ML24036A007
| ML24036A007 | |
| Person / Time | |
|---|---|
| Site: | 99902016 |
| Issue date: | 02/12/2024 |
| From: | Licensing Processes Branch |
| To: | Constellation Energy Generation, Electric Power Research Institute |
| References | |
| BWRVIP-100, BWRVIP-100, Rev 2, EPID L-2023-TOP-0018, EPID L-2023-NTR-0005 | |
| Download: ML24036A007 (1) | |
Text
OFFICIAL USE ONLY - PROPRIETARY INFORMATION Enclosure 2 OFFICIAL USE ONLY - PROPRIETARY INFORMATION REQUEST FOR ADDITIONAL INFORMATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REVIEW OF BWRVIP-100, REVISION 2, UPDATED ASSESSMENT OF THE FRACTURE TOUGHNESS OF IRRADIATED STAINLESS STEEL BOILING WATER REACTOR (BWR)
INTERNAL COMPONENTS BOILING WATER REACTOR VESSEL AND INTERNALS PROJECT (BWRVIP)
ELECTRIC POWER RESEARCH INSTITUTE (EPRI)
DOCKET NO. 99902016 (EPID L-2023-NTR-0005)
Division of New and Renewed Licenses (DNRL)/Vessels Internal Branch (NVIB) 1.0
Background
By letter dated July 11, 2023 (Agencywide Documents Access Management System (ADAMS)
Accession No. ML23198A234), the EPRI submitted licensing topical report BWRVIP-100, Revision 2, "Updated Assessment of the Fracture Toughness of Irradiated Stainless Steel for BWR Internals Components" for U.S. Nuclear Regulatory Commission (NRC) review. The revised topical report proposes new fracture toughness correlations for irradiated stainless steel weld and base materials, based upon data collected since publication of BWRVIP-100, Revision 1-A. This toughness and strength correlations in this topical report provide important inputs to flaw evaluation procedures that underpin licensee aging management programs for the vessel internals.
2.0 Regulatory Basis The regulations in Title 10 of the Code of Federal Regulations, Part 54 (10 CFR Part 54) address the requirements for nuclear power plant license renewal. 10 CFR 54.21, Contents of application-technical information, requires that each application for a renewed operating license contain an integrated plant assessment (IPA) and an evaluation of time limited aging analyses. As stated in 10 CFR 54.21(a), the IPA shall identify and list those structures and components subject to an aging management review (AMR) and demonstrate that the effects of aging (cracking, loss of material, loss of fracture toughness, dimensional changes, loss of preload) will be adequately managed so that their intended functions will be maintained consistent with the current licensing basis for the period of extended operation as required by 10 CFR 54.29(a). In addition, 10 CFR 54.22 requires that applications for renewed operating licenses include any technical specification changes or additions necessary to manage the effects of aging during the period of extended operation as part of the application.
Structures and components subject to an aging management program (AMP) shall encompass those structures and components that: (1) perform an intended function, as described in 10 CFR 54.4, Scope, without moving parts or without a change in configuration or properties and (2) are not subject to replacement based on a qualified life or specified time period. These structures and components are referred to as passive and long-lived structures and components, respectively. The scope of components considered for inspection under a BWRVIP AMP meet these criteria.
The BWRVIP RVI AMP may be informed by the 10 program elements described in NUREG-2192, Standard Review Plan for Subsequent License Renewal Applications (SRP-SLR) (ADAMS Accession No. ML17188A158), and NUREG-2191, Generic Aging Lessons Learned Report for Subsequent License Renewal (GALL-SLR) (ADAMS Accession
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Nos. ML17187A031 and ML17187A204). GALL-SLR AMP XI.M9, BWR Vessel Internals, includes various BWRVIP reports that address aging management strategies for RVI components that serve an intended safety function pursuant to criteria in 10 CFR 54.4(a)(1).
The scope of the program does not include consumable items such as fuel assemblies, reactivity control assemblies, and nuclear instrumentation because these components are not typically within the scope of the components that are required to be subject to an AMP, as defined by the criteria set out in 10 CFR 54.21(a)(1).
The regulations in 10 CFR 50.55a(g)(4) state, in part, components that are classified as American Society of Mechanical Engineers (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 Boiler and Pressure Vessel Code,Section XI, to the extent practical within the limitations of design, geometry, and materials of construction of the components. The examination requirements for vessel internals of Section XI are specified in IWB-2500, Table IWB-2500-1, Examination Categories B-N-2 and B-N-3. These requirements specify that visual examinations must be performed on core support structures and interior attachments to the reactor vessel each inspection interval.
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.
Licensees have requested through 10 CFR 50.55a(z) to implement BWRVIP guidance for examination of RVI as an alternative to the examination requirements of Section XI (e.g., see example alternative request at ADAMS Accession No. ML17305B279). Therefore, BWRVIP topical reports may impact licensees inservice inspection programs.
The staff has determined that it requires additional information to complete its review of BWRVIP-100, Revision 2.
3.0 Request for Additional Information RAI #1 In Section 1.6 of BWRVIP-100, Revision 2, the BWRVIP stated that the new flaw evaluation guidance overrides all existing flaw evaluation guidance. The existing flaw evaluation guidance for reactor vessel internals resides in multiple BWRVIP topical reports that span a broad range of components. Given the complications associated with existing component-specific flaw evaluation guidance, explain how licensees will incorporate the new guidance of BWRVIP-100, Revision 2 in aging management programs and inservice inspection programs.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION RAI #2 In Section 1.4, the BWRVIP states that the base metal correlations should only be used when the flaw is sufficiently far from the weld. Clarify whether specific guidance will be developed to aid licensees in determining whether the base metal or weld metal correlations should be used, where the specific guidance is, and what the specific guidance is.
RAI #3 In Section 1.3.2, while describing results of a sample problem, the BWRVIP states that the total stress that produces unstable crack extension is 66.7 MPa and that the nominal stress multiplied by the margin is 66.7 MPa, implying a margin of 1.0 between the nominal stress and the critical stress. In Section 5.3, while describing the results of another sample problem, the BWRVIP stated that ((
]. The staffs interpretation of the flaw evaluation procedure in BWRVIP-100, Revision 2 is that licensees should calculate the margin and then compare that value to a criterion. The criterion may be found in component-specific BWRVIP topical reports. However, BWRVIP-100, Revision 2, does not actually specify what constitutes an acceptable margin. Confirm whether staffs interpretation of the flaw evaluation procedure is correct. Clarify whether there is specific guidance to instruct the user more explicitly on determining the appropriate margin for flaw evaluation, where the specific guidance is, and what the specific guidance is.
RAI #4 In Section B.1 of BWRVIP-100, Revision 2, the BWRVIP presents an example elastic-plastic fracture mechanics analysis. However, the BWRVIP used a K solution that is not valid for the R/t of interest, leading to extrapolation of the A parameter. However, K solutions exist for large R/t, so this extrapolation is not necessary. Illustrative examples should demonstrate good engineering practice. Explain why a more appropriate K solution was not applied for the example problem described in Section B.1.
RAI #5 BWRVIP-100, Revision 1-A provides a fracture toughness recommendation for brittle fracture conditions (i.e., KIc = 50 ksiin). BWRVIP-100, Revision 2, does not address KIc for the base metal and weld metal. Justifying why Revision 2 does not such guidance or direct the NRC staff to the guidance.
RAI #6 The n-fluence correlation for the weld metal represents a conservative lower bound of the data, while the n-fluence correlation for the base metal is not a lower bound of the data. The BWRVIP stated that the correlation for the base metal was developed to ensure that predicted J-T curves were a conservative or close match to the experimental J-T curves. (a) Explain why the two correlations were not developed consistently. (b) Explain what criteria the BWRVIP employed to ensure that the predicted J-T curves were appropriately conservative for use in fitness-for-service determinations.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION RAI #7 The BWRVIP proposes to maintain the neutron fluence threshold of ((
)) for irradiated stainless steel base metal, based upon the fracture toughness results shown in Tables 3-1 through 3-4. However, in Table 3-1, specimen 12, CT 304 HAZ (3rd from the bottom) exhibited brittle fracture at a fluence of ((
)), suggesting uncertainty in the threshold value. (a) Provide justification for disregarding this data point when determining the fluence threshold. (b) Given that the BWRVIP rejected a data point to maintain ((
)) as a threshold, provide justification for why it is appropriate to have one threshold for both base metal and weld metal. Otherwise, provide additional justification for why the threshold is different for the two materials.
RAI #8 In Sections 3.1.2 and 3.1.3, the BWRVIP describes the correlations for base metal C and n. The BWRVIP excluded a data point at ((
)), stating that the data point lies outside of the apparent trend of the rest of the data. However, a large gap in the data exists between unirradiated data and ((
)), such that it is difficult to judge what the real trend is based upon the data set. Justify neglecting this data point when developing the C and n correlations for the base metal.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Division of Safety Systems/Nuclear Methods and Fuel Analysis Branch (SFNB) 1.0
Background
By letter dated July 11, 2023 (ADAMS Accession No. ML23198A234), the EPRI submitted licensing topical report BWRVIP-100, Revision 2, "Updated Assessment of the Fracture Toughness of Irradiated Stainless Steel for BWR Internals Components" for U.S. NRC review.
The revised topical report proposes new fracture toughness correlations for irradiated stainless steel weld and base materials, based upon data collected since publication of BWRVIP-100, Revision 1-A. This toughness and strength correlations in this topical report provide important inputs to flaw evaluation procedures that underpin licensee aging management programs for the vessel internals.
2.0 Regulatory Basis Regulatory Guide (RG) RG 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence (ADAMS Accession No. ML010890301), provides guidance on methods for calculating pressure vessel neutron fluence, including qualifying the methods and estimating the calculational uncertainty, which are acceptable to the NRC staff based on General Design Criteria (GDC) 14, 30, and 31 contained in Appendix A of Title 10 of the Code of Federal Regulations (10 CFR) Part 50. GDC 14, 30, and 31 contain requirements regarding the design, fabrication, erection, and testing of the reactor coolant boundary. At the time RG 1.190 was developed, fluence methods were used to estimate fluence in a beltline region that axially extended along the vessel from the bottom of the active fuel to the top of the active fuel.
Outside of the traditional beltline, in areas such as the reactor vessel internals, which is the subject of this topical report, the methods that RG 1.190 recommends for beltline fluence estimates may require refinement, because the geometry becomes more complex, the neutron transport distance is longer, and the neutron energy spectrum changes.
3.0 Request for Additional Information RAI #9 BWRVIP-100 Revision 2 topical report provides newly acquired data through testing of irradiated materials harvested from the decommissioned José Cabrera (also known as Zorita) and Barsebck nuclear power plants. These data are a function of neutron fluence.
BWRVIP-100, Revision 2, does not provide any details as to how the neutron fluence was calculated for the materials at either plant. As a result, the staff is unable to assess the accuracy and associated uncertainty of the fluence estimates and therefore cannot determine the acceptability of the fluence estimates to the Zorita and Barsebck materials provided in the topical report.
For the Zorita data, the topical report states that testing of the material removed from Zorita was performed at Argonne National Laboratory (ANL) as presented in ANL 19/45 (ADAMS Accession No. ML20198M503). ANL 19/45 states that the fluence estimates for the Zorita materials was detailed in EPRI report, MRP-392, Materials Reliability Program: Zorita Internals Research Project (MRP-392): Radiation and Temperature Analysis of Zorita Baffle Plate and Core Barrel Weld Material. MRP-392 has not been reviewed by the NRC staff and appears to be based upon a methodology that has not been qualified.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Please justify the adequacy of the fluence estimates (e.g., comparison to experimental data, other codes, etc.) and provide estimates of associated uncertainties for the Zorita and Barsebck irradiated materials.