ML20092A754

Evaluation of PTS for Braidwood Unit 2

ML20092A754
Person / Time
Site:	Braidwood
Issue date:	03/31/1995
From:	Peter P WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20092A747	List: ML20092A745 ML20092A754 ML20092A761
References
WCAP-14229, NUDOCS 9509110180
Download: ML20092A754 (18)

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9509110180 ADOC h56 J

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.P emm;st55? del' _ - . .

'O WESTINGHOUSE CLASS 3 (Non-Proprietary)

WCAP-14229 O

O EVALUATION OF PRESSURIZED THERMAL SHOCK FOR BRAIDWOOD UNIT 2 O

P. A. Peter

'O March 1995 O W rk Performed Under Shop Order BWSP-108A i

Prepared by Westinghouse Electric Corporation O for the Commonwealth Edison Company Approved by: ~

lk-- ~

AJb '

O R. D. Rishel, Manager Metallurgical & NDE Analysis O

WESTINGHOUSE ELECTRIC CORPORATION Nuclear Technology Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 O

O 1995 Westinghouse Electric Corporation All Rights Reserved O

J

O PREFACE This repon has been technically reviewed and verified by:

E. Terek O

O O

O O

O O

O i

O

3 TABLE OF CONTENTS m

~~'

Section Title Pace LIST OF TABLES . ... . .. . ... . . .... . iii LIST OF FIGURES . . . . . . .. . ... .. . . iii

1.0 INTRODUCTION

. . . . . .. . .. . 1 s

2.0 PRESSURIZED THERMAL SHOCK .. .. . . . .. . ... 2 3.0 METHOD FOR CALCULATION OF RTm . - . - .... 4 m

J 4.0 VERIFICATION OF PLANT-SPECIFIC MATERIAL PROPERTIES . . . . 5 5.0 NEUTRON FLUENCE VALUES . . . . . . . ..... . . . .. 8 J

6.0 DETERMINATION OF RTm VALUES FOR ALL BELTLINE REGION MATERIAL .. . . ........ .... . . .. . . .. .... 9 O

7.0 CONCLUSION

S . . . . . . ... ...... . . . . . .... 12

8.0 REFERENCES

. . .. .. .. ... .. . .. . .. .. ... 13 0

O O

kk O

l LIST OF TABLES l 1

1 g

Table Title Pace i Braidwood Unit 2 Reactor Vessel Beltline Region Material Properties Used in l l

Calculations .. . . .. . . . 7

-)

J l 2

2 Neutron Exposure (n/cm , E>l.0 MeV) Projections at Peak Location on the i Braidwood Unit 2 Pressure Vessel Clad / Base Metal Interface . . . . . 8 3 Calculation of Chemistry Factors Using Braidwood Unit 2 Surveillance Capsule Data . . . .. . . . . . . . ... 10

. 4 PTS Calculations for Braidwood Unit 2 . . 11 O

LIST OF FIGURES O Ficure Title Pace 1 Identification and Location of Beltline Region Material for the Braidwood Unit 2 Reactor Vessel . . .. . 6 0

2 RTers versus EFPY of Operation for Braidwood Unit 2 Limiting Material - Weld Metal 12 O

O iii O

A O

SECTION

1.0 INTRODUCTION

O 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:

O severe overcooling of the inside surface of the vessel wall followed by high repressurization; O -

significant degradation of vessel material toughness caused by radiation embrittlement; and the presence of a critical-size defect in the vessel wall.

0 In 1985 the Nuclear Regulatory Commission (NRC) issued a formal ruling on FTS. It established screening criteria on pressurized water reactor (PWR) vessel embrittlement as measured by the

.O nil-ductility reference temperature, termed RTnsN. RTns 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 screening criteria were determined using conservative fracture mechanics analysis techniques. All PWR vessels in the United States have been required to evaluate vessel embrittlement O in accordance with the criteria through end of license. The NRC recently amended its regulations for light water nuclear power plants to change the procedure for calculating radiation embrittlement. The revised FTS 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 RTns values consistent with the O

methods given in Regulatory Guide 1.99, Revision 2m The purpose of this report is to determine the RT ys values for the Braidwood Unit 2 reactor vessel to o address the revised FTS Rule. Section 2.0 discusses the Rule and its requirements. Section 3.0 provides the methodology for calculating RTys. Section 4.0 provides the reactor vessel beltline region material properties for the Braidwood Unit 2 reactor vessel. The neutron fluence values used in this analysis are presented in Section 5.0. The results of the RT ns calculations are presented in Section O 6.0. The conclusions and references for the FTS evaluation follow in Sections 7.0 and 8.0, respectively.

I

'O

~_ . _ _ _ _ _ . _ _ . . _ . . _ _ - .- _. _

O SECTION 2.0 PRESSURIZED THERMAL SHOCK O

The IrTS Rule requires that the FTS submittal be updated whenever there are changes in core loadings, l

surveillance measurements or other information that indicates a significant change in projected RTm l values. The Rule outlines regulations to address the potential for I'TS events on pressurized water O reactor vessels in nuclear power plants that are operated with a license from the United States Nuclear Regulatory Commission (USNRC), ITS 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 l'TS concern arises if one of these transients acts on the O 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.

O The Rule establishes the following requirements for all domestic, operating PWRs:

All plants must submit projected values of RTm for reactor vessel beltline material by 9 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 RTm for any material is projected to exceed the screening O 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 O

the following:

1) the bases for the projection (including any assumptions regarding core loading O P "**S)'""d

2) copper and nici el content and fluence values used in the calculations for each beltline material. (If these values differ from those previously submitted to the O NRC, justification must be provided.)

2 O

B 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 weld material.

The following equations must be used to calculate the RTm values for each weld, plate 9 or forging in the reactor vessel beltline:

Equation 1: RTrrs = 1 + M + A RTers Equation 2: """80 A RTrrs = CF

• f ""8 -

l O

• All values of RTm must be verified to be bounding values for the specific reactor vessel. l In doing this each plant should consider plant-specific information that could affect the level of embrittlement n

./

Plant-specific l'TS safety analyses are required before a plant is within 3 years of reaching the screening criteria, including analyses of alternatives to minimize the PTS concern.

• NRC approval for operation beyond the screening criteria is required.

3 n

a O

O 3

0

J SECTION 3.0 METHOD FOR CALCULATION OF RT,,n D

In the I'rS Rule, the NRC Staff has srlected a conservative and uniform method for determining plant-specific values of RT,ys at a given time. For the purpose of comparison with the screening criteria, the value of RT,.n for the reactor vessel must be calculated for each weld and plate or forging

.D in the beltline region as follows.

RTru = 1 + M + ART,n, where ARTrn = CF

• FF O I= Initial reference temperature (RTym)in F of the unirradiated material M= Margin to be added to cover uncertainties in the values of initial RTsm, copper and nickel

, contents, fluence and calculational procedures, per Regulatory Guide 1.99, Revision 2, in F.

J M = margin = 2 0 3 + 2 o,2, F o i = 0 F when I is a measured value 9 o, = 17 F when I is a generic value For plates and forgings:

03 = 17 F when sun'elllance capsule data is not used c3 = 8.5 F when sun'elllance capsule data is used O For welds:

03 = 28 F when surveillance capsule data is not used c3 = 14 F when surveillance capsule dcta is used o 3not to exceed 0.5* ARTym Li

'*2"'

FF = fluence factor = f '"8 0, where f= Neutron fluence (E>l.0 MeV) at the clad / base metal interface divided by 10" n/cm 2 O

CF = Chemistry Factor in F from the tabled 2; for welds and base metals (plates and forgings). If plant-specific surveillance data from two or more suneillance capsules has been deemed credible per Regulatory Guide 1.99, Revision 2, it should be considered in the calculation of O the chemistry factor.

4 0

3 SECTION 4.0 VERIFICATION OF PLANT-SPECIFIC MATERIAL PROPERTIES a_-

Before performing the pressurized thermal shock evaluation, a review of the latest plant-specific material properties for the Braidwood Unit 2 vessel was performed. The beltline region is defined by the PTS Rulem to be "the region of the reactor vessel (shell mwilal including welds, heat-affected O 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 limiting material with regard to radiation damage." Figure 1 identifies and indicates the location of all beltline region material for the Braidwood Unit 2 reactor O vessel.

Material 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 testing O,

program. The copper and nickel values used in the calculations were obtained from Reference 5. A summary of the pertinent chemical and mechanical properties of the beltline region forgings and weld material of the Braidwood Unit 2 reactor vessel are given in Table 1.

O O

O O

O 5

O

O O

3 90' Upper Shell Forging 49D963-1/49C904-1 0

0* 180*

O

@U C WP562 270*

90' Lower Shell Forging 50D102-1/50C97-1 0- 180*

O 270' O

O

'O Figure 1 Identification and Location of Beltline Region Material for the Braidwood Unit 2 Reactor Vessel'*1 6

.O

O TABLEI Braidwood Unit 2 Reactor Vessel Beltline Region Material Properties Used in Calculations Material Cu% Ni% Chemistry Factor Initial RTum"'

Upper Shell Forging 0 03 0.71 200*' -30

!>>wer Shell Forging 0.06 0.77 37.0*' -30 13.3 -30 g _ using S/C data Weld Metal 0.03 0.71 41.0*' 40 using S/C data 12 8"' 40 NOTE j I

(a) Initial RTum values are measured values.

g (b) Chemistry Factor calculated per Regulatory Guide 1.99, Revision 2, Position 1.1.

(c) Chemistry Factor calculated per ReFulatory Guide 1.99, Revision 2, Position 2.1.

3 O

O i

O

!O 1

l l

l 7

O

.O SECTION 5.0 NEUTRON FLUENCE VALUES O

The calculated peak fast neutron fluence (E>1.0 MeV) values at the inner surface of the Braidwood Unit 2 reactor vessel are shown in Table 2. These values were projected using the results of the surveillance Capsule X radiation analysis'4 The RTru calculations were performed using peak O fluence values, which occur at the 25 azimuthal angle of the Braidwood Unit 2 reactor vessel.

TABLE 2 0 Neutron Exposure (n/cm', E>1.0 MeV) Projections at Peak Imation on the Braidwood Unit 2 Pressure Vessel Clad / Base Metal Interface O 25' EFPY Base & Weld Metal 4.215 2.8% x 10

32 2.199 x 10" O 48 3.298 x 10" O

O O

O 8

O

-O SECTION 6.0 DETERMINATION OF RTm VALUES FOR ALL BELTLINE REGION MATERIAL

'O Using the prescribed PTS Rule methodology, RTm values were generated for all beltline region material of the Braidwood Unit 2 reactor vessel for fluence values at the present time (4.215 EFPY per Capsule X analysis), end-of-license (32 EFPY), and 48 EFPY. The PTS Rule requires that each plant O assess the RTrrs 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 0

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, exceed the screening criteria prior to the expiration of the operating license, including any renewed O

term, if applicable, for the plant.)

Although the RTers value changes are not significant for Braidwood Unit 2, plant specific surveillance '

'O capsule data for the lower shell forging and weld metal is provided 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.

-0

2) The surveillance capsule materials are representative of the actual vessel forging and weld material.

O The chemistry factors for the lower shell forging and weld metal were calculated using the surveillance capsule data as shown in Table 3. The chemistry factors were also calculated using Tables 1 and 2 from 10 CFR 50.61m. Table 4 provides a summary of the RTm values for all beltline region material O for 4.215,32, and 48 EFPY.

9 O

O TABLE 3 Calculation of Chemistry Factors Using Braidwood Unit 2 Surveillance Capsule Data W v"

Material Capsule f FF A RTwin PMTuor W Lower Shell Forging (Tangential) U 0.3933 0.741 0 0.000 0.550 X 1.126 1.033 3 3.099 1.067

' *er shell Forging (Axial) U 0.3933 0.741 5 3.707 0.550 0

X 1.126 1.033 35 36.160 1.067 Sum: 42.996 3.234 2

CF = [(FF

• RTum) + [(FF ) = 13.3 O weid Metai u 0.3933 0.74i 0 0.000 0.350 X 1.126 1.033 20 20.663 1.067 Sum: 20.663 1.617 CF = [(FF

• RTurn) + [(FF )2 , 3 2,g Q NOTES:

f = Iluence + 10 n/cm ;2 All values taken from Section 6.0 of Reference 5.

FF = fluence factor = f ""8 ' **

• O Chemistry Factor for the Lower Shell Forging based on surveillance capsule data = 13.3 F Chemistry Factor for the Weld Metal based on surveilfance capsule data = 12.8 F O

O O

O 10 0  ;

1

O TABLE 4 I'TS Calculations for Braidwood Unit 2 O Material cu% Ni% CF f FF I M ARTm RTm 4.215 EFPY Upper Shell Forging 0.03 0.71 20.0 0.2896 0.6612 -30 13.22 13.22 -3.6 Lower Shell Forging 0.06 0.77 37.0 0.2896 0.6612 -30 24.46 24.46 18.9

.O using S/C data 13.3 0.2896 0.6612 -30 8.79 8.79 -12.4 Weld Metal 0.03 0.71 41.0 0.2896 0.6612 40 27.11 27.11 94.2 using S/C data 12.8 0.2896 0.6612 40 8.46 8.46 56.9 0 32 EFPY Upper Shell Forging 0.03 0.71 20.0 2.199 1.214 30 24.28 24.28 18.6 Lower Shell Forging 0.06 0.77 37.0 2.199 1.214 -30 34 44.92 48.9 using S/C data 13.3 2.199 1.214 -30 16.15 16.15 2.3 O 40 49.77 49.77 139.5 Weld Metal 0.03 0.71 41.0 2.199 1.214 using S/C data 12.8 2.199 1.214 40 15.54 15.54 71.1 48 EFPY Q Upper Shell Forging 0.03 0.71 20.0 3.298 1.313 -30 26.26 26.26 22.5 Lower Shell Forging 0.06 0.77 37.0 3.298 1.313 -30 34 48.58 52.6 using S/C data 13.3 3.298 1.313 -30 17 17.46 4.5 Weld Metal 0.03 0.71 41.0 3.298 1.313 40 53.83 53.83 147.7

'O using S/C data 12.8 3 298 1.3i3 40 i6.8i 16.8i /3.6 NOT[;,;

initial RT,arr values are measured values.

All of the beltline rnaterial in the Braidwood Unit 2 reactor vessel is below the screening criteria at 32 and 48 EFPY. The following plot of RTrrs versus fluence (Figure 2) also illustrates the available margin, where RTrrs = I + M + ARTers-O lO 11 0

lO -

SECTION

7.0 CONCLUSION

S o

As shown in Table 4, all RTm values remain benw the NRC IrFS screening criteria values using fluence values for the present time (_4.215 EFPY), end-of-license (32 EFPY), and 48 EFPY. A plot of the RTm values versus fluence, shown in Figure 1, illustrates the available margin for the most O limiting material in the Braidwood Unit 2 reactor vessel beltline region, the weld metal WF-562.

8 O

Screening Criteria 300 -

O' 250 -

C w

O u) p 200 -

CL, E

150 -

O -

Weld M 100 g O e=a- * * '

weld Metal usina surv 1 ;.

50 2.896E+18 2.199E+19 3.298E+19 Fluence (n/cm2)

O O Figure 2 RTm versus EFPY of Operation for Braidwood Unit 2 Limiting Material - Weld Metal 12 O

O SECTION

8.0 REFERENCES

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[1] 10 CFR Part 50, " Analysis of Potential Pressurized Thermal Shock Events," July 23,1985. ,

[2] 10 CFR Pan 50.61, " Fracture Toughness Requirements for Protection Against Pressurized O Thermal Shock Events," May 15,1991. (IYTS Rule)

[3] Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials,"

U.S. Nuclear Regulatory Commission, May 1988.

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[4] WCAP-11188, " Commonwealth Edison Company Braidwood Station Unit No. 2 Reactor Vessel Radiation Surveillance Program". L. R. Singer, December 1986.

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[5] WCAP-14228. " Analysis of Capsule X from the Commonwealth Edison Company Braidwood Unit 2 Reactor Vessel Radiation Surveillance Program", P. A. Peter, et al., January 1995.

O 16] WCAP-13070, " Evaluation of Pressurized Thermal Shock for Braidwood Units 1 & 2", M. A.

Ramirez, et al., September 1991.

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