ML20216D014
| ML20216D014 | |
| 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: |
| References | |
| WCAP-14781, WCAP-14781-R03, WCAP-14781-R3, NUDOCS 9803160260 | |
| Download: ML20216D014 (21) | |
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,,,,,,,,3 EVALUATION OF J
. PRESSURIZED THERMAL SHOCK i
FOR P.RAIRIE ISLAND
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WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-14781, Revision 3 Evaluation of Pressurized Thermal Shock for Prairie Island Unit 1 S. L. Abbott February 1998
(
Work Performed Under Shop Order NLTP-106 Prepared by Westinghouse Electric Company for Northem States Power Company Approved:
b C'H. Boyd,~ Managqr ' '
Engineering & Materials Technology Approved:
D. M. Trombola, Acting Manager Mechanical Systems Integration l
l WESTINGHOUSE ELECTRIC COMPANY Nuclear Service Division P.O. Box 355 i.
Pittsburgh, Pennsylvania 15230-0355 4
C 1998 Westinghouse Electric Company i
All Rights Reserved yn
j I_
PREFACE Revision 3:
I Added Exr:utive Summary.
Revised Table 1 to include additional core region materials with Cu and Ni content data.
l l
Added Table 2 to include nozzle shell base metal and weld Cu and Ni content data.
Renumbered Tables 3 through 8.
l Revised Table 3 to include nozzle shell forging and weld chemistry and material property data.
i l
Revised text in Section 6 (P.10) to justify use of not credible survellance capsule data.
Revised Table 4 to update fluence data.
Revised Table 5 to update chemistry factors and fluence accordant with surveillance capsule data.
Revised Table 6 to update chemistry factor calculations based on surveillance capsule data.
Revised Table 7 to include nozzle shell materials and to update RTns calculations based on surveillance capsule data.
Revised Reference 5 to WCAP-14779, Rev. 2.
Added References 8 through 13 as sources of additional material properties.
Revision 2:
)
Revised initial RT,a and RTns in Tables 2 and 6.
Revision 1:
Revised Tables 3 and 6 per updated fluences given in reference 5.
I Verified By:
T. J. LaulMJr.
l s
Evaluation of PTS for Prairie Island Unit 1
ii EXECUTIVE
SUMMARY
The purpose of this report is to determine the RTers values for the Prairie Island Unit No.1 reactor vessel beltline based upon the results of the Surveillance Capsule S evaluation.
However, the surveillance capsule data is deemed to be not credible per Regulatory Guide 1.99, Revision 2 criteria. Furthermore, the nozzle (upper) shell to intermediate shell weld seam is located below the top of the active core, and 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 RTets values are calculated in accordance with Regulatory Guide 1.99, l
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.1 l
beltline is the intermediate shell forging with a projected EOL RTns value of 122*F using the l
Prairie Island Unit No.1 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.1 reactor vessel beltline is found in nozzle to intermediate shell weld seam W2 with an EOL RTns value of 162*F. The maximum generic margin value was used to this RTns value since the weld seam W2 material is not included in the surveillance capsule program. This RTns value is also well below the acreening criteria value of 300*F for circumferential welds at EOL (35 EFPY).
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l Evaluation of PTS for Prairie Island Unit 1 l
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l TABLE OF CONTENTS l
j PREFACE............................................................................
i EXEC UTIVE S U M MARY................................................................................
ii i
l LI ST OF AB LE S......................................
iv 1
1 N TR O D U C TI O N.............................................................................
1 l
2 PRESSU RIZED TH ERMAL SHOCK...................................................................
2 4
3 METHOD FOR CALCULATION OF RTns.......................................................
3 I
4 VERIFICATION OF PLANT-SPECIFIC MATERIAL PROPERTIES........................
5 l
5 N EUTRON FLU ENCE VALUES........................................................
9 6
DETERMINATION OF RTns VALUES FOR ALL BELTLINE REGION MATERIALS...............................................................................
10 7
C O N C LU S I O N S...........................................................
14 8
R E F E R EN C E S.................................................
15 l
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t Evaluation of PTS for Prairie Island Unit 1
iv LIST OF TABLES Table 1 Calculation of Average Cu and Ni Weight Percent Values for Beltline Region Materials..............................
6 I
Table 2 Calculation of Average Cu and Ni Weight Percent Values for Materials Near Beltline Region.......................................
7 Table 3 Prairie Island Unit 1 Reactor Vessel Beltline Region Material Pro pe rtie s.........................,..,
8 Table 4 Fluence (10" n/cm, E > 1.0 MeV) on the Pressure Vessel Clad / Base 2
Metal Interface for Prairie Island Unit 1.................
9 Table 5 Interpolation of Chemistry Factors Using Tables 1 and 2 of 10 CFR P a rt 50. 61.............,,,..........,,,,,,.,,,,,,,,,,,,,,,,,,,,,,,
11 Table 6 Calculation of Chemistry Factors Using Surveillano Capsule Data Per 10 CFR Part 50.61.....
12 Table 7 RTns Calculations for Prairie Island Unit 1 Beltline Region Materials at E O L (3 5 E F P Y)...................................................................
13 o
Evaluation of PTS for Prairie Island Unit 1
1 1
INTRODUCTION 1
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 concem arises if one of these transients acts on the j
I 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 i reactor vessel using the results of the surveillance Capsule S evaluation. Section 2 discusses the PTS Rule and its requirements. Section 3 provides the methodology for calculating RTprs. Section 4 provides the reactor vessel beltline region material properties for the Prairie Island Unit 1 reactor vessel. The neutren fluence values used in this analysis are presented in Section 5. The results of the RTp13 calculations are presented in Section 6. The conclusion and references for the PTS evaluation follow in Sections 7 and 8, respectively.
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a Evaluation of PTS for Prairie island Unit 1
2 2
PRESSURIZED THERMAL SHOCK The Nuclear Regulatory Commission (NRC) recently amended its regulations for light-water-cooled nuclear power plants to clarify severalitems related to the fracture toughness requirements for reactor pressure vessels, including pressurized thermal shock requirements.
Th2 revised PTS RuleW,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 maketi binding by rule, the method for determining the reference temperature, RTer, including treatment of the unirradiated RTer value, the margin term, and the explicit definition of " credible" surveillance data, which is currently described in Regulatory Guide 1.99, Revision 2A.
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, RTpys.
2 3.
Thermal annealing is identified as a method for mitigating the effects of neutron irradiation, thereby reducing RTprs.
The PTS Rule requirements consist of the following:
For each pressurized water nuclear power reactor for which an operating license has o
been issued, the licensee shall have projected values of RTers, accepted by the NRC, for each reactor vessel beltline material for the EOL fluence of the material.
The assessment of RTers must use the calculation procedures given in the PTS Rule, o
and must specify the bases for the projected value of RI rs for each vessel beltline material. The report must specify the copper and nickei ;;,atents and the fluence values used in the cale'llation for each beltline material.
\\
n This assessment must be updated whenever there is a significant change in projected o
values of RTpys or upon the request for a change in the expiration date for operation of the facility. Changes to RTprs values are significart 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 RT,rs screening criterion values for the beltline region are:
e 270*F for plates, forgings, and axial weld materials, and 300 'F for circumferential weld materials.
t Ev luation of PTS for Prairie Island Unit 1
3 3
METHOD FOR CALCULATION OF RTers RTpys 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 o' RTuor for each weld and plate or forgi.1g in the reactor vessel beltline.
RTwor = RTworw> + M+ o RTwor (1)
RTuorm =
reference temperature for a reactor vessel material in the pre-service or unirradiated cordite M
=
Margin to be added to account for uncertainties in the values of RTuorm, ;opper and nickel contents, fluence and calculational procedures. M is evaluated from Equation 2.
M = 2dai,+ oi (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
og 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 og = 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 RTwor = (CF)
- fe2:4 oion (3)
Evaluation of PTS for Praine Island Unit 1 j
4 CF (*F) is the chemistry factor, which is a function of copper and nick 21 cont:nt. 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.
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 = RTwrm + M+ A RTm (4)
To verify that RT for each vessel beltline material is a bounding value for the specific reactor vessel, licensees shall consider plant-specific information that could affect the level of embrittlement. This in'ormation includes but is not limited to the reactor vessel operating i
temperature and any related surveillance program results. Results from the plant specific f
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.
l CF = I[Ai
- fp2umgr.)j mm,,,q (5)
In Equation 5, "A," is the measured value of ART, and "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 measure values of ARTer 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.
Evaluatsn of PTS for Prairie Island Unit 1
I l
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4 VERIFICATION OF PLANT-SPECIFIC MATERIAL PROPERTIES j
Before performing the pressurized thermal shock evaluation, a review of the latest plant-specific l
material properties for the Prairie Island Unit 1 vessel was performed. The beltline region of a I
reactor vessel, per the PTS Rule, is defined 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".
l 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 i surveillance capsule testing program. The average copper and nickel values m
were calculated for each beltline region material using all of the available material chemistry information as shown in Tables 1 and 2. A summary of the pertinent chemical and mechanical properties of the Delt!ine region forgings and weld material of the Prairie Island Unit 1 reactor vesselis given in Table 3.
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Evaluation of PTS for Prairie Island Unit 1
6 Table 1 Calculation of Average Cu and Ni Weight Percent Values for Beltline Region Materials Intermediate Shell Lower Shell interJLower Shell Correlation N
Forging C Forging D Circumferential Weld (* bl Monitor Material Ref.
Cu %
NI %
Cu %
NI %
Cu %
Ni %
Cu %
Ni %
6 0.06 0.72 6
0.06 0.72 7
0.07 0.66 7
0.065 0.66 3
0.13 0.14 0.68 4
0.13 0.09 5
0.078 0.956 0.149 0.138 5
0.138 0.118 5
0.143 0.091 Average 0.07 0.80 0.07 0.66 0.14 0.11 0.14 0.68 Best Estimate Chemistry Average of 0.14 0.11 Surv. Weld N
Chemistry Heat 0.14 0.14 1752/1263*
Heat 0.14 0.17 1752/1230*
Heat 0.105 0.11 1752/1180"'
Best 0.13 0.13 Estimate NOTES:
(a) Surveillance program base metal material.
(b) The surveillance weld specimens were made of the same wire and flux as the intermediate to lower shell circular seam (Wire Type UM40, Heat 1752 and Flux Type UM 89, Lot 1230).
(c) Per Reference 8.
(d) Per Reference 9.
(e) Per Reference 10.
Evaluation of PTS for Prairie Island Unit 1 I
7 TaNe 2 Calculation of Average Cu and Ni Weight Percent Values for Materials Near Beltline Region Material
% Cu
% Ni Nozzle Shell Forging B(*)
0.075 0.68 Nozzle to intermediate Shell 0.17 0.15 Circumferential Weld" 0.12 0.14 Weld Material Average 0.15 0.15 NOTES:
(a) Per Reference 13 (b) Weld Seam W2 was fabricated with Weld Wire Type UM40, Heat No. 2269, Flux Type UM89, Lot No.1180 (References 11,12 and 13) i 1
Evaluation of PTS for Prairie Island Unit 1
8 Table 3 Prairie Island Unit 1 Reactor Vessel Beltline Region Material Properties Material Description Cu (%) (*)
Ni (%) (*)
RTecia (*F) *)
Nozzle Shell Forging B 0.08 0.68
-4 Intermediate Shell Forging C 0.07 0.80 14 1
Lower Shell Forging D 0.07 0.66
-4 Nozzle / Inter. Shell Circumferential 0.15 0.15 0(*
Weld (Heat 2269)
Inter / Lower Shell Circumferential 0.13 0.13
-13(*
Weld NOTES:
(a) Average values of copper and nickel as indicated in Tables 1 and 2 on preceding pages.
(b) The RTecTu values for the forgings and weld are measured values.
(c) Per NSP, an initial RTmi of -13 'F was used.
(d) Estimated per Standard Review Plan Section 5.3.2 (See Ref.13).
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Evaluation of PTS for Prairie '.sland Unit 1
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S 5
NEUTRON FLUENCE VALUES The calculated fast neutron fluence (E > 1.0 MeV) values at the inner surface of the Prairie Island Unit i reactor vessel are shown in Table 4. These values were projected using the i
results of the Capsule S radiation analysis. See Section 6.0 of the Capsule S analysis report, WCAP-14779 Rev. 2*
(
Table 4 Fluence (10" n/cm', E > 1.0 MeV) on the Pressure Vessel -
Clad / Base Metal Interface for Prairie Island Unit 1 unummmmmmusu EFPY 0*
15' 30' 45' Ca!culated Fluences @ Core Mid-Plane 18.12 2.45 1.55 1.10 0.943 24 2.97 1.93 1.40 1.20 35 3.95 2.62 1.95 1.69 Calculated Fluences @ Top of Core 18.12 1.37 0.865 0.613 0.526 24 1.66 1.07 0.778 0.670 35 2.20 1.46 1.09 0.940
)
Evaluation of PTS for Prairie Island Unit 1
10 6
DETERMINATION OF RTpTs VALUES FOR ALL BELTLINE l
REGION MATERIALS Using the prescribed PTS Rule methodology, RTpts values were generated for all beltline j
region materials of the Prairie Island Unit 1 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 1, the related surveillance program results have been included in th's PTS evaluation. Specifically, the Prairie island Unit 1 plant-specific surveillance capsule data for the intermediate shell forging C and weld metal is provided for the following reasons:
- 1) There have been four capsules removed from the reactor vessel.
- 2) The surveillance capsule forging data is deemed not credible per Regulatory Guide 1.99, Revision 2 (See Reference 5). However, the data was used with a e4 margin of 17'F.
- 3) The surveillance capsule weld data is deemed not credible per Regulatory Guide 1.99 Revision 2 (See Reference 5). However, the data was used with a c margin of 28'F.
a
- 4) The surveillance capsule materials are representative of the actual vessel forgings and circumferential weld metal.
- 5) The resulting RTns values are well below the PTS Rule screening criteria.
As presented in Table 5, chemistry factor values for Prairie Island Unit 1 based on average W
copper and nickel weight percent were calculated using Tables 1 and 2 from 10 CFR 50.61.
Additionally, chemistry factor values based on the surveillance capsule data are calculated in i
Table 6. Table 7 contains the RTers alculations for all beltline region materials at 35 EFPY.
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Evaluation of PTS for Prairie Island Unit 1
11 Table 5 Interpolation of Chemistry Factors Using Tables 1 and 2 of 10CFR Part 50.61 Material Ni, wt %
Chemistry Factor, 'F Intermediate Shell Fomina C 0.80 44 Given Cu wt% = 0.07 Lower Shell Foraina D 0.66 44 Given Cu wt % = 0.07 Lower to inter. Shell Cire. Weld Metal 0.00 58 Given Cu wt % = 0.13 0.13 69.7 0.20 76 Surveillance Weld Metal 0.00 61 Given Cu wt % = 0.14 0.11 70.9 0.20 79 Correlation Monitor Material 0.60 100 Given Cu wt % = 0.14 0.68 102 0.80 105 Nozzle Shell Foraina B 0.68 51 Given Cu wt % = 0.08 Nozzle to Inter. Shell Cire. Weld Metal 0.00 66 Given Cu wt% = 0.15 0.15 79.5 0.20 84 Bold Values are the Interpolated Chemistry Factors Ratio Procedure CFv
,wm = 69.7 CFsu vwm =
70.9 Ratio =
(CFvuwm + CFsor,w.io) = 69.7 + 70.9 = 0.98 l
Evaluation of PTS for Prairie Island Unit 1
l 12 Table 6 Calculation of Chemistry Factors Using Surveillance Capsule Data Per 10CFR Part 50.61 Material Capsule Capsule FF )
ARTuot FF*ARTuot FF S
(*
2 gia)
Intermediate Shell V
0.6267 0.869 24.07 20.92 0.755 Forging C (Axial)
P 1.314 1.076 33.98 36.56 1.158 R
4.000 1.356 84.18 114.15 1.839 S
4.338 1.373 74.27 101.97 1.885 Intermediate Shell V
0.6267 0.869 56.36 48.98 0.755 Forging C (Tangential)
P 1.318 1.076 23.11 24.87 1.158 R
4.000 1.356 95.85 129.97 1.839 S
4.338 1.373 101.46 139.* J 1.885 SUM 616.72 11.27 8
CFw.,m.a.m.sn.n F-oing = I(FF
- ARTwot) + I(FF )
= 54.7'F Weld Metal (*
V 0.6267 0.869 34.38 33.69(*)
29.28 0.755 P
1.314 1.076 45.15 44.25(*
47.61 1.158 R
4.000 1.356 122.47 120.02(*
162.75 1.839 S
4.338 1.373 160.43 157.22(*
215.86 1.885 SUM 455.50 5.64 8
CFw.o u.,, = I(FF
- ARTuot) + I(FF )
= 80.8'F NOTES:
(a) f = fluence (10 n/cm', E > 1.0 MeV). All updated calculated fluence values were taken from f
the Capsule S analysis f2e-oimCAP-14779 *).
FF = fluence factor = f <
(b)
(c)
ARTwot values were obtained from CVGRAPH Version 4.1 plots (See WCAP-147799 (d)
The reactor vessel intermediate to lower shell circular weld seam was made with the same weld wire and flux as the surveillance weld specimens (Wire UM40, heat number 1752, UM 89 flux, batch no.1230. The ratio procedure is used since the average Cu and Ni content of the vessel weld differs from that of the surveillance weld material (See Table 5).
(e)
Ratio of 0.98 applied (See ' Ratio Procedure' calculation on previous page).
Evaluation of PTS for Prairie Island Unit 1
13 i
i Table 7 RTm Calculations for Prairie Island Unit 1 Beltline Region Materials at EOL (35 EFPY)
--mmmmmmmmmmme mamma summmmum Material CF f*)
FF" RTworf/4 M
(*F)
(*F
('F)
(*F)
('F)
Inter. Shell Forging C 44.0 3.95 1.35 14 34 59.4 107 l
Using Surveillance 54.7 3.95 1.35 14 34(*)
73.8 122 Capsule Data Lower Shell Forging D 44.0 3.95 1.35
-4 34 59.4 89 Inter. to Lower Shell 69.7 3.95 1.35
-13 56 94.1 137 Cire. Weld Using Surveillance 80.8 3.95 1.35
-13 56(8) 109.1 152 Capsule Data Nozzle to Inter. Shell 79.5 2.20 1.21 0'o 66 96.2 162 Cire. Weld Nozzle (Upper) Shell 51 2.20 1.21
-4 34 61.7 92 Forging B i
NOTES:
(a) f = peak clad / base metalinterface fluence (10 n/cm, E > 1.0 MeV) at 35 EFPY 2
(b) FF = fm2e.oio%n (c) RTwo,dYl'o, margin of 17'F for the forging and 28'F for the weld was used since the surveillan values are measured values.
(d) The f was deemed not credible.
l (e) Initial RTwor was estimated per Standard Review Plan Section 5.3.2 (See Ref.13).
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l Evaluation of PTS for Prairie Island Unit 1 i
l 14 7
CONCLUSIONS l
As shown in Table 6, all of the beltline region materials in the Prairie Island Unit i reactor vessel l
have EOL RTns values well below the screening criteria values of 270 F for plates or forgings i
and longitudinal welds and 300'F for circumferential welds at EOL (35 EFPY).
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i Evaluation of PTS for Praine Island Unit 1 L.
15 8
REFERENCES 1.
10 CFR Part 50.61," Fracture Toughness Requirements for Protection Against l
Pressurized Thermal Shock Events", Federal Register, Volume 60, No. 243, dated l
December 19,1995, effective January 18,1996.
2.
Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, May 1988.
3.
WCAP-8086, "Northem States Power Co. Prairie Island Unit No.1 Reactor Vessel Radiation Surveillance Program", S.E. Yanichko and D.J. Lege, June 1973.
4.
WCAP-11006, " Analysis of Capsule R from the Northem States Power Company Praine Island Unit 1 Reactor Vessel Radiation Surveillance Program", R.S. Boggs, et. al.,
February 1986.
5.
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. Abbott, February 1998.
6.
Societe Des Forges et Ateliers Du Creusot usines Schneider, Chemical Analysis Report No.17-9-2, NSP shell course C, heat 21918/38560.
j 7.
Societe Des Forges et Ateliers Du Creusot usines Schneider, Chemical Analysis Report No.15-8-1, NSP shell course D, beat 21887/38530.
8.
Proces Verbal De Recette De Produits De Soudre, S.A.F., Code No. 118 716 AS4, Spec
- PS 308/R, Designation UM 40 (fil)- UM 89 (flux), Lot No. 1752-69 (wire)-1263 (flux),
Dated 3/26/70.
9.
Proces Verbal De Recette De Produits De Soudre, S.A.F., C ode No.11B 716 A54, Spec
- PS 308/R, Designation UM 40 (fil)- UM 89 (flux), Lot No. 1752-69 (wire) -1230 (flux),
Dat6d 3/31/70.
10.
Proces Verbal De Recette De Produits De Soudre, S.A.F., Code No. 118 716 AS4, Spec
- PS 308/R, Designation UM 40 (fil) - UM 89 (flux), Lot No. 1752-69 (wire) -1180 (flux),
Dated 10/13/68.
11.
Procer Verbal De Recette De ProduPs De Soudre, S.A.F., Code No. 118 716 AS4, Spec
- PS 308/R, Desigreon UM 40 (fil) - UM 89 (flux), Lot No. 2269 (wire) -1230 (flux),
Dated 3/31/70.
12.
Proces Verbal De Recette De Produits De Soudre, S.A.F., Code No. 118 716 AS4, Spec
- PS 308/R, Designation UM 40 (fil) - UM 89 (flux), Lot No. 2269 (wire)-1180 (flux),
Dated 3/31/70.
I 13.
MM-SME-2925, "NRC Request for Information on Prairie Island Unit No.1 and 2 Reactor Vassel Materials Surveillance Program," S. E. Yanichko, 10/19/77.
1 Evaluation of PTS for Prairie Island Unit 1