ML23159A230
| ML23159A230 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf, River Bend  |
| Issue date: | 06/08/2023 |
| From: | Couture P Entergy Operations |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| Shared Package | |
| ML23159A229 | List: |
| References | |
| CNRO2023-00015 | |
| Download: ML23159A230 (1) | |
Text
NOTICE: Enclosure 4 to this letter contains Proprietary Information to be withheld from public disclosure per 10 CFR 2.390. Upon separation from Enclosure 4, this letter is DECONTROLLED.
Phil Couture Senior Manager Fleet Regulatory Assurance - Licensing 601-368-5102 Entergy Operations, Inc., 1340 Echelon Parkway, Jackson, MS 39213 CNRO2023-00015 10 CFR 50.55a(z)(1)
June 8, 2023 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
Subject:
Response to Request for Additional Information concerning Request to Update ASME Code Relief Request Safety Evaluations with NRC-Approved Revision of BWRVIP Guidelines (GG-ISI-020, and RBS-ISI-019)
Grand Gulf Nuclear Station, Unit 1 NRC Docket No. 50-416 Renewed Facility Operating License No. NPF-29 River Bend Station, Unit 1 NRC Docket No. 50-458 Renewed Facility Operating License No. NPF-47 By letter dated December 6, 2022 (Reference 1), Entergy Operations, Inc. (Entergy) submitted a proposed revision to the U.S. Nuclear Regulatory Commission (NRC)-approved alternatives (GG-ISI-020 and RBS-ISI-019) to use Boiling Water Reactor (BWR) Vessel and Internals Project (BWRVIP) guidelines for Grand Gulf Nuclear Station, Unit 1 (Grand Gulf), and River Bend Station, Unit 1 (River Bend). Specifically, pursuant to 10 CFR 50.55a(z)(1), Entergy requested to revise these NRC-authorized alternatives to allow the use of BWRVIP-48, Revision 2, "BWR Vessel and Internals Project BWR Vessel ID [Inside Diameter] Attachment Weld Inspection and Flaw Evaluation Guidelines," as applicable, instead of the currently authorized versions of these guidelines.
The NRC staff has reviewed the request and determined that additional information was required to complete their review (Reference 2). An extension to the response due date was subsequently granted by the NRC (Reference 3). provides the Entergy response to the Request for Additional Information (Reference 2). Enclosure 2 is the letter from Electric Power Research Institute (EPRI) to the NRC providing a Request for withholding on information supporting this response. Enclosure 4
NOTICE: Enclosure 4 to this letter contains Proprietary Information to be withheld from public disclosure per 10 CFR 2.390. Upon separation from Enclosure 4, this letter is DECONTROLLED.
CNRO2023-00015 Page 2 of 3 provides Section G, "Screening of BWRVIP-48, Rev. 2 in accordance with Appendix C of NEI 03-08, Rev. 4, 'Document Screening'" and Section H, "Revision Details (BWRVIP-48, Revision 2)." Some information provided in Enclosure 4 is considered proprietary to EPRI and request it to be withheld from public disclosure in accordance with 10 CFR 2.390 of the Commissions regulations. The proprietary information is identified by highlighted yellow shading and the pages with proprietary information are also marked with the letters "TS" in the page footer indicating that the information is considered trade secrets in accordance with 10 CFR 2.390.
The non-proprietary version is provided in Enclosure 5.
This information is supported by an affidavit, signed by Steve Swilley, Vice President and Deputy Chief Nuclear Officer at Electric Power Research Institute, Inc., (EPRI, 1300 West W.T.
Harris Boulevard, Charlotte, NC 28262-8550), the owner of the information. The affidavit sets forth the basis by which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (a)(4) of 10 CFR 2.390 of the Commission's regulations. The affidavit is included in Enclosure 3.
This letter contains no new regulatory commitments.
Should you have any questions or require additional information, please contact me at (601) 368-5102.
Respectfully, Phil Couture PC/gpn
Enclosures:
- 1. Response to Request for Additional Information for GG-ISI-020 and RBS-ISI-019
Philip Couture Digitally signed by Philip Couture Date: 2023.06.08 13:16:04 -05'00'
NOTICE: Enclosure 4 to this letter contains Proprietary Information to be withheld from public disclosure per 10 CFR 2.390. Upon separation from Enclosure 4, this letter is DECONTROLLED.
CNRO2023-00015 Page 3 of 3
References:
- 1. Entergy, letter to the U. S. Nuclear Regulatory Commission (NRC),
"Request to Update ASME Code Relief Request Safety Evaluations with NRC-Approved Revision of BWRVIP Guidelines (GG-ISI-020, and RBS-ISI-019)" (ML22340A672), dated December 6, 2022.
- 2. NRC email to Entergy, "Entergy Fleet (Grand Gulf and River Bend) - Final RAI for Request to Update ASME Code Relief Request Safety Evaluations with NRC-Approved Revision of BWRVIP Guidelines (GG-ISI-020, and RBS-ISI-019) (EPID L-2022-LLR-0090), dated April 26, 2023.
- 3. NRC email to Entergy, "Entergy Fleet (GGNS and RBS) - Update ASME Code Relief Request Safety Evaluations with NRC-Approved Revision of BWRVIP Guidelines (GG-ISI-020, and RBS-ISI-019) (EPID L-2022-LLR-0090), dated May 26, 2023.
cc:
NRC Region IV-Regional Administrator NRC Senior Resident Inspector-Grand Gulf Nuclear Station NRC Senior Resident Inspector-River Bend Station NRC Project Manager-Grand Gulf Nuclear Station NRC Project Manager-River Bend Station NRC Project Manager-Entergy Fleet
ENCLOSURE 1 CNRO2023-00015 Response to Request for Additional Information for GG-ISI-020 and RBS-ISI-019
CNRO2023-00015 Page 1 of 2 Response to Request for Additional Information for GG-ISI-020 and RBS-ISI-019 Request for Additional Information (RAI)-1 Issue:
Nuclear Energy Institute (NEI) 03-08, "Guideline for the Management of Materials Issues,"
Revision 4 (ADAMS Accession No. ML20315A536) provides the industrys process for determining when an updated industry inspection and evaluation guideline may be implemented by an industry licensee without NRC prior approval of the guideline. Grand Gulf and River Bend propose to use BWRVIP-48, Revision 2, as the basis for inspecting reactor pressure vessel (RPV) interior attachment welds in lieu of applicable ASME Section XI, Examination Category B-N-2 requirements. BWRVIP-48, Revision 2 was not submitted to the NRC staff for review and approval.
Request:
Describe the changes between BWRVIP-48-A and BWRVIP-48, Revision 2. Based on these changes, provide the plant-specific bases for Grand Gulf and River Bend to demonstrate that the proposed implementation of BWRVIP-48, Revision 2 provides an acceptable alternative to ASME Code requirements for the inspection of the applicable RPV interior attachments welds.
Entergys Response to RAI No. 1:
As described in the safety evaluation (Reference 6), the core spray piping bracket to vessel Internal Dimension (ID) attachment welds and heat affected zones (HAZs) are identified as the applicable RPV interior attachment welds with revised inspection frequencies applied. The associated changes between BWRVIP-48-A and BWRVIP-48 Revision 2, applicable for the core spray piping bracket welds, are for these inspection frequency requirements. The reinspection requirements in BWRVIP-48-A are 100% every four refueling cycles (every eight-year intervals for units with 24 month refueling cycles or every six years for plants with 18-month refueling cycles) as shown in Table G-3 of BWRVIP 2023-034 (Enclosure 4). BWRVIP-48 Revision 2 provides a revised periodic inspection frequency that changes the inspection interval to 25%
every 12 years. Tables G-10 and G-11 in BWRVIP 2023-034 (Enclosure 4) summarize the proposed changes to inspection requirements for BWRVIP-48 Revision 2 and provides a summary of the technical bases for these changes, respectively. Note that Table G-10 contains two rows associated with Core Spray Piping. Entergy is not proposing to adopt any of the other changes associated with this revision. The technical justification for changing the inspection frequency is reported in the Qualitative Risk Assessment for BWRVIP-48, Revision 2, Appendix G Sections G.4 and G.5 as provided in BWRVIP 2023-034 (Enclosure 4). The revision details for BWRVIP-48 Revision 2 are listed in Table H-1 of BWRVIP 2023-034 (Enclosure 4).
Grand Gulf Nuclear Station (GGNS) core spray piping bracket weld inspections from 1997 to 2017 resulted in no recordable indications (Reference 4).
River Bend Station (RBS) core spray piping bracket weld inspections from 1998 - 2016 resulted in only one indication, with discovery in 2012 (14-year period) (Reference 5). In 2014, the weld was inspected with no change in indication and no others discovered.
CNRO2023-00015 Page 2 of 2 Aligning with the inspection frequency of BWRVIP-48 Revision 2 of 25% every 12 years fall within an acceptable period for detecting recordable indications based on GGNS and RBSs historical data.
References:
- 1) BWRVIP-48-A, "BWR Vessel and Internals Project Vessel ID Attachment Weld Inspection and Flaw Evaluation Guidelines".
- 2) BWRVIP-48, Revision 1-A, "BWR Vessel and Internals Project, Low Pressure Coolant Injection (LPCI) Coupling Inspection and Flaw Evaluation Guidelines".
- 3) BWRVIP-48, Revision 2, "BWR Vessel and Internals Project, Vessel ID Attachment Weld Inspection and Flaw Evaluation Guidelines".
- 4) Proposed Alternative with 10 CFR 50.55 (z)(1) Use of Boiling Water Reactor Vessel and Internals Project (BWRVIP) Guidelines at Grand Gulf Nuclear Station, GNRO-2017-00009, Enclosure 3 to Attachment - Grand Gulf Reactor Vessel Internal Inspection History (ML17074A625), dated March 15, 2017.
- 5) Request for Alternative in Accordance with 10 CFR 50.55a (z)(1) Use of Boiling Water Reactor Vessel and Internals Project (BWRVIP) Guidelines in Lieu of Specific ASME Code Requirements (RR-RBS-ISI-019), RBG-47827, Attachment 3 - RBS Reactor Vessel Internal Inspection History (ML18045A151), dated February 13, 2018.
- 6) Letter from NRC to Entergy "Grand Gulf Nuclear Station, Unit 1 and River Bend Station, Unit 1 - Request to update American Society of Mechanical Engineers Code Relief Request Safety Evaluations with NRC-approved Revision of Boiling Water Reactor Vessel and Internals Project Guidelines (EPID L-2020-LLR 0079)" (ML21258A408),
dated September 21, 2021.
ENCLOSURE 2 CNRO2023-00015 BWRVIP 2023-034, Attachment 1, Transmittal of EPRI Proprietary Affidavit (1 Page)
BWRVI P 2023-034, Attachment 1 EPl21 Ref. EPRI Docket No. 99902016 May 4, 2023 Document Control Desk Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 STEVE SWILLEY Vice President and Deputy Chief Nuclear Officer
Subject:
Request for Withholding of the following Proprietary Information Included in:
Response to Request for Additional Information for Request to Update ASME Code Relief Request Safety Evaluations with NRG-Approved Revision of BWRVIP Guidelines, GG-ISl-020 and RBS-ISl-019 (CNRO2022-00026, ML22340A672)
Grand Gulf Nuclear Station, Unit 1 River Bend Station, Unit 1 Renewed Facility Operating License Nos. NPF-29 and NPF-47 NRC Docket Nos. 50-416 and 50-458 Excerpts from BWRVIP-48, Revision 2: Vessel ID Attachment Weld Inspection and Flaw Evaluation Guidelines (EPRI ID 3002018321)
To Whom It May Concern:
This is a request under 10 C.F.R. §2.390(a)(4) that the U.S. Nuclear Regulatory Commission ("NRC") withhold from public disclosure the information identified in the enclosed Affidavit consisting of the proprietary information owned by Electric Power Research Institute, Inc. ("EPRI") identified above in the attached report. Proprietary and non-proprietary versions of the Report and the Affidavit in support of this request are enclosed.
EPRI desires to disclose the Proprietary Information in confidence to assist the NRC review of the enclosed submittal to the NRC by Entergy. The Proprietary Information is not to be divulged to anyone outside of the NRC or to any of its contractors, nor shall any copies be made of the Proprietary Information provided herein. EPRI welcomes any discussions and/or questions relating to the information enclosed. In case the NRC rejects this affidavit for protection for any reason, EPRI requests that the NRC contact the EPRI project manager and EPRI Order Center (vvaughn@epri.com) with an offer of an opportunity for EPRI to withdraw this submittal.
If you have any questions about the legal aspects of this request for withholding, please do not hesitate to contact me.
Questions on the content of the Report should be directed to Nathan Palm of EPRI at (724) 288-4043.
Sincerely, Attachment(s)
Together... Shaping the Future of Energy 1300 West W.T. Harris Boulevard, Charlotte, NC 28262-8550 USA
- 704.595.2630
- Mobile 704.562.8135
- sswilley@epri.com
ENCLOSURE 3 CNRO2023-00015 BWRVIP 2023-034, Affidavit for request for withholding proprietary information (3 Pages)
Subject:
Request Withholding of the Following Proprietary Information Included Response to Request for Additional Information for Request to Update ASME Code Relief Request Safety Evaluations with NRG-Approved Revision of BWRVIP Guidelines, GG-ISl-020 and RBS-ISl-019 (CNRO2022-00026, ML22340A672)
Grand Gulf Nuclear Station, Unit 1 River Bend Station, Unit 1 Renewed Facility Operating License Nos. NPF-29 NPF-47 NRC Docket Nos. 50-416 and 50-458 Excerpts BWRVIP-48, Revision 2: Vessel ID Attachment Weld Inspection and Flaw Evaluation Guidelines (EPRI ID 3002018321)
I, Steve Swilley, being duly sworn, depose and state as follows:
I am the Vice President and Deputy Chief Nuclear Officer at Electric Power Research Institute, Inc., whose principal office is located at 3420 Hillview Avenue, Palo Alto, California ("EPRI"), and I have been specifically delegated responsibility for the above-listed Report which contains EPRI Proprietary Information that is sought under this Affidavit to be withheld "Proprietary Information". I am authorized to apply to the U.S. Nuclear Regulatory Commission ("NRC")
for the withholding of the Proprietary Information on behalf of EPRI.
EPRI Proprietary Information is identified in the above referenced excerpts from the report with highlighted yellow shading. The pages with the proprietary information are also marked with the letters "TS" in the page footer indicating that information is considered trade secrets in accordance with 10 CFR 2.390.
EPRI requests that the Proprietary Information be withheld from the public on the following bases:
Withholding Based Upon Privileged and Confidential Trade Secrets or Commercial or Financial Information (see e.g. 10 C.F.R. §2.390(a)(4)):
- a. The Proprietary Information is owned by EPRI and has been held in confidence by EPRI. All entities accepting copies of the Proprietary Information do so subject to written agreements imposing an obligation upon the recipient to maintain the confidentiality of the Proprietary Information. The Proprietary Information is disclosed only to parties who agree, in writing, to preserve the confidentiality thereof.
- b. EPRI considers the Proprietary Information contained therein to constitute trade secrets of EPRI. As such, EPRI holds the information in confidence, and disclosure thereof is strictly limited to individuals and entities who have agreed, in writing, to maintain the confidentiality of the Information. EPRI made a substantial economic investment to develop the Proprietary Information, and, by prohibiting public disclosure, EPRI derives an economic benefit in the form of licensing royalties and other additional fees from the confidential nature of the Proprietary Information. If the Proprietary Information were publicly available to consultants and/or other businesses providing services in the electric and/or nuclear power industry, they would be able to use the Proprietary Information for their own commercial benefit
and profit and without expending substantial economic resources required of Information.
to develop the
- c. EPRl's classification of the Proprietary Information as trade secrets is justified by the Uniform Trade Secrets Act, which California adopted in 1984 and a version of which has been adopted by over forty states. The California Uniform Trade Secrets Act, California Civil Code §§3426 - 3426.11, defines a "trade secret" as follows:
"Trade secret' means information, including a formula, pattern, compilation, program device, method, technique, or process, that (1) Derives independent economic value, actual or potential, from not being generally known to the public or to other persons who can obtain economic value from its disclosure oruse;and
- 2) Is the subject of efforts that are reasonable under the circumstances to maintain its secrecy."
- d. The Proprietary Information contained therein are not generally known or available to the public. EPRI developed the Information only after making a determination that the Proprietary Information was not available from public sources. EPRI made a substantial investment of both money and employee hours in the development of the Proprietary Information. EPRI was required to devote these resources and effort to derive the Proprietary Information.
As a result of such effort and cost, both in terms of dollars spent and dedicated employee time, the Proprietary Information is highly valuable to EPRI.
- e. A public disclosure of the Proprietary Information would be highly likely to cause substantial harm to EPRl's competitive position and the ability of EPRI to license the Proprietary Information both domestically and internationally.
The Proprietary Information and Report can only be acquired and/or duplicated by others using an equivalent investment of time and effort.
I have read the foregoing, and the matters stated herein are true and correct to the best of my knowledge, information, and belief. I make this affidavit under penalty of perjury under the laws of the United States of America and the laws of the State of North Carolina.
Executed at 1300 W WT Harris Blvd, Charlotte, NC being the premises and place of business of Electric Power Research Institute, Inc.
Steve Swilley
(State of North Carolina)
(County of Mecklenburg) before me on this /y day
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ENCLOSURE 5 CNRO2023-00015 BWRVIP 2023-034, Attachment 3, BWRVIP-48, Revision 2, Sections G and H (NON-PROPRIETARY)
(51 Pages)
G-1 G
SCREENING OF BWRVIP-48, REV. 2 IN ACCORDANCE WITH APPENDIX C OF NEI 03-08, REV. 4, DOCUMENT SCREENING G.1 Introduction and Purpose The purpose of this appendix is to document the results of a screening evaluation conducted to determine if BWRVIP-48, Revision 2 BWR Vessel and Internals Project: Vessel ID Attachment Weld Inspection and Flaw Evaluation Guidelines can be generically released for implementation by U.S. BWRVIP members using the screening process defined in Appendix C Document Screening of NEI 03-08, Revision 4, Guideline for the Management of Materials Issues [G1].
G.2 Summary of Technical Revisions Addressed by Screening Evaluation This screening evaluation addresses changes to periodic inspection and scope expansion requirements contained in Section 3 of BWRVIP-48. Section G.4.5 contains a summary of these changes.
G.3 NEI 03-08 Screening Determination The following sections provide responses to the questions asked in the Applicability Evaluation Flow Chart [G1, Figure C-1] and the Screening Evaluation Flow Chart [G1, Figure C-2].
G.3.1 Applicability Evaluation The Applicability Evaluation Flow Chart from NEI 03-08, Rev. 4 [G1, Appendix C, Figure 1] is shown in Figure G-1 below. Table G-1 presents the results and bases for evaluation of BWRVIP-48, Rev. 2., A full document screening evaluation is required based on Yes answers to all of the applicability evaluation questions. Section G.3.2 contains the screening evaluation.
BWRVIP 2023-034, Attachment 3
Screening of BWRVIP-48, Rev. 2 in Accordance with Appendix C of NEI 03-08, Rev. 4, Document Screening G-2 Figure G-1 Applicability Evaluation Process (From [G1, Appendix C])
Screening of BWRVIP-48, Rev. 2 in Accordance with Appendix C of NEI 03-08, Rev. 4, Document Screening G-4 Figure G-2 Screening Evaluation Process [G1, Appendix C]
Screening of BWRVIP-48, Rev. 2 in Accordance with Appendix C of NEI 03-08, Rev. 4, Document Screening G-6 Table G-2 (continued)
Screening Evaluation - Responses to Questions from Screening Evaluation Flow Chart Figure G-2
[G1]
Flow Chart Question
Response
(2e)
Sensitivity to New Data Are the evaluation results potentially sensitive YES, but addressed by normal BWRVIP activities fleet the emergent issues protocol contained in provide a method for so
-48 or issue interim guidance to address There is no need for a commitment to specifically revisit
-48 in the future.
G.4 Qualitative Risk Assessment for BWRVIP-48, Rev. 2 G.4.1 Background The inspection requirements for reactor vessel ID attachment welds contained in BWRVIP-48 were originally based on the potential susceptibility of attachment welds to SCC given the existing state of knowledge. At the time BWRVIP-48 [G4] was initially issued (1998), SCC of BWR internals was still largely in a discovery phase, with the frequency and ultimate extent of cracking largely unknown. As a result, the inspection program specified by BWRVIP-48 was purposely conservative. Over twenty years have elapsed since the initial issue of BWRVIP-48 and it is reasonable to revisit the inspection requirements in BWRVIP-48 based on the current state of knowledge regarding performance in the field and understanding of the progression of SCC in BWRs.6 G.4.2 Approach The evaluation documented in this Appendix is a qualitative risk assessment consistent with the requirements of NEI 03-08, Rev. 4 [G1] screening step 2d which permits disposition by qualitative risk assessment. As identified in screening step 2d, the evaluation should ideally be consistent with methods previously applied and either directly or indirectly accepted by NRC. To ensure that the intent of the screening guidance is met, the evaluation (1) considers all of the elements typically included within inspection optimization evaluations (i.e., field performance, NDE capability, residual stress state, and flaw tolerance), (2) includes a level of rigor consistent with prior inspection optimization evaluations and (3) applies risk principles consistent with 6 Revision 1 to BWRVIP-48 addressed a small change to the reinspection frequency for core spray piping bracket welds, from reinspection every 4 refueling cycles (every 8 years for units with 24 month refueling cycles or every 6 years for plants with 18-month refueling cycles) to reinspection every 10 years. However, this revision was not intended as a comprehensive update to inspection guidance. Revision 2 to BWRVIP-48 represents a comprehensive update of inspection guidance.
Screening of BWRVIP-48, Rev. 2 in Accordance with Appendix C of NEI 03-08, Rev. 4, Document Screening G-7 precedent inspection optimization evaluations used to provide technical bases for modifications to inspection program requirements.
Section G.4.3 defines the scope of components included in the evaluation. Sections G.4.4 through G.4.8 summarize the inputs considered relevant to the risk assessment. These include inspection history and NDE capability (Section G.4.4), field performance (G.4.5), evaluations of SCC and fatigue cracking susceptibility (Sections G.4.6 and G.4.7), and evaluation of flaw tolerance (Section G.4.8). Section G.4.9 contains the risk assessment results associated with the revisions to the inspection program made in BWRVIP-48, Revision 2.
Additionally, it is noted that the general methodology is consistent with the approach taken to optimize BWRVIP requirements for other reactor internals components found not to have significant SCC susceptibility. As such, precedent exists for the approach taken, consistent with NEI 03-08, Rev. 4 screening step 2c. See Section G.6.
G.4.3 Scope The components addressed by this evaluation are the vessel ID attachment welds for core spray piping (primary and supplemental brackets), jet pump riser braces (primary only), steam dryer supports, and feedwater spargers. Additionally, the inspections specified by BWRVIP-48 are limited to the groove weld of the bracket or brace attachment to the weld buildup pad or cladding on the vessel wall and the associated HAZ induced by the groove weld in the weld buildup pad or cladding. The only exception is for steam dryer support brackets for which the inspection requirement includes the bracket-side HAZ. Operating experience for feedwater sparger brackets is discussed in this Appendix; however, there are no BWRVIP-48 examinations required for feedwater sparger brackets based on their non-safety related function and existing ASME Section XI examinations (see Table 2-1 for additional discussion).
G.4.4 Summary of Inspections Section G.4.4.1 summarizes the inspection history of RPV attachment welds in the U.S. BWR fleet. As of 2020, baseline inspection of all attachment welds using a detailed visual examination technique (either MVT-1 or EVT-1) are complete and all attachment weld types except jet pump riser brace attachment welds have had the entire weld population reinspected at least once.
Section G.4.4.2 summarizes inspection coverage attained for all bracket types. All bracket types are accessible and high coverage values are reported. Overall, these data indicate relatively high confidence in the condition of welds within the fleet (i.e., it is very unlikely that significant cracks exist in service undetected).
G.4.4.1 Inspection History The initial version of BWRVIP-48 [G4] was published in 1998 and the NRC approved version, BWRVIP-48-A [G2], was subsequently published in 2004. The baseline and reinspection requirements contained in BWRVIP-48 and BWRVIP-48-A for core spray piping, jet pump primary riser brace and steam dryer support bracket attachment welds are shown in Table G-3 below.
Screening of BWRVIP-48, Rev. 2 in Accordance with Appendix C of NEI 03-08, Rev. 4, Document Screening G-8 Table G-3 BWRVIP-48, BWRVIP-48-A Baseline / Reinspection Requirements (EVT-1)
Figure G-3 provides a timeline overview of RPV attachment weld examination requirements and the status of U.S. fleet exams. The upper portion of the timeline in Figure G-3 illustrates the timing of requirements related to weld examinations and the lower portion of the timeline in Figure G-3 illustrates the timing of inspections required by BWRVIP-48 for the attachment welds being evaluated. Although BWRVIP-48 initially specified examination using MVT-1, a technique based on a one mil resolution capability, the BWRVIP committed to the use of EVT-1 for BWRVIP-48 examinations in 1999 and formalized this commitment within BWRVIP-48-A in 2004 [G2]. The initial EVT-1 standard required 1/2 mil resolution capability. In 2008, the BWRVIP published Revision 10 to BWRVIP-03 [G5] which made significant changes to the EVT-1 standard including 1) replacement of the 1/2 mil resolution standard with the ASME Code Section XI VT-1 standard of resolving 0.044 inch (1.12 mm) high characters, 2) requiring that camera motion (speed) during examination not exceed 0.5 inch/sec (12.7 mm/sec), and 3) limiting coverage credit to surf to the examination surface. All exams performed in outages occurring in late 2009 onward would have been performed using the more rigorous EVT-1 standard contained in BWRVIP-03, Revision 10 and subsequent versions of BWRVIP-03.
Content Deleted - EPRI Proprietary Information
Screening of BWRVIP-48, Rev. 2 in Accordance with Appendix C of NEI 03-08, Rev. 4, Document Screening G-9 Figure G-3 Timeline of BWRVIP-48 Revisions and Attachment Weld Examination Requirements All core spray piping bracket attachment welds have been examined at least three times, with a portion of the 1st reinspection and all of the 2nd reinspection performed using the updated EVT-1 standard initially published in late 2007 as part of BWRVIP-03, Revision 10. There are between 4 and 8 core spray piping brackets per unit and 33 U.S. BWR/3-6s that have vessel ID attached core spray piping brackets, with all of these plants contributing inspection data. As a result, over 450 detailed visual examinations of core spray piping attachment welds have been completed and all welds have been examined using EVT-1 based on the current standard.
All steam dryer support attachment welds have been inspected at least twice. Since there are between 4 and 6 steam dryer supports per unit and 35 U.S. BWRs contributing inspection data, over 300 detailed visual examinations of steam dryer support welds have been completed.
Almost all of the 1st reinspection exams would have been performed using the updated EVT-1 standard.
All jet pump riser brace attachment weld baseline exams are now complete, as are the first 6-year 25% reinspection. Section 3 of BWRVIP-266 [G20] indicates that there were 1,180 individual riser brace to vessel welds (RB-1 and RB-3 welds) in the operating U.S. fleet as of 2012. This entire population of welds has been inspected at least once using EVT-1 visual examination. At least 25% of the population has been re-inspected. All re-inspections and some baseline examinations would have been performed using the updated EVT-1 standard.
G.4.4.2 lity Table G-4 presents the results of a review of available coverage data for attachment weld inspections. All bracket types are found to be generally accessible for examination. Note that the results presented in Table G-4 represent coverage values reported for examinations completed using the updated EVT-1 standard initially included in BWRVIP-03, Rev. 10 that restricts coverage credit to examination surfaces where the camera is not more than 30 from perpendicular to the examination surface.
Content Deleted - EPRI Proprietary Information
Screening of BWRVIP-48, Rev. 2 in Accordance with Appendix C of NEI 03-08, Rev. 4, Document Screening G-11 Table G-5 Summary of RPV Attachment Weld Age-Related Degradation Attachment Weld Type Inspection Findings Core Spray Piping Primary and Supplemental Attachments There are no reported occurrences of cracking (due to either SCC or fatigue).
Steam Dryer Support Bracket Attachments SCC:
The only well documented occurrences of structurally significant SCC occur in one U.S. BWR/2 unit for which there are unique fabrication-related factors and two Swedish plants which contained slag inclusions and were repaired. See Section G.4.5.1.
Fatigue Cracking:
Only one occurrence reported. See Section G.4.5.2.
Wear:
A significant percentage of U.S. BWRs have reported occurrences of steam dryer support wear due to steam dryer motion. Identifications span multiple years, with some occurrences identified relatively recently. See Section G.4.5.3. Wear cannot be relevant to the attachment welds themselves. However, wear can provide an indication of dryer movement that could induce fatigue cracking in the associated attachment weld.
Feedwater Bracket Attachments Wear:
A significant percentage of U.S. BWRs have reported occurrences of wear occurring at the feedwater sparger support bracket. These instances are associated wear of the pin to end bracket connections and are not relevant to evaluation of attachment weld integrity. See Section G.4.5.4.
Jet Pump Riser Brace Primary Attachments There are no reported occurrences of cracking (due to either SCC or fatigue) for primary riser braces. BWRVIP-301 describes a secondary riser brace indication at Dresden Unit 3 at the riser brace arm to yolk weld [G6, Table A-8]. At Monticello there was an indication on the secondary brace to block weld that appeared to propagate into the block-vessel pad weld [G6, Table A-8].
G.4.5.1 Steam Dryer Support SCC Table 3-1 in BWRVIP-48, Revision 1 [G22] describes two instances of SCC associated with steam dryer support welds.
Swedish Plant Operating Experience The first item describes SCC occurring in two Swedish plants. In the first plant, linear indications were found by penetrant testing (PT). A weld repair was performed for one of the indications. Other indications were removed by grinding and reinspection of repaired areas did not show new cracking. In the second plant, the Alloy 182 attachment weld contained a series of
Screening of BWRVIP-48, Rev. 2 in Accordance with Appendix C of NEI 03-08, Rev. 4, Document Screening G-12 slag inclusions located through the length and depth of the weld. There were also star or crater cracks exposed to the surface. Reactor coolant entered through the crater cracks and contacted the slag, which resulted in a crevice filled corrosive material. Cracking propagated to join up with other slag inclusions, so that the subsurface length of the cracking was greater than the exposed length. The weld was ground to remove the cracking, but grinding revealed additional cracking. A boat sample confirmed that the cracking was interdendritic and therefore indicative of SCC. An ultrasonic technique was used to verify that no crack tips had penetrated into the low alloy steel of the vessel head.
US Plant Operating Experience The second instance of SCC included in BWRVIP-48, Revision 1, Table 3-1 [G22] relates to identification of flawed cladding in the vicinity of a steam dryer support weld. BWRVIP-48, Section 3.1 notes that this indication was found in the heat-affected region of the vessel clad adjacent to the bracket weld. However, the cracking identified was not related to the bracket weld and is one of many cracks in the stainless steel clad found throughout the vessel. The cladding SCC occurrences were associated with fabrication issues, specifically the result of weld metal dilution due to the nature of the clad operation used. One additional U.S. plant is also known to have similar occurrences of SCC occurring in stainless steel cladding applied to the RPV upper head. For all cases of SCC occurring in stainless steel cladding, cracking was contained in the stainless steel clad and there was no indication of crack extension into the low alloy steel vessel material [G21]. Subsequent examination results reported in the BWRVIP inspection database do not include any identification of new cracks or growth of existing cracks.
Subsequent to completion of BWRVIP-48-A, structurally significant SCC was identified in a U.S. BWR/2. SCC indications were reported in all four steam dryer support bracket attachments and were evaluated in detail [G19]. In three of the four attachments, multiple SCC indications were identified. All of the cracking of significance occurred in the stainless steel support bracket HAZ material and not the Alloy 182 attachment weld itself. The location of cracking is notable since this plant is one of a select number of early vintage plants for which it is likely that the bracket attachments were welded to the RPV prior to PWHT7. The PWHT applied to relieve stress in the vessel itself resulted in thermal sensitization of the 304SS base metal. This condition is termed furnace sensitization and is known to result in high susceptibility to SCC in austenitic stainless steel base metal. As this unit is the only BWR/2 remaining in operation, this condition is likely a unique situation. For other U.S. BWRs, the cracking does not raise new or increased concerns for at least two reasons:
- 1) No cracking was detected in the Alloy 182 attachment weld.
- 2) UT examination indicated that the cracks remained relatively shallow. Figure G-4 illustrates the cracking identified in the C support bracket in 2014 by both EVT-1 and UT. This result is consistent with predictions based on weld residual stress evaluation that stresses in the center region of the bracket are compressive, tending to cause cracks to slow and eventually arrest as they grow deeper into the bracket. Section G.4.6.3 provides a summary of conclusions reached from an evaluation of bracket weld residual stresses. Figure G-4 is a 7 See EPRI NP-7139-D [G9, Section 5, page 5-1], which identifies plants constructed prior to 1968 as potentially having furnace sensitized bracket attachments.
Screening of BWRVIP-48, Rev. 2 in Accordance with Appendix C of NEI 03-08, Rev. 4, Document Screening G-13 representative example of the SCC observed in one bracket for the purpose of illustrating the nature of the SCC observed.
Note: The End of Interval flaw sizes assume conservative crack growth in the depth and length direction for the purpose of ensuring adequate structural margins. The region shaded in dark red should not be treated as the expected flaw size.
Figure G-4 Example of Steam Dryer Support Attachment SCC Indications [G19]
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Screening of BWRVIP-48, Rev. 2 in Accordance with Appendix C of NEI 03-08, Rev. 4, Document Screening G-16 details, the primary difference between the various configurations is weld material. One group of plants have most bracket welds fabricated using Alloy 182, while the other set of plants have the welds fabricated using stainless steel weld metal (Type 308, 308L or 309L). One exception to this statement is that some steam dryer holddown bracket welds made by Babcock and Wilcox (B&W) and Chicago Bridge and Iron (CBI/CBIN) were performed using low alloy steel weld metal. The relative SCC susceptibility of these weld materials will be discussed separately below.
Alloy 182 Attachment Welds Although there has been no occurrence of SCC in Alloy 182 bracket attachment welds to date, Alloy 182 exposed to the BWR NWC environment is known to be susceptible to SCC in both laboratory testing and in operating plants. Field observations of SCC began in 1984 [G8]. Since that time, significant cracking has been reported in nozzle butters, safe end-to-nozzle welds, instrument penetration J-groove welds, access hole cover welds, and shroud support welds.
Alloy 182 also exhibits significant SCC CGRs in laboratory studies [G29][G36]. Based on this experience, it is concluded the bracket welds fabricated from Alloy 182 weld metal are at least nominally susceptible to SCC. Review of EPRI NP-7139-D [G9] indicates that only reactor vessels constructed by Combustion Engineering have bracket welds fabricated from Alloy 182.
Although the SCC events reported to date have been associated with NWC conditions, nickel-base alloys are known to have some SCC susceptibility even in low electrochemical corrosion potential (ECP) environments and to have long initiation periods [G37].
Stainless Steel Attachment Welds (308 / 309 or 308L / 309L)
Bracket attachment welds made with stainless steel filler metal represent a different case from the nickel alloy welds with respect to SCC susceptibility. Materials having a duplex cast stainless steel microstructure have shown significant resistance to SCC in the BWR environment.
This has generally been attributed to the presence of ferrite intermixed with the predominately austenitic structure [G10]. The ferrite breaks up continuous austenite-austenite grain boundary pathways for crack propagation. In addition, the solubility of carbon in ferrite is significantly greater than in austenite. The presence of ferrite limits the amount of carbon available at austenite boundaries to form chromium carbides which causes chromium depletion in the material matrix. It has been observed that the presence of as little as 3% to 4% ferrite is sufficient to substantially increase SCC resistance [G11]. Most, if not all, stainless steel weld metal used to fabricate the bracket welds may be expected to contain at least that much ferrite because ferrite is also essential to prevent hot cracking of stainless steel weld deposits. This was recognized by the NRC in Regulatory Guide 1.31 and ASME Section III which both require a minimum of 5% (or 5 FN) ferrite to prevent hot cracking. The presence of at least some ferrite in stainless steel weld deposits as an SCC deterrent is supported by the fact that no spontaneous SCC initiation has been observed in stainless steel welds in BWR piping or internals. The only exceptions are rare instances of cracking in weld surfaces heavily cold worked by machining.
Even in these rare instances, crack initiation occurred in the adjacent stainless steel HAZ material. Heavy grinding of bracket attachment welds would only be plausible in the case of defects that required extensive repairs and associated cracking would have most likely been manifested earlier in plant life. Consequently, there is no evidence that that the stainless steel weld metal used to fabricate the bracket welds should be considered susceptible to SCC. This position is consistent with the BWRVIP position taken for other internals components in NRC
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Screening of BWRVIP-48, Rev. 2 in Accordance with Appendix C of NEI 03-08, Rev. 4, Document Screening G-18 For the stainless steel attachment welds, the arrangement is similar with a stainless steel weld buildup pad or cladding applied to the vessel wall. This results in the HAZ on the vessel side of the attachment weld being in a stainless steel weld deposit. The discussion above on SCC susceptibility of the stainless steel attachment welds remains generally applicable to these stainless steel HAZs even though the weld buildup pads were subjected to post weld heat treatment. Further, performance of stainless steel cladding and weld buildup in BWRs has been excellent. Defects identified in cladding have been generally attributed to fabrication issues and a limited number of instances have reported as SCC; however, these flaws are self-limiting due to the lack of driving force and have not propagated into the vessel or attachment weld [G13].
Environment Water Chemistry The various attachment welds are exposed to different water chemistry environments depending on the elevation in the reactor vessel. There is also some variation depending on whether NWC or some version of HWC is applied in the individual plant. The distinct water chemistry regimes can be described as follows:
Upper reactor vessel (approximately above the top of the shroud) - In this region of the reactor vessel, the welds are exposed to a blend of feedwater and the drains coming from the steam separators and the steam dryer, or for the steam dryer hold-downs, steam exiting the dryer. The reactor water or steam at these elevations is very oxidizing and a high ECP (electrochemical corrosion potential) exists. ECP reduction is not possible by any form of hydrogen water chemistry. The environment is conducive to SCC during normal power operation, regardless of HWC technology implementation.
Downcomer region between the shroud and vessel wall (NWC) - The only brackets in this region are the jet pump riser braces and the surveillance capsule holder brackets. Under NWC conditions the water in this region remains a mix of feedwater and drains from the steam separators and dryer. The environment is conducive to SCC during normal power operation under NWC conditions.
Downcomer region between the shroud and vessel wall (HWC) - Implementation of effective HWC through either moderate HWC (HWC-M) or HWC with noble metal catalyst addition is capable of reducing surface ECP to values sufficiently low to mitigate SCC.
Effective HWC implementation provides similar reductions in CGR for stainless steels and nickel-base alloys. There is also evidence that SCC initiation becomes much less likely
[G39]. Finally, in these regions, the potential for crack extension into the low alloy steel vessel is extremely unlikely, even in the case of chloride transient conditions [G38].
Irradiation Effects The attachment welds in the downcomer region (jet pump riser brace and surveillance capsule holder) are exposed to some fast neutron irradiation being located near the core mid-plane.
Neutron irradiation of stainless steel causes material property changes that can result in increased susceptibility to SCC. However, even for these brackets, typical fast neutron fluence associated with extended operations (operation beyond 60 years) will remain at least an order of magnitude below the nominal threshold fluence for consideration of irradiation effects, i.e., 5x1020 n/cm2 (E > 1 MeV). This conclusion is supported by recent RPV evaluations that included estimation
Screening of BWRVIP-48, Rev. 2 in Accordance with Appendix C of NEI 03-08, Rev. 4, Document Screening G-19 of 80-year vessel ID peak fluence values [G45]. Consequently, neutron irradiation is not considered to be a significant factor in promotion of SCC in attachment welds.
During steady state operation, the only applied load to the attachment welds is the dead weight of the component being supported. The stresses associated with dead weight are minimal in most cases. For the steam dryer support, deadweight loads can become somewhat significant; however, these welds have a large shear area which reduces the applied stress. The main stress applied to the welds is weld residual stresses. An assessment of weld residual stress is provided below.
Review of numerous experimental and analytical weld residual stress (WRS) studies [G27, G28, G29, G30, G31, G32, G33] conducted on flat plates, piping, and round posts, and considering both single and double bevel weld preparations provide valuable insight into the anticipated WRS distributions and trends for the RPV attachments, as a function of section thickness.
Considering the uncertainties associated with the present state of the art in analytical and experimental WRS methods, multiple independent studies were reviewed for this work. The objective of the review was to determine whether general trends can be identified across multiple studies rather than attempting to extract specific WRS values. If the same trends are supported by a majority of the studies reviewed, then the trend is considered to be credible. Both single bevel and double bevel weld types were included in the study since the data did not show substantial variation between weld preparation designs [G27, G28, G29, G30]. These trends assist in understanding the relative sensitivity of various BWR RPV bracket designs to SCC:
both likelihood of crack initiation and flaw tolerance. Crack initiation and flaw tolerance are affected by parameters in addition to WRS; however, when WRS characteristics are taken into consideration with additional relevant parameters such as bracket dimensions, other bracket loads, materials of fabrication, and environment, then insight into relative risk associated with bracket degradation caused by SCC can be gained. The two stress directions relevant for discussion of SCC are the:
- 1. Stress component acting to open a crack oriented parallel to the longitudinal axis of the weld, and the
- 2. Stress component acting to open a crack that is oriented transverse to the longitudinal axis of the weld.
These stresses are referred to as Parallel and Transverse in the context of this discussion. Figure G-7 illustrates the orientation of SCC for which each stress direction is relevant.
Screening of BWRVIP-48, Rev. 2 in Accordance with Appendix C of NEI 03-08, Rev. 4, Document Screening G-20 Figure G-7 Schematic of SCC Orientation Important observations from the studies reviewed for this work are [G32]:
Parallel stresses An increase in surface stress magnitude (typically in the surface opposite to the final weld bead) is observed when increasing the thickness of the attachment.
This increase in the surface stress magnitude usually makes the zone of material that experiences tensile stresses larger which tends to shift the location at which the WRS is a minimum towards the mid-thickness of the section.
This behavior can be explained by the increased constraint imposed by a thicker plate.
All cases except the t < 0.5 inch (12.7 mm) thin wall piping case exhibit a central compressive area with a minimum stress value which seems to be not strongly influenced by member thickness.
Transverse stresses The stress tends to be tensile through the section thickness, in all cases.
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Screening of BWRVIP-48, Rev. 2 in Accordance with Appendix C of NEI 03-08, Rev. 4, Document Screening G-22 occurrences early in service life, followed by a decline to low rates of occurrence at longer operating times. This trend continues to be consistent with field observations of SCC in BWR components. Observations that can be made from review of the BWRVIP inspection database include:
Many cracks were identified within the first 10 years of operation (or at the time of initial inspections using high resolution visual or volumetric techniques), followed by a declining number of new crack occurrences with additional operating time.
For many components, virtually no new observations of cracking are being reported with continued operation.
For reactor internals, most of the SCC occurrences are associated with locations that are likely to have fabrication issues related to fit-up (i.e., these are welds made in the field and often are final assembly welds requiring the introduction of cold bending or draw bead welding to achieve component alignment). [G20, G40].
Most newly identified cracking indications in recent years can be attributed in some way to improved NDE capabilities or procedures. There is little evidence of any ongoing trend of new SCC initiation.
To date, SCC has not been identified in any weld material associated with any RPV bracket attachment weld. It is reasonable to conclude that a sudden increase in cracking probability is extremely unlikely.
The only possible exception to the trends described above would be Alloy 182 weldments.
Recent leakage events associated with SCC of Alloy 182 / 600 instrument penetrations at Limerick [G41], Quad Cities [G42] and Peach Bottom [G43] suggest that late occurring SCC initiation or crack progression is plausible in the NWC environment for nickel-base alloys. These events are particularly notable since these flaws most likely were not old flaws that were only recently detected. Through-wall cracking in a pressure-retaining component is much more likely to be detected closer to the time of initiation and active growth than cracking in reactor internals components.
G.4.7 Fatigue Cracking Susceptibility Cyclic loading as a result of steam dryer movement could potentially result in fatigue cracking in a steam dryer support bracket weld. There has been one instance of what is hypothesized to be fatigue cracking in the Susquehanna steam dryer support bracket described in Section 4.5.2.
Since the cyclic loading could only be caused by steam dryer assembly movement, it was likely caused by FIV. This hypothesis has not been validated through metallurgical evaluation of the failed support bracket. Substantial dryer motion is often indicated by support wear without any trend of fatigue cracking identified. As a result, the potential for this form of cracking to occur in a steam dryer support attachment weld is judged to be unlikely unless there have been significant changes in operating conditions and / or evidence of significant dryer assembly motion causing wear of a support.
There have also been a number of instances of fatigue cracking attributable to FIV in welds attaching the jet pump riser to the riser brace (welds RS-8 and RS-9). As such, it is reasonable to consider the potential for FIV to promote cracking in the riser brace RPV attachment weld.
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Screening of BWRVIP-48, Rev. 2 in Accordance with Appendix C of NEI 03-08, Rev. 4, Document Screening G-25 G.4.9.1 SCC is the only degradation mode that requires a detailed occurrence assessment.8 Table G-7 presents an assessment of SCC occurrence in a matrix format. The evaluation considers material susceptibility to SCC occurence, the effect of environment (NWC vs. HWC), and weld residual stress. Scores are assigned to each box from 1 to 5, with 1 corresponding lower likelihood of SCC occurence (bottom left) and 5 corresponding to higher likelihood of SCC occurence (top right). The SCC susceptibility categories are based on material type and environment. The notes following Table G-7 provide the rationale for the scores. See Sections G.4.6.1 and G.4.6.2 for additional discussion regarding the effect of material and environment on SCC susceptibility.
The residual stress rankings are based on the observed trend that thicker brackets are associated with higher surface tensile weld residual stresses that would promote SCC initiation. Since the steam dryer support brackets are substantially thicker than other bracket configurations evaluated, this bracket type is categorized as higher residual stress. There are two types of core spray piping brackets: cylindrical post and rectangular plate. The cylindrical post design is larger and will lead to higher residual stress and is categorized as moderate residual stress.
This result is conservatively applied for all core spray line bracket locations. Finally, the jet pump riser brace is a thin rectangular beam that has the thinnest cross section of the attachment welds evaluated and relatively lower surface tensile weld residual stress than other bracket configurations addressed in this evaluation. This bracket type is categorized as lower residual stress. See Section G.4.6.3 for additional discussion.
8 The potential for crack occurrence due to fatigue is highly dependent on location, plant design, and operating history. Section G.5.2 summarizes new one-time inspection requirements that address crack occurrence due to fatigue.
Screening of BWRVIP-48, Rev. 2 in Accordance with Appendix C of NEI 03-08, Rev. 4, Document Screening G-26 Table G-7 SCC Occurrence Assessment Matrix G.4.9.2 Table G-8 presents an assessment of SCC severity in matrix format. The severity assessment considers FIV potential and flaw tolerance. Scores are assigned to each box from 1 to 5, with 1 corresponding to lower severity (bottom left) and 5 corresponding to higher severity (top right). For this assessment, severity is defined as the potential for a crack initiated by SCC to grow to a size that could result in a loss of bracket integrity.
Brackets having lower flaw tolerance are associated with smaller critical crack sizes than brackets having higher flaw tolerance. As a result, there would be greater significance associated with SCC occurrence in a bracket with lower flaw tolerance, since less SCC growth would be required to reach a critical flaw size. The flaw tolerance categorizations are based on the results of a generic flaw tolerance assessment summarized in Section G.4.8. This assessment concluded that in terms of relative flaw tolerance, a moderate flaw tolerance categorization is appropriate for jet pump riser brace attachment welds and a higher flaw tolerance categorization is appropriate for the steam dryer support and core spray piping bracket attachment welds.
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Screening of BWRVIP-48, Rev. 2 in Accordance with Appendix C of NEI 03-08, Rev. 4, Document Screening G-27 Although material type (stainless steel vs. Alloy 182) does have some effect on flaw tolerance, the impact is not significant enough to justify further refinement of the flaw tolerance scores in Table G-8.
Higher FIV loads can also result in a reduction in critical flaw size, thus increasing the significance of an SCC occurrence. With regard to the FIV potential of the brackets evaluated, the core spray piping brackets experience little to no FIV and are categorized as lower FIV potential. Steam dryer support brackets and jet pump riser braces have some potential for FIV to occur and are categorized as moderate FIV potential.
Table G-8 SCC Severity Assessment Matrix G.4.9.3 Results A risk number is calculated for each bracket type and material combination using the occurrence and severity assessment scores from Table G-7 and Table G-8, respectively. The risk number is calculated as: (Occurrence Score) x (Severity Score). Table G-9 shows the resulting risk numbers for bracket type and material, sorted based on descending risk number. These results form a basis for the revisions to the periodic inspection program described in Section G.5.
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Screening of BWRVIP-48, Rev. 2 in Accordance with Appendix C of NEI 03-08, Rev. 4, Document Screening G-43 G8.
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G26.
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EPRI TR-100651, Residual Stress Analysis in BWR Pressure Vessel Attachments, June 1992.
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Proposed Relief Request Associated with Reactor Pressure Vessel Nozzle Repairs, Limerick Generating Station, Unit 2, May 15, 2017, NRC ADAMS ML17135A423.
G42.
Additional Information Supporting Relief Request I5R-11 Associated with Reactor Pressure Vessel Penetration N-11B Repair, Quad Cities Nuclear Power Station, Unit 2, December 20, 2013, NRC ADAMS ML13358A401.
G43.
Proposed Relief Request Associated with N-16A Reactor Pressure Vessel Instrument Nozzle Repairs, Peach Bottom Atomic Power Station, Unit 2, November 4, 2020, NRC ADAMS ML20309B020.
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BWRVIP-329: BWR Vessel and Internals Program, Updated Probabilistic Fracture Mechanics Analyses for BWR RPV Welds to Address Extended Operations. EPRI, Palo Alto, CA: 2019. 3002015930.
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Overview of 1999/2000 Shroud Support Cracking in Tsruruga 1, Presentation to U.S.
NRC by General Electric Nuclear Energy (Tom Caine), June 30, 2000. NRC ADAMS ML003745473.
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Revision Details (BWRVIP-48, Revision 2)
H-3 32 BWRVIP-99-A, BWR Vessel and Internals Project, Crack Growth Rates in Irradiated Stainless Steels in BWR Internal Components 33 BWRVIP-190 Revision 1: BWR Vessel and Internals Project, Volume 1: BWR Water Chemistry Guidelines - Mandatory, Needed and Good Practice Guidance.
34 EPRI Materials Degradation Matrix, Revision 4 3002013781.
BWRVIP-233, Revision 2: BWR Vessel and Internals Project: Updated Evaluation of Stress Corrosion Crack Growth in Low Alloy Steel Vessel Materials in the BWR Environment BWRVIP-62-A (2018 Update): BWR Vessel and Internals Project, Technical Basis for Inspection Relief for BWR Internal Components with Hydrogen Injection
Revision Details (BWRVIP-48, Revision 2)
Table H-1 Revision Details BWRVIP-48, Revision 2 Revision Source of Revision Description of Revision Implementation Section 1.2, "Objective and Scope" 10CFR50 Appendix B Revised 1 0CFRS0 Appendix B requirements to denote that Revision 2 was not prepared in accordance with the requirements of 1 0CFRS0 Updated content.
Appendix B.
Section 2, "Vessel ID Attachment Weld Design and Editorial Clarify safety and non-safety components.
Susceptibility Information" Minor edits.
Table 2-1 Content added.
Discussion updated to identify steam dryer assemblies as being within the scope of license renewal for most all U.S. BWRs.
Table 2-2, "Weld Configurations for Jet Pump Riser Brace, Content added.
Plants that are no longer in operation have been removed. Figure Core Spray Piping, and Steam Dryer Support references have been added to show plants with supplemental core Attachments" spray piping brackets. Table 2-2 now provides information only for jet pump riser braces, core spray piping bracket attachment welds and Updated content.
steam dryer support attachment welds. A new Table 2-3 was added to address configuration details for the remaining bracket types.
Figures 2-9 through 2-12 BWRVIP-15 New figures added for welded plate core spray piping brackets and for steam dryer support brackets.
Added content.
Section 2.1.1, "Jet Pump Riser Brace Attachments" Editorial Removed old plants and updated the discussion to more clearly describe primary vs. secondary riser braces and the various riser brace Editorial changes.
designs.
Section 2.1.2, "Core Spray Piping Bracket Attachments" Content added.
Removed old plants and updated the discussion to more clearly describe supplemental core spray piping bracket designs.
Added content.
Section 2.1.3, "Steam Dryer Support Brackets" Content added.
New section added to describe steam dryer support bracket configurations.
Added content.
Section 2.2, "Susceptibility Factors" Appendix G, This section was updated throughout to bring the content up to date BWRVIP-181-A based on the conclusions reached in the risk evaluation documented in Updated content.
Appendix G. The sec susceptibility assessment was updated based on the current state of the art and understanding of SCC mechanisms related to RPV attachment locations. Added discussion of vibratory loading on steam dryer support brackets per BWRVIP-181-A.
Section 3.1, "Inspection Methods" Editorial This section was moved from Section 3.1.2 in Revision 1. Other sections have been moved or deleted.
Moved.
H-4
Revision Details (BWRVIP-48, Revision 2)
Table H-1 (continued)
Revision Details BWRVIP-48, Revision 2 Revision Source of Revision Description of Revision Implementation Section 3.2.1, "Periodic Inspection" BWRVIP-62 New sections have been added to provide revised inspection BWRVIP-41, Rev. 4-A requirements based on the results of a risk assessment documented in Added sections on HWC-Based Inspection Strategy Appendix G. Additionally, HWC inspection criteria are added for jet pump and FIV Susceptibility Categorization.
riser braces.
Section 3.2.2, "One-Time Examination" BWRVIP-41, Rev. 4-A, Added new set of requirements for one-time inspections to address BWRVIP-139, Rev. 1-A, fatigue cracking due to FIV.
Added new Section.
Appendix G Section 3.2.3, "Supplement Examination" Editorial This section was reorganized from Section 3.2.1 in Revision 1.
Requirements for scope expansion examinations were strengthened.
Added new Section.
Section 3.2.4, "Scope Expansion" Appendix G Scope expansion requirements have been updated based on conclusions in Appendix G.
Updated guidance.
Table 3-1, "Bracket Attachment Inspection Editorial Historic inspection information (Table 3-1 in Revision 1) has been Recommendations" Appendix G removed.
Comprehensive update to inspection Bracket attachment inspection recommendations have been updated recommendations.
based on conclusions in Appendix G.
Section 4.2, "Load Combinations" BWRVIP-303 Added reference to BWRVI P-303.
Added reference.
Section 4.3, "Flaw Evaluation Approach" Editorial Updated wording to be more concise and added a reference to a more current ASME Section XI edition.
Content updated / editorial.
Section 5, "References" Editorial New References for BWRVIP-303 and NEI 03-08, Rev. 4 have been added.
Added references.
Appendix E, Section 4.5 Editorial Wording adjusted to not directly quote BWRVIP-06, Rev. 1-A.
Editorial change.
Appendix G, "Screening of BWRVIP-48, Rev. 2 in NEI 03-08, Rev. 4 This appendix documents a comprehensive risk assessment that Accordance with Appendix C of NEI 03-08, Rev. 4, supports release of BWRVIP-48, Rev. 2 to members for immediate "Document Screening" implementation.
New content.
End of Revisions H-5