ML20029B353

Amend 77 to License DPR-22,revising Pump & Valve Surveillance Testing Requirements to Be Consistent W/Asme Code Requirements & Eliminating Requirements for Immediate Surveillance Testing of Redundant Inoperable Equipment

ML20029B353
Person / Time
Site:	Monticello
Issue date:	02/15/1991
From:	Marsh L Office of Nuclear Reactor Regulation
To:
Shared Package
ML20029B354	List: ML20029B353 ML20029B355
References
NUDOCS 9103070089
Download: ML20029B353 (37)

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UNITED STATES <

NUCLEAR REGULATORY COMMISSION -

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NORTHERN STATES POWER COMPANY DOCKET NO. 50 263 MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO FACILITY OPERATING LICENSE u

'm Amendment No. 77-i c

License No. DPR-22 h

1.-

The Nuclear Reguletory Commission (the Comission) has found that:

1 L

A.

The application' for amendment by Northern States Power Company (the

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licensee)datedNovember 13, 1990,-complies with the standards-and requirements'of the Atomic Energy Act of 1954, as amended (the Act),

l and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The f acility will operate in conformity with the application, the-provisions of the Act, and the rules and' regulations of the Commission; l

C.-

Thereisreasonableassurance(1)thatthe.activitiesauthorized by this-amendment-can'be conducted without endange*ing'the-

.t 11 health and safety of the public, and (ii) that such activities

will be conducted in compliance with the Commission's: regulations; 1:

o D.

-The issuance of this amendment will not be inimical to the common-L-

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' L

of the Comission's regulations'and all: applicable requirements have been satisfiedL Lh 2.-

.Accordingly, the license is amended by. changes to the' Technical Specifica-

.tions as indicated in the attachment to this license amendment, and para--

Egraph 2.C.2.of Facility Operating-License No. DPR-22-is hereby amended to-read as follows::

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!TechnicaliSpecifications-

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.The' Technical Specifications contained-in Appendix A, as-revised through Amendment <No. 77, are hereby incorporated in the license.-

The-licensee-shall operate the facility in accordance with the t

Technical Specifications.

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.3; : This license amendment is effective' as of the date of it.s issuance. _

l FOR THE NUCLEAR REGULATORY COMMISSION' S

L.'B. Marsh, h;setor.

9 Project Directorate 111-1 Division of Reactor Projects III/IV/V

0ffice of Nuclear Reactor Regulation

Attachment:

TChanges to the Technical

-i

-Specifications

- Date offissuance:( February 15, 1991

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-, ;3 LATTACHMENT-TO LICENSE AMENDMENT NO.=77 FACILITY OPERATING-LICENSE NO.0PR-22

- DOCKET NO. 50-263:

- Revise AppeiMix A Technical-Specifications by removing the pages identified below and inst:rting the attached'pages.

The-revised pages are identified by amendment number and contai_n marginal lines. indicating the area of change.

- Remove Insert l

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h3 4: and. 4.4 '- Standby L' quid Control System -

93' L

~A.- Normal operation 93 B.. Volume Concentrar. ion Requirements 95 l

3.41and 4,4 Bases 99

3.5.and 4.5 Core and containment. cooling Systems 101 i

A.: Core; Spray System 101-B.-

LPCI: Subsystem-103

C.

1000 Service Water System -

106 D.-

HPCI' System 108 E; ' Automatic Pressure Relief System 109 Fl RCICcSystem

'111 C.: Minimum Core;and Containment-Cooling

_ l:

System' Availability:

113 H._

Recirculation-System 114

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Deleted; p

3.5: Bases -

115 4.5. Bases 120

3i6LandL4 61 Primary System Boundary 121 A L_. Reactor-Coolant Heatup and.Cooldown 121-

'B. ' Reactor Vessel Temperature and Pressure

~122 1

- C,' Coolant Chemistry:

123 D.

Coolant Leakage:

126

'E.

Safety / Relief Valves 127 JF.

Deleted.

C. iJe t Pumps :

128

.H.

Snubbers' 129'-

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+,. :

3.6 and 4.6 Bases.

~144 H

$3.7 Land 4.7 L Containment Systems-156:

1

-~ A.'. ^ Primary -Containment 156-TJ

g B.;' Standby Cas Treatment-System 166 C. -Secondary Containment 1169

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D; : Primary' Containment; Isolation 1 Valves' 11701 j

s.E.

Combustible Cas Contro1~' System-172 1

-3.7-Bases 1751

' 4.'71 Bases 183-e Amendment.No.yjy7 47, 3

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'3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIANCE. REQUIREMENTS

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3.4 STANDBY LIOUID CONTROL SYSTEM 4.4 STANDBY LIOUID' CONTROL SYSTEM Applicability:

Applicability:

Applies to the operating status of the Applies to the periodic testing require-standby 11 quid control system.

ments for the standby 11guld control system.

Objective Objective:

To assure the availability of an To verify the operability of the standby independent reactivity control mechanism.

11guld control system.

SPECIFICATION:

SPECIFICATION A.

System Operation A.

The ope' ability of the ' standby liquid r

control system shall be verified by 1.

The standby liquid control system performance of the following tests:

shall be operable at all times when fuel is in the reactor and the 1.

At least ance per month -

reactor is not shut down by control rods, except as.specified in 3.4.A.2.

Demineralized water shall be i

recycled to the test tank.

Pump 2.

From and after the date'that a redun-minimum flow rate of 24 gpm shall dant component is made or found to be be verified against a system head g

inoperable, reactor operation is permissible-of 1275 psig. Comparison of the g

only during the following 7 days provided measured pump flow rate against 3

that the redundant component is operable.

equation 2 of paragraph 3.4.B.1 shall

'l be made to demonstrate operability of the system in accordance with the AWS Design Basis.

2.

At least once durinz each operatine evcie -

a.

Manually initiate one of the two standby liquid control systems and pump demineralized water into the reactor vessel.

This test checks explosion of the charge associated with the tested system. proper operation of the valves and pump capacity.

Both systems shall be tested and inspected, including each explosion valve in the course of two operating cycles.

3.4/4.4 93 Amendment No. 56, 57.

3.0 ' LIMITING CONDITIC.tlS FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS b.

Explode one of two primer assemblies manufactured in the same batch to verify-proper function, Then install, as a replacement, the second primer assembly in the explosion valve-of the system tested for operation.

c.

Test that the setting of the system pressure relief valves is between 1350 and 1450 psig.

L 3.4/4.4 94 kiendment No. 56, 77

3.0 LIMITINC CONDITIONS FDR OFERATION 4.0 SURUEYLIANCE REQUIREMENTS '

l B.

Boron Solution Requirements B.

Boron Solution Surveillance l-At all times when the Standby Liquid Control The availability of the proper boron System is required to be operable:

bearing solution shall be verified by.

1.

The liquid poison tank shall contain a boron bearing solution that satisfies the volume, 1.

At least once per cycle -

concentration and enrichment requirements of Figure 3.4.1, or compliance can be demon-Boron enrichment shall be determined.

strated by satisfying the following equations:

The laboratory analysis to determine enrichment shall be obtained within Equation 1 (Original Design Basis):

30 days of sampling or chemical addition.

71.18 4821 1 (.0.0051xC + 0.998}I *1101-E (19.8 (100

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C

+ 128 gal Boron concentration shall be determined.

In addition, the boron concentration Equation 2 (ATUS Design Basis):

shall be determined any time water or 86 boron are added or if the solution C > 8.28(Q )(19.8)

E temperature drops below the limits specified by Figure 3.4.2.

where:

3.

At laast once per day -

V-indicated Boron solution tank volume (gal)

E - measured Boron solution enrichment (atom %)

a.

Solution volume shall be checked.

C - measured Boron solution concentration (ut%)

Q - measured pump flow rate (gpm) at 1275 psig b.

The solution temperature shall be checked.

If Equation-1 is satisfied but Equation 2 cannot be met, continued plant operation is c.

The room temperature shall be permissible, provided that:

checked in the vicinity of the standby liquid control system a.

Compliance with Equation 2 is demonstrated pumps.

within 7 days or b.

The Commission shall be notified and a special report provided outlining the actions taken and the plans and schedule for demonstrating compliance with the ATUS Design Basis.

2.

The temperature shall not be less than the solution temperature presented in Figure 3.4.2.

3.

The heat tracing on the pump suction lines shall be operable whenever the room temperature is less than the solution temperature presented in Figure 3.4.2.

3.4/4.4 95 Amendment No. 56. 57. 77

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-3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS lC.

If Specification 3.4.A.through B -

are not: met. an orderly ' shutdown shall be initiated and the reactor shall be in flot Shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b 3.4/4.4 96-Mendment No. 56, 77

Basis 3.4 and 4.4:

The design objective of the standby liquid control system is to provide the capability of bringing the A.

reactor from full power to a cold, xenon-free shutdown assuming that none of the withdrawn control rods can be inserted. To meet this objective,. the liquid control system is designed to inject a quantity of boron which produces a concentration of boron in the reactor core in less than 125 minutes sufficient to bring the reactor from full power to a 31 delta k suberitical condition considering the hot to cold reactivity swing, xenon poisoning and an additional 25% boron concentration margin for possible imper-feet mixing of the chemical solution in:the reactor water and dilution from the water in the cooldown circuit.

The time requireme%t'(125 minutes) fer insertion of the boron solution was selected to override the rate of reactivity insertion due to cooldown of the reactor following the xenon poison peak.

The ATUS Rule (10CFR50.62) requires the addition of a new design requirement to the generic SLC System design basis.- Changes to flow rate, solution concentration or boron enrichment to meet the ATUS Rule do not invalidate the original system desige basis.

Paragraph (c)(4) of 10CFR50.62 states that:

"Each bolling water reactor must have a Standby Liquid Control System (SLCS) with a minimum flow capacity and boron 'cantent equivalent in control capacity to 86 gallons per minute of 13 weight percent sodium pentaborate solutson" (. natural boron enrichment).

The described minimum system parameters toquivalent to 24 gpm, 10.7% concentration and 55 atom percent Boron-10 enricheerit) will ensure an equivalent injection capability that meets the ATUS rule requirement.

Boron enrichment concentration, solution temperature, and volume (including check of tank heater and pipe heat tracing system) are checked on a frequency to assure a high reliability of operation of the system should it ever be required. Only one of the two standby liquid control pumping circuits is needed for proper operation of the system.

If one pumping circuit is found to be inoperable, there is no immediate threat to shutdown capability, and reactor operation may continue while repairs are being made.

A reliability analysis indicates that the plant can be operated safely in this manner for ten days.

For additional margin, the allowable out of service time has been reduced to seven days.

The only practical time to test the standby liquid control system is' during a refueling outage and by initiation from local stations. Component:- of the system are checked periodically as described above and make a functional test of the entire system on a frequency of less than once each refueling outage unnecessary. A test of explosive charges from one manufacturing batch is made to assure that the replacement charges for the tested system are satisfactory. A continual check of the firing circuit continuity is provided by pilot lights in the control room.

The relief valves in the standby liquid control system protect the system piping and positive displacement pumps which are nominally designed for 1500 psi from overpressure. The pressure relief valves discharge back to the standby liquid control solution tank.

3.4/4.4 BASES 99 Amendment No. 56, 57, 77

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c B.

The solution saturation temperature varies.with the concentration of-sodium' pentaborate.

The' solution s

v111~ be maintained' at least 5*F above: the. saturation temperature within -the' tank and suction piping to

. guard againstl precipitation.. The.5*F margin'.is included in: Figure 3.4.2.

Temperature.and liquid level.

alarms for..the system are annunciated in the: control room.

Pump operability is checked:en :a frequency to ' assure:a high reliability of_ operation of the system should it'ever be required..

l Once the salution has been made up, boron concentration will not -vary unless more boron or more water is added. : Level' indication'and alarm Indicate whether.the solution volume has changed which'might indicate a possible. solution' concentration change.- Boron Enrichment will.not. vary unless'more Boron is added. - No deterioration of-the Boron-10 enrichment level should ' occur =- during system ' standby operation.

considering these factors the ' test intervals have been established.

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Anendment No. 56, 77 L

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am m-3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIANCE REQUIREMDITS 3.5 Cj]RE AND CONTAINMENT COOLING SYSTEMS 4,5 CORE AND CONTAINMENT COOLING SYSTEMS Applicability:

Applicability:

Applies to the operational status of the Applies to periodic testing of the emergency emergency cooling systems.

cooling systems.

l Objective:

Obiectivel To insure adequate cooling capability for heat To verify the operability of the emergency removal in the event of a loss of coolant cooling. systems.

accident or isolation from the normal reactor heat sink.

Specification:

Specification:

Low Pressure Core Cooline Capability Low Pressure Core Cooline Capability A.

Core Spray System A.

Surveillance of the core spray system shall be performed as follows:

1.

Except as specified in 3.5.A.2.,

3.5.A.3.,

and 3.5 A.S. below, both core 1.

Testing spray subsystems shall be operable when-ever irradiated fuel is in the reactor Item Frequency vessel and reactor coolant water tempera-Simulated automatic Each refueling ture is greater than 212*F.

actuation test outage l

3.5/4.5 I01 kiendment No. 77

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 - SURVEILIANCE REQUIREMENTS l

i IcJm Frecuency l

l Pump Operability' Pursuant to Specification 4.15.B Valve operability Pursuant to Specification 4.15.B Core Spray heyfer Ao instrumentation Check Once/ day Test Once/ month-Calibrate Once/3 months 2.

From and after the date that one of the core spray systems is made or found to be inoperable for any reason, reactor opera-tion is permissible only during the suc-l ceeding 7 days unless such system-is sooner made operable,'provided that l

during such 7 days all active'compo-nents of the other core spray system and the LPCI mode of the RilR system and the diesel generators' required for operation of such components (if no external source of power were available)'shall be operable.

3.

From and after the date that both core spray systems are made or found to be inoperable for any reason, reactor 3.5/4.5 Amendment.102pa. 77

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-3.0 -LIMITING" CONDITIONS.FOR OPERATION

.- 4.0 ~ SURVEII.IANCE REQUIREMENTS operationJis permissible only 'during

.l Ithe succeeding 72. hours'unless at7

-least. one;- of, such systems 11s ' sooner:

made operable,:provided that'during-

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-such.72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> all actEveLcomponents..

of. the LCPI mode :of RHR systen and the.

. diesel generators required!for operation-of'such components:(if no external ~l source

of' power werelavaLlable)'shall be.: opera--

'ble.

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4.

Each corerspray system shall be capable i.

. of delivering 3,020.. gpm against a reactor

_ pressure of' 130 psig.-

If this rate;of-l delivery requirement.cannot be met; the-system shall:be considered inoperable.

3.

If the requirements of 3.5.A.1-3 cannot be met ' an orderly shutdown of. the' reacter will be initiated and the reactor water temperature. shall-be reduced to less than j'

212*F within~24' hours.

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3.0 LIMITING. CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS B.

Low Pressure Coolant Injection (LPCI)

B.

Surveillance of the Low Pressure. Coolant Subsystem (LPCI Mode of RIIR System)

Injection (LPCI) Subsystem (LPCI Mode of RIIR System) shall be perforced as follows:

1.

Except as specified in 3.5.B.2 and 1.

Testing 3.5.B.3 below, the LPCI shall be operable whenever irradiated. fuel Item Frecuency is in the reactor vessel and reactor.

Pump Operability Pursuant to

' coolant temperature is greater than Specification 4.15.B 212*F.

Valve operability Pursuant to Specification 4.15.B Cycling of RIIR Cnce/ Quarter Intertie Line Valves i

Simulated automatic Every refueling actuation test outage 2.

From and af ter the date that one of the 2.

Deleted LPCI pumps or admission valves is made or found to be inoperable for any reason, reactor operation is permissible only l

during the. succeeding 30 days for an I

inoperable LPCI pump or the succeeding 7 days for an inoperable admission valve, unless such pump or admission valve is sooner made operable, provided that during such 30 ' days for an inoperable LPCI pump or during such 7 days for an inoperable admission valve the remaining active components of the LPCI and containment cooling sub-system and all active components of both core spray systems and the diesel genera-

.cors required for operation of such com-ponents (if no external source of power were available) shall be operable.

3.5/4.5 104 Amendment No. 27, 77

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4.0. : SURVEILIANCE REQUIREMENTS 3.

From.and after'the date that'two'of ihe LPCI 3.: 1 Deleted

. pumps or admission valves'are made:-orJfound.-

to'be Inoperable for anyfreason,'(reactor, operation is' permissible only"during the-succeeding 7 daysefor'twol inoperable;LPCI' pumps.

or the' succeeding.72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> L for two inoperable

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'addmission valves unlesslsuch pumps or-admission ~ valves are made. operable souier, provided that:duringisuchT71 days (for two inoperableeLPCI' pumps.or during'such.72' hours.

'for-two[ inoperable admission valves all'

. active components of ' both ' core.. spray sys- -

tems, the: containment cooling Esubsystem

.(including 2.LPCILpumps) and:the diesel ~

generators required for. operation of such components '(if no external source fof; power ?

were1available).shall be operable.

4.

A maximum of one. drywell spray loop (contain-

4.. During each five year period, an

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ment cooling' mode'of.RilR) may be; inoperable air test shall be performed'on the l

for 7. days.when.the reactor water tempera-g drywell spray headers :and nozzles.

ture is greater-than: 212*F. ~ If. theiloop

.l' is not returned to; service within 7 days,-.

the orderly shutdown of the reactor will be initiated'and"thelreactorfwater_temperiture i

shall be reduced ~ to less than 212* F.2 i

S.

Each LPCI. subsystem -(R11R) ~ pump 'shall L be.

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capable of delivering 4,000.gpm 110%.

against a system. head correspondingLto-

-.three pumps delivering '12,000 gpm. at a reactor pressure of 20 -psi above the suppression chamber pressure. 'If;this i

rate. of delivery requirement cannot' be -

met.1the. pump shall be considered inoper--

able.

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3.5/4.5 105 Amendment No. 27, 77, l

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3.0 LIMITING CONDITIONS.FOR OPERATION _

4.0 SURVEILLANCE REQUIREMENTS Containment Coolint Capability Containment Cooline Capability C.

Residual IIeat Removal (Rl!R) Service Water C.

Surveillance of the' R11R service water System system shall be performed as fellows:

1.

Except as specified in 3.5.C.2 and 3.5.C.3 1.

Testing below, both Ri!R service water system loops shall be operable whenever irradiated fuel Item Frecuency is in the reactor vessel and reactor coolant Pump and valve Pursuant to temperature is greater than 212*F.

operability Specification 4.15.B 2.

From and after the date that one of the RIIR service water system pumps is made or found to be inoperable for any reason, reactor. operation is permissible only during the succeeding thirty days unless such pump is sooner-made operable, pro-vided that during such. thirty days all other active components of the R11R service water system are operable.

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3.0 { LIMITING. CONDITIONS-FOR' OPERATION 4.0". SURVEILLANCE REQUIREMENTS

3. ~From and-afterJthe'date that oneJofithe~~

'RHR service water systems'is'made'or:found.

to be ' inoperable.. for-any. reason, reactor.

. operation is'permissibleionly during.the-

.. succeeding 77 - days ~ forf 2. inoperable pumps l

powered from'different divisionsfor during

' the succeeding 72. hours for 2? inoperable pumps powered from the'same division unless-such system:isisooner made operable,'provided' that during'such time :all:. active components -

of the' operable RHR service water system l

' system shall be; operable.

4.

To be considered operable, a RHR service water pump shall be capable of delivering '

i 3500 gpa against a head. of 500. feet.

I 5.

If. the requirements. of 3.5.C.1-3 cannot -

be met, ' an. orderly shutdown.of' thel reactor:

will be initiated and the reactor water:

l.

temperature.shall be reduced to,less than:

.212*F within.24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />' l

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3.0 LIMITItr CONDITIONS FOR OPERATION-4.0 : SURVEILIANCE REQUIREMENTS Ilieh Pressure Core Cooline Capability llich Pressure Core Cooline Capability D.

liigh Pressure _ Coolant Injection.(llPCI) System D.

Surveillance of IIPCI System shall be performed as follows:

1.

Except as specified in 3.5.D.2 below, 1.

Testing the IIPCI system shall be operable when-ever the reactor; pressure is greater than Item Frecuency 150 psig and irradiated fuel is in the-Pump-operability Fursuant to reactor vessel, except during reactor Specification 4.15.B vessel hydrostatic or leakage tests.

Valve operability Pursuant to Specification 4.15.B'

. Simulated ~ automatic Each refueling actuation test outage (testing valve oper-ability) 2.

From and after the date that the IIPCI system is made or found to be Inoperable for any reason, reactor operation is per-1 missible only during the succeeding 14 days unless such system is_ sooner;made 1

operable, provided that during such 14 days all of the Automatic Pressure Relief systems, the RCIC system, both of the core spray systems, and the LPCI subsystem and containment cooling mode of the RIIR system are operable.

3.5/4.5 108 Amendment No. 63, 77 e

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4.01lSURVEILIANCE REQUIREMENTS 3.0 LIMITING CONDITIONS _.FOR OPERATION E.

' Automatic ' Pressure Relief System E. ~ Surveillance of the AutomaticL Pressure Relief

~

System shall beiperformed as follows:-

1.

Except as specified in 3.5.E.2 and :

1.

Testing:

3.5.E.~3 below..the~ entire automatic

. pressure. relief system lshall_be' Item-

"Frecuency-operable whenever the reactor pressure Valve operability-Each operating cycle is above 150 psi' andairradiated fuel.-

is the reactor. vessel. except during Simulated automa-Each operating cycle reactor vessel hydrostatic or leakage.

tic l actuation test-

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' tests.

ADS Inhibit Switch Each operating cycle -

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2.

From and after the date that one of the l

automatic pressure; relieff system.valvesL NOTE: Safety / relief. valve operability.

is made-or found to be' inoperable for any, is verified by cycling the valve reason, reactor operation'is permissible and observing a compensating I

only during the succeeding 14 days' change in turbine bypass valve unless such valve:is sooner made operable,.

position.

l provided that during such 14 days both remaining? automatic relief system I

valves and the 'llPCI system ' are operable.

3.

'From and after:the date that'more than one of-the automaticLpressure. relief valves are made or:found to be Inoperable for_any-reason,:reactorioperation isz permissible'only during the succeeding l

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> unless repairs are made and-provided that during such time the IIPCI.

system is operable.

4.

If the requirements of 3.5.E.1 -

cannot be met, an orderly reactor 110 3.5/4.5 Amendment No. 62, 63. 77

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'3.0-. LIMITING CONDITIONS FT)R: OPERATION:

.4.0..

SURVEILIANCE REQUIREMENTS

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shutdown-shall.be initilatedJimmediately.

-and the reactor pressure.'shall-be reduced-1

to1150 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter-i t

F.. Reactor Core: Isolation Cooling System-(RCIC)'

. Surveillance of Reactor Core Isolation :

F.

Cooling System (RCIC)'

s

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1.

. Except as. specified -in 3.5.F.2 below, s the L Surveillance of the RCIC System shall be RCIC system shall ne operable whenever'

' performed as follows:

the reactor pressure is greater than'150 psig and irradiated fuel is'in the reactor!

1.

Testing vessel, except during reactor vessel hydrostatic or 1eakage. tests.

To be considered operable, the Item.

Frecuency

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RCIC system shall meet the following conditions:

Pump operability' Pursuant to.

Er Specification 4.15.B' 4

a.

The RCIC shall be capable of;de: ivering

. Pursuant to i

400 gpm into the reactor-vessel at 150 psig.

Valve operability Specification 4.15..B b.

The controls for automatic transfer of i

the,RCIC pump suction from the condensate Simulated automatic Once/ Operating Cycle, i

< storage tanic to the suppression chamber actuation, transfer:

shall be operable.

of' suction'to sup-pression pool, and c.

The controls for automatic restart on' automatic restart I

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-subsequent low reactor levelJafter it.

on subsequent low has been terminated by a high reactor..

reactor water level level. signal'shall be: operable t

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3.5/4.5 111 l

Amendment No. 37, 63, 77 i

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>b 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS 2.

From and after the date that the RCIC system is made or found to be inoperable for any reason, except automatic transfer of pump suction, reactor operation is permissible only during the succeeding 14 days unless such system is sooner made operable. L'ith the controls for automatic transfer of pump'~ suction inoperable, operation for up to 30 days is permissible if the pump suction is aligned to the. suppression pool.

If these conditions cannot be met, an orderly shutdown shall be initiated and the reactor pressure reduced to1150 psig witbin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

I i

112 3.5/4.5 Amendment No. 77,77 l

'3.0.

LIMITING CONDITIONS FOR OPERATION _.

4.0 SURVEILLANCE REQUIREMSNTS G.

Minimum Core and Containment Cooling System Availability l

1.

.When Irradiated fuel is in the reactor vessel'and reactor coolant temperature is less than 212* F, call low pressure ' core and' containment cooling subsystems may be inoperable provided no work is being done which has the potential for draining the i

reactor vessel except as allowed by I

specification 3.5.G.2 below.

l 2.

When irradiated' fuel is in the reactor vessel and the vessel-head'is removed, the suppression chamber may be drained completely and no more than one control rod drive housing or instrument thiable opened at any one time provided that the spent fuel pool gates -are open and -

the fuel pool water level is maintained at a level of greater than or equal to 33 feet.

s 3.5/4.5 113 g

Amendment No. 47, 77

s Bases Continued:

3.5 The allowable repair times are established so that the average risk; rate for repair would be no greater than the basic risk rate.

The cethod and concept are described in reference (1). Using the results developed in this reference, t5s repair period is found to be Iess than 1/2 the test interval.

This assumes that the core spray and LPCI subsystems constitute a 1 out of 3 system; however, 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 pursuant to Specificatien 4.15.B, Inservice Testing, and Specification 4.15.B references ASME Code Section XI which is 3 months. Therefore, an

. allowable repair period which maintains the basic risk should be less than 45 deys and this specification is within this period. Although it is recognized that the information given in reference (1) provides a i

quantitative method to estimate allowable repair times, the lack of operating data to support the

. analytical approach prevents complete acceptance of this method aC this tlwe.

There fore, the times stated in the specific items were established with due regard to judgmenc.

Should one core spray subsystem beccse inoperable, the remaining core spray and tha entire LPCI system are available should the need for core cooling arise.

Baseo on judgments of the reliability of the remaining. systems, i.e., the care spray and LPCI, a 7 day repair per iod was obtained.

If both core spray subsystems become inoperable, only the LPCI is available for low pressure cooling.

Based on the fact that the LPCI is available, a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> repair period was obtain'd.

g Should the loss of one. LPCI pump occur, a..early full complement of core and tantainment cooling equipment is available. Three LPCI pumps in conjunctian with the core spray subsystem will perform the core cooling function.

Because of the availability of the majority of the core cooling equip.

ment, a 7 day repair period is justified.

If the LPCI subsystem is not available, at least 2 LPCI pumps must be available to fulfill the containment cooling function. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> repair period was set on this basis.

(1) APED 5736 - Guidelines for Determining Safe Test Incervals and Repair Times for Engineered Safeguards -

April, 1969, I. M. Jacobs and P. U. Marriott.

3. 5 BAS ES 116 Amendment No. 77 i

1 Bases Continued 3.5:

C.

EllR Service Unter 4

The. containment heat removal portion of the Rl!R system is provided to remove heat energy from the containment in the event of a loss of coolant accident.

For the flow specified, the containment longterm pressure is limited to less than 5 psig and, therefore, is more than ample to provide the required heat removal capability. Reference Section 6.2.3.2.3. FSAR. The repair periods specified were arrived at as in 3. 5.B ai>ove.

The containment cooling subsystem consists of two sets,f 2 service water pumps, I heat exchanger, and 2 R11R pumps.

Either set of equipment is capable of performing the containment cooling fune-tion.

Loss of one RiiR service water pump does not seriously jeopardize the containment cooling capability as two of the remaining three pumps can satisfy the cooling requirements.

Since there is some redundancy left, a 30 day repair period is adequate.

less of I containment cooling subsystem leaves one remaining system to perform the containment cooling function.

Based on the fact that when one containment cooling subsystem becomes inoperable only one system remair.s.

a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> repair period was specified.

k The RIIR service water system provides cooling for the RIIR heat exchangers and.can thus mair:tain the suppression pool water within limits. With the flow specified, the pool temperature limits are maintained as cpecified in Specification 3.7.A.I.

D.

Ilich Pressure Coolant Iniection l

The high pressure coolant injection system is provided to adequatelv cool the core for all pipe breaks smaller than these for which the LPCI or core spray subsystems can protect the core.

The !!PCI meets this requirement without the use of off-site AC power.

For the pipe breaks for which the llPCI is intended to function, the core never uncovers and is continuously cooled and i

thus no clad damage occurs. -Reference Section 6.2.4.3 FSAR.

The IIPCI system is backed up by the automatic pressure relief system and either cf two core spray l

systems or the LPCI system. Therefore, when the flPCI system is out of service, the automai.ic i

j pressure relief and core spray systems and LPCI system are required to be opePable.

For additional l

l 3.5 BASES 117 Amendment No. 75, 77 l

!u

n B ses Continued-3t margin. the RCIC system (a non-safeguard system) has been recuired to be operable durf ng this time, since tie RCIC system is capable of supplying significant water makeup to the reactor (400 gpm).

E.

Automatic Pressure Relief.

The relief valves of the autornatic pres ure relief subsystem are a backup to *he itPCI subsystem.

They enable the core spray system or IECI to provido protectio, againsc ti.e small pipe breat in the event of "PCI fallare by depressurizing the reactor vessel repl31) enough to actuare the core sprays or 12CI. Either of the twe m.' spray systems or ISCI provide sufficient flow of coolant to limit fuel clad temperatures to wet av clad melt and to assure that core geometry remains intact.

Three safety /relicf. valves are includo in the automat!c pressure relief systes. Of these three, l

only two are required to provide sufficient capacity for the automatic pressure relief system. See section 4.4 and 6.2.5.3 FSAR.

F.

RCIC The RCIC system is provided to supply continuo s makeup water to the rea.ctor core when the reactor is isolated from the turbine and when the feedvater system is not available. The peneping caliacity of the RCIC system is sufficient to maintain the water level ebove the core without any other water system in operation.

If the water level in the reactor vas.el decreases to the RCIC initiation level, tb system automatically starts. The system may also be manually initiated at j

any time.

The HPCI system provides an alternate method of supplyirr; makeup water to th, reactor should the normal feedvater tecet-e unav311able. Therefore, tI f specification calls for the IIPCI system to bc operable should the RCIC system be found to be i.eope rable.

3.5 BASES 118 Ar:en@ent No. 77 i

L_.

Bases Continued 3.5:

G.

Emergency Cooling Availability The purpose of Specification G is to assure that sufficient core cooling equipment is available at all times.

It is during refueling outages that major maintenance is performed and during such ti=e that all core and containment cooling subsystems may be out of service. Specification 3.5.G.1 allows all l

core.and containment cooling subsystems to be inoperable provided no work is being done which has the f

potential for draining the reactor vessel. Thus events requiring core cooling are precluded.

Specification 3.5.G.2 recogalzes thae concurrent with control rod drive maintenance duries the l

refueling outage, it may be necesssrv to drain the suppressian chamber for maintenance or for the inspection required by SpecificatiL.

4.7.A.I.

In this situation, a rifficient inventory of water is maintained to assure adequate core cooling in the unlikely eves.. of loss of control rod dribe housing or instrument thimble seal integrity.

I II. Recirculation System Specification 3.5.H_1 is based upon providing essurance that neutron flux limit cycle cscillations, which have a small probability of occurring in the high power / low flow corner of the operating domain, are detected and suppressed. Under certain high power / low flow conditions that could occur during a recirculation pump trip and subsequent Single Loop Operation (SLO' where reverse flow occurs in inactive jet pumps, a hydraulic / reactor kinetic feedback mechanism can be enhanced such that sustained limit cycle oscillations of flow noise with peak to peak levels several times normal values are exhibited. Althcugh large margins to safety limits are maintained when these 11elt cycle oscillations occur, they are to be monitored for, and suppressed when flux noise exceeds the three time baseline valve by insertir.g rods and/or increasing coolant flow.

The line in Figure 3.5.1 is based on the 804 rod line below which the probability of limit cycle oscillations occurring is regligible.

APRM and/or LPRM oscillations in excess of those specified in Specification 3.5.H.I.e ceuld be an indication that a condition of thermal hydraulic instability exists and that appropriate remedfel action should be taken.

By restricting core flou to greater than or equal to 39% of rated, which corresponds to the core flow at the 80% rod line with 2 recirculation pumps running at minimum speed.

the region of the power / flow map where these oscillations are most likely to occur is avoi<ed (Ref. 1).

Above 45% of rated core flew in Single Loop Operation there is the potential to set up high flos-induced noise in the core.

Thus, surveillance of core plate AP noise is required in this.cgion cf the l

power / flow map to alert the operators

• _o take appropriate remedial action 11 such a condition exists.

Specification 3.6.A.2 governs the restart of the pu=p in an idle recirculation loep.

Adherence to this specification limits the probability of excessive flux transients and/or thermal stresses.

I.

Deleted

References:

1. General Electric Service Information Letter No. 380. Rev. 1. Febe nary 10, 1984 II9

3. 5 BASES kend ent No. 27. 37. 77 i

v Bases 4.5-The ' testing interval for the core and containment coaling systems is baset on a quantitatise reliebility analysis, judgment, and practicality. The core coollog systems have not en designed to be fully testable during operation.

For example, the core spray final admission u tves <!o not open until reactor pressure has fallen to 450 psig; thus, during operation even if high drywell pressure were simulated, the final valves would not open.

?n the case of the HPCI, automatic initiation during power operation would result in pumping cold water into the reactor vessel, which is not desirable.

The systems can be automatically actuated during a refeeling outage and this will be done.. To increate the availability of the individual components of th= core and containment cooling systems, the components which make up the system, _i.e.,

iustrumentation, pumps, salve operators, etc., are tested more frequently.

The pumps and motor-operated valves are tested pursuant to Specification 4.15.B. Inservice Testing, to assure their operability. The combiraation of a simulated automatic actuation test each refueling cycle and Section XI testing of the pumps and valve operatoics is dee*3 to be adequate testing of these systeur.

With components or subsystems out-of-service, overall core and centainment cooling reliability is main-l tained by' periodic testing of the remaining cooling equipment. The degree of operability to be demon-I strated depends on the nature of the reason for the out-of-service equipment. For toutine out-of-service I

periods caused by preventative maintenance, etc., the periodic pump end valve operability checks assure the reliability of the remaining components. However, if a failure, design deliciency, etc., caused the out-of-service period, then the demonstration of operability should be thorough enough to assure that a similar problem does not exist on the remaining components.

l 4.5 BASES 120 Acondment fb. 77

+

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3.0 LIMITING CONDITIONS FOR OPERATION 5.0 SURVEILIANGE REQUIREMENTS c.

Except for inerting and deinerting operations permitted in (b) above, all containment purging and venting above cold shutdown shall be via a 2-inch purge and vent valve bypass line and the StanJby Cas Treanuent System.

InertingI and deinerting operations may be via the 18-inch purge and vent valves (equipped with 40-degree limit steps) aligned to the Reactor Building plenum and vent.

l l-6.

If the specifications of 3.7.A cannot be met - the reactor shall be placed in a cold shutdown conditica within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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E.

Standby Cas Treatment System B..

Standby Cas Treatment System 1.

Two separate and independent standby 1.

At least once per month, initiate from gas treatment system circuits shall be the control room 3500 cfm (1104) flow operable at all times when secondary throur)h both circuits of the standby containment. Integrity is required, gas treatment system.

except as specified in sections 3.7.B.I.(a) and (b).

a.

After one of the standby gas treatment system circuits is made

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or found to be inoperable for.,y l'

reason, reactor operation and fuel handling is permissible only during the succeeding seven days, provided that all active components in the other standby gas treatment system l

are operable. Within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> follow-Ing the 7 days, the reacter shall be pieced in a conditica for which the standby gas treatment system is iot required in accordance wit's Specification 3.7.C.2.(a) through (d).

166 3.7/4.7 Amendmont No. FK, 77

1 3.0 LIMITING CONDITIONS m R OPERATION 4.0 SURVEILIANCE REQUIRD1ENTS b.

If bott. standby gas treatment system circuits.are not operable, within 3" hurs the reactor shall be placed in a condition for which the standby gas treatment system is not required in accordance with Specifim tion 3.7.C.2.(a) through (d).

2.

Performance Requirement Tests 2.

Performance Requirements.

a.

At least once per 723 hours0.00837 days <br />0.201 hours <br />0.0012 weeks <br />2.751015e-4 months <br /> of system operatian; or once per cperating cycle, n.

Periodic Requirements but not to exceed 18 months, whichever occurs first; or following painting.

(1) The results of the in-place fire or chemical release in any vent-DOP tests et 3500 cfm (ilot) 11ation zone communicating with the on HEPA filters shall show system while the system is operating

$1% DOP penetration.

that could contaminate the HEPA filters or charcoal absorbe~rs, perform the (2) The results of in-place halo.

following:

genated hydrocarbon tests at 3500 cfm (110%) on charcoal (1) In-place DOP test the HEPA filter banks shall show $1% penetra-banks.

tion.

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(2) In-place test the charcoal adsorber (3) The results of laboratory banks with halogenated hydrocarbon carbon sample analysis shall i

t. race r.

show >901 methyl lodine re-moval efficiency when tested (3) Remove one cartmn test canister at 130*C, 95% R.H.

I from the charco.1 adsorber.

Sub-j ject this sample to a laboratory l

analysis to verify methyl lodine i

removal efficiency.

I i

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3.7/4.7 167 l

Amendment ib. 60 77 i

3.0.1.IMITING CONDITIONS FOR OPERATION 4.0 SURVEII.1ANCE REQUIREMENTS reactor core, operations with a potential for reducing the shutdown L

margin below that.specified in specification 3.3.A. and handling of irradiated fuel or the' fuel cask i

in 'the secondary containment are to

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be immediately suspended if secondary containment integrity is not main-tained.

t D.

Primary Containment Automatic Isolation Valves D.

Primary Containment Automatic Isolation Valves 1.

During reactor power operating conditions, 1.

The primary containment automatic isolation valve l all Primary Containment automatic isolation surveillance shall be performed as follows:

valves and all primary system instrument line flow check valves shall be operable except At least once per operating cycle the a.

as specified in 3.7.D.2.

operable isolation valves that are power operated and automatically initiated shall be tested for simulated automatic initiation and closure times.

b.

At least once per operating cycle the primary system instrument line flow check valves shall be tested for proper ope ration.

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c.

All normally open power-operated isolation l

valves shall be tested pursuant to specification 4.15.B.

Main Steam isolation valves shall be tested (one at a time) with

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the reactor power less than 75% of rated.

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4-3.7/4.7 170 l

Amendment No. 3. 71, 77

7 d

j 3. 0' LIMITING CONDITIONS FOR OPERATION 4.0 SURVEIIh NCE REQUIREMENTS d.

At least once per week the main steam-line power-operated isolation valves shall be exercised by partial closure and subsequent reopening.

2.

In the event any Primary Containraent automatic 2.

Whenever a Primary Containment automatic isolation valve becomes inoperable, reactor isolation valve is inoperable, the position of operation in the run mode may continue at least one fully closed valve in each line provided at least one valve in each line having an inoperable valve shall be recorded having an inoperable valve is closed.

daily.

3.

If Specification 3.7.D.1 and 3.7.D.2 cannot 3.

Deleted be met, initiate normal orderly shutdown and have reactor 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 />.

4 The. seat seals of the drywell and suppression chamber 18-inch purge and vent valves shall be replaced at least once every five years.

a 171 3.7/4.7 Amendment fio. O, 71, 77

3.0 LIMITING CONDITIONS FOR OPEFATION 4.0 SURVEILIANCE REQUIREMENTS service providing both the emergency diesel generators are operable.

2.

Reserve Transformers If offsite power sources are made or found to be inoperable for any reason such '. hat Specification 3.9.A.1 is not met, reactor operation is permissible only during the succeeding 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> unless such offsite sources are sooner made operable, provided that either 1R or 2R Transformer is operable.

3.

Standby Diesel Generators B.

3.

Standby Diesel Generators Each diesel generator shall be manually a.

From and after the date that one of the a.

diesel generators is made or found to stcrted, loaded and operated at I

be inoperable, reactor operat. ion is approximately fated load for at least 60 permissible only during the succeeding minutes once every month to demonstrate 7 days provided that the operable diesel operational readiness.

generator is demonstrated to be operable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

I I

This test is required to be completed l

regardless of when the inoperable diesel generator is restored to operability.

The operability of the other diesel generator need not be demonstrated if the diesel generator 1 wperability was due to preplanned preventative maintenance or' testing.

b.

If both diesel generators become inoper-S.

During the monthly generator test, tne able during power operation, the reactor diesel starting air compressor shall F4 shall be placed in the cold shutdown checked for operation and their ab(11ty condition.

to recharge air receivers.

201 3.9/4.9 Amendment No. 23, 51, 77

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1 Bases 3.9:

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The general objective is to assure an a,desuate supply of power with 'at least one active and one standby f

source of powt r available for operation of equipment required for a safe plant shutdown, to maintain the plant in a safe shutdown condition, ar*1 to operate the required engineered safeguards equipraent fcllowing an accident.

I t

AC for shutdown requirements and operation of engineered safeguards equipament can be provided by either f

of the two standby sources of power (tha diesel generators) or any of the three active sources of power (No. IR, No. 2R, or No. IAR transformers).,

Refer to Section 8 of the USAR.

t To provide for maintenance and repair of equipment and still have redun.lancy of power sources, the

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requirement;of one active and one standby source of power was established. The plant's main generator is not given credit as a source since it is not available during shutdown.

The plant 250 V de power is supplied by two batteries. Most station 250 v loads are supplied by the f

original station 250 V battery. A new 250 V battery has been installed for HPCI loads and may be usv

[

for other station loads in the future. Each battery is maintained fully charged by two associate' r

i chargers which also supply the normal de requirements with the betteries as a standby source during

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emerg w y conditions. The plant 125 V de pcwer is normally supplied by two batteries, each with an i

assoct 2ted charger. Backup chargers are available.

The minimum diesel fuel supply of 32,500 gallons will supply one diesel generator for a minimum of seven

[

days of full load operation. The diesel fuel oil requirement of 32.500 gallons ensures that one emergency j

diesel generator can run for seven days at full load (2500 KU).

The amount of fuel oil necessary to run one 4

j emergency diesel generator for seven days is 31,248 gallons. The difference between these two volumes allows for instrument inaccuracy, tank volume uncertainties, and the location of the suction pipe. Additional diesel i

fuel can normally be obtained within a few hours. Maintaining at least seven days supply is therefore conservative.

In the normal mode of operation, power is available from the off-site sources. One diesel may be allowed f

out of service based on the availability of off-site power provided that the remaining diesel generator is l

3 1

demonstrated to be operable within 24 hcurs.

This test is required even if the inoperable diesel is

[

restored to operability within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Thus, though one diesel generator is temporarily out of f

service, the off-site sources are available, as well as the remaining diesel generator. Based on a monthly testing period (Specification 4.9), the seven day repair period is justified. (1) j (1) "Rellability of Engineered Safety Features as a Function of Testing Frequency *, I.M. Jacobs,

)

Nuclear Safety, Volume 9, 76. 4. July - August 1968.

s 5

J 3.9/4.9 BASES 204 i

j kor& ent tb. at, 51, 75, 77 l

2 b:

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I 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS 2.

It is permissible to have one of the. F.smps e.

Every three months verify that a required by Specification 3.13.B.I.a sample of fuel from the diesel oil inope rable provided that the redundanc pumps storage tank, obtained in accordance are operable. Restore the inoperable pump to with ASTM-D2!0-65, is within the operable status within seven days or provide a acceptable limits specified in l

[

special report to the Commission within 30 Table 1 of ASTM D975-74 when checked l

days outlining the plans and procedures to bn for viscosity, water, and sediment.

used to provide for the loss of redundancy f u the Fire Suppression Water system.

f.

Every 18 months subject the diesel-driven fire pump engine to an inspection 3.

With the fire suppression water system other-in accordance with procedures prepared wise inoperable:

in conjunction with the manufacturer's recommendations for this class of a.

Establish a backup fire Suppress 1 ort Water standby service.

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

g.

A simulated automatic actuation of b.

Provide a special report to the Commission each fire pump and the screen wash / fire within 14 days outlining the actions taken pump, including verification of and the plans and schedule for restoring the ptmp capability, shall be conducted system to operable status.

every 18 months.

h.

The yard main and the reactor bulIding and turbine building headers shall be flushed every 12 months.

i. System flow tests shall be performed every three years.

J.

Valves in flow paths supplying fire suppression water to safe *y related structures, systems, and component shall be cycled every 12 months.

3.13/4.13 225 Amendment No. /, 33, 77 h__...__.__.____

~ = - - - - -.,

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1 3.0-LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIMCE REQUIREMENTS k.

Each valve (manual, power operated, or automatic) in the flow path that is not I

electrically supervised locked, realed i

or.otherwise secured in position, shall l

be verified to be in its correct position every month.

C.

Hose Stations C

Hose Stations 1.

Whenever equipment protected oy hose 1.

The hose stations specified in 3.13.C.1 stations in the following areas is shall be s

' monstrated operable as follows:

required to be operable, the hose station (s) protecting equipment required j

Each month a visual inspection shall I

to be operable in those areas shall be operable:

g

'e conducted to assure all equipment

[

s available.

a.

Diesel generator rooms t

.very 18 months the hose shall be b.

Safety related areas of the turbine I

'oved for inspection and re-racking

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building

c.. Safety related areas of the screen-nd all gaskets in the couplings shall l

e inspected and replaced if necessary.

house d.

Reactor building c.

Every 3 years each hose station valve

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Reactor building addition shall be partially opened to verify

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e.

f.

Safety related areas of the valve operability and no flow blockage.

Administration building j

i d.

Every 3 years each hose shall be I

l hydrostatically tested at a pressure at least 50 psig greater. than the

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maximum pressure available at any j

hose station.

i t

n 3.13/4.13 r

226 l

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toen@ent No. 7, 61, 77 i

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4.13 BASES:

Fire _ detectors -are tested in accordance with the manufacturer's recommendations

'All tests and inspections are performed by the plant staff.- Every six months each detector is functionally tested.. Combustion generated smoke is_not used in these tests.

Alarm circuits are functionally checked every six months.

In addition, all circuitry is automatically supervised +:r open wiring and ground faults.

Fire pumps are tested each ne -h to verify operability. Test starting of the screen wash / fire pump is not required since it is norma 11;

.n service.

Each fire pump is manually started ar.d operated for at least 15 minutes with pump flow directed through the recirculation test line.

Every 18 months the operability of the automatic actuation logic for the fire pumps and the screen wash / fire pump is verified snd the performance of each pump is verified to meet system requirements. The specified flush and valve checks provide assurance that the piping system is capable of supplying fire suppression water to all safety related l

^

areas.

l A system flow test is specified every three years.. This test verifles the hydraulic performance of the fire suppression fire water header system. The testing will be performed using Section II. Chapter 5 of the Fire Protection Handbook, 14th Edition, as a procedural guide. This test is generally performed in conjunction with a visit from insurance company inspectors.

Ilose stations and yard hydrant hose houses are inspected monthly to verify tha all required equipment is in place. Gaskets in hose couplings are inspected periodically and the hose is pressure tested.

Pressure testing of outdoor hose is conwiucted more frequently than indoor hose because of the less favorable storage conditions. Operability of hose station isolation valves is verified every three years by partially opening each valve to verify flow. All of. these tests provide a high degree of assurance that each hose station and yard hydrant hose house will perform satisfactorily after periods of standby service.

Simulated automatic actuation tests are conducted each 18 months to confirm the operability of the sprinkler and Italon systems. These tests consist of verification that all valves dampers (Halon system only).

alarms, and flow paths are functional.

Plant fire barrier walls are provided with seals for pipes and cables where necessary. Where such seals are installed, they must be maintained intact to perform their function. Visual inspection of each installed seal is required every 18 months and after seal repair. A visual inspection following repair of a real is sufficient to assure that seal Integrity will be within acceptable limits.

4.13. Bases-228b Amendoent No. 67, 77

- - - - - d

+

Bases 3.15 and 4.15 The inservice inspection program for the Monticello plant conforms to the requirements of 10 CFR 50, Section 50.55a(g). Where practical, the inspection of components classified into NRC Quality Groups A, B, and C conforms to the requirements of ASME Code Class 1, 2, and 3 components, respectively, contained in Section XI of the ASME Boiler and Pressure Vcssel Code.

A program of inservice testing of Quality Group A, B, and C pumps and valves is also in effect at the Monticello plant, that conforms to the requirements contained in Section XI of the ASME Boiler and Pressure vessel Code or where alternate testing is justified in accordance with Generic Letter 89-04 If a Code required inspection is impractical for the Monticello facility, a request for a deviation from that requirement is submitted to the coeraissien in accordance with 10 CFR 50, l

Section 30.55a(g)(6)(i).

Deviations which are needed from the procedures prescribed in Section XI of the ASME Code and applicable Addenda will be reported to the Commission prior to the beginning of each IO-year inspection period if they are known to be required at that time.

Deviations which are identified during the coursa of inspection will be reported quarterly throughout the inspection period.

i 229 3.15/4.15 BASES A endrent M. F.g 77. 77 m

3.0 LIMITING (MNDITIONS FT)R OPERATION L.0 SIEVEILLANCE REQUIREMENTS 2.

LTelds in austenttic stainless steel piping four inches or larger in diameter containing reactor coolant at a temperature above 200 degrees F during power operation, including reactor vessel attachments and appurtenances, shall be included in an augmented inspection program pacting the requirements of Generic Letter 88-01.

B.

Inserrice Testing 1.

Inservice Testing of Quality Group A, B,

and C pumps and valves shall be performed g

in accordance with the requirements for ASME Code Class 1, 2 and 3 pumps and valves, respectively, contained in Section XI of the ASME Boiler and Pressure Vessel l

Code and applicable Addenda as required by 10 CFR 50 Section 50.55a(g) except where relief has been granted by the Comreission pursuant to 10 CFR 50 Section i

50.55(a)(g)(6)(1), or where alternate I

testing is Justified in accordance with l

Generic Ixtter 89-04 2.

Nothing in the ASME BoIIer and Pressure i

vessel code shall ba construed to supersede l

the requirements of any Technical Specification.

3.15/4.15 229ff