ML20071A951

Evaluation of PTS for Byron Unit 1

ML20071A951
Person / Time
Site:	Byron
Issue date:	01/31/1994
From:	Peter P WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20071A952	List: ML20071A951 ML20071A953
References
WCAP-13881, NUDOCS 9403040275
Download: ML20071A951 (18)

Text

_.

.o WESTINGHOUSE CLASS 3 (Non-Proprietary)

WCAP-13881 EVALUATION OF PRESSURT7Fn THERMAL SHOCK FOR BYRON UhTT 1 i

P. A. Peter January 1994 Work Performed Under Shop Order BPPP-108 l

Prepared by Westinghouse Electric Corporation for Commonwealth Edison Company l

l Approved by:MM -

- -3 (T/L.

T. A. Meyer, Manager \\

't Structural Reliability & Plant Life Optimization WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 i

C 1994 Westinghouse Electric Corporation All Rights Reserved 9403040275 940223 PDR ADOCK 05000454 P

PDR l

.~

l 1

i 1

PREFACE l

t f

This report has been technically reviewed and verified by-J. M. Chicots 8,/f 8/E g i

a l

r I

l l

l 1

l 1

l l

l l

i l

r>-

_. _ ~

TABLE OF CONTENTS LIST OP TAB LES..................................................

ili LIST OF FIGURES.................................................. iii l

1.0 INTRODUCTION

I 2.0 PRESSUR f7Fn THERMAL SHOCK...................................... 2 3.0 METHOD FOR CALCULATION OF RTrn.. -

4 1

l l

4.0 VERIFICATION OF PLANT-SPECIFIC MATERIAL PROPERTIES............... 5 l

1 1

5.0 NEUTRON FLUENCE VALUES 9

i 6.0 DETERMINATION OF RTen VALUES FOR ALL BELTLINE REGION MATERIALS 10 I

1 l

7.0 CONCLUS I ONS................................................... 13

)

i 2

4

8.0 REFERENCES

14 j

11

3-i 1;

LIST OF TABLES i

i.

i J

i TABG 1 BYRON UNIT 1 REACTOR VESSEL BELTLINE REGION MATERIAL i

i PROPERTIES................................................. 7 1

J l

TABLE 2 CALCULATION OF ' AVERAGE CU AND NI WEIGHT % USING ALL 1

i PREVIOUS BYRON UNIT 1 CHEMISTRY TEST RESULTS.............

-8 i

I TABE 3 NEUTRON EXPOSURE PROJECTIONS AT KEY LOCATIONS ON THE

!'t BYRON UNIT 1 PRESSURE VESSEL CLAD / BASE hETAL INTERFACE FOR 5.64 AND 32 EFPY........................................ 9 TABE 4 CALCULATION OF CHEMISTRY FACTORS USING BYRON UNIT 1 SURVEILLANCE CAPSULE DATA............................... 11 1

TABG 5 RTm VALUES FOR BYRON UNIT 1 FOR 5.64 EFPY.................. 12 i

TABG 6 RT VALUES FOR BYRON UNIT 1 FOR 32 EFPY.................. 12 en l

5 l

i i'

LIST OF FIGURES i

FIGURE 1 IDENTIFICATION AND LOCATION OF BELTLINE REGION MATERIALS FOR THE BYRON UNIT 1 REACTOR VESSEL............ 6 FIGURE 2 RTm VERSUS FLUENCE CURVES FOR BYRON UNIT 1 LIMITING MATERIAL - INTERMEDIATE SHELL FORGING SP-5933.........

13 1

1 4

!J lii 1

(

1.0 INTRODUCTION

A limiting condition on reactor vessel integrity known as Pressurized Thermal Shock (PTS) may occur l

during a severe system transient stich 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; 1

significant degradation of vessel material toughness caused by radiation 1

embrittlement; and the presence of a critical-size defect in the vessel wall.

l l

In 1985 the Nuclear Regulatory Commission (NRC) issued a fonnal mling on PTS. It established screening criteria on pressurized water reactor (PWR) vessel embrittlement as measured by the 1

W I

nil-ductility reference tem,mrature, termed RT RTm screening values were set for beltline axial m.

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 in accordance with the criteria through end oflicense. The NRC recently 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 14,1991m. This amendment makes the procedure for calculating RTm values consistent with the N

methods given in Regulatory Guide 1.99, Revision 2 The purpose of this report is to detennine the RTm values for the Byron Unit I reactor vessel to address the revised FTS 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 properties for the Byron Unit I 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 FTS evaluation follow in Sections 7 and 8, respectively.

1

i-2.0 PRESSURT7FD THERMAL SHOCK J

f The FIS Rule reqmres that the FTS submittal be updated whenever there are changes in core loadings, i

surveillance measurements or other information that indicates a significant change in projected RTp13 values. The Rule outlines regulations to address the potential for PTS events on pressunzed water reactor vessels in nuclear power plants that are operated with a license from the United States Nuclear I

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 FTS concem arises if one of these transients acts on the

]

beltline region of a reactor vessel where a reduced fracture resistance exists because of neutron j

irradiation. Such an event may result in the propagation of flaws postulated to exist near the inner j

wall surface, thereby potentially affecting the integrity of the vessel.

i i

The Rule establishes the following requirements for all domestic, operating PWRs 1

All plants must submit projected values of RT,73 for reactor vessel beltline materials by giving values for time of submittal, the expiration date of the operating f

license, and the projected expiration date if a change in the operating' license or l

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 criteria. Otherwise,it must be submitted with the next update of the pressure-temperature limits, or the next reactor vessel surveillance capsule repon, or within 5 years from the effective date of this Rule i

change, whichever comes first. These values must be calculated based on the 4

j-methodology specified in this rule. The submittal must include the following:

l 1) the bases for the projection (including any assumptions regarding core loading patterns), and i

l 2) copper and nickel content and fluence values used in the calculations I

for each beltline material (If these values differ from those previously submitted to the NRC, justification must be provided.)

4 4

1 2

- - -, ~ -,

c s

n -

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

270 'F for plates, forgings, axial welds; and r

300 'F for cunimferential weld 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 Equation 2:

ARTm = CF

• f *2* 40 All values of RTm must be verified to be boundmg 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 of alternatives to minimize the PTS concern.

NRC approval for operation beyo"' the screening criteria is required.

1 1

3

3.0 METHOD FOR CALCULATION OF RTm In the FIS Rule, the NRC Staff has selected a conservative and uniform method for determining i

plant-specific values of RTm at a given time. For the purpose of comparison with the screening l

+

criteria, the value of RTm for the reactor vessel must be calculated for each wold and plate or forging in the beltline region as follows.

t RTm = I + M + ARTm, where ARTm = (CF)

• FF I=

Initial reference temperature (RTm) in 'F of the unirradiated material i

M=

Margin to be added to cover uncertamties in the values of initial RTm, copper and t

nickel contents, fluence and calculational procedures in *F.

i M = 66 F for welds and 48 'F for base metalif generic values of I are used.

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

FF =

fluence factor = f "28-*2*"8 0, where j

l f=

Neutron fluence (E>1.0 MeV at the clad / base metal interface), divided by 10 '

l 2

i n/cm j

2 i

CF =

Chemistry factor in F from the tablef for welds and base metals (plates and i

forgings). If plant-specific surveillance data has been deemed credible per Reg.

Guide 1.99, Rev. 2 and is significant, it may be considered in the calculation of the chemistry factor.

I i

i E

4 e ---

r-w

.r

-<---w-w mr-

~+n e

-w---.-

n--

n

= - -, ~,

4.0 VERIFICA~l10N OF PLANT-SPECIFIC MATERIAL PROPERTIES Before performing the pressunzed thermal shock evaluation, a review of the latest plant-specific material properties for the Byron Unit 1 vessel was performed. The beltline region is def"med by the FTS Rulem to be "the region of the reactor vessel (shcIl 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 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 materials for the ByTon Unit I reactor vessel.

Material property values were obtained from material test certifications from the original labrication as well as the additional material chemistry tests performed as part of the sun'eillance capsule testing W

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 peninent chemical and mechanical properties of the beltline region forgings and weld materials of the ByTon Unit I reactor vessel are given in Table 1. All of the initial RTm values (I) are also presented in Table 1.

l

{

l 5

1

l N

l l

t

(

WR-34 l

5 Intermediate Shell Forging SP-5933 1

.5 Core d

WR-18 l

l 5

5 i

g Lower Shell Forging 5P-5951 k

3 4

WR-29 FIGURE 1.

IDENTIFICATION AND LOCATION OF BELTLINE REGION MATERIALS U

FOR THE BYRON UNIT 1 REACTOR VESSEL *1 6

TABLE 1 BYRON UNIT 1 REACTOR VESSEL BELTLINE REGION MATERIAL PROPERTIES Material Desaiption Cu (%)

• Ni (%)

• I (*F)f'l ")

i Intermediate Shell, Forging $P-5933 0.0364 0.747 40 1.ower Shell, Forging 5P-5951 0.04 0.64 10 J

Circumferential Weld Metal, WR-18 0.022.

0.690

-30 J

(a)

Initial RT, values were ><+i==*-d per U.S. NRC Standard Review Plan. The initial RTm j

values for the fonpngs and welds are measured values.

• Average values of copper and nickel as indicated in Table 2 on the following page i

i 7

l TABLE 2 CAlfULATION OF AVERAGE CU AND NI WEIGHT % USING ALL PREVIOUS BYRON UNIT 1 CHEMISTRY TEST RESULTS l

Inter. shen 1.ower shen

(

Forging SP-5933 Forging $P-5951 Weld Metal **

j Rekrence Cu (wt%)

Ni (wt%)

Da (wt%)

Ni(wt%)

Cu (wt%)

Ni (wt%)

i t2 0.05*

0.73*

0.026 0.71 Surveillance Program Capsule U Report'8 0.034 0.73 0.023 0.67 ie 0.032 0.791 0.022 0.665 l

Capsule U Report 1

ie

. 0.021 0.714 Capsule U Report ie 0.021 0.741 l

Capsule U Report l

te 0.022 0.713 Capsule U Report 0.021 0.714 Capsule U Reporge.

I8 0.020 0.704 Capsule U Report l

0B20

' O.694 Capsule U ReporP e

e.020 0.706 Capsule U Report 38 0.021 0.677 Capsule U Report IS 0.023 0.677 -

f l

Capsule U Report 0.021 0.680 Capsule U Report'8 t

I ie 0.021 0.680 Capsule U Report l

4 te 0.021 0.667 Capsule U Report i

te 0,024 0.677 Capsule U Report l

ie 0222 0.697 l

Capsule U Report Capsule U Report'8 0.021 0.634 m

0.04 0.64 Material Cert. Report B&W Quali6catiorP O.024 0.70 m

0.03 0.75 Material Cert. Report Material Cert. Report 0.05 0.73 m

Oerrucal Analysi/tu 0.036 0.735 0.024 0.682 Chemical Analysig u 0.022 0.678 Chemical Analysiliu 0.025 0.705 Average 0.0364 0.747 0.04 0.64 0.022 0.690 bot used m average calculation smcz same values as those trom Keterence W reported only 1or completeness.

The core region girth seam weld metal (WR-18) is type Linde hMoNi, heat number 442002, with a Linde 80 type flux, lot number 8873.

8

5.0 NEUTRON FLUENCE VALUES The calculated fast neutron fluence (E>1.0 MeV) at the inner surface of the Byron Unit I reactor vessel is shown in Table 3. These values were projected using the results of the Capsule X radiation surveillance program"3. The RTns calculations were performed using the peak fluence value, which occurs at the 25' azimuth in the Byron Unit I reactor vessel.

TABLE 3 NEUTRON EXPOSURE PROJECTIONS

• AT KEY LOCATIONS ON THE BYRON UNIT 1 PRESSURE VESSEL CLAD / BASE hETAL INTERFACE FOR 5.64 AhT _^*. FFPY1'l EFPY O'

15 25*

35' 45 5.64 0.2275 0.3434 0.3807 0.3007 0.3418 32 1.290 1.947 2.159 1.705 1.939 1

2

• Fluence in 10" n/cm (E>1.0 MeV) 4 i

1 9

l f

6.0 DETERhENATION OF RTm VALUES FOR ALL BELTLIhT REGION MATERIALS Using the prescribed FTS Rule methodology, RTm values were generated for all beltline region materials of the Byron Unit I reactor vessel for fluence values at the present time (5.64 EFPY per Capsule X analysis) and end of license (32 EFPY). The FTS 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 RT values change significantly. (Changes to RT values are considered significant m

m if the value determmed 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 RTm value changes are not significant for Byron Unit 1, plant specific surveillance capsule data for intermediate shell forging SP-5933 and the circumferential weld metal WR-18 is provided because of the following reasons:

1)

'Rere have been two capsules removed from the reactor vessel, and the data is deemed credible per Regulatory Guide 1.99, Revision 2.

2)

The sutveillance capsule materials are representative of the actual vessel forgings and circumferential weld materials.

The chemistry factors for intermediate shell forging 5P-5933 and the circumferential weld metal were calculated using the surveillance capsule data as shown in Table 4. The chemistry factors were also calculated using Tables 1 and 2 from 10 CFR 50.61*. Tables 5 and 6 provide a summary of the RTm values for all beltline region materials for 5.64 EFPY and 32 EFPY, respectively, using the FTS Rule.

10

,r

a TABLE 4 CALCULATION OF CHEMISTRY FACTORS USING BYRON UNIT 1 SURVEILLANCE CAPSULE DATAW 2

Material Capsule Fluence FF ARTm FF* ART m FF Inter. Shell, Forging U

3.58 x 10'8 0.716 0

0 0.513 5P-5933 l

(Tang.)

X 1.443 x 10 '

1.102 30 33.051 1.214 2

Inter. Shell, Forging U

3.58 x 10

O.716 0

0 0.513 SP-5933 X

1.443 x 10

1.102 30 33.051 1.214 (Mial)

Sum:

66.102 3.454 l

Chemistry Factor = 66.102 + 3.454 = 19.1 28 Circumferential U

3.58 x 10 0.716 0

0 0.513 Weld Metal 2

X 1.443 x 10 '

1.102 35 38.559 1.214 l

(WR-18)

Sam:

38.559 1.727 Chemistry Factor = 38.559 + 1.727 = 22.3 1

11

/

1 TABLE 5 i

RTm VALUES FOR BYRON UNIT 1 FOR 5.64 EFPY Muerid CF ('F)

FF" I ( F)

M ( F)

RTm ( F)

Intermediate Shell Forging 23.8 0.7328 40 34 91.4

$P-5933 Using surv. capsule data" (19.1) 0.7328 40 34 (88.0)

Lower Shell Forging 26 0.7328 10 34 63.1 5P-5951 Weld Metal 29.8 0.7328

-30 56 47.8 Using sury. capsule data" (22.3) 0.7328

-30 56 (42.3) l l

FF (Fluence factor) based upon peak inner surface neutron fluence of 3.807 x 10" n/cm 2

(E>1.0 MeV)!'3 Numbers were calculated using a chemistry factor (CF) based on surveillance capsule data per Regulatory Guide 1.99, Revision 2, Position 2.

TABLE 6 RTm VALUES FOR BYRON UNIT 1 FOR 32 EFPY l

Material CF ( F)

FF" 1( F)

M (*F)

RTm ( F)

Intermediate Shell Forging 23.8 1.209 40 34 102.8 j

5P-5933 Using surv. capsule data" (19.I) 1.209 40 34 (97.1)

Irrer Shell Forging 26 1.209 10 34 75.4 5? 5951 Weld Metal 29.8 1.209

-30 56 62.0 Using sury, capsule data" (22.3) 1.209

-30 56 (53.0)

FF (Fluence factor) based upon peak inner surface neutron fluence of 2.159 x 10" n/cm2 (E>1.0 MeV)t'l Numbers were calculated using a chemistry factor (CF) based on surveillance capsule data per Regulatory Guide 1.99, Revision 2 Position 2.

12

Y.

l 7.0 CONCLUSIDNS -

As shown in Tables 5 and 6, all RTm values remain below the NRC screening values for PTS using l

fluence values for the present time (5.64 EFPY) and projected fluence values for the end of license (32 EFPY). A plot of the RTm values versus fluence shown in Figure 2 illustrates the available margin for the most limiting material in the Byron Unit I reactor vessel beltline region. Intermediate Shell Forging 5P-5933. The surveillance capsule analyses results are also shown.

t I

1 350

.i 300 3

l SCREENING CRITERIA 250 i

f 7'>0 cn H 150 K

l

=

l 100

, = = = = = = e-............ ; - - u n M C C C C C =

50 e 5.64 EFPY A 32.0 EFPY i

i i

i i

i g

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

FLUENCE (neutrons /cm )

f INTERMEDIATE SHELL INTERMEDIATE SHELL USING SURV. CAPSULE DATA 1

FIGURE 2.

RT VERSUS FLUENCE CURVES FOR BYRON UNIT 1 LIMITING m

MATERIAL - INTERMEDIATE SHELL FORGING 5P-5933 13 i

i.

8.0 REFERENCES

[1]

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

[2]

10CFR Part 50.61, " Fracture Toughnns 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-13880, " Analysis of Capsule X from Commonwealth Edison Company Byron Unit 1 Reactor Vessel Radiation Surveillance Program", P. A. Peter, December 1993.

[5]

WCAP-9517. " Commonwealth Edison Co. Byron Unit No.1 Reactor Vessel Radiation Surveillance Program", J. A. Davidson, July 1979.

[6]

WCAP-11651, " Analysis of Capsule U from Commonwealth Edison Co. Byron Unit 1 Reactor Vessel Radiation Surveillance Program", S. E. Yanichko, et al., November 1987.

[7]

Babcock & Wilcox Materials Cenification Report, Westinghouse -NES, hE-24-3 Contract

1. 640000451240, Lower Shell Forging 5P-5951.

[8]

Babcock & Wilcox, " Record of Filler Wire Qualification Test", Test No. WF-336,5/6/74.

[9]

Babcock & Wilcox Materials Certification Repon, Westinghouse -NES, hE-24-2 Contract

1. 640000451240, Intermediate Shell Forging 5P-5933.

[10]

CAE Weld File: B&W Letter No. 004-741 to Westinghouse, " Weld Metal Qualifications",

February 17,1976 and " Byron Unit 1 Reactor Vessel Submerged Arc Weld Seams" table.

[11]

AR #15197, " Alloy Analysis - Steel, Commonwealth Edison Company ByTon Nuclear Plant, Unit 1", L Kardos,11/5/93.

14