NUREG-0821, Forwards Safety Evaluations of SEP Topics,Vols I-III,per App E of NUREG-0821, Integrated Plant Safety Assessment Rept for Re Ginna Nuclear Power Plant

Forwards Safety Evaluations of SEP Topics,Vols I-III,per App E of NUREG-0821, Integrated Plant Safety Assessment Rept for Re Ginna Nuclear Power Plant

ML17309A271
Person / Time
Site:	Ginna
Issue date:	06/04/1982
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:	Shewmon P Advisory Committee on Reactor Safeguards
References
RTR-NUREG-0821, RTR-NUREG-821, TASK-***, TASK-RR NUDOCS 8206070270
Download: ML17309A271 (68)

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50-10 50-237 I

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Q Commonwealth Edison Company ATTN:

Hr. Cordell.Reed Assistant Yice Pres'ideht

'Post Office Box 767 Chicago, Illinois 60690 Gentlemen:

, ~ s At our meeting with you on tray 31,'978,,'.Qe'.ft}dfcated ghat our ravfew of several SEP topics was essential.Yy'domp$ jte:.'..

We also,',stat'ed'that completed topic assessments would be 'sent'-to you for fhfoirmhtion and review and would be placed fn the Public Document, Rooms.".

Our initial evaluation of efght of these essentially complete topfcs is enclosed.

You are requested to carefully examfpe the facts upon-whfch the staff has'ased fts evaluation and respond either'by -"

confirmfng that the facts defining your plants are correct, oj'".by identifying any errors.

If fn error, please supply cory'ercted:"

'nformation for the docket.

We encourage you to supply, anyether material for the docket related to these topics that ydu b'elfeve

'o be helpful.

C At the Nay 31 meeting, the SEP Owners Group requested clarification of SEP documentation procedures and made severa'I suggestions in that

. regard.

Enclosure I is our response to the request and suggestions.

It contains the documentation procedures to be used throughout the SEP program and discusses the bases for these procedures..

Our documentation of the eight essentially complete topics fn Attachment I illustrates the documentation procedure to be used.

We woul.d appreciate any comments you may have to improve documentation of topic assessments.

Enclosure:

Response

to the SEP Owners Group Suggestions, Darrell G. Eisenhut, Assistant for Systems 4 Projects Division of Operatfng Reactors

~gt8 60-ZA Cwtral 8 R,zoec vazzc n~

Director

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Commonweal th Edison Company CC Hr. John M. Rowe Isham, Lincoln 5 Beale Counselors at Law One First National

Plaza, 42nd Floor Chicago, Illinois 60603 Hr. B. B. Stephenson Plant Superintendent Dresden Nuclear Power Station Rural Route 81 Horris, Illinois 60450 Anthony Z. Roisman Natural Resources Defense Council 917:15th Street, N.

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TOPIC VI-7.A.2 - Upper Plenum Injeotion SEP Plants Affected - Ginna OBEs Affected - Loss-of-Coo1ant Accident Discussion

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On Hay 1, 1978, NRC issued Amendment No. 19 to operating license No. OPR-18 The staff Safety Evaluation Report which supported the license amendment addressed the upper plenum injection topic.

Ginna submitted ECCS performance analyses for the Westinghouse and new Exxon Nuclear Company (EHC) fuels.

The Westinghuse analysis was perfor ~d for Cycle 7 fuel which the staff believes is a conservative evaluation for the Westinghouse fuel during Cycle 8.

The ENC analysis was performed

'for Cycle 8 using the EHC WREN-.II ECCS evaluation model.

The ENC

.evaluation model'has been revieiced and approved cond'.tionally by +".e NRC.

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The staff has recently considered whether the Westinghouse generic evaluation adequately represented the floe characteristics of Westinghouse two loop units.

The generic evaluation model assumes that all safety injection water is introduced directly into the lower plenum.

For the Oe loop units, the safety injection water is injected into the upper plenum.

Thus, the staff was concerned that the Mestinghosue model did not consid r interaction between UPI water and steam flow. After plant specific submittals by licensees operating two loop plants were reviewed, the staff concluded that the calculations provided by the licensees (with certain

.:>difications to the staff's model) are acceptable on an interim basis or

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'tehii e forts continue for developing.a model specifically tr'eating UP[..

Forthe Ginna.plant the calculations which specificallylconsidered..UPI:.using the modified version of the staff model, resulted in a change of only 15oF

.from those using the generic model in which the UPI-core interaction was not specifically considered.

In the interim, before these models are developed, Ginna has provided a modification to the current Mestinghouse model which accounts for UPI-core interaction. It was demonstrated that the modification resulted in the increase of peak clad temperature by 15 F.

Since for the Ginna plant both EHC MRE!)-II and Mestinghouse models predict similar PCT's (1922 F for EHC MREH-II and 1957 F for Mestinghouse) it can be expected that the UPI modification, when applied to the ENC MRS'f-II model, would allow about the same increase in PCT.

The licensee has drawn a similar conclusion.

Concl us ion The staff has concluded'hat although the Mestinghouse and Exxon two-loop generic-evaluation models should be changed to consider upper plenum injection (unless the plant is modified), analys'es't the specific operating conditions applicable to the Ginna plant demonstrate that ~

the effect of disregarding upper'plenum injection interaction on refill and reflood conditions will not be significant (less than 20 F PCT).

Therefore, the staff believes. that. for the limited range to which the models do not deviate from the requirements of 10 CFR 50 Appendix K item 1.0.3; and the calculations are acceptable.

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Mr. Leon D. Mhite, Jr.

cc Harry'H. Voigt, Esquire

LeBoeuf, Lamb, Leiby and MacRae 1333 New Harpshire Avenue, H.

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Suite 1100 Mashington, D. C.

20036 Mr. Michael Slade 12 Trailwood Circle Rochester, New York 14618 Rochester Committee for Scientific Information Robert E. Lee, Ph.D.

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Box 5236 River Campus Station Rochester, Hew York 14627 Jeffrey Cohen New York State Energy Office

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Swan Street Building Core 1, Second Floor Enqire State Plaza

Albany, New York 12223 Director, Technical Development Programs State of Hew York Energy Office Agency Building 2 Erpire State Plaza Alba',

New. York 12223 Rochester Public Library 115 South Avenue Rochester, New York 14604 Supervisor of the Town of Ontario 107 Ridge Road Mest

Ontario, Hew York 14519 Resident Inspector R. E. Ginna Plant,

.c/o U. S.

NRC 1503 Lake Road'ntario, New York 14519 R. E.

GINHA NUCLEAR POWER PLANT DOCKET NO. 50-244

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g Director, Technical Assessment Div",'sion Office of Radi ation Programs (AM-459)

U. S. Environmental Protection Agency Crystal Mall f2 Arlington, Yirgi ni a 20460 U. S. Environmental Protection Agency Region II Office ATTN:

EI S COORD IHATOR 26 Federal Plaza

~ New York, New York 10007 Herbe& Grossman, Esq.,

Chairman A'tomic Safety and Licensing Board

.. U. S. Nuclear Regulatory Comnission Mashington, D. C.

20555 Dr. Richard F. Cole Atomic Safety and Licensing Board U. S.'uclear Regulatory Cormission Mashington, D. C.

20555 Dr.

Earn th A. Luebke Atomic Safety and Licensing Board U. S. Nuclear Regulatory Comnission Washington, D. C.

20555 Mr. Thomas B. Cochran Natural Resources Defense Council, Inc.

.1725 I Street, N. M.

Suite 600 Mashington,.D.

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20006 Ezra I. Bialik Assistant Attorney General Environmental Protection Bureau New York State Department of Law 2 World Trade Center New York, New York 10047

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%g*y4 UNITED STATES NUCLEAR R EGULATORY COMMISSION WASHINGTON, D. C. 20555 November 27, 1981 Docket N>.

50-244 LS05 11-067 Nr. John E. Maier Vice President Electric and Steam Production Rochester Gas 8 Electric Corporation 89 East Avenue Rochester, New York 14649

Dear Hr. Naier:

RE FINAL TECHNICAL ASSESSMENT,OF SEP TOPICS VI-10.A AND VI-7.A.3 FOR GINNA We have enclosed a revised final staff evaluation for SEP Topics VI-10.A and VI-7.A.3.

The revised report reflects the additional information provided in your September 25, 1981 letter.

As noted in our revised safety evaluation, the staff considers SEP Topics VI-7.A.3 and VI-10.A to be acceptably resolved except for the concern with regard to bypassing of manual initiation.'ecause this concern is being pursued as a part of SEP Topic VI-4, the staff concludes that SEP Topics VI-7.A.3 and VI-10.A have been completed.

Sincerely, I

Dennis H. Crutchfield, Chief Operating Reactors Branch No.

5 Division of Licensing

Enclosure:

Topic VI-10.A and VI-7.A.3 cc war'enclosure:

See next page

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TEC NICKEL ASSESSEE.'3T OF 740 SAFETY TOP jCS FOR Gf,'NA 1.

VI-10.A:

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VI-7.A.3:

T sting of Reactor Trip Including Response Time ECCS Actuation System and Engineered Safety Features, Testing TABLE OF CONTE?)TS jv.

V.

VT

<ntro "ucticn Review CriTeria Related Sa ety Topics and Interfac Review Guideline Testing of RTS nd. ESF at Ginna?lant 1.

Reactor pro.ection sys m general descr ption 2.

Reac:cr. protection trip function 3.

Reactor protection system 'sting 4.

Ergineered sai ty features general description 5.

Engineered safety feature testing Table 1.

Ginna Tech.

Spec.

require.",ents for Reactor Trip System.

Table 2.

Ginna Tech.

Spec.

requirements for ESFAS.

Evaluation and Conclusion

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TOPIC VI-10.A TESTING CF REACTOR IRIP SY5TEN AMO E'<GINEERED S~F TY

""-'iNCLUDIi'(G RES?CiNSE TIDE TESTIf/G

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TOPIC YI-7 A 3 ECCS ACTlJATION SYS~IM I'ntroduction These

.~o SFP safety topics deal.with the tes.ability and operability of the Reac.or Protection Sys em (RPS) and the Enciineer d Safety Fea ures (E=F) ystems.

Sirce the ECCS actua icn is par. of.he.Engine red Safety Fee uw Sys.am, these two topics will be treated in one evalua.ion r por..

The RPS and E F tes procram should demonstra e

a high decree of avail Qilit of t.e systans and the rospcnse times assi-...ed in the accident analyses o

be withir. the design specifications, This report reviews the plant desicn to assure tha. all ECCS ccrponents, ircluding the pumps and valves, are included in.he component.and system.es..

.he frequency and scopo'f the periodic testing is adequa.e, and the ~s: procram cz ts "e requir-ments of the General Design Cri. ria and the Reculatory Guides defined in Sec fon II oi this report.

This evaluation report is limited to a compar',soh o

the RPS and ESF est'.ng progran with the r view criteria, and the review guidelines defined in Sec.ion II and IV.

Fur.her.detail of the test program for pumps and vaTves can. be found in the.

"in-service valve test program and r lief reques.'afety evalua ion regni.

II.

R view Cri eria The following General Design Cri.aria govern the topic review:

GDC 21 - Protec.ion sys em reliability ard es.abili"y GQC 37 - Tes.ing of emergency core cooling system The following Regulatory Guides and Sranch Technical ?ositions provide accept ble basis for RP5 and ESF es.ing progr~~:

0 RG 1.22 -

P riodic testing of pro ection sys.em actuation,uncticns.

RG 1.118 - Periodic tas-ing of lec ric power and protection systems.

o RG 1.105 - Instrur;,ent setpoint Branch Technical Position IC53 24 - Tes.ing of Reactor Trip System and Engineered Safety Feature Actuation System Sensor response times.

-"ranch Technical Position ICSB 25 - Guidarce.or Interpretation of Genera1 Oesign Criterion 37 for testing and cper&ility of the ECCS as a whole

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Standard Review?lan Section 7.2 and 7.3.

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Related Safety Topics and Inter aces; YI-7.C -.ECCS Single Failure Criteria and require. ants.or lockout pow r to valves.

YI-7.F - Accumulator isola.ion valves power and c"ntrol sys.em design.

III Environmental gualification of Safety Rela. d:-quipment.

VI Containment isolation.

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Review Guidelines

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6DC 21 states that.'.he redundancy and indepen(en e

C sigred in o he protection sys

=-m shall be sufficient to ass r that (1) no single

~ ailure results in loss oi the protec:ion function and (2) the prot tion sys.em shall be designed to. permi" periodic testing of its func ioning when the reactor is in operation, including a capa"i'.ity to test channels independently.to determine failures and losses of redundarcy that may have occurred.

GOC 37 requires that the ECCS be ~esigned to p rmi. apprcpri

.a "eriodic pressure and functional testing to verify the performance of the full operational sequence tha brings the system in,o opera ion, including opera.ion oi.applicaole portions of the protaciicn sys e."., the rans-,ar'etween normal and emergency power sources, and the opera:ion of the associated cooling water system.

Regulatory Guide 1.22 provides the acceptable methods for es ing actuation devic s

and ac.ua.

d equipment.

Regula cry Guide 1.105 States tha. Ins -..men.s should be calibrated so as to ensure the required accuracy at he setpoint.

one ac uracy of all setpoin s should be equal to or b t.ar than the accuracy assumed in the safety analysis.

Re ulatory Guide 1.118 describes the me.hod

<<ccaptable to the NRC staf, of complying wi h he Com"..ission's regulat'ions wi-h respect to the periodic testing o. the pro a tion syst~~

and elec.ric power system. for systms impor.ant to safety.

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Systems impor.ant.o safety as defined by R.G. 1.165 are as follows:

Those systems tha.

are ne essa"..y

.o osure (1),the integri.y of he reac or coolant pressure bound ry, (2) the capability to shu" down the reactor,.or (3) the capability to prevent or mitigat the consequence~

of accidents.

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Branch Technical Position ICSB 24 states that periodic tests for veriii-.

cation of system response ti...es of RTS and ESFAS should include the re-sponse tine of the sensors whenever practical.

B.

Br nch Technical Position ICSB 22 s ates tha. all portions of the protection system shculd be designed in accordance with IE=E Std.

i79-1 71 and all actuated equ'.'p-.,ent tha. is no-tested durino reac.or

'era:

cn should. be identified and just'.f'ed to the provisions cf posi ~

>cn 0 4 in R G. 1.2"..

9.

Branch.echnical Position ICSB 25 s"ates that all '"CS "umps should be incluCed in the syst m tes..

10.

taodard Review Plan Section 7.2 Appendix A

"ams 9, '.0, 11 and

'rovice

..ore spe i ic,uioance to review Re~ctor Tr ip System Tes

ing.

11.

Staodard Review Plan Sec.ion 7.3 Appendix A

It~>.s ll, '.2, 13 and.l4 provide more specific guidance to review Enqire red Safety feature system testing.

12.

'leri>fy.he following:

A.

Tes. conditions co"..e as. close as possible to'he actual performance requ'red by RTS and ESF.

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Compliance with the s'.ngle >ailure criterion dur'ng s

i>ng.

C.

The results of lic nsee response tine tasting data (if available)

.or the RTS and ESF are within the delay times used in the FSAR accident analysis.

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Test can be made

.o ensure the readiness or the operability of system ccmponents.

The Auto Bode of ac uation does not inhibit the Manual

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actuation, and vice versa, at any time.

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The power s'upplies'atisfy the Single Failure Criterion.

G.

The overl.appinc tes-s inde d overlap fron one

-.as segment to another.

H.

Transducer calibrations are adequate.

Comoarator calibr tions are adequate.

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Testin of RPS and ESF at Ginna Plant 1.

Reactor Protection System general description.

The RPS automatically trips the reactor to protect against reactor coolant system (RCS) damage caused by high system pressure and to protect the reactor core against fuel rod cladding damage caused by a departure from nucleate boiling (DNB) under the following conditions:

A.

Reactor power reaches a preset limit.

B.

Excessive temperature rise across the core.

C.

Pressurizer pressure reaches an established minimum or maximum limit.

D.

Pressurizer level reaches an established maximum E.

Loss of reactor coolant flow.

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The basic re c.or ripping ph:losophy is to defire a ".egion of power.'and coolant te-,perature aod pressure conditions allowed by the primary tripping functicns (overpcwer hich hi.rip, overta~perature high BT trip, and nuclear overpower trip).

one allowable opera.ing region wi hin these trip settings is provided to proven.

any combination of powe.,'emperature, and pressure which would result in a

Di18 with all reac.or coolant punps in operation.

Additional trippIng functions such as a high pressurizer. pressure trip low pressurizer pressure trip, hich press ri= r wa er level.rip, loss-.

of-flow trip, s.

am and fe dwater flow mis-atch trip, st am generaton low-low.water level trip, turbine trip, safety injec~ion trip, nuclear source and inter;.ediate rang

trips, and manual trip are provided to back up the primary tripping unctions for specific.accident condition ard nechanical fail'ur s.

Th Ginna reac.or possesses high-spe d Westinghouse magnetic-type cont". ol rod drive (CRD) mechanics,w.

The reac".or in.ernal c"rponents fuel assemblies, rod cluster con.rol (RCC) asse,ilies, and drive sys.ems components are designed as Class l.euip.-..en".

Two reac or trio breakers are. provided to in.arrupt pcwer to the CRD me hanis,".s.

The breaker main contacts are connected in ser~es wi.h the power s pply to the 'e~aanism coils.

one trip bre'kers are opened by the undervoltage coils on both br akers.(nor;.ally anergi d) which beccres de ner",ized.bony one of he several trip signals.

E'ch pro=

tec.ion channel actuates

.~o separate trip logic trains, one for each reac oz trip breaker undervol age trip coil.

The ele

".ical s-ate of the devices providing signals to.h circui bre <er undervoltage trip coils causes these coils to trip the br akar in the event of reactor trip or power loss.. 'Opening either. breaker intarr pis po~er to the magnetic latch mechanisms on.e ch CRD, causing them.to release the rods and allowing the rod clus ars to inser. by gr vity into the core.,

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The reactor shutdcwn furc.ion of the rods is cci:pletaly independent of the nor.al control,unc-ions since the trip breakers cc-.pie.aly in.erupt he power supply to the rod.~echanis,"..s and thereby raga.a any possibili='y

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r. sponse to control signals.

The control rods nus" he" energi-ad ta rer.:ain withCrawn fran the care.

An autar,"atic reacta'r trip occurs

-r:pan the loss oi power to the control rods..

me RPS is designed cn a chanrelized basis to achieve '.so)ation and

'ndepenCence be-..ween redundan prst~ctian channels.

The coincicet tr-:p philosophy is carried out ta provide a saia and reli ble sys-.e sire a singl'e.ailur wirl not defeat:.,e iunc-.-'on oi the crranra]

and will also not causa a spurious plant trip.

Chanreliindependence is carvied

.hroughcut the systen extending free'toe sensor to the relay providing th logic.

me chanrelized design that applies to.he analo as well as the logic portions o-, the pro.ecticn system is discussed e ow.

Isolation of r Cundan.

analog channels criginates at.he proc ss.siensars an continues back throuah tne.field wi>ring and csnta-,";.ent pener ticn a tha analog protect'.an racks.

'~hen the saiety and can:rol unc ions are coribined, bo.h >one.ions are fully isolated in the re.-.>airing part oi the channel, control being derived.frc;,.the pri."..ary safe.y sic-.>a1 pa, h

through an isolation a.-.,plifier.. As such, a failure

>n the car>trol circ;itry Caes not ai>act the safety channel.

'This approach is sed ior pressurizer pressure and water level channels,'sta~~

.generatcr.

water level, and

~iT chanrels, s

e~~ flow-e Cwater flow and nucle'r power rarge channels.

1 Ph>ys cal sapara.ion is used to achieve isolat>on.oi. redundant trans.-..it~re, Separaticn of field wirina is. achieved using separate wireways, able trays, conduit runs, and containrient pane.ratians d'or each reduncant c..ar,nel.

Analog equipr;.ent is.separa ed by loca.ing redurdant c~one~ts in diiferer>t protection racks.

The pawer supplies to the channels are.iad.=".ca aur ins.rur:ant buses.

Two of the buses ar.e supplied by cons.ant voltage transformers ard.~w are supplied by irverters.

ach charm 1 is enargiz d fram a separate a-c power feed.

iach reactor trip circuit is designed so.hat a'rip occurs when

.he circui-is Ceenergized.

An cpen circuit or the lass.a-.

~ channel power,.hereior, causes

".he system ta go into its trip ;~de.

Reliability and independence are obtained by reducoancy within e* 1 tripping iur.c-ion.

?n a two-cut-oi-three circuit, the.three c!ia:~els are equipped wi.h separate prir ary.sensors and each channel is a"ergizQ raa an inaependent electrical bus.

A singl failure may be applied in which a channel fails to de nergize when required haweve.,

such a ml-func ion can ai a. only one channel.

one trip signal.urnished by ~Ha two remaining charnels is unirpair d in iis event.

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' All reactor pro.ec.ion charnels are supplied. with sufficient redu'ndancy proviCe the capability for cbannel calibration and tes.ing a

.power.

8ypass r moval.of one trip circui is acccmplished by placing. that circuit in a half-.ripped mode, i.e.,

a two-out-of-three circuit

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becom s

a ore-out-of-a-b o circuit.

Testing does not i

h

.r p. e sys en unless a trip condition concurrently exis s in a redundanx channel.

Cer ain reac'.ar irip channels are au.amatically bypassed at lcw power to allow or suc!i condi.ions as s a'rtup a.,d shutdown and where

.e no requir d

or sa>e y.

.Nuclear source range and inter..ediate range trips, which specifically provide protec.ion at low pcwer or subcri.ical.cp raticn, are.bypassed at pcwer operation to prevent spurious reactor trip signals and to improve r liability.

The reac.or trip bis ables are mounted in the pro.ec icn racks

~"d a..e final operational components in an analog protection channel.

="ach bistable driv s two logic relays (C and D}.

Tne contac.s fram¹ C

relays are.interconnec-ed to form the required ac.uaticn lcgic for trip breaker Ho.

1 through d-c power feed Ho..l.

The transi.>on 'cm 0

1 identi.t 1

. y o logic id n ity is made at the logic relay coillrelay contact n.erface.

As such,

.here are both elec.rical and physica,l sepa". ation between

-he analog and the logic por.ions of the protection system.

The above locic ne.wor k is duplicated for trip.breaker

>)o.

2 using d-c power feed No. 2 and the cantac s frcm the 0 relays.

7nerefore, the.<<o redundant reactor..rip logic channels will.be physically separat d and e ec.rically isolated rcm one another.

Overall, the RRS is cc.";.prised of identi iable channels which are physically, elec rically, and func.ionally separated and isolated frcm one aaother.

A typ>cal trip logic channel is shown in Figure 7.2-8 of the FSN.,

2.

Reactor Protec.ion Trio Function A.

manual Tr-;p A manual re c.or. trip is provided ta permit the operators to trip th reac.or.

Tne manual ac:ua.inc devices are indeoenden. of.the automatic reactor trip. circuitry and ar not subjec to failures

<<hich cauld ;,'ake the automatic circuitry inoperable.

3.

High Nuclear Flux (?ower Rance) Trip This circui t"..ips the reac.or

<<hen t'<<o-of-the-four power range c!ianne1s re'd above the t". ip set?oint.

Tnere are

-<<o se.points associated wi.h.his trip.

Th low set.ing can be'manually bypassed

'<<ben.<<o-of-the-Four power rana

.channels are above approxima:el 1C>> power.

Thr ee-of-the-four channels reading below 10>> pcwer autc-atically reirs.a es Jie trip.

The hich set ing is always ac.ive.

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5 Hioh nuclear Flux (Inter...ed-ata Rance) Trio This circuit trips the reactor. when ore-oi-t"e-two intermediate range channels reads above the rip setpoint.

This. r5'p can be manually bypassed if two-of-tbe-four power range channels are above.aporox:ma.ely 10" power..Thre -of-the four channels below this value automatically reinsta as

.he trip.

Tne inter;.ediate channels (including.detec ors) are separate from the. power range charnels in this plant design.

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Hich lluclear Flux (Source Rance) Trio This circuii. trips the reactor when one-of-the-two source rance channels reads above the trip setpoint.

I can be manually bypassed when ore-of-the-two intei-.. diate rarge channels.re ds above the source rane cuto=-, value.

Both inta....ediate pange channels below this value automatically reinsta.e the trip.

I This.trip is also bypassed by two-of-the-four high power rance signals..

The trip point is set between he source range cutoff pointer level and Ne maxim'ource rarge power level.

Overtamoe. at'e dT Trio This circui. trips the reactor on coincidence. of two-o -the-four signals, with two channels per loop o protect the core agains=

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Overocwe.

4T Trio This circuit trips the reactor on coincidence of two-of-the-four

signals, with, two channels per 1oop to protect agains.

excessive power (i.e fuel rod rating protection}.

G.

Lcw Reactor Coolant Pressure Trio 7nis circui. irips the reac or on coincidence oi two-of-the-four pressurizer pressure sicnals to protac against excessive voids and resultart high iuel temperature.

High Reactor Coo)ant Pressure Trio This circui rips the reac or on coircidenc of two-of-the-thr pressurizer pressure siignals to limit the range of required pro-tac ion from the overtarrperatur e dT trip and to protect aoainst overpressure.

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Hi h Pressurizer Mater Level Tri This circuit trips the reactor on coincidence of two-of-the-three high pressurizer water level signals to trip the reactor.

It is provided to limit water relief from the pressurizer.

Low Reac.or Coolant Flow Trio K ~

This circuit r p signal is ac.ua d 5y the coincidenc Gf wo 0-the-three signals fcr each reactor coclant lcop.

The less cf flew in either locp c*uses a reac: r trip.

This trip protects

.~e ccrc from a "NB oHcwing a loss of coolant flew'.

Sa.etv In'ect',on System Actuation Trio L.

This reac.cr trip oc urs on he actuation of the sa.ety irjec ion system (SIS), i,e.,

when here is I

1)

Low primary syste~ pressure (twc-o -the-three signals);

2)

Hich contairr;.en pressure (two-cf-the-three signals);

I 3)

Coincidence of low pressure in ei.her s.eami generator (two-o,-

the-thre signals).

Turbine Trio This trip is sensed bv two-of-the-thro e signals

rem the au estop oi 1 pressure.

Tnis is an anticipatory trip which prctects the reac.or rom a.sudden loss of heat sink.

S:e~~/Feedwater Flow Mismatch Trio This trip is actuated bv a ste~~/fe

'water flow mismatch (one-of-the-two signals) in coincidence with low water level (one-of-the-two signals) in either s

earn generator.

This trip protects the reactor frcn a sudden loss of heat sink.

Low-Low Stem Generator Mater Level Trio inis trip is ac uated on two-of-the-thr low-low wa:er level signals in ei her stems generator.

Tnis trip prot cts the re ctor frcm a loss of heat sink.

3.

Reactor protection System Testing A.

Protec=ive S stems Capability for Testing and Cali ration Tne his.able oor.ions of the pro ective system (e.g., relays, bistabl s, etc.) provide trip signals only fter signals

~ rm the analcg portions of the system have r ched a,"res t value.

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me capability is provided for calibrating and testino.the per-.ormance of the bistable portion of protective channels and various ccnbinz-tions of the logic networks Curing reactor operation.

The analog p:rtion of a protec ive channel (e.g.,

sensors and a.,.pli iers) provides analog. signals of reac.or or plant parameters.

The following means are provided to permit checking o

the analog portion of a protective channel during reac.zr opera.i'on:

I)

Varyinc.he moni:ored variable 2)

.'ntroducing and varying. a substitute transmitter signal 3)

Cross-checking between iden ical channels or between channels wnich bear.a known relationship to each other and which have readou.s avail able.-

This design permits administrative control of the:

I)

Kems for manually bypassing channels or protective func.ions.

2)

Acc ss to all trip set ings, module calibration ad'us ~ents, tes: points, and signal injection points.

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Reactor Trio Sicnal Tes inc Provisions are made to manually place the ou.pu oi the bistable in a tripped conditicn for "at power" testing o= all portions of each trip circuit, including the reactor trip breakers.

Administra-t've procedure requires the inal element in a rip charinel (required dur ing pcwet operation) to be placed in the trip mode before that channel is taken out of service for repair or testing so that th single failure cri terion is miet by the remaining channels.

Provision is made for the inser ion o test signals in each analog loop.

Verification of the test signal is made by station ins:ru-ments a. test points specific lly provided.or this purpose.

This allows es-ing and caI:bration oi meters and bis:ables.

Transmitters and sensors're checked against each other and against pr cision r ad-cut equipment during norma1 power operation.

C.

RPS Analoo Channel Tes inc ine basic ele."ents compris',ng an RPS analog protac.ion channel are sho~n in Figure 7.2-7 of he FSM, and consist of a transmitter, po~er supply, bistable, bis able irip switch and proving lamp, tes. sicnal injection swi.ch, tes-signal injection jack, and tes point.

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Each protec.ion rack incluCes a test panel ccntaining the switches, test jacks, and related equipment ne ded. to test the channels con-tained in the rack.

A hinged cover enclcses the test panel.

Opening the cover or, placing the test-operate switch in the

" es."-.

position will initia.e an alamo.

These alarms are arranged on a

rack basis to preclude entry to more thah one redundant prote ticn rack (or channel) at any time.

The test panel cover is desicned such tha. it cannot be closed and the. alarm cle red unless ie test signal. plugs

{Cesar'.bed below) are re.-,cvea.

Closing the's-canel cover will mechanically return the test ski ches

".o the "oper'ate'"'osition

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Adr~n>strative procedures rewire that the bistable in the channel under test be placed in the tripped mcde prier to test.

This places a proving larp across the bistable output so that he bi-

'stable trip point can be checked during channel calibration.

Tne is. able trip switches. rust be.ranually reset. after. completion of test.

Closing.ne t st panel cover will not restcre

".hese switches to the untripped r ode.

,Administrative controls prevent the nucle r irstrument".ion source range and intermediate range protec.ion.channels rom beino disabled uring periodic tes-ing.

Power. range over-cower protec.ion does not have',an administrative control provision because there are. sufficient channels to satis y the single failure criterion during the testinq of circu ts.

Administrative con rois also prevent the power range Crcpped-rod protection from being Cisabled by testing.

In adoition, the rod position system will provide indication of an associated corrective ac icns for a dropped rcd condition.

Actual channel calibration will consist of injecting a tes signal frcm an ex.ernal calibration signal into he signal.injection jack.

Mhere applicable, the channel power supply will serve as a power sour for the calibration sourc and permit verification of tho output load c pacity of the po~er supply.

Test pain s are located in the analog channel and.provide an independent means of measuring he calibr ation signal level.

0.

RPS Locic Channel Testino The general Cesign featur'es of the RPS lcgic system~

are Cesc".i'ed belo~.

Tne trip logic channels for a typical two-out-of-cur trip

'unction are shown in Figure 7.2-8 of the FSN.

The analog cortions of these channels are shown in Figure 7.2-9 of the FSAR.

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bistable drives two relays:

The A and S relays -or level and the C and 0 relays =or pressure.

Ccn ac s from the A and C relays are arranged in a two-out-of-three and two-out-of-four trip matrix fo trip breaker No. 1.

The above configuration is duplicated for trip breaker No.

2 using contacts from the 8 and 0 relays.

A series

~ ~ 1 t

con, iguraticn is used for the trip breakers since they are actuated (i.e., opened) by undervoltage coils.

This approach is consis ent with a deenergize-to-trip preferred failure mode.

The planned loaic system testing includes exercising the individual reactor trip breakers to de...cns rate sys m.integrity.

One bypass breaker is used in conjunc ion with tes ing of the reac.cr trip breakers.

It is installed to allow opening the normal trip br aker.

To tes both reac.or trip br aker s, the bypass breaker must be used in.cne cell or reactor trip breaker A af.ar which it is physically moved to the cell associated with reactor trip breaker B.

One annunciator window on the main control board will indica.a tha. ¹ bypass breaker is closed in either cell.

Ouring normal operation,.

the bypass breaker is, physically 'remov d {racked out).

As shown in Ficure 7.2-8 of the FSN, the trip signal fram tha: logic network is simultaneously applied to the main trip breaker associated with the specific logic chain as well as the bypass 'breaker associated with the alternate trip breaker.

If.a valid trip signal occurs while bypass breaker AB-1 is bypassing trip br ker Na.

1, he trip--

breaker No.'

will be opened through its associated logic train; The trip signal applied to trip breaker No.

2 is simultanecusly applied ta bypass breaker RB-l, thereby opening the bypass around trip breaker No. 1.

Trip breaker No.

1 would either have be n oper, d

manually as part of.he test cr opened through its associated'ogic train which would be operational or tripped during a test.

An auxiliary relay is located in parallel with the undervoltage coils of the trip breakers.

This relay is tied to an event raccrder which is used to indicate. transmission of a signal through the logic net-work during testing.

Lights are also provided cn.he main control board to indicate

.he status of the indiv'idual logic relays.

In order to minimize the possibili.y of operational errors from either the standpoint af trippirg he reactor inadvertently or only par.ially checking all logic cambinatians, each logic netwcrk includes a logic channel tes panel.

This panel includes those

switches, indicators, and recorders needed to perform the logic system tes The arrangement is shown in Figure 7.2-10 of the FSAR.

The st swi.ches used to deenergize the.rip bis. ble relays opera.e through in.erposing relays as shown in Figures 7.2-7 and 7.2-9 oi the FSAR.

This approach avoids violating -he separation philosophy used in the analog channel design.

Thus, although tas switches for redunaant channels are conveniently grouped on a single panel to facilitate testing, physical and elec.ri>>

cal isolation of redundant pro action channels are maintained by the inclusion of the interposing relay which is ac.uated by the logic tes't switches..

Identification or the instr~entation protac-.ion system are provided by colared namepl tas on the cabinets.

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Enoineered Safet Features Ce.".eral Oescriotion Engineered safety feat.res (ESF) are provided in the facility to mitigate the consequence of the design bases accidents.

ESFs have been designed to cope wi.h any size reactor coolant pipe breaks, up to and including the circum erential rupture of any pipe assigning unobstructed dischzrg from both ends.

They are also Co signed to cope with zny s

earn or,eed-water line break, up to znd including the main steam or feedwzter headers.

ESFs in the Ginna plant are comprised o

the iollcwing sys ems:

Safety Injec.ion System (ECCS)

Contain-ent Sprzy 'ystea Contairiment

.Air Recirculation, Cooling znd Fil.ration Sys-em A.

Containment Isolation Sys-em Safety Injec.ion Svstem Emergency core cooling is provided by the SIS wnich cons itu.es

'he ECCS.

The SIS'compon nts operate in three modes delineated as passive accumulator inj ction,'ctive safety injec.ion, and residual heat removal (RYR) recirculation.

The primary purpose of the SIS is to automz, icz,lly Celiver cooling water to the reactor core to limit the fuel clad temperature, znd thereby ensur that the core will.

re..ain intact and in place wi h its he t transfer geometry preserved.

This protection is prescribed for all br zks (up to and including a hypothetical instantaneous double enCed rupture of the reactor coolant pipe), for a rod ejection accident, znd for a ste>s cenera or tube rupture.

For any rupture of a stem pipe and the associated uncontrolled heat removal from. the core, the SIS adds conc ntrat d boron solution to provide negative rezc ivity to acccmmodata the reactivity increase due to th to perature Crop znd a possible stuck rod.

ne principal SIS components tha provide core cooling irr.ediztely following a LOCA zre the two accumulators (one for ach loop), the

.hr SC,".-capacity safety injection (high-he d) pumps, and the two 100"-capzcity fiHR (low-he d) pumps.

For large breaks,'the accumulators which are passive components discharge into ihe cold legs or,. the reactor coolant piping, thus rapidly ensuring core cooling.

The safety injec.ion pumps are zctuated by two-of-the-thre lcw

'ressurizer pr ssures, or by t~o-of-he-three low s eamline pressures, or by two-of-he-hre high contair;ent pressures, or manually.

The pressuri-er pr ssure is monito". ed by pressure transmi".t rs with belier cap sul es.

13 The safety injection signal will open the SIS isolation valves and start the high-head safety injection pumps and low-head safety injection pumps.

Suction for the safety injection pumps will be aligned initially to a tank containing boric acid.

The suction for these pumps is transferred to the refueling water storage tank when the boric acid in the tank is nearly expended.

During normal plant operation, the two boric acid tanks are aligned to the suction of the high-head safety injection pumps.

The piping from the boric acid tank to the suction of the high-head safety injection pumps contains two independent parallel flow paths, each with two motor-operated valves (HOVs) in series.

The safety injection signal is applied to the HOVs in the suction line to assure concentrated boric acid flow to the suction of the safety injection pumps.

When a low level is reached in the boric acid tanks, the suction valves from the refueling water storage tank open and the suction valves from the boric acid tanks close.

The suction to the safety injection pump is then aligned from the refueling water storage tank.

In the event that the suction valves from the boric acid tanks do not open within two seconds after receiving the safety in-jection actuation signal, the suction valves from the refueling water storage tank open.

Redundant level instrumentation to the boric acid tank are used to switch the safety injection pump suction flow from the boric acid tanks to the refueling water storage tank.

The refueling water storage tank is equipped with one level indicator and two differen-tial pr essure switches which, together with the level indication system, initiate audible and visual alarms.

Each level channel has two alarms, i.e., the first low-level alarm and the second low-low-level alarms, respectively.

During reactor operation, the RHR pumps are aligned to the refueling water storage tank.

Because the injection phase of the LOCA is terminated before the refueling water storage tank is emptied, all pipes are kept filled with water before recirculation is initiated.

The level indicator and alarms on the refueling water storage tank warn the operator to terminate the injection phase.

Two additional level indicators and alarms are provided in the containment sump which also indicate when injection can be terminated and recirculation initiated.

After the injection operation, tpe coolant that spilled from the break and the water that was collected from the containment spray are cooled and recirculated to the RCS by the SIS.

When the break is large, depressurization occurs due to the high rate of mass and energy loss through the break to the containment.

When the break is small, the depressurization of the RCS can be augmented by a steam dump and auxiliary feedwater addition.

'L~) If the necessary RCS depressuri=aticn occurs be ore the injection mode of the SIS is terminated, the RHR pumps take suc ion frcii the containment sump, circulate tl:e spilled coolant throuch the r esidual heat exchangers, and 'return

.he coolant to the reactor.

If depres-surization of the RCS proceeds slowly, the safety in'ec ion pumps may ba used to aug,ent the head capacity of the RP3 pumps in returning the spilled coolant'o

.he reactor.

The recirculation surp lines c

...rise two independent lines which penetrat tha contair,-..ent.

Each line has a remo:a POV located insi(e and outside the'ontainment.

Each line is run independently to tha suction of a R:-:R pump..one system permi s long-t rm rec'~rcu-lation in tne event of a passive or active component failure.

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o ine.re-.o e-operated SIS valves which are under manual control {i.e.,

valves which nor.ally are. in their r ady position and. dc not receive a safety injection signal) have:Keir positions indic tad on a

conron portion of the control board.

Con a'.nment Spray S st n Tne coo.ainment spray system consists of two pumps, one spray additiv aok,.two spray.headers, spray nozzles, and the necessary piping and valves.

The system ini"ially tak s suction frcm the refueling wa.ar s.orace

.ank.

'~hen a low level is reached in the

'e-.u ling water storace, the spr y p

.-..p suction is fed from the discharge of the RHR pur:ps i continued spray is re"uired.

The syst~a Casicn condit~ons wer select d to be cc.",.patible wi h :he design conditions for the low pressure injection system since both of these systems share the same suc.ion line.

'ur',ng the period o

time that the spray pumps Craw frcm the refueling water.storage tank, approximately 20 gpm of spray additive (sodium hydroxide) will be added to the refualir,-" water by using a

liquid aouc.or motivated by the spray pump discharce pressure.

The fluid passing from the tank will then mix withthe fluid entering the pump suction.

The result will be a solution suitable for the removal of iodine.

7n spray sys em will be actua d by the coincidence of two sets of two-ou -of-three high coniainment pressure signals.

7nis s. ruing signal, entitled "Cortanment Hi-Hi-?ressure", NiII star. Ae pumps and open the discharge valves to the sprav header.

Tne valves associated wi h the spray addi:ive tank will be open d au.orna"<cally two minutes af er the containment spr y signal is ac:uated.

Sodit~ hydroxide will flow due

.o the suction of th spray oumps and mix wi:h refueling water prior.to being discharged scrouch the spray nozzle into the containment.

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~ A,.er the contain'ent spray signal is actuated, the cperator has the capability to stop the timer ii it has be n deteraineC that actuation of the sodiua hydroxide addition is not warranted.

The operator also has the capability to reinitiate the sod-:um hydroxide addition if required.

- E,"..ergency.procedures set

. orth guidelines for his ac.ion.

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C ntair.-..ent Air Recirculaticn Cool-;nc ard Fil raticn Svstm one contain;..ent air recirc"lation sys. i consisis of four air-handling systems, each includinc a mo-or, fan, coolina coils moisture separatcrs and high-efficiency particulate air (H=."A,'ilters, duct distribu ion system, ins.ru,".eqtation, ard con:rois.

The units are located cn the inter..ediate floor be. ween tne ccn:ain-rent wall and he prirary cc.-par=ent shield walls.

Two-oi-the-iour air-handling sys.ems are equipped with activate" charcoal-filter uni.s, which are ncr;..ally isola".

d frcc the ~~in air re ircu-la.ion stre~a and through which the air-ste~s mixture is bypassed to remove vola.ile iodine following an accident.

two of the air-handlirg assa...b1ies are required during '."e jos-.-

acciden. period d'or Cepressurization oi the c"n-.ainment vessa1..

Local ilow and

.e. perature indication of service water at each air-handling unit and

.he alaj-...s.indica ing abner.,zl service wa:e.

i low, te"..perature, and rad',oactivity ara provided in the control roc'.

Upon receipt o

either hich contain-ent pressure cr auto."..atic sa ety injection signal, the butterfly valves in the ccntain.-..ent recircula-tion sys.ems are tripped to the acc',dent position.

Accident position is also ie "fail-safe" close position.

~l Sutter ly valves are used to route the air flow through the c!i rcoal filters; these valves have only two positions, full open or full closed.

These valves are air oper. d and spring loaded..Upon loss of control signal cr control air, the spring ac:uates the valve to the acciCent position.

Redundant elec.rically operated three-way solenoid valves are used a

each bu aefly valve to ccn rol the instrument air supply (control air}.

These valves are arranged so hat.ailure of a single solenoid valve to respond to he acciden signal will not prevent act ation of he butterfly valve to the accident position.

The ccntairurent pressure is sensed hrouch six separat pr ssure transducers loca.ed outside the contairment.

Containment pressure is conmicated to the transduc rs.by three 3/8" s ainless st 1

li es pene.ratirg he contaiment vessel.

The hich ccntairr.;eni:

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pressure sigcal.fro>m these sensors trip:he containment isolation Ca~>pers and valves and sends' signal to start the fan motcrs - th remaining tuo motors not opera ing.under normal conditions, or. all four mo.ors in the case of a loss of outside power.

The autcmat-:c safety inje.ion signal is that resulting rca.uo-out-GT%.re' lcw pressure in the pressurizer, or froth hich contair>>-.ent pressure.

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Ccntai;...en.'.sol ation S

Cortai>rmen isola ion is initiat d automa ically by a sa ty.

lnJ c ion s >c"..al or ranual ly by one of t'uo switches on the main....

con rol board.

Contai.".-..ant isola.ion trips..he,cortainment sump pumps and closes all contain;.~nt isolation valves that are not r'equired to be open during an accident condition, which includes contairment

.sumo pu=p discharge isolat->on valves; stem~

generator bl>owdcun isolation.valves, reactor coolant drain tank vent header and pump suction valve.

gabe contair~ent isola". on signal also isolates four cortait.-..ent y rtilatioa purge vzlv s, two contair.- nt Cepressuri"ation'. valves, c"ntainm n'ir -.es. supply valve, two containm nt.air '- vent valves, and tr-'ps.the purge supply and exhaust fans.

The cont inment ven ilation valves also are isolated on high containmant activity or on manual contain.-..an spray.

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Re...ote opera.

d con+a'.nmen isolation. valves are ei.her air or motor opera=ed.

'~hen one air operated isolation valve is sed,:here are uo r lays in series to energize the solenoid..ach relay is operated

>rom a separate con.rol channel, each of which bas.an indeperdent dc pcwer source.

'Ahen two air cperated isolation valves in series are

used, ther is one solenoid. for each valve, each of which has an independent dc power source.

Ken a ro.or operated valve is used, he ac power is fed from one of two mo or con.rol centers, and each NCC is ied frcm a diesel powered bus.

In the FSAR, Sec.ion 5.2.2, the lic nse has stated that if,.in an emergency, only one diesel s.arts, then both NCCs.are autom>atically loaded onto the oper ating diesel.

This desicn devia as frcm current licensing criteria because this design challenges he inCependence of the redundant e.-..ergency power sourc s.

Tne contairi;ent isolaticn sys em can be reset by a manual sw-tch in the control room.

Some equipment would return autom>ai;ica'.ly

.o.he position prior. to the isolation signal.

Presectly, proc dur s require that he operator plac contair."..e... isolation valve suitches in the

'closed" position prior to reset ing containment isolation.-

Tnis current design on reset capability does not satisfy the HRC Lessons Le'rned Task Fore position, which requires that resetting of the contain en" isolation signa1 will not eesult in the autcma ic reopening of contair;ent isolation valves.

The 1'.censee has c"wit ed o modify the control circuitry o preclude the reopening of isolation valves.

ine modified design will be reviewed in Topic YI-4, "Containment Isolation'.

Encineered Safety Features Testing Safety Injection Sys em test is performed at each reactor refueling interval, with the r actor coolant syste.

pressure le)s than or equal to 3-"0 psig and t=-..perature less than or equal to 350 F.

A test sicnal is applied to initiate operation of the system~.

.The safety injection and residual heat removal pumo motors are prevented from s-arting dur'.ng the test.

The system is considered satisfactory if control board indication and visual'observa:

ons ',ndicate that all valves have roc ived he Safety Injection Signal ano have c"...pieced their travel.

xcep; during cold or refueling shut"owns, the s iety in'ec:icr pumps and residual heat removal. pumps are started at in.arvals not to exceed one month.

Acc ptable levels of perfor.-ance

.or the RHR pu.-..ps-wi ll be

".00 cpm at the minimum dischar"e-pressure of 140 psig.

Acc ptable-level oi perfonance for.he Si p~-...os will be

~O gpm ai the minirom discharge pressure of 1420 psig.

The spray. ad"itive valves ar tested at intervals not o exc ed one month.

With the puros shut dowo and the valves ups.-. earn and downstream of :he spray a"ditive valves closed, each valve. is opened and closed by operator ac,ion.

The accu",uIator cree'< valves are chec'<ed

=or operability at

r. =ueling shutdown.

B.

Containment Spray System test is performed at each reactor failure interval.

The test is performed with the isolation valves, in the spray supply lines, at the containment blocked closed.

Operation of the system is initiated by tripping the normal actuation instru-mentation.

The spray nozzles are checked for proper functioning at least every five years.

The test is considered satisfactory if visual observations indicate all components have operated satisfactorily.

Acceptable Iev'el of performance for containment spray pumps is 35 gpm at the minimum discharge pressure of 240 psig.

V7.

Evaluation Based on the information available on the docket, the Ginna plant testing program for the Reactor Trip System in general is in conformance with the reliability and testability criteria discussed in Section II of this report.

However, there are several areas in the Engineered Safety Feature System which are not in conformance with the criteria discussed in Section II of this report.

The following listed items sumnarize the major deviations based on the staff's audit review.

e 18 1.

The instrumentation strings from sensors thru bistable devices are not response time tested.

As a result, the testing required by IEEE Std 279-1968 Section 4.10 is not satisfied because the response time design basis (IEEE 279-1968 Section 3 (i)

) is not verified.

However, in a letter dated September 25, 1981, the licensee has committed to developing a response time testing program that will test all of the channels that are used to initiate reactor trip and engineered safety featur es except for the nuclear instrumentation, reactor coolant flow and the anticipatory trips that are not required for safety.

2.

The test procedures require that certain equipment be removed from service by racking out breakers and by pull to stop switches as well as the use of jumpers and removal of fuses discussed above.

These test methods violate Section 4.20 of IEEE Std 279-1968 because they are not annunciated to the operator in a timely manner such as to provide him with an unambiguous indication of the status of equipment needed to protect the public health and safety.

3.

As noted in Topic VI-4 we have also discovered that the override of an automatic ESF actuation signal incapacitates the system level manual actuation features.

VII.

CONCLUSION 1.

The licensee is in the process of establishing a suitable response time testing program as a result of the THI Lessons Learned Program.

2.

Plant procedure A-1103 provides an acceptable alternative to IEEE

.Std 279-196$ :Section 4.20 for annunciation of disabling tests, at older plants such as at R.

E. Ginna.

3.

The question of bypassing manual initiation of Safety Systems is.

being pursued under SEP Topic VI-4.and is of no further inter est under SEP Topics VI-7.A-3 and VI-lO.A.

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QNIl'ED STATQS NUCI EAR REGULATORYCOMM(SSlQN WASNING)QN,S Cg0500 Docket No. 50-244 LS05 O7 O65 Mr. John E. Maier Vice President Electric and Steam Production Rochester Gas 'g Electric Corporation 89 East Avenue Rochester, New York 14649

SUBJECT:

SEP TOPIC II-1.A, "EXCLUSION:AREA AUTHORITY ANO CONTROL" (GINNA)

Enclosed is the staff's final eva1uation.of SEP. -Topic;II.-1.A, ':Exc1usion Area Authority and Control" for..the R. E. Ginna',Nuclear;Power Plant.

This assessment incorporates the revision to your. Exclusion Area Boundary pro-'ided in your letter dated June 26, 1981.

Although;the area:-has been

changed, your authority to control 'that area. has.,not'.and therefore:the conclusions reached in our May-18, 1981.safety;.eval;uation:remain the.-same.

0 This evaluation will be a basic input to the':.'int'egrate4 safety assessment foi your fad$ ity unl'es's'ou identify changes needed to'.reflect the as-built conditions at your facility.

This assNessment may be revised in the future if your facili'4y design. is changed or if NRC criteria relapsing to

, this subject are modified before the integrated assessment is completed.

Si.ncerel.j,; -.

Dennis M. Crutchfield, 'Chief Operating Reactors Branch Nq; 5 Division of. Licensing Encl osure:,

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GINNA TOPIC II-l.A EXCLUSION AREA AUTHORITY AND CONTROL INTRODUCTION The safety objective of this topic is to assure that appropriate exclusion'rea authority and control are maintained by the licensee as required by 10 CFR Part 100.

V.

REVIEW CRITERIA Section 100.3(a) of 10 CFR Part 100 requires that a reactor licensee have the authority to determine all activities within the designated

area, including the exclusion and removal of personnel and property.

RELATED SAFETY TOPICS Topic XIII-l, "Conduct of Operations" will assure that the licensee can adequately specify proper operation in routine, accident and emergency conditions.

The topic is being covered as part of. the NRC TMI Task Action Plan.

Topic XIII-2, "Safeguards/Industrial Security" will evaluate the licensee's capability to protect the operating unit(s).

REVIEW GUIDELINES V.

The review was conducted in The capability of the plant the exclusion area boundary of the SEP review.

EVALUATION a

accordance with the. guidance given in.SRP.2.'1.2:-

to meet the dose criteria of 10 CFR Part 100 at will be'valuated in the Design Basis Event phase

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~.c >a~ q The R. E. Ginna plant is located on the south shore of Lake Ontario 16 miies east of Rochester, New York, a city of 241,539 peopIe (Rochester metropolitan area:

population is 701,745)

The site exclusio'n area, is completely within the plant boundaries.

The distance from the containment to the nearest site boundary (excluding the boundary on the lake front) is 1550 feet but the minimum exclusion distance is assumed to be 450 meters or 1476 feet.

The site boundary is shown in the attached Figure 1 of this

'valuation (note that additional land to the west has been purchased since Figure 2.2-3 of the FSAR was submitted).

No public hig'hways or railroads

'traverse the exclusion area.

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Rochester Gas and Electric Corp.

(RG8E) owns 'and Cpntro10;all.of cthe.land, including mineral rights, within the exclusion area.

Re'g'y'rdiog.tlie lake-,....

shore, frontage within the exclusion area,'GKE, by New York Stat'e,proceduvies,'wns the land above 243.8'sl.

This is well below the average lake stage of 246 feet msl, but is above the Extreme Low WaterlLeye1.of242;23,feel msl and the lowest regulated level of 243 feet msl...HoWeves'i,'.sirice,4he.,

low period is'enerally in the winter and the high. period. in,the sum'mer.,',it is not expected that there would be any "beach use" of thfs area.',.Previous experience confirms this; thus, no special precautions foi.,portentisa1 beach users are in place.

The exclusion area is not defined. over, the waters of Lake Ontario adjacent to the R. E. Ginna site.

The NRC st'aff,in recent,.

cases involving shore front sites has interpreted the, definition,of. an ex-.

clusion area in 10 CFR Part 100.as applying to the entire area. surrounding a reactor including the overwater portion.

In these

cases, aspplica'nt's,:.

have been required to make appropriate arrangements,to

.control wa'ter traf-fic within the exclusion area in the event of a plant emergenc'y'..'While,,...,.

RGEE has not, specifically defined an exclusion area over the w'ater'.,'.

asrra'nge-'ents have been made with the (B. Coast Guard, as document'e'd in' h e 'G'i,nna Plant Radiation Emergency Plan

, for the control of water traffiic'n the event of a plant emergency.

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'he lack of a defined exclusion area over the water adjacent',.to

.the pla'nt...,

site is a deviation from the staff's current interpretat'ion of:.the. criteiia in 10 CFR Part 100.

However, the arrangements made by RG5E wi'th,the U.S.,

Coast Guard meet the intent of the criteria.and, therefore, it is considered that the lack of a defined exclusion area over the water does not constitute a significant safety issue for the SEP. review'.

VI. CONCLUSIONS Based on the above evaluation we conclude that RG8E has the proper...authority to determine all activities within the exclusion area,'s required by 10 CFR Part 100.

This completes the evaluation of this SEP topic.

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REFERENCES 1.

Rochester Gas and Electric Corporation, Robert Emmett...anna,Null.ear,.Power Plant Unit No.

1 - Final Facility Description'and.Safety.'.Ana1ysis R'eport (FSAR), Sections 1.1, 2.1, 2.2 and Appendix 2C.

2.

Rochester Gas and Electric Corporai.ion, R. E; Ginna. Nuclear,:Power., P)ant Unit No. 1, Environmental

Report,

.olume 1, Section 2.1".'.

Nuclear Regulatory Comnission NUREG-75/087, Standard Review Plan;.,Section 2.1.2, September 1975.

4.

,",Official Preliminary 1980 Census Figures," taken from",the,Rochester Times-Union," September 15, 1980.

5.

RGSE Procedure "Radiation Emergency Plan" (SC-1).

6.

New York State Policy as established by. the Land Utilization.Department within the NYS Office of General Services.

7.

Letter from John E. Maier to Denni M. Crutchfield; June:-26;:.;1981.

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INFORMATION DISTRIBUTION ll'STEM (RIDS)

AOCBSSION NBR:8206070270 DOC ~ DATE: 82/06/04 NOTARIZED:

NO DOCKET FACIL,'50-244 Robert'mmet Ginna Nuclear Planti,Uni;t 1r,Rochester G

05000240 AUTH BYNAME AUTHOR AFFILIATION ORUTCHF IELDi D ~

Operating Reactors Branch 5

'RECIP ~ NAME RECIPIENTA)FILIATION SHEKMON~P.S.

ACRS - Advisor y Committee on Reactor Safeguards

SUBJECT:

Forwards safety evaluations of SEP TopicsiYols I-IIIiper App. E of NUREG"0821'Integra'ted-'Pl.ant,'Safety.:Assessment Rgpt for RE Gin a Nuclear Power Plant ~ "

DISTRIBUTION CODE:

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UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 June 4, 1982 Dr. Paul S.

Shewmon, Chairman Advisory Committee on Reactor Safeguards U. S. Nuclear Regulatory Commission Washington, D.C.

20555

Dear Dr. Shewmon:

Enclosed are three sets of the Safety Evaluation Reports referenced in Appendix E:of NUREG-0821, The Integrated Plant Safety Assessment Report for the R.

E. Ginna Nuclear Power Plant.

These references are presented for the use of the ACRS and its staff in the review of NUREG-'0821.

Complete sets of the Safety Evaluation Reports are also being placed in the NRC Public Document Room and the Local Public Document Room for ease of reference by the general public.

Sincerely, Dennis M. Crutchfield Chief Operating Reactors B anck No.

5 Division of Licensing, NRR

Enclosure:

As stated cc w/enclosure.:

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NC DOCKET FACIL:50-244 Robert Enmet Ginna Nuclear Planti Unit 1i Rochester G, 05000244 AUTH'AME AUTHOR AFFILIATIGN j

CRUTCHF IELO, D.

Operating Reactors Branch 5

RECIP.NAME RECIPIENT AFFILIATION Sl;EHMOfu,P ~ S.

ACRS - Advisory Commi t tee on Reactor Safeguaf ds

SUBJECT:

Forwards safety evaluations of SEP TopicsiVols I III<per App E of NUREG-V821~ "Integrated Plant Safety Assessment Rept for RE Ginna Nuclear Power Plant ~

DISTRIBUTION CODE:

HOOLS COPIES RECEIVED:LTR I ENCL ~ SIZE:

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 June 4, 1982 Dr. Paul S.

Shewmon, Chairman Advisory Committee on Reactor Safeguards U. S. Nuclear Regulatory Commission Washington, D.C.

20555

Dear Dr. Shewmon:

Enclosed are three sets of the Safety Evaluation Reports referenced in Appendix E of NUREG-0821, The Integrated Plant Safety Assessment Report for the R.

E.

Ginna Nuclear Power Plant.

These references are presented for the use of the ACRS and its staff in the review of NUREG-0821.

Complete sets of the Safety Evaluation Reports are also being placed in the NRC Public Document Room and the Local Public Document Room for ease of reference by the general public.

Sincerely,

Enclosure:

As stated

'N Dennis M. Crutchfield Chief Operating Reactors 8 anch No.

5 Division of Licensing, NRR cc w/enclosure:

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SUBJECT:

Forwards safety evaluations of SEP Topics~Vols I-III<per App E of NUREG Oedii "Integrated Plant Safety Assessment Rept for RE Ginna Nuclear Power Plant ~

DISTRIBUTION CODE:

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 June 4, 1982 Dr. Paul S.

Shewmon, Chairman Advisory Committee on Reactor Safeguards U. S. Nuclear Regulatory Commission Washington, D.C.

20555

Dear Dr. Shewmon:

Enclosed are three sets of the Safety Evaluation Reports referenced in Appendix E of NUREG-0821, The Integrated Plant Safety Assessment Report for the R.

E. Ginna Nuclear Power Plant.

These references are presented for the use of the ACRS and its staff in the review of NUREG-0821.

Complete sets of the Safety Evaluation Reports are also being placed.in the NRC Public Document Room and the Local Public Document Room for ease of reference by the general public.

Sincerely,

Enclosure:

As stated

'N Dennis M. Crutchfield Chief Operating Reactors B anch No.

5 Division of Licensing, NRR cc w/enclosure:

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