ML16355A429
| ML16355A429 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 01/06/2017 |
| From: | Stephen Koenick Plant Licensing Branch 1 |
| To: | Brian Sullivan Entergy Nuclear Operations |
| Render D, NRR/DORL/LPL1 | |
| References | |
| CAC MF8361 | |
| Download: ML16355A429 (12) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 6, 2017 Mr. Brian Sullivan Site Vice President Entergy Nuclear Operations, Inc.
James A. FitzPatrick Nuclear Power Plant P.O. Box 110 Lycoming, NY 13093
SUBJECT:
JAMES A. FITZPATRICK NUCLEAR POWER PLANT - RELIEF REQUEST FOR PROPOSED ALTERNATIVE FOR THE IMPLEMENTATION OF BWRVIP-05 (CAC NO. MF8361)
Dear Mr. Sullivan:
By letter dated September 8, 2016, as supplemented by letter dated November 9, 2016, Entergy Nuclear Operations, Inc. (the licensee) submitted a request to the U.S. Nuclear Regulatory Commission (NRC) for relief from the use of alternatives to certain American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI requirements at the James A. Fitzpatrick Nuclear Power Plant (JAFNPP).
Specifically, pursuant to Title 1O of the Code of Federal Regulations (10 CFR) 50.55a(z)(1 ), the licensee requested to use the proposed alternative on the basis that the alternative provides an acceptable level of quality and safety.
As set forth in the enclosed Safety Evaluation, the NRC staff determined that the licensee's proposed alternative provides an acceptable level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(1 ). Therefore, the staff authorizes the proposed alternative for the remainder of the fourth inservice inspection interval and through the period of extended operation at JAFNPP, which ends on October 17, 2034.
All other requirements of Section XI of the ASME Code for which relief was not specifically requested and approved in the subject relief requests remain applicable, including third party review by the Authorized Nuclear lnservice Inspector.
B. Sullivan If you have any questions, please contact the Project Manager, Diane Render, Ph.D., at 301-415-3629 or by e-mail to Diane.Render@nrc.gov.
Sincerely, Stephen S. Koenick, Acting Chief Plant Licensing Branch 1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-333
Enclosure:
Safety Evaluation cc w/encl: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST NO. 19 ENTERGY NUCLEAR FITZPATRICK. LLC AND ENTERGY NUCLEAR OPERATIONS. INC.
JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333
1.0 INTRODUCTION
By application dated September 8, 2016 (Reference 1), as supplemented by letter dated November 9, 2016 (Reference 2), Entergy Nuclear FitzPatrick, LLC (the licensee) submitted a proposed alternative to the inservice inspection (ISi) requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) for the reactor pressure vessel (RPV) shell welds at the James A. FitzPatrick Nuclear Power Plant (JAFNPP), pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Part 50, paragraph 50.55a(z)(1 ).
Specifically, the licensee proposes to permanently eliminate the volumetric examination requirements of Section XI of the ASME Code for RPV circumferential welds (ASME Code,Section XI, Examination Category B-A, "Pressure Retaining Welds in Reactor Vessel," Item No. B1 .11 ), for the remainder of JAFNPP's fourth ISi interval and through the period of extended operation (PEO). Details of the licensee's proposed alternative are in Section 3.3 of this safety evaluation (SE). Section 50.55a(z)(1) of 10 CFR requires the licensee to demonstrate that the proposed alternative provides an acceptable level of quality and safety. Section 2.0 of this SE provides a discussion of the attendant regulations and requirements. The proposed alternative is based on Electric Power Research Institute proprietary report TR-105697, "BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05)" (Reference 3), and follows the implementation guidance in the NRC staffs July 28, 1998, SE of the report (Reference 4).
2.0 REGULATORY REQUIREMENTS AND GUIDANCE 2.1 Requirements of 10 CFR The RPV shell welds at JAFNPP are ASME Code, Class 1 components, whose ISi requirements are performed in accordance with Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," of the ASME Code and applicable edition and addenda, as required by 10 CFR 50.55a(g). 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 Enclosure
provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, to the extent practical, within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(a)(1 )(ii), 12 months prior to the start of the 120-month interval, subject to the limitations and modifications in 10 CFR 50.55a(b)(2). The Code of record for JAFNPP for the fourth 10-year ISi interval is the 2003 Addenda to the 2001 Edition of the ASME Code,Section XI.
2.2 NRC Staff's Safety Evaluation of BWRVIP-05 The technical basis for the proposed alternative is BWRVIP-05, which calculates conservative conditional probabilities of failures for RPV welds. The basic principle for justifying the proposed alternative is to demonstrate that the conditional probabilities of failures of specific RPV welds are lower than the conservative values determined in BWRVIP-05. By letter dated February 22, 2000 (Reference 5), the NRC staff issued an SE that approved the proposed alternative for JAFNPP's initial licensing term. However, Section 3, "Conclusions," of the July 28, 1998, SE of BWRVIP-05 stated that since the failure frequency for the limiting circumferential weld could significantly increase through the PEO, the NRC staff will be requesting plants to perform plant-specific assessments that consider weld chemistry and neutron fluence at the end of the PEO. In addition, the July 28, 1998, SE stated that licensees may also request relief from the requirements of the ASME Code,Section XI, Examination Category B-A, Item No. 81 .11 applicable through the end of the PEO by demonstrating the following:
(1) At the expiration of the license, the circumferential welds will continue to satisfy the limiting conditional failure probability for circumferential welds in the NRC staff's July 28, 1998, SE of BWRVIP-05.
(2) Licensees have implemented operator training and established procedures that limit the frequency of cold over-pressure events to the amount specified in the NRC staff's July 28, 1998, SE of BWRVIP-05.
Section 4, "Implementation," of the July 28, 1998, SE stated that if the axial weld examinations reveal an active mode of degradation, the examination of the circumferential welds shall be performed.
3.0 TECHNICAL EVALUATION
3.1 ASME Code Requirements The specific examination requirement for RPV shell welds is volumetric examination of essentially 100 percent of the weld length of the volume defined in Figure IWB-2500-1, "Vessel Shell Circumferential Weld Joints," of the ASME Code,Section XI, as specified in Table IWB-2500-1, "Examination Categories," of the ASME Code,Section XI, Examination Category B-A, Item No. 81 .11.
3.2 ASME Code Components Affected
ASME Code Class: 1 Examination Category: B-A Item Number: 81 .11 Component: RPV Circumferential Welds Component Numbers: VC-1-2, VC-2-3, VC-3-4, and VC-4-BH-1 3.3 Licensee's Proposed Alternative Pursuant to 10 CFR 50.55a(z)(1 ), the licensee proposes the following alternative:
The alternative plan would require performance of RPV vertical weld examinations and incidental examination of 2 to 3 percent of the intersecting circumferential shell welds to the maximum extent possible based on accessibility. The circumferential welds would be permanently deferred until plant renewed operating license expiration. This alternative aligns with BWRVIP-05.
The axial weld seams (Examination Category B-A, Item No. 81 .12) and their intersection with the associated circumferential weld seams will be examined in accordance with ASME Section XI except where specific relief is granted when essentially 100% (>90%) coverage cannot be obtained.
3.4 Licensee's Basis 3.4.1 Satisfying the Conditional Probability of Failure for Welds Through the Period of Extended Operation The licensee performed a time-limited aging analysis (TLAA) of the RPV circumferential welds through the PEO in its 2006 license renewal application (LRA) (Reference 6) for JAFNPP. The NRC staff approved the LRA in 2008, as documented in NUREG-1905, "Safety Evaluation Report Related to the License Renewal of James A. FitzPatrick Nuclear Power Plant" (Reference 7). The NRC staff's evaluation of the TLAA of the RPV circumferential welds through the end of the PEO in enclosed in Section 4.2.5 of NUREG-1905, "Reactor Vessel Circumferential Weld Inspection Relief'.
The licensee included information from Section 4.2.5.2 of NUREG-1905, "Staff Evaluation," in Section 5 of its submittal. Specifically, the licensee presented information from Table 4.2.5-1 of NUREG-1905, "Comparison of NRC and JAFNPP 54 EFPY Mean RT NDT Calculations to the 64 EFPY Mean RT NDT Calculations for the Limiting Combustion Engineering Owners Group Case Study on BWRVIP-05," in Table 1 of the submittal. Table 1 of the submittal compares conditional probabilities of failures for the RPV circumferential welds, computed for a bounding case of 64 effective full power years (EFPY), from the SE of BWRVIP-05, dated July 28, 1998, with those from NRC staff's and licensee's calculations for JAFNPP for 54 EFPY. JAFNPP's operational condition at the end of the PEO is represented by 54 EFPY. Note 2 of Table 1 of the submittal includes the acceptance criterion for the conditional probability of failure that the NRC staff approved: if JAFNPP's mean reference temperature (nil ductility transition) (RT NDT) for the limiting RPV circumferential weld (weld 1-240) is less than the mean RT NDT of the
bounding case, then JAFNPP's conditional probability for failure for weld 1-240 is less than that of the bounding case. Table 1 shows that since JAFNPP's mean RTNDT of 81.1 degrees Fahrenheit (°F) for weld 1-240 is less than that of the bounding mean RTNDT of 128.5 °F, the conditional probability of failure for weld 1-240 is less than that of the bounding case. Note 2 in Table 1 of the submittal further states that plants that meet the above acceptance criteria may conclude that the conditional probability of failure for the limiting RPV circumferential weld is low enough to justify elimination of the required ASME Code volumetric examinations.
The examination of the circumferential welds shall be performed if the axial weld examinations reveal an active mode of degradation. In the submittal, the licensee included its response to a previous request for additional information (RAI) regarding confirmation on whether previous volumetric examinations of the RPV axial welds showed any indication of cracking or any other age-related degradation. The licensee's response to RAI 4.2.5-2 from NUREG-1905 stated that no unacceptable inservice examination indications have been found on the RPV circumferential or axial welds.
The NRC staff noted that the licensee referenced the NRC staff's evaluation (Reference 5) for the conditional probabilities of failures for the RPV circumferential welds applicable to JAFNPP's initial licensing term.
3.4.2 Lower Head Events Section 6 of the licensee's submittal includes information on two recent lower head events that violated the JAFNPP Limiting Conditions for Operation (LCOs) in Section 3.4.9 of the Technical Specifications (TSs). Specifically, during any 1-hour period, the subject lower head events caused a violation of the TS requirement for reactor coolant system temperature change during heatup or cooldown of s 100 °F. The licensee stated that the two lower head events were entered in its correction action program, evaluated, and found to be acceptable. Evaluations of the two lower head events were also included as enclosures in the submittal.
3.5 NRC Staff's Evaluation 3.5.1 Satisfying the Conditional Probability of Failure for Welds Through the Period of Extended Operation Circumferential Welds As mentioned in Section 3.4.1 of this SE, the licensee included Table 4.2.5-1 of NUREG-1905 as Table 1 in its submittal. Table 1 contains the information to justify the elimination of the JAFNPP RPV circumferential welds during the PEO, based on BWRVIP-05. Furthermore, the licensee has shown that the JAFNPP conditional probability of failure for the limiting RPV circumferential weld is bounded by the BWRVIP-05 analysis and, therefore, elimination of the required ASME Code volumetric examinations through the PEO is justified.
The NRC staff approved the licensee's evaluation in Table 1 in 2008. The NRC staff notes that surveillance capsule test data that are withdrawn and/or tested after 2008 can potentially impact the values in Table 1 and invalidate them, especially values of mean RT Nor. One of the conditions
listed in Section 2.V of the JAFNPP Renewed License No. DPR-59, "Capsule withdrawal schedule," is that any changes to the capsule withdrawal schedule must be approved by the NRC prior to implementation. Furthermore, any changes to the capsule withdrawal schedule may also impact the values in Table 1. JAFNPP's capsule withdrawal schedule is contained in the integrated surveillance program (ISP) in BWRVIP-116 (Reference 8), as indicated in JAFNPP's updated final safety analysis report (UFSAR). Section 16.10.1.26, "Reactor Vessel Surveillance Program," indicates that JAFNPP's surveillance schedule includes the PEO.
On October 18, 2016 (Reference 9), the NRC staff issued an RAI to confirm whether there have been any changes to the JAFNPP surveillance capsule withdrawal schedule or any surveillance test results since 2008 that could invalidate the technical basis for the proposed alternative. By letter dated November 9, 2016 (Reference 2), the licensee responded that there have been no changes to the capsule withdrawal schedule since the JAFNPP LRA in 2006 and clarified that JAFNPP's capsule withdrawal schedule is in accordance with Table 4-8, "ISP Test Matrix Results," in Revision 1-A of BWRVIP-86 (Reference 10). Revision 1-A of BWRVIP-86 merged the information in BWRVIP-116 into a single updated ISP. The NRC staff verified that Table 4-8 in Revision 1-A of BWRVIP-86 is equivalent to Table 3-3 in BWRVIP-116. In addition, the licensee stated that it evaluated an ISP representative surveillance plate material in 201 O and documented the results in the ISP data source book, BWRVIP-135 (Reference 11). The licensee determined in its evaluation that there have been no changes to the JAFNPP RPV material properties nor to pressure-temperature limit curves. The NRC staff verified that testing of the 2010 ISP representative surveillance plate is in Table 4-7, "Detailed Test Plan by Plant,"
in Revision 1-A of BWRVIP-86 for JAFNPP.
The NRC staff determined that the values and evaluation in Table 1 of the licensee's submittal for the RPV circumferential welds are valid through the end of JAFNPP's PEO for the following reasons, as supported by the licensee's response to the NRC staff's RAI:
- The licensee has evaluated and documented surveillance capsule tests since 2008 and has determined that the tests do not impact the JAFNPP material properties, among which is mean RT NDT, the material property used directly in the criterion for acceptability of the conditional probability of failure values for the JAFNPP RPV circumferential welds.
Therefore, the licensee has satisfied item 1 in the SE of BWRVIP-05.
Axial Welds According to Section 3.3 of this SE, the JAFNPP RPV axial welds will be examined in accordance with the requirements of Section XI of the ASME Code. With respect to conditional probabilities of failure, the licensee also performed a TLAA for the RPV axial welds through the PEO in its 2006 LRA. The NRC staff evaluated and approved it in Section 4.2.6 of NUREG-1905, "Reactor Vessel Axial Weld Failure Probability," in 2008 (Reference 7). The licensee has shown in the TLAA that the JAFNPP's conditional probability for failure for the axial
welds through the PEO is acceptable. The NRC staff's review of the axial welds TLAA consisted of the validity of the analysis since the 2008 NRC staff approval. The NRC staff determined that the TLAA that evaluated the conditional probabilities of failure for RPV axial welds are still valid through JAFNPP's PEO for the same reasons the TLAA for the RPV circumferential welds are valid, as discussed above in Section 3.5.1 of this SE, under "Circumferential Welds."
Regarding any signs of degradation in the RPV axial welds, the licensee referred to its response to RAI 4.2.5-2 in NUREG-1905, which stated that, "no unacceptable inservice examination indications have been found on reactor vessel welds (circumferential or axial)." The NRC staff has found the response acceptable. Therefore, the licensee has satisfied the implementation section of the SE of BWRVIP-05.
3.5.2 Lower Head Events In the submittal, the licensee included two events in the RPV lower head that violated the JAFNPP TS requirements of s 100 °F over any 1-hour period during heatup and cooldown. In both events, instrumentation in the RPV bottom head recorded a maximum increase in metal temperature over a 1-hour period of 124.9 °F for one event and 125.7 °F for the other event.
The NRC staff considers both events as heatup events because the temperature increased in both events. The NRC staff's evaluation of these two lower head events focuses on its potential impact on the proposed alternative to eliminate the ASME Code examination requirements of the RPV circumferential shell welds through the PEO.
Heatup events generate compressive stresses on the inside surface of the RPV shell welds. A heatup event would, therefore, have no adverse impact on the values of conditional probability of failure for the type of flaws postulated in the BWRVIP-05 probabilistic fracture mechanics analysis. Therefore, the NRC staff determined that the two RPV lower head events included in the licensee's submittal have no relevant impact on the technical basis for the proposed elimination of the ASME Code volumetric examination requirements for the RPV circumferential shell welds at JAFNPP.
3.5.3 High Pressure Sources For BWRVIP-05, item 2, the licensee identified the high pressure sources, which include the feedwater system, high pressure coolant injection (HPCI) system, reactor core isolation cooling (RCIC) system, control rod drive (CRD) system, reactor water cleanup (RWCU) system, and the standby liquid control (SLC) system. The NRC staff reviewed JAFNPP's UFSAR and determined that the licensee has identified the high pressure injection sources at the plant.
The licensee stated that the feedwater system, HPCI system, and RCIC system are steam turbine driven and, therefore, it is not plausible for these systems to contribute to an over-pressurization event while the unit is in cold shutdown. The NRC staff has reviewed the JAFNPP's UFSAR and determined that these systems identified by the licensee are steam turbine driven and, therefore, it is not plausible that they will contribute to an over-pressurization event while the unit is in cold shutdown.
The licensee stated that the SLC system has no automatic starts associated with the system and requires the operators to manually start the system from the control room or from the local test station. The licensee also stated that the injection rate of the SLC pump is approximately 50 gallons per minute, which gives the operators ample time to control reactor pressure in the case of an inadvertent injection. The NRC staff reviewed JAFNPP's UFSAR and determined that the SLC system, which is driven by two positive-displacement pumps, requires a manual operator action to initiate the system, and the injection rate provided by the licensee is consistent with the UFSAR. Given that the SLC is only operated in plant emergency situations or during testing conditions, and the injection rate is small relative to the total volume of the reactor vessel, the NRC staff determined that the licensee adequately justified that the design, training, and procedures, limit the cold over-pressure event with respect to the SLC system.
The licensee stated that during normal cold shutdown conditions, RPV level and pressure are controlled with the CRD and RWCU systems using a "feed and bleed" process and the RPV is not taken water solid during these times. Additionally, "feed and bleed" is used during pressure testing of the RPV. The NRC staff reviewed the applicable sections of the UFSAR, TSs, and TS Basis and determined that the design, training, and procedures limit the cold over-pressure event with respect to the CRD and RCWU systems.
The NRC staff determined that the licensee identified all sources of high pressure injection and adequately justified that the procedures and training limit the cold over-pressure events.
Additionally, the NRC staff compared this relief request to a previous relief request (Reference 5), and determined that there are no additional high pressure injection sources or changes to the procedures or training that would increase the likelihood of the cold over-pressure event. Based on its review, the NRC staff determined that the licensee satisfied item 2 in the SE of BWRVIP-05.
4.0 CONCLUSION
As set forth above, the NRC staff determined that the licensee's proposed alternative provides an acceptable level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(1 ). Therefore, the staff authorizes the proposed alternative for the remainder of the fourth ISi interval, and through the PEO at JAFNPP, which ends on October 17, 2034.
All other requirements of Section XI of the ASME Code for which relief was not specifically requested and approved in the subject relief requests remain applicable, including third party review by the Authorized Nuclear lnservice Inspector.
5.0 REFERENCES
- 1. Drews, W.C., Entergy Nuclear Operations, Inc., letter to the U.S. Nuclear Regulatory Commission, "Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1 ),
Implementation of BWRVIP-05 at James A. FitzPatrick Nuclear Power Plant," dated September 8, 2016 (Agencywide Documents Access and Management System (ADAMS)
Accession No. ML16252A473).
- 2. Drews, W.C., Entergy Nuclear Operations, Inc., letter to the U.S. Nuclear Regulatory Commission, "Response to Request for Additional Information - Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1), Implementation of BWRVIP-05 at James A.
FitzPatrick Nuclear Power Plant," dated November 9, 2016 (ADAMS Accession No. ML16314E532).
- 3. Electric Power Research Institute Report TR-105697, "BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05),"
September 1995 (EPRI Proprietary) (ADAMS Accession No. ML032200246 (non-proprietary version)).
- 4. Lainas, Gus C., U.S. Nuclear Regulatory Commission, letter to Carl Terry, BWRVIP Chairman, "Final Safety Evaluation of the BWR Vessel and Internals Project BWRVIP-05 Report (TAC No. M93925)," dated July 28, 1998 (ADAMS Legacy Library No. 9808040037).
- 5. Gamberoni, Marsha K., U.S. Nuclear Regulatory Commission, letter to James Knubel, Power Authority of the State of New York, "Relief Request No. 17 - Request for Relief from the Requirements of 10 CFR 50.55a(g)(6)(ii)(A)(2) for Augmented Inspection of the Circumferential Welds in the Reactor Vessel of the James A. FitzPatrick Nuclear Power Plant (TAC No. MA6215)," dated February 22, 2000 (ADAMS Accession No. ML003685801).
- 6. James A. FitzPatrick Nuclear Power Plant License Renewal Application, dated July 31, 2006 (ADAMS Package Accession No. ML062140129).
- 7. U.S. Nuclear Regulatory Commission Report NUREG-1905, "Safety Evaluation Report Related to the License Renewal of James A FitzPatrick Nuclear Power Plant," April 2008 (ADAMS Accession No. ML081510826).
- 8. Electric Power Research Institute Report TR-1007824, "BWRVIP-116: BWR Vessel and Internals Project, Integrated Surveillance Program (ISP), Implementation for License Renewal," July 2003 (EPRI Proprietary).
- 9. Render, Diane, U.S. Nuclear Regulatory Commission, letter to the James A FitzPatrick Nuclear Power Plant Site Vice President, Entergy Nuclear Operations, Inc., "James A FitzPatrick Nuclear Power Plant- Request for Additional Information Re: Relief Request for Proposed Alternative for the Implementation of BWRVIP-05 (CAC No. MF8361 ),"dated October 18, 2016 (ADAMS Accession No. ML16280A573).
- 10. Electric Power Research Institute Report TR-1025144, "BWRVIP-86, Revision 1-A: BWR Vessel and Internals Project, Updated BWR Integrated Surveillance Program (ISP),
Implementation Plan," October 2012 (EPRI Proprietary).
- 11. Electric Power Research Institute Report TR-3002003144, "BWRVIP-135, Revision 3: BWR Vessel and Internals Project, Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations," December 2014.
Principal Contributors: D. Dijamco J. Borromeo Date: January 6, 2017
B. Sullivan If you have any questions, please contact the Project Manager, Diane Render, Ph.D., at 301-415-3629 or by e-mail to Diane.Render@nrc.gov.
Sincerely, IRA/
Stephen S. Koenick, Acting Chief Plant Licensing Branch 1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-333
Enclosure:
Safety Evaluation cc w/encl: Distribution via Listserv DISTRIBUTION:
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DATE 12/16/2016 1/06/2017 11/30/2016 12/05/2016 1/06/2017 1/06/2017 OFFICIAL RECORD COPY