ML17309A130

Notifies That Scope of NRC Review for Amend 7 to License DPR-18 Is Identical to SEP Topic VI-7.C Re ECCS Single Failure Criterion for Locking Out Power to Valves & Topic VI-7.C.2 Re Failure Mode Analysis.No Addl Review Required

ML17309A130
Person / Time
Site:	Ginna
Issue date:	02/20/1981
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:	Maier J ROCHESTER GAS & ELECTRIC CORP.
References
TASK-06-07.C, TASK-06-07.C2, TASK-6-7.C, TASK-6-7.C2, TASK-RR LSO5-81-02-037, LSO5-81-2-37, NUDOCS 8102260627
Download: ML17309A130 (52)

Text

UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555

'0 tgei Docket No. 50-244 LS05-81-02-037

~c Mr. John E. Maier Vice President Electric and Steam Production Rochester Gas 8 Electric Corporation 89 East Avenue Rochester, New York 14649

SUBJECT:

SEP TOPICS VI-7.C, ECCS SINGLE FAILURE CRITERION AND REQUIREMENTS FOR LOCKING OUT POWER TO VALVES AND VI-7.C.2, FAILURE MODE ANALYSIS (R.

E.

GINNA)

The staff has determined that the scope of review and evaluation performed for Amendment No.

7 to DPR-18 for ECCS Single Failure is identical to the subject SEP topics.

Additional review and evaluation is therefore not required; Accordingly, we consider these topics to be completed with the exception of publishing the integrated assessment for your plant.

Sincerely, cc:

See next page Dennis M. Crutchfield, ief Operating Reactors Branch No, 5

Division of Licensing gEol 5

,/j QSQ

@f6 6x C07)

Hr. John E. Maier R. E.

GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244 CC Harry H. Voigt, Esquire

LeBoeuf, Lamb, Leiby and NacRae 1333 New Has@shire

Avenue, N.

W.

Suite 1100 Washington, D. C.

20036 Hr. Michael Slade 12 Trailwood Circle Rochester, New York 1461B Rochester Committee for Scientific Information 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 l., Second Floor Empire State Plaza

Albany, New York 12223 Director, Technical Development Programs State of New York Energy Office Agency Building 2 Eopire 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 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 Hall S2

'rlington, Virginia 20460 U. S. Environmental Protection Agency Region II Office ATTN:

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

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

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

20555 Dr.

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

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

1725 I Street, N.

W.

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

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+**y4 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 November 27, 1981 Docket No. 50-244 LS05-81-11-066 Mr. John E. Maier Vice President Electric and Steam Production Rochester Gas 8 Electric Corporation 89 East Avenue Rochester, New York 14649

SUBJECT:

SEP TOPIC VI-7.C.1, APPENDIX K - ELECTRICAL INSTRUMENTATION AND CONTROL (EI&C) RE-REVIEWS, SAFETY EVALUATION FOR R.

E.

GINNA Enclosure 1 is our contractor's final evaluation of this topic.

The eval-uation has been revised to reflect the additional information provided in your July 14, 1981 letter.

Enclosure 2 is the staff safety evaluation that is based upon Enclosure 1, and your letter, and supplements our contractor's evaluation.

Enclosure 2

notes that your design provides an acceptable alternative to current criteria.

Accordingly, the staff considers Topic VI-7.C.l for your plant to have been completed acceptably.

Sincerely,

Enclosures:

As stated Dennis M. Crutchfield, Chief Operating Reactors Branch No.5 Division of Licensing cc w/enclosures:

See next page

Nr. John E. Maier CC Harry H. Voigt, Esquire LeBoeuf, Lamb, Leiby and MacRae 1333 New Hampshire Avenue, N. M.

Suite 1100 Mashington, D. C.

20036 Hr. Michael Slade 12 Trailwood Circle Rochester, New York 14618 Ezra Bialik Assistant Attorney General Environmental Protection Bureau New York State Department of Law 2 Morld Trade Center New York, New York 10047 Jeffrey Cohen New York State Energy Office Swan Street Building Core 1, Second Floor Empire State Plaza

Albany, New York 12223 Director, Bureau of Nuclear Operations 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 Mr. Thomas B. Cochran Natural Resources Defense Council, Inc.

1725 I Street, N. M.

Suite 600 Mashington, D. C.

20006 U. S. Environmental Protection Agency Region II Office ATTN: "EIS COORDINATOR 26 Federal Plaza New York, New York 10007 Herbert Grossman, Esq.,

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

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

20555 Dr. Eometh A. Luebke Atomic Safety and Licensing Board U. S. Nuclear Regulatory Comnission Mashington, D. C.

20555

fOPIC:

YI-7.C.1 APPENDIX K - ELECTRICAL INSTRUMENTATION AND CONTROL lKC RE-REVIEWS INTRODUCTION During the Appendix K reviews of some facilities initially considered, a detailed EIEC review wa not performed.

Accordingly we intended to re-review the modified ECCS of these facilities to confirm that it is designed to meet the most limiting single failure.

Several types of failure were considered as candidates for designation as the most limiting.

Because of the scope of the other SEP Topics, it was decided that, for the purpose of this study (and to reduce replication of effort on other SEP Topics),

the loss of a single ac or dc onsite power system was the most limiting failure.

Accordingly, this topic was limited to an evaluation of the independence between the onsite power systems.

REVIEW CRITERIA The review criteria are presented in Sectior, 2 of EGKG Report EGG-EA-5641 "Independence of Redundant Onsite Power Systems."

RELATED SAFETY TOPICS AND INTERFACES The scope of review for this topic was limited to avoid duplication of effort since some aspects of the review were performed under related topics.

The related topics and the subject matter are identified below.

Each of the related topic reports contain the acceptance criteria and review guidance for its subject matter.

YI-4 VI-7.A.3 YI-7.B VI-7.C.2 VI-7.D VI-10.A VII-1.A VII-3 VIII-2 VIII-3 YIII-4 IX-6 Bypass and Reset of Engineered Safety Features (B-24)

ECCS Actuation System ESF Switchover from Injection to Recirculation Failure Mode Analysis-ECCS Lohg Term Cooling Passive Failures (e.g., flooding)

Testing of Reactor Protection Systems Reactor Trip System Isolation Systems Required for Safe Shutdown Onsite Emergency Power Systems Emergency dc Power Systems Electrical Penetrations Fire Protection The conclusion that suitable isolation devices are 'provided is a basic assumption for Topics YI-7.C.2 and VII-3.

IV.

REVIEW GUIDELINES The review guidelines are presented in Section 3 of Report EGG-EA-5641 "Independence of Redundant Onsite Power Systems".

Y.

EVALUATION As noted in Report EGG-EA-5641, "Independence of Redundant Onsite Power Systems",

the separation between redundant systems does not satisfy the review criteria.

However, the short circuit analysis provided in the licensee's July 14, 1981 letter shows that (1) fusing has been coordinated so that faults will be cleared prior to dc bus transfer; (Z) the automatic transfer schemes for buses 14, 16, 17, and 18, DGlA control panel and DGlA control panel have electrical interlocks to prevent the paralleling of the two dc systems; (3) the two dc systems can be paralleled when the two systems are purposely tied together during the test of one set of batteries or during the maintenance or repair of a main 150 ampere charger unit; (4) no credible component failure can cause the paralleling of the two dc systems through the manual switches on the 4KV non-class IE buses;

and, (5) the automatic transfer scheme used for the main control board annunci-ators is designed so that only one of the two dc sources can be connected.

VI.

CONCLUSION As a result of our review of our contractor's work the staff concludes that the subject ac and dc onsite systems do not satisfy the review criteria.

From our review of the licensee's calculations and after consultation with our contractor, we also conclude that the present design and administrative controls provide an acceptable alternative to our criteria provided that fuse types and sizes, battery capacity, and electrical loads are not changed.

EGG-EA-5641 NOVEMBER 1981 SYSTEMATIC EVALUATION PROGRAM, TOPIC VI-7.C.1, INDEPENDENCE OF REDUNDANT ONSITE POWER

SYSTEMS, R.

E.

GINNA NUCLEAR STATION S.

E.

Mays Revised by R.

VanderBeek Prepared for the U.S. Regulatory Commission Under DOE Contract No. DE-AC07-76ID01570 FIN Ho. A6425

+Q ECHE&idaho Inc

>AHMfGSC 595 il<ev )1 19)

C'NTERlMREPORT Accession No.

Report No.

EGG-EA-5641 Contract Program or Project Tille:

Electrical, Instrumentation, and Control Systems Support for the Systematic Evaluation Program (III)

Subject of this Document:

Systematic Evaluation Program, Topic VI-7.C. 1, Independence of Redundant Onsite Power Systems, R.

E. Ginna Nuclear Station Type of Document:

Informal Report Author(s):

S.

E.

Mays Revised by R. VanderBeek Date of Document:

November 1981 Responsible NRC Individual and NRC Oflice or Division:

Ray F. Scholl, Jr., Division of Licensing This document was prepared primarily for preliminary or internal use. It has not received fullreview and approval. Since there may be substantive changes, this documenf should not be considered final.

EGB G Idaho, Inc.

Idaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission Washing ton, D.C.

Under OOE Contract No.

DE-AC07-761D01 670 NRC FIN No. A6425

~C INTERIM REPORT

0094 J SYSTEf'lATIC EVALUATION PROGRAM TOPIC V1-7.C.1 INDEPENDENCE OF REDUNDANT ONSITE POWER SYSTEMS R. E.

GINNA NUCLEAR STATION Docket No. 50-244 November 1981 S.

E.

Mays Revised by R. Vanderbeek EGEG Idaho, Inc.

11-5-81

ABSTRACT This SEP Technical Evaluation, for the R.

E. Ginna Nuclear Station, reviews the electrical independence between redundant standby (onsite) power sources and their distribution systems.

FOREWARD This report is supplied as part of the "Electrical, Instrumentation, and Control Systems Support for the Systematic Evaluation Program

( II) being conducted for the U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Division of Licensing by EG8G Idaho, Inc.,

Reliability

& Statistics Branch.

The U.S. Nuclear Regulatory Commission funded the work under the authorization B8R 20-10-02-.05, FIN A6425-1.

CONTENTS

1.0 INTRODUCTION

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SYSTEMATIC EVALUATION PROGRAM TOPIC V1-7.C.1 INDEPENDENCE OF REDUNDANT ONSITE POWER SYSTEMS FINAL DRAFT R. E.

GINNA HUCLEAR STATION 1.0 INTRODUCTIOH The objective of this review is to determine if the onsite electrical power systems (AC and DC) are in compliance with current licensing 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 sufficient inde-pendence to perform their safety function assuming a single failure.

Regulatory Guide l.o,

" Independence Between Redundant Standby (Onsite)

Power Sources and Between Their Distribution Systems,"

and IEFE Stan-dard 308-1974, "IEEE Standard Criteria for Nuclear Power Generating Sta-tions" provide a basis acceptable to the NAC staff for meeting GDC 17 in reaards to electrical independence of onsite power systems.

2.0 CRITERIA operating from standby

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

l.

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 be provided to prevent an operator error that would parallel their standby power sources..

2.2 DC Supplies.

As stated in Regulatory Guide 1.6, Section 0.3, each d-c load group should be energized by a battery and battery charger.

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

3.0 DISCUSSION ANO EVALUATION 3.1 AC Supplies Discussion.

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

Diesel generator lA (OGlA) supplies 480 V buses 14 and 18 while diesel generator 1B (OG1B) supplies buses 16 and 17.

f1anual means exist to tie buses 17 and 18 through 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 OG is closed on either bus or if the normal feeders to the bus are closea.

Additionally, if the tie breakers are closed, they will trip open upon restoration of normal

power, OG closing on the bus, or any safety injection signal.

Heans 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 that shutting of one breaker prevents shutting the other breaker so that paralleling the redundant OGs is prevented.

Instrument buses lA, 1B, 1C, and 10 are capable of being supplied by multiple sources; Each bus is supplied by a pair of mechanically inter-locked breakers such that paralleling of redundant sources is prevented.

Evaluation.

The redundant onsite AC power trains have no automatic 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 inadvertent paralleling of redundant sources.

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

3.2

~OC S

stenos Discussion.

Ginna Nuclear Station has two redundant battery and charger trains to supply 125 V

OC emergency loads.

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

150-amp charger.

fleans 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 fuse fromm one of the buses feeding the tie breaker prior to 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 par-alleling of the redundant OC trains.

Automatic transfer of 125 V

OC load groups from train A to 8 (or vice versa) occurs in seven locations.

Control power for 480 V switchgear on buses 14, 16, 17, and 18, DGlA control panel, OG1B control panel, and the rod drive MG set control panel automatically transfer to the redundant train'n a loss of power from the normal source.

Each load will automati-cally transfer back to the normal supply when it is regained.

Evaluation.

The 125 Y

OC system has one manual tie between redundant train seven ant oniatic transfers of povrer from one redundant train to the other.

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

There-

fore, the 125 V

OC system is not in compliance with Regulatory Guide 1.6 Section 0.4.C for current licensing criteria with respect to independence of onsite power systems.

4.0 SUHtlARY The review of docketed information and plant electrical drawings indi-cate that the Ginna Nuclear Station onsite AC redundant power sources and distribution system meet the current licensing criteria for independence of onsite power systems.

The 125 V

OC system has seven automatically trans-ferred loads and one manual tie breaker which are not in compliance with current criteria for independence of onsite power systems.

5.0 REFERENCES

1.

General Design Criterion 17, "Electrical Power System," of Appendix 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 Guiae 1.6.

3.

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

4.

RG&E Corp. drawings 10905-59, 62, 63, 74, and 75.

5.

RGEE Corp.

drawings 21489-269, 33013-652 and '33013-756.

4

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'ocket.'No.

50%44 LS05 04-035 UNITEDSTATES NUCL'EAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 April 24, 1981 J'

~ ~

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;

SUBJECT:

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

SYSTEMS, V-ll,B, RHR INTERLOCK REQUIREMENTS AND VI-7>C.1, INDEPENDENCE OF REDUNDANT QWSITE POWER SYSTEMS R.

E, GINNA NUCLEAR POWER PLANT

~c:

We have reviewed your letter of March 27, 1981 and agree with resolving open items during topic evaluations rather than deferring a decision to the Integrated Assessment.

To this end, we are enclosing a revised safety eval'uation of Topic V-ll.A.

We have also reviewed your comments on the draft Technical Evaluation Report (TER)

SEP Topic V-ll.B dated January 8, 1981, Your comments on SEP Topic V-ll.B are covered by Sections

3. 1 and 3.2 of our safety evalua-t'EP Topic V-ll A.

We are enclosing a revised Technical Evaluation Report, on Topic V-ll.B which incorporates a reference to Sect>on 3

1 and 3.2 of our safety evaluation report on Topic V-ll,A.

We are enclosing a request for additional information on SEP Topic VI-7,C,l where we do not have sufficient information to reach an independent safety assessment.

Sincerely,'nclosure:

SER for SEP Topic V-ll.A Quest)ons for SEP Topic VI-.7.C,l cc w/enclosureI See next page

~ 9f.

Dennis M, Crutchfield, ief Operating Reactors Branch No, 5

Division of Licensing

RE(UEST FOR ADDITIONAL INFORMATION ON SEP TOPIC VI-7.C.1 FOR R. E.

GINNA 1.

For each of the seven automatic transfers from one dc train to the other, provide the short circuit analyses and the protective device coordination curves.

Short circuit analyses should be provided for each of the following initial conditions:

a.

Full battery charge with equilizing charge in progress b.

Battery near discharge with chargers not available.

2.

Describe the methods used to assure that fault interrupting devices remain within the curves provided in response to guestion l.

Your answer to this question should address breaker test frequency used vice that recommended by the breaker manufacturer and production lot verifica-tion of fuse characteristics.

3.

Provide electrical schemhti~

from load to dc bus for each transfer and the other drawings given in References 4 and 5 of Enclosure 2 to your letter of March 27, 1981.

S

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 F~r[

a IG~B1 gcg~g~

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

~g 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 b'e pointed out that the currently approved

'ersion 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:

'raft SEP Topics V-II.A and VI-7.C.1 Dennis M. Crutchfield, ief Operating Reactors Branch 85 Division of Licensing cc w/enclosure:

See next page

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 3.1 AC Supplies 3.2 DC Supplies

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4

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

If means exist for manually connecting redundant load groups together, at least one interlocx should ne provided to prevent an operator error that would parallel their 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 Discussion Ginna onsite emergency AC power system consists of two redundant diesel-generator power trains.

Diesel generator lA (DG1A) supplies 480 V buses 14 and 18 wnile diesel generator 1B (DGIB) sup-plies buses 16 and 17.

Manual means exist to tie buses 17 and 18 through 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, tney will trip open upon restoration of normal power, DG closing on the bus, or'ny 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 tne 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 po~er trains 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 inadvertent 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 3 seems 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 fuse from one of the buses feeding the tie breaker prior to 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, DGlA control panel, DGlB control

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

Each load will automatically transfer baca 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 otner.

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.

4o0

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 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 RG&E Corp. drawings 10905-59, 62, 63, 74, and 75.

5.

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

TOPIC VI-7,C,2 SEE TOPIC VI-7.C

TOPIC VI-7,C,3 SEE TOPIC II-4.E

TOPIC VI-7D - Long Term Cooling Pressure Failures SEP Plants Affected - All PMRs DBEs Affected - Loss-of-Coolant Accidents Discussion This issue was raised by Mr. Ronald M. Fluegge in an October 24, 1976 letter to then Chairman Rowden. It was later defined in the Office of Nuclear Reactor Regulation as follows:

"The General Design Criteria require that the Emergency Core Cooling Systems (ECCS) shall be capable of providing adequate core cooling following a Loss of Coolant Accident, assuming a

single failure in Emergency Core Cool'ing Systems.

The staff assumes the single failure to be either an active failure during the injection phase, or an active or'passive failure during the long-term recirculation phase.

The physical layouts of engineered safety feature pumps and components on.some pressurized water reactors makes them vulnerable to floodino that might result from large passive failures in system piping, although they are protected for more likely events, such as sudden seal failure.

Large pipe ruptures are not required to be protected against because of their low probability during the ECCS recirculation mode."

P a As stated in the "NRR Reports on Allegations Made by Mr. Ronald M.

Fluegge" (ll/76):

'The General Design Criteria (Appendix A to 10 CFR 50) include the following

- footnote regarding single failures:"

'single failures of passive components in electrical systems should be assumed in designing against a single failure.

The conditions. under which a single failure of a passive component in a fluid system should be considered in designing the system against a single.

0 failure are under development.'Thus, the General Design Criteria do not provide an explicit

'equirement for the treatment of failures of passive components.

Appendix K to 10 CFR 50 pertains to ECCS performance requirements and also does not provide explicit guidelines on the treatment of failures of passive components after a loss-of-coolant

'ccident (LOCA).

Present plants are reviewed, however, to ass'ure

'that the plant arrangement and design features provide the

'ecessary protection of essential systems and components (such as shutdown cooling and 'pressurized por'tions of emergency core cooling systems) due to poten ial piping failures as an initiating event (not concurrent with or consecutive to a LOCA).

3

~C I

l

~

~

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~

Pip.- ng. failures auts ide. containment,. are.:postulated

..ia accordance with Branch Technical Positions llEB 3-1 and APCSB 3-1 in th USHRC Standard Review Plan Section 3.6.

Longitudinal or circumferential breaks in high energy fluid system piping or leakage-cracks in a moderate energy*

0 fluid system piping are considered separately as a single postulated event occurring during normal plant conditions.

The crack size assumed. for a moderate energy pipe is equi-valent to a slot of dimensions (1/2 x pipe thickness) x (1/2 x diameter).

The plant must be designed such that the effects of such a postulated piping failure, including'the environmental conditions resulting from the escape of container fluids, do not affect function of equipment essential to safe shutdown of the reactor.

Mith regard to postulation of failures in emergency core cooling systems subsequent to a loss-of-coolant accident, the USNRC Standard Review Plan on Emergency Core Cooling System {Section 6.3) provides additional. guidance with the statement that:

'The ECCS should retain its capability to

'*Subsequent to a LOCA, all pipes of relevance are moderate energy pipes defined as a piping system carrying fluid at a temperature below 200'F ar'd at a pressure below 275 psig.

~ I,'

~

e

~ ~

I'

~

"cool the: core in the-event, of c;:fai1upeef;any.single active

'r pas'sive failure during'he. long-term recirculation-cooling phase following an accident.'ased on this guidance, the staff assures the ECCS design and layout satisfies the requirement for redundancy in such systems.'he imple-mentation of the passive failure statement does not require significant ruptures of moderate-energy piping subsequent to LOCA, as this combined event would be extremely unlikely.

The more credible passive failure is at pump or valve seals, or measurement devices.

The staff review of the effects of such a postulated leak rate includes consideration of:

(1) the flow paths of the radioactive fluid through floor drains, sump pump discharge piping, and the auxiliary building; (2) the operation of the auxiliary systems that would receive this radioactive fluid; (3) the ability.of the leakage detection system to detect the passive failure; and (4) the ability of the operator to isolate the ECCS passive failure.

Therefore, the ECCS passive failure criterion being implemented by the staff requires the consideration of additional leakage but not pipe breaks beyond the initiating LOCA.

The basis for this is the staff's judgment that the probability of

~

rr

~

serious.mul tinge pipe'aiJurgs.

is. suf.iciently.llaw." that.Whey.

%Bed not be: considered

a. desiga basis.,event. since: when operating in the long-term recirculation mode, the ECCS is subjected to temperatures and pressures much less than those for which the system is designed.

In addition, after long-term cooling has been initiated, the need for recirculation diminishes due to the decrease in available core decay heat.

For example, or a 3500 tilt reactor, the amount of core decay heat which is being produced at the beginning of a normal shutdown, is 203 lilt; after one week it has decreased to 13 Slt; and after eight weeks it is only 5.7 t5lt.

This means that significantly less coolant recirculation wou'ld be necessary a ter several week-.

The reeded cooling wa er to prevent core overheating can be provided by the RHR system even considering leakage in the suction or discharge side of the piping.

In addition, should recirculation cooling be temporarily interrupted at the end of one week,'he core would be adequately cooled by the heat transfer effected by vessel boiloff.

To r

maintain vessel

level, a makeup of only about 100 gpm would I'e necessary."

COti'CLUSTOttS

~r

)le consider this issue to be closed..

The effect of ECCS leakage will be assessed on the SEP plants during the OGE evaluation of LOCAs.

~

,i0'

))I

," ')':,(ril

+**<<+

, UN!TEDSTATES NUCLEAR REGULATORYCOiv1MISSION V(ASHINGTON, O. C. 20555

'June 1,

1977 Docket No. 50-244 Rochester Gas

& Electr'ic Corporation ATTN:

Mr. Leon D. White, Jr.

Vice President Electric and Steam Production 89 East Avenue Rochester, New York 14604 Gentlemen:

The Commission has issued the enclosed Amendment No. 14. to Provisional Operating License No. DPR-18.

This amendment consists of changes to the Technical Specifications in response to your requests dated March 10,

1975, and February 1, 1977.

This amendment revises the Technical Specifications to clarify the surveillance specification for diesel-generator starting and breaker closing times under test conditions, to establish specifications for equipment designed to mitigate the consequences of flooding of safety-related equipment due to the failure of non-seismic piping and to establish specifications for safety-related shock suppressors (sn'ubbers).

Copies of the related Safety Evaluation and the Notice of'ssuance are also enclosed.

Sincerely

Enclosures:

l.

Amendment No.

14 to DPR-18 2.

Safety Evaluation 3.

Notice A. Schwencer, Chief Operating Reactors Branch 81 Division of Operating Reactors cc w/encl:

See next page

UNITEDSTATES NUCLEAR REGULATORYCOMMISSION INASHINGTON,D. C. 20555 A

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 14 TO PROVISIONAL OPERATING LICENSE NO. DPR-18 ROCHESTER GAS AND ELECTRIC CORPORATION R. E.

GINNA ATOMIC POWER STATION DOCKET NO. 50-244 Introduction By letters dated March 10, 1975, and February 1, 1977, Rochester Gas

=and Electric Corporation {the licensee) requested amendments to License'No.

DPR-18 to revise the R.

E. Ginna Plant {the facility)

Technical Specifications.

These amendment requests propose changes to clarify the surveillance specification for diesel-generator starting and breaker closing times under test conditions, to establish Technical Specifications for equipment designed to mitigate the consequences of flooding of safety-related equipment due to the failure of non-seismic piping and to establish Technical Specifications for hydraulic shock suppressors

{snubbers).

Discussion Diesel-Generator Testin The March 10,. 1975 request for license amendment proposed changes to the Technical Specifications to clai ify the required start and breaker closure times fop, diesel-generator trains A and B under a simulated safety infection signal and proposed a specification for maximum closure times for "all breakers closed."

Flood Protection E ui ment Testin The licensee'p request for license amendment dated March 10, 1975, also proposes Technical Specification surveillance requirements for the flood protection related, circulating water pump trip equipment.

~(.

>hock Su ressors Snubbers

// 'The licensee's request for license amendment dated February 1, 1977.,

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proposes Technical Specifications for Hydraulic Snubbers based on NRC Model Technical Specifications.

Evaluation Oiesel-Generator Testin

~c The Engineered Safety Features Actuation Sequence specified in Table 8.2-4 of the Final Safety Analysis Report (FSAR) provides the design basis for the delay time for residual heat removal (RHR) pumps and for safety injection (SI) pumps reaching full rated flow when powered by the emergency diesel-generators.

A 28 second total time delay in reaching rated flow for these pumps was assumed in the Emergency Co) e Cooling System Analysi's.

Section 6.2.3 of the FSAR provides'he event sequence which constitutes this 28 second delay.

Other required loads automatically start later in the loa'ding sequence for the diesel-generator.

Those other loads include the service water pumps (total delay of 33 seconds),

the containment fans (43 seconds) and the Auxiliary Feedwater Pumps (48 seconds).

P The current Technical Specifications require that "the diesel-generator start and assume the required load in less than 30 seconds after the initial starting signal."

The proposed change would specify the allowable time delay for each emergency diesel-generator train.

The licensee included a period of 5 seconds from breaker closure time to reaching rated flow.for'ach of these pumps.

The 28 second total delay time assumed in the ECCS analysis for the RHR and SI pumps to reach rated flow also includesa 1 second time delay to initiate a Safety Injection Signal (SIS) and to account for instrument time delay.

Subtracting the 1 second and 5 second portions noted above, leaves 22 seconds from actuation of a SIS until the train 8

RHR pump breaker must close:

(Train A RHR Pump breaker must close in 20 seconds).

The licensee has therefore proposed diesel start and SI and"RHR pump breaker closure times of 20 seconds and 22 seconds for Train A and Train 8 respectively.

Since these times are those assumed in the previous acceptable safety analysis, we find this Technical Specification change to be acceptable.

e licensee has proposed to include all other safety load circuit brea ers w ic mus k

h h

t automatically close during the diesel-generator

.2-4 the diesel-loading sequence following an SIS.

FSAR Table 8. -,

generato~

loading sequence, identifies the auxiliary feedwater pump for each train as the last load required to be automatically started.

Considering the 5 second delay from breaker closure to reaching full rated flow, the times for 1A and 1B Auxiliary Feed Pumps breaker closure are 40 seconds and 42 seconds respectively.

Since these times are the same as those specified in the FSAR, we find this revision of this specification acceptable also.

Flood Protection E ui ment The licensee has made modifications to prevent the loss of function of engineering safety features (safeguards) equipment due to flooding that could be caused by a circulating water pipe or expansion joint failur'e.

Redundant water level instrumentation channels have been installed in the condenser pit and in the screen-house pit to automatically trip the circulating water (CW) pumps.

Each redundant t

't ha nel consists of three float switch level detectors arranged in a two-out-of-three logic to actuate the p

p p

circuitry.

The circulating water pumps will be tripped if'ny two of three level switches at any of the four locations sense a water level two feet above their respective pit floor elevations.

Th 1

see's letter dated March 10, 1975, proposed surveillance Technical Specifications for circulating water flood protect e

icensee ion equipment.

The proposed surveillance frequency is each refueling shutdown.

The inherent high reliability of the redundant float switch level detectors and the fail-safe character of the pump trip re ays result in a highly reliable circulating water pump trip circui Additionally, independent trip circuits are provided for each potential flood location.

Sur veillance testing at each refueling shutdown, provides additional assurance of the continued operability of this flood protection equipment and is therefore acceptable; The. licensee initially proposed to perform a functional test on each circu a n

1 ti g water pump trip channel.

The licensee subsequently agreed to revise the proposal to a calibration of each channel.

This would insure that the float switches properly actuate at 2 feet above their respective pit floors and that each channel is functiona

The proposed change to the Technical Specification for circulating water flood protection equipment did not include limiting conditions for operation.

During discussions with the staff, the licensee also agreed to limiting conditions for operation (LCO's) for the circulating water flood protection equipment in Technical Specification Table 3.5.1.

The LCO's would require redundant circulating water pump trip channels in the screen house pit and in the condenser pit to be operable.

Power operation for a period of up to seven days with one redundant channel inoperable or for a period of up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with two channels inoperable would be allowed.

Additionally a channel would be considered operable if it functions on either a

one-out-of-two logic or a two-out-of-three logic.

The licensee's proposed LCOs for circulating water flood protection equipment will assure that (1) the circulating water pumps will be tripped as a

result of flooding in either the condenser pits or the screen well house,'(2) sufficient redundancy is maintained to permit a'hannel to be out of service for testing or maintenance, and (3) that the specified, coincidence logic is maintained.

Continued power operation for a period of up to seven days with one redundant channel inoperable or for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with two redundant channels inoperable, provides a brief period for maintenance prior to requiring plant shutdown and is acceptable because of low probability of a circulating water pipe fail'ure during the allowable maintenance period.

Additionally, the operator has the capability to manually trip the circulating water pumps from the control room upon receipt of a flood alarm.

In the event'f a failure of one of three level detectors (float switches) in a single channel, the channel is considered operable ifit will perform its safety function with a one-out-of-two logic using the remaining two operable level detectors (float switches).

Sufficient detector redundancy is maintained using a one-out-of-two logic to insure that the channel is capable of performing its safety function

'onsidering a failure of the redundant channel.

Therefore, channel operation on either a two-out-of-three or one-out-of-two tripping logic is acceptable.

Shock Su ressors Snubbers Shock suppressors (snubbers) are installed at strategic locations to assist in maintaining the structural integrity of the reactor coolant system and other safety related systems against postulated seismic events or other events that could initiate an abnormal dynamic load.

It is, therefore,'ecessary, that shock suppressors installed to protect such safety system piping and components be operable during reactor operation and b'e inspected at appropriate intervals to assure their operability.

~

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xamination of defective hydraulic shock suppressors at reactor facilities has shown that the high incidence of failures observed during the summer of 1973 were caused by severe degradation of seal materials and sub-sequent leakage of hydraulic fluid.

The basic seal materials used in Bergen Paterson hydraulic shock suppressors were two types of polyurethane; a millable gum polyester type containing plasticizers and an unadulterated molded type.

Haterial tests performed at several laboratorie established that the millable gum polyurethane deteriorated rapidly under the temperature and moisture conditions present in many snubber locations at operating reactor facilities.

The molded polyurethane exhibited greater resistance to these conditions, however, it also may be unsuitable for application in higher temperature environments.

The investigation indicated that seal materials are available, primarily ethylene propylene compounds, which give satisfactory performance under the most severe conditions expected in reactor installations.

An extensive seal replacement program has been carried out at many reactor facilities.

Experience with ethylene propylene seals has been very good with no serious degradation reported thus far.

Although the seal replacement program has significantly reduced the incidence of

failures, some failures continue to occur.

These failures have severally been attributed to faulty assembly and installation, loose fittings and connections and excessive pipe vibration.

The failures have been observed in both PHRs and BHRs and have not been limited to uni s't manufactured by Bergen Paterson.

Because of the continued incidence of hydraulic shock suppressor

failures, we have concl

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uded that operability and surveillance requirements for hydrualic shock suppressors should be included in the Technical S'pecifications in all reactor facilities, regardless of manufacturer.

Our October 2, 1973 letter required the licensee to initiate a monthly inspection of the R.E.

Ginna hydraulic shock suppressors.

No hydraulic shock suppressors were.found inoperable during these monthly inspections.

The licensee's proposed Technical Specifications provide additional assurance of satisfactory shock suppressor operation and reliability.

The proposed specifications require that shock suppressors be operable during reactor operation and prior to start up.

Additionally, because protection is only required during low probability events, a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is allowed for repair or replacement of defective units before reactor shutdown must be initiated.

The licensee also proposed that the Technical Specifications allow continued reactor operation beyond 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after, showing by analysis that the integrity of the system with one or more shock suppressors inoperable can be maintained under design loading conditions.

This option for continued operation beyond 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> without specific NRC review and prior approval is not acceptable.

~(,

>ne licensee therefore concurred in deletin this

~

1 i

)ng ss provssson for continued se ec nacal Specification.

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The licensee's proposed surveillance rovides s

a surance that safety-related h

h 1

b 1 t.

Th

, th in opera le.

The inspection fre inversely with observed failures.

The ongest snspectson interval R

E 6'

'1't d

th t

C r operating facilities has shown that the ance program wi<<provide an acceptable level of p

provided that the seal material p rating environment; Hydraulic shock suppressors with the operating environment will continue d

To f th 1

p orm nce, th 1 censee hd 1

hok o

t h

f 1

re ue ing shutdown, not to exceed 1

h The proposed Technical Specifications for h d

't t th dl 1975 d

t t th o S

t ar Techn al Spec at o

fo changes which were discussed with and a reed to g

by the licensee and for S fo fo h d 1

h k

satisfactory perfonqance and; reliabi

. nce an

re >

ability and are therefore acceptable.

For the purpose of initiating the survei g

llance required by Techn cal the unit was previously at the 12 month in n...

the inspection schedule wi val.

The zni teal znspectvon ctive date of the g

suppressor.

surveillance.

Thss snitsal nth 1

p opo e

er whic was based on our Model Technical

'ased on over three years of monthl ins e

is onger initial inspection int y inspections of hydraulic shock the hydraulic shock suppressors installed in ou a

az ure.

Additionally, the seal ma n

e R. E.

anna ac lz y a nce unng prior inspections.

e op ra ng environment, based upon

4l

(vironmental Consideration Me have determined that the amendment does not authorize a change in

~~ effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact.

Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 551.5(d)(4), that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.

Conclusion I

Me have concluded, based on the considerations discussed above, that:

(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the publicwill not be endangered by operation in the proposed

manner, and (3) such activities will be conducted'in compliance with the Commission's regulations.and the issuance of this amendment will not be inimical to the common defense and security

, or to the health and safety of the public..

~(:

Date:

June 1, 1977

UNITED STATES NUCLEAR REGULATORY COMMISSlON WAS'HIMGTON, O.

C.

20555 Docket Ho. 50-244 SEP >9 >~~~

Karl R. Goller, Assistant Director for Operating Reactors, RL TECHNICAL ASSISTANCE REQUEST NO. ORB-1-167, R.

E.

GIHNA - MODIFICATIONS TO ELECTRICAL CONTROL CIRCUITS OF POST-LOCA FLOODED VALVES (TAR 1747) c.M Plant Name:

R.

E. Ginna VI 1c Docket Number:

50-244 Responsible Branch ORB-1 vl-1 ~

and Project Leader:

T.

Mambach Technical Review Branch Involved:

EI8CS Branch Review Status:

Complete Your memorandum of July 22, 1975 to R.

E. Heineman requested our evaluation of modifications proposed by Rochester Gas and Electric Corporation (RGImE). to electrical control circuits of post-LOCA flooded valves for the R.

E. Ginna Nuclear Power Station.

The modifications were described in letters from L. D. White (RGEE) to R. A. Purple (NRC) dated May.20 and May 30, 1975.

These 'modifications were accepted on an interim basis by letter dated July 3, 1975 from R. A. Purple (HRC) to L. D. White (RG8E).

The RGEE report on post-LOCA flooded valves submitted by letter dated June 16, 1975 provided supplemental information and concluded that "...no additional actions beyond those outlined in our (previous) letters.... are required."

Your request specifically asked that we determine the acceptability of the proposed modifications to the electrical control circuits for valves 852A and 852B and the report conclusion that no further modifications to post-LOCA flooded valves are required.

The Electrical, Instrumentation and Control Systems Branch has reviewed the proposed valve electrical contro) circuit modifications and the report on post-LOCA flooded vaIves for conformance to the following:

1.

Single failure criterion.

2.

Regulatory Technical Position EICSB 18 in Appendix 7A of the Standard Review Plan.

I 3.

The staff position on removing d.c. control power to a,c. motor operated valves (flOVs) as stated in Section I.B.2 of Amendment Number 7 to Provisional Operating License Number DPR-18 for the R. E. Ginna Nuclear Power Plant.

Based on the information provided, we conclude the following:

1.

The proposed modifications to the d.c. control circuits for'alves

K. R; Goller PEP ].9 18IS, 2.

MOV 852A and MOV 852B are in conformance with the stated re-quirements and will prevent improper closing of the valves despite post-LOCA flooding.

The modifications would open the positive and negative sides of the d.c. close control circuits for each valve with key switches with the valves in the closed position.

These valves are in the low head injection lines from the residual heat removal pumps to the reactor vessels.

The valves are closed during normal operation but are required to open and remain open in the event of a LOCA.

The modification will eliminate the possibility of a single short in either the a.c. or d.c. circuits spuriously closing either valve.

The control and power for the open circuits for both valves are unchanged and will perform their intended automatic functions.

The removal of a.c.

power from valves MOV 700, MOV 701 and MOY 721 is in conformance with the stated requirements and will prevent improper movement of the valves despite post-LOCA flooding.

These valves are normally closed, do not receive a safety injection signal and are required only during normal shutdown procedures.

Removal of a.c.

power to the valve motor will eliminate the possibility of a spurious opening of the.valves.

The scope of our review included only the proposed modifications to the electrical power and control circuits for valves MOY 700, MOY 701, MOV 721, MOV 852A and MOY 8528.

Based on our review, we conclude that the proposed modifications are acceptable.

It should be noted that the Reactor Systems Branch has reviewed and accepted these modifications and we have reported this in a separate memorandum.

cco S.

Hanauet R. Heineman R. Boyd R. Purple T. Rambach T. Ippolito F.

Rosa C. Berlinger R. Naventi for Reactor Safety Division of Technical Review Office of Nuclear Reactor Regulation j

f49 ~g

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=- ROCHESTER GAS AMD ELcCTI'IC COB'~i~r.'

lC(1 89 CAST AV. HUE, RQCHESTEP tc.Y. I-f849 Nil ~ >

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iitiC e. ~~( b ibi CL:3 a o-z-++

April. 1, 975

~:

Hr. Renal<>

C+ Rusche, Director Office of LJucl8ar RcecCor Regulation U.S. Nuc1cer Regulatory Conuaission Via8hinga on@

Doc ~

20555 Des Hr. Rusche:

Following zeceipC o< a 3.eater from f).

Robex'C C. Puxpl=, Chief, Oper<~t-ing RcacCo

~ Branch -.l datorl bl~roh l4, 1975, representative."-

of RUTE met with r eshbc:r'. of your staff on March 21 Co Qi::cu s EW. Purple's letter 'and to revisit Cho px'@lection against single faj.lure in the Ginna Emergency Core Coo3.in<

~~y."Oem

(~CCS)

~

-. engineered safety features pz'ouidec't Ginna to m tip te Che

~-.insequences of accidcnC~

axe c~~i cussed in detail in Chc;p"..er 6 of the Cinna PS~M.

The.-ingle failure design critcrior. which is met iG Chc.t:

each engineerei safcCy-feature provides suffi'cicnC perfowiance Capability to accommodatv any sit>gle failure of an active Component and still funcCion ia a manner to avoid undue xi,k to the healCh and

~alecky of the public".

The revio~t of the Ginna system inclucl!~A active failures o pa",.ive devi cob in 81&, ZcCG.

Spe cifically, +he reviev addressed "hc postulated spurious closing of ihe following valve r.

1)

HOV 841,

865, ln the delivery line froa the accumulators.

2)

HOV 896',

8960.

Xn the 3.inc from the re ueX ng water storage tank (RhST) to the safety injection pumps (8 P).

3)

HOV 856.

Xn the linc irom the R~'ST to the residual hcag removal (RIfg) pump 4)

NOV 078D, 8VGD..Xn +he delivery line from abc SIP to -ie reacto cool ~nt system (Res) cold 1cqs.

~

i Xn.addition the possible syuriou~ opening t'e dolive"-y line from the SIV eo ehE.

aCS of liOV 878k or 87AC (in hat legs} wes diecas-~~.

~l

~ yjjeag4 ~ g to vrla hitV pl C '4 I rC ~ I leVCKI o fore April 1, 3,975 vo e'r, aenard C.

Ru ~, Director JttCC I IIV' He have revS.eved the probability and consequence'-

of such failures end the -allowing..ctions could be taken EOE these valves, algiough the probability of uch failure'ee ls extren>el@ ternate.

A.C power could be removed f am the folio["$,ng valves:

NOV 841; 865; 8"6; and 878K, 8, C,and D.

c~ll va3,ver vouch be in thcid open position except 87SA and C, which vi1.1 be closed in accorOanci: kith Technical Spocifica-Cioh 3+3 1 i 3 g

'c~

i<a be1ievc that it would not be prudent to remove A.C. power fX'0N V)OV 896A or B.

These valves

~which a e normally in the open po.".it$ an'>

are open during thc induc{:tion phase of a &~CA and 'clo~od duping t:he recirculation phase if high head, recirculation is used.

Since this eCtion (:ould be required as early as 20 minutes followinq a LOCA, 3.t is not appropriate to remove porkier to Qe valve" since an operator Youl(7 il8ve to go to Auxiliary Building to restore power to'he valves so they could bc halo.".c4.

Recognizing this and the desirc: to Zurhhe" reduce the probability of spurious action of ~ho"c valves, a suitable switc)1 could be installed in the con"rol room which mould interrupt the D.C. control circuil.

",lith D.C. control removed, t.;io.,Box'I.'s vould he nece,..wry in tho three phase A.C. poser supply to cause spurious valve ac. '7 g-'con D. 4chite, Jr.

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TOPIC VI-7,E SEE TOPIC II-2.B