ML22307A246
| ML22307A246 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 11/10/2022 |
| From: | Nancy Salgado Plant Licensing Branch III |
| To: | Rhoades D Constellation Energy Generation |
| Wiebe J | |
| References | |
| EPID L-2021-LLR-0035, EPID L-2021-LLR-0036 | |
| Download: ML22307A246 (1) | |
Text
November 10, 2022 Mr. David P. Rhoades Senior Vice President Constellation Energy Generation, LLC President and Chief Nuclear Officer Constellation Nuclear 4300 Winfield Road Warrenville, IL 60555
SUBJECT:
BRAIDWOOD STATION, UNITS 1 AND 2 AND BYRON STATION, UNIT NOS. 1 AND 2 - PROPOSED ALTERNATIVE TO THE REQUIREMENTS OF THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS BOILER & PRESSURE VESSEL CODE (EPIDS L-2021-LLR-0035 and L-2021-LLR-0036)
Dear Mr. Rhoades:
By letter dated May 12, 2021, as supplemented by letters dated November 16, 2021; March 10, 2022; and April 08, 2022, Constellation Energy Generation, LLC (the licensee) submitted requests to the U.S. Nuclear Regulatory Commission (NRC) for the use of alternatives to certain American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), section XI, requirements at Braidwood Station, Units 1 and 2, (Braidwood) and Byron Station, Unit Nos. 1 and 2 (Byron).
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1), the licensee requested to use the proposed alternatives on the basis that they provide an acceptable level of quality and safety. The regulations in 10 CFR 50.55a(z) allow the NRC staff to authorize alternatives to the requirements in paragraphs (b) through (h) of 10 CFR 50.55a.
The requirements for 10-year inservice inspection (ISI) intervals after the initial 120-month inspection interval are established by 10 CFR 50.55a(g)(4)(ii), and require licensees to comply with the Code requirements incorporated by reference in 10 CFR 50.55a(a) 18 months prior to the start of the ISI interval.
As set forth in the enclosed safety evaluation, the NRC staff has determined that the proposed alternatives provide an acceptable level of quality and safety for the fourth inspection interval.
Accordingly, the NRC staff concludes that the licensee has adequately addressed the regulatory requirements set forth in 10 CFR 50.55a(z)(1) for the fourth inspection interval. Since the ASME Code requirements beyond the fourth 10-year ISI intervals at Braidwood and Byron have not yet been established in accordance with 10 CFR 50.55a(g)(4)(ii), no authorization can be made at this time for subsequent ISI intervals following the fourth ISI interval. Therefore, the NRC staff authorizes the proposed alternative RR-I4R-15 at Braidwood and RR-I4R-21 at Byron only for their respective fourth 10-year ISI intervals.
The enclosed safety evaluation documents the technical basis for the NRCs verbal authorizations given April 15, 2022, for RR-I4R-15 at Braidwood and RR-I4R-21 at Byron.
D. Rhoades All other ASME Code, section XI, requirements for which relief was not specifically requested and authorized remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
If you have any questions, please contact Joel S. Wiebe at 301-415-6606 or Joel.Wiebe@nrc.gov.
Sincerely, Digitally signed by Scott P. Wall Scott P. Wall Date: 2022.11.10 09:34:13 -05'00' Nancy L. Salgado, Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. STN 50-456, STN 50-457, STN 50-454, and STN 50-455
Enclosure:
Safety Evaluation cc: Listserv
November 10, 2022 BRAIDWOOD STATION UNITS 1 AND 2; AND BYRON STATION UNITS 1 AND 2 -
AUTHORIZATION AND SAFETY EVALUATION FOR ALTERNATIVE REQUEST NOS. I4R-15 AND I4R-21 (EPID L-2021-LLR-0035 AND 0036)
LICENSEE INFORMATION Licensee: Constellation Energy Generation, LLC (Constellation)
Plant Name and Unit: Braidwood Station Units 1 and 2 (Braidwood, Units 1 and 2)
Byron Station Units 1 and 2 (Byron, Unit Nos. 1 and 2)
Docket Nos.: 50-456 and 50-457 50-454 and 50-455 APPLICATION INFORMATION Submittal Date: May 12, 2021 Submittal Agencywide Documents Access and Management System (ADAMS) Accession No.: ML21133A297 Supplement Dates: November 16, 2021; March 10, 2022; and April 08, 2022 Supplement ADAMS Accession Nos.: ML21320A242 (hereafter, RAI [request for additional information] response), ML22069A580, and ML22098A179 (hereafter, supplemental RAI response)
Applicable Inservice Inspection (ISI) Program Interval: The applicable ISI program intervals are listed in the following table.
Plant ISI Interval Start Date End Date Braidwood Units 1 August 29, 2018 (Unit 1) July 28, 2028 (Unit 1)
Fourth and 2 November 5, 2018 (Unit 2) October 16, 2028 (Unit 2)
Byron Units 1 and 2 Fourth July 16, 2016 July 15, 2025 Alternative Provision: The applicant requested an alternative under Title 10 of the Code of Federal Regulations (10 CFR), paragraph 50.55a(z)(1).
ISI Requirements: For American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME BPV Code) Class 1 welds, the ISI requirements are those specified in paragraph IWB-2411 of the ASME Code, section XI, which requires the licensee to perform volumetric examinations of the following pressurizer (PZR) welds as specified in ASME Code, section XI, table IWB-2500-1 once every 10-year ISI interval.
Enclosure
Examination Category B-B, item No. B2.11, PZR shell-to-head welds, circumferential.
Examination Category B-B, Item No. B2.12, PZR shell-to-head welds, longitudinal.
Examination Category B-D, Item No. B3.110, PZR nozzle-to-vessel welds.
Applicable Code Edition and Addenda: The applicable ASME Code, section XI, editions are listed in Table 1.
Table 1: Applicable ASME Code,Section XI, Editions Plant Edition Braidwood Units 1 and 2 2013 Edition Byron Station Units 1 and 2 2007 Edition, through 2008 Addenda Brief Description of the Proposed Alternative: In section 1 of attachment 1 to the submittal dated May 12, 2021, the licensee stated that the proposed alternative is to increase the ISI interval from the current ASME Code, section XI, requirement of 10 years to the end of period of extended operation for the following PZR welds at Braidwood, Units 1 and 2 (Table 2); and Byron, Unit Nos. 1 and 2 (Table 3). Table 1 of attachment 1 to the submittal dated May 12, 2021, shows that the proposed inspection interval varies from 14.4 to 32.1 years, depending upon the specific component and unit. This was amended in the supplemental RAI response as discussed in Section 11.
Table 2: Braidwood Component ID's ASME ASME Unit Component ID Component Description Category Item No.
1 B-B B2.11 1PZR-01-08A Shell - Lower Head 1 B-B B2.11 1PZR-01-08E Shell - Upper Head 2 B-B B2.11 2PZR-01-08A Shell - Lower Head 2 B-B B2.11 2PZR-01-08E Shell - Upper Head 1 B-B B2.12 1PZR-01-09A Shell Longitudinal Weld 1 B-B B2.12 1PZR-01-09D Shell Longitudinal Weld 2 B-B B2.12 2PZR-01-09A Shell Longitudinal Weld 2 B-B B2.12 2PZR-01-09D Shell Longitudinal Weld 1 B-D B3.110 1PZR-01-N1 Surge Nozzle 1 B-D B3.110 1PZR-01-N2 Spray Nozzle 1 B-D B3.110 1PZR-01-N3 Relief Nozzle 1 B-D B3.110 1PZR-01-N4A Safety Nozzle 1 B-D B3.110 1PZR-01-N4B Safety Nozzle 1 B-D B3.110 1PZR-01-N4C Safety Nozzle 2 B-D B3.110 2PZR-01-N1 Surge Nozzle 2 B-D B3.110 2PZR-01-N2 Spray Nozzle 2 B-D B3.110 2PZR-01-N3 Relief Nozzle 2 B-D B3.110 2PZR-01-N4A Safety Nozzle 2 B-D B3.110 2PZR-01-N4B Safety Nozzle 2 B-D B3.110 2PZR-01-N4C Safety Nozzle
Table 3: Byron Component ID's ASME ASME Unit Component ID Component Description Category Item No.
1 B-B B2.11 1RY-01-S/PC-01 Shell - Bottom Head 1 B-B B2.11 1RY-01-S/PC-05 Shell - Upper Head 2 B-B B2.11 2RY-01-S/PC-01 Shell - Bottom Head 2 B-B B2.11 2RY-01-S/PC-05 Shell - Upper Head 1 B-B B2.12 1RY-01-S/PL-01 Lower Longitudinal Weld 1 B-B B2.12 1RY-01-S/PL-04 Upper Longitudinal Weld 2 B-B B2.12 2RY-01-S/PL-01 Lower Longitudinal Weld 2 B-B B2.12 2RY-01-S/PL-04 Upper Longitudinal Weld 1 B-D B3.110 1RY-01-S/PN-01 Surge Nozzle 1 B-D B3.110 1RY-01-S/PN-02 Spray Nozzle 1 B-D B3.110 1RY-01-S/PN-03 Relief Nozzle 1 B-D B3.110 1RY-01-S/PN-04 Safety Nozzle 1 B-D B3.110 1RY-01-S/PN-05 Safety Nozzle 1 B-D B3.110 1RY-01-S/PN-06 Safety Nozzle 2 B-D B3.110 2RY-01-S/PN-01 Surge Nozzle 2 B-D B3.110 2RY-01-S/PN-02 Spray Nozzle 2 B-D B3.110 2RY-01-S/PN-03 Relief Nozzle 2 B-D B3.110 2RY-01-S/PN-04 Safety Nozzle 2 B-D B3.110 2RY-01-S/PN-05 Safety Nozzle 2 B-D B3.110 2RY-01-S/PN-06 Safety Nozzle For additional details on the licensees request, please refer to the documents located at the ADAMS Accession Nos. identified above.
STAFF EVALUATION The NRC verbally authorized requests RR-I4R-15 at Braidwood and RR-I4R-21 at Byron on April 15, 2022 (ML22105A072).
1 Licensees Basis for Proposed Alternative The licensee referred to the results of the probabilistic fracture mechanics (PFM) analyses in the following Electric Power Research Institute (EPRI) report as the primary basis for proposing to increase the ISI interval for the requested components until the end-of-license for the subject units: non-proprietary EPRI Report 3002015905, Technical Bases for Inspection Requirements for PWR [pressurized-water reactor] Pressurizer Head, Shell-to-Head, and Nozzle-to-Vessel Welds, December 2019 (ML21021A271). This report will be referred to as EPRI Report 15905 from this point forward.
The NRC staffs review first focused on evaluating the PFM analyses in section 8.3 of EPRI Report 15905 and verifying whether the deterministic fracture mechanics (DFM) and PFM analyses in the report support the proposed alternative. The NRC staff reviewed the proposed alternative request for Braidwood, Units 1 and 2; and Byron, Unit Nos. 1 and 2, as a plant-specific alternative. The NRC did not review EPRI Report 15905 for generic use, and this alternative request does not extend beyond the Braidwood, Units 1 and 2; and Byron, Unit
Nos. 1 and 2, plant-specific authorization. The staff previously reviewed a request based on the EPRI Report 15905 in support of a Salem Generating Station, Units 1 and 2, submittal (hereafter Salem submittal). The staff documented its review in Salem Generating Station Units 1 And 2 - Authorization and Safety Evaluation For Alternative Request No. Sc-I4r-200 (EPID L-2020-LLR-0103) (ML21145A189; hereafter Salem SE [safety evaluation]). As part of the Salem review, the NRC staff conducted a thorough review of the generic aspects of the EPRI report and documented its review in the Salem SE. Consequently, for the Braidwood, Units 1 and 2; and Byron, Unit Nos. 1 and 2 review the staff focused on the plant-specific application of the EPRI document.
Consistent with the key principles of the NRC risk-informed approach, the staff also confirmed that the alternative provides sufficient performance monitoring. In addition, the NRC staff reviewed the impacts of the proposed alternative on the subject fourth ISI intervals, and not until end-of-license as requested. This is because the pressurizer weld ISI requirements beyond the fourth 10-year intervals at Braidwood and Byron have not yet been established in accordance with 10 CFR 50.55a(g)(4)(ii), consequently no authorization can be made at this time for subsequent ISI intervals following the fourth inspection interval. This review was to ensure the adequacy of the proposal in the risk-informed context presented by the applicant.
2 Degradation Mechanisms The NRC staff reviewed the submittal for plant-specific circumstances that may indicate degradation mechanism presence and activity sufficiently unique to Braidwood, Units 1 and 2; and Byron, Unit Nos. 1 and 2, to merit additional consideration. The staff found no evidence that conditions at the subject units would require unique degradation mechanism consideration beyond application of the EPRI Report 15905. Specifically, the staff reviewed the subject materials, stress states, and consistency of chemical environment (i.e. reactor coolant) and found them to be consistent with the assumptions made in the EPRI report.
3 Overall PFM Approach The NRC staff confirmed that the overall PFM approach for the Braidwood and Byron application is consistent with the approach taken in the Salem submittal. The review of this approach is documented in the Salem SE. Consequently, the staff confirmed that the overall PFM approach is acceptable because it is consistent with the approach reviewed previously for the Salem submittal.
The NRC staff noted that the acceptance criterion of 1E-06 failures per year (also termed probability of failure, PoF) is tied to that used by the NRC staff in the development of 10 CFR 50.61a, Alternate fracture toughness requirements for protection against pressurized thermal shock events and other similar reviews. In that rule, the reactor vessel through-wall crack frequency (TWCF) of 1E-06 per year for a pressurized thermal shock event is an acceptable criterion because reactor vessel TWCF is conservatively assumed to be equivalent to an increase in core damage frequency, and as such meets the guidance in Regulatory Guide (RG) 1.174, An Approach to for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis. This assumption is conservative because a through-wall crack in the reactor vessel does not necessarily increase the likelihood of core damage. The discussion of TWCF is explained in detail in the technical basis document for 10 CFR 50.61a, NUREG-1806 Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61), August 2007 (ML072830074).
The NRC staff also noted that the TWCF criterion of 1E-06 per year was generated using a very conservative model for reactor vessel cracking. In addition, this criterion exists within a context of reactor pressure vessel surveillance programs and inspection programs. The NRC staff finds that the licensees use of 1E-06 failures per year based on the reactor vessel TWCF criterion is acceptable for the requested PZR welds of Braidwood, Units 1 and 2; and Byron, Unit Nos. 1 and 2, because the impact of a PZR vessel failure is less than the impact of a reactor vessel failure on overall risk; because the subject welds have substantive, relevant, and continuing inspection histories and programs; and because the estimated risks associated with the individual welds are mostly much lower than the system risk criterion (i.e., the system risk is dominated by a small sub-population which can be considered the principal system risk for integrity). The NRC staff further noted that comparing the probability of leakage to the same criterion is conservative because leakage is less severe than rupture. The generic use of a PoF criteria such as 1E-06 per year for individual welds may not be appropriate but based on the discussion above the staff find the application of this criterion acceptable for this plant-specific review for these PZR welds for Braidwood, Units 1 and 2; and Byron, Unit Nos. 1 and 2.
Lastly, the NRC staff noted that acceptance criterion of 1E-06 failures per year is lower and, thus more conservative, than the criterion the NRC staff accepted in proprietary report BWRVIP-05, BWR [boiling-water reactor] Vessel and Internals Project: BWR Reactor Pressure Vessel Weld Inspection Recommendation, September 1995; non-proprietary report BWRVIP-108NP-A, 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 Blend Radii, October 2018 (ML19297F806); and non-proprietary report BWRVIP-241NP-A, BWR Vessel and Internals Project: Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii, October 2018 (ML19297G738). These EPRI reports were developed prior to or around the time the rules for PTS were reevaluated, and as such the acceptance criterion for failure frequency in the reports is based on the guidelines for PTS analysis in RG 1.154, Format and Content of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors that were available at the time. RG 1.154 was later withdrawn in 2011. Both BWR vessel and internal projects included substantive inspection aspects, which was critical to the NRCs findings.
Based on the discussion above, the NRC staff finds the use of the acceptance criterion of 1E-06 failures per year for PoF acceptable for the Braidwood, Units 1 and 2; and Byron, Unit Nos.1 and 2, plant-specific alternative request.
4 Parameters Most Significant to PFM Results NRC staff reviewed the submittal for plant-specific aspects that may diverge from the Salem SE concerning parameters most significant to PFM results. The staff confirmed that the Salem SE review conclusions applied to the subject submittal and found that the parameters most significant to PFM results would be the same and consistent with the staff review documented in the Salem SE, and consequently the approach taken in that review appropriately applies to this review as well.
As discussed in the Salem SE, the sensitivity analysis (SA), sensitivity study (SS), and the NRC staffs observations on the PROMISE software thus identified the following significant parameters or aspects of the PFM analyses that warrant a close evaluation: stress analysis, fracture toughness, flaw crack growth (FCG) rate coefficient (or simply FCG rate), and effect of ISI schedule and examination coverage. The NRC staff discussed and closely evaluated each in
the next four sections of this SE. The NRC staff also evaluated other parameters or aspects of the analyses in section 9 of this SE.
5 Stress Analysis 5.1 Selection of Components and Materials In appendix A of attachment 1 to the submittal, the licensee evaluated the plant-specific applicability of the components and materials selected and analyzed in EPRI Report 15905 to the PZR welds of Braidwood, Units 1 and 2; and Byron, Unit Nos. 1 and 2. The licensee stated that the subject units met the component configuration and material criteria as specified in the EPRI Report 15905. The acceptability of meeting the criteria, however, depends on the acceptability of the component and material selection described in EPRI Report 15905, which the NRC staff evaluated below.
In sections 4.3 and 4.5.1 of EPRI Report 15905, EPRI discussed the variation among PZR designs and selection of the shell-to-head, vessel head, and vessel-to-nozzle welds of a representative PZR vessel. EPRI used this selection for finite element analyses (FEA, see section 5.4 of this SE) to determine stresses in the analyzed PZR welds, in which the licensee referenced for the corresponding PZR welds requested for Braidwood, Units 1 and 2; and Byron, Unit Nos. 1 and 2. In selecting the components, EPRI considered geometry, operating characteristics, materials, field experience with respect to service-induced cracking, and the availability and quality of component-specific information.
The NRC staff reviewed sections 4.3 and 4.5.1 of EPRI Report 15905, and finds that the PZR configurations selected in the report for stress analysis are acceptable representatives for the corresponding PZR shell-to-head welds and nozzle-to-vessel welds requested for the Braidwood, Units 1 and 2; and Byron, Unit Nos. 1 and 2, plant-specific alternative request because differences in R/t ratios are small, and therefore, differences in stresses would be reasonably addressed through the SS on stress in EPRI Report 15905. To verify the dominance of the R/t ratio, the NRC staff reviewed the through-wall stress distributions in sections 7.1 and 7.2 of EPRI Report 15905 to confirm that the pressure stress is dominant, which would confirm the dominance of the R/t ratio. The NRC staff finds that EPRIs conclusion about the R/t ratio being the dominant parameter in evaluating the various configurations to be acceptable for the Braidwood, Units 1 and 2; and Byron, Unit Nos. 1 and 2, plant-specific alternative request since the pressure stress is the dominant stress as evidenced in figures 7-10, 7-11, and 7-22 through 7-25 of EPRI Report 15905.
Section 9.4 of EPRI Report 15905 addresses criteria for plant-specific applicability of the generic analysis. The licensee addressed these criteria in appendix A, tables A-1 through A-4 of attachment 1 to the submittal for Braidwood, Units 1 and 2; and Byron, Unit Nos. 1 and 2, respectively. Table A-1 and A-3 states that the Braidwood, Units 1 and 2; and Byron, Unit Nos.
1 and 2, PZR upper heads, lower heads, and shells are made from SA-533 Grade A, Class 2, a material sufficiently similar to SA-533, Grade B, Class 1 (used in the EPRI Report 15905), and the surge, spray, and safety/relief valves are made from SA-508, Class 2, material. These materials all have specified minimum yield strength of 50 ksi (kilo pound per square inch)
(higher in the case of SA-533, Grade A, Class 2), which is in conformance with the requirements of ASME Code, section XI, Nonmandatory Appendix G. Therefore, the NRC staff finds that the materials for Braidwood, Units 1 and 2; and Byron, Unit Nos. 1 and 2, meet the material applicability criterion of EPRI Report 15905.
Tables A-2 and A-4 of the licensees submittal state that the Braidwood, Units 1 and 2; and Byron, Unit Nos. 1 and 2, PZR nozzles and bottom head weld configurations conform to the figures and the diameter criteria specified in EPRI Report 15905. The licensee illustrated the PZR nozzle and bottom head weld configurations in figures A-1 through A-8 of its submittal. The NRC staff compared the licensees figures to those referenced in EPRI Report 15905. Based on this comparison the staff finds that the PZR bottom head and nozzle weld configurations meet the applicability criterion of EPRI Report 15905.
5.2 Selection of Transients In section 5.2 of EPRI Report 15905, EPRI discussed the thermal and pressure transients under normal and upset conditions considered relevant to PZR shell-to-head welds. EPRI developed a list of transients for analysis, shown in Table 5-6 of EPRI Report 15905, that is applicable to all PZR shell-to-head welds analyzed in the report, based on transients that have the largest temperature and pressure variations. EPRI stated that additional cycles of the loss of load transient addressed the transients not explicitly selected for analysis in EPRI Report 15905.
EPRI also developed a list of insurge/outsurge transients, shown in table 5-9 of EPRI Report 15905, that is applicable to the welds in the PZR bottom head, in addition to the general transients in table 5-6 of EPRI Report 15905. Insurge/outsurge transients are events due to changes in the inventory of reactor coolant within the PZR resulting from the PZRs control of pressure of the reactor coolant system; these changes in reactor coolant inventory cause reactor coolant to flow in and out of the surge nozzle at the bottom of the PZR vessel.
The NRC staff evaluated the EPRI Report 15905 transient selection in detail, as discussed in the Salem SE. The staff confirmed that the generic aspects of the Salem review apply equally to this review. The NRC staff reviewed the discussion of transients in section 5.2 of EPRI Report 15905, and determined that the transients defined in tables 5-6 and 5-9 of EPRI Report 15905 selected for analysis are reasonable for the Braidwood, Units 1 and 2; and Byron, Unit Nos. 1 and 2; plant-specific alternative request because the transient selection was based on large temperature and pressure variations that are conducive to FCG that are expected to occur in PWRs, including a set of insurge/outsurge transients applicable to the welds in the PZR bottom head to account for reactor coolant inventory changes within the PZR. The staff then compared the generic analysis in EPRI Report 15905 to plant-specific information.
In appendices C and D of attachment 1 to the submittal, the licensee evaluated the plant-specific applicability of the transients selected in EPRI Report 15905 to the PZR welds of Braidwood, Units 1 and 2; and Byron, Unit Nos. 1 and 2. EPRI Report 15905 stated that the plant-specific general transients should be bounded by table 5-6 the report, which allows 300 cycles of heatup/cooldown transients and 360 cycles of loss of load transients. The EPRI Report 15905 also stated that plant-specific insurge/outsurge transients must be bounded by table 5-10 of EPRI Report 15905, which allows up to 3,000 cycles depending upon the temperature differential of the transient.
The licensee projected the number of Braidwood heatup/cooldown and loss of load transients to 60 years of operation in tables C-1 and C-2 of the licensees submittal, based upon the number of cycles as of calendar year 2020. According to table C-2 of the licensees submittal, the 60-year projection for Braidwood Unit 1 is 72 heatup/cooldown cycles and 269 loss of load cycles, which is bounded by EPRI Report 15905, table 5-6. The 60-year projection for Braidwood Unit 2 is 94 heatup/cooldown cycles and 281 loss of load cycles, which is bounded by EPRI Report 15905, table 5-6. In Tables D-1 and D-2 of the licensees submittal, the licensee projected the number of insurge/outsurge transients to 60 years of operation, based upon the
number of cycles as of calendar year 2020. According to Table D-2 of the licensees submittal, the 60-year projection for Braidwood, Unit 1, is 62 insurge/outsurge cycles, which is bounded by the strictest criterion of 600 cycles given in Table 5-10 of EPRI Report 15905. The 60-year projection for Braidwood, Unit 2, is 52 insurge/outsurge cycles, which is bounded by the strictest criterion of 600 cycles given in table 5-10 of EPRI Report 15905. Given this discussion, the NRC staff finds that the projected cycles for general transients and insurge/outsurge transients are appropriately bounded by the criterion in the EPRI technical basis report.
The licensee projected the number of Byron heatup/cooldown and loss of load transients to 60 years of operation in tables C-3 and C-4 of the licensees submittal, based upon the number of cycles as of calendar year 2020. According to table C-4 of the licensees submittal, the 60-year projection for Byron, Unit 1, is 66 heatup/cooldown cycles and 21 loss of load cycles, which is bounded by EPRI Report 15905, table 5-6. The 60-year projection for Byron, Unit No.
2, is 62 heatup/cooldown cycles and 27 loss of load cycles, which is bounded by EPRI Report 15905, table 5-6. In tables D-3 and D-4 of the licensees submittal, the licensee projected the number of insurge/outsurge transients to 60 years of operation, based upon the number of cycles as of calendar year 2020. According to Table D-2 of the licensees submittal, the 60-year projection for Byron, Unit No. 1, is 252 total insurge/outsurge cycles, which is bounded by the strictest criterion of 600 cycles given in table 5-10 of EPRI Report 15905. The 60-year projection for Byron, Unit No. 2, is 297 total insurge/outsurge cycles, which is bounded by the strictest criterion of 600 cycles given in table 5-10 of EPRI Report 15905. Given this discussion, the NRC staff finds that the projected cycles for general transients and insurge/outsurge transients are appropriately bounded by the criteria in the EPRI technical basis report.
Based on the review of appendices C and D of attachment 1 to the submittal, the NRC staff finds that the Braidwood, Units 1 and 2; and Byron, Unit Nos. 1 and 2, PZR welds will be appropriately bounded by the transient analyses in EPRI Report 15905; therefore, the analyzed transient loads for the requested PZR components at Braidwood, Units 1 and 2; and Byron, Unit Nos. 1 and 2, are acceptable.
5.3 Other Operating Loads Weld residual stress and cladding stresses are addressed in EPRI Report 15905. The NRC staff documented the review of these aspects in the Salem SE. The staff confirmed that no plant-specific aspects of this submittal required additional consideration, noting in particular the relatively low sensitivity of the EPRI results on residual stress (EPRI Report 15905 table 8-14) and sensitivity studies conducted on stress. Based on this the staff find that there is a very low probability that plant-specific aspects of other operating loads would have a significant effect on probability of leakage or rupture beyond the studies documented in the EPRI report.
Based on the discussion above, the NRC staff finds the treatment of other loads described in this section of the SE acceptable for the requested PZR components of Braidwood, Units 1 and 2; and Byron, Unit Nos. 1 and 2.
5.4 Finite Element Analyses The NRC staff reviewed the FEA conducted in EPRI Report 15905 and documented their review in detail in the Salem SE. The staff confirmed that no plant-specific aspects of this application warranted further review. Based on this, the NRC staff determined that the pressure and thermal stresses calculated through FEA in the EPRI Report 15905 are acceptable for
referencing for the requested PZR welds of Braidwood, Units 1 and 2; and Byron, Unit Nos. 1 and 2.
6 Fracture Toughness In sections 8.2.2.6 and 8.3.2.7 of EPRI Report 15905, EPRI assumed for fracture toughness of ferritic materials an upper-shelf KIC value of 200 ksiin based on the upper-shelf fracture toughness value in the ASME Code, section XI, A-4200. EPRI treated KIC as a random parameter normal distribution with a mean value of 200 ksiin and a standard deviation of five ksiin, stating that these assumptions are consistent with the Boiling Water Reactor Vessels and Internals Project (BWRVIP)-108 project. Further discussion of this topic as it relates to the EPRI Report 15905, and to plant-specific applications, is contained in the Salem SE. The NRC staff confirmed that the evaluation documented in the Salem SE applies to the Braidwood and Byron application without further plant-specific considerations. As discussed in section 5 of this SE, Braidwood, Units 1 and 2; and Byron, Unit Nos. 1 and 2; meet the material criteria in EPRI Report 15905 and, thus, the NRC staff determined that the fracture toughness parameters above are applicable to Braidwood, Units 1 and 2; and Byron, Unit Nos. 1 and 2.
Based on the discussion referenced above and the discussion in section 5 of this SE which confirmed that the materials and transient loads are acceptable for the requested PZR welds of Braidwood, Units 1 and 2; and Byron, Unit Nos. 1 and 2, the NRC staff finds the fracture toughness model in EPRI Report 15905 acceptable for the requested PZR components of Braidwood, Units 1 and 2; and Byron, Unit Nos. 1 and 2.
7 FCG Rate The NRC staff reviewed the FCG rate used in the EPRI Report 15905 and documented its review in detail in the Salem SE. The staff confirmed that no plant-specific aspects of this application warranted further review. Based on the discussion referenced above, the NRC staff finds that the ASME Code, section XI, A-4300, FCG rate used in the EPRI 15905 analyses is acceptable for the requested PZR components of Braidwood, Units 1 and 2; and Byron, Unit Nos. 1 and 2.
8 ISI Examination Coverage EPRI discussed the analyzed examination schedules in chapter 8 of EPRI Report 15905. The NRC staff reviewed the generic ISI schedule and examination coverage modeling used in the EPRI report and documented its review in detail in the Salem SE. In section 5 of the licensees submittal, the licensee described the inspection history of Braidwood, Units 1 and 2; and Byron, Unit Nos. 1 and 2. The NRC staff evaluated the inspection histories as described below.
The licensee provided the inspection history of Braidwood, Units 1 and 2; and Byron, Unit Nos.
1 and 2; in tables 2 and 3, of the submittal. These tables shows that examinations for the subject components were performed preservice and for the first, second, third, and to a limited extent in the fourth ISI intervals. Tables 2 and 3 of the submittal shows that there is no evidence of unacceptable flaws in these components, which is consistent with known operating history.
Table 2 of the submittal indicates that some of the examinations did not meet the ASME Code, section XI, examination coverage requirements. However, licensees are required to submit a relief request under 10 CFR 50.55a(g)(5)(iii) for ASME Code, section XI, examination requirements that are determined by the licensee to be impractical. Therefore, the NRC staff has reviewed and approved the identified cases of lack of coverage.
To assess the potential impact of the lower reported coverages, the NRC staff compared the subject low coverage components to the results reported in the EPRI report. The licensee stated that the lowest coverage listed in tables 2 and 3 of the submittal is 40 percent, and all of the lowest coverages including the 40 percent coverage were reported for component item No. 3.110 components. EPRI performed a SS assuming 50 percent examination coverage and an examination schedule of PSI+20+40+60 (PSI being preservice inspection) showing that the acceptance criterion (PoF of 1E-06 events/year) was met by several orders of magnitude, as described in Section 8.3.5 of EPRI Report 15905. This case is somewhat more severe than that proposed (PSI+10+20+30+50+70 in the supplemental RAI response) and, consequently, provides a reasonable basis to assess the effect of limited coverage. The EPRI Report 15905 indicated no increase in the probability of leakage for item No. 3.110 components due to coverage in table 8-33 for the Combustion Engineering geometry, and a relatively small increase for Babcock &Wilcox geometries.
The likelihood that a generic degradation mechanism or plant-specific flaw is located only and uniquely in the unexamined portion of the subject welds is low based on current operating experience. Furthermore, the likelihood that any unique degradation is significant and growing but has not occurred in an inspected volume of material is quite low. Based on this, the NRC staff found that the likelihood that coverage limitations would present an unacceptable level of risk relative to the current ISI requirements was acceptably low.
Therefore, because examination coverage was otherwise adequately implemented and PFM results for less than essentially 100 percent examination coverage were included in the PoF calculations as discussed above, the NRC finds that the licensee adequately addressed the effect of examination coverage on the PoF values for the requested PZR welds of Braidwood, Units 1 and 2; and Byron, Unit Nos. 1 and 2.
9 Other Considerations The NRC staff reviewed the application and associated references concerning initial flaw depth and length distribution; probability of detection; models; uncertainty; convergence; flaw density; and DFM analysis. The staff previously reviewed the generic aspects of these topics as used in EPRI Report 15905 and documented their review in detail in the Salem SE. The staff confirmed that no plant-specific aspects of this application warranted further review. Based on the discussion referenced above, the NRC staff finds that the application is acceptable as regards initial flaw depth and length distribution; probability of detection; models; uncertainty; convergence; flaw density; and DFM analysis used in the EPRI analyses and is acceptable for the requested PZR components of Braidwood, Units 1 and 2; and Byron, Unit Nos. 1 and 2.
10 PFM Results Relevant to Proposed Alternative In section 8.3.4.1.1 of EPRI Report 15905, EPRI stated that based on the PFM results, after PSI, no other inspections are required for up to 60-80 (depending on component) years of plant operation to meet the acceptance criterion of 1E-06 failures per year. The NRC staff does not find this conclusion acceptable since it does not account for the effect of the combination of the most significant parameters or the added uncertainty of low probability events. More significantly, the NRC staff considers this conclusion to be a risk-based approach inconsistent with NRC policy against reliance on solely risk-based approaches. Post fabrication examinations are critical in supporting necessary performance monitoring goals including
monitoring and trending; bounding uncertainties; validating/confirming analytical results; and providing timely means to identify novel and/or unexpected degradation.
As discussed in section 8 of this SE, the licensee is seeking the alternative ISI schedule of PSI+10+20+30+50+70. Therefore, the NRC staff reviewed the EPRI Report 15905 PFM results for PSI+10+20+30+50+70 and noted that this case was bounded by the base case of PSI+20+40+60 for which results were reported for all sensitivity analyses and studies (not all sensitivity analyses and studies included the PSI+10+20+30+50+70 case).
Table 8-32 of EPRI Report 15905 shows the probability of rupture results for the SS on the combined effect of fracture toughness and stress. These probability of rupture results are for an ISI schedule of PSI+10+20+40+60, which bound the licensees proposed alternative of PSI+10+20+30+50+70 for the reasons the NRC staff previously stated. As shown in Table 8-32 of EPRI Report 15905, the limiting probability of rupture is 3.18E-07 per year, which is below the criterion of 1E-06 per year. The NRC staff noted that if fracture toughness was set to the base case values of 200 ksiin with standard deviation of five ksiin, which the NRC staff found acceptable in Section 6 of this SE, the limiting case would have much more margin from the criterion of 1E-06 per year. This larger margin is shown in the SS on stress in table 8-17 of EPRI Report 15905, which shows that even with a stress multiplier of 1.80, the limiting probability of rupture is 2.50E-09 per year.
The results in tables 8-17 and 8-32 of EPRI Report 15905 discussed above assume 100 percent examination coverage. As mentioned in section 8 of this SE, the licensee showed in table 4 of the submittal that the examination coverage for several welds were well below the ASME Code coverage requirements. As discussed in section 8 of this SE, the impact of weld coverage was quite low relative to the modeled probabilities. Based on the relatively low probabilities calculated for the subject item No. 3.110 components for which lesser coverages were obtained and the SA and studies conducted, the staff finds that the EPRI Report 15905 conclusions regarding these components are acceptable as regards PFM results.
Finally, the NRC staff noted that since the licensees proposed alternative is through 60 years of operation, the probability values should be based on 60 years of operation. Tables 8-17 and 8-18 of EPRI Report 15905 are for 80 years of operation and at 60 years of operation, the results could be up to 80/60 = 1.3 times larger since the number of failures would be divided by 60 years instead of 80 years (assuming the number of failures have been reached by 60 years).
As discussed in section 3 of this SE, PoF at a given time is estimated as the fraction of the total number of realizations that the computed failure time is less than the given time (i.e., PoF is the number of failure times within a given time divided by the total number of realizations). Since the number of failure times could be reached before 60 years, the PoF value could be the same at 60 years and at 80 years. And since the licensees proposed alternative is through 60 years of operation, this PoF value should be divided by 60 years instead of 80 years to obtain the PoF per year value. The NRC staff determined that this factor of 1.3 has no impact on the NRC staffs discussion of the PFM results in EPRI Report 15905 in the preceding paragraphs. Thus, the NRC staff determined that the PFM analyses in EPRI Report 15905 adequately address uncertainties in the PoF values relevant to the licensees proposed alternative of PSI+10+20+30+50+70 for the requested PZR welds of Braidwood, Units 1 and 2; and Byron, Units 1 and 2.
Based on the discussion above, the NRC staff finds that the proposed alternative of PSI+10+20+30+50+70 for the requested PZR welds of Braidwood, Units 1 and 2; and Byron,
Unit Nos. 1 and 2; would result in a PoF per year that is below the acceptance criterion of 1E-06 per year.
11 Performance Monitoring and Associated Considerations Performance monitoring such as ISI programs, is a necessary component per the NRC five principles of risk-informed decision making. Analyses, such as PFM analyses, work in concert with performance monitoring to provide a mutually supporting and diverse basis for maintaining reasonable assurance of adequate safety. In the context of this submittal, an adequate performance monitoring program must provide direct evidence of the presence and/or extent of degradation; validation/confirmation of continued adequacy of associated analyses; and a timely method to detect novel/unexpected degradation. These characteristics were presented, for example, at a March 4, 2022, public meeting (ML22053A171 and ML22060A277); agenda and slides respectively.) The NRC staff reviewed the application accordingly.
The NRC staff noted that the application did not directly address performance monitoring. The staff issued an RAI (the supplemental RAI) requesting that the licensee clarify the relationship between their proposed alternative and the adequacy of the resulting performance monitoring.
The licensee, in their supplemental RAI response, stated that they believed their analysis demonstrated that the inspections already performed constitute adequate performance monitoring for the requested PZR welds for the submittal period. The supplemental RAI response also provided an inspection program in addition to the original proposal that would ensure that no more than 20 years elapses between performance of ASME Code, section XI, examinations for Categories B-B and B-D components included in the performance monitoring plan. The licensee noted that inspections at other plants will continue both domestically and internationally, providing a continuing sampling basis to confirm the general assumptions in the PFM analysis in EPRI Report 15905. Finally, the licensee provided how they would handle any indications that would be detected in the proposed inspections, and how the inspection program may be expanded if new unacceptable indications were found.
The NRC staff reviewed the submitted information in the context of the characteristics of adequate performance monitoring. The staff examined the proposed alternative as it relates to the three functions of performance monitoring noted above. The staff noted that the submittal makes no change regarding the first two functions, and consequently focused its review on the third, namely whether the submittal would allow for timely identification of novel and/or unexpected degradation. The staff noted that this requires a sampling schema that can provide reasonable assurance regarding handling of several potential uncertainties and that these uncertainties can be expressed as parameter, model, and completeness (discussed, for example, in Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking, NUREG-1855, ML17062A466). The parameter and model uncertainties were considered by the staff in the review of the analysis documented above (see section 9 of this SE). The licensee stated in their supplemental RAI response that based on the analyses, the timely identification of degradation in the subject components could be properly conducted on a timescale an order of magnitude longer than the requested period of relief.
Using the methodology described above, the staff verified this assertion and finds it acceptable as it relates to the parameter and model uncertainties because there would need to be an unlikely number of adverse differences between the modeling and the subject components to forecast an inappropriate level of risk-based on the modeling approach.
Regarding completeness uncertainty (e.g., the uncertainty related to what is not modeled, not yet understood, or not yet found) the NRC staff considered several pertinent considerations.
The staff noted that a component-specific degradation mechanism would likely have become apparent during the already performed inspections. Further, the staff noted that any novel mechanisms likely to emerge later in the lifespan of a component would also likely be slow-growing in the subject environment(s) (hence, the late emergence of the mechanism).
Consequently, the time between detection of the initiation of degradation and the degradation threatening the integrity of the components would be relatively long. This long-lead time between initiation of degradation and potential component loss of function would reasonably provide time for this mechanism to be identified provided that the proposed number of inspections had sufficient sensitivity to emerging degradation mechanisms in the sampled population(s). The NRC staff finds that the assumptions listed above are acceptable to support a review of the sensitivity of the submittal.
The supplemental RAI included that the following welds would be examined by the end of 2034:
Table 4: Supplemental RAI Response 2034 Inspections Unit ASME ASME Component ID Component Description Category Item 2 B-B B2.11 2RY-01-S/PC-05 Shell - Upper Head 2 B-B B2.12 2RY-01-S/PL-04 Upper Longitudinal Weld 2 B-D B3.110 2RY-01-S/PN-02 Spray Nozzle 2 B-D B3.110 2RY-01-S/PN-03 Relief Nozzle 2 B-D B3.110 2RY-01-S/PN-04 Safety Nozzle 2 B-D B3.110 2RY-01-S/PN-05 Safety Nozzle 2 B-D B3.110 2RY-01-S/PN-06 Safety Nozzle The licensee states that these inspections would provide sufficient sampling of representative welds covering all subject units. This approach, uniquely proposed here for two very similar sites, required the NRC staff to evaluate both the program sensitivity, the handling of any potential indications, and potentially necessary inspection program expansion.
To evaluate the proposed inspection program sensitivity to emerging detectable degradation the NRC staff conducted independent Monte Carlo based inspection simulations (a computational technique allowing the staff to forgo using complex statistical methods and their related situational assumptions). This entailed modeling simulated plant and fleet inspection scenarios to establish the likelihood of detection if a new generic mechanism is active in similar materials in similar components, and is detectable by the inspection technique. The staff found that while the sensitivity to component-specific degradation would be lower under the proposal (for degradation occurring during the period of the proposal), it would be very unlikely that a novel degradation mechanism occurring in the family of similar material and component designs analyzed in EPRI Report 15905 would go undetected. The staff found that that this approach is acceptable due to the unusually close design, fabrication, and operational characteristics of the Braidwood and Byron units (e.g., inspecting components in one unit to represent four at two sites). As such, the staff concludes that the proposal would enable timely detection of novel degradation sufficient to meet the completeness uncertainty goals of an adequate performance monitoring program.
The licensee also described follow-on actions if an indication was found in the seven proposed inspection components. Specifically, any indications exceeding the acceptance standards of IWB-3500 will be evaluated as required by ASME Code, section XI, and the site corrective action program. Further, the additional examination and successive inspection requirements of
ASME Section XI also apply. Finally, any new unacceptable indications identified as part of the performance monitoring plan ... will result in the same population of welds being examined at Byron, Unit 1 and Braidwood, Units 1 and 2 during the next regularly scheduled refueling outage.
The NRC staff reviewed these proposals and found them both appropriate and necessary to conclude that the proposed performance monitoring plan would appropriately disposition any potential indications; and expand inspections if new unacceptable indications were found. The staff finds that this expansion is a critical component of accepting inspections at one unit as representative for all four units. If no indications are found, this confirms the modeling assumptions and results. If unacceptable indications were to be found, these would be contrary to the modeling results and an expansion of inspections would be necessary to determine the true state of the components. Based on the above the NRC staff determined that the proposed inspections, indication resolution, and inspection expansions proposed together support an acceptable performance monitoring plan.
Based on the discussion above the NRC staff concludes that the applicants proposed alternative will, for the purposes of providing adequate performance monitoring, provide an acceptable level of quality and safety because it will continue to provide direct evidence supporting the assurance of component integrity in a timely fashion, appropriately disposition indications, and critically expand inspections to all four subject units if warranted.
CONCLUSION As set forth above, the NRC staff has determined that the proposed alternative in the licensees application, as supplemented, would provide an acceptable level of quality and safety.
Accordingly, the NRC staff concludes that the licensee has adequately addressed the regulatory requirements in 10 CFR 50.55a(z)(1). Therefore, the NRC staff authorizes the use of proposed alternatives RR-I4R-15 at Braidwood and RR-I4R-21 at Byron for the remainder of the fourth 10-year interval at the subject units. Verbal approval of these alternatives was provided on April 15, 2022. This safety evaluation provided final documentation of the staffs review.
The NRCs authorization of the proposed alternative does not infer or imply the approval of EPRI Report 15905 for generic use.
All other ASME Code, section XI, requirements for which an alternative was not specifically requested and authorized remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
Principal Contributors: Dan Widrevitz, NRR Michael Benson, NRR Date: November 10, 2022
ML22307A246 OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DEX/ESEB/BC NRR/DORL/LPL3/BC NAME JWiebe SRohrer ABuford NSalgado (SWall for)
DATE 11/03/2022 11 / 8 /2022 9/27/2022 11/10/2022