ML20207H773
| ML20207H773 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 11/30/1998 |
| From: | Christopher Boyd, Laubham T, Trombola D WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20207H753 | List: |
| References | |
| WCAP-15125, WCAP-15125-R, WCAP-15125-R00, NUDOCS 9907210207 | |
| Download: ML20207H773 (24) | |
Text
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Revismn 0 i
Evaluation of Pressurized Thermal Shock for Byron Unit 1
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Rension 0 Evaluation of Pressurized Thermal Shock for Byron
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WESTINGHOUSE NON PROPRJETARY CLASS 3 WCAP-15125 Evaluation of Pressurized Thermal Shock for Byron Unit I l
I T.J. Laubham Neveraber 1998 Work Performed Under Shop Order C8QP-108 i
Prepared by the Westinghouse Electric Company for the Commonwealth Edison Company Approved:
C. H. Boyd, Manager Equipment & Materials Technology
/
Approved:
j D. M. Trombola, Manager Mechanical Systems Integration Westinghouse Electric Company Energy Systems P.O. Box 355 Pittsburgh, PA 15230-0355 O1998 Westinghouse Electric Company All Rights Reserved
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111 TABLE OF CONTENTS
.iv LIST OF TABLES.............................
Y LIST OF FIGURES...........
... ~.....
PREFACE....................................................................
... vi EXECUTIVE S EN1 MARY..................................................
............ vii 1
INTRODUCTION....................................................
..... 1-1 2
PRESSURIZED THERMAL SHOCK RULE.,.........................
............2-1 i
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3 METHOD FOR CALCULATION OF RTm..................................
............3-1 i
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. 4-1 l
~4 VERIFICATION OF PLANT SPECIFIC MATERIAL PROPERTIES..........
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5
. NEUTRON FLUENCE VALUES.......................................
.......5-1 L
6 DETERMINATION OF RTm VALUES FOR ALL BELTLINE REGION MATERIALS...... 6-1 I
7 C ON C LUS I ON................................................................................. 7-1 i
8 REFEREN C ES......................................................................................... 8-1 Revision 0
j iv j
LIST OFTABLES l
.43
' Byron Unit 1 Reactor Vessel Beltline Unitradiated Material Properties.
Table 1 Fluence (E > 1.0 MeV) on Pressure Vessel Clad / Base Interface for Bvron Table 2
. 5-1 Unit 1 at 32 (EOL) and 48 (Life Extension) EFPY..
.6-2 Interpolation of Chemistry Factors Using Tables 1 and 2 of 10 CFR 50.61.
Table 3 Table 4 Calculation of Chemistry Factors using Surveillance Capsule Data Per
.6-3 Regulatory Gnide 1.99, Revision 2, Position 2.1.....
Table 5 RTm Calculation for Byron Unit 1 Beltline Region Materials
.6-5 at EOL(32 EFPY)........
Table 6 RTm Calculation for Byron Unit 1 Beltline Region Materials at Life
... 6-6 Extension (48 EFPY).....
Revision 0 l
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v LIST OF FIGURES Figure 1 -
Identification and Location of Beltline Region Materials for the Byron Unit 1 Reactor Vessel..,,
... 4-2 4
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Revision 0 i
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- PREFACE -
Thn report has been technicall.y reviewed and verified by:
Reviewer:
Ed Terek t
,k s
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i Revision 0 c
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EXECUTIVE
SUMMARY
The purpose of this report is to determine the RTnsvalues for the Byron Unit I reactor ves:.el beltline based upon the results of the Surveillance Capsule W evaluation. The conclusion of this report is that all the beltline materials in the Byron Unit I reactor vessel have RTns values below the screening criteria of 270*F for plates, forgings or longitudinal welds and 300'F for circumferential welds at EOL (32 EFPY) and life extension (48 EFPY). Specifically, the intermediate shell forging was the most limiting material with 32 and 48 EFPY PTS values of 110 F and 113*F respectively.
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i Revision U
1-1
~ NTRODUCTION-I 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 concern arises if one of these transients acts on the beltline region of a reactor vessel where a reduced fracture resistance exists because of neutron irradiation. Such an event may produce a flaw or cause the propagation of a flaw postulated to exist near the inner wall surface. thereby potentially affecting the integrity of the vessel.
. The purpose of this repon is to determme the RTm values for the Byron Unit I reactor vessel using the
- I results of the surveillance Capsule W evaluation. Section 2.0 discusses the PTS Rule and its requirements.
Section 3.0 provides the ekulalogy for calculating RTm. Section 4.0 provides the reactor vessel beltline
- region matenal properties for the Byron Unit I reactor vessel.: The neutron fluence values used in this W
analysis are presented in Section 5.0 and were obtained from Section 6 ofWCAP-15123. The results of the RTm calculations are presented in Section 6.0. De conclusion and references for the PTS evaluation follow in Sections 7.0 and 8.0, respectively.-
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Stroduction '
Revision 0
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- PRESSURIZED THERMAL SHOCK RULE J
The Nuclear Regulatory Commission (NRC) amended its regulations for light-water-cooled nuclear power plants to chrify several items related to the fracture toughness requirements for reactor pressure -
. vessels, including pr:ssurized thermal shock requirements. The latest revision of the PTS Rule.10 CFR m
l Part 50.61, was publisied la 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 makes binding by rule, the method for determining the reference temperature, RT., including treatment of the unirradiated RT value, the margin j
tenn, and the explicit definition of" credible" surveillance data, which is also described in Regulatory Guide 1.99, Revision 2m, 2.
The rule is restructured to improve clarity, with the requirements section giving only the requirements for the value for the reference temperature for end oflicense (EOL) fluence, RTm-3.
Thermal annealing is identified as a method for mitigating the effects of neutron irradiation, thereby reducing RTm.
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The PTS Rule requirements consist of the following:
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.o-For each pressurized water nuclear power reactor for which an operating license has been issued, the licensee shall have projected values of RTm, accepted by the NRC, for each reactor vessel beltline material for the EOL fluence of the material.
L The assessment ofRTm must use the calculation procedures given in the PTS Rule, and must o
specify the bases for the projected value of RTm for each beltline material. The report must specify the copper and nickel contents and the fluence values used in the calculation for each beltline material.
This assessment must be updated whenever there is a significant change in projected values of o
RTm or upon the request for a change in the expiration date for operation of the facility.
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- Changes to RTm 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 RTm screening criterion values for the beltline region are:
o 270*F for plates, forgings and axial weld materials, and 300'F for circumferential weld materials.
Pressurized Th.ww.al Shock Rule Revision 0
3-1 I
3 METHOD FOR CALCULATION OF RTers RTrrs must be calculated for each vessel beltline material using a fluence value, f, which is the EOL fluence for the material. Equation 1 must be used to calculate values of RTm1 or each weld and plate or j
f forging in the reactor vessel beltline.
{
R&r = Rhrcu)+ M + ARhr (1)
)
)
- Where, Reference Temperature for a reactor vessel material in the pre-service or unirradiated RTercu)
=
condition Margin to be added to account for uncertainties in the values of RTercu>, copper and 1
M
=
nickel contents, fluence and calculational procedures. M is evaluated from Equation 2 M=Jcry2,32 (2) ou is the standard deviation for RTercu).
0*F when RTercu)is a measured value, ou
=
17'F when RTercu)is a generic value.
=
ou a6 is the standard deviation for RTer.
For plates and forgings:
17'F when surveillance capsule data is not used.
=
c4 8.5'F when surveillance capsule data is used.
e4
=
For welds:
28'F when surveillance capsule data is not used.
ca
=
ca 14'F when surveillance capsule data is used.
=
c3not to exceed one half of ARTer ARTer is the mean value of the transition temperature shift, or change in ARTer, due to irradition, and must be calculated using Equation 3.
AR&r = (CF)
- f("-"'"' A (3)
~
Method For Calcualtion of RTm Revision 0
3-2 CF (*F)is the chemistry factor, which is a function of copper and nickel content. CF is determined from Tables 1 and 2 of the PTS Rule (10 CFR 50.61). Surveillance data deemed credible must be used to
' determine a material-specific value of CF. A material specific value of CF is determined in Equation 5.
The EOL Fluence (f) is the higher of the best estimate or calculated neutron fluence, in units of 10" n/cm:
(E > 1.0 MeV), at the clad-base-metal interface on the inside surface of the vessel at the location w here the material in question receives the highest fluence. The EOL Duence is used in calculating RT,n.
Equation 4 mu'st be used for determining RTns using Equation 3 with EOL fluence values for determining RTns.
RTm = RTxorw>+ M+ ARTm (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 information includes but is not limited to the reactor vessel operating temperature and any related surveillance program results. Results from the plant specific surveillance program must be integrated into the RT, estimate if the plant-specific surveillance data has been deemed credible.
A material-specific value of CF for surveillance materials is determined from Equation 5.
7 " g e p oas-o.ioios h)
(5) gjo.56-o.2cm3) in Equation 5, "Aj" is the measured value of ART, and "f" is the fluence for each surveillance data i
point. If there is clear evidence that the copper and nickel conient of the surveillance weld differs from the vessel v> eld, i.e., differs from the average for the weld wire heat number associated with the vessel weld and the surveillance weld, the measured values of RT, must be adjusted for differences in copper and nickel content by multiplying them by the ratio of the chemistry factor for the vessel material to that for the surveillance weld.
Method For Calcualtion of RTns -
Revision 0
r 4-1 4-
' VERIFICATION OF PLANT SPECIFIC MATERIAL PROPERTIES Before performing the pressurized thermal shock evaluation, a review of the latest plant-speci6c material
. properties for the Byron Unit i vessel was performed. The beltline region of a 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" Figure 1 identi0es and indicates the location of all beltline region materials for the Byron Unit I reactor vessel.
The best estimate copper and nickel contents of the beltline materials were obtained from WCAP-14824 Revision 2I'l and from the testing of Charpy specimens in Capsule Wl3. The best estimate copper and nickel content is documented in Table I herein. The average values were calculated using all of the available material chemistry information. Initial RTor values for Byron Unit 1 reactor vessel bMiline material properties are also shown in Table 1.
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i Verification of Plant Specific Mage,i,; p,,p, Revision 0
4-2 N
WF-501 (Heat # 442011)
T>$
l Intermediate Shell Forging SP-5933 i
a Core d
WF-336 (Heat # 442002)
=:
E b
L wer Shell Forging SP-5951 0
WF-472 (Heat # 31401)
Figure 1:
Identi5 cation and Location of Beltline Region Materials for the Byron Unit 1 Reactor Vessel Verification of Plant Specific Material Properties Revision 0
4-3 Table 1 Byron Unit 1 Reactor Vessel Beltline Unirradiated Material Properties Material Description Cu(%)
Ni(%)
Initial RTer"'
0.74 60 Closure Head Flange 124K358VA]
0.73 10 Vessel Flange 123J219 val Nozzle Shell Forging 123J218*)
0.05 0.72 30 Intermediate Shell Forging SP-5933 0.04 0.74 40 Lower Shell Forging 5P-5951 0.04 0.64 10 Intermediate to Lower Shell Forging Cire.
0.04 0.63
-30 Weld Scam WF-336 (Heat # 442002)
- Nozzle Shell to Intermediate Shell Forging 0.03 0.67 10 Cire. Weld Seam WF-501 (Heat # 442011)*)
Byron Unit 1 Surveillance Program 0.02 0.69 Weld Metal (Heat # 442002)
Byron Unit 2 Surveillance Program 0.02 0.71 Weld Metal (Heat # 442002)
Braidwood Units 1 & 2 Surveillance Program 0.03 0.67, 0.71 Weld Metals (Heat # 442011)
Notes.
(a) TheinitialRTmrrvalues for the plates and welds are based on measured data per reference 5 and 6.
(b) Best Estimate Cu% / Ni% and Initial RTun Per Reference 5 and/or 6.
I Verification of Plant Specific Material Properties Revision 0 i
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NEUTRON' FLUENCE VALUES
' The calculated fast neutron fluence (E > 1.0 MeV) values at the inner surface of the ByTon Unit I reactor j
vessel for 32 and 48 EFPY are shown in Table 2. These values were projected using the results of the Capsule W radiation analysis. See Section 6.0 of the Capsule W dosimetry analysis report. WCAP-15123"!(48 EFPY values were obtained from interpolation of fluences between 32 EFPY and 54 EFPY).
TABLE 2 Fluence (E > 1.0 MeV) on the Pressure Vessel Clad / Base Interface for Byron Unit 1
)
at 32 (EOL) and 48 (Life Extension) EFPY
' Material Location 32 EFPY Fluence 48 EFPY Fluence i
Nozzle Shell Forging 123J218
'45' 5.83 x 10" n/cm 8.70 x 10" n/cm' Intermediate Shell Forging SP-5933 45' l.95 x 10" n/cm2 2.91 x 10"n/cm2 Lower Shell Forging SP-5951 45' l.95 x 10"n/cm:
2.91 x 10" n/cm Intermediate to Lower Shell Forging 45'
' l.95 x 10" n/cm2 2.9.1 x 10" n/cm Cire. Weld Seam WF-336 (Heat #
442002) 8.70 x 10" n/cm
'l Nozzle Shell to Intermediate Shell 45' 5.83 x 10" n/cm2 Forging Cire. Weld Scam WF.501 (Heat # 442011) 1 I
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Neutron Fluence Values Revision 0
6-1 DETERMINATION OF RTrrs VALUES FOR ALL BELTLINE 6
REGION MATERIALS Using the prescribed PTS Rule methodology, RTns values were generated for all beltline region m of the Byron Unit I reactor vessel for fluence values at the EOL (32 EFPY) and life extension (48 EFPY).
Each plant shall assess the RTn, values based on plant-specific surveillance capsule data. For Byron
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Unit 1, the related surveillance program results have been included in this PTS evaluation.
(See Reference 5 for the credibility evaluation of the Byron Unit I surveillance data.)
As presented in Table 3, chemistry factor values for Byron Unit I based on average copper and nickel weight percent were calculated using Tables 1 and 2 from 10 CFR 50.61IO. Additionally, chemistry factor values based on surveillance capsule data from Byron Units 1 and 2, and Braidwood Units 1 and 2 are calculated in Table 4. Tables 5 and 6 contain the RTns calculations for all beltline region materials at EOL (32 EFPY) and life extension (48 EFPY).
Determination of RTns Values For All Beltline Region Materials Revision 0
6-2 TABLE 3
' laterpolation of Chemistry Factors Using Tables 1 and 2 of 10 CFR Part 50.6i Material Ni,wt %-
Chemistry Factor, "F Intermediate Shell Forcine 5P-5933 0.74 26.0 F Given Cu wt% = 0.04 Lower Shell Forninn SP-5951 0.64 26.0 F Given Cu wt% = 0.04 Nozzle Shell Forninn 123J218 0.72 31.0*F Given Cu wtv. = 0.05 joigrmadi='a to Lower Shell Cirs,_Wsid WF-336 0.63 54.0 F Given Cu ws% = 0.04 Nozzle Shell to Intermediate Shell Cire. Weld WF-501 0.67 41.0'F Given Cu wt% = 0.03 Byron Unit I and 2 Surveillance Prearom Weld Metal 0.69, 0.71 27.0'F Given Cu wt% = 0.02.
Braidwood Unit I and 2 Surveillance Program Weld Metal 0.67, 0.71 41.0'F j
Given Cu wt% = 0.03
[ -
Determmation of RTm Values For All Beltline Region Materials Revision 0
6-3 TABLE 4 Calculation of Chemistry Factors using Surveillance Capsule Data Per Regulatory Guide 1.99. Revision 2. Position 2.1 Material Capsule Capsule fa' FF*'
ARTsn "'
FF*ARTsm FF Intermediate Shell U
0.404 0.748 28.55 21.36 0.560 Forging 5P 5933 X
1.57 1.124 9.82 11.04 1.263 (Tr.ngential)
W 2.43 1.239 49.20 60.96 1.535 Intermediate Shell U
0.404 0.748 18.52 13 85 0.560 Forging SP 5933 X
1.57 1.124 53.03 59.61 1.263 (Axial)
W 2.43 1.239 29.34 36.35 1.535 SUM' 203.17 6.716 CFsp.3,33 = Z(FF
0.404 0.748 11.22 8.39 0.560 (5.61)'d' Material X
1.57 1.124 80.22 90.17 1.263 (40.11)'d' (Heat # 442002)
W 2.43 1.239 102.68 127.22 1.535 (5) 34)^
Byron Unit 2 Surv. Weld U
0.405 0.749 0'd' O
0.561 Matenal W
l.27 1.067 60.0 (MO)'d' 64.02 1.!38 (Heat # 442002)
SUM:
289.80 5.057 CFs= w.ia,u2002 = I(FF
- RTm) + I( FF ) = (289.80) + (5.057) = 57.3*F Notes:
(a)
Byron Unit 1 & 2 capsule fluences were updated as a part of the capsule W dosimetry analysis results Ref. 5.
2 (x 10" n/cm, E > 1.0 MeV).
(b)
FF = fluence factor = (":'** *
(c)
ARTmvalues are the measured 30 ft-lb shift values taken from Ref. 5.
(d)
The Byron 1 & 2 surveillance weld meta! ARTmvalues have been adjusted by a ratio factor of 2.0.
No temperature adjustment is required.
Determination of RTrrs Values For All n ltline Region Materials Revision 0 e
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I 6-4 TABLE 4 - Continued Calculation of Chemistry Factors using Surveillance Capsule Data Per Regulatory Guide 1.99, Revision 2, Position 2.I Material Capsule Capsule (*
FF*
ARTsm "
FF*ARTsn FF i
T l
Weld Heat 442011. WF-501 U
0.3814 0.733 10 7.3 0.538 Using Braidwood 1 Sury.
X 1.144 1.038 25 26.0 1.077 Data Weld Heat 442011, WF-501 U
0.3933 0.741 0
0 0.550 Using Braidwood 2 X
1.126 1.033 20 20.7 1.067 Sury. Data SUM:
54.0 3.232 CFw w.w on = Z(FF
- RTmn) + I( FF ) = (54.0) + (3.232) = 16.7'I'M Notes.
(a)
Braidwood Units 1 & 2 fluences were taken from WCAP-14824 R.2 (Ref. 4)(x 10 n/cm, E > 1.0 MeV).
2 (b)
FF = fluence factor = ta23 oi wo,
(c)
ARTmyrvalues are the rneasured 30 ft-lb shift values taken from Appendix B of Ref. 4.
(d)
The Braidwood I & 2 surveillance weld metal ARTmyrvalues do not require a ratio factor or temperature adjustment.
(c)
Per Reference 7. Comed reported to the NRC a chemistry factor of 17.0. The difference is a result of rounding and is negligible when used in the calculation of the PTS values.
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l Determination of RTrrs Values For All Beltline Region Materials Revision 0
F
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6-5 TABLE 5 RTm Calculation for Byron Unit i Beltline Region Materials at EOL (32 EFPY)
Margin RTsurcel
RTm"'
(n/ctn'.
(*F)
('F)
(*F)
(*F)
(*F)
E>1.0 MeV)
Intermediate Shell Forging 1.95 x 10
l.18 26.0 30.7 30.7 40 101 l'itermediate Shell Forging 1.95 x 10
1,18 30.3 35.8 34 40 110
-+ using S/C Data"'
Lower shell Forging 1.95 x 10
l.18 26.0 30.7 30.7 10 71 Inter. to Lower Shell Cire. Weld 1.95 x 10
l.18 54.0 63.7 56
-30 90 Metal Inter. to Lower Shell Cire. Weld 1.95 x 10
l.18 57.3 67.6 28
-30 66 Metal
-+ using S/C Data Nozzle Shell Forging 5.83 x 10'8 0.849 31.0 26.3 26.3 30 83 N:zzle Shell toInter. Shell Cire.
5.83 x 10
O.849 41.0 34.8 34.8 10 80 Wald Metal N:zzle Shell to Inter. Shell Cire.
5.83 x 10
O.849 16.7 14.2 14.2 10 38 W;ld Metal
-+ using S/C Data Notes:
(a)
Initial RTmvalues are measured values (See Table 1)
(b)
RTm = RTatto + ARTm + Margin ('F)
(c)
ARTm = CF
- FF (d)
Surveillance data is considered not credible, however, since the chemistry factor (CF) from the Reg. Guide Tables (Pos.1.1) is considered not conservative (i.e. CF via Pos. 2.1 > CF via Pos.1.1), then this data will be used to determine PTS with a full c4 margin term.
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Deternunation of RTm Values For All Beltline Region Materials Revision 0
6-6 TABLE 6 RTm Calcula+jon for Byron Unit 1 Beltline Region Materials at Life Extension (48 EFPY)
Margin RTsum"'
RTm" (n/cm'.
(*F)
(*F)
(*F)
('F)
(*F)
F21.0 MeV)
Intermediate Shell Forging 2.91 x 10
1.28 26.0 33.3 33.3 40 107 Intermediate Shell Forging 2.91 x 10" 1.28 30.3 38.8 34 40 113
-+ using S/C Data
- Lower shell Forging 2.91 x 10" 1.28 16.0 33.3 33.3 10 77 Inter. to Lower Shell Cire. Weld 2.91 x 10" 1.28 54.0 69.1 56
-30 95 Metal Inter. to Lower Shell Cire. Weld 2.91 x 10" 1.28 57.3 73.3 28 30 71 Metal
- using S/C Data Nozzle Shell Forging 8.70 x 10" 0.%1 31.0 29.8 29.8 30 90 Nozzle Shell to Inter. Shell Cire.
8.70 x 10" 0.961 41.0 39.4 39.4 10 89 j
Weld Metal Nozzle Shell to Inter. Shell Cire.
8.70 x 10" 0.961 16.7 16.0 76.0 10 42 Weld Metal
-+ using S/C Data Notes.
(a)
Initial RTervalues are measured values (See Table 1)
(b)
RTm = RTemn + ARTm + Margin ('F)
(c)
ARTm = CF
- FF (d)
Surveillance data is ettisidered not credible, however, since the chemistry factor (CF) from the Reg. Guide Tables (Pos.1.1) is conddered not conservative (i.e. CF via Pos. 2.1 > CF via Pos.1.1), then this data will be used to deternune FTS xvith a full c margin term.
a Deternunation of RTm Values For All Beltline Region Materials Revision 0
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7-1 i
L 7.
CONCLUSIONS l-As shown a Tables 5 and 6. all of the beltline region materials in the Byron Unit I reactor vessel have EOL (32 EFPY) RTnsand Life Extensid. (48 EFPY) RTns values below the screening criteria values of 270 F for plates, forgings and locgitudinal welds and 300 F for circumferential welds. Specifically, the l-intermediate shell forging was the most limiting material with 32 and 48 EFPY PTS values of 110 F and
. I13'F respectively.
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REFERENCES l
1.
10 CFR Part'50.61, " Fracture Toughness Requirements For Protection Against Pressurized Thermal i
Shock Events", Federal Register, Volume 60. No. 243, dated December 1911995.
l l
2 Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Matenals". U.S.
~ Nuclear Regulatory Commission, May,1988.
3 WCAP-9517, " Commonwealth Edison Co. Byron Station Unit No. I Reactor Vessel Radiation Surveillance Program", J. A. Davidson, July 1979.
4 WCAP-14824, Revision 2, " Byron Unit 1 Heatup and Cooldown Limit Curves for Normal Operation and Surveillance weld Metal Integration for Byron & Braidwood", T. J.Laubham, et. al..
I November,1997.
5-WCAP-15123, " Analysis of Capsule W from the Commonwealth Edison Company Byron Unit i Reactor Vessel Radiation Surveillance Program", T. J. Laubham, et al., November 1998.
6 Letters CAE-97-231, CCE-97-314, " Comed Response to NRC Question to WCAP-14940, WCAP 14970 and 14824 Rev. 2", From C.S. Hauser to Mr. Guy DeBoo (of Comed), Dated l
January 6,1998.
7.
Comed Le:ter to U.S. Regulatory Commission, " Response to Additional Information Regarding Reactor Pressure Vessel", From R.M. Krich, Dated September 3,1998.
i f
i References Revision 0
Attachment F Byron Station WCAP-15177, Rev. O," Evaluation of Pressurized Thermal Shock for Byron Unit 2" p:WbytrsWOO95 doc
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