ML20149F940

Revised Nuclear Operations Procedure Manual NOP8301, Conduct of Operations. Related Info Encl

ML20149F940
Person / Time
Site:	Pilgrim
Issue date:	06/17/1983
From:	Harrington W, Howard J BOSTON EDISON CO.
To:
Shared Package
ML20149F593	List: ML20149F589 ML20149F667 ML20149F940
References
FOIA-87-644 NOP8301, NUDOCS 8801140397
Download: ML20149F940 (23)

Text

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NUC1. EAR OPERATIONS PROCEDURE MAWAL COND' JCT OF OPERATIONS Title N008331 FRO:LDt?I No.

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, CONDUCT OF OPERATIONS 1.0 References e

1.1 ANSI W18.7 - Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants.

1.2 Regulatory 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 Specifi:a-tions, rules, regulations, and Orders of the NR 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 emergin:p coNiitions shall be documented.

The safety and general welf are 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 the reactor down when, in his judgement, continued operation would jeopardize the health and welf are of the general putic or the safety of plant equipment. In an emergency, any lit,ensed operatoe has the saw authority.

3.0 Shift Turnover Shif t turnover shall be diligently performed in a conscientious professional and thorough menner. As a minimum, shift turnove-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 o annunciator status l

o major work in progress o status of surveillance testing o planned operational activities including power maneuvers l

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- R m sto 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 ktivities should be minimized during turnover.

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

These shall be reviewed frequently by appropriate managers.

Shift records such as operating log books, data sheets, check lists, recorder charts, computer printouts, and maintenance requests that describe or record op2 rating information and actions must be legible, accurate, complete and understandable.

When these records require correction, a single line shall be drawm 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 Plant ,5ystems and Equipment ,

Measures shall be established and implemented to assure that:

- Only licensed operators are permitted to manipulate the controls that directly affect reactivity (100FR50.54 (1))

- Licensed operators are required to be present at the controls at all times during the operation of the facility (10;FR50.54 (k))

- Operation of mechanisms 2nd apparatus other than controls which may indirectly affect the poser level or reactivity of a rea: tor l

' shall only be accomplisned w'th the knowledge and consent of an operator licensed in accordance with Part 55 (100FR50.54 (j))

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' - A licensed senior operitor shall be present at the facility or readily available on call r,t 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 pub 1'c health and

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• NDP8301

., REVISED 6/17/83 safety and no action consistent with 1.tcense/ Technical Specifi-cation conditions that can provide adequate or equivalent pro-tection is immediately apparent. Prier to taking these permitted actions (which may be contrary to 1T. eense/ Technical Specification conditions) approval, as a minimum, shall be by a licensed senior

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operator (10CFR50.54 (x),(y)).  ;

• Notification of protective actions taken in accordance wit'n 100FR 50.54 (m) 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 thereafter (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 actions 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 f acility which would have an adverse affect on the health and safety of the public, plant workers or the individuals).

Nt 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 implesenting 100rR23.

The status of plant systems and equipment shall be continuously monitored. Monitoring meth9ds and techniques employed at Pilgriri l include, but are not limited to personnel tests /inspe:tions/eaamina-l tions, annunciations, indicating lights, indicators, recorders and CRT/ computer.

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

Operators shall have confidence in the instrument readings'annun:ia-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 implewented 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.

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• The station systems, structures, and components are divided into A list for Q sys-three categories. These are Q, non-Q, and 1/Q.

tems, structures, and components is provided as defined by the BE0a*.

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• - REVISED 6/17/83 C* *

. A second list is provided by the Station Manager and is the 1/0 list. The remainder of the systems, structures, and components are non-Q and are not specifically listed.

• Modification is any change to a station system, structurk, or com-ponent that requires a change to a controlled desigfdocment. A modification for Q and non-Q can only be implemented by gn approved design change or Temporary Modification.

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

• For 1/Q 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 wort controls specified herein do not apply.

• permission to release any system or equips nt for maintenance or testing shall be given by the SRO in charge of the watch. The granting of the peruission shall be documented. Prior to granting release of equipment, a determination 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 tag;. The tags will be placed at locations where the equipment could be operated. Tags may only be placed by qualified individuals frm the group responsible for the operation of the equipment. Tags placed on control room components shall be placed so 45 not to obstruct instruments, controls or indicating lights. Equipment released for testing or returned to service after 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 109 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.

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• All Q-listed equipment which has failed required tests or lacks i

dxumentation attesting to its operability will be identified as

• l being deffetent and logged in the Control Room logbook and/or the l

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NOP8301

• REVISEO

- 6/17/83 i ,

Operators Shif t Turnover sheet. When Q-listed equipment is lacking proper documentation, it is considered to be inoperable unless otherwise resolved under the guidance of the BEco QA Manual.

r 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 perfomed.

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• When an ECCS, ECCS Subsystem, RPS or Primary Containwnt Isolation Systes 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 align 6ents will be verified prior to startup after 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 aligWnt.

' Note (1):

Independent verific!tions may be perfomed by:

a. Direct method - che: king appropriate equipment and/or control s.

or,

b. Indirect method - obtervation of indicators and/or status lights.

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

The status of safety-related (Q tist) systems shall be maintained at all times, and abnormal system alignment should be avoided ence.'t when absolutely necessary.

6.0 Unit Trips / Reactor Shutdown All unit trips shall be thoroughly investigated, and of Cause a report the trippre-shall pared and submitted to the Station Manager.

be determined, and meant 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 a:-

tor, reportable occurrences, and actions taken outside of plant 1

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, REVISED l

. 6/17/83 l

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• 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 co'aply with the station license (PNPS Technical Specifjcation),

actions shall be initiated .t that the specified condition is' attained within the prescri4 4 time period.

The cause of a scram or an nexplained power reduction must be investigated and determined before the reactor is returned to powea. 1 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 fortn-righ,tness is evident in the investigation.

Operating personnel shall be encouraged to express their concern or suggestions on how to improve performan:e 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 maintenan:e l

or refueling, will be issued by the Station Manager or his designated alternate.

7.0 Instruc_tions, Procedures. Drawings The use of prMedures shall comply with the Tech. Spe: 5. and with Reg. Guide 1.33.

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

Adherence to procedures is an essential ingredient of good station

, perfomance and configuration control. Adherence recognizes that procedures are not blindly followed without esercising 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 judpent. In these instances the indivikal should seek help and advice and whenever time pemits the procedure must be changed or, if one does not exist, a procedure must be provided. In all cases where evolutions were performed without a procedure or deviations from a procedure were performed, the event shall be evaluated by appropriate levels of management.

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V Et,V V M ' O %#1 8 b.66 v r.* 3 rage s or y NOP8301 a-REVISED

• 6/17/83 8.0 Temporary Procedures (TPs)

Approved temporary procedures may be issued to direct operations during testing, refueling, maintenance and modificationk 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 w5en the plant, a systes cria component of a system is performing in a manner not covered by existing det' ailed procedures or has been modified or extended in such a manner that portions of existing procedures 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 implewntation is necessary. These "SRO 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) Itcense.

These "5RD changes" shall be administrative 1y controlled so as to allow for proper reviews, within a specific time frame, subsequent to implementation.

10.0 C,ontro1_ Room Work Atmosphere Strict discipline and attendance to instruments and alarms as we'1 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 i' it does not detract from attendance of the controls, and is limitet to those positions which most remain in the Control Room due to license requirements or operational considerations.

"e Watch Engineer is responsible for overall Control Room Super-

v. en. Access to the Control Room shall be limited to persons wt:

have a need or requirement to be there. Examples are as follows:

- Persons required to perform 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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. k'Yl$ED 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 Cogunications '

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-tioH, the receiver shall be required to write the directions down and repeat them back to the sender.

The text of written coenunications 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 perschnel responsible for the described activity.

12.0 Behavior Observation Supervisors shall be aware of and observe employees and contrac- "

tors for aberrant behavior, including argumentative hostility towa-d authority, irresponsibility poor reaction to stress, and suspicion of being under the influence of alcohol or drugs while on Company .

property.

A Continual Behavior Observation Training brochure / package will be prepared by the Nuclear Training Department and provided to supe -

visors so as to give them guidance on situations to be aware of and As a minimum, the following two basic situs-what actions to take.

tions/ actions will be addressed:

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

Action:

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

- The Station Manager should then discuss the situation with l the VPWO 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.

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N0P8301 REYlSED 6/17/83 Action:

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

herself, others or equipment by talking with them, and if l necessary subduing them. 3 .

- Request Security Assistance.

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- Arrange for transportation to the hospital.  ;

- Notify Station Manager.

- The Station Manager should notify the VPWO and the Medical Departoent.

• 13.0 Ov,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 and the use of overtime:

o dequate 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 shif t turnover time).

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

Recognizing that unusual circumstances may arise requiring deviatio*.

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 redu:-

tions in the effectiveness of operating personnel would be highly unlikely.

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. ATTACHMENT B

J ASPENDIX 2 GEPAC 4020 I/O LIST ANALOG INPUTS DIGITAL INPUTS l

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4 ANALOG INPUTS (0-160mv)

40. OF SIGNALM POINT.ID SIGMAL 120 A000-A119 LPM Level 6 8000-R005 APM Level 2 B006-8007 R9M A&B 4 9008-3011 TIP Level, A-D 1 3013 Reactor Pressure 1 3014 Reactor Core Pressure Drop 1 3015 total Jet Pump Flow (Core Flow) 4 B016,3038,9039,5060 Rocirculation Drive Flow, Loops A1, A2,81,82 1 R01.7 ccntrol Rod Drive System Flow 2 3018, 8019 Reactor Feedwater Inlet Flow, A&R 2 9020, 8021 Cloanup Flow, A & B 2 9022, 8023 Rceirculation Pump Motor Power A&B 1 3024 Recotor Water Level. 1 B025 l Reactor Outlet steam Flow 2 M026, 3027 l Clocnup Rystem Temp, inlet 6 outlet 4 B028-R031 Rocetor Feedwater Inlet Temp. , A1, A2,R1,82 4 B052-3055 R:circul ation Inlet Temp. , Loops A1, A2,31,32 1 E001 l Main Transformer Hot Spot Temp. 1&2 2 E002, E003 -

Aux. Transformer Hot Spot temp. , 2 E004, E005 Startup transformer Mot Spot Temp. ,162 1 E006 Shutdown Transformer Not 4 pot Temp. 4 E007-E010 345 KV Switchyard 1, Line 342 V, A,W, VAR 4 E011-E014 345 Kv Switchyard 1, Line 355 V, A,W, VAR 1 E015 l

' Isolated Phase Bus Air Temp. 2 E016, E017 Racire. Pump Motor Winding temp. ,162 2 E018, E019 Recire. Pump M-G Set Winding Temp. ,142 2 F000, ?001 Drain Cooler Temp., Drain A,9 1 F002 Candensate Domin. Dif f. Pressure 1 7003 Stoam Real Meader Pressure 1 F004 Glond Real condenser Pressure 1 F005 Reactor Bldg. Vent Exhaust Dif f. Press. - 1 F006 Barometric Pressure 1 F007 Reactor Feedpump Suction Press. 3 F008-r010 Recctor Feedpump Discharge Press. , A,M,C 1 F011 Condensate Pump Discharge Reader Press. 2 F012, F013 Condenser pressure, west & east 10 F014-F023 Feedwater Meater Extraction Press 1A-5A,15-5B 1 F024 i Air Itjector Of fgas Flow 1 F025 Makeup Water Flow to Demin. Water Storage tank 2 F026, F027 Radwaste Chemical waste tank Level, A&B 2 F028, F029 I

Cecctor Bldg. Closed Cooling water Flow, A&B 2 F030, F031 i BER Cool.ing Water Flow, A&B 2 F032, F033 l RBR Water Flow, A&B 1 F034 Bjector Condensate Outlet Temp. 1 F035 Drain Cooler A-B Inlet gesp. 1 F038 Gland Seal Condensate Outlet Temp. 2 F039, F040 20s Water Inlet temp. ,1-1 & 1-2 4 F041-F044 Som Water Outlet temp.,1W,2W,3E,4E 2 F045-F046 Condenser Motwell Outlet temp. , east 6 west 10 F047-F056 Noater Inlet temp. (tsw) , 1st-5th points l

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ANALOG INPUTS (0-160mv)

40. OF SIGNAL SIGNALS POINT'ID N: ster Outlet Temp. (EnW) , 1st & 3rd points 4 F057-F060 Naater Drain Temp. (Train A&B) , 1st-5th points 10 F061-F070 Reactor Bldg. CCW Serv. Water Inlet Temp. A&B 2 F071-F072 Reactor 31dg. CCW Serv. Water outlet Temp. AEB 2 F073-F074 Re:ctor Midg. PCW Heat Exch. Inlet Temp, A&B 2 F075-F076 R:a3 tor 51dg. CCW Reat Exch. Outlet Temp, A&M 2 F077-F078 RRR CW Inlet Temp. , A&B 2 F079, F080 RRR CW Outlet Temp. , A&M 2 F081., F082 RRR Water Inlet Temp. to Meat Exch. A&B 2 F083, F004 RNR Water Outlet ' temp., Meat Exch. AEB 2 F085, F086 833 Mater Pump I,evel, AaB 2 F087, F088 Radwaste Disch. went tracing Flow 1 F089 Reactor Feedwater Dif f. Press. , AEB 2 7092, F093 G;terator Voltage 1 G000 Generator Mtator Current, Phase A,M,C 3 G006-G008 Main Transformer , Net Amos 1 E000 G:nerator Gross Power 1 B012 GGrerator M, Seal Oil Temp. 1 G002 G nerator Field voltage 1 G004 G nerator Field Anps 1 G005 Gsnerator VARs 1 G001 Ccnerator Stator Temp., Phase A,B,C 3 G009-G011 l Stator Cooling Meader Inlet Temo. 1 G012 Stator Cooling Reader Outlet Temp. 1 G013 G:nerator collector Air Inlet Temp. 1 G014 Generator collector Air outlet Temp. 1 0015 Alterex Stator Winding 'emp. , 1-3 3 G016-G018 Alternator Air Inlet & Outlet Temp. , Point 1 2 G019, G020 Diosel-Generator winding Temp. ,162 2 G022, G023 Alterex Diode Cooling Water Outlet Temp. 1 C024 R: actor Feedpump Motor Winding temp. ,1,2,3 3 G025-0027 ccndensate Pump Motor Winding Temp. , 1,2,3 3 C028-G030 . ,

! 803 Water Pump Motor Winding Temp. , 1&2 2 G031, G032 Ejector Radication Monitor 1 M000 stack Gas Radiation Monitor 1 M001 Radwaste Discharge Radiation Monitor 1 M003 Radwaste Disch. to Circulating Water PR 1 M004 Reactor Bldg. Vent. Exhaust Temp. 1 M005 Reactor totton Mead Water Temp. 1 M077 Relief Valve Temp. for I,eak Detection 6 M006-M010,M078 Drywell Containment Temp., 1-7 7 M011-M017 Suppr e s s ion Ch ambe r Temp. , poin t s 1,2,3 3 M018-M020 M021-M0 2 4,M 0 6 5-M067 Drywell Containment Dew Point, 1-7 7 Reactor Vessel Metal Temp. ,1-8 8 M025-M032 Plant Meating Steam Flew 1 M033 Prossure Torus - Mrus Reference vessel 1 M034 Pressure Drywell - Drywell Reference Vessel 1 M035 Reactor Feedwater conductivity, A&B 2 M037, M048 Reactor Feedwater FM, A&R 2 ,M038, M049 Reactor Feedwater Turbidity, A&R 2 M039, M080 Coodensate Domin water to FW Meaters 02 content 1 M036

4MALOG INDUTS (0-160mv) .

NO. Or 31GNALg pogyt go SIGMAL 1 M040 Cind Direction 1 M041 Wind velocity Reactor tidq. CCW Serv. Water Flow, A&B 2 M042, M043 1 M044 Generator M Pressure 4045, M046 Ra8vaste Didch. Flow to Cire. Water Disch., A4B 2 1 M047 Total Condensate Flow 8 M050-M057 Reference Vessel Metal Temp., 1-8 M058 Ambient Temp. 1 1 M064 l Of fgas Temp. M0A8-M070 Suppression Chamber Dew Point, 1- 3 3 Suppression chamber Level a 1 M071 Radwaste Monitoring tank Level, Tanks A,B,C 3 M072-M074 R3 fueling Floor Vent. Exh. Rad. Mon. 1 M079 1 7000 turbine steam Pressure 1 T001 Turbine First Stage Pressure l

Low Pressure Turbine Inlet Press.,1-4 4 T002-7005 g Turb.-Gen. Oil Temp. to cooler 1 T006 Turb.-Gen. Oil Temp. from rooler 1 T007 l '

Roactor Feedpump Bearing Temp., 1,2,3 9 W000-W008

! 6 W009-W014 Reactor Feedoump Motor Temp., 1,2,3 W 015-16,W 018-19,W 021-2 2 ccndensate pump Motor temp., 1,2,3 6 ccndensate Pump Motor Thrust Brg. Temp., 1,2,3 3 WO17,W020,WO23 WO 2 4,W 02 5,W 0 2 7,W 028 saa Water Pump Motor 'emp., 162 4 WO26, W029 Rea Water Pump Motor Thrust 4rg. Temo. ,162 2 centrol Rod Drive Pump Motor Temp., 162 4 WO30, WO33 Turbine Bearing temp. 2 WO34-W037 Turbine Thrust Bearing Drain 'emp. 2 WO38, WD39 Turbine Generator 011 Drain Temp. ,1-8 8 W040-WO47 2 WO48, WO49 Alterex 011 Drain Temp.

Accirculation M-G Sat Motor Bearing Temp. , A&B 4 WO50-WO53 RCcirculation M-G Set Gen. tearing Temo. , A&M 4 W054-WO57 Accire. M-G set Impeller / Runner arg. Temp. , Als 8 WO58-W065 2 WO66, WO67 Rccire. Pump Motor Thrust Brg. temo. , AE8 WO68-WO71 R;cire. Pump Cavity seal Temp., A&B 4

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DIGl*AL INPUPS (CONTACTS except *L MO. OF _..

SIGNALA POINT ID RIGNAL 4 D558-D561 4 A504-A507 8 team Line High Flow, A,M,C ,0 1 A515 Steam Line Leak Detection, A,B,C,D A516 Scram Discharge volume Not Drained 1 1 A517 Rofueling Interlock 1 A542 control Rod Timer Malfunction 1 A558 Rod out Block

• 1 A546 RPIS Inoperative 1 A547 Cod Orif t Alarm 1 A548 Rod Selected and Driving 1 A518 Control Rod withdrawal 1 A519 RWM Block 1 A520 SRM Detector Retracted 1 A521 SRM Migh Count 1 A522 SRM Inoperative 1 A523 IRM Detector Hot in Full Position 1 A524 IRM Downscale 1 A525 IRM Inoperative 1 A533 IRM Migh Flux 1 A534 SRM Sypassed 8 A550-A557 IRM Myoassed 1 A526 IRM Flux Trip Mi-Mi , A-M 1 A527 APRM Downscale 1 A528 APRM Migh Flux 1 A529 APRM Inonerative 1 A530 l Flow converter Upscale /Inoper ative 1 A531 l

l QN4 Downscale 1 A532 DBM Migh Flux 6 A535-A540 RMM Inoperative 1 AS41 APRM typassed, A-F 1 A543 RBM typass 6 DS44-D549 l Flow Converter comparator Alarm 2 A549-A550 APRM Flux Mi-Mi, A-F 1 A551 Rod Sequence telect, A&B 1 A552 Shutdown Margin Select 1 A556 RWM Opeeating 1 A557 Low Power Level Alarm 1 A561 Low Power Level Set Point 1 A553 System Diagnostic 1 A554 RWM mod Select Permissive Echo 1 A555 RWM ' Rod Withdraw Permissive Echo 4 D500-D503 RWM Rod Insert Permissive EchoDisch.44 Vol. Eigh Water Level D504..D507 D508-D511 Condenser Low vacuum, A-D 4 D512-D515 Isolation value Not Fully Open, A-D 4 D516-D519 Drywell sigh Pressure, A-D 4 D520-D523 Reactor Righ Pressure, A-D 4 D524-0527 Reactor Water Low Level, A-D 4 n528-D531 Steam Line Migh Radiation, A-D 2 0534,D535 Reactor Meutron Mon. Trip, A-D 4 0536-0539 Reactor Scram, A,3 Stop Valve closed, A-D

I a . '. ' .

• DIGITAL INPUTS (CONTACTS except *)

M0. OF SIGNALM PC IMP ID MIGNAL 4 D540-D543 t-G Load Rejection Scram Trip, A-D 2 D532, D533 Roactor Manual Meram, A,5 1 D562 Roactor Full Scram 1 D580 Generator Differential Trip 1 D582 Generator Meutral overvoltage Trip 1 D583 Generator Megative sequence Overcurrent Trip 1 D584 Conerator Loss of Field Trip 1 D586 Gonerator Overcurrent Distance trip 1 D587 Gon. Startup Overcurrent and Overvoltage Trip 1 D581 Unit Dif ferential Trip 1 D588 Aux. Transformer Differential Trip 1 D589 Aux. Transf. Ground overcurrent Trip 1 D590 Aux. Transf. Overcurrent Trip 1 D585 Generator overexcitation 1 D591 Startup Transf. Dif ferential Trip 1 D592 startup Transf. Ground overcurrent Trip 1 0593 Startup Transf. Overcurrent Trip 1 D594 Generator Startor coolant Trip 1 D595 Thrust seating Wear

• rip 2 0596, 0597 vaccuum erip, 1,2 1 D598 Emerg. Turbine Manual /overspeed Trip 1 D599 Reactor Migh Water Level Trio 1 0600 closed Turbine Valves Generator Protection 4 0601-D604 Moisture Reparator Nigh Level, Tank A-D 1 Y503

cMain Generator Watt-Mour Gross Output 1 Y504 oMain Transformer Watt-Mour Net output 1 D607

(

Rofueling Floor Vent. Exhaust Radiation 30 E504-E533 Main, Aux. , Startup Transf. and Gen. Watt-Mr. 2 E500, E501 Loss Generator Potential transf., 1,2 2 E502, E503 Loss Main and Startup Transf. Pot. Transf. 1 F500 Condensate Demin Ef fluent Strainer Dif f Dress Mi 1 F501 cendensate Domin Regeneration Trouble 1 F502 Condensate Demin Exhausted 1 F503 Condensate Demin Ef fluent conductivity Ri 2 F504, F505 ,

Cleanuo sludge Rec. Tank Level Mi, Tanks 1&2 2 .F506 Cleanup Backwash Rec. Tank Level Mi/ Low 3 F508-F510

' condensate Pump 1,2,3 3 F511-F513 Reactor Feedpump 1- 1, 1- 2, 1- 3 2 F514, F515 Sea Water Pump 1-1,1-2 1 F516 Makeup Domin. Anton Regen Tank D Cond. Mi 1 F517 Makeup Domin. Mixed Bed Disch. Cond. Mi 4 M510-M513 Radweste Clean Waste Tank A,3 Level Mi/ Low 8 M514-M521 Radwaste Trtd Wtr Moldup Tank A-D Level Mi/ Low 19 M540-M558 Torus and Drywell Abs. Press. 1 M562 Instrument Servo Motor Disabled 2 M560, M561 Drywell or suppression Chamber Status 8 1500-5507 Control Rod Selected 10 1508-2517 Control Rod Position 4 A562-A565 TIP Machine Readv A-D 4 1518, t.5 2 2,25 2 6, f.5 3 0 TIP Guide Tube LSD, A-D l

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PDC 86-75, Rev. 0 Safety Evaluat on. Rev. 0

" '1 'AttachmentQ. .

113 f Sheet iof 3 GENERAL h ELECTRIC J

l STANDBY LIQUID CONTROL SYSTEM .,

- CONTROL CAPACITY .

EQUIVALD8CY REPORT PREPARED FOR THE BOSTON EDISON COMPANY PILGRIM NUCLEAR POWER STATION JANUARY.29, 1987 f, - -

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PREPARED BY: R.T. EARLE 2'T%bwh &h .

. .g VF.RIFIED BY: J.K. SAWABE

),% Q I 4 M7 Verification Material in ORF C41-00095/2, Section L4.

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PDC 86-75, Rev. 0 3, 3 Safety Evaluati n, Rev. 0 l Attachment 1 l Sheet p_ of c) 2g2 / )

DISCLAIMER OF RESPONSIBILITY ,

This document was prepared by or for the General Electric Company. ,

Neither the General Electric Company nor any of the contributors to l

this document: -

l l

A. Makes any"warranty or representation, express or implied, with respect to the accuracy, completeness, or usefulness of the  !

information contained in this document, or that the '::-e of any - i information disclosed in.this document may not infri.ss privately owned rights; or, .

B. Assumes any responsibility for liability or damage of any kind which may result from the use of any information disclosed in this document.

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'1 PDC 86-75 Rev. 0 Safety Evaluation

.' \\ $ '

1 Attachment 2 Q , Rev. 0

.) Sheet 3 of *) 2.11l

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3 .

1.

ABSTRACT .

This documdnt was prepared for the Boston Edison Company to address the requirem nts of the Standby Liquid Control System (SLCs) at the s Pilgrim Nuclear Power Station for compliance with the NRC ATWS. Rule 10CFR50.62. The plant specific values used to demonstrate compliance with the NRC A'NS Rule are the same as the minimum values provided in tha system Technical Specifications. .

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PDC 86-75, Rev. 0 Safety Evaluation Rev. 0 ,

l

' i Attachment 1 Q-2.sg; l Sheet f of 3 1

~

TABLE OF CONTENTS 11 Abstract 1

1. Introduction . .

1 ,

2. Discussion 1

2.1 SLC System Design Basis .

. 1 p 2.2 NRC Mws Rule * .

.O v

1

3. inalysis .

2 3.1 Equivalent Control Capacity Definition .

3 3.2 Equivalent control Ca'paci,ty, Calculation -

4

4. Surernary -

5

5. R.eferences e

I ,

e e

111 - .

PDC 86-75, Rev. O

' Safety Evtluatiop, Rev. 0 a i Attachment 1, SheetSof9

. ( #83 i 1.0 Iw!1toDUCTIO'I .

Boston Edison has requested an evaluation of the minimum required l concentration (weight percent) of sodium pentaborate for the Pilgrim .

St'andby Liquid Control System to comply with the NRC AWS rule rcquirements in 10CFR50.62 (Reference 3. ) . The minimum concentration l is to be based on equivalency to the minimum 86 gym,13-weight percent '

codium pentaborate control capacity requirement ~ stated in the NRC AWS rule. Equivalency.is calculated using the ratio of the specific Pilgrim mir?. anum values to reference plant values that the rdle is based on.

- For Pilgrim, a minimum solution concentration of 8.42 percent is required. This is based on the assumptions that the ninfamm ~

• "ri*h**"* ' 'h* * **"* **"* *" ' * ** * ** *hY * ** * *'*"d " 5 4 - 5 bf5 3 "atom percent, one pump is required to operate and the actual capacity of each pump exceeds the required minimum pump flow rate. .

3, 2.0 DISCUSSION h 2.1 SLC System Design Basis The generic design basis for the. SLC System is to provide a specified cold boron shutdown concentration The SLC to the reactor System wasvessel typically as described designed in NEDE-24222 (Ref erence 4. ) .

to provide the specified cold shutdown concentration in about one or two hours. During reload licensing evaluations, this shutdown concentration is verified by analysis to be adequate to render the core subcritical. The considerations' in the reload evaluation are -

independent of AW S and injection rate is not directly considered.

The AWS rule requires the addition of a new design requirement to the -

generic SLC System design basis. Changes to flow rate, solution concentration or boron enrichment, to meet the AWS Rule, must not invalidate the original system design basis.

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PDC 86-75. Rev. 0 Safety Eva1uatiqn, Rev* o Att:chment i

{h9

,- Sheet (,org 2.2 NRC A'IVS Rule Paragraph (c)(4) of 10CFR50.62 states, it. part:

1. Each boiling water reactor must have a Ltandby Liquid control ,

system (SLCS) with a minimum flow capacity and boron content equivalent in control capacity to 86 gallons per minute of 13-weight percent sodium pentaborate solution."

The NRC Staff has provided clarification of equivalent: cont ol capacity (' Reference 5.) as follows:

(1) The "equivalen't"In control capacity" wording was choosen to allow flexibility in the implementation of the requirement. For example, the

~

Gquivalence can belob'tained by incYeasing flow rate, boron -

concentration or boron enrichment. ,

"12T The 86 gallons per minute and 13-weight percent sodium pentaborate were values used in NEDE-24222, "assessment of BWR Mitigation of ATWS, Voluines I and II", December 1979, for BWR/4, BWR/5 and BWR/6 plants .

with wouldabe251-inch equivalentvessel -inside for smallpr plants diameter.

was recognised TheinfNEDE-24222.

act that differen "The flow rates given here are normalised frca a 251-inch i.e.,

diameter vesse1~ plant to a 218-inch diameter vessel plant, the 66 gym control liquid injection rate in a 218 is equivalent l

to 86 gpm in a 251. This is done to bound the analysis....(pp.

2-15 [NEDE-24222))." '

are vessel b'oron concentration required to achiev

~

time required to achieve that vessel boron concentration. The minimally acceptable system should show an equivalence in the .

parameters to the 251-inch diameter vessel studied in NEDE-24222.

3.0 ANALYSIS l

3.1 Equivalent Control Capacity The NRC equivalent control capacity concept of the A'IVS rule is a very l simple, direct cri'terion that does not require consideration ific core nuclear of the l ih l

I

. mixing efficiency or to account for plant-speccharacteristics

. , PDC 86-75 Rev. 0

' Safety Evaluation, Rey , o Attachment 1 Sheet J ofc) 2tgy i is shown to the equivalency requirement if the following relationsh p .

be true:

• >= 1 (Equati,on 1) p*U251

• C E __ .

86 M 13 19.8 where the plant-specific parameters are defined as:

s 0 = minimum SLCS flow rate (one or two pump operation appropriate), gym.

5 M = mass of water in the reactor vessel and recircu ation system at the hot rated conditions, lbs. ,

c = minimum sodium pentaboratu solution concentration, weight '

percent.

isoto'pe enrichment (19.8% for natural E = minime.n expected B -

boror,), atom percent. .:. . . .

. . . . .c ~ .

g (the. mass of water in the reactor vesselis and . ..

' w

• ne va%e of M d recirculation hlten at rated conditions in the refer'enc 628,300 t l rods l l temperature, rated void content, normal water leve .

! internals dimensions.

3.2 Equivalent Control Capacity calculation he NRC requires the use of minimum ' plant-specific values'to i l .

daronstrate compliance with the equivalency requirement. i For s ng e pump operation, 54.5 atom percent boron enrichment, Pilgr m can demonstrate compliance if the following relationship is true:

(Equation 2.) ,

C>= 13

• 5

• 8.6

• 19.8 M 251 O i (weight whe're C is this case is the minimum allowed h i imum allowed concentrat on l

percent) of the sodium pentaborate solution, Q is t e m n d l individual pump flow rate, M is the mass of water in i hthe ent reactor an recirculation lines, and E is the minimum allowed boron; enr c m

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PDC 86-75. Rev. 0 Safety Evaluatiop, Rey, o

.P. Attachment t

. Sheetgofg {2fy, was level. The water mass is based plent water mass. The minimum allowed individual pump flow on the same conditions l

rate as the cbtained will become part from of theReference system Technical1. The minimum Specicication andallowed i

Des gn boron enr Specification (Reference 1).

g = 39 (minimum rated) gyn .

M = 507,850 lbs ,

E = 54.5 % *

2) gives Using the current Pilgrim plant-specific values (in Equa *

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pentaborate. < .

(Equation 3.)

C >=

13

• 507,850

• 86

• y

• 628,300 39 54.5 , _

6 W- c.>= 8.42 .

4.0

SUMMARY

hd te When the concentration of enriched sodium )pentagrate is e rual to, deca or y ra (enrichment exceMing 54.5 atom percent boron BPilgrip meets ,or exc6eds th

, greater than 8.42 percent, ,

equivalency requirements. .

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PDC 86-75 Rev. 0

. .s . Safety Evaluati n. Rev. 0 Attachment 1 of9 2l3j Sheet

5.0 REFERENCES

1. Doc. No. 257BA169, Rev. O, Standby Liquid Control System Design Specification. . .

2. Doc. No. 257HA169AV, Rev. 2, DesignStandbySpecification Liquid controlData System Sheet.

3.10CFR50.62, NRC AWS Rule, June 1984.

4. NEDE-24222, Assessment of BWR Mitigation of AWS,' December 1979. .

5. USNRC. Generic Letter 85-03, Clarification Control of for capacity Equivalent Star.dby

"'"~~~

Liquid control Systems, January -

' 28, 1985.

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PDC 86-75 Rev. O Safety Evaluation No.163 Rev. O Attachment 3 Sheet of ATTACHMENT 3 Recomended Technical Specification Changes The pages of the following sections and figures of the Technical Specifications that need to be updated due to St.CS modification (PDC 86-75) have been marked with suggested updates and included in this attachment for your review.

Technical Specifications Sections 3.4 & 4.4 Technical Specifications Figures 3.4.1 & 3.4.2 .

l

.' LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.4 A 4.4 STANOBY LIOUID CONTROL SYSTEM SLA.NDBY_LIOUID CONTROL SYSTEM 6pplicability:

Anolicability:

Applies to the operating status of the Standby Liquid Control Applies to the surveillance Systee- requirements of the Standby Liquid Control System.

To assure the availability of a Obiective:

system with the capability to To verify the operability of the shutdown the reactor and maintain Standby Liquid Control System, the shutdown condition without the use of control rods.

Snecification: -

Moreal System Availability Snecification:

A.

A. Normal System Availability

1. During periods when fuel is in the reactor and prior t The operability of the Standby h S andby Liquid Liquid Control System shall be cond on verified by the performance of Control System shall be the following tests:

operable, except as specified in 3.4.B below. This system 1. At least mee per m'm g need not be operable when the h pg Mp sM k con rol functionally tested by co ion and a d n e!watertothe ci ica n 3. s t

2. At least once during each operating cycle:

a. Check that the systes relief valves trip full open at pressures less than 1800 psig, and reseat on a falling pressure greater than 1275 psig.

b. Manually initiate the system, except explosive valves.

Pump boron solution through the recirculation path and back to the Standby Liquid Control Solution Tank. Check that each pump flow rate exceeds 39 GPM against a system head of 1275 psig.

' LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS

' 3.4 STANDBY LIOUID CONTROL SYSTEM 4.4 STANDBY LIOUID CONTROL SYSTEM

c. Manually initiate one of the Standby Liquid Control System loops

~

and pump demineralized water into*the reactor vessel.

This test checks explosion of the charge associated with the tested loop, proper, operation of the valves, and pump

. operability. The replacement charges to be installed will be selected from the same manufactured batch as the tested charge.

d. Both systems.

including both explosive valves.

A shall be tested in the .

G' course of two l operating cycles.

! B. Ooeration with Inocerable B. Surveillance with Inocerable Comoonents: Comoonents:

1. From and after the date 1. When a component is found that a redundant to be inoperable, its component is made or redundant component shall found to be inoperable, be demonstrated to be Specification 3.4.A.1 operable immediately and shall be considered daily thereafter until the fulfilled and continued inoperable component is  ;

operation permitted repaired.

provided that the component is returned to ,

! an operable condition l within seven days.

l l

l 96 i

--e - - - - ----

~1 1

' ,, . LIMI, TING CONDITIONS FOR OPERATION

. SURVEILLANCE REOUIREMENTS 3.4 STANDBY LIOUID CONTROL SYSTEM 4.4 STANDBY LIOUID CONTROL SYSTEM C. Sodium Pentaborate Solution C. Sodium Pentaborate Solution At all times when the Standby The following tests shall be Liquid Control System is performed to verify the required to be operable the availability of the liquid following conditions shall be Control Solution:

. met: .

1. Volume: Check at least

1. The net volume - once per day.

concentration of the Liquid Control Solution 2. Temperature: Check at in the liquid control least once per day.  ;

tank shall be natntained as required in Figure 3. Concentration: Check at l 3.4.1. least once per month.

Also check concentration

2. The temperature of the anytime water or boron is 11guld control solution added to the solution, or shall be maintained above the solution temperature 48'F. is at or below 48'F.

3. The enrichment of the 4. Enrichment: Check liquid control solution Boron-10 enrichment level

- shall be maintained at a by test anyttee boron is boron 810 isotope added to the solution and enrichment exceeding 54.5 prior to restarting from each refueling outage.

h atos percent, Enrichment analyses shall D. If specification 3.4.A 8, or be received within 30 days C.1 or C.2 cannot be met, the of test performance. Men reactor shall be placed in a M:t n c :=  :::" P ,

Cold Shutdown Condition with F a check shall be nace to m all operable control rods ensure that Doron levels fully inserted within 24 seet the original design hours. If the enrichment criteria by comparing the requirements of specification # enrichment, concentration 3.4.C.3arenotmetfbeseg '

and volume to established

' criteria. IT the Semb lgvels alo hofAtos meef- ^ SN rf +he Bra' [theBoron-10Isotonidenrichment percent within seven days than tof54.5 L 4t. or:StM ; gj 1,#3h cr$,s a p I g leve4 hect from the line of enrichment Oe Pegetor shull be f aceal l m9 recess.fsubatt Treport to /

# ' W l

, the R and advise them of Gld Shgfdown CoHd O with d* S i')b p ans o bring the solution gli o Twll .

CriteW9 g up to a demonstratable 54.5 O grj6fe.+gehfwl  ?' tods'" 24 go es .

hvb atos percent Boron-10 .ff ef4ep th 7 tt,n A fch od _

Isotopic Enrichment. g g y y g

\ w', A l n r e v e n c h y s d SF 'M 4re. s O U hof M8b 3,4, c 3 O 97 b

I y . .

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e gygg _

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\ wanwon snatw.wwEuT wng = 9% fE h'T*M Tuo%1Mut' . . O 'Pum.eAlsT .

f W" -

tas e .

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d'oea , soon h o eooo tooo seen ,

s .

- v. mar %

ved>Ns - ( ou..M Figure 3.4-1 Sodium Penta rSte Solution

. Volume and Concentration Requirements.

._ e-

Safety Evaluation i

no.: 2IS I

} 3AFETY EVALUAT10N PILGRIM MUCLE AR POWER iTATION .-

i -

' Rev. No. O Initiator: Dept: Group:

PDC PCW System Calc.

No.:

w. % ske. neb F S L M c.

Name: No.:

! Date: 5/4/g7 8(a-75 &bd,)f4S@D3 of P 87-S04-

gescriptior)ksc)roposed Enne change, test gr experiment

! Beyon

% ,huie t Spec MarJ 19e. ifs *cs + s n' u 4e SLC

+ ten chanses, and i SAFETY f YALUATION CONCLU$10NS: "

assoc n an 4 e at ~

~

ne proposed change, test or experiment:

1. ()$ Does Not consequences o(f an accident er asifunctica safety previously evaivated in the FSAA.

r ant to of eq 3.

3.

of a different type than any evaluated .

unction p@viou for any technical specification.M Does Not ( asis * '

) Does redu h ,

l

_34525 FOR SAFETY EVKluATION CONCLUSIONS:

.5e>

A44ae kord %kee+s .

Change -

. \

$4 Reconnended Change N ) SE performed by _ ( ) Not Recommended >

N, d Data _4!G ff 7

- Exhibit 3.07-A I Sheet 1 of 3 Rev. 3 I

_ - - ~ " ' '

\, Safety Evaluation No. .'?i.h Ct.A . c . 5lA 157 A.

Descriotion of Procosed Chance. Test or Eroeriment:

This modification replaces the Standby Liquid Control Systems's (SLCS) existing sodium pentaborate solution (natural boron with 19.8 atom perc B

IO) of 9.4solution pentaborate to 16(bo weight percent concentration with an menriched sodl 8.42 to 9.22 weight perce$i enriched greater than 54.5 atos percent By0) or nt concentration.

recalibrates and revises the setpoints of the level and tempera,tureIn add sensing instruments, relocates the St.C pump 2078 test button near the button for pump 207A and locates a new pressure gauge near the test buttons.

instrumentation is required, due to the solution conce The SLC's Technical Specification change includes the solution new enriched sodium pentaborate solution. concentration requiremen B. Puroose of Chance comply (10CFRSO.62).

with the NRC's Anticipated Transient Nithou from 16 to 9.22 weight percent reduces the maximus solu temperature from 70*F to 38'F.

h Specifications requiring reactor shutdown as a result of solutionThis temperature requirements.

The additional system changas are being performed loss. to simplify testing and minimize enriched sodium pentaborate I C.

Systems. Subsv.etems. Comoonents Affected:

Standby  !! quid Control System - This modification affects the SLCS in the following manner:

l The performance of the system is improved by,,this modification. The meet the NRC ATHS Rules equivalence requireme percent of normal sodium pentaborate solution.

• Ifmeet

, the the enrichment option was not used, two pumps would be required to NRC's ATHS Rule.

With the enrichment option, the reliability of the the NRC's system is maintained, since only one pump is required to satisfy ATHS Rule.

The system retains one redundant pump.

The low crystallization temperature (38'F corresponding to a 9.22 weight percent concentration) of the enriched sodium pentaborate solution will further improve the system reliability.

reactor shutdown because of solution temperature requirements.This reduce t . . . . . -

l>alety LValustavu No. d ' hi kJ. O = %fa 1%)

The storage tank high and lowlevel alarms are being maintained at their original volume setpoints. The high level alarm alerts the operator to a solution volume near the storage tank overflow. The original volume concentration requirements were such that, should evaporation occur, a low level alarm would annunciate before the temperature-concentration requirements were exceeded. For the original solution, the maximum possible attainable concentration at the low level alarm was 14 weight percent. This corresponded to a saturation temperature of 60*F which is less than the original 65'F setpoint of the heat tracing. This ensured the operator was given an alarm before crystallization could occur from high solution concentration. The requirement for a low level alarm to annunciate before '

temperature-concentration requirements are exceeded is not needed because of the new lower solution concentration requirements (8.42'to 9.22 weight percent). Since the maximum concentration of the new solution is 9.22 weight percent, the maximum possible solution concentration. obtainable from evaporation without a high or low level alara is approximately 10.3 weight percent *, This corresponds to approximately a 44*F solution temperature (Iow solution temperature alara setpoints is 48*F). Due to the 53*F setpoint of the tank heater and heat tracing and the design room temperature of 60*F to

_ 100*F, solution concentration changes due to evaporation would be slow. The operator would be alerted to a solution concentration change from evaporation by either the low level alare or the technical specification monthly surveillance requirements before the crystallization point is reached.

• Hioh Level Alara (9.22) - (4430) (9.22) - 10.3 weight percent Low Level Alarm G950) h G. Summary The SLCS by itself cannot cause an accident and it does not interact with any other systen whose malfunction could cause an accident. Hence, this modification on the system does not increase the probability of occurrence of an accident.

This modification increases the system's control capacity to satisfy MRC ATHS rule requirements. The modified system is more effective than the existing system in bringing the reactor to the cold shutdown condition from rated power. Hence, the modification does not increase the consequences of an accident.

This modification does not call for the safety equipment of the system to work at higher pressures, temperatures and more severe conditions than the existing levels. The modification makes the SLCS pumps redundant and it does not change the logic of the system. Hence, the modification does not increase the probability of the malfunction of the equipment important to l

safety.

This modification increases the margin of safety fcr system availability by reducing the possibility of system unavailability from solution temperature requirements.

This modification increases the margin of safety for flow rate requirements (required 39 GPH; available 78 GPH) and for minimum volume of l solution requirements (required 2068 gallons at a mid-range concentration

of 8.82 percent; available 3960 gallons).

1 l

l 1

Safet4 Evaluation No. *8 0 ks.e . < . 5/A 'S '

The upper limit, 9.22 weight percent, concentration of enriched sodium pentaborate has a saturation temperature of 38'F. To preclude precipitation, the minimum solution temperature will be maintained above 48'F, which is 10'F above the saturation temperature of the maximum concentration. In order to ensure a solution temperature greater than 48*F, the technical specifications will require determination of the solution temperature daily. This frequency is considered adequate because the room minimum design temperature is 60*F and any temperature change 1 would be gradual. In addition, the daily monitoring will be backed up by l the tank heater, heat tracing, and low temperature alaras. If the solution temperature in either the tank or pump suction lines reaches 53*F, the tank heater or heat tracing will commence operation. If the solution temperature in either the tank or suction lines cont 1~nues to drop to 48'F the operator will receive an alara in the control room, l Technical specifications will then require that the reactor be placed in a i cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of a solution temperature less

- ]

than 48'F.

~

In order to comply with the ATHS Rule (10CFR50.62), the boron in the sodlue penta percent of Bgrate. The solution must be enriched technical specifications to greater will requlre that the than B10 54.5 atos enrichment be greater than 54.5 atos percent. If the 8 u enrichment is found to be less than or equal to 54.5 atos percent, the Technical Specifications will require the operator to determine if the original shutdown criteria (equivalent of 700 ppe of natural boron) can be met. If

~

the original shutdown criteria can not be met Technical Specifications will require that the reactor be placed into cold shutdown within 24 g hours. If the original shutdown criteQa can be met, Technical Specifications will require that the B' enrichment be returned to greater than 54.5 Atos percent within seven days. If at the end of this seven day period, Bl v enrichment is still less than or equal to 54.3 atos percent, Technical Specifications will require that the NRC be notified within seven days with BECo's pla01 to bring the enrichment into compliance.

This ensures that if the BlU enrichment is less than or equal to 54.5 atos percent, the operator will shutdovn if the original shutdown criteria cannot be met, or bring the enrichaent into compliancg if the original In order to ensure a B'O enrichment grea r

. shutdown @ teria can the beTechnical met. Specifications will require that the Bpv than 54.5 percent enrichment be determined prior to restart from a refueling outage or any time boron is addep to the storage tank. This frequency is considered

. adequate because B'O is a stable isotope and enrichment changes can only occur when additional boron is added. In addition to ensure that the 10 boron added is enriched properly, station procedures will require that B enrichment be determined as part of the receipt inspection before release the matrial for use. Technical Specifications will also require that o{0 B

enrichment test results be known within 30 days of sampling the material in the Standby Liquid Control Storage Tank. The 30 day time period allows sufficient time to perform the enrichment test and receive the test results. It is considered adequate from a safety point, due to the station procedure requirement to determine enrichment as part of the receipt inspection before release of the material for use. The requirement to determine enrichment after the addition of boron to the storage tank functions as a backup check to the station procedures.

F Safety Evaluation

' . ,. No. AU G w . s - sja ' 1,7

=

Theresponsetim{0ofinjection the system is improved by this modification due to higher rate of 8 into the reactor.

• The relocation of the test button for SLC pump 207B and the addition of the pressure gauge will facilitate the system testing and does not affect the safety performance of the system.

~

0. Safety Function of Affected Systems /Comoonents .

The safety function of the SLC system is to provide a backup method, which is independent of the control rods, to maintain the reactor subcritical as the n'uclear system cools, in the event that not enough of the control rods can be inserted to counteract the positive reactivity affects of a colder moderator (Ref PNPS-FSAR, Rev. 6, Section 3.8.1). This modification has an impact on the safety analysis (Ref. PNPS-FSAR Section 3.8.4) and the Technical Specification Section 3.4 which need to be updated to include the NRC ATHS Rule (10CFR50.62) requirements.

E. Effects on Safety Function The enriched sodium pentaborate modification to the SLC will upgrade the system to the reactivity control capacity requirements of the NRC's ATHS Rule (10CFR50.62) and still provide the equivalent of 700 ppe of natural boron to maintain the original system shutdown requirement.

~

The low crystallization temperature (38'F corresponding to a a 9.22 weight

' percent concentration) of the enriched sodium pentaborate solution allows the reduction of the tank heater and heat tracing setpoint to 53*F. This

! temperature is 5'.F a~oove the low temperature alara setpoint of 48'F. The low solution crystallization temperature and the new tank heater and heat tracing setpoint will reduce the possibility'of reactor shutdown because of solution temperature requirements.

The addition of the pressure gauge and the relocation of the safety related test button for SLC pump 2078 will not have any adverse effects on the safety functions of the SLC system. Materi-Is for these changes will be procured, installed and tested in accordance with safety related requirements.

F. Analysis of Effects on Safety Functions As per GE analysis (see Attachment 2 to this safety evaluation), use of an 8.42 or greater percent concentration of enrighed sodium pentaborate (enriched to greater than 54.5 atos percent bio) will meet or exceed the NRC ATHS Rule 10CFR50.62 requirements of the SLCS at P11gris Nuclear Power Station. This analysis is based on an injection rate of 39 gallens per minute (Ref.1 & GE Calc. No. DRF C41-00095/2 L4, Sht.13A, SLCS Volume &

Concentration Chart). As each pump of the system has a minimum discharge capacity of 39 gallons per minute, the design is adequate to satisfy the NRC ATHS Rule requirements. The minimum concentration of 8.42 percent and aB10 enrichment greater than 54.5 percent provides a total margin of 136 percent beyond the amount needed to shutdown the reactor.

Safet Evaluation

• , * No. i 3: I

' Gm, c, - 6 /.* (51 The technical specification changes will provide adequate operational and surveillance requirements for the SLCS modification and will not reduce the margin of safety. i l

This modification does not involve an unreviewed safety question.

References

1. General Electric Company letter, Pilgrim ATMS SLC System Modification, R. G. Ferguson to R. N. Swanson, dated 2/2/87.

W e

l 1

i

~~

Safety Evaluation olG (

SAFETY EVALUATION PILGRIM WUCLEAR POWER 5TATION Rev. me. (0 )

A. APPROVAL

()

This proposed change does not involve a change in,the Technical Specifications.

@ 50.59(a)(2)

This proposed change, test or experiment

@ 50.7)(e) and is reportable under 10CFt50.5g(b).T (4 Comments: IWO/we XE 8/d-eV The safety evaluation basis and caeclesien is:

M Approved () het Approved '

A PrA ch/' q isE1,11ne ishne Leaserradte

1) ffAfg7 .

supportsas Discipline areap Lander / tete

s. atv1tw ApptovAt

) ([ Comments:la crn Pa.r4 2t a.pphes.6kt modvbh N h treeurem# e f bonn S/9 n hU h h sasa$roup Leader /nate

1. 4/t7' C. OSC REVIEW

() This preeIesed change involves an unreviewed safety geestien and a rogmest for authorization of this change must he filed with the Mrectorate of Licensing. NRC prior to implemsstation.

()

This proposed change does not involve an unreviewed safety geesttee.

Ott Chairuns gate ,

Ott Meeting number cc:

- Exhibit 3.07-A Rev. 3 Sheet 2 of 3 n-- ---- _ _ - -- --_ __-____ _____,_ _,_,-.___,,,-,,. _.____ ,_., _

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

• ai ny oeiuation No.: _kb' SAFETY EVALUATIOW WORK SHEET Rev. No. _O A.

Systee Structure Component Failure and Consequence Analyses.

System Structure Component Failure Modes Effects of Failure Comments

1. - SLES ,

Sics 5.id'm See, A fhc. heel 'Shech Y P 43t*F '

2. _SLCS 5 .Glih .h, See Anckai skaa.+

5 % +e w h 4Le Reac4&

by SLC

~

General Reference Material Review FSAR CALCULATI WS REGULATORY SECT 13 ENPS TECHNICAL SPECS. DESIGN SPECS PROCEDURES GU10ES STAND

. l.2 3. + / 4. 4- G E SPK. P f 7H A 149 T* 23 lo c F e s'o 4 2 3.1

,h 5. L

14. L

~

A PfEnaEG _

t.

i For the proposed hardware change, identify the failure modes that are I likely for the components consistent with FSAR assumptions. For each failure mode, show the consequences to the system, structures or related components. Especially show how the failure (s) affects the assigned safety basis (FSAR Text for each system) or plant safety functions FSAR Chapter 14 and Appendix 8).

• Prepared by b '

d- Seto 4!6!i7 NOTE:

It is a requirement to include this work sheet with the safety Evaluation.

Exhibit 3.07-C tev. 2 l

I

Safety Evaluation

, s l(o. 2 U Cw. c - Q 4 U ]

SAFETY EVALUATION HORK SHEET A. System / Structure /Comoonent Failure and Consecuence Analyses

1. System / Structure / Component: Standby Liquid Control System Failure mode: SLCS solution temperature less than 38'F.

Effects of Failure: Enriched sodium pentaborate solution crystallizes in the pump suction pipe rendering the system inoperative.

Comments: In order to ensure a solution temperature greater than 38'F;

1. Solution temperature will be determined daily.

2. Tank heater and heat tracing commence operation when the solution temperature reaches 53*F.

3. Solution temperature of 48'F will alars in control room, reactor must be placed into Cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of alarm.

t 2. System / Structure /Comoonent: Standby Liquid Control Systen Failure Modei Inability to shutdown the reactor.

Effects of Failure: Core damage, release of radioactive materials.

Comments: Analysis performed by General Electric to assure the modified Standby Liquid Control System will provide the equivalent of 700 ppa of natural boron to maintain the original shutdown requirement.

FSAR etvitu SHLtT Ref erences:

Safety Evaluation: Pbc B6 75 Rev. No.: O oate: f/4h7 Support a change List FSAR test, diagran, and indices af fected by this change and ccrresponding FSAL revision.

Af fected FSAR Revision to affected FSAR Section is shown on:

Section Preliminary Final

.5Ec7)oa)$ 2 0 33.B.4 33.1,5,&3.8.6 Attaehoent i RG. 3.t.1 M2413 Rer, f.7 FsC. 3 .* E M- 1F - 2.3 F#s. 3.%. 3 & J.%. (,

Attachment i j Attachment I Attachment l PSELININARY FSAR t[ VISION (to be completed at time of Safety Evaluation l

', preparation). .

Prepared by:bD.b/Date: 4[ 87 Reviewed by: -

to: V.i/E 7 l Approyed by: M ' /Date: 8 /f7  !

I FINAL FSAR REV1510N (Prepared following operational turnover of related systees structures of components for use at PNPS). (1) l Prepared by: -

/ Bate: Reviewed by: /Date:

i (1) Attach completed FSAR Change Request Form (Refer to NOP).

Exhibit 3.07-A Rev. t Sheet 3 of 3 4

e

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

PDC 86-75, Rev.

Safety Evaluation,p Rev. O 24h I Attachment 1 SheetIof !G ATIACH W2r 1 RECO9 ENDED FSAR OWGES

'the pages of the following sections, & figures of the FSAR that need to be updated due to StrS modification 86-75) have been marked with suggested updates and included in this attachment foc your review.

FShR Sections: 3.8.3, 3.8.4, 3.8.5, 3.8.6 (re== r--r  : -2. 0.0 : W FSAR Figures: 3.8-3, 3.8-6 .

The following drwings will be revised as part of the Plant Desir Change package (PDC 86-75) but are not included herein.

Dwg. I. D. FSAR FIGURE TIT 12 -

M249 3.8-1 PEID SEC System M-lF-213 3.8-2 Str Syst e Process Diagram 9

I k

PDC 86-75, Rev. 0 I ..

Safety Evaluat on. Rev. 0 Attachment 1 A /3 t Sheet 2 of IG -

PNPs-FSAR 3.8 STANDsY L1 QUID CONTROL SYSTEM .

3.s.1 safety Objective The safety objective of the stan&y Liquid Control System (Stes) is -

to provide a backup method, which is independent of the control rods, to maintain the reactor suberitical as the nuclear system cools in the event that not enough of-the control rods can be inser4ed to '

counteract the positive reactivity effects of a colder moderator.

3.8.2 Safety Design Basis

1. Backup capability for reactivity control shall 'bd provided, independent of normal reactivity control provisions in the

- nuclear reactor, to be able to shut down the reactor if the normal control ever becomes inoperative. ,

The' backup system shall beve the capacity for controlling the 2.

re, activity difference between the steady state rated operating  :

condition of the reactor with voids and the cold shotaewn *

• - ' ' condition ~, inclu61ng shutdown margin, to assure esmplete shutdown from the most reactive condition, at any time in the core life. ,

The time required for actuation and effectiveness of the backup h- 3.

control shall be consistent with the auclear reactivity rate of change predicted between rated operating and, cold shutdown

", i conditions. . A fast scram of the reactor or operational control ,

of fast reactivity transients is not specified to be accomplished .-

by this system.

4. Means shall be provided by which Uw ' functional performance i capability of the backup control system components can be verified periodically under conditions approaching actual use l

requirements, & substitute solution, rather than the actual - I

. neutron absorber solution, may be injected into the reactor to test.the operation of all components of the sedundant., Centrol .

systes. l

5. The neutron ' absorber shall be dispersed within the reactor core

( in sufficient quantity to provide a reasonable margin for imperfect =hhg or leakage.

6. The system shall be reliable to a degree consistent with its role as a special safety system the possibility of unintentional or .

accidental, shutdown of the reactor by this system shall be minimised.

- 3.3.3 De scription .

The' piping and instrumentation for the SLC5 is shown on Figure 3.5-1.

Figure 3.5 2 is a process diagram for the system. The SLC5 is manus 11y initiated from the main control room to pg a boron neutron l

absorber solution into the reactor if the operator believes the reacter cannot be shut down or kept shut down with the control rods.

~ ~ ~ = - = ~ ~ - - . ~ . - - - - . _ _ , , _ _ _ _ _ _ _ _ _

PDC 86-75, Rev. 0 Safety Evaluatipn, Rev. O l Attackwnt 1 %

pgp3 73g Sheet J of16 M/

e

. .g  ; ,

1. However, insertion of control rods is expected to always assure d prompt shutdown of the reactor should it be required. The boron nuclear fission

• y absorbs thermal neutront. and thereby terminates the j [ chain reaction in the uranim fuel. -

The SLCS is needed only in the improbable event that not enough control rods can be inserted in the reactor core to accomplish g g

[8v shutdown and cooldown in the at normal a steadymanner. The SLcs therefore ist8the Y rate within

• I sized only of to shut the reactor downthe shutdown Cooling systems, and kee

,l* capacity going critical again as it cools. '

W The boron solution tank, the test water tank, andthe two positive associated local

{

g b displacement pumps, the two explosive valves, controls are mounted in f valves and The liquid is piped into the reactor vessel and g prianary coctainment.the bottom of the core shroud so that it mises with 4) discharged near See Section 3.3, Reactor Jy the cooling water rising through the core. Mechanical Design, and Secpon'4.2, Internals M

a wg,tgeA ,,

--)g = *yi I f vesselg and appbrtenances Mechanical l Design.

The specified neutron absorber solution is aisodimpentabor' ate -

a - ri t - Mri: A 2:u

.% ,,* &# solution. git is prepared byQ _ _ e:ralized water. An air sparger is, W _x:- = b i; .J0 in demin To ' prevent system plugging, the .

provided in the tank for mixing. tank outlet is raised above the ' bottom

• Tef.8 '

( a strainer. f' "'" cere at all times when it is por,sible to make the reactor' W gal of k critical, the SLCS shall be able to deliver at least d -

D pentaborate solution er equivalent into the it 3.t1, 6~h percent 4 sodism~ N denW eifr's( %Mw,Cm%r .# *r 4 Me si: tor.

a contro tank and lling w th

- m..co a e in r tosta "I tj liqui 1 ( alized w a east low 1- 1 alarm olme. ~

are po and e

3 .

J '

demi design oncentra on -at lov level all _ Joy, ge sol ion is o W erflow vel .vol to {is E) g' e 4 be di 1edssts up 5$ e va ai r to low _ the=-sit stimet tamperatura_# -

__ -- 9 r-temperature of the specified solution isWF so the The n saturation j

equipment containing the solution is ^?installed in a room ' " "  :^^int which tha a air temperature is to be controlled ' .

tank, and a 2 -' -*'3. An electric immersion beater in therit$ S :t in i :4' M T contro11ere;M;a;;po  :

hl temperature n:~r n q is shoo used to elevate the g -- + - a t:- q , .and n.. : . assure that the boron dissolves when first added to or a sTET )' temperature the water. High or low tesperature, high or low liquid level, shorted heater causes an alarm in the control roen$.e a g aaLa h

's sited to inject the solution into gach positive displacentnt p ^^= .. W solution level in the the resetor in 50 to 125 min, _,

The pung and system design tank, at all reactor operating pressures.The two relief valves are set the to exceed pressure is 1,500 psig. pressure by a suf ficient margin to avoid valve reactor operating The relief valves are installed with the discharge flooded le akage .

3.8 2

Safety Evalua on Rev.0

' Attachment 1

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PDC 86-75, Rev. 0 ]

Safety Evaluation, Rev. 0

' - Attachment 1 Sheet S of $

Q Al3j PNPS-FSAR

. to prevent eva'pora tion and precipitation within the valve. To .

prevent bypass flov'from one pump in case of relief valve failure in' the Inne from the other pump, a check valve is installed downstresa -

of each relief valva line in each pump discharge line.

The two emplosive actuated injection valves provide high assurance of .

opening when needed and ensure that boron will not leak lato the reactor even when the pumps are being tested. The valves have a firing reliability la excess of 93.93 percent. Each emplosive valve is closed by a plug la the inlet chamber. The plug is circumscribed with a deep groove so the end will readily shear off when pushed. by the valve plunger. This action opens the inlet hole through the plug. The sheared end is pushed out of the way la the chamber, and is shaped so it vill not block the ports after release. -

The shearing plunger is actuated by an explosive charge with dual ignition pfleers, inserted in the side chamber of the valve. 4 Ignition circuit contimulty is monitored by a teletle current, and an

- alars occurs in the control room if either ti,rcutt opens. Indicator 11thts show' which channel primer circuit opened. To servita a valve. . -

af ter firing, a 6 in length pipe - (spool place) must be removed *

- tamediately upstreas of the valve to gain access to the shear plup. }

g The 51,CS is actuated by a' three' poiltion keylock switch .on the control roca console. This assures that switching from the 'off" .

w Settching to either side starts one

, i pos'ltion is a dellberate act. -

lajection pump, opens an explosive valve, and closes the Reactor j

~

Cleanup System isolation valves to prevent loss or diluttor of the boron. 3 A green light in the control race ladicates that power is 6vallable fr to the pump motor contactor, but that the contactor is open (pump not j j

l runntrig). A red light ladicates that the ,contactor is closed (pump running). 3 g

h ,

~ '

I Liquid flow is confirmed by a decrease la reactlvity, storage tant drawdown and pump runnleg Indication. A red light beside the keylock switch turns on when valve 1101-1 downstream of the analosive valves is open. If the pump 11thts or explosive valve light Lndicates that the 11guld may not be flowing, the operator can lamediately turn the ,)$

keylock sultch to the other side; this switch actuates the alternate equipment. Crosspiping and check valves assure a flow path through -

either pump and either esplosive valve. The chosen pump will start fj J even though Sta local switch at the pump is in the 'stop" position

.for test or malatenance. pump discharge pressure lodication is, also y

' provided la the control room. $

I Equipment drains and tank overflows are piped not to the Waste System but to separate containers (such as 55 gal druesb7 Ec h --- -M J l J '. 4 y . - ; . .M to prevent any trace of boron from #

laadvertently reaching the reactor. 1%cs N w %. ww.wa4. sal rene si m & q u i a n h g , er- em -

Instiveentation is provided locally at the standby liquid control tant and consists of solution temperature indication and control.

I 3 e5

1 PDC 86-75, Rev. 0 y Safety Evaluat}on, Rev. o ,

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1 Instrumentation and control logic is

-f- tank level, and heater status.

a presented on Figure 3.8-4.

3.8.4 Safety Evaluation The St.CS, although not necessary for plant operation. is required to '

j be operable when the reactor 15 in other than cold condition.

Jr Despite this precaution, the system is espected never to be needed for plant safety because of the. large number of independent control rods available two to shot To further assere this

• setsdown of thethe reactor. required components W acteste the

• availability,losive pumps and exp valves are provided in parallel rede my. ap g

The systen is designed to bring the reactor from rette power to a Y cold shutdown at any t1ee la core life. The reactivity campensation g b-d

• provided coolin will reduce reatter power from rated to aero and allowthe Nuc contro rods remainlag withdrawn la the rated power patters. {t4r .I tecludes the reactivity galas due to completa decay of the seson g J leventory. .It also 'lacludes the positive twattivity effects fron s hetMto cold,g eltalnatteg steam volds, changing water denstty from reduced Doppler effect la wrantua, reduction of neutron leakage fromb 2..

i holling to cold, and decreastnp control rod worth as the moderator 5U cools. The. Speelfled alaisua final concentration of boros to the4 4  :-+

@d l .c 4 ,.s -

( *gjassures g (reactor plus a margin of -4.05 A k for calculational uncertalattes ang core margin.provider a reactivity x< lh,f Uwo e-s a substantial shutdown

' Pee 7en h 3J f lThe reactor, to specified alnlaus aveshutdown ecified ge concentratt,on of natural margin af ter aperation of ytoren J la theT 4 g

the St.C5, is pen.1

^^

, . . . . c 5. The alnlema quantity 5f j T esodlus pentat> orate H lajected into the reactor is calculated , _

( tased on the required an recirculation 5

) ppa average concentration la the reactor 9 h result is increased by 25 percent to allow for taperfect at Ing, 1eakage, and volume la other small piping connected to the reactor.

Cooldown of the nuclear lystem will take several hours as a stalaus, -

to remove the thermal energy stored in the reactor, coollag water, and associated equipment and to remove most of the radlesctive decay

. beat. The controlled 11 alt for the reactor vessel cooldoen is 100'F/hr, and normal operating temperature is about 550*f. Usually.

shutting down the plant with the main condenser and wartons shutdovr.

coollag systent will takc 10 to 24 hr before the reactor vessel is

' opened, and auch lobger to reach room temperature 00'F) which is the  !

condition of maalaus reactivity and therefore, the condition which requires the maataua boron concentration.

The injection rate is .lletted to the range of 39 to 79 galletn.

The lower rate assures that the boron gets into the reactor in about 1 1/2 hr, consleerably eulcker than the cooldown rate. The upper a*m 4 " n. M Ft td be tw2uA b% , Gj M

g %

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- l limit injection rate assures that there is suffletent etsing so the -

boron does not rectreulate through the core in uneven concentrations '

which could possibly cause the nuclear power to rise and f611

. cyc1;cally. -

The systen piping The SLCS is designed as a Class I seisale system.

and equipment are designed. Installed,Nonprocess and testedequipment in accordance such as with USAS 831.1.0 Section I and Appendla A.

the ~ test tank is designed as Class II. -

if The $LCS 13 required to be operable in the event of a station power failure, so the pumps. valves, and controls are powered from the The . pumps j

a standby ac power supply la the absence of normal power. separate buses and and velves are powered and :ontrolled from *'

"- circuits so that a single failure will lights mot prevent.systea. operation.

are powered from the essential instruments and The - -

120 V ac instrument power supply. '

r 1965

~11ae SLCS angl pumps have sufficient pressure margin. wp to the systeminjection reitef valve setting of 1.400 psig, to assure solutt ..

the reactor above the normal pressure 5.0.~ psig in of .about

.theThe nuclear systes rettef ande safe .,

bottom of the reactor.

' begin to retteve pressure'above 4 bout 1.100 psig;'.therefo -

! f posittve i.etealed % de,( .Ato provlielconcentrationl .a.

s of anhand boron in a 17 The systealts the reactor Aof in 700 ppa The shutdown marata from this concentra9 ton J 1 *I pilgria's Supplemental Reload License Submittal in can be found The analysts and models for the reload core are

! Appendt Q. ,

described in the GE Standard Appilcation for Reactor Fuel."'

! *g -

.e*r 3,s.5 Inspection and Testing

)j l avoid inadvertently injecting boron into theUEI -

ng reactor. /sy l

Yet ra N lts /

t Mir[at 1 W ch n the valves to and from the solut'on tank closed and theope

)g ~three valves (two locked closed)la the test tank can be rectrculate J

• f the deatneralized water turntng on either pump _ locally. g m

,m..- _ _ _ _

gepgggy ~

unctional testing of the la.jection portion of the ' system is accon tshed by closing the locked open valve fron'the solution tank.

~ opening the locked glosed valve free the test tank, and actuating the keylock witch in the control room to either the A or 5 circuit.

This circutt.

starts the pump and blows open the injection valve system is ope.*ating.

= * *

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PDC B6-75, Rev. 0 Safety Evaluat. n, Rev. 0

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• Attachment 1

• PNPS-FSAR Sheet 6of16 2/3/

closing a local locked open valve to the reactor in the 8f leakage through the injection valves can be detected at contatnment, 3 a test connection la the line between the contatnment tsola valves. valve is closed for tests, or open and ready for A the local Leakage from the reactor through the first thcck valve sj

. operation.1 can be detected by opening the same test connection whenever the ,

reactor is pressurized. j After the functional tests, the injection valves and emplosive cj 5, charges must be replac2d and all valves returned to te,etr F 4 normal p

posttions, as f adicated on Figure 3.81. . '

I The test tank contains waterdeatnerallred water for d *)6about thre

.lleatnerallred from the sakeup or cendensate

. pump operation. storage system is available at 30 gal / min for raf tllts the system. , ,

i f,hould the boron solution ever be injected into theL.,-

~ reactor, eith

,, Latentionally or inadvertently, then, af ter making certata the+j ta i normal . reactivity controis will keep the reactor subcritical, for

.. 3 y-on' is removed from the Reactor Coolant r

} f. lystem there is practically no effect on r'enctor operations them the .

concentration has been reduced below approntmately 50 pga.. "#

'O The concentration of the sodlus pentandate'in the solutton* tank is ,

deteralned by cheetcal analysis periodically. "th s Mmad at A The gas pressure in 'the two accumulators is measured periodica A pressure gage and portable altrogen supply are detect leakage.

required to test and recharge the accumulators. M N

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- s.7 Current Operational Nuclear Safety Requirements current 1taiting condition for operation, survelliance -

The in tiu Technical requirements , and their bases are conteined 5pecifications referenced la Appendia G.

3.8.8 References NEDE-24011-P-A, General fluctrical Standard Application for

1. .

Reactor Fuel, app 11 cable revlr,lon.

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DISIRIBUTICH:

M1aB Wr APR 181985 DES WP (2) anxher 1crocker DEhtn MEMORANDUM FOR: Dennis Crvtchfield, Assistant Director for Safety Assessment Division of Licensing FROM: Don H. Beckham, Acting Deputy Director Division of Human Factors Safety

SUBJECT:

SAFETY EVALUATION REPORT FOR PILGRIM STATION, GENERIC LETTER 83-28. ITEM 1.1 (POST-TRIP REVIEW)(TACNO.52786)

By letter dated November 7,1983, the licensee of Pilgria Station responded to Generic Letter 83-28 with regard to required actions based on generir. implications of Salem ATWS events. The enclosed SER and SALP were prepared by the Licensee Qualifications Branch based upon TER input from its contractor, Science Applications International Corporation (SA!), after having reviewed the applicable portions of the licensee's response to Generic Letter 83-28.

We find that the licensee's response does nct meet the guidelines developed for post-trip review in the following areas:

1. The criteria for detennining the need for independent assessment of the event.

2. The method and criteria for comparing the event with expected plant performance.

3. A systeestic safety assessment program to assess unscheduled reactor trips.

We reccesnend that the licensee take necessary action to correct the deficiencies noted above, which are discussed in greater detail in Section III of the SER.

Acceptable responses to the above noted deficiencies are required before we can complete our review of the licensee's Post-Trip Review Program and Procedure for Pilgrim Station. We will review these responses when received and report our findings in a supplement to this SER.

Information in this record was deleted

[:0 ^ g e5ggg (.13:cordance T with the Freedom of Information y ..:!,

m.2cxem4tions-q49 DIE 6/MDO FDR CRUIC11-PIIIRIM //p/ h '

/

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a ')b.. .....EE........ .......og... ..................... .................... ................

-> .am.A.... .. . u ...... ... m e r....... = = = . ..................... ..................... . . . . . . . . . . . . . . . .

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. . . . .d . . . .a. s. v. . . . . #. . . . /.8.s.

.v. . /.7. . . . /.8.s. ..................... .................... ................

.-...  %, ,,,-c.

7 - _

l I

l l

D. Crutchfield APR 18 lb This review has been conducted by D. Shum (x24906) and SAI. There are no known dissenting professional opinions on this matter.

Don /9 Acting Deputy Director H Beckham, Divis on of Human Factors Safety

Enclosures:

1. Safety Evaluation -

2. SALP Evaluation ec: A. Bournia J. Zwolinski G. Holahan .

i l

l l

\

Enclosure 1 SAFETY EVALUATION REPORT FOR GENERIC LETTER 83-28, ITEM 1.1 - POST-TRIP REVIEW (PROGRAM DESCRIPTION AND PROCEDURE)

PILGRIM STATION DOCKET NO.: 50-293

!. INTRODUCTION On February 25, 1983, both of the scram circuit breakers at Unit 1 of the Salem Nuclear Power Plant f ailed to open upon an automatic reactor trip signal from the reactor protection system. This incident occurred during the plant start-up and the reactor was tripped manually by the operator about 30 seconds after the initiation of the automatic trip signal. The failure of the circuit breakers has been determined to be related to the sticking of the under voltage trip attachment. Prior to this incident, on February 22, 1983, at Unit 1 of the Salem Nuclear Power Plent, an automatic trip signal was generated based on steam generator low-low level during plant start-up. In this case, the reactor was tripped manually by the operator almost coincidentally with the automatic trip. Following these incidents, on February 28, 1983, the NRC Executive Director for Operations (E00), directed the staff to investigate and report on the generic implications of these occurrences at Unit 1 of the Salem Nuclear Power Plant. The results of the staff's inquiry into the generic implications of the Salem unit incidents are reported in NUREG-1000, "Generic Implications of ATWS Events at the Salem Nuclear Power Plant." As a result of this investigation, the Comission (NRC) requested (by Generic Letter 83-28 dated July 8,1983) all licensees of operating reactors, applicants for an operating license, and holders of construction permits to respond to certain generic concerns. These concerns '

are categorized into four areas: (1) Post-Trip Review (2) Equipment Classification and Vendor Interface, (3) Post-Maintenance Testing, and l (4) Reactor Trip System Reliability Improvements. l The first action item, Post-Trip Review, consists of Action Item 1.1,  ;

"Program Description and Procedure" and Action Iter 1.2. "Data and a r7 mi qar l >1> 4r 9 @D

,y ,

V 2

Infonnation Capability." This safety evaluation report (SER) addresses Action item 1.1 only.

II. REYlEW GUIDELINES The following review guidelines were developed after initial evaluation of various utility responses to Item 1.1 of Generic Letter 83-28, and incorporate the best features of these submittals. As such, these review guidelines in effect represent a "good practices" approach to post-trip review. We have reviewed the licensee's response to item 1.1 against these guidelines: .

A. The licensee or applicant should have systematic safety assessment procedures established that will ensure that the following restart criteria are met before restart is authorized.

• The post-trip review team has determined the root cause and sequence of events resulting in the plant trip.

l

• Near term corrective actions have been taken to remedy the cause of l the trip.

l The post-trip review team has performed an analysis and determined that the major safety systems responded to the event within specified limits of the primary systen parameters.

l The post-trip review has not resulted in the discovery of a j potential safety concern (e.g., the root cause of the event occurs with a frequency significantly larger than expected).

• If any of the above restart criteria are not inet, ther. an independent assessment of the event is perfonned by the Plamt Operations Review Comenittee (PORC), or another designated group with similar authority and experience.

l

3 B. The responsibilities and authorities of the persennel who will perfonn the review and analysis should be well defined.

The post-trip review team leader should be a member of plant management at the shif t supervisor level or above and should hold or should have held an SRO license on the plant. The team leader should be charged with overall responsibility for directing the post-trip review, including data gathering and data assessment and he/she should have the necessary authority to obtain all personnel and data needed for the post. trip review.

• A second person on the review team should be an STA or should hold a relevant engineering degree with special transient analysis training.

• The team leader and the STA (Engineer) should be responsible to concur on a decision /reconnendation to restart the plant. A nonconcurrence from either of these persons should be sufficient to prevent restart until the trip has been reviewed by the PORC or equivalent organization.

C. The licensee or applicant should indicate that the plant response to the.

trip event will be evaluated and a determination made as to whether the plant response was within acceptable limits. The evaluation should include:

• A verification of the proper operation of plant systems and equipant by comparison of the pertinent data obtained during the post . rip review to the applicable data provided in the FSAR.

• An analysis of the sequence of events to verify the proper functioning of safety related and other imortant equipment. Where possible, comparisons with previous similar events should be made.

4 D. The licensee or applicant should have procedures to ensure that all physical evidence necessary for an independent assessment is preserved.

E. Each licensee or applicant should provide in its submittal, copies of the plant procedures which contain the infomation required in Items A through D. As a minimum, these should include the followinc:

• The criteria for detemining the acceptability of restart The qualifications, responsibilities and authorities of key personnel involved in the post-trip review process

• The methods ard criteria for determining whether the plant variables and system responses were within the limits as described in the FSAR

• The criteria for determining the need for an independer.t review.

l 111. EVALUATION AND CONCLUSION i

l By letter dated November 7,1983, the licensee of Pilgrim Statfori provided infomation regarding its Post. Trip Review Program and Procedures. We have evaluated the licensee's program and procedures against the review guidelines developed as described in Section II. A brief description of the licensee's response and the staff's evaluation of the response against each of the review guidelines is provided below:

A. The licensee has established the criteria for determining the acceptability of restart. Based on our review, we find that the licensee's criteria conform to the guidelines as described in the above Section !!.A and, therefore, are acceptable.

l l

I l

B. The qualifications, responsibilities and authorities of the personnel who will perfonn the review and analysis have been clearly described.

We have reviewed the licensee's chain of command for responsibility for post-trip review and evaluation, and find it acceptable.

C. The licensee has not addressed the methods and criteria for comparing the event information with known or expected plant behavior. We recomend that the ' pertinent data obtained during the post-trip review be compared to the applicable data provided in the FSAR. Where possible, comparisons with previous similar events should be made.

D. The licensee has indicated that if the cause of the trip is unknown, an independent assessment conducted by the Operation Review Connittee is required for the event. We find that this action to be taken by the licensee is not sufficient to ensure safe plant operation. We reconnend that if any of the restart criteria are not met an independent assessment of the event should be perfonned. The licensee has established procedures to ensure that all physical evidence necessary for an independent assessment is preserved.

,E. The licensee has not provided for our review a systematic safety assessment program to assess unscheduled reactor trips. We reconnend that the licensee develop a systematic safety assessment program to handle unscheduled reactor trips.

l Acceptable responses to the above noted deficiencies are required before we can con 19 ete our review of the licensee's Post-Trip Review Program and Procedures for Pilgrim Station. We will review these responses when received and report our findings in a supplement to this SER.

i l

l l

Enclosure 2 SALP EVALUATION

'I PILGRIM STATION i DOCKET NO.: 50-293 i GENERIC LETTER 83-28 ITEM 1.1 POST TRIP REylEW l

I i

\

t

! I i

?

I

00T 101986 Docket No.: 50-293 MEMORANDbMFOR: P. Leech, Project Manage:', Project Directorate Al Division of BWR Licensing FROM: Gus C. Lainas, Assistant Director Division of BWR Licensing

SUBJECT:

SAFETY EVALUATION REPORT FOR GENERIC LETTER 83-28, ITEM 2.1 (PART 1) (EQUIPMENT CLASSIFICATION) FOR PILGRIM NUCLEAR POWER STATION (SRP SECTION 7.2,17.2)

Plant Name: Pilgrim Nuclear Power Station Utility: Boston Edison Company Docket No.: 50-293 TAC Ho.: 52867 Licensing Status: OR Resp. Directorate: PAD #1/DPLA Project Manager: P. Leech Review Branch: PAE!/DPA Review Status: . Incomplete The licensee was required by Generic Letter 83-28 Item 2.1 (Part 1) to confinn that all components whose functioning is required to trip the reactor are identified as safety-related on all plant documentation and in information handling systems that are used to control all activities perfonned on this safety-related equipment. The licensee has responded and our review of the responses as documented in the enclosed contractor's report (EG&G-NTA-7188) finds the licensee's responses to Generic letter 83-28 Item 2.1 (Part 1) to be acceptable. The enclosed SER documents our concurrence with the contractor's findings and also finds the licensee's responses for this item to be acceptable. We therefore consider Part 1 of item 2.1 to be closed by this action. SALP input for the review of Item 2.1 (Part 1) of Generic letter 83-28 is enclosed.

Generic Letter 83-28 Item 2.1 and its associated TAC number remain incomplete because the Yendor Interface portion (Part 2) of this item has not been resolved.

g!,-y e :d By:

G.C. Laires

Enclosures:

Gus C. Lainas, Assistant Director As stated Division of BWR Licensing cc: R. W. Houston Distribution:

T. Novak 6 it ffle Wof 50-293 C. E. Rossi PAE! Rdg.

J. Zwolinski D. Lasher (PF)(2) c M. Srinivasan J. E. Knightinformation in this record was defelpd -

' Conta t: .C ainasin accordance with the Freedom of Inform D. Lasher, EICSB/DPA Pilgrim S/F st,A exemptions #~

X27200 F0!A M- d W PAEI SL/PAE! BC/PAEI BC/BWEl AD DLa :ct JEKnightpl FRosa MSrinivasan GC a s y v 10/J/86 10/ P /86 10/f/86 10/le/86 10/0B6

/

0FFICIAL RECORD COPY d/ ' N) b$ , h

SAFETY EVALUATION REPORT GENERIC LtITER 53-25. ITEM 2.1 (PART 1)

EQUIPMENT CLAS5IFICATION (RT5 COMPONENTS)

PIL6 RIM NUCLEAR POWER STATION DOCKET NOS. 50-293 INTRODUCTION AND SUM 4ARY On February 25, 1983, both of the scram circuit breakers at Unit 1 of the Salem Nuclear Power Plant failed to open upon an automatic reactor trip signal frts the reactor protection system. This incident was tenninated manually by the operator about 30 seconds after the initiation of the automatic trip sigval.

The failure of the circuit breakers was determined to be related to the sticking of the undervoltage trip attachment. Prior to this incident, on Febreary 22, 1983, at Unit 1 of the Salem Nuclear Power Plant, an auttmatic trip signal was generated based on steam generator low-low level during plant start-up. In this case, the reactor was tripped manually by the operator almost coin-cidentally with the automatic trip.

Following these incidents, on February 28, 1983, the NRC Executive Director for Operations (E00), directed the staff to investigate and report on the geveric implic&tions of these occurrences at Unit 1 of the Salem Nuclear Power Plant. .

The results of the staff's inquiry into the generic implications of the Salem l

unit incidents are reported in h0 REG-1000. "Generic Implications of the ATW5 l Events at the Salem Nuclear Power Plant." As a result of this investigation, the Counission (NRC) requested (by Generic Letter 83-28 dated July 8,1983I )

l all licensees of operating reactors, applicants for an operating license, and holders of construction pennits to respond to generic issues raised by the analyses of these two ATWS events.

{

q: ir w ' 6 W

l i

l This report is an evaluation of the response submitted by Boston Edison Company, the licensee for the Pilgrim Nuclear Power Station, for Item 2.1 (Part 1) of Generic Letter 83-28. The actual documents reviewed as part of this evaluation are listed in the references at the end of the report.

Item 2.1 (Part 1) requires the licensee to confirm that all Reactor Trip System components are identified, classified and treated as safety-related as indicated in the following statement:

Licensees and applicants shall confinn that all components whose functioning _is required to trip the reactor are identified as safety-related on documents, procedures, and information handling system used in the plant to control safety-related activities, in-cluding maintenance, work orders, and parts replacement.

EVALUATION The licensee for the Pilgrim Nuclear Power Station responded to the requirements of Item 2.1 (Part 1) with submittals dated November 7,19832and June 28, 19853 .

The licensee stated in these submittals that all components that are required to perfonn the reactor trip function were reviewed to vedfy that these components are classified as safety-related equipment in the plant "Q-list." The licensee further confirmed that documents used to control activities associated with this equipment are identified as "Q' which designates the use of safety-related procedures.

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CONCLUSION Based on our review of these responses, we find the licensee's statements confirm that a program exists for identifying, classifying and treating components that are required for performance of the reactor trip function as safety related. This program meets the requirements of item 2.1 (Part 1) of the Generic letter 83-28, and is therefore acceptable.

REFERENCES

1. NRC Letter D. G. Eisenhut to all Licensees of Operating Reactors, Applicants for Operating License, and Holders of Construction Perwits, "Required Actions Based on Generic Implications of Salem ATWS Events (Generic letter 83-28)," July 8,1983.

2. Letter, W. D. Harrington, Boston Edison Co., to D. B. Vassallo, NRC, Noventer 7,1983.

3. Letter, W. D. Harrington, Boston Edison Co., to D. B. Vassallo, NRC, June 28, 1985.

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CONFORMANCE TO GENERIC LETTER 83-28 ITEM 2.1 (PART 1) EQUIPMENT CLASSIFICATION (RTS COMPONINTS)

SELECTED GENERAL ELECTRIC BOILING WATER REACTOR PLANTS HOPE CREEK PEACH BOTTOM 2 AND 3 PERRY 1 AND 2 PIL6 RIM 1 R. HAROLOSEN Published September 1986 EG46 Idaho, Inc.

Idaho Falls, Idaho 83415 e

Prepared for the U.S. Nuclear Regulatory Commission Washington, D. C. 20555 Under 00E Contract No. OE-AC07-761001570 FIN Nos, 06001 and 06002 ge_-

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• 1 ABSTRACT

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i This EG&G Idaho, Inc. report provides a review of the submittals from

, selected operating and applicant 8 oiling Water Reactor (BWR) plants for conformance to Generic Letter 83-28. Ites 2.1 (Part 1). The following

. plants are included in this review, l Plant Name Docket Number TAC Number Hope Creek 50-354 OL )

Peach Bottom 2 50-277 52865 Peach Bottom 3 50-278 52866 Perry 1 50-440 61705 Perry 2 50-441 OL Pilgrim 1 50-293 52857 FOREWORD This report is supplied as part of the program for evaluating licensee / applicant conformance to Generic letter 83-28, ' Required Actions Based on Generic Implications of Salem ATWS Events.' This work is being conducted for the U. S. Nuclear Regulatory Commission, Office of Nuclear Regulation, Division of PWR Licensing-A, by the EG66 Idaho, Inc.

The U. S. Nuclear Regulatory Commission funded this work under the authorization 86R 20-19-10-11-3 and 20-19-40-41-3, FIN tos. 06001 and 06002.

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CONTENTS .

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MBSTRACT .............................................................. 11 F 6R E W O R D . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11  ;

1. INTRODUCTION AND

SUMMARY

......................................... 1

, 2. PLANT RESPONSE EVALUATIONS ....................................... 3 2.1 Hope Creek ................................................. 3 2.2 C o n c l u s i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2.3 P e a c h B o t t on 2 a n d 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . $

2.4 C o n c l u s i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 2.5 P e r r y 1 a nd 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 2.6 C o n c l u s i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 2.7 Pilgrim 1 .................................................. 7 2.8 Conclusion ................................................. 7

3. GE N E R I C R E F E R E NC E S . . . . .' . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8. . . . . . .

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1. INTRODUCTION AND SUMLARY On February 25, 1983, both of the scram circuit breakers at Unit 1 of the Salem Nuclear Power Plant f ailed to open upon an automatic reactor trip signal from the reactor protection system. This incident was terminated manually by the operator about 30 seconds af ter the initiation of the

, aiutomatic trip signal. The f ailure of the circuit breakers was determined to be related to the sticking of the undervoltage trip attachment. Prior to this incident, on February 22, 1983, an automatic trip signal was generated at Unit 1 of the Salem Nuclear Power Plant based on steam generator low-low level during plant startup. In this case, the reactor was tripped manually by the operator almost coincidentally with the automatic trip.

Following these incidents, on February 28, 1983, the NRC Executive 31 rector for Operations (E00), directed the staff to investigate and report om the generic impitcations of the occurrences at Unit 1 of the Salen nuclear Power Plant. -The results of the staf f's inquiry into the generic implications of the Salem Unit 1 incidents are reported in NUREG-1000

• Generic Implications of the ATWS fvents at the Sales Nuclear Power Plant.b As a result of this investigation, the Commission (NRC) requested (by Generic Letter 83-28, dateo July 8,1983 ) all licensees of eperating reactors, applicants for an operating license, and holders of construction permits to respond to generic issues raised by the analyses of these two ATWS events.

This report is an evaluation of the responses submitted from a selected' group of Boiling Water Reactors (BWRs) for Ites 2.1 (Part 1) of Generic Letter 83-28.

The results of the review of four individual plant responsas are combined and reported on in this document to enhance review ef ficiency.

The specific plants reviewed in this report were selected based on the

e convenience of review. The actual documents which were reviewed for each evaluation are insted at the end of each plant evalutten. The generic documents referenced in this report are listed at the and of the report,

, Part 1 of Ites 2.1 of Generic Letter 83-28 requires the licensee or applicant to confirm that all reactor trip system components are

. identified, classified, and treated as safety-related, as indicated in the f ollowing statement:

Licensees and appitcants shall confirm that all components whose functioning is required to trip the reactor are identified as safety-related on documents, procedures, and infor1 nation handling systems used in the plant to control safety-related activities, including maintenance, work orders, and parts replacement.

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2. PLANT REJPONSE EVALUATIONS 2.1 Hooe Creek 50-354 (OLI

. The applicant for Hope Creek (Public Service Electric and Gas Company) provided responses to the requirements of Ites 2.1 (Part 1) of Generic Letter 83-28 in submittals dated March 30, 1984, December 17, 1944 and May 21, 1985. In the first submittal the applicant described their plan to develop a Master Equipment List (MEL) which would identify the components required to initiate reactor trip and designate these components as safety-related. The MEL imposes quality assurance requirements for the safety-related components and is the controlling document for safety-related activities. The applicant stated intentions to be in compliance with Iten 2.1 (Part 1) prior to September 1984.

The secor,d submittal reviewed progress to December 17, 1984 and outlined a revised program which would meet the requirements of Ites 2.1 (Part 1) prior to March 1985., The applicant confirmed la their May 21,

1985 submittal that review of the reactor trip system had been completed and that reactor trip system components were verified to be classified safety-related on appropriate design documents, however, the MEL had not been completed for all components of the reactor trip system. The applicant stated that this ef fort would be completed by September 30, 1945.

2.2 Conclusion

Based on a review of the applicant's subelttals, we find that the l applicant's responses confirm that components required to trip the reactor

! have been designated safety-related and that the MEL is used to control all activities relating to safety-related components. ide, therefore, find that the applicant's responses meet the requirements of Ites 2.1 (Part 1) of Generic Letter 83-28, and are acceptable.

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REFERENCES

1. Letter, R.L. Mitti, Public Service Electric and Gas Co., to A.

Schwacer. NRC, March 30, 1984,

2. Letter, R.L. Mitt 1, Public Service Electric and Gas Co., to A.

Sch w ncer, NRC, Decefaber 17, 1984.

3. Letter, R.L. Mitti, Pubile Service Electric and Gas Co., to W. Butler'

. NRC, May 21, 1985.

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2.3 Peach Botton 2 00-277 TAC N0. 52665 Peach Bottos 3 50-278 TAC NO. 52866 The licensee fcr Peach bottos 2 and 3 (Philadelphia Electric Co.)

provided. responses to the requirements of Item 2,1 (Part 1) of Generic Letter 83-28 in submittals dated November 4,1983 April 23,1984 and May 29, 1985.

The responses state that all systems that contribute to the reactor trip function have been identified as safety-related in the current 'Q' list and that all components of safety-related systems are safety-related unless specifically excluded by safety evaluation. The 'Q' 11st is used to identify the applicable codes, standards and procedures to be used for activities relating to the safety-related components.

Each item or service to be procured is reviewed to determine tf it is safety-related. The review is performed by a congntrant member of the plan staf f or the Engineering and Research Departaent.

2.4 Conclusion l

Item 2.1 (Part 1) requires licensees to confirm that all caq>enents whose functioning is required to trip the reactor are identified as safety-related on documents, procedures, and inforestion handling systems used in the plant to control safety-related activities, including maintenance, work orders, and parts replacement.

Based on the licensee's submittal we find that the list of camponents l

l required to trip the reactor is incomplete. We also find that the licensee's program does not *dentify safety-related components en relevant plant documents. The response, therefore, does not meet the regstrements of Item 2.1 (Part 1) of Generic Letter 83-28 and is unacceptable.

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REFERENCES

1. Letter, S.L. Daltroff, Philadelphia Electric Co., to D.G. Eisenhut, NRC, November 4, 1983.

. 2. Letter, S.L. Daltroff, Philadelphia Electric Co., to 0.G. Eisenhut, NRC, April 23, 1984.

3. Letter, S.L. Daltroff, Philadelphia Electric Co., to J.F. Stolz, NRC May 29,1985.

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2.5 Perry 150-440 (Oli and _ Perry 2 50-44i (0L1 The applicant for Perry 1 and 2 (Cleveland Electric Illuminating Co.)

provided responses to the requirements of Item 2.1 (Part 1) of Generic

. Letter 83-28 in submittals dated April 6,1984 and August 28, 1985. The applicant reported in the first submittal that the 'Q'-list for the plants was undergoing review to verify the correct classification of safety-related components. The 'Q'-list is to be used to determine classification for maintenance, work orders and procurement activities.

The second submittal reported that the 'Q'-list evaluation had been completed and that all nimbered components from the 5 systems that contribute to the reactor trip function had been reviewed and classified as safety-related or nonsafety-related. The 'Q'-list is the saf ety-related subset of the Perry Equipment Master Files System (PEMS) used to determine the classification for woti arders, maintenance and parts procurement.

2.6 Conclusion Based on the review of the applicant's submittals, we find that the applicant has verified that the components necessary to perform reactor trip are classified as safety-related and that this classification program imposes safety-related procedures on work orders, maintenance, and procurement activities. We, therefore, find that the applicant's response meet the requirements of Ites 2.1 (Part 1) of Generic Letter 83-28 and are acceptable.

REFERENCES

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1. Letter, M.R. Edelman, Cleveland Electric Illuminating Co. to D.G. I Eisenhut, NRC, April 6,1984.

2. Letter, M.R. Edelman, Cleveland Electric Illuminating Co., to 8.1.

Youngblood, August 28, 1985.

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2.7 Pilgrim 1. 50-293. TAC No. 52867 The Itcensee for Pilgrim 1 (Boston Edison Co.) provided responses to the requirements of Item 2.1 (Part 1) of Generic Letter 83-28 in submittals

, dated November 7,1983 and June 28, 1985. In the submittals the licensee confirmed that the components required to function for reactor trip are identified in the plant 'Q'-list and are controlled at a quality level which reflects the safety-related functions. Documents (Purchase Orders, Maintenance Requests) usea to control activities associated with the

'Q' listed equipment are identified as 'Q* which designates the use of safety-related procedures.

l 2.8 Conclusion Based on the review of the licensee's submittals, we find that the licensee has verified that the components necessary to perform reactor trip are classified as safety-related and that the classification program

! taposes safety-related procedures on maintenance and procurement activities relating to the components. We, therefore, find that the licensee's l

l response meet the requirements of Items 2.1 (Part 1) of Generic l Letter 83-23 and are acceptable.

REFERENCES l

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1. Letter, W.D. Harrington, Boston Edison Co., to 0.8. Vassallo, NRC, November 7, 1983.

2. Letter, W.D. Harrington, Boston Edison Co., to 0.8. Vassallo, NRC, June 28, 1985.

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3. GENERIC REFERENCES 1.

Generic Imol a tions of ATWS Events at the Salem Nuclear Power Plant.

NUREG-1000, Wune 1, Apr11 1983; Volume 2, July 1983.

2. NRC Letter. I.L. Eisenhut to all Licensees of Operating Reactors, Applicants fr Dperating License, and Holders of Construction Permits,

' Required Ac tens Based on Generic Impilcations of Salem ATWS Events (Generic Let:er 83-28),' July 8,1983.

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SEP 031985 L. crocker D. Shum Gus C. Lainas, Assistant Director MEMORANDUM FOR:

for Operating Reactors 5k7-Division of Licensing FROM: Dennis L. Zie: nann, Acting Deputy Director Division of Human Factors iafety

SUBJECT:

REVISED SAFETY EVALUATION REPORT FOR PILGRI.M STATION, GENERIC LETTER 83-28, ITEM 1.1 (POST-TRIP REVIEW)

(TAC NO. 52786)

Reference:

Memorandum from D. Beckham to D. Crutchfield "Safety Evaluation Report for Pilgrim Station, Generic Letter 83-28. Item 1.1 (Post-Trip Review)," dated April 18, 1985.

By the above referenced memorandum, we forwarded the LQB SER with regard to Pilgrim Station. In our SER, we stated that the licensee's response did not meet the guidelines developed for post-trip review in the following areas:

1. The criteria for determining the need for independent assessment of the event.

2. The method and criteria for comparing the event with expected plant performance, j

3. A systematic safety assessment program to assess unscheduled reactor trips.

By 4tter dated August 13, 1985, the licensee provided responses for the

.soove cited open issues. We have reviewed the licensee's responses and find that the above open issues have been resolved. Thus, we conclude that the Post-Trip Review Program and Procedures for Pilgrim Station are acceptable.

Our revised SER and SALP evaluation are enclosed.

nformation in this record was deleted

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d Gus C. Lainas This review has been conducted by D. Shum (x24906). There are no known dissenting professional opinions on this matter.

% LaelsEsasent Deads L r-,

Dennis L. Ziemann, Acting Deputy Director Division of Human Factors Safety

Enclosures:

1. Safety Evaluation

2. SALP Evaluation ec: D. Vassallo A. Bournia G. Holahan P. Leech DW/DHS10/ REVISED MEMO FOR LAINAS DILGR H A N8 .

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s Enclosure 1 SAFETY EVALUATION REPORT FOR GENERIC LETTER 83-28, ITEM 1.1 - POST-TRIP REViry

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(PROGRAM DESCRIPTION AND PROCEDURE)

.. PIL6 RIM STATION DOCKET N05.: 50-293

1. INTRODUCTION On February 25, 1983, both of the scram circuit breakers at Unit 1 of the Salem Nuclear Power Plant (SNPP) failed to open upon an automatic reactor trip signal frv i the reactor protection system. This incident occurred during the plat start-up and the reactor was tripped manually by the operator about 30 seconds after the initiation of the automatic trip signal.

'The failure of the circuit breakers has been determined to be related to the sticking of the under voltage trip attachment. On February 22, 1983, during start-up of SMPP, Unit 1, an automatic trip signal occurred as the result of steam generator low-low level. In this case, the reactor was tripped nanually by the operator almost coincidentally with the automatic trip.

Following these incidents, on February 28, 1983, the NRC Executive Director for Operations (EDO) directed the staff to investigate and report on the generic implications of these occurrences. The results of the staff's inquiry into these incidents are reported in NUREG-1000, "Generic

! Implications of ATWS Events at the Salem Nuclear Power Plant." As a result of this investigation,' the Comission requested (by Generic Letter 83-28 e dated July 8,1983) all licensees of operating reactors, applicants for an operating license, and holders of construction permits to respond to certain generic concerns. These concerns are categorized into four areas: (1)

Post-TripReview,(2) Equipment Classification and Yendor Interface, (3)

Post-Maintenance Testing, and (4) Reactor Trip System Reliability l Improvements.

The first action item, Post-Trip Review, consists of Action Item 1.1, "Program Description and Procedure," and Action Item 1.2, "Data and Information Capability." This safety evaluation report (SER) addresses Action Item 1.1 only.

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11. REVIEW GUIDELINES The following review guidelines were developed after initial evaluation of various utility responses to Item 1.1 of Generic Lecter 83-28, and incorporate the best features of these submittals. As such, these review guidelines in effect represent a "good practices" approach to post-trip review. We have reviewed the licensee's response to item 1.1 against these guidelines:

A. The licensee or applicant should have systematic safety assessmemt procedures established that will ensure that the following restart criteria are met before restart is authorized.

The post-trip review team has determined the root cause amd sequence of events resulting in the plant trip.

• Near term corrective actions have been taken to remedy the cause of the trip.

• The post-trip review team has perfonned an analysis and determined that,the major safety systems responded to the event withis specified limits of the primary system parameters.

• The post-trip review has not resulted in the discovery of a potential safety concern (e.g.3 the root cause of the event occurs with a frequency significantly larger than expected).

• If any of the above restart criteria are not met, then an independent assessment of the event is perfonned by the Plant Operations Review Comittee (PORC), or another designated group with similar authority and experience.

B. The responsibilities and authorities of the personnel who will perform

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the review and analysis should be well defined.

The post-trip review team leader should be a member of plant management at the shift supervisor level or above and should hold or should have held an SR0 license fcc the plant. The team leader should be charged with overall responsibility for directing the post-trip review, including data gathering and data assessment and he/she should have the necessary authority to obtain all personnel and data needed for the post-trip review.

A second person on the review team should be an STA or should hold a relevant engineering degree with special transient analysis training.

The team leader and the STA (Engineer) should be responsibie to concur on a decision /recosinendation to restart the plant. A nonconcurrence from either of these persons should be sufficient to prevent restart until the trip has been reviewed by the PORC or equivalent organization.

C. The licen$ee or applicant should indicate that the plant response to the trip event will be evaluated and a determination made as to whether the plant response was within acceptable limits. The evaluation should include:

A verification of the proper operatiam of plant systems and equipment by comparison of the pertirent data obtained during the post-trip review to the applicable data provided in the FSAR,

• An analysis of the sequence of events to verify the proper functioning of safety related and other important equipment. Where possible, comparisons with previous similar events should be made.

D. The,11censee or applicant should have procedures to ensure that all physical evidence necessary for an independent assessment is preserved.

E. Each licensee or applicant should provide in its submittal, copies of the plant procedures which contain the information required in Items A through D. As a minimum, these should include the following:

The criteria for determining the acceptability of restart The qualifications, responsibilities and authorities of key personnel involved in the post-trip review process The methods and criteria for determining whether the plant variables and system responses were within the limits as described in the FSAR The criteria for determining the need for an independent review.

III. EVALUATION AND CONCLUSION By letters dat,ed November 7,1983, and August 13, 1985, the licensee of Pilgrim Statio'n provided infonnation regarding its Post-Trip Review Program and Procedures. We have evaluated the licensee's program and procedures against the review guidelines described in Section II. A brief description of the licensee's response and the staff's evaluation of the response against each of the review guidelines is provided below:

A. The licensee has established the criteria for determining the acceptability of restart. Based on our review, we find that the licensee's criteria conform to the guidelines described in Section II.A and, therefore, are acceptable.

i 5-B. The qualifications, responsibilities and authorities of the personnel who will perform the review and analysis have been clearly described.

We have reviewed the licensee's chain of conmand for responsibility for post-trip review and evaluation, and find it acceptable.

C. The licensee has described the methods and criteria for comparing the event infonnation with known or expected plant behavior. Based on our review, we find them to be acceptable.

D. The licensee has established the criteria for determining the need for an independent assessment conducted by the Operation Review Connittee.

In addition, the licensee has established procedures to ensure that all physical evidence necessary for an independent assessment is preserved.

We find these actions conform with the guidelines described in Sections !!.A and D.

E. The licensee has provided a systematic safety assessment program to assess unscheduled reactor trips. We have reviewed this program and find it acceptable.

Based on our r,eview, we conclude that the licensee's Post-Trip Review Progras and Procedures' for Pilgrim Station are acceptable.

Enclosure 2 SALP EVALUATION PILGRIM STATION DOCKET NO.: 50-293 GENERIC LETTER 83-28, ITEM 1.1, POST-TRIP REVIEW i

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Docket No.: 50-293 JUL 101986 50 .293 i

MEMORANDUM FOR: Paul Leech, Senior Project Manager, BWR Project Directorate il Division of BWR Licensing FROM: Gus C. Lainas, Assistant Director for BWR Division of BWR Licensing

SUBJECT:

SAFETY EVALUATION REPORT FOR PILGRIM STATION, GENERIC LETTER 83-28 ITEM 1.2 (POST-TRIP REVIEW -

DATA AND INFORMATION CAPABILITY) DOCKET NO. 50-293 Plant Name: Pilgrim Station Docket No.: 50-293 Utility: Boston Edison Company Licensing Status: OR Resp. Directurate: PD #1/ DBL Project Manager: P. Leech Review Branch: EICSB/DPA l Review Status: Complete By letter dated November 7,1983, the Boston Edison Canpany responded to Generic Letter 83-28 with regard to required actions based on generic implications of Salem ATWS events. The enclosed SER and SALP were prepared by the Electrical, Instru-mentation and Control Syste..rs Branch of DPA based upon TER input from its con-tractor, Science Applications International Corporation (SAIC), af ter having re-viewed the applicable portions of the licensee's response to Generic Letter 83-28 for Pilgrim Station. The TER was discussed with the licensee in telephone con-versations between Mr. W. Lobo, the licensee's representative, and Mr. Raj Auluck and Mr. Joel Kramer of the NRC on February 12 and 26,1986.

Based on our review, we conclude that the licensee's post-trio review data and information capabilities for Pilgrim Station are acceptable.

l This review has been conducted by J. J. Kramer (X28408) and SAIC. There are j no known dissenting professional opinions on this matter.

I Origumi signed la

b. - [ctW l c s C. Lainas, Assistant Director l

Division of BWR Licensing l

l

Enclosures:

) 1. Safety Evaluation 1

2. SALP Evaluation ec: See attached list l EICSB/DPA SL/EICSB/DPA BC/EIC A SB/ DBL A 1 1 r:ct SHWeis FRos C ainas MSrinivasan g.4 7[ 86 7/ 1 /85 7/ 9/86 7//0 /86 7/ d/86 0F/ CIAL RECORD COPY

..__ . , . Information in this record was deleted M

g - ( G f _r W F * *1 in accordance with the Freedom of Informadon

,- -- - - = .T o ff- Act, exemntions M

" F0IA- If- 4 N $.

P. Leech . 2-cc: R. Liner, SAIC

0. Gallagher, SAIC C. E. Rossi J. Zwolinski E. Adensam T. Bournia Distribution: $3 Docket File No. 50-N4 EICSB Rdg.

J. Kramer (PF)(2)

5. Weiss F. Rosa M. Srinivasan G. Lainas Pilgrim S/F r ~ s- w ,-+-w - , - - - , - - --- ---.- -- -

e

• SAFETY EVALUATION REPORT GENERIC LETTER 83-28, ITEM 1.2 - POST-TRIP REVIEW (DATA AND INFORMATION CAPABILITT)

PILGRIM STATION DOCKET NO. 50-293 I. INTRODUCTION On February 25, 1983, both of the scram circuit breakers at Unit 1 of the Salem Nuclear Power Plant (SNPP) failed to open upon an automatic reactor trip signal from the reactor protection system. This incident occurred during the plant start-up and the reactor was tripped manually by the operator about 30 seconds after the initiation of the automatic trip signal. The failure of the circuit breakers has been detennined to be related to the sticking of the undervoltage trip attachment. On February 22, 1983, during start-up of SNPP, Unit 1, an automatic trip signal occurred as the result of steam generator low-low level. In this case, the reactor was tripped manually by the operator almost coincidentally with the automatic trip. Following these incidents, on February 28, 1983, the NRC Executive Director for Operations (E00) directed the staff to investigate and report on the generic implications of these occurrences. The results of the staff's inquiry into these incidents are re-ported in NUREG-1000, "Generic Implications of ATWS Events at the Salem Nuclear Power Plant." As a result of this investigation, the Consission requested (by Generic Letter 83-28 dated July 8,1983) all licensees of operating reactors, applicants for an operating license, and holders of construction permits to respond to certain generic concerns. These concerns are categorized into four areas: (1) Post-Trip Review, (2) Equipment Classification and Vendor Interface, (3) Post-Maintenance Testing, and (4) Reactor Trip System Reliability Improve-ments, w sw . . _. .- .. .

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The first action item, Post-Trip Review, consists of Action Ites 1.1, "Program Description and Procedure" and Action Item 1.2, "Data and Information Capabili ty. " This safety evaluation report (SER) addresses Action Item 1.2 only.

II. REVIEW GUIDELINES The following review guidelines were developed after initial evaluation of the various utility responses to Item 1.2 of Generic Letter 83-28 and incorporate the best features of these submittals. As such, these review guidelines in effect represent a "good practices" approach to post-trip review. ble have re-viewed the licensee's response to Item 1.2 against these guidelines:

A. The equipment that provides the digital sequence of events (SOE) record and the analog time history records of an unscheduled shutdown should provide a reliable source of the necessary information to be used in the post-trip review. Each plant variable which is necessary to determine the cause and progression of the events following a post trip should be monitored by at least one recorder (such as a sequence-of-events recorder or a plant process computer) for digital parameters; and strip charts, a plant process computer or analog recorder for analog (time history) variables. Performance characteristics guidelines for SOE and time history recorders are as follows:

o Each sequence of events recorder should be capable of detecting and recording the sequence of events with a sufficient time discrimination capability to ensure that the time responses associated with each monitored safety-related system can be ascertained, and that a detennination can be made as to whether the time response is within acceptable limits based on FSAR Chapter 15 Accident Analyses. The reconsnended guidelines for the SOE time discrimination is approximately 100 milliseconds. If current SOE recorders do not have this time discrimination capability the licensee should show that the current time discrimination capability is sufficient for an adequate reconstruction of the course of the reactor trip and post-trip events. As a minimum this should include the ability to adequately reconstruct the transient and accident scenarios presented in Chapter 15 of the plant FSAR.

, o Each analog time history data recorder should have a sample interval small enough so that the incident can be accurately i

reconstructed following a reactor trip. As a minimum, the licensee should be able to reconstruct the course of the transient and accident sequences evaluated in the accident 1

4 analysis of Chapter 15 of the plant FSAR. The recammended guideline for the sample interval is 10 seconds. If the time history equipment does not meet this guideline, the licensee should show that the time history capability is sufficient to accurately reconstruct the transient and accident sequences presented in Chapter 15 of the FSAR. To support the post-trip analysis of the cause of the trip and the proper functioning of involved safety related equipment, each analog history data recorder should be capable of updating and retaining infonnation from approximately five minutes prior to the trip until at least ten minutes after the trip.

o All equipment used to record sequence of events and time history infonnation should be powered from a reliable and non-interruptible power source. The power source used need not be Class 1E.

B. The sequence of events and time history recording equipment should monitor sufficient digital and analog parameters, respectively, to assure that the course of the reactor trip and post-trip events can be reconstructed. The parameters monitored should provide sufficient infonnation to determine the root cause of the unscheduled shutdown, the progression of the reactor trip, and the response of the plant parameters and protection and safety systems to the unscheduled shutdowns. Specifically, all input parameters

- S-associated with reactor trips, safety injections and other safety-related systems as well as output parameters sufficient to record the proper functioning of these systems should be recorded for use in the post-trip review. The parameters deemed necessary, as a minimum, to perfonn a post-trip review that would detennine if the plant remained within its safety limit design envelope are presented in Table 1. They were selected on the basis of staff engineering judgrnent following a complete evaluation of utility submittals. If the licensee's SOE recorders and time history recorders do not monitor all of the parameters suggested in these tables the licensee should show that the existing set of monitored parameters are sufficient to establish that the plant remained within the design envelope for the accident conditions analyzed in Chapter 15 of the plant FSAR.

C. The infonnation gathered by the sequence of events and time history recorders should be stored in a manner that will allow for data retrieval and analysis. The data may be retained in either hardcopy (e.g., com-puter printout, strip chart record), or in an accessible memory (e.g.,

magneticdiscortape). This information should be presented in a read-able and meaningful fonnat, taking into consideration good human factors practices such as those outlined in N1JREG-0700.

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I D. Retention of data from all unscheduled shutdowns provides a valuable

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reference source for the determination of the acceptability of the plant vital parameter and equipment response to subsequent unscheduled l

shutdowns. Infonnation gathered during the post-trip review is to be retained for the life of the plant for post-trip review comparisons of subsequent events.

III. EVALUATION AND CONCLUSION By letter dated November 7,1983, the Boston Edison Company provided information regarding its post-trip review program data and infonnaticei capabilities for Pilgrim Station. We have evaluated the licensee's submittal against the review guidelines described in Section II. Deviations from the pidelines of Section II were discussed with representatives of the licensee by telephone on February 12 and 26,1986. A brief description of the licensee's responses and the staff's evaluation of the responses against each of the review guidelines follows:

A. The licensee has described the perfonnance characteristics of the equipment used to record the sequence of events and time history data needed for post-trip review. Based on our review of the licensee's submittal and the information provided by the licensee during the above telecons, we find that the sequence of events reconke and time history recorder characteristics conform to the guidelines described in Section II A, and are acceptable.

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8. The licensee has established and identified the parameters to be monitored and recorded for post-trip review. Based on our review, up find that the parameters selected by the licensee include all but one of those identified in Table 1. While Diesel Generator Status is not included as a parameter, there are indicators in the control room that provide this infomation. The staff finds this acceptable. Con-sequently, we find that the licensee's selection of r'rameters meets the intent of the guidelines described in Section II.B and is, there-fore, acceptable.

C. The licensee described the means for storage and retrieval of the infonnation gathered by the sequence et events and time history recorders, and for the presentation of this information for post-trip review and analysis. Based on our review, we find that this infonsation will be presented in a readable and meaningful format, and that the storage, retrieval and presentation confonn to the guidelines of Section II C.

I D. During the February 26, 1986 telecon, the licensee stated that the data and infonnation used during post-trip reviews are being retained is an l accessible manner for the life of the plant. Based on this infoniation, i

we find that the licensee's program for data retention conforms to the guidelines of Section II D, and is acceptable.

8 Based on our review of the applicant's submittal and the telecons with the licensee, we conclude that the licensee's post-trip review data and infonna-tion capabilities for Pilgrim Station are acceptable.

9 TABLE 1 BWR PARAMETER LIST SOE Time History Recorder Recorder Parameter / Signal x Reactor Trip ,

x Safety Injection x Containment Isolation x Turbine Trip

-x Control Rod Position x(1) x Neutron Flux, Fower x(1) Main Steam Radiation (2) Containment (Dry Well) Radiation x(1) x Drywell Pressure (Containment Pressure)

(2) ppre.sion Pool Temperature

, x(1) x Primary System Pressure x(1) x Primary System Level x MS!Y Position

. x(1) Turbine Stop Valve / Control Valve Position x Turbine Bypass Valve Positics x Feedwater Flow x Steam Flow (3) Recirculation; Flow. Pump Jtatus x(1) Scram Discharge Level..

x(1) Condenser Yacuum

r SOE Time History Recorder Recorder _Pa ramete r/ Signal x -

AC and DC System Status (Bus Voltage)

(3) (4) Safety Injection; Flow. Pump /Yalve Status x ,. Diesel Generator Status (on/Off, Start /Stop)

(1) Trip parameters (2) Parameter say be recorded by either an SOE or time history recorder.

(3) Acceptable recorder options arej (a) system flow recortled on an SOE recorder, (b) system flow recorded on a time history recorder, or (c) equipment status recorded on an SOE recordar.  !

(4) Includes recording of parameters for all applicable systems from the following: HPCI, LPCI, LPCS,1C. RCIC.

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ENCLOSURE 2 SALP EVALUATION PILGRIM STATION DOCKET NO 50-293 GENERIC LETTER 83-28, ITEM 1.2, POST-TRIP REVIEW l

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