ML20078R818

Forwards Response to Generic Ltr 83-28, Required Actions Based on Generic Implications of Salem ATWS Events. Investigation of Each Reactor Trip Will Be Documented in Incident Investigation Rept

ML20078R818
Person / Time
Site:	Catawba
Issue date:	11/04/1983
From:	Tucker H DUKE POWER CO.
To:	Eisenhut D Office of Nuclear Reactor Regulation
References
GL-83-28, NUDOCS 8311150312
Download: ML20078R818 (42)

Text

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DUKE POWER GOMPANY P.O. foox 33180 CHAltL(FrrE, N.C. 28242 HAL H. TUCKER Tet.nenown

,;",l" L November 4, 1983 (704)o73-4538 Mr. D. G. Eisenhut, Director Division of Licensing U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Re: Catawba Nuclear Station Docket Nos. 50-413, 50-414

Dear Mr. Eisenhut:

NRC Generic Letter No. 83-28. " Required Actions Based Upon Generic Implications of Salem ATWS Events" requested information concerning the status, plans, and schedules for conformance with the positions contained in the letter. Attached are responses for Catawba Nuclear Station, Un'its 1 and 2, to the applicable sections of the letter.

I declare under penalty of perjury that the statements set forth herein are true and correct to the best of my knowledge.

Very truly yours,

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ey Hal B. Tucker REH:jfw Attachments cc: Mr. Robert Guild, Esq. NRC Resident Inspector Attorney-at-Law Catawba Nuclear Station P. O. B<,x 12097 Charleston, South Carolina 29412 Mr. Jesse L. Riley Palmetto Alliance Carolina Envircnmental Study Group 2135 Devine Street 854 Henley Place Columbia, South Carolina 29205 Charlotte, North Carolina Mr. James P. O'Reilly, Regional Administrator

'U. S. Nuclear Regulatory Commission Region II 101 Marietta Street NW, Suite 2900 Atlanta, Georgia 30303 8311150312 831104 DR ADOCK 05000

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1 D. G. Eisenhut i November 4, 1983 '

Page 2 bec: . K.' S. Canady N. A. Rutherford J. W. Cox (CNS)

R. O. Sharpe L. R, Bledsoe. (Toddville)

P. L. Abraham

.G. t ' , Hallman G. B. Caldwell

- B. T..Faulkenberry R. C. Futrell P. H. Barton

' J. S. Warren R. S. Howard (W)

G.'A. Copp R..E. Hatris J. W. Hampton (CNS)

R. W. Ouellette.

C. A. Little

. G. J. Pollak S. T. Rose T. C. McMeekin J.-E. Thomas Section File MC-801.01

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• 1.1 POST-TRIP REVIEW (PROGRAM DESCRIPTION AND PROCEDURE)

.This report describes the Reactor Trip Investigation Program established at Duke Power-Company's Oconee, McGuire, and Catawba Nuclear Stations. In particular, that portion of the Program that is performed immediately following a trip and prior to restart (e.g. , post-trip review) is described.

This~ report describes the Program's essential elements and basic requirements; details can be found in the implementing station-directives and procedures.

Purpose The purpose of the teactor trip investigation is to provide a mechanism whereby each reactor trip is systematically and thoroughly investigated in order to:

1. Determine the immediate and root causes of the trip.

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2. Identify unexpected, abnormal responses to the trip by plant systems, equipment, and personnel.

3. Assess the impact of identified abnormalities on nuclear safety, equipment reliability, system performance, and availability.

4. Confirm the readiness of the unit to restart, or bring it to readiness.

5. Develop corrective actions to prevent the recurrence of the trip and mitigate abnormal responses commensurate with their significance.

-6. Document observed plant behavior for use in subsequent evaluations.

7. Satisfy reporting requirements (10 CFR 50.73).

Scope The Reactor Trip Investigation Program implemented at Duke nuclear stations applies to every reactor trip, planned and unplanned. However, planned reactor trips need not undergo all phases of the investigation if response is no rmal . The scope of the information reviewed under the program is sufficient to accomplish its objectives and includes data on plant system behavior, actuation and sequence of operation of equipment, records of operator actions, and plant activities-affecting the event. .The program prescribes activities that are performed immediately following a trip, prior to restart, and continues through a subsequent in-depth evaluation that supports preparation of internal and external reports.

Roles and Responsibilities Several groups, including station and general office staff, participate in the reactor trip investigation program. The responsibilities and authority of each participant have been clearly defined.

., 2 At each station, the Operations group is responsible for operating the plant.

Under the program, they notify station management and the Performance and Licensing groups of the occurrence of a reactor trip.

The reactor operators are responsible for _ diagnosing and controlling the event and thus will have firsthand knowledge of the particulars of the event. This information is to be promptly documented to help ensure that a complete record of the event is obtained. Operations is responsible for deciding when and how

'the unit is to be. restarted.

At Duke nuclear stations, the Reactor Engineer, or an Engineer trained and-qualified by the Reactor Engineer is responsible for performing the post-trip review.

The general office Reactor Safety Section staff provides analytical support.to the Station staff as . requested. - The section provides expertise in event and trainsient analysis, safety analysis, and probabilistic risk analysis.

Program Phases The Reactor Trip Investigation Program consists of four distinct phases:

1. Post-trip review

2. Restart decision

3. Independent review

4. Subsequent investigation Every reactor trip will be subjected to a post-trip review and restart decision. Planned reactar trips, where no abnormalities have been identified, need not proceed to the subsequent investigatlen phase. The major elements of each of these phases is described below.

Post-trip Review The post-trip review is performed immediately following the trip and. completed prior to restart. The purpose of the post-trip review is to:

1. Determine the immediate cause of the trip.

2. Identify other-than-expected performance of operators, systems, and equipment.

3. Assess the impact of identified abnormal performance on safe operation.

4. Ensure continued availability of information and data pertaining to the event.

. The scope of the post-trip review has been established to ensure that abnormal

- performance in important systems will be identified. Guidelines and criteria, which define the range of expected system response, are used in the process.

The major elements of the post-trip review are:

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- 1. . Determination of.-immediate cause of the trip.

2. Determination of the reactor trip sequence events.

3. Review of pre- and-pos't-trip behavior of key parameters that reflect overall plant performance and will identify abnormal perforeance of important systems.

4. Review of- the performance of important plant systems and equipment

'(safety and, control) to identify'other-than-expected response to the trip. These systems include ESF, ,AFW/EFW, RCS pressure relief and control, RCS inventory control, main steam relief and control, SG level control,~and radiological control' .

5. Determination of the need for an Independent Review prior to restart.

6. Approval for restart, or identification of corrective actions that must be completed prior to restart.

Restart Decision Prior,to restarting the unit,. Operations must ensure that:

1. The immediate cause of the trip is known or has been investigated to the fullest extent possible while shutdown.

2. The plant's transient response either did not identify any problems that impact the ability of the unit to be safely restarted and operated, or that the problems.have been corrected.

3. Any problems with equipment subject to Tech Spec LCO requirements are corrected as required.

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4. The recommendations for corrective actions identified during the post-trip review are resolved.

-Independent Review I

Under certain conditions, an additional independent review must be performed prior to restart to ensure that all questions regarding the ability to safely restart and operate the plant are resolved. Criteria have been established as to when the additional independent review is required. They are as follows:

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1. If the immediate cause of the reactor trip cannot be determined, or

2. If any unresolved safety issues exist, or

3. If compliance with licensing requirements is in question.

- The 'need for an independent review will be identified by the individual performing the Post-Trip Review, concurred with by'the Station Manager (or designee), and will be performed by. a group of knowledgeable individuals

-designated by the station manager. They will report their results to Station

-Manager or his designee.

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e. 4 Subsequent Investigation Every unplanned reactor trip will be subjected to a follow-up, in-depth investigation. In addition, planned reactor trips which show abnormalities in plant response will also be investigated. The purpose of the subsequent investigation is to ensure that all aspects of the events are fully investi-gated, evaluated, and documented. The subsequent investigation takes the knowledge gained from the post-trip review and expands upon it in areas of identified abnormal response. It ensures that the more subtle aspects of system performance, even though they did not significantly affect the plant response, are evaluated and needed corrective action identified. This report need not be completed before restart. .The scope of the subsequent investig-ation is prescribed to ensure that all reporting requirements can be met.

Implementation The Reactor Trip Investigation Program is implemented at each station via a station directive with supporting guidance contained in a manual to be used by personnel involved in the process.

Attachment A to this report is a copy of a generic station directive, upon which individual station directives are based.

Attachment B is a copy of the form used to document the post-trip review. The form is structured to provide guidance and direction, to help the reviewer "ask the right questions" in collecting and recording necessary information.

Training and Qualification Requirements The investigation of reactor trips is not an exact science, it involves considerable amounts of interpretation and judgement. Consequently, a successful, thorough investigation is dependent upon the participation of knowledgeable individuals who understand plant design, operating character-istics, safety requirements and who are familiar with plant specific transient behavior.

The Reactor Trip Investigation Program will be conducted by sufficiently trained and experienced personnel in order to ensure high quality results. A list of individuals qualified to perform post-trip reviews will be maintained at each station. Only individuals on the list will perform post-trip reviews.

As a minimum, these individuals will possess five years nuclear experience, of which four years may be satisfied by a bachelor's degree in physical science

- or engineering.

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Data Acquisition The essential element of the Reactor Trip Investigation Program is the acquisition of necessary plant data. At Oconee, McGuire, and Catawba these data are obtained from many sources. Post-trip reviewers are directed to alarm typer and events recorder printouts for data to be used in assemblying the sequence of events and documenting operator actions. The plant transient monitor is used to record analog and digital parameters for analysis of system responses. The parameters recorded by these devices include:

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(a) _ RCS pressure, temperature, flow; pressurizer level; and power level; (b) Steam generator pressure, level; condensate /feedwater flow, pressure, temperature; relief valve position; (c)' Protection system trip and actuation signals; reactor trip breaker status; generator breaker status; control system signals.

The sufficiency of parameters selected for recording has been proven by experience in performing reactor trip investigations.

Additional information on data acquisition capabilities can be found in the response to Item 1.2.

Schedule The Reactor Trip Investigation Program described herein will be implemented...

at Oconee by January 1, 1984 at McGuire by January 1, 1984 at Catawba by initial criticality.

At that time the implementing station directive, operating procedures, and performance manual guidance will be in place.

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= ATTACHMENT A, RESPONSE TO' GENERIC LETTER 83-28, ITEM 1.1

, Station Directive DUKE POWER COMPANY lWUCLEAR STATION INVESTIGATION OF REACTOR TRIPS 1.0 PURPOSE This' directive defines the action to be taken in investigating reactor trips to eneure full understanding of the cause of-the trip; the plant transient behavior before and after the trip; the trip's impact on nuclear safety, power production.and performance; and to identify necessary corrective action. In addition, this directive prescribes the criteria that must be satisfied in order to-restart the unit.

2.0 APPLICABILITY Every reactor trip shall be reviewed as set forth in this directive, except as noted below. A reactor trip is-defined as an event wherein the reactor trip breakers open and the (shutdown bank) control rods fall to their fully inserted position. Specifically excluded from this

.. definition are control rod drop tests performed at power levels below 5%

full power.

A reactor trip need not undergo Subsequent Investigation (Sectior. 4.4) as prescribed herein if:

(a) The trip was initiated and specifically called for in an in progress test procedure, or was part of a planned unit shutdown, and

.(b) No' anomalies in plant behavior were identified during the Post-Trip Review (Section 4.1).

3.0 RESPONSIBILITIES Upon a reactor trip, the Superintendent of Operations, Shif t Supervisor, or Operations Duty Engineer (hereinafter " Operations"), shall notify the Station Reactor Engineer or Performance Reactor Unit / Duty Engineer of

'the trip. Operations'shall ensure that appropriate notification is made to the NRC and Nuclear Production Department Duty Engineer, and that appropriate notation of the event is entered into the SRO Logbook.

The Station Reactor Engineer, or an engineer trained and qualified by the Reactor Engineer, (hereafter Reactor Engineer) shall perform .a c Post-Trip Review. The Reactor Engineer shall.make a recommendation to Operations on whether or not to restart the unit, including identifica-tion of any items that should be . resolved prior to criticality.

Operations shall be responsible for deciding when and how the unit is to be. restarted. The recommendation (s) from the Reactor Engineer shall be considered in this decision.

s 4.0 IMPLEMENTATION s An' investigation as described herein shall be initiated immediately following a reactor trip. Prompt action shall be taken to ensure that relevant information and data are retained. In particular, necessary

- interviews of personnel knowledgeable of the reactor trip should be conducted promptlyi The investigation process described ir. this directive has four (4) distinct phases:

(a) Post-Trip Review (b)' Restart Decision

_ (c) Independent Review (d) Subsequent Investigation 4.1 Post-Trip Review ,

- A Post-Trip Review shall be performed immediately following a reactor trip and completed prior.to restart of the unit.

The purpose of the Post-Trip Review is to:

(a) Determine the immediate-cause of the reactor trip. It is not required that the root cause (e.g. , the cause of a component

-failure' leading-to!the trip) be determined at this time.

. (b) Identify other-than-expected performance of operators, systems, and equipment and assess its impact on safe plant operation.

' The actions to be performed as the Post-Trip Review are listed on Enclosure 4.1.

The' Post-Trip Review shall be performed by the Reactor Engineer with assistance as needed from other personnel. The results'of the review shall be adequately documented and provided to personnel performing a Subsequent Investigation (Section 4.4). The Reactor Engineer shall retain a record of the Post-Trip-Review.

Written guidelines shall be prepared for use in performing the Post-Trip

. Review. These. guidelines shall describe the various aspects of the trip event that should be considered in order to ensure that any pact on safe operation ic identified and resolved and shall provide criteria and guidelines defining the range of expected plant responses.

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4.2 Restart Decision

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Prior to restart of the unit, Operations shall ensure that the following criteria are met.

1. The immediate cause of the reactor trip is known or has been investi-

gated to the fullest extent possible while remaining in the shutdown condition.

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.2. . The plant transient behavior, immediately preceding and until

. stabilization following the trip, does'not identify any unresolved problems that impact the ability of-the unit to be safely restarted and operated.

'3. - Any malfunctions or failures in equipment or components subject to Tech Spec LCO requirements are evaluated and corrected as required prior to restart.

Operations shall ensure that the Reactor Engineer's recommendation (s) are resolved prior to restart and shall obtain his/her concurrence with restart. Concurrence shall be indicated by the Reactor Engineer's signature on the trip recovery operating procedure. The Station Manager, or his designee, shall resolve disagreements between Operations and the Reactor Engineer regarding necessary corrective action.

4.3 Independent Review The.need for an Independent Review will be identified by the Reactor Engineer and concurred with by the Station Manager, or his designee.

An independent review shall be performed prior to restart if:

1. The immediate cause of the reactor trip cannot be determined, or

2. Any unresolved safety issues exist, or

3. Compliance with licensing requirements is in question.

The independent review shall be performed by a group of individuals identified by the Station Manager who are knowledgeable of the issue in question. This group shall make a recommendation on restart to the Station Manager, or his designee, who shall resolve the restart decision. Documentation of the results of this review, including the bases for conclusions reached, shall be included in the documentation of the event.

4.4 Subsequent Investigation The investigation of each reactor trip shall be documented in an Incident Investigation Report. This report shall be prepared in accordance with Station DItective 2.8. .

As a minimum, the Incident Investigation Report for a reactor trip event shall include,

1. Information and data collected during the Post-Trip Review.

2. A sequence of Events describing all action taken and system /

equipment performance during the event that materially affected the l course of the event.

3. As assessment of the plant transient behavior that identifies any deviations from expected plant performance and that documents the behavior of key plant parameters.

4. An assessment of the performance of protection and Engineered Safety systems during the transient, identifying any malfunctions or failures to perform as expected.

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5. As asseesment of other equipment failures that contributed to the f ,- event.

6. As assessment of personnel performance and procedure adequacy l relevant to the event.

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7. A description of corrective actions taken or planned as a result of the event.

The Subsequent Investigation need not be completed prior to restart.

5.0 PERSONNEL QUALIFICATIONS Personnel performing Post-Trip Reviews shall be qualified by training and/or experience to ensure that they are sufficiently knowledgeable of system design and operating characteristics and nuclear safety consider-  !

ations to identify other-than-expected performance and to assess its  !

significance. As a minimum, they shall have five (5) years experience, {

of which four (4) years may be satisfied by a bachelors degree in engi- )

neering or physical science. The Reactor Engineer shall maintain a list of personnel qualified to perform Post-Trip Reviews. Only personnel on the list shall perform Post-Trip Reviews.

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Enclosure 4.1 Post-Trip Review Actions

.(1) Verify unit stability following trip

'(2). . Service Transient Monitor (3) Determine or verify immediate cause of trip; confirm reactor trip sequence of events through review of Events Recorder, Alarm Typer, and Transient Monitor.

The . reactor trip sequence of events includes:

(a) type and time of initiating RPS/SSPS signal (b) time of each Reactor Trip Breaker opening

'(c) time of manual reactor trip signal (d' time of turbine trip (4) Review pre- and' post-trip behavior for the following key parameters:

Primary Seconda ry .

RCS Tave, each loop SG Pressure

'"russurizer Level- SG Level RCS Pressure

-RCS Cooldown Limit Reactor Power Identify deviations from expected behavior and perform in-depth investigation as appropriate.

(5) . Review performance of the following systems / equipment:

(a) Safety systems including Emergency / Auxiliary Feedwater; Emergency Core Cooling systems; Containment Isolation, cooling and spray; Reactor Protection System including trip breakers.

(b) RCS PORV and Code Relief valves (c) Emergency power supply (source to 4160 v bus) -

(d)4 SPDS " problem" indication (when installed)

(e) Main Steam (code) relief valves (f) . SG pressure control ;(TBV or steam dump)

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U (g) SG 1evel control (h) RCS inventory control (i)' RC Pumps including seal injection and cooling (j) ~ Sequence of operation for turbine runback, FW isolation, AFW start (k) RIA/ EMF (response as expected)

Identify other-than-expected performance. Abnormal behavior requires i in-depth evaluation and resolution prior to restart. If performance in all areas was as expected, unit may be safely restarted.

(6) Ensure continued availability of data from the following sources:

Transient Monitor, Alarm Typer, Events Recorder / Summary, Control Room Logbooks.

(7) Conduct interview of involved personnel as necessary.

(8) Identify the need for an (additional) Independent Review. If the criteria specified in Section 4.3 are not met, then an (additional)

Independent Review is required. Indicate need on Post-Trip Review Report and notify Operations and Station Manager, or his designee.

(9) Identify corrective action that should be taken to resolve identified problems. Specifically identify those that should be completed prior to restart.

(10) Sign off operating procedure.if in agreement with restart. If not in agreement, advise Operations of actions that should be completed prior to criticality to ensure continued safe operation and DO NOT sign operating procedure until completed.

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1 Attachment B, Response to Generic Letter 83-28, Item 1.1 DUKE POWER COMPANY Nuclear Production Department Reactor Trip Investigation Program NCGUIRE/ CATAWBA NUCLEAR STATION DATE:

POST-TRIP REVIEW REPORT PREPARED BY:

I. Identification of Tr'nsienta (a) Unit Date of Occurrence Time (b) Cause of Transient:

'(c) Description of Transient:

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Revised 10/83

2 HCGUIRE/ CATAWBA NUCLEAR STATION POST-TRIP' REVIEW REPORT II. Initial Conditions (a) Reactor Power (b) No. of. NC Pumps Operating (c) No. of CF Pumps Operating Thru ( ) Upper ( ) Lower Nozzles (d) Turbine Load (e) No.' of Charginig Pumps Operating (f) . Status of Control Stations (Manual / Auto):

.1. Reactor Control

2. PZR PORV

3. PZR Heater

4. PZR Spray

5. PER Level (RCS Makeup)

6. MFW Pump Speed Control'

7. SG Level Control

8. SG Pressure Control-

9. Turbina Control

10. PZR PORV Block Valve (0/C)

(g) Off-normal Status of Any Trains / Portions of an Safety Systems:

1. RPS

2. ECCS

i. 3. Containment Cooling & Spray 4

4. Auxiliary Feedwater

5. Emergency Power

(h) . Testing-in' Progress

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. MCGUIRE/ CATAWBA NUCI. EAR STATION POST-TRIP REVIEW REPORT ,

~III. Plant Response A. Safety System Actuation and- Performance

1. Sol'id State Protection System Trip Initiating Signal Time

2. Reactor Trip Switchgear Reactor Trip Breaker A open-time Reactor Trip Breaker B open-time

3. Manual Trip Signal-Time

4. ~ Turbine Trip Trip Initiating Signal Time

5. Engineered Safeguards

' Safety Injection . Initiating Signal Time NI Actuated:

UHI Actuated:

Accumulators Actuated:

ND Actuated:

DG Sequencer - Initiating Signal Time Feedwater. Isolation - Initiating Signal Time Steamline Isolation - Initiating Signal Time Phase A Isolation - Initiating Signal Time Phase B Isolation - Initiating Signal Time Containment Spray- - Initiating Signal Time 3

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. 4 MCGUIRE/ CATAWBA NUCLEAR STATION POST-TRIP REVIEW REPORT Cont. Vent. Isolation

- Initiating Signal Time Ice Condenser Doors Open Time

6. Auxiliary Feedwater - Initiating Signal Time MDAFWP #1 Started at Stopped at

1. 2 Started at Stopped at TDAFWP Started at Stopped at

7. Pzr Code Safety Valves -

Actuated Time

8. Main Steam Safety Valves -

Actuated Time Which Valves?

9. SPDS Problem Indication -

Type Time B. Control Systems Actions

1. Reactor Runback (Yes/No) Runback Signal Power Level / Time Beg.: / End: / _

2. Turbine Runback (Yes/No) Runback Signal Load / Time Beg.: / End: /

3. Primary Pressure Control Was Pzr PORV Actuated? (Yes/No) Time Was Pzr Spray response normal? (Yes/No)

Was I r Heater response normal? (Yes/No)

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MCGUIRE/ CATAWBA NUCLEAR STATION POST-TRIP REVIEW REPORT b

4. NC Inventory Control Was Pressurizer Level Control normal? (Yes/No)

Was letdown isolated? (Yes/No) Time Was coolant released from the' quench tank?

Were any additional NV pumps started? (Y/N) Which?

Auto Start On what signal?

Manual Start Time Started Time Stopped

5. NC Flow Response Did/were any NC pumps trip (ped)? (Yes/No)

Pump (s) Time

6. NC Temperature Response Was the NC System Cooldown' rate within the Technical Specification Limit? (Yes/No)

7. Secondary Pressure Control-Did the following systems respond properly:

Atmospheric Dump Valves?

Turbine Bypass Valves?'

SG PORV?

Was steam pressure control satisfactory?

8. Secondary Level Control Was the steam generator level response as expected?

Feedwater Isolation of Reactor Trip and Low Tave-Time

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HCGUIRE/ CATAWBA NUCLEAR STATION POST-TRIP REVIEW REPORT f

C. ' Manual Actions - were any control stations taken from auto to manual? (Specify station and N time / sequence.)

Other manual actions:

D. Radiological Response - (Describe any unexpected behavior, such as unexpected EMF response)

E. . Transient Data for Key Plant- Parameters MAX MIN Reactor Power NC Tave Loop A Loop B. Loop-A Loop B Loop C Loop D Loop C Loop D.

-NC Cooldown Rate Pressurizer Pressure j Pressurizer. Level l

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7 MCGUIRE/ CATAWBA NUCLEAR STATION POST-TRIP REVIEW REPORT SG Pressure SG A SG B SG A SG B SG C SG D SG C SG D SG Level SG A SG B SG A SG B SG C SG D SG C SG D (Attach Transient Monitor Plots of Key plant parameters when available.)

IV. Description of Unexpected Responses A. Discussion of Any Unexpected Behavior of Key Parameters (Refer to III.E.)

s B. Discussion of Unexpected Personnel Response (Refer to III.C)

C. Identification of Systems with Inadequate Performance (Discuss the Nature of the Deficiency) (Refer to III.A and B) 1 a

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8 MCGUIRE/ CATAWBA NUCLEAR STATION POST-TRIP REVIEW REPORT V.. Identification of Needed Action - Prior to Restart Is an Independent Review Required? ( )Yes ( )No VI. Identification of Recommended Action - Following Restart VII. Recovery Operating Procedure signed off, time Reactor Critical, time Turbine On-Line, time i

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1.2 POST-TRIP REVIEW - DATA AND INFORMATION CAPABILITY Each unit- has three primary sources for collecting data for analyzing unscheduled reactor shutdowns. These are the:

(1) Events Recorder, (2) Alarm Typer, and (3) Transient Monitor.

The first two are used to determine the sequence of events, while the third is used for analog data trending.

Events Recorder The events recorder is a dedicated, computer system. Inputs from electrical devices are wired to the events recorder cabinets. Shan an input alarms, this fact is printed out on a paper tape. The time of each event is recorded, with discrimina-tion down to 1 millisecond, along with the parameter number. Some of the inputs also print out when the alarm clears. This tape is removed (torn off) daily and is stored in the station master file. A representative list of parameters moni-tored by this device is attached. As shown, the Events Recorder input sources include the equipment and components pertinent to reactor trip investigations.

This device is normally powered by a non-class IE non-interruptible AC source.

In addition, the events recorder is supplied by a non-interruptible non-class IE DC power supply as a backup.

Alarm Typer The alarm typer is an output device associated with the plant computer. All digital alarms and safety-related pumps, valves, and motors change of status indications that are received in the control room are printed on the typer.

Event times are printed out to the nearest second. The data are printed out giving the time, the alarm number, and a description of the alarm or device.

The alarm typer printouts are removed daily and stored in the station master file. This device is powered by a non-class IE, non-interruptible AC source. y Transient Monitor The transient monitor is a program on the plant computer at Catawba.

The' data are recorded once a second. A list of parameters routinely delogged following a trip is attached. Many additional points are available. Data is captured over a 40 minute window, 10 minutes before the trip and 30 minutes afterwards. The data is stored in two ways. The data from the plant computer is stored on floppy disks, as well as being shipped to an offsite computer located at the General Office. This device is powered by a non-class IE non-interruptible source.

The transient monitor can output data either in graph or ta'ule form. For graphs, the timescale and parameter ranges can be specified. Time windows can be as long or as short as desired, within the bounds of collection capability. (See attached example,s.) Hard copies of graphs of major plant parameters are normally kept.

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c r Tabular data can be printed, at specified intervals, and stored as well.

The selection of parameters available on the transient monitor was made based on Duke Power's experience with transient analysis. We have several years experience with in-depth analysis of transients at Oconee, and McGuire transients are undergoing similar analysis. The parameters recorded by the transient monitor document the pre- and post-trip behavior of the primary system, secondary system, important BOP systems and equipment, and Safety Systems. The parameters listed have been proven sufficient to analyze plant response and to identify and resolve abnormal system performance.

In addition to the data available from the events recorder, alarm typer,'and transient monitor, information is available from the control room strip chart recorders. Operator interviews are routinely conducted as part of the post-trip review process and the control room logbooks are availabic for additional information.

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Attachment 1 Event Recorder Inputs Representative List of Points Available for Post-Trip Review Reactor Protection System Trips - each channel and parameter Reactor Trip Breakers - each breaker Turbine Trips - each signal Feedwater Pump Trips - each signal and pump Condensate Booster Pump Trips - each signal and pump Generator Lockout - each bus or train Station Battery Trouble Alarm - each battery 4KV Bus Transfer - each bus 7KV Bus Transfer - each bus Generator Loss of Excitation Diesel Generator Status - each D/G Diesel Generator Load Sequencer - each D/G Manual Safety Injection Reactor Trip Actuation - each train Reactor Coolant Pump Underfrequency - each pump Reactor Coolant Pump Trip - each pump Reactor Coolant Pump Undervoltage - each pump llotwell pump Trip - each pump In addition, some Condensate Booster Pump and Main Feedwater Pump status alarms  ;

(e.g. -low suction pressure) are on the events recorder.

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Attachment 2

• Catawba Nuclear Station-Standard Transient Monitor Plot List -

A0890 AUX FEEDWATER FLOW TO S/G h A0884 AUX FEEDWATER FLOW TO S/G B A0878 ' AUX FEEDWATER FLOW TO S/G C A0872 AUX FEEDWATER FLOW TO S/G D A0555 STEAM GENERATOR A CA INLET TEMP A0561 STEAM GENERATOR B CA INLET TEMP A0567 STEAM GENERATOR C CA INLET TEMP A0573 STEAM GENERATOR D CA INLET TEMP A0938 STEAM GEN A FEEDWATER FLOW I A0950 STEAM GEN B FEEDWATER FLOW I A0867 STEAM CEN C FEEDWATER FLOW I A0879 STEAM GEN D FEEDWATER FLOW I A0454 CF TO S/G A TEMP A0455 CF TO S/G B TEMP A0449 CF TO S/G C TEMP A0448 CF TO S/G D TEMP

' A1109 NC LOOP A OVERPOWER D/T SETPOINT A1115 NC LOOP B OVERPOWER D/T SETPOINT A1121 NC LOOP C OVERPOWER D/T SETPOINT A1127 NC LOOP D OVERPOWER D/T SETPOINT A1059 STEAM CEN A MARROW RANGE LEVEL IV A1065 STEAM GEN B NARROW RANGE LEVEL IV A1071 STEAM GEN C NARROW RANCE LEVEL IV A1077 STEAM GEN D NARROW RANGE LEsJ.L IV

-A0628 POWER RANGE AVG LEVEL QUANDRANT 1 A0627 POWER RANGE AVG LEVEL QUANDRANT 2 A0629 POWER RANGE AVG LEVEL QUANDRANT 3 A0626 POWER RANCE AVG LEVEL QUANDRANT 4 -

A1061 NC LOOP A WIDE RANGE COLD LEG TEMP A1067 NC LOOP B WIDE RANGE COLD LEG TEMP A1073 NC LOOP C WIDE RANGE COLD LEG TEMP A1079 NC LOOP D WIDE RANCF COLD LEG TEMP A1058 NC LOOP A D/T A1070 NC LOOP B D/T A1082 NC LOOP C D/T -

A1094 NC LOOP D D/T A1064 NC IDOP A NARROW RANGE COLD LEC TEMP A1076 NC LOOP B NARROW RANGE COLD LEG TEMP A1088 NC LOOP C NARROW RANGE COLD LEG TEMP A1100 NC LOOP D NARROW RANCE COLD LEG TEMP 1

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. Attachment 2 (Cont)

A1085 NC LOOP A OVERTEMP D/T SETPOINT A1091 NC 140P B OVER11EMP D/T SETPOINT J

A1097 NC LOOP C OVERTEMP D/T SETPOINT

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A1103 'NC 140P D OVERTEMP D/T SETPOINT l A0965 NC LOOP A WIDE RANCE HOT LEC TEMP A0971 NC IAOP B WIDE RANCE HOT LEC TEMP A0977 NC LOOP C WIDE RANCE HOT LEC TEMP A0983 NC 140F D WIDE RANCE HOT LEC TEMP A1083 NC LOOP A FLOW I A1089 NC LOOP B FLOW I A1095 NC LOOP C FLOW I A1101 NC LOOP D FLOW I A1107 ST?.AM CEN A STEAM PRESS I A1113 STEAM CEN B STEAM PRESS I A1119 STEAM CEN C STEAM PRESS I A1125 STEAM CEN D STEAM PRESS I A1060 STEAM GEN A MAIN STEAM FLOW I A1066 STEAM CEN B HAIN STEAM FLOW I A1072 STEAM GEN C MAIN STEAM FLOW I A1078 STEAM CEN D MAIN STEAM FLOW I A1118 PRESSURIZER PRESS I ,

A1124 PRESSURIZER LEVEL I A1108 CF PUMP A DISCHARGE PRESSURE A1114 CF PUMP B DISCHARGE PRESSURE D1238 REACTOR TRIP BREAKER A D0084 REACTOR TRIP BREAKER B ,

A1112 NC LOOP HICHEST AVERACE TEMPERATURE A1106 REFERENCE TEMPERATURE, T-REF D0301 CF PUMP A TURBINE D0302 CF PUMP B TURBINE This list is taken from McGuire. The program at Catawba will be similar.

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Attachment 3 Note: This plot is taken from a McGuire event; however, "i the program at catawba MCGUIRE TRANSIENT MONITOR UNIT 1 will be similar.

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1200J [1200 1180J l-1180 1160-j '

[1160 1140j [1140 1120 5 F1120 AIII3 il A1107 1100-j [1100 h; 6 6 A PressiAre. M 1980 5 f1980 095 % ) 1960-2 E-1060

-1948 A1125 .

A1119 1040 5 5(.s D 5(s C.  ! 5-1920 1920 E E ,

1000-5 pEgangsmemir, E-1000 ,

980 980 viiigisi,juisigivingissigivingssing,,,,

0 5 10 15 20 25 30 35 40 I ' .

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Leac%< T;p haaker B START TIME OF TRANSIENT MONITOR PLOT: 18:04:19.6 DATE:08/12/83 .

LEGEND A1107 = SOLID LINE A1113 = DOTTED LINE LEGEND A1119 = DOT DASH A1125 = DASHED LINE j

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2.1 EQUIPMENT CLASSIFICATION AND VENDOR INTERFACE (REACTOR TRIP SYSTEM COMPONENTS) .

All components of the Reactor Trip System whose functioning is required to trip the reactor are identified as safety-related on documents, pro-cedures, and information handling systems used in the plant to control safety-related activities.

To address the issue of vendor interface for Reactor Trip System components, a distiaation is made between components supplied by Westinghouse and components supplied by other suppliers. Most of the Reactor Trip System components, including the Reactor Trip Breakers, are supplied by Westing-house. Vendor interface with suppliers other than Westinghouse is addressed in the response to Section 2.2. The following description addresses the interface with Westinghouse.

Duke Power has a continuing program in place for interface with Westinghouse which ensures that information concerning safety-related equipment is complete, current, and controlled throughout the life of the plant and appropriately referenced or incorporated in plant instructions and procedures.

Westinghouse transmits important information to customers via Technical Bulletins. These are received at Duke Power and processed by a single, vendor information coordinator. Duke confirms receipt of these transmittals using a receipt acknowledgement form. Transmittals not acknowledged by Duke Power are retransmitted. At least once per year, Westinghouse transmits a Technical Bulletin containing a list of current Technical Bulletins. Duke reviews this list to verify receipt of all applicable information.

Upon receiving a technical bulletin from Westinghouse, the vendor information coordinator transmits the information to the appropriate organization within Duke Power to determine applicability and safety significance. If applicable, the information is transmitted to the station staff for implementation. The station staff would incorporate the information into plant procedures and instructions, as appropriate. The vendor information coordinator maintains records of the disposition of each technical bulletin and provides follow-up to ensure timely action.

Letters received from Westinghouse which transmit technical recommendations are reviewed by the vendor information coordinator and, if appropriate, are processed using the same controls as for ,chnical bulletins.

Vendor service on Reactor Trip System components is either (1) controlled completely by the Duke Quality Assurance Program, or (2) the service is requisitioned using the procurement procedures for materials, parts, and components and performed under the vendor's quality assurance program.

In the latter case, Duke Power is responsible for identifying to the vendor that QA is required for controlling the activities and that 10 CFR 21 reporting requirements are applicable.

Most vendor service on Reactor Trip System components is performed by Westinghouse, since Westinghouse supplied most of the RTS components.

Within Westinghouse, all orders are screened to determine whether nuclear safety-related equipment, parts, or service are invclved. If so, the

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- orders are processed through the Water Reactor Divisions (WRD). All safety--

- related-service orders will be placed with divisions who are on an. approved supplier-list for the particular services requested. If no approved supplier is listed for requested services, the' order will ise processed through WRD.

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2.2 -EQUIPMENT CLASSIFICATION AND VENDOR INTERFACE (PROGRAMS FOR ALL SAFETY RELATED COMPONENTS) 2.2.1 EQUIPMENT CLASSIFICATION 1 &'2. The Catawba Nuclear Station Nuclear Safety-Related Structures, Systems, ~ and Components document provides the mechanism for the determination of whether or not a given station structure, system, or component is safety-related. This document will be superceded in 1984 with the Catawba Nuclear Station Quality Standards Manual for Structures, Systems, and Components which will incorporate additional quality standards.

This document and its revisions are approved and issued by the Vice President, Nuclear Production Department, with appropriate administrative procedu'res implemented to control distribution. Before operating licen'ses are. issued, the manual is also approved by the Vice President, Design Engineering and the Corporate QA Manager.

LThis manual provides the criteria to be utilized in determining safety-

.related structures. systems, and components (both mechanical and electrical).

.In general, an item is considered to be safety-related if necessary to assure:

(a) the integrity _of the reactor coolant pressure boundary, (b) The capability to shut down the reactor and maintain it in a safe

. shutdown condition, or (c)' the capability to prevent or mitigate the consequences of accidents-which could result in potential.off-site exposures comparable to the guideline exposures of 10 CFR Part 100.

In addition to the general criteria, the following considerations have been established as guidance in the determination of safety-related mechanical systems and components and are contained in the manual.

(a) Reactor coolant pressure boundary.

(b) Systems or portions of systems that.are designed for emergency core cooling, post-accident containment heat removal or post-accident fission product removal.

(c)~ Systems or portions of systems required for reactor shutdown or residual heat' removal.

L(d) Portions of'the main steam and feedwater systems extending from and including the secondary side of the steam generator and including the outermost containment isolation valve.

(c) Cooling water and seal water systems required for functioning of safety-related systems or components.

(f) Radioactive waste treatment, handling, and disposal systems, except those portions of these systems whose postulated failure would not result in off-site doses that exceed 0.5 rem whole-body or equivalent.

(g) Instrumentation that is essential to the performance of reactor protection and/or engineered safeguards function.

(h) Instrument piping connected to a nuclear safety-related system.

However, the connected instrument is not safety related unless it performs a necessary safety function.

(i)' Containers / packages utilized by a station for off-site shipment of fissile material and quantities of licensed materials in excess of type A quantities, as defined in 10 CFR 71.4(q), are considered safety-related.

(j) Piping and equipment with Duke classification A, B, C.

(k) First seismic restraint beyond the seismic category I boundary where non-seismic category I piping is connected to seismic category I pipirig.

(1) Required for control room habitability.

(m) Required for auxiliary feedwater to the steam generators.

(n) Containment Hydrogen Control System (hydrogen recombiners).

The classification of mechanical systems and components primarily considers the functional aspects of a system. In some cases, however, a system or portion of a system is considered safety-related because it serves as a pressure boundary. This would include such items as vessel shells, heads and nozzles, pipes, tubes and fittings, valve bodies, bonnets and discs, pump casings and covers, and bolting which joins pressure-retaining items.

Specifically excluded are items not associated with the pressure-retaining function of a component such as shafts, stems, trims, bearings, bushings, wear plates nor seals, packing, gaskets and valve seats unless these items perform a safety-related function other than pressure-retaining.

Systems and components which may contain radioactive material but whose postulated failure would not result in off-site doses that exceed 0.5 rem whole-body or equivalent, and that are not otherwise related to safe shut-down or accident mitigation, are classified as non-safety-related.

In addition to the general criteria, the following considerations have been established as guidance in the determination of safety-related electrical systems and components and are contained in the manual.

(a) Electric systems and components that are essential to shut down the reactor and limit the significant release of radioactive material

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following a design basis event.

(b) Controls and instrumentation systems and components that are utilized to develop a signal which is essential to initiate reactor protection and/or engineered safeguards functions.

(c) Instrumentation systems and components that are essential for post-accident monitoring as defined in the FSAR, Chapter 7.

(d) Cables and their support systems that provide power to or control safety-related components.

In addition to the general criteria, the following considerations have been established as guidance in the determination of safety-related structures and are contained in the manual.

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(a) The integrity of the reactor coolant pressure boundary.

(b) The capability to shut down the reactor and maintain it in a safe shutdown condition, or (c) The capability to prevent or mitigate the consequences of accidents which could result in potential off-site exposures comparable to the guideline exposures of 10 CFR 100.

3. Based upon the general criteria, and the specific criteria for all systems, components, and structures listed above, Duke has predetermined the safety-related status of many systems, structures, and major components.

This information is provided in the following formats as part of the manual:

(a) Tabulated listings of the plant systems showing the overall safety-related designation of each system and major components which are exceptions to the overall system designation.

(b) Tabulated listings of the plant structures showing individual areas as necessary with the safety-related status of each item indicated.

(c) Reference to station-specific drawings or documents (pipi*ag drawings, instrument list, etc.) which identify safety-related items at the component level.

(d) A criteria checklist for use when an item under consideration is not contained in either of the above formats (see Figure 1 attached).

Station personnel use the manual as the basis for determining the safety-related functions of structures, systems, or components. To expedite this process, station procedures (Station Directives) define the use of the manual. The Station Directives provide guidance in the following areas:

o Applicability of the manual.

o Responsibilities of the station organizational units in determining safety-related classification of a particular structure, system, or component.

o The procedure for use in applying the specific information contained in the manual or in using the QA Condition 1 (Nuclear Safety-Related) checklist (Figure 1 attached).

o Personnel Qualifications.

o Review and Approval Process If a structure, system or component is identified as being safety-related, in accordance with the provisions of the Quality Standards Manual for Structures, Systems, and Components, then the activities affecting the quality of the identified item are controlled to the extent consistent with 10 CFR 50, Appendix B. The Duke Power Company Topical Report " Quality Assurance Program Duke-1" describes the Duke program for conformance to 10 CFR 50, Appendix B.

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FIGURE 1

.QA CONDITION 1 (NUCLEAR SAFETY RELATED) CHECKLIST

. Page 1 of 2 (1) STATION:

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~(2) CHECKLIST APPLICABLE T0:

~ (3)' EVALUATION:

(3.1) Is ~ the component part of the pressure boundary of any of the following?

'Yes No a. Reactor Coolant System Yes No b. . Any_ system connected to the Reactor Coolant System out to the second isolation valve Yes -No c. An Engineered Safety Feature System as listed in Section 3.2 below Yes No d. The reactor containment Yes No e. A post-accident containment atmosphere cleanup system Yes No- f. A system which treats, handles and/or disposes of radioactive material, the failure of which could cause off-site doses greater than 0.5 rem whole body or equivalent (3.2) Is the component required for the functioning of or actuation of any of-the following?

Yes- No a. Emergency core cooling .

Yes No b. Residual heat removal from the reactor or. spent fuel pool Yes No c. Containment isolation Yes No d. Post-accident containment heat removal or atmosphere cleanup Yes No . e. Reactor shutdown Yes No f. A ' cooling or ' seal water system required for functioning of safety-related systems or ' components Yes No g. ~ Control room habitability

g. Yes No h. . Post Accident Monitoring-(as defined in FSAR, Chapter 7)

Yes No 1. Main steam and normal feedwater systems extending from and

. including the secondary side of the steam generator-and.

e including the outermost Containment isolation valves.

j . Auxiliary feedwater to the steam generators

k. Containment Hydrogen Recombiners (3.3) Is the' component required to assure any of the following?

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Yes No a. The integrity of the reactor coolant pressure boundary

'Yes No b. The capability to shutdown the reactor and maintain it in a safe shutdown condition

- Yes No c. "The capability to prevent or mitigate the consequences of accidents which could result in potential off-site exposures comparable'to the guideline exposures of 10CFR100 i

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8 FIGURE 1 (Cont'd.)

QA CONDITION 1 (NUCLEAR SAFETY RELATED) CIIECKLIST Page 2 of 2 (3.4) Is the item:

Yes ~ No a. A component significant to safety of a shipping container /

i-package for, pursuant to 10CFR71, fissile material of quan-tities of licensed material of excess of Type A quantities.

-Yes No b. A consumable / expendable item which is part of, or contained within, and affects the safety function of any component iden-tified in Section 3.1 above, j Yes No c. The first seismic restraint beyond the seismic category I boundary where QA Class A, B or C piping connects to QA Class E, F, G, or H piping.

If the answer to any of the above questions is "yes", then the component must Jbe treated as QA Condition 1.

(4) PREPARED BY: DATE:

! (5) REVIEWED BY: DATE:

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4. During the development of the quality standards manual, an inter-departmental review of the draft document was conducted to ensure completeness, technical accuracy, and application of the safety-related designation criteria.

Additionally, an established procedure exists in the Licensing Section Manual concerning revisions to the quality standards document. This procedure defines the followins aspects of the revision process.

(a) Organizational Responsibilities (b) Determination of Necessary Revisions (c) Frequency of Revision (d) Development of the Document Revision (c) Determination of Safety-Related Designation.

(f) Approval Process (g) Distribution and Documentation Duke Power Company has developed management procedures and controls which cover the utilization of the quality standards documents at each nuclear station. The Administrative Policy Manual for Nuclear Stations contains requirements for determining the safety-related status of an affected item in each of the following activities.

(a) Station Modifications.

(b) Station Maintenance Activities (c) Procedure Development and Revision (d) Procurement Each activity requires the determination and documentation of the safety status of the affected item. Appropriate management approval is received in the documentation package. If an item cannot be shown to be non-safety-related, based on the established criteria, it is considered to be safety-relate.d with all applicable QA requirements enforced.

All documentation packages (procedures, work requests, etc.) for activities affecting a safety-related structure, system, or component receive an inter-disciplinary station review including the Quality Assurance Department.

5. The program for ensuring that the appropriate design verification and qualification testing is specified during procurement of safety-related components is being reviewed to determine whether changes are needed. A description of the program, including any changes resulting from this review, will be provided to the NRC by January 15, 1984.

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. 2.2.2. VENDOR INTERFACE A Nuclear Utility Task Action Committee (NUTAC) has been established by.the; institute for Nuclear Power Operations _for the specific purpose of.~ defining an appropriate industry-wide vendor interface program.

- Duke Power is participating in the activities of NUTAC. Following

completion of this activity, currently planned for February,1984, Duke

, Power will submit _ a ' description 'of the vendor interface program.

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3.L': POST-MAINTENANCE TESTING (REACTOR TRIP SYSTEM COMPONENTS)

Existing procedures and programs require.that all safety-related and Technical Specification-related components be tested after. maintenance before being returned to service. This testing demonstrates that the-equipment;is capable of performing its intended safety function.

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_ Duke Power has reviewed the vendor technical information for the Reactor Trip Breakers and has verified that the information is appropriately incorporated'into plant procedures and Technical Specifications. For the remainder of the Westinghouse-supplied Reactor Trip System components, Duke is reviewing previous Westinghouse recommendations to ensure that they are appropriately incorporated into plant procedures. This review will'be completed by December 31, 1984. The response to Section 3.2 addresses the Reactor Trip System components not. supplied by Westinghouse.

Duke Power has no knowledge of any post-maintenance testing requirements

-in Technical Specifications which can be demonstrated to degrade safety.

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3.2 POST-MAINTENANCE TESTING (ALL OTHER SAFETY-RELATED COMPONENTS)

Existing procedures and programs require that all safety-related and Technical Specification-related componints be tested after maintenance before being returned to service. This testing demonstrates that the equipment is capable of performing its intended safety function.

Existing administrative procedures are established to control the distribution of vendor instruction manuals and to ensure their guidance is appropriately incorporated into plant procedures. However, changes to these administrative procedures are planned to resolve weaknesses identified during internal audits. Duke Power is implementing a substantial review to verify that (1) all vendor instruction manuals are complete and consistent, and (2) where inconsistencies are discovered, the correct information is appropriately incorporated into plant procedures.

This review will involve comparing all in-house copies of vendor manuals to ensure that all copies of a given manual are complete and consistent.

If differences are identified between the copies of a given vendor manual, the differences will be resolved - the vendor may be contacted, if necessary to resolve differences. Any changes made as a result of this review will be appropriately incorporated into plant procedures.

Duke Power has no knowledge of any post-maintenance test requirements in Technical Specifications which can be demonstrated to degrade safety.

,e 4.1 . REACTOR TRIP SYSTEM RELIABILITY (VENDOR-RELATED MODIFICATIONS)

.A Westinghouse letter dated March 31,-1983 recommended that several critical dimensions on the..undervoltage. trip attachment (UVTA) be verified. 1 0n April 21, 1983, a Westinghouse letter was issued requiring

'UVTA replacement with devices having modified shaft grooves for the retaining ring. !This replacement will be completed at Catawba before fuel loading'and_all UVTA's will be verified.to comply with the dimen-

.sional tolerances identified by Westinghouse.

No other vendor-related modifications have been identified for the Reactor-

. Trip System components supplied-by Westinghouse.

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f 4.2~ REACTOR TRIP SYSTEM RELIABILITY (PREVENTIVE MAINTENANCE AND SURVEILLANCE

-PROGRAM FOR REACTOR TRIP BREAKERS) 1.

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Preventive maintenance is performed on the Catawba Reactor Trip Breakers in accordance with'the current manufacturer's recommendations

.as described in the technical manual. This maintenance is currently performed once per 6 months. This frequency is under review and changes

'may be proposed later.

2. Trending of parameters is not currently recommended by the manufac-turer. -However, the cyclic life testing program described in item 3.

below, will provide information to determine the need for trending of parameters.

3. Life cycle testing of the shunt trip attachment and the undervoltage trip attachment of the reactor trip switchgear is being conducted by Westinghouse for the Westinghouse Owners Group. This program is aimed toward establishing 1the service life of these devices and substantiating periodic test requirements with proper maintenance. The results of this program will be incorporated into maintenance, replacement, and qualifica-tion programs. The test program is scheduled for completion in the second quarter of 1984.

4. Breakers or components will be replaced consistent with demonstrated life cycles.

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' 4.3 " REACTOR TRIP SYSTEM RELIABILITY (AUTOMATIC ACTUATION OF SHUNT TRIP ATTACHMENT FOR WESTINGHOUSE AND B&W PLANTS)

A modification will be implemented before fuel loading to provide automatic' actuation of the reactor _ trip breaker shunt trip attachments.

The modification was described in a letter from H. B. Tucker to Harold R. Denton dated April 18, 1983 and was discussed in'a Duke Power-NRC

-staff meeting on April 19, 1983. McGuire Safety Evaluation Report, Supplement 7 included a discussion of the modification. The modifica-tion will include an independent fusing design for safety-related and 11 'non-safety-related circuits similar to the design planned for McGuire.

The shunt trip attachments will be considered safety-related (Class 1E) for all future replacements, modifications, maintenance, and testing.

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.4.5 REACTOR TRIP SYSTEM RELIABILITY (SYSTEM FUNCTIONAL TESTING)

1. The capability to perform on-line functional testing of the Reactor Trip System will be provided for both Catawba units, including independent testing of the breaker undervoltage and shunt trip features. Testing will be performed in accordance with the Techni-cal Specifications, when issued with the facility operating license.

2. Not applicable.

3. The Westinghouse Owners Group in January, 1983, submitted WCAP-10271 to the NRC for review. WCAP-10271, " Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System" documents an evaluation of the impact on RPS unavailability of current and extended surveillance intervals.

The WCAP considered common mode failure, operator error, reduced redundancy during testing and equipment bypass. WCAP-10271 also considered correlative effects on plant operation and safety including the manpower expenditure associated with surveillance, the number of inadvertent trips which occur during testing and the distraction from plant monitoring on the part of the control room operator and shift supervisor associated with testing. Supplement 1 to WCAP-10271 which was submitted to the NRC in September 1983 is an extension of the evaluation and provides a discussion of component wearout caused by testing.

The NRC review of WCAP-10271 to date has resulted in a request for additional information the NRC felt necessary to complete the review. Information that will be submitted to the NRC in response to that request will include an.overall evaluation of the impact on plant safety of RPS surveillance, a discussion of the uncertainty

-of failure rates and common mode failure and more detail concerning the impact of surveillance intervals on RPS unavailability. WCAP-10271, Supplement 1 and the information provided to the NRC in defense of WCAP-10271 provides in a comprehensive form the information requested by item 4.5.3. The conclusion of WCAP-10271 and Supplement 1 is that although RPS unavailability is increased, less frequent testing of RPS components is warranted and will result in an improvement in overall plant safety and equipment reliability.

Upon completion of the NRC review of WCAP-10271, it is anticipated that appropriate changes to Reactor Trip System surveillance require-ments will be incorporated into the standard Technical Specifications for Westinghouse PWR's and that those changes will be incorporated into the Catawba Technical Specifications.

- _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _