ML20037C691

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Provides Supplemental Matls to Resolve Issues Identified by Reactor Sys Branch.Issues Include Internally Generated Missiles,Control Rod Scram Discharge Vol,Control Return Line Mod & Intersys Leakage
ML20037C691
Person / Time
Site: LaSalle  Constellation icon.png
Issue date: 02/12/1981
From: Delgeorge L
COMMONWEALTH EDISON CO.
To: Youngblood B
Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737 LOD-81-40-19, NUDOCS 8102190446
Download: ML20037C691 (42)
v • d • e


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i Ch.ca;:J.Wa n; V,.31 February 12, 1991 Mr.

3.

J.

Youn, blood, Chief Licensing Branch 1 Division of Licensin.7 U.S. Nuclear Regulatcry Commission Washington, D.

C.

20555 s

S u bj ec t : LaSalle County Station Units 1 and 2 Resolution of Reactor Systeus Eranch Qu e s t i o r.s NRC Docket Nos. SC-373/374 LOD Cl-40-19

Dear Mr. Youngblood:

The purpose of this letter is to provide supplemental materials requested by the NRC Staf f in order to resolve issues identified by the Reactor Sys tems Branch (RSB).

3ased on meetings with that Granch on February 5 and February li, 1931 the questions requiring further action by the applicant have been reduced to the following:

1. Internally Ger.erated Missiles - Enclosure 1
2. ODYN Reanalysis and MCPR Assessment - Enciosure 2
3. Control Pod Scram Discharge Volume System -

^

4.

Safety-Relief Valves - Enclosure 4

5. Post LOCA ECCS Leakage - Enclosure 5
6. LOCA Reanalysis - Enclosure 5
7. NUREG - 0737 Items - Enclosure 7 a.

II.D.1 S/RV Testing

b. II.K.l(10) & (23) -

I.E.

Bulletin 79-08

c. II.K.3 - B&O Task Force Recommendations (13) RCIC Auto Restart (Implementatica Schedule)

(21) HPCS Auto Restart (22) RCIC Suction (Implementation Schedule)

(25) Emergency Power on Pump Seals (46) Michelson Concerns All other issues raised by this Cranch have been resolved based on material already documented by the applicant and are, therefore, not addressed in this letter.

g g) [

S

/ /

my1021oo'\\'60 h

Fr.

B.

J.

Youngblood Page Tuo February 12, 1981 Uith respect to NUREG-0737 Itens II.K.3(13) and II.K.3(22) there exists a need to request relief frca the implementation schedules defined in t h e ::'l D E G.

Tb, bases for the relic f are de fined in Sections Li"'3nd L.3'T[spectively of the FSAR and were reviewed with the NRC S ta f f management at the meeting of December 22, 1900.

That justification is restated in the enclosure to this letter for purposes of completeness.

It has come to our attention that four issues previously under review by RSS now fall within the respcnsibility of the Auxiliary Systems Branch (ASB).

These four issues are:

1.

Internally Generated Missiles

2. Control Rod Scram Discharge Volume
3. Control Rod Return Line Modification 4

Inter-System Leakage The first two items in this list are discussed in the enclosures and will not be addressed here.

Iten 3 had been reviewed and resolution reached on the ba sis o f the commi tmen t documented by the applicant in Section 4.6.1 o f the FS AR

f. c perform CRD seal leakage testing as part of the LaSalle County Unit 1 initial test program.

Item a was ultimately resolved on the basis of periodic surveillance in the hi-point vents on all ECCS lines.

This surveillance is documented in Section 4.5.1.a.1 of the Technical Specification.

These la tter two items are judged by the applicant to be resolved, and no further information is being prepared at this time for submittal to the NRC.

In the event you have any questions in this regard, please direct them to this of fice.

Very t r u l.y y,o u r s,

ll "t

e l 95

.44rk,uL-L.

O.

DelGeorge' Nuclear Licensing Administrater Enclosures 1-7 cc: NRC Resident Inspector-LSCS

ENCLOSunE 1 INTERNALLY CENERAICD MISSILE',

The NRC Staf f requested.that an assessr.:ent be cade to identify the limiting missiles with the potential for beinq generated in the coa taircant.

The potential for such missile generation hns been discu; sed in Section 3.5.1.2 of the FSAR lt was cencludLJ cn th; basis of the ecvic..

discussed thercin that no safety concern en pres <:ntac by such potential missiles due to the segregation design of ESF systems.

This segragation s

design was briefly revierced by the "RC Reactor Systems Branch and has been discussed with the Auxiliary System.s Branch.

The a.'p l i can t is currently in the proc:ss of identif,-ing, by a bounJing missile survey, representa.ive r

missile energies v.hich will be compared to the allowc:)ic cmergy absorption characteristics of the centainment.

The r sults of this essessment will be documented in Amendment 55 of the FSAR.;hich will be submitted by March 1, 1981.

o w

1 1

ODYN Reanalysis anc.vCPR Assessment i

Extensive reanalysis o f the LaSalle County oressuriration transients na~s'oeen cone using tne GDYN Ccce-to resolve concerns expressec by-tne NRC Staff.

Tne results of tnese analyses are provicec as an attacnment to tnis' enclosure in the form of Tacle 15.0-2 ano new Figure 15.0-3.

The results reportec that were calculatec using ODYN are notec.

The remainoer of the transients, with the excection of tne Roc Witncrawal Error event for anica the cooe PANACEAla) was useo, were evaluatec using~tne REDY coce.

These are~all-approveo ccces.

Furthermore, to resolve a question raised oy the Staff relative to certain results enaracterirac as estimates, all transient snc accioent shown in tne Cnapter 15 summary tacle were simulatec by coces approvec by the NRC for'One specific type. event.

A secono ccce used to calculate CPR's was not always employec.

If the event was (1) Coviously counceq oy a more severe event of ne same type (see Tacle 15cO-3 Note 2] or (2) there was ooviously no threat to tne MCPR safety limit cue to the event being initiatec at less than 1005 power Use_e Tacie 15.0-3 Note 3].

Therefore, eacn'of the CPR values reported represents an actual or councing value for the event.

3 In addition, in response to a question from the Staff relative-to tne pea < vessel pressure vs. S/RV capacity shown in Figure 5.2-3 the following infcrmation is provideo:

1.

The LaSalle County analysis was done for tne all valve i

case.

2.

Previous studies assure us that in every case (i.e.

valve capacity-?o:Lany number of valves inclucing all valves), the ODYN result will be more conservative than the REDY result.

Therefore, One result snoan in Figure 5.2-3.is justifiably ccnservative for the 3/a valve criterion allowec Dy tne ASMI ccce.

In other scrcs a 13 valve case fMr LaSalle County run oy 00YN acuic remain significantly below the Coce allowaole because the REDY result for 13 valves is celow the coce allowable.

1 4

(a) PANACEA:

J.

A. Wooley, "3 Dimensional SWR Core Simulator" January, 1977.

NE00-20953A.

)

t A

/w g

~w....

ODYf Reanalysis anc MCPR Assessment Extensive reanalysis of the LaSalle County cressuriration transients nas oeen cone using tne COYN Ccce to resolve concerns expressec cy tne NRC Staff.

Tne results c f tnese analyses are proviceo as an attacnment to tnis enclosure in tne fctm o f Tacle 15.0-2 and new Figure 15.0-3.

The results reccrtec tnat were calculatec using CDYN are notec.

Tne remaincer of tne transients, with the exceo;icn of tne Roc Witncrawal Error event for.vnicn :ne coce PANACEA (al was usec, aere evaluated using tne REDY ccce.

These are all approved ccces.

Furthermore, to resolve a question raisec cy tne Staff relative to certain results enaracterirec as estimates, all transient cnc accicent snown in tne Cnapter 15 summary taole were simulateo cy. coces approvec cy tne NRC for the sceCific type event.

A seconc ccce.(ued to calculate CPR's was not always emcloyed.

If the event was (1) obvicusly acunced cy a more severe event of ne same tyce (see Tacle 15.C-3 Note 2] or (2) there was c0ViOusly no threat to the MCPR safety limit cue to tne event being initiatec at less than 100S pcwer tsee Tacle 15.C-3 Note 3].

Tnerefore, eacn of the CPR values reportec represents an actual or councing value fc the event.

In adcition, in response to a question from the Staff relative to the pea < vessel ;ressure vs. S/RV cacacity sncwn in Figure 5.2-3 the follcwing information is provicec:

1.

The LaSalle County analysis was cone for the all valve Case.

2.

Previous stuales assure us that in every case (i.e.

valve capacity for any number of valves inclucing all valves), tne CDYN result will ce more conservative tnan the REDY result.

Inerefore, the result snown in Figure 5.2-3 is justifiaoly conservative for tne 3/4 valve criterion allowec cy ne ASME ccce.

In otner worcs a 13 valve case ter LaSalle County run cy 00YM acula remain significantly celad tne Coce allowacle because the REDY result for 13 valves is celow tne coce allowable.

(a)

PANACEA:

J.

A. Wooley, "3 Dimensional BWR Core Simulator" January, 1977.

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A",ENcxt:T 27 1:0VEMMR 1977 inc lud d.nq th.! th trmal li.iits given previoucly.

The core and fuel deuign basia for ateady-sta:e op tration, i.d., !:s'PR and LUGR limitu, hn b,en dc>iined to provide nare;in hEtveen the steady-state operating condition and anv fy, t.l__.'..:-N.---cor '.+::.i_ce.s accon odat.t 1:nc h:-.ainuiw

.:.Pt'o ensur3 that no fual dansere o,.

renultu, even durini. '..- Ucrs t anticira tsi transient c.ond3.tions N.

3. 4

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(S e

.io-n y v

4. 2).

Machanis -s that cause fuel da. mage in reactor transients a.-n..

severe overheating of fuel cladding caused by a.

inadequate cooling, and b.

fracture of the fuel cladding causad by relative expansion of the uranium dic::ida pellet inside une fuel cladding.

I re.-

o s.ign,ou cosa_s,

+.ba W a n s.i e r.'.

.i. d. ' - a w- " '.. e m. = r...i s..a...'

  • m' ~

least 99.9 ' of the fuel rods in the cera do not experience b o.i. _' 4..'.,...'. n s _i _ _4 *

  • u'._.4.~..',

."... v, c." p.r,,_-.a _' c p _ a '.i.'. ~., _ - a r. s _i.s....

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  • u e.' c'.a m. ', =_.4. s cx ~...a d. o c -u a_ v.n. -.' >

.c".._'

_ o '. a

."..u.-' ",

experiences a boa.g. :.ng trans:. t:..cn.

A value of 1, c.lastic ctrnin of "ircalov. claddinc. is c e.v,,a_. v a 4.

u. _' ", F. s.>..'..' +.. s. d a s

..n o.

_'.d.... i. ", a_ _' m. w h. i. m-s.. " u.s i d. m... g a_

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_,'.i.n n,..". a_

f o a.. l c l a d d.d..~ 4 y

s..o ' e_ x. - _s c _ a_ d o ^c u..

m e

Available data indicate that the threshold fer damage is in excess of this value.

The linear heat ganeration rate required

  • o cau,>. - 5.4.

m.. o ".. o c _l a d.*. _i..

.-.. T _i.n A' o e y e. o "..-....

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d.=_.c.-a. a s,

u.i

5. b e..- %".. o.. y.~. _4...a. a..' v,

.o 20,5 r.uf: c. t a local e::occure c: 40,000 s c/ c.

e~,. G A,,. W*".'.'u'k,

,. a - e u.4.1.5 se:- a ry o f Dasien tw ses s -

a p/w a :-

a.,-=.

In cummary then, the steady-staic c erating limits have been e

e s ". a"....^... s A

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v.o.-

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most severe abncrmal operational t ar.i~:-t.

There is no cteady-3 tat > desir:n ov.-rpov r basis An overpower which Occurs during an abnor: ai on.? rational, transien: must.'b i d e th e. o. l e n t, e w.-,---s- ---:

-l.

..r.-J 3-- '.

D.,ca.4 ra tion that the tra.uien: lj.u to are no t e::ce>mied is su f ficient no conclude that the dysign

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... i 15.0.1 Appron:5 to Safety Analysis This safety anal"... sic evaluates the abilite of the plant to 4

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The initiating events tere assigned One Of :he follovring e:[pected

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(CECO) Operating experiences with seven n" Clear pOwcr sta*; ions @

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upper bounds On the radiolO.pCal Consequ.:nCes.

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Operation would not be allo..ad 2: anf point

.p. u...

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y '.p,e..-c. a.. :..

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or v.e th.

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The N

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m'o.r

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Amendment 55 iMSERT B (p. 15.0-6)

The i...ady-s tate reactor cperating limi t ir. determined as follows:

1)

The change in the critical power ratio ( CPR) which would result in the safety limit CPR (1.06) being reached, is calculated for cach event.

These cpi values are shc.m in Table 15.0-2 2)

Thc4CPR value for each event is then addad to the Safety Limit CPR value (1.06) to yield the event-based t;CPR, except for thosc events.. hose dCPR is calculated using ODYN.

3) 'For events whose.'.CPR is determined by OYD!! (all rapid pres-surization events), the event-based MC?R is determined in conjunction wi th f.RC-additive correctica f actors, the LCPR, and the Safety Limit CPR.

These correction factors are in Table 15.0-1.

These results are given in Figure 15.0-3 for limiting transients and accidents.

The operating limit MCPR is the maximum locus of values f rom this event MCPRs calculated with the above method.

The maximum calculated MLPP. is depicted 'y the solid lina in Figure 15.0-3.

Maintaining c

the CPR operating limit at or abcve this solid line assures that the LaSalle Safety Lim.it CPR of 1.C6 is never violated.

I rf/626 2/L/01 p.

15.0-6a

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y P e n. f u e '. c n tha '.py (dir.cuse.ed in Subh tion 0. 3) is used to eva'.uate u!.e-her rwetor Coolant pres.mrc boundary d: m w CeCurs a s-a....1, e

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st6-----I.45 0- t-4.y l.1;/a_;2
C, O r, i.,,.

1

--c u -rs.:n:. t_ m - r<a.r.ct v.:e,v 6ac... x.r.,.

T-w.'. -,.-e u e: m'n-h,1--

remainc t+10a 2 00 cal /q, ' no reactor co<>1a nt prc.mure bounb.y-r v. /C A d damage r:c

..uc, 44.,,'s.u lt s. f. cm nucle a r.e r.c urc ion acc ilc r.ts. x / *,..:.:4 e"r ~ g

..a u r.v:s c.~:

c...- w.e ~;.w i

Padiolonical coneecuunces

,i

,.e In thic section, the conocquences of radioactivity rele. ace during.I.)

f ' f... r.i--

t ynes o f eve n ta :

(1) -ine; ' r "

- '= -

3- --

j fe,hm.,

..!....,. {. p i.. --...

r;._._._ y..n..z_p.pg...

i

_,-1., { s)., ; -.- -

- -.,)--

u., wi..

~_

1

-?.,-

4 s g s u.o

.g*

~.....

gg...... t

-[ 30 sign-ba s iO SCcilentd,' are Considered.

For all.eV*fntG

  • h o v.:

consequ-.nces are limiting, a de'a iled quantitative evaliation is presented.

For non-lin!t inc mn e.;I a cualitativa evaluatian is s

presented or resalts are referenced fron a more li:-iting or enveloping case or event.

{

/.

analyses are conside,' red:idesign-basis accidents; two quantitative

(

For limiting faults i

IS j

a.

The first is based en conservative assungtions

'i considered to be acceptable to the *:?C for the purposes of the worst case bounding event which determines the ade:Iuacf ot the.nlant -iesign to m+-t 10 CFR Part 100 guldelines.

This analysis is referred to as the " design-basis analysis".

b.

The second is based on realistic asstraptiens considered to reflect ex.cected radi:1:c. ical consec:aence s.

This analysis is referred to as the "rea.*nstic ana3ysis".

.ecults for both are shown to be within ':.:C guidelines.

f

-Wf.&.. i. g v..,

"g rf. ""

8 y;4.:.g jQg c, ju,n.

,a.,t.a.,k

.q C.'.. "m::

oesu1ta

~

4 E,

The results of analy,_ical evaluations are ptovided for.each event.

I n a c., t t i on. v u-- s._v+

.+.--.- -r - 3 h c wn i n. a b.,. e f

15.0.2.

Froml. '

w- -l-n~xe.re

-T. ' eut n-r r-br+:;-M.( o i th e 1intt:ng e7ent.

or

  • a.s] pa :ti cui a r a tdga r y,*._.M.ye2.. - --t.r. -e.e n,

s 1

e-~=,

-u.i<.

.~

~

.(

.1 -- -.m,wc-~~.)

15.0.5

'oteorolo ical Paran9 tars 1,

m...~.s..,,..,...

a.4 - _. 4.3,

.e.3 c a. s (.,,/o ' s )

.1...*.,_.....'.. i.". A

..u'.'.

.' a -.... -

a.

in this chapter.

The atm9spherte dilution factors for the con-servative ana!yses arc l ased on the dif f usio: -- dels pres:.:n cd U.S. NRC PeJulator./ Cu'de 1.3, Sovision 2 (J une 1974).

1 f

The atm n;t:ric diiution factors at the 50th I ;. l e for the c e. : l t.s - i c

-analy s.; h..v2 b.in derived from 2 yearc of ancite mot..orolcgical data.

15.0-8

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

-~

~

d LSCS-FSA::

AME!!DME!;T.4 ; ~

FEB RUA P.Y l979

.L TA3LE 15.0-1 I!!PUT PARAMETERS A!!D I::ITIAL C0::DITIC::S FOR TRA::SIE!!TS A :D ACCIDE::TS I

i i -

1.

Ther~al power level,

",it

->.,,./

.p., -

N.

5(lUy.N N69)

Analysi value 2:~~

2.

Steam flow, Ib per hr 14.31 x 10 6.(1041 ::3R) I i,

3.

Core flow, lb per hr 108.36 x 10 6 9

4.

Feedwater flow rata, Ib per sec 4115 5.

Feedwater temperature,

  • F 420

]

[

4 6.

Vessel done pressure, psig 1020 1

1 7.

Vessel core pressure, psig 1031 8.

Turbine bypass capacity, i::BR 25 j

9.

Core coolant inlet enthalpy, Btu per 15 529 10.

Turbine inlet cressure, psig 962 i

i 11.

Fuel lattice 8x8 1

12.

Core average gap conductance, Stu/ce:-f:'

  • F 0.1662 Q,g,.2 6 / 5 L 3 13.

Core leakage ficw, i 12 IEE' C$1DN876EY t

14.

Required MCPR cperating limit c.

F' " ~

l 15.

MCPR Safety I.ini:

1.06 16.

Doppler coefficient (-)c/'F t

i l

I l

l

!;ominal E00-1 0.221 l

i i

Analysis data 0.221 f.-

i

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ENCLOSURE 3 CONTROL RCD SCRAtt DISCl!ARGE VlLUttE SYSTEM The LaSalle County Scram Discharge Volume (SDV) System has been reviewed in depth to asccr ain the extent to which system modification it required to satisfy the design criteria developed by the BWR Caners Group and accepted by the NRC Staff.

The LaSalle County design represents, a marked imprevement over that installed at Browns Ferry and is, therefore, judged by the applicant to justify the operation of LaSalle County Unit 1.

The SDV sys:cm as it will exist at the time of Unit 1 fuel icading is shown in Figure E-1, and will !,creaf ter La referred to as the " current" design.

The current design consists of two separate discharge volume hraders, the piping for which is 10" dia.

There exists an intergral instrumented volume 12" dia" at the end of t hese 10" headers.

The instrument volume nos provides redundan: high level alarm and scran instrumentation.

The scram instruments being of the Magnetrol design.

The drain frcm the tuo instrumented vo!umes, which is a 2" dia pipe, is commen :o the two volumes, as is the single vent line.

The vent line is vented to atmosphere.

Both the ven t and drain lines currently have a single valve to isolate the SDV.

Modifications to the system are planned for completion upon receipt'of qualified equipment to provide (1) diverse level instrumentation MP Transmit:ces) on each of the instrumented volumes, and (2) a second valve on the sys cm vent line.

The taps for the additional level instruments will be completed prior to fuel loading to facilitate the completion of the modi ficai ton later.

Because of the si:e of the header system, the fact that slopes are provided to assure proper drainage, that each header has an integral instrumented volume with rede dont ine t rumentation, and the vent, which vents to atmosphere, is routec separately from the drain results in the anolicants conclusion that the system is adequated to justi fy initial operation on LaSalle County Unit 1.

Furthermore, be.ause of the size and location of the instrumented volume, UT monitoring of header le/el is judged to be unnecessary.

This discussion with details concerning component design and performance characteristic, will be documen" i in Amendment 55 of the FSAR to be submit:cd by March 1, 1931.

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ENCLOSURE 4 1

SAFETY-RELIEF vftygg A qucStion arosc relative to the 'dc use

~

-4 at LaSalle County Station for operati n uri 3 h

'4-tornate shutdo.n" scenario.

This issue w dS i

S depth in questien Q212.146.

The NRC Staff requested ha ' P I I inf rm ti n be provided relative j

to the manner in which valve f This informatiori wasprovid2dinfor.allyonDecer(er ho I,

'.s attached here m

copy i for co..pic teness.

Theattachedinfor'$ti 7Ibedocumentedina We future antandmant to the TsA3, l

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-2 OCTCd.C3 1930 h

g...a.._ t,e. 7 3 i.,. 3., o -

v u

T_

.s "SectIon 15.2.9~of the La Salle F3.u connid3:2d al. ternate nhutdoun cool.ing r!.f thods in the' event tha residual haat rc= oval (iuin) sys:s: in tt:2 suction lina may not ha usad because of valve failu b.i sin the analynis, valves in the at:cmatic dipressurigation Eyntaa (7 DS) were used to transfer fluids L tea:a, untar or a combination 02 theaa) f:Ca thair23ctof v$u's*1 to tha cu.ca.rasaion pcol.

The P22 system rocc,ven tha 4.dded haa:

s s by ccoling watur removed f:ca the :upp:cscion'p ol and injacting it into the reacto: vassal.

Me require that you pirform a test or cito previous test resultu to demonstrate that the 103 valvea can discharge the fluid 21o-7 ditiona uhan tha fluid is all,encer thi r. tout limiting con-natar.

Show that th !.c alterngt2 : ethod~i's c 'vMalg :a+uas of chu dcun ecoling by comparing tha systen hyd'raulic leasen uith the avail -

able cuaC. head.

Hydrau'.ic losses chould ba provided be de' rived frem edyr1:aental results."for cach syntan cyaponen s

s.

I RESPOMSZ

[

s w

Elevation head of 'u~ater\\ l'h the reactor vessel uas tahen

~

to ba 5 fn,et above t.'c'a st'e&7 line Mo::les.

4 S u c. o. :2ssica c s.

s is pu= ped intA the vessel b?'a leu--pressure guap.ool water s

(TSR-C) until the staa Li enthalpy per poundCof hos,ne.s are ficoded.

The resultant s

7geneous vessel water is 204 Utu/lb.

Saturation temperadu)2 of 'thi; uater is 205' 7, and the saturation. (dead pNsat.:e is 23.'.'5 psia.

The elevation of the :elief. valva outlet no::les is abc'('t 2 '

u feat b21cw the stea: aline d tscharge D'c:':les,.orovidinc. h. small amount s

of subecoling R s

the relief valve _. inlet no :la.

N

~-

Calculation of flow par val */e qssumes-no liqui.d back-pressure from tha felief valve" discharge).line;as t.he line-flow area is much larger.than the pcrt site. ;

liquid vill cause a recuco;;6n in tha. expyctedSome f::inor flashing liquid flow rate.

The reduced floc rLte fc: 4his slightly subccolod licuid is handled _as suct oy.a comtection fac c:

obtained feca Fisher cont:cl endac'c'i pagi S1, first edition, ubich i

provides a calculatt;nal cethc4.for such, a fluid stata.

s The flow s

s rate per va'lve is,calculhted to Us about 1400 gpa under tha condi tions stated.

\\

s.

\\

g.

The suppression'. cool ecolin:s =cde of the j!!2 oump has a total flow

cciatar.,ce-head of appro4 m:tely 330 f a-s t including line losues.. static'haad,?idat encha.get losses, inlet and s

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OCTO3Ea 19G0-s.._.

outlet lossa, at the pu.7.a and stainera.

At thi.c held'

  • h'a t

pt!:,1p capacicy 1; over /000 95;a for one Rii2 pu.r.o.

bn'2'00c~n b-valva low rato uhan fod f O;a tha ~ cain stea:aliilus iO 1 per va ve.

uagiCullYi then five open I.LS valVC3 ~ 9iven om"II valen, coolant flow to the regula: LiCI cyatea, i'

as

' Mco cea thg response to Question 212.132 ubich chows that' c atn.aun oc tuo ADS valvea uill cluays be availabla' mn, coolanc clo9 for that rainical case is therefore.2355 gp[i.'

1 y' "

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i 0212.146-2

/

I 81 ternate ShutJosn 00termine Flev Pcr Valve A.

Evaluate using valva data and equation below.

Cv=

\\1 When LP > 1/2 P upstreata 3 P,

'E-0 C

=-1.03 X 10.,f.SM Stesa Flcw for S/RV = 560.7 -> uss 500 V

3 x 6z55 > Stoea Design 2

Inlet Press Valva Ocme Inlet Outlet press press press Sat.Tomp Act Te.rp AP PSI a

Psia Psia Psia AP F

F ATemo Ki

-K, (inlet) 15 30

' 20 10 250 219 31

.30 9

105 120 20 100 341 219 122

.675 81 85 100 20 80 323 219 103

.64 64 65 E0 23 60 312 219 93

.50

-48 45 60 20 40 293 219 74

.53 31.8 25 40 20 20 267 219 48

.41 16.4 i

m i

j Q=C AP.

= 500 % P y

i

%

  • r-is,a.G r, Valve Indicated.

Avail a?

Q/Value

  1. From

'a?

PSI So. Gr.

GPM each valve 2 valves 3 valves 10 9

.953 1534 12229 24458 36587 100 81

.958 4602 36087 73374 110061 i

60 64

.959 4091 32613 65227 97839 60 48

.958 3543 28245 56439 84735

-40 31.8

.958 2984 22991 45982 68973 i

l Crossplot this en pump input plot.

SW r /;,~r,y,y p p y g j g g y,y;. g y-i, f

I

.hmuary 2), 1931 B.

Reading Cros: plot 1 Value 2 Valus 3 valua Done pressure psia 122 3:

. Valve flow, #rin 39500 52000 550C0 Pump f1:.< in g;m 4795 6313 6793

-C.

Check to sea if we can keep dome from filling up.

Elevation of steam lines = ?.[' 9" Elevation of Celief valve outle = 16.5" + 10" + 13" + 27'9" s 30' i

.6.T rp 7.?.>;f

.g;',- ? ' -::$%

, j.11. i'?",(h If C = 550/ valve & water is 0 200*F, evaluate flew at 23.75' y

s elevation haad cnd 0 back pressure (psig).

P.eevaluate vessel water temperature figuring 1/2 vol to fill to steam lines S

Decav heat s 9.5 X 10 BTU.W ed VeselenthdpF=(/

!1095ft[218.8 E

1

~dIT01 f t*/

6 142.7 x 10 BTU

=

~

\\

6 94[3750 New added fluid =

( 3T623)= 21.7 x 10 g7g l

j New total enthalpY = 142.7 x 106 + 21. 7 X 10*-9.5 X 10S f

6-

= 173.9 X 10 6"

New Mass = 11095 3750 =.8B X10 '

+

.01701 707323 4

Enthalp;//it = 179.3 x 106/.83 X 106 = 203.7 STU/i! =

SAT ted; = 235 F SAT press 23.25 psia done press.

24 ft elevation head s 9.37 psi + vessel dome press SS. /QWd

.01639 x 144 SAT temp at 33.13 psia = 255 F.

ACT temp in vessel 236 F zFF i

Frem Fisher book p 81, K3 =.225

}

.225 (33.13) = 7.5 psi S.G =.950 4

i Q = 550 l.9 i

EG.

Q = 550iI'i.50 cr 1547 g n/valua F.ved i

I

,-,,_m

,m.

..,,,,,n

J,inuary 2), IW1 C.

(Cont' d. )

C 550 Fill to 5' ov w st m 1inas 1vaYv2 Ts47 cp:a 1610 l

=

2 valves 3093 g;a 3220

=

3 valves 4540 gpa 4830

=

4 valv2s 6137 ep 6440

=

5 valves 7734 gpm 8050

=

Sensible heat /1 F At !! da C 233 plant s

residual 'r.elt sensible heat 0100*F/hr + RHR pu,rp heat

(143.1 X 10") 2 locps

- remove residual heat and caic remaindar 1 RHR pu.rp 0 1200 EH? X 42.44 BTU X 60 nin min nr 6

Qp = 3.05 X 10 GT0/hr pump 6

Res; dual heat = 3579 x 10 Uatts x.011 x 3.419 STU 6

Resid ht = 134.3 X10 BTV/hr 6

6 Total heat = 143.1 X10 x 2 = 226.2 x 10 STU/hr 6

6 6

Sensible heat = 286.2 x 10 - 134.3 X 10 - 2 (3.05 X10 )

6 6

Sens. = 145.8 X 10 BTU

=

1.458 x 10 BTU 100*

'r D: pes /a?9-451 1/29/81

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=

EllCLGSU2E 5 POST-LOCA ECCS LEA" ACE Revicu procedure 111.20 of SRP 6.3 requires that long term cooling capability folicwing a LOCA shculd be adequate in the event of failure of any sing!c active or past,ive compo,cnt of the CCCS.

The NnC Staff required the applicant to iliscuss how leakage fror. the first isolation valve in an ECCS suction line frem tha supprc:

an pool during post LOC #

long term cooling will be contained. The expresserl concern is drainage of the suppression pool (heat sink) in view of the possible inaccessablity for repair of leaking valve due to local contamination.

The scenaria postulated has been reviewed and is judged not to present a safety concern in light of the LaSalle County design.

All of the %CCS suction valves are designed with a back seat to allow volve repacking at full system pressure with the valve open.

In addition, the limit suitches and torque switches have bacn adjusted at the factory to l

1isure that in closing and opening, the travel is stopped t f the torque switch with the limit switch set to shut of f for backup safety.

This means that anytime the ECCS suctica valves are in the full open position (normal posi tion) that the packing g!snds arc isolated from the suppression pcol water by the valve back scots.

For these reascas, leakage of the type postulated is not expected to occur.

In cJdir, ion, the leakage monitoring program defined in Section L.37 of the FSAR to address NUREG-0737 Iten Itl.D.1.; will r.taintain leakage rates for these valves to an "as low as practicable" levnl. A! cenate sources of water (condensate storage tank, service water. etc,) are available to assur2 adequate heat sink capacity by refilling the pool should excessive leakage occur.

This information 4.ill be documented as an addenden to the response provided to Q212.37 of the FSAR.

4

ENCLOSUR5 6 LOCA REANALYSIS Tha revision to NP.C Qucstien 212.143 subnitted uith kendment 50, October 13:0, providerl the ?CT inpact fer a realistic conscrvative FCV closure ra te of 11's second follcuing LOCA.

As previculsy stated, the probability of any closure is extcremely unlikely.

However, to respond to !;RC cencerns, the impact of an 11% second closure was provided.

It should be noted that t'.co failurcs wculd be required to initiate this closura.

They are a failure of dryuell high pressura signal, which locks the valvc in position follouing a major LOCA, and a failurc of the electronic valve centrol.

Recent telephone conversations with the NRC have led to the belief that an.11% second closure rate would be considered realistically conservative if the pressure sensors were properly qualified for a LOCA environment.

It has been determined that these drywell pressure sensors are located on the instrument racks mounted outside the dry..all and are connected to instru..ient lincs originating in the drywell.

The sensors are also indentical in design and manufacture to the pressure transmitters which provide a high drywell pressure signal to the RPS and ECCS systems.

The nroposed ravision to Q212.143 which documents this conclusion is provided as an ottachment and will be formally sub:aitted in a future amendment to the FSAR.

l 4

(

NUREG-0737 Items a.

II.D.1 S/RV Testino Commonwealth Eoisen is a particicant in the BWR Safety / Relief valve Testing Program now unoer active revien by the NRC Staff.

The formal commitmen; to participation in anc acherence to the results of this test program are occumentec in the L.

O.

DelGeorge letter to S.

J.

Youngblood (LCD 81-40-21) dated Feoruary 12, 1961.

b.

II.K.1 (10) & (23) - IE Bulletin 79-08 Commonwealth Edison nas performed a comprehensive review of the procecures associa:ec witn coeracility status (II.K.l(10) and reactor vessel dater level (II.K.l. (23)). This' review was

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documenteo anc crovicec to tne Regional Office of Inspection and i

Enforcement (Resicent Inspector).

It is jucgeo, therefore, tna LaSalle County Station is in conformance with tnis task action item.

c.

II.K.3-8 & O Task Force Recommencations Commonwealth Edison has committed to provi.f3 additional information in support of the following responses now documented in the FSAR:

(i) L.34-13 "HPCS Auto Restart" (II.K.3 (21))

(ii) L.34-16 " Emergency Power on Pump Seals" (II.K.3 (25))

(iii) L.34.24 "ACRS Consultant Cuestions" (II.K.3.45)

With respect to item (i), the Staff excressec concern that the HPCS system, thougn reviewed as a part of tne SWR Owners analysis of this task, was not accressec explicitly in the conclusions stated for LaSaile County in Section t.3a-13 of the FSAR.

The FSAR nas ceen reviewea ano acoraorlate cnanges mace to resolve :nis concern.

rne proocsec cnange to L.34-13 is attachec anc aill ce sucmittec formally in a future amenament to One FSAR.

Also wortny of note is the fact tnat the LaSalle HPCS will cuto-restart after a manual termination.

With respect to item (i!}, the Commonwealth Eoison design oces not now provice emergency power to tne recirculation pump seai cooling oater.

Therefore, tne current cosition statec in L.34-16 of t9e FSAR can not yet be changed.

However, it is nortn notinc :nat these cooling water systems can ce manually transferred tu an emergency cus if requirco.

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. With recpect to item (iii), the Commonwealth Edison review is expecteo to be completed cefore Feoruary 18, 1981.

The'results

.of that review will De transmitted immeciately upon completion.

Two other items are ciscussed here only to re-emphasize the point that-altnough commitments have oeen mace to mocify the plant as recommenced in NUREG-0737, equipment.celivery scheoules are expectec to preclude satisfying the requi~ rec implementation cates.

The two items in question are:

(1)

RCIC Auto Restart-(L.34-6. II.K.3(13))

Availability of qualified components will prohibit-final implementation'by 1/1/81.

System cesign for tnis mocification was initiatec in the Fall, 1980 prior to imposition-of tne requirement in Novemoer, 1980.. Final design was comcletec in Decemoeri 1980'ano equipment-purchases will ce expedited to assure installation at the first outage of. sufficient duration to complete the work.

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(ii) RCIC Suction Transfer (L.3414 II.K.3(22))

The situation for this modification is the same as described for the RCIC auto-restart. -For that reason, it.

is judgec that tne implementation cated 1/1/81 will not be met.

For both items II.K.3.13 and II.K.3.22, equipment celivery schedules'can be proviced to the Staff upon acceptance of firm proposals from eculpment sucoliers.

These senecules would reflect the impact of equipment qualification testing as well.

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T h i c ):e vicy han inclu %d a co.icidrat /.v.: c2 i.'.1. acpecto of "?Cl, Lin'C ani' fit Ci nyu tor, operac. o: :: t:5!ch t ould b-inflt.c ac d b y a ny n:.:pti:tdad cutenit hic ror, tart capaby li ty.

It is cou :Lud.!d that the s

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fle :ibility and tn?4 other undesirable 0:.lecto discuuced in thic ncnorandan.

The "SGS Vendor Lind the BUR C'.4ners: G roi.' 7 beli tva the current DWR ECS design, uhc-n coupled uith rAgmroius and contirtuotts o.n. o ra ting u t e ' t' t aining n. roc.;rws, rc.70s n ti: thu o p '. i nt2:a approach to T43 c u L.' C 'f.

' 'o n o.l i. 7. C 't t.'.a n 3 o f e::iutia3 LPCI.

.ef '. LPCO need to ha undertaken.

S.--$J D N PC 5 2.

One of the factors which su.cc. orts the current adec.uac.v.

cor.cluu.un ::.a the D&ciod of grace-Pin i'.*/ailable to th=-

operator betten i:h e inchant :h : th oi nator thould have startad an idle ECC s.aten (but does o >t) c'id th.: instant uhen predefined core cooling difficulties uculd occur (fuel clad attains 22000 F).

Given a 3*43 core initially at natur ttion to: pa catura conditiona 9.ithout any cource of makeup reactor water (becauce the operator has cr.conecualy ' erninated ECCG pump c

f lo.s), the foltouinc; tui.u]ation cit:estrines th" ti:re internal bot. wen pu:ap flo h rmin.ttion..:nd occur.*:ence of 2200" F fuel ciud. t e:.1p 'r,t tura:

I:ini. mum Tin s In ter'/al

_P_.O:; t M1.t h e d. _C_a. G'. 3 TO P.%1Ch 22000 P Cac 1.

Hail of f f rom T.c' vel 1 30 minuto:,

31 5 inches above top Of factiVU fuOl Cac2 2.

Poil o.'f fro" top of j t PI::.['

15 to 20 :: 41.nueoa I n C i;. -

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