ML20311A658
| ML20311A658 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 11/10/2020 |
| From: | Nancy Salgado Plant Licensing Branch III |
| To: | Payne F Energy Harbor Nuclear Corp |
| Wall S | |
| References | |
| EPID L-2020-LLR-0002, EPID L-2020-LLR-0003, EPID L-2020-LLR-0005 | |
| Download: ML20311A658 (13) | |
Text
November 10, 2020 Mr. Frank R. Payne Site Vice President Energy Harbor Nuclear Corp.
Perry Nuclear Power Plant P.O. Box 97, Mail Stop A-PY-A290 Perry, OH 44081-0097
SUBJECT:
PERRY NUCLEAR POWER PLANT, UNIT NO. 1 - ISSUANCE OF RELIEF RE: PROPOSED ALTERNATIVE REQUEST ASSOCIATED WITH FOURTH 10-YEAR INSERVICE INSPECTION INTERVAL (EPIDS L-2020-LLR-0002, -
0003, AND -0005)
Dear Mr. Payne:
By letter dated January 6, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20006D984), and supplemented by letters dated June 2, July 15, and November 2, 2020 (ADAMS Accession Nos. ML20154K444, ML20197A212, and ML20308A436, respectively), FirstEnergy Nuclear Operating Company (FENOC) submitted relief requests (RRs) Nos. PT-001, Revision 3, IR-054, Revision 2, IR-056, Revision 3, and IR-060, Revision 0, to the U.S. Nuclear Regulatory Commission (NRC) for the use of alternatives to the pressure test requirements in American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code)Section XI, Rules for lnservice Inspection of Nuclear Power Plant Components, at Perry Nuclear Power Plant (PNPP) associated with the fourth 10-year interval inservice inspection (ISI) interval.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1), the licensee requested to use the proposed alternative in RRs PT-001, Revision 3, IR-054, Revision 2, and IR-060, Revision 0, on the basis that the proposed alternative will provide an acceptable level of quality and safety.
The NRC staff has reviewed the subject request and concludes, as set forth in the enclosed safety evaluation, that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(1). Therefore, the NRC staff authorizes the proposed alternatives in RRs PT-001, Revision 3, IR-054, Revision 2, and IR-060, Revision 0, for the fourth 10-year ISI interval at PNPP which began May 18, 2019, and is scheduled to expire May 17, 2029.
All other ASME Code requirements for which relief was not specifically requested and approved in the subject requests for relief remain applicable.
The RR identified as IR-056, Revision 3, will be handled under separate NRC correspondence.
F. Payne If you have any questions, please contact the Project Manager, at 301-415-2855 or e-mail at Scott.Wall@nrc.gov.
Sincerely, Digitally signed by Joel S.
Joel S. Wiebe Date: 2020.11.10 Wiebe 10:49:07 -05'00' Nancy L. Salgado, Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-440
Enclosure:
Safety Evaluation cc: Listserv
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION 10 CFR 50.55a REQUESTS PT-001, IR-054, AND IR-060 RELIEF REQUEST FOR PUMP AND VALVE INSERVICE TESTING PROGRAM FOURTH 10-YEAR INTERVAL INSERVICE TESTING INTERVAL PERRY NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-440
1.0 INTRODUCTION
By letter dated January 6, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20006D984), and supplemented by letters dated June 2, July 15, and November 2, 2020 (ADAMS Accession Nos. ML20154K444, ML20197A212, and ML20308A436, respectively), FirstEnergy Nuclear Operating Company (FENOC) submitted relief requests (RRs) Nos. PT-001, Revision 3, IR-054, Revision 2, IR-056, Revision 3, and IR-060, Revision 0, to the U.S. Nuclear Regulatory Commission (NRC) for the use of alternatives to the pressure test requirements in American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code)Section XI, Rules for lnservice Inspection of Nuclear Power Plant Components, at Perry Nuclear Power Plant (PNPP) associated with the fourth 10-year interval inservice inspection (ISI) interval which began May 18, 2019, and is scheduled to expire May 17, 2029.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1), the licensee requested to use the proposed alternative in RRs PT-001, Revision 3, IR-054, Revision 2, and IR-060, Revision 0, on the basis that the proposed alternative will provide an acceptable level of quality and safety. The alternative contained in IR-056, Revision 3, will not be discussed further in this safety evaluation (SE) and will be handled under separate correspondence.
By order dated December 2, 2019 (ADAMS Accession No. ML19303C953), the NRC staff approved the direct and indirect transfers of several FENOC-owned and operated plants, including PNPP. By letter dated December 3, 2019 (ADAMS Accession No. ML19337B181),
FENOC indicated that the entities taking control of the plants which had previously been referred to as New Hold Co, OwnerCo, and OpCo, would be named Energy Harbor Corp.,
Energy Harbor Nuclear Generation LLC, and Energy Harbor Nuclear Corp., respectively. Under this new set-up, Energy Harbor Corp. would indirectly own the plants as a parent company, Energy Harbor Nuclear Generation LLC would directly own the plants, and Energy Harbor Nuclear Corp. would have authority to operate the plants.
Enclosure
On February 20, 2020, FENOC informed the NRC (ADAMS Accession No. ML20054B733) that:
Upon completion of the license transfer, Energy Harbor Nuclear Corp. will adopt and endorse the outstanding commitments, licensing actions, applications, and similar items on the aforementioned docket numbers. Energy Harbor Nuclear Corp. requests NRC continuation of the regulatory reviews and actions on these items.
On February 27, 2020, Energy Harbor Nuclear Corp., informed the NRC that the transaction closed on February 27, 2020, and that it adopted and endorsed the outstanding commitments, licensing actions, applications and similar items on dockets submitted by FENOC on behalf of the licensees (ADAMS Accession No. ML20058D315). On February 27, 2020 (ADAMS Accession No. ML20030A440), the NRC staff issued Amendment No. 187 to reflect the license transfer. Accordingly, Energy Harbor Nuclear Corp. is now authorized to act as agent for Energy Harbor Nuclear Generation, LLC, and has exclusive responsibility and control over the physical construction, operation, and maintenance of the facility at PNPP.
2.0 REGULATORY EVALUATION
The NRC regulations in Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Section 55a(g)(4, requires that ASME Code Class 1, 2 and 3 components meet the ISI requirements, except the design and access provisions, set forth in Section XI of editions and addenda of the ASME Code, to the extent practical within the limitations of design, geometry, and materials of construction of the components.
Regulation 10 CFR 50.55a(z) states, in part, that alternatives to the requirements of 10 CFR 50.55a(g) may be used, when authorized by the NRC, if the licensee demonstrates:
(1) the proposed alternatives would provide an acceptable level of quality and safety, or (2) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
3.0 TECHNICAL EVALUATION
The information provided by the licensee in support of the requests for alternatives to ASME Code requirements has been evaluated and the bases for disposition are documented below.
For clarity, the licensee's requests have been evaluated in several parts according to ASME Code Examination Category.
Applicable Code Edition and Addenda
The applicable Code edition and addenda for the fourth ISI interval of the PNPP is the 2013 Edition of ASME Code,Section XI.
Duration of the Alternative:
The duration of the proposed alternatives is for the fourth 10-year ISI interval, which began on May 18, 2019, and is scheduled to end on May 17, 2029.
3.1 Proposed Alternative PT-001, Revision 3 ASME Code Requirements Table IWC-2500-1, Examination Category C-H, Item No. C7.10, requires all ASME Code Class 2 pressure-retaining components to be visually examined for evidence of leakage each inspection period (i.e., VT-2 examination during system leakage tests). The system leakage tests are also subject to the system pressure test requirements of IWC-5210, which also references IWA-5000. Specifically, IWA-5213(a)(3) requires ASME Code Class 2 systems to be in operation for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for insulated components prior to commencing the system leakage test.
ASME Code Components Affected The affected components are ASME Code Class 2 components connected to the reactor coolant pressure boundary (ASME Code Class 1) that are not provided with pressure or test isolation. The specific components are documented in Enclosure A of the January 6, 2020, application. These components include instrumentation, test connection, vent and drain lines and the associated piping components (e.g., reactor pressure vessel level instrumentation line drain valves and recirculation pump discharge line drain valves).
Reason for Request
The subject components cannot be isolated from the reactor coolant pressure boundary (ASME Code Class 1). Conducting the examinations of the ASME Code Class 2 components during the ASME Code Class 1 system leakage tests would eliminate the 4-hr hold time in accordance with the ASME Code Class 1 system leakage test requirements. This approach provides an acceptable level of quality and safety for the system leakage tests of the ASME Code Class 2 components that are not isolable from the ASME Code Class 1 boundary.
Licensees Proposed Alternative As previously discussed, IWA-5213(a)(3) and IWC-5210 require ASME Code Class 2 systems to be in operation for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to commencing VT-2 examinations on insulated components during system leakage tests. The subject insulated ASME Code Class 2 components and associated pipes are connected to the ASME Code Class 1 boundary that are not provided with either pressure or test isolation. In lieu of using IWA-5213(a)(3) and IWC-5210, the licensee proposed to conduct system leakage tests on the subject components in accordance with IWA-5213(a)(1) and IWB-5210, which do not require a hold time after attaining the test pressure. By using these provisions, the subject components, which are not isolable from the ASME Code Class 1 boundary, will be examined during the ASME Code Class 1 system leakage tests without a hold time. The system leakage test frequency and pressure will be those required for the ASME Code Class 2 system leakage test.
The subject pipes are less than 1 inch in diameter. These pipes and associated components were constructed to the requirements of ASME Code,Section III, Subsection NC, and identified as ASME Code Class 2 for ISI. The ASME Code Class 2 pipes and components will be pressurized during the ASME Code Class 1 system leakage test and VT-2 visual examinations will be performed for detection of leakage. Although the system would not have been in operation for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to commencing the VT-2 examination, the time required to bring the reactor coolant system up to the test pressure would be sufficient to allow for detection of leakage. This basis is consistent with the system leakage test conditions for ASME Code
Class 1 components that do not require a hold time after attaining the test pressure, as described in IWA-5213(a)(1).
NRC Staff Evaluation
Examination Category C-H, Item No. C7.10, requires system leakage tests to be conducted on ASME Code Class 2 components each inspection period. In these tests, the ASME Code Class 2 components are visually examined to detect evidence for leakage (VT-2 examination).
For insulated ASME Code Class 2 components required to operate during normal plant operation, IWA-5213(a)(3) specifies that no hold time is required if the system has been in operation for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The intent of the requirement (operation for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to VT-2 examination) is to allow any leakage that may exist to penetrate the insulation.
The licensee explained that the subject ASME Code Class 2 components are connected to and non-isolable from the reactor coolant pressure boundary (ASME Code Class 1 boundary).
These ASME Code Class 2 components will be pressurized during the ASME Code Class 1 system leakage tests. IWB-5221(a) requires that during the ASME Code Class 1 system leakage tests the test pressure shall not be less than the pressure corresponding to 100-percent rated reactor power. In addition, IWB-5240(a) requires that the minimum test temperature for the ASME Code Class 1 system leakage tests shall not be lower than the minimum temperature for the associated pressure specified in the plant technical specifications (TSs). Therefore, the ASME Code Class 1 system leakage tests will require a time to bring the ASME Code Class 1 systems to the pressure and temperature corresponding to the 100-percent normal operating conditions.
With respect to the pressure and temperature requirements for ASME Code Class 1 system leakage tests, the NRC staff notes that (a) the time required to achieve the test pressure and temperature is sufficient to reveal any leakage that may exist during the tests; and (b) the test pressure will be maintained at a level not less than the normal operating pressure. Therefore, the NRC staff finds that the pressure and temperature requirements ensure effective detection of leakage from ASME Code Class 1 components even though no hold time is required after attaining the test pressure.
The licensee proposed to perform system leakage tests on the subject ASME Code Class 2 components during the ASME Code Class 1 system leakage tests with no hold time prior to VT-2 examination of the subject components after attaining the test pressure. As discussed above, the NRC staff finds that the proposed alternative provides an acceptable level of quality and safety because (a) the subject ASME Code Class 2 components are connected to and non-isolable from the ASME Code Class 1 boundary; (b) the non-isolable nature of the subject components supports the use of the provisions for ASME Code Class 1 system leakage tests, which allows no hold time; and (c) the time required to pressurize the piping systems to the test conditions is sufficient to reveal any leakage during the system leakage tests. Therefore, the NRC staff finds that the proposed alternative provides reasonable assurance of the structural and leak-tight integrity of the subject components.
3.2 Proposed Alternative IR-054, Revision 2 ASME OM Code Requirements ASME Section XI, Table IWB-2500-1, Examination Category B-D, requires a volumetric examination of all nozzles with full penetration welds to the vessel shell (or head) and integrally cast nozzles each 10-year interval.
ASME Code Components Affected The affected components at PNPP belong to Examination Category B-D, Full Penetration Welded Nozzles in Vessels under Examination Item No. B3.90, Nozzle-to-Vessel Welds and B3.100, Nozzle Inside Radius Section. The specific components are documented in Enclosure B of the January 6, 2020, application, and summarized in Table 1, below.
Table 1 RPV Nozzle-to-Vessel Welds and Inner Radii Subject to this Request Identification Total Minimum Number Description Number Number to be examined N1 Recirculation Outlet 2 1 N2 Recirculation Inlet 10 3 N3 Main Steam 4 1 N5 Core Spray 2 1 N6 Low Pressure Core Injection 3 1 N7 Top Head Spray Spare 1 0 (see Note)
N8 Top Head Spray 1 1 (see Note)
N9 Jet Pump Instrumentation 2 1 Note: Nozzles N7 and N8 are an equivalent design, but the spare nozzle (N7) is subjected to loading conditions equal or less severe than the nozzle in use (N8). The licensees request states that nozzle N8 is bounding for these two nozzles, and that nozzle N8 will be inspected during the duration of the proposed alternative.
Reason for Request
For all reactor pressure vessel (RPV) nozzle-to-vessel shell welds and nozzle inner radii, ASME Code,Section XI, requires 100 percent inspection during each 10-year ISI interval.
However, Code Case N-702 provides an alternative, which reduces the inspection of RPV nozzle-to-vessel shell welds and nozzle inner radii areas from 100 percent to 25 percent of the nozzles for each nozzle type during each 10-year interval. This Code Case was conditionally approved in Regulatory Guide (RG) 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1,"Revision 18, which was the current revision at the time of the licensees proposed alternative request. For application of Code Case N-702, the licensee is required to address the conditions specified in RG 1.147, Revision 18 for ASME Code Case N-702:
The applicability of Code Case N-702 must be shown by demonstrating that the criteria in Section 5.0 of NRC Safety Evaluation regarding BWRVIP [boiling water reactor vessel and internal project]-108 dated December 19, 2007 (ADAMS Accession No. ML073600374) or Section 5.0 of NRC Safety Evaluation regarding BWRVIP-241 dated April 19, 2013 (ADAMS Accession No. ML13071A240) are met. The evaluation demonstrating the applicability of the Code Case shall be reviewed and approved by the NRC prior to the application of the Code Case.
BWRVIP-108, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radii (ADAMS Accession Nos. ML023360232 and ML023360234; non-publicly available) and BWRVIP-241, Probabilistic Fracture Mechanics [PFM] Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii (ADAMS Accession Nos. ML11119A042, non-publicly available, and ML11119A043, publicly available) contain PFM analysis results supporting Code Case N-702. Both reports are for 40 years of operation.
Code Case N-702 allows that VT-1 visual examination may be performed in lieu of volumetric examination for Examination Item No. B3.100 nozzle inner radius sections. Code Case N-648-1, as conditionally accepted by RG 1.147, Revision 18, requires that nozzle inner radius examinations must use the allowable flaw length criteria of ASME Code, Table IWB-3512-1, with limiting assumptions on the flaw aspect ratio.
Licensees Proposed Alternative The licensee proposed to implement ASME Code Case N-702 and reduce the ASME Code-required volumetric examinations for all RPV nozzle-to-shell welds and inner radii, to a minimum of 25 percent of the nozzle inner radii and nozzle-to-shell welds, including at least one nozzle from each system and nominal pipe size during each inspection interval. The required examination volume for the reduced set of nozzles remains at 100 percent of that depicted in Figures IWB-2500-7 (a) through (d), as applicable in the ASME Code.
In addition, the licensee stated it may perform a VT-1 visual examination, as outlined in ASME Code Case N-648-1, Alternative Requirements for Inner Radius Examination of Class 1 Reactor Vessel Nozzles,Section XI Division 1, in lieu of a volumetric examination for Category B-D, Item No. B3.100.
Licensees Bases for Alternative The alternative is based on the PFM results documented in BWRVIP-108. The licensee proposed that it met the evaluation criteria in the SE for BWRVIP-108 as follows:
(1) Max RPV Heatup/Cooldown Rate The maximum RPV heatup/cooldown rate is limited to < 115°F/hr.
PNPP TSs SR 3.4.11, reactor coolant system heatup and cooldown rates are limited to a maximum of 100 °F in any 1-hour period and thus meet the requirement of Criterion 1.
(2) Recirculation Inlet (N2) Nozzles (pr/t) /Ci-RPV < 1.15, where p = RPV normal operating pressure (psi),
r = RPV inner radius (inch),
t = RPV wall thickness (inch), and Ci-RPV = 19332.
The plant-specific result based on the input parameters for this nozzle per the licensee submittal is (pr/t) /Ci-RPV = ([(1045)(120.2)/7]/19332) = 0.93 < 1.15, thus meeting the requirements of Criterion 2.
(3) Recirculation Inlet (N2) Nozzles
[p(ro2+ri2)/(ro2-ri2)]/Ci-NOZZLE < 1.15, where ro = nozzle outer radius (inch),
ri = nozzle inner radius (inch), and Ci-NOZZLE = 1637.
The plant-specific result based on the input parameters for this nozzle per the licensee submittal is [p(ro2+ri2)/(ro2-ri2)]/Ci-NOZZLE = ([1045(11.1252 + 5.8132)/(11.1252 - 5.8132)]/1637) = 1.12 < 1.15, thus meeting the requirements of Criterion 3.
(4) Recirculation Outlet (N1) Nozzles (pr/t)/Co-RPV < 1.15, where r = RPV inner radius (inch),
t = RPV wall thickness (inch), and Co-RPV = 16171.
The plant-specific result based on the input parameters for this nozzle per the licensee submittal is (pr/t)/Co-RPV = ([(1045)(120.2)/7]/16171) = 1.11 < 1.15, thus meeting the requirements of Criterion 4.
(5) Recirculation Outlet (N1) Nozzles
[p(ro2 + ri2)/(ro2 - ri2)]/Co-NOZZLE < 1.15, where ro = nozzle outer radius (inch),
ri = nozzle inner radius (inch), and Co-NOZZLE = 1977.
The plant-specific result based on the input parameters for this nozzle per the licensee submittal is [p(ro2 + ri2)/(ro2 - ri2)]/Co-NOZZLE = ([1045(17.5942 + 102)/(17.5942 - 102)]/1977) = 1.03 < 1.15, thus meeting the requirements of Criterion 5.
NRC Staff Evaluation
The licensee proposed an alternative to implement ASME Code Case N-702 for all PNPP RPV nozzle-to-vessel shell penetration welds and nozzle inner radii using the criteria in BWRVIP-108. For each of the nozzle designs listed in Table 1 of this SE, the licensee proposed inspection of a minimum of 25 percent of the nozzle inner radii and nozzle-to-shell welds, including at least one nozzle from each system and nominal pipe size during each inspection interval. The licensee states that the N7 and N8 nozzles are equivalent with respect to geometry and materials, but that nozzle N7 (which is a spare) experiences equivalent or less loading than nozzle N8 (which is currently in use). The licensee also states that nozzle N8 bounds nozzle N7 and that nozzle N8 is scheduled to be inspected during the duration of the proposed alternative. The NRC finds the licensee inspection plan acceptable since it is consistent with ASME Code Case N-702.
In general, the applicability of BWRVIP-108 report to an ASME Code Case N-702 alternative is demonstrated by showing that Criteria 2 through 5 within Section 5.0 of the NRC SE for BWRVIP-108 are met for the bounding nozzles (recirculation inlet and outlet nozzles), and that Criterion 1 is met for all components included in the proposed alternative.
The NRC staff confirms that Criterion 1 (applicable to all nozzles within the scope of ASME Code Case N-702) is satisfied because PNPP TSs SR 3.4.11 limits the maximum
heatup/cooldown rate to less than or equal to 100 °F/hour, well below the 115 °F/hour criterion limit.
For Criteria 2 through 5, the licensee provided plant-specific data and its evaluation of the driving force factors, or ratios, using the criteria established in Section 5.0 of the BWRVIP-108 SE. The licensee showed that Criteria 2, 3, 4 and 5 are satisfied. The NRC staff reviewed the licensee's calculations and confirms that they show that Criteria 2 through 5 are satisfied.
Therefore, the BWRVIP-108 report applies to PNPP, and the basis for using Code Case N-702 is demonstrated for the PNPP RPV nozzle-to-vessel welds and inner radii listed in Table 1 above.
For the Examination Item No. B3.100 nozzle inner radius sections, the NRC staff finds the licensee proposal to perform VT-1 visual examination in lieu of ultrasonic examination to be acceptable since the licensee will comply with ASME Code Case N-648-1, with associated required conditions specified in RG 1.147, Revision 18, which was the current revision at the time of the licensees proposed alternative request.
3.3 Proposed Alternative IR-060, Revision 0 ASME OM Code Requirements ASME Code,Section XI, Table IWB-2500-1, Examination Category B-G-1, Item No. B6.40, requires volumetric examination of the RPV threads in flange. The required examination volume is specified in Figure IWB-2500-12 of the ASME Code,Section XI.
ASME Code Components Affected The specific components are documented in Enclosure D of the January 6, 2020, application.
Duration of the Alternative:
The duration of the proposed alternative is for the fourth 10-year ISI interval, which began on May 18, 2019, and is scheduled to end on May 17, 2029.
Reason for Request
The proposed alternative is consistent with the opinion stated in ASME Code Case N-864, "Reactor Vessel Threads in Flange Examinations,Section XI, Division 1. The affected ASME Code components identified in request IR-060 are components within the scope of ASME Code Case N-864. ASME Code Case N-864 was approved by the ASME Board on Nuclear Codes and Standards on July 28, 2017; however, it has not been incorporated into RG 1.147, and is not available for application at nuclear power plants without specific NRC approval.
Licensees Proposed Alternative In request IR-060, Revision 0, the licensee proposed to eliminate the ASME Code-required volumetric examination of the threads in the reactor vessel (RV) flange stud holes for the duration of the fourth 10-year ISI interval at PNPP. As an alternative to the ASME Code-required volumetric examination, the licensee described maintenance activities and technical evaluations for ensuring the structural integrity of the threads in the RV flange stud holes. The licensee reported that these activities and evaluations will provide an acceptable level of quality and safety.
As part of its routine maintenance activities, the licensee reported that it performs visual inspections of the RV closure head studs and threaded RV flange stud holes each time the closure head is removed. Care is taken to inspect the threads in the RV flange for damage and to protect the threads from damage when the studs are removed. Prior to reinstallation, the studs and stud holes are cleaned and lubricated. The studs are then reinserted and tensioned into the RV flange. These activities are performed in accordance with plant procedures and provide assurance that degradation is detected and mitigated prior to returning the reactor to service.
The licensee cited Electric Power Research Institute (EPRI) Report No. 3002007626, Nondestructive Evaluation: Reactor Pressure Vessel Threads in Flange Examination Requirements, March 2016 (ADAMS Accession No. ML16221A068), as the generic technical basis for elimination of the volumetric examination of the threads in the RV flange. The EPRI report includes the following generic evaluations:
A review of inservice examination data for threaded RV flange stud holes at 94 U.S.
reactors showed no reportable indications based on 10,662 examinations performed; An evaluation of the susceptibility of the threads in the RV flange to potential inservice degradation mechanisms showed no active degradation mechanisms; A stress analysis determined operating stresses due to design pressure, stud preload, and thermal transients at the critical locations; and Based on the operating stresses, a flaw tolerance evaluation determined that a postulated flaw would not be expected to challenge the integrity of the RV during the design life.
The licensee described its plant-specific evaluation for demonstrating that the evaluations documented in the EPRI report are applicable to PNPP. For additional details on the licensees submittal, please refer to the document located at the ADAMS Accession number identified above.
NRC Staff Evaluation
The NRC staff reviewed PNPP request IR-060, Revision 0, pursuant to 10 CFR 50.55a(z)(1).
The staff noted that mechanical and thermal fatigue are the only potential degradation mechanisms for the threads in the RV flange. Section 6 of EPRI Report 3002007626 documents generic stress, flaw tolerance, and fatigue crack growth analyses for the threaded regions in the RV flange. The EPRI analyses demonstrate that, for the geometry and boundary conditions assumed, a 360-degree elliptical flaw will not reach the maximum flaw size allowed by the ASME Code,Section XI, considering mechanical and thermal fatigue cycles for at least an 80-year operating term.
The NRC staff previously reviewed the generic analyses in Section 6 of EPRI Report 3002007626 and found them acceptable based on evaluations documented in Sections 3.2.1.1 and 3.2.1.2 of its January 26, 2017, SE (ADAMS Accession No. ML17006A109) for the pilot application of this methodology at Vogtle Electric Generating Plant, Units 1 and 2 and Joseph M. Farley Nuclear Plant, Unit 1. However, since this plant-specific SE does not constitute a generic endorsement of the EPRI report, licensees must provide plant-specific evaluations of its applicability to their RVs in submittals of proposed alternatives, per 10 CFR 50.55a(z)(1).
The NRC staff compared the relevant design and operating parameters for PNPP, including operating pressure, RV diameter, flange geometry, and bolt geometry, to those used as inputs for the generic analyses documented in the EPRI report. The staff confirmed that the PNPP
design and operating parameters are bounded by those used in the EPRI report. The staff independently confirmed the 28 thousand pounds per square inch (ksi) stud preload stress that the licensee calculated for PNPP and noted that this value is significantly less than the 42.338 ksi stud preload stress used in the EPRI report. The staff also identified that there are several conservatisms in the EPRI analytical methods, including the use of linear elastic fracture mechanics for flaw tolerance calculations during high temperature operations; use of ASME Code,Section XI structural factors; and a larger number of assumed fatigue cycles than expected for normal operation of the plant. Based on these considerations, the NRC staff finds that the referenced EPRI analyses are conservative and bounding relative to the actual design and operating conditions at PNPP.
The licensee also cited the EPRI reports 2016 survey that showed no indications from 10,662 examinations of the threads in RV flange at 94 U.S. power reactors; the licensee confirmed that PNPP has shown no indications as a result of these exams. Therefore, considering the licensees evaluation demonstrating the applicability of the EPRI report for PNPP, and the routine maintenance and visual inspection activities for these threaded components, the NRC staff finds that the licensees proposal to eliminate the volumetric examination of the threaded stud holds in the RV flange will provide an acceptable level of quality and safety at PNPP.
4.0 CONCLUSION
As set forth above, the NRC staff finds that the proposed alternatives described in RRs PT-001, Revision 3, IR-054, Revision 2, and IR-060, Revision 0, provide an acceptable level of quality and safety for components listed in Enclosure A, B, and D of the January 6, 2020 application.
Accordingly, the NRC staff concludes that the licensee has adequately addressed all the regulatory requirements set forth in 10 CFR 50.55a(z)(1). Therefore, the NRC staff authorizes the proposed alternatives in RRs PT-001, Revision 3, IR-054, Revision 2, and IR-060, Revision 0, for the fourth 10-year ISI interval at PNPP which began on May 18, 2019 and is scheduled to end on May 17, 2029.
All other ASME Code requirements for which relief was not specifically requested and approved in the subject requests for relief remain applicable.
Principle Contributors: S. Min, NRR J. Jenkins, NRR C. Sydnor, NRR Date: November 10, 2020
ML20311A658 *via e-mail ** via memo OFFICE NRR/DORL/LPL3/PM* NRR/DORL/LPL3/LA* NRR/DNRL/NVIB/BC**
NAME SWall SRohrer HGonzalez DATE 11/6/2020 11/9/2020 9/11/2020 OFFICE NRR/DNRL/NPHP/BC** NRR/DORL/LPL3/BC* NRR/DORL/LPL3/PM*
NAME MMitchell NSalgado (JWiebe for) SWall DATE 7/21/2020 11/10/2020 11/10/2020