ML25220A279
| ML25220A279 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 08/25/2025 |
| From: | Ilka Berrios NRC/NRR/DORL/LPL3 |
| To: | Tony Brown Energy Harbor Nuclear Corp |
| Haeg, LE | |
| References | |
| EPID L-2024-LLR-0082 | |
| Download: ML25220A279 (1) | |
Text
August 25, 2025 Mr. Terry J. Brown Vistra Operations Company LLC c/o Davis-Besse Nuclear Power Station Mail Stop P-DB-3080 5501 N. State Route 2 Oak Harbor, OH 43449-9760
SUBJECT:
DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1 - PROPOSED RELIEF REQUEST RR-A1 RELATED TO PRESSURE RETAINING WELDS AND NOZZLE INNER RADII OF THE STEAM GENERATOR AND PRESSURIZER (EPID L-2024-LLR-0082)
Dear Mr. Brown:
By \
[[letter::L-24-233, Nuclear Power Station, Unit 1, Proposed Alternative Request to Defer ASME Code Section XI Inservice Inspection Examinations for Pressurizer and Steam Generator Pressure-Retaining Welds and Full Penetration Welded Nozzles|letter dated December 18, 2024]], Vistra Operations Company LLC (the licensee) requested relief from the inspection requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI for Davis-Besse Nuclear Power Station, Unit 1 (Davis-Besse). Pursuant to Title 10 of the Code of Federal Regulations (10 CFR),
Part 50, 50.55a(z)(1), the licensee requested U.S. Nuclear Regulatory Commission (NRC) review and approval of Alternative Request RR-A1 (hereafter referred to as RR-A1) to defer the inservice inspection (ISI) of the pressure-retaining welds and nozzle inner radii of the steam generators and pressurizer at Davis-Besse.
The NRC staff has reviewed the subject request and concludes, as set forth in the enclosed safety evaluation, RR-A1 provides an acceptable level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all the regulatory requirements set forth in 10 CFR 50.55a(z)(1). The NRC staff noted that the sixth 10-year inspection interval at Davis-Besse is currently scheduled to end on September 20, 2042; however, the current 60-year operating license expires on April 22, 2037. Therefore, the NRC staff authorizes the use of the Alternative Request RR-A1 at Davis-Besse for the remainder of the fifth 10-year ISI interval, through a portion of the sixth 10-year ISI interval, to the end of the 60-year operating license which expires on April 22, 2037.
All other ASME Code,Section XI requirements for which relief has not been specifically requested and approved in this alternative request remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.
T. Brown If you have any questions, please contact the Project Manager, Robert Kuntz, at 301-415-3733 or via email at Robert.Kuntz@nrc.gov.
Sincerely, Ilka Berrios, Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-346
Enclosure:
Safety Evaluation cc: Listserv ILKA BERRIOS Digitally signed by ILKA BERRIOS Date: 2025.08.25 14:57:09 -04'00'
Enclosure SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR RELIEF RR-A1 FOR THE FOURTH 10-YEAR INSPECTION INTERVAL VISTRA OPERATIONS COMPANY LLC DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1 DOCKET NO. 50-346 1.0 INTRODUCTION By \
[[letter::L-24-233, Nuclear Power Station, Unit 1, Proposed Alternative Request to Defer ASME Code Section XI Inservice Inspection Examinations for Pressurizer and Steam Generator Pressure-Retaining Welds and Full Penetration Welded Nozzles|letter dated December 18, 2024]], (Agencywide Documents Access and Management System Accession No. ML24353A315), Vistra Operations Company, Inc. (the licensee) requested relief from the inspection requirements of the American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel Code,Section XI for Davis-Besse Nuclear Power Station, Unit 1 (Davis-Besse). Pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Part 50, 50.55a(z)(1), the licensee requested U.S. Nuclear Regulatory Commission (NRC) review and approval of Alternative Request RR-A1 (hereafter referred to as RR-A1) to defer the inservice inspection (ISI) of the pressure-retaining welds and nozzle inner radii of the steam generators (SGs) and pressurizer at Davis-Besse.
2.0 REGULATORY EVALUATION
The pressure-retaining welds and nozzles of the SGs and pressurizer at Davis-Besse are ASME Code Class 1 and Class 2 components, whose ISIs are performed in accordance with the applicable edition of the ASME Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, as required by 10 CFR 50.55a(g).
The regulations in 10 CFR 50.55a(g)(4) state, in part, that components that are classified as ASME Code Class 1, 2, and 3 must meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, to the extent practical within the limitations of design, geometry, and materials of construction of the components.
The regulations in 10 CFR 50.55a(z) state, in part, that alternatives to the requirements in paragraphs (b) through (h) of 10 CFR 50.55a may be used when authorized by the NRC if the licensee demonstrates that: (1) the proposed alternative would provide an acceptable level of quality and safety, or (2) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
3.0 TECHNICAL EVALUATION
3.1 Licensees Relief Request 3.1.1 ASME Code Components Affected For the SGs, the affected components are Code Class 1 and Class 2 pressure-retaining shell welds and full penetration welded nozzles (nozzle-to-shell welds and nozzle inside radius sections).
For the pressurizer, the affected components are Code Class 1 pressurizer vessel head, shell-to-head, and nozzle-to-vessel welds.
The specific examination categories, Inspection Items, weld and nozzle identifications, and component descriptions are provided in Section 1 of RR-A1.
The NRC staff noted that the nuclear steam supply system at Davis-Besse is of the Babcock and Wilcox (B&W) design. Davis-Besse has two Once-Through SGs and one pressurizer that are of the B&W design.
3.1.2 ASME Code Requirements The fifth 10-year inservice inspection (ISI) interval Code of record for Davis-Besse is the 2017 Edition of the ASME Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components.
The ASME Code,Section XI, IWB-2500(a), Table IWB-2500-1, Examination Categories B-B and B-D, and IWC-2500(a), Table IWC-2500-1, Examination Categories C-A and C-B, require examination of the following Item Nos.:
B2.11 which states that volumetric examination of both circumferential shell-to-head welds during each inspection interval. The examination volume is shown in Figure IWB-2500-1.
B2.12 which states that volumetric examination of one foot of all longitudinal shell-to-head welds that intersect circumferential welds during the first interval and one foot of one longitudinal shell-to-head weld per head that intersects a circumferential weld during successive intervals. The examination volume is shown in Figure IWB-2500-2.
B2.40 which states that volumetric examination of essentially 100 percent of the weld length of all welds during the first Section XI inspection interval. For successive inspection intervals, the examination may be limited to one vessel among the group of vessels performing a similar function. The examination volume is shown in Figure IWB-2500-6.
B3.110 which states that volumetric examination of all full penetration nozzle-to-vessel welds during each inspection interval. The examination volume is shown in Figures IWB-2500-7(a), (b), (c), or (d).
C1.30 which states that volumetric examination of essentially 100 percent of the weld length of the tubesheet-to-shell welds during each Section XI inspection interval. In the case of multiple vessels of similar design, size, and service (such as SGs, heat exchangers), the required examinations may be limited to one vessel or distributed among the vessels. The examination volume is shown in Figure IWC-2500-2.
C2.21 which states that volumetric and surface examination of all nozzle welds at terminal ends of piping runs during each Section XI inspection interval. In the case of multiple vessels of similar design, size, and service (such as SGs, heat exchangers), the required examinations may be limited to one vessel or distributed among the vessels.
The examination area and volume are shown in Figures IWC-2500-4(a), (b), or (d).
C2.22 which states that volumetric examination of all nozzle inside radius sections at terminal ends of piping runs during each Section XI inspection interval. In the case of multiple vessels of similar design, size, and service (such as SGs, heat exchangers), the required examinations may be limited to one vessel or distributed among the vessels.
The examination volume is shown in Figures IWC-2500-4(a), (b), or (d).
3.1.3 Reason for Request The licensee stated that RR-A1 is based on the following EPRI topical reports:
Technical Bases for Inspection Requirements for PWR Steam Generator Feedwater and Main Steam Nozzle-to-Shell Welds and Inside Radius Sections, 3002014590 (ML19347B107) (EPRI Report 14590).
Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds, 3002015906 (ML20225A141) (EPRI Report 15906).
Technical Bases for Inspection Requirements for PWR Pressurizer Head, Shell-to-Head, and Nozzle-to-Vessel Welds, 3002015905 (ML21021A271) (EPRI Report 15905).
The licensee stated that EPRIs assessments include a survey of inspection results from 74 domestic and international nuclear units and flaw tolerance evaluations using probabilistic fracture mechanics (PFM) and deterministic fracture mechanics (DFM) as discussed in these three EPRI reports. The licensee further stated that these three EPRI reports conclude that the current ASME Code,Section XI ISI examinations can be deferred for some time with no impact to plant safety. Based on the conclusions of the three EPRI reports, supplemented by plant-specific evaluations, the licensee requested an ISI examination deferral for the SG and pressurizer components at Davis-Besse. The licensee stated that the three EPRI reports were developed consistent with the recommendations provided in EPRIs White Paper (ML19241A545) on suggested content for PFM submittals, NRC Regulatory Guide 1.245, Preparing Probabilistic Fracture Mechanics Submittals, (ML21334A158), and NUREG/CR-7278, Technical Basis for the use of Probabilistic Fracture Mechanics in Regulatory Applications, (ML22014A406) January 2022.
3.1.4 Proposed Alternative The licensee proposed to increase the inspection interval for the in-scope welds and nozzle inside radius section in the SGs and pressurizer, thereby deferring examination for two 10-year ISI intervals from the last examination performed for each item number. Section 5 of RR-A1 provides the proposed examination schedule for the required SG and pressurizer examinations.
The licensee stated that SG No. 1 is the selected vessel to comply with the examination requirements associated with Inspection Item Nos. B2.40, C1.30, C2.21, and C2.22. The licensee further stated that Steam Generator No. 2 equivalent weld examinations may be substituted to satisfy the requirements.
The NRC staff noted that for Examination Category B-B (Items No. B2.40), the footnote in the ASME Code,Section XI, Table 2500-1, states that [t]he examination may be limited to one vessel among the group of vessels performing a similar function For Examination Category C-A (Item No. C1.30) and Examination Category C-B (Item Nos. C2.21 and C2.22), the footnote in the ASME Code,Section XI, Table IWC-2500-1, states that [i]n the case of multiple vessels of similar design, size, and service (such as steam generators, heat exchangers), the required examinations may be limited to one vessel or distributed among the vessels During the first period of the fourth 10-year ISI interval in 2014, the licensee replaced both SGs.
The licensee inspected the new SG welds and nozzle inside radius sections prior to service (preservice inspection, PSI) followed by ISI examinations through the third period of the fourth inspection interval.
Davis-Besse is currently in the first period of the fifth ISI interval. The licensee requested the proposed alternative for the remainder of the fifth 10-year inspection interval and through the sixth 10-year inspection interval. The sixth 10-year inspection interval is currently scheduled to end on September 20, 2042, recognizing that the current 60-year operating license expires on April 22, 2037.
The licensee stated that the proposed alternative is to increase the inspection interval for the applicable SG and pressurizer components from the current ASME Code,Section XI 10-year requirement, thereby deferring examinations for two 10-year ISI intervals from the last examination performed for each component. The licensee further stated that the subject welds will be reexamined prior to the end of the current 60-year operating license for Davis-Besse.
3.2
NRC Staff Evaluation
3.2.1 Background and Evaluation Approach The licensees submittal contains six attachments. Attachment 1 is the proposed alternative RR-A1. Attachment 2 contains the technical basis showing that the Davis-Besse SGs meet the guidance of the EPRI Reports 14590 and 15906. Attachment 3 contains the technical basis showing that the Davis-Besse pressurizers meet the guidance of the EPRI Report 15905. contains the survey of inspection results of SGs and pressurizers of PWRs. contains the finite element analysis of the pressurizer at Oconee. Attachment 6 contains DFM and PFM analyses of the pressurizer at Oconee. All the attachments mentioned in this safety evaluation are associated with December 18, 2024, submittal unless specifically noted.
In RR-A1, the licensee used the PFM and DFM methodology of EPRI Reports 14590, 15905, and 15906 to perform PFM and DFM analyses for Davis-Besse SGs and pressurizer components. The industry has not submitted EPRI Reports 14590, 15905, and 15906 for NRC review and approval as topical reports. Therefore, the NRC did not review the EPRI reports for generic use, and this current review for Davis-Besse does not extend beyond the plant-specific authorization. However, the NRC staff has reviewed these three EPRI reports as part of the review and approval of alternative requests from other licensees as discussed below.
Steam Generator The PFM analysis in the EPRI Report 14590 was performed using the PRobabilistic OptiMization of InSpEction (PROMISE) Version 1.0 software. As part of the NRCs review of an alternative request submitted by the Southern Nuclear Company for Vogtle Electric Generating Plant (Vogtle), Units 1 and 2, the NRC performed an audit of the PROMISE Version 1.0 software as discussed in the NRCs audit plan dated May 14, 2020 (ML20128J311), and the NRC audit summary report dated December 10, 2020 (ML20258A002). The PFM analysis in the EPRI Report 15906 was performed using the PROMISE Version 2.0 software, which the NRC has not audited. The only technical difference between the two versions is that in PROMISE Version 1.0, the user-specified examination coverage is applied to all inspections, whereas in PROMISE Version 2.0, the examination coverage can be specified by the user uniquely for each inspection. In both Versions 1.0 and 2.0, the software assumes 100 percent coverage for the PSI examination.
The EPRI Report 14590 provides DFM and PFM analyses for SG, feedwater and main steam nozzle-to-shell welds and nozzle inside radius sections. The EPRI Report 15906 provides DFM and PFM analyses for nozzle-to-shell welds and other shell welds in the SG. The NRC staff has approved similar requests from Vogtle, Units 1 and 2, and Millstone Power Station (Millstone),
Unit 2, that were based on these two EPRI reports as documented in respective safety evaluations (ML20352A155 and ML21167A355, respectively).
For RR-A1, the NRC staff reviewed the plant-specific information of Davis-Besse SGs with respect to the relevant information in the EPRI reports to determine that the EPRI report methodology is applicable to Davis-Besse SGs. The NRC staff also evaluated the plant-specific ISI scenarios in Davis Besse PFM and DFM analyses.
Pressurizer The licensee for Davis-Besse performed DFM and PFM analyses of the pressurizer welds, specifically the head-to-shell, shell-to-head, and nozzle welds based on a bounding evaluation of the Oconee pressurizer, as documented in Attachments 5 and 6 of the submittal. These analyses rely on the technical methodology described in EPRI Report No. 15905, which provides both DFM and PFM evaluations for pressurizers in Westinghouse, Combustion Engineering (CE), and Babcock & Wilcox (B&W) designed plants.
The NRC staff previously reviewed and approved the technical approach and conclusions of EPRI Report 15905 in support of an alternative request submitted by Duke Energy for B&W-designed pressurizers. The staff's evaluation and acceptance of the report are documented in the NRC safety evaluation dated October 19, 2023 (ML23264A853), which concluded that the methodologies used in EPRI Report 15905 were technically sound and acceptable for demonstrating the structural integrity of pressurizer welds.
Steam Generator and Pressurizer Consistent with the key principles of the NRC risk-informed approach for performing reviews, the NRC staff evaluated the proposed performance monitoring of the SG and pressurizer components at Davis-Besse.
The NRC staff reviewed the following four major areas of Davis-Besse SGs and pressurizer analyses: degradation mechanisms, PFM analysis, DFM analysis, and performance monitoring as discussed below.
3.2.2 Degradation Mechanisms The licensee evaluated the degradation mechanisms including stress corrosion cracking (SCC),
environmental assisted fatigue (EAF), microbiologically influenced corrosion (MIC), pitting, crevice corrosion, erosion-cavitation, erosion, flow accelerated corrosion (FAC), general corrosion, galvanic corrosion, and mechanical/thermal fatigue as discussed in Section 5 of RR-A1. The licensee stated that other than the potential for EAF and mechanical/thermal fatigue, there were no active degradation mechanisms identified that significantly affect the long-term structural integrity of the SG and pressurizer welds and nozzle components.
The licensee stated that the materials and operating conditions for the Davis-Besse SGs and pressurizer welds and components considered in RR-A1 are similar to those in the three EPRI reports and therefore, the conclusions of the EPRI reports apply to the components covered in RR-A1. The licensee further stated that as discussed in the EPRI topical reports, an industry survey showed that no examinations identified any unknown degradation mechanisms (i.e.,
mechanisms other than those listed above). Based on this exhaustive industry survey, the licensee concluded that although the emergence of an unknown degradation mechanism cannot be completely ruled out, the possibility of the occurrence of such an unknown degradation mechanism is highly unlikely.
The NRC staff reviewed the plant-specific circumstances that may indicate the presence of a degradation mechanism and circumstances sufficiently unique to Davis-Besse to merit additional consideration. The NRC staff found that the degradation mechanisms described by the licensee for Davis-Besse are addressed in a manner sufficient for the applicability of the EPRI Reports and that no unknown degradation mechanisms were identified for the SGs and pressurizer at Davis-Besse.
3.2.3 PFM Analysis The PFM analysis is to demonstrate that the probability of failure (PoF) for a postulated flaw in the weld or nozzle inside radius of the SG or in the pressurizer welds, under applied loading, is low. The acceptance criterion for the PoF applicable to the SG or pressurizer components is 1x10-6 failures per year.
The NRC staff noted that the acceptance criterion of 1x10-6 failures per year 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 1x10-6 events 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 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 1x10-6 per year was generated using a conservative model for reactor vessel cracking. In addition, this criterion exists within the context of reactor pressure vessel surveillance programs and inspection programs. The NRC staff finds that Davis-Besses use of 1x10-6 failures per year for the acceptable PoF based on the reactor vessel TWCF criterion is acceptable for RR-A1 because (a) the impact of an SG or pressurizer weld and/or nozzle failure would be less than the impact of a reactor vessel failure on overall risk, (b) the subject SG and pressurizer welds and nozzles have substantive, relevant, and continuing inspection histories and programs, and (c) the estimated risks associated with the individual welds and nozzles 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).
Based on the discussion above, the NRC staff finds that the use of the acceptance criterion of 1x10-6 failures per year for PoF acceptable for the Davis-Besse SGs welds and nozzle inside radius and pressurizer welds covered under RR-A1.
Steam Generator The licensees basis for the proposed alternative relies upon the PFM analyses presented in the EPRI Reports 14590 and 15906. Section 5 of RR-A1 states that both EPRI reports perform finite element analyses (FEA) to determine the stresses in the SG welds and components covered in RR-A1. The EPRI analyses use representative SG geometries, bounding transients, and typical material properties. The results of the stress analyses were used in a flaw tolerance evaluation.
The NRC staff noted that the flaw tolerance evaluation consists of PFM evaluations and DFM evaluations to demonstrate the structural integrity of the SG. The results of the PFM analyses should indicate that the NRCs safety goal of 1x10-6 failures per year will be met.
The NRC staff confirmed that the licensees plant-specific analysis for Davis-Besse SG is consistent with the approach taken in the Vogtle and Millstone submittal. The original NRC evaluation of this approach is documented in the Vogtle and Millstone SEs (ML20352A155 and ML21167A355, respectively). The NRC reviewed the application of this approach, as proposed in the licensees request, and determined that the PFM analysis is consistent with the previously approved precedents in the Vogtle and Millstone submittals. Therefore, the NRC staff finds the proposed PFM analysis to be appropriate for the Davis-Besse submittal.
Pressurizer Section 5 of RR-A1 states that the geometric configuration of the pressurizer used in the EPRI Report 15905 stress analysis, while consistent with the Westinghouse/Combustion Engineering (CE) pressurizer designs, is not appropriate for the B&W pressurizer design at Davis-Besse.
The licensee stated that a plant-specific stress analysis was previously performed for the B&W pressurizers at Oconee Nuclear Station, Units 1, 2, and 3, and is documented in the Duke Energy fleet relief request for the pressurizer weld examination dated February 17, 2023 (ML23048A148). The NRC approved the Duke Energy fleet relief request in a safety evaluation (ML23264A853). Table 3 of RR-A1 provides a comparison of key plant-specific parameters for the Davis-Besse and Oconee pressurizers that demonstrates consistency of the B&W pressurizers at the two plants. The licensee concluded that the plant-specific stress analysis performed for Oconee in Attachment 5, together with additional information provided in, demonstrates that all plant-specific criteria in Attachments 5 and 6 are met for the Davis-Besse pressurizer. Based on the comparison of key parameters (e.g., material of construction and component geometries) between the Oconee pressurizer and Davis-Besse pressurizer and plant-specific loading and transients meeting the applicability criteria in EPRI Report 15905, the NRC staff determined that the analyses for the Oconee pressurizer in are applicable to the Davis-Besse pressurizer. The NRC staff also verified that PFM analysis in uses the acceptance criterion of 1x10-6 failures per year for PoF for the Davis-Besse pressurizer.
In the following subsections, the NRC staff reviewed the parameters and aspects most significant to the Davis-Besses PFM analysis: selection of components and materials, selection of transients, residual stresses, finite element analysis, fracture toughness, flaw density, fatigue crack growth rate, examination history and coverage, other considerations, and PFM results.
3.2.3.1 Selection of Components and Materials Steam Generator EPRI Reports 14590 and 15906 evaluate representative component geometries, materials, and loading conditions that were used in the PFM and DFM analyses. The EPRI reports also defined plant-specific applicability criteria related to component geometries, materials, and loading conditions that must be evaluated and met by each plant to determine the applicability of the EPRI reports. Section 4 of both EPRI Reports discusses the variation among SG shell designs.
EPRI used this information for finite element analyses to determine stresses in the analyzed SG welds. In selecting components, EPRI considered geometry, operating characteristics, materials, field experience with respect to service-induced cracking, and the availability and quality of component-specific information. presents the plant-specific applicability of the components and materials selected and analyzed in the EPRI Reports 14590 and 15906 to the Davis-Besse SG welds. The NRC staff reviewed Section 4 of both EPRI Reports and determined that the SG configurations selected in the EPRI reports for stress analysis are acceptable representatives for the corresponding Davis-Besse SG welds and nozzle components. The NRC staff verified that the radius-to-thickness (R/t) ratios of the Davis-Besse SG welds are bound by the stress multiplier used in the licensees plant-specific PFM analysis. To verify the dominance of the R/t ratio, the NRC staff reviewed the through wall stress distributions in Section 7 of the two EPRI reports to confirm that the pressure stress is dominant, which would confirm the dominance of the R/t ratio. Accordingly, the NRC staff finds that the EPRIs conclusion about the R/t ratio being the dominant parameter in evaluating the various component configurations to be acceptable for Davis-Besse SGs.
Section 9.4 of both EPRI reports addresses criteria for plant-specific applicability of the PFM analysis and indicates that materials are acceptable if they conform to the ASME Code,Section XI, Appendix G, paragraph G-2110. The NRC staff verified that the materials of Davis-Besse SG construction conform with the material property requirements of the ASME Code,Section XI, Appendix G, paragraph G-2110.
The NRC verified that the Davis-Besse SGs meet the plant-specific applicability criteria of EPRI Reports 14590 and 15906 and that the results and conclusions of the EPRI reports are applicable to Davis-Besse SGs. Based on the above, the NRC staff finds that Davis-Besses SGs meet the component geometry and materials applicability criteria in the EPRI Reports 14590 and 15906. Based on its evaluation, the NRC staff determined that the analyzed geometries and materials of the EPRI reports are applicable to and acceptable for the Davis-Besse SG welds.
Pressurizer Section 5 of RR-A1 states that the technical approach used in the stress analysis for the B&W pressurizer design in Attachment 5 is consistent with Section 7 of the EPRI Report 15905 and is based on the Oconee plant-specific geometry and operating conditions. Table 3 of RR-A1 compared the material specification and geometries of the pressurizer between Oconee and Davis-Besse. The NRC staff verified that the material and geometries of the Davis-Besse pressurizer are consistent with that of the Oconee pressurizer. The NRC staff determined that the analyzed geometries and materials in Attachment 5 are applicable to and acceptable for the Davis-Besse pressurizer welds.
3.2.3.2 Selection of Transients Steam Generator Section 5.2 of both EPRI Reports 14590 and 15906 discuss the thermal and pressure transients under normal and upset conditions which are considered relevant to the SG shell welds. The EPRI developed a list of transients for flaw analysis applicable to the SG shell welds based on transients that have the largest temperature and pressure variations. In Attachment 2, the licensee evaluated the plant specific applicability of the transients selected and analyzed in the two EPRI Reports to the Davis Besse SG welds. Tables 1-2, 1-3 and 1-4 of Attachment 2 presents the plant-specific transients associated with the Davis-Besse SGs in comparison with the transients from the EPRI Reports 14590 and 15906. The NRC staff confirmed that the Davis-Besse transient projections to 60 years are bound by the transients used in the two EPRI Reports. The NRC staff noted that the Davis-Besse projected 60-year cycles are substantially below the number of cycles assumed in the EPRI analysis. Therefore, the NRC staff finds that the transients assumed in the EPRI analysis bound the plant-specific transients for the Davis-Besse SGs.
The EPRI Reports 14509 and 15906 do not have separate test conditions included in the transient selection. Section 5 of RR-A1 states that pressure tests (i.e., system leakage tests) are performed at normal operating conditions and no hydrostatic testing has been performed since the plant began operation. The NRC staff noted that because the pressure tests are performed at normal operating conditions and are part of normal heatup/cooldown transients which have already been included in the flaw analysis, the system leakage test conditions need not be analyzed as a separate transient.
The NRC staff determined that transients in Section 5.2 of both EPRI Reports 14509 and 15906 are reasonable for the Davis-Besse SGs because the selection was based on large temperature and pressure variations that are conducive to FCG and expected to occur in PWRs. The NRC staff finds that Davis-Besse has met the transient applicability criteria in the two EPRI Reports.
Therefore, the NRC staff determined that the analyzed transients for the Davis-Besse SG welds and nozzle inside radius sections are acceptable and that the licensee sufficiently addressed applicable aspects of the transients for Davis-Besse SGs.
Pressurizer Tables 2-1 and 2-2 of Attachment 3 present applicable transients affecting the Davis-Besse pressurizer and demonstrate that the general transients and insurge/outsurge transients, respectively, at the Davis-Besse pressurizer are bounded by the transients used in the flaw evaluations in Attachments 5 and 6. The flaw evaluations in Attachments 5 and 6 are related to the Oconee pressurizer. The NRC staff verified that the Davis-Besse pressurizer transient cycles are much lower than those used in the flaw evaluations for Oconee pressurizer as shown in Attachments 5 and 6. In addition, the NRC staff noted that the Davis-Besse pressurizer transients as shown in Tables 2-1 and 2-2 are much lower than the transient used in the EPRI Report 15905. Therefore, the NRC staff determined that the pressurizer transients assumed in Attachments 5 and 6 appropriately bound the Davis-Besse pressurizer, and therefore, acceptable.
3.2.3.3 Residual Stresses Steam Generator The EPRI Reports 14590 and 15906 address weld residual stress and cladding stress. The NRC staff confirmed that no plant-specific aspects of this submittal beyond those evaluated in the Vogtle and Millstone SEs warranted additional consideration because of (1) the relatively low sensitivity of the weld residual stress on the PoF of SG as shown in Tables 8-17 through 8-20 of the EPRI Report 15906; and (2) the small impact of clad residual stress on the SG PoF.
Based on this determination, the NRC staff finds that there is a very low probability that plant-specific aspects of residual stress would have a significant effect on the SG probability of leakage or rupture beyond the studies documented in the EPRI Reports 14590 and 15906.
Pressurizer Section 5 of RR-A1 states that the weld residual stress from Figure 8-1 in the EPRI Report 15905 and the 30 ksi clad residual stress discussed in Section 8.2.2.4 of the EPRI Report 15905 were considered in the evaluation for B&W pressurizer welds as part of DFM analysis. The NRC staff verified that the flaw evaluation in Attachment 6 includes weld residual stress distribution which is consistent with the weld residual stress distribution in Figure 8-1 of the EPRI Report 15905. The NRC staff also verified that Section 8.2.2.4 of the EPRI Report 15905 states that clad stress at room temperature is the yield stress of 30 ksi, which is conservative for use across the thickness of the clad. Therefore, the NRC staff finds that the weld residual stress and clad residual stress are reasonable and acceptable to be considered in the flaw evaluation for the Oconee pressurizer and are applicable to the Davis-Besse pressurizer.
3.2.3.4 Finite Element Analysis Steam Generator The objective of the FEA is to derive stress in components of the SGs and pressurizer. The FEAs in EPRI Reports 14590 and 15906 were performed using representative SG geometries, bounding transients, and typical material properties. The resulting stresses from FEA were used in the flaw tolerance evaluation in the EPRI reports. The NRC staff verified in Attachment 2 that the key geometric parameters of Davis-Besse SGs are bound by the geometric parameter used in the finite element model of EPRI Reports 14590 and 15906. Additionally, in sections 3.2.3.1 and 3.2.3.2 of this SE, the NRC staff determined that the plant-specific applicability criteria in the EPRI Reports 14590 and 15906 regarding material and loading conditions were met. The NRC staff, therefore, determined that the FEAs in these two EPRI reports are applicable to the Davis-Besse SGs because all plant-specific requirements are bound within the two EPRI reports. The NRC staff determined that the applicability of the FEAs of the EPRI reports to the Davis-Besse SGs is demonstrated in Attachment 2 and confirmed that Davis-Besse SGs have met all plant-specific requirements specified by the EPRI reports.
Pressurizer Based on the results in the EPRI Report 15905, the pressurizer bottom head is controlling from a stress point of view due to the insurge/outsurge transients experienced in that region. Hence, performs the plant-specific stress analyses for the B&W pressurizer bottom head based on the Oconee pressurizer. Because of the relatively complicated geometry of the B&W pressurizer design as shown in Figure 1 of Attachment 5, the licensee chose thirteen critical stress paths to obtain stresses and compared them to two stress paths in the EPRI Report 15905. The NRC staff noted that the licensees thirteen stress paths provide more detailed information on the stress distribution in the pressurizer shell and welds and cover more areas of the pressurizer shell than the two stress paths in the EPRI Report 15905 and therefore are acceptable.
The NRC staff reviewed material and component selection, transient selection, and assumed residual stresses to determine whether the FEA in Attachment 5, which analyzed the Oconee pressurizer welds, is applicable to the Davis-Besse pressurizer welds. The NRC staff determined that the FEA of Attachment 5 is applicable to the Davis-Besse pressurizer welds because the geometries, bounding transients, and typical material properties of the Davis-Besse pressurizer are either consistent or bound by the Oconee pressurizer. Based on the above, the NRC staff finds that the FEA performed in Attachment 5 adequately represents the Davis-Besse pressurizer welds.
3.2.3.5 Fracture Toughness Steam Generator The licensee stated that the materials of the Davis-Besse SG welds and associated nozzles conform to the requirements of the ASME Code,Section XI, Paragraph G-2110. The EPRI Reports 14590 and 15906 assume for fracture toughness of ferritic materials an upper-shelf KIc (fracture toughness) value of 200 ksiin based on the KIc curve in the ASME Code,Section XI, A-4200. The A-4200 fracture toughness curve refers to the same fracture toughness curve in the ASME Code,Section XI, Paragraph G-2110. The NRC staff verified that the Davis-Besse SG materials conformed to the requirements of the ASME Code,Section XI, Paragraph G-2110.
The NRC staff determined that the fracture toughness of the Davis-Besse SG components is conformed to the requirements of the ASME Code,Section XI, Paragraph G-2110 and is within the bounds of the fracture toughness values used in the EPRI Reports.
Pressurizer Table 3 of RR-A1 shows that the Davis-Besse pressurizer is made of SA-516, Grade 70 (shell or head) and A 508, Class 1 (nozzle). Attachment 6 states that the pressurizer materials under consideration are low carbon ferritic steels (SA-516 Grade 70 and SA-212). The fracture toughness provided in the ASME Code,Section XI, Appendix A only applies to low alloy ferritic steels such as SA-533 Grade B Class 1, SA-508 Class 2, and SA-508 Class 3 (typically used in the fabrication of Westinghouse and CE pressurizers) and is therefore not applicable to the B&W pressurizers. Attachment 6 states that the fracture toughness for low carbon ferritic steel piping components is provided in the ASME Code,Section XI, Appendix C. The licensee recognized that Appendix C is for the flaw evaluation of piping, not a pressurizer. However, the licensee used lower bound values of fracture toughness (106 ksiin as compared to 200 ksiin) for ferritic piping that are provided in the ASME Code,Section XI, Appendix C, for postulated circumferential and axial flaws. The NRC staff finds that use of a lower bound upper shelf fracture toughness of 106 ksiin for the pressurizer weld material is conservative and, thus, acceptable. The licensee also performed sensitivity studies that demonstrated that the PoF of 1 x 10-6 per year did not change using a fracture toughness from 72 ksiin to 106 ksiin. The NRC staff determined that use of 106 ksiin for fracture toughness to analyze pressurizer welds is acceptable because it is lower than and thus conservative as compared to the use of the fracture toughness value of 200 ksiin. Therefore, the NRC staff finds that use of 106 ksiin for fracture toughness is applicable to the Davis-Besse pressurizer.
3.2.3.6 Flaw Density Steam Generator Section 5 of RR-A1 states that Section 8.2.2.2 of the EPRI Report 14590 and Section 8.3.2.2 of the EPRI Report 15906 assume a flaw density of 0.001 flaws per nozzle for the SG nozzle inside radius sections. The NRC staff performed a detailed evaluation of the flaw density used in the PFM analysis as discussed in its previous safety evaluations for Vogtle and Millstone submittals. In its Vogtle safety evaluation, the NRC staff indicated that a nozzle flaw density of 0.1 flaws per nozzle should have been used. Sensitivity studies performed in Section 8.2.4.3.4 of the EPRI Report 15906 indicated that by changing the flaw density (number of flaws) in the nozzle inside radius sections from 0.001 to 0.1, the probabilities of leak and rupture increased by two orders of magnitude but were still significantly below the acceptance criterion of 1.0x10-6 per year.
Section 8.3.2.2 of the EPRI Report 15906 stated that a flaw density of 1.0 flaw per weld was used in the SG weld. The NRC staff noted that, so long as the component materials applicability criteria are met, the use of a flaw density of 0.1 flaws for the SG nozzle inside radius and a flaw density of 1.0 for the SG weld is sufficient for the PFM analysis as stated in the NRCs safety evaluation for the Millstone submittal.
As stated in the NRCs safety evaluation for Millstone, the NRC staff previously determined that the technical basis for evaluating flaw density at the nozzle inner radius and nozzle-to-vessel shell welds was acceptable. Specifically, the NRC staff concluded that an acceptable flaw density of 0.1 flaw per nozzle at the inner radius is supported by the evaluation documented in the NRCs safety evaluation dated December 19, 2007 (ML073600374) for Topical 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). In that same review, the staff also determined that a flaw density of 1.0 flaw per weld is a conservative assumption for the nozzle weld configurations analyzed in BWRVIP-108NP-A.
This same technical basis remains applicable to the Davis-Besse evaluation. The NRC staffs conclusions regarding flaw density and acceptability of the BWRVIP-108NP-A methodology, as documented in the Millstone safety evaluation, continue to provide the basis for the staffs acceptance for RR-A1.
Pressurizer Tables 6 to 10 of Attachment 6 show that the flaw density is assumed to be 1.0 in the PFM analysis for the pressurizer weld with various stress multiplier of 1.0 to 1.4. The NRC staff noted that the use of a flaw density of 1.0 for the Davis-Besse pressurizer weld is consistent with the flaw density of 1.0 for the SG weld that the NRC staff has found acceptable in the NRCs safety evaluation for the Duke Energy submittal (ML23264A853) and in the EPRI Report 15905. Based on the above, the NRC staff finds that the flaw density assumed in the PFM analysis of the is appropriate and applicable to the Davis-Besse pressurizer.
3.2.3.7 Fatigue Crack Growth Rate Steam Generator and Pressurizer The NRC staff noted that the fatigue crack growth (FCG) rate used in the three EPRI Reports 14590, 15905, and 15906 and in Attachment 6 is based on the FCG rate of the ASME Code,Section XI, A-4300. The NRC staff noted that FCG rate depends on component material and environmental conditions. Per the code of record for the current fifth ISI interval at Davis-Besse, the FCG rate of the 2017 Edition of the ASME Code,Section XI, A-4300 may be used for low alloy ferritic steels in air and reactor water environments. The NRC staff determined that the Davis-Besse SG and pressurizer materials meet the provisions of A-4300 and may use the FCG rate of A-4300. Therefore, the NRC staff determined that the FCG rate of the ASME Code,Section XI is appropriate for Davis-Besse SGs and pressurizer.
3.2.3.8 Examination History and Examination Coverage Steam Generator The examination history of the Davis-Besse SG welds and nozzle inner radii is in Table 1-5 of. Table 1-5 indicates that no reportable indications were found during the examinations. As shown in Table 1-5, the NRC staff noted that the examinations of Davis-Besse SG welds and nozzle inside radii have satisfied the ASME Code,Section XI, requirement of 90 percent or greater examination coverage of the required volume.
The licensee replaced SGs at Davis-Besse in 2014 and performed associated PSI examinations. Subsequently, the licensee performed ISI examinations over one complete 10-year interval following the SG replacement. Section 5 of RR-A1 states that the licensee considered PSI plus one set of 10-year ISI examinations to be followed by two 30-year ISI deferrals (PSI+10+40+70) as part of the PFM analysis for Davis-Besse SGs. The NRC staff noted that the licensee proposed to defer SG examination for 20 years even though it analyzed a deferred examination of 30 years. In terms of the PFM results, the 30-year interval analyzed is more conservative than the proposed 20-year inspection interval because there is more time for postulated flaws modeled in the analysis to grow. Therefore, the NRC staff finds that the licensee sufficiently accounted for plant specific examination history in the Davis-Besse PFM analysis. Based on the above discussion, the NRC staff finds that the ISI scenarios in the PFM analysis of the Davis-Besse SG welds and nozzle inside radii to be acceptable because the scenarios are adequately addressed plant-specific examination history and bound the proposed inspection interval.
Pressurizer The NRC staff reviewed the examination history of the Davis-Besse pressurizer in Table 2-3 of which indicates that no reportable indications were found during these examinations. The licensee stated in Section 5 of RR-A1 that the Davis-Besse pressurizer received PSI examinations followed by ISI examinations over four completed 10-year intervals.
This ISI history is the same as that analyzed in the PFM analysis, as indicated in Table 6 of. Therefore, the NRC staff finds that the examination history of the Davis-Besse pressurizer welds are appropriately analyzed as the same ISI scenario modeled in the PFM analysis.
The NRC staff noted that Table 2-3 of Attachment 3 indicates that the examination coverages of some of the Davis Besse pressurizer welds did not meet the ASME Code,Section XI, examination coverage requirement of 90 percent or greater. Table 2-3 shows that the lowest examination coverage for the pressurizer welds was 52.48 percent. The NRC staff evaluated this examination coverage in section 3.2.3.10 of this SE.
3.2.3.9 Other Considerations The PFM analysis of the EPRI Reports 14590, 15905, and 15906 also involves other parameters such as initial flaw depth and length distribution, probability of detection, models, uncertainty, and convergence. The NRC staff noted that these considerations of the PFM analyses in the three EPRI Reports do not depend on plant-specific information from Davis-Besse SG and pressurizer components.
The NRC staff noted that initial flaw depth and length distribution do not depend on plant-specific information because the flaw distribution used was based on fabrication flaws instead of service induced flaws. Probability of detection, which is associated with volumetric examinations, does not depend on plant-specific information because the corresponding welds in SGs and pressurizer in different plants are subject to the same volumetric examination requirements of the ASME Code,Section XI. The models (e.g., the stress intensity factor models) used do not depend on plant-specific information because they are widely used models in PFM analyses. Uncertainty and convergence do not depend on plant-specific information because these are part of the overall PFM analyses that were addressed in the sensitivity studies and sensitivity analyses in the three EPRI Reports. Because these considerations are not dependent on plant-specific information, the NRC staff finds that the Davis-Besse submittal is acceptable in terms of these considerations.
3.2.3.10 PFM Results Steam Generator The licensee performed a plant-specific PFM analysis for three critical components in the Davis-Besse SGs. The limiting component analyzed for Inspection Item Nos. C2.21 and C2.22 in the EPRI Report 14590 is the feedwater nozzle. However, the licensee stated that Davis-Besse SGs do not have Item No. C2.21 and C2.22 feedwater nozzle components because of the B&W design. Instead, the licensee analyzed the main steam nozzle inside radius section for the Davis-Besse SGs. From the EPRI Report 14590, the critical Case Identification for the main steam nozzle inside radius section is SGB-P1N. The licensee performed a PFM analysis for the Davis-Besse SGs similar to that shown in Table 8-28 of the EPRI Report 14590 for the main steam nozzle assuming a nozzle flaw density of 0.1, a fracture toughness of 200 ksiin and a standard deviation 5 ksiin. The licensee used a relatively high-stress multiplier of 2.35, which represents the highest stress multiplier that can be applied without exceeding the acceptance criterion of 1.0 x10-6. The licensee analyzed the main steam nozzle inside radius section up to 80 years of plant operation using the PSI/ISI inspection scenario of PSI+10+40+70. The results of the PFM evaluation demonstrate that the PoF of the main steam nozzle inside radius section of the Davis-Besse SGs is less than 1.0x10-6 failures per year as shown in Table 4 of RR-A1.
The second critical component evaluated in the licensees PFM analysis was the main steam nozzle-to-shell weld of the Davis-Besse SGs. Table 5 of RR-A1 shows that the licensee calculated the PoF of the main steam nozzle-to-shell weld to be less than 1.0x10-6 failures per year. The third critical component that the licensee calculated was for the remaining SG welds in the Davis-Besse SGs. Table 6 of RR-A1 shows that that the PoF of the remaining SG welds is also less than 1.0x10-6 failures per year.
The NRC staff determined that the licensees PFM analysis of the Davis-Besse SGs demonstrates (1) that the proposed inspection interval (evaluated in Section 3.2.5 of this SE) will maintain the structural integrity of the Davis-Besse SG welds and SG nozzles, and (2) the PoF will satisfy the acceptance criterion of 1x10-6 failure per year to the end of the 60-year operating license.
Pressurizer The PFM evaluation of the Davis-Besse pressurizer is based on the PFM evaluation that was performed using the pressurizer at Oconee Units 1, 2, and 3, as shown in Attachment 6.
Applying the PFM analysis of the Oconee pressurizer to the Davis-Besse pressurizer, the NRC staff noted that:
the Oconee PFM analysis used higher number of transient cycles than the Davis-Bessel pressurizer would experience. This indicates that Oconee pressurizer analysis provides conservative results than if the actual Davis-Besse pressurizer transient cycles were used, Oconee PFM analysis did a sensitivity analysis using a range of lower (thus conservative) bound fracture toughness from 72 ksiin to 106 ksiin that are lower than the 200 ksiin that is normally used in the PFM analysis for vessel materials, The Oconee analysis did sensitivity studies by increasing the stresses by 10 percent and by a factor of 1.4, and the parameters in the Oconees PFM analysis are either applicable to or bound the Davis-Besse pressurizer.
The licensee also performed a sensitivity study using an examination coverage of 25.2 percent of pressurizer weld volume as shown in Attachment 6. The NRC staff noted that Table 10 of shows that with a low examination coverage of 25.2 percent, the PoF is still lower than the acceptance criterion of 1.0 x 10-6 failures per year. Therefore, the NRC staff determined that calculating the PoF using the lowest examination coverage of 52.48 percent as an input would still result in a PoF that is within the acceptance criterion of 1 x 10-6 failures per year.
The NRC staff noted that the Oconee PFM evaluation results show that the probabilities of rupture and leakage at Oconee pressurizer are below the acceptance criterion of 1.0 x10-6 failures per year after 80 years of operation. The NRC staff determined that because Oconees PFM evaluation is applicable to the Davis-Besse pressurizer, Oconees PFM analysis results are applicable to Davis-Besses pressurizer. In addition, the NRC staff has approved a similar alternative request for the deferral examination of pressurizer welds, which were based on the EPRI Report 15905, for the Duke Energy plants as documented in the NRCs safety evaluation (ML23264A853). Therefore, the probabilities of rupture and leakage for the Davis-Besse pressurizer are below the acceptance criterion of 1.0 x10-6 failures per year to the end of the 60-year operating license.
3.2.4 Deterministic Fracture Mechanics Analysis Steam Generator Section 5 of RR-A1 states that the DFM evaluations in Table 8-31 of the EPRI Report 14590 and Table 8-3 of EPRI Report 15906 verify the above PFM results for Davis-Besse SGs by demonstrating that it takes significantly more than 80 years for a postulated flaw with an initial depth equal to the ASME Code,Section XI, acceptance standards to grow to an unacceptable depth where the maximum stress intensity factor exceeds the ASME Code,Section XI allowable fracture toughness.
The NRC staff noted that in Section 8.3 of the EPRI Report 14590 and Section 8.2 of the EPRI Report 15906, EPRI performed a DFM analysis using an initial flaw depth of approximately 5.2 percent of the component thickness and average values of all other parameters considered in the PFM analysis. The initial flaw depth is obtained from the ASME Code,Section XI, Tables IWB-3510-1 and IWC-3510-1. The DFM analysis in the two EPRI reports indicates that all analyzed locations in the SG welds and nozzles resulted in more than 80 years before reaching leakage, where leakage is defined as the point at which the flaw depth has reached 80 percent of the component thickness. The DFM analyses also indicate that no locations of the SGs reached an applied stress intensity factor of greater than 200 ksiin which indicates that the SG welds and nozzle will not fail. The NRC staff determined that the DFM analyses in both EPRI reports are applicable to the Davis-Besse SGs and nozzle components because the parameters for the DFM analyses are either the same or similar to that of the PFM analyses. As discussed above, PFM analyses in the EPRI reports are applicable to the Davis-Besse SGs and nozzle components. In addition, the NRC staff approved the PFM methodology when it reviewed a similar submittal from Vogtle, Units 1 and 2, and Millstone, Unit 2, that were based on these two EPRI reports as documented in safety evaluations. Thus, the NRC staff determined that the DFM analysis performed in both EPRI reports demonstrates that the structural integrity of the Davis-Besse SGs will be maintained to the end of 60-year operating license.
Pressurizer Section 5 of RR-A1 states that the technical approach used in the DFM evaluation for the B&W pressurizer in Attachment 6 is consistent with Section 8.2 in EPRI Report 15905. The design inputs used in the DFM evaluation are summarized in Table 1 of Attachment 6. The licensee assumed an initial flaw size of 5.2 percent of the wall thickness, equivalent to the most conservative ASME Code,Section XI acceptance standard for these components. The licensee used the FCG rate of the ASME Code,Section XI, Appendix A, Paragraph A-4300 in the DFM analysis using the through-wall stress distributions from the stress analyses in Attachment 5.
The licensee used the fracture mechanics models in Section 8.2.2.4 of the EPRI Report 15905 to determine the length of time for the postulated initial flaw to grow to a depth of 80 percent of the wall thickness (assumed to equate to leakage) or the depth at which the allowable fracture toughness was reached, whichever time period was less. The allowable fracture toughness used in the DFM analysis is the upper shelf value of KIC equal to 106 ksiinch reduced by structural factors of 2.0 for primary stresses and 1.0 for secondary stresses, which are consistent with ASME Code,Section XI, Appendix G.
As shown in Table 4 of Attachment 6, the DFM analysis results show that the period required for hypothetical postulated flaws to leak are very long (in excess of 200 years). As Attachment 6 is applicable to the Davis-Besse pressurizer, DFM analysis results indicate that the Davis-Besse pressurizer welds are flaw tolerant. Thus, the NRC staff determined that the DFM analysis supports the PFM analysis and demonstrates that the structural integrity of the Davis-Besse pressurizer welds will be maintained to the end of 60-year operating license.
3.2.5 Performance Monitoring The NRC staff noted that performance monitoring, such as ISI programs, is a necessary component described by the NRC five principles of risk-informed decision making. Analyses, such as PFM, combine with performance monitoring to provide a mutually supporting and diverse basis for facility condition and maintenance that is within its licensing basis. The NRC staff determined that an adequate performance monitoring program must provide direct evidence of the presence and extent of degradation, validation of continued appropriateness of associated analyses, and a timely method to detect novel/unexpected degradation. The NRC staff described these characteristics at various public meetings (ML22060A277, ML23033A667, and ML23114A034).
The licensee proposed alternative examination consists of performing the subject SGs and pressurizer examinations once every other ISI interval, in lieu of the ASME Code,Section XI requirement of once every ISI interval. The NRC staff noted that the proposed alternative examination in RR-A1 is equivalent to inspecting a 50 percent sample of full component SGs and pressurizers required to be inspected by the ASME, Code Section XI in the fifth and sixth ISI intervals. The NRC staff determined that a 50 percent sample population is adequate in terms of binomial statistics to monitor the conditions of the SGs and pressurizer. Therefore, the NRC staff finds that the proposed alternative provides sufficient performance monitoring for the in-scope components of SGs and pressurizer at Davis-Besse.
Section 5 of RR-A1 states that if during the examination schedule proposed in RR-A1, indications are detected that exceed the applicable ASME Code,Section XI, acceptance standards of IWB-3500 or IWC-3500, the licensee will address detected indications as required by the ASME Code,Section XI. The licensee will apply the additional examination and successive inspection requirements of the ASME Code,Section XI. The licensee stated that the number of additional examinations shall be the number required by the ASME Code,Section XI, IWB-2430/IWC-2430. In addition to the ASME Code,Section XI, Davis-Besse uses the Corrective Action Program to review and evaluate industry Operating Experience (OE) to determine the appropriate actions required based upon the specific OE. If the OE indicates that a new or novel degradation mechanism may exist in SG and pressurizer welds or components, appropriate examinations will be performed to ensure that no such mechanism is occurring at Davis-Besse.
Based on its evaluation, the NRC staff finds that the licensees proposed future inspections and the scope expansion plans provide an adequate level of performance monitoring for the subject SG and pressurizer components because a 50 percent sample of full component SGs and pressurizers required to be inspected by the ASME Code,Section XI will be inspected during the proposed alternative period (i.e., during the fifth and sixth ISI intervals).
4.0 CONCLUSION
As set forth above, the NRC staff determined that RR-A1 provides an acceptable level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all the regulatory requirements set forth in 10 CFR 50.55a(z)(1). The NRC staff noted that the sixth 10-year inspection interval at Davis-Besse is currently scheduled to end on September 20, 2042; however, the current 60-year operating license expires on April 22, 2037. Therefore, the NRC staff authorizes the use of the Alternative Request RR-A1 at Davis-Besse for the remainder of the fifth 10-year ISI interval, through a portion of the sixth 10-year ISI interval, to the end of the 60-year operating license which expires on April 22, 2037.
All other ASME Code,Section XI requirements for which relief has not been specifically requested and approved in this alternative request remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.
Principal Contributors: John Tsao, NRR Cory Parker, NRR David Dijamco, NRR Date: August 22, 2025
ML25220A279 OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DNRL/NVIB/BC NAME RKuntz SLent (KEntz for)
ABuford DATE 8/8/2025 8/11/2025 7/28/2024 OFFICE NRR/DORL/LPL3/BC NAME IBerrios DATE 8/25/2025