ML20092A761

Evaluation of PTS for Braidwood Unit 1

ML20092A761
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-14242, NUDOCS 9509110182
Download: ML20092A761 (18)

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O WESTINGHOUSE CLASS 3 (Non-Proprietary)

WCAP-14242 O

O EVALUATION OF PRESSURIZED THERMAL SHOCK FOR BRAIDWOOD UNIT 1 O

P. A. Peter O

March 1995 O W rk Performed Under Shop Order BWSP-108 Prepared by Westinghouse Electric Corporation O for the Commonwealth Edison Company Approved by: UW 1l.

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

WESTINGHOUSE ELECTRIC CORPORATION Nuclear Technology Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 0

01995 Westinghouse Electric Corporation All Rights Reserved O

O PREFACE O

This report has been technically reviewed and verified by:

./

P. L. Strauch 8 /wd soA r. L 5.

O O

O

'O

.O f

,O l

lO i

O i

O

i l

-O. j i

l TABLE OF CONTENTS i l

O' Section Title Pgg LIST OF TABLES . ........ . . .... ....... ...... ... ..... iii O LIST OF FIGURES . . . . . . . . . ... .. . . ...... . . . . . . . . . . . . . . iii

1.0 INTRODUCTION

.. . .. . . ..... .. . ............. 1 0

2.0 PRESSURIZED THERMAL SHOCK . . . . . . . . . ..... ...... 2 3.0 METHOD FOR CALCULATION OF RTm- - - - -- 4 0

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

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

7.0 CONCLUSION

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

8.0 REFERENCES

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

I O ,

'O ii  !

O

O LIST OF TABLES TabJs Title Eags O

1 Braidwood Unit i Reactor Vessel Beltline Region Material Properties Used in Calculations ..... .. .. . . . ..... ... ... ........ ... . 7

'O 2 Neutron Exposure (n/cm2 , E>1.0 MeV) Projections at Peak Location on the Braidwood Unit 1 Pressure Vessel Clad / Base Metal Interface . . . . . . . . . . . . . . . . . . . 8 0 3 Calculation of Chemistry Factors Using Braidwood Unit 1 Surveillance Capsule Data . . .. . . ..... .... ... ... ........ . .......... ... 10 4 FTS Calculations for Braidwood Unit 1 ........ . . ........ .... .... 11 O ,

LIST OF FIGURES i

O Ficure Title Eage 1 Identification and Location of Beltline Region Material for the Braidwood Unit 1 Reactor Vessel ... . . .. .. . ..... . .......... ............. 6 l O '

l 2 RTm versus Fluence Curves for Braidwood Unit i Limiting Material - Weld ,

M etal . . . . . .. ... . . . . ...... .............. . ...... 12  !

-O O  :

i l

1 kki 1

O l i

C

O SECTION

1.0 INTRODUCTION

O A limiting condition on reactor vessel integrity known as Pressurized Thermal Shock (IrrS) 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.

In 1985 the Nuclear Regulatory Commission (NRC) issued a formal ruling on IrFS. It established screening criteria on pressurized water reactor (PWR) vessel embrittlement as measured by the nil-ductility reference temperature, termed RTerslH. RTm screening values were set for beltline axial O l 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 0 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 IrrS Rule was published in the Federal Register, May 15,1991 with an effective date of June 14,1991W This amendment makes the procedure for calculating RTm values consistent with the O

methods given in Regulatory Guide 1.99, Revision 2*

The purpose of this report is to determine the RTm values for the Braidwood Unit I reactor vessel to address the revised IrrS Rule. Section 2.0 discusses the Rule and its requirements. Section 3.0

-O ,

provides the methodology for calculating RTns. Section 4.0 provides the reactor vessel beltline region material properties for the Braidwood Unit I reactor vessel. The neutron fluence values used in this analysis are presented in Section 5.0. The results of the RTrrs calculations are presented in Section O 6.0. The conclusions and references for the IrTS evaluation follow in Sections 7.0 and 8.0, respectively.

I  !

O l I

l

O SECTION 2.0 PRESSURIZED THERMAL SHOCK

'O The IYrS Rule requires that the IrTS 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 regulations to address the potential for PTS 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). IrFS 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 FTS 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 O Si ving 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 the following:

1) the bases for the projection (including any assumptions regarding core loading g 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 O NRC, justification must be provided.)

2 0

O

• The RTm (measure of fracture resistance) screening criteria for the reactor vessel beltline region is:

270 F for plates, forgings, axial welds; and

.O 300*F for circumferential weld material.

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

Equation 1: RTm = I + M + A RTrrs Equation 2: $2s aioica n A RTm = CF

• f O

• All values of RTm 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.

O Plant-specific PTS safety analyses are required before a plant is within 3 years of reaching the screening criteria, including analyses of attematives to minimize the FTS concem.

'O

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

.O O

O

.O 3

lO

O-SECTION 3.0 METHOD FOR CALCULATION OF RTm O

In the IrrS Rule, the NRC Staff has selected a conservative and uniform method for determining plant-specific values of RTm at a given time. 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 O in the beltline region as follows.

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

• FF O I= Initial reference temperature (RTsm)in F of the unirradiated material M= Margin to be added to cover uncertainties in the values of initial RTsm, copper and nickel c ntents, fluence and calculational procedures, per Regulatory Guide 1.99, Revision 2, in *F.

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

03 = 17 F when surveillance capsule data is not used o3 = 8.5*F when surveillance capsule data is used O For welds:

03 = 28 F when surveillance capsule data is not used o3 = 14 F when surveillance capsule data is used 03 not to exceed 0.5* ARTsm FF = fluence factor = f a2s aioio,o, where f= Neutron fluence (E>l.0 MeV) at the clad / base metal interface divided by 10" n/cm2 O

CF = Chemistry Factor in *F from the tables t2: for welds and base metals (plates and forgings). If plant-specific surveillance data from the surveillance program'53 has been deemed credible per Regulatory Guide 1.99, Revision 2, it may be considered in the calculation of the chemistry O factor.

4 O

l

.O l SECTION 4.0 l l

VERIFICATION OF PLANT-SPECIFIC MATERIAL PROPERTIES O i Before performing the pressurized thermal shock evaluation, a review of the latest plant-specific ,

material properties for the Braidwood Unit i vessel was performed. The beltline region is defined by the PTS Rulet21 to be "the region of the reactor vessel (shell material 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 l identifies and indicates the location of all beltline region material for the Braidwood Unit I 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 program"I. 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 I reactor vessel are given in Table 1.

O  :

O  ;

O t I

O

)

O  ;

i 5 i

'O

O l

o 0 Forging 90 49D383/ 49C344/ l-1 O

w

.O 0 -

180 0 g 0 W

m 5

0 $

0 270 CORE ( WF562 0

0 90 Forging 49D867/ 49C813/ l-1 d

W

-O E 0 0 0

-- 180 a

0 0

270

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

i Vessel'63 6

O

~O  !

TABLE 1 Braidwixxl Unit i Reactor Vessel Beltline Region Material Properties Used in Calculations O Material Cu% Ni% Chemistry Factor Initial RTyn*'

intermediate Shell Forging 0.05 0.73 31.0 * -30 lewer Shell Forging 0.l4 0.74 26.0* -20 using S/c data i s 8*' -20 0

Weld Metal 0.03 0 67 41.0* 40 using S/C data 20.6"* 40 NOII..

(a) Initial RTmy values are measured values.

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

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

O O

O 1 O

O

O 7

O

l O

SECTION 5.0 '

NEUTRON FLUENCE VALUES O

The calculated peak fast neutron Buence (E>l.0 MeV) values at the inner surface of the Braidwood Unit 1 reactor vessel are shown in Table 2. These values were projected using the results of the Capsule X radiation surveillance program"1 The RTrrs calculations were performed using peak O fluence values, which occur at the 25 azimutbal angle of the Braidwood Unit I reactor vessel.

TABLE 2 O Neutron Exposure (n/cm2 , E>l.0 MeV) Projections at Peak Location on the .

Braidwood Unit 1 Pressure Vessel Clad / Base Metal Interface o 25-EFPY Base & Weld Metal 4.234 2.963 x 10

32 2.239 x 10

O 48 3.358 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 17FS Rule methodology, RTrrs values were generated for all beltline region material of the Braidwood Unit I reactor vessel for fluence values at the present time (4.234 EFPY per Capsule X analysis), end-of-license (32 EFPY), and 48 EFPY. The FTS Rule requires that each plant O 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 0

RTm values change significantly. (Changes to RT m 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 term, if applicable, for the plant.)

Although the RTrrs value changes are not significant for Braidwood Unit 1, 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.

O

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 50.6112' Table 4 provides a summary of the RT m values for all beltline region material o for 4.234,32, and 48 EFPY.

O 9

O

~

O 1

TABLE 3 Calculation of Chemistry Factors Using Braidwood Unit i Surveillance Capsule Datal "

O Material Capsule f FF ARTuy FF* A RTm FF2 l l

lower Shell ForF ing (Tangential) 11 0.3814 0.733 5 3.666 0.538 X 1.144 1.038 30 31.127 1.077

wer Shell Forging (Axial) 11 0.3814 0.733 0 0.000 0.538 O

X 1.144 1.038 25 25.939 1.077 Sum: 60.733 3.228 CF = [(FF

• RTmn) + E(FF2) , 3g g Q Weld Metal U 0.3814 0.733 10 7.333 0.538 X 1.144 1.038 25 25.939 1.077 Sum: 33.272 1.614 2

CF = E(FF

• RTen) + [(FF ) = 20.6

/Q NOTF.5:

f = fluence + 10 n/cm ;2 All values taken from Section 6 0 of Reference 5.

FF = fluence factor = f '" '"

O

. Chemistry Factor for the Lower Shell Forging based on surveillance capsule data = 18.8 F

. Chemistry Factor for the Weld Metal based on surveillance capsule data = 20.6 F O

O

-O

~O 10 0

O TABLE 4 l'TS Calculations for Braidwood Unit i O uateriai Cu* Nis CF r rF i u a nin, arm, 4.234 EFPY inter. Shell Forging 0.05 0.73 31.0 0.2963 0.6671 -30 20.68 20.68 11.4 L wer Shell ForF ing 0.M 0.74 26.0 0.2963 0.6671 -20 17.34 17.34 14.7 O

using S/C data 18.8 0.2963 0.6671 -20 12.54 12.54 5.1 Weld Metal 0.03 0.67 41.0 0.2963 0.6671 40 27.35 27.35 94 7 using S/C data 20.6 0.2963 0.6671 40 13.74 13.74 67.5 O 32 EFPY Inter Shell Forging 0.05 0.73 31.0 2.239 1.218 -30 34 37.77 41.8 lower Shell Forging 0.04 0.74 26.0 2.239 1.218 -20 31.68 31.68 43.4 using S/C data 18.8 2.239 1.218 -20 17 22.90 19.8 O

Weld Metal 0.03 0.67 41.0 2.239 1.218 40 49.95 49.95 139.9 using S/C data 20.6 2.239 1.218 40 25.10 25.10 90.2 48 EFPY Z

O Inter. Shell Forging 0.05 0.73 31.0 3 358 1.317 -30 34 40.83 44.8 lower Shell Forging 0 04 0.74 26.0 3.358 1.317 -20 34 34.25 48.3 using S/C data 18.8 3.358 1.317 -20 17 24.76 21.8 Weld Metal 0.03 0.67 41.0 3.358 1.317 40 54.00 54.00 148.0 O using S/C data 20.6 3.358 i .3 i7 40 27.i3 27.13 94.3 NTO I .

Initial RTer values are measured values.

O All of the beltline material in the Braidwood Unit I 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 RT i ,s = 1 + M + A RTers.

O O

l l

11 O

O SECTION

7.0 CONCLUSION

S O

As shown in Table 4, all RTers values remain below the NRC PTS screening criteria values using fluence values for the present time (4.234 EFPY), end-of-license (32 EFPY), and 48 EFPY, A plot of the RT ers values versus fluence, shown in Figure 2, illustrates the available margin for the most

'O limiting material in the Braidwood Unit I reactor vessel beltline region, the weld metal WF-562.

l 350 O

Screening Criteria 300 -

0 250 -

v

• ~

O CL F-(E 150 -

=

O Weld M 100 W .o.

'O so 2.963E+18 2.239E+19 3.358E+19 Fluence (n/cm2)

.O O Figure 2 RTrrs versus Fluence Curves for Braidwood Unit 1 Limiting Material - Weld Metal 12

'O t

O SECTION

8.0 REFERENCES

l O

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

[2] 10 CFR Part 50.61, " Fracture Toughness Requirements for Protection Against Pressurized O 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.

O

[4] WCAP-9807, " Commonwealth Edison Company Braidwood Station Unit No.1 Reactor Vessel Radiation Surveillance Program". S. E. Yanichko, et al., February 1981.

O

[5] WCAP-14241, " Analysis of Capsule X from the Commonwealth Edison Company Braidwood Unit 1 Reactor Vessel Radiation Surveillance Program", P. A. Peter, et al., December 1994.

O [6] WCAP-13070, " Evaluation of Pressurized Thermal Shock for Braidwood Units 1 & 2", M. A.

Ramirez, et al., September 1991.

O O

O O

13

.O