ML20090L974

Amend 152 to License DPR-46,changing Tech Specs to Clarify Usage of Terms Sys & Subsystem, Assuring Proper & Correct Interpretation of Tech Specs

ML20090L974
Person / Time
Site:	Cooper
Issue date:	03/11/1992
From:	Bevan R Office of Nuclear Reactor Regulation
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ML20090L975	List: ML20090L974 ML20090L976
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NUDOCS 9203230359
Download: ML20090L974 (41)

Text

,

'o UNITED STATES

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NUC dAR REGULATORY COMMISSION 5

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NEBRASKA PUBLIC POWER DISTRICT DOCKET NO. 50-E8 COOPER NUCLEAR STATION eMENDMENT T0 FACILITY OPERATING LICENSE Amendment No.152 License No. DPR-46 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Nebraska Public Power District (the licensee) dated September 30, 1991, as supplemented by letter dated January 20, 1992, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance:

(i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations:

D.

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

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

9203230359 920311 DR ADOCK 0500 8

o

s - 2.

Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and

-Paragraph 2.C.(2) of Facility Operating License No. DPR-46 is_hereby amended to read as follows:

2.

Technical' Specifications The' Technical Specifications contained in Appendix A, as revised through Amendment No.152,-are hereby incorporated in the license.

The' licensee _shall operate the facility in accordance with the Technical Specifications.

3.

The license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Ys John T. Larkins, Director Project Directorate IV-1 Division of Reactor Projects - III/lV/V Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: March 11, 1992

ATTACHMENT TO llCENSE AMENDMENT N0.152 FAClllTY OPERATING LICENSE N0. DPR-46 DOCKET NO. 50-298 Replace'the following pages of the Appendix A Technical Specifications with the enclosed pages.

The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

REMOVE PAGES INSERT PAGES ii-iii 11-111 21 21 52-52a 52-52a 83 83

-107-108 107-108 110 110 114-122 114-122 124-128 124-128 131 131 165-166 165-166 180 180 182-183 182-183 205a 205a 209a-209b 209a-209b 215b-215e 215b-215e 216b2 216b2

s TABLE OF CONTENTS (cont'd) 1 Pace No-.

SURVEILLANCE LIMITING CONDITIONS FOR OPERATION REOUIREMENTS 3.5 - CORE AND CONTAINMENT COOLING 1 SYSTEMS 4.5 114 - 131 A.

Core Spray and LPCI Systems A

114 B.

RHR Service Water System B

116 C,

HPCI System C

117 D.

RCIC System.

D 118-E. -Automatic Depressuritation System E

119 F.

Minimum Low Pressure Cooling System Diesel Generator Availability F

120 G.

Maintenance of Filled Discharge Pipe G

122 H.

Engineered Safeguards Compartments Cooling H

123 3.6 PRIMARY SYSTEM BOUNDARY 4.6 132 - 158 A.

Thermal and Pressurication Limitations A

132 B.

-Coolant Chemistry B

133a C.

Coolant Leakage C

135 D.

Safety and Relief Valves D

136 E.

-Jet Pumps E

137 F.

Recirculation Pump Flow Mismatch F

137 G.

Inservice Inspection G

137 H.- Shock Suppressors (Snubbers)

H 137a 3.7 CONTAINMENT SYSTEMS 4.7 159 - 192 A.

Primary' Containment A

159 B.

Standby Gas Treatment System B

165 C.

Secondary Containment.

C 165a D.

Primary Containment Isolation Valves D

166

-3.8 MISCELLANEOUS RADIOACTIVE MATERIAL SOURCES 4.8 185 - 186 3.9=

AUXILIARY ELECTRICAL SYSTEMS 4.9 193 - 202-A.- Auxiliary Electrical Equipment A

193 B.

Operation with Inoperable Equipment B

195 3.10 CORE ALTERATIONS 4.10 203 209 A.

Refuelin5 Interlocks A

203 3.

Core Monicering 3

205 C.

Spent Fuel Pool Water Level C

205 D.

Time Limitation D

206 E.

Spent-Fuel Cask Handling E

206 3.11 FUEL RODS 4.11 210 - 214e A.

Average Planar Linear Heat Generation Rate (APLHCR)

A 210 B.

Linear Heat Generation Rate (LHCR)

B 210 C.

Minimum Critical Power Ratio (MCPR)

C 212 Amendment No. 9#,97/,100,152..

6 TABLE OF QQNTENTS (cont'd)

Pare No _

SURVEILLANCE LIMITING CONDITIONS FOR OPERATION REOUIREMENTS 3.12 ADDITIONAL SAFETY RELATED PLANT CAPABILITIES 4.12 215 - 215f A.

Main Control Room Ventilation A

215 B.

Reactor Equipment Cooling System B

215b l

C.

Service Water System C

215e D.

Battery Room Vent D

215c 3.13 RIVER LEVEL 4.13 216 3.14 FIRE DETECTION SYSTEM 4.14 216b 3.15 FIRE SUPPRESSION WATER SYSTEM 4.15 216b 3.16 SPRAY AND/OR SPRINKLER SYSTEM (FIRE PROTECTION) 4.16 216e 3.17 CAR 3ON DIOXIDE AND RALON SYSTEMS 4.17 216f 3.18 FIRE HOSE STATIONS 4.18 216g 3.19 FIRE BARRIER PENETRATION FIRE SEALS 4,19 216h 3.20 DELETED 2161 3.21 ENVIRONMENTAL / RADIOLOGICAL EFFLUENTS 4.21 216n A.

Instrumentation 216n B.

Liquid Effluents 216x C.

Gaseous Effluents 216a4 D.

Effluent Dose Liquid / Gaseous 216all E.

Solid Radioactive Waste 216a12 F.

Monitoring Program 216a13 G.

Interlaboratory Comparison Program 216a20 3.22 SPECIAL TESTS / EXCEPTIONS 4.22 216b1 A.

Shutdown Margin Demonstration 216bl

~

B.

Training Startup 216b2 C.

Physics Tests 216b3 D.

Startup Test Program 216b3 5.0 MAJOR DESIGN FEATURES 5.1 Site Features 217 5,2 Reactor 217 5.3 Reactor Vessel 217 5.4 Containment 217 5.5 Fuel Storage 218 5.6 Seismic Design 218 5.7 Barge Traffic 218 6.0 ADMINISTRATIVE CONTROLS 6.1 Organitation 219 6.1,1 Responsibility 219 6.1.2 Offsite 219 6.1.3 Plant Staff Shift Complement 219 6.1.4 Plant Staff - Qualifications 219a Amendment No. 89,97,98,127/, 152

-lii-

2.1 Eases

(Cont'd) 5.

Main Steam Line Isolation Valve Closure on Low Presaure The low pressure isolation of the main steam -lines (Specifica-tion 2.1.A.6) was provided to protect against rapid reactor depressuriza-tion.

B, Reactor Unter level Trio Settings Which initiate Core Standby Cooling Systems l

(CSCS)

The core standby cooling systems are designed to provide sufficient cooling to l

the core to dissipate the energy associated with the loss-of-coolant accident and to. limit fuel clad temperature, to assure that core geometry remains intact and to limit any clad metal-water reaction to less than 1%.

To accomplish their intended function, the capacity of each Core Standby Cooling System component was established based on the reactor low water level scram set point.

To lower the set point of the low water level scram would increase the capacity requirement for each of the CSCS components.

Thus, the reactor vessel low water level scram was set low enough to permit margin for operation, yet will not be set lower because of CSCS capacity requirements.

The design for the CSCS components to meet the above guidelines was dependent upon three previously set parameters: The maximum break size, low water level scram set point and the CSCS initiation set point. To lower the set point for initiation of the CSCS may lead to a decrease in effective core cooling.

To raise the -CSCS initiation set point would be in a safe direction, but it would reduce-the margin established to prevent actuation of the CSCS during normal operation or during normally expected transients.

Transient and accident analyses reported in Section 14 of the Safety Analyses Report _ demonstrate that these conditions result in adequate safety margins for the fuel.

C.

References for 2.1 Basqg 1.

"Ceneric Reload Fuel Application," NEDE-24011-P, (most current approved submittal).

2.

" Cooper Nuclear Station Single-Loop Operation," NEDO 24258, May 1980.

3,

" Supplemental Reload Licensing Submittal for Cooper Nuclear Station Unit 1," (applicable reload document).

4.

Safety Analysis Report (Section XIV).

Amendment No. 9#,152.

=.

NOTES FOR TABLE-3.2.A 1.

.Vhenever Primary Containment integrity is required there shall be two operable or tripped trip systems for each function.

2.

If the minimum number of operable instrument channels per trip system requirement cannot be met by a trip system, that trip system shall be tripped.

If the requirements cannot be met by both trip systems, the appropriate action listed below shall be taken.

A.

Initiate an orderly shutdown and have the reactor in a cold shutdovn condition in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

B.

Initiate an orderly load reduction and have the Main Steam Isolation Valves shut within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

C.

Isolate the Reactor Water Cleanup System.

D.

Isolate the Shutdown Cooling mode of the RER System.

l 3.

Two required for each steam line.

4.

These signals also start the Standby Cas Treatment System r.nd initiate Secondary Containment isolation.

5.

Not required in the refuel, shutdown, and startup/ hot standby modes (interlocked with the mode switch).

,6.

Requires one channel from each physical location for each trip system.

7.

Low vacuum isolation is bypassed when the turbine stop is not full open, manual bypass switches are in bypass and mode switch is not in RUN, 8.

The instruments on this table produce primary containment and system isolations. The following listing groups the system signals and the system isolated.

Group 1 Isolation Signals:

1.

Reactor Low Low Low Water Level (2-145.5 in.).

2.

Main Steam Line High Radiation (3 times full power background) 3, Main Steam Line Low Pressure (2825 psig in the RUN mode) 4.

Main Steam Line Leak Detection (5200'F) 5.

Condenser Low Vacuum (27" Hg vacuum) 6.

Main Steam Line High Flow (5150% of rated flow)

Isolations:

1.

MSIV's 2.

Main Steam Line Drains-1 l

Amendment No. 55,83,88,115, 152,

~

.--. - - - -. _..... ~

= _ ~

NOTES FOR TABLE 3.2.A'-(cont'd.)

Group 2 IsolationLSignals:

1.

Reactor Low Water Level (24.5 inches)

-2.

High Dry Well Pressure ($ 2 psig)

-Isolations:

1.

RHR Shutdown Cooling mode of the RHR system.

l 2.

Drywell floor and equipment drain sump discharge lines.

3.

TIP ball valves 4.

Group 6 isolation relays Group 3 Isolation Signals:

1.

Reactor Low Water Level (24.5 inches) 2.

Reactor Water Cleanup System High Flow ($200% of system flow) 3.

Reactor Water Cleanup System High Area Temperature (s 200'F)

Isolations:

1.

Reactor Water Cleanup System Group 4 Isolation Signals:

Provided by instruments on Table 3.2.B (HPCI)

Isolations:

LIsolates the HPCI steam line Group 5-Isolation Signals:

Provided by-_ instruments on Table 3.2.B (RCIC)

Isolations:

Isolates the RCIC steam line.

Group.6 Isolation' Signals

1.

Group 2 Isolation Signal 2.

Reactor Building H&V Exhaust Plenum High Radiation ((100 mr/hr)

Amendment No. 103,109, f45,152

-52a-

?

3.2 MSfd In addition to reactor protection instrumentation which initiates a reactor scram,

-protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious consequences. This set of specifications provides-the limiting conditions of operation for the primary system isolation function, initiation of the core cooling systems, control rod block and Standby Cas Treatment System. The objectives of the specifications are (1) to assure j

the effectiveness of the protective instrumentation when required even during periods when portions of such systems are out of service for maintenance, and (2) to pt acribe the trip settings required to assure adequate performance. When necessary.

one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.

Some of _ the settings on the instrumentation that initiate or control core and containment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety, The set points of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the normal operating range to prevent inadvertent actuation of the safety system involved and exposure to abnormal situations.

A.

primary Containment Isolation Functions Actuation of primary containment valves is initiated by protective instrumentation shown in Table 3.2.A which senses the conditions for which isolation is required.

Such instrumentation must be available whener primary containment integrity is required.

The instrumentation which initiates primary s) *em isolation is connected in a dual D

bus a,rrangement.

The low water level instrumentation, set to trip at 168.5 inches (+4.5 inches) above the top of the active fuel, closes all isolation valves except those in Groups 1, 4, 5,

and 7.

Details of valve grouping and required closing times -are given in Specification 3.7.

For - valves which isolate at this level this trip setting is adequate to prevent core uncovery in the case of a break in the largest line assuming a 60 second valve closing time. Required closing times are less than this.

The low low low reactor water level instrumentation is set to trip when the water level im 19 inches (-145.5 inches) above the top of the active fuel.

This trip closes Groups 1 and 7-Isolation Valves (Reference 1), activates the remainder of the CSCS subsystems, and starts the emergency diesel generators.

These trip level settings were chosen to be high enough to prevent spurious actuation but low enough to initiate CSCS operation and primary system isolation so that post accident cooling can be accomplished, Amendment No. 5,83,83,102,16, 152 83-l

Lilillll{G CONDITIONS FOR OPERATION EllyLILI ANCE RLQUIREMI NTS 3.4 STANDBY LIQUID CONTROL. SYSTEM 4,4 STANDhY LIOU1D C01(TROL SYSIf3 Applistilit y!

g lienbility:

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

ments of the Standby Liquid Control Systes.

pblectivet Oldg ni m To assure the availability of a To verify the operability of the system with the capability to shut-Standby Liquid Control Systern.

down he resctor and maintain the shutdown condition without the use of control rods.

Epteification:

Epecificn ipm A.

Normal System Availability A.

Normal System t.vailability During periods when fuel is in.he The operability of the Standby Liq.

reactor and prior to startup from a uid Control Systern shall be shown by Cold Condition, the Standby Liquid the performante of the following Control System shall bo operable, tests:

except as specified in 3.4.B below.

This system need not be operable 1.

At least once per month each subsys-when the reactor is in the Cold tem shall be tested for operability condition and all control rods are by recirculating demineralized water fully inserted and Specification to the test tank.

3. L. A 'i s tne t.

2.

At least. once during each operating cycle:

a.

heck that the settings of the sub-system relief valves are 1450 ( p

( 1680 psig and the valves will rese t at P 2 1300 psig.

b.

Manually initiate the system, except explosive valves, and pump boron solution from the Standby Liquid Control Storage Tank through the recirculation path.

Miniinum pump flow rate of 38.2 gpm against a system head of 1300 psig shall be verified. Af ter pumping boron solu-

}

tion the system will be flushed with demineralized water.

c, Manually initiate one of the Standby Liquid Control System Pumps and Amendment No. 4.!23, 152

-107-

i lp

• ~

LIMITING CO@lIDNS FOR OPERATION

$URVEILIANCE REOUIREMENTS 3.4 4.4.A.2.c (Cont'd.)

e pump dek'neralized water into the reactor vs ssel from the tisat tank.

These tests check the actuation of the explosive charge of the tested

loop, proper operation of the valves, and pump operability. The i

replacement c'iarges to be installed r

will be selected from the same manu-factured batch as a previously tent-ed charge.

t d.

Both subsystems, including both l

explosive valves, shall be tested in the course of two operating cycles.

i B.

Operation with inoperable B.

Surveillance with Inonerable Components:

Connonents:

+

1.

From - and af ter the date ' that r.no 1.

When a subsystem -is found to be subsystem is made or found to be inoperable, the operable subsystem inope,rable. Specification 3.4.A.1 shall be verified to be operable shall be considered fulfilled ar.d immediately and daily thereafter continued operation permitted pro-until the inoperable subsystem is vided that the operable ' subsystem returned to an operabic condition.

remains operable and the inoperable subsystem is returned to an operable condition within seven days.

4 C.

Sodium Pentaborate Solution At all times when the Standby Liquid C.

Sodium Pentaborate Solution Control -System is required to be operable - the following conditions The following tests - shall be per-

-shall be met:

formed to verify the availability of 1.

The net volume versus concentration of the ' Liquid Control Solution in the liquid control tank shall be 1.

Volume:

Check and record at least main:ained as required-in Fig-once per day, ure 3.4.1.

2.

The temperature of the liquid con-trol-solution shall be maintained above--the curve shown in Fig-2.

Temperature:

Check and record at ure 3.4.2.

least once per day.

3.

Concentration: Check and record at least once per month.

Alao check concentration anytime water or buon is Amendment No. 80,139,1#6,152 108-

.- _ _ u _ __.______,____;.- _

3.4 PASES

[

t l

STANDBY LIOUID CON 1ROL SYSTEM i

A.

The Standby Liquid Control System consists of two, dietinct subsystems, each containing one positive displacement pump and independent suction from the liquid control tank, and discharge to a common injection header through 3

parallel squibb valves.

The purpose of the Standby Liquid Control System is to provide the capability of bringing the reactor from full power to a cold, xenon free shutdown condition assuming that none of the withdrawn control rods can be inserted.

To meet this objective, the system is designed to inject a i

quantity of boron that produces a concentration of 600 ppm of boron in the reactor core in less than 125 minutes.

The 600 ppm concentration in the j

reactor core is required to bring the reactor from full power to a 3.0 percent Ak suberitical condition, considering the hot to cold reactivity difference, xenon poisoning, etc.

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

The conditions under which the Standby Liquid control System must provide shutdown capability are identified via the Plant Nuclear Safety Operational j

Ar.alysis (Appendix G).

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

Standby Liquid Control System is not required.

Thus, the basic reactivity requirement for the core is the primary determinant of when the Standby Liquid Control System is required.

I The minimum limitation on the relief valve setting is intended to prevent the recycling of liquid control solution via the lif ting of a relief valve at too low a pressure.

The upper limit on the relief valve setting provides system l

protection from overpressure.

B.

Only one of the two Standby Liquid Control subsystems is needed for operating i

the system.

One inoperable subsystem does not immediately threaten shutdown capability, and reactor operation can continue while the inoperable subsystem

+

is being repaired.

Assurance that the remaining subsystem will perform its i

intended function and that the long term average availability of the system is not reduced is obtained for a one out of two system by an allowable equipment out of service time-of one third of the normal surveillance frequency.

This method de.rmines an equipment out of service time of ten days.

Additional conservatism is introduced by reducing the allowable out of service time to seven days.

C.

-Level indication and alarm indicate whether the solution volume has changed, which might indicato a possible solution coacentration change.

The test interval has been established in consideration of these factors,. Temperature and liquid level alarms for the system are annunciated in the control room.

The solution-is kept at least-10'F above the-saturation temperature to guard against boron precipitation.

The margin is-included in Figure 3.4.2.

• 110'

-Amendment No. 32,146,152 i

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.,_._,.__,-_.,.,.-,_a.,m_..-.-.

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..mm.,

LIMITING CONDITIONS FOR OPERATION-

$NRVEILLANCE REOUIREMENI 3.5 CORE AND CONTAINMENT COOLING SYSTEMS _

4.5 CORE AND CONTAINMENT COOLING SYSTEMS

[

Anp.licabiliev:

Aralicability:

i Applies to the operational status of Applies to the Surveillance Require-the core and containment cooling ments of the core and containment l

systems.

cooling systems which are required l

when the corresponding Limiting r

condition for Operation is in ef-feet.

Obicetive:

Objective:

l To assure the operability of the To verify the operability of the

- core and containment cooling systems core and containment cooling systems l

under all conditions for which this under all conditions for which this cooling capability is an essential cooling capability is an essential response to-station abnormalities, response to station abnormalities.

Soccification:

Specificariea:

l A.-

Core Sorav and LPC1 Systems A.

Core Sorav and LPCI Systems l

l 1.

Both Core Spray subsystems shall be

. 1.

Core Spray Subsystem Testing.

operable:

(

11cm Frequency (1) prior to reactor startup from a Cold Shutdown. or a.

Simulated Once/ Operating Automatic Cycle (2) when there is'irre.diated fuel Actuation Test.

in the vessel and when the reactor' vessel pressure is b.

Pump Operability Once/ month _

greater than atmospheric pressure, excapt as specified c.

Motor Operated Once/Honth in 3.5. A.2 and 3.5.F.3 below.

Valve Operability d.

Pump flow rate.

Once/3 months Both loops shall

- deliver at least 4720 gpm against a system head corresponding to a differential pressure of ;t 113 psi between the reactor vessel and_the primary containment, e.

Core Spray Header AP Instrumentation

' Check Once/ day Calibrate Once/3 months Amendment No. 152' 114-wy gvm

.y-w--+--w----

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-,w-pr,-v~,-,---v nr-w w~,w--re-v.,---*.e,--e v-n-.-----

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L1]i1TINC CONDITIONS FOR OPERATION

$1tfWflLIANCE REOUIREMENTS 3.5.A (cont'd.)

4.S.A (cont'd.)

2.

From and af ter the date that one of 2.

When it is determined that ono Cora

_l the Core Spray subsystems is made or Spray subsystern is inoperable, the found to be inoperable for any rea-operable Core Spray subsystem.nd son, continued reactor operation is both LPCI subsystems shall be veri-permissible during the succeeding fled to be operable immediately.

seven days provided that during such The operable Core Spray subsystem l

seven days all active components shall be verified to be operable i

that affect operability of the oper-daily tijereafter.

able Core Spray subsystem and all active components that effect opera-3.

LPCI subsystern testing shall be as bility of both LPCI subsystems and follows:

the diesel generatoss are operable, lleg Frecuency 3.

h LPCI subsystems shall be opera-a.

Simulated once/ Operating Automatic Actuation Cycle Test (1) prior to reactor startup from a Cold Condition, except as b.

Pump Operability Once/ month specified in 3.22.B.1, or c.

Motor Operated Once/ month (2) when there is irradiated fuel Valve Operability in the vessel and when the reactor vessel pressure is d.

Pump Flow Rate once/3 months greater than atmospheric pressure, except as specified During single pump LPCI, each RllR in 3.5 A.4 and 3.5 A.5 below, pump shall deliver at least 7700 CPM but no more than 8400 CPM against a system head equivalent to a reactor vessel pressure of 20 psid above 4

drywell pressure with water level below the jet pumps.

At the same conditions, two pump LPC1 flow shall l

be at least 15,000 CPM.

e.

Recirculation pump discharge valves shall be tested each refueling out-age to verify full open to full closed in t s 26 seconds, f.

An airtest shall be performed on the drywell and torus headers and noz-zies once/5 years.

4.

From and after the date that one of tb RHR (LPCI) pumps is made or 4.

When it is-- determir ed that one of found-to be inoperable for any rea-the RilR (LPCI) pumps is inoperable son, continued reactor operation is at a time-when it is required to be permissible only during the succeed-operable the remaining active compo.

ing thirty days provided that during nents that affect operability of the such thirty _ days the remaining ac.

LPCI subsystem containing the inop-tive components that affect opera-erabic pump, all active components bility of ' the LPCI subsystem con-that affect operability of the opor-taining the inoperable pump and all able LPCI subsystem, and both t: Je active components that affect opera.

Spray subsystems

t. hall be verified bility of the operable LPCI subsys-to be operable immediately and the tem both Core Spray subsystems and operable LPCI pumps daily thereaf-

~

both diesel generators are operable.

ter.

Amendment No. 57/.77,P3,H,97/,

115 -

137/. 152

L'1gITING CONDITIONS FOR OPERATION SURVEf tlANCE RE0yREMENTS 3.5,A (Cont'd.)

4.5 A. (Cont'd.)

I

.3 From and after t) e date that one 5.

When it is determined that LPCI sub.

LPCI subsystern is made or found to system is inoperable, the operable be ingerable for any reason, con.

LPCI subsystem, both Core Spray sub.

tinued reactor operation is permis.

systems and the RHR Service Water sible only during the succeeding 7 subsystein associated with the opera.

days, unless it is sooner made oper-ble 1.PCI subsystem, shall be verified l

able, provided that during such 7 to be operable immediately and daily days all active components that thereafter.

affect operability of the operable LPCI subsystem, both Core Spray subsysters, the RRR Service Water subsystem associated with the opera-ble LPCI subsystem and both diesel generators shall be operable.

6.

All recirculation pump discharge 6.

All recirculation pump dischar6e valves. shall be operable prior to valves shall be tested for operabil-l reactor startup (or closed if per.

ity during any period of Reactor cold mitted elsewhere in these specifica.

- shutdown exceeding 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, if oper.

i tions),

ability tests have not been performed during the preceding 31 days.

7.

The reactor shall not be started up

.vith the RHR system supplying cool.

i ing to the fuel pool.

8.

If the requirements of 3.5.A 1,2,3,4,5,6 or 7 cannot be met, an orderly shutdown of the reactor shall be initiated and the reactor shall be in the cold shutdown condi.

tion-vithin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

B.

Residual Heat Removal (RllR) Service

-l -

A-Residual Heat Removal (RHR)-Servlee Water System Water System 1.

RHR Service Water System testing 1.

Except as specified in' 3.$.B.2, shall be as follo'es:

3.5.B.3', and 3.5.F.3 below, both RHR Service - Water subsystems shall be item -

Freauenev cperable whenever irradiated fuel is in the reactor vessel and reactor a.

pump ( Valve once/3 months coolant temperature is greater than Operability 212'F, and prior to reactor startup from a Cold Condition, b.

Pump Capacity Test.- After pump Each RHR service maintenance and water booster pump every 3 months shall deliver 4000.

gpm.

Auendmunt No'. 36,92,95,146, 152 116-

.a m

. a

L1HITING CS DITIONS FOR OPERATION SURVEILIANCE REQ 1)JREMENTS 3.5.B (Cont'd.)

4.5.B (Cont'd.)

2, Froin and af ter the date that any RilR 2.

When it is deterinined that any RilR Service Water booster pump is inade Service Water booster pump is inop.

}

f or found to be inoperable for any erable, the rennining active con:po-reason, continued reactor operation nents that affect operability of the l

1s permissible only during the sue.

RllR Service Water subsystem contain-f ceeding thirty days, unless such ing the inoperable pump and all pump is sooner r.ade operable provid.

active components that affect opera.

ed that during such thirty days the bility of the operable RilR Service remaining active components that Water subsystem shall be verified to affect operability of the RilR Ser.

be operable irmediately ' and weekly vice Water subsystem containing the thereafter.

inoperable pump, and all active com.

ponents that affect operability of the operable RilR Service Water sub.

systern are operable.

3.

Frote and af ter the date that one RitR 3.

Vhen one RilR Service Water subsystern j

Service Water subsystem is made or becomes inoperable, the operable RilR found to be inoperable for any rea-Service Water subsystem and its son, continued reactor operation is associated LPCI subsystem shall be permissible only during the succeed.

verified to be operable irmediately ing seven days unless such subsystem and daily thereafter.

is cooner made operable, provided that all active components that affect _ operability of the operable RilR Service-Water subsystem, its I

associated LPCI subsystem, ~ and its

_3 associated diesel generator, are

operable, 4

If the requirements of 3.5 B.1, 3.5.B.2 or 3.5,B.3 cannot be met, an i

orderly shutdown sball be initiated and the reactor shall-be in a ev1d shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

C.

HPCI Systern C.

HPCI System l

5 1.

The llPCI System shall be operable 1.

IIPCI Systein testing shall be per-l whenever there is irradiated fuel in formed as follows:

the reactor vessel, reactor pressure is greater than 113 psig, and prior its IIgguency to reactor startup from a Cold Con-

dition, except as specified in a.

Simulated once/ operating 3.5.C.2 and 3.5.C 3 below..

-Automatic cycle Actuation Test b.

Pump Operability once/ month c.

Motor Operated Once/ month Valve Operability Amendment No. M,95,247/,152 117

~.

i r

LIMITING CONDITIONS FOR Rff,M U QJ

$URVE1LIANCE REOUIREMENT l

3.5.C llPCI System (cont M.)

4 5.0. llP01 System (cont'd.)

l f

4 2.

From and af ter the date that the 1tta Ereauency l

llPc1 System is made or found to be i

inoperable for any reason, continued d.

Flow Rate at once/3 monthu reactor operation la permissible approximately 1000 enly during the succeeding seven psig Steam Press.

j days unless such syctem is sonner made operable, providing that during e.

Flow Rate at once/ operating such seven day 6 all active compo-approximately 150 cycle nents that affect operability of the psig Stoam Praas.

ADS, the IC System, both LPCI sub.

systems and both Core Spray subsys.

S llPCI pump shall be demonstrated 1

{

tems are operable.

- se capable of delivering at least j

4.:w gpm for a system head corre*

3.

-With the survelilance requirements sponding to a reactor pressure of of 4.5.0 not performed at the re.

1000 to 150 psig.

quired intervals due to reactor shutdown, a teantor startup-ray be 2.

When it is determined that the llPCI conducted provided the appropriate System is inoperabla. the RCIC Sys-surveillance is performed within tem, both LPCI subs, tars, and both 49 hours5.671296e-4 days <br />0.0136 hours <br />8.101852e-5 weeks <br />1.86445e-5 months <br /> of achieving 150 psig reac-Core Spray subsystems shall be veri-tor steam pressure, fled to be operable immediately.

The RCIC System shall be verified to l

4 If the requirements of 3.5 C.1 can-be operable daily thereaf ter.

In net be met. an orderly shutduvn addition, the ADS. logic shall be l

shall be-initiated and the reactor demonstrated to be operable immedi.

--pressure Anil be reduced. to 113 ately and daily thereafter, psig or 1 s within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

.i D.

Reactor Core Isolation Cooling D,

Reactor Core Isolation Cooling l

fRCIC) System (RCIC1 Svateg

.l

.[

1.

The RCIC System shall be operable 1.

RCIC System testing shall be per.

I whenever there is irradiated fuel in formed as follows:

the reactor vessel, the reactor-pressure is greater than 113 psig.

Rea frecuency and prior to reactor startup from a cold condition, except as specified a.

Simulated once/ operating in 3.5.D 2-and 3.5.D 3 below.-

Automatic cycle.

Actuation Test Amendment No. 34,36,H6,152

. 118-

-~_

..-.,a-._

-_.___..__.u-_.._-_,

LIMITINC CONDITIONS POR OPERATJDH SURVEILIANCE R_EDylP1timI 3.5.D (cont'd.)

4.5.D (cont'd.)

lieg frequensy b.

pwnp operability once/ month c.

Motor Operated once/ month Valve Operability d.

Flow Rate at Once/3 months approximately 1000 psig Steam Pressure c.

Flow Rate at Once/operatin6 approxirmately 150 cycle psig Stearn Pressure The RCIC pump shall be demonstrated to be capable of deliverin6 at leaut 400 gpm for a rys t ern head corre-sponding to a reactor pressure of 1000 to 150 psig.

2.

From and after the date that the 2.

Vhen it is deterrnined that the RCIC

{

RCIC Systern is made or found to be System is inoperable, the lipCI Sys-inoperable for any reason, continued tem shall be verified to be operable reactor power operation is pertaissi-immediately.

ble only during the succeeding seven days provided that during such seven l

days the lipC1 Systein is operable.

3.

With the survcillance requirements of 4.5.D not perforined at the re-quired intervals due to reactor shutdown, a reactor startup may be conducted provided the appropriate surveillance is performed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of achieving 150 psig reac-tor stearn pressure.

4.

If the requirements of 3.5.D 1 6 2 cannot be met, an orderly shutdown shall be initiated and the reactor pressure shall be reduced to 113 psig or less within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

E.

butomatic Depressurization Sysica E.

Antputic Dentngntip1Lipn System (ADS 1

.($jil 1.

The Automatic Depressurization 1.

During each operating cycle the l

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

reactor vessel and the reactor pres-sure is greater than 113 psig and A simulated automatic actuation test prior to a startup from a Cold Con-shall be performed prior to startup

dition, except as specified in after eacl. refueling outage.

3.5.E.2 and 3.5.E.3 below.

Amendment No. 76,146,152

-119-

LIMITING CONDITIONS FOR OPERATION JURt'EILIANCE REOUIREMEN'[g 3.5.E (cont'd) 4 5.E (cont'd)_

2.

From and -af ter the Jate that one 2.

When it is determined that one valve valve in the Automatic Depressuriza-of the ADS is inoperable, the ADS tion System is made or found to be actuation logic for the other ' ADS l

inoperable for any reason, continu$d valves shall be demonstrated to be reactor operation is permissible operabl6 immediately.

In addition.

.only-during the succeeding seven

- the llPCI System shall be verified to l

days unless such valve is sooner be opetabic immediately.

made operable, provided that durirg l

such seven days the llPCI System is operable.

3.

With the surveillance requirements of 4.6.D.5 not performed at the required intervals due to reactor shutdown, a reactor startup may be conducted provided.the appropriate surveillance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of achieving 113 psig reactor steam pressure.

4, If the requirements of 3.5.E.1 or 3.5.E.2 cannot be met, an orderly shutdown shall be initiated and the reactor pressure shall be reduced to at least 113 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

F.

Minimum Low Pressure cooline and F.

Minimum Low Pressure Coolinr and Diesel Generator Avg 11 ability Diesel Generator Availability 1..

During any period when one-diesel 1.

When it is determined that one die.

generator is inoperable, continued sel generator is inoperable, the reactor operation is permissible LPCI,. Core Spray, and PJIR Service only during the ' succeeding sever.

Water subsystems associated with the

. days unless such diesel generator is operable diesel generator shall be sooner made operable, provided that verified to be operable immediately the operable diesel generator and and daily thereafter.

In add): ton, its associated LPCI, Core Spray, and the operable diesel generatot shall PJiR Service Water subsystems are be demonstrated to be operable imme-

.g operable and the requirements of diately and every three days there.

3.9.A.1'are met.

If this require-

efter, ment cannot be met,' the requirements of 3.5.F.~2 shall be met.

A*nendent No. 36,95, fH,152

-120-

.a,-,,.-..

-1,IMITIllG_C.ONDITIONS FOR ^PERATION SURVE1 LIM 4CE REQUIREMENTS 3.5.F (cont'd.)

4.5 F (cont'd.)

2.

During any period ahen both diesel generators are in;perable, continued reactor operation is permissible only during the succeeding 24 hourt unless one di.esel generator is soon -

t er made operable, provided that both 1.PCI subsystems, both Core Spray subsystems, and both RllR Service Water subsystems are operable and

~the reactor power level is reduced to 25% of rated power and the re.

quirements of 3.9.A.1 are met.

If I

this requitement cannot be met, an orderly shutdown shall be initiated and the reactor 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 />.

3.

Any combinstion of inoperable compo+

nents 'n the LPCI, RiiR Service Wa-ter, and Core = Spray systems chall not defeat the capability of the ronining operable components to fulfill the cooling functions.

4 When irradiated fuel is in the reac-tor vessel and the reactor is in the Cold Shutdown Condition, both core spray subsystems, both LPCI subsys-

-tems, and both RllR-' Service Water subsystems may be inoperable, pro.

vided no work 'is being done which has the potential for draining the reactor vessei. Refueling require.

ments are as specified in Specifica-tion 3.10.F.

5.

With irradiated fuel in the reactor vessel,4 one control rod drive hous-

-ing may be open while the suppres-sion enamber is completely drained provided that:.

a.

The reactor vessel head is removed.

b.

The spent fuel pool gates are open and the fuel pool water

. level is maintained at a lev-el- ;t 33. feet.

c.

The condensate transfer - sys-l tem is operable and a minimum of 230,000 gallons of water is in the condensate storage tank.

Amendment No. 88,97/, 152

-121

^

i

~,

-LIMITING CONDITIONS FOR OPERATION SURVEILIANCE R_.EOUIREMENT j

i 3.5.F (cont'd) 4,5.P (cont'd)-

d.

The automatic mode of the drywell sump pump is dis-

abled, e.

No maintenance is being con-ducted which will prevent filling-the suppression cham-ber to a level above the Core Spray and LPCI suctions.

f.

With the exception of the suppression chamber water supply.. both Core Spray sub-systems and both LPCI subsys-I tems are operable.

~,.

The control rod is withdrawn to the backseat.

h.

A special flange, capable of sealing a leaking control rod

housing, is available for immediate use.

The control rod housing is covered with the special nange following the removal of the control rod drive.

j.

No work is being performed in the vessel while the housing is open.

C.

Maintenance of Filled Discharge Pine Whenever the Core Spray subsystems' G.

ligintenance of Filled Discharge Ping LPCI subsystems, HPCI System, or RCIC System are required to be oper*

The following surveillance require-able, the -discharge piping from the ments shall be adhered to, to assure pump discharge of these systems t that the discharge piping of the I

ast block valve shall be Core Spray subsystems, LPCI subsys-te.as, HPCI System and RCIC System are filled:

1.

Whenewr a Core Spray subsystem,

.LPCI subsystem, the HPCI System, RCIC System is made operable, the discharge piping shall be vented from the high point of the syster -

and water ilow observed initially and on a monthly basis.

l-2.

The pressure switches which monitor I

the LPCI, Core Spray, HPCI and RCIC System lines to ensure they are full shall be functionally tested and calibrated every three months.

Amendment No-57/,80,88,97/,'152

-122-o l ^

.m_

_ _ _. _ _ _ _ _ _ _. _ _ m _.__ _

3.5 EASES A.

Core Sprav and LPCI Subsystems This specification assures that adequate emergency cooling capability in available whenever irradiated fuel is in the reactor vessel.

The liiniting conditions of operation in Specifications 3.5.A.1 through 3.5.A.8 I

specify the combinations of operable subsystems to assure the availability of the minimurn required cooling systems.

During reactor shutdown when the residual heat removal system. is realigned from LPCI to the shutdown cooling mode, the LPCI subsystems are-considered operable.

l

}

The Core Spray System is a low pressure coolant system which is comprised of two, distinct subsysterns and is designed to provide emergency cooling to the core by spraying in the esent of a loss of coolant accident.

This sys tern functions in combination with the LPCI System to prevent excessive fuel clad temperature.

l The LPCI System is an operating mode of the RilR System and is cornprised of two, distinct subsysterns. The LPCI System is designed to provide ernergency cooling 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 and the Core Spray aystems provide adequate cooling for break areas of approximately 0.2 square feet up to and including the double ended recirculation line break without assistance from the high pressure emergency core cooling subsystems.

The allowable repair times are established so that the average risk rate for repair would be no greater than the basic risk rate, The method and concept are described in reference (1). Using the results developed in this reference, the repair period ir found to be slightly greater than 1/2 the test interval. This assumes that the f

-i (1)

Jacobs, I.M.. " Guidelines for Determining Safe Test Intervals and Repair Times for Engineereo Safeguards", General Electric Co.

A.P.E.D.,

April, 1969' (APED $736).

e l

p.

Amendment No. 57/,80, 152

-124-

t 1

3.5.A BASES.(cont'd.)

Core Spray subsystems and LPCI subsystems constitute a 1 out of 4 system; however, l

i the combined effect of the two systems to limit excessive clad temperatures must also be considered. The test interval specified in Specification 4.5 is 1 month.

Should one Core Spray subsystem become inoperable, the remaining Core Spray subsystem and the LPCI subsystems are available should the need for core cooling arise.

To assure that the remaining Core Spra they r.re verified to be operable immediately.y and LPCI subsystems are available, should the loss of one LPCI pump. occur, a nearly full complement of cora cooling equipment is available.

Thtee LPCI pumps in conjunction with the Core Spray subsystems will perform the core cooling function.

Because of the availability of thz majority of the core cooling equipment, which vill be verified to be operable, a thirty day repair period is justified. If one LPCI subsystem is not available, at l

l 1 east 1 LPCI pump must be available to fulfill the containment cooling function. The 7 day repair period is set on this basis.

B.

PfR service Water system The RilR Service Water System consists of two, distinct subsystems designed to provide heat removal for the containment cooling function. Each RilR Service Water subsystem contains two RHR Service Water booster pumps serving one side of one of two RHR Heat Exchangers, while two RllR (LPCI) pumps serve the other side. The RilR Service Water System operates in conjunction with the RHR~ System to provide the containment cooling function.

The. design of the PJtR Service Water System is predicated upon the use of one RHR Service Water booster pump and one RllR heat exchanger for heat removal after a design Thus, there are ample spares for margin above design conditions.

basis accidetit.. should be avoided and the equipment maintained in a state of Loss of margin operation.

So a 30 day out of service time is chosen for this equipment.

If one loop is out of service reactor operation is permissible for seven days, The requirements for availability of the RHR System for support of the containment cooling function are reflected in the associated action statements for the LPCI System.

With components or subsystems out of service, overati core and containment cooling reliability is maintained by verifying the operability of the remaining cooling equipment.

For routine out of service periods caused by preventive maintenance, etc., the operability of other systems and components will be verified as given in the Technical Specifications. However, if a failure, design deficiency, etc., caused the out of service period then a demonstration of operability may be needed to assure that a similar problem does not exist on the remaining components.

For example, if an out of service period were caused by failure of a pump tv deliver rated capacity, the other pumps of this type might be subjected to a capacity test.

The pump capacity test is a comparison of measured pump performance parameters Amendment No. 80,75,H6,152

,_ =

_u__~.,_____

D 3.5,B BASIS (cont'd.)

r to shop performance tests.

Tests during norm.1 operation will be performed by measuring the flow and/or the pump dischar6e pressure.

These pa rameters. and its power requirement will be used to establish flow at that pressure.

C.

HPCI System l

The litniting 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 Systern (Section VI.).

The HPCI System is provided to assure that the reactor core is adequately cooled l

to limit fuel clad temperature in the event of a small break in the nuclear system and loss.cf. coolant which does not result in rapid depressurization of the reactor vessel.

The itPCI --System permits the reactor to be shut down while maintaining j

sufficient reactor vessel water level inventory until tl.e vessel is depressurized.

The llPCI System continues to operate until reactor vessel pressure is below the l

pressure at which LPCI operation or Core Spray System operation maintains core cooling.

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

The IIPCI pump - is designed to pump 4250 gpm at reactor pressures between 1120 and 150 psig. Two sources of water are available.

Initially, domineralized water from the emergency condensate stcrage tank is used instead-of injectin6 water from the suppression pool into the reactor.

When the HPCI Lystem begins operation, the reactor depressurizes more rapidly than would occur if HPCI was not initiated due to the condensation of stearn by the cold fluid pumped into the reactor vessel by the llPCI System.

As the reactor vessel pressure continues to decrease, the HPCI flow momentarily reaches equilibrium with the flow through the break, Continued depressurization causes the break flow to decrease be.ow the llPCI 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 System.

l 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 System serves as an alternate to the HPCI System only for decay heat removal when feed water is lost. Considering the HPCI System and the ADS plus the RCIC System as redundant paths, reference:(1) methods would give an estimated allowable repair time of 15 days based on the one month testing frequency.

Ilowever, a maximum allowable repair time.of 7 days is selected for conservatium.

The HPCI and RCIC Systems as j

vell as all other Core Standby Cooling Systems must be operable when starting up from a Cold Condition.

It is realized that the HPCI System is not designed to operate l

until reactor pressure exceeds 150 psig and is automatically isolated before the r

I Amendment No. ' N,152 126

3.5.0 MEfa (cont'd.)

reactor pressure decreases below 100 psig.

It is the intent of this specification to assure that when the reactor is being started up from a Cold Condition, the HPCI System is not known to be inoperable, j

D.

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

..lanned operation for periods when the main condenser is unavailable.

The nuclear safety analysin, FSAR Appendix C, shows that the RCIC System provides water to cool l

the fuel when feed s iter is lost. In all other postulated accidents and transients, the ADS provides redundancy for the HPCI System. Based on this and judgements on the reliability of the HPCI System, an allowable repair time of 7 days is specified.

Immediate verifications of HPCI System operability during RCIC System outage is l

considered adequate based on judgement and practicality.

E.

Automatic Depressurization System (ADS)

The limiting conditions for operating the ADS are derived from the Station Nuclear Operational Analysis (Appendix C) and a detailed functional analysis of the ADS (Section VI.).

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 abnnrmalities.

The nuclear system pressure relief system provides automatic nuclear system depressurization for small breaks in the nuclear system so that the LPCI and Core Spray Systems can operate to prot at the fuel barrier.

Because the Automatic Depressurization System does not provide makeup to the reactor primary vessel, no credit 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 the LPCI or Core Spray Systems. There are I

six valves provided and each has a capacity of 800,000 lb/hr at a set pressure of 1080 psig.

The allowable out of service time for one ADS valve is determined as seven days because of the redundancy and because the HPCI System is verified to be operable

,l during this period. Therefore, redundant protection for the core with a small break in the nuclear system is still available.

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

F.

_ Minimum Low Pressure Cooling and Diesel Generator Availability The purpose of Specification F is to assure that adequate core ecoling equipment is available at all times.

It is during refueling outages that major maintenance is performed and during such time that all low pressura core cooling systems may be out Amendment No. 16,146, 152

-127-1

_, ~.

-.. ~

3.5 11111 (cont'd) of service. Specification 3.5.F.4 provides that :,hould this occur, no work will be performed-or 4.e primary system which could lead to draining the vessel. This work would inclod., work on certain control rod drive components and recirculation system.

Thus, the specification precludes the events which could require core coolin6-Specification 3.5.F.5 recognizes that, concurrent with control rod drive maintenance during the refueling outage, it may be necessary to drain the suppression chamber for maintenance or for the inspection required by Specification 4.7. A.2.h.

In this case, if excessive control rod housing leakage occurred, three levels of nrotection against loss of core cooling would exist. First, a special flange would be used to stop the leak.

Second, sufficient inventory of water is maintained to provide, under worst case leak conditions, approximately 60 minutes of core cooling while attempts to secure the leak are made. This inventory includes water in the reactor well, spent funi pool, and condensate storage tank.

If a leak should occur, manually operated valves in the condensate transfer system can be opened to supply either the Core Spray System or the spent fuel pool.

Third, sufficient inventory of water is maintained to permit the water which has drained from the vessel to fill the torus to a level above the Core Spray and LPCI suction strainers. These systems could then

[

recycle the water to the vessel.

Since the system cannot be pressurized during refueling, the potential need for core flooding only exists and the specified combination of the Core Spray or the LPCI subsystems can provide this.

This i

specification also provides for the highly unlikely case that both diesel generators are found to be inoperable. The reduction of rated power to 25% will provide a very stable operating condition.

The allowable repair time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> will provide an oppcrtunity to repair the diesel and thereby prevent the necessity of taking the plant down through the less stable shutdown condition, If the necessary repairs cannot be made in the allowed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the plant will be shutdown in an_ orderly fashion. This will be accomplished while the two off site sources of power required by Specification 3.9.A.1 are available, G.

Maintenance of Filled Discharre Pine If the discharge piping of the Core Spray, LPCI, i!PCI, and RCIC systems are not i

filled, a water hammer can develop in this piping when the pump and/or pumps are started.

If a water hammer were to occur at the time at which the system were required, the system would still perform its design functions, llowever, to minimico damage to the discharge piping and to ensure added margin in the operation of these systems, this Technical Specification requires the discharge lines to be filled whenever the system is in an operable condition.

H.

Engineered Safeguards Compartments Cooling The unit cooler in each pump compartment is capable of providing adequate ventilation flow _and cooling.

Engineering analyses indicate that the temperature rise in safeguards compartments without adequate ventilation flow or cooling is such that continued operation of the safeguards equipment or associated auxiliary equipment cannot be assured.

Amendment No. 57,97/, 152

~

128-x.-

- -. ~ _ _

4.5 MSES 1

Core and Containment Cooling Systems Surveillance Frecuenefes The testing intervals for the core and containment cooling systems are based on 1

industry practice, quantitative reliability analysis, judgement and practicality.

The core cooling systems have not been designed to be fully testable during operation. For example, in the case of the HPCI System, automatic initiation during j

power operation would result in pumping cold water into the reactor vessel, which is not desirable.

Complete ADS testing during power operation causes an undesirabic 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, p.:mps, 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. frequent tests of the - pumps and i

injection valves is deemed to be adequate testing of these systems.

When components and subsystems are out of servico, overall core and containment cooling reliability is maintained by verifying the operability of the remaining equipment.

For routine out of s,ervice periods caused by preventative maintenance, etc., the operability of other systems and components will be verified as given in the Technical Spec.ifications. However, if a failure or design deficiency caused by -

outage, ttan a demonstration of operability may be needed to assure that a generic problem does not exist.

For example, if an out of service period were caused by failure of a pump to deliver rated capacity due to a design deficiency, the other pumps of this type might be subjected to a flow rate test.

i t

l.

Amendment No. 16,146,152 131 l

-.-.-,.,-,.,..-.,,,,.,_.m.-

---

,_.-.-,-__-._,,,-......._...,_.--.__.-._.-_,-_-.,__-~_.---.--.~..,m.-..,~---

L1MITib'G CONDITION FOR OPERATION SURVEILIANCE REOU1HEMENT 3.7.A (cont'd.)

4.7.A (cont'd.)

6.

Low Low. Set Re li e f Func ti.RD 6.

Low-Low Set Relief func1Lqu

a. The low low set function of the

a. The low low set safety /rellei valves safety relief valves shall be opera-shall be tested and calibrated as ble wnen there is irradiated fuel in specified in Table 4.2.B.

the_ reactor vessel and the reactor coolant temperature is h 212*F, and 2 below.pecified in 3.7. A.6.a.1 except as s 1.

With the low low function of one safety / relief valve (S/RV) inopera.

ble, restore the inoperable LLS S/RV to OPERABLE within 14 days or be in the ll0T STANDBY mode within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SIIUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

2.

With the low low set function of both S/RVs inoperable, be in at least 110T STANDBY within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD Sl!UTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. The pressure switches which control the low low set safety / relief salves shall have the following settings.

NB1-PS-51A Open Low Valve 1015 1 20 psig (Increasing)

NBI PS 51B Close Low Valve 875 1 20 psig (Decreasing)

NB1-PS-51C Open Iligh Valve 1025 1 20 psig (Increasing)

B.

Standby Cas Treatment System NBI-PS 51D Close liigh Valve 875 1 20 psig (Decreasing) 1.

At least once per operating cycle the following conditions shall be B.

Standhv Cas Treatment System demonstrated.

1.

Except as specified in 3.7.B.3 be-

a. Pressure drop across the combined low, both Standby Gas Treatment liEPA filters and charcoal adsorber subsystems shall be operable at all banks is less than 6 inches of water times when secondary containment at the system design flow rate.

integrity is required.

b. Inlet heater input i:: capable of 2.a. The results of the in place cold DOP leak tests on the llEPA filters shall reducing R.H. from 100 to 70% R.ll.

show 2991 DOP removal. The results 2,a. The tests and sample analysis of of the halogenated hydrocarbon leak Specification 3.7.3.2 shall be per-tests on the charcoal adsorbert shall show 299% halogenated hydro-formed at least once every 18 months I

for standby service or after every carbon removal. The LOP and haloge-720 hours of system operation and nated hydrocarbon tests shall be performed at a Standby Cas Treatment following significant painting, fire flowrate of $1780 CFM and at a Reac-or chemical release in any ventila-tion zone communicating with the I

tor Building pressure oi s.25" Vg.

system.

Amendment No. 80,P2,152 165-

=

a____m..

LIMITINC CONDITION PQ}LpFERATION SURVEILIANCE REDEIREMEtil 3.7.B (cont'd) 4.7.B (cont'd)

b. The results of laboratory carbon

b. Cold D0p testing shall be perforined sample analysis shall show h99%

after each complete or partial re-radioactive methyl ir,dide removal placement of the llEPA filter bank or with inlet conditions of: velocity after any structural rnaintenance on 3

227 FPM, 21.75 mg/,n inlet methyl the systern housing, iodide concentration, 270% R.!!. and

$30'C.

c. Halogenated hydrocarbon testing shall be perfortned after each com-

c. Each fan shall be shown to provide plete or partial replacernent of the i

1780 CHF 110%.

charcoal adsorber bank or after any structural maintenance on the system housing.

3.

From and after the date that one

d. Each subsystem shall be operated l

l '

made or found to be inoperable for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> every month.

Standby. Cas Treatinent subsystern is with the heaters on at least any reason,- reactor operation is permissible only during the succeed-

e. Test sealing of gaskets for housing l

ing seven days unless such subsystem doors downstream of the llEPA filters is sooner inade operable,_provided and charcoal adsorbers shall be that _ during - such seven days all performed at, and in confortnance

. active components that affect opera-with, each test performed for corn-bility of the operable-Standby Cas pliance with Specification 4.7.B.2.a Treatment subsystem, and its associ-and Specification 3.7.B.2.a.

ated diesel generator, shall be operable.

3.

System drains where present shall be inspected quarterly for adequate Fuel handling requirements are spec-water level in loop seals.

ified in Specification 3,10.E.

4.a. At least once por operating cycle 4.

If these conditions cannot be snet, autoinatic initiation of each Standby procedures shall be initiated imme.

Cas Treatment subsystem shall be diately to establish reactor condi-demonstrated.

tions for which the Standby Cas Treatment System is not required, b.

At least once per operating cycle manual operability of the bypass valve for filter coolin6 shall be detnonstrated.

c.

When one Standby Cas Treatment sub.

system becomes-inoperable, the operable Standby Cas Treatment sub-system shall be verified to be oper-able immediately and daily thereaf-ter.

A demonstration of diesel generator operability is not re-quired by this specification.

C.

Secondary Containment C.

Secondary Containment 1.

Secondary containment surveillance 1,

Secondary _ containment integrity shall be performed as indicated shall be maintained during all modes below:

of plant operation except when all of-the following conditions are met, cendment No.- 80,95,14,1H,.152 165a.

=.

- - - - - -. - ~. -.. -. - _ - _.

1 JJjil?ING CONDITIONS FOR OPERATION SURVEILIANGE REOUIREMENTS 3.7,C (cont'd.)

4.7.C (cont'd.)

a.

The reactor is suberitical and Spec-a.

A preoperational secondary contain.

ification 3.3.A is met, ment capability test shall be con

• ducted after isolating the reactor b.

The reactor water temperature is building and placing either Standby below 212'F and the reactor coolant Gas Treatment subsystem filter train system is vented.

in operation.

Such tests shall demonstrate the capability to main-

)

c.

No activity is being performed which tcIn 1/4 inch of water vacuum under can reduce the shutdown margin below calm vind (2<E <5 mph) conditions that specified in Specification with a filter train flow rate of not 3.3.A.

more than 100% of building volume per day.

(E - wind speed) d.

No irradiated fuel is being handled in the secondary containment and no b.

Additional tests shall be performed loads which could potentially damage during the first operating cycle irradiated fuel are being moved in under an adequate number of differ-the secondary containment..

ent environmental wind conditions to enable valid extrapolation of the e.

If secondary containment integrity test results.

cannot be maintained, restore sec-ondary containment integrity within c.

Secondary containment capability to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or; maintain 1/4 inch of water vacuum under calm wind (2<E <5 mph) condi.

a.

Be in at least Hot Shutdown tions with a filter train flow rate within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and of not more than 100% of building in cold shutdown within the volume per day, shall be demonstrat-following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, ed at each refueling outage prior to refueling,

b..

dling operations in the sec-d.

Af ter a secondary containment viola-Suspend irradiated fuel han-ondary containment, movement tion is determined, the Standby Gas of-loads which could potenti-Treatment System will be operated ally damage irradiated fuel in.

immediately af ter the.affected zones the secondary containment, and are isolated from the remainder of all core alterations and ac-the secondary containment to confirm tivities which could reduce its ability to maintain the remain-the shutdown margin. The pro-der of the secondary containment at

. visions of Specification 1.0.J 1/4 inch of water negative pressure are not applicable, under calm wind conditions.

D.

Primary Containment Isolation Valves D.

Primary Containment Isolation Valves 1.

During reactor power ' operating con-1.

The primary containment isolation ditions, all isolation valves listed valves surveillance shall be per.

in Table 3.7.1 and all instrument formed as follows:

line-flow check valves shall be a.

At least once per operating cycle operable except as specified in the operable isolation valves that 3.7.D.2.

are power operated and automatically initiated shall be tested for simu-lated automatic initiation and clo-sure times.

Amendment No. 49,80,SS.167.152

-166-

3.7.A 6 4.7.A 14ASES (cont'd)

-The pritnary containment is normally slightly pressurized during periods of reactor operation.

Nitrogen used for inerting could leak out of the containment but air could not leak in to increase oxygen concentration.

Once the containment is filled with nitrogen to the required concentration, no monitoring of oxygen concentration is necessary.

However, at least twice a week the oxygen concentration will be determined as added assurance.

The 500 gallon conservative limit on the nitrogen storage tank assures that adequate time is available to get the tank refilled assuming nor a1 plant operation.

The estimated maximum snakeup rate is 1500 SCFD which would require about 160 gallons for a 10 day rankeup requirement.

The norrnal leak rate should be about 200 SCFD.

3.7.A.6 6 4.7.A.6 -LOV LOW SET RELIEF FUNCTION The low low set relief logic is an automatic safety relief valve (SRV) control system t

designed to initigate the postulated thrust load concern of subsequent actuations of SRV's during certain transients (such as inadvertant MSIV closure) and small and intermediate break loss of-coolant accident (14CA) events.

The setpoints used in Section 3.7.A.6.b are based upon a minimum blowdown range to provide adequate time between valve actuations to allow the SRV discharge line high water leg to clear, coupled with consideration of instrument inaccuracy and the main stearn isolation valve isolation setpoint.

The as found setpoint for NBI PS 51A, the pressure switch controlling the opening of RV-71D, must be 51040 psig. The as found closing setpoint for NB1 pS 51B must be at least 90 psig less than 51 A, and must be 2 850 psig.

The as found setpoint for NBI*pS 51C, pressure switch co'., trolling the opening of RV-71F mus,t be s 1050 psig.

The as found closing setpoint for NBI pS 51D aust be at least 90 psig below 510, and must be 2 850 psig.

This ensures that the analytical upper limit for the opening setpoint (10$0 psig), the analytical lower limit on the closing setpoint (850 psig) and the analytical lin t on the blowdown range (2 90 psig) for the Low Low Set Relief Function are not exceeded. Although the specified instrument setpoint tolerance is 1 20 psig, an instrument drift of 125 psig was used in the analysis to ensure adequate margin in determining the valve opening and closing setpoints.

The opening setpoint is set such that, if both the lowest set non LLS S/RV and the highest set of the two LLS S/RVs drif t 25 psig in.the worst case directions, the LLS S/RVs will still control subsequant S/RV actuations. Likewise, the closing setpoint is set to ensure the LLS S/RV closing setpoint remains above the MSIV low pressure trip. The 90 psig blowdown provides adequate energy release from the vessel to ensure time for the water leg to clear between subsequent S/RV actuations.

3.7.B_6 3.7.C STANDBY CAS TREATMENT SYSTEM AND E CONDARY CONTAINMENT The secondary containment is designed to minimize any ground level release of radioactive materials which tnight result from a serious accident.

The reactor building provides secondary containment during reactor operation when the drywell is sealed and in service.

The reactor building provides primary containment when the reactor is shut down and the drywell is open, as-during refueling.

Because the secondary containment is an integral part of-the complete containment system.-secondary containment is required at all times that primary containment is required as well as during refueling, and during movement of loads which could potentially _ damage irradiated fuel in - the secondary containment, Secondary containment may be broken for short periods of time to allow access to the reactor building roof to perform necessary inspections and maintenance, The Standby Gas Treatment System consists of two, distinct subsystems, each containing one exhaust fan and associated filter train, which is designed to filter and exhaust the reactor building atmosphere to the stack during secondary containment isolation conditions.

Both Standby. Gas Treatraent Systern fans are designed to automatically start l

upon containment isolation-and to maintain the reactor building pressure to the design

. negative pressure so that all leakage should be in leakage.

Should one subsystern fail to start, the redundant subsystem is designed to start automatically.

Each of the two fans has 100 percent capacity.

e Amendment No. -38r bW141,147,152 180-

a t

3.7.8 6 3.7.C

]}ASES (cont'd)

I liigh efficiency particulate absolute (HEPA) filters are installed before and after the charcoal adsorbers to ininimize potential release of particulates te the environment and to prevent clogging of the iodine adsorbert.

The charcoal adsorbers are installed to reduce the potential release of radioiodine to the environment. The in. place test results should indicate a system Icak tightness of less than 1 percent bypass leakage for the charcoal adsorbers and llEPA filteta.

The laboratory carbon sainple test results should indicate a radioactive methyl iodide removal ef ficiency of at least 99 percent for expected accident conditions.

If the perfortmance of the llEPA filters and charcoal adsorbers are as specified, the resulting doses will be less than the 10 CFR 100 guidelines for the accidents analyzed.

Only one of the two Standby Cas Treatment subsystems is needed to cleanup the reactor building atmosphere upon contaitunent isolation.

If one subsystem is found to be inoperable, there is no immediate threat to the containment system perfortnance and reactor operation or refueling operation may continue while repairs are being reade.

If both subsystems are inoperable, the plant is brought to a condition where the Standby Gas Treatment System is not required.

I 4.7.B & 4.7.0 BASES Sjandby Cas Treatrent System and Secondary Contalnment i

Initiating reactor building isolation and operation of the Standby Gas Treatment System i

to maintain at least a 1/4 inch of water vacuum within the secondary containment provides an adequate test of the operation of the reactor building isolation valves, leak tightness of the reactor building and performance of the Standby Gas Treatinent System. Functionally I

testing the initiating sensors and associated trip channels demonstrates the capability

- for automatic actuation. Periodic testing gives sufficient confidence of reactor building integrity and Standby Gas Treatment Systets perfortnar.ce capability.

l Pressure drop across the combined llEPA filters and charcoal adsorbers of less than 6 inches of water at the system design flow rate will indleate that the filters and

. adsorbers are not clogged by excessive amounts-of foreign matter.

A 7.8 kw heater is capable of maintaining relative humidity below 70%.

Ileater capacity and pressure drop should be determined at least once per operating cycle to show system perfortnance capability.

The frequency of tests and sample analysis are necessary to show that the llEPA filters and charcoal adsorbers can perforin as evaluated.

Tests of the charcoal adsorbers with halogenated hydrocarbon refrigerant shall be performed in accordance with ANSI N5101980.

The test canisters that are installed with the adsorber trays should be used for the charcoal adsorber efficiency test. Each sample should be at least two inches in diameter and a length equal to the thickness of the bed.

If test results are unacceptable, all t

adsorbent in the system shall be replaced with ar, adsorbent qualified according to Amendment No. 80,82,BS,' O2, IE0,152 i

182 e-~-.,-

,,,--w..-,.4

.,,-,ww

.,.m,v.mn

.,--.,,.-----,.,m-m-w

.--,pyyv.,+.4,---

w-,.,.~.pm.,wyy--,,-..%..,m.,..w.,..my...,-%e,r m.w.- y y n,-,m.,,w - p e b-

4'.7.B & 4.7.C EASES Table 5,1 of ANSI N$09 1980. The replacement tray for the adsorber tray removed for the test should meet the same adsorbent quality. Tests of the HEPA filters with DOP aerosol shall be performed in accordance to ANSI N510 1980.

Any filters found defective shall i

be replaced with filters qualified pursuant to Regulatory Po ition C.3.d. of Regulatory i

Guide 1.52, Revision 2, March, 1978.

All elements of the heatcr should be demonstrated to be functional and operable during the test of heater capacity. Operation of the heaters will prevent moisture buildup in the filters and adsorber system.

j With doors closed and fan in operation, DOP aerosol shall be sprayed externally along the full linear periphery of each respective door to check the gasket seal.

Any detection of DOP in the fan exhaust shall be considered an unar

'able test result and the gaskets repaired and test repeated.

If systera drains are present in the filter /adsorber bas

, loop seals must be used with aderuste water level to prevent by pass leakage from the banks.

If s ignificant painting, fire or chemical release occurs such that the HCPA filter or chat coal adsorber could become contaminated from the fumes, chemicals or foreign material, the same tests and sample analysis shall bc performed as required for operational use.

The dete mination of significance shall be made by the operator on duty at the time of the stuident.

Knowledgeable staff members should be consulted prior to making this deterroination.

Demonstration of the automatic initiation capability and operability of filter cooling is -necessary to assure system performance capability.

If one Standby Gas Treatment i

subsystem is inoperable, the operable subsystem's operability is verified daily.

This substantiates the availability of the operable subsystem and thus reactor operation or refueling operation can continue for a limited period of time.

3.7.D 6 4.7.D BASES Primary Containment Isolation Valves Double isolation valves are provided on lines penetrating the primary containment and open to the free space of the containment.

Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system.

Automatic initiation is required to minimize the potential leakage paths from the containment in

- the event of a loss of coolant accident.

The maximum closure times for the-automatic isolation valves of the primary containment and reactor vessel isolation control system have been selected in consideration of the design intent to prevent core uncovering following pipe breaks outside the primary containment and the need to contain released fission products following pipe breaks inside the primary containment.

These valves are highly reliable, have a low service requirement, and most are normally closed.

The - initiating sensors and associated trip channels are also checked to demonstrate the capability for automatic isolation.

The test interval of once per operating cycle for automatic-initiation Amendment No. 22,102,136, N6,152 183-

mW-LIMITING CONDIT104S FOR OPTRATION SJRE1L1ANCE RPOUIREMENTS 3.10.B (Cont'd) 4.10 (Cont'd)

4. During spiral reload, SRM operability will be verifAed by using a portable external source every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> until the required amount of fuel is lonied to maintain 3 cps. As an alternative i

to the above, two fuel assemblies will be loaded in different cella contain.

ing control blades around each SRM to obtain the required 3 eps.

Until these two assemblies have been loaded, the 3 cps requirement is not neces.

sary.

C. Spent ruel Pool Water Level Whenever irradiated fuel is stored in the spent fuel pool, the pool water C.

Spent _ruel Pool Water Level level shall be maintained at or above 8h' above the top of the fuel.

When irradiated fuel is stored in D. Time Limi. tail 2D the spent fuel pool, the water level

. Irradiated fuel shall not be handled shall be recorded daily.

-in or above the reactor prior to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reactor shutdown.

E. Standby Cas Treatment Sverem E.

St.andby Cas Treatment System From and after the date that one j

Standby Cas Treatment subsystem is When one Standby Cas Treatment sub.

made or found to be inoperable for any system becomes inoperable, the oper.

j reason, handling of irradiated fuel, able Standby Cas Treatment subsystem and movement of loads which could shall be v*erified to be operable potentially damage irradiated fuel in immediately and daily thereafter. A the secondary containment is permissi.

demonstration of diesel generator ble only during the succeeding seven operability is not required by this

}

days unless such subsystern is sooner specification, made oparable, provided that during such seven days all active components that affect operability of the opera-ble_ Standby Cas Treatment subsystem, and its associated diesel generator, shall be operable.

At least one diesel generator shall be operable during fuel handling opera-tions. This one diesel shall be capa.

ble of supplying power to an operabic i

Standby Cas_ Treatment subsystem.

F. Core standby Cooling Systems x

During a refueling outage, refueling i

operation with fuel in the vessel may continue with one Core _ Spray and one LpC1-subsystem inoperable, or - with l.

both Core Spray subsystems inoperable.

Refueling is permitted with the sup.

pression chamber-drained provided an operable Core' Spray or LPCI subsystem

-of RHR is aligned to take a ru tion on the condensate storage tank containing at least 150,000 gallons (214 ft.

indicated level).

Amendment No. 61W46rM;'r 152 205a.

9.

1 3.10 BASES (Cort'd)

D.

U ge Lireitatin The radiological consequences of a fuel handling accident are based upon the accident occurring at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ter reactor shutdown.

E.

Siendby Cas TreatamLjiy11m Only one of the two Standby Cas Treatment subsysterns is needed to cican up the reactor building atmosphere upon containment isolation.

If one subsystein is found to be inoperable, there is no immediate threat to the ec~tainment system performance and refueling operation may concinue while repairs are being nade.

If both subsystems are inoperable, the plant is brought to a condition where the Standby Cas Treatment System is not required.

F.

Core Standby Cooline 5.3rdcF.1 During refueling the systern cannot be press.urized, so only the potential need for core flooding exists and the specified cotabination of the Core Spray or LpCI subsysterns can provide this. A more detailed discussion is contained in the bases for 3.5.F.

G.

Control Room Air Treatment If the system is found to be inoperable, there is no immediate threat to the control room and refueling operation may continue for a limited period of time while repairs are being made.

If the system cannot be repaired within seven days, tefueling operations will be terminated.

11.

Spent Fuel Cask liandline The operttion of the redundant crane in the Restricted Mode during fuel cask handling operations assures that the cask remains within the controlled area once it has been removed from its transport vehicle (i.e., once it is above the 931' elevation).

llandling of the cask on the Refueling Floor in the Unrestricted Mode is allowed only in the case of equipment failures or emergency conditions when the cask is already suspended. The Unrestricted Mode of operation is allowed only to the extent necessary to get the cask to a suitable stationary position so the required repairs can te inade.

Operation with a failed controlled area rnicroswitch will be allowed for a 48-hour period providing an Operator is on the floor in additio.1 to the crane operator to assure that the cask handling is limited to the controlled area as marked on the floor.

This will allow adequate time to make repairs but still will not restrict cask handling operations unduly.

4.10 EASES A.

Reigeling Interlocks Cotaplete functional testing of all refueling interlocks before any refueling outage will provide positive indication that the interlocks operate in the situations for which they were designed.

By loading each hoist with a weight equal to the fuel assembly, positioning the refueling platform and withdrawing control rods, the interlocks can be subjected to valid operational tests.

Where redundancy is provided in the logic circuitry, tests can be performed to assure that each redundant logic element can independently perform its i

functions.

Amendment No. 35,61,97/, 152 209a-

I s

4.10 BASES (Cont'd)

B, Core Monitoring i

Requiring the SRM's to be functionally tested prior to any core alteration assures that the SRM's will be operable at the start of that alteration.

The daily response check (or 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> check for spiral reload) of the SRM's ensures their continued operability.

E.

Standby Cas Treatpent System Only one of the two Standby Cas Treatment subsystems is needed to clean up the reactor building atmosphere upon containment isolation.

If one subsystem is f ound to be inoperable, there is no immed' ate threat to the containment system performance and refueling operations may continue while repairs are being made.

If both subsystems are inoperable, the plant is brought to a condition where the Standby Gas-Treatment System is not required, 11.

Spent Fuel Cask llandline The Surveillance Requirements specified assure that the redundant crane is adequately inspected in accordance with the accepted tNSI Standard (B.30.2.0) and manufacturer's recommendations to determine that - the equipment is in satisfactory _ condition.

The testing of the controlled area limit switches assures that the crane operation will be limited to the designated area in the Restricted Mode of operation. The test of the "two block" linit switch assures the power to the hoisting motor will be interrupted before an actual "two blocking" incident can occur. The test of the inching hoist assures that this mode of load control is available when required.

Requiring the lifting and holding of the cask for 5 minutes during the initial lif t of each series of cask handling operations puts a load test on the entire crane lifting mechanism as well as the braking system.

Performing this test when the cask is being lifted initially from the cask car assures that the system is operable prior to lifting the load to an excessive height.

f' b

Amendment No. 35,6f,97/,152 209b.

1

--.~.-m__,..-.m-.,-.----

4

- ~

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

.,,_v.-,,,.-.n

y t

CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS t

3.12 (cont'd) 4.12 (cont'd)

B.

Reactor Equitment Cooline (REC)

B.

Reactor Eaultment Cooling,,,JEf41 Syggtg Svatem 1.

Both Reactor Equipment Conling sub-1.

REC System Testing systems and their associuted pumps i

shall be operable whenever irradiat.

ltca Freauency ed fuel is in the vessel or the spent fuel pool, except as specified a.

Pump Operability Once/*'.onth in 3.12.B.2 and 3.12.B.3 below, b.

Motor operated Oncy/ Month Valve Operability c.

Pump flow rate once/3 months Each pump shall and after pump deliver 1175 gpm maintenance at 65 paid, d.

System head tank Daily level shall be monitored.

2.

From and af ter the date that any 2.

When it is determined that any ac.

component-that affects operability tive component that affects_ opera-in one RBC subsystem becomes inop-bility of an REC subsystem is inop+

etable, continued reactor operation erable, all active components that is permissible during the succeeding affect operability of the operable thirty days provided that during REC nubsystem shall be verified such thirty. days all the active operable immediately and weekly components that affect operability thereafter.

of the operable REC subsystem, the active components that affect opera.

bility of the engineered safeguards compartment cooling systemt, the diesel generator associated with the i

operable subsystem are operable.

The allowable repair time does not apply when the reactor is in the shutdown mode and reactor pressura is less than 75 psig.

3.

Both REC subsystems with one pump per subsystem shall be operable as stated in 3.12.B.1-- and 3.12.B.2 above during reactor head of f opera-tions requiring LPCI or Core Spray j

system availability or Service Water cooling shall be available.

4.

If the requirements of 3.12.B.1 through 3.12.B.3 cannot be met, the reactor shall be shutdown in an orderly manner and in the Cold Shut-down condition within 24. hours or operations requiring LPCI or Core Spray system availability shall be halted.

Atiendment No. 80,U6,152 215b.

e w

LIMITING CONDITIONS FOR OPERATION Sl%VJ,JLMSCE REOUIREMFNTS 3.12 (cont'd) 4.12 (cont'd)

C.

$_treice Water System C.

S.tyyice Water System

]

1.

Both Service Water sybsystems with 1.

Service Water System Testing both pumps in each subsystem shall be operable whenever irradiated fuel Ifug Functional is in the vessel or spent fuel pool and prior to reactor startup except a.

Pamp Operability once/ Month an specified in 3.12.C.2 below.

b.

Motor Operated once/ Month Valve Operability c.

Pump discharge Once/3 months head tests 2.

From and after the date that any 2.

When it is determined that any re-active component that affects opera-quired Service Water System compo-bility of one Service Water subsys-nent is inoperable, all active com-t,em is made or found to be inopera-ponents that affect opersbility of ble for any ceason, continued reac-the operable Service Water subsystem l

tar operation is permissible during components shall be verified to be the succeed hg thirty days provided operable inmediately and weekly

r. hat durie.6 such thirty days all thereafter.

active components that affect opera-bility of the opereblo Service Water subsystem and its associated diesel generator are operable.

3.

If the requirement uf 3.12.C.1 and 3.12.C.2 cannot be met, an orderly shutdown of the reactor shall be initiated and the reactor shall be in the Cold Shutdown condi. ion with-in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

D.

Batterv Room Ventilation D.

Battery Room Ventilatf ort 1.

Battery room ventilation shall be 1.

The spare battery room ventilation operable on a continuous basis when-fan shall be checked for operability ever specification 3.9. A is required once/ week, to be satisfied.

2.

-Fron and after the date that either

- of te two battery room vent fans it.

made or found to be inoperable for any reason, continued reactor opera-tion is permissible during the suc-ceeding 7 days.

3.

If the requirements of 3.12.D 1 & 2 cannot be met, an orderly shatdown of the reactor chall be initiated and the reactor shall be in Cold Shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Amendment No. 16,1#6,152

-215c-

o

.c 7,12 BASES A.

Main Control Room Ventilation Sysits-The control room ventilation system is designed to filter the control room atmosphere for intake air and/or for recirculation during control room isolation conditions.

The system is designed to automatically start upon control room isolation and to maintain the control room pressure to the design positive pressure so that all leakage should be out leakage.

High efficiency particulate absolute (HEPA) filters are installed before the charcoal adsorbers to prevent clogging of the iodine adsorbers.

The charcoal adsorbers are installed to reduce the potential it. cake of radioiodine to the control room. The in-place test results should indicate a system leak tightness of less than 1 percent bypass leakage for the charcoal adsorbers and HEPA filters.

The laboratory carbon sample test results should indicate a rad.

.ac. tve methyl iodide removal efficiency of at least 99 percent for expected accis.n:

(..aditions.

If the performaace of the HEPA filters and charcoal adsorbers ar. <,

-pecified, the resulting doses will be less than the allowable leve.

.. ed in criterion 19 of the General Design Criteria for Nuclear Power Plants, Appendix A to 10 CFR Part 50.

If the system is found to be inoperable, there is no immediate threat to the control room and reactor operation may continue for a limitr. period of time while repairs are being made.

If the system cannot be repaired within seven days, the reactor is shutdown and brought to cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

B.

Reactor Eautoment Cooling (REC) System The Reactor Equipment Cooling System consists of two, distinct subsystems, each containing two pumps and one heat exchanger. Each subsystem is capable of supplying the-cooling r&quirements of the essential services followin5 design accident l.

conditions with only one pump in either subsystem.

[

The REC System has additional flexibility provided by the capability' of intorconnec-tion of the two subsystems and the backup water supply to the critical cooling loop by.the Service Water System. This flexibility and tne need for only one pump in one critical cooling loop to meet the design accident requirements justifies the 30 day repair -time during normal operation and the reduced requirements during head +off

-l operations requiring the availability of the LPCI or Coro Spray *ystems.

C.

Service Water System

}.

The Service Water System' consists of two, distinct subsystems, each containing two vertical Service Water pumps located in the ' intake structure, and. associated strainers, piping, valving and instrumentation.

The pumps discharge to a common header from which independent piping supplies two seismic Class I cooling water loops and one turbine building loop. Automatic valving is provided to shutoff all supply to the turbine building loop on drop in header pressure thus assuring supply to the

'j-Seismic Class I loops each of which feeds one diesel generator, two RHR Service Water Amendment No. 82,88,102,152

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e 3.12 BASES (cont'd) l booster pumps, one control room basement fan coil unit and one REC heat exchanger.

Valves are included in the common discharge header to permit the Seismic Class 1 l

Service Water System -to be operated as two independent - subsystems.

The heat exchangers are valved such that - oney can be individually backwashed without interrupting system operation.

During normal operation two or three pumps will be required.

Three pumps are used for a normal shutdown.

[

The loss of all a-c power will crip all operating Service Water pumps. The automatic emergency diesel generator start system and emergency equipment starting sequence I

will then start one selected Service Water pump in 30-40 seconds.

In the meantime, the_ drop in Service Water header pressure will close the turbine building cooling water isolation valve guaranteeing supply to the reactoc building, the control room l

basement,. and the diesel generators from the one Service Water pump.

Due to_the redundance of pumps and the requirement of only one to meet the accident requirements, the 30 day repair time is justifiad.

.D.

Batterv Room Jegitilation The temperature rise and hydrogen buildup in the battery rooms without adequate ventilation is such that continuous safe operation of equipment in these rooms cannot be assured, 4.12 EASES A.

Main Control Room Ventilation System Pressure drop across the combined HEPA filters and charcoal adsorbers of less than 6 inches of water at the system design flow rate will indicate that the filters and adsorbers ' are not clogged by excessive amounts of foreign matter.

Pressure drop should be-determined at least once per operating cycle to show system performance capability, Tests of the charcoal adsorbers with halogenated hydrocarbon refrigerant should be performed in accordance with ANSI N510 1980.

The frequency of tests and sample analysis are necessary to show that the HEPA filters and charcoal adsorbers can perform as evaluated. The test canisters that are installed with the adsorber trays should be used for the charcoal adsorber efficiency test.

Each. sample should be at least two inches in diameter and a length equal to the thickness of-the bed.

If test results are unacceptable, all adsorbent in the system snall be replaced with an adsorbent qualified according to Table 5.1 of ANSI

~

N509-1980. The replacement tray for the absorber tray removed for the test should meet the same adsorbent quality.

Tests of the HEPA filters with DOP aerosol shall

. be performed in accordance to ANSI N510-1980. Any HEPA filters found defective shall be replaced with filters qualified pursuant to Regulatory Position C.3.d of Regulatory Guide 1.52.

L Operation of the system for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> every month will demonstrate operability of the L

filters and adsorber system and remove excessive moisture built up on the adsorber.

Amendment No. 82,136,152

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LIMITING CONDITIONS FOR OPEPATION SURVEILLANCE REOUIREMENTS 3.22 SPECIAL TESTS /EXCEP1 E (CONT'D) 4.22 SPECIAL TESTS / EXCEPTIONS (CONT'D) 2.

Rod Sequence control System 2.

~ 'h e n the constraints imposed on

=

(RSCS) control rod groups by the RSCS' are bypassed, verify:

The sequence constraints im-posed on control rod groups a.

That the RWM is OPERABLE.

by the RSCS may be suspended by means of the individual b.

Conformance with this specifi-rod position bypass switches cation and procedures by a sec-or jumpers, provided that the ond licensed operator or other rod worth minimizer is OPERA-qualified employee.

BLE, for this and the follow-ing special tests, a.

Control rod scram timing.

b.

Control rod friction mea-surements, c.

Startup test program with thermal power less than 20% of rated thermal pow-er.

If. the above requirement is not satisfied, the RSCS shall be operable.

3.

RRR System The RHR system may be aligned in the shutdown cooling mode with at least one shutdown cooling mode loop OPERABLE vhile performing the Shutdown Margin Demonstration.

4.

Containment Systems Primary containment is not required while performing the

- Shutdown Margin Demonstration when reactor water tempera-ture is equal to or less than 212*F.

B.

Trainine Startun g

g 1.

LPCI System The reactor vessel shall be verified The LPCI System is required to be t be unpressurized and the thermal operable with the exception that the Power verified to be less than 1% or, RHR system may be aligned in the rated thermal power at lens once l

shutdown cooling mode while perform-per h ur during training startups.

ing training startups at atmospheric pressure at power levels less than 1%-of rated thermal power.

Amendment No. 97/,152

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