ML20085C148

Evaluation of PTS for Callaway

ML20085C148
Person / Time
Site:	Callaway
Issue date:	05/31/1991
From:	Chicots J, Meyer T, Ray N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20085C142	List: ML20085C139 ML20085C144 ML20085C148 ML20085C152
References
WCAP-12948, NUDOCS 9108300036
Download: ML20085C148 (15)

Text

s...........y . . . . ~ . , , ,. . . .. . .s a g.-.+.--...-+-** ' . ...

,n.. .- p ~ ' - ' ' . * . ,' : ."

. . \' . ?

, #l* ' . ,* ,j

. g , ' ..$*

.;<,, , .W +

g ,.

.- 4 ,. ,, ,, . . .

.

• s

,t, . ..

. . . . 8 - . ,

c .L s ,' '. ' -

g .

', ' ,y

'.., . . . ?. .

- . ~

J .. . ,. . . . # .-

. . , . /

9

'3.,.

i

- , p. . , * . . ..

, ., ,.y

.,,g. ,g l

~

,, , ,t, .

.l , . . . ,

,.~a-v' -. , , . . ,. ._ ,$'g

.s < . . < , .: ,; , . , ' ' ' *.. ' ' ',, , . ,- ..

.'- ., s, ..

,f,,,'.,.,,'.

^ * - .. . . . ' ';

. G ' ' '

. l ' ' ,, i ,_ l* ,

, .'?', , *

'^

r ,y i ' ,

t s,.. /

i. . - . - , . .

0 ' , '; '.:

) S

','y+,

(*

, l ,, .

- , 4 -..

,.*y. l' f, * ' , .

.,.gg .. , , , , . , 3; . .

. qc' -

c.

,., ~ - t . ,. ,

'g' ,,' ' . . . , <.. +f.

. i * ' '

4.y- .1.- ,',.-.$ . * '* ll

. . . .J. .

- ..  :. .af., .. . .,

l' . , ,- ,...s . . .. . .

. . , , . ' . . . . :! . % ,.,,u . , , ' , . ' ,l,, ,,

N ~; l. .. . . ,;

.a- .

l, . i . e. .<.. *'

....,',_[.__.._7^..,,'....

6 - ^ G' ', . ._ ;. . . , . ._ ,;.'..',.s.

, , ' . < q~ ; '~, * ! .._t  ;, ,f:..'.. . .. .N n

. , . . * .;,_y'.

.'.4- -i'-s...' ***.,'#p .,'._- [. '. 3w  ;. ...-. * ' . . .

+. -

' .  ?,

-* , , . , . , *- , $'., r - g k (. .t, l '8 +

'. i, d;s. . " - , ' . . 'w .,'c'w-*J..;

s  ; #. ,

4  ;. ',g ' * ,' .V . -

ob

- h ; Y,! A * - .,*et ' . .; b=;=C. ,. ',$' .'e,9.i h .' * . U' , , '.7 . ' , ,, / ;.*...*

Q ' .g ' '.i , .' '.k;._ b (ld i 'a '[

. .[ .

.; },. ' ', 3[q.

's '

,v .

,, +.. , *k

,+ . , . .s

. .* , i

- , . 9 g .b '

_,', .,g .t' ,

r. . ,r ,. . ..

., . G e

. .s ..

.'. I ' 1,fb .. *

• ' . + , *. -' ' ' . .,, , , .

,. , 4

[ " .' .

, . .3i..' - , ..

• 9 *, ',- 3 ,

• k , .

• b.. .-. g n' , ,

g', , . 4.' l. ,,

4

.,- , . , .'.**(h - .

(j . ,

-E ,., , - , y,,. .v . .%' .

,; l ,3 e., ( f ' . ';t .

.' si g ,. . 4. ,

'.,.,e.., $ * ., -

p,o'.'l,* .c. %e

.sE. .g * ..,1 . -

1 -

? . .*, .. '

j, ,- ; , J. { ). i

'4,.'-

t -

C,

) .,,. ic....

-p- -

,. ' . , ', e . .; , ,"~

[.

• ' 4 s[. - .

.,<,'w.

v

.: c.

. u.; .: ,o...n yy .

}l . (' , -. , ' , < - i ,. [.,

. i - .ps  !' ,.. ' i'_ T. ' '.'.',,

<. . .- ' . a .i h.c . ., ,

, .. M' , ' I . ..;-

.<,;* ,,{. =.* ' - . . *.' . - +'y

• _ , , , . . t u & y g

~ -

q 5 . . .

~

q,- .

3,, ,;a. 3. , p y. ,,, .5. . n y ,, .o I* 'g

• " +

. .y. .i ; p. m_w .:x.g, , %. .. .

' ii

. q . .. : .

D - . $ " .,

.h"

4. $

g.-

t ,,

.. ,c . s .. ... . . . n . . . , , . . < .. ,, .. - ;, . . . , . .

,...,',f'.

, + .e

.9-q

{-

i . . ,,.i + .. . s, , . . . , . ' ,; .'.=. -.: . s,5

,, - - ,. ,. , c'.- - ".is .  ;

5. .'.._ .* j.,- * *.- ?gq

, '- .' y* I' .,8,,- '

J - -- - - 4

+ ,'-- -

.',..g'..

I

. . . g g

{

jb

.' .4- s - . - . - ; .

,. 'g'. k+ ,."*. ,.'*- .

, - . ' .. ;7 , c . . . -

s I.; , . 7 ., . .

l.

.Id *. .* ,',. .' , .. ', . ' , ., . .'*.., ',

p, -

l . .

' '.~b.,'.

'a--

e - . . .. . '

- . ..g

.. . . ~ .** . . .

v J. J ,', ,,.

t

-.. -,. p ,*~

,. , . ' .0'.', ,

9

-' 4 , ,.

- . 5 - . , ~i,s .

. ,, , ',* _i - , .,

.l .

i 7 - . , ' . , . ,~ ' . .. . .. :<:..........' ,  ;. ~ ,.. >. ..

-~ . .

./, . .

, " d 9  %

• c -s .

" .yi >. . .' ' . . . . .

g

.,a ye . . - '

- ' ', * ' b", g ,- -

• f . , '

• a ' ,.

9 ',

4 ,

. . . . . . .; . . . . . - ,. 7, m , .

g

, I , . ., ' " . 6 . . ^ . . * -

g

. . . , * . ',? -g

,f .l g- . ' - ' . .

'* . .- ' s

. 's y j '.. ;l . ,, * . . . , -. ,, '

4 , .' . ' . ' -

.; (,

vr -' '.

',,.: . ' . ' . - , ., . , , u.s* ..s.

. . .s

,f., , . . .. . ,

'c' e

?

h s

• ' , T

-s

• .,g '+

','.,-)

, , [ ' ,i

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

, j; - . . ,

I '* s ,- - ,

0 .

h *g *

,' ' g '

.,'9 , ,. . , .

, [.',I',, *. s,, I *

,#ee%- g.

' [ -_ J _'* - *

,6_

' #I

• . g** 'W

.I g'

,*'y.,

'e _o g +. . = .

-

• t,4 -

-.i.,,, M- * ' . i'. ,. .

[

- .,:-',. - .* .j, y

'.- t ,, (,..'+

. , ' gy

',,( .,

'";i<,- -

$a -

. .' , , ' . -ep- .

- " . . -3...

- '

• N- - .' * . s L % ,

- - . . . . . .L '

, . , .3 , , . . , <

l

x. , .*.

.. * ' s .c. . . . , . *y... . ,- . . * *.*, ' '. .- ... '. ,- - - .g g ...;

s -

g -

a

,gs. ' -- - ' . c.P.f .

, 8

,1* '. . '

~ *

. .. (.

.v. . .' . ,

,-.,..#'. .,s  ; ., y - .-.

. , , . -s t.

y.s

.g

, ,,. , . . . . . . , 3,

. .i.y.'

. .3 ,

' , .. . . . , ' -.n.. .

- 7 ., -. . ..,.s.. -

g..

c. ; ,

+ '.' J ' . Et ,

' . ,. .. .' ' ,'. ,, -. . '. I ,' -

t-4 * , _, - . ,. 4

._7....-.. e 4.

, ,. ; , . / - .- . . .

! ' , . . t . . , , , 7 c. . s,,. e.

.,..;g,,. ,

o',. \ -- ,

..c(. .

s.

. , , . . ' . .. .. .. ... .J.. -

. 4 .. ,. ... e, . - .. ..v, '. . ..,.

c t.. - ,. 3. ,

., ,- . . .; - ...y ,e , . . . ,

-. . .? - - .

.g _ ,

+ b .. *  %

• *..,.4 * " ' ,. ,

y..,.,'.'.

", g

, " g #-' ,O.

, , [ ' . . , .

4 ,

y . - .- ... ....... .;.. (. . -.;..n

.. . , , . . + . . = .

.. , .v ... - ..,* .+.;*, ..,. . - . . -;

.. . . ., . . .a,...,. ..',- -

• ,, ' j . . -. .i g

j. * .- ,. t* ':

gs-

, . . , I..

p i

- c., *, ' .i _* *, . - . , . . , , ,

, , - g , y , ' -* - - W 9..

. _. ~

?f G _g%gf.G l ?.g...l'+

7

_ ...,7,'

. A

.gn,e.,,. .

g.n i ,..;:... .

N 7-. .

g,,n ,,v-... .7.. ,c a <;

y;..l. . . .' ..

• - -v a.

u; .

.c

b. yO,H f

.--...+%.f+ . ya '. $'y;gm~..y sc ..i..

. '.50

a. s,,2,,

q- -m. .., .

A

. h . . ;l, +k.

yy 1

.a 3 ig_"c.

c.

(,y .

.; f..n. , t.+?;Q.Wg .. .m' f .

.9 u . . [...c.':u . ..,' .n. . . ,n . ~ -

w. s w .-

z. ' .;*:

t- C ...y-  : ..c .e

.% n4 <. y:. '

~~.

. . . .* R :K* . (U.:%.. S

.r .. '.>.

y. . .

.' ' p y;* b y '- ' h..yN: . . . c s}. Q. J }.. :..! :.i lf;,%. >% 2 ' . j ; e h j...[9

., A.yff .%.. .f.. . .bh.h. . '.j.?: Q

. . ,a. ' 6: ,7 . ' - -

.. ~

. l. f. :f,f.,. :.: .j h,, -.  ;. .,p. %;".:d,:h..)n ... h]). . h, h;h.-f

. 7

' m., . . . , . , ,

y . . .

h. ,.f, " 3 . g

{f. h't'.h, .fh ,%. . ph.,,p . '.h.f'. -Q'

~

[ .j .' E h.w

,.c. ,. a.c 4 . ;. % k,..,.g..,.*' e. ,,.p o ,r h' . m..r'.W m $ q*. e .'%-..;.. %;% m.pg .w m.4g.

1. . ; w.. ,, %. .(. v,.wt4 i. s%

.% m.

[c e$ Ei ? ..V.l.].b Q,.c.., , > . 4,.;pW Mg @ N [,.

m cy _

yJ 3,....- .

f.z t, .9 # @t,r#... '.i .,. ,f yf ee y t 1,7c.p:.Yfptis?.[x-- .ig.3. % 9%Q 4 T4.i.' <%O. 4* .a.% D $ ':G.n .:.Dgi/Mb.W4 I ;j.)'j

WCAP-12948 EVALUATION OF PRESSURIZED THERMAL SHOCK FOR CALLAWAY J. H. Chicots N. K. Ray May 1991 Work Performed Under Shop Order VMSP-108 Prepared by Westinghouse Electric Corporation for the Union Electric Company Approved by: - '

N -N T.A.Heyer,Madager Structural Reliability & Plant Life Optimization WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 o 1991 Westinghouse Electric Corp.

TABLE OF CONTENTS Pace Table of Contents i List of Tables ii List of figures ii

1. Introduction 1

2. Pressurized Thermal Shock 2

3. Methods of Calcula'. ion of RTPTS 3

4. Verification of Plant-Specific Material Properties 4

5. Neutron Fluence Values 6

6. Determination of RTPTS Values for All Beltline 8 Region Materials

7. Conclusions 9

8. References 11 i

LIST OF TABLES lab.lg Tit 1e EAgg

1. Callaway Reactor Vessel Beltline Region Material Properties 6

2. Neutron Exposure Projections at Key Locations on the 7 Callaway Pressure Vessel Clad / Base Metal Interface for 32 EFPY

3. RTPTS Values for Callaway 9 LlST OF Fl_(il!RES Fiaure lille EAgg

1. Identification and Location of Beltline Region Material for the Callaway Reactor Vessel 5

2. Fluence vs. Effective Full Power Years for Callaway 7

3. RTPTS versus Fluence Curves for Callaway Lower Shell 10 Plates, R2708-1 and P.2708-3 il

l. INTR 0000110N 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 f ollowing 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 the presence of a critical-size defect in the vessel wall Fracture mechanics analysis can be used to evaluate reactor vessel integrity under severe transient conditions, in 1985 the Nuclear Regulatory Commission (NRC) issued a formal ruling on pressurized thermal shock. It established screening criterion on pressurized water reactor (PWR) vessel embrittlement as measured by the nil-ductility reference temperature, termed RTPTS I. RTPTS screening values were set for beltline axial welds, forgings and plates and for beltline circumferential weld seams for end-of-life plant operation. The screening criteria were determined using conservative fracture mechanics analysis techniques. All PWR vessels in the United States havo been required to evaluate vessel embrittlement in accordance with the criteia through end-of-life. The Nuclear Regulatory Commission 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 14, 1991 l23 This amendment makes the procedure for calculating RTPTS values consistent with the methods given in Regulatory Guide 1.99, Revision 2[3] ,

_1

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

The purpose of this report is to determine the reference temperatures for pressurized thermal shock (RTPTS) values for the Callaway Unit I reactor vessel to address the Pressurized Thermal Shock (PTS) Rule. Section 2 discusses the Rule and its requirements. Section 3 provides the methodology for calculating RTPTS. Section 4 provides the reactor vessel beltline region material properties for the Callaway Unit I reactor vessel. Th6 neutron fluence values used in this analysis are presented in Section 5. The results of the RTPTS calculations are presented in Section 6. The conclusions and references for the PTS evaluation follow in Sections 7 and 8, respectively.

2. PRESSURIZED THERMAL SH0CK 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 values.

The Rule outlines regulations to address the potential for PTS events on pressurized water reactor (PWR) 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 concern arises if one of these transients acts on the beltline region of a reactor vessel where a reduced fracture resistance exists because of neutron irradiation. Such an event may produce 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:

All plants must srbmit projected values of RTPTS f r reactor vessel beltline materials 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 by six months after the efective date of this Rule if the value of RTPTS for any material is projected to exceed the i

\

screening criteria. Otherwise, it should be submitted with the l next update of the pressure-temperature limits, or the next '

reactor vessel surveillance report, or 5 years from the ,

effective date of this Rule, 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 patterns), ,

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

• The RTPTS (measure of fracture resistance) Screening Criterions for the reactor vessel beltline region is 270*F for plates, forgings, axial welds l 300*F for circumferential weld materials l The following values equations for each should weld, plate be usedintothe or forging calculate reactorthe RTpI3 vesse beltline.

Equation 1: RTPTS = 1 + M + ARTPTS l Equation 2: ART PTS = (CF)f(0.28-0.10 log f)

All values of RTp13 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.

• Plant-specific PTS safety analyses are required before a plant is within 3 years of reaching the Screening Criterion, including analyses of alternatives to minimize the PTS concern.

• NRC approval for operatirn beyond the Screening Criterion is required.

3. MLTHOD FOR CALCUl.ATION OF RTPTS In the PTS Rule, the NRC Staff has selected a conservative and uniform method for determining plant-specific values of RTPTS at a given time.

For the purpose of comparison with-the Screening Critierion, the value of RTPTS for the reactor vessel must be calculated for each weld and plate or forging in the beltline region as given below.

RTpyg = 1 + M + ARTPTS, where ARTPTS - (CF)f(0.28-0,10 log f) 1= Initial reference temperature (RTNDT) of the unirradiated material M- Margin to be added to cover uncertainties in the values of initial RTNDT, copper and nickel contents, fluence and calculational procedures. 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.

f- Neutron fluence, n/cm2 (E > IMeV at the clad / base metal interface), divided by 10 19 CF - Chemistry factor from tables [2] for welds and for base metal (plates and forgings). If plant-specife 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. VERIFICATION OF PLANT-SPECIFIC MATERIAL PROPERTIES Before performing the pressurized thermal shock evaluation, a review of the latest plant-specific material properties was performed.

The beltline region is defined by the PTS RuleI23 to be "the region of the reactor vessel (shell 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 Callaway reactor vessel.

Material property values were derived from vessel fabrication test certificate results. Fast neutron irradiation-induced changes in the tention, fracture and impact properties of reactor vessel materials are

N a

N w

W b

E 90 0

101-124A' g R2707-2 2707-3

+ f

• fJ M 0 A 180 0 101-124C 101-124B

$ ' 2707-1 270 0 CORE d 101-171 90 0

[ p 101-142A R2708-2 R2708-1 a

k- 0 0

=[

0 180 5

101-142B 101-142C .c N R2708-3 l

0 270 Figure 1. Identification and Location of Beltline Region Material for the Callaway Reactor Vessel d - .- --- . - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

largely dependent on chemical composition, particularly in the copper j concentration. The variability in irradiation-induced property changes, which exists in general, is compounded by the variability of copper I concentration with the weldments. I A summary of the pertinent chemical and mechanical properties of the beltline region plate and weld niaterials of the Callaway Unit I reactor vessel are given in Table 1. All of the init!al RTNOT values (1-RTNDT) are also presented in Table 1.

TABLE 1 CALLAWAY REACTOR VESSEL BELTLINE REGION MATERIAL PkOPERTIET I CU N1 1-RIN01 Material Description '

(%) (%) (*F)

Intermediate Shell, R2707-1 0.04 0.57 40 Intermediate Shell, R2707-2 0.05 0.59 10 Intermediate Shell, R2707-3 0.06 0.61 .- 10 Lower Shell, R2708-1 0.07 0.59 50 Lower Shell, R2708-2 0.05 0.57 10 Lower Shell, R2708-3 0.07 0.59 20 Intermediate and Lower Shell 0.04 0.06 -60 Longitudinal Welds, G2.03 Circumferential Weld 0.06 0.07 -60  :

i I

5. NEUTRON FLUENCE VALUES The caiculated fast neutron fluence (E>l MeV) at the inner surface of the {

Callaway reactor vessel is shown in Figure 2. These values were projected 3 using the results of the Capsule Y radiation surveillance programf 43 and s are presented in Table 2. The peak fluence (25' location) was used in i Figure 2 as well as for all PTS calculations.

TABLE 2 NEUTRON EXPOSURE PROJECTIONS AT KEY LOCATIONS ON THE CALLAWAY PRESSURE VESSEL CLAD / BASE METAL INTERFACE FOR 32 EfPY O' 15' 25' 35' 45' i

Fluence x 10 19 n/cm 2 1.50 2.16 2.39 2.02 2.38 (E > 1 MeV) 26 24 -

}

22 - .

20 -

18 -  !

16 -

g Q 14 - /

7 c /

g 12 -

7 EU 3 6 -< /

/

/

u.

4-2-

0 i , , , , ,

0 5 10 15 20 25 30 EFPY Figure 2. Fluence vs. Effective Full Power Years for Callaway (Based on peak fluence, 25' location)

6 DETERMINATION OF RTPTSVALUES FOR ALL BELTLINE REGION MATERIALS Using the prescribed PTS Rule methodology, RTPTS values were generated for all beltline region materials of the Callaway reactor vessel as a function of end-of-life (32 EFPY) and 43 EFPY fluence values. The fluence data were generated ba ed on the most recent surveillance capsule program resultsl4l.

Table 3 provides a summary of the RTPTS values for all beltline region materials for the end-of-life (32 EFPY) and 48 EFPY using the PTS Rule.

The PTS Rule requires that each plant assess the RTPTS values based on plant specific surveillance capsu.e 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 RTPTS values change significantly. (Changes to RTPTS values are considered significant if the value determined with RTPTS 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.)

For Callaway, the use of plant specific surveillance capsule data arises because of the following reason:

!) There have been two capsules removed from the reactor vessel, hence the data is credible per Regulatory Guide 1.99, Revision 2.

TABLE 3 Rip;S VALUES FOR CALLAWAY  ;

s .. w Screening Material Description Criteria 32 EfPY 48 EfPY Inter. Shell, R2707-1 270 106 109 Inter. Shell, R2707-2 270 82 85 Inter. Shell, R2707-3 270 70 73 Lower Shell, R2708-1 270 138 143 (116) (119)

Lower Shell, R2708-2 270 82 85 '

Lower Shell, R2708-3 270 108 113 Longitudinal Welds 300 33 36 Circumferential Weld 300 30 45 (65) (71)

( ) ."dicates numbers were calculated using surveillance capsule data.

7. CONCLUSIONS As shown in Table 3, all the RTPTS values remain below the NPC screening values for PTS using the projected fluence values for both the end-of-life (32 EfPY) and 48 EFPY. The highest RTPTS values for the upper bound fluence case at 32 EfPY and 48 EfPY for the lower shell plate, (R2708-1) are 138'f and 143*F, respectively. A plot of the RTPTS values versus the fluence are shown in figure 3 for the two most limitng materials in the Caliaway reactor vessel beltline region, the lower shell plates, R2708-1 and R2708-3. This plot Indicates that neither of the two limiting materials in the beltline region of the Callaway reactor vessel are expected to exceed the screening criteria based on the current fluence projections.

( \

I 300 -

I NRCRT PTS GCREENING VALUE (270F) 250 -

200 -

C . i

'A

{ 150 -

LOWER GHELL, R27081

$ sn 1

. - . pa=e

--<#' l 100 LOWER SHELL, R2708 3 50 -

O' 1 ' ' ' I 1E+10 1E+20 NEUTRON FLUENCE (n/cm')  ;

Figure 3. RTPTS versus Fluence Curves for Callaway Lower Shell Plates, R2708-1 and R2708-3,

8. REFERENCES

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

• July 23, 1985.

[2] 10CfR Part 50, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, June 14,1991.

(PTS Rule)

[3] Regulatory Guide 1.99, Revision 2,

• Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission.

Hay 1988.

[4] WCAP-12946, " Analysis of Capsule Y frotn the Union Electric Company Callaway Unit 1, Reactor Vessel Radiation Surveillance Program," E. Terek, et al . , April 1991.

r s

_ , . - , _ , . . _ . __ _ ._ ..-. __. _ . . _ _ _ . . _ . _ _ _ _ _ _ _ . . . _ _ _ . _ _ . . - _