Text

Application for Alternative Request - Extension of Steam Generator Primary Inlet Nozzle Dissimilar Metal Weld Inspection Interval (Volumetric Examination)
	ML22069B117
Person / Time
Site:	Summer
Issue date:	03/10/2022
From:	Lawrence D; Dominion Energy South Carolina
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML22069B116	List: ML22069B117;
References
	22-025 LTR-SDA-21-023-NP, Rev 0
	Download: ML22069B117 (48)
	v • d • e

PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 Dominion Energy South Carolina, Inc.

5000 Dominion Boulevard, Glen Allen, VA 23060

=--= Dominion Dominion Energy.com ~ Energy~

March 10, 2022 Attn: Document Control Desk Serial No.: 22-025 U.S. Nuclear Regulatory Commission NRA/YG: RO Washington, DC 20555-0001 Docket No.: 50-395 License No.: NPF-12 DOMINION ENERGY SOUTH CAROLINA VIRG IL C. SUMMER NUCLEAR STATION UNIT 1 APPLICATION FOR ALTERNATIVE REQUEST - EXTENSION OF STEAM GENERATOR PRIMARY INLET NOZZLE DISSIMILAR METAL WELD INSPECTION INTERVAL (VOLUMETRIC EXAMINATION)

Pursuant to the provisions of 10 CFR 50.55a(z)(1 ), Dominion Energy South Carolina, Inc.

(DESC), hereby submits the attached one-time extension request for approval to use an alternative to the volumetric examination In-Service Inspection (ISi) requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV)

Code Section XI for the Virgil C. Summer Nuclear Station (VCSNS) Unit 1 Class 1 PWR pressure retaining dissimilar metal (DM) piping and vessel nozzle butt welds.

Specifically, DESC proposes a one-time extension of the VCSNS steam generator (SG) primary inlet nozzle dissimilar metal weld (DMW) inspection interval from 5 years to nominally 9 years for the volumetric examination. No changes are requested to the visual examination requirements applicable to the SG inlet nozzle DMW. DESC is seeking to defer performance of the volumetric examination from the current scheduled date in the spring of 2023 to the fall of 2027.

DESC has determined that the proposed alternative would provide an acceptable level of quality and safety.

A detailed description of the proposed alternative request RR-4-26, including basis for use, is provided in Attachment 1 to this letter. DESC requests NRC review and approval of this request by March 10, 2023, to support planning for the Spring 2023 refueling outage (RF-27) with a 60-day implementation period. , the technical report which supports this submittal, contains information proprietary to Westinghouse Electric Company, LLC ("Westinghouse"). Enclosure 1 is supported by an affidavit (Enclosure 2) signed by Westinghouse, the owner of the information. 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 NRC's regulations. The non-proprietary version of the technical report is provided in Enclosure 3.

Accordingly, it is respectfully requested that the information which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10 CFR Section 2.390 of the NRC's regulations.

Enclosure 1 contains information that is being withheld from public disclosure under 10 CFR 2.390. U on se aration from Enclosure 1, this a e is decontrolled.

Serial No.22-025 Docket No. 50-395 Page 2 of 3 Correspondence with respect to the copyright or proprietary aspects of the information provided in Enclosure 1 and/or Enclosure 3 of this letter or the supporting Westinghouse Affidavit should reference CAW-21-5230 (Enclosure 2) and should be addressed to Anthony J. Schoedel, Manager, eVinci Licensing & Configuration Management, Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, Pennsylvania 16066.

In accordance with 10 CFR 50.91, a copy of this request, with Enclosures (w/o Enclosure 1), is being provided to the designated South Carolina State Official.

Should you have any questions related to this submittal, please contact Yan Gao at (804) 273-2768.

RespectfullYJ Douglas C. La Vice Presiden - lear Engineering and Fleet Support Commitments made in this letter: None.

Attachment:

1. Alternative Request RR-4-26 Description

Enclosures:

1. LTR-SDA-21-023-P, Revision O (Proprietary)

2. Westinghouse Affidavit for Withholding Proprietary Information

3. LTR-SDA-21-023-NP, Revision O (Non-proprietary)

Serial No.22-025 Docket No. 50-395 Page 3 of 3 cc: U.S. Nuclear Regulatory Commission, Region II Marquis One Tower 245 Peachtree Center Avenue, NE Suite 1200 Atlanta, Georgia 30303-1257 Mr. G. Edward Miller NRC Senior Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 09 E-3 11555 Rockville Pike Rockville, Maryland 20852-2738 NRC Senior Resident Inspector V.C. Summer Nuclear Station Ms. Anuradha Nair-Gimmi (w/o Enclosure 1)

Bureau of Environmental Health Services South Carolina Department of Health and Environmental Control 2600 Bull Street Columbia, SC 29201 Mr. G. J. Lindamood (w/o Enclosure 1)

Santee Cooper - Nuclear Coordinator 1 Riverwood Drive Moncks Corner, SC 29461

Serial No.22-025 Docket No. 50-395 Attachment 1 Alternative Request RR-4-26 Description Extension of Steam Generator Primary Inlet Nozzle Dissimilar Metal Weld Inspection Interval (Volumetric Examination)

Virgil C. Summer Nuclear Station (VCSNS} Unit 1 Dominion Energy South Carolina, Inc. (DESC}

Serial No.22-025 Docket No. 50-395 Attachment 1 RR-4-26: Page 1 of 7 TABLE OF CONTENTS 1.0 ASME CODE COMPONENT(S) AFFECTED ................................................................... 2 2.0 APPLICABLE CODE EDITION AND ADDENDA ............................................................. 2 3.0 APPLICABLE CODE REQUIREMENT AND PROPOSED ALTERNATIVE ...................... 3 4.0 REASON FOR THE REQUEST .................. ........................................... .............. ..... .... .. 3 5.0 BASIS FOR THE PROPOSED ALTERNATIVE REQUEST .................. ........................... 4 6.0 DURATION OF PROPOSED ALTERNATIVE. ................................................................. 6 7.0 PRECEDENTS ................................................................................................................ 6

8.0 REFERENCES

..... ........................................ ....... .............................. ... ........................... 6

Serial No.22-025 Docket Nos. 50-395 Attachment 1 RR-4-26: Page 2 of 7 Alternative Request RR-4-26 Description Extension of Steam Generator Primary Inlet Nozzle Dissimilar Metal Weld Inspection Interval (Volumetric Examination)

Pursuant to 10 CFR 50.55a(z)(1)

Acceptable Level of Quality and Safety 1.0 ASME CODE COMPONENT(S) AFFECTED Table 1 - ASME Code Components Affected Elements Description

~SME Code Class Code Class 1 References ~SME Code Case N-770-5

~lass 1 PWR Pressure Retaining DM Piping and Vessel Nozzle Butt Welds Examination Category Containing Alloy 82/182 Inspection Item ~-2 1-4100A-31 (OM)- 'A' SG Inlet Nozzle to Safe End DMW Components 1-4200A-28 (OM)- 'B' SG Inlet Nozzle to Safe End DMW 1-4300A-29 (DM) - 'C' SG Inlet Nozzle to Safe End DMW Table 2 - Materials [8.1]

Elements Description SG Inlet Nozzle SA-508 CL. 3a Weld Alloy 82 filler material Inlay Alloy 152 filler material Stainless Steel Safe End SA-336 CL. F316LN 2.0 APPLICABLE CODE EDITION AND ADDENDA American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, Rules for lnservice Inspection of Nuclear Power Plant Components, 2007 Edition through 2008 Addenda [8.2].

VCSNS is currently in the Fourth lnservice Inspection Interval (January 1, 2014, to December 31, 2023). VCSNS will enter the Fifth lnservice Inspection Interval on January 1, 2024.

Serial No.22-025 Docket No. 50-395 Attachment 1 RR-4-26: Page 3 of 7 3.0 APPLICABLE CODE REQU IREMENT AND PROPOSED ALTERNATIVE ASME Code Case N-770-5, Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated With UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities,Section XI, Division 1 (8.3]:

Inspection Item A-2: Unmitigated butt weld at Hot Leg operating temperature < 625 °F (329 CC) requiring visual examination each refueling outage and volumetric examination every five (5) years in accordance with ASME Code Case N-770-5 Table 1.

Per 10 CFR 50.55a(g)(6)(ii)(F), holders of operating licenses of pressurized water reactors shall implement the requirements of ASME Code Case N-770-5.

Pursuant to 10 CFR 50.55a(z)(1 ), DESC proposes to extend the VCSNS steam generator (SG) primary inlet nozzle dissimilar metal weld (DMW) inspection interval from 5 years to nominally 9 years for the volumetric examination only. No changes are requested to the visual examination requirements of the SG inlet nozzle DMWs. This proposal is a one-time extension request. The VCSNS SG inlet nozzle DMWs were examined in the fall of 2018 using inside surface phased array ultrasonic volumetric examination and automated eddy current surface examination methods with no indications identified. The next scheduled volumetric examination for these welds is in the spring of 2023. DESC is seeking to defer the volumetric examination to the fall of 2027 (nominally 9 years).

4.0 REASON FOR TH E REQUEST The requested extension would allow coordination of the SG inlet nozzle DMW volumetric examinations with the SG outlet nozzle DMW volumetric examinations. This coordination would result in draining of the reactor coolant system to low levels and opening the SG manways once, instead of twice, thereby supporting nuclear, radiological, and industrial safety. The next scheduled volumetric examination for the SG outlet nozzle DMWs is in the fall of 2027.

Technical justification for deferral of the SG inlet nozzle DMW volumetric examinations is based on a proprietary VCSNS-specific crack growth rate analysis (8.4] which is provided in Enclosure 1 of this submittal (the non-proprietary version of the analysis [8.5] is provided in Enclosure 3 of this submittal.) The crack growth rate analysis, (8.4] or (8.5],

supports extension of the SG inlet nozzle DMW inspection interval (volumetric examination) to nominally 9 years while maintaining an acceptable level of quality and safety.

Serial No.22-025 Docket No. 50-395 Attachment 1 RR-4-26: Page 4 of 7 5.0 BASIS FOR THE PROPOSED ALTERNATIVE REQUEST VCSNS has three (3) Delta 75 replacement stream generators that were installed in the fall of 1994. The SGs were fabricated with factory welded stainless steel safe ends attached to the SG primary nozzles with Alloy 82 DMWs. The inside surface of the welds and adjacent base material were clad with resistant Alloy 152 during fabrication. Minor linear indications and defects were discovered during initial fabrication in two of the three inlet nozzle welds. All indications were removed by grind out, rewelded, and reinspected to acceptable conditions. The analysis considered in this proposed alternative bounds those repairs.

In the fall of 2018, ASME Section XI, Appendix VIII-qualified automated phased array ultrasonic examinations and ASME Section XI, Appendix IV-demonstrated automated eddy current examinations were performed on the inside surfaces of all three (3) inlet nozzle welds. No indications were observed. The use of both phased array ultrasonic testing (PAUT) and eddy current test (ECT) methods ensured that neither sub-surface flaws nor surface-breaking flaws were located within the inner 1/3 of the weld thickness.

In fact, for the entirety of the VCSNS replacement SGs service history, there have been no indications observed in the inlet nozzle or outlet nozzle DMWs.

The technical justification which demonstrates the acceptability of extending the SG inlet nozzle DMW inspection interval is contained in the VCSNS-specific crack growth rate analysis which is provided in Enclosure 1 (Enclosure 3 for non-proprietary version) of this submittal. The crack growth rate analysis demonstrates that, for the VCSNS SG inlet nozzle DMWs, it would take more than 8.48 effective full power years (EFPYs) (9 calendar years) for a postulated 0.065-inch-deep flaw (axial or circumferential) to grow to the maximum allowable end-of-evaluation flaw size.

The crack growth rate analysis in Enclosure 1 (Enclosure 3 for non-proprietary version) considers minimum detectable postulated flaws that could hypothetically propagate through the Alloy 152 inlay and the Alloy 82 DMW materials to the maximum allowable end-of-evaluation flaw size over the span of nominally 9 years. A postulated initial flaw depth of 0.065 inch (half the inlay weld thickness) was used for the flaw evaluation, and both axial and circumferential flaws were evaluated. The crack growth rate for the Alloy 82 DMW material was determined per MRP-115 (8.6] and is consistent with ASME Section XI, Appendix C. The maximum allowable end-of-evaluation flaw size was based on the ASME Section XI, Appendix C methodology. In accordance with ASME Section XI, a maximum allowable end-of-evaluation flaw size of 75% wall thickness was used.

The primary water stress corrosion cracking (PWSCC) crack growth rate for the Alloy 152 inlay material was based on determining the minimum factor of improvement (FOi) over the Alloy 182 PWSCC crack growth rate required to provide at least nominally 9 years of operation to reach the maximum allowable end-of-evaluation flaw size. Applying this methodology, the VCSNS-specific crack growth rate analysis used an FOi of 29 over the Alloy 182 PWSCC crack growth rate for the Alloy 152 inlay material. This FOi is used to justify extending the SG inlet nozzle DMW inspection interval to nominally 9 years from the last inspection which was performed in the fall of 2018.

Serial No.22-025 Docket No. 50-395 Attachment 1 RR-4-26: Page 5 of 7 Additional details of the VCSNS-specific parameters used in the crack growth rate analysis are described in Enclosure 1 (Enclosure 3 for non-proprietary version). The VCSNS-specific parameters include nozzle geometry and material properties, normal operating parameters, and applied stresses including piping loads and welding residual stresses. Assumptions are defined where applicable.

The applied FOi of 29 for the Alloy 152 inlay material is supported by extensive laboratory test data. In April 2010, the NRG Staff conducted a general confirmatory review of flaw tolerance evaluations of inlays as a mitigation technique for PWSCC concerns [8.7). The study considered a reactor coolant nozzle with an Alloy 82/182 DMW and an Alloy 52/152 inlay on the inside surface. While a FOi of 100 was used as the baseline case, the study determined that a FOi of 30 represented the 95th percentile of the laboratory data for Alloy 52/152 available at that time. More recently, in April 2019, NUREG/CR-7103, Volume 4 presented laboratory test data for an NRG-sponsored study of Alloy 52/152 crack growth rates from Pacific Northwest National Laboratory (PNNL), Argonne National Laboratory (ANL), General Electric Global (GEG), and Energy Environmental and Technological Investigations Centre (CIEMAT) [8.8). As shown in Enclosure 1 (Enclosure 3 for non-proprietary version), Figure 6-1, an FOi of 29 for the Alloy 152 inlay material bounds a significant amount (greater than 75%) of the laboratory test data. In January 2021, PNNL presented additional crack growth rate test data on Alloy 52/152 with 15%

cold forging [8.9). Also shown in Enclosure 1 (Enclosure 3 for non-proprietary version),

Figure 6-2, an FOi of 29 for the Alloy 152 inlay material bounds all the 15% cold forging data presented by PNNL. In addition to the crack growth rate test data, Enclosure 1 (Enclosure 3 for non-proprietary version), Section 6.2, provides a comparison of PWSCC initiation data for Alloy 52/152 materials versus Alloy 82/182 materials. The findings of the initiation test data indicated that Alloy 52/152 material will take significantly longer than Alloy 82/182 before any evidence of crack initiation is detected.

A collection of industry operating experience further supports the superior service history of Alloy 52/152 materials as compared to Alloy 82/182 materials. A review of service history, summarized in Enclosure 1 (Enclosure 3 for non-proprietary version), Section 6.3, has found that many Alloy 52/152 welds have been in operation at hot leg conditions longer than Alloy 82/182 welds with no evidence of PWSCC initiation or cracking. Further, as indicated above, since the installation and operation of the VCSNS replacement SGs in the fall of 1994, there have been no indications observed in either the inlet nozzle or outlet nozzle DMWs.

In summary, the use of an FOi of 29 for the Alloy 152 inlay material in the VCSNS-specific crack growth rate analysis supports extension of the SG inlet nozzle DMW inspection interval (volumetric examination) to nominally 9 years, until the fall of 2027 refueling outage. This proposal cites several studies which have produced laboratory test data that are bounded by an FOi of 29 for the Alloy 152 inlay material used in the VCSNS-specific crack growth rate analysis. The use of this proposed alternative will provide an acceptable level of quality and safety.

Serial No.22-025 Docket No. 50-395 Attachment 1 RR-4-26: Page 6 of 7 6.0 DURATION OF PROPOSED ALTERNATIVE This request is applicable until the VCSNS fall 2027 refueling outage. This includes the remainder of the VCSNS Fourth lnservice Inspection Interval (January 1, 2014, to December 31, 2023) and the beginning of the VCSNS Fifth lnservice Inspection Interval which will start on January 1, 2024.

7.0 PRECEDENTS Plant Relief Request NRC Safety Evaluation Approval Date ML15225A104 Point Beach Unit 2 ML16063A058 3/22/2016 ML15324A152 Point Beach Unit 2 ML19241A492 ML19339H747 12/13/2019

8.0 REFERENCES

8.1 1MS-07-460 Sheet 2, VCSNS Replacement Steam Generator Lower Shell Assy Weld Information Map (6146E68) 8.2 American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, Rules for lnservice Inspection of Nuclear Power Plant Components, 2007 Edition through 2008 Addenda 8.3 ASME Code Case N-770-5, Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities,Section XI, Division 1 8.4 Westinghouse LTR-SDA-21-023-P Revision 0, V. C. Summer Unit 1 Steam Generator Hot Leg Inlet Nozzle to Safe-End Weld PWSCC Growth Analysis, October 2021 8.5 Westinghouse LTR-SDA-21-023-NP Revision 0, V. C. Summer Unit 1 Steam Generator Hot Leg Inlet Nozzle to Safe-End Weld PWSCC Growth Analysis, October 2021 8.6 EPRI Materials Reliability Program: Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 82, 182, and 132 Welds (MRP-115), 2004 8.7 Evaluation of the Inlay Process as a Mitigation Strategy for Primary Water Stress Corrosion Cracking in Pressurized Water Reactors, Rudland, David L. et al, April 2010

Serial No.22-025 Docket No. 50-395 Attachment 1 RR-4-26: Page 7 of 7 8.8 U.S. NRC NUREG/CR-7103, Volume 4, Pacific Northwest National Laboratory Investigation of Stress Corrosion Cracking in Nickel-Base Alloys: Behavior of Alloy 152 and 52 Welds, April 2019 8.9 Update on sec Growth Rate Testing Research for Alloy 152(M)/52(M) Welds at PNNL. Presentation provided by PNNL at EPRI Alloy 690/52/152 PWSCC Research Collaboration Meeting, Toloczko, M., Olszta, M., Ziqing, Z., Bouffious, R., and Bruemmer, S., January 2021

Serial No.22-025 Docket Nos. 50-395 Enclos ure 2 Westin ghouse Affidavit for Withholding Proprietary Information Virgil C. Summer Nuclear Station {VCSNS) Unit 1 Dominion Energy South Carolin a, Inc. (DESC)

Serial No.22-025 Docket Nos. 50-395 Enclosure 2: Page 1 of 3 Westinghouse Non-Proprietaiy Clas~ 3 CAW-21-5230 Page 1 of3 CO~IMONWEALTI-1 OF PENNSYLVANIA:

COu'"NTY OF Bun.ER:

(1) L Anthony J_ Schoedel, ha,;e been specifically deleg,at.ed and authorized to apply for withholding and execute 1his Affidavii on b ehalf of Westinghouse Electric Company LLC (Westinghouse).

(2) I am ~ting the proprietary portions ofLJR-SDA-21--023-P. Re,i.sion Obe 'l\,i.tbheld from public disclosure under 10 CFR 2.390.

(3) I have penona1 blo"'iedge of the criteria :md prncedures utiliz.ed by Westinghouse in desigrurting information as a trade secret, prrvileged, or 3'l confidential commercial or financial information.

{4) PurnJant to 10 CFR 2390. the follo'll'ing is fumislled fur consideration by ~ Canmission in detennini.ng wbetl=- the infonnation sought to be withheld from public disclosure YJCuld be wit.hbeld.

(i) Toe infonnation sought to be withheld :from public di;domre is owned and bas been held in confidence by Westinghouse and is not cmtomarily disclO".aed to the public.

(u.) Toe infonnatiro sought to be Ri.tbheld is being t.r.uBmiited to the C-ommi~on in confidence and, to \Vestinghou.se's mowledg~ is not available in public sources.

(iii.) w..,.tin~hause notes that a showing of~ubstanlial harm is no longer :m applicable criterion for anal}'Zing whetbec a document should be \\i.thheld from _public dis closure. Nevertheless, public disclo = ofthB proprietary infoo:nationi,- likely to cause substantial hann to the competitfr-e position of Westinghouse because it would enhance the ability of campetitotS to pro1tide similar technical e\*aluation

_j1J5ti.fication;; and licensing defense sen-ices for commercial power reactors \\itbout commensurate expense;. Aho, public disclorore of the infotmation would enable

Serial No.22-025 Docket No. 50-395 Enclosure 2: Page 2 of 3 Westinghouse Non-Proprietary Class 3 CAW-21-5230 Page 2 of3 others t o u~ the infomiation to meet NRC reqniremenh foc liceming documentation

,.,ithout pllf'ChMing the right lo u;,e the infonnatioo.

(5) Westi:nghou.;.e has policies in place to identify proprietary infonnation. Under that S)-mem, infOIIllation is held in confidence if it falls in one or more of seveml types, the release of which might re.suh .in the loos of an existing oc potential competitive. advantage, as follows:

(a) The infonnation re\-eals the distinguishing aspects of a process (oc component structure, tool, method, etc.) v,:here prevention of im llie by any ofWes~use's competiiOIS without license from Westinghouse constitutes a competitive economic am-antage O\*er other companies.

(b) It consists of snpporti:ng data. including test data, relafu-e to a process ( or component. structure, tool method. etc.), the applicatioo of which dala secures a competiti,-e economic am-antage (e.g., by optimization or impro,-ed marketability").

( c) Its use by a competil<< would cednce his expenditure of resources or impro,-e his competitive position in the design. manufacture. shipment. installation.

aSS'UflU!Ce of quality, or licensing a similar product (d) Ir re,*eals cost or price infomlatioa, production capacities, budget le\-el~, or commercial stt:itegies of Westinghouse. its customer:, or suppliers.

(e) It reveals aspects of past. prese!ll, or future W~ghome oc customer fonded de,-elopme.nt plaru and programs of potential COlllJile£Cia1 ,'alue to

\Vestingho~.

(t) It contains patentable ideas. for which patent protection may be desirable.

Serial No.22-025 Docket No. 50-395 Enclosure 2: Page 3 of 3 CAW-21-5230 Pagelofl (6) ~ auacbcd documtnb arc ~led and marked lo indicat.e the bases for withhold* g. The j ustification for wilhholding is imlieatcd in bodl vcruom by means of lower-case letters (a)

Uut>ugb (i) IORtcd as a 11ij)CnCript immcdi.a.tcly following the bttckets enclosing each item of information ~ms identified 1.1 popnctmy or in the 11111'&in oppQSim gdl infomiation. ThCIC fowg--c;asc letters mer to the ~ of k!fonnation Westinghouse customarily bolds in confidence idmti.tk:d in Sections {SXa) through (t) ofthls AffidaviL I declare 1hal thii llvc:mle!rts offKt set f011h Ill, this Affidavit are true and eom:ct to lbo bc$1 ofmy knowkd~ intbrmation. and belief.

I ~ lllldQ- pcnaUy of perjury tha11be foregoing is true and corm:t.

~-14//

AnthODy J. SdloedeJ. ~

cVlliCI Llccosing & ConfiguraliOD Management

Serial No.22-025 Docket Nos. 50-395 Enclosure 3 LTR-SDA-21-023-NP, Revision 0 (Non-Proprietary)

Virgil C. Summer Nuclear Station (VCSNS) Unit 1 Dominion Energy South Carolina, Inc. (DESC)

Serial No.22-025 Docket Nos. 50-395 Enclosure 3: Page 1 of 32 Westinghouse N on-Proprieta1y Class 3 LlR.-SDA-21-023-NP RevisionO Y. C. Summer Unit 1 Steam Generator Hot Leg Inlet ~ozzle to Safe-End ,veld PWSCC Growth Analysis October 2021 All1h01: Brady Cameron*, RV!CV ~gn & Analy~is Anee& Udyawar. RWCV Design & Analysis.

Verifiers: :Maria Rizzilli*. R\-XV Dwgn & Analysis ApprO\-ed: L)'llllA Patterson*, l\fanager, RWCVDe.ign& Analysis

..Thl:tronica11_r approwd records m~ anth1mricatro i11 the eltct1"011ic dor11111~1t ,r1a11agt'n1tt1t systenl

@ Westinghouse

"'This r,;cora v,as fna l apµo'led IOil!/2021 8:52:57 AM. tTos s = nt was a.."'ced b y the PRll,IE system upon :tr, alid.ric )

Serial No.22-025 Docket No. 50-395 Enclosure 3: Page 2 of 32 Westinghouse Non-Proprietary Cla~s 3 LTR-SDA-21-023-NP Re._,i;ion O FOREWORD This document contains Westinghouse E lectric Company LLC proprietary information and data "'ilich ha:.

been identified by brackets. Coding (a.<.*) associated v.i.th the brnclrets sets forth the basis on which the information is considered proprietary based on Westinghouse policy BMS-LGL-84.

The proprietary infonnation and data contained in this rep ort were obtained at c011Siderable Westinghouse expense and ili release could seriously affect our competitive position. This infoanation is to be withheld from public disclosure in accordance with the Rul5 of Practice 10CFR2.390 and the infonnation presented herein is to be safeguarded in accordance with 10CFRl.390. Withholding of thi; information d oes not advet"Sely affect the public intuest.

Thi; infOfJlllltion Im been pro'\-ided for your intemal use only and rdiould not be released to persons or organizati= outside the Directorate of Regulation and the ACRS without the expre;s written appro,'lll of

\Vestinghouse Electric Cawpany LLC. Should it becollll" necessary to release this infom:iation to such persons as part of the review procedure, please contact Westinghouse Electric Cawpany LLC, which will make ibe necessary arrangements required to protect the Company';; proprietary interests.

The proprieta1y infomiati011 in the brackets is pro,ided in the proprietatyvemon of this report (LTR.-SDA-21-023-P Re"\.UtOll 0).

P:ige 2 of32

Serial No.22-025 Docket No. 50-395 Enclosure 3: Page 3 of 32 Westinghouse Non-Proprietary C1a;; 3 LlR-SDA-21..023-NP Re..,ision0 1.0 Introduction The V. C. Summer Unit 1 replacement Steam Generators (SO) v.-ere fabricated withfactotywelded stainless steel safe e11d, attached to the SG primary nozzles with Alloy 81 Dissimilar Metal (D~{) weld,. The replacement steam genentors were in;talled during the fall 1994 outage. The inside smface of the welds and adjacent base material:. were cladded with Primary Water Stress Corrosion Cracking (PWSCC) resi;tant Alloy 152 material during fabrication. In Fall 2018, these welds were examined using inside smface phased array ultrasonic volumetric and automated eddy cwrent ,;udace examination method; _The Fall 2018 impection did not m*eal any indications on any of the SG DM weld;_

Per Code Case N-770-5 Inspection Item A-2 [l J unmitigated butt welds at hot leg operation temperature

'.:: 625°F, such as th.e SG primary inlet nozzle DM weld, are volumetric ally inspected every 5 years. While, per N-770-5 Inspection Item B-2, umnitigated butt weld at cold leg temperatures (between 525°F and 580°F), such as the SG outlet nozzle DM weldarein:.pectedonce per interval (i.e. 10 years).

Tue ne,._i scheduled volumetric examination for V. C. Summer Unit 1 Steam Generator inlet dissimilar metal weld, is planned for Spring 2023, which is 4.5 years ~ce the last examination inf all 2018. HoweYer, th.e site is seeldngto defer the inspection to Fall 2027, which is 9 years since Fall 2018. The steamgeiierator outlet nozzle, per N-770-5, are now scheduled to be inspected once ei.-eiy interval; therefore, no relaxation i; requested fur the cold leg temperature ~team generator outlet nozzle weld impections.

The technical justification for the defenal of the SG inlet nozzle dir.;imilar metal weld;. to 9 years is perfOlllled in thi; report based on a detailed Primary Water Stress C-Olrosion Crack (PWSCC) growth analy;is_ The crack growth analysis v.ill consider roioiom.rn detectable postulated flaw; that could h}'J>Othetically pl"opagate through the Alloy 152 inlay and the Alloy 82 dis;imilar metal weld mat.eruili of the Steam Getierator nozzle for duration of9 years from Fall 2018. The analysis consi~ rs the latest plant specific temperature, loadings, geometry, and welding residual stresses fo£ the ft:icture mechanic-~ and PWSCC growth of the postulated axial and circumferential flaws in the nozzle D~-! weld region.

It is known that Alloy 690 and its family of weld metili such as Alloy 152, 52 and variants are highly resistant to PWSCC as compared to Alloy 600:t 82i82 material;. Excellent operating experience with no o~ervatiom of P\VSCC in almost 30 years of service supports the superiority of Alloy 690 and its weldme nls relative to Alloy 600 and its associated weldmenfs (i.e. Alloy 82, Alloy 182 and variants) in P\VR primary water emir0llllle1lfs, as does extensive laboratory testing.. Thus, o, -ei* the past few years, an i:ntemational expert panel v.-as organized by EPRI to determine the factors of impro\*ements (FOI) for PWSCC of Alloy 52i152!690 materials o\-er Alloy 182/600 crack growth rates that could be u.;ed in PWSCC crack grnwth anal:y;is_The investigation of the EPRI expert panel led to the publication oflMR.P-386 [2] in December 2017. Based on }.fRP-386, it was recommended that for Alloy 52i152 materials, a PWSCC factor of improvement of 324 over the Alloy 182 PWSCC growth equations from ~IRP-115 [3]

can be used in crack growth ei.'aluations. The factors of imprm:ement based on MR.P-386 were then recently codified by the AS..'\{E Code Committee Working Group on Flaw Evaluation Reference Cur.es as a ASl.'.IE Sectioo XI appro\*ed Code Case N -909 [4].

The NRC Staff has conducted re..iews of 11.IRP-386 and Code Case N-909 and detennined that generic approval of these doc=nts will not be provided Instead, the Staff will re.iew plant-~pecific submittals Page3 of32

"'This reoord vras final al)proved on 10/9/2021 8:52:57 AM. (ih:s statem,nt was added bytlte PRD,I E system u;,on ~ validnon)

Serial No.22-025 Docket No. 50-395 Enclosure 3: Page 4 of 32 Westinghouse Non-Proprietary Qa;_~ 3 LJR-SDA-21-023-~l' Revi,--ion O based on a :fracture mechanics evaluation with the me of factors of improvement o..-er the Alloy 182 curre to represent the P\VSCC gro'-"ih through tbe Alloy 152 material.

Foe V. C. Summer Unit 1, the analysis in this report herein will determine ihe minimum FOI required to provide at least 9 yellfi of operation for a postulated flaw to grow through ihe Alloy 152 inlay and SG DM

\1,eldmaterial and reach ihe maxinmmallowable end-of-evalu.ation flaw 5ize. The maximum allowable end-of-evaluation flaw size and P\VSCC crack gmwth for Alloy 82 is perfonned based on ASl\iE Section XI, Appendix C methodology [5]. The PWSCC growth through the Alloy 152 is based on the mininmro back-calculated FOI over the Alloy 182 rafe as specified in Appendix C [5], which will pro--ide 9 yean of operation between ihe Fall 2018 inspection and the proposed Fall 2027 outage inspection. The PWSCC analj"ill and resulting roiniroum FOI for Alloy 152 is provided in Section 7 of this report, while the general methodology for the fuicture mechanics is prmi.ded in Section 2, with accompanying reference;, in Section 9.

Page4 of32

... This """"° vras final accro < maximum allowable end-of-ei.1Huation period flaw size is detennined in accordance with the 2007 Edition with 2008 Addenda of tbe ASME Section XI Code [5]. To detemiine the maximum allowable end-of-evaluation period flaw sizes and the crack tip 5tress intensify factors u._<ed for th.e P\\'SCC an.alysi,;, it is nece;sary to establish.the stresses, crack geomeliy, and the material properties at the loeatiOIB of interest.

The applicable loadings, which must be cons.idered, consist of piping loads acting at the D...'-1 weld regions and the llielding re.;idoal slies= which exist in the region of intere;t.

The loadings considered in the anal;,is ifl.cluded the latest piping loads for the replacement steam generator; taking into con;ideration the Stretch Power Uprate (SPU). In addition to the piping loads, the effects of welding residual stJe,;ses are also considered. The nozzle geometry and piping loads used in the fuicfnre mechanics anal)~ file shown in Section 3.0. A discussion of the plant specific welding residual stres, distributions used for the DM welds is pro,ided in Section 4.0. The detennination of the maximum allowable end-of-e,-aluation period flaw sizes is discussed in Section 5.0.

The flaw grO\\ih is detennined based on !he PWSCC growth mechanism in the SG primary nozzle Alloy 82 D).! weld & Alloy 151 inlay material. The PWSCC gr°"'ih model for the Alloy 82 DM \\'eld material i; perMRP-115 (3), whichi;conmtent withAS:ME SectionXIAppen.dix C (5]. For the Alloy 152 inlay, the PWSCC grom:h considers a factor of improvement of29 o-,'t'r the ).IRP-115 (3] grO\\ih rate for the i\lloy 182 material (see discussion in Section 6). This factor of improvement is used to ju.;tify perfomiiog the vohlmetric examination of the SG DM welds in Fall 2027 refueling outage, which is 9 years (8.48 EFPY) from the pre,ious Fall 2018 inspection. PWSCC gro,11ih is calculated ba..<ed on the nonnal. operating temperature and the crack tip stre;.s intensity factor; resulting from the nonna1 operating steady state piping loads and welding residual stres;.es aa; discussed in Section 6.0. Section 7.0prmides the .ilaw gro\\ih cur1:es used in determining the allowable iflspection interval for the V. C. Summer L"nit 1 SG inlet nozzle DM

\\'elds.

Page :5 of32

"' This record was fina a~ on 10."912021 8.52:57 AM. m *s staterr~ni was a * "d by me PRiME system Ul)Oll

valida1io )
Serial No.22-025 Docket No. 50-395 Enclosure 3: Page 6 of 32 We,twgl.lou.se Non-Pfopuetary Chm 3 LTR-SDA-21--02:J.NP Re\'isionO 3.0 Nozzle Geometry and Loads Geornetrv. MateriHI Properties. and Normal Operntme: Parameters Tue V. c. Sllllllller U1111 I SG iolet nozzle dil.silnilaf ID.eta! wcld _geollletnes were ba~ed on SG draw.Lil(! {7].
I lle dini.ensi ns are sllown mTable 3-1. Tut lirnitingmarerial propcrtifs IBed wcre bastd on ihose for the lower strength stainless steel safe aid material iu lieu of the DM wdd 1D.1tcrial. The l 00"/2 normal opa ating rempernmre was based 011 customer com:spondeuce. which also considers any latest uprate conditions. A nonnal operating prtssurc of 2250 psia was tL'iCd in the analysis.
TabL 3-1 V. C. Summn Unit I Steam Gtnera tor Inlet Nozzk Gt'omtlry, Normal Oprr-atin2 Temperatur~.
and :\Iate11al Properti<< a.c.e Pipiil2 Loads The piping loods d~ to prc-ssun:. deam*;eight. 100% power nonnal operating thcm 1al expansion. seismic evrnts. and Loss of Coolani Accident (LOCA) events were c0ll5ldered for the a alysis of the SG inlet nozzles The axial furce and moment components for vanous loadm~ axe summ_arized m Table 3~2 The loadlngs considered in this analysis induded !he effects of the replacement siealll genetator program and the SPU. The loads in Table 3~1 bolllld all loops ofV. C. SlllllDler L'nit 1 SO primary inkt nozzle D~f wdd location.
Page 6 of J2
Serial No.22-025 Docket No. 50-395 Enclosure 3: Page 7 of 32 We,tiughou.!!e Non-Proptietaty Cla~~ 3 LTR-SDA*2 1..023*NP Re\i siau 0 Table l-2 V. C:. Summet Uult I Piping Loads
Serial No.22-025 Docket No. 50-395 Enclosure 3: Page 8 of 32 Westinghouse Non-Proprietary aas~ 3 L1R-SDA-11-023-~
Re-\'liionO 4.0 Disc;imilar :\Ietal Weld Rec;idual Sh*ess Distribution The residual ,tresses used in the generation of the crack grov.1h evaluation charts are obtained from the finite elemMlf residual ,tress analysis for the V. C. Summer Cnit 1 Steam Generator oos-imilar metal w eld geometry in C-8842-00-01 [8].
The finite element anal:,= in C--8842-00-01 [8] wei-e petlonned to simulate the weld fabrication process for the nozzle safe end weld region .B;Ullling an initial. 50'% inside surloce weld repair in accordance with the guidelines in :MRP-287 (9]. A two-dimemional axi,:ymmetric finite element model of the nozzle \l,as used in the finite elem.ent analysis. The finite element model geometry inchldes a portion of the low alloy steel nozzle, the stainless steel safe e:nd, a ponion of the stainless steel piping, the DM weld attaching the nozzle to the 5.afe end (along with an inlay on the inside surface). and the stainless steel weld attaching the safe e:nd to tbe piping. Figure 4-1 shoW5 a sketch of the final nozzle DM weld configura1ion. The following fabrication sequence was ~ ulated in the finite element residual 5fress analysis:
The SG nozzle is buttered with weld-deposited Alloy 82 mate!-ial. The inside ;umce of the buttering and the nozzle is dadded "'-ith weld deposited Alloy 152 material.
The nozzle and buttering are po;t weld heat treated at 11 00"F.
The nozzle is welded to the s-afe end ring forging *with an Alloy 82 weld, with a hrjer of Alloy 152 on the inside snrfitce.
A repair ca.ity of 50'!f, of the w-all thicl::ness is machined out of the weld region as per the guidance in MRP-287 [9]. The repair cavity i; filled with Alloy 82 weld metal, with a layer of Alloy 152 on the inside surface.
The outside and inside diametl!fl of the weld region are machined to final MZe.
A shop hydrotest is performed at a pre;,5ure of3110 psig and tempemture of3 00°F.
The safe end is machined v.-ith the piping side v.-eld prep.
The machined safe end ifl welded to a long segment of stainle5-;, '!,teel piping using a stainless steel l\"eld.
A plant hydroiest is petlooned at a pressure of2485 psig and a temperature of300°F.
After the plant hydrotest, 11oanal operating temperature and pressure was uni:founly applied three times to comider any shakedown effects, after ,vhich the model \v-as ;et to nonn.al operating conditions.
The resulting hoop and axial welding residual stresses under normal operating conditions in the DM weld region are shown in Figures 4-2 and 4-3 respedi*.-ely. [
Page 8of32
"'This r.,roni w.,s approved o 101912021 8:52:57 AM. ~ s~ n t wa,; aooe:I byt'ie PR1ME system upon ~ validation)
Serial No.22-025 Docket No. 50-395 Enclosure 3: Page 9 of 32 Westinghouse Non-Proprietary aa~s 3 LJR-SDA-21-023-NP RevrnonO The 11.wtl stresses used in the P\VSCC grov.1h analysis are based on the combination of !he axial welding residual stresses and the stres-se;. due to pre;.;ure, ll011llal operating thennal expansion loads, and the deadweight loads.
Page9of32
... This n>eoni w;is lin;,I appro*eo on 10!9/2021 8:52:57 AM. (Tit_s stJten-""nt was acdE-d bylhe PRIME 5YStem upon h va :d.alio )
Serial No.22-025 Docket No. 50-395 Enclosure 3: Page 10 of 32 Westi.n.gboo.sc No11-Propne1aty Cla<;s 3 LTR~"IDA-2 l-OH -NP Rerisiou O SiiJ End - Stainless Steel BMe Metal Wald Inlay -
Safe End lo ~ e W91d
Alloy 82 Fl er Mllffll Noule Suttaring
Alloy 82 F!Uer >. 8181 NoZZ!~ (Chanlll!I H211d FO~ing)-Catl10nSteel Base Metil Ch8nnel Haad ~Ing
Vt,'e!d Cladding -
Stalnlest. S~ Ftllet Alloy 152 F l!f Me1al Metal fl g,n"f 4-1: V. C. Snmmu Coit I Stram Gt'nrrator Dissimilar Mt'tal Wfld Configuration
Serial No.22-025 Docket No. 50-395 Enclosure 3: Page 11 of 32 Westinghouse Non-Pioprietary Class 3 LTR-SDA-21.023-NP Re,iti<,:, 0 a..c.e Fi gtln 4-l: Tbro11gb-WaII Hoop Resld!lal Stress at tbt SG lnlrt :Xonlt Dl\l Welm t:ndrr :Xorm al Opu:i tl g Co dittons
~~II of32
Serial No.22-025 Docket No. 50-395 Enclosure 3: Page 12 of 32 We.tiilghous.: Non*Pl"Op!loalaiy Class 3 LTR*SOA-2 1-023,.NP Re,u.on 0 Figure 4--3: Through-Wall ADaJ R~ldaal Stft'Ss ar rhe SG In! 1 :-i"ouJr D'.\I Weith r; der Xofflllll Operatlttg Conditions
Serial No.22-025 Docket No. 50-395 Enclosure 3: Page 13 of 32 Westinghouse Non-Proprietary Chm 3 LTR-SDA-21--023-:NP Re\ci;ion 0 5.0 ~larimum Allowable End-of-Ernluation Period Flaw Size Determination In ord& to develop the technical justification fur a longer interval between examination of the SG primary inlet nozzle D:\{ weld, the fir,t step is the determination of the nwtimum allowable end-of-evaluation period flaw sizes. Toe maximnm allowable end-of-e.*aluation period flaw size is the size to 'lmich an indication can grow to until the next impection or evaluation period. 1llli particular flaw size h determined based on the piping load;, geometty, and the material properties of the component. The evaluation guidelines and procedures for calculating the maximum allowable end-of-evaluation period flaw siz6 are described in paragraph IW13-3640 and Appendix C of the AS..vlE Section XI C-Ode [5].
Rapid, nonductil.e failure is pos.sible for fenitic material,; at low temperature, but is not applicable to the nickel-ba,e alloy material. In nickel-base alloy material, the higher ductility leads to t\vo possible modes of failure, plastic collapse or unstable ductile tearing. The -~ ond mechanism can occur when the applied J integral exceeds the Jk fracture toughness, and some stable tearing occurs prior to failure. If thi;. mode of failure is dominant, then the load-canying c.apacity is less than that predicted by the plastic collapse mechanism. Th~ muirnum allowable end-of-e'l..lluation period fla\v sizes of paragraph I'WB-3640 for high toughness materials are detemiined ba,ed on the assnmption that plastic collapse would be achieved and would be the dow.ina:nt mode of failure. However, due to the reduced toughness of the DM welds, it is pos.sible that crack extension and UIBtable ductile tearing could occuc and be the dominant mode of failure.
To account for this effect, penalty factors called "Z factors" ,..-ere developed in AID.-lE Code Section XI, w hich lll"e lo be mnltiplied by the loadings at these welds. In the current analyill for V. C. Summer Unit 1, Z factors ba,ed on :\iRP-216 [l 0) a.re used in the ana1ysi~ to pro,ide a more repre'lentative approximation of the effects of ihe DM weld,. The Z-factois fur Alloy S-2/182 from}.fRP..216 [10) h.a'l.-e been incoqxirated into the latest NRC approved ASME Section XI 2013 Edition per 10CFR50.55a.. The us,e of Z factors effectively reduces the ma.itimnm allowable end-of-evaluation period flaw sizes for flux weld, and thus has been incorporated directly into the evaluation performed in accordance with the procedure and acceptance criteria given in n\lB-3640 and *.\ppendi.,;: C of ASME Code Section XL It should be noted that the maximum allowable end-of-evaluation period flaw sizes i, 75% of the \\'-all ihicl:ne.;., in accordance \\'ith the requirements of ASl\.fE Section XI paragraph I'WB-3640 [5).
The !lla'Wllll1ll allowable end-of-evaluation period flaw sizes detemiined for both axial and circumferential fla\v-s ha,*e incotporated the rele-,;ant material properties, pipe loading-3/4 and geometry. L oadings under normal, upset, emergency, and faulted conditions a.re considered in conjunction \\'ith the applicable safety fact= for the corre.ponding sen-ice conditions required in the .AS).-ffi Section XI Code. For circ:umferential fl.a,vs, axial ;,tress due to the pressure, dea~ght, thermal expan,ion, seismic, and pipe break loads are considered in the evaluation. For the axial flav.s, hoop stren resulting from pressure loading is used.
The ma.'Wlllltll allowable end-of-evaluation period flaw sizes for the a.-ual and circumferential flaws at the SG primary inlet nozzle D:\I weld are prO\ided in Table 5-1. The ma.icimum allowable end-of-evaluation period axial flaw ; ize was calculated with an aspect ratio (flaw length/flaw depth) of2. The aspect ratio of 2 is reasonable becai>>e the a.-ual flaw growth due to P\VSCC is limited to the *width of the D:\I *weld configuration. For the circumferential flaw, an aspect ratio of 10 is n..sed.
Page 13 of32
"' This record was final approved o IO!P/2021 8:52.57 AM. (Th:s s!3ten:ent was aclcal by the F ME system upon ts va!ida:i )
Serial No.22-025 Docket No. 50-395 Enclosure 3: Page 14 of 32 Westinghouse Non-Proprietary C1a.;.~ 3 LTR.-SDA-21-023-:'.'ll' Re\:mo:n 0 It should be noted, again, that the resultine m-"l<iomro !ll.lov,,able end-0f-evaluation period flaw me-s were limited by the ASME Code limit of 75%, of the ,veld thickness for both flaw configurations.
Table5-l lla:s:imum End-of-Ernluation Period Allon-able Flaw Sizes (Flaw DepthfWall Thickoef>S Ratio - alt)
A-..ialFtaw Circuo:rli!f"ential F1aw (Aspe<:t Ratio= 2) (Aspect Ratio= 10) 0.75 0.75 Page 14of32
"'This r~ coni wa,; w ap~ O'l 10/11/2021 8:52:57 AM. (Th;s Slate= nt was a aced !>y e PRIME svstem uixm 'ls val'da
n)
Serial No.22-025 Docket No. 50-395 Enclosure 3: Page 15 of 32 Westinghouse Non-Proprietary Clli~ 3 LlR-SDA-21-023-~
Re\isionO 6.0 Primary \Yater Stress Corro'5ion Cracking Discussions This ~tion provi des the detaili of the P\\1SCC growth correlatiOfil n-sed for V. C. Summer Unit 1 SG DM for Alloy 152 and Alloy 82 (Section 6.1). General discussion of P\VSCC initiation for Alloy 152/ 52 as related to Alloy 182 is pro"ided in Section 6.2 to qualitati.:ely describe the significantly long time needed for a PVlSCC crack to initiate before it can even propagate due to PW'SCC. La>!ly, a re\ciew ofthe operating experience for Alloy 600.182/182 as compared to Alloy 690.152/152 is gi,-en in Section 6.3 to discuss the excellent service histoty for Alloy 690 base material and its weldments.
6.1 PWSCC CR.\.C.K GROWTH .-\.X-\LYSIS The PWSCC grov.1h analysis invol\-e; po;tnlating an inside flaw in the dissimilar metal weld inlay for the nozzles of interest. The objecti\-e ofthi~ anal},is i~ to detennine the sen.ice life required for a postulated inside ;urface flaw to propagate to a size that exceeds the maximum allowable end-of-evaluation period flaw depth as described in Section 5.0. An initial flaw depth of0.065 inch into the inlay "'ill be used in the crack growth evaluation. Note that for all postulated iruide sutfuce flaw:,. the governing crack growth mechanism for the SG inlet nozzle DM v.-eld is PWSCC.
Crack grov.1h due to PWSCC was calculated for both axial and circumferential flaw-s based on the llCtmal operating condition iteady-state str~ses combined with the v.-elding residU.'ll strei.;e.;_ For axial flaws, lhe hoop '>tresse.; are due to pressure and residual strei;es. For cii-cumferential flaws, the axial stresses considered are due to pressure, *thetmal expans.ion, deadweight, and re>idual ,;ti:esses. The crack growth analy,is requires calculation of the crack tip stress intensity factor {Ki), which depends on the geometry of the crack, its surrounding strocture, and the applied stresses_ The geomeliy and loadings for the nozzles of interest are discussed in Section 3_0 and the applicable residual stresses n-sed are discussed in Section 4.0.
Once K: i;; calculated, PWSCC gro\\1h due to the applied itresses can be calculated u.sing the crack growth rate for the Alloy 82 nickel-base alloy from :MRP-115 [3]. For PWSCC through the Alloy 152 inlay lhid::ne;,;, a factor ofimprm:ement of19 o\--erthe MRP-115 [3} crack growth for the Alloy 182 material is conservatively used. Tim FOI is n.;ed to justify perl'onning the volwnetric examination of the SG DM welds in Fall 2027 refueling ou.tage, which is 9 years (8-48 EFPY) from the pre\-io1JS Fall 2018 inspection :Kote, the length of the postulated axial flaw is assumed to be limited by the width of the Alloy 152 weld inlay remlting in an aspect ratio (flaw length/flaw depth) of 2, while for circumferential postulated flaws, the a;,pect ratio is assumed as 10_
Using the applicable stresses at the D1{ welds, the crack tip str~ intefility fact= can be detemili:ted based on the stress intensity factor expressions frnmNASAdatabase [11} andAPI-579 2007 Edition[12]. [
]"c.- A 41!o order polynomial stress distribution profile is defined as:
Page 15 of32
... This reooni was '<11 apflr()ve<i oo 10!0,2021 8:52:57 AM. {Th:S mtemicru w as a<iced bv me PR ME system uJ)OO 'ts v * )
Serial No.22-025 Docket No. 50-395 Enclosure 3: Page 16 of 32 W5tingb.ouse Non-Proprietary C1as~ 3 LTR-SDA-11-013-NP Re..-ision O Where:
~ . 01, <>2, u1, and <H are the stte;;<; profile cun*e fitting coefficient~;
x is the distance from the u.ill surface where the crack ini:tiate8; t is the wall thickness; and cr is the s!IBs peipe:ndiculac to the plane of the crack.
The stre;,;. intensity factor calculations for semi-elliptical inside i.utface axial and circumrerential :flawa are expressed in the general fOfDl as follows:
Where:
a Crack depth C Half crntl:: length along suri'ace t Thickness of cylinder R Inside radius 4> Angular position of a point on the cnck front n Order of polynomial fit Gj Gj is the inflllence coefficient far j,i,. ~ires!; distribution on crack surfuce (i.e., ~.
Gi,~.~G4)
Gj cr1 is stress profile cmve fitting coefficient foe j 8 stress distribution (i.e .. cro. 01, a2.
G3, v4)
Q ~ shape fuctoc of an elliptical crack i.,; approximated by:
Q= 1 + 1.464{a!c) 1 El for a!c ,;: 1 oc Q = 1 + 1.464{cfa) 1 tl for a'c > 1 Once the crack tip stress intere:ity facton are detennioed, P\\7SCC grouth calculations can be performed using the crack g,:o\\1h rate discussed below with the applicable norm.a.I operating temperature.
The V. C. Summer Unit 1 SG inlet nozzle to safe e1ld dissimilac metal weld regions are conitmcted primarily of Alloy 82 \\-iih an Alloy 152 .inlay on the im.ide surl"ace. The Alloy 152 inlays were installed :is a protecth:e barrier foe the Alloy 82 weld against P\\'SCC. Alloy 152 v.'eld metal is known to be moce resistant to P\VSCC tban the lower chcomium confelli Alloy 82. Current industiy dalll fi'om ~ IRP-386 [2) and N-909 [4] provide s technical baYS that the PWSCC crack growth for Alloy 152 has a factor of
.improvefllfflt of 314 01:er the PWSCC crack growth of Alloy 182.
How~._-er, foc the e'\-aluation contained herein, a conservati\c-e impt*o\-ement fuctor of29 over the Alloy 182 crack growth rate \\-ill be used to repcesent the crack grow1h rate of Alloy 152, in ofller to achie1;e 9 )'-Carn Page 16 of32
"'This recoo1 was f,rol ap~-wwd oo Ulil!i2021 8:52:57 AM. (Thls S!ale!l'enl was added by lhe PRlME ~y,;tem ul)Oll 1IS vali<b *o-:)
Serial No.22-025 Docket No. 50-395 Enclosure 3: Page 17 of 32
\Vestinghouse Non-Proprietary C135s 3 LTR-SDA-21-05-~
Re-,, iiion 0 (8.48 EFPY) of operation for a postulated flaw to reach the maximllm allowable end-of-evaruation period
lla,v size.
The PWSCC grov.1h rate fur the Alloy 82 (and Alloy 152 FOI) material based on MRP-115 [3] is:
da di
=exp[-~ (~- _!_)]~(K),
R T Tru FOi
'Where:
da Crack gi-m.-ih rate in m!sec (in:br) dt
<4 Thermal activation energy for crack growth = 130 kJimole (31.0 kc aL'mole)
R Universal ga;; constant= 8-314 x 10-3 k.J.(mole-K ( l.103 x 10*3 kcal'mole5R)
T Absolute opernting tl'wperature at the location of crack. K ( 0 R)
Tm Ab!iolute reference temperature u.5ed to normalize daia = 598.15 K (1076.67°R) a Crack gi-owth amplitude 1.50 X 10-11 at 325eC (2.47 X 10-' at 617°F) p Exponent= 1.6 K Crack tip stress imen;.ity factor MPa v'm (l:si \'in.)
FOI Improvement Factor = 2.6 for Alloy 82 ~IRP-115 and AS~IE Section XI Appendi.." C)
FOI 19 considered for Alloy 152 (;;ee discussion below)
As discussed previomly, based on :l'.IRP-386 (2) and N-909 [4], the PWSCC crack growth fur Alloy 152 is a factor ofimprm:emeot of324 over the P\VSCC crack gro\\"1h of Alloy 182 from l\.IRP-115 (31-In April 2010, the N""R.C Staff conducted a general confumat oty re1.iew of flaw tolerance e-..tluations of inlay as a mitig;ttion technique f.or PWSCC concerns in PWR. The cowinnatory study by the Staff was published in ML101260554 [6], which considered reactor coolant nozzle "'ith Alloy 81 f181 dissimilar metal weld v..ith an Alloy 152/52 inlay 011 the in.side surface . The PWSCC rn1ck growth model in Section 5.1.2 ofMI..101260554 [6], considered a FOI oflOO for Alloy 152152 o,wthe PWSCC rate of *.\lloy 182 from MRP-115. The FOi of 100 was used as the primary baseline ca,;,e in the study, but also considered FOiof30and IOOOas sensiti'.ity case.s , ree Section5.4-2 ofRef[6]. The dio;c1m.ionin Section5.42 of[6]
stated that the FOI of30 represented 9 5~ percentile of the a1;-ailable labocatoiy data tested foe Alloy 521152 at the time.
In this evaluation, the Alloy 152 improvement factor of19, although slightly higher than that recommended in. the NRC Public Meeting Summary (13), is a more realistic improvement factor for Alloy 152 welds based on more recent data documented in 1-.ilJ°REG/CR-7103 Volume 4 [14]. Nu"'REGICR-7103 pro-.ides the latest NRC ; ~ed study of Alloy 52/152 mock-ups SCC growih rates from PNNI.., ANL, GEG. and CIEMAT laboratory data.As shownin.Fignre6--L which is Figure 3-132of.iUREGiCR-7103 Volume 4, nearly all the measured stress corro;ion propagation rates fur the Alloy 511152 welds are less than lx10-s Page 17 of32
"' This recool was final ;io ~ "" 10.f{l/:lll21 8,52:57 AM. ffr~-s stJ!err""1t was add,ed tvlhe PR!ME s-.stem IIDOR ts valida!!<ml
Serial No.22-025 Docket No. 50-395 Enclosure 3: Page 18 of 32 Westinghouse Non-Proprietary Q33;; 3 LIR-SDA-21-023-~
Re,-rnon O mmfs with most less than 3xl o mmts. There i, also no significant difference between the crack gro\vth rates of Alloy 52 and 152 welds as .; hown in Figure 6-1. Included in Figure 6-1 are additional cunes for FOI = 29 and FOi = 324 (m;er the l\.fRP-115 Alloy 182 cutVe). These curves are used to illustrate the con;eivatism in the FOi of29 which bound a significant amount (e.g. more than 75'%) of the Alloy 152152 test data per l\'UREG,CR.-7103 Volume 4.
More recent SCC data on Alloy 152/52 with 15% cold forging v.-ere presented by PNNL at the Jan 11-14.
2021 EPRI Alloy 690/51/152 Prima,')' WatBr Stress Corrosion Cracking ResearcJ, CollaboratiOII Meeting
[15]. k discu.,sed in the Pl\1'"1. presentation [15), even with some amoum of cold forging.. the SCC rates remain below h:10-S mm..'s growthr:rte (seeFigure 6-2). Furthe-nnore, theuseofaFOiof29 over the :11.IRP-115 Alloy 182 will bound all the 15% cold forging SCC data prodded in Figure 6-2 Ilm,, the use of a FOI of29 over the ll.fRP-115 Alloy 182 PWSCC cmve for the V.C. Summer Unit 1 SG DM has sufficient justification based on EPRI e,q>ert panel (MRP-386, Code Case N-909). It aho has ju~tification based on the more recent NRC spomored study ptO\ided in NVREG!CR.-7103 and the latest Pl\'NL laboratory data present at EPRI Alloy 690/52/1 52 PWSCC research collaboration meeting, in January 2021 [15].
Page l&of32
"'This l'rcorti was a~rovec on 10;11/202 1 8:52:!i7 AM. (Tus 51a eroont was aoded by e PRiME S)"'tem upoo 'ts v d.rio )
Serial No.22-025 Docket No. 50-395 Enclosure 3: Page 19 of 32 Westinghouse Non-Proprietary Qa,;.,; 3 LIR-SDA-21-023-!-W Re\"rion O
'Y UJ~:q 1 ? A
lt' ,i-l'i'I t
M :i;* l ~' A Alllly 52 A Allu:,528 AllDy52. A
[,;ii l\lk,y 5;! ' 8!
11!".i Al!Oy"'2M8Z I} ABoy 521.' 83
A!!oy 52}.ISS
- MRP-i;,;
Gt:G 1~'1l GEG 152 0 GEG 15~ i:
GEG l 52F C EG 152G GEG 52>.tA CEU 1:1:rM t!
. GEG52 !l GEG52 C
-y fiEG52 ::l 0 GEG52MC
'1 )
'-l * ..
"l
I
1 " l...2 ;:
I
-: 1 1- - ,.
tl >t.~ lo.
~
a:
...... 1~:.,,.
C JU.'I\T 152 [l C!EMAT 2C 20 30 40 50 60 Stress Intensity, MPa'l'm Figure 3-132 Summary of PNNL-. GEG-. ANL- and CIEMAT*Measured sec Crack-Growth RatC!S for Alloy 152J152M!S2/52M Weld Metals as a Function of Stress Intensity Figure 6-1: SCC Crack Gro\1.ih Rates for Alloy 152'52/52M Weld Metah (Figure 3-132 ofNlJ"'REG!C'R.-
7103 Volume 4 [14D Notes:
The cun-es for FOI = 29 andFOI = 324 are added to the original Figure 3-132 of?-.u'REG,CR.-7103 Volume 4 for info!lllational ~ e s as a comparison with the laboratory data. The FOI = 29 curre is used in the analysis herein and lhe FOI = 324 is improvement :factorbasedon?,,fRP-386 [2] andN-909 [4].
In the above figure, the solid black n>n-e represents. the :MRP-115 Alloy 182 PWSCC cmw at 32YC of Ref [3] (Fig. 3-132 of [14] inacrnrately labels the solid cm..-e as MRP-55).
Page 19 of32
"' This ~ was linal .m~ °" 10i9/202I 8:52.57 Md. ffrh; statM".enl was aoc'?d bv li:e PRIME svstern uoon ,t s V3lads:icr.l
Serial No.22-025 Docket No. 50-395 Enclosure 3: Page 20 of 32 Westinghouse N on-Propnetary C1a.ss: 3 LTR-SDA-21-023-:NP Re.i;ion 0 MRP-115 Ntoy 182 360'C75%
C I*
t>.l!o, 152 t>.
PNN L As---we lded ,,
A and 5% CF Data I vss
.!.-,, I ;
, *-rr-1 ~
1s.; .-....... . *I I
~ .J, *
) ,,
f ,,,
t,"i, ~.: :
20 JO 40 50 tlO Stti ss Int nslty, MP-a~m Figure 6-2: Th'NL SC'C Test Results for Alloy 152i52 \lrith 15% cold fotging [15]
Note: The CllITa for FOi = .!9 and FOI = 324 are added to the original figure [15] for informational putposes as a comparison with the laboraroiy data.
Page 20 of32
Serial No.22-025 Docket No. 50-395 Enclosure 3: Page 21 of 32
\Vestinghoooe Non-Proprietary C1a;_;_ 3 LTR-SDA-21-023--~1' Re\'isiOD 0 6.2 PWSCC CRACK D,"ITI...\TIO~ OF ALLOY 1:52 ~L.\TERL-\1.S This section pro"i~ a compari.;on of PWSCC initiation data for Alloy 690/52/ 152 materials and data fur Alloy 600182/ 182 materials.
Based on the dim>>;ions provided in Section 3.2.1.3 of11RP-375 [16], laboratory tMts by EDF (Electric de France) shov.."ed that Alloy 182 cracked after 95 hours0.0011 days 0.0264 hours 1.570767e-4 weeks 3.61475e-5 months , and Alloy 82 cracked after 570 hours0.0066 days 0.158 hours 9.424603e-4 weeks 2.16885e-4 months . In coruparison, Alloy 521152 still had not cracked after > 21,000 hours0 days 0 hours 0 weeks 0 months . This resulted in a FOI for Alloy 52i1 52 of 37 compared lo Alloy 82 and over 150 compared to Alloy 182. Also, per MRP-375, KAPL (Knolls Atomic Power L:iboratoi:y) tested Alloy 521152 weld; fur 2300 h = at 640"F (338'C) ruld 5300 hours0.0613 days 1.472 hours 0.00876 weeks 0.00202 months at 680°F (36o=q_ The fonner tests showed no indication;; of PWSCC, v.-ilile the latter only had a few, isolated
~pod..<!ts." KAPL estimated the FOI ofAlloy 521152 over Alloy 82 to be approximately 100. Tests by ~fiII (Mitsubishi &a"'Y Industries) ha.-e dem=trated specimens of both Alloy;_ 52 and 152 that have not cracked after 107,000 houn [Seetion 3.2.1.3 of[l6)). Furthemiore, ba5-ed on more recent data by ~fiII for ihe Alloy 52 and 152 \Rids specimen. there w as no evidence of cracking after >122535 hours at temperature of360°C; this translates to 71 years of operation at 325°C (617°F) [Table 6 of Reference 17].
During the recent January 13-14, 2021 EPRI-NR.C Cooperative Re.search Project: PWSCC Crack Initiation Cliaracterfr:ation ofAll.Dys 600/182 and Alloys 690/521152, the presentation pro,ided jointly by EPRI and N""RC [18] discussed the status of the ongoing cooperative re;earch on PWSCC initiation t~ting at PNNL Based on the preseotation [Slides 12 and 13 of Reference 18], temperature-adju.sted median PWSCC initiation times provided based on the latest test data show that Alloy 690152.'152 materials ha\-e not experienced any PWSCC initiation at 360°C for more than >35,414 hours0.00479 days 0.115 hours 6.845238e-4 weeks 1.57527e-4 months (4.05 yearn) and longer than
>277,140 hours0.00162 days 0.0389 hours 2.314815e-4 weeks 5.327e-5 months (>31 )"ean) for an adju.;ted temperature of 325°C (617°F). \\'hen compared to the ., lowest P\VSCC initiation time for Alloy 182 weld [Slide 13 ofReference 18], this represents a FOi of greater than 48 for P\VSCC initiation of Alloy 152152 welds.
Therefore, the discussion prmided here for P\VSCC initiation support; that the Alloy 152 OM weld at V. C. Summer Unit 1 will take a significantly long time (more than 31 yeaB up to e\'en 71 yeaJS) before any e\'idence of crack initiation i; detected at the SG OM locations.
Pro*.ided in the nm section is a qualitative re1.iew of the operation experience for cracked Alloy 600/82/182 materials as a comparison to the operation time for V. C. Summer Unit 1 Alloy 690/52/152 materials to demonstrate the greater P\VSCC resi,tance of the latter materials a;_ compared to the former "1-b.en operating in the same en\-ironment and temperature.
Page 21 of32
"' This reooro was fi apl-""IM'(! 0<1 10/9/2021 8:52:fl AM. flit s statern<>nt as a d by the PRIME system upan 'ls ~ jc )
Serial No.22-025 Docket No. 50-395 Enclosure 3: Page 22 of 32 Westinghouse Non-Proprietary Oass. 3 LTR-SDA-21-023-)W Re*,i."ion 0 6.3 OP:ER.-\110:'.'> E.UERIE::-.CE OF _-\LI.OY 821182/600 YER.Sl'S All.OY 51l151!690 WELDS This section prmides a comparison of the oper-aling experience of components installed with the highly PWSCC resBtani: materials of Alloy 521152/690 in relationship to the operating experience of Alloy 821182/600 in PWR.reactor coolant ~ystem service_ To date, there ha...-e been no occurrences ofP\VSCC cracl:inginAlloy 51i151!690componentsinPWR.em,-ironment_ InfomutioncelatedtoAlloy~ 690, 52, and 151 inP"\\'Rs is locatedin11RP-110 [19] and inll.mP-111 [20].
Steam Gem;:rators Replaced steamgenerat= uithAlloy 690 tubing have been in operation since 1989 at D. C. Cook 2. Indian Point 3, md R.ingha1s 2. No corrosion induced flaws hai.--e been detected at these plams or any subsequent plants that ba,.-e st&1ed up ~ e that time uith Alloy 690 tubes in either replacement or original steam generatocs. In c01ltfasl. P\VSCC was detected after one cycle of operation at several units v:ith Alloy 600MA tubes. e.g., Doel 3, Tihange 1 and V . C. Summer Unit 1. and after the second cycle at a number- of other plants with Alloy 6001-fA tubes. This experience indicates that there is a senice demonstrated fac tor of impro\~ i of about 30 or more, with the value increasing as the PWSCC-ftee senice life of Alloy 690 tubes ronl:innes to accumnlate.
Steam geoerator tubes are joined to the tube Mieet using autogen.ous ,1.-elds between the tube and Alloy 52/152 cl.adding on the prim.aiy fuce of the ttJbesheet. Thus, each ;,team generator has thou.sands of welds and heat affected zones. 'Il=e have been no repoits of PWSCC initiation or crad:i:ng detected at the;,e weld joims b em*een Alloy 690 tubes and the cladding on the tubesheet (the cladding on early Alloy 690 steam generators w:n Alloy 82/182, while for later units ha~ been Alloy 52tl52). \\,"bile the _,\lloy 690 tube to tubesbeet ..-eld Joints are not routinely inspected with sensitive methods, significant cracking ..-ould likely have been deiected as result ofleakage or \isible crad:s, as has occuned occasionally with Alloy 600 tube to tubesh.eet v,..,elds. The earliest U. S. steam generator v,,ith Alloy 152 welds installed was in 1994 for V. C.
Summec Cnit 1. which has over 27 years of e.~ence with IID cracking. Point Beach 2 Steam Generator Alloy 52 welds have been operating at hot leg temperatures form-er 19 EFPY v.ith no e--.iden.ce ofPWSCC initiaticn or cracking.
Many ,;team generator tubes h:r,-e also been plugged using Alloy 690IT tube plugs since the late 1980;,_
There ha\-e been no reports of PWSC-C being detected in t1Rr..e p:lngs. 1he plugs ba"\-e been of two main kinds, the "ribbed" plug and the "roll-expanded" plug. In both cases, the plu~ are made from tbick w-all cod material rather than from thin tubes. In comrast to the o...-e£ 30 yezs of trouble-free 5er\'ice \\ith Alloy 690TT plugs, plugs made of Alloy 600).L-'\ and even Alloy 600IT e."'qJCrienced P\VSCC "\"\ithin one to two
r-ears of senice. Therefore. m.-er 30 ye.ars of experience for s.team generators indicates that there i5 a ~ ice dem:o~trated fadoc of improvement of about 30 or more for Alloy 52/152*690. v.ith the value increasing a'> the PWSCC-fiee senice life of Alloy 690 tube plugs contimles to accumulate.
Pressurizers Alloy 690 and its weld metak were also used to cepaicpre;;,,--urizer componeots in more than 17 plant~ since 1994. The hig.h senice temperature in the pressurizer-. (t)11ically 653°F) is partieul.arly aggressi.u~ regarding PWSCC initiation and grO\"\"-th. Many of these pressurizer components that have been repaired with Alloy 690 and its weld metals, ha....-e the equi\-alent of more than 100 EDY (Effective Degradation Y ean) ofnme-Page22of32
"' Tits !i!COfd ?las ma! 3!00>~ et1 t0:~.'21121 8:52:57 Al,1* !Th's st3ten:ml ..,,,. a ~ by t!:e PR iE SYS!em uron els y;Jl.'d.rlo \
Serial No.22-025 Docket No. 50-395 Enclosure 3: Page 23 of 32 Westinghouse Non-Proprietary Oas; 3 LTR-SDA-21-023-NP Re.,i.ion 0 temperature e.."}JO;,ure since installation without cracking. Inspections performed have shol\n no evidem:e of PWSCC Uldications.
Reacior l'essels Heads ~jt/i Alloy 690 No:=le.s New and repaired reactoc pi-essuce vessel heads with Alloy 690 nozzle;, started to be used in the indu.,tty from 1991. From 1992 on\l'-ards, a significant m.unber of reactoc presf>ure ve.;,;,el beads have been replaced with wrought Alloy 690 tubes and Alloy 52i152 weld metal (early replacements in France). In the USA, over 45 heads have been replaced with Alloy 52/152/690 materiah and the senice perfonnance has been excellent. MO!.t of there heads have over 40 to 100 penetration nozzles made of Alloy 690 and welded \,ith Alloy 52/152 welds. The;,e replacement heads have operated o,w 25 years with no inspection findings, and most all have been inspected at least once. In France, detailed ins-pections are R quired at 10 year interrnls for the reactor \*essel heads and in the Lnited States volumetric inspections are required enry 20 year, per Code Case N-729-6 [21].
Repait"s with Alloy 51/151 Material Alloy 52 and 152 v.-eld lllrfah started to be u.sed in repairs and in replacement componenis starting in the mid-1990;,. There ha,-e been no reports of senice indu.ced P\VSCC in thef>e welds.
Weld inlays were installed at Ringhal, 3 and 4 reactor ,-essel outlet nozzles in Sweden. For the hol leg nozzle at Ringha1s 4, the inlay was applied in 2002 and then inspected with vltrasonic Testing (U1) and Eddy Current Testing (ECT) in 2005 with no indications. Funhennore, UT and ECT inspections in 2010 also demonstrated no indicatiom after 11.7 EDY; morem.w, no indications were detected after UI and ECT ins-pecti= in 2020. This particular location is currently on a 10 year re-inspection frequency. Similarly, at Ringhals 3, the inlay was applied in 2003 and then inspected with li"T and ECT in 2006 with no indications.
Subsequently, this location wa~ re-illipected with UI and ECT in 2010 wiihno indications after- 10.4 EDY; moreover, no indications were detected after ur and ECT inspections in 2020. Thi~ location is now aho on a 10 )-ear re-inspection frequency. Tue experience of Ringhal; inlay~ on hot leg nozzle DM welds demonstrate;, a good precedence for the beneficial use of Alloy 152i52 inlay material applied to the Alloy 82weld.
Next, a comparative discussion will be prm.i ded for cracking of hot leg nozzles with Alloy 82/182 welds based on ope.rating conditions and duration as compared to the tinle V. C. Sllllllller Unit 1 SG D:M wel~
v.ith .\Hoy 152 inlay,; ha\*e been in operation. .Based on cunent operating data, the V. C. Summer l:nir 1 SG DM v.-elds with Alloy 152 inlay. ha,-e been operating for 27 :ye=, with no cracking or PWSCC initiation at hot leg tempernlures of 619°F since the time the replaced steam generators v.ere put in sen-ice in 1994. In contrast, V. C. Summer lJnit 1 di,covered through-wall axial cracking in a reactor ve;,sel hol leg DM \,-eld in October 2000 after being in-service since 1984 ,,..-hen the p lant was commercially started.
One of the leading factors of cracking at V. C. S= r Unit 1 was extensi'\-e weld repairs on the inside soc.face and cold worl: of the weld due to grinding [22].
Another- example of cracking in Alloy 182/82 welds at hot leg temperature is that of Seabrook. In October 2009, the 10-year in-;enice inspectioo at Seabrook identified an a.'rial indication in the Alloy 182182 DM at one ofthe reactor outlet nozzles. Tue plant ha, been in operation &ince 1990 and was at 16.53 EFPY with hot leg operating temperatures of 621 cF when the flaw was discovered in 2009.
Page23 of32
"' This ~ was final ac:irov<<! a 10/9/2021 8:52:57 AM. lThs Slaten:ent was add~ bv the PRIME S\'Stem uoon .ls validnonl
Serial No.22-025 Docket No. 50-395 Enclosure 3: Page 24 of 32 Westinghouse Non-Proprietary Oas.s 3 L1R-SDA-21-023-~
Re,ision 0 In 2012, North Anna 'C"nit 1 discovered m--o through-wall a~ al flaws at the SG inlet nozzle Alloy 182 u~lds w hen preparing to apply full strucrural weld overlays at the location during a planned outage. Note that no leakage v,,as obser.-ed during operation prim to the outage. Fabrication record; indicated extensive ID (Inside Diame ter) weld repair; were petlonned for the SG in question which had the through-wall flaw s
[23].
The discussion in thi; section demonstrates the excellent ;.ervice histoiy of Alloy 52/1521690 materials, v.-hether in'ltalled or used a; repair, as compared to Alloy 82/182/600 materiali. Based on ,enice history, mm.y of the Alloy 521152 welds have been in operation longer than Alloy 82/182 welds at hot leg conditions, \vith no e"liidence of PWSCC initiation or cracking.
Page24of32
"'This r;;roro was final acll!O £<I on 10/9/2021 8:52:57 AM. ffn's slaten:rnt was aa<:';;d bvth.e PRiME S'5tem uoon :ts v31id:;:icnl
Serial No.22-025 Docket No. 50-395 Enclosure 3: Page 25 of 32 Westinghouse Non-Proprietary Cla;.~ 3 L1R-SDA-21-023-~
P-.e.i,ion 0 7.0 Technical Justification for Defening the Yolumetric Examination V. C. Summer -Unit 1 previouslyperfomied qualified automated phased array u!trarocic (PA-UI) and ECT inspections of the;;e welds in Fall 2018 and no indications were idemified during that inspection. V. C.
Summer Unit l is seeking rela'Oltion from the AID.-IE Code Case N-770-5 [l Jrequirement in order to defer the volumetric examination of the SG prima1y inlet nozzle DM v.-elds. This inspKtion is planned for the Spring 2023 refueling outage but will be postponed until Fall 2027. Therefore, the technical bMis herein is to justify acceptable P\VSCC growlh for an undetected postulated flaw for an operation duration of 8.48 EFPY (9 j-"<'an) for the SG primary nozzle DM welds. A 0.065 inch initial flaw depth i~ postulated for the flaw growth evaluation and an aspect ratio of 2 is used for the a.--cial flaw and an aspect ratio of 10 for the circumferential flaw. This initial flaw depth is acceptable as eddy current terung has demomtrated the ability to detect a surface flaw as shall.ow as 0.3 mm (0.012 in.) in depth and E short as 1.5 mm (0.06 in.)
in length at the weld inlay The P\VSCC growth was calculated in two stages. The first stage is growth through the Alloy 152 inlay material. As di~u.,;;;ed in Section 6.0. for Alloy 152 an impmvement factor of29 m-er the crack growth rate for Alloy 182 in ~IRP-115 is used for grov.th through the inlay. The second stage is growth through the Alloy 82 DM weld ba5ed on :MRP-115. The Cfllck growth through each material is then combined to deten:nine the len.:,<>th oftime it would take for the po-stulated initial flaw to grow to the ma.'WllW!l allowable end-of-evaluation period flaw size.
Deadweight, normal thermal expansion and pressure (2.25 ksi) loadings along with through wall axial residual stresses u-ere u.sed to generate the through wall axial stref.;es used in the PWSCC analysB for the circumferential flaw. Since the axial v.-elding residual ;;tresses are compres<<..ive at locations through the wall, PWSCC analyses were performed with and without residual stress in order to detemiin.e the m o;t limiting PWSCC growth remits. Only through wall normal operating hoop r~idual stresses were used in the PWSCC analysis for the a.ual flaw. It should be noted that no fatigue crack grov.th calculation was pafonned since crack: growth doe to PWSCC is the conttnl1in3 crack growth mechanism.
The PWSCC gro\\th cun-es for an axial flaw and a circumferential flaw are iliown in Figures 7-1 and 7-2 respectively for the SG inlet nozzle DM weld. The horizontal a.~s displays senice life in Effective Full Power Years (EFPY), and the '\-ertical a.' lh shows the flaw depth to wall thickness ratio (alt). The ma.rjmum allowable end-of-evaluation period flaw sizes are also shown in the:ae figures for the respective flaw configurations. The SG inlet nozzle crack growth results in team of EFPY are based on a tempemture of 619.4'F.
The PWSCC growth rate is highly depeiident on the temperature at the location of the flaw, futthennore, t:he crack growth rate increases as the temperature increases. Therefore, duruig periods when the plant is not in operation, such as refueling outages or shutdowm, the temperature at the SG nozzles i, low mch that crack gro,..,.th doe to PWSCC is insignificant. Therefore, PWSCC grov.th calculation should be detennined for the time interval when the plant is operating at full po,..,.-er_ The amount of time when the plant is operating at full pov.-er i'> determined based on pre-~iou.s plant operation data and the anticipated outages scheduled nnti1 the next inspections. This operating duration at full power is referred to as Effective Full Power Years (EFPY). For V. C. SUDlDler Unit 1, based on operational data, the time inten.-al between the Page 25 of32
Serial No.22-025 Docket No. 50-395 Enclosure 3: Page 26 of 32 Westinghouse N on-Proprietary Oas.;; 3 LTR-SDA-21-023-NP P-.e,ision 0 previous inspection in Fall 2018 and the proposed future inspection in Fall 2027, is consenraiiwly set to 8.48 EFPY (9 calendar years). For V. C. Summer Unit 1, this transl.ate;; to a power availability factor of 94% to account for the time the plam is opelllting at 100% pov.-w and normal operating temperature of 619.4°F.
Based on the crack growth results from Figures 7-1 and 7-2, it is demonstrated that it would take more man 8.48 EFPY (9 years) for the postulated 0.065 inch deep a.'lial and circumferential flaw in the V. C. Summer Unit 1 SG inlet nozzle DM v:eld inlay to grow to the maxinmm allowable end-of-evaluation period flaw sizes with consideration of Alloy 151 with a FOI of29 over the Alloy 182 PWSCC rate from :l\-fRP-115.
Therefore, the plant ;,pecific PWSCC growthresu1ts, in this letter, provide technical justification that V. C.
Summer Unit 1 SG inlet nozzle DM welds can be examined after a duration ofat least 8.48 EFPY (9 years) from the pte\iou.~ refueling outage inspections in Fall 2018.
Page26of32
"'This re-rord was final ,11nrove<i on 10,'1!12021 8:52:57 AM. !llt:s statem;;nl was added bv
e PRiME svstem uron 'ls v;,!icb:iool
Serial No.22-025 Docket No. 50-395 Enclosure 3: Page 27 of 32 Westlugl1ous.: Non*Prnprietat)' Cla~i 3 LlR*SDA-21--023-NP 1<<:,*faion 0 Ma.'timum Allowable l:i:>ck,f-E"'.tluation Period Fl....- Size G.7 Allowab le f law Size Reached in 9.58 EFPY I
0.2
()
!l 1 3 4 S 6 7 8 1 Elfttfln JfuIJ Pul\tr Yn ,s (EfP\')
lgu r<' 7-1 Flaw Growth E,*alo ~tion Chart for SG Inlet Xozzlt' (T = 6l!J.4"F) Axial Flaw (AR=?)
For ,ll/oy I 52 PWSCC growth rote. a eom en-atiwi FOI of19 orcr tl:c Alloy I 82 crack gru111h rate is used Page27of32
Serial No.22-025 Docket No. 50-395 Enclosure 3: Page 28 of 32 Westinghouse Non-Piop1i,:tary C'las; 3 LTR-SDA-21.023-NP
~\uiouO l ~
lllXllllllill Allowable End-of*faal~ 1iou Period f law
,!().l;I I
- 0.5 t
1
~
i= - ~ - - - - - - - - - + - - - ~ 1 - - - - -- - - - - - - -! -- - - - - ,
L ________ _
(!_~ --- - - -
I I
0 0
r 3 4 ., f, EITm in Full Pow,r Yun {Efl'l')
7 8 '-'
~"llre 7-2 Flaw Growth Eval :u lon Chll rt ror SG In!M Xo:ztlt (I - 619.f'F) Clrto.mferenlfal Flaw (AR- 10)
For Alloy J51 PWSCC gro111/i rote, r. co11.<en*,1rii*<? FOJ of19 on>;- tl~ Alloy 1SJ at1di: gron th rate is used Pa~,: 28of32
Serial No.22-025 Docket No. 50-395 Enclosure 3: Page 29 of 32 Westinghouse Non-Prnprietary Cl:ns 3 LlR.-SDA-11-023-~
Re\':isonO 8.0 Summary and Condmiom A volumetric examination of the SG primary nozzle DM butt "'--elds *was perfonned in Fall 2018 at V . C.
Summer Unit 1, with no indications detected on any of the three inlet n022le DM \\'elds. The next required volumetric examination is planned during the Spring 2023 refueling outage, which is 4-5 ;--ears past Fall 2018. V. C. Summer Unit 1 is seeking rela.'Gltion from AS~ffi Code Ca;e N-770-5 [ l ], beyond the 5 }'!!:US to at least 8.48 EFPY (9 years) by perfonning a flaw evaluation to dem.omtrnte that the SG DM welds possess adequate thickn5~ to protect against fui1u.re due to PWSCC.
The P\v'SCC grow1h analyi.i,, herein is co!Lillitent v.ith the methodology in J\-Il..101260554 [6]. A 0.065 inch w ide surface flaw is po;rulated in the inlay and the amoilllt of time is detennined for the flaw to reach the ma:xinmm <tllowable end-of-e>-11.lnation period flaw size. This ma.'timum allowable end-of-enluation period flaw size would be the largest flaw size that could exi;t in the DM welds and be acceptable according to the Afil.{E Section XI Code [5]. Crack gro\\1h was calculated based on the PWSCC grm1;th mechanism through both the Alloy 152 inlay and the Alloy S-2 D).f \\'eld. The evaluation herein considered the P\VSCC crack gro\\1h through the Alloy 82 D).-! weld based on).IRP-115 [3], while fur the P\VSCC gr0\\1h though the .'\.lloy 152 wdd inlay, a factor of improvement of 29 over the crack growth rate for Alloy 182 based on
MRP-115 [3] was used. The justification for the use of the factor of improvement of29 for the Alloy 152 weld material is ba;ed on Yarious sourc5 from CUtTent industry data and laboratory rerults. The EPRI international expert panel provided the technical basis in MRP-386 [2] to demon..-trate that the Alloy 152 has a factor of impro\seillfll.t of324 oYer the PWSCC crack gro,;,ihof Allo-j 182. This FOI of324 has been incmporated into the ASME Section XI Code Case N-909 [4], based on the supporting data from l\.lRP-386. Separately, the staff's review of inlays as a mitigation tecb:niqlle for .'\.lloy 182, had consider a factor of improYement of 100 o \*er the P\VSCC rate of Alloy 182 [6] a; a baseline case for their coniinnatory siudy . The FOI of 29 also bounds more than 75% of the Alloy 152/52 test data per NUREG/CR-7103 Volume 4 [14], which is an NRC i.poru:ored ;tudy of Alloy 52/152 sec rates from PNNL .'\.i.'U., GEG, and CIEM..\.T laboratory resuln. Furthennore, the most recent PNJ\l data prffifnted at the Jan 13-14, 2021 EPRI Alloy 690/ 521152 PrimMy Water Stre;;; Corrosion Cracking Rfiearch Collaboratiw Meeting [15],
dem.on;trates that the use ofFOI of29 bounds all the sec data even \\ith 15'H, cold forging.
The results in Figures 7-1 and 7-2 demonstrate that it would take more than 8.4.S EFP¥ (9 years) for the postulated 0.065 inch deep axial and circumferential flaw in the V. e. Summer Unit 1 SG inlet nozzle DM weld inlay to grow to the maxinnim allowable end-of-evaluation period flaw size. Therefore, the V. C.
Summer Unit 1 plant specific analy,is performed herein jnstifi.5 an examination inter\lll of 8.48 EFPY (9
}--ears).
Operating experience and cutreni data indicate that Alloy 690 and associated weld metal Alloy 521152, are highly PWSCC resistant materials. These materials are also extremely resistant to PWSCC initiation with
,:e1y ,;low P\VSCC gro,,1h from starter indications. PWSCC initiatioo data from laboratory rh-ults ;upports that the Alloy 152 D)..f weld at V. C. Summer Cnit 1 \\ill take a significantly long time (more than 31 years up to e\'en 71 years) before any e>-idence of crack propagation is detected at the SG inlet nozzle DM locations.
Page29 of32
"' Th6 r..cool was iinal ap:,rov-< Materials Reliability Program: R.ecOlllll.li'llded FactOB of Improvement for Evaluating Primary Water Stre;;~ Conos.ion Cracking (PWSCC) Growth Rates of Thick-Wall Alloy 690 Materials and Alloy 52, 152. and Variants Welds (MRP-386). EPRI, Palo Alto, CA: 1017, 3002010756.
3. ~teri.als Reliability Program: Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (P\VSCC) ofAlloy 82, 182, and 132 Welds (MRP-115), EPRI, Palo Alto, CA: 1004. 1006696.
4. AS.ME Code Case N-909, "Primaty Waler Stres.s Corrosion Crack Gro,;\-ih Rate Curve5 for Alloy 690
~faterials and Associated Weld Material~ Alloy;; 52, 151, and Variant \Velds bpo.s.ed to Pressurized Wafer Reactor Environmem~ Section XI, Di1.ision 1,~ Approval Date Au,,."llst 10, 2020.
5. ASME Boiler & Pressure Vessel Code,Section XI, Rnles for ln..."ai.ice Inspection ofl\'uclear Power Plant Components, 1007 Edition with 2008 Addenda.
6. Rndl.and., Da,id Let al '*Evaluation of the Inlay Process as a l.\.fitigation Strategy for Primary Water Stress Corrosion Cracking in Pressuriz.ed Wale£ Reactors;* April 2010 (ML101260554).
7..
S.
9. .Malaials Reliabilit-y Pro~ Primary Water- Stress COll"Chion Cracking (PWSCC) Flaw Evaluation Guidance (?,,mP-287). EPRL Palo Alto. CA: 2010, 1021023.
10. Materials Reliability Program: Adt-anced FEA Evaluation of Gro\"\1h of P0o.tulated Circumferential P\VSCC Flaws in Pres.~ Nozzle D i ~ l\fetal Weld-, (}.fRP-216. Rev. I): Evaluatiom Specific to Nine Subject l'lani;;_ EPRL Palo Alto, CA: 1007. 1015400.
11. S. R. Mettu, l S. Raju, "Stress Inteniity Factorr. for Part-through Surface Cram in Hollow Cytindent Jointly developed under Grants N AS.A-ISC 25685 and Lockheed ESC 3012-t Job Order number 85-130. Call number 961'112214 (NASA-Thf-111707}. July 1992.
12. American Petroleum Institute. API 579-1:'ASME FFS-1 (API 579 Second Edition). ~Fime--;'1-For-Service.~ June 2007.
13. U. S. 1':"RC Memorarutum To: Timothy Lupold, Office of Nuclear Reactor Regulation, From: Jay Collins, Office of:t-.1JCl!!M Reactor Regulation, "Summaiy Of Public Meeting Between The Nuclear Regulatory Commission Staff And Irui~ b:j* Representativtt On Jmplen1emarion Of AS).fE Code Case N-770-t:* July 12. 2011 (MI.1122-t0SlS).
Page31 of32
"' llis n!CO<d was 3"0:tWed c 10.W2021 B.:52:57 AM. fllfs , titeree nt was a~ l;y tn.e PRll, E s"5!.em uoon 1s v~ l
Serial No.22-025 Docket No. 50-395 Enclosure 3: Page 32 of 32 Werunghouse Non-Proprietary Oass 3 LTR-SDA-21-023-~
Revi;ion O
14. U.S. N"RC. NUREG!CR-7103 Vol. -t *"Paci.fie Northwest National l.aboratoiy fu\*estigation of Stress Corro;ion Crad.:ing in Nickel-Base Alloys: Beh:11.ior of Alloy 152 And 52 Welds,~ April 2019 (M:L19099A200).
15. Toloczl:o, M .. Olszta, M., Ziqing, Z., Bouffioux, R, and 8llfellllller, S., -update on SCC Growth Rate Testing Research for Alloy 152(:M)/52(M) Welds at ~'L.'" Presentation pro'i.ided by l'Nll<'L at EPRl
,"-lloy 690.152.:1 51 PWSCC Researcli Collaboration Meeting. January 13-14, 2011 Web Meeting.
16. };faferials Reliability Program: Technical Ba.iis for Reexamination futerval Extension for Alloy 690 P\VR Reactor Vessel Top Head Penetration Nozzles (MRP-375). EPRl Palo Alto. CA: 201 4.
3002002441.
17. Sakima, K., Maegu.chi, T., et. al, ""An Update on Alloys 690f52i152 PWSCC Initiation Testing," 17th futemational Conf&eoce on F.nviromnental D egradation of?.faterials in Nuclear Power Systems Water Reacton. Augmt 9-13, 2015, Ontario. Cll1!ada
18. Wa.-... C., Foclit, E, Toloczko, 11,l, Ziqing, Z., Jenks, A., ""EPRI-NRC Cooperative Rfiearch Project:
P\VSCC Crack Initiation Characteazation of Alloys 600/182 and Alloys 690/52.152 - 2020 Statu.s Lpdate." Presentation pro,.ided at EPRI Alloy 690!52i152 PWSCC Re;,earch Collaboration :Meeting, Ja:mwy 13-14, 2011. Web Meeting.
19. Materials Reliability Program: Reactoc Vessel ClOfillre H ead P enetration Safety Assessment for U.S.
Pressurized Water Reactor (PWR) Plants (MRP-110): E,:allllliiom Supporting the ri.fRP Inspection Plan, EPRL Palo Alto, CA: 2004, 1009807.
20. }.,fateriaJs Reliability Program: Resistance to P rimary W ater Stress Comr,..ion Cracking of Alloys 690, 52, and 152 in Pres.surized Water Reactois 0-,mP-111). EPRI, Palo Alto, CA: 2004, 1009801.
21. ASME Code Case N-719-6, GAltema.ti.e E.'Wllination Requirements for PWRReactor Vessel Upptt Heads Wiih Nozzle"> Having Pressure-Retaining Pa:rtial-Peoerration WeldsSection XI, Division l,~
Approval Date ~farch_ 3, 2016.
22_ OEl 1505 - Hairline Crack Found in Weld Connecting RCS Hot Leg Pipe to Reactoc Vessel Nozzle_
Event No. 395-001007-1, faent Date: October 7. 2000.
23. License Event Report No. 50-338.'2012-001-00, "North Anna Power Station, Unit 1, Degraded Reactor Coolant System Piping Due To Primary Water Stress Corrooon Cracking." (MI..111.51A441).
Page31of32
" ' Th;!, F-"""1 was f nal l<!C 01ml co I0!w.1021 8:52:57 AM. !Tris stat-nl was a<l,'..ed l>v t!le PR J E svstem uron *1s vJl;datic.'ll

v t e Letter Sequence Request
EPID:L-2022-LLR-0033, Application for Alternative Request - Extension of Steam Generator Primary Inlet Nozzle Dissimilar Metal Weld Inspection Interval (Volumetric Examination) (Approved, Closed)
Initiation Request, Request, Request Acceptance Administration Withholding Request Acceptance Questions Answers Results Approval Other: ML22224A035
MONTHYEAR2022-03-10 [Table View]

ML22069B117

Text

Attachment:

Enclosures:

8.0 REFERENCES

8.0 REFERENCES

Navigation menu

Search