ML20128C375

Requests Listed Addl ECCS & LOCA Info for Review of Application for Monticello Reactor

ML20128C375
Person / Time
Site:	Monticello
Issue date:	03/24/1969
From:	Rosen M US ATOMIC ENERGY COMMISSION (AEC)
To:	Boyd R US ATOMIC ENERGY COMMISSION (AEC)
References
NUDOCS 9212040375
Download: ML20128C375 (2)

Text

. _ - -

(.

(agz i

l NAR 2 41959 l

66-26 1 R.' 8. Boyd, A/D for Roseter Projects. DEL QM slped t Wap

8. Levine, A/D for Raaster Yeahnology, DEL i te+,

WesTICELLO - Ab81T10EAL BOCs AND lac & Ispenes&TIcu 1 -

The following additiemal informaties is regetred for one review of the

\\

POL application for the Meetime11e naaster.

E ~

(1) Dneeribe the provietans that have hoes ande to essene the evail;'

e

.,. #m

~

ability of the B005 is a post-socident perted following a passive failure (water leakage) of the ring header er the accs piping.

Describe the emans availabis to the operater to detect and isolate such failures under post-accident conditions. How much time omst elapse following the design basis LOCA before these previsions aan adequately limit the someequences of this postulated ECCS failure modef i

1 (2)

Design basis "s" on page 6-2.1 of the PSAR is someoomervative for cladding exidation and cladding skatter concerns.

Provide a more realistic clad temperature limit for the LOCA.

l (3)

For clarification of design basis "f" en page 6-2.1 of the PSAR Are the individual systems designed to withstand the sembined LOCA and seismic loadet (4)

The ds. sign basis for the auto-relief system is not clear.

cive the relief capacity, member of valves required, and number of valves to be actuated.

Is the auto-relief permissive of "no NPCIS sod /or feed-water flow" to be removed as in other usoest applications?

If so, how does the combined acties of auto-relief and HPCIS ebange the small-break thCA analyseat following any baeak so as to preclude vessel repressartaatiaatWill the am We currently require an interlmak to prevent amte-solief metenties 'in the event a.s. power la met available to drive the les paesasse BCC e

seesystems--will oesh an interleek he peweidad for this plantt (5) surisatian taflammes of the EPCI System for all aut's.We end the 1g.

The 5 -

tapiaal report, APID-5608,-deserthee testa skish are intended to

'J peevide seek verifiaaties by mid-1969. Will the test resalto-be-available te samfits your analyses? If not, them othe wise demon.

strate the adequesy of the BCCB design for that range of internadiata '

breaks for which EPCIS depressariaation of the vessel is most important.

w.4.,

u _ _.....

.2.~_...

__ - a._ m...w f._

.y g..-

9212040375 690324 PDR ADOCK 05000263 A

PDR

. -. - -. - - -. -... -. -.. -. ~ - -. - - -

~. -. -.... - -

l I

1 l

4 R. S. Boyd !

1 to a meaningful clad temperature limit (not the Zircaloy melting l

targe rature).

i

{

(6) It appears from the drawissa la the FgM that a break in the RCIC steam feed line could result in loss of HPCI esoability due to prosimity of systems (the "high eres temperaters" signal veuld isolate both systems).

Describe the likelihood sad eensequeness l

of this event.

(7) A bubble rise velocity of 1 f t/see is cited on page 6-2.35 for the liquid break analysis, while a velocity of 2 ft/sec'io etted en i

j page 6-2.40 for the steam break analysis. The 1 ft/ses assumption j

is less senservative for determining time of core sneevery, while the 2 f t/sec is less conservative for determining taatimum icvel I

swell in the vicinity of the steam line noszieo.

Give the reaan i

for using the two velocities and establish the conservatis:a of your i

selection in each case.

(S) Provide a quantitative assessment of the analyticci conservatist.,

that has been retained in the core heetup ca culations for inter-1 mediate and large break sises (0.1 to 5.5 f t}).

Include consideration 1

of such things as transition and film boiling af ter MCurR <1, steam j

cooling, residual water levci, and steam availability for the :wtal-water reaction.

1 gging syndbt 31,lloscu,

Morris Rosen, Chief l

LT-260A Nuclear 7echnology. Branch j

DRL

• NTE :RJ'l Division of Reactor Licensing cc:

D. Vassallo, DEL i

l 4

Dis tribution :

Suppl.

DRL Reading ADRT Reading NIB Reading j

bec:

S. Levine M. Rosen R. J. Mattson t

omct >.. DEL:KTB......

. D D.: NIB _.. D DRI..

..I1RL:ADRI...

[M b

j sURMME > RJMattson:rbw......MRosca....

oting..

...SLevine.....

.;.)lYl$.~_.)IN. 09..

- ?.Y!.N?....

Dnt>.)l 2Nl$..

~......

. ~. -

n,.

m..inim.mn I

.