Text

Rev 1 to WCAP-15080, Evaluation of Pressurized Thermal Shock for Wolf Creek
	ML20153G285
Person / Time
Site:	Wolf Creek
Issue date:	09/30/1998
From:	Christopher Boyd, Spragg S, Trombola D; WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	;
Shared Package
ML20153G260	List: ML20153G257; ML20153G269; ML20153G277; ML20153G285;
References
	WCAP-15080, WCAP-15080-R01, WCAP-15080-R1, NUDOCS 9809300024
	Download: ML20153G285 (21)
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W:stinghouse Non-Proprietary Class 3 WCAP - 15080 Revision 1 Evaluation of Pressurized Thermal Shock

, for Wolf Creek 1

S. Spragg l

September 1998 I Work Perfomied Under Shop Order K6TP - 108 Prepared by the Westinghouse Electric Company for the Wolf Creek Nuclear Operating Corporation Approved: .

CTH. B6yb, Mana' der i

Equipment & Materials Technology '

Approved:

D.M. Trombola, Manager Mechanical Systems Integration WESTINGHOUSE ELECTRIC COMPANY Nuclear Services Division P.O. Box 355 Pittsburgh, Pennsylvania 15230 - 0355

.. 1 11 l

TABLE OF CONTENTS l l

PREFACE .. .

................................................................ ....................ii TAB LE OF CONTENTS . ........... . . ..... ... ..... . .... .. ... .. ... . ... . .. .... .... . ....................... ... iii

, . LI ST OF TABLES . ...... ..... . . ........ .. ..... ... ........ . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . iv LIST OF FIGURES... . . . ........................................................ ....................y EXECUTIVE S UMMARV. .. . ..... ...... . . .... ......... ... . .. .... ... ...... . .. . . .... .

.................................vi 1 INTRODUCTION = ............................................ . . . . .11 2

PRESS URIZED THERMAL SHOCK RULE ..... .. ......... .. .... . ....... .... ... ......... . .... . ..... 2-1 >

3 METHOD FOR CALCULATION OF RTPTS...........................................................3-1 4

VERIFICATION OF PLANT SPECIFIC MATERIAL F ROPERTIES. .... .. ........ . .. . . . . 4-1 5

NEUTRON FLUENCE VALUES .......... .. . . .. .. .. .. ....... .. . ... .. . .. .. .... ..... . . .. . .. . 5. l 6

DETERMINATION OF RTPTS VALUES FOR ALL BELTLINE REGION MATERIALS... 6-1

? CONCLUSION .. .... .... ................ ...... ... ...................................................................7-1 8

REFERENCES .. .. . ............... .. . ........... .... . .. .. . .......................................81 Oi Introduction

F i

PREFACE

' This report has been technically reviewed and verified by: T.J.Laubham , s 4 m .

7-I I

i 1

IV LIST OF FIGURES l

l l

Figure 1 Identification and Location of Beltline Region Materials for Wolf Creek

, Reactor Vessel ........ ............................................................................42 t

i i

1 4

i e

4

( -.

I-I r-t I-I i

i J

a i . . - . . . , __ . _ . _ __ ..

iii LIST OF TABLES Table 1 Reactor Vessel Beltline Material Unirradiated Toughness Properties... . ..... ... 4-3 Table 2 Fluence (E> 1.0MeV) on the Pressure Vessel Clad / Base Interface for Wolf Creek @ 35 EFPY (EOL) ........... . ... ......... .. ....... ... ... ............. . . . . . 5-1 ;

I Table 3 Calculations of Chemistry Factors using the Wolf Creek Surveillance i i

Capsule Data.- .............................................................................6-1 Tabla 4 Calculation of Chemistry Factors using Surveillance Capsule Data per

Regulatory Guide 1.99, Revis'on 2, Position 2.1.... . ... .... .. ........ .... .... ...... .... 6-2 i

l l

l Tabis 5 RTers Calculations for Wolf Creek Beltline Region Materials

@ 3 5 E F PY .. .... .... ...... . . ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 6-3 Introduction

v EXECUTIVE

SUMMARY

The purpose of this report is to determine the RTns values for the Wolf Creek reactor vessel beltline based upon the results of the Surveillana Capsule V evaluation. The surveillance program test results are desmad cr:dible per the 10 CFR Part 50.61 criteria (see WCAP-15078, Appendix D). Hence, the RTyrs values pr s;nted in this report were cal:ulated per the methodology given in 10CFR50.61 utilizing surveillance data wh;ra applicable. Based upon these conservative assumptions, all o.'the beltline materials in the Welf Creek r:cctor vessel have RTns values below the screening criteria values of 270 F for plates, forgings or longitudinal welds and 300 F for circumferential welds at EOL (35 EFPY).

e p.

e e

Introduction

.O

6 i

l s

G O

l 6

_ _ _ _ _ _ _ _ . il at

1-1 1 INTRODUCTION A Pressurized Thermal Shock (PTS) Event is an event or transient in Pressurized Water Reactors (PW causing severe overcooling (thermal shock) concurrent with significant pressure or followed by a significant re-pr;ssurization in the reactor vessel. A PTS concem arises if one of these transient acts on the beltline region of the reactor vessel where a reduced fracture resistance exists because of neutron irradiation. Such an even may produce a flaw or cause the propagation of a flaw postulated to exist near the inner wall surface, thereby pot;ntially affecting the integrity of the vessel.

The purpose of this report is to determine the RTers values for the Wolf Creek reactor vessel using the results of the surveillance Capsule V evaluation. Section 2.0 discusses the PTS Rule and its requirements. Sections 3.0 provides the methodology for calculating RTers. Section 4.0 provides the reactor vessel beltline region mitsrial properties for the Wolf Creek reactor vessel. The neutron fluence values used in this analysis are pr;sented in Section 5.0. The results of the RTers calculaticr.3 are presented in Section 6.0. The conclusion tnd references for the PTS evaluation follow in Sections 7.0 and 8.0, respectively.

f 9

Introduction

_ J

2-1 2 PRESSURIZED THERMAL SHOCK CRITERIA Tha Nuclear Regulatory Commission (NRC) recentn nended its regulations for light-water-cooled nuclear power plants to clarify several items related to the fracture toughness requirements for reactor pressure

, VIssels, including pressurized thermal shock requirements. The revised PTS RuleD I,10 CFR Part 50.61, was published in the Federal Register on December 19,1995, with an effective date of January 18,1996.

Tha amendment to the PTS Rule makes three changes:

1.

The Rule incorporates in total, and therefore makes binding by rule, the method for determining the reference temperature, rte, including treatment of the unirradiated RTm value, the margin term, and the explicit definition of " credible" surveillance data, which is currently described in Regulatory Guide 1.99, Revision 2:23, 2.

The rule was restructured to improve clarity, with the requiremente section giving only the requirement for the value of the reference temperature for End Of Life (EOL) fluence, RTns.

3. Thermal annealing is identified as a method for mitigating the effects of neutron irradiation, thereby

- rrducing RTns Th3 PTS Rule requirements consist of the following:

-o For each pressurized water nuclear power reactor for which an operating license has been issued, the licensee shall have projected values of RTns, accepted by the NRC, for each reactor vessel beitline material for the EOL fluence of the material.

i o The assessment of RTns must use the calculatior. procedures given in the PTS Rule, and must specify the bases for the projecied value of RTns for each beltline material. The report must specify the copper and nickel contents and the fluence values used in the calculation for each beltline material.

o This assessment must be updated whenever there is a significant change in projected values or of RTns

- upon the request for a change in the expiration date for operation of the facility, Changes to RTns values tre significant if either the previous value or the current value, or both values, exceed the screening criterion prior to the expiration of the operating license, including any renewal term, if Epp!icable for the plant.

Pressurized TMiroal Shock Rule

_ a

2-2 o The RTns screening criterion values for the beltline region are:

270*F for plates, forging and axial weld materials, and 300'F for circumferential weld materials. .

o All available surveillance data must be considered in the evaluation. All credible plant specific surveillance -

data must also be used in the evaluation.

l l

l i

I I

l I

Pressurned Thermal Shock Rule

31 3 METHOD FOR CALCULATION OF RTpts RTers must be calculated for each vessel beltline material using a fluence value, f, which is the EOL fluence for tha material. Equation 1 must be used to calculate values of RTuorfor each weld and plate or forging in the reactor vessel beltline.

RTnr = RTwrcu> + M + ARTxor (1)

Where, RTuorm = Reference Temperature for a reactor vessel material in the pre-service or unirradiated condition.

M =

Margin to be added to account for uncertainties in the values of RTuorm, copper and nickel contents, fluence and calculat;on procedures. M is evaluated from Equation 2.

M = 2dm2 # g 2 (2) ouis the standard deviation for RTuorm.

ou= 0 F when RTuorm is a measured value.

ou= 17'F when RTuorg is a generic value.

ca is the standard deviation for ARTuor.

For plates and forgings:

ca = 17*F when surveillance capsule data is not used.

c6 = 8.5'F when surveillance capsule data is used.

For welds:

c6 = 28 F when surveillance capsule data is not used.

cA = 14 F when surveillance capsule data is used.

can t to exceed one half of ARTuor Method for Calculation of RTm

_J

3-2 ART,or is the mean value of the transition temperature shift, or change in ARTer, due to irradition, and must be calculated using Equation 3.

ARTer = (CF)

f(-'" 8/) (3)

CF (*F) is the chemistry factor, which is a function of copper and nickel content. CF is determined from Tables 1 End 2 of the PTS Rule (10 CFR 50.61). Surveillance data deemed credible must be used to determine a mtterial-specific value of CF. A material-specific value of CF is determined in Equation 5.

fis the higher of the best estimate or calculated neutron fluence, in units of 10" n/cm2 (E > 1.0 MeV), at the clid-base-metal interface on the inside surface of the vessel at the location where the material in question receives the highest fluence. The EOL fluence is used in calculating RTns-in equation 4, calculate a bounding value of RTns using Equation 3 with EOL fluence values for determining ART ns.

RTm = RTar(u> + M + ARTm (4)

To verify that RT,e7for each vessel beltline material is a bounding value for the specific reactor vessel, licensees shall consider plant-specific information that could affect the level of embrittlement. This information includes but is not limited to the reactor vessel operating temperature and any related surveillance program results. Results from the plant-specific surveillance program must be integrated into the RTer estimate if the -

plint-specific surveillance data has been deemed credible.

' s A material-specific value of CF is determined from Equat!on 5.

CI

g, e j(o2s-oaoiosj))

g(o.56-caolosj>) (5)

Method for Calculation of RTns i

. ~ . . _ _ . _ - __ _

3-3 in Equation 5, "A," is the measured value of ARTer and "f," is the fluence for each surveillance data point. If there is clear evidence that the copper and nickel content of the surveillance weld differs from the vessel weld, l.o., differs from the average for the weld wire heat number associated with the vessel weld and the surveillance weld, the measured values of ARTe7 must be adjusted for differences in copper and nickel

' content by multiplying them by the ratio of the chemistry factor for the vessel material to that for the surveillance

. weld.

6 9

4 l

Method for Calculation of RTns

4-1 4 VERIFICATION OF PLANT SPECIFIC MATERIAL PROPERTIES Before performing the pressurized thermal shock evaluation, a review of the latest plant-specific material properties for the Wolf Creek vessel was performed. The beltline region of a reactor vessel, per the PTS Rule,

' is defined as "the region of the reactor vessel (shell material including welds, heat-affected zones and plates and forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor-vsssel that are predicted to experience sufficient neutron radiation damago 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 Wolf Creek reactor vessel.

Th3 best estimate copper and nickel contents of the Lo.'tline neaterials were obtained from WCAP-10015, Rev.

il 1, CE Report NPSD-1039, Rev. 21'la nd letter WCAP-15078ts). The best estimate copper and nickel content is documented in Table 1 herein. These average values were calculated using all of the available material chtmistry information. Initial RTc values for the Wolf Creek reactor vessel beltli a materials are also shown in Table 1.

)

Verifkation of Plant Specific Material Properties

4-2 CIRCUWERENTIAL SEAMS VERTICAL SEAMS 270*

R2OOS-

- 4 101-124B 101-124C 20 s" I

180* /k g , j O*

144" e

~

R2OOS-2 d %(! ,1 P2OOS-l IOI-124A ,

q __ g 90*

= 101-171 15.1" 270*

R2SOO-I 101-1428 IO1-842C 50.8" d f */ 100* J- --

O*

\ \ !

N'13- 3 ,

R2SOO-2 lOI-142A 90-Figure 1 Identification and Location of Beltline Region Materials for Wolf Creek Reactor Vessel Verification of Plant Specific Material Properties

4-3 Tchl31 R 2 Wor Vcssil Biltlins Mat: rial Unirradiat d Toughness Properties e m

. U i MaMrialsDescripti6Q(iA ? Cu wt% ' ;Ni wt%

XLML M 5 Initial RTm(a)7F l Clomre Head Flange.R?501- 1 0.66 20

, . + . -_ .

==

l8 Vessel Flange R2501 -1 0.70 20

_.- .. i-

! Intennedirte Shell Plate R2003 . 0.lM 0.66 -20

=r Intermediate Shell Plate R20t3 - 2 0.(M 0.61 -20 Inten.ediate Shell Plate R2D5-3 Amhe se .h 6 0.05 0.63 -20 N Lowe. She" Plate__f'25A- 1 0.09 0.67 0 Lower SheU Plaie R2508-2 0.06 0.64 10 lower Shell Plate R2508-3 0.09 0.58 40 Intermediate and IowerShell 0.(M 0.08 -50 Iongitudmal Weld Seams (b)

Intermediate toIowerSimil Weld t 0.01 0.08 -50 Seam (b)

Surveillance Weld Metal (b) 0.07 0.10 Notes:

(a) Based on measured data.

(b) All vessel beltline weld seams were fabricate with weld wire heat number 90146. The intermediate to lowe circumferential weld seam 101-171 was fabricated'with Flux Type 124 Lot Number 1061. The intermediate shell longitudinal weld seams 101-124A, B & C and lower shell longitudinal weld seams 101-142A , B & C were fabricated with Flux Type 0091 Lot Number 0842. The surveillance weld metal was fabricated with weld wire heat number 90146, Flux Type 124 Lot Number 1061. Per Regulatory Guide 1.99, Revision 2, " weight-percent copper" and " weight percent nickel

are the best-estimate values for the material, which will normally be the mean of the measured values for a plate or forging or for weld samples made with the weld wire heat number that matches the critical vessel weld. The surveillance weld material was made with the same weld wire hea the vessel beltline v' eld seams and is therefore representative of all of the beltline weld seams.

b Verification of Plant Specific Material Propenies 1

51 5 NEUTRON FLUENCE VALUES The calculated fast neutron fluence (E>1.0 MeV) values at the inner surface of the Wolf Creek reactor vessel are shown in Table 2. These values were projected using the results of the Capsule V radiation analysis. See

, Section 6.0 of the Capsule V analysis report, WCAP -15078f5l

Table 2 Fluence (E> 1.0MeV) on the Pressure Vessel Clad / Base Interface for Wolf Creek @ 35 EFPY (EOL)

' Maferidle m s (llo~c stion(*h l 7 Fluence : (n/crd)':

Intermediate Shell Plate R2005 - 1 30 2.18E+19 Intermediate Shell Plate R2005 - 2 30 2.18E+19 Intermediate Shell Plate R2005 - 3 30 2.18E+19 Lower Shell Plate R2508 - 1 30 2.18E+19 Lower Shell Plate R2508 - 2 30 2.18E+19 Lower Shell Plate R2508 - 3 30 2.18E+19 Intermediate Shell o LongitudinalWeld Seam 101-124A 0 1.15E+19 Intermediate Shell LongitudinalWeld Seams 101- 30 2.18E+19

-142B&101-142C Intermediate to Lower Shell o 30 2.18E+19 Circ. Weld Seam 101-171 Lower Shell Longitudinal Weld 0 1.15E+19 Seams 101-142A Lower Shell Longitudinal Weld o 30 2.18E+19 Seams 101-142B&101-142C (a) These locations are shown graphically in Figure 1.

Neutron Fluence Values

__o

6-1 1

6 DETERMINATION OF RT pts VALUES FOR ALL BELTLINE REGION MATERIALS Using the prescribed PTS Rule methodology, RT n3 values were generated for all beltline regions materials of ths Wolf Creek reactor vessel for fluence values at EOL (35 EFPY).

. Erch plant shall assess the RT ns values based on plant specific surveillance capsule data. For Wolf Creek, surveillance data is only available from the Wolf Creek surveillance program and this data was included in the PTS Evaluation.

As presented in Table 3, chemistry factor values for Wolf Creek based on Average Copper and Nickel weight percent were calculated using Tables 1 and 2 from 10 CFR 50.61D! Additionally, chemistry factor values b:sid on credible surveillance data are calculated in Table 4. Table S contains the RT ns calculations for all beltline region materials at EOL (35 EFPY).

Table 3

. . - . -Calculations of Chernistry Factors using the Wolf Creek Surveillance Capsule Data is

,wy;g# :-FR,L Mater..2alA-y O WVS '

u 4 ya -

" > t/nMW.%y4 (Chemistry; p ; .3,3.4y 3; , wa Compositiont 4

~m --

,e .c

,; . - 7: , & - YFactor;,o- .F;

, p/ a . o

, , mx.'

kNi M di CChiwtW S Note'(a)' '

Intermediate Shell Plate R2005 - 1 0.66 0.04 26 Intermediate Shell Plate R2005 - 2 0.64 0.04 26 Intermediate Shell Plate R2005 - 3 0.63 0.05 31 Lower Shell Plate R2508 - 1 0.67 0.09 58 Lower Shell Plate R2508 - 2 0.64 0.06 37 Lower Shell Plate R2508 - 3 0.58 0.09 58 Beltline Region Weld Metal 0.08 0.04 31.6 Surveillance Weld Metal 0.10 0.07 43.5 (a) Chemistry factors were determined per the Chemistry factor tables in 10CFR50.61.

J l Determination of RTm Values

6-2 Table 4 Calculation of Chernistry Factors using Surveillance Capsule Data per 10CFR50.61

< . , . ..,,_..a..,. g,,., ; g - _ ._ ., _ . _ ._ .

_ -l' ;; $ ? .

Lower Shell U 36.46 25.7 0.5 l0.3429 l 0.705 Plate R2500-3 16.03 17.23 1.16 Y - l 1.308 l 1.075 (Longitudinal) V 2.528 52.03 64.99 1.56 l l 1.249 Lower Shell U 23.79 16.77 0.5 l 0.3429 l 0.705 Plate A9154-1 35.39 38.04 1.16 Y

l 1.308 l 1.075 V 54.53 68.11 1.56 (Transverse) l 2.528 l 1.249 SUM: 230.84 6.44 2

CFe30 3 = E(FF

RTwor) + Z( FF ) = (230.84) + (6.44) = 35.8 F Surveillance Weld U 0.3429 0.705 19.75 13.92 0.5 Material Y 1.308 1.075 32.74 35.2 1.16 V 2.528 1.249 33.64 42.02 1.56 SUM: 91.14 3.22 CFsipm = Z(FF

RTuor) + Z( FF2) = (91.14) + (3.22) = 28.3 F Nstes:

(a) f = Calculated fluence from capsule V dosimetry analysis results (51, (x 10 " n/cm , E > 1.0 MeV).

(b) FF = fluence factor = f m2s..onoso (c) ART my values for lower shell plate R2508-3 are the measured 30 ft-lb shift values given in the capsuls V -

analysis report (WCAP-15078).

(d) The surveillance weld metal ARTuor values have been adjusted by a ratio factor of 0.726 (CF vw / CFsw = 31.6 / 43.5 = 0.726) l l

Detennination of RTm Values

6-3 Tcbl]5 I RT prs Calculations for Wolf Creek Beltline Region Materials @ 35 EFPY Intermediate Shell Plate 2.18E+19 1.21 26.0 31.5 31.5 -20 43 R2005-1 Intermediate Shell Plate 2.18E+19 1.21 26.0 31.5 -20 l 31.5 43 R2005-2 l

Intermediate Shell Plate 2.18E+19 1.21 31.0 37.5 34.0 -20 52 R2005-3 Lower Shell Plate R2508-1 2.18E+19 1.21 58.0 70.2 34.0 0 104 Lower Shell Plate R2508 2 2.18E+19 1.21 37.0 44.8 34.0 10 89 Lower Shell Plate R2508-3 2.18E+19 1.21 58.0 70.2 34.0 40 143 Using S/C Data 2.18E+19 1.21 35.8 43.3 17 40 100 Inter. and Lower Shell Long. 1.15E+19 1.04 31.6 32.9 32.0 -50 16 Weld Seams 101-124A &

101-142A (90* Azimuth)

~ ~ ~ ~ ~ ~ ~ ~ ~

~~~~~ITsIng 5IC Data j.15E+19 504 ~~5.3~ "

2 2954 28I0 -50 7 Inter, and Lower Shell Long. 2.18E+19 1.21 31.6 38.2 38.2 -50 26 Weld Seams 101-124B/C &

101-142B/C (210' & 330' Azimuth)

- ~~

~~~~~ITsing 5/C DaIa~~~~~~ ~2.18E+19 5 21 25.3 34 2

~

28I0 -50 12 Intermediate to Lower Shell 2.18E+19 1.21 31.6 38.2 38.2 -50 26 CircumferentialWeld Seam 101-171

- Using S/C Data 2.18E+19 1.21 28.3 34.2 28.0 -50 12 N:tes:

(a) Initial RT NDT values are measured values (b) RT ns = RT mm + Margin + ARTns (c). ART, .s = CF

FF Determination of RTm Values

7-1 7 CONCLUSIONS All of the beltline materials in the Wolf Creek reactor vessel are well below the screening criteria values of 270'F and 300*F at 35 EFPY.

9 4

4 4

4 Conclusions

8-1 8 REFERENCES

1. 10 CFR Part 50.61, " Fracture Toughness Toughness Requirements for Protection Against Pressurized Thermal Shock Evunts", Federal register, vol. 60, No.243, December 19,1995.

2. Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materia's," US Nuclear Regulatory Commission, May 1988.

3. WCAP-10015, Revision 1, " Kansas Gas and Electric Company Wolf Creek Generation Station Unit 1 Reactor Vessel Radiation Surveillance Program", L.R. Singer, June 1982.

4. CE Report NPSD-1039, Revision 2," Updated Analysis for Combustion Engineering Fabricated Reactor Vessel Welds Best Estimate Copper and Nickel Content", ABB Combustion Engineering Nuclear Operations, June 1998

5. WCAP-15078," Analysis of Capsule V from the Wolf Creek Nuclear Operating Corporation Wolf Creek Reactor Vessel Radiation Surveillance Program", Ed Terek, et.al., August 1998.

Conclusions

.

ML20153G285

Text

SUMMARY

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