ML20044B141

Forwards Addl B&W Rept 77-1159832-00 to Facilitate Completion of Reviews & Closeout of Pressurized Thermal Shock Issue,Per NRC Request

ML20044B141
Person / Time
Site:	Byron, Braidwood
Issue date:	07/12/1990
From:	Hunsader S COMMONWEALTH EDISON CO.
To:	Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation
References
NUDOCS 9007170424
Download: ML20044B141 (1)

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\\ Comranwealth Edison _ /1400 Crus Ptee %~ rr "

s 2 Downers Grove, Illinois 60515 a

V' July 12, 1990 Mr. T.E. Hurley Office of Nuclear Reactor Regulation i U.S. Nuclear Regulatory Commission Hashington, D.C. 20555 i ATTN' DOCUMENT CONTROL DESK

Subject:

Byron Station Units 1 and 2 Braidwood Station Units 1 and 2 Pressuriu d Thermal Shock l RRC__Dathtt Nos. 50-454/455 and 50-456/452 1

Reference:

(a) January 19, 1986 G.L. Alexander letter to H.R. Denton (b) November 26, 1986 L.N. Oishan letter to D.L. Farrar

Dear Dr. Hurley:

Reference (a) provided the Commonwealth (Edison) response to meet the l requirements iof 10 CFR 50.61(b)(1). For the Byron and Braidwood Stations the required information was contained in Babcock and Hilcox Report 77-1159832-00. This response indicated that all of the RT PTS values were found to be below the NRC screening values. Consistent with the requirements of 13 CFR 50.61 no further submittals are necessary to be made unless there is a significant change in those projected values. Reference (b) provided the NRC staff acceptance of reference (a) for 7 Byron Unit 1. However, no correspondence has been received to date regarding Byron Unit 2 and Braidwood Units 1 and 2. At the request of the NRC staff,

l Edison irJ providing an additional copy of the Babcock and Wilcox report 77-1159832-00. This has been requested to help facilitate the completion of revtews and close-out of this issue for Byron Unit 2 and Braidwood Units 1 and 2.

\\ Please address any que Mions that you may have concerning this submittal to this office, i; Very truly yours, ( u w der S.C. Hunsader Nuclear Licensing Administrator

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Attachments cc: T. Boyce - NR't S. Sands - NRR H. Shafer - RIII Byron Resident. Inspector o Braidwood Resident Inspector 0 ZNLD/ID101/TLM i{ L. ,${f@EA$25$g4 g vt Li~'

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NUCLEAR REGULATORY COMMISSION ( 3 W ASHINo TON. D. C. 20666 (' 74 November 26, 1986 l'ocket No. STN-50-454 Mr. D. L. Farrar Director of Nuclear Licensing Commonwealth Edison Company Post Office Box 767 Chicaao, Illinois 60690

Dear Mr. Farrar:

SUB.1ECT: PROJECT VALUE5 0' + ' A PWERTIES FOR FRACTURE TOUGHNESS REQUIREMENTS FO. F e c't0N O AINST PRESSURIZED THERMAL SHOCK EVENTS u, 17, 1986 which was submitted in We have reviewed your letter dated January (PTS) Rule,10 CFR 50.61, for Byron response to the Pressurized Thermal Shock Stition,' Unit 1. We have found the material properties of reactor vessel be,'itlirs materials, the projected fluence at the inner surface of the reactor vessel for the end of life of the plant and the calculated RT for the end The calculated RT U3below the of life of the plant to be acceptable. screeninacriteriaof?70' Fat 3?EFPY.whichisbevondtNbaxpiration date of the licensa and, therefore, meets the requirements of the PTS Rule. The PTS Rulefrequires that the projected assessment of the RT must he rundatedwheneverchangesincoreloadinas,surveillancemeasuNbntsorother information (including changes in capacity factori indicate a sianificant change in the' projected values. This ensures that the licensee will track the accumulated fluence for the limting beltline materials throughout the life of the plant to verify that their assumptions remain valid. In this regard w_a_ request that you submit a re-evoluation of the RTme and comparison of the predicted value in any future Pressure-TemperaturF fubmittals which are submitted as required by 10 CFR 50. Appendix G, Our associated Safety Evaluation is enclosed., Sincerely, m g yi Leonard N. 01shan, Project Manager Project Directorate #3 Division of PWR Licensing-A

Enclosure:

As stated cc: See next page (A Eg

l' L Mr. Dennis L. Farrar Pyron Station Commonwealth Edison Company Units 1 and 2 cc: Mr. William Vortier Ms. Diane Chavez Atomic Power Distribution 5?8 Gregory Street Westinghouse Electric Corporation Rockford, Illinois 61108 Post Office Box 355 Pittsburgh, Pennsylvania 15?30 Regional Administrator, Reoion !!! ll. S. Nuclear Regulatory Commission Michael Miller 799 Roosevelt Road Isham, t.ircoln A Peale Glen Ellyn,1111ncis 60137 One First National Plaza 42nd floor iloseph Gallo. Esq. Chicago, Illinois 60603 1 sham, Lincoln A Beale Suite 1100 Mrs. Phillip B. Jchnson 1150 Connecticut Avenue, N.W. 1907 Stratford Lane Washington, D. C. 20036 Rockford, Illinois 61107 Douglass Cassel Esq. Dr. Bruce von 7 ellen 109 N. Dearborn Street .c ite 1300 Department of Biolonical Sciences u Northern Illinois t!niversity Chicaoo, Illinois 60602 DeKalb, Illinois 61107 Ps. Pat Morrison Mr. Edward R. Crass 5568 Thunderidge Drive. Nuclear Safeguards & Licensing Rockford, Illinois 61107 Saraent & Lundy Engineers 35 East Monroe Street Ms. Lorraine Creek Chicaco,'1111nois 60603 -Rt. 1. Box 182 Panteno. Illinois 60950 Pr. Julian Hinds

11. S. Nuclear Regulatory Commission Pyron/ Resident Inspectors Offices 4448 German Church Road nyron, Illinois 61010 Pr.. Michael C. Parker, Chief Division of Engineering Illinois Department of Nuclear Safety 1035 Outer Park Drive Springfield, Illinois 62704 l

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he* % I /,' ' 'o, UNITED STATES NUCLE AR REGULATORY COMMISSION WASHINGTON, D..C. 20h56 j %,...../ i t SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOP REGULATION i i REGARDING PROJECTED VAltlE5 0F MATERIAL PROPERTIE5 FOR FRACTURE T0llriHNE55 REQUIREMEN15 FOR PQOTECTION AGAINST PRES 5URIZED THEPMA'_ 5 HOCK EVENTS CO'N0 WEALTH EDISON COMPANY BYPDF STATION, UNIT 1 i DOCFET NO.' STN 50-454 i INTRODUCTION As required bv 10 CFR 50.61 " Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock" (PTS Rule) which was published in the FIDfRAL PEGISTER July 23, 1985, the licensee for each operating pressurized ] surfacei of reactor vessel beltline materials by qNkno(at the inner vessel i water reactor "shall subnit pro.iected values of RT values from the time j of submittal to the expiration date of the operating Ilcense. The assessment must specify the bases for the projection including the assumptions regarding core loading patterns. This assessment must be submitted by January 23, 1986, and nust be updated whenever changes in core loadings, surveillance measurements or other information indicate a significant change in projected values." By letter dated Januarv 17, 1986, the Commonwealth Edison Company, licensee for the Byron Unit 1 plant, submitted information on the material properties and the fast neutron fluence (E E 1.0 PeV) of the reactor pressure vessel in compliance with the requirements of 10 CFR 50.61. The following evaluation. concerns the estimation of the fluence to the pressure vessel for 32 effective full power years of operation and the corresponding value of RTPTS' EVALUATION OF THE MATERIALS ASPECTS The controlling beltline material from the standpoint of PTS susceptibility was identified to be the upper shell forging 5P-5933. The material properties of the controlling material and the associated margin and chemistry factor were reported to be: Utility Submittal Staff Evaluation Cu (copper content, %) 0.05 0.05 Ni (nickel content, %) 0.73 0.73 I (Initial RT F) 40 40

NDT, M(Margin,'F) 48 48 26.3 CF (Chemistry Factor, 'F)

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~2-The controlling material has been properly identified, The justifications are acceptable. given for'the cooper'and nickel contents and the initial RTg . The margin has been derived from consideration of the bases N r these values, following the PTS Rule, Section 50.61 of 10 CFR Part 50. Assuming.that the reported values of fluence are correct, Equation 1-of the PTS rule governs, and-the; chemistry factor is as shown above. EVAltJATION OF THE FLllENCE ASPECTS AND THE CALCULATED RTnyy The methodology of the' fluence calculation was based on the discrete ordinates code 00T using " Sailor" an ENDF-R/IV hased cross section set. The scattering. is treated with a P approximation. The power distribution utilized in the 3 analysis was derived from statistical studies of lono-term operation-of Westinahouse 4-loop plants..These distributions include the out-in fuel . management and are conservative. - The upper shell forging 5P-5033 was I-identified as the controlling. material. The peak axial and azimuthal fluence 6 isL the" aoolicable value for the estimation of the RT . :The code was bench- . marked by Westinghouse.. The methodolooy, the cross Nhtions and the , approximations used are acceptable, The equation specified in 10 CFR 50.61, as applicable'for the Byron tinit 1, m plant is:- PTS = I+M+(-10+470xCu+350xCuxNi)xf.'27 0 RT where: I = Initial RT M =Uncertainthargin = 40'F Cu = w/o Copper in shell foraing SP-5933 = 48'F r - Ni = w/o Nickel in shell. forgino SP-5933 = 0.05 f = peak azimuthal fluence for 32 effective = 0.73 fuQpowegvearsiE 1.0 MeV) in units of g = 2*80' 10 n/cm m Therefore:- RTPTSY = 40+48+(-10+470x0.05+350x0.05x0.73) x 2.8.'27 0 L = 88+26.3x1.32 = 88+34.7 = 122.3*F-which is less than 270*F, the applicable screening criterion 'and is acceptable. r l: t i lll

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w, CONCLUSIONS of 12?.3'F for plate material at the end The. licensee-has calculated a RT of the 32 EFPY considered in the$velopment of the PTS Rule. Thic is less -then ??O'F which is the screenina criteria for the limiting material at expiration date of the license. This is acepetable. However in order for o throuchout the the staff < to confirm' the licensee's pro.iected estimated RT life of the-. Byron Unit 1 operating license, the licensee iS9equired to submit a re-evalstation of the RT and comparison to the predicted value with future B Pressure-Temperature submSIkals which are required by 10 CFP 50, Appendix G. A PRINCIPAL CONTRIBUTORS L. Lois L. 01shan P. N..Randall Cate: 1 i k' 8 p' y" 4

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m PRESSURIZED. THERMAL SH0CK EVALUATION 3 -.- IN ACCORDANCE-WITH 10CFR50.61 FOR'THE REACTOR VESSELS IN ..e M< RYRON UNITS 1 & 2 i AND l% BRAIDWOOD: UNITS 1 & 2 ) y \\& .\\. l o January 13, 1986 .ti l B&W Contract #583-7497 ' Task 001 L x I-J;t, 2 l M,, i, Prepared by ~ Babcock & Wilcox Company-Nuclear Power Division P. O. Box 10935 ...o 4 L '; Lynchbarg, Virginia 24506-0935 u j:H j g M OL'hur t Q ? fk q ((f l Babcock &Wilton a Meoermou company: e t ,, whw ~ +

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Q-REPORT 000UMENTAT10N s k DOCUMENT IDENTIFIER: 77-1159832-00 TITLE: PRESSURIZED THERMAL SHOCK EVALUATION IN ACCORDANCr WITH 10CFR50.61 FOR THE REACTOR VESSELS IN fl T-BYRON UNITS 1 & 2 AND*BRAIDWOOD UNITS 1 & 2 DATE: 'JNNUARY 13, 1986 1

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CUSTOMER:

COMMONWEALTH EDISON

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} g(. . CUSTOMER OR' DER N0.:- P.-0. #303654 j. B&W CONTRACT NO.: '583-7497, TASK 001 e,.. 2 I a,: 4 4 i::= u ii-: -i-4:i; Babcock &Wilcox i a McDermott company

a,1 -,7 4 9. t .ts,*~ j . TABLE OF CONTENTS-Page -l,- INTRODUCTION. 1-1 2. BACKGROUND.............................. 2-1 , 3. INPUT DATA 3.1. Materials Data...............'.',, ..... 3-1 3 3.2. Neutron Fluence Estimates 3-2 ly 4.- -RT CALCULATIONS...............,,......, 4-1. PTS 4 v 5.- CONCLUSIONS 5-1 6. REFERENCES,.,, 6-1 o List of Tables . (2-Table-1. Evaluation of Byron 1 Reactor Pressure Vessel in Accordance with Pressurized Thermal Shock Criteria..........-...... .3 2. Evaluation of Byron 2 Reactor Pressure Vessel in Accordance with Pressurized Thermal Shock Criteria.............. 3-4 3. Evaluation of Braidwood 1 Reactor Pressure Vessel in

Accordance with Pressurized Thermal Shock Criteria 3-5 4.

Evaluation'of Braidwood 2 Reactor Pressure Vessel in-Accordance with Pressurized Thermal Shock Criteria 3-6 ....... Babcock &Wilcox a McDerrnott company

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o \\, K. E. Moore, S. H. Campbell, and A..L. Lowe, Jr. c. Pressurized Thermal Shock Evaluations in Accordance with 10CFR50.61 for the Reactor Vessels in Commonwealth Edison Company's ? Byron Units 1 & 2 and Braidwood Units 1 & 2 - i 4 ABSTRACT Pressurized thermal shock evaluations were performed in accordance with / t" 10CFR50.61 Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," for the' reactor vessel-(RV) beltline region materials q inLthe Commonwealth Edison Company's' Byron Units 1 & 2 and Braidwood Units 1 & 2. .The projected values of RT f r all.these materials are'below the ~ PTS screening criteria for fast neutron'fluences projected to 32-effective full-poweryears(2180x.1019; E>1 MeV).

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C 't 1. INTRODUCTION The. Nuclear Regulatory Conrnission's pressurized thermal shock (PTS) rules for pressurized water reactors (PWRs)' are contained in 10CFR50.61. This document 1 -i requires-that licensees submit projected values of refer'ence temperature for i each of the reactor vessel beltline materials. Thesevalues,(RTPTS). as-determined for the Coninonwealth Edison Byron Units 1 and 2 and Braidwood Units l'and 2, are presented in this report. It also contains PTS background information, a description of the reacter vessel beltline materials, the - source of the materials, neutron fluenc6 data, and a review of the t-calculational methods employed, i s_.- L 1.; i x y I b V 1-1 Babcock &Wilcox a McDermott company n.

). O = y 2. BACKGROUND ? The Nuclear Regulatory Commission (NRC) amended its regulations for light ~ water nuclear power plants, effective July 23,1985f to (1) establish a screening criterion related to the fracture resistance of pressurized water . reactor (PWR)vesselsduringpressurizedthermalshock('PT'S) events; (2) require analyses and schedule for implementation of flux reduction. programs that are reasonably practicable to avoid exceeding the screening criterion; and (3)' require detailed safety evaluations to be performed before l plant operation beyond the screening criterion will be considered. These j amendments are intended to produce an improvement in the safety of PWR vessels by identifying those corrective actions that may be required to prevent or mitigate potential PTS events. Transients and accidents can be postulated to occur in pressur'ized water ~ reactors (PWRs) that result in severe overcooling (thermal shock) of the - l reactor vessel concurrent with high pressure. In these pressurized thermal _ shock (PTS) events, rapid cooling-of the reactor vessel internal surface causes a temperature distribution across the reactor vessel wall. This -temperature-distribution produces a thermal stress on the reactor ressel with a maximum tensile stress at the inside surface of the vessel. The magnitude -of the thermal stress varies with the rate of change in temperature and with- -time-during the transient, and its effect is compounded by coincident pressure stresses. Severe reactor system overcooling events with pressurization of the reactor ' vessel-(PTS events) are postulated to result from a-variety of causes. These include system transients, some of which are initiated by instrumentation and control system malfunctions (including stuck open valves in either the primary or secondary system), and postulated accidents such as small break loss-of-coolant accidents, main steam line breaks, and feedwa ter line breaks. As long as the fracture resistance of the reactor vessel material is 5 relatively high, these events are not expected to cause vessel failure. 2-1 Babcock &WHcom a McDermott company

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However, the fracture _ resistance of the reactor _ vessel' material decreases with

,the i_ntegrated exposure to fast neutrons during the life of a ' nuclear power plant. The rate of decrease depends on the chemical composition of the vessel wall and weld materials. If the fracture resistance of the vessel is reduced- ? sufficiently by neutron irradiation, seere PTS events.could cause small flaws. .that might exist... N 'he inner surface to propagatt into the vessel wall. {, Th'eTassumed initial flaw might be enlarged into a cred through the vessel ._ wall of sufficient extent to threaten vessel integrity _ and, therefore, core L ' cooling capability.- r The toughness state of-reactor vessel materials can be characterized by a " reference temperature for nil ductility tra'nsition" (RTHDT). At normal operating temperatures, vessel materials ara quite tough and resistant to . crack propagation.- As the temperature decreases, the metal gradually loses toughness over a temperature range of about 100*F. RT is a measure of.the-NDT -temperatum range at which this toughness transition occurs. Its value depends on the specific material in the vessel wall and the integrated neutron ' ( ' irradiation received by the vessel. - These effects are determined by destructive tests of material specimens. Correlations, based on-tests _of irradiated specimens, have been developed to calculate the shift inRT HDT as'a function of neutron fluence for various material compositions. The value of RTHDT at a given time-in a vessel's life is used in fracture mechanics calculations lto determine whether assumed pre-existing flaws would propagate -when the vessel is subjected to overcooling events. On the basis of the studies of severe overcooling events that have occurred, generic calculations of-postulated PTS events that could occur, and vessel y integrity calculations, the NRC concluded that a value of RT HDT can be - sekcted so that the risk from PTS events for reactor vessels with smaller RT values is acceptable. (The risk of vessels with higher values of RT NDT might 'also be shown to be acceptable but the demonstration would require NDT detailed plant-specific evaluations and possibly modifications to existing equipment, systems,andprocedures.) The NRC approach to selection of the 'RTNDT screening criterion is described in detali in SECY-82-465.2 In summa ry,. the approach was_to use a deterministic fracture mechanics algorithm to f calculate the value of RTNDT f r which assumed pre-existing flaws in the n, c 2-2 1x Babcock &WHcom 4 McDermott company cp e

~- _.m . 's'^ _-. ; - r { reactor; vessel would be_ predicted to initiate (grow deeper into the vessel'- wall). assuming occurrence of-one of the severe overcooling events that have 4 been experienced. These'" critical" values of RT were related to the HDT expected frequency of_the experienced severe overcooling events based on a 1 limitsd data base, consisting of eight events in 350 reactor-years, h The designation RTPTS (reference temperature for pressurized thermal shock)-is' 'the nil ductility temperature of the material as defined by 10CFR50.61, j Paragraph (b)(2) for.use as~a screening criterion. This designation is used to avoid confusion with the RTNDT used to characterize the toughness state of: ' reactor pressure vessel materials, i On the basis of these studies, the NRC concluded that the PWR reactor pressure vessels with conservatively calculated values of.RT less than 270'F for ~ P.TS ( plati'and forcjing'initifialind axial weld,[an[less,ti[an 3,00'ilor j circumferential welds present an acceptably low risk of vessel failure from PTS ev_ents. i.. -The requirements of 10CFR50.61 further state the following: i "For each. pressurized water nuclear power reactor for which an operating-J license has been issued, the' licensee shall submit projected values of k RTPTS (at the inner vessel surface) of the reactor vessel beltline 9 materials by giving values from.the time of submittal to the. expiration 3 date of the operating license. The assessment must specify the bases for ( the projection, including the. assumptions regarding core loading j patterns. This assessment must be submitted by January 23, 1986, and must-be updated whenever changes in cora loadings, surveillance measurements, or other'information indicate a significant change in projected values." i L e 2-3 Babcock &WHcom a uconmois company

N 4 1 ~A-g l e 3 INPUT DATA The pressurizad thermal shock regulations require that the data used to l} -perform the stecified calculations be traceable by including the source of all values included in the assessment. The relationship ofithe material on which any measurements are made to the actual material in the reactor vessel (RV) must be described. For the fluence values, the assessment must specify the bases for all projections including the assumptions regarding core loading patterns such as standard vs. low-leakage cores, f d The following describes the sources for all data used to evaluate the R.E beltline materials in the Byron and Braidwood units. 3.1. Materials Data (+ The R.V. beltline materials of all four of the Byron and Braidwood units met 2 the requirements of Appendix G of 10CFR50. This included the use of materials with prescribed levels of copper and fracture toughness properties. The chemical compositions.and reference temperature data shown in Tables _1-4 were obtained from the Quality Assurance records available at The Babcock &- Wilcox Company, the manufacturer of these vessels. Either SA 508 C12 mod, or SA 508 C13 forgings were used in the beltline of these plants.- The. test data u were obtained from coupons of the actual forgings in accordance with Section III, Article NB-2000 of the 1971 Edition of the ASME Code and the following Addenda: F Byron I All Addenda through Sumer 1972 ' J Byron II, All Addenda through Braidwood I, II Summer 1973 L 3-1 Babcock &WHcox l' a McDermott company

j': y,'.. r 1The test data shown in these tables for the beltline welds were obtained from . weld metal qualification test samples which also met the requirements of iSectionIII/ArticleNB-2000oftheASMECode'. As can be seen, measured values of RT and copper and nickel concentrations were-available:for each of'the. NDT beltline-materials. l' 3.2. Neutron Fluence Estimates The peak-fast neutron flux at the inside surface of each reactor vessel is= q 1 2.77 x 10 0 nyt-as presented in the FSAR for each plant. The estimated peak 10 D- - neutron fluence is 2.77 x 10 x 32 effective full power years or 2.80 x 109 2 ln/cm (E > 1 MeV). -The value of 32 EFPY is based on assumed 40-year licensed' operating period and 80% ' full power operation during this period.. This fluence was applied to'the upper and lower shell fnrgings and the -circumferential weld Joining these shells, The projected pea'k neutron fluences at other R.V. beltline locations were (> based on the following: s e The relationship. for relative axial variation of fast neutron flux and fluence within the pressure vessel wall (E > 1 MeV) for Zion Units 1 3 '&-2. The relative positions of the beltline materials with respect to the. core e midplane. These positions are virtually identical for all four reactor vessels. 'As shown in Tables 1-4, this value was applied to this weld and the. nozzle bel t, forging. .( 4 3-2 ), a Babcock &Wilcox

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,y = a ~ e' Table 1. Evaluation of Byron 1 Reactor' Pressure Vessel in Accordance with Pressurzized Thermal Shock Criteria Material Description Chemical . Constants for PTS-Calculated-2 Reactor Vessel Heat Composition, w/o RT Calculations. F Inside Surface Fluence, n/cm Screentng RTPTS, F PTS Bettline Region Location Number Type Copper -Michel Initial RT Margin 32 EFPY Criteria, F 32 EFPY I NDT Lower Nozzle Belt 123J218 SA 503 C1 2 mod..05 .72 ' +20 46 6.30E18 270 - 91 Upper Shell SP-5933 SA 508 C1 2 mod..05 .73 +40 48 2.80E19 270 . 123 w Lower Shell SP-5951 SA 508 C1 2 mod. .04 .64' +10 48 - 2.80E19 270 81 Upper Circumferential deld WF501 ASA/Linde 80 028 .63 +10 48 6.30E18 300 66 Middle Circumferential Weld WF336 ASA/Linde 80 .031 46 -30 48 2.80E19 300 31 Lower Circumferential Weld WF472 ASA/Linde 80 .23 .57 +10 48 - .- < E17 300 F s P 5 D ' ng g .o

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= ~:, 4. y .. x -- v --- 4 n._ g:. =l. ~ .~ -[ j f- ^ n. Table 2. Evaluation of Byron 2 Reactor Pressure Yessel in Accordance with Pressurzized Thermal ~5 hock Criteria Chemical Constants for 'Inside Surface FIS ' Calculated ~' Material Description Screening. RTPTS, F Reactor Vessel Heat Composition, w/o; RT Calculations. F Fluence, c/cm PTS Beltline Region Location Number Type Copper Nickel Initial RT Margin 32 EFFT Criteria, F' - 32 EFPf MDT 8.ower Nozzle 8elt 4P.6107 5A 508 C1 2 mod. .05 .74 +10 48 6.30E18 270 81 Upper Shell 490329) 49C297) SA 508 C1 3 01 .70 -20 48 2.80E19 270 30 ~I Y l"# 'II 1 SA 508 C1 3 .05 .73 -20 48 2.80E19 270 63 4 Upper Circumferential Weld WF562 ASA/Linde 80 .03 .65 +40 48 6.30E18 .300 96 Middle Circumferential Weld WF447 AS?./Linde 80 .053 .62 +10 48 2.80E19 300 93 Lower Circumferential Weld WF614 ASA/Linde 80 .18 .54 +40 48 < E17 300 >T 4 .M ee a a iio asg 2-3@ M _.m-. bN

l- : ( Table 3. Evaluation of Braidwood 1 Reactor Pressure Vessel f in Accordance with Pressur:tred Thermal Shock Criteria I Material Description Chemical Constants for Inside Surface PTS Calculated 2 Reactor Vessel Heat Cornposition, w/o RT Calculations. F F19ence, c/cm Screemng RTPTS* I PTS 6?ltline Region tocation Number Type Copper Nickel Initial RT rgin 32 EFPY Criteria. F 32 EFPY NDT Lower Norzle Pelt SP-7016 SA 508 C1 2 mod. .04 .7' +10 48 6.30E18 270 75 Upper shell 49C344) I~I 490383)~ SA 508 C1 3 .05 .73 -30 48 2.80E19 270 53 b Lower Shell 490867) 49C813)~ SA 508 C1 3 .03 .73 -20 48 2.80E19 270 44 Upper Cbcueterential Weld liF645 ASA/tinde 80 .033 .50 -30 48 6.30C18 300 28 Middle Circu-ferential Weld WF562 ASA/ttnde 50 .03 .65 +40 48 2.80E19 300 102 icwer Circumferent tal ' Weld Wr653 f.5A/ttrde 80 .19 .56 -40 88 < E 17 300 6 ae 9 cq ^ C4 2w

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.. ~ - +,, '"'y" ,[ L -w. - 'i j '"(. L... n e-s. i- ~ Table 4.. Evaluatton of Braldwcod 2 9eactor Pressure Yessel' ~ -in Accordsace teith Pressor: fred Thermal Shock Criterie' Cheatica'1 - Constants'for- .Inside Surfaca-Pii " ~ Calculated Material Descriptt m Reactor Yessel.. Heat Compositica, w/o PT Calets13tions, F - Fleience, n/ca.I $creening RT[73gF PTS Peltline Region tocation W e er Type Cepper Nickel Initial RT !Eargin 32 EFPY - Criteria,' F ._32 EFPY-MT Lower Morrie Belt $P-7056 SA 508 C1 2. mod. .04 .90 - + 3C - 49 6.30E18 :- -270 .$7 g.; Upper Shel? -49D963) 1-1 SA 508 C1 3 ~ . 03. .71 _ 48 7.80E19 -270 33 - 49C954}- w Lower Shell 500:02)1-1. .A 508 C1 3 .06 .75 -30 4S 2.80t19 270 ' 63 - 50C97) Upper Circumferential Weld WF645 'A5A/Linde 80 .033 .50 -30 43 -6.30Ela c ' 78 Middle Circumferenttal Weld siF552 ASA/Linde 80 .0{, .65 440 48 2.80E19 .300

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O j { 4. RT CALCULATIONS PTS For the purpose of comparison with the PTS criterion, the value of RT IU" PTS each of the reactor vessel beltline materials must be calculated as described R in the following paragraphs. The calculation must be made for each weld, 2 plate, and forging in the reactor vessel beltline. For each material, the RT PTS is the lower of the results given by Equations 1 and 2. Equation 1 was applicable to the beltline materials in the four reactor vessels in the Byron and Braidwood plants. 0 PTS = I+M+[-10+470Cu+350CuNi]f.270 ,.P Equation 1: RT b 0 PTS = 14M+283f.1M Equation 2: RT "I" means the initial reference temperature of the unirradiated material i a., measured as defined in the ASME B&PV Code Section Ill, Paragraph NB-2331. If a measured value is not available, the following generic mean value must be used: 0*F for weld made with Linde 80 flux (as stated in Part 3, measured values were available for all traterials). 3 b. "M" means the margin to be added to cover uncerttinties in the values of initial RTNOT, copper and nickel content, fluence and the calculctional procedures. In Equation 1, M=48 F if a measured value of I was used at:d M=59*F if the generic mean value of I was used. (Since measured values were available M=48 F was employed in these calculations.) "Cu" and "Ni" mean the best estimate weight percent of copper and nickel c. in the material. 4 4-1 Babcock & Wilcox 1 McDermott company

.= m i I; B .c i 19 Q d. "f" means the best estimate neutron fluer,ce, in units of 10 n/cm2 (E q -1 { greater than or equal to 1 MeV), at the clad-base retal interface on the i l J inside surface of the vessel at the location where the material in question receives the highest fluence for the period of service considered. ] The results of the reactor vessel specific PTS calculatiens using Equation 1 and the date sources described in Part 3 are included in Tables 1 through 4. [ E I m b t P V l i g E = i e i i i = T F 6 i i -si d i 4-2 l Babcock &Wilcox )\\ e ueoerman ecmpany 1 3

Q(('(.,Dg,. -{ p ry;g , ; ', ? yl.',: r l -=fw e./n' g' 5. CONCLUSIONS 3 5( The Byron. Units l' and'2 and Braidwood Units 1 and 2 reactor vessel beltline. ' y materials met the requirements of Appendix G to 10CFR50.. The projected j Sl 40-year RTPTSl values for these materials are well under the PTS scr'ening c criteria. All of the calculated RT values were $ 23'at the estimated peak 1 P 2(TSE > 1 MeV). J 19 neutron fluence of 2.80x10 n/cm n. l to If Ih - (.. E i l r a ) f L i :. u %]j! ,a li Un } S.; 'w .D l4' ',l ? A Lo m,. t-lp Jr. w 'l li /4 iljj ]hl d<1 f;} l 3 k 5.h Babcock &Wilcox y a McDettnott company Lh, fQ 'o! ~ ~ ~ - , n,,

i l ls, ' (_ e ,,,1; - r'

.. t., j

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6. REFERENCES l, 1. 10CFR50,61, " Analysis of Potential Pressurized Thermal Shock Events," July 23,.1985, i j, 2. Appendix G,100FR Part 50, " Fracture Toughness Requl'rements," March 1.- l

1973, 1

l, 3.- S.L. Anderson, Plant Specific' Neutron Fluence Evaluation for Zion Units 1 I and 2," WCAP-10902 August 1985. l p E 1It 3 - (m> ; 4 g. n 6 2t s, q ,.i ll 't i l j. 'sA i r$v .g ]v9. hg 6-1 Babcock &WHcom - G.. a McDermott company 3 1 ; lp.: f - .? s p (_ f i ~ v- - ~ - - --}}