ML20073L788
| ML20073L788 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 08/31/1994 |
| From: | Peter P WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20024J087 | List: |
| References | |
| WCAP-14054, NUDOCS 9410130289 | |
| Download: ML20073L788 (19) | |
Text
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WESTINGHOUSE CLASS 3 (Non-Proprietary)
WCAP-14054 EVALUATION OF PRESSURT7FD THERMAL SHOCK FOR BYRON UNIT 2 P. A. Peter i
August 1994 Work Performed Under Shop Order BPPP 108A I
l Prepared by Westinghouse Bectric Corporation for Commonwealth Edison Company l
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[,,n Approved by: /
R.!D Rish, Manager Metallurgical & NDE Analysis WESTINGHOUSE ELECTRIC CORPORATION Nuclear Technology Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 C 1994 Westinghouse Bectric Corporation All Rights Reserved 04f 0 { 50 LY(
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WESTINGHOUSE CLASS 3 (Non-Proprietary)'
WCAP-14054 l-l EVALUATION OF PRESSURT7FD THERMAL SHOCK -
FOR BYRON UNIT 2 P. A. Peter August 1994 Work Performed Under Shop Order BPPP-108A l
l Prepared by Westinghouse Bectric Corporation for Commonwealth Edison Company -
l Approved by: ik
,n mv-(
R.!D Rish, Manager j
Metallurgical & NDE Analysis i
l WESTINGHOUSE ELECTRIC CORPORATION l
Nuclear Technology Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 O 1994 Westinghouse Bectric Corporation All Rights Reserved l
kd/Cl301N
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PREFACE This report has been technicaUy reviewed and verified by:
j b///[hy-te J. M. Chicots
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TABLE OF CONTENTS LIST OF TABLES..................................................iii I
1 LIST OF FIGURES................................................. iii i
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1.0 INTRODUCTI ON.................................................
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l 2.0 PRES SUR T7FD THERMAL SHOCK...................................... 2 i
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l 3.0 METHOD FOR CALCULATION OF RTrn................................ 4 4.0 VERIFICATION OF PLANT-SPECIFIC MATERIAL PROPERTIES 5
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- NEUTRON FLUENCE VALUES 10 6.0 DETERMINATION OF RTen VALUES FOR ALL BELTLINE REGION i
MATERIALS 11 l
7.0 CONCLUSION
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REFERENCES........
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LIST OF TABLES
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i TABLE 1 BYRON UhTT 2 REACTOR VESSEL BELTLINE REGION MATERIAL PROPERTIES................................................. 7 TABLE 2 CALCULATION OF AVERAGE CU AND NI WEIGHT % USING ALL PREVIOUS BYRON UhTT 2 CHEMISTRY TEST RESULTS.............
8 TABLE 3 NEU' IRON EXPOSURE PROJECTIONS AT KEY LOCATIONS ON THE BYRON UhTT 2 PRESSURE VESSEL CLAD / BASE METAL INTERFACE FOR 4.634 AND 32 EFPY....................................... 10 TABLE 4 CALCULATION OF CHEMIS*IRY FACTORS USING BYRON UNIT 2 SURVEILLANCE CAPSULE DATA............
12 TABLE 5 RTm VALUES FOR BYRON UNIT 2 FOR 4.634 EFPY,................ 13 1
l TABLE 6 RTm VALUES FOR BYRON UNTT 2 FOR 32 EFPY......
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LIST OF FIGURES FIGURE 1 IDENTIFICATION AND LOCATION OF BELTLINE REGION MATERIALS F' R THE BYRON UhTT 2 REACTOR VESSEL............ 6 J
FIGURE 2 RTm VERSUS FLUENCE CURVES FOR BYRON UNIT 2 LIMITING MATERIAL - CIRCUMFERENTIAL WELD METAL WF447............. 14 iii J
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1.0 INTRODUCTION
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i 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.
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Such transients may challenge the integrity of a reactor vessel under the following conditions:
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severe overcooling of the inside surface of the vessel wall followed by high l
repressurization; significant degradation of vessel material toughness caused by radiation embrittlement; and the presence of a critical-size defect in the vessel wall.
l In 1985 the Nuclear Regulatory Commission (NRC) issued a formal ruling on PTS. It established j
screening criteria on pressurized water reactor (PWR) vessel embrittlement as measured by the nil-ductility reference temperature, termed RTen"1 RTrn 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 criteria were determined using conservative fracture mechanics analysis l
techniques. All PWR vessels in the United States have been required to evaluate vessel embrittlement in accordance with the criteria through end of license. The NRC recently amended its regulations for light water nuclear power plants to change the procedure for calculating radiation embrittlement. The revised PTS Rule was published 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 2W.
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l The purpose of this report is to determine the RTen values for the Byron Unit 2 reactor vessel to l
address the revised PTS Rule. Section 2 discusses the Rule and its requirements. Section 3 provides the methodology for calculating RTpn. Section 4 provides the reactor vessel beltline region material properties for the Byron Unit 2 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 references for the FTS evaluation follow in Sections 7 and 8, respectively.
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2.0 PRESSURT7Fn THERMAL SHOCK i
The FTS Rule requires that the FTS submittal be updated whenever there are changes in core loadings, surveillance measurements or other infonnation that indicates a significant change in projected RTm values. The Rule outlines regulations to address the potential for FIS 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 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 RTm 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 RTm 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 repon, 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 any assumptions regarding core loading patterns), and 2) copper and nickel content and fluence values used in the calculations for each beltline material. (If these values differ from those previously submitted to the NRC, justification must be provided.)
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'Ibe RTpn (measure 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 RTrn values for each weld, plate or forging in the reactor vessel beltline:
Equation 1:
RTrn = I + M + ARTrn Equation 2:
ARTrn = CF
- f "l* "80 t
All values of RTm must be verified to be bounding values for the specific reactor vessel. In doing this each plant should consider plant-specific information that f
could affect the level of embrittlement.
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.
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NRC approval for operation beyond the saeening criteria is required.
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e 3.0 hEITIOD FOR CALCULATION OF RT,3 In the IrrS Rule, the NRC Staff has selected a conservative and uniform method for determining plant-specific values of RTrn 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.
RTrn = I + M + ART,3, where ARTrn = CF
- FF I=
Initial reference temperature (RTm) in F of the unirradiated material M=
Margin to be added to cover uncertainties in the values of initial RTm, copper and nickel contents, fluence and calculational procedures in *F.
M = 66*F for welds and 48'F for base metal if generic values of I are used.
M = 56*F for welds and 34 F for base metal if measured values of I are used.
FF =
fluence factor = f * "*8 0, where f=
Neutron fluence (F>l.0 MeV at the clad / base metal interface), divided by 10" 2
n/cm CF =
Chemistry factor in 'F from the tabled 2j for welds and base metals (plates and forgings). If plant-specific surveillance data has been deemed credible per Regulatory Guide 1.99, Revision 2 and is significant, it may be considered in the calculation of the chemistry factor.
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4.0 VERIFICATION OF PLANT-SPECIFIC MATERIAL PROPERTIES Before performing the pressurized thermal shock evaluation, a review of the latest plant-specific l
material properties for the Byron Unit 2 vessel was performed. The beldine region is defined by the l
l r2 PTS Rule ) to be "the region of the reactor vessel (shell material meluding 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 irradiation 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 ByTon Unit 2 reactor vessel.
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 surveillance capsule testing f
program '1 The average copper and nickel values were calculated for each of the beltline region l
materials using all of the available material chemistry information. A summary of the pertinent chemical and mechanical properties of the beltline region forgings and weld materials of the ByTon Unit 2 reactor vessel are given in Table 1. All of the initial RTym values (I) are also presented in Table 1.
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90 Forging
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49D329/ 49C297/ l-1 W
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0 1800
~~
d5 5
Io 2700 CORE
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WF447 t
0 90 i
Forging I
49D330/ 49C298/ l-1 Jd5 5
0 0
E O
180 a
0 270 l
l FIGURE I IDENTIFICATION AND LOCATION OF BELTLINE REGION MATERIALS FOR THE BYRON UNIT 2 REACTOR VESSEll83 6
e TABLE 1 BYRON UNIT 2 REACTOR VESSEL BELTLINE REGION MATERIAL PROPERTIES j
Material Description Cu (%)
- Ni (%)
- I (*F)"3 ")
Iower Shell Forging MK 24-3 0.052 0.720
-20 Intet. Shell Forging MK 24-2 0.007 0.703
-20 Weld Metal 0.024 0.709 10 (a)
Initial RT, values were estimated per U.S. NRC Standard Review Plan. The initial RTm values for the forgings and welds are measured values.
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TABLE 2 CALCULATION OF AVERAGE CU AND NI WEIGHT % USING ALL PREVIOUS BYRON UNIT 2 CHEhBSTRY TEST RESULTS I
Inter. Shell 14wer Shell Forging 49D329/
Forging 49D330/
49C297 1 MK 24-2 49C298 -1 1 MK 24 3 Weld Metal **
Cu Ni Cu Ni Cu Ni Reference (wt.%)
(mt. %)
(wt.%)
(wt.%)
(wt.%)
(wt.%)
6 0.0l*
0.70*
0.0$*
0.73*
0.06*
0.62*
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0.07' O.65*
0.03*
0.65*
7 0.07 0.65 0.03 0.65 7
0.022 0.689 0.024 0.740 7
0.024 0.786 7
0.022 0.704 7
0.020 0.681 7
0.021 0.706 7
0.020 0.697 7
0.019 0.668 7
0.022 0.759 7
0.021 0.714 7
0.020 0.678 7
0.020 0,695 7
0.019 0.689 7
0.021 0.744 7
0.022 0.738 7
0.022 0.771 8
0.01 0.70 8
0.006 0.70 8
0.006 0.71 9
0.059 0.62 10 0.05 0.73 10 0.05 0.73 10 0.05 0.75 11 0.067 0.772 0.024 0.705 11 0.023 0.706
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Inter. Shell Lower Shell Forging 49D329/
Forging 49D330/
49C297 1 MK 24 2 49C298 -11 MK 24 3 Weld Metal **
Cu Ni Cu Ni Cu Ni Reference (wt.%)
(wt.%)
(wt.%)
(wt. %)
(wt.%)
(wt.%)
11 0.023 0.698 11 0.024 0.696 11 0.023 0.711 11 0.024 0.708 11 0.024 0.716 11 0.024 0.715 11 0.024 0.707 11 0.024 0.720 11 0.024 0.717 11 0.024 0.711 11 0.024 0.706 11 0.024 0.707 11 0.025 0.717 Average 0.007 0.703 0.052 0.720 0.024 0.709 Not used in Average calculanon; reported only for completeness.
The surveillance weld is identical to that used in the core region girth seam. The weld sire is type Linde MnMoNi (Low Cu-P), Heat Number 442002, with a Linde 80 type flux. let Number 8064
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5.0 NEUTRON FLUENCE VALUES The calculated fast neutron fluence (E>1.0 MeV) at the inner surface of the Byron Unit 2 reactor vessel is shown in Table 3. These values were projected using the results of the Capsule W radiation surveillance program"1 The RTrn calculations were performed using the peak fluence value, which occurs at the 25' azimuth in the Byron Unit 2 reactor vessel.
TABLE 3 NEUTRON EXPOSURE PROJECTIONS
- AT KEY LOCATIONS ON TIE BYRON UhTT 2 PRESSURE VESSEL CLAD / BASE hETAL INTERFACE FOR 4.634 AhT 32 EFPYH1 l
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EFPY 0
15' 25' 35' 45*
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l 4.634 1.959 x 10" 2.867 x 10" 3.174 x 10" 2.566 x 10" 2.934 x 10" 32 1.353 x 10" 1.979 x 10" 2.192 x 10" 1.772 x 10" 2.026 x 10*
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- Fluence in n/cm (E>1.0 MeV)
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6.0 DETERhi1 NATION OF RTrn VALUES FOR ALL BELTLINE REGION MATERIALS Using the prescribed PTS Rule methodology, RTrn values were generated for all beltline region materials of the Byron Unit 2 reactor vessel for fluence values at the present time (4.634 EFPY per Capsule W analysis) and end oflicense (32 EFPY). The PTS Rule requires that each plant assess the RTrn values based on plant specific surveillance capsule data whenever:
Plant specific surveillance data has been deemed credible as defined in Regulatory Guide 1.99, Revision 2, and j
l RTrn values change significantly. (Changes to RTrn values are considered significant if the value determined with RTrn equations (1) and (2), or that using capsule data, or j
toth, exceed the screening criteria prior to the expiration of the operating license, 1
including any renewed term,if applicable, for the plant.)
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Although the RT,n value changes are not significant for Byron Unit 2, plant specific surveillance capsule data for lower shell forging MK 24-3 and the weld metal is provided because of the following reasons:
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There have been two 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 forgings and circumferential weld materials.
l The chemistry factors for lower shell forging MK 24-3 and the weld metal were calculated using the surveillance capsule data as shown in Table 4. The chemistry factors were also calculated using Tables 1 and 2 from 10 CFR 50.61m. Tables 5 and 6 provide a summary of the RTrn values for all beltline region materials for 4.634 EFPY and 32 EFPY, respectively, using the FTS Rule.
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TABLE 4 CAlfULATION OF CHEhDSTRY FACTORS USING BYRON UhTT 2 SURVEILLANCE CAPSULE DATA"1 2
Material Capsule Fluence FF ARTm7 FF*ARTay FP lower Shell U
3.996 x 10" 0.746 0
0 0.556 Forging MK 24-3 (Tang.)
W
' 1.2110 x 10" 1.053 5
5.267 1.110 lower Shell U
3.996 x 10" 0.746 25 18.643 0.556 f
W 1.2110 x 10" 1.053 40 42.136 1.110 Sum:
66.046 3.332 l
Chemistry Factor = 66.046 + 3.332 = 19.8
^
Weld Metal U
3.996 x 10" 0.746
-0 0
0.556 4
W 1.2110 x 10" 1.053 30 31.602 1.110 Sum:
31.602 1.666 Chemistry Factor = 31.602 + 1.666 = 19.0 a
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TABLE 5 4
RTm VALUES FOR BYRON UNIT 2 FOR 4.634 EFPY
~
M RTm Screening Material
('F)
(*F)
(*F)
('F)
Criteria ('F) j Imwer Shell Forging MK 24-3 32.2 0.6849
-20 34 36.1 270 Using surv. capsule data" 19.8 0.6849
-20 34 27.6 270
)
Inter. Shell Forging MK 24-2 20.0 0.6849
-20 34 27.7 270 Cire. Weld Metal WF447 32.6 0.6849 10 56 88.3 300 1
j Using sury. capsule data" 19.0 0.6849 10 56 79.0 300 a
FF (Fluence factor) based upon peak inner surface neutron fluence of 3.174 x 10" n/cm 2
(E>l.0 MeV)").
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Numbers were calculated using a chemistry factor (CF) based on surveillance capsule data per Regulatory Guide 1.99. Revision 2, Position 2.
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TABLE 6 RTm VALUES FOR BYPCF UNIT 2 FOR 32 EFPY
..,_,m CF Ps" I
M-RTm Screening 4
Material
(*F)
(*F)
('F)
(*F)
Criteria (*F) j e
Lower ShcIl Forging MK 24-3 32.2 1.213
-20 34 53,1 270 Using sury. capsule data" 19.8 1.213
-20 34 38.0 270 Inter. Shell Forging MK 24-2 20.0 1.213
-20 34 38.3 270 Circ. Weld Metal WF447 32.6 1.213 10 56 105.5 300 Using surv. capsule data" 19.0 1.213 10 56 89.0 300 e
2 FF (Fluence factor) based upon peak inner surface neutron fluence of 2.192 x 10" n/cm (E>1.0 MeV)"3 Numbers were calculated using a chemistry factor (CF) based on surveillance capsule data j
per Regulatory Guide 1.99, Revision 2, Position 2.
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7.0 CONCLUSION
S As shown in Tables 5 and 6 all RTpn values remain below the NRC screening values for PTS using fluence values for the present time (4.634 EFPY) and projected fluence values for the end oflicense (32 EFPY). The plot of RTrn values versus fluence shown in Figure 2 illustrates the available margin for the most limiting material in the Byron Unit 2 reactor vessel beltline region, Circumferential Weld Metal WF447. The surveillance capsule analyses results are also shows.
350 EEN CRUERM 300 250
[200 en g150 100
- ........e
-- --- -~........-- -
- ~~~~'''
____e--
---____g_-
50
1E+18 2E + 18 3E + 18 SE + 18 1E+19 2E + 19 3E + 19 SE+19 1E + 20 FLUENCE (neutrons /cm )
2 L
AL WELD METAL USING SURV. CAPSULE DATA FIGURE 2 RT,,3 VERSUS FLUENCE CURVES FOR BYRON UNIT 2 LIMITING MATERLAL - CIRCUMFERENTIAL %TLD MEAL MP447 14 l
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8.0 REFERENCES
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[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 Pressurized Thermal Shock Events," May 15,1991. (PTS Rule)
[3]
Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials,"
U.S. Nuclear Regulatory Commission, May 1988.
[4]
WCAP-13880, " Analysis of Capsule W from Commonwealth Edison Company Byron Unit 2 Reactor Vessel Radiation Surveillance Program", M. J. Malone, May 19M.
[5]
WCAP-13069, " Evaluation of Pressurized Thermal Shock for Byron Units 1 & 2", M. A.
Ramirez, et al., September 1991.
[6]
WCAP-10398, " Commonwealth Edison Company Byron Unit No. 2 Reactor Vessel Radiation Suneillance Program", L. R. Singer, December 1983.
[7]
WCAP-12431 " Analysis of Capsule U from the Commonwealth Edison Company Byron Unit 2 Reactor Vessel Radiation Surveillance Program". E. Terek, et al., October 1989.
[8]
Japan Steel Works, Ltd. Material Test Results, Contract # 640-0012, Upper Shell Forging 49D329/49C297 1 MK 24-2.
[9]
Babcock & Wilcox, " Record of Filler Wire Qualification Test" Test No. WF-447,8/8/74.
[10]
Japan Steel Works, Ltd. Material Test Results, Contract # 640-0012, Lower Shell Forging 49D330/49C298 1 MK 24 3.
[11]
AR #15344, " Low Alloy Steel Analysis, Commonwealth Edison Company Byron Nuclear Plant, Unit 2", L. Kardos, 03/31/94.
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