ML20252A135
| ML20252A135 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 09/04/2020 |
| From: | Rafferty-Czincila S Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Relief Request I4R-24 | |
| Download: ML20252A135 (31) | |
Text
200 Exelon Way 200 Exelon Way Kennett Square, PA 19348 www.exeloncorp.com 10 CFR 50.55a September 4, 2020 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Limerick Generating Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-39 and NPF-85 NRC Docket Nos. 50-352 and 50-353
Subject:
Relief Request I4R-24 Associated with Residual Heat Removal Heat Exchanger Category C-A and C-B Examinations Attached for your review is a relief request associated with the Inservice Inspection (ISI)
Program for Limerick Generating Station (LGS), Units 1 and 2. Specifically, this relief request is associated with Residual Heat Removal heat exchanger Category C-A and C-B examinations.
LGS, Units 1 and 2 is currently in the fourth 10-year interval, which began on February 1, 2017.
We request your approval by September 4, 2021.
There are no regulatory commitments in this letter.
If you have any questions concerning this letter, please contact Tom Loomis at (610) 765-5510.
Respectfully, Shannon Rafferty-Czincila Director - Licensing Exelon Generation Company, LLC
Attachment:
Relief Request I4R-24 cc:
USNRC Region I, Regional Administrator USNRC Senior Resident Inspector, LGS USNRC Project Manager, LGS W. DeHaas, Pennsylvania Bureau of Radiation Protection Rafferty-Czincila, Shannon Digitally signed by Rafferty-Czincila, Shannon DN: cn=Rafferty-Czincila, Shannon Date: 2020.09.08 10:49:58 -04'00'
Attachment Relief Request I4R-24
10 CFR 50.55a RELIEF REQUEST I4R-24 Revision 0 (Page 1 of 29)
Relief Request I4R-24 Associated with Residual Heat Removal Heat Exchanger Category C-A and C-B Weld Examinations In Accordance with 10 CFR 50.55a(z)(1)
--Alternative Provides Acceptable Level of Quality and Safety--
1.0 ASME CODE COMPONENTS AFFECTED Code Class:
Class 2
==
Description:==
Residual Heat Removal (RHR) heat exchanger Examination Categories: Class 2, Category C-A, pressure retaining welds in pressure vessels Class 2, Category C-B, pressure retaining nozzle welds in pressure vessels Item Numbers:
C1.10 - Shell circumferential welds C1.20 - Head circumferential welds C2.21 - Nozzle-to-shell (nozzle-to-head or nozzle-to-nozzle) weld C2.22 - Nozzle inside radius section Component IDs:
Unit ASME Category ASME Item Comp ID Description 1
C-A C1.10 RHR-HXAR-4 Shell Ring 1 to Flange Weld 1
C-A C1.10 RHR-HXBR-4 Shell Ring 1 to Flange Weld 2
C-A C1.10 2AE-205 SG-1 Shell (Ring #1) to Flange Weld 2
C-A C1.10 2BE-205R W20 Shell Ring 1 to Flange Weld 1
C-A C1.20 RHR-HXAR-1 Head to Shell Ring 3 Weld 1
C-A C1.20 RHR-HXBR-1 Head to Shell Ring 3 Weld 2
C-A C1.20 2AE-205 SG-5 Shell Head to Shell Weld (Ring
C-B C2.21 RHR-HXAR-N3 Nozzle to Head Weld 1
C-B C2.21 RHR-HXAR-N4 Nozzle to Shell 1 Weld 1
C-B C2.21 RHR-HXBR-N3 Nozzle to Head Weld 1
C-B C2.21 RHR-HXBR-N4 Nozzle to Shell 1 Weld 2
C-B C2.21 2AE-205 N-3-1 Inlet Nozzle (N-3) to Shell Head Weld 2
C-B C2.21 2BE-205R N-4-1 Outlet Nozzle (N-4) to Shell Weld 2
C-B C2.21 2AE-205 N-4-1 Outlet Nozzle (N-4) to Shell Weld (Ring #1) 1 C-B C2.22 RHR-HXAR-N3IR Nozzle N3 Inner Radius 1
C-B C2.22 RHR-HXAR-N4IR Nozzle N4 Inner Radius 1
C-B C2.22 RHR-HXBR-N3IR Nozzle N3 Inner Radius 1
C-B C2.22 RHR-HXBR-N4IR Nozzle N4 Inner Radius 2
C-B C2.22 2AE-205 N-3-1 Inner Radius Inlet Nozzle (N-3) Inner Radius 2
C-B C2.22 2BE-205R N-4-1 IR Outlet Nozzle (N-4) Inner Radius 2
C-B C2.22 2AE-205 N-4-1 Inner Radius Outlet Nozzle (N-4) Inner Radius
10 CFR 50.55a RELIEF REQUEST I4R-24 Revision 0 (Page 2 of 29) 2.0 APPLICABLE CODE EDITION AND ADDENDA The fourth Inservice Inspection (ISI) interval Code of record for Limerick Generating Station (LGS), Units 1 and 2 is the 2007 Edition with 2008 Addenda of the ASME Boiler and Pressure Vessel Code,Section XI, Division 1, Rules for Inservice Inspection of Nuclear Power Plant Components.
3.0 APPLICABLE CODE REQUIREMENTS ASME Section XI IWC-2500(a), Table IWC-2500-1, Examination Categories C-A and C-B require examination of the Item Nos. as follows:
Item No. C1.10 - Volumetric examination of circumferential cylindrical-shell-to-conical-shell-junction welds and shell (or head)-to-flange welds during each inspection interval.
The examination volume is shown in Figure IWC-2500-1.
Item No. C1.20 - Volumetric examination of circumferential head-to-shell welds during each inspection interval. The examination volume is shown in Figure IWC-2500-1.
Item No. C2.21 - Surface and volumetric examination of nozzle-to-shell welds at terminal ends of piping runs each inspection interval. The examination volume is shown in Figure IWC-2500-4(a), (b), or (d).
Item No. C2.22 - Volumetric examination of nozzle inside radius section at terminal ends of piping runs each inspection interval. The examination volume is shown in Figure IWC-2500-4(a), (b), or (d).
4.0 REASON FOR REQUEST The Electric Power Research Institute (EPRI) performed assessments in Reference 1 of the basis for the ASME Section XI examination requirements specified for the above listed ASME Code,Section XI, Division 1 Examination Categories for BWR heat exchanger components. The assessments include a survey of inspection results from 24 BWR units as well as flaw tolerance evaluations using probabilistic fracture mechanics (PFM) and deterministic fracture mechanics (DFM). The Reference 1 report supports the conclusion that the current ASME Code,Section XI, Division 1 inspection interval of ten years for these components can be increased with no impact to plant safety. It is upon the basis of this conclusion that an alternate inspection interval is being requested.
5.0 PROPOSED ALTERNATIVE AND BASIS FOR USE Exelon is requesting an inspection alternative to the examination requirements of ASME Code,Section XI, Division 1, Table IWB-2500-1 for the following Examination Categories and Item Nos.:
10 CFR 50.55a RELIEF REQUEST I4R-24 Revision 0 (Page 3 of 29)
Item No.
Description ASME Category C1.10 Heat Exchanger, shell circumferential welds C-A C1.20 Heat Exchanger, head circumferential welds C-A C2.21 Heat Exchanger, nozzle-to-shell (nozzle-to-head or nozzle-to-nozzle) welds C-B C2.22 Heat Exchanger, nozzle inside radius section C-B The proposed alternative is to increase the inspection interval for these Item Nos. to the end of the current operating license (from the current ASME Code,Section XI Division 1 10-year requirement) which is October 26, 2044 and June 22, 2049 for LGS Units 1 and 2, respectively. This equates to an extension of 27 years, 8 months, 25 days for Unit 1 and 32 years, 4 months, 21 days for Unit 2 from the end of the third Inservice Inspection Interval (January 31, 2017) at which time all ASME Section XI, Division 1 requirements were satisfied. A summary of the key aspects of the technical basis for this request are summarized below. The applicability of the technical basis to LGS, Units 1 and 2 is shown in Appendix A.
Degradation Mechanism Evaluation An evaluation of degradation mechanisms that could potentially impact the reliability of the heat exchanger components was performed in Reference 1. Evaluated mechanisms included stress corrosion cracking (SCC), environmentally assisted fatigue (EAF),
microbiologically influenced corrosion (MIC), pitting, crevice corrosion, erosion-cavitation, erosion, flow accelerated corrosion (FAC), general corrosion, galvanic corrosion, and mechanical and thermal fatigue. Other than the potential for EAF and mechanical/thermal fatigue, there were no active degradation mechanisms identified that significantly affect the long-term structural integrity of the heat exchanger components.
Stress Analysis Finite element analyses (FEA) were performed in Reference 1 to determine the stresses in the heat exchanger components. The analyses were performed using representative Boiling Water Reactor (BWR) heat exchanger geometries, bounding transients, and typical material properties. The results of the stress analyses were used in a flaw tolerance evaluation. The applicability of the FEA to LGS, Units 1 and 2 is shown in Appendix A and confirms that all plant-specific requirements are met. Therefore, the evaluation results and conclusions of Reference 1 are applicable to LGS, Units 1 and 2.
Flaw Tolerance Evaluation Flaw tolerance evaluations were performed in Reference 1 consisting of PFM and DFM evaluations. The results of the PFM analyses indicate that, after a preservice inspection (PSI), no other inspections are required for up to 80 years of plant operation to meet the U.S. Nuclear Regulatory Commissions (NRCs) safety goal of 10-6 failures per year. For the specific case of LGS, Units 1 and 2, some of the heat exchangers have been replaced as a result of tube degradation. For Unit 1, the RHR heat exchangers were replaced in 1994 during 1R05. Therefore, for Unit 1, both heat exchangers would have received the baseline examination plus examinations consistent with two intervals. For Unit 2, the A RHR heat exchanger is from original plant construction and the B RHR heat exchanger was replaced in 2011. Therefore, for Unit 2, the A heat exchanger
10 CFR 50.55a RELIEF REQUEST I4R-24 Revision 0 (Page 4 of 29) would have received an initial examination plus exams for three intervals while the B heat exchanger would have only received its initial examinations plus one intervals worth of examinations. Table 8-10 of Reference 1 indicates that if a preservice examination is performed with no subsequent inservice examinations, the NRC safety goal is met for up to 80 years of plant operation. All the heat exchangers at LGS, Units 1 and 2 have received a preservice examination plus at least one inservice examination which would support an inspection interval equal to the remainder of the current operating license for Unit 1 (October 26, 2044) and Unit 2 (June 22, 2049). The DFM evaluations provide verification of the PFM results by demonstrating that it takes approximately 600 years for a postulated flaw with an initial depth equal to the ASME Code,Section XI, Division 1 acceptance standards to grow to 80% of the wall thickness without exceeding the ASME Code,Section XI, Division 1 allowable fracture toughness.
Note that, on December 11, 2019, Southern Nuclear submitted a proposed alternative to increase the inspection interval for ASME Section XI, Table IWC-2500-1, exam Category C-B, Item No. C2.21 and C2.22, exams from 10 years to 30 years based on the analysis performed in EPRI Technical Report 3002014590 (ADAMS Accession No. ML19347B105). The probabilistic fracture mechanics (PFM) analysis in that EPRI report was performed using the PRobabilistic OptiMization of InSpEction (PROMISE) Version 1.0 software, developed by Structural Integrity Associates. As part of the NRCs review of Southern Nuclears request, the NRC performed an audit of the PROMISE Version 1.0 software as discussed in the NRCs audit plan dated May 14, 2020 (ADAMS Accession No. ML20128J311). Note that this Request for Alternative for LGS, Units 1 and 2 uses Version 2.0 of the PROMISE software, as described in Reference 1. The only technical difference between the two versions is that in PROMISE Version 1.0, the user-specified examination coverage is applied to all inspections, whereas in PROMISE Version 2.0, the examination coverage can be specified by the user uniquely for each inspection. In both Versions 1.0 and 2.0, the software assumes 100% coverage for the PSI examination.
Inspection History Plant operating experience (including examinations performed to-date, examination findings, inspection coverage, and Relief Requests) is presented in Appendix B. As shown in Appendix B, some of the previous examinations for the affected heat exchanger components have had limited coverage. Also, as shown in Appendix B, no flaws that exceeded the ASME Code,Section XI, Division 1 acceptance standards were identified during any prior examinations.
Industry-wide inspection history for the affected heat exchanger weld components (obtained from an industry survey) is presented in Appendix C. The results of the survey indicate that these components are very flaw tolerant.
Conclusion It is concluded that the RHR heat exchanger components contained in this Request for Alternative are very flaw tolerant. PFM and DFM evaluations performed as part of the technical basis (Reference 1) demonstrate that, after PSI, no other inspection is required for at least 80 years of operation to meet the NRC safety goal of 10-6 failures per reactor year. Plant-specific applicability of the technical basis to LGS, Units 1 and 2 is demonstrated in Appendix A. Increasing the inspection interval for the RHR heat
10 CFR 50.55a RELIEF REQUEST I4R-24 Revision 0 (Page 5 of 29) exchanger examinations identified in this relief request to the end of the current operating license provides an acceptable level of quality and safety in lieu of the current ASME Code,Section XI, Division 1 10-year inspection frequency.
Operating and examination experience demonstrates that these components have performed with very high reliability, mainly due to their robust design. As shown in Appendix B, to-date Exelon Generation Company, LLC (Exelon) has performed more than 65 inspections of heat exchanger components at LGS, Units 1 and 2. No flaws that exceeded the ASME Code,Section XI, Division 1 acceptance standards were identified during any examinations, as shown in Appendix B. Two of the examinations listed in Appendix B involved limited coverage that was less than the ASME Code required coverage of essentially 100% (which is 90% or greater). The coverage achieved in the examination for one weld for Item C2.21 for Unit 2 during the 2R13 refueling outage was 53%, and the coverage achieved in the examination for one component for Item C2.22 for Unit 1 during the 1R15 refueling outage was 70%. Section 8.3.5 of Reference 1 discusses limited coverage and determines that the conclusions of the report are applicable to components with limited coverage. In addition, Table 8-9 in Section 8.3.4.1 of the technical basis report (Reference 1) presents probabilities of rupture and leakage for the base case (PSI only examination). This table shows that the probabilities of rupture for all Cases are lower than the acceptance criteria by three orders of magnitude, and the probability of leakage is lower than the acceptance criteria for the limiting Case (BHX-TS-P3A) by an order of magnitude. Therefore, any subsequent inservice examinations, including examinations with limited coverage of any extent, would further reduce these acceptability values of probabilities of rupture and leakage.
Because Exelon performed the required ASME Code,Section III and Section XI PSI examinations, which involved 100% coverage, these coverage results are applicable to the LGS, Units 1 and 2 RHR heat exchangers. Finally, it is important to note that all other inspection activities, including the ASME Code,Section XI, Division 1, Examination Category C-H system leakage test conducted each inspection period, will continue to be performed, providing further assurance of safety.
Finally, as discussed in Reference 2, for situations where no active degradation mechanism is present, it was concluded that subsequent inservice inspections do not provide additional value after PSI has been performed and the inspection volumes have been confirmed to be defect-free. The LGS, Units 1 and 2 RHR heat exchanger components have received the required PSI examinations and more than 65 follow-on inservice inspections with no flaws that exceeded the ASME Code,Section XI, Division 1 acceptance standards.
Therefore, Exelon requests that the NRC authorize this proposed alternative in accordance with 10 CFR 50.55a(z)(1).
6.0 DURATION OF PROPOSED ALTERNATIVE The proposed Alternative is requested for LGS, Units 1 and 2 for the fourth inspection interval and the remainder of the current operating license for Unit 1 (October 26, 2044) and Unit 2 (June 22, 2049).
10 CFR 50.55a RELIEF REQUEST I4R-24 Revision 0 (Page 6 of 29) 7.0 PRECEDENT No previous submittals have been made requesting relief from the ASME Code,Section XI, Division 1 Examination Categories C-A and C-B (Item Nos. C1.10, C1.20, C2.21, and C2.22) volumetric examinations on the basis of the Reference 1 Technical Report.
However, Code Case N-716-1 (Reference 11) was approved by ASME in January 2013 and provides alternate piping classification and examination requirements for Class 1, 2, 3, and non-Class piping welds as well as for Category C-A, C-B, C-D and C-G components, including the Item No. C1.10, C1.20, C1.30, C2.21, C2.22, and C2.32 BWR heat exchanger components in the scope of this evaluation. This Code Case is unconditionally approved for use in RG 1.147 (Reference 10). Low safety significant (LSS) components, typically including the BWR heat exchanger components in the scope of this evaluation, are exempt from the volumetric, surface, and VT-1 and VT-3 visual examination requirements of Section XI. Within the Exelon fleet, Peach Bottom Atomic Power Station, Units 2 and 3, James A. FitzPatrick Nuclear Power Plant, and the Nine Mile Point Nuclear Station, Units 1 and 2 have implemented this Code Case.
Below is a list of Relief Requests and other precedents related to inspections of BWR heat exchanger welds and components:
Letter from D Terao (NRC) to R. K. Edington (Nebraska Public Power District),
Cooper Nuclear Station RE: Risk-Informed Inservice Inspection Program for the Fourth 10-year Interval: Relief Request No. RI-34 (TAC No. MD0283), dated October 23, 2006, ADAMS Accession No. ML062850051.
Letter from E. C. Marinos (NRC) to L. M. Stinson (Southern Nuclear Operating Company Inc.), Edwin I. Hatch Nuclear Plant, Unit Nos. 1 and 2, - Evaluation of Third 10-year Interval Inservice Inspection Program Plan Request for Relief Nos.
RR-42, RR-43, RR-44, RR-45, RR-51, RR-58, RR-59, RR-60, and RR-62 (TAC Nos. MD2587, MD2588, MD2589, MD2590, MD2591, MD2597, MD2609, MD2610, MD2611, and MD2613), dated June 5, 2007, ADAMS Accession No. ML071360297.
Letter from M. T. Markley (NRC) to J. J. Hutto (Southern Nuclear Operating Company Inc.), Edwin I. Hatch Nuclear Plant, Unit Nos. 1 and 2 - Relief Requests ISI-RR-16, ISI-RR-17, ISI-RR-21, and ISI-RR-22 for Relief from Inservice Inspection Requirements (CAC Nos. MF9027, MF9030, MF9031, MF9034, MF9035, and MF9036; EPID Nos. L-2016-LLR-0008, L-2016-LLR-0010, L-2016-LLR-0011, and L-2016-LLR-0009), dated October 20, 2017, ADAMS Accession No. ML17268A044.
Letter from S. J. Campbell (NRC) to C. G. Pardee (Exelon Nuclear), Lasalle County Station, Units 1 and 2 - Relief Request CR-26, Inservice Inspection Program Relief Regarding Examination Coverage for the Second 10-Year Inservice Inspection Interval (TAC Nos. MD9817 - MD 9818), dated September 21, 2009, ADAMS Accession No. ML092570321.
In addition, other studies have been performed by the industry to extend the inspection interval for various components and have been accepted by the NRC.
Based on studies presented in Reference 3, the NRC approved extending PWR reactor vessel nozzle-to-shell welds from 10 to 20 years in Reference 4.
10 CFR 50.55a RELIEF REQUEST I4R-24 Revision 0 (Page 7 of 29)
Based on work performed in BWRVIP-108 (Reference 5) and BWRVIP-241 (Reference 7), the NRC approved the reduction of BWR vessel nozzle-to-shell weld examinations (Item No. B3.90 for BWRs from 100% to a 25% sample of each nozzle type every 10 years) in References 6 and 8. The work performed in BWRVIP-108 and BWRVIP-241 provided the technical basis for ASME Code Case N-702 Reference 9, which has been conditionally approved by the NRC in Revision 18 of Regulatory Guide 1.147 (Reference 10).
ASME Code Case N-706-1 (Reference 12) allows for alternate examinations of stainless steel PWR regenerative and residual heat exchangers. This is approved for use in NRC RG 1.147 (Reference 10).
8.0 ACRONYMS ASME American Society of Mechanical Engineers BWR Boiling Water Reactor BWRVIP Boiling Water Reactor Vessel and Internals Program CFR Code of Federal Regulations DFM Deterministic fracture mechanics EAF Environmentally assisted fatigue EPRI Electric Power Research Institute FAC Flow accelerated corrosion FEA Finite element analysis HX Heat exchanger ISI Inservice Inspection LSS Low safety significant MIC Microbiologically influenced corrosion NPS Nominal pipe size NRC Nuclear Regulatory Commission OD Outside diameter PFM Probabilistic fracture mechanics PSI Preservice inspection PWR Pressurized Water Reactor SCC Stress corrosion cracking
9.0 REFERENCES
- 1.
Technical Bases for Inspection Requirements for Class 2 BWR Heat Exchanger Nozzle-to-Shell Welds, Nozzle Inside Radius Sections, and Vessel Head, Shell, and Tubesheet-to-Shell Welds, EPRI, Palo Alto, CA: 2020. 3002018473.
- 2.
American Society of Mechanical Engineers, Risk-Based Inspection: Development of Guidelines, Volume 2-Part 1 and Volume 2-Part 2, Light Water Reactor (LWR)
Nuclear Power Plant Components. CRTD-Vols. 20-2 and 20-4, ASME Research Task Force on Risk-Based Inspection Guidelines, Washington, D.C., 1992 and 1998.
- 3.
B. A. Bishop, C. Boggess, N. Palm, Risk-Informed extension of the Reactor Vessel In-Service Inspection Interval, WCAP-16168-NP-A, Rev. 3, October 2011.
10 CFR 50.55a RELIEF REQUEST I4R-24 Revision 0 (Page 8 of 29)
- 4.
US NRC, Revised Safety Evaluation by the Office of Nuclear Reactor Regulation; Topical Report WCAP-16168-NP-A, Revision 2, Risk-Informed Extension of the Reactor Vessel In-service Inspection Interval, Pressurized Water Reactor Owners Group, Project No. 694, July 26, 2011, ADAMS Accession No. ML111600303.
- 5.
BWRVIP-108-A: BWR Vessels and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Shell Welds and Nozzle Blend Radii, EPRI, Palo Alto, CA: 2018. 3002013092.
- 6.
US NRC, Safety Evaluation of Proprietary EPRI Report, BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius (BWRVIP-108), December 19, 2007, ADAMS Accession No. ML073600374.
- 7.
BWRVIP-241-A: BWR Vessel and Internals Project, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii, EPRI, Palo Alto, CA: 2018. 3002013093.
- 8.
US NRC, Safety Evaluation of Proprietary EPRI Report, BWR Vessel and Internals Project, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Shell Welds and Nozzle Blend Radii (BWRVIP-241), April 19, 2013, ADAMS Accession Nos. ML13071A240 and ML13071A233.
- 9.
Code Case N-702, Alternate Requirements for Boiling Water Reactor (BWR)
Nozzle Inner Radius and Nozzle-to-Shell Welds, ASME Code Section XI, Division 1, Approval Date: February 20, 2004.
- 10. U. S. NRC Regulatory Guide 1.147, Revision 19, Inservice Inspection Code Case Acceptability, ASME Code Section XI, Division 1, dated October 2019.
- 11. Code Case N-716-1, Alternative Classification and Examination Requirements, ASME Code,Section XI, Division 1, Approval Date: January 27, 2013.
- 12. Code Case N-706-1, Alternate Examination Requirements of Table IWB-2500-1 and IWC-2500-1 for PWR Stainless Steel Residual and Regenerative Heat Exchangers, ASME Code Section XI, Division 1, Approval Date: January 10, 2007.
10 CFR 50.55a RELIEF REQUEST I4R-24 Revision 0 (Page 9 of 29)
PLANT-SPECIFIC APPLICABILITY
10 CFR 50.55a RELIEF REQUEST I4R-24 Revision 0 (Page 10 of 29)
Plant-Specific Applicability Section 9 of Reference A1 provides requirements that must be demonstrated in order to apply the representative stress and flaw tolerance analyses to a specific plant. Plant-specific evaluation of these requirements for LGS, Units 1 and 2 is provided in Table A-1.
Table A-1 indicates that all plant-specific requirements are met for LGS, Units 1 and 2. Therefore, the results and conclusions of the EPRI report are applicable to LGS, Units 1 and 2.
Table A-1. Plant-Specific Applicability of Reference A1 Representative Analyses to Limerick Units 1 & 2 Category Requirement from Reference A1 Applicability to LGS, Units 1 and 2 General Requirements The HX must operate within the maximum and minimum temperatures and pressures given in Table 5-4 for all known thermal transients.
In Appendix D of this Relief Request, the Limerick Units 1 & 2 transient temperatures are compared to those listed in Table 5-4 of Reference A1. As shown by Table D-1, the Limerick Units 1 & 2 transients are bounded by those from Table 5-4 of Reference A1.
The HX must not experience more than the number of cycles shown for all transients in Table 5-4 over an 80-year operating life. The projected count of actual operating transients, as obtained from the plant Fatigue Management Program, should be used for this comparison; otherwise, projection of the design transients from the FSAR may also be used.
In Appendix D of this Relief Request, the Limerick Units 1 & 2 numbers of transients are compared to those listed in Table 5-4 of Reference A1. As shown by Table D-1, the Limerick Units 1 & 2 transients are bounded by those from Table 5-4 of Reference A1.
10 CFR 50.55a RELIEF REQUEST I4R-24 Revision 0 (Page 11 of 29)
Category Requirement from Reference A1 Applicability to LGS, Units 1 and 2 The materials of the HX shell and nozzles must be low carbon steels operating at temperatures of 70°F or greater.
The Limerick Units 1 & 2 HX nozzles are fabricated of SA-105 material (with the exception of the HX 2B nozzles, which are fabricated of SA-350 LF2 material). The HX shell is SA-516, Grade 70 for all cases. All materials above are low carbon steels and operate at temperatures greater than 70°F, as identified by the temperatures included in Appendix D.
Specific Requirements -
Shell Welds The HX shell welds must conform to the configuration shown in either Figure 1-1 (C1.10 and C1.20) or Figure 1-2 (C1.30) of Reference A1.
The Limerick Units 1 & 2 weld configurations are shown in Figures A-1 and A-2 of this Appendix and show conformance with the figures in Reference A1.
The OD of the vessel shell must be in the range of 45 to 70 inches, and the shell thickness must be in the range of 0.65 to 1.5 inches.
The OD of the vessel shell for Limerick Units 1 & 2 is 55.25 with a thickness of 1.25, which are within the values specified in Reference A1.
Specific Requirements -
Nozzle-to-Shell Welds and Inside Radius Sections The HX nozzle-to-shell weld and inside radius sections must conform to the configuration shown in either Figure 1-3, 1-4, 1-5 or 1-6 of Reference A1.
The Limerick Units 1 & 2 weld configurations are shown in Figures A-3 and A-4 of this Appendix and show conformance with the figures in Reference A1.
The NPS of the piping attached to the HX nozzle must be in the range of 14 inches to 26 inches.
The NPS of the inlet/outlet HX piping is 20 for LGS, Units 1 and 2, which is
10 CFR 50.55a RELIEF REQUEST I4R-24 Revision 0 (Page 12 of 29)
Category Requirement from Reference A1 Applicability to LGS, Units 1 and 2 within the values specified in Reference A1.
REFERENCE A1. Technical Bases for Inspection Requirements for Class 2 BWR Heat Exchanger Nozzle-to-Shell Welds, Nozzle Inside Radius Sections, and Vessel Head, Shell, and Tubesheet-to-Shell Welds, EPRI, Palo Alto, CA: 2020. 3012018473.
10 CFR 50.55a RELIEF REQUEST I4R-24 Revision 0 (Page 13 of 29)
Figure A-1: Limerick Unit 1 RHR HX A and B Vessel
10 CFR 50.55a RELIEF REQUEST I4R-24 Revision 0 (Page 14 of 29)
Figure A-2: Limerick Unit 2 RHR HX A (left) and B (right) Vessels
10 CFR 50.55a RELIEF REQUEST I4R-24 Revision 0 (Page 15 of 29)
Figure A-3: Limerick Unit 1 RHR HX A and B Inlet/Outlet (N3/N4) Nozzles
10 CFR 50.55a RELIEF REQUEST I4R-24 Revision 0 (Page 16 of 29)
Figure A-4: Limerick Unit 2 RHR HX A (top) and B (bottom) Inlet/Outlet (N3/N4) Nozzles (Note: Unit 2 RHR HX B Inlet Nozzle N3 has a reinforcing plate that makes the nozzle-to-vessel weld inaccessible. )
10 CFR 50.55a RELIEF REQUEST I4R-24 Revision 0 (Page 17 of 29)
LIMERICK UNITS 1 & 2 INSPECTION HISTORY
10 CFR 50.55a RELIEF REQUEST I4R-24 Revision 0 (Page 18 of 29)
Table B-1. LGS, Units 1 and 2 Inspection History for Item C1.10 (Heat Exchanger, Shell Circumferential Welds)
Item Date (Outage) Interval/Period Component ID Exam Results (1)
Coverage Relief Request C1.10 3/3/2012 (1R14) 3rd Interval 2nd Period RHR-HXAR-4 NRI 100%
None 3/9/2002 (1R09) 2nd Interval 2nd Period RHR-HXAR-4 NRI 100%
None 2/1/1994 (1R05) 1st Interval 3rd Period RHR-HXAR-4 NRI 100%
None 2/4/1994, 2/7/1994 (1R05) 1st Interval 3rd Period RHR-HXBR-4 NRI 100%
None 3/30/2011 (2R11) 3rd Interval 2nd Period 2AE-205 SG-1 NRI 100%
None 3/25/2009 (2R10) 3rd Interval 1st Period 2AE-205 SG-1 NRI 100%
None 3/6/2003 (2R07) 2nd Interval 1st Period 2AE-205 SG-1 NRI 100%
None 2/24/1995 (2R03) 1st Interval 2nd Period 2AE-205 SG-1 NRI 100%
None 4/15/2015 (2R13) 3rd Interval 3rd Period 2BE-205R W20 NRI 96%
None 2/18/2011 (2R11) 3rd Interval 2nd Period 2BE-205R W20 NRI 100%
None Note: NRI = No Recordable Indications
10 CFR 50.55a RELIEF REQUEST I4R-24 Revision 0 (Page 19 of 29)
Table B-2. LGS, Units 1 and 2 Inspection History for Item C1.20 (Heat Exchanger, Head Circumferential Welds)
Item Date (Outage) Interval/Period Component ID Exam Results (1)
Coverage Relief Request C1.20 3/25/2016 (1R16) 3rd Interval 3rd Period RHR-HXAR-1 NRI 100%
None 3/18/2006 (1R11) 2nd Interval 3rd Period RHR-HXAR-1 NRI 100%
None 2/1/1994 (1R05) 1st Interval 3rd Period RHR-HXAR-1 NRI 100%
None 2/4/1994, 2/7/1994 (1R05) 1st Interval 3rd Period RHR-HXBR-1 NRI, GEO 100%
None 3/26/2009 (2R10) 3rd Interval 1st Period 2AE-205 SG-5 NRI 100%
None 3/6/1997 (2R04) 1st Interval 3rd Period 2AE-205 SG-5 NRI 100%
None 4/18/2019 (2R15) 4th Interval 1st Period 2BE-205R W26 NRI 100%
None 4/15/2015 (2R13) 3rd Interval 3rd Period 2BE-205R W26 NRI 100%
None 2/19/2011 (2R11) 3rd Interval 2nd Period 2BE-205R W26 NRI, GEO 100%
None Notes: NRI = No Recordable Indications GEO = Geometry
10 CFR 50.55a RELIEF REQUEST I4R-24 Revision 0 (Page 20 of 29)
Table B-3. LGS, Unit 1 Inspection History for Item C2.21 (Heat Exchanger, Nozzle-to-Shell Welds)
Item Date (Outage)
Interval/Period Component ID Exam Results (1)
Coverage Relief Request C2.21 4/8/2020 (1R18) 4th Interval 1st Period RHR-HXAR-N4, UT NRI 100%
None 4/8/2020 (1R18) 4th Interval 1st Period RHR-HXAR-N4, MT NRI 100%
None 3/25/2016 (1R16) 3rd Interval 3rd Period RHR-HXAR-N3, UT NRI 100%
None 3/25/2016 (1R16) 3rd Interval 3rd Period RHR-HXAR-N3, MT NRI 100%
None 3/18/2006 (1R11) 2nd Interval 3rd Period RHR-HXAR-N3, UT NRI 100%
None 3/18/2006 (1R11) 2nd Interval 3rd Period RHR-HXAR-N3, MT NRI 100%
None 2/1/1994 (1R05) 1st Interval 3rd Period RHR-HXAR-N3, UT NRI 100%
None 2/1/1994 (1R05) 1st Interval 3rd Period RHR-HXAR-N3, MT NRI 100%
None 4/6/2020 (1R18) 4th Interval 1st Period RHR-HXAR-N4, UT NRI 100%
None 4/6/2020 (1R18) 4th Interval 1st Period RHR-HXAR-N4, MT NRI 100%
None 4/5/2010 (1R13) 3rd Interval 1st Period RHR-HXAR-N4, UT NRI 100%
None 4/5/2010 (1R13) 3rd Interval 1st Period RHR-HXAR-N4, MT NRI 100%
None 3/9/2002 (1R09) 2nd Interval 2nd Period RHR-HXAR-N4, UT NRI 100%
None 3/9/2002 (1R09) 2nd Interval 2nd Period RHR-HXAR-N4, MT NRI 100%
None 2/1/1994 (1R05) 1st Interval 3rd Period RHR-HXAR-N4, UT NRI 100%
None 2/1/1994 (1R05) 1st Interval 3rd Period RHR-HXAR-N4, MT NRI 100%
None 2/7/1994 (1R05) 1st Interval 3rd Period RHR-HXBR-N3, UT NRI 100%
None 2/7/1994 (1R05) 1st Interval 3rd Period RHR-HXBR-N3, MT NRI 100%
None 2/4/1994, 2/7/1994 (1R05) 1st Interval 3rd Period RHR-HXBR-N4, UT NRI 100%
None 2/7/1994 (1R05) 1st Interval 3rd Period RHR-HXBR-N4, MT NRI 100%
None Note: NRI = No Recordable Indications
10 CFR 50.55a RELIEF REQUEST I4R-24 Revision 0 (Page 21 of 29)
Table B-4. LGS, Unit 2 Inspection History for Item C2.21 (Heat Exchanger, Nozzle-to-Shell Welds)
Item Date (Outage)
Interval/Period Component ID Exam Results (1)
Coverage Relief Request C2.21 3/26/2009 (2R10) 3rd Interval 1st Period 2AE-205 N-3-1, UT NRI 100%
None 3/25/2009 (2R10) 3rd Interval 1st Period 2AE-205 N-3-1, MT NRI 100%
None 3/6/1997 (2R04) 1st Interval 3rd Period 2AE-205 N-3-1, UT NRI 100%
None 3/6/1997 (2R04) 1st Interval 3rd Period 2AE-205 N-3-1, MT NRI 100%
None 3/30/2011 (2R11) 3rd Interval 2nd Period 2AE-205 N-4-1, UT NRI 100%
None 3/29/2011 (2R11) 3rd Interval 2nd Period 2AE-205 N-4-1, MT NRI 100%
None 3/6/2003 (2R07) 2nd Interval 1st Period 2AE-205 N-4-1, UT NRI 100%
None 3/6/2003 (2R07) 2nd Interval 1st Period 2AE-205 N-4-1, MT NRI 100%
None 2/24/1995 (2R03) 1st Interval 2nd Period 2AE-205 N-4-1, UT NRI 95%
None 2/24/1995 (2R03) 1st Interval 2nd Period 2AE-205 N-4-1, MT NRI 100%
None 4/15/2015 (2R13) 3rd Interval 3rd Period 2BE-205R N-4-1, UT NRI 53%
ML17275A202 (2) 4/18/2015 (2R13) 3rd Interval 3rd Period 2BE-205R N-4-1, MT NRI 91%
None 2/19/2011 (2R11) 3rd Interval 2nd Period 2BE-205R N-4-1, UT NRI 100%
None 2/17/2011 (2R11) 3rd Interval 2nd Period 2BE-205R N-4-1, MT NRI 100%
None Notes: 1. NRI = No Recordable Indications
- 2. Relief was requested because the examination covered less than 90% of the required volume and/or area.
10 CFR 50.55a RELIEF REQUEST I4R-24 Revision 0 (Page 22 of 29)
Table B-5. LGS, Units 1 and 2 Inspection History for Item C2.22 (Heat Exchanger, Nozzle Inside Radius Section)
Item Date (Outage) Interval/Period Component ID Exam Results (1) Coverage Relief Request C2.22 3/25/2016 (1R16) 3rd Interval 3rd Period RHR-HXAR-N3IR NRI 100%
None 3/18/2006 (1R11) 2nd Interval 3rd Period RHR-HXAR-N3IR NRI 100%
None 2/1/1994 (1R05) 1st Interval 3rd Period RHR-HXAR-N3IR NRI, GEO 100%
None 3/28/2014 (1R15) 3rd Interval 3rd Period RHR-HXAR-N4IR NRI 70%
ML17275A202 (2) 3/9/2002 (1R09) 2nd Interval 2nd Period RHR-HXAR-N4IR NRI 100%
None 2/1/1994 (1R05) 1st Interval 3rd Period RHR-HXAR-N4IR NRI 100%
None 2/4/1994 (1R05) 1st Interval 3rd Period RHR-HXAR-N3IR NRI, GEO 100%
None 2/4/1994 (1R05) 1st Interval 3rd Period RHR-HXBR-N4IR NRI 100%
None 3/26/2009 (2R10) 3rd Interval 1st Period 2AE-205 N-3-1 IR NRI 100%
None 3/6/1997 (2R04) 1st Interval 3rd Period 2AE-205 N-3-1 IR NRI, GEO 100%
None 2/10/1995 (2R03) 1st Interval 2nd Period 2AE-205 N-4-1 IR NRI 100%
None 3/6/2003 (2R07) 2nd Interval 1st Period 2AE-205 N-4-1 IR NRI 100%
None 4/26/2015 (2R13) 3rd Interval 3rd Period 2AE-205 N-4-1 IR NRI 100%
None 4/15/2015 (2R13) 3rd Interval 3rd Period 2BE-205R N-4-1 IR NRI 100%
None 2/19/2011 (2R11) 3rd Interval 2nd Period 2BE-205R N-4-1 IR NRI 100%
None Notes: 1. NRI = No Recordable Indications GEO = Geometry
- 2. Relief was requested because the examination covered less than 90% of the required volume and/or area.
10 CFR 50.55a RELIEF REQUEST I4R-24 Revision 0 (Page 23 of 29)
RESULTS OF INDUSTRY SURVEY
10 CFR 50.55a RELIEF REQUEST I4R-24 Revision 0 (Page 24 of 29)
Overall Industry Inspection Summary The results of an industry survey of past heat exchanger inspections are summarized in Reference C1. Table C-1 provides a summary of the combined survey results for Item Nos. C1.10, C1.20, C1.30, C2.21, C2.22, and C2.32, which includes the BWR Heat Exchangers. The results identify that examination of the items adversely impact outage activities including worker exposure, personnel safety, and radwaste. A total of 86 domestic and international BWR units responded to the survey and provided information representing most BWR plant designs currently in operation. A total of 390 examinations for the components of the affected Item Nos. were conducted on BWR Heat Exchanger components.
Out of a total of 390 examinations, two flaws were identified during the examinations which exceeded the acceptance criteria of ASME Code,Section XI and required flaw evaluation. A subsequent fracture mechanics and fatigue crack growth evaluation showed flaw acceptability, since the calculated maximum acceptable flaw sizes were much larger than the existing flaw sizes including growth until the end of plant life. The welds were examined in two subsequent outages and no changes were noted. No other indications were identified in any in-scope components.
10 CFR 50.55a RELIEF REQUEST I4R-24 Revision 0 (Page 25 of 29)
Table C-1 Summary of Survey Results Item No.
No. of Examinations No. of Reportable Indications C1.10 140 0
C1.20 54 0
C1.30 32 0
C2.21 106 2 (1)
C2.22 58 0
C2.32 0
0 Note: 1. Flaw evaluations were performed to show acceptability of these indications and follow on examinations showed no change in flaw sizes since the original inspections.
REFERENCE C1. Technical Bases for Inspection Requirements for Class 2 BWR Heat Exchanger Nozzle-to-Shell Welds, Nozzle Inside Radius Sections, and Vessel Head, Shell, and Tubesheet-to-Shell Welds, EPRI, Palo Alto, CA: 2020. 3012018473.
10 CFR 50.55a RELIEF REQUEST I4R-24 Revision 0 (Page 26 of 29)
COMPARISON OF LIMERICK GENERATING STATION, UNITS 1 AND 2 TRANSIENTS TO REFERENCE 1 REQUIREMENTS
10 CFR 50.55a RELIEF REQUEST I4R-24 Revision 0 (Page 27 of 29)
Comparison of Limerick Generating Station, Units 1 and 2 Transients to Requirements in Reference [D1]
The LGS, Units 1 and 2 transients are tracked by the plant Fatigue Management Program and the number of cycles encountered as of 2020 are provided in Table D-1 (Reference D1). Table 5-4 of Reference D1 identified the transient cycles used in the underlying PFM and DFM evaluations. The comparison of applicable LGS, Units 1 and 2 transients to the transient limits from Reference D1 is shown in Table D-1.
10 CFR 50.55a RELIEF REQUEST I4R-24 Revision 0 (Page 28 of 29)
Table D-1. Comparison of Limerick Generating Station, Units 1 and 2 General Transients to the Requirements in the EPRI Report [D1] (1)
Transient Maximum Transient Temperature for Limerick Units 1 & 2
[value from D1]
(°F)
Minimum Transient Temperature for Limerick Units 1 & 2
[value from D1]
(°F)
Maximum Transient Pressure for Limerick Units 1 & 2
[value from D1]
(psig)
Minimum Transient Pressure for Limerick Units 1 & 2
[value from D1]
(psig)
No. of Cycles Projected to 60 Years of Operation (2)
Limerick Unit 1, 2 60-Year No. of Thermal Transient Cycles Tabulated in [D1] (2)
Shutdown Cooling 281 [360] (3) 70 [70] (3) 75 [500] (3) 0 [0] (3) 102, 94 (4) 300 Fuel Pool Cooling Assist (5)
N/A [212]
N/A [70]
N/A [500]
N/A [0]
N/A 300 Suppression Pool Cooling (6)
N/A [120]
N/A [70]
N/A [500]
N/A [0]
N/A 480 Steam Condensing (7)
N/A [390]
N/A [70]
N/A [500]
N/A [0]
N/A 10 Notes: 1. The applicable transient information was obtained from Table 5-4 of Reference [D1].
- 2. Per Note 3 of Table 5-4 of the EPRI report [D1], the 60-year cycle counts are shown and compared; the numbers of cycles from the EPRI report were multiplied by 1.33 (= 80/60) to conservatively project them to 80 years for use in the PFM evaluations in Section 8.3 of the EPRI report [D1].
- 3. The maximum and minimum transient temperatures and pressures were obtained for Mode D (normal shutdown after blowdown to the main condenser) for the RHR system from Reference [D3].
- 4. Values obtained from Reference [D2]. Shutdown Cooling is activated once per shutdown of the reactor pressure vessel; therefore, the numbers of shutdown events from Reference [D2] are used to project the number of events until the end of plant life.
- 5. Fuel pool cooling assist has not been performed at either Limerick unit in the past.
- 6. Suppression Pool Cooling is not performed at either Limerick unit as defined by the description for this event in Section 5.2 of the EPRI report [D1]. Therefore, the values shown for this event are all not applicable (N/A) for both Limerick units. Suppression pool cooling using the RHR system is at nearly constant temperatures and pressures (the pressure difference is small, equal to the pump pressure difference); therefore, the fatigue impacts of these events on the heat exchangers are insignificant so these events are not counted and tracked by Reference [D2].
- 7. Per Section 5.4.7.1.1.5 of the Limerick Updated Final Safety Analysis Report (UFSAR), the reactor steam condensing mode is no longer functional. With the exception of an isolated vent line off each Unit 2 heat exchanger, components required for the reactor steam condensing mode have been either abandoned in place or physically removed from the plant. As a result, the Steam Condensing transient evaluated in the EPRI report [D1] is not applicable to Limerick, so the values shown for this event are all not applicable (N/A) for both Limerick units.
10 CFR 50.55a RELIEF REQUEST I4R-24 Revision 0 (Page 29 of 29)
REFERENCES D1. Technical Bases for Inspection Requirements for Class 2 BWR Heat Exchanger Nozzle-to-Shell Welds, Nozzle Inside Radius Sections, and Vessel Head, Shell, and Tubesheet-to-Shell Welds, EPRI, Palo Alto, CA: 2020. 3012018473.
D2. Surveillance Test, Reactor Vessel Thermal Transient Monitoring, ST-1-107-640-1(2).
D3. Process Diagram Residual Heat Removal System, E11-1020-G-002, Sheet 1, Rev. 8, 9/5/06.