ML18151A644

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Response to GL 92-01,Rev 1,Suppl 1 for Virginia Power North Anna Units 1 & 2 Beltline Matls & Surry Units 1 & 2 Rotterdam Beltline Weld Metals.
ML18151A644
Person / Time
Site: Surry, North Anna  Dominion icon.png
Issue date: 10/31/1995
From: Devan M
BABCOCK & WILCOX CO.
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Shared Package
ML18151A645 List:
References
BAW-2260, GL-92-01, GL-92-1, NUDOCS 9511280151
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BAW-2260 OCTOBER 1995 RESPONSE TO GENERIC LETTER 92-01, REVISION 1, SUPPLEMENT 1, FOR VIRGINIA POWER'S NORTH ANNA UNITS 1 AND 2 BELTLINE MATERIALS AND SURRY UNITS 1 AND 2 ROTTERDAM BELTLINE WELD METALS Fi?>!7!/*{1nn*nn B&WNUCLEAR fJ ? ) ~ TECHNOLOGIES

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  • BAW-2260 October 1995 Response to Generic Letter 92-01, Revision 1, Supplement 1 for Virginia Power's North Anna Units 1 and 2 Beltline Materials and Surry Units 1 and 2 Rotterdam Beltline Weld Metals by M. J. DeVan BWNT Document No. 77-2260-00 (See Section 4 for document signatures.)

Prepared for Virginia Power Prepared by B&W Nuclear Technologies, Inc.

Engineering and Project Services Division 3315 Old Forest Road P. 0. Box 10935 Lynchburg, Virginia 24506-0935

  • 1.0 CONTENTS INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 2.0 STATEMENT OF RESPONSE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 2.1 Part (1) to NRC Generic Letter 92-01, Revision 1, Supplement 1 ........ 2-1 2.2 Part (2) to NRC Generic Letter 92-01, Revision 1, Supplement 1 ........ 2-2 2.3 Part (3) to NRC Generic Letter 92-01, Revision 1, Supplement 1 ........ 2-9 2.4 Part (4) to NRC Generic Letter 92-01, Revision 1, Supplement 1 ....... 2-10

3.0 REFERENCES

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 4.0 CERTIFICATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 APPENDIX A . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1 List of Tables

  • 2-1.

2-2.

Chemistry Data for North Anna Unit 1 Beltline Base Metals . . . . . . . . . . . . . . . 2-3 Chemistry Data for North Anna Unit 1 Beltline Welds . . . . . . . . . . . . . . . . . . . 2-4 2-3. Chemistry Data for North Anna Unit 2 Beltline Base Metals . . . . . . . . . . . . . . . 2-5 2-4. Chemistry Data for North Anna Unit 2 Beltline Welds . . . . . . . . . . . . . . . . . . . 2-6 2-5. Chemistry Data for Surry Unit 1 Rotterdam Weld . . . . . . . . . . . . . . . . . . . . . . 2-7 2-6. Chemistry Data for Surry Unit 2 Rotterdam Weld . . . . . . . . . . . . . . . . . . . . . . 2-8 A-1. North Anna Unit 1 -- Data Summary for Pressurized Thermal Shock Calculation . A-2 A-2. North Anna Unit 1 -- Data Summary for Upper-Shelf Energy Calculation . . . . . A-4 A-3. North Anna Unit 2 -- Data Summary for Pressurized Thermal Shock Calculation. A-6 A-4. North Anna Unit 2 -- Data Summary for Upper-Shelf Energy Calculation . . . . . A-8 A-5. Surry Unit 1 -- Data Summary for Pressurized Thermal Shock Calculation .... A-10 A-6. Surry Unit 1 -- Data Summary for Upper-Shelf Energy Calculation . . . . . . . . . A-13 A-7. Surry Unit 2 -- Data Summary for Pressurized Thermal Shock Calculation .... A-16 A-8. Surry Unit 2 -- Data Summary for Upper-Shelf Energy Calculation . . . . . . . . . A-19

  • 11

1.0 INTRODUCTION

This report provides a response to the Nuclear Regulatory Commission (NRC) Generic Letter 92-01, Revision 1, Supplement 1, for the Virginia Power's North Anna Units 1 and 2 beltline materials and Surry Units 1 and 2 beltline weld materials that were fabricated by Rotterdam Dockyard Company.

Generic Letter 92-01, Revision 1, Supplement 1, was issued by the NRC on May 19, 1995 and addressed to all holders of nuclear power plant operating licensees. The generic letter was issued to require the licensees to identify, collect, and report any new data pertinent to the analysis of structural integrity of their reactor vessels and assess the impact of that data on their reactor

  • vessel integrity analyses relative to the requirements of current regulations .
  • 1-1
  • ~

2.0 STATEMENT OF RESPONSE 2.1 Part (1) to NRC Generic Letter 92-01, Revision 1, Supplement 1 Part (1) of the Generic Letter requires the Addressees to provide the following information:

"a description of those actions taken or planned to locate all data relevant to the determination of RPV integrity, or an explanation of why the existing data base is considered complete as previously submitted;"

The North Anna Units 1 and 2 reactor vessels, and portions of the Surry Units 1 and 2 reactor vessels, were fabricated by the Rotterdam Dockyard Company. To date, the following sources of information have been identified and utilized in the evaluation of the North Anna Units 1 and

  • 2 reactor vessel beltline materials and the Surry Units 1 and 2 Rotterdam weld materials:

11 Plant-specific reactor vessel materials surveillance reports Virginia Power documentation on reactor vessel materials (i.e., vendor correspondence, reports, etc.)

Because several other U.S. nuclear reactor vessels were fabricated by the Rotterdam Dockyard Company, additional information on Rotterdam weld materials and weld material surrogates may exist for the weld wire heats used in the fabrication of the North Anna and Surry reactor vessels.

To ensure that all relevant chemical and mechanical test data have been identified, confirmatory searches of all known data sources for these reactor vessel beltline materials will be reviewed.

Potential sources of additional data include:

11 Westinghouse Electric Company records

  • Rotterdam Dockyard Company fabrication records
  • Industry data bases (i.e., PREP3, PR-EDB, RVID, etc.)

2-1

  • The portions of the Surry Units 1 and 2 reactor vessels that were not fabricated by Rotterdam Dockyard Company were fabricated by Babcock & Wilcox (B&W). The response to Part (1) of Generic Letter 92-01, Revision 1, Supplement 1, for the B&W-fabricated portions of the Surry Units 1 and 2 reactor vessels will be included in the forthcoming B&W Owners Group (B&WOG) Reactor Vessel Working Group (RVWG) integrated Generic Letter response.

2.2 Part (2) to NRC Generic Letter 92-01, Revision 1, Supplement 1 Part (2) of the Generic Letter requires the Addressees to provide the following information:

"an assessment of any change in best-estimate chemistry based on consideration of relevant data;"

To develop and/or verify the best-estimate chemical contents for the North Anna Units 1 and 2 beltline materials and the Surry Units 1 and 2 beltline Rotterdam weld materials, confirmatory searches were performed that included the following sources:

Westinghouse Electric Company records Virginia Power internal documentation Industry data bases (i.e., PREP3, PR-EDB, RVID)

  • Plant-specific reactor vessel materials surveillance reports The results of these searches are listed in Tables 2-1 through 2-6. The best-estimate mean copper and nickel chemical contents and their respective standard deviations for each beltline material are also presented in these tables.
  • The Data Summary Tables for Pressurized Thermal Shock and Upper-Shelf Energy presented in documents BAW-2224 1 and BAW-22222 for the North Anna Units 1 and 2 beltline materials and the Surry Units 1 and 2 Rotterdam weld metals respectively have been revised to include the best-estimate copper and nickel chemical compositions described above. These revised tables are presented in Appendix A with the modified values shown in nonshaded boxes.

2-2

  • Table 2-1. Chemistry Data for North Anna Unit 1 Beltline Base Metals Forging 05 (Nozzle Belt Forging)

Heat No. 990286/295213 ROM" Analysis ASTM A 508 Cl. 2 0.16 0.74 Docket 50-3383

& ROM Analysis Forging 04 (Intermediate Shell Forging)

Heat No. 990311/298244 ROM" Analysis ASTM A 508 Cl. 2 0.12 0.82 Docket 50-3383

& ROM Analysis Forging 03 (Lower Shell Forging)

Heat No. 990400/292332

  • ROM" Analysis Surveillance Data Surveillance Data (Specimen VT71)

ASTM A 508 Cl. 2 ASTM A 508 Cl. 2 ASTM A 508 Cl. 2 Mean:

0.15 0.16 0.158 0.16 0.80 0.79 0.893 0.83 Docket 50-3383

& ROM Analysis WCAP-8771 4 WCAP-11777 5 Std. Dev.: 0.004 0.046

  • - RDM = Rotterdam Dockyard Company .
  • 2-3

Table 2-2. Chemistry Data for North Anna Unit 1 Beltline Welds Weld Wire 25295, Type SMIT 40 Weld 05A - Nozzle Belt to Intermediate Shell Circ. Weld ROM" Weld Qualification SMIT 89 1170 0.30 Docket 50-3383 &

Weld Qualification ROM" Weld Qualification SMIT 89 1170 0.46 Weld Qualification RDM. Weld Qualification SMIT 89 1135 0.25 Weld Qualification -

Surveillance Data SMIT 89 2275 0.33 0.17 WCAP-82336 Surveillance Data (Specimen TW58) SMIT 89 2275 0.42 0.08 WCAP-10340 7 Mean: 0.35 0.13 Std. Dev.: 0.077 0.045 Weld Wire 4278, Type S4Mo Weld 058 - Nozzle Belt to Intermediate Shell Circ. Weld ROM" Weld Qualification Surveillance Data SMIT 89 SMIT 89 1211 1211 Mean:

0.11 0.13 0.12 0.11 0.11 Docket 50-3383 &

Weld Qualification WCAP-8513 8 Std. Dev.: 0.010 Weld Wire 25531, Type SMIT 40 Weld 04 - Intermediate Shell to Lower Shell Circ. Weld Surveillance Data SMIT 89 1211 0.086 0.11 WCAP-8771 4 Surveillance Data (Specimen VW71) SMIT 89 1211 0.124 0.152 WCAP-11777 5 Mean: 0.11 0.13 Std. Dev.: 0.019 0.021

  • - RDM = Rotterdam Dockyard Company.
    • - Data obtained from RPVDATA, Reactor Vessel Materials Database, Version 1.09 2-4
  • Table 2-3. Chemistry Data for North Anna Unit 2 Beltline Base Metals Forging 05 (Nozzle Belt Forging)

Heat No. 990598/291396

}:oo~iiil!P¢:§iNP; : : :

1 ROM' Analysis ASTM A 508 Cl. 2 0.08 0.77 Docket 50-3383

& ROM Analysis Forging 04 (Intermediate Shell Forging)

  • Heat No. 990496/292424 ROM' Analysis ASTM A 508 Cl. 2 0.09 0.83 Docket 50-3383

& ROM Analysis Surveillance Data ASTM A 508 Cl. 2 0.11 0.86 WCAP-8772 10 Mean: 0.10 0.85 Std. Dev.: 0.010 0.015

  • Forging 03 (Lower Shell Forging)

Heat No. 990533/297355 ROM' Analysis ASTM A 508 Cl. 2 0.13 0.83 Docket 50-3883

& ROM Analysis

' '"*I'>'~

  • - RDM = Rotterdam Dockyard Company .
  • 2-5

Table 2-4. Chemistry Data for North Anna Unit 2 Beltline Welds Weld Wire 4278, Type S4Mo Weld 05A - Nozzle Belt to Intermediate Shell Circ. Weld 0

RDM Weld Qualification SMIT 89 1211 0.11 Docket 50-3383 &

Weld Qualification Surveillance Data SMIT 89 1211 0.13 0.11 WCAP-8513 8 Mean: 0.12 0.11 Std. Dev.: 0.010 Weld Wire 801, Type S4Mo Weld 058 - Nozzle Belt to Intermediate Shell Circ. Weld 0 0 RDM Weld Qualification SMIT 89 1211 0.18 0.1 f Weld Qualification Weld Wire 716126, Type S3Mo Weld 04 - Intermediate Shell to Lower Shell Circ. Weld 0

RDM Weld Qualification LW320 26 0.064 0.04 Weld Qualification***

0 RDM Weld Qualification LW320 26 0.062 0.08 Weld Qualification***

0 RDM Weld Qualification LW320 26 0.079 0.04 Weld Qualification***

0 RDM Weld Qualification LW320 26 0.061 0.03 Weld Qualification***

0 RDM Weld Qualification LW320 26 0.062 0.03 Weld Qualification***

Surveillance Data LW320 26 0.088 0.084 WCAP-8772 10 Mean: 0.07 0.05 Std. Dev.: 0.010 0.023

  • - RDM = Rotterdam Dockyard Company.
    • - Conservative estimate (mean plus one standard deviation) determined using data from other plants with similar materials to the beltline material, i.e., data from reactor vessels fabricated to the same material specification in the same shop and in the same time period. 1
      • - Data obtained from RPVDATA, Reactor Vessel Materials Database, Version 1.09 2-6
  • Table 2-5. Chemistry Data for Surry Unit 1 Rotterdam Weld Weld Wire 25017, Type SMIT 40 (J726)

Nozzle Belt to Intermediate Shell Circ. Weld 0.1 o**

0 RDM Weld Qualification SMIT 89 1197 0.33 Docket 50-281 11 &

Weld Qualification --

  • - RDM = Rotterdam Dockyard Company.
    • - Conservative estimate (mean plus one standard deviation) determined using data from other plants with similar materials to the beltline material, i.e., data from reactor vessels fabricated to the same material specification in the same shop and in the same time period. 2
      • - Data obtained from RPVDATA, Reactor Vessel Materials Database, Version 1.09 -:***

2-7

Table 2-6. Chemistry Data for Surry Unit 2 Rotterdam Welds Weld Wire 4275, Type S4Mo (L737)

Nozzle Belt to Intermediate Shell Circ. Weld Weld Wire 0227, Type S3Mo (R3008)

Intermediate Shell to Lower Shell Circ. Weld Surveillance Data Grau Lo LW320 0.19 0.56 WCAP-8085 12 Surveillance Data (Specimen W14) Grau Lo LW320 0.184 0.53 WCAP-11499 13 Mean: 0.19 *0.55 Std. Dev.: 0.003 0.015

  • - Conservative estimate (mean plus one standard deviation) determined using data from other plants with similar materials to the beltline material, i.e., data from reactor vessels fabricated to the same material specification in the same shop and in the same time period. 2 2-8
  • 2.3 Part (3) to NRC Generic Letter 92-01, Revision 1, Supplement 1 Part (3) of the Generic Letter requires the Addressees to provide the following information:

"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;"

If the mean copper and nickel concentrations of surveillance materials are less than the mean copper and nickel concentrations of the beltline materials which they are intended to represent, the use of the surveillance data to calculate ARTNDT may under-predict the embrittlement of the reactor vessel beltline. Therefore, Regulatory Guide 1.99, Revision 2, 14 requires measured values of ARTNnT to be adjusted when there is clear evidence that the mean copper and nickel concentrations of the surveillance material differ from those of the vessel material. Position 2.1 of Regulatory Guide 1.99, Revision 2, specifies that measured values of ARTNnT should be multiplied by the ratio of the chemistry factor for the vessel material (determined in accordance with Position 1.1 of Regulatory Guide 1.99, Revision 2) to that of the surveillance material (also

  • determined in accordance with Regulatory Guide 1.99, Revision 2, Position 1.1). When the mean copper and nickel concentrations of the surveillance material are greater than the mean copper and nickel concentrations of the beltline material, it is not necessary to apply this ratio procedure, since its application would result in less limiting values of ARTNDT for the beltline material.

The following is a list of the beltline materials for North Anna Units 1 and 2 and Rotterdam weld materials for Surry Units 1 and 2 that have used surveillance data in evaluations of reactor vessel integrity:

  • 2-9

North Anna Unit 1 990400/292332 25531 Lower Shell Forging 03 Interm. to Lower Shell Circ. Weld North Anna Unit 2 990496/292424 Interm. Shell Forging 04 716126 Interm. to Lower Shell Circ. Weld Surry Unit 1 Surry Unit 2 0227 Interm. to Lower Shell Circ. Weld An examination of the chemical composition data which substantiate the mean values for the beltline materials reveals that the mean copper and nickel concentrations for the surveillance materials are equal to or greater than the mean copper and nickel concentrations determined herein for the corresponding beltline materials. (Values used in the determination of mean copper and nickel concentrations for surveillance materials are presented in bold in Tables 2-1 through 2-6). Therefore, it may be concluded that prior uses of surveillance data in evaluations of reactor vessel integrity are conservative with or without application of the Regulatory Guide 1.99, Revision 2, Position 2.1, ratio procedure.

2.4 Part (4) to NRC Generic Letter 92-01, Revision 1, Supplement 1 Part (4) of the Generic Letter requires the Addressees to provide the following information:

"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 10CFRS0.60, 10CFRS0.61, Appendices G and H to 10CFR 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 should include consideration of Position 2.1 of Regulatory Guide 1.99, Revision 2, as applicable, and any new data."

Best-estimate (mean) copper and nickel concentrations were previously provided in the Generic Letter 92-01, Revision 1 responses for North Anna Units 1 and 2 1 and Surry Units 1 and 2. 2

. After consideration of all relevant data, the mean copper and nickel concentrations previously provided for the North Anna Unit 1 and 2 beltline materials and the Surry Unit 2 beltline Rotterdam weld metal are being updated as follows:

2-10

  • North Anna Unit 1 Lower Shell 0.15 0.80 0.16 0.83 Forging (03)

Nozzle to Int. 0.30 0.17 0.35 0.13 Shell Circ.

Weld 05A Nozzle to Int. 0.11 0.11 0.12 0.11 Shell Circ.

Weld 05B Int. to Lower 0.09 0.11 0.11 0.13 Shell Circ.

Weld 04 North Anna Unit 2 Intermediate 0.09 0.83 0.10 0.85 Shell Forging (04)

Nozzle to Int. 0.11 0.11 0.12 0.11 Shell Circ.

Weld 05A Int. to Lower 0.09 0.08 0.07 0.05 Shell Circ.

Weld 04

  • 2-11

Surry Unit 2 Int. to Lower 0.19 0.56 0.19 0.55 Shell Circ.

R3008 There are no changes to previously reported mean copper and nickel concentrations for Surry Unit 1 Rotterdam beltline weld metal.

Pressurized Thermal Shock Evaluations As a result of the data collected, the mean copper and nickel concentrations for the North Anna Unit 1 lower shell Forging 03, nozzle belt/intermediate shell circumferential Welds 05A and 05B, and the intermediate/lower shell circumferential Weld 04 beltline materials have increased.

Despite the chemical composition changes, the lower shell Forging 03 remains the limiting material with respect to pressurized thermal shock events North Anna Unit 1. (The limiting material determination is based on the relative margin between the 10CFR50.6l 15 pressurized thermal shock reference temperature (RTPTs) and the applicable screening criterion.) After consideration of all relevant chemical composition data, the end-of-license RT PTs value for North Anna Unit 1 lower shell Forging 03 increased from 227.7°F to 238.9°F, versus a screening criterion of 270°F. (This value is based on an end-of-license fluence of 3.95E+l9 n/cm2 , an initial RTNDT of 38°F, a margin term of 34°F, and the copper and nickel chemical composition described above.) Surveillance data for Forging 03 indicates that the amount of embrittlement is less than the 10CFR50.61 projected value. (When surveillance data is considered, the applicable chemistry factor is reduced from 115°F to 73.5°F.) It is concluded that the chemical composition changes determined herein do not cause the calculated RTPTs value for any North Anna Unit 1 beltline material to exceed the PTS screening criteria.

Similarly, the mean copper and nickel concentrations for the North Anna Unit 2 intermediate shell Forging 04 and nozzle/intermediate shell circumferential Weld 05A have increased, while the mean copper and nickel concentrations for the intermediate/lower shell circumferential Weld 2-12

l

  • 04 have decreased. Despite the chemical composition changes, the lower shell Forging 03 remains the limiting material for North Anna Unit 2 with respect to PTS. Because no new data is currently available for North Anna Unit 2 Forging 03, the PTS evaluation for North Anna Unit 2 is conservative and remains valid.

For Surry Unit 1, the limiting material with respect to PTS is the B&W-fabricated Linde 80 weld metal, SA-1585. The evaluation of Surry Units 1 and 2 Linde 80 weld materials is presented under separate cover. 16 For Surry Unit 2 the limiting material with respect to PTS is the intermediate/lower shell circumferential Rotterdam Weld R3008. Because the mean copper concentration for this weld is unchanged, and the mean nickel concentration is less than that previously reported,2 it is concluded that the current PTS evaluation for Surry Unit 2 is conservative and remains valid.

Low Temperature Overpressure Protection System (LTOPS) and Pressure-Temperature (P-T)

Limit Evaluations

  • The current LTOPS and P-T limit evaluations for North Anna Unit 1 are based on the limiting weld identified as Weld 04 (intermediate/lower shell circumferential weld).

surveillance data.

The adjusted

_ reference temperature (ART) for the applicable time period was calculated on the basis of Since the recorded copper and nickel concentrations for the surveillance material are greater than the best-estimate copper and nickel concentrations for the beltline material, the current LTOPS and P-T limit evaluations for North Anna Unit 1 are conservative and remain valid.

The current LTOPS and P-T limit evaluations for N9rth Anna Unit 2 are based on the limiting material identified as Forging 03 (Lower Shell Forging). The ART for the applicable time period was calculated in accordance with the Regulatory Guide 1.99, Revision 2, Position 1.1 guidelines.

Since no new data is currently available for this limiting material, the current LTOPS and P-T

. limit evaluation for North Anna Unit 2 remain valid.

Recently submitted LTOPS and P-T evaluations for Surry Units 1 and 2 were based on the Surry

  • Unit 1 intermediate/lower shell circumferential weld SA-1585, 17 which is a Linde 80 weld material. This material remains the limiting material for both Surry Units with respect to LTOPS 2-13

and P-T limit evaluations. The response to Part (4) of NRC Generic Letter 92-01, Revision l,

  • Supplement 1, for the B&W Owners Group (B&WOG) Reactor Vessel Working Group (RVWG)

Linde 80 weld metals is included in the B&WOG RVWG integrated response. 16 Upper-Shelf Energy Evaluations The response to the NRC closure letter for Generic Letter 92-01, Revision 1, for the North Anna Unit 1 and 2 reactor vessel beltline materials2 demonstrated compliance with the upper-shelf energy (USE) requirements of I0CFR50 Appendix G. 18 Regulatory Guide 1.99, Revision 2 models the percent drop in USE as a function of fluence and mean copper content only. As described above, the mean copper concentrations for several North Anna beltline materials have increased. However, after consideration of the revised copper concentrations, the T/4 end-of-license USE values for the North Anna Units 1 and 2 beltline materials remain above the 50 ft-lb level.

The limiting materials with respect to USE for Surry Unit 1 and 2 are the Linde 80 weld metals.

Because the mean copper concentrations for the Rotterdam weld metals in the Surry Units 1 and

  • 2 reactor vessels beltline remain unchanged from those previously reported, the Linde 80 weld metals remain the limiting materials with respect to USE. The equivalent margin analyses for these weld materials as documented in Topical Reports BAW-2178PA 19 and BAW-2192PA20 remain valid.

2-14

3.0 REFERENCES

1. M. J. DeVan, "North Anna Units 1 and 2 Response to Closure Letter for NRC Generic Letter 92-01, Revision 1," BAW-2224, B&W Nuclear Technologies, Inc., Lynchburg, Virginia, July 1994.
2. M. J. DeVan and K. K. Yoon, "Response to Closure Letters to Generic Letter 92-01~

Revision l," BAW-2222, B&W Nuclear Technologies, Inc., Lynchburg, Virginia, June 1994.

3. Letter from C. M. Stallings, Virginia Electric and Power Company to Harold B. Denton, Office of Nuclear Reactor Regulation, "Pressure Vessel Fracture Toughness Properties, North-Anna Power Station, Unit Nos. 1 and 2," Docket No. 50-338, December 11, 1978, Public Document Accession No. 7812150277.
4. J. A. Davidson and J. H. Phillips, "Virginia Electric and Power Company North Anna Unit No. 1 Reactor Vessel Radiation Surveillance Program," WCAP-8771, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, September 1976.
5. S. E. Yanichko, et al., "Analysis of Capsule U from the Virginia Electric and Power Company North Anna Unit 1 Reactor Vessel Radiation Surveillance Program," WCAP-11777, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, February 1988 .

. 6. S. E. Yanichko, D. J. Lege, and J. H. Phillips, "Tennessee Valley Authority Sequoyah Unit No. 1 Reactor Vessel Radiation Surveillance Program," WCAP-8233, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, November 1973.

7. S. E. Yanichko, et al., Analysis of Capsule T from the Tennessee Valley Authority Sequoyah Unit 1 Reactor Vessel Radiation Surveillance Program," WCAP-10340, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, May 1983.
  • 8. J. A. Davidson, J. H. Phillips, and S. E. Yanichko, "Tennessee Valley Authority Sequoyah Unit No. 2 Reactor Vessel Radiation Surveillance Program," WCAP-8513, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, November 1975.
9. RPVDATA, Reactor Vessel Materials Database, Version 1. 0, developed for Westinghouse Owners Group by ATI Consulting, dated July 28, 1995.

.10. J. A. Davidson, J. H. Phillips, and S. E. Yanichko, "Virginia Electric and Power Company

  • North Anna Unit No. 2 Reactor Vessel Radiation Surveillance Program," WCAP-8772, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, November 1976.

3-1

11. Letter from C. M. Stallings, Virginia Electric and Power Company to E.G. Case, Office
  • of Nuclear Reactor Regulation, "Reactor Vessel Material Surveillance Program," Docket No. 50-281, January 23, 1978, Public Document Accession No. 780260154.
12. S. E. Yanichko and D. J. Lege, "Virginia Electric and Power Company Surry Unit No.

2 Reactor Vessel Radiation Surveillance Program, WCAP-8085, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, June 1973.

13. S. E. Yanichko and V. A. Perone, Analysis of Capsule V from the Virginia Electric Power Company Surry Unit 2 Reactor Vessel Radiation Surveillance Program," WCAP-11499, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, June 1987.
14. U. S. Nuclear Regulatory Commission, "Radiation Embrittlement of Reactor Vessel Materials," Regulatory Guide 1.99. Revision 2, May 1988.
15. Code of Federal Regulations, Title 10, Part 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events."
16. M. J. DeVan, "B&W Owners Group Reactor Vessel Working Group Response to Generic Letter 92-01, Revision 1, Supplement 1," BAW-2257. Revision 1, B&W Nuclear Technologies, Inc., Lynchburg, Virginia, October 1995.
17. M. J. Malone, "Surry Units 1 and 2 Heatup and Cooldown Limit Curves for Normal

.Operation," WCAP-14177, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania,

  • October 1994.
18. Code of Federal Regulations, Title 10, Part 50, "Domestic Licensing of Production and Utilization Facilities," Appendix G. Fracture Toughness Requirements.
19. K. K. Yoon, "Low Upper-Shelf Toughness Fracture Analysis of Reactor Vessels ofB&W Owners Group Reactor Vessel Working Group for Level C & D Service Loads," BAW-2178PA, B&W Nuclear Technologies, Inc., Lynchburg, Virginia, April 1994.
20. K. K. Yoon, "Low Upper-Shelf Toughness Fracture Analysis of Reactor Vessels ofB&W Owners Group Reactor Vessel Working Group for Load Level A & B Conditions," BAW-2192PA, B&W Nuclear Technologies Inc., Lynchburg, Virginia, April 1994.

3-2

4.0 CERTIFICATION This report accurately responds to the information requested in Generic Letter 92-01, Revision 1, Supplement 1.

1n£P,vC-M. J. ~ ' Engineer III Date Materials & Structural Analysis Unit This report has been reviewed for technical content and accuracy.

  • & ~ r y EngineerOc,f_

Materials & Structural Analysis Unit Date Verification of independent review.

K. E. Moore, Manager Date Materials & Structural Analysis Unit This report is approved for release.

10/~1/qs D. L. Howell ' Date Program Manager

  • 4-1

APPENDIX A

  • Revised Pressurized Thermal Shock and Upper-Shelf Energy Data Summary Tables A-1

Table A-1. North Anna Unit 1 -- Data Summary for Pressurized Thermal Shock Calculation IS Neutron Method of Method of Beltline Fluence at IRTNDT Determin. Chemistry Determin.

Material Heat No. 30.7 EFPY F IRTNDT Factor CF %Cu %Ni i!iH.f~t!l/il:§hili

rgt§.1ns::g1
ij2wit!§'.ij;il!::I 0.13<9) 0.13<9)

L_. _ *

  • Table A-1. (cont.) North Anna Unit 1 -- Data Summary for Pressurized Thermal Shock Calculation NOTES:
a. Values obtained from WCAP-11777.A-I (Nozzle belt shell forging and nozzle belt shell-to-intermediate shell circumferential weld fluences are 7% of maximum vessel inner surface fluence.)
b. Initial reference temperature was determined in accordance with MTEB 5-2A-2 guidelines for cases where the reference temperature was not determined using ASME Boiler and Pressure Vessel Code,Section III, NB-23 31,A-3 methodology.
c. Initial reference temperature was determined from tests on material fabricated from the same heat of the beltline material.
d. Initial reference temperature was determined from the mean value of tests on material of similar types.
e. Chemistry factor was determined from the chemistry factor tables in Regulatory Guide 1.99, Revision 2.A-4
f. Chemistry factor was determined from surveillance data (WCAP-11777) via procedures described in Regulatory Guide 1.99, Revision 2.
g.
  • Best-estimate values.

A-3

Table A-2. North Anna Unit 1 -- Data Summary for Upper-Shelf Energy Calculation 1/4T Method of 1/4T USE Neutron Determin.

Beltline Material at 30. 7 Fluence at Unirrad. Unirrad.

Material Heat No. Type EFPY 30. 7 EFPY USE USE

~:::::::::::::::::::::::::::::::::::::::::::::*.;::::::
  • Table A-2. (cont.) North Anna Unit 1 -- Data Summary for Upper-Shelf Energy Calculation NOTES:
a. End-of-life neutron fluence at T/4 from inner wall calculated using Regulatory Guide 1.99, Revision 2/4 neutron fluence attenuation methodology from ID value. (Vessel thickness= 7.667 in.)
b. The unirradiated USE was determined on the basis of a comparison with similar materials to the beltline material.
c. The unirradiated USE for the forgings was determined from weak oriented specimens. The unirradiated USE for the weld was determined from test data.
d. The unirradiated USE was determined using reported data from other plants with the same weld wire heat number (Sequoyah Units 1 and 2A-5,A-6).
e. Weld 05A is 94% of the thickness of the nozzle belt shell-to-intermediate shell circumferential weld and Weld 05B is the remainder. Therefore, it is not necessary to evaluate the end-of-life USE for Weld 05B because it is not at the T/4 location.

A-5

Table A-3. North Anna Unit 2 -- Data Summary for Pressurized Thermal Shock Calculation IS Neutron Method of Method of Beltline Fluence at IRTNDT Determin. . Chemistry Determin.

Material Heat No. 32 EFPY F IRTNDT Factor CF %Cu %Ni

Table A-3. (cont.) North Anna Unit 2 -- Data Summary for Pressurized Thermal Shock Calculation NOTES:

a. Values obtained from WCAP-12497.A-7 (Nozzle belt shell forging and nozzle belt shell-to-intermediate shell circumferential weld fluences are 7% of maximum vessel inner surface fluence.)
b. Initial reference temperature was determined in accordance with MTEB 5-2A-2 guidelines for cases where the reference temperature was not determined using ASME Boiler and Pressure Vessel Code,Section III, NB-23 31 ,A-3 methodology.
c. Initial reference temperature was determined from tests on material fabricated from the same heat of the beltline material.
d. Initial reference temperature was determined from the mean value of tests on material of similar types.
e. Chemistry factor was determined from the chemistry factor tables in Regulatory Guide 1.99, Revision 2.A-4
f. Chemistry factor was determined from surveillance data (WCAP-12497) via procedures described in Regulatory Guide 1.99, Revision 2.
g. Best-estimate values.

A-7

Table A-4. North Anna Unit 2 -- Data Summary for Upper-Shelf Energy Calculation 1/4T Method of 1/4T Neutron Determin.

Beltline Material USE at Fluence at Unirrad. Unirrad.

Material Heat No. Type 32 EFPY 32 EFPY USE USE 1

11i11:111:1:::1111.

ffUMt§!§/% %

t 1

1:1~tll~l !!!!l!;!i!i ili!/i:i !!i!

?VY~iij 9.§§\\@:tt f NB to IS*.** :* ..

,,ret¥Jera:: :

(ID 6%)

'i}

Table A-4. (cont.) North Anna Unit 2 -- Data Summary for Upper-Shelf Energy Calculation NOTES:

a. End-of-life neutron :fluence at T/4 from inner wall calculated using Regulatory Guide 1.99, Revision 2/4 neutron :fluence attenuation methodology from ID value. (Vessel thickness= 7.667 in.)
b. Letter from W. L. Stewart, Virginia Electric and Power Company, to U.S. Nuclear Regulatory Commission,

Subject:

Virginia Electric and Power Company North Anna Power Station Unit 2 Selection of Limiting Forged Material for Low Upper-Shelf Energy Considerations, dated December 29, 1992.A-s

c. The unirradiated USE for the forgings was determined from weak oriented specimens. The unirradiated USE for the weld was determined from test data.
d. The unirradiated USE was determined using reported data from other plants with the same weld wire heat number (Sequoyah Unit 2A-s).
e. Weld 05A is 94% of the ,thickness of the nozzle belt shell-to-intermediate shell circumferential weld and Weld 05B is the remainder. Therefore, it is not necessary to evaluate the end-of-life USE for Weld 05B because it is not at the T/4 location.

A-9 i . '

Table A-5. Surry Unit 1 -- Data Summary for Pressurized Thermal Shock Calculation IS Neutron Method of Method of Beltline Fluence at IRTNDT Determin. Chemistry Determin.

Material Heat No. 28.8 EFPY F IRTNDT Factor CF %Cu %Ni

Beltline IS Neutron Fluence at IRTNDT

  • Method of Determin. Chemistry Method of Determin.

Material Heat No. 28.8 EFPY F Factor CF %Cu %Ni A-11

Table A-5.. (cont.) Surry Unit . 1 -- Data Summary. for Pressurized Thermal . Shock Calculation NOTES:

a. Values obtained from WCAP-11015, Revision 1,A-9 or WCAP-11017, Revision 1,A-io (Upper shell forging and intermediate-to-upper shell circumferential weld fluences are 10% of maximum vessel inner surface fluence.)
b. Values determined from data in Material Test Report.
c. Since the initial RTNDT data for the Rotterdam welds are similar to the Linde 80 class welds, the initial RTNDT for the Rotterdam welds are bounded by Linde 80 weld metal generic values. Therefore, IRTNDT = OF, and a cr1 = 20F are estimated for the Rotterdam welds. See SECY 82-465,A- 11 "Pressurized Thermal Shock."
d. Mean value from data in BAW-1803, Revision 1,A- 12
e. Initial RTNDT* cr1, and Chemistry Factor for weld metal SA-1526 are based on NRC Safety Evaluation related to Amendment No. 176 to Facility Operating License No. DPR-50, Three Mile Island Nuclear Station, Unit No. 1, Docket No. 50-289.A- 13 Chemistry Factor weld metal WF-25 was determined using TMil and B&WOG surveillance data for weld metal WF-25 and SI surveillance data for weld metal SA-1526. These surveillance welds were fabricated with the same wire heat. The TMII and B&WOG surveillanc~ data were obtained from BAW-1803, Revision 1.
f. Chemistry Factor for plate C4415-1 was determined using Sl surveillance data as reported in WCAP-11415.A- 14
g. Chemistry Factor for weld metal SA-1585 and weld metal SA-1650 was determined using B&WOG surveillance data for weld metal SA-1585 and PB1 surveillance data for weld metal SA-1263. These surveillance welds were fabricated with the same wire heat. The B&WOG surveillance data were obtained from BAW-1803, Revision 1. The PBI 30 ft-lb transition temperature shift data were also obtained from BAW-1803, Revision 1, while the fluence data for the capsules were obtained from WCAP-12794, Revision 2.A-t 5 *
h. Values obtained from BAW-2150.A- 16
1. Best-estimate value.

J. Values obtained from BAW-212IP.A-I 7

Table A-6. Surry Unit 1 -- Data Summary for Upper-Shelf Energy Calculation

\,.. .,.

1/4T Method of 1/4T USE Neutron Determin.

Beltline Material at 28.8 Fluence at Unirrad. Unirrad.

Material Heat No. Type EFPY 28.8 EFPY USE USE A-13

1/4T Method of 1/4T USE Neutron Determin.

Beltline Material at 28.8 Fluence at Unirrad. Unirrad.

Material Heat No. Type EFPY 28.8 EFPY USE USE

  • Table A-6. (cont.) Surry Unit 1 -- Data Summary for Upper-Shelf Energy Calculation NOTES:
a. Rotterdam welds were considered in the equivalent margin analyses documented in Topical Reports BAW-2I78PAA-IS and BAW-2I92PAA* 19 and found to be acceptable by the NRC, as stated in the applicable Safety Evaluation Reports.
b. The approved equivalent margin analyses in the Topical Reports BAW-2192P A and BAW-2 l 78P demonstrates compliance with requirements of I0CFR50, Appendix G.A-zo
c. Values obtained from BAW-2I92A, based on a conservative estimated wall thickness of 7.75 inches. (Rotterdam fabrication report indicates that the minimum vessel wall thickness is 8.0 inches.) Fluence estimates do not include the benefit of the installation of flux suppression inserts beginning with Surry Unit I Cycle 13. End-of-life extension flux reduction targets for Surry Unit I are 2.5E+I9 n/cm2 at 0° l/4T location.and 7.5E+I8 n/cm2 at the 45° inner surface location.
d. Unirradiated USE is 65% of the USE from a longitudinal oriented specimens as defined in MTEB 5-2.A-z
e. Unirradiated USE is determined from transverse oriented specimens (WCAP-11415A- 14).
f. Unirradiated USE is determined from surveillance weld specimens (BAW-1803, Revision 1A- 12).
g. Unirradiated* USE is determined using data from other plants with similar materials to the beltline material.

A-15

Table A-7. Surry Unit 2 -- Data Summary for Pressurized Thermal Shock Calculation IS Neutron Method of Method of Beltline Fluence at IRTNDT Determin. Chemistry Determin.

Material Heat No. 29.4 EFPY F IRTNDT Factor CF %Cu %Ni

Beltline IS Neutron Fluence at IRTNDT Method of Determin. Chemistry Method of Determin.

  • \. ,.,_

Material Heat No. 29.4 EFPY F Factor CF %Cu %Ni A-17

Table A-7. (cont.) Surry Unit 2 -- Data Summary for Pressurized Thermal Shock Calculation NOTES:

a. Values obtained from WCAP-11015, Revision 1,A-9 or WCAP-11017, Revision 1.A- 10 (Upper shell forging and intermediate-to-upper shell circumferential weld fluences are 10% of maximum vessel inner surface fluence.)
b. Values determined from data in Material Test Report.
c. Since the initial RTNDT data for the Rotterdam welds are similar to the Linde 80 class welds, the initial RTNDT for the Rotterdam welds are bounded by Linde 80 weld metal generic values. Therefore, IRTNDT = OF, and a cr1 = 20F are estimated for the Rotterdam welds. See SECY 82-465,A- 11 "Pressurized Thermal Shock."
d. Mean value from data in BAW-1803, Revision LA- 12
e. Chemistry Factor for plate C4339-1 was determined using S2 surveillance data as reported in WCAP-11499.A-21
f. Chemistry Factor for weld metal R3008 was determined using S2 surveillance data as reported in WCAP-11499.
g. Chemistry Factor for weld metal SA-1585 was determined using B&WOG surveillance data for weld metal SA-1585 and PBl surveillance data for weld metal SA-1263. These surveillance welds were fabricated with the same wire heat. The B&WOG surveillance data were obtained from BAW-1803, Revision 1. The PBl 30 ft-lb transition temperature shift data were also obtained fromBAW-1803, Revision 1, while the fluence data for the capsules were obtained from WCAP-12794, Revision 2.A-ts
h. Values obtained from BAW-2150.A- 16
1. Best-estimate value.

J. Values obtained from BAW-2121P.A-t 7

  • I Table A-8. Surry Unit 2 -- Data Summary for Upper-Shelf Energy Calculation
  • I, .,

1/4T Method of 1/4T USE Neutron Determin.

Beltline Material at 29.4 Fluence at Unirrad. Unirrad.

Material Heat No. Type EFPY 29.4 EFPY USE USE A-19

~ - - - - - - - - - - - - - - - -- - - - ----

1/4T Method of 1/4T USE Neutron Determin.

Beltline Material at 29.4 Fluence at Unirrad. Unirrad.

Material Heat No. Type EFPY 29.4 EFPY USE USE

+,.,,.,===~====~

Table A-8. (cont.) Surry Unit 2 -- Data Summary for Upper-Shelf Energy Calculation

  • \, '

NOTES:

a. Rotterdam welds were considered in the equivalent margin analyses documented in Topical Reports BAW-2178PAA-is and BAW-2192PAA- 19 and found to be acceptable by the NRC, as stated in the applicable Safety Evaluation Reports.
b. The approved equivalent margin analyses in the Topical Reports BAW-2192PA and BAW-2178P demonstrates compliance with requirements of I0CFR50, Appendix G.A-zo
c. Values obtained from BAW-2192A, based on a conservative estimated wall thickness of 7.75 inches. (Rotterdam fabrication report indicates that the minimum vessel wall thickness is 8.0 inches.)
d. Unirradiated USE is 65% of the USE from a longitudinal oriented specimens as defined in MTEB 5-2.A-z
e. Unirradiated USE is determined from transverse oriented specimens (WCAP-11499A-21 ).
f. Unirradiated USE is determined using data from other plants with similar materials to the beltline material.
g. Unirradiated USE is determined from surveillance weld specimens (BAW-1803, Revision 1A- 12).

A-21

"* t '

  • ...i.1 A-1.

APPENDIX A References S. E. Yanichko, et al., "Analysis of Capsule U from the Virginia Electric and Power Company North Anna Unit 1 Reactor Vessel Radiation Surveillance Program," WCAP-11777, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, February 1988.

A-2. U. S. Nuclear Regulatory Commission, Standard Review Plan, Branch Technical Position MTEB 5-2, Revision 1, "Fracture Toughness Requirements, NUREG-0800, July 1981.

A-3. American Society of Mechanical Engineers, Boiler and Pressure Vessel Code,Section III, "Nuclear Power Plant Components," NB-2331, 1989 Edition.

A-4. U. S. Nuclear Regulatory Commission, "Radiation Embrittlement of Reactor Vessel Materials," Regulatory Guide 1.99, Revision 2, May 1988.

A-5. J. A. Davidson, J. H. Phillips, and S. E. Yanichko, "Tennessee Valley Authority Sequoyah Unit No. 2 Reactor Vessel Radiation Surveillance Program," WCAP-8513, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, November 1975.

A-6. S. E. Yanichko, D. J. Lege, and J. H. Phillips, "Tennessee Valley Authority Sequoyah A-7.

Unit No. 1 Reactor Vessel Radiation Surveillance Program," WCAP-8233, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, November 1973.

E. Terek, S. L. Anderson, and L. Albertin "Analysis of Capsule U from the Virginia Electric and Power Company North Anna Unit 2 Reactor Vessel Radiation Surveillance Program," WCAP-12497, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, January 1990.

A-8. Letter from W. L. Stewart, Virginia Electric and Power Company, to U. S. Nuclear Regulatory Commission, "Virginia Electric and Power Company North Anna Power Station Unit 2 Selection of Limiting Forged Material for Low Upper-Shelf Energy Considerations," December 28, 1992.

A-9. C. C. Heinecke, et al., "Surry Units 1 and 2 Reactor Vessel Fluence and RTPTs Evaluations," WCAP-11015, Revision 1, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, April 1987.

A-10. C. C. Heinecke, et al., "Surry Units 1 and 2 Reactor Vessel Fluence and RTPTs Evaluations for Consideration of Life Extension," WCAP-11017. Revision

  • 1, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, April 1987.

A-11. U. S. Nuclear Regulatory Commission, "Pressurized Thermal Shock (PTS)," SECY 465, Nuclear Regulatory Commission, Washington, D. C., November 23, 1982.

  • A-22
  • A-12. A. L. Lowe, Jr. and J. W. Pegram, "Correlations for Predicting the Effects of Neutron Radiation on Linde 80 Submerged-Arc Welds," BAW-1803, Revision I, B&W Nuclear Technologies, Inc., Lynchburg, Virginia, May 1991.

A-13. Letter from R. W. Hernan, Nuclear Regulatory Commission, to T. G. Broughton, GPU Nuclear Corporation, Issuance of Amendment No. 176 for TSCR No. 207 (TAC No.

M86085), dated August 16, 1993.

A-14. S. E. Yanichko and V. A. Perone, "Analysis of Capsule V from the Virginia Electric Power Company Surry Unit 1 Reactor Vessel Radiation Surveillance Program," WCAP-11415, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, February 1987.

A-15. S. L. Anderson and A. H. Fero, "Reactor Cavity Neutron Measurement Program for Wisconsin Electric Power Company Point Beach Unit 1," WCAP-12794, Revision 2, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, December 1992.

A-16. C. A. Ouellette, "Materials Information for Westinghouse-Designed Reactor Vessels Fabricated by B&W," BAW-2150, B&W Nuclear Technologies, Inc., Lynchburg, Virginia, December 1990.

A-17. L.B. Gross, "Chemical Composition ofB&W Fabricated Reactor Vessel Beltline Welds,"

BAW-2121P, B&W Nuclear Technologies, Inc., Lynchburg, Virginia, April 1991.

  • A-18. K. K. Yoon, "Low Upper-Shelf Toughness Fracture Analysis of Reactor Vessels ofB&W Owners Group Reactor Vessel Working Group for Level C & D Service Loads," BAW-2178P, B&W Nuclear Technologies, Inc., Lynchburg, Virginia, February 1993.

A-19. K. K. Yoon, "Low Upper-Shelf Toughness Fracture Analysis of Reactor Vessels ofB&W Owners Group Reactor Vessel Working Group for Load Level A & B Conditions," BAW-2192P, Revision 1, B&W Nuclear Technologies Inc., Lynchburg, Virginia, December 1993.

  • A-20. Code of Federal Regulations, Title 10, Part 50, "Domestic Licensing of Production and Utilization Facilities," Appendix G, Fracture Toughness Requirements .

A-21. S. E. Yanichko and V. A. Perone, "Analysis of Capsule V from the Virginia Electric Power Company Surry Unit 2 Reactor Vessel Radiation Surveillance Program," WCAP-11499, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, June 1987.

  • A-23
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