ML17250A632

Forwards Current SEP Topics 3-10.A,6-7.C.1,6-7.F & 3-3.B. Assessments Compare Facility W/Current NRC Licensing Criteria.Requests Proposed Tech Spec Change to Prevent Transfer of All Four Instrument Buses to Same Source

ML17250A632
Person / Time
Site:	Ginna
Issue date:	08/20/1980
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:	White L ROCHESTER GAS & ELECTRIC CORP.
References
TASK-03-10.A, TASK-06-07.C1, TASK-06-07.F, TASK-08-03.B, TASK-3-10.A, TASK-6-7.C1, TASK-6-7.F, TASK-8-3.B, TASK-RR NUDOCS 8009230757
Download: ML17250A632 (26)

Text

Docket No. 50-244 AUG P g )ggg Mr. Leon.D. Mhite, Jr.

Vice Presi dent Electric and Steam Production Rochester Gas 8I Electric Corporat1on-89 East Avenue Rochester, New York DISTRIBUTION Docket NRC PDR LPDR TERA NSIC NRR Reading ORB ¹5 Reading DEisenhut J 01 shi nkki SEP BC /

SEP PF GLainas Dcrutchfield HS>>th Project Manager OELD OI8IE (3)

ACRS (16)

SEP File TNovak RTedesco SEP. TF

Dear Mr. White:

RE:

SEP TOPICS III-lO.A,V-ll-A, VI-7.C.l, VI-7.F AND III>>3.B (GINNA)

Enclosed are copies of our current evaluations of Systematic Evaluation Program Topics III-lO.A, Thermal-Overload Protection for Motors of Motor-Operated Valves; V-ll.A, Electrical, Instrumentation and Control Features for Isolat1on of H1gh and Low Pressure Systems; VI-7.C.l, Independence of Redundant Ons1te Power Systems; VI-7.F, Accumulator Isolat1on Valves Power and Control System Design; and III-3.B, D.C. Power System Bus Voltage Monitoring and Annunciation.

These assessments compare your fac1lity, as described in Docket No. 50-244, with the criteria currently used by the regulatory staff for li.censing new facilit1es.

Please inform us if your as-bu1lt fac111ty differs from the licensing basis assumed 1n our assessments within 45 days of receipt of this letter.

You are also requested to submit a proposed technical spec1fication change wh1ch would prevent the transfer of all four 1nstrument buses".tboth'e"sameesource.

These evaluations w111 be basic 1nputs to the 1ntegrated safety assessments for your facility unless you identify changes needed to reflect the as-built conditions at your fac1lity.

These top1c assessments may be revised in the future 1f your facility des1gn is changed or 1f NRC cr1teria relat1ng to these topics are modif1ed before the integrated assessments are completed.

S1ncerely, Denn1s M. Crutchfield, Chief Operating Reactors Branch ¹5 Division of Operating Reactors Encl osure:

Conpleted SEP Topics III-10.A, V-1l.A, VI-7.C.l, VI-7.F and III-3.B cc w/encl osures:

P OFFICE f>.

SURNAME D '.0 5/LA

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$ /y/80 DL:

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NRC FORM 318 (9 76) NRCM 0240 4U.S. GOVERNMENT PRINTING OFFICE: 1979 289 369

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+g*~4 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 August 20, 1980 Docket No. 50-244 Mr. Leon D. White, Jr.

Vice President Electric and Steam Production Pochester Gas E Electric Corporation 89 East Avenue Rochester, New York

Dear Mr. White:

RE:

SEP TOPICS III-10.A, Y-ll.A, VI-7.C.1, VI-7.F AND III-3.B (GINr A)

Enclosed are copies of our current evaluations of Systematic Evaluation Program Topics III-10.A, Thermal-Overload Protection for Motors. of Motor-Operated Yalves; Y-ll.A, Electrical, Instrumentation and Con.rol Features for Isolation of High and Low Pressure Systems; YI-7.C.l, Independence of Redundant Onsite Power Systems; V1-7.F, Accuaulator Isolation Valves Power and Control System Design; and 111-3.B, D.C.

Power System'Bus Yoltag Monitoring and Annunciation.

These assessments compare your facility, as described in Docket Iio. 50-244, with the criteria currently used by the regulatory st~ 'or licensing new facilities.

Please in orm us if your as-built facili~y differs from the licensing basis assumed in our assessments w",.hin 45 days of receipt o, this letter.

You are also requested to submit a proposed technical specification change wh',ch wo~ld prevent the transfer of all four instrument buses to the same source.

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

These topic assess

=nts may be revised in the future if your facility design is changed or if NRC criteria relating to these topics are modified before the integrated assessments are conqleted.

Enclosure:

Completed SEP Topics III-10.A, V-ll.A, VI-7.C.1, VI-7.F and III-3.B Sin erely, Operating Reactors Branch d5 Division of Operating Reactors cc w/enclosures:

See next page

Hr. Leon D. White, Jr. August 20, 1980 cc w/enclosures:

Harry H. Voigt, Esquire

LeBoeuf, Lamb, Leiby and tlacRae 1333 New Hampshire Avenue, N.

M.

Suite 1100 Washington, D.

C.

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

P.

0.

Box 5236 River Campus Station Rochester, New York 14627 Jeffrey Cohen New York State Energy Office Swan Street Building Core 1,

Second Floor Empire State Plaza

Albany, New York 12223 Director, Technical Development Programs State of New York Energy Office Agency Building 2 Esquire State Plaza

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

Ontario, New York 14519 Resident Inspector R.

E. Ginna Plant c/o U. S.

NRC 1503 Lake Road

Ontario, New York 14519 Direct or, Techni ca 1 As ses sment Division Office of R adi at i on Programs (AW-459)

U.

S.

Environmental Protection Agency Crystal Hall d2 Ar1i ngton, Virgi ni a 20460 U.

S. Environmental Protection Agency Region II Office ATTN:

E I S COORDINATOR 26 Federal Plaza New York, New Yorg 10007 Herbert Grossman, Esq.,

Chairman Atomi c Saf ety and Licensing Board U.

S. Nuclear Regulatory Comnission Washington, D.

C.

20555 Dr. Ric rd F. Cole Atomic Safety and Licensing Board U.

S. Nuclear Regulatory Comiission Mashingt.on, D.

C; 20555 Dr. Emeth A.

Luebk e Atomi c Saf ety and L icensing Board U. S. Nuclear Regulatory Cormission Washington, D.

C.

20555 Nr.

Thomas B. Cochran Natural Resources Defense Council', Inc.

1725 I Street, N.

M.

Suite 600 Washington,'.

C.

20006 g

C

SYSTEMATIC EVALUATION PROGRAM TOPIC IIX-lO.A THERMO-OVERLOAD PROTECTION POR MOTORS OP MOTOR-OP

~ED V ~VES Robert Emmet t Ginna, Unit No.

1 TOP IC III-10. A ThermalWverload Protection for Motors of Motor-Operated Valves The objective of this review is to provide assurance that the application of thermal-overload protection devices to motors associated with safety-".elated motor-operated valves do not result in needless hind".ance of the valves to perform their safety functions.

In accordance wi h this objective, the application of either one of the two recommendations contained 'n Regulatory Guide 1.106, IIThermal&verload Protection for Electric Motors on Motor-Operated Valves," is adequate.

These recommendations are as follows:

(1)

Provided that the completion of the safety function is not jeopardized or that other safety systems are not degraded, (a) the thermal-overload protection devices snould be continuously bypassed and'tempor-arily placed in force only when the valve motors are undergo'ng periodic. or maintenance

testing, or (b) those the. al-overload protection devi.ces that are normally in force during plant operation should be bypassed under accident conditions.

(2)

The trip se poin oi the thermal-overload protec-tion devices should be established with all uncer-tainties resolved in favor of completing the safety-"elated action.

With respect to those uncertainties, consideration should be given to (a) variations in the ambient temperature at the installed location of the overload protection

devices and the valve motors, (b) inaccuracies in motor hea'ting data and the overload protection device trip characteristics and the matching of these two 'items, and (c) setpoint drift.

In order to ensure continued functional reliability and the accuracy of the trip point, the thermal-overload protection devce should be periodically tested..

In addition, the requirements fo" TOLs as outlined in R.G.

1.106, the current licensi.ng criteria, require that:

(3) in MOV designs that use a torque s~itch to limit the opening or closing of the valve, the automatic opening or closing signal should be used in con-

]unction (se=ies) with a corresponding limit s~itch.

DISCUSSION A review of the dockets and the plant safety-related motor-s

~

operated valve (MOV) schematics d'd not disclose any thermal-overload (TOL) bypasses

'or the MOVs.

There was no information in the NRC docket ind'cating that t'ne trip point settings of the safety-related MOVs were estzbl'shed in favor of completing the safety function.

The MOV schemat'cs indicate that, on all automat cally-or manually-operated

MOVs, a limit switch is in series with a torque switch in the open ng circuitry or the valves.

These switcnes open the opening cir-curt when the valve is in the full-open position.

However, on valves that are automatically-closed, on'y z torque switch is used to stop the closing of the valve when it is fully closed.

EVALUATION Accordingly, we concluded that the R.

E, Ginna Unit No.

1 Nuclear Plan" does not satisfy current licensing, requirements for (a) bypassing

or, establishing t at the TOl. trip setpoints are set in favo" of com-pleting the safety-related function and (b) use of a limit switch in conjunction with the torque switch to limit the automatic closure of valves.

REr ERENCES 1.

Rochester Gas and Electric Corp.

Dwg. No.

10905, "MotorWperated Valve 'Elementary Miring Diagrams."

2.

IE.E Standard 179-1971, "Criteria for Protection Systems for Nuc-lear Power Generating Stations."

3.

Branch Technical Position EICSB-27, "Design Criteria for Thermal Ove load Protection for Motors of MotorWperated Valves."

Regulatory Guide 1.106, "Thermal Overload Protection for Electric Motors on MotorWperated Valves."

S AL SEP TOPIC V-,l1.A S""P TiCHHZCAL ""VALU.-".T10N R~T

""L~~TR1CAL,:>>S RL:.."-.N

~ A

~ ON. ZiD CO'iTROL:"-AT'R:.S 0 -'.T:ON 0" HTGil. 'iD LOli 23.:-SSUR"-

S"-'ST:".1S

"OR R. ". G"NNA NUCLZ ~.

STAT 0!t

'.0

STRODUCTTON The purpose of th s rev'ew

's to detertline '" the e ectrical,

'I

%>>+>>>>

kaav l 1 I Qcl >>al 9

~ )

2nd cont o'ZX5C) fe" s usec to isolate svstesls pressure rating than the reactor coolant ori~ary ".ste-,

'-., Co-o ance

~ >>>>- n c rren

'c nsing require=-"ents as outlined in Topic 'It-1 lA.

Current guidance for isola".ion of high and low pres" su e

$ ) s te~~s

"$ cortained in branch Technical position (BTP) ".ECS3-3, 3TP RSB-5-1, the Standard Review p ant (SRP),

Sect on 6.3.

2.0 CRTZ=R A l,as: ru 1

1e Rea oval (RHR) S)'stegs

~

o R'R sjstevs contalnec in 3T RSS 5

1 are l T so lat'

-.equi" events (1)

The sue ticn sice must oe provided with the =ollowing

'so at'on features:

(a)

Two power-ope'rated valves in series with posi-tion 'ndicatec in the contro1 roo~.

(b)

The va'ves rust

'have incepe-..dent and d'verse inta" loc';.s to nrevent open'ng if the reactor coolant system (RCS) pre sure 's

-bove the design pressu e of the R:-:R system.

(c) ':

v ves ~ust have i; de'Dende lt a..c cive se inter'ocks to ensure at east one va ve c oses upon an 'ncrease

n PCS pressu' bove ti:e desi.gn p-essure of

.le

+3,

$ Ste~

(2)

T':le c~scha

~o c41ng featur ce nust be provided with one o"" the

~S:

(a)

The v descr 2 ves)

'oos>> tron,nc~ cators

~ anc l>>terlcc'rs ibed in (l)(a) through (1)(c) -bove.

(b)

One or Ror e normal lv c 'he"'c valves n series w'n a

osed powe.-operated va ve wr.ich has

pos'tion ind'ca ed in the col.

ol room.

i i

e i

1 1r ti' va Cool.'..g S:

OP ell UPor.

(S=S)

-;h=n P~-'P s vs etl Eve 1s usea sac=.

("-CCS) ecelpt oz a

PCS pressur Ges1gn p es 0

Ul i su e

~

an

""emergency Core c t 1oil ) the va lve ~us t 1r.~ ec 0

1 s'na-as decreased below (c)

Th check valves

'r. se" 'es.

(d) '0

'i check valves in series, prov'dec that both be, per oa'ca ly checkea

".o" eak, tightness a"e checked at least annual

~ y.

=~e""- ""~ Co-e Coo~g cy~".

solation reauire. eats Cov

.CCS are conta'ned

'n S."-~~ 6.3.

s'u>>

~ za<<lon Rust

'Beet aisle 0<<<<ne 0

ov ng 0= "CCS to preven- 0;erpres-reatures One o" ore cneck valves in c'osed

=o"or-cperated v 've cpened upon

- ece1'pt 0r a S' less than the

.CCS cesign pr series>>itn a nor-..a'ly-(!!OV) "nich is to be hen nCS pressure

's essure (2)

Znree check valves in series (3) wo check

<<. 'v s in se ies, be pe"ioaically checkea

='or check ec a t

~ ea s t annua 1ly.

nroviced that both a y 2.3 Othe S =ate-s.

.11 o her o" "" ssure sy tens interiacirg

<<W I<<

1

<<itil the PCS i Ust u.ee the Zolaowing 'ole icn reoui e ants r0 BTP;".ZCSB-3:

..t 1 ast t;o va ves

'n se"'es

.Ust be provided to iso'ate the system l"hen RCS pressure is above the s.:stem design pressure and valve positon should bc prov=dec

'n the control roon (2) tor sy 1ilaeDe ste"..s.>>'h tvo Y~OVS, each i!OV i

ncent ana diverse inte"Locks snou Q have o prevent open 1Ges RCS g unt'1 PCS p" ssure

's

=ressure and should au" below o~atac essu.e

'nc eases above svstei the y s. ea al y c'052 <<hei design press re (3)

For s' shou d

p essu au too~a <<

svs"e=

s one check be interlockea to p" e 's above sys<<e.il de ca ly c'os

~neneve ces':

l p essUre

~

va've and a

."!OV the

'AOV vent opening i=.

BCS s'gn p essu e

cnd should RvS pressu e ezceeds

".G

""TSC"SSXON...iD ="-" U-'"

-'ilCi2 c sys at R.

~. Ginna

.~'

=-"-" Stat'on wh~cn..ave C

I pressU ante p.ce rating oi with the RCS pressure bounda"y

-nd ha;e a cesi n

2 0

part oi the 8;i s" Q..l wn cil Ls

'Qs5

'- nan tnP t 0i RCS (CVCS),

<<gs C

~

Sc c

) ngec thon Systetl (S:S),

ana the Rss d a'.".Qat Re=.oval These syste

.s are the Chem cal anc Vo'."-2 Con roi System (R'P))

SyS )teil i

.Qsicua i.'.Qat Re-.o:a.

Vstem.

he R~R syste=.

takes

- suc-ion cn tne RCS loop A ho" leg, circuicres the -ater through the R11R system he t Q.lchanger, and discharges to the RCS loop B cold leg.

'wo.~otor-ope aied v" ves '

se

')es p ov 02 solat~ cn c-'pab '- i2s

. Tl bo't 1

th)e s ct'on 2nd d'scharge 1'nes.

Cach Oi

<< leSC i'OVS ilP 5 pC'S~ 'n

. ild'a tion in the control roo~.

The nboard (closest to the RCS) valves are

'nterlockec to p=event openin>> " 'RCS pressure

.'s above RER systez des'gn pres ure.

hoWever Oo.;1 VP VQS use 1'ls Sa...e:2 5 Su 2

SWltC:1

-Tld relay to p"ov'e th's 'n"erlock.

T'ne outooard valves have no pressure None o= the 'a ves w'll auto a"'c y c ose i= RCS pres-sure increases aoove

~>".R system desi'ressure curing R=R systetl Ope atl.oTli

~ ne

<<'".R sys" 2-.

's not 'n co-pli-nce wi"='h the eouh QLlisntS 0

3 ';.S3-5 1

si?lce

'none 0). the 'o i P t 0;1 v" ves wi 1 1 auto=atica v cicse iP RCS pressure evceeds R:":R design pressure.

Also, the cut'ooard iso =".'n valves

'na'e no interioc' to preveT't R:":R overp es '

z 0 1 aTld the in'ooard valve inte" locks P, I ie d verse nor independent.

3.2 accu-.ula tots

>re ssur ized w'-th Tl' ogen

<< irn a:s=y ngectxcn Svste~.

One S

S sUDs pste..l consLsts oi two Qac)l i".CC" u ' t0 3.501P te>>

OL the 0"

QPC:1 Ther

""2 CCT.'.i C t LOns uPS t Qa)Q chieck va 'e t).P r. can al 'ow tnD to oe tested

~

ncr.aiiv-oven r

RCS by pa'r 0=.

check va ves

~

~tor op2 a).ed

') 50 'tion valve upstrea~

o" tne check va ves or each accuicuip. tor iles pcs li~ C.l indic-t'on in the control.

<<ooi

~

>>acn Ov 's

'pened aut0

=

11

'ign 1.

l c osed1 uDG'1 ece

'pt c

sa etv ',1'le t Gn The other S:S subsyste

. consists Of

-= 'o

cops, eacn " pplied by a or one RCS loop.

Isolation s-"ety ngect on pu-p.

"ach pu=p discharges to the hot and cold

'egs is prov ded Dy t"o che "c va 'es in se" les

=Gr eac'n branch of tne safety iniection loop.

The check valves in the lines supp y:ng tne RCS hot leg for each SiS 'oop are not testabie since there are no locat:ons wnere leakage cou c be determined between the va 'ves or 'here tne outboard va1ve (far hest frot1 the RCS) could be pressurized.

Ti<<oto ope'ted isolation va ve 'h Doslt on ind cat 01 in

. ne control ron~ 's prov dec

'n each branch Gf the S:S iooDs.

These va 'es 0 pe:i upcn receipt of a sa ='y injec io.. signal, out have no

'-nter ocks preventing opening when RCS pressure is above S:iS design p"ess. re.

lne S

S ls i.ot 'n co.iip 'nce wltn

-ents of SBP 6.3 since t'ne check valves tne current iicens in"- re Gu' e-the safety ngection pump to the RCS hot legs are not testabl.e, and the HOVS 'n 1ines have no nte" 1o..'. s overpressurization.

to p eve 1

sys teR 3

~ 3 Chew c - 'nd VG iupe Con i ol S

steep.

~

.'he CVCS t"-'r.es wate or. the RCS and passes through a regenerat':e heat

xchanger, or 'ice to reduce its pressure, and

= nonregenerative heat ezchan" De. Gre va lve r reducing ltS 1 te"ng a..d cleanup, the water wa:

be returned pressure further by the use of a pre sure co.lt 01 to the PCS by the use of tne charg'ng pu-ps,

.>>h'ch increase the water p essu and Dass it Ough

"- 't 1ve h e s. t.

t.'1e ho t c" co'd legs o= t'ne RCS or to the p essu-"er au>:iliarv spr-y line.

~PCS s c"on:ne isolation

's prov'd

'Gy a =anua ly-operatea so e>>G' v"- lve ln 52 's wltn t?i ee p -

a le 'olenoid-operated v-ves.

-.ach or hese valves ls 0'Dera ed 2 ois the co.it 0 1 Lon 1

nd DGSlt-0 1 lndlCate

~

None OX the Va iV S

haVe 'te 10C~RS to aasVa i Ve prevent

ona:1 'g ol to rat ng G: the

=,atica i v c cse i-tha pressu

=':caeds the dasSn

're ressu Q.

uo Gils or the

}ste 1 ~

he ChCS disch=

g

ne isolation

's provided by co on dis-CA.i gd tla check va lve ai d a

o

-ncn check va va in e"- ch c L he th b

nenes cob. s rea o

the co;.,G l c lack va 've.

D a n f'

'~'5 on the cisch Q

l ine Ups a 'H o f each check valve can

-. i'v the valves to

=e "asted.

.hera

's no pcs.'t'on indication avai zola in the control roo=

ror tne c..eck valves

~

"here are so anoid iso'at'on valves

n each dischar"-e

'ne or-..ch;n:ch have poson ind'caton the control

rcoa, but these valves lava no inta" locks to ovaroressurlzat'n o

prevect syste~

~he CVCS is not

'".. Comnlianca vith curra..t 1 censing reouir

.ants for 'solaton of 'n'h

"-nd 1 0 w

'9ra s s U << e systa

.s cG.tamed in 3P:TCSB-3 s nce the s ct'on and discharge 1;.e solenoic-opera ad va'ves have no in e" ocks to prevent system ovarpress ri-ation, and the d scharge 1'ne.

c'neck valves inava no pcs~ tion incicaticn a'a'labia in tha control roo..

4.0 S'Z~~'CRY The R.:". Ginn

!i clear Staton has three s ist =s vi"h a lo"er design pressure ratin~~ than "he RCS, -hicn a=e directly connected to the RCS.

ha

CVCS, STS, and 3 2 syste do nct aet cu rant icensin~

rec ire=ants

-ed be 'Gw.

for isolation of high and io" pressUre syste~s as speci The CVCS solenoid cp re ared in arlocks, va,ives have no oos't contre roo; as rec crated valves have no nressure-nd the cischar~

'ne cneck cn l.na.c t.on -va. Laoia xn the rad by i"'P ?:CSB-3 ines

"-Ging to the RCS d "ne ="otor-operated su a related in er

<< i lsolat cii va 'vas have

..0- p as locks as raGU'ed by S.iP 6

(3)

'On va VQS auto ati-ncraases bove P"8 s) ~ teD operati<<on

)

Pone o= the

?-:3. systesl 'so'at cally c osa i=

RCS pressure SVSte.l O<s<<bn prQSSUra CU--n~

5

~ eA eel)v r 'cc~s:o n~ x,one ce lo 0 la1'eon va lve s flave

.,o Q~es sl's reo'ed

"'." i""

RS5-5-1.

ne

'-:.board iso a"'on 'alves are indeoenden".

5. 0 K:"-H,Rib~:-S G 0/3/087

)

rancil 'cnn ca 'osh<

ons S iznda c PievlevT Plagal 0

o 0 ~

". T CS"=-3 ) P.SB-5-1

)

t)ada G~-<<

Vgc 1 ea.

- c.' '. y De, s cr'"='o;.

Boxer Plan-) Unir No.

~a'e"y za ys's 4e ))or 4 )

3.

RG~~ crawn~s 33013-<<22)

-~2-.') -

,'a) "<<26) -+2)7) -<28) -432)

-433)

-43<<I ) -<<35) and -436.

I,<<

~

10.05-280)

-285) -287) -295

-296, -300, and

SEP TOPIC YI-7.C.1 Tr.CMICAL EVALUATION INDEPENDENCE O.= R=DUNDAVZ ONSITE POWER SYSTEMS R. "". GINNA NUCLEAR STATION

1.0 INTRODUCTION

The objective'f this review is to determine if the onsite elec-trical powe" systems (AC and DC) are in compliance with current licen-sing criteria for electrical independence between redundant standby (onsite) power sources and their distribution systems.

General Design Criterion 17 requires that the onsite electrical power supplies and their ons te distribution systems shall have suf-ficient independence to perform their safety function assuming a single failure.

Regulatory Guide 1.6, "Independence Betwen Redundant Standby (Onsite)

Power Sources and Between Their Distribution System,"

and IEEE Standard 308-1971<,

"IEEE Standard Criteria for.Nuclear Power Gen-crating Stations" provide a basis acceptable to the NRC staff for meeting GDC 17 'n regards to electrical independence of onsite po~er systems.

2. 0 CRITERIA 2.1 AC Supplies.

When operating from standby

sources, redundant load groups and redundant standby sources should be independent of each othe" at least to the following extent.

(1)

The standby source of one load group should not be automatically pa=alleled with the standby source of another load group under accident conditions (2)

No provisions should exist for automatically trans-ferring one load g oup to anothe load g oup or loads between redundant power sources

(3) If means exist for manuaLLy connecting redundant load groups together>

at least one interlock should be provided to prevent an operator erzoz that would parallel thei" standby power sources.

battery and battery charger.

The battery-charger -combination should have no automatic connection to any other redundant d-c load group.

3.0 DISCUSSION AND EVALUATION 3.1 AC Supplies Discussion Ginna onsite emergency AC power svstem consists of two redundant diesel-generator power trains.

Diesel generator lA (DG1A) suppl'es 480 V buses 14 and 18 while dieseL genezatbr 1B (DG1B) sup-plies buses 16 and 17.

Manual means exist to tie buses 17 and 18 through a "ie bzeakez and to t'e buses 14 and 16 through a tie breaker.

The control circuit for each breaker provides interlocks such that the breake cannot be shut if eithe DG is closed on either bus or if the normal feedezs to the bus are closed.

Additionally, if the tie breakers,are

closed, they will trip open upon restoration of normal power, DG closing on the ous, or any safety injection signal.

Manual means exist to power safety injection pump SI-LC from eithez bus 14 o" 16.

The control circuit fo" the breaker from each bus is des-'gned such that shutting of one breake" prevents shutting the other breaker so that paralleling the redundant DGs is prevented.

Instrument buses 1A, 1B, 1C, and 1D are capable of being supplied by multiple sou.ces.

Each bus is supplied by a pai-of mechanically interlocked b ea'kers such that pa"alleling of redundant sources

.is prevented.

Evaluation.

The redundant onsite AC power t ains have no auto-matic transfers of loads and/or load groups.

The manual transfer of load groups or manual interconnection of emergency buses have the required interlocks to prevent inadve tent paralleling of redundant sources.

Therefore, the onsite emergency AC system is in compliance with current licensing requirements for independence of onsite power systems.

3.2 DC Systems Discussion.

Ginna Nuclear Station has two redundant battery and charger trains to supply 125 V DC emergency loads.

Each train consists of a battery, a 75-amp charger, and a 150-amp cnarger.

Means exist to interconnect ooth trains by manually shut t'ng a tie breaker.

This breaker is padlocked open and the k'ey is maintained by the shift foreman.

Current operating procedures require removal of the feede" fuse from one of the buses feeding the tie breaker prior to closing the tie breaker

Powever, no interlocks exist to prevent closure of the tie br'eaker if the feeder fuse has not.been removed.

This would allow paralleling of the redundant DC trains.

Automatic transfer of 125 V DC load groups from train A to B (or vice versa) occurs in seven locations.

Control power for 480 V switch-gear on buses 14, 16, 17, and 18, DG1A control panel, DG13 control

panel, and the rod drive )fG set control panel automatically transfers to the redundant train on a loss of power from the normal source.

Each load twll automatically transfer back to the normal supply when it is regained.

Evaluation.

The 125 U DC system has one'anual tie between redun-dant.trains and seven automatic transfers of power from one redundant train to the othe=.

Although administrative controls are provided to prevent paralleling redundant trains via the tie breake, no physical or electrical interlocks exist to prevent parallel operation of the two

trans.

Therefore, the 125 V DC system is not in compliance with cu-rent licensing requirements with respec to independence of onsite power systems.

4.0 S~MY The review of docketed information and plant electrical drawings indicate that the Ginna Nuclear Station onsite AC redundant power sources and distribution system meet the current licensing requirements for independence of onsite power systems.

The 125 V DC system has seven automatically transferred loads and one manual tie breake which are not in compliance with current criteria for independence of onsite pave" systems.

5.0 REPERENCES 1.

General Design Criterion 17, "Electrical Power System," of Appen-dix A, "General Design Cri.teria of Nuclear Power Plants," to 10'PR Part 50, '"Domestic Licensing of Production and Utilization Pacilities."

2.

"Xndependence.ietween Redundant Standby (Onsite)

'Power Sources and Between Their Distribution Systems,"

Regulatory Guide 1.6.

3.

Rochester Gas and Electric Corp. letter (@hite) to NRC (Ziemann) dated April 18, 1979.

4.

RGSG Corp. drawings 10905-59, 62, 63, 74, and 75.

5.

RGGG Corp. drawings D-206-51, 21489-269, and 33013-652.

c D

~ C" j C" 'T

""V-"LU-'T~O" TOPTC Vi-7 ACCv"'A3i='TGR TSOLATiON VPLV""S DO~iiR P'i3 CONTROL SIST:M D=STGH R.

GlNMK 1.0 NTPODUCTiOH Th objective of

'solation va)ve powe"

censing criter'a.

th.s review s to deter'ine '= the acc,u'ator anc control system, is in compliance ""th current The speci" ic requirements fo" acc -ul ato" isolation valve power and control system" design cer've frow i"=:.:-

27 - 971, the bvnass o

a protective function w:11 be re-oved

-nlch st" tes that autoaatically when-eve Deraissive conditions are not et and wh'h a

so assures that s ngle elec c

1 allQ e or c pab lity of the accunulato criter'a

.2 furth

= de "ned 2nd iCS3 lS ln 3ranch 'cn !leal Dos ooerator e r:.r will not to pe for% 's sa e

result in loss of 1

function.

The ltlons iCS3 2

~ O CRT:" RTA Current licensin" criteria ron iCS3 0 are:

when e"=her -r.

) ary

<<Qtor!at c 0Pen ng or the vc ves coolant s~ ster pressure exceeds (to be specif'ed in the Tecnnic or

" safety injection signal is pr ima v cool nt sys tee pressure tion signals should be provided OPc 2tor

~

a prese a'pec

'"esent anc sa to

" 'the lected value f1C2'tlonS

) l jot'h

'1

% %cl

)J 3 v21ve V'sual 'ndication in the control rooa of the open or closed status of th)e va ve.

)"-n 2Q C lb 1 e

cbove, tha" when the va ls lve nO ). l)) the v sual cia 2y

~~ cp cc ).Qc-s en so on the valve pos 1

". 'n incepende..t of item 2.

Utilization auto~a 'al Oi 2

(=v in jec n y e ride) bypass ea' e 'that si+nal to ezove

-zy be p divided to c'osea

cr short pe='c='s o

su coo '2"ilt s" s ap- '

at prcvi 'ons o='ne Te i ~..1r C o'at'on v-to be title when thie reactor re {'n zccorcznce with Spec'ica.:cns) a Current licens'ng c" ter'a

=rem

~ CSD 18 zre:

"211U e

IIU'n l.he el ac tr1 system agz oth e z'Gnc s

.'n born he "=ail to =Unction" sense and desirable

=unct'on" se..se oz co..ponents in co

'po Leii s sinoula De cons laa ea 1

Qe. s gn a sin~le failure even though the vz've or 1 u' svstam cc pcnant zlzy no-be cal led Upon t'on 'n a "-ive.. Szzety operat cn" 1 sea e..ce.

ause Unaesl-e ec....

~ '

'la svs em co'"'Dcnen 1 css 0

tho svs am 52 r

eel el 1 r lieu 0'z ae Sign cceptable, to c'scon..ect 2

s o" the va've or ot!lar he plant.echniczl Speci-

~ ist oz all electrical lv eau'red pos 'ticns 0~

reau're=ent zor removal ed

'-n oraer to satis=y on i LLnere i" 's ceterm'e'ha cz system co ponen" czn c

mot on oz" a v"lve or 0=her ana titlS <<0'tloil reSults 11 ety funct on,

~ t s

-ccapt cilznges th2t

"-1 so mizy be.

2

-ower to the e ec tr.'

sy s t zl id svs e

comDon -nt.

zlc"- tlons S.ici'lc

.nc luce 2

oDe'ated valves ana "ne r these

valves, to whic." the o= electr.'c powe" is aopi.i the sinsle zzilure cr'ter'.

-lac trically-oDerated valves that zre clzssi<i ea as 4

active valves

i. e, 2

e eaui d

0 DDe L or close. in various sa=e=y 5-stem operat'o..al seauen-ces bu-zre "znuall"-ccntrol'ed shoild be opera-tea zrcm the =-.zlr contro'oom.

Such va ves may rot be:ncluaaa among trose valves

'ro.". whic'o power is removed in or 'e to =eat the s"'n".'e

=a ure criterion Unless (a) electrical power can be re-stored to the valves

=rom the main cont ol.ocm, (0) v2'e DDe 5 i.ior. '

rot necessary

=or a" lezs t-an m'n

- s 'ol Owing occurrence cz t..e event re-oui ring such operatlon1 and (c) 1 ls demonstrated there is rezsonaD o assurance that a

neces-5 v oue 2 0 2c lo s wi1 1

'De De

. 0 ed w th i~ tile t,ime sho~~

to be zaac zte by the 2".ialysis.

~he p'iz..t ecnnlc 1 Speclzlcc tlors sho'u a lilclllQe 2 list Oz t."Le ecu'ea pcs 1 l.oils oz manna cont 0

c cz '-one at Q va '

s ard hould r

ide..t'=y hose v-'ves to wh'ch the rea re. ant 'or ratlovzl c= e'ectric power 's applied

'";. or e

Sat S 1V i iie sinuose

~-1' C - ~ation

ngie

. p 2:ovaLO 2 2Ct 1

Cc,.it Po c10n rc c..lC 'e

'DOS t 1011 eet he, s'ng e

\\

1GTl '

Sa~. 'd Dy 1U 2

c 1 pQwer cv otl va ves described es should have redun-iDa 1<i cori 0 1 roGTS ste~ should,

1tsaiz, cat1Gil 1.1 nd caticn I

~ ~

Sv iiu 2 criterion

~

D ~

r1cailv-0 ec d'rect le nfl

"-.Se

= 'C t Qotn ValveS OP2rc.t cev'ce (e.g.)

a -0 ope ared valve) by an elec rical d vp 'e whose a1r su ca 20 12.!0) c vc.lve tor 002 a tnose v

av'ce (e.

ppiy s c

nc luces 1

c<no~ d-irectly d

valves,"

p I.e pe-lv

< co valve G"

a S01 c ives all O~.

Ql ooera tea nd P 'ne<<ai 2

ed by an electri-

"=.0 D: CUSS:ON

.-'.AD L.U. LUP "ON D ~

~

Discussicn.

he G'nna lant uses

"- o -cc

. U!ators.

eacn o=

wh chi has a =otor-ope=ated 1solaticn valve.

"hase valves are MOU SLI and

~OU OOE1 ~

-'.:5.:-ach va va has

- s)ngle position indication

'T. the control

. ha Ginna, echnical Snec i-.'a r iGT-.s e oui 2 th t.

Dr1Qr to 2 c Gr c 1tlca 1 sty

) the Accu U 1 to isol at1011 Ua ives

( 1 s

ec

/

by valve num'Dar) =ust be opened and that power to the valve t.otors rust p-t a1n1ng be r o'ed;

=ethod o= res!oval (open brea):er>

rac. -out breaker, o

d's-

.6 COrcneCt

-,.OtOr PGWer CableS) iS nOt SPeCi-ied Ba OVal O-Valve one accu. u ctor to

'De out-o=.-serv ce or no Taore t'nan one hcur Quring po>>er operation w =hout going to hot shutdom; a

sa ety njection signal Roto pOW2 CoeS Ot C1saol 2 VP 'Ve posl t) 0! lnC1Cc't10 1>

Wn'il iS po' eC1

~l rrora a separate 1" 5 U DC bus..he ec'rn'cai Spec'=ic t'ons 2'low 1s p ov1cea to pUQQ>a t1ca

'y 0'pe 1 the vP ives U.de suc.

c cut!st n 2$

~

Teel il1Cc. 1 SPeC 1 rica'0nS 00 inoi.

P 1 }O'W Dl-i 'Pera't 10'n u,lL2SS both DC buses are ope" "'ral.

accutlulato 1.50 iation va ve pc;.

2nd ccn=rol systara design meets tha

". CoulreQetlt 0:

LCSB '8, part 2.

Lhe ign does not> ho;ever, aeec th raqLire ent o=

CSZ>

L8) part 4) whicn..ust be CG~Dliea with wnen reU!QvaL 0 va!ve tlo<<or Gower 1s us ed

=o ~eat the single =ailu=e criterio<<. orlly cne pos.'cion 1no lca t10il p va 've s

-'=- 'ble

.'.. tne "o""-ol roon) a scheme which

's, 'nhe"ently

Glnna ace U~U 1 ator isolat 'on valve po'er ana control s.:ste-.,

design Qo S not e~eet cnr ent lice..s n<

c l la ln that co.l'l root1 valve position i;. 'cat'on

's neither reQUnoant nor sing e-fail re free 5.0 2"-:"3"-HCiS 1 I

~~Zt S4 Power G

andarQ

>7Q "Cr~ terl enerating Stat ons.

for wrote tion Syste s for NUclear 2

~

Branch V-ives in the,

"-CCS ccU;QU1 ator echnical 'Pos'tion TCSZ Re QUl e'Ken t s ines."

oz

>oto -Overatec 3 e 3'nc'1

~~',1U.o Va1 ves

'c."1'n 1c2 1 Cr te' on os't'on i to h>anU2

~

CS3 l8.

An'Dl c t o. Of the S nile Con-rolled

-lee tricai lv-Oneratec G'nna Braving 33013

> Revlsloa ca Q 7~.

Ginna Dr ving 10905-242 h'o ?evision dateQ 2-25-75 5.

"Technical S ecificati ons for R.

=". Ginna NUC lear Po"er 'Plant 1

1Q t:n t Qo.

1>

"Ch nge Q

March 1>

1Q11 '1<<-az.ash

).3,1.1.v, 7

~

Gi'. 1" D a@i lg 10905 287

>.'io:,evis lo'.1

> Qatar 8"22-78

SEP TECHNICAL EVALUATION TOPXC VIII-3.B DC POWER SYSTEM BUS VOLTAGE MONITORING AND ANNUNCIATION R. E. GINNA

1.0 INTRODUCTION

The objective of this review is to determine if the DC power sys-tem bus voltage monitoring and annunciation are in compliance with current licensing criteria for Class IE DC power systems.

The specific requirements for DC power system monitoring derive from the general requirements embodied in Sections 5.3.2(4), 5.3.3(5),

and 5.3 '(5) of IEEE Standard 308"1974 and in Regulatory Guide 1.47 In summary, these general requirements simply state that the DC system (batteries, distribution systems, and chargers) shall be monitored to the extent that it is shown to be ready to perform its intended function.

2.0 CRITERXA As a minmum, the following indications and alarms of tne Class IE DC power system(s) status shall be provided in the control room: 3 e

Bat tery current (ammeter-charge/discharge) e Battery charger output current (ammeter)

~

DC bus voltage (voltmeter)

~

Battery, charger output voltage (voltmeter) o Bat tery high discharge rate alarm o

DC bus undervoltage and overvoltage alarm

~

DC. bus ground alarm (for ungrounded system) e Battery breaker(s) or fuse(s) open alarm

~

~

~

Battery charger output breaker(s) or fuse(s) open alarm a

Battery charger trouble alarm (one alarm for a number of abnormal conditions which are usually indicated locally).

3.0 DISCUSSION AND EVALUATION

3. l Discussion.

Tvo l25 V batteries, four battery chargers, and two DC buses comp".ise the Ginna Class IE pover systems.

Control room indication consists of DC bus voltmete'rs, bus ground alarms, charger failure alarms, battery undervoltage

alarms, an "Annunciator Normal Supply Off" alarm, and "Diesel Generator lA Panel" and "Diesel Gener-ator lB" alarms. 4 3.2 Evauation.

The Ginna control room has no indication of 'oat-tery current, charger output current, charger output voltage, battery high discharge ratep bus under/overvoltage, battery breaker/fuse

status, or charger oreaker/fuse status.

Therefore, the Ginna DC pover systems monitoring is not in compliance vith current licensing criteria.

4.0 SUM~Y Of ll parameters currently required to be indicated or alarmed in the control room, only three are monitired in the Ginna control room.

Therefore, the Ginna DC pover systems are not monitored in compliance with current licensing criteria.

5.0 REPRENCES l.

I'""EE Standard 308-5974, "Standard Criteria for Class I: Povez Systems for Nuclear Power Generating Stations."

2.

Regulatory Guide l.'74, "Bypassed and Inoperable Status Indi" cation for Nuclear Power Plant Safety Systems."

3.

NRC Memorandum,

'PSB (Rosa) to SFPB (Crutchfield),

"DC System Monitoring and Annunciation," dated October 16, 1979

~

4.

Letter, Roches er Gas and Electric Corporation

('wnite) to NRR (Ziemann),

"SEP Topic VIII-3.B, DC Power System Bus Voltage Monitoring and Annunciation," dated July 6, 1979.

WESTINGHOUSE PRESSURIZED WATER.REACTOR LICENSEES Docket Ho. 50-334 Beaver Valley Unit 1

Docket No. 50-315 D.

C.

Cook Unit 1

Docket Ho. 50-316 D.

C.

Cook Unit 2 Docket No. 50-348 Farley Docket No. 50-3 Indian Point Unit 1

Docket No. 50-247 indian Point Unit 2 Docket Ho. 50-286 Indian Point Unit 3 Docket Ho. 50-305 Kewaunee Docket Ho.

50-338 North Anna 1

Docke. Ho.

50-266 Point Beach Unit 1

Docket No. 50-301 Point Beach Unit 2 Docket Ho. 50-282 Prairie island Unit 1

Docket No. 50-306 Prairie Island Unit 2 Docke'i Ho.

50-261 H.

B. Robinson Unit 2 Docket No. 50-272 Sa 1 em Unit 1

Docket No. 50-280 Surry Unit 1

Docket No. 50-281 Surry Unit 2 Docket No. 50-344 Trojan Docket Ho. 50-250 Turkey Point Unit 3 Docket No. 50-251 Turkey Point Unit 4 Docket No. 50-295 Zion Unit 1

Docket No. 50-304 Zion Unit 2 Doc et No. 50-244 Ginna Docket No. 50-213 Haddam Neck Docket No. 50-206 San Onofre Unit 1

Docket No. 50-29 Yankee-Rowe