ML18038A064

Review of Licensee & Applicant Responses to NRC Generic Ltr 83-28 (Required Actions Based on Generic Implications of Salem ATWS Events),Item 1.2, Post-Trip Review:Data & Info Capabilities ..., Technical Evaluation Rept

ML18038A064
Person / Time
Site:	Nine Mile Point, 05000000, Shoreham
Issue date:	10/11/1985
From:	SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY
To:	NRC
Shared Package
ML082320175	List: ML18038A064 ML20205B068 ML20205D971 ML20205D996 ML20205E018 ML20205E031 ML20205E037 ML20205E048 ML20205E061 ML20205E065 ML20205E092 ML20205E102 ML20205E129 ML20205E135 ML20205E151
References
CON-IIT07-447-91, CON-IIT7-447-91, CON-NRC-03-82-096, CON-NRC-3-82-96 GL-83-28, NUREG-1455, SAIC-85-1521-4, NUDOCS 8510170187
Download: ML18038A064 (60)

Text

SAIC-85/1521-4 REVIEW OF LICENSEE AND APPLICANT RESPONSES TO NRC GENERIC LETTER 83-28 (Required Actions Based on Generic Implications of Salem ATWS Events),

Item 1.2 "POST-TRIP REVIEW:

DATA AND INFORHATION CAPABILITIES" FOR NINE HILE POINT NUCLEAR STATION, UNIT 2 (50-410)

Technical Evaluation Report Prepared by Science Applications International Corporation

'710 Goodridge Drive HcLean, Virginia 22102 Prepared for U.S. Nuclear Regulatory Commission Washington, D.C.

20555 Contract No. NRC-03-82-096

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FOREWORD This report contains the technical evaluation of the Nine Nile Point Nuclear Station response to Generic Letter 83-28 (Required Actions Based on Generic Implications of Salem ATWS Events), 'Item 1.2 "Post Trip Review:

Data and Information Capabilities."

For the purposes of this evaluation, the review criteria, presented in part 2 of this report, were divided into five separate categories.

These are:

1.

The parameters monitored by the sequence of events and the time hi story recorders, 2.

The performance characteristics of the sequence of events recorders, 3.

The performance characteristics of the time history recorders, 4.

The data output format, and 5.

The long-term data retention capability for post-trip revi ew material.

For this plant no information was provided in response to item 1.2 of Generic Letter '83-28.

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TABLE OF CONTENTS Section Page Introduction.

1.

Background...

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2 2.

Review Criteria

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3. Evaluation...............

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8 4.

References.

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INTRODUCTION SAIC has reviewed the material prepared in response to Generic Letter 83-28.

The response (see references) failed to provide any information regardi ng the post trip revi ew data and information capabilities at this plant.

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1.

~Back round On February 25, 1984, both of the scram circuit breakers at Unit 1 of the Salem Nuclear Power Plant failed to open upon an automatic reactor trip signal from the reactor protection system.

This incident occurred during the plant startup and the reactor was tripped manually by the operator about 30 seconds after the initiation of the automatic trip signal.

The failure of the circuit breakers has been determined to be related to the sticking of the under voltage trip attachment.

Prior to this incident; on February 22, 1983; at Unit 1 of the Salem Nuclear Power Plant an automatic trip signal was generated based on steam generator low-low level during plant startup.

In this case the reactor was tripped manually by the operator almost coinci-dentally with the automatic trip.

At that time, because the utility did not have a requirement for the systematic evaluation of the reactor trip, no, investigation was performed to determine whether the reactor was tripped automatically as expected or manually.

The utilities'ritten procedures required only that the cause of the trip be deter mined and identified the responsible personnel that could authorize a restart if the cause of the trip is known.

Following the second trip which clearly indicated the problem with the trip breakers, the question was raised on whether the circuit breakers had functioned properly during the earlier incident.

The most useful source of information in this case, namely the sequence of events printout which would have indicated whether the reactor was tripped automatically or manually during the February 22 incident, was not retained after the incident.

Thus, no judgment on the proper functioning of the trip system during the earlier incident could be made.

Fol'lowing these incidents; on February 28, 1983; the NRC Executive Director for Operations (EDO), directed the staff to investigate and report on the generic implications of these occurrences at Unit 1 of the Salem Nuclear Power Plant.

The results of the staff's inquiry into the generic implications of the Salem Unit incidents is reported in NUREG-1000, "Generic Implications of ATWS Events at the Salem Nuclear Power Plant."

Based on the results of this study, a set of required actions were developed and included in Generic Letter 83-28 which was issued on July 8, 1983 and sent to all licensees of operating reactors, applicants for operating license, and construction permit holders.

The required actions in this generic letter consist of four categories.

These are:

(1) Post-Trip Review, (2) Equipment

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Classification and Vender Interface, '(3) Post Maintenance

Testing, and (4)

Reactor Trip System Reliability Improvements.

The first required action of the generic letter, Post-Trip Review, is the subject of this TER and consists of action item 1.1 "Program Description and Procedure" and'action item 1.2 "Data and Information'Capability."

In the next section the review criteria used to assess the adequacy of the utilities'esponses to the requirements of action item 1.2 will be discussed.

2.

Review Criteria The intent of the Post Trip Review requirements of Generic Letter 83-28 is to ensure that the licensee has adequate procedures and data and information sources to understand the cause(s) and progression of a reactor trip.

This understanding should go beyond a simple identification of the course of the event.

It should include the capability to determine the root cause of the reactor trip and to determine whether safety limits have been exceeded and if so to what extent.

Sufficient information about the reactor trip event should be available so that a decision on the acceptability of a reactor restart. can be made.

The following are the review criteria developed for the requirements of Generic Letter 83-28, action item 1.2:

The equipment that provides the digital sequence of events (SOE) record and the analog time history records of an unscheduled shutdown should pro-vide a reliable source of the necessary information to be used in the post trip review.

Each plant variable which is necessary to determine the cause(s) and progression of the event(s) following a plant trip should be monitored by at least one recorder

[such as a sequence-of-events recorder or a plant process computer for digital parameters; and strip charts, a plant process computer or analog recorder for analog (time history) variables].

Each device used to record an analog or digital plant variable should be described in sufficient detail so that a determination can be made as to whether the following performance characteristics are met:

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Each sequence-of-events recorder, should be capable of detect'ing and recording the sequence of events with a sufficient time discrimination capability to ensure that the time responses asso-ciated with each monitored safety-related system can be ascer-

tained, and that a determination can be made as to whether the time response is within acceptable limits based on FSAR Chapter 15 Accident Analyses.

The recommended guideline for the SOE time II discrimination is approximately 100 msec.

If current SOE recorders do not have this time discrimination capability the licensee or applicant should show that the current time discrimi-nation capability is sufficient for an adequate reconstruction of the course of the reactor trip.

As a minimum this should include the ability to adequately reconstruct the accident scenarios pre-sented in Chapter 15 of the plant FSAR.

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Each analog time history data recorder should have a sample inter-val small enough so that the incident can be accurately reconstructed following a reactor trip.

As a

minimum, the licensee or applicant should be able to reconstruct the course of the accident sequences evaluated in the accident analysis of the plant FSAR (Chapter 15).

The recommended guideline for the sample interval is 10 sec.

If the time history equipment does not meet this guideline, the licensee or applicant should show that the current time history capability is sufficient to accurately recon-struct the accident sequences presented in Chapter 15 of the FSAR.

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To support the post trip analysis of the cause of the trip and the proper functioning of involved safety related equipment, each analog time history data recorder should be'capable of updating and retaining information from approximately five minutes prior to the trip until at least ten minutes after the trip.

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The information gathered by the sequence-of-events and time history data collectors should be stored in a manner that will allow for retrieval and analysis.

The data may be retained in either hardcopy (computer printout, strip chart output, etc.) or in an accessible memory (magnetic disc or tape).

This information should be presented in a readable and meaningful format, taking

into consideration good human factors practices (such as th'ose outlined in NUREG-0700).

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All equipment used to record sequence of events and time history information should be powered from a reliable and non-interruptible power source.

The power source used need not be safety related.

The sequence of events and time history recording equipment should monitor sufficient digital and analog parameters, respectively, to assure that the course of the reactor trip can be reconstructed.

The parameters monitored should provide sufficient information to determine the root cause of the reactor trip, the progression of the reactor trip, and the response of the plant parameters and systems to the reactor trip.

Specifically, all input parameters associated with reactor trips, safety injections and other safety-related systems as well as output parameters sufficient to record the proper functioning of these systems should be recorded for use in the post trip review.

The parameters deemed necessary, as a minimum, to perform a

post-trip review (one that would determine if the plant remained within its design envelope) are presented on Tables 1.2-1 and 1.2-2.

If the appli-cants'r licensees'OE recorders and time history recorders do not monitor all of the parameters suggested in these tables the applicant or licensee should show that the existing set of monitored parameters are sufficient to establish that the plant remained within the design envelope for the appro-priate accident conditions; such as those analyzed in Chapter 15 of the plant Safety Analysis Report.

Information gathered during the post trip review is required input for future post trip reviews.

Data from all unscheduled shutdowns provides a

valuable reference source for the determination of the acceptability of the plant vital parameter and equipment response to future unscheduled shut-downs.

It is therefore necessary that information gathered during all post trip reviews be maintained in an accessible manner for the life of the plant.

Table 1.2-1.

PWR Parameter List SOE Recorder Time History Recorder Parameter / Si nal x

(1) x x

(1) x x

(1) x x

(2)

(1) x (1) x (1) x (1) x (1) x (3) x x

(1) x (1) x (1) x (3)

Reactor Trip Safety Injection Containment Isolation Turbine Trip Control Rod Position Neutron Flux, Power Containment Pressure Containment Radiation Containment Sump Level Primary System Pressure Primary System Temperature Pressurizer Level Reactor Coolant Pump Status Primary System Flow Safety Inj.; Flow, Pump/Valve Status MSIV Position Steam Generator Pressure Steam Generator Level Feedwater Flow Steam Flow Auxiliary Feedwater System;

Flow, Pump/Value Status AC and DC System Status (Bus Voltage)

Diesel Generator Status (Start/Stop, On/Off)

PORV Position (1): Trip parameters (2): Parameter may be monitored by either an SOE or time history recorder.

(3): Acceptable recorder options are:

(a) system flow recorded on an SOE

recorder, (b) system flow recorded on a time history recorder, or (c) equipment status recorded on an SOE recorder.

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Table 1.2-2.

BMR Parameter List SOE Recorder Time History Recorder Parameter / Si nal x

x x

x x

x (1) x (1)

(2) x (1)

(2) x (1) x (1) x x (1) x (3) x (1) x (1)

(3)(4)

Reactor Tr ip Safety Injection Containment Isolation Turbine Trip Control Rod Position Neutron Flux, Power Hain Steam Radiation Containment (Dry Mell) Radiation Drywell Pressure (Containment Pressure)

Suppression Pool Temperature Primary System Pressure Primary System Level HSIV Position Turbine Stop Valve/Control Valve Position Turbine Bypass Valve Position Feedwater Flow Steam Flow Recirculation;

Flow, Pump Status Scram Discharge Level Condenser Vacuum AC and DC System Status (Bus Voltage)

Safety Infection; Flow, Pump/Valve Status Diesel Generator Status (On/Off, Start/Stop)

(1): Trip parameters.

(2): Parameter may be recorded by either an SOE or time history recorder.

(3): Acceptable recor der options are:

(a) system flow recorded on an SOE

recorder, (b) system flow recorded on a time history recorder, or (c) equipment status recorded on an SOE recorder.

(4): Includes recording of parameters for all applicable systems from the following:

HPCI, LPCI, LPCS, IC, RCIC.

3.

Evaluati on Additional information is needed before an adequate evaluation of the post-trip review data and information capabilities for the plant can be performed.

To date, little or no information has been provided in response to action item 1.2 of Generic Letter 83-28.

Any information provided by the licensee should addre'ss the evaluation criteria set forth in part 2 of this report.

The information should detail how the data and information capabilities at this nuclear power plant ful-fill the intent of the evaluation criteria.

If current capabilities do not meet the intent of the evaluation criteria, the licensee should either show that the data and information capabilities are sufficient to meet the intent of the evaluation criteria in part 2 of this report or detail future modifi-cations that will enable the licensee to meet these criteria.

REFERENCES NRC Generic Letter 83-28.

"Letter to all licensees of operating

reactors, applicants for operating license, and holders of construction permits regarding Required Actions Based on Generic Implications of Salem ATMS Events."

July 8, 1983.

NUREG-1000, Generic Implications of ATMS Events at the Salem Nuclear Power Plant, April 1983.

Letter from C.V. Mangan, Niagara Mohawk Power Corporation, to D.G.

Eisenhut, NRC, dated September 6, 1983, Accession Number 8309120389 requesting time extension for response to Generic Letter 83-28.

Letter from G.K.

Rhode, Niagara Mohawk Power Corporation, to A.

Schwencer, NRC, dated April 10, 1984, Accession Number 8404160070, submitting response to Generic Letter 83-28 for Nine Mile Point Unit 2, with attachment.

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07.749OZOO LOG NO.

SUBJECT:

NRC CORRESPONDENCE DISTRIBUTION LIST NINE MILE POINT UNIT 2 LETTERS FROM:

A. F. Zallnick, Jr.

TO:

DISTRIBUTION Explanation of Action to be Taken:

Distribution Made:

DISTRIBUTION:

R.

ABBOTT G.

AFFLERBACH J.

ASH W.

BAKER J.

BEBKO C.

BECKHAM W.

BRYANT J.

GALLAGHER G. GRIFFITH W.

HANSEN B.

HOOTEN J.

KROEHLER, JR.

By:

T.

LEMPGES J.

MACKENZIE C.

MANGAN C. MILLIAN T.

PERKINS J.

PERRY R.

PLANT D.

QUAMME M.

RAY (PSC-2 copies)

C.

STUART C.

TERRY J.

THOMAS D.

VANDEPUTTE J.

VOUGHT K.

WARD J.

WHEELOCK S.

WILCZEK, JR.

G.

WILSON CONNER 5 WETTERHAHN D. HILL(w/attachments)

PACKET'ttachments transmitted Attachments on file in Unit 2 Licensing Files (too bulky for transmission)

Letter Only w/transmittal lo/BS

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RA&

o7.14 NTER~AAL CORRESPONDENCEttop~ fl ~~ ~

FORM 11~2 A 0240 88-0t.013 INFO+III LI7 NIAGARA H v MOHAWK t'~OM A. F. Zallnick, Jr.

M. Drews DlsTRlcT Sys tern DATE September 27, 1985 FILE CODE SUBJECT Review of Technical Specifications In Accordance with Generic Letter 83-28 Niagara Mohawk's NMP2 has been requested to perform several reviews of the NMP2 Technical Specifications via Generic Letter 83-28, which concerns the "Required Actions Based on Generic Implications of Salem ATNS Events" (see attached).

Licensing requests that in preparation of the Technical Specifications, you perform these reviews.

Please make note that the review of the maintenance procedures will be performed by another group.

If you have any questions,.

please contact Mr. T. Loomis (X-6168) of my staff.

AF2/rla 0983G Attachment a

n>c Man

-Nuclear Licensing xc: R. Randall S. Nicolaous (NMPl Site)

K. Korcz Project File (2)

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10-3.1 POST-MAINTENANCE TESTING (REACTOR TRIP SYSTEM COMPONENTS)

Position The following actions are applicable to post-maintenance testing:

1.

Licensees and applicants shall submft the results of their review

, of test and maintenance procedures and Technical Specifications to

~~sp assure that post-maintenance operabtlity testing of safety-related components in the reactor trip system fs required to be conducted Q~'nd that the testing demonstrates that the equipment is capable of performi ng its safety functions before being returned to service.

C 3a g 5f

,gissi Licensees and applicants shall submit the results of their check of vendor and engineering, recommendatfons to ensure that any appropriate test guidance is included fn the test and maintenance procedures or the Technical Speci ficatfons, where requf red.

Licensees and applicants shall identify, if applicable, any post-maintenance test requirements fn existing Technical Specifications which can be demonstrated to 4fegrade rather than enhance safety.

Appropriate changes to these test requirements, with supporting justification, shall be submitted for staff approval.

(Note that action. 4.5 discusses on-line system functional testing.)

A plicabilit This action applies to all licensees and OL applicants.

~l'R For licensees, a post-implementation jevfew will be conducted for actions 3.1.1 and 3.1.2 above.

The Regions will perform these licensing reviews and issue Safety Evaluations.

Proposed Technical Specification changes. resulting from action 3.1.3 above will receive a pre-implementation review by NRR.

For OL applicants, the review will be performed consistent with the licensing schedule.

Oocumentation Re ufred Licensees and applicants should submit a statement confirming that actions 3.1.1 and 3.1.2 of the above position have been implemented.

Technical S ecfffcatfon Chan es Re ufred Changes to Technical Specfffcatfons, as a result of action 3.1.3, are to be determined by the licensee or applicant-and submitted for staff approval, as necessary.

Reference

,Section 2.3.4 of NUREG-1000.

11 3.2 POST-MAINTENANCE TESTENG (ALL OTHER SAFETY-RELATED COMPONENTS)

Position The following actions are applicable to post-maintenance testing:

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Licensees and applicants shall submit a report documenting the extending of test and maintenance procedures and Technical Specifications revi ew to assure that post-maintenance operability testing of all safety-related equipment is required to be conducted and that the 'testing demonstrates that the equipment is capable of performing its safety functions before being returned to service.

Licensees and applicants shall submit the results of their check of vendor. and engineering recommendations to ensure that any appropriate test guidance is included in the test and maintenance procedures or the Technical Specifications where required.

Licensees and applicants shall identify, if applicable, any post-maintenante test requiremehts in existing Technical Specifications which are perceived to degrade rather than enhance safety.

Appropriate changes to these test requirements, with supporting justification, shall be submitted for staff approval, A

1 icabi 1 it This action applies to all licensees and OL.applicants.

~Tf For licens'ees, a post-implementation review-will be conducted for actions 3.2.1 and 3.2;2 above.

The..Regions will perform these licensing reviews and issue Safety Evaluations.

Proposed Technical Specification changes resulting from action 3.2.3. above wi11 receive a pre-implementation review by NRR.

I For OL applicants, the review will be performed consistent with the licensing schedule.

e Documentation Re uired Licensees and applicants should submit a statement confirming that actions 3.2.l and 3;2.2 of the above position have been implemented.

Technical Specification Chan es Re ufred Changes to Technical Specifications, as a result of action 3.2.3, are to be determined by the licensee or applicant for staff approval, as necessary.

'eference.

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4. 5 REACTOR TRIP SYSTEM RELIABILITY (SYSTEM FUNCTIONAL TESTING)

Position hJo hip~

y potpie~

verse tr. p e

1.

The diverse trip features to be tested include-the breaker undervoltage and shunt trip features on Westinghouse, B&W (see h)o l<<g~

Action 4.3 above) and CE plants; the cfrcuftry used for power interruption with the silicon controlled rectfffers on B&W plants (see Action 4.4 above);

and the scram pilot valve and backup scram valves (including all initiating,circuitry) on GE plants.

2.

Plants not currently designed to permit periodic on-line testing shall justify not making modifications to pemait such testing.

Alternatives to on-line testing proposed by licensees will be i cuir acf considered where special circumstances exist and where the oh/ective of high relfabflfty can be met fn another way.

3.

Existing-intervals for on-line functional testing required by Technical Specifications shall be reviewed to determfne that the intervals are consistent with achieving high reactor" trip system availability whe'n accounting for considerations such Pd bn ipe,d 1.

uncertainties fn component failure rates qual 2.

uncertainty fn common mode failure rates 3.

reduced redundancy during testing gddl'+'

4.

operator errors dui ing testing

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caaponent "wear-out" caused by the testing Licensees currently not performfng periodic on-line testing shall

. determine appropriate test intervals:as described above.

Changes to existing required intervals for on-line testing as well as the'ntervals to be determined by licensees currently not performing on-line testing shall be justified by information on the sensitivity of reactor trip system availability to parameters such as the. test intervals, component failure rates, and common'mode failure rates.

A licablit This action applies to all licensees and OL applicants.

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For licensees, a post-implementation review will be conducted for action 4.5.1.

The Regions will perform tllese licensing reviews and Rfssue Safety Evaluations.

Actions 4.5.2 and 4.5.) will requfr e a pre-implemen-tation review.by NRR.

Results will be issued fn a Safety Evaluation.

17 For OL applicants, the NRR revfew should be performed consistent with the licensing schedule.

Oocumentatfon Re uired For item 4.5.1, licensees and applicants should submit a statement confirming that this action has been implemented.

For item 4.5.2, licensees and applicants should submit a report describing the modifications for staff review.

For item 4.5.3, lfcensees and applicants should submit proposed Technical Specification changes for staff review.

Technfcal S ecfffcation Chan es Re ufred For licensees, Technical Specification changes are required.

For OL applicants, Technical Specifications will be incorporated as part of the license.

Reference Section 3 of NUREG-1000.

(The staff finds that modifications are not required to permit on-T~ne te~nq.

of the backup scram valves.

However,. the staff concludes that testing of the

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backup scram valves (including fnftfatang cfrcuztrJJ.1 at a refueling outage

.frequency, in lieu of on-line testing, is appropriate a

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and should 8 inc u in the technical specification surveillance requirements.'he 1icensee needs to address this conclusion.)

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INTERNALCORRESPONDENCE Foa+M +12.2 R 0240 66%1-013 pygmy <ec L'I1 NIAGARA H O MOHAWK, FROM A. F. 2allnick, Jr.

H. Barrett DI$TRIQT Syr acus e DATE May 16, 1985 FILE GQDE

$UBJEQT Generic Letter 83-28 Attached is a tentative list of actions and identified individuals responsible for closing the Generic Letter 83-28 concerns.

Licensing requests that you review this list and provide comments and a schedule for closure of your items by May 24, 1985.

Oue to a recent request by the NRC Project Manager for a schedule for completion of the 83-28 concerns, your immediate attention is requested.

AFZ/TL:ja Atta chtren t Project File (2) a n>c,

Manager, Nuclear Licensing

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07,749Q j "RE VIREO ACTIONS BASED ON GENERIC IMPLICATIONS OF SALEM ATWS EVENTS 1.1 POST-TRIP REVIEW (PROGRAM DESCRIPTION AND PROCEDURE)

Position Licensee and applicants shall describe their progr am for ensuring that unscheduled reactor shutdowns are analyzed and that a determination is made that the plant can be restarted safety.

Action a.

Describe administrative controls relating to post-trip review and revise as necessary to meet the intent of the requirement.

g. Rachel A report describing the program for review and analysis of such unscheduled reactor shutdowns should include, as a minimum:

1.

The criteria for determining the acceptability of r estart.

Action Refer (again) to the above procedures and revise as necessary.

R RonQ 2.

The responsibilities and authorities of personnel who will perform the review and analysis of these events.

Action Refer to Administrative Procedures (NMPl uses AP1.2, Conduct of 0 erations an Com osition and Res onsibilities of Station or Unit Or anization and APl.l, om os>tron an es ons>bi itches o

i e r anszation ev>ew and revise th procedures as necessary.

3.

The necessary qualifications and training for the responsible personnel.

Action Refe~ to ANSI/ANS 3.1-1981 and describe the qualifications of the on-shift (responsible)"personnel.

Review and revise procedures as necessary.

4.

The sour ces of plant information necessary to conduct the review and analysis.

The sources of information should include the measures and equipment that provide the necessar y detail and type of information to reconstruct the event accurately and in sufficient detail for proper understanding.

(See Action 1.2)

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07/490 79 7;

1.

Capability for assessing sequence of events (on-off indications),

1)

Brief description of equipment (e.g., plant computer.,

dedicated computer, strip chart)

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2)

Parameters monitored 3)

Time discrimination between events 4)

Format for displaying data and information 5)

Capability for ~etention of data and information 6)

Power source(s)

(e.g.,

Class lE, non-Class lE, noninterruptible) 2.

Capability for assessing the time history of analog variables'eeded to determine the cause of unscheduled reactor shutdowns and the functioning of safety-related equipment.

1)

Brief description of equipment (e.g., plant computer, dedicated computer, strip charts) 2)

Parameters monitored, sampling rate and basis for selecting parameters and sampling rate 3)

Duration of time history (minutes before trip and minutes after trip) 4)

Format for displaying data including scale (readability) of time nistories.

5)

Capabi 1.ity for retention of data, information and physical evidence (both hardware and software) 6)

Power source(s)

(e.g.,

Class lE, non-Class lE, noninterruptible) 3.

Other data and information provided to assess the cause of unscheduled reactor shutdowns.

4.

Schedule for any planned changes to existing data and information capability.

Action The response should be in a description format and relatively brief.

The onl item which could cause a problem would be item 1.2.4.

See the NMPl response.

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': 2.1 EQUIPMENT CLASSIFICATION AND VENOOR INTERFACE (REACTOR TRIP SYSTEM COMPONENTS)

Position Licensees and applicants shall confirm that all components whose functioning dt 1

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1-1 1

1 procedures and information handling system used in the plant to control safety-related activities, including maintenance, work orders and parts replacement.

In addition, for these components, licensees and applicants shall establish, implement and maintain a continuing program to ensure that vendor information is complete, current and controlled throughout the life of the plant, and appropriately referenced or incorporated in plant instructions and procedures.

Vendors of-these components should be contacted and an interface established.

Where vendors cannot be identified, have gone out of

business, or will not supply the information, the licensee or applicant shall assure that sufficient attention is paid to equipment maintenance, replacement and repair to compensate for the lack of vendor backup to assure reacto~ trip system reliability.

The vendor interface program shall include periodic communication with vendors to assure that all applicable information has been received.

The program should use a system of positive feedback with vendors for mailings containing technical information.

This could be accomplished by licensee acknowledgement for receipt of technical mailings.

The program shall also define the interface and division of responsibilities among the licensees and the nuclear and nonnuclear divisions of their vendors that provide service on reactor trip system components to assure that requisite control of and applicable instructions for maintenance work are provided.

Documentation Requi~ed Licensees and applicants should submit a statement confirming that they have reviewed the Reactor Trip System components and confor m to the position regarding equipment classification.

In addition, a summary report describing the vendor interface program shall be submitted for staff review.

Vendor lists of technical information, and the technical information itself, shall b

available for inspection at each reactor site.

Actions a.

Review maintenance procedures to ensure that RTS components in various systems are identified as SR (Unit 1 action).

b.

Review g-List to ensure that all RTS components are identified as SR (Uni D g 1 action).

c.

Ensure that work requests and other "work assigning" documents contain classification information.

Identify other, documents and parties that identify correct classification.

d.

Review all SILs, SALs, TILs, PERs to ensure that they have been incorporated into the plant equipment.

In addition, review Bulletins, Circulars, Notices, Significant Event Reports, Significant Operating Expel ience Reports to determine their effect on the RTS.

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~ e.

Ensure "technical manuals" are up to date and complete.

P&nJw>i f.

Review maintenance procedures to ensure that the "control, copy" tech'nical g

manual is referenced or incorporated in the maintenance'procedures..

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g.

Evaluate General Electric's,method of classifying equipment and revise engineering procedures, if appropriate.

h.

Obtain information on General Electric's information programs.

This information can be submitted to the NRC to demonstrate our review of the programs.

i.

Evaluate BWR Owners Group options pertaining to tne reacto~ trip system.

(The Unit 1 conclusion can be duplicated in our response.)

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~gomez 2.2 E(UIPMENT CLASSIFICATION ANO VENOOR INTERFACE (PROGRAMS FOR ALL SAFETY-RELATEO COMPONENTS)

Position Licensees and applicants shall submit, for staff review, a description of their programs for safety-related equipment classification and vendor interface as described below:

1.

For equipment classification, licensees and applicants shall describe their progr am for ensuring that all components of safety-related systems necessary for accomplishing required safety functions are identified as safety-related on documents, procedures and information handling systems used in the plant to control safety-related activities, including maintenance, work orders and replacement parts.

Actions Identify more administrative procedures for handling the g-List.

(This is in addition to our pr evious response.)

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1.

The criteria for identifying components as safety-related within systems currently classified as safety-r elated.

This shall not be interpreted to require changes in safety classification at the systems level.

Action In addition to what was previously submitted, submit administrative procedure e names.

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2.

A description of the information handling system used to identify safety-related components (e.g.,

computerized equipment list) and the methods used for its devel opment and val idati on.

/gOnnS g Ranks

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3.

A description of the process by which station personnel use this information handling system to determine that an ac'tivity is safety-related and what procedures for maintenance, surveillance, parts replacement and other activities defined in tne introduction to 10 CFR 50, Appendix B, apply to safety-related components.

Action Describe the process of determining if an activity is safety related.

This description should include the role of the shift super visor, gA department reviews and any administrative procedures governing this function (Unit 1

response).

4.

A description of the management controls utilized to verify that the procedures for prepar ation, validation and routine utilization of the information handling system have been followed.

Action Previous NMP2 response should suffice.

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5.

A demonstration that appropriate design verification and qualification testing is specified for procurement of safety-related components.

The specifications shall include qualification testing for expected safety service conditions and provide support for the licensees'eceipt of testing documentation to support the limits of life recommended by the supolier.

Action Describe engineering procedures which govern design control and design ver ification.

In addition, state the engineering procedure which contr ols procurement activities.

6.

Licensees and applicants need only to submit for staff review the equipment classification p~ogram for safety-related components.

Although not required to be submitted for staff review, your equipment, classification p~og~am should also include the broader class nf structures, systems and components important to safety required by GDC-1 (defined in 10 CFR Part 50, Appendix A, "General Design Criteria,

-- Iatroduction").

Action

Response

provided.

Ho additional information is needed.

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0 For vendor interface, licensees and applicants shall establish, implement and maintain a continuing program to ensure that vendor information for.

safety-related components is complete, current and controlled throughout'.

the life of their plants and appropriately referenced or incorpor ated in plant instructions and procedures.

Vendors of safety-related

-equiprhent should be contacted and an interface established.

Wher e vendor s cannot be identified, have gone out of business, or will not supply information, the licensee or applicant shall assure that sufficient attention is paid to equipment maintenance, replacement and repair to compensate for the lack of vendor backup to assure reliability commensurate with its safety function (GOC-1).

The program shall be closely coupled with action 2.2.1 above (equipment qualification).

The program shall include periodic

'communication with vendors to assure that all applicable information has been received.

The program should use a system of positive feedback with vendors for mailings containing technical information.

This could be accomplished by licensee acknowledgment for receipt of technical mailings.

It sha'll also define the interface and division of responsibilities among the licensee and the nuclear and nonnuclear divisions of their vendors that provide service on safety-related equipment to assure that requisite control of and applicable instructions for maintenance work on safety-related equipment are provided.

r Action The previous response submitted by NMP1 can be duplicated for NMP2; however, the response was rejected by the NRC.

Additional industry action is war< anted 3.1 POST-MAINTENANCE. TESTING (REACTOR TRIP SYSTEM COMPONENTS)

Position The following actions are applicable to post-maintenance testing:

Licensees and applicants shall submit the results of their review of test and maintenance procedures and Technical Specifications to assure that post-maintenance operability testing of safety-related components in the reactor trip system is required to be conducted and that the testing demonstrates that the equipment is capable of performing its safety functions before being returned to service.

Action

'a ~

b.

c ~

Review the Tech.

Specs.

with respect to post-maintenance testing for reactor trip components.

Review I5C Oepartment Procedures (7) with respect to post-maintenance testing for reactor trip components.

Review maintenance procedures with respect to post-maintenance testing fo reactor trip components.

2.

Licensees and applicants shall submit the results of their check of vendo and engineering recommendations to ensure that any appropriate test guidance is included in the test and maintenance procedures or the Technical Specifications, where required.

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0 Action a.

The Tech.

Specs. will be reviewed for the above concerns.

The

~ staff's new position is that backup scram valves should be.tested during a refueling outage.

o.

A response similar to Unit 1's should be prepared.

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IATERNALCORRESPONDENCE FOAM 112.2 A02M 66<1-013 0

FILE LOPE TO A. F. Zallnick, Jr.

Mr. C.

V. Mangan DI$TRIQT Syracus e

.DATE April 26, 1985 fILE CODE

$IjBJEST NMP2 Response to Gener ic Letter 83-28.

Attachment 1 is a March 19, 1985 letter to NMPC requesting additional information on our previous response to Generic Letter 83-28.

A response is requested by May 19, 1985.

In'review of this letter, many of the NRC questions are not applicable to the previous NMP2 response (Attachment 2).

In addition, information needed to respond to the questions that do pertain to NMP2 will not be gathered in time for the May date.

After conferring with another utility, it appears that some of the questions are "boiler-plate" questions forwarded to many utilities.

It is our recommendation that a short letter which states, in part, that a response to the concerns stated in Generic Letter 83-28 will be submitted prior to startup.

Regardless of the interim letter, it is obvious that a significant amount of work rust be completed by startup to avoid a licensing condition attached to the NMP2 license.

It is conceivable that failure to address many of the 83-28 concerns could be detrimental towards granting an operating license to NMP2.

Licensing requests that an individual in the NMP2 Operations Department be assigned the responsibility of addressing the concerns of 83-28 on an expeditious basis.

a nic r.

AFZ/TL:ja xc:

R. Randall T.

E.

Lempges R.

B. Abbott Project File (2)

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