nuclear regulatory commission7 niagara v mohawk niagaramohawkpower'corporation/300 erie...

7 NIAGARAV MOHAWK

NIAGARAMOHAWKPOWER'CORPORATION/300 ERIE BOULEVARDWEST, SYRACUSE, N.Y. 13202/TELEPHONE (315) 474-1511

Apri1 15, 1986(NMP2L 0687)

Ms. Elinor G. Adensam, DirectorBWR Project Directorate No. 3U.S. Nuclear Regulatory Commission7920 Norfolk AvenueWashington, DC 20555

Dear Ms. Adensam:

Re: Nine Mile Point Unit 2Docket No. 50-410

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Attached is the Niagara Mohawk Power Corporation's Nine Mile, Point Unit, 2response to Generic Letter 83-28 which concerns the "Required Act"ionsBased on Generic Implications of Salem ATWS Events.." This letter-,supersedes. the previous responses sent to the Nuclear RegulatoryCommission on April 10, 1984 (G. K. Rhode (NMPC) to A. Schwencer (NRC)),and on December "20; 1985 (T. E. Lempges to you). This response doe-s notcontradict the previous responses, but does ..augment previous„., statements.It is presented here -in total for our mutual convenience.

'a'<~aThis response'lso addresses the „ letter . from M. Maughey .(NRC) , to,

'.

G. Mooten (NMPC) dated March 20, 1986, concerning a request foradditional informati'on on this subject.

Very truly yours,v ~ ~

Vice PresidentNuclear Generation

Attachments "

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xc: R. A. Gramm,'RC Resident InspectorProject File (2')

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NIAGARAMOMAWKPOWER CORPORATION/300 ERIE BOULEVARDWEST. SYRACUSE. N.Y. 13202/TELEPHONE (31S) 4T4.1311

April 15, 1986(NMP2L 068T)

Ms. Elinor G. Adensam, DirectorBWR Project Directorate No. 3U.S. Nuclear Regulatory Commission7920 Norfolk AvenueWashington, DC 20555

Dear Ms. Adensam:

Re: Nine Mile Point Unit 2Docket No. 50-410

Attached is the Niagara Mohawk Power Corporation's Nine Mile Point Unit 2response to Generic Letter 83-28 which concerns the "Required ActionsBased on Generic, Implications of Salem ATWS Events." This lettersupersedes the previous responses sent to the Nuclear RegulatoryCommission on April 10, 1984 (G. K. Rhode (NMPC) to A. Schwencer (NRC)),and on December 20, 1985 (T. E. Lempges to you). This response „does notcontradict the previous responses, but does augmen't previous statements.It is presented here in total for our mutual convenience.

This response also addresses the letter from M. Haughey (NRC) toB. G. Hooten (NMPC) dated March 20, 1986, concerning a request foradditional information on this subject.

Very truly yours,

Vice PresidentNuclear Generation

Attachments

xc: R. A. Gramm, NRC Resident InspectorProject File (2)

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Preface,

Throughout ,this document are references to NMPC or SWEC procedures.Except as specified in the response, these procedures are a'ttached to thisletter (listed below) and are included to facilitate NRC's review of thisdocument. . These procedures are, and must be "living documents" that willundoubtedly be revised in the future. Their" inclusion here does notconstitute any commi,tment by NMPC or SWEC to maintain these proceduresverbatim as presented here. However, NMPC does commi*t to maintainingcompliance with the -intent of Generic Letter 83-28 as specified herein.

Enclosures ¹1 AP-1. 1--

Enclosures ¹2 AP-1.2

Enclosures ¹3~AP-1.3

Enclosures ¹4 AP-2- - .

Enclosures ¹5 ,AP-3;-4.1

Enclosures ¹6 AP-3.4.2

Enclosures ¹7 .AP-4.0




Enclosures ¹11 TDP-5

Composition and Respo'nsibili'ty of SiteOrganization

Composition and Responsibility of UnitOrganization

Personnel Responsibilities a'nd Authority

Production and Control of Procedures

Administration of Technical and Safety Reviews— Site Operations Review Committee

Operations Experience Assessment

Administration of Operations

Procedure for Repair

Procedure for Modification and Addition-Unit 2(Draft)

Manag'ement 'of Station Records

Administration of Operational EngineeringAssessment'tems




Enclosures ¹15 NTP-10

Nuclear Plant Reliability Data System (NPRDS)Failure Reporting

Post-Maintenance Testing Criteria

Independent Safety Engineering Group

Training & Licensed Operator Candidates

Enclosures ¹16 NTP-ll Licensed Operator Retraining

Enclosures ¹17 N2-IOP-101A

Enclosures ¹18 N2-RAP-6

Plant Startup

Post Reactor Scram Analysis and Evaluation

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Enclosures ¹19 SNEC Procedure Method. for Handling .-Supplier TechnicalPP-81 Documents

Enclosures ¹20 NEL-014.G

Enclosures ¹21 SNEC ProcedureC-3

Control and Distribution 'of Vendors'Documents

Equipment Identification Codes'.

Enclosures ¹22 SHEC Procedure Project Specification and ProcurementPP-3 Procedures

Enclosures ¹23 SNEC Procedure Review of changes and.their effect 'onPP-94 Environmental/Mechanical/Seismi'c/Hydrodnamic

Qualification of Equipment

Enclosures ¹24 '-MI-GEN-'002 Maintenance Instructions for Hri ting, Procedures

Enclosures ¹25 MI-4.0 Maintenance Instructions for'eview andImplementation of Technical Requirements inMaintenance Procedures

Enclosures ¹26 S-IDP-PO

Enclosures ¹27 EngineeringAssuranceProcedure 3.1

Outline for IEC Procedures

Verification of Nuclear Power Plant Designs

Enclosures ¹28 INPO Letter on NUTAC Recommended Enhancements'

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Section 1.1

Generic Letter 83-28

Post-Trip Review (Program Description and Procedure)

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RE UIRED ACTIONS BASED ON GENERIC IMPLICATIONS OF SALEM ATNS EVENTS

POST-TRIP REVIEH (PROGRAM DESCRIPTION AND PROCEDURE)

'ositionLicensees and applicants shall describe their program for ensuringthat unscheduled reactor shutdowns are. analyzed and that adetermination is made that the plant can be restarted safely. Areport describing the program for review and" analysis of suchunscheduled reactor shutdowns should include, as a minimum:

The criteria for determining the acceptability of restart.

NMP2 Res onse

Nine Mile Point Unit 2's criteria for determining the acceptabilityof restart are contained in Procedure N2-RAP-6, Post Reactor ScramAnal sis and Evaluation, and in (Interim) Operating Procedure

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prior to startup). N2-RAP-6 provides a revieW and evaluation ofspecific parameters associated with a Reactor Scram from all.operating conditions'f after the completion of this'rocedure,there is a condition which is not fully understood, The SiteOperations Review Committee (SORC) must review this report before theStation Superintendent can authorize a restart. In the operatingprocedure N2-IOP-101A, valve instrumentation, system and componentcheckoff sheets must be completed prior to reactor startup. Thesepre-startup checkoff sheets are used to ensure that all equipment,necessary for safe operation is operable in accordance with plantTechnical Specifications. This procedure also states that N2-RAP-6must be completed (following a Scram) prior to restart.

The Administrative Procedure which identifies the criteria that theStation Superintendent will use for determining the acceptability ofrestart is, AP-4, Administration of 0 erations. Section 7.4 of thisprocedure will be changed to state as follows:

7.4

7.4.1

7.4.2

The criteria in which the Station Superintendent willuse for determining the acceptability of restart,after an unscheduled shutdown, shall be as follows:

The plant is shown to be in a safe condition.

The cause of the event is either understood or SORChas reviewed and authorized a restart.

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7.4.3 The need for corrective action has been determined andappropriately implemented.

7.4.4 The expected automatic operation of plant safetyrelated systems has been verified.

Therefore, Unit 2 is currently in compliance with the intent ofSection 1.1.1.

The responsibilities and authorities of personnel who. iill performthe review and analysis of these events.

NMP2 Res onse

The Superintendent Operations, , Station Shift Supervisor,. ShiftTechnical Advisor, and Technical Department personnel will performthe Post-Trip Review analysis, Their duties are specifically statedin Administrative Procedures AP-1.2 and AP-1.3, Com osition andRes onsibilit of Unit Or anization and Personnel Res onsibilities

which is directly responsible for the completion of N2-RAP-6, PostReactor Scram Anal sis and Evaluation Procedure. , This procedure ,states specifically that "the Reactor Analyst Department will .bedirectly responsible for data gathering and process evaluation. Theanalysis will be completed by the Unit Reactor Analyst or SiteReactor Analyst. In the event that those individuals areunavailable, the analysis will be conducted by a senior member ofTechnical Services and/or operations". Their duties are alsosupported by Site Administrative Procedure AP-l.l, Com osition andRes onsibilit of Site Or anization.

Therefore, admi ni strati ve controls whichresponsibilities and authorities. of personnelPost-Trip Review meet the intent of Section 1.1.2.

regulate theevaluating the

The necessary qualifications and training for the responsible .

personnel.

The analysis of unscheduled shutdowns at Nine Mile Point Unit 2 willbe performed by a select group of trained and qualified individuals.,The individuals currently in the positions of SuperintendentOperations, Site Reactor Analyst, Unit Reactor Analyst, and the .

Station Shift Supervisors all have experience at the ,operatingfacilities at Nine Mile Point Unit 1 and/or James A. Fitzpatrick.The education, training, and job related experience qualify thes'epeople to make the Post-Trip Review and restart recommendation.

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The qualifications and training for the positions of StationSuperintendent, Superintendent Operations, Site Reactor AnalystSupervisor and Unit Reactor Analyst Supervisor comply with therequirements of ANSI/ANS 3.1-1978. Additionally, the current siteReactor Analyst Supervisor and the Unit Reactor Analyst Supervisorboth hold Senior Reactor Operator licenses at Unit 2. The StationShift Supervisors and Shift Technical Advisors also will meet thequalification requirements of ANSI/ANS 3.1-1978. The Shift TechnicalAdvisor meets the Commission's Policy Statement on engineeringexpertise described in 50FR43621.

p dLicensed 0 erator Candidates and NTP-11, Licensed 0 erator~Retratnln . These procedures formally establish the procedures,programs, responsibilities and requirements necessary for thequalifications of NRC Licensed Reactor Operators and Senior Operators .

at Nine Mile Point 2.

Therefore, the existing Nine Mile Point Unit 2 administrativecontrols currently meet the intent of Section 1.1.3.

The sources of plant information necessary to conduct the review andanalysis. The sources of information should include the measures andequipment that provide the necessary detail and type of informationto reconstruct the event accurately and in sufficient detail forproper understanding. (See Action 1.2)

NMP2 Res onse

Information necessary to conduct the review and analysis isavailable, to the responsible personnel, through a number ofdifferent sources. The main source of data will come from, N2-RAP-6,Post Reactor Scram Anal sis and Evaluation Procedure. This procedureis designed to evaluate system performance from an initiation orisolation standpoint. The determination of safety system initiation,proper flow paths and system operation will be done using post triplogs, control room instrumentation, recorders, alarms, indicatinglights, and the General Electric Transient Analysis Recording System-(GETARS), as well as the Unit 2 Process Computer System. Thesesystems provide Operators with essential plant performanceinformation through a variety of logs, trends, summaries,. and datadisplays. More information on these systems is provided in section1.2 (Post-Trip Review — Data and Information Capability).

The methods and criteria for comparing the event information withknown or expected plant behavior (e.g., that safety-related equipmentoperates as required by the Technical Specifications or otherperformance specifications related to the safety function).

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NMP2 Res onse

As stated in Section 1.1.3, the individuals responsible for the eventanalysis are qualified per ANSI/ANS 3.1-1978 and currently holdSenior Reactor Operator licenses (both Unit and Site ReactorAnalysts). At their disposal are records of previous reactor trips(when history records exist), Technical Specifications, Final SafetyAnalysis Report data, and reload licensing analyses which are used attheir discretion for comparing the transient to expected responses.

The criteria for determining the need for independent assessment ofan event (e.g., a case in which the cause of the event cannot bepositively identified, a competent group such as the Site OperationsReview Committee, will be consulted prior to authorizing restart) andguidelines on the preservation of physical evidence (both hardwareand software) to support independent analysis of the event.

NMP2 Res onse

Unit 2's criteria for determining the need for independent assessmentis contained in Reactor Analysis Procedure N2-RAP-6, Post ReactorScram Anal sis and Evaluation. This procedure specifically, states(on the Final Assessment Sheet) that "If there is a condition notfully understood, the Station Superintendent should be so notified,and the appropriate staff members called in to assist in theevaluation. If after further evaluation the scram is still notunderstood, SORC must review this report before authorization torestart". Also, AP-3.4.1, Administration of Technical and SafetReviews (SORC) states: "Scram reports need not be reviewed by SORCprior to restart unless the cause of the scram or'he plan't transientresponse is not fully understood. Under these conditions SORC willprovide the independent assessment per generic letter 83-28 Section1.1.6, and SORC approval is required prior to restart". Section1.1.1 (of this response) states specific criteria contained inAdministrative Procedure AP-4 which the Station Superintendent mustfollow prior to authorizing a restart.

Unit 2's procedure established to assure that all physical evidence(necessary for an independent assessment) is preserved is AP-10.1,Mana ement of Station Records. This procedure provides an outlinefor the collection, storage and maintenance of site records andtechnical information. This procedure states that all Scram Reportsand Scram Analysis data (N2-RAP-6) remain in plant archives for thelife of the plant. This enables operating personnel to compare eventinformation with known or expected plant behavior at any time.

Therefore, the Administrative Controls provide a systematic method todetermine the need for independent assessment and NMP2 meets theintent of Section 1.1.6.

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1.1.7 Our systematic safety assessment procedures which addressesSection 1.1 Post-Trip Review, are as follows:

Site Administrative Procedures

AP-1.1AP-1.2AP-1.3AP-3.4.1AP-4.0AP-10.1

Composition and Responsibility of Site OrganizationComposition and Responsibility of Unit OrganizationPersonnel Responsibilities and AuthorityAdministration of Technical and Safety Reviews (SORC)Administration of OperationsManagement of Station Records

Nuclear Trainin Procedures

NTP-10NTP-ll

Training of Licensed Operator CandidatesLicensed Operator Retraining

Reactor Anal st Procedure

N2-RAP-6 Post Reactor Scram Analysis and Evaluation

0 eratin Procedure

N2-IOP-101A Plant Startup

The administrative controls currently being implemented at Nine Mile-Point Unit 2 contain procedures and data collection requirementsrelated to Post-Trip Review. These requirements provide assurancethat the cause for unscheduled reactor shutdown is analyzed and adetermination made as to the cause prior to plant restart. Inaddition, the general response of safety related equipment isreviewed prior to plant restart.

Nine Mile Point Unit 2's Administrative Controls adequately addresses-Sections 1.1 on Post-Trip Review.

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Section 1.2


Post-Trip Review (Data and Information Capability)

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Section 1.2

Post-Tri Review — Data and Information Ca abilit

Unit 2's Computer equipment which is capable of recording, recalling anddisplaying data and information necessary to diagnose the cause of unscheduledreactor shutdowns, is comprised of three different systems: The ProcessComputer System, General Electric's Transient Analysis Recording System, andthe Safety Parameter Display System. Each system works independent of oneanother, but has many redundant data ID points which provide crucialinformation during a system failure.

The following three sections discuss each system in detail and anser thequestions generated in Generic Letter 83-28.

Section 1.2A


Post-Trip Review — Data and Information Capability

Process Computer System (PCS)

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Capability for assessing sequency of events (on-off indications).

Brief description of equipment (e.g., plant computer,'edicatedcomputer, strip chart).

NMP2 Res onse

The Process Computer installed at Nine Mile Point Unit 2 consists ofdual Honeywell 4500 C.P.U.'s with the General Electric ProcessManagement System (PMS) software package. Each processor contains128K word memory for core storage and dual ported Ampex large corestores for bulk devices. In addition, the system utilizes an 80MBdisk drive for additional storage capacity, and for back-upcapability. Two magnetic tape units are utilized for eitherhistorical recording retention or for back-up capabilities.

For peripherals, the computer room is equipped with two color graphicvideos, two input keyboards, two input/output terminets, one outputonly terminet, a cardreader, and a high speed line printer.

The control room is equipped with four color graphic videos, twoinput keyboards, one input/output terminet, two output terminets, sixtrend recorders, and five digital displays. Attachment A contains alist of all the main control room dedicated strip charts.

Additionally, the remote shutdown room is equipped with one colorgraphic video and one keyboard.

Parameters monitored.

NMP2 Res onse

Niagara Mohawk has reviewed the Technical Evaluation Report of .

October 18, 1985 (from N.R. Butler to B.G. Hooten) pertaining toPost-Trip Review Criteria. Nine Mile Point 2 has investigated itsSequence of Events and Historical Recording parameters and determinedthat Unit 2 adequately addresses the digital and analog parametersspecified in the report (Table 1.2-2).

Attachment 1 is a copy of all the sequence of event points that existon the system to date. There has been a considerable amount ofspares created so that points may be added in the future. Thesepoints reflect trip points associated with electrical breaker status,water levels, relief valve positions, IRM and APRM upscale levels,and the Neutron Monitoring System.

Time discrimination between events.

NMP2 Res onse

SOE (sequence of event) points are alarmed and recorded on anautomatic interrupt driven basis on a change of state. Temporalresolution is 4 milliseconds between events. Events occurring withinthis time period may not be recorded in sequence.

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NMP2 Res onse

Attachment 2 is a copy of the(SOE) log is printed out to arecording 64 contact changeschange. Each change of stateterminet in the Control Room:changed from 1-60 seconds.

format used when a sequence of eventterminet. The log will print afteror 30 seconds after first contactwill also be alarmed to the alarmThe time period to printout can be

Capability for retention of data and information.

NMP2 Res onse

Retention of all sequence of event (SOE) data is controlled by theHistorical Recording System. The HRR (Historical RecordingRetention) system will record all changes of state including SOEpoints. This data can then be retrieved at any time from either thedisk drive or from magnetic tape depending on the time frame.Attachment 3 contains the data format viewed by the user. Thedistance back in time a user may go depends. on the retention cycle ofmagnetic tapes used. The data can be printed to a terminet or viewedon the CRT screen.

Niagara Mohawk has revised Administrative Procedure AP-10.1,Mana ement of Station Records to meet the commitment of GenericLetter 83-28 that Scram Reports and Scram Analysis Data will remainon site (Plant Archives) for the life of the plant.

Power source(s) (e.g., Class lE, non-Class lE, noninterruptible).

NMP2 Res onse

Power to the Unit 2 Process Computer is provided by anUninterruptible Power Supply 2VBB-UPSlG Non-Class lE. This supply isfed from a 600V power panel 2VBB-PNL301, which is supplied by eitherthe Station Generator 13.8KV line, (2NJS-US3, during normaloperation) or from an off-site Scriba 115KV line (2NJS-US4, during ashutdown condition). The process computer is also supplied by analternate 600V bus 2NJS-US6. In the condition which all power islost, backup power is supplied by a 125V DC battery supply,2BYS-SNG001C.

In summary, upon loss of normal power, a static transfer switchtransfers power from the normal source to the alternative source. Ifboth normal and alternate sources are lost, the DC source willautomatically pickup the loads by means of a DC auctioneering circuit.Capability for assessing the time history of analog variables neededto determine the cause of unscheduled reactor shutdowns and thefunctioning of safety-related equipment;".

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

NMP2 Res onse

A brief description of the equipment comprising the Unit 2 ProcessComputer was given in Section 1.2.1.1.

In addition, the Post Trip data can be obtained in the Control Roomon two of three terminets, or in computer room on the line printer oron either teriminet.

Historical data may be obtained on either the color video or theterminet. Also, historical data can be obtained from the list ofdedicated strip charts which are located in the control room (seeSection 1.2.1.1).

Parameters monitored, sampling rate and basis for selectingparameters and sampling rate.

NMP2 Res onse

The parameters monitored by the Process Monitoring System (PMS) arelocated on Attachments 4 "NSSS Post Trip Log" and 5 "BOP Post Triplog".

These NSSS points provide information to enable the system tocalculate and display or printout, a variety of nuclear system data

'rrays(LPRM readings, sensitivities, and calibration constants; APRMgain adjustment factors and trip levels; control rod positions; fuelbundle isotopic compositions,

etc'�

).

The BOP points provide data to enable the system to performcalculations, evaluations of the status and efficiency of variousplant systems not directly related to the nuclear steam supply. Thecalculations include turbine cycle performance, condenserperformance, unit electric performance, and feedwater heaterperformance.

Selected Nuclear Steam Supply System and Balance of Plant digitalsignals are scanned once each second for the purposes of monitoringprocess variable alarms. The sampling rate for the analog variablesare in the process of being reviewed to determine what scan ratewould be most effective. They can presently be varied (to scan)every 1, 5, 15, 30, or 60 seconds. The system is capable of scanning100 points per second. Each time an input is scanned, it is comparedto its previous state and if it is different, the program willdetermine the nature of the change, (e.g., alarm or return-to-normal)and a descriptive message wi 11 be logged.

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1.2.2.3 Duration of time history (minutes before trip and minutes after trip).NMP2 Res onse

Two Post-Trip Logs are used at Nine Mile Unit 2. The first log is anaccumulation of points associated with the Nuclear Steam SupplySystem, and the second is an accumulation of Balance of Plantpoints. The NSSS log data interval is undergoing software changesand will be completed prior to startup. These changes will allow thelog to record data 5 minutes before a trip until 10 minutes after theevent. The BOP log is made up of a maximum of 48 predeterminedpoints. The data collection period ranges from 30 minutes before thetrip until 30 minutes after the event. Each point is scanned every15 seconds. Both logs will be initiated upon completion of theirrecording constraints. Recording and scan rate times are changeablevia software change routines to allow plant operations to vary theprocess monitoring function.

1.2.2.4 Format for displaying data including scale (readability) of timehistories.

NMP2 Res onse

Attachment 6 is a representation of the NSSS and BOP Post-Trip logs.This attachment is self-explanatory as to the data that is containedon these logs.

1.2.2.5 Capability for retention of data, information and physical evidence(both hardware and software).

NMP2 Res onse

Post-Trip logs can be recovered in. the same manner as discussed inSection 1.2.1.5. The only difference being that the logs can berecovered and reprinted exactly as the original log. Post-Trip logscan be demanded at a later time if no other event has generatedanother new Post-Trip log to overlay existing data.

As stated in Section 1.2.1.5, AP-10.1 commits Unit 2 to maintain hardcopies of Scram Reports and Scram Analysis Data for the life of theplant.

1.2.2.6 Power source(s) (e.g., Class lE, non-Class lE, noninterruptible).

Power sources are the same sources discussed in Section 1.2.1.6.

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

Other data and information available to assess the cause ofunscheduled reactor shutdowns include operator logs, trend recorders,

meter indications, surveillance test data sheets, seismic recordingequipment, operator interviews, and occurrence reports. Othercomputer systems available to assist in the evaluation of unscheduled.shutdowns are the Safety Parameter Display System (Attachment 1.2 B),and the General Electric's Transient Analysis Recording System(Attachment 1.2 C). In addition, previous scram report data andinformation is available at the operator's disposal enabling them tocompare event information with known or expected plant behavior.

Schedule for any planned changes to existing data and informationcapability.

NMP2 Res onse

The SOE printout wil,l be changed from 30 to 5 seconds to be moreconsistent with timelines of plant data for operator response. Also,points will be added to SOE 5 Alarm displays as required for moreeffective plant operations.

In addition, the duration of time history associated with the NSSSlog is undergoing software changes so that it will be capable ofrecording data 5 minutes before a trip until 10 minutes after theevent.

ATTACEKNT A (to Section 1.2)

Main Control Room Pen Recorders (stri charts)

Reactor Vessel Level-Fuel ZonePost Accident Monitor, channels (A5B) Rx level, Rx PressureReactor Water Cleanup F/D Inc. Conductivity 5 Oxygen sampleService Water/RHR TemperatureRECIRC Pumps Suction TemperatureTotal RECIRC FlowReactor Pressure; Turbine Steam FlowCore Pressure Drop; Total FlowReactor Steam Flow; Feedwater FlowReactor Water LevelCondensate Demineralized Conductivity in/out 6 oxygen outInlet Conductivity High; out Conductivity High6th Point Heater Outlet Conductivity 6 Oxygen/PGGenerator Turbine Component PositionCore MonitoringBearing Metal TemperaturesTurbine TemperaturesTurbine VibrationBearing Drain 5 Thrust Bearing TemperaturesCRD Pump Discharge Conductivity and OxygenIRM/APRM Recorders (4 units)SRM channel (records two of 4 channels)Main Steam Reheater Reheating Steam Supply Temperature 1A 5 1BCirculating Water System Return Water Conductivity 5 PHMain Generator Frequency34SKV Line Main Generator Volts

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COMPUTER POINTS

For NSSS Post Trip Log

FWSLA'101RCSFB01FWSB01FWSFA103FWSFA104NSSPB01FWSPA101CNMFB02CNMFB01RCSTA103RCSTA105CMSPA01CMSTA01CMSPA02CMSPA04CMSTA07CMSAA02MSSBC20CWSTB10SWPTA53SWPTA74CNSLA03TMLPA02CNMPA02

Reactor Water LevelRecirc Flow TotalFeedwater Flow TotalMain Steam Flow TotalMain Turbine Steam FlowReactor PressureSteam Dome PressureCond Booster FlowCond Pump FlowRecirc Pump Suet Temp ARecirc Pump Suet Temp B

Drywell PressuresDrywell TempDrywell High Range PressureSuppression Pool PressureSuppression Pool TempDrywell OxygenBypass Valve PositionAvg Cond Temp RiseService Water Inlet 'FService Water Disch 'FHotwell LevelTurbing Big Oil PressCond Vac

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COMPUTER POINTS

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NMP2A273NMP2A274NMP2A275NMP2A276NMP2A277NMP2A278CNMPA04

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APRM C

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APl.((H.HIT 6 (to section 1.2)

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Attachment 1.2 6


Post-Trip Review - Data and Information Capability

Safety Parameter Display System (SPDS)

Capability for assessing the time history of anglo variables neededto determine the cause of unscheduled reactor shutdowns, and thefunctioning of safety-related equipment.

Brief description of equipment (e.g., - plant computer, dedicatedcomputer, strip charts).

NMP2 Res onse

The Nine Mile Point 2 Liquid Radwaste System and Emergency ResponseFacility computer system (LWCS/ERF) consists of dual Honeywell 4500CPU's on which the standard Honeywell SEER software package has beenimplemented and modified as needed. Each processor contains 25K ofcore memory and dual ported Ampex large core stores used for bulkdevices. The system also utilizes two 80MB disk drives foradditional storage capacity. A magnetic tape unit is used forhistorical recording and additional back-up capabilities. Videomonitors, types/printers and keyboards are located in the computerroom, the control room, the Technical Support Center, and theEmergency Operations Facility to enable operators in recording andviewing event data.

1.2.2.2 Parameters monitored, sampling rate, and basis for selectingparameters and sampling rate.

NMP2 Res onse

There can be up to 690 data points contained on the Event HistoricalRecording System. These points can be contained in one of 115 groupsmade up of 6 data points each. Depending on which group the datapoint is contained in. The sampling rate can be every 1, 5, or 30seconds. The only exception to these rates is with the two week postevent recording. This data is collected every 15 minutes.

1.2.2.3 Duration of time history (minutes before trip and minutes after trip).NMP2 Res onse

There are three types of historical event recording on the LWCS/ERFcomputer system. Two hours of pre-event data is collected on acontinuous basis in a circular buffer. When a pre-defined event isdetected, the buffer is frozen. Twelve hours of post-event data iscollected immediately following the occurrence of a predefinedevent. This is done by the use of two 1 hour buffers which are usedin a switching process for 12 hours. Two weeks of additional datawill be collected immediately after the 12 hour collection iscompleted.

Format for displaying data including scale (readabi 1 i ty) of timehistories.

NMP2 Res onse

The trend history will consist of five secondary displays which arereactivity control, core cooling, coolant system integrity,containment integrity, (Attachments A-D) and a radioactive releasedisplay (future). Each secondary display consists of a number oftrend plots covering a 6-minute time span. The reactivity controldisplay consists of trend plots of APRM power, IRM power, and SRM logcount rate. The core cooling display consists of a trend plot of RPVwater level. The coolant system integrity display consists of trendplots of RPV pressure and drywell pressure. The containmentintegrity display consists of trend plots for drywell pressure,drywell oxygen concentration, suppression pool temperature, andsuppression pool water level, as well as the containment isolationvalve groups, The radioactive release display is a composite ofstack, off-gas and containment rad monitor parameters. Although thefinal design is not complete, we anticipate that the radioactiverelease display will be a composite of off-gas and containment radmonitor parameters.

All displays contain safety function blocks at the bottom, which maybe green or red depending upon whether the function is considered"normal" or "in alarm/unknown." The color of the safety functions isdetermined by the status of the variables associated with thosesafety functions.

Attachments (E-H) are the formats used for viewing the EventHistorical data. This data can be accessed by a display or aprinter. The operator can view the data based on time for eachsample taken. This display/printout can be based on any of thegroups and ranged over all or any of the time period of the eventrecording. The operator can also display a trend of the variousgroups. This trend can also be based by group and consist of dataover a specified period of the data recording.

Capability for retention of data, information, and physical evidence(both hardware and software).

NMP2 Res onse

Primary retention of data is done on disk buffers for all three typesof historical recording. Two buffers are used for the pre-eventrecording, which is able to hold 1 hour of data. Upon the detectionof an event, the pre-event buffers will be frozen and can be saved tomagnetic tape by a programmer. The 12 hour post-event data iscollected with the use of two, 1 hour buffers on disk. These buffersare used in a switching mode, and are dumped to magnetic tape whenthey become full. The 14 day post event collection also uses twobuffers on the disk. These buffers are dumped to magnetic tapedaily, by the programmer. The programmer has the option of dumpingthe data to tape manually or initiating an automatic mode. The datacan be restored from tape to a review buffer on the disk to allow aprogrammer to view or trend the data.

1.2.2.6 Power source(s) (e.g., Class IE, non-Class IE, non-interruptable).

NHP2 Res onse

Power to the Unit 2 LWCS/ERF computer is provided by anuninterruptible power supply 2VBB-UPSlA non class lE. This supply isfed from a 600V power panel 2VBB-PNL301, which is supplied by eitherthe station generator 13.8kv line (2WJS-US3, during normal operation)or from an off-site Scriba 115kv line (2NJS-US4, during a shutdowncondition). Back-up power is supplied by a 125v DC battery supply,2BYS-SWG001A.

In summary, upon loss of normal power, a static transfer switchtransfers power from the normal source to the alternative source. Ifboth normal and alternate sources are lost, the DC source willautomatically pickup the loads by means of a DC auctioneering circuit.

ATt r A (tp 1.2 B)

NHP-2

HOOE ENTER

4227

1254PRH

REACT I V I TY CONTROL04TE: 03:18:86TINK: 00: 00: 90

/ 0

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TIME NMP2Ui00 NMP2U181 NMP2U182 NMP2U103%P IlR /PWR %PWR /PWR

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56. 668

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Attachment 1.2 C


Post-Trip Review — Data and Information Capability

'eneral Electric's Transient Analysis Recording System

GETARS — 1

Capability for assessing sequency of events (on-off indications).

Brief description of equipment (e.g., plant computer, dedicatedcomputer, strip chart).

NMP2 Res onse

The General Electric Transient Analysis Recording System (GETARS-1)is a high-speed data acquisition system developed for startupoperations but is a permanent plant system. The system operates on aHewlett-Packard 2117F computer system. The processor 'contains 128Kwords of high-speed memory, dual-channel direct memory access, and adynamic mapping system. The system utilizes a Hewlett-Packard 7920moving head disk, which has a capability of 50 mbytes for program anddata storage. A Hewlett-Packard 7970E magnetic tape drive is usedfor historical recording. The system also utilizes the ValidyneHD310 Expanded Multiplexer system as the analog to digitalconverter. This system can contain up to 4096 analog inputs.

Peripherals contained on the system include the Versatec V80printer/plotter, and one HP2645A black and white video display.

The operating system uses two sets of supervisory software. Therealtime executive system is the RTE-IVB. This system executes alldata entry programs, data reduction programs, and utility programs.A lower overhead executive called RTEM is used to permit interfacingbetween peripheral devices and (control for high-speed data)acquisition programs.

These are also 21 permanent remote multiplexes used for analogscanning.

1.2.1.2 Parameters monitored.

The GETARS system presently contains approximately 500 Analog points.Attachment 1 contains a list of the systems and the ID points whichare monitored by the GETARS system.

1.2.1.3 Time discrimination between events.

NMP2 Res onse

The remote multiplexers (MC370AD) each contain 32 analog channels.The scan rates range from 23,810 scans/second when monitoring onechannel, to 2,100 scans/second when monitoring all 32 channels. In areal time environment, groups may not be scanned more rapidly thanonce each millisecond i.e., 1,000 samples per second.

Format for displaying data and information.

NMP2 Res onse

There are a number of functions contained on the GETARS system thatproduce various reports and plots.

The control rod timing function indentifies the status of eachcontrol rod and evaluates control rod scram performance against testtime criteria. Attachments 2 through 7 provide format examples forthe reports generated by this function.

The off-line Print/Plot program provides on-site verification andanalysis of data recorded by the data acquisition system.Attachments 8 and 9 represent the format associated with the printerand the plotter.

The dynamic noise frequency analysis function is a time seriesanalysis package which allows time history data to be analyzed in thefrequency domain. Attachment 10 provides this function's outputformat.

The histogram function provides a display of signal data in eitherengineering units, millivolts, or engineering units. Attachments 11& 12 provides a sample of this function's output.

The Run analysis function provides a statistical analysis. of a givendata acquisition Run. A sample format is provided on Attachment 13.

Capability for retention of data and information.

NMP2 Res onse

Data retention for the GETARS system is contained on either the DiskDrive or Magnetic Tape. System utilities are available on thissystem to save the data to or from tape. The data acquisition systemautomatically writes analog data to the disk.

Power source(s) (e.g., Class 1E, non-Class lE, noninterruptible).

NMP2 Res onse

Power to the General Electric Transient Analysis Response System(GETARS) is supplied by an Uninterruptible Power Supply 2VBB-UPSlGNon-Class 1E. This supply is fed from a 600V power panel2VBB-PNL301, which is supplied by one of two sources, either theStation Generator 13.8KV line (2NJS-US3, during normal operation) orfrom an off-site Scriba 115'ine (2NJS-US4, during a shutdowncondition). The GETARS system is also supplied by an alternate 600VBUS 2NJS-US6. In a condition by which all power is lost, backuppower is supplied by a 125V DC battery supply 2BYS-SWG001C.

1,2,1.6 t ')In summary, upon loss of normal power, a static transfer switchtransfers power from the normal source to the alternative source. Ifboth normal and alternate sources are lost, the DC source willautomatically pickup the loads (by means of a DC auctioneeringcircuit) and supply power panel 2VBS-PNLC102 which feeds GETARS.

Capability for assessing the time history of analog variables neededto determine the cause of unscheduled reactor shutdowns and thefunctioning of safety-related equipment.

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

NMP2 Res onse

A description of the equipment making up the GETARS system isprovided in the response for Section 1.2.1.1.

1 ~ 2.2.2 Parameters monitored, sampling rate and basis for selectingparameters and sampling rate.

NMP2 Res onse

All system inputs contained in the systems described on Attachment 1

are continually being monitored. As stated in Section 1.2.1.3, theabsolute maximum recording speed is 1,000 samples per channel.

1,2,2.3 Duration of time history (minutes before trip and minutes after trip).NMP2 Res onse

Upon a trip condition, data recording continues until the disk dataarea becomes full or the operator terminates the data recording.This disk area will hold a maximum of ll minutes of data of whichone-sixth is pre-trip data.

1.2.2.4 Format for displaying data including scale (readability) of timehistories.

NMP2 Res onse

Description of the formats for displaying the recorded data iscontained in Section 1.2.1.4.

1.2.2.5 Capability for retention of data, informatin and physical evidence(both hardware and software).

NMP2 Res onse

Description of the capability for retention of data is contained inSection 1.2.1.5.

-3-

1.2.2.6 Power source(s) (e.g., Class 1E, non-Class lE, noninterruptible).

NMP2 Res onse

A description of the power sources is contained in Section 1.2.1.6.

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

Schedule for any planned changes to existing data and informationcapability.

NHP2 Res onse

See section 1.2.4.A.

Attachment 1 (to Section 1.2 C)

I

Unit 2's.GETAR's System (parameters Monitored) and ID Points

Main Steam

Steam Line FlowMain Steam Header PressureMain Steam Line IsolationMSIV Position

RPS

Manual Reactor ScramAuto Reactor Scram

Rx Instrumentation

Rx Dome PressureRx Water LevelRx Core Plate DPRx Bottom Head Drain TemperatureRx Vessel Level (WR)

Neutron Monitorin

APRM'sLPRM'sThermal Heat FluxFlow-Biased Thermal Upscale Trip Setpoin

Residual Heat Removal

RHR Hx LevelRHR Sys FlowRHR Hx Lvl Cont OutputRHR Hx PressureRHR Pump Trp Brkr.'osn.CRD System PlowSelectable CRD Position

Attachment 1 (Cont'd) (to Section 1.2 C)

Recirculation S stem

Recirc Loop FlowRecirc Loop Flow Control Valve PositionRecirc Pmp Trip BkrRecirc Master Controller OutputLoad Demand ErrorRecirc Sys Flux ErrorRecirc Sys Flux Estimator OutputRecirc Loop Suction TempRecirc Pump D/PJet Pump Double Tap D/PJet Pump Flow LoopsTotal Core FlowLFMG Drive Motor BkrRecirc Flow Control Funct Generator Inputs

RCIC

RCIC InitiationRCIC Suction Pressure Controller OutputRCIC Elbow Tap D/PRCIC Turbine SpdRCIC Flow

Feedwater

Fdwtr Line TempFdwtr FlowStm/Fdwtr Flow MismatchFdwtr Master Controller OutputFdwtr Pump Suction PressureFdwtr Pump Disch PressureFdwtr Pump Byp Low Flow Control Valve PosFdwtr Pump TripFdwtr Low Flow Valve PosFdwtr High Flow Control Valve PosFdwtr Recirc Valve PosHigh Flow Funct Generator OutputLow Flow Master Controller Level SetpointSTep Generator Output

HPCS/LPCS

HPCS Pump TripHPCS InitiationHPCS Discharge FlowHPCS Discharge PressureHPCS Diesel Generator Bkr TripHPCS MCC Feeder Bkr TripLPCS Injection Valve Pos

Attachment 1 (Cont'd) (to Section 1.2 C)

Control Rod Drive

CRD Flow Controller OutputRCIC Trip/Throttle Valve Pos

Safet Relief Valve

SRV'sADS Initiation

BOP/Emer enc Bus Breaker

Normal Brkr PosAlternate Brkr PosDiesel Brkr PosMCC Feeder Breaker Pos

BOP (Turbine/Generator)

Main Turbine SpeedAuto Load FollowingLoad Reference OutputMain Turbine tripTransient Auto Pressure SetpointEHC Pressure SetpointPower/Load UnbalanceStop Valve PosBypass Valve PosMain Turbine Stm FlowMain Generator MWE

Grid VoltageGrid FrequencyRCIC Ramp Gen Signal Converter OuputRCIC EGM OutputRCIC Steam Control (Governor) Valve PosTotal Byp Valve Posn

BOP (Condenser Extract Stm Service Water FW Heater

Cond Bstr Pmp Disch Hdr PressService Water Pump TripsSpent Fuel Pool Cooling Pump TripService Water Pump TripMain Condenser VacuumCond Pump Disch Hdr PressHtr Drn Pmp Disch PressLp Htr StringsA&B Isolation ValvesBypass Rx FWP BypassESS LP/HP Htr StringsWarming ValvesMain Gen Bkr PosPress Reg OutputTotal Cont Valve Pos

NTAQ{fBf2 (to Section 1.2 C)

GETARS-I SAMPLE INPUT AND OUTPUTCRD - CONTROI. ffOD TIMING

Page 4"709 Aug 84

4.2.3 Sample Output - Channel And Subchannel Number

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ATTACHI'ENT 3 (to Section 1.2 C)

GETARS-I SAMPLE INPUT AND OUTPUTCRD - CONTRO'OD TiMING

~ 3

Sample Output Con't - Channel and Subchinnel Number

Page 4-809 Aug 84

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ATT4(@Pl[4 '(to Section 1.2 C):

G- TARS- I SAMPLE I NPiJT ANO OUTPUTCRG - CohTsROL ROD Tlb'slNG

Page 4-909 Aug 84

4.2.4 Sample Outpu: - Roc! Timming

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(to Section l-.2 C)

GETARS I S Af'iPLE I NPUT AND OUTPUTCRO - CONTROL ROD TII'ING

Page 4-II09 Aug 84

4.2.5 Sample Input - CRDSV~ ~

iRU.CROSVLIST DEVICE LUTTHIS RROCRAh VILL HECK SCRAN DATES ACAINST SCRANSURVEILLANCE REQUIREHENTS *HO VILL PRINTOUT THE RESULTSTO HELf THE USER IN IOENTIFYINC RODS DUE FOR TESTINC,INRUT F II.E NANET

4.2.6 Sample Output - CRDSV

CRD'SV OUTtVT lzi27 AN TH4. 26 JUNE> 1964

A' RODS HAVE SEEh SCRAH TESTED VITHIH 1294 DAYS.

SURilEILLAHCE IHTERVALS HAVE SEEH EX EEDED

THE FOLLOVINC RODS HAVE SCRAll TINE DATES LESS THAN 128 OA~S

DATE ROD COORDINATE STATUS

THE FOLLOVIHC 28K OF THE R3CS HAVE THE'LDEST SCRAN TEST DATES

DATE12/12/881/le/814 "/2 8/ 8 1

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Page 4-1409 Aug 84

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Page 4-13Oc Aug 84

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Page 4-2209 Aug 84

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Section 2.1 2.2

Generic l.etter 83-28

Eguipment Classification and Vendor Interface

EQUIPMENT CLASSIFICATION AND VENDOR INTERFACE (REACTOR TRIP SYSTEMCOMPONENTS)

Position

Licensees and applicants shall confirm that all components whosefunctioning is required to trip the reactor are identified assafety-related on documents, procedures and information handlingsystems used in the plant to control safety-related activities,including maintenance, work orders and parts replacement. Inaddition, licensees and applicants shall establish, implement andmaintain a continuing program .to ensure that vendor information iscomplete, current and controlled throughout the life of the plant,and appropriately referenced or incorporated in plant instructionsand procedures. Vendors of these components should be contacted andan interface established. Hhere vendors cannot be identified, havegone out of business, or will not supply the information, thelicensee or applicant shall assure that sufficient attention is paidto equipment maintenance, replacement, and repair to compensate forthe lack of vendor backup and to assure reactor trip systemreliability. The vendor interface program shall include periodiccommunication with vendors to assure that all applicable informationhas been received. The program should use a system of positivefeedback with vendors for mailings containing technical information.This could be accomplished by licensee acknowledgement for receipt oftechnical mailings. The program shall also define the interface anddivision of responsibilities among the licensees and the nuclear andnon-nuclear divisions of their vendors that provide service onreactor trip system components to assure that requisite control of,and applicable instructions for maintenance work are provided.

NMP2 Res onse

Niagara Mohawk does not currently plan to develop a specific list ofcomponents that would comprise a reactor trip system. The reactortrip function is accomplished at Nine Mile Point Unit 2 by utilizingredundant plant process instrumentation that input to aone-out-of-two taken twice logic system. These signals initiate areactor trip (rapid control rod insertion i.e. scram) by deenergizingsolenoid operated scram pilot valves that vent air from the reactorscram valves.

The components that contribute to the reactor trip function arecontained in several systems rather than one reactor trip system.Those systems whose components contribute to the reactor tripfunction include the reactor protection system, reactor vesselinstrumentation system, neutron monitoring system and control roddrive system. Therefore, a new system identified as the reactor tripsystem would cause unnecessary inconsistencies with existing NineMile Point Unit 2 system nomenclature. This would require extensiverevision to existing documentation and train.ing program with noenhancement of safety.

(Cont'd)

However, a task is currently underway to upgrade the details of ourequipment classification list (Q-List, See Response 2.2.1.2). Thiswill provide additional assurance that those components whichcontribute to the reactor trip function are appropriately classifiedas safety-related.

Administrative controls consisting of documents, procedures andinformation handling systems are used in the station to controlsafety-related activities including maintenance, work requests (workorders), parts replacements and modifications.

The work request form (AP-5, Page 15) contains the classificationinformation, which is derived from the equipment classification list(Q-List) by the work request originator or the approving supervisor.A Quality Assurance representative checks the classification againusing the equipment classification list (Q-List) (AP-5, Page 6 and 7).

Maintenance procedures are in the process of being reviewed to assurethat any classification information is correct. The review ofMaintenance Department maintenance procedures is complete. Thereview of IKC Department maintenance procedures is ongoing and willbe completed prior to startup.

Nine Mile Point Unit 2 has an ongoing program to ensure that vendorinformation is complete, current and controlled throughout the lifeof the plant, and appropriately referenced in procedures. Thisprogram is conducted in three parts. The first part is the AP-3.4.2,0 erations Ex erience Assessment program which receives, reviews andacts on applicable information from the reactor trip system supplierfor Nine Mile Point Unit 2. The information consists of GeneralElectric Service Information Letters (SILs) which the IndependentSafety Engine'ering Group (ISEG) receives and reviews to determineapplicability. These documents provide recommendations for equipmentmodification, plant design improvements or changes to procedures toimprove plant performance. They are distributed through the GEDomestic Apparatus and Engineering Service Operations (DAESO) or GENuclear Services Operation Regional Offices and are normally followedup by discussion during periodic service plan conferences. TDP-5,Administration of 0 erational En ineerin Assessment Items, providesguidance to the ISEG for the handling of OEA items to assure completeand accurate closeout of potential operating problems.

In summary, Niagara Mohawk receives SILs from General Electric. TheIndependent Safety Engineering Group investigates each one todetermine its applicability to the plant and incorporates accordingly(via, Operations Experience Assessment Program). A response form isthen completed (Attachment A) and returned to General Electric whereit's logged in and recorded. A copy of the GE SIL Log is availablethrough GE for plant review. This Log enables plants to review theSILs that have been transmitted, and act on any they have eithermissed or haven't received'his program provides an open line ofcommunication between the reactor trip system vendor and NiagaraMohawk, hence improving relations between the two.

(Cont'd)

In addition, the Operations Assessment Program addresses informationfrom the Nuclear Regulatory Commission (NRC) such as I&E Notices,Circulars and Bulletins, as well as information from the Institute ofNuclear Power Operations (INPO) such as Significant Event Reports andSignificant Operating Experience Reports. Collectively, thesesources of information provide a comprehensive and timely mechanismto assure that information pertaining to problems with safety-relatedequipment are identified and corrected.

Niagara Mohawk currently participates in the General ElectricOperations Engineer (OE) Program. This results in a GE seniorengineer being assigned on a resident basis to the Nine Mile PointStation. This individual is SRO Certified by GE and has a companysenior engineer position. This resident engineer program providesNiagara Mohawk as well as GE with a number of benefits, such as:

l. Improve fuel performance through assistance with core managementand PCIOMR implementation.

2. Contribute to availability and capacity factor improvements.

3. Assist in general plant operations, such as maintenance andoperations.

4. Increase information flow between Niagara Mohawk and GeneralElectric.

S. Assist with interpretation of SILs, backfits, and othermodifications.

6. Provide operating plant data to GE to improve future designs,backfit designs and modification recommendations.

7. Provide access to GE technical expertise on an informal basis.

This engineer has a computerized communication system connecting allthe staffed sites within the U.S. Plant status, good practices,current plant concerns and expedited data requests are handled ontypically a 24 to 48 hour turn around.

GE also provides the site with a (Service Project Manager) companyrepresentative. This individual handles all commercial communciationbetween NMPC and GE. Through these particular programs, a high levelof communication, feedback and equipment performance improvement isachieved.

2.1 (Cont.d)

The second part of this program is the Administrative Control oftechnical manuals. The current method used to control the flow oftechnical information is Stone & Webster's Project Procedure PP-81,Method for Handlin Su lier Technical Documents. This procedurestates a specific program in which technical documents are receivedand transmitted to the appropriate personnel for proper channeling.All technical documents are received by Stone & Webster where aresponsible engineer is assigned to review it and monitor itsprogress until it is issued as a controlled document. This reviewconsists of a detailed investigation to ensure that the informationand specifications are technically adequate and applicable to theequipment purchased. It is then transmitted to the site (Nine MilePoint's Document Control) where it is checked for comments, issued asa controlled document and maintained throughout the life of theplant. This procedure will stay in effect until a similar program,such as the one being implemented at Unit 1, NEL-014G, Control andDistribution of Vendor Documents can be developed. This procedurewi 11 define specific instructions on handling vendor documentsreceived by Nuclear Engineering and Licensing. It will contain listsof responsibilities for the responsible engineer enabling him/her toensure that vendor documents undergo proper reviewing. It will then,only after all comments are resolved and reviews completed, betransmitted to the Administrator/Engineering clerk who in turn willlog the document in the Master Drawing Index, stamp the manuals"Controlled" in red ink, and issue each as a controlled document.This procedure will provide proper guidance for the control anddistribution of vendor documents.

2.2

The third part of the program is Niagara Mohawk's Technical Reviewand Control of maintenance procedures per Section 6.5.2 of TechnicalSpecifications, which is administered through AP-2, Production andControl of Procedures. This is a unique feature of the Nine MilePoint Technical Specifications which assures that a thoroughtechnical review is performed on all safety-related procedures,rather than a cursory review and approval by the Site OperationsReview Committee as could occur at nuclear stations with StandardTechnical Specifications.

These three parts provide Unit 2 with an improved method ofevaluating and controlling technical information which subsequentlyenhances Nine Mile's position on safety.

EQUIPMENT CLASSIFICATION AND VENDOR INTERFACE (PROGRAMS FOR ALLSAFETY-RELATED COMPONENTS)

Position

Licensees and applicants shall submit, for staff review, adescription of their programs for safety-related equipmentclassification and vendor interface as described below:

1. For equipment classification, licensees and applicants shalldescribe their program for ensuring that all components ofsafety-related systems necessary for accomplishing required

2.2 (Cont'd)

2.2.1.1

safety functions are identified as safety-related on documents,procedures, and information handling systems used in the plant tocontrol safety-related activities, including maintenance, work ordersand replacement parts. This description shall include:

The criteria for identifying components as safety-related withinsystems currently classified as safety-related. This shall not beinterpreted to require changes in safety classification at thesystems level.

NMP2 currently does not have a program for classifying subcomponentsof safety-related components. All subcomponents of safety-relatedcomponents are considered safety-related.

NMP-2 utilizes the quality group classification system forclassifying the water, steam, and radioactive waste containingcomponents important to the safety of water-cooled nuclear powerplants. This system established by NRC Regulatory Guide 1.26,"Quality Group Classification and Standards," defines the QualityGroup Classification System consisting of four Quality Groups A, 8,C, and D. The definition of Quality Group A (Class 1) is provided by10CFR50.2 (V) under "Reactor Coolant Boundary". The definitions ofGroups 8, C, and D are provided by Regulatory Guide 1.26.

Niagara Mohawk's architect engineer, Stone & Webster, used this guideto develop a detailed "Equipment and Structure Classification List"located in Section 3.2 (Classification of'tructures, Systems, andComponents) of the FSAR. This section states that, "Seismic CategoryI structures, systems and components are necessary to ensure:

1. The integrity of the reactor coolant pressure boundary (RCPB).

2. The capability to shut down the reactor and maintain it in asafe shutdown condition.

3. The capability to prevent or mitigate the consequences ofaccidents that could result in potential offsite exposurescomparable to the guideline exposures of 10CFR100."

The criteria used for identifying equipment as safety-related ondocuments, drawings, and information handling systems is Stone &Webster's procedure C-3, E ui ment Identification Codes. Thisprocedure describes a format and application by which the equipmentis identified in such a manner to allow control during all phases ofplant design and construction. Each piece of equipment is identifiedby an equipment code number. This code number is divided in two byeither an asterisk(*) for safety-related equipment, or a dash(-) forall other equipment. This provides a systematic way in whichsafety-related equipment can be identified by operating personnel inquick concise manner. Therefore, NMP2 meets the intent of Section2.2.1.1.

A description of the information handling system used to identifysafety-related components (e.g. computerized equipment list) and themethods used for its development and validation.

NMP2 Res onse

The current listing of safety-related equipment is provided in theQ-List, Table 3.2-1 of the FSAR. Currently, this document is beingused by plant and engineering personnel to identify safety-relatedcomponents. As mentioned in Section 2.1, a task is currentlyunderway to upgrade the details and accessibility of the equipmentclassification list. This upgrade is described as follows and willbe implemented when the data is fully validated.

The Information Handling System that will be used to identifysafety-related components is the Master Equipment List (MEL). TheMEL is a computer data base which will ultimately consist of on-lineinformation on all equipment installed at NMP2. This data base formsthe nucleus of an information system that ties engineered componentattributes to (1) installed component attributes, (2) activecomponent documents, (3) spare parts necessary to maintaincomponents, and (4) archived component documents. Eventually, itwill form an operational authority file which interfaces with othercomputer data bases which track scheduled and unscheduledmaintenance, equipment qualification requirements, in-serviceinspections, and modifications of plant components, thus ensuringconfiguration integrity for NMP2 as well as ready access for stationsupervision.

The MEL was developed from all major existing computerized designinformation systems on cables, raceways, equipment, pipe linessupports etc., and then i.ntegrated into one data base. The designinformation provided by the NSSS vendor and A/E was developed fromengineering evaluations performed by GE and Stone 5 Webster engineersusing the criteria of FSAR Section 3.2.

Niagara Mohawk is in the process of reviewing a Project GuidelineProcedure that will provide instructions for the control, use andupdating the MEL system.

Section 6.1 of this procedure identifies a specific ModificationGroup who's responsibility is to input, modify and verify informationwithin the MEL data base. They are the only individuals authorizedto modify data base information. This is done only under the directsupervision of the Lead Modification Engineer. If and when a userbecomes aware of the information pertaining to plant equipment whichis not present in the MEL, is an authorized data field, or is inconflict with existing MEL data, he/she is required to file a MELData Input Form. This form allows the modification group toinvestigate new additions and corrections for verification andvalidation. If the information is valid, the modification engineersigns-off on the input form and a change is made to the data base ~

Validation of the MEL for safety-related components is accomplishedon a system basis by an extensive check of the componentidentification number against drawings, existing data bases, testinginformation, name pl'ate serial numbers and if necessary, physicalinspection in the plant. This effort is. currently continuing.

Personnel using the MEL data base will have access to a number ofterminals available at several locations throughout the plant.Selected terminals will have printers available enabling the user tomake "hard" copies of requested information. In addition, a MELusers manual will be made accessible to the user to assist in usingthe terminals.

A description of the process by which station personnel use thisinformation handling system to determine that an activity issafety-related and what procedures for maintenance, surveillance,parts replacement and other activities defined in the introduction to10CFR50, Appendix 8, apply to safety-related components.

NMP2 Res onse

The following is a description of the process of determining if anactivity is safety related. The supervisor of the departmentresponsible for the activity has the responsibility to utilize theEquipment Classification List (Q-List) to determine the equipmentclassification. Documents such as work requests and purchaserequisitions are reviewed and approved by the Quality AssuranceDepartment. Activities such as surveillance or preventativemaintenance are covered by procedures which are reviewed and approvedper Section 6.5.2 of the Unit 2 Technical Specifications. Theseattributes are specified in various administrative procedurescurrently in place. The final administrative control before workoccurs is approved by the Shift Supervisor. Based on the training,experience and knowledge of Technical Specifications required to fillthe position, the Shift Supervisor can determine if the correctpractices are to be used. This control includes sign-offs in theprocedures, work requests and markups (tags) to be used. It is theintent of the process at Nine Mile Point Unit 2 to have checks andbalances on the system to assure that an error on the part of 'anindividual will not result in "non-safety related practices" beingapplied to safety-related equipment.

A description of the management controls utilized to verify that theprocedures for preparation, validation and routine utilization of theinformation handling system have been followed.

NMP2 Res onse

Safety-related activities are governed by various administrativecontrols which implement the Quality Assurance Program. Adherence tothe Quality Assurance Program is monitored primarily through the useof audits and inspections. These audits and inspections encompassed

(Cont'd)

the various safety-related activities and are performed at variousfrequencies. For example, maintenance activities on safety-relatedequipment are subject to quality assurance inspections on a routinebasis. Other audits or inspections are performed less often butcover a longer period of operation or activity. Items ofnon-compliance identified as a result of these audits and inspectionsare documented in accordance with provisions of the quality assuranceprogram and are carried as open items until resolved.

The Project Guildline Procedure (Management Control) for utilizingthe Master Equipment List (MEL) has been described in Section2.2.1.2. This procedure will be governed by the Quality AssuranceProgram to assure validation and compliance of standards.

A demonstration that appropriate design verification andqualification testing is specified for procurement of safety-relatedcomponents. The specifications shall include qualification testingfor expected safety service condi tions and provide support for thelicensees'eceipt of testing documentation to support the limits oflife recommended by the supplier.

NMP2 Res nse

Currently, Equipment Qualification and Design Verification areperformed in accordance with the Project Manual. The Project Manualincludes Project Procedures (PP), Project Guidelines (PG) and otheradministrative documents that control activities at Nine Mile PointUnit 2.

Design Verification and Equipment Qualification requirements arespecified in procedures and specifications for all safety-relatedprocured items. Project Procedures 3, 94 and Engineering AssuranceProcedure 3.1 describes the review, control and updati.ng of thesespecifications. Independent review is performed in accordance withSection H.l.e(3) of PP-3.

As required by the aforementioned procedures, the specificationsinclude requirements for qualification testing, review, receipt andapproval of testing documentation and vendor manuals which supportthe limits of life recommended by the supplier.

The testing .documentation and vendor manuals are reviewed,maintenance and surveillance data is extracted in accordance withPP-131 and transmitted via Equipment Qualification MaintenanceProgram Data Sheet (EQMPDS) to Niagara Mohawk Project Engineering.

(Cont.d)

This information (EQMPDS) is transferred to on site maintenancemanagement for incorporation into maintenance procedures asappropriate in accordance with Maintenance Instruction MI-4.0.

Administrative Procedure AP-6.1 is in the final stages of signoff andwill be approved prior to fuel load. Once approved, it will controlengineering support for design modifications after fuel load. Thisprocedure permits the use of the project -manual and proceduresdescribed above or the NMPC Nuclear Engineering and Licensingprocedures will be updated and used when approved for use at Unit 2.Until the NMPC Engineering Procedures are approved for use at Unit 2,the Project Procedures will be implemented in accordance with AP-6.1.

A provision is also included in AP-6.1 for procurement of exactreplacements. Exact replacements procured in accordance with theapplicable NMP2 Quality Assurance Program Topical Report(December 1985), Sect. 7.2.5 and/or ASME, Section XI, IWA-7210 (a) or(b) may be installed without recourse to a new design safetyanalysis. Applicable procurement and Quality Assurance requirementsshall be met and station documentation of these replacements shall beupdated to provide a current record of station components andconfiguration.

The NMPC Nuclear Engineering and Licensing Procedures currently usedat Unit 1, which will be updated for Unit 2 Design Verification andEquipment Qualification, include:

NEL 014D

NEL 015NEL 027N.D. 100N.T. 100.CN.T. 015. I

Control 5 Distribution of Calculations, Specifica-tions/System Descriptions/Design VerficationProcurement of Material ServicesDesign VerificationPlant ModificationsEquipment QualificationCommercial Grade Procurement and Dedication

These procedures will ensure that appropriate design verification andqualification testing is specified for procurement of safety-relatedcomponents. These procedures will ensure the receipt of testingdocumentation which supports the limits of life recommended by thesupplier.

Licensees and applicants need only to submit for staff review theequipment classification program for safety-related components.Although not required to be submitted for staff review, yourequipment classification program should also include the broaderclass of structures, systems and components important to safetyrequired by GDC-1 (defined in 10CFR Part 50, Appendix A, "GeneralDesign Criteria, Introduction" ).

(Con't)

NMP2 Res onse

With respect to the equipment classification program in use atNiagara Mohawk for structures, systems and components Important toSafety, we are participating in the Utility Safety ClassificationGroup and are seeking a generic resolution to the Staff's concern inthis regard through the efforts of the Group. We do not agree thatthe plant structure and components important to safety constitute abroader class than the safety-related set. Nevertheless, we believethat non-safety related plant= structures, systems and components havebeen designed and are maintained in a manner commensurate with theirimportance to the safety and operation of the plant.

For vendor interface, licensees and .applicants shall establish,implement and maintain a continuing program to ensure that vendorinformation for safety-related components is complete, current andcontrolled throughout the life of their plants, and appropriatelyreferenced or incorporated in plant instructions and procedures.Vendors of safety-related equipment should be contacted and aninterface established. Where vendors cannot be identified, have goneout of business, or will not supply information, the licenseeorapplicant shall assure that sufficient attention is paid toequipment maintenance, replacement, and repair, to compensate for thelack of vendor backup, to assure reliability commensurate with itssafety function (GDC-1). The program shall be closely coupled withaction 2.2.1 above (equipment qualification). The program shallinclude periodic communication with vendors to assure that allapplicable information has been received. The program should use asystem of positive feedback with vendors for mailings containingtechnical information. This could be accomplished by licenseeacknowledgment for receipt of technical mailings. It shall alsodefine the interface and division of responsibilities among thelicensee and the nuclear and nonnuclear divisions of their vendorsthat provide service on safety-related equipment to assure thatrequisite control of and applicable instructions for maintenance workon safety-related equipment are provided.

NMP2 Res onse

Niagara Mohawk was an active participant in the Nuclear Utility TaskAction Committee formed to address control and utilization ofinformation regarding safety-related components. At the outset theCommittee recognized that individual utilities have the greatestexperience with, and are most cognizant of, the application ofsafety-related equipment. Based on this recognition, the Committeeinvestigated the mechanisms currently available to facilitateinformation exchange among utilities. These included the routineutility/vendor and utility/regulator interchanges and the SignificantEvent Evaluation and Information Network (SEE-IN) and Nuclear PlantReliability Data Systems (NPRDS) programs managed by the Institute ofNuclear Power Operations (INPO). The committee concluded that these

-10-

(Cont'd)

existing activities, coupled with a coordinated program wi thin eachutility, constituted an overall program to ensure the disseminationand utilization of technical information regarding reliability ofsafety-related equipment. Additional information describing thisoverall program was provided to the Nuclear Regulatory Commission inMarch 1984 by the Committee.

A key element of the vendor equipment technical information programis a utility program to contribute information to the NPRDS andSEE-IN programs and to use the results of these programs. Theadministrative controls currently being implemented at Nine MilePoint Unit 2 contain procedures and data collection requirementsrelated to these programs. AP-3.4.2, "0 erations Ex erienceAssessment", TDP-5, "Administration of 0 erational En ineerinAssessment Items", and TDP-9, "Inde endent Safet En ineerin Grou "

define the administrative controls for handling information fromSEE-IN, NRC, GE, etc. TDP-6, "Nuclear Plant Reliabilit Data S stem(NPRDS) Failure Re ortin ", describes the steps used to input data toSEE-IN via NPRDS. These requirements provide assurance thatinformation regarding safety-related equipment is handled in anefficient, timely manner. No specific change to these existingadministrative controls is deemed necessary at this time. A minimumof 5 dedicated engineers (comprising the ISEG) are responsible forhandling the SEE-IN information. Another dedicated individual isresponsible for coordinating NPRDS activities for both Unit 1 andUnit 2, with technician and clerical assistance assigned asnecessary. This action, coupled with the existing administrativecontrols, meets the intent of Section 2.2.2 of Generic Letter 83-28addressing vendor information and interface.

The following are responses to the NRC's review guidelines forSection 2.2.2:

NMPC has obtained from INPO a status on the NPRDS and SEE-IN programenhancements. This letter is attached to this response. Inaddition, NMPC controls currently in place with regard to theguidelines for the SEE-IN program are described below:

Guideline:

Reports should be generated for potential failures caused byfaulty or missing vendor supplied information or other EquipmentTechnical Information (ETI). Such occurrences should bereported over NUCLEAR NETNORK.

Response:

TDP-5 requires that reports be submitted to INPO via NuclearNetwork for any occurrence with generic applications. Potentialfailures caused by faulty or missing vendor supplied informationor other Equipment Technical Information (ETI) would fall intothis criteria.

-11-

2.2 (Cont'd)

Guideline:

Licensee response should describe briefly how their program willaccomplish the implementation responsibilities recommended inSection 4.1.1 of the NUTAC/VETIP Report. These include:

Establishment and maintenance of vendor interface with NSSSsupplier.

Response:

Vendor interface has been established with General Electric, theNSSS vendor for Unit 2. This interface consists principally ofthe Service Information Letter (SIL) program, augmented by lessformal information exchange programs such as Service AdviceLetters and Technical Information Letters.

Guideline:

Have a program of seeking assistance from other vendors ofsafety-related equipment when found necessary.

Response:

Guideline:

Assistance is routinely sought and obtained from vendors ofsafety-related as well as non-safety related equipment. Thisassistance ranges from telephone contact to bringing the vendorservice representative on site to assist in servicing thecomponents. This is a basic part of the maintenance program andis implemented any time that the staff cannot resolve acomponent performance problem with existing procedures ortechnical manuals. Technical Specification operabilityrequirements and post-maintenance testing requirements assurethat components are not returned to service until it is proventhat they can meet their intended function. Therefore, it isnot necessary to formalize a program to seek assistance from avendor because existing programs indirectly require it.

Have procedures for processing all incoming Equipment TechnicalInformation (ETI) regardless of source to assure prompt review,evaluation, and distribution of results so that:

(1) Key personnel are promptly warned of possible problems.

(2) New or revised information is incorporated into plantprocedures and programs.

(3) Significant Equipment Technical Information (ETI) isshared with other utilities via NUCLEAR NETNORKreports.

-12-

2.2 (Cont'd)

Response:

The Operations Assessment Program (AP-3.4.2, TDP-5) containthese attributes.

Guideline:

(4) Administrative procedures should require that plantprocedures at least reference appropriate EquipmentTechnical Information (ETI).

(5) Appropriate Equipment Technical Information (ETI) should beincorporated into the performance and quality reviewsections of safety-related procedures.

Response:

Guideline:

S-MI-GEN-002, Maintenance Instructions for Writin Procedures,and S-IDP-PO, Outline for I&C De artment Procedures containprovisions that require the use of Equipment TechnicalInformation (ETI) in writing maintenance procedures.

(6) Vendors or outside contractors who perform or providesafety-related services shall be subject to adequateutility control and shall conform to utility orutility-approved QA procedures and controls.

Response:

All work performed on safety-related equipment at Nine MilePoint Unit 2 must be performed with NMPC approved procedures,regardless of whether it is performed by NMPC employees oroutside vendors or contractors. Consequently all work isperformed in conformance with NMPC or NMPC approved QAprocedures and controls.

Guideline:

Licensee response should show that interfaces have been or arebeing established with at least two or more major vendors ofsafety-related equipment other than the NSSS. Examples of suchvendors include: diesel generator vendor, switchgear vendor,major pumps vendor, or vendor of motor-operated valves.

-13-

2.2.2 (Cont'd)

Response:

NMPC strongly endorses the NUTAC report on Generic Letter 83-28NMPC will attempt to establish a vendor interface program withtwo major vendors of safety-related equipment other than theNSSS. It is our intention to develop this relationship with thediesel generator vendor, and/or with the major vendor of motoroperators for valves, and/or with the major vendor of valves,and/or with the vendor of safety-related switchgear. Thisrelationship will be developed expeditiously, however due touncertainties in the willingness of these companies toparticipate, no commitment date can be specified at this time.It should be noted, however, that the existing SIL programcovers the components in the GE scope of supply, which includescomponents in the following major systems: ECCS, includingRCIC; ADS; SLC; RHCU; RPS; Recirculation; Neutron monitoring,including RSCS; RRCS; and Fuel Handling among others. Thereforethe intent of this guideline is met without establishing twoadditional vendor interface programs.

Guideline:

Response:

Licensee response should show that they have committed to workwith INPO to ensure accomplishment of INPO ImplementationResponsibilities as described in Sections 3.2, 4.1.2, and4.2.2.1 of the NUTAC/VETIP report.

INPO has prepared revisions to NPRDS and SEE-IN as described inthe attached letter.

Guideline:

Response:

The vendor interface program should include periodic contactwith the NSSS vendor to assure that the latest versions ofmaintenance, test, service, and modification recommendations arein the licensee's possession.

TDP-5 has been revised to require annual contact with GeneralElectric regarding the SIL program and an audit of the resultsto assure that all the SILs are in NMPC's possession.

-14-

2.2 (Cont'd)

Guideline:

The licensee should show that contact has been attempted withmajor vendors of their safety-related equipment other than theNSSS to establish continuing, periodic interfaces with them forexchange of service, test, maintenance, and modification

" information. Evidence of such attempts and their results shouldbe retained for audit.

Response:

Guideline:

As described above, NMPC routinely consults vendors for thepurpose of exchange of service, test, maintenance, andmodification information. However, no attempt was made todevelop a formal vendor interface program with all vendors ofsafety-related equipment because NMPC strongly endorses theNUTAC report on Generic Letter 83-28, section 2.2.2, andconsiders a formal program unnecessary.

The vendor interface program should use a system of positivefeedback such as licensee acknowledgement of receipt oftechnical information mailings to assure that licensee hasreceived all current information.

Response:

Guideline:

The SIL program utilizes a SIL feedback form which is used byGeneral Electric to update a computerized status log. This formis .sent to GE at the time of closeout by NMPC, not at the timeof receipt. TDP-5, as described above, assures, on an annualbasis, that NMPC has received all current information.

Program description shall define the interface and describe thedivision of responsibilities among the licensee and the nuclearand non-nuclear divisions of their vendors that provide serviceon safety-related equipment. This is interpreted to mean thatthe licensee shall remain responsible for controlling thecontent and application of procedures, instructions, and qualityassurance activities to maintenance, test, service, andmodification work on safety-related equipment performed by otherthan licensee organizations and personnel.

-15-

2.2 (Cont'd)

Response:

As described above, all work performed on safety-relatedequipment at Nine Mile Point must be performed with NMPC

approved procedures, regardless of whether it is performed byNMPC employees or outside vendors or contractors. Consequentlyall work is performed in conformance with NMPC or NMPC approvedprocedures and controls. NMPC always remains responsible forcontrolling the content and application . of procedures,instructions, and quality assurance activities to maintenance,test, service, and modification work on safety-related equipmentperformed by other than NMPC personnel.

-16-

S I 0 2

01

G ENE IiALE LC CT II I C COMI:ANY

IVIANAGER, UTILITYSUPPORT SERVICES

175 CURTNER AVENUESAN JOSE, CA 95125

M/C S9O

Attachment A(To Section 2.1)

SERVICE INFORMATION LETTER STATUS RESPONSF-

SIL NO.

pc

PROJECT14

ACTION TAKEN AS OF

MO DY YR

16

6

24

CHECK ONE

UNDER INVESTIGATION

NOT APPLICABLE

DO NOT PLAN TO IMPLEMENT

AI.READY IN COMPLIANCE

PLAN TO PARTIALLYIMPLEMENT

PLAN TO FULLY IMPLEMENT

PARTIALLYIMPLEMENTED —NO FURTHER ACTION

PARTIALLYIMPLEMENTED —PLAN TO COMPLETE

FULLY IMPLEMENTFD

COMMENTS: IHAND PRINTED COMMENTS MAYBE ENTERED BELOW OR USE

REVERSE SIDE OF SHEET FOR TYPED COMMENTS)

26

60

96

130

I.ROM DATE

Section 3.1 & 3.2


Post-Maintenance Testing (Safety Related Systems)

3.1 5. 3.2 POST-MAINTENANCE TESTING

Positions

3.1.1

The following actions are applicable to post-maintenance testing:

Licensees and applicants shall submit the results of their review oftest and maintenance procedures and Technical Specifications toassure that post-maintenance operability testing of safety relatedcomponents in the reactor trip system is required to be conducted andthat the testing demonstrates that the equipment is capable ofperforming its safety functions before being returned to service.

NMP2 Res onse

AP-2, Production and Control of Procedures requires review of testand maintenance procedures and Technical Specifications to assurethat post-maintenance operability testing of safety-relatedcomponents in the reactor trip system is conducted. Additionally,this procedure requires that the testing demonstrates that theequipment is capable of performing its safety functions before beingreturned to service. The AP-2 review is conducted in two parts:1) an interdisciplinary review, and 2) a cross disciplinary review.The interdisciplinary review is the portion that involves assuringthat the test procedure demonstrates that the equipment is capable ofperforming its safety functions. All tests in maintenance proceduresand Technical Specifica'tion changes under go this review prior toimplementation.

S-IDP-PO, Outline for I&C Procedures and S-MI-GEN-002, Ma>ntenanceInstructions for Nritin Procedures are the two departmentalprocedures that control the development of maintenance procedures.These two procedures require post-maintenance testing and are used bythe reviewer to assure that appropriate post-maintenance testing hasbeen incorporated.

At this time not all maintenance and test procedures have beenapproved. Those that have been approved have been reviewed foradequacy of post-maintenance testing via the procedures describedabove. Those that have not yet been approved will'e reviewed foradequacy of post-maintenance testing with these same administrativecontrols.

Therefore Unit 2 has complied with the requirements of Section 3.1.1for those procedures approved to date, and has administrativecontrols in place to comply with these requirements in the future.

Further, AP-5, Procedure for Re air contains additional controls toassure that these requirements are met. These controls are describedin the response to Section 3.2.1 below.

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

NMP2 Res onse

Vendor and Engineering reccomendations are currently being reviewedto ensure that any appropriate test guidance is included in the testand maintenance procedures or the Technical Specifications.

As stated in Section 2.1, the General Electric SIL programconstitutes the RTS Vendor Interface Program. The post-maintenancetesting recommendations contained in the SILs have been identifiedand are being handled via the Operations Assessment Program asdescribed in the response to Section 2.1 above. As of this writing,approximately 90% have been reviewed. A few instances have beenfound where procedure changes are required. These changes are beingtracked to assure completion. The remainder (which number under 10)have been assigned to an engineer for incorporation of applicableinformation. Di.sposition of all of these will occur prior to fuelload.

Engineering recommendations are in general sent to the StationSuperintendent, who assigns them to the appropriate department headfor disposition. However, documentation of this process is notformalized, so at this time it is not possible to state the status.A thorough search and review of engineering testing recommendationshas been initiated and will be completed by 12-1-86, with anyprocedure modifications completed by 1-31-87.

Licensees and applicants shall indentify, if'pplicable, anypost-maintenance test requirements in existing technicalspecifications which can be demonstrated to degrade rather thanenhance safety. Appropriate changes to these test requirements, withsupporting justification, shall be submitted for staff approval.(Note that action 4.5 discusses on-line system functional testing.)

NMP2 Res onse

Technical Specifications have been reviewed for Post-MaintenanceTesting Requirements that can be demonstrated to degrade safetyrather than enhance it. None were identified.

Licensees and applicants shall submit a report documenting theextending of test and maintenance procedures and TechnicalSpecifications review to assure that post-maintenance operabilitytesting of all safety related equipment is required to be conductedand that the testing demonstrates that the equipment is capable ofperforming its safety functions before being returned to service.

(Cont'd)

NMP2 Res onse

Niagara Mohawk has made improvements to administrative andimplementing procedures to more clearly satisfy the Post-MaintenanceTesting (PMT) requirements of Generic Letter 83-28. AP-5 "Procedure5... 5» 5

PMT fol lowing any maintenance of Safety Related equipment. TDP-8"Post-Maintenance Testin Criteria" provides guidance on the type oftesting required based on the type of component and the type ofmaintenance performed.

This process applies to systems that have been turned over to NMPC

from Construction and is summarized as follows: The departmentsupervisor receiving the Work Request (AP-5.0, page 15) determinesif the departmental procedure for accomplishing the maintenance task,or another departmart('s) procedure, incorporates a maintenance testthat meets the requirements given in TDP-8. If so, he denotes theprocedure number on the WR (line ¹15) and on the PMT requirementsline (line 37). If not, line ¹37 is left blank. Upon completion ofthe work, the WR is returned to the Control Room, where the StationShift Supervisor or the Assistant Shift Supervisor review the WR

including line ¹37. If the Operations Department has a procedurewhich meets the testing requirements of TDP-8, it is denoted on line¹37, and performed. Successful performance results in the StationShift Supervisor or Assistant Station Shift Supervisor accepting thesystem/component for return to service. An unsuccessful test resultsin the initiation of another WR.

If no procedure exists for testing the system/component in relationto the maintenance performed, (which could be the case for a safetyrelated component or system that is not in Technical Specifications)a PMT Test Report is completed per AP-5 and attached to the WR.

Generally, this will involve placing the component in service andwitnessing proper operation.

Further, maintenance procedures which do not contain post-maintenancetests generally contain steps to notify the appropriate department toconduct a test. However, the WR is the administrative control.

Thus, Nine Mile Point Unit 2 is currently in compliance. withPost-Maintenance Testing requirements of Generic Letter 83-28.

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

(Cont.d)

NMP2 Res onse

Vendor and Engineering reccomendations are currently being reviewedto ensure that any appropriate test guidance is included in the testand maintenance procedures or the Technical Specifications.

As stated in Section 2.2.2, the General Electric SIL programconstitutes the Safety Related Systems Vendor Interface Program. Thepost-maintenance testing recommendations contained in the SILs and inany other vendor reccomendations not contained in technical manualshave been identified and are being handled via the OperationsAssessment Program. As of this writing, 901. have been reviewed. Afew instances have been found where procedure changes are required.These changes are being tracked to assure completion. The remainder(which number under 20) have been assigned to an engineer forincorporation of applicable information. It is expected thatdisposition of all of these will occur prior to fuel load. Thestatus of engineering testing recommendations is given in theresponse to 3.1.2 above.

Licensees and applicants shall identify, if applicable, anypost-maintenance test requirements in existing TechnicalSpecifications which are perceived to degrade rather than enhancesafety. Appropriate changes to these test requirements, withsupporting justification, shall be submitted for staff approval.

NMP2 Res onse

Technical Specifications have been reviewed for Post-MaintenanceTesting Requirements that can be demonstrated to degrade safetyrather than enhance it. None were identified.

Section 4.5


Reactor Trip System Reliability (System Functional Testing)

4.5 REACTOR TRIP SYSTEM RELIABILITY (SYSTEM FUNCTIONAL TESTING)

Position

4.5.1

On-line functional testing of the reactor trip system, includingindependent testing of the diverse trip features, shall be performedon all plants.

The diverse trip features to be tested include the breakerundervoltage and shunt trip features on Westinghouse, BEW and GEplants; the circuitry used for power interruption with the siliconcontrolled rectifiers on B&W plants; and the scram pilot valve andbackup scram valves (including all initiating circuitry) on GE plants.

NMP2 Res onse

Generic Letter 83-28, Section 4.5 recommends on-line functionaltesting of scram pilot valves and scram backup valves. At Nine MilePoint Unit 2, the scram pilot air system controls and supplies air tooperate the scram valves and the scram discharge volume vent anddrain valves. The control air is supplied through two backup scram,and two Redundant Reactivity Control System (RRCS) solenoid operatedair valves to the scram pilot valves, at the individual control roddrive Hydraulic Control Units (HCU) and the scram discharge volumevent and drain valves, per each of two HCU Air Headers. The backupscram valves receive signals from the reactor protection system, asdo the pilot solenoids, to each scram, vent and drain valve,providing redundancy and increasing system reliability. In the eventthat the scram pilot valves fail to function, the action of thebackup scram valves assure that the control rods insert, thus,enhancing the reliability of the reactor trip function.

The backup scram valves are normally de-energized, DC solenoidoperated valves. When at least one pair of channel sensor relays inboth trip systems de-energize (one out of tw'o taken twice logic),both backup scram valve solenoids energize and reposition the backupscram valves to block the instrument air supply and exhaust the scramair header. This action alone will cause the insertion of allcontrol rods. The check valve around backup scram valve B allows thepilot air header to bleed down even if backup scram valve B fails tochange position. Thus, the fai lure of one backup scram valve tooperate will not prevent a scram, and the operation of one backupscram valve will cause a scram of the one half of the control rods.

Current testing of the scram pilot valves is accomplished through theexisting surveillance program. The surveillance tests, takent'ogether, functionally test the trip system from the sensinginstrument, through the trip logic circuitry, to the scram pilotvalves. The surveillance procedures are written to test theone-out-of-two taken twice logic in such a manner that the channelsare tested independently. This allows one-half of the necessarylogic to "makeup," actuating the entire trip channel up to andincluding one out of the two scram pilot valves on every controlrod's scram inlet and discharge valves.

4.5.1 (Cont'd)

Scram testing will be performed during each operating cycle. Thisscram time testing demonstrates the action of the pilot scram valvesand scram inlet and discharge valves. The frequency of testing is asfollows: *

1. For all control rods prior to THERMAL POWER exceeding 401. ofRATED THERMAL POWER following CORE ALTERATIONS or after areactor shutdown that is greater than 120 days.

2. For specifically affected individual control rods followingmaintenance on or modification to the control rod or control roddrive system which could affect the scram insertion time ofthose specific control rods, and

3. For at least 101. of the .control rods, on a rotating basis, atleast once per 120 days of POWER OPERATION.

In series with the backup scram valves are two normally deenergizedDC RRCS solenoid operated Alternate Rod Insertion (ARI) valves.Similar to the B backup scram valve, each RRCS valve has a checkvalve in a bypass line so its failure will not prevent the other RRCSor the backup scram valves from depressurizing its scram air header.The ARI function of RRCS is actuated on failure to scram symptoms,i.e. high reactor vessel pressure or low-low reactor water level.

Because of the design of the system, on-line testing of one backupscram or one RRCS valve would result in a full scram of one half thecontrol rods, This would be an unacceptable situation which wouldresult in an automatic or a manual full scram of all the controlrods. Therefore, on-line testing of backup scram valves or RRCSvalves will not be performed. However, backup scram valves and RRCSvalves will be tested.

4.5.2

A plant specific reliability study was performed by GE in NEDE 22157for RRCS and ARI. The results of this study showed that thesesystems are highly reliable.

Plants not currently designed to permit periodic on-line testingshall justify not making modifications to permit such testing.Alternates to on-line testing proposed by licensees will beconsidered where special circumstances exist and where the objectiveof high reliability can be met in another way.

*NOTE: This frequency is currently specified in the Unit 2 TechnicalSpecifications. The scram timing program will always be based on theTechnical Specifications.

(Cont'd)

NMP2 Res onse

As described in 4.5.1, Nine Mile Point Unit 2 is not designed foron-line testing of the backup scram or the ARI valves. The currentdesign would result in scram of one half of the rods, if one of thebackup scram, or ARI valves were energized while on-line. However,,due to the multiple redundancy of the system, ie. the backup scramvalves are redundant to the scram pilot valves, and are alsoredundant to each other, modifications to permit on-line testing arenot warranted.

Additionally, the ARI valves are a redundant scram system, utilizingindependent sensors from the Reactor Protection System and capable ofcompleting a scram with the total failure of the normal scramsystem. The ARI valves are controlled by the Redundant ReactivityControl system, which is also redundant.

NMPC endorses the following excerpt from NEDC-30505 "ResponseGuidelines for NRC Generic Letter 83-28" prepared by General Electricfor the BWR Owners Group.

"The Nine Mile Point Unit 2 Reactor Protection System design complieswith all applicable regulatory requirements for the RPS.

The remainder of this paragraph is a summary of the on-linefunctional testing and testing intervals performed on the RPS.Consistent with the Technical Specifications, on-line channelfunctional testing is performed on the multiple and diverse reactortransient trip sensors [Average Power Range Monitor (APRM) andintermediate Range Monitor (IRM) Reactor trip signal channels, andmultiple and diverse Scram Discharge Volume High water level trips).During the required trip sensor channel tests discussed above, eachscram contactor which actuates the scram pilot solenoid valves istested. The simple operation of the scram contactors minimizesconcerns of wear, and frequent testing assures that any failures aredetected early. The Scram Pilot Solenoid Valves which are actuatedby the scram contactors are all tested regularly. RedundantElectrical Protection Assemblies (EPAs) which protect the Scram PilotSolenoid Valves from low voltage chattering (and the associatedpotential consequence of accelerated wear) are also functionallytested. These surveillance testing requirements related to the ScramPilot Solenoid Valves assure that the probability of undetectedfailure of these solenoid valves is small. In summary, the currentRPS on-line surveillance requirement, in conjunction with multipleand diverse scram sensors, assure that the probability of failure of

.enough control rods to prevent scram is negligible.

(Cont'd)

Channel functional tests are performed on-line for the followingsensor trips:

Reactor Vessel Dome Pressure-HighReactor Vessel Water Level-LowMain Steam Line Isolation Valve-ClosureMain Steam Line Radiation-HighDrywell Pressure-HighTurbine Control Valve Fast Closure, Control Oil Pressure-LowTurbine Stop Valve-Closure

Channel functional tests are also performed for APRMs and IRMs.

In References 1 and 2, it is shown that each of the above plantvariables used to initiate a protective function is backed up by acompletely different plant variable. In fact, it can be seen fromTable 1 that for the most frequent transients, scram is initiated bythree diverse sensors in all but one case (regulator failure-primarypressure increase which is intitiated by two diverse sensors). Thisindicates that adequate redundancy exists in the design to provideprotection against multiple independent sensor failures. Also,diversity among sensor types reduces the potential for common causefailures, failures due to human error, and increases in failure ratedue to wearout. A pictorial representation of the RPS logicconfiguration is provided in Figure 1.

Each sensor channel functional test includes full actuation of theassociated logic, the two output scram contactors in each channel,and the individual CRD scram air pilot valve solenoids for theassociated logic division (solenoids from both logic Division A and 8are required for scram initiation).

The most credible failures within the RPS logic will de-energize aset of scram solenoids which causes a half scram, i.e., one of thetwo scram solenoids required for scram initiation is de-energized atsome or all hydraulic control units. These failures would be "SAFE"failures that would increase the probability of plant shutdown.

The less credible logic failures which prevent a channel fromde-energizing will be detected during channel functional test incompliance with Technical Specification requirements. The testsdescribed above ensure that an increase in failure rate due to awearout condition or a common cause failure potential could bedetected early and corrective action taken before the failurecondition becomes systematic.

Other channel functional tests include testing of the Scram DischargeVolume (SDV) Water Level-High trip and manual scram trip and test ofthe reactor mode switch in the shutdown position every refueling.The first two trips involve on-line testing and the latter modeswitch test can only be conducted during reactor shutdown. Themanual scram trip can be tested on-line without creating a scram.

(Cont'd)

The testing of the SDV Hater Level-High trip is considered adequatebased on the current designed redundancy and diversity incorporatedinto the system. There are two diverse and redundant sets of levelsensors which scram the reactor in the unlikely event of high waterlevel in the SDV during power operation. These trips are designed toallow sufficient scram water discharge volume given the scram trippoint is reached.

Reference 2 concluded that reactor shutdown can be achieved if atleast 50'/ of the control rods in a checkerboard pattern and 69K in arandom pattern are inserted in the core. The probability ofindependent failure of enough rods to prevent shutdown isnegligible. The most unlikely type of failure would be some commoncause mechanism that if undetected over a long period of time wouldcause unsafe shutdown. The Technical Specification surveillancerequirements adequately ensure that a failure mechanism affectingseveral individual drives (considered to be very remote) would not goundetected. One of the major features that ensures that severaldrives do not fail at one time due to wearout or a common cause isthe staggared maintenance and overhaul of selected degraded CRDs orHydraulic Control Units (HCUs) at refueling outages. This ensures amix of drives by age, component lot, maintenance time and servicingpersonnel, and testing.

The scram insertion time tests include, in addition to drive timingand insertion capability, a test of operability of the HCU scraminsert and discharge valves including associated scram air pilotvalves. As stated in the previous paragraph, the required frequencyof testing given in the Technical Specification ensures that asystematic failure mechanism in the HCUs would be detected earlyenough and corrective action taken before the condition becomes acritical failure preventing scram."

Therefore, since the scram pilot valves are tested weekly during APRMhalf scram tests, and since the backup scram valves and the ARIvalves will be tested once a refueling cycle, and since rod scramtime testing is performed at on a refueling cycle or more frequentlyin accordance with Standard Technical Specifications, on-line testingof the backup scram and ARI valves is not warranted.

ExistingTechnicalintervalsavailabili

l.2.3.4,5.

intervals for on-line functional testing required bySpecifications shall be reviewed to determine that theare consistent with achieving high reactor trip systemty when accounting for considerations such as:

uncertainties in component failure ratesuncertainty in common mode failure ratesreduced redundancy during testingoperator errors during testingcomponent "wear-out" caused by the testing

(Cont'd)

Licensees currently not performing periodic on-line testing shalldetermine appropriate test intervals as described above. Changes toexisting required intervals for on-line testing as well as theintervals to be determined by licensees currently not performingon-line testing shall be justified by information on the sensitivityof reactor trip system availability to parameters such as the testintervals, component failure rates, and common mode failure rates.

NMP2 Res onse

Nine Mile Point Unit 2 on-line functional testing and testingintervals are performed consistent with the Technical Specificationswhich are based on Standard Technical Specifications. The followingreactor trips are functionally tested on-line.

I

Manual ScramHigh Reactor PressureHigh Drywell PressureLow Reactor Water LevelHigh Water Level Scram Discharge VolumeMain Steam Line Valve PositionHigh Radiation Main Steam LineNeutron Flux

Intermediate Range Monitor (IRM) (when required)Average Power Range Monitors

Turbine Valve ClosureGenerator Load Rejection

In addition, the shutdown position of the reactor mode switch scramfunction is tested during refueling outages. During the testingdiscussed above, the scram pilot solenoid valves are tested, in thatone of the two scram pilot valves on. every control rod scram inletand outlet valves are activated. Also, overvoltage, undervoltage andunderfrequency protection is provided for the reactor trip busincluding power to the scram pilot valves.

For the major transients evaluated, the number of independent scramfeatures which are available to terminate a particular transient arelisted in the response to Section 4.5.2 above. Therefore, it can bedemonstrated that adequate redundancy exists in the Nine Mile PointUnit 2 design to provide protection against multiple independentsensor failures.

Further, NMPC participated in and endorses the "BWR Owners Groupresponse to NRC Generic Letter 83-28, Item 4.5.3" NEDC-30844. Thisdocument contains analyses performed by General Electric thatconcluded that the current on-line functional testing intervals areadequate to achieve high reactor trip system availability.

(Cont'd)

In summary, the current reactor protection system on-linesurveillance program requirements, in terms of scope and testingintervals, in conjunction with multiple and diverse scram sensorsassures the probability and reliability of the reactor trip system tofunction to effect control rod insertion and resulting reactorshutdown.

Further, for Unit 2 an automatic standby liquid control system isinstalled which provides redundant means to shut down the reactor.

Scram Signals — Order of Occurrence

Inputs FromPressure orDifferentialPressureTransmittersand Trip Units

Inputs FrcmInputs From Pressure Neutron Pt.ux

Position or Micro Switch or RadiationContact Opening Sensors

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Pres. Ilegulator Failure (primary pressureincr«ase)

F.W. Flow Control, Failure (reactor waterinve»toly i»cll.,lse)F.W. Flow Co»t.rol, Failure (reactor waterinventory d«crease)

l.oas o I (:o»d e»se r V, cuum

Loss of Normal AC Power

Figure 1 RPS Relay Logic Configurarion (to Secgion 4.5)

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REFERENCES

1. NE00-1-189, "An Analysis of Functional Common-Mode Failures in GE BWR

Protection and Control Instrumentation," L. G. Frederick, et al, July1970.

2. "BWR Scram System Reliability Analysis," W. P. Sullivan, et al,September 30, 1976 (Transmitted in letter from E. A. Hughes (GE) to0. F. Ross (NRC), "General Electric Company ATWS Reliability Report,"September 30, 1976).