United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Relating To Categorization Of Recent Regulatory Guides By The Regulatory Requirements Review Committee (Generic Letter 79-05)


GL79005 

                              UNITED STATES 
                      NUCLEAR REGULATORY COMMISSION 
                         WASHINGTON, D. C. 20555 

                             January 26, 1979 

Docket No.:    50-471 

     Mr. R. M. Butler 
     Nuclear Projects Manager 
     Boston Edison Company 
     800 Boylston Street 
     Boston, Massachusetts 02199 

     Dear Mr. Butler: 

     SUBJECT:  INFORMATION RELATING TO CATEGORIZATION OF RECENT REGULATORY 
               GUIDES BY THE REGULATORY REQUIREMENTS REVIEW COMMITTEE 
               PILGRIM STATION, UNIT 2 

     We have recently advised utilities with plants in the post-CP phase of 
     the reactor licensing process of the status of NRC staff review and use
     of recently-approved regulatory guides, and have indicated how these 
     guides would be used in the Operating License review of their Final 
     Safety Analysis Reports. Such information, while not directly 
     applicable to you at this time, may nonetheless be useful to you for 
     your future planning. The text of our letter to these utilities is the 
     following: 
     
          "SUBJECT: IMPLEMENTATION OF STAFF REVIEW REQUIREMENTS - (Name of 
                    Plant) - OPERATING LICENSE REVIEW 

          During the last several years, we have reviewed and approved 
          several new regulatory guides and branch technical positions or 
          other modifications to existing staff positions. Our practice is 
          that substantive changes in staff positions be considered by the 
          NRC's Regulatory Requirements Review Committee (RRRC) which then 
          recommends a course of action to the Director, Office of Nuclear 
          Reactor Regulation (NRR). The recommended action includes an 
          implementation schedule. The Director's approval then is used by 
          the NRR staff as review guidance on individual licensing matters. 
          Some of these actions will affect your application. This letter is
          intended to bring you up to date on these changes in staff 
          positions so that you may consider them in your Final Safety 
          Analysis Report (FSAR) preparation. 
.

                                  -2- 

     "The RRRC applies a categorization nomenclature to each of its actions.
     (A copy of the summary of RRRC Meeting No. 31 concerning this 
     categorization is attached as Enclosure 1.) Category 1 matters are 
     those to be applied to applications in accordance with the 
     implementation section of the published guide. We have enclosed lists 
     of actions which are either Category 2 or Category 3, which are defined 
     as follows: 
     
     Category 2:    A new position whose applicability is to be determined 
                    on a case-by-case basis. You should describe the extent 
                    to which your design conforms, or you should describe an
                    acceptable alternate, or you should demonstrate why 
                    conformance is not necessary. 

     Category 3:    Conformance or an acceptable alternative is required. If
                    you do not conform, or do not have an acceptable 
                    alternate, then staff-approved design revisions will be 
                    required. 

     "We believe that providing you with a list of the Category 2 and 3 
     matters approved to date will be useful in your FSAR preparation, and 
     they will be an essential part of our operating license review. 
     Enclosure 2 is a list of the Category 2 matters. Enclosure 3 is a list 
     of the Category 3 matters. 

     "In addition to the RRRC categories, there also exists an NRR Category 
     4 list which are those matters not yet reviewed by the RRRC, but which 
     the Director, NRR,, has deemed to have sufficient attributes to warrant
     their being addressed and considered in ongoing reviews. These Matters 
     will be treated like Category 2 matters until such time as they are 
     reviewed by the RRRC, and a definite implementation program is 
     developed. A current list of Category 4 matters is attached (Enclosure 
     4). These also should be considered in your FSAR. 

     "In some instances the items in the enclosures may not be applicable to
     your application. Also, we recognize that your application may, in some
     instances, already conform to the stated staff positions, in your FSAR 
     you should note such compliance. 

     "If you have any questions please let us know." 
.

                                  -3- 

     For your information, I am enclosing a set of the enclosures that 
     accompanied these individual letters. These enclosures list the present
     Category 1-4 matters discussed in the letter. 

                                             Sincerely, 


                                             Roger S. Boyd, Director 
                                             Division of Project Management 
                                             Office of Nuclear Reactor 
                                             Regulation 

Enclosures: 
As stated 

cc: See next page 
.

Mr. R. M. Butler 

cc:  Mr. William Griffin                Charles Corkin, II, Esq. 
     Project Engineer                   Assistant Attorney General 
     Boston Edison Company              Commonwealth of Massachusetts 
     800 Boylston Street                One Ashburton Place, 19th Floor 
     Boston, Massachusetts 02199        Boston, Massachusetts 03105 

     Dale G. Stoodley, Counsel          Henry Herrman, Esq. 
     Boston Edison Company              151 Tremont Street, 27K 
     800 Boylston Street                Boston, Massachusetts 02111 
     Boston, Massachusetts 02199  
                                        Mr. & Mrs. Alan R. Cleeton 
     George H. Lewald, Esq.             22 Mackintosh Street 
     Ropes & Gray                       Franklin, Massachusetts 02038 
     225 Franklin Street 
     Boston, Massachusetts 02110        W. M. Sides  
                                        Quality Assurance Manager 
     William S. Abbott                  Boston Edison Company 
     Attorney & Counsellor at Law       800 Boylston St. 
     50 Congress Street, Suite 925      Boston, Massachusetts 02199 
     Boston, Massachusetts 02109  
                                        Mr. R. A. Fortney 
     B. N. Pushek                       EDS Nuclear 
     Bechtel Power Corp.                220 Montgomery Street 
     P. O. Box 3695                     San Francisco, California 94104 
     San Francisco, California 94119  
                                        Edward Luton, Esq., Chairman 
     John D. Fassett                    Atomic Safety and Licensing Board 
     Vice President and General         U. S. Nuclear Regulatory Commission 
          Counsel                       Washington, D. C. 20555 
     United Illuminating Company 
     80 Temple Street                   Dr. Dixon Callihan 
     New Haven, Connecticut 06506       Union Carbide Corporation  
                                        P. O. Box Y 
     Mr. R. Newman                      Oak Ridge, Tennessee 37830 
     Combustion Engineering, Corp. 
     1000 Prospect Hill Road            Dr. Richard F. Cole 
     Windsor, Connecticut 06095         Atomic Safety and Licensing Board 
                                        U. S. Nuclear Regulatory Commission 
     W. C. Tallman, President           Washington, D. C. 20555 
     Public Service Company of 
          New Hampshire  
     1000 Elm Street 
     Manchester, New Hampshire 03105 
.

Mr. R. M. Butler 

cc:  Richard S. Salzman, Esq., Chairman 
     Atomic Safety and Licensing Appeal Board 
     U. S. Nuclear Regulatory Commission 
     Washington, D. C. 20555 

     Dr. John H. Buck 
     Atomic Safety and Licensing Appeal Board 
     U. S. Nuclear Regulatory Commission 
     Washington, D. C. 20555 

     Michael C. Farrar, Esq. 
     Atomic Safety and Licensing Appeal Board 
     U. S. Nuclear Regulatory Commission 
     Washington, D. C. 20555 
.

                             UNITED STATES 
                       NUCLEAR REGULATORY COMMISSION 
                         WASHINGTON, D. C. 20555  

                                SEP 24 1975 

Lee V. Gossick 
Executive Director for Operations 

REGULATORY REQUIREMENTS REVIEW COMMITTEE MEETING NO. 31, 
JULY 11, 1975 

1.   The Committee discussed issues related to the implementation of 
     Regulatory Guides on existing plants and the concerns expressed in the 
     June 24, 1974 memorandum, A. Giambusso to E. G. Case, subject: 
     REGULATORY GUIDE IMPLEMENTATION, and made the following recommendations
     and observations: 

     a.   Approval of new Regulatory Guides and approval of revisions of 
          existing guides should move forward expeditiously in order that 
          the provisions of these regulatory guides be available for use as 
          soon as possible in on-going or future staff reviews of license 
          applications. The Committee noted that over the recent past, the 
          approval of proposed regulatory guides whose content is acceptable
          for these purposes has experienced significant delays in RRRC 
          review pending the determination of the applicability of the guide
          to existing plants, often requiring significant staff effort. To 
          avoid these delays, the Committee concluded that, henceforth, 
          approval of proposed regulatory guides should be uncoupled from 
          the consideration of their backfit applicability. 

     b.   The implementation section of new regulatory guides should 
          address, in general, only the applicability of the guide to 
          applications in the licensing review process using, in so far as 
          possible, a standard approach of applying the guide to those 
          applications docketed 8 months after the issuance date of the 
          guide for comment. Exceptions to this general approach will be 
          handled on a case-by-case basis. 
          
     c.   The regulatory position of each approved proposed guide (or 
          proposed guide revision) will be characterized by the Committee as
          to its backfitting potential, by placing it in one of three 
          categories: 

          Category 1 - Clearly forward fit only. No further staff 
          consideration of possible backfitting is required. 

                                                               ENCLOSURE 1 
.

                                   -2- 

          Category 2 - Further staff consideration of the need for 
          backfitting appears to be required for certain identified items of
          the regulatory position--these individual issues are such that 
          existing plants need to be evaluated to determine their status 
          with regard to these safety issues in order to determine the need 
          for backfitting. 

          Category 3 - Clearly backfit. Existing plants should be evaluated 
          to determine whether identified items of the regulatory position 
          are resolved in accordance with the guide or by some equivalent 
          alternative. 

          From time to time, for a specific guide, there will probably be 
          some variation among these categories or even within a category, 
          and these three broad category characterizations will be qualified
          as required to meet a particular situation. 

     d.   It is not intended that the Committee categorization appear in the
          guide itself. The purpose of the categorization is to indicate 
          those items of the regulatory position for which the Committee can
          make a specific backfit recommendation without additional staff 
          work (Categories 1 and 3), and to indicate those items for which 
          additional staff work is required in order to determine backfit 
          considerations (Category 2). 

     e.   The Committee recommends that for approved guides in Category 2, 
          staff efforts be initiated in parallel with the process leading to
          publication of the guide in order that specific backfit 
          requirements for existing plants be determined within a reasonable
          period of time after publication of the guide. 

     f.   The Committee observed that more attention needs to be given to 
          the identification of acceptable alternatives to the positions 
          outlined in the guides in order to provide additional options and 
          flexibility to applicants and licensees, with the possible 
          benefits of additional innovation and exploration in the solution 
          of safety issues. 

2.   The Committee reviewed the proposed Regulatory Guide 1.XX: THERMAL 
     OVERLOAD PROTECTION FOR MOTORS ON MOTOR-OPERATED VALVES and recommended
     approval. This guide was characterized by the Committee as Category 1 
     - no backfitting, with the stipulation that as an appropriate occasion 
     presented itself in conjunction with the review of some particular 
     aspect of existing plants, the thermal overload protection provisions 
     be audited. 

                                                     ENCLOSURE  1  (CONT'D) 
.

                                   -3-

3.   The Committee reviewed the proposed Regulatory Guide I.XX: INSTRUMENT 
     SPANS AND SETPOINTS and recommended approval subject to the following 
     comment:       

          Paragraph 5 of Section C (page 4 of the proposed Guide) should be 
          reworded in light of Committee comments, to the satisfaction of 
          the Director, Office of Standards  Development. This guide was 
          characterized by the Committee as Category 1 - no backfit. 

4.   The Committee reviewed Proposed Regulatory Guide 1.97: INSTRUMENTATION 
     FOR LIGHT WATER COOLED NUCLEAR POWER PLANTS TO ASSESS PLANT CONDITIONS 
     DURING AND FOLLOWING AN ACCIDENT  and deferred further consideration to
     a later meeting in order to permit incorporation of recent comments by 
     the Division of Technical Review. 


                                        Edson G. Case, Chairman 
                                        Regulatory Requirements Review 
                                             Committee 

                                                       ENCLOSURE 1 (CONT'D) 
.

                                                       September 15, 1978 

                            CATEGORY 2 MATTERS 

Document 
Number         Revision       Date           Title 

RG 1.27          2            1/76           Ultimate Heat Sink for Nuclear 
                                             Power Plants 

RG 1.52          1            7/76           Design, Testing, and 
Maintenance 
                                             Criteria for Engineered-Safety-
                                             Feature Atmosphere Cleanup 
                                             System Air Filtration and 
                                             Adsorption Units of Light Water
                                             Cooled Nuclear Power Plants 
                                             (Revision 2 has been published 
                                             but the changes from Revision 
                                             1 to Revision 2 may, but need 
                                             not, be considered. 

RG 1.59          2            8/77           Design Basis Floods for Nuclear 
                                             Power Plants 

RG 1.63          2            7/78           Electric Penetration Assemblies
                                             in Containment Structures for 
                                             Light Water Cooled. Nuclear 
                                             Power Plants 

RG 1.91          1            2/78           Evaluation of Explosions 
                                             Postulated to Occur on 
                                             Transportation Routes Near 
                                             Nuclear Power Plant Sites 

RG 1.102         1            9/76           Flood Protection for Nuclear 
                                             Power Plants 

RG 1.105         1            11/76          Instrument Setpoints 

RG 1.108         1            8/77           Periodic Testing of Diesel 
                                             Generator Units Used as Onsite 
                                             Electric Power Systems at 
                                             Nuclear  Power Plants 

RG 1.115         1            7/77           Protection Against 
                                             Low-Trajectory  Turbine 
                                             Missiles 

RG 1.117         1            4/78           Tornado Design Classification 
                                             
RG 1.124         1            1/78           Service Limits and Loading 
                                             Combinations for Class 1 Linear 
                                             Type Component Supports 

RG 1.130         0            7/77           Design Limits and Loading 
                                             Combinations for Class 1 Plate- 
                                             and Shell-Type-Component 
                                             Supports 

                               (Continued)  

                                                                ENCLOSURE 2 
.

                      CATEGORY 2 MATTERS (CONT'D) 

Continued 

Document 
Number         Revision       Date           Title 

RG 1.137          0           1/78           Fuel Oil Systems for Standby 
                                             Diesel Generators  (Paragraph 
                                             C.2) 

RG 8.8            2           3/77           Information Relevant to 
                                             Ensuring that Occupational  
                                             Radiation Exposures at Nuclear 
                                             Power Stations Will be as Low 
                                             as is Reasonably, Achievable 
                                             (Nuclear Power Reactors) 
                                             
BTP ASB                                      Guidelines for Fire Protection 
9.5-1             1                          for Nuclear Power Plants (See 
                                             Implementation  Section, 
                                             Section D)  

BTP MTEB 5-7                  4/77           Material Selection and 
                                             Processing  Guidelines for BWR 
                                             Coolant Pressure  Boundary 
                                             Piping   

RG 1.141          0           4/78           Containment Isolation 
                                             Provisions for Fluid Systems 

                                   -2-  

                                                       ENCLOSURE 2 (CONT'D) 
.

                                                       September 15, 1973 

                            CATEGORY 3 MATTERS 

Document 
Number         Revision       Date           Title 

RG 1.99          1            4/77           Effects of Residual Elements on
                                             Predicted Radiation Damage to 
                                             Reactor Vessel materials 
                                             (Paragraphs C.1 and C.2. 

RG 1.101         1            3/77           Emergency Planning for Nuclear 
                                             Power Plants 

RG 1.114         1            11/76          Guidance on Being Operator at 
                                             the Controls of a  Nuclear 
                                             Power Plant 

RG 1.121         0            8/76           Bases for Plugging Degraded PWR
                                             Steam Generator Tubes 

RG 1.127         1            3/78           Inspection of Water-Control 
                                             Structures Associated  with 
                                             Nuclear Power Plants 

RSB 5-1          1            1/78           Branch Technical Position: 
                                             Design Requirements of  the 
                                             Residual Heat Removal System 
RSB 5-2          0            3/73           Branch Technical Position: 
                                             Reactor Coolant System 
                                             Overpressurization Protection 
                                             (Draft copy attached) 

RG 1.97          1            8/77           Instrumentation for Light Water
                                             Cooled Nuclear Power Plants to 
                                             Assess Plant Conditions During 
                                             and Following an Accident 
                                             (Paragraph C.3 - with 
                                             additional guidance on 
                                             paragraph C.3.d to be provided 
                                             later) 

RG 1.68.2        1            7/78           Initial Startup Test Program to
                                             Demonstrate Remote Shutdown 
                                             Capability for Water-Cooled 
                                             Nuclear Power Plants 

RG 1.56          1            7/78           Maintenance of Water Purity in 
                                             Boiling Water Reactors 

Attachment: 
BTP RSB 5-2 (Draft) 

                                                                ENCLOSURE 3 
.

                    BRANCH TECHNICAL POSITION RSB 5-2 
        OVERPRESSURIZATION PROTECTION OF PRESSURIZED WATER REACTORS 
                   WHILE OPERATING AT LOW TEMPERATURES 

A.   Background 

     General Design Criterion 15 of Appendix A. 10 CFR 50, requires that 
     "the Reactor Coolant System and associated auxiliary, control, and 
     protection systems shall be designed with sufficient margin to assure 
     that the design conditions of the reactor coolant pressure boundary are
     not exceeded during any condition of normal operation, including 
     anticipated operational occurrences." 

     Anticipated operational occurrences, as defined in Appendix A of 10 CFR
     50, are "those conditions of normal operation which are expected to 
     occur one or more times during the life of the nuclear power unit and 
     include but are not limited to loss of power to all recirculation 
     pumps, tripping of the turbine generator set, isolation of the main 
     condenser, and loss of all offsite power. 

     Appendix G of 10 CFR 50 provides the fracture toughness requirements 
     for reactor pressure vessels under all conditions. To assure that the 
     Appendix G limits of the reactor coolant pressure boundary are not 
     exceeded during any anticipated operational occurrences, Technical 
     Specification pressure-temperature limits are provided for operating 
     the plant. 
     
     The primary concern of this position is that during startup and 
     shutdown conditions at low temperature, especially in a water-solid 
     condition, the reactor coolant system pressure might exceed the reactor 
     vessel pressure-temperature limitations in the Technical Specifications 
     established for protection against brittle fracture. This inadvertent 
     overpressurization could be generated by any one of a variety of mal-
     functions or operator errors. Many incidents have occurred in operating
     plants as described in Reference 1. 
     
     Additional discussion on the background of this position is contained 
     in Reference 1. 

                                                              ENCL 3 (CONT) 
.

                                   -2-

B.   Branch Position 

     1.   A system should be designed and installed which will prevent 
          exceeding the applicable Technical Specifications and Appendix G 
          limits for the reactor coolant system while operation at low 
          temperatures. The system should be capable of relieving pressure 
          during all anticipated overpressurization events at a rate 
          sufficient to satisfy the Technical Specification limits, 
          particularly while the reactor coolant system is in a water-solid 
          condition. 

     2.   The system must be able to perform its function assuming any 
          single active component failure. Analyses using appropriate 
          calculational techniques must be provided which demonstrate that 
          the system will provide the required pressure relief capacity 
          assuming the most limiting single active failure. The cause for 
          initiation of the event, e.g., operator error, component 
          malfunction, will not be considered as the single active failure. 
          The analysis should assume the most limiting allowable operating 
          conditions and systems configuration at the time of the postulated 
          cause of the overpressure event. All potential overpressurization 
          events must be considered when establishing the worst case event. 
          Some events may be prevented by protective interlocks or by 
          locking out power. These events should be reviewed on an 
          individual basis. If the interlock/power lockout is acceptable, it 
          can be excluded from the analyses provided the controls to prevent 
          the event are in the plant Technical Specifications. 
          
     3.   The system must meet the design requirements of IEEE 279 (see 
          Implementation). The system may be manually enabled, however, the 
          electrical instrumentation and control system must provide alarms 
          to alert the operator to: 

          a.   properly enable the system at the correct plant condition 
               during cooldown, 

          b.   indicate if a pressure transient is occurring. 

     4.   To assure operational readiness, the overpressure protection 
          system must be tested in the following manner: 

          a.   A test must be performed to assure operability of the system 
               electronics prior to each shutdown. 

          b.   A test for valve operability must as a minimum be conducted 
               as specified in the ASME Code Section XI. 

          c.   Subsequent to system, valve, or electronics maintenance, a 
               test on that portion(s) of the system must be performed prior
               to declaring the system operational. 

                                                              ENCL 3 (CONT) 
.

                                   -3-

     5.   The system must meet the design requirements of regulatory Guide 
          1.26, "Quality Group Classifications and Standards for Water-, 
          Steam-, and Radioactive-Waste-Containing Components of Nuclear 
          Power Plants" and Section III of the ASME Code. 


     6.   The overpressure protection system must be designed to function 
          during an Operating Basis Earthquake. It must not compromise the 
          design criteria of any other safety-grade system with which it 
          would interface, such that the requirements of Regulatory Guide 
          1.29, "Seismic Design Classification" are met. 

     7.   The overpressure protection system must not depend on the 
          availability of offsite power to perform its function. 

     8.   Overpressure protection systems which take credit for an active 
          component(s) to mitigate the consequences of an overpressurization
          event must include additional analyses considering inadvertent 
          system initiation/actuation or provide justification to show that 
          existing analyses bound such an event. 

C.   Implementation 

     The Branch Technical Position, as specified in Section B, will be used 
     in the review of all Preliminary Design Approval (PDA), Final Design 
     Approval (FDA), Manufacturing License (ML), Operating License (OL), and
     Construction Permit (CP) applications involving plant designs 
     incorporating pressurized water reactors. All aspects of the position 
     will be applicable to all applications, including CP applications 
     utilizing the replication option of the Commission's standardization 
     program, that are docketed after March 14, 1978. All aspects of the 
     position, with the exception of reasonable and justified deviations 
     from IEEE 279 requirements, will be applicable to CP, OL, ML, PDA, and 
     FDA applications docketed prior to March 14, 1978 but for which the 
     licensing action has not been completed as of March 14, 1978. Holders 
     of appropriate PDA's will be informed by letter that all aspects of the
     position with the exception of IEEE 279 will be applicable to their 
     approved standard designs and that such designs should be modified, as 
     necessary, to conform to the position. Staff approval of proposed 
     modifications can be applied for either by application by the 
     PDA-holder on the PDA-docket or by each CP applicant referencing the 
     standard design on its docket. 
     
     The following guidelines may be used, if necessary, to alleviate 
     impacts on licensing schedules for plants involved in licensing 
     proceedings nearing completion on March 14, 1978:   

                                                              ENCL 3 (CONT) 
.

                                  -4- 

     1.   Those applicants issued an OL during the Period between March 14, 
          1978 and a date 12 months thereafter may merely commit to meeting 
          the position prior to OL issuance but shall, by license condition,
          be required to install all required staff-approved modifications 
          prior to plant startup following the first scheduled refueling 
          outage. 

     2.   Those applicants issued an OL beyond March 14, 1979 shall install 
          all required staff-approved modifications prior to initial plant 
          startup. 

     3.   Those applicants issued a CP, PDA, or ML during the period between
          March 14, 1978 and a date 6 months thereafter may merely commit to
          meeting the position but shall, by license condition, be required 
          to amend the application, within 6 months of the date of issuance 
          of the CP, PDA, or ML, to include a description of the proposed 
          modifications and the bases for their design, and a request for 
          staff approval. 

     4.   Those applicants issued a CP, PDA, or ML after September 14, 1978 
          shall have staff approval of proposed modifications prior to 
          issuance of the CP, PDA. or ML. 

D.   References 

     1.   NUREG-0138, Staff Discussion of Fifteen Technical Issues Listed in
          Attachment to November 3, 1976 Memorandum from Director, NRR, to 
          NRR Staff. 

                                                              ENCL 3 (CONT) 
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