ML20216D063

Non-proprietary Evaluation of Pressurized Thermal Shock for Prairie Island Unit 2

ML20216D063
Person / Time
Site:	Prairie Island
Issue date:	02/28/1998
From:	Abbott S, Christopher Boyd, Trombola D WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20216C995	List: ML20216C984 ML20216D014 ML20216D033 ML20216D063
References
WCAP-14638, WCAP-14638-R02, WCAP-14638-R2, NUDOCS 9803160268
Download: ML20216D063 (22)

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Revision 2 b.

EVALUATION OF PRESSURIZED THERMAL SHOCK FOR PRAIRIE ISLAND UNIT 2

/-

Etsiin gb o use E ner g y S y stem s m

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-14638, Revision 2

~

Evaluation of Pressurized Thermal Shock for Prairie Island Unit 2 S. L. Abbott February 1998 Work Performed Under Shop Order NLTP-106 Prepared by Westinghouse Electric Company for Northern States Power Company Approved:

d 4. d C A

C. H. Boyd, Manager

• Engineering & Materials Technology Approved:

D. M. Trombola, Acting Manager Mechanical Systems Integration WESTINGHOUSE ELECTRIC COMPANY Nuclear Service Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 C 1998 Westinghouse Electric Company All Rights Reserved as

I PREFACE Revision 2:

Added Executive Summary.

Revised Table 1 and added new Table 2 to add upper shell Cu and Ni content data.

Renumbered Tables 2,3,4,5 and 6 to Tables 3,4,5,6 and 7.

Revised Table 3 to add upper shell initial RT,er data.

Revised Table 4 to add " top of core" fluence data.

Revised paragraph (1) in Section 6.0 to indicate that the surveillance capsule data is deemed not credible.

Revised Table 5 by adding upper shell and revising chemistry factors per RG 1.99, Rev.

2, Position 1.1 Revised Table 6 to reflect lower shell and RG 1.99, Rev. 2 Position 2.1 including ratio procedure in weld metal calculation.

Revised Table 7 to add upper shell and update chemistry factors and RTpy, results.

Revision 1:

Revised format to fit updated WCAP standards.

Revised Tables 3 and 6 per updated fluences given in reference 5.

1 r

Verifed By: MM E. Terek Evaluation of PTS for Prairie Island Unit 2

ii EXECUTIVE

SUMMARY

The purpose of this report is to determine the RTers values for the Prairie Island Unit No. 2 reactor vessel beltline based upon the results of the Surveillance Capsule P evaluation. However, the surveillance capsule data is deemed to be not credible per Regulatory Guide 1.99, Revision 2 criteria. Since the upper shell to intermediate shell weld seam is located below the top of the active core, the upper shell forging material and the weld material were considered in the evaluation as beltline materials. Since the surveillance capsule data is not credible, the RTers values are calculated in accordance with Regulatory Guide 1.99, Revision 2 using the procedures and margins which give the most conservative results. Based upon these conservative results, the limiting forging material in the Prairie Island Unit No. 2 beltline is the lower shell forging with a projected EOL RTers value of 110 F using the Prairie Island Unit No. 2 surveillance capsule data. This value is well below the screening criteria of 270 F for plates and forgings in Regulatory Guide 1.99, Revision 2. The limiting weld material in the Prairie Island Unit No. 2 reactor vessel beltline is found in upper to intermediate shell weld seam W2 with an EOL RTers value of 143 F using the Prairie Island Unit No.

1 surveillance capsule data (which has surveillance weld material fabricated from the same heat of weld wire as Unit No. 2 weld seam W2). This RTrys value is well below the screening criteria value of 300 F for circumferential welds at EOL (35 EFPY).

Evaluation of PTC for Prairie Island Unit 2

iii TABLE OF CONTENTS PREFACE.......................................................................................................i EX E C U TIVE S U M MA RY............................................................................... ii LI ST O F TAB LES....................................

....................iv 1.0 l N TR O D U C TI O N....................................................................................

1 2.0 PRESSURIZED THERMAL SHOCK...................................

...2 3.0 METHOD FOR CALCULATION OF RTns.................................

3 4.0 VERIFICATION OF PLANT-SPECIFIC MATERIAL PROPERTIES.............

5 5.0 N E UTRON FLU ENC E VALU E S...................................................... 9 6.0 DETERMINATION OF RTy3 VALUES FOR ALL BELTLINE REGION MATE R lA L S................................................................................. 1 0

7.0 CONCLUSION

S........................................................................................14

8.0 REFERENCES

.........................15 Evaluation of PTS for Prairie Island Unit 2

iv LIST OF TABLES Table 1 Calculation of Average Cu and Ni Weight Percent Va!ues for Be!tline Region Base Materials............................................

.. 6 Table 2 Calculation of Average Cu and Ni Weight Percent Values for Beltline Region Weld Materials...................................................... 7 Table 3 Prairie Island Unit 2 Reactor Vessel Beltline Region Material P ro pe rtie s.............................................................................................. 8 Table 4 Peak Fluence (10" n/cm, E > 1.0 MeV) on the Pressure Vessel 2

Clad / Base Metal Interface for Prairie Island Unit 2.............................. 9 Table 5 Interpolation of Chemistry Factors Using Tables 1 and 2 of 10 CFR Pa rt 5 0. 61.................................

11 Table 6 Calculation of Chemistry Factors Using Surveillance Capsule Data Per Regulatory Guide 1.99, Revision 2, Position 2.1.......................... 12 Table 7 RTers Calculations for Prairie Island Unit 2 Beltline Region Materials at EOL (35 EFPY)..............................................

.... 13 5

e i

Evaluation of PTS for Prairie Island Unit 2

1

1.0 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 the propagation of flaws 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 RTers values for the Prairie island Unit 2 reactor vessel using the results of the surveillance Capsule P evaluation. Section 2.0 discusses the PTS Rule and its requirements. Section 3.0 provides the methodology for calculating RTprs-Section 4.0 provides the reactor vessel beltline region material properties for the Prairie Island Unit 2 reactor vessel. The neutron fluence values used in this analysis are presented in Section 5.0. The results of the RTers 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 Evaluation of PTS for Prairie Island Unit 2

2 2.0 PRESSURlZED THERMAL SHOCK 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 requirements for reactor pressure vessels, including pressurized thermal shock requirements.

The revised PTS Rule"3,10 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 three changes:

1.

The rule incorporates in total, and therefore.nakes binding by rule, the method for determining the reference temperature, RTu, including treatment of the unirradiated RTa value, the margin term, and the explicit definition of " credible" surveillance data, which is currently described in Regulatory Guide 1.99, Revision 2r21, 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 of life (EOL) fluence, RTns.

3.

Thermal annealing is identified as a method for mitigating the effects of neutron irraoistion, thereby reducing RTns.

The PTS Rule requirements consist of the following:

For each pressurized water nuclear power reactor for which an operating license has been issued, the licensee shall have projected values of RTns, accepted by the NRC, far each reactor vessel bettiine material for the EOL fluence of the material.

The assessment of RTprs must use the calculation procedures given in the PTS Rule, and must specify the bases for the projected value of RTns for each vessel 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 a 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 RTpts 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 RTpys screening criterion values for the beltline region are:

270 F for plater, forgings, and axial weld materiais, and 300 F for circumferential weld materials.

Evaluation of PTS for Prairie Island Unit 2

3 3.0 METHOD FOR CALCULATION OF RTers RTpts must be calculated for each vessel beltline material using a fluence value, f, which is the EOL f'uence for the material. Equation 1 must be used to calculate values of RTuor for each weld and plate or forging in the reactor vessel beltline.

RTwor = RTuorm + M+ A RTuor (3)

RTuorm =

reference temperature for a reactor vessel material in the pre-service or unirradiated condition e

M

=

Margin to be added to account for uncertainties in the values of RTuorcu>, copper and nickel contents, fluence and calculational procedures. M is evaluated from Equation 2.

M = 2do + ai 2u (2) o is the standard deviation for RTuorm.

u o = 0 F when RTuorm is a measured value u

o = 17'F when RTuorm is a generic value u

or is the standard deviation for ARTuor.

For plates and forgings:

og = 17'F when surveillance capsule data is not used og = 8.5 F when surveillance capsule data is used For welds:

og = 28 F when surveillance capsule data is not used oA = 14 F when surveillance capsule data is used og not to exceed one-half of ARTuor.

ARTuor is the mean value of the transition temperature shift, or change in RTuor, due to irradiation, and must be calculated using Equation 3.

A RTuor = (CF)

• f** "" '

(3)

{

i Evaluation of PTS for Prairie Island Unit 2

4 CF ('F) is the chemistry tct:r, which is a function of coppet end nicksl cont:;nt. CF is givsn in Table 1 for welds and Table 2 for base metal (plates or forgings) of the PTS Rule (10 CFR 50.61). Surveillance data deemed credible must be used to determine a mateiial-specific value of CF. A material-specific value of CF is determined in Equation 5.

f is the best estimate neutron fluence, in units of 10" n/cm (E > 1.0 MeV), at the clad-base-2 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 RTns.

Equation 4 must be used for determining RTns using Equation 3 with EOL fluence values for determining ARTns.

RTm = RTiemn + M+ A RTm (4)

To verify that RT,e7 or each vessel beltline material is a bounding value for the specific reactor f

VIssel, licensees shall consider plant-specific information that could affect the level of smbrittlement. This information includes but is not limited to the reactor vessel operating i

t;mperature and any related surveillance program results. Results from the plant specific surveillance program must be integrated into the RT, y estimate if the plant-specific surveillance d;ta has been deemed credible.

A material-specific value of CF is determined from Equation 5.

CF = I[Ai

• fp2moint.>j o.u zoigr,>)

(5) in Equation 5, "A,"is the measured value of ART,or and "f,"is the fluence for each surveillance dita point. If there is clear evidence that the copper and nickel content of the surveillance weld differs from the vesse! weld, i.e., differs from the average for the weld wire heat number casociated with the vessel weld and the surveillance weld, the measure values of ART,e7 must i

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.

Evaluation of PTS for Prairie Island Unit 2

5 4.0 VERIFICATION OF PLANT-SPECIFIC MATERIAL PROPERTIES Before performing the pressurized thermal shock evaluation, a review of the latest plant-specific material properties for the Prairie Island Unit 2 vessel was performed. The beltline region of a reactor vessel, per the PTS Rule, is definod as "the region of the reactor vessel (shell material including welds, heat-affected zones and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage". For Prairie Island Unit 2, the upper shell forging and the upper shell to intermediate shell weld seam are included in the beltline region since the weld seam is 6.833 inches below the top of the active fuel stack.

Material property values were obtained from material test certifications from the original fabrication as well as the additional material chemistry tests performed as part of the Prairie Island Unit 2 surveillance capsule testing program % The average copper and nickel values were calculated for each beltline region material using all of the available material chemistry l

information as shown in Tables 1 and 2. A summary of the pertinent chemical and mechanical properties of the beltline region forgings and weld material of tha Prairie Island Unit 2 reactor vesselis given in Table 3.

O Evaluation of PTS for Prairie Island Unit 2 1

6 l

l Table 1 Calculation of Average Cu and Ni Weight Percent Values for Beltline Region Base Materials Ref.

Intermediate Shell Lower Shell Upper Shell A533 Gr. 8, CL1 Forging 22829 Forging 22642(*)

Forging Correlation Monitor 22231/39088 )

Naterial th (HSST Plate 02)

Cu %

Ni %

Cu %

Ni %

Cu %

NI %

Cu %

Ni %

3 0.085 0.70 0.14 0.68 4

0.068 0.694 5

0.068 0.575 6

0.075 0.75 6

0.070 0.75 7

0.085 0.700 7

0.085 0.700 8

0.065 0.73 8

0.065 0.73 Avg.

0.0725 0.75 0.0782 0.6738 0.065 0.73 0.14 0.68 NOTES.

(a) Surveillance program base metal material.

(b) Forging B is included since it extends 6.833 inches below the top of the active fuel stacs.

Evaluation of PTS for Prairie Island Unit 2

7 Table 2 Calculation of the Average Cu and Ni Weight Percent Values for the Prairie Island Unit 2 Weld Materials Surveillance Program Inter to Lower Shell Upper to inter Shell Weld Metal (*)

Weld Seam W3(*

Weld Seam W2" Ref.

Cu %

Ni %

Cu %

Ni %

Cu %

Ni %

3 0.082 0.072 4

0.076 0.071 5

0.094 0.103 a

5 0.081 0.087 J

5 0.078 0.081 9

0.090 0.130 I

0.0822(d 0.0828("

10 0.14 0.14 11 0.14 0.17 12 0.105 0.11 13 0.14(*

0.11 #

Average 0.0822 0.0828 0.0861 0.1064 0.13125 0.1325 NOTES:

(a) The Surveillance Program weld metal was fabricated with Weld Wire Type UM40, Heat No. 2721, Flux Type UM89, Lot No.1263 and is identical to the intermediate to lower shell girth weld seam W3.

(b) Weld seam W2 is being included here, since it is 6.833 inches below the top of the active fuel stack.

Weld seam W2 was fabricated with Weld Wire Type UM40, Heat No.1752, FluxType UM89, Lot No.

1263.

(c) These values are the average of all data points from the Prairie Island Unit 2 surveillance weld which was fabricated with the same heat of weld wire (Ref. 5).

(d) These values are the average of all data points from the Prairie Island Unit 1 surveillance weld which was fabricated with the same heat of weld wire (Ref.14).

Evaluation of PTS for Prairie Island Unit 2

8 Table 3 Prairie Island Unit 2 Reactor Vessel Beltline r lion Material Properties

.. ou Material Description Cu (%) (*)

Ni(%)(*)

RT ottu) ( F)

N i

intermediate Shell Forging 22829 0.07 0.75

-4*)

Lower Shell Forging 22642 0.08 0.67

-6*)

)

Upper Shell Forging 22231/39088 0.07 0.73

-13(4 Inter / Lower Shell Girth Weld 0.frJ 0.11

-31*)

Upper / Inter. Shell Girth Weld 0.13 0.13

-13(*

NOTES:

(:) Average values of copper and nickel are from Tables 1 and 2 and rounded per the procedure given in ASTM E 29.

The RT ottu) values for the intermediate and lower shell forgings and the inter./lawer shell girth weld (b)

N are measured values and were obtained from Prairie Island Unit 2 FSAR.

(c)

The RT ottu) value for the upper shell forging is a measured value obtained from Reference 8.

N (d)

The RT oT(u) value for the upper / inter, shell girth weld is a measured value obtained from Reference N

14.

Evaluation of PTS for Prairie Island Unit 2

9 5.0 NEUTRON FLUENCE VALUES The calculated fast neutron fluence (E > 1.0 MeV) values at the inner surface of the Prairie Island Unit 2 reactor vessel are shown in Table 4. The core rnid-plane and top of cora values were projected using the results of the Capsule P radiation analysis. See Section 6.0 of WCAP-14613W l

Table 4 Peak Fluence (10" n/cm', E > 1.0 MeV) on the Pressure Vessel Clad / Base Metal Interface for Prairie Island Unit 2

)

EFPY 0'

Core Mid-Plane 17.24 2.44 24 3.11 32 3.89 35 4.18 Top of Core 17.24 1.39 24 1.76 32 2.21 35 2.38

\\

i l

Evaluction 6f PTS for Prairie Island Unit 2 J

10 6.0 DETERMINATION OF RTers VALUES FOR ALL BELTLINE REGION MATERIALS Using the prescribed PTS Rule methodology, RTns values were generated for all beltline region materials of the Prairie Island Unit 2 reactor vessel for fluence values at the EOL (35 EFPY).

Each plant shall assess the RTns values based on plant-specific surveillance capsule data. For Prairie Island Unit 2, the related surveillance program results have been included in this PTS evaluation. Specifically, the Prairie Island Unit 2 plant-specific surveillance capsule data for the lower shell forging D and weld metal is provided and applied as follows:

1) There have been four capsules removed from the reactor vessel.

2) The data for the surveillance program forging material is deemed not credible. However, the data was used with a og margin of 17 F.

3) The data for the Unit 2 surveillance program weld materialis deemed credible.

4) The data for the Unit i surveillance program vveld material used to evaluate RTprs for the Unit 2 upper to intermediate shell weld metal is deemed not credible. However, the data was used with a og margin of 28'F.

5) The surveillance capsule materials are representative of the actual vessel forgings and circumferential weld metal.

6) The resulting RTns values are well below the PTS Rule screening criteria.

As presented in Table 5, chemistry factor values for Prairie Island Unit 2 based on average copper and nickel weight percent were calculated using Tables 1 and 2 from Regulatory Guide 1.99, Revision 2 (Position 1.1)l4. Additionally, chemistry factor values based on the surveillance capsule data are calculated in Table 6. Table 7 contains the RTns calculations for all beltline region materials at 35 EFPY.

Evaluation of PTS for Prairie Island Unit 2

11 Table 5 Prairie Island Unit 2 Chemistry Factors per Table 1 and 2 of Regulatory Guide 1.99, Revision 2, Position 1.1 Material Ni, wt %

Chemistry Factor, 'F Upper Shell Foraina 22231/39088 0.73 44 Given Cu wt% = 0.07 Intermediate Shell Foraina 22829 0.75 44 Given Cu wt% = 0.07 g

Lower Shell Foroino 22642 0.67 51 Given Cu wt% = 0.08 Voper to Intermediate Shell Girth Weld 0.13 69.7 Given Cu wt% = 0.13 Intermediate to Lower Shell Girth Weld 0.11 51.6 Given Cu wt% = 0.09 Surveillance Weld (Unit 2) 0.08 44.8 Given Cu wt % = 0.08 Eurveillance Weld (Unit 1) 0.11 70.9 Given Cu wt % = 0.14 Evaluation of PTS for Prairie Island Unit 2

12 Table 6 Calculation of Chemistry Factors Using Surveillance Capsule Data Per Regulatory Guide 1.99, stevision 2, Position 2.1 Material Capsule Capsule f)

FF" ARTuot*

FF*ARTuor FF 2

Lower Shell V

0.6206 0.866 35.28 30.55 0.75 Forging 22642 (Axial Orientation)

T 1.199 1.051 29.93 31.46 1.10 R

4.376 1.375 84.73 118.50 1.89 P

4.165 1.365 103.87 141.78 1.86 Lower Shell V

0.6206 0.866 32.89 28.48 0.75 Forging 22641 (Tangential T

1.199 1.051 55.69 58.53 1.10 Orientation)

R 4.376 1.375 90.02 123.78 1.89 P

4.165 1.365 99.91 136.38 1.86 SUM 667.46 11.2 2

CFw sneu rnoing = I(FF

• ARTuoT) + I(FF )

= 667.46 + 11.2 = 59.6 *F Weld Metal

• V 0.6206 0.866 80.58" 69.78 0.75 T

1.199 1.051 66.39" 69.78 1.10 R

4.376 1.375 115.36*

158.62 1.89

]

P 4.165 1.365 110.68*

151.08 1.86 SUM 449.26 5.6 2

CFw.m u.i = I(FF

• ARTwoT) + I(FF )

= 449.26 + 5.6 = 80.2*F NOTES:

(a) f = fluence (10 n/cm, E > 1.0 MeV). All updated fluence values were taken from the Capsule 2

1 51 P analysis. WCAP-14613 (b)

FF = fluence factor = f

• 28'*

(c)

ARTuot values are measured values.

l53 (d)

These measured ARTuo7 values obtained from the Capsule P analysis were multiplied by a ratio factor of 1.15.

(CFv i + CFscw.m = 51.6 + 44.8 = 1.15)

Evaluation of PTS for Prairie Island Unit 2

13 Table 7 RTets Calculations for Prairie Island Unit 2 Beltline Region Materials at EOL (35 EFPY)

--mmmmmmpummma-ummmme Material CF f)

FF*)

RT oTru/4 M

ARTeis RTpTs N

Upper Shell Forging 44*F 2.38 1.234

-13*F 34*F 54.3*F 75*F 22231/39088 Upper to Inter. Shell 70*F 2.38 1.234

-13*F 56*F 86.4*F 129 F Weld Seam W2 Using* Unit 1 S/C 81*F 2.38 1.234

-13*F 56*FM 100.0 F 143*F Data Intermediate Shell 44*F 4.18 1.366

-4*F 34*F 60.1 *F 90 F Forging 22829 Lower Shell Forging 51*F 4.18 1.366

-6*F 34*F 69.7*F 98 F 22642 Using Surveillance 60 F 4.18 1.366

-6*F 34*FW 82.0*F 110*F Capsule Data Inter. to Lower Shell 52"F 4.18 1.366

-31*F 56*F 71.0*F 96*F Weld Seans W3 Using S/C Data 80*F 4.18 1.366

-31*F 28'FM 109.3 F 106*F NOTES:

(a) f = peak clad / base metal interface fluence (10" n/cm, E > 1.0 MeV) at 35 EFPY, Best-Estimate 2

values are used since the Best-Estimate values are greater than the calculated values (See Table 6-15 in Ref. 5}io m FF = f o re.

(b)

(c)

RTNottu) values are measured values.

(d)

This calculation uses the chemistry factor based on the surveillance capsule weld data from the Prairie Island Unit 1 surveillance program. The margin used is the full o3 value of 28'F since the surveillance weld data is not credible (See App. D in Ref.16).

(s)

Margin used is the full c3 value of 17*F since the forging surveillance data is not credible (See App. D in Ref. 5).

(f)

See Appendix D in Ref. 5.

l Evaluation of PTS for Prairie Island Unit 2

14 1

7.0 CONCLUSION

S As shown in Table 7, all of the beltline region materials in the Prairie Island Unit 2 reactor vessel have EOL RTers values well below the screening criteria values of 270 F for plates or forgings and longitudinal welds and 300 F for circumferential welds at EOL (35 EFPY).

i

(

1 l

Evaluation of PTS for Prairie Island Unit 2

m 15

8.0 REFERENCES

1.

10 CFR Part 50.61," Fracture Toughness Requirements for Protection Against i

Pressurized Thermal Shock Events", Federal Register, Volume 60, No. 243, dated i

December 19,1995, effective January 18,1996.

2.

Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel i

Materials," U.S. Nuclear Regulatory Commission, May 1988.

j 3.

WCAP-8193, "Northem States Power Co. Prairie Island Unit No. 2 Reactor Vessel Radiation Surveillance Program", S.E. Yanichko et. al., September 1973.

4.

WCAP-11343, " Analysis of Capsule R from the Northem States Power Company Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program", S.E. Yanichko et. al.,

December 1986.

5.

WCAP-14613 Rev. 2, " Analysis of Capsule P from the Northem States Power Company Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program", S.L. Abbott, et.

al., February 1998.

6.

Societe Des Forges Et Ateliers Du Creusot Usines Schneider, Report No. 3.5.7, Rev.1, item C, Heat No. 22829, Order No. 700 814/54, Dated September 8,1970.

)

7 Societe Des Forges Et Ateliers Du Creusot Usines Schneider, Report No. 4.5.7, Rev.1, item D, Heat No. 22642, Order No. 700 814/54, Dated June 24,1970.

8.

Societe Des Forges et Ateliers Du Creusot Usines Schneider, Report No. 2.5.7, Rev.1, ltem B, Heat No. 22231/39088, Order No. 700 814/54, Dated February 27,1970.

j 9.

Proces Verbal De Recette De Produits De Soudure, SFAC, Code No.119 957 TN 54 -

119 937 TN 54, Designation UM 40 (fill) Lot No. 2721, UM 89 (flux), Lot No.1263, Specification PS 308/R, Dated 7/31/70.

10.

Proces Verbal De Recette De Produits De Soudure, S.A.F., Code No. 118 716 A 54 -

119 937 TN 54, Designation UM 40 (fill) Lot No. 1752-69, UM 89 (flux), Lot No.1263, Specification PS 308/R, Dated 3/26/70.

11.

Proces Verbal De Recette De Prod"t De Soudure, S.A.F., Code No. 118 716 A 54, Designation UM 40 (fill) Lot No.17/. W, UM 89 (flux), Lot No.1230, Specification PS 308/R, Dated 3/31/70.

12.

Proces Verbal De Recette De Produits De Soudure, S.A.F., Code No. 118 716 A 54, Designation UM 40 (fill) Lot No. 1752-69, UM 89 (flux), Lot No.1180, Specification PS

'l 308/R, Dated 10/13/68.

l 13.

Calc No.1, " Prairie Island Unit No. 2 Reactor Vessel Toughness Properties," S. E.

Yanichko, 5/4/77.

14.

WCAP-14781, Revision 3, " Evaluation of Pressurized Thermal Shock for Prairie Island Unit 1," S.L. Abbott, February 1998.

Evaluation of PTS for Prairie Island Unit 2

l 16 15.

SAE-REA-98-190, " Calculated and Best Estimate Reactor Vessel Fast Neutron Exposures for Prairie Island Units 1 and 2," Letter from REA to MSI dated January 6, 1998.

16.

WCAP-14779 Rev. 2, " Analysis of Capsule S from the Northem States Power Company Prairie Island Unit 1 Reactor Vessel Radiation Surveillance Program", S.L.

1 Abbott, et al., February 1998.

i i

l Evaluation of PTS for Prairie Island Unit 2