ML17258A192

Forwards Draft Safety Evaluations of SEP Topics XV-16 & XV-20 & Radiological Aspects of SEP Topics XV-2,XV-12 & XV-17.Recommends Tech Spec Amend to Incorporate Max Primary Coolant Activity During Spiking

ML17258A192
Person / Time
Site:	Ginna
Issue date:	09/24/1981
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:	Maier J ROCHESTER GAS & ELECTRIC CORP.
References
TASK-15-02, TASK-15-12, TASK-15-16, TASK-15-17, TASK-15-2, TASK-15-20, TASK-RR LS5-81-9-64, LSO5-81-09-064, LSO5-81-9-64, NUDOCS 8110010360
Download: ML17258A192 (25)

Text

gpiI AC<II

~O UhllTED STATES NUCL AR REGULATORY COMMlSSION WaSHINGTON, O. C. 2q55~

September Z4, I981 Docket No. 50-244 LS05-81-09-064 Mr. John E. Maier Vice President Electric and Steam Production Rochester Gas 6 Electric Corporation 89 =ast Avenue Rochester, New York 14649

SUBJECT:

SEP TOPICS, XY-Z, SPECTRUM OF STEAM SYSTEM PIPING FAILURES INSIDE AND OUTSIDE CONTAINMENT, XV-12, SPECTRUM OF ROD ='T;OIi ACC'.O=l TS, XV-16, RADIOLOGICAL CONSEQUENCES OF FAILURE OF SHALL LINES CARRYING PRIMARY COOLANT OUTSIDE CONTAINMENT, XY-17, STEAM GENERATOR TUBE FAILURE, AND XV-20, RADIOLOGICAL CONSEQUENCES OF FUEL DAINGING ACCIDENTS - 'R. E.

GINNA Enclosed are the staff's draft safety evaluations of Topics XY-16 and XY-20 and the radiological aspects of Topics XV-2. XY-12 and XV-17 for the R. E.

Ginna Nuclear Power Plant.

The systems portions of Topics XY-2, XY-12 and XY-17 were previously sent by letter dated September 4, 1981.

Inese assess-ments compare your facility as described in Docket No. 50-244 wi.h the criteria currently used for licensing new facilities.

Please inform us ~ithin 30 days if your as-built facility differs from the licensing bases assumed in our assessments.

If no comments are received within 30 days we will assume the topic is coIrpleteo The results of the tooic assessments indicate that Ginna does not conform to current criteria in some areas.

Soecifically, we have found tha the existing plan-. technica I specif cations limiting the primary coolant activ..'ty are inadequate to meet the current dose criteri a for a steam generator

.ube failure accident.

Accordingly, we recomend that you incorporate technical speci fica-tfons in the following areas:

l.

The maxinam primary coolant activity during spiking should be 60 uCi/gm I-131 dose equivalent.

Current specifications do not limit the maximum activity.

2.

The duration of all spikes during the year should be 10 of the operating time by incorporating the Standard Technical Speci,ications for duration.

3.

Increase surveillance of the coolant activity during conditions of expected iodine spiking by incorporating the sampling requirements in the Standard Technical Specifications.

in order to expedite the resolution of these items we request that you inform us of your intent to comply with the staff's recommendations and your proposed schedule within 30 days of receipt of this letter.

These evaluations will be a basic input to the integrated safety assessment for your facility unless you identify changes needed to reflect the as-built conditions at your facility.

The assessments may oe revised in the futur if your facility design is changed or if NRC criteri a relating to these subjects are modified before the integrated assessment is completed.

Sincerely, Oennis N. Crutchfield, hief Operating Reactors Branch Ho.

5 Oivision of 'censing

Enclosure:

As stated cc w/encl osure:

See next page

CC Harry H. Voigt, Esquir L Boeuf, Lamb, Leiby and 'NacRae 1333 New Hampshire

Avenue, N. 'A.

Suite 1100 Washington, O. C.

20036 Nr. Michael S'iade 12 Trailwood Circle Roches er, New York

'l4618 Ezra Bialik Assistant Attorney General Environmen.al Protection Bureau New York State Oepartment of Law 2 World Trade Center New York, New York 10047 Jeffrey Cohen New York State Energy Office Swan Street Building Core 1, Second ".-loor Empiro State Plaza

Albany, New York 12223 Director, Bureau of Nuclear Operations State of New York Energy Offic

" Agency Building 2 Empire State Pla a

Albany, New York 12223 Rochester Public Library

,115 South Avenue Rochester, New York 14604 Supervisor of the Town of Ontario 107 Riage Road 'Jest

Ontario, New York 14519 Resident Inspector R. E. Ginna Plant c/o U. S.

NRC 1503 Lake Road

Ontario, New York 14519 4lr. Thomas B. Cochr an Natural Resources Oerense Council, Inc.

1725 I S"re t, N. u.

Suite 600 Mashington, O. C.

20006

~ ~

U. S. Environmental Prot ction Agency Region II Office ATTN:

Regional Radiation Representative 26 Federal Plaza New York, New York 10007 Herbert Grossman,

=sq.,

Chai rman Atomic Sarety and Licensing Board U. S. Nuclear Regulatory Coamission

'Aashington, O. C.

20555 Or. Richard F. Cole Atomic Safety and Licensing Board U. S. Nuclear Regulatory Cormission

'Aashington, 0. C.

20555 Or. Ejtmeth A. Luebke Atomic Safety and Licensing Board U. S. Nuclear Regulatory Cottmission

'Aashington, 0. C.

20555

R

~

GE>'iNA NUCL:-V PC'iH >CATION X'( P S?

CTRL'N OF ST"=4iii SYSi-9 P'IRING FAILURES I'f<IGK ANO OUTSIDE CON i AIiib'6'fT iR INTRODUCTION Rupture. of a stean line outside contain-,ant will allow radioactivity contained

\\

in the primary and secondary coolant to scape to the nvirorzznt.

ScTopic XV-?.

is intended to r view the radiological onsequences of such fai'.ures, The r view will encompass those design feat res which limit the release of 4

radioactivity in the released coolant x~ d the amount of primary to secondary leakage.

II. R VI-M CRI=R:A Sec.ion =O,A'f li0 CFR?art SO r.qu'.res that each applicant for a cons ructicn permit or'cpe. ating license provide an analysis and evaluaticn of the design and per'.ormance of structur s,

systems, and ccmocnents of the faci Hty with the objective of assess'ng the risk to public health and safety resulting from cper-ation of the facili y.

ine steam line 5re k ac"ident is cne of the postulated ac"idents used to valuate the adequacy of these s ructur s, sys

ems, anc components wi h respect to the public heal".h nd saf ty.

-'n addition'0 CrR Par-100.11 provides ms'. gu--e, ines fr r actor sit ng against wh'ich calcula.ad accident dose consequences may 5e c=moaroc.

'I~ RE'DL =O SAF=..

TOP'.C".

Topic II-.2.C, Atmospheric Transport and Qif,usicn Characteristics fcr Acc dent Analys's" provides "e metacr log'.c 1

ata usec to valuata

'.he cffsi:

-csas.

Tcpic r,r

~ A

~gffects of?ipe ~reaks on Struc-ures~

S<<tams anc C"mpcrents Ins'.de contaiwznt and T Dic 'I.'-i'g "!? pe 3reak Cuts ide An-.airmen " will cover the dynamiic affec.s of the "os ula.ad pipe ailur insi~m and outs

~ de ccntainaant.

The review of the r adiclogical consequences of these failures was conducted fn accordance with the Appendix of Standard Review Plan 1S.1.5, "Radiolcgfcal Consequences of <fain Steam Line Failures Qutsice Containment."

inis review cetarmines he affac.s cf fuel damage, icdfne spiking and primary o second-ary steam generator tube leakage cn exclusion area boundary

(=AB) anc lcw population zone (i.?

) boundary doses after a steam line break, then ccmoares these doses to the dcsa guideline values in 10 CFR Part 100.

The plant is considered ade"uat ly desigred agains s

earn line failures if the calculations show that the r suiting oiisita doses are less than a small free.ion (1G") of the 10 C.=R Part 1GO exposure guidelines and ar within the exposure guidelines of 10 CFR?ar 1CO for the casa of a pr accident iodine spike cr cne rcd held cut cf the car V

.VAL"ATION Rale<<sa of ac.ivity follcwfng this accident is via the se ondary system, but consists cf ac.ivf:y originally c"ntafned in the sac"ndary sys:am

<<nd that which passes fr".m the primary system through steam generator tube leaks.

Two cases have been eva':uatad: first, assuming that no f el failures occur resu 1

oi 'he ac"ident, but

<<as semi ng th a the primary

<<nd sac-ndary activit;es are at their present technical specification limits anc second;

<<assumfng that fuel r<<flures occur.

The plant char c:aristics u.fliz d in ".his aval ation are suamari ed in Table,'(V-2-1.

ine assumed pr mary ac:ivi:y limit, ~Ci/p, is :he pr esant Tacnn!cal Spe fficatfcn limit. It is further assumed that during the first

two hours I gaIlon per minute of primary coolant is released directly to the atrasphere and that one steam generator is blown dry within 60 secords follow-ing the accident.

The exclusion area boundary doses for this case are IQ Rem thyroid and 0.1 Rem wnole body.

The present primary coolant activity Technical Specifications allow optimum operation with coolant activity above the "equilibrium" value.

No limit is

~131 presently set 'on the maximum coolant activity during Oose Equivalent soikes.

As a check on the maximum value recommended in the report on Topic XV-17, a calculation of the steam line break accident with the primary coolant activity at the r corarended 60uCi/gm level has been performed; a.hyroid dose of 19 Rem has been calculated.

Although recent generic studies of PWR core response to the steam line break accident indicate that Oeparture from Nucleate Boiling Conditions would not be reached, it, was assumed in this evaluation that Ylof the core rods failed at the time of the accident.

A thyroid dose of 62 Rem and a whole body dose of O. I Rem were calculated for these conditions.

V:. CONCLUSIONS The plant is adequately designed against steam line break accidents outside of containment.

'ML= XV-Z-T ASSUi iPT ONS i AOE IN ANALYSIS OF THE RAO IOLOG ICAL CCNSEOUENCES OF THE MAIN S<cAN LINE 3R~4K OUTSIGE CQNTA:NMENT ACCIOENT Reactor Pcwer

• 1525 With 2.

Loss of Offsite

.-" wer follcwing the accident 3.

Reactor Coolant System (RCS) volume

~ 5236 ft 4.

RCS Operating Conditions

• 2250 psia, 580oF 5.

Normal Latcown =lcw

~ 40 gpm 6.

Steam Gene. stars Operating Ccnditions

~ 770 psia, 514oF 7.

Steam Gener tors sa cndary side wat r volume

• 1581 f.

1 8.

Stean Generators secondary side st am volume

• 2398 9.

Emergency Fe dwatar -

1 s ean oriven pep (4QO gpm)>>and 2 motor driven pumos (250 g"m,'0.

Steam Generator Flow Rata

• 3.13 x

10 lb/hr per ste>>am generator

11. Operating 3lcwdcwn Flew * '0 tc 50 g"m per steam generator

12. Primary ol&>>t c ivi "y prlol tc the accident o

3..uCi/gr>>m of Oose Equivalent I-131 nd 54/E ~i/g or totai ac.ivity.

13. Secondary coolant activity pr.'or to the 'ccident of 0.1 uCi/gram of ose

="quivalant I-13i.

14. Icdine 'econ-vina"ion f*c=cr of 10 '.etween water>>nd s-earn

15. Meteorolocical c"nditicns ccr".espcnd'.ng to a ground ievel r olaase of 4.8 x 10-4 sac/m3 at

>> distance of 45~ m.

(Sa~

co c <:-2.C; lo.

No add ticnal fuel melting.

17.

10" of ruei ac'vity in.he gaos for c'sas

~here "lad f'flure is assumed.

18. ?rimary-to-secondary ia>>> r>>te cf 1

g"m.=er steam generator

R. =. Ginrra Auciem~

Pavier St tian XV-1,2 5>~CTPU". CF R00 "'cCT{nr!'>CC708?!TS li'ITROOUC.TCH "rection of a control element asse...bly irm '.he care can occur if the control element drive mechanis..

housing or the noz le an the reactor vessel head bre>Ds oif circ~far n:ially.

7ne e'ectioa af a control ele. ent assembly by

."he reac:or coolant system pi.essure can c use a seve.

e re ct vity excursion.

This accident may resul in radioactivity beirg released ta the enviroment thr cugn the ste m ger.'eratar a.r.d cantair:-..ent 1 eaks.

SV Taaic X'I-L2 is intended ta evaluate he radiological sansequenc s of this accident.

The revim ~ill enca pass those plant desicn features

~hich limit Ne release includina t? e plant technical speci ficatians on prieary ta sacond-ar.

system leakage.

r.r.

RFl;.A CREr-RrA Section:0.34 of rC C;~ par 5(i re"ui. es that each acalicant For a construction pe~it or caerating license provide an aralysis and evaluation af:he design ard perfar-.arc af structures sys.

ms, ard components of:he,aciliay A h the objective a

assessing the risk a pub'ic heal h and safety resul ing Fram aper-ation af the faciligi.

Tih..e central rcd erection acc dert is ane or tne pasz-lated acciden

" used:a evaluate the ade,uacy O'

hese sect res,

systems, and capcne's ~i h respect ta the public health and safety.

.General "esign riterian {"~) 2S, "Reac.iviay Limni a', ai Apndix K ta

'.0 C"":. Par:"0, r~uires

.he reac=ivi p antral system ta "e desicn~;~i

". appra-pria

~ I mi s an he patent al amau..t and rate ar reactiv',"g incr ase.

GX 29 al sa r~uires

".at these postulated react vi".I accidents include cars deratian af the rad ejec:ian acc dent unless srrch an accident is prevent~

by pasitive means.

[n addition,

>0 CFR Par IOO.I~ provides dose guidelines fcr.reac ov.siting=

agains-wnich calculated accident cse consa uenczs may be c"moar d.

Topic I I-Z.C, "A~-..osoheric Transqort and "irfusion Character'.'s'.ics for 'cci-dent Analysis" Prcvides the met orolcnical data used to evaluat the offsit.

doses.

Various other S.=P topics evaluate such itens as cont irsent fsolat on, contairr nt I ak tasting,

=SF syster's, and steam venerator intecri".g.

IV R~Vr~a GU'~

i rv:

The revim o:

-.he r adiol cical conse".,uorces cf a control rcd e'ec.ion ac.,dent was conducted ir. accordance with the Aocendix to Stardard:",eviaw Plan 1=.4.8 and Regulatory Cuide L.77.

"-xistirg Plan. technical specir',cations will be taken into account in calculating the radiolcgical ccrsaquences.

The plant is considered adequa

~Iv designed acainst a control rod e'ection accident if the resulting doses a

he exclusicn area ard la pooulacion

.cne boundaries are within the guideline values or L0 CFR Part LGC.

I<L

'"I'"

The specific assumptions made in this evaluation are listed in Table XV-12-1.

The presently existing Technical Specification limit of 3uCi/gm I dose 131 equivalent was assumed.

In addition, a primary to secondary leakage of I

gpm for each steam generator has been

assumed, which is in excess of the plant's Technical Specification of 0. I gpm.

Jt has been conservatively assmed that the ac"ident is followers by a.

canplete loss of offsite power, thereby requiring the olant to be cooled

~

I down by r leasing secondary stean t" the envirorurent through the safety and relief valves.

An additional assumpt:on nas been the ailure of 1G" of the rods due to he accident with release of the gap inventory of those rods to the

.-".CS.

Ine r su I s

0 h

evaluation ar 03 l Ml thyl Qid dose and l'Kl whole bocy (5-2 hour at the ~3).

The LP: doses are lower.

V a.

CQl)CL~JSTQN5'he plant is adequately desi-ned against a control rod ejecticn acciden-.

Assumpt',ons Hade in Analysis of the Radiological Consequences of a Control Rod "=jection Accident 1.

Reactor Paver I:29 Hath 2.

Loss of Oa =sit Power t'oIIcwfng the ac"idents 3.

6.

7.

Reac:or Coolant System (RCS) volume

= 6236;.

RCS Operating Conditions

~

2ZSO psfa, 58CoF a

abnormal L tdcwn Flew

~ 40 gpm

'St~am Ceneratcrs Opera ing ""nditions

• 77O psfa, 614 F

S"aaw Gar.erators sec"ndarv side water volume

~ 1581 8.

9.

IQ.

13.

14.

16.

Steam gererators secondary side s.earn volume

• 2898 rt emergency Feedwa.ar

- I steam driven pump (4CO gpm) ard 2 motor drfven -umps (2OQ gpm)

Steam "=enera.or Flow Rate

~ 3.13 x IO lb/hr per steam generator Qperat'.ng 3lcwdcwn Flew

~ 40 to SQ gpm per steam generator Primary coolant ac.ivfty prier to the accident o= 3. uCf/gram or Case

=quivalen.

.'-131 and 84/= uCi/9 o-totaI actfvf:y.

1odfne spfking aactor oi

=.QO arter.he accident Seconcary coolant ac.ivity pr'.'or to the acciden of O.l uCi/gram o= Bose

-quival nt I-I31.

a iodine "ec"ntaminatfon 'ac.or or

!Q bet'~e n water and s.earn 16.

17.

18 a 19.

Heteoro'.oclp!

congftfcns cor. es-cnd'ng t" a "round level r lease o=

4.0 x 10 aac/ca a-a dfa-. cc cr a:0 e.

l!ee Tcpdc ip -2.0j le additional

-.uei meltfng as a

r su!= o-.

any o,

.he ac=ident.

IG~

o-.

=uel activfty in the gacs -or cases where c!ad iailure is assumed.

All releases thr.ugh the secondary side sa-.aty and railer valves.

P. ima7 y io saroradary Iaà rate a

9Pm Per s

earn generator,

R.=. Gfnna Huclear,".ewer S~ation XY-16 RNOICLCGIC'L COHS="UF"IC.6 Ql',=AfLURc 0; ':"AL'

.fES CARRYING PRL."PRY CQOLA'IT OUTSiOK ~HTAiM"~HT IHTROOUCTNH Rupture e'fnes car.jfng pri;.ary coolant cuts Ide contalnnant can aIIor pr,Wry coolant and the rad',cac.ivlty contained the~in to escape

.0 he ewircnmn-..

S:-~ Topic ÃI-16 is fntended:o 'vier

~e radfoiogfcaI consequenc s of suca '.aflur s.

The r vier cf thfs topfc enccnaassec

-1ose 1fnes ~hfch carry prf;wry coolant cutsfde =ntafnnant during parer ccera.icn.

The sccoe

!ncluded those lines that ar nct ncr.-.aIIy ex" c:ec to 5e cpen o the prfwry system 5u can

".e cpened ~ring pcwer cperatfcn (f.e., r.ac:or coolant sarmle I!nes, fnstrvaent lf..es, etc.).

l'.. Rrc =a CR>

"=R.'OH All smII lines carrying prf.-cry coolant cu.side contalnnant ~e~,

vfwed to ensu, that any release of rad'.cac-.ivfty circa their pcstula.ed iaflur was a s<II frac-.fon of the 10 C,=R Par-ICO e

csure guidelfnes.

S@II

.rac:icn fs defined fn the SRo:o be no wre than

'.Q~ c.

the '.0 CFR ?ar.

TOO e-csur

@fidel f nes.

Lll~

R 'l 3 SAF I

e IP <<5 AHC 'i RF+a <o S'2 T><ics, Y-'.0.3,

'RHR System Ral!aci i y'rd V-:.'.3,

'RHR.'nc rlcck R ui.

ma.. s

)

Tne review was anduct d in accardanco with SRP 15.8.Z.

me license was r eques:ad ta provide pIant-specific iniarmation such as the I entlfi-cation oi lines covered by this topic, the si~e ai these lines, break locaticns and flew, tc.

one licensee r spcnded ta this reques.

in a letter a ad June 18, 1880:

'I.

-"VAL'JAT';Gi'r An evaluation ci:he June 18, 1980 Rcchestar gas

'nd =lectric submi".al inclucing an independent review o-all lines cannec.ed directly tc the primary system, was canduc.ed.

The analysis assumed that the frash fr:ctian oi the fissian pracuc-s can-.

'.ned in the leaked coolant

~as re',eased ta the envirarmen The -'.cw r a a salac.ed was 50 gpm, car, espcnding ta the mas.

severe casa oi a C'ICS lat"awn line break.

Assuming a previous iodine spike, the arimary cao'ant activity was sa at 80 Cl/gn, l-131 "asa Cuival nt.

Ai.er a 20-a'outa delay, operator ac:ion ta isola"e the break

~as assumed a acc r.:ne cperatar de.action capabilities far the C'ICS let"awn line break include the vclume control tank level, the letdown line pressure and ilow, and cantinucus air.;,cnitar in charging "ump racm.

Using these assumptions, the resultant thyroid and whole body doses are 12 rem and 1

rem, respectively.

V!.

CONCLUS!OllS The doses calculated above are below the 105 of the 10 CFR Part 100 exposure guidelines and, therefore, ccmply with the SRP criterion.

The plant is adequately designed againsi failure oi small 1-ines carrying primary caolant outside containment.

R E. Gfnna.'Iuclear Power Stat an X'I-17 Sr=-'I". Ce.HEBATOR:UBE FAir URK'IPiRCCUCT!GN Steam generat.r tube failures aIIcw the escape

o. radioacti vfty 'rom

<<he r actor c:clant system ta the environrrent.

ScP T pfc X'I 17 is irtended to r view Ne radioiogicai consequenc s o; a st m generatcr tube aiiure.

The r view wiii encompass those design f atur s whicn I mit the release of radfcact.vfty includfng th plant technical specificatfans associated wi:h ccoiant activity ccncentratfons.

rr. p~II~~ C~r:-era Sec ',on:"0.3-" of 10 C.=R Par-.

."G r quir s that each a,plicant fcr a cons-'.ruc fon permf <<or apera.fng lic nse pr"vide an analysis ard evalua ian of the design ard per-.crrznc a-sm.ctures, sys:ams, ard components o.

're f cility wi r

the obective af assessing the risk a public heal "h.and sa.=oty resulting fran operation af the facility.

Tne ste m gener"ar

. be failure acciden.

is one cf the postulated accidents used to evalua.a the adequacy of these s.lectures,

systems, and ccmpanents wi:h."ospec to,ublic heaith and safety, In addition,

,0 CFR Parr. !"Q.II Provides dose

",uide',ines for reac: r si ing agains.

which calcuiated

• cciden: Case canse"uenc s ~ay be cmoared.

Tocic i.-2.C, "A--.cspner'.'c =ranspar-and Qif-.us'.an Character st.'cs fcr Acc'.den Analys's" pr vices

".;,.etaorcrcgicai data used to evaiua e

e of.si

~ doses.

Topic I-3, "Steam Cereratar.'ntac.i:y" ansur s

oat ac" ptabie

levels o

integrity o; the steam enerator ar maintained in accordance with current cri:aria.

'I I1 A "J1 'l S The review of the radiological consequences was concucted in accordance wi Standard Review Plan 1.=.6,3.

ne olart is consider d adequac ly des.'gned against a steam generator tube failur i= calculations show Cha-ae r sult-fng doses at the exclusion ar a and Icw poculation:cne boundaries are less than a small fraction o.

the 10 C;2, Pa.t 100 exposure guidelires, and ar within the 10 CR,?art 100 guidelines

-.or the case of a preaccident iodine sof ke.

'l.

='lAL"AT~OH The offsita "csa calcula.ions are based on.he x'sting plant tachnical specification limit For Ne maximum secondary coolant concentration, Tne plan-T hnical Specifications on primary coolant concentraticns allows

~1 an equilibrium value of 3 uCi/gm '! "ose equivalent and addi: onally allows operation in excess of.his value For oeriocs up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

';.'owever, since ro max',mum ccoIwt concentr'tion in the spike is set anc no yoarly limit is astablisned For.he numrer of such spikes, the licensee

'nas, effect'.veiy, no limit on the primary coolant conca,"t'i-ation.

'<le recomend that the licensee adopt a technical specification limiting the maximum allowable spike value to be 60uCi/gm 1 dose equivalent which has been established as the limit 131 in the Standard Technical Speci ications.

Further, the Standard

,echnical Soecifications for Westinghouse Pressurized Mater Reactors limit the amount of time per year when the equilibrium value of the coolant activity can be exceeded to 10" because of the higher consequences that result from a

oostulated ac ident.

The surveillance requirements incorporat d in the Standard Technical Specifications monitor the coolant activity at times of expected spi king conditions, namely, following pressure or power transients and during a spike in excess of the equilibrium value.

It is recommended that the duration and surveillance requirements of the Standard Technical Speci ica-tions also be incorporated.

Calculations were performed For the olant's existing "equilibrium" value of 3uCi/gm I dose equivalent assuming a concurrent spike from the amount of pre-existing failed fuel that would sustain the "equilibrium" value.

The dose calculated was 70 rem.

Nole body doses were also considered and found to be not limiting, in each case.

yI CO~lCl USI0~lc To sumnarize our recommendations on primary coolant activity resulting from our review of th s accident, the fo'llowing changes should be made to the plant' Technical Specifications:

l.

The maximum primary coolant activity during spi king should be 60uCi/gm 131 I dose equivalent.

2.

The duration of all soi kes during the year should be 10" of the operating time by incorporating the Standard Technical Specifications for duration.

3.

increase surveillance of the coolant activty during conditions of expected iodine spiking by incorporating the samoling requirements in the Standard Technical Specifications.

The conditions assumed for all the above calculations are summarized in Table XV-17-l.

Assumptions Hade in Anaiysfs of the Radiological C.nsequences

.t of Postulated S

am Gener tor Tube Failur 1.

Reac.or power

~ 1520 If+5 2.

3.

6.

7.

8.

9.

12.

Loss of Of.sita power =o'ilowfng the accidents Reactor Ccolant System (RCS) volume

~ o236 ft3 RCS Goerating Condi fons

• 22 0 psfa, 58G Homal Letdcwn Flow

~ 40 gpm Steam Generators Operating Ccnditions

= 770 psia, 514 F

1 Steam Cenera ors secondary sfde wa er volume

• 1681 It3 Steam generators secondary sfde steam volume

~ 2898 ft mer gency Feedwater 1 steam driven pump (4GO gpm) and 2 nator driven pumps (200 gpm)

Steam Generator Flow Rate

~ 3.13 x 10o lb/hr per steam generator Operating Slowdown Flew ~ 40 to 50 gpm per steam generator P imarf c"olant activity prior to the accident or 3. uCi/gram c= Cosa Equivalent

?-131 and 84/'5 uCi/g of total ac ivity, 13.

lodfne spiking factor or 500 abater the accfdent 14.

15.

Sec"ndary colant ac fvity pr'.or to he accident o= 0.1 uCi/gram o, Cose

"-"ufvalent.'-131.

fcdfne cecon~>anatfon fac or of 1G between wa er and steam 16.

17, 18.

19.

20.

Hetac~logfgal con/i.;cns carr spending to a ground level release o-,

4.8 x 10 " sac/m" at a disznc~e oi 450 m.

(Se ooic ii-Z.C)

Ho additicnaI

-,uel mel".:ng as a r suit o; any or the accident.

IC~ o$ ruel ac:fvity in Ne gaps,or cases wnere clad -;aflur is ass"med.

A11 releases

'.hr ugh the secondary side sa;ety and relfei valves.

1CO,CCG lbs. oi primary c olant leakage "o t.he secondary -i e o=

he ;ailed s:earn generator thr"ugh the,'afiad tub~e during the i frst 60 minutes

XV-ZQ MOIO<OGICAL CCnSKOUDCKS QF FU:-L O~~G:.?G ACCIOKaTS I.'fTRQOUCTI'3N

?ne saiety objective oi this topic is ta assure that the of-sita doses iron fuel da???aging accidents

<<s a result oi fuel handling inside

<<na aut-side ccntainrrent r

<<ell <<fthin the guidelire value or 10

".FR ?<<r.

100.

II.

R"-'/i:-'8 CRiT.R I~

Section:0.34 of 10 C'R Part

.=0, "Cantents of Applfcatfcns:

Technical Information," requires tha-e<<ch

<<pal icant for a canstructicn pem'.t or operating license provide an analysis and eraluaticn ar the design and periarnance ci struc:ur s,

systems, nd ccmocnents of the racility <<i:h the cb'ective of assessing the risk ta public heal:h and sarety resulting fr~

aperation oi the racility.

A uel handling <<ccident in the fuel handling

<<and st"rage racilfty resulting fn datnage ta ruel cladding and subseouert release ci radioactive material is ane ai the postulared ac icents used a

evaluate

.he ace",uacy oi these s:ructures,

systems,

<<and carpanents

<<ith respect:a the public health nd safety.

in addit';on, 10 CFR Par-

',00 provices "ose guidelines far ~ac. r si=ing agains.

<<nich calculated'c=ident dcse cansel,uences

.~ay be ca@pared.

R-'7"=0.:~t =i! TOP'.CS

oafc '.i-Z.", "A~mspher r Tr<<anspart ec Ofrfusicn Char c=erfs:ics.ar Acc;cent

.naiysfs prov'Ides

.he 1e ora lcglcal ata "sec or c

! utatfng the a=isite dose consequences.

The review o=

he fuel damaging accioen s did not consider fuel damage as a

result of drops of the spent fuel c sk or other neavy objects which can be carried either over

<<an open reac or vessel or the spent fuel pool.

Review of the droos of casks and heavy objects is covered in two SKP Topics, IX-2, "Overhead

'Handling Systems-Cranes" and XV-21, "Spent Fuel Cask Groo Accideats."

IV RK'll'A uU<QKLf.'iK>

Accidents resulting from the movement of fuel inside and cutsice containment were reviewed fol'owing the assumptions

<<xo-procedures outlined in Standard Review Plart (SRP)

Se tion 15.7.4 nd Regulatory "=uide 1.25.

The dose:o an individual frcm a postulated fuel handling accident should be "well within" the xposure guidelines of lQ CFR Part 1CQ.

('dhole body coses are also

<<xamined but are not con"rolling due o the decay o-the short-lived radio-isotopes pr ior to fuel handling.)

ignis is based on the probaoility of this event relative to other events wnich ar evaluated against 10 CFR Par.

1GO exposure guidelines.

The r eview considers single failure, seismic design md equipment qualification cnly when t'h e potential cense'quences mignt xceed the guidelines oi 10 CFR P rt 1CQ in the asence cf cont inment isolation

<<nd <<ffluent fiL=raticn.

The sys em design is considered to be acceptab:e if tne Iimi ing doses r

<<el 1 within the 1Q.FR lQQ ;u!oelines V.

VIAL'JATLQN ine assumptions used fn th!s evaluation are sumarized in Table X'I 2Q 1, Two cases of the fuel handling accident ~ere considered.

The plant's Technical Specifications related to fuel handling in the auxiliary building provide for tne required filtration of radloiodines.

That is, non-cSF charcoal filters are requited to be operable when irradiated fuel is handled in he building.

The surveillance requirements are sufficient:o provide reasonable C

assurance that the efficiency will be as high as the 90~ assumed in the staff' calculations.

Assuming that filters with an efficiency of 90~ for elemental iodine were used and that the fuel was damaged 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> af.er shutdown, a

limiting thyroid dose at the exclusion area boundary of less than 34 Rem was calculated.

For "refueling," inside the containment the plant's Technical Specifications require that personnel and equipment doors be closed and radiation levels be continuously monitored.

Ho filters are required to be ooerable; the HEPA and charcoal filters in the purge exhaust are "optional" and are not subject to surveillance to confirm their efficiency.

The staff, therefore, calculated the offsite dose consequences assuming that 100",. of the activity released from the fuel oool is released to the atmosphere.

At the same handling time of 100

hours, the calculated dose for release of the activity unfiltered would be 96 Rem at the

~MB.

In both cases, whole body doses were also considered, but are not limiting due to the decay of the short-lived radio isotopes.

Low population zone doses are lower due to the lower atmospheric dispersion actor.

- Vi.

CONCLUSIONS The limiting doses for the fuel damaging accidents indicate that the plant is adequately designed to mitigate the consequences of this type of accident.

TABlZ XV-20-1 ASSUNPTEONS NAOS EH A'QLYSES OF s riE FUBL OAW"E,'lG ACCEOKHTS ESSEOK A@0 OUTSEOE CGWTAE.'IDENT 1.

Reac.or icwer "bema 1

Z.

Clad failure of all rods in one o=

120 modules.

3.

Release oi gap inventory or all railed rods:

1G" E

10" Apole Gas 3G" B-'Kr 1.6o 4.

P aking

.." ctor Meteorological gonditions corresponding to a ground lev 1

." le se of 4.8x10-4 sac lrr'. a distance o-4:0 a.

(See Tooi" EE-Z.c},