ML020100111
| ML020100111 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  (DPR-080, DPR-082) |
| Issue date: | 04/22/2002 |
| From: | Shukla G NRC/NRR/DLPM |
| To: | Rueger G Pacific Gas & Electric Co |
| Shukla G | |
| References | |
| TAC MB3515, TAC MB3517 | |
| Download: ML020100111 (14) | |
Text
April 22, 2002 Mr. Gregory M. Rueger Senior Vice President, Generation and Chief Nuclear Officer Pacific Gas and Electric Company Diablo Canyon Nuclear Power Plant P. O. Box 3 Avila Beach, CA 94177
SUBJECT:
DIABLO CANYON NUCLEAR POWER PLANT, UNIT NOS. 1 AND 2 -
ISSUANCE OF AMENDMENT RE: REVISION OF TECHNICAL SPECIFICATIONS SECTION 5.5.16, FOR A ONE-TIME EXTENSION OF THE 10 CFR PART 50, APPENDIX J, INTEGRATED LEAK RATE TEST INTERVAL (TAC NOS. MB3515 AND MB3517)
Dear Mr. Rueger:
The Commission has issued the enclosed Amendment No. 150 to Facility Operating License No. DPR-80 and Amendment No. 150 to Facility Operating License No. DPR-82 for the Diablo Canyon Nuclear Power Plant (DCPP), Unit Nos. 1 and 2, respectively. The amendments consist of changes to the Technical Specifications (TSs) in response to your application dated November 16, 2001.
The amendments revise Technical Specification Section 5.5.16, "Containment Leakage Rate Testing Program," to allow a one-time extension of the 10 CFR Part 50, Appendix J, Type A integrated leak rate test interval from the required 10 years to a test interval of 15 years.
A copy of the related Safety Evaluation is enclosed. The Notice of Issuance will be included in the Commissions next regular biweekly Federal Register notice.
Sincerely,
/RA/
Girija S. Shukla, Project Manager, Section 2 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-275 and 50-323
Enclosures:
- 1. Amendment No.150 to DPR-80
- 2. Amendment No.150 to DPR-82
- 3. Safety Evaluation cc w/encls: See next page
April 22, 2002 Mr. Gregory M. Rueger Senior Vice President, Generation and Chief Nuclear Officer Pacific Gas and Electric Company Diablo Canyon Nuclear Power Plant P. O. Box 3 Avila Beach, CA 94177
SUBJECT:
DIABLO CANYON NUCLEAR POWER PLANT, UNIT NOS. 1 AND 2 -
ISSUANCE OF AMENDMENT RE: REVISION OF TECHNICAL SPECIFICATIONS SECTION 5.5.16, FOR A ONE-TIME EXTENSION OF THE 10 CFR PART 50, APPENDIX J, INTEGRATED LEAK RATE TEST INTERVAL (TAC NOS. MB3515 AND MB3517)
Dear Mr. Rueger:
The Commission has issued the enclosed Amendment No. 150 to Facility Operating License No. DPR-80 and Amendment No. 150 to Facility Operating License No. DPR-82 for the Diablo Canyon Nuclear Power Plant (DCPP), Unit Nos. 1 and 2, respectively. The amendments consist of changes to the Technical Specifications (TSs) in response to your application dated November 16, 2001.
The amendments revise Technical Specification Section 5.5.16, "Containment Leakage Rate Testing Program," to allow a one-time extension of the 10 CFR Part 50, Appendix J, Type A integrated leak rate test interval from the required 10 years to a test interval of 15 years.
A copy of the related Safety Evaluation is enclosed. The Notice of Issuance will be included in the Commissions next regular biweekly Federal Register notice.
Sincerely,
/RA/
Girija S. Shukla, Project Manager, Section 2 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-275 DISTRIBUTION:
and 50-323 PUBLIC RidsNrrPMGShukla PDIV-2 R/F RidsNrrLAEPeyton
Enclosures:
- 1. Amendment No.150 to DPR-80 RidsNrrDlpmLpdiv DBujol, RIV
- 2. Amendment No.150 to DPR-82 GHill (4)
LHurley, RIV
- 3. Safety Evaluation RidsOgcRp JHannon RidsAcrsAcnwMailCenter RidsRgn4MailCenter(KBrockman) cc w/encls: See next page RidsNrrWBeckner ACCESSION NO.: ML020100111 RBarrett PACKAGE: ML021260734 & TECH SPEC PAGES: ML021140131 OFFICE PDIV-2/PM PDIV-2/LA SPLB/SC SPSB/SC EMEB/SC OGC PDIV-2/SC NAME GShukla:lf EPeyton JHannon RBarrett DTerao SBrock SDembek DATE 1/30/02 1/30/02 2/4/02 3/12/02 3/22/02 4/19/02 4/19/02 OFFICIAL RECORD COPY
Diablo Canyon Power Plant, Units 1 and 2 cc:
NRC Resident Inspector Diablo Canyon Nuclear Power Plant c/o U.S. Nuclear Regulatory Commission P.O. Box 369 Avila Beach, CA 93424 Dr. Richard Ferguson, Energy Chair Sierra Club California 1100 11th Street, Suite 311 Sacramento, CA 95814 Ms. Nancy Culver San Luis Obispo Mothers for Peace P.O. Box 164 Pismo Beach, CA 93448 Chairman San Luis Obispo County Board of Supervisors Room 370 County Government Center San Luis Obispo, CA 93408 Mr. Truman Burns Mr. Robert Kinosian California Public Utilities Commission 505 Van Ness, Room 4102 San Francisco, CA 94102 Mr. Steve Hsu Radiologic Health Branch State Department of Health Services P.O. Box 942732 Sacramento, CA 94327-7320 Diablo Canyon Independent Safety Committee ATTN: Robert R. Wellington, Esq.
Legal Counsel 857 Cass Street, Suite D Monterey, CA 93940 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission Harris Tower & Pavilion 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 Christopher J. Warner, Esq.
Pacific Gas & Electric Company Post Office Box 7442 San Francisco, CA 94120 Mr. David H. Oatley, Vice President, Diablo Canyon Operations Diablo Canyon Nuclear Power Plant P.O. Box 3 Avila Beach, CA 93424 Telegram-Tribune ATTN: Managing Editor 1321 Johnson Avenue P.O. Box 112 San Luis Obispo, CA 93406 Mr. Ed Bailey, Radiation Program Director Radiologic Health Branch State Department of Health Services P.O. Box 942732 (MS 178)
Sacramento, CA 94327-7320 Mr. Robert A. Laurie, Commissioner California Energy Commission 1516 Ninth Street (MS 31)
Sacramento, CA 95814
PACIFIC GAS AND ELECTRIC COMPANY DOCKET NO. 50-275 DIABLO CANYON NUCLEAR POWER PLANT, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 150 License No. DPR-80 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Pacific Gas and Electric Company (the licensee) dated November 16, 2001, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions 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. DPR-80 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No.
150, are hereby incorporated in the license. Pacific Gas and Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Stephen Dembek, Chief, Section 2 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: April 22, 2002
PACIFIC GAS AND ELECTRIC COMPANY DOCKET NO. 50-323 DIABLO CANYON NUCLEAR POWER PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 150 License No. DPR-82 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Pacific Gas and Electric Company (the licensee) dated November 16, 2001, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions 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. DPR-82 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No.
150, are hereby incorporated in the license. Pacific Gas and Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days from the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Stephen Dembek, Chief, Section 2 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: April 22, 2002
ATTACHMENT TO LICENSE AMENDMENT NO. 150 TO FACILITY OPERATING LICENSE NO. DPR-80 AND AMENDMENT NO. 150 TO FACILITY OPERATING LICENSE NO. DPR-82 DOCKET NOS. 50-275 AND 50-323 Replace the following page of the Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.
REMOVE INSERT 5.0-24 5.0-24
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 150 TO FACILITY OPERATING LICENSE NO. DPR-80 AND AMENDMENT NO. 150 TO FACILITY OPERATING LICENSE NO. DPR-82 PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON NUCLEAR POWER PLANT, UNITS 1 AND 2 DOCKET NOS. 50-275 AND 50-323
1.0 INTRODUCTION
By application dated November 16, 2001, Pacific Gas and Electric Company (the licensee) requested changes to the Technical Specifications (TS) (Appendix A to Facility Operating License Nos. DPR-80 and DPR-82) for the Diablo Canyon Nuclear Power Plant (DCPP), Units Nos. 1 and 2. The proposed changes would modify TS Section 5.5.16, "Containment Leakage Rate Testing Program," to allow a one-time extension of the 10 CFR Part 50, Appendix J, Type A test interval from the required 10 years to a test interval of 15 years.
This application did not request to modify the existing containment building inspections under the requirements of the American Society of Mechanical Engineers (ASME)Section XI Subsections IWE and IWL, or the existing leakage rate testing programs under 10 CFR Part 50, Appendix J, Type B and Type C.
2.0 BACKGROUND
Appendix J, Option B of 10 CFR Part 50 requires that a Type A containment integrated leakage rate test (ILRT) be conducted at a periodic interval based on historical performance of the overall containment system. TS Section 5.5.16 requires that a program be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR Part 50, Appendix J, Option B, as modified by approved exemptions. TS 5.5.16 further requires that this program be in accordance with the guidelines contained in Regulatory Guide (RG) 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995.
RG 1.163 references 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, based upon two consecutive successful tests. There is also a provision for extending the test interval an additional 15 months under certain circumstances.
The licensee has requested a revision to TS 5.5.16 which would allow an exception to the guidelines of RG 1.163 regarding the Type A test interval. Specifically, the proposed TS change would allow the 10-year interval between ILRTs to be extended to a 15-year interval from the last ILRTs completed for DCPP. Unit 1s last ILRT was completed on May 4, 1994, and Unit 2s ILRT was completed on April 30, 1993.
3.0 EVALUATION Appendix J of 10 CFR Part 50 specifies the leakage rate test requirements for primary reactor containments. The test requirements ensure that: (a) leakage through containment or systems and components penetrating containment does not exceed allowable leakage rates specified in the TS; and (b) integrity of the containment structure is maintained during its service life. The licensee has adopted Option B of 10 CFR Part 50, Appendix J, which requires that the ILRT be performed at periodic intervals based on performance of the containment system. Adoption of the Option B performance-based containment leakage rate testing program allowed a change in the frequency of measuring primary containment leakage in Type A, B, and C tests.
The licensee provided a risk impact assessment of extending the Type A test interval from 10 years to 15 years. In performing the risk assessment, the licensee followed the guidelines of NEI 94-01, and the guidelines of RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis."
The basis for the current 10-year test interval is provided in Section 11.0 of NEI 94-01 which states that NUREG-1493, Performance-Based Containment Leak-Test Program, September 1995, provides the technical basis to support the revised leakage rate testing requirements contained in Option B to Appendix J. The basis consisted 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. NEI 94-01 provides methods acceptable for complying with the provisions of Option B of 10 CFR Part 50, Appendix J. NEI 94-01 contains the criterion that allows Option B Type A testing to be performed at an interval of at least once per 10 years.
The surveillance interval for Type A testing specified in NEI 94-01 is at least 1-in-10 years based on an acceptable performance history (i.e., two consecutive periodic Type A tests at least 24 months apart where the calculated performance leakage rate was less than 1.0 La, where La is equal to 0.1 percent of the containment air weight per day at the peak calculated containment internal pressure for the design basis loss-of-coolant accident [47 psig]) and consideration of the performance factors in Section 11.3 of NEI 94-01, related to past component performance, service, design, safety impact, and cause determination.
The two most recent Type A tests performed at DCPP have been successful, so the licensees current interval of performing the ILRT is 10 years. The licensee has performed six Type A tests, including preoperational and operational testing, on Unit 1, and two Type A tests, including preoperational and operational testing, on Unit 2. The results have indicated that a considerable margin exists between the Type A test results and the TS 5.5.16 limit of 0.75 La.
Therefore, these test results have demonstrated that both units have low leakage containments.
The containment leakage rate testing interval is based upon an evaluation which looks at the "as found" leakage history to determine the interval for leakage rate testing which provides assurance that leakage rate limits will be maintained. The Type A test interval was changed from 3-in-10 years to 1-in-10 years in Amendment No. 110 for DCCP Unit 1 and Amendment No. 109 for DCPP Unit 2.
The licensee has performed a risk assessment of the proposed one-time extension of the containment Type A leakage rate test interval from a 10-year interval to a 15-year interval. In performing the risk assessment, the licensee considered the guidelines set forth in NEI 94-01, the methodology used in Electric Power Research Institute (EPRI) Report TR-104285, and the NRC regulatory guidance on the use of PRA findings and risk insights for changes to a plants licensing basis as discussed in RG 1.174. The updated Level 1 internal events PRA model was used for this risk assessment calculation in combination with the individual plant examination (IPE) Level 2 model.
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. The EPRI study estimated that relaxing the test interval from 3 in 10 years to 1 in 10 years, would increase 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 based on industry leakage rate data gathered from 1987 to 1993), this results in a 10 percent increase in the overall probability of leakage. The risk contribution of leakage, in percent of person-rem/year, for the pressurized-water reactor representative plant was estimated to increase from.032 percent to.035 percent. This 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.
Building upon the methodology of the EPRI study, the licensee assessed the change in the predicted person-rem/year frequency. The staff considers the licensees assessment an improvement of the EPRI study because the leakage from sequences that have the potential to result in large releases if a pre-existing leak were present were quantified. Since the Option B rulemaking in 1995, the 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 during 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 which the licensee estimated. 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 the defense-in-depth philosophy, are met. The licensee estimated the increase in the conditional containment failure probability which helps to ensure that the defense-in-depth philosophy is maintained.
The licensee provided an analysis which estimated all of these risk metrics and whose methodology is consistent with previously approved submittals. The following conclusions can be drawn from the analysis associated with extending the Type A test frequency:
1.
A slight increase in risk is predicted when compared to that estimated from current requirements. Given the change from a 10 year test interval to a 15 year test interval, the increase in the total integrated plant risk in person-rem/year is estimated to be 0.03 percent. The increase in the total integrated plant risk, given the change from a 3 in 10 year test interval to a 15 year test interval, was found to be 0.17 percent. This is reasonable when compared to the range of risk increase, 0.02 to 0.14 percent, estimated in NUREG-1493 when going from a 3 in 10 year test interval to a 10 year interval. NUREG-1493 concluded that a reduction in the frequency of tests from 3 per 10 years to 1 per 10 years leads to an imperceptible increase in risk. Therefore, the increase in the total integrated plant risk for the proposed change is considered small and supportive of the proposed change.
2.
RG 1.174 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. RG 1.174 defines very small changes in the risk-acceptance guidelines as increases in CDF less than 10-6 per reactor year and increases in LERF less than 10-7 per reactor year. Since the Type A test does not impact CDF, the relevant criterion is LERF. The increase in LERF resulting from a change in the Type A test interval from 1 in 10 years to 1 in 15 years is estimated to be 1 x 10-8/year. The increase in LERF resulting from a change in the Type A test interval from the original 3 in 10 years to 1 in 15 years is estimated to be 3 x 10-8/year. Increasing the Type A interval to 15 years is considered to be 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 philosophy.
Consistency with the defense-in-depth philosophy is maintained if a reasonable balance is preserved among prevention of core damage, prevention of containment failure, and consequence mitigation. The change in the conditional containment failure probability was estimated to increase by 0.001 for the proposed change and 0.003 for the cumulative change of going from a test interval of 3 in 10 years to 1 in 15 years. The staff finds that the defense-in-depth philosophy is maintained based on the very small change in the conditional containment failure probability for the proposed change.
The staff recognizes the limitations of a conditional containment failure probability approach. For plants, such as DCPP, with core damage frequency estimates well below 10-4, the ability of the containment to withstand events of even lower probability becomes less clear. Therefore, it is important to consider other risk metrics in conjunction with the conditional containment failure probability, such as total LERF. The licensee has sufficiently demonstrated that the total LERF for internal events is less than 10-5 for the purpose of this evaluation.
Based on these conclusions, the staff finds that the increase in predicted risk due to the proposed change is within the acceptance guidelines while maintaining the defense-in-depth philosophy of RG 1.174 and, therefore, is acceptable.
The DCPP, Units 1 and 2 are Westinghouse pressurized-water reactors with a large, dry reinforced concrete primary containment structure. The containment pressure boundary consists of the steel containment shell structure, containment access penetrations, and process piping and electrical penetrations. The integrity of the penetrations is verified through Type B and Type C local leak rate tests (LLRT) 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 essentially leak-tight characteristics of the containment structure at the design basis accident (DBA) pressure. As stated in the request, six Type A tests, including preoperational and operational testing, have been performed on Unit 1 (December 1975, November 1978, February 1982, April 1985, May 1988, and April 1994) and two Type A tests, including preoperational and operational testing have been performed on Unit 2 (August 1984 and November 1993). Based on these successful Type A tests and the requirements of 10 CFR Part 50, Appendix J, Option B, the current interval requirement is 10 years for both units.
With the requested extension of the ILRT time interval, the next overall verification of the containment leak-tight integrity will be performed in April 2009 for Unit 1 and November 2008 for Unit 2. Because the leak rate testing requirements (ILRT and LLRTs) of Option B of Appendix J, and the containment inservice inspection (ISI) requirements mandated by 10 CFR 50.55a complement each other in ensuring the leak-tightness and structural integrity of the containment, the licensee provided in the request information related to the ISI of the primary containment and potential areas of weaknesses in the containment that may not be apparent in the risk assessment. In addition, the licensee provided information to explicitly address the five issues raised by the staff during its ILRT review of other plants. The staffs evaluation of the licensees discussion of the five issues is provided in the following paragraphs.
- 1. Regarding the ISI performed on the containment, the licensee stated that the ISI of the DCPP containment buildings is conducted in accordance with the requirements of the 1992 Edition with 1992 Addenda of Subsections IWE and IWL of the ASME Code,Section XI, including the NRC-approved requests for relief from certain code requirements. The results for the in-service examination indicate that there are some minor areas of coating degradation occurring, but nothing significant that would adversely impact either the containment structural integrity or leak tightness. All identified areas of minor coating degradation were evaluated and found to be limited in scope, with no significant liner material loss, and no potential for precursors to significant containment liner failures.
- 2. For the issue related to the application of any augmented examination (required by IWE Table-2500-1, Examination Category E-C) and findings at DCPP, the licensee stated that based on the results of documented historical issues, the sump liner for each unit has been classified as augmented examination areas. These augmented examination areas were overlaid with a grid system and ultrasonically examined for wall thinning. This process allowed an assessment of the condition of the backside of the liner in this area, which was determined to be sound and free from detectable wastage. A visual examination of the augmented inspection area noted no significant attack of the liner.
- 3. With regard to the issue related to the ISI of seals, gaskets and pressure retaining bolted connections, the licensee stated that with the approved requests for relief in these areas, under Option B of Appendix J, the containment leak-tight integrity will be pressure tested periodically during Type B tests. As stated in the approved relief requests, the alternate examinations of Appendix J testing will be performed at least once during each containment ten-year inspection interval. Thus, the extension requested for Type A testing does not affect the interval of these alternate examinations. Inspection results of seals, gaskets, and bolts have shown that Appendix J Type B component seals have had an excellent performance record. On this basis, the licensee concluded that the schedule for examination of the seals, gaskets, and bolts will continue to provide assurance that the integrity of the containment pressure boundary is maintained.
- 4. As for the integrity of stainless steel bellows, the licensee stated that the DCPP containment isolation system does not contain any stainless steel bellows that are credited as a containment pressure boundary. Therefore, the concerns of Information Notice 92-20, "Inadequate Local Leak Rate Testing," are not applicable to DCPP and no ISI of these bellows is required.
- 5. Because ILRTs help to identify areas of through-wall degradations when the containment vessel is pressurized, the staff reviewed how the licensee addressed the potential leakages due to age-related degradation in the uninspectable areas (areas that cannot be visually examined) were considered in risk assessment of the extended ILRT. The licensee stated that the potential for containment leakage was explicitly included in the risk assessment. By definition, the intact containment case, EPRI Containment Failure Class 1, includes a leakage term that is independent of the source of the leak. The assessment also includes specific containment failure classes due to extending the ILRT interval (Classes 3a, 3b and 7). These cases include the potential that the leakage is caused by a liner failure. The assessment shows that even with the increased potential to have an undetected containment flaw or leak path, the increase in risk is small.
Based on the information provided in the TS change request and the licensee discussion of the above five issues, the staff finds that (1) the structural integrity of the containment vessel is verified through the periodic inservice inspections conducted as required by Subsections IWE and IWL of the ASME Code,Section XI, (2) the integrity of the penetrations and containment isolation valves are periodically verified through Type B and Type C tests as required by 10 CFR Part 50, Appendix J and DCPP TS, and (3) the potential for large leakage from the areas that cannot be examined by the ISI has been explicitly modeled in performing the risk assessment. In addition, the system pressure tests for containment pressure boundary (i.e.,
Appendix J tests, as applicable) are required to be 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 is required to be reported under 10 CFR 50.72 and 10 CFR 50.73.
Additionally, the one-time increase from a 10-year interval to a 15-year interval for the Type A leakage rate test does not affect any accident parameters discussed in the DCPP Final Safety Analysis Report (FSAR) Update. The Type A test interval does not affect the operation of any safety related equipment credited for accident mitigation. A change in the Type A test interval does not affect the containment initial conditions nor the leak rates that are assumed in the accident analyses. Therefore, the one-time increase from a 10-year interval to a 15-year interval for the Type A leakage rate test does not affect the results of accident analyses results in the DCPP FSAR Update.
Based on the foregoing evaluation, the staff finds that the interval until the next Type A testing at DCPP may be extended to 15 years, and that the proposed changes to TS 5.5.16 are acceptable.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the California State official was notified of the proposed issuance of the amendments. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
These amendments change a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration and there has been no public comment on such finding (67 FR 930). Accordingly, the amendments meet 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 amendments.
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 amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: G. S. Shukla Date: April 22, 2002