ML17212A508

Forwards Addl Info Necessary to Continue Review of Util Stretch Power Request

ML17212A508
Person / Time
Site:	Saint Lucie
Issue date:	07/28/1981
From:	Clark R Office of Nuclear Reactor Regulation
To:	Robert E. Uhrig FLORIDA POWER & LIGHT CO.
References
NUDOCS 8108130321
Download: ML17212A508 (8)

Text

~ 4 l

JUL 2 8 1981 Docket No. 50-335 Dr. Robert E. Uhrig Yice Pres,i'dent Advanced Systems 5 Technology Florida Power 8 Light Company P. 0.

Box 529100 tliami, Florida 33152

Dear Dr. Uhrig:

Based on our review of you stretch power request for St. Lucie Unit 1

we have determined that the additional information identified in the enclosure is necessary to continue our review.

Please provide this information as soon as possible.

Sincerely, Original signed by Robert A. Clark Robert A. Clark, Chief Operating Reactors Branch ¹3 Division of Licensing

Enclosure:

As stated cc:

See next page DISTRIBUTION:

Docket File NRC PDR L

PDR NSIC TERA ORB¹3 Rdg DEisenhut OELD ICE-3 ACRS-10 RAClark PMKreutzer-3 CNelson JHeltemes Gray File

~ 8i08i303ai 8i072d PDR ADDCK 05000335

(

P PDR OG Qz p U~

~IQ+

y,S. ~pglSS11 g

~R oFFlcEI1 SURNAME)

OATE/

ORB¹3:DL Pjgjeu zei'-

7 "7 81""""

ORB¹3'DL

~ G QQQ%pne

~ ~ ~

~ Ijt/9'/Q10 ~ ~ ~ ~ ~ ~ ~

B¹ Anila pg o ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~.7jQ/81" "

1NRC FORM 318 ll0/80) NRCM 8240 OFFICIAL RECORD COPY

• USGPO: 1980-329.824

4 4

(l'

~

'I l J y

P U

I E

1

~"

Florida Power 8 Light Company CC:

Robert Lowenstein, Esquire Lowenstein,

Hewman, Reis 8, Alexrad 1025 Connecticut Avenue, H.W.

Washington, D. C.

20036 Horman A. Coll, Esquire NcCarthy, Steel, Hector 8 Davis 14th Floor, First Hational Bank Building Ni ami Florida 33131 Indian River Junior College Library 3209 Virginia Avenue Fort Pierce, Florida 33450 Administrator Department of Environmental Regulation Power Plant Siting Section State of Florida 2600 Blair Stone Road-Tallahassee, Florida 32301 Nr. Weldon B. Lewis County Administrator

'St. Lucie County 2300 Virginia Avenue, Room 104 Fort Pierce, Florida 33450 U.S. Environmental Protection Agency Region IV Office ATTH:

EIS COORDIHATOR 345 Courtland Street, H.E.

Atlanta, Georgia 30308 Mr. Charles B. Brinkman Manager - Washington Nuclear Operations C-E Power Systems Combustion Engineering, Inc.

4853 Cordell Avenue, Suite A-1

Bethesda, Maryland 20014 Nr. Jack Schreve Office of the Public Counsel Room 4, Holland Building Tallahassee, Florida 32304 Resident Inspector/St.

Lucie Huclear Power Station c/o U.S.H.R.C.

P. 0.

Box 400 Jensen Beach, Florida 33457 Bureau of Intergovernmental Relations 660 Apalachee Parkway Tallahassee, Florida 32304

0

EtlCLOSURE Tne -:~verse bo. on worth values listed in Table 7.1.1-1 are incre=sed fo.

ai I:-.oces of c=-.. ation.

increased inver se ooron worth mean=:.-:-=

.;":.=. b:ron m.s bc dilut d for a given chango in reactivity, which is

~ess conservative.

mesc> ibe the oasis for and justify the new values of inverse boron worth for each mode of operation.

2.

The re ueling shutdown margin listed in Table 7.1.1-1 has been ci ar.".ed from (7.1.1)

\\t ~

7.1.1)

(7.1. 4) 9.45'ubcri tical to 6.28~ subcritical, which reduces the dilution tim s to reach criticality.

Hhat is the borcn concentration that corresponds with the. new shutdown margin?

Compare this v i th tho previous refueling boron concer tration, Tn results of the. boron. dilution events shown in Table 7.1.1-2 iist the times tc lose prescribed shutdown margin for each mode.

Please be a>>ar that SRP section 15.4.6 specifies minimum times from when an alarm makes the operator aware of an unplanned di'ution event as acceptance criteria.

Rhat alarms make the operator aware of boron dilution in each mode?

! hat are the setpoints, time delays, and errors associated with detection and alarm systems, and how are these accounted

=or -in the time -,or the operator to react to a boror, dilution event?

T.-e oarameters shown in Table 7.1.4-1 are stated io maximize the cai uiated peak PCS pressure for a loss of load even

However, the initial pressure of 2200 psia is lower than the value previously utilized(2250 osia) to maximize RCS peak pressure.

Provide further discussion on.why a lower initial prcssure ls conservative, or evaluate the effects of a nigher initial pre'sure cn tne calculated peak pressure.

'pl c /-

r I "r-I-

r-i ~

~

q i "q

~

1

~ )~r 4 10 h5

~

c

~

~

~

t

~

i I

>> I Ii'

~

C>>

vl g ch Q

~

-coressed and:.<'.v o= n".r -conservative.

P Iease discuss the following

) tho initial core;=~.".;-" is at 1005 rather than 102" as required by SRP 1

=ection 10

~ 3. i '

> w;:e::;er the most reacti ve control element assemhl.v i'

held out of the core:.

no bases are orovided to iustify the oump (7 2;3) coastdown curve.

The Loss of Non-Emerc ncy AC Power event i'Ilizes the same DfiH analysis used for the Loss of Coolant Flow transient (7.2.2).

The items in auestion 5 mus.

be satisfac~or'.y resolved before the analysis for Loss of AC Power will be considers-va7 "'.

in addi.tion, th= value of 1.15 used

=or.he dop"ler coef icient mul tipl'ier.-;."s. be justi ied's conserVative considering tne previous value of 0.85 used in the FSAR.

Provide justification for the values of the initial cor>> coolant temperature 3 2) and pressure to snow that they are conservative for the Steam Line Break analyses

~

Also discuss the basis for the ini "ial core fIow rates a.sumed and::"e delaved n u-.ro;. fraction C

~

I 8

A)

';o D"S analysis v!as oerformed for tne S.=a-Line Break accideni.

c so the..

ra~i'ystem depressu

':.a ion. 'at he minimum Bl>B I atios calcu!ated.

(AD.a!

Th S 'am Generator Tube Pu "iur'.

ven.

snows a rapi d d. oz '

RCS ores sul e and

-.emp rapture a. about 600 s conds in figures 7.3.3-3 and 7.3.3-':.

Please orovide figures with firer detail in this region (approximately 550 to c50 seconds) and evaluate tre charces of and effects of steam bubble formation in the vessel head or not legs.

The effects of ste:m bubble formation on the radiological evaluation should also be considered.

1

Sei ed Rotor analysis ooes

"=-. 'ncluo

=- c~lcuiaieo '"., D::-;.

(zo account for statistical uncertainties w-.

~ the new CE methodology),

or a oeak clad temoeraiure as re".uired bv SR.= section 15.3.3.

1n addition, the SRP requires that tho seized rotor event is analvzed with a coincident turbine trip, loss of offsite oower, and the most reactive control assembly held out of the core.

Please provide this inforr ation which is needed to meet the SRP.

The maximum Tech.

Spec.

steam g nerator tube leakage rate should bo used in evaluating the radiological conseouences of a seized rotor accident.

0 4

4 l

'4

,n I

y't h