ML20141E922

Fulfills Utils Commitment to Provide Final Rept IAW Requirements of 10CFR50.54(f) Re Response to GL 92-01,Rev 1, Suppl 1 to Rv Structural Integrity

ML20141E922
Person / Time
Site:	Maine Yankee
Issue date:	05/13/1997
From:	Meisner M Maine Yankee
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-92-01, GL-92-1, MJM-97-004, MJM-97-4, MN-97-66, NUDOCS 9705210146
Download: ML20141E922 (73)

Text

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t i MaineYankee REllABLE ELECTRICITY SINCE 1972 329 BATH ROAD e BRUNSWICK, MAINE 04011 + (207) 798-4100 May 13,1997 MN-97-66 MJM-97-004

- UNITED STATES NUCLEAR REGULATORY COMMISSION 4 Attention: Document Control Desk Washington, D.C. 20555 ,

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References:

(a) License No. DPR-36 (Docket No. 50-309)

(b) USNRC Generic Letter 92-01, Revision 1, Supplement 1 dated May 19,1995 -

Reactor Vessel Structural Integrity l (c) MY Letter to USNRC dated August 17,1995 (MN-95-98)-Response to Generic Letter 92-01, Revision 1, Supplement 1: Reactor Vessel Structural Integrity (d) MY Letter to USNRC dated November 14,1995 (MN-95-124) -Response to Generic Letter 92-01, Revision 1, Supplement 1: Reactor Vessel Structural Integrity (e) MY Letter to USNRC dated October 28,1991 (MN-91-151) - Update of PTS Assessment to Address the Revised PTS Rule (10 CFR 50.61)

(f) MY Letter to USNRC dated June 22,1995 (MN-95-72) - Response to Generic Letter 92-01, Revision 1 (Reactor Vessel Structural Integrity) Maine Yankee Review Status (TAC No. M83479)

(g) MY Letter to USNRC dated October 29,1992 (MN-92-111) - Proposed Change No.173: Incorporation of Cycle-Specific Pressure-Temperature Limits into the Core Operating Limits Report (h) MY Letter to USNRC dated April 8,1996 (MN-96-043)- Proposed Change No.

173: Incorporation of Cycle-Specific Pressure-Temperature Limits into the Core Operating Limits Report

Subject:

Response to Generic Letter 92-01, Revision 1, Supplement 1 I Reactor Vessel Structural Integrity - Final Report Gentlemen:

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The USNRC Staffissued Generic Letter 92-01, Revision 1, Supplement 1, to resolve issues relating to reactor vessel integrity including weld chemistry variability, Reference (b).

Maine Yankee's response to Part 1 of the supplement (90 day response). Referer.ce (c), committed to respond to Part 2 (6 month response) as follows:

Provide an interim report addressing Items 2,3, and 4 by November 19,1995.

Provide a final report addressing Items 2,3, and 4 by April 30,1997.

We provided Maine (ankee's interim report in Reference (d). This letter fulfills Maine Yankee's commitment to provide a final report in accordance with the requirements of10 CFR 50.54(f).

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MaineYankee UNITED STATES NUCLEAR REGULATORY COMMISSIO'N MN-97-66 Attention: Document Control Desk Page 2 Maine Yankee's final report is provided in Enclosure 1. The report address Items 2,3, and 4 identified in Supplement 1 to GL 92-01, Revision 1. The report reflects the results of participation in Combustion Engineering Owner's Group (CEOG) tasks to (a) resolve the issue associated with weld chemistry variability and (b) develop generic Initial Upper Shelf Energy (USE) values.

Included in Enclosure 1 is Revision 2 to Maine Yankee's PTS Report which documents our evaluation of Chemistry Factors and Margins for beltline materials. Although we have made numerous adjustments to the parameters used to establish Adjusted Reference Temperatures (ARTS) and USE, the conclusions regarding Maine Yankee's RTersand USE at Expiration of License (EOL) remain valid. Specifically, we conclude the following:

our conclusion in Section 6.0 of the PTS Report, Reference (e), that "the PTS screening criterion would not be reached until 2037 A.D. if the plant continued operation,28 years beyond expiration of the current license," remains conservative.

the 1/4 T USE is not predicted to drop below 50 f1-lb before Expiration of License (EOL)

We also conclude that the ART relationships used to determine Appendix G limits in Technical Specification 3.4 and Low Temperature Overpressure Protection (LTOP) limits remain conservative.

Please note that the following items identified in Enclosure 1 require additional action by your stafr:

We assumed that your staff approved a modified surveillance capsule withdrawal schedule and revised the Final Safety Analysis Report (FSAR) accordingly. However, your staff's Safe;y Evaluation Report (SER) did not explicitly address the withdrawal schedule. Picase notify us if this assumption is contrary to your belief.

We have requested NRC approval of the use of the CEOG Generic Initial USE for Weld 2-203 A, B, C (Heat No. 51989 Linde 124). Your prompt review of Maine Yankee's use of the generic USE value would be appreciated. Following a favorable finding, we plan to revise our relief request from the minimum USE requirements of 50 fl-lb in 10 CFR 50, Appendix G, Reference (f), to a value of 42 ft-lb..

Maine Yankee submitted Proposed Change No.173 to incorporate cycle-specific Pressure-Temperature Limits into the Core Operating Limits Report (COLR) in 1992, Reference (g). In Early March 1996, Mr. Howard F. Jones, Jr. of our staff discussed the proposed change with Mr. E. H. Trottier and Mr. B. J. Elliot of your staff. Although we consider the proposed change administrative, your staff informed us that the NRC technical staff needs to review the basis for the limits before approval. We agreed in Reference (h) to postpone your review of the proposed change until completion of our response to Generic 92-01, Revision 1, Supplement 1. We request your stafi's review of Proposed Change No.173 since our commitment to respond to Generic 92-01, Revision 1, Supplement 1 is now fulfilled.

Please see Enclosure 1 for details relating to the above items.

k MaineYankee i i

UNITED STATES NUCLEAR REGULATORY COMMISSION MN-97-66 ,

Attention:- Document Control Desk Page 3 We look forward to your staffs prompt review of the enclosed information, resolution of outstanding

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. iss es and a favorable SER. ~

l We trust this information is satisfactory.

.V truly urs, y

l ael I. Meisner  !

i esident, Nuclear Safety & Regulatory Affairs l

Enclosure c: Mr. H. J. Miller Mr. D. H. Donnan i Mr. J. T. Yerokun  !

Mr. Patrick J. Dostie' 1 Mr. Uldis Vanags

- STATE OF MAINE Then personally appeared before me, Michael J. Meisner, who being duly swom did state that he is Vice President of Maine Yankee Atomic Power Company, that he is duly authorized to execute and l i file the foregoing response in the name and on behalf of Maine Yankee Atomic Power Company, L and that the statements therein are true to the best of his knowledge and belief. i

j. Notary Public M nl:a W. s Por.' ice~ rm ry Pubh.c l Mycomn$t '# i " ?n' Gx!Prca Gl3/ga ,

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l Enclosure 1 i Page1of10 MN-97-66 l

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l Maine Yankee Final Report Items 2,3, and 4 ,

Generic Letter 92-01, Revision 1, Supplement 1  ;

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Enclosure 1 Page 2 of 10 MN-97-66 Maine Yankee Final Report Items 2,3, and 4 .j Generic Letter 92-01, Revision 1, Supplement I l i

I BACKGROUND i The USNRC Staffissued Generic Letter 92-01, Revision 1, Supplement 1, Reference (b), to i

" require that all addressees identify, collect and report any new data pertinent to analysis of structural integrity of their reactor pressure vessels (RPVs) and to assess the impact of that data on their RPV integrity analyses relative to the requirements of Section 50.60 of Title 10 of the Code ofFederal Regulations (10 CFR 50.60),-

10 CFR 50.61, Appendices G and H to 10 CFR Part 50, (which encompass pressurized thermal shock (PTS) and upper shelf energy (USE) evaluations) and any 3

potential impact on low temperature overpressure (LTOP) limits or pressure-

[ temperature (P-T) limits." ,

The letter required written responses to Part (1) within 90 days and Part (2) within 6 months.

A Part 1 (90 day resnonsel Maine Yankee responded to Part 1 (90 day response) of the su)plement in Reference (c). A j Ascription of actions taken or planned by Maine Yankee to :ocate all data relevant to the l t : termination of RPV integrity was included.

Maine Yankee informed the staff that it was participating in a CEOG task expected to provide the -

basis for resolving the USNRC Staff's concerns relative to weld chemistry variability. Included in

- this effort would be a collection of relevant data for reactor vessel welds fabricated by CE.

A primary goal of the CEOG task was to establish uniform best-estimate copper and nickel contents -  !

and standard deviations for weld consumables. This infonnation should be able to address certain inconsistencies in the USNRC's Reactor Pressure Vessel Integrity Database (RVID). The volume of material which required reviewing, collation, and compiling into a comprehensive database and report dictated a schedule for the CEOG task of about eighteen (18) months.

Maine Yankee also informed your staffofplans to (a) meet with licensees that share similar welds, and (b) review industry and USNRC RPV databases, to locate all data relevant to the mechanical properties (Initial RTer and Initial USE) of Maine Yankee's beltline welds. The purposes of these activities would be to (a) use all relevant data to determine these properties, and (b) attempt unifonnity in these properties between licensees with similar welds. l B Part 2 (6 month resoonse) l Part 2 (6 mon *h response) of Generic Letter 92-01, Revision 1, Supplement 1, Reference (b) requires addressees to respond to the following items within six(6) months ofissuance: ]

hem Descriotion (2) an assessment of any change in best-estimate chemistry based on consideration of all relevant data; a

Enclosure 1 l

Page 3 of10 MN-97-66 Maine Yankee Final Report Items 2,3, and 4  :

Generic Letter 92-01, Revision 1, Supplement 1 (3). a determination of the need for use of the ratio procedure in accordance with the established Position 2.1 of Regulatory Guide 1.99, Revision 2, for those licensees that use surveillance data to provide a basis for the RPV integrity evaluation; and ,

-(4) a written report providing any newly acquired data as specified above and (1) the results of any necessary revisions to the evaluation of RPV integrity in  ;

accordance with the requirements of 10 CFR 50.60,10 CFR 50.61, Appendices G and H to 10 CFR Part 50, and any potential impact on the LTOP or P-T limits in the technical specifications or (2) a certification that previously submitted ,

evaluations remain valid. Revised evaluations and certifications should include -

consideration ofPosition 2.1 of Regulatory Guide 1.99, Revision 2, as applicable, and any new data. ]

Maine Yankee's response to Part 1, Reference (c), committed to respond to Part 2 of the supplement  ;

as follows:

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Provide an interim report addressing Items 2,3, and 4 by November 19,1995.

Provide a final report addressing Items 2,3, and 4 by April 30,1997.

.We provided Maine Yankee's interim report in Reference (d). This enclosure fulfills Maine ]

Yankee's commitment to provide a final report in accordance with the requirements of 10 1

' CFR 50.54(f). .i

, .II DISCUSSION A Item (21 - Assessment of Any Change in Best-estimate Chemistry Based on Consideration of All Relevant Data

-l This section summarizes changes in best-estimate chemistry based on consideration of all relevant i data. Included are both beltline welds and plates. j Beltline Welds Maine Yankee participated in C 50G Task 902 M to provide the basis for resolving the USNRC Staff's concerns relative to veld chemistry variabihty. The CEOG effort included a collection of relevant data for rcactor vessel welds fabricated by CE.

The primary goal of the CEOG task to " establish uniform best-estimate copper and nickel

. contents and standard deviations for weld consumables" was achieved. The resulting best-estimate copper and nickel contents and standard deviations should address pertinent inconsistencies in the USNRC's Reactor Pressure Vessel Integrity Database (RVID). The table '

. below compares the best-estimate copper and nickel contents previously provided in References (f), (g) and (k) for Maine Yankee's beltline welds to the CEOG Task 902 results M.

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Enclosure 1 Page 4 of10 MN-97-66 Maine Yankee Final Report Items 2,3, and 4 Generic Letter 92-01, Revision 1, Supplement 1 Previous CEOG

- Heat Number Flux Lotfl'ype ValuesM8 4 Task 902M l Weld Seam Cu,% Ni,% Cu,% Ni,%

2-203 A, B, C 51989 Linde 0.17 0.17 0.17 0.16 l 124/3687 3-203 A, B, C 12008,13253 Linde 0.22 0.84 0.21 0.87 1092/3833 .

13253m Linde .m .m 0.22 0.73 1092/3833 )

I 9-203 IP3571 Linde 0.31 0.76 m 0.28 0.75  ;

1092/3958 33A277W Linde W W 0.26 0.16 0091/3922 (1) Earlier PTS submittals did not identify heats 13253 and 33A277 in these weld seams since they were not the limiting heat.

(2) The best estimate Ni content for Weld seam 9-203 (1P3571) used in Reference (g) for P/r limits (0.74 %) was revised in Reference (k) to 0.76 % to reflect results of ar.:. lysis of Wall Capsule 253.

l Beltline Plates Maine Yankee also reviewed the basis for best-estimate copper and nickel contents used for the  !

beltibe plates. We conclude, based upon our review, that the best estimate Cu and Ni conn 'Rtions f~ /,a) Plates D-8406-1 & 2 and (b) Plates D-8407-1 & 2 should be the averages

ofindividual measurements conducted on the test block for each plate. This is because these two plates originated from the same heat. The table below compares the best-estimate copper and -

nickel contents previously provided for Maine Yankee's beltline platesR$ to the revised values.

Previous Revised Values Weld Seam Heat Number Values 58) (See Attachment A, Table 5.1)

Cu,% Ni,% Cu,% Ni,%

D-8406-1 B7955-1. 0.15 0.59 0.16 0.58 D-8406-2 B7955-2 0.17 0.56 D-8407-1 B8330-1 0.24 0.62 0.24 0.62 D-8407-2 B8330-2 0.23 0.62 s

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Enclosure 1 Page 5 of10

• MN-97-66 '

Maine Yankee t Final Report Items 2,3, and 4 Generic Letter 92-01, Revision 1, Supplement 1

' B Item (3)- Determination of the Need for Use of the Ratio Procedure -

We believe that a process similar to the ratio procedure is necessary to " normalize" the CF tables in

. . RG 1.99, Revision 2 (predictions) to credible surveillance measurements. This allows determination of the CFs for beltline materials with different Cu/Ni content using surveillance results. -

In the past, Maine . Yankee used the ratio procedure described in Regulatory Guide (RG) 1.99,

- Revision . 2 to determine the- Chemistry Factors (CF) for the limiting beltline material  ;

(Circumferential Weld 9-203 and Plate D-8407-2). We used the resulting Chemistry Factors to - i determine the heatup/cooldown limits in accordance with 10 CFR 50 Appendix G.(8M The ratio ,

procedure was not explicitly used for the remaining beltline materials.  !

We recently reviewed surveillance data applicable to Maine Yankee's beltline welds and plates to +

determine whether we should apply the ratio procedure to determine Chemistry Factors for  ;

other beltline weld seams and plates for determination of heatup/cooldown limits, and  !

beltline weld seams and plates for determination of RTns- #

Our review included data obtained from (a) Maine Yankee's surveillance program, (b) irradiation of Maine Yankee's beltline materials in test reactor, and (c) sister plant surveillance programs. We ,

concluded that we should apply the ratio procedure to determine the Chemistry Factor for four (4) i ofMaine Yankee's beltline plates and four (4) heats used in Maine Yankee's beltline welds, for both a

)" heatup/cooldown limits and PTS.  ;

1 Maine Yankee's revised PTS Report (Revision 2), Attachment A, documents the application of the  ;

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ratio procedure to determine the Chemistry Factors for beltline materials. Included is the .

determination of the margin term, M. The resulting Chemistry Factors and margins are applicable l

, to the determination of both RTns and heatup/cooldown limits. Please note that we did not apply i the ratio procedure in the previous PTS Report M since it preceded the latest PTS Rule change. The i

, previous PTS Rule was interpreted to require that licensees determine the Chemistry Factor from the tables rather than a fitting process similar to RG 1.99, Revision 2. Table I compares the Chemistry -j Factors used in Maine Yankee's revised PTS Report m to the previous values.

j C L Item (4)- Written Report Providing Anv Newiv Acauired Data L Item (4) of Part 2 (6 month response) of Generic Letter 92-01, Revision 1, Supplement 1, Reference I (b) requires addressees to provide the following: )

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"a written report providing any newly acquired data as specified above and (1) the results of any necessary revisions to the evaluation of RPV integrity in accordance with the requirements of10 CFR 50.60,10 CFR 50.61, Appendices G and H to 10 CFR Part 50, and any potential impact on the LTOP or P-T limits in the technical specifications or (2) a certification that previously submitted evaluations remain valid. Revised evaluations and I certifications should include consideration of Position 2.1 of Regulatory Guide 1.99, Revision 2, as applicable, and any new data."

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The sections below fulfill the requirement for this item.

Enclo.aie 1 Page 6 of10 MN-97-66 Maine Yankee Final Report Items 2,3, and 4 Generic Letter 92-01, Revision 1, Supplement 1 1.10 CFR 50.61. PTS Rule We have evaluated the impact of the best estimate Cu and Ni concentrations obtained from (a)

CEOG Task 902 N and (b) the 1996 revision to the PTS Rule, on Maine Yankee's PTS Submittal. W The results of our evaluation have been incomorated into Revision 2 of Maine Yankee's PTS Report, Attachment A. The conclusion in Section 6.0 of the previous PTS

. Report M that "the PTS screening criterien would not be reached until 2037 A.D. If the plant continued operation,28 years beyond expiration of the current license" remains conservative.8 Maine Yankee provided a tabular comparison of PTS licensing values between the USNRC's RVID and Maine Yankee's PTS ReportM in Reference (i). We have revised this table to include Maine Yankee's licensing values provided in the revised PTS Report, Attachment A.

Attachment B provides a copy of the comparison for your use in resolving discrepancies in licensing values. We look forward to a favorable SER for PTS.

2.10 CFR 50.60 and Apoendices G and H to 10 CFR Part 50 Aopendix H to 10 CFR Part 50 Maine Yankee concluded in Reference (j) that our surveillance program meets the requirements of Appendix H to 10 CFR Part 50. In addition, the test procedures and reporting requirements for all withdrawn capsules meet the requirements of E 185-82 to the extent practicable for the configuration of the specimens in the capsules.

We submitted a proposed Technical Specification change and a Surveillance Capsule Removal

' Schedule Report in Reference (k) to reflect the results of testing the last withdrawn capsule. The Surveillance Capsule Removal Schedule Report included a withdrawal schedule based upon the recommended withdrawal schedule provided in E 185-82. We also provided a modified withdrawal schedule, which postponed the removal of the next capsule by 4 EFPYs, for NRC review and approval. We justified the revised schedule based upon our ability to predict the RTwr of the limiting vessel materials. These prediction * :tse the results of both dosimetry (surveillance capsule and cavity measurements) and behavim f surveillance materialirradiated in four capsules removed from Maine Yankee as well as test reactor data on the same surveillance material. We received an amendment to our license and a Safety Evaluation Report (SER) relating to the submittal in Reference (1). However, the SER did not explicitly address the Surveillance Capsule Removal Schedule Report. We assumed that your staff approved the modified schedule, and revised the Final Safety Analysis Report (FSAR) accordingly. Please notify us if this assumption is contrary to your belief. Please note that adequate time remains to remove a capsule before the next scheduled date in the original schedule,2003 A.D.

Accendix G to 10 CFR Part 50 (USE)

We have evaluated the impact of the best estimate Cu concentrations of beltline materials

, obtained from CEOG Task 902 N on Maine Yankee's projected 1/4 T Upper Shelf Energy (USE). We have also reevaluated the initial USE for several beltline welds including the use of generic values based upon CEOG Task 839. ""M Table 2 provides the results ofour evaluation.

1 The revised year the PTS screening criterion is reached is 2038.

Enclosure 1 Page 7 of10 MN-97-66 Maine Yankee Final Report Items 2,3, and 4 Generic Letter 92-01, Revision 1, Supplement 1 We conclude, based upon the results of our evaluation summarized in Table 2, that the l 1/4 T USE is not predicted to drop below 50 ft Ib before Expiration of License (EOL).

l Please note that we have requested NRC approval of the use of the CEOG Generic Initial USE** for Weld 2-203 A, B, C (Heat No. 51989 Linde 124) in Reference (m).

Maine Yankee provided a tabular comparison of USE licensing values between the USNRC's RVID and Maine Yankee's values in Reference (m). We have revised this table to include Maine Yankee's licensing values provided in Table 2. A copy of the comparison is provided in Attachment C for your use in resolving discrepancies in licensing values. We look forward to a favorable SER for USE.

Anpendix G to 10 CFR Part 50 (P/T Limits) l Maine Yankee's Appendix G limits are implemented in Technical Specification 3.4. The i

. relationships for Adjusted Reference Temperature (ART) in Technical Specification 3.4 are j based upon measurement results from four (4) surveillance capsules and the application of RG 1.99, Revision 2. W We compared these relationships to the ART relationships derived in Attachment A using the current PTS Rule2 to determine whether the Technical Specification is ,

conservative. The results of this comparison demonstrate that Technical Specification 3.4 is conservative for projection ofART for the Surface,1/4T and 3/4 T locations into the future. We conclude that the ART relationships used to determine Appendix G limits in Technical -

Specification 3.4 and Low Temperature Overpressure Protection (LTOP) limits remain l valid. j l

l Maine Yankee submitted Proposed Change No.173 to incorporate cycle-specific Pressure- 1 Temperature Limits into the Core Operating Limits Report (COLR) in 1992, Reference (p).  ;

Your staffinformed us in early 1993 that they could not review the proposed change until the l NRC made a policy decision on removal of similar limits from the CE Standard Technical Specification (STS). Although many differences existed between our proposed change and the proposed improved STS, we acknowledged the need for the policy decision. The NRC made this i policy decision in January 1996. 1 In Early March 1996, Mr. Howard F. Jones, Jr. ofour staff discussed the proposed change with Mr. E. H. Trottier and Mr. B. J. Elliot ofyour staff. Although we consider the proposed change administrative, your staffinformed us that the NRC technical staff needs to review the basis for the limits before approval. Furthermore, your staffinformed us that the technical review would include issues identified in Generic 92-01, Revision 1, Supplement 1. We agreed to postpone your review of the proposed change until completion of our response to Generic 92-01, Revision ,

1, Supplement 1, in Reference (q). We request your staff's review of Proposed Change No.

173 since our commitment to respond to Generic 92-01, Revision 1, Supplement 1 is now fulfilled.

2 Note the current PTS Rule essentially mirrors RG 1.99, Revision 2

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( Enclosure 1 3 Page 8 of10 L

l MN-97-66 Maine Yankee

. Final Report Items 2,3, and 4 .

. Generic Letter 92-01, Revision 1, Supplement 1 -

REFERENCES  !

l- - (a) . = License No. DPR-36 (Docket No. 50-309) '

L  ; (b) . USNRC Generic Letter 92-01, Revision 1, Supplement I dated May 19,1995 - Reactor Vessel StructuralIntegrity (c). MY Letter to USNRC dated August 17,1995 (MN-95-98)-Response to Generic Letter 92-01, t Revision 1, Supplement 1: Reactor Vessel Structural Integrity

.(d) < MY Letter to USNRC dated November 14,1995 (MN-95-124)-Response to' Generic Letter l 92-01, Revision 1, Supplement 1: Reactor Vessel Structural Integrity ,

! (e) "Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds," CEOG  ;

I. Task 902, CE NPSD-1039, December 1996 l l . (f) MY Letter to USNRC dated October 28,1991 (MN-91-151) - Update of PTS Assessment to i Address the Revised PTS Rule (10 CFR 50.51) l (g) Letter MYAPCo to USNRC dated December 2,1988 (MN-88-116) - Proposed Change No.

145 - Combined Heatup, Cooldown and Pressure-Temperature Limitations (h) Letter USNRC to MYAPCo dated November 17,1989 - Issuance of Amendment No.114 to i Facility Operating License No DPR-36-Maine Yankee Atomic Power Station (TAC No.

71511): .

(i) Letter MYAPCo to USNRC dated May 20,1994 (MN-94-50) - Response to Generic Letter 92- -

l 01, Revision 1 (Reactor Vessel Structural Integrity) Maine Yankee Review Status (TAC No.

. M83479) .

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(j) Letter MYAPCo to USNRC dated July 2,1992 (MN-92-65) - Response to Generic Letter 92 . )

l . 01, Revision 1 (Reactor Vessel Structural Integrity) l

! .(k) Letter MYAPCo to USNRC dated September 30,1991 (MN-91-138) - Proposed Change No.

l 163: Revision to Combined Heatup, Cooldown and Pressure Temperature Limitations to j Reflect Analysis of Wall Capsule 253 1

_(1) Letter USNRC to MYAPCo dated February 5,1992, - Issuance of Amendment No.128 to.  !

Facility Operating License No. DPR-36 ' Maine Yankee Atomic Power Station (TAC NO. j

. M81815)-

L (m) MY letter to USNRC dated July 26,1995 (MN-95-85) - Response to Generic Letter 92-01, L Revision 1 (Reactor Vessel Structural Integrity) Maine Yankee Review Status (n)- NRC Letter to CEOG dated September 25,1996 - Safety Evaluation of Report CEN-622, a L ." Generic Upper Shelf Values for Linde 0091,124, and 1092 Reactor Vessel Welds," June i 1995: " Supplemental Information to C-E Owners Group CEN-622," June 1996 (o) CEOG Letter to USNRC dated December 4,1996 (CEOG-96-499) - Submittal of CEN-622 Approved Version . .

l (p) MY Letter to USNRC dated October 29,1992 (MN-92-111) - Proposed Change No.173:

l Incorporation of Cycle-Specific Pressure-Temperature Limits into the Core Operating Limits

! Report '

.(q) . MY Letter to. USNRC dated April 8,1996 (MN-96-043) - Proposed Change No.173:

Incorporation of Cycle-Specific Pressure-Temperature Limits into the Core Operating Limits

_ . : Report

_ (r) - : Letter MYAPCo to USNRC dated July 28,1994 (MN-94-77)- Response to Generic Letter 92-

' 01, Revision 1 (Reactor Vessel Structural Integrity) Maine Yankee Review Status (TAC No.

4 M83479) i j

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Enclosure 1 <

Page 9 of10 i MN-97-66 ,

Table 1 ,

i Comparison ofChemistry Factors

Revised PTS Report versus Previous Values i I .

Chemistry Factor, *F '

l Material Heat Number Flux Lot / Type Previous Revised PTS

., PTS Report (0 Report, Change ,

Attachment A i D-8406-1 B7955 Plate 109C) 127(4 +18 t D-8406 2 B7955-2 Plate 1240) 127(U +3

- D-8406-3. C3982-5 Plate 83 ") 83 C) 0 i D-8407-1 ' B8330-1 Plate 174C) 165C) -9 '

!. D-8407-2 - B8330-2 Plate 1690) 16548 -4 D-8407-3 B8324 Plate 92 0) 920) 0 I

2-203 A, B, C 51989 Linde 124/3687 89 0) 89 U) 0 I i

3-203 A, B, C 12008,13253- Linde 1092/3833 2220) 224(D +2 l 13253(2) Linde 1092/3833 - (') 203 (# .  !

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. 9-203- IP3571 Linde 1092/3958 206 (') 214 (4 +8 33A277(4 Linde 0091/3922 0) 99(4 -

(1) Chemistry Factor per paragraph (cXIXivXA) of 10 CFR 50.61 (Tables)

(2) Chemistry Factor per paragraph (cK2) of 10 CFR 50.61 (Derived)

(3) Earlier pts submittals did not identify heats 13253 and 33A277 in these weld seams since they were not the limiting heat.

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.i l Enclosure 1 '

• Page 10 of10 MN-97-66 Table 2-Upper ShelfImpact Energy (USE) Data for Reactor Vessel Beltline Plates and -

. Weld Seams for the Maine Yankee Atomic Power Plant

- End-of-Life' Predicted EOL -

L Plate or J - Heat No. Flux Type , Initial USE

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USE Source:  : 1/4T Fluence i 1/4T USEE

= 2

. Weld Seam . (ft-lbs)1  !(x10"n/cm )" (ft-Ibs)i i

D-84 %-1 B7955-1 N/A 115 Reference (i) 1.073 86 D-8406-2 B7955-2 N/A 86 Reference (i) 1.073 64 D-8406-3 C3982-5 N/A -90 Reference (i) ~1 .073- 71 D-8407-1 B8330-1 N/A 86- Reference (i) 1.073 - 57

- D-8407-2 B8330-2 N/A 82 Reference (i) 1.073 56 D-8407-3 B8324-1 N/A 82 Reference (i) -1.073 64 2-203 A,B,C 51989 Linde 124 84 Reference (m)(Generic) 0.882 - 59-3-203 A,B,C 12008,13253 Linde 1092 119 Reference (r) 0.882 79 3-203 A,B,C 13253 Linde 1092 111 Reference (o) 0.882 72 l 9-203 IP3571 Linde 1092 107 Reference (i) 1.073 -61 9-203 33A277 Linde 0091 160 Reference (o) 1.073 94 I

r:\ne\hfj\97-011

. . . - . . . _ . ~ . _ . , ,- - - - . . . - - _

- .. - . . ~ . - . . - - - - - - . . . . . - . . - . - . - . . . . . - . . ~ . .- -....

l~

i  ;

Attachment A of Enclosure 1 MN-97-66 l-r i l

i l l t

k Attachment A (

Maine Yankee '

Final Report Items 2,3, and 4 - i Generic Letter 92-01, Revision 1, Supplement 1 Results of an Assessment of i Reactor Pressure Vessel Beltline Materials Required by 10 CFR 50.61 (Pressurized Thermal Shoc~k Rule) >

for the Maine Yankee Atomic Power Plant Rev.2

' April 1997 -I t

i 1

1 l

l-

-r:\se\hfj\97-011

-nm.. ,y p-. -, u ~

I Results of an Assessment o'f Reactor Pressure Vessel Beltline Materials Required by 10 CFR 50.61 (Pressurized Thermal Shock Rule)

.for the Maine Yankee Atomic Power Plant Rev. 2 l l-April 1997 l l l l

Prepared by.: copy signed by Stephen T. Byrne ABB-Combustion Engineering Prepared by: h hbM V H. F. Jones. Jr P Principal Nuclear Engineer

(

i Reviewed by: t P.'J. Plante ' l Principal Engineer  !

-Reviewed by: 4<

<= 4

1. &am / c- //

E. C. Biemiller Principal Engineer. Yankee Atomic Electric Company Approved by: * '

/#7

! Robert Fraser. Manager Design Engineering r:\ne\nfji97-014

i ABSTRACT l

This' report provides results of an assessment of Reactor Pressure Vessel (RPV) materials to respond to the requirements of the latest revision of 10 CFR 50.61, " Pressurized Thermal Shock. Rule" uz) Included are values of t

RTp13 for April 30, 1997 and expiration of License on October 21, 2008. The j report also specifies the bases for values of RTp13 These results demonstrate that at expiration of License, a minimum margin of 34 F exists l I to the Pressurized Thermal Shock (PTS) screening criterion. Furthermore, the results demonstrate that the PTS screening criterion would not be ,

reached until 2038 A.D. if the plant were to continue operation 30 years l beyond expiratien of current License. The United States Nuclear Regulatory Commission Staff (USNRC) concluded in Reference (1) that values

of RTers below the PTS screening criterion present an acceptably low risk of vessel failure from PTS events. Therefore, PTS is not a significant safety concern for the Maine Yankee Atomic Power Plant.

l I

11 r;\ne\hfj\97-014 l

i

l l

l' TABLE OF CONTENTS I

P_a92 L ABSTRACT ii LIST OF TABLES iv

! LIST OF FIGURES v ,

l 1

l

1.0 INTRODUCTION

2.0

SUMMARY

OF RESULTS 2 .,

3.0 REACTOR PRESSURE VESSEL DRAWING AND WELD MAP 3 l

6 4.0 BEST ESTIMATE PROJECTIONS OF FAST NEUTRON FLUENCE 4.1 Basis for Projections of Fast Neutron Fluences 6 4.2 Assumptions About Core loadings for Future Operation 7 5.0 MATERIALS PROPERTIES 9 5.1 9 Initial RTuor Values 5.2 Chemistries 9 5.3 RTris Values for April 30. 1997 and 10 l Expiration of License 5.4 Time to Reach Screening Criterion 10

6.0 CONCLUSION

14

7.0 REFERENCES

15 APPENDICES A. Excerpts from Maine Yankee Cycle 15 Core Performance A-1 l Analysis Report B. Derivation of Chemistry Factor and Margins for Maine B-1 l Yankee Vessel Beltline Material l I

I

.r:\ne\hfj\97-014 iii

-- -- .- . . ~ . . . -- -

LIST OF TABLES Mpp_br Title Paae 5.1 Initial Data for Reactor Vessel Plates and Weld 11 Seams for the Maine Yankee Atomic Power Plant 5.2 RTp;3 Data on April 30, 1997 and Expiration 12 l of License for Beltline Plates and Welds for the Maine Yankee Atomic Power Plant ,

5.3 RTp1s Projections for Beyond Expiration of License for 13 l Beltline Plates and Welds for the Maine Yankee Atomic l Power Plant l B-1 Chemistry Factor Derivation Maine Yankee Weld 9-203 B-16 l B-2 Chemistry Factor Derivation Maine Yankee Experimental B-17 l Weld Data l B-3 Chemistry Factor Derivation Using Kewauee Data B-18 l B-4 Chemistry Factor Derivation Maine Yankee Plates B-19 _l D-8406-1 and D-8406-2 l B-5 Chemistry Factor Derivation Maine Yankee Plates B-20 l  ;

D-8407-1 and D-8407-2 l  :

B-6 Chemistry Factor Derivation Maine Yankee Weld B-22 l  ;

3-203 A. B. C with !! eat 13253 - l l l

B-7 Chemistry Factor Derivation Maine Yankee Weld B-23 l 9-203 with Heat 33A277 l B-8 Chemistry Factor Derivation Maine Yankee Weld B-24

[

3-203 A B. C with Heats 12008 & 13253 l j I

, I i l L

I l

r\ne\hfj\97-014 iv l

LIST OF FIGURES Number Title Ease 3.1 Section Drawing Depicting Material and Weld 4 Seam Identification for the Maine Yankee Reactor Pressure Vessel 3.2 Maine Yankee Reactor Pressure Vessel Map 5 4.1 Maine Yankee Atomic Power Plant Maximum Fast 8 i

Neutron Fluence (E greater than 1.0 MeV.)

Versus Year of Operation B-1 Comparison of Experimental and Surveillance Data B-25 l for Maine Yankee RPV Plates l l

B Comparison of Surveillance and Experimental B-26 l Irradiation Data for Maine Yankee Weld 3-203 A. B. C l l  ;

l I

l l

l

(

r:\ne\hfj\97 014 V L -  :

1.0-INTRODUCT10N-t i

. This report provides results of an assessment of RPV materials to respond to the requirements of the latest revision of 10 CFR 50.61(b)(1) uz) This  !

includes projected-values of RT,13 (at the inner vessel surface) of RPV  !

beltline materials ~ for April ' 30,1997, and for expiration of License I (October 21. 2008) as well as the date when the critical weld reaches the .

PTS screening criterion. .This report also specifies' the bases for the projection including the following: ,

(a) Fluence _ values for materials at specified times. during plant 1 lifetime [the values of "f" specified in 10 CFR 50.61(c)(1)(iv)(B)]. l  !

(b) The bases for the fluences for future operation, including .

assumptions about core loadings, required by 10 CFR 50.61(b)(1).

1 l

.(c) Initial reference temperatures. (Initial RTuoi values) of the unirradiated materials [the value of "RTwouu)" specified in 10 CFR l 50.61(c)(1)(i) and (ii)]. l l q

(d)- The chemistry factors based on best estimate weight percent copper t and nickel in the materials [the values of CF specified in 10 CFR 50.61(c)(1)(;v)(A) and (c)(2)(ii)(A)]. l (e) The equation specified in 10 CFR 50.61(c)(1)(v) used to determine l RTns for each material, and l-(f) Margin added to cover uncertainties [the values of "M" specified in 10 CFR 50.61(c)(1)(iii) and (c)(2)(iii)]. l i

l l

p 4

4

~

a.

l rune \hfj\97-014 1

,P .r

- +

i l

i 2.0

SUMMARY

-OF RESULTS The results of.this assessment demonstrate that at expiration of License, f a minimum margin of 34*F exists to. the. PTS screening _ criterion. l l.

Projection beyond expiration of current License indicate that the PTS -  ;

screening criterion would.not be reached until approximately 2038 A.D.. -l assuming continued plant operation. -This is the time the Longitudinal l. l Weld 3-203'A..B. C (Heat-12008. 13253) reatnes 270*F. Furthermore, the [ l circumferential weld seam between the middle and lower shells (9-203) l.

~

reaches 300*F .in approximately 2113 A.D. Finally, the results indicate j the first- plate projected to reach the screening criterion is plate -l

< D-8407-2 in approximately 2193 A.D. l l

The USNRC staff concluded in Reference (1) that values of RTpu below the I

screening criterion present an acceptably low risk of vessel failure from PTS events. Therefore. PTS is not a significant safety concern for the  !

{

Maine Yankee Atomic Power Plant. ,

-I 4 i

i 1

N l

l f

l .

i ,

s rdne\hfj\97014 2

3.0 REACTOR PRESSURE VESSEL DRAWING AND WELD MAP A section drawing depicting the material and weld seam identification of the Maine Yankee RPV is provided in Figure 3.1. A vessel weld map is provided in Figure 3.2. Included in the vessel weld map is the outline of the active region of the core which defines the beltline of the RPV.

There are six plates within the beltline (D-8406-1, 2. and 3: and D-8407-1, 2 and 3). These plates are joined by seven weld seams (2-203 A.

B. C: 3-203 A. B. C: and 9-203). Subsequent sections of this report will concentrate on these materials.

l l

l I

1

)

i i

I i

I l

l-s

. I i

r:\ne\hfj\97 014 3

FIGURE 3.1 SECTION DRAWING DEPICTING MATERIAL AND WELD SEAM i IDENTIFICATION FOR THE MAINE YANKEE REACTOR PRESSURE

j. VESSEL

{ REACTOR VESSEL BELTLINE MATERIALS 4

NOT SHOWN m m INTERMEDIATE SHELL WELD SEAM No. 2-203C f -

pQN LOWER SHELL #

WELD SEAM No. 3-2038 g

WELD SEAM No. 3-203C -

G , C

j. PLATE No. 08407-1 l , ,

c" Q g - ;5 i

A L

~

l 1

) . (

. V UPPER TO INTERMEDIATE SHELL GIRTH SEAM 1' 3

./ INTERMEDIATE SHELL LONGITUDINAL WELD l

WELD No. 8-203 s #

N SEAM No. 2-2038 INTERMEDIATE SHELL PLATE No. 08406-3

)N _* d g\p INTERMEDIATE SHELL PLATE No. 08406-2

/ /

INTERMEDIATE SHELL /

LONGITUDINAL WELD / " -

p INTERMEDIATE TO LOWER SHELL GIRTH SEAM <

l SEAM No. 2-203A WELD No. 9-203 INTERMEDIATE SHELL PLATE No. C8406-1

' LOWER SHELL

' LOWER SHELL PLATE - 7 ' PLATE No. 08407-3 No. 08407-2 g

~ '

LOWER SHELL LONGITUDINAL WELD , s SEAM No. 3-203A

\

~

REACTOR VESSEL

. _ . . . _ . . _ _ _ . . _ . ._ _ . _ . . _ _ _ . _ . _ . _ . . _ _ _ . _ . _ _ . _ _ . . ~ . _ _ .

L FIGURE 3.2 Maine Yankee Reactor Pressure vessel Map bw w hb n o e- o u

55

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4

'4.0 BEST ESTIMATE PROJECTIONS OF tLT NEUTRON FLUENCE ]

Best estimate projections _ of fast neutron fluences (E greater than 1.0 MeV.) of the RPV are provided in Figure 4.1. These values correspond to )

the clad base metal interface on the inside surface of the vessel where the materials receive the highest fluence. Included are projections for

- both (a) plates and circumferential weld seam 9-203, and (b) longitudinal weld seams-2-203 A. B. C and 3-203_A. B. C. The projections of fluence for the plates end circumferential weld seam 9-203 correspond to the peak .

vessel fluence to which a portion of: the . material is exposed. The projections for the longitudinal weld seams corresponds to the fluence 10 degrees f. rom the peak fluence location (location opposite core flats) to which one of three seams (A. B or C) are exposed. The remaining two seams for each type weld have somewhat lower fluences.

The projection of fluences provided in Figure 4.1 include consideration of )

.the flux reduction program committed to in Reference (2). These measures were demonstrated in Reference (?. to be adequate to meet the screening criterion at expiration of License.

4.1 Basis for Projections of Fast Neutron Fluences The basis for projections of fast neutron fluences used in this assessment is the result of. an analysis of fast neutron flux (E greater than 1.0 MeV.) levels in the RPV performed by Westinghouse Electric Corporation (W). These results are reported in Raference (4).

+

W calculation results have been used in previous PTS submittals. 3 Reference (9) and (17). Changes to the projections from Reference J L (9) to Reference - (17) reflect surveillance capsule and cavity I measurements through Cycle 11 operation and a modified treatment of the . axial fluence profile. The W calculation results used in l f

Reference (17) were approved by the USNRC for application to l Pressure / Temperature Limits'in Reference ((16). The estimated Flux l 1 Reduction Factors (FRF) achieved in design of Cycles 7 through 15 l h are provided in Section 4.10 of Reft..ence (5). Excerpts from

~

l r:\ne\hfj\97 014 6

l Reference (5) are provided in Appendix A. The FRFs assumed in Cycles

, 16 and all subsequent cycles correspond to the target FRFs provided l 1 in Reference (2).

The projections assume a licensed core power level of 2700 MWth, a bounding Cycle 8 source distribution, and an average capacity factor for the remainder of operation of seventy five percent (75%). The average capacity factor of seventy-five percent (75%) is based upon

-historical data'and is consistent with the capacity factor for the period beginning October 25, 1985 (startup of cycle 9), and ending June 30.1990 (startup . of cycle 12). This ' assumption remains j conservative based upon. review of recent operation (Cycles 12 l through 15). I

~4.2 Assumptions About Core loading for Future Operation The projections of fast neutron fluences provided in Figure 4.1 assume the target FRFs provided in Reference (5) are met or exceeded for Cycle 15 and-all subsequent cycles. A comparison of estimated l FRFs for Cycle 15 to the target FRFs is provided in Section 4.10 of l Reference (5). Excerpts from Reference (5) are provided in Appendix A. Results of.this comparison indicate-that the target FRFs are exceeded ' by a significant margin. Similar results have been obtained in fuel management. scoping calculations for future cycles.

, Therefore, the projections adequately envelope future operation. We will continue to assess the effectiveness of flux reduction measures in future cycles to assure the target FRFs are achieved on the average.

t r:\ne\hQ\97014' 7

.-- - .= .. . . - - - - . ~ . . . . - . . -.. .

l *

)

l l 1

l 1

i FIGURE 4.1 i

Maine Yankee Atomic Power Plant l Maximum Fast Neutron Fluence (E> 1.0 MeV)

L - Versus Year of Operation l

l 4 -.

[

3.5 -

3 - /

/.

~

/

5 /

/

/

2.5 -

/

/

/

x / '

, :: . / l

>"- 2 - / I

/ )

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• 1.5 -

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/

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'yy' 1 -

FluenceA(Year Where and 8)are = A + prov(ded 8 below. Year - 1991) l i 3 v y '

Legend:

gefficlentg:

3 N

-0.5 - All Plates & Weld 9.203 UT(8 0 SIG .

--. velds 2-203A 8, & C & 0.780 0.039 1 3-203A, 5, & C j t,..

0 . . .

1990 2000 2010 2020 2030 2040 2050 Beginning of Calender Year

( .

l

-I I

5,0 MATERIALS PROPERTIES l 1

5.11 Initial RT 7 Values l l I l A summary of the initial rte 1 values for all plates and weld seams i in the beltline is presented in Table 5.1. (RTer values and other l properties for plates and welds immediately above the beltline are l also provided in Table 5.1 for information.) The initial rte 1 l values for the plate materials are based on test results evaluated l using Reference (6). The initial RTer values for all but one of the l weld seam materials are based on the generic mean of -56 F for Linde l 124 and Linde 1092 flux welds in accordance with paragraph l (c)(1)(ii) of Reference (12). A measured value based on drop weight I and Charpy data from Reference (14) was used for heat 33A277 in weld l seam 9-203. l l

5.2- Chemistries l l

A summary of the best estimate copper and nickel chemical content l for the plates and weldments is shown in Table 5.1. Sources for- l chemistry values for the plates are References (7) and (8). In the l cases of plates D-8406-3 and D-8407-3, each plate was from a unique j heat of steel. Therefore, the best estimate copper and nickel was l based on the measured values as previously reported (References 7 l and 8). In the case of plates D-8406-1 and D-8406-2, both plates l were from the same heat of steel. #B7955: samples for chemical l analysis were taken from one section of- each plate used .for ASME l Code qualification and the surveillance program, Given that both l plates D-8406-1 and D-8406-2 originated from the same heat, the best l estimate for the two plates in the reactor vessel intermediate shell l 1s the average of the two measurements (i.e., an average for the l heat). [ Note: The plate-specific measurements are_ given_ in a l footnote to Table 5.1. The heat average provides a more l conservative estimate for Plate D-8406-1 than the copper and nickel l based on the plate-specific measurement.) For plates D-8407-1 and l D-8407-2, both plates were from the same heat. #B8330: the samples l w

for chemical analysis were taken from one section of each plate (as l

, described above). Following the~ same, logic the best estimate for l r;\ne\hfj\97 014 9

= _ - _ _ - - _ _ _

plates D-8407-1 and D-8407-2 in the reactor vessel lower shell is l also the average of the two measurements representing that heat. l Therefore, the best estimate copper and nickel content reported in l Table 5.1 for these plates are the average of the two measurements. l l

The source for the best estimate chemistry values for the welds is l Reference (13). These best estimate copper and nickel values are l the mean of measured values in accordance with paragraph j (c)(1)(iv)(A) of Reference (12). l l

5.3 RTris Values for April 30, 1997 and Expiration of License l l

A summary of RTp13 values predicted using Reference (12) for the RPV l beltline plates and weld seams is given in Table 5.2. Values are j shown for April 30. 1997 and at expiration of license (October 21. I 2008). The limiting materials of the vessel with respect to PTS l considerations are weld seams 3-203 A. B. C (Heat 12008, 13253). I weld seam 9-203 (Heat IP3571) and plate D-8407-2 (Heat B8330-2) {

because they come the closest to the PTS screening criteria at l expiration of license. However, the most limiting of those l l materials is still 34 F below the screening criterion. l l

5.4 Time to Reach Screening Criterion l l

Projections of the time for the critical materials to reach the j screening criterion are summarized in Table 5.3. Weld seam 3-203 l A.B.C (Heats 12008 & 13253) is projected to have a RTp13 of 270 F in l the year 2038 A.D. Furthermore, circumferential weld seam 9-203 l ,

(Heat 1P3571) is projected to have a RTp13 of 300*F in the year 2113 l A.D. Finally. Plate D-8407-2 (Heat B8330-2) is projected to have a l RTpis of 270*F in the year 2193 A.D. l l

r:\ne\hfj\97-014 10

TABLE 5.1 Initial Data for Reactor Vessel Plates and Weld Seams for the Maine Yankee Atomic Power Plant Initial RTc Best Estimate Chemistry l j

Plate or Weld Heat No. Flux Type / Lot No. RT. Source  % Cu %Ni Source Seam Beltline- l D-Ba0f-1 B7955-1 N/A -10*F ceasured. Ref. 7 0.16'2' O.58'2' Ref. 7.8 D-8496-2 B7955-2 N/A 0'F Ref. 6.7 0.16'2' O.58'2' Ref. 7.8 D-f'%6-3 C3982-5 N/A ll F Ref. 6.7 0.12 0.62 Ref 7.8 D-8407-1 B8330-1 N/A -20*T Ref. 6.7 0.24"' O.62'3' Ref. 7.8 D-8407-2 B8330-2 N/A 2'F Ref. 6.7 0.24'3' O.62"' Ref. 7.8 {

D-8407-3 B8324-1 N/A 0*F Ref. 6.7 0.13 0.65 Ref. 7.8 j 2-203 A.B.C 51989 LINDE 124/3687 -56*F Ref.12, para 0.17 0.16 Ref. 13 (c)(1)(ii) 3-203 A.B.C 12008.13253 LINDE 1092/3833 -56'F Ref. 12. para 0.21 0.87 Ref. 13 (c)(1)(ii) i Ref. 13 13253 LINDE 1092/3833 -56*F Ref. 12. para 0.22 0.73 (c)(1)(ii) 9-203 1P3571"' LINDE 1092/3958 -56*F Re' M 16 0.28 0.75 Re f. 13 33A277"' LINDE 0091/3922 -60*F measurco, icf. 14 0.26 0.16 Ref. 13 Above the Beltline l D-8405-1 C3918-1 N/A 0*F Ref. 6.7 I 0.17 0.51 Ref. 7 D-8405-2 C3918-2 N/A 40*F Ref. 6.7 0.17 0.54 Ref. 7 D-8405-3 C3932-2 N/A 20*F Ref. 6.7 0.17 0.58 Ref. 7 1-203 A.B.C 12008.13253 LINDE 1092/3791.3833 -56*F Ref.12. para 0.21 0.87 Re f. 13 (c)(1)(ii) 13253 LINDE 1092/3791.3833 -56*F Ref.12. para 0.22 0.73 Ref. 13 (c)(1)(ii) 8-203 20291 LINDE 1092/3833 -56*F Ref. 12. para 0.22 0.74 Ref. 13 (c)(1)(ii)

"' Host of wcid seam 9-203 was deposited using heat IP3571 with LINDE 1092 flux.

<r' Average for heat B7955, plate specific results are 0.15% Cu and 0.59% Ni for D-8406-1. anj 0.171 Cu and 0.56% Ni for D-8406-2.

"' Average far heat B8330: plate specific results are 0.24% Cu and 0.62% Ni for D-8407-1, and 0.23% Cu and 0.62% Hi for D-8407-2.

r:\ne\hfj\97-014 11

^

f' _

' TABLE 5.2 -.

RT,n Projections for April 30. 1997 and Expiration of License for Beltline Plates and Welds for the Maine Yankee Atomic Power Plant

-April.30. 1997 . Dpiration of-License (10/21/2003)

Plgteor Initia .M Fluence" RTm ' Fluence""' RTm RTm Screening R I ,. ,1 )- 4 (.f.

or Seam F)' l CF -' 1 (x10"n/cm') (*F) (x10"n/cm') -(*F) Criteria ('F),

l

, D-8406-1 -10 17 127"' 1.20 140 1.80 154 270 l l D-8406-2 0 17

127"' 1.20 150 1.'80 164 270 l D-8406-3 0 34 83"' 1.20 121 1.80 130 270 l I

D-8407-1 -20 17"' 165"' 1.20 170 1.80 189"' 270 l 0-8407-2 2 17"' 165"' 1.20 192 1.80 211'" 270 l D-8407-3 0 34 92"' 1.20 131 1.80 141 270 l 1

2-203 A.B.C -56 66~ 89"' O.99 99 1.48 109 270 l 3-203 A.B.C"' -56 44"' 224"' O.99 211 1.48 236 270 l 3-203 A.B.C<r' -56 44"' 203"' O.99 190 1 48 213 270 l

. I 9-203 -56 44"' 214"' 1.20 213 1.80 -237 300 l 9-203"' -60 28"' 99"' 1.20 72 1.80- 83 300 l l

Notes:

(1) Weld with heats 12008 and 13253 (6) RT,n=147'F based on test reactor data (CF = 128.8 and M=17)

(2) Weld with heat 13253 (7) RT,n-169"F based on test reactor data (CF = 128.8 and M=17)

(3) Weld with heat 1P3571 (8) See Appendix B for derivation based on surveillance data.

(4) Weld with heat 33A277 (9) Chemistry factor per paragraph (c)(1)(iv)(A) of Ref.12 (5) Margin per paragraph (c)(1)(iii)- (10) See Footnote 1 of Table 5.3 for estimation basis.

of Ref. 12.

I r:\ne\hfj\97-014 12

TABLE 5.3 l l

RTru Projections for Beyond Expiration of License l for Beltline Plates and Welds for the Maine Yankee Atemic Power Plant l l

l Fluence 9- Screening l Dates (AD) (x10"n/cm )

2 EFPYs Material Criterion ( F) l 2038- 2.63- 48 Critical Longitudinal 270 l weld (3-203 A.B.C. Heat l 12008,13253) l l

2113 6.80 105 Critical circumferential 300 l weld (9-203. Heat 1P3571)  !

l

'2193 10.7 164' Critical Plate (D-8407-2) 270 -l l-l l

(1)' ' Fluence values are calculated from the following equations based on a j licensed core power level of 2700 MWth, a bounding Cycle 8 source l distribution, and an average capacity factor of 74% for the remainder .l of operation: l l

(a) Plates and circumferential welds - peak vessel fluence l l

F(x10"n/cm2 ) = 0.948 + 0.048 (YEAR - 1991), j l

(b) Axial (longitudinal) welds - 10 from flats. l l

2 F(x10"n/cm ) " 0.780 + 0.039 (YEAR - 1991). l l

Where:

l

~ l t YEAR is the year of interest rounded to the nearest 1/10th of a l year. For example. 10/1/91 is 1991.7 and 10/21/2008 is 2008.8. l r:\ne\hfj\97 014 13

6.0 CONCLUSION

S The USNRC Staff concluded in Reference (1) that values of RTers below the Pressurized Thermal Shock (PTS) screening criterion present an acceptab'ly low risk of vessel failure from PTS events. Results of the assessment provided in Section 5.0 demonstrate that the PTS screening criterion would not be reached until 2038 A.D. if the plant continued operation. 30 years l beyond expiration of the current License. Therefore. PTS is not a significant safety concern for the Maine Yankee Atomic Power Plant.

1 r:\ne\hfj\97-014 14

J 7;0 REFERENCES

1. " Policy 1ssue for Pressurized Thermal Shock (PTS)", SECY-82-465, November 23, 1982.

2. MYAPCo Letter to USNRC dated April 22. 1983 (MN-83-76).

3. MYAPCo Letter to USNRC dated June 1. 1984 (MN-84-88).

4. S. L. Anderson. " Summary of Fast Neutron Exposure Evaluation for the Maine Yankee Reactor Pressure Vessel" WCAP-11335. Rev.1. April 1991.

5. " Maine Yankee Cycle 15 Core Performance Analysis", YAEC-1907. Revision l

3. April 1996. I

6. USNRC Standard Review Plan - Branch Technical Position MTEB 5-2, NUREG-0800 Rev. 1, 1981.

7. " Evaluation of Pressurized Thermal Shock Effects Due to Small Break LOCA's with Loss of Feedwater for the Maine Yankee Reactor Vessel".

CEN-189, Appendix C, December 1981.

i

8. " Summary Report on Manufacture of Test Specimens and Assembly of  ;

Capsules for Irradiation Surveillance of Maine Yankee Reactor Vessel Materials", CENPD-37, December 30, 1971.

1

9. MY Letter to USNRC dated December 6,1988 (MN-88-118) Update of Assessment 10CFR50.61 Fracture Toughness for Protection Against l Pressured Thermal Shock. -

L

10. MYAPCo Letter to USNRC dated January 21. 1986 Response to Requirement l of 10CFR50.61.

I

11. S.T. Byrne. H.F. Jones. Jr. and P.J. Plante. " Reactor Pressure Vessel Beltline Materials Assessment Maine Yankee Atomic Power Plant". Rev. 1. l August, 1991.

I I

r:\ne\hfj\97 014 15 .

l l

12. 10 CFR 50.61 " Fracture Toughness Requirements for Protection Against l Pressurized Thermal Shock Events", Federal Register Vol. 60. No. 243 l dated Tuesday, December 19, 1995 .pages 65468-65472. l

13. "Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel l Welds." CE-NPSD-1039. December 1996. l l

14. J. A. Davidson, et. al. " Southern Alabama Power Co. Joseph M. Farley l Nuclear Plant. Unit No.1, Reactor Vessel Surveillance Program", WCAP- l 8810 December 1976. l l

15. Attachment F to MYAPCo Letter to USNRC, " Proposed Change No.145 - l Combined Heatup. Cooldown and Pressure-Temperature Limitations " dated l December 2, 1988 (MN-88-116). l l

16. USNRC Letter to MYAPCo. " Issuance of Amendment No.114 to Facility l Operating License No. DPR Maine Yankee Atomic Power Station (TAC l No. 71511) " dated November 17, 1989. l l

17 MY Letter to USNRC dated October 28, 1991 (MN-91-151) - Update of PTS l Assessment to Address the Revised PTS Rule (10 CFR 50.61) l l

l r:\ne\hfj\97 014 16

h E

APPENDIX A I

l Excerpts From Maine Yankee Cycle 15 l q Core Performance Analysis Report l

l i

1 YAEC-1907. Rev. 3 l l

l (Reference 5) l

)

l i

l i

1 r:\ne\hfj\97 014 A-1

L Excerpts from Maine Yankee Cycle'15 l Core Performance Analysis Report  ;

YAEC-1907. Rev. 3 l  :

(Reference 5) l l i

, 4.10 Pressure Vessel Fluence A program for reduction in pressure vessel fluence has been in place since Cyde 7 to address Pressunzed Thermal Shock (PTS) concerns. The Cycles 7 through 15 core designs have been a progression of low-leakage loading pattems, with particular emphasis on the high fluence area from 0 to 10' from a perpendicular line to the core shroud flats. The core shroud flats are the j core boundary lines defined by assembly locations 1 and 2 (or 45 and 54) in Figure 3.3. j j

'l

, The program for fluence reduction has been detailed in (5.7,5.8), with target fluence reductions for Cycles 7,8 and subsequent cycles, relative to the Cycle 6 fluence rate as a reference. Given these target fluence reductions, the materials assessment submitted in (5.9) and updated in (5.10)

. concludes that the circumferential weld seam between the middle and lower shells will not reach l l

the PTS screening criteria until well beyond the expiration of the current plant license.

The fluence reductions are shown in Table 4.13. The Fluence Reduction Factors (FRFs) in the table are the fluence rate reductions, relative to the Cycle 6 average fast fluence rate (greater than :  ;

1 MeV), at the stated azimuth's peak axiallocation on the surface of the pressure vessel. The inverse of the FRF is the fraction by which the fluence rate is reduced relative to the Cycle 6 average fluence rate. The target fluence reductions in (5.7) for Cycles 7 through 15 are compared i

j -.

to the actual core design fluence reductions, obtained by a Fluence View Factor (FVF) weighting technique of the average quarter-assembly powers for the cycles. The FVF calculation methods are unchanged for Cycle 15.

The condusion in Table 4.13 is that the cumulative FRF target to EOC 15 has been achieved and exceeded for both the 0 and 10' azimuthal angles. At the critical circumferential weld at 0*, the as-designed cumulative FRF due to low-leakage fuel management is 1.33, or a cumulative fluence reduction to 75% of the Cyde 6 accumulation rate. This is achieved by a Cycle 15 FRF of 1.94, or a cyde fluence rate reduction to 52% of the Cyde 6 accumulation rate. Similar FRFs are expected for future cydes to meet and exceed the targets established in (5.7).

4.11 Methodology and Methodology Revisions A summary of the reference reports, NRC approvals and supplemental documentation for the application of physics methodology to Maine Yankee since Cycle 3 is given in Tabla 4.14. The reference physics methods reports are (1.7,1.8,1.22,1.23). All of the physics analysis for Cyde 15 utilizes the advanced nodal methods, which were described in (1.22,1.23) and approved for use on Maine Yankee in (2.10,2.11). There are no revisions to the physics methods for Cycle 15.

l l>

l TABLE 4.13 I

Maine Yankee Cycles 1 through 15 Relative Pressure Vessel Fluence Comparisons l

Total Effective Fluence Reduction FactorsG) at Azimuthal Angle Full-Power from Perpendicular to Core Shroud Flats ,

Years

( EFPYs 0* - 10*

Cycles to EOC ))U Taraeted Desianed Taraeted Desianed L  :

1 0.86 - 1.21 - 1.19

• 1A 1.23 -

1.22 - 1.21 -

2 2.66 - 1.08 - 1.08 3 3.59 - 1.01 -

1.03 4 4.47 - 0.95 - 0.97 5 5.37 - 0.96 - 0.96 6 ' 6.34 - 1.00 1.00 1.00 1.00 L 7 7.36 1.02 1.07 1.28 1.18 8 8.38 1.35 1.35 1.51 1.36 9 9.55 1.35 1.41 1.51 1.44 10 10.60 1.35 1.73 1.51 1.61 11 11.75 1.35 1.66 1.51 1.56 12 13.04 1.39(3) 1.72 1.55( ) 1.57

! 13 14.19 1.39 1.91 1.55 1.65 14 15.32(') 1.39 1.98 1.55 1.69 15 16.59(*) 1.39 1.94 1.55 1.70 Future Cycles 1.39 1.55 Cycles 1-15 I

CumulativeB) 1.18 1.33 1.26 1.29 l

I 1)' Based on 2,700 MWt full ' power operation

i. 2) Inverse of fractional cycle average flux relative to Cycle 6

3) Increase in fluence reduction factor target for Cycle 12 and later cycles due to power uprate from 2630 to 2700 MWt

4) Estimated cycle lengths of 13,500 and 15,000 mwd /Mt for Cycles 14 and 15, respectively

5) Inverse of the EFPY-weighted fractional cycle average fast fluxes of each cycle l

o

APPENDIX B l l

I Derivation of Chemistry Factor and Margins for l l

Maine Yankee Vessel Beltline Material l l  :

)

i l

l l

l l \

t i

r:\ne\hfj\97-014 B-1

Appendix B l I

Derivation of Chemistry Factor and Margin for j Maine Yankee Vessel Beltline Materials l l

In accordance with 10 CFR 50.61 (Reference 1), paragraph (c)(2), an evaluation l was performed of post-irradiation test measurements for materials applicable to l the Maine Yankee reactor vessel beltline. The applicable data include reactor l vessel survaillance program results from Maine Yankee and three sister vessels. l It also includes test reactor results on samples from two plates and welds from j i the Maine Yankee vessel. The test reactor Charpy shift measurements on one of l the weld heats and one of the plate heats are very consistent with Maine Yankee j surveillance data for those same heats. (The other plate and weld do not have l corresponding Maine Yankee surveillance data.) Therefore, all the test reactor i data are being used to augment the surveillance data for the Maine Yankee reactor l vessel beltline materials analysis. The evaluations are described below and the j results are utilized in Table 5.2 for projections of RTpn. l l

1. Weld 9-203. Heat 1P3571 l I

Three sets of post-irradiation test measurements are available for heat 1 1P3571 and are presented in Tables B-1. B-2 and B-3. Table B-1 provides l the determination of chemistry factor using credible data from the Maine l Yankee surveillance program (References B-2 through B-5), Credibility was l established as follows: (1) The surveillance weld material is identical to l the weld material, heat 1P3571 with LINDE 1092 flux, used to fabricate weld l seam 9-203. (2) The scatter in the plots of Charpy energy versus l l temperature of the irradiated and unirradiated conditions is small enough l to permit the determination of the 30 ft-lb index temperature l unambiguously. (3) There are four sets of surveillance weld measurements: l the adjusted shifts are within 18 F of the predicted shifts using the l derived chemistry factor. (4) The irradiation temperature of the Charpy l specimens in the capsule was within 225 F of the vessel base metal / cladding l interface; this is based on the observed surveillance capsule temperature l monitor results (melting of the 536 F monitors no melting of the 558 F l l

monitors) and the nominal vessel cold-leg temperature during the l m ne % 97.o14 B-2

y / .

~

irradiation periods for the four capsules (estimated as 533, 542, 522 and j t

533 F for capsule W-263. W-253. A-25. and A-35. respectively). (5) The ,

l surveillance data for the correlation monitor material from capsules A-25 [

and W-253 fall within the scatter band.of the data base for the material. J I

The Maine Yankee surveillance weld data were . analyzed using the " ratio l method" to account for differences in copper and nickel content between the l surveillance weld and the vessel weld. The best_ estimate given for the l Vessel weld 9-203 is that based on all valid measurements for heat 1P3571 l as given in Reference B-6. The best estimate given for the surveillance l 1 weld is that based on all valid measurements for that specific weldment ]

(sample) as given in Reference B-6. The measured shifts were adjusted by l ,

the ratio method and the resultant chemistry factor applicable to weld seam l 9-203 was determined to be 213.7*F. l l

Table B-2 and Table B-3 present similar analyses for related post- l irradiation test measurements. Table B-2 provides the determination of l

. chemistry factor using the ratio method for test reactor irradiations of l j

the Maine Yankee surveillance weld. The data are from. References B-7 and B-8. The chemistry factor based on the test reactor data as applied to l r weld seam 9-203 was determined to be 217.4 F. Table B-3 provides the. l determination of chemistry factor using the ratio method for data from the l j Kewaunee surveillance weld. The data are from References B-9 B-10. B-11 l l and B-12. The best estimate copper and nickel for the Kewaunee l )

surveillance weld is that based on all valid measurements for that specific l .J weldment (sample) as given in Reference B-6 The chemistry factor based on l l l

the Kewaunee data as applied to weld seam 9-203 was determined to be l 207.5 F. l l l The chemistry factors < derived using the Maine Yankee plant-specific - l surveillance measurements (Table B-1) are comparable to the chemistry l factor derived using the related post-irradiation test measurements (Table- l B-2) and the Kewaunee plant-specific surveillance measurements (Table B-3). l

! Therefore, the chemistry factor of 213.7 F based on the Maine Yankee l surveillance data can be applied for use in predicting the RTm for weld l i

seam 9-203. This chemistry factor is valid because it is based on credible l 1

r:\ne\hfj\97 014 B-3 i

l -- _ .-- - --

. plant-specific surveillance data and is supported by close agreement with l two sets of related post-irradiation test measurements. Furthermore, in

~

j accordance with paragraph (c)(2)(iii) of Reference B-1 when using the l results from a credible surveillance program. the value of o3 to be used is l 14 F for Weld Seam 9-203. The value of o,is 17 F based on the use of the l l generic initial RT,,1 of -56 F. [ Note: The initial RTm of -56*F was proposed l in Reference B-18 and approved in Reference B-19.] The margin. M. is 44 F l in accordance with paragraph (c)(1)(iii)(A) of Reference B-1. l l

2. Plates D-8406-1 and D-8406-2 l l l l Post-irradiation surveillance test measurements (References B-2 through B- l

5) and test reactor measurements (Reference B-7) are available for the -l Maine Yankee reactor vessel plates 0-8406-1 and 0-8406-2. These two plates l are from the same heat no. B7955 and were processed in the same fashion. l The credibility criteria for material traceability. Charpy data scatter. l irradiation temperature, and correlation monitor material were satisfied l for the surveillance plate as discussed for the surveillance weld (see l preceding section). The criteria for material traceability. Charpy data l scatter, and irradiation temperature were also satisfied for the test l i reactor plate: there were no correlation monitor materials included in the l test reactor irradiation. The credibility criterion for predictability for l l the surveillance and test reactor data was satisfied as follows: (1) The l predicted shifts based on the derived chemistry factor of 126.6'F (see l Table B>4) are within 16 F of the measured (adjusted) shifts for three of l l the four capsules and for the test reactor results. (2) The 146.2 F l predicted shift for capsule A-25 was within 18.8 F of the 127.4 F measured l l (adjusted) shift. The adjusted shifts for data irradiated to fluences l comparable to the A-25 capsule were overpredicted by 7.0 and 10.6 F. (3) l The fluence range of the seven measurements extends more than one order of l magnitude (i .e. 0.567 to 7.13x10"n/cm 2) . The seven measurements for the l two plates exhibit limited scatter relative to predictions, and those data l are conservatively predicted within the range of application (i.e., the l 2

l four measurements below the fluence of 5.9x10"n/cm were less than the l f predicted shifts.) l I l l

r:\ne\hfj\97-014 8-4

. .c

> Table B-4 provides the determination of chemistry factor for heat B7955. l which encompasses plates.0-8406-1 and D-8406-2. using the four surveillance l

-measurements- from plate D-8406-1 and two test reactor measurements from l plate D-8406-2. Those measurements were adjusted by the ratio method to- l account for differences in copper and nickel content between the two l plates. The plate D-8406-1 samples were represented by a chemistry factor j of 109.5 F: the plate D-8406-2 samples were represented by a chemistry 1 factor- of 123.6*F. The data were adjusted to the chemistry factor of l 116.25 F based on the best estimate copper and nickel content of the two l

' plates (i.e., the heat average discussed in Section 5.2) which was obtained l by averaging the individual D-8406-1 and D-8406-2 plate measurements. -The l latter were from chemical analyses made e. samples from the same section of l each test plate used to fabricate specimens for irradiation (surveillance l or test reactor). Therefore, the measurements represent the best estimate l copper and nickel for the post-irradiation test measurements. -The l resultant' chemistry factor applicable to plates D-8406-1 and D-8406-2 was l determined to be 126.6*F: this is more conservative (i.e. higher) than the l chemistry factors based on the copper and nickel content of the individual l plates. l l

In accordance with paragraph (c)(2)(iii) of Reference B-1 when using the l results from a credible surveillance program, the value of c3 to be used is l 8.5*F for the plates. -The value of o, is 0*F because there are measured l values of initial.RT, available for both plates. Therefore, the margin. l M. is 17*F in accordance with paragraph (c)(1)(iii)(A) of Reference B-1. I l

3. Plates 0-8407-1 and D-8407-2 l l

The chemistry factor and margin (165*F and 17'F. respectively) for plate D- l 8407-2 given in Table 5.2 was derived previously. as reported in Reference l B-18. and accepted (Reference B-19) for use in setting the Maine Yankee l

. pressure / temperature limits. A review was performed of the derivation from l Reference B-18. It was determined that the conclusions regarding chemistry l factor and margin remain valid 'for application to the Maine Yankee vessel l pressure / temperatures limits and are acceptable for the determination of l RTers . The derivation reported in Reference B-18 for plate D-8407-2 is l r:\ne\hfD97014 B-5

provided below for completeness, and the conclusion regarding chemistry l factor and margin is extended to plate D-8407-1. l l-

-(a) In Reference B-18. an analysis of the plates was performed using l Regulatory _ Guide 1.99. Revision 02. Position 1.1 to identify 'the l controlling plate in terms of adjusted reference temperature. Plate D- l 8407-2 was identified as the controlling plate. l 1

-(b) Available post-irradiation Charpy measurements applicable to the Maine l Yankee vessel beltline plates were assembled as shown in Table B-5. l They include' data for the surveillance plate, D-8406-1 (References B-2 l through B-4) plate D-8406-2 (plate 1. Low Copper Program, Reference B- l 7), and plate D-8407-1 (plate 2. Low Copper Program. Reference B-7). l These shift data are also shown in Figure B-1. l l

(c).In Table B-5 a chemistry factor was derived for the surveillance plate l (D-8406-1) following Regulatory Guide 1.99. Revision 02. Position 2.1. l A chemistry factor of 118.2 F was obtained. l-l (d) In Table B-5, a second chemistry factor was derived from all seven l measurements using the ratio method of Position 2.1 to adjust to plate l D-8407-2. - A chemistry factor of 165.3 F was obtained which was l intended to conservatively represent plate D-8407-2 using shift l measurements from three other Maine Yankee plates. l l

(e) In Figure B-1, five trend curves were drawn using the Regulatory Guide l  !

1.99. Revision 02 fluence function and the following chemistry factors l (CF): l l

- 109.5 F based on Position 1.1 for plate D-6406-1 l

- 118.2 F based on Position 2.1 for plate C-8406-1 l

- 165.3 F based on Position 2.1 for plate D-8407-2 l

- 168.7 F based on Position 1.1 for plate D-8407-2 l

- 173.9 F based on Position 1.1 for plate D-8407-1 l l

r.:\ne\hfj\97-014 B-6

! [ Note: Position 1.1 refers to'the use'of Table 2 from Regulatory Guide l- {

1.99. Revision 02. .to ~ determine the chemistry factor. Position 2.1 l l refers to the' derived values in Table B-5.]- j. j l' I l

-(f) Observations based on Figure B-1 and supporting information were: -l i i

l .

L - Plates 0-8406-1. and D-8406-2 were cast from the same heat. of l .l

-steel and display similar Charpy shift trends. l l' j

-- Plates 0-8407-1 and 0-8407-2 were also from the same heat.'and~1t l  ;

i 5 .is likely that the Charpy shift trends for both would also be -l .

similar. -The Position 2.1 chemistry factor derived from'the D- l- l 8407-1 data is 129 F. l  :

I l

' - -The position 2.1 mean trend curve using CF-165'F (based on data l- l from three plates adjusted to plate D-8407-2) exceeds the l l unadjusted measurements by an average of 2.9 c3 and a minimum of l l 27'F. The unadjusted measurements include two shift measurements .l {

from the same neat as for plate D-8407-2 which the CF-165 F_ trend l l curve bounds by 48 to 50 F. l l

+

l'  !

- The two trend curves based on Position 1.1 chemistry factor l  ;

values for D-8407-1 and D-8407-2 (168.7 F and 173.9*F.

l respectively) are even more conservative than the CF-165 F trend l l curve relative to the measurements. l ,

I

- The measurements are bounded by a trend curve using CF-146*F l (i.e., the 0-8406-2 shift of ' 210"F divided by the fluence l ,

function)- . l l

_( g) Conclusions drawn. from Figure B-1. Table B-5 and the preceding l observations were that the seven measurements " provide direct evidence l to support the premise that the test measurements can be used to bound l L

the anticipated' irradiation' behavior of the six Maine Yankee reactor l L

l pressure vessel beltline plates". (Reference B-18. Attachment F. page l

! 5-5.') Based on the seven adjusted measurements, the bounding chemistry l I

f

.r:\ne\hfj\97-014. B-7

[ Note: Position 1.1 refers to the use of Table 2 from Regulatory Guide l 1.99, Revision 02, to determine the chemistry factor. Position 2.1 l refers to the derived values in Table B-5.] l 1

(f) Observations based on Figure B-1 and supporting information were: l l

Plates D-8406-1 and D-8406-2 were cast from the same heat of l steel and display similar Charpy shift trends. l l l

- Plates D-8407-1 and 0-8407-2 were also from the same heat, and it [

is likely that the Charpy shift trends for both would also be I similar. The Position 2.1 chemistry factor derived from the 0- l 8407-1 data is 129*F. l 1

- The position 2.1 mean trend curve using CF=165 F (based on data l from three p?ates adjusted to plate D-8407-2) exceeds the l unaajusted measurements by an average of 2.9 o, and a minimum of l l 27 F. The unadjusted measurements include two shift measurements l l from the same heat as for plate D-8407-2 which the CF-165 F trend l curve bounds by 48 to 50 F. l 1 l l

- Th2 two trend curves based on Position 1.1 chemistry factor l values for D-8407-1 and D-8407-2 (168.7 F and 173.9 F. l respectively) are even more conservative than the CF=165 F trend l curve relative to the measurements. l l

- The measurements are bounded by a trend curve using CF-146 F l (i.e. the D-8406-2 shift of 210 F divided by the fluence l function). l

I l (g) Conclusions drawn from Figure B-1. Table B-5 and the preceding l observations were that the seven measurements " provide direct evidence l l

to support the premise that the test measurements can be used to bound l the anticipated irradiation behavior of the six Maine Yankee reactor l pressure vessel beltline plates" (Reference B-18. Attachment F. page l

5-5.) Based on the seven adjusted measurements the bounding chemistry l r:\ne\hfj\97 014 B-7

._ . _ . . ~ . , _ _ _ - __ _ _ _ _ _ _ . . _ - , _ _ _ . _ .

l.

L factor of 165 F was derived. Given the seven credibl'e measurements l which were bounded (i.e.., conservatively predicted) using CF-165*F. and l ,

the availability of a measured value of initial. RTm. the value of o,, l l 1s 0 F and the value of o, to be used is 8,5*F for plate D-8407-2. ' The l l margin. M. is then 17*F (Reference B-18. Attachment F. Section 6).

l l l 1 (h) Subsequent to the preparation of Reference B-18. an additional l

measurement was .obtained for the Maine Yankee surveillance plate l [

(Reference B-5). This was evaluated relative to the original results 'l l and the W-253 surveillance capsule data were found to support the l l conclusions presented above. l- l' l

(1) There are two post-irradiation Charpy measurements (Reference B-7) on l samples from plate D-8407-1. Both measurements are very conservatively l j bounded by a mean trend curve prediction based on a CF-165 F. Given l the availability of credible post-irradiation measurements for plate D- l 8407-1 and the availability of a measured value of initial RTm the l  !

value of ou is 0*F and the value of c6 to be used is 8.5 F for plate D- l  ;

8407-1. 'The margin is, therefore. 17"F in accordance with paragraph l (c)(1)(iii)(A) of Reference B-1. l q 1

.l

4. Weld 3-203. Heat 13253 l l

Table B-6 provides the determination of chemistry factor using measurements  !

from the Salem Unit 2 surveillance program (References B-13 through B-15). l

-which are related to the portions of Maine Yankee vessel weld seam 3-203 l J A.B.C which were fabricated with heat 13253. The credibility criteria for l j material traceability. Charpy data scatter, and predictability were l l satisfied for the Salem Unit 2 surveillance weld relative to the Maine l l

- Yankee weld. Satisfaction of the irradiation temperature criterion is l l based _ on the reported (Reference B-16) cold-leg temperature. 541*F. for l i Salem Unit 2 being within 25*F of the Maine Yankee vessel wall temperature. l l LTherewerenojcorrelationmonitormaterialdataavailable. l l

In Table B-6, the data were analyzed using the ratio method to account for l g differences in copper and nickel content between the Salem 2 surveillance l l l

r:\ne\hfj\97014: B-8 l

~

j f

- weld and Maine Yankee weld 3-203. . The best estimate copper and nickel l l

given for weld 3-203. with heat 13253 1s that based on all valid l  ;

measurements as reported in Reference B-6. The best estimate copper and l [

l nickel given for the surveillance weld is that based on all valid l  ;

measurements for that specific weldment (sample) as given in Reference B-6. l l' l The measured shifts were adjusted by the ratio method and the resultant l chemistry factor applicable to weld seam 3-203 A.B.C was determined to be l l 202.8*F. l I l l Measurements applicable to heat 13253 are also available from the Cook Unit l- j

. 1 surveillance program (Capsules T. Y and U). These results were evaluated l l f~ 'and were found to be conservatively bounded by predictions based on the l ,

- chemistry factor derived for weld 3-203. Therefore, the chemistry factor l l

- of 202.8*F based.on the Salem Unit 2 surveillance weld data can be applied l

! for use in predicting the RT 1 for portions of weld seam 3-203 A.B.C l containing heat 13253 because it is based on credible surveillance data. l l  :

In accordance with paragraph (c)(2)(iii) of Reference B-1, when using the j j f results from a credible surveillance program, the value of o, to be used is l 1

14 F. The value of o, is 17*F based on the use of the generic initial RTer i

of -56 F. The margin. M. is then 44*F in accordance with paragraph l

-(c)(1)(iii)(A) of Reference B-1. l l l l S. Weld 9-203. Heat 33A277 l l

'l Table B-7 provides the determination of chemistry factor using measurements l l from the Calvert Cliffs Unit 1 surveillance program (Reference B-17). which

. l are related to the portion of Maine Yankee vessel weld seam 9-203 which may l have been fabricated with heat 33A277. The credibility criteria for the l l material traceability. Charpy data scatter, predictability, irradiation l

temperature, and correlation monitor material data were satisfied. l  ;

l )

In Table B-7. the data were analyzed using the ratio method to account for -l c differences in copper and nickel content between the Calvert Cliffs 1 l l

surveillance weld and Maine Yankee weld seam 9-203. The best estimate l copper and nickel given for weld 9-203 with heat 33A277 is that based on l i

i i

r:snesntjs97.ot4 B-9 9

L 4- .

i  !

L .

all valid measurements as reported in Reference B-6. The best estimate l l L copper and nickel given for the surveillance weld is that based on all l L valid ' measurements for that specific weldment (sample) as given in l  :

l Reference B-6. The measured shifts were adjusted by the ratio method and l  ;

the resultant chemistry factor applicable to weld seam 9-203 for portions l  !

L containing heat 33A277 was determined to be 99.2*F. -l

^

1 In accordance with paragraph (c)(2)('iii) of Reference B-1, when using the l

results from a credible surveillance program, the value of at to be used is l l 14 F. The value of o, is 0*F because there are two measured values of l j initial RTer available for this heat. (Initial RT.1 is -80 F for the l Calvert Cliffs Unit 1 surveillance weld and -60 F for the Farley Unit 1 l-surveillance weld; the lower value of -60*F is being used for Maine Yankee l l weld 9-203 with heat 33A277.) The margin. M. is then 28 F in accordance l-with paragraph (c)(1)(iii)(A) of Reference B-1. l t

l j

l

6. Weld 3-203. Heat 12008 and 13253 l l

l The determination of chemistry factor for the portions of Maine Yankee l vessel weld seam 3-203 A.B,C which were fabricated with heat 13253 is l presented in Section 4 and Table B-6. This section addresses determination l of a chemistry factor for the balance of weld 3-203 A.B.C which was l fabricated with a combination of heats 12008 and 13253. The analysis.uses l ,

the Salem Unit 2 surveillance weld data (References B-13 through B-15). l

[ which .were discussed previously in Section 4 of Appendix B. and test l L reactor data obtained on a Maine Yankee vessel weld material (Reference B- l-20). l l

The Salem Unit 2_ data represent a weld fabricated using heat 13253. These l j data were analyzed in Table B-8 using the ratio method to represent the l l l chemistry of weld 3-203 A.B.C made using a combination of heats 12008 and l L 13253- . The best estimate copper and nickel given for weld 3-203 with both l heats is that based on all valid measurements as reported in Reference B-6. l

! The best estimate copper and nickel given for the Salem Unit 2 surveillance l l

I weld is that based on all valid measurements for that specific weldment l (sample) as given in Reference B-6. The measured shifts for the Salem data  ;

l

.r;\ne\hfj\97014 B-10

. _ _ - . _. _._ _-__ _ __ _ . _ _ . - . _ _ _ . . _ . _ _..__..__.m . _ _ . _ .

l were adjusted by -the ratio method and the resultant chemistry factor I.

derived for weld seam 3-203 ,A.B.C was 224.2 F. l.

L . l l Test reactor post-irradiation measurements are also available for a weld I fabricated using heats 12008. and 13253 in tandem. Material " Code Y" l  ;

results given in Reference B-20 were obtained on samples _ from a Maine l i Yankee weld seam 1-203 A nozzle drop-out. Two sets of CharpyLsamples were l )

irradiated- to a' neutron fluence 'of 7.9 x 1018 n/cm2 at two different 'l -l

-temperatures. 500 F and 600 F The measured Charpy shifts were 215 F and 'l 100 F. respectively. The shifts were -then adjusted to an irradiation' l l

temperature of' 540*F using the assumption that 1*F elevation of the l

irradiation. temperature reduces the shift by 1*F. (540 F was selected for l i j the normalization temperature because both Salem 2 and Maine Yankee operate l with approximately a 540*F cold leg temperature.) The normalized shifts l are then

l l  !

-215*F + (500 - 540) F = 175 F l -

100 F + (600 - 540)*F - 160 F. .l j l 1 In Figure B-2. the as-irradiated and the normalized shifts for the test l t

l reactor data are shown plotted together with the' adjusted shifts for the l I

Salem 2 surveillance data from Table B-8. Also shown in Figure B-2 are_ l ,

l mean shift prediction curves for two cases. In the first case, the l -l chemistry factor (224 F) is from Table B-8 to represent Maine Yankee weld l l 3-203 with heats 12008 and 13253

• ed on the Salem 2 adjusted data. In l ,

the second case. the second chemisk factor (203 F) is from Table B-6 to l

' represent weld 3-203 with just heat 1, '3. -l  ;

I Observations based on Figure B-2 are as follows: l l.

l j -

The normalized shifts for the test reactor data are bounded by a l l j chemistry factor of 187.4 F (i.e. , the 175 F shift divided by the l fluence -function) . l The mean predicted shift based on CF=224*F (Table B-8) exceeds the two l  ;

L normalized test reactor measurements by 34 to 49*F. l 3

i l ,

r:\ne\nfj\97-014 B-11

The mean predicted shift based on CF-203 F (Table B-6) exceeds the two l l

normalized test reactor , measurements.by 14 to 29*F. l l l

The material irradiated in the test reactor is from Maine Yankee weld seam l l i

1-203 A wh'ch was fabricated using the same weld electrodes. flux type and- l fabrication procedures as used for weld seams 3-203 A.B.C. The test l  ;

I reactor data should, there# m . be. a precise representation of the j irradiation response for weld seams 3-203 A.B.C and the normalized shift l' l

.should be a good approximation of the irradiation response at 540 F. In l conclusion, the derived chemistry factor of. 224 F from an analysis of l I

credible surveillance data in Table B-8 will provide a conservative shift l l prediction method for weld seams 3-203A.B.C.- This conclusion about l prediction conservatism is based on the availability of test reactor l 3 measurements on. samples from an equivalent Maine Yankee reactor vessel l weld. l i

l In ~ accordance with paragraph (c)(2)(iii) of Reference B-1, when using l results from a credible surveillance program, the value of o3 to be used is l 14 F. The value of o, is 17 F based on the use of the generic initial RT, l 1 of -56 F. The margin. M. is then 44*F in accordance with paragraph l ,

(c)(1)(iii)(A) of Reference B-1. l l I

)

l i

t I

i I

r:\ne\hfji97014 B-12 i i

l

The mean predicted shift based on CF-203*F (Table B-6) exceeds the two l normalized test reactor , measurements by 14 to 29 F. l l

The material irradiated in the test reactor is from Maine Yankee weld seam l 1-203 A which was fabricated using the same weld electrodes, flux type and l fabrication procedures as used for weld seams 3-203 A.B.C. The test l reactor data should. therefore. be a precise representation of the l irradiation response for weld seams 3-203 A.B.C and the normalized shift l

..should be a good approximation of the irradiation response at 540 F. In l conclusion. the derived chemistry factor of 224*F from an analysis of l credible surveillance data in Table B-8 will provide a conservative shift l

prediction method for weld seams 3-203A.B.C. This conclusion about l prediction conservatism is based on the availability of test reactor j measurements on. samples from an equivalent Maine Yankee reactor vessel l weld. l l

In accordance with paragraph (c)(2)(iii) of Reference B-1. when using l results from a credible surveillance program, the value of a3 to be used is j 14*F. The value of o, is 17*F based on the use of the generic initial RT, l of -56 F. The margin. M. is then 44*F in accordance with paragraph l (c)(1)(iii)(A) of Reference B-1. l l

l I

l r:\ne\hfj\97 014 B-12

Accendix E References l I

B-1 10 CFR 50.61. " Fracture Toughness Requirements for Protection Against l Pressurized Thermal Shock Events." Federal Register Vol. 60. No. 243 dated l

~

Tuesday. December 19. 1995, pp. 65468-65472. l l.

B-2 " Evaluation of the First Maine Yankee Accelerated Surveillance Capsule." l Effects Technology. Inc. Report CR 75-317. August 15. 1975. l l

B-3 " Maine Yankee Nuclear Plant Reactor Pressure Vessel Surveillance Program - l l Capsule 263." Battelle Columbus Laboratories Report l

. BCL-585-21. December 12, 1980. l l

B-4 " Analysis of the Maine Yankee Reactor Vessel Second Accelerated l Surveillance Capsuis " Westinghouse Report WCAP-9875. March 1981. l l

B-5 E. Terek. et. al.. ' Analysis of the Maine Yankee Reactor Vessel Second Wall l Capsule Located at 253." Westinghouse Report WCAP-12819. March 1991. l l

B-6 "Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel  !

Welds." CE NPSD-1039. December 1996. l l

B-7 J. R. Hawthorne. J. J. Koziol, and S. T. Byrne. " Evaluation of Commercial l Production A533-B Steel Plates and Weld Deposits with Extra-Low Copper l Content for Radiation Resistance." NRL Report 8136. October 21. 1977. l l

B-8 J. R. Hawthorne. " Notch Ductility Degradation of Low Alloy Steels with Low- l to-Intermediate Neutron Fluence Exposures." NRL Report 8357 (NUREG/CR- l 1053). January 14. 1980. l l

B-9 S. E. Yanichko, et. al.. " Analysis of Capsule V from the Wisconsin Public l Service Corporation Kewaunee Nuclear Plant Reactor Vessel Radiation l Surveillance Program." Westinghouse Report WCAP-8908. January 1977. l l

r:\ne\hfj\97-014 B-13

B-10 S. E. Yanichko. et. al., " Analysis of Capsule R from the Wisconsin Public l Service Corporation Kewaunee :iuclear Plant Reactor Vessel Radiation l Surveillance Program." Westinghouse Report WCAP-9878. March 1981. l l

B-11 S. E. Yanichko, et. al , " Analysis of Capsule P from the Wisconsin Public l )

Service Corporation Kewaunee Nuclear Plant Reactor Vessel Radiation [

Surveillance Program". Westinghouse Report WCAP-12020. November 1988. l l

l B-12 E. Terek. et. al., " Analysis of Capsule S from the Wisconsin Public Service l l Corporation Kewaunee Nuclear Plant Reactor Vessel Radiation Surveillance l l Program." Westinghouse Report WCAP-14279. March 1995. l 1

B-13 R. S. Boggs, et. al. " Analysis of Capsule T from the Public Service l Electric and Gas Company Salem Unit 2 Reactor Vessel Surveillance Program." l March 1984. Westinghouse Report WCAP-10492. l I

B-14 S. E. Yanichko. et. al., " Analysis of Capsule U from the Public Service l l Electric and Gas Company Salem Unit 2 Reactor Vessel Radiation Surveillance j Program." September 1987. Westinghouse Report WCAP-11554. l l

B-15 J. M. Chicots et. al. . " Analysis of Capsule X from the Public Service l Electric and Gas Company Salem Unit 2 Reactor Vessel Radiation Surveillance l  !

i Program." June 1992. Westinghouse Report WCAP-13366. l l

l B-16 Public Service Electric and Gas Company Letter NLR-N92081 dated June 30. l 1992. " Response to Generic Letter 92-01. Revision 1. Reactor Vessel l Structural Integrity 10CFR50.54(f). Salem Gener ting Station Units No.1 l l and 2.' l l l B-17 R. E. Denton. "Calvert Cliffs Nuclear Power Plant Units No.1&2 Request l l

l for Approval of Updated Values of Pressurized Thermal Shock Reference l Temperature Values (10CFR50.61)." Baltimore Gas & Electric Company letter l dated July 21. 1995. l l

l r:\ne\hfj\97-014 8-14

B-18 Att6chment F to MYAPCo Letter to USNRC. " Proposed Change No.145 - Combined l Heatup. Cooldown and Pressure-Temperature Limitations." dated December 2. l 1088 (MN-88-U.6). l l

B-19 USNRC Letter to MYAPCo. " Issuance of Amendment No. 114 to Facility l Operating License No. DPR Maine Yankee Atomic Power Station (TAC No. l 71511)," dated November 17. 1989. l l

B-20 J. R. Hawthorne. " Exploratory Studies of the Effects of Irradiation l Temperature on Mechanical Properties of Structural Steels and Welds." l Effects of Radiation on Materials 16th International Symoosium. ASTM STP l 1115. A. S. Kumar. D. S. Gelles. R. K. Nanstad, and E. A. Little. Eds. , l American Society for Testing and Material. Philadelphia, 1993. l l

B-21 S. T. Byrne and R. O Hardies. " Refinements to Pressure Vessel Steel l Embrittlement Correlations." Effects of Radiation on Materials- 18tn l International Svrnoosium. ASTM STP 1325 Randy K. Nanstad. M. L. Hamilton. l F. A. Garner, and A. S. Kumar. Eds . American Society for Testing and l Materials. Philadelphia,1997 (to be published). l f

r:\ne\hfj\97-014 B-15

__ _ . _ - _ . _ . . . _ _ _ _ _ _ . _ _ _ . _ ~ . . . _ . _ . _ . _ _ _ _ . . - . . _ _ _

l TABLE B-1 l

. l Chemistry Factor Derivation .

l Maine Yankee Weld 9-203 l l i 1

l Weld-9-203: Cu - 0.28% Ni = 0.75% CF - 210.2 F [ ,

Surveillance Weld: Cu - 0.35% Ni - 0.77% CF = 236.6 F (Best Estimate l Chemistry for MY l l Surveillance Weld) l l l  !

l l

Data;Sourced JMeasuredL. i Adjusted (*( 1 Fluence?2 ?Fluencei , l

,; Shifty (F)? (Shiftf(*F)c 5(n/cm )? , 2 IFacitoid?^- l W-263 222 197.1 0.567 0.841 l

.W-253 -260 230.9 1.25 1.062 l A-25 270 239.8 1.76 1.155 l A-35 345 306.4 7.13 1.466 l l

l Fluence Factor'x. (FluenceL- Predicted: Adjusted Minus:: l Adjusted Shift.( F) Factor)2' . Shift ( F) Predicted ( F) l

-165.76 0.708 180 +17 .l 245.22 1.128 227 +4 -l l 276.97- 1.335 247 -

7 l 449.18 2.149 313 -7 l  !

Total 1137.13 5.320 l l l

.CF (Adjusted) = 1137.13 F / 5.320 - 213.7 F l l I

l  !

(a) Adjusted shift - Measured x (210.2/236.6) = M'easured x 0.888 l

-(b) Fluence in units of 10" n/cm2 . l (c) -Fluence factor per Reference B-1. l l- I i

'r:\ne\hfj\97-014 8-16 I

E __ -_ _ . _ - _ _ _ _ _ _ . . . _ . _ _

i TABLE B-2 l l l  !

Chbmistry Factor Derivation l Experimental Weld Data .

l  ;

l i l

Weld 9-203: Cu - 0.28% Ni = 0.75% CF = 210.2 F l l

' Surveillance Weld: Cu - 0.35% N1 - 0.77% CF - 236.6*F (Best Estimate l i Chemistry for MY _ l .

Surveillance Weld) l j l '

l sDataiSource} ..MeasuredL, i

j;AdjustedW w flue'nce*' 3 Fluence l l 1 Shift 3(*F)! 4 Shift (*F)!  ;(n/cm2 )' Factor (c'z ,

l j B 315 279.72 3.0 1.291 l  !

B 350 310.80 5.3 1.414 l

-B-8 55 48.84 0.12 0.454 l I i

B-8 240 213.12 0.66 0.884 l l 1

l Fluence factor =x (Fliterice ' -Predicted- _~AdjustedMinus-' l Adjusted- Shift ':(*F) Factor)2 Shift'( F) Predicted.("F) l 361.12' 1.666 281 -1 .l 439.47 -1.998 307 +4 l 22.17 0.206 99 -50* l 188.40 0.781 192 +21 l Total 1011.16 4.651 l l l l CF (Adjusted) - 1011.16 F- / 4.651 - 217.4*F l 1

l I

(a)- .

Adjusted shift - Measured x (210.2/236.6) - Measured x 0.888 l ,

(b) Fluence in units of 109 n/cm 2. l (c). Fluence factor'per Reference B-1. l j (d)- Large (but conservative) overprediction is 1imited to the lowest fluence l l exposure measurement. l ;

i

]

\

l r:\ne\hfj\97 014 B-17  :

' TABLE B-3 l

. I Chemistry Factor Derivation l Using Kewaunee Data l l

l

' Weld 9-203: Cu - 0.28% Ni - 0.75% CF - 210.2*F: .

l Kewaunee Surveillance Weld: Cu - 0.23% Ni'= 0.74% CF = 192.8*F(Best Estimate l Chemistry for Kewaunee l Surveillance Weld) l 1

l-

D'ata;Sourcel .; Measured; [Adjtisted!') ! xFluence %  ! Fluence l w -

(Shift? (.*F)J ;Shifti(*FD , ~ j(n/bm 2)? Fadtor!c'i l Capsule V 17.5 190.75 0.6295 0.870 l Capsule R 235 256.15 1.94

• 1.181 l Capsule P 230 250.70 2.89 1.282 l Capsule S 250 272.50 3.45 1.323 l l

l

' Fluence Factor ~xn -(Fluence Predicted Adjusted Minus l Adjusted Shift-(*F) Factor)2- Shift?(F) Predicted (*F) l '

k 165.95 0.757 181 +10 l

' 't 302.51 1.395. 245 +11 l 321.39 1.643 266 -15 l 360.52 L130 275 -2 j Total 1150.37 5.545 l f CF (Adjusted) - 1150.37 F / 5.545 - 207.5*F l l

l (a) . Adjusted shift - Measured x (210.2/192.8) - Measured x 1.090 l (b) Fluence in units of 10" n/cm2 . -l (c) Fluence factor per Reference B-1. l (d)- . Fluence per Reference B-12. l l

rnne\hfj\97-014 B-18

L.

l TABLE B-4 l

. l Chemistry Factor Derivation l Maine Yankee Plates D-8406-1 and D-8406-2 1 3 l i l D-8406-1 Cu - 0.15% N1 - 0.59% CF = 109.5 F (Surveillance Plate)

D-8406-2 Cu = 0.17% Ni = 0.56% CF = 123.6 F L Heat Average: Cu - 0.16% Ni = 0.575% CF - 116.25 F L' Datia Source 1 1Measured . l Adjusted ('E LFluence"- FluenceL l l LShiftf( F)j iShift:(*F)-  :-l(n/cm )?

2 l Factor" l W-263 95

• 100,85 0.567 0.841 l l W-253 120 127.39 1.25 1.062_ l A-25 120 127.39 1.76 1.155 l  !

A-35 190* 201.70- 7.13 1.466 l Ref. B-7 145 136.37 1.8 1.161 l Ref. B-7 210 197.50 5.9 1.434 l l

Fluence Factor x (Fluence Predicted Adjusted.Minus. l

'Adjus'ted Shift *( F) Factor)2 Shift-(*F) . Predicted. -(

• F) l L 84,81 0.708 106 -5 l 135.29 1.128 134 -7 l l 147.14 1.335 146 -19 l 295.69 2.149 186 +16 l L 158.33 1.349 147 -11 l 283.21 P.L_QM 182 +16' l Total 1104.47 8.725 l l

.l l CF (Adjusted) - 1104.47 F / 8.725 - 126.6*F l l

1 (a) Adjusted shift - measured shift x 116.25 / CF l Where CF = 109.5 for the surveillance plate data and CF-123.6 for the l Reference B-7 data; l l (b) Fluence in units of 10 9 n/cm. 2 l

l (c) F.luence factor per Reference B-1. l (d) Average of longitudinal-and transverse orientation results. l r

r:\ne\hfj\97-014 B-19

i

l. TABLE B-5 l )

l

~l Chimistry Factor Derivation l Maine Yankee Plate 0-8407-1 and 0-5407-2 .

l l

L . I A -'P1 ate D-8406-1 Cu = 0.15% Ni = 0.59%. CF = 109.5 F (Surveillance P1 ate) l B . Plate D-8406-2 Cu = 0.17% Ni = 0.56% CF u 123.6*F -l

-C - Plate D-8407-1 Cu = 0.24% Ni = 0.62% CF = 173.9'F- l.

i D.- Plate D-8407-2 Cu - 0.23% Ni = 0.62% CF = 168.7 F l 1 l l 1

!DataVSource"E DMeasured. Fluence * ) Fluence- )

l - <f JShift ( F) 2 1(n/cm )v ( Factor'd f  !

1 1 .

W-263(A) . 95(* 0.567 0.841 l A-25 (A)~ 120 1.76 1.155 l A-35 (A) ^ 190(85 7.73 1.478 l LCP (B) 145 1.8 1.161 l l

l LCP (B) 210 5.9 1.434 l LCP (C) 165 -3.2 1.306 l 1

LCP (C) 185 5.3 1.414 1 I

L 'l l Surveillance Data. D-8406-1. Curve Fit l l

l Fluence Factor xtShift (*F) .(Fluence Factor)2 .l 79.9. .0.707 l l -!

L 138.6 1.333 l. -!

23{L3 2.184 l  !

Total- 499.3 4.224 l l

I

CF = 499.3 F / 4.224 = 118.2 F l l

l I I i

r:\ne\hfj\97-014: B-20

)

l '

L ,: ....

TABLE B-5 (CONT'D) l

.. - . I

' Low Coooer Droaram. D-8407-1. Curve Fit l- )

l _

l  !

Fluence Factor' ( Shif R F) (Fluente Factor)2 l l

215.49 .1.706 l L 261.59- 1.999 j.

Total 477.08 3.705 l l '

l CF = 477.08 F /3.705 = 128.8 F- l 1

Surveillance and Exoerimental Data Adiusted to Plate D-8407-2 IdjustedlShifti(FNA Fluence Factorix3 Adjusted I(F10ence1 Factor)2;

Shift.

( F):-

146.4 123.1 0.707 l

, 184.9 213.5 1.333 l 1

292.7 432.6- 2.184 l.

197.9 229.8 1.348 l 286~.6 411.0 2.056 l-160.1 209.0 1.706 l 4

179.5 ?13.J1 1.999 l-Total 1872.8 11'.333 l-CF (Adjusted) - 1872.8 F /'.11.333 -'165.3*F  !

J

-- ( a ) Surveillance capsule or Low Copper Program (LCP, Reference B-7), and

(b) material Fluence in(A. B or units of C) 10 " n/cm 2

'(c)- Fluence factor per Reference B-1 (d) Avera e of longitudinal and transverse orientation results

'(e)- Adjus ed-Shift - Measured Shift x 168.7/CF where CF is for material A B

! 4 I

(

r:\ne\hfj\97-014 B-21 U - . . . . . . . --

TABLE B-6 Chemistry Factor Derivation Maine Yankee Weld 3-203 A.B.C with Heat 13253 l

I Weld 3-203 A.B C (Ht. 13253): Cu = 0.22% Ni = 0.73% CF = 188.4'F Salem 2SurveillanceWeld: Cu = 0.25% Ni = 0.73% CF = 197.4 F y

!- Data Source' Measured. Adjusted ('!-' Fluence2 *)  : Fluence

Shift ( F)- -Shift

-( F).

s < (n/cm ) - Factor!c)

Capsule T 155 147.9 0.275 0.648 l-Capsule U 190 181.3 0.55 0.833 l

~

Capsule X 195 186.1 1.07 1~.019 l 1

l

. Fluence FactorJx Pred'icted AdjustddiMihus l Adjusted Shift'(.*F) Factor) ,Shifti( F)' .PredictedL(*F)t

.(Fluench i L )

95.8 0.420 131 +16 l  !

151.0 0.694 169 +12 l 189.6 1 038 207 -21 l l

l l Total- 436.4 2.152 l CF (Adjusted) = 436.4*F / 2.152 - 202.8 F

'(a) Adjustedshift-MeasuI9ed x (188.4/197.4) - Measured x 0.954-(b) Fluence in units of 10 n/cm' u l.

(c) Fluence factor per Reference1. B pdated in WCAP-14044.

l l' '

I l I i

i i

U r M e\hf.1\97 014 B-22 l

A TABLE B-7

. . Chemistry Factor Derivation Maine Yankee Weld 9-203 with Heat 33A277 Weld 9-203 A.B.C'(Ht. 33A277): Cu - 0.26% Ni = 0.16% CF = 126.6 F Calvert Cliffs Surveillance Weld: Cu = 0.18% Ni = 0.16% - CF = 91.8 F DataLSource. Measured - Adjusted *J Fluence

• Fluence

. Shift (*F)  : Shift:.( F)' d

(n/cm ) Factor

• s W263 59 -81.4 0.620 0.866. l W97 93 128.2 2.64 1.260 1

~

. Flusnce Factorix .. Predidted!- ? AdjustsdlMinus-:

Adjusted.' Shift-(*F) (Fluencb.-

Factor) Shift:( F) PredictedL(*F)-

70.5 0.750 86 -5 l 15LE L5a8 125 +3 l Total 232.0 2.338 l CF (Adjusted) = 232.0"F / 2.338 - 99.2 F (a) Adjustedshift-Measuredx'{126.6/91.8)- Measured x 1.379 (b)- Fluence in units of 10 9 n/cm.

(c) . Fluence factor per Reference B-1.

r:ine\hfj\97 014 B-23 1

l

. . ~ . _. .. .- - - . - -- - . - -_ .-

i TABLE B-8 l

Chemistry Factor Derivation Maine Yankee Weld 3-203 A.B.C with Heats 12008 & 13253 l l

I Weld 3-203 A.B.C: Cu = 0.21% Ni = 0,87% CF 208.2*F j Salem Unit 2 Surveillance Weld: Cu - 0.25% Ni = 0.73% CF = 197.4 F l l

. Data. Source- Measured: Adjusted (*). Fluence *) . Flue'nce l Shift:(?F)- Shift ( F) 2 (n/cm)J JFactork)c l .

Capsule T 155 163.5 0.275 0.648 l Capsule U 190 200.4 0.55 0.833 l Capsule X 195 205.7 1.07 1.019 l l

l Fluence Factor x .(Fluence iPredicte'd Adjusted.Minus-- -l Adjusted Shift ( F) Factor)2 ShiftL( F) Predicted ( F).- l.

105 9 0.420 145.3- +18 l 166.9 0.694 186.8 .+14 l 201Ji LD3B 228.5 -23 l Total 482.4 2.152 l l

CF (Adjusted) = 482.4 F / 2.152 = 224.2 F l

'l I

(a) Adjusted shift - Measured x (208.2/197.4) - Measured x 1.055 l (b) Fluence in units of 10" n/cm updated 2

in WCAP-14044. l (c) Fluence factor per Reference B-1. l j l

I i

l I

r:\ne\hfj\97-014 B-24

FIGURE B-1 COMPARISON OF EXPERIMENTAL AND SURVEILLANCE IRRADIATION DATA FOR MAINE YANKEE RPV PL ES .  :

3 i ,

i i i i i i i i -

280 -

0 SURVEILLANCE (D-8406-1) ~

PRED CTION D 8407-1.

A PLATE ILCP (D-8406-2) CF = 173.9 l 260 - ~

Q PLATE 2LCP (D-8407-1) 240 -

TABLE 2/REV. 2 ~

PREDICTION D-8407-2, i l CF = 168.7 l 220 - -

A 200 - -

u. .

P O E 180 -

Q . -

I .

sn Uz 160 - -

Et 140 -

a -

[

120 -

D TABLE 2/ REY. 2 -

PREDICTION D-8406-1, i 100 -

CF = 109.5 _

80 -

POSITION 2.1 CURVE _

j

^

. POSITION 2.1 CURVE SURVEILLANCE DATA, 60 -

FIT TO D-8407-2, CF = 118 -

CF = 165 r

40 I I 8 I I f I i I '

O.1 0.2 0.4 0.6 0.8 1.0 2.0 4.0 6.0 8.0 t 0.0 FLUENCE (E > 1MeV),10 I3 n/cm 2  ;

t

i FIGURE B-2 COMPARISON OF SURVEILLANCE AND EXPERIMENTAL IRRADIATION DATA FOR MAINE YANKEE WELD 3-203 A,B,C 250 CF=224*F (Table B-8) 225 >

0 CF=203*F (Table B-6) 200 X i-

.-175 e

~

x. x

$ A 5

x 150 o X Salem 2 Survemence Weld Adjusted to Heats 12008,13253 i O MY Weld 1-203, Test Reactor, Irradiated 500*F -

e MY Weld 1-203, Irradiated at 500*F and 125 Adp'usted to 540*F A MY Weld 1203 Test Reactor, Irradiated i 600*F 6 A MY Weld 1-203, Irradiated at 600*F and '

Adjusted to 540*F 100 A r

75 O.1 1 10 FLUENCE (E > 1MeV),10" n/cm' ,

I i

i i

Attachment B of Enclosure 1 MN-97-66 l

l l

t l

L t

i i

Attachment B Maine Yankee l Final Report Items 2,3, and 4 Generic Letter 92-01, Revision 1, Supplement 1

[

L Comparison of PTS Licensing Values l

USNRC's RVID and Maine Yankee's PTS Report (Attachment A) l l

l~

l l

'rs\ne\hfj\97-011 L

L. __ ..

l 4

Attachment 8 of Enclosure 1 TA8tf 8.1 MN 97-66 Conparison of Pressurized Thermal Shock Values Page 1 of 2 USNRC Staff

• arul Maine Yankee'"

Cheatstry Factor Method of Detersin. CF Cu Ni Plant Beltline ' Heat No. ID heut. IRT, Method of Concentration. 4 Concentration. 8 E* W Report Date

• haue Ident. Ident. Fluence at Determin. W RM Date ( NNTS NRC m ECL/EFPV IRT , g g,97 , ,g, p,gg gg,q, g,gg og;qy Maine Int. Shell 8-7955-1 1.80E19 -10*F P* ant -119.53 7109$ $ ~12h ' Table - ' Calc. ; 0.15  ; 9.16 ? 4.59 ' 8.58 ; See Section II.81.a of MN 94 50 i Cate.c' Ya8* ee Plate specjgic

'c p fe ' Best Estimate Cu/N1 Average of 0-8406 1 ,. + - M. ' 0-8406 1 & 2 Int. Shell See section II.8.1.a of MN-94-50 EOL: B-7955-2 1.80E19 0*F MTER 5-2 119.53 ' - 124' . 127 5 Nate.] .TYable .~ Calc. - 0.17 J 8.16 c 8.56  :. 0.58 10/21/ Plate

^

W '

Best Estimate Cu/N1 Average of 2008 0 8406-2 U D-8406-1 & 2 Int. Shell C-3982-5 1.80E10 0*F hTES 5-2 83.3 83 Table 0.12 0.62 Insignificant difference in CF likely Plate due to rounding.

0 8406 3 __

Laver 8-8330-1 1.80E29 -20*F M118 5-2 Table 0.24 0.62 Best Estimate Cu/N1 Average of Shell 174 165 0-8407 1 & 2 Plate 0 8407 1 _

Lower B-8330-2 1.80E19 2*F MTE8 5-2 169 165 Table e.23 0.24 - 0.62 Best Estimate Cu/ut Avem of Shell , 0-6407-1 & 2 Plate D-8407-2 Lower B-8324-1 1.80E19 0*F MTES 5-2 92.25 92 Tacle 0.13 4.65 Insignificant difference in CF likely Shell due to row-fing.

Plate D 8407-3 ,

Cire. Weld IP3571 -56'T Generic 240.96 l L222 ': i 214i k, Calc Q: .. Table -

- Cate'. . 0.31 ' g.28 . 0.76 6.75 See section II 8.1.h of MN-94 30 9 203 1.80E19 x, - Best Estimate Ca/21 obtained from CE

' '~

J NP5D-1039 3D277 -60*F Messured _ h 99 - - -

I Calc [ - 0.26 ;  : 8.16 .

~

Best Estimate Cu/41 obtained fran CE

-i...

nesD.1039 r \ne\hf j\M-011 O

l

] Attachment B of Enclosure 1 TABLE S.1 E97 66 Page 2 of 2 l

Casparison of Pressurized Thermal Shock Values

• USNRC Staff"8 and Maine Yankee

• i Chemistry Factor Method of Determin. CF Cu N1 J

i Plant Beltline Heat 50. ID Neut. IRT., Method of Concentration. 4 Concentration. 4 E* W Re Date

• p Report Date

• coeggys Name ILit. Ident. Fluence at Determin.

IRT*' 05/94 64/97 05/94 04/97 05/94 64/97 05/94 04/97 Maim Antal 51989 1.48E19 -56*F Generic 99.45 89 Table 0.17 0.17 9.16 ; Insignificant difference in CF likely Yankee Welas due to rounding.

2-203 Best Estimate Cu/N1 obtained from CE 4 - MPSD-1039 ECL: Axial 13253 - 56*F Generic 206.4 206 224 Table Calc. 0.22 9$21.I 0.84  ; DI-~ Insignificant difference in CF likely 20/211 Welds and ese to rounding.

2008 3-203 12008 1.49E19

- 8est Estimate Cu/Ni obtained from CE j

NPSD 1339 13253 -56*F Generic -

? 20(. -

[ Calc.: ~0.22 l - 6.2 . Best Esttaate Cu/W1 obtatned from CE

). NPSD-1639 "VEtt staff values are obtatned from Enclosure 1. *Pressurued Thermal $ rock Table". of USNRC Letter to MY dated April 12. 1994 "Haine Ye8*ee Values are obtained from Matre Yankee letter to USNRC dated May 20.1994 (N94 50) and Attacteient A f

r r an. wi n s7-o11 O

- _ _ _ - _ - _ _ _ _ _ _ _ - . _ . _ . - _ _ - _ . . ._._._-__._s .__._m -_.m__--- __-u. _ _ - . _ . . _ . - _ . - _ _ . . _ _ _ _ _ . . - _ _.__-.m _ _ _ . - . _ m-__________.__m. __._-______-_m. .__m_mm.._.___-m._ _ _-..[..-._.. A.______m-s- w i *b ' _ _ _ . _ _ _ _ _ _ _ . - - . _ _

==- % .,

TABL Comparison of Pressuri USNRC Staff'" a mumummune mummmmmmmmmme mummmmmmmum - nummmmmum mumummmmmmm Chemistry Factor Plant Beltline Heat No. ID Neut. IRTe Method of Ident. Ident. NRC "' MY Report Date d)

Name Fluence at Determin.

EOL/EFPY IRT* 05/94 04/9 Maine Int. Shell B 7455 1 1.80E19 10*F Plant 119.53:  :

109.-  ::127).

Yankee Plate Specific D 8406-1 EOL: Int. Shell B 7955 2 1.80E19 0*F MTEB 5 2 119.53( l- 124 i 1127 f.

M/21/ Plate 2008 D 8406 2 Int. Shell C 3982 5 1.80E19 0*F MTEB 5 2 83.3 83 Plate D 8406 3 Lower B 8330 1 1.80E19 20*F MTEB 5 2 Shell 174 165 Plate D 8407 1 Lower B 8330 2 1.80E19 2*F MTEB 5 2 169 165 Shell Plate D 8407 2 Lower B 8324 1 1.80E19 0*F MTEB 5 2 92.25 92 Shell Plate D 8407 3 Cire. Weld IP3571 56*F Generic 240.96'- :222- L214-9-203 1.80E19 33A277 60*F Measured - 99 ;

r:\ne\hfj\97-011

ARSTEC __

APERY GE 1 {C Attachment B of Enclosure 1 MN 97 66

,hermal Shock Values g /ggg kVSgg gg on.

• Page 1 of 2

.aine Yankee cr> ,, gggre Method of Determin. CF Cu Ni Concentration, 2 Concentration.

• MY Report Date '2' COMMENTS NRC '"

05/94 04/97 05/94 04/97 05/94 04/97 Calc.- ' Table : -: Calec 0.15 - 0.15, 0.59 . 0.5S E See Section II.B.I.a of MN 94 50 Best Estimate Cu/N1 Average of D 8406 1 & 2 Calc.' Table- .Cale 0.17 ' O.16 i 0.56 0.68: See Section II.B.I.a nf MN 94 50 Best Estimate Cu/Ni Average of D 8406-1 & 2 Table 0.12 0.62 Insignificant difference in CF likely due to rounding.

Table 0.24 0.62 Best Estimate Cu/N1 Average of D 8407 1 & 2 Table 0.23 0.24 - 0.62 Best Estimate Cu/Ni Average of D 84C/.1 & 2 Table 0.13 0.65 Ins 19nificant difference in CF likely due to rounding.

Calc . ITable : c Calc. 0.31  ; 0'.23 - 0.76 - 0.75 ) See Section II.B.1.b of MN 94 50 Best Estimate Cu/N1 obtained frce CE NPSD 1039

. Calc. 0.'2'6 I- 0.16 2 Best Estimate Cu/Ni obtained from CE NPSD 1039 l

7652/0/ K

L.

TABL Comparison of Pressuriz USNRC Staff"' a i asemamammunummmmmmmmmmm mummmmmmmmmen -

argummmuumm Chemistry Factor Plant Beltline Heat No. ID Neut. IRT , Method of ident. Ident. Fluence at NRC "' MY Report Date

• Name Determin.

EOL/EFPY IRT* 05/94 04/9 Maine Axial 51989 1.48E19 56*F Generic 89.45 89 Yankee Welds 2 203 EOL: Axial 13253 56*F Generic 206.4 206 224 10/21/ Welds and 2008 3 203 12008 1.48E19 13253 56*F Generic 203

"'USNRC Staff Values are obtained from Enclosuce 1. " Pressurized Thermal Shock iable" of USNRC Letter to

• Maine Yankee Values are obtained from Maine Yankee letter to USNRC cated May 20. 1994 (MN-94-50) and Att I

r:\ne\bfj\97-011

. * ~ ~ ~

Attachment B of Enclosure 1 B.1 MN 97-66 d Thers;l Shock Values Page 2 of 2 d Maine Y2nkee:21 Hetnod of Determin. CF Cu Ni Concentration.

• Concentration.

• NRC "'

05/94 ~'04/97 05/94 04/97 05/94 04/97 Table 0.17 0.17

-: 0.16 f- Insignificant difference in CF likely due to rounding.

Best Estimate Cu/N1 obtained from CE NPSD 1039 Table Calc. 0.22 1-0.21 0.84 . 0.87.:' Insignificant difference in CF likely due to rounding.

Best Esti.aate Cu/Ni obtained from CE NPSD 1039 Calc.- .' O . 22 'O.73 Best Estimate Cu/Ni obtained from CE NPSD 1039 If dated April 12, 1994

hment A ANSTEC APERTURE

CARD j Also Available on Aperture Card 9% Mol 90 -

l-l- ,

l.

l l

~

Attachment C of Enclosure 1 MN-97-66 s

, Attachment C Maine Yankee Final Report Items 2,3, and 4  :

Generic Letter 92-01, Revision 1, Supplement 1  ;

l' Comparison of USE Licensing Values ,

USNRC's RVID and Maine Yankee's Reported Values l l

l 4

l.

l.

l l-i l

i h

I f

ri\ne\hfj\97-011 l

l

~

Table C.1 Attactament C of Enclosure 1 Comparison of Upper-Shelf Energy values Mi 97-M Page 1 of 2 USWIC Staff

• and naine Yankee

• 1/4T USE at untrrad. Method of Deterein.

E1/ETPY 1/4T USE Unt rrad USE Plant 8eltitne Material  % tron Ihre MY Report Date

• PV 5.eport Date

• My Report Date

• coments Ident . Heat it). Type F W eat NHC* gg,gypy NRC* NRC*

07/% 04/97 07/% 04/97 07/% 04/97 Matre lot, Shell g.7%51 A S3]B 1 .': 80 3 (hk ' 36 ? 1073E19 115 Dirxt NRC s 1/4 Y USE at E1/EFPY appears to be based on USE decreaw for wid (301)

Yankee Plate ratt;er than plate (242). See Section II B 2.a of los 04 50.

D4406- 1 Best Estimate tu Average of0 8406-1 & 2 Ea: trt Stelt 8 - 7%5- 2 A 53381 62 ' 5 63 ; E' 64 2

3 #73E19 84 86 '.54 MY's Int USE based on average of 2 test results with 1953 svar (137 and 129 10/21I Piete - ft - 1bs) . NRC's value appears to teclude a 3rd test with or.ly 901 shear (122 2008 D 6406 2 f t - Ibs) .

Best Istteste Cu Average of D-84M-1 & 2 Int Snell C-39R2 5 A 53381 74 3 G7X19 90 653 Plate D 4406- 3 Lwr B 8330-1 A 53381 57 1.G7X19 86 653 8est Estteate Cu Average of D 8a071 & 2 Shell Plate 0 6457-1 Lwr 8 8330-2 A $118-1 56 1.07X19 82 65s aest Estimate Cu Average of0-8407-1 & 2 Shell Piete D 8407-2 tcner S-8324-1 A 5338-1 64 1.07X19 B2 653 Snell Plate D 9407 3 I:\ne\htj\97-011 g

O

,= _ . . -

i 4

1 l

Table C.1 Attachsent C cf Enclosure I rr=narison of Upper-Shelf Energy values PN 974 Page 2 of 2 USNRC Staff'" and Maine Vant.ee **

1/4T USE at Umrred. Method of Deteratn.

E('LIEFPY 1/4T USE Untrrad USE Plant SeltlinP Mater 1al Ntstron hame ident. Heat 40. Type Mr Report Date "'

Fluente at m Date

• m W Me

• Wts Nfe" E1/EFPY WRCS NRC*

07/95 04/97 07/95 04/97 07/95 C4/97 Maine Circ. held 193571 Linde T

56 1073EI9 335 7 '. 107 1, surv. wid NRC value of 105 ft-1bs arcears to be based on fit rather than average of all

? 61:l~

YaMee 9-2C3 1042. 5And ~ $

' i . tests with : %I shear (107 ft-ibs). Maine Yankee s 1/4 Y USE at E0L has been

. - detersineu ssmg a conservative decrease in USE (433) octained from R S.199 p

• r g .* w Rev . 2. The NRC Staff's Value (478) appears everly conservative see

Sections II B 2 b of 181-94-50.

s 3.3 Sest Estimate Cm obtatned from CE HPSD-1839

[

EOL: 33A277 Linde -

?Q } 94 : - .*

(3Mk Sister Best Esttente Cu obtained from E NP50-1839 10/21/ 0091. 5 4 29- Plant Surv. Weld Initial USE obtstned from EM-622. Reference (o) 2008 s AUal 51989 L*nde .' 53 ; f 59 ' s 82Ela (.?$  ; a3.5 o ' 84 i  :-iaCl  :: CE 4 ' .' CE : see Sections 1.8 and II 8 2 c of m 44 50 Welds 124, , < g3 Mk Gener4- ' WM - E i(M%9M3 -W d gsCE N w se W & W

^

W 3 . CEN-622 - CE. savporeing MF) andsief (3E of 84pA Tnds solme sgpare NF% eof. (3E A.B.C  %' . gy,p

' 'W -'

8est Estimate Cu obtained from E WPSD-1039 Aaial 13253 Lirde ' to s

7e : } [3 - 8 82E18 [107 j l' N7 7 cN9Y $5ssh'~ . DSM - ' Direct '. MV has identified seasurements results for the sane weld wsre heet evid fLa helds and 1992. SM A2 , 3. 't 7. pig ^f lot . See Sections 1.C and II B 2 e of MN 94-50.

3-203 12008 Nw

_. (7.@ , uw77 4 4,.ima t.sEgit9A a.B C -- .. ,;. 4 - Best Estiente Cu obtained frcs E NP5D-1039 i., -f:1 m.s72 ; ';. :i" 31sterO Best Estimate Cu obtained from E HPSD 1639 13?53 -

4.;

'[11d '

s- ~ ^4 d U. . Plant s faitial USE cbtained fran EN 622. Reference (e)

> gW nsurv.:

se - -

.,  ; Weld -

"tSNRC Staff ifakes are obtained from Enclosure 2. Pressurtzed Thermal Shed Table *. of USMC Letter to Nv dated April 12. 1994

• Metw Yartee values are cDtained from Mame Yartee Letter to LEMC dated May 20.199a (m 94-50) and Attactment A

• .r \r.e\bfj\S7-oti e

4

Table C.1 Comparison of Upper Shelf USNRC Staff (D and Maine

-ummuumununu unmumumumum -

, 1/4T USE at Unirrad.

EOL/EFPY 1/4T USE Plant Beltline Material Neutron Name Ident. Heat No. Type MY Report Date NY Report Date m Fluence at NRC(D NRC(D E1/EFPY 07/95 04/97 07/95 04/97 Maine Int. Shell B-7955-1 A 5338 1  :.: 80l:  :;-f87 6 L 86 1.073E19 115 Yankee Plate D-8406 1 EOL: Int. Shell B-7955-2 A 5338-1 "l62 : [ 63I 3 64 :'- 1.073E19 84 86 10/21/ Plate 2008 D 8406-2 Int. Shell C-3982-5 A 5338-1 71 1.073E19 90 Plate D-8406-3 Lower B-8330-1 A 533B 1 57 1.073E19 86 Shell Plate D-8407 1 Lower B-8330-2 A 5338-1 56 1.073E19 82 Shell Plate D 8407-2 Lower B-8324-1 A 533B 1 64 1.073E19 82 Shell Plate D-8407-3

. r:\ne\hfj\s7-o11

Attachment C of Enclosure 1 gy Values E97 65 tkee "' Page 1 of 2 h as h e s= de Method of Determin.

Unirrad. LISE

{

MY Report Date "'

Coments 07/95 04/97 Also Avaliable i n MMure Card Direct NRC's 1/4 T USE at EOL/EFPY appears to be based on USE decrease for Weld (302) rather than plate (243). See Section II.B.2.a of MN 94 50.

Best Estimate Cu Average of D 8406-1 & 2 65% MY's Int. USE based on average of 2 test results with 2951 shear (137 and 129 ft-lbs). NRC's value appears to include a 3rd test with only 90% shear (122 ft-lbs).

Best Estimate Cu Average of D 84061 & 2 658 653 Best Estimate Cu Average of D 84071 & 2 653 Best Estimate Cu Average of D 84071 & 2 65%

I 0 -

w --

Table C.1 Comparison of Upper Shelf USNRC Staff (" and Main mummmmmm amummmmmmmmmm smusammmmmm ammmmmmmmmm -

1/4T USE at Unirrad.

EOL/EFPY 1/4T USE Plant Beltline Material Neutron Name Ident. Heat No. Type MY Report Date '8' MY Report Date cri Fluence at NRC") NRC'"

E1/EFPY 07/95 04/97 07/95 04/97 Maine Circ. Weld 193571 Linde :556 i 1:6# 1.073E19 '105 ! -

107.

Yankee 9 203 1092, SAW EOL: 33A277 Linde .

F: 194 - :E  ::>::-- :f160j 10/21/ 0091. SAW 2008 Axial 51989 Linde - 53 / i 69 : 8.82E18 75 i  :::83.5 ; s 84 J Welds 124 2-203 SAW A.B C so Axial 13253 Linde  : 70 ' i-- 70 ' i 19 - 8.82E18 -1 107 : J1872 '!119 t Welds and 1092. SAW 3-203 12008 A.B.C 13253 - + 2 '72 ;; 4 :-

X-::: ill1E

'"USNRC Staff values are obtained from Enclosure 2. " Pressurized Thermal Shock Table", of USNRC Letter to MY dated April

Maine Yankee Values are obtained from Maine Yankee Letter to USNRC dated May 20.1994 (MN-94-50) and /,ttachment A l -

r:\ne\hfj\97-o11 4% -,

1

Attachment C of Enclosure 1 nergy values E 97-66 c2 Yankee > Page 2 of 2 i

Method of Determin.

Unirrad. USE MY Report Date '"'

Coments CARD NRC"'

07/95 04/97 g jg g A y gll8 blt Oli

,g Sury. Weld NRC Value of 105 ft-lbs appears to be based on fit rather than average of all tests with 2 953 shear (107 ft-lbs). Maine Yankee's 1/4 T USE at E0L has been detennined using a conservative decrease in USE (43%) obtained from R.G.1.99.

Rev. 2. The NRC Staff's Value (47%) appears overly conservative. See Sections II.B.2.b of MN-94 50.

Best Estimate Cu obtained from CE NPSD 1039 Sister Best Estimate Cu obtained from CE NPSD 1039 Plant Surv. Weld Ir.tial USE obtained frte U.N 622, Referer.:e (o)

NRC. 1CEi . Cf J See Sections I.B and II.B.2.c of MN-94-50.

Genkrib (Gbnttric1 .l Generic l Enclosure 1 ofhiN-95-85 summart:es the results ofa CEOGtask to supplement the original

- CEN-622_' CE evala.ution supporting hit"s laisial!!SE of 84p-Ib. This value supports hf1"s EOL USE L ,l :- off9p-lb.

Best Estimate Cu obtained from J F.D 1039 p S1 ster? EDfrectj -LDirectI MY nas identified measurements results for the same weld wire heat and flux

.y$ Plant} lot. See Sections I.C and II.B.2.c of MN 94-50.

bi:l? '

htN-94-77 veripedanInisla!USEofi19p-lbs.

, Best Estimate Ce obtained from CE NPSD 1039

' $1 ster ' Best Estimate Cu obtained from CE NPSD 1039 TPlanti: Initial USE obtained from CEN 622. Reference (o) fSut;v?

WeldY

2. 1994 i

9 05;2lCI 6-00 -