ML20003B480

Submits Response to NUREG-0737 Items I.C.1,II.C.1,II.F.2, II.E.4.2,II.K.3.27,II.K.3.3 & II.K.3.30.Schedule to Comply w/NUREG-0654,Revision 1, App 2,will Adopt NUREG-0737 Alternative Schedule.Mod Proposal Expected by 810601

ML20003B480
Person / Time
Site:	Oyster Creek
Issue date:	02/10/1981
From:	Finfrock I JERSEY CENTRAL POWER & LIGHT CO.
To:	Eisenhut D Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-1.A.1.1, TASK-1.C.1, TASK-2.C.1, TASK-2.E.4.2, TASK-2.F.2, TASK-2.K.3, TASK-2.K.3.03, TASK-2.K.3.27, TASK-2.K.3.30, TASK-TM NUDOCS 8102120228
Download: ML20003B480 (47)

Text

Jersey Central power & Light Company Madison Avenue at Puncn Bowl Road Mornstown. New Jersey 07960 (201)455-8200 February 10, 1981

'/r. Darrell G. Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Wasnington, D. C.

20555

Subject:

Oyster Creek Nuclear Generating Station Docket No. 50-219 NUREG 0737 NUREG-0737 was transmitted to us by your letter dated October 31, 1980.

This letter required certain information be supplied by January 1981.

Our responses to several items are given below:

1.

l.C.1 - Reanalysis of Transients and Accidents - JCP&L as part of the BWR Ouners' Group has already provided suf ficient analysis regarding transients and accicents expected of the EWR design.

As stated q second fulI paragraph on page 1C1-3, NUREG 0737, the guidelines submi-ted by the BWR Owners' Group are being implemented on a trial basis. The results of this interim program shoul d indicate ^he adequacy of the guidelines submitted.

2.

l.C.1 - Reanalysis of inadequate Core Cooling - The guidelines submitted by the SWR Osners' Group have included provisions to ensure the adequacy of core cooling. As stated above, +hese guidelines are being implemented on a trial basis; and untii the comp letion of the trial program, no additional effort is planned.

l 3.

fl.F.2 - Instrumentation for the Detection of inadequate Core Cooling - On April 4, 1980 JCP&L responded to item 2.1.3B of NUREG 0578 (Enclosure B) by stat i ng that "...the ex i sti ng saf ety grade mu ltip le water level instrumentation is a direct and umambiguous indication of inadequate core cooling".

JCP&L is still in agreement with this position; and as such, no new modifications are l

planned.

1 4

II.E.4.2 - Con air. ment Isolation Dependability - Position 2 of this item requested that ne reconsider the classification of essential and nonessential systems.

We have completed this reassessment and have concluded that o - h%

l original evaluation as stated in our January 4, 1980 le.

'r is still valid.

Position 5 of this topic asks that we rev jew the containment p.cssure setpcInt g

that initiates containment isolation. Our present pressure setpoi,t is 2 psig l

//

l 1

t$10 9 Jerse, Cnal Pcwer & L 7! Company,s a MenTer et Ta Generm Ahc Wmas System

and is witnin 1 psig of normal operating containment pressure.

Position 6 of stipulates certain requirements for containment vent and purge valves.

this item As stated in our December 17, 1980 letter and accepted by your October 3, 1980 letter.e intend to comply.ith the staff interim position of October 23, 1979.

5.

II.K.3.2'7 - Common Peactor Vessel Reference Level - The description of the method used to accomplish tnis es described in my June 23, 1980 and September 26 1980 letters to you. The changes described in that letter have all been accomplished and all appropr i ate procedural changes have been made to reflect the original instrument ref erence level

c. well as the new common reference level.

6.

I I.K.3.3 - Repor t i ng SV i RV Failures and Chal lenges - There have been no chal l enges or failures associated with the safety valves during the period from April 1,

1980 to present.

one challenge to 'n electromatic relief valve during the period There has been from April 1,

1980 to present.

.nat incident took place on Jufy 17, 1980 at its setpoint approximately 8:35 p.m. when an electromatic relief valve lif ted at of 1050 psig and restarted at setpoint of 1000 psig.

7.

11 K.3.30 - Revised Small Brean LOCA model - 11 K3.44 Evaluation of Transients with Single Fail ures; and ll.K.3.45 Manual Depressurization with other than ADS. These items have all been assessed by the BWR cwners group and for arded to you by the group. We have reviewed their responses and endorsed them.

3.

Our response to the following items are included as Attachments to this letter:

1 1.A.1.1 Shift Technical Advisor - (a) Short Term - Attachment 1 (b) Long Term -

?

II.K.3.17 ECCS System Outages 3

II.K.3.21 Restart of Core Spray 4

5 til.D.3.3 Modifications to Accurately Measure f 6

Ill.D.3.4 Control Room Habitability 7

ll.F.1.2.8 Accident Monitoring instrumentation 8

lit.A.2 Meteorological Data would also like to use this letter as the vehicle to update my January I

6, 1981 response to you. A further investigation by my staff has revealed that we may be unable to meet the NUREG 0737 schedule on two additional topics. describes the schedule delay for commiting to the plans and schedule for meeting the meteorological data requirements of NUREG 0654, Rev. 1, Appendix 2. describes our position on NUREG 0737, item II.E.4.1,

[

(

Dedicated Hycrogen Penetrations.

Our emergency plan Citem ill.A.2) will be submitted in acccrdance with the equiremen+ of 10CFR50.47 under a separate cover letter.

I This suomittal reflects to the best of my knowledge the status of those NUREG 0737 items discussed above.

l If you should have any questions concerning this submittal please l

contact Yr. James Knutel (201-455-8753) of my staff.

l i

Very truly yours,

\\

4 71. 4

\\

Ivan R. Finfr ;k, Vice Presic e'

$ worn and subscribed to before me this

/O _ cay of h 1981.

YA

. Of>

Notary Pu5iig///

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l 1931 STA Recualificatien ?rcera:

vill censist of a =inimum of fif:y (50) hours of for=al classrec The progra:

1981 with classes scheduled approxi=ately twice training presented th:cughou:

l a menth.

I

!n addi:1cn, self study assignmen:s will be given and reviewed on a periodic basis.

training.

the f ollowing topics will be covered during :he classroc:

As a =ini=u=,

1.

Plan Isergency Frecedures 2.

Energency Plan and l=plementing Procedures 1

3.

Transien: and Accident Analyses 4

Mitigating Ccre Da: age 3.

Review of significant LIR's 6.

Modifications to plant systens af f ecting saf ety vill be used to deter =ine additional : pics to Results of the 1930 STA exa l

supple en: the above.

Lecture Series Qui::es Periedic qui::es vill be ad=inistered c evaluate the eff ectiveness of the retraining.

STA's who A performance standard of 507, or more should be established for each qui:.

his perfor:ance standard should cc=plete a remedial review process did not mee:

casisting of:

Review of thc lecture :sterial :cvered in the classrec= : raining or 1.

self study.

instructor.

Review of associated reference ca:erial identified by the 2.

3.

STA shall undergo another wri: ten qui: or an oral evaluatien administered If the reexamination by the Saf ety Review Manager and/or the instrue cr.

is :cepleted satisf ac:orily, :he STA should receive credit for cc=pletion of the lecture or self study caterial.

If the reexaminatien is unsatisfac: cry, the SIA l

the evalua:Or should rece==end to the Safety Review Manager that should be rencved frc the duty rooster unit until satisfactory knowledge l

is demonstrated.

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ENCLCSURE 1 to ATTAC.91ENT 1.-.

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L5 Aug. 27 Fx. 7essel, Inter als, Fuel 'esign Ccat&-ant PRI/SEC 16 Sept. 3 Centa4--ant PRI/SEC Electrical Distributien 17 Sep*,. 10 Electrical Distributien tegineered Safeguar.s 16 Sept. 17 Engicea d Safeguards Fx. 7essel In;tr=ents 13 Sept. Eh Ec.

.ssel I:str=entatic:

Heat dans./?.F. Refr-sher 20 Oct.

1 E n t hansfer Eest dans./F.F. Ref. sher m

21 Oct.

S

• he:.:al Hydraul.1:s Eeat dans./?.F. Ref.msher 22 Oct. 15

~he.~al ~inits Ifue & Frecess Instru:ents 23 Cet. 22 iuclear k Prccess Instrunents Fx. Frctectica Systen ik Oct. 29 Re. Protecticn Systen ECCS 25 icy.

5 ECCS Plt. Sansient Eehavice 26

Tcv. 12 Plt. Sansient Echavier-Fit. Sansient 3ehavier-27

Tev. 19 Fit. dansie t 3ehavier Plt. dansient schavice i

26

Icv. 26

?x. ~heery Review 29 Dec.

3 I:erge:cy Plan & Frecedur-s Reviev 1

30 rec. 10 ficce Reviev i

1 31

'ec. 17 Exanizatice 22 rec. 2h icce l

33 Dec. 31 Re-Exa d nation All classes are scheduled for 'a'ednesday 3:00 A.M. - 3 :00 F.M.

With sc._.al lunch treal at 12:00 ccen.

(Rev.2) 1 l

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-e NI~ACHMENT :

OESCRI?TICN OF ??.E GPU I.CNG-TE?M STA ?RCCRAM The GFU STA program is described in the folleving enclosures:

Enclesure 1 - Positica Oescriptien - Qualifications of the Shif: Technical Advisor - Educa:1:n and Training Requirements.

Since the selection and ::sining of STA's for the G7U program began pric to the issuance of the I:20 document, " Nuclear ? cue Plan: Shift Technical Advisor - Reccc=endations f or ?csition Oescriptica, Qualificati:ns, Educa:ica and !:aining" (Revision 0, April 30, 1980), sc=e of the require =ents of our the intent progra differ f cm the I GO program, although we feel that we eet of :he program :ecc== ended by 1570. Spe.cific cc=: arisen to the !!GO docu=ent is =ade only in the education and ::aining sec_ien sinca sc:a pc::icus of this secticn are still being develcted for our program.

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? siti:n: Shif t Technical Advisor Recorts to: Plan: Analysis Fanager Accountabiliev Cb4ctive:

The Shif t Technical Advisor will be respcusible f or advising :he shif t super-visor en technical questi:ns and for senitoring operation to ensure safe, reliable and efficien perf:rnance.

~he STA vill evaluate the technical perf or ance of the reac:cr plant during pcVer, s:s :up, shu:det.m and refueling i

nodes as well as peri:ds of cold shu:devn when special nain:enance evolu:icas

~he STA vill be available :o advise :he on-shif:

ce :ests are in progress.

operation supervisor of :he ac: ices necessa:7 :o naintain or regain a safe, stable cpera:ing state.

~he najer functi:n of this posi:ica is :o use addi:1:nal engineering and

ecH4 cal expertise available to che shif: cperations s:aff.

~he STA vill l

perfor= the functions of :he posi:1:n independent of the opera:1:ns shif t organization.

• he STA vill no: displace any of the presen: functions of the shif organi:a:icn bur ra:her supplanen:s the capabili:7 of :he shif t :o Operase :he plant safely and reliably. A seconda.:7 func:icn is :o assist in nain:a ning and incroving plant saf ety by evalua:ing plan: perfor=ance and pr: vide eif active feedbac!: =c Technical ?~~"ns.

31:enstens

• his posi.ica centributes to :he safe, reliable and efficies: cpera: ices of l

nuclear genera:ing s:a:1cus with a value of SfC0.?..

Nature and Sccee to the Six Shif t, Technical Advisers fr:s each etclear generating sta:icn report Plan: Analysis Manager along vi:h :hree iTA trainees 2r.:n each sta:1:n, two Plan: Analysis hgineers frem each sta:icn and five ?lan: Analysis ~ngineers a:

headquarte rs.

~he STA vill seni::: the response of the plat.: during ::ansien:s and acciden:s, de:em4"e if :he plan: is respending vi:hin sau:7 lini:s and advise the shif t supe: rise: Of abnornal s:a us and the correc:1ve ac:icns needed :s prevent adverse c:nsecuences.

~he STA vill acci:or :he :echnical perfor ance of the plan: during steady s: ace apera:1:n and during planned evoluti:ns such as startup, shu:dev. and pcVer changes.

~he STA vill seni:Or :he trends of cri:1 cal parane:n.rs to de:ernine if the plan: is operating within design limits. S e STA vill review and eval-ua:e all Opera:icns conducted during :ne shif: that could jeapcrdize the safe i

cpera: ice of the plan:.

l eM I

2 The STA vill advise :he Shif: Supervisor i==ediately of any rece==enda:icus or evalua:1:ns during the shift and will keep :he ?las: Analysis Manager apprised of abnor=al plan: condi: ices in a ti=ely =anner appr:priate to the condi:ica bu: not :o interfere with the safe operati:n of :he plan:.

The STA vill meni:or the' readiness of :he Plan: ? ocection Sys:e=s, i.e., :he R?S, ICCS, ESyAS, :o perfor= their design func:icn and be especially alert :o tes:ing or =ain:enance evolu:1cas :ha: could preven: these syste=s f rm perforaing.

The STA vill :ake a :achnical evaluacion of all significas: events : hat occur during the shift. The cbjec: ef :his evaluati:n vill be :o de:er=ine :he cause of the even m:d determine :he : rrec:i"e ac:icns necessary to prevent r e-occurrence.

The STA vill peri:dically =cnitor selected plant evolutions m:d surveillance

es:s :o evalua:e :he effec:1veness of plant pr:cedures.

The STA vill be aware of significa:: Operating even:s :ha: occur a: other nuclear power plan:s and :he pctential applicabili:7 Of : hose even:s :: the Operation of the plan:.

'"he STA vill reviev TCN's and Special Operating ?::cedures :ha: are i=plemented before receiving nor=al Technical Fu=c:icns review for :achnical adequacy and l

safaty censiderati es.

T=e STA vill pericr= a shif relief af:e: each shif t eff ec.1 rely ce==u 1 cat 1=g a c:=ple:e plant status t: :he en==ing STA.

?rier to assuaisg the respensibilities Of :his positien, each individual will be required :o c:=plete a ::sising progra= and be certified by writ:en and oral ev'-d a:1:n to =ee: the qualifica:icns sca:ed is A::sch en: A.

After initial certifica:ica, each STA vill be required to participa:e in a ::aining progra=

and fulfill :he require =ents f or ob:aining an NRC Seni:r Reac:c Cpera:Or's license. After 0b:21:1:g a license s:stus each STA vill ps::icipate in ::aining l

pr:gra=s to 9a n:ai Licensabili:7 and : increase : heir proficiency as STA's.

d This posi:1cn performs i:s responsibilities en ::: acing shif:s en :he si:e of l

the nuclear generating s:a:1:n ei:her in :he :::::01 ::c= or in :he plan vithin ready access to the c:nt:01 ::c=.

l

? rim ital A : untabilities 1.

Meni::r and evalua:e :he perf:r:ance of :he plan: duris; s:eady s: ate cpera:1:n, planned evolu:1:ns, ::ansients and ac:idents.

2.

Advise :he Shift Supervisor of abn:r=al :endi:icns and of ec ective' acti:ns needed :o preven: adverse : rsecuences.

3.

M:ni::: :he readiness Of plan protec:ive sys:e=s.

4.

Ivalua:e significa:: even:s :ha Occur en shift.

a 7

3 helosure 1 Meni:or selec:ed evolucions :o evaluate :he effectiveness of plan:

5.

procedures.

6.

Se aware of significan: events :ha: cc ur a:. ocher nuclear power plants and of :he po en:ial applicability of :hese even:s.

7.

Review TC:!'s, Special Opera:ing ?;ocedures and o:her procedures that are needed en an urgent basis for :echnical adequacy.

l 3.

Perfors a shif: relief with :he encoming STA each shif: affec:17ely cc==unicating a cecple:e plan: sta:us.

9.

?e:fors special analyses of plan: :echnical perfor=ance as he considers var an:ed or directed by Plan: Analysis.

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? 1or to entering :he full ti e STA ::aining pr:gra= STA candidates vill have the folloviri qualifiestiens:

A 3achelor's Oegree in Engineering or Science vi h sufficien:

c urses :o p : vide a sound backg; und for unders:anding :he design and operation of a Nuclea: ?:ve: Plan:.

Two years of experience in the Opera:icn, design er : her engineering l

supper: Of a Nucles: ?:ver Plan:.

i Equivalency of the above :vo require ents shall be dete::ined by individual evalua:1cn of each candidate's specific qualifica:icas.

Af:er ::cpleting the ::aining pr:gra= and prior to assu=ing duties cc shif: the STA vill have the f alleving qualifica:1:ns demenstrated by sa:isf actory ccm--

ple:1:n of a writ:en ev' d nati:n and oral board:

tech-

- Eave an unders:anding of the fundanen:als of nuclear pcuer plant l

nology such as:

Reacter Heat Transf er - Mechanis=s of heat transfer fr:s nuclear procsss and accident to fuei and f uei :o ecolanc during steady state, ::ansien:

f condi:1:n.

l Core Neutrenies and ?:ver Cistributien. - Neutren Transport, Nuclear Of Global ? ver and local ? cue: Ois:ribu-Hea F : cess and Measureces:

1:n, d :atiens of Analysis Medels and Measuring Oevices.

7uel Damaze Mechanisms - Fuel Mel: Causes and Threshold, Clad Cxidaci:n and Hyd:1 ing, Ini:ia:icn Mechanis=s, Oa= age C.,csequences j

and Mitigation.

I Materials and Chemistrv C:n rol - Ma:erials used in fuel, 1

i Rese::: ?lan:

clad, RCS, Sr.,

c:ter sajor cenpenen:s and bases for s41ec:1:n; che=1 cal cenditiens cecessary to preven: degrada:ica of cc=penents; corresten l

j mech...is=s both local and general; ef f ect of :henis: 7 c:n:rol er radio.:gical :en::al.

Svsta= her edv-.amics and ?luid 71:v Charse te:-is tics her:ody~.2=1c

nci:1:ns 2: Ort:ical par:s :s pri=ary and sec:ndary systers. S ingle phase and rvo phase flev characteristics; fl:v c nditions Of f::ced and natural cir:ula:1:n; unusual condi:i:ns : hat exist during accidents.

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T Encicsure 2 systes used in Instru=ents:1:n Systems and Cemeenents_- Measure =ent signal :::ns=1331:n and p cessing :echniques, nucles pcue plan:3, devices, c:=puter applica-principals and construe:1 n of measure =en:

l tiens and limita:1:ns.

>!echani:sl Svste-s and C:ecenents - Cc=penen:s used in nuclear pcwer plan:a, basis f or applica:1:n)sy2:es charac: eristics and in:eractica.

Elect:1:al Distributien Svstems and Cc.renen:s - Internal and external f

dis:ribution sys:ess; serv:ce, vi:a1 and ins::ucen:a:icn pcuer distrihu-reliabili:/ and failure :cdes.

tien and ::cpenen:s; sys:en and cecpenen:

r and Control - Radiati n scur:es, ce:heds of :: anspor:

Radia:1:n 7:snsect:radia:1:n hasards and safe:7 me:heds.

anc c:n:ainment; respense and be able to recognize deviation fr:s Unders:and expected plan:

respcuse during s:eady sta:e operation and durin? planned l

ner:a1 plan:

f

ansi:1 =s f := cne steady s:ste cenditien to another, e.g., star:tps, Understand li=1:a:icas of redels er plan:

shutdevna and. lead changes.

f da:a used :o ;;edian plaa: respcase.

respense during Have :he abili:y c ree:gnize deviatien f :s expected plan:

10CA's.

unplanned ::ansients and accidents, e.g.,

reactor crip, iden:ify :he causes of abncr:al response based en sy=pcots 3e able :

f derived fr:s basic plann pars =e:ers and rec==end :he proper :crrective ac:1:ns :: Le: urn :he plan: to a safe stable c ndi:icn.

Have a knculedge of :he functions of the plant p:ctactive sys:ess, :he 1

lineup of :hese sys:e=s in :he standby : de and when called upon to opers:e, the effec these systa=s have en plan: respense when opera:ing, cc ec:

l al:ernate at:i:ns required if these systa:s =alfune:1:n, and 0:ndi:

1:ns I

necessary to :ermina:e pro:ective func:1:ns.

Unders:and :he effec:s :ha: $jc cen::ci sys:e=s have en plan: respense and =anual mede and when :alfunc:icns ec:ur to the in nor=al au::=a:1:

con : 1 sys:e=s and principle inpu:s :o the c:n::ci sys:e=s.

i principle plant sys:e=s have en plan: respense Unders:and :he effec:s :ha:

during nc::a1 Operation and during =alfune:icns of =af cr c =penen:s in the sys:e=s.

Have an unders:anding and kncvledge of basis of plant e=e:gency, abnor=al and in:egra:ed plan: Opera:ing pr:cedures and a sufficien: kn:wledge !

ha: unders anding.

sys:e= and :=penen: Opera:ing precedures :o supper:

Have a kn:wledge of :he plan: License :equire=en:s and Technical Ipecifi-l

a:i:=s.

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Have an unders:sedi:3 of the 54fe:7 Anal 7 sis perfc=ed := su~ or-f Opera:icn especially vi:h regard to the 11:1:3 assu=ed n plan-ph :

para e:ers. *Jeders:and the assu=c:1:ns used in :he 54ft:7 Analysis

he lici:stions of the a:alysis.cdels.

and i

Have a k=culedge of the :: di:1:ns tha: require i. ple=en:a:1:n of :he l

var.cus levels of the E=e:genc7 Plan and.the ac:1 cs required :o

• ncu:e 1 as well as :he specific ac:icns required of :he STA during I

-p le=en:a:1:n.

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G7" SHIM ICCi!CAi $CVISCR EDUCAT!CNAL RECUI2_"I';TS INFO has p cpesed educational require =ents for Shif: Technical Advisors in 1:3 April 30,1930 Nucles: ?cwer Plan: Shift '.'echnical Adviser - Recc==enda-

1cns for ?csition Oescription, Qualifica:icns, Educati:n and T:sining -

Revisien 0.

Inis docu=ent has been reviewed by the NRC staff and has been spp;cved as an acceptable :ethed of i=plementing an STA program that will zee: NRC require =ents.

The following lis: delinea:es the al:ernate methods that GPU will utill:e in their STA prograns.

Ci?O See:1cn 6.1.1 Ci?O Equivalent Sub4ect Cen2c: Heurs Semester Heurs Mathematics 90 6

Che=1s:ry 30 2

Physics if0 10 The 0?U STA's will be required to have as a =ini=um college credits ree ing :he INFO guidelines.

IN?O Section 6.1.2 Mathe=stics - through 90 6

c dinary differential equations and use of La? lace transfor=s The 07U STA's will be required to have as a nini=us cc11ege credi:s

=eeting the IN?O guidelines.

Rese:cr Theory - nuclear 100 6.7 pnysics, 11ffision :heory kinetics and reactivity feedback The GPU STA's will be required Oc have as a mini =us ecliege credits see:ing the IN?O guidelines.

Resctor Che=1strv - as 30 2

rela:ed :o reactor systems Oue :c the specific applica:ica of :his course, G?U 3TA's will meet the require:ents by a ecurse tailored to the specific chemistry applica:icns of thei: particular plant.

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IN?O Equivalen:

Semester Hours _

Contac: *dours Sub4ect 1.7 Nuclear Materials -

40 l

Streng h of =aterials; Reactor :sterial properties the strength of sacerials requiremen: by a college G7U STA's vill meet The reactor =aterials requirement will be =et either by a credi: course.

college credit course or by a special course covering both general and plant specific reactor racerial properties.

Thermal Sciences 120 3

(for nuclear systems)

Ther:odynaaics Laws of S er:odynamics Properties of Water & Steam Steam Cycles & Efficiency Fluid Dynamics Bernoulli's Equatica Fluid Fric: ion & Head Less Elevatien Head Pu=p & System Characteristics Hest Transf er Metheds of Heat Transfer 3 oiling Heat Transfer Heat Exchangers this requirement either by separate general engineering GPU SIA's will =eet ecurses covering the above subjects or by an integrated Reactor Syste=s Heat Transport course or courses.

4 Elec:rical Sciences 60 Electronics (Circui:

theory, digital elec:ronics)

Motors, Generators,

' ~

Transfor=ers, Switchgear Instrumentation & Con:rol Theory this requirement either by college credi: courses G?U STA's vill meet covering :he subjects or by ccmbinatiens of cellege courses and specific courses :silored :s nuclear plant elec:rical, electronic acn-credi:

and ins:: :::en: systems.

1 I

3 1N70 Equivalen:

Sub4ee:

Centact Meurs Senester Monts Nuclear Ins:rrentatien and Centrcl 40 2.7 Radiation Detec crs Reac:ct Instr =en:stica Reactivity Con:rcl & Feedback G?U STA's will nee: this requirenen: either by a ecliege credit course

(

in Nuclear Reac:cr Instr =en:ation er by a general instrunentatica course supplerected by ncn-credit nuclear and radiation instru=ent instrue:1cn.

Nuclear Radiatien

?rceecticn 5 Health Physics 40

2. 7 31clagical Effects Radiation Surrey Instrunentaticn Shielding GPU STA's vill neet this requirenent ei:her by a ecliege credit course or by a certified Health Physics course.

All the abcVe contact hours are censidered as guidelines only and nay be altered dependirg en course applicability, level of instrue:1cn and the STA's previcus educatica and e:cperience.

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4 IN?O Sectien 6.2 Applied Fundamentals - ?lan: 3pecific The GPU STA ::sining program will meet or exceed :he requirenents of this sec:1cn.

IN?O Section 6.3 F.anagement/ Supervisory Skills I

l Since the position dese:1pti:n for GPU STA's dces not include any or supervisory functions this section vill no: be nanagement censidered as a requirement except for the probles and Decisional Analysis course which vill be required.

!N70 Sectiens 6.4, 6. 5, 6. 6 and 6. 7 Syste=s, Ad=inistrative Centrols, General Cperating ?rocedures Plan:

f' and Transient / Accident Analysis and Emergency Frecedures.

II.e G?U STA training program will meet or exceed :he requirements i

(

of these secticus.

l 1570 Sectica 6.3 Simulator Training The GPU STA training p::gra= will :eet or exceed the requirements of this section in regard to the at=ber of sinulation hours, however, the list of sL=ulator exercises is considered :nly as a guideline as to the type and spectru= of events to be included in the simulator pr gra= and specific list of events that must be included.

not a l

l EN?O Secticn 6.9 Annual Requalificacion ?:ogras Ic.e G7U STA training program vill eet or enceed :he requirements of

• his section.

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o ATTACHMENT 3 NUREG 0737 ITEM li..<.3.17 - ECOS CUTAGES -

The Emergency Core Cooling Systems at Cyster Creek censist of the Ocntainment Spray Systems, Core Spray Systems, Auto Cepressuri:stien System and the Emergency Diesel Generaters. The Information provided below will provide a cuantification of histcrical unreliability due to maintenance and test cutages i

ter the ECOS systems. Preventive maintenance activities are performed when l

technical speci fications either de not requie the availability of a system, or l

do allow Oceratien of the system in a cegraded mode in ceder ?c accomplish preventive maintenance activities. Since no recceds exist en the duration of such cutagos, *ney are emi+*ed frcm Tne data given below.

The following are explanations and descriptions of sources fcr the data to follow.

Main + mance Activi?v Outages - These cutages and times were ecmpiled from rec':cas en Reocetaole Occurrences (RO) cated S/75 tnrougn 3/80.

Its cut ge times cculd not be determined frcm centrol, ccm legs iney were estimatad to the oest of our ability.

Estimated times are i denti f ied with an asterisk *. See Attachment I for date, duration, cause ccmpenent, and corrective action. Note:

Duratien is time of discovery to time corrective acticn was ccmoleted.

Surveillance Activity Outages - Surveillance testing which renders the asscc i at*c systems cut of service were determined and average performance times were cetainec frcm tne respective depar*ments responsi_ble for performance of sucn testing.

The f requency requirements of each surveillance was determined and total cut of service times for a five year period were calcutafec.

)

Total Outace Time - Sum of the maintenance and surveillance activity cutages.

The following is a summary by system cf total cutage times and a breakdcwn of survei l l ance activity cutage time. Enclosure 1 will serve as a breakdown of maintenece activity cutages.

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Cottl AltifilillT SPilAY "0UltCli DA'lli IHiltATioff CAllSli Cutti'0Hl.fi r Coltitiri l VI: AC i loti Ito 26 9/23/75

• 2 lifts l' allure of I:mergency service water GI! Itelay tJew selay Instatie.l pump t o automatically start Cit 2820 I1111 AAll 110-76-9 4/1/76

• 1 Ilit The trip point for drywell high Ilarton ITF The switches were resel to pressure switches IP15ft anI 150 278 within tech spec Iimits found out of Iech spec Iimit's were ItO-76-10 4/1/76

• 411!!S.

Hisaligned starter contacts on the Cli7700 1he contacts were renligneil.

460V breaker which result ed in firea k e r disconnect ing one of the three phase for the motor on valve V 15.

110-78-11 6/22/78 2 lillS.

DG load sceptencer t imer (16K4B)

Gli Cit 2820lt Timer was reset to trip within limits failed to start containment spray

_emer. spray _jnimp

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110-78-30 11/26/78 3.5 lifts I* allure of containment spray 11 Cl! Ilreaker Circuit brealers for t he AK-2A-50 affected pumps were cleaned pump Sic to start during auto actuatioidest

_ and disassempted.___ _ _ _ _

I 110-78-32 12/4/78

+2 liitS Failure of tortis spray valve tJA Va lve wit or lire 11< er was sacked in pioperly and V-21-18 to funct ion motor overload seset Ito-79-4 3/1/79 21 lifts.

During surveillance, containment f4A 1he leaky piping was replaced j

spray system It was found to f>e

'l in degraded mode dine to leak in

__s.y s t em 110-79-21 6/20/79

• l litt Iligh drywell pressure sensor IPISA ITI 11ai ton Inst a oment IPISA wae ucsted i t rip set t ing 2 psig ind press to trip at tech npv ve switch 110-79-22 6/27/79 I litt Containment spray system I was f4A 1he leaky nipples werc found to be in degraded mde rep _l ac ed 110-79-32 8/14/79

• 3 liitS Containment spray'/small pipe leaks FJA

't h e leaking nigiples were replaced s

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SullitCII DIiftA f lON itO - 7 6 -.1 1/73/76 I l 6.SilitS I!mergency diesel shutilown due to Itaill a t o r The 1-1 emergency diesel t rip of low cooling wat er pressure l'ilD make generator railiator will sene.or due to leak in raillator be repaireil.

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lto,76-5 3/3/76 12 liitS l'a ilur e o f D.G, # 2 t o e.uppl y Ivower l'I CO Contact wipe was ine s easeil due to excitation not being Westinghouse to overcome v iili at ion i nih ered a

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supplled causeil by fault y cont act Voltage ite-by tlu diesel generator

- - -.O lay-1876091- - - - -

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ItO-76-16 6/8/76

') littS Sequence fault which was initiateil NA Iliesel generator mainsf actui er was contacted to investigate by no pinion engagement caused #2 l>G to not start t h i s_ p r,ob l_em. _ _ _ _

110-76-71 9/17/76

• 2 lift 9 (TK3) inst rument tirl f t causing load IIl a st i c

'limer was i ee.c t to within sequencing time to be out of tech s t oli no t tech sper limit s l

spec limits Corp. Plodel 2114PD itO-7 7-1 3/18/77 105. 5111tS Trip of IC breaker was caused by Cable lhree cables wit hin t he e phase leads were ground between Gli Vull<ene grounded conluit wer e IC and DG#1 breaker 51-58061 replaced.

l e

NA lil ec t r ica l tests showed 110-77-17 7/2R/77 13 liitS Operatlon in a degraded mode when i

no inhesent problem power was lost to core spray 1"Jnips connected to C bus f(0-78-1 1/3/78 5.5 lilts DG#1 t ripped due to engine high Square D Switch was replaced by t emp condit ion caused by fault y 9025 new switch BGW352 i

temp switch l

series B 110-78-31 11/30/78 7.5 liitS l'ailure of diesel gen. #1 to ITCO ItGD-10 was replaced by assume loa 1 during loss of power West a nebuilt unit t est due to mal funct ion of 1:.Gil-10 1876091 GI! type AV-2A151 Dreaker i

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0tt A. I IOrl S(Hilttii l>Alli ItO 1 1/2/76

' 5. 5 1111 S Failur e of one higli iltywell pressuti Itartim IIT Itep l a c eme nt of mercury switch associateil with core spray ikulel 278 swi t ch bulb in t he syst em t o act ivat e c< pial t o or less pressure switch than 2 psig litV4611)

Ito-76-8 3/26/76

• 2 liit's l'ailure of one high ilrywelI pressine lia r t on l~lT lteset ItV1fC to trip at swit ch associat eil wi t h core spray th >ilo l 178 2.0 psin syst em to act ivat e e<1ual t o or less than 2 psig itV46C 110-76-14 4/23/76 63 IlitS Operat ion of core spray syst em i Ingersoll

,l' ump sea l wa s r epl ac eil in elegraileil motic due to c acke.1 Ita nil f

inboaril pump seal on N7.Ollt mechanical packing carbon rotating l

washer 110-76-15 4/25/76 28.5 IlllS Operat ion of Core Spray Syst em li flA freplaceil af fec t eil por t ion in ilegratteil moile as r esult of leak of sensing line anil ilye in sensing line to sen3or itV40ll penet rant test all sensing iine fi t t ings for con e spiay system igigic insttument i

It0-77-5 3/23/77 1.5 littS

' Core Spr sy System I parallel flA Iru reaseil t est inn f a cituency isolat ion valve V 15 became in-to weekly until refueling operable in the open position when outage when more extensive elect rical anil mechanical motor operator breaker t rippeil.

maintenance activitles can Valve operator was checkett an<l ilet ermine cause of breaker bench test eil wit h no appaient cause fotenti for t ripping of breaker to t ip. _.. _ _ _ _

ltO-77-16 7/27/77 11.5 liitS Core Spray System I parallel NA Clicuit breakers for motor isolat lon valve V-20-40 falleil to opei n t oi s on V-70.10 anil V 15 we r e t est eil anil fully open whea t he motor operator circuit breaker t rippeil.11reaker r c <.c t. POftC ili rec t eil va l ve overloail set t ing was founil to be V- /O-tu t o be i nspec t eil sie t t outage anil to investi-24 amps below recommenileil set t isig.

. gate ta} sing set pnint ut

' hs cakes' a

l' age 2 CORii SPRAY SYSTDI (CotTJ)_ __ _ __ __

' 1 50ilitCli DA fli DuttATION CAllSli CotlP0fil:tJI rDHitirliVli ACi1OtJ

[t0-77-23 9/8/77 5 liitS Core Spray System I parallel Ilreak er Iteplaceil bi caler isolation valve V-20-15 motor General eitcuit breaker trippest.

InvestI-til ec t ric gat ion revealeil magnet ic t ri -

Tl!I:-116A i

t i

for "n" phase of breaker -as In amp t ripping at 45 amps 3:.st cail of 60 ampp__ _ _,_ ___,_, _ ____,___

1t0-77-11 4/19/79 5 lilts Core spray l a,p in01 A faileil to One t ime l'use was replaceil anil all start iluting routine pump class K-5 fuses on engincen eil sa fe i

operability test.

A fuse failure fuse maile guanil systems will be creat eil an open in the closing liy Shawnut repl aceil ilur ing. nex t re fuel i nn circuit of the control power useil 15 amps outage.

'j to close the pump motor breaker 120V or les-j 110-79-17 5/17/79 3. 5 1111 5 Core Spray System Il parallel 11reaker Itepl a cement of circuit i

isolation valvo V-20-21 became Cencral bre.sker with new one of same

l Inoperable in the open position lilec t r ic t ype anil moilet when the motor operator breaker til00 line t r i ppeil. Apparent cause for circuit breaker lo trip was circuit lireakei breaker internal centacts making poor t ype l111

~

contact in the "ll" phase thus with ailjust creat Ing an overcurrent conilit ion able magneti trip only flotlel 'llir,

g 136t11010

[

I Snublier was replaccil wit h 110-79-18 5/17/79

• 6 littS Failure of hyilraulic scrubber to Bergen I

an opesable spare.

lock up in compression on Core Patterson Spray System II. ~!he compression flyilrau l ic poppet seat lug surface anil poppet Sl;ock anil i

spring were founit to be ilamageil.

rway arresti e

type llSSA-If 6

i RO-79-27 8/7/79

• 18 liitS Six seismic rest raint s for the 6 flA Itest ra int s wer e v est or eil inch Core Spray System 11 test to t heir slesign conilit ion line were founit to be in posit lons anil/or elesign moili ficit t o other than requireil by original provlile casier placement slesign criteria, or hail f a i t eil.

o f a nc ho r bolts.

h.

Page 3

'?

COltli SPitAY SYSTCil (COf1T)

COfII'Ofit til ColtItl CIIVli ACiitn1

_ CAllSl!_ - - - _ -

SOIlitCl!

OATli DilltATIOff 110-79-78 8/7/79 1.5 lillS Core Spray System i parallel flA Suiveillance proceilures weie reviscil to inc l uile isolation valve V-20-15 liccame 4pecific gulilance to i nope ra!>1 e in flie open posit ion more a vo lil inailvert ait operat ional when tiic motor tireaker t rippeil.

Apparent catise of litcaker to trip mailes.

was the inailv er t ant initlation of a valve close signal slu Ing the perloil when t he vaIve was st 1iI j~

st roking open thus canising great er than normal start ing current s to t

hU S?n by lu eaker FIA lose was i cplaceit anil job RO_79-37 11/3/79

• 1.5 lilts Core Spray lloost er l' ump f1ZU3It Sys.

i oriter init int eil t o check other fuse: which an c manipulat eil l

11 failure to restart iluring sur-veillance test ing.

Apparent cause was loose fuse in t he breal,er iluring surveliIance iest ing

<nf safety relateil systems.

cont rol power.

110-80-19 5/17/80

% liitS Core Spray System I anil if itclief Helief Itclief valve set point s weie Valves.V-20-21 anil V-20-25 setroint-valve reset to 350 psig. Itot h valves founit to be less conservative 2" x 3"-300t are incli eleil in new ISI pump than specificit. Set point s shoutil I.onergan an.I valve inspection program.

.~

were have been 350 psig, they were founit to be 300 ant!.175 psig respectively_

110-80-29 7/11/80 9 liitS Core Spray Syst ems i f.11 al t ernat e fla r t on ITI' immeiliat e ly veri ficil high core spray pumps were prevent eil lloilel 288A ilrywell press iliil not from starting automatically when

.5-9.5psig exist.

liese t pressure switch ilrywell high pressure sensor ItV16ft Atlj ust abl e to 1.93 psig.

acistateil at a more conservat Ive range setpoint. Operator stoppeil primary aiul alternate core spray pumps ilne to ilesign of logic ciretilt alter-nate pumps were prevent eil f rom starting automatically until trip j

signal cleareil allowing breakers to be reset, I

l'a ge 1 1

C0111! SI'IIAY SYSilill(COffi)_ _____ _ _ _ _ _ _ _ _

50111tC11 11Nili fit fitAT loff CAllSli Corleofil:ta l' Coltiti:0Iiyr At:ll Ort replaccil.

ilA l'i pe nipple was 1t0- 80_41 9/5/80

• 3 l111S Core Spray System II was removeil past memo's liave licen i s socil from service to repair learing concerning t his pintelem.

vent line on core spray Imoster Conc ee n wi ll 'bc aililresseil pump flZi13D. I'robable cause was in contractor iniloct r ina t ion personnel error when pipe useil as cove r ses.

j footholil which loosencil the joint.

110-75-20 7/17/75

• 2 lilts 1: allure of one high if rywell lla r t on itese t instrument s e t po i nt pre,sure swtich to actuate 2 psig I'IT ikulel e

278

_L-______.___..___

{

lia r k sila t e Iteset set point

[t0- 7 5-27 10/8/75

• 2 lil!S Core Spray valve permissive 1121 til 2SS pressure switch set point ilrl f t.

lla rk sila t e Iteset setpoint 110-75-30 10/6/75

• 2 IIRS li21' Pfl 2SS I?2T Al2SS I

t i

l a

l e

i i

AliS StHiltCl!

DNIII IlliitM IOtl CAllSli Cof-ll'Oril fil' Coltitl'CllVII At:1 tori RO-77-19 8/2/77 13 IIRS Striker plate was not reset Con so l iita t i I

'the clearance tietween prevent Ing IIRV e from not 6 in-1525 striker plate anil solenoid opening iluring t est VX ilresser toils was increaseil to I

s t op t h i. --ob l em._ _ _ _ _ _.

R0-79-42 11/6/79 7.5 Ill1S t4R108A 1.i f t eil Durisig Operat ion Ita r kila t e l'ressiin e switch was

]

i 11-25-111255 replaceil ley spare RO-80-1 1/5/80

• 8 IIRS liffilV D diil not open when t est ed Dresser

'the grule screws weie diaring shutdown improper st aking

(> x8 reset.

of grule screw RO-80-30 7/16/80

• l 2 lilts Operat ion in a ilegraded mode when NA

'Ibc clearance between D electromalle relief valve sailed the D 1.fIRV solenolit pluns'er anil pilot valve to open was ailjusted.

R0-80-38 8/26/80

• 2 111t5 lteactor t riple low wat er level l'IT Ita r t on Sensor was reset sensor Illi 1811 exceedeil setpoint 288A s

I I

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_. _ _ _ _ - - = - -. _ -..

J 1

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a y

Containment Spray Svstem i

42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br />

'4aintenance Activity Outages -

i Surveillance Activity Outages 360 hours0.00417 days <br />0.1 hours <br />5.952381e-4 weeks <br />1.3698e-4 months <br /> l

- Containment Spray System Automatic Actuation Tes 402 Hours Total Curage Time -

I Emergency Diesel Generat'as_

i 28.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> i

Maintenance Activity Ou'. ages Surveillance Activity Outages 480 hours0.00556 days <br />0.133 hours <br />7.936508e-4 weeks <br />1.8264e-4 months <br />

-Monthly Inspection of the Diesel Generators 1440 hours0.0167 days <br />0.4 hours <br />0.00238 weeks <br />5.4792e-4 months <br />

-Diesel Generator Battery Discharge Test 2204 hours0.0255 days <br />0.612 hours <br />0.00364 weeks <br />8.38622e-4 months <br /> Total Outage Time -

Core Spray System j

i 180 hours0.00208 days <br />0.05 hours <br />2.97619e-4 weeks <br />6.849e-5 months <br /> Maintenance Activity Outages -

Surveillance Activity outages Core spray System Instrument Channel Calibration 4

and Tes:

960 hours0.0111 days <br />0.267 hours <br />0.00159 weeks <br />3.6528e-4 months <br /> Core Spray Isolation Valve Actuation Test and Calibration 480 hours0.00556 days <br />0.133 hours <br />7.936508e-4 weeks <br />1.8264e-4 months <br /> 120 hours Core Spray Motor Operated Valve Operability Test 1740 Hours Total Outage Time i

Auto Decressuri:stion System i

The outage times for the electromatic relief valves Note:

are given only if the auto depressuri:stion mode was degraded.

43 hours4.976852e-4 days <br />0.0119 hours <br />7.109788e-5 weeks <br />1.63615e-5 months <br /> 4aintenance Activity Outages None

'urveillance Activity Outages 43 hours4.976852e-4 days <br />0.0119 hours <br />7.109788e-5 weeks <br />1.63615e-5 months <br /> Total Outage Times

' '~~*~Due to the high availability of the ECC systems as computed using maintenance activity cutage times; containment spray system -

99.904*4 Emergency Diesel Generators - 99.35%

?'

99.539%

Core Spray System

~

Auto Depressuri:ation System 99.901's

=.'

~

Thersarenosuggestedchangesthatwouldi=provethe. availability}

of these systems.

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

i ATTAC4 VENT 4 NUPEG 0737 Item iI.K.3.21 - RESTART CF CCRE SPCAY -

1.0 INTDCCUCTICN The Oyster Creek Cere Spray injection it initiated autematically or manually may be stepped y tne centrol recm operators.

This system.Ill not restart autematically or manually frem control reem if an Initiation signal persists.

The NRC requires, as per item it.K.3.21 of NUPEG 0737, that the ccre spray system Icgic ce modified tc crevice tne caoacility to restart tne system, if required, to assure scequate ccre ecoling.

The cere scray Icgic is at present so set up tnat it cannot be reset to a standby state after initiation i t either one of the t.o autcstart signals, namely "reacter Lc-Lc-water level" anc the "hi-drywell pressure" is continuously j

present.

With a centinous ccre spray start signal present the parallel isolation valves in the core scray locp cannet be closed by the coerators from the centrol r ec6a, but tne ccre spray and becster pumps can be stcpped.

If s?cpped, the pumps nill not restart autcmatically or by the control. room coeraters due to the antipump and spring energing design features of the pump breakers centrol circuits. The pumps can enly be restarted nen the core spray logic is reset to a stancby state.

2.0 DISCUSSICN 2.1 Normally during loss of coolant accident (LCCA) the ccre spray ficw

.ill not be ccmpletely stcoped by the coersters hen the core spray logic cannot be put back in standby mode. But there are situaticns wnich will require the coeraters te stem the flew even if the ECCS logic cannet be put in standby mode. A few of these situations are listed here:

2.1.1 Gross leakage from pump seal i

2.1.2 Eculpment malfunction 2.1.3 Need to shed the lead to start scme other important equipment such as emergency s4rvice arer pumes.

2.1. 4 Reacter "Le-Lc.ater level" signal, in mcs? type Of LOCA's','

will reset long before the "hi-crywell pressure" signal. Par?Icularly in smalI si::e break LOCA, tne water level in the vessel w i I l - b e.

2 n

_ on

+

AWAChMENT 4 (CONT) tiew i s not stepped in time restored quicxly and if tre cere sceay mater tevel can go too nign.

I 2.2 From the above the need ter sicpoing ine core spray tiew partially or ccmp letely, even without Oeing able to reset the start logic, can te seen.

start going in case of an uninsolatacle LOCA, the reacter.ater level will down again scon af ter the core spray injection is interrupted. The ECCS logic may ce may not be reset and i n stancby mode,. hen the water level initiation level. The signal will I

gets down to "Lc-Lo water level" signal scund an alarm anc light up an annunciater en panel 1F in the control recm.

But the core scray injection nill not restart due to the existing design

{

nct be able

~

feature Oescribed in Section 1.0 accve. Even the cperatcrs will m restar* the ficw trem ne centrol recm.

2.3 The need for providing restart cacabil ity i s obvious frem the above discussicn. The icgic can be designec to previce One et the tclicwing:

Fully autcmatic restart cacaDility i

a.

D.

Semi aut:matic restar capaollief c.

Yanual restart cacaDility t

Automatic restart capacility wnlle desiracle is not suitacle 2.3.1 l

ter the fellcwing reascns:

1 2.3.1.1 The long term core ecoling after'a LCCA should be centrolled Dy tne coersters Oecause many unforeseen events can occur during an actual LOCA.

It is very ditticult te design an automatic circuit which will take i nto account all possible ccmbinations of events.

2.3.1.2 The ccre spray Icgic is part of an i nterdependent ECCS system. Autematic design shculd consider impact on all these systems.

2.3.1.3 The design seculd taxe inte acccunt pctential impact en plant sucocrting systems suen.as standby pc.er supplies and service nater.

2.3.1.A The design shculd c'ensider c?ner pricei*y actions at the time et restart.

? ~

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2.3.1.5 ft snculd reccgnize; and di f f erenti ate 'te~ tween ene two l

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n ATTACEEVT 4 (CONT)

LCCA cere soray initiating si gnal s i.e.

reactces "Lc-Lo nater level", and "hi-crywell pressure".

2.3.1.6 Reacter ccre and vessel gecmetry maxes selection of the auto restart signal difficult.

Because, ter large break accidents at cer+ain locations the water level in the vessel may not be brought nign enougn to reset Lo-Lo water level signal, althougn the ccre is acequately ccoled.

Based on the above arguments it can be said that a fully autcmatic restart cacaDility nill be very ecmplex,.iil reduce system reliabi t ity and dec. ease cperater flexibility to a great extent.

2.3.2 A semi-automatic restart feature can be designed in the This wilI have existing logic for the core spray system pumps.

security of an automatic circuit initially and when everridden will provide the flexibility needed ter the operater centrol of the system.

This design is described in greater detail in Section 3.0.

2.3.3 A manual restart feature is thcugh most su itab l e w ith regards to flexibiIity of Operation vill still have greater peccability of inadvertent Operater mistake than the semi-autematic circuit.

3.0 REC 0KVENCED CHANGES After considering all the arguments discussed above, a semi autcmatic restart f eature is reccmmended to oe inecepcrated in the cere soray logic. The proposed changes are shown en logic senematics at* ached as Figure #1 & Figure #2.

Since the lack of restart feature in the existing logic is due to the. design of ccre spray and becster cumo centrol circuits, the design aporoacn taken here is to change the pumos centrol icgic witncut toucning the overal l ECCS automatic start logic.

The centrol switen contacts in tne pump trip circuit will be relocated to initiate a set of T.D.D.C. timing relays and auxiliary relays *nen the swi cn is actuated to trip the pumo, as snown in Figure #2.

In creer to prevent the in the presence of a "Lo-Le water levef" ccndi?Icn

'b' l

tripoing Of the pump centacts frem Octh "Lc-Le aater fevel" auxiliary reIays ina? can actuate the ECOS cnannel censisting of the particular pumo vill be placed in series with the timing relay cells. The trip auxiliary relays will be sealec-in when actuated.

6 I

~_-

I l

4 AI~ACBiENT 4 (CONT)

I i

Operation cf tne

• rip auxiliary relays will trip the sumo and disconnect the i

ECCS au*c star + signal fecm the ciese circuit of the pump ;cwer circuit Dreaner.

Centacts frem ECCS enannel reset switen allt be placed in T.te seal-in ;crtion of feature.

tne trip auxiliary relays circuit *c previce a ccar.cn reset The time celay in Orep cut feature of the circui* elli serve two purposes. It stagger *ne res* art of the ccre spray anc Occster pumps therecy preven *ing

.ill inem ccming en simultanecusly and cegrading the emergency sc.er cistribution system caused by the nigh starting inrusn currents. Secencly it.ilf revide encugn time for tne spring motor to reenarge ine closing spring if *ne reset-ing et Lc-Lo signal is Oniy mementary.

A manually c;ers*ed keyleck typass nill be previced around fne "Lc-Lo.ater level" interlock in *ne trip circuit *c give *ne crerster tne flexibill*y to trip tne pume even in the presence of *he signal. The bypass swl*ch.ift be key l

coerated and will light up wi*n flashing signal.nen in typass pcsitlen. It will al so annunciate an alarm in the cypass posi*len. The access te the bypass swi*cn key.ill te acministratively centrcllec.

Oceration cf any of *ne **ic auxiliary relays will annunciate and alarm in the centrcl recm en Panel IF.

4 i

The logic mccification is so designec tnat a single failure will nc* prevent the system f ecm perf orming its intenced safety functicn. Apprcpriate core spray l

secticns et Oyster Cree < technical specifica* lens anc c era-Ing precedures.iI!

te revised *o specify, the requirements anc prerecuisites for ?ne use Of the new au*c restar? Sypass feature, *ne description of the new restar* feature and tne eculpment surveillance requirements. The *ime celays set on tne new *iming relays afil te as fellcws:

1 Pumo NZO1A 10 secs.

Dumo NZ018 20 secs l

Pume N;01C 10 secs.

Dump NICID 20 secs.

Dumo N203A f* secs.

Pumo NZO3B 25 secs Pump N203C 1* secs.

Pumo NZO30 25 secs.

l 4.0 00NCLUSICN i

The exi sti ng Oyster Creek Cc'e's 'S:ra'y Sys3em :'ces not meet tne NRC recuirement i.

l l

t 4

g.

~-

i

.. AMACF21.N~~ J (CONT)

er ite-ll <.3.21 et N'UPEG 0737. The Ocre 3cray Sys em Icgic will have tc ce mccitiec ?c incer

crete ?ne NRC recuiriment.

semi-au?cmatic restar* teature escribec in Section 3.0 3 cve is reccmmendec A

tc te incer:cratec In the i cgic.

A ?ctal su?:riatic restart feature fil not enhance plant sate?y as Olscussec in See?Icn 2.0. The reccmmencec mocitiestion

.311 ?.e e? IEEE 279-1971 criteria 4.2, 4.12, 4.13 & 4.16.

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I TY?! CAL FUMP CONTROL LOGIC.(EXISTING)

CcRE SFRAY SYSTEM

(

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PAGE

\\ REVISiCN NO. I CATE o

m m 80 l

SC CCRE SFRAY RjsTAR T

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TITLE I

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ANAChuENT 6 NLFEG 0737 ITEM 111.0.3.4 CCNTRCL RCCM hhBITABILITT ns l

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TICRM.C:CN RKUIRED FOR CON ~ROL RCCM EA3I~A3Ei!~! E7ALUA!!CN Control Roce =ede of operation, i.e., pressurization and

es 1.

isolation filter recirculation for radiological accident or c lorine release.

h Fasconse:

The present Control Room does not meet the standard review plan criteria for habitability of :he Control Room in the event of an accident.

• he basic modification requi ed to bring the Control Rocm into ec=pliance is a new redundant ventilation system capable of positive pressure (1/3 inch water gage), with two air intakes with the necessary detectors for chlorine, radiation, and smoka, redundant air coed 1:1oning

• -ation and heating trains, redundant recirculation a"-

units, smoke re= oval system, and the necessary automatic dampers for the system to operate in the required modes.

Refer to :he attached flow diagram for the proposed system.

• he =edes of operation vould be as follows:

Normal - Intake air and recirculated air would bypass the filtration unit and =aintain the Control Roca envelope at desired conditions.

Accident - With the detection of radiation or smoke, the air intake detecting the radiation er. smoke vould close, and air would be drave frem the alter: ate air intake. At the sa=e ti=e, damper actuation would bring one of the recirculation air filtration units online.

The recirculation air train would filter a portion of the existing area air and intake air to ensure area habitability from any conta=inants.

Accident - With the detection of chlorine, both air intakes and all exhausts vould be isolated and damper actuation would put the ventilation system in a recir-culation mode, -ith a portion of the recirculated air passing through one of the filtration units to ensure filtering of any conta=inants that may be in the area.

I e= 2.

Control Race Characteristics Air volu=e Control Roce a.

?msponse:

• he Control Room air volume is 15,700 cubic feet.

- - - ~ -

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Control Rocs emergency :ene (Control Rocm, Critical b.

Files, Ritchen, "asbroc=, Computer Rocm, etc.)

Resconse:

The emergency zone of the Control Roca censists of the Control Rocm itself, the Shift Supervisors Office, Toilet Room No. 3 and the K.itchen.

Control Rocm ventilation system schematic with normal and c.

emergency air flow rates.

Response

Refer to attached si=plified flow diagram for normal and emergency flow conditions. (Ovg.No. 13432.25). Flow rates shown on flow diagram indicate preliminary design only, d;

Ilfiltfationleakagerate

Response

Since the new Control Roca ventilation system vill be designed to maintain the Control Roca =ene under positive pressure, there would be no infiltration of the outside atmosphere.

High Efficiency Farriculate Air (EEPA) filter and charcoal e.

adsorben efficiencies.

Response

The EEPA and charcoal filters will comply with the guidance of USNRC Regulatory Guide 1.32.

Mien new, the HE?A filter removal ef ficiency for particulates will be and the charcoal adsorber fecontamination 99.97 percent, In service, efficiency for iodines vill be 99 percent.

the HEPA filters will be tested in place to assure an efficiency of 99 percent, and the charcoal adsorber vill be laboratory tested to assure a decontamination ef ficiency for iodines of 95 percent.

f.

Closest distance between contaiment and air intake.

Response

. ~.

distance between the conta1=ent and the local The close:

air intaka vill be approxi=ately 100 feet, g.

Layout of Control Room, air-intabs, Containment ~3uild1ng, and chlorine, or other chemical se cage facility with dimensions.

- _ :3

Response

Refer to attached conceptual arrangement 3uilding i and Building 2 ter two proposed locations for the Control Roce 27AC and filter units (Dwg. No. 13432.23-1 and 13432.23-2). For general arrange-and d1=ensions of the Oyster Creelc facilities, see Dwg. JC19508.

ment

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Control Room shielding including radiation streaming frem penetrations, doors. ducts, stairways, etc.

Response

Control Room shielding including radiation streaming from various sources is discussed in a report prepared by EDS Nuclear, Inc. The only significant source contributing to an elevated radiation dose rate in the control room is from the core spray booster pu=p sue:Lon and discharge piping located at Elevation 51'3" in the Reac:or Building.

A shield wall is under design to reduce the Control Room dose rate belov the 10CFR, Part 50, Appendix A Criterion 19 limit.

1.

Automatic isolation capability-da=per closing time, damper leakage and area.

Resconse:

The automatic dc=pers which are required to isolate upon receipt of high chlorine, toxic gas, and radiation levels will close within 5 to 15 seconds. They will be designed for bubble-tight leakage in accordance with ANSI N509.

j. Chlorine detectors or toxic gas (local or remote).

Resconse:

Chlorine and radiation detection instruments will be located on the local and remote air intakes. Smoke detector in-struments will be located in the filtration unic Control Room exhaust and intake ducrwork.

k.

Self-contained breathing apparatus availabill:y (number).

Response

There are 30 Scot: Air Packa available for Health Physics use, 10 units reserved for the Fire 3rigade, and 10 units on order--a total of 50 uni:s.

l l

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3ottled air supply (hours supply)

Response

There are 30 bottles available for Health Physics use, 32 7

bottles reserved for the Fire 3rigade, and 30 bot:les on i

l order--a total of 152. bot:les (76 hours8.796296e-4 days <br />0.0211 hours <br />1.256614e-4 weeks <br />2.8918e-5 months <br /> of breathing air).

Upon reques:, Allied. Fire.& Safety in Redbank, New Jersey, has the capacity to fill;20-100 be::les and have them on size at Oyster Creek approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> af ter notification.

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Fsergency food and potable water supply (how sany days s.

and how nany people).

Resconse:

The water supply is plant potable water and :here is no e=ergency f ood available, Control Room personnel capaci:y (normal and e=ergency) n.

Resconse:

The Control Room has a desired nor=al operating capacity of 7 people and an e=ergency capacity of 15 people.

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?otassium iodine drug supply.

o.

Resconse:

The potassium iodine supply is 10,000 pills, 130 ng.

each with a shelf lif e of 2 years.

4 t

Item 3.

On-site storage of chierine and other hazardous chemicals Total amount and size of container; and, a.

I b.

Closest distance from Centrol Room air inzake.

r Resconse:

Closest

v. -

I Total Container Distance Iros l

Chemical A=ount Size CR Inle:

l l

Sulfuric acid 3000 gal.

3000 gal.

250 ft.

(L/S Radwas:e 31dg.)

Sulfuric acid 3000 gal.

5000 gal.

350 ft.

(Fre-Treatmen: Tank) chlorine 10 tons

,1 :en 225 fr.

Sodium hypochlori:e 300 gal.

15 gal.

450 f:.

l I:e= A: Off-si:e anufacturing, storage or transportation facili:les of hazardous chemicals.

l

.u Identify facili:ies wi:hin a 5 mile.radies;

=

a.

da b.

Distance frem Control Room; a Frequencyofha:ardouschemicak::ansper:ationtraffic c.

(truck, rail, and barge).

4 un a.1AHi u

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Response

There is no off-site =anufacturing, storage or transportation of hazardous che=icals within a 5 mile radius of the OCiGS.

1 Item 5: Technical specifications a.

Chlorine Detection Systes b.

Control Room emergency filtration system including the capability to maintain the Control Room pressurization at i

1/8 inch water gage, verification of isolation by test signals and damper closure times, and filter testing requirements.

Resocnse:

The required technical specifications will be based on t

the guidance.provided in NUREG~073 C They will'be provided on a_ schedule consistent 7 ith the recuired nodifications.

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ATTACHMENT 7 NUPEG 0737 ITEM ll.F.1.2 ACCICENT MCNITCRING INSTRUMENTATICN - RADIGIC0!NE & PARTICULATE EFFLUENT VCNITCRS Oyster Creek Nuclear ' Station of Jerscy Central Power & Light Co. i s installing a new gaseous effluent menitor en the isekinetic prebe. NUREG 0737 (Sections 11.F.1-2 through 11.F.1-9) and Regula? cry Guide 1.97, proposed revision 2, drafts, October 1980 (Table 1) give a requirements ter the ccerational accer ranges of the particulate iccine anc noble gas Onannels of e f f l uent monitors.

The new monitor at Cyster Creek, supplied by Science Applications, Inc. is capaole of menitoring ef fluent activity concentrations up to the requirec upper limits. However, the requirement for the iodine and particulate channels (100 uC1/CC) presents scme real conceptual and coerational problems that are ecmmen to any ef fluent monitor. We believe that this required uoper limit for gasecus effluent is unnecessary. Therefore, we request tnat the NRC reconsider the use of the value of 100 uCi/cc of iodine as a design upper liml? for cperation. We feel that an upper liml? cf 1 uCi/cc for botn iccine and particulate concentrations, although still conservative, is a mere realistic and i

practical value.

Basis ter Preccsed Reduction Measured concentrations of airbcrne radioicdine and racicparticulates from the TMI-2 accident have been publisned. The maximum airborne lodine concentration in the auxiliary building was aticut 10 uCi/cc anc that in the centainment atmcsphere was about 10 uCi/cc.

The corresponding particulate concentrations were lower than these values. These levels were the highest that

.ere reali:ec in spite of tne f act that nearly all of the lodine was released tecm the core early in tne accident.

Concentrations of airbcene iodine are controlled by natural phenomena that limit the dispersal of fission product activities.

These phencmena are def ineated in a recent presentation given at the AMS meeting in nashington and include chemical reactions, aerosol tenavicr, condensation and surface deposition. The establishment of an equilibrium airborne concentration by ceposition and resuspension from surfaces i s especially important.

As a consequence of the natural limiting of airbcene iodine, we consider 1 uCi/cc to :e a conservative upper limit (ten times higher than the maximum concentratien cbserved anywhere at TMI) to the upper limit ter iccine anc carticulate cencentra?Icns in airecrne streams.

m n mA e

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ATTACHMENT 3 NLREG 0737, ITEM i i I.A.2 VETECR0 LOGICAL CATA -

.VETERCLOGICAL PROGRAM DESCRIPTICN Existing veteorological v nitering Svstem e

The existing meterological data collection system consists of a 400 foot tcwer Iccatec approixmately 2400 feet northwest of tne Oyster Creek Station.

The tower is located en flat terrain in a i:leared area, no natural er man-made ecstructicns exist in the vicinity of the tcwer nicn eculd influence instrument measurements.

The tower is instrumented at 33,150 and 380 feet aoove grade. The meterological parameters monitored at these levels include: Wind speed and direction, ambient tem:erature, tem erature ber.een 150' - 33 ' and 380 ' - 33'.

Dew Point is monitored at the 33 and 380 foot levels and precipitation at tne tower base.

Data acquisition censists of botn digital and analcg recording systems.

are made bet at the meteroicgical shed Analeg recceding (centinues strip charts) tne tower base and the Oyster Creek Centrol Recm. Digital recceds consist of at data sampled at an interval et once every 60 seconds, with the exception of The precipitation which is recorded on a cumul ative basis ence every hour.

digital cata is averaged every 15 minutes, stored by a mini-ccmputer Iccated at and then transmitted via telephone communicatiens to a central tne tower base, ccmputer which archieves *ne cata.

Voceeded v tecrotocical Mcniterina Svstem e

A schedule will be adopted to fully ccmply with the four essential elements of the NUREG 0654, Rev.

1, Appendix 2 utilizing the alternative implementation schecule described in the NUREG 0737.

Because NUREG C654 Rev. I has just recently been issued, we have not yet been sole to fully review its contents and decide how best to meet those We will complete our review of recuirements centained within this document.

this cccument and previce you witn cur preposal fer any modifications recuired by Accendix 2 to the NLPEG Oy June 1, 1981.

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4 ATTACHMENT 9 NUREG 0737 ITEM ll.E.4.1 - DEDICATED HYDROGEN PENETRATIONS Backcround NUREG 0578 item 2.1.5.a issued in September 1979 established additional requirements for post accident hydrogen control.

As you are aware there was l

1nitiaiiy a great deal of debate and confusion regarding this issue. This topic was further clarified by your October 30, 1979 letter.

Through a misinterpretation on our part, the need to make the vent and purge pathway single failure proof from an operational standpoint was not recognized until the spring of 1980.

Status l

Un June 10, 1980 we submitted to you our conceptual design for modifications that would be performed to accomplish the requirements of NUREG 0578.

Presently work is proceeding on the design, engineering and procurement-packages to support these modifications.

Due to projected delivery dates of cer tain components we will be unable to meet the NUREG 0737 completion date of July 1, 1981. Listed below are three of the controlling components and their expected delivery dates.

Item Projected Delivery Date l

Pressure and Flow Transmitters August I, 1981 Remotely-Operated Pneumatic isolation and Control Valves August 15, 1981 Control Room instrumentation November 20, 1981 I'

Justification It should also be noted that the Oyster Creek plant utilizes an inerted containment. This eliminates the possibility of creating an explosive hydrogen l

mixture until a significant time af ter any situation that results in a metal to water reaction. The method that would be utilized to ensure that an explosive mixture did not occur, would be to add nitrogen to the containment while venting of f at the same time.

The Oyster Creek containment does presently have an oxygen analyzer installed which monitors the containment atmosphere.

The containment hydrogen monitor will be installed in accordance with the NUREG 0737

(

schedule of January 1, 1982.

Based upon the unlikeliness of an event which would generate significant amounts _ of hydrogen, the f act that the Oyster Creek Containment is inerted, the projected delivery date of key components and the NUREG 0737 schedule for i

hydrogen monitor, we feel that a delay until our next i

installation of a refueling outage which will commence no later than December 1, 1981 is-i justifled.

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