ML17309A221

Forwards Comments on NRC 811103 Safety Evaluation & Draft Evaluation of SEP Topics VI-2.D, Mass & Energy Release Inside Containment, & VI-3, Containment Pressure & Heat Removal Capability.

ML17309A221
Person / Time
Site:	Ginna
Issue date:	02/01/1982
From:	Maier J ROCHESTER GAS & ELECTRIC CORP.
To:	Crutchfield D Office of Nuclear Reactor Regulation
References
TASK-06-02.D, TASK-06-03, TASK-6-2.D, TASK-6-3, TASK-RR NUDOCS 8202090299
Download: ML17309A221 (116)

Text

cebruary l, 1982 y..rf ~

Dizec"cr of Huclear Reactor Regulation At>>ent'cn: Hr. Dennis M. Cwtchf eld, Chief 4 Operatinq Reactors Branch Ho- 5 U.S. Huclear Regulatory Commission Washing on,. D.C. 20555 gQii,'M '?

Subject:

. SZP Topics VI-2.D and VX-3 R. Z. Ginna Huclea= Pcwer Plant ~j( t ?

Docket Ho. 50-244 Dea iW. Cwtchfield:

This letter is in response to the diai evaluation of SZ?

Topic VX-2.D, "Bass and Energy Release Inside Containmen " and

SZP Topic VI-3, "Containment P essure and Heat Removal Capability,"

which was ""ansmitted by your letter dated Ho@ember 3, l981.

We have reviewed. the draft evaluation and. have identi ied several ccnservatisms in the analysis or Ne loss of coolant accident (I,OCR). These ccnservat'sms and a qualitative c?iscussion of

. their impact on the L'OCA results, as well as a neer of general

~ ccmments on the evaluaticn, aze dicussed in At achment A. We also identi ied a number- of conservatisms in the analysis fo=

the ma'n steam line break (A? SLB). Because of "Ne degree o conservatisms in the HRC evaluation, we pe formed a sensitivity studv w"'x. "he cede CONTEMPT-FI/2SB. This code is verv similar to CONTZKPZ-LTOZS, as discussed in Attachment B. Therefore, wh'le "his., cade has not been completely auali ied oz use as a l'censina cede, we believe'that i 's accurate and adecuately represen s Ginna. The sensitivi"y study, presen ed in A>> achment B to this letter, confirmed the conservatism cf "he HRC results foz MSLB.

Regarding the LOCA analysis, since the Ginna design bas's pressure envelopes the HRC resul s, we conclude that "he Ginna ces gn 'oas's pressure prof.'e rema ns acceptable. The Ginna design basis temperature profile exceeds the HRC resul"s except bet'ween IO,OOQ seconds ard approximately 20,000 seconds a"-e "he des'cn bas's ' even". We-propose Cat the Ginna c?es'an 'oasis

~ " mperatur prciile rema'n as shcwn in "he "=SAR =or times 'ess

~Man O,GQO seconds and be revised beycrd IQ,QQO sec nds as ccws:

.from ',000 seconds "o 20,QQO seconds, tempe atu e ~ 250c= bevond

. 20,000 seccrds, temperature < lQOcp. Since -'e containment tem-oerature 's alreadv c?ecreasing at this "'me, '= 's nct ccnside ed

.ROCHESTER GAS MD ELZCTRXC CORP ~. SEEET HOe 2

'DATE February 1', 1982 TO M e Dennis Hi C utchf ield that this zevised profile would invalidate any conc3.us'ons drawn as par of ouz review of envizonmental qualification of electrical equipment, since the affec" ed equipment temperatures would also be dec easing at this time from their peak temperatures.

Regarding the HRC steam break analyses, the HRC results foz pressure and temperatuze exceed the Ginna design basis as revised above. However, as shown in Attachment B, acceptable results foz containment pressure are obtained when more reasonable assumptions aze used. Xt is our conclus'on, based on Attachment that the Ginna design basis pressure profile exceeds the pressure profile calculated for main steam line break and, therefore, remains acceptable. We conclude that the temperature result'ng f om a steam line bzeak in but containment may exceed the Ginna temperature profile, that this is of no consequence design.'asis due to the short: duzation of this exceedance and may therefore be ignored. This. concLusion is based on HRC guidance provided in the Division of Operating Reactors (DOR) Guide3.ines which in turn is based, for example, on the shor duration of the tem-pezatuze spike, 3.ower heat t ans er coefficient, and the elevation of the steam Lines rela 've. to equipment. Thus, we conclude that the Ginna design basis temperature profi3.e, as rev"sed based on LOG, results d'scussed above, remains val'd.

Very truly yours,

~ ~

Jo Z. Aaier

ATTACHMENT A CO5R~S OH HRC. SACHET EVALUATXOH'.

5 Containment conditions. resulting fzam a main steam break were assessed in a grass fashion in FSAR page 14.2.5-10.

That analysis does not, however, comply with curren c iteria po 5 The LOCA analysis underestimated'he effectiveness of the containment. an.caolers by a heat removal rate of approximately

25. to 30 million BTU per hour. This resulted f om an incor ect set'falsodataLERbeing provided by RQ&E to the HRC consultan 81-022 transmitted by latter dated January 4, (see 1982 fram John E. Maier, RQ&E, to Ronald C. Haynes, HRC).

We have- estimated the impact af correcting the fan cooler heat removal rate to be on the order of a 1 to 2 psi reduc>>'an in ceak containment pressure. This, thezefore, is an additional conservation in the analysis

p. 6 The SER d'scussion of the resu3.t 'af a main steam 3ine break shou'1d be revised to reflect the sensitivity study presented in Attachment B; po 7 The last paragraph of the SER should be revised ta re 3.ec the conclusions presented in cur 3.ettez t ansmi"ting this attachment.

See our comment above regarding FSAR Section 3.4.2.5.

The assWption of a canstant containment pressure of 14.7 psia throughau the transien"- wi13. result in an overestimate of the mass and energy re3.ease and, therefore, an overestimate of containment pressure and temperature.

p. 16 A1X. information provided in conversations was also avai3.able on the docket.

p.18 Xt is 'cur understanding. that only accumulator water was (conservatively) set at. 272.9oF, not all ECCS flow. This should be c3.arified'. 3:f our understand'ng. is not carrec>>,

then Table 1; which ind'cates RhST temperature, shou3.d.

be revised.

p. 21 The ccntaimen design pressure is incorrec>>Dy stated at.

the top of the page as being 74 ps'a; i is 75 psia.

la. General A number o other'anservat'sms are discusCed in the LOCA evaluaticn. A mare reasonable assessment would nat reaui e the level of conservatism employed here.

l l. p 21.-25

~ See ccmments provided in Attachmen. B.

Attachment B Containment Temperature/Pressure Pollcwing a Main Steam Line Break Xntroduction

'The purpose of 'vis study:s to provide. a reevaluation of dxe contaizunent conditions following a. main steam line break.

The fix'st step w'll be to reconstzuc>> the worst case containment temperature/pressure transients presented by the HRC, in Reference 1 for a large steam break. Once the Reference 1'- results have, been reproduced, the assumptions necessary to repxoduce those results can be evaluated. Xt may then be poss'ble to remove some of the conservatism and calculate a more reasonable result-Discussion The containment tempe atuxes and pressures presented in this study were calculated using Ae CO~ET-"=X/28B computer code. The xesults presented in Reference 1 were calculated using CONTZMPT-LT/028. The CONTACT--X/28B code is quite solar to the COHT~MT-LT/028, code. wi"4 changes'hich allow more user flex'bilitv.

Hot Ze o Powe Case The highes containment pressure was ca'culated in Reference to occur foz a la ge s earn line break at H"P wit'.r failuxe or one sprav train- The inpu" or this, case was run using CORP&>T-BX/28B. Piguze 1. and 2 illust=ates the esults of this run and points taken f om Reference 1. The fol'owing peak temperature and pressu e was obtained; Reference 1 Case <<5 85.8 psia 8 91 sec.

413o 9 34 sec.

COHT~T-BX/28B 83.4 ps'a 9 99.8 sec. 403.9o 8 35 sec.

While epzoduc'ng this cise fram the Refexence 'nput one inconsis encv was noted. Reference 1 states that spray was ini 'ted 35 seconds af er the setpoint at 30 psig was reached.

Xn genex'al, "'vis pressure seapoint is reached at approximately 10 sec. Therefore, spray would start at apprcxima ely 45 sec.

Sinae "he temperatu e 'se 's "e donated by sprav, the peak "emperature would ochen spx'ay s-a s. All curves in Reference 1 'llus"-ate aeak temple ature at aocroximately 35 sec. Ther ore, appea s 'eat "he Refe enc 1 ana'vs's neglec"s "he ime "o

=each "he spray setpoi.z when ac= at'ng sprav.

Using the reconstruction of Reference 1 case 05 as the base case, seveza3. cases were run to determine the sensitivity of containment temperature and pressure to various parameters.

The results of these sensitivi't9.es are 3.is ed on Table 1 and are discussed below.

Q/V, where Q is the total energy re3.eased ta the time of oeak containment pressure and is the canta'ment volume, is a parameter associated with the Tagami firn heat, t=ansfer caz'elation. This shau3.d represent total energy release to the time of oeak pressure.

A Q/V of 165 results zom the ene qy released ta containment up to the time of peak pressuze. however, a bet er aporaximation of ~he Reference 1 resu3.ts can be obtained by reducing this parameter. The effect o" reducing this parameter can be seen by camparing.

~L and 43 on Table 1. Xnczeasing Q/V results 'n 'ncreasing the ilm heat transfer coefficient. Manginq Q/V zom 87 to 165 esults in aporoxima ely -1.0 psi p essuze change and approximately -3.1 change in temperatu e

(~3 temoezature -. 400.8a 9 35.0 sec.). Therefore, the Q/V term in Tagami may. he doubled and st'l3. have

'only a smal3. ef ect an cantainme'nt temperatu e'ard.

pressure~

The Uchida.film heat transfer corzela ion has ""adit'anal3.y been used far steam breaks. When. Uchida 's used 'n the Ref. 1 Model a 3.7 psi pressure reduction ard a 15,8o temperature reduction results ( 1 versus ~4 on able 1).

~mxon Nuclear Company (='fC) mass. and energy'elease faz the most Limi"inq la qe steam line break (R f. 3) was used in the evaluat'on. The ~TC mass and ene gy was nozmali ed ta the total mass in the braken steam genezato at FZP plus the mass eleased fzcm the una fec ed steam qeneratar unt'3. main steam iso3.ation occurs. The narmalixed - fC mass and energy is illus=ated on Piqure 3 with the mass and ene gy release used in he Reference ana3.ysis. The mass associated with aux'lazy feed was nat included. The effec of auxil'ary eed on oeak conta~cent pressure and temperatu e uould be

..eql'qible since the mass added duzinq the time zame of interest" 's a very small zact'on of "he secondary s"'de inven ary (<1%) ~

The e f c o" using "he norma3.ised Zxxon mass and ene gy

=elease 's a pressure "educ 'an'" 3.3 ps'nd a tempera-

-ure zeduc='an of 17.4o (-4 and >~5 on Table 1).

A comparison between the RGB containment mode3. and the containment model used in Rex. 1 is i13.ustrated on Table 2. The ma]or di" erence between the models is the inclusion of the accumulators and ducting in the RGaE model. The area of the ducting is an assumed value based on values used by other plants,, i.e.,

Palisades ~ 20,072 9 0.10 in.

Indian Point. ~ 22,000 9 0.3.382 in.

Prairie Island ~ 22,000 6 0.1875 in Therefore, an. assumption was. made tha Ginna had 20;000 sa. ft. 9 0.10 in.

The CO~~T codes axe sensitive to node spacing.

A Large spacing will. result in Lower surface temperatures which will result in emoving too much energy zrom containment The ef ec of node spacing can be seen by comparing ~6 and >7. The effect of inclusion of accumulators and, duc ing is a3.so i13.us ated on Tab3.e L.

the process of doing this study it was determined that the. heat removal capac'tv of the fan coolers used in the Reference' analysis was the capacity of one an cooler at a service watex'temperatuxe of

'35oP. This cor esponds to maximum cooling capabi'ity

=or one cooler. The minimum capability should be used in this ana3ysis ~ The minimum capability 's assoc'ated wiD the maximum service wa er tempexature (80oP). Refexence 4 pxesents a curve of heat removal versus containment pressure and eauipnent speci"ications presents, the capaci v at. 120 and 286 The ollowing illus -ates the heat removal capacity used in'case >LL of. Table 1: The capacit'es =epresent .

the minixzm values of Refexence 4 and ecuipment specifica-tions; therefoz'e, the va}.ues a e consevrative.

containment heat removal total heat temperatuxe pe fan emoval (4 ~ans) op NBTU/hr NBTU/hr L20 286 1.575 50.0 6

200

'0 308 Q4 ~ I 2 '18.9 320 <<6 Q2 e 226 '

The exxec of us'.".g the appropriate fan cooler capac'ty can be seen by c=mpa .'ng ~' and 411 cn Table 1. h's "ep esents a 2.2 Ps"'duction in c ntainment aressure and a. 4.Io eduction in containment temaerature.

The effect of containment volume is i3.3.ust=atad on Table 1. Xnczeasing the volume by 28,000 cu. f" zasu3. s

~

in a.1.4 psi. reduction in pressure and a 1.9o reduction in temperature. Since the gross 'volume of containment is approx~taly 1.13E6; 28,000 cd. ft. reprasen s

.approximately 2.5% of the gross volume. Calculations show a, net vo3.ume of approximate3.y 1.037E6 cu. ft- Based

'on the PSAR the net volume of 9-72K'5 cu. ft. represents a conservative small volume contain'ng at 3.east 3%

mazgin. Therefore, a best estimate vo3,ume would be be"ween 9.72Z5 and 1.037E6 cu. ft. This raprasen s available margin that was not used in des study.

Pigures 4 and 5 illust=ate the effect of Uchida, ETC mass and energy, and the RGSZ'ontainment heat sink mode3. on the worst case containment response presented in Reference 3. (fit 04, 05, and 011 of Tab3.e 1).

Using the RG&E containment heat sink model (volume ~ 9.72E5),

~4C'ass and ane gy release, Uch'da correlation and fan ecole capacity of our fans r'asults, inc 72.4 ps'a 9 128.6 sec. 356.1o 9 20.3 sec.

Pull Power Case.

The highest containment tampa ature was calculated. in Rezaranca 1 to occur "or a large steam break occurring at ull power wi 'x fa'3.uze of one spray "-ain. The mass and energy release presented in Reference 3.. for this case was coupled with the Reference 1 model discussed pravious3.y and containmant tempera ura and prassu a was ca3.culatad using the CO~~T-EX/2SB cede. The following results were obtained:

Reference 1 Case ~3 75 ps'a 9 60 sec.

42lo 9 34 sec.

CO~PT-E3:/28B 73 ' psia 9 59.0 sec.

43.2.3.o 9 35.0 sac.

.he mass and ana cy ra3.ease presented 'n Reference 1 was used with the RGS.<< containmen< heat sink mode3. (volume ~ 9 '2ES) previously described, Uchida corze3.at'on, and =an cooler caaaci"y of fouz fans. This resulted '.". the fo3.low'ng peak temperature and pressure:

63-2 psia 8 51.8 sec.

374:Oo 3 32.0 sec.

~ ~

~ ~ ~ e, The temperatuze and pressure versus time is illust ated on Pigures 6 and 7 tcgethez with the reproduction of Reference

3. Case ~3 using. the CONT MT-EX/288 cade.

Sleazy In summary, the ollawing compares the Reference 1 worst case with the camparabLe worst case calculated, by RGEE as previously desc ibed:

case Reference 1 Results ..RGSE Results Steam Break - HZP 85.8 psia 9 91 sec 72.4 psia 9 128.6 sec 413a 9 34 sec 356<<1o 9 20.3 sec Steam Break - EPP 7S asia 9 60 sec 63.2 psia 9 51.8 sec 421o Q 34 sec 374.0a 9 32.0 sec Por the worst case t ansients the calcu3.ated aeak pressure

's less than the. des'gn pressure far the Ginna containment (60 ps'g). The temperature is above <<De design temperature for Ginna containment (286 p). However, the tempezatuze is exceeded only for a shor period of time.

The RGGE calculations do not accaunt "or revaparizatian, ent=ain-ment, oz: bes< estimate containment volume. Inc3.usion af these effec<<s would result in additional margin to design Limits.

Revapori ation - Reference 2 presents a discuss'cn on evaparizatian. A temperature response 's presented for a, 3.arge s earn 3.ine break us'ng he Uchida heat t=ansfez coefficient.. When revapori=at'on is used the temperature profiLe is reduced by approx~~ tely 40o. This would also esu3. in a reduction in canta'nment pressure.

2 0 --ailment - In reaL' the steam flaw~ag out of the break would not be dry s earn but would contain same moisture.

As De moisture canten of ~he steam increases, the energy assoc ated with the steam dec eases; therefore, the enercy added to containment decreases. This would resu3.t in a dec ease'n containment pressure and temperatu e.

been estimated that the decrease 'n containment p essure It has and temperature resulting f=cm accounting far e.. rainment wou3.d be similar to the decrease assoc"'atec w.' revapor"'-

"at'on.

3 ~ Can ainmen" Uo3.ume - 'As prev'aus'y desc 'bed, a bes" es-'-

mate cansai;ment. volume wauld be between 1.03756 and 9.72ES cu. = . nc easing he can ainment volume "sed 'n =he

~

ana3ys's would resu3. in a s'gh" "ressure 'decrease.

~ ~

References

l. HRC letter from D. M- Crutchfield to Z. E. Maier, Systematic Evaluation Program (SEP) for the R. E. Ginna Huclear Power Plant - Evaluation Repor on Topics VZ-2.D and VX-3," November 3, 1981.

2.. HRC let er from R. Tedesco to R. Hattson, V. Stello, and R. Boyd, "Bes< Estimate Evaluation for Environmental Qualifica-

'ion or Equipment Enside Containment Pollowing A Hain Steam Line Break," February 21, 1978-3.. Exxon Report XM-NP-77-40 Supplement 1, ."Plant Transient AnaLysis for the R; E. Ginna. Unit 1 Nuclear Power Plant',"

Harch 1980-

4. Roches er Gas and Elect 'c Corporation, R.E. Ginnai -SAR.

t ~ ~ ~ ~

TABLE 1 S~SXTXUXT STUDY Hot: Zero Power Case.

Peak P essure Peak Temperature Case (asia) (asia)

L; Ref. 3. Model, .Q/V ~ 87 83.4 9 99.8 sec. 403.9 9 35.0 sec.

2. ReZ. 1 Model, Q/V ~ 87, 83.7 9 99.4 sec 404.8 9 35.0 sec. +0'.3 + 0.9 no fan 'coolers

3. Re~. 1.Model, Q/V ~ 82.4 9 84.4 sec. 408.0 9 44.6 sec. -!.'.0 + 4;1 165, spray 9 45 sec.

4. Rex. 1 Model, Uch'da 79.7 9 100.4 sec. 388.1 8 34.8 sec. -3 ' -15.8 S. Re . 1 Model, Uchida, 76.4 9 129.4 sec. 370.7 9 32.6 sec. 3 -17.4 ENC mass/energy 6 ~ RGZ Model (Large 73.5 8 129 sec.. 359.7 9 32.8 sec. -2.9 -11.0 spacLTlg) s Uchidag

~MC mass/energy RGE Mode3. (small 75-1 I 129.2'ec. 365.4 4J 32.6 sec. +3..6' '5.7 spacing ), Uchida,

=fC mass/energy

8. same as 7 wi"5 con- 73-7 9 129.2 sec. 363.5 9 32.8 sec. -L.4 - L.9 "ainment. volume ~

'1 ~ OE6 9 same as 7 wi"4 74.9 '9'28.8 sec. 365.4 9 34.6 sec. -0.2 - 0.0 accumulat:ors 10 same as 9 wi"'x duc ing 74..6 8 L29.0 sec. 360.8 9 30.0 sec. -0.3 - 4.6

11. same as 3.0 wi"5 4 fan 72.4 e 128-6 sec. 356 ~

1 20. 29 sec. -2. 2 -4. 7 coolers

~ ~ ~ ~ ~

TABLE 2 CONTAiNMENT MODEL Reference 1 RGecZ Containment Volume. 9.72ES cu. "t. Containment Volume 9.72ES cu. ft Lxsulated Dome and 36,181 sq. same 36,181 sg.

Halls U~sulated and Walls Dome 12i474 sq. ft same .12,474 sa.

S!n!!p %falls 2,342 sq. . Same 2342 Sar ~

Sump Ploor 2>I 7

• 2,639 sc ~

Basement Ploor <<7,955 sc ~

Ca Re

'if~!

celwq Cavi y 6,400 sq. ft. same 6,400 sg. ft ~

and Floor Outs.de Refuelinq 21,800 sa. ~. 'ame 21,800 Cav ty and. S.G".'omp Operat ng Floor 9,162 sg. 2 . same 9, 162 30,962 QxtermecL'ate Ploor 6,170 sq Cw a Iran same bu' 2 X Area ~ 12,340 sq "t 1 5 in Beams 9,174. Sa. same 9,174 sa. C4o A I ~

L.O m. Beams 5,016 sa. F4,I ~ same 5,016 sa.

0'-5 in.. Beams 8,586 sa. sane 8, 586 sa ~ ~ ~

C=ane Supports 5,756 sa. Ca same 5,756 sc. ~ ~

Grat nq etc. 7, 000 SC ~

r< r

~ I 7, 000 SC ~

C4I

~

-Accumu3.ators 1> I56 SC ~

Duc"'nc 8 0.'0 ~n ~ 20,000 sc. Ck

Su==acee assumed "o 'oe in t t wish.pool i con"ac o 1

PXGVRE L CONTENT TEiM~~ATURE SM~ 3~~ HZP 1/2 SPRAY SYS~

Ref erence Recons&mcwon of Ref. 3. Case T-ze (sec) .

~ ~ ~

Re e"ence l e on w ~' f, Case $ 5.

C4 04 Cl gl

~ 4i

~0 CJ.

Co CI Yci V

Vr~

Cg c ZKuK 2 0<<

COHTAZ~iV-" PRZSSUEK X

S~c~ 3RZRZ HZP

<<C 1/2 SPHAY SYS~~C

~ 7-"ze ( sec)

/0 ~d

~ ~

' ~

PEGURE 3 MESS cLOR ST~ BREAK HZP

~

.g Z~ ~ .

%l 4, V

~ 'i4 o

a a

b 0

a

+8Fcamp~gcs C 4fass "-Xc~

~ ~ ~ i S i

~ ~

FIGURE 4 CQNTAZ~iHT TKLEBATURE ST~ S~~ HZP 1/2 SPRAY SYSTEM Reconst~c~'on of Ref. 1 Case 05 Refe ence 1, Uchida CV IV ICE (Q l Reference 1, Uchida, NC Mass/Zne cry VI

~A e

~ 5 e

Ca RGaE Model, Uchida, WC Mass/Ene~, 4 Fan Coolers X

QI CJ gj

~ ea C~

VIV X

t~ ~

~ ~ ~ ~ ~ ~

I L

~

'-:-'- --" - ~ ~

- - ~'ae N (sec~

Jd <y gd <4' gd'y id@ /i'4 ~ I'~ ~A ~ c5

~ ~ ~ ~

r

~~

Recons wc"ion o" Ref. 1 Case ~5!'-

Ref eceac'e j. Uchida

.RGa>> Nocel, Uch'da, ZHC Mass/~exp@

'4 9'an Coolers ~

Reference 1, Uchida, MC Mass/oner 4

4 0

5i>

e ZC Ct:

R'Z

<<4 4<<

X 4<<

V =:GuK 5 CONTAZ~~T PRESSURE~

ST~~ a~~ RZW 3./2 SPAZ SIST=-8

~~

Reference 1

=XGURE 6 CQNTAZ~~iT T~MZ3MUEK STEAM B~~

1/2 SPHAY SZST~i Reconst ~chion of 'Ref. 1 Case 43 RG~Z Mode1, Uchida, 4 Pan Coolers 4d q'd; Qd

~'-e'secl Cd ?0 Fd >0 lad ld JO

PZQQBZ 7 CQNTKZHiiilT PPZSSMZ STEAM 3~~ 2ZP 1/Z Seam SZST ~

Rezerence 1 Reconst uct." on or'e ". 1 Case 03 RGaZ +oce3., Uchida, 4 "-an Coolers ~

e (sec) cr~ 5~ Cd 'FO Pd pd JCO

TOPIC YI-3 SEE TOPIC YI-2.D

av

+

'+p,R Rangy'r.

P 0

g0~ STATES /p2z~ kkeL C o NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 v March 19, 1982 n

+a*++

LS05-82-03"088'NITED Docket No. 50-244 Mr, John E, Maier, Vice President Electric and Steam Production Rochester Gas 8 Electric Corporation 89 East Avenue Rochester, New York 14649

Dear Mr,

Maier:

SUBJECT; ELECTRICAL, INSTRUMENTATION AND CONTROL ASPECTS OF THE OVERRIDE OF CONTAINMENT PURGE VALVE ISOLATION (INCLUDING RESOLUTION OF SEP TOPIC VI-4, CONTAINMENT ISOLATION)

R. E. GINNA I

The staff has determined that the scope of review and evaluation performed for multi-plant generic activity B-24 addresses the electrical aspects of SEP Topic VI-4, Additional electrical review and evaluation is, therefore, not required.

Enclosed is a copy of our revised evaluation of the electrical override portion of'eneric activity 8-24 for Ginna. This assessment compares your facility as described in Docket No. 50-244, with the criteria currently used by the regulatory staff for presently operating facilities., Our report replaces that issued by my letter of January 12, 1982 and reflects the infor-provided in your letter of March 2, 1981 and your responses to IIlE 'ation Bulletin 80-06.

Itte require that by September 30, 1982, you provide physical features to augment existing administrative controls for each manual override. With regard to radiation monitoring, should further reviews of operating plants and/or additional requirements be deemed necessary, the Ginna plant will be included with that operating plant action.

Sincerely, Dennis M, Crutchfield, Chief Operating Reactors Branch No, 5 Division of Licensing Enclosures; As'tated cc w/enclosures:

See next page

Mr. John E. Maier CC Harry H. Yoigt, Esquire U. S. Environmental. Protection Agency LeBoeuf, Lamb, Leiby and MacRae Region II Office 1333 New Hampshire Avenue, N. M. ATTN: Regional Radiation Representative Suite 1100 26 Federal Plaza Mashington, D. C. 20036 New York, New York 10007 Mr. Michael Slade t Herber Gr ossman, Esq., Chairman 12 Trailwood Circle Atomic Safety and Licensing Board Rochester, New York 14618 U. S. Nuclear Regulatory Commission Mashington, D. C; 20555 Ezra Bialik Assistant Attorney General Ronald C. Haynes, Regional Administrator Environmental. Protection Bureau Nuclear Regulatory Commission, Region I New York State Department of Law Office of Inspection and Enforcement

. 2 World Trade Center 631 Park Avenue New .York, New York 40047 King of Prussia, Pennsylvania 19406 Resident Insp'ector" R. E. Ginna Plant c/o U. S. NRC 1503 Lake"Road Ontario, New York 14519 Director, Bureau of Nuclear Operations State of New York Energy. Office Agency Building 2 Empire State Plaza Albany, New York 12223 Rochester Public Library 115 South Avenue Rochester, New York 14604-Supervisor of the 'Town of Ontario Ridge Road West

'07 Ontario; New York 14519 Dr. Emmeth A. Luebke Atomic Safety and Licensing Board U. S. N'uclear Regu1ato'ry Commission Washington, D.. C. 20555 ..

Dr. Richard F. Cole Atomic Safety and Licensing Board U.. S. Nuclear Regulatory Commission Washington, D..C. 20555

REVISED SAFETY EVALUATION REPORT GINNA PLANT OVERRIDE OF CONTAINMENT PURGE ISOLATION AND OTHER ENGINEERED SAFETY E TURf ACTU TION SIGNALS INTRODUCTION As a result of Abnormal Occurrence 878-5, the NRC issued a generic letter requesting each licensee to take certain actions.

EVALUATION The enclosed revised report (EGG-EA-5723, Rev. 2) was prepared for us by EG&G, Idaho, as part of our technical assistance. program for SEP. Also, enclosed is EG&G Report 1183-4204, "Technical Evaluation of the Licensee's Response to I&E Bulletin 80-06." -These reports provide a technical'evalua-tion of the electrical,:, instrumentation and control design aspects of the override of containment purge valves isolation and other engineered safety feature actuation signals and is based upon review of these design aspects the six NRC criteria provided for the review. The technical evalua- 'gainst tion .concludes that the modifications made by the licensee at the plant have not brought the designs of the engineered safety features into confor-mance with our review criteria.

The reports identify several areas in which the present'plant does not satisfy -the review criteria. The most important design problems are that the radiation monitors have not been demonstrated to satisfy Class lE requirements and some ESF reset pushbuttons are unguarded.

Me have reviewed the licensee's justification for not modifying the Con-tainment Spray Additive Tank Discharge valves and find it acceptable on the basis that, following reset actuation, these valves close and that this would likely be the desired position. Further, the containment spray pumps remain in operation and chemical additives can be reinstated manually if required.

Me have also reviewed the licensee's justification for not modifying the Main Feedwater Isolation and Bypass'alves and find it acceptable on the basis that once the Feedwater Isolation reset has been actuated, the Feed-water Isolation and Bypass valves will not assume the position called for by their. controllers unless, the valves are in manual control, Since the plant is not operated at power levels above 15K with the valves in manual control, there is little consequence in the feedwater valves reopening.

(Reopening of the Feedwater Isolation and Bypass valves may result in the addition of eedwater to a failed steam generator. This condition would occur if the pump discharge valves fail to close or fail to remain closed and the condensate booster pumps remain in operation.)

4

CONCLUSION Sased upon our review of the consultant's technical evaluations, we conclude that the electrical, instrumentation and control design aspects of the over-ride of engineered safety feature actuation signals are acceptable, except for a lack of adequate physical protection for some of the ESF reset push-buttons. The licensee must modify such pushbuttons to provide protection against inadvertent actuation.

With regard to radiation monitoring, should further reviews of operating plants and/or additional requirements be deemed necessary, the Ginna plant will be included with that operating plant action.

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EGG-EA-5723; Rev. 2

' JANUARY 1982 SYSTEMATIC EVALUATION PROGRAM TOPIC VI-4, ELECTRICAL, INSTRUMENTATION, AND CONTROL ASPECTS OF THE OVERRIDE

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Report No. EGG-EA-5723 Rev. 2 Contract Program or Project

Title:

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Electrical, Instrumentation, and Control Systems Support for the Systematic Evaluation Program ( II) h

l Subject of this Document:

il Systematic Evaluation Program Topic YI-4, Electrical, Instrumentation, and Control Aspects of the Override of Containment Purge Valve Isolation, R. E. Ginna Nuclear.

1 Power Plant

.I Type of Document:

'Informal Report Author(s):

A. C. Udy

Date of Document:

January 1982 esponsible NRC individual and NRC Office or Division:

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Ray F. Scholl, Jr., Division of Licensing This document was prepared primarily'or preliminary or internal use. It has not received review and approval. Since there may be substantive changes, this document should 'ull not be considered final.

EG8 G Idaho, Inc.

Idaho Falls, Idaho 63416 Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C.

Under DOE Contract No. DE-AC07-76 I D01670 NRC FIN No. A6425 INTERIM REPORT

0288J SYSTEMATIC EVALUATION PROGRAM TOPIC VI-4 ELECTRICAL, INSTRUMENTATION, AND CONTROL ASPECTS OF THE OVERRIDE OF CONTAINMENT PURGE VALVE ISOLATION R. E. GINNA NUCLEAR POWER PLANT Revision 2 Docket No. 50-244 January 1982 A. C. Udy Reliability and Statistics Hranch Engineering Analysis Division EGK Idaho, Inc.

1-7-82

CONTENTS 1.00

~ IINTRO ROOUCT ION ~ ~ ~

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

2.0 EVALUATION OF THE R. E. GINNA NUCLEAR POWER PLANT ...............

2.1 Review Guidelines ...... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

2.2 Containment Ventilation Isolation Circuits Design Description ............ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~, ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ t ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

2.3 Containment Ventilation Isolation System Design Evaluation ............. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3 2.4'ther Related Engineered Safety Featur e System Circuits ~ ~ ~ 4 3 .0

SUMMARY

4 .0 .REFERENCES 111

SYSTEMATIC EVALUATION PROGRAM TOPIC VI-4 ELECTRICAL, INSTRUMENTATION, AND CONTROL ASPECTS OF THE OVERRRIDE OF CONTAINMENT PURGE VALVE ISOLATION R. E. GINNA NUCLEAR POWER PLANT

1. 0 INTRODUCTION Based on the information supplied by the Rochester Gas and Electric Company (RGE) this report addresses the electrical, instrumentation, and control system design aspects of the Containment Ventilation Isolation (CVI) system and other r elated Engineered Safety Feature (ESF) functions for the Ginna plant.

Several instances have been reported where the automatic closure of the containment ventilatidn or purge isolation valves would not have occurred because the safety actuation signals were manually overridden or blocked dur-ing normal plant pperations. Lack of proper management controls, procedural inadequacies, and; circuit design deficiencies contributed to these instances.

These events also brought into question the mechanical operability of the valves themselves. These events were determined by the Nuclear Regulatory Commission (NRC) to be an Abnormal Occurrence (878-05) and accordingly, were reported to Congress.

'he NRC is now reviewing the electrical. override aspects of containment purging and venting for all operating reactors. On November 28, 197(, the NRC issued a letter, "Containment Purging During Normal Plant Operation"'o all Boiling Water Reactor and Pressurized Water Reactor licensees, which required a reyien of these systems by the licensee. RGE responded on February 16, 1979, March 30, 1979, and March 17, 19805. The Final Safefy Analysis Report (FSAR) and Westinghouse Drawing No. 882D612, Sheet 6, also contain design informatioIa reviewed for this reIIort. RGE letters of March 2, 1981, November 19, 1979 and December 1, 1981 also contain information on the control systems that was reviewed for this report.

2.0 EVALUATION'OF THE R. E. GINNA NUCLEAR POWER PLANT 2.1 Review Guidelines. The intent of this evaluation is to determine if the actuating signals for the ESF equipment meet the following NRC criteria:

1. Guideline No . 1--In keeping with the requirements of Gen-eral Design Criteria 55 and 56, the overridea of one

a. The following definitions are given for clarity of use in this evaluation:

Override: the signal is still present, and it is blocked in order to perform a function contrary to the signal.

Reset: the signal has come and gone, and the circuit is being cleared in order to return it to the normal condition.

1. High containment radiation

2. Safety injection signal (high containment pressure can initiate a safety injection signal).

RGE has indicated that these signals a~e derived from equipment "designed and constructed as a Class 1E system." However, the radiation channels have not been shown to be Class lE.

These eight valves (except for the radiation monitor valves) are air-operated butterfly valves and are used. so that one is redundant for another on the same air line. Valve position lights show the actual valve posi-tion. The solenoid valves fail closed on loss of air or on loss of power.

The radiation monitor valves are air-operated diaphragm valves which have.

either a check valve or a manual valve for redundancy.

The logic of the containment isolation and the CVI valves is shown in reference 6. In both systems, the manual actuation is overridden along with the automatic actuation signals by operation of a reset switch (one per safeguards train). This logic has since been modified as outlined below.

As a result of the short-term lessons learned, the,CVI valve control circuits have been modified to provide individual resetting of each isola-tion valve. Resetting a valve after automatic closure now requires opera-tion of a key-locked reset switch and a valve reset (guarded) pushbutton switch. -The valve then goes to the position the valve control circuit requires. Administrative procedure requires the valve controller to be in the closed position before resetting the valve logic.

2.3 Containment Ventilation Isolation S stem Desi n Evaluation.

Guideline requires that no signa override can prevent another safety actuation signal from functioning. Ginna has override provision in the reset switches. The circuits involved have been modified to comply with this guideline.

Guideline 2 requires that reset and override switches have physical provisions to aid iii the administrative control of these switches. The reset switches're keylocked. 'he individual valve reset switches are guarded. This guideline is satisfied; Guideline 3 requires system level annunciation whenever an override affects the performance of a safety system. The literal intent of this guideline is not satisfied by the Ginna design; however, individual status lights monitor the status of each individual valve override. Thus, oper-ators will be aware of the status of any overrides.

Guideline 4 requires that isolation of the CVI valves be actuated by several diverse signals. This criterion is'et in that:

1. Safety injection will initiate isolation.

2. High pressure in the reactor building will initiate safety in jection.

4. 0 REFERENCES

1. NRC/DOR letter, A. Schwencer, to RGE and all.BWR and PWR licensees, "Containment Purging During Normal Plant Operation," dated November 28, 1978.

2. RGE letter, L.'. White, Jr., to Director of Nuclear Reactor Regula-tion, U.S. NRC, "Containment Purging During Normal Plant Operations,"

, January 2, 1979.

3. ~ RGE letter, L. O. White, Jr., to Director of Nuclear Reactor Regula-tion, U.S. NRC, "Review of Safety Actuation Circuits with Overrides,"

February 16, 1979.

4. RGE letter, L. D. White, Jr., to Director of Nuclear Reactor Regula-tion, U.S. NRC, "Review of Safety Actuation Circuits with Overrides,"

March 30, 1979.

5. RGE letter, L. O. White, Jr., to Director of Nuclear Reactor Regula-tion, U.S. NRC, "SEP Topic VI-4, Containment Isolation System,"

March 17, 1980.

6. Drawing, Westinghouse Logic Diagram No. 8820612, Sheet 6, Revision 7, "Safeguards Actuation Signals."

7. RGE letter, J. E. Maier to Director of Nuclear Reactor Regulation, NRC, "SEP Topic VI-4, Containment Isolation (Purge Valve Reset),"

March 2, 1981.

8. RGE letter, L. D. White to Director of Nuclear "Reactor Regulation, NRC, "Discussion of Lessons Learned Short Term Requirements,"

November 19, 1979.

9. RGE letter, L. D. White to Director of Nuclear Reactor Regulation, NRC, "SEP Topic VI-4,'ontainment Isolation (Electrical)," December 1, 1981.

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NRC TAC No.

Report No. EGG 1183-4204 Contract Program or Project

Title:

Electrical, Instrumentation, and Control System Support Subject of this Document:

Technical Evaluation of the Licensee's Response to IIEE Bulletin 80-06 Concerning ESF Reset Controls for the R. E. Ginna Nuclear Power Plant, Unit 1 Type of Document:

Informal Report Auth or(s):

D. H. Laudenbach Date of Document:

May 1381 Responsible NRC individual.and NRC Office or Division:

P. Bender/R. Wilson, ICSB This document was prepared primarily for preliminary or internal use. it has not received full review and approval,. Since there may be substantive changes, this document should not be cons>dered final.

EGKG Energy Measurements Group San Ramon Operations San Ramon, CA 94583 Prepared ,or the U.S. Huclear Regulatory Commission Washington, O.C.

'Jnder QOE Contract Ho. BZ R 201904031 HRC F iH Ho. A 0250 INTERIM REPORT

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0 EGG 1183-i Energy Meoaurementa Group May .

San Ramon Operations TECHNICAL EVALUATION OF THE LICENSEE'S RESPONSE TO l8cE BULLETIN 80-06 CONCERNING ESF RESET CONTROLS FOR THE R. E. GINNA NUCLEAR POWER PLANT, UNIT 1 (DOCKET NO. 50-244) by O. H. Laudenbach Approved .for Publication t

J. R. Raaosevic Oepartment Manager I

This document is UNCLASSIFIED Oerivative Classifier: 7 ilicno a~~. roaericx Oepartment Manager Work Performed for Lawrence Liverfnore National Labor tory under V.S. Department of Energy Contract No. DE-ICOS-76 NVO 0" S3.

I HTRODUCT IOH On March 13, 1980, the USHRC Office of Inspection and Enforcement

( IBE), issued 18E Bulletin 80-06, entitled ".Engineered .Safety Feature .(ESF)

Reset Controls," to all PMR and BQR facilities with operating licenses.

IhE Bulletin 80-06 requested that the following actions be taken 'by the licensees:

(1) Review the drawings for all systems serving safety-related functions at the schematic/elementary diagram level to determine whether or not upon the reset of an ESF actuation signal all associated safety-related equipment remains -in its emergency mode.

(2) Yerify that the actual installed instrumentation and controls at the facility are consistent with the

.i schematics reviewed in Item 1 above by conducting a

'test to demonstrate that all equipment remains in its emergency mode upon removal of the actuating signal and/or manual resetting of. the various isolating or actuation signals. Provide a schedule for the per-formance of the testing in your response to this bulletin.

(3) If any safety-related equipment does not remain in its emergency mode upon reset of an ESF signal at your facility, describe proposed system modification, design change, or other corrective action planned to resolve the problem.

(4) Report in writing within 90 days the results of your review, include a lis of all devices which respond as discussed in Item 3 above, actions taken or planned to assure adequate equipment control, and a schedule for implementation of corrective action.

C This technical evaluation addresses the licensee's response to IhE Bulletin 80-06 and he licensee's proposed system modification, design cn'ange, and/or other correcti ve action planned to resolve the problem. In evaluating the licensee's response to the four Action Item requirements of the bulletin, the following HRC staff guidanc is also used:

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Upon the reset of ESF signals, all safety-related equipment shall remain in 'its emergency mode.' Multiple reset sequencing shall not cause the affected equipment to deviate from i: s emergency mode. Justification should be provided for any exceptions.

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C EVALUATION AHD COHCLUS IOHS if'n a letter dated June 3, 1980 (Ref. I], Rochester Gas "and Elec-tric Corporation, the licensee for R. E. Ginna Nuclear Power Plant, Unit 1, replied to I&E Bulletin 80-06. In a telephone conference call conducted on Harch 3, 1981 )Ref. 2], the licensee provided additional information and clarification to their written response.

The licensee reported [Ref. 1] that a drawing review has been completed at Ginna station for all systems serving safety-related functions. This review was conducted at the schematic level to determine

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1 h th ll. associated safety-related equipment would r main in its emer-gency mode upon the reset of an engineered safety feature actuation si'g nal.

The licensee i'dentified [Ref. 1] the following equipment as not remaining in the emergency mode'pon FSF reset:.

1. Containment Spray additive tank discharge valves.

2. > Hain Feedwater isolation and bypass valves, 1

Me conclude that the licensee has complied with the requirements of Action conc Items 1 and 4 of I&E .Bulletin 80-06 by completing the drawing review o al systems serving safety-related functions and by identifying the devices that do not remain in their emergency mode upon ESF reset.

The licensee reported )Ref. 1] that testing to verify that actual installed instrumentation and controls were consistent with the schematics reviewed was completed during the Nay 1980 refueling outage. Me conclude that the licensee has complied with the requirements of Action Item 2 of I&E Bulletin 80-06 by providing a schedule and completion date for the performance of testing.

The licensee indicated {'.Ref. 1] that no modifications or design changes were planned for the Containment Spray additive tank discharge valves nor for the .Hain Feeedwater isolation and bypass valves. The licensee offered justification )Ref. 1] for not modifying these devices and also provided )Ref. 2] a verbal explanation to enhance the justification offered in reference 1.

The licensee offered )Ref. 1] the following justification for not modifying the Containment Spray additi've tank discharge valves:

The Containment Spray circuit has a reset switch which gives the operator the means of resetting containment spray. Once the reset switch has been actuated, the spray additive tank discharge valves will return automatically to the position called for by their .controllers. The containment spray pumps and their discharge valves would require operator action to change state. This capability is necessary so tne operator has the flexibility in. dealing with pos--accident conditions within containment (i.e., LOCA or steam line break).

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The liconsee offered LRef. 2] the following additional justifi-cation or not modifying the Containment Spray additive tank discharge valves:

The valves associated with 'the spray 'additive tank 'will be opened automatically two minutes after the containment spray signal's actuated. The s'odium hydroxide will flow due to the suction of the spray pumps and mix with refueling water prior to being discharged through the SPray nozzle into the containment. After the containment spray signal is actuat-ed, the operator has the capabi'lity to stop the timer has been detemined that actuation of the sodium hydroxide if it addition is not warranted. The operator also has the cap-ability to reinstate the sodium hydroxide addition, if required. Emergency procedures set forth guidelines for this action based on one or more of the following:

(1). High containment pressure in combination with a total loss of RCS pressure.

I (2) High radiation levels in combination with elevated containme'nt pressure.

(3) Pressure signals indicative of accumulator dis-charge into the RCS.

The licensee offered t.Ref. 1] the following justification for not modifying the Main Feedwater isolation and bypass valves:

The Feedwater Isolation circuit has a .reset switch which gives the operator the means of resetting the isolation signal to the feedwater bypass valves. Once the reset switch is actuated, the feedwater bypass valves will a'ssume the position called for by thei r controllers. The main feedwater valves wi 11 remain closed until the isolation logic clears, and then they will automatically assume the position called for by their controllers. lt should be noted that a safety injection signal also causes the main feedwater pumps to be tripped and their discharae valves to automatically close; -therefore, closing 'the main fe dwatef valves on a safety injection signal is redundant.

The licensee offered {Ref. 2] the following additional justifi-cation for not modifying the Hain Feedwater isolation and bypass valves:

While reset. will result in the feedwater isolation valves returning to their demand position, reset does not affect the status of the feedwater pumps or the pump discharge valves. Thus, re-openina of the feedwater isolation (and bypass) valves would not result in the addition of fe dwater to the steam generator via the fe dwater lines.

The above justifications were offered by the license in lieu of any .system modification, design change, or other corrective action. We have reviewed the justifications submitted by the. licensee to insure that sufficient information is provided as a basis for the NRC staff to prepare a Safety Evaluation Report.

F INOINGS Based on our review of the information and documents provided by the license , we find that, the ESF reset controls for R. E. Ginna Nuclear Power Plant, Unit 1, satisfy the requirements of Action Items 1, 2, and 4 of IKE Bulletin 80-06.

In response to Action Item 3 of IhE Bulletin 80-06, the licensee identified several valves as not remaining in their emergency mode upon FSF

'eset and offered justification in lieu of any system modification, design change, or other corrective action.

REFERENCES

1. Rochester Gas and Electric Corporation letter (L.O. White, Jr.) to NRC 18E (B.H. Grier), "Response to IhE Bulletin 80-06," dated June 3, 1980.

2. Telephone conference call, NRC (P. Be'nder); Rochester Gas and Electric Corporation (R. HcCready, G. Oaniels); EG5G San Ramon (0. Hackett, 0.

Laudenbach), March 3, 1981.

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Operating Reactors Branch No. 5 U.S. Nuclear Regulatory Commission Washington,- D.C. 20555

Subject:

SEP Topics VI-2.D and VI-3

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R. E. Ginna Nuclear Power Plant Docket No. 50-244

Dear Mr. Crutchfield:

This letter is in response to the draft evaluation of SEP Topic VI-2.D, "Mass and Energy Release Inside Containment" and SEP Topic VI-3, "Containment Pressure and Heat Removal Capability,"

which was transmit ed by your. letter dated November 3,-1981.

We have reviewed the draft evaluation and hive identified several conservatisms,in the analysis for the loss of coolant accident their impact on the L'OCA results, asqualitative (LOCA). These conservatisms and a discussion of well as a number of general comments on the evaluation, are dicussed in Attachment A. We also identified a number of conservatisms in the analysis foi the main steam line break (MSLB) . Because of the degree of conservatisms in the NRC evaluation, we performed a sensitivity studv with the code CONTEMPT-EI/28B. This code is very similar to CONTEMPT'-LT028, as discussed in Attachment B. Therefore, while this code has not been completely qualified for use as a licensing code, we believe'that it is accurate and adequately represents Ginna. The sensitivity study, presented in Attachment B to this letter, confirmed the conservatism of the NRC results or MSLB.

Regarding the LOCA analysis, since the Ginna design basis pressure envelopes the NRC results, we conclude that the Ginna design basis pressure profile remains acceptable. The Ginna design bas's temperature profile exceeds the NRC results except between 10,000 seconds and approximately 20,000 seconds after the" design basis event. We propose that the Ginna design basis temperature profile remain as shown in the FSAR for times- less than 10, 000 seconds and be revised beyond 10, 000 seconds as ollows:

.from 10,000 seconds to'0,000 seconds, temperature = 250oF, beyond

. 20,000 seconds, temperature < 100oF. Since the containment tem-perature is already decreasing at this time, 'it is not considered 8 C QSQ 'R 820205 P DR ADCCN. OSCOO244

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~CEESTER GAS AHD ELECTRIC CORP. SHK'T NOo 2

'DATE February 1', 1982 TO Mr. Dennis M. Crutchfield that this revised profile would invalidate any conclusi.ons drawn as part of our review of environmental qualification of electrical equipment, since the affected equipment temperatures would also be decreasing at this time from their peak temperatures.

Regarding the NRC steam break analyses, the NRC results .

for pressure and temperature exceed the Ginna design basis as revised above. However, as shown in Attachment B, acceptable results for containment pressure are obtained when more reasonable assumptions are used. It is our conclusion, 'based on Attachment B, that the Ginna design basis pressure profile exceeds the pressure profile calculated for main steam line break and, therefore, remains acceptable. We conclude that the temperature resulting from a steam line break in containment may exceed the Ginna temperature profile, but that this is of no'onsequence design.'asis due to the short duration of this exceedance and may therefore be. ignored. This conclusion is based on NRC guidance provided in the Division of Operating Reactors (DOR) Guideli.nes which in turn is based, for example, on the short durati.on of the t'em-perature spike, lower heat transfer coefficient, and the elevation of the steam lines relative to equipment. Thus, we conclude that the Ginna design basis temperature profile, as revised based on LOCA results discussed above, remains valid.

Very truly yours, Zo n E. Maier

~

0 ATTACHMENT A CO%CENTS ON NRC SAFETY EVALUATION

p. 5 Containment conditions resulting from a main steam break weie assessed in a gross fashion in FSAR page 14.2.5-10.

That analysis does not, however, comply with current criteria.

p. 5 The LOCA analysis underestimated'he effectiveness of the containment fan coolers by a heat removal rate of approximately 25 to 30 million BTU per hour. This resulted from an incorrect set'f data being provided by RG&E to the NRC consultant (see also LER 81-022 transmitted by latter dated January 4, 1982 from John E. Maier, RG&E, to Ronald C. Haynes, NRC) .

We have estimated the impact of correcting the fan cooler heat removal rate to be on the order of a 1 to 2 psi reduction in peak containment pressure. This, therefore, is an additional conservation in the analysis.

p. 6 .The SER discussion of the result'of a main steam line break should be revised to reflect the sensitivity study presented in Attachment B;

p. 7 The last paragraph of the SER should be revised to reflect the conclusions presented in our letter transmitting this attachment.

p.15 See our comment, above regarding FSAR Section 14.2.5.

'.l6 The assumption of' constant containment pressure of 14.7 psia throughout the transient will result in an overestimate of the mass and energy release and, therefore, an overestimate of containment pressure and temperature.

p.16 All information provided in conversations was also available on the docket.

p.18 It is our understanding that only accumulator water was (conservatively) set at 272.9 F, not all ZCCS flow. This should be clarified. If our understanding is not correct, then Table 1, which indicates RWST temperature, should be revised.

po21 The containment design pressure is incorrectly stated at the top of the page as being 74 psia; it is 75 psia.

lo General A number of other conservatisms are discusSed in the LOCA-evaluation. A more reasonable assessment would not require the level of conservatism employed here.

l l. p 21;25

~ See comments provided in Attachment B.

~ ~ I Attachment B Containment Temperature/Pressure Pollowing a Main Steam Line Break Introduction

'The purpose of this study is to provide a reevaluation of the containment conditions following a main steam line break.

The first step will be to reconstruct the worst case containment temperature/pressure transients presented by the NRC. in Reference 1 for a large steam break. Once the Reference 1'- results have been reproduced, the assumptions necessary to reproduce those results can be evaluated. Xt may then be possible to remove "some of the conservatism and calculate a more reasonable result.

Discussion The containment temperatures and pressures presented in this study were calculated using the CONTEMPT-EX/28B computer code. The results presented in Reference 1 were calculated using CONTEMPT-LT/028. The CONTEMPT-EI/28B code is quite similar to the .CONTEMPT-LT/028 code.'with changes which allow more user flexibility.

Hot Zero Power Case The highest containment pressure was calculated in Reference 1 to occur for a large steam line break at HZP with failure of one spray train- The input for this case was run using CONTEMPT-EI/28B. Figure 1 and 2 illustrates the results of this run and points taken from Reference 1. The following peak temperature and pressure was obtained'.

Reference 1 Case <5 85.8 psia 8 91 sec.

413o 9 34 sec.

CONTEMPT-EI/28B 83.4 psia 9 99.8 sec. 403.9o 9 35 sec.

Nhi3.e reproducing this case from the Reference 1. input.

one inconsistency was noted. Reference 1 states that spray was initiated 35 seconds after the setpoint at 30 psig was reached.

Xn general, .this pressure setpoint is reached at approximately 10 sec. Therefore, spray would start at approximately 45 sec.

Since the temperature rise is:terminated by spray, the peak temperature would occur when spray starts. All curves in Reference 1 illust ate peak temperature at approximately 35 sec. Therefore, i" appears that the Reference 1 analysis neglects "he time to reach the spray setpoint when actuating spray.

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Using the reconstruction of Reference 1 case f5 as the base case, several cases were run to determine the sensitivity of containment temperature and pressure to various parameters.

The results of these sensitivities are listed on Table 1 and are discussed below.

Q/V, where Q is the total energy released to the time of peak containment pressure and is the containment volume, is,a parameter associated with the Tagami film heat transfer correlation. This should represent total energy

" release to the time of peak pressure.

A Q/V of 165 results from the energy released to containment up- to the time of peak pressure. However/

a better approximation of the Reference 1 results can be obtained by reducing this parameter. The effect of reducing this parameter can be seen by comparing fl and 43 on Table 1. Encreasing Q/V results in increasing the film heat transfer coe fficient. Changing Q/V from 87 to 165 results in approximately -1.0 psi pressure change and approximately -3.1 change in temperature

( 3 temperature . 400.8 9 35.0 sec.). Therefore, the Q/V term in Tagami may. be doubled and still

'.only a small effect on containment temperature 'and.

have pressure.

The Uchida .film heat transfer correlation has traditionally been used for steam breaks. When Uchida is used in the Ref. 1 Model a 3.7 psi pressure reduction and a 15,8o temperature reduction results (~1 versus ~4 on Table 1).

Exxon Nuclear Company (ENC) mass and energy release

,for the most limiting large steam line break (Ref. 3) was used in the evaluation. The ENC mass'nd energy was normalized to the total mass in the broken steam generator at HZP plus the mass released from the unaffected steam generator until main steam isolation occurs. The normalized ENC mass and energy is illustrated on Figure 3 with the mass and energy release used in the Reference 1 analysis. The mass associated with auxilary feed was not included. The effect of auxiliary feed on peak containment pressure and temperature would be negligible since the mass added during the time frame of interest is a very small fraction of the secondary side inventory (<1%).

The effect of using the normalized Exxon mass and energy release is a pressure reduction of 3.3 psi and a tempera-ture reduction of 17.4o (-.4 and ~5 on Table 1).

A, comparison between the RGRE containment model and the containment model used in Ref. 1 is illustrated on Table 2. The major difference between the models is the inclusion of the accumulators and dhcting in the RGSE model- The area of the ducting is an assumed value based on values used by othe'r plants, i.e.,

Palisades ~ 20,072 9 0.10 in.

Indian Point = 22,000 9 0.1382 in.

Prairie Island ~ 22,000 9 0.1875 in.

Therefore, an. assumption was made that. Ginna had 20;000 sa. ft. 9 0.10 in.

The CONTEMPT codes are sensitive to node spacing.

K large spacing will result in lower surface temperatures which will result in removing too much energy from containment. The effect of node spacing can be seen by comparing >6 and ~7. The effect of inclusion of accumulators and ducting is also illustrated on Table 1.

In the process of doing this study it was determined.

that the heat removal capacity of the fan coolers used in the Reference 1 analysis was the capacity of one fan cooler at a service water temperature of

'35 P. This corresponds to maximum cooling capability for one cooler. The minimum capability should be used in this analysis. The minimum capability '

associated with the maximum service water temperature (80 0 P). Reference 4 presents a curve of heat zemova3; versus containment pressure and eauipment specifications presents the capacity at 120 and 286 The following illustrates the heat removal capacity used in case Ill of. Table 1: The capacit's represent .

the minimum values of Reference 4 and eauipment specifica-tions; therefore, the values are conservative.

containment hea removal total heat temperature per fan removal (4 fans) op MBTU/hr MBTU/hr 120 1.575 50 '

6 200

'0 286 308 54.72 '18.9 320 56.52 226.0 The effect of using the appropriate fan cooler capacity can be seen by comparing 410 and ~11 on Table 1. This represents a 2.2 psi reduction in containment pressure and a 4.7o reduction in containment temperature.

The effect of containment volume is illustrated on Table 1. Increasing the volume by 28,000 cu. ft. results in a 1.4 psi reduction in pressure and a 1.9o reduction in temperature. Since the gross volume of containment is approximately 1.13E6; 28,000 cd. ft. represents

.approximately 2.58 of the gross volume. Calculations-show a net volume of approximately 1.037E6 cu. ft. Based on the FSAR the net volume of 9.72E5 cu. ft. represents a conservative small volume containing at least 3%

margin. Therefore, a best estimate volume would be between 9.72E5 and 1.037E6 cu. ft. This represents available I margin that was not used in this study.

Figures 4 and 5 illustrate the effect of Uchida, ENC mass and energy, and the RGRE containment heat sink model on the worst case containment response presented in Reference 1 (<1,

~4, ~5, and ~ll of Table 1).

Using the RG&E containment heat sink model (volume = 9.72E5),

ENC mass and energy release, Uchida correlation and fan cooler capacity of four fans z'esults, in: .

72.4 psia 9 128 ' scca 356.1o 9 20.3 sec.

Full Power Case The highest containment temperature was calculated in Reference 1 to occur for a large steam break occurring at full power with failure of one spray tz'ain. The mass and energy release presented in Rezerence 1 for this case was coupled with the Reference 1 model discussed previously and containment temperature and pzessure was calculated using the CONTEMPT-EI/2SB code. The following results were obtained:

Reference 1 Case <3 75 psia 9 60 sec.

23o g 34 sec.

CONTEMPT-EX/28B 73.3 psia 9 59.0 sec. 412.1o 9 35.0 sec.

The mass and energy release presented in Rezerence 1 was used with the RG&E containment heat sink model (volume = 9.72E5) previously described, Uchida correlation, and fan cooler capacity of four fans'his resulted 'n the following peak temperature and pressure:

63.2 psia 9 51.8 sec. 374.0o 9 32.0 sec.

• ca<<'I The temperature and pressure versus time is illustrated on Figures 6 and 7 together with the reproduction of Reference 1 Case f3 using. the CONTEMPT-EI/28B code.

S~ary In summary, the following compares the Reference 1 worst case with the comparable worst case calculated by RG&E as previous'ly described:

case Reference 1 Results '..RGRE Results Steam Break - HZPI 85.8 psia 9 91 sec 72.4 psia 9 128.6 sec 413 9 34 sec 356.1o 9 20.3 sec Steam Break HFP 75 psia 9 60 sec 63.2 psia.9 51.8 sec 421o 9 34 sec 374.0 9 32.0 sec For the worst case transients the calculated peak pressure

. is less than the design pressure for the Ginna containment (60 psig). The tempe'rature is above the design temperature for .

Ginna containment (286 F). However, the temperature is exceeded only for a short period of time.

The RGSE calculations do not account for revaporization, entrain-ment, or best estimate containment volume. Inclusion of these effects would result in additional margin to design limits.

1. Revaporization - Reference 2 presents a discussion on revaporization. A temperature response is presented for a large steam line break using the Uchida heat transfer.

coefficient. Nhen revaporization is used the temperature profile is reduced by approximately 40o. This would also result in a reduction in containment pressure.

2 ~ Entrainment - In reality the steam flowing out. of the break would not be dry steam but would contain some moisture.

As the moisture content of the steam increases, the energy associated with the steam decreases; therefore, the energy added to containment decreases. This would result in a decrease in containment pressure and temperature.

been estimated that the decrease in containment pressure It has and temperature resulting from accounting for entrainment would be similar to the decrease associated with revapori-zation.

3. Containment Volume - As previous'y described, a best es mate containment volume wo'uld be between 1.037E6 and 9 '2E5 i-cu. ft ~ Increasing the containment volume used in the

~

analysis would result 'n a slight pressure decrease.

~~ I

References 1 ~ NRC letter from D. M. Crutchfield to J. E. Maier, "Systematic Evaluation Program (SEP) for the R. E. Ginna Nuclear Power

~

Plant Evaluation Report on Topics VI-2.D and Vl.-3," November 3, 1981.

2.. NRC letter from R. Tedesco to R. Mattson, V. Stello, and R..Boyd, "Best Estimate Evaluation for Environmental Qualifica- .

tion of Equipment Xnside Containment Following A Hain Steam Line Break," February 21, 1978.

3 ~ Exxon Report XH-NP-77-40 Supplement 1, "Plant Transient Analysis for the R. E. Ginna Unit 1 Nuclear Power Plant,"

March 1980."

4 ~ Rochester Gas and Electric Corporation,'.E. Ginna, PSAR.

~ ~

TABLE 1 SENSITIVITY STUDY Hot Zero Power Case Peak Pressure Peak Temperature Case ( sia) ( sia) 1'ef. 1 Model, .Q/V = 87 83.4 9 99.8 sec. 403.9 9 35.0 sec.

2. Ref. 1 Model, Q/V ~ 87, 83.7 9 99.4 sec. 404.8 9 35.0 sec. +0'.3 + 0' no fan 'coolers

3. Ref. 1 Model, Q/V ~ 82.4 9 84.4 sec. 408.0 9 44.6 sec. -1'.0 + 4'1 165, spray 9 45 sec.

4~ Ref. 1 Model, Uchida 79.7 8 100.4 sec. 388.1 9 34.8 sec.. -3.7 -15.8

5. Ref. 1 Model, Uchida, 76.4 9 129.4 sec. 370.7 9 32.6 sec. -3.3 -17.4 ENC mass/energy 6 ~ RGE Model (large 73.5 9 129 sec. 359.7 8 32.8 sec. -2.9 -11.0 spacing), Uchida, ENC mass/energy

\ /

7 RGE Model (small 75.1 8 129.2'ec. 365.4 9 32.6 sec. +1.6 + '5.7 spacing), Uchida, ENC mass/energy

8. same as 7 with con- 73.7' 129.2 sec. 363.5 9 32.8 sec. -1.4 - 1.9 tainment volume ~

1.0E6

9. same as 7 with 74.9 8 128.8 sec. 365.4 9 34.6 sec. -0.2 0.0 accumulators

10. same as 9 with ducting 74.6 9 129.0 sec. 360.8 9 30.0 sec. -0.3 - 4.6

11. same as 10 with 4 fan 72.4 9 128.6 sec. 356.1 9 20.29 sec. -2.2 -4.7 coolers

~ ~ all I ~

~ TABLE 2 MODEL

'ONTAINMENT Reference 1 RG&E Containment Volume. 9. 72E5 cu. ft. Containment Volume 9.72E5 cu. ft.

Insulated Walls Dome and 36,181 sq. ft. same 36,181 sq. ft.

Uninsulated and Walls Dome 12,474 sq. ft. same .12,474 sq. ft.

Sump Walls 2,342 sq. ft,. same 2,342 sa. ft.

Sump Floor 297

• 2,639 sq. ft.

Basement Floor <<7,955 sq. ft..

Refueling Cavity 6,400 sq. ft. same 6,400 sq. ft.

Wall and Floor I

" Outside Refueling 21,800 sq. ft. 'ame 21,800 Cavity and S.G.'ompo Operating Floor 9,162 sq. ft.. same 9,162 30,962 Intermediate Floor 6,170 sq. ft. same but 2 X Area = 12,340 sq. ft.

1.5 in. Beams 9,174 sq. ft. same 9,174 sq. ft.

1.0 in. Beams 5,016 sq. ft,. same 5,016 sq. ft.

0. 5 in. Beams 8,586 sq. ft. same 8, 586 s'qt fto Crane Supports 5,756 sq. ft. 5,756 sq. ft.

Grating etc. 7,000 sq. ft. same 7,000 sa. ft.

Accumulators 1,756 sa. ft.

Ducting 8 0.10 in. 20,000 sq. ft.

Sur ace assumed to be in contact with pool water.

r

~ ~ '

~ ~ i i,) ~

FIGURE 1 CONTAINMENT TEMPERATURE STEAM BREAK HZP 1/2 SPRAY SYSTEM Reference

~Q

'i

Qi Reconstructio n of Ref. 1 Case ~

O O

~g

.. 8 Q.

0

~a O4 f4 ~

X4 O4 t4 O Time (sec) yo yo m co . ro s o so roo'l< ii~ rgb

~,Ie 4. c'

~ 's ~ ) ~

efere nce

~ ~

l.

Recon f, .Case 05 e

Xe FIGURE 2 CONTAINMENT PRESSURE STEAM BREAK HZP 1/2 SPRAY SYSTEM

~

~

t PXGURE 3 MASS FLOW STEAM BREAK HZP

't 't, ~

~ V X 0 Reference l ENC Mass Flow

~ M Time (sec) ~ I g lo ag go siO rg dc> pu go qo pg uo Z=o /'& /e-

I

~ ~ ~

C'IGURE

{

CONTAZNMENT TEMPERATURE STEAM BREAK HZP 1/2 SPRAY SYSTEM 4

Pl ~ ' \

'0 Reconstruction of Ref. 1 Case $5 f~

f I

~

Reference 1, Uchida Ol CV Q7 Reference 1, Uchida, ENC Mass/Energy

~ RGGE Model, Uchida, ENC Mass/Energy, 4 Fan Coolers P ' '

~

]

' I ~ ~

I

~ ~ I ~ ~ ~ ~ ~ ~ ~

I Time (sec) g qg pd Jd pO gd 4'u

~ per Far yy /dO //C' /~

~ ~

~dl, h A 4.

d.

~ ~

~ \~ '

~,,

Recons truct ion o f Ref 1 Case ~ 5 <"-

Reference 1, Uchida C4 Cl<

~ H tD

.RG&E Model, Uchida, ENC Mass/Energy y 4 Fan Coolers )

Re ference 1, Uchida, ENC Mass/Ener O

Z~

r g

O; zoIJ ld IA 0

~<<

O big ll X

O~

vY FIGURE 5 CONTAINMENT PRESSURE STEAM BREAK HZP 1/2 SPRAY SYSTEM 1 ' ' ~

P I~ ~ ~ ~

cJ g' (Time (sec) 4 /dG'/d / "4 /Zd r'5Q P

~ ~

Reference 1 FIGURE 6 CONTAINMENT TEMPERATURE STEAM BREAK HFP 1/2 SPRAY SYSTEM Recons truction of 'Ref. 1 C ase 43 RG&E Model, Uchida, 4 F an Coolers Time'sec) l Zo iso. ~o co ro Fo

. PO r<d

FZGURE 7 CONTAINMENT PRESSURE STEAM BREAK HFP 1/2 SPRAY SYSTEM Reference 1 Reconstruction of Ref. 1 Case <

RG&E Model, Uchida 4 Fan Coolers

<a R

jc O+

W t T~e (sec) a to zo Za Yd rc Co ro ro. ro cdC

f ff

"~

February 1, 1982 Director of Nuclear Reactor Regulation Attention: Mr. Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

SEP Topics VZ-2 AD and VZ-3 R. E. Ginna Nuclear Power Plant lJ~

~ I $ ~

Docket No. 50-244

Dear Mr. Crutchfield:

This letter is in response to the draft evaluation of SEP Topic VZ-2.D, "Mass and Energy Release Znside Containment" and SEP Topic VZ-3, "Containment Pressure and Heat Removal Capability,"

which was transmitted by your letter dated November 3,. 1981.

We have reviewed'the draft evaluation and have identified several conservatisms,in the analysis for the loss of coolant accident (LOCA). These conservatisms and a qualitative discussion of their impact on the LOCA results, as well as a number of general

~ comments on the evaluation, are dicussed in Attachment A. We also identified a number of conservatisms in the analysis for the main steam line break (MSLB) . Because of the degree of conservatisms in the NRC evaluation, we performed a sensitivity study with the code CONTEMPT-EZ/28B. This code is very similar to CONTEMPT'-LT028, as discussed in Attachment B. Therefore, while this code has not been completely qualified for use as a licensing code, we believe'that it is accurate and adequately represents Ginna. The sensitivity study, presented in Attachment B to this letter, confirmed the conservatism of the NRC results for MSLB.

Regarding the LOCA, analysis, since the Ginna design basis pressure envelopes the NRC results, we conclude that the Ginna design basis pressure profile remains acceptable. The Ginna design basis temperature profile exceeds the NRC results except between 10,000 seconds and approximately 20,000 seconds after the design basis event. We propose that the Ginna design basis temperature profile remain as shown in the FSAR for times less than 10,000 seconds and be revised beyond 10,000 seconds as follows:

.from 10,000 seconds to 20,000 seconds, temperature = 250oF, beyond 20,000 seconds, temperature < 100oF. Since the containment tern>>

perature is already decreasing at this time, 'it is not considered 820201 82020 102'R 05000244 PDR ADOCK PDR F p

l 0'HEET

~ ~ ~ ~

ROCHESTER GAS AND ELECTRIC CORP. NO 2

'DATE February 1', 1982 TO Mr. Dennis M. Crutchfield that this revised profile would invalidate any conclusions drawn as part of our review of environmental qualification of electrical equipment, since the affected equipment temperatures would also be decreasing at this time from their peak temperatures.

Regarding the NRC steam break analyses, the NRC results for pressure and temperature exceed the Ginna design basis as revised above. However, as shown in Attachment B, acceptable results for containment pressure are obtained when more reasonable assumptions are used. It is our conclusion, based on Attachment B, that the Ginna design basis pressure profile exceeds the pressure profile calculated for main steam line break and, therefore, remains acceptable. We conclude that the temperature resulting from a steam line break in containment may exceed the Ginna design.

basis temperature profile, but that this is of no consequence due to the short duration of this exceedance and. may therefore be ignored. This conclusion is based on NRC guidance provided in the Division of Operating Reactors (DOR) Guidelines which in turn is based, for example, on the short duration of the t'em-perature spike, lower heat transfer coefficient, and the elevation of the steam lines relative to equipment. Thus, we conclude that the Ginna design basis temperature profile, as revised based on LOCA results discussed above, remains valid.

Very truly yours, Jo n E. Maier

ATTACHMENT A COMMENTS ON NRC SAFETY EVALUATION Containment conditions resulting frorq a main steam break were assessed in a gross fashion in FSAR page 14.2.5-10.

That analysis does.not, however, comply with current criteria.

p. 5 The LOCA analysis underestimated'he effectiveness of the containment fan coolers by a heat removal rate of approximately 25 to 30 million BTU per hour'. This resulted from an incorrect set: of data being provided by RGSE to the NRC consultant

{see also LER 81-022 transmitted by letter dated January 4, 1982 from John E. Maier, RG&E, to Ronald C. Haynes, NRC).

Ne have estimated the impact of correcting the fan cooler heat removal rate to be on the order of a 1 to 2 psi reduction in peak containment pressure. This, therefore, is an additional-conservation in the analysis.

p. 6 .The SER discussion of the result'of a main steam line break shou1d be revised to reflect the sensitivity study presented in Attachment B:

p. 7 The last paragraph of the SER should be revised to reflect the conclusions presented in our letter transmitting this attachment.

p.l5 See our comment above regarding FSAR Section 14.2.5

'.l6 The assUmption of a'constant containment pressure of 14.7 psia throughout the transient will result in an overestimate of the mass and energy release and, therefore, an overestimate of containment, pressure and temperature.

p.l6 All information provided in conversations was also available on the docket.

~

p.18 It is our understanding that only accumulator water'as

{conservatively) set at 272 ' F, not all ECCS flow. This should be clarified. If our understanding is not correct, then Table 1, which indicates RUST temperature, should be revised.

p-21 The containment design pressure is incorrectly stated at the top of-the page as being 74 psia; it is 75 psia.

lo General A number of othe'r conservatisms are discusSed in the LOCA

'valuation. A more reasonable assessment would not require the level of conservatism employed here.

ll. p.21.-25 See comments provided in Attachment B.

~ '

Attachment B .

Containment Temperature/Pressure Following a Main Steam Line Break Introduction

'The purpose of this study is to provide a reevaluation of the containment conditions following a main steam line break.

The first step will be to reconstruct the worst. case containment temperature/p'ressure transients presented by the NRC. in Reference 1 for a large steam br'eak. Once the Reference 1 results have been reproduced, the assumptions necessary to reproduce those results can be evaluated. It may then be possible to remove

'some of the conservatism and calculate a more reasonable result.

Discussion The containment temperatures and pressures presented in this study were calculated using the CONTEMPT-EI/28B computer code. The results presented in Reference 1 were calculated using CONTEMPT-LT/028. The CONTEMPT-EI/28B code is quite similar to the .CONTEMPT-LT/028 code. with changes which allow more:user flexibility.

Hot Zero Power Case The highest containment pressure was calculated in Reference 1 to occur for a large steam line break at HZP with, failure of one spray train. The input for this case was run using CONTEMPT-EI/28B. Figure 1 and 2 illustrates the results of this run and points taken from Reference 1. The following peak temperature and pressure was h

obtained;,

Reference 1 Case $5 85.8 psia 9 91 sec.

413o 9 34 sec.

CONTEMPT-EI/28B 83.4 psia 9 99.8 sec. 403.9o 9 35 sec.

While reproducing this case from the Reference 1 input, one inconsistency was noted. Reference 1 states that spray was initiated 35 seconds after the setpoint at 30 psig was general, '.this pressure setpoint is reached at approximately reached'n 10 sec. Therefore, spray would start at approximately 45 sec.

Since the temperature rise is terminated by spray, the peak temperature would occur when spray starts. All curves in Reference 1 illustrate peak temperature at approximately 35 sec. 'Therefore, i" appears that the Reference 1 analysis neglects- the time to reach the spray setpoint when actuating spray.

Using the reconstruction of Reference 1 case $ 5 as the base case, several cases were run to determine the sensitivity of containment temperature and pressure to various parameters.

The results of these sensitivities are listed on Table 1 and are discussed below.

Q/V, where Q is the total energy released to the time of peak containment pressure and is the containment volume, is a parameter associated with the Tagami film heat transfer correlation. This should represent total energy release to the time of peak pressure..

A Q/V of " 165 results from the energy released to containment up to the time of peak pressure. However, a better approximation of the Reference 1 results can be obtained reducing this parameter. The effect

'of reducing this byparameter can be seen by comparing 01 and t3 on Table 1. increasing Q/V results in increasing the film heat transfer coefficient. Changing Q/V from 87 to 165 results in approximately -1.0 psi pressure change and approximately 1 change in temperature

(<3 temperature 400.8o 9 35.0 sec.). Therefore, the Q/V term in Tagami may. be doubled and still have

only a small effect on containment temperature 'and, pressure.

The Uchida.film heat transfer correlation has traditionally been used for steam breaks. $ fhen Uchida is used in the Ref. 1 Model a 3.7 psi pressure reduction and a 15,8o temperature reduction results (gl versus 44 on Table 1).

Exxon Nuclear Company (ENC) mass and energy release

,for the most limiting large steam line break (Ref. 3) was used in the evaluation. The ENC mass and energy was normalized to the total mass in the broken steam generator at HZP plus the mass released from the unaffected steam generator until main steam isolation occurs. The normalized ENC mass and energy is illustrated on Figure 3 with the mass and energy release used in the Reference 1 analysis. The mass associated with auxilary feed was not included. The effect of auxiliary feed on peak containment pressure and temperature would be negligible since the mass added during the time frame of interest is a very small fraction of the secondary side inventory (<1%).

The effect of using the normalized Exxon mass and energy release is a pressure reduction of 3.3 psi and a tempera-ture reduction of 17.4o (~4 and >5 on Table 1).

~ 3 ~

A comparison between the RG&E containment model and the containment model used in Ref 1 is illustrated on

~

Table 2. The major difference between the models is the inclusion of the accumulators and ducting in the RGRE model ~ The area of the -ducting is an assumed value based on values used by other plants, i.e.,

Palisades = 20,072 9 0.10 in.

Zndian Point = 22,000 8 0.1382 in.

Prairie Zsland = 22,000 9 0.1875 in.

Therefore,.an. assumption was made that Ginna had 20,000 sq. ft. 6 0.10 in.

The CONTEMPT codes are sensitive to node spacings A large spacing will result in lower surface temperatures which will result in removing too much energy from containment. The effect of node spacing can be seen by comparing 06 and C7. The effect of inclusion of accumulators and ducting is also illustrated on Table 1.

Zn the process of doing this study it"was determined that the heat removal capacity of. the fan coolers used in the Reference 1 analysis was the capacity of one fan cooler at a service water 'temperature of

'35 F. This corresponds to maximum cooling capability for one cooler. The minimum capability should be used in this analysis. The minimum capability is associated with the maximum service water temperature (80 F). Reference 4 presents a curve of heat removal versus containment pressure and equipment specifications presents the capacity at 120 and 286 The following illustrates the heat removal capacity used in case ill of. Table 1: The capacities represent the minimum values of Reference 4 and equipment specifica-

~

tions; therefore, the values are conservative.

containment heat removal total heat temperature per fad removal (4 fans) oF MBTU/hr MBTU/hr 120- 1.575 6.30 286 50 ' 200.

308 54.72 218.9 320 56.52 226.0 The effect of using the appropriate fan cooler capacity can be seen by comparing 410 and Oil on Table 1. This represents a 2.2 psi reduction in containment pressure and a 4.7O reduction in containment temperature.

The effect of containment volume is illustrated on Table 1. Increasing the"volume by 28,000 cu. ft. results in. a 1.4 psi reduction in pressure and a 1.9 reduction in temperature. Since the gross volume of containment is approximately 1 13E6; 28,000 cd. ft. represents

~

.approximately 2.58 of the gross volume. Calculations show a net volume of approximately 1.037E6 cu. ft. Based on the FSAR the, net volume of 9.72E5 cu. ft. represents a conservative small volume containing at least 3%

margin. Therefore, a best estimate volume would be between 9.72E5 and 1.037E6 cu. ft. This represents available margin that was not used in this study.

Figures 4 and 5 illustrate the effect of Uchida, ENC mass and energy, and the RGEE containment heat sink model on the worst case containment response presented in Reference 1 44, 05, and Oil of Table 1).

(tl, Using the RGRE containment heat sink model (volume = 9.72E5),

ENC mass and energy release, Uchida correlation and fan cooler capacity of four fans r'esults. in:

72.4 psia 9 128.6 sec. 356.1o Q 20.3 sec.

Full Power Case The highest containment temperature was calculated in Reference 1 to occur for a large steam break occurring at full power with failure of one spray train. The mass and energy release presented in Reference 1 for this case was coupled with the Reference 1 model discussed previously and containment temperature and pressure was calculated using the CONTEMPT-EI/28B code. The following results were obtained:

Reference 1 Case 03 75 psia 9 60 sec.

421o 9 34 sec.

CONTEMPT.-EZ/28B 73.3 psia 9 59.0 sec. 412.1o Q 35.0 sec.

The mass and energy release presented in Reference 1 was used with the RGEE containment heat sink model (volume = 9 '2E5) previously described, Uchida correlation, and fan cooler capacity of four fans. This resulted in the following peak temperature and pressure:

63.2 psia 9 51.8 sec. 374 0o 9 32.0 sec.

), ~

The temperature and pressure versus time is illustrated on Figures 6 and 7 together with the reproduction of Reference 1 Case 43 using, the CONTEMPT-EI/28B code.

~summa In summary, the following compares the Reference 1 worst case with the comparable worst case calculated by RG&E as previously described:

case Reference 1 Results RG&E Results Steam Break - HZP 85.8 psia 9 91 sec 72.4 psia 9 128.6 sec 13o 9 34 sec 356.1o 9 20.3 sec I

Steam Break - HFP 75 psia 9 60 sec 63.2 psia.9 51.8 sec 421 6 34 sec 374.0 9 32.0 sec For the worst case transients the calculated peak pressure

. is less than the design pressure for the Ginna containment (60 psig). The temperature is above the design temperature for Ginna containment (286 F). However, the temperature is. exceeded only for a short period of time.

The RG&E calculations do not account. for revaporization, entrain-ment, or best estimate containment volume. Inclusion of these effects would result in additional margin to design limits.

Revaporization Reference 2 presents a discussion on revaporization. A temperature response is presented for a large steam line break using the Uchida heat transfer coefficient. Nhen revaporization is used the temperature profile is reduced by approximately 40o. This would also result in a reduction in containment pressure.

2 ~ Entrainment In reality the steam flowing out of the break would not be dry steam but would contain some moisture.

As the moisture content of the steam increases, the energy associated with the steam decreases; therefore, the energy added to containment decreases. This would result in a decrease in containment pressure and temperature.

been estimated that the decrease in containment pressure It has and temperature resulting from accounting for entrainment would be similar to the decrease associated with revapori-zation.

3 ~ Containment Volume 'As previously described, a best esti-mate containment volume would be between 1.037E6 and 9.72E5 cu. ft. Increasing the containment volume used in the analysis would result in a slight pressure decrease.

References 1 ~ NRC letter from D. M. Crutchfield to J. E. Maier, "Systematic Evaluation Program (SEP) for the R. E. Ginna Nuclear Power Plant Evaluation Report on Topics VI-2.D and Vl-3," November .

3, 1981.

2; NRC letter from R. Tedesco to R. Mattson, V. Stello, and R. Boyd, "Best Estimate Evaluation for Environmental Qualifica-tion of Equipment inside Containment Following -A-Main Steam Line Break," February 21, 1978.

~

3. Exxon Report XN-NF-77-40 Supplement 1, "Plant Transient Analysis for the R. E. Ginna Unit 1 Nuclear Power Plant,"

March 1980.'.

Rochester Gas and Electric Corporation, R.E. Ginna, FSAR.

TABLE 1 SENSZTXVITY STUDY Hot Zero Power Case Peak Pressure Peak Temperature Case ( sia) ( sia) P

'1; Ref. 1 Model, Q/V = 87 83.4 9 99.8 sec. 403.9 9 35.0 sec.

2. Re f. 1 Model, Q/V = 87, 83.7 9 99.4 sec. 404.8 9 35.0 sec. +0.3 + 0.9 no fan 'coolers 3 ~ Re f . 1 Model, Q/V ~ 82.4 9 84.4 sec. 408.0 8 44.6 sec. -1.0 +-4'1 165, spray 9 45 sec.

4. Ref. 1 Model, Uchida 79.7 8 100.4 sec. 388.1 9 34.8 sec. -3.7 >>15.8

5. Ref. 1 Model, Uchida, 76.4 9 129.4 sec. 370.7 9 32.6 sec. -3.3 -17.4 ENC mass/energy 6 RGE Model (large 73.5 9 129 sec. 359.7 8 32.8 sec. -2.9 -11.0 spacing), Uchida, ENC mass/energy

\

7. RGE Model (small 75.1 9 129.2'ec. 365.4 9 32.6 sec. +1.6 + '5.7 spacing), Uchida, ENC mass/energy

8. same as 7 with con- 73.7 9 129.2 sec. 363.5 9 32.8 sec. -1.4 1.9 tainment volume =

1 ~ OE6

9. same as 7 with 74.9 9 128.8 sec. 365.4 9 34.6 sec. -0.2 0.0 accumulators

10. same as 9 with ducting 74.6 9 129.0 sec. 360.8 9 30 ' scca -0 ' 4.6

11. same as 10 with 4 'fan 72.4 9 128.6 sec. 356.1 8 20.29 sec. -2.2 -4.7 coolers

~ ~

~ ~ ~

0 TABLE 2 CONTAINMENT MODEL Reference 1 RG&E Containment Volume. 9. 72E5 cu. ft. Containment Volume 9.72E5 cu. ft.

Insulated Walls Dome and 36,181 sq. ft. same 36,181 sq. ft.

Uninsulated Dome 12, 474 sq. ft. same ., 12,474 sq. ft.

and Walls I Sump Walls 2,342 sq. ft. same 2,342 sq. ft.

Sump Floor 297

• 2,639 sq. ft.

Basement Floor <<7,955 sq. ft.

Refueling Cavity 6,400 sq. ft. same 6,400 sq. ft.

I~

Wall and Floor

'utside Refueling 21,800 sq. ft. 'ame 21,,800 Cavity and S.G.'omp.

Operating Floor 9,162 sq. ft. same 9,162 30,962 Intermediate Floor 6,170 sq. ft. same but 2 X Area = 12,340 sq. ft. ~

1.5 in. Beams 9,174 sq. ft. same 9,174 sq. ft.

1.0 in. Beams 5, 016 sq. ft. same 5,016 sq. ft.

0.5 in. Beams 8,586 sq. ft. same 8,586 sq. ft.

Crane Supports 5,756 sq. ft. same 5,756 sq. ft.

Grating etc. 7,000 sq. ft. same 7,000 sq. ft.

~

Accumulators 1, 756 sq. ft.

Ducting 8 0.10 in. 20,000 sq. ft.

Surface assumed to be in contact with pool water.

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REGUlATORY 'RMATION DISTRIBUTION SYS j (RIDS)

AGCESSION NBR:8202090299 DOC ~ DATE: 82/02/01 NOTARIZED; NO ,

DOCKET FACIL:50-244 BYNAME Robert Emmet Ginna Nuclear Planti Unit 1P Rochester G 05000244 AUTH AUTHOR AFFILIATION MAIERP J,E ~ Rochester Gas L Electr ic Corp ~

RECIP ~ NAME RECIPIENT AFFILIATION CRUTCHFIELDPD ~ Operating Reactors Branch 5

SUBJECT:

Forwards comments on NRC 811103 safety evaluation 8 draft evaluation of SEP Topics VI-2Di "Mass 8 Energy Release "Capability' Inside Containmenti" 8 VI-3P "Containment Pressure 8 Heat Removal "

DISTRIBUTION CODE: A03BS TITLE: SEP Topics COPIES RECEIVED:LTR L ENCL l SIZE:

NOTES: 1 copy:SEP Sects Ldr ~ 05000200 RECIPIENT COPIES REC IP lENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL ACTION: ORB P5 BC 01 7 7 INTERNAL: IE 06 2 2 NRR/DE/ADMQE 13 1 NRR/DE/HGEB 10 2 2 NRR/DL/ORAB 11 1 NRR/DL/SEPB 12 3 3 NRR /AEB 1 NRR/DS I/CSB 07 1 1 IL 00 EXTERNAL: ACRS 14 10 10 LPDR 03 NRC PDR 02 1 1 NTIS 5

'33 s3 TOTAL NUMBER OF COPIES REQUIRED: LTTR g ENCL

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////////t//// irzrizu tg I////'i Ii/, //Zl/1////I' ROCHESTER GAS AND ELECTRIC CORPORATION ~ 89 EAST AVENUE,'OCHESTER. N Y. 14649 d

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JQHM f. MAlfR TCLKPHONf.

Vice President ARt:* coDK Tie 546-2700 e

l February 1, 1982 Director of Nuclear Reactor Regulation Oi'-

cQpgy~@

Attention: Mr. Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 IE8 8 U.S. Nuclear Regulatory Commission 6 eve@ 198 Washington, D.C. 20555

Subject:

SEP Topics VI-2.D and VI-3 R. E. Ginna Nuclear Power Plant Docket No. 50-244

Dear Mr. Crutchfield:

This letter is in response to the draft evaluation of SEP Topic VI-2.D; "Mass and Energy Release Inside Containment" and SEP Topic VI-3, "Containment Pressure and Heat Removal Capability,"

which was transmitted by your letter dated November 3, 1981.

We have reviewed the draft evaluation and have identified several conservatisms in the analysis for the loss of coolant accident (LOCA). These conservatisms and a qualitative discussion of their impact on the LOCA results, as well as a number of general comments on the evaluation, are dicussed in Attachment A. We also identified a number of conservatisms in the analysis for the main steam line break (MSLB) Because of the degree of ~

conservatisms in the NRC evaluation, we performed a sensitivity study with the code CONTEMPT-EI/28B. This code is very similar to CONTEMPT-LT028, as discussed in Attachment B. Therefore, while this code has not been completely qualified for use as a licensing code, we believe that it is accurate and adequately represents Ginna. The 'sensitivity study, presented in Attachment B to this letter, confirmed the conservatism of the NRC results for MSLB.

Regarding the LOCA analysis, since the Ginna design basis pressure envelopes the NRC results, we conclude that the Ginna design basis pressure profile remains acceptable. The Ginna design basis temperature profile exceeds the NRC results except between 10,000 seconds and approximately 20,000 seconds after the design basis event. We propose that the Ginna design basis temperature profile remain as shown in the FSAR for times less than 10,000 seconds and be revised beyond 10,000 seconds as follows:

from 10,000 seconds to 20,000 seconds, temperature = 250 F, beyond 20,000 seconds, temperature < 100oF. Since the containment tem-perature is already decreasing at this time, it is not consid'ered 8202090299 820201 oS<

r 'PDR ADDCK 05000244- I.:

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ROCHESTER GAS AND ElTRIC CORP SHEET NO 2 DATE February 1, 19M TO Mr. Dennis M. Crutchfield that this revised profile would invalidate any conclusions drawn as part of our review of environmental qualification of electrical equipment, since the affected equipment temperatures would also be decreasing at this time from their peak temperatures.

Regarding the NRC steam break analyses, the NRC results for pressure and temperature exceed the Ginna design basis as revised above. However, as shown in Attachment B, acceptable results for containment pressure are obtained when more reasonable assumptions are used. It is our conclusion',.based on Attachment "B, that the Ginna design basis pressure profile exceeds the pressure profile calculated for main steam lin'e break and, therefore, remains acceptable. We conclude that the temperature resulting from a steam line break in containment may,exceed the Ginna design basis temperature profile, but that this is of no consequence due to the short duration of this exceedance and may therefore be ignored. This conclusion is based on NRC guidance provided in the Division of Operating Reactors (DOR) Guidelines which in turn is based, for example, on the short duration of the tem-perature spike, lower heat transfer coefficient, and the elevation of the steam lines relative to equipment. Thus, we conclude that the Ginna design basis temperature profile, as revised based on LOCA results discussed above, remains valid.

Very truly yours, Jo n E. Maier

P ~

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ATTACHMENT A COMMENTS ON NRC SAFETY EVALUATION

p. 5 Containment conditions resulting from a main steam break were assessed in a gross fashion in FSAR page 14.2.5-10.

That analysis does not, however, comply with current criteria.

p. 5 The LOCA analysis underestimated the effectiveness of the containment fan coolers by a heat removal rate of approximately 25 to 30 million BTU per hour. This resulted from an incorrect, set of data being provided by RG&E to the NRC consultant (see also LER 81-022 transmitted by letter dated January 4, 1982 from John E. Maier, RGSE, to Ronald C. Haynes, NRC).

We have estimated the impact of correcting the fan cooler heat removal rate to be on the order of a 1 to 2 psi reduction in peak containment pressure. This, therefore, is an additional conservation in the analysis.

p. 6 The SER discussion of the result of a main steam line break should be revised to reflect the sensitivity study presented in Attachment B.

po 7 The last. paragraph of the SER should be revised to reflect the conclusions presented in our letter transmitting this attachment.

p.l5 See our comment above regarding FSAR Section 14.2.5.

p. 16 The assumption of a constant containment pressure of 14.7 psia throughout the transient will result in an overestimate of the mass and energy release and, therefore, an overestimate of containment pressure and temperature.

p.16 All information provided in conversations was also available on the docket.

p.l8 It is our understanding that only accumulator water was (conservatively) set at 272.9 F, not all ECCS flow. This should be clarified. If our understanding is not correct, then Table 1, which indicates RWST temperature, should be revised.

p.21 The containment design pressure is incorrectly stated at the top of the page as being 74 psia; it is 75 psia.

General A number of other conservatisms are discussed in the LOCA evaluation. A more reasonable assessment would not require the level of conservatism employed here.

p.21-25 See comments provided in Attachment B.

Attachment B Containment Temperature/Pressure Following a Main Steam Line Break Introduction The purpose of this study is to provide a reevaluation of the containment conditions following a main steam line break.

The first. step will be to reconstruct the worst case containment temperature/pressure transients presented by the NRC in Reference 1 for a large steam break. Once the Reference 1 results have been reproduced, the assumptions necessary to reproduce those results can be evaluated. It may then be possible to remove some of the conservatism and calculate a more reasonable result.

Discussion The containment temperatures and pressures presented in this study were calculated using the CONTEMPT-EI/28B computer code. The results presented in Reference 1 were calculated using CONTEMPT-LT/028. The CONTEMPT-EI/28B code is quite similar to the CONTEMPT-LT/028 code with changes which allow more user flexibility.

Hot Zero Power Case The highest containment pressure was calculated in Reference 1 to occur for a large steam line break at HZP with failure of one spray train. The input for this case was run using CONTEMPT-EI/28B. Figure 1 and 2 illustrates the results of this run and points taken from Reference l. The following peak temperature and pressure was obtained.

Reference 1 Case 55 85.8 psia 9 91 sec.

413o g 34 CONTEMPT-EI/28B 83.4 psia 9 99.8 sec. 403.9o 9 35 sec.

While reproducing this case from the Reference 1 input one inconsistency was noted. ,Reference 1 states that spray was initiated 35 seconds after the setpoint at 30 psig was reached.

In general, this pressure setpoint is reached at approximately 10 sec. Therefore, spray would start at approximately 45 sec.

Since the temperature rise is terminated by spray, the peak temperature would occur when spray starts. All curves in Reference 1 illustrate peak temperature at approximately 35 sec. Therefore, it appears that the Reference 1 analysis neglects the time to reach the spray setpoint when actuating spray.

~

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Using the reconstruction of Reference 1 case 55 as the base case, several cases were run to determine the sensitivity of containment temperature and pressure to various parameters.

The results of these sensitivities are listed on Table 1 and are discussed below.

Q/V, where Q is the total energy released to the time of peak containment pressure and is the containment volume, is a parameter associated with the Tagami film heat transfer correlation. This should represent total energy release to the time of peak pressure.

A Q/V of 165 results from the energy released to containment. up to the time of peak pressure. However, a better approximation of the Reference 1 results can be obtained by reducing this parameter. The effect of reducing this parameter can be seen by comparing gl and 43 on Table 1. increasing Q/V results in increasing the film heat transfer coefficient. Changing Q/V from 87 to 165 results in approximately -1.0 psi pressure change and approximately -3 1 change in temperature (g3 temperature 400.8 Q 35.0 sec.). Therefore, the Q/V term in Tagami may be doubled and still have only a small effect on containment temperature and pressure.

The Uchida film heat transfer correlation has traditionally been used for steam breaks. When Uchida is used in the Ref. 1 Model a 3.7 psi pressure reduction and a 15.8o temperature reduction results (gl versus g4 on Table 1).

Exxon Nuclear Company (ENC) mass and energy release for the most limiting large steam line break (Ref. 3) was used in the evaluation. The ENC mass and energy was normalized to the total mass in the broken steam generator at HZP plus the mass released from the unaffected steam generator until main steam isolation occurs. The normalized ENC mass and energy is illustrated on Figure 3 with the mass and energy release used in the Reference 1 analysis. The mass associated with auxilary feed was not included. The effect of auxiliary feed on peak containment pressure and temperature would be negligible since the mass added during the time frame of interest is a very small fraction of the secondary side inventory (<1%).

The effect. of using the normalized Exxon mass and energy release is a pressure reduction of 3.3 psi and a tempera-ture reduction of 17.4o (g4 and 45 on Table 1).

P

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A comparison between the RG&E containment model and the containment model used in Ref. 1 is illustrated on Table 2. The major difference between the models is the inclusion of the accumulators and ducting in the RGRE model. The area of the ducting is an assumed value based on values used by other plants, i.e.,

Palisades = 20,072 9 0.10 in.

Indian Point, = 22,000 9 0.1382 in.

Prairie Island = 22,000 9 0.1875 in.

Therefore, an assumption was made that Ginna had 20,000 sq. ft. 9 0.10 in.

The CONTEMPT codes are sensitive to node spacing.

A large spacing will result in lower surface temperatures which will result in removing too much energy. from containment. The effect of node spacing can be seen by comparing ¹6 and ¹7. The effect. of inclusion of accumulators and ducting is also illustrated on Table l.

In the process of doing this study 'it 'was determined that the heat removal capacity of the fan cooler's used in the Reference 1 analysis was the capacity of one fan cooler at a service water temperature of 35 F. This corresponds to maximum cooling capability for one cooler. The minimum capability should be used in this analysis. The minimum capability is associated with the maximum service water temperature (80 F). Reference 4 presents a curve of heat removal versus containment pressure and equipment specifications presents the capacity at 120 and 286 The following illustrates the heat removal capacity used in case ¹ll of Table 1: The capacities represent the minimum values of Reference 4 and equipment specifica-tions; therefore, the values are conservative.

containment heat removal total heat temperature per fan removal (4 fans)

OF MBTU/hr MBTU/hr 120 1.575 6.30 286 50 ' 200.

308 54.72 218.9 320 56.52 226.0 The effect of using the appropriate fan cooler capacity can be seen by comparing ¹10 and represents a 2.2

¹ll on Table 1. This psi reduction in containment pressure and a 4.7o reduction in containment temperature.

I I

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The effect of containment volume is illustrated on Table l. Increasing the volume by 28,000 cu. ft. results in a 1.4 psi reduction in pressure and a 1.9o reduction in temperature. Since the gross volume of containment is approximately 1.13E6; 28,000 cu. ft. represents approximately 2.5% of the gross volume., Calculations show a net volume of approximately 1.037E6 cu. ft. Based on the FSAR the net volume of 9.72E5 cu. ft. represents a conservative small volume containing at least 3S margin. Therefor'e, a best estimate volume would be between 9.72E5 and 1.037E6 cu. ft. This represents available margin .that was not used in this study.

Figures 4 and 5 illustrate the effect of Uchida, ENC mass and energy, and the RGGE containment heat sink model on the worst case containment response presented in Reference 1 (gl, g4, 45; and Oil of Table 1).

Using the RGEE containment heat sink model (volume = 9.72E5),

ENC mass and energy release, Uchida correlation and fan cooler capacity of four fans results in:

72.4 psia 9 128.6 sec. 356.lo Q 20 3 sec Full'Power Case The highest containment temperature was calculated in Reference 1 to occur for a large steam break occurring at full power with failure of one spray train. The mass and energy release presented in Reference 1 for this case was coupled with the Reference 1 model discussed previously and containment temperature and pressure was calculated using the CONTEMPT-EI/28B code. The following results were obtained:

Reference 1 Case g3 75 psia 9 60 sec.

421o 9 34 sec.

CONTEMPT-EI/28B 73.3 psia 9 59.0 sec. 412.1 9 35.0 sec-The mass and energy release presented in Reference 1 was used with the RGSE containment heat sink model (volume = 9.72E5) previously described, Uchida correlation, and fan cooler capacity of four fans. This resulted in the following peak temperature and pressure:

63.2 psia 9 51.8 sec. 374.0o 9 32.0 sec.

l1 pl 8 'g]

g I'

~

The temperature and pressure versus time is illustrated on Figures 6 and 7 together with the reproduction of Reference 1 Case g3 using the CONTEMPT-EI/28B code.

~Summa r In summary, the following compares the Reference 1 worst case with the comparable worst case calculated by RGEE as previously described:

case Reference 1 Results RGSE Results Steam Break HZP 85.8 psia 9 91 sec 72.4 psia 9 128.6 sec 413 9 34 sec 356.1o 9 20.3 sec Steam Break HFP 75 psia 9 60 sec 63.2 psia 9 51.8 sec 421 9 34 sec 374 Oo Q 32 0 s For the worst case transients the calculated peak pressure is less than the design pressure for the Ginna containment (60 psig). The temperature is above the design temperature for Ginna containment (286 F). However, the temperature is exceeded only for a short period of time.

The RG&E calculations do not. account for revaporization, entrain-ment, or best estimate containment volume. Inclusion of these effects would result in additional margin to design limits.

1 ~ Revaporization Reference 2 presents a discussion on revaporization. A temperature response is presented for a large steam line break using the Uchida heat transfer coefficient. When revaporization is used the temperature profile is reduced by approximately 40 . This would also result in a reduction in containment pressure.

2 ~ Entrainment In reality the steam flowing out of the break would not be dry steam but would contain some moisture.

As the moisture content of the steam increases, the energy associated with the steam decreases; therefore, the energy added to containment decreases. This would result in a decrease in containment pressure and temperature. It has been estimated that the decrease in containment pressure and temperature resulting from accounting for entrainment would be similar to the decrease associated with revapori-zation.

Containment, Volume As previously described, a best esti-mate containment volume would be between 1.037E6 and 9.72E5 cu. ft. Increasing the containment volume used in the analysis would result in a slight pressure decrease.

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References NRC letter from D. M. Crutchfield to J. E. Maier, "Systematic Evaluation Program (SEP) for the R. E. Ginna Nuclear Power Plant Evaluation Report on Topics VI-2.D and VI-3," November 3; 1981.

NRC letter from R. Tedesco to R. Mattson, V. Stello, and R. Boyd, "Best Estimate Evaluation for Environmental Qualifica-tion of Equipment. Inside Containment Following A Main Steam Line Break," February 21, 1978.

Exxon Report XN-NF-77-40 Supplement 1, "Plant Transient Analysis for the R. E. Ginna Unit 1 Nuclear Power Plant,"

March 1980.

Rochester Gas and Electric Corporation, R.E. Ginna, FSAR.

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TABLE 1 SENSITIVITY STUDY Hot Zero Power Case Peak Pressure Peak Temperature Case ( sia) ( sia)

1. Ref. 1 Model, Q/V = 87 83.4 9 99.8 sec. 403.9 9 35.0 sec.

2. Ref. 1 Model, Q/V = 87, 83.7 9 99.4 sec. 404.8 9 35.0 sec. +0.3 + 0.9 no fan coolers 3 ~ Ref. 1 Model, Q/V = 82.4 8 84.4 sec. 408.0 9 44.6 sec. -1.0 + 4.1 165, spray 9 45 sec.

4. Ref. 1 Model, Uchida 79.7 9 100.4 sec. 388.1 9 34.8 sec. -3.7 -15.8

5. Ref. 1 Model, Uchida, 76.4 9 129.4 sec. 370.7 9 32.6 sec. -3.3 -17.4 ENC mass/energy

6. RGE Model (large 73.5 9 129 sec. 359.7 9 32.8 sec. -2.9 -11.0 spacing),'chida, ENC mass/energy I

7. RGE Model (small 75.1 9 129.2 sec.,365.4 9 32.6 sec. +1.6 + 5.7 spacing), Uchida, ENC mass/energy

8. same as 7 with con- 73.7 9 129.2 sec. 363.5 9 32.8 sec. -1.4 1.9 tainment volume =

1-OE6

9. same as 7 with 74.9 9 128.8 sec. 365.4 9 34.6 sec. -0.2 0.0 accumulators

10. same as 9 with ducting 74.6 9 129.0 sec. 360.8 9 30.0 sec. -0.3 4.6 ll. same as coolers 10 with 4 fan 72.4 9 128.6 sec. 356.1 9 20.29 sec. -2.2 -4.7

TABLE 2 CONTAINMENT MODEL Reference'1 RGScE Containment Volume 9.72E5 cu. ft. Containment Volume 9.72E5 cu.

Insulated Dome and 36,181 sq. ft. same 36,181 sq. ft.

Walls Uninsulated Dome 12,474 sq. ft. same 12,474 sq. ft.

and Walls Sump Walls 2,342 sq. ft.. same 2,342 sq. ft.

Sump Floor 297

• 2,639 sq. ft.

Basement Floor *7,955 sq. ft.

, Refueling Cavity 6,400 sq. ft. same 6,400 sq. ft.

Wall and Floor Outside Refueling 21,800 sq. ft. same 21,800 Cavity and S.G.

Comp.

Operating Floor 9,162 sq. ft. same 9,162 30,962 Intermediate Floor 6,170 sq. ft. same but 2 X Area = 12,340 sq. ft.

1.5 in. Beams 9,174 sq. ft. same 9,174 sq. ft.

1.0 in. Beams 5,016 sq. ft. same 5,016 sq. ft.

0.5 in. Beams 8,586 sq. ft. same 8,586 sq. ft.

Crane Supports 5,756 sq. ft. same 5,756 sq. ft.

Grating etc. 7,000 sq. ft. same 7,000 sq. ft.

Accumulators 1,756 sq. ft.

Ducting 9 0.10 in. 20,000 sq. ft.

• Surface assumed to be in contact with pool water.

Is FIGURE 1 CONTAINMENT TEMPERATURE STEAM BREAK HZP l/2 SPRAY SYSTEM e ~S P>y$ V)P Vg I

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i ~

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Reference l FIGURE 6 CONTAINMENT TEMPERATURE STEAM BREAK HFP 1/2 SPRAY SYSTEM ec on st ruc Re Ca s e 03 p,IQ g'

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