ML20154E724

Rev 0 to WCAP-15069, Evaluation of Pressurized Thermal Shock for Vogtle Electric Generating Plant Unit 1

ML20154E724
Person / Time
Site:	Vogtle
Issue date:	09/30/1998
From:	Laubham T WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20154E714	List: ML20154E707 ML20154E724
References
WCAP-15069, WCAP-15069-R, WCAP-15069-R00, NUDOCS 9810080272
Download: ML20154E724 (21)

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15069 Evaluation of Pressurized Thermal Shock for Vogtle Electric Generating Plant Unit 1 T. J. Laubham September 1998 Approved: ( C. H. Boyd, Manager ( \\ Equipment & Materials Technology Approved: D. M. Trombola, Manager Mechanical Systems Integration Westinghouse Bectric Company Energy Systems P.O. Box 355 l Pittsburgh, PA 152304355 @1998 Westinghouse Bectric Company All Rights Reserved o:\\WCAP15069. doc:lt4)91598 i

iii l TABLE OF CONTENTS - LIST OFTABLES...... .iv g l l LIST OF FIGURES. .v PREFACE.. ... vi EXECUTIVE

SUMMARY

(OR) ABSTRACT.. ....vii i INTRODUCTION. ..1-1 2 PRESSURIZED THERMAL SHOCK RULE... .2-1 3 METHOD FOR CALCULATION OF RTm...... ... 3-1 4 VERIFICATION OF PLANT SPECIFIC MATERIAL PROPERTIES., . 4-1 5 NEUTRON FLUENCE VALUES... .... 5-1 6 DETERMINATION OF RTm VALUES FOR ALL BELT

• INE REGION MATERIALS.... 6-1 7

CONCLUSION...... . 7-1 ( '8 REFERENCES. . 8-1 )l, 0:\\WCAP15069. doc 1b-071598 Revision 0

iv LIST OF TABLES ' TaHe 1 Vogtle Unit 1 Reactor Vessel Beltline Unirradiated Material Properties.. .4-3 Table 2 Fluence (E > 1.0 MeV) on Pressure Vessel Clad / Base Interface for Vogtle Unit I at 36 (EOL) and 54 (Life Extension) EFPY, . 5-1 Table 3 Interpolation of Chernistry Factors Using Tables 1 and 2 of 10 CFR 50.61. .6-2 Table 4 Calculation of Chemistry Factors using Surveillance Capsule Data Per R.egulatory Guide 1.99, Revision 2, Position 2.1. .6-3 Table 5 .RTers Calculation for Vogtle Unit 1 Beltline Region Materials at EOL (36 EFPY).. .6-4 Table 6 RTvrs Calculation for Vogtle Unit 1 Beltline Region Materials at Life Extension (54 EFPY). .6-5 o:\\WCAP15069. doc:lt4r71598 Revision 0

v LIST OF FIGURES l Figure 1 Identification and Location of Beltline Region Materials for Vogtle Unit 1 Reactor Vessel .4-2 o:\\WCAP15069.docim598 Revision 0 l

vii PREFACE This report has been technically reviewed and verified by: l Reviewer: Ed Terek s l l l o:\\WCAP15069. doc:1h071498 Revision 0

vii EXECUTIVE

SUMMARY

The purpose of this report is to determine the RTers values for the Vogtle Electric Generating Plant Unit I reactor vessel beltline based upon the results of the Surveillance Capsule V evaluation. The conclusion of this report is that all the belthne materials in the Vogtle Electric Generating Plant Unit I reactor vessel have RTrrs alues below the screening criteria of 270 F for plates, forgings or longitudinal welds and 300 F for v circumferential welds at EOL (36 EFPY) and life extension (54 EFPY). 1 1 0:\\WCAP15069. doc 1h091598 Revision 0

1-1 1 INTRODUCTION A Pressurized Thermal Shock (PTS) Event is an event or transient in pressurized water reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by significant pressure in the reactor vessel. A PTS concern arises if one of these transients acts on the beltline region of a reactor vessel where a reduced fracture resistance exists because of neutron irradiation. Such an event may produce a flaw or cause the propagation of a flaw postulated to exist near the inner wall surface, thereby potentially affecting the integrity of the vessel. The purpose of this report is to determine the RTrrs values for the Vogtle Electric Generating Plant Unit I reactor vessel using the results of the surveillance Capsule V evaluation. Section 2.0 discusses the PTS Rule and its requirements. Sections 3.0 provides the methodology for calculating RTyrs. Section 4.0 provides the reactor vessel beltline region material properties for the Vogtle Electric Generating Plant Unit I reactor vessel. The neutron fluence values used in this analysis are presented in Section 5.0 and were obtained from Section 6 ofWCAP-15067. The results of the RTvrs calculations are presented in Section 6.0. The conclusion and references for the PTS evaluation follow in Sections 7.0 and 8.0, respectively. l Introduction o:\\WCAP15069.docil>.091598 Revision 0

2-1 2 PRESSURIZED THERMAL SHOCK RULE The Nuclear Regulatory Commission (NRC) recently amended its regulations for light-water-cooled nuclear power plants to clarify several items related to the fracture toughness reqairements for reactor W pressure vessels, including pressurized thermal shock requirements. He revised PTS Rule,10 CFR Part 50.61, was published in the Federal Register on December 19,1995, with an effective date of January 18, 1996. This amendment to the PTS Rule makes Sree changes: 1. The rule incorporates in total, and therefore makes binding by mle, the method for determining the reference temperature, RTm, including treatment of the unirradiated RTm value, the margin term, and the explicit defmition of" credible" surveillance data, which is currently described in Regulatory Guide 1.99, Revision 2m 2. The rule is restructured to improve clarity, with the requirements section giving only the requirements for the value for the reference temperature for end oflife (EOL) fluence, RTvrs. 3. Thermal annealing is identified as a method for mitigating the effects of neutron irradiation, thereby reducing RTns. The PTS Rule requirements consist oithe following: For each pressurized water nuclear power reactor for which an operating license has been issued, the licensee shall have projected value: of RTns, accepted by the NRC, for each reactor vessel beltline material for the EOL fluence of the material. The assessment of RTvrs must use the calculation procedures given in the PTS Rule, and must specify the bases for the projected value of RTns for each beltline material. The report must specify the copper and nickel contents and the fluence values used in the calculation for each beltline material. This assessment must be updated whenever there is significant change in projected values of RTns or upon the request for a change in the expiration date for operation of the facility. Changes to RTns values are significant if either the previous value or the current value, or both values, exceed the screening criterion prior to the expiration of the operating license, including any renewal term, if applicable for the plant. The RTns screening criterion values for the beltline region are: 270 F for plates, forgings and axial weld materials, and 300 F for circumferential weld materials. 1 1 Pressurized Thermal Shock Rule o:\\WCAP15069. doc 1b-071498 Revision 0

3-1 3 METIIOD FOR CALCULATION OF RTers RTers must be calculated for each vessel beltline material using a fluence value, f, which is the EOL fluence for the material. Equation 1 must be used to calculate values of RTer for each weld and plate or forging in the reactor vessel beltling. RTer = RTercu> + M + ARTer (1)

Where, RTenu)

Reference Temperature for a reactor vessel material in the pre-service or unirradiated = condition M Margin to be added to account for uncertainties in the values of RTenu), copper and = nickel contents, fluence and calculational procedures. M is evaluated from Equation 2 M= Joy 2 + oi2 (2) o is the standard deviation for RTenu). u 0 F when RTenu)is a measured value. ou = 17 F when RTmnu)is a generic value. ou = c3 is the standard deviation for RTmr. For plates and forgings: c3 17 F when surveillance capsule data is not used. = 8.5 F when surveillance capsule data is used. os = For welds: c3 28 F when surveillance capsule data is not used. = c3 14 F when surveillance capsule data is used. = c not to exceed one halfof ARTmr a ARTmr is the mean value of the transition temperature shift, or change in ARTer, due to irradition, and must be calculated using Equation 3. ARTer = (CF)

• f( 28- "" 8')

(3) l Method For Calcualtion of RTrrs o:\\WCAP15069. doc:S -071498 Revision 0 l

3-2 CF ( F) is the chemistry factor, which is a function of copper and nickel content. CF is determined from Tables 1 and 2 of the PTS Rule (10 CFR 50.61). Surveillance data deemed credible must be used to determine a material-specific value of CF. A material-specific value of CF is determined in Equation 5. 2 F is the higher of the best estimate or calculated neutron fluence, in units of 10" n/cm (E > 1.o MeV), at the clad-base-metal interface on the inside surface of the vessel at the location where the material in question receives the highest fluence. The EOL fluence is used in calculating RTm. Equation 4 must be used for determining RTrrs using Equation 3 with EOL fluence values for determining A RTm R&rs = RTertv> + M + ARTm (4) To verify that RTer for each vessel beltline material is abounding value for the specific reactor vessel, licensees shall consider plant-specific information that could affect the level of embrittlement. This information includes but is not limited to the reactor vessel operating temperature and any related surveillance program results. Results from the plant-specific surveillance program must be integrated into the RTer estimate if the plant-specific surveillance data has been deemed credible. A material-specific value of CF is determined from Equation 5. U = g

• j(o2:-o 20tos A)]

(5) g(o 36-o 2oiosf> ) In Equation 5, "A,"is the measured value of ARTer and if"f,"is the fluence for each surveillance data point. If there is clear evidence that the copper and nickel content of the surveillance weld differs from the vessel weld, i.e., differs from the average for the weld wire heat number associated with the vessel weld and the surveillance weld, the measured values of RTer must be adjusted for differences in copper and nickel content by multiplying them by the ratio of the chemistry factor for the vessel material to that for the surveillance weld. Method For Calcualtion of RTm o:\\WCAP15069. doc:1b-071498 Revision 0

4-1 4 VERIFICATION OF PLANT SPECIFIC MATERIAL PROPERTIES Before performing the pressurized thermal shock evaluation, a review of the latest plant-specific material properties for the Vogtle Electric Generating Plant Unit I vessel was performed. The beltline region of a l reactor vessel, per the PTS Rule, is defined as "the region of the reactor vessel (shell material including welds, heat-affected zones and plates and forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience suflicient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage" Figure 1 identifies and indicates the location of all beltline region materials for the Vogtle Electric Generating Plant Unit I reactor vessel. The best estimate copper and nickel contents of the beltline materials were obtained from WCAP-13931, Rev lH1 and CE Report NPSD-1039, Rev. 2 'l The best estimate copper and nickel content is documented l in Table I herein. The average values were calculated using all of the available material chemistry information. Initial RTwr alues for Vogtle Electric Generating Plant Unit I reactor vessel beltline v material properties are also shown in Table 1. Verification of Plant Specific Material Properties o:\\WCAPl%69. doc:1b.091598 Revision 0

4-2 $w ? o 0 88805-2 90 101-124B k 88805-3 0 j 0 180 1 E j 101-124A 4 s Core B8805-1 2700 101-i24C 4 6 101-171 0 101-142A 30 88606-1 0 0 180 1 0 .J 'J00 # 101-1428 88606-3 0 101-142C 270 88606-2 l Figure 1: Identification and Location of Beltline Region Materials for the Vogtle Unit 1 Reactor Vessel Verification of Plant Specific Material Properties o:\\WCAP15069. doc 1b471498 Revision 0

4-3 Table 1 H Vogtle Unit 1 Reactor Vessel Beltline Unirradiated Material Properties *O Material Description Cu(%) Ni(%) Initial RTum* Intermediate Shell Plate B8805-1 0.083 0.597 0F Intermediate Shell Plate B8805-2 0.083 0.61 20'F Intermediate Shell Plate B8805-3 0.062 0.598 30 F Lower Shell Plate B8606-1 0.053 0.593 20 F Lower Shell Plate B8606-2 0.057 0.60 20 F Lower Shell Plate B8606-3 0.067 0.623 10 F Intermediate Shell Longitudinal Welds, 0.042N 0.102 -80 F 101-124A, B & C*) l Lower Shell Longitudinal Welds, 0.042M 0.102 -80 F 101-142A, B & C*) Circumferential Weld 101-171*) 0.042M 0.102 -80 F Surveillance Program Weld Metal 0.040 0.102 Notes: (a) The initial RTuor values for the plates and welds are based on measured data. (b) All welds, including the surveillance weld, were fabricated with weld wire heat number 83653, Linde 0091 Flux, Lot No. 3536. Per Regulatory Guide 1.99, Revision 2, " weight percent copper " and " weight percent nickel" are the best-estimate values for the material, which will normally be the mean of the measured values for a plate or forging or for weld samples made with the weld wire heat number that matches the critical vessel weld." (c) The copper weight percent of 0.042 was obtained using all available data for that heat of weld wire per reference 6. This value is more conservative than that documented (0.039) in Vogtle Electric Generating Plant's " Pressure and Temperature Limits Report" Verification of Plant Specific Material Properties o:\\WCAP15069. doc:1two91598 Revision 0

5-1 5 NEUTRON FLUENCE VALUES The calculated fast neutron fluence (E > 1.0 MeV) values at the inner surface of the Vogtle Electric Generating Plant Unit I reactor vessel for 36 and 54 EFPY are shown in Table 2. These values were projected using the results of the Capsule V radiation analysis. See Section 6.0 of the Capsule V analysis W report, WCAP-15067 TABLE 2 Fluence (E > 1.0 MeV) on the Pressure Vessel Clad / Base Interface for Vogtle Unit I at 36 (EOL) and 54 (Life Extension) EFPY Material Location 36 EFPY Fluence 54 EFPY Fluence Intermediate Shell Plate B8805-1 30 2.09 x 10 n/cm2 3.10 x 10 n/cm2 Intermediate Shell Plate B8805-2 30 2.09 x 10 n/cm2 3.10 x 10 n/cm2 Intermediate Shell Plate B8805-3 30 2.09 x 10 n/cm2 3.10 x 10 n/cm2 Lower Shell Plate B8806-1 30 2.09 x 10 n/cm2 3.10 x 10 n/cm 2 Lower Shell Plate B8806-2 30 2.09 x 10 n/cm2 3.10 x 10 n/cm2 Lower Shell Plate B8806-3 30 2.09 x 10 n/cm 2 3.10 x 10n/cm 2 l I Intermediate Shell Longitudinal go 1,17 x 10 n/cm 3,74x19 n/cm 19 2 19 2 Weld Seam 101-124A (O Azimuth) Intermediate Shell Longitudinal 30' 2.09 x 10 n/cm2 3.10 x 10 n/cm2 Weld Seam 101-124B & C (120 & 240 Azimuth) Intermediate to Lower Shell 30 2.09 x 10 n/cm2 3.10 x 10 n/cm2 Cirumferential Weld Seam 101-171 Lower Shell Longitudinal 30* 2.09 x 10 n/cm2 3.10 x 10 n/cm 2 Weld Seams 101-142A & C (60 & 300 Azimuth) Lower Shell Longitudinal 0 1.17 x 10 n/cm2 1.74 x 10 n/cm2 Weld Seam 101-142B (l80* Azimuth) Neutron Fluence Values o:\\WCAP15069. doc:1b&#1598 Revision 0 w

6-1 6 DETERMINATION OF RT s VALUES FOR ALL BELTLINE PT REGION MATERIALS Using the prescribed PTS Rule methodology, RTm values were generated for all beltline region materials of the Vogtle Electric Generating Plant Unit I reactor vessel for fluence values at the EOL (36 EFPY) and life extension (54 EFPY). Each plant shall assess the RTm values based on plant-specific surveillance capsule data. For Vogtle Electric Generating Plant Unit 1, the related surveillance program results have been included in this PTS evaluation. As presented in Table 3, chemistry factor values for Vogtle Electric Generating Plant Unit I based on average copper and nickel weight percent were calculated using Tables I and 2 from 10 CFR 50.6110 Additionally, chemistry factor values based on credible surveillance capsule data are calculated in Table 4. Tables 5 and 6 contain the RTm calculations for all beltline region materials at EOL (36 EFPY) and life extension (54 EFPY). I l 1 Determination of RTm Values For All Beltline Region Materials o:\\WCAP15069. doc 1M)91598 Revision 0

6-2 TABLE 3 Interpolation of Chemistry Factors Using Tables 1 and 2 of 10 CFR Part 50.61 l Material Ni,wt % Chemistry Factor, F Intermediate Shell Plate B8805-1 0.597 53.1 F Given Cu wt% = 0.083 Intermediate Shell Plate B8805-2 0.61 53.1 F Given Cu wt% = 0.083 Intermediate Shell Plate B8805-3 0.598 38.4 F Given Cu wt% = 0.062 Lower Shell Plate B8606-1 0.593 32.8 F Given Cu wt% = 0.053 Lower Shell Plate B8606-2 0.60 35.2 F Given Cu wt% = 0.057 Lower Shell Plate B8606-3 0.623 41.9 F Given Cu wt% = 0.067 Intermediate Shell Longitudinal Welds. 101-124A. B & C 0.102 34.5 F Given Cu wt% = 0.042 Lower Shell Longitudinal Welds.101-142A. B & C 0.102 34.5 F Given Cu wt% = 0.042 Circumferential Weld 101-171 0.102 34.5 F Given Cu wt% = 0.042 Surveillance Proaram Weld Metal 0.102 33.7 F Given Cu wt% = 0.040 Determination of RTm Values For All Beltline Region Materials o:\\WCAP15069. doc 1MDIS98 Revision 0

6-3 TABLE 4 Calculation of Chemistry Factors using Surveillance Capsule Data Per Regulatory Guide 1.99, Revision 2, Position 2.1 Material Capsule Capsule f* FF ARTm FF*ARTer FF W 2 Intermediate Shell U 0.3691 0.725 13.6 9.9 0.526 Plate B8805-3 Y 1.276 1.068 31.9 34.1 1.141 (Longitudinal) V 2.178 1.211 42.7 51.7 1.467 " Intermediate Shell U 0.3691 0.725 ON 0.0 0.526 Plate B8805-3 Y 1.276 1.068 15.2 16.2 1.141 1 (Transverse) V 2.178 1.211 33.8 40.9 1.467 SUM 152.8 6.268 2 CFannos.3 = Z(FF

• RTer) + Z(FF ) = (152.8) + (6.268) = 24.4 F Surveillance Weld U

0.3691 0.725 25.5(d) 18.5 0.526 Metal Y 1.?.76 1.068 7.9 ) 8.4 1.141 Cd 40) 0.0 1.467 V 2.178 1.211 SUM 26.9 3.134 2 g CFw.ia = Z(FF

• RTer) + I(FF ) = (26.9) + (3.134) = 8.6 F Notes:

(a) f = Calculated fluence from capsule V dosimetry analysis results(5) (x 10 n/cm, E > 1.0 MeV). 2 (b) FF = fluence factor = f( 28' " 8') (c) ARTervalues are the measurec 30 ft-lb shift values. (d) The surveillance weld metal ARTmp values have been adjusted by a ratio factor of 1.02. (e) Actual values for ARTmn are -9.58 (Plate) and -1.34 (Weld). This physically should not occur, therefore for conservatism a value of zero will be used for this calculation. Determmation of RTns Values For All Beltline Region Materials o:\\WCAP15069. doc:1b.091598 Revision 0

6-4 TABLES RTm Calculation for Vogtle Unit 1 Beltline Region Materials at EOL (36 EFPY) Material RG 1.99 R2 CF FF IRTuorm>"' ARTm") M argin RTm" Method (*F) Intermediate Shell Plate B8805-1 Position 1.1 53.1 1.20 0 63.7 34 98 Intennediate Shell Plate B8805-2 Position L1 53.1 1.20 20 63.7 34 118 Intermediate Shell Plate B8805-3 Position 1.1 38.4 1.20 30 46.1 34 110 Position 2.1 24.4 1.20 30 29.4 17 76 Lower Shell Plate B8606-1 Position 1.1 32.8 1.20 20 39.4 34 93 Lower Shell Plate B8606-2 Position 1.1 35.2 1.20 20 42.2 34 96 Lower Shell Plate B8606-3 Position 1.1 41.9 1.20 10 50.3 34 94 ~ Inter. Shell Longitudinal Weld ~ Seam 101-124A(0* Azimeh) Position 2.1 8.6 1.04 -80 8.9 8.9 -62 Inter. Shell Long. Weld Seams Position 1.1 34.5 1.20 -80 41.4 41.4 3 101-124B,C (120*,240* Azimuth) Position 2.1 8.6 1.20 -80 10.3 10.3 -59 Intermediate to Lower Shell Position 1.1 34.5 1.20 -80 41.4 41.4 3 Girth Weld Seam 101-171 Position 2.1 8.6 1.20 -80 10.3 10.3 -59 Lower Shell Long. Weld Seams Position 1.1 34.5 1.20 -80 41.4 41.4 3 101-142A,C (60,300* Azimuth) Position 2.1 8.6 1.20 -80 10.3 10.3 -59 Lower Shell Long. Weld Seam Position 1.1 34.5 1.04 -80 35.9 35.9 -8 101-142B (180* Azimuth) Position 2.1 8.6 1.04 -80 8.9 8.9 -62 Notes: (a) Initial RTnor values are measured values. (b) RTm = Initial RTwar(u) + ARTm + Margin ( F) (c) ARTrrs = CF

• FF Determination of RTm Values For All Beltline Region Ma:edals o:\\WCAP15069. doc:11>O91598 Revision 0

6-5 l TABLE 6 RTm Calculation for Vogtle Unit 1 Beltline Region Materials at Life Extension (54 EFPY) Material RG 1.99 R2 CF FF 1RTm>mg ARTm") Margin RTm" Method ('F) Intermediate Shell Plate B88051 Position 1.1 53.1 1.30 0 69.0 34 103 Intermediate Shell Plate B8805-2 Position 1.1 53.1 1.30 20 69.0 34 123 Intermediate Shell Plate B8805-3 Position 1.1 38.4 1.30 30 49.9 34 114 Position 2.1 24.4 1.30 30 31.7 17 79 Lower Shell Plate B8606-1 Position 1.1 32.8 1.30 20 42.6 34 97 Lower Shell Plate B8606-2 Position 1.1 35.2 1.30 20 45.8 34 100 Lower Shell Plate B8606-3 Position 1.1 41.9 1.30 10 54.5 34 99 Position 1.1 34.5 1.15 -80 39.7 39.7 0 Inter. Shell Longitudinal Weld Seam 101-124A(0" Azimuth) position 2.1 8.6 1.15 -80 9.9 9.9 -60 g Inter. Shell Long. Weld Seams Position 1.1 34.5 1.30 -80 44.9 44.9 10 101-124B,C (120*, 240 Azimuth) Position 2.1 8.6 1.30 -80 11.2 11.2 -58 Intermediate to Lower Shell Position 1.1 34.5 1.30 -80 44.9 44.9 10 Girth Weld Seam 101-171 Position 2.1 8.6 1.30 -80 11.2 11.2 -58 Lower Shell Long. Weld Seams Position 1.1 34.5 1.30 -80 44.9 44.9 10 101 142A,C (60,300* Azimuth) Position 2.1 8.6 1.30 -80 11.2 11.2 -58 j Lower Shell Long. Weld Seam Position 1.1 34.5 1.15 -80 39.7 39.7 0 101-142B (180' Azimuth) Position 2.1 8.6 1.15 -80 9.9 9.9 -60 Notes: (a) Initial RTsm values are measured values. RTm = Initial RT mm) + ARTm + Margin ( F) (b) 3 (c) ARTm = CF

• FF Determination of RTm Values For All Beltline Region Materials o:\\WCAP15069. doc:ll>.091598 Revision 0

7-1 7 CONCLUSIONS 1 As shown in Tables 5 and 6, all of the beltline region materials in the Vogtle Electric Generating Plant Unit I reactor vessel have EOL (36 EFPY) RTers and Life Extension (54 EFPY) RTns values below the screening criteria values of 270 F for plates, forgings and longitudinal welds and 300 F for circumferential welds. ~ Conclusions o:\\WCAP15069. doc 1M)91598 Revision 0

8-1 8 REFERENCES l 1 10 CFR Part 50.61, " Fracture Toughness Requirements For Protection Against Presuurized Thermal Shock Events", Federal Register, Volume 60, No. 243, dated December 19,1995. 2 Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials", U.S. Nuclear Regulatory Commission, May,1988. 3 WCAP-110ll, " Georgia Power Company Alvin W. Vogtle Unit No. I Reactor Vessel Radiation Surveillance Program", L.R. Singer, February 1986. 4 WCAP-13931, Rev.1, " Analysis of Capsule Y from the Georgia Power Company Vogtle Unit 1 Reactor Vessel Radiation Surveillance Program", M.J. Malone, et al., August 1995. 5 WCAP-15067, " Analysis of Capsule V from the Georgia Power Vogtle Electric Generating Plant Unit 1 Reactor Vessel Radiation Surveillance Program", T. J. Laubham, et al., September 1998. 6 CE NPSD-1039, Rev. 2, "Best Estimate Copper and Nickel Values m CE Fabricated Reactor Vessel Welds, Appendix A, CE Reactor Vessel Weld Properties Database, Volume 1," CEOG Task 902, June 1997. References o;\\WCAP15069. doc:1b-o91598 Revision 0}}