ML20081K283

Forwards Response to Generic Ltr 83-28.Procedure Update Program for Operations & Maint Procedures Will Be Completed by 841031

ML20081K283
Person / Time
Site:	Pilgrim
Issue date:	11/07/1983
From:	Harrington W BOSTON EDISON CO.
To:	Vassallo D Office of Nuclear Reactor Regulation
References
83-275, GL-83-28, NUDOCS 8311090331
Download: ML20081K283 (41)

Text

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q BDSTON EDISDN COMPANY B00 BOYLSTON STREET BDSTON, M AssAcHusETTs 02199 WILLIAM D. HARRINGTO N c::ssion wees pasassent w t aksan November 7, 1983 BECo 83-275 Mr. Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.

20555 License No. DPR-35 Docket No. 50-293

Subject:

Response to Generic Letter 83-28

Dear Sir:

In Generic Letter 83-28, dated July 8, 1983, NRC requested that we submit the status of current conformance with the positions contained in that letter, and plans and schedules for any needed improvements for conformance with the posi-tions.

We requested an extension to April, 1984 to respond because of our participation in industry groups formed to address GL 83-28.

Our request was denied in an NRC letter dated October 19, 1983.

That letter requested that we submit the requested information to the extent practical, pro-vide commitments and schedules for items where final implementation is known, and to supply our plans and schedules for developing programs where needed.

The attachments accompanying this letter responds to the October 19, 1983 letter.

We believe it satisfies your request.

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8311090331 831107 PDR ADOCK 05000293 P

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COSTON EDCON COMPANY Mr. Domenic B. Vassallo, Chief Page 2 November 7,1983 Should you require further information concerning this response, please contact us.-

Very truly yours, PMK/ mat Attachments cc:

Mr. Darrell G. Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.

20555 Commonwealth of Massachusetts)

County of Suffolk

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Then personally appeared before me W.D. Harrington, who, being duly sworn, did state that he is Senior Vice President - Nuclear of Boston Edison Company, the applicant herein, and that he is duly authorized to execute and file the submittal contained herein in the name and on behalf of Boston Edison Company and that the statements in said submittal are true to the best of his knowledge and belief.

My Commission expires: V/4N/

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73 Notary Public

Generic Letter 83-28 1.1.1 1 Restart of Pilgrim Station is acceptable only if the following criteria are satisfied:

1) The cause of the unscheduled reactor shutdown has been determined and appropriately corrected or the Operations Review Committee (0RC) has

. determined 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-istrative controls (0RC, 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 requirei.nents 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 NUREG-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 surroundii ' a trip occurrence. Recorder strip charts are uti-lized for evaluation o? long-term trends which may not be indicated by the alarm typer output. The balance of essential information is provided by plant. personnel, whose combined knowledge is relied upon for reconstruc-tion of human activities prior to and during t?.a trip event (maintenance activities, operator actions, etc.).

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

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e 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 Committee is convened to provide independent assessment of the event. The ORC utilizes the same information as the task force. Both groups may call on the technical and engineering expertise of other personnel in the Nuclear Orgar.ization or other appropriate group, internal or external.

1.1.7 We are in the process of reviewing a draft INP0 procedure concerning post-trip reviews for possible incorporation into a Nuclear Operations Pro-cedure (N0P). Currently, Section 6 of N0P8301, " 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 alarm indications and messages if monitored analog values exceed predefined limits or if digital trip signals occur.

It performs calculations with selected input data to provide the operator with essential core performance information through a variety of logs, trends, summaries, and other typewriter data arrays. The Sequence of Events printer responds to digital signals, the Data 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 information is as follows:

Sequence of Events:

Time Cycle Point ID Name Status XXXXXX XX XXXX XXXX Date Recall Log Time Point ID 1.......................... Point 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).

0 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 80P 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 poi,nts, 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 Reactor Feedwater Flow "A" Loop Reactor Feedwater Flow "B" Loop Reactor Water Level (inches)

Outlet Steam Flow Feedwater Temperature "A" Loop Feedwater Temperature "B" Loop Recirculation Flow "A" Loop Recirculation Flow "B" Loop Reactor Saturation Temperature Calculated Seawater Flow Hotwell Outlet Temperature Drywell 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)

Reactor Building Closed Cooling Water System (RBCCW) "A" Loop Flow RBCCW "B" Loop F1ow 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-

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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:

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

2)

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

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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 6 lbs/ hour.

pen records feedwater flow 0-10x10

5) 640-27 Two pen recorder: black pen records wide range vessel pressure 6 lbs/ hour.

l 0-1500 PSIG; red pen records reactor steam flow 0-10x10 6

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6) 640-28 Two pen recorder:

black pen records turbine steam flow 0-10x10 lbs/ hour; red pen records narrow range pressure 950-1050 PSIG.

7) 750-10 A, B, C, D APRM Neutron Flux 0-125%, Two pen recorder for 6 channels of APRM.

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2.1.1 Safety-related systems, structures, and components (SS&C) are identified I

in the Pilgrim Nuclear Power Station (PNPS) Q-List which is described.in the item 2.2.1.2 response. Documents (Purchase Orders, Maintenance Requests) used to control activities associated with the Q-Listed equip-ment are identified as "Q" and subject to the requirements of 10CFR50, Appendix B and the Boston Edison Quality Assurance Manual -(BEQMi) (see response to 2.2.1.4).

Components which are required to function for a reactor trip ara identified in the Q-i.ist and are, therefore, contro!!ed

'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),

Volume II. This encompasses documentation for equipment which serves a reactor trip function.

Nuclear Operations Procedure (N0P83A1) defines Pilgrim's Technical Group as responsible, when requested, for station evaluations of, among other 1

externally generated information, Bulletins, Circulars, Service Informa-tion Letters, and Technical Information Letters.

We realize 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 i

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 Committee (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 commitments 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 i

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.

This criteria is applied as follows:

1.- Civil Structures which are required to maintain their integrity to 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) directly (Reactor Building perimeter provides Secondary Contain-4 ment), (b) indirectly through support of safety-related equipment 4

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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 design requirements. This includes those systems or por-4

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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, the Functional. Class I Breaks are clearly identified on Piping and Instrument Diagrams (PAID's)_ and/or piping isometrics, whereas those Class I: portions of the latter type are identified in piping iso-metrics (with notes on PAID's to indicate that some portions-are Class I). All n=chanical piping and components determined to be Seismic Class I.are designated as safety-related either passively (pressure boundary only) or actively. Also considered safety-related 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 functi_on are considered safety-related. This includes such l

-items as cables, penetrations, junction boxes, conduits, cable trays, panels (and associated internals), supports, and power sup-i plies.-

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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,LSpecial. Local Control Panels, and all of the asso-ciated cables, junction boxes, etc. which are required to transmit power between these stations.

2.2.1.2 Operable safety-related SS&C's are identified in the PNPS Q-List. This list was originally developed by Bechtel from a criteria similar to that described in 2.2.1.1.

In February,1983, BECo completed an effort to

--verify the contents of the Bechtel-generated list against the latest approved engineering drawings and against the criteria of 2.2.1.1.

This t'

-assured accuracy of the'Q-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

in PDC packages which ide uify the work controls applicable to the asso-ciated SS8C's. These are dispositioned on a case-by-case basis. A system is now in effect to identify safety-related SSAC's on a Bill of

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Materials in the design phase of plant changes.

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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 well-documented engineering reviews of requests for Q-List revision.

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-gorically 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 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 SS&C's are identified in the PNPS Q-List, any activity associated with a Q-Listed item is designated as a "Q" activity and controlled appropriately in accordance with the QA Program work controls.

2.2.1.4 The Boston Edison Quality Assurance Program is defined in the Boston Edison QA Manual (BEQAM), Volume II, and applies to quality-related and 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 QA Program management controls, be identified on the Q-List.

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

The Nuclear Engineering Manager implements this responsibility through NED Procedure 6.07, " Maintaining the Q-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 0-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 (Q), or non-safety-related (non-Q) according to i;he 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 drawings, safety evaluations, and available procurement documents are also checked for consistent 0 or non-Q designations.

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

2.

Procurement Document Review

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

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

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

3.

Maintenance Request Review Currently, the Operations Quality Control (00C) Group reviews all-Maintenance Requests for work at Pilgrim Station under the BECo QA Program using QC instruction 5.01, Revision 1, " Quality Control Review of PNPS Maintenance Requests." A checklist is used to ensure proper review and classification of the work based on Station pro-cedures and the.PNPS Q-List.

The 00C Group does not review Maintenance Requests for work per-formed by contractors under their own BECo-approved 0A programs.

4 In addition to the in-process checks, the QAD performs random surveil-lance inspections and periodic scheduled audits of all QA Program related activities. The preparation, validation, and routine use of the Q-List is within the scope of these inspections and audits. Details of these functions are as follows:

1.

Periodic Audits

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

followed and are effective. These audits. are performed by qualified l

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 by management, and followup i

action on 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 (S&SA) Group (via a DRN, Revision to "Q" Request (RQR), Plant Design Change Bill of Materials, etc.), which is responsible for maintaining the Q-List.

o An index is maintained by the S&SA Group of requested changes received.

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o Requested changes are reviewed and approved by appropriate per-sonnel for inclusion in a Q-List revision.

o The Q-List is updated, as required, to reflect approved Q-List changes.

o The Q-List is controlled and distributed to required personnel.

2.

Surveillance 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 QAD Procedure 10.03. Surveillance inspection reports are issued to document these surveillances.

Based on recommendations from the NRC and INP0, 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 performed in support of, and as supplements to, 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 unplanned (e.g., response to INP0 S0ER's, NRC I&E Information Potices and Circulars, and other relevant nuclear industry reports and information).

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, N0P8301, Section 5 for further definition). The "Q" designation applies to all safety-related equipment and activities.

The PNPS Q-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 Organization recognizes a level within "Q" of those piece-parts which are not engineered for spet.'fic nuclear application and require no vendor-certified qualifica-tions testing. These are designated as Commerical Quality control Items and are specifically identified in a section of the 0-List. Quality Controls are being established in Specifications for these items.

2.2.2 Addressed above in our response to 2.1.3.

3.1.1 Plant maintenance at Pilgrim 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).

The MR process incorporates a series of steps and check-offs prior to beginning maintenance. These steps are to ensure that the necessary disciplines (Operations, Maintenance, Quality Control, and Fire Protec-tion, as necessary), may review the request and designate any steps or procedures necessary to satisfy existing requirements.

The specific issue of post-maintenance testing is determined by both the Maintenance Staff Engineer (MSE) and the Operating Supervisor (0S). The OS 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 determines 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 OS 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 familia 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 pro-cedures and Vendor Manuals are to be handled by existing organization procedures in an ongoing, timely manner. At this time, the PUP is ongoing, and is expected 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 priorities are determined by cognizant operations personnel based on their experience and knowledge of plant systems.

The inputs to this program are listed below:

o Plant Design Change Information 0

1.C.6 Independent Valve Verification Requirements

o Operator Experience Feedback o Training Department Feedback o INP0 Recommendations o Modification Management Group Feedback o Procedure Classification Changes (Safety-related or Non-safety related l

detennination) 1 I

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 I

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 frequencies 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.

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 to 3.1.3.

4.5 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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6/17/83 CONDUCT OF 00ERATIONS 1.0 References 1.1 ANSI N18.7 - Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants.

1.2 Regulatcry Guide 1.33 - Quality Assurance Program Requirements (operation).

2.0 General Boston Edison Co.'is responsible for assuring that the facility is operated within the requirements of the license, Technical Specifica-tions, rules, regulations, and Orders of the NRC and for the actions of their employees.

The Watch Engineer has the authority and responsibility to direct all activities of Pilgrim Station that effect safe operation.

Trans-fer of authority and responsibility to plant staff members above the Watch Engineer level shall be predetermined. Transfer of this authority and responsibility for routine shif t turnover and energen:y conditions shall be docanented.

s The safety and general welfara of employees, general public and the facility shall be of prime concern during the operational phase of Pilgrim Station.

The watch engineer has the authority to shut x

the reactor down when, sin his judgement, continued operation would jeop3rdize the health and welfare of the general public or the safety'of plant equipment.

In an emergency, any licensed operator has the same' autMority.

3.0 Shift Turnover Shif t turnover shall be diligently performed in a conscientious professional and thorough manner.

As a minimum, shift turnover sheets are to be prepared by the off-going shif t and information transferred to the on-coming personnel shall include:

o status of Safety-systems o off-normal lineups s

o annunciator status o.. major work'.in progress o status of surveillance testing o planned operational activities including power maneuvers s

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UNCOWTROLiED G G PV99p;;J REVISED 6/17/83 Shift responsibility remains with the off-going personnel until the on-coming shif t personnel have individually accepted, in writing, responsibility for their watch station.

Shift change should be a dedicated task. Other planned Ictivities should be minimized during turnover.

4.0 Shift Records Each operating station, e.g., watch engineer, shif t supervisor, control room operator, shift chemist, shif t HP, shall keep a journal or log of the significant activiiies of their operating station.

These shall be reviewed frequently by appropriate managers.

Shift recceds such as operating log books, data sheets, check lists, recorder charts, computer printouts, and maintenance requests that describe or record operating information and actions must 5e legible, accurate, complete and understandable.

When these records require correction, a single line shall be drawn through the incorrect data in a manner such that it will not be obliterated. The correct data will be in a space adjacent to the lined out data along with the date and initials of the person making the correction.

5.0 Control and Status of P1_ ant _S_ystens and_ Equipment Measures shall be established and implemented to assure that:

Only licensed operators are permitted to manipulate the controls that directly affect reactivity (10CFR50.54 (1))

Licensed operators are required to be present at the controls at all times during the operation of the facility (10CFR50.54 (k))

Operation of mechanisms and apparatus other than controls which may indirectly affect the power level or reactivity of a reactor shall only be accomplished with the knowledge and consent of an operator licensed in accordance with Part 55 (10CFR50.54 (j))

- A licensed senior operator shall be present at the facility or readily available on call at all times during its operation, and shall be present at the facility during initial startup and approach to power, recovery from an unplanned or unscheduled shut-down or significant reduction in power, and refueling or as otherwise prescribed in the facility license (10CFR50.54(m)).

• Reasonable action that departs from a condition of the License /

Technical Specifications may be taken in an emergency when such action is immediately needed to protect the public health and

• Revision J

tit' CONT ROi LED G G P i9e s of 9

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NOP8301 REVISED 6/17/83 safety and no action consistent with 1.icense/ Technical Specifi-cation conditions that can provide adequate or equivalent pro-tection is immediately apparent. Prior to taking these permitted actions (which may be contrary to License / Technical Specification conditions) approval, as a minimum, shall be by a licensed senior operator (10CFR50.54 (x),(y)).

• Notifkation of protective actions taken in accordance with 100FR 50.h (x) and (y), shall be made to the NRC Operations by tele-phone. When time permits, the notification must be made before protective action is taken; otherwise, the notification must be made as soon as possible thereaf ter (10CFR50.72(c)).

The NRC licensed individual shall observe all applicable rules, regulations and orders of the Commission, whether or not stated in the license (100FR55.31 (d))

NRC licensed individuals are responsible for taking timely and proper action so as not to create or cause a hazard to " safe opera-tion of the facility" (i.e. actions or activities, including failure to take action, related to the facility which would have an adverse affect on the health and safety of the public, plant workers or the individuals).

NRC licensed individuals shall comply with the requirements per-taining to the operation of the facility and manipulation of its controls and with radiation safety procedures implementing 100FR20.

The status of plant systems and equipment shall be continuously moni tored. Monitoring methods and techniques employed at Pilgrim include, but are not limited to personnel tests / inspections /examina-tions, annunciations, indicating lights, indicators, recorders and CRT/ computer.

Measures shall also be established for indicating the operating status of structures, systems, and components of the station.

Operators shall have confidence in the instrument readings /annuncia-tions and response to these indications are to be as specified in the applicable procedure / instruction.

Should the indication not meet specified operating parameters, or be suspect for any reason, an alternate means of monitoring should be implemented until such time as the adverse, or suspected adverse condition is rectified.

In the unique event that alternate means is not available, the most conservative value is to be assumed.

• The station systems, structures, and components are divided into three categories.

These are Q, non-0, and 1/Q.

A list for Q sys-tems, structures, and components is provided as defined by the BE0A'i.

• Revision

9' 4 GI 9 NOP8301 REVISED 6/17/83 A second list is provided by the Station Manager and is the 1/Q list.

The remainder of the systems, structures, and components are non-Q and are r M specifically listed.

• Modification is any change to a station system, structurh, or com-ponent that requires a change to a controlled design document.

A modification for Q and non-Q can only be implemented by an approved design change or Temporary Modification.

• Maintenance is the act of maintaining, i.e., keeping the existing stete of repair, ef ficiency, and quality. Maintenance for Q and non-Q can only be implemented by an approved Maintenance Request (MR).

+For 1/0 systems, structures, and components, any level of quality is sufficient if the system, structure, or component is capable of performing its intended function and the modification and main-tenance work controls specified herein do not apply.

• Pennission to release any system or equipment for maintenance or testing shall be given by the SRO in charge of the watch. The granting of the permission shall be documented.

Prior to granting release of equipment, a deternination shall be made to assure that equipment or a system may be released, how long it may be released, and what functional testing of redundant system is required prior to and during the out-of-service period.

Upon completion of such activity or, when a change in the scope of such activity is contemplated, the SRO in charge of the watch shall be notified.

• Equipment removed from service for maintenance will be identified by tags. The tags will be placed at locations where the equipment could be onerated. Tags may only be placed by qualified individuals froin the group responsible for the operation of the equipment. Tags placed on control room components shall be placed so as not to obstruct instruments, controls or indicating lights.

Equipment released for testing or returned to service af ter testing need not be so identified.

The test status of equipment (i.e., when the last tett was performed or when next test is to be performed) will be identified in plant surveillance records.

• Temporary modifications shall be identified by tags placed at the location of the modifications.

Current status of temporary modifi-cations shall be maintained in a log.

Installation or removal of temporary modifications shall be verified (see Note (1) next page) by a second person having knowledge in the system being modified or by a functional test which will prove the installation or removal.

• All Q-listed equipment which has failed required tests or lacks documentation attesting to its operability will be identified as being deficient and logged in the Control Room logbook and/or the

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NOP8301 REVISED 6/17/83 Operators Shift Turnover sheet. When 0-listed equipment is lacking proper documentation, it is considered to be inoperable unless otherwise resolved under the guidance of the BECo QA Manual.

Technical Specifications Action Statements are entered when items required to be operable by limiting conditions for operation are known to be inoperable.

Items may be determined inoperable (1) during use, (2) during a surveillance test, or (3) when surveillance requirements are not performed within the specified time intervals (af ter applying the allowable tolerance).

Action Statements are entered under item (3) when the surveillance requirements should have been performed rather than at the time it is discovered that tests were not performed.

• When an ECCS, ECCS Subsystem, RPS or Primary Containment Isolation System is placed back into service after maintenance or testing has been performed, an independent verification (see Note (1) below) that equipment has been placed in its proper configuration shall be made by a qualified person. All station system alignnents will be verified prior to startup af ter each refueling outage.

In addition to this, all ECCS, PCIS and RPS will be verified independently by a second qualified person.

A qualified person is defined as a person who would be qualified to perform the initial component alignment.

• Note (1):

Independent verifications may be perfomed by:

a.

Direct method - checking appropriate equipment and/or control s, or, b.

boi ect method - observation of indicators and/or datus lights.

• An independent verification need not be performed if a person has the potential of receiving greater than 25 mrem whole body while performing the verification.

The status of safety-related (Q-List) systems shall be maintained at all times, and abnormal system alignment should be avoided except when absolutely necessary.

6.0 Unit Trips / Reactor Shutdown All unit trips shall be thoroughly investigated, and a report pre-pared and submitted to the Station Manager. Cause of the trip shall be detemined, and means to preclude repetition shall be implemented.

Likewise, challenges to safety systems and personnel errors are to be investigated and corrective action implemented to prevent recurrence.

Personnel actions responsible for unit trips, reduced capacity f ac-tor, reportable occurrences, and actions taken outside of plant

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UC kge 6 cf 9 N3P8301 REVISED 6/17/83 instructions or Technical Specifications shall be thoroughly investi-gated and reported to the Station Manager.

Personnel, whose flagrant, careless, or repetitive erroneous actions shall be subject to disci-plinary action.

When a condition exists that the reactor must be shutdown'in order to comply with the station license (PNPS Technical Specif.ication),

actions shall be initiated so that the specified condition is attained within the prescribed time period.

The cause of a scram or an unexplained power reduction must be investigated and determined before the reactor is returned to power.

Following any scram for which the cause cannot be determined quickly and without reasonable doubt, the Chief Operating Engineer or the Day Watch Engineer shall proceed to the site and take charge of the investigatory process until such time as the cause of the scram is known.

Essential in the investigation process is the encouragement of employees to reveal their specific actions at the time of the unit trip or shutdown.

Disciplinary action shall be tempered when forth-rightness is evident in the investigation.

Operating personnel shall be encouraged to express their concern or suggestions on how to improve performance whether it be from a viewpoint of safety, hardware, or personnel duties and responsi-bilities.

Permission to startup the reactor and the systems required for power operation, or to shut the reactor down for planned maintenance or refueling, will be issued by the Station Manager or his designated alternate.

7.0 l_nstructions, Procedures, Drawings The use of procedures shall comply with the Tech. Spe s. and with Reg. Guide 1.33.

Instructions, procedures or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining thet important activities have been satisfactorily accomplished.

Adherence to procedures is an essential ingredient of good station performance and configuration control.

Adherence recognizes that procedures are not blindly followed without exercising good judg-ment.

Individuals must recognize when adherence to procedures will result in degrading conditions and in emergencies take appropriate action utilizing their knowledge and judgment.

In these instances the individual should seek help and advice and whenever time permits the procedure must be changed or, if one does not exist, a procedure must be provided.

In all cases where evolutions were nerformed without a procedure or deviations from a procedure were performed, the event shall be evaluated by appropriate levels of management.

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[Yage 7 of 9 N0P8301 REVISED 6/17/83 8.0 Temporar,y Procedures (TPs)

Approved temporary procedures may be issued to direct operations during testing, refueling, maintenance and modifications. These are necessary to provide guidance in unusual situations not within the scope of the normal procedures and to ensure orderly ~ and uniform operations for short periods when the plant, a system or:a component of a system is performing in a manner not covered by existing detailed procedures or has been modified or extended in such a manner that portions of existing proceditres do not apply. Temporary procedures shall include designation of the period of time during which they may be used and shall be subject to the appropriate review process.

These TPs are reviewed and approved in the same manner as permanent PNPS procedures or as temporary changes to procedures.

9.0 Temporary Changes to Procedures (SRO changes)

Temporary changes to procedures may be appropriate when immediate implementation is necessary. These "SR0 changes" are allowed pro-viding 1) the intent of the original procedure is not altered and

2) the change is approved by two members of the plant management staff, one of whom holds a Senior Reactors Operator's (SRO) license.

These "SRO changes" shall be administratively controlled so as to allow for proper reviews, within a specific time frame, subsequent to implementation.

10.0 Control Room Work Atmosphere Strict discipline and attendance to instruments and alarms as well as proper behavior shall be exercised in the Control Room at all times.

Non-technical material and audio / video entertainment devices are not permitted in the Control Room. Eating is permitted only if it does not detract from attendance of the controls, and is limited to those positions which must remain in the Control Room due to license requirements or operational considerations.

The Watch Engineer is responsible for overall Contral Room Super-vision. Access to the Control Room shall be limited to persons who have a need or requirement to be there.

Examples are as follows:

Persons required to perfonn duties in the Control Room.

Persons who provide technical support to Operations.

Persons who are required during emergencies.

Persons requiring authorization for tests, maintenance, or moni-toring.

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!t ge 8 of 9 N0P8301 REVISED 6/17/83 Operators shall be cognizant of changes in instrument indications and annunciations that detect abnormal conditions or changes in equipment performance.

Logged parameters should be compared to previous readings to detect any trends in equipment performance.

11.0 Communications All communications, both verbal and written, shall be clear and precise.

All verbal communications of a directive nature (i.e., verify valve position, re-position valves, etc.) shall be repeated back by the receiver to the sender prior to the directions being carried out.

If the directions are complex or involve more than a routine evolu-tion, the receiver shall be required to write the directions down and repeat them back to the sender.

The text of written communications shall contain only essential information and shall be factual, specific, concise, comprehen-sive, and nonambiguous.

It shall be clearly worded as to be readily understandable by personnel responsible for the described activity.

12.0 Behavior Observation Supervisors shall t,e aware of and observe employees and contrac-tors for aberrant behavior, including argumentative hostility toward authority, irresponsibility, poor reaction to stress, and suspicion of being under the influence of alcohol or drugs while on Conpany property.

A Continual Behavior Observation Training brochure / package will be prepared by the Nuclear Training Department and provided to super-visors so as to give them guidance on situations to be aware of and what actions to take.

As a minimum, the following two basic situa-tions/ actions will be addressed:

12.1 Situation #1 A change in personality or behavior is noted where there is no immediate threat of changes.

Action:

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- Call the situation to the attention of the Station Manager.

- The Station Manager should then discuss the situation with the VPN0 and the Medical Department.

12.2 Situation #2 An individual displays a significant degree of aberrant behavior that is considered to be a threat to plant or personal safety.

WiiGOWTROiiEF W P Tage 9 of g NOP8301 REVISED 4

6/17/83 Action:

- Ensure that the individual will cause no harm to himself/

herself, others or equipment by talking with them and if necessary subduing them.

- Request Security Assistance.

- Arrange for transportation to the hospital.

- Notify Station Manager.

- The Station Manager should notify the VPN0 and the Medical Department.

• 13.0 0,v,ertime Guidelines In an effort to prevent situations where fatigue could reduce the ability of operations personnel to keep the reactor in a safe condition, the following guidelines has been adopted in regards to shift staffing aad the use of overtime:

o Adequate shift coverage should be maintained without routine heavy use of overtime.

o An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> straight (excluding shif t turnover time).

o An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in any 24-hour period, nor more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48-hour period, nor more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any seven day period (all excluding shift turnover time).

o A break of at least eight hours should be allowed between work periods (including shift turnover time).

o The use of overtime should be considered on an individual basis and not for the entire staff on a shift.

Recognizing that unusual circumstances may arise requiring deviation from these shif t staffing overtime guidelines, such deviation shall, as a minimum, be authorized by the Station Manager. The primary consideration in such authorization shall be that significant reduc-tions in the effectiveness of operating personnel would be highly unlikely.

• Revision

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ATTACHMENT B J

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w ADPENDIX 2 GEPAC 4020 I/O LIST ANALOG INPUTS DIGITAL INPUTS i

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ANALOG INPUTS (0-160mv)

NO. OF SIGNAL SIGNALS POINT ID 120 A000-A119 LPRM Level 6

B000-B005 APRM Level 2

9006-B007 RBM A&B 4

B008-B011 TIP Level, A-D 1

8013 Roactor Pressure 1

9014 R actor Core Pressure Drop Total Jet Pump Flow (Core Flow) 1 B015 Recirculation Drive Flow, Loops Al,A2,B1,82 4

B016,B 0 3 8,B 039,B060 Control Rod Drive System Flow 1

B 01.7 Reactor Feedwater Inlet Flow, A&B 2

B018, B019 Cleanup Flow, A&B 2

B020, B021 Recirculation Pump Motor Power A&B 2

B022, 8023 1

8024 Roactor Water Level.

Roactor Outlet Steam Flow 1

B025 Cleanup System Temp, inlet & outlet 2

B026, 8027 Roactor Feedwater Inlet Temp., A1., A2,81,B2 4

B028-R031 ROcircul ation Inlet Temp., Loops Al, A2,B1,B2 4

B052-B055 1

E001 Main Transformer Hot Spot Temp.

Aux. Transformer Hot Spot Temp., 1&2 2

E002, E003 Startup Transformer Hot Spot Temp., 1&2 2

E004, E005 Shutdown Transformer Hot 9 pot Temp.

1 E006 345 KV Switchyard 1, Line 342 V,A,W, VAR 4

E007-E010 4

E0ll-E014 i

345 KV Switchyard 1, Line 355 V,A,W, VAR Isolated Phase Bus Air Temp.

1 E015

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necirc. Pump Motor Winding Temp., 1&2 2

E016, E017 l

accirc. Pump M-G Set Winding Temp., 1&2 2

E018, E019 2

F000, P001 l

Drain Cooler Temp., Drain A,8 Condensate Demin. Diff. Pressure 1

F002 Steam Seal Header Pressure 1

F003 Gland Seal Condenser Pressure 1

F004 Roactor Bldg. Vent Exhaust Diff. Press.

1 F005 1

F006 Barometric Pressure 1

F007 Reactor Feedpump Suction Press.

Raactor Feedpump Discharge Press., A,B,C 3

F008-F010 Condensate Pump Discharge Header Press.

1 F0ll 2

F012, F013 Condenser pressure, west & east Fcedwater Reater Extraction Press lA-5A,lB-5B 10 F014-F023 1

F024 Air Ejector Of fgas Flow Mckeup Water Flow To Demin. Water Storage Tank 1

F025 Rcdwaste Chemical Waste Tank Level, A&B 2

F026, F027 RGactor Bldg. Closed Cooling Water Flow, A&B 2

F028, F029 2

F030, F031 RHR Cool.ing Water Flow, A&B 2

F032, F033 RHR Water Flow, A&B l

1 F034 Ejector Condensate Outlet Temp.

1 F035 Drain Cooler A-B Inlet Temp.

1 F038 l

Gland Seal Condensate Outlet Temp.

S a Water Inlet Temp.,1-1 & 1-2 2

F039, F040 Sea Water Outlet Temp., 1W,2W,3E,4E 4

F041-F044 Condenser Hotwell Outlet Temp., east & west 2

F045-F046 Haater Inlet Temp. (E&W), 1st-Sth points 10 F047-F056 1

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l' ANALOG INPUTS (0-160mv) i NO. OF SIGNAL SIGNALS POINT ID Hcater Outlet Temp. (S&W), 1st & 3rd points 4

F057-F060 H ater Drain Temp. (Train A&B), 1st-5th points 10 F061-F070 Raactor Bldg. CCW Serv. Water Inlet Temp. A&B 2

F071-F072 ROactor Bldg. CCW Serv. Water Outlet Temp. A&B 2

F073-F074 R2 actor Bldg. CCW Heat Exch. Inlet Temp, A&B 2

F075-F076 Raactor Bldg. CCW Heat Exch. Outlet Temp, A&B 2

F077-F078 RHR CW Inlet Temp., A&B 2

F079, F080 RHR CW Outlet Temp., A&R 2

F081, F082 RHR Water Inlet Temp. to Heat Exch. A&B 2

F083, F084 RHR Water Outlet Temp., Reat Exch. A&B 2

F085, F086 Sca Water Pump Level, A&B 2

F087, F088 Redwaste Disch. Meat Tracing Flow 1

F089 Roactor Feedwater Diff. Press., A&B 2

F092, F093 Generator Voltage 1

G000 l

l Generator Stator Current, Phase A,B,C 3

G006-G008 Mcin Transformer Net Amos 1

E000 Generator Gross Power 1

B012 Generator H Seal Oil Temp.

1 G002 2

Generator Field Voltage 1

G004 G2nerator Field Amps 1

G005 Generator VARs 1

G001 Gsnerator Stator Temp., Phase A,B,C 3

G009-G011 Stator Cooling Header Inlet Temo.

1 G012 Stator Cooling Header Outlet Temp.

1 G013 Gonerator Collector Air Inlet Temp.

1 G014 G:nerator Collector Air Outlet Temp.

1 G015 Alterex Stator Winding Temp., 1-3 3

G016-G018 Alternator Air Inlet & Outlet Temp., Point 1 2

G019, G020 Diesel-Generator Winding Temp., 1&2 2

G022, G023 Alterex Diode Cooling Water Outlet Temp.

1 G024 Rocctor Feedpump Motor Winding Temp., 1,2,3 3

G025-G027 Condensate Pump Motor Winding Temp., 1,2,3 3

G028-G030 Sta Water Pump Motor Winding Temp., 1&2 2

G031, G032 Ejector Radication Monitor 1

M000 Stcck Gas Radiation Monitor 1

M001 Rcdwaste Discharge Radiation Monitor 1

M003 Rndwaste Disch. to circulating Water PH 1

M004 Racctor Bldg. Vent. Exhaust Temp.

1 M005 R3cctor Bottom Head Water Temp.

1 M077 Rolief Valve Temp. for Leak Detection 6

M006-M010,M078 Drywell Containment Temp., 1-7 7

M0ll-M017 Suppression Chamber Temp., Points 1,2,3 3

M018-M020 Drywell Containment Dew Point, 1-7 7

M021-M024,M 065-M067 ROcctor Vessel Metal Temp., 1-8 8

M025-M032 Plcnt Heating Steam Flow 1

M033 Pressure Torus - Torus Reference Vessel 1

M034 Pressure Drywell - Drywell Reference Vessel 1

M035 ROcctor Feedwater Conductivity, A&B 2

M037, M048 Racctor Feedwater PH, A&R 2

M038, M049 Rocctor Feedwater Turbidity, A&R 2

M039, M080 Condensate Demin Water to FW Heaters 02 content 1

M036

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d ANALOG INPUTS (0-160mv)

NO. OF SIGNAL SIGNALS POINT ID Wind Direction 1

M040 Wind velocity 1

M041 R0 actor Bldg. CCW Serv. Water Flow, A&B 2

M042, M043 G;nerator H Pressure 1

M044 7

Rcdwaste Disch. Flow to Circ. Water Disch., A&B 2

M045, M046 Total Condensate Flow 1

M047 Reference Vessel Metal Temp., 1-8 8

M050-M057 Ambient Temp.

1 M058 Offgas Temp.

1 M064 Suppression Chamber Dew Point, 1-3 3

M068-M070 Suppression Chamber Level 1

M071 Rcdwaste Monitoring Tank Level, Tanks A,B,C 3

M072-M074 Refueling Floor Vent. Exh. Rad. Mon.

1 M079 Turbine Steam Pressure 1

T000 Turbine First Stage Pressure 1

T001 Low Pressure Turbine Inlet Press., 1-4 4

T002-7005 Turb.-Gen. Oil Temp. to Cooler 1

T006 Turb.-Gen. Oil Temp. from rooler 1

T007 Reactor Feedpump Bearing Temp., 1,2,3 9

WOOO-WOO 8 Reactor Feedoump Motor Temp., 1,2,3 6

WOO 9-WO14 condensate pump Motor Temp., 1,2,3 6

W O 15-16,W O 18-19,W O 21-2 2 Condensate Pump Motor Thrust Brg. Temp., 1,2,3 3

WOl7,WO20,WO23 Saa Water Pump Motor Temp., 1&2 4

WO24,WC25,WO27,WO28 Saa Water Pump Motor Thrust Brg. Temo.,1&2 2

WO26, WO29 Control Rod Drive Pump Motor Temp., 1&2 4

WO30, WO33 Turbine Bearing Temp.

2 WO34-WO37 Turbine Thrust Bearing Drain memp.

2 WO38, WO39 Turbine Generator Oil Drain Temp., 1-8 8

WO40-WO47 Alterex Oil Drain Temp.

2 WO48, WO49 Rccirculation M-G Set Motor Bearing Temp., A&B 4

WO50-WO53 R; circulation M-G Set Gen. Bearing Temo., A&R 4

WO54-WO57 RCcirc. M-G Set Impeller / Runner Brg. Temp., A&B 8

WO58-WO65 RCcirc. Pump Motor Thrust Brg. Temp., A&B 2

WO66, WO67 R; circ. Pump Cavity Seal Temp., A&B 4

WO68-WO71

DIGI *AL IMPUPS (CONTACTS except

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NO. OF SIGNALR POINT ID RIGNAL 4

D558-D561 4

A504-A507 Steam Line High Flow, A,R,C,D Steam Line Leak Detection, A,B,C,D 1

A515 Scram Discharge Volume Not Drained 1

A516 Rafueling Interlock 1

A517 Control Rod Timer Malfunction 1

AS42 Rod Out Block l

A558

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RPIS Inoperative 1

A546 Rod Drift Alarm 1

A547 Rod Selected and Driving 1

A548 Control Rod Withdrawal 1

A518 RWM Block 1

A519 SRM Detector Retracted 1

A520 SRM High Count 1

A521 1

A522 SRM Inoperative in Full Position 1

A523 IRM Detector Not IRM Downscale 1

A524 IRM Inoperative 1

A525 IRM High Flux 1

A533 SRM Bypassed 1

A534 IRM Bypassed 8

A550-A557 IBM Flux Trip Ri-Hi, A-H 1

A526 APRM Downscale 1

A527 APRM High Flux 1

A528 1

A529 APRM Inocerative Flow Converter Upscale / Inoperative 1

A530 RBM Downscale 1

A531 RBM High Flux 1

A532 RBM Inoperative 6

A535-A540 APRM Bypassed, A-F 1

A541 1

A543 RBM Bypass Flow Converter comparator Alarm 6

D544-D549 APRM Flux Hi-Hi, A-F 2

AS49-A550 A&B 1

A551 Rod Sequence Select, Shutdown Margin Select 1

A552 RWM Operating 1

A556 Low Power Level Alarm 1

A557 Low Power Level Set Point 1

A561 System Diagnostic 1

A553 RWM Rod Select Permissive Echo 1

A554 RWM Rod Withdraw Permissive Echo 1

A555 4

D500-D503 RWM Rod Insert Permissive Echo Disch. Vol. High Water Level Scram Trip, A-D 4

D504-D507 Condenser Low Vacuum, A-D 4

D508-D511 Isolation value Not Fully Open, A-D 4

D512-D515 4

D516-D519 Drywell High Pressure, A-D Reactor High Pressure, A-D 4

D520-D523 4

D524-D527 Reactor Water Low Level, A-D 4

n528-D531 Steam Line High Radiation, A-D Reactor Neutron Mon. Trip, A-D 2

D534,D535 Reactor Scram, A,B 4

D536-D539 Stop Valve Closed, A-D

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DIGITAL INPUTS (CONTACT 9 except

• )

NO. OF MIGNAL SIGNALS POINT ID T-G Load Rejection Scram Trip, A-D 4

D540-D543 Recctor Manual Scram, A,B 2

D532, D533 Rscctor Full Scram 1

D562 Gsnerator Differential Trip 1

D580 Ganerator Neutral Overvoltage Trip 1

D582 GGnerator Negative Sequence Overcurrent Trip 1

0583 Ganerator Loss of Field Trip 1

D584 GGnerator Overcurrent Distance Trip 1

D586 Gen. Startup Overcurrent and Overvoltage Trip 1

D587 Unit Differential Trip 1

D581 Aux. Transformer Differential Trip 1

D588 Aux. Transf. Ground Overcurrent Trip 1

0589

-Aux. Transf. Overcurrent Trip 1

DS90 Gonerator Overexcitation 1

D585 Startup Transf. Differential Trip 1

D591 Startup Transf. Ground Overcurrent Trip 1

D592 Startup Transf. Overcurrent Trip 1

0593 Generator Startor Coolant Trip 1

D594 Thrust Bearing Wear Prip 1

DS95 vcccuum Trip, 1,2 2

DS96, DS97 Emerg. Turbine Manual /Overspeed Trip 1

DS98 Reactor High Water Level Trip 1

DS99 closed Turbine Valves Generator Protection 1

D600 Moisture Separator High Level, Tank A-D 4

D601-D604

• Main Generator Watt-Hour Gross Output 1

Y503

• Main Transformer Watt-Hour Net Output 1

YSO4 Rofueling Floor Vent. Exhaust Radiation 1

D607 Main, Aux., Startup Transf. and Gen. Watt-Hr.

30 E504-E533 Loss Generator Potential Transf., 1,2 2

E500, E501 Loss Main and Startup Transf. Pot. Transf.

2 E502, E503 Condensate Demin Effluent Strainer Diff Press Hi 1 F500 Condensate Demin Regeneration Trouble 1

F501 Condensate Demin Exhausted 1

F502 Condensate Demin Ef fluent Conductivity Hi 1

F503 Cleanup Sludge Rec. Tank Level Hi, Tanks 1&2 2

F504, F505 Cloanup Backwash Rec. Tank Level Hi/ Low 2

. F506 Condensate Pump 1,2,3 3

F508-F510 Rocctor Feedpump 1-1, 1-2, 1-3 3

F511-F513 Sac Wa te r Pump 1-1, 1-2 2

F514, F515 Mgkeup Demin.' Anion Regen Tank D Cond. Mi 1

F516 Mckeup Demin. Mixed Bed Disch. Cond. Hi 1

F517 REdwaste Clean Waste Tank A,B Level Hi/ Low 4

M510-M513 Radwaste Trtd Wtr Holdup Tank A-D Level Hi/ Low 8

M514-M521 Torus and Drywell Abs. Press.

19 M540-M558 Instrument Servo Motor Disabled 1

M562 Drywell or Suppression Chamber Status 2

M560, M561 Control Rod Selected 8

%500-Z507 Control Rod Position 10 2508-Z 517 TIP Machine Readv A-D 4

A562-A565 TIP Guide Tube ~.SD, A-D 4

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