EA-05-103 - LaSalle 1 and 2 (Exelon Nuclear)
September 7, 2005
Mr. Christopher M. Crane
President and Chief Nuclear Officer
Exelon Generation Company, LLC
4300 Winfield Road
Warrenville, IL 60555
FINAL SIGNIFICANCE DETERMINATION FOR A WHITE
FINDING AND NOTICE OF VIOLATION (NRC INSPECTION REPORT 05000373/2005010;
05000374/2005010), LASALLE COUNTY STATION, UNITS 1 AND 2
Dear Mr. Crane:
The purpose of this letter is to provide you with the final results of our significance determination of the preliminary White finding identified in the subject inspection report issued June 22, 2005. This finding was assessed using the Significance Determination Process (SDP) and was preliminarily characterized as White (i.e., a finding with low to moderate increased importance to safety, which may require additional NRC inspection). The White finding involved a single point vulnerability that could result in a loss of all onsite and offsite power sources to both 4160 Vac Division 1 and Division 2 safety-related buses at either of your LaSalle County Station units.
In our letter dated June 22, 2005, the Nuclear Regulatory Commission (NRC) provided Exelon Nuclear with an opportunity to address the White finding documented in the inspection report by either requesting a Regulatory Conference or by providing a written response before we made our final risk significance determination. On July 7, 2005, in a telephone conversation between the LaSalle County Station Plant Manager, Mr. Daniel Enright, and Mr. Bruce Burgess of the NRC Region III Division of Reactor Projects, you informed us that you did not intend to request a Regulatory Conference, and did not intend to provide a written response.
After considering the information developed during the inspection, and in the absence of any additional information provided by you, the NRC has concluded that the inspection finding was appropriately characterized as White (i.e., an issue with low to moderate safety significance which may require additional NRC inspection).
You have 30 calendar days from the date of this letter to appeal the staff’s determination of significance for the identified White finding. Such appeals will be considered to have merit only if they meet the criteria given in NRC Inspection Manual Chapter (IMC) 0609, Attachment 2. Appeals to reduce the significance of an inspection finding will be considered as having sufficient merit for review by this appeal process only if the contention falls into one of the following categories: (1) actual (verifiable) plant hardware, procedures, or equipment configurations were not considered by the staff; or (2) the staff’s significance determination process was inconsistent with the applicable SDP guidance or lacked justification.
The NRC has also determined that the single point vulnerability within your offsite power transformer circuits and the associated failure to assure that applicable regulatory requirements and the design basis for safety-related systems were correctly maintained and controlled commensurate with the standards applied to the original design is a violation of 10 CFR Part 50, Appendix B, Criterion III, as cited in the enclosed Notice of Violation (Notice). The circumstances surrounding the violation are described in detail in the subject inspection report.
You are required to respond to this letter and should follow the instructions specified in the enclosed Notice when preparing your response.
Because the finding involved an issue with low to moderate safety significance (White), the NRC would normally use the Reactor Assessment Program Action Matrix to determine the appropriate NRC response to the finding. However, the NRC may also exercise discretion and refrain from considering a safety significant finding in the assessment program if the finding involves design-related engineering calculations or analysis, associated operating procedures, or the installation of plant equipment. In addition, the finding must have been: (1) licensee- identified as a result of a voluntary initiative such as a design basis reconstitution; (2) corrected, or will be corrected, to include immediate corrective action and long term comprehensive corrective action to prevent recurrence, within a reasonable time following identification; (3) unlikely to have been previously identified by recent ongoing licensee efforts such as normal surveillance, quality assurance activities, or evaluation of industry information; and (4) not reflective of a current performance deficiency associated with existing licensee programs, policies, or procedures. In these cases, the NRC may characterize the finding as an “old design issue.” A finding determined to be appropriately characterized as an “old design issue” will not be aggregated in the NRC Action Matrix with other performance indicators and inspection findings, nor will the finding individually result in a change from one column to another in the Reactor Assessment Program Action Matrix.
As documented in the subject inspection report, the NRC previously determined that: 1) the inspection finding should be considered licensee-identified as a result of the licensee’s immediate review of an operating events experience; 2) the licensee took both immediate and long-term corrective actions to address the inspection finding; 3) the licensee’s normal quality assurance and surveillance activities were not likely to have identified the vulnerability associated with the inspection finding; and 4) the performance errors that caused the inspection finding were not reflective of the licensee’s existing programs, policies, or procedures. Therefore, based upon a consideration of the facts described above and in the subject inspection report, the NRC has determined that the inspection finding should be characterized as an “old design issue” and the NRC is exercising discretion to not consider the finding as a part of the Reactor Assessment Program. However, consistent with the guidance in IMC 0305, the NRC is considering an inspection, such as a supplemental inspection in accordance with NRC Inspection Procedure 95001, to review your root cause evaluation and corrective actions for the finding. We will notify you, by separate correspondence, of that determination.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a
copy of this letter, its enclosure, and your response will be made
available electronically for public inspection in the NRC Public
Document Room or from the Publicly Available Records (PARs) component
of the NRC’s document system (ADAMS), accessible from the NRC
Web site at http://www.nrc.gov/reading-rm/adams.html (The Public
NRC Library). To the
extent possible, your response should not include any personal privacy, proprietary, or safeguards information so that it can be made available to the public without redaction. The NRC also includes significant enforcement actions on its Web site at www.nrc.gov; select What We Do, Enforcement, then Significant Enforcement Actions.
|/RA by Geoffrey Grant Acting for/
|James L. Caldwell
Docket Nos. 50-373; 50-374
License Nos. NPF-11; NPF-18
Enclosure: Notice of Violation
Site Vice President - LaSalle County Station
LaSalle County Station Plant Manager
Regulatory Assurance Manager - LaSalle County Station
Chief Operating Officer
Senior Vice President - Nuclear Services
Senior Vice President - Mid-West Regional
Vice President - Mid-West Operations Support
Vice President - Licensing and Regulatory Affairs
Director Licensing - Mid-West Regional
Manager Licensing - Clinton and LaSalle
Senior Counsel, Nuclear, Mid-West Regional
Document Control Desk - Licensing
Assistant Attorney General
Illinois Department of Nuclear Safety
State Liaison Officer
Chairman, Illinois Commerce Commission
NOTICE OF VIOLATION
Exelon Generation Company, LLC
LaSalle County Station, Units 1 and 2
|Docket Nos. 50-373; 50-374
License No. NPF-11; NPF-18
During an NRC inspection conducted from February 1 through May 31, 2005, a violation of NRC requirements was identified. In accordance with the NRC Enforcement Policy, the violation is listed below:
Title 10 CFR Part 50, Appendix B, Criterion III, (Design Control) requires, in part, that design changes, including field changes, shall be subject to design control measures commensurate with those applied to the original design.
10 CFR 50, Appendix A, General Design Criterion 17, (Electric Power Systems) requires, in part, that onsite electric power supplies, including the onsite electric distribution system, shall have sufficient independence, redundancy, and testability to perform their safety functions assuming a single failure.
Contrary to the above, the licensee made modifications to the emergency diesel generator (EDG) output circuit breakers that were completed on December 21, 1988, for Unit 2, Division 1; September 26, 1989, for Unit 1, Division 1; March 8, 1991, for Unit 1, Division 2; and February 1, 1992, for Unit 2, Division 2 that were not subject to design control measures commensurate with those applied to the original design. Specifically, the modifications introduced a single failure vulnerability such that a failure (i.e., open circuit) of the common current transformer circuit would have resulted in a loss of all alternating current, including the EDG supplied feeds, for the Division 1 and Division 2 safety buses on both units.
This violation is associated with a White Significance Determination Process finding.
Pursuant to the provisions of 10 CFR 2.201, Exelon Nuclear is hereby required to submit a written statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555 with a copy to the Regional Administrator, Region III, and a copy to the NRC Resident Inspector at the LaSalle County Station, within 30 days of the date of the letter transmitting this Notice of Violation (Notice). This reply should be clearly marked as a "Reply to a Notice of Violation; EA-05-103" and should include for each violation: (1) the reason for the violation, or, if contested, the basis for disputing the violation or severity level, (2) the corrective steps that have been taken and the results achieved, (3) the corrective steps that will be taken to avoid further violations, and (4) the date when full compliance will be achieved. Your response may reference or include previous docketed correspondence, if the correspondence adequately addresses the required response. If an adequate reply is not received within the time specified in this Notice, an order or a Demand for Information may be issued as to why the license should not be modified, suspended, or revoked, or why such other action as may be proper should not be taken. Where good cause is shown, consideration will be given to extending the response time.If you contest this enforcement action, you should also provide a copy of your response, with the basis for your denial, to the Director, Office of Enforcement, U. S. Nuclear Regulatory Commission, Washington, DC 20555-0001.
Because your response will be made available electronically for public inspection in the NRC Public Document Room or from the NRC’s document system (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams/html, to the extent possible, it should not include any personal privacy, proprietary, or safeguards information so that it can be made available to the public without redaction. If personal privacy or proprietary information is necessary to provide an acceptable response, then please provide a bracketed copy of your response that identifies the information that should be protected and a redacted copy of your response that deletes such information. If you request withholding of such material, you must specifically identify the portions of your response that you seek to have withheld and provide in detail the bases for your claim of withholding (e.g., explain why the disclosure of information will create an unwarranted invasion of personal privacy or provide the information required by 10 CFR 2.390(b) to support a request for withholding confidential commercial or financial information). If safeguards information is necessary to provide an acceptable response, please provide the level of protection described in 10 CFR 73.21.
In accordance with 10 CFR 19.11, you may be required to post this Notice within two working days.
Dated this 7th day of September 2005