Text

Nonproprietary Rev 1 to WCAP-12789, RTD Bypass Elimination Licensing Rept for Vogtle Electric Generating Plant
	ML20070S344
Person / Time
Site:	Vogtle
Issue date:	03/31/1991
From:	Di Tommaso S, Sterrett C; WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	;
Shared Package
ML19310E740	List: ELV-02627, Forwards Nonproprietary WCAP-12789 & Proprietary WCAP-12788, RTD Bypass Elimination Licensing Rept for Vogtle Electric Generating Plant. WCAP-12788 Withheld; ML20070S344;
References
	WCAP-12789, WCAP-12789-R01, WCAP-12789-R1, NUDOCS 9104020246
	Download: ML20070S344 (83)
	v • d • e

.

g, .. ~ . . .-

<r

. g

. J '] . . * . . ,- -.

- .. . - ..,?, :

. Nr ~ - - , ' . . .; .,,.y; T i p,Jg ' i4'c' . , ,.y , . . ,

, w. 7.

, s.. w.

j ..,.

. . . . .; . p.....

..,,.3- .. , , , -

, . , . ! .d . <-l. ' '. ' *

' J i.' ' - . e,.. " . . ' - ; -',..'j,. ..:lY [ ' .,' . ;: \ * . - -3, ,,: ":.

f).,. . '..' ,* - -h.Y

' N[*' b ' . . *;, ' , _

' hQ,-.*; ,:

~ ^%%, .,'{f" '

. : *. . ; s T y:;p;;. .

. .~.

s e

.sj( ,' '. .' ,I 2 _

3l '- - .- ' N; *, - '4',. : - . ' ':/ ..

' . . ' . , ,. -g .f-. ,, ' ,4.h, 5

.. ( ' , i **. ' i , .- -

, jl - ef . . .

'; .gf; . '...i P g,Y p) g.*h I. , * . c,r . ' ,' : - ,, .- . ., , ,. * * - - . . 7 . 4 ,.-' . , ;.,, .

.' -W. .;,..,

. -* - , ?/ Q .

T$,

- ' ) E '*'~ - ;, - I. .

, . ; [ '.k ' ;' . - .

Q..,' .

h W(:' .' 3 -; ' +- 0, ' "-

;,(- ',,'

) '.

, f, .

~

-(. - ; ' * ,. [- k' . '

,[ , k ., .

.] '!. .

f/ , 4 .

h.hb.b.h.h .J~c

N.ht .h[gh h h (b: .'[.';*.

1 -

h; Y

-.?' 'M[j.h,d.*, i ' ' . f. dl .0 't,t-h) (h.<

. u- .

'.".-5 s ' ' :

. ' . d L .h - ' o W ? 5 I.

I D * ;. * * * . . -' "- ,,i #

I *"'

S '.' i d'

. ', H( ' .' / ' f . .j ' ' . . ., ."'- @i

!"f M

.' L.f. l' . '

- [W

'f

. tyd.pd.j3

'I 6* 8

s. 'J.

N

...r. ','.f gc

<.,f-c :

. :p f@Ia p 'i ; -

r.<.% -

s : .n .

f, ' e

-- 2

..,+[,,\.

,.- 4 , 9 ', . . - ,-p

,.Q iN j .s . , , . _.

J . .. ' f ,g . _ , . q t q.\s .', ,' , j -

..rg 8 gy cy - . = , l .: ,

g-

.c. 4 =,. s i j.% . . ,y, g q

. .. .. .c e1 .

?f;Y. ..

.~ .. y W ,

~

r,f.,. ,Wh.: b 'O'. *'

,~ $,f,7.. .; nfl.N(1. -

N[h+h.W#f-l.l.O. :.,

.? ; ' Nik.hlyN. N Q?'s: i &g,h N g,l! lO.

.' .,Y:"..',lf Q f*f. . ' ; , ' ' ~

O 6 - &m {,n,\Q. o.f. v,W. c.' . .' u

.. .WY.f.,7...',f.Q: ' \h ll,? . ,.h 'S ,,9 .?W

1 " . . ... ."%.f. ., g.f.,',',:A. . h, .?, ,!y?v. .

a .

a . 25 . ,_ . . . '. . , = . ..i e- -

bQ. A'# N1.,.,*. a * ,e fj :;

. - .i. ; .. g p*. pgi.......

g.m. . v. m, . .

r

. 4,i,.

. . 4. ,, = ,' w. ;x . _". ,- .

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

, 9H.& 5; J. : ,s.. . - :- . -

lc ,* .

. . .;: , '. .U . ;. g . .

+.y Q ;f .;.;, ..,'^.,;,.,,f ]. '

, . yg*.(.P.l. ,. ,,

U: 4.t h;$ ' . -l '} f 3 ' 9 ..%

- ' ' A'**

- f "A%f . :f.:3 ' J./. / p* , ' f, . *; 9.T.,i.J7*, ,"m';" .g. ., . ..

~ '

m- . .4

.. .- , .. . - 3

, . . y ;!..M,,,,? 4 . n. , . . :.p'... -

I'

? .;;

l h;b,

' _' ~

h . h.~y: ','1 r

69.p.

.1. J .. 'y ...~ .' . . , - - .-

?.;; ? '-

Qe' k '. Y, .

. s e.

, :;;;l: ;.

' ~

p.% :,,.h;;Q.. g.J;. c,: ,s .:.~ .. ,i', - * -

),'

d[ . : , h _ ' ;' . ,. .- .-

,' *, v. . v N e . . -(- s.. : c...e%'s p .t ., . :e. . -

--a

t.

.....g W, ; *1r.go s.q- .?

.4 n:.'.,--.n'.:

- ; 4 * ,*

,r+s t %, i: . :0. . . -

e

.l, .-

.a g.t.;,.. ,

g ' :S.ys rg- 4 - p. e , . - - -

'. . . ;.,~ , rw . . y [, .Jy . 7 .,..

o QF.3, ; , ., t>m ;.. ,

o

g. . l.'. . . , * . .,'

- t 7, ,

p.a y,. 'l. :

,<.f. c, %] ( , p . y, ,m . .y/. e .. h. ,. , ' '

c

, ,g ,

( L ;'.. : ,- -.. l' .**

y.A .

.:,-; 6 ,,3' . yMa.>* y . ;. 3;; - l; @.

,v ,'.,y

-i

, - - . . . ,f. ,- -,, g :gf . , : ; , .

it , . . . - .

c. ; g yNA . : .i; , ,g ,,

j, f ' , $*"'. ,.,;.' i , U S .'., 3.) d I ' '

,~ '

. . . p?gf.

i

.k , ,f[ .g 5 'e

.a

'-[..

,4 , :.

' , ~ s [}-. -l ;- I ,. . ' ,. y I - f ' -,

- / ;. J.3 .,, r li v . '- _ 3 .. . -

.i s ;-i , , (y. ,. ',' .- . , '. , ' '

e' - ' '. ., , , ,

/nh' M@,'M* ft4

'. c .:' b q %

1.-i,'.s ('; 4 h. ) ,, . ., T (.(ji4 [.h ilij,.Q.- ;. '..

i J e.."

.,.e . . ; , '4([ -

M '

'h ) T .,t[,h,i, ,( j , . , :'!,).' . , , , ', -) . L .: ..F.i - " :I/ :

..w-.U:N.-.

,? .

- [ - ' ,M. ' .' y'g . 9 '?i, E .h. .I-g ..' - -

a ; - , i 5

q4 . )~ J ., ?

f M [- Jl' S'; k$f. i '..., . I . i;

[. ih.M[. h '. ,, . *,. .

T .T-i' ,'a

I.j,{ ,?$ G , ,, 3 , . v - , !, .k(j . f.a. ,gI Ql 'f :. *; .

'; , r,_ ..'.l'.

hh\g . ' ,: , '. . . .- - . : \if. t:.

3 -

a ,n l :- - . ;,. ,

. 'p p g/,;, g.g .

- * ; - , 14 .. i . -- , . . , .

. . 4 . i, i P ] yn . , ' , . . ,. 4 9 c

'.'..',..'t. .y Kg, r .:

. ' . P i'l

.e.,*.,.s-

?, :

".. i..spP t, I ' . ', - g ' , -

r..

. ., t ' -

. ,) . 8 4 . l y: .. . . - l .' y y' .s,. y.'.., .

N' ;.M i

4t d1 ';.

.: 4

-; Q- ,' * .; 8

.1 b- W :.

j 1,,), .k,gp. .. g f j.II,, . " .e.g.,..

t,.- *',l m..

g , .-

. ;, . . . i , i i

,g

, ' , ' ,,. .. . . k. . ...,

., - .i.- ,,,,,')

- , < * . '. ( '" ' - - '

. f, . 6 - J,,,,., 'i,,,

..,l-,. .,, *(

i D., e r, .)

.y. .

,,* i.acy .- ..-' --5 ,Jy, c,4 ; .i , . - l .;. , , . .

p.. . ,,Q , -

. ..! j a, j - <

s

4. . , ..if , ; ., - . . , . . .

,, , .' > a g ',1!,,'1, e ,.i, ty,Wp , . f. , . , ( ..' ., ,,3 . ,,

,1 ~.!

f. '" 5 N;', ' ' . ;u .u C,j .: '..,*,Q' :, '\ f' .' i ' ' . , . ' .

l' ;':y , ,. ,,:. . ' . l - ' '

'. : ' :' - -h:

- -. - ~'*<..: Ia.p .. , f k[ '.

} , i (, ,:y' .

. ,; i 't ,,

.) - ' . - a

.r. . .. . ., r .. . ,e ... s. .. . . .

C .f 3 l. ... _ f . \ .w.. ) j.*-[.'." ' *^ ' ' -' . ' .] \ . } .g& l .'. - f . 4,,,e.

, . V ) ' fu '. . . . :

. ';' "g s.[.3,s. .U :' . . l !. ' , ! ,' , A - %,e

=, . v:. ..: ...u( . . y -3,g.s . , ..% . ' ; 3. ,;-. A 9 }'y.; [';'_ ,,.; ' - . ._. f. f v..

q 1 - . .

~-

? %.. ., ,.- . . T. f ',l % . \.,( cR.: ._ :- ) Q n, . G. ' * ',

t ,. .- -

. .;1 ., ,

i .., -: A s, Q ; ;; ; . . J .'.' >. . s. _ ' .

.tl- o ', , , : -- , .. , q)s,. .. ;- '.c.* *_ ; ' - )..-! , .. - .

., .. s

. ..,f,-'.3 .c *v . e _. ._.y ,i,.

Q. f. i. .' q)' *

.et.*

-9., . . 534 ..-. ? ..,',s'. - .: p,i. .-

c..

.I

, , . g..V ;yc.. v, -.' .

. v -v._- .- : . .

"y- t Q.,.

.i.. .

a

$ . . . d. f,

,. 3 ,...

- s - '-

,......,,g. M .i,

.' .* .6

gA*) 'F . ?... ..

. ; . .. - - .-

f. l . ,F,. . s.

. t'. - - a *,( . * ,

r *-. -' , . .f . .

.a ,s ' - :

(': . ' "lL ' '- '

~

.< T...

.:.>:.t',

.s., . +

$ ,. / , ". - -

.l',.. ,ff . ' Q 6,lg., la, ' * * ' - i' ' ' . . , . , ..y,,,, t

- %,, ' . .' , ' *. , . -. , 1, '. p g' .. ; I'

, ' v, . ,'/ *. Y. ,

,'Y *C , ' ' . ' , ,y

.,,:' f .' [f ',? . ... .

' . .;. : 3, 1y,..- - '. ,g .u s;.... ., . - . , ,

n. -.'. .

. . yu. 4. ... * , , . . , . g . . . . . ,g s . . -

e- : * -

. . . .. . g ,y . - . , .. . . . . . . .. , ., ; ,a

*. c .

'e N 'l. y.', f.. ' , f . h I't .

'* . ' ' .. ' y ,hl, ,,g j . ' _, ,_ E = Vl . ':. '?aj '. . _ .

' . . ' [ ' :. - .

-[ .. ' . .z.. ' -_-~. ., '

a _... y..iyi.s' . e:.'n. ' M;.'.,'N

~~*^. ' ' - *

-~,.;._ ;\ .

t,;

_1 y ;l,; 'r,r3 e

> , n. . .,. c. : ...A ,

y. ..,,....t.,-

y .

' . !I.Q, , k ,,;g

',.'[

,'.f . p .'f. \ ; * . .

. ::M ly { p

.'.'*A y.

. . . ; p . g. , g .. ..c,.* ,: -- .

. , %,.L.

q\ ,','.

.= .

,y ._' ' . ' ' . , - .

-'.',_,..*..'-'.f

. .r 4* .

~'../

e ., . .- '..4-r,,... - ,

.if....s .",,,l ;,? , '..;

.i

..' . . - '. . - ' Vj. . . . l % );. g\a . '.

^ q q y . .. . . . , r ;9k:._ . .. ' :.

h,;3.

q- ....: y ).f,7

- .(

.;,e;a . 7 ; .c: . . -

^

._ ., ~ .;

.i.

.c n....i.

',.S . -'.* ... :. .

. ' ,. , ' , . ., ..[.y : . v t ..., , , .q. .. .m,m 'SE .-.. . '..? . n. .

- - f*

.- ~ ..

'* , h: ,

_. ' 'm

. ; M k c W .. Q ' ;:::x '.( W .; : l.f:n C ' , .' f: . y g ..n f .l. m =,.. : h- - y.. .h . u . . .: . 4.' v.

a.;y n.,p . .. ,;. ..

. - -.;... q: % . .. ./. .,:. 4. .~ _

n, %ow ww ly) 9..;+; . ..s.~ n m,c.. <f

%,. ..,.ee..M 8 ;r,. -, J r . e ;.. ..: :,5 - . .

.i p.

r

. < . .~ . :: r 3. .

-m.4 m$ . .n p i .... .y : *

. .c .....; 0 ,

r d . :. ' .. .

. f- N. . '. . - * ', p, _ ..;.. .' . - [ ,,: .-.'. ..l,'s, $:h, ,k ff g. .f j,gjs'. '..' " ~ ,.. : ' .; '- - .' : p ev<.'<} ;b. s .1 .,h ,y . m

& , ,e. . r

- ,f ,Q< m. f *-. . =.':,. *' . m' -a- ' :,

. e ...'. m. :. '"> -

.f ' <

M. y3&..n 4: g. C <dA . 3. w s.. + b.o3. . . t.. - e -.*r.:

' /,

.' - s ay u pW{ f, n* .

I.:- 3... ..h(

1e V.:: . .

pr g..:s . g? " . tr.9 t-A'......s

% f..d..-

P 9. .,". .s. th . ,- e '~.< Ar :r- a '* " -..*-

C%%q (. i sd' ' {-

i.. .

. ;UQ Q !. l. ;;)$9::lE ?).l.:<.~ .

~. . 4,& ] &. - f .' 'l .Q ;, ' ' l':W.Q : ,

' ' W M N W .'.. ..

.,l.'

w* ,', m?;l - s .:,::.

-* ' ";' .. ';%> c i e . w . .w . n '1. yet

3 *t

' W, } .( W/

c. Q h.ye,px

/l, ../.'14 .,.. w;< ( . r.i i . .. , y y. Ae.g ; ,.

,, m.

7, q c .m. .:m . ;, .

4 _ a ; ,; j':1.

r..

n.. . . . . . . , .

. .- _ ;- Q .- -

. -~ .

t.. . - 4 s . . . . . .

i .

WESTINGHOUSE CLASS 3 WCAP-12789 Rev. I t

RTO BYPASS ELIMINATION LICENSING REPORT FOR V0GTLE ELECTRIC GENERATING PLANT i

S. H. Di TOMMASO C. R. STERRETT MARCH 1991 s

i .

ACKNOHLEDGEMENT 4 The authors wish to recognize contribution by the following individuals:

H. G. Lyman G. E. Lang J. H. Brennan G. O. Barrett H. H. Hoomau J. A. Kolano R. N. Lewis H. G. Hillis I

0935D:10/021891

~.

i. ,

l a

TABLE OF CONTENTS Section Egg List of Tables 111 l List of Figures iv 1.0 Introduction 1.1 Historical Background 1 1.2 Mechanical Modifications 2 1.3 Electrical Modifications 4 2.0 Testing 2.1 Response Time Test- 9 2.2 Streaming Test 9 3.0 Uncertainty Considerations 3.1 Calorimetric Flow Measurement Uncertainty 12 3.2 Hot Leg Temperature Streaming Uncertainty 12-3.3 Control and Protection Function Uncertainties 15 4.0 Safety Evaluation 4.1 Licensing Basis-4.1.1 Response Time 27-4.1.2 RTO Uncertainty. 27 4.2 Evaluations 4.2.1 Non-LOCA Evaluation 28-4.2.2 LOCA and LOCA-related Evaluation '30 4.2.3- Instrumentation.and Control Evaluation 32 4.2.4 Mechanical Evaluation 35.

4.2.5 Steam Generator Tube Rupture Evaluation 37 09350:10/021891 i 1-- - -

,___m__,__,_ _ _ _ _ , _ _ _ , , , , , ,,,______,,,,_m _ _ _ ,a __,, _,___,

i .

TABLE Of CONTENTS (Cont)

Section Eagt 4.2.6 Containment Integrity Evaluation 38 4.2.7 Dose Evaluation 38 4.2.8 Technical Specification Evaluation 38 5.0 Determination of No Unreviewed Safety Question 39 6.0 Conclusions 42 7.0 Control System Evaluation 43 8.0 References 44 Appendix A - Definition of An Operable Channel And 45 Hot Leg RTO Failure Compensation Procedure Appendix B - Acronyms for Uncertainty Calculations 54 0935D:10/021891 11 I

1

'I l

LIST _ OF TABLES Tabla Title Ptgg 2.1-1 Response Time Parameters for RCS Temperaturo Measurement 11 3.1-1 Rod Control System Accuracy 16 3.1-2 Flow Calorimetric Instrumentation Uncertainties 17 3.1-3 Flow Calorimetric Sensitivities 18 3.1-4 Calorimetric RCS Flow Heasuremant Uncertainties 19 3.1-5 Overtemperature Delta-T Trip 21 3.1'6 Overpower Delta-T 1 rip 22 3.1-7 Low RCS Tgyg (Coincident with Reactor Trip, TurLine 23 !

Trip and Feedwater Isolation) 3.1-8 Cold Leg Elbow Tap Flow Unr.ertainty 24 f

3.1-9 Low Flow Reactor Trip 25 3.1-10 Technical Specification Table 2.2-1 Hodifications 26 4

4 l

, 09350:1D/021891 iii i

l t _ . _ - _ _ _ _ _ _ - - _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - I

I e LLSlfl FIGURES Fioure Iltle Pjtqn l 1.2-1 Hot leg RTO Scoop Modification for Dual Element fast-Rasponse RTD Installation 6 1.2-2 Cold Leg Pipe Nozzle Modification for Dual Element s' >

~

Fest-Response RTO Installation 1.3-1 RTO Averagir g i, lock D agram, Typical f r Each of 4 8 Char.nels i

l J

t i

0935D:10/123190 i '/

i .

7 !

[

1.0 INTRODUCTION

Westinghouse Electric Corporation has been centracted by Georgia Power to remove the existing Resistance Temperature Detector (RTO) Bypast System and replace this hot leg and cold leg temperature peasurement niethod with fast response thermowell mounted RTDs in:talh;# 'n the rertsr @i,Snt loop piping. This report is submitttd ror ne orpote of supyceting operation c,f the Vogtle Elettric Generatirg Plant (vtGP) utilizing be new temowg ?

mounted RTDs. j

, 1.1 blSTORICAL Bi<f/,t "ND

Pr cr to 19%, PWR aesigns ha1 been basid on the assumption that the not 1 0 i h teriperhtute as ur t.orm across the pipe. Theref0 e, p.acement of the g temperhture instruments was not cens.!dereo to be c. faro e affectla; f.;e ,

at:uracy of the measurement. The hot leg temperature wr.s measured witn direct immersica a.TDs enttendtrig a short distante into the pipo at one loc 3. ion By L =

the late 1960s, as a result of accumulated operating experiencs at sever &l plahts, the following problems associated with direct immersion RTD: were 1 identified:

o Temperatu're streaming conditions; the incomplete mixing of the coolant leaving regions of the reactor core at different temperatures produces F significant temperature gradients within the pipe.

r o The reactor coolant loops required cooling and draining before the r RTDs could be replaced.

The RTD bypass system was designed to resolve these problems; however, operating plant experience has now shown that operation with the RTD bypass

- loops has created its own obstacles b eh as:

o Plant shutdowns caused by excessive primary leakage thrcagh valves, flanges, etc., or by interruptions of bypass flow due to valve stem failure.

h 1

I 09350:1D/021891 1 f , 1

o Increased radiation exposure due to maintenance on the bypass line and to crud traps which increase radiation exposure throughout the loop compartments.

The proposed temperature measurement modificati0n has been developed in response to both sets of problems encountered in the past. Specifically:

o Removal of the bypass linet Aliminates the components which have been a major source of plant outages as well as Occupational Radias1on Exposure (ORE).

o Three thermowell mounted hot leg RTDs provide an average measurement (equivalent to the temperature meas" ed by the bypass system) to account for temperature str3aming, o Use of thermowells permits RTD replacement without draining the reactor coolant loops. ,

following is a detailed description of the effort required to perform this modi fication.

1.2 MECHANICAL HODIFICATIONS The individual loop temperature signals required for inp;t to the Reactor Control and Protection System will be obtained using RTDs installed in tact.

reactor coolant loop.

1.2.1 Bol_Lig a) The hot leg temperature measurement on each loop will be accomplished with three fast response, narrow range, dual element RTDs mounted in thernowells. One element of the RTD will be considered active and the other element will be held in reserve as a spare. To accomplish the sampling function of the RTD bypass manifold system and minimize the need for additional hot leg piping penetrations, the thermowells will be 09350:10/021891 2

m -

located within the three existing RTD bypass manifold scoops. A hole will be made through ti,e end of each scoop so that water will flow in through the existing holes in the leading edge of the scoop, past the RTD rnermowell, and out through the new hole (Figure 1.2-1). These three RTDs will measure the hot leg temperature which is % sed to calculate the reactor coolant loop differential temperatur6 (AT) and average temperature (T,yg).

t)) This modification will not affect the single wide range RTD currently installed near the entrance of each steam generator. This RTO will l continue to provide the hot leg temperature used to monitor reactor coolant temperature durir, startup, normal operation, shutdown, and post accident conditions.

1.2.2 Cold Les l

f a) One fast response, narr v range, dual-element RTD will be located in each I

cold leg at the discharge of the reector coolant pump (as replacements for the cold leg RTDs located in the bypass manifold). These RTDs will measure the cold leg temperature; which is used to calculate reactor coolant loop delta T and T avg. The existing cold leg RTD bypass penetration nozzle will be modified (Figure 1.2-2) to accept the RTD thermowell. One element of the RTD will be considered active and the other element will be held in reserve as a spare. '

b) This modification will not affect the single wide range RTD in each cold leg currently installed at the discharge of the reactor coolant pump. This RTD will continue to provide the cold leg temperature used to monitor reactor coolant temperature during startup, normal operation, shutdown, and post accident conditions.

09350:10/021891 3 l

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ i

1.2.3 Crossover Leo The RTO bypass manifold return line will be capped at the nozzle on the crossover leg.

1.3 ELECTRICAL MODIFICATIONS 1.3.1 Control & Protection System Figure 1.3-1 shows a block diagram of the modified protection system electronics. The hot leg RTD measurements (three per loop) will be electronically averaged in the process protection system. Tr.e averaged T hot signal will then be used with the T signal to calculate reactor coolant cold loop AT and T avg which are used in the reactor control and protection systems. This will be accomplished by additions to the existing process protection systM equipment.

1.3.2 Ouali fi cation The 7300 Process Electronics modifications will be qualified to the same level as the existing -7300 electronics. RTD qualification will be verified to support Georgia Power's compliance to 10CFR50.49.

The Westinghouse qualification program entailed a review of the Heed Instrument Company's qualification documentation for testing performed on these RTDs. It was concluded that the equipment's qualification was in compliance with IEEE Standards 344-1975 and 323-1974 as discussed in Regulatory Guide 1.89 with one exception. Specifically, requirements relative to flow induced vibration were not addressed. To demonstrate that flow inriuced vibration would not result in significant aging mechanisms that could cause common mode concerns during a seismic event, Westinghouse performed flow indu.ed vibration tests followed by pipe vibration aging and.a simulated seismic event. These tests confirmed that the Weed RTDs comply with the above IEEE standards.

09350:10/021891 4

1.3.3 RTD Ooerability Indication Existing control board of and T avg indicators and alarms will provide the means of identifying RTD failures. The spare cold leg RTD element provides sufficient spare capacity to accommodate a single cold leg RTD failure per loop. Failure of a hot leg RTD can be handled in twc ways. In the first, mihlal action by I&C personnel defeats the failed signal and rescales the electronics to average the remaining signals (see Figura 1.3-1 and Ser.ti on 4.2.3). The second method requires I&C personnel to disconnect the fa' led tiement and connect the second element of that same RTD.

}

i i

I j

0935D:10/021891 5

_ o., c w

Figure 1,2-1 Hot Leg RTD Scoop Modification for Dual Element fast Response RTD Installation 09350:lD/123190 6

, CL > C.

1 l

Figure 1.2-2 Cold Leg Pipe Nozzle Modification for Dual Element fast Response RTD Ia;tallation 09350:10/123190 7 l

i

-_-_ - ,-----,a--- - - -

l l , G. , C.

l Figure 1.3-1 RTD Averaging Block Diagram.

Typical for Each of 4 Channels 09350:10/123190 8

2.0 TESTING There are two specific types of tests which are performed to support the installation of the thermowell mounted fast-response RTDs in the reactor coolant piping: RTD response time tests and a hot leg tempvature streaming test. The response time for the VEGP application will be verified by testing at the RTD manufacturer and by in-situ testing. Data from thermowell/RTD performance at operating plants provide additional support for the system.

2.1 hSPONSE TIME 1(ST The RTO manufacturer, seed Instruments Inc., will perform time response testing of each RTD and thermowell prior to installation of the RTDs at VEGP.

These RTD/thermowells must exhibit a response time bounded by the values shown in Table 2.1-1. The response time has been factored into the transient analyses discussed in Section 4.2.

In addition, response time testing of the Heed RTDs will be performed in-situ. This testing will demonstrate that the Weed RTDs can satisfy the response time requirement when installed in the plant.

2.2 STREAMING TEST Past testing at Westinghouse PHRs has established that temperature stratification exists in the hot leg pipe w'.b a temperature gradient from maximum to minimum of ( l b,c.e. A iist program was implemented at an operating plant to confirm the temperature streaming magnitude and stability with measurements of the RTD. bypass branch line temperatures on two adjacent hot leg pipes. Specifically, it-was intended to determine the magnitude of the differences between branch line temperatures, confirm the short-term and long-term stability of the temperature streaming patterns and evaluate the impact on the indicated temperature if only 2.of the 3 branch line temperatures are used to determine an average temperature. This plant specific data is used in conjunction with data taken from other Westinghouse designed plants to. determine an appropriate temperature error for use in the I

09350:10/021891 9

safety analysis and calorimetric flow calculations. Section 3 will discuss the specifics of these uncertainty considerations.

The test data was reduced and characterized to answer the three objectives of the test program. First, it is conservative to state that the streaming pattern ( l b.c.e. Steady state data taken at 100% power for a period of four months indicated that the streaming pattern

[ l b.c.e. In other words, the temperature gradient ( ]b,c.e . This is inferred by ( Jb .c.e observed between branch lines. Since the (

l b.c.e into the RTD averaging circuit if both elements of a hot leg RTD fail and only 2 RTDs are available for obtaining an average hot leg temperature. The operator can review temperatures recorded prior to the RTD failure and determine an (

]b c.e into the "two RTD" average to obtain the "three RTD" expected reading. A generic procedure is included as Appendix A which specifies how these ( l b.c.e are to be determined. This significantly reduces the error introduced by a failed RTD.

Both the test data and the operating data support previous calculations of streaming errors determined from tests at other Westinghouse plants. The temperature gradients defined by the recent plant operating data are well within the upper bound temperature gradients that characterize the previous data. Differences observed in the operating data compared with the previous data indicate that the temperature gradients are smaller, so the measurement uncertainties are conservative. The measurements at the operating plants, obtained from thermowell RTDs installed inside the bypass scoops, were expected to be, and were found to be, consistent with the measurements obtained previously from the bypass loop RTDs.

09350:1D/021891 10 i

1

________________J

TABLE 2.1-1 RESPONSE TIME PARAMETERS FOR RCS TEMPERATURE HEASUREMENT RTD fast Response Byoass System Thermowell RTD Systra

~ ~ ~ ~

RTD Bypass Piping and Thermal Lag (sec)

RTD Response Time (sec)

Electronics Delay (sec) _ _

Total Response Time (sec) 6.0 sec 6.0 sec l

0935D:10/021891 11

3.0 USCERTAINTY_(ONSIDERATIONS This method of hot leg temperature measurement has been analyzed to determine the magnitude of the two uncertainties included in the safety analysis:

etiorimetric flow measurement uncertainty and hot leg temperature streaming uncertainty.

3.1 CALORIMETRIC FLOW MEAbVREMENT UNCERTAINTY Reactor coolant flow is verified with e calorin.etric measurement performed after the return to power operation following a refueling shutdown. The two most important process parameters required for the calorimetric measurement of RCS flow are the narrow range hot leg and cold leg coolant temperatures. The accuracy of the RT05 therefore has a major effect on the accuracy of the flow

, measurement.

With the use of three T hot RTDs (resulting from the elimination of the RTD bypass lines) and the latest Westinghouse RTO cross-calibration procedure (resulting in low RTD calibration uncertainties at the beginning of a fuel cycle), the VEGP RCS flow calorimetric uncertainty is estimated to t,e (

)"'C including use of cold leg elbow taps (see Tables 3.1-2, 3, 4 and 5). This estimate is based on the standard Wettinghouse tethodology previously approved on earlier submittals of other plants associated with RTD bypass elimination and the use of the Westinghouse Revised Thermal Design Procedure. Tables 3.1-1 through 3.1-10 were generated specifically for VEGP and reflect plant .,ecific measurement uncertainties and operating conditions. The setpoints presented on Tables 3.1-7 and 3.1-9 do not

, represent changes as a result of the VANTAGE 5 fuel program or this RTD bypass elimination. It.was determined that for these setpoints, the existing VEGP analyses were bounding and no technical specification changes are necessary.

3.2 HOT LEG TEMPERATURE STREAHING VNCERTAINTY The safety analyses incorporate an uncertainty to account for the difference between the actual hot leg temperature and the measured hot leg temperature 09350:10/021891 12

caused by the incomplete mixing of coolant leaving regions of the reactor core at different temperatures. This temperature streaming uncertainty is based on an analys'is'of test data fiom other Westinghouse plants, and on evaluations of temperature measurement accuracy for numerous possible temperature distributions within the hot leg pipe. The test data has shown that the circumferential temperature variation is no more than (

l b c.e , and that the inferred temperature gradient within the pipe is limited to about (

)b,c.e The evaluations of numerous possible temperature distributions have shown that, even with margins applied to the .teasured temperature gradients, the three-point temperature measurement (scoops or thermowell RTDs) is very effective in deternining the average hot leg temperature. The most recent calculations for the thermowell RTO system have established an overall streaming uncertainty of ( l bc.e for a hot leg measurement. Of this total, (

plants. -

t

)b,c.e ,

The new method of measuring hot leg temperatures, with ine three hot leg thermowell RTDs, is at least as effective as the existing RTD bypass system,

[

a Jc .

Although the thermowell mounted RTDs measure temperature at one point at the RTD/thermowell tip, compared to the five sample points in a S-inch span of the scoop measurement, the thermowell measurement point is opposite the center hole of the scoop and therefore measures the equivalent of the average scoop sample if a linear radial temperature gradient exists in the pipe. The thermowell measurement may have a small error relative to the scoop measurement if the temperature gradient over the 5-inch scoop span is nonlinear. Assuming that the maximt.m hferred temperature gradient of (

l bc.e exists from the center to the end of the scoop, the difference between the thermowell and scoop measurement is limited to

( l b,c.e. Since three RTD measurements are averaged, and the nonlinearities at each scoop are random, the effect of tnis error on the hot 09350:10/021891 13

leg temperature measurement is limited to ( l b ,c.e. On the other hand, imbalanced scoop flows can introduce temperature measurement uncertainties of up to (

j a .C, In all cases, the flow imbalance uncertainty will equal or exceed the

[ ]b,c.e sampling uncertainty for the thermowell mounted RTDs, therefore the thermowell mounted RTDs tend to be a more accurate measurement with respect to measuring streaming uncertainties.

Temperature streaming meu,ure.cnts have been obtained from tests at 2, 3 and 4-loop plants and from thermowell RTD installations at 4-loop plants.

Although there have been some differences observed in the orientation of the individual loop temperature distributions from plant to plant, the magnitude of the differences have been (

)b.c.e.,

Over the testing and operating periods, there were only minor variations of les: than ( l b.c.e in the temperature differentials between scoops, and smaller variations in the average value of the temperature differentiels. (

)b c.e, Provisions were made in the RfD electronics for operation with only two hot leg RTDs in service. The two-RTD measurement must be biased to correct for the difference from the three-RTD average. Based on test data, the bias value would be expected to range between ( lb c.e . Data comparisons show that the magnitude of this bias-varied less than ( l b,c.e over the test period. Appendix A provides a procedure for utilizing the actual plant bias data, Note that this procedure only allows the use of positive (or zero) '

bias values. This assures that the measured hot leg temperature for the 2-RTD average is maintained at or above the true hot leg temperature. A negative bias adjustment could result in a nonconservative adjustment in T-hot at reduced power. 4 0935D:10/021891 14

-_ - _ - - - - - - _ - - - - - - - - _ _ - - _-----------J

3.3 CONTROL AND PROTECTION FUNCTION UNCERTAINTIES Calculations were performed to determine or verify the instrument uncertainties for the control and protection functions affected by the RTD bypass elimination. Table 3.1-1 (Rod Control System Accuracy) notes that an acceptable value for control was calculated. Table 3.1-2, 3.1-3 and 3.1-4 I provide the uncertainties, sensitivities and final result of the precision RCS flow calorimetric. The total uncertainty for the measurement of RCS flow, as noted on Table 3.1-8, is less than the value noted in the VEGP Technical Specifications with and without VANTAGE S fuel. Table 3.1-5 provides the uncertainty breakdown for Overtemperature AT. As noted on this table, TA is greater than CSA, thus "iceptable results are calculated for this function.

Table 3.1-6 provides the 'oreakdown for Overpower AT, with the same conclusions as for Overtemperature AT. Table 3.1-7 notes the uncertainty 1 breakdown for Low RCS Tavg coincident with Reactor Trip, Turbine Trip and Feedwater Isolation. Again acceptable results are calculated. Table 3.1-9 is concerned with the RCS Low Flow reactor trip. Based on the earlier calculations for the RCS flow calorimetric and the rod control system accuracy, acceptable results are determined. Finally. Table 3.1-10 summarizes the resultant reactor trip setpoint changes necessary to the VEGP Technical Specifications. As noted, these changes were submitted with the VANTAGE 5 fuel technical specification changes (Reference 3). Note that Appendix B is included to provide definitions of the acronyms used in Tables 3.1-1 thru 3.1-10.

09350:10/021891 15

TABLE 3.1-1 ROD CONTROL SYSTEM ACCURACY Tavg TURB PRES

- +A*C pg , _

SCA -

H&TE-STE .

SD .

BIAS.

RCA -

H&TE.

RTE =

RO -

CA -

BIAS.

RTDs USEO - TH = 2 TC - 1

. +a.c ELECTRONICS CSA =

ELECTRONICS SIQ4A =

CONTROLLER.SIG4A =

CONTROLLER BIAS =

CONTROLLER CSA =

09350:10/021891 16

TABLE 3.1-2 i

FLON CALORIMETRIC INSTRUMENTATION UNCERTAINTIES O f

(% SPAN) FH TEMP FH PRES FH OP STM PRESS TH TC PRZ PRESS

+a.C l SCA =

M M&TE-SPE - !

l STE -

SD -

R/E =

i RDOT.

BIAS =

CSA =

OF INST USED 3 1 1 DEG F(l) PSIA (2) top (3) psia (4) DEG F(5) DEG F(5) PSIA (6)

INST SPAN 500. 1500, 120. 1500. 120. 120. 800.

+a.C INST UNC. ~ ~

(RANDOM) -

INST UNO.

(BIAS) -

i NOMINAL =

Notes:

(1) Final feedwater temperature from plant computer. A consorvative uncertainty value of 2.00F is used.

(2) Assumed. constant in precision heat balance (not measured).

A conservative value is used. .

(3) Measured with test d/p gauge. Does not include venturi uncertainty.

(4) Heasured with test pressure gauge.

(5) Uncertainty assumes temperature measured with DVH at output of R/E converters and the averaging of the three hot leg temperatures.

(6) Assumed constant in precision heat balance. The values are based on permanently installed plant instrumentation.

0935D:1D/021891 17

TABLE 3.1-3 FLOM CALOR! METRIC SENSITIVITIES FEEDMATER FLOW FA __

TEMPERATURE = '~]+a,C MATERIAL =

DENSITY '

TEMPERATURE =

PRESSURE =

DELTA P =

FEEDMATER ENTHALPY TEMPERATURE =

PRESSURE

=_ _

h5 = 1192.9 BTV/LBM hF = 419.5 BTU /LBM Oh(SG) = 773.4 BTU /LBM STEAM ENTHALPY

--- -- + a C p _

MOISTURE =

HOT LEG ENTHALPY TEMPERATURE =

PRESSURE =

hH = 640.2 BTU /LBN hC = 558.3 BTV/LBM Oh(VESS) = 81.9 BTU /LBM Cp(TH) = 1.548 BTU /LBM-DEGF COLD LEG ENTHALPY

- - +&,C TEMPERATURE =

PRESSURE -

Cp(TC) = 1.266 BTU /LBM-DEGF COLD LEG SPECIFIC VOLUME

., . +4,C TEMPERATURE =

PRESSURE =

0935D:10/021891 18.

l TABLE 3.1-4 CALOR! METRIC RCS FLOW HEASUREMENT UNCERTAINT!!S COMPONENT INSTRUMENT ERROR FLDH UNCERTAINTY

(% FLOW)

+8.C FEEDHATER FLOW ---

VENTURI THERMAL EXPANSION COEFFICIENT TEMPERATURE MATERIAL DENSITY TEMPERATURE PRESSURE DELTA P FEEDRATER ENTHALPY TEMPERATURE PRESSURE STEAM ENTHALPY PRESSURE MOISTURE NET PUMP HEAT ADDITION HOT LEG ENTHALPY TEMPERATURE STREAMING, RANDON STREAMING, SYSTEMATIC PRESSURE COLD LEG ENTHALPY TEMPERATURE PRESSURE COLD LEG SPECIFIC VOLUME TEMPERATURE PRESSURE a

09350:1D/021891 19 l

TABLE 3.1 4 (continued)

"* ~

CALOR! METRIC RCS FLOW HEASUREMENT UNCERTAINTIES BIAS VALUES +a,e FEEDRATER PRESSURE DENSITY ENTHALPY STEAM PRESSURE ENTHALPY PRESSURIZER PRESSURE ENTHALPY - HOT LEG ENTHALPY - COLD LEG SPECIFIC VOLUME - COLD LEG FLOH BIAS TOTAL VALUE

** +,++ INDICATE SETS OF DEPENDENT PARAMETERS N LOOP UNCERTAINTY (HITH BIAS VALUES) l 09350:10/021891 20

I i

l TABLE 3.1-5 s OVERTEMPERATURE DELTA-T TRIP DLLTA-T Tavg PRESS DELTA-I -

- +&,C SCA -

H&TE -

STE -

SD -

BIAS -

RCA -

H&TE -

RCSA -

RTE -

RD -

SA -

Of RTD USED TH - 2 TC - 1 INSTRUMENT SPAN - 88.9 DEGF

+4,C i SAFETY ANALYSIS LIMIT -[ ]

ALLOKABLE VALUE - 3.10% EELTA-T SPAN NOMINAL SETPOINTS K1 - 1.12 K3 - 0.001152 VESSEL DELTA-T - 59.3 DEGF DELTA-I GAIN - 1.97

+4,C PRESSURE GAIN -( )

~ +4,C +&,C ,,,,,,

+&,C Z - S - T =

TA CSA _ ,,,

MAR ,,,,, ,,,,,,

09350:10/021891 21

TABLE 3.1-6 OVERPOWER DELTA-T TRIP DELTA-T Tavg

+4.C PHA .

SCA .

SD -

BIAS .

RCA . .

M&TE =

RCSA .

RTE =

RD =

~

OF RTD USED TH = 2 TC = 1 INSTRUMENT SPAN . 88.9 DEGF

+4 C SAFETY ANALYSIS LIMIT =( )

ALLOHABLE VALUE = 1.90% DELTA-T SPAN i

NOMINAL SETPOINTS 1.08 VESSEL DELTA-T 59.3 DEGF 1

- +4,C . +4,C , _ 8,

+ C Z . S . T =

TA . CSA = -

MAR =

_.=

09350:10/021891 22

TABLE 3.1-7 Low RCS T&vg (Coincident with Reactor Trip, Turbine Trip and Feedwater Isolation

_. ,_. +a.c PKA -

SCA .

50 -

BIAS -

RCA =

M&TE .

RCSA -

RTE =

RD .

OF RTD USED TH - 2 TC - 1 INSTRUMENT SPAN - 100.0 DEGF

+R C LIMIT -( )

ALL0HABLE VALUE . 561.5 DEGF NOMINAL TRIP SETPOINT . 564.0 DEGF

- - +&,C +4,C ~ - +&,C Z - S - T .

TA = . CSA . MAR -

09350:10/021891 23

TABLE 3.1-8 COLD LEG ELB0H TAP FLUH UNCERTAINTY INSTRUMENT UNCERTAINTIES

+4,C

% DP SPAN % FLOW

~

PMA =

PEA =

SCA =

SPE .

STE =

SD .

RCA .

M&TE -

RTE =

RD -

ID =

A/D = 1 RDOT -

FLOW CALORIN. BIAS -

FLON CALORIMETRIC =

INSTRUMENT, SPAN =

- - +a c SINGLE LOOP.ELB0H TAP FLOH UNC = 1 FLOH N LOOP ELB0H TAP FLON UNC =

~ ~

H LOOP RCS FLON UNCERTAINTY +a.c (HITH BIAS VALUES) 09350:10/021891 24

TABLE 3.1-9

.> . LOW RCS FLOH REACTOR TRIP INSTRUMENT UNCERTAINTIES

+4,c

% OP SPAN % FLON SPAN 1

~ ~

PHAL =

PMA2 -

PEA =

SCA =

SPE =

STE =

SD =

BIASF.

BIASl=

BIAS 2

, RCA =

M&TE =

RCSA = ,

RTE =

RD =

BIAS =

FLOH SPAN = 120.0 % FLOW

- +4;C SAFETY ANALYSIS LIMIT =[ ]

ALL0HABLE VALUE - 89,4 % FLOH NOMINAL TRIP SETPOINT = 90.0% FLOW -

- +4,C - - +4,C ,- - +4,C Z = S = T =

TA = CSA = MAR =

1 09350:10/021891 25 l

TABLE 3.1-10 4e TECHNICAL SPECIFICATION TABLE 2.2-1 MODIFICATIONS Overtemperature oT Reactor Trip ,

Z - 7.04 S - 1.96 (AT) + 1.17 (pressure)

Allowable Value 1 3.1 % AT span Delete "by RTO Manifold Instrumentation" in NOTE 1.

Overpower AT Reactor Trip Z - 1.54 5 - 1.96 Allowable Value i 1.9 % AT span Delete "by RTD manifold instrumentation" in NOTE 3.

2 09350:10/021891 26

4.0 SAFETY EVALUATION g

The primary effect of the RTD bypass elimination on the FSAR Chapter 15 3 (Reference U oafety analyses are the differences in response time characteristics and instrumentation 9ncertainties associated with the fast response thermowell RTD system. The affects of these differences are discussed in the following sections.

4.1 LICENSING BASIS 4.1,1 Etigante Time The response time parameters of the VEGP RTD bypass system assumed in the safety analyses are shown in Table 2.1-1. For the fast response thermowell RTD system, the overall response time will consist of (

1."'C This makes the total RCS temperature measurement response time to be 6.0 seccads which is the same as that assumed for the RTD bypass system. However, the components of the 6.0 seconds have changed and this must be evaluated to confirm the effect of the modified response characteristics. This response time is factored into the Overtemperature AT trip and Overpower of trip performance. Section 4.2 includes a discussion of the evaluations performed for these events.

4.1.2 ElrLyncertainty The propraed fast response thermowell RTD system will make .use of RIDS, I

manufactured by Weed Instruments Inc., witn a total uncertainty of 9- ( ]C assumed for the analyses.

The FSAR analyses make explicit allowances for inst *umentation errors for the reactor protection system setpoints. In addition, allowr. aces are made for the average RCS temperature, pressure and power. These allowances are made explicitly to the initial conditions.

0935D:10/021891 27

The following protection and control system parameters were evaluated and determined to be unaffected by the change from one hot leg RTD to three hot l leg PTDs; the Overtemperature AT (OTDT), overpower AT (OPDT), and Low RCS Flow reactor trip functions, RCS loop T,yg measurements used for input to j the rod control system and safety injection, and the calculated value of the j RCS flow uncertainty. System uncertainty calculations were performed for j these parameters to determine the effect of the change in the number of hot

} leg RTOS. The results of these calculations, noted in 3.3, indicate i sufficient margin exists to account for known instrument uncertainties, j

4.2 EVALUATIONS 4.2.1 !!on_LQCA EvalVit120 The changes in the composition of the RTD response time discussed in Section 2.1 and the instrumentation uncertainties discussed in Section 3.3 have been considered for the VEGP non-LOCA safety analysis design basis. The VEGP units will replace the RTD bypass system with Weed RTDs that have a total response time of 6.0 seconds in the thermowell system (Table 2.1-1). The-RTDs (3 hot leg and 1 cold leg) in each loop are all dual element. The setpoint uncertainties are consistent with the uncertainties assumed in the analyses performed to support the VEGP VANTAGE 5 program (Reference 3).

The primary effect of the RTD bypass elimination on the non-LOCA licensing basis safety analyses is the response time associated with the thermowell RTD system. The total response time for the thermowell RTD system is 6.0 seconds.

RTD response times result in delays from the time when the fluid conditions in the RCS indicate,the need for an overtemperature AT (OTAT).or overpower AT (OPAT) reactor trip until a trip signal is actt,4lly generated; therefore, those transients that rely on the 'above mentioned trips must be analyzed with the appropriate response time incorporated. The affected transients are the following:

09350:10/021891 20

- _ - - _ _ - _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ - - _ _ _ - - _ = _ .

o Sceam Syste.s Piping failure (FSAR Section 15.1.5) o loss of Electrical Load (FSAR Section 15.2.2) o Turbine Trip (FSAR Section 15.2.3) o Uncontrolled Rod Clutter Control Assembly (RCCA) Dank Hithdrawal at Power (FSAR Section 15.4.21 o Chemical and Volume Control System /*VCS) Halfunction that Results in a Decrease in the Boron Concentraite en the Reactor Coolant (FSAR Section 15.4.6) o Steamline Break Hith Coincidental Rod Hithdrawal at Power (Reference 3 Section 15.4.9) o Inadvertent Opening of a, Pressurizer Safety or Relief Valvi (FSAR Section 15.6.1) o The VEGP specific steamline break superheat mass / energy cases (outside containment) reported in WCAP ll285 (Reference 2).

With the exceptions of the VEGP specific steamline break superheat mass / energy cases (outside containment) reported in NCAP-11285 and the Steam System Piping Failure (FSAR Section 15.1.5), the transient'. cffe :ted by the total response time were reanalyzed with the analytes to support tiA use of VANTAGE 5 fuel.

Th6 analysis of the VEGP specific steamline break supt rheat mass / energy cases (outside contain. tent) reported in NCAP-11285 supports i.he 6.0 second response time associated with the RTD bypass elimination for both OPAT and OTAT reactor trips. The at-power analysis for the Steam Systve Piping Failure (FSAR Section 15.1.5) summarized in WCAP-9226 (Holligsworth, S. D. and i

1 3

09350:10/021891 29 Y

Wood, D. C. Reactor Core Response to Excessive Secondary Steam Releases, January 1978) supports the 6.0 second response time associated with the RTD bypass elimination for the OPAT reactor trip.

Since a 6.0 second response time is incorporated in tiir mst r,: rent M.7-LOCA safety analyses that model the OTAT and and OPAT reactor trip setpi.:ati and these analyses calculated acceptable results, the removal of the RTD bypass system in favor of the thermowell RTD system that has a total response t he of 6.0 seconds is acceptable.

4.2.2 LOCA and LOCA-Related Evaluation The LOCA analyses must be reviewed in order to determine that the uncertainties in various parameters for RTO bypass elimir,ation are bounded by LOCA analyses. The FSAR as curr itly updated for VEGP contair.s vnly the 17x17 LOPAR fuel analyses results. VANTAGE 5 fuel analysis results whi;h include documentation of the effects of RTD bypass elimination on the LOCA-related analyses are included in Reference 3. It should be noted, therefore, that I

this evaluation assumes that RTD bypass. elimination c' curs no earlier than the first reload utilizing VANTAGE 5 fuel. The followli.g presents summaries of -

evaluations performd to assess-the effect of revised parameter uncertainties for VEGP for the LOCA-related analyses performed by Westinghouse.

4.2.2.1 Large Break LOCA (FSAR Section 15.6.5)

The VANTAGE 5 fuel large break ~ LOCA analyses for VEGP contained in Reference 3 was performed using the NRC-approved 1981 ECCS Evaletion Model with BASH and resulted in a Peak Clad Temperature (PCT) of 2108'F (including the 50'F transition core penaltyN for a double-ended cold leg guillotine (DECLG) break with a discharge coeffMh nt (CD ) of 0.6. This value for the'li ge break LOCA PCT includes the effects of RTO bypass elimination and . .1o additional effects need be accounted for. -

09350:10/021891 30 l

4.2.2.2 Small Break LOCA (FSAR Section 15.6.5)

The VANTAGE 5 fuel small break LOCA analysis for VEGP performed in Reference 3 was performed using the NRC-approved Small Break LOCA ECCS Evaluation Model with NOTRUMP which resulted in the most limiting PCT of 2056"F for a 3 inch equivalent diameter break. This value includes the effects of 9TO bypass elimination and so no additional penalties need be applied.

4.2.2.3 Blowdown Reactor Vessel and Loop Forces (FSAR Section 3.9)

The blowdown hydraulic loads resulting "com a loss of cooiant accident are considered in Section 3.9 of tne VEGP FSAR. The maximum loads are generated within the first few seconds after break initiation, ind depend mainly on T

cold. The VANTAGE 5 analysis utilized the uncertainties appropriate for RTO bypass elimination and so that analysis remains valid.

4.2.2.4 Post-LOCA Long Term Core Cooling Subtriticality Requirement; Westinghouse Licensing Fosition (FSAR Section 15.6.5)

The typical Westinghouse licensing position for sctisfying the requirements of 10CFR Part 50 Section 50.46 Paragraph (b) Item (5) "Long Term Cooling" is defined in WCAP-8339 (Bordelon, F. M., et. al. " Westinghouse ECCS Evaluation Model - Summary" (flon-Proprietary), June 1974). The Westinghouse commitment is that the reactor will remain shutdown by borated ECCS water residing in the sump following a LOCA. Since credit for the control rods is typically not taken for large break LOCA, the borated ECCS water provided by the accumulators and the RWST must have a concentration that, when mixed with other sources of-borated and non borated water, will result in the reactor core remaining rubcritical assuming all control rods are out. The revised uncertainties to account for RTD bypass elimination will have no effect on those volumes and boron concentrations assumed for this calculation. An assessment of taking credit for centrol rods for post-LCCA subcriticality is provided in Section 5.2.5 ef Enclosure 4 in Reference 3. Therefore, operation of VEGP with these revised values doe; not result in a violation of this requirement.

09350:10/021891 31

4.2.2.5 Hot leg Switchover to Prevent Potential Boron Precipitation (FSAR Sections 6.3.2.8 and 6.3.3.1)

Post-LOCA hot leg recirculation switchover time is determined for inclusion in ,

emergency procedures to ensure no boron precipitation in the reactor vessel following boiling in the core. This time is dependent on power level, and the RCS, RHST and accumulator water volumes and boron concentrations. The revised uncertainties will have no effect on the volumes assumed for the RCS, RHST and accumulators, and will have no effect on the boron concentrations. Therefore, there was no effect on the post-LOCA hot leg switchover time for VEGP.

4.2.2.6 Post-LOCA Long Term Core Cooling Minimum Flow Requirement (FSAR Section 15.6.5)

In accordance with the requirements of 10 CFR 50.46 for long term core cooling, a calculation is performed to determine the minimum safety injection flow rate required for hot leg injection mode. This flow rate then becomes the basis for the minimum technical specification flow rate for hot leg injection mode. None of the revised values considered herein affect this flow rate and so this r;quirement is not violated.

4.2.2.7 LOCA and LOCA-Related Conclusion The effect of revised uncertainties due to RTD bypass elimination for VEGP has been evaluated. The potential effect of the revised parameters on the VANTAGE 5 fuel analysis results for each of the LOCA-related accidents was considered and it was shown in all cases that the effect of the revised parameters had no adverse effect. Therefore, it can be concluded that the a revised uncertainties for VEGP are acceptable for the LOCA analyses discussed in this safety evaluation.

4.2.3 Instrumentation and Control (ISC) Evaluation The RTD bypass elimination nadificatior. for VEGP does not functionally change the AT/T avg protection channals. The implementation of the fast response 09350:10/021891 32

RTDs n the reactor coolant piping will change the inputs into the AT/T,yg Protection Sets I, II, III and IV as follows: ,

1

1. The Narrow Range (NR) cold leg RTD (used in the protection system) in the cold leg manifold will be replaced with a fast response NR dual element well mounted RTD in the RCP pump discharge pipe. The signal from this fast response NR RTD will perform the same function as the existing RTD T

cold signal. One element of the RTD will be held in reserve as a spare.

2. The NR hot leg RTD in the bypass manifold will be replaced with 3 fast response NR dual element well mounted RTDs in ecch hot-leg that are electronically averaged in the process protection system.

3. Identification of failed signals will be by the same means as before the i modifications, i.e., existing control board alarms and indications.

4. Signal process and-the added circuitry to the protection set racks will_be accomplished by additions -to the process control (Hestinghouse Model- 7300) racks using 7300 technology. When one T hot signal is removed from the averaging process, the electronics will allow a bias to be manually added to a 2-RTD average Thot (as opposed to a 3-RTD average Thot) in order to obtain a value comparable with the 3-RTD average Thot prior to the failed RTD. In the event of a cold leg RTD failure, the spare cold leg RTD element will be manually connected to the 7300 circuitry !.n place of.

the failed RTD.

Existing control board AT and Tavg indicators and alarms will provide the means of identifying RTD failures. Upon identification of a failed hot leg RTD,- the operator would place that protection channel in trip (consistent with the time requirements specified in the Technical Specifications), identify and disconnect the failed RTD, and rescale-the summing amplifier for a two RTD input condition. Specifically, if one T

hot signal is removed from the averaging process, the electronics will allow a bias to be manually added to a 2-RTD average Thot (as pp sed to a 3-RTD average Thot) in order to obtain a value comparable with the 0935D:10/021891 33

3-RTD average Thot prior to the failed RTD. An alternative procedure would be to utilize the spare hot leg element within the dual element RTD by manually connecting the spare element to the 7300 circuitry in place of the failed element. In the event of a cold leg RTD failure, the spare cold leg RTD element will be manually connected to the 7300 circuitry in place of the failed RTD. After this process, the channel would then be returned to service. As noted, during this rescale process the plant will be in a partial trip mode and will therefore be in a safe condition. For more details on this procedure see Appendix A.

Other than the above changes, the Reactor Protection System will remain the same as that previously utilized. For example, two out of four voting logic continues to be utilized for protection functions, with the model 7300 process control bistables continuing to operate on a "de-energize to actuate" principle. Non-safety related control signals will continue to be derived from isolated protection channels.

The above principles of the modification have been reviewed to evaluate.

conformance to the-requirements of IEEE Standard 279-1971 criteria and associated 10CFR 50 General Design Criteria (GDC) 1, 2, 4 and 23, Regulatory Guide 1.89, and other applicable industry standards. IEEE Standard 27 '971 requires documentation of a design basis. Following is a discussion v design basis requirements in conformance to pertinent I&C criteria:

a. The single failure criterion continues to be satisfied by this change because the independence of redundant protection sets is maintained,

b. The quality of the components and modules being added is consistent with use in a nuclear generating station protection system._ For the Westinghouse Quality Assurance program, refer to Chapter 17 of the FSAR.-

c. The changes will continue to maintain the capability of the protection system to initiate a reactor trip during and follovf ng natural phenomena credible to the plant site-to the same extent as the existing system.

09350:10/021891 34

d. Channel independence and electrical separation is maintained because the protection set circuit assignments continue to be Loop 1 circuits input to Protection Set I; Loop 2 to Protection Set II; Loop 3 to Protection Set III; and loop 4 to Protection Set IV, with appropriate observance of field wiring interface criteria to assure the independence. Output circuits are the same as before except that there will be one T cold and three T hot outputs to the computer sent through Class lE isolators in each protection set,

e. Equipment seismic and environmental qualification will be to IEEE standards 344-1975 and 323-1974, respectively, as described in NCAP 8587 Rev. 6-A "Hethodology for Qualifying Westinghouse HR0 Supplied NSSS Safety Related Electrical Equipront".

f. The compliance of the hardware to IEEE 279-1971 Section 4.7 and GDC requirements concerning Control and Protection interaction has not been changed.

On the basis of the foregoing evaluation, it is concluded that the compliance of VEGP to IEEE Standard 279-1971, applicable GDCs, and industry standards and regulatory guides has not been changed with the IIC modifications required for RTD bypass removal.

4.2.4 Mechanical Evaluation The presently installed RTD bypass manifold system is to be replaced with fast acting narrow range dual element thernovell mounted RTDs installed in the reactor coolant loop piping. This change requires modifications to the hot leg scoops, the crossover leg bypass return nozzle, and the cold leg RTD bypass nozzle.

All machining operations performed during modification of the hot and cold leg penetrations, as well as the machining and capping of the crossover leg bypass return line, will be performed in a manner that minimizes debris entering the reactor coolant system.

09350:10/021891 35

{

__ --- - - 4

Hot Leg The original hot leg RTD bypass piping which provides RCS flow to the bypass manifold must be removed, and the scoops modified to accept fast response RTD thermowells. A hole will be machined through the tip of each scoop which will provide the proper flow path. The field machined surfaces will be examined prior to welding as required by ASME Code Section XI. The install 4 tion described above will be performed using Gas Tungsten Arc Held (GTAW) for the root pass, and finished out with either GTAH or Shielded Metal Arc Weld (SMAW). All of the welds will be examined by Penetrant Test (PT) per ASME Code,Section XI.

Cold leg The cold leg RTD bypass piping must be removed, and the cold leg RTD bypass nozzle modified to accept the fast response RTD thermowell. The RTD thermowell will be installed into the nozzle and will extend approximately 3.5 inches into the flow stream. The thermowell will be fabricated in accordance with Section III (class 1) of the ASME Code. The machined turfaces of the nozzle to be welded will be examined prior to welding as required by ASME Code,Section XI. The root weld joining the RTD thermowells to the modified nozzles will utilize GTAN for the root pass, and will be finished out with either GTAN or SMAH. The welds will be examined by Penetrant Test (PT) per ASME Code,Section XI.

Cross-over Leg The cross-over bypass return piping must be removed and the nozzle capped.

The cap will be fabricated to meet the pressure boundary criteria of the ASME Code,Section III (class 1). The machined surfaces will be examined prior to welding as required by the ASME Code,Section XI. The cap will be root welded to the nozzle by GTAM and fill welded by either GTAH or SMAN.. The welds will be examined by Penetrant Test (PT) and Radiography Test (RT) per ASME Code,Section XI.

0935D:10/021891 36

Pressure Testing A system hydrostatic test will be performed af ter completion of the bypass elimination modification. This test will be performed in accordance with ASME Code,Section XI, which requires that a hydrostatic test of new pressure boundary welds be performed when the connection to the pressure boundary is larger than one (1) inch in diameter.

In summary, the integrity of the reactor coolant piping as a pressure boundary component, is maintained by adhering to the applicable ASME Code suctions and Nuclear Regulatory Commission General Design Criteria. Further, the pressure retaining capability and fracture prevention characteristics of the piping is not compromised by these modifications.

4.2.5 Steam Generator Tube Ruoture (SGTR) Evaluation The SGTR analysis in the VEGP FSAR was perform 3d using the LOFTTR2 program and includes the modeling of the operator actions required for SGTR recovery. The FSAR SGTR analysis included an analysis to demonstrate margin to steam generator overfill and an analysis to demonstrate that the offsite radiological consequences are within the allowable guideline values in 10CFR100 and Standard Review Plan 15.6.3. A SGTR reanalysis was performed to support the change to VANTAGE 5 fuel and the results of the reanalysis are presented in the VANTAGE 5 submittal (Reference 3). The results of the VANTAGE 5 reanalysis indicated that margin to steam generator overfill will be maintained and that the offsite radiation doses are well within the allowable guideline values for a design basis SGTR. The changes associated with the RTO bypass elimination were incorporated into the SGTR reanalysis, and thus the results in the VANTAGE 5 program are also applicable for operation with RTD bypass elimination. On this basis, it is concluded that the SGTR analysis results will be acceptable for operaHon of VEGP with RTD bypass loops eliminated.

09350:10/021891 37

4.2.6 Containment Intearity Evaluation RTD bypas's elimination does not adversely affect the short and long term LOCA mass and energy releases and/or the containment response analyses following a LOCA H&E release or main steamline break mass and energy release containment analysis. The condition does not affect the normal plant operating parameters, system actuations, accident mitigating capabilities or assumptions important to the containment analyses, or create conditions more limiting than those assumed in these analyses. Therefore, the conclusions presented in the FSAR remain valid with respect to the containment.

4.2.7 Dose Evaluation Based on the fact that no additional mass releases are predicted due to the bypass system elimination, in addition to the fact that fuel integrity and mitigating equipment integrity is maintained, there are no additional radiological consequences associated with this modification.

4.2.8 Technical Soecification Evaluation As a result of the calculations summarized in Section 3.0, two protection functions' Technical Specifications must be modified. The affected functions and their associated trip setpoint information, are noted on Table 3.1-10.

The flexibility to account for elimination of the bypass loops has been previously factored into the analyses in support of the transition to VANTAGE 5 fuel (Reference 3). Therefore, the two identified protection functions have been incorporated into the VANTAGE 5 technical specification changes and bound the plant configuration with the bypass loops removed.

09350:10/021891 38

.)

5.0 DETERMINATION OF NO UNREVIEWED SAFETY OUESTION Based upon the evaluations presented in Section 4.0, the criteria for determination of an unreviewed safety question can specifically be addressed.

5.1 The removal of the bypass system and installation of the fast response RTDs will not increase the probability of a previously analyzed l accident. The integrity of the reactor coolant pressure boundary is i maintained by design and installation procedures adhering to appropriate codes and standards. In addition, an instrumentation and control evaluation has concluded that the fast response RTO system remains in compliance with industry standards and criteria for singla failure, independence, separation and qualification considerations. No new accident initlators are created by this modification. Therefore, the l

modification has no effect on the probability of previously analyzed ,

accidents involving the integrity of the reactor coolant pressure boundary or performance of the control and protection system.

5.2 This modification does not increase the radiological consequences.of any previously evaluated accident. Although the pressure boundary will be modified, power welding techniques and tests will ensure the integrity of the pressure boundary and thus not contribute to any additional radiological consequences. Protection systems.will continue to operate as assumed in the safety analyses so that no additional challenges to fuel integrity or mass / energy release calculations will result.

5.3 This modification does not create the possibility of an accident which is different than any already evaluated in the FSAR. The modification creates no riew accident initiators.and no new single failures have been identified. The installation operations minimize the. potential-for debris escaping into the RCS and the small amount of debris introduced has been determined to be inconsequential. Additionally, the reliability and performance of the protection system remains consistent with that assumed in the safety analyses. Therefore, no new accidents have been created.

1 09350:10/021891 39 ,

l

. - - - . - . - , . . - - - . . - A

5.4 This modification does not increase the probability of a malfunction of equipment important to safety previously evaluated in the FASR. The removal of the RTD bypass system eliminates the components which have been a contributor to plant outages as well as occupational radiation exposure. The installation of fast response thermowell mounted RTDs in the reactor coolant loop piping will continue to provide the required loop temperature signals for input to the reactor control and protection systems. The removal of the RTD bypass piping and the installation of a modified temperature measurement system (fast response RTDs mounted in thermowells) does not affect the integrity of the reactor coolant system. This is due to the reactor coolant piping (pressure boundary component) modifications adhering to the ASME Code (Sections III (Class

1) and XI] and to the Nuclear Regulatory Commission General Design Criteria.

5.5 The modification will not increase the consequences of a nalfunction of equipment important to safety previously evaluated in the FSAR. The arrangement of the RTD measurement system is not assumed nor credited in the mitigation of the radiological consequences of any accident. Since this installation will not lead to any additional- electrical or mechanical equipment malfunctions, the consequences of a malfunction of equipment is not affected.

5.6 The possibility of a malfunction of equipment important to safety different than any already evaluated in the FSAR is not created'as-a result of this modification. The installation will create a malfunction different from those previously evaluated. Reactor coolant pressure boundary integrity will be maintained. Reactor coolant loop temperature inputs for control and protection functions will continue to be supplied as assumed in the safety analyses in accordance with appropriate criteria for single failure, independence, separation and qualification considerations. Other equipment important to safety will- be unaffected-and will continue to function as designed.

0935D:1D/021891 40 l

5.7 The margin of safety defined in the bases to any technical specification will not be reduced as a result of this modification. The resultant eff(cts on the Overtemperature delta-T and Overpower delta-T reactor trip setpoints defined in Technical Specification Table 2.2-1 have been accounted for in the safety analyses. It has been confirmed that the changes identified for the VANTAGE 5 fuel program (Reference 3) remain bounding for elimination of the bypass system and replacing it with the fast respon=e RTDs. Additionally, the integrity of the reactor coolant boundary is maintained. Therefore, no safety margin is degraded.

Based upon this assessment it can be concluded that the modifications to the plant configuration to eliminate the RTD bypass system and replace it with fast response RTDs do not involve an unreviewed safety question with respect to mechanical integrity, setpoint determination or safety analysis results.

~

i

{

v 0935D:10/021891 41

6.0 CONCLUSION

S The method of utilizing fast-response RTDs installed in the reactor coolant loop piping as a means for RCS temperature-indication has undergone tytensive analyses, evaluation and testing as described in this report. The incorporation of this system into the VEGP design meets all safety, licensing and control requirements necessary for safe operation of this unit. The analytical evaluation has been supplemented with in-plant and laboratory testing to further verify system performance. The fast response RTDs installed in the reactor coolant loop piping adequately replace the present hot and cold leg temperature measurement system and enhances ALARA efforts as well as improves plant reliability.

t 09350:10/021891 42 i

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

7 ,- , , - . . _ . . . , . . .. . _ _ . , .., ... m .- ..._. ~ . .,_ _.,_.,._,m...-.,~,._......_,,-,w~- m...

7.0 CONTROL SYSTEM EVALUATION A prime fnput to the various NSSS control systems is the RCS average temperature, T(avg). This is calculated electronically as the average of the measured hot and cold leg temperatures in each loop.

The effect of the new RTD temperature measurement system is to potentially change the time response of the T(avg) channels in the various loops. This in turn could affect the response of (

]"'C As previously noted, the naw dTO system (RTO + thermowell) will have a time response the same as that of the current system (RTO + bypass line). Therefore, there will be no effect on the T(avg) channel response and no need, as a result of implementing the new system, to revise any of the control system setpoints. However, VEGP always has the option of making setpoint adjustments. If desired, system performance can be verified by performing a series of plant tests (e.g., step load changes, load rejections,.

etc.) following installation of the new RTD system. Control system setpoints can then be adjusted based on the results of the tests. It should be recognized that control systems do not perform any protective function in the FSAR accident analysis. Hith respect to accident analyses, control systems.

are assumed to operate only in cases in which their action aggravates the consequences of an event, and/or as required to establish initial plant conditions for an analysis. The modeling of control systems for accident analyses is based on nominal system parameters as presented in the Precautions, limitations, and Setpoint document.

09350:1D/021891 43

i

8.0 REFERENCES

1. Vogt'T'e Electric Generating Plant Final Safety Analysis Report Update, March 1990.

2. WCAP-ll285, "MSLB Information Used for Superheat Study for Vogtle units 1 and 2", September 1986.-

3. Georgia Power Submittal to the NRC, ELV-02166, November 1990, Request for

Technical Specification Changes - VANTAGE 5 Fuel Design.

09350:10/021891 44 3.., .. .. .

d APPENDIX A DEFINITION OF AN OPERABLE CHANNEL AND HOT LEG RTD FAILURE COMPENSATION PROCEDURE -

0935D:10/021891 45 j

\

r RTD BYPASS ELIMINATION FOR V0GTLE ELECTRIC GENERATING PLANT (VEGP)

OEFINITION OF AN OPERABLE CHANNEL AND HOT LEG RTD FAILURE COMPENSATION PROCEDURE Hestinghouse Electric Corporation Pittsburgh, PA 09350:10/021891 46

DEFINITION OF AN OPERABLE CHANNEL The RTD bypass elimination modification uses the average of 3 RTDs in each hot leg to provide a representative temperature measurement. In the event one or more of the RTDs fails, steps must be taken ',o compensate for the loss of that RTD's input to the averaging function. VEGP will have dual element RTDs installed in each hot leg thermowell location. The second element may be used when the first element fails and the three RTD average maintained. In the event of the second element falling in the same RTD, then this procedure could be envoked.

Sinale RTD Failure Hot Leg: All three hot leg RTDs must be operable during the period following refueling from cold to hot zero power and from hot zero power to full power.

During the heat up period the plant operators will be [ -

]"'C Typically this data is recorded at initial 100% power and, thereafter, during the normal protection channel surveillance interval.

Once [ l a.c -any hot leg can then tolerate failure of both elements of a single dual element RTD and still remain operable. If the situation arises where such a failure occurs a bias value must be applied to the average of the remaining two valid RTDs. [

.j a.c 09350:10/021891 47

_ _ __ _ _ _ _ -_ - _ A

The plant may operate with a failed hot leg RTD at any power level during that samefuejcycle. It is permissible to shutdown and startup during the cycle without requiring that the failed RTD be replaced. [

3a ,c In order to eliminate any control system concerns, the Tavg and AT signal

( associated with the loop containing the failed hot leg RTD will be defeated as l an input to the control system. This will prevent the control system from using a Tavg or AT at power levels less than 100% which may be offset due to i the fixed biased. If another hot leg RTD fails in a different loop the utility should operate using manual control. Manual control is recommended because only one control channel at a time can be defeated. If-automatic operation is continued the control system may choose the biased channel due to the positive (or zero) bias application. This means the control system will perceive a higher Tavg than actually exists at reduced power and the plant -

will operate at reduced temperatures. While this is not necessarily undesirable it does reduce the total plant megawatt output. The use of automatic control can be considered based on utility power requirements.

Cold Leg: If the active cold leg RTD fails, then that RTD should be disconnected from the 7300 cabinets. The installed spare RTD should then be connected in the failed RTO's place.

Double RTD Failure: Inocerable Channel Hot Leg or Cold Leg: If two or more of the three hot leg RTDs or both cold i leg RTD elements fail in the same protection channel then that channel is considered inoperable and should be placed in trip. Operation with only one valid hot leg RTD is not presently analyzed as part of the licelising basis.

0935D:10/021891 48

m PROCEDURE FOR OPERATION WITH A HOT LEG RTD OUT OF SERVICE The hot leg temperature measurement is obtained by averaging the measurements from the three thermowell RTDs installed on the hot leg of each loop. (

j a .C In the event that one of the three RTDs fails, the failed RTD will be disconnected and the hot leg temperature measurement will be obtained by averaging the remaining two RTD measurements. (,

jc a

The bias adjustment corrects for [

s a

Jc To assure that the measured hot leg t

temperature is maintained at or above the truo hot leg temperature, and thereby avoid a reduction in safety margin at reduced power, [

.j a.C 09350:10/021891 49

_ _ _ _ _ _ _ - - _ _ _ _ - - - - _ - - - - _ _ _ _ J

An RTD failure will most likely result' in an offscale high or low indication and will,,be detected through the normal means in use today (i.e., T and AVG AT deviation alarms). Although unlikely, ' RTD (or its electronics channel) can fail gradually, causing a grad 1 change in the loop temperature measurements. [

3a,C The detailed procedure for correctir.g for a failed hot leg RTD is presented below:

- a,c 09350:1D/021891 50

, a,C 4

e 51

a.C 09350:10/021891 52

APPENDIX CALCULATION OF HOT LEG TEMPERATURE BIAS a,c s

09350:10/021891 53

f. .

APPENDIX B ACRONYMS FOR UNCERTAINTY CALCULATIONS l

l l

0935D:10/021891 54

a,C e

t 4

e e

a h

I 55 1

m - + oeo &e- a,m. .--,,s 4A 4.- .s 4 cs. , e,- a 1 e a

a,C s

(

e 4

56

- ~""*-e-=,- .,, , _ _ , _ " '" gw--w- ,

A w eer.u.E-cn_aw.4._ma. m- _d"m-. --%m -me pn - _ mm - _ w...- +, .,_q .. ___.1. n_4dw,e,nm, _ , 4 .A a ,,t., ~<J ,4_3, 8 i B,C l

i f

4

+

4 9

6 57

O $

8,C 1

t

(

b 58

. .*- 8,C t

'N 4

59 J________ ______ - - - - - . - - _ . - - - . - _ _ . - - - -

4 6

% 6,C a

f 4

e 60

4 0

4) aiC k

i e

61

8 e a,C i

6 4

m c 62

s. .a - w - ._ .,a w ENCLOSURE 4 V0GTLE ELECTRIC GENERATING PLANT AMENDMENT TO VANTAGE 5 TECHNICAL SPECIFICATIONS CHANGES RTD BYPASS MANIFOLDS ELIMINATION PROPRIETARY INFORMATION E4-1

i .

4 Proprietary Information Notice Transmitted herewith are proprietary and/or non proprietary versions of ,

documents furnished to the NRC in connection with requests for generic and/or plant-specific review'and approval.

in order to conform to the requirements of 10 CFR 2.790 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non proprietary versions, only_the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated-in both versions by means of lower case letters (a) through (g) contained within parentheses located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(g) of the affidavit accompanying this transmittal pursuant to 10 CFR 2.790(b)(1).

t t

s 4

0055J sMD/032091

_ _ . . . _ _ . _ _ _ _ . _ , _ _ _ _ _ . _ . , _ . _ _ . - . __ ,.- _ _._ _ ___._,u.- . _ __. , . - . , - -

I' l

Copyright Notice

The reports transmitted herewith each bear a Westinghouse copyright notice.

The NRC is permitted to make the number of copies of the information contained !

in these reports which are necessary for its internal use in connection with generic and plant specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.790 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection not withstanding. With respect to the non proprietary varsions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. The NRC is not authorized to make copies for the personal use of members of the public who make use of the NRC public document rooms. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.

9

-__m--- , , ~ . . ._m,..-. .. ....w...,, --.,, ,__,_,_,,,--.--,,.m.,..m..., , , , ...y,. . 3 - . . , . . - _ -y .. g .y v,, ,-,

4 4

y.

1 l

l Westinghouse Electric Corporation Energy Systems

$' 3bem om March 20, 1991 CAW 91-140 Document Control Desk US Nuclear Regulatory Commission 1

Washington, DC 20555 Attention: Dr. Thomas Murley, Director j APPLICATION FOR WITHHOLDING PROPRIETARY l JR[pRMATION FROM PUBLIC DISCLOSURE

Subject:

WCAP-12788, Rev. 1 "RTD Bypass Elimination Licensing Report for Vogtle Electric Generating Plant"

Dear Dr. Murley:

The proprietary information for which withholding is being requested in the above referenced letter is- further identified in Affidavit CAW 91-140 signed by the owner of the proprietary information,- Westinghouse Electric Corporation.

The affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in caragraph:(b)(4) of 10 CFR Section 2.790 of the Commission's regulations. ..

t Accordingly, this letter authorizes the utilization of the accompanying Affidavit by Georgia Power Company.

Correspondence with respect to the proprietary aspects of the application for

withholding _or_ the Westinghouse affidavit :hould reference this--letter, CAW-91 140, and should be addressed to the undersigned.

Very truly yours,

. P. DiPiazza, Manag Enclosures Operating Plant Licens ng Support cc: M. P. Siemien, Esq.

Office of the General Counsel, NRC ,

00$$J sse/032091 w - ,- ,e =- -w +- y*-wy-*'mwy'a- T*P f -g w eey-- m

l CAW 91 140 AFFIDAVIT l COMMONWEALTH OF PENNSYLVANIA:

l ss COUNTY OF ALLEGHENY:

i Before me, the undersigned authority, personally appeared l Ronald P. DiPlazza, who, being by me duly sworn according to law, j s deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Corporation (" Westinghouse") and that l the averments of fact set forth in this Affidavit are true-and correct !

to the best of his knowledge, information, and belief:

W Ronald P. DiPiazza, Manah 1 Operating Plant Licensing. Support. I Sworn to and subs;ribed . i before me this f2 # day !

of W m b , 1991.

OfSWAf.Y& Notary Public iCTAA;AL $!M. .

i LoAAA!NE M PIPLCA,tioTARY PUFJc MO*4RoEY:LLE ECAo. ALLEh!NYCoVNTY ,

MY cow l$t'oN ExPAES CEC.14,1991 l Mew Pennsylvau h4%n of Nvet,s

- _ _ . _ _ _ . _ _ = , . . -,_ _ , , . . _ . . . , . _

I

L ,

! 2 CAW 91-140 (1) I am Manager, Operating Plant Licensing Support. in the Nuclear and Advanced Technology Division, of the Westinghouse Electric Corporation and as such, I have been specifically delegated the function of reviewing the proprietary informati:n sought to be withheld from public disclosure in connection with nuclear power plant licensing and rulemaking proceedings, and am authorized to apply for its withholding on behalf of the Westinghouse Energy Systems Business Unit.

(2) I am making this Affidavit in conformance with the provisions of 10CFR Section 2.790 of the Commissio:i's reggiations and in conjunction with the Westinghouse application for withholding accompanying this Affidavit.

(3) I have personal knowledge of the criteria and procedures utilized by the Westinghouse Energy Systems Business Unit in designating information as a trade secret, privileged or as confidential commercial or financial information.

1 (4) Pursuant to the provisions of paragraph (b)(4) of Section 2.790 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.

i

_.._..._,..-.__.___-.._.,__.a_ . _ _ _ . _ _ . - _ . , _ - _ _ _ . . . . . _ _

9

CAW 91 140 1

(ii) The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public, Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that systc.a and the substance of that system constitutes Westinghouse policy and provides the rational basis required.

Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse #s competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or component, structure, tool. method, etc.), the appilcation of which data secures a competitive economic advantage, e.g., by optimization or improved. marketability.

i 1

4 1

f 4 CAW-91 140 I,

d (c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its P customers or suppliers.

(e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f) It contains patentable ideas, for which patent protection may be desirable.

(g) It is nct tne property of Westinghouse, but must be treated as proprietary by Westinghouse according to agreements with the owner.

There are sound policy reasons behind the Westinghouse. system which >

include the following:

(a) The use of such information by Westinghouse gives' Westinghouse a competitive advantage over its competitors. It is, therefnre,

withheld from disclosure to protect the Westinghouse competitive

, position.

?

6 l

I -

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

- 5- CAW 91 140 (b) It is information which is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving ,

the use of the-information.

(c) Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.

(d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable at

( the total competitive advantage. If competitors acquire

' components of proprietary information, any one component may be the key to the entiro puzzle, thereby depriving Westinghouse of a competitive advantage.

(e) Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.

L (f) The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage. ,

i

'l 9

4-, -,cw----,vrr- w +--ev+--wer- +,-*e" - - - - ,~-,-v, * - -- -,r- ,--

i 6- CAW 91 140 ,

(iii) The information is being transmitted to the Commission in confidence and, under the provisions of 10CFR Section 2.790, it is to be received in confidence by the Commission.

(iv) The information sought to be protected is not availablo in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.

1 (v) The proprietary information sought to be withheld in this submittal is that which is appropriately marked in RTD Bypass Elimination Licensing Report for Vogtle Electric Generating Plant", WCAP 12788, Revision 1, (Proprietary) for Vogtle Units 1 and 2, being transmitted by the Georgia Power Company (GPC) letter and Application for Withholding Proprietary Information from Public Disclosure, C. K. McCoy, GPC, to Document Control Desk, to the Attention Dr. Thomas Murley, March,1991. The proprietary information as submitted for use by Georgia Power Company for the Vogtle Units 1 and 2 is expected to be applicabic in other-licensee submittals in response to certain NRC requirements for justification of actions to remove the existing Resistance Temperature Detector (RTD) bypass loop _ temperature measurement system and replace'it with fast response lthermowell mounted RTDs in the reactor ecolant loop piping.

l l

l j .

)

i 7 CAW 91 140 Thi: information is part of that which will enable Westinghouse tot

_(a) Provide documentation of the analyses, methods, and testing for reaching a conclusion in support of removing the existing Resistance Temperature Detector (RTD) bypass loop '

temperature measurement system and replacing it with fast response thermowell mounted RTDs.

(b) Support the continued validity of safety analysis initial condition assumptions. ,

(c) Establish the effects of the fast response thermowell RTD :

system on instrumentation and reactor coolant temperature measurement uncertainties.

(d) Assist the customer to obtain NRC approval for operation with RTD bypass loop elimination.

further this information has substantial commercial value as follows:

(a) Westinghouse plans to sell the asn of stallarzinformation to its customers for purposes of satisfying NRC requirements for licensing documentation.

4

_.A

3 i

CAW 91 140 (b) Westinghouse can sell support and defense of the RTO Bypass Elimination technology to its customers in the licensing process.

Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Vestinghouse because it would enhance the ability of competitcrs to provide similar analytical documentation and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the inforniation would enable others to use the information to meet NRC requirements for-licensing documentation without purchasing the right to use the information. ;

The development of the technology described in part-by the information is the result of applying the results of many years I

of cxperience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.

In order for competitors of Westinghouse to duplicate this information, similar technical- programs would have to be ,

performed and a significant manpower effort, having the requisite talent and experience, would have to be expended for developing testing and analytical methods and performing tests.

Further the deponent sayeth not.

, , . , . . . . , .r, -

, , . . - ,-r:-r,-- -c .. -. ,ewr...,n , m-,. %..

f ENCLOSVRE 5 V0GTLE ELECTRIC GENERATING PLANT AMENDMENT TO VANTAGE 5 TECHNICAL SPECIFICATIONS CHANGES RTD BYPASS MANIFOLDS EllMINATION ENVIROL :TAL EVALUATION 4

manifolds have been shown to be a The significant resistance source t6mperature of occupationaldetector radiat (RTD) ion exposure as well as a contributor to plant outages. The radiation exposure affects personnel who work on the RTD-bypass manifolds as well as those working on other systems such as reactor coolant pumps and steam generators locued it. the vicinity of the manifolds.

Data recorded on other plants indicates that the RTD bypass manifolds contribute 80 to 90 raan-rem / outage. Georgia Power Company currently expects to conduct approximately 47 refueling outages on Vogtle Electric Generating Plant (VEGP)

Units 1 and 2. This expected dose savings should be compared to the expected exposure for removal of the RTD bypass manifold piping.

Similar modifications have been performed on other >1 ants. A review of the radiation exposure recorded during the removal of tie bypass manifolds at seven other four-loop plants shows that total exposure ranged from a low of 74 rem to a high of 178 rem. The average was about 128 rem.

On the basis of the experience recorded above, Georgia Power Company. expects to achieve significant savings in personnel exposure over the operating life of the two VEGP units.

Actual exposure during the removal of the RTD bypass manifolds will be maintained as low as reasonable achievable (ALARA). Preplanning in order to achieve ALARA radiation exposure will include classroom training, hands-on training, and mockup training with tools that will be used to perform the work.

Specialized tools will be tested and qualified on mockups prior to use.

Temporary shielding and decontamination will be used to reduce exposure.

The tasks involved with removing the RID manifolds and installing the new RTD thermowells will be divided into discrete parts and conducted by teams that are qualified to perform each task, thus reducing the amount of time spent in radiation areas. The main activities expected to result in exposure to radiation include erection of scaffolding, interference and piping removal, shielding installation, installation of the RTD's, installation of cable, and _

. weld inspection. ,

The primary waste to be generated by the removal of the manifolds will be in the form of approximately 160 ft3 of piping weighing about 3000 lbs for each loop. '

Current plans are to ship the scrap material to an offsite vendor to be recycled. The piping and valves will-be cut into pieces and shipped in standard B-25 containers. Approximately 1000 ft3 of solid waste is expected to be generated per plant.-

Based on the above considerations, GPC expects to achieve significant reductions in doses over the lifetime of plant operations, with a minimum cost in-the E5-1

ENCLOSURE 5(CONTINUED)

V0GTLE ELECTRIC GENERATING PLANT AMENDMENT TO VANTAGE 5 TECHNICAL SPECIFICATIONS CHANGES RTO BYPASS MANIFOLOS ELIMINATION generation of radwaste or exposura during the modification. Therefore, GPC has concluded that the removal of the RTD manifolds will not result in a significant-increase in any adverse environmental impact and does not involve any unreviewed environmental questions, l

i ES-2

. . . . . _ . . . . . . . . ..-..... ._

ML20070S344

Contents

Text

1.0 INTRODUCTION

6.0 CONCLUSION

8.0 REFERENCES

Subject:

Dear Dr. Murley:

Navigation menu

ML20070S344

Text

1.0 INTRODUCTION

6.0 CONCLUSION

8.0 REFERENCES

Subject:

Dear Dr. Murley:

Navigation menu

Search