ML20078R502

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Forwards Response to Generic Ltr 83-28, Required Actions Based on Generic Implications of Salem ATWS Events. Reactor Trip Investigation Program Will Be Implemented by 840101
ML20078R502
Person / Time
Site: McGuire, Mcguire  Duke Energy icon.png
Issue date: 11/04/1983
From: Tucker H
DUKE POWER CO.
To: Eisenhut D
Office of Nuclear Reactor Regulation
References
GL-83-28, NUDOCS 8311150172
Download: ML20078R502 (42)
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$ ~ e . a ng DUKE POWEn GOMPANY P.O. ItoX 33180 CHA54LOTTE, N.C. 28242 HAL H. TUCKER tes.erneown vers possament (704) 073-41531

.= u s =>.- November 4, 1983

' Mr. D. G. Eisenhut, Director Division of Licensing U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Re: McGuire Nuclear. Station

~ Docket Nos. 50-369, 50-370 3 3 ,.g~. Dear Mr. Eisenhut;

' .p a ~. ' 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 McGuire Nuclear Station, Units 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, o

D N ,

Hal B. Tucker REH:j fw Attachments cc: Mr. James P. O'Reilly, Regional Administrator Mr. W. T. Orders U. S. Nuc1 car Regulatory Commission NRC Resident Inspector

, Region II McGuire Nuclear Station 101 Marietta Street NW, Suite 2900

- Atlanta, Georgia 30303 Mr. R. A. Birkel Division of Project Management Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 8311150172 831104

< PDR ADOCK 050G0369

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_____-___ --__________-_._-__ _- -__-_______ J

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D. G. Eisenhut November 4, 1983 Page 2'

'bec: .K. S. Canady-N. A. Rutherford W. M. Sample-(MNS).

P. B. Nardoci L. R. Bledsoe (Toddville)

P. M. Abraham G. W. Hallman G. B. Caldwell B. T. Faulkenberry R. C. Futrell P. H. Barton J. S. Warren R. S. Hosard (W)

G. A. Copp R. E. Harris M. D. McIntosh (MNS)

W. H. McDowell C. A. Little

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

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."'&4 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 Nuc1 car 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 reactor 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.
2. Identify unexpected, abnormal responses to the trip by plant systams, 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 reactor trips, where no abnormalities have been identified, need not proceed to the subsequent investigation 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 periormance 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 post-trip behavior of key parameters that reflect overall plant performance and will identify abnormal performance 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 1evel 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.
4. The recommendations for corrective actions identified during the post-trip review are resolved.

Independent Review 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:

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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'- 4 Subsequent Investiaation 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 fons 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 lor engineering.

Data Acquisition The . essential element of the Reactor Trip Investigation Prearam 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:

. S (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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g ATTACHNENT A, RESPONSE TO GENERIC LETTER 83-28, ITEM 1.1 Station Directive DUKE POWER COMPANY NUCLEAR STATION INVESTIGATION OF REACTOR TRIPS 1.0 PURPOSE This directive defines the action to be taken in investigating reactor trips to ensure 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 l to their fully inserted position. Specifically excluded from this definition are control rod drop tests performed at power levels below 5%

full power.

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

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l (a) The trip was initiated and specifically called for in an in-progress

. test procedure, or was part of a planned unit shutdown, and l (b) No anom'alies 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, l or Operations Duty Engineer (hereinafter " Operations"), shall notify the Station Reactor Engineer or Performance Reactor Unit / Duty Engineer of the, trip. Operations sha11' 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 SFO Logbook.

The Station Reactor Engineer, or an engineer trained and qualified by the Reactor Engineer, (hereaf ter Reactor Engineer) shall perform a 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.

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o 4.0- IMPLEMENTATION

'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 promptly.

The investigation process described in 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.

fhe 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 is identified and resolved and shall provide criteria and guidelines defining the range of expected plant responses.

4.2 Restart Decision 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

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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.

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'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.

4.41 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 Directive 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 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.
5. As assessment of other equipment failures that contributed to the event.
6. As assessment of personnel performance and procedure adequacy 4: 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 L

Personnel performing Post-Trip Reviews chall 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. i c

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 frip 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 Secondary RCS Tave, each loop SG Pressure Pressurizer 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) SPDS " problem" indication (when installed)

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

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m (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 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 MCGUIRE/ CATAWBA NUCLEAR STATION DATE:

POST-TRIP REVIEW REPORT PREPARED BY:

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

(c) Description of. Transient:

Revised 10/83

2 HCGUIRE/ CATAWBA NUr"'AR STATION POST-TRIP REVIEW IT 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. PZR Level (RCS Makeup)
6. MFW Pump Speed Control

'7. SG Level Control

.8. SG Pressure Control

9. Turbine Control
10. PZR PORV Block Valve (0/C)

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

1. RPS
2. ECCS
3. Containment Cooling & Spray
4. Auxiliary Feedwater
5. Emergency Power (h) Testing in Progress:

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1 MCGUIRE/ CATAWBA NUCLEAR STATION POST-TRIP REVIEW REPORT III. Plant Response A. Sa'fety System Actuation and ' Performance

'I . Solid 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. Tur'oine 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 Contairaent Spray - Initiating Signal Time

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Cont.IV ent. Isolation 9

Initiating. Signal- Time Ice Condenser Doors Open Time

, 6. Auxiliary Feedwater - Initiating Signal- Time MDAFWP #1 Started at Stopped at I
  1. 2 . Started at Stopped at TDAFWP. Started at Stopped at
7. Pzr Code Safety Valves -

Actuated Time

8. Main Steam Safety Valves -

f Actuate'd Time i

Which Valves?

9. SPDS Problem' Indication.-

Type Time

-B. Control Systems Actions t

.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 Pzr Heater response normal? (Yes/No)

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MCGUIRE/ CATAWBA NUCLEAR STATION

~ POST-TRIP REVIEW REPORT e i

- 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?

JAuto 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 4

Low Tave-Time

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6 HCGUIRE/ CATAWBA NUCLEAR STATION POST-TRIP REVIEW REPORT 1

L C. ' Manual Actions - were any control stations taken from auto to

, manual? . (Specify station and s time / sequence.)

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Other manual actions:

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13. Radiological Response - (Describe any unexpected behavior, such as

< unexpected EMF response) 4 i

a 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-4 Pressurizer Pressure Pressurizer Level 4

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y 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.)

B. Discussior 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)

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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 Precedure signed off, time __

Reactor Critical, time Turbine On-Line, time

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. When an input alarms, this fact is printed mit 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 tine, 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.

Transient Monitor The transient monitor is a program on the plant computer at McGuire.

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 table 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 .) Hard copies of graphs of major plant parameters are normally kept.

d-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 available 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 In addition, some Condensate Booster Pump and Main Feedwater Pump status alarms (e.g. low suction pressuce) are on the events recorder.

Attachment 2

. 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 GEN A NARROW RANGE LEVEL IV A1065 STEAM GEN B NARROW RANGE LEVEL IV A1071 STEAM GEN C NARROW RANGE LEVEL IV A1077 STEAM GEN D NARROW RANGE LEVEL IV A0628 POWER RANGE AVG LEVEL QUANDRANT 1 A0627 POWER RANGE AVG LEVEL QUANDRANT 2 A0629 POWER RANGE AVG LEVEL QUANDRANT 3 A0626 POWER RANGE AVG LEVEL QUANDRANT 4 A1061 NC LOOP A WIDE RANGE COLD LEG TEMP A1067 NC LOOP B WIDE RANCE COLD LEC TEMP A1073 NC LOOP C WIDE RANCE COLD LEC TEMP A1079 NC LOOP D WIDE RANGE 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 LOOP A NARROW RANGE COLD LEG TEMP A1076 NC LOOP B NARROW RANGE COLD LEG TEMP A1088 NC LOOP C NARROW RANGE COLD LEG TEMP A1100 NC LOOP D NARROW RANGE COLD LEG TEMP T

-, _ , . _-- ,, . . . . - - .,.n . - _ _ . _ . . . _ .-. . - . , , - . . , - - - - - . . , - , . , , ,

Attachment 2 (Cont)

A1085 NC LOOP A OVERTEMP D/T SETPOINT A1091 NC LOOP B OVERTEMP D/T SETPOINT A1097 NC LOOP C OVERTEMP D/T SETPOINT A1103 NC LOOP D OVERTEMP D/T SETPOINT A0965 NC LOOP A WIDE RANCE HOT LEC TEMP A0971 NC LOOP B WIDE RANGE HOT LEG TEMP A0977 NC LOOP C WIDE RANGE HOT LEG TEMP ,

A0983 NC LOOP D WIDE RANGE HOT LEG TEMP A1083 NC 140P A FLOW I A1089 NC LOOP B FLOW I A1095 NC LOOP C FLOW I A1101 NC LOOP D FLOW I A1107 STEAM GEN A STEAM PRESS I A1113 STEAM GEN B STEAM PRESS I A1119 STEAM GEN C STEAM PRESS I A1125 STEAM GEN D STEAM PRESS I A1060 STEAM GEN A MAIN STEAM FLOW I A1066 STEAM CEN B MAIN STEAM FLOW I A1072 STEAM GEN C MAIN STEAM FLOW I A1078 STEAM GEN 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 HIGHEST AVERAGE TEMPERATURE A1106 REFERENCE TEMPERATURE, T-REF l D0301 CF PUMP A TURBINE l D0302 CF PUMP B TURBINE l

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- Attachment 3 s

MCGUIRE TRANSIENT MONITOR UNIT 1 1220 - 'buire. b i\ \ % dror int oa %l\1 33 5-1220 1200-j l-1200 1180-j l-1180 1160-j -1160 1140-j l-1140 1120-j l-1120 A1107 1100-j A1113 l-1100 5 6 A ?resstAre. 1880

- 1080 N6 LPS\G) 1060-5 g- 10 A1119 1040 -j A1125 l-1040 5(s C 1820_j l-1820

%D 1000[ w

~ l-1000 980 iiiigviiiivisiquiiigivingiviijiiiijiii, 980 0 5 10 15 20 25 30 35 40 MINUTES 0-hach< T;p Braake r 3 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 i

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6 2.1 . EQUIPMENT CLASSIFICATION AND VENDO2 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 pctivities, s To address the issue of vendor interface for Reactor Trip System components, a distinction 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.

DukePowerNasacontinuingprograminplaceforinterfacewithWestinghouse 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 technical 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,s 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 involved. If so, the

J 4 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 j is listed'for' requested services, the order will be 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 McGuire 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 about November, 1983 with the McGuire Nuclear Station Quality Standards Manual for Structures, Systems, and Components which will incorporate additional quality standards.

This document and its revisiens are approved and issued by the Vice President, Nuclear Production Department, with appropriate administrative procedures implemented to control distribution.

This manual provides the criteria to be utilized in determining safety-related structures, systems, and components (both mechanical and electriccl).

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.

(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.

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

(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.

I (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 piping.

(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, 7 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.

'ystems and components which may contain radioactive material but whose S

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

'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 prcvide 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 neceasary with the safety-related status of each item indicated.

(c) Reference to station-specific drawings or documents (piping 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 camponent 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.

FIGURE 1 QA CONDITION 1 '(NUCLEAR SAFETY RELATED) CIECKLIST Page 1 of 2 (1) STATION:

(2) CHECKLIST APPLICABLE TO:

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

Yes No i. Main steam and normal feedwater systems extending from and including the secondary side of the steam generator and 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?

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

FIGURE 1 (Cont'd.)

QA CONDITION 1 (NUCLEAR SAFETY RELATED) CHECKLIST Page 2 of 2 P.

(3.4) Is the item:

Yes___No a. A component significant to safety of a shipping container /

d 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.

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 be treated as QA Condition 1.

(4) PREPARED BY: DATE:

- (5) REVIEWED BY: DATE:

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,n. - - - - - , . , , - - . . . . . - . , - , ~ . , -- . , . , . - .

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4. .During.the development ofLthe quality standards manual, an inter-departmental review-of the draft document was conducted to ensure completeness, technical accuracy, and application of the' safety-related 1 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 precess.

(a) Organizational Responsibilities.

(b) -- Determination of Necessary Revisions (c) Frequency of Revision (d) - Development of the Document Revision (e) ' 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 l(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-related 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 .ts being reviewed to determine whether changes are needed. A

~ description af the program, including any changes resulting from this review, will be provided to the NRC by January 15, 1984.

- 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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3.1 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.

Duke Power has reviewed the vendor technical information for the Reactor s 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 components 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. arc 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 ir. formation 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.

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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. On April 21, 1983, a Westinghouse letter was issued requiring UVTA replacement with devices having modified shaft grooves for the retaining ring. This replacement has been completed at McGuire and all UVTA's have been verified to comply with the dimensional tolerances iden-tified by Westinghouse.

. Westinghouse issued Technical Bulletin NSD-TB-75-02 on February 20, 1975 concerning a circuit breaker wiring bundle problem. Preoperational testing of balance of plant DS breakers at a nuclear plant in construction prior to 1975 had, revealed an area of a possible circuit breaker wiring bundle problem that could also apply to reactor trip breakers that were shipped from the manufacturer prior to 1975. This would have applied to McGuire. Breakers shipped after that time had the problem resolved prior to delivery.

Due to the size of the control wiring bundle in the upper left hand corner of the breaker and the use of panel mounting screws of excessive length at this point, the control wiring insulation may have been damaged. This potential defect was noted at the factory and a change was made whereby the 1/4 inch diameter open Tinnerman nut was changed to a pressed-fit round nut with a .192-32 blind tapped hole. The rear projection of this nut was of such depth that a large bundle of control wiring could be jammed into the area between the nut and the shunt trip device of the breaker (if so equipped) and apply pressure to the wire bundle. There was no abrasive damage to the wiring insulation as a result of this latter condition.

The recommendation for the remedial action was as follows:

If the size of the wire bundle in'the area in question is too large to fit loosely into the area, it should be sub-divided into two or more bundles.

On those breakers'having the 1.4 inch diameter Tinnerman nuts, all front panel screws should be 1/2 inch long. Duke has verified that the recommended remedial action of Technical Bulletin NSD-TB-75-02 has been taken for the Reactor Trip System at McGuire Units 1 and 2.

No other vendor-related modifications have been identified for the Reactor Trip System components supplied by Westinghouse.

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- 4.2 REACTOR TRIP SYSTEM RELIABILITY (PREVENTIVE MAINTENANCE. AND ' SURVEILLANCE

' PROGRAM FOR REACTOR' TRIP-BREAKERS)

1. Preventive maintenance is performed on the McGuire 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 as required inr the McGuire Unit 2 operating license. 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 the 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 l 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 was implemented in April, 1983 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 Apri1~18, 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. As a result of the NRC staff review, a modification will be implemented to provide for independent fusing for safety-related and non-safety-related circuits.

This modification will be implemented before startup after the first refueling outage for each McGuire unit.

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 REACTCR TRIP SYSTEM RELIABILITY (SYSTEM FUNCTIONAL TESTING)

1. On-line functional testing of the Reactor Trip System is performed for both McGuire units as described in the McGuire Unit 2 operating license.
2. Not applicable.
3. Duke Power plans to submit proposed Technical Specification changes which are ccnsistent with the conclusions of WCAP-10271. The Westinghouse Owners Group in January, 1983, submitted WCAP-10271 to the N"C 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 I to WCAP-10271 which was submitted to the NRC in September 1983 is an extension of the evaluation and provides a discussion of com-ponent 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 unavail-ability is increased, less frequent testing of RPS components is warranted and will result in an improvement in overall plant safety and equipment reliability.

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