ML20058H777
| ML20058H777 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 11/29/1993 |
| From: | Devan M, Moore K, Yoon K BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20058H749 | List: |
| References | |
| 77-1228963, 77-1228963-00, NUDOCS 9312130222 | |
| Download: ML20058H777 (20) | |
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{{#Wiki_filter:'. I PRESSURIZED THERMAL SHOCK (PTS) EVALUATION t FOR ZION UNITS 1 AND 2 77-1228963-00 F Prepared for Commonwealth Edison Company Prepared By B&W Nuclear Service Company Engineering and Plant Services Division l P. O. Box 10935 Lynchburg, Virginia 24506-0935 NONPROPRIETARY t P PreparedBy'2/20$ff f r /29 /qs M. f.' DeVan Dat e Reviewed By [b ll[2 h i K. K. Y o,f Date ApprovedBy[ / f ll Sk h $ K. E."Idoore Da'te '/ Approved By ll!7.7 fi3 D. L Howell Date l i [DR332130222 931203 ) ADOCK 05000295 i p PDR
... ~ l ? p e 1 1. INTRODUCTION I r The Pressurized Thermal Shock (PTS) rule, 10CFR50.61,") became { effective on July 23, 1985, to (1) establish a screening criterion related to the fracture resistance of pressurized water reactor -l (PWR) vessels during pressurized thermal shock events; (2) require f f an analysis and schedule for implementation of flux reduction programs that are reasonably practicable to avoid exceeding the l screening criterion; and (3) require detailed safety evaluations to be performed before plant operation beyond the screening criterion I 1 will be allowed. The intent of this regulation was to improve the safety of the PWR vessels by identifying those corrective actions [ that may be required to prevent or mitigate potential PTS events. Effective June 14, 1991, the U.S. Nuclear Regulatory Commission f (NRC) amended this regulation to change the procedure for calculating the RTn, value. The amendment updates the calculation procedure and makes it consistent with the calculation p'rocedure l ~ for RT described in Regulatory Guide 1.99, Revision 2. m m i . Transients and accidents can be postulated.to occur in PWRs resulting in severe overcooling (thermal shock) of the reactor l vessel concurrent with high pressure. In these PTS events, rapid cooling of the reactor vessel internal surface causes a temperature j distribution across the reactor vessel wall which produces a l thermal stress on the reactor vessel with a maximum tensile stress at the inside surface of the vessel. The magnitude of the thermal stress varies with the rate of change of temperature and with time i during the transient, and its effect is compounded by coincident l pressure stresses. l 1 Prepared By: M. J. DeVan Date: 11/29/93 77-1228963-00 Reviewed By: K. Y. Yoon Date: 11/29/93 Page 2 x i l 1 i
_m t On the basis of. studies of severe overcooling events that have. occurred, the NRC concluded that the PWR reactor pressure vessels i with conservatively calculated' values of RTrrs less than 270*F for plate and forging material and axial welds, and less than 300 F for f circumferential welds present an acceptably low risk of vessel j failure from PTS events. I The purpose of this report is to determine the RTns values for the l Zion Unit 1 and Zion Unit 2 reactor vessels in accordance with n 10CFR50.61 as amended. l l f t I l l t 4 Prepared By: M.
- 3. DeVan Date:
11/29/93 77-1228963-00 Reviewed By: K. K. Yoon Date: 11/29/93 Page 3 ~ ""
Il I 'i l ') 2. BASIS OF INPUT DATA i I The 10CFR50.51 regulation requires that the data used to perform' l the specified calculations must be traceable by including the -i source of all values included in the assessment. The following sections describe the sources for all data used in the evaluation. i 2.1 Material Identification f I The beltline materials (base metals and weld metals) included in this report conform to the beltline definition specified in 10CFR50, Appendix G.l') The Zion Unit 1 and Unit 2 reactor vessel' beltline materials and - their locations were obtained from the reactor vessel schematics contained in WCAP-10962, Revision 3.") l I 2.2 Chemical comoosition The reported Zion Unit 1 and Unit 2 beltline base metal copper and nickel compositions were obtained from the material certifications and are documented in BAW-2166.N The' copper and nickel compositions for the beltline weld metals were obtained from BAW-j 2121P."' j 2.3 Fluence Estimates The peak inside surface fluences for the Zion Unit 1 and Unit 2 reactor vessel beltline materials were obtained from WCAP-10962, i Revision 3. Additional flux reductions have been implemented at l Zion Units 1 and 2 which are not reflected in this analysis. 1 i l Prepared By: M. 3. DeVan Date: 11/29/93 77-1228963-00 Reviewed By: K. K. Yoon Date: 11/29/93 Page 4 ) l l v
i 2.4 Initial Reference Temperature l The measured initial reference temperature (RTm) values for the i base metal beltline materials were-determined using drop weight and Charpy impact data from the material certifications and are documented in BAW-2166. These base metal RT values were m I determined using the method specified in Paragraph NB-2331 of i Section III of the American Society of Mechanical Engineers (ASME)- Boiler and Pressure Vessel (B&PV) Code.W For all the beltline' weld metals except WF-70, a generic mean value of initial RT, are used based on the data in BAW-1803, Revision 1. '88 The initial RTm value for weld metal WF-70 was determined using an alternate method j which is described in Appendix A. f I i i i f 1 ll I I f i Prepared By: M. 3. DeVan Date: 11/29/93 77-1228963-00 \\ Reviewed By: K. K. Yoon Date: 11/29/93 'Page 5 i i h I 9 m r
. ~ -. E ) 3. CALCULATION OF PRESSURIZED THERMAL SHOCK ) REFERENCE TEMPERATURE For the purpose of comparison with the PTS screening criterion, the value of RTyrs for the Zion Unit 1 and Unit 2 reactor vessels. must l be calculated for each beltline base metal and weld metal as i follows: RT,.n = Initi i RT, + Margin + ART,,n (I) I i where: Initial reference temperature-of the Initial RTm = unirradiated material. Margin = Margin to be added to cover uncertainties in the values of initial RTag copper and j nickel
- contents, fluence and the calculational procedures.
If generic values of initial RTm are used, Margin = i 66 F for welds and 48 F for base metals. If measured values of initial RTm are ~l ~ 56 F for welds and 34 F used, Margin = for base metals. (ff) (CF) ARTm = chemistry factor determined from CF = Tables in 10CFR50.61 for base metals and weld metals. Surveillance data may be used in the chemistry factor' calculation if it is deemed credible per Regulatory Guide 1.99, Revision 2. l ff = fluence factor = f o.2w.a los n where f equals fluence (10" n/cm, E > 1 MeV). 2 Prepared By: M. 3. DeVan .Jate: 11/29/93 77-1228963-00 Reviewed By: K. K. Yoon Date: 11/29/93 Page 6 i )
+ i .l The results of the reactor pressure vessel-specific PTS calculations using the Equation 1 and the data sources described in j i Section 2 which meet the requirements of 10CFR50.61 are included in Tables 1 and 2. i For weld metal WF-70, an ' alternative method for determining the RTp.rs value is presented and justified in BAW-2202.
- A direct l
comparison with weld metal WF-70 fracture toughness data to the ( Code reference toughness curves justifies the use of Tm values l instead of the Charpy transition temperature as defined in Paragraph NB-2331 of Section III of the ASME B&PV Code. A summary of this alternative method is described in Appendix A. t i r I I i I h + ? f i 1 l l i Prepared By: M. J. DeVan Date: 11/29/93 77-1228963-00 Reviewed By: K. K. Yoon Date: 11/29/93 Page 7 i f ~
= - =_ 4 Table 1. Evaluation of Zion Unit 1 Pressurized Thermal shock Criterion for 32 EFPY . Estimated Inside Chemical Surface Composition, Fluence 32 EFPY PTS Evaluation Reactor Vessel Heat w/o e 3 2 EFPY initial Chemistry RTm Screening Beltline Region Location Number Tyre Cu Ni n/cm: &L Factor Shift Martrin RT,m Criterica Nozzle Belt Forging ANA 102 A 508, C1.2 0.06 0.83 1.21E+19 +20" 37 39 34 93 270 Interinediate Shell C3795-2 SA-533, Gr.B 0.12 0.49 1.73E+19 -10* 81 93 34 117 270 Intermediate Shell B7835-1 SA-533, Gr.B 0.12 0.49 1.73E+19 -20* 81 93 34 107 270 Lower Shell C3799-2 SA-533, Gr.B 0.15 0.50 1,73E+19 -20* 105 121 34 135 270 Lower Shell B7823-1 SA-533, Gr.B 0.13 0.48 1.73E+19 -20* 87 100 34 114 270 5* 197 207 66 268 300 NB to IS Cire. Weld (ID B2%) WF-154 ASA/Linde 80 0.31 0.59 1.21E+19 + M NB to IS Cire. Weld (OD 18%) SA-1769 ASA/Linde 80 0.26 0.61 - S 182 66 300 1S to LS Circ. Weld (100%) WF-70 ASA/Linde 80 0.35 0.59 1.73E+19 - 2 6 '* 198'* 228'* 2 8 '* 230'* 300 LS to Dut. Circ. Weld (100%) WF-154 ASA/Linde 80 0.31 0.59 <1.00E+17 - 5* 197 66 300 IS Longit. Weld (100%) WF-4 ASA/Linde 80 0.20 0.55 6.29E+18 - 5* 152 132 66 193 270 IS Longit. Weld (ID 39%) WF-8 ASA/Linde 80 0.20 0.55 6.29E+18 - 5" 152 132 66 193 270 IS Longit. Weld (OD 61 %) WF-4 ASA/Linde 80 0.20 0.55 - S 152 66 270 M M LS Longit. Weld (Both 100%) - WF-8 ASA/Linde 80 0.20 0.55 6.29E+18 - S 152 132 66 193 270 '*' WCAP-10962, Revision 3, September 1991.
- BAW-2166, June 1992.
) BAW-1803, Revision 1, May 1991. j See Appendix A of this report. OD not a concern. i ' I Prepared By: M. J. DeVan Date: -11/29/93 77-1228963-00 l Reviewed By: K. K. Yoon -Date: 11/29/93 Page 8 i I i v- .v,m~ ,.w.,-,..,m-w-.- .~vr~. c r ..w.-#,---,. -m--- - ~ - - - - - ..m+ - ~ ~.. -- -~.r,--- -. ~.
.-. - - - =. i Table 2. Evaluation of Zion Unit 2 Pressurized Thermal Shock Criterion for 32 EFPY Estimated Inside Chemical Surface Composition, Fluence 32 EFPY PTS Evaluation Reactor Vessel Heat w/o e 32 EFPY'*' Initial Chemistry RTm Screening Peltline Penion Locatf ort Number TYPE Cu _Ni n/cm' &L Facter Shift Marqin R Criterion h Nozzle Belt Forging ZV-3855 A 508, C1.2 0.09 0.66 1.30E+19 + 10
- 58 62 34 106 270 Intermediate Shell B8006-1 SA-533, Gr.B 0.12 0.54 1.69E+19
+ 10
- 82 94 34 138 270 Intermediate Shell B8040-1 SA-533, G.B 0.14 0.52 1.69E+19
-10" 96 110 34 134 270 Lower Shell B8029-1 SA-533, Gr.B 0 12 0.51 1.69E+19 -10* 81 93 34 117 270 Lower Shell C4007-1 SA-533. Gr.B 0.12 0.53 1.69E+19 + 10
- 82 94 34 138 270 NB to IS Cire. Weld (100%)
WF-200 ASA/Linde 80 0.24 0.63 1.30E+19 - 5 '" 178 191 66 252 ~ 300 IS to LS Cire. Weld (100%) SA-1769 ASA/Linde 80 0.26 0.61 1.69E+19 - S 182 208 66 269 300 LS to Dut. Cire. Weld (100%) WP-154 ASA/Linde 80 0.31 0.59 <1.00E+17 - 5 "' 197 66 300 IS Longit. Weld (Roth 100%) WF-70 ASA/Linde 80 0.35 0.59 6.04E+18 -26 "' 198"' 170'* 2 8 'd' 172'* 270 LS Longit. Weld (Both 100%) WF-29 ASA/Linde 80 0.23 0.63 6.04E+18 - D '" 174 149 66 210 270 4 I '*' WCAP-10962, Revision 3, September 1991. BAW-2166, June 1992. BAW-1803, Revision 1, May 1991. See Appendix A of this report. t i L i Prepared By: M. J. DeVan Date: 11/29/93 77-1228963-00 l Reviewed By: K. K. Yoon Date: 11/29/93 Page 9 _. _,.. _. _. _ - _... _ _ =.._. _.. _ _ _ _. _.. _ _ _.. _.. _ _ _
!.i f -l l 5 1 4.
SUMMARY
i j 4.1 Zion Unit 1 The projected end-of-1.ife (32 EFPY) values of RTp73 for all materials in the reactor vessel beltline region are below the applicable screening criteria. Weld metal WF-154 is the limiting material with a RTrrs value of 268*F (screening criteria is 300*F). 4.2 Zion Unit 2 The projected end-of-life (32 EFPY) values of RTrrs for all-l { ~ materials in the reactor vessel beltline region are below the applicable screening criteria. Weld metal SA-1769 is the limiting { material with a RTrrs value of 269 F (screening criteria is 300 F). -i .i 1 i f i 1 i k l P Prepared By: M. 3. DeVan Date: 11/29/93 77-1228963-00 i' Reviewed By: K. K. Yoon Date: 11/29/93 Page,10-i i l i I
l l i 5.O REFERENCES i 1 1. Code of Federal Regulations, Title 10, Part 50.61, Fracture I Toughness Requirements for Protection Against Pressurized j Thermal Shock Events. l 2. U.S. Nuclear Regulatory Commission, Radiation Damage to' l 1 Reactor Vessel Material, Reculatory Guide 1.99. Revision 2, j May 1988. i 'l l 3. Code of Federal Regulations, Title 10, Part 50, Domestic .{ Licensing of Production and Utilization Facilities, Appendix G, Fracture Toughness Requirements. 4. J. M. Chicots, et al., Zion Units 1 and 2 Reactor Vessel Fluence and' RTrrs Evaluations, WCAP-10962. Revision '3, i Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, September 1991. i I 5. M. J.
- DeVan, L.
B. Gross, and A. L.
- Lowe, Jr.,
B&W Owners Group Response to Generic Letter 92-01, BAW-2166, B&W Nuclear -{ { Technologies, Lynchburg, Virginia, June 1992.. 6. L. B. Gross, Chemical Composition of B&W Fabricated Reactor Vessel Beltline Welds, BAW-2121P, B&W Nuclear Technologies, .!1 Lynchburg, Virginia, April 1991. j
- 7.
American Soc 3 r cy of Mechanical Engineers (ASME) Boiler and Pressure Vessel
- Code, Section III, ~ Nuclear Power -Plant Components,. Division 1, Subsection NB, Class.1 Components.
,{ I 1 Prepared By: M. J. DeVan Date: 11/29/93 27-1228963 ;_ Reviewed By: Y. Y. Yoon Date: 11/29/93 Page il .I .,,. =
a n a ~> .a . i t 8. A. L. Lowe, Jr. and J. W. Pegram, Correlations for Predicting the Effects of Neutron Radiation on Linde 80-Submerged-Arc
- Welds, BAW-1803.
Revision 1, B&W Nuclear Technologies, Lynchburg, Virginia, May 1991. f t i 9 K. K. Yoon, Fracture Toughness Characterization of WF-70 Weld l
- Metal, BAW-2202, B&W Nuclear Technologies, Lynchburg,-
i Virginia, September 1993. l l I 1 i f' i i ? 4 h j l' i W i 9 a i i i Prepared By: M. G. DeVan Date: 11/29/93 77-1228963-00 ' 1 Reviewed By: K. K. Yoon Date: 1_1/29/93 Page 12 ,v.* e
i j l APPENDIX A ALTERNATIVE INITIAL RTm DETERMINATION A.1 Introduction l Commonwes.lth Edison Company submitted a PTS evaluation for the Zion Units 1 and 2 reactor vessel limiting beltline weld metal, WF-70, on May 22, 1992. A revised evaluation based on further WF-70 weld metal testing is described below. This submittal differs from the j previous submittal in (1) the method for determination of the ) initial RTm, and (2) the determination of the chemistry factor i i which is now solely based on Zion specific surveillance data. i Previously, the initial RTm for WF-70 weld metal was determined from the Charpy transition tenperature. An alternative method for l determination of initial RTm has been developed by the B&W' Owners i Group (B&WOG)'which is validated by a direct toughness curve. The [ methodology is provided in a separate B&WOG report BAW-2202. { i 7 A.2 Evaluation of Initial RTm j t In BAW-2202, it is demonstrated that use of the Charpy transition ] temperature-based initial RTm results' in two significant problems: .j (1) it introduces a large scatter in the data and (2) it introduces l widely fluctuating values through the thickness of a Midland WF-70 l weld metal piece. However, based on actual fracture toughness i I testing, BAW-2202 justifies the use of drop weight. test data for determination of initial RTm for welds'made'from wire heat 72105 [ I All'available B&WOG and Oak Ridge National Laboratories (ORNL) drop j weight test data on Heat 72105 are listed in Table A-1. From the available drop' weight data, a mean Tm temperature value of. -56*F L Prepared By: M.
- 3. DeVan Date:
11/29/93 77-1228963-00 Reviewed By: K. K. Toon Date: 11/29/93 Page 13 l ) I i ,f
i is established with a standard deviation (0 ) of 14.8*F. Using 2 this mean Tm temperature an d the s?.andard deviation, the initial l RTm value for WF-70 is determined as follows: i -26.4*F (A1) Initial RTa=Tm+20 = 1 i (For conservatism, a value of -26*F will be used.) l A.3 Evaluation of Shift of RT I m The surveillance data available for Zion Units 1 and 2 weld metals I fabricated of wire heat 72105 include the data from the reactor vessel surveillance programs of both units. Seven individual data sets (seven neparate surveillance capsule irradiations) are presented in Table A-2 and constitute all the irradiated surveillance data for Zion Units 1 and 2 that are currently available. The data have been verified by the B&WOG by a thorough review of all sou rce documents; the fluence values are based-on re-l ~ evaluations of the capsule fluence. t i The irradiated shift data were analyzed using the procedure described in Regulatory Guide 1.99, Revision 2, Position 2, for l credible surveillance data. A chemistry factor of 198 was calculated as shown in Table A-3, which is the relationship of RT m shif t to fluence that fits the surveillance data of both Zion units l t combined. A chemistry factor of 198 will be used for calculating l the RTm for the weld metal WF-70. j i A.4 Marai.n f The Margin term was calculated using the following expression from Regulatory Guide 1.99, Revision 2: l
- h l
i i l Prepared By: M. 3. DeVan Date: 11/29/9)_ 77-1228963-00 Reviewed By: K. K. Yoon Date: 11/29/33 Page 14 t i 1 i ~
I r l I ~ Margin = 2 lo3 + of (* I l C where: o = standard deviation for the initial RTm . [ r o, = standard deviation for ARTm i Since the standard deviation of the available drop weight data is used in the determination of the initial RTmy value, at in the above expression is 0. Since credible surveillance data is used in the i determination of the chemistry factor for the WF-70 weld metals, j the standard deviation for ARTm is 14 F based on Regulatory Guide f 1.99, Revision 2, Position 2. Using these values in the above l expression, the margin factor is calculated to be 28*F. { A.5 RT g Evaluation i n Using the initial RTm, ARTm, and Margin values determined above . l I' and Equation 1 in 10CFR50.61, the RTns values were calculated for weld nitsteil WF-70 at the end-of-life (32 EFPY) fluence values for l Zion Units 1 and 2 and are presented in Table A-4. l t i l i i I l Prepared By: M. 3. DeVan Date: 11/29/93 77-1228963-00 Reviewed By: K K. Toon Date: 11/29/93 Page 15 i i t 4 ,,n +, r v w -r,~ ~ ,v., a -e e-.
i l A.6 Summarv i 1. Initial RT,m - The initial RTm value to be used for Zion Units 1 and 2 WF-70 beltline welds is calculated as follows: l t Tm (mean) = -56 F Standard Deviation (0 ) = 14.8'F 3 4 t Initial RTm = Tm (mean) +20 = -26 F 2 l 4 2. Shift of RT,w - Using the methodology of Regulatory Guide 1.99, Revision 2, Position 2, a chemistry factor of 198 was -{ calculated for the seven surveillance' data sets with wire heat number 72105. This chemistry factor is used in the determination of the RTm shift. l 3. Marctin - The margin term was deternined using the expression i defined in Regulatory Guide 1.99, Revision 2. Because the standard deviation of the available drop weight data is already used in the determination of the initial RTm value, is considered to be O for the purpose of calculating the j or margin term for weld metal WF-70. Therefore, only the standard deviation for ART (14*F) is used to calculate the 'l m margin value (based on Regulatory Guide 1.99, Revision 2, l Position 2). 4. RTpre for WF Based on the preceding values of initial
- j RTm, RTm shift, and margin, the RTprs at 32 EFPY are as
.j follows: } Zion Unit 1, RTprs = 230 F Zion Unit 2, RTrrs = 1726F' { i Prepared By: M. 3. DeVan Date: 11/29/93 77-1228963-00 l Reviewed By: K. K. Yoon __ Date: 11/29/93 Page 16 i
^ t A Table A-1. Summary of Available Drop Weight Data for Weld Metals Fabricated with Weld Wire 72105 and Linde 80 Flux f i S.R. Drop Wt. Source Material Time. Hrs. T,y, Fm l Oconee 2-RVSP WF-209-1 33 -20 Oconee 3-RVSP WF-209-1 30 -20 l Zion 1-RVSP-WF-209-1 23 -70 Zion 2-RVSP WF-209-1 30 -70 B &WOG-RVSP WF-70 48 -50 r B&W-NBD WF-70 48, t HSST-Series 3 WF-70 48 -50 s Midland Beltline WF-70 23 -60 Midland Beltline WF-70 23 -60 l Midland Beltline WF-70 30 -60 l 4 1 Midland Beltline WF-70 40 -50 l Midland Beltline WF-70 50 -60 ORNL Midland Beltline WF-70 23 -76 f ORNL Midland Beltline WF-70 23 -58 f ORNL Midland Beltline WF-70 23 -76 ORNL Midland Beltline WF-70 23 -58 l ORNL Midland Beltline WF-70 23 -76 h t ORNL Midland Beltline WF-70 23 -49 t t ORNL Midland Beltline WF-70 23 -49 l ORNL Midland Beltline WF-70 23 -67 ORNL Midland Nozzle Belt WF-70 28 -40 ] ORNL Midland Nozzle Belt WF-70 28 -49 ORNL Midland Nazzle Belt WF-70 28 -58 ORNL Midland Nozzle Belt WF-70 28 -67 i (1) Drop weight temperatures of ORNL data were converted from [ degrees centigrade. I Prepared By: M. J. DeVan Date: 11/29/93 77-1228963-00 .l Reviewed By: Y. K. Toon Date: 11/29/93 Page 17 l 5 h e uW -m-,iv-r - + _. --rr n y = g i--,r
s e Table A-2. Zion Units 1 and 2 Surveillance Transition Temperature Data (Data available through 11/17/93) 30 ft-lb Transition Capsule Weld Fluence Temperature.
- F Plant Ident.
Metal n/cm Initial Irradiated Charige T 2 Zion Unit 1 T WF-209-1D 2.53E+18 +4 116 112 Zion Unit 1 U WF-209-1D 8.49E+18 + 4 203 199 ' l Zion Unit 1 X WF-209-1D 1.26E+19 +4 203 199 Zion Unit.1 Y WF-209-1D 1.56E+19 + 4 209 205 Zion Unit 2 U WF-209-1E 2.57E+18 -23 122 145 Zion Unit-2 T WF-209-1E B.04E+18 -23 168 191 Zion Unit 2 Y WF-209-1E 1.48E+19 -23 208 231 I i I i i 1 Prepared By: M. J. DeVan Date: 11/29/93 77-1228963-00 Reviewed By: K. K. Yoon Date: 11/29/93 Page 18 i ______________,,,m._.._-._m.-.._. .,....,,,,4.,,...m~._ ,,,,,m-... ,-,_...-,,.,-,mm...-,.._..,- , -,, ~,..,..... .e
t 4 Table A-3. calculation of Chemistry Factor Usino Zion Units 1 and 2 Surveillance Capsule Data Capsule Fluence Fluence Measured Plant Ident. .n/cm2 Factor ARTa_. ff* ART,r (ff)^2 Zion Unit 1 T 2.53E+10 0.627 112 70.224 0.393 Zion Unit 1 U B.49E+18 0.954 199 189.846 0.910. Zion Unit.1 X 1.26E+19 1.064 199 211.736 1.132 Zion Unit 1 Y 3.56E+19 1.123 205 230.215 1.261 Zion Unit 2 U i.57E+18 0.631 145 91.495 0.398-Zion Unit 2~ T e.04E+18 0.939 191 179.349 0.882-Zion Unit 2-Y 1.48E+19 1.109 231 256.179 1.230 1229.044 6.206 198.04 Chemistry Factor 1229.004 = = 6.206 J 4 Prepared By: M. J. DeVan Date: 11/79/93 77-1228963-00 . Reviewed By: K. K. Yoon Date: 11/79/93 Page 19 a,-....-..- _ ~,.....
4 Table A-4. Evaluation of Zion Cnits 1 and 2 WF-70 Pressurized Thermal Shock Criterion for 32 EFPY Per 10CFR50.61 (May.1991) with RG 1.99. Rev. 2. Pos. 2. Chemistry Factor and Marcin i I t chemical Estimated Composition,. IS Fluence 32 EPPY PTS Evaluation Reactor Vessel liest w/o ___ e 32 EFPY Initial RT,n Screening R Criterien Beltline Recien tocation Nu h r Type Cu Ni n /cm' &1 Shift Marcin M Zion Unit i WF-70 ASA/Linde 80 0.35 0.39 1.73E+19 -26 228 28 230 300 - IS to 1.S Circum. Weld (1005) t Zion Unit 2 WF-70 ASA/Linde 80 0.35 0.'59 6.04E+18 -26 170 28 172 270 Interm, t,ongit. Weld (100%) l 4 6 ) Prepared By: M. J. DeVan Date: 11/29/93 77-1228963-00 Reviewed By: K. K. Yoon Date: 11/29/93 Page 20 -m . m -mm_m__m_<,m._._.Emwm___m_m. _. m mm ..-.m.. .,,-,mm,m,.,, ,w7 ,,,,,ce, .,w,,ws..,mwe.,,.s.,w .e.,,....w,,,,....%..,,. _m.w..sm,w%,,...,,ymwe,. ,m.,m}}