ML20149F667

Proposal ATWS-Recirculation Pump Trip (ATWS-RPT),Pilgrim

ML20149F667
Person / Time
Site:	Pilgrim
Issue date:	03/31/1979
From:	GENERAL ELECTRIC CO.
To:
Shared Package
ML20149F593	List: ML20149F589 ML20149F667 ML20149F940
References
FOIA-87-644 416-4208-HK1, NUDOCS 8801140298
Download: ML20149F667 (14)

Text

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l PROPOSAL ANTICIP.\TED TRAE!Dif$ WITHOUT $0tAf6 ItECIRCULATION PUPF TRIP (ATWS-RPT)

PILGRIM 4164206-NK1 March 1979 o

O w GDIERAL ELECTRIC COPFANY NUCLEAR ENutGY P90KCTS O!V!$10N BWR SERVICES DEhu mqENT

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PROPflETARY INFONET10N HOTICE .

',l This proposal dKunen contains propri ry infocustion of the General Electric y and is fWnished f in confidence solely for use in consi ing tyscrits of the proposal and for no other' direct or i rec ( vse. By accepting this dece-arrot from General Electricy recipient agrees: (1) to use thisdocumentandtheinfdras on it contains esclusively for theabovestatedpuM.(2) avoid publicatfor, or other ,

unrestricted disc 1 pre of this vment or the internation it contains. (3) make no copies of ny part hereof withovi the

- prior writ ssion of General tric, and (4) to return to this t when it is no longer for the purpose for n which f 1shed or upon the request of Gene 1 Electric.

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1. Background The ist has issued the requirement for installation of an Anticipated Transients Without Screm.necirculation Pump Trip (A1WS-IPT) modificatten for all operating plants within tuo years. Further, they have indicated '

licensing acceptance of the General Electric Company *Monticallo ATWS.IPT" l l

design for all operating plants. Finally, they have requestad the vtility's implementation schedule by April g. Ig7g.

The General Electric Company proposes to furnish the plantisprovement

• progren described in Section 4.5 of Topical Report NE00-25016, *lvalvation of Anticipated Transients Without Scree for the Monticello Nuclear Generating Plant", to meet this requirement.

2. Deste Ob.iectives, The intent of the A1WS-RPT design is to meet the following objectives:

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1. Shut down all recirculation pump motor generator sets with redundant .;

" logic from the following inputs:

LD O M3 ykh b) Reactor Vessel liigh Pressure c) Ibactor Low Water Level

- 2. The systeni is to be diverse from the Reactor Protection System (RPS).

4 3. The systas is to be ttstable in service.

4. The system is to be designed so that as much as possible no O single component failure can prevent the tripping of both o recirculation pumps.

5. The hardware should be high quality and environmentally qualified.

6. The system's per'orssaca characteristics are as follows:

a) Logic delay for trip, including dynamic response of the sensors. logic, action of < 0.53 seconds the breakers and collapse of the generator field

' b) Low level delay timer (to be confirmed by plant unique analysis}

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3. Justification Economic justification for the utility to accept the proposal results from successful realization of the objectives discussed ta Section 2. '

Further, the A1W5-IFT system taprovement package is a standard, generic design, with appropriate licensing report. ,

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4. Har here Description

1. Transmitterf, The reacter pressure and reactor unter level will be sensed by analog transmitters. The pressure and level tremonitters used are designed and manufactured to General Electric specifications.

Calibretten and shetoff valves. 3-way balancing ammifolds and fittings are provided for field assembly. See Figares (1) and (2) en for typical sketches of a pressure Transmitter and Differential Pressure Transmitter Valve seasses61y.

e t, 2. Trip units o The trip unit used to generate the trip signal is designed and manufactured to Geners) Electric specifications. Features of this system are:

1. Trip units are functionally tested or calibrated in place

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Vseens of a portable readout asses 61y.

• 2. Trip unit calibration current is a controlled rag of 1 N/sec.

< 3. 1he output of the trip unit is a voltage to a Class 1E relay.

c. 4. De master trip unit has a display meter stich monitors the treassitter current for gross failums as mall as failures o

la the current to voltage and filtering secticas of the easter trip unit.

5. Trip unit "out-of-card file annunciaties" is provided.

6. The trip unit provides long'-tem trip potat stability.

7. The calibrator is capable of introducine a step current for transient testing of the logic.

8. De readout anee61y is a portable unit thich may be used with all calibration units. This provides for portability of the secondary standard from card file to card file to seesure the value of the trip point in various cabinets.

9. Card extenders, a bench test unit, and calibration units are

, sw11ed to facilitate anintenance as regstred.

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10. The trip relay is a general purpose miay with four fore C contacts which is rated for the aetent environment at its mounting location.

3. Cabinets General Electric designed cabinets which house the trip unit card file, DC power supplies, and trip relays are illustrated in Figsms 3 and 4. The large cabinet (Fipwre 3) has space to house up to three trip unit card files thereby permitting use of the cabinet for future installation of the Analog Trip System or other plant improvosents. The small cabinet (Figure 4) has space for only one card file. Each carif file has space to accommodate twelve trip unt,ts. Additional card files can be factory supplied for large c'abinets per the price tabulation in the quotation letter, e or added later by field installation.

a3 The 25 volt DC power supply used is a ferro resonant type which to is recognized for its high reliability. Two complete power o supplies are mounted in each trip unit cabinet and connected in parallel. With this arrangement a single power supply can fail without affecting the trip function. The power supplies in the

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large cabinet have sufficient capacity to supply power for the maximum nueer of trip units (36) tht may eventually be housed 4

in a large cabinet. The power supplies for the small cabinet o have sufftcient power to swply the 12 trip unit capacity of that a cabinet. .

O The cabinets any be located in the control room, auxiliary equipment ,

room (1stion of existing logic cabinets), or in the vicinity of '

the local instrument rock in the reactor building. The cabinets mest be located in an area where seximum adtent temperature will be between 40-145'F. The electrical components within the cabinet

. are qualified to operate up to 150'F at 995 relative humidity.

The cabinets are seismically qualified per Itti-344,1975, to the safe Shutdown Eartigske ($$t) acceleration response spectra shoun on Figure 5. The large cabinets am floor mounted; the small cabinets are voll mounted.

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An additional DC to AC inverter is required is each ECCS division to sinable the Ahts wT trips to have their backup prer sewce from the station batteries. The inverter is to be located and

• installed by the purchaser extemal to the trip cattants.

5. System Description and Ace 11 cation _

$1nce normal scres is assmed to be unavailable for redmetas the r power, and since the transient event is one in which peuer reduction is ;

necessary, the Ahf5-WT system provides another method of reGucing pouer

1. for the first 15 wccads of the event. The trip of both sectreulation This e pumps causes a qwick reduction in core flow thus reducing the pouer. '

quick power reduction briags the reactor pressure, newtres flux and fWel tn surface heet flux doun in time to acceptably limit the peak pressure, o clad oxidatim and peak fuel enthalpy.

The initiating vertables, reactor pressure and reacter stor level, are sensed by analog transmitters. The transmitters are monsted on reactor pressure instroentation racks and connected through catihrstion and isolation manifolds to existing instrument sensor piping to the reactor l

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flanifold and calibration valves are applied to facilitate insta11at".on of the transmitter to existfag process piping.

C o Figure 6 is a schematic diagram sheising the arrangement W the transm and the divistoast separation. The trip units are mounted is cabinets f

dich conform ta ECCS separation criteria.1he trip suits provide the I inputs to the logic which trips both Recirculation Pump IIstor. Generator Sets' field braiskers thus shutting down the recirculation pops.

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6. General Electric Responsib111 ties

1. Prepare cabinet outline drawingh

• 2. Design, build and wire check the trip unit cabinets. -

3. Provide the hardware listed in Table 1.

4. Prepare Elementary Diagress showing devices added with location, termination numbers and intartonnection per the optians in the quotation letter.

5. Provide instruction annuals.

6. Provide an ATW5-RPT system specification.

3 7. Provide transmitter and related valve assembly drawings.

' 8. Provide an installation specification with installation, calibration.

and startup instructions.

3 9. Provide an ATW5-RPT system description suitable for licensing submittal .

• Cabinet design will vary with the cabinet option purchased.

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7. Purchaser Responsibilities

1. Physically locate and fasten the cabinets securely to the floor or well. as required (if cabinets are supp11e1). ~

2. Run raceways from the trip units to the transmitter racks, as required. ,

3. Mount. assemble and pipe the transmitter / valve subassembly to the process piping. Leak test the new transmitter installation.

4. Make electrical connections between tfm transmitter to the trip unit and the trip unit relay to trip coils and vent solenoids.

5. Run power cables from the existing logic cabinets to each trip l unit cabinet. Existing spare cables that have been made available from other plant modifications may be used.

p 6. Run annum;iator loop cables from the control room to each trip unit cabinet.

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7. Revise the plant unique P&l0's. FCO's. Elementaries. ID's, etc.

8. Provide labor for installation. -

9. Provide all Itcensing for the application of the hardware into the power plant. .

- 10. Checkout and placing in' service. .

11. Verify the seismic spectrum the cabinets were tested to (Figure 5) o is greater than the floor /ws11 response for the mounted location of the cabinets.

c4 12. Perfore seismic analysis as may be required to verify the seismic O conditions at the locationswhere the transmitters will be installed do not exceed the alloweble equipment design limits.

13. Verify the environmental conditions at the locations where the equipment will be installed do not exceed the allowable equipment design Itaits.

14. Incorporate new surveillance testing requirements for the analog trip units and transmitters into the technical specification's.

15. Provide detail test and calibration procedures based won GE selled instructions.

16. Locate and install the DC to AC inverters external to the trip cabinets.

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8. Ma.1or Ceg:r.;;ts Included in this Proposal j Table 1 is a list of anjor har& art items which will be supplied by the j General Electric Company.

- TABLE 1 hjor Mar &are items $v901104 1

Description Itas_ Qantity p.<'#p p 4 's

+I Level Transmitter. Manifbid Valve & Fittings 1

4 Pressure Transadtter. Entwo and Fittings 2

8 Trip Unit itith Analog estput g 3 p 4 gf2- Trip and Alare Relays Ih - N+c ^ ^ - ~

" 6 4 Manual Initiation Pusl6stion C 7 As Required Light & Holder 8 As Required Fuse 9 5 Trip Coil for M Set Fleid Breakers 2

AC/DC Power Supply 10*

3 Readout Assedly 11 12 3 Card Extender 13 1 8ench Test Unit C 14 2 Card File C. 15 2 Calibrator Cabinet j 16* 2 17 2 DC/AC Inverter 18 4 Isolation valve Z-D k ki ge b M f r Y 0 > fses 4

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• Capacity will depend on cabinet option purchased.

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9. Quality Assurance l 1. Equipment and Services shall be provided in accordance with the ,

General Electric IWt plity assurance program as described in  !

Topical Report E00-11209-04A. *

2. The provisions of 10 CFR part 21 apply. l

3. A product Quality certification (poc) sha11 be provided by i General Electric as the primary que11ty assurance record for Equipment classified as "taportant to safety". l

4. SWR ouner access requirements for audits and/or witness of inspection points for supplied Equipment and Services shall be arranged upon request at autua11y 4Feeable tems.

, 5. All General Electric teclear Energy Divisions' work at the BWR M omer's plant site shall be under the cognizance of the SWR ower .

> quality assurance proram.

17 g 10. Licensing E A description of the ATWS-RPT plant improvement retrofit package suitable for Itcensing submittal will be s@ plied. Additional detailed assistance t 'I can be provided on a cons'ulting bcsis at our comercial rates.

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.;_ , June 25, 1985 -

Docket No. 50-293 n

DISTRIBUTION Docket File ORBl2 Reading OR8#2 Wrkg File MEMORANDUM FOR: Richard W. Starostecki, Director Pleech Division of Reactor Projects DVassallo

', Region I Glainas HThompson FROM: RXarsch Hugh L. Thompson, Jr., Director Division of Licensing JCarter Office of Nuclear Reactor Regulation

SUBJECT:

LICENSING ACTION REVIEW Re:

Pilgrim Nuclear Power Station Your assistance is requested in conducting a review 3.1.2, 3.2.1 and 3.2.2, and 4.5.1.

TAC numbers 52948, 53785 and 54095, respectively.This work should be con Enclosed 1984, which are copies address of BEco the above items.letters dated November 7,1983 and June 2 If further infomation is needed (FTS)492-4952.from the licensee, it should be obtained via Project Manager P In accordance with NRR Office Letter No. 44, each safety evaluation perfomed by a technical division shall have a separate SALF input provided.

For the purposes of these reviews, the Regional personnel are considered part of the technical divisions. Therefore, we are evaluation performed. requesting that your forwarding memorandum contain a As discussed with Don Haverkamp of your staff, the requested target date for completion of your review is March 31, 1986. The period from now until that date was established in anticipation of the need for additional information from the licensee.

a revised date as soon as possible.Please confirm this target date or provide Ori-inal Signed by FJMiragl f a for/

Hu-h L Thompson,Jr.

Hugh L. Thompson, Jr. , Director Division of Licensing Office of Nuclear Reactor Regulation

Enclosures:

As stated DL:0RB#  : RBf2 OL- -OR Pleech:a s DVassallo G nj s 06/A',r85 064y/85 '

T son 06/g/85 06 85 1 _ -

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l Generic Letter 83-28 1.1.1 Restart of Pilgrim Station is acceptable only if the following criteria are satisfied:

1) The cause of the unscheduled reactor shutdown has been detemined and appropriately corrected or the Operations Review Committee (ORC) has )

detemined that the safety of equipment, site personnel, and the pub-lic are not threatened by a restart, based on an independent review (see response to 1.1.6).

2) Components within systems designed for automatic response to abnormal  !

parameters did indeed respond properly to the appropriate initiating signals, or exceptions evaluated and approved by the appropriate admin- i istrative controls (ORC, Nuclear Engineering Department, Safety Evalua-  !

tion).  !

3) The station manager, or his designated alternate, has given approval to commence restart.

1.1.2 Post-trip review activities are conducted by the on-duty Nuclear Operating Supervisor and Shift Technical Advisor under the direction of the Nuclear Watch Engineer. If the cause of the trip is not readily apparent, or cannot be determined beyond a reasonable doubt, the Chief Operating Engi-neer or Day Watch Engineer will take charge of the investigation until the cause of the trip has been determined. The individual in charge of the investigation will take responsibility for making appropriate recom-mendations to the station manager, or his designated alternate, based on the criteria of 1.1.1 above.

1.1.3 The qualifications of the facility staff are addressed in Section 6.3 of Pilgrim's Technical Specifications. This requires that the requirements of ANSI N18.1-1971, "Selection and Training of Personnel for Nuclear power Plants" be met.

The Shift Technical Advisor is, as a minimum, qualified to the requirements described in NtREG-0737 1.1.4 Output of the alarm typer on the plant process computer provides the pri-mary information source for chronological reconstruction of the sequence of events surrounding a trip occurrence. Recorder strip charts are uti-11 red for evaluation of long-ters trends which may not be indicated by the alam typer output. The balance of essential infomation is provided by plant personnel, whose combined knowledge is relied upon for reconstruc-

. tion of human activities prior to and during the trip event (maintenance activities, operator actions, etc.).

l l 1.1.5 Technics 1 Specifications identify the trip settings for Reactor Protection, Containment Isolation, and Energency Core Cooling Systems actuation. The information sources identified in 1.1.4 above indicate the presence of any l

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I signals exceeding those trip settings. Expected plant behavior is based on system design responses described in the FSAR and on the experience and training of operators and supervisors.

1.1.6 If the post-trip task force identified in 1.1.2 above is unable to estab-lish the cause of the trip, the Operations Review Comittee is convened to provide independent assessment of the event. The ORC utilizes the same infomation as the task force. Both groups may call on the technical and engineering expertise of other personnel in the Nuclear Organizatien or other appropriate group, internal or external.

1.1.7 We are in the process of reviewing a draft INPO procedure concerning post-trip reviews for possible incorporation into a Nuclear Operations Pro-cedure (NOP). Currently, Section 6 of NOP8301, "Conduct of Operations,"

deals with this issue (Attachment A).

1.2.1.1 The GE-PAC 4020 Process Computer System provides on-line monitoring of several hundred input points (digital, analog, and pulse) representing significant plant process variables. The system scans digital and analog inputs at specified intervals and issues appropriate alam indications and messages if monitored analog values exceed predefined limits or if digital trip signals occur. It perfoms calculations with selected input data to provide the operator with essential core performance infomation through a variety of logs, trends, sumaries, and other typewriter data arrays. The Sequence of Events printer responds to digital signals, the Osta Recall Log is analog.

1.2.1.2 The monitored parameters are listed in Attachment B.

1.2.1.3 The log gives a time field for events in hours, minutes, seconds and then to the nearest 1/60 (.0166) second.

1.2.1.4 The format for displaying data and infomation is as follows:

Sequence of Events:

Time Cycle Point 10 Name Status XXXXXX XX XXXX XXXX Date Recall Log Time Poi nt 10 1 . . . . . . . . . . . . . . . . . . . . . . . . . . Poi nt ID 19 (Value)

XXXXXX XXXXXX ............................ XXXXXX The Data Recall Log prints values preceding the event in black (2.5 minutes before event), values following the event are printed in red (2.5 minutes after the event).

.i i 1.2.1.5 The computer has core storage to record the status of the first 80 events in sequence for the NSS log and the first 20 events in sequence for the B0P log. The data is typed on the alarm typer. During this interval of time, (approx. 5 min.) all change of contact status is ignored. At the termination of the typer routine, the program rescans all digital points again and outputs any change of status on the typer, and reinitializes the program again.

The hard copy of the printout is controlled by the Document Control Group and is retained in the records vault at Pilgrim.

1.2.1.6 The power source for the process computer is non-interruptible and is non-Class IE.

1.2.2.1 As described above, the GE-PAC 4020 Process Computer System provides the major information source for assessing the time history of analog vari-ables. Additionally, several variables (see 1.3 response) are recorded on strip-charts for trend evaluation.

1.2.2.2 The Data Recall Log monitors up to 38 preselected analog points, which are scanned continuously at 5-second intervals.

The Data Recall Log currently supplies information on the following:

APRM Channel 'A' APRM Channel "C" Reactor Pressure Core Plate P Reactor Core Flow Control Rod Drive Flow Reactc,r Feedeater Flow "A" Loop Reactor Feedwater Flow "B" Loop Reactor Water Level (inches)

Outlet Steam Flow

Feedwater Teg erature 'A' Loop l Feedwater Temperature "B" Loop

Recirculation Flow "A" Loop Recirculation Flow "B" Loop Reactor Saturation Tegerature i Calculated Seawater Flow l Hotwell Outlet Temperature Dry; ell Temperature (64' elevation)

Suppression Chamber Level Stator Cooler Header Inlet ('C)

Stator Cooler Header Outlet ('C)

Alternator Air to Cooler ('C)

Alternator Air from Cooler ('C)

Condensate Demineralizer Differential Pressure Reactor Feedpump Suction Pressure Condensate Pump Discharge Header West Condenser Pressure (inches Hg)

East Condenser Pressure (inches Hg)

8 Reactor Building Closed Cooling Water System (RBCCW) "A" Loop Flow RBCCW "B" Loop Flow RBCCW Residual Heat Removal (RHR) Heat Exchanger Loop "A" Flow RBCCS RHR Heat Exchanger Loop "B" Flow ~~'

RBCCW "A" Outlet Temperature RBCCW "B" Outlet Temperature Torus Pressure Drywell Pressure Service Water Loop "A" Flow Service Water Loop "B" Flow 1.2.2.3 The above parameters are stored in a special Scan Table section of com-puter memory for 2.5 minutes. Upon occurrence of a designated plant trip event, the data currently in the Scan Table is transferred to a special Output Table and frozen, while the program continues to collect and save data at the same 5-second scan rate for the next 2.5 minutes.

This five minutes of data is then displayed on the alarm typer. Strip chart information is continuously displayed, so that time history is dependent only on the requirements of the evaluation team (see 1.3 res-ponse).

1.2.2.4 The format for displaying data and information is standard for the GEPAC 4020 (see Response to 1.2.1.4).

1.2.2.5 Retention and retrievability is provided by the Document Control Group.

The hard copy of the printout is controlled by this group and retained in the records vault at Pilgrim.

1.2.2.6 The power supply to the process computer is non-interruptible and non.

Class IE.

1.3 The following is a list of instrumentation available in the main Control Room which may be used as needed for the assessment of unscheduled shut-downs:

l 1) PR-3392 Condenser Yacuum Strip Chart (reads in inches of mercury).

2) PR-3050 Turbine Main Steam Pressure (850-1050 PSIG).

3) VR-3000 Turbine Vibration Trip T/G at 12 mils.

4) 640-26 Two pen recorder: black pen records vessel level 0"-60"; the red pen records feedwater flow 0-10x10 6 lbs/ hour.

5) 640-27 Two pen recorder: black pen records wide range yessel pressure 0-1500 PSIG; red pen records reactor steam flow 0-10x100 lbs/ hour.

6) 640-28 Two pen recorder: black pen records turbine steam flow 0-10x106 lbs/ hour; red pen records narrow range pressure 950-1050 PSIG.

7) 750-10 A, 8, C, D APRM Neutron Flux 0-1251, Two pen recorder for 6 channels of APRM. l l

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1,4 General Electric has been contracted to 'eplace the existing plant process computer, with expected completion by the end of 1986.

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2.1.1 Safety-related systems, structures, and components (SS&C) are identified i in the Pilgrim Nuclear Powar Station (PNPS) Q-List which is described in l the item 2.2.1.2 response. Documents (Purchase Orders, Maintenance l Requests) used to control activities associated with the Q-Listed equip- i 1

ment are identified as "Q" and subject to the requirements of 10CFR50, Appendix B and the Boston Edison Quality Assurance Manual (BEQAM)'(see l response to 2.2.1.4). Components which are required to function for a reactor trip are identified in the Q-List and are, therefore, controlled at a quality level consistent with their safety-related functions.

2.1.2 Records documenting the original qualification and testing of existing safety-related equipment are retained as quality assurance records and controlled in accordance with the 10CFR50, Appendix B, Criterion XVII requirements described in Boston Edison's Quality Assurance Manual (BEQAM), i Volume II. This encompasses documentation for equipment which serves a reactor trip function. l Nuclear Operations Procedure (N0P83A1) defines Pilgrim's Technical Group as responsible, when requested, for station evaluations of, among other externally generated information, Bulletins, Circulars, Service Informa-tion Letters, and Technical Information Letters.

We realire'that there is an effort by GE (through the BWROG) and a NUTAC (see 2.1.3) on vendor interface which may require changes to our existing systems. We shall inform the NRC of plans for such changes, if necessary, after we have assessed what GE and the NUTAC have provided to us.

2.1.3 We are participating in the Nuclear Utility Task Action Consnittee (NUTAC) on vendor interface, which is expected to provide results and recommenda-tions in February,1984 We wish to review this material, assess what impact it has on current BEco programs and procedures, and develop appro-priate comitments and schedules. Based on the NUTAC's February date, we_.

shall provide our commitments and date of completion in April,1984._

2.2.1.1 Components within systems classified as safety-related are themselves considered safety-related if they function in some capacity to assure (1) the integrity of the reactor coolant pressure boundary, (2) the capability to shut down the reactor and maintain it in a safe shutdown condition, or (3) the capability to prevent or mitigate the consequences ,

of accidents that could result in potential offsite exposures comparable to the guideline exposure of 10CFR Part 100. This will henceforth be )

referred to as a safety-related function, j

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This criteria is applied as follows:

1. Civil Structures which are required to maintain their integrity to l assure performance of a safety-related function are considered safety-related. This includes all elements of the structure which  !

are essential to maintenance of its structural integrity. A safety- ,

related structure may provide its assurance of safety either (a) l (Reactor Building perimeter provides Secondary Contain- l directly(b) ment), indirectly through support of safety-related equipment l

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(specific block walls), or (c) indirectly through housing of safety-related equipment, such that failure of the structure could threaten the performance of a safety-related function (Main Control Room perimeter). , , ,

2, Mechanical piping and components which are considered safety-related have been and are designed and installed in accordance with Seismic Class I desip requirements. This includes those systems or por- ,

tions of systems which either directly serve a safety-related func-tion or are in such close proximity to safety-related equipment that failure of the pressure boundary could potentially affect a safety-related function. For those Class I portions of the former type, ,

1 the Functional Class I Breaks are clearly identified on Piping and Instrument Diagrams (PSID's) and/or piping isometrics, whereas those j Class I portions of the latter type are identified in piping iro- i metrics (with notes on PalD's to indicate that some portions are ClassI). All mechanical piping and components determined to be  ;

Seismic Class I are designated as safety-related either passively (pressure boedary only) or actively. Also considered safety-related l l

are the supports and hangers which provide the Seismic Class I pro-tection.

3. For each mechanical component above that serves its safety-related function by actively responding to some electrical stimulus, the electrical assemblies critical to the performance of that safety-related function are considered safety-related. This includes such items as cables, penetrations, junction boxes, conduits, cable trays, panels (and associated internals), supports, and power sup-t plies.

4 Power supplies to safety-related devices are traced back to the originating emergency supply (Battery, Diesel Generator) through applicable switchboards, transformers, switching and breaking devices (including controllers) Motor Control Centers, Distri-bution Panels, Special Local Control Panels, and all of the asso-ciated cables, junction boxes, etc. which are required to transmit power betwee9 these stations.

This 2.2.1.2 Operable safety-related SSAC's are identified in the PNPS Q-List.

list was originally developed by Bechtel from a criteria similar to that described in 2.2.1.1. In February,1983, REco completed an effort to verify tne contents of the Bechtel-generated list against the latest This approved engineering drawings and against the criteria of 2.2.1.1.

assured accuracy of the 0-List for completed plant design changes, as reflected in those engineering drawings. Plant design changes which have been implemented, but are not yet reflected on engineering drawings, and hence are not yet incorporated into the PNPS Q-List, are described l

l in PDC packages which identify the work controls applicable to the Aasso-ciated SS&C's. These are dispositioned on a case-by-case basis.

l l system is now is effect to identify safety-related SS&C's on a Bill of Materials in the design phase of plant changes.

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~ l Validation of the results of the February,1983 effort was made through  !

independent review by representatives from each engineering discipline, Operations, Maintenance, and Quality Assurance prior to its release.

Subsequent revisions are validated by independent and wel.1-documented  !

engineering reviews of requests for Q-List revision. -

l 2.2.1.3 BEQAM, Volume II is applied to all activities affecting safety-related SS&C's. Activities falling within the scope of the QA Program cate-gorica11y include: designing, purchasing, fabricating, handling, shipping, storing, cleaning, erecting, installing, inspecting, testing,  ;

operating, maintaining, repairing, refueling, training, and modifying. ,

Any procedure, maintenance request, work order, purchase order, design l change, or other document used to control one of the above activities is '

required by procedure to indicate either "Q" or "Non-Q" control of the applicable activity. Since safety-related S$8C's are identified in the ,

PNPS 0-List, any activity associated with a 0-Listed item is designated '

as a "Q" activity and controlled appropriately in accordance with the OA Program work controls.

2.2.1 A The Boston Edison Quality Assurance Program is defined in the Boston Edison QA Manual (BE0AM), Volume II, and applies to quality-related and l quality assurance activities.  :

The BEQAM requires that structures, systems, and components designated as safety-related, and other items for which the Vice Presidents agree to use the OA Program management controls, be identified on the 0-List.

The Q-List is the "information handling system" referred to in NUREG-1000. The BEQAM requires that the 0-List be established and maintained by the Nuclear Engineerir.g Manager.

The Nuclear Engineering Manager implements this responsibility through NED Procedure 6.07, "Maintaining the 0-List." The Q-List is controlled, and the latest revision is distributed to the locations of use.

Three in-process checks are done by the Quality Assurance Department to ensure proper routine use of the Q-List. ,

1. Plant Design Change Review The QAD reviews and approves all proposed plant modifications accord-ing to QAD Procedure 3.02, "Review of Plant Design Changes and Major Field Revision Notices." All changes are designated safety-related l (Q), or non-safety-related (non-0) according to the Q-List Classifi-cation of the system or component being modified.

The validity of the Q or non-Q designation is checked by comparing each system or component to be modified with the 0-List. Associated l

drawings, safety evaluations, and available procurement documents l are also checked for consistent 0 or non-0 designations.

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The QAD is required to signify approval by signing the Plant Design l l Change or Major Field Revision Notice.

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2. Procurement Docuwnt Review The QAD reviews and approves all preliminary procurement documents according to QAD Procedure 4.01, "Review of Preliminary. ?rocurement Documents Prepared by BECo." All procurement documents are.desig-nated Q or non-Q according to the Q-List classification of the itta or end use of the service being purchased.

The validity of the 0 or non-Q designation is checked by comparing each item or and use service application to be purchased with the Q-List.

The QAD is required to signify approval by signing the preliminary procurement document.

3. phinter.ance Request Review Currently, the Operations Quality Control (000) Group reviews all Maintenance Requests for work at Pilgrim Station under the BEco QA Progran using QC instruction 5.01, Revision 1 "Qualtt: Control Review of PNPS Maintenance Requests." A checklist is used to ensure proper review and classificaticn of the work based r,n Station pro-cedures and the PNPS Q-List.

The 00C Group does not rev*ew Maintenance Reque',ts for work per-formed by contractors under their own BEco-approved QA p-ograms.

In addition to toe in-process checks, the QAD performs random surveil-lance inspections and periedic scheduled audits of all QA Program related activities. The preparation, validation, and routine use of the Q-List is writhin the scope of these inspections and audits. Details

of these functions are as follows

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1. Periodic Audits Planned periodic audits are performed to verify that procedures for preparation, validation, and routine use of the Q-List have been l

followed and are effective. These audits are performed by qualified personnel not having responsibilities in the areas being audited and using written checklists according to 0AD Procedure 18.01. Audit results are documented and reviewed 'oy management, and followup action en deficient areas is taken.

QA audits evaluate the entire Q-List update process to assure that:

. o Required changes are forwarded to the Systems and Safety Analysis (SASA) Group (via a DRN, Revision to "Q" Request (RQR), Plant Design Change Bill of Materials, etc.), which is responsible for l

' meintaining the 4-List.

1 o An index is maintained by the $4SA Group of requested changes l

received.

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Requested changes are reviewed and approved by appropriate per-o sonnel for inclusion in a Q-List revision, o The Q-List is updated, as required, to reflect apprend Q-List ~

changes. ,

o The 0-List is controlied and distrhed to required personnel.

2. Survet11ance Inspections Currently, surveillance inspections of various plant activities are performed by the 00C Group on a random, unscheduled basis in accord-ance with QAD Procedure 10.03 and QC Instructions 7.02 and 10.04 Checklists are generally not used. Selection criteria is not for-malized expect as delineated in QA0 Procedure 10.03. Surveillance inspection reports are issued to document these surveillances.

Based on recommendations from the NRC and INPO, a draft change to QAD Procedure 10.03 has been prepared to redefine the scope, purpose, and implementation of the surveillance (monitoring) function. The QAD Procedure will require that surveillances be scheduled and un-scheduled, random and selective, and with sufficient detail to effectively monitor and report the conditions at PNPS. The sur-veillances are porformed in support of, and as supplements tu, audits and inspections to provide quality assurance coverage of station in-process activities. The scope of monitoring includes verification that procedures for preparation, validation, and rou-tine use of the Q-List have been followed.

This expanded surveillance would be both planned (on a monthly basis),

and unpionned (e.g., response to INPO SOER's, NRC I&E Information Potices and Circulars, and other relevant nuclear industry reports andinformation).

2.2.1.5 Attachment C provides a sample Production Order for the purchase of safety-related equipment.

2.2.1.6 The Boston Edison Nuclear Organization recognizes three levels of major classification, 'Q", "non-Q" and 1/Q (See Attachment A, NOP8301, Section 5 for further definition). The "Q" designation applies to all safety-related equipment and activities. The PNPS 0-List identifies safety-related SS&C's at a level which does not recognize the classification of piece-parts within listed assemblies. These are generally dealt with on a case-by-case basis, however, the Organizstion recos,;izes a level within "Q" of those piece-parts which are not engineered for specific nuclear application and require no vendor-certified qualifica-tions testing. These are designated as Comerical Quality control Items and are specifically identified in a section of the Q-List. Quality Controls are being established in Specifications for these items.

2.2.2 Addressed above in our response to 2.1.3.

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3.1.1 Plant maintenance at Pilgets is done in accordance with the requirements of Procedure 1.5.3, "Maintenance Requests" and is tracked by the Mainte-nance Request (MR) form which reflects 1.5.3. This procedure is in accord-ance with ANSI 18.7 (1976).

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The p5t process incorporates a series of steps and check-offs prio'r to  ;

beginning maintenance. These steps are to ensure that the necessary l disciplines (Operations, Maintenance, Quality Control, and Fire Protec- l tion, as necessary), may review the request and designate any steps or procedures necessary to satisfy existing requirements. l The specific issue of post-maintenance testing is determined by both the Maintenance Staff Engineer (MSE) and the Operating Supervisor (05). The 05 determines what tests are required prior to beginning work, for example, testing a redundant system prior to removing its duplicate for maintenance.

The parameters for acceptance are contained in Technical Specifications or surveillance procedures.

The OS also determinds what tests must be performed before the system can be returned to service. In some cases, where QC has indicated necessary during the MR review, QC must be notified prior to the performance of the test. .

After the OS has made his determination, the Watch Engineer reviews the MR  %

and, should the Watch Engineer disagree, the MR is returned to the 05 for resolution prior to the start of work. ,

Post-maintenance testing other than for surveillance is determined by the MSE. This testing is to demonstrate that the maintained item performs in accordance with procedures or vendor information.

We believe this process allows appropriate determinations to be made by those most familiar with plant conditions at the time work will take place.

We also believe this process adequately ensures appropriate post-mainte-nance testing; therefore we plan no further action at this time.

3.1.2 As part of our Performance Improvement Plan, we instituted a Procedure Update Program (PUP) for operations and maintenance procedures. The PUP is a one time effort. After completion of PUP, future revisions to prc>

cedures and Vendor Manuals are to be hanMed by existing organization procedures in an ongoing, timely manner. At this time 3 the PUP is ongoing,

and is expceted to be completed by October 31, 1984 The PUP has been implemented using a systems approach, with work assigned and scheduled by the PNPS Operations Department Management. System pro-cedure update prioriti n are determined by cognizant operations persennel based on their experience and knowledge of plant systems.

The inputs to this program are listed below:

o Plant Design Change Information o 1.C.6 Independent Yalve Verification Requirements 4

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e o Operator Experience Feedback o Training Department Feedback o INP0 Reconnendations o Modification Management Group Feedback o Procedure Classification Changes (Safety-related or Non-safety related determination) o Vendor Manual Information The vendor manual validation process is being incorporated into a Nuclear Organization Procedure. This N0P is in its review cycle and is expected to be emplaced by January 1,1984 We believe the PUP, while not specifically initiated in response to Generic Letter 83-28, satisfactorily addresses its concerns.

3.1.3 Surveillance freqvsncies contained in Pilgrim's Technical Specifications for both the Reactor Protection System and other systems were initially formulated using vendor information and established probability techniques.

As operating experience and new information has developed, we have amended the Technical Specifications, after careful review and with NRC concur-rence, to emplace changes which would enhance safety.

We are considering an evaluation of relevant nuclear industry and Pilgrim failure rate data to assess appropriate actions concerning Technical Speci-fication post-maintenance testing requirements.

3.2.3 Addressed above in 3.1.3.

l 3.2.1 Addressed above in our response to 3.1.1, 3.2.2 Addressed above in our response to 3.1.2.

3.2.3 Addressed above in our response t: 3.1.3.

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45 0 PNPS performs on-line functional testing of the reactor protection system, including independent testing of the diverse trip features. Initiating circuitry is tested in accordance with appropriate Technical Specifica-tion requirements. For this testing, the logic is checked from~ process parameter input through to the actuating device.

General Electric, through the BWROG, is reviewing the adequacy of existing surveillance and the periodic testing of Backup Scram Valves.

The results of this effort is expected in March,1984 After reviewing GE's recommendations and results, we shall submit any appropriate actions and completion dates. Based on March,1984 as our receipt from GE of their findings, we will submit our results in June,1984 The results of the BWROG may also indicate a need to change Technical Speci-fications, and we will assess such recommendations at that time. However, we wish to reinforce our response in 3.1.3 of this letter that our Technical Specifications is a "living" document which has been and continues to be refined by operating experience.

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• ATTACHMENT A

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