ML20238A369

Forwards Complete Response to Question 7 of NRC 861216 Request for Info Re Util Response to NUREG-0737,Item II.D.1, Relief & Safety Valve Test Requirements

ML20238A369
Person / Time
Site:	Rancho Seco
Issue date:	08/27/1987
From:	Andognini G SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:	Miraglia F Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM GCA-87-522, NUDOCS 8709090264
Download: ML20238A369 (2)

Text

-- --

x .

$SMUD

' SACRAMENTO MUNICIPAL UTIUTY DfSTRICT O P. O. Box 15830, Sacramento CA 95852-1830,(916) 452-3211 AN ELECTRIC SYSTEM SERVING THE HEART OF CALIFORNIA August 27, 1987 GCA 87-522 U. S. Nuclear Pegulatory Commission Attn: Frank J. Miraglia, Jr.

Associate Director for Projects Philips B1dg.

7920 Norfolk Avenue Bethesda, MD 20014 DOCKET 50-312 RANCHO SECO NUCLEAR GENERATING STATION UNIT #1 NUREG 0737, ITEM II.D.1 REQUEST FOR INFORMATION

Dear Mr. Miraglia:

Your letter dated December 16, 1986, requested additional information to complete your review of NUREG 0737, Item II.D.1, " Relief and Safety Valve Test Requi rements. " The District responded to that request on March 3,1987 (JEW 87-172) and informed ,you that additional information would be forthcoming. I Our May 14,1987 (GCA 87-027) letter updated our commitment.

Attached is our conplete response to question number 7 of your December 16, 1986 request for information. The reanalysis of the pressurizer i safety and relief valve discharge piping is complete. I If you have any questions, please contact John Atwell of Licensing at extension 3906.

Sincerely, l tLfd xwo

^

.- - f G. Carl Andogfiini Chief Executive Officer, Nuclear Attachment ,

cc: G. Kalman, NRC, Bethesda A. D'Angelo, NRC, Rancho Seco J. B. Martin, NRC, Walnut Creek B709090264 070027 /

PDR ADOCK 05000312 PDR

. plOTsth P g I

RANCHO SECO NUCLEAR GENERATING STATION O 1444o Twin Cities Road, Herald, CA 95638 9799;(209) 333-2935

d

.c NUREG 0737, ITEM II.D.1 QUESTION 7: The submittal indicated the limiting transient for piping i stresses was the steam discharge transient and not subcooled '

water discharge. This result is different from the results of a similar B&W plant (TMI-1) which found, with similar initial conditions, that subcooled water discharge was the limiting case. In addition, EPRI/CE testing found water seal discharge

} resulted in greater piping loads than steam dischcrge without water seals. Therefore, provide more details on the differences in the stress analysis between the steam discharge case and the subcooled water discharge case. Include reasons why the steam discharge case was the limiting transient. Also include a table comparing the calculated stress with the allowable stress for the most highly loaded pipes and supports for both the steam and water discharge cases.

RESPONSES: As discussed in our April 12, 1985 submittal (Question 8) the Rancho Seco thermal-hydraulic analyses considered transients involving both steam and subcooled water discharge through the SRY's and PORV. Under these transients, the saturated steam discharge cases generally produced higher thermal-hydraulic piping reactions. It was judged that these would be the more credible transients especially when combined with seismic ,

forces. The stress analyses for the subcooled cases were j therefore not completed, l 1

The concern of greater piping loads from water seal discharge l is not applicable to Rancho Seco since there are no loop seals i at the safety and relief valves. Thus the extremely high j dynamic forces due to a slug of cold water being forced down I the discharge piping by high pressure steam were not analyzed I at Rancho Seco.

Based on these analyses, the piping system was upgraded in 1965 to meet the anticipated thermal-hydraulic and other loads. During the performance of confirmatory analyses to verify piping stresses under the as-built support locations it was determined that the system modeling required additional refinements to accurately predict the system response to I dynamic loads. A reanalysis effort was conducted to improve the model and fully verify system adequacy under all the design basis loadings. The results of the reanalysis showed  ;

that all but two of the thirty-five supports were qualified under design basis loadings including the sub-cooled cases.

These two supports will be modified to satisfy design basis loading requirements prior to plant restart from the present outage.

IOl Ch u,

\l y y

.~ :-g l

b s

<,~s

• .: 3, 39, , i cg4o r.. .n e " s, e.fne,,

g ,

f FS (* 'd / 4- ')() 9 ( f g ghg, d'

. 4 f* r ' ] l d'g b I '

t . .. T . , ,,7 i

3 c- <

..f-. it . i ., n-t y,,..,, ,q.

'; , , ,,, y, =4 ,

,p .s ... "+ ,

s' ,

, ;tp -s i t . - 4 ,

' * ' "r, '8 ., [ !m

!~ e *. +

• r , r<- ['>, s i * ** I i;

. ., r A '? '. u , s * .t, , ., e  : 1*<

qr a f,,;; feo1 v r, 4 c , i i, ) 'l ,

[? ?r.+ " 9\ 7 O S it o {

• I fsh Y i.,

t a

r 1 a 21 s *t F I ' m 's 'i ! s 'gr v - 4 9  !} f( I f)f j h,'$er ' t , 'j r" r ) s r' A Is r y4 [ ~ [ ', 6

. 1

'E '

i. i4 ) f)t S 1 f \i i g '7 (g . $ *j '. f t

, *4 4

N ' f g ;# ht f r4( a^ 5 q fi S / '

• J  ;) ) r I * - 9 t, ='t S r }4g ( I 'l rs '{

• 9 '= "r I

.,4

% .g rs ' 4 s [,, n T c. 3 r .. ?m- }!, .,.t f, .# l ,3 f.>i , f i , - , I& f,2in < : e, e

! d"" '

", .,

• k rd } 44ff

• f' I g " ,^g ( ' g e (

,5 5 } . .,, . , . ,,; ,

r. s e r - ,M y, , <,7 t c a. , , p, . , 7

*e4. ri , e t , i k. i  ; i..g'. * ., i . 4, ,7 ,, : T -fT.

t y [- $ , e  %,

g 4 9 ,3 _

5 I

l

.I 3 [  % r A e' \ l 45 * ' # I ' ' '

t . i '

1_  ?

.s ty 4 9  ! l  !

{

+

• 4 ., -

h fp . $. }

. r; o ,' '

3 . s ,

., < , r: ,

n I 1 p , k4 1 - 9 f

, f

^

9

• t 4 e p p .-

g

-t a s

% N l

SE ib y 8- d3S L9bi S0-08HSD

- -___ _ ___- ____-_____ - . _ _ _ _ . _.