ML17334B468

Evaluation of PTS for DC Cook Unit 2.

ML17334B468
Person / Time
Site:	Cook
Issue date:	02/28/1993
From:	Chicots J, Meyer T WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML17331A266	List: ML17331A265 ML17334B468
References
WCAP-13517, NUDOCS 9304150164
Download: ML17334B468 (30)

Text

~vv WCAP-13517 WESTINGHOUSE PROPRIETARY CLASS 3 A TTA c//gEwl g EVALUATION OF PRESSURIZED THERMAL SHOCK FOR D.C.COOK UNIT 2 J.M.Chicot February 199 AMERICAN ELECTRIC POWER SERVICE CORPORATION APPROVED IN GENERAL 0 APPROVED EXCEPT AS NOTED E3 NOT APPROVED 0 FOR REFERENCE ONLY BY m'r i kd~DATE Work Performed Under Shop Order AFFP-108 Prepared by Westinghouse Electric Corporation for the Indiana Michigan Power Company Approved by: T.A.Meyer, Manager Structural Reliability 8 Plant Life Optimization WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division P.O.Box 355 Pittsburgh, Pennsylvania 15230-0355 4 1993 Westinghouse Electric Corporation All Rights Reserved Table of Contents List of Tables List of Figures TABLE OF CONTENTS~Pa e Introduction Pressurized Thermal Shock Methods of Calculation of RTPTS Verification of Plant-Specific Material Properties Neutron Fluence Values Determination of RTPTS Values for All Beltline Region Materials Conclusions References 13 Table LIST OF TABLES Title Pacae 1.D.C.Cook Unit 2 Reactor Vessel Beltline Region Haterial Properties 2;Neutron Exposure Projections at Key Locations in the D.C.Cook Unit 2 Pressure Vessel Clad/Base Hetal Interface for 8.65 and 32 EFPY 3.Calculation of Chemistry Factors Using D.C.Cook Unit 2 Surveillance Capsule Data 10 4.RTPTS Values for D.C.Cook Unit 2 for 8.65 EFPY 5.RTPTS Values for D.C.Cook Unit 2 for 32 EFPY 12~iciure LIST OF FIGURES Title~Pa e 1.Identification and Location of Beltline Region Haterials for the D.C.Cook Unit 2 Reactor Vessel N 2.RTPTS versus Fluence Curves for D.C.Cook Unit 2 Limiting Haterials-Intermediate Shell Plate, C5556-2 13 INTRODUCTION A limiting condition on reactor vessel integrity known as Pressurized Thermal Shock (PTS)may occur during a severe system transient such as a Loss-Of-Coolant-Accident (LOCA)or a steam line break.Such transients may challenge the integrity of a reactor vessel under the following conditions:

severe overcooling of the inside surface of the vessel wall followed by high repressurization; significant degradation of vessel material toughness caused by radiation embrittlement; and the presence of a critical-size defect in the vessel wall.In 1985 the Nuclear Regulatory Commission (NRC)issued a=formal ruling on PTS.It established screening criteria on pressurized water reactor (PWR)vessel embrittlement as measured by the nil-ductility reference temperature, termed RTPTS.RTPTS screening values were set for[$1 beltline axial welds, forgings or plates and for beltline circumferential weld seams for end-of-license plant operation.

The screening criteria were determined using conservative fracture mechanics analysis techniques.

All PWR vessels in the United States have been required to evaluate vessel embrittlement in accordance with the criteria through end-of-license.

The NRC has amended its regulations for light water nuclear power plants to change the procedure for calculating radiation embrittlement.

The revised PTS Rule was published in the Federal Register, May 15, 1991 with an effective date of June 14, 1991~~.This amendment makes the procedure for calculating RTPTS values consistent with the methods given in Regulatory Guide 1.99, Revision'~

~.

The purpose of this report is to determine the RTPTS values for the 0.C.Cook Unit 2 reactor vessel to address the revised PTS Rule.Section 2 discusses the Rule and its requirements.

Section 3 provides the methodology for calculating RTPTS.Section 4 provides the reactor vessel beltline region material properties for the 0.C.Cook Unit 2 reactor vessel.The neutron fluence values used in this analysis are presented in Section 5.The results of the RTPTS calculations are presented in Section 6.The conclusions and references for the PTS evaluation follow in Sections 7 and 8, respectively.

2.PRESSURIZED THERMAL SHOCK The PTS Rule requires that the PTS submittal be updated whenever there are changes in core loadings, surveillance measurements or other information that indicates a significant change in projected RTPTS values.The Rule outlines regulations to address the potential for PTS events on pressurized water reactor vessels in nuclear power plants that are operated with a license from the United States Nuclear Regulatory Commission (USNRC).PTS events have been shown from operating experience to be transients that result in a rapid and severe cooldown in the primary system coincident with a high or increasing primary system pressure.The 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 result in the propagation of flaws postulated to exist near the inner wall surface, thereby potentially affecting the integrity of the vessel.The Rule establishes the following requirements for all domestic, operating PMRs:*All plants must submit projected values of RTpTS for reactor vessel beltline materials by giving values for time of submittal, the expiration date of the operating license, and the projected expiration date if a change in the operating license or renewal has been requested.

This assessment must be submitted within six months after the effective date of this Rule if the value of RTPTS for any material is projected to exceed the screening criteria.

Otherwise, it must be submitted with the next update of the pressure-temperature limits, or the next reactor vessel surveillance capsule report, or within 5 years from the effective date of this Rule change, whichever comes first.These values must be calculated based on the methodology specified in this rule.The submittal must include the following:

I)the bases for the projection (including any assumptions regarding core loading patterns), and 2)copper and nickel content and fluence values used in the calculations for each beltline material.(If these values differ from those previously submitted to the NRC,'justification must be provided.)

• The RTPTS (measure of fracture resistance) screening criteria'for the reactor vessel beltline region is 270'F for, plates, forgings, axial welds;and, 300 F for circumferential weld materials.

• The following equations must be used to calculate the RTPTS values for each weld,.plate or forging in the reactor vessel beltline: Equation I: RTPTS I+H+hRTPTS Equation 2: hRTPTS (CF)f(0.28-0.10 log f)*All values of RTPTS must be verified to be bounding values for the specific reactor vessel.In doing this each plant should consider plant-specific information that could affect the level of embrittlement.

• Plant-specific PTS safety analyses are required before a plant is within 3 years of reaching the screening criteria, including.analyses'f alternatives to minimize the PTS concern.*NRC approval for operation beyond the screening criteria is required.

3~METHOD FOR CALCULATION OF RTP TS In the PTS Rule, the NRC Staff has selected a conservative and uniform method for determining plant-specific values of RTPTS at a given time.For the purpose of comparison with the screening criteria, the value of RTPTS for the reactor vessel must be calculated for each weld and plate or forging in the.beltline region as follows.RTPTS+H+~PTS where hRTPTS (CF)f('Initial reference temperature (RTNDT)in'F of the unirradiated material M Margin to be added to cover uncertainties in the values of initial RTNDT, copper an'd nickel contents, fluence arid calculational procedures.

H-66'F for welds and 48'F for base metal if generic values of I are used.H=56'F for welds and 34'F for base metal if measured values of I are used.f Neutron fluence, n/cm2 (E>1HeV at the clad/base metal interface), divided by 10 CF Chemistry factor from tables~2~for welds and for base metal (plates and forgings).

If plant-specific surveillance data has been deemed credible per Reg.Guide 1.99, Rev.2~~, it may be considered in the calculation of the chemistry factor.

VERIFICATION OF PLANT-SPECIFIC HATERIAL PROPERTIES Before performing the pressurized thermal shock evaluation, a review of the latest plant-specific material properties was performed.

The beltline region is defined by the PTS Rule~2~to be"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 irradiation damage to be'onsidered in the selection of the most limiting material with regard to radiation damage." Figure I identifies and indicates the location of all beltline region materials for the D.C.Cook Unit 2 reactor vessel.Haterial property values were obtained from material test certifications from the original fabrication as well as the additional material chemistry tests performed as part of the surveillance capsule program~~.The average copper and nickel values were calculated for each of the beltline region materials using all the available material chemistry information.

A summary of the pertinent chemical and mechanical properties of the beltline region plate and weld materials of the D.C.Cook Unit 2 reactor vessel are given in Table l.All of the initial RTNDT values (I-RTNDT)are also presented in Table l.

C5521-2 4J I Ci VJ CZ!Lal I 270 10'80 10'0'5556-2 CORE 0~C5540-2 270'0'C C)180'5592-1 Figure l.Identification and Location of Beltline Region Materials for the D.C.Cook Unit 2 Reactor Vessel I

TABLE 1 D.C.COOK UNIT 2 REACTOR VESSEL BELTLINE REGION MATERIAL PROPERTIES Materi al Descri pti on CU (>)NI (%%d)I-RTNDT ('F)Intermediate Shell, C5556-2 0.15 Intermediate Shell, C5521-2*0.125 0.57 0.58 58 38 Lower Shell, C5540-2 Lower Shell, C5592-1 Longitudinal'elds

• Circumferential Weld*0.11 0.14 0.052 0.052 0.64 0.59 0.967 0.967-20-20-35-35*Mean values of copper and nickel as indicated below Material Plate, C5521-2 Data Source Original Hill Test Report Surveillance Program[5]Mean value Copper~wt.t', 0.14 O.ll 0.125 Nickel~wt.I 0.58 0.58 0.58 Weld Original Mill Test Report Surveillance Program[5]Surveillance Program[5]Mean value 0.05 0.055 0.05 0.97 0.97 0.96 0.052 0.967 5.NEUTRON FLUENCE VALUES The calculated fast neutron fluence (E>1 MeV)at the inner surface of the D.C.Cook Unit 2 reactor vessel is shown in Table 2.These values were projected using the results of the Capsule U radiation surveillance program~"~.

TABLE 2 NEUTRON EXPOSURE PROJECTIONS*

AT KEY LOCATIONS IN THE D.C.COOK UNIT 2 PRESSURE VESSEL CLAD/BASE METAL INTERFACE FOR 8.65 AND 32 EFPY[I EFPY 8.65 32 0~0.179 0.663 10'.244 0.902 30'.309 1.14 0.465 1.71*F1 uence x 10 n/cm (E>1.0 MeV)

6.DETERMINATION OF RTPTS VALUES FOR ALL 8ELTLINE REGION MATERIALS Using the prescribed PTS Rule methodology, RTPTS values were generated for all beltline region materials of the 0.C.Cook Unit 2 reactor vessel as a function of present time (8.65 EFPY per Capsule U analysis)and end-of-license 32 EFPY)fluence values.The fluence data were generated based on the most recent sut veillance capsule program results~"~.

The PTS Rule requires that each plant assess the RTPTS values based on plant specific surveillance capsule data under certain conditions.

These conditions are: Plant specific surveillance data has been deemed credible as defined in Regulatory Guide 1.99, Revision 2, and RTPTS values change significantly.(Changes to RTPTS values are considered significant if the value determined with RTPTS equations (I)and (2), or that using capsule data, or both, exceed the screening criteria prior to the expiration of the operating license, including any renewed term, if applicable, for the plant.)For D.C.Cook Unit 2, the use of plant specific surveillance capsule data arises for the intermediate shell plate, C5521-2 and the welds because of the following reasons: I)There have been three capsules removed from the reactor vessel, and the data is deemed credible per Regulatory Guide 1.99, Revision 2.2)The surveillance capsule materials are representative of the actual vessel materials.

The chemistry factors for the intermediate shell plate, C5521-2 and welds were calculated using the surveillance capsule data as shown in Table 3.All other chemistry factor values for the remaining beltline materials were calculated using the Tables I and 2 from Regulatory Guide 1.99, Revision 2.

TABLE 3 CALCULATION OF CHEMISTRY FACTORS USING D.C.COOK UNIT 2 SURVEILLANCE CAPSULE DATA~~Carponent CaPsule F Luence FF DRTNDT FPDRTNDT (FF)"2 Intermediate Shell PLate C5521-2 (Long.)intermediate Shell Plate C5521-2 (Trans.)0.264 0.683 1.06 1.58 0.264 0.683 1.06 1.58 0.638 0.893 1.016 1.126 0.63S 0.893 1.016 1.126 55 90 95 95 80 100 103 138 35.072 80.378 96.548 107.004 51.013 89.309 104.679 155.438 0.407 0.798 1.033 1.269 0.407 0.798 1.033 1.269 719.442 7.012 Chemistry Factor~719.442/7.012 i 102.61 Meld Metal T 0.264 0.638'T 0.683 0.893 X 1.06 1.016 U 1.58 1.126 40 50 70 75 25.507 44.655 71.141 84.477 0.407 0.798 1.033 1.269 225.779 3.506 Chemistry Factor~225.779/3.506~64.40 Tables 4 and 5 provide a summary of the RTPTS values for all beltline region materials for 8.65 EFPY and end-of-life (32 EFPY), respectively, using the PTS Rule.TABLE 4 RTPTS VALUES FOR D.C.COOK UNIT 2 FOR 8.65 EFPY Haterial hRTgpT('F)

+Initial RTNOT+Ha>gin RTPTS (CF x FF*)('F)('F)(F)Intermediate Shell Plate, C5556-2 Intermediate Shell.Plate, C5521-2 Lower Shell Plate, C5540-2 Lower Shell Plate, C5592-1 Intermediate Shell Longitudinal Welds Lower Shell Longitudinal Welds Circumferential Meld Seam 108.35 0.787 86.50 0.787 (102.61)0.787 74.60 0.787 99.55 0.787 70.80'0.618 (64.40)0.618 70.80 0.543 (64.40)0.543 70.80 0.787 (64.40)0.787 58 38 38-20-20-35-35-35-35-35-35 34 34 34 34 34 56 56 56 56 56 56 177 140 (153)73 92 65 61 60 56 77 72 ()Indicates numbers were calculated using surveillance capsule data.*Fluence factor based upon peak inner surface neutron.fluence of 4.65 x 10 n/cm[4], except for the longitudinal welds.For the intermediate shell longitudinal welds, the fluence factor is based on a , neutron fluence of 2.44 x 10 n/cm[4]at the inner surface of the weld at the 10'ocation.

For the lower shell longitudinal welds, the fluence factor is based on a neutron fluence of 1.79 x 10 n/cm[4]at the inner surface of the weld at the 0'ocation.

P 0 TABLE 5 RTPTS VALUES fOR D.C.COOK UNIT 2 FOR 32 EFPY Material hRTNDT('F)

'Ini ti al RTNDT+(CF x FF*)('F)Margin RTPTS ('F)('F)Intermediate Shell Plate, C5556-2 Intermediate Shell Plate, C5521-2 Lower Shell Plate, C5540-2 Lower Shell Plate, C5592-1 Intermediate.

Shell Longitudinal Welds 108.35 1.148 86.50 1.148 (102.61)1.148 74.60 1.148 99.55 1.148 70.80 0.971 (64.40)0.971 58 38 38-20-20-35-35 34 34 34 56 56 216 171 (190)100 128 90 85 Lower Shell Longitudinal Welds Circumferential Weld Seam 70.80 (64.40)70;80 (64.40)0.885 0.885 1.148 1.148-35-35-35-35 56 56 56 56 84 78 102 95 ()Indicates numbers were calculated using surveillance capsule data.*Fluence factor based upon peak inner surface neutron fluence of 1.71 x 10 n/cm[4], except for the longitudinal welds.For the intermediate shell longitudinal welds, the fluence factor is based on a neutron fluence of 9.02 x 10 n/cm[4]at the inner surface of the weld at the 10 location.For the lower shell longitudinal welds, the fluence factor is based on a neutron fluence of 6.63 x 10 n/cm[4]at the inner surface of the weld at the 0'ocation.

gl 7.CONCLUS IONS I As shown sn Tables 4 and 5, all the RTPTS values remain below the NRC screening values for PTS using the fluence values for the present time (8.65 EFPY)and the projected fluence values for the end-of-life (32 EFPY).A plot of the RTPTS values versus the fluence is shown in Figure 2 for the most limiting material, the intermediate shell plate, C5556-2 in the D.C.Cook Unit 2 reactor vessel beltline region.300 250 200 LI 0 M 150 t-CL CC 100 SCREENING CRITERIA~~~~~~~~~~~~~~~0~~~~~~~....k"~~~~~~~~~~~~~.0'"~~~~~~~~~~~~~~~~508.65 EFPY 4 32EFPY 1E+18 2E+'I8 3E+18 5E+18 1E+19 2E+19 3E+19 5E+19 FLUENCE (NEUTRONS/CM)INTER.SHELL PIATE, C5556-2 1E+20 Figure 2.RTPTS versus Fluence Curves for D.C.Cook Unit 2 Limiting Material-Intermediate Shell Plate, C5556-2 eg l J REFERENCES

[1]10 CFR Part 50,"Analysis of Potential Pressurized Thermal Shock Events," July 23, 1985.[2]10 CFR Part 50,"Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," May 15, 1991.(PTS Rule)[3]Regulatory Guide 1.99, Revision 2,"Radiation Embrittlement of Reactor Vessel Materials," U.S.Nuclear Regulatory Commission, May 1988.[4]WCAP-13515,"Analysis of Capsule U from the Indiana Michigan Power Company D.C.Cook Unit 2 Reactor Vessel Radiation Surveillance Program," page 6-28, J.M.Chicots, et al., October 1992.(Westinghouse Proprietary Class 3)[5]WCAP-8512,"American Electric Power Company Donald C.Cook Unit No.2 Reactor Vessel Radiation Surveillance Program", J.A.Davidson, et al., November 1975.[6]MT/SMART-090(89),"D.C.Cook Unit 2 Reactor Vessel Heatup and Cooldown Limit Curves for Normal Operation", N.K.Ray, April 1989, Table l.

ENCLOSURE 2 TO AEP:NRC:1173C MARKUP OF PRESSURIZED THERMAL SHOCK AND UPPER SHELF ENERGY

SUMMARY

TABLES Enclosure.2 to AEP:NRCall73" SucTRary File for Pressurized Theraal Shack PLant Naaa 0 C Cock 1 EOLa}0/25/2014$eltL}r ident.Na-Le Shell$445 Mat" L e Shel L$4405.2 Mc-Le Shel L$4'05 3 Meet Xc.ident.C3$n}0 Xeut fLuence at.ER/EFPT 1ehkfc&l.lO c 18 1ehhakh 0'4'F i LO'E}g~9 8 40-F 1 I}OE}9 Xethcd of Oetarein.}RT XTEX S-Z Xta S-2 NTH 5.Z thaaiatry factor Xechad of Deterge}n CF Table TabLe Table 0.14 0.}4 0.46 0.'5 0.4S int.SheLL~>>-I int.SheLL$4'06-2 int.Shell$4406-3 Lever ShelL$4407 1 Lover SheL L 5447.2 Lover SheLL$4407.3 1.41E19 1.41E19 1.41E19 1.4}E19 1 41E19 1.41E19 SOP 33'F 40'F'2$'F-1Z'F 3S'F Plant Saecif ic Plant Scecific Plant Soecif ic Plant Speci f ic P lant Speci f ic Plant Speci f ic$1.4 104,5 102.94 95.5 TabL~Tabl~Calmlatad Table Tab'Le Table 0.12 0.15 0.15 0.14 0 12 0 14 0.52 0.50 0.49 0 55 0.59 0.50 Xa='.e SheL}.Axial Ve Lda 1-4-2 A 1(.'c-'.e/int.ShelL C}rc.Veld s-c'z 13Z53 and 12OCS (T)20291 8.56'F 1.NK15-56'F i'0'E1$Ceneric Cene ric 206,4 Table Table D,ZS 0.35 0.7>>O.i>>tp Lnt to Lover Shell Circ.Veld 9-4'2 1P 3571 1.4}E19'-56'F Ceneri c 219 Table Q.7'n.ShelL Axial Ve}d 2-4'A/C Lover Shell Axi al Veld 3-4'ZA/C 13Z53 120~&(T)13Z53 and 12OCS (T)56'F O qo r=14~QZ-S&F 0 clcj, 1R Ccneric Ceneric ZO&.4 206.4 Table Tabl e Qe27 8' Sumitary File for Pressurized Theraal Shock Plant Naeke Beltl ine Ident.Heat No Ident.IO Neut.FLuence at EOLIEFPT IRT Hethod of Oetenafn.IRT Chos istry Factor H<<thod of Oetanafn.CF C.C.Caak'I ALL data except aa noted below wre frere the July G, 1992 Letter frere E.E.Ffapatrfck:to T.E.Hurley,"OanaLd C.Cook Nuclear PLant Units 1 and 2...generic I.atter 92-0t, Revfaian 1, Reactar Vessel gtruceuaL Integrity." Information regarding chersfcaL ccapoaftfcn, initial RT, and aethodalagy of detemfnatfan far the lIJSE for'he wide are frere the Novedaer 29, 1993 Letter free E.E.Ffttpatrfck to T.E.Hurley, eganald C.Cook Nuclear plant Units 1 ard 2...Response ta Request for Nitfanal Infonaatian far generfc letter 92-Ot Revisicn t." No value far I Cu af veld 5-442@as pravided, therefore the default value of 0.35 Nas Used.Da t Ld, Sumary File for Pressurized Thermal Shock P lant Xaaa O.C.Cook 2 EOL: 12/23/2017 Bel tl fne ident.int.SheLL 10-1 int.Shell 10-2 Lrwer Shell 9-1 Lo~er SheLL 9-2 int.ShelL Axial Velds L~r Shell Axial'Melds Cf rcus.'Meld Heat Xo.ident.C5556-2 05521-2 C5540 2 C5592-1 53986 S3986 10 Xeut.Fluence at EOLIEFPY'1.71E19 1.71E19~~0

~~~i)1 l~~%~~C~TC',.'~eL,I,,~~4'C',,~RtWRSl%t I 8 SLfmflary File for Upper Shelf Energy PLant Nate Beltlfne ident.Heat No.Natarial 1/4T USE ac EOL 1/4T Xeutran Fluence at EOL Unir red.USE Hethad of Determfn.Unirrad.USE~Rf one D.C.Caak L All data'xŽcapt as noted balan uere frca the July 13, 1992 Letter free E.E.Flttpatrfck ta T.E.Hurley, 49onald C.Cook Nuclear Plant Units,'1 and 2...Generic Letter 92-0t, Revision 1, Reactor VesseL StructuraL integrity."" Infarnatfon regardfng WSE and eethodalogy of detersinatfcn for the LOSE far the mfds are fry the Xavier 29, 1993 Letter frca E.E.Fit~trick to T.E.Hurley, Donald C.Cook Nuclear'Plant Units 1 and 2...Response to Request for Ahfftfcnal fnfomatfan for Gcnerfc Letter 92.01, Revisicn 1.5 O 8~~used'he LJJSE values far sids t-442, 2-442 and 3 442 represent the NRC staff calculated average of D.C.Caak and Sfstar PLant data<NcGufre, Unit 1 and Diablo Canyon, Unit 2)for veLd ufre heats containing heat nas.13253 and f20'.The NSE far uelds 9-44Z fs free MCAP.12519,"Analysis of the Nafne Yankee Reactor Vessel Second Matt Capsule located at 253'," Narch 1991.The QJSE for zelda 0-442 fa free CAP.10756, the Report for SurvefLlance Capauf~U free Ncafre Unit t.

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