ML20064C820

Nonproprietary Evaluation of Pressurized Thermal Shock for Catawba Unit 2

ML20064C820
Person / Time
Site:	Catawba
Issue date:	02/28/1994
From:	Peter P WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20064C806	List: ML20064C802 ML20064C820
References
WCAP-13874, NUDOCS 9403100250
Download: ML20064C820 (21)

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WESTINGHOUSE CLASS 3 (Non Proprietary)

WCAP-13874 I

w EVALUATION OF PRESSURIZED TIIERMAL SilOCK FOR CATAWBA UNIT 2 l

P. A. Peter February 1994 Work Perfonned Under Shop Order XARP-93209 Prepared by Westinghouse Electric Corporation for Duke Power Company Approved by: lhec_f doGu.

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T. A. Meyer, Manager Structural Reliability & Plant Life Optimization WESTINGilOUSE ELECTRIC CORPORATION Nuclear Technology Division P.O. Box 355 Pittsburgh, Pennsylvania 152304)355 01994 Westinghouse Electric Corporation All Rights Reserved

PREFACE This report has been technically reviewed and verified by:

J. M. Chicots

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TAllt.E OF CONTENTS LIST OF TABLES iii LIST OF FIGURES iii 1.0 INTRODUCrlON I

2.0 PRESSURIZED TiiERMAL SIIOCK 2

3.0 METHOD FOR CALCULATION OF RTn.

4 4.0 VERIFICATION OF PLANT-SPECIFIC MATERIAL PROPERTIES 5

5.0 NEUTRON FLUENCE VALUES 9

6.0 DLTERMINATION OF RTn3 VALUES FOR ALL IAELTLINE REGION MATERIALS 10

7.0 CONCLUSION

S 14

8.0 REFERENCES

15 ii

LIST OF TABLES l

TABLEI CATAWBA UNIT 2 REACTOR VESSEL BELTLINE REGION MATERIAL PROPERTIES..........

7 TABLE 2 NEUTRON EXPOSURE PROJECflONS' A'i KEY LOCATIONS ON THE CATAWBA UNIT 2 PRESSURE VESSEL CLAD / BASE METAL INTERFACE FOR 4.516 AND 32 EFPY 9

TABLE 3 CALCULATION OF CHEMISTRY FACTORS USING CATAWBA UNIT 2 I1 SURVEILLANCE CAPSULE DATA 4

TABLE 4 RTm VALUES FOR CATAWBA UNIT 2 FOR 4.516 EFPY 12 13 TABLE 5 RTm VALUES FOP CATAWBA UNIT 2 FOR 32 EFPY 1

LIST OF FIGURES FIGURE 1.

IDENTIFICATION AND LOCATION OF BELTLINE REGION MATERIALS FOR THE CATAWBA UNIT 2 REACTOR VESSEL 6

FIGURE 2.

RTm VERSUS FLUENCE CURVES FOR CATAWBA UNIT 2 LIMITING L

14 MATERI AL - INTERMLDIATE SHELL, B8605-2 iii I

.. ~ -

1.0 INTRODUCTION

A limiting condition on reactor vessel integrity known as Pressurized Themial 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 iractor 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.

i in 1985 the Nuclear Regulatory Commission (NRC) issued a formal ruling on PTS. It established screening criteria on pressurized water reactor (PWR) vessel embritdement as measured by the i

W nil-ductility reference temperature, termed RTm RTers screening values were set for beltline axial welds, forgings or plates and for beltline circumferential weld seams for the end-of-license plant operation. The sctrening 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 I'rS Rule was published in the Federal Register, May 15, 1991 with an effective date of June 14,1991m. This amendment makes the procedure for calculating RTm values consistent with the methods given in Regulatory Guide 1.99, Revision 2m The purpose of this report is to determine the RTm values for the Catawba 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 RTm. Section 4 provides the reactor vessel beltline region material propenics for the Catawba Unit 2 reactor vessel. The neutron fluence values used in this analysis are presented in Section 5. The results of the RTm calculations are presented in Section 6. The conclusions and references for the PTS evaluation follow in Sections 7 and 8, respectively.

1

2.0 PRESSURIZED THERMAL SHOCK fhe FTS Rule requims that the FTS submittal be updated whenever there are changes in core loadings, surveillance measurements or other information that indicates a significant change in projected RTm values.

The Rule outlines reg':lations to address the potential for FTS events on pressurized water reactor l

vessels in nuclear power plants that are operated with a license from the United States Nuclear Regulatory Commission (USNRC). FTS events Lave been shown from operating experience to be transients that asult in a rapid and severe cooldown in the primary system coincident with a high or increasing primary system pressure. The PTS concem arises if c.ne of these transients acts on the beltline ret; ion of a reactor vessel where a reduced fracture resistance exists be ause 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 PWRs:

All plants must submit projected values of RTm for reactor vessel beltline materials by giving values for time of submittal, the expiratien date of the operating license, and the projected expiration date if a change in the operating ?icense or renewal has been requested. This assessment must be ',ubmitted widdn six months after the effective date of this Rule if the value of RTr:s for any matenal 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:

1) the bases for the projection (including any assumptions mgarding core j

loading pattems), and 2) copper and nickel content and fluence values used in the calculations for each beltline material. (If the values differ from those previously submitted to the NRC.justificati.. must be provided.)

2

-l

The RTm (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 wcld materials.

The following equations must be used to calculate the RTm values for each weld, plate or forging in the reactor vessel beltline:

Equation 1:

RTm = I + M + ARTm 82 * ** 0 Equation 2:

ARTm = CF

• f All values of RTm mttst 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 17FS safety analyses are required before a plant is within 3 years of reaching the screening criteria, including analyses of alternatives to minimize the IrrS concern.

NRC approval for operation beyond the sc:rening criteria is required.

1 3

t

- 3.0 MET 110D FOR CALCULATION OF RTm In the l'I'S Rule, the NRC Staff has selected a conservative and uniform method for determining plant-specific values of RTm at a given ame.

For the purpose of comparison with the screening criteria, the value of RTm for the reactor vessel must be calculated for each weld and plate or forging in the beltline region as follows.

RTm = 1 + M + ARTm, where ARTm = CF

• FF I=

Initial reference temperature (RTm)in F of the unirradiated material M=

Margin to be added to cover uncertainties in the values of initial RTm, copper and nickel contents, iluence and calculational procedures.

M = 66 *F for welds and 48 F for base metal if generic values of I are used.

M = 56 'F for welds and 34 F for base metal if measured values of I are used.

FF =

fluence factor = f p2s.o.so w r)a, where 2

f=

Neutron fluence, n/cm (E > 1 MeV at the clad / base metal interface), divided by 10" CF =

Chemistry factor in F from the tablesW for welds and base metals (plates and forgings). If plant-specific surveillance data has been deemed credible per Regulatory Guide 1.99, Revision 2,it may be considered in the calculation of the chemistry factor.

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I 4.0 VERIFICATION OF PLANT-SPECIFIC MATERIAL PROPERTIES l

Before performing the pressurized thermal shock evaluation, a review of the latest plant-specific material properties was performed.

A The beltline region is defined by the I'rS Rule to be "the region of the reactor vessel (shell material including welds, heat-ufceted 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 considered in the selection of the most lirniting material with regan!

to radiation damage " Figure 1 identifies and indicates the location of all beltline region materials for the Catawba Unit 2 reactor vessel.

Material property values were obtained from material test cenifications from the original fabrication as wcil as the additional material chemistry tests performed as part of the surveillance capsule testing H31 program The average copper and nickel values were calculated for each of the beltline region materials using all of the available material chemistry information.

A summary of the pertinent chemical and mechanical properties of the beltline region plate and weld materials of the Catawba Unit 2 reactor vessel are given in Table 1. ' All of the initial RT, values (1-RTm) are also presented in Table 1.

0 l

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(

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f B8806-2 l

0 is 180

.3 oo 101-142B 30o y N

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270 Figure 1. Identification and locanon of Beltline Region Materials for the Catawba Unit 2 Reactor Vessel 6

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TABLE 1 CATAWB A UNIT 2 REACTOR VESSEL BELTLINE REGION MATERIAL PROPERTIES Material Description Cu (%)

• Ni (%)

• I RTa( F)

Intermediate Shell, B8605-1 0.082 0.618 15 using S/C data 15 Intermediate Shell, B8605-2 0.080 0.613 33 Intermediate Shell, B8616-1 0.045 0.595 12 Lower Shell, B8806-1 0.057 0.560 6

Lower Shell, B8806-2 0.057 0.593

-10 Lower Shell, B8806-3 0.057 0593 8

Longitudinal Welds 0.042 0.153

-80 using S/C data

-80 Circumferential Weld 0.042 0.153

-80 using S/C data

-80 Average values of coppe and nickel as indicated in tl'e following tables Reference Inter. Shell, B8605-1 Inter. Shell, B8605-2 Inter. Shell, B8616-1 wt % Cu wt % Ni wt % Cu wt % Ni wt % Cu wt % Ni x

W Surveillance Program 0.071 0.590 W

Capsule Z Report 0.085 0.640 W

0.080 0.610 Chemical Analysis W

0.090 0.630 Chemical Analysis W

0.080 0.610 Chemical Analysis m

Chemical Analysis 0.090 0.620 Letter from CE D23 0.070 0.610 W

Chemical Analysis 0.040 0.600 m

Chemical Analysis 0.050 0.590 Average 0.082 0.618 0.080 0.613 0.045 0.595 7

i l

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I Reference Lower Shell, B8806-1 Lower Shell B8806 2 Lower Shell, B8806-3 wt % Cu wt % Ni wt % Cu wt % Ni wt % Cu wt % Ni I

Chemical Analysis 'l 0.060 0.570 l

IM 0.060 0.550 Chemical Analysis D1 Chemical Analysis 0.060 0.590 i

Chemical Analysis 0.060 0.600 j

DI Chemical Analysis "I

'O.060 0.590 j

t Chemical Analysis 0.060 0.600 ou Letter from CE 02) 0.050 0.560 0.050 0.590 0.050 0.590 1

l Average 0.057 0.560 0.057 0.593 0.057 0.593 Reference Surveillance Weld

• wt % Q m % Ni 141 Surveillance Pmgram 0.N0 0.140 141 Surveillance Program 0.036 0.140 D1 Capsule Z Report 0.051 0.180 Letter fmm CE "1 0.040 I

Average 0.N2 0.153 Per Reference 4, the core region beltline welds are considered to include the intermediate j

1 and lower shell plate longitudinal seams and the joining intermediate to lower shell girth scam. All core region (beltline) welds were fabricated using Weld Wire Heat No. 83648, Linde 0091 Flux, Lot No. 3536.

l 8

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5.0.

NEUTRON FLUENCE VALUES The calculated fast neutron fluence (E>l.0 MeV) at the mer surface of the Catawba Unit 2 reactor vessel is shown in Table 2. These values were projected using the results of the Capsule X radiation surveillance pmgram '1, 'The RTns calculations were performed using the peak fluence value, which D

occurs at the 25 azimuth (except for the longitudinal welds which are located at 30 azimuth) in the Catawba Unit 2 reactor vessel.

TABLE 2 NEUTRON EXPOSURE PROJECTIONS

• AT KEY LOCATIONS ON Tile CATAWBA UNIT 2-PRESSURE VESSEL CLAD / BASE METAL INTERFACE FOR 4.516 AND 32 EFPY "'I EFPY O'

15 25 30 35 45 4.516 0.2M 0.328 0.347 0.209 0.263 0.289 32 1.66 2.32 2.46 L48 1.86 2.05

• Fluence x 10" n/cm (E>1.0 MeV) 2 j

j 9

~.

~

6.0 DETERMINATION OF RTns VALUES FOR A.LL BELTLINE REGION MATERIALS Using the prescribed PTS Rule methodology, RTns values were generated for all beltline region materials of the Catawba Unit 2 reactor vessel as a function of present time (4.516 EFPY per Capsule X analysis) and end-of-life (32 EFPY) fluence values. The fluence data was generated based on the 04 most recent surveillance capsule program results The NS Rule requires that each plant assess the RTm values based on plant specific surveillance capsule data whenever:

Plant specific surveillance data has been deemed credible as defined in Regulatory Guide 1.99, Revision 2. and RTm values change significantly. (Changes to RTm values are considered significant if the value determined with RTm equations (1) and (2), or that using capsule data, or both, cxceed the screening criteria prior to the expiration of the operating license, including any renewed term,if applicable, for the plant.)

For Catawba Unit 2, the use of plant specific surveillance capsule data arises for the Intermediate Shell, B8605-1 and Surveillance Welds because of the following reasons:

1) There have been two 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 intermediate shell and surveillance weld materials.

The chemistry factors for the Intermediate Shell, B8605-1 and Surveillance Welds were calculated using the suiveillance capsule data as shown in Table 3. The chemistry factors for the lower shells

]

and other intermediate shells were calculated using Table 2 from 10 CFR 50.61.

l W

Tables 4 and 5 provide a summary of the RTm values for all beltline region materials for 4.516 EFPY and end-of-license (32 EFPY), respectively, using the MS Rule, l

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TABLE 3 CALCULATION OF CHEMISTRY FACTORS USING CATAWB A UNIT 2 SURVEILLANCE CAPSULE DATA "'I I Capsule 2

Fluence FF ARTm FF*ARTm FF Material L

Inter. Shell, B8605-1 2

3.435 x 10" 0.706 20 14.12 0.498 X

1.19 x 10" 1.055 45 47.48 1.113 Inter. Shell, B8605-1 Z

3.435 x 10" 0.706 40 28.24 0.498 (Trans.)

X 1.19 x 10" 1.055 55 58.03 1.113 Sum:

147.87 3.222 Chemistry Factor = 147.87 + 3.222 = 45.89 Weld Metal Z

3.435 x 10" 0.706 0

0 0.498 X

1.19 x 10" 1.055 35 36.93 1.113 Sum:

36.93 1.611 Chemistry Factor = 36.93 + 1.611 = 22.92 W

11'

TABLE 4 RTn3 VALUES FOR CATAWBA UNIT 2 FOR 4.516 EFPY Material ART 3w ( F)

Initial RT m Margin RTm 3

(CF X FF*)

( F)

( F)

(F)

Intermediate Shell, B8605-1 52.4 0.7083 15 34 86.1 using S/C data (45.89) 0.7083 15 34 (81.5) i Intermediate Shell, B8605-2 51.0 0.7083 33 34 103.1 Intermediate Shell, B8616-1 28.5 0.7083 12 34 66.2 Lower Shell, B8806-1 35.2 0.7083 6

34 64.9 Lower Shell, B8806-2 35.2 0.7083

-10 34 48.9 1

Lower Shell, B8806-3 35.2 0.7083 8

34 66.9 Longitudinal Welds 40.1 0.5800

-80 56

-0.7 using S/C data (22.9~2) 0.5800

-80 56

(-10.7)

Circumferential Weld 40.1 0.7083

-80 56 4.4 using S/C data (22.92) 0.7083

-80 56

(-7.8)

()

Indicates numbers were calculated using surveillance capsule data.

Fluence factor based upon peak inner surface neutron fluence of 3.47 x 10 : n/cm (2.09 x 1

2 10 n/cm for the longitudinal welds) D'l 2

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.. - ~. -.

.- ~.

TABLE 5 RTm VALUES FOR CATAWBA UNIT 2 FOR 32 EFPY Material ARTsw ( F)

Initial RTsw Margin RTm (CF X FF*)

( F)

(F)

( F)

Intermediate Shell, B8605-1 52.4 1.2422 15 34 114.1 using S/C data (45.89) 1.2422 15 34 (106.0)

Intermediate Shell, B8605-2 51.0 1.2422 33 34 130.4 Intermediate Shell, B8616-1

'28.5 1.2422 12 34 81.4 Lower Shell, B8806-1 35.2 1.2422 6

34 83.7 Lower Shell, B8806-2 35.2 1.2422

-10 34 67.7 Lower Shell, B8806-3 35.2 1.2422 8

34 85.7 Longitudinal Welds 40.1 1.1086

-80 56 20.5 using S/C data (22.92) 1.1086

-80 56 (1.4)

Circumferential Weld 40.1 1.2422

-80 56 25.8 using S/C data (22.92) 1.2422

-80 56 (4.5)

()

Indicates numbers were calculated using surveillance capsule data.

Fluence factor based upon peak inner surface neutmn fluence of 2.46 x 10" n/cm (1.48 x 2

10 n/cm for the longitudinal welds) D'l 2

13

7.0 CONCLUSION

S As shown in Tables 4 and 5, all RTm values n main below the NRC screening values for FTS using fluence values for the present time (4.516 EFPY) and projected fluence values for the end-of-license (32 EFPY). A plot of the RTm values versus fluence is shown in Figure 2 for the most limiting material in the Catawba Unit 2 reactor vessel beltline region, Intermediate Shell, B8605-2.

l l

300 l

SCREENING CRITERIA 250 200 iC p 150

...A e

H tr 100 50 e 4.516 EFPY A 32.0 EFPY 0

1E+18 2E + 18 3E + 18 SE+18 1E 419 2E + 19 3E + 19 SE + 19 1E+20 2

FLUENCE (neutrons /cm )

INTER SHELL, B8605-2 Figure 2. RTm versus Fluence Curves for Catawba Unit 2 Limiting Material - Intermediate Shell, B8605 7 l

l 14.

l a

i

8.0 REFERENCES

[1]

10CFR Part 50, " Analysis of Potential Pressurized Thermal Shock Events," July 23,1985.

{

[2]

10CFR Part 50.61, " Fracture Toughness Requirements for Protection Against Pressurized 1

Thennal 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-10868, " Duke Power Company Catawba Unit No. 2 Reactor Vessel Radiation Surveillance Program", L. R. Singer, November 1985.

[5]

WCAP-11941, " Analysis of Capsule Z from the Duke Power Company Catawba Unit No. 2 Reactor Vessel Radiation Surveillance Program" S. E. Yanichko, September 1988.

[6]

Combustion Engineering, Inc. Metallurgical Research and Development Dept. Materials Certification Repon, Revision 2, Contract No. 8871, Job No. 707124-001, dated May 22,1974 and Lukens Steel Company Test Certificate, Mill Order No. 10553-1, dated April 14, 1972.

[7]

Combustion Engineering, Inc. Metallurgical Research and Development Dept. Materials Cenification Report, Revision 1, Contract No. 8871, Job No. 707124-003, dated May 22,1974 arid Lukens Steel Company Test Certificate. Mill Order No. 10553-1, dated April 14, 1972.

[8]

Combustion Engineering, Inc. Metallurgical Research and Development Dept. Materials Cenification Repon, Contract No. 8871, Job No. 707124-011, dated February 22,1973 and Lukens Steel Company Test Cenificate, Mill Order 27495-1, dated January 17,1973.

[9]

Combustion Engineering,Inc. Metallurgical Research and Development Dept. Materials Certification Report, Contract No. 8871, Job No. Rej. Notice 03664-001, dated October 3, 1973 and Lukens Steel Company Test Certificate, Mill Order No. 26602-1, dated April 18, 1973.

15

[10]

Combustion Engineering, Inc. Metallurgical Research and Development Dept. Materials Cenification Repon, Contract No. 8871, Job No. 707124-009, dated August 16,1973 and Lukens Steel Company Test Cenificate, Mill Order No. 26602-1, dated April 18,1973.

[11]

Combustion Engineering, Inc. Metallurgical Research and Development Dept. Materials Cenification Report, Revision 2. Contract No. 8871 Job No. 707124-011, dated August 16, 1973 and Lukens Steel Company Test Certificate, Mill Order No. 26602-1, dated April 26, 1973.

[12]

Letter from Combustion Engineering Metallurgical & Materials Laboratory - Chattanooga to Westinghouse, " Westinghouse Electric Corp. Surveillance Test Material", W. A. House, Contract No. 8871, Pmject No. 960001, dated April 9,1976.

[13]

Letter from Combustion Engineering, Inc. to Westinghouse, " Core Region Weld Materials Contract 161053/8871 & 161054/8971", M. C. Herder, dated April 8,1974 and Combustion Engineering Metallurgical Research and Development Department - Chattanooga, Welding Material Qualification to requirements of ASME Section 111, Job Number D32255, Project Number 960009, December 12,1972. (Code G1.45)

[14]

WCAP-13875, " Analysis of Capsule X from the Duke Power Company Catawba Unit 2 Reactor Vessel Radiation Surveillance Pmgram", E. Terek, dated February,1993.

16

,