ML21253A010
| ML21253A010 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 10/15/2021 |
| From: | Joel Wiebe Plant Licensing Branch III |
| To: | Peter Dietrich DTE Electric Company |
| Arora S | |
| References | |
| EPID L-2020-LLR-0161 | |
| Download: ML21253A010 (15) | |
Text
October 15, 2021 Mr. Peter Dietrich Senior Vice President and Chief Nuclear Officer DTE Energy Company Fermi 2 - 260 TAC 6400 North Dixie Highway Newport, MI 48166
SUBJECT:
FERMI 2 - REVISED RELIEF REQUEST RR-A39 FOR THE FOURTH 10-YEAR INSERVICE INSPECTION INTERVAL (EPID L-2020-LLR-0161)
Dear Mr. Dietrich:
By letter dated December 30, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20365A043), as supplemented by letters dated May 27, 2021 and July 1, 2021 (ADAMS Accession Nos. ML21147A495 and ML21182A136, respectively),
DTE Energy Company (the licensee), submitted revised relief request RR-A39 to the U.S. Nuclear Regulatory Commission (NRC) for the use of alternative requirements to certain specific examination categories in use of the American Society of Mechanical Engineers Boiler and Pressure Vessel (ASME Code),Section XI, Rules for lnservice Inspection of Nuclear Power Plant Components, at Fermi 2.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Section 50.55a(z)(1), the licensee requested to use the proposed alternative in RR-A39 on the basis that the proposed alternative will provide an acceptable level of quality and safety.
The NRC staff has reviewed the subject request and concludes, as set forth in the enclosed safety evaluation, that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(1). Therefore, the NRC staff authorizes the use of proposed alternative at Fermi 2 power plant for the fourth 10-year ISI interval, which began on May 2, 2019, and is scheduled to end on May 1, 2029.
All other ASME Code requirements for which relief was not specifically requested and approved in the subject requests for relief remain applicable.
If you have any questions, please contact the Project Manager, at 301-415-1421 or e-mail at Surinder.Arora@nrc.gov.
Sincerely, Nancy L. Salgado, Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-341
Enclosure:
Safety Evaluation cc: ListServ Joel S.
Wiebe Digitally signed by Joel S.
Wiebe Date: 2021.10.15 15:51:05 -04'00'
Enclosure SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REVISED RELIEF REQUEST RR-A39 FOURTH 10-YEAR INTERVAL INSERVICE TESTING FERMI 2 DOCKET 50-341
1.0 INTRODUCTION
By letter dated December 30, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20365A043), as supplemented by letters dated May 27, 2021 and July 1, 2021 (ADAMS Accession Nos. ML21147A495 and ML21182A136, respectively), DTE Energy Company (DTE, the licensee), submitted revised relief request (RR) No. RR-A39 to the U.S. Nuclear Regulatory Commission (NRC or Commission) for the use of alternative requirements to certain specific examination categories in the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, Rules for lnservice Inspection of Nuclear Power Plant Components, at Fermi 2, associated with the fourth 10-year interval inservice inspection (ISI) interval, which began on May 2, 2019, and is scheduled to expire on May 1, 2029.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1), the licensee requested to use the proposed alternative in Revised RR-A39 on the basis that the proposed alternative will provide an acceptable level of quality and safety.
2.0 REGULATORY EVALUATION
The NRC regulations in Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(g)(4),
require that ASME Code Class 1, 2, and 3 components meet the ISI requirements, except the design and access provisions, set forth in Section XI of editions and addenda of the ASME Code to the extent practical, within the limitations of design, geometry, and materials of construction of the components. The interior accessible areas, welded attachments, and the welded core support structures in the reactor pressure vessel are categorized as ASME Code Class 1 components.
Section 50.55a(z) of 10 CFR states, in part, that alternatives to the requirements of 10 CFR 50.55a(g) may be used, when authorized by the NRC, if the licensee demonstrates:
(1) the proposed alternatives 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 The Licensee's Request for Alternative The proposed alternative is being submitted as an update to a previous RR dated February 28, 2019 (ADAMS Accession No. ML19059A327) which was approved by a safety evaluation (SE) dated July 20, 2019 (ADAMS Accession No. ML19204A253), here in referenced as July 2019, SE. The licensee proposes to utilize later versions of applicable Boling Water Reactor Vessel and Internals Program (BWRVIP) Inspection and Evaluation (I&E) Guidelines as an alternative to the ASME Code,Section XI requirements for the fourth ISI interval at Fermi 2. Specifically, the licensee requests the use of BWRVIP-25-R1-A, BWRVIP-41-R4-A, and BWRVIP-48-R1, instead of the older versions approved in the July 2019, SE.
The licensee stated that any deviations from the referenced BWRVIP guidelines for the duration of the proposed alternative will be appropriately documented and communicated to the NRC per the BWRVIP deviation disposition process. The licensee also stated that it currently has an active deviation for the core plate bolting under BWRVIP-25, dated December 1996. This deviation was resubmitted to the BWRVIP and the NRC by letter dated December 22, 2015 (ADAMS Accession No. ML15356A601). The licensee had originally stated that the applicability of this deviation will be extended until the revised BWRVIP-25 is approved by the NRC or until some other NRC-approved solution is implemented. In its current submittal, the licensee now stated that the guidance in BWRVIP-25-R1-A will be considered for the future resolution of this deviation.
3.1.1 ASME Code Components Affected ASME BPV Code,Section XI, Class 1, Examination Category B-N-1 (Interior of Reactor Vessel), and B-N-2 (Welded Core Support Structures and Interior Attachments to Reactor Vessels), Item Nos.:
B13.10 - Vessel Interior B13.20 - Interior Attachments within Beltline Region B13.30 - Interior Attachments beyond Beltline Region B13.40 - Core Support Structure 3.1.2 Applicable Code Edition and Addenda The applicable Code edition and addenda for the fourth ISI interval of Fermi 2 is the 2013 Edition of the ASME Code,Section XI.
3.1.3 Applicable Code Requirement Section XI of the ASME BPV Code requires the visual examination (VT) of certain components.
These examinations are included in Table IWB-2500-1, Categories B-N-1 and B-N-2. The scope and method of examination for these items are described as follows:
B13.10 - Examine accessible areas of the reactor vessel interior each inspection period using a technique, which meets the requirements for a VT-3 examination, as defined in paragraph IWA-2213 of the ASME BPV Code,Section XI.
B13.20 - Examine accessible interior attachment welds within the beltline region each interval using a technique which meets the requirements for a VT-1 examination as defined in paragraph IWA-2211 of the ASME BPV Code,Section XI.
B13.30 - Examine accessible interior attachment welds beyond the beltline region each interval using a technique which meets the requirements for a VT-3 examination, as defined in paragraph IWA-2213 of the ASME BPV Code,Section XI.
B13.40 - Examine accessible surfaces of the core support structures each interval using a technique which meets the requirements for a VT-3 examination, as defined in paragraph IWA-2213 of the ASME BPV Code,Section XI.
3.1.4 Duration of the Alternative The duration of the proposed alternatives is for the fourth 10-year ISI interval, which began on May 2, 2019, and is scheduled to expire on May 1, 2029.
3.1.5 Reason for Request Proposed alternative RR-A39 documents a request for the staffs approval to implement later versions of BWRVIP I&E guidelines from those versions approved in the July 2019, SE.
Specifically, the licensee is proposing the use of the following:
- 3. BWRVIP-48, Revision 1, Vessel ID Attachment Weld Inspection and Flaw Evaluation Guidelines The licensee did not otherwise propose to revise or amend the content of the proposed alternative regarding the referencing of other BWRVIP I&E guidelines that were approved in the July 2019, SE. The licensee stated that no other changes to the previously approved submittal are being proposed. Therefore, this SE will constitute a supplement to the July 2019, SE. In addition, during the review of proposed alternative RR-A39, the NRC staff became aware of a possible non-conservatism associated with BWRVIP-100, Revision A, Updated Assessment of the Fracture Toughness of Irradiated Stainless Steel for BWR Core Shrouds, and BWRVIP-76, Revision 1-A, BWR Core Shroud Inspection and Flaw Evaluation Guidelines. The use of BWRVIP-100-A and BWRVIP-76-Revision 1-A, were approved in the July 2019, SE. Therefore, the NRC staff is also addressing the licensees response to this new development within the purview of the proposed alternative.
3.2
NRC Staff Evaluation
3.2.1 Proposed Use of BWRVIP-25, Revision 1-A The licensee proposed to utilize BWRVIP-25, Revision 1-A, BWR Core Plate Inspection and Flaw Evaluation Guidelines updating from the earlier version, BWRVIP-25, as approved in the July 2019, SE. BWRVIP-25, Revision 1-A, has now been approved by the NRC staff in a SE dated March 23, 2020 (ADAMS Accession No. ML19290G703). In the July 2019, SE, the NRC staff reviewed multiple BWRVIP guidelines that addressed the scope of Section XI, B-N-1 examinations, and determined that these guidelines including BWRVIP-25, provided an equivalent visual examination method on a more frequent basis than required by the ASME Code. Therefore, as documented in the July 2019, SE, the NRC staff found it acceptable to utilize BWRVIP-25 to replace the applicable Section XI, B-N-1, examination. The NRC staff reviewed and confirmed that the recommended inspections for the various core plate subcomponents have not changed relative to BWRVIP-25, Revision 1-A. Therefore, the NRC staff concludes that the examinations in BWRVIP-25, Revision 1-A, continue to provide an equivalent visual examination as documented in the March 23, 2020, SE, and approved in the July 2019, SE. The NRC staff find it acceptable to utilize BWRVIP-25, Revision 1-A, for applicable examinations addressing the scope of B-N-1 examinations.
The NRC staff reviewed the applicability of BWRVIP-25, Revision 1-A, regarding the examinations of the core plate rim hold-down bolts (CP bolts) associated with Note (1) of Table 1 in the licensees submittal. This note states:
Based on the renewed license for Fermi 2 Enhancement 3 of [license renewal application] LRA Section B.1.10 and SER [safety evaluation report] acceptance criteria for continued examination per BWRVIP-25. The analysis is to be submitted to the NRC no later than 2 years prior to the period of extended operation). The NRC-approved guidance in BWRVIP-25-R1-A will be considered when addressing this license renewal commitment.
This enhancement is captured as commitment 7c in Appendix A of the Fermi 2 SER (ADAMS Accession No. ML16194A067) for license renewal. Additional to the content of Note (1), the commitment also provides the option of installing core plate wedges prior to the period of extended operation. The licensee stated that currently Fermi 2 has an active deviation for the core plate bolting under BWRVIP-25. This deviation is to extend the interval of applicability until the revised BWRVIP-25 is approved by the NRC or some other NRC-approved solution is implemented. As BWRVIP-25, Revision 1-A has now been published and approved by the NRC staff, the licensee has stated that this guidance will be considered for the resolution of the stated deviation. Specifically, Appendix I, Evaluation to Justify Core Plate Bolt Inspection Elimination has been added to BWRVIP-25, Revision 1-A, and has now been reviewed and accepted as documented in the March 23, 2020, SE. The NRC staff concurs that with the staffs approval of BWRVIP-25, Revision 1-A, it is now appropriate to use this guidance in lieu of BWRVIP-25, as stated in the deviation, as consideration to address commitment 7c of Appendix A.
Based on the above, the staff find that the applicants proposed implementation of BWRVIP-25, Revision 1-A, is acceptable because its use is consistent with the NRC staffs evaluation of BWRVIP-25, Revision 1-A, and it continues to provide the same recommended inspections in lieu of Section XI examinations as previously approved.
3.2.2 Proposed Use of BWRVIP-41, Revision 4-A The licensee proposed to utilize BWRVIP-41, Revision 4-A, BWR Jet Pump Assembly Inspection and Flaw Evaluation Guidelines updating from the earlier version, BWRVIP-41, Revision 3, as previously approved in the July 2019, SE. BWRVIP-41, Revision 4-A has now been approved by the NRC staff in an SE dated July 2, 2018 (ADAMS Accession No. ML18130A024), and has incorporated the NRC staffs conditions identified in this SE. The NRC staff approved the use of BWRVIP-41, Revision 3, as documented in the July 2019, SE, by doing a comparison to BWRVIP-41, Revision 4. The NRC staff concluded that examinations conducted under BWRVIP-41, Revision 3, provided an equivalent detectability and characterization result as a B-N-1 examination of the vessel interior, in conjunction with the other BWRVIP I&E guidelines credited by the licensee. As BWRVIP-41, Revision 4, has been approved and was the basis for approval in the July 2019, SE, the NRC staff finds it acceptable to now utilize the NRC-approved version (BWRVIP-41, Revision 4-A) as an alternative to the Category B-N-1 examination.
The NRC staff considered whether any circumstances existed that were not included in BWRVIP-41, Revision 4-A, including the incorporated NRC conditions, that would preclude the use of its guidelines at Fermi 2. The staff found one instance which addresses the efficacy of hydrogen water chemistry (HWC), in combination of noble metal chemical addition (NMCA), or on-line noble chemical addition (OLNC). All these methods provide chemical mitigation for reducing the occurrence of intergranular stress corrosion cracking (IGSCC) in boiling water reactor (BWR) units.
By letter dated January 24, 2018 (ADAMS Accession No. ML18033A323), the BWRVIP issued interim guidance for BWRVIP-62-A associated with the implementation of an OLNC program to take certain inspection credits for jet pump components addressed within the purview of BWRVIP-41, Revision 4-A. By letter dated July 6, 2018 (ADAMS Accession No. ML18142A019), the NRC staff found the use of ONLC acceptable subject to the limitations specified within the SE. Specifically, the NRC staff concluded that plants which apply the OLNC method must meet the criteria of a Category 3a plant in BWRVIP-62-A to be able to take the applicable inspection credit.
In its supplemental response to the proposed alternative dated May 27, 2021, the licensee stated that Fermi 2 employs an OLNC chemistry injection system for mitigating IGSCC in BWR Internals and has done so since March 2011. Therefore, the NRC staff notes that Fermi 2 must meet the requirements for a Category 3a NMCA plant, in addition to the applicable criteria that must be met for a plant applying noble metal chemistry, to credit inspection relief for the applicable jet pump components.
In its review of the licensees response to a request for additional information (supplemental letter dated May 27, 2021), the NRC staff confirmed that Fermi 2 met the criteria in accordance with the applicable requirements in BWRVIP-62-A and those in the supplemental SE for BWRVIP-62-A. Therefore, the NRC staff finds it acceptable for Fermi 2 to take inspection credit for the applicable jet pump components addressed in BWRVIP-41, Revision 4-A. Based on the above, the NRC staff find that the licensees proposed implementation of BWRVIP-41, Revision 4-A, is acceptable because the licensee continues to implement an equivalent examination program as previously approved and has demonstrated that it meets the requirements to implement an OLNC program.
3.2.3 Proposed Use of BWRVIP-48, Revision 1 The licensee proposed to utilize BWRVIP-48, Revision 1, Vessel ID Attachment Weld Inspection and Flaw Evaluation Guidelines updating from the earlier version, BWRVIP-48-A, approved in the July 2019, SE. The NRC has not reviewed BWRVIP-48, Revision 1, in its entirety. To justify the use of this report without submitting it for review, the licensee references a technical evaluation developed by BWRVIP as justification to revise the inspection frequency for the primary and supplemental core spray (CS) piping brackets. This technical evaluation was submitted to the NRC by Entergy as part of a supplemental letter (ADAMS Accession No. ML20314A125) for a RR (ADAMS Accession No. ML20160A032) proposing to update to BWRVIP-48, Revision 1, for Grand Gulf Nuclear Station and River Bend Station. The evaluation constitutes Appendix E, Section E.4 of BWRVIP-48, Revision 1. The supplemental letter also included Table F-1 of BWRVIP-48, Revision 1, which includes a summary of changes in Revision 1 of the document. The licensee references Entergys supplemental letter dated November 9, 2020, (ADAMS Accession No. ML20314A125), as part of the justification for approving the use of BWRVIP-48, Revision 1, for Fermi 2. The licensee stated that the principle change in BWRVIP-48, Revision 1, is the update to the periodic inspection strategy of the CS piping bracket attachments which is contained in Table 3-2, "Bracket Attachment Inspection Recommendations" of BWRVIP-48, Revision 1. This revision changes the inspection interval from 100 percent every four refueling cycles to 100 percent every 10 years. Appendix E, Section E.4, of BWRVIP-48, Revision 1, addresses the following seven items related to aging management of CS piping brackets:
- 1. Historical Performance of CS Piping Brackets
- 2. SCC Susceptibility Discussion
- 3. Weld Metal Susceptibility
- 4. HAZ on the (vessel side) Susceptibility
- 6. SCC Risk Assessment
- 7. Qualitative Risk Assessment Summary and Conclusion The NRC staff has addressed all these items within this SE.
The NRC staff has reviewed the referenced supplemental letter and found it acceptable to support the licensees (DTE) proposed alternative.
In the July 2019, SE, the NRC staff had assessed the use of BWRVIP-48-A as an alternative examination for the jet pump riser brace (primary) and CS piping primary and supplemental brackets. Regarding the jet pump riser brace, the NRC staffs evaluation found that the use of BWRVIP-48-A was an acceptable alternative to the applicable Section XI requirements. The NRC staff reviewed the changes in BWRVIP-48, Revision 1, contained in Table F-1, and confirmed that the evaluation provided in the July 2019, SE still remained valid. The NRC staff confirmed that the changes between the revisions did not change the technical justification previously reviewed and approved in the July 2019, SE. The NRC staff notes that the only applicable revision to the jet pump riser brace inspection is a clarification of the areas of interest for vessel attachment welds. Specifically, Table 3-2 of BWRVIP-48, Revision 1, was revised to denote the attachment welds and the heat affected zones (HAZs) on the vessel side of the weld instead of welds and heat-affected zones. Therefore, the NRC staff finds it acceptable to use BWRVIP-48, Revision 1, as an alternative to applicable Section XI requirements for the jet pump riser brace.
The BWRVIP stated in its evaluation that the vessel attachment welds for CS piping brackets are susceptible to SCC. The NRC staff notes that the susceptibility to SCC is caused by the combined effects of stresses in the weld, high oxygen concentration in the CS systems due to lack of hydrogen water protection, and susceptible materials. The NRC staff further notes that the susceptibility to SCC in these welds is confined to the groove weld joints. The following addresses the NRC staffs evaluation of each section of Appendix E, Section E.4, of BWRVIP-48, Revision 1.
3.2.3.1 Historical Performance of CS Piping Bracket Attachment Welds The BWRVIP stated that to date, there have been 450 detailed visual examinations performed in the US BWR fleet using Enhanced Visual Testing (EVT-1) for these welds. BWRVIP further stated that coverage obtained in these welds ranged from 80-100 percent for the primary CS brackets and from 60-95 percent for CS supplemental brackets. The NRC staff notes that these coverage values have been achieved with the current and more conservative interpretation of EVT-1 coverage. The NRC staff performed a review of known operating experience to date and found no results indicating that SCC cracking has been reported in the CS piping backet attachment welds. The NRC staff acknowledges that the EVT-1 inspection technique provides an adequate inspection for identifying small indications in the required coverage. In the absence of SCC in CS bracket attachment welds, the NRC staff can reasonably conclude that if any SCC cracks were in existence they would likely have been identified.
3.2.3.2 SCC Susceptibility Discussion The NRC staff notes that there are three factors for the initiation of SCC which include the environment, material susceptibility, and sufficient tensile stresses. These factors are discussed in the BWRVIP evaluation specific to the CS backet welds. The NRC staff notes that, as specified in Figure 2-6 of BWRVIP-48-A (ADAMS Accession No. ML043290381), the weld material applicable to the CS piping backet attachment welds at Fermi 2 is Alloy 182. The NRC staffs evaluation is addressed below.
(1) Environment: The CS piping bracket attachment welds are exposed to a highly oxidizing environment with no effective protection from hydrogen. Therefore, these welds are considered susceptible to SCC regardless of hydrogen water chemistry (HWC). The NRC staff notes that a sample population in the CS piping system is inspected to monitor the potential of SCC during each refueling outage for the BWR fleet.
(2) Material Susceptibility: The CS piping bracket welds are fabricated using Nickel Alloy 182 as specified in BWRVIP-48, Revision 1. The NRC staff concurs with the BWRVIP that Alloy 182 weld material is susceptible to SCC.
(3) Tensile Stresses: The BWRVIP stated that the only applied load to the CS brackets is the dead weight of the CS pipe. It is further stated that the loading due to the dead weight is minimal, therefore, the predominant drivers for SCC are residual stresses in the weld joint between the CS attachment brackets and the vessel wall. The NRC staff concurs that residual stresses are the most likely factor for initiating SCC. The NRC staff notes that these residual stresses may stem from the original welding process and/or weld repairs during fabrication. The NRC staff further notes that any residual stress added from a weld repair could accentuate the initiation of cracking during service.
3.2.3.3 Weld Metal SCC Susceptibility BWRVIP states that reactor vessels fabricated by Combustion Engineering (CE) have CS piping bracket welds fabricated from Alloy 182, and are therefore, at least nominally susceptible to SCC. NRC notes that the reactor vessel at Fermi 2 was fabricated by CE. Therefore, for consideration to the applicability of BWRVIP-48, Revision 1, the NRC staff confirmed that the CS piping bracket welds for Fermi 2 are susceptible to SCC.
3.2.3.4 Weld Metal and HAZ (vessel side) Assessment BWRVIP states that, for CE plants, the weld buildup pad or inlay for the CS piping bracket attachment are fabricated with Alloy 182 weld material. The NRC staff notes that the HAZs on the vessel side of the weld and the groove welds that join the bracket attachments are both fabricated with Alloy 182 weld material and susceptible to SCC. Therefore, for consideration to the applicability of BWRVIP-48, Revision 1, the NRC staff confirmed that the CS piping bracket welds for Fermi 2 are susceptible to SCC.
3.2.3.5 State of Stress (Weld and HAZ on Vessel side)
The BWRVIP stated that the applied loads on the CS piping brackets during normal operations are low, hence it can be concluded that the predominant driving force for SCC is the residual weld stresses associated with the double welded groove joint of the CS bracket welds. These bracket welds are shown in Figure 2-6 of BWRVIP-48-A (ADAMS Accession No. ML043290381). The BWRVIP stated that shrinkage stresses associated with the solidification cooling of the weld metal pool could increase the residual stresses in the weld joint. This increase in the weld residual stress is directly proportional to weld joint restraint offered by the thicker member in the weld joint assembly. The NRC staff notes that residual stresses can generally be minimized by using a proper welding sequence at the top and bottom portions of a double groove weld joint to balance the stresses and minimize distortion. The NRC staff further notes that during fabrication of the CS piping brackets at Fermi 2, the piping brackets were not completely restrained allowing them to move freely during the welding process. Therefore, the NRC staff determined that weld residual stresses in the CS piping bracket groove joints could not have exceeded the required threshold limit to initiate SCC.
3.2.3.6 SCC Risk Assessment The NRC staff determined that to date, no SCC has been discovered in the CS piping bracket welds even though the CS piping systems are exposed to highly oxygenated reactor coolant.
The NRC staff further notes, that as a matter of operating experience, the industry identified several cracks in the CS piping systems during the initial ten-year interval of operation followed by a declining trend with additional operating time. The staff notes that this is a similar trend that has been observed in many other BWR components. With respect to CS piping brackets welds, given that no cracks have been observed to date and non-destructive examination (NDE) capabilities have improved, the staff concludes that the initiation of a new crack or propagation of an existing crack during the fourth ISI interval at Fermi 2 would be identified before reaching an unacceptable size. The CS piping brackets would be routinely inspected in accordance with the inspection criterion addressed in the BWRVIP-48, Revision 1. The staff determined that due to the time dependent nature of SCC growth rates, if any new cracks were to occur in these welds, there is reasonable assurance that they will be identified and corrective actions would be taken by the licensee in a timely manner. Hence, the NRC staff determines that the risk factor for failing to identify SCC in a timely manner is low.
3.2.3.7 Qualitative Risk Assessment and Conclusions The NRC staff notes that BWRVIP has implemented a conservative approach to monitor SCC in CS piping bracket welds. The NRC staff further notes that this is corroborated by the performance of approximately 450 inspections of these welds with no evidence of cracking.
Therefore, the NRC staff finds the licensees proposed inspection frequency addressed in BWRVIP-48, Revision 1 acceptable based on the following observations, (1) If any detection of SCC in the CS piping welds were to occur in the BWR fleet, it would prompt the industry and the NRC to further evaluate the applicable weld inspection frequencies. In addition, recent examination data will be regularly available to the BWR fleet due to the frequency of inspections and refueling outages.
(2) Previous inspection results addressed in BWRVIP-18, Revision, 2-A, indicated that over time, the average crack growth rate because of SCC tends to decrease. Since SCC is a time dependent phenomenon, based on item (1) above, there is reasonable assurance that emerging cracks can be identified in a timely manner.
(3) All the requirements addressed in BWRVIP-48-A to include re-inspection and scope expansion criteria continue to be applicable to the CS piping bracket welds.
Based on these observations, the NRC staff concludes that the proposed inspection frequency provides reasonable assurance that the aging degradation due to SCC in CS piping bracket welds is being adequately managed by the licensee during the fourth ISI interval. The NRC staffs approval of the proposed inspection frequency for CS piping bracket welds contained in BWRVIP-48, Revision 1, is on a plant-specific basis and only applicable for the duration of the fourth ISI interval at Fermi 2.
3.2.4 Potential Non-Conservatism in Electric Power Research Institute (EPRI) Reports The NRC staff became aware of a possible non-conservatism associated with the fracture toughness model published in BWRVIP-100-A, Updated Assessment of the Fracture Toughness of Irradiated Stainless Steel for BWR Core Shrouds which is incorporated into BWRVIP-235, BWR Vessel and Internals Project, Structural Analysis Software for BWR Internals, DLL Version 3.1. This issue was communicated to the licensee through a 10 CFR Part 21, transfer of information notice, dated February 19, 2021, and was updated on March 19, 2021. This information was also communicated to the NRC in EPRIs letter dated March 22, 2021 (ADAMS Accession No. ML21084A164). The licensee states that the transfer of information notice was promptly entered into Fermi 2s Corrective Action Program (CAP) for review and evaluation.
The NRC staff is aware that this possible non-conservatism exists in a fluence range of 5 x 1020 n/cm² to 3 x 1021 n/cm². More specifically, the NRC staff notes that current experimental data suggests the lower bound fracture toughness of 50 ksi-in is applicable at fluences of or greater than 5 x 1020 n/cm² as opposed to the defined threshold of 3 x 1021 n/cm² in BWRVIP-100-A. Similarly, the NRC staff also became aware that this non-conservatism could potentially impact BWRVIP-76, Revision 1-A, BWR Core Shroud Inspection and Flaw Evaluation Guidelines. In the 10 CFR Part 21, transfer of information notice, EPRI communicated that this document could not be used in its entirety as written and provided recommended actions. The NRC staff evaluated how this possible non-conservatism affected the current and previously approved proposed alternatives.
3.2.4.1 BWRVIP-100, Revision A By supplemental letter dated July 1, 2021, the licensee is withdrawing the request to utilize BWRVIP-100-A, from proposed alternative RR-A39. The NRC staff had approved the use of BWRVIP-100-A as documented in the July 2019 SE and shown in Table 1, Table 1-BWRVIP Guidelines Referenced in RR-A39 of the SE. The NRC staff notes that the approval to use BWRVIP-100-A allowed the licensee to reference fracture toughness values for flaw evaluations, but it did not affect the performance of the required inspections as outlined in the proposed alternative. Regarding the 10 CFR Part 21, transfer of information notice, the licensee stated that EPRI is working to revise the fracture toughness correlations currently present in BWRVIP-100-A. The licensee further stated that once BWRVIP-100-A is revised, and if the licensee opts to use this guidance as part of the proposed alternative, it will resubmit its request in accordance with 10 CFR Part 50.55a. Therefore, the NRC staff finds that the request to withdraw the use of BWRVIP-100-A, is acceptable because it does not affect the performance of inspections and evaluation of results under the proposed alternative.
3.2.4.2 BWRVIP-76, Revision 1-A The NRC staff notes that the use of BWRVIP-76, Revision 1-A, was approved in the July 2019 SE and included in Table 1 of the SE. By supplemental letter dated July 1, 2021, the licensee communicated to the NRC staff that it plans to continue to implement the inspection criteria provided in BWRVIP-76, Revision 1-A, as modified by EPRI recommendations and applicable to the proposed alternative. The licensee states that as part of the transfer of information notice, EPRI provided preliminary recommendations related to revisions of applicable inspection intervals and flaw evaluation guidance. The NRC staff notes that there are two separate items related to the EPRIs recommendations that the staff has evaluated:
- 1. Inspection Criteria
- 2. Flaw Evaluation Guidance Inspection Criteria - The licensee states that a review of the EPRI recommendations applicable to Fermi 2 shows that the recommended inspection intervals have not resulted in any missed inspections. The NRC staff understands this change to be a conservative approach, and therefore, requires inspection intervals to be shortened. The licensee further states that it plans to programmatically incorporate revised inspections after they are reviewed by the BWRVIP Executive Committee. The NRC notes that the licensee had already been implementing BWRVIP-76, Revision 1-A, as approved by the July 2019, SE, and that it had recently credited the ultrasonic inspection of multiple welds in RF18 (2017) in lieu of Section XI requirements. The NRC staff notes that no flaws were identified in these inspections. As documented in the July 2019, SE, the NRC staff concurred that the BWRVIP guidelines used the same examination method or a more stringent examination method than Section XI of the ASME Code. In addition, the NRC staff noted that examinations were conducted at an equal or shorter interval and were therefore acceptable. The NRC staff has reasonable assurance that if the EPRI-recommended revised inspections are performed within a shorter interval than previously performed using the same detectability and characterization methods, it will continue to meet the applicable inspection requirements as approved in the July 2019, SE.
Flaw Evaluation Guidance - In response to the EPRI-recommended actions, the licensee stated that Fermi 2 has proactively updated its core shroud flaw analysis and no IGSCC flaws have been identified to date. The licensee further stated that if an emergent reevaluation is required to address a condition in the core shroud, the preliminary assessment criteria presented in the meeting between EPRI and the NRC on May 27, 2021, will be utilized by Fermi
- 2. The NRC staff has discussed this analytical approach with EPRI on multiple occasions. The NRC staff concludes that until such time permanent guidance is implemented by EPRI, based on those discussions, finds it acceptable for Fermi 2 to use the preliminary assessment criteria proposed by EPRI during an emergent flaw analysis reevaluation.
3.2.5 Applicable BWRVIP Guidelines The licensee has communicated that it will be withdrawing the utilization of BWRVIP-100-A and continue to use BWRVIP-76, Revision 1-A which are referenced in Table 1 of the July 2019, SE.
Note 3 of this Table associates the flaw evaluations performed under BWRVIP-76, Revision 1-A, to the fracture toughness values of BWRVIP-100-A. Since BWRVIP-100-A will be withdrawn, the licensee stated that Note 3 of Table 1 will also be withdrawn. The NRC staff finds this change acceptable given the licensees withdrawal of BWRVIP-100-A and as evaluated above. Therefore, commensurate with the supplemental response dated July 1, 2021, the NRC staff understands that the following Table supersedes Table 1 of the July 2019, SE.
Updated BWRVIP Guidelines Referenced in RR-A-30 BWRVIP-03NP BWR Vessel and Internals Project, Reactor Pressure Vessel and Internal Examination Guidelines BWRVIP-06-R1-A BWR Vessel and Internals Project, Safety Assessment of BWR Internals BWRVIP-14-A BWR Vessel and Internals Project, BWR Evaluation of Crack Growth in BWR Stainless Steel RPV Internals BWRVIP-18-A-R2 BWR Core Spray Internals Inspection and Flaw Evaluation Guidelines, (Licensee Renewal (LR) Safety Evaluation Report (SER) updated to Revision 2 in the annual update of May 9, 2016)
BWRVIP-25-R1-A (1) BWR Core Plate Inspection and Flaw Evaluation Guidelines BWRVIP-26-A BWR Top Guide Inspection and Flaw Evaluation Guidelines BWRVIP-27-A BWR Standby Liquid Control System/Core Plate P Inspection and Flaw Evaluation Guidelines BWRVIP-38 BWR Shroud Support Inspection and Flaw Evaluation Guidelines BWRVIP-41-R4-A BWR Jet Pump Assembly Inspection and Flaw Evaluation Guidelines BWRVIP-47-A BWR Lower Plenum Inspection and Flaw Evaluation Guidelines BWRVIP-48-R1(2)
Vessel ID Attachment Weld Inspection and Flaw Evaluation Guidelines BWRVIP-49-A Instrument Penetration Inspection and Flaw Evaluation Guidelines BWRVIP-74-A BWR Reactor Vessel Inspection and Flaw Evaluation Guidelines BWRVIP-76-R1-A BWR Core Shroud Inspection and Flaw Evaluation Guidelines (LR) SER updated to Revision 1-A in the annual update of May 9, 2016)
BWRVIP-94NP BWR Vessel and Internals Project Program Implementation Guide Notes:
(1) Based on the renewed license for Fermi 2 Enhancement 3 of LRA Section B.1.10 and SER Section 3.0.3.2.3, BWRVIP-25 shall be met by submittal of an analysis justifying the elimination of inspections for the core plate bolting or an analysis determining acceptance criteria for continued examination per BWRVIP-25. The analysis is to be submitted to the NRC no later than 2 years prior to the period of extended operation.
(2) Currently, there are no existing BWRVIP guidelines or ASME Code Section XI requirements regarding the feedwater spargers except for BWRVIP-48-A which governs inspection of reactor vessel internal attachment welds, namely the feedwater sparger brackets. Fermi 2 will continue to use inspections modeled after the guidance of NUREG-0619 on the feedwater spargers outside of this request
4.0 CONCLUSION
As set forth above, the NRC staff has determined that the proposed alternative, RR-A39, provides an acceptable level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed the regulatory requirements set forth in 10 CFR 50.55a(z)(1). Therefore, the NRC staff authorizes the use of proposed alternative RR-A39 at Fermi 2 for the fourth 10-year ISI interval, which began on May 2, 2019, and is scheduled to end on May 1, 2029.
All other ASME Code,Section XI, requirements for which an alternative was not specifically requested and authorized remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
Principal Contributor: Isaac Anchondo-Lopez, NRR Date: October 15, 2021
- by memo OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DNRL/NVIB/BC* NRR/DORL/LPL3/BC NAME SArora SRohrer ABuford NSalgado (JWiebe for)
DATE 09/08/2021 09/20/2021 08/19/2021 10/15/2021