAD) of interest
	IRF_1 �m	=	Total body IRF for inhalation of 1 �m AMAD aerosols (these IRFs are given in Appendix B to NUREG/CR-4884)
	_T	=	Summation over all tissues (and organs) T
	N-P, T-B, P	=	The compartments or regions of deposition of the respiratory tract: the nasopharyngeal passage region (N-P), the tracheobronchial region (T-B), and the pulmonary region(P)
	f_N-P,T, f_T-B,T, f_P,T	=	The fraction of committed dose equivalent in the tissue T resulting from deposition in the N-P, T-B, and P regions, respectively. (Values for individual radionuclides are contained in the Supplements to Part 1 of ICRP-30)
	H_50T	=	Committed dose equivalent for tissue (or organ) T per unit intake
	W_T	=	Tissue (or organ) weighting factor, from 10 CFR 20.1003
	D_N-P, D_T-B, D_P	=	Regional deposition fractions for an aerosol entering the respiratory system. (Values presented in Table 1 below.)

Equation 10 may not provide valid corrections for time periods shortly following intakes. The time after intake for which Equation 10 begins to yield satisfactory results is less than 1 day for Class D compounds. For Class W compounds, this time is about 7 days following intake, and for Class Y compounds, it is about 9 days following intake.

Table L.3 Aerosol AMAD

	0.2 �m	0.5 �m	0.7 �m	1.0 �m
D_N-P	0.05	0.16	0.23	0.30
D_T-B	0.08	0.08	0.08	0.08
D_P	0.50	0.35	0.30	0.25
Total Deposition	0.63	0.59	0.61	0.63
	2.0 �m	5.0 �m	7.0 �m	10.0 �m
D_N-P	0.50	0.74	0.81	0.87
D_T-B	0.08	0.08	0.08	0.08
D_P	0.17	0.09	0.07	0.05
Total Deposition	0.75	0.91	0.96	1.00

Equation 10, for revising the IRF for different particle sizes, is applicable for the total body IRF. ICRP-54 provides graphs of IRF values for 0.1 �m, 1 �m, and 10 �m AMAD particles for other tissues and excreta. Intake retention and excretion functions are derived for other AMAD particles based on the acceptable biokinetic modeling, as discussed in Regulatory Positions 4.2 and 4.3 of Regulatory Guide 8.9.

It is acceptable to take into account particle size distribution and its effect on lung deposition and transfer in evaluating an individual's dose. ICRP-30 (with supplements) provides data and methods for use in evaluating the lung deposition and resultant doses for particle sizes between 0.1 and 20 �m AMAD. For particles with AMADs greater than 20 �m, complete deposition in the N-P region can be assumed.

It is acceptable to compare the estimate of intake for different particle sizes with the ALIs in Appendix B to ��20.1001-20.2401 for demonstrating compliance with intake limits. The ALIs are based on a particle size of 1 micrometer; however, modifying the ALI values for different particle size distributions requires prior NRC approval (10 CFR 20.1204(c)(2)).

Use of Individual Specific Biokinetic Modeling

Individual specific retention and excretion rates may be used in developing biokinetic models that differ from the reference man modeling (10 CFR 20.1204(c)). The quality and quantity of data used for this type of individual specific modeling should be sufficient to justify the revised model. Licensees should not attempt to develop individual specific retention and excretion fractions in the absence of actual biochemical and particle size information. Individual specific modeling is not required but may be developed; the modeling as presented above is acceptable for evaluating regulatory compliance.

Calculating Dose From Estimates of Intake

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Regulatory Guide 8.34, "Monitoring Criteria and Methods to Calculate Occupational Radiation Doses," contains additional guidance on determining doses based on calculated intakes once the intake is determined.

Recordkeeping

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Records of measurement data, calculations of intakes, and methods for calculating dose must be maintained as required by 10 CFR 20.1204(c), 20.2103(b), and 20.2106(a). For additional information on recordkeeping and reporting occupational exposure data, including intakes, refer to Revision 1 of Regulatory Guide 8.7, "Instructions for Recording and Reporting Occupational Radiation Exposure Data."

Implementation

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The purpose of this section is to provide information to applicants and licensees regarding the NRC staff's plans for using this appendix.

Except in those cases in which an applicant proposes an acceptable alternative method for complying with specified portions of the Commission's regulations, the methods described in this appendix will used by the NRC staff for evaluating compliance with 10 CFR 20.1001-20.2401.

Intake Retention Fraction Examples

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Examples illustrating the use of retention and excretion functions for calculating intakes based on bioassay measurements are presented in Appendix A of Regulatory Guide 8.9. The data used for these examples are taken from NUREG/CR-4884, "Interpretation of Bioassay Measurements." These examples do not illustrate the use of all possible bioassay or health physics measurements that are available (e.g., excreta and air sampling measurements) during an exposure incident. The examples demonstrate the use of retention and excretion factors to:

Estimate intake from one or several bioassay measurements;
Adjust intake estimates for multiple or continuous intakes; and
Correct intake estimates for particle size differences.

Appendix M: Guidance for Demonstrating that Individual Members of the Public Will Not Receive Doses Exceeding the Allowable Limits

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Licensees must ensure that:

The radiation dose received by individual members of the public does not exceed 1 mSv (100 mrem) in one calendar year resulting from the licensee's possession and/or use of licensed materials.

Members of the public include persons who live, work, or may be near locations where color="#008000"> licensed material is used or stored and employees whose assigned duties do not include the use of licensed materials and who work in the vicinity where it is used or stored.

The radiation dose in unrestricted areas does not exceed 0.02 mSv (2 mrem) in any one hour.

Typical unrestricted areas may include offices, shops, laboratories, areas outside buildings, property, and nonradioactive equipment storage areas. The licensee does not control access to these areas for purposes of controlling exposure to radiation or radioactive materials; however, the licensee may control access to these areas for other reasons, such as security.

Licensees must show compliance with both portions of the regulation. For areas adjacent to facilities where licensed material is used or stored, calculations or a combination of calculations and measurements (e.g., using an environmental TLD) are often used to show compliance.

Calculation Method ⁽⁵⁾

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The calculational method takes a tiered approach, going through a four-part process starting with a worst case situation and moving toward more realistic situations. It makes the following simplifications:

licensed material is a point source;
typical radiation levels encountered when the source is in the shielded position are taken from the manufacturer's literature; and
credit is taken for any shielding found between the licensed material and the unrestricted areas.

Part 1 of the calculational method is simple but conservative. It assumes that an affected member of the public is present 24 hours a day and uses only the inverse square law to determine if the distance between the licensed material and the affected member of the public is sufficient to show compliance with the public dose limits. Part 2 considers not only distance, but also the time that the affected member of the public is actually in the area under consideration. Part 3 considers the distance, the portion of time and dose rate during source exposure, the portion of time and dose rate while the source is in the shielded position, and the portion of time that the affected member of the public is present. Part 4 considers the approach in Part 3 plus any additional shielding between the licensed material and the unrestricted area. Using this approach, licensees make only those calculations that are needed to demonstrate compliance. The results of these calculations typically result in higher radiation levels than would exist at typical facilities, but provide a method for estimating conservative doses that could be received.

Example 1

To better understand the calculational method, we will examine Therapy Clinic, an HDR Afterloading Device licensee. Yesterday, the clinic's president noted that the new HDR facility is close to his secretary's desk and he asked Joe, the RSO, to determine if the clinic is complying with NRC's regulations.

The secretary's desk is near the wall separating the reception area from the designated, locked HDR room where the clinic has located its unit. Joe measures the distance from the HDR unit to the wall and assumes that the device would have the maximum dose rate when the source is exposed: 5000 mrem per hour at one meter. This is the maximum dose rate during treatment time. Figure M.1 is Joe's sketch of the areas in question, and Table M.1 summarizes the information Joe has on the HDR unit.

Figure M.1 Diagram of Office and HDR Facility. This sketch shows the areas described in Examples 1 and 2.

Table M.1 Information Known About the HDR Unit

Description of Known Information	Ir-192/HDR/10 Ci
Dose rate in mrem/hour encountered at specified distance from the HDR unit (from manufacturer)	5000 mrem/hour at 1 meter (3.28 ft)
Distance in ft to secretary's chair	15 ft

Example 1: Part 1

Joe's first thought is that the distance between the HDR unit and the secretary's chair may be sufficient to show compliance with the regulation in 10 CFR 20.1301. So, taking a worst case approach, he assumes: (1) the HDR unit is constantly present with a patient being treated; and (2) the secretary is constantly sitting in the desk chair (i.e., 24 hours/day). Joe proceeds to calculate the dose she might receive hourly and yearly from the HDR unit as shown in Table M.2 below.

Table M.2 Calculational Method, Part 1: Hourly and Annual Dose Received From the HDR Unit

Step No.	Description	Input Data	Results
1	Dose received in an hour at known distance from the HDR unit (e.g., from the manufacturer), in mrem/hour	5000	5000
2	Square of the distance (ft) at which the Step 1 rate was measured, in ft²	(3.28)²	10.8
3	Square of the distance (ft) from the HDR unit to the secretary's desk in an unrestricted area, in ft²	(15.0)²	225
4	Multiply the results of Step 1 by the results of Step 2 (this is an intermediate result)	5000 x 10.8	54000
5	Divide the result of Step 4 by the result of Step 3 to calculate the dose received by an individual at the secretary's desk, HOURLY DOSE RECEIVED FROM THE HDR UNIT, in mrem in an hour	54000/225	240
6	Multiply the result of Step 5 by 24 hours/day x 366 (leap year) days/year = MAXIMUM ANNUAL DOSE RECEIVED FROM THE HDR UNIT, in mrem in a year	240 x 24 x 366	2108000

Note: The result in Step 5 does not demonstrate compliance with the 2 mrem in any one hour limit. Also, if the result in Step 6 exceeds 100 mrem/year, proceed to Part 2 of the calculational method.

At this point, Joe notes that the total dose that an individual could receive in any one hour greatly exceeds 2 mrem in an hour (i.e., 240 mrem in an hour) and notes that an individual could receive a dose of 2108000 mrem in a year, much higher than the 100 mrem limit.

Example 1: Part 2

Joe reviews his assumptions and recognizes that the secretary is not at the desk 24 hours/day. He decides to make a realistic estimate of the number of hours the secretary sits in the chair at the desk, keeping his other assumptions constant (i.e., the HDR unit is constantly present and in use (i.e., 24 hours/day). He then recalculates the annual dose received.

Table M.3 Calculational Method, Part 2: Annual Dose Received From the HDR Unit

Step No.	Description	Results
7	A. Average number of hours per day that individual spends in area of concern (e.g., secretary sits at desk 5 hours/day; the remainder of the day the secretary is away from the desk area copying, filing, etc.)	5.0
	B. Average number of days per week in area (e.g., secretary is part time and works 3 days/week)	3.0
	C. Average number of weeks per year in area (e.g., secretary works all year )	52
8	Multiply the results of Step 7.A. by the results of Step 7.B. by the results of Step 7.C. = AVERAGE NUMBER OF HOURS IN AREA OF CONCERN PER YEAR	5.0 x 3.0 x 52 = 780
9	Multiply the results in Step 5 by the results of Step 8 = ANNUAL DOSE RECEIVED FROM THE HDR UNIT CONSIDERING REALISTIC ESTIMATE OF TIME SPENT IN AREA OF CONCERN, in mrem in a year	240 x 780 = 187000

Note: If Step 9 exceeds 100 mrem in a year, proceed to Part 3 of the calculational method.

Although Joe is pleased to note that the calculated annual dose received is significantly lower, he realizes it still greatly exceeds the 100 mrem in a year limit.

Example 1: Part 3

Again Joe reviews his assumptions and recognizes that the HDR unit is not constantly in use when the secretary is seated at the desk. As he examines the situation, he realizes he must take these factors into account.

Table M.4 Calculational Method, Part 3: Summary of Information

Step No.	Description	Input
10	Dose rate while the source is in the shielded position, in mrem per hour at 3.28 ft from the HDR (from manufacturer)	0.02
11	Dose rate while patient is being treated, in mrem per hour at 3.28 ft from the HDR	5000
12	Maximum number of patients treated per hour	2
13	Maximum treatment time, in minutes	1
14	From Table M.1, distance from HDR to secretary, in feet	15
15	From Step 8, average number of hours that secretary is in area of concern, per year	780

Table M.5 Calculational Method, Part 3: Annual Dose Received from HDR

Step No.	Description	Result
16	[60 minus the input from Step 12 multiplied by (the input from Step 13 )] divided by 60 = [60 - 2 x (1)] / 60 = [60 - 2] / 60 = FRACTION OF TIME THE SOURCE IS IN THE SHIELDED POSITION	0.97
17	1.0 minus the result from Step 16 = 1 - 0.97 = FRACTION OF TIME THE HDR UNIT IS USED	0.03
18	(The input from Step 10 multiplied by the result from Step 16) plus (the input from Step 11 multiplied by the result from Step 17) = (0.02 x 0.97) + (5000 x 0.03) = 0.02 + 150 = AVERAGE DOSE ENCOUNTERED AT 3.28 FEET FROM THE HDR UNIT, in mrem in an hour.	150
19	The result from Step 18 multiplied by (3.28 squared divided by the input from Step 14 squared) = 150 x (3.28² / 15²) = 150 x (10.8 / 225) = AVERAGE DOSE RATE ENCOUNTERED BY THE SECRETARY, in mrem per hour.	7.20
20	The result from Step 19 multiplied by the input from Step 15 = 780 x 7.20 = ANNUAL DOSE RECEIVED FROM HDR UNIT CONSIDERING REALISTIC ESTIMATES OF TIME SPENT IN AREA OF CONCERN, DOSE RATES, AND HDR UNIT USAGE, in mrem in a year.	5600

Note: If the result in Step 20 is greater than 100 mrem/yr, the licensee must take corrective actions. Corrective action may include shielding the HDR treatment room.

Although, Joe notes that the result in Step 20 is significantly lower, he realizes that the result still exceeds the 100 mrem in a year limit and that he must consider additional corrective actions. As he reviews the situation, he realizes that the walls of the treatment room consist of approximately 10.5 inches of concrete. He decides to take this into account.

Example 1: Part 4

Table M.6 Calculational Method, Part 4: Summary of Information

Step No.	Description	Input
21	Tenth Value Layer for Iridium 192 in concrete in inches	3.5
22	Thickness of concrete in wall	10.5
23	Annual dose received from HDR unit from Step No. 20, in mrem in a year.	5600

Table M.7 Calculational Method, Part 4: Annual Dose Received from HDR

Step No.	Description	Result
24	[Input from Step 22 divided by the input from Step 21] = [10.5/3.5] = NUMBER OF TENTH VALUE LAYERS	3.0
25	[Input from Step 23 divided by 10 raised to the result from Step 24] = 5600/10³ = ANNUAL DOSE RECEIVED FROM HDR UNIT CONSIDERING REALISTIC ESTIMATES OF TIME SPENT IN AREA OF CONCERN, DOSE RATES, HDR UNIT USAGE, AND TREATMENT ROOM SHIELDING, in mrem in a year	5.6

Joe is glad to see that the results in Step 25 show compliance with the 100 mrem in a year limit. Had the result in Step 25 been higher than 100 mrem in a year, then Joe could have done one or more of the following:

Consider whether the assumptions used to determine occupancy are accurate, revise the assumptions as needed, and recalculate using the new assumptions
Calculate the effect of any additional shielding⁽⁶⁾ located between the HDR unit and the secretarial workstation
Take corrective action (e.g., move the secretarial workstation) and perform new calculations to demonstrate compliance
Designate the area outside the use area as a restricted area and the secretary as an occupationally exposed individual. This would require controlling access to the area for purposes of radiation protection and training the secretary as required by 10 CFR 19.12

Note that in the example, Joe evaluated the unrestricted area outside only one wall of the HDR facility. Licensees also need to make similar evaluations for other unrestricted areas and to keep in mind the ALARA principle, taking reasonable steps to keep radiation dose received below regulatory requirements. In addition, licensees need to be alert to changes in situations (e.g., changing the secretary to a full-time worker, or changing the estimate of the portion of time spent at the desk) and to perform additional evaluations, as needed.

RECORDKEEPING: 10 CFR 20.2107 requires licensees to maintain records demonstrating compliance with the dose limits for individual members of the public.

Combination Measurement - Calculational Method

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This method, which allows the licensee to take credit for shielding between the HDR unit and the area in question, begins by measuring radiation levels in the areas, as opposed to using manufacturer-supplied rates at a specified distance from the unit. These measurements must be made with calibrated survey meters sufficiently sensitive to measure background levels of radiation; however, licensees must exercise caution when making these measurements, and they must use currently calibrated radiation survey instruments. A maximum dose of 1 mSv (100 mrem) received by an individual over a period of 2080 hours (i.e., a "work year" of 40 hours/week for 52 weeks/year) is equal to less than 0.5 microsievert (0.05 mrem) per hour.

This rate is well below the minimum sensitivity of most commonly available G-M survey instruments.

Instruments used to make measurements for calculations must be sufficiently sensitive. An instrument equipped with a scintillation-type detector (e.g., NaI(Tl)) or a micro-R meter used in making very low gamma radiation measurements should be adequate.

Licensees may also choose to use environmental TLDs⁽⁷⁾ in unrestricted areas next to the HDR unit for monitoring. This direct measurement method would provide a definitive measurement of actual radiation levels in unrestricted areas without any restrictive assumptions. Records of these measurements can then be evaluated to ensure that rates in unrestricted areas do not exceed the 1 mSv/year (100 mrem/year) limit.

Example 2

As in Example 1, Joe is the RSO for Therapy Clinic a HDR licensee. The clinic has one HDR unit in a designated, locked area that adjoins an unrestricted area where a secretarial work station is located. See Figure M-1 and Table M-2 for information. Joe wants to see if the clinic complies with the public dose limits at the secretarial station.

Joe placed an environmental TLD badge in the secretarial work space for 30 days. The TLD processor sent Joe a report indicating the TLD received 1 mSv (100 mrem).

Example 2: Part 1

Table M.8 Combination Measurement-Calculational Method

Step No.	Description	Input Data and Results
1	Dose received by TLD, in mrem	100
2	Total hours TLD exposed	24 hours/day x 30 days/month = 720
3	Divide the results of Step 1 by the results of Step 2 = HOURLY DOSE RECEIVED, in mrem in an hour	100/720 = 0.14
4	Multiply the results of Step 3 by 366 days/year [leap year] x 24 hours/day = 8760 hours in one year = MAXIMUM ANNUAL DOSE RECEIVED FROM THE HDR UNIT, in mrem in a year	366 x 24 x 0.14 = 8784 x 0.14 = 1230

Note: For the conditions described above, Step 3 indicates that the dose received in any one hour is less than the 2 mrem in any one hour limit. However, if there are any changes, then the licensee will need to reevaluate the potential doses that could be received in any one hour. Step 4 indicates that the annual dose received would be much greater than the 100 mrem in a year allowed by the regulations.

Example 2: Part 2

At this point Joe can adjust for a realistic estimate of the time the secretary spends in the area and the time the HDR unit was operating, as he did in Parts 2 and 3 of Example 1.

Example 2: Part 3

If the results of Joe's evaluation in Part 2 show that the annual dose received in a year exceeds 100 mrem, then he may have to consider moving the secretary's desk or adding additional shielding to the wall.

Appendix N: Emergency Procedures

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Model Spill Procedures - Low and High Dose Unsealed Sources

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Minor Spills of Liquids and Solids

Notify persons in the area that a spill has occurred.
Prevent the spread of contamination by covering the spill with absorbent paper.
Clean up the spill using disposable gloves and absorbent paper. Carefully fold the absorbent paper with the clean side out and place in a plastic bag for transfer to a radioactive waste container. Also put contaminated gloves and any other contaminated disposable material in the bag.
Survey the area with a low-range radiation detector survey meter. Check the area around the spill. Also check your hands, clothing, and shoes for contamination.
Report the incident to the RSO.

Major Spills of Liquids and Solids

Clear the area. Notify all persons not involved in the spill to vacate the room.
Prevent the spread of contamination by covering the spill with absorbent paper, but do not attempt to clean it up. To prevent the spread of contamination, limit the movement of all personnel who may be contaminated.
Shield the source if possible. Do this only if it can be done without further contamination or a significant increase in radiation exposure.
Close the room and lock or otherwise secure the area to prevent entry.
Notify the RSO immediately.
Decontaminate personnel by removing contaminated clothing and flushing contaminated skin with lukewarm water, then washing with mild soap. If contamination remains, induce perspiration by covering the area with plastic. Then wash the affected area again to remove any contamination that was released by the perspiration.

The decision to implement a major spill procedure instead of a minor spill procedure depends on many incident-specific variables, such as the number of individuals affected, other hazards present, likelihood of contamination spread, types of surfaces contaminated and radiotoxicity of the spilled material. For some spills of short-lived radionuclides, the best spill procedure may be restricted access pending complete decay.

Use Table N-1 as general guidance to determine whether a major spill procedure or a minor spill procedure should be implemented.

Estimate the amount of radioactivity spilled. Initiate a major or minor spill procedure, based on the following information. Spills above these mCi amounts are considered major, and below are considered minor.

Table N.1 Relative Hazards of Common Radionuclides

Radionuclide	Millicuries	Radionuclide	Millicuries
P-32	1	Tc-99m	100
Cr-51	100	In-111	10
Co-57	10	I-123	10
Co-58	10	I-125	1
Fe-59	1	I-131	1
Co-60	1	Sm-153	10
Ga-67	10	Yb-169	10
Se-75	1	Hg-197	10
Sr-85	10	Au-198	10
Sr-89	1	Tl-201	100

Spill Kit

You may also want to consider assembling a spill kit that contains:

6 pairs of disposable gloves, 1 pair housekeeping gloves
2 disposable lab coats
2 paper hats
2 pairs shoe covers
1 roll absorbent paper with plastic backing
6 plastic trash bags with twist ties
"Radioactive Material" labeling tape
1 china pencil or marking pen
3 pre-strung "Radioactive Material" labeling tags
Supplies for 10 contamination wipe samples
Instructions for "Emergency Procedures"
Clipboard with one copy of Radioactive Spill Report Form
Pencil

Emergency Surgery of Patients Who Have Received Therapeutic Amounts of Radionuclides

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If AUs or other personnel involved in the surgical procedure are likely to receive exposures exceeding the nonoccupational permissible dose limits specified in 10 CFR 20.1301, we will follow the procedures below:

If emergency surgery is performed within the first 24 hours following the administration of I-131 sodium iodide, fluids (e.g., blood, urine, etc.,) will be carefully removed and contained in a closed system.
The surgeon and the personnel involved in the surgical procedures will wear protective gear for the protection of the eyes from possible splashing of foreign materials, as well as from beta radiation.
The RSO will direct personnel in methods to keep doses ALARA during surgical procedures.
If an injury occurs during surgery that results in a cut or tear in the glove used, the individual involved will be monitored to determine if radioactive material was introduced into the wound. The RSO will be informed of any possible radiation hazard.

Autopsy of Patients Who Have Received Therapeutic Amounts of Radionuclides

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If AUs or other personnel involved in the autopsy are likely to receive exposures exceeding the nonoccupational permissible dose limits specified in 10 CFR 20.1301, we will follow the procedures below:

Upon the death of the therapy patient, the AU in charge and the RSO will be notified immediately.
An autopsy will be performed only after consultation and permission from the RSO.
Protective eye wear will be worn by the pathologist and his assistants for protection from possible splashing of foreign materials and exposure from beta radiation.
If an entire block of tissue containing the radionuclide can be removed during autopsy, this will be done first. The remainder of the autopsy can then proceed as usual.
The RSO will evaluate the radiation hazard(s), direct personnel in safety and protection, and suggest suitable procedures in order to keep doses ALARA during the autopsy.
When possible, separate organs will be promptly removed from the body, and detailed dissection will be carried out a safe distance away from the body.
After selected small samples have been removed, the radioactive tissues that are retained will promptly be either placed in appropriately shielded vessels for storage or disposed of according to procedures deemed appropriate by the RSO and in accordance with the regulations.
If an injury occurs during the autopsy which results in a cut or tear in the glove, the individual will be monitored to determine if radioactive material was introduced into the wound. The RSO will be informed on any possible radiation hazard.

Model Emergency Procedures for Teletherapy Units Containing Sealed Sources-Emergency Procedures for Beam Control Failure or Malfunction

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If the light signals or beam-on monitor indicates that the beam control mechanism has failed to terminate the exposure at the end of the pre-set time (e.g., if the red light stays on and the green light is off, or if both the red and the green lights stay on for more than a few seconds), the source may still be in the exposed position. The following steps are to be carried out promptly:

Open the door to the treatment room.
Tell an ambulatory patient to leave the room.
If the patient is not ambulatory, enter the treatment room but avoid exposure to the direct beam. Pull the treatment table as far away from the direct beam as possible. Transfer the patient to a stretcher and remove the patient from the room.
Close the door and secure the area by locking the door to the treatment room or posting a guard at the entrance.
Turn off the main switch at the control panel.
Notify the AU and RSO at once.
Conspicuously post a sign in the area to warn others of the problem.


>Authorized User:	________________________________________________________

>Phone No.:	On Duty:	______________	Off Duty:	______________

>Radiation Safety Officer:	________________________________________________________

>Phone No.:	On Duty:	______________	Off Duty:	______________

Appendix O: Model Procedures for Ordering and Receiving Packages

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Model Guidance

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We will, through a designee (e.g., RSO), authorize each order of radioactive materials and ensure that the requested materials and quantities are authorized by the license for use by the requesting AU and that possession limits are not exceeded.
We will establish and maintain a system for ordering and receiving radioactive material. The system will contain the following information:

Written records that identify the AU or department, radionuclide, physical and/or chemical form, activity, and supplier
Confirmation, through the above records, that material received was ordered through proper channels.

For deliveries during normal working hours, we will inform carriers to deliver radioactive packages directly to a specified area.
For deliveries during off-duty hours, we will inform security personnel or other designated persons to accept delivery of radioactive packages in accordance with procedures outlined in the sample memorandum for delivery of packages to the Nuclear Medicine Division, provided below. We will develop a similar memorandum for delivery of packages to other divisions.

Sample Memorandum

MEMO TO:   Chief of Security
FROM:   Radiation Safety Officer
SUBJECT:   Receipt of Packages Containing Radioactive Material

The security guard on duty shall accept delivery of radioactive material that arrives outside normal working hours. Packages should be placed on a cart or wheelchair and taken immediately to the Nuclear Medicine Division, Room ________. Unlock the door, place the package on top of the counter, and relock the door.

If the package appears to be damaged, immediately contact one of the individuals identified below. Ask the carrier to remain at the hospital until it can be determined that neither the driver nor the delivery vehicle is contaminated.

If you have any questions concerning this memorandum, please call our hospital Radiation Safety Officer, at extension ________.

	Name	Home Telephone
Radiation Safety Officer:
Director of Nuclear Medicine:
Nuclear Medicine Technologist Supervisor:
Nuclear Medicine Technologist on call (call page operator at extension _________ )
Nuclear Medicine Physician on call (call page operator at extension _________ )

Appendix P: Model Procedure for Safely Opening Packages Containing Radioactive Material

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Special requirements must be followed for packages containing quantities of radioactive material in excess of the Type A quantity limits specified in 49 CFR 173.433 or in Table A-1 of 10 CFR Part 71 (e.g., 13.5 curies of Mo-99, Cs-137, Ir-192; 54.1 curies of I-125; 541 curies of Xe-133, or 216 curies of Tc-99m). Such packages must be received expeditiously when the carrier offers it for delivery or when the carrier notifies the licensee that the package has arrived at the carrier's terminal. For these and other packages that are so required, monitoring for external radiation levels and surface contamination must be performed within 3 hours after receipt (if received during working hours) or no later than 3 hours from the beginning of the next working day (if received after working hours), in accordance with the requirements of 10 CFR 20.1906(c). The NRC Regional Office and the final delivery carrier must be notified if:

Removable radioactive surface contamination exceeds the limits of 10 CFR 71.87(i) [i.e. 22 dpm/centimeter squared (cm²) of beta and gamma emitting photons and 2.2 dpm/cm² of alpha]; and
External radiation levels exceed the limits of 10 CFR 71.47.

We will implement the following procedure for opening each package containing radioactive material received pursuant to our NRC license:

Put on gloves to prevent hand contamination.
Visually inspect the package for any sign of damage (e.g., wet or crushed). If damage is noted, stop the procedure and notify the RSO immediately.
Monitor the external surfaces of a labeled⁽⁸⁾ package for radioactive contamination, unless the package contains only radioactive material in the form of a gas or in special form, as defined in 10 CFR 71.4.
Monitor the external surfaces of a labeled⁶ package for radiation levels, unless the package contains quantities of radioactive material that are less than or equal to the Type A quantity, as defined in 10 CFR 71.4 and Table A to 10 CFR Part 71.
Monitor all packages known to contain radioactive material for radioactive contamination and radiation levels, if there is evidence of degradation of package integrity, such as packages that are crushed, wet, or damaged.
Remove the packing slip.
Open the outer package, following any instructions that may be provided by the supplier.
Open the inner package and verify that the contents agree with the packing slip.
Check the integrity of the final source container. Notify the RSO of any broken seals or vials, loss of liquid, condensation, or discoloration of the packing material.
If there is any reason to suspect contamination, wipe the external surface of the final source container and remove the wipe sample to a low-background area. Assay the wipe sample to determine if there is any removable radioactivity. An appropriate instrument with sufficient sensitivity will be used to assay the sample. For example, a NaI(T1) crystal and ratemeter, a liquid scintillation counter, or a proportional flow counter may be used for these assays. The detection efficiency will be determined to convert wipe sample counts per minute to disintegrations per minute. Note: a dose calibrator is not sufficiently sensitive for this measurement. Take precautions against the potential spread of contamination.
Check the user request to ensure that the material received is the material that was ordered.
Monitor the packing material and the empty packages for contamination with a radiation detection survey meter before discarding. If contaminated, treat this material as radioactive waste. If not contaminated, remove or obliterate the radiation labels before discarding in in-house trash.
Make a record of the receipt.

For packages received under the general license in Section 31.11, we will implement the following procedure for opening each package:

Visually inspect the package for any sign of damage (e.g., wet or crushed). If damage is noted, stop the procedure and notify the RSO immediately.
Check to ensure that the material received is the material that was ordered.

Appendix Q: Model Leak Test Program

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Facilities and Equipment

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To ensure achieving the required sensitivity of measurements, leak tests should be analyzed in a low-background area.
Consider using a NaI(Tl) well counter system with a single or multichannel analyzer to analyze samples obtained from gamma-emitting sources (e.g., Cs-137).
Consider using a liquid scintillation or gas-flow proportional counting system to analyze samples obtained from beta-emitting sources (e.g., Sr-90).
Instrumentation used to analyze leak test samples must be capable of detecting 0.005 microcurie of radioactivity.

Model Procedure for Performing Leak Testing and Analysis

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For each source to be tested, list identifying information such as sealed source serial number, radionuclide, and activity.
Use a separate wipe sample (e.g., cotton swab or filter paper) for each source.
Number each wipe to correlate identifying information for each source.
Wear gloves.
Obtain samples at the most accessible area where contamination would accumulate if the sealed source were leaking.
Measure the background count rate and record.
Check the instrument's counting efficiency, using either a standard source of the same radionuclide as the source being tested or one with similar energy characteristics. Accuracy of standards should be within + 5% of the stated value and traceable to a primary radiation standard, such as those maintained by NIST.
Calculate efficiency of the instrument.

For example:	[(cpm from std) - (cpm from bkg)] activity of std in microcuries	=	efficiency in cpm/microcurie


>where:
	cpm	=	counts per minute
	std	=	standard
	bkg	=	background

Analyze each wipe sample to determine net count rate.
For each sample, calculate the activity in microcuries and record.

For example:	[(cpm from wipe sample) - (cpm from bkg)] efficiency in cpm/microcurie	= microcuries on wipe sample

Leak test records will be retained in accordance with 10 CFR 35.2067 for three years. The records must contain:

The model number and serial number (if assigned) of each source tested,
The identity of each source radionuclide and its estimated activity,
The measured activity of each test sample expressed in microcuries,
A description of the method used to measure each test sample,
The date of the test, and
The name of the individual who performed the test.

If the wipe test activity is 0.005 microcurie or greater:

>	1.	Immediately withdraw the sealed source from use and either store the source, dispose of the source, or cause the source to be repaired, in accordance with the requirements in 10 CFR Parts 20 and 30.
	2.	File a report within five days of the leakage test with the appropriate NRC Regional Office listed in 10 CFR 30.6.

Appendix R: Model Procedure for Area Surveys

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Facilities and Equipment

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To ensure achieving the required sensitivity of measurements, survey samples should be analyzed in a low-background area.
Consider using a gamma counter system with a single or multi-channel analyzer to count samples containing gamma-emitters (e.g., Cs-137, Co-60).
Consider using either a liquid scintillation or a gas-flow proportional counting system to count samples containing alpha-emitters, beta-emitters, and gamma-emitters (if efficiency is great enough to achieve the required sensitivity for measurements).

Ambient Radiation Level Surveys

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We will implement the following procedure for ambient radiation level surveys:

Dose-rate surveys, at a minimum, will be performed in locations where:

workers are exposed to radiation levels that might result in radiation doses in excess of 10% of the occupational dose limits, or
an individual is working in an environment with a dose rate of 2.5 mrem/hour or more (5 rem/year divided by 2,000 hour/year).

10 CFR 20.1301 requires that the TEDE to an individual member of the public from the licensed operation does not exceed 1 mSv (0.1 rem) in a year, and that the dose in any unrestricted area from external sources does not exceed 0.02 mSv (0.002 rem) in any one hour. Appropriate surveys will be conducted to assure that the requirements of 10 CFR 20.1301 are met.
Radiation level surveys will consist of measurements with a survey meter sufficiently sensitive to detect 0.1 milliroentgen (mR) per hour. The following areas and frequencies will be followed:

Survey at the end of each day of use for all radiopharmaceutical elution, preparation, assay and administration areas (except patient rooms should be surveyed at the end of the therapy instead of on the day of administration) when using radiopharmaceuticals requiring a written directive (e.g., all therapy dosages and any iodine-131 dosage exceeding 30 �Ci).
Survey monthly all laboratory areas where only small quantities of gamma-emitting radioactive material are used (< 200 �Ci at a time).
Survey weekly all radionuclide use, storage, and waste storage areas. If diagnostic administrations are occasionally made in patients rooms (e.g., bone scan injections, Tc-99m heart agents, etc.) and special care is taken to remove all paraphernalia, those rooms need not be surveyed.
Survey quarterly all sealed source and brachytherapy source storage areas.

The RSO will be notified immediately of radiation levels that exceed trigger levels. Trigger levels for restricted and unrestricted areas are presented in Table R-1.

Table R.1 Ambient Dose Rate Trigger Levels

Type of Survey	Area Surveyed	Trigger Level
Ambient Dose Rate	Unrestricted	0.05 mR/hr
Ambient Dose Rate	Restricted	5.0 mR/hr

Contamination Surveys

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We will implement the following procedure for contamination surveys:

Contamination surveys should be sufficient to identify areas of contamination that might result in doses to workers or to the public. Combined removable and fixed contamination should be surveyed using appropriate radiation detection equipment. Removable contamination can be detected and measured through use of a wipe test of the surface, counted in an appropriate counting instrument, such as a liquid scintillation counter, a sodium iodide or germanium gamma counter, or a proportional alpha/beta counter.
Contamination surveys should be performed:

to evaluate radioactive contamination that could be present on surfaces of floors, walls, laboratory furniture, and equipment
after any spill or contamination event
when procedures or processes have changed
to evaluate contamination of users and the immediate work area, at the end of the day, when licensed material is used
in unrestricted areas at frequencies consistent with the types and quantities of materials in use but not less frequently than monthly
in areas adjacent to restricted areas and in all areas through which licensed materials are transferred and temporarily stored before shipment.

Personnel will survey for contamination in locations where individuals are working with an unsealed form of radioactive material, in an amount greater than or equal to 10% of the smallest annual limit on intake (ALI) (either the inhalation or ingestion ALI) listed for that radionuclide in 10 CFR Part 20. These surveys will be done at a frequency appropriate to the types and quantities of radioactive materials in use, but at least monthly. If amounts are used that are greater than or equal to the smallest ALI listed for that radionuclide in 10 CFR Part 20, then detailed documented surveys should be performed at least weekly.
The method for performing removable contamination surveys must be sufficiently sensitive to detect the most restrictive isotope used and listed in Table R.2 for restricted areas and R.3 for unrestricted areas (e.g., 200 dpm/100 cm² for isotopes of iodine-131 in unrestricted areas). Removable contamination survey samples should be measured in a low-background area. The following areas and frequencies will be followed:

Removable contamination surveys weekly for radiopharmaceutical elution, preparation, assay, and administration areas. If diagnostic administrations are occasionally made in patients rooms (i.e., bone scan injections, Tc-99m heart agents, etc.), with special care taken to remove all paraphernalia, those rooms need not be surveyed.
Removable contamination surveys monthly of laboratory areas where only small quantities of photon-emitting radioactive material are used (< 200 microcuries at a time).
Removable contamination surveys weekly for radionuclide storage and radionuclide waste storage areas.

A radioactive source with a known amount of activity will be used to convert sample measurements (usually in cpm) to dpm.
The area will be either decontaminated, shielded, or posted and restricted from use if unable to decontaminate.
The RSO will be immediately notified of contamination levels in excess of the trigger levels. Trigger levels for restricted areas are presented in Table R-2. Contamination found in unrestricted areas and on personal clothing will be immediately decontaminated to background levels. When it is not possible to get to background levels, we will ensure that the amounts do not exceed the contamination levels listed in Table R-3.

Table R.2 Acceptable Surface Contamination Levels in Restricted Areas in dpm/100 cm²

Area, clothing, skin if indicated	P-32, Co-58, Fe-59, Co-60, Se-75, Sr-85, In-111, I-123, I-125, I-131, Yb-169, Au-198	Cr-51, Co-57, Ga-67, Tc-99m, Hg-197, Tl-201
Restricted areas, protective clothing used only in restricted areas, skin	2000	20000

Table R.3 Acceptable Surface Contamination Levels in Unrestricted Areas in dpm/100 cm²

Nuclide¹	Average^{2, 3, 6}	Maximum^{2, 4, 6}	Removable^{2, 5, 6}
I-125, I-129, Transuranics	100	300	20
I-126, I-131, I-133, Sr-90	1,000	3,000	200
Beta-gamma emitters (nuclides with decay modes other than alpha emission or spontaneous fission) except Sr-90 and others noted above.	5,000	15,000	1,000
¹ Where surface contamination by both alpha- and beta-gamma-emitting nuclides exists, the limits established for alpha- and beta-gamma-emitting nuclides should apply independently. ² As used in this table, dpm means the rate of emission by radioactive material, as determined by correcting the counts per minute observed by an appropriate detector for background, efficiency, and geometric factors associated with the instrumentation. ³ Measurements of average contaminant should not be averaged over more than 1 square meter. For objects of less surface area, the average should be derived for each such object. ⁴ The maximum contamination level applies to an area of not more than 100 cm². ⁵ The amount of removable radioactive material per 100 cm² of surface area should be determined by wiping that area with filter or soft absorbent paper, applying moderate pressure, and assessing the amount of radioactive material on the wipe with an appropriate instrument of known efficiency. When removable contamination on objects of less surface area is determined, the pertinent levels should be reduced proportionally and the entire surface should be wiped. ⁶ The average and maximum radiation levels associated with surface contamination resulting from beta-gamma emitters should not exceed 0.2 millirad/hour at 1 cm and 1.0 millirad/hour at 1 centimeter, respectively, measured through not more than 7 milligrams per square centimeter of total absorber.

When equipment or facilitates that are potentially contaminated are to be released to unrestricted areas, the above table provides the maximum acceptable residual levels. To the extent practicable and consistent with the ALARA principle, it is appropriate to decontaminate to below these levels. Surface contamination surveys will be conducted for both removable and fixed contamination before these facilities or equipment are released from restricted to unrestricted use to ensure that they meet these limits.
A standardized method for smear testing of a relatively uniform area should be used to aid in comparing contamination at different times and places. A smear taken from an area of about 100 cm² is acceptable to indicate levels of removable contamination.

Alternate Survey Frequency

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Classification of Laboratories:

Group	Low	Medium	High
Survey Frequency Category
1	< 0.1 mCi	0.1 mCi to 1 mCi	> 1 mCi
2	< 1 mCi	1 mCi to 10 mCi	> 10 mCi
3	< 100 mCi	100 mCi to 1 Ci	> 1 Ci
4	< 10 Ci	10 Ci to 100 Ci	> 100 Ci

Proportional fractions are to be used for more than one isotope.

Modifying Factors	Factors
Simple storage	x 100
Very simple wet operations (e.g., preparation of aliquots of stock solutions)	x 10
Normal chemical operations (e.g., analysis, simple chemical preparations)	x 1
Complex wet operations (e.g., multiple operations, or operations with complex glass apparatus)	x 0.1
Simple dry operations (e.g., manipulation of powders) and work with volatile radioactive compounds	x 0.1
Exposure of non-occupational persons (including patients)	x 0.1
Dry and dusty operations (e.g., grinding)	x 0.01

The object is to determine how often to survey the laboratory. To do this, multiply the activity range under LOW, MEDIUM, and HIGH survey frequency by the appropriate Modifying Factor to construct a new set of mCi ranges for LOW, MEDIUM, and HIGH survey frequency.

Survey Frequency:

Low -- Not less than once a month
Medium -- Not less than once per week
High -- Not less than once per normal working day.

>Group 1	Pb-210 Po-210 Ra-223 Ra-226 Ra-228 Ac-227 Th-227 Th-228 Th-230 Pa-231 U-230 U-232 U-233 U-234 Np-237 Pu-238 Pu-239 Pu-240 Pu-241 Pu-242 Am-241 Am-243 Cm-242 Cm-243 Cm-244 Cm-245 Cm-246 Cf-249 Cf-250 Cf-252
Group 2	Na-22 Cl-36 Ca-45 Sc-46 Mn-54 Co-56 Co-60 Sr-89 Sr-90 Y-91 Zr-95 Ru-106 Ag-110m Cd-115m In-114m Sb-124 Sb-125 Te-127m Te-129m I-124 I-126 I-131 I-133 Cs-134 Cs-137 Ba-140 Ce-144 Eu-152 (13 y) Eu-154 Tb-160 Tm-170 Hf-181 Ta-182 Ir-192 T1-204 Bi-207 Bi-210 At-211 Pb-212 Ra-224 Ac-228 Pa-230 Th-234 U-236 Bk-249
Group 3	Be-7 C-14 F-18 Na-24 C1-38 Si-31 P-32 S-35 A-41 K-42 K-43 Ca-47 Sc-47 Sc-48 V-48 Cr-51 Mn-52 Mn-56 Fe-52 Fe-55 Fe-59 Co-57 Co-58 Ni-63 Ni-65 Cu-64 Zn-65 Zn-69m Ga-72 As-73 As-74 As-76 As-77 Se-75 Br-82 Kr-85m Kr-87 Rb-86 Sr-85 Sr-91 Y-90 Y-92 Y-93 Zr-97 Nb-93m Nb-95 Mo-99 Tc-96 Tc-97m Tc-97 Tc-99 Ru-97 Ru-103 Ru-105 Rh-105 Pd-103 Pd-109 Ag-105 Ag-lll Cd-109 Cd-115 In-115m Sn-113 Sn-125 Sb-122 Te-125m Te-127 Te-129 Te-131m Te-132 I-130 I-132 I-134 I-135 Xe-135 Cs-131 Cs-136 Ba-131 La-140 Ce-141 Ce-143 Pr-142 Pr-143 Nd-147 Nd-149 Pm-147 Pm-149 Sm-151 Sm-153 Eu-152 Eu-155 Gd-153 Gd-159 Dy-165 Dy-166 Ho-166 Er-169 Er-171 (9.2 hr) Tm-171 Yb-175 Lu-177 W-181 W-185 W-187 Re-183 Re-186 Re-188 0s-185 Os-l91 0s-193 Ir-190 Ir-194 Pt-l91 Pt-193 Pt-197 Au-196 Au-198 Au-l99 Hg-197 Hg-197m Hg-203 Tl-200 Tl-201 Tl-202 Pb-203 Bi-206 Bi-212 Rn-220 Rn-222 Th-231 Pa-233 Np-239
Group 4	H-3 0-15 A-37 Co-58m Ni-59 Zn-69 Ge-71 Kr-85 Sr-85m Rb-87 Y-9lm Zr-93 Nb-97 Tc-96m Tc-99m Rh-103m In-113m I-129 Xe-131m Xe-133 Cs-134m Cs-135 Sm-147 Re-187 Os-19lm Pt-193m Pt-197m Th-232 Th-Nat U-235 U-238 U-Nat.

Survey Record Requirements

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Each survey report should include the following:

A diagram of the area surveyed
A list of items and equipment surveyed
Specific locations on the survey diagram where wipe tests were taken
Ambient radiation levels with appropriate units
Contamination levels with appropriate units
Make and model number of instruments used
Background levels
Name of the person making the evaluation and recording the results and date.

Licensees should record contamination levels observed and procedures followed for incidents involving contamination of individuals. The record should include names of individuals involved, description of work activities, calculated dose, probable causes (including root causes), steps taken to reduce future incidents of contamination, times and dates, and the surveyor's signature.

Appendix S: Procedures for Developing, Maintaining, and Implementing Written Directives

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Written Directive Procedures

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This model provides guidance to licensees and applicants for developing, maintaining, and implementing procedures for administrations that require written directives (WD). This model does not limit you from using other guidance in developing procedures for administrations requiring a WD, e.g., information available from the Joint Commission on Accreditation of Healthcare Organizations or the American College of Radiology.

Discussion

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The administration of radioactive materials can be a complex process for many types of diagnostic and therapeutic procedures in nuclear medicine or radiation oncology departments. A number of individuals may be involved in the delivery process. For example, in an oncology department when the AU prescribes a teletherapy treatment, the delivery process may involve a team of medical professionals such as an AMP, a dosimetrist, and a radiation therapist. Conducting the treatment plan may involve a number of measurements, calculations, computer-generated treatment plans, patient simulations, portal film verifications, and beam-modifying devices to deliver the prescribed dose. Therefore, instructions must be clearly communicated to the professional team members with constant attention devoted to detail during the treatment process. Complicated processes of this nature require good planning and clear, understandable procedures. In addition, 10 CFR 35.27(c) requires that licensees establish a policy for all supervised individuals to request clarification, as needed, about instructions, including procedures requiring a WD.

The administration of radioactive materials can involve a number of treatment modalities, e.g., radiopharmaceutical therapy, teletherapy, brachytherapy, gamma stereotactic radiosurgery, and future emerging technologies. For each modality, the licensee shall develop procedures for WDs to meet the objectives of 10 CFR 35.40 and 35.41 (as applicable to the type of medical use), outlined below:

Have an authorized user prepare, date, and sign a written directive prior to the administration,
Verify the patient's or human research subject's identity prior to each administration,
Verify that the specific details of the administration are in accordance with the written directive and the treatment plan,
Check both manual and computer-generated dose calculations,
Verify that any computer-generated dose calculations are correctly transferred into the consoles of therapeutic medical devices, and
Record the radiopharmaceutical dosage or radiation dose actually administered.

The WD must be prepared for any administration of I-131 sodium iodide greater than 1.11 MBq (30 �Ci), any therapeutic dosage of a radiopharmaceutical, and any therapeutic dose of radiation from byproduct material. The WD must contain the information described in 10 CFR 35.40. The licensee shall retain the WD in accordance with 10 CFR 35.2040.

Suggested Policies and Procedures for Any Therapeutic Dose or Dosage of a Radionuclide or Any Dosage of Quantities Greater than 30 Microcuries of Sodium Iodide I-131

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We will establish the following policies and procedures:

A.	An AU must prepare, date, and sign a WD prior to the administration of any dose or dosage. This is required by 10 CFR 35.40.
B.	Prior to administering a dose or dosage, the patient's or human research subject's identity will be verified as the individual named in the WD. Examples of patient identity verification include the patient's ID bracelet, hospital ID card, driver's license or social security card.
C.	Before administering the dose or dosage, the specific details of the administration will be verified in accordance with the WD or treatment plan. All components of the WD (radionuclide, total dose or dosage, etc.) will be confirmed by the person administering the dose or dosage to verify agreement with the WD. Appropriate verification methods include: measuring the activity in the dose calibrator, checking the serial number of the sealed sources behind an appropriate shield, using color-coded vials or sealed sources, or using clearly marked storage locations. The verification will be performed by at least one qualified person (e.g., an oncology physician, AMP, nuclear medicine technologist, or radiation therapist) preferably other than the individual who prepared the dose or dosage or the treatment plan.
D.	All workers will be instructed to seek guidance if they do not understand how to carry out the WD. Specifically, workers should ask if they have any questions about what to do or how it should be done, prior to administration, rather than continuing a procedure when there is any doubt.

Additional Suggested Policies and Procedures for Sealed Therapeutic Sources and Devices Containing Sealed Therapeutic Sources

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We will establish the additional following policies and procedures:

A.	To ensure that the dose is delivered in accordance with the WD, the AU (and the neurosurgeon for GSR therapy) must approve a treatment plan that provides sufficient information and direction to meet the objectives of the WD. Suggested guidelines for information to be included in the treatment plan may be obtained from the American College of Radiology.

B.	For sealed sources inserted into the patient's body, radiographs or other comparable images (e.g., computerized tomography) will be used as the basis for verifying the position of the nonradioactive dummy sources and calculating the administered dose before administration; however, some brachytherapy procedures may require the use of various fixed geometry applicators (e.g., appliances or templates) to establish the location of the temporary sources and calculate the exposure time (or, equivalently, the total dose) required to administer the prescribed brachytherapy treatment. In these cases, radiographs or other comparable images may not be necessary, provided the position of the sources is known prior to insertion of the radioactive sources and calculation of the exposure time (or, equivalently, the total dose).

C.	Dose calculations will be checked before administering the prescribed therapy dose. An AU or a qualified person under the supervision of an AU (e.g., an AMP, oncology physician, dosimetrist, or radiation therapist), preferably one who did not make the original calculations, will check the dose calculations. The responsibilities and conditions of supervision are contained in 10 CFR 35.27. Suggested methods for checking the calculations include the following:

	1.	Computer-generated dose calculations will be checked by examining the computer printout to verify that correct input data for the patient were used in the calculations (e.g., source strength and positions).
	2.	The computer-generated dose calculations for input into the therapy console will be checked to verify correct transfer of data from the computer (e.g., channel numbers, source positions, and treatment times).
	3.	Manual dose calculations will be checked for:

		a.	Arithmetic errors,
		b.	Appropriate transfer of data from the WD, treatment plan, tables and graphs,
		c.	Appropriate use of nomograms (when applicable), and
		d.	Appropriate use of all pertinent data in the calculations

	If possible, the therapy dose will be manually calculated to a single key point and the results compared to the computer-generated dose calculations. If the manual dose calculations are performed using computer-generated outputs (or vice versa), particular emphasis should be placed on verifying the correct output from one type of calculation (e.g., computer) to be used as an input in another type of calculation (e.g., manual). Parameters such as the transmission factors for wedges and applicators and the source strength of the sealed source used in the dose calculations should be checked.

D.	After insertion of permanent implant brachytherapy sources, we will have an AU promptly record the actual number of radioactive sources implanted and sign or initial the patient's chart or other appropriate record.

E.	Acceptance testing will be performed by a qualified person (e.g., an AMP) on each treatment planning or dose calculating computer program that could be used for dose calculations. Acceptance testing shall be performed before the first use of a treatment planning or dose calculating computer program for therapy dose calculations. Each treatment planning or dose calculating computer program will be assessed based on our specific needs and applications. Acceptance testing will also be considered after each source replacement or when spot-check measurements indicated that the source output differed by more than 5% from the output obtained at the last full calibration corrected mathematically for radioactive decay.

F.	Independent checks on full calibration measurements will be performed when possible. The independent check will include an output measurement for a single specified set of exposure conditions and will be performed within 30 days following the full calibration measurements. The independent check will be performed by either:

	1.	An individual who did not perform the full calibration (the individual will meet the requirements specified in 10 CFR 35.51 using a dosimetry system other than the one that was used during the full calibration (the dosimetry system will meet the requirements specified in 10 CFR 35.630), or
	2.	An AMP (or an oncology physician, dosimetrist, or radiation therapist who has been properly instructed) using a thermoluminescence dosimetry service available by mail that is designed for confirming therapy doses and that is accurate within 5%.

G.	Full calibration measurements will include the determination of transmission factors for trays, wedges, applicators, etc. Transmission factors for other beam-modifying devices (e.g., nonrecastable blocks, recastable block material, bolus and compensator materials, and split-beam blocking devices) will be determined before the first medical use of the beam-modifying device and after replacement of the source.

H.	For GSR, particular emphasis will be directed toward verifying that the stereoscopic frame coordinates on the patient's skull match those of the treatment plan.

I.	A physical measurement of the teletherapy output will be made under applicable conditions prior to administration of the first teletherapy fractional dose, if the patient's treatment plan includes (1) field sizes or treatment distances that fall outside the range of those measured in the most recent full calibration or (2) transmission factors for beam-modifying devices (except nonrecastable and recastable blocks, bolus and compensator materials, and split-beam blocking devices) not measured in the most recent full calibration measurement.

J.	If possible, a weekly chart check will be performed by a qualified person under the supervision of an AU (e.g., an AMP, dosimetrist, oncology physician, or radiation therapist) to detect mistakes (e.g., arithmetic errors, miscalculations, or incorrect transfer of data) that may have occurred in the daily and cumulative dose administrations from all treatment fields or in connection with any changes in the WD or treatment plan.

Review of Administrations Requiring a Written Directive

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The licensee should consider establishing procedures to conduct periodic reviews of each applicable program area, e.g., radiopharmaceutical therapy, high-dose-rate brachytherapy, implant brachytherapy, teletherapy, gamma stereotactic radiosurgery and emerging technologies. The number of patient cases to be sampled should be based on the principles of statistical acceptance sampling and should represent each treatment modality performed in the institution, e.g., radiopharmaceutical, teletherapy, brachytherapy and gamma stereotactic radiosurgery.

For example, using the acceptance sampling tables of 10 CFR 32.110 and assuming an error rate (or lot tolerance percent defective) of 2%:

{ } Lot Size	{ } Sample Size	{ } Acceptance No.
1 to 75	All	0
76 to 100	70	0
101 to 200	85	0
201 to 300	95	0
301 to 400	100	0
401 to 600	105	0
601 to 800	110	0
801 to 4000	115	0

In order to eliminate any bias in the sample, the patient cases to be reviewed should be selected randomly. If the number of errors in the sample does not exceed the acceptance number in the appropriate Sampling Table, the lot should be accepted. For each patient's case, a comparison should be made between what was administered versus what was prescribed in the WD. If the difference between what was administered and what was prescribed exceeds the criteria for a medical event, that comparison is unacceptable. The number of unacceptable comparisons allowed for each sample size and lot tolerance percent defective is provided in the acceptance sampling tables of 10 CFR 32.110.

If feasible, the persons conducting the review should not review their own work. If this is not possible, two people should work together as a team to conduct the review of that work. The licensee or designee should regularly review the findings of the periodic reviews to ensure that the procedures for administrations requiring a WD are effective.

For each patient case reviewed, the licensee shall determine whether the administered radiopharmaceutical dosage or radiation dose was in accordance with the WD or treatment plan, as applicable. For each patient case reviewed, the licensee should identify deviations from the WD, the cause of each deviation, and the action required to prevent recurrence.

Records of Medical Events

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You shall maintain a record of medical events for three years as required by 10 CFR 35.2045. The licensee shall notify by telephone the NRC Operations Center^⁽⁹⁾ no later than the next calendar day after discovery of the medical event and shall submit a written report to the appropriate NRC Regional Office listed in 10 CFR 30.6 within 15 days after the discovery of the medical event, as required by 10 CFR 35.3045.

Appendix T: Model Procedures for Safe Use of Licensed Material

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Model Procedures for Safe Use of Radionuclides

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Wear laboratory coats or other protective clothing at all times in areas where radioactive materials are used.
Wear disposable gloves at all times while handling radioactive materials.
Either after each procedure or before leaving the area, monitor your hands for contamination in a low-background area using an appropriate survey instrument.
Use syringe shields for reconstitution of radiopharmaceutical kits and administration of radiopharmaceuticals to patients, except in those circumstances in which their use is contraindicated (e.g., recessed veins, infants). In these exceptional cases, consider the use of other protective methods such as remote delivery of the dose (e.g., through use of a butterfly needle.)
Do not eat, store food, drink, smoke, or apply cosmetics in any area where licensed material is stored or used.
Wear personnel monitoring devices, if required, at all times while in areas where radioactive materials are used or stored. These devices shall be worn as prescribed by the RSO. When not being worn to monitor occupational exposures, personnel monitoring devices shall be stored in the work place in a designated low-background area.
Wear a finger exposure monitor (TLD), if required, during the elution of generators; during the preparation, assay, and injection of radiopharmaceuticals; when performing camera quality control; when holding patients during procedures; and when handling radioactive material.
Dispose of radioactive waste only in designated, labeled, and properly shielded receptacles.
Never pipette by mouth.
Consider wipe testing unsealed byproduct material storage, preparation, and administration areas weekly for contamination. If necessary, decontaminate or secure the area for decay.
Consider surveying with a radiation detection survey meter, all areas of licensed material use, including the generator storage, kit preparation, and injection areas daily for contamination. If necessary, decontaminate or secure the area for decay as appropriate. Areas used to prepare and administer therapy quantities of radiopharmaceuticals must be surveyed daily in accordance with 10 CFR 35.70 (except when administering therapy dosages in patients' rooms).
Store radioactive solutions in shielded containers that are clearly labeled.
Radiopharmaceutical multidose diagnostic and therapy vials must be labeled in accordance with 10 CFR 35.69 and 10 CFR 20.1904, with the radionuclide, the activity, the date for which the activity is estimated, and the kinds of materials (i.e., radiopharmaceutical).
Syringes and unit dosages must be labeled in accordance with 10 CFR 35.69 and 10 CFR 20.1904, with the radionuclide, the activity, the date for which the activity is estimated, and the kinds of materials (i.e., radiopharmaceutical). If the container is holding less than the quantities listed in Appendix C to Part 20, the syringe or vial need only be labeled as containing radioactive material and the radiopharmaceutical (10 CFR 35.69). Consider also labeling the syringe with the type of study and the patient's name.
For prepared dosages, assay each patient dosage in the dose calibrator (or instrument) before administering it (10 CFR 35.63).
Do not use a dosage if it is greater than +20% off from the prescribed dosage, except for prescribed dosages of less than 30 microcuries or as approved by an authorized user.
When measuring the dosage, you need not consider the radioactivity that adheres to the syringe wall or remains in the needle.
Consider checking the patient's name and identification number and the prescribed radionuclide, chemical form, and dosage before administering. If the prescribed dosage requires a written directive, the patient's identity must be verified and the administration must be in accordance with the written directive (10 CFR 35.41).
Always keep flood sources, syringes, waste, and other radioactive material in shielded containers.
Consider using a cart or wheelchair to move flood sources, waste, and other radioactive material, since sources with small amounts of radioactivity exhibit a high dose rate on contact.
Secure all licensed material when not under the constant surveillance and immediate control of the authorized user(s).

Appendix U: Release of Patients or Human Research Subjects Administered Radioactive Materials

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Section 35.75, "Release of Individuals Containing Radiopharmaceuticals or Permanent Implants," of 10 CFR Part 35, "Medical Use of Byproduct Material," permits licensees to "authorize the release from its control of any individual who has been administered radiopharmaceuticals or permanent implants containing radioactive material if the total effective dose equivalent to any other individual from exposure to the released individual is not likely to exceed 5 millisieverts (0.5 rem)."

In this appendix, the individual or human research subject to whom the radioactive material has been administered is called the "patient."

Release Equation

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The activities at which patients could be released were calculated by using, as a starting point, the method discussed in the National Council on Radiation Protection and Measurements (NCRP) Report No. 37,"Precautions in the Management of Patients who Have Received Therapeutic Amounts of Radionuclides."

NCRP Report No. 37 uses the following equation to calculate the exposure until time t at a distance r from the patient:


Equation U.1:

Where
	D(t)	=	Accumulated exposure at time t, in roentgens,
	34.6	=	Conversion factor of 24 hrs/day times the total integration of decay (1.44),
		=	Specific gamma ray constant for a point source, R/mCi-hr at 1 cm,
	Q₀	=	Initial activity of the point source in millicuries, at the time of the release,
	T_p	=	Physical half-life in days,
	r	=	Distance from the point source to the point of interest, in centimeters,
	t	=	Exposure time in days.

This appendix uses the NCRP equation (Equation U.1) in the following manner to calculate the activities at which patients may be released.

The dose to an individual likely to receive the highest dose from exposure to the patient is taken to be the dose to total decay. Therefore, (1-e ^-0.693t/Tp ) is set equal to 1.
It is assumed that 1 roentgen is equal to 10 millisieverts (1 rem).
The exposure rate constants and physical half-lives for radionuclides typically used in nuclear medicine and brachytherapy procedures are given in Supplement A to this Appendix.
Default activities at which patients may be released are calculated using the physical half-lives of the radionuclides and do not account for the biological half-lives of the radionuclides.
When release is based on biological elimination (i.e., the effective half-life) rather than just the physical half-life of the radionuclide, Equation U.1 is modified to account for the uptake and retention of the radionuclide by the patient, as discussed in Supplement B.2.
For radionuclides with a physical half-life greater than 1 day and no consideration of biological elimination, it is assumed that the individual likely to receive the highest dose from exposure to the patient would receive a dose of 25% of the dose to total decay (0.25 in Equation 2), at a distance of 1 meter. Selection of 25% of the dose to total decay at 1 meter for estimating the dose is based on measurements discussed in the supporting regulatory analysis that indicate the dose calculated using an occupancy factor, E, of 25% at 1 meter is conservative in most normal situations.
For radionuclides with a physical half-life less than or equal to 1 day, it is difficult to justify an occupancy factor of 0.25, because relatively long-term averaging of behavior cannot be assumed. Under this situation, occupancy factors from 0.75 to 1.0 may be more appropriate.

Thus, for radionuclides with a physical half-life greater than 1 day:

Equation U.2:

For radionuclides with a physical half-life less than or equal to 1 day, and if an occupancy factor of 1.0 is used:

Equation U.3:

Equations U.2 and U.3 calculate the dose from external exposure to gamma radiation. These equations do not include the dose from internal intake by household members and members of the public, because the dose from intake by other individuals is expected to be small for most radiopharmaceuticals (less than a few percent), relative to the external gamma dose (see "Internal Dose,"of Supplement B). Further, the equations above do not apply to the dose to breast-feeding infants or children who continue to breast-feed. Patients who are breast-feeding an infant or child must be considered separately, as discussed in Regulatory Position 1.1, "Release of Patients Based on Administered Activity."

Regulatory Position

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1 Release Criteria

Licensees should use one of the following options to release a patient who has been administered radiopharmaceuticals or implants containing radioactive material in accordance with regulatory requirements.

1.1 Release of Patients Based on Administered Activity

In compliance with the dose limit in 10 CFR 35.75(a), licensees may release patients from licensee control if the activity administered is no greater than the activity in Column 1 of Table U.1. The activities in Table U.1 are based on a total effective dose equivalent of 5 millisieverts (0.5 rem) to an individual using conservative assumptions of

administered activity;
physical half-life;
occupancy factor of 0.25 at 1 meter for physical half-lives greater than 1 day, and, for conservatism, an occupancy factor of 1 at 1 meter for physical half-lives less than or equal to 1 day, and
no shielding by tissue.

The total effective dose equivalent is approximately equal to external dose because the internal dose is a small fraction of the external dose (see Section B.3, "Internal Dose," of Supplement B). In this case, no record of the release of the patient is required unless the patient is breast-feeding an infant or child, as discussed in Regulatory Position 3.2, "Records of Instructions for Breast-Feeding Patients." The licensee may demonstrate compliance by using the records of activity that are already required by 10 CFR 35.40 and 35.63.

If the activity administered exceeds the activity in Column 1 of Table U.1, the licensee may release the patient when the activity has decayed to the activity in Column 1 of Table U.1. In this case, a record is required by 10 CFR 35.75(c) because the patient's release is based on the retained activity rather than the administered activity. The activities in Column 1 of Table U.1 were calculated using either Equation U.2 or U.3, depending on the physical half-life of the radionuclide.

If a radionuclide not listed in Table U.1 is administered, the licensee can demonstrate compliance with the regulation by maintaining, for NRC inspection, calculation of the release activity that corresponds to the dose limit of 5 millisieverts (0.5 rem). Equation U.2 or U.3 may be used, as appropriate, to calculate the activity Q corresponding to 5 millisieverts (0.5 rem).

The release activities in Column 1 of Table U.1 do not include consideration of the dose to a breast-feeding infant or child from ingestion of radiopharmaceuticals contained in a patient's breast milk. When the patient is breast-feeding an infant or child, the activities in Column 1 of Table U.1 are not applicable to the infant or child. In this case, it may be necessary to give instructions as described in Regulatory Positions 2.2 and 2.3 as a condition for release. If failure to interrupt or discontinue could result in a dose to the breast-feeding infant or child in excess of 5 millisieverts (0.5 rem), a record that instructions were provided is required by 10 CFR 35.75(d).

1.2 Release of Patients Based on Measured Dose Rate

Licensees may release patients to whom radionuclides have been administered in amounts greater than the activities listed in Column 1 of Table U.1, provided the measured dose rate at 1 meter (from the surface of the patient) is no greater than the value in Column 2 of Table U.1 for that radionuclide. In this case, however, 10 CFR 35.75(c) requires a record because the release is based on considering shielding by tissue.

If a radionuclide not listed in Table U.1 is administered and the licensee chooses to release a patient based on the measured dose rate, the licensee should first calculate a dose rate that corresponds to the 5 millisievert (0.5 rem) dose limit. If the measured dose rate at 1 meter is no greater than the calculated dose rate, the patient may be released. A record of the release is required by 10 CFR 35.75(c). The dose rate at 1 meter may be calculated from Equation U.2 or U.3, as appropriate, because the dose rate at 1 meter is equal to Q/10,000 cm².

1.3 Release of Patients Based on Patient-Specific Dose Calculations

Licensees may release patients based on Dose Calculations using patient-specific parameters. With this method, based on 10 CFR 35.75(a), the licensee must calculate the maximum likely dose to an individual exposed to the patient on a case-by-case basis. If the calculated maximum likely dose to an individual is no greater than 5 millisieverts (0.5 rem), the patient may be released. Using this method, licensees may be able to release patients with activities greater than those listed in Column 1 of Table U.1 by taking into account the effective half-life of the radioactive material and other factors that may be relevant to the particular case. In this case, a record of the release is required by 10 CFR 35.75(c). If the dose calculation considered retained activity, an occupancy factor less than 0.25 at 1 meter, effective half-life, or shielding by tissue, a record of the basis for the release is required by 10 CFR 35.75(c).

Supplement B contains procedures for performing patient-specific dose calculations, and it describes how various factors may be considered in the calculations.

Table U.1 Activities and Dose Rates for Authorizing Patient Release^�

Radionuclide	COLUMN 1 Activity at or below Which Patients May Be released		COLUMN 2 Dose Rate at 1 Meter, at or below Which Patients May Be released*
	(GBq)	(mCi)	(mSv/hr)	(mrem/hr)
Ag-111	19	520	0.08	8
Au-198	3.5	93	0.21	21
Cr-51	4.8	130	0.02	2
Cu-64	8.4	230	0.27	27
Cu-67	14	390	0.22	22
Ga-67	8.7	240	0.18	18
I-123	6.0	160	0.26	26
I-125	0.25	7	0.01	1
I-125 implant	0.33	9	0.01	1
I-131	1.2	33	0.07	7
In-111	2.4	64	0.2	20
Ir-192 implant	0.074	2	0.008	0.8
P-32	**	**	**	**
Pd-103 implant	1.5	40	0.03	3
Re-186	28	770	0.15	15
Re-188	29	790	0.20	20
Sc-47	11	310	0.17	17
Se-75	0.089	2	0.005	0.5
Sm-153	26	700	0.3	30
Sn-117m	1.1	29	0.04	4
Sr-89	**	**	**	**
Tc-99m	28	760	0.58	58
Tl-201	16	430	0.19	19
Y-90	**	**	**	**
Yb-169	0.37	10	0.02	2
� The activity values were computed based on 5 millisieverts (0.5 rem) total effective dose equivalent. * If the release is based on the dose rate at 1 meter in Column 2, the licensee must maintain a record as required by 10 CFR 35.75(c), because the measurement includes shielding by tissue. See Regulatory Position 3.1, "Records of Release," for information on records. ** Activity and dose rate limits are not applicable in this case because of the minimal exposures to members of the public resulting from activities normally administered for diagnostic or therapeutic purposes.
*Notes*: The millicurie values were calculated using Equations U.2 or U.3 and the physical half-life. The gigabecquerel values were calculated based on the millicurie values and the conversion factor from millicuries to gigabecquerels. The dose rate values are calculated based on the millicurie values and the exposure rate constants. In general, the values are rounded to two significant figures; however, values less than 0.37 gigabecquerel (10 millicuries) or 0.1 millisievert (10 millirems) per hour are rounded to one significant figure. Details of the calculations are provided in NUREG-1492. Although non-byproduct materials are not regulated by the NRC, information on non-byproduct material is included in this regulatory guide for the convenience of the licensee. Agreement State regulations may vary. Agreement State licensees should check with their State regulations prior to using these values.

2 Instructions

2.1 Activities and Dose Rates Requiring Instructions

Based on 10 CFR 35.75(b), for some administrations, the released patients must be given instructions, including written instructions, on how to maintain doses to other individuals as low as is reasonably achievable after the patients are released.⁽¹⁰⁾ Column 1 of Table U.2 provides the activity above which instructions must be given to patients. Column 2 provides corresponding dose rates at 1 meter, based on the activities in Column 1. The activities or dose rates in Table U.2 may be used for determining when instructions must be given. If the patient is breast-feeding an infant or child, additional instructions may be necessary (see Regulatory Position 2.2, "Additional Instructions for Release of Patients Who Could be Breast-Feeding After Release").

When patient-specific calculations (as described in Supplement B) are used, instructions must be provided if the calculation indicates a dose that is greater than 1 millisievert (0.1 rem).

If a radionuclide not listed in Table U.2 is administered, the licensee may calculate the activity or dose rate that corresponds to 1 millisievert (0.1 rem). Equation U.2 or U.3, as appropriate, may be used.

2.2 Additional Instructions for Release of Patients Who Could be Breast-Feeding After Release

The requirement in 10 CFR 35.75(b) that a licensee provide instructions on the discontinuation or the interruption period of breast-feeding, and the consequences of failing to follow the recommendation, presumes that the licensee will inquire, as appropriate, regarding the breast-feeding status of the patient.⁸ The purpose of the instructions (e.g., on interruption or discontinuation) is to permit licensees to release a patient who could be breast-feeding an infant or child when the dose to the infant or child could exceed 5 millisieverts (0.5 rem) if there is no interruption of breast-feeding.

If the patient could be breast-feeding an infant or child after release, and if the patient was administered a radiopharmaceutical with an activity above the value stated in Column 1 of Table U.3, instructions on discontinuation or on the interruption period for breast-feeding and the consequences of failing to follow the recommendation must be provided. The patient should also be informed if there would be no consequences to the breast-feeding infant or child. Table U.3 also provides recommendations for interrupting or discontinuing breast-feeding to minimize the dose to below 1 millisievert (0.1 rem) if the patient has received certain radiopharmaceutical doses. The radiopharmaceuticals listed in Table U.3 are commonly used in medical diagnosis and treatment.

If a radiopharmaceutical not listed in Table U.3 is administered to a patient who could be breast-feeding, the licensee should evaluate whether instructions or records (or both) are required. If information on the excretion of the radiopharmaceutical is not available, an acceptable method is to assume that 50% of the administered activity is excreted in the breast milk. The dose to the infant or child can be calculated by using the dose conversion factors given for a newborn infant by Stabin.

2.3 Content of Instructions

The instructions should be specific to the type of treatment given, such as permanent implants or radioiodine for hyperthyroidism or thyroid carcinoma, and they may include additional information for individual situations; however, the instructions should not interfere with or contradict the best medical judgment of physicians. The instructions may include the name of a knowledgeable contact person and that person's telephone number, in case the patient has any questions. Additional instructions appropriate for each modality, as shown in examples below, may be provided (refer to 2.3.1 and 2.3.2).

Table U.2 Activities and Dose Rates Above Which Instructions Should be Given When Authorizing Patient Release*

Radionuclide	COLUMN 1 Activity above Which Instructions are Required		COLUMN 2 Dose Rate at 1 Meter above Which Instructions are Required
	(GBq)	(mCi)	(mSv/hr)	(mrem/hr)
Ag-111	3.8	100	0.02	2
Au-198	0.69	19	0.04	4
Cr-51	0.96	26	0.004	0.4
Cu-64	1.7	45	0.05	5
Cu-67	2.9	77	0.04	4
Ga-67	1.7	47	0.04	4
I-123	1.2	33	0.05	5
I-125	0.05	1	0.002	0.2
I-125 implant	0.074	2	0.002	0.2
I-131	0.24	7	0.02	2
In-111	0.47	13	0.04	4
Ir-192 implant	0.011	0.3	0.002	0.2
P-32	**	**	**	**
Pd-103 implant	0.3	8	0.007	0.7
Re-186	5.7	150	0.03	3
Re-188	5.8	160	0.04	4
Sc-47	2.3	62	0.03	3
Se-75	0.018	0.5	0.001	0.1
Sm-153	5.2	140	0.06	6
Sn-117m	0.21	6	0.009	0.9
Sr-89	**	**	**	**
Tc-99m	5.6	150	0.12	12
Tl-201	3.1	85	0.04	4
Y-90	**	**	**	**
Yb-169	0.073	2	0.004	0.4
* The activity values were computed based on 1 millisievert (0.1 rem) total effective dose equivalent. ** Activity and dose rate limits are not applicable in this case, because of the minimal exposures to members of the public resulting from activities normally administered for diagnostic or therapeutic purposes.
*Notes:* The millicurie values were calculated using Equations U.2 or U.3 and the physical half-life. The gigabecquerel values were calculated based on millicurie values and the conversion factor from millicuries to gigabecquerels. The dose rate values were calculated based on millicurie values and exposure rate constants. In general, values are rounded to two significant figures; however, values less than 0.37 gigabecquerel (10 millicuries) or 0.1 millisievert (10 millirems) per hour are rounded to one significant figure. Details of the calculations are provided in NUREG-1492. Although non-byproduct materials are not regulated by the NRC, information on non-byproduct material is included in this regulatory guide for the convenience of the licensee. Agreement State regulations may vary. Agreement State licensees should check with their State regulations prior to using these values.

Table U.3 Activities of Radiopharmaceuticals that Require Instructions and Records When Administered to Patients Who are Breast-Feeding an Infant or Child

Radionuclide	COLUMN 1 Activity above Which Instructions are Required		COLUMN 2 Activity above Which a Record is required		COLUMN 3 Examples of Recommended Duration of Interruption of Breast-Feeding*
	(MBq)	(mCi)	(MBq)	(mCi)
I-131 NaI	0.01	0.0004	0.07	0.002	Complete cessation (for this infant or child)
I-123 NaI	20	0.5	100	3
I-123 OIH	100	4	700	20
I-123 MIBG	70	2	400	10	24 hr for 370 MBq (10 mCi) 12 hr for 150 MBq (4 mCi)
I-125 OIH	3	0.08	10	0.4
I-131 OIH	10	0.30	60	1.5
Tc-99m DTPA	1000	30	6000	150
Tc-99m MAA	50	1.3	200	6.5	12.6 hr for 150 MBq (4 mCi)
Tc-99m Pertechnetate	100	3	600	15	24 hr for 1,100 MBq (30 mCi) 12 hr for 440 MBq(12 mCi)
Tc-99m DISIDA	1000	30	6000	150
Tc-99m Glucoheptonate	1000	30	6000	170
Tc-99m MIBI	1000	30	6000	150
Tc-99m MDP	1000	30	6000	150
Tc-99m PYP	900	25	4000	120
Tc-99m Red Blood Cell In Vivo Labeling	400	10	2000	50	6 hr for 740 MBq(20 mCi)
Tc-99m Red Blood Cell In Vitro Labeling	1000	30	6000	150
Tc-99m Sulphur Colloid	300	7	1000	35	6 hr for 440 MBq (12 mCi)
Tc-99m DTPA Aerosol	1000	30	6000	150
Tc-99m MAG3	1000	30	6000	150
Tc-99m White Blood Cells	100	4	600	15	24 hr for 1,100 MBq (30 mCi) 12 hr for 440 MBq (12 mCi)
Ga-67 Citrate	1	0.04	7	0.2	1 month for 150 MBq (4 mCi) 2 weeks for 50 MBq (1.3 mCi) 1 week for 7 MBq (0.2 mCi)
Cr-51 EDTA	60	1.6	300	8
In-111 White Blood Cells	10	0.2	40	1	1 week for 20 MBq (0.5 mCi)
Tl-201 Chloride	40	1	200	5	2 weeks for 110 MBq (3 mCi)
* The duration of interruption of breast-feeding is selected to reduce the maximum dose to a newborn infant to less than 1 millisievert (0.1 rem), although the regulatory limit is 5 millisieverts (0.5 rem). The actual doses that would be received by most infants would be far below 1 millisievert (0.1 rem). Of course, the physician may use discretion in the recommendation, increasing or decreasing the duration of interruption.
*Notes:* Activities are rounded to one significant figure, except when it was considered appropriate to use two significant figures. Details of the calculations are shown in NUREG-1492, "Regulatory Analysis on Criteria for the Release of Patients Administered Radioactive Material." If there is no recommendation in Column 3 of this table, the maximum activity normally administered is below the activities that require instructions on interruption or discontinuation of breast-feeding. Although non-byproduct materials are not regulated by the NRC, information on non-byproduct material is included in this regulatory guide for the convenience of the licensee. Agreement State regulations may vary. Agreement State licensees should check with their State regulations prior to using these values.

2.3.1 Instructions Regarding Radiopharmaceutical Administrations

For procedures involving radiopharmaceuticals, additional instructions may include the following:

Maintaining distance from other persons, including separate sleeping arrangements.
Minimizing time in public places (e.g., public transportation, grocery stores, shopping centers, theaters, restaurants, sporting events).
Precautions to reduce the spread of radioactive contamination.
The length of time each of the precautions should be in effect.

The Society of Nuclear Medicine published a pamphlet in 1987 that provides information for patients receiving treatment with radioiodine. This pamphlet was prepared jointly by the Society of Nuclear Medicine and the NRC. The pamphlet contains blanks, for the physician to fill in the length of time that each instruction should be followed. While this pamphlet was written for the release of patients to whom less than 1,110 mega becquerels (30 millicuries) of iodine-131 had been administered, the NRC still considers the instructions in this pamphlet to be an acceptable method for meeting the requirements of 10 CFR 35.75(b), provided the times filled in the blanks are appropriate for the activity and the medical condition.

If additional instructions are required because the patient is breast-feeding, the instructions should include appropriate recommendations on whether to interrupt breast-feeding, the length of time to interrupt breast-feeding, or, if necessary, the discontinuation of breast-feeding. The instructions should include information on the consequences of failure to follow the recommendation to interrupt or discontinue breast-feeding. The consequences should be explained so that the patient will understand that, in some cases, breast-feeding after an administration of certain radionuclides should be avoided. For example, a consequence of procedures involving iodine-131 is that continued breast-feeding could harm the infant's or child's thyroid. Most diagnostic procedures involve radionuclides other than radioiodine and there would be no consequences; guidance should simply address avoiding any unnecessary radiation exposure to the infant or child from breast-feeding. If the Society of Nuclear Medicine's pamphlet is given at release to a patient who is breast-feeding an infant or child, the pamphlet should be supplemented with information specified in 10 CFR 35.75(b)(1) and (2).

The requirement of 10 CFR 35.75(b) regarding written instructions to patients who could be breast-feeding an infant or child does not in any way interfere with the discretion and judgment of the physician in specifying the detailed instructions and recommendations.

2.3.2 Instructions Regarding Implants

For patients who have received implants, additional instructions may include the following:

A small radioactive source has been placed (implanted) inside your body. The source is actually many small metallic pellets or seeds, which are each about 1/3 to 1/4 of an inch long, similar in size and shape to a grain of rice. To minimize exposure to radiation to others from the source inside your body, you should do the following for _____ days.

Stay at a distance of ______ feet from_______.
Maintain separate sleeping arrangements.
Minimize time with children and pregnant women.
Do not hold or cuddle children.
Avoid public transportation.
Examine any bandages or linens that come into contact with the implant site for any pellets or seeds that may have come out of the implant site.
If you find a seed or pellet that falls out:
Do not handle it with your fingers. Use something like a spoon or tweezers to place it in a jar or other container that you can close with a lid.
Place the container with the seed or pellet in a location away from people.
Notify _________ at telephone number _____________.

3 Records

3.1 Records of Release

There is no requirement for recordkeeping on the release of patients who were released in accordance with Column 1 of Table U.1; however, if the release of the patient is based on a dose calculation that considered retained activity, an occupancy factor less than 0.25 at 1 meter, effective half-life, or shielding by tissue, a record of the basis for the release is required by 10 CFR 35.75(c). This record should include the patient identifier (in a way that ensures that confidential patient information is not traceable or attributable to a specific patient), the radioactive material administered, the administered activity, and the date of the administration. In addition, depending on the basis for release, records should include the following information:

For Immediate Release of a Patient Based on a Patient-Specific Calculation: The equation used, including the patient-specific factors and their bases that were used in calculating the dose to the person exposed to the patient, and the calculated dose. The patient-specific factors (see Supplement B of this Appendix) include the effective half-life and uptake fraction for each component of the biokinetic model, the time that the physical half-life was assumed to apply to retention, and the occupancy factor. The basis for selecting each of these values should be included in the record.
For Immediate Release of a Patient Based on Measured Dose Rate: The results of the measurement, the specific survey instrument used, and the name of the individual performing the survey.
For Delayed Release of a Patient Based on Radioactive Decay Calculation: The time of the administration, date and time of release, and the results of the decay calculation.
For Delayed Release of a Patient Based on Measured Dose Rate: The results of the survey meter measurement, the specific survey instrument used, and the name of the individual performing the survey.

In some situations, a calculation may be case-specific for a class of patients who all have the same patient-specific factors. In this case, the record for a particular patient's release may reference the calculation for the class of patients.

Records, as required by 10 CFR 35.75(c), should be kept in a manner that ensures the patient's confidentiality, that is, the records should not contain the patient's name or any other information that could lead to identification of the patient. These recordkeeping requirements may also be used to verify that licensees have proper procedures in place for assessing potential third-party exposure associated with and arising from exposure to patients who were administered radioactive material.

3.2 Records of Instructions for Breast-Feeding Patients

If failure to interrupt or discontinue breast-feeding could result in a dose to the infant or child in excess of 5 millisieverts (0.5 rem), a record that instructions were provided is required by 10 CFR 35.75(d). Column 2 of Table U.3 states, for the radiopharmaceuticals commonly used in medical diagnosis and treatment, the activities that would require such records when administered to patients who are breast-feeding.

The record should include the patient's identifier (in a way that ensures that confidential patient information is not traceable or attributable to a specific patient), the radiopharmaceutical administered, the administered activity, the date of the administration, and whether instructions were provided to the patient who could be breast-feeding an infant or child.

4 Summary Table

Table U.4 summarizes the criteria for releasing patients and the requirements for providing instructions and maintaining records.

Table U.4 Summary of Release Criteria, Required Instructions to Patients, and Records to Be Maintained

Patient Group	Basis for Release	Criteria for Release	Instructions Needed?	Release Records Required?
All patients, including patients who are breast-feeding an infant or child	Administered activity	Administered activity < Column 1 of Table U.1	Yes, if administered activity > Column 1 of Table U.2	No
	Retained activity	Retained activity < Column 1 of Table U.1	Yes, if retained activity > Column 1 of Table U.2	Yes
	Measured dose rate	Measured dose rate < Column 2 of Table U.1	Yes, if dose rate > Column 2 of Table U.2	Yes
	Patient-specific calculations	Calculated dose < 5 mSv (0.5 rem)	Yes, if calculated dose > 1 mSv (0.1 rem)	Yes
Patients who are breast-feeding an infant or child	All of the above bases for release		Additional instructions required if: Administered activity > Column 1 of Table U.3 OR Licensee calculated dose from breast-feeding >1 mSv (0.1 rem) to the infant or child	Records that instructions were provided are required if: Administered activity > Column 2 of Table U.3 OR Licensee calculated dose from continued breast-feeding > 5 mSv (0.5 rem) to the infant or child

Implementation

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The purpose of this section is to provide information to licensees and applicants regarding the NRC staff's plans for using this appendix. Except in those cases in which a licensee proposes an acceptable alternative method for complying with 10 CFR 35.75, the methods described in this appendix will be used in the evaluation of a licensee's compliance with 10 CFR 35.75.

References

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National Council on Radiation Protection and Measurements (NCRP), "Precautions in the Management of Patients Who Have Received Therapeutic Amounts of Radionuclides," NCRP Report No. 37, October 1, 1970. (Available for sale from the NCRP, 7910 Woodmont Avenue, Suite 800, Bethesda, MD 20814-3095.)

S. Schneider and S. A. McGuire, "Regulatory Analysis on Criteria for the Release of Patients Administered Radioactive Material," NUREG-1492 (Final Report), NRC, February 1997.

M. Stabin, "Internal Dosimetry in Pediatric Nuclear Medicine," in Pediatric Nuclear Medicine, Edited by S. Treves, Springer Verlag, New York, 1995.

"Guidelines for Patients Receiving Radioiodine Treatment," Society of Nuclear Medicine, 1987. This pamphlet may be obtained from the Society of Nuclear Medicine, 136 Madison Avenue, New York, NY 10016-6760.

Supplement A

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Table U.5 Half-Lives and Exposure Rate Constants of Radionuclides Used in Medicine

Radionuclide¹	Physical Half-Life (days)²	Exposure Rate Constant³ (R/mCi-h at 1 cm)
Ag-111	7.45	0.150
Au-198	2.696	2.3
Cr-51	27.704	0.16
Cu-64	0.529	1.2
Cu-67	2.578	0.58
Ga-67	3.261	0.753
I-123	0.55	1.61
I-125	60.14	1.42
I-125 implant	60.14	1.11⁴
I-131	8.04	2.2
In-111	2.83	3.21
Ir-192 implant	74.02	4.59⁴
P-32	14.29	NA⁶
Pd-103 implant	16.96	0.86⁵
Re-186	3.777	0.2
Re-188	0.708	0.26
Sc-47	3.351	0.56
Se-75	119.8	2.0
Sm-153	1.946	0.425
Sn-117m	13.61	1.48
Sr-89	50.5	NA⁶
Tc-99m	0.251	0.756
Tl-201	3.044	0.447
Yb-169	32.01	1.83
Y-90	2.67	NA⁶
Yb-169	32.01	1.83
¹ Although non-byproduct materials are not regulated by the NRC, information on non-byproduct material is included in this regulatory guide for the convenience of the licensee. ² K.F. Eckerman, A.B. Wolbarst, and A.C.B. Richardson, "Federal Guidance Report No. 11, Limiting Values of Radionuclide Intake and Air concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," Report No. EPA-520/1-88-020, Office of Radiation Programs, U.S. Environmental Protection Agency, Washington, DC, 1988. ³ Values for the exposure rate constant for Au-198, Cr-51, Cu-64, I-131, Sc-47, and Se-75 were taken from the Radiological Health Handbook, U.S. Department of Health, Education, and Welfare, pg. 135, 1970. For Cu-67, I-123, In-111, Re-186, and Re-188, the values for the exposure rate constant were taken from D.E. Barber, J.W. Baum, and C.B. Meinhold, "Radiation Safety Issues Related to Radiolabeled Antibodies," NUREG/CR-4444, U.S. NRC, Washington, DC, 1991. For Ag-111, Ga-67, I-125, Sm-153, Sn-117m, Tc-99m, Tl-201, and Yb-169, the exposure rate constants were calculated because the published values for these radionuclides were an approximation, presented as a range, or varied from one reference to another. Details of the calculation of the exposure rate constants are shown in Table A.2 of Appendix A to NUREG-1492, "Regulatory Analysis on Criteria for the Release of patients Administered Radioactive Material," U.S. NRC, February 1997. ⁴ R. Nath, A.S. Meigooni, and J.A. Meli, "Dosimetry on Transverse Axes of ¹²⁵I and ¹⁹²Ir Interstitial Brachytherapy Sources," Medical Physics, Volume 17, Number 6, November/December 1990. The exposure rate constant given is a measured value averaged for several source models and takes into account the attenuation of gamma rays within the implant capsule itself. ⁵ A.S. Meigooni, S. Sabnis, R. Nath, "Dosimetry of Palladium-103 Brachytherapy Sources for Permanent Implants," Endocurietherapy Hyperthermia Oncology, Volume 6, April 1990. The exposure rate constant given is an "apparent" value (i.e., with respect to an apparent source activity) and takes into account the attenuation of gamma rays within the implant capsule itself. ⁶ Not applicable (NA) because the release activity is not based on beta emissions.

Supplement B

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Procedures for Calculating Doses Based on Patient-Specific Factors

A licensee may release a patient who has been administered an activity higher than the values listed in Column 1 of Table U.1 of this supplement, if dose calculations using patient-specific parameters, which are less conservative than the conservative assumptions, show that the potential total effective dose equivalent to any individual would be no greater than 5 millisieverts (0.5 rem).

If the release of a patient is based on a patient-specific calculation that considered retained activity, an occupancy factor less than 0.25 at 1 meter, effective half-life, or shielding by tissue, a record of the basis of the release is required by 10 CFR 35.75(c). The following equation can be used to calculate doses:

Equation B-1:

Where
	D(t)	=	Accumulated dose to time t, in rems
	34.6	=	Conversion factor of 24 hrs/day times the total integration of decay (1.44)
		=	Exposure rate constant for a point source, R/mCi x hr at 1 cm
	Q₀	=	Initial activity at the start of the time interval
	T_p	=	Physical half-life, in days
	E	=	Occupancy factor that accounts for different occupancy times and distances when an individual is around a patient
	r	=	Distance in centimeters. This value is typically 100 cm
	t	=	exposure time in days

B.1 Occupancy Factor

B.1.1 Rationale for Occupancy Factors Used to Derive Table U.1

In Table U.1 in this Appendix, the activities at which patients could be released were calculated using the physical half-life of the radionuclide and an occupancy factor at 1 meter of either 0.25 (if the radionuclide has a half-life longer than 1 day) or 1.0 (if the radionuclide has a half-life less than or equal to 1 day). The basis for the occupancy factor of 0.25 at 1 meter is that measurements of doses to family members, as well as considerations of normal human behavior (as discussed in the supporting regulatory analysis (Ref. B-1)), suggest that an occupancy factor of 0.25 at 1 meter, when used in combination with the physical half-life, will produce a generally conservative estimate of the dose to family members when instructions on minimizing doses to others are given.

An occupancy factor of 0.25 at 1 meter is not considered appropriate when the physical half-life is less than or equal to 1 day, and hence, the dose is delivered over a short time. Specifically, the assumptions regarding patient behavior that led to an occupancy factor of 0.25 at 1 meter include the assumption that the patient will not be in close proximity to other individuals for several days; however, when the dose is from a short-lived radionuclide, the time that individuals spend in close proximity to the patient immediately following release will be most significant because the dose to other individuals could be a large fraction of the total dose from the short-lived radionuclide. Thus, to be conservative when providing generally applicable release quantities that may be used with little consideration of the specific details of a particular patient's release, the values calculated in Table U.1 were based on an occupancy factor of 1 at 1 meter when the half-life is less than or equal to 1 day.

B.1.2 Occupancy Factors to Consider for Patient- Specific Calculations

The selection of an occupancy factor for patient-specific calculations will depend on whether the physical or effective half-life of the radionuclide is used and whether instructions are provided to the patient before release. The following occupancy factors, E, at 1 meter, may be used for patient-specific calculations.

E = 0.75 when a physical half-life, an effective half-life, or a specific time period under consideration (e.g., bladder holding time) is less than or equal to 1 day.
E = 0.25 when an effective half-life is greater than 1 day, if the patient has been given instructions, such as:

Maintain a prudent distance from others for at least the first 2 days,
Sleep alone in a room for at least the first night,
Do not travel by airplane or mass transportation for at least the first day,
Do not travel on a prolonged automobile trip with others for at least the first 2 days,
Have sole use of a bathroom for at least the first 2 days,
Drink plenty of fluids for at least the first 2 days.

E = 0.125 when an effective half-life is greater than 1 day if the patient has been given instructions, such as:

Follow the instructions for E = 0.25 above,
Live alone for at least the first 2 days,
Have few visits by family or friends for at least the first 2 days.

In a two-component model (e.g., uptake of iodine-131 using thyroidal and extrathyroidal components), if the effective half-life associated with one component is less than or equal to one day but is greater than one day for the other component, it is more justifiable to use the occupancy factor associated with the dominant component for both components.

Example 1: Calculate the maximum likely dose to an individual exposed to a patient who has received 2,220 megabecquerels (60 millicuries) of iodine-131. The patient has been provided with instructions to maintain a prudent distance from others for at least 2 days, lives alone, drives home alone, and stays at home for several days without visitors.

Solution: The dose to total decay (t = ) is calculated based on the physical half-life using Equation B-1. (This calculation illustrates the use of physical half-life. To account for biological elimination, calculations described in the next section should be used.)

Since the patient has been provided with instructions for reducing exposure as recommended for an occupancy factor of E = 0.125, the occupancy factor of 0.125 at 1 meter may be used.

D () = 4.59 millisieverts (0.459 rem)

Since the dose is less than 5 millisieverts (0.5 rem), the patient may be released, but 10 CFR 35.75(b) requires that instructions be given to the patient on maintaining doses to others as low as is reasonably achievable. A record of the calculation must be maintained, pursuant to 10 CFR 35.75(c), because an occupancy factor less than 0.25 at 1 meter was used.

B.2 Effective Half-Life

A licensee may take into account the effective half-life of the radioactive material to demonstrate compliance with the dose limits for individuals exposed to the patient that are stated in 10 CFR 35.75. The effective half-life is defined as:

Equation B-2:

Where


T_b	=	biological half-life of the radionuclide
T_p	=	physical half-life of the radionuclide.

The behavior of iodine-131 can be modeled using two components: extrathyroidal iodide (i.e., existing outside of the thyroid) and thyroidal iodide following uptake by the thyroid. The effective half-lives for the extrathyroidal and thyroidal fractions (i.e., F1 and F2, respectively) can be calculated with the following equations.

Equation B-3:

Equation B-4:

Where


T_b1	=	biological half-life for extrathyroidal iodide
T_b2	=	biological half-life of iodide following uptake by the thyroid
T_p	=	physical half-life of iodine-131

However, simple exponential excretion models do not account for (a) the time for the iodine-131 to be absorbed from the stomach to the blood and (b) the holdup of iodine in the urine while in the bladder. Failure to account for these factors could result in an under estimate of the dose to another individual. Therefore, this supplement makes a conservative approximation to account for these factors by assuming that, during the first 8 hours after the administration, about 80% of the iodine administered is removed from the body at a rate determined only by the physical half-life of iodine-131.

Thus, an equation to calculate the dose from a patient administered iodine-131 may have three components. First is the dose for the first 8-hours (0.33 day) after administration. This component comes directly from Equation B-1, using the physical half-life and a factor of 80%. Second is the dose from the extrathyroidal component from 8 hours to total decay. In this component, the first exponential factor represents the activity at t = 8 hours based on the physical half-life of iodine-131. The second exponential factor represents the activity from t = 8 hours to total decay based on the effective half-life of the extrathyroidal component. The third component, the dose from the thyroidal component for 8 hours to total decay, is calculated in the same manner as the second component. The full equation is shown as Equation B-5.

Equation B-5: + _e^{-0.693(0.33)/Tp}E₂ F₁ T_1eff + e^{-0.693(0.33)/Tp} E₂ F₂ T_2eff

Where


F₁	=	Extrathyroidal uptake fraction
F₂	=	Thyroidal uptake fraction
E₁	=	Occupancy factor for the first 8 hours
E₂	=	Occupancy factor from 8 hours to total decay

All the other parameters are as defined in equations B-1, B-3, and B-4. Acceptable values for F₁, T_{1 eff}, F₂, and T_2eff are shown in Table U.6 for thyroid ablation and treatment of thyroid remnants after surgical removal of the thyroid for thyroid cancer. If these values have been measured for a specific individual, the measured values may be used.

The record of the patient's release required by 10 CFR 35.75(c) is described in Regulatory Position 3.1 of this appendix.

Example 2, Thyroid Cancer: Calculate the maximum likely dose to an individual exposed to a patient who has been administered 7,400 megabecquerels (200 millicuries) of iodine-131 for the treatment of thyroid remnants and metastases.

Solution: In this example, we will calculate the dose by using Equation B-5 to account for the elimination of iodine-131 from the body, based on the effective half-lives appropriate for thyroid cancer. The physical half-life and the exposure rate constant are from Table U.5. The uptake fractions and effective half-lives are from Table U.6. An occupancy factor, E, of 0.75 at 1 meter, will be used for the first component because the time period under consideration is less than 1 day; however, for the second and third components, an occupancy factor of 0.25 will be used, because (1) the effective half-life associated with the dominant component is greater than 1 day, and (2) patient-specific questions were provided to the patient to justify the occupancy factor (see Section B.1.2, "Occupancy Factors to Consider for patient-Specific Calculations," of this Supplement B).

Table U.6 Uptake Fractions and Effective Half-Lives for Iodine-131 Treatments

	Extrathyroidal Component		Thyroidal Component
Medical Condition	Uptake Fraction F₁	Effective Half-Life T_1eff (day)	Uptake Fraction F₂	Effective Half-Life T_2eff (day)
Hyperthyroidism	0.20¹	0.32²	0.80¹	5.2¹
Post thyroidectomy for Thyroid Cancer	0.95³	0.32²	0.05³	7.3²
¹ M.G. Stabin et al., "Radiation Dosimetry for the Adult Female and Fetus from Iodine-131 Administration in Hyperthyroidism," Journal of Nuclear Medicine, Volume 32, Number 5, May 1991. The thyroid uptake fraction of 0.80 was selected as one that is seldom exceeded by the data shown in Figure 1 in this referenced document. The effective half-life of 5.2 days for the thyroidal component was derived from a biological half-life of 15 days, which was obtained from a straight-line fit that accounts for about 75% of the data points shown in Figure 1 of that Journal of Nuclear Medicine document. ² International Commission on Radiological Protection (ICRP), "Radiation Dose to Patients from Radiopharmaceuticals," ICRP Publication No. 53, March 1987. (Available for sale from Pergamon Press, Inc., Elmsford, NY 10523.) The data in that document suggest that the extrathyroidal component effective half-life in normal subjects is about 0.32 days. Lacking other data, this value is applied to hyperthyroid and thyroid cancer patients. For thyroid cancer, the thyroidal component effective half-life of 7.3 days is based on a biological half-life of 80 days (adult thyroid), as suggested in the ICRP document. ³ The thyroidal uptake fraction of 0.05 was recommended by Dr. M. Pollycove, M.D., NRC medical visiting fellow, as an upper-limit postthyroidectomy for thyroid cancer.

Substituting the appropriate values into Equation B-5, the dose to total decay is:

+ e^{-0.693(0.33)/8.04} (0.25)(0.95)(0.32) + e^{-0.693(0.33)/8.04} (0.25)(0.05)(7.3)

D() = 4.53 millisieverts (0.453 rem)

Therefore, thyroid cancer patients administered 7,400 megabecquerels (200 millicuries) of iodine-131 or less would not have to remain under licensee control and could be released under 10 CFR 35.75, assuming that the foregoing assumptions can be justified for the individual patient's case and that the patient is given instructions. Patients administered somewhat larger activities could also be released immediately if the dose is not greater than 5 millisieverts (0.5 rem).

In the example above, the thyroidal fraction, F₂ = 0.05, is a conservative assumption for persons who have had surgery to remove thyroidal tissue. If F₂ has been measured for a specific patient, the measured value may be used.

Example 3, Hyperthyroidism: Calculate the maximum likely dose to an individual exposed to a patient who has been administered 2,035 megabecquerels (55 millicuries) of iodine-131 for the treatment of hyperthyroidism (i.e., thyroid ablation).

Solution: In this example, we will again calculate the dose using Equation B-5, Table U.5, and Table U.6, to account for the elimination of iodine-131 from the body by using the effective half-lives appropriate for hyperthyroidism. An occupancy factor, E, of 0.25 at 1 meter will be used for the second and third components of the equation because patient-specific instructions were provided to justify the occupancy factor (see Section B.1.2, "Occupancy Factors to Consider for Patient-Specific Calculations").

Substituting the appropriate values into Equation B-5, the dose to total decay is:

+ e^{-0.693(0.33)/8.04} (0.25)(0.20)(0.32) + e^{-0.693(0.33)/8.04} (0.25)(0.80)(5.2)

D() = 4.86 mSv (0.486 rem)

Therefore, hyperthyroid patients administered 2,035 megabecquerels (55 millicuries) of iodine-131 would not have to remain under licensee control and could be released under 10 CFR 35.75, once the occupancy factor of 0.25 in the second and third components of the equation is justified.

In the example above, the thyroidal fraction, F₂ = 0.8, is a conservative assumption for persons who have this treatment for hyperthyroidism. If F₂ has been measured for a specific patient, the measured value may be used.

B.3 Internal Dose

For some radionuclides, such as iodine-131, there may be concerns that the internal dose of an individual from exposure to a released patient could be significant. A rough estimate of the maximum likely committed effective dose equivalent from internal exposure can be calculated from Equation B-6.

Equation B-6: D_i = Q (10^-5)(DCF)

Where


D_i	=	Maximum likely internal committed effective dose equivalent to the individual exposed to the patient in rems
Q	=	Activity administered to the patient in millicuries
10^-5	=	Assumed fractional intake
DCF	=	Dose conversion factor to convert an intake in millicuries to an internal committed effective dose equivalent (such as tabulated in Reference B-2).

Equation B-6 uses a value of 10^-5 as the fraction of the activity administered to the patient that would be taken in by the individual exposed to the patient. A common rule of thumb is to assume that no more than 1 millionth of the activity being handled will become an intake to an individual working with the material. This rule of thumb was developed in reference B-3 for cases of worker intakes during normal workplace operations, worker intakes from accidental exposures, and public intakes from accidental airborne releases from a facility, but it does not specifically apply for cases of intake by an individual exposed to a patient. However, two studies (Refs. B-4 and B-5) regarding the intakes of individuals exposed to patients administered iodine-131 indicated that intakes were generally of the order of 1 millionth of the activity administered to the patient and that internal doses were far below external doses. To account for the most highly exposed individual and to add a degree of conservatism to the calculations, a fractional transfer of 10^-5 has been assumed.

Example 4, Internal Dose: Using the ingestion pathway, calculate the maximum internal dose to a person exposed to a patient who has been administered 1,221 megabecquerels (33 millicuries) of iodine-131. The ingestion pathway was selected since it is likely that most of the intake would be through the mouth or through the skin, which is most closely approximated by the ingestion pathway.

Solution: This is an example of the use of Equation B-6. The dose conversion factor DCF for the ingestion pathway is 53 rems/millicurie from Table 2.2 of Reference B-2.

Substituting the appropriate values into Equation B-6, the maximum internal dose to the person is

D_i = (33 mCi)(10^-5)(53 rem/mCi)

D_i = 0.17 mSv (0.017 rem)

In this case, the external dose to the other person would-be no greater than 5 millisieverts (0.5 rem), while the internal dose would-be about 0.17 millisievert (0.017 rem). Thus, the internal dose is about 3% of the external gamma dose. Internal doses may be ignored in the calculations if they are likely to be less than 10% of the external dose, since the internal dose would be significantly less than the uncertainty in the external dose.

The conclusion that internal contamination is relatively unimportant in the case of patient release was also reached by the NCRP. The NCRP addressed the risk of intake of radionuclides from patients' secretions and excreta in NCRP Commentary No. 11, "Dose Limits for individuals Who Receive Exposure from Radionuclide Therapy Patients" (Ref. B-6). The NCRP concluded, "Thus, a contamination incident that could lead to a significant intake of radioactive material is very unlikely." For additional discussion on the subject, see Reference B-1.

References for Supplement B

B-1. S. Schneider and S.A. McGuire, "Regulatory Analysis on Criteria for the Release of Patients Administered Radioactive Material," USNRC, NUREG-1492, February 1997.

B-2. K.F. Eckerman, A.B. Wolbarst, and A.C.B. Richardson, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," Federal Guidance Report No.11, U. S. Environmental Protection Agency, Washington, DC, 1988.

B-3. A. Brodsky, "Resuspension Factors and Probabilities of Intake of Material in Process (or 'Is 10^-6 a Magic Number in Health Physics?')," Health Physics, Volume 39, Number 6, 1980.

B-4. R.C.T. Buchanan and J.M. Brindle, "Radioiodine Therapy to Out-patients - The Contamination Hazard," British Journal of Radiology, Volume 43, 1970.

B-5. A.P. Jacobson, P.A. Plato, and D. Toeroek, "Contamination of the Home Environment by Patients Treated with Iodine-131," American Journal of Public Health, Volume 68, Number 3, 1978.

B-6. National Council on Radiation Protection and Measurements, "Dose Limits for Individuals Who Receive Exposure from Radionuclide Therapy Patients," Commentary No. 11, February 28, 1995.

Regulatory Analysis

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"Regulatory Analysis on Criteria for the Release of Patients Administered Radioactive Material" (NUREG-1492, February 1997), provides the regulatory basis for this guide and examines the costs and benefits. A copy of NUREG-1492 is available for inspection and copying for a fee at the NRC Public Document Room, 2120 L Street NW., Washington, DC. Copies may be purchased at current rates from the U.S. Government Printing Office, P.O. Box 37082, Washington, DC 20402-9328 (telephone (202)512-2249), or from the National Technical Information Service by writing NTIS at 5285 Port Royal Road, Springfield, VA 22161.

Appendix V: Guidance for Mobile Services

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Mobile Services

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Type and Location of Use

There are three classes of mobile service. Class 1 mobile service is defined as transportation and use of byproduct material within a transport vehicle (e.g., in-van use). Class 2 mobile service is defined as transportation of byproduct material to a client's facility and use within a client's facility by the mobile service's employees (i.e., transport and use). Class 3 mobile service is defined as transportation of byproduct material to a client's facility and use of the byproduct material by the client's employees (i.e., transport only).

Class 1 and 2 mobile service providers must apply for full service authorization. Class 3 mobile service providers need only apply for authorization for possession and transport of the byproduct material. For Class 3 mobile services, the client must possess a license for medical use of the byproduct material. For Class 3 mobile services, the client is authorized to provide the patient treatments and is responsible for all aspects of the byproduct material use and patient treatments upon transfer of the byproduct material to their possession.

For all classes, list the medical institutions, hospitals, or clinics and their addresses that comprise the client sites for mobile services. Licensed activities must be conducted in accordance with the regulations for compliance with 10 CFR 35.80(a), which states that we will obtain a letter signed by the management (i.e., chief executive officer or delegate) of each of our clients for which services are rendered. The letter will permit the use of byproduct material at the client's address and clearly delineate the authority and responsibility of each entity. This agreement must be applicable for the entire period of time over which the service is to be provided. The letter will be retained for three years after the last provision of service. Additionally, we will develop and implement survey procedures to ensure that all byproduct material, including radiopharmaceuticals, sealed sources, and all associated wastes have been removed before leaving each location of use as required by 10 CFR 35.80(d).

Applicants for self-contained byproduct material services (i.e., Class 1) must provide the following facility information:

A detailed description and diagram(s) of the proposed use facility (e.g., van) and associated equipment, in accordance with Items 8.16 through 8.20 of this report. The description and diagram of the proposed use facility must demonstrate that the facility is of adequate construction and design to protect its contents from the elements (e.g., high winds, rain), to ensure security of licensed material including controlling access to the van and storage area keys, and to ensure that radiation levels in unrestricted areas are in compliance with 10 CFR 20.1301. Include a diagram showing the location of the equipment, receipt, and use areas, and identify all areas adjacent to restricted areas. The description of the van must address radiation levels in the van driver's compartment to demonstrate compliance with 10 CFR 20.1201, "Occupational dose limits for adults."
For therapy treatments with byproduct material, submit a separate drawing for each client site showing the location of the treatment device/vehicle in relation to all nearby roads, sidewalks, structures, and any other locations accessible by members of the public.
A signed agreement that location of the device/vehicle will be on client-owned or controlled property.
Protection from vehicular traffic that could adversely affect patient treatment(s) should be provided. This could be accomplished either by locating the facility away from all vehicular traffic or by using barriers. Any protective measures must be shown on the facility/site drawings provided.
A description of the emergency lighting system that automatically activates on detection of the loss of primary power during patient remote afterloader treatments. The system must provide sufficient light to perform any possible emergency procedures, including the removal of a detached or stuck source that remains within the patient.

Applicants who will provide transportable services to the client's site and use within the client's facility (i.e., Class 2) must provide the following facility information and commitment:

A detailed description and diagram(s) of the proposed use facility (e.g., client site) and associated equipment in accordance with Items 8.16 through 8.20 of this report. The description and diagram of the proposed use facility must demonstrate that the facility is of adequate construction and design to protect its contents from the elements (e.g., high winds, rain), to ensure security of licensed material to prevent unauthorized access, and to ensure that radiation levels in unrestricted areas are in compliance with 10 CFR 20.1301. Include a diagram showing the location of the equipment, receipt, and use areas, and identify all areas adjacent to restricted areas.
A commitment to retain a record establishing that the mobile service licensee has full control of the treatment room during byproduct material use for each client. This proof may be in either the form of a signed agreement or a lease agreement with the client, establishing full control of the treatment room by the applicant during all periods of use.
The initial installation records of a remote afterloader device for each site of use, and function checks of the remote afterloader device.

Applicants for a Class 3 mobile service must ensure the following:

Each client is properly licensed for medical use of byproduct material. If applicable, ensure that each client has received the necessary initial and, if appropriate, recurrent training for the specific make and model of the remote afterloader device being provided. If the above applicable conditions are not met, the mobile service licensee must not transfer the remote afterloader device to the client.
No signed agreement with a client may state nor imply any assumption of responsibility on the part of the mobile service for the use of byproduct material for patient treatments. This includes such activities as dosage measurements, source calibrations and remote afterloader device operational checks. Although these and other services may be provided to the client by the mobile service if specifically licensed to provide such services, the client (licensee) retains all of the responsibilities related to the use of the byproduct material for patient treatments. The responsibilities for supervision of individuals in the use of the byproduct material, set forth in 10 CFR 35.27, transfer to the client's AUs upon transfer of the device to the client by the mobile service provider.
The initial installation of a remote afterloader device at the client site may be performed by either the mobile service provider or the client, but all device function checks are the responsibility of the client (i.e., the licensee authorized to provide patient treatments at the client site).
A formal record of the transfer of control of the byproduct material from the mobile service provider to the client, and from the client back to the mobile service provider, must be made for each transfer of the byproduct material. A signed receipt of each transfer must be made and retained for inspection for a period of three years.

Base Location

The base location (e.g., central radiopharmaceutical laboratory) for the mobile service must be specified. The base facility may be located in a medical institution, non-institutional medical practice, or commercial facility. As required by 10 CFR 30.33, submit a detailed description and diagram(s) of the proposed base facility and associated equipment in accordance with Items 8.16 through 8.20 of this report. The description and diagram of the proposed facility must demonstrate that the building is of adequate construction and design to protect its contents from the elements (e.g., high winds, rain), to ensure security of licensed material to prevent unauthorized access, and to ensure that radiation levels in unrestricted areas are in compliance with 10 CFR 20.1301. Include a diagram showing the location of the equipment, receipt, and use areas, and identify all areas adjacent to restricted areas, including areas above and below the restricted areas.

The applicant may request multiple base locations. Radioactive material must be delivered only to a facility licensed to receive the type of radioactive material ordered.
Base locations can include the use of a mobile van. In those cases in which the base facility is in the van, and there is no permanent structure for the byproduct material storage, we will have provisions for the following:

Secured off-street parking under licensee control. Public rights-of-way are not considered part of the address of the client;
Secured storage facilities will be available for storage of byproduct material and radioactive waste should the van be disabled; and
Byproduct material will be delivered (if necessary) directly to the van only if the van is occupied by licensee personnel at the time of delivery.

If a base facility is located in a residential area, we will provide the following information:

Evidence of the requirement for the private residence location rather than for a commercial location.
Documentation of the agreement between the residence owner and the licensee in the event of disagreements between these two entities. It is essential that the mobile service have access to the facility in the event of contamination. Included will be provisions for decontamination of the mobile service van, etc., on the client property, if necessary. Documentation from both parties will illustrate the agreement between the client and the mobile service.
A description of our program demonstrating compliance with 10 CFR 20.1301, "Dose limits for individual members of the public."
Verification that restricted areas shall not contain residential quarters.

Training And Experience

We will require the supervised individual to:

follow the instructions of the supervising authorized user for medical uses of byproduct material;
follow the written radiation protection and written directive procedures established by the licensee; and
comply with the regulations of 10 CFR 35.80, 10 CFR 35.647 (if applicable) and the license conditions with respect to the mobile use of byproduct material.

Training for Individuals Working in or Frequenting Restricted Areas

In addition to the training requirements of 10 CFR 19.12, 10 CFR 35.27, 10 CFR 35.310, 10 CFR 35.410, and 10 CFR 35.610 (as applicable), drivers and technologists (or therapists) will be properly trained in applicable transportation regulations and emergency procedures. The training records for these individuals will include (at a minimum) dates, topics discussed (e.g., DOT regulations, shielding, ALARA, basic radiation protection), attendees, and the instructor's name, and shall be maintained for 3 years for NRC review. Licensees may choose to retain records for longer periods.

Survey Instrument & Dose Measurement Instrument Checks

We will check survey instruments for proper operation with a dedicated check source before use at each address of use. We will check dose measurement instruments (e.g., dose calibrator) as described in 10 CFR 35.60 or 10 CFR 35.62, as applicable, before medical use at each address of use or on each day of use, whichever is more frequent. Additionally, all other transported equipment (e.g., cameras) should be checked for proper function before medical use at each address of use.

Order and Receipt of Byproduct Material

Byproduct material will be delivered by a supplier to the base location or to the client's address, if the client is licensed to receive the type of byproduct material ordered. Delivery of byproduct material to a van that is not occupied by the mobile service personnel will not be permitted.

Alternatively, we will pick up the byproduct material (e.g., radiopharmaceuticals) from the supplier (e.g., nuclear pharmacy) en route to client facilities.

Emergency Procedures

We will develop and implement emergency procedures, in accordance with 10 CFR 20.1101, that, in part, will indicate that the RSO, AU, or a responsible designee, can be physically present at the client's address in response to incidents (e.g., accidents, spills, medical events) that occur at client facilities. We will indicate typical response times of the RSO and AU in the event of an incident. We will develop and implement procedures that include emergency response regarding an accident scenario. An accident is defined as a vehicle collision or other events, such as, wind, water or fire damage that results in damage to exterior or interior portions of the vehicle or the byproduct material used in the mobile service. The transportation emergency response plan will cover both the actions to be taken by the mobile service provider's headquarters emergency response personnel and the "on scene" hazmat trained personnel, and it will be readily available to both transport vehicle personnel and headquarters emergency response contacts. At a minimum, this plan will include:

A 24-hour emergency contact telephone number for the mobile service provider's emergency response personnel.
Emergency contact number for NRC's Operation Center and all appropriate state radiological protection agencies.
Procedures for restricting access to the transport vehicle until surveys have been made to determine if any radiological hazards exist.
Procedures for retrieving and securing any byproduct material, including a sealed source that may become detached and/or dislodged to the extent that a radiological hazard is created. This may require one or more emergency shielded source containers.
Predetermined (calculated) exposure rates for an unshielded therapy source (if applicable) as a function of distance for use in controlling the exposures of emergency response personnel to the maximum extent possible under various emergency response scenarios.
Preplanned decontamination procedures including ready access to all necessary materials.
Procedures to ensure that, following any accident, no patient treatments will occur until all systems pertaining to radiation safety have been tested and confirmed to be operational by the RSO or AMP. If any problem is found, including remote afterloader device interlocks and operation, the remote afterloader device or facility will be repaired and re-certified by the device vendor prior to return to service. In addition, a copy of the report, generated in accordance with 10 CFR 30.50, will be provided to any Class 2 or 3 clients following any accident in which there is actual or possible damage to the facility or device.
A calibrated, operational survey meter should be maintained in the cab of the transporting vehicle. Such a survey meter may be used at an accident scene for conducting surveys.

Note: The type of response should be consistent with the level of the incident. The response may range from phone contact for minor spills, to prompt on-site response (less than 3 hours) to events such as a medical event or lost radioactive material.

Transportation

We will develop procedures to assure that:

Radioactive material is transported in accordance with 49 CFR Parts 170-189. Procedures will include:

use of approved packages
use of approved labeling
conduct of proper surveys
complete and accurate shipping papers
bracing of packages
security provisions

Management (or their designee) will perform audits (at least annually) of transportation documentation (e.g., shipping papers and survey reports) and activities at client facilities.
Licensed material is secured during transport and use at the client's facilities.
Radioactive waste is handled properly during transport. We will describe the method of storage and final disposal.
The transport vehicle including the driver's compartment, if separate, will be secured at all times from any unauthorized access when the vehicle is unattended.

Note: The necessary DOT Type 7A package certification for remote afterloader devices is established via prior approval of the appropriate sealed source and device sheets; however, if the remote afterloader device is damaged in any way during use or transport, then the integrity of the DOT Type 7A packaging may be compromised, and the device must not be used or transported until checked by the vendor and certified as retaining its integrity as a Type 7A package.

Radioactive Waste Management

If waste will be stored in vans, the vans will be properly secured and posted as byproduct material storage locations. We will ensure that the van will be secured against unauthorized access and that the waste storage location will be posted as a byproduct material storage area.

We will develop and implement final waste disposal procedures in accordance with Section 8.44 of this report.

Excreta from individuals undergoing medical diagnosis or therapy with radioactive material may be disposed of without regard to radioactivity if it is discharged into the sanitary sewerage system, in accordance with 10 CFR 20.2003. However, excreta collected from patients in a rest room in a van with a holding tank, is not considered direct disposal into the sanitary sewerage system. If we will provide rest-room facilities on the van for patient use, we will submit the following information for NRC review:

A description of the structure of the tank holding facility and the location of the tank in relation to members of the public, workers on the van, and the driver of the van. A description of procedures to assess the tank for possible leakage. A description, of any ventilation associated with the rest-room, will be provided if any I-131 will be held in the tank.
A description of procedures to ensure doses to occupational workers and members of the public will not exceed the exposure limits in 10 CFR 20.1201 and 20.1301. A description of procedures ensuring that doses to the general public do not exceed 100 mrem/year, that the external surfaces of the van do not exceed 2 mrem/hour, and that doses to members of the public and workers are maintained ALARA. We will include considerations of external dose rates in the rest-room due to the proximity of the holding tank to the toilet.
A description of procedures for emptying and disposing of the contents of the holding tank, including the frequency of disposal, who empties the tank into the sanitary sewer system, and the location of disposal into the sanitary sewer. Precautions taken to minimize contamination in this process will be included.

Mobile Services With Remote Afterloader Devices

In addition to the above procedures addressed in the mobile services section, we will develop and implement the following procedures regarding mobile remote afterloader service operations:

Since the movement of the remote afterloader device from one location to another increases the risk of electro/mechanical component failures or misalignments, it is important that the proper operation of the device be fully checked after each such relocation. Therefore, we will develop and implement the following procedures used to determine if the device is operating properly prior to the commencement of patient treatments:

Safety checks conducted on a remote afterloader device and facility. The procedure will, as a minimum, include the periodic spot checks required by 10 CFR 35.643 (or 10 CFR 35.644) and the additional spot checks required by 10 CFR 35.647, prior to medical use following any relocation of the device. Additionally, the procedure must include provisions for prompt repair of any system identified as not operating properly.
The pretreatment operational function checks after each device move must also include a review of any device alarm/error messages and, if necessary, a resolution of problems indicated by such messages.
Such tests, as indicated above, must be established by the AMP identified on the license authorizing patient treatments. (This would be the mobile service provider in Class 1 and 2 applications and the client in Class 3 applications.)
The applicant must maintain records showing the results of the above safety checks for NRC inspection and review for a period of three years.

For Class 1 and 2 services, we will develop and implement a radiation survey procedure which includes results of radiation surveys conducted prior to initiation of the treatment program (baseline surveys). These baseline surveys must include the source housing, with the source in the shielded position and all areas adjacent to the treatment room with the source in the treatment position. Immediately following any relocation of the remote afterloader device, and prior to patient treatments, the licensee must conduct such surveys, as necessary, with a portable, calibrated survey meter, to assure the radiation dose rates obtained are comparable to those obtained from the baseline surveys. Any significant increase in the dose rates found after relocation from the baseline survey results suggests that the shielding integrity may have been compromised and must be investigated and resolved prior to use of the remote afterloader device.

Appendix W: Transportation

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The major areas in the DOT regulations that are most relevant for transportation of Type A or Type B quantities of licensed material are:

Table of Hazardous Materials and Special Provisions 49 CFR 172.101, and App. A, Table 2: Hazardous materials table, list of hazardous substances, and reportable quantities
Shipping Papers 49 CFR 172.200-204: General entries, description, additional description requirements, shipper's certification
Package Markings 49 CFR 172.300, 49 CFR 172.301, 49 CFR 172.303, 49 CFR 172.304, 49 CFR 172.310, 49 CFR 172.324: General marking requirements for non-bulk packagings, prohibited marking, marking requirements, radioactive material, hazardous substances in non-bulk packaging
Package Labeling 49 CFR 172.400, 49 CFR 172.401, 49 CFR 172.403, 49 CFR 172.406, 49 CFR 172.407, 49 CFR 172.436, 49 CFR 172.438, 49 CFR 172.440: General labeling requirements, prohibited labeling, radioactive materials, placement of labels, specifications for radioactive labels
Placarding of Vehicles 49 CFR 172.500, 49 CFR 172.502, 49 CFR 172.504, 49 CFR 172.506, 49 CFR 172.516, 49 CFR 172.519, 49 CFR 172.556: Applicability, prohibited and permissive placarding, general placarding requirements, providing and affixing placards: highway, visibility and display of placards, RADIOACTIVE placard
Emergency Response Information, Subpart G, 49 CFR 172.600, 49 CFR 172.602, 49 CFR 172.604: Applicability and general requirements, emergency response information, emergency response telephone number
Training, Subpart H, 49 CFR 172.702, 49 CFR 172.704: Applicability and responsibility for training and testing, training requirements
Radiation Protection Program for Shippers and Carriers, Subpart I, 49 CFR 172.800, etc.
Shippers - General Requirements for Shipments and Packaging, Subpart I, 49 CFR 173.403, 49 CFR 173.410, 49 CFR 173.411, 49 CFR 173.412, 49 CFR 173.413, 49 CFR 173.415, 49 CFR 173.416, 49 CFR 173.433, 49 CFR 173.435, 49 CFR 173.441, 49 CFR 173.471, 49 CFR 173.475, 49 CFR 173.476: Definitions, general design requirements, industrial packages, additional design requirements for Type A packages, requirements for Type B packages, authorized Type A packages, authorized Type B packages (including package certification requirements), requirement for determining A1 and A2�, table of A1 and A2 values for radionuclides, radiation level limit, requirements for USNRC-approved packages (Type B), quality control requirements prior to each shipment� approval of special form radioactive materials
Carriage by Public Highway 49 CFR 177.816, 49 CFR 177.817, 49 CFR 177.834(a), 49 CFR 177.842: Driver training, shipping paper, general requirements (secured against movement), Class 7 (radioactive) material.

Appendix X: Model Procedure for Waste Disposal by Decay-In-Storage, Generator Return, and Licensed Material Return

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Model Procedure for Decay-In-Storage

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10 CFR 35.92 describes the requirements for DIS. Short-term storage should be designed to allow for segregation of wastes with different half-lives (e.g., the use of multiple shielded containers). Containers should have shielded covers to maintain occupational exposure at ALARA levels. Storage area must be in a secure location.

We will consider using separate containers for different types of waste, e.g., needles and syringes in one container, other injection paraphernalia such as swabs and gauze in another, and unused dosages in a third container. Because the waste will be surveyed with all shielding removed, the containers in which the waste will be disposed of must not provide any radiation shielding for the material.
When the container is full, we will seal it and attach an identification tag that includes the date sealed, and the longest-lived radionuclide in the container. The container may then be transferred to the DIS area.
Prior to disposal as in-house waste, we will monitor and record the monitoring of each container as follows:

Choose a survey instrument that is appropriate for the type and energy of the radiation being measured.
Check the radiation detection survey meter for proper operation.
Plan to monitor in a low-level radiation (< 0.05 millirem per hour) area away from all sources of radioactive material, if possible.
Remove any shielding from around the container or generator column.
Monitor, at contact, all surfaces of each individual container.
Remove or deface any radioactive material labels (unless the containers will be immediately incinerated and instruction has been provided to the incinerator operator regarding radioactive labels and potential hazards)⁽¹¹⁾.
Discard as in-house waste only those containers that cannot be distinguished from background. Record the disposal date, the radionuclides disposed, the survey instrument used, the background dose rate, the dose rate measured at the surface of each waste container, and the name of the individual who performed the disposal.
Containers that can be distinguished from background radiation levels must be returned to the storage area for further decay or transferred to an authorized byproduct material recipient.

Model Procedure for Returning Generators to the Manufacturer

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Used Mo/Tc-99m generators may be returned to the manufacturer. This permission does not relieve licensees from the requirement to comply with 10 CFR Part 71 and DOT regulations. We will perform the following when returning generators:

Retain the records needed to demonstrate that the package qualifies as a DOT Specification 7A container (see DOT regulations, paragraph 173.415 (a) of 49 CFR Part 173).
Assemble the package in accordance with the manufacturer's instructions.
Perform the dose rate and removable contamination measurements required by paragraph 173.475 (i) of 49 CFR Part 173.
Label the package and complete the shipping papers in accordance with the manufacturer's instructions.
Retain records of receipts and transfers in accordance with 10 CFR 30.51.

Model Procedure for Return of Licensed Material to Authorized Recipients

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We will perform the following when returning licensed material to authorized recipients:

In accordance with 10 CFR 30.41(a)(5), confirm that persons are authorized to receive byproduct material prior to transfer (e.g., obtain a copy of the transferee's NRC license or Agreement State license that authorizes the byproduct material).
Retain the records needed to demonstrate that the package qualifies as a DOT Specification 7A container (see DOT regulations, paragraph 173.415 (a) of 49 CFR Part 173).
Assemble the package in accordance with the manufacturer's instructions.
Perform the dose rate and removable contamination measurements required by paragraph 173.475 (i) of 49 CFR Part 173.
Label the package and complete the shipping papers in accordance with the manufacturer's instructions.
Retain records of receipts and transfers in accordance to with 10 CFR 30.51.

Appendix Y: NRC Form 314

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NRC Form 314 - Certificate of Disposition of Materials

NRC Form 314 - Front

NRC Form 314 - Back

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1. See "Amendments and Renewals to a License" later in this document. Licensees may request an amendment to an existing license to add authorization for other uses of byproduct material. As described in 10 CFR 35.12(c), a separate application and license is required for the addition of 10 CFR 35.600 uses, with the exception of remote afterloaders.

2. The requirement for an examination and the limitation that an AU can only be named the RSO for the types of use for which the individual has training and experience will become effective two years after the publication of the final rule (10 CFR Part 35).

3. See the Notice of Availability (inside front cover of this draft report) to obtain copies of RG 3.66, "Standard Format and Content of Financial Assurance Mechanisms Required for Decommissioning Under 10 CFR Parts 30, 40, 70, and 72," (dated June 1990).

4. Co-57 and Ra-226 are not subject to NRC licensing; the appropriate State agency should be consulted to determine its requirements for possessing this material.

5. For ease of use, the examples in this Appendix use conventional units. The conversions to SI units are as follows: 1 foot (ft) = 0.305 meter; 1 mrem = 0.01 mSv.

6. NCRP Report No. 49, " Structural Shielding Design and Evaluation for Medical Use of X Rays and Gamma Rays of Energies Up to 10 MeV," contains helpful information. It is available from the National Council on Radiation Protection and Measurements, 7910 Woodmont Avenue, Suite 800, Bethesda, Maryland 20814. NCRP's telephone numbers are: (301) 657-2652 or 1-800-229-2652.

7. TLDs used for personnel monitoring (e.g., LiF) may not have sufficient sensitivity for this purpose. Generally, the minimum reportable dose received is 0.1 mSv (10 mrem). Suppose a TLD monitors dose received and is changed once a month. If the measurements are at the minimum reportable level, the annual dose received could have been about 1.2 mSv (120 mrem), a value in excess of the 1 mSv/year (100 mrem/year) limit. If licensees use TLDs to evaluate compliance with the public dose limits, they should consult with their TLD supplier and choose more sensitive TLDs, such as those containing CaF2 that are used for environmental monitoring.

8. Labeled with a Radioactive White I, Yellow II, or Yellow III label as specified in DOT regulations, 49 CFR 172.403 and 172.436-440.

9. The commercial telephone number of the NRC Operations Center is (301) 951-0550.

10. The NRC does not intend to enforce patient compliance with the instructions nor is it the licensee's responsibility to do so.

11. Note: Contact the appropriate NRC Regional Office for guidance in this area