ML20046D236

Nonproprietary Evaluation of PTS for Catawba Unit 1

ML20046D236
Person / Time
Site:	Catawba
Issue date:	08/31/1993
From:	Meyer T, Peter P WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20046D237	List: ML20046D236
References
WCAP-13763, NUDOCS 9308170039
Download: ML20046D236 (17)

Text

. ffW%ifR77MWK, KW.": ,

7;M W aim m a p

J ggy j

7 3

M r g,; o,

,y,:x ~ %sw,g.4.a )y m

y; e v c

c,.

e 3

y e...

- p.

s -

r

,~

c

,y;,;

, #QW j f

~

cl, }.MLuauswQ,ip

,j!f <

q a<

~

m 3a

~

L

'l%

Y

~

[, ' Mmwrd wM

?

" (':; ' lh

~

~~

k.3

^

Wi l n ~[.. '

o

J ' ;g;; ppa

y 3:g

~

V-x',

AW A,

Qa w

g c.

w eq (Qlqp;,g,$gs

, 5:sp;;N

>s

' y\\-

o m :,

g; % s 8p3

+ tr.

_y:

~yy N.;:

.,9 oy

^y--

j A

..y %pha 3

y

~

v r

y

~

v

• • ""~ "Mwa-*Q W," " * " ~ " * " - "

+c

's t

j

> g;;pg

%K 4a R$!vM(S.

yA y

d Q qQyy Q a;

h

, p@f pp

~

gl

^ ',

]wf. Q.,

l x<.

n m

m$,'

,, s,,,.

9-s:

a.

w.

+

ba.

wg 1.

,y p'l y tm,

a

.i u

mh.Dj s u

/

<,, b 4. --.

i y

w'[F E '. ,.'U n,.

'+JE A

g

=

j\\

%. y.p p

y.... a. 3

.., m'

,r b m, "p*

' w f_

'r.

1

~

^

7-g; y

g

[f)

-f A

W eY g

p "

wwrw(g 3_,;3., 9; g,.,

n a.

s y.

p1 y

n rm: m-m v,,

m:v y -

-m.m u

s

?

V

$Mk I:b N

W

(

yi Q

_cn s,

9 1

j.l y2

.c e,

.gn ;;ys.,

p.

s

~

gy:g

_n

,-r

~

< ' ( Mp. gg 3 -

y7

~,s s

S s,.

+<'

,+

r' Y-

~-.e i

.*l. ' fk

,.4 I'[;,le.

c

-:pm W]m a.!:, l.

, L.:r u

, +

7 y l<

a~

.., o 7

t

..yg gy C

n:] p 1G ',

u ' Q ;,h, u

,. e m

aw

ufw, 3

s s,nv

- w -s. r

~.

J,. -

e,

M, _:VV 1

b,. p.-

+ &

w o n :/ pnr v

. ~

W, s

w y; I ;,

s%

wa Q,

. }

s c

-r

yn ; x g,(

gy.

___'L-- ' '. '. -

, g,

'i q,,

p.

~

~

',hfj

,.a,.

a

.-}.

' [Q:

e t

g

~

m r.

3,,,,~.373. q M.

+

ig +

4.s +

% g-y g

g g$l s

4.

~r

+e*,

& m! ~ wl_.&,( ?:Y' s

am':.; h.

t

^ ' '

A

,' l A-t;

~

L e

i,

(

_ * -' %: b :

,3. D. W.%e.

1?

,3

&g,1l QR??, Y % 'nlW>y,' %~,,, N., >

p m.<

x.

-M @g:.,%.p"MAM. ~MMM-n H' pgp c.3.,.

.. _. s p e

f t

s g $m,N. w 7( n J

- 7*. m,

q eur

)]g;c NW

,t c

i. s,s.

• n &,

i,.

n~

w u

g

,.i_%j y rh Vtt' f, j'.

p+

.y gy 7 i -

y

}l gg ;

g g

Q*

w

,~

oc

~ lh.?D.Q::l

hj; %r &,.,,.Q$%'V,f L e ::, L %,. >

h y

m u, y.

a.wa m -. m an RQ^ $y& 4Ml1Mc $ N M} yukg./ f& Q,.',, Q W Q22 ff, UR:::x,

a 3

J@ -ppm 4

%, h n ;r WJ h, k^

• l[%w#apph h b N N 'h [ Y k )T

• W k

N R Mg,%, l N, & m% y &( p 9,

m$32 w;,, g%,

^

as g

9 y

-+,

3, %, g g& g;a/;p.n ~

- ~r y Q y,' ;

' M( q Q*y j g.t p;,

l Qj %

a pq pp'9 4;

g;w L

3g, y sw y

p.3,

g:g, wgg+ w; < ggAY, my L'

%7yy.

Mi%

DD.$' WMMIhk#E 'k 9308170039 930812 PDR ADOCK 05000413 A

• %$g.

ny l

w a

p pyg l

gs eghqs e--~

ny W

W n

,s

~ 5)

.I

'k 5

.. }j, 3

9

--p w

j3 z.; y.;

yd

+

j Ai

~

7 o

p

?

fI15 r

. SY dNN N'

Jh

^

@l ~ 1 l

z l

ne a

G,il L

(

TFml W H

.....X_

r i

t E"'

pf M.*

. - - -m-

_.-;:y fhkk Nb?W.$l';

5 I

O

+

h

?.l., fWhiin;:

hdIn 0

apag.

y

\\,94.+N r~

i 1

ae n.p., h fh ff.

Y

' a 1

-l

• 0,

.m n1e

m

'l9 4

xy 3-

+ 8, ;

pgggg g g:g g g 7

~

anne.au.n.w:

l

_, _ m;._

QC

~

.' e, *2 ;Q. I[,;h y

a...

s.

;, ; y g.3.

n%,si?'

L, f S)-

L

~?

l [, h~

. p.a

+9 AY

[:khi[.

uh,,uMWimTe mis 4

^

W-NTM

'j

(

4,g. g 1.

g

%3 ggggg g,

n m, % g@g%wwww ggy 3

y eg; %

w ap;j;Ynp w;p x: w '

jmp w

M s

w I

~

l' n

, e % mwAm s===mi$,;M M M S M

, z 1;

NE* 7.-MsMN;OC@ Sa 9308170039 930si; 8tl$ll0lk?Y!;I h

WCAP-13763 WESTINGHOUSE CLASS 3 (Non-Proprietary)

EVALUATION OF PRESSURIZED THERMAL SHOCK FOR CATAWBA UNIT 1 j

i P. A. Peter i

August 1993 i

Work Performed Under Shop Ortier DHSP-108 Prepared by Westinghouse Electric Corporation for Duke Power Company Approved by:

kWd I

T. A. Meyer, Manager Structural Reliability & Plant Ufe Optimization WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Techrx> logy Division -

P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 -

i O 1993 Westinghouse Electric Corporation All Rights Reserved i

s

.)

5 PREFACE This report has been technically reviewed and verified.

Reviewer:

J. M. Chicots W-V 4

I

- i 5

i 5

i

..f A

3 4

4 I

.l r

i 1

1 i

4

+

TABLE OF CONTENTS

.[

Paae Table of Contents ii i

List of Tables iii l

List of Figures lii 1.

Introduction 1

+

2.

Pressurized Thermal Shock 2

i e

3.

Methods of Calculation of RTpw 3

-i 4.

Verifcation of Plant-Specific Material Properties 4

5.

Neutron Fluence Values 7

[

6.

Determination of RT,3 Values for All Beltline 7

Region Materials 7.

Condusions 11 e

8.

References 12 It i

t I

i t

?

f P

O f

t l

'I i

f l

LIST OF TABLES i

Table Trtle Pace 1.

Catawba Unit 1 Reactor Vessel Beltline Region 6

Material Properties 2.

Neutron Exposure Projections at Key Locations on the 7

Catawba Unit 1 Pressure Vessel Clad / Base Metal interface for 4.08 and 32 EFPY 3.

Calculation of Chemistry Factors Using Catawba 8

Unit 1 Surveillance Capsule Data 4.

RTym Values for Catawba Una 1 for 4.98 EFPY 9

5.

RT,3 Values for Catawba Unit 1 for 32 EFPY 10 i

i LIST OF FIGURES Fioure Trtle Paae 1.

Identification and Location of Be!line Region S

Materials for the Catawba Unit 1 Reactor Vessel 2.

RTen versus Ruence Curves for Catawba Unit 1 11 Urniting Material - Intermediate Shell Forging 05 I

1.

INTRODUCTION A limiting condition on reactor vesselintegrity 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:

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.

In 1985 the Nuclear Regulatory Commission (NRC) issued a formal ruling on PTS.11 established screening criteria on pressurized water reactor (PWR) vessel embnttlement as measured by the nil-ductility reference temperature, termed RT,13tn.

RT,1, 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 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 PTS Rule was published in the Federal Register, May 15,1991 with an effective date of June A

14,1991. This amendment makes the procedure for calculating RT,73 values consistent with the 2

methods given in Ragulatory Guide 1.99, Revision 2.

The purpose of this report is to determine the RT, values for the Catawba Unit 1 reactor vessel to py address the revised PTS Rule. Section 2 discusses the Rule and its requirements. Section 3 provides the methodology for calculating RT,. Section 4 provides the reactor vessel beltline region material p7 properties for the Catawba Unit 1 reactor vessel. The neutron fluence cues used in this analysis are presented in Section 5. The results of the RTpys calculations are presented in Section 6. The conclusions and references for the PTS evaluation follow in Sections 7 and 8. respectively.

I r

2.

PRESSURIZED THERMAL SHOCK i

The PTS Rule requires that the PTS submittal be updated whenever there are changes in core loadings, surveillance measurements or other information that indicates a significant change in projected RT values.

m The Rule outlines regulations to address the potential for PTS 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). 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 PTS concem arises if one of these transients acts on the i

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.

{

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

i f

All plants must submit projected values of RT for reactor vessel beltline materials m

by 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 RT for any materialis projected to f

ers 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 i

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 i

in this rule. The submittal must include the following:

1) the bases for the projection (including any assumptions regarding core

[

loading 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 I

to the NRC, justification must be provided.)

f k !

i

}

~

The RTp3 (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 materials.

l l

The following equations must be used to calculate the RTp3 values for each weld, plate or forging in the reactor vessel beltline:

Equation 1: RTp3 = 1 + M + ART l

p3 Equation 2: ARTyn = (CF)

• f S'" *8 9 All values of RTp3 must be venfied to be bounding values for the specific reactor vessel. In doing this each plant should consider plant-specfic 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, induding analyses of altematives to minimize the PTS concem.

NRC approval for operation beyond the screening criteria is required.

3.

METHOD FOR CALCUl.ATION OF RTp7, in the PTS Rule, the NRC Staff has selected a conservative and uniform method for determining plant-specific values of RT at a given tirne.

p3 For the purpose of comparison with the screening criteria, the value of RTp3 for the reactor vessel must be calculated for each weld and plate or forging in the beltline region as follows.

RTp3 = 1 + M + ARTp3, where ARTp3 = (CF)

• FF I=

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

Margin to be added to cover uncertainties in the values of initial RTum, copper and nickel contents, fluence and calculational procedures. _ _ _ -

i 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.

i a

FF = fluence factor = f **"8 0, where 2

f=

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

CF=

Chemistry factor in *F from tables for welds and for base metal (plates and forgings). If plant-speci5c surveillance data has been deemed credible per Reg. Guide 1.99, Rev. 2, it may be considered in the calculation of the chemistry factor.

4 4.

VERIFICATION OF PLANT-SPECIFIC MATERIAL PROPERTIES j

+

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

5 A

The beltline region is defined by the PTS Rule to be 'the region of the reactor vessel (shell material induding 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 Catawba Unit 1 reactor vessel.

j Material property values were obtained from material test certifications from the original fabrication as j

well as the additional material chemistry tests performed as part of the surveillance capsule testing i

prograrri'M The average copper and nickel values were calculated for each of the beltline region j

materials using all the available material chemistry information.

l A summary of the pertinent chemical and mechanical pn>perties of the beltline region plate and weld -

materials of the Catawba Unit 1 reactor vessel are given in Table 1. All of the initial RT values a

(I-RT,y) are also presented in Table 1.

I i

I

.t

~

j t

B E

h b

I~

. INTERMEDIATE SHELL FORGING 05, e

HEAT NUMBER 411343 -

CIRCUMFERENTIAL WELD SEAM, R747 CORE E

WIRE HEAT NUMBER 895075,.

i FLUX GRAU LO, FLUX LOT NUMBER P46'-

2

. LOWER SHELL FORGING 04,

[

HEAT NUMBER 527708 l

,t l

1 I

Figure 1. Identification and Location of Beltiine Repon Materials for the Catawba '

f Unit 1 Reactor Vessel

.l

.1 '

'I

.c}

?

I r

a

.l 1

l J

TABLE 1 CATAWBA UNIT 1 REACTOR VESSEL BELTLINE REGION MATERIAL PROPERTIES c

CU Ni l-RT,

Material Description

(%)

(%)

('F)

Intermediate Shell Forging 05*

0.086 0.858

-8 l

Lower Shell Forging 04*

0.050 0.825

-13 l

Circumferential Weld

• 0.049 0.717

-51

• Mean values of copper and nickel as indicated below Copper Nickel Material Data Source (wt. %)

(wt. %)

Forging 05 Surveillance Program [4]

0.10 0.84 Capsule Z Report [7]

0.079 0.85 Chemical Analysis [5]

0.08 0.85 Chemical Analysis [5]

0.89___

Mean Average Value 0.086 0.858 Forging 04 Chemical Analysis [8]

0.04 0.83 Chemical Analysis [8]

0.06 0.82 Mean Average Value 0.05 0.825 W eld Surveillance Program [4]

0.066 0.71 Capsule Z Report [7]

0.031 0.74 Nuclear Safety Task Sheet [6]

0_05 0.70 Mean Average Value 0.049 0.717 - - -

5.

NEUTRON FLUENCE VALUES t

The calculated fast neutron fluence (E>1 MeV) at the inner surface of the Catawba Unit 1 reactor vesselis shown in Table 2. These values were projected using the results of the Capsule Y radiation surveillance prograrrit The RTp3 calculations were performed using the peak fluence value, which occurs at the 25. azimuth in the Catawba Unit 1 reactor vessel.

TABLE 2 NEUTRON EXPOSURE PROJECTIONS

• AT KEY LOCATIONS ON THE CATAWBA UNIT 1 PRESSURE VESSEL CLAD / BASE METAL INTERFACE FOR 4.98 AND 32 EFPW EFPY O'

15' 25' 35' 45' 4.98 0.265 0.370 0.392 0.309 0.346 32 1.70 2.38 2.52 1.99 2.22

• Fluence x 10$ n/cm (E>1.0 MeV) 2 6.

DETERMINATION OF RT,3 VALUES FOR ALL BELTLINE REGION MATERIALS Using the prescribed PTS Rule methodology, RT values were generated for all beltline region pn materials of the Catawba Unit 1 reactor vessel as a function of present time (4.98 EFPY per Capsule Y analysis) and end-of-life (32 EFPY) fluence values. -The fluence data were generated based on the M

most recent surveil!ance capsule program results.

The PTS Rule requires that each plant assess the RTem values based on plant specific surveillance.

mpsule data under certain conditions. These conditions are:

Plant specific surveillance data has been deemed credible as defined in Regulatory Guide 1.99, Revision 2, and - _ - - _ _ _

i RT,, values change significantly. (Changes to RT,3 values are considered significant if the value determined with RT,3 equations (1) and (2), or that using l

capsule data, or both, exceed the screening criteria prior to the expiration of the operating license, including any renewed term, if applicab!e, for the plant.)

For Catawba Unit 1, the use of plant specific surveillance capsule data arises for the intermediate shell forging 05 and circumferential weld because of the following reasons:

l

\\

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 forging and weld materials.

The chemistry factors for the intermediate shell forging 05 and circumferential weld were calculated i

using the surveillance capsule data as shown in Table 3. The chemistry factor value for the lower shell forging 04 was calculated using the Table 2 from 10 CFR 50.61.

TABLE 3 CALCULATION OF CHEMISTRY FACTORS USitG CATAWBA UNIT 1 SURVE!!. LANCE CAPSULE DATAM Component Capsule Fluence FF DRTNDT FF*DRTNDT (FF)^2 FORGING 05 (Long) TANG Z

0.343 0.705 0

0.000 0.497 Y

1.35 1.083 15 16.251 1.174 FORGING 05 (Trans) AXIAL Z

0.343 0.705 10 7.052 0.497 Y

1.35 1.083 35 37.920 1.174 -

61.223 3.342 Chemis:ry Factor =

61.223 /

3.342 =

18.32 Weld Metal Z

0.343 0.705 5

3.526 0.497 Y

1.35 1.083 15 16.251 1.174 19.777 1.671 Chemistry Factor =

19.777 /

1.671 =

11.84,

Tables 4 and 5 provide a summary of the RT,nvalues for all beltline region materials for 4.98 EFPY and end-of-license (32 EFPY), respectively, using the PTS Rule.

TABLE 4 RT VALUES FOR CATAWBA UNIT 1 FOR 4.98 EFPY m

Material ART,(oF)

Initial RT, Margin RTen (CF X FP)

(F)

(F)-

(oF)

Intermediate Shell 55.2 0.7406

-8 34 67 Forging 05 (18.3) 0.7406

-8 34 (40)

Lower Shell 31.0 0.7406

-13 34 44 Forging 04 Circumferential 66.6 0.7406 56 -

54 Weld Seam (11.8) 0.7406

-51 56 (14)

( ) Indicates numbers were calculated using surveillance capsule data.

Fluence factor based upon peak inner surface neutron fluence of 3.92 x 10 n/cm (4].

2

.g.

~

TABLE 5 RTem VALUES FOR CATAWBA UNIT 1 FOR 32 EFPY L

Material ART,(oF)

Initial RT, Margin RT,n (CF X FP)

(eF)

(oF)

(oF) l intermediate Shell 55.2 1.2482

-8 34 95 Forging 05 (18.3) 1.2482

-8 34 (49)

Lower Shell 31.0 1.2482

-13 34 60 Forging 04 Circumferential 66.6 1.2482

-51 56 88 l

Weld Seam (11.8) 1.2482

-51 56 (20) l

() Indicates numbers were calculated using surveillance capsule data.

2 Fluence factor based upon peak inner surface neutron fluence of 2.52 x 10 n/cm [4].

l l

i.I J

7.

CONCLUSIONS r

As shown in Tables 4 and 5, all the RTpr3 values remain below the NRC screeniro values for PTS using the fluence values for the present time (4.98 EFPY) and the projected fluence values for the end-of-license (32 EFPY). A plot of the RTets values versus the fluence is shown in Figure 2 for the most firniting material, the intermediate shell forging 05, in the Catawba Unit 1 reactor vessel beltline region.

t 300 SCREENING CRITER A 250 200 S

Uv 3 150 Q.

F--

U~

100 A'

' ~

..g.

50

.......------*-,,,,,,,,,,,,,,....g.........-----

G EFPY A 32 EFPY i

0 1E + 18 2E + 18 3E+ 18 SE+ 18 1E+ 19 2E+19 3E+19 5E+19 1E+20 FLUENCE (NEUTRONS /CM )

2 EMONM INTER SHELL FORGING USING SURV. CAPSULE DATA i

Figure 2. RTp7, versus Fluence Curves for Catawba Unit 1 Limiting Material - Intermediate Shell Forging 05 8.

REFERENCES T

i

[1]

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

t

[2]

10CFR Part 50.61,

• Fracture Toughness Requirements for Protection Against i

Pressurized Thermal Shock Events," May 15,1991. (PTS Rule)

[3]

Regulatory Guide 1.99, Revision 2,

• Radiation Embrittlement of Reactor Vessel 4

Materials," U.S. Nuclear Regulatory Commission, May 1988.

[4]

WCAP-9734,

• Duke Power Company Catawba Unit No.1 Reactor Vessel Radiation Surveillance Program *, S. E. Yanichko, July 1980.

[5]

Potterdam Order No. 30743-66100, Chemical Analysis, Drawing No. 30738-1515.

1

[6]

Rotterdam Order No. 92104, " Test Report of Wekling Materials", Lab # R.747.

[7]

WCAP-11527, " Analysis of Capsule Z from the Duke Power Company Catawba Unit 1 Reactor Vessel Radiation Surveillance Program *, S. E. Yanichko, et al., June 1987.

[8]

Rotterdam Order No. 30743-61520, Chemical Analysis, Drawing No. 30738-1514.

[9]

WCAP-13720, " Analysis of Capsule Y from the Duke Power Company Catawba Unit 1 Reactor Vessel Radiation Surveillance Program *, J. M. Chimts, et al., June 1993.

4 e

1 i

r

?

-l.