CNRO-2020-00021, Response to Request for Additional Information Regarding Relief Requests GG-ISI-020 and RBS-ISI-019

Response to Request for Additional Information Regarding Relief Requests GG-ISI-020 and RBS-ISI-019

ML20314A125
Person / Time
Site:	Grand Gulf, River Bend
Issue date:	11/09/2020
From:	Gaston R Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML20314A124	List: CNRO-2020-00021, Response to Request for Additional Information Regarding Relief Requests GG-ISI-020 and RBS-ISI-019
References
CNRO-2020-00021
Download: ML20314A125 (24)

Text

Entergy Operations, Inc.

1340 Echelon Parkway Jackson, MS 39213 Tel 601-368-5138 Ron Gaston Director, Nuclear Licensing Proprietary Information - Withhold from Public Disclosure Under 10 CFR 2.390 The balance of this letter may be considered non-proprietary upon removal of Attachments 1 and 2.

10 CFR 50.55a(z)(1)

CNRO-2020-00021 November 9, 2020 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

Response to Request for Additional Information Regarding Relief Requests 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

References:

1. Entergy Operations, Inc. (Entergy) letter to 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)," (ADAMS Accession Number ML20160A032), dated June 8, 2020

2. NRR E-mail, "Entergy Fleet (Grand Gulf and River Bend) - Official RAIs for RR Associated with Updating ASME Code RR SEs with NRC-Approved Revision of BWRVIP Guidelines (GG-ISI-020 and RBS-ISI-019)

(EPID L-2020-LLR-0079))," (ML20260H442), dated September 8, 2020 In Reference 1, Entergy Operations, Inc. (Entergy) requested a revision to U.S. Nuclear Regulatory Commission (NRC) Safety Evaluations (SEs) concerning the use of the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Guidelines in lieu of specific ASME Code requirements on Reactor Pressure Vessel internals and components inspection (i.e., Relief Request GG-ISI-020 for Grand Gulf Nuclear Station, Unit 1 (GGNS) and Relief Request RBS-ISI-019 for River Bend Station, Unit 1 (RBS)).

CNRO-2020-00021 Page 2 of 3 By email correspondence dated September 8, 2020 (Reference 2), the NRC staff informed Entergy that they had reviewed the relief requests and had determined that additional information would be required to complete the review.

The Enclosure to this letter provides the additional information requested by the NRC, and includes the following Attachments.

Attachment 1 provides BWRVIP-48, Revision 1, Section E.4, "Qualitative Risk Assessment for Extension of the Core Spray Piping Bracket Attachment Weld Examination Interval in Rev. 1 to BWRVIP-48." This attachment contains information proprietary to EPRI, which is supported by an Affidavit signed by EPRI, the owner of the information.

Attachment 2 provides BWRVIP-48, Revision 1, Table F.1, "Revision Details BWRVIP-48, Revision 1." This attachment contains information proprietary to EPRI, which is supported by an Affidavit signed by EPRI, the owner of the information.

Attachment 3 provides a non-proprietary, redacted version of BWRVIP-48, Revision 1, Table F.1.

Attachment 4 provides a non-proprietary, redacted version of BWRVIP-48, Revision 1, Section E.4.

Attachment 5 provides the EPRI Affidavit in support of Attachments 1 and 2. The Affidavit sets forth the basis on which the information may be withheld from public disclosure by the NRC and addresses, with specificity, the considerations listed in paragraph (b)(4) of Section 2.390 of the Commission's regulations.

There are no regulatory commitments contained in this letter.

Should you have any questions or require additional information, please contact Ron Gaston, Director, Licensing at (601) 368-5138.

Respectfully, Ron Gaston RWG/jls

CNRO-2020-00021 Page 3 of 3

Enclosure:

Response to Request for Additional Information Regarding Relief Requests GG-ISI-020 and RBS-ISI-019 Attachments to Enclosure

1. BWRVIP 48, Revision 1, Table F.1 (Proprietary)

2. BWRVIP 48, Revision 1, Section E.4 (Proprietary)

3. Redacted Version BWRVIP 48, Revision 1, Table F.1 (Non-proprietary)

4. Redacted Version BWRVIP 48, Revision 1, Section E.4 (Non-proprietary)

5. EPRI Affidavit dated October 27, 2020, in support of Attachments 1 and 2 cc:

NRC Region IV Regional Administrator NRC Senior Resident Inspector - Grand Gulf Nuclear Station, Unit 1 NRC Senior Resident Inspector - River Bend Station, Unit 1 NRC Project Manager - Grand Gulf Nuclear Station, Unit 1 NRC Project Manager - River Bend Station, Unit 1

ENCLOSURE CNRO-2020-00021 Grand Gulf Nuclear Station, Unit 1 River Bend Station Unit 1 Response to Request for Additional Information Regarding Relief Requests GG-ISI-020 and RBS-ISI-019

CNRO-2020-00021 Enclosure Page 1 of 3 RESPONSE TO REQUESTS FOR ADDITIONAL INFORMATION By letter dated June 8, 2020, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20160A032), Entergy Operations Inc. (Entergy) submitted a proposed revision to the NRC approved alternatives (i.e., Relief Request (RR) GG-ISI-020 for Grand Gulf Nuclear Station, Unit 1 (GGNS) and RR RBS-ISI-019 for River Bend Station, Unit 1 (RBS)).

In these two RRs, Entergy requested to revise these NRC-authorized alternatives to allow the use of BWRVIP 18, Revision 2 A, "BWR Vessel and Internals Project BWR Core Spray Internals Inspection and Flaw Evaluation Guidelines" (NRC Approval Letter and Safety Evaluation (SE)

ADAMS Accession No. ML16011A), BWRVIP-41, Revision 4-A, "BWR Vessel and Internals Project BWR Jet Pump Assembly Inspection and Flaw Evaluation Guidelines" (NRC Approval Letter and SE ADAMS Accession No. ML18130A024),and BWRVIP-48, Revision 1, "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 application for the subject relief requests (RRs) and concluded that additional information is required for complete evaluation.

RAI-1

Provide a summary of the revised changes between BWRVIP-48-A and BWRVIP-48, Revision 1. Based on these changes, justify why BWRVIP-48, Revision 1 may be implemented as an acceptable basis for inspection of the applicable RPV interior attachments welds without prior NRC staff approval of the methods in BWRVIP-48, Revision 1.

Entergy Response:

The technical changes to BWRVIP-48, Revision 1 are listed in Table F-1, "Revision Details BWRVIP-48, Revision 1," of that document. Attachment 1 provides Table F-1. The principal change in BWRVIP-48, Revision 1 is the update to the periodic inspection strategy of the Core Spray Piping bracket attachment in Table 3-2, "Bracket Attachment Inspection Recommendations." This revision changes the inspection interval from 100% every four refueling cycles to 100% every 10 years. The technical justification for changing the inspection interval is documented in BWRVIP-48, Revision 1, Appendix E, Section E.4, "Qualitative Risk Assessment for Extension of the Core Spray Piping Bracket Attachment Weld Examination Interval in Rev. 1 to BWRVIP-48." Attachment 2 provides Section E.4.

RAI-2

Provide Information regarding how the licensee is meeting the NRC staffs Conditions on use of OLNC additions as issued in the NRC staffs SE for BWRVIP-62-A (e.g., conditions on ECP values and Platinum content of the test coupons).

CNRO-2020-00021 Enclosure Page 2 of 3 Entergy Response:

Entergy is able to satisfy all the BWRVIP-62-A, "BWR Vessel and Internals Project, Technical Basis for Inspection Relief for BWR Internal Components with Hydrogen Injection," Table 3-8, "Implementation steps for category 3a and 3b (NMCA) plants," for hydrogen water chemistry inspection relief at RBS. The station recently completed used fuel sampling to verify the catalyst loading is greater than 0.1 g/cm2.

Entergy is able to satisfy all the BWRVIP-62-A, Table 3-8 implementation steps for hydrogen water chemistry inspection relief at GGNS, except for the catalyst loading. Entergy expects to complete used fuel sampling to verify catalyst loading at GGNS in 2021.

Entergy will not implement BWRVIP-41, Rev.4-A at GGNS unless vessel internal artifact sampling can verify the catalyst loading is greater than 0.1 g/cm2. The earliest opportunity to implement BWRVIP-41, Rev.4-A at GGNS would be during the 2022 refueling outage in-vessel inspections.

RAI-3

(a)

Provide information regarding the type of inservice inspections (ISI) and the frequencies of ISI (consistent with ASME Code,Section XI criteria) that have been performed to date, on the aforementioned RPV internal lower plenum components.

(b)

Justify the basis for using the I&E bases in BWRVIP-47-A to inspect the RPV internal lower plenum components for the requested ISI interval given that the BWRVIP report does not include any re-inspection criteria for RPV internal lower plenum components.

Entergy Response:

(a) Entergy has completed the BWRVIP-47, "BWR Vessel and Internals Project, BWR Lower Plenum Inspection and Flaw Evaluation Guideline," baseline inspections of the Control Rod Guide Tubes (CRGT) at GGNS and RBS. No other baseline inspections were required by the BWRVIP to be completed. All subsequent inspections of components in the Lower Plenum have been opportunistic, meaning that if there is bottom head access as a result of normal refueling outage activities, the ASME Section XI code requires performance of visual inspection of accessible areas in the region, to the extent practical. BWRVIP-47, Section 3.2.5 also includes the requirement to perform inspections when there is access to the lower plenum due to maintenance activities that are not part of normal refueling outage activities. In such cases, licensees will perform a visual inspection to the extent practical. No indications have been identified from any of the Lower Plenum inspections at either GGNS or RBS. The following summarizes the CRGT inspections and other Lower Plenum inspections that have been completed for both GGNS and RBS.

CNRO-2020-00021 Enclosure Page 3 of 3 GGNS Spring 1998 - 34 CRGT-1 exams completed with no indications Spring 2001 - 12 guide tubes. 12 FS/GT-ARPIN-1 and CRGT-1 Fall 2002 - CRGT-2 & 3 (10 places). FS/GT-ARPIN-1 (2 places)

Spring 2008 - Completed remaining baseline exams on 10 CRGTs Spring 2016 - VT-3 examinations of control rod drive housings and in-core housings accessible during the Lower Plenum Below Core Plate Examinations through cells 28-29 & 32-33 RBS Spring 2003 - Completed baseline exams on 15 CRGTs Spring 2008 - Viewed to the extent possible the CRDHs within view through JP19 & 20 to examine the cap to tube assembly weld (CRDH-1), tube to tube assembly weld (CRDH-4), and tube to reactor pressure vessel weld (CRDH/RPV-1). Best effort attempt to examine the ICH within view through JP19

& 20 on the in-core housing to reactor pressure vessel weld (ICH/RPV-1)

Spring 2017 - VT-3 examinations of control rod drive housings and in-core housings accessible during the Lower Plenum Below Core Plate Examinations through cell 24-25 (b) The following components should be inspected to the extent practical as made accessible, including maintenance activities not part of normal refueling activities: all penetration welds to the Vessel Lower Head, CRD Housing assemblies, CRD Guide Tubes, In-core housing Dry Tubes and support hardware, Shroud Support Legs and Vessel Lower Head for foreign material. Determination of reinspection intervals and frequencies will be based on flaw evaluation results. No indications have been identified from any of the Lower Plenum inspections at either GGNS or RBS; therefore, no reinspection intervals or frequencies have been required.

ATTACHMENT 3 to ENCLOSURE CNRO-2020-00021 Grand Gulf Nuclear Station, Unit 1 River Bend Station Unit 1 BWRVIP 48, Revision 1, Table F.1 (Non-proprietary)

Revision Details (BWRVIP-48, Revision 1)

F-3 Table F-1 Revision Details BWRVIP-48, Revision 1 Revision Source of Revision Description of Revision Implementation In Abstract, clarified inspection area of interest to include attachment welds and heat affected zones on the vessel side of the welds.

Need to clarify inspection areas of interest for vessel attachment welds Added the following to Abstract: The recommended inspections contained herein apply to the attachment welds and the heat affected zones the vessel side of the welds. The inspection recommendations for the brackets themselves (HAZs on the bracket side) are provided in the Clarify applicable Sections of ASME B-N-2 ASME Section XI Section 1.2 revised to specify applicable Sections of ASME Code Section XI B-N-2 Add Section 1.3 this revision as NEI-03-08 was not published at the time of publication of BWRVIP-48-A.

NEI-03-08, Revision 3 Added new Section 1.3:

1.3 Implementation Requirements In accordance with Nuclear Energy Institute (NEI) 03-version of this report has been reviewed by the NRC, this report revision was evaluated using the NEI 03-08, Appendix C document screening process. Using this process, it was determined that the report may be generically released for implementation without NRC approval. This determination was based a step 2d of the screening evaluation. Appendix E provides the details of the screening result and basis.

Change plant name from WNP-2 to Columbia Editorial As plant designation has changed, plant name changed from WNP-2 to Columbia in Table 2-2.

Revise Section 3.1 to reflect the current state of knowledge regarding vessel ID attachment weld inspection history.

BWRVIP-301 BWRVIP-251 BWRVIP-266 Added section headings to separate historical OE (associated with inspections performed before initial development of BWRVIP-48) from more recent OE. Added new sub-sections to describe recent inspection trends based on the sources listed and to provide an updated performance summary.

10623290 BWRVIP 2020-090, Att. 5 CNRO-20-00021 Enclosure - Attachment 3

Revision Details (BWRVIP-48, Revision 1)

F-4 Table F-1 Revision Details BWRVIP-48, Revision 1 (Continued)

Revision Source of Revision Description of Revision Implementation requirements.

BWRVIP requirement 3.2.2, 4.1, 4.2, Table 3-2 footnote 1 and Table 4-1 footnote 1.

In Table 3-2, clarified inspection area of interest to include attachment welds and heat affected zones on the vessel side of the welds.

Need to clarify inspection areas of interest for vessel attachment welds In Table 3-

-1 inspections of jet pump riser brace and core spray bracket to vessel ID attachment welds recommended in BWRVIP-48, Revision 1 was clarified vessel side of the Revise Table 3-2 to reflect revised periodic inspection interval for Core Spray bracket to vessel ID welds and HAZs Qualitative Risk Assessment performed in accordance with NEI 08, Revision 3, Appendix C (see Appendix E)

Add SI unit conversion EPRI requirement Added SI conversion from Kips to KN on page 4-3 Update definition of VT-1 in Section 3.1.2 ASME Section IX Changed to: VT-1 is defined using the ASME Section XI criteria for the

-service inspection program.

Update definition of EVT-1 in Section 3.1.2 BWRVIP-03 Rev 19 Changed to: Enhanced VT-1 (EVT-1) is defined in the latest revision to BWRVIP- 03.

Add/update References Editorial Updated References 2 and 5 and added new References 10, 11 and 12.

on BWRVIP-basis for removal.

BWRVIP position on LR appendices implemented in revisions to other I&E guidelines BWRVIP-and replaced with explanation of basis for removal.

Appendix on BWRVIP-explanation of basis for removal.

BWRVIP position on LR appendices implemented in revisions to other I&E guidelines.

BWRVIP-Add documentation of screening to determine whether BWRVIP-48, Revision 1 can be generically released for implementation without NRC approval.

NEI-03-08, Revision 3, Appendix C Added new Appendix E: Screening of BWRVIP-48, Revision 1 in Accordance with NEI.

Add Appendix F Record of Revisions (BWRVIP-48 Revision 1)

BWRVIP Practice Added new Appendix F Record of Revisions (BWRVIP-48 Revision 1)

End of Revisions 10623290 Inserted revised inspection guidance in Table 3-2 as follows:

Content Deleted - EPRI Proprietary Information

]

[

CNRO-20-00021 Enclosure - Attachment 3

ATTACHMENT 4 to ENCLOSURE CNRO-2020-00021 Grand Gulf Nuclear Station, Unit 1 River Bend Station Unit 1 BWRVIP 48, Revision 1, Section E.4 (Non-proprietary)

Screening of BWRVIP-48, Revision 1 in Accordance with Appendix C of NEI 03 -

E-7 E.4 Qualitative Risk Assessment for Extension of the Core Spray Piping Bracket Attachment Weld Examination Interval in Rev. 1 to BWRVIP-48 E.4.1 Introduction The examinations specified for core spray piping bracket to reactor vessel wall welds in BWRVIP-48 [E4] were deemed to be needed primarily because of concerns regarding potential susceptibility of RPV ID attachment welds to stress corrosion cracking (SCC). The scope of inspection in BWRVIP-48 of this weld is confined to the groove weld of the bracket to the weld buildup pad or cladding on the vessel wall and the associated heat affected zone (HAZ) induced by the groove weld in the weld buildup pad or cladding. The following discussion is limited to material in this inspection scope.

At the time BWRVIP-48 was initially issued, 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. About twenty years have elapsed since the initial issue of BWRVIP-48 and it is reasonable to revisit the specified inspection intervals based on the current state of knowledge regarding performance in the field and understanding of the progression of SCC in BWRs. This qualitative change in risk evaluation addresses a change from E.4.2 Historical Performance of Core Spray Piping Attachment Welds and HAZs BWRVIP-48 [E4] (published in 1998) and BWRVIP-48-A [E2] (published in 2004) required BWRVIP-48-A required examination by EVT-1. In Revision 10 of BWRVIP-03, published in December 2007, the requirements for EVT-1 were updated to require character height resolution requirements and to impose limits on camera travel speed and camera angle. Examinations performed to this standard have been demonstrated capable of detecting very small indications. Based on that implementation date, it is known that all core spray piping bracket attachment welds and HAZs in US BWRs currently remaining in operation have been inspected at least once using this technique. Additionally, at least one, and potentially two, prior exams have been performed for all locations, either by the initial EVT-1 standard (the examination method specified in BWRVIP-48-A) or by MVT-1 (examination technique specified in the original version of BWRVIP-48). There are between 4 and 8 core spray piping brackets per unit and 34 US BWR units with core spray bracket to vessel ID welds that have performed inspections as required by BWRVIP-48. There have been over 450 detailed visual examinations completed to date, with most of these exams using EVT-1. No reportable conditions have been identified in these inspections [E5]. In addition, a general review of examination coverage was performed. EVT-1 inspection coverage for the primary core spray brackets is typically in the range from 80 to 100% and for supplemental core spray brackets from 60-95% (Note: These coverage values are associated with the current, more conservative interpretation of coverage for EVT-1 exams). Given that examination coverage is good and iíç Content Deleted - EPRI Proprietary Information Content Deleted - EPRI Proprietary Information BWRVIP 2020-090, Att. 3

[

[

]

]

CNRO-20-00021 Enclosure - Attachment 4

Screening of BWRVIP-48, Revision 1 in Accordance with Appendix C of NEI 03 -

E-8 exams have not detected any indications, the performance of these welds to date is considered to be excellent with no SCC found.

E.4.3 SCC Susceptibility Discussion SCC Factors As with all SCC-related phenomena, it is necessary to consider the three essential factors for cracking to occur. These are: 1) a conducive environment, 2) material susceptibility, and 3) sufficient tensile stress. The absence or significant reduction of any one of these three factors is sufficient reason to conclude the risk of SCC is very low. The discussion below is specific to the core spray bracket welds based on the materials of construction, weld design and application, and the location of the welds in the reactor vessel:

Environment:

With respect to environment, the core spray piping brackets are located in the mixing plenum or upper down comer region of reactor vessel at an elevation just below the feed water spargers. In this region of the reactor vessel, the bracket welds are exposed to a blend of feedwater and the drains coming from the steam separators and the steam dryer.

The reactor water at this elevation is oxidizing and ECP (electrochemical corrosion potential) reduction is unlikely by any form of hydrogen water chemistry (HWC). It must therefore be concluded that the core spray bracket welds are exposed to an environment conducive to SCC during normal power operation, regardless of HWC technology implementation.

Material Susceptibility:

The materials of construction and the general configuration of the welds are shown in Figures 2-6 and 2-7 of this document (BWRVIP-48-Rev. 1), along with supplemental bracket welds present in some plants, which are shown in Figure 2-8. For the purposes of this evaluation, the supplemental bracket welds are included within the discussion of the stainless steel bracket welds shown in Figure 2-7. Table 2-2 lists figures in BWRVIP-15

[E6] that show the specific configuration applicable to each plant. Although there are some differences in weld preparation details, the primary difference between the two configurations is weld material. One group of plants have core spray piping bracket welds fabricated using Nickel Alloy 182, while the other set of plants have the welds fabricated using stainless steel weld metal (Type 308 or 308L). In general, Alloy 182 weld metal deposits have been found to be susceptible to SCC in the field and laboratory studies, while stainless steel weld deposits generally have not. The relative SCC susceptibility of these two weld materials will be discussed separately below.

Tensile Stresses:

During steady state operation, the only applied load to the bracket welds is the dead weight of the core spray pipe. The stresses associated with dead weight are minimal. The main stress applied to the weld is weld residual stress. An assessment of weld residual stress is provided below.

iíç CNRO-20-00021 Enclosure - Attachment 4

Screening of BWRVIP-48, Revision 1 in Accordance with Appendix C of NEI 03 -

E-9 Weld Metal SCC Susceptibility Alloy 182 weld metal exposed to the BWR environment has been seen to be susceptible to SCC in both laboratory testing and in operating plants. Field observations of SCC began in 1984 [E7].

Since that time, significant cracking has been reported in nozzle butters, safe end-to-nozzle welds, and shroud support structure welds. Based on this experience, it must be concluded the core spray piping bracket welds fabricated from Alloy 182 weld metal are at least nominally susceptible to SCC. Review of Table 2-2 and supporting document NP-7139-D [E8] show that only reactor vessels constructed by Combustion Engineering have core spray piping bracket welds fabricated from Alloy 182.

Core spray piping bracket 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 [E9]. The ferrite serves two significant purposes. It breaks up continuous austenite-austenite grain boundary pathways for crack propagation, and the solubility of carbon in ferrite is significantly greater. Therefore, the presence of ferrite limits the amount of carbon available at austenite boundaries to form the chromium carbides that generate chromium depletion. It has been seen that the presence of as little as 3 to 4% ferrite is sufficient to render the weld deposit essentially immune to SCC [E10]. Most, if not all, stainless steel weld metal used to fabricate the core spray piping 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 deposit 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 (excepting rare instances of cracking in weld surfaces heavily cold worked by machining). There is no heavy machining of core spray piping bracket welds. Consequently, it may be concluded that the stainless steel weld metal used to fabricate the core spray bracket welds can be considered not susceptible to SCC. This position is consistent with the BWRVIP position taken for other internals components in NRC approved BWRVIP guidelines. A directly applicable example of this case is BWRVIP-18, Revision 2-A [E3] where the examination requirements specifically state that the inspection requirements for locations where stainless steel welds are used to join fully austenitic stainless steel base materials are applicable only to the HAZs associated with the stainless steel weld base materials (and not the welds themselves).

Similarly, within BWRVIP-03, Rev. 19 [E11], the areas of interest defined for thick stainless steels specifically include only the toe of the weld and one-half inch of the adjacent base material on each side of the weld. The stainless steel weld itself is excluded.

HAZ (Vessel side) Material SCC Susceptibility The bracket attachment weld on the vessel side of the joint is applied to a weld buildup pad on the vessel wall or in some cases to the cladding. For the Combustion Engineering plants, the weld buildup pads are Alloy 182. Therefore, the HAZs on the vessel side of the weld are in iíç CNRO-20-00021 Enclosure - Attachment 4

Screening of BWRVIP-48, Revision 1 in Accordance with Appendix C of NEI 03 -

E-10 Alloy 182 and have the same SCC susceptibility as the groove welds. For the stainless steel bracket 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 groove weld being in stainless steel weld deposit. The discussion above on SCC susceptibility of the stainless steel bracket attachment welds is equally applicable to these stainless steel HAZs.

Stress State (Weld and HAZ on Vessel Side)

Since no cracking has been observed to date irrespective of weld filler type or HAZ material on the vessel side [E5] even though the Alloy 182 bracket welds/HAZs must be considered SCC susceptible material and operate in a very aggressive environment, rationalization of a lack of cracking in what, for many plants, is more than 40 years of operation must focus on stress state.

As noted above, applied loads on the core spray bracket welds/HAZs are minimal during steady state operation. That leaves weld residual stress as the likely source of any sustained tensile stress in the weld. Weld residual stresses can be substantial for large internal structures such as shrouds or the shroud supports as a result of weld size and restraint imposed by adjacent thick base metal. These residual stresses have been judged to be sufficient to explain the observed SCC in core shroud stainless steel weld heat affected zones and shroud support Alloy 182 welds. While the weld is being made, the weld deposit tries to shrink as it cools down after solidification. However, this natural shrinkage is constrained by the adjacent base metal. This occurs in part because the base metal parts being joined are not free to move, being constrained by configuration and thickness. This results in residual stresses that, in some cases, approach the yield strength of the material. Such stresses are more than sufficient to promote initiation and growth of SCC in Alloy 182 when exposed to an oxidizing reactor water environment.

However, fabrication of the Alloy 182 core spray piping bracket welds present a substantially different case. For the piping bracket design using Alloy 182, the diameter of the bracket is typically only about 1.5 inches (38.1 mm). Also, and very importantly, the bracket is completely unrestrained as the weld is being made. That is, the end of the bracket was free to move as shrinkage of the weld deposit occurred. As can be seen in Figure 2-6 of BWRVIP-48-A [E2]

(Figure 2.13.2.23 of BWRVIP-15 [E6]), the weld preparation of the bracket was designed as a two-sided weld to be made from above and below the bracket post. In order to complete the weld with the post level within the specified tolerance, it was necessary for the welder to apply one or two passes from the top side. Weld shrinkage would pull the end of the post upward. At that point the welder would have back gouged the unfused root to sound metal and applied weld passes from the underside to pull the end of the post back level. Welding would have proceeded in this alternating fashion until the weld groove was filled to the specified dimensions and the post was level within the drawing tolerance. Throughout this process, the post was free to move with the weld shrinkage such that the final residual tensile stresses were limited. Based on the results of the repeated inspections over the last 20 years noted above (in many cases accounting for more than 40 years of operation), it is reasonable to conclude the residual tensile stresses in the core spray piping bracket welds are insufficient to promote SCC.

E.4.4 SCC Risk Assessment As described above, the overall risk of SCC occurring in a given component depends on convergence of the three essential conditions, i.e. susceptible material, conducive environment, and tensile stress. The relative risk in turn is dictated to a large extent by the level or intensity of iíç CNRO-20-00021 Enclosure - Attachment 4

Screening of BWRVIP-48, Revision 1 in Accordance with Appendix C of NEI 03 -

E-11 these three conditions. Over the history of SCC in BWRs, it has been observed that not all components in a given system experience cracking even though they are fabricated from the same material and operate in the same water chemistry environment. Further, it has been determined that the probability of cracking as a function of time is more appropriately described by a Weibull distribution than a normal distribution [E12]. Frequently, the available data indicate the cracking probability distribution is characterized by a rapid peak in occurrences early in service life followed by a decline to low probabilities at longer operating times. This trend continues to be consistent with field observations of SCC in BWR components. 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 trend with additional operating time. For many components, virtually no new observations of cracking are being reported with continued operation. Additionally, most recent new cracking indications can be attributed in some way to improved NDE capabilities or procedures. There is little evidence of any ongoing trend of new SCC initiation. Assuming these trends are also applicable to the core spray piping bracket welds where no cracking has been observed, it is reasonable to conclude that a sudden increase in cracking probability is extremely unlikely.

E.4.5 Qualitative Risk Assessment Summary and Conclusions Reinspection intervals for the core spray piping bracket welds were established with the initial issue of BWRVIP-48 [E4]. At that time, SCC of reactor internals was in the early stages of detection and evaluation. There were virtually no reliable inspection data for the core spray piping bracket welds to establish a likelihood and frequency of cracking and the welds were judged to be at least moderately susceptible to SCC based on materials of construction, local environment, and likely stress state [E8]. Consequently, a conservative approach was used to set baseline inspection and reinspection requirements. However, since the initial issue of BWRVIP-48, over 450 inspections of core spray bracket welds and HAZs on the vessel side have been performed. Despite the early predictions of moderate risk of SCC, no cracking has been identified to date. The stainless steel bracket welds and SS HAZs on the vessel side are fabricated of a material that has been shown to be highly resistant to SCC irrespective of residual stress and environment. Although the Alloy 182 bracket welds and Alloy 182 HAZs on the vessel side must be acknowledged to be fabricated from a susceptible material, no cracking of these welds has been observed even though the welds/HAZs are exposed to an aggressive local environment. The lack of cracking can most likely be attributed to a lack of sufficient tensile stresses. Relatively low weld residual stresses are related to the size, sequencing, and lack of restraint in these specific welds. Observation of SCC occurrences in other BWR internals and piping continue to demonstrate a declining trend in cracking probability with continued operation.

Regarding CS piping to vessel ID attachments, BWRVIP-06 [E16] states

. However, since each unit has between 4 and 8 CS piping to vessel ID attachments, there is significant redundancy to mitigate a single attachment failure. With the proposed change in the inspection interval, and there have been no reportable indications in 40 years (>1300 years of fleet reactor operation). It is therefore concluded that Content Deleted - GE-Hitachi Proprietary Information - Refer to last paragraph of Section 2.4.2 of BWRVIP-06 [E16] for quoted information Content Deleted - EPRI Proprietary Information Content Deleted - EPRI Proprietary Information

]

[

]

[

[

CNRO-20-00021 Enclosure - Attachment 4

Screening of BWRVIP-48, Revision 1 in Accordance with Appendix C of NEI 03 -

E-12 an

, the revised aging management guidance can be released for implementation by the BWRVIP utilities without NRC approval.

E.6 References E1 Document Screening, February 2017.

E2 BWRVIP-48-A: BWR Vessel and Internals Project, Vessel ID Attachment Weld Inspection and Flaw Evaluation Guidelines. EPRI, Palo Alto, CA: 2004. 1009948.

E3 BWRVIP-18, Revision 2-A: BWR Vessel and Internals Project, Core Spray Piping Inspection and Flaw Evaluation Guidelines. EPRI, Palo Alto, CA: 2016. 3002008089 E4 BWRVIP-48: BWR Vessel and Internals Project, Vessel ID Attachment Weld Inspection and Flaw Evaluation Guidelines. EPRI, Palo Alto, CA: 1998. TR-108724.

E5 BWRVIP-301: BWR Vessel and Internals Project, BWRVIP Inspection Trends, 2016 Update. EPRI, Palo Alto, CA: 2017. 3002008102.

E6 BWRVIP-15:BWR Vessel and Internals Project, Configurations of Safety-Related BWR Reactor Internals. EPRI, Palo Alto, CA: 1996. TR-106368 E7 BWR alloy 182 stress Corrosion Cracking Societe Francaise d'Energie Nucleaire - SFEN, 75 - Paris (France); (v.1-2).

1175 p; 2002; p. 55-67; Fontevraud 5 International symposium; Fontevraud Royal Abbey (France); 23-27 Sep 2002.

E8 Reactor Pressure Vessel Attachment Welds: Degradation Assessment. EPRI, Palo Alto, CA: 1991. NP-7139-D.

E9 N. Hughes et. al., Intergranular Stress-Corrosion Cracking Resistance of Austenitic Stainless Steel Castings, ASTM STP756 Stainless Steel Castings, 1982.

E10 G. Nakayama, K. Yoshida, and M. Akashi, Effects of Carbon and Delta Ferrite on Stress Corrosion Cracking Susceptibility of Type 309 Weld Metal in Simulated BWR Environment, NACE Corrosion 93, Paper 171.

iíç NEI-03-08 Revision 3, Guideline for the Management of Mateirals Issues, Appendix C, Content Deleted - EPRI Proprietary Information is entirely reasonable and appropriate given the current state of knowledge. This conclusion is applicable to both the stainless steel and Alloy 182 core spray piping bracket attachment welds and stainless steel/Alloy 182 HAZs on the vessel side.

E.5 NEI 03-08 Document Screening Conclusion

]

CNRO-20-00021 Enclosure - Attachment 4

Screening of BWRVIP-48, Revision 1 in Accordance with Appendix C of NEI 03 -

E-13 E11 TR-105699-R19 (BWRVIP-03) Revision 19: BWR Vessel and Internals Project, Reactor Pressure Vessel and Internals Examination Guidelines. EPRI, Palo Alto, CA: 2016.

3002008095.

E12 BWRVIP-315: BWR Vessel and Internals Project, Reactor Internals Aging Management for Extended Operations. EPRI, Palo Alto, CA: 2019.3002012535.

E13 Generic Aging Lessons Learned (GALL) Report Final Report. (NUREG-1801, Revision 2), December 2010 E14 Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report.

(NUREG 2191), July 2017 E15 BWRVIP-94NP, Revision 2: BWR Vessel and Internals Project, Program Implementation Guide. September 2011, EPRI Report 1024452 E16 BWRVIP-06, Revision 1-A: BWR Vessel and Internals Project, Safety Assessment of Reactor Internals. December 2009. EPRI Report 1019058.

CNRO-20-00021 Enclosure - Attachment 4

ATTACHMENT 5 to ENCLOSURE CNRO-2020-00021 Grand Gulf Nuclear Station, Unit 1 River Bend Station Unit 1 EPRI Affidavit dated October 27, 2020, in support of Attachments 1 and 2

2020-090 BWR Vessel & Internals Project (BWRVIP)

October 29, 2020 Todd Sherman Entergy Services, LLC 220 Hickory Glen Madison, MS 39110

Subject:

Transmittal of EPRI Proprietary Affidavit to Entergy

Dear Mr. Sherman:

The purpose of this letter is to transmit the EPRI proprietary affidavit for transmittal of the following document to the NRC:

Entergy Nuclear, Inc.

CNRO-2020-00021: Response to Request for Additional Information Regarding Relief Requests GG-ISI-020 and RBS-ISI-019 Excerpts from BWRVIP-48, Revision 1 Please note that the document referenced above contains EPRI proprietary information. A letter requesting that the report be withheld from public disclosure and an affidavit describing the basis for withholding this information are provided as Attachment 1. The Proprietary version (Attachment 2) and the Non-Proprietary version (Attachment 3) of the BWRVIP-48 Revision 1 Section E.4 and Proprietary version (Attachment 4) and Non-Proprietary version (Attachment 5) of the BWRVIP-48 Revision 1 Table F-1 Revision Details are also provided.

If you have any questions on this subject, please contact me by telephone at 724-288-4043 or by e-mail at npalm@epri.com.

Sincerely, Nathan Palm EPRI, BWRVIP Program Manager

BWRVIP 2020-090, Attachment 1

-=~121 1 ELECTRIC POWER a=,-

RESEARCH INSTITUTE Ref EPRI Docket No. 99902016 October 27, 2020 Document Control Desk Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

Request for Withholding of the following Proprietary Information Included in:

Entergy Nuclear, Inc.

CNRO-2020-00021: Response to Request for Additional Information Regarding Relief Requests GG-ISl-020 and RBS-ISl-019 Excerpts from BWRVIP-48, Revision 1 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 report 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 NRG review of the enclosed submittal to the NRC by Entergy. The Proprietary Information is not to be divulged to anyone outside of the NRG 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.

If you have any questions about the legal aspects of this request for withholding, please do not hesitate to contact me at (704) 595-2885. Questions on the content of the Report should be directed to Nathan Palm of EPRI at (724) 288-4043.

Sincerel4/

Attachment(s)

Together... Shaping the Fu ture of Electr icity 1300 Wes1 W. T. Horris 6oulevord, Chorlo11e, NC 28262*8.S.SO USA

• 704..59..5.2732

• Mobile 704..490.26.53

• nwilm3liuri1@epri.com

-=~121 1 ELECTRIC POWER a=,1-RESEARCH INSTITUTE AFFIDAVIT RE:

Request for Withholding of the Following Proprietary Information Included In:

Entergy Nuclear, Inc.

CNRO-2020-00021: Response to Request for Additional Information Regarding Relief Requests GG-ISl-020 and RBS-ISl-019 Excerpts from BWRVIP-48, Revision 1 I, Steve Chengelis, being duly sworn, depose and state as follows:

I am the Director of Plant Support 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 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.

c.

The information sought to be withheld is considered to be proprietary for the following reasons. 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 the substantial economic resources required of EPRI to develop the Proprietary Information.

d.

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 or use; and (2) Is the subject of efforts that are reasonable under the circumstances to maintain its secrecy."

e.

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 Propretary 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.

f.

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 under 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 Chengeli~

(State of North Carolina)

(County of Mecklenburg)