ML030030827
| ML030030827 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 12/31/2002 |
| From: | Stephen Monarque NRC/NRR/DLPM/LPD2 |
| To: | Christian D Virginia Electric & Power Co (VEPCO) |
| References | |
| TAC MB3611 | |
| Download: ML030030827 (12) | |
Text
December 31, 2002 Mr. David A. Christian Sr. Vice President and Chief Nuclear Officer Virginia Electric and Power Company 5000 Dominion Blvd.
Glen Allen, Virginia 23060
SUBJECT:
NORTH ANNA POWER STATION, UNIT 1 - ISSUANCE OF AMENDMENT RE: PROPOSED RISK-INFORMED TECHNICAL SPECIFICATION CHANGE TO EXTEND THE TYPE A TEST INTERVAL BY 5 YEARS (TAC NO. MB3611)
Dear Mr. Christian:
The Commission has issued the enclosed Amendment No. 234 to Facility Operating License No. NPF-4 for the North Anna Power Station, Unit No. 1. The amendment changes the Technical Specifications in response to your letter dated December 7, 2001, as supplemented June 28, 2002, and July 25, 2002.
The amendment permits a one-time, 5-year extension of the 10-year performance-based Type A test interval established in Nuclear Energy Institute 94-01, Industry Guideline For Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, Revision 0, July 26, 1995.
A copy of the Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Sincerely,
/RA by K.Cotton for/
Stephen Monarque, Project Manager, Section 1 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-338
Enclosures:
- 1. Amendment No. 234 to NPF-4
- 2. Safety Evaluation cc w/encls: See next page
Mr. David A. Christian December 31, 2002 Senior Vice President - Nuclear Virginia Electric and Power Company 5000 Dominion Blvd.
Glen Allen, Virginia 23060
SUBJECT:
NORTH ANNA POWER STATION, UNIT 1 - ISSUANCE OF AMENDMENT RE: PROPOSED RISK-INFORMED TECHNICAL SPECIFICATION CHANGE TO EXTEND THE TYPE A TEST INTERVAL BY 5 YEARS (TAC NO. MB3611)
Dear Mr. Christian:
The Commission has issued the enclosed Amendment No. 234 to Facility Operating License No. NPF-4 for the North Anna Power Station, Unit No. 1. The amendment changes the Technical Specifications in response to your letter dated December 7, 2002, as supplemented June 28, 2002, and July 25, 2002.
The amendment permits a one-time, 5-year extension of the 10-year performance-based Type A test interval established in Nuclear Energy Institute 94-01, Industry Guideline For Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, Revision 0, July 26, 1995.
A copy of the Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Sincerely,
/RA by K.Cotton for/
Stephen Monarque, Project Manager, Section 1 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-338
Enclosures:
- 1. Amendment No. 234 to NPF-4
- 2. Safety Evaluation cc w/encls: See next page Distribution:
PUBLIC ACRS GHill (2 paper copies)
OGC PDII-1 R/F SMonarque (paper copy)
RDennig KLandis HBerkow EDunnington (paper copy)
OTabatabai JNakoski FILENAME - C:\\ORPCheckout\\FileNET\\ML030030827.wpd ADAMS ACCESSION NO. ML030030827 OFFICE PM:PDII/S1 LA:PDII/S2 OGC SC:PDII/S1 NAME KCotton for SMonarque EDunnington JNakoski DATE 12/31/02 12/11/02 12/16/02 12/31/02 COPY Yes/No Yes/No Yes/No Yes/No OFFICIAL RECORD COPY
VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-338 NORTH ANNA POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 234 License No. NPF-4 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Virginia Electric and Power Company (the licensee) dated December 7, 2001, as supplemented June 28, 2002, and July 25, 2002, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-4 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 234, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
John A. Nakoski, Chief, Section 1 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: December 31, 2002
ATTACHMENT TO LICENSE AMENDMENT NO. 234TO FACILITY OPERATING LICENSE NO. NPF-4 DOCKET NO. 50-338 Replace the following page of the Appendix "A" Technical Specifications with the enclosed revised page. The revised page is identified by amendment number and contains a vertical line indicating the area of change.
Remove Page Insert Page 5.5-18 5.5-18
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.234 TO FACILITY OPERATING LICENSE NO. NPF-4 VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION, UNIT NO. 1 DOCKET NO. 50-338
1.0 INTRODUCTION
By letter dated December 7, 2001, as supplemented by letters dated June 28, 2002, and July 25, 2002, Virginia Electric and Power Company (VEPCO), the licensee for the North Anna Power Station, Unit 1 (North Anna 1), requested Technical Specification (TS) changes to Section 5.5.15, Containment Leakage Rate Testing Program. The licensee proposed to add the following words to the end of TS 5.5.15.a as modified by the following exception:
NEI 94-01-1995, Section 9.2.3: The first Type A test performed after the April 3, 1993, Type A test shall be performed no later than April 2, 2008.
The proposed change would allow the licensee to defer the Appendix J Type A Containment Integrated Leak Rate Test (ILRT) from the scheduled April 2003 date until April 2008, which is 15 years after the last ILRT, performed on April 3, 1993. The extension would enable North Anna 1 to optimize refueling activities and save critical-path time in the upcoming Refueling Outage 16, scheduled to begin in March 2003. The requested TS change is based on a risk-informed approach and follows the guidelines in Regulatory Guide (RG) 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis.
The June 28, 2002, and July 25, 2002, supplements contained clarifying information and did not change the initial no significant hazards consideration determination or expand the scope of the initial application. The initial request from the licensee requested changes to TS Section 4.6.12 that was formatted as Custom TS and TS Section 5.5.15 formatted as Improved TS (ITS). Per the letter dated August 26, 2002, North Anna implemented the IST as of September 2, 2002; therefore, this amendment will only address the change to TS Section 5.5.15 that is in the ITS format.
2.0 REGULATORY EVALUATION
Option B of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix J requires that a Type A test be conducted at a periodic interval based on the historical performance of the overall containment system. TS 5.5.15 for North Anna 1 requires that leakage rate testing be performed as required by 10 CFR Part 50, Appendix J, Option B, as modified by approved exemptions, and in accordance with the guidelines of RG 1.163, Performance-Based Containment Leak-Test Program, dated September 1995. This RG endorses, with certain exceptions, Nuclear Energy Institute (NEI) 94-01, Revision 0, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, dated July 26, 1995.
A Type A test is an overall (integrated) leakage rate test of the containment structure.
NEI 94-01 specifies an initial test interval of 48 months, but allows an extended interval of 10 years, after two consecutive successful tests. There is also a provision for extending the test interval an additional 15 months in certain circumstances.
The most recent two Type A tests at North Anna 1 have been successful, so the current interval requirement is 10 years.
The licensee is requesting additions to TS 5.5.15 to allow an exception from the guidelines of RG 1.163 regarding the Type A test interval. Specifically, the proposed TS states that the first Unit 1 Type A test performed after the April 3, 1993, Type A test shall be performed no later than April 2, 2008.
3.0 TECHNICAL EVALUATION
The licensee provided its evaluation to the NRC staff in the December 7, 2001, application for license amendment. The licensee provided additional analysis and information in letters dated June 28, 2002, and July 25, 2002.
3.1 Containment Pressure Boundary Degradation North Anna 1 is a Westinghouse pressurized-water reactor (PWR) with a large reinforced concrete primary containment structure. The containment pressure boundary consists of the steel liner, containment access penetrations, and penetrations for process piping and electrical wiring. The integrity of the penetrations is verified through Type B and Type C local leak rate tests (LLRTs) as required by 10 CFR Part 50, Appendix J, and the overall integrity of the containment structure is verified through an ILRT. These tests are performed to verify the leaktightness of the containment structure at the design-basis accident (DBA) pressure. As stated in the request, North Anna 1 has performed two ILRTs during the period of its operating license. The first ILRT was completed in June 1989, and the second in April 1993. Based on the two successful Type A tests at North Anna 1 and the requirements of 10 CFR Part 50, Appendix J, Option B, the current interval requirement is 10 years. In requesting an extension of the ILRT interval, the licensee proposed to perform the next overall verification of the containment leaktight integrity by April 2, 2008.
The leak rate testing requirements of Option B of 10 CFR Part 50, Appendix J, and the containment inservice inspection (ISI) requirements of 10 CFR 50.55a complement each other in ensuring the leaktightness and structural integrity of the containment. After reviewing NRC safety evaluations of Type A test interval extension requests for other plants, the NRC staff identified several general concerns about the ISI of the containment and potential weaknesses in the containment and requested the licensee to address these concerns. The NRC staffs evaluation of the licensees responses is discussed in the following paragraphs.
Regarding the ISI program for the containment and the schedule for implementation, the licensee stated that the ISI of the North Anna 1 containment building is conducted in accordance with its ISI program that is based on the requirements of the 1992 edition through the 1992 addenda of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),Section XI, Subsections IWE and IWL. According to the licensee, North Anna 1 completed its first interval IWE examinations on March 14, 2000, and its first-interval (5-year) IWL examinations on August 31, 2001. The next two IWE examination intervals are scheduled to run from March 15, 2000, to March 14, 2004, and from March 15, 2004, to March 14, 2007. Based on the current inspections and the associated engineering evaluations, the licensee stated that no areas of the containment liner required augmented examinations according Subsection IWE, Subarticle IWE-1240. Based on the information provided by the licensee, the NRC staff finds that the schedule for implementing the containment ISI program will not be affected by the requested extension of the ILRT interval to 15 years.
With regard to the issue related to the ISI of seals and gaskets, the licensee stated that since the approval of relief requests (RR-IWE2), the leaktightness of seals and gaskets at North Anna 1 and 2 is Type B-tested in accordance with 10 CFR Part 50, Appendix J at least once each inspection interval. The licensee performs some of the Appendix J Type B tests each refueling outage, staggering the testing to balance the outage work scope. Currently, these tests are completed at approximately 60-month intervals. The licensee stated that Type B tests will continue to be completed within the 120-month interval and will adequately test the applicable seals and gaskets. The proposed one-time Appendix J Type A extension will not affect this relief basis.
With respect to the issue of torque or tension tests on bolted connections not disassembled and reassembled during the inspection interval, the licensee stated that with the approval of the alternative requested in RR-IWE5, the following examinations and tests required by Subsection IWE ensure the structural integrity and leaktightness of Class MC pressure-retaining bolting:
(1)
Exposed surfaces of bolted connections shall be visually examined in accordance with requirements of Table IWE-2500-1, Examination Category E-G, Pressure Retaining Bolting, Item E8.10.
(2)
Bolted connections shall meet the pressure test requirements of Table IWE-2500-1, Examination Category E-P, All Pressure Retaining Components, Item E9.40.
The Appendix J Type A test frequency extension will not affect the relief request basis or the frequency of Type B testing on components with bolting. On this basis, the NRC staff finds that the licensees ISI program for seals, gaskets, and bolted connections provides reasonable assurance that the integrity of the containment pressure boundary will be maintained.
In its response to the issue of the integrity of stainless steel bellows, the licensee stated that a stainless steel bellows is located inside the containment on the outer tube of the fuel transfer tube containment penetration. This bellows merely compensates for any differential motion and is not a part of the containment boundary. The containment boundary is formed by the welded connection of the containment liner to the inner and outer tubes and by the double O-ring blank flange on the inner tube. The blank flange is Type-B tested every refueling outage and the welded connection is tested during the ILRT. A manual isolation valve isolates the inner tube from the spent fuel pool in the spent fuel building. The NRC staff finds that the licensees justification and examination approach provide reasonable assurance that the integrity of the containment pressure boundary will be maintained.
On the basis of its review of the information provided by the licensee in its TS amendment request and its response to the NRC staffs questions, the NRC staff finds that (1) the structural integrity of the containment vessel is verified through the periodic ISIs conducted as required by Subsections IWE and IWL of the ASME Code,Section XI, and (2) the integrity of the penetrations and containment isolation valves is periodically verified through Type B and Type C tests as required by 10 CFR Part 50, Appendix J. In addition, the system pressure tests for containment pressure boundary (i.e., Appendix J tests, as applicable) are performed following repair and replacement activities, if any, in accordance with Article IWE-5000 of the ASME Code,Section XI. Serious degradation of the primary containment pressure boundary must be reported under 10 CFR 50.72 and 50.73.
On this basis, the NRC staff concludes that a one-time extension for performing the integrated leak rate testing, as proposed by the licensee in Section 5.5.15 of the proposed TS amendment request, is acceptable.
3.2 Risk Evaluation The licensee has performed an assessment of the risk impact of extending the Type A test interval to 15 years. In performing the risk assessment, the licensee considered the guidelines of NEI 94-01, the methodology used in Electric Power Research Institute (EPRI) TR-104285, Risk Impact Assessment of Revised Containment Leak Rate Testing, and RG 1.174.
The basis for the current 10-year test interval is provided in Section 11.0 of NEI 94-01, Revision 0, and was established in 1995 during development of the performance-based Option B to Appendix J. Section 11.0 of NEI 94-01 states that NUREG-1493, Performance-Based Containment Leak-Test Program, September 1995, provided the technical basis for rulemaking to revise the leakage rate testing requirements in Option B to Appendix J. The basis consists of qualitative and quantitative assessments of the risk impact (in terms of increased public dose) associated with a range of extended leakage rate test intervals. To supplement the NRCs rulemaking basis, NEI undertook a similar study. The results of that study are documented in EPRI Research Project Report TR-104285.
The EPRI study used an analytical approach similar to that presented in NUREG-1493 for evaluating the incremental risk associated with increasing the interval for Type A tests.
Decreasing the test frequency from 3 in 10 years to 1 in 10 years increases the average time that a leak detectable only by a Type A test goes undetected from 18 to 60 months. Since Type A tests only detect about 3 percent of leaks (the rest are identified during local leak rate tests, according to industry leakage rate data gathered from 1987 to 1993), NUREG-1493 estimated a 10-percent increase in the overall probability of leakage. The risk contribution of preexisting leakage, in percent of person-rem/year, for the representative PWRs and boiling water reactors confirmed the NUREG-1493 conclusion that a reduction in the frequency of Type A tests from 3 per 10 years to 1 per 10 years leads to an imperceptible increase in risk (ranging between 0.02 and 0.14 percent).
Using the methodology of the EPRI study, the licensee assessed the change in the predicted person-rem/year frequency. The licensee quantified the risk from sequences that have the potential to result in large releases if a preexisting leak is present. Since the Option B rulemaking in 1995, the NRC staff has issued RG 1.174 on the use of probabilistic risk assessment (PRA) in risk-informed changes to a plants licensing basis. The licensee has proposed using RG 1.174 to assess the acceptability of extending the Type A test interval beyond that established by the Option B rulemaking. RG 1.174 defines very small changes in the risk-acceptance guidelines as increases in core damage frequency (CDF) less than 10-6 per reactor year and increases in large early release frequency (LERF) less than 10-7 per reactor year. Since the Type A test does not impact CDF, the relevant criterion is the change in LERF.
The licensee has estimated the increase in LERF for the proposed change and the cumulative increase resulting from decreasing the test frequency from three tests in 10 years to one test in 15 years. RG 1.174 also discusses defense-in-depth and encourages the use of risk analysis techniques to help ensure and show that key principles, such as defense-in-depth, are met.
The licensee estimated the change in the conditional containment failure probability for the proposed change in order to demonstrate that the defense-in-depth principle is met.
The licensee provided an analysis that estimated all of these risk metrics and using a methodology that is consistent with previously approved submittals. The following conclusions can be drawn from the analysis decreasing the Type A test frequency:
- 1. Changing the test interval from 10 years to 15 years is estimated to increase the total integrated plant risk, in person-rem/year by 0.001 percent. The total integrated plant risk will increase 0.003 percent as a result of doing one test in 15 years instead of three in 10 years. NUREG-1493 concluded that a reduction in the frequency of tests from three per 10 years to one per 10 years leads to an imperceptible increase in risk, ranging from 0.02 to 0.14 percent. Therefore, the increase in the total integrated plant risk for the proposed change is considered to be small and supportive of the proposed change.
- 2. The increase in LERF resulting from a change in the Type A test frequency from the original three in 10 years to one in 15 years is estimated to be 1.1 x 10-7/year.
However, there is some likelihood that the undetected flaw in the containment liner estimated as part of the Class 3b frequency would be detected as part of the IWE visual examination of the containment liner. Eighty-five percent of the inner containment liner can be visually inspected. Assuming the visual inspections are capable of detecting large flaws in the visible regions of the containment, the increase in LERF would drop from 1.1 x 10-7/year to 1.7 x 10-8/year. Therefore, increasing the Type A interval to 15 years is considered to be a very small change in LERF, using the guidelines of RG 1.174.
The licensee performed additional risk analysis to consider the impact of hypothetical corrosion in inaccessible areas of the containment liner on the proposed change. The inaccessible areas included the back of the containment liner. The risk analysis considered the likelihood of an age-adjusted liner flaw that would lead to a breach of the containment.
The risk analysis also considered the likelihood that the flaw was not visually detected but could be detected by a Type A test. When possible corrosion of the containment liner is considered, the increase in LERF resulting from a change in the Type A test frequency from the original three in 10 years to one in 15 years is estimated to be 2.1 x 10-8/year. This additional risk analysis provides added assurance that increasing the Type A interval to 15 years is a very small change in LERF.
- 3. RG 1.174 also encourages the use of risk analysis techniques to help ensure and show that the proposed change is consistent with the defense-in-depth principle. Defense-in-depth is maintained if a reasonable balance is preserved among prevention of core damage, prevention of containment failure, and consequence mitigation. Based on information provided by the licensee, the NRC staff estimates the change in the conditional containment failure probability to be an increase of 0.1 percent for the proposed change and 0.3 percent for the cumulative effort of one test in 15 years instead of three in 10 years. The NRC staff finds that the defense-in-depth principle is maintained based on the change in the conditional containment failure probability for the proposed amendment.
Based on these conclusions, the NRC staff finds that the increase in predicted risk due to the proposed change is within the acceptance guidelines and maintains the defense-in-depth principle of RG 1.174; therefore, the change is acceptable.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Virginia State official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a surveillance requirement. The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released off site, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on this finding (67 FR 21295). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: J. Pulsipher, SPLB/DSSA M. Snodderly, SPLB/DSSA T. Cheng, DE/EMEB Date:
December 31, 2002
Mr. David A. Christian North Anna Power Station Virginia Electric and Power Company Units 1 and 2 cc:
Mr. C. Lee Lintecum County Administrator Louisa County P. O. Box 160 Louisa, Virginia 23093 Ms. Lillian M. Cuocco, Esq.
Senior Nuclear Counsel Dominion Nuclear Connecticut, Inc.
Millstone Power Station Building 475, 5 th floor Rope Ferry Road Rt. 156 Waterford, Connecticut 06385 Dr. W. T. Lough Virginia State Corporation Commission Division of Energy Regulation P. O. Box 1197 Richmond, Virginia 23209 Old Dominion Electric Cooperative 4201 Dominion Blvd.
Glen Allen, Virginia 23060 Mr. Stephen P. Sarver, Director Nuclear Licensing & Operations Support Virginia Electric Power Company Innsbrook Technical Center 5000 Dominion Blvd.
Glen Allen, Virginia 23060-6711 Office of the Attorney General Commonwealth of Virginia 900 East Main Street Richmond, Virginia 23219 Senior Resident Inspector North Anna Power Station U. S. Nuclear Regulatory Commission 1024 Haley Drive Mineral, Virginia 23117 Mr. David A. Heacock Site Vice President North Anna Power Station P. O. Box 402 Mineral, Virginia 23117-0402 Mr. Richard H. Blount, II Site Vice President Surry Power Station Virginia Electric and Power Company 5570 Hog Island Road Surry, Virginia 23883-0315 Mr. Robert B. Strobe, M.D., M.P.H.
State Health Commissioner Office of the Commissioner Virginia Department of Health P. O. Box 2448 Richmond, Virginia 23218 Mr. William R. Matthews Vice President-Nuclear Operations Virginia Electric and Power Company Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, Virginia 23060-6711