ML17258A867

Forwards Final Evaluation of SEP Topic V-II.A Re Isolation of High & Low Pressure Sys & SEP Topic VI-7.C.1 Re Independence of Redundant Onsite Power Sys

ML17258A867
Person / Time
Site:	Ginna
Issue date:	02/27/1981
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:	Maier J ROCHESTER GAS & ELECTRIC CORP.
References
TASK-05-02, TASK-06-07.C1, TASK-5-2, TASK-RR LSO5-81-02-060, LSO5-81-2-60, NUDOCS 8103120306
Download: ML17258A867 (44)

Text

Docket No. 50-,244 g Gcn -Fi-aa.-S 6 FE8 2V 198t DISTRIBUTION SEO-1 internal) 07 (external)

I Mr. John E. Mafer Vice President.

Electric and Steam Production Rochester Gas 8 Electr ic Corpor atf on 89 East Avenue Rochester, New York 14649

~eo6 )ge)

COA9ELTS 82M 8

Dear Mr. Maier:

V RE:

SEP TOPICS V-II.A, ISOLATION OF HIGH AND LOW PRESSURE

SYSTEMS, AND VI-T.C.l,'NDEPENDENCE OF REDUNDANT ONSITE POWER SYSTEMS,-

R.E.

GINNA NUCLEAR POWER PLANT Enclosed are final evaluations of SEP Topics V-II.A and VI-7.C.1 for R.E.

Gfnna Nuclear Power Plant.

These assessments compare your facility, as described in Docket No. 50-244, with the criteria currently used by the regulatory staff for licensing new facilities.

These reports have been revised to reflect the factual comments provided by your January 8, 1981 letter.

Your observations with regard to the acceptability of alternative designs and the use of administrative controls will be considered during our preparation of the integrated safety assessment for your plant.

However, it must be pointed out that the currently approved version of Regulatory Guide 1.139 is Revision 0.

Revision 0 requi as diverse interlocks.

These evaluations wfll be basic inputs to the integrated safety assess-ment for your facility.

As previously stated, these assessments may be revised in the future if your facility design is changed or if NRC criteria relating to this subject are modified before the integrated assessment is completed.

Sincerely,

Enclosure:

Draft SEP Topics V-II.A and VI-T.C.l Dennis M. Crutchffeld, Chief Operating Reactors Branch 85 Division of Licensing P 58 HAs OFFICE fj SURNAME(

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UNITED STATES NUCLEAR REGULATORY COMM(SSION WASHINGTON, D. C. 20555 Docket No. 50-244 LS05-81-02-060 Mr. John E. Maier Vice President Electric and Steam Production Rochester Gas 8 Electric Corporation 89 East Avenue Rochester, New York 14649

Dear Mr. Maier:

RE:

SEP TOPICS V-II.A, ISOLATION OF HIGH AND LOW PRESSURE

SYSTEMS, AND VI-7.C.l, INDEPENDENCE OF REDUNDANT ONSITE POWER SYSTEMS-R.E.

GINNA NUCLEAR POWER PLANT Enclosed are final evaluations of SEP Topics V-II.A and VI-7.C.l for R.E.

Ginna Nuclear Power Plant.

These assessments compare your facility, as described in Docket No. 50-244, with the criteria currently used by the regulatory staff for licensing new facilities.

These reports have been.revised to reflect the factual comments provided by your January 8, 1981 letter.

Your observations with regard to the acceptability of alternative designs and the use of administrative controls will be considered during our preparation of the integrated safety assessment for your plant.

However, it must be pointed out that the currently approved version of Regulatory Guide 1.139 is Revision 0.

Revision 0 requires diverse interlocks.

These evaluations will be basic inputs to the integrated safety assess-ment for your facility.

As previously stated, these assessments may be revised in the future if your facility design is changed or if NRC criteria'relating to this subject are modified before the integrated assessment is completed.

Sincerely,

Enclosure:

Draft SEP Topics V-II.A and VI-7.C.l Dennis M. Crutchfield, ief Operating Reactors Branch k'5 Division of Licensing cc w/enclosure:

See next page

sA h

Nr.. John E. Maier R.

E.

GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244 cc w/enclosure:

Harry H. Voigt, Esquire

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

M.

Suite 1100 Mashington, D.

C.

20036 Mr. Michael Slade 12 Trailwood Circle Rochester, New York 14618 Rochester Committee for 5ci ent ific Informat i on Robert E. Lee, Ph.D.

P. 0. Box 5236 River Campus Station Rochester, New York 14627 J effrey Cohen New York State Energy Office Swan Street Building Core 1, Second Floor Errpire State Plaza

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

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

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

.c/o U. S.

NRC 1503 Lake Road

Ontario, New York 14519 Richard E. Schaffstall, Executive Director for SEP Owners Group 1747 Pennsylvania

Avenue, NW Washington, D.C.

20006 Director, Technical Assessment Division Office of Radiation Programs (AW-459)

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

E IS COORD INATOR 26 Federal Plaza New York, New York 10007 Herbert Grossman, Esq.,

Chairman Atomic Safety and Licensing Board U. S. Nuclear Regulatory Conmission Washington, D. C.

20555 Dr. Richard F. Cole Atomic Safety and Licensing Board U. S. Nuclear Regulatory Comission Mashington, D. C.

20555 Dr.

Emmeth A. Luebk'e Atomic Safety and Licensing Board U. S. Nuclear Regulatory Cormission Mashington, D. C.

20555 Mr. Thomas B. Cochran

~ Natural Resources Defense Council, Inc.

1725 I Street, N.

M.

Suite 600 Washington,.D.

C.

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

0 0130J SEP TECHNICAL EVALUATION TOPIC V-11.A ELECTRICAl, INSTRUMENTATION, AND CONTROL FEATURES FOR ISOLATION OF HIGH AND LOW PRESSURE SYSTEMS FINAL DRAFT R. E.

GINNA NUCLEAR STATION Docket No. 50-244 January 1981 S. E. Mays 1-26-81

CONTENTS

1. 0 INTRODUCTION

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2.0 CRITERIA 2.1 Residual Heat Removal (RHR) Sys 2.2 Emergency Core Cooling System 2.3 Other Systems tern

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3 tern tern 3.1 Residual Heat Removal (RHR) Sys 3.2 Safety Injection System 3.3 Chemical and Volume Control Sys

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SEP TECHNICAL EVALUATION

.TOPIC V-11.A ELECTRICAL, INSTRUMENTATION, AND CONTROL FEATURES FOR ISOLATION OF HIGH AND LOP PRESSURE SYSTEMS FINAL DRAFT R

E. GINNA NUCLEAR STATION

1.0 INTRODUCTION

The purpose of this review is to determine if the electrical, instrumentation, and control (EISC) features used to isolate systems with a lower pressure rating than the reactor coolant primary system are in compliance with current licensing requirements as outlined in SEP Topic V-llA. Current guidance for isolation of high and low pres-sure systems is contained in Branch Technical Position (BTP) EICSB-3, BTP RSB-5-1, and the Standard Review Plant (SRP),

Section 6.3.

2 0 CRITERIA 2.1 Residual Heat Removal (RHR)

S stems.

Isolation requirements for RHR systems contained in BTP RSB-5-1 are:

1.

The suction side must be provided with the following isolation features:

a.

Two power-operated valves in series with posi-tion indicated in the control room.

b.

The valves must have independent and diverse interlocks to prevent opening if the reactor coolant system (RCS) pressure is above the design pressure of the RHR system.

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The valves must have independent and diverse interlocks to ensure at least one valve closes upon an increase in RCS pressure above the design pressure of the RHR system.

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The discharge side must be provided with one of the following features:

a.

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

b.

One or more check valves in series with a normally-closed power-operated valve whicn has its position indicated in the control room.

If this valve is used for an Emergency"Core

'ooling System (ECCS) function, the valve must open upon receipt of a safety injection signal (SIS) when RCS pressure has decreased below RHR system design pressure.

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Three check valves in series.

d.

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

2.2 Emer enc Core Coolin S stem.

Isolation requirements for ECCS are contained in SRP 6.3.

Isolation of ECCS to prevent overpres-surization must meet one of the following features:

One or more check valves in series with a normally-closed motor-operated valve (MOV) which is to be opened upon receipt of a SIS wnen RCS pressure is less than the ECCS design pressure 2.

Three check valves in series 3.

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

with the RCS must meet the following isolation requirements from BTP EICSB-3:

1.

At least two valves in series must be provided to isolate the system when RCS pressure is above the system design pressure and valve position should be provided in the control room 2.

For systems with two MOVs, each MOV should have independent and diverse interlocks to prevent opening until RCS pressure is below the system design pressure and should automatically close when RCS pressure increases above system design pressure 3.

For systems with one check valve and a MOV, the MOV should be interlocked to prevent opening if RCS

pressure is above system design pressure and should automatically close whenever RCS pressure exceeds

,system design pressure.

3.0 DISCUSSION AND EVALUATION There are three systems at R. E. Ginna Nuclear Station which have a direct interface with the RCS pressure boundary and have a design pressure rating of all or part of the system which is less than that of the RCS.

These systems are the Chemical and Volume Control System (CVCS), the Safety Injection System (SIS),

and the Residual Heat Removal (RHR) system.

3.1 Residual Heat Removal S stem..

The RHR system takes a suction on the RCS loop A hot leg, circulates the water through the RHR system heat exchanger, and discharges to the RCS loop B cold leg.

Two motor-operated valves in series provide isolation capabilities in both the suction and discnarge lines.

Each of these MOVs has position indica-tion in the control room.

The inboard (closest to the RCS) valves are interlocked to prevent opening if RCS pressure is above RHR system design pressure.

However, both valves use the same pressure switch and relay to provide this interlock.

The outboard valves have no pressure interlocks.

None of the valves will automatically close if RCS pres-sure increases above RHR system design pressure during RHR system operation.

The RHR system is not in compliance with the current licensing requirements of BTP RSB-5-1 since none of the isolation valves will automatically close if RCS pressure exceeds RHR design pressure.

Also, the outboard isolation valves have no interlocks to prevent RHR overpressurization, and the inboard valve interlocks are neither diverse nor independent.

3.2 Safet In ection S stem.

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

There are connections upstream of each check valve that can allow them to be tested.

A normally-open

motor-operated isolation accumulator has position opened automatically, if signal.

valve upstream of the check valves for each indication in the control room.

Each MOV is

closed, upon receipt of a safety injection The second SIS subsystem consists of two loops, each supplied by a safety injection pump.

Each pump discharges to the hot and cold legs of one RCS loop.

Isolation is-provided-by two check valves in series for each branch of the safety injection loop.

The cold leg check valves are testable.

The check valves in the lines supplying the RCS hot leg for each SIS loop are not testable.

However, the MOV in each hot leg is locked shut with power removed and is not required for accident mitigation.

A motor-operated isolation valve with position indication in the control room is provided in each branch of the cold le+discharge lines.

These valves open upon receipt of a safety injection signal, but have no interlocks preventing opening when RCS pressure is above SIS design pressure.

The third SIS subsystem uses the RHR system to provide low pres-sure water from the refueling water storage tank to the reactor vessel head (core deluge).

Isolation is provided by a MOV in series with a check valve in each of two branches.

The MOVs open upon receipt of a

, safety injection signal but have no interlocks to prevent opening when RCS pressure is above SIS design pressure.

The SIS is not in compliance with the current licensing require-ments of SRP 6.3 since the MOVs for the low pressure injection lines have no interlocks to prevent opening when RCS pressure exceeds SIS design pressure.

3.3 Chemical and Volume Control S stem.

The CVCS takes water from the RCS and passes it through a regenerative heat exchanger, an orifice to reduce its pressure, and a nonregenerative heat exchanger before reducing its pressure further by the use of a pressure control valve.

After filtering and cleanup, the water may be returned to the RCS by the use of the charging

pumps, which increase the water pressure

and pass it through the regenerative heat exchanger to either the hot or cold legs of the RCS or to the pressurizer auxiliary spray line.

The CVCS suction line isolation is provided by a manually-operated solenoid valve in series with three parallel solenoid-operated valves.

Each of these valves is operated from the control room and has valve position indicated.

None of the valves have interlocks to prevent opening or to automatically close if the pressure exceeds the design rating of the low pressure portions of the system.

The CVCS discharge line isolation is provided by a common dis-charge line check valve and a branch check valve in each of the three branches downstream.of the common check valve.

Drain fittings on the discharge line upstream of each check valve can allow the valves to be tested.

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

There are solenoid isolation valves in each discharge line branch which have position indication in the control room, but these valves have no interlocks to prevent system overpressuriza tion.

The CVCS is not in compliance with current licensing requirements for isolation of high and low pressure systems contained in BTP EICSB-3 since the suction line solenoid-operated valves have no interlocks to prevent system overpressurization, and the discharge line check valves have no position indication available in the control room.

4. 0

SUMMARY

'The R. E. Ginna Nuclear Station has three systems with a lower design pressure rating than the RCS, which are directly connected to the RCS.

The CVCS, SIS, and RHR system do not meet current licensing requirements for isolation of high and low pressure systems as speci-fied below.

1.

The CVCS solenoid-operated valves have no pressure-related interlocks, and the discharge line check

'alves have no position indication available in the control room as required by BTP EICSB-3 2.

The MOVs in the low pressure SIS lines have no pressure-related interlocks required by SRP 6.3 3.

None of the RHR system isolation valves automati-cally close if RCS pressure increases above RHR system design pressure during RHR system operation, and the outboard isolation valves have no pressure-related interlocks as required by BTP RSB-5-1.

The interlocks for the inboard isolation valves are neither diverse nor independent.

5.0 REFERENCES

1.,

NUREG-075/087, Branch Technical Positions EICSB-3, RSB-5-1; Standard Review Plan o.3.

2.

Updated Final Facility Description and Safety Analysis Report, Ginna Nuclear Power Plant, Unit No.

1.

3.

RGGE drawings 33013-422,

-424, -425, -426, -427, >>428, -432, -433,

-434, -435, and -436.

4.

RG&E drawings 10905-280,

-285, -287, -295, -296, -300, and -301.

0094J SEP TECHNICAL EVALUATION TOPIC VI-7.C.1 INDEPENDENCE OF REDUNDANT ONSITE POWER SYSTEMS FINAL DRAFT R. E.

GINNA NUCLEAR STATION Docket No. 50-244 January 1981 S. E. Mays 1-26-81

CONTENTS

1.0 INTRODUCTION

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2 ~ 0 CRITERIA

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2.1 AC Supplies 2.2 DC Supplies

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2 3.1 AC Supplies 3.2 DC Supplies

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SEP TECHNICAL EVALUATION TOPIC VI-7.C.1 INDEPENDENCE OF REDUNDANT ONSITE POWER SYSTEMS FINAL DRAFT R. E.

GINNA NUCLEAR STATION

1.0 INTRODUCTION

The objective of this review is to determine if the onsite elec-trical power 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 onsite 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 308-1974, "IEEE Standard Criteria for Nuclear Power Gen-erating Stations" provide a basis acceptable to the NRC staff for meeting GDC 17 in regards to electrical independence of onsite power systems.

2. 0 CRITERIA load groups and redundant standby sources should be independent of each other at least to the following extent.

1.

The standby source of one load group should not be automatically paralleled with the standby source of another load group under accident conditions 2.

No provisions should exist for automatically trans-ferring one load group to another load group or loads between redundant power sources

3.

If means exist for manually connecting redundant load groups together, at least one interlock should oe provided to prevent an operator error that. would parallel their standby power sources.

"RR

'attery and battery charger.

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

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

Diesel generator 1A (DGlA) supplies 480 V buses 14 and 18 while diesel generator 1B (DG1B) sup-plies buses 16 and 17.

Manual means exist to tie buses 17 and 18 througn a tie breaker and to tie buses 14 and 16 through a tie breaker.

The control circuit for each breaker provides interlocks such that the breaker cannot be shut if either DG is closed on either bus or if the normal feeders 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 bus, or any safety injection signal.

Means exist to power safety injection pump SI-1C from either bus 14 or 16.

The control circuit for the breaker from each bus is designed such tnat shutting of one breaker 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 sources.

Each bus is supplied oy a pair of mechanically interlocked breakers such tnat paralleling of redundant sources is prevented.

Evaluation.

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

The manual transfer of load groups or manual interconnect'ion of emergency buses have the-required interlocks to prevent inadvertent paralleling of redundant sources.

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

Means exist to interconnect

'ooth trains by manually shutting a tie breaker.

This breaker is padlocked open and the key is maintained by the shift foreman.

Current operating procedures require removal of the feeder ruse from one of the buses feeding the tie breaker prior to closing the tie breaker

However, no interlocks exist to prevent 3

closure of the tie breaker 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, DGlB control

panel, and tne rod drive MG set control panel automatically transzers to tne redundant train on a loss of power from the normal source.

Each load will automatically transfer back to tne normal supply when it is regained.

Evaluation.

The 125 V DC system has one manual tie between redun-dant trains and seven automatic transfers of power from one redundant train to the other.

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

c

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Therefore, the 125 V

DC system is not in compliance with cur-rent licensing requirements with respect to independence of onsite J'ower systems.

4. 0

SUMMARY

The review of docketed information and plant electrical drawings indicate tnat the Ginna Nuclear Station onsite AC redundant power sources and distribution system meet the cuxrent licensing requirements for independence of onsite power systems.

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

5 0 REFERENCES General Design Criterion 17, "Electrical Power System," of Appen-dix A, "General Design Criteria of Nuclear Power Plants,"

to 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities."

2.

"Independence Between Redundant Standby (Onsite)

Power Sources and Between Their Distribution Systems,"

Regulatory Guide 1.6.

3.

Rochester Gas and Electric Coxp. letter (White) to NRC (Ziemann) dated April 18, 1979 RG&E Corp. drawings )0905-59, 62, 63, 74, and 75.

5.

RG&E Corp. drawings D-206-51, 21489-269, and 33013-652.

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0 REGULATORY,IIRNATION DISTRIBUTION SYSl+ (RIDE)

ACCESSION NBRt8103120306I DOC ~ DATE4 Sf/02/27 NOTARIZEDt NO DOCKE>>T' FACIL't50 244 Robert E'mmet Ginna>> Nuclear, Planti Unit lt Rochester; G.

05000244 AUTH~ NAME>>

AUTHOR AFFILIATION CRUTCHFIELD'iD,'perating. Reactors Branch 5

REC IP ~ NAh1E~

RECIPIENT AFFILCAT'ION MAIERiJ ~ E ~

Rochester Gas t>> Electric Corp<

SUBJECT:

For wards final>> evaluation of SEP Topic V<<I'I,A r ei isolation>>

of hi gh L l ow pr essur e>> sys L SEP Topic QI 7,C ~ 1 r e

'ndependehce.

of redundant" onsite power>> sys.

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 FEg 2 "I I981 Docket No. 50-244 LS05-8')-02-060 Mr. John E. Maier Vice President Electric and Steam Production Rochester Gas 8 Electric Corporation 89 East Avenue Rochester, New York 14649

Dear Mr. Maier:

'/g t'[g

~

><< ~es COIOjII~ION

~R REC~g RE:

SEP TOPICS V-II.A, ISOLATION OF HIGH AND LOW PRESSURE

SYSTEMS, AND VI-'7.C.l, INDEPENDENCE OF REDUNDANT ONSITE POHER SYSTEMS-R.E.

GINNA NUCLEAR POWER PLANT Enclosed are final evaluations of SEP Topics V-II.A and VI-7.C.1 for R.E.

Ginna Nuclear Power Plant.

These assessments compare your facility, as described in Docket No. 50-244, with the criteria currently used by the regulatory staff for licensing new facilities.

These reports have been revised to reflect the factual comments provided by your January 8,

1981 letter.

Your observations with regard to the acceptability of alternative designs and the use of administrative controls will be considered during our preparation of the integrated safety assessment for your plant.

However, it must be pointed out that the currently approved version of Regulatory. Guide 1.139 is Revision 0.

Revision 0 requires diverse inter locks.

These evaluations will be basic inputs to the integrated safety assess-ment for your facility.

As previously stated, these assessments may be revised in the future if your faci'lity design is changed or if NRC criteria relating to this subject are modified before the integrated assessment is completed.

Sincerely,

Enclosure:

Draft SEP Topics V-II.A and VI-7.C.l cc w/enclosure:

See next page Dennis M. Crutchfield, ief Operating Reactors Branch 85 Division of Licensing ggt 5 i(l

~5~

INCISE EÃ g~)

I i

gr uu if'

'1. t H'y r

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Mr. John E. Maier R.

E.

GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244 cc w/enclosure:

Harry H. Voi gt, Esquire

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

M.

Suite 1100 Mashington, D. C.

20036 Mr. Michael Slade 12 Trailwood Circle, Rochester, New York 14618 Rochester Committee for Sci ent ifi c Informat i on 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 F loor Enpire State Plaza

Albany, New York 12223 Director, Technical Development Programs State of New York Energy Office Agency Building 2 Empire 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 Richard E. Schaffstall, Executive Director for SEP Owners Group 1747 Pennsylvania

Avenue, NW Washington, D.C.

20006 Director, Technical Assessment Division Office of Radiation Programs (AW-459)

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

E IS COORDINATOR 26 Federal Plaza New York, New York 10007 Herbert Grossman, Esq.,

Chairman Atomic Safety and Licensing Board U. S. Nuclear Regulatory Comission Washington, D. C.

20555 Dr. Richard F. Cole Atomic Safety and Licensing Board

. U. S. Nuclear Regulatory Comission Washington, D. C.

20555 Dr.

Emmeth A. Luebke Atomic Safety and Licensing Board U. S. Nuclear Regulatory Coranission Washington, D. C.

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

1725 I Street, N.

M.

Suite 600 Mashington,-D.

C.

20006 r

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

0130J SEP TECHNICAL EVALUATION TOPIC V-11.A ELECTRICAl, INSTRUMENTATION, AND CONTROL FEATURES FOR ISOLATION OF HIGH AND LOW PRESSURE SYSTEMS FINAL DRAFT R. E.

GINNA NUCLEAR STATION Docket No. 50"244 January 1981 S. E. Mays 1-26-81

CONTENTS

1. 0 INTRODUCTION

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2.0 CRITERIA 2.1 2.2 2.3 Residual Heat Removal (RHR) System Emergency Core Cooling System Other Systems 1

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3.0 DISCUSSION AND EVALUATION

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3 3.1 Residual Heat Removal (RHR) System 3.2 Safety Injection System 3.3 Chemical and Volume Control System 3

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SUMMARY

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5.0 REFERENCES

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6 1.1

SEP TECHNICAL EVALUATION

,TOPIC V-11.A ELECTRICAL, INSTRUMENTATION, AND CONTROL FEATURES FOR ISOLATION OF HIGH AND LOW PRESSURE SYSTEMS FINAL DRAFT R. E.

GINNA NUCLEAR STATION

1.0 INTRODUCTION

The purpose of this review is to determine if the electrical, instrumentation, and control (EIGC) features used to isolate systems with a lower pressure rating than the reactor coolant primary system are in compliance with current licensing requirements as outlined in SEP Topic V-llA. Current guidance for isolation of high and low pres-sure systems is contained in Branch Technical Position (BTP) EICSB"3, BTP RSB-5-1, and the Standard Review Plant (SRP),

Section 6.3.

2.0 CRITERIA 2.1 Residual Heat Removal (RHR)

S stems.

Isolation requirements for RHR systems contained in BTP RSB-5-1 are:

1.

The suction side must be provided with the following isolation features:

a.

Two power-operated valves in series with posi-tion indicated in the control room.

b.

The valves must have independent and diverse interlocks to prevent opening if the reactor coolant system (RCS) pressure is above the design pressure of the RHR system.

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The valves must have independent and diverse interlocks to ensure at least one valve closes upon an increase in RCS pressure above the design, pressure of the RHR system.

2.

The discharge side must be provided with one of the following features:

a.

The valves, position indicators, and interlocks described in (1)(a) through (1)(c) above.

b.

One or more check valves in series with a normally-closed power-operated valve which has its position indicated in the control 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) when RCS pressure has decreased below RHR system design pressure.

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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 Emer enc Core Coolin S stem.

Isolation requirements for ECCS are contained in SRP 6.3.

Isolation of ECCS to prevent overpres-surization must meet one of the following features:

l.

One or more check valves in series with a normally-closed motor-operated valve (MOV) which is to be opened upon receipt of a SIS wnen RCS pressure is less than the ECCS design pressure 2.

Three check valves in series 3.

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

with the RCS must meet the following isolation requirements from BTP EICSB-3:

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At least two valves in series must be provided to isolate the system when RCS pressure is above the system design pressure and valve position should be provided in the control room 2.

For systems with two MOVs, each MOV should have independent and diverse interlocks to prevent opening until RCS pressure is below the system design pressure and should automatically close when RCS pressure increases above system design pressure 3.

For systems with one check valve and a MOV, the MOV should be interlocked to prevent opening if RCS

pressure is above system design pressure and should automatically close whenever RCS pressure exceeds system design pressure.

3.0 DISCUSSION AND EVALUATION There are three systems at R. E. Ginna Nuclear Station which have a direct interface with the RCS pressure boundary and have a design pressure rating of all or part of the system which is less than that of the RCS.

These systems are the Chemical and Volume Control System (CVCS), the Safety Injection System (SIS),

and the Residual Heat Removal (RHR) system.

3.1 Residual Heat Removal S stem..

The RHR system takes a suction on the RCS loop A hot leg, circulates the water through the RHR system heat exchanger, and discharges to the RCS loop B cold leg.

Two motor-operated valves in series provide isolation capabilities in both the t

suction and discnarge lines.

Each of these MOVs has position indica-tion in the control room.

The inboard (closest to the RCS) valves are interlocked to prevent opening if RCS pressure is above RHR system design pressure.

However, both valves use the same pressure switch and relay to provide this interlock.

The outboard valves have no pressure interlocks.

None of the valves will automatically close if RCS pres-sure increases above RHR system design pressure during RHR system operation.

The RHR system is not in compliance with the current licensing requirements of BTP RSB-5-1 since none of the isolation valves will automatically close if RCS pressure exceeds RHR design pressure.

Also, the outboard isolation valves have no interlocks to prevent RHR overpressurization, and the inboard valve interlocks are neither diverse nor independent.

3.2 Safet In'ection S stem.

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

There are connections upstream of each check valve that can allow them to be tested.

A normally-open

motor-operated isolation accumulator has position opened automatically, if signal.

valve upstream of the check valves for each indication in the control room.

Each MOV is

closed, upon receipt of a safety injection The second SIS subsystem consists of two loops, each supplied by a safety injection pump.

Each pump discharges to the hot and cold legs of one RCS loop. --Isolation is provided by two check valves in series for each branch of the safety injection loop.

The cold leg check valves are testab'le.

The check valves in the lines supplying the RCS hot leg for each SIS loop are not testable.

However, the MOV in each hot leg is locked shut with power removed and is not required for accident mitigation.

A motor-operated isolation valve with position indication in the control room is provided in each branch of the cold lepdischarge lines.

These valves open upon receipt of a safety injection signal, but have no interlocks preventing opening when RCS pressure is above SIS design pressure.

The third SIS subsystem uses the RHR system to provide low pres-sure water from the refueling water storage tank to the reactor vessel head (core deluge).

Isolation is provided by a MOV in series with a check valve in each of two branches.

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

The SIS is not in compliance with the current licensing require-ments of SRP 6.3 since the MOVs for the low pressure injection lines have no inter3.ocks to prevent opening when RCS pressure exceeds SIS design pressure.

3.3 Chemical and Volume Control S stem.

The CVCS takes water from the RCS and passes it through a regenerative heat exchanger, an orifice to reduce its pressure, and a nonregenerative heat exchanger before reducing its pressure further by the use of a pressure control valve.

After filtering and cleanup, the water may be returned to the RCS by the use of the charging

pumps, which increase the water pressure

and pass it through, the regenerative heat exchanger to either the hot or cold legs of the RCS or to the pressurizer auxiliary spray line.

The CVCS suction line isolation is provided by a manually-operated solenoid valve in series with three parallel solenoid-operated valves.

Each of these valves is operated from the control room and has valve position indicated.

None of the valves have interlocks to prevent opening or to automatically close if the pressure exceeds the design rating of the low pressure portions of the system.

The CVCS discharge line isolation is provided by a common dis-charge line check valve and a branch check valve in each of the three branches downstream.of the common check valve.

Drain fittings on the discharge line upstream of each check valve can allow the valves to be tested.

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

There are solenoid isolation valves in each discharge line branch which have position indication in the control room, but these valves have no interlocks to prevent system overpressurization.

The CVCS is not in compliance with current licensing requirements for isolation of high and low pressure systems contained in BTP EZCSB-3 since the suction line solenoid-operated valves have no interlocks to prevent system overpressurization, and the discharge line check valves have no position indication available in the control room.

4.0 SVmmV The R. E. Ginna Nuclear Station has three systems with a lower design pressure rating than the RCS, which are directly connected to the RCS.

The CVCS, SIS, and RHR system do not meet current licensing requirements for isolation of high and low pressure systems as speci-fied below.

2.

The CUCS solenoid-operated valves have no pressure-related interlocks, and the discharge line check valves have no position indication availab'le i'ri the

control room as required by BTP EICSB-3 The MOVs in the low pressure SIS lines have no pressure-related interlocks required by SRP 6.3 3.

None of the RHR system isolation valves automati-cally close if RCS pressure increases above RHR system design pressure during RHR system operation, and the outboard isolation valves have no pressure-related interlocks as required by BTP RSB-5-1.

The interlocks for the inboard isolation valves are neither diverse nor independent.

5 0 REFERENCES 1.,

NUREG-075/087, Branch Technical Positions EICSB-3, RSB-5-1; Standard Review Plan o.3.

2.

Updated Final Facility Description and Safety Analysis Report, Ginna Nuclear Power Plant, Unit No. l.

3.

RGB drawings 33013-422,

-424,

-425> -426, -427, -428, -432, -433,

-434, -435, and -436.

4.

RGSE drawings 10905-280,

-285, -287, -295, -296, -300, and -301.

0094J SEP TECHNICAL EVALUATION TOPIC VI-7.C.1 INDEPENDENCE.OF REDUNDANT ONSITE POWER SYSTEMS FINAL DRAFT R.

ED GINNA NUCLEAR STATION

'ocket No. 50-244 January 1981 S. E. Mays 1-26-81

C CONTENTS

1.0 INTRODUCTION

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1 2 ~ 0 CRITERIA

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2.1 2-2 AC Supplies DC Supplies

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2 3.0 DISCUSSION AND EVALUATION

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2 3.1 AC Supplies 3.2 DC Supplies

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SUMMARY

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SEP TECHNICAL EVALUATION TOPIC VI-7.C.1 INDEPENDENCE OF REDUNDANT ONSITE POWER SYSTEMS FINAL DRAFT R. E.

GINNA NUCLEAR STATION

1.0 INTRODUCTION

The objective of this review is to determine if the onsite elec-trical power 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 onsite 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 308-1974, "IEEE Standard Criteria for Nuclear Power Gen-erating Stations" provide a basis acceptable to the NRC starf for meeting GDC 17 in regards to electrical independence of onsite power systems.

2.0 CRITERIA load groups and redundant standby sources should be independent of each other at least to the following extent.

1.

The standby source of one load group should not be automatically paralleled with the standby source of another load group under accident conditions 2.

No provisions should exist ror automatically trans-ferring one load group to another load group or loads between redundant po~er sources

If means exist for manually connecting redundant load groups together, at least one interlock should oe provided to prevent an operator error that. would parallel thei'r standby po~er sources.

battery and battery charger.

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

3.0 DISCUSSION AND EVALUATION W

Discussion Ginna onsite emergency AC power system consists of two redundant diesel-generator power trains.

Diesel generator 1A (DG1A) supplies 480 U buses 14 and 18 while diesel generator 1B (DG1B) sup-plies buses 16 and 17.

Manual means exist to tie buses 17 and 18 through a tie breaker anB to tie buses 14 and 16 through a tie breaker.

The control circuit for eacn breaker provides interlocks such that the breaker cannot be shut if eitner DG is closed on either bus or if the normal feeders 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 bus, or any safety injection signal.

Means exist to power safety injection pump SI-1C from either bus 14 or 16.

The control circuit for the breaker from each bus is designed such tnat shutting of one breaker prevents shutting the other breaker so that paralleling the redundant DGs is prevented.

Instrument buses lA, 1B, 1C,

'and 1D are capable of being supplied by multiple sources.

Each bus is supplied by a pair of mechanically interlocked breakers such tnat paralleling of redundant sources is prevented.

Evaluation.

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

The manual transfer of load groups or manual interconnection of emergency buse's have the required interlocks to prevent inadvertent paralleling of redundant sources.

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

Means exist to interconnect both trains by manually shutting a tie breaker.

This breaker is padlocked open and the key is maintained by the shift foreman.

Current operating procedures require removal of the feeder ruse from one of the buses feeding the tie breaker prior to 3

closing the tie breaker

However, no interlocks exist to prevent closure of the tie breaker 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, DG1B control panel and tne rod drive MG set control panel automatically transzers to tne redundant train on a loss of power from the normal source.

Each load will automatically transfer back to tne normal supply when it is regained.

Evaluation.

The 125 V DC system has one manual tie between redun-dant trains and seven automatic transfers of power from one redundant train to the other.

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

trains.

Therefore, the 125 V

DC system is not in compliance with cur-rent licensing requirements with respect to independence of onsite power systems.

4. 0

SUMMARY

The review of docketed information and plant electrical drawings indicate tnat the Ginna Nuclear Station onsite AC redundant power

~t 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 breaker which are not in compliance with current criteria for independence of onsite power systems.

5 0 REFERENCES General Design Criterion 17, "Electrical Power System," of Appen-dix A, "General Design Criteria of Nuclear Power Plants,"

to 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities."

2.

"Independence Between Redundant Standby (Onsite)

Power Sources and Between Their Distribution Systems,"

Regulatory Guide 1.6.

3.

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

4.

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

5.

RG6E Corp.

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

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