ML22195A025

Authorization and Safety Evaluation for Alternative Request No. ISI-05-016

ML22195A025
Person / Time
Site:	Calvert Cliffs
Issue date:	01/03/2023
From:	Plant Licensing Branch 1
To:	Rhoades D Constellation Energy Generation
Marshall M, NRR/DORL/LPLI, 415-2871
References
EPID L-2021-LLR-0036
Download: ML22195A025 (15)

Text

January 3, 2023 CALVERT CLIFFS NUCLEAR POWER PLANT, UNITS 1 AND 2 - AUTHORIZATION AND SAFETY EVALUATION FOR ALTERNATIVE REQUEST NO. ISI-05-016 (EPID L-2021-LLR-0036)

LICENSEE INFORMATION Licensee: Constellation Energy Generation, LLC (Constellation)

Plant Name and Unit: Calvert Cliffs Nuclear Power Plant Units 1 and 2 (Calvert Cliffs, Units 1 and 2)

Docket Nos.: 50-317 and 50-318 APPLICATION INFORMATION Submittal Date: May 12, 2021 Submittal Agencywide Documents Access and Management System (ADAMS) Accession No.: ML21133A297 (Note: the incoming request was for Braidwood Station, Units 1 and 2, Byron Station, Units 1 and 2, and Calvert Cliffs, Units 1 and 2, however this authorization and safety evaluation is only for the request for Calvert Cliffs, Units 1 and 2, disposition of the other plants will be handled by separate correspondence).

Supplement Dates: November 16, 2021, and April 29, 2022 Supplement ADAMS Accession Nos.: ML21320A242 (hereafter, request for addition information (RAI) response); ML22119A110 (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 Calvert Cliffs, Units 1 Fifth July 1, 2019 June 30, 2029 and 2 Alternative Provision: The licensee 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 BPV Code,Section XI, which requires the licensee to perform volumetric examinations of the following pressurizer (PZR) welds as specified in ASME BPV Code,Section XI, Table IWB-2500-1 once every 10-Year ISI interval.

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 BPV Code,Section XI editions are listed in Table 1, below.

Table 1: Applicable ASME Code,Section XI Editions Plant Edition Calvert Cliffs, Units 1 and 2 2013 Edition 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 for the following PZR welds at Calvert Cliffs, Units 1 and 2:

Table 2: Calvert Cliffs, Units 1 and 2 Component ID's ASME ASME Weld Component Unit Examination Component ID Item No. Description Category 1 B-B B2.11 3-401 Shell - Lower Head 1 B-B B2.11 8-411 Shell - Upper Head 2 B-B B2.11 3-401 Shell - Lower Head 2 B-B B2.11 8-411 Shell - Upper Head 1 B-B B2.12 2-401A Upper Shell At 270 Deg.

1 B-B B2.12 2-401B Upper Shell At 90 Deg.

1 B-B B2.12 2-401C Lower Shell At 180 Deg.

1 B-B B2.12 2-401D Lower Shell At 0 Deg.

2 B-B B2.12 2-401A Upper Shell At 270 Deg.

2 B-B B2.12 2-401B Upper Shell At 90 Deg.

2 B-B B2.12 2-401C Lower Shell At 180 Deg.

2 B-B B2.12 2-401D Lower Shell At 0 Deg.

1 B-D B3.110 4-404 Surge Nozzle 1 B-D B3.110 4-405 Spray Nozzle 1 B-D B3.110 16-405A Safety & Relief Nozzle 1 B-D B3.110 16-405B Safety & Relief Nozzle 2 B-D B3.110 4-404 Surge Nozzle 2 B-D B3.110 4-405 Spray Nozzle 2 B-D B3.110 16-405A Safety & Relief Nozzle 2 B-D B3.110 16-405B Safety & Relief Nozzle For additional details on the licensees request, please refer to the documents located at the ADAMS Accession Nos. identified above.

STAFF EVALUATION 1.0 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, dated December 2019 (ML21021A271). This report will be referred to as EPRI report 15905 from this point forward.

The U.S. Nuclear Regulatory Commission (NRC) staffs review 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 Calvert Cliffs, Units 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 Calvert Cliffs, Units 1 and 2, plant-specific authorization. NRC staff previously reviewed a request based on EPRI report 15905 in support of a Salem Generating Station, Units 1 and 2, submittal (hereafter Salem submittal). The NRC 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). As part of the Salem submittal 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 Calvert Cliffs review the NRC staff focused on the plant-specific application of the EPRI document.

In addition, the NRC staff reviewed the impacts of the proposed alternative on the sufficiency of performance monitoring for the subject components relative to the extent of time this alternative covers. For the subject components at the two Calvert Cliffs units, this time ranged from 14.4 to 18.5 years and thus represents more than one ASME BPV Code ISI interval as indicated in Table 3, shown on the next page.

Table 3: Proposed period between previous examination and end-of-license Length of End of Time Until Current Next Date of ASME Item Licensed Inspection Station Description Last Category No. Operating for Inspection Period This (60 Years) Request (Years)

Calvert Pressurizer, Shell-to-Head 07/31/2034 Cliffs 1 B-B B2.11 02/25/2020 14.4 Welds, Circumferential Pressurizer, shell-to-Head B-B B2.12 02/25/2020 14.4 Welds, Longitudinal Pressurizer, Nozzle-to-B-D B3.110 02/21/2016 18.5 Vessel Welds Calvert Pressurizer, Shell-to-Head 08/13/2036 Cliffs 2 B-B B2.11 03/13/2021 15.4 Welds, Circumferential Pressurizer, shell-to-Head B-B B2.12 03/12/2021 15.4 Welds, Longitudinal Pressurizer, Nozzle-to-B-D B3.110 03/10/2021 15.4 Vessel Welds 2.0 Degradation Mechanisms The NRC staff reviewed the submittal for plant-specific circumstances that may indicate degradation mechanism presence and activity sufficiently unique to Calvert Cliffs, Units 1 and 2 to merit additional consideration. The NRC staff found no indication that conditions at the subject units would warrant unique degradation mechanism consideration beyond application of the EPRI report 15905. Specifically, the NRC staff reviewed the subject materials, stress states, and consistency of chemical environment (i.e., reactor coolant) and determined they are consistent with the assumptions made in the EPRI report.

3.0 Overall PFM Approach The NRC staff confirmed that the overall PFM approach for the Calvert Cliffs application is consistent with the approach taken in the Salem submittal. The NRC staff review of this approach is documented in the Salem SE. Consequently, the NRC staff confirmed that the overall PFM approach is acceptable because it is consistent with the approach reviewed previously for the Salem submittal and for the reasons stated in the Salem SE.

The NRC staff notes that the acceptance criterion of 1E-06 failures per year (also termed Probability of Failure or 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 criteria in NRCs 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 (ML17317A256). 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), dated 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 Calvert Cliffs, Units 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 and relevant 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. Generically the 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 Calvert Cliffs, Units 1 and 2.

Lastly, the NRC staff notes that the 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, dated 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, dated 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, dated 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 Internals Project topical reports 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 Calvert Cliffs, Units 1 and 2 plant-specific alternative request.

4.0 Parameters Most Significant to PFM Results NRC staff reviewed the submittal for plant-specific conditions that may diverge from those considered in the Salem SE concerning parameters most significant to PFM results. The NRC staff confirmed that the Salem SE review conclusions applied to the subject submittal and determined that the parameters most significant to PFM results would be the same and consistent with the NRC staff review documented in the Salem SE. Consequently, the approach taken in that review 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 computer software thus identified the following significant parameters or aspects of the PFM analyses that warrant a close evaluation. These include stress analysis, fracture toughness, fatigue crack growth (FCG) rate coefficient (or simply FCG rate), and effect of ISI schedule and examination coverage. The NRC staff discusses each of these matters in the next four sections of this SE. The NRC staff evaluation of other parameters or aspects of the analyses is documented in Section 9.0 of this SE.

5.0 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 Calvert Cliffs, Units 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 as described 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, which the licensee referenced for the corresponding PZR welds requested for Calvert Cliffs, Units 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, nozzle-to-vessel welds, and head welds requested for the Calvert Cliffs, Units 1 and 2 plant-specific alternative request. Specifically, the differences in radius-to-thickness ratios (R/t) 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 confirmed that the pressure stress is the dominant stress as shown by Figures 7-10, 7-11, and 7-22 through 7-25 of EPRI report 15905.

Accordingly, 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 Calvert Cliffs, Units 1 and 2 plant-specific alternative request.

Section 9.4 of EPRI report 15905 addresses criteria for plant-specific applicability of the generic analysis and indicates that materials are acceptable if they conform to ASME B&PV Code,Section XI, Nonmandatory Appendix G. The licensee addressed these criteria in Appendix A, Table A-5 of Attachment 1 to the submittal for Calvert Cliffs, Units 1 and 2, respectively. Table A-5 states that the Calvert Cliffs, Units 1 and 2 PZR upper heads, lower heads, and shells are made from SA-533, Grade B, Class 1, 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, which is in conformance with ASME Code Section XI, Nonmandatory Appendix G.

Therefore, the staff finds that the materials for Calvert Cliffs, Units 1 and 2, meet the material applicability criterion of EPRI report 15905.

Table A-5 of the licensees submittal states that the Calvert Cliffs PZR nozzles and bottom head weld configurations conform to the figures and the diameter criterion specified in EPRI report 15905. The licensee illustrated the PZR nozzle and bottom head weld configurations in Figures A-9 through A-12 of its submittal. The NRC staff compared the licensees figures to those referenced in EPRI report 15905. The NRC staff finds that the PZR bottom head and nozzle weld configurations meet the applicability criteria 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 NRC staff confirmed that the generic aspects of the Salem submittal 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 Calvert Cliffs, Units 1 and 2, plant-specific alternative request. Specifically, the EPRI 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 NRC staff then compared the generic analysis in the 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 Calvert Cliffs, Units 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. EPRI report 15905 also stated that plant-specific insurge/outsurge transients must be bounded by Table 5-10 of the report, which allows up to 3,000 cycles depending upon the temperature differential of the transient.

In Tables C-3 and C-4 of the licensees submittal, the licensee projected the number of heatup/cooldown and loss of load transients to 60 years of operation, 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 Calvert Cliffs Unit 1 is 66 heatup/cooldown cycles and 21 loss of load cycles, which is bounded by Table 5-6 of EPRI report 15905. The 60-year projection for Calvert Cliffs Unit 2 is 62 heatup/cooldown cycles and 27 loss of load cycles, which is also bounded by Table 5-6 of EPRI report 15905.

The licensee stated that Calvert Cliffs does not track thermal transients and instead provided fatigue usage factors and environmental-assisted fatigue usage factors. The NRC staff noted that compliance with design requirements related to fatigue usage is not a basis to use the PFM

results of the EPRI report to justify relief from ASME Code,Section XI examination requirements. The NRC staff issued an RAI on this topic (ML21287A032). In its response (ML21320A242), the applicant stated that they derived the number of projected insurge/outsurge transients for Calvert Cliffs, Units 1 and 2 (Table RAI1-1 of the RAI response) based on a ratio of the actual Calvert Cliffs, Units 1 and 2 heatup/cooldown cycles. The licensee compared the ratio of insurge/outsurge transients derived in this way to the EPRI report simulation of 500 cycles of each. These Calvert Cliffs transients were then further divided by the size of the thermal delta and compared to the thermal deltas used in EPRI report 15905 (Table RAI1-2 of the RAI response). The NRC staff confirmed that this approach would provide a reasonable basis to compare Calvert Cliffs operation to the conditions assessed in EPRI report 15905. The NRC staff noted that the projected transients were bounded by the EPRI report simulation by a reasonable margin. Based both on this margin and the SS and SA discussed in the EPRI report and Salem SE, the NRC staff finds that the projected cycles for general transients and insurge/outsurge transients at Calvert Cliffs are bounded by the criteria in EPRI report 15905.

Based on the review of Appendices C and D of Attachment 1 to the submittal described above, the NRC staff finds that the Calvert Cliffs, Units 1 and 2 PZR welds will be bounded by the transient analyses in EPRI report 15905; therefore, the analyzed transient loads for the requested PZR components at Calvert Cliffs, Units 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 of the EPRI report in the Salem SE. The NRC staff confirmed that no Calvert Cliffs plant-specific aspects of this submittal warranted 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 NRC staff finds that there is a very low probability that plant-specific aspects of other operating loads would have a significant effect on the 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 Calvert Cliffs, Units 1 and 2.

5.4 Finite Element Analyses The NRC staff reviewed the FEA conducted in EPRI report 15905 and documented its review in detail in the Salem SE. The NRC staff confirmed that no Calvert Cliffs 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 Calvert Cliffs, Units 1 and 2.

6.0 Fracture Toughness In Sections 8.2.2.6 and 8.3.2.7 of EPRI report 15905, EPRI used for fracture toughness of ferritic materials, an upper-shelf KIC value of 200 kilopound-force per square inch square root inches (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 5 ksiin, stating that these assumptions are consistent with the 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 Calvert Cliffs application as well without further plant-specific considerations. As discussed in Section 5 of this SE, Calvert Cliffs, Units 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 Calvert Cliffs, Units 1 and 2.

Based on the discussion referenced above and the discussion in Section 5.0 of this SE, which confirmed that the materials and transient loads are acceptable for the requested PZR welds of Calvert Cliffs, Units 1 and 2, the NRC staff finds the fracture toughness model in EPRI report 15905 acceptable for the requested PZR components of Calvert Cliffs, Units 1 and 2.

7.0 FCG Rate The NRC staff reviewed the FCG rate used in EPRI report 15905 and documented its review in detail in the Salem SE. The NRC 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 Calvert Cliffs, Units 1 and 2.

8.0 ISI Schedule and 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 for Calvert Cliffs, Units 1 and 2. The NRC staff evaluated the inspection histories as described below.

The licensee provided the inspection history of Calvert Cliffs, Units 1 and 2 in Table 4 of the submittal. This table shows that examinations were performed pre-service and at 10, 20, 30, and 40 years following the initial commercial operation for most of the components in scope of this relief request. Some of the components were inspected in the fifth ISI interval. Table 4 of the submittal shows that there is no evidence of detrimental flaws in these components, which is consistent with known operating history.

Table 4 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 previously reviewed and approved the identified cases of lack of coverage. Furthermore, the licensee stated that the lowest coverage listed in Table 4 of the submittal is 28%. EPRI performed a SS assuming 50% examination coverage and an examination schedule of PSI+20+40+60 (PSI being pre-service inspection, and each subsequent number being the year after commencement of operation in which the inspection is repeated) showing that the acceptance criterion (PoF of 1E-06 events/year) was met, as described in Section 8.3.5 of EPRI report 15905. However, the staff notes that the acceptance criterion was not met for PRSHC-BW-2C under the Base Case with 50% Coverage scenario listed in Table 8-33 of EPRI Report 15905 and consequently the impacts of coverage may be significant. Table 4 of the submittal also included four PZR head welds that were not inspected (or that records were not made of any inspection) after PSI. The NRC staff issued an RAI on this (ML21287A032).

In its response, the licensee provided further analysis summarized in RAI Table RAI2-2 analyzing the effects of limited examination coverage at Calvert Cliffs, Units 1 and 2. This analysis compared the probability of rupture and leakage with and without ISI relief (PSI + ISI

every ten years versus PSI + 10 + 20 +30 + 40 + 70) and found that the change in leakage probability, based on limited coverage, from the model analysis to be very low. Specifically, a change in leakage probability was only apparent for the 4-404 components (see Component IDs in Table 2 of this SE) and was on the order of 10-8 between the two scenarios. The NRC staff notes that the most significant change in leakage probability was associated with components that were not the most significant in leakage probability for the sensitivity studies (that being Examination Category B2.22 components; versus welds 4-404 identified in Table 2 of this SE which are Examination Category B3.110 components).

Similarly, welds 2-401B and 2-401C identified in Table 2 of this SE for both units have not been sampled under the ISI program. The NRC staff issued an RAI concerning this. The applicant provided simulations summarized in RAI Table 3-1 pertaining to the coverage of the weld type to which welds 2-401B and 2-401C belong to (Examination Category B-B, Item B2.12). The NRC staff reviewed these results and finds that the risk delta due to missed examination coverage is on the order of 10-8 consistent with the scenarios discussed above.

The likelihood that a generic degradation mechanism or plant-specific flaw is located only in the unexamined portion of the subject welds is low based on current operating experience. In particular, any generic degradation would likely also occur in portions of the PZR welds that are inspected. The additional analysis provided in the RAI responses support that the modeled risk impact of the proposed alternative for these welds is likewise small. Based on this, the NRC staff finds that the likelihood that omission of examination coverage for these welds would present an unacceptable level of risk associated with this alternative request is 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 staff finds that the licensee adequately addressed the effect of examination coverage on the PoF values for the requested PZR welds of Calvert Cliffs, Units 1 and 2.

9.0 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 NRC staff previously reviewed the generic aspects of these topics as used in the EPRI report and documented its review in detail in the Salem SE. The NRC staff further reviewed the application for any potential plant-specific considerations. The NRC staff issued an RAI concerning inspection coverage, which relates to uncertainty. The resolution of this RAI is discussed in Section 8.0 of this SE. 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 Calvert Cliffs, Units 1 and 2.

10.0 PFM Results Relevant to Proposed Alternative In Section 8.3.4.1.1 of reference 1 to the submittal, 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 calling for defense-in-depth, rather than 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 + 40 + 70. Therefore, the NRC staff determined that PFM results for PSI + 10 +

20 + 30 + 40 + 70 are the results relevant to the licensees proposed alternative.

The NRC staff noted that even though EPRI 15905 report does not have PoF results for PSI +

10 + 20 + 30 + 40 + 70, it has results for PSI + 20 + 40 + 60 or PSI + 10 + 20 + 40 + 60, which provide some context. Therefore, the NRC staff evaluated the PFM results in the SS in Section 8.3.4.3 of EPRI report 15905 relevant to the proposed alternative ISI schedule by assessing the results for PSI + 20 + 40 + 60 or PSI + 10 + 20 + 40 + 60. Note, the NRC staff also reviewed results for several plant-specific cases presented in the RAI response. These plant-specific cases are discussed in Section 8.0 of this SE providing more detailed assessment of the plant-specific circumstances.

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 + 40 + 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 a standard deviation of 5 ksiin, which the NRC staff found acceptable in Section 6.0 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 8.0 of this SE, the licensee showed in Table 4 of the submittal that the examination coverage for weld 4-404 (see Table 2 of this SE) of Unit 2 PZR was as low as 28.2 percent. The licensee provided additional sensitivity runs including the weld type with a 28.2 percent examination coverage for this weld in the RAI response. This is discussed further in Section 8.0 of this safety evaluation.

Finally, the NRC staff notes 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 3.0 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 failures within a given duration divided by the total number of realizations). Since failure 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 + 40 for the requested PZR welds of Calvert Cliffs, Units 1 and 2.

Based on the discussion above, the NRC staff finds that the proposed alternative of PSI + 10 +

20 + 30 + 40 for the requested PZR welds of Calvert Cliffs, Units 1 and 2 would result in a PoF per year that is below the acceptance criterion of 1E-06 per year.

11.0 Performance Monitoring 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 a facility within its licensing basis with respect to a particular set of safety issues. In the context of this submittal, an adequate performance monitoring program must provide direct evidence of the presence and 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 noted that the application did not directly address performance monitoring. The NRC staff issued an RAI (ML22091A239) requesting that the licensee clarify the relationship between its proposed alternative and the adequacy of the resulting performance monitoring.

The licensee, in its supplemental RAI response (ML22119A110), stated that its analysis demonstrated that the inspections already performed constitute adequate performance monitoring for the requested PZR welds for the submittal period. In a letter from David T.

Gudger, Constellation Energy Generation, LLC to Document Control Desk, NRC, dated April 2022, the licensee cited a performance monitoring plan submitted in response to an RAI for a similar pressurizer alternative (ML22098A179; or April 2022 letter) that was accepted by the NRC staff as sufficient. This letter provided that no subject pressurizer component will operate for more than 20 years without an associated inspection. The licensee proposed that the acceptability of the April 2022 letter applies to the Calvert Cliffs alternative given that the Calvert Cliffs request encompasses a shorter span of time than the other proposal. Specifically, that approval of this alternative would not result in a longer gap in inspections than that proposed in the April 2022. Finally, 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.

The NRC staff reviewed the submitted information in the context of the characteristics of adequate performance monitoring. The NRC staff examined the proposed alternative as it relates to the three functions of performance monitoring noted above. The NRC 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 NRC staff notes that timely identification of novel or unexpected degradation calls for a sampling schema that can account for several potential sources of uncertainty and that these uncertainties can be expressed as parameter, model, and completeness uncertainties (discussed, for example, in NUREG-1855, Guidance on the Treatment of Uncertainties Associated with PRAs [Probabilistic Risk Assessments] in Risk-Informed Decisionmaking, dated March 31, 2017 (ML17062A466)).

The parameter and model uncertainties were considered by the NRC staff in the review of the analysis documented above (see Section 9.0 of this SE). The licensee stated in its 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 inspection period requested in the alternative proposed for Calvert Cliffs 1 and

2. Using the methodology described above, the NRC 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 understood, or not found) the NRC staff considered several pertinent considerations. The NRC staff noted that a component-specific degradation mechanism would likely have become apparent during the already performed inspections. Further, the NRC 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 inspection techniques, given 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 inspection techniques.

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 universal mechanism is active in similar materials in similar components and is detectable by the inspection technique. The NRC staff determined 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. As such, the NRC 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.

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.

CONCLUSION The NRC staff determined that the proposed alternative in the licensees request referenced above would provide an acceptable level of quality and safety.

The NRC staff concludes that the licensee has adequately addressed the regulatory requirements set forth in 10 CFR 50.55a(z)(1).

The NRC staff authorizes the use of proposed alternative RS-21-056 at Calvert Cliffs, Units 1 and 2 until the end of the current licensed periods, which is July 31, 2034, and August 13, 2036, respectively.

The NRCs authorization of the proposed alternative does not infer or imply the approval of EPRI report 15905 for generic use.

All other ASME BPV 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 and Michael Benson, NRR Date: January 3, 2023 Digitally signed by Hipolito J.

Hipolito J. Gonzalez Date: 2023.01.03 11:22:51 Gonzalez -05'00' Hipólito J. González, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

ML22195A025 OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/LA NRR/DNRL/NVIB/BC NAME MMahoney KEntz ABuford DATE 07/14/2022 07/21/2022 06/23/2022 OFFICE NRR/DORL/LPL1/BC NAME HGonzález DATE 01/03/2023