ML20070F114

Amend 135 to License DPR-35,changing Surveillance Requirements for Redundant Core & CCS & Allowing out-of-svc Period for CCS & LPCI Pumps

ML20070F114
Person / Time
Site:	Pilgrim
Issue date:	03/04/1991
From:	Shankman S Office of Nuclear Reactor Regulation
To:
Shared Package
ML20070F116	List: ML20070F114 ML20070F122
References
NUDOCS 9103080117
Download: ML20070F114 (26)

Text

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UNITED sT ATEs NUCLEAR REGULATORY COMMISSION e

W ASHING T ON. D. C. 20bb6 BOSTON E0150t! COMPANY DOCKET NO. 50-?93 PILGRIM NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Anendment No.135 License No. DPR 35 1.

The Nuclear Regulatory Comission (the Comission or the NRC)-has found that:

A.

The application for amendment filed by the Boston Edison Company (the licensee) dated March 15, 1990 complies with the standards and requirements of the Atomic Energy Act of 1954,-as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter 1; B.

The facility will o)erate in conformity with the application, the provisions of tie Act, and the rules'and regulations of the Comission!

C.

There is reasonable assurance: (i)thattheactivitiesauthor-ized by this amendment can be conducted without endangering the i

health and-safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regula-tions set forth in 10 CFR Chapter 1;

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D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this emendment is in accordance with 10 CFR Part 51 of the Comission's regulations.and all applicable requirements have been satisfied.

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2.

Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and partgraph 3.B of f acility Operating License No. DPR 35 is hereby amended to read as follows:

Technical Sjecifications The Technical Specifications contained in Appendix A, as revised through Amendnint No.135, are hereby incorporated in the license.

The_ licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendr.tu is effective as of its date of issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY CO Wi$510tl flb.n bl%sb% awn-susan F. Shankman, Acting Director Project Directorate 1-3 Divition of Reactor Projects - 1/11 Office of Nuclear Teactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

March 4,.1991 v--e,-,

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ATTACHMENT TO LICENSE AMEtitPENT NO. 135 FACILITY OPERATINO L1Ctl4SE NO. DPR-35 DOCKET NO. 50-293 Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

Remove Insert i

i 100 100 103 103 104-104 105 105 106 106 107 107 108 100 109 109 110 110 111 111 112 112 113 113 114 114 115 115 116 116 117 117 118 118 121 121 122 122 148 148 201 201

1ABLE OF CONTEN M ggg h 1.0 DEFINITIONS 1

2.0 SAFETY LIMITS 2.1 Safety Limits 6

2.2 Safety Limit Violation 6

Limitina Conditions For Ooeratio.n Survelliance Reautrement 3.1 REACTOR PROTECTION SYSTEM 4.1 26 3.2 PROTECTIVE INSTRUMENTATION 4.2 42 3.3 REACTIVITY CONTROL 4.3 80 A. Reactivity Limitations A

B0 B. Control Rods B

81 C. Scram Insertion Times C

83 D. Control Rod Accumulators 0

84 E. Reactivity Anomalies E

85-F. Alternate Requirements

'85 G. Scram Discharge Volume G

85 3.4 STANDBY LIQUID CONTROL SYSTEM 4.4 95 A. Normal System Availability A

95 B. Operation with Inoperable Components B

96 C. Sodium Pentaborate Solution C

97 D. Alternate Requirements 97 3.5 CORE AND CONTAINHENT COOLING SYSTEMS 4.5 103 A. Core Spray and LPCI Subsystems A

103 B. Containment Cooling Subsystem B

106 C. HPCI Subsystem C

107 D. RCIC Subsystem 0

108 E. Automatic Depressurization System E

109 F. Minimum Low Pressure Cooling System F

110 and Diesel Generator Availability G. (Deleted)

G 111 H. Maintenance of filled Discharge Pipe-H 111 l

3.6 PRIMARY SYSTEM BOUNDARY 4.6 123

-A. Thermal and Pressurization Limitations A

123 B. Coolant Chemistry B

124-C. Coolant Leakage C

125 D. Safety and Relief Valves D

126 E. Jet Pumps E

127 F. Jet Pump Flow Hismatch F

127a G. Structural Integrity G

127a H. Deleted H

I. Shock Suppressors (Snubbers)

I 137a Amendment No. 15, 45, 65, 133, 135 j

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-3.4 & 4.4 STANDBY L10UID CONTROL SYSTEN A.

The requirements for SLC capability to shutdown the reactor are identified via the station Nuclear Safety Operational Analysis d

(Appendix G to the FSAR, Special Event 45).

If no more than one operable control rod is withdrawn, the basic shutdown reactivity requirement for the core is satisfied and the Standby Liquid i

Control system is not required. Thus, the basic reactivity requirement for the core is the primary determinant of whe7 the

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standby liquid control system is required.

The design ob.iective of i

L the standby liquid control system is to provide the capabllity of bringing the reactor from full power to a cold, xenon-free shutdown condition assuming that none of the withdrawn control rods can be 4

inserted. To meet this objective, the Standby Liquid Control system is designed to inject a quantity of boron that produces a minimum concentration equivalent to 675 ppm of natural boron in the i

i reactor core. The 675 ppm equivalent concentration in the reactor core is required to bring the reactor from full power to at least a three percent Ak subtritical condition, considering the hot to cold reactivity difference, xenon poisoning etc.

The system will inject this boron solution in less than 125 minutes.

The maximum time requirement for inserting the boron solution was selected to override the rate of reactivity insertion caused by cooldown of the reactor following the renon poison peak.

The Standby Liquid Control system is also required to meet 10CFR50.62 (Requirements for Reduction of Risk from Anticipattd i '

-Transients Hithout Scram (ATHS) Events for Light-Hater-Cooled

' Nuclear Power Plants). The Standby Liquid Control system must have theequivalentcontrolcapacity(injectionrate)of86gpmat13 percent by wt -natural sodium pentaborate for a 751" diameter reactor pressure vessel in order to satisfy 10CFR50.62 requirements. This equivalency requirement is fulfilled by a combination of concentration, BM enrichment and flow rate of 4

sodium pentaborate solution. A minimum 8.42% concentration and 54.5% enrichment of B4 isotope at a 39 GPH pump flow rate satisfies the ATHS Rule (10CFR50.62) equivalency requirement.

Becausetheconbentration/volumecurvehasbeenrevisedtoreflect i

the increased B isotopic enrichment, an additional requirement has been added to evaluate the solution's capability to meet the original design' shutdown criteria whenever the BM enrichment requirement is not met.

Experience with pump operability indicates that the monthly test, in combination with the; tests during each operating cycle, is sufficient to maintain pump performance.

The only practical'. time to' fully test the liquid control system is during a refueling outage.- Various components of the system are individually tested periodically, thus making more frequent testing of-the entire system unnecessary.-

Amendment No in?,135 100 L

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LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT l

3.5 CORK AND CONTAINMENT C00L MG 4.5 CORE AND CONTAINMENT COOLING SYS" EMS SYSTEMS Anolicability Annlicability Applies to the operational status of Applies to the Surveillance the core and suppression pool cooling Requirements of the core and l systems, suppression pool cooling systems which l

)

are required when the corresponding Limiting Condition for operation is in etftct.

ll Obinettve Obinettve To assure the operability of the core l and suppression pool cooling systems To verify the operability of the core and suppression pool cooling systems l

under all conditions for which this under all conditions for wh'ch this cooling capability is an essential cooling capability is an essential response to station abnormalities.

response to station abnormalities.

Snecification l A.

Core Sorav and LPCI Systems A.

Core Sorav and LPCI Systemt l

l1.

Both core spray systems shall be operable whenever irradiated fuel 1.

Core Spray System Testing, l

is in the vessel and prior to reactor startup from a Cold Ltta Er.tutEy Condition, except as specified in 3.5.A.2 below, a.

Simulated Once/ Operating Automatic Cycle Actuation test, b.

Pump Operability Once/ month c.

Motor Operated Once/ month Valve Operability d.

Pump flow rate Once/3 months Each pump shall deliver at least 3300 gpm against a system head corresponding to a reactor vessel pressure of 104 psig.

e.

Core Spray Header 6pInstrumentation Amendment No. 43, 62. 114, fal, 135 103

SURVEILLANCE REOUIREMENT LIMITING CONDITION FOR OPERATION l

3.5.A Core Sorav and LPCI Systems 4.5.A Core Sorav and_LPCI Systems (cont'd)

(cont'd)

Check Once/ day Calibrate Once/3 months Test Step Once/3 months 2.

From and after the date that one 2.

This section intentionally left I

of the core spray systems is made blank or found to be inoperable for any reason continued reactor 3

LPCI system Testing shall be as operatIonispermissibleduring follows:

the succeeding seven days, provided that during such seven

a. Simulated Once/ Operating days all active components of the Automatic Cycle other core spray system and active Actuation i

components of the LPCI system and Test the diesel generators are operable.

b. Pump Once/ month l

3.

The LPCI system shall be operable Operability whenever irradiated fuel is in the reactor vessel, and prior to

c. Motor Operated Once/ Month reactor startup from a Cold valve Condition, except as specified in operability 3.5.A.4 and 3.5.F.5.

d. Pump Flow Once/3 months 4.

From and after the date that the LPCI system is made or found to be Each LPCI pump shall pump 4800 inoperable for any reason, gpm at a head across the pump continued reactor operation is of at least 380 ft.

permissible only during the-succeeding seven days unless it is sooner made operable, provided that during such seven days the l

containment cooling system (including 2 LPCI pumps) and active components of both core l

spray systems, and the diesel generators required for operation of such compcients if no external source of power were available shall be operable.

l 5.

If the requirements of 3.5.A cannot be s.et, an orderly shutdown of the reactor shall be initiated and the reactor shall be in the Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Amendment No. 4.2 N, 177r 174, 135 104

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This Page Intentionally left Blank e,.

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Amendment No.135 105 l

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1.

LIMITING CONDITION FOR OPERATION SURVEILLANCE REOUIREMENT i

l 3.5.B -Containment Coolina Systg3 4.5.B Containment Coolina System 1.

Except as specified in 3.5.B.2 1.

Containment Cooling system and 3.5.F.3 below, both Testing shall be as follows:

containment cooling system loops shall be operable whenever 1.tta Freauenev irradiated fuel is in the reactor vessel and reactor coolant 4.

Pump & Valve Once/3 months temperature is greater than Operability 212'F and prior to reactor startup from a Cold Condition, b.

Pump Capacity After pump 1

Test Each RBCCH maintenance i

2.

From and after the date that one pump shall and every 3 containment cooling system loop deliver 1700 gpm months is made or found to be inoperable at 70 ft. iDil.

for any reason, continued reactor Each SSHS pump operation is permissible only shall deliver 2700 during the succeeding 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> gpm at 55 ft. TDH.

Unless such-system loop is sooner made operable, provided that the c.

Air test on Once/5 years l_

other containment cooling system drywell and loop, including its associated torus headers diesel generator, is operable, and nozzles 3.

If the requirements of 3.5.B-cannot_be met, an orderly shutdown shall be initiated and the reactor shall be in a Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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Amendment No. f?$Af,JJf, 135 106 l

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'4 LIMITING CONDITION FOR OPERATION

$1)RVEILLANCE REOUIREMENT C.

HPCI System C.

SPCI System l

l 1.

The HPCI system shall be operable 1.

HPCI system testing shall be l

whenever there is irradiated fuel perf,..ned as follows:

in the reactor vessel, reactor pressure is greater than 150

a. Simulated Once/ operating psig, and reactor coolant Automatic cycle temperature is greater than Actuation 365'F; except as specified in Test 3.5.C.2 below, b.' Pump Oper-Once/ month 2.

From and after the date that the ability l

HPCI system is made or found to be inoperable for any reason,

c. Motor Operated Once/ month continued reactor operation is Valve Oper-permissible only during the ability

$Ucceeding seven days unless such system is sooner made operable,

d. Flow Rate at Once/3 months providing-that during such seven 1000 psig days all active components of the ADS system.the RCIC system, the

e. Flow Rate at Once/ operating LPCI. system and both core spray-150 psig cycle systems are operable.

The HPCI-pump shall deliver at least 3.

If the requirements of 3.5.C 4250 gpm for a system head cannot be met, an orderly corresponding to a reactor pressure of shutdown shall be initiated and 1000 to 150 psig.

the reactor pressure shall be reduced to or below 150 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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Amendment No. 42, //, JM JJ/, 135 107

_ _ _ _ _., _ _ _. ~..

LIMIVING CONDITION FOR OPERA 110N SURVEIL 1MCE __RE0'JIREMENT Retttor Core Isolation Conling l 3.5.D Reattor Core Isolation CQQling 4.5.0

(

1RCIC) System IRCICLSystem l

1.

The RCIC system shall be operable 1.

RCIC system testing shall be l

whenever there is irradiated fuel performed as follows:

in the reactor vessel, reactor pressure is greater than 150 a.

Simulated Once/ operating psig, and reactor coolant Automatic cycle temperature is greater than Actuation 365'F; except as specified in Test 3.5.D.2 below.

b.

Pump Once/ month 2.

From and after the date that the Operability RCICS is made or found to be inoperable for any reason, c.

Motor Once/ month continued reactor power operation Operated is permissible only during the Valve succeeding seven days provided Operability that during such seven days the HPCIS is operable.

d.

Flow Rate at Once/3 months 1000 psig 3.

If the requirements of 3.5.D cannot be met, an orderly e.

Flow Rate at Once/ operating shutdown shall be initiated and 150 psig cycle the reactor pressure shall be reduced to or below 150 psig The RCIC pump shall deliver at within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

leatt 400 gpm for a system head corresponding to a reactor pressure of 1000 to 150 psig.

Amendment No.

A2, M, W, 13r 103

l LIMITING CONDITION FOR OPERATION SURyG LLANCE REOUIREMENT 3.5.E Automatic Deeressurizalina 4.5.E Mtiolatic_ Den.ttuurization System (ADS)

System ( ADil The Automatic Depressurization 1.

During each operating cycle the l1.

System shall be operable whenever following tests shall be performed there is irradiated fuel in the on the ADS:

reactor vessel and the reactor pressure is greater than 104 psig a.

A simulated automatic actuation and prior to a startup from a test shall be performed prior Cold Condition, except as to startup after each refueling specified in 3.5.E.2 below.

outage.

The ADS manual inhibit switch will be included in this 2.

From and after the date that one test.

valve in the Automatic l

Depressurization System is made b.

With the reactor at pressure, or found to be inoperable for any each relief valve shall be reason, continued reactor manually opened until a operation is permissible only corresponding change in reactor during the succeeding seven days pressure or main turbine bypass unless such valve is sooner made valve positions indicate that operable, provided that duritig steam is flowing from the valve.

l such seven days the HPCI system is operable.

3.

If the requirements of 3.5.E cannot be met, an orderly shutdown shall be initiated and the reactor pressure shall be reduced to at least 104 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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Amendment No. ??, #2, J0f>,107, JJA,135 109

B LIMITING CONDITION LOR OPERATION SURVEILLANCE REOUIREMENT 3.5.T ttinimum Low Pressure Cooling _e.nd 4.5.F liinin!L1.ow Pressure Cooling Diesel Generator Availability and Diesel Generator Availability 1.

When it is determined that one 1.

During any period wten one diesel diesel generator is inoperable, the generator is inopetable, operable diesel qenerator shall be continued react n operation is demonstrated to be operable permissible cMy during the immediately and daily thereafter l

succeeding /2 hours unless such untit the inoperable diesel is diesel generator is sooner made repaired.

operable, provided that all of the low pressure core and I

containment cooling systems and the remaining diesel generator shall be operable.

If this requirement cannot be met, an orderly shutdown shall be initiated and the reactor shall be placed in the Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

2.

Any combination of inoperable components in the core and containment cooling systems shall not defeat the capability of the remaining operable components to fulfill the cooling function;.

3.

When irradiated fuel is In the re-actor vessel and the reactor is in the Cold Shutdown condition, both core spray systems, the LPCI and l

containment cooling systems may be inoperable, provided no work is being done which has the potential for draining the reactor vessel.

4.

During a refueling outage, for a period of 30 days, refueling oper-ation may continue provided that one core spray system or the LPCI system is operable or Specification 3.5.F.5 is met.

5.

When irradiated fuel is in the reactor vessel and the reactor is in the Refueling Condition with the torus drained, a single control rod drive mechanism may be removed, if both of the following conditions are satisfied:

Amendment No.

JE, 135 110

LIMITING CONDITION FOR OPERATION SURVEILLANCE REOUIREMENT 3.5.F Mhnimum low Pressure Coolina and D1etal Generator Availability a) No work on the reactor ves-sel, in addition to CR0 re-moval, will be performed which has the potential for exceededing the maximum

. leak rate from a single control blade seal if it became unseated, b) i) the core spray systems are operable and aligned with a suction path from the condensate storage

tanks,

11) the condensate storage tanks shall contain at least 200,000 gallons of usable water and the refueling cavity and dryer /

separator pool shall be flooded to a least elevation 114'-O" 3.5.G (Intentionally left blank) 3.5.H Baintenance of Filled Discharat 4.5.H Baintenance of Filled Discharoe Eint P_lat Whenever core spray systems LPCI The following surveillance requirements system, HPCI or RCIC are required to shall be adhered to to assure that the l

be operable, the discharge piping from discharge piping of the core spray the pump discharge of these systems to systems, LPCI system, HPCI and RCIC are l

the last block valve shall be filled.

filled:

1.

Every month prior to the testing of the LPCI system and core spray systems, the discharge piping of l

i these $ stems shall be vented from the hig point and water flow observe.

2.

Following eny period where the LPCI system or core spray systems have l

not been required to be operable, the discharge piping of the inoperable system shall be vented from the high point prior to the return of the system to service.

Amendment No. 79 135 111

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 4.5.H Maintenance of rilled Diteharoe Ping (Cont'd) 3.

Whenever the HPCI or RCIC system is lined up to take suction from the torus, the discharge piping of the HPCI and RCIC shall be vented from the high point of the system and water ficar observed on a monthly basis.

4.

The pressure switches which monitor the discharge lines to ensure that they are full shall be functionally tested every month and calibrated every three months.

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Amendment No, 37. 135 112 i

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3.5.A Core Sorav and LPCI System This specification assures that adequate emergency cooling capability is available whenever irradiated fuel is in the reactor vessel, Based on the lost of coolant analysis performed by General Electric in accordance with Section 50.46 and Appendix K of 10CF350, the Pilgrim I Emergency Core Cooling Systems are adequate to provide sufficient cooling to the core to dissipate the energy associated with the loss of coolant accident, to limit calculated fuel clad temperature to less than 2200'F, to limit calculated local metal water reaction to less than or equal to 17%, and to limit calculated core wide metal water reaction to less than or equal to 1%.

General Electric Company Proprietary Report EAS-65 0989, " Safety Evaluation for Interim Operation of Pilgrim Nuclear Power Station with Reduced Core Spray System Flow Rate" (September 1999) calculates a peak fuel clad temperature of less than 2200'F with a Core Spray pump flow of 3240 gallons per minute (gpm). A flew rate of 3300 gpm ensures adequate flow for events involving degraded voltage.

Core spray distribution has been shown, in full-scale tests of systems similar in design to that of Pligrim, to exceed the minimum requirements by at least 25%.

In addition, cooling effectiveness has been demonstrated at less than half the rated flow in simulated fuel assemblies with heater rods to duplicate the decay heat characteristics of irradiated fuel.

The accident analysis takes credit for core spray flow into the core at vessel pressure below 205 psig. However, the analysis is conservative in that no credit is taken for spray cooling heat transfer in the hottest fuel bundle until the pressure at rated flow for the core spray (104 psig vessel pressure) is reached.

The LPCI system is designed to provide emergency cooling to the core by I

flooding in the event of a loss-of-coolant accident.

This system functions in combination with the core spray system to prevent excessive fuel clad temperature.

The LPCI system and the core spray system provide adequate l

cooling for break areas of approximately 0.2 square feet up to and including the double-enderi recirculation line break without assistance from the high i

pressure emergency core cooling systems.

I The combination of the core spray systems and the LPCI system assures that adequate core cooling is achieved assuming any coincident single failure of an active safety-related component. Core Standby Cooling System (CSCS) performance evaluations consider only the most severe single failure for each break size range.

These single failures include the LPCI injection valve, one diesel generator, the HPCI system or one ADS valve.

With these single failures, the combinations of analyzed low pressure CSCS capacity include two core spray pumps, one core spray pump and two LPCI pumps, or two core spray and four LPCI pumps.

Each core spray system consists of one pump and associated piping and valves with all active components required to be operable. The LPCI system consists of four LPCI pt.mps and associated piping and valves with all active components required to be operable.

Amendment No. 75, 109, 131, 135 113 I

IL% n:

I 3.5.A Core Sorav and LPCI System This specification assures that adequate emergency cooling capability is available whenever irradiated fuel is in the reactor vessel.

Based on the loss of coolant analysis performed by General Electric in accordance with Section 50.46 and Appendix K of 10CFR50, the Pilgrim I Emergency Core Cooling Systems are adequate to provide sufficient cooling to 4 e the core to dissipate the energy associated with the loss of coolant accident, to limit calculated fuel clad temperature to less than 2200*F, to limit calculated local metal water reaction to less than or equal to 171, and to limit calculated core wide metal water reaction to less than or equal to 11.

Core spray distribution hat been shown, in full-scale tests of systems similar in design to that of Pilgrim, to exceed the minimum requirements by at least 25%.

In addition, cc311ng effectiveness has been demonstrated at less than half the rated flow in simulated fuel assemblies with heater rods to duplicate the decay heat characteristics of irradiated fuel. The accident analysis takes credit for core spray flow into the core at vessel pressure below 205 psig.

However, the analysis is conservative in that no credit is taken for spray cooling heat transfer in the hottest fuel bundle until the pressure at rated flow for the core spray (104 psig vessel pressure) is reached.

The LPCI system is designed to provide emergency coolirg to the core by flooding in the event of a loss-of-coolant accident.

This system functions in combination with the core spray system to prevent excessive fuel clad temperature.

The LPCI system and the core spray system provide adeo, ate l

cooling for break areas of approximately 0.2 square feet up to and '.ncluding tne double-ended recirculation line break without assistance from t'.te high pressure emergency core cooling systems.

The combination cf the core spray systems and the LPCI system assures that adeq w te core cooling is achieved assuming any coincident single failure of an active safety-related component.

Core Standby Cooling System (CSCS) performance evaluations consider only the most severe single failure for each break size range.

These single failures include the LPCI injection valve, one diesel generator, the HPCI system or one ADS valve. With these single failures, the combinations of analyzed low pressure CSCS capacity inciude two core spray pumps, one core spray pump and two LPCI pumps, or two core spray and four LPCI pumps.

Each core spray system consists of one pump and associated piping and valves with all active components required to be operable.

The LPCI system consists of four LPCI pumps and associated piping and valves with all active components required to be operable.

Amendment No. JE, 107, 137,135 113

BASIS:

3.5.A Core Sorav and LPCI Systems (Cont'd)

I Should one core spray system become inoperable, the remaining core spray and the LPCI system are a Wilable should the need for core cooling arise.

Based on judgments of the reliability of the remaining systems; i.e., the core spray and LPCI, e seven-day repair period was obtained.

l If the LPCI system is not available, at least 2 LPCI pumps must be available to fulfill the containment cooling function.

Based on judgments of the reliebility of the remaining core spra) systems, a 7-day repair period was set.

The LPCI system is not considered inoperable when the RHR System is operating in the shutdown croling mode.

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m Amendment No. 135 jj4 l

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BASES:

3.5.B Containment Coolino System The containment cooling system for Pilgrim I consists of two independent loops each of which to be an operable loop requires one LPCI pump, two RBCCW pumps, and two SSH pumps to be operable. There are installed spares for margin above the design conditions.

Each sys function;i.e., removing 64x10gemhasthecapabilitytoperformits Btu /hr (Ref. Amendment 18), even with some system degradation.

If one loop is out-of-service, reactor operation is permitted for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

With components or systems out-of-service, overall core and containment cooling reliability is maintained by the operability of the remaining cooling equipment.

Since some of the SSH and RBCCH pumps are required for normal operation, L

capacity testing of individual pumps by direct flow measurement is impractical.

The pump capacity test is a comparison of hieasured pump performance parameters to shop performance tests combined with a comparison to the performance of the previously tested pump. These pumps are rotated during operation and performance testing will be integrated with this or performed during refueling when pumps can be flow tested individually.

Tests during normal operation will be performed by measuring the shutoff head, Then the pump under test will be placed in service and one of the previously operating pumps secured.

Total flow indication for the system will be compared for the two cases. Where this is not feasible due to changing system conditions, the pump discharge pressure will be measured and its power requirement will be used to establish flow at that pressure.

4 Amendment No. 135 1j$

ESES:

3.5 C ILPCI The limiting conditions for operating the HPCI System are derived from the Station-Nuclear Safety Operational Analysis (Appendix G) and a detailed functional analysis of the HPCI System (Section 6).

The HPCIS is provided to assure that the reactor core.s adequately cooled to limit fuel clad temperature in the event of a small break in the nuclear system and loss-of-coolant which does not result in rapid depressurization of the reactor vessel.

The HPCIS permits the reactor to be shut dcwn while maintaining sufficient reactor vessel water level inventory until the vessel is depressurized.

The HPCIS continues to onerate until reactor vessel pressure is below the pressure at which LPCI operation or Core Sprey System operation maintains core cooling.

The capacity of the system is selected to provide this required core cooling.

The HPCI pump is designed.to pump 4250 gpm at reactor pressures between 1100 and 150 psig.

Two sources of water are available.

Initially, demineralized

. water from the condensate storage tank is used instead of injecting water from the suppression pool into the reactor.

When the HPCI System begins operation, the reactor depressurizes more rapidly than would occur if HPCI'was not initiated due to the condensation of steam by the cold fluid pumped into the reactor vessel by the HPCI System. As the reactor vessel pressure continues to decrease. the HPCI flow momentarily reached 3quilibrium with the flow through the break. Continued depressurization causes.the break flow to decrease below the HPCI flow and the liquid inventory begins'to rise. This type of response is typical of the small breaks.

The core never uncovers and is continuously cooled throughout the transient so that no core damage of any kind occurs for breaks that lie within the capacity range of-the HPCI,

-The analysis in=the FSAR, Appendix G. shows-that the-ADS provides a single failure proof' path for'depressurization for postulated transients and accidents.

The RCIC is required as an alternate source of makeup-to-the HPCI

.only in the case of loss of all offsite A-C power. Considering the HPCI and the ADS plus RCIC as redundant. paths, and considering judgments of the reliability of the ADS and RCIC systems, a 7-day allowable repair time is-specified.

The requirement that HPCI be operable when reactor coolant temperature is greater that 365'F is included in Specification 3.5.C.I to clarify-toat HPCI-need not be operabla-during certain testing.(e.g., reactor. vessel hydro testing at high reactor pressure and low reactor coolant temperature).

365'F is approximately equal to the saturation steam temperature at 150 psig.

Amendment No.

199, 135 116

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MiiS:

3.5.0 RCIC System The RCIC is designed to provide makeup to the nuclear system as part of the i

planned operation for periods when the normal heat sink is unavailt.ble.

The nuclear safety analysis, FSAR Appendix'G, shows that RCIC also serves as 3

redundant makeup system on total loss of all offsite power in-the event that HPCI is unavailable.

In all other postulated accidents and transients, the

~ ADS provides redundancy for-the HPCI.

Based on this and judgments-on the reliability of the HPCI-system, an allowable repair time of seven days is specified.

- The requirement that RCIC be operable when. reactor coolant temperature is greater than 365'F is included in Specification 3.5.D.1 to clarify that RCIC need not-be operable during certain testing (e.g., reactor vessel hydro testing at high reactor pressure and low reactor coolant temperature). 365'F is approximately equal _to the saturation steam temperature at 150 psig.

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Amendment No. g, 135

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3.5 E Automatic Deoressurization System (ADSJ The limiting conditions for operating the ADS are derived from the Station Nuclear Operational Analysis (Appendix G) and a detailed functional analysis of the ADS (Section 6).

This specification ensures the operability of the ADS under all conditions for which the automatic or manual depressurization of the nuclear system is an essential' response to station abnormalities.

The nuclear system pressure relief system provides automatic nuclear system depressurization for small breaks in the nuclear system so that the low pressure coolant injection (LPCI) and the core spray systems can operate to l

protect the fuel barrier.

Because the Automatic Depresscriration System does not provide makeup to the reactor primary vessel, no creatt is taken for the steam cooling of the core caused by the system actuation to provide further conservatism to the CSCS.

Performance analysis of the Automatic Depressurization System is considered only with respect to its depressurizing effect in conjunction with LPCI or Core Spray.

There are four valves provided and each has a capacity of 800,000 lb/hr at a reactor pressure of 1125 psig.

The allowhble out of service time for one ADS valve is determined as seven days because of the redundancy and because of HPCIS operability; therefore, redundant. protection for the core with a small break in the nuclear system is still available.

The /DS test circuit permits continued surveillance on the operable relief valves to assure that they will be available if required.

H, US Amendment No.

118

BASES:

3.5.H Maintenance of Filled Discharae Ping If.the discharge piping of the core spray, LPCI system, HPCI, and RCIC are not filled, a_ water hammer can develop in this piping when the pump and/or pumps are-started. An analysis has been done which shows that if a water hammer were to occur at the time at which the system were required, the system would still perform its design function.

However, to minimize damage to the discharge piping and to ensure added margin in the operation of these systems,_

c this Technical Specification requires the discharge lines to be filled whenever the system is in an operable condition.

Amendment-No. 135-121 a.

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4 4.5 Core and Containment Coolino Systems Surveillance FreaktDIl11 i

The testing interval for the core and-containment cooling systems is based on industry practice, quantitative. reliability analysis, judgment and

' practicality.

The core cooling systems have not been designed-to be fully testable during operation.

For example, in the case of the HPCI, automatic initiation during power operation would result in pumping cold water into the reactor vessel which is not. desirable. Complete ADS testing during power operation causes an undesirable loss-of-coolant inventory.

To increase the availability of the core and. containment cooling systems, the components which make up the system; i.e., instrumentation, pumps, valves, etc., are tested frequently.

The pumps-and motor operated injection valves are also tested each month to assure their operability.

A simulated automatic actuation test once each cycle combined with monthly tests of the pumps and injection valves is deemed to be adequate testing of these systems..

. The surveillance requirements provide adequate assurance that the core and containment cooling systems will be operable when required.

i Amendment No, 135 122

1 kLSES:

3.6 F and 4.6.F 2et Pumo Flow Hismatch The LPCI loop selection logic has been previously described in the Pilgrim Nuclear Power Station FSAR.

For some limited low probalilicy accidents with the recirculation loop operating with large speed differtices, it is possible for the logic to select the wrong loop for injection.

For these limited conditions the core spray itself is adequate to prevent fuel temperatures from exceeding allowable limits. However, to limit the probability even further, a procedural limitation has been placed on the allowable variation in speed between the recirculation pumps.

The licensee's analyses indicate that above 80% power the loop select logic could not be expected to function at a speed differential of 15%. At or below 80% power the loop select logic would not be expected to function at a speed differential of 20%. This specification provides a margin of 51 in pump speed differential before a problem could arise.

If the reactor is operating on one pump, the loop select logic trips that pump before making the loop selection.

The flow mismatch restriction also derives from the " Core flow Cosstdov concern. This concern postulates that if the recirculation loop with the higher flow is broken, the " effective core flow" is determined by the loop with the lower flow. Compared to a matched flow condition, this would start pump coastdown from a lower flow / speed with the reactor power effectively above the rated rod line, Therefare, boiling transition may occur earlier during a postulated LOCA event, which could result in higher calculated peak cladding temperatures (PCTs).

Therefore, the purpose of the " Core Flow Coastdown" flow mismatch restriction is to maintain Pilgrim within its analyzed conditions.

Spacification 3.6 F allows 30 minutes to correct a mismatch in recirculation pump. seeds in order to take manual control of the recirculation pump MG set scoop tube positioner in the event that its control system should fail.

Amendment No. 71,135 148 l

l

BASES:

(Cont'd)

-4.9 The diesel fuel oil quality-must be checked to ensure proper operation of the.

diesel generators. Water content should be minimized because water in the fuel could contribute to excessive dange to the diesel engine.

The Electrical Protection Assemblies (EPAs) on the RPS inservice power supplies (either two motor generator sets-or one motor generator and the alternate supply), consist of protective relays that trip their incorporated circuit breakers on overvoltage, undervoltage or underfrequency conditions.

There are 2 EPAs in series per power source.

It is necessary to periodically test the relays to ensure the sensor is operating correctly and to ensure the trip unit is operable.

Based on experience at conventional and nuclear power plants, a six month frequency for the channel functional test is established.

This frequency is consistent Mith the Standard Technical Specifications.

The EPAs of the power sources to the RPS shall be determined to be operable by performance of a chanhel Calibration of the relays once per operating cycle.

During calibration, a transfer to the alterrate power source is required; however, prior to switching to alternate feed, de-energiration of the applicable MG set power source must be accamplished.

This results.in a half-scram on the channel being calibraten until the alternate power source is connected and the half scram is cleared.

Based on operating experience, drift of the EPA protective relays is not sign'ficant.

Therefore, to avoid possible spurious scrams, a calibration frequency of once per cycle is established.

Amendment No. 127, 135 201

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