ML18046A166

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Forwards Evaluation of SEP Topics V-11.A, Electrical Instrumentation & Control Features for Isolation of High & Low Pressure Sys, V-11.B, RHR Interlock Requirements & VIII-2, Onsite Emergency Power Sys
ML18046A166
Person / Time
Site: Palisades Entergy icon.png
Issue date: 09/15/1980
From: Crutchfield D
Office of Nuclear Reactor Regulation
To: Hoffman D
CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
References
TASK-05-11.A, TASK-05-11.B, TASK-08-02, TASK-5-11.A, TASK-5-11.B, TASK-8-2, TASK-RR NUDOCS 8011190002
Download: ML18046A166 (34)
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Text

. DISTRIBUTION Dockaet

ACRS (16)

Docket No~ 50... 255

NRR Reading DEisenhut RPurple

  • TNovak RTedesco.

Glainas

. _JHe ]~emes, ~OD Mr. David P. Hoffman Nuclear licensing Administrator

. Consumers Power Company 212 L~est Michigan Avenue Jacl<?Oh, Michigan

  • 49201 *

Dear Mr. Hoffman:

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DCrutchfi e 1 d"-

TWambach

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'0 Enclosed is a cqpy of our evaluatfon of Systeinatic Eval1,iation Program *Topic V-11.A, Electrical Instrumentation and Control Fea~ures. for ~solat1or:i of High, and lo~1 Pressure Systems; v... *11.B, RHR Interlock Requirements; VIII-.2, Onsite.

Emergnecy Power Systems.

This assessment*compares.your.fatility, as described in Docket No. 50-255, with* th~ criteria currently used by the regulatory staff for Ncensing new* facilities.

  • Please. inform us. if your as~b.uilt facility differs from the lfcen$ing basis assuf(led in-our -asses~merit withi.n 45 days of receipt._;*.

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  • of this* letter.

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This eyaluat1on will. be. a* basic input to 'the integrated safety assessment for your facility unless *you identify*changes needed to reflect the as-built condi:-:

tions at your facility. This.topic:assessment may b.e revised in the fut_ure -*.

if your facility.de~ign is changed or.if NRC criteria relating'to this t_opii:

is modif1ed-"before the inte9rated asses~ment is *completed.;

With* regard to Topics V-1'1.A and V-H.B, you are aware *that there-15 an _ongotng*

generic act1v1ty for WASH 140.0 Event V intersystern LOCA's as described in our.

letter to.< all LWR licensees on February 23, 1980. Resolution of this Event V* con--

cern is expec;ted 1n the near future.

Please also.be advised that the broader issue of intersystem lOCA's is presently under -staff review and that direction regarding**

  • this concern may be expected in the near'fi.lture.

The effect of the resolution of these activities *along with any corrrnents you may have on this evaluation wilJ be included in the final report on these topics prior to the.integr.ated assessment*

s*o111 0002 Sincerely, "

Dennis M. Crutchfield, Chief Operating Reactors Branch #5

  • Division of *Licensing J~t~.<.1~~-~~-----******************* *.. P;Sjz9~? __ *.P~...

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NftC FORM 318 (9*76) NRCM 0240

  • u.s-. GOVERNMENT PRINTING OFFICE: 1979-289-369.

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IT.ED ST.AT.ES NUCLEAR RrEGULATORY COMMISSION w~isHINGTON, D. c. 20555 J

I Mr. David P./Hoffman Nuclear L i/~nsing Administrator ConsumervP0111er Company September 15, 1980

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212 \\tJe7fMichigan Avenue.

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Jacjs5on, Michigan 49201

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Dear Mr. Hoffman:

I RE:

SEP TOPICS V-11.A ELECTRICAL, INSTRUMENT!i.TION AND CONTROL FEATURES

.FOR ISOLATION OF ~UGH AND LO\\~; PRESSURE SYSTEMS; V-11.B RHR INTERLOCK REQUIREMENTS; VIII-2 ON-SITE EMERGENCY POWER SY STE~~S (Palisades)

Enclosed i~ a copy of our evaluation of Systematic Evaluation Program Topic V-11.A, Electrical Instrumentation and Control Features for Isolatinn of High and Low Pressure Systems; V-11.B, RHR Interlock Requirements; VIII-2, Onsite Emergnecy Power Systems.

This assessment compares your facility, as described in Docket No. 50-255, with the criteria currently used by the regulatory.staff for licensing ne\\'J faciliti:es.

Please inform us if your as-built facility differs from the licensing basis assumed in our 2ssessment within 45 days of.receipt of this letter.

This evaluation \\'!ill be a.basic inpu": to the integrated safety assessment for your facility unless you identify ch2nges needed to reflect the 2s-built condi-tions at your facility:. This topic assessment may be revised in the future if your facility design is changed or if NRC criteti*~ relating to this topic is modified before the integrated assessment is completed.

~ith regard to Topics V-11.A and V-11.B, y6u are aware that there is an ongoing generic activity for WASH 1400 Event V intersystem LOCA's as described in our letter to all L\\4R licensees on February 23, 1980.

Resolution of this Event V con-cern is expected in the near future.

Please also be advised that the broader issue of intersystem LOCA's i:? presently under staff revie1*1 and that direction regarding (

tb.is conc2rn may be expected ir1 the near future. *The effect of the resolution of

  • these activities along v:ith any co:::-::ents you may have on this evaluation will be included in the final report on these topics prior to the integrated assess~ent.

~....

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Dennis M. Crutchfield, Ope:--at i ng Reactors 5ra 1ch Division of Licensing

I Mr. *David P. Hoffman cc M. I. Mi l l er, Esq u i re Isham, Liricoln & Beale Suite 4200 O~~ First National Plaz~

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Chi)f:ago, Illinois 60670

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Mr.' Paul A. P~rry, Secretary.*

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Consumers Power Company. ~***;..>,_.,,

212 West ~~*cfil*gai'd"\\-vt:r;'l:r~'****

JacksQq;.-~ichigan 49201.

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~):!'~~ L. Bacon; Esquire Ir/Consumers Power Company w

,f 212 West Michigan Avenue

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Jackson, Michigan 49201 Myron M~ Cherry, Esquire Suite 4501 One IBM Plaza Chicago, Illinois 60611 Kalamazoo Public Library 315 South Rose Street Kalamazoo, Michigan 49006 Joseph Gallo, Esquire Isham, Lincoln & Beale 112J Connecticut Avenue Roo;;i 325 r/ashington, D. c.

20036 Peter W. Steketee, Esquire 505 Peoples Building Grand Rapids, Michigan 49503 Chcrlevoix Public Library 107 Clinton Street Charlevoix, Michigan 49720 Ms. Mary P. Sinclair Great Lakes Energy Alliance 571 l Summe~set Drive V.idland, Michigan 48640 Res1dent Inspector Bi<; Rock Point Plant c/o U.S. NRC RR

~3, Box 600 Che:rlevoix, Michigan 49720 '.

2 -

September 15, 1980 Charles Bechhoefer, Esq., Chairman Atomic Safety and Licensing Board Panel U. S. Nuclear Regulatory Commission Washington, D. c.

20555 Dr. George C. Anderson Department of Oceanography University of Washington Seattle, Washington 98195 Or. M. Stanley Livingston 1005 Calle Largo Santa Fe, New Mexico 87501 Sheldon, Harmon and Weiss 1725 I Street, N.

  • w.

Suite 506 Washington, D.. c.

20006 Mr. John O'Neill, II Route 2, Box 44 Maple City, Michigan* 49664 Herbert Grossman, Esq., Chairman Atomic Safety and Licensing Board

u. S. Nuclear Regulatory Commission Washington, D. c.

20555 Dr. Oscar H. Paris Atomic Safety and Licensing Board U.S. Nuclear Regulatory CofT1llission W as h i n gt on, D

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  • 2 0 5 5 5 Mr. Frederick J. Shon Atcimic Safety and Licensing Board
u. S. Nuclear Regulatory Corm1ission Washington, o. c.

20555 Big Rock Point Nuclear Power Plant ATTN:

Mr. C. J. Hartman Plant Superintendent Charlevoix, Michigan 49720

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. :. > * * * (H:r:;*~e *,of the G:bvernor (2).:; *

  • Rcom.1 *Capitol Building;::~.-*

Laris.ing.~ Michi_gan. 4~913.";;.

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Di re~tor, T~chntca]. Asses*Jm~nf i'..

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(AW-459)

U. S. Enyirorirnental Prot~cti-on * *

  • -~Agency.

Crystai Mall* #2' A~lington, Vifginia 20450 *. '

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lJ; s. Env*ironmental Protect.ion r "enc'y'.

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Federal Activities Branch~*

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El S COORDINATOR 230 Sout~ D~arborh Sttee~

.U.ia:c:go,'111in.6is 60504 i\\2si0ent Insp*ector c/o U. S. NRC P. O. Sox 8.7

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)o;.;.* n._ave!'),.;1c 19an F.
H sades Pl a nt

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Mr. J. G. Lewis*

Pl a'nt Manager Covert, Michiga~ 49043

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  • Toh~ship.Supervisor Co**ert Township r.-:.:.::e 1, Box 10

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V~~ Sure~ County, Michigen.

Christa-Maria

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  • S ca n lo n, E s q u ) re 22~~ ;auline.Boulevard *
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SEP TECHNICAL EVALUATION REPORT ELECTRICAL, INSTRUMENTATION, AND CONTROL FEATURES FOR ISOLATION.OF HIGH AND LOW PRESSURE SYSTEMS PALISADES NUCLEAR STATION Consumers Power Company Docket No.

50-~5~

Januaryl980 1373F 6-19-80

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CONTENTS

1.0 INTRODUCTION

2. 0 CRITERIA 3.0 4.0 5.0
2. l 2 *. 2 2.3

~e~idual Heat Removal (RHR) System * * *

  • E~ergency Core Cooling System O~~er Systems **.******

DISCUSSION AND EVALUATION 3.1 3.2 3.3 Residual Heat Removal (RHR) System

  • Safety Injection System ******* -* **

Chemical and Volume Control System *

SUMMARY

REFERENCES ii l

l 1

2 2

3 3

3 4

5 6

SEP TECHNICAL EVALUATION REPORT El,.ECTRICAL, INSTRUMENTATION, AND CONTROL FEATURES FOR L

. ISOLATION OF HIGH AND LOW PRESSURE SYSTEMS PALISADES NUCLEAR STATION l

  • 0 INTROD'!JCT ION The p~rcpose of this review is to determine if t:he electrical, instrum¢ntation, and control (EI&C) features used to isolate systems with a lower pressure rating t~an the reactor coolant primary system

~

are in comP!iance with current licensing requirements as outlined in SEP Topic VillA. Current guidance for isolationi of high and low pres-sure sxstems.j~ contained in Branch Technical P~sition (BTP) EICSB-3,

\\. :. *'

BTP RSB~S~l~ an~ the Standard Review Plant (SRP), Section 6.3.

z.o CRITERIA 2.1 R~sid~al Heat Removal (RHR) Systems.

Isolation requirements for RHR systems cont;a.ined in BTP RSB-5-1 are:

(1)

The suction side must be provide~ with the fol10wing i'~olatfon feature~:

(a,) ' TWo power-operated valves in series with posi-

. tion indicated. in t.he control* room.

(b). *The valves must have independent and diverse in:t~rlocks *to prevent opening if t:he rea1=tor

  • ~oolant syst~m (RCS) presstire is above the
  • desigQ pressure of the RHR system.

~c) The valves must have independent'and diverse interlocks to en~ure at least on~ valve closes upon an increase in RCS pressur~: above the

  • design p~~ssu~e of the RHR ~yste~.

(it The-di-scliatge- -side--mi.fst be pt6vided-wi-th one of--the --

following features:

(a)

The valves, position indicators, and'interlocks described in (l)(a) through (l)(c) above.

  • '(~) One or more check valves in series 'l!i'ith a normally-closed power-operate4 valve which has l

if position indicated in the conll room.

If this valve is used for an Emergency Core Cooling.System (ECCS) function, the valve must open upon receipt of a safety injection signal (SIS) w~en RCS pressure has decreased below RHR system design pressUTe.

(c)

Three check valves in series.

(d)

Two check valves in series, provided that both may be periodically checked for leak tightness and are checked at least annually.

2.2 Emergency Core Coo~ing System.

Isolation requirements for

~CCS are contained in SRP 6.3.

Isolation of ECCS to prevent overpres-surization must meet ope of the following fe~tures:

(l) 0n*e or more check valves in series with a normally-closed motor-operated valve (MOV) ~hich is to be opened upon receipt of a SIS when RCS pressure is l~ss than the ECCS design pressure (2) 111ree check valves. in series (3)

Two check valves in series, provided that both may be periodically checked for leak tightness and are c~ecked at least annually.

2.3 Other Systems.

All other low pressure.systems interfacing with the RCS must meet the following isolation requirements from BTP E;ICSB-;r:

(l)

At least two valves in series must be provided to jsolate the system when RCS pressure is above the

~ystem design pres~ure and valve position should be

~rovided in the control room (2)

  • ~or systems with two MOVs, each MOV should have
  • independent and diverse interiocks to prevent.

opening until RCS pressure is below the system design pressure and should automatical"ly close when

  • RCS pressure incr~ases above system de*sign pressure en "For systems with one check valve and a MO\\!, the MOV should be interlocked to prevent openitig if RCS pressure is above system design pressure and sho~ld automatically close w,henever RCS pressure exc.eeds system design pressure.

2

          • ---*/*

3.0 DISCUSS ION

  • EVALUATION There are three systems at Palisades Nuclear Station which have a

~irect interface ~ith tl1e RCS pr~ssure boundary and have a design pres-sure rating of all or part of the system which is less than that of the RCS.

Thes¢ systems are the Chemical and Volume Control System (CVCS),

the Safety Injection System (SIS), and the Residual Heat Removal (RHR) system.

3.1 R~sidual Heat Removal System.

Evaluation of the isolation functions qf the RHR system *is performed in Topic V-11.B.

Refer to that Topic for the evaluation.

3.2 Safety Injection System.

One SIS subsystem consists of four pressurized accumulators with each accumulator isolated from the RCS by a pair of check valves.

There arc connections upstream of each check valve that can allow them ~o be tested.

A normall~-open motor-operated isolation val~e upstream of the check.valves for each accumulator has position indication in the control room.

Each MOV is locked* open during reactor operation.

The second SIS subsystem consists of two loops, each supplied by a high pressure safety injection pump.

Each loop discharges through a common header to each of th:e four RCS cold legs.

_Isolation is provided by two*check valves in series for each safety injection loop.

The

~heck valves in each high p~essure SIS loop are not testable since

~ --

there are no loc-ations where leakage could be determined from the out-board va1ve (farthest from the RCS).

A motor-opera~ed isolation valve with position indi~ation in the control room is provided in each branch of ~he SIS loops.

These valves open upon receipt of i safety injection 3

. *********---... -------.. -.... -*------~--*-**-~~--

  • .l* k

. eh signal, but have~ inter oc s preventing opening _w ~n RCS pressure is above SIS design pressure.

The third SIS subsystem use~ the low pressure -inj~ci:ipl) pumps to provide low pressure water from the refueling water storage tank to the reactor vessel through the same lines used for high pressure injection.

Isolation i~ provided by a MOV in series with two ch~ck valves in each of the branches.

The MOVs open upon receipt of a safety injection signal but ~ave no interlocks to prevent opening when RCS pressure is above SIS pesign pressure.

The outboard (farthest from ~CS) check valves ar~ not testable. since there are no locations to determine leakage.

The SIS is not in compliance with the current lic~qsin~ require-ments of SRP 6.3 since the 12 outboard check valves ip the high and low pressure safety injection lines to the RCS are not testable, and the MOVs in the.high and low pressure injection lines have no iriterlocks to prevent opening when RCS pressure exceeds system design pressure.

3.3 Chemical and -Volume Control System.

The CVCS takes "ater from the RCS and.passes it through a regenerative heat ~xchanger, an orifice to redu~e its press~re,. and a nonregenerative ~eat exchanger before sending it to the filte~ing and cleanup portions of t~e system.

After filte~ing and cleanup; the water may be return~d to t~e RGS by th~ use of the charging pumps, which_ increa~e t~e water pressuI)e and pas~ it through th~ regenerative heat exchanger~to either RCS loop lA, RCS loop 2A~ or to the pressurizer auxiliaiy spray line.

The CV~S suction line isolation is provided by a_solenoid air-op~rated ~alve in series with three parallel solenoid-controlled air-operated valves.

Each of t'hese valves is operated from the control room and has valve "posi tioif indicated-_- - Norie--o'f' the va'fves have.in-ter-------

locks to_prevent opening or to automatically close jf the pressure exceeds the design rating of the low pressure portions of the. system.

4

'..*. ****--*-- -* -...... -* ***-***-. -~*-** -....

The CVCS ~*arge line isolation is provid.by a common dis-

, char*e line check valve, a branch check valve in each of the three branches.dO\\o.'TIStre~m of the common check valve, and s check valve at.

each pump discharge.

There is no position indication available in the coptrol room for the check valves.

There are solenoid-controlled air-operated i~olation valves in each discharge line ~ranch which have position ihdication in the control room, but these yalves have no inter-locks tb ptevent system overpressurization.

The C\\ICS is not in compliance with current licensing requirements for isqlat~ori of high and lo.w pressure systems contained in BTP EICSB-3 since the suction and discharge line solenoid-controlled air-operated valves have no interlocks t°- prevent system overpressurization, and the discharge line check valves ;have no position indication available in the controi room.

  • 4.'0

SUMMARY

The P,alisades Nuclear Station has three systems with a lower design pressure r'tirig than the RCS, which are directly connected to the RCS.

The *eves,* SIS, and RHR system do not meet current licensing requirements for isolation of high and Jow pressure systems ~s ~p,cified below.

  • -..,..~--**********-

'"~*******

(1)

The eves solenoid-controlled air-operated valves have no pre~sure-related interl~cks, and the dis-bharge line check.valves have no posiiion indica-tion availabl~ in;the control room as required by BTP EICSB-3 (2)

The SIS outboard (farthest from RCS) check valves in the lines going to the RCS from the high and low pr~ssure injection pumps are not testable, and the motor-operated isolation valves in these lines have no pressure-rel~t~d interlocks as required by SRP 6.3

--(3)

The RHR system isolation deviations from current criteri~ are d~scribed in Topic V-11.B.

  • 5

5.0 REFERENCES

1.

NUREG-075/087, Branch Technical Positions EICSB-3, RSB-5-1; St~ndard Review Plan 6.3.

2.

NRC Memo (Knox to Crutchfield) dated 31 Mar 1980, SEP Review Topi~ V-llB: RHR Interlock Requirements~-Palisades.

3.

Updated Final Safety Analysis Report, Palisades Plant.

4.

Pali~ades Drawings M-201, *202, -203, -204, and -219.

6

I.

INTRODUCTION ENCLOSURE EI&C EVALUATION REPORT OF SEP TOPIC V-llB RHR INTERLOCK REQUIREMENTS PALISADES Docket # 50-255 The purpose of this evaluation is to ascertain the degree to which the Palisades design complies with review criteria that deal with the interface between the high_ pressure primary coolant system and the low pressure shutdown cooling system {RHR system).

Review criteria,. review guidelines, and review areas, to be covered in this evaluation, are defined in section II and IV.

Review areas that are not covered, but are related and essentia 1 to the completion of this topic, are covered by other SEP topics defined in Section III. SEP topics are defined in the Report on the Systematic Evaluation of Operating Facilities, dated November 25, 1977.

This topic evaluation.report is limited to identification of compliance to review criteria, identification of deviations *from review criteria, and identification of any viable corrective measures for each deviation r;

identified.

An integrate'd system assessment of the identified deviation's significance :and recorrunendations as to the imposition of

  • ** cc:rr*rective measures are outside the scope of th-is report and.will..

be the subject of a subsequent report.

The implementation of corrective design measures will be evaluated and reported by others.

- 2 ~

II. REVIEW CRITERIA Review criteria that govern the subject safety topic are identified. in section

.7.6 (part II) of NRC standard review plan~

III.

RELATED SAFETY TOPICS AND INTERFACES Related Safety Topics The follqwfng listed review areas are not covered in this report, but are related and essential to the completion of this topic. These review areas are covered by.other SEP topics as indicated below:

1.

The capability of the isolation components (motor operator, interlock, sensor fof interlock, and associated cables} to function during and after.

design basis events, such as earthquakes and anticipated operational occurrences, are addressed by SEP ~opics III-6 and III-12;

2.

the capacity and reliability of the.isolation components to perfonn their intended safety function on demand is* add.ressed by SEP to pi cs XVII, XVI, and VI-lOA;

3. Capability of the containment heat removal arid pressure control system to.maintain a controJled environment for *safety related instrumentation and electrical equipment located inside containment is covered by topic VI-3;
4.

Technical specification requirements are covered by topic XVI;

5.

System testing and s:urveil lance requirements are covered by topic VI-1 OA~

-~..,. *-*. -**.... *,** -*~-.. - *~*-.

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6. The adequacy of the quality assurance program is covered by topic XVII;
7. The ECCS actuation system is covered by topic VI-7A3 and VII-2; and
8. Protection of the Class lE components from design bQ~is ev~nts such as flooding, missiles, pipe breaks, and fires are covered by topics II-3, III-4, III-5, and IX-6.
  • Interfaces The following SEP topics interface with and are dependent on this topic's information for their completion:

Topic II-3.B, II-4, III-1, III-3.B, III~4, III-5, III-6, III-12, V~lO.B, VI-7.C, Vl-10, VII-3, XVI, and XVII.*

IV.

REV! EW GU IDE LINES

  • The purpose of this evaluation is to ascertain the degree to which the Palisades design complies with review criteria that deal with the interface between the high pr:-essur_e pr'imary cool ant system and the low pressure* shut-down cooling system (RHR system). The review area tQ be covered in this report includes the evaluation of electrical isolation components (motor-operated valve controls, interlocks, sensors for'interlocks, position indicators, power sources, and associated cables). The review establishes 1) that the sensors for the interlocks are suitably independent and diverse, 2) that a trip signal closes the motor-operated va*l-ves when the pressure' iS too high, 3) that the 'iriter10-cks prevent the motor-operated ~alves from op~ning u~less the high pressure system pressure is below the low pressure system's design pressure, 4) that the valves a re powered from redundant power sources, 5) that the AC and DC power source for

~~

-.. each valve's associated interlock. controls. position indic~tiQn, and motor are powered from the same Class I division, and 6) that the associated fedundant control power, instrument. and interlock cables ha~e adequate physical separation.

Review guidelines that address this topic's defined review area are delineated in the following sections of NRC Standard Review Plan:

Sections 5.4.7 (Branch Technical Position RSB 5-1), 7,p (part III),

and Appendix 7A (Branch Technical Position ICSB-~).

Review guidel{nes that addr~ss ph~sical separation of power, instrumentation~

and control cables are defined in Regulatory Guide 1.75 and IqE standard C*

384.

These physical separation guidelines aJ?e included with this topic review.

Additional review guidelines for. cable separation are defined in section 9.5.1, Appendix A, positions D.1.(a).(2) and D,1.* (c} of the NRC Standard Review Plan.

Deviations from 9.5.1 position guidelines have been judged significant.

The NRC staf~ requires backfit tp correet the deviations for all operating plants.

Identification of deviations and implementation of corrective design measures are being done by others.

Refer to the Palisades fire hazards anal~s1s, the NRC Fire Profection Safe~y Evaluation Report, and any supplements, thereto for i dent i fi cat ion of compliance to and deviations from 9.5.1 guidelines.

V.

EVALUATION The Palisades shutdown cooling system is a low pressure residual heat removal system located outside of containment.

The system. is isolated from the high pressure system by two motor-operated valves onrthe suction side and by two check valves in series with a motor operated valve on the qi~charge side of the system.

. A simplified diagram of the shutdown cooling system is presented in figure

1.

The diagram is based on Palisades piping and in~trument diagrams M-201, M-203, and M-204.

The following evaluation is based on this simplified diagram and information presented in the Palisades FSAR.

Suction Side Isolation To isolate the suction side of the low pressure sy~tem from the high pressure

~ystem, two motor-operated valves (MO 0316 and MO 0315) are provided in series.

Valve position is indicated in the control room.

Power is provided from re-dundant divisions.

Failure of the power supply will not cause any valve to change position.

Two administratively controlled key locked hand switches are

provided in the control room., This meets the review guidelines defined in

. section IV of thi~ report.

The review gµidelines also require independent and diverse interlocks to prevent these valves from being opened unless the high pressure system pressure is below the low pres~ure system design pressure.* The Palisades design has two systems.of interlocks.

The first interlock is provided by a single pressure switch for both valves.

The pressure switch (PS 0103} senses the press1,1re in the primary system pressurizer and is normally open so that control power is disconnected when pressurizer pressure is greater than the design pressure of the.low pressure

  • system.

Control power is required to energize th~ main ~ontactor and thus open the valve.

The second interlock is provided by a torque switch on the valve itself.

The torque switch is normally closed but opens tq interrupt control power when the valve motor attempts to open against a pressure di fferentia 1 greater than the designpressure of the low pressure system.

The torque switch interlock lacks the required setpoint accuracy and is difficul~ if not impossible, to calibrate.

The Palisades design, therefore, has only the single pressµre switch interlock. This is a deviation from review guide-fines.

This'c:ieviation, could be corrected by the installation of an

~dditional pressure switch interlock that is independent and diverse from the existing pressure switch interlock.

Th~ review guidelines also require that both valves receive a signal to close automatically whenever primary system pressure exceeds the low pressure system's design pressure.

The Palisades design does not provide this automatic close signal and is, therefore, a d~viation from the review guidelines defined in section IV.

This deviation could be corrected by the inst~llation of the appropriate pressure. sensor and associated automatic close signal.

The Palisades design currently depends on administrative control to assure that the two isolati6n valves are closed. : The automatic close signal eliminates dependence on administrative controls to assure isolation.

The review guidelines also req,uire that the cables.(power, instrumentation, and control) associated with these redundant isolation valves be separated by a three-hour fire ratei'd barrier or eq'uivalent.

The Palisades cable' raceway design routes cables in a two. raceway system separ~ted by or:ily on~ f9ot ver_ti_~~l air sp~~e_

separation.

The one foot separation is a deviation from review guidelines defined in section IV.

The NRC staf~ has judged this ~eviation significant and has required corrective design measures. The implementation of corrective measures is being evaluated by others as part of our generic fire protection review.

Refer to Palisades fire hazards analysis,* the NRC Fire Protection Safety Evaluation Report, and any supplements thereto for identification of compliance to, deviation from, and corrective design measures required for this deviation.

Discharge Side Isolatidn To isolate the di.scharge side of the low pressure shutdown cooling system from the high pressure primary coolant system, the Palisades design provides two check valves in series with a normally closed motor-operated valve.

The position of the motor operated valve is indicated in the control room.

This meets review guidelines defined in section IV.

/

An interlock is currently required on the motor-operated valve to prevent its opening upon receipt of a safety injection signal when the high pressure system pressure is abbve the design pressure of the low pressure system. The Palisades design does n9t provide the subject interlock. This is a deviation from review

~uidelines defined in section IV.

This deviation. could be corrected by installa-tion of a pressure switch sensor ind interlock.

It is also currently required.that the motor-operated valve receive a signal to close automatically whenever the high pressure system pressure exceeds the low

---pressure system's desfgn* 1:>ressµre.* --The-PaliSades design *does-not*provide this --

automatic close signal.

This is a deviation from review guidelines defined in section IV.

This deviation could be corrected by the installation of the l

appropriate pressure switch sensor and associated automatic close signal.

The Palisades design currently depends on administrative control and the two check valves. to assure isolation.

The automatic close signal eliminates dependence on administrative control to assure isolation.

VI.

CONCLUSION The following listed items summarize deviations from review guidelines identified and described in section V.

1.

The Palisades design currently depends on a single interlock to provide isolation on the s~ction side of the low pressure system.

This is a deviation from review guidelines.

2.

The Palisades design currently depends on administrative control to assure that isolation valves are:closed. This is a deviation from review guidelines.

3.

When there is a safety inJection signal, the Palisades design depends on two check valves in series to assure isolation on the discharge side of the low pressure system.

This is a deviation from review guidelines.

4.. The Palisades design currently routes cable associated with the two suction side isolation va.lves in a two raceway system separated by only one foot separation~-. This one-foot-separation is a deviati.on from review guidelines.

Refer to the Palisades fire hazards analysis, the NRC Palisades Fire Protection Safety Evaluation Report, and any supplements thereto for identification of compliance to, deviation from, and corrective design measures required for this deviation.

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SEP TECHNICAL EVALUATION TOPIC VIII-2 DIESEL GENERATORS PALISADES Docket No. 50-255 June 1980 07 llF

, "".. *'!'*'"' *,*: M

  • '*'* *:,,,."-*-**::**~..,"*!'";"
  • !'".*" ** ;. 'J *
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June 25,. 1980

e e

CONTENTS l.O INTRODUCTION.*

l 2.0 CRITERIA

  • 2 2.1 Diesel Generator Loading...

2 2.2 Bypass of Protective Trips 2

2.3 Diesel Generator Testing 2

3.0 DISCUSSION ANP EVALUATION 6

3.1 Diesel Generator Loading 6

3.2 Bypass of Prote~tive Trips 6

3.3 Diesel Generator Testing 7

4.0

SUMMARY

9

5.0 REFERENCES

9 J

1.0 INTRODUCTION

SEP TECHNICAL EVALUATION TOPIC VIII-2 DIESEL.GENERATORS.

PALISADES The objective of the review is to determine if the onsite AC generator for the Palisades Nuclear Station has sufficient capacity and capability to supply the required automatic safety loads duri~g anticipated occurrences and/or in the event of postulated accidents after loss of offsite power.

The requirement that the onsite electric power supplies have capacity and capability to complete the required safety functions is contained in General Design Criterion 17.

Criterion III, "Design Control," of Appendix B, "Quality Assurance Criteria for Nuclear Power Plani.'and Fuel Reprocessing Plan!,

11 to.

10 CFR Part 50 includes a requirement that measures be provid.ed for veri-fying or checking the adequacy of design by design reviews, by the use of alterIJate or simplified calculational methods, or by the performance ~f a suitable testing program.

Regulatory Guides, IEEE Standa~ds, and Branch Technical Positions which provide a basis acceptable to the NRC staff for compliance with 'GDCl 7 and Criterion III include:

Regulatory Guide 1.9, "Selection of Diesel Generator Set Capacity for Standby Power Supplies;" Regulatory Guide 1.108,.

"Periodic Testing of Diesel Generators Used as Onsite Power Systems at Nuclear Power Plants"; IEE.E Standard 387-1977, "Criteria for Diesel Generator Units Applied as Standby Power Supplies for Nuclear Power Stations;" BTP ICSB2, "Die-sel-Generator Reliability Qualification Testing";

and -BTP I CS Bl 7' "Diesel Generator Protective Trip Circuit Bypa.sses. II Specifically, this review evaluates the loading of the diesel~

generator, bypasses of protective trips during accident conditions and periodic testing.

The SEP reviews for Topics III-1 and III-12 will evaluate the diesel_;,generator qualification~

1.

. 2.0.CRITERIA 2.1 Diesel Generator Loading.

Regulatory Guide 1.9, "Selection of Diesel-Generator Set Capacity for Standby Power Sup-plies," provides the basis acceptable to the NRC staff for loading diesel-generator units.

The following criterion is used in this report to determine compliance with current licensing requirements:

(1)

The automatically-connected loads on each diesel-generator unit should not exceed the 2000-hour rating.

(Loads must be conservatively estimated utilizing the nameplate r_atings of motors and transformers with motor efficiencies of 90% or less.

When available, actual measured *loads can be used.)

2.2 Bypass of Protective Trips.

Branch Technical Position (BTP)

ICSB 17, "Diesel-Generator'.Protective Trip.Circuit Bypasses," specifies that:

(1)

The design of standby diesel generator systems should retain only the engine overspeed and the generator differential trips and bypass all other trips under an accident condition (2) If other trips, in addition to the engine overspeed and generator differential, are retained for.accident con-ditions, an acceptable design should provide two or more independent' measurements of each of these trip.

parameters.

Trip logic should be such that diesel-g~nerator trip would require specific coincident logic.

2.~

Diesel Generatdr Testing.

Regulatory Guide 1.108, hperiodic.

Testing Of Diesel Generator Units Used as Onsite El 1ectrical Power Systems at Nuclear Power* Plants", 'states that:

I 2

(1)

Testing of diesel-generator units, at least once every 18 months, should:

  • I.

(a)

Demonstrate proper startup operation* by simulating (b).

loss of all ac voltage and demonstrate that the diesel gener~tor unit can startautomatically and attain the r~quired voltage and frequency within acceptable limits and time.

Demonstrate proper ~operation for design-accident-loading sequence to design-load requiremen~s and verify that voltage and frequency ar.e maintained within required limits.

(c)

Demonstrate.full-load-carrying capability for an interval of ;not less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, of which 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> should be at a load eq~ivalent to the continuous ~ating of the diesel generator and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at a load equivalent to.*the 2-hour rating of the diesel generator.

Verify that voltage and frequency r~quirements are maintained.

The test should also.verify that the cooling system func-tions within design limits.

(d)

Demonstrate.proper operation during diesel-generator l~ad shedding, including a test of the loss of the. largest single load and of complete loss of loap, and verify that the voltage require-ments are met and that the overspeed limits are not exceeded.

(e)

Demonstrate functional capability at full-load temperature conditions by rerunning the test phase outlined in (a) and (b), immediately following (c), above~.,*

)

(f) Demonstrate the ability to synchronize the.diesel generator unit with offsite power while the unit is connected to the emergency load, tran~f.~~ this load to the offsite power, isolate the diesel-generator unit, and restore it to standby status.

(g)

Demonstrate that the engine will perform properly if switching from one fuel-oil supply system to another is a part of the nornial *operating proced-ure to satisfy the 7-day_ storage requirement.

(h). Demonstrate' that the capability of the diesel-gerierator unit to supply emergency power within the required time is not impaired during periodic testing und-er (3), below.

(2)

Testing of redundant diesel-generator units during normal plant opera_tion should be performed indepen-dently (nonconcu'rrently) to minimize common failure modes resulting '.from undetected interdependences among diesel-generator units.

However, during reliability demonstration of diesel-generator units during plant preoperational testing and testing subsequent to any plant modification where diesel-generator unit ipter-.

dependence may have been affected or every 10 years (during a plant shutdown), whichever is the shorter, a test should be conducted in which redundant units are started simultaneously to help identify certain common failure modes undetected in single diesel-generator unit test_s.

(3) Periodic testing of di~sel~generator units during nor~

mal plant operation should:

(a)

Demonstrate proper startup and,verify that the required voltage and frequency*are automatically 4

attained within acceptable limits and time, This test should also verify that the components of the diesel-generator unit required for auttomat:ic startup are operable.

(b)

Demonstrate full-load-carrying capability (contin-uous rating)** for an interval of not less than o1'e hour.

The test should also ve~ify that the cooling system functions within design limits.*

This test could be accomplished by syn~hronizing the generator with the offsite power and assuming a load at the maximum practical rate.

(4)

The interval for *periodic testing under (3), above (on a per diesel-generator unit basis) should be no more than 31 days and.should depend on demonstrated perfor-mance.

If more than one failure has occurred in the last 100 tests (on aper nuclear unit basis), the test.

interval should be shortened in accordance with the following schedule:

(a) If the number,of failures in the last 100 valid tests is one or zero, the test interval should be not more than 31 days.

(b) If the number of failures in the last 100 valid tests is two, the test interval*should be not more than 14 days.

(c)_

If the number of failures in the last 100 valid tests is three, the test interval should be not more than 7 days.

(d)

If the number of failures in the last 100 valid tests is four or more, the test interval should be not more than 3 days.

5

3.O DISCUSSION AND EVALUATION 3.1 Diesel-Generator Loading Discussion.

The Palisades FSAR lists the worst-case loading sequence for the diesel ge~erators.

1 Diesel Generator 1-2 is slightly more heavily loaded than Diesel Generator l~l (2,714 Hp versus 2,640 Hp).

'The maximum step load change is 530 Hp* ( 18% of* capacity).

Evaluation.

Palisades Technical Specifications require veri-fication, by test, of diesel generator emergency load capability during each ~efueling outage. 2 Maximum lo~ds of the diesel gen-erators, at 90%

._,.motor efficiency, are 2,lS8 KW for DG 1-1 and 2,250 KW for DG 1-2.

The continuous rating for each diesel generator is 2,500 KW at *0.8 Power Factor.

Therefore, the t1ital diesel generator loads (at 88% and 90% of capacity for DG 1-1 and DG 1-2, respectively) are within the requirements of Regulatory Guide 1.9.

3.2 Bypass of Protective Trips Discussion.

On:May 16, 1977 and July 12, 1977, CPCo provided a list of protective trips which render the diesel generators 1 3,4 incapable of responding to an automatic emergency start signa

  • Further conversation with'* CPCo determined that the diesel generators have protective trips which are not bypassed in accident conditiqns on high differential, overspeed and bearing oil pressure.

None of these trips has redundant parameter sensors and coincident logic.

Evaluation.

The diesel generator bearing oil pressure pro-

- -- --. tecd.ve tripX is peither ?ypa_ssed durin_g _accident coqditions nor provided with multiple parameter sensors and coincident logic.

Therefore, the use*

of this diesel generator protective trip is not; in agreement_ with current NRC staff guidelines as listed in BTP ICSB 17~

6

3.3 Diesel Generator Testing Discussion.

Palisades Technical Specifications, paragraph 4.7.1, require diesel-generator test~ng as follows.:

(1)

Each diesel generator shall be manually started each month and demonstrated to be ready for loading within 10 seconds.

The signal initiated to'start the diesel shall be varied ~rom one test to another to verify all j starting circuit~ are operable.

The generator shall be synchronized from the control room; and loaded to 2400 + 100 kW.

(2)

A test shall be conducted during each refueling outage to demonstrate the overall automatic'operation of the emergency power system.

The test shall be initiated by a simulated simultaneous loss of normal and standby power sources and a si11Ulated SIS signal.

Proper operations shall be verified by bus load sheding and automatic starting of selected meters and equipment to establish that restoration with emergency power has been accomplished within 30 seconds.

(3)

Each diesel-generator shall be subjectE!d to an inspection, in accordance with procedures prepar~d in conjunction with the manufacturer's recommendations for' this class of standby service, at least once per 18 months during plant shutdown.

'lbe licensee sh al 1 utilize his best efforts to conduct additional.major diesel-generator inspections and overhauls quring shutdown per.iocis.

_Diese_~:-:ge_nerator._ electric loads shall not be increased beyond the continuous rating of 2500 kW.

7

(4)

The fuel transfer pumps shall be verified to be operable each month.

Evaluation.

Diesel-generator testing defined in the plant Technical Specifications address the criteria listed in paragraph 2. 3 to the following extent:

(l)

(a)

Acceptable (b) Acceptable (c)

Duration not specified (d)

Not addressed (e)

Not addressed

( f)

Not addressed (g.). Covered under monthly test (h)

Not addressed (2)

Not addressed (3)

(a)

Acceptable (b)

Duration not specified

- (4-) - Not addreued.

The Technical Specifications do n~t meet current licensing criteria. for. -..

diesel-generator testing.

Diesel-generator failure data will be extracted by NRC from Licensee Event Reports and will be considered in the final evaluation of testing adequacy.

8

4.0.

SUMMARY

Automatic diesel-generator loading is in compliance. with current 1 icensing criteria.

The bypass of diesel-generator* protective trips is not in agreement with current NRC staff guidelines.

Diesel-generator testing, as specified by plant Te.chriical Specifications, doe*s not meet current licensing criteria.

Th¢ review of qualification of the diesel-generators will be completed with SEP :Topics III-1, Sehmic Qualification, and III-12, Enviro~mental Qualification.

5. 0 REFERENCES
1.

Final Safety Analysis Report, updated throu~h Amendment 32, Sep-tember 4, 1975, page 8-20b.

2.

Technical Specifications for the.Palisades Plant, April 16, 1976, paragraph 4.7.1.

3.

Letter CPCo to NRC, dated May 16, 1977.

4.

Letter CPCo to NRC, dated July 12, 1977.

5.

'l'elephone conversation CPCo (J. Kuemin) and EG&G Idaho, I,nc.*

(F. Farmer), April 26, 1979.

6.

General Design Criterion 17, "Electric Power System," of Appendix A, "General Design Criteria of Nuclear Power Plants," to 10 CFR Part SO, "Domestic Licensing of Production and Utilization Facilities."

7.

General Design Criterion III, ~'Des.ign Co~tr~.l~" o_~ _Appendix B, "Qual-ity Assurance Criteria *for Nuclear Power Plants and Fuel Reprocessing Plants," to 10 CRF Part 50, "Domestic Licensing.. o.f. Production and.. _.....

Utiliza~ion Facilities."

)

9

.. 8*. * "Standard Criteria for Class IE Power Systems and Nuclear Power Gener... ating Stations", IEEE Std. 308, 1974, paragraph 5.2.4.

9.

"Criteria for Diesel-Generator Units Applied as Standby Power Supplies for Nuclear Power Stations," IEEE.Std. 387, 1977.

10~

"Selection of Diesel Generator Set capacity for Standby Power Sup-plies", Regulatory Gµide 1.9.

11.

"Periodic Testing of D.iesel Generators Used as Onsite Power systems at Nuclear Pian ts, II Regulatory Guide 1.108.

  • 12.

"Diesel-Generator Reliability Qualification Testing," BTP ICSB2 (PSB).

13.

"Diesel-Generator Protective Trip Circuit Bypasses," BTP ICSB17 (PSB).

10

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