ML20064M452

WCAP-13948, Evaluation of Pressurized Thermal Shock for McGuire Unit 1

ML20064M452
Person / Time
Site:	McGuire
Issue date:	02/28/1994
From:	Meyer T, Peter P WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20064M423	List: ML20064M421 ML20064M452
References
WCAP-13948, NUDOCS 9403280224
Download: ML20064M452 (26)

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WESTINGIIOUSE CLASS 3 (Non Proprietary)

WCAP-13948 l

EVALUATION OF PRESSURIZED TilERSfAL SIIOCK FOR MCGUIRE UNIT 1 P. A. Peter i

February 1994 Work Performed Under Shop Order DXBP-108A Prepared by Westinghouse Electric Corporation for Duke Power Company P

Approved by:. _ n i I L T. A. Meyer, Manager Structural Reliability & Plant Life Optimization WESTINGHOUSE ELECTRIC CORPORATION Nuclear Technology Division P.O. Box 355 Pittsburgh. Pennsylvania 15230-0355 i

@ 1994 Westinghouse Electric Corporation All Rights Reserved l

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.- _ _.-------_.---_-----------_-----.__-_-__a

l PREFACE 4

This repon has been technically reviewed and verifted by:

J. M. Chicots h. M7M v -

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TABLE OF CONTENTS LIST OF TABLES . . . . . . . . . . ......... ........... ... .. . . ..... . ill .

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LIST OF FIGURSS . . . . . . . . . . . ...... . .. .. .. ..... ........ . . . iii

1.0 INTRODUCTION

. . . . . . . . . . . . . . . ... .......... .............. ... I 2.0 PRESSURIZED THERMAL SHOCK , . .... .... ... .. . .... .... 2 3.0 METHOD FOR CALCULATION OF RTm . . . . . . ...... . 4 4.0 VERIFICATION OF PLANT-SPECIFIC MATERIAL PROPERTIES ............... 5 5.0 NEUTRON FLUENCE VALUES .. .. ..... .. ....... ...... .. 12 6.0 DETERMINATION OF RTm VALUES FOR ALL BELTLINE REGION MATERIALS ..... ..... .. . ... ...... . ...... .... .. . . 13

7.0 CONCLUSION

S . . . . . . .. ....... . .. ... . .. .. ... . 17

8.0 REFERENCES

... .. .. .. ...... ..... ..... . .. ...... ........ 18 i

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LIST OF TABLES I

TABLE 1 CALCU1ATION OF AVERAGE CU AND NI WEIGIIT % FOR MCGUIRE UNIT A . ..... . . . . . ... ... . . . 7 TABLE 2 MCGUIRE UNIT 1 REACTOR VESSEL BELTLINE REGION MATERIAL PROPERTIES . ... ... . . .. . . . ... . 11 TABLE 3 NEUTRON EXPOSURE PROJECTIONS AT KEY LOCATIONS ON TIIE MCGUIRE UNIT I PRESSURE VESSEL CLAD / BASE METAL INTERFACE FOR 7.241 AND 32 EFPY . . . . . . . 12 TABLE 4 CALCULATION OF CHEMISTRY FACTORS USING i4CGUIRE UNTT 1 SURVEILLANCE CAPSULE DATA . . . . . ... . 14 TABLE 5 RTyn VALUES FOR MCGUiRE UNIT 1 FOR 7.241 EFPY . ... 15 1

TABLE 6 RTen VALUES FOR MCGUIRE UNIT 1 FOR 32 EFPY . . . 16 LIST OF FIGURES FIGURE 1 IDENTIFICATION AND LOCATION OF BELTLINE REGION MATERIALS FOR T71E MCGUIRE UNIT 1 REACTOR VESSEL . 6 FIGURE 2 RTrn VERSUS Fi UENCE CURVES FOR MCGUIRE UNIT 1 LIMITING MATERIAL - LOWER SIIELL LONGITUDINAL WELDS 3-442A&C 17 iii i

1.0 INTRODUCTION

A limiting condition on reactor vessel integrity known as Pressurized Thermal Shock (PTS) may occur during a severe system transient such as a Loss-Of-Coolant Accident (LOCA) or a steam line break.

Such transients may challenge the integrity of a reactor vessel under the following conditions:

revere overcooling of the inside surface of the vessel wall followed by b8

t. i.ressudzation:

significant degradation of vessel material toughness caused by radiation embrittlement; and the presence of a critical-size defect ir 'be vessel wall.

In 1985 the Nuclear Regulatory Commission (NRC) issued a formal ruling on PTS. It established screening criteria on pressurized water reactor (PWR) vessel embrittlement as measured by the nil-ductility reference temperature, termed RTpn i ". RTen screening values were set for beltline axial welds, forgings or plates and for beltline circumferential weld seams for the end-of-license plant operation. The screening enteria were determined using conservative fracture mechanics analysis techniques. All PWR vessels in the United States have been required to evaluate vessel embrittlement in accordance with the criteria through end oflicense. The NRC recently amended its regulations for light water nuclear power o *. to change the procedure for calculating radiation embrittlement. The revised I'TS Rule was pub > ,.:u. in the Federal Register, May 15, 1991 with an effective date of June 14,1991m This amendment makes the procedure for calculating RTen values consistent with the methods given in Regulatory Guide 1.99, Revision 2N The purpose of this report is to determine the RTen values for the McGuire Unit I reactor vessel to address the revised FTS Rule. Section 2 discusses the Rule and its requirements. Section 3 provides the methodology for calculating RT,n. Section 4 provides the reactor vessel beltline region material properties for the McGuire Unit I reactor vessel. The neutron fluence values used in this analysis are presented in Section 5. The results of the RTen calculations are presented in Section 6. The conclusions and refere. as for the PTS evaluation follow in Sections 7 and 8, respectively.

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d 2.0 PRESSURT7Fn THERMAL SIlOCK The FTS Rule requires that the FTS submittal be updated whenever there are changes in core loadings, surveillanc: measurements or other information that indicates a significant change in projected RTru values. The Rule outlines regulations to address the potential for PTS events on pressurized water reactor vessels in nuclear power plants that are operated with a license from the United States Nuclear Regulatory Commission (USNRC). PTS events have been shown from operating experience to be transients that result in a rapid and severe cooldown in the primary system coincident with a high or increasing primary system pressure. The FTS 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 l

irradiation. Such an event may result in the propagation of flaws postulated to exist near the inner wall surface, thereby potentially affecting the integrity of the vessel.

The Rule establishes the following requirements for all domestic, operating PWRs:

All plants must submit projected values of RTrn for reactor vessel beltline materials by giving values for time of submittal, the expiration date of the operating license, and the projected expiration date if a change in the operating license or renewal has been requested. This assessment must be submitted within six months after the effective date of this Rule if the value of RTru for any material is projected to exceed the screening criteria. Otherwise,it must be submitted with the next update of the pressure-temperature limits, or the next reactor vessel surveillance capsule report, or within 5 years from the effective date of this Rule change, whichever comes first. These values must be calculated based on the methodology specified in this rule. The submittal must include the following:

1) the bases for the projection (including my assumptions ret rding core loading patterns), and 1

2) copper and nickel content and fluence values u ;cd in the calculations for each beltline material. (If these values dif.er from those previously submitted to the NRC, justification must bc provided.)

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The RTen (mea ure of fracture resistance) screening criteria for the reactor vessel beltline region is:

270 'F for plates, forgings, axial welds; and 300 'F for circumferential weld materials.

The following equations must be used to calculate the RT,n values for each weld, plate or forging in the reactor vessel beltline:

Equation 1: RTrn = I + M + ART,3 ,

Equation 2: ARTrn = CF

• f nusaioin r3 All values of RTrn must be verified to be bounding values for the specific reactor vessel. In doing this each plant should consider plant-specific information that could affect the level of embrittletuent.

Plant-specific PTS safety analyses are required before a plant is within 3 years of reaching the screening criteria. including analyses of alternatives to minimize the PTS concern.

NRC approval for operation beyond the screening criteria is required. '

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i j 3.0 METHOD FOR CALCULATION OF RTp33 l

i In the FTS Rule, the NRC Staff has selected a conservative and uniform method for determming plant-specific values of RTen at a given time. For the purpose of comparison with the screening criteria, the value of RTrn for the reactor vessel must be calculated for each weld and plate or forging in the beltline region as follows.

RTen = 1 + M + ART,3, where ARTrn = CF

• FF I= Initial reference temperature (RTxor) in F of the unirradiated material M= hfargin to be added to cover uncertainties in the values of initial RTxo7, copper and nickel contents, fluence and calculational procedures in 'F.

M = 66 *F for welds and 48 'F for base metalif generic values ofI are used.

M = $6 F for welds and 34 'F for base metal if measured values of I are used.

l FF = fluence factor = f a23.aio win, where i

f= Neutron fluence (E>l.0 MeV at the clad / base metal interface), divided by 10 '

2 I n/cm 2 l

CF = Chemistry factor in F from the tablef for welds and base metals (plates and forgings). If plant-specific surveillance data has been deemed credible per Regulatory Guide L99, Revision 2 and is significant, it may be considered in the calculation of the chemistry factor.

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l 4.0 VERIFICATION OF PLANT-SPECIFIC MATERIAL PROPERTIES i

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Before performing the pressurized thermal shock evaluation, a review of the latest plant-specific material properties for the McGuire Unit I vessel was performed. The beltline region is defined by the FTS Rulem to be "the region of the reactor vessel (shell material including welds, heat-affected j zones and plates or forgings) that directly surrounds the effective height of the active core and adjacent '

4 regions of the reactor vessel that are predicted to experience sufficient neutron irradiation damage to j be considered in the selection of the most limiting material with regard to radiation damage." Figure 1 identifies and indicates the location of all beltUne region materials for the McGuire Unit I reactor vessel.

i Material property values were obtained from material test certifications from the original fabrication as

}.

well as the additional material chemistry tests performed as pan of the surveillance capsule testing I

) programm. Pertinent chemistry test results were also obtained from other plants with similar weld l materials. The average copper and nickel values were calculated for each of the beltline region materials using all of the available material chemistry information as shown in Table 1. A summary

of the pertinent chemical and mechanical propenies of the beltline region plates and weld materials of ,

j the McGuire Unit I reactor vessel are given in Table 2. All of the initial RTm values (1) are also l

l presented in Table 2.

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FIGURE 1. IDENIIFICATION AND LOCATION OF BELTLINE REGION MATERIALS FOR THE MCGUIRE U.=TT 1 REACTOR VESSELi' 1 l

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- -- ---- - n

TABLE 1 CALCULATION OF AVERAGE CU AND NI WEIGHT % FOR MCGUIRE UNIT 1 Intermediate Shell Longitudinal Weld Seams,2-442A, B & C (Ht. 20291 & 12008, Linde 1992, Flux Lot No. 3854)

Reference wt. % wt. %

Cu Ni 5 0.21 0.88 6 0.20 0.91 32 0.195 0.87 32 0.191 0.848 32 0.193 0.863 Average 0.20 0.87 Intermediate to Lower Shell Circumferential Weld. 9-442 (Ht. 83640, Linde 0091, Flux Lot No. 3490)

Reference wt. % wt. % 4 Cu Ni 7 0.050 --

8 0.050 0.120 Average l 0.050 0.120 7

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TABLE I (CONT'D.)

Lower Shell lennitudinal Weld Seams. 3-442A. B & C (lit. 21935 & 12008, Linde 1092, Flux Lot No. 3889)

Reference wt. % wt. %

Cu Ni 9 0.220 --  ;

1 10 0.200 ---

11 0.22 0.83 12 0.23 0.90  ;

1 0.21 0.76 {

l 0.22 0.90 l l

13 0.219 0.86  :

0.212 0.88 0.213 0.90 14 0.225 0.875 0.213 0.856 1 0.225 0.877 l Average 0.22 0.86 l -

Lower Shell Lonnitudinal Weld Seams. 3-442A (Root Weld) *

(lit. 305424, Linde 1092, Flux Lot No. 3889)

Reference wt. % wt. %

Cu Ni 15 0.300 Of>40 16 0.260 0.620 33 0.230 0.637 Average 0.263 0.632 The lower shell longitudical welds also contained a different weld wire heat in the double U root area of the weld. Since the root weld chemistry is not more limiting than the above weld data, it was not utilized in the pressurized thermal shock evaluations. l l

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TABLE 1 (CONT'D.)

Lower Shell Longitudinal Weld Seams. 3-442H & C (Roct Weld)* j (lit. 21935, Linde 1092, Flux Lot No. 3889)

Reference wt. % wt. %

Cu Ni 17 0.200 ---

12 0.21 0.68 0.71 Average 0.21 0.70 Intermediate Shell Plate. B50121 (lit. C4387 2)

Reference wt. % wt. %

Cu Ni 18.19 0.13 0.60 5 0 087 --

20 --

0.58 32 0.117 0.643 Average 0.11 0.61 Intermediate Shell Plate. B5012-2 (Ilt. C4417-3)

Reference wt. % wt. %

Cu Ni 18,21 0.14 0.62 22 ---

0.60 Average 0.14 0.61 1

The lower shell longitudinal welds also contained a different weld wire heat in the double U root area of the weld. Since the root weld chemistry is not more limiting than the above weld data, it was not utilized in the pressurized thermal shock evaluations.

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TABLE I (CONT'D.)

Intermediate Shell Plate, H5012 3 (Ilt. C4377-2) 1 Reference wt. % wt. % I Cu Ni 18, 23 0.11 0.66 20 --

0.65 Average 0.11 0.66 Lower Shell Plate, B5013-1 (IIt. C43151)

Reference wt. % wt. %

Cu Ni 18,24 0.14 0.56 25 ---

0.59 Average 0.14 0.58 Lower Shell Plate. B5013-2 (IIt. C4374 2)

Reference wt. % wt. %

Cu Ni 18,26 0.10 0.52 27 --

0.50 Average 0.10 0.51 Lower Shell Plate. B5013-3 (III. C43712)

Reference wt. % wt. %

Cu Ni 18,28 0.10 0.55 29 ---

0.54 Average 0.10 0.55

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_- -_ _ _ _ _ _ _ _ _ . _ - - . - _ - . _ _ _ - . ________._a

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TABLE 2 MCGUIRE UNIT 1 REACTOR VESSEL BELTLINE REGION MATERIAL PROPERTIES i

(')

Material Description Cu (%)

• Ni (%)

• I (*F)!'"

Intermediate Shell Plate B50121 0.11 0.59 34 Intermediate Shell Plate, B5012-2 0.14 0.61 0 Intermediate Shell Plate, B5012-3 0.11 0.66 -13 lewer Shell Plate, B50131 0.14 0.58 0 Lower Shell Plate, B5013-2 0.10 0.51 30 Lower Shell Plate, B5013-3 0.10 0.55 15 Inter. Shell Longitudinal Welds,2-442A, B & C 0.21 0.90 -50 Lower Shell Longitudinal Welds,3-442A, B & C 0.22 0.86 -56 Circumferential Weld, 9-442 0.05 0.12 -70 (a) Initial RT, values were estimated per U.S. NRC Standard Review Plan. The initial RT, values fcr the plates and welds are measured values, except for the lower shell longitudinal welds (which is a generic value).

Average values of copper and nickel as indicated in Table 1.

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5.0 NEUTRON FLUENCE VALUES The calculated fast neutron fluence (E>1.0 MeV) at the inner surface of the McGuire Unit I reactor vessel is shown in Table 3. These values were projected using the results of the Capsule V radiation surveillance prograni'l. The RTers calculations were performed using the peak fluence value, which occurs at the 45 azimuth for all materials in the McGuire Unit I reactor vessel beltline region, except for the intermediate and lower shell longitudinal welds, which are located at the 0* and 30 azimuths, and do not experience the peak fluence.  :

TABLE 3 l

NEUTRON EXPOSURE PROJECTIONS

• AT KEY LOCATIONS ON TIIE MCGUIRE UNIT 1 PRESSURE VESSEL CLAD / BASE METAL INTERFACE FOR 7.241 AND 32 EFPYt'l  !

I EFPY 0* 15* 30 45 5

7.241 0.3015 0.4457 0.3301 0.4562 i

32 1.332 1.%9 1.459 2.016  !

2

• Fluence in 10 ' n/cm (E>1.0 MeV) l i

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6.0 DETERMINATION OF RTrn VALUES FOR ALL BELTLINE RFBION MATERLALS Using the prescribed PTS Rule methodology, RTen values were generated for all beltline region materials of the McGuire Unit i reactor vessel for fluence values at the present time (7.241 EFPY per Capsule V analysis) and end oflicense (32 EFPY). The FTS Rule requires that each plant assess the l RTrn values based on plant specific surveillance capsule data whenever; i

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Plant specific surveillance data has been deemed credible as defined in Regulatory Guide l 1.99, Revision 2, and t

i RTrn values change significantly. (Changes to RTen values are considered significant if the value determined with RTpn equations (1) and (2), or that using capsule data, or both, exceed the screening criteria prior to the expiration of the operating license, including any renewed term, if applicable, for the plant.)

Although the RTen value changes are not significant for McGuire Unit 1, plant specific surveillance capsule data for intermediate shell plate B50121 and intermediate shell longitudinal welds is provided >

because of the following reasons:

1) There have been three capsules removed from the reactor vessel, and the data is deemed credible per Regulatory Guide 1.99. Revision 2.

2) The surveillance capsule materials are representative of the actual vessel plates and intermediate shell longitudinal weld materials.

The chemistry factors for intermediate shell plate B5012-1 and intermediate shell longitudinal welds

( were calcula:ed using the suneillance capsule data as shown in Table 4. The chemistry factors were  ;

l also calculated using Tables 1 and 2 from 10 CFR 50.61W. Tables 5 and 6 provide a summary of the j RTru values for all beltline region materials for 7.241 EFPY and 32 EFPY, respectively, using the l PTS Rule. I i

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TABLE 4 CALCULATION OF CHEMISTRY FACTORS USING MCGUIRE UNIT 1 SURVEILLANCE CAPSULE DATAW Material Capsule Fluence FF ART3m FF* ART3m FF2 Inter. Shell Flate, U 4.719 x 10 O.790 45 35.550 0.624 B5012-1 (Longitudinal) X 1.4091 x 10 l.095 45 49.275 1.199 V 2.1858 x 10 ' l.212 85 103.020 1.469 Inter. Shell Plate, U 4.719 x 10 O.790 50 39.500 0.624 B5012-1 X 1.4091 x 10 l.095 65 71.175 1.199 (Trmerse)

V 2.1858 x 10 l.212 85 103.020 1.469 Sum: 401.54 6.584 Chemistry Factor = 401.54 + 6.584 = 61.0 = 61 Intermediate Shell U 4.719 x 10'8 0.790 160 126.400 0.624 Iengitudinal Welds, 2442A, B & C X 1.4091x 102' l.095 165 180.675 1.199 V 2.1858 x 102' l.212 175 212.100 1.469 Sum: 519.175 3.292 Chemistry Factor = 519.175 + 3.292 = 157.7 = 158 i

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TABLE 5 j RTpn VALUES FOR MCGUIRE UNIT 1 FOR 7.241 EFPY i

l Material CF (*F) FI? I( F) M ( F) RTen ( F) i t

Inter. Shell Plate, B5012-1 74.2 0.7815 34 34 125.99 Using S/C data" 61 0.7815 34 34 115.67 l

Inter. Shell Plate, B5012-2 100.25 0.7815 0 34 112.35 j Inter. Shell Plate, B5012-3 74.9 0.7815 -13 34 79.54 i

j Lower Shell Plate, B50131 99.1 0.7815 0 34 111.45 IAwer Shell Plate, B5013-2 65.0 0.7815 30 34 114.80 i

j Ixwer Shell Plate, B5013 3 65.0 0.7815 15 34 99.80 l Inter. Long. Weld, 2-442A 204.2 0.6716 -50 56 143.13 l Using S/C data" 158 0.6716 -50 56 112.11

Inter. Ieng. Weld,2-442B&C 204.2 0.6951 -50 56 147.94 l Using S/C data" 158 0.6951 -50 56 115.83 i Lower Long. Weld,3-442B 209.6 0.6716 -56 66 150.76

] Imwer Long. Weld,3-442A&C 209.6 0.6951 -56 66 155.69

$ Circumferential Weld 39.8 0.7815 -70 56 17.10 l

)

• FF (Fluence factor) based upon peak inner surface neutron fluence of 4.562 x 10" n/cm 2 i (E>l.0 MeV)W at the 45* azimuthal angle, except for the longitudinal weld seams (which are j located at the 0 and 30 azimuthal angles with fluences of 3.015 x 10' n/cm2 and 3.301 x 10' n/cm 2, respectively).

{

i Numbers were calculated using a chendstry factor (CF) based on sulveillance capsule data.

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TABLE 6 RTrn VALUES FOR MCGUIRE UNIT 1 FOR 32 EFPY Material CF (*F) FF" I( F) M ( F) RTen ( F)

Inter. Shell Plate, B50121 74.2 1.1912 34 34 156.39 Using S/C data" 61 1.1912 34 34 140.66 Inter. Shell Plate, B5012-2 100.25 1.1912 0 34 153.42 Imer. Shell Plate. B5012-3 74.9 1.1912 -13 34 110.22 14wer Shell Plate, B5013-1 99.1 1.1912 0 34 152.05 Imwer Shell Plate, B5013-2 65.0 1.1912 30 34 141.43 Lower Shell Plate. B5013 3 65,0 1.1912 15 34 126.43 Inter. Long. Weld, 2-442A 204.2 1.0797 -50 56 226.48 Using S/C data" 158 1.0797 -50 56 176.60 Inter. Long. Weld, 2-442B&C 204.2 1.1047 -50 56 231.58 Using S/C data" 158 1.1047 -50 56 180.54 Iower Long. Weld. 3-442B 209.6 1.0797 -56 66 236.31 Ixwer Long. Weld,3-442A&C 9.6 1.1047 -56 66 241.54 Circumferential Weld 39.8 1.1912 -70 56 33.41 FF (Fluence factor) based upon peak inner surface neutron fluence of 4.562 x 10 n/cm2 (E>l.0 MeV)"1 at the 45' azimuthal angle, except for the longitudinal weld seams (which are located at the 0 and 30' azimuthal angles with fluences of 3.015 x 10' n/cm' and 3.301 x 10" n/cm 2, respectively).

Numbers were calculated using a chemistry factor (CF) based on surveillance capsule data.

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7.0 CONCLUSION

S As shown in Tables 5 and 6, all RTenvalues remain below the NRC screening values for ITS using fluence values for the present time (7.241 EFPY) and projected fluence values for the end of license (32 EFPY). A plot of the RTrn values versus fluence shown in Figure 2 illustrates the available margin for the most limiting material in the McGuire Unit I reactor vessel beltline region Lower Shell Longitudinal Welds,3-442A & C.

350 l

300 -

SCREENING CRITERIA 250 -

p 200 -

n.  ;

p150 -

l l

100 -

50 -

• 7.241 EFPY A 32.0 EFPY 0

1 E + 18 2E + 18 3E+18 SE + 18 1 E + 19 2E + 19 3E + 19 SE+19 1E + 20 2

FLUENCE (neutrons /cm )

LOWER SHELL LONG. WELDS 3-442A & C FIGURE 2. RTy n VERSUS FLUENCE CURVES FOR MCGUIRE UNIT 1 LIMITING MATERIAL - LOWER SHELL LONGITUDINAL WELDS,3-442A&C 17

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8.0 REFERENCES

[1] 10CFR Part 50, " Analysis of Potential Pressurized Thermal Shock Events," July 23,1985.

[2] 10CFR Part 50.61, " Fracture Toughness Requirements for Protection Against IYessurized

'Ibermal Shock Events," May 15,1991. (I FS Rule)

[3] Regulatory Guide 1.99, Revision 2 " Radiation Embrittlement of Reactor Vessel Materials,"

U.S. Nuclear Regulatory Commission, May 1988.

[4] WCAP-13949," Analysis of Capsule V Specimens and Dosimeters and Analysis of Capsule Z Dosimeters from the Duke Power Company McGuire Unit 1 Reactor Vessel Radiation Surveillance Program", E. Terek, January 1994.

[5] WCAP-9195, " Duke Power Company William B. McGuire Unit No.1 Reactor Vessel l Radiation Surveillance Program", J. A. Davidson and S. E. Yanichko, November 1977.

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[6] WCAP-10786," Analysis of Capsule U from the Duke Power Company McGuire Unit 1 Reactor Vessel Radiation Stuveillance Program", S. E. Yanichke, et al., February 1985, i

[7] Combustion Engineering, Inc., Metallurgical Research & Developmen', " Chemical Analysis of Wire-Flux Test Coupon", Job Number D32255, 815-72.

[8] WCAP-8819. " Central Nuclear de Almaraz, Almaraz Unit No.1 Reactor Vessel Radiation Surveillance Program", R. A. Smith, et al., December 1976.

[9] Combustion Engineering, Inc., Welding Matenal Certification and Release for Section III,

" Chemical Analysis of Test Weld Sample", August 12,1%9.

[10] Combustion Engineering. Inc., Metallurgical Research & Development, " Chemical Analysis of Wire-Flux Test Coupon", Job Number X-32255,10-14-69.

[11] WCAP-8783, " Pacific Gas and Electric Company Diablo Canyon Unit No. 2 Reactor Vessel Radiation Surveillance Program", J. A. Davidson and S. E. Yanichko, December 1976.

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[12] WCAP-10472, " Evaluation of Diablo Canyon Units 1 and 2 Reactor Vessel Beltline Weld Chemistry", S. E. Yanichko and M. K. Kunka, September 1983. (Proprietary)

[13] WCAP-il851, " Analysis of Capsule U from the Pacific Gas and Electric Company Diablo Canyon Unit 2 Reac: Vessel Radiation Surveillance Program", S. E. Yanichko, et al., May 1988.

~

[14] WCAP 12811. " Analysis of Capsule X from the Pacific Gas and Electric Company Diablo Canyon Unit 2 Reactor Vessel Radiation Surveillance Program", E. Terek, et al., December 1990.

[15] Combustion Engineering,Inc., Metallurgical Research & Development 'vhemical Analysis of Wire-Flux Test Coupon",2-10-70. ,

[16] WCAP-8457, "Duquesne Light Company Beaver Valley Unit No.1 Reactor Vessel Radiation

~

Surveillance Program". J. A. Davidson, et al., October 1974.

[17] Combustion Engineering. Inc., Welding Material Certification and Release for Section III,

" Chemical Analysis of Test Weld Sample", Sample No. D 7279, August 12,1%9.

m

[18] Lukens Steel Company letter from John A. Soltesz to S. E. Yanichko dated December 6, 1973 Ladle Copper Analysis for Lukens Vessel Materials.

[19] Combustion Engineering,Inc., Metallurgical Research & Development Department, " Materials Cenification Repon", Job No. V-70333-001, March 5,1%9.

[20] Lukens Steel Company, " Test Cenificate", Mill Order No. 20241 2, corrected copy dated 7 68.

[21] Combustion Engineering, Inc., Metallurgical Research & Development Dep etment, " Materials -

Cenification Repon", Job No. V-70333404, corrected copy dated 3/450.

[22] Lukens Steel Company, " Test Certificate", Mill Order No. 20241-2, dated 7-24-68.

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1 CE? ,

i

[23] Combustion Engineering, Inc., Motallurgical Research & Development Department, " Materials <

l l Certification Report", Job No. V "0333-007, corrected copy dated 3/31/70.

[24] Combustion Engineering, Inc Metallurgical Research & Development Department, " Materials Certification Report". Job No. 5 J0334-001, September 4,1%9.

4

[25] Lukens Steel Company, " Test Certificate", Mill Order No. 20241-3, dated 5 25-68.

[26] Combustion Engineering, Inc., Metallurgical Research & Develo ment Department, " Materials Certification Report", Job No. V-70334-005, September 4,1969.

[27] Lukens Steel Company, " Test Certificate", Mill Order No. 20241-3, dated 7-10-68.

s

[28] Combustion Engineering Inc., Metallurgical Research & Development Department, " Materials s

Certification Report", Job No. V-70334 009, September 4,1969.

[29] Lukens Steel Company, " Test Certificate" Mill Order No. 20241-3, dated 6-24-68.

[30] Combustion Engineering, Inc, Drawing 235-464-3, " Material Identification Vessel for Westinghouse Elecuic Corporation 173" 1.D. Reactor Vessel".

[31] WCAP-12354, " Analysis of Capsule X from the Duke Power Company McGuire Unit 1 Reactor Vessel Radiation Surveillance Programs", S.E. Yanichko, et al., August 1989.

[32] Westinghouse Electric Corporation Nuclear Service Division CMT - Analytical Laboratory, Waltz Mill Site, Analytical Request #15211 Alloy Analysis - Steel, Duke Power Company McGuire Nuclear Plant Unit 1 Lawrence Kardos, November 15,1993.

)

[33] WCAP-10867, " Analysis of Capsule U from the Duquesne Light Company Beaver Valley Unit 1 Reactor Vessel Radiation Stuveillance Program", R. S. Boggs, et al., September 1985.

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