ML17252A722

Forwards NRC Final Safety Evaluation & Contractor Technical Evaluation Re SEP Topic VII-3,re Sys Required for Safe Shutdown.Facility Needs Addl Ac Onsite Electrical Source in Order to Satisfy All Safe Shutdown Requirements

ML17252A722
Person / Time
Site:	Dresden
Issue date:	07/07/1981
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:	Abel J COMMONWEALTH EDISON CO.
References
TASK-07-03, TASK-7-3, TASK-RR LSO5-81-07-010, LSO5-81-7-10, NUDOCS 8107130403
Download: ML17252A722 (19)

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• Docket No. S0*237 LS05-8l-07=010 Mr. J

• S. Abe 1 Director of Nuclear"Llcensfog Commonwealth Edison Company Post Office Box 767 Chicago. 11 linois 60690

SUBJECT:

SEP TOPIC VII-3 11 SAFE SHUTDOWN SYSTEMS SAFETY EVALUATION REPORT (DRESDEN POWER STATION. UNIT NO. 2)

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\\ is the staff's final safety evaluat.ion report (SER) on SEP Topi~

VII-3. This SER replaces the SER--fomarp~d by lt\\Y. letter of March 11, 1981. is our contractor's technical evaluation of your plant. This technical evaluation is basis for Enclosure. 1. The revisions to Enclosures

.i 1 and 2 have been made as a result of several telephone conferences between our staff's. and the contractor's evaluation of SEP Topic VI-10.B support the staff's conclusion that diesel generator 2/3 is essential to the proper func-tioning of the engineered safety features.

The need to modify the ac and de circuits associated with diesel generator 2/3 will be aqdressed in the staff's final safety evaluation of Topic VI-7.C.1.

Enelosure 1 is the staff's posit ton. with regard to your fac11 ity in the gen-eti.Jll su~ject area of safe shutcto~ except for diesel generator 2/3. With this exception. the staff has concluded that your facility meets current licensing criteria.

Enclosures:

As stated Sincerely, Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 Division of Licensing

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~TEMATIC EVALUATION PROGRAM

• • • • • • -*. TOPIC VII-*3

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TOPIC VII~3 SYSTEMS REQUIRED FOR SAFE'SHUTDOWN.

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• The sy.ste~.s:~aspects of the revie~i of.:.;S'.Ystems***Requir~d for Safe Shutdown*

was co.nducted as part of.Topic V-10.J3 CRHRReliabfl]ty).* This safety

• eval~ation is limi.ted to the e].,e.c:trical' instrumentation arid control sys-tem's-i.dentified as* bei:ig requff:'.'ed for safe shutdo1,;n;~

Plant systems that

~re ne~ded to ac~ieve and maintai~ a safe ~hutdo~~ ~ondition of the pl~nt, inc 1 udi ng the ca pa bi li ty for prompt *hoJ *shutdm\\*n of* the reactor. from out-side the control r'oom we*re reviewed:*: included also,~ 1.,ras a revieh' of the design capability and methpd of brtn~ing,the. pl~nt from a high pressure condition to* low* pressure cooling' assum:h1g the use of only safety grade equipment.

Th_e objectiv-es of the review ~;-~re tb assure:

(1)

The desig~ adequacy of the safe ~hutdown syste~ to (a) initiate automatically the operation of appr6priate systems, including the

-reactivity tohtrol systemsi s0ch that specified acceptable fuel*

design limits are not exceeded as a resul~ of a~ticipated operat-ional occurrences or postulated accidents and lb) initiate the

( 2)

( 3)

. operation of systems and componerits"required to bring, the plant to a safe shutdown..

That the required syste~s and equipment, including necessary in-strumentation and contfols to maintain the Dnit in a safe condition

• during hot sh~tdown; are* loca~ed at appropriate p~aces outside the contr61 room and have a potential capability for subsequent cold shutdown of.the reactor through the *suitable procedures.

. That only safety grade equipment is required to bring the reactor coolant syit~m f~o~ a high.press~pe condition to a low pressure cooling.condition.

II.

Reviev1 Cr1teria The review criteria are presented in Section 1 of EG&G Report bl21J

• "Electrical, Instrumentation, and Control Features of Systems Required for,Safe Shutdown;."

III.. Related Safety.Topics and Int~rfaces Review are~s outside the scope of this topic and safety topics that are dependent on the piesent topic information for completion are identified in Section 3 of EG&G Report.Ql2-J~~-

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'Review Guidelines*

The review guidelines are p~esented )n Section.4 of EG&G Report 0121J~

V.

• Eval uatfon

.. As noted ;~* EG&G Repor't 012lJ,. the :systems re~l)ired to take Dresden 2 from hot shutdown to cold shutdown,.assuming only offsite power is a-.

vailable or ~nly onsite power.is ~vailable and a single failure, are

• capabl.e of initiation t6 bring the plant to safe shutdown and ate in. *

• compliance with'.current licensing criteria and the safety objectives

. of SEP Topic VII-3, except that long-term co'oling (RHR) is susceptible to singl~ EI&C failures which render* this form of long-term cooling in~

operable if diesel generator 2/3 is not available to Unit 2 (e.g.-LOCA in Unit 3).

• v1;. Concl*usions Dresden 2 satisfies all of the requirements for Safe Shutdown except for a lack of.ade~uate onsite ~lectrical supply.

An additional ac onsite source is reqDired. so th~t Dresden 2 can be shutdown with a single failure in Unit 2 and a LOCA or similar single failure induced shutdown in Unit 3.

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SEP TECHNICAL EVALUATIO'N--RE'f>ORT

• ELECTRICAL, INSTRUMENTATION AND,CO,NTROL.JEATURES OF

. SYSTEMS REQU IRE'o* FOR SAFE SHUTDq~N..

FINAL DRAFT DRESDEN NUCLEAR STATION, UNIT 2

• Commonwealth Edison Company

s. E. Mays June 1981 0121J.

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• INTRODUCTION

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2.0 REVIEW.CRITERIA

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3. 0 RELATED.SAFETY TOPICS AND :INTERFACES >.. :. *.. '..*. '*..

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. 4.0, REVIEW GUIDELINES* * **

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. 5 ~O DISCUSSION AND.. EVALUATIO.N

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Instrumentatjon. ".... * ** : * *.*..-

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• 5.* J. l. Eva lUation

• 5.2* Safe Sh~tdown Systems......

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• 5.2.l. Onsite Power Una~~tlable I.,:

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• Ev a* 1 u at i on * *... * *

.5;2~2 Offsite. Pow~r Una~ailable=

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• l : Ev al u a_t i o ri*

5.3. Shutdown and Cpoldown Capabifity'01.Jts*ide ttie

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• Contrql Room

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• . 5.3.l Evaluation *

5. 4 RHR Reliability and I.nterl ocks. *

• 5.4.l Evaluatjon

6. 0..

SUMMARY

7. 0. SAFE SHUTDOWN EI&C FEATURES FOR CONS IDE.RATION BY SEP TOPIC III-1

8. 0 REFERENCES

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1 ELECTRICAL, INSTRUMENTATION AND CONTROL P"fATURES OF<!:

SYSTEMS REQUIRE~ FOR SAFE S~UTDOWN.

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• .. DRESDEN. NUCLEAR STATION,. UNIT 2

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• is part of the* Systemat*i C',:Eva'l uat 1on *Program (SEP) re~iew of***fopic VII--3, ;,Systems Requff~d for Safe Shutdowni". The objective of this review.is to determ,ne whether the electrical,

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instrumentation, and control (EI&C) *. feature*5**.,of the.systems required

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for safe shutdown, -including. suppp~t syste_111s, fl}eet current*.licensing

. requirements.

The systems required for safe -shutdown have been id.entified by the NRC SEP.

The systems were reviewed io ~nsJre the follo~in~ safety objectives are met:

(l) Assure th~ design ade~uacy of th~ safe shutdown system to automatically initiate qperation of.*

appropriate systems, including reactivity control systems, ~uch that fuel design limits are not exc~eded as a result of operational occurrence~ and postulated accidents, and to automatically initiate systems required to bring the plant to a safe s*hut-( 2)

• down Assure ~hat required systems, ~quipment, and con-tra 1 to maintain the unit in a safe '.cond it icin dur-ing hot shutdown are. appropriately* located outside...

the control room, and*hav~ the capability for sub-sequent cold; shutdown 6f the reactor ~sing suitable procedures (3)

Assure only safety grade equipment is tequi~ed* to bring primary coolant:systems ~rom a high pressure to low pressure cooling condition.

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The scope of this review specifical.ly includes an evaluation -of the electrical, instrumentation, and control features necessary for operation 6f the identified safe shutdow~ systems.

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rh:~' rev'iew:-evaluate?s~: the :systems :for Qper:abi lity *with an_d witho'ut.

offs ite power and* the ab11 ity *to.*.operate *:wlth. an*y single fafl ure.

The 1

EI&C review:.*af safe shu.td,own ~systems* only includes those -fea'tures not

• covered-under othe; SEP 1opici.

~pecifi~ i~ems ~hi~h will be covered.

under.other SEP reports are. id~nlif)~d* in Section 4.0~ Re~-iew: d:uide-

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. 2.0 REVIEW 'CRITERIA...,,.

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Current licensing ~~iteria for safe ~hutdo~n is 2o~t~ined in the f_ol lowing: *

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( 2)

( 3)

. (4)

( 5)

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IEEE Standard 279-197i,.

11 Cr.iteria for Prot.ect1on Sy?tems for Nuclear Power Generating _Stations" GDC-5,. "Sharing 'of Structures, Systems, and_ Com--*

panerits" -

GDC-13, II Instrumentation and Control:"

GDC-17, "Ele-ctric Power. Systems.

GDC-19, "Control Room"'.

• (6)

GDC'-26, "Reactivity Control -System Redundancy and*

Capability"

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GDC-34, "Residual Heat Remov*a l" ':

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GDC-35, "Emergency Core Cooling

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GD C -44, " Coo l i n g W ate r. "

3.0-_RELATED SAFETY TOPICS AND INTERFACES

. The following list of SEP*topics are related* to the safe shutdown topic with respect to EI&C features, but~are not being specifically_*

reviewed under th i.s. top i.c:

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SEP III-10.A, ~'Thermal Ov1:rload Protection for

- Motors of Motor-Operated Va 1 ves".

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SEP*rv."-2; .R~activity.Control.. Systenis.Inclu.ding.- **:.~..

Functional Design. and Protectioh:.:Against Single *.:.<' * :.

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. (3). SEP VL-7A3, "ECCS Actuation *system

(4) *SEP VI-'Cl,, Appendix K*~ "EI&C.Re~re_vi~w~".... *

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S.EP VI-10A,.. "Testing6f RTS-andESF. lnc*luding'.

• -_Response Time;! Testing".....

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• SEP;. VI:'."lDB, "Shared ESF, Ons i te ETrierge.ncy Power-;.:--'.

anei ~~rvice Systems for Multi'ple*_.Uhit Facilities.

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. * (7) *sEP... yU-1, "Re'actor Trip -System"

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SEP VII-2, ".ESF _*contro,1 Logic and."Design"

.. (9). SEP VIII-2, "On.site Eme,r:*_gency Power Systems--Diesel.*

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Generators" (10) SEP '.vrII-3, "E~ergency be Power Systems" (11) :SEP IX-3, "Station Service and Cooling Water:

Systems" (12) SEP IX-6, "Fire Protection."

Where safe shutdown system EI&C response ~s affected*by the above-men~ioned topics, ~hat p~rticular SEP review ha~ been torisulted for determination of overall safe shutdown ~ystem ~erformance. Where the SEP topic review is not available, the affe~t on safe shutdown system performance ha~ been*identified as being based on ~n assumed operating condition of the affecting system.

The safe shutdown review will be conside~ed preliminary unt~l r~solutibn of th~. affecting topic is com-pleted and found to*be in accordance with assumptions made in.. ~his review.

The completion of this review impacts upon the f6llowirig SEP top-ics, since capabilitie~ relating to safe shutdown is requir~d in the topic:

(l) *SEP VI-10.B, "Shared ESF, Onsite Emergency Power*

and Service Systems for Multipl.e"."Unit Facilities" 3

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• (2), SEP.virI-lA_, "Potential Equipment Failures Associ-ated* with a Degraded Grid Vo 1 tage*"

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SEP VIII-2,. "Onsite Emergency Power Systems_-~Di;E!sel..

Generators."*

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The capability to attain a safe shutdo~n has. been' re.viewed by evalu~ting.the syste~s:*used for normal*shutdci~n (o~site power not avail'"'.

able) and ~mergen~y shutdown (offsite power:notavailab:le).

SRP 7.4.

was 'applied. to each. s.ystem to e~sure ttie following gu i dE!l i nes were met: *

(l)

They have the required redundancy_ (SRP 7)

(2)

They meet th~ single fail~re cri~er~on (RG 1.53, ICSB BTP 18)

(3)

They,~ave the required capacity arid reliability -to perform int~nded safety functions on demand (SRP 7).

Additional.ly, SRP 5.4 requirements cont~ined in BTP RSB 5-1 were reviewed to determine if the systems _required for residual heat ~emoval met the following cr~teria:

( l). The system_s a-re capable of being operated from the control room with only offsite or only 6nsite power available (2)

The systems a~e capable of bringing the reactor to cold shutdown'with only offsit~ or only onsite power available within a reasonable period, assuming the most limiting _single failGre (3)

The RHR sys_tem has* the* required* isolation features to prevent OVE!rpressurization when RCS pressure is above RHR design pressure (4)

Protection from RHR pump overheating, cavitation,.

or loss of suction is provided (5)

Isolation and *interlock circui.try is testable during RHR operation and is tested in ~reoperational and initi~l startup test programs.

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at ion are bein.9 'Y.eviewed'.undet oth~r top1t'i*;~--as is _the seisni:fq\\equip-'

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t ment qualificat-ion,. -~nd*,are not* reviewed in;cthis re"por.L *;:,Sec'r:'ion,~i.O.

consiits 6f a l~st df safety.related EI&C.equipme~t ~~tessary for safe shutdown to be ~sed :in res.olving.SEP:Topi.~:.r~J..;.l,~,*~Class\\f~1_c~tion of

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Struc:tures, Components, and Systems."**,.*

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The SEP Revie*w of Safe Shutdown Systems, identified the' instrumen-

• tat ion avail ab 1 e in the<control rQo_m necessary Jo.bring the reactor.

frpm the hot shutdown i~ c~ld s~~tdo~n conditi~n.~'.:This review also.

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evaluated the nuclear instrumentation~ since t.. his parameter. must be*

monitored to ensure the reactor achieves. and* maint.ains shutdown condi-*

• tions!

V~rious ~ystem ~atameters, such-as LPCI pump runriing or valve posit_ion indications, are not included in the list of... safe shutdown instrumerits of Table 4.2 of the SEP Review of_ Saf_e Shutdown Systems.

This is due to the fact that indfcation is provid~d by.the.control/

operate circuitry.

Availabi.lity of cont'rol/ope~ate circui.try_to run the system also means availability' of the. re.quired indication.

Simi-larly, if the tontrol/operate circuitr~ is unavailable such that system operation is not possibl_*e~ then system ind.icatio'h is not mandatory.

The nuclear instrumentation consists* of t.wo independent buses providing redundant.indi~ation of each range bf p6wer level.

The buses are *powered by battery/charger *arrangements a*nd have no single failures that.would result in f.ailure of both systems.

The reactor parameter indicators (level, ~ressure/temperature)

• avai.lable in the control room are powered from different buses with pressure and level indic~tors receiving power from the 120 V AC essen-tial bus ~nd tempetaturS ~nd redundant level indication powered from the 120 V AC instrument bus.

Sin.ce the reactor operates in a saturated steam/water environment, knowledge of either temperature or pressure 5,

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• • imp'iies.*know~1edge.*of.the"othe~ *cus:e of,steam tables.:'cfr su.itab.le. cha~f

\\.. r~quiredLr No'..ele~-trica*i s.ing.le.fai.1dre.'would/r~:sult;n*f~;°l~re of'the l'eve 1 ~*** pre*s sure, *and i temperature

• i ~struments... Therefore,

• op~~ators would have the necessary information* avai.lable for determining reactor l,eve_l ~-

t-~mperat~re, *and pressure. '--...

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The. -i~strumerits -in the control room *providing indicaticin of reactor*..:~.;.r.

pressure, temperature,* and level~ as'we11*;as thoseproviping indication

• of ~arioui systems, flow, tempe~at~re; p~essu~e, ;alve ~osition, etc.,

- -~~e 1ndependent of th*e instrume11ts used to initiate Reactor Prbtecti6~

System (RPS) and En,gi.neered Safety° Features_- ( ESF} actions. *. *Fai 1 ure *of any of the instruments providing indication in the control room has no effect on ~he oper~tion of.the RP~ an~ ESF.*actuation systems:.

The indications for ~ewer to.the Narious AC and DC buses is sup-plied by lighti, meters, or alarms powe~ed from the bus' being monitored.

Loss of power to the bus would be indicated in the control room, and no single failures of indications would effect the ability to monitor any other bus.

Indication of Service Water, Co~ta i-nm_ent Coo 1 i ng Water, Core Spray,

• Torus, Isolation Condenser, Shutdown Cooling, and LPCI system parameter~

such as flow, temperafure, level,. and pressure available in the control rodm, are_ powered by_~he AC instrument bus.

Whil~ loss of the AC instrument bus would cause a loss of this _indication, each. of these systems has direct reading indicators available.at its lo~al control station. Status of fiow for some systems such as LPCI can be inferred from.the pump running/valve open indicators (not powered by the AC

• instrument bus) and by reactor parameters of level, pressure, and

• temperature.

5.1.1 Evaluation.

The instrumentation necessary for reaching ~nd maintaining cold shutdown at Dresden 2'meets curreni licensing criteria since no potential single failures of electrical equipme~t _could render vital i ndi cations necessary for maintaining p 1 ant cont ro*l i noperab 1 e ~

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The SEP review.of; Safe Shutdown°Systems ident.ffied't~e.* ~*ystems**;

• required for short-term cooling (imme'd.iately.. af.ter:reacto*r.:shutdown) an*d long-term *COO l i'ng<{-when the *reactor.js* COOled.to' th~>*:*s~DCS *design I

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temp~rature ~imit of 350°F) *with orily off$ite and; orily~n~ite power

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NormaJ *short-term cooling.is prov._'ided."bY ldumping *steam* from..the:.*

. r*eactor* to th.e main condenser vi a the~ tu.rbi ne bypass valve$.

The Ser-*

. vice Water Sy-~tem (Sws) rem~ves heat by conpens:ing the st~am.. The feedwater :system then returns the water the '.the reactor.~: This cooling method is only avai.lable when offs.He power: is_.,available.*.Failure o.f the feedwater control system, turbine hydraulic s,o'ntrol system, AC...

essential bus, or loss of SWS flow to the cond~ris~r can render this

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method of ~ooling inoper~tive.

The~ystems in thi~ metho~ are not cl~ss lE but are being considered as an available means to remove decay heat..

Emergency or alternate short...:term cooling* i nvo*1 ve*s operation of the isolation condenser (ICS),. autoblowdown $,YSte~ (ADS)~ HPCI system, or safety valves.

The !CS consists of a steam line from the reactor immersed in a large tank of ~ater which is vented to the* atmosphere.

The line ret~rns from the tank to the feedwatef line.

Flow through the system is initi-ated by opening a single MOV which allows th.e reactor to be co.a._led by

.boiling the water in the tank and returning condensed steam-to the reactor by natural circulation.

Failu~e of the ~ontrQl power or motive

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power for the *MOV.wo1,Jld disable the system although the. va.lve may be manually operated.

Failure of t.he radiation monitors or control cir-cuits for the normally-open MOVs could also cause system isolation.

The !CS i~ not a class lE system but is being considered as an avail-able means to remove decay heat~

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The safety ~valves: ppe~ate to :pr-ev.ent _overpressurization oLt~e, reactor~. The,:safety: valves *an:i.mech.. ~nical relief va.lves with *no.. eiec- :

tricaJ inpuis *and no manuaT operating capability.

They relieve pressure by dumping steam-into':the ~rywell.. The ADS_is<eiectrically initi~ted.

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and controlled and.. dumps steam in the suppr.ession.poo.l '(Torus) *. Ho~~:

ever, the ADS may be manually fnitiated to ~educe -reactor pressure and remove heat by dumping steam to the suppression* poo 1.. It wi 11 automat-

-**.J ically dump steam if.a high dr,Yweii pressure.or low reactor:level*

occurs *.

• In this case,<amea~~of.adding*water to the reac*tor (HPCI,°

• LPCI, or Core Spray) *is necessary to ma1ntain reactor level and.provide coolin.g.

The five ADS relief*vaives_.~re power~d by an autotran.sfer circuit for each valve :which prov~des 1~~V:DC power frbm redundant sources.

Thus, there are no.single fai,1ureswhich disable more ttian*

• one valve.

The HPGI system u~es steam from the reactor to turn a turbine-dri ven pump which prbvides water to the reactor while the turbine exhaust steam goes to the suppression pool.

A single MOV ppens to initiate the steam flow** to the turbine.

Failure -Of the contr-C:il power,

.motive power, or the co~trol circuitry for the val~e prevents HPCI operation.

Fai 1 ure of other normally-open va 1 ves *cou 1 q i so 1 ate feed flow to the reactor.

The SOCS is the system normally u~ed fbr long-term cooling (below 350°F).

It consists of a single. suction line; three parallel pump *and heat exchanger loo.ps, and a common discharge line..

There are multiple single failures which c~n render SOCS inoperable, including Joss of. the AC instrument bus which provides:cqntrol power for all SOCS MOVs, l~ss 6f 250 V DC R~actor Bus which powers the.pump suction and dis~harges normally-~losed MOVs, ~r loss of power at b~s 28-1 which p~wers the

~uction and discharge li~e normally-closed MOVs.

Other failures such as loss of bus 23-1, can: reduce system capacity to one-twelfth. its

• design capacity due to loss of SOCS, Reacto~ Building Closed Cooling ~ater System (RBCCWS), and. SWS.pumps.

8

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ThE! L1?-t:*ls* $Y.§ t~w..,'§~n§~ ~t;~ i oit:; t~*Q i; l'JQe-P.~:~£;;Qi~:nliqoP.,~i:"r~ctci~*rWi ~ it~

P~.~ 11 eJ @!1~5,.,.;Whl ch-= pump. W~terrfit:OIJ!~:~he ctQ~*§~tt~.t~f:iR: ~\\tt,ore.*~l(ji a a..

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heat* exchp.riger.T coo.1~ -by th Co.n:t'ail'1rnel'Jt-cGQo:l fri.g-:W:a:tet :s;i~~E(-C:CW.SJ.

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5. 2.. J _ _.O:n sjte P mve r-Uo a..V aj Jab le..::. ~;p-.re sden 2 JJ.b.rma l;l Y ope r a.te,s.::_wj th

• .half.O.~ j;!;~~~§9. y': ~~S~S: s~e~l.~ ~uY:~*!hi~~*~ i.i._*~~n'~-*~~~b).~tj);~*~~-E~~>;.

. a: ux: i 1 i ary t r9-n sf orm~r. _and. b alL f 1;,,om_tb,e * *reser;y.~ cwxi.li.ary,t,r..aii~former*.

-: -...._...... ~.. : -:... : r -*.--:. -..... :_* ~-. *.,. *.... *-.. *... :..

powered by the* l38-kV - grid~"'= Loss -of. th$. main -9e11.e.rator:.p,ower-,d.ur.i rig

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opera ti on Yf i J l,,fes ult in. a.. re actor.s_cram.-an.d_: _t.\\JJ:b 'Lne 0t r-i p. ~ '- -The0 bu.ses.*

normally p;w~r~d ':.by.. tbe -~~e~e~a-to'~.. wt 1~l.. tr:an-$-f~r~~w' _t.~e. ~~-r-v,e- ~

i 1.-i~ry t-ransformer whi.ch_ is-.~Jr~~dy:po~eri~g~t_h~ *~~sJ~:~o{_th~_=-b~u_s~~s_.~: O_n~l_y

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l.

si*ngle failures_jnvolving buses, svlitchgear, _e'-t_c.

do~nstr.gam ot the._*

t ran sf armer feed-lines. to_ the_ qi St.ri but i Or). syst~~ '_a?e_ *~o.ns~i_de_r_e-d_.~ *:.

- :.... = -

Single failures_Qf_O&C f~atu.~~s, such as-:-los_s __ of t~e.. _fe,.ed_wa,ter ~.

cont ro 1. sys tern, ex is( wh i c I;. co.~ 1 d g { s ~b 1 ~- the_~~

r::nJ~ -~tiort-t,e_rm ~c_o_o_l -

dowh methods.*

However, no EI.&C. *sing 1 e fa i. l ur_e_.r.end.e r_i_ng t.he. norma.1-_ _

• short-term cooldown methods.inop~rabl~ can.Als~ cau~e fail~re of the iso:lation condenser, ADS, and/or HPCI ;ystem.

There are multiple single failures previously* mentioned w~ich cah render the SOCS inoperable.

However, since only~one LPCI pump and loop i~ needed to maint~in core level and.provide c~oling water, there are no single.*f~ilures \\'Jhich can disable both the SOCS and lPCI systems.

Therefore, the short-term and fong-term coo 1 ing cap9~j 1 ity -~e~ts the cur.rent l i cehs in.g criteria \\'1-ith

. only offsite po_wer av.ailablE:*_.-.. ~xC.i~di~g_the. n*o_n-c:_las_s lE' ~et_h_ods (ICS and' normal cooldown),, the-requireci _ih~r:t-te_rm and long-term cooling is still provided.

.,~

5.2. J. 1.. Evaluation *.. The systems re~u..i r.e._d f_or s!1ort-_term. and l.ong-term*c~~li~~g--at*~o;~sd-e~ t~~e capa-b-1~- -o/prov-idi,ng 0

the r.equi*r~d~=

cooling assuming*no o~site power is available and a single failure.

~-

-~~:~=~~ *~~~- ~*'.

-- 9

• ~-... '**~_:)

\\'

'. ~5.2~2' Off site Power Uriavai.lable.*.. During normal oper'atien, *a* lo'ss.

. ~of off'site powe.r wiJl result, in a r'eactor. sctam,. turbin,e t_~ip~: arid

-~omentar/ioss* of pow.er to the AC d{strib~:tion* syste~.

S.u.bsequ~~tly;.

diesel generator 2 ( DG2') and DG2/ 3 will be aut~mati ca 11 y started to..

s.upp iy power *at bu.ses 24-l and 23~ l ~*respect i 1/e ly.

I

.1 DG20 is. shared between Units 2 ~and 3 and is oniy capable of pro*

,... '*. \\..

v'iding power t.o one unit at. a-fi~e.*~' Th~ *c~ntrol circu.itry will p~efer-.**

entially cause.DG2/3 to provide.power**t~* ~hichever unit is experien'cing *. *,*.

.. *.an accident..

Also, each unit has a normal/bypass* switch which will a*How/not allow DG2/3.to automatically provide power to that unit's buses.

There are ~o LCO conditions requi~ing that this switch be in *

"normal" during plant operations or describing conditions. to be met when in "bypass*."

However operating. procedure DOP 6600-4 prescribes.the proper.s~itch position.

Thus, if DG2jj is already s~pplying Unit 3 or Unit 3 has a LOCA coincident with Unit 2.losing of,fsite power or the*

normal/bypass switch is not in "normal", then DG2/3 will be unavailable to automatically provide power to Unit 2.

In that case, single failures exist that would prevent DG2 from supplying-power to Unit 2 resulting

• in a sustained loss of AC power~ The. JCS is designed to provide short~

term cooling in this case, but is not ~.class lE system.

The ADS and HPCI system are also designed to provide short-term cooling ind ~ater makeup without AC pow~r. Long-term cooling by the SOCS*~~ LPCI system would not be available due t6 the *1ack *of At power..

Assuming both diesel generators are available, there are no single failures.which would disable the lCS, ADS, and HPCI system.

Failure bf the 250 V DC Reactor ~us would-prevent -~nitiation of* the ICS and HPCI

• system.

However, the ADS, in conjunction with either the Core Spray (CS) or LPCI systems, would provide the required short-term cooling.*

As before, there are no single failures which ~ould disable both long-term cooli~g systems (SOCS and LPCI) if* AC power from one or both diesel generators is a~ailable~

• 5.2.2.l Evaluation.

The systems required for short-term and long-term cooling at Dresden 2*are capable ~f providing the required 10

a **

cooling assuming offsit~ power.is not ava.ilable.and.a single failure if DG*2/3 is available.

If DG 2/3 is not available to Unit 2, then single failures.can be postulated which disable the long-term c6bling capabil-ity necessary to reach and maintain cold shutdo~n.' However, short-term cooli_ng would still be available..,This is*a.deviation from the current

~icensing iequirements.

• I l'

5.3 *Shutdown and Cooldown Capability 00tside th~ Control Roo~

The capability to maintain the plant in hot shutdown from outside the control. room exists at Dresden 2.. Reactor parameters such as leve~,

press~re, and temperatu~e cah be m6nitbred at local stations outside the control room.

Reactor level has been previously discussed.* Reactor pressure (th~refore,. temperature) can be determined at.the ICS system local indication or by local indications such as system discharge pres-sure for LPCI or SOCS.

Local control stations exist for ~he pumps and valves of the systems required for.safe shutdown described in Sec-tion 5.2~ Additionally, many of the vatves are _also capable of being manually operated (such as the ICS return isQlation valve and the SOCS

~solation valves).

However, no procedures for taking th~ plant from hot to cold shutdown from outside the control room exist.

5.3.l Evalu~tion. Adequate capability e~ists to maintaih the reactor at hot shutdown from outside the.control room; No procedures exist for t~king the reactor from hot to cold shutdown from outside the control room.

5.4. RHR System Reliability and Interlocks The SOCS at Dresden 2 is designed to withstand RCS design pressure.

Therefore, the isolatio~ valve interlocks required by BTP RSB 5-1 are*

not applicable.

The isolation valves have interlock~ to prevent opening and to automatically close when RCS temp~rature exceeds the 350°F design temperature of the SOCS.

11

~***

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. ~*

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A.bypass* line p'rovid~s a flow path f'rdm-each SOCS pump discharge to its suction to provide the necessary f"1ow to'*prevent*purnv:;~ye~h~ating

'f:';:.. J

~..:

due to a discharge isolation valve being closed.. "A. lm:1 suet.run pressure

~1 trip.. \\vill stop the pump~ if.a suction valv-ehas c*losed.(j~r'ing operation, to prevent pump damage due to cavitatiOn:~:_ **:/:_..

J:\\;{l'*::

' ~

"t,._

I

'*{: --** j,.-:*~I There:.~ re. no requ i remerits *in the presd~n.. 2_ :~ec~~D=~.T}pec if i cations for testing *th'e ~~ocs interlocks and iSolatjofi *circuHi*Y*~u'ring socs

')

operation.. <Tile: electrical circuitry is not.desjgned to permit testing_

1-1hile.the'system is ope~ating 111ithocfl;a m6mentary i.nte*;ruplion in *system operation. A'lthough.licensed prior to the iSS!Jance of RG :1.68, -con-

~

ce rn i ng -preopera ti on al and st a rt up testing:, l;-.b'resden 2 conducted such

~

tests and t:ias demonstrated SDC.S.q_perabii.ity ori _.~;"~veral occasions as

~.

-~:

noted by the SEP Rev i ei*1 cif Safe Shutdown* Sys terns,,* 5ect ion 4. 5.

-~--~

5.4.l Evaluation.

The SOCS meets the current-licensing.require-ments of STP RSB 5-l in accordance viith.SEP Topics V-10.-B C:nd V-11.B.

6.0

SUMMARY

The systems required to take the react.or from hot shutdO\\*m to cold

. shutdowni assuming only off sit~ po0er is avail~b1e or ohly onsite power is available and a single failure, are capab1e of automatic ioitiation to bring the plant to a safe shutdown a~d are in ccimpliance with current licensing criteria, and the safety objectives of SEP 1opic VII-3 except.

that long-term cooling can be lost if DG 2/3~is not available to Unit 2 for the onsite power 6nly available scenario..

The instrumentation available to co~trol room operators to reach and maintain the reactor in cold shutdoi\\1n conditions meets curre-nt licensing criteria.

The capabil~ty to maintain the reactor in hot shutdown from out~ide the control room exists and is in compliance 1*;ith the safety objectives of SEP Topic VII-3.

NO procedures exist to take the plant from hot to 12

.. ~

. ~: :~ : '* "

cold shutdown from outside the cont~61 joom tri-satisfy the sajety.

crbject~~es of SEP Topi~ VII-3~

The SOCS satisfies the safety.requfre;ne~ts of SEP Topic V-10.B ahd v.,:.11.B for RHR*Systeni R:eliability,and Interloc_ks.... *

.7.0 SAFE SHUTDOWN El&C FEATµRES FOR CONSibERAT10N'.

BY SEP TOPIC III-1 ELECTRICAL DISTRIBUTION {including s~~port structure, but not iridivid~al loads)

l.

2.

ALL A~ 6USES (except 25-3: 25-4~' 26~3, 26-5~ 26-6; 27-2 27-3 27-4 27-5 29-4 29-5 29-6)--including a*l l feeders, incomirig.or outgoing, control c_ircuits, indicating circuits, bus work ahd support structures ALL DC BU~ES--including 12~ V, '~50 V, 24 V batteries~

chargers, ~reakers, bu~ work, and support structures 3:

DIESEL GENERATOR 2--including control and indicating circuitry, and control and indicatjon Df vital DG auxiliaries such as lube oil, fuel, and.cooling

4.

DIESEL GENERATOR 2/3--same as above

!NST~UMENTATION (including suppo~t structures) l~

REACTOR LEVEL.*

2:

REACTOR PRESSURE

3.

REACTOR TEMPERATURE 4:.

REACTOR PROTECTION SYSTEM

5.

NEUTRON MON !TOR ING (including in-core rnon it or i rig)

6.

AREA AND SYSTEM RADIATION MONITORING SYSTEMS (includes pumps,. valves, control, indication, and support structures)

l.

SHUTDOWN. COOLING SYSTEM l 3

-.~.

• r : ~*

.' ~*'

~...

"*/:.. *.

.*.1\\

~ '.. "

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• ...... \\

2.

REACTOR BUILOING CLOSED COOLIN~ ~ATER 4

SERVICE WATER SYSTEM LPC I.

5.

CONTAINMENT COOLING SERVICE WAT£~ :

i

.. l*..

6.-

CORE SPRAY

7.

HPCI,

-~-.

8.. ADS.

• ! J

_;Y,'

... ~~~"

9..

lS_OLATION CONDENSER 1 O.

TORUS (suppression poo 1) _.

~1 *

• .*~**.

~;;

11.

CONTROL ROD DRIVE SYS1tM (scram~funct~6n only).

..~**

':/,:.

8. 0 REFERENCES.

l.

Fin~l Safety Analysis Report, Dresden Nuclear Power Station, Units 2 arid 3.

2.

CE ltr (Peoples) to NRC (Keppler) dated February 28, 1980 concern-ing-control room instrumentation.

3.

CE Hr (Janeelk) to NRC (O/Connor) dated January 28, 1981 concern-ing OG2/3 bypass S\\11itch and oper<Jt ing pi-ocedure _OOP6600-4.

4.

Code of Federal Regulations, 10 CFR 50, i~.ppendix A, "General Design Criteria for Nuclear Po1*,1er Plants."

5.

iEEE Standard 279.-1971, "Criteria for Pi*otection Systems for Nuc-1 ear Po1*,1er Genera t 1 ng St at ions.".

6.

NUREG 75/087, Nuclear Regulatory Commis~~on Standard Review Plan 7.4; Systems Required for Safe Shutdovm" and 5.4,* "Residual Heat Removal_."*

14

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• ,...