ML20115A377

Evaluation of Pressurized Thermal Shock for VC Summer

ML20115A377
Person / Time
Site:	Summer
Issue date:	03/31/1992
From:	Chicots J, Meyer T, Munoz Frances Ramirez WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20115A367	List: ML20115A365 ML20115A372 ML20115A377
References
WCAP-13209, NUDOCS 9210140267
Download: ML20115A377 (17)

Text

{{#Wiki_filter:w= q . O ' ~ E 8

4......

.s. s .-l. l e p.' ..1 9. *9.'. l- . ~., _. ,p. * - .e., f c. .. y ,($...

• A

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

L r.

p r . *.e w.. -....

• ',4..

'!:; ;3 ".,4 g'.....;. i ', ;_. s -.., h. ' ; e; [ '-. . ~ .<d I '

L = . ' '..,. '. :. *. ? g.

1 [. p.., .g . r.. n -. a ...'..:'z.- .m,..

,t

,.../.- ..-4 P..- ,1 W.g ,* '.. 3'& ,.f%g .,f'.,,,...s.s- ....._.z.....:,-.... . ; s.,. <**.. ;. g.

• ?....,-...
• M,. *.* ' *.

j...

• (....,., -_,, g

..,.,.......4..,'..,, o' ..,..:..-..~. 4 :g- .(. .A s. ,.I + r g s,. s ..4'~- .n. k. c.

s.g - -

m.

^ ^ \\

• S N
• F r.3,.g%. e'%.; V. g.g y...,

_ Lx;. ' ' ~%g. ' .. y '. 3..;yl : 7,d..' .. f.,.f (L.?. ;W.,..-., . ; '.): -?'; l f 3

)

4. j l. - p*

.Q

• z..,..: ': '.: ;,,,,...;._.--

..u s,t 3..

q.~:

c..

.y n

S f,,.

s ;.s.y;. ?.... - l: y s; ;...

y.'

.z ,.s. e ...9lv..,[;.;:::;cq;,:.f.n:?.'.% ; y.g :.:4...j.). g .. * : :.V.' s y s.., . s, 'Jn., 7

.... V ;,

7..

.4.. q.T. 5..~.v.. u,..- c ',.....-- ;.' *w

... '

.. v. ; '

5.... .y L. 3)'.' 3,Q;.3(. ;;).....g-{~. g ~ sth V r t[ .s

• .q.:. L,. $.f : ;' y..p.;_.~ :

,3.*,. y w ;; n y= :y:.p.4 3..^'..y.. y ge z. .?_ s 'a...* .,. - ev. c..,'. t.;. f. ...,...,.;W g Z ..e-

• 4

...,s a 98.p;,%[.,.>.....g..%T.;:j.:a9, .}, '.. y...,i. 3;.., g xt 4, y g,, . -: g 1. 2* ..c . =.;

• p Ag,... : x..

. g; y{ y c

k.. b _.^:..},. n.s;g.5. g % n

.2.;. * ; e g. $,,....'.f;..., v w..':.: q. q p,, V, :.: ..v+ ,y e : p 4.:O. ?.: [,:{f..=>n'... ' l..f,.f. y. . e '..N. '. ' y. EM. ,dl, v -. f. jf -.- A (.,,,; je. [w.-. ppj .l, .c 3 ..%.;..y y....... d.. c... .c.. -E...<.e -. g y p.... %... ,,g n 3.... .. ~ - r.,,* g.(, y,,._., +.. .>...,,.s ,. s,. w..v .Wf.: : %,..;... v..: :. _. >.. 7 Q~. i ;.. p: v ;::g. -.- ;,_ w,-,, Q ?;.y...; , v. q. . L..% g.. .. ~ _ b .e s .. yI.-...,. :.., ,,4.3. -. -?, ., A,; V, ; k,i =..Q. g. 1- -,. f. 3.,y 5 .. g.. :.., (. 3.; .g' '" i. , 7.- - ' '. ;,{, : ':){4.p.Q ;

ff&;.., g'%,'.

y e

.- _

?__

.), _, ):;. a

w. _.

q.m... :.....M .. amt,. ;;. - n'. i. W ...g7....'..s gg ~ .w j. ^ >_ ' ' * ' \\.?. p... <: G.Q.: 2 ' -?... :.. 'N ,7Q. ' y _I. , p..(($ ~[ 7 ',).z. M,..: j h;,q).[.&,J

h.. ; f, A..y,@ '. I

. ;,;{ .v.

...

. v

+ y. t. ' ; ~a: w. n.n y. l &.{.g.,;p ;,:..fi.j::JQ %q c. b_..

u. u

..t c ;. w ?. L. -.; t. * . J.; n. . :... s. .. ~.,.1 W. .:. ::. A ? &. ;c? 'Q y - -.e.> %.: .,..v. J '....'..c... }..,5..ai :.:. 9.,.,) Q &.c

:' -%.p y ':j >.

..'.t'. ll.O a. '.., h, L ...l r" %- '..* - '. + Y * '...:,....[.. '. ' _,* ';. h ;. Pi 'h'

• ? " U

}, : f' N'.' .f.., ' ?',,, 1.. ;-) - [ _ ; +, * '., ;. ; _.. ;+ y,, f, J *..?.- - ' (, ~..O ,.....y: . ya' >s.*: l. .. l-. ; '... {'l', w A '.

• • 6

.g' 'tg* ry V V 'h,,- N<>- : - - * * '.,", ). j (f ?, 'f 7 .fj [ $.: I j ' [.1, 7.k,g.' O,. ,j.2.;.,kh k.lL. l f l ' '.. '.[ k J. ', h. \\f : ' l-Y' ' ( 3 ?- l . ;. j "5:j. '..?? ' ' (,Y,$..

• N. 's * ; } a ;. \\ Y 5.

..:l 'e.X> I,t ? N.;$.:?ft. f';Y. (w,G.ll&. 3, '1 " ',..!-l%

h., '....' 'e.L, C..,.,..' nQ: [' ' *'

2 ? o I f + Y.T N., g - 3 t.~ s ....e .g

s. 3

g. -..

s I 'k, hh Ih .3 d -[. ,L .l.[* y -.' c [ ; ' [:d, D k'..[,'[i, [.h"; k [ *< $., [ U; ',...h 1 -lq .k*'ts4...' '".c;%g. -,,'..E Y,. - ,,'.h;. 14.' pc. Y %,..-7.f.'e ;,;;;., *,.,y '.,g.. .) g. .y.. c: .,. ;s. *. >, . 4 gg ? . T .~- .] d . y _. :y_.. ;:g' 8,.9 p g. 3,,A?.^..n,y,.':_ g Q yy. ;g.,:: ; . y.. j q_.._.,,wjff. ~ 'y ' t ;. ',. '. .gk l V'9] h [. f ' 4y ;.jf j ' j,. ; - '. ', h;]. t, p - [r .]'. f#A.. % d.. [q '.i ' ' [ [.l* L ' l. { *. '.l { $ ': $ '.Q, h .?~.'+..L .l & l ^. E ~ }. j.: '_ ) '?: . '. '._2.; ,S..

I

.' A.' b }. _ (._*; 'l-5 }- }[ q

t..,,.4 ?

. c '..:.-., ~.,-, *.... ', - - 1. 6 i .. L %q%g,. 4 '...~ ' y ';.,. yf... , 2 8,.4,, 3%,y. 7. - *..,... .P >ldy?, Q, ' 1 .j f... L.*

• i ;3... +.

f. 2

. g. A.. ' : - :

f.. - qapq. ; 7t c

p'. w '.*; .+ \\ 's'. , f- \\,.

• r,

.i : ? n; f '. ' '. : - ; '.. : j. ..:.['64}A (.':

'.; $.,s 4..

. +... y %f-.' . ;* 4 * 's.f, ' ' ' '. '. ' ' } ' 7,,.- [. r,,5,l g l, g :I - e. ? c' .m ? <

y...

.',,1. ' ',[b,.. l [, ',;..,c, ~ e

c.. : )' <. '?.;.;;[ :.,

T -,.* W 1,.i.,. : ,. (- L. ~,,

-s

.. q:;, *... . v.

.v.

v

Q y

f..

.. 4,., -(, _ ty ;.. ., ?. w.,.,.:; ;..,. O;.,. g... [.. f. -.....-.. :,%... ,y

4...........,.

e. . v. l -1... n- . c 3 % *f....f.,J.v

k.' f?,l'll ; k;. y,? l4

..G ' ' p'l.~' w:- % 1~. Q '..:.l. '. ' ' ( e-c *.Mr.-. T. *,..,*., r? _. '0' ...'%:... i 8....' ,L. 4. *.. -' * -.- Q.'.. ;l.. '.. h, , J 6 ,.4 1 . T n J. '.+ 7. ?.l A: ::' r.. '.: >.. ^ ~..,...,.. 1. .y

3.. s. p..

t < %p'f,[. - T.; ~r '.' V.. \\ > ' ;..;.'..,, '- M ; /

u. s ' k ' -

e.,, f., h. v? t.. '.. :,2;g

, 't. 7,.s'.. ~ . C ' *' ^ .'A, ^ . [ 't,f-' wi 3 - : ~; ; '. '. 4..... ), *. .,,', ;..: 1 * ? .... %,.(y

. '.i.,.,.

w ., -. ~.. m X.y ? '.

}Nc

g*N.

l.Q?\\( L g%>*:.._...- };g*":.*.; g]'..;. ' ,.[*~'.., g , } } ' *%..a.' ' + . -~_ _? g... ? g '.,*]..c. .. *v % ~.; \\. 'l -'_'__.v . ' i.. \\, I I' }l e, 3 +' } .4,'- .*\\., '.- '. ' ',..,, -y..,. 3. ,,._n. m i'*:.., ~.,

b. :. 'y\\ ;. "..',, ;'.,,

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

• 9.,

. sr.- ; .,. f.g ;,. g,..,...: ),. L t. e., =,. -{., y .v y - ,.. sp f ^ --g...., f, ,4., ,a ' ;, - ....i..'1-- . e. 27 e_

• ,,g *' 7

'..; ; y., l *,,,, i % h. a Y, c - Z.'. A. * ,\\l', ...-l. ,' :;. l '; * (% T5 )*,*- L 'e l'... 'C:'. ' -._ I - )' , ;.. a.

• ,.. m6

~*,'.). 2 d . ' '+., iQ i n.. % :- -,}g. y 3 r ' 4.\\ v,' ~' ' '.,. g. ..4

f.,

.y* h,. ; V '. z %, T

y

.. :v. i. T. y;.,%,.yA.&M,3:;:. y#.M ::m:gG,,., c g y n% r.y.. : a., 1 :,;..L. v ... M...::..c.; - :.,).,. --. QW s:g.7 . s_.7.,7 y'l ~, -

. - i ;

i.l.... ;....n. s....

y.,.

c^ y':.> 4 l. l O, n n. w v%y..,.3 p c:,;f.9.L<,. ),. Q ^s .. '...p ~ m 1:..n J . c - 9 j Q : y. !.. ~ 'i

n{ ^

;Q

n.,.,... 3 ;v;.. '..,Mer.:.

,'7:,q.% :. g' i *, ~l '. ':- . t., ; ;7..'.h ;;.; r,. i j _,.,

l. ' ' Q

_. y.:. ( :- . y a : 7.., .z

_ _ _ - n

*..,

.n\\ r :: . w. :.. x. ::x *c.y, a. s_ ' } * ; f,,. [ Q..... y, f; - ;;. ; .. ~.. c' '_;, V :.,\\, [.. ,s.*:: y l[ - }. " ?f.; ' - j ^ f _ [y., ' g [.; . ; f:_ R. 3.; ;.1 ;,, yW ; L.'.;- a, }.;; '* .a 7 +. [;7.';; QQ._ y:;. l. ;%[$. < j j Q: Y '; '!;:l;. jff ..'.y ;. Q:3_.[ --, S g:y.};;e [,,..g ;y[. t.g.. 3 + e ne e,s.:.. - a.# p.; y. ' q,.s a,le yq;;..;.. $ - f. - 7y y m A 7... ...3 _e... ; : a. W.. O n m- -Q.?:.5 yg:. wn a.i : a:... . y.:.y.: 7 :p., ::g:.:.b.e :...; g. .g. & ?p.k.;e. sy glff %0.e 0[&.&yr;%q?)g].;L;:_d %dhyd.6 G f.: y.' b lt

w. M M::::. u f.

pi. 7 p...y.. 4y:~.c:. :;. p..y M.e.n. -

s. g ; : y

?* - ; ng M n wgw # - + o *....

m. mw

.: g f.y ;;..:A;. <m. m. 9. gn.. W.n... y w.. C - a/.3 g 7;0 5.: p*.:,q :.'m y :._.q.. n (; y y. c W.. s r .g '. Cl.. 9210)40267 92i006 .73 g...... < (. -.; w. Q% 4:: & :._a. PDR ADOCK OD000395 _ 4;:% :: n':i. ? r

s. w..

y i.J;f. )A.F G :- . p. 3.. y%. :.3.4.y... - e-ik e ?? y .,c s ; Q y. ;_ *yj PDR f .[(_ap.. _:? .: J.; m - xn o.. .J. 9:-; g:. w a e ..f e - p;4.M.:( a: y v_

WESTINGHOUSE PROPRIETARY CLASS 3 WCAP-13209 EVALUATION OF PRESSURIZED THERMAL SH0CK FOR V. C. SUMMER J. M. Chicots M. A. Ramirez March 1992 kork Per.ormed 'Jnder Shop Oroer VCSP-9001 Preparad by Westinghouse Electric Corporation for South Carolina Electric snd Gas Company Approved by: 7k M M T.A.Meyer,Makager ) Structural Reliability & Plant Life Optimization ? WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 C 1992 Westinghouse Electric Corporation Ali Rights Reserved ) __ m.

3 TABLE OF CONTENTS P19st Table of Contents ii List of Tables 11 List-of Figures it 1. Introduction- -1 ~ 2.- Pressurized Thermal Shock 2 3. Methods of Calculation of RTPTS 3 - 4. Verification of Plant-Specific Material Preperties 5 i' - 5. Neu troit Fluence Values 7 6. -Determination of RTPTS Values for All Beltline 8 Region Materials 7. Conclusiors 12 8.- References 13 i .i t 4 ~- w .n .n- ---..A

LIST OF IABLES -11hlg Title Eggs 1. V. C. Summer Reactor Vessel Beltline Region Material 7 Properties 2. Neutron Exposure Projections at Key Lu;aticas on the 7 V. C. Summer Pressure Vessel Clad / Base Metal Interface for 32 and 48 EFPY 3. Calculation of Chemistry Factors Using V. C. Summer 9 Surveillance Capsule Data 4. RTPTS Values for V. C. Summer for 32 EFPY 10 5, RTPTS Y"iacs for V. C. Summer for 48 EFPY 11 LIST OF FIGURES Fiqure Title East 1. Identification and Location of Beltline Region 6 Mater1 for the V. C. Summer Reactor Vessel 2. RTPTS.versus Fluence Curves for V. C. Summer 12 Limiting Material - Intermediate Shell Plate, A9154-1 ii

v 'l. INTRODUCTION A limiting condition on reactor vessel integrity known as pressurized thermal 1 shock (PTS) may octor during a severe system transient such as a Such transients may loss-of-coolant-accident (LOCA) or a steam line break. challenge the integrity of a reactor vessel under the follwing conditions: severe overcouling 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. 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 (FWR) vessel embrittlement as measured by the Ill. RT m nil-ductility referance temperature, termed RTPTS screening values were set for beltline axial welds, forgings and plates and for beltline circum erential weld seams for end-of-life plant r The screening criteria were determined using conservative operation. fracture mechanics analysis techniques. All PWR vessels in the United States have been required to evaluate vessel embrittlement in accordance with these criteria through end-of-life. The Nuclear Regulatory Commission has amended its regulations for light wa er 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 Jane 14, 1991 f23 lnis amendment makes the procedure for calculating RTPTS values consistent with the methods given in Regulatory Guide 1.99, Revision 2I33 __. - + - - ~ _ _ - _ _. _ _ _ _ _ _ _ _ _ _ _

4 The purpose of_thi : report is to detcrmine the'.RTPTS values for_the V.:C. Summer reactor vessel and address the. revised Pressurized Thermal Shock;(PTS). Rule. Sectioa 2.'discussas 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 V. C. Summer reactor vessel. The neutron fluence values used in this analysis are presented ir. Section S. The-results of the RTPTS calculations are presented in Section 6. The e.onclusions and references for the PTS evaluation follow in Sections 7 and 8, respectively. 2. PRESSURIZED THERMAL SHOCK 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 potatial 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 cxperience 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 postulateo 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 submit projected values of RT PTS for reactor /essel beltline materials by giving values for time of submittal, the expiration data 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 efective c 'te of this Rule if the value of RTPTS for any material is projected to exceed .g.

l the screening criteria. Otherwise, it must be submitted with the next update of the pressure-temperature limits, or the next raactor vessel surveillance capsule report, or within 5 years from the efier.tive date of this Rule enange, whichever comes first. These valucs niust be calculated ta.;ed on the methodology specified in this rule. The submittal must include the following:

1) the bases for the projection (including any assumotions regarding core loading,natterns),

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 NRC, justification trust be provided.)

• The RTPTc. (measure of fracture resistance) Screaning Criteria for the reactor vessel beltline region is 270'F for platos, forgings, axial weld,; and, 300*F for circumferential weld materials.

The following equations must be use to caltelate the RT PTS values for each weld, plate or forging in the reactor vessel beltline: t;uation 1: RTPTS - I + M + ARTPTS s Equation ? ARTPTs - (CF)f(0.28-0.10 log f) Ali values of RTPTS must be verified tn be bounding values for the spacific reactor vessel. In doing this each plant should consicer plant-specific information that could affect the level of et brittlement. 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 operation beyond the Screening Criterion is renuired. 3. MtTHOD TOR CALCULATION OF RT PTS In the PTS Rule, the NRC Staff has selected a conservative and uniform method for c;eter c r j plant-specifi' values of RTPTS at a given time. For the purpose of comparison with ine Screening Critieria, the value of RTPTS for the reactor vessel must be calculated for each weld and plate or forging in the beltline region as given be'ow. ' _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ = _ ____-___-_- -

RTPTS - I+M4 ARTp;3, whero 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 uncertaintie.c 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 rocasured values of I are used, f-Neutron fluence, n/cm2 (E > IMeV at the clad / base metal interface), 19 divided by 10 CF - Chemistry factor from tibles[2] for welds and for base metal (plates and forgings). If plant-specife surveillance dc?c has been deemed credible per Reg. Guide 1.99, Rev. 2, it may be considered in the calculation of the chemistry factor. v 9, __-__i-_---_-_--__------------------------ ^

j 4 VERIFICATION OF PLANT-SPECIFIC MATERIAL-PROPERTIES Before performing the pressurized thermal snock evaluation, a review cf the latest plant-specific n:attrial properties was performed. The beltline region is defined by the PTS RuleI23 to be "the region of th> ecactor vessel (shal material including welos, heat affected zones and plates or forgings) that directly surrounds the effective height of the active core and adjacent ragions of the reactor vessel that-are predicted to experience sufficient neutron irradit.' ion damage tte be c.onsidered in the selection of the most limiting material with regard to radiation damage," Figure 1 identifies and indicates the location of all beltline reg;on materials for the V, C. Summer reactor vessel. l' Material property values were derived from vessel fabrication material-ceri;fications, past neutron irradiation-induced changes in the tension, fr cture and impact properties of reactor vessel materials are largely-L dependent on chemical composition, particularly in the sopper i i concentration, The variability in irradiation-induced property changes, which exists in general, is compounded by the variability of copper concentration with the weldments. A summary of the pertinent chemical and mechanical properties of the beltline region plate and weld materials of the V. C. Summer reactor ve n a!.re given in 'able 1 [4]. All of the initial RTNDT values (I-RTNOT) are al t presented in Table 1. l: l. l i p l

m. _..,

r 4 -.,

4. VERIFICATION OF PLANT-SPECIFIC MATERIAL PROPEF. TIES Before performing the pressurized thermal shock evaluation, a review of the latest plant-specific naterial properties was performed. 1 i The beltline region is defincd by the PTS Sule(2) to be "the reoion of the reactor vessel (shell material including welds, heat affected zones ano plates or forgings) that directly surrounds the effective height of the active core and adjacent regicns of the reactor vessel that are predicted to exrc~ien. ;ufficient 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 V. C. Summer reactor vessel. Material property values were derived from vessel fabrication material certifications. Fast neutron irradiation-induced changes in the tension, fracture and impact properties of reactor vessel materials are largely depenc'ent on chemical composition, part icularly in the copper concentratita. The variability in irradiation-induced property changes, which exists in general, is compounded by the variability of copper concentration with the weldments. A summary of tha portinent chimiral and mechanical p.aperties of the beltline region plate and weld materiah of the V. C. Summer reactor vessel are given in Table 1 (4]. All of the initial RTNDT values (1-RTNOT) are also presented in TaLle 1. I

.. _ = _.

+-.

t CinCUMFERENTI AL WELDS LONGITUDIN AL WELDS 0* A9153-2 .s [ 4 t 13.1" f 86* b E CORE u 0 2700 90 ~ CORE -j w Q L _I 144.0" g \\ s A9154-1 m W 1so* E ' MIDPLANE 3.0" M a o* C9923-2 y i I i J d J l 5 .(_,5o \\ coaE-E 90o g 270* o 1- .a 49.0" 1 I I C9923-1 " i 8 180 Figure 1. Identification and Location'of. Beltline Region Material for the V. C. Summer Reactor Vessel l- _s_ t -.c ~

TABLE 1 V. C. SUMMER REACTOR VESSEL BELTLINE REGION MAlERIAL PROPERTIES CU NI I-RTNDT Material Description (%) (%) (*F) l Intermediate Shell, A9154-1 0.10 0.01 30 Intermediate Shell, A9153-2 0.09 0.45 -20 Lower Shell, C9923-1 0.08 0.41 10 Lower Shell, C9923-2 0.08 0.41 10 Longitudinal Welds 0.06 0.89 -44 Circumferential Weld 0.06 0.89 -44 5. NEUTRON FLUENCE VALUES The calculated fast neutron fluenca (E>l MeV) at the inner surface of the V. C. Summer reactor vessel is shown in Table 2. These values were projected using the results of the Capsule X radiation surveillance programl43 TABLE 2 NEUTRON EXPOSURE PROJECTIONS

• AT KEY LOCnTIONS ON THE V. C. SUMMER l' NIT 1 PRESSURE VESSEL CLAD / BASE METAL INTERFACE FOR 32 AND 48 EFPY EFPY O'

12' 30' 45' 32 3.87 2.85 2.04 1.33 48 5.81 4.27 3.06 1.99

• Fluence x 1019 n/c.c2 (E>l.0 MeV) l 1

m

~6. DETERMINATION OF RTPTS VALUES FOR ALL BELTLINE REGION MATEP.lALS 'Jsing the prescribed PTS F.ule methodology, RTPTS values were generated for all beltline region materials of the V. C. Summer reactor vessel as a function of end-of-life (32 EFPY) od 48 EFPY fluence values. The fluence data ware generated based on the most recent surveillance capsule prngram resultsl4l, i lhe PTS Rule requires that each plant assess the RTPTS values be. sed on plant specific surveillance capsule data under certa'n conditions. These conditions are: Plant specific surveillance data has been deemed credible as defined in Regulatory Guide 1.09. Tsevision 2, and e RTPTS values change sign.ficantly. (Changcs to RTPTS values are considered significant if the value determined with RT PTS equations (1) and (2), or that using capsule data, or both, exceed the screening criteria prior to the expiration rf the operating ~ license, including any renewed term, if applicable, for the plant.) for V. C. Summer, the use of plant specific surveillance capsule data arises because of the fcllowing reason: 1) There have been taree capsules removed from the reactor vessel, hace the data is credible per Regulatory Guide 1.99, Rev,sion 2. ~

2) The surveillance capsule materials are representative of the actual vessel materials.

The chemistry factors for the intermediate shell plate, A9154-1 and the walds were calculated using the surveillance capsule data as shown in Table C. 91 Tables 4 and 5 provide a summary of the RTPTS values for all beltline region materials for the end-of-life (32 EFPY) and 48 EFPY, respectively, using the PTS Rule. b6 l l l

TABLE 3 CALCULATION Oc CHEMISTRY FACTORS USING I V. C. SUMMER SURVEILLANCE CAPSULE DATAI43 Corponent Capsule Fluence FF DRTWDT FF*DRfkDT (FF)"2 Int. $hett, A9154-1 U 0,639 0.874 40 34.980 0.765 ('.one. ) Y 1.470 1.107 60 66.405 1.225 X 2.460 1.242 50 62.138 1.543 Int. 5hett, A9154-1 U 0.639 0.874 30 26.235 0.763 (Trans.) V 1.470 1.107 40 44.270 1.225 x 2.460 1.242 35 43.476 1.543 277.473 7.065 Chemistry Factor = 277.473 / 7.065 39.273 Weld Metet U 0.639 0.874 30 26.235 0.765 y 1.47C . 107 45 49.804 1.225 x 2.460 1.242 35 43.476 1.543 119.514 3.533 Chemistry Factor - 119.514 / 3.533 - 33.832 y E d .g.

-TABLE 4 RTPTS VALUES FOR V. C. SUMMER FOR 32 EFPY. ARTNDT(*F) + -Ir.itial RT + ' Margin RTPTS-NDT liaterial (CF x FF*) () (*F) ('F) Intermediate Shell -65 1.35 30 34 152 Plate, A9154-1 (39.3) 1.35 30 34 (117) Intermediate Shell 58 1.35 -20 34~ 92 Plate, A9153 Lower Shell 51' l.35 10 34 113' Plate,- C9923-1 Lower Shell 51 1.35 10 34 113 Plate, C9923-2 Circumferential 82 1.35 -44 56 123 Weld Seam (33.8) 1.35 -44 56 (58) longitudinal Welds 82 1.08 -44 56 101 (33.8) 1.08 -44 56 (49) () Indicates numbers were calculated using-surveillance capsule data. Fluence ' factor based upon peak inner surface neutron fluence of 3.87-x 10l9 2 n/cm [4],.except for longitudinal welds. For lonoitudinal welds, the fluence factor is based on a neutron fluence of.l.33 x 1019 n/cm2 [4] at the inner surface of the weld. 4, f '

TABLE-5 .RTPTS VALUES-FOR V. C. SUMMEP FOR 48 EFPY ARTNDT(*I). NDT + Margin' +' Initial RT RTpyS; Materi.1 (CF x FF*) (*F) ('F) ('F) Intermediate Shell 65 1.43 30 34 157 Plate, A9154-1 (39.3) 1,43 30 34 (120) Intermediate Shell 58 1,43 -20 34 97 Plate, A9153-2. Lower Sheli-51 1.43 10 .117-Plate, C9923-1 Lower Shell 51 1.43 10 34 117 Plate, 09923-2 Circumferential 82 1,43 -44 56-1291 Weld Seam (33.8)- 1.43 -44 56 -(60) longitudinal Welds-82 1.89 -44 56 109-(33.8) 1.89 -44 56 (52) () Indicates numbers were calculated using surveillance capsu'e data. Fluence factor based upon peak inner surface neutron fluence of 5.81 x 10I9 2 n/cm [4], except for longitudinal welds. For longitudinal welds, the fluence factor is based on a neutran fluence of 1.99 x 1019 n/cm2 [4] at the inner surface of the weld.

7. CONCLUSIONS As shown in_ Tables 4 and 5, all the RTPTS values. remain be;ce the NRC screening values for PTS using the projected fit:ence values for both the-end-of-life (32 EFPY) -and 48 EFPY, A plot of the RTPTS values versus the fluence are shown in Figure 2 for the most limiting material, the intermediate shell_ plate, A9154-1, in the V. C. Summer reactor vessel beltline region, 300 SCPEENING CRITERIA '250 20o C D j M 150 / USING SURVEILL. CE CAPSULE DATA 100 / 50 4 o 2E + 19 4E+19 6E+'9 8E+19 1E+20 2 FLUENCE (X 1E19 n/cm ) Figure 2. RTPTS versus Fluence Curves for V. C. Summer Limiting Material - Intermediate Shell Plate, A9154-1.. .M-

F 8. "~FERENCES-1 [1].10CFR Part-50, "Analfsis of Potential Pressurized; Thermal i Shock Events," July 23, 1985. [2] 10CFR Part 50, " Fracture Toughness Requirements for Protection Against Pressurized Thermal ShocP 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-12867, " Analysis of Capsule X from the South Carolina Electric'and Gas. Company Virgil C. Summer Unit l-Reactor Vessel Radiation Surveillance Program," J. M. Chicots, et al., March 1991. (Westinghouse Proprietary Class 3) E' .}}