ML20207H777
| ML20207H777 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 06/30/1999 |
| From: | Christopher Boyd, Laubham T, Trombola D WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20207H753 | List: |
| References | |
| WCAP-15177, WCAP-15177-R, WCAP-15177-R00, NUDOCS 9907210208 | |
| Download: ML20207H777 (23) | |
Text
{{#Wiki_filter:---
, J Westingbouse Non-Proprietary Class 3 . -,
i '
+ + + + + + + + + + + "
Ll ,T Evaluation of Pressurized E 1 Thermal Shock for Byron .
.J Unit 2 '
i
+
.}
i l l I 1 i 1 t
.- . i s -{
Westinghouse Electric Company LLC W -
~
9907210208 990712 PDR ADOCR 05000454 P PDR (;i .
r '6
- Westinghouse Non-Proprietary Class 3 I
a
+ + + + + + + + + + + -
i m. - "..~ m l i Evaluation of Pressurized - - Thermal Shock for Byron Unit 2 l i
~!
l l l i l \ l
. ~w [h.
l V ( , i
- ~
l Westinghouse Electric Company LLC W' - l i l 9907210208 990712 PDR ADOCK 05000454 (-, P PM ,
WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15177 Evaluation of Pressurized Thermal Shock for Byron Unit 2 T.J. Laubham June 1999 4 Work Performed Under Shop Order C8QP-108A 1 Prepared i,3 ?in Westinghouse Electric Company LLC for the Commonwealth Edison Company Approved: i-C. H. Boyd, ManagM Equipment & Materials Technology iqvroved: ibm D. M. Trombola, fdanager Mechanie.al Systems Integration l Westinghouse Electric Company LLC I Energy Systems P.O. Box 355 Pittsburgh, PA 15230-0355 01999 Westinghouse Bectric Company LLC All Rights Reserved
ni TABLE OF CONTENTS l LIST OFTABLES. . . .. . .. .. .. . .iv OST OF FIGURES.. . . . . . .. . . v PREFACE , . . . . . . . . . . . . . . . . . . . . . . . . . . . . vi EXECUTIVE
SUMMARY
... . . . . . . . .. . .. .. . .sii i INTRODUCTION . .......... . ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 2 PRESSURIZED THERMAL SHOCK RULE... . . . . . . . . . . 2-1
.. . . . . . . l l
3 METHOD FOR CALCULATION OF RTm.. . .. .. . . . . . . . 3-1 '
- 4. VERIFICATION OF PLANT SPECIFIC MATERIAL PROPERTIES.. . . . . . . . . . . 4-1 5 NElJrRON FLUENCE VALUES.... ..... . ....... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 6 DETERMINATION OF RTm VALUES FOR ALL BELTLINE REGION MATERIALS. .6-1 7 CONCLUSION ..... .. ... .. .... . ....... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7- 1 8 REFE REN CES . . . . . . .. .. .. . . . . . . . . . . .... . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . .8-1 Revision 0
iv LIST OFTABLES Table 1 Byron Unit 2 Reactor Vessel Beltline Unirradiated Material Properties. .43 1 1 i i Table 2 Fluence (E > 1.0 MeV) on Pressure Vessel Clad / Base Interface for Byron Unit 2 at 32 (EOL) and 48 (Life Extension) EFPY.. . .51 f Table 3 Interpolation of Chemistry Factors Using Tables 1 and 2 of 10 CFR 50.61.. . . . 6-2 i Table 4 Calculation of Chemistry Factors using Surveillance Capsule Data Per j Regulatory Guide 1.99, Revision 2, Position 2.1. .. . . .6-3 ) Table 5 RTm Calculation for Byron Unit 2 Beltline Region Materials at EOL (32 EFPY) ........ . ... . . . . . . ... . . ... .6-5 l
}
l Table 6 RTm Calculation for Byron Unit 2 Beltline Region Materials at Life Extension (48 EFPY) . . . . . . . . . . . . . . . . . . . . . . . . .. .6-6 l l Revision 0
v LIST OF FIGURES
- Figure 1 Identification and Location of Beltline Region Materials for the Byron Unit 2 Reactor Vessel . . . .. .. . . . . . . . . . . . .. .. .. . ..4-2 l
l l l 4
)
l Revision 0 1 1
r-- . , c. vi PREFACE i This report has been technically reviewed and verified by: Reviewer: Ed Terek !M <
)
l i i I
-)
l l l 1 Revision 0
vii l EXECUTIVE
SUMMARY
- De purpose of this report is to determine the RTrrs values for the Byron Unit 2 reactor vessel beltline materials based upon the results of the Surveillance Capsule X evaluation. He conclusion of this repon is that all the beltline materials in the Byron Unit 2 reactor vessel have RTm values below the screening l
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 to lower shell circumferential weld, WF-447 was the most limiting material with 32 and 48 EFPY PTS values of 116 F and 123*F respectively. i i I l l i I 4 l [ Revision 0 L
- s. .
1-1 1 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 concem 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 report is to determine the RTns values for the Byron Unit 2 reactor vessel using the results of the surveillance Capsule X evaluation. Section 2.0 discusses the PTS Rule and its requirements. Section 3.0 provides the methodology for calculating RTns. Section 4.0 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 5.0 and were obtained from Section 6 of WCAP-15176"I The results of the RTns 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 Introduction Revision 0 t i b
2-1 2 . PRESSURIZED THERMAL SHOCK RULE The Nuclear Regulatory Commission (NRC) 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 latest revision of the PTS ~ Rule"I,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. He rule incorporates in total, and therefore makes binding by rule, the method for determuung the
. reference temperature, RTm, including treatment of the unirradiated RTa value, the margin term, and the explicit definition of" credible" surveillance data, which is also described in Regulatory Guide 1.99, Revision 2m, ,
i
~ 2. The rule is restructured to improve clarity, with the requirements section giving only the requirements for the value for the reference temyerature for end oflicense (EOL) fluence, RTm.
i
- 3. Dermal annealing is i&ntified as a method for mitigating the effects of neutron irradiation, l
thereby reducing RTm. l I l 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 RTm, accepted by the NRC, for each reactor vessel beltline material for the EOL fluence of the material, u
-. The assessment of RTm must use the calculation procedures given in the PTS Rule, and must specify the bases for the projected value of RTm for each beltline material. The report must j 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 RTm or upon the request for a change in the expiration date for operation of the facility. 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: ,
270*F for plates, forgings and axial weld materials, and 300*F for circumferential weld materials. Pressurized Thermal Shock Rule Revision 0
3a
'3 METHOD FOR CALCULATION OF RTns 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 RTm for each weld and plate or forging in the reactor vessel beltline.
RTer = RTercu> + M + ARTer (l) Where, RTem = Reference Temperature for a reactor vessel material in the pre-senice or unirradiated t condition M = Margin to be added to account for uncertainties in the values of RTam, copper and nickel contents, fluence and calculational procedures. 'M is evaluated from Equation 2 a M = dcro + os (2) o uis the standard deviation for RTecu). ou = 0*F when RTercu> is a measured value. ou = 17'F when RTecu)is a generic value. c6 si the standard deviation for RTm. For plates and forgmgs e4 = 17'F when surveillance capsule data is not used. e4 = 8.5'F when surveillance capsule data is used. For welds: os = - 28'F when surveillance capsule data is not used. c6 = 14'F when surveillance capsule data is used. c anot to exceed one half of ARTm ARTm is the mean value of the transition temperature shift, or change in ARTm, due to irradition, and must be calculated using Equation 3. ARTvor = (CF)
- f(o.2 -oice (3) i I
Method For Calcualtion of RTrrs Revision 0
v 3-2 CF (*F) is the chehstry 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.
)
l
- 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 where the J
material in question receives the highest fluence. The EOL fluence is used in calculating RTm.
. Equation'4 must be dsed for determuung RTmusing Equation 3 with EOL fluence values for determining RTm .
RTm = RTacu)+ M + ARTm . (4) To verify that RTm for each vessel beltline material is a boundmg value for the specific reactor vessel, licensees shall consider plant-speci6c 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 msults. Results from the plant-specific surveillance program must be integrated into the RTm estunate if the plant-specific surveillance data has been deemed credible.
'A material-specific value of CF for surveillance materials is determmed from Equation 5.
7 = g.jo2s-onowA)) (5) gou-o aci.:f>) In Equation 5, "A"is i the measured value of ARTm 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 measured values of RTm must be adjusted.for differences in copper and nickel j content by multiplying them by the ratio of the chemistry factor for the vessel material to that for the surveillance' weld.
' Irradiation temperature and fluence (or fluence factor) are first order environmental variables in assessing irradsation damage. To account for differences in temperature between surveillance specimens and vessel, an adjustment to the data must be performed. Studies have shown that for temperatures near 550*F, a 1*F ' decrease in irradiation temperature will result in approxunately a l'F increase in ARTm. For capsules with irradiation temperature oft % and a plant with an irradiation temperature of T,i., an adjustment to i normalize ARTm. -., to T,i,is made as follows:
Temp. Adjusted ARTm = ARTm===.d + 1.0*( Tw. - T,i.) . Note that the temperature adjust +klogy has been reinforce by the NRC at the NRC Industry
' Meetmas on November 12,1997 and February 12,13 of 1998. - Method For Calcualtion of RTm Revision 0 !
4-1 4' VERIFICATION OF PLANT SPECIFIC MATERIAL PROPERTIES Before performing the pressurized thennal shock evaluation, a review of the latest plant specific material S
, properties for the Byron Unit 2 vessel was performed. 'Ihe beltline region of a reactor vessel, per the PT Rule, is defmed as "the region of the reactor vessel (shell material including welds. heat affceted 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 identifies and indicates ' the location of all beltline region materials for the Byron Unit 2 reactor vessel.
W The best estimate copper and nickel contents of the beltline materials were obtained from WCAP-15176 The best estimate copper and nickel content is also documented in Table I herein. The average values were calculated using all of the available material chemistry information. Initial RTsw values for Byron Unit 2 reactor vessel beltline material properties are also shown in Table 1. As a note, per WCAP-15176, Weld 2 WF-614 experience less the 10" n/cm for both 32 and 48 EFPY l 1 Verification of Plant Specific Material Properties Revision 0 L
4-2 l l 1 i V 0 90 0 WF-562 (Heat # 442011) 49 29 / 49C2971-1-1
/
-- 180 0 g 0 W
m 5 a 270 0 CORE ( WF-447 (Heat # 442002) 0 90 Forging [49D330 / 49C298] 'l-1 m 0 0 0
-- 180 a
WF-614 (Heat # 31401) 0 270 Figure 1: Identification and Location of Beltline Region Materials for the Byron Unit 2 Reactor Vessel Verification of Plant Specific Material Properties Revision 0
4-3 Table 1 Byron Unit 2 Reactor Vessel Beltline Unitradiated Material Properties
- Material Description Cu(%) Ni(%) Initial RT,sm
--- 0.71 0
' Closure Head Flange SP7382 / 3P6407 Vessel Flange 124L556VA1 --- 0.70 30 Nozzle Shell Forging 4P-6107*)- 0.05 0.74 10 Intermediate Shell Forging [49D329/49C297]-1-1 0.01 0.70 -20 Lower Shell Forging [49D330/49C298]-1-l*) 0.06 0.73 -20 Intermediate to Lower Shell Forging Cire. Weld 0.04 0.63 10 Seam WF-447 (Heat # 442002)*)
Nozzle Shell to Intermediate Shell Forging Cire. 0.03 0.67 40 Weld Seam WF-562 (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) TheinitialRTmvalues for the plates and welds are based on measured data per reference 5 and 6. (b) Best Estimate Cu% / Ni% and laitial RTm Per Reference 5 and/or 6. Verification of Plant Specific Material Properties Revision 0
5-1 5 NEUTRON FLUENCE VALUES The calculated fast neutron fluence (E > 1.0 MeV) values at the inner surface of the Byron 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' X analysis. See Section 6.0 of the Capsule X analysis report. WCAP-15176DI l TABLE 2 Fluence (E > 1.0 MeV) on the Pressure Vessel Clad / Base Interface for Byron Unit 2
]
at 32 (EOL) and 48 (Life Extension) EFPY
. Material Location 32 EFPY Fluence 48 EFPY Fluence i Nozzle Shell Forging 4P-6107 45* 5.03 x 10 n/cm 7.53 x 10 n/cm
' Intermediate Shell Forging 45* 1.99 x 10 n/cm 2.98 x 10" n/cm
[49D329/49C297]-1-1 Lower Shell Forging [9D330/49C298]-1-1 45* 1.99 x 10 n/cm 2.98 x 10 n/cm2
]
Intermediate to Lower Shell Forging Cire. 45' l.96 x 10n/cm2 2.93 x 10n/cm Weld Seam WF-447 (Heat 442002) ) i Nozzle Shell to Intermediate Shell Forging 45' 5.03 x 10" n/cm 7.53 x 10 n/cm2 i Cire. Weld Seam WF-562 (Heat 442011) f 4 Neutron Fluence Values Revision 0
6-1 l 1 1 6 DETERMINATION OF RTns VALUES FOR ALL BELTLINE 1 REGION MATERIALS. J Using the prescribed PTS Rule methodology, RTns values were generated for all beltline region materials of the Byron Unit 2 reactor vessel for fluence values at the EOL (32 EFPY) and life extension (48 EFPY). i Each plant shall assess the RTns values based on plant-specific surveillance capsule data. For Byron Unit 2, the related surveillance program results have been included in this PTS evaluation. l (See Reference 8 for the credibility evaluation of the Byron Unit 2 surveillance data.) As presented in Table 3, chemistry factor values for Byron Unit 2 based on average copper and nickel weight percent were calculated using Tables I and 2 from 10 CFR 50.6101 Additionally, chemistry factor values based on credible surveillance capsule data from Byron Units 1 and 21 'i, and Braidwood Units I 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). 1 I I i i l Determmation of RTns Values For All Beltline Region Materials Revmon 0 i t.
i 6-2 TABLE 3 Interpolation of Chemistry Factors Using Tables I and 2 of 10 CFR Part 50,61 Material Ni,wt % Chemistry Factor. "F 1 Intermediate Shell Forcine 149D329/49C2971-1-1 0.70 20.0 F Given Cu wt% = 0.01
]
Lower Shell Forninc I49D330/49C2981-1-1 0.73 37.0 F 1 Given Cu wt% = 0.06 Nozzle Shell Forzine 4P-6107 0,74 31,oop Given Cu wt% = 0.05 Intermediate to Lower Shell Circ Weld WF-447 0.63 54.0 F Given Cu wt% = 0.04 Nozzle Shell to Intermediate Shell Cire Weld WF-562 0.67 41.0 F Given Cu wt% = 0.03 Byron Unit I and 2 Surveillance Pronram 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 Given Cu wt% = 0.03 l i Deternunation of RTrrs Values For All Beltlme Region Materials Revision 0 ) I
6-3 TABLE 4 Calculation of Chemistry Factors using Sun'eillance Capsule Data Per Regulatory Guide 1.99, Revision 2, Position 2.1 Capsule Capsule l'" FF) ARTm'" FF*ARTm FF' Material 0.405 0.749 0. 0" ' o 0.561 Lower Shell Forging U W l.27 1.067 3.65 3.89 1138 l49D330/49C298]-1-1 1.225 15.75 19.29 1.500 (Tangential) X 2.30 U 0.405 0.749 19.76 14.80 0.561
' Lower Shell W l.27 1.067 31.88 34.02 1.138 Forging 149D330/
X 2.30 1.225 38.91 47.66 1.500 l 49C298]-1-1 l SUM: 119.66 6.398 i l 2 l CFr = Z(FF
- RTm) + I( FF ) = (!!9.66) + (6.398) = 18.7'F U 0.404 0.748 11.22 8.39 0.560 Byron Unit 1 Sury. Weld Material (5.61)'d' I
80.22 90.17 1.263 (Heat # 442002) X 1.57 1.124 (40.ll)'d' W 2.43 1.239 102.68 127.22 1.535 (51.3' )'d' U 0.405 0.749 16.88 12.64 0.561 Byron Unit 2 Sury. Weld. lvsterial (8.44) i I (Heat # 442002) W l.27 1.067 57.76 61.63 1.138 (28.88)'d' X 2.30 1.225 108.02 1 132.32 1.500 (54.01)'d) I SUM: 432.37 6.557 2 CFs. w.a m2002 = E(FF
- RTm) + I( FF ) = (432.37) + (6.557) = 65.9'F Enis (a) Byron Unit 2 capsule fluences wert updated as a part of the capsule X dosimetry analysis results Ref. 5, (x 10 Mcm 2, E > 1.0 MeV).
(b) FF = fluence factor = f'' * * * 0 (c) ARTmvalues are the measured 30 ft-lb shiA values taken from Ref. 5. (d) The Byron I & . surveillance weld metal ARTmvalues have been adjusted by a ratio factor cf 2.0. No temperature adjustment is required. l
~
Deternunation of RTrrs Values For All Beltline Region Matenals Revision 0
6-1 TABLE 4 - Continued Calculation of Chemistry Factors using Surveillance Capsule Data Per Regulatory Guide 1.99, Revision 2. Position 2.1 i Material Capsule Capsule (*' FF* ARTsm" FF*ARTsn FF T Weld Heat 44201 L WF 501 U 0.3314 0.733 10 7.3 0.538 Using Braidwood 1 Surv. X 1.144 1.038 25 26.0 1077 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 Surv. Data SUM: 54.0 3.232 CFS,, w.ia 442o i = I(FF
- RTm) + I( FF ) = (54.0) + (3.232) = 16.7'F'"
Notes. 2 (a) Braidwood Units 1 & 2 fluences were taken from WCAP-14824 R.2 (Ref. 4)(x 10 n/cm . E > 1.0 MeV). (b) FF = fluence factor = f 28 * * '** 0 (c) ARTmvalues are the measured 30 ft-lb shift values taken from Appendix B of Ref. 4. (d) The Braidwood 1 & 2 surveillance weld metal ARTm alues v 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. l t ( Determmation of RTm Values For All Eeltime Region Matenals Revmon 0 i
I 65 1 TABLE 5 1 l RTns Calculation for Byron Unit 2 Beltline Region Materials at EOL (32 EFPY) Fluence FF CF ARTm'" Margin RTm m* RTm" Material (n/cm'. (*F) (*F) (*F) (*F) ('F) E>l.0 MeV) 1.19 20 23.8 23.8 -20 28 1.99 Intermediate Shell Forging 1.19 37 44.0 34 -20 5g 1.99 Lower Shell Forging ------------------------------ ---- - ------- - - - - - .
- - . - - - - - - - - - - - - - . - 20 1.99 l 19 18 7 22.3 17 19 Lower ShcIl Forging .+ Using S/C Data 0.808 31 25.0 25 10 60 0.503 Nozzle Shell Forging 1.18 54 63.7 56 10 130 Inter, to Lower Shell Cire. Weld 1.96 1.18 65.9 77.8 28 10 116 1.96 Inter. to Lower Shell Circ. Weld -+ Using S/C Data 0.503 0.808 41 33 } 33, j 40 106 Nozzle Shell to Inter. Shell Cire.
Weld ---------- - - .--- ---------- ------------ -. 0.808 16.7 13.5 13.5 40 67 0.503 Nozzle Shell to Inter. Shell Cire. Weld
.+ Using S/C Data (p.
Notes: (a) Initial RTmvalues are measured values (See Table 1) (b) RTns = RTam + ARTns + Margin (*F) (c) ARTm = CF
- FF 1
Deternunation of RTm Values For All Beltline Region Materials Revision 0
66 i TABLE 6 RTm Calculation for Byron Unit 2 Beltline Region Materials at Life Extension (48 EFPY) Material Fluence FF CF ARTm"' Margin RT.mmi RTm*' (n/cm 8, (*F) (*F) (*F) (*F) (*F) 3 J E>1.0 MeV) Intermediate Shell Forging 2.98 1.29 20 25 8 25 8 -20 32 l I' Lower Shell Forging 2.98 1.29 37 47.7 34 20 62 Lower Shell Forging 2.98 1.29 18.7 24.I 17 -20 2g
-+ Using S/C Data j Nozzle Shell Forging 0.753 0.920 31 28.5 28.5 10 67 huer. to Lower Shell Cire. Weld 2.93 1.29 54 69.7 56 10 136 Inter. to Lower Shell Cire. Weld 2.93 1.29 65.9 85 28 10 123 -+ Using S/C Data j Nozzle Shell to Inter. Shell Cire. 0.753 0.920 41 37.7 37.7 40 gg5 Weld Nozzle Shell to Inter. Shell Cire. 0.753 0.920 16.7 15.4 15.4 40 7i Weld l -+ Using S/C Data ]
Notes: { i (a) Initial RTm values are measured values (See Table 1) (b) RTm = RTam + ARTm + Margin (*F) ; (c) ARTm = CF
- FF l
{ i i i l l l l Deternunation of RTns Values For All Beltline Region Matenais Revision 0
D, 7-1
. 7. - CONCLUSIONS .
~ As shown in Tables 5 and 6, all of the beltline region materials in the Byron Unit 2 reactor vessel have EOL H.- (32 EFPY) RTm and Life Extension (48 EFPY) RTm values below the screening criteria values of 270'F jh for plates, forgings and longitudinal welds and 300'F for circumferential welds. Specifically, the 1: ; intermediate to lower shell circumferential Weld, WF-447, was the most limiting material with 32 and 48
' EFPY PTS values of Il6*F and 123*F respectively.
l l l. l 4 1 Conclusions Revision 0 l
7 1 3-1
)
8 REFERENCES I I 1 10 CFR Part 50.61, " Fracture Toughness Requirements For Protection Against Pressunzed Thermal Shock Events", Federal Register. Volume 60, No. 243, dated December 19,1995. l 1 2 Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials". U.S. Nuclear Regulatory Commission, May,1988. I WCAP-10398," Commonwealth Edison Company Byron Station Unit No. 2 Reactor Vessel 4 3 Radiation Surveillance Program", L. R. Singer, December,1983. 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 J November,1997. 5 WCAP-15176, " Analysis of Capsule X from the Commonwealth Edison Co. Byron Unit 2 Reactor Vessel Radiation Surveillance Program", T. J. Laubham, et al., March 1999.
-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 January 6,1998.
- 7. Comed Letter to U.S. Regulatory Commission, " Response to Additional Information Regarding Reactor Pressure Vessel", From R.M. Krich, Dated September 3,1998.
- 8. WCAP-15180, " Commonwealth Edison Co. Byron Unit 2 Surveillance Program Credibility Evaluation", T. J. Laubham, et al., March 1999.
- 9. WCAP-15123, Revision 1, " Analysis of Capsule W from Commonwealth Edison Company Byron )
Unit 1 Reactor Vessel Radiation Surveillance Program", T.J. Laubham, et. al., Jan! ary 1999. l 4 I References Revision 0
Attachment G Byron Station WCAP-15183, Rev. O," Commonwealth Edison Company Byron Unit 1 Surveillance Program Credibility Evaluation" 1 4 4}}