ML102590586
| ML102590586 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 10/01/2010 |
| From: | Chawla M L, Pascarelli R J Plant Licensing Branch III |
| To: | Jennifer Davis Detroit Edison |
| Chawla M L, NRR/DORL, 415-8371 | |
| References | |
| TAC ME3119 | |
| Download: ML102590586 (9) | |
Text
UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 October 1, 2010 Mr. Jack M. Davis Senior Vice President and Chief Nuclear Officer Detroit Edison Company Fermi 2 -210 NOC 6400 North Dixie Highway Newport, MI 48166 RELIEF REQUEST RR-A30 FOR FERMI 2 -RE:
EVALUATION OF INFORMED INSERVICE INSPECTION PROGRAM FOR THE THIRD 1 a-YEAR INTERVAL (TAC NO. ME3119)
Dear Mr. Davis:
By letter dated January 20, 2010, with supplement dated August 13, 2010, Detroit Edison (the licensee) submitted for staff review and approval Relief Request RR-A30 for Continuation of Risk-Informed Inservice Inspection application on circumferential welds in Class 1 piping for the third ten-year interval of the inservice inspection program at Fermi 2. The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the licensee's submittal and finds the request provides an acceptable level of quality and safety as discussed in the attached safety evaluation.
Therefore, pursuant to Title 10 of the Code of Federal Regulations 50.55a(a)(3)(i) the NRC staff authorizes the use of Relief Request RR-A30 at Fermi 2 for the third 1 a-year inservice inspection interval which began on May 2, 2009 and ends on May 1, 2019. The NRC staff review and evaluation is contained in the enclosed safety evaluation. If you have any questions, please contact Mahesh Chawla of my staff at (301) 415-8371.
Sincerely, Robert J. Pascarelli, Branch Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-341
Enclosure:
Safety Evaluation cc w/encl: Distribution via ListServ UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RISK-INFORMED INSERVICE INSPECTION PROGRAM FOR THIRD 10-YEAR INTERVAL RELIEF REQUEST NUMBER RR-A30 DETROIT EDISON FERMI 2 DOCKET NUMBER 50-341
1.0 INTRODUCTION
By letter dated January 20, 2010 (Reference 1), supplemented by a letter dated August 2010 (Reference 2), Detroit Edison (the licensee) requested U.S.
Nuclear Commission (NRC) authorization to extend the risk-informed inservice inspection program plan for Fermi 2 to the third ten-year inservice inspection (lSI) interval. The Fermi 2 lSI program for the second ten-year lSI interval was submitted to the NRC in a letter April 30, 2001 (Reference 3). The Fermi 2 RI-ISI program was reviewed and approved by NRC for use in the second 10-year lSI interval in a letter dated September 10, (Reference The licensee considered relevant information since the development of the implemented during the second 1O-year lSI interval and reviewed and updated the program. The licensee's January 20, 2010 submittal requests authorization to extend the program for the third 10-year 151 interval, which began on May 2, 2009, and ends on May
2.0 REGULATORY EVALUATION
Paragraph 50.55a(g) of Title 10 of the Code of Federal Regulations (10 CFR) specifies that lSI of nuclear power plant components shall be performed in accordance with the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). Title 10 CFR 50.55a(a)(3) states that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if (i) the proposed alternatives would provide an acceptable level of quality and safety, or (ii) compliance with the specified Enclosure
-2 requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1,2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the pre-service examination requirements, set forth in the ASME Code,Section XI, "Rules for Inservice Inspection of Nu.clear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of components.
The regulations require that inservice examination of components and system pressure tests conducted during the first 1O-year interval and subsequent intervals comply with the requirements of the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) twelve months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The applicable code of record for the third 10-year lSI interval for Fermi 2 is the ASME Code,Section XI, 2001 Edition through the 2003 Addenda.
The licensee's RI-ISI program, as outlined in Reference 3 was developed in accordance with the methodology contained in the Electric Power Research Institute's (EPRl's) topical report (TR) EPRI TR-112657, Rev. B-A, "Revised Risk-Informed Inservice Inspection Evaluation Procedure, Final Report" (Reference 5), which was reviewed and approved by the NRC staff. The Fermi 2 RI-ISI program is an alternative pursuant to 10 CFR 50.55a(a)(3)(i).
In Reference 1, the licensee requests NRC authorization to extend the RI-ISI program, previously approved for use in the second lSI interval, to the third lSI interval at Fermi 2. The scope of the RI-ISI program is limited to the inspection of ASME Code Class 1, Examination Category B-F and B-J piping welds. The proposed relief is sought for the third 10-year lSI interval, which began on May 2, 2009, and ends on May 1, 2019.
The information provided by the licensee in support of the request has been evaluated and the basis for disposition is documented below. 2.1 Component Identification Code Class:
1 Examination Category:
B-F (Pressure Retaining Dissimilar Metal Welds in Vessel Nozzles) and B-J (Pressure Retaining Welds in Piping) Item Number: B5.10, B5.20 and B9.11 Component Number: All non-exempt Class 1 circumferential piping welds 2.2 Code Requirement for which Relief is Requested Class 1 circumferential piping welds are subject to volumetric and surface examinations as stipulated in ASME Code,Section XI, Table IWB-2500-1, Examination Categories B-F and B-J.
2.3 Licensee's
Basis for Relief Pursuant to 10 CFR 50.55a(a)(3)(i) relief is requested on the basis that the proposed alternative to current ASME Code,Section XI inspection requirements for all Class 1 circumferential piping welds that are subject to volumetric and surface examinations as stipulated in ASME Code,
-Section XI, Table IWB-2500-1, Examination Categories B-F and B-J will provide an acceptable level of quality and safety. The RI-ISI program has been developed in accordance with the NRC Safety Evaluation Report of the EPRI methodology contained in EPRI TR-112657, Informed Inservice Inspection Evaluation Procedure," Revision B-A (Reference 5). The initial Fermi RI-ISI Program was submitted (Reference 3) and approved by NRC (Reference 4) during the second period of the second ten-year lSI interval. Changes between the ASME Code,Section XI, 1989 Edition and the ASME Code,Section XI, 2001 Edition through the 2003 Addenda of the Code were reviewed by the licensee for applicability to this request and found to have no impact on the RI-ISI program. Additionally, the proposed RI-ISI program does not impact the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Report No.
BWRVIP-75-A program. Specifically, the RI-ISI program subsumes the Class 1 integranular stress-corrosion cracking (IGSCC) Category "A" welds. The BWRVIP-75A augmented inspection program for the other piping welds subject to IGSCC is not affected by the RI-ISI program, and are examined at the percentages established in 75-A. TECHNICAL EVALUATION The licensee is requesting relief for continued use of the approved RI-ISI program plan in the third 10-year lSI interval as an alternative to the current ASME Code,Section XI, 2001 Edition through the 2003 Addenda, inspection requirements for Class 1 Examination Category B-F and B-J piping welds. The licensee's process used to develop the RI-ISI program is consistent with the methodology described in EPRI TR-112657. An acceptable RI-ISI program plan is expected to meet the five key principles of risk-informed decisionmaking, discussed in Regulatory Guide (RG) 1.178, "An Approach for Plant-Specific Risk-Informed Decision Making: Inservice Inspection of Piping" (Reference 6), Standard Review Plan 3.9.8 (Reference 7), NUREG-0800 Chapter 19 (Reference 8), and the EPRI TR-112657, Rev. B-A, as stated below. The proposed change meets the current regulations unless it is explicitly related to a requested exemption or rule change. The proposed change is consistent with the defense-in-depth philosophy. The proposed change maintains sufficient safety margins. When proposed changes result in an increase in core damage frequency and/or large early release frequency (LERF), the increases should be small and consistent with the intent of the Commission's Safety Goal Policy Statement. The impact of the proposed change should be monitored by using performance measurement strategies.
-The first principle is met in this relief request because an alternative lSI program may be authorized pursuant to 10 CFR 50.55a(3)(i) and therefore, an exemption request is not required. The second and third principles require assurance that the alternative program is consistent with the defense-in-depth philosophy and that sufficient safety margins are maintained, respectively.
The methodology used to develop the third 1O-year RI-ISI program interval is unchanged from the methodology approved for use in the second 1O-year RI-ISI interval program.
Assurance that the second and third principles are met is based on the application of the approved methodology and not on the particular inspection locations selected. Therefore, the second and third principles are met. The fourth principle requires an estimate of the change in risk between the proposed risk-informed program and the program the licensee would otherwise be required to implement. The topical report (EPRI TR-112657) requires that a change in risk measurement must consider the discontinuance of ASME Code required inspections, as well as any new inspections resulting from the application of its methodology. The licensee stated in Reference 1 that it had completed a Gap Analysis using RG 1.200 Revision 1 (Reference 9) and the associated ASME Standard (Reference 10) in 2008. The reported results indicate that many (i.e., 67) of the ASME standard supporting requirements were not-met or assigned a capability category less than II. The licensee reported that it had reviewed these gaps and concluded that none of the gaps was deemed significant with respect to the RI-ISI program. The NRC staff recognizes that the EPRI method in Reference 5 uses the quantitative results of the probabilistic risk assessment (PRA) as order-of-magnitude estimates for several risk and reliability parameters and to support the assignment of segments into three broad consequence categories.
Inaccuracies in the models or in assumptions large enough to invalidate the broad categorizations developed to support RI-ISI or the change in risk estimates should have been identified during the gap analyses or the licensee's evaluation of the gaps. Minor errors or inappropriate assumptions could potentially affect only the consequence categorization of a few segments and will not invalidate the general results or conclusions. Therefore, the NRC staff finds the licensee has assessed the technical adequacy of its PRA using the appropriate version of RG 1.200 and the quality of the PRA is sufficient to support the proposed RI-ISI program. The NRC staff has previously determined that it is not necessary to develop a new deterministic ASME program for each new 1O-year interval but, instead, it is acceptable to compare the new proposed RI-ISI program with the last deterministic ASME program. The licensee states in Reference 1 that a new Risk Impact Analysis was performed, and the revised program continues to satisfy the acceptance criteria of RG 1.174 and EPRI TR-112657 when compared to the last deterministic Section XI inspection program. Thus, the staff finds that the licensee's analysis provides assurance that the fourth key principle is met. The fifth principle of risk-informed decisionmaking requires that the impact of the proposed change be monitored by using performance measurement strategies. The Fermi 2 Relief Request RR-A30 and the associated response to the NRC's Request for Additional Information (RAI) state that the Fermi 2 RI-ISI program was developed in accordance with the EPRI methodology contained in EPRI Topical Report TR-112657, Revision B-A, with identified differences, and with additional guidance taken from ASME Code Cases N-578 and N-578-1. This program is a living program and, as such, is subject to periodic reviews. The licensee states that, to satisfy the periodic review requirements, an evaluation and update was performed
-5 in accordance with the Nuclear Energy Institute (NEI) document 04-05, "Living Program Guidance To Maintain Risk-Informed Inservice Inspection Programs for Nuclear Power Plant Piping Systems," published in April, 2004. In Reference 1, the licensee states that, in accordance with NEI 04-05, the following aspects were considered during the review: Plant Examination Results Piping Failures Plant Specific Failures Industry Failures PRA Updates Plant Design Changes Physical Changes Programmatic Changes Procedural Changes Changes in Postulated Conditions Physical Conditions Programmatic Conditions The licensee provided a table that identified the number of welds added to or deleted from the originally approved RI-ISI program. Though there were a limited number of "like-for-like" substitutions that were made during the implementation of the program due to inaccessibility and as low as reasonably achievable, changes to the program were attributed to the following three specific actions (as stated): Condition Assessment Resolution Document (CARD) 07-25329 evaluated limitations of Weld 1 02-304A and determined that it was not inspectable. There was no other weld in Jet Pump Instrumentation (JPI) that was a suitable replacement, so an additional weld in Reactor Coolant Recirculation (RCR) with the same degradation mechanism and a higher Risk Category ([Weld] 2-303J) was selected as a replacement. This caused the total for the JPI Risk Category 4(2) to decrease by one and the total for RCR Risk Category 2(2) to increase by one. Condition Assessment Resolution Document (CARD) 07-263467 evaluated limitations of Weld 4-303A and determined that it was not inspectable. It was replaced by Weld 203A, which has the same degradation mechanism, but a higher risk. This caused the total for RCR Risk Category 4(2) to decrease by one and the total for RCR Risk Category 2(2) to increase by one. Condition Assessment Resolution Document (CARD) 07-26900 reported that a first time Risk-Informed Inservice Inspection (RI-ISI) weld selection was scheduled for examination (FW-RD-2-A1-W1, 4" sweepolet to cap). The examiner reported the scan would be limited due to cap configuration and could not be credited as a full examination for the RI-ISI Program. This weld was not examined, and another selection was required. There were no 4" selections that were not also limited. The action taken to correct the condition was to select two adjacent single side access welds (SW-RS-2-A3-W4 and SW-RS-2-A3-W5) to replace the original 4" selection in the Recirculation system. The new selections have the same degradation mechanism and
-6 consequence of failure as the original welds. CARD 07-26900 also reported that a weld selection previously examined (FW-RD-B1-W1) was likely impacted by current expectations for weld flatness. It would be limited due to cap configuration and should not be credited as a full examination for the RI-ISI Program.
Another selection was required. There were no other 4" selections that were not also limited. The action taken to correct the condition was to select two adjacent single side access welds (SW-RS-2-B3-W4 and SW-RS-2-B3-W5) to replace the original 4" selection in the Recirculation System. The new selections have the same degradation mechanism and consequence of failure as the original welds. The actions caused the total for the RCR Risk Category 4(2) to increase by two. In its RAI, the NRC staff asked the licensee to explain why the combination of coverage for two welds as discussed in action 3 above is an adequate/appropriate substitution for one weld that was originally selected for the RI-ISI program. The licensee responded in Reference 2 that it is the intent of the risk program to effectively examine not less than 10 percent of the welds with applicable degradation mechanisms and, that in the case mentioned above, one weld examined from two sides is technically equivalent to two welds being examined from one side. Since the pipe size is the same for all the welds involved, an equivalent amount of weld volume is examined using this approach.
Additionally, the new selections are in the same system, and have the same degradation mechanism and consequence of failure as the original welds.
Therefore, the staff finds this substitution acceptable as it provides an acceptable level of quality and safety.
Though the overall number of elements inspected changed as a result of the above three actions, the NRC staff has determined that the licensee's approach of ensuring that new selections have the same or higher Risk Category provides an acceptable level of quality and safety. The analyses and changes reported by the licensee in its submittal demonstrate that the RI-ISI program is a living program that is being periodically updated and therefore the NRC staff concludes that the fifth key principle which provides that risk-informed applications should include performance monitoring and feedback provisions is met. Based on the above discussion, the NRC staff concludes that the five key principles of informed decisionmaking are ensured by the licensee's proposed third 10-year RI-ISI program, and therefore the proposed program for the third 10-year lSI interval is acceptable.
4.0 CONCLUSION
S Based on the information provided in the licensee's submittals, the NRC staff has determined that the proposed alternative provides an acceptable level of quality and safety, and therefore the proposed alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(i) for the third 10-year lSI inspection interval at Fermi 2, which began on May 2, 2009, and ends on May 1, 2019. All other ASME Code,Section XI requirements for which relief was not specifically requested and approved remain applicable.
-5.0 Letter NRC-1 0-0004 from J. H. Plona, Detroit Edison to U.S.
Nuclear Commission, "Submittal of the Inservice Inspection/Nondestructive Program Relief Requests for the Third Ten-year Interval," dated January 20, Letter NRC-10-0060 from J. H. Plona, Detroit Edison to U.S.
Nuclear Regulatory Commission, "Response to Request for Additional Information Regarding Relief Requests RR-A37 and RR-A30 for the Inservice Inspection/Nondestructive Examination Program Third Ten-Year Interval," dated August 13, 2010. Letter NRC-01-0038 from P. Fessler, Detroit Energy, to U.S.
Nuclear Regulatory Commission, "Submittal of Relief Request (RR-A30) for Applying the Risk-Informed Inservice Inspection Program Along with BWRVIP-75 Weld Examination Schedule," dated April 30, 2001. Letter from U.S.
Nuclear Regulatory Commission to W. T. O'Connor, Jr., "Fermi 2 -Relief Requrest (RR-30) Concerning Risk-Informed Inservice Inspection Program and Boiling Water Reactor Vessel and Internals Project Report 75 75) for the Second Inservice Insepction Interval (TAC Number MB1813)," dated September 10, 2001. EPRI TR-112657, Revision B-A, "Revised Risk-Informed Inservice Inspection Evaluation Procedure, Final Report," December 1999. Regulatory Guide 1.178, "An Approach for Plant-Specific Risk-Informed Decisionmaking for Inservice Inspection of Piping," September 2003. NRC NUREG-0800, Chapter 3.9.8, "Standard Review Plan for the Review of Informed Inservice Inspection of Piping," September 2003. NRC NUREG-0800, Chapter 19, "Use of Probabilistic Risk Assessment in Plant-Specific, Risk-Informed Decisionmaking:
General Guidance," November 2002. Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 1, January 2007. 10. ASME RA-Sb-2005, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," Addendum B to ASME RA-S-2002, ASME, New York, New York, December 30, 2005.
Principal Carol Nove, NRR Steve Dinsmore, NRR Date: October 1, 2010 October 1, 2010 Mr. Jack M. Davis Senior Vice President and Chief Nuclear Officer Detroit Edison Company Fermi 2 -210 NOC 6400 North Dixie Highway Newport, MI 48166 RELIEF REQUEST FOR FERMI 2 -RE:
EVALUATION OF INFORMED INSERVICE INSPECTION PROGRAM FOR THE THIRD 10-YEAR INTERVAL (TAC NO. ME3119)
Dear Mr. Davis:
By letter dated January 20, 2010, with supplement dated August 13, 2010, Detroit Edison (the licensee) submitted for staff review and approval Relief Request RR-A30 for Continuation of Risk-Informed Inservice Inspection application on circumferential welds in Class 1 piping for the third ten-year interval of the inservice inspection program at Fermi 2. The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the licensee's submittal and finds the request provides an acceptable level of quality and safety as discussed in the attached safety evaluation.
Therefore, pursuant to Title 10 of the Code of Federal Regulations 50.55a(a)(3)(i) the NRC staff authorizes the use of Relief Request RR-A30 at Fermi 2 for the third 1O-year inservice inspection interval which began on May 2, 2009 and ends on May 1, 2019. The NRC staff review and evaluation is contained in the enclosed safety evaluation. If you have any questions, please contact Mahesh Chawla of my staff at (301) 415-8371.
Sincerely, IRA! Robert J. Pascarelli, Branch Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-341
Enclosure:
Safety Evaluation cc w/encl: Distribution via ListServ DISTRIBUTION:
PUBLIC LPL3-1 RlF RidsNrrDorlLpl3-1 Resource RidsRgn3MailCenter Resource RidsNrrDirsltsb Resource RidsNrrPMFermi2 Resource SDinsmore, NRR RidsAcrsAcnw_MailCTR Resource RidsNrrLABTully Resource DMerzke, EDO Rill RidsNrrDorlDpr Resource RidsOgcRp Resource CNove, NRR SDinsmore, NRR ADAMS Accession Number" ML*Memo -ML 102500106 OFFICE NRRlLPL3-1/PM NRRlLPL3-1/LA NRRlCPNB/BC NRRlDRAAlAPLA NRRlLPL3-1/BC NAME MChawla BTullV TLupold* DHarrison RPascarelii DATE 09/22/10 09/22/10 09/03/10 09/22/10 10101/10 OFFICIAL RECORD COpy