Text

Application for Amend to License NPF-21,revising TS to Relocate Component Lists in Accordance W/Gl 91-08 & Selected Implementation of Commission Policy Statement of TS Improvements
	ML17290A702
Person / Time
Site:	Columbia
Issue date:	10/21/1993
From:	Parrish J; WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML17290A703	List: ML17290A702; ML17290A704;
References
	GL-91-08, GL-91-8, GO2-93-256, NUDOCS 9310260394
	Download: ML17290A702 (84)
	v • d • e

ACCEI ERATO DOCUMENT DIST UTION SYSTEM REGULATO~ INFORMATION DISTRIBUTION STEM (RIDS)

ACCESSION NBR:9310260394 DOC.DATE: 93/10/21 NOTARIZED: NO DOCKET FACIL:50-397 WPPSS Nuclear Project, Unit 2, Washington Public Powe 05000397 AUTH. NAME AUTHOR AFFILIATION PARRISH,J.V. Washington Public Power Supply System RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)

SUBJECT:

Application for amend to License NPF-2l,revising TS to relocate component lists in accordance w/GL 91-08 S selected implementation of Commission policy statement of TS improvements.

DISTRIBUTION CODE: AOOZD TITLE: OR COPIES RECEIVED:LTR Submittal: General Distribution L ENCL g SIZE: 7+

NOTES:

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PDV LA 1 1 PDV PD 1 1 CLIFFORD,J 2 2 INTERNAL: ACRS 6 6 NRR/DE/EELB 1 1 NRR/DORS/OTSB 1 1 NRR/DRCH/HICB 1 1 NRR/DSSA/SPLB 1 1 NRR/DSSA/SRXB 1 1 NUDOCS-ABSTRACT 1 1 O~CMEDCB 1 0 OGC/HDSl 1 0 EG FIL 01 1 1 EXTERNAL: NRC PDR 1 1 NSIC 1 1 NOTE TO ALL"RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTEl CONTACI'HE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 504-2065) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEEDl TOTAL NUMBER OF COPIES REQUIRED: LTTR 21 ENCL 19

WASHINGTON PUBLIC POWER SUPPLY SYSTEM P.O. Bm96'8 ~ 3000George Wasbtngton Way ~ Rtcbkrnd, Wasbtngton 99352-0968 ~ (5093372-5000 October 21, 1993 Docket No. 50-397 G02-93-256 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555 Gentlemen:

Subject:

NUCLEAR PLANT NO. 2, OPERATING LICENSE NPF-21 REQUEST FOR AMENDMENTTO TECHNICAL SPECIFICATIONS TO RELOCATE COMPONENT LISTS IN ACCORDANCE %1TH GENERIC LETTER 91-08 AND SELECTED IMPLEMIPlTATION OF THE COMMISSIO¹S POLICY STATEMENT ON TECHNICAL SPECIFICATIONS IMPROVEMENTS In accordance with the Code of Federal Regulations, Title 10, Parts 50,90 and 2.101, the Supply System hereby submits a request for amendment to the WNP-2 Technical Specifications.

Specifically, the Supply System requests revision of the WNP-2 Technical Specifications consistent with the guidance of Generic Letter 91-08. These changes will implement the relocation of component lists from the Technical Specifications. Additionally, several of these component specific lists are part of Technical Specifications that do not meet the criteria for inclusion in the Technical Specifications as defined in the Commission's Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors as noticed in the Federal Register on July 22, 1993. These Technical Specification requirements are also not included in NUREG-1433, "Standard Technical Specifications, General Electric Plants, BWR/4," dated September 28, 1992. The identified Technical Specifications, as well as the component specific lists, are being relocated to a new plant procedure which will be controlled pursuant to the requirements of 10CFR50,59 and Technical Specification 6.8.1.

The relocation of these Limiting Conditions For Operation (LCOs) is made in accordance with the Commission policy that licensees may adopt portions of the improved Standard Technical Specifications (STS) without fully implementing all Technical Specification improvements as stated in the supplementary information provided with the Final Policy Statement in the Federal Register. The Commission also stated that LCOs which do not meet any of the four criteria cited below may be proposed for removal from the Technical Specifications and relocation to licensee-controlled documents, such as the FSAR. The criteria may be applied to either standard or custom Technical Specifications. The identified LCOs are proposed for relocation based on failure to meet the four criteria for inclusion in Technical Specifications.

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9310260394 93i02i PDR ADOCK 05000397 PDR

0 Page Two REQUEST FOR AMEND TO TS TO RELOCATE COMPONENT LISTS IN ACCORDANCE WITH( GL 91-08 AND SELECTED IMPLEMENTATION OF THE COMMISSION'S POLICY STATEMENT ON TSI The specific outline for Technical Specifications required to meet 10 CFR 50.36a is provided by Generic Letter 91-08.t The guidelines provided by Generic Letter 91-08 nnd the Policy Statement are followed in this change request in that the component'pecific lists, and the relocated LCOs, willbe included in a plant procedure as recommended in Generic Letter 91-08.

This Technical Specification amendment request includes the following proposed relocations from the Technical Specifications to a Supply System controlled plant procedure:

1) Relocation of Table 3.4.3.2-2, Reactor Coolant System Interface Valves Leakage Pressure Monitors, and the associated Action statement 3.4.3.2.d and Surveillance Requirement 4.4.3.2.3. This relocation is performed in accordance with the Commission's Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors. Relocation of Table 3.4.3.2-2 also meets the relocation guidance of Generic Letter 91-08. This change also deletes the note at the bottom of page 3/4 4-11 since this note was only applicable until the start of the first refueling outage.

2) Relocation of Table 3.6.3-1, Primary Containment Isolation Valves, to a plant procedure in accordance with the guidance of Generic Letter 91-08. This change includes deletion of references to this table. It also includes the addition to the Bases section the expectations for opening under administrative control. locked or sealed containment isolation valves. These expectations have been modified from the wording included in Generic Letter 91-08 to provide additional clarification relative to plant specific details.

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3) Relocation of Technical Specification 3.8.4.1, A.C. Circuits Inside Primary Containment, in accordance with! the Commission's Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors. The listing of components in the Limiting Condition For Operation section of this Specification meets the purpose of Generic Letter 91-08 and thus could be relocated in accordance with the guidance of the Generic Letter.

4) Relocation of Te'chnical Specification 3.8.4.2, Primary Containment Penetration Conductor Overcurrent Protective Devices, including Table 3.8.4.2-1, in accordance with the Commission's Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors. Relocation of Table 3.8.4.2-1 also meets the guidance for relocation per Generic Letter 91-08.

5) Relocation of Technical Specification 3.8.4.3, Motor-Operated Valves Thermal Overload Protection, including Table 3.8.4.3-1, in accordance with the Commission's Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors.

Relocation of Table 3.8.4.3-1 also meets the guidance for relocation per Generic Letter 91-08.

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REQUEST FOR AMEND TO TS TO RELOCATE COMPONENT LISTS IN ACCORDANCE WITH GL 91-08 AND SELECTED IMPLEMENTATION OF THE COMMISSION'S POLICY STATEMENT ON TSI The Commission's Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors contains four criteria against which Technical Specifications are to be evaluated for inclusion in the Improved Technical Specifications. Meeting any of the four criteria results in inclusion in the Improved Technical Specifications. These four criteria are:

Criterion 1 - Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

Criterion 2 - A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or prevents a challenge to the integrity of a fission product barrier.

Criterion 3 - A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient analysis that either assumes the failure of or prevents a challenge to the integrity of a fission product barrier.

Criterion 4 - A structure, system, or component which operating experience or prob'abilistic safety assessment has shown to be significant to public health and safety.

As detailed below, the Supply System has evaluated each of the proposed changes identified above per the requirements of 10 CFR 50.92 and determined they do not represent an unreviewed safety question or a significant hazard.

With respect to the proposed relocation of Table 3.6.3-1, Primary Containment Isolation Valves, this change is made in accordance with the guidance provided in Generic Letter 91-08. The Supply System has evaluated this proposed change per the requirements of 10 CFR 50.92 and determined it does not represent an unreviewed safety question or a significant hazards consideration because it does not:

1) Involve a significant increase in the probability or consequences of an accident previously evaluated. The relocation of Table 3.6.3-1 from the Technical Specifications to a licensee controlled document is administrative in nature. The surveillance and operability requirements remain in the Technical Specifications. The Technical Specification restrictions and actions are not being revised or relaxed by this change. The accident analyses that rely on these components are not affected. Plant procedures will only be revised where necessary to reflect the relocation of the Table. Changes to the relocated valve information will be controlled and made in accordance with the administrative controls required by Technical Specification 6.8.1 and 10 CFR 50.59. Since this change does not affect the content, control, or adherence to the Technical Specification, the probability or consequences of previously evaluated accidents are not impacted.

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Page Four REQUEST FOR AMEND TO TS TO RELOCATE COMPONENT LISTS IN ACCORDANCE WITH GL 91-08 AND SELECTED IMPLEMENTATION OF THE COMMISSION'S POLICY STATEMENT ON TSI

2) Create the possibility of a new or different kind of accident from any previously evaluated. The relocation of information from Table 3.6.3-1 to a licensee controlled document serves to consolidate information on affected components. This information willbe controlled under appropriate administrative requirements. This relocation is made in accordance with the guidance of Generic Letter 91-08. The proposed revision does not involve a change in the manner in which these valves will be operated, maintained, or tested. No components or systems are physically added, removed, or modified as a result of the proposed change. Therefore, a new or different kind of accident as a result of this change is not credible.

3) Involve a significant reduction in a margin of safety. The margin of safety associated with these valves is unaffected by this proposed change since the applicable operability and surveillance requirements are not being revised except administratively as recommended by the Generic Letter. No technical changes to valve oyeration or maintenance will result due to this change. The incoryoration of this information into a new plant procedure will ensure changes are made in accordance with the controls of Technical Specification 6.8.1 and 10 CFR 50.59.

The relocation of Table 3.4.3.2-2, Reactor Coolant System Interface Valves Leakage Pressure Monitors, and the associated Action statement 3.4.3.2.d and Surveillance Requirement 4.4.3.2.3, and the relocation of Technical Specifications 3.8.4.1, A.C. Circuits Inside Primary Containment, 3.8.4.2, Primary Containment Penetration Conductor Overcurrent Protective Devices, and Technical Syecification 3.8.4.3, Motor-Operated Valves Thermal Overload Protection; have been evaluated by the Supyly System against the requirements of 10 CFR 50.92. It has been determined that these changes do not reyresent a significant hazards consideration since they do not:

1) Involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed changes are consistent with the Commission's Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors.

Specifically, the Specifications being proposed for relocation to Supply System controlled documents do not meet any of the four criteria specified in the Policy Statement for equipment to be included in the Technical Specifications.

Regarding the Reactor Coolant System Interface Valves Leakage Pressure Monitors, these instruments do not meet Criterion 1 since the Policy Statement specifically excludes "instrumentation to identify the source of actual leakage." These Pressure Monitors are designed to identify the source of actual leakage.

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'EQUEST FOR AMEND TO TS TO RELOCATE COMPONENT LISTS IN ACCORDANCE WITH GL 91-08 AND SELECTED IMPLEMENTATION OF THE COMMISSION'S POLICY STATEMENT ON TSI The relocation of these Technical Specification requirements to a new plant procedure does not change the requirements. The required testing and associated Actions for out of service equipment will continue to be met. The current restrictions and actions are not being revised or relaxed by this change. The accident analyses that rely on, these components are not affected. Relocation results in licensee control of future changes under the requirements of 10 CFR 50.59. Removal of the note at the bottom of page 3/4 4-11 is administrative in nature since this note was applicable only until the first refueling outage. Thus, there is no significant increase in the probability or consequences of an accident previously evaluated as a result of these changes.

2) Create the possibility of a new or different kind of accident from any previously evaluated. This relocated information willbe controlled under appropriate administrative requirements. This relocation is made in accordance with the guidance of the Commission's Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors. The proposed revisions do not involve a change in the manner in which the equipment will be operated, maintained, or tested, nor in the actions that will be taken should the equipment be out of service or incapable of performing their intended safety functions. No components or systems are physically added, removed, or modified as a result of the proposed change. Therefore, this change will not result in the possibility of a new or different type of accident than those previously evaluated.

3) Involve a significant reduction in a margin of safety. The requested changes are not the result of a physical change to the plant or the manner in which the plant will be operated.

The accident analyses for the plant as described in the FSAR are not affected by this proposed change. The operation, maintenance, and testing of equipment is not affected.

Therefore, the margin of safety for the plant is not significantly reduced as a result of these changes.

As discussed above, the Supply System considers that the proposed changes do not involve a significant hazards consideration, nor is there a potential for a change in the types or increase in the amount of any effluents that may be released offsite, nor do they involve an increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed changes meet the eligibilitycriteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, per 10 CFR 51.22(b), an environmental assessment of these changes is not required.

The approved amendment will be implemented within thirty days of receipt. Numerous rocedural proc u changes will be required to delete the reference to the Technical Specification requirements that will be relocated by this request. Since the Technical Specification requirement relocation does not result in a change in requirements, the resulting reference changes in the procedures willbe treated as administrative in nature and willbe made at the next

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Page Six REQUEST FOR AM~22lD TO TS TO RELOCATE COMPONENT LISTS IN ACCORDANCE WITH GL 91-08 AND SELECTED IMPLEMENTATION OF THE COMMISSION'S POLICY STATEMENT ON TSI available opportunity. Special procedure revisions to delete reference to the relocated Technical Specifications will thus not be made. As stated above, the specified testing will continue to be performed at the current periodicity with any future changes to these requirements made per the requirements of 10CFR50.59.

This Technical Specification change has been reviewed and approved by the WNP-2 Plant Operations Committee and the Supply System Corporate Nuclear Safety Review Board (CNSRB). In accordance with 10 CFR 50.91, the State of Washington has been provided a copy of this letter.

Very yours,

. Parrish, Assistant Managing Director, Operations DAS/ds Attachments CC: BH Faulkenberry - NRC RV NS Reynolds - Winston 8t: Strawn RG Waldo - EFSEC JW Clifford - NRR DL Williams - BPA NRC Site Inspector - 901A

STATE OF WASHINGTON )

Subject:

Amendment to Tech Specs to

) Relocate Com onent List COUNTY OF BENTON )

I, J. V. PARRISH, being duly sworn, subscribe to and say that I am the Assistant Managing Director, Operations for the WASHINGTON PUBLIC POWER SUPPLY SYSTEM, the applicant herein; that I have the full authority to execute this oath; that I have reviewed the foregoing; and that to the best of my knowledge, information, and belief the statements made in it are true.

DATE Z0 , 1993 J . Parrish, Assistant Managing Director Operations On this date personally appeared before me J. V. PARRISH, to me known to be the individual who executed the foregoing instrument, and acknowledged that he signed the same as his free act and deed for the uses and purposes herein mentioned.

GIVEN under my hand and seal this ~~day of 1993.

Notary Public in and for the STATE OF WASHINGTON

~15 Residing at Kennewick W . hin n MyC i i Epi

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lJ INDEX

'IMITING CONDITIONS FOR OPERATION ANO SURVEILLANCE RE UIREMENTS SECTION PAGE ELECTRICAL POWER SYSTEMS (Continued) 3/4.8. 3 ONSITE POWER DISTRIBUTION SYSTEMS Distribution Operating...;.......................... 3/4 8-16 Distribution " Shutdown............................... 3/4 8-18 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Reactor Protection System Power Supply Monitoring .... .3/4 8-QCP 3/4. 9 REFUELING OPERATIONS 3/4. 9. 1 REACTOR MODE SWITCH................................... 3/4 9"1 3/4.9. 2 INSTRUMENTATION....................................... 3/4 9-3 3/4. 9. 3 CONTROL ROD POSITION.................................. 3/4- 9-5 3/4.9.4 D ECAY TIME............................................ 3/4 9-6 3/4. 9. 5 CO~MU~ICATIONS........................................ 3/4 9-7 3/4. 9. 6 REFUELING PLATFORM.................................... 3/4 9"8 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL................ 3/4 9-9 3/4.9.8 WATER LEVEL - REACTOR VESSEL.......................... 3/4 9-11 3/4.9.9 WATER LEVEL - SPENT FUEL STORAGE POOL................. 3/4 9-12 3/4.9.10 CONTROL ROD REMOVAL Single Control Rod Removal............................ 3/4 9-13 Multiple Control Rod Removal.......................... 3/4 9-15 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION H sghh Mater Level...................................... 3/4 9-17 Low Water Level.......................-............... 3/4 9-18 WASHINGTON NUCLEAR - UNIT 2 9310260394

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CONT ROlkED COPY LIST OF TABLES Continued TABLE PAGE 3.3.7. 5" 1 ACCIDENT MONITORING INSTRUMENTATION.................. 3/4 3-71 4.3.7. 5-1 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.................;.......... 3/4 3"74 3.3.7. 12" 1 EXPLOSIVE GAS MONITORING INSTRUMENTATION............. 3/4 3-80 4.3.7. 12-1 EXPLOSIVE GAS MONITORING INSTRUMENTATION SURVEILLANCE REQUIREHENTS............................ 3/4 3"81

3. 3. 9-1 FEEDWATER SYSTEM/MAIN TURBINE TRIP SYSTEH ACTUATION INSTRUHENTATION............................ 3/4 3-85 3.3. 9-2 FEEDWATER SYSTEM/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SETPOINTS.................. 3/4 3"86 4;=3. 9. 1-1 FEEDWATER SYSTEM/HAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS.............. m......................... 3/4 3"87 3.4.3.2-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES..... 3/4 4-11 O~/etc) 3.4.3. 2-2

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3. 4. 4-1 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS.............. 3/4 4-14

4. 4. 5-1 PRIMARY COOLANT SPECIFIC ACTIVITY SAHPLE ANO ANALYSIS PROGRAM..................................... 3/4 4-17 WASHINGTON NUCLEAR - UNIT 2 XXlll Amendment No. $7, 98

CGMI RC f l;":D COPY INDEX LIST OF TABLES Continued TABLE PAGE 4.4. 6.1. 3-1 ' 3/4 4"22 O ELETEDo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ t ~

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3. 6. 3" 1

3. 6. 5. 2-1 SECONOARY CONTAINHENT VENTILATION SYSTEM AUTOMATIC ISOLATION VALVES....................... 3/4 6-39 3.7. 8-1 AREA TEHPERATURE MONITORING ...................... 3/4 7-31

4. 8.1.1. 2-1 OIESEL GENERATOR TEST SCHEDULE ................... 3/4 8-9 4.8.2.1"1 BATTERY SURVEILLANCE RE(UIREMENTS ................ 3/4 8-14 QELETSQ

3. 8. 4. 2-1 PEL.mme

3. 8. 4. 3-1 B3/4.4,6-1 REACTOR VESSEL TOUGHNESS ......................... B 3/4 4-6

5. 7. 1" 1 COMPONENT CYCLIC OR TRANSIENT LIHITS ............. 5-7

6. 2. 2-1 MINIMUM SHIFT CREW COMPOSITION "

SINGLE UNIT FACILITY ............................ 6-6 WASHINGTON NUCLEAR " UNIT 2 Xxiv Amendment No. 87, 98, 107

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DEFINITIONS OPERABLE " OPERABILITY 1.28 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s) and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or othe~ auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function(s) are also capable of performing their related support function(s).

OP ERAT IONA L CONDITION " CONDITION 1.29 An OPERATIONAL CONDITION, i.e., CONDITION, shall be any one inclusive combination of mode switch position and average reactor coolant temperature as specified in Table 1.2.

PHYSICS TESTS 1.30 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation as (1) described in Chapter 14 of the FSAR, (2) author ized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE QQ 1.31 PRESSURE BOUNDARY LEAKAGE shall be leakage through a non-isolable fault G4~" <>-tao in a reactor coolant system component body, pipe wall, or vessel wall.

PRIMARY CONTAINMENT INTEGRITY q

)q't 1 1.32 PRIMARY CONTAINMENT INTEGRITY shall exist when:

a. All primary containment penetrations required to be closed during accident conditions are either:

1. Capable of being closed by an OPERABLE primary containment automatic isolation system, or

2. Closed by at least one manual valve, blind flange, or deacti-vated automatic valve secured in its closed position. except es-Specification 3.6.3.

vahvcs >~<> <<~c op4w Aedce'd~'<<'0>~Yacc co+~) o% per~'it~ t,~

b. All primary containment equipment hatches are closed and sealed.

c. Each primary containment air lock is in compliance with the requirements of Specification 3.6. 1.3.

d. The primary containment leakage rates are within the limits of Specification 3.6. 1.2.

e. The suppression chamber is in compliance with the requirements of Specification 3.6.2.

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f. The sealing mechanism associated with each primary containment penetration; e.g., welds, bellows, or O-rings, is OPERABLE.

WASHINGTON NUCLEAR " UNIT 2 1-5 Amendment No. 28

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CONTROLLED COPY TABLE 3.3.2-3 (Continued)

ISOLATION SYSTEM INSTRUMENTATION RESPONSE T!HE TABLE NOTATIONS (a)The isolation system instrumentation response time shall be measured and

.recorded as a part of the ISOLATION SYSTEM RESPONSE TIHE. Isolation system instrumentation response time specified includes the diesel generator starting and sequence loading delays assumed in the accident analysis.

(b)Radiation detectors are exempt from response time testing. Response time shall be measured from detector output or the input of the first electronic component in the channel.

"Isolation system instrume'ntation response time for HSIVs only. No diesel generator delays assumed.

""Isolation system instrumentation response time for associated valves except MSIVs.

PIsolation system instrumentation response time specified for the Trip Function actuating each valve group shall be added to isolation time ~en-

\ ~ e in each valve group to obtain ISOLATION SYSTEH RESPONSE TIHE for each valve.

&This response time does not include the 45-second time delay.

for each power operated or automatic primary containment I

isolation valve and secondary containment ventilation system automatic isolation valve (Table 3.6.5.2-1)

WASHINGTON NUCLEAR

- UNIT 2 3/e 3-2I.

CONTROLLED COPY REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.3.2 Reactor coolant system leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE.

b. 5 gpm UNIDENTIFIED LEAKAGE.

c. 2 gpm increase in UNIDENTIFIED LEAKAGE within any 24-hour or less period.

d. 25 gpm total leakage averaged over any 24-hour period.

e. 1 gpm leakage at a reactor coolant system pressure of 950 k 10 psig from any reactor coolant system pressure isolation valve specified in Table 3.4.3.2-1.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION: I ao With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLO SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. With any reactor coolant system leakage greater than the limits in b.

and/or d. above, reduce the leakage rate to within the limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLO SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

With any reactor coolant system pressure isolation valve leakage

~~t" greater than the above limit, isolate the high pressure portion of o1 < the affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one other closed (manual or deactivated automatic) pcs (or check") valve, or be in at least HOT SHUTDOWN within the next 1 and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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d. ith one r more o the high/ ow pressu e 'nterfa valve le age ressure m nitors s own in Tab e 3.4.3.2 1 inopera le, restor the i operable onitor(s to OPERAB E status ithin 7 d s or veri the pr ssure to be less t an the al setpoin at least once per 1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ; re tore the 'noperabl monitor(s to OPERA E status ithin 30 d ys or be n at leas HOT SHU DOWN withi the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months a d in COLD SHU OWN with> the fol owing 24 h urs.

e. With any reactor coolant system UNIDENTIFIED LEAKAGE increase greater than 2 gpm within any 24-hour or less period, identify the source of leakage increase as not service sensitive Type ~~ or 31G .-'stenitic stainless steel within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLO SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

"Which has been verified not to exceed the allowable leakage limit at the last

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refueling outage or after the last time the valve was disturbed, whichever is more recent. ~

WASHINGTON NUCLEAR " UNIT 2 3/4 4-9 Amendment No. 111

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REACTOR COOLANT SYSTEM SURVEILLANCE RE UI REMENTS 4.4.3.2.1 The reactor coolant system leakage shall be demonstrated to be within each of the above limits by:

a. Monitoring the primary containment atmospheric particulate and gaseous radioactivity at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. Monitoring the primary containment sump flow rate at least once per shift, not to exceed 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c. Monitoring the reactor vessel head flange leak detection system at least once per 24 I hours.

4.4.3.2.2 Each reactor coolant system pressure isolation valve specified in Table 3.4.3.2-1 shall be demonstrated OPERABLE by leak testing pursuant to Specification 4. 0.5 and verifying the leakage of each valve to be within the specified limit:

a. At least once per 18 months.

b. Prior to returning the valve to service following maintenance, repair or replacement work on the valve which could affect its leakage rate.

The provisions of Specification 4.0.4 are not applicable for entry into OPERATIONAL CONDITION 3.

.4.3. .3 The h gh/low pre sure int rface val e leaka e press re mon'tor s all b demonst ted OPERAB E with a arm setpo nts per Table 3 4. 3. 2- b pe forma ce of a:

a. HANNEL F CTIONAL T T at le st once pe 31 day, and C gNNEL CAL BRATION at least on e per 18 m nths.

WASHINGTON NUCLEAR " UNIT 2 3/4 4" 10 Amendment No. g, i1i

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3/4.6 CONTAINMENT SYSTEMS 3/4.6. 1 PRIMARY COHTAINMENT PRIMARY CONTAINMEHT INTEGRITY LIMITIHG COHDITIOH FOR OPERATION 3.6.1.1 PRIMARY CONTAINMENT IHTEGRITY shall be maintained.

APPLICABILITY: OPERATIOHAI CONDITIONS 1, 2" and 3.

ACTION:

Mithout PRIMARY CONTAINMENT INTEGRITY, restore PRIMARY CONTAINMENT INTEGRITY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT SHUTDOMN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE RE UIREMENTS 4.6.1.1 PRIMARY CONTAINMENT INTEGRITY shall be demonstrated:

a~ After each closing of each penetration subject to Type B testing, pele>64 ~y except the primary containment air locks, if opened followin Type A or 8 test, b leak rate testing the seals with gas at P a"34. psig,

~ gg-q3-t PO and verifying that when the measured leakage rate for these seals is added to the leakage rates determined pursuant to Surveillance Requirement 4.6. 1,2.d for all other Type B and C penetrations, the combined leakage rate is less than or equal to 0.60 L .

At least once per 31 days by verifying that all primary containment penetrations",~ not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated auto-a4~i~isb Rim ca~4 7 matic valves secured in position, except as per~'d, bg ~ -e$ - Specification 3.6.3. d'or vaKvw ~p yw oyco u~cte.~

C. By verifying each primary containment air lock is in compliance with the requirements of Specification 3.6.1.3.

d. By verifying the suppression chamber is in compliance with the re-quirements of Specification 3.6.2.1.

"See Special Test Exception 3. 10. 1.

""Except valves, blind flanges, hnd deactivated atuomatic valves which are with-in the primary containment or other areas administratively controlled to pro-hibit access for reasons of personnel safety (i;e., radiation and temperature) and are locked, sealed, or otherwise secured in the closed position (l~ inch and smaller valves connected to vents, drains or test connections must be closed but need not be sealed). Valves inside containment shall be verified closed following primary containment de-inerting, but verification is not required more often than once per 92 days. Valves in other administratively controlled areas shall be verified closed during each COLD SHUTDOWN, but verification is not required more often than once per 31 days.

WASHINGTON NUCLEAR - UNIT 2 3/4 6"1 Amendment Ho. 22

CONTAIN)1ENT SYSTEMS PRIMARY CONTAINMENT LEAKAGE LINITIHG CONDITION FOR OPERATION

3. 6.1. 2 Primary containment leakage rates shall be limited to:,

han or equal to L, r

a. An overall integrated leakage ra e of less 0.50 ercent by weioht of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at P, 34.7 ps>g.

A combined leakage rate of less than or equal to 0.60 L for and a'll valves

' ." (except for main all'enetrations e

i'd steam line isolation valves"poland valves which are hydrostatically gO gY~~~ 85~ ressurized to P , 4. 7 pslg.

1 C. "Less than or equal to 11.5 scf per hour for any one main steam line isolation valve when tested at Pt, 25.0 psig..

d. A combined leakage rate of less than or equal to 1 gpm times the total number os ECCS and RCIC containment isolation valves in hydrostatically tested lines which penetrate thy primary containment, when tested at l. 10 P , 8. 2 psso APPLICABiLITY: Mhen PR:MARY COHTAINHEHT INTEGRITY is reouired per ACTION:

Specification 3.6. 1- -.

Mi th:

a. The measured overall integrated primary containment leakage rate exceeding 0.75 La'r

b. The measured combined leakage rate for all penetrations and a11 valves valves"pand valves which are hydrostaticaliy leak teste$ ~

." ,(except for main steam line isolation The measured leakage rate exceeding 11. 5 scf per hour sor any one main steam line isolation valve, or The measured'ombined leakage rate for ail ECCS and RCIC containment isolation valves in hydrostatically tes ed lines which penetrate the ~

primary containment exceeding 1 gpm times the -otal number of such valves, restore:

a. The overall integrated leakage rate(s), o less than or equal to 0.75 L,'nd

b. The combined leakage rate for all penetra-ions and ail valves,+H~&-

(

valves which are hydrostati cally leak ~'tesTed) subject to Type B and i". tests to less han or equal to O.oO L, and cxemot>on to Appendix .) o 10 CFR Pa,"t 50.

WASHINGTON NUCLEAR UNIT 2 3/4 6-2

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CONTAINMENT SYSTEMS SURVEILLANCE RE UIREMENTS (Continued

d. Type B and C tests shall be conducted with gas at P , 34.7 psig,* at no greater than 24""* months except for tests involving:

a'ntervals

1. Air Locks

2. Main steam line isolation valves,

3. Valves pressurized with fluid from a seal system, 4 ECCS and RCIC containment isolation valves in hydrostatically tested lines which penetrate the primary containment, and Purge supply and exhaust isolation valves with resilient seals.

Air locks shall be tested and demonstrated OPERABLE per Surveillance Requirement 4.6. 1.3.

Main steam line isolation valves shall be leak tested at least once per 18 months'.

Leakage from isolation valves that are sealed with fluid from a seal system may be excluded, subject to the provisions of Appendix J,Section III.C.3, when determining the combined leakage rate provided the seal system and valves are pressurized to at least 1.10 P ,

a' 38.2 sig and the seal system capacity is adequate to maintain system pressure for at leas 30 days.

h. ECCS and RCIC containment isolation valves in hydrostatically tested lines which penetrate the primary containment shall be leak tested at least once per 18 months.

Purge supply and exhaust isolation valves with resilient material seals shall be tested and demonstrated OPERABLE per Surveillance Requirements 4.6. 1.8.2 and 4.6. 1.8.3.

j. The provisions of Specification 4.0.2 are not applicable to'4-month or 40 + 10-month surveillance intervals.

qg,>V; )14<4'.

"Unless

"""For those tests conducted during refuel ng outages, the P4-month interval mav be exceeded by no more than 3 months.

WASHiNGTON NUCLEAR - UNIT 2 3/4 6-4 Amendment No. 4I

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CONTAINMENT SYSTEMS 3/4. 6.

~ ~ 3 PRIMARY CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION

%ch

3. 6. 3 -The-primary containment isolation valve/ and +he- reactor instrumentation line excess flow check valve( shall be OPERABLE.~

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

a. With one or more W~ primary containment isolation valves ~ewn-inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

1. Restore the inoperable valve(s) to OPERABLE status, or 2, 'Isolate each affected penetration by use of at least one de-activated automatic valve secured in the isolated position,"

or

3. Isolate each affected penetration by use of at least one closed manual valve or blind flange", and

4. The provisions of Specification 3.0.4 are not applicable pro-vided that the affected penetration is isolated in accordance with ACTION a.2. or a.3. above, and provided tha the associ-ated system, if applicable, is declared inoperable and the appropriate ACTION statements for that system are performed.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHVTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. With one or more w~e- reactor instrumentation line excess flow check valves . . inoperable, ooeration may continue and the provisions of Specifications 3. 0. 3 and 3. O. 4 are not applicable provided that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either;

1. The inoperable valve is returned to OPERABLE status, or

2. The instrument line is isolated and the associated instrument is declared inoperable'.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

"Isolation valves closed to satisfy these requirements may be reopened on an intermittent basis under adminis rative control.

WASHINGTON NUCLEAR - UNIT 2 3/4 6-19 Amendment No. 31

'S 4

4l All 4

CONTAINMENT SvST=MS SURVEILLANCE REOUIREMENTS 4.6.3.1 Each primary containment isolation valve shal 1 be demonstrated OPERABLE prior to returning the valve to service after mainte-nance, repair, or replacement work is performed on'he valve or its associated actuator, control or power circuit by cycling the valve through at least one complete cycle of full travel and verifying the specified isolation time.

4. 6. 3. 2 Each primary containment automatic isolation valve Mew~

at least once per 18 months by verifying that on a containment isolation test signal each automatic isolation valve actuates to its isolation position.

4,6.3.3 The isolation time of each primary containment power operated or utomatic valve shall be determined to be within its limit when tested pursuant to Specification 4.0.5.

4.6.3.4 Each reactor instrumentation line excess flow check valve M~~

shall be demonstrated OPERABLE at least once per 18 months by verifying that the valve checks flow at grea er than a 10 psid differential pressure in hydraulic service and l.5 psid differential pressure in pneumatic ser vice.

4.6.3.5 Each traversing in-core 'probe system explosive isolation valve shall be demonstrated OPERABLE:

a. At leas once per 31 days by verifying the continuity of the explosive charge.

b. At least once per l8 months by removing the explosive squib from at least one explosive valve, such that each explosive squib in each explosive valve will be tested at least once per 90 months, and initiating the explosive squib. The replacement charge for the exploded squib shall be from the same manufactured batch as the one fired or from another batch which has been certified by having at least one of that batch successfully fired. No explosive squib shall remain in use beyond the expiration of its shelf-life and operating life, as applicable.

WASHINGTON NUCLEAR - UNIT 2 3/4 6"20

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Page 3/4 6-21 through 3/4 6-33 DELETED

.Next Page is 3/4 6-34

1 TABLE 3. 6. 3" 1 PRIMARY CONTAINMENT ISOLATION VALVES 5

Cl HAXIHUH n ISOLATION TIME m VALVE FUNCTION AND NUMBER VALVE GROUP. a Seconds

Ã) a. Automatic Isolation Valves C:

Hain Steam Isolation Valves M

HS"V-22A,B,C,D(b)

HS-V-28A,B,C,D(b)

Hain Steam Line Drains HS-V-16 25 HS-V"19. 25 HS"V"67A,B,C,D(b) 15 Reactor Recirc. Cooling Sampl alves RRC-V-19 RRC-V-20 Containment Purg xhaust 8 n

CO Supply'EP-V-lA,,3A,4A ln CEP-V- ,2B,38,4B CSP- 1 4 CL C -V-2 4 SP-V-3 CSP-V-4 Cl W CSP-V-93 4 Al Xl

'D A CSP-V"96 4 Qp fP CSP-V-97 4 N0 rt CSP-V-98 4

~ C+

(D 5

~ CL gjp in cf in ro

i TABLE 3 '.3-1 (Continued PRIMARY CONTAINMENT ISOLATION VALVES MAXIM ISOLATI TIME VALVE FUNCTION AND NUMBER VALVE GROUP a) conds)

a. Automatic Isolation Valves (Continued)

Equipment Drain (Radioactive)

EDR-V-19 EDR-V-20 Floor Drain (Radioactive)

FDR-V"3 FDR-V-4 Fuel Pool Cooling/Suppression Pool Cleanup 35 FPC-V-149 FPC-V-153(f)

FPC-V-154(f)

FPC" V-156 Reactor Recirculation Hydraulic Con ol(e) 15 HY-V-17A,B HY-V-18A,B HY-V-19A,B HY-V-20A,B HY-V-33A,B HY-V-34A,B HY-V-35A,B HY-V-36A,B Traversi Incore Probe T -V-1,2,3,4,5 TIP-V-15

TABLE 3. 6. 3-1 Continued)

PRIHARY CONTAINMENT ISOLATION VALVES IHUM I ATION TINE VALVE FUNCTION AND NUMBER VALYE GROUP a) (Seconds

a. Automatic Isolation Valves (Continuedf Reactor Closed. Cooling 60 RCC"Y-5 RCC-V-21 RCC=V-40 RCC- Y-104 Radiation Monitoring Supply 8 Return PI-VX"250 PI-YX-251 PI-VX-253 PI-VX-256 P I"YX-257 P I-VX-259 Residual Heat Removal RHR-V"123A, B(g) 5 15 RHR- Y-8(g) (k) 6 40 RHR-V-9(g) 6 40 RHR-V-23(g) 6 90 RHR-V-53A, g) 6 40 RHR-V-2 ,8(c) 10 270 RHR-V 1 10 270 RH -27A,B(c) 10 36 eactor Hater Cleanup System RMCU-V"1(d) 30(i)

RMCU-V-4 21(j)

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TABLE 3. 6. 3-1 (Continued)

PRIHARY CONTAINMENT ISOLATION VALVES HAXI ISOLA N TIME VALVE FUNCTION AND NUMBER VALVE GROUP a econds)

a. Automatic Isolation Valves (Continued)

Reactor Core Isolation Cooling RCIC-V-8 13( j)

RCIC-V-63 l6(j)

RCIC-V-76 22 Low Pressure Core Spray LPCS-V-12 10 180 High Pressure Core Spray HPCS"V"23 180

b. Excess Flow Check Valves e Containment Atmosphere N.A.

PI" EFC" X29d P I-EFC-X29 f PI-EFC-X30a P I-EFC-N30 f PI-EFC-X4 PI" EFC- 42f PI- -X6lc

-EFC-N62b PI-EFC-N69f PI-EFC-X78a

TABLE 3.6.3-1 (Continued)

PRIMARY CONTAINMENT ISOLATION Vhl VES MAXIMUM SOLATION TIME VALVE FUNCTION ANO NUMBER VALVE GROUP(a (Seconds)

b. Excess Flow Check Valves (e) (Continued)

Containment Atmosphere (Continued) ' N.A.

I-EFC-X66 Pl-EFC"X67 P!-EFC-X82b Pl-EFC-XQ4a P I"EFC-X86A, 8 PI"EFC-X87A,B PI-EFC-X119 Reactor Pressure Vessel N.A.

PI "EFC-X18A, 8, C,O PI-EFC-X37e,f PI-EFC-X38a,b,c,d,e,f P I- EFC-X39a, b, d, e P I-EFC-X40c, d PI-EFC-X4lc, d PI-EFC-X42a,b PI-EFC-X44Aa,Ab,Ac d,he,hf,hg,hh,hj, Ak ,Am PI-EFC-X440a ,Bc,Bd,Be,Bf,Bg,Bh,Bj, Bk,81,8m PI-EFC- la,b P 1-E -X62c, d P FC-X69a,b,e I-EFC-X70a,b,c,d,e,f P I-EFC-X7la, b, c, d, e, f PI-EFC-X72a Pl-EFC-X73a PI-EFC-X74a,b,e,f

TABLE 3.6. 3-1 (Continued}

PRIHARY CONTAINHENT ISOLATION VALVES HAXIHUH C) ISOLATION TI VALVE FUNCTION ANO NUHBER VALVE GROUP a Secon n Excess Flow Check Valves (e) (Continued) m Reactor Pressure Vessel (Continued) N.A.

PI-EFC-X75a,b,c,d,e,f Pl-EFC"X70b,c,f PI-EFC-X79a,b PI-EFC-X106 PI-EFC" X107 P I-EFC-X108 PI-EFC-X109 PI-EFC-X110 PI-EFC-X111 PI-EFC-X112 PI-EFC-X113 Pl-EFC-X114 PI-EFC-X115 Other N.A.

PI-EFC-X40e, f PI-EFC-X4le, f C. Hanual Containmen sol ation Valves Oemineraliz Mater N.A.

OM- 156

-V-157 O Containment Air System N~ A.

CAS-VX-02e CAS-V-730

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TABLE 3.6.3-1 (Continued PRIHARY CONTAINHENT ISOLATION VALVES M

HAXIHUH C) I SOLAT I ON HE VALVE FUNCTION AND NUHBER VALVE GROUP a (Se nds)

C C)

I m C. Hanua'l Containment Isolat.ion Valves (Continued)

Service Air N.A.

SA-V-109 Residual Neat. Removal N.A.

RNR-V-llA,B RNR-V-12D RNR-V-121 RNR-V-124A,B Rl<R-V-125A, 0

~4 Reactor Core Isolation Cooling ll N.A.

RCIC-V-64 RCIC-V- f42(g) (b)

Air Supply to Test. able Check Va s N.A.

A~iir Su 1 Check ve P I-VX-42d Rl -50A P I-VX-216 P I-VX-69c RllR-V-500 PI-VX-2 PI- -61f RIIR-V-41A

-VX-219 P I-VX-540f RllR-V-410.

Pl-VX-218 O

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TABLE 3. 6. 3-1 (Continued PRIHARY CON1AINHENT ISOLATION VALVES HAKIHUH ISOLATION Tl VALVE FUNCTION AND NUHBER VALVE GROUP(a (Secon

c. Hanual Containment Isolation Valves (Continued)

Air Supply to Testable Check Valves (Continued) N.A.

PI-VX-62 f RHR-V-41C PI"VX-220 LPCS-V-66 LPCS-V-6 LPCS-V-67 HPCS-V-65 HPCS-V-5 HPCS-V-68 RCIC-V-184 RCIC-V-66 RCIC-V-740

d. Other Containment Isolation Valves Hain Steam Leakage Control(b) N.A.

HSLC-V"3A,B,C,D Reactor Feedwater/RWCU Return N.A, RFM-V-lOA,B RFM-V-32A,G RFM-V-65A,D RWCU-V-40 High Pressur ore Spray N.A.

HPCS-V- g)(b)

~iIC -5(g)(b)

S-V-12

)PCS-V-15(f)(b)

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TABLE 3.6. 3-1 (Continued PRIMARY COHTAIHMEHT ISOLATION VALVES IMUH I LATION TIME VALVE fUHCTION ANO NUMBER VAI VE GROUP(a (Seconds)

d. Other Containment Isolation Valves (Continued) bligh Pressure Core Spray (Continued) N. A.

HPCS-RV-14(e)(h)

HPCS-RV-35(e)(h)

Low Pressure Core Spray N.A.

LPCS-V-l(f)(b)

LPCS-V-5(g)(b)

LPCS-V-6(g)(b)

LPCS-RV-18(e) (h)

LPCS- RV-31(e) (h)

LPCS-FCV-ll Standby Liquid Control N~ A.

SLC-V-.7 SLC-V-IA,O Reactor Core Isolat' Cooling H.A.

RCIC-V"13( )

RCIC-V-1 RCIC- 8 RC V-31(f)(b)

IC-V"40 RCIC-V-66(g)(t )

RCIC-V-6B O RCIC-V-69 PO

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TMLE 3.6. -1 (~Conl.inued PRIMARY CONTAINMENT ISOLATION VALVES MAXIMUM ISOLATION TIH VALVE FIINCTION AND NUtEDER VALVE GROuP a Seconds

d. Ottier Containment Isolation Valves (Continued)

Residual Heat Removal/Low Pressure N.A.

In,jection REER-V-4A,D,C(f)(b)

RHR-V-16A,D REER-V-17A,D RHR-V-41A, 8(g) (b)

RHR-V"42A,B,C(g)(b)

REIR-V"50A,D(g)(b)

RHR-V-73A,B RHR-V-134A, D(c)

RIIR-V-209(g)(b)

RHR-RV- lA,D(e) (h)

RtlR-RV-5(e)(h)

REER-RV-25A,D,C(e)(h)

RtlR-RV-30(e)(h)

RHR-RV-36(e)(h)

REIR-RV-88A,D,C(e)(h)

RHR-FCV-64A,B,C Containment Atmospher ontrol(c)(i) N.A.

(H2 Recombiner)

CAC-V"2 CAC-FCV- A,B CAC- 15

-FCV" lA,8 CAC-V-11

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TABLE 3. 6. 3-1 Continued)

PRIMARY CONTAINMENT ISOLATION VALYFS MAXIH ISOLA H TINE VALVE FUNCTION AHD NUllBER VALVE GROUP(a). econds)

d. Other Containment Isolation Valves (Conti<<ued)

Containment Atmosphere Control(c)(i) H.A.

(Il2 Recombiner) (Continued)

CAC-V-6 CAC-V-4 CAC-FCV-4A,B CAC-V-13 CAC-V-17 CAC-FCV-3A,B'AC-V-8 CSP-V-5 CSP-Y-6 CSP-V-7 Containment Purge System N.A.

CSP-V-8 CSP-V-9 CSP-V-10 Reactor Recircula (Seal Injection) H.A.

RRC-V-3.3A RRC-V-1 ,B Con i<<ment Instrument Air N.A.

CIA-V-20 CIA-V-21

TABLE 3.6. -1 Con~tinued PRIMARY CONTAINMENT ISOLATION VALVES MAXIHUH I SOLAT IO I ME VALVE FUNCTION ANO NUMBER VALVE GROUP a Se ds

d. Other Containment Isolation Valves (Continued)

Containment Instrument Air (Continued) N.A.

CIA-V-30A,B CIA-Y-31A,B Post-Accident Sampling System(c) N.A.

P SR" V-X73-1 PSR- Y-X73-2 PSR-Y-X77Al PSR-V-X77A2 PSR-V-X77A3 PSR-V-X77AO PSR-V-X80-1 PSR-V-X80-2 PSR-V-X82-1

'SR-V-X02-2 PSR-V-X82-7 PSR-V-X02-8 PSR"V-X83-1 PSR-V-X83-2 PSR-Y-X84-1 PSR-V-X04-PSR-V- 8-1 PS -X88-2

TABLE 3.6. 3-1 Continued M PRIMARY CONTAINMENT ISOLATION VALVES F)

C)

MAXI ISOL ON TIME

?: VALVE FUNCTION AND NUMBER VALVE GROUP a econds

d. Other Containment Isolation Valves (Continued)

Radiation Monitoring N.A.

PI-EFCX-72 I-EFCX-73e Transversing Incore Probe System N.A.

TIP-V-6 TIP-V-7,8,9,10,11(e)

LE NOTATIONS "But greater than 3 seconds.

NProvisions of Technical Specification 3 .4 are not applicable.

(a) See Technical Speci fication 3 .2 for the isolation signal(s) which operate each group.

(b) Valve leakage not included n sum of Type 8 and C tests.

(c) Hay be opened on an int mittent basis under administrative control.

(d) Not closed by SLC ac ation signal.

(e) Not subject to Ty C Leak Rate Test.

(f) Hydraulic leak est at 38.2 psig.

(g) Not subject o Type C test. Test per Technical Specification 4.4.3.2.2 (h) Tested a part of Type A test.

(i) Hay b ested as part of Type A test. If so tested, Type C test results may be excluded from sum of ott Type B and C tests.

O (j) eflects closure times for containment isolation only.

( During operational- conditions 1, 2 8 3 the requirement for automatic isolation does not apply to RUR-V-8.

Except that RIIR-V-8 may be opened in operational conditions 2 8 3 provided control is returned to the control room, with the interlocks reestablished, and reactor pressure is less than 135 psig.

we I

ELECTRICAL POWER SYSTEMS 3/4.8.4 ELECTRICAL E UIPMEHT PROTECTIVE DEVICES A.C. CIRCUITS IHSIDE PRIMARY CONTAINMEHT LIMITIHG CONDITION FOR OPERATION 3.8.4.1 At least the following A.C. circuits inside primary contain ent shall be deenergized":

a. Circuits supplied by breakers 2AR and 8AR, MCC E"MC-8C

b. Circuits supplied by panel E-LP-SBAG.

c. Circuits supplied by panel E-LP-3DAG.

d. Circuits supplied by breakers in cubicles 2BL 10, and 2CR of 9C-3DA.

APPLICABILITY: OPERATIONAL COHDITIOHS 1, 2, and 3.

ACi10N:

Vith any of the above required circuits energi d, trip the associated circuit .

breaker(s} in the specified panel(s} within 1 our.

SUP VEI LLAHCE REQUIREMENTS 4.8. 4. l. Each of the above required . C. circuits shall be determined to be deenergized at least once per 24 ho rs"" by verifying that the associated circui breakers are in the tripp condition.

=he drywell.

--.": a. !oast once 3} days if ]o-ked, sealed, pl p +F '~'. <<e ~ I>>eC in

~r ~ Onc't'eon

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CONTROL1ED COPY ELECTRICAL POWER SYSTEMS PRIMARY CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES LIMITING CONDITION FOR OPERATION 3.8.4.2 All primary containment penetration conductor overcurrent prot ctive devices shown in Table 3.8.4.2-1 shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION'.

With one or more of the primary containment penetrati conductor overcurrent protective devices shown in Table 3.8.4. -1 inoperable, declare the affected system or component inoperabl and apply the appropriate ACTION statement for the affected sys em and:

1. For 6.9 kV circuit breakers, de-energize t e 6.9 kV circuit(s) by tripping the associated redundant cir it breaker(s) within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and verify the redundant circu breaker to be tripped at least once per 7 days thereafter.

2. For 480 volt circuit breakers, remov the inoperable circuit breaker(s) from service by removin the fuses within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and verify the fuses associated w th the inoperable breaker(s) to be removed at least once per days thereafter.

Otherwise, be in at least HOT SHUT WN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the folio ng 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. The provisions of Specification .0.4 are not applicable to over current devices in 6.9 kV cir its which have their redundant circuit breakers tripped or to 480 v t circuits which have the fuses asso-ciated with the inoperable ircuit breaker removed.

SURVEILLANCE RE UIREMENTS 4.8.4.2 Each of the primary con ainment penetration conductor overcurrent protective devices shown in Tab e 3.8.4.2-1 shall be demonstrated OPERABLE:

a. At least once per months:

1. By verifyin that the mediumvoltage, 6.9 kV, circuit breakers are OPERAB by selecting, on a rotating basis, at least 10K of the circu't breakers of each voltage level and performing:

a) A HANNEL CALIBRATION of the associated protective relays, d

b) An integrated system functional test which includes simulated automatic actuation of the system and verifying that each relay and associated circuit breakers and overcurrent control circuits function as designed.

.) For each circuit breaker found inoperable during these functional tests, an additional representative sample of at least 10~ of all the circuit breakers of the inoperable

<'v type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested.

SHINGTON NUCLEAR - UNIT 2 3/4 8-2'

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ELECTRICAL POMER SYSTEMS SURVEILLANCE RE UIREMENTS Continued

2. By selecting and functionally testing a representativ sample of at least lOX of each type of lower voltage circui breakers.

Circuit breakers selected for functional testing s ll be selected on a rotating basis. Testing of these c cuit breakers shall consist of injecting a current with a valu equal to 300K of the pickup of the longtime delay trip eleme and 150K of the pickup of the short time delay trip eleme , and verifying that the circuit breaker operates within the ime delay band-width for that current specified by the ma facturer. The instantaneous element shall be tested by 'ecting a current equal to +2(C of the pickup value of the element and verifying that the circuit breaker trips instant eously with no inten-tional time delay. Molded case circu breaker testing shall also follow this procedure except th generally no more than

. two trip elements, time delay and i stantaneous, will be involved. Circuit breakers found noperable during functional testing shall be restored to OPE BLE status prior to resuming operation. For each circgit br aker, found inoperable during these functional tests, an ad tional representative sample of at least 10K of all the circ t breakers of the inoperable type shall also be functionally ested until no more failures are found or all circuit brea rs of that type have been.

functionally tested.

b. At least once per 60 months by subjecting each circuit breaker to an

.inspection and preventive aintenance in accordance with procedures prepared in conjunction ith its manufacturer s recommendations.

HINGTON NUCLEAR " UNIT 2 3/4 8-22 Amendment No. 42

CONTROLl.ED COPY TABLE 3.8.4.2-1 PRIMARY CONTAINMENT PENETRATION CONDUCTOR OV RCURR N R C IV D V CES

~EUIPMENT PRIMARY PROTECTION BACKUP PROTECTION

a. 6900V Circuit Breakers RRC-P-1A E-CB-RRA (Relay) E-CB-S5 {Relay) E-CB- /5 (Relay)

RRC-P-18 E-CB-RRB (Relay E-CB-S6 (Relay) E-CB N2/6 (Relay)

b. 480VAC Fused Disconnects MS-V-16 MC-88"A Fused MC-88 Fuse RWCU"V"1 MC-88-A Fused MC-88 Fus d RHR-V-9 MC"88-A Fused MC-SB F ed RCIC"V-63 MC-88-A Fused MC-SB . sed RCC-V-40 MC-SB-A Fused MC-SB Fused RHR-V- 1238 .,MC-88-A Fused MC-SB Fused RCIC-V"76 MC-SB-A Fused MC-88 Fused RHR-V-123A MC-SB"A Fused MC-8 Fused WA INGTON NUCLEAR - UNIT 2 3/4 8-23 Amendment No. 72

THIS PAGE INTENTIONALLY LEFT BLANK MA INGTON NUCLEAR - UNIT 2 3/4 8"24 Amendment No. 42

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P ""TR <AL POWER 5Y~i.5 ROTOR-OPERATE VALV 5 THERMAL OV RLQAQ PRO~>, GH LIMING CQHQITIOH FOR OPERA ON 3.9.4.3 The 'Newel over lead prctac ion of each valve shayn fn Tab 3.8.4.3-'hall be OPERABLE.

APPLICASI~ ~i: 'whenever "-5e actor operand valve is rs~uf~ "e OP ~BL=.

@COLON:

Mf 5 ~"le >areal overload protection fcr cne cr mare of th above recuired valves inoperable, continuously bypass the fnoperable the . 1 overload wfDfn S hours or declare the af'f'ected valve(s) inoperable and ply Ne accrcpriata ACTIQH statanent(s) for Ne af aced sys~(s).

5URVEILLAHCE REGUIR fBPi5 4.3.4,3 The -Nerval overload prctac ion,c Ne above required valves shall be daaans rated QP RABL a leas once per 8 eon~as ano cl!cuing .zfntanance cn -ae aatcr starter by "'le per~c~nce a CFAHHEL CALIBRAT.QH c~ a represan" itive sanple o at leas 2Q o all =a l overloads fcr ~ze above recufW valves.

'.<<A".H GiQH HUC~R - UHIT

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TABLE 3.8.4.3"1 MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION SYSTEM(S) SYSTEM(S}

VALVE NUMBER AFFECTED VALVE NUMBER AFFECTED a0 CAC-V"2. Containment HSLC" V-1A Main Stea CAC-V"4 Atmospheric HS LC-V-1B Isolatio Valve CAC"V-6 Control System MS LC" Y-lc Leakag Control CAC-V"8 MSLC-V"1D "

Syste CAC-V-ll MSLC"V-2A CAC-V"13 MSLC-V-2B CAC"V"15 HSLC-V-2C CAC-V-17 HSLC-Y-2D MSLC"V-3A CIA"V"20 Containment MSLC"V-3B CIA-V-30A Instrument Air MSLC-V"3C CIA-V"30B System HSLC-Y"3D MSLC"V-4 C. F PC-V-149 Fuel Pool Cooling MSLC-V-FPC"V-153 System HSLC" 9 FPC-V"154 MSLC -10 FPC-V-156 FPC-V-172 FPC-V"173 FPC-Y-175 FPC-V" 181A F PC" Y-181B F PC-V" 184

d. HPCS" Y-1 High Pressure Co e h. RCC-V-5 Reactor Closed HPCS-V-4 Spray System RCC-Y-21 Cooling Mater HPCS-V-10 RCC-V-40 System HPCS"V"ll RCC"Y-104 HPCS-V-12 RCC-V-129 HPCS-V"15 RCC"V-130 HPCS"V-23 RCC-V-131

e. LPCS-V" 1 Low P essure Core i. RCIC"V-1 Reactor Core LPCS"Y-5 Spr System RCI C" Y-8 Isolation Cooling LPCS-FCV-ll RCIC"V-10 System LPCS-Y-12 RC I C" V-13 RCIC"Y-19 MS"V"1 Main Steam System RCIC"Y-22 MS-V-2 RCI C" V"31 MS-Y"5 MS-V"16 MS-Y-19 MS-V-20 MS-V- A MS"V 7B HS" -67C M V-67D S-V-146 W HINGTON NUCLEAR " UNIT 2 3/4 8-26 Amendment No. 27

'I CONTROLl.ED COPY TABLE 3.8.4.3"1 (Continued)

MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION SYSTEM(S) SYSTEM(

VALVE NUMBER AFFECTED VALVE NUMBER AFFECT RCI C" V-45 Reactor Core, RHR-Y-42C RCIC" V"46 Isolation Cooling RHR"V-47A RCIC-V-59 System RHR-V-478 RCIC-V-63 RHR-Y-48A RCIC-Y-68 RHR-V-488 RCIC-V-69 RHR-V-49 RCIC-V-76 RHR-V-53A RCIC-Y-110 RHR"Y-538 RCIC-V"113 RHR-V-64A RHR-V"64 RHR-V-6 C RFM"V-65A Reactor Feedwater RHR-V- 8A RFM"V-658 System RHR" 688 RHR -73A RH -Y-738

k. RHR-V-3A Residual Heat R-V-74A RHR-Y-38 Removal System HR-V-748 RHR"V"4A RHR-V-115 RHR-V"48 RHR-V-116 RHR-Y-4C RHR-V-123A RHR"V-GA RHR"V-1238 RHR-V-68 RHR-V-134A RHR-V-8 RHR-Y-1348 RHR-Y-9 RHR-V-16A RRC-V-16A Reactor Recirculation RHR-V-168 RRC-V-168 System RHR" V-17A RHR" V-178 RHR-V-21 RHR-V-23 RMCU-V-1 Reactor Mater RHR-V-24A RMCU-Y"4 Cleanup System RHR-V-248 RMCU-V-40 RHR-V-27A RHR-V-278 RHR-V-40 RHR-V"42A RHR-V"428 WASH GTON NUCLEAR

- UNIT 2 3/4 8-27 Amendment No. 27

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TABLE 3.8.4.3-1 (Continued)

MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION SYSTEM(S) SIST+(S)

VALVE NUMBER AFFECTED VALVE NUMBER AFFECTED

n. SGT-V-1A Standby Gas o. AS-V-68A Auxili y Steam SGT"Y-1B Treatment System AS-V-68B Syste SGT-V"3A1 SGT-V-3A2 p. SM-V-2A St dby Service SGT-V-3Bl SM-V-2B M ter System SGT-Y-3B2 SM"Y-4A SGT-V-4Al SM-Y-4B SGT-V-4A2 SM-V-4C SGT-V-4Bl SM-V-12A SGT"V-4B2 SM" V-12B SGT-V-5A1 SM-V-24A SGT"V"5A2 SM"V-24

'SGT" V-5B1 SM-Y-2 C SGT-V-5B2 SM-Y" 9 SM- 44 SM -54

-V-75A M-V-75B SM-V-90 SM-V-187A SM"V-187B SM-V-188A SM"V-188B

" UNIT 2 3/4 8"28 Amendment No. 27 Mi HINGTON NUCLEAR I

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CONTROLLED COPY ELECTRICAL POWER SYSTEMS REACTOR PROTECTION SYSTEM ELECTRIC POWER MONITORING LIMITING CONDITION FOR OPERATION 3.8.4.4 Two RPS electric power monitoring channels for each inservice RPS MG set or alternate source shall be OPERABLE.

APPLICABILITY: At all times.

ACTION:

With one RPS electric power monitoring channel for an inservice RPS MG set or alternate power supply inoperable, restore the inoperable power monitoring channel to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or remove the associated RPS MG set or alternate power supply from service.

With both RPS electric power monitoring channels for an inservice RPS MG set or alternate power supply inoperable, restore at least one electric power monitoring channel to OPERABLE stat~s within 30 minutes or remove the associated RPS MG set or alternate power supply from service.

SURVEILLANCE RE UIREMENTS 4.8.4.4 The above specified RPS power monitoring channels instrumentation shall be determined OPERABLE:

a. By performance of a CHANNEL FUNCTIONAL TEST each time the plant is in COLD SHUTDOWN for a period of more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , unless performed within the previous 6 months.

b. At least once per 18 months by demonstrating the OPERABILITY of overvoltage, undervoltage and underfrequency protective instru" mentation by performance of a CHANNEL CALIBRATION including simulated automatic actuation of the protective relays, tripping logic and output circuit breakers and verifying the following setpoints.

l. Overvoltage < 132 VAC,

2. Undervoltage > 108 VAC, and

3. Underfrequency > 57 Hz.

WASHINGTON NUCLEAR " UNIT 2 3l4 8"P

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X.

CONTAINMENT SYSTEMS BASES DEPRESSURIZATION SYSTEMS (Continued)

~k~)o c~~~

Should it be necessary to make the suppression chamber inoperable, this shall only be done as specified in Specification 3.5.3.

Under full power operating conditions, blowdown from an initial suppres-sion chamb w temperature of'90 F results in a water temperature of approxsmatel 135'F immediately following blowdown which is below the 200'F used for comp e e condensation via quencher devices. At this temperature and

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atmospheric pressure, the available NPSH exceeds that required by both the RHR and core spray pumps, thus, there is no dependency on containment overpressure during the accident injection phase. If both RHR loops are used for contain-ment cooling, there is no dependency on con ainment overpressure for post"LOCA operations.

Experimental data indicate that excessive steam condensing loads can be avoided if the peak bulk temperature of the suppression pool is maintained below 200'F during any period of relief valve operation with sonic conditions at the discharge exit for quencher devices. Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to avoid the regime of potentially high suppression chamber loadings.

Because of the large volume and thermal capacity of the suppression pool, t

the volume and temperature normally changes very slowly and monitoring these arameters daily is sufficient to establish any temperature trends. Ey requiring the suppression pool temperature to. be frequently recorded during periods of significant heat addition, the temperature trends will be closely

~

followed so that appropriate action can be taken. The requirement for an external visual examination following any event where potentially high loadings could occur provides assurance that no significant damage was encountered.

In addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event a safety/relief valve inadvertently opens or sticks open. As a minimum this action shall include: (1) use of all available means to close the valve, (2) initiate suppression pool water cooling, (3) initiate reactor shutdown, and (4) if other safety/relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open safety/relief valve to assure mixing and 'uniformity of energy insertion to the pool.

3/4. 6. 3 PRIMARY CONTAINMENT ISOLATION VALVES The OPERABILITY of the primary containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmospheie or pressurization of the containment. Containment iso1ation within ahe~ime

.limits spec-if'+ ensures for those isolation valves designed to close auto-matically that the release of radioactive material to the environment will

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e consistent with the assumptions used in the analyses for a LOCA.

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WASHINGTON NUCLEAR - UNIT 2 B .3/4 6-4 Amendment No. 85, 100

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CMTAINMENT SYSTEMS . CONTROLl.ED COPY BASES 3/4;6.4 VACUUM RELIEF Vacuum relief breakers are provided to equalize the pressure between the suppression chamber and drywell and between the reactor building and suppres-sion chamber. This system will maintain the structural integrity of the Bm primary containment under conditions of large differential pressures.

The vacuum breakers between the suppression chamber and the drywell must not be inoperable in the open position since this would allow bypassing of the suppression pool in case of an accident. There are nine pairs'of valves to provide redundancy and capacity so that operation may continue indefinitely with no more than two pairs of vacuum breakers inoperable in the closed position.

3/4.6.5 SECONDARY CONTAINMENT Secondary containment is designed to minimize any ground level release of

'adioactive material which may result from an accident. The reactor building and associated structures provide secondary containment during normal opera-tion when the drywell is sealed and in service. At other times the drywell may be open and, when required, secondary containment integrity is specified.

Establishing and maintaining a vacuum in the reactor building with the standby gas treatment system once per 18 months, along with the surveillance of the doors, hatches, dampers, and valves, is adequate to ensure that there are no violations of the integrity of the secondary containment.

The OPERABILITY of the standby gas treatment systems ensures that suf-

~

ficient iodine removal capability will be available in the event of a LOCA.

The reduction in containment iodine inventory reduces the resulting SITE BOUNDARY radiation doses associated with containment leakage. The operation of this system and resultant iodine removal capacity are consistent with the assumptions used in the LOCA analyses. Continuous operation of the system with the heaters OPERABLE for l0 hours during each 31 day period is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters.

3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL The OPERABILITY of the systems required for the detection and control of hydrogen gas ensures that these systems will be available to maintain the hydrogen concentration within the primary containment below its flammable limit during post-LOCA conditions. Either drywell and suppression chamber hydrogen recombiner system is capable of controlling the expected hydrogen generation associated with (1) zirconium-water reactions, (2) radiolytic decomposition of water, and (3) corrosion of metals within containment. The hydrogen control system is consistent with the recommendations of Regulatory Guide l.7, "Control of Combustible Gas Concentrations in Containment Following a LOCA," September 1976 T'ai> sec 7iotl sk tl Le ~~ver ++ a. tlecu P~ga. 5 >// 8-g

<< ~A~ revePse. s'le oF this j'4g~

WASHINGTON NUCLEAR - UNIT 2 B 3/4 6"5 Amendment No. 100

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PRIMARY CONTAINMENTISOLATION VALVES (Continued)

The opening of locked or sealed closed (i.e. manual) containment isolation valves on an intermittent basis under administrative control includes the following considerations: (I) stationing an operator, who is in constant communication with the control room, at the valves, (2) instructing this operator to close these valves in an accident situation, and (3) assuring that environmental conditions will not preclude access to close the valves and that this action will prevent the release of radioactivity outside the containment.

Editing Note: The above information shall be inserted at the top of page 8 3/4 6-5.

A r o 0

, ELECTRICAL POWER SYSTEMS BASES A.C. SOURCES D.C. SOURCES and OHSITE POWER DISTRIBUTION SYSTEMS (Continue

'Operation with a battery cell's parameter outside the normal limit but within the allowable value specified in Table 4.8.'2.1-1 is permitted for up to 7 days. During this 7-day period: (1) the allowable values for electrolyte level ensures no physical damage to the plates with an adequate electron transfer capability; (2) the allowable value for the average specific gravity of all the cells, not more than 0.020 below the manufacturer's recommended full charge specific gravity ensures that the decrease in rating will be less than the safety margin provided in sizing; (3) the allowable value for an individual cell's specific gravity, ensures that an individual cell's specific gravity will not be more than 0. 040 below the manufacturer's full charge specific gravity and that the overall capability of the battery will be maintained within an acceptable l.imit; and (4) the allowable value for an individual'cell's float voltage, greater than 2.07 volts, ensures the battery's capability to perform its design function.

3/4.8.4 ELECTRICAL E UIPMEHT PROTECTIVE DEVICES Primary con ainment electraca penetrations a d penetration co uctors are protected by either deenergiz'ng circuits not required during r actor operation or de onstrating the 0 RABILITY of pr mary and backup ercurrent protection cir uit breakers by riodic surveil ance.

The surv illance requirem nts applicable o lower voltage ircuit breakers provide assu ance of breaker eliability by sting at least e representativ sample of ea h manufactur ers rand of circui breaker. Each anufacturer's molded case and metal case rcuit breakers are grouped int representative samples wh ch are then tes d on a rotatin basis to ensur that all breake s are tested If a wide var ety exists wit in any manufact er's brand of c r-cuit brea ers, it is nece sary to divide that manufactur 's breakers int groups a d treat each gr up as a separa e type of break for sutveilla e purposes T e bypassing of he motor-oper ed valve therm overload prot ction contin ously or durin accident cond tions ensures at the thermal verload prote tion will not revent safety- elated valves rom performing eir junc on. The surv illance requir ments for demo strating the b ssing of the hermal overlo protection c ntinuously and during accident conditions are in accordance ith Regulator Guide 1.106 " ermal Overload rotection for fl tric Motors o Motor Operat Valves," Rev sion 1, Harch 1 7.

the motor-generator set The RPS electric power monitoring system isolates the RPS bus from This or alternate power source in the event of overvoltage, undervoltage, or underfrequency.

voltage or frequency conditions.

system protects the RPS components against unacceptable Isolation of the RPS power supplies is the fail-safe condition.

WASHINGTON NUCLEAR

- UNIT 2 B 3/4 8"3 Amendment Hn.

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