Text

Tech Spec Changes Re Auxiliary Feedwater Flow Rate Instrumentation Surveillance Requirements & Containment Isolation Valves
	ML18139B504
Person / Time
Site:	Surry
Issue date:	08/21/1981
From:	; VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	;
Shared Package
ML18139B503	List: ML18139B502; ML18139B504;
References
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PDR ADOCK 05000~DR p

e ATTACHMENT 1 SUPPLEMENT TO PROPOSED TECHNICAL SPECIFICATION CHANGE

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TS 3.1-1 3.1 REACTOR COOLANT SYSTEM Applicability Applies to the operating status of the Reactor Coolant System.

Objectives To specify those limiting conditions for operation of the Reactor Coolant System which must be met to ensure safe reactor operation.

These conditions relate to:

operational components, heatup and cooldown, leakage, reactor coolant activity, oxygen and chloride concentrations, minimum temperature for criticality, and reactor coolant system overpres-sure mitigation.

A.

Operational Components Specifications

1.

Reactor Coolant Pumps

a.

A reactor shall not be brought critical with less than two pumps, in non-isolated loops, in operation.

a

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b.

If an unscheduled loss of one or more reactor coolant pumps occurs while operating below 10% rated power (P-7) and results in less than two pumps in service, the affected plant shall be shutdown and the reactor made subcritical by inserting all control banks into the core.

The shutdown rods may remain withdrawn.

c.

When the average reactor coolant loop temperature is greater than 3S0°F, the following conditions shall be met:

1.

At least two reactor coolant loops shall be operable.

2.

At least one reactor coolant loop shall be in operation.

d.

When the average reactor coolant loop temperature is less than or equal to 350°F, the following conditions shall be met:

1.

A minimum of two non-isolated loops, consisting of any combination of reactor coolant loops or residual heat removal loops, shall be operable, except as specified in Specification 3.10.A.6.

2.

At least one reactor coolant loop or one residual heat removal loop shall be in operation, except as specified in Specification 3.10.A.6.

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e.

It TS 3.1-3 Reactor power shall not exceed 50% of rated power with only two pumps in operation unles's the overtemperature b.T trip setpoints have been changed in accordance with Section 2.3, after which power shall not exceed 60% with the inactive loop stop valves open and 65% with the inactive loop stop valves closed.

f.

When all three pumps have been idle for> 15 minutes, the first puinp shall not be started unless:

(1) a bubble exists in the pressurizer or (2) the secondary water temperature of each steam generator is less than 50°F above each of the RCS cold leg temperatures.

2.

Steam Generator A minimum of two steam generators in non-isolated loop shall be operable when the average reactor coolant temperature is greater than 350°F.

3.

Pressurizer Safety Valves

a.

One valve shall be operable whenever the head is on the reactor vessel, except during hydrostatic tests.

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TS 3.1-4

b.

Three valves shall be operable when the reactor coolant average temperature is greater than 350°F, the reactor is critical, or the Reactor Coolant System is not connected to the Residual Heat Removal System.

c.

Valve lift setting$ shall be maintained at 2485 psig +/- 1 percent.

4.

Reactor Coolant Loops Loop stop valves shall not be closed in more than one loop unless the Reactor Coolant System is connected to the Residual Heal Removal System and the Residual Heat Removal System is operable.

5.

Pressurizer

a.

The reactor shall be maintained subcritical by at least 1%

until the steam bubble is established and necessary sprays and at least 125 Kw of heaters are operable.

b.

With the pressurizer inoperable due to inoperable pressurizer heaters, restore the inoperable heaters within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least hot shutdown within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and the reactor coolant system temperature and pressure less than 350°F and 450 psig, respectively, within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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6.

Basis e

TS 3.1-5

c.

With the pressurizer otherwise inoperabl~, be in at least hot shutdown with the reactor trip breakers open within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and the reactor coolant system temperature and pressure less than 350°F and 450 psig, respectively, within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Relief Valves

a.

Two power operated relief valves (PORVs) and their associated block valves shall be operable whenever the reactor keff is ~0.99.

b.

With one or more PORVs inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the PORV(s) to operable status or close the associated block valve(s) and remove power from the block valve(s); otherwise, be 0 in at least hot shutdown within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in cold shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c.

With one or more block valve(s) inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the block valve(s) to operable status or close the block valve(s) and remove power from the block valve(s); otherwise, be in at least hot shutdown within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in cold shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Specification 3.1.A-1 requires that a sufficient number of reactor coolant pumps be operating to provide coastdown core cooling flow in the event of a loss of reactor coolant flow accident. This provided flow will maintain the

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TS 3.1-Sa DNBR above 1.30.(l) Heat transfer analyses also show that reactor heat equiva-lent to approximately 10% of rated power can be removed with natural circulation; however, the plant is not designed for critical operation with natural circulation or one loop operation and will not be operated under these conditions.

When the boron concentration of the Reactor Coolant System is to be reduced th_e process must be uniform to prevent sudden reactivity changes in the

reactor. Mixing of the reactor coolant will be sufficient to maintain a uni-form concentration if at least one reactor coolant pump or one residual heat removal pump is running while the change is taking place.

The residual heat removal pump will circulate the equivalent of the reactor coolant system volume in approximately one half hour.

One steam generator capable of performing its heat transfer function will provide sufficient heat removal capability to remove core decay heat after a normal reactor shutdown.

The requirement for redundant coola~t loops ensures the capability to remove core decay heat when the reactor coolant system average temperature is less than or equal to 350°F.

Because of the low-low

steam generator water level reactor trip, normal reactor criticality cannot be achieved without water in the steam generators in reactor coolant loops with open loop stop valves.

The requirement for two operable steam generators, combined with the requirements of Specification 3.6, ensure adequate heat removal capabilities for reactor coolant system temperatures of greater than 3S0°F.

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TS 3.1-5b Each of the pressurizer safety valves is designed to relieve 295,000 lbs.

per hr. of saturated steam at the valve setpoint.

Below 350°F and 450 psig in the Reactor Coolant System, the Residual Heat Removal System can remove decay hea-t and thereby control system temperature and pressure.

There are no credible accidents which could occur when the Reactor Coolant System is connected to the Residual Heat Removal System which could give a surge -rate exceeding the capacity of one pressurizer safety valve.

Also, two safety valves have a capacity greater than the maximum surge rate resulting from complete loss of load. C2)

The limitation specified in item 4 above on reactor coolant loop isolation will prevent an accidental isolation of all the loops which would eliminate the capability of dissipating core decay heat when the Reactor Coolant System is not connected to the Residual Heat Removal System.

The requirement for steam bubble formation in the pressurizer when the reactor has passes 1% subcriticality will ensure that the Reactor Coolant System will not be solid when criticality is achieved.

The requirement that 125 Kw of pressurizer heaters and their associated controls be capable of being supplied electrical power from an emergency bus provides assurance that these heaters can be energized during a loss of offsite power condition to maintain natural circulation at hot shutdown.

e TS 3.1-Sc The power operated relief valves (PORVs) operate to relieve RCS pressure below the setting of the pressurizer code safety valves.

These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable.

The electrical power for both the relief valves and the block valves. is capable of being supplied from an emergency power source to ensure the ability to seal this possible RCS leakage path.

References:

(1)

FSAR Section 14.2.9 (2)

FSAR Section 14.2.10

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e 3.7 INSTRUMENTATION SYSTEMS Operational Safety Instrumentation Applicability:

e Applies to reactor and safety features instrumentation systems.

Objectives:

TS 3.7-1 To provide for automatic initiation of the Engineered Safety Features in the event that principal process variable limits are exceeded, and to delineate the conditions of the plant instrumentation and safety circuits necessary to ensure reactor safety.

Specification:

A.

For on-line testng or in the event of a sub-system instrumentation channel failure, plant operation at rated power shall be permitted to continue in accordance with TS Tables 3.7-1 through 3.7-3.

B.

In the event the number of channels of a particular su,b-system in service falls below the limits given in the column entitled Minimum Operable Channels, or Minimum Degree of Redundancy cannot be achieved, operation shall be limited according to the requirement shown in Column 4 of TS tables 3.7-1 through 3.7-3.

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e TS 3.7-2 C~

In the event of sub-system instrumentation channel failure permitted by Specification 3.7-B, Tables 3.7-1 through 3.7-3 need not be observed during the short period of time and operable sub-system channel are tested where the failed chann.el must be blocked to prevent unnecessary reactor trip.

D.

The Engineered Safety Features initiation instrumentation setting limits shall be as stated in TS Table 3.7-4.

E.

Automatic functions operated from radiation monitor alarm shall be as stated in TS Table 3.7-5.

The requirements of Specification 3.0.1 are not applicable.

F.

The accident monitoring instrumentation for its associated operable components listed in TS Table 3.7-6 shall be operable in accordance with the following:

1.

With the number of operable accident monitoring instrumentation channels less than the total number of channels shown in TS Table 3.7-6, either restore the inoperable channel(s) to operable status within 7 days or be in at least hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2.

With the number of operable accident monitoring instrumentaton channels less than the minimum channels operable requirement of TS Table 3.7-6, either restore the inoperable channel(s) to operable status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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TS 3.7-3 Basis Instrument Operating Conditions During plant operations, the complete instrumentation system will normally be in service. Reactor safety is provided by the Reactor Protection System, which automatically initiates appropriate action to prevent exceeding established limits.

Safety is not compromised, however, by continuing operation with certain instru-mentation channels out of service since provisions were made for this in the plant design.

This specification outlines limiting conditions for operation necessary to preserve the effectiveness of the Reactor Control and Protection System when any one or more of the channels is out of service.

Almost all reactor protection channels are supplied with sufficient redundancy to provide the capability for channel calibration and test at power.

Exceptions are backup channels such as reactor coolant pump breakers.

The removal of one trip channel on process control equipment is accomplished by placing that channel bistable in a tripped mode; e.g., a two-out-of-three circuit becomes a one-out-of-two circuit *. The nuclear instrumentation system channels are not intentionally placed in a tripped mode since the test signal is superimposed on the normal detector signal to test at power..

Testing of the NIS power range channel requires:

(a) bypassing the Dropped Rod protection from NIS, for the channel being tested: and (b) placing the aT/T

protection channel set that is avg being fed from the NIS channel in the trip mode and (c) defeating the power mismatch section of T control channels when the appropriate NIS channel is avg

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TS 3.7-4 being tested. However, the Rod Position System and remaining NIS channels still provide the dropped-rod protection. Testing does not trip the system unless a trip condition exists in a concurrent channel.

Instrumentation has been provided to sense *accident conditions. and to initiate operation of the Engineered Safety Features. (1)

Safety Injection System Actuation

. Protection against a Loss of Coolant or Steam Break Accident is brought about by automatic actuation of the Safety Injection System which provides emergency cooling and reduction of reactivity.

The Loss of Coolant Accident is characterized by depressurization of the Reactor Coolant System and rapid loss of reactor coolant to the containment.

The Engineered Safeguards Instrumentaton has been designed to sense these effects of the Loss of Coolant accident by detecting low pressurizer pressure to generator signals actuating the SIS active phase.

The SIS active phase is also actuated by a high containment pressure signal brought about by loss of h_igh enthalpy coolant to the containment.

This actuation signal acts as a backup to the low pressurizer pressure actuation of the SIS and also adds diversity to protection against loss of coolant.

Signals are also provided to actuate the SIS upon sensing the effects of a steam line break accident.

Therefore, SIS actuation following a steam line.

break is designed to occur upon sensing high differential steam pressure

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TS 3.7-5 between the steam header and steam generator line or upon sensing high steam line flow in coincidence with low reactor coolant average temperature or low steam line pressure.

The increase in the extraction of RCS heat following a steam line break results in reactor coolant temperature and pressure reduction.

For this reason pro-tection against a steam line brea accident is also provided by low pressurizer pressure actuating safety injection.

Protection is also provided for a steam line break in the containment by actuation of SIS upon sensing high containment pressure.

SIS actuation injects highly borated fluid into the Reactor Coolant System in order to counter the reactivity insertion brough about by cooldown of the reactor coolant which occurs during a steam line break accident.

Containment Spray The Engineered Safety Features also initiate containment spray upon sensing a high-high containment pressure signal.

The containment spray acts to reduce containment pressure in the event of a loss of coolant or steam line break accident inside the containment.

The containment spray cools the containment directly and limits the release of fission products by absorbing iodine should it be released to the containment.

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e TS 3.7-6 Containment spray is designed to be actuated at a higher containment pressure (approximately SO% of design containment pressure) than the SIS (10% of design).

Since spurious actuation of containment spray is to be avoided, it is initiated only on coincidence of high-high containment pressure sensed by 3 out of the 4 containment pressure signals provided for its actuation.

Steam Line Isolation Steam line isolation signals are initiated by the Engineered Safety Features closing all steam line trip valves.

In the event of a steam line break, this action prevents continuous, uncontrolled steam release from more than one steam generator by isolating the steam lines on high-high containment pressure or high steam line flow with coincident low steam line pressure or low reactor coolant average temperature. Protection is afforded for breaks inside or outside the containment even when it is assumed that there is a single failure in the steam line isolation system.

Feedwater Line Isolation The feedwater lines are isolated upon actuation of the Safety Injection System in order to prevent excessive cooldown of the reactor coolant system.

This mitigates the effects of an accident such as steam break which in itself causes excessive coolant temperature cooldown.

Feedwater line isolation also reduces the consequences of a steam line break inside the containment, by stopping the entry of feedwater.

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e TS 3.7-7 Auxiliary Feedwater System Actuation The automatic initiation of auxiliary feedwater flow to the steam generators by instruments identified in Table 3.7-2 ensures that the Reactor Coolant System Decay Heat can be removed following loss of main feedwater flow.

This is consistent with the requirements of the "TMI-2 Lesson Learned Task Force Status Report", NUREG-0578, item 2.1.7.b.

Setting Limits

1.

The high containment pressure limit is set at about 10% of design containment pressure. Initiation of Safety Injection protects against loss of coolant (2) or steam line break (3) accidents as discussed in the safety analysis.

2.

The high-high containment pressure limit is set at about 50% of design containment pressure. Initiation of Containment Spray and Steam Line Isolation protects against large loss of coolant (2) or steam line break accidents (3) as discussed in the safety analysis.

3.

The pressurizer low pressure setpoint fo.r safety injection acutation is set substantially below system.operating pressure limits. However, it is sufficiently high to protect against a loss-of-coolant accident as shown in the safety analysis. (2)

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4.

The steam line high differential pressure limit is set well below the differential pressure expected in the event of a large steam line break accident as shown in the safety analysis. (3)

5.

The high steam line flow differential pressure setpoint is constant at 40% full flow between no load and 20% load and increasing linearly to 110% of full flow at full load in order to protect against large steam line break accidents.

The coincident low T setting limit for SIS and avg.

steam line isolation initiation is set below its hot shutdown value.

The coincident steam line pressure setting limit is set below the full load operating pressure.

The safety analysis shows that these settings provide protection in the event of a large steam line break. (3)

Automatic Function Operated from Radiation Monitors The Process Radiation Monitoring System continuously monitors selected lines containing or possibly containing, radioactive effluent. Certain channels in this system actuate control valves on a high-activity alarm signal. Additional information on the Process Radiation Monitoring System is available in the FSAR. (4)

Accident Monitoring Instrumentation The operability of the accident monitoring instrumentation is Table 3.7-6 ensures that sufficient information is available on selected plant parameters to monitor and assess these variables during and following an accident.

On the pressurizer PORV's, the pertinent channels consist of limit switch indication and acoustic

TS 3.7-9 monitor indication.

The pressurizer safety valves utilize an acoustic monitor channel and a downstream high temperature indication channel.

This capability is consistent with the recommendations of Regulatory Guide 1.97, "Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident", December 1975, and NUREG-0578, "TMI.:.2 Lessons Learned Task Force Status Report and Short Term Recommendations".

References (1)

FSAR Section 7.5 (2)

FSAR - Section 14.5 (3)

FSAR Section 14.3.2 (4)

FSAR Section 11.3.3

1.

2.

3.

4.

s.

FUNCTIONAL UNIT Manual Nuclear Flux Power Range Nuclear Flux Intermediate Range Nuclear Flux Source Range Overtemperature aT

6.

Overpower aT

7.

Low Pressurizer Pressure

8.

Hi Pressurizer Pressure 1

MIN.

OPERABLE CHANNELS 1

3 1

1 2

2 2

2 TABLE 3. 7-1 REACTOR TRIP INSTRUMENT OPERATING CONDITIONS 2

DEGREE OF REDUN-DANCY 2

1 1

3 PERMISSIBLE BYPASS CONDITIONS Low trip setting when 2 of 4 power channels greater than 10% of full power 2 of 4 power channels greater than 10% full power 1 of 2 intermediate rang~10 channels greater than 10 amps 3 of 4 nuclear power channels and 2 of 2 turbine load channels less than 10% of rated *power Same as Item 7 above

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4 OPERATOR ACTION IF CONDITIONS OF COLUMN 1 OR 2 EXCEPT AS CONDI-TIONED BY COLUMN 3 CANNOT BE MET Maintain hot shutdown

Maintain hot shutdown Maintain hot shutdown Maintain.hot shutdown Maintain hot shutdown Maintain hot shutdown Maintain hot shutdown Maintain hot shutdown

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MIN.

OPERABLE CHANNELS

9.

Pressurizer-Hi Water Level

2

10.

Low Flow 2/ operable loop

11.

Turbine Trip 2

12.

Lo-Lo Steam Generator 2/non-iso-TABLE 3.7-1 REACTOR TRIP INSTRUMENT OPERATING CONDITIONS 2

DEGREE OF REDUN-DANCY 1

1 I/non-3 PERMISSIBLE BYPASS CONDITIONS 3 of 4 nuclear power channels and 2 of 2 turbine load channels less than 10%*of rated power If inoperable loop channels are not in service they must be placed in the tripped mode Water Level lated loop isolated loop

13.

Underfrequency 4KV Bus 2

1

14.

Undervoltage 4KV Bus 2

1 4

OPERATOR ACTION IF CONDITIONS OF COLUMN 1 OR 2 EXCEPT AS CONDI-TIONED BY COLUMN 3 CANNOT BE MET Maintain hot shutdown Maintain hot shutdown Maintain less than 10% rated power Maintain hot shutdown e:

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shutdown Maintain hot shutdown

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FUNCTIONAL UNIT

15.

Control rod misalignment Monitor**

a) rod position deviation b) quadrant power tilt monitor (upper and lower excore neutron detectors)

16.

Safety Injection 1

MIN.

OPERABLE CHANNELS 1

1 TABLE 3.7-1 REACTOR TRIP INSTRUMENT OPERATING CONDITIONS 2

DEGREE OF REDUN-DANCY 3

PERMISSIBLE BYPASS CONDITIONS See Item 1 of TS Table 3~7-2 4

OPERATOR ACTION IF CONDITIONS OF COLUMN 1 OR 2 EXCEPT AS CONDI-TIONED BY COLUMN 3 CANNOT BE MET -

Log individual rod positions once/hour, and after a load change

> 10% or after> 30 inches of control rod motion.

Log individual upper upper and lower ion chamber currents once/

hour and after a load change> 10% or after

> 30 inches of control rod motion.

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TABLE 3.7-1 REACTOR TRIP INSTRUMENT OPERATING CONDITIONS 1

2 FUNCTIONAL UNIT MIN.

OPERABLE CHANNELS DEGREE OF REDUN-DANCY

17.

Low steam generator water level with steam/feedwater mismatch flow I/non-iso-lated loop I/non-iso-lated loop

- If both rod misalignment monitors (a and b) inoperable for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or more, the nuclear overpower trip shall be reset to 93 percent of rated power in addition to the increased surveillance noted.

3 PERMISSIBLE BYPASS CONDITIONS

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4 OPERATOR ACTION IF CONDITIONS OF

coLUMN 1 OR 2 EXCEPT AS CONDI-TIONED BY COLUMN 3 CANNOT BE MET Maintain hot shutdown e

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FUNCTIONAL UNIT

1.

SAFETY INJECTION

a.

Manual

b.

High Containment Press.

c.

High Differential Press.

between any Steam Line and the Steam Line Header

d.

Pressurizer Low-Low Press.

e.

High Steam Flow in 2/3 Steam Lines with Low T or Low Steam Line Pres~~g

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TABLE 3.7-2 ENGINEERED SAFEGUARDS ACTION 1

2 DEGREE MIN.

OF OPERABLE REDUN-CHANNELS DANCY 1

0 3

1 2/non-iso-1/non-lated loop isolated loop 2

1 1/steamline 2 T signals 1 2 sti§m Press.

1 Signals 3

PERMISSIBLE BYPASS CONDITIONS Primary Pressure less than 2000 psig except when reactor is critical Primary Pressure less than 2000 psig except when reactor is critical

.Reactor Coolant aver-age temperature less than 543°F (nominal) during heatup and

.cooldown 4

OPERATOR ACTION IF CONDITIONS OF COLUMN 1 OR 2

.EXCEPT AS CONDI-TIONED BY COLUMN 3 CANNOT BE MET Cold shutdown Cold shutdown Cold shutdown Cold shutdown Cold shutdown

- - With the specified minimum operable channels the 2/3 high steam flow is already in the trip mode.

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b.

FUNCTIONAL UNIT

2.

CONTAINMENT SPRAY Manual High Containment Press.

(Hi-Hi Setpoint)

3.

AUXILIARY FEEDWATER

a.

Steam Generator Water Level Low-Low

i. Start Motor Driven*Pumps II. Start Turbine Driven Pumps

b.

RCP Undervoltage Start Turbine Driven Pump 1

MIN.

OPERABLE CHANNELS 2

3 2/Stm.

2/Stm.

2 TABLE 3.7-2 ENGINEERED SAFEGUARDS ACTION INSTRUMENT OPERATING CONDITIONS Gen.

Gen.

2 DEGREE OF REDUN-DANCY 1

1 1

1 3

PERMISSIBLE BYPASS CONDITIONS Loop Stop Valve in res-pective loop closed

c.

Safety Injection (All safety injection initiating functions and requirements)

Start Motor Driven Pumps

.**Must actuate 2 switches simultaneously.

4 OPERATOR ACTION IF CONDITIONS OF COLUMN 1 OR 2 EXCEPT AS CONDI-TIONED BY COLUMN 3 CANNOT BE MET Cold shutdown Cold shutdown Place inoperable channel in Tripped condition within one hour Place inoperable channel in Tripped condition within one hour e

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FUNCTIONAL UNIT Station Blackout Start Motor* Driven Pump 1

MIN.

OPERABLE CHANNELS 2

Trip of Main Feedwater Pumps 1/Pump Start Motor Pumps TABLE 3.7-2 ENGINEERED SAFEGUARDS ACTION INSTRUMENT OPERATING CONDITONS 2

DEGREE OF REDUN-DANCY 0

I/Pump 3

PERMISSIBLE BYPASS CONDITIONS

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4 OPERATOR ACTION IF CONDITIONS OF COLUMN 1 OR 2 EXCEPT AS CONDI-TIONED BY COLUMN 3 CANNOT BE MET Restore inoperable channel within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in hot shutdown within next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in cold shutdown within the follow-ing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Restore inoperable channel within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in hot shutdown within next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in cold shutdown within the follow-ing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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a.

b.

c.

d.

FUNCTIONAL UNIT CONTAINMENT ISOLATION Safety Injection Manual High Containment Press.

(Hi Setpoint)

High Containment Press.

2.

STEAM LINE ISOLATION

a.

b.

c.

3.

a.

High Steam Flow in 2/3 lines and 2/3 Low T or 2/3 avg Low Steam Pressure High Contaimnent Pres.s.

(Hi-Hi Level)

Manual FEEDWATER LINE ISOLATION Safety Injection

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TABLE 3.7-3 INSTRUMENT OPERATING CONDITIONS FOR ISOLATION FUNCTIONS INSTRUMENT OPERATING CONDITIONS 1

2 3

MIN.

OPERABLE CHANNELS DEGREE OF REDUN-DANCY PERMISSIBLE BYPASS CONDITIONS See Item No. 1 of Table 3.7-2 1

3 3

1/steamline 2/T av2 signars 2 Stm. Press.

signals 3

1/line 1

1 1

See Item No. 1 of Table 3.7-2 4

OPERATOR ACTION IF CONDITIONS OF COLUMN 1 OR 2 EXCEPT AS CONDI-TIONED BY COLUMN 3 CANNOT BE MET Cold shutdown Hot shutdown Cold shutdown Cold shutdown Cold shutdown Cold shutdown Hot shutdown Cold shutdown

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./

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_., ~*,.... :'-.....,. *-

TABLE 3.7-4 ENGINEERED SAFETY FEATURE SYSTEM INITIATION LIMITS INSTRUMENT SETTING No.

FUNCTIONAL UNIT 1

High Containment P_ressure (High Contain-.

ment Pressure Signal) 2 3

4 5

High High Containment Pressure (High High Containment Pressure Signals)

Pressurizer Low Low Pressure High Differential Pressure Between Steam Line and the Steam Line Header High Steam Flow in 2/3 Steam Lines f..

Coincident with Low T or Low Steam

. a\\7g

. Line Pressure CHANNEL ACTION SETTING LIMIT a) Safety Injection

~5 psig b) Containment Vacuum Pump Trip c) High Press. Containment Iso.

d) Safety Injec.tion Contain. Iso.

e) F.W. Line Isolation a) Containment Spray

~25 psig b) Recirculation Spray c) Steam Line Isolation d) High High Press. Contain. !so.

a) Safety Injection

~l, 700 psig b) Safety Injection Cont. !so.*

c) Feedwater Line Isolation a) Safety Injection b) Safety Injection Contain. !so.,

c) F.W. Line Isolation a) Safety Injection b) Steam Line Isolation c) Safety Injection Contain. !so.

d) F.W. Line Isolation

~150 psi

~40% (at zero load) of full steam flow

~40% (at 20% load) of full steam flow

~110% (at full load) of full steam flow

~541°F T avg

~500 psig steam line pressure -

-I

(/)

w.

I _,

CX)

.*':l

No.

6

a.

b.

c.

d.

e.

TABLE 3.7-4 ENGINEERED SAFETY FEATURE SYSTEM INITIATION LIMITS INSTRUMENT SETTING FUNCTIONAL UNIT AUXILIARY FEEDWATER Steam Generator Water Level RCP Undervoltage Safety Injection Station Blackout Main Feedwater Pump Trip Low-Low CHANNEL ACTION Aux. Feedwater Initiation S/G Blowdown Isolation Aux. Feedwater Initiation Aux. Feedwater Initiation Aux. Feedwater Initiation Aux. Feedwater Initiation SETTING LIMIT

~5% narrow range

~70% nominal All S.I. setpoints

~46.7% nominal e

N.A.

-I

(/)

w j"

I __,

U)

MONITOR CHANNEL

1.

Pr.6cess vent particulate and gas monitor~.

. (RM,-GW-101 & RM-GW-102)

2.

Component cooling water radiation monitors (RM-CC-105 & RM-CC-106) 3.. Liquid waste disposal radiation monitors

-(RM-LW-108)

4.

Condenser air ejector radiation monitors (RM-SV-111 & RM-SV-211)

5.

Containment particulte and gas monitors (RM-RMS-159 & RM-RMS-160, RM-RMS-259 & RM-RMS-260)

6.

Manipulator crane area monitors (RM-RMS-162 &

RM-RMS-262)

TABLE 3.7-5 AUTOMATIC FUNCTIONS OPERATED FROM RADIATION MONITORS ALARM AUTOMATIC FUNCTION AT ALARM CONDITlONS Stops discharge from contain.

vacuum systems and waste gas decay tanks (shuts Valve Nos. RCV-GW-160, FCV-GW-260, FCV-GW-101)

Shuts surge tank vent valve HCV-CC-100 Shuts effluent discharge valves FCV-LW-104A and FCV-LW-104B Diverts flow to the contain-ment of the affected unit (Opens TV-SV-102 and shuts TV-SV-103 or opens TV-SV-202 and shuts. TV-SV-203)

Trips affected unit's purge supply and exhaust fans, closes affected unit's purge air butterfly valves (MOV-VS-lOOA, B, C & Dor MOV-VS-200A, B, C & D)

Trips affected unit's purge supply and exhaust fans, closes affected unit's purge air butterfly valves (MOV-VS-lOOA, B, C & Dor MOV-VS-200A, B, C & D MONITORING REQUIREMENTS See Specifications 3.11 and 4.9 See Specifications 3.13 and 4.9 See Specifications 3.11 and 4.9 See Specification 3.11 and 4.9 See Specifications 3.10 and 4.0 See Specifications 3.10 and 4.9 ALARM SETPOINT

µCI/cc

-8 Particula!~ ~4x10 Gas ~9x10

~Twice Backgroun~

-3

~1.SxlO

~1.3 Particula!s ~9xl0-9 Gas ~lxlO -

~50 mr:em/hr

-i I

\\

d

<.n

~ -

w.

-....J I

N 0

!.~

TABLE 3.7-6 ACCIDENT MONITORING INSTRUMENTATION I*

INSTRUMENT I

1.

Auxiliary Feedwater Flow Rate

2.

Reactor Coolant System Subcooling Margin Monitor

3.

PORV Position Indicator (Primary Detector)

4.

s.

6.

7.

PORV Position Indicator (Backup Detector)

PORV Block Valve Position Indicator Safety Valve Position Indicator (Primary Detector)

Safety Valve Position Indicator (Backup Detector)

TOTAL NO.

OF CHANNELS 1 per S/G 2

I/valve I/valve I/valve I/valve I/valve MINIMUM CHANNELS OPERABLE 1 per S/G 1

I/valve 0

I/valve I/valve 0

e

--i

(/)

w.

I N

--1 t* '

_, --*- -~.-- -~*.*. ",. -*--**--*'... -~- *,.,;._."-:":-...

e TS 3.8-1 3.8 CONTAINMENT Applicability Applies to the integrity and operating pressure of the reactor containment.

Objective To define the limiting.operating status of the reactor containment for unit operation.

Spedf ication A.

Containment Integrity and Operating Pressure

1.

The containment integrity, as defined in TS Section 1.0, shall not be violated, except as specified in Specification 3.8.A.2, below, unless the reactor is in the cold shutdown condition.

2.

The reactor containment shall not be purged while the reactor is operating, except as stated in Specification 3.8.A.3.

3.

During the plant startup, the remote manual valve on the steam jet air ejector suction line may be open, if under administrative control, while containment vacuum is being established.

The Reactor Coolant System temperature and pressure must not exceed 350°F and 450 psig, respectively,.~til the air partial pressure in the containment has been reduced to a value equal to, or below, that specified in TS Fig. 3.8-1.

4.

The containment integrity shall not be violated when the reactor vessel head is unbolted unless a shutdown margin greater than 10 percent l::.k/k is maintained.

B.

Basis e

TS 3.8-2

5.

P:bsitive reactivity changes shall not be made by rod drive motion or boron dilution unless the containment integrity is intact.

6.

The containment isolation valves shall be listed in Tables 3.8-1 and 3.8-2.

Internal Pressure

1.

If the internal air partial pressure rises to a point 0.2:S psi above the allowable value of the air partial pressure (TS Fig. 3.8-1),

the reactor shall be brought to the.hot shutdown condition.

2.

If the leakage condition cannot be corrected without violating the containment integrity or if the internal partial pressure continues to rise, the reactor shall be brought to the cold shutdown condition utilizing normal operating procedures.

3.

If the internal pressure falls below 8.25 psia the reactor shall be placed in the cold shutdown condition.

4.

If the air partial pressure cannot be maintained greater than or equal to 9.0 psia, the reactor shall be brought to the hot shutdown condition.

The Reactor Coolant System temperature and pressure being below 350°F and 450 psig, respectively, ensures that no significant amount of flashing steam will be formed and hence that there would be no significant pressure build-up in the containment if there is a loss-of-coolant accident.

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

e TS 3.8-3 The shutdown margins are selected based on the type of activities that are being carried out.

The 10 percent l:lk/k shutdown margin during refueling precludes criticality under any circumstance, even though fuel and control rod assemblies are being moved.

The allowable value for the containment air partial pressure is presented in TS Fig. 3.8-1 for service water temperatures*from 25 to. 90°F.

The allowable value varies as shown in TS Fig. 3.8-1 for a given containment average temperature.

The RWST _water shall have a maximum temperature of 45°F.

The horizontal limit lines in TS Fig. 3.8~1 are based on LOCA peak calcu~

lated pressure criteria, and the sloped line is based on LOCA subatmospheric peak pressure criteria

The curve shall be interpreted as follows:

The horizontal limit line designates the allowable air partial pressure value for the given average containment temperature.

The horizontal limit line applies for service water temperatures from 25°F to the sloped line intersection value (maximum service water temperature).

From TS Fig. 3.8-1, if the containment average temperature is 112°F and the service water temperature is less than or equal to 83°F, the allow-able air partial pressure value shall be less thari. or equal to 9.65 psia.

If the average containment temperature is 116°F and the service water temperature is less than or equal to 88°F, the allowable air partial pressure value shall be less than or equal to 9.35 psia. These horizontal limit lines are a result of the higher allowable initial containment average temperatures and the analysis of the pump suction break.

.. ' :*.. : ~:. :*,......

e TS 3.8-4 If the containment air partial pressure rises to a point 0.25 psi above the allowable value, the reactor shall be brought to the hot shutdown condition.

If a LOCA occurs at the time the containment air partial pressure is 0.25 psi above the allowable value, the maximum containment pressure will be less than 45 psig, the containment will depressurize in less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and the maximum subatmospheric peak pressure will be less than 0.0 psig.

If the containment air partial pressure cannot be maintained greater than or equal to 9.0 psia, the reactor shall be brought to the hot shutdown condition.

The shell and dome plate liner of the containment are -capable of withstanding an internal pressure as low as 3 psia, and the bottom mat liner is capable of withstanding an internal pressure as low as 8 psia.

References FSAR Section 4.3.2 FSAR Section 5.2 FSAR Section 5.2.1 FSAR Section 5.5.2 Reactor Coolant Pump Containment Isolation Design Bases Isolation Design

A.

e TABLE 3.8-1 TS 3.8-5 UNIT NO. 1 CONTAINMENT ISOLATION VALVES VALVE NUMBER FUNCTION PHASE I CONTAINMENT ISOLATION (SAFETY INJECTION SIGNAL)

L MOV-1867C Boron Injection Tank Outlet

2.

MOV-1867D Boron Injection Tank Outlet

3.

MOV-1289A Charging Line

4.

MOV-1381 Reactor Coolant Pump Seal Water Return

5.

HCV-1200A Letdown Orifice Isolation

6.

HCV-1200B Letdown Orifice Isolation

7.

HCV-1200C Letdown Orifice Isolation

8.

TV-SI-101A Accumulator N2 Relief Line

9.

TV-SI-IOIB Accumulator N2 Relief Line

10.

TV-SI-100 Accumulator N2 Relief Line

11.

TV-VG-109A Primary Drain Transfer Tank Vent

12.

TV-VG-I09B Primary Drain Transfer Tank Vent

13.

TV-DG-108A Primary Drain Transfer Pump Discharge

14.

TV-DG-108B Primary Drain Transfer Pump Discharge

15.

TV-CC-109A*

Component Cooling from RHR's

16.

TV-CC-109B*

Component Cooling from RHR's

17.

TV-SS-IOOA Pressurizer Liquid Sample

18.

TV-SS-lOOB Pressurizer Liquid Sample

19.

TV-SS-lOlA Pressurizer Vapor Sample

20.

TV-SS-10IB Pressurizer Vapor Sample

. i

~***-***-- *-*---. *- -..... _' *~* -

e TABLE 3.8-1.

TS 3.8-6 UNIT NO. 1 CONTAINMENT ISOLATION VALVES (Continued)

VALVE NUMBER

21.

TV-SS-103

22.

TV-SS-106A

23.

TV-SS-106B

24.

TV-SS-102A

25.

TV-SS-102B

26.

TV-SS-104A

27.

TV-SS-i04B

28.

TV-CH-1204

29.

TV-PG-1519A

30.

TV-BD-lOOA*

31.

TV-BD-lOOB*

32.

TV-BD-lOOC*

33.

TV-BD-lOOD*

34.

TV-BD-lOOE*

35.

TV-BD-lOOF*

36.

TV-DA-lOOA

. 37.

TV-DA-lOOB

38.

TV-MS-109*

39.

TV-MS-110*

40.

TV-LM-lOOA

41.

TV-LM-lOOB

42.

TV-LM-lOOC FUNCTION Residual Heat Removal System Sample Reactor Coolant Hot Leg Sample Reactor Coolant Hot Leg Sample Reactor Coolant Cold Leg Sample Reactor Coolant Cold Leg Sample Pressurizer Relief.Tank Vapor Sample Pressurizer Relief Tank Vapor Sample-Letdown Isolation Valve Primary Grade Water to Pressurizer Relief Tank Steam Generator Blowdown Valve Steam Generator Blowdown Valve Steam Generator Blowdown Valve Steam Generator Blowdown Valve Steam Generator Blowdown Valve Steam Generator Blowdown Valve Containment Sump Pump Isolation Containment Sump Pump Isolation Main Steam Drain Trip Valve Main Steam Drain Trip Valve Containment Isolatio~ Monitoring Containment IsoJation Monitoring Containment Isolation Monitoring

- --- ¥**- **-*.

e TABLE 3.8-1 TS 3.8-7 UNIT NO. 1 CONTAINMENT ISOLATION VALVES (Continued)

VALVE NmJBER

43.

TV-,LM-lOOD

44.

TV-LM-lOOE

45.

TV-,LM-lOOF

46.

TV-LM-lOOG

47.

TV-LM-lOOH

48.

TV-:CV-150A

49.

TV-CV-150B

50.

-TV-LM-101A

51.

TV-LM..-lOlB

52.

TV-CV-150C

53.

TV-CV-150D

54.

TV-SV-102A B.

PHASE II CONTAINMENT ISOLATION (HI CLS SIGNAL)

1.

TV-RM-lOOA

2.

TV-RM-lOOB

3.

TV-RM-lOOC

. 4.

TV-IA-lOlA

5.

TV-IA-101B FUNCTION Containment Isolation Monitoring Containment Isolation Monitoring Containment Isolation Monitoring Containment Isolation Monitoring Containment Isolation Monitoring Containment Vacuum Suction Valve Containment Vacuum Suction Valve Leakage Monitoring Sealed Reference Leakage Monitoring Sealed Reference Containment Vacuum Suction Valve Containment Vacuum Suction Valve Condenser Air Ejector Vent Trip Valve Containment Air & Particulate Rad. Mon. TV's Containment Air & Particulate Rad. Mon. TV's Containment Air & Particulate Rad. Mon. TV's Containment Instr. Air Compressor Suction Containment Instr. Air Compressor Suction

C.

D.

I I I I

e TABLE 3.8-1 TS 3.8-8 UNIT NO. 1 CONTAINMENT ISOLATION VALVES (Continued)

VALVE NUMBER FUNCTION PHASE III CONTAINMENT ISOLATION (HI-HI.CLS SIGNAL)

1.

TV-MS-101A*

Main Steam Trip Valve

2.

TV-MS-lOlB*

Main Steam Trip Valve

3.

TV-IA-100 Containment Instr. Air Compressor Disch. Vlv.

4.

TV-MS-lOlC*

Main Steam Trip Valve

s.

TV-CC-107*

cc from RCP Thermal Barriers

6.

TV-CC-lOlA*

cc from A Air Recirc.

7.

TV-CC-101B*

cc from B Air Recirc.

8.

TV:..cc-101C**

cc from C Air Recircr

9.

TV-CC-lOSA*

cc from "A" RCP

10.

TV-CC-lOSB*

cc from "B" RCP

11.

TV-CC-lOSC*

cc from "C" RCP CONTAINMENT PURGE & EXHAUST

1.

MOV-VS-lOOC R.C. Purge Exhaust MOV's 2 *.

MOV-VS-lOOD R.C. Purge Exhaust MOV's

3.

MOV-VS-101 R.C. Purge Exhaust Bypass MOV

4.

MOV-VS-lOOA R.C. Purge Supply MOV's

s.

MOV-VS-lOOB R.C. Purge Supply MOV's

6.

MOV-VS-102 Contain. Vacuum Breaker Atmos. Supply MOV

...... ~-.. -*-*-** -.*...... -.........

.1 i

_,.*.~**.* *"-***-~' '***--

.**.,.-~* ~*

-~-- --* **-

r e

TABLE 3.8-1

-**** -* :*:.. :*:, -:... :.:..... *.. *._ --~ --: *-*--.: :_ *.......... -

TS 3.8-9 UNIT NO. 1 CONTAINMENT ISOLATION VALVES (Continued)

VALVE NUMBER E.

REMOTE MANUAL VALVES

1.

MOV-CS-101A

2.

MOV-CS-101B

3.

MOV"'.'CS-101C

4.

MOV-CS-lOlD

s.

MOV-RS-lSSA

6.

MOV-RS-lSSB

7.

MOV-RS-156A

8.

MOV-RS-156B

9.

MOV-1842

10.

MOV-RH-100

11.

FCV-1160

12.

MOV-1890A

13.

MOV-1890B I

14.

MOV-1890C

15.

MOV-1869A

16.

MOV-1869B

17.

MOV-1860A

18.

MOV-1860B

19.

MOV-SW-I04A*

20.

MOV-SW-104B*

21.

MOV-SW-104C*

FUNCTION Containment Spray Discharge Valve Containment Spray Discharge Valve Containment Spray Discharge Valve Containment Spray Discharge Valve Outside Recirc. Spray Suction Valve Outside Recirc. Spray Suction Valve Outside Recirc. Discharge Valve Outside Recirc. Discharge Valve Bypasses Boron Injec. Tank to Cold Leg Injec.

Resi. Heat Remov. to RWST Loop Fill Header Flow Valve Lo Header S. I. Pump Disch. from Hot Leg Lo Header S. I. Pump Disch. from Hot Leg Lo Header S. I. Pump Disch. from Cold Leg

!so. from Hot Leg to Hi Header S. I. Line A

!so. from Hot Leg to Hi Header S. I. Line B Iso. from Sump to Lo Header S. I.

!so. Valve from Sump to Lo Header S. I.

SW to "A" HX' s SW to "B" HX' s SW to "C" HX' s

e TABLE 3.8-1 e

TS 3.8-10 UNIT NO. 1 CONTAINMENT ISOLATION VALVES (Continued)

VALVE NUMBER

22.

MOV-SW-104D*

23.

MOV-SW-105A*

24.

MOV-SW-105B*

25.

MOV~SW-105C*

26.

MOV-SW-105D*

27.

HCV-CV-100 F.

MANUAL VALVES

1.

2.

3.

4.

1-SI-150 1-SI-32 1-GW-182 1-GW-183

5.

1-SA-60

6.

1-SA-62

7.

1-IA-446

8.

1-VA-1 9.*

1-VA-6

10.

2-IA-446

11.

1-GW-175

12.

1-GW-166

13.

1-GW-.174

14.

1-FP-151

15.

1-FP-152

... ~.. *-... *...

FUNCTION SW to "D" HX's SW from "A" HX's SW from "B" HX's SW from "C" HX's SW from "D" HX' s Cont. Vacuum Isolation Boron Injection Tanlc 1" line Accumulator Fill Valve Discharge from Hydrogen Analyzer Discharge from Hydrogen Analyzer Service Air to Containment Service Air to Containment Instrument Air to Containment Outside Isolation from Primary Vent Pot Inside Isolation from Primary Vent Pot Cross Tie from #2 Instrument Air Header Suction from Containment to H2 Analyzer Suction from Containment to H2 Analyzer Inlet to Cont. from H2 Analyzer Outside Cont.

Outside Iso. Vlv for Cont. Fire Protection Outside Iso. Vlv for Cont. Fire Protection

G.

VALVE NUMBER

16.

l-RL-3

17.

1-RL-S

18.

l-RL-13

19.

1-RL-15

20.

1-SI-73

21.

1-SI-174

22.

1-SW-208

23.

l-SW-106

24.

1-cv-2 e

TABLE 3.8-1 e

TS 3.8-11 UNIT NO. 1 CONTAINMENT ISOLATION VALVES (Continued)

FUNCTION Inlet Vlv to Cavity from RCS Outside Cont.

Inlet Vlv to *cavity from RCS Inside Cont.

Suction Vlv to 1-RL-P-IA Inside Containment Suction Vlv to 1-RL-P-IA Outside Containment Accumulator N2 Fill Vlv Outside Containment Bypasses MOV-1869A RS HX SW Drain RS HX SW Drain Cont. Vacuum Isolation CONTAINMENT CHECK VALVES

1.

1-FP-153

2.

1-VP-12

3.

1-RS-17

4.

1-RS-11

5.

l-CS-13

6.

l-CS-24

7.

1-IA-938

8.

2-IA-446

9.

1-SI-234

10.

1-IA-939

11.

1-IA-446 Inside Cont. - Fire Protection Header Inside Cont. - Air Eject Disch to Cont.

Inside Cont. - RS Disch to Cont. A Inside Cont. - RS Disch to Cont. B Inside Cont. - Discharge of 1-CS-P-IA Inside Cont. - Discharge of 1-CS-P-IB Inside Cont. - Disch of Cont. IA Component Manuai Valve - Disch. of IA Component Unit #2.

Check Inside Cont. - N2 to Accumulator Check Inside Cont.~ Disch. of Cont. IA Component Unit #1 Manual Vlv - Disch. of Unit 1 Instr. Air Comp.

e TABLE 3.8-1 e

TS 3.8-12 UNIT NO. 1 CONTAINMENT ISOLATION VALVES (Continued)

VALVE NUMBER FUNCTION

12.

1-RC-160 Check Valve Inside Contain. from PG Supply

13.

1-RM-3 Check Valve Inside Contain. - Rad. Monitoring Sue.

14.

1-IA-939 Instr. Air Check Valve to Containment

15.

1-SA-446 Service Air Check Valve to Containrrient

16.

1-CC-177*

cc to 11 A II rum HX

17.

1-CC-176*

cc to 11B II rum HX

18.

1-SI-225 IIlISI from BIT

19.

1-CC-242*

CC to "A" Air Recirc.

20.

1-CC-233*

cc to 11B11 Air Recirc.

21.

1-CC-224*

cc to 11C" Air Recirc.

22.

1-CH-309 Normal Chg. Hdr

23.

1-CC-1*

CC to "A" RCP

24.

1-CC-58*

CC to "B" RCP

25.

1-CC-59*

CC to "C" RCP

26.

1-SI-224 HHS! BIT Bypass

27.

1-SI-226 HHS! to Hot Legs

28.

1-SI-228 LHSI Pp Discharge*

29.

1-SI-229 LHSI Pp Discharge

30.

1-SI-227 LHSI to Hot Leg

- Not subject to Type "C" Testing.

- - Modifications to this table should be submitted to the NRC as part of the next license amendment.

- - -*--**...,.:.._.-;.,. *:.:.. *.. :._. ;..,-,*.~*.*,*...* ;:* _,.;:.;._,...,.,.... ~: *..,..:.. _.:.:,:b,.:<<.:'..

_,.:...',. : *.. b,_..........

e TABLE 3.8-2 e

TS 3.8-13 UNIT NO. *2 CONTAINMENT ISOLATION VALVES VALVE NUMBER A..

PHASE I CONTAINMENT ISOLATION (SAFETY INJECTION SIGNAL)

1.

MOV-2867C

2.

MOV-2867D

3.

MOV-2289A

4.

MOV-2381

5.

HCV-2200A

6.

HCV-2200B

7.

HCV-2200C

8.

TV-SI-201A

9.

TV-SI-201B

10.

TV-SI-200

11.

TV-VG-209A

12.

TV-VG-209B

. 13.

TV-DG-208A

14.

TV-DG-208B

15.

TV-CC-209A*

16.

TV-CC-209B*

17.

TV-SS-200A

18.

TV-SS-200B

19.

TV-SS-201A

20.

TV-SS-201B

,.v..

~

FUNCTION Boron Injection Tank Outlet Boron Injection Tank Outlet Charging Line Reactor Coolant Pump Seal Water Return Letdown Orifice Isolation Letdown Orifice Isolation Letdown Orifice Isolation Accumulator N2 Relief Line Accumulator N2 Relief Line Accumulator N2 Relief Line Primary Drain Transfer Tank Vent Primary Drain Transfer Tank Vent Primary Drain Transfer Pump Discharge Primary Drain Transfer Pump Discharge Component Cooling from RHR's Component Cooling from RHR's Pressurizer Liquid Sample Pressurizer Liquid Sample Pressurizer Vapor Sample Pressurizer Vapor Sample

-~--- -

. -*--*--**..,,... *.. ~-- -

..~.

e TABLE 3. 8.-2 e

TS 3.8-14 UNIT NO. 2 CONTAINMENT ISOLATION VALVES (Continued)

VALVE NUMBER

21.

TV-SS-203

22.

TV"."'SS-206A

23.

TV-SS-206B 24..

TV-ss~202A

25.

TV'.'"SS-202B

26.

TV-SS-204A

27.

TV-SS-204B

28.

TV-CH-2204

29.

TV~PG-2519A

30.

TV-BD-200A*

31.

TV-BD-200B*

32.

TV-BD-200C*

33.

TV-BD-200D*

34.

TV-BD-200E*

35.

TV-BD-200F*

36.

TV-DA-200A

37.

TV-DA-200B

38.

TV-MS-209*

39.

TV-MS-210*

40.

TV-LM-200A

41.

TV-LM-200B

42.

TV-LM-200C FUNCTION Residual Heat Removal System Sample Reactor Coolant Hot Leg Sample Reactor Coolant Hot Leg Sample Reactor Coolant Cold Leg Sample Reactor Coolant Cold Leg Sample Pressurizer Relief Tank Vapor Sample Pressurizer Relief Tank Vapor Sample Letdown Isolation Valve Primary Grade Water to Pressurizer Relief Tank Steam Generator Blowdown Valve Steam Generator Blowdown Valve Steam Generator Blowdown Valve Steam Generator Blowdown Valve Steam Generator Blowdown Valve Steam Generator Blowdown Valve Containment Sump Pump Isolation Containment Sump Pump Isolation Main Steam Drain Trip Valve Main Steam Drain Trip Valve Containment Isolation Monitoring Containment Isolation Monitoring.

Containment Isolation Monitoring

e TABLE 3.8-2 e

TS 3.8-15 UNIT NO. 2 CONTAINMENT ISOLATION VALVES (Continued)

VALVE NUMBER

43.

TV-LM-200D

44.

TV-LM-200E

45.

TV..:LM-200F

46.

.TV-LM-200G

47.

TV-LM-200H

48.

TV-CV-250A

49.

TV-CV-250B so.

TV-LM-201A

51.

TV-LM-20IB

52.

TV-CV-250C

53.

TV-CV-250D

54.

TV-SV-202A B.

PHASE II CONTAINMENT ISOLATION (HI CLS SIGNAL)

1.

TV-RM-200A

2.

TV-RM-200B

3.

TV-RM-200C

4.

TV-IA-201A.

s.

TV-IA-20IB FUNCTION Containment Isolation Monitoring Containment Isolation Monitoring Containment Isolation Monitoring Containment Isolation Monitoring Containment Isolation Monitoring Containment Vacuum Suction Valve Containment Vacuum Suction Valve Leakage Monitoring Sealed Reference Leakage Monitoring Sealed Reference Containment Vacuum Suction Valve Containment Vacuum Suction Valve Conde~ser Air Ejector Vent Trip Valve Containment Air & Particulate Rad. Mon. TV's Containment Air & Particulate Rad. Mon. TV's Containment Air & Particulate Rad. Mon. TV's I

Containment Instr. Air Compressor Suction Containment Instr. Air Compressor Suction

C.

D.

TABLE 3.8-2 e

TS 3.8-16 UNIT NO. 2 CONTAINMENT ISOLATION VALVES (Continued)

VALVE NUMBER FUNCTION PHASE III CONTAINMENT ISOLATION (HI-HI CLS SIGNAL)

1.

TV-MS-201A*

Main Steam Trip Valve

2.

TV-MS-201B*

Main Steam Trip Valve

3.

TV-IA-200 Containment Instr. Air Compressor Disch~

4.

TV-MS-201C*

Main Steam Trip Valve

s.

TV-CC-207*

CC from RCP Thermal Barriers

6.

TV-CC-201A*

-CC from A Air Recirc.

7.

TV-CC-201B*

CC from B Air Recirc.

. 8.

TV-CC-201C*

CC from C Air Recirc.

9.

TV-CC-205A*

CC from "A" RCP

10.

TV-CC-20SB*.

CC from "B" RCP

11.

TV-CC-205C*

CC from "C" RCP CONTAINMENT PURGE & EXHAUST

1.

MOV-VS-200C R.C. Purge Exhaust MOV's

2.

MOV-VS-200D R.C. Purge Exhaust MOV's

3.

MOV-VS-201 R.C. Purge Exhaust Bypass MOV

4.

MOV-VS-200A R.C. Purge Supply MOV's

s.

MOV-VS-200B R.C. Purge Supply MOV's Vlv.

6.

MOV-VS-202 Contain. Vacuum Breaker Atmos. Supply MOV

.. -*, -~*-,*-* -.-..

1*...

I I

I i

I I

I.*

I

'.1 e

'. --~

....... -.. *-*.... *... =*~.:..

l,,*

TABLE 3.8-2

cc.*-.-*

e TS 3.8-17 UNIT NO. 2 CONTAINMENT ISOLATION VALVES (Continued)

VALVE NUMBER E.

REMOTE MANUAL VALVES

1.

MOV-CS-201A

2.

MOV-CS-201B

3.

MOV-CS-201C

4.

MOV-CS-201D

5.

MOV-RS-255A

6.

MOV-RS-255B

7.

MOV-RS-256A

8.

MOV-RS-256B

9.

MOV-2842

10.

MOV-RH-200

11.

FCV-2160

12.

MOV-2890A

13.

MOV-2890B

14.

MOV-2890C

15.

MOV-2869A

16.

MOV..;2869B

17.

MOV-2860A

18.

MOV-2860B

19.

MOV-SW-204A*

20.

MOV-SW-204B*

21.

MOV-SW-204C*

FUNCTION Containment Spray Discharge Valve Containment Spray Discharge Valve Containment Spray Discharge Valve Containment Spray Discharge Valve Outside Recirculation Spray Suction Valve Outside Recirc. Spray Suction Valve Outside Recirc. Discharge Valve Outside Recirc. Discharge Valve Bypasses Boron Injec. Tank to Cold Leg Injec.

Resi. Heat Remov. to RWST Loop Fill Header Flow Valve Lo Header S.I. Pump Disch. from Hot Leg Lo Header S.I. Pump Disch. from Hot Leg Lo Header S.I. Pump Disch. from Cold Leg

!so. from Hot Leg to Hi Header S. I. Line A

!so. from Hot Leg to Hi Header S. I. Line B Iso. from Sump to Lo Header S. I.

!so. Valve from Sump to Lo Header S. I.

SW to "A" HX's SW to "B" HX's SW to "C" HX's

e TABLE 3.8-2 e

TS 3.8-18 UNIT NO. 2 CONTAINMENT ISOLATION VALVES (Continued)

VALVE NUMBER

22.

MOV-SW-204D*

23.

MOV-SW-205A*

24.

MOV-SW-205B*

25.

MOV-SW-205C*

26.

MOV-SW-205D*

27.

HCV-CV-200 F.

MANUAL VALVES

1.

2-SI-150

2.

2-SI-3Z

3.

2-GW-182

4.

2-GW.l.183

5.

2-SA-60

6.

2-SA-62

7.

2-IA-446

8.

2-VA-1

9.

2-VA-6

10.

2-IA-446

11.

2-GW-175

12.

2-GW-166

13.

2-GW-174

14.

2-FP-151

15.

2-FP-152 FUNCTION SW to IID" HX' s SW from "A" HX's SW from "B" HX's SW from "C" HX's SW from "D" HX's Cont. Vacuum Isolation Boron Injection Tank 1" line Accumulator Fill Valve Discharge from Hydrogen Analyzer Discharge from Hydrogen Analyzer Service Air Service Air Instrument Air to Containment Outside Isolation from Primary Vent Pot Inside Isolation from Primary Vent Pot Cross Tie from #1 Instrument Air Header Suction from Cont. to H2 Analyzer Suction from Cont. to H2 Analyzer Inlet to Cont. from H2 Analyzer Outside Cont.

Outside Iso. Vlv for Cont. Fire Protection Outside Iso. Vlv for Cont. Fire Protection

l**.--c~.--~

VALVE NUMBER

16.

2-RL-3

17.

2-RL-5

18.

2-RL-13

19.

2-RL-15

20.

2-SI-73

21.

2-SI-174

22.

2-SW-208

23.

2-SW-106

24.

2-cv-2 e

TABLE 3.8-2

~ -'*"** ~..

e TS 3.8-19 UNIT NO. 2 CONTAINMENT ISOLATION VALVES (Continued)

FUNCTION Inlet Vlv to Cavity from RCS Outside Cont.

Inlet Vlv to Cavity from RCS*Inside Cont.

Suction Vlv to 2-RL-P-lA Inside Containment Suction Vlv to 2-RL-P-lA Outside Containment Accumulator N2 Fill Vlv Outside Containment Bypasses MOV-1869A RS HX SW Drain RS HX SW Drain Cont. Vacuum Isolation G.

CONTAINMENT CHECK VALVES

1.

2-FP-153

2.

2-VP-12

3.

2-RS-17

4.

2-RS-11

5.

2-CS-13

6.

2-CS-24

7.

2-IA-938

8.

2-IA-446.

9.

2-SI-234

10.

2-IA-939

11.

2-IA-446 Inside Cont. - Fire Protection Header Inside Cont. - Air Eject Disch to Cont.

Inside Cont. - RS Disch to Cont. A Inside Cont. - RS Disch to Cont. B Inside Cont. - Discharge of 2-CS-P-lA Inside Cont. - Discharge of 2-CS-P-lB Inside Cont. - Disch of Cont. IA Component Manual Valve - Disch. of IA Component Unit #2 Check Inside Cont. - N2 to Accumulator Check Inside Cont. - Disch. of Cont. IA Component Unit #2 Manual Vlv - Disch. of Unit 2 Instr. Air Comp.

.I

... ; :.:..,:: ~:*,**.:"..'......

e TABLE 3.8-2 e

TS 3.8-20 UNIT NO. 2 CONTAINMENT ISOLATION VALVES (Continued)

VALVE NUMBER FUNCTION

12.

2-RC-160 Check Valve Inside Contain. from PG Supply

13.

2-RM-3 Check Valve Inside Contain. - Rad. Monitoring Sue.

14.

2-IA-939 Instr. Air Check Valve to Containment

15.

2-SA-446 Service Air Check Valve to Containment

16.

2-CC-177*

cc to "A II RllR HX

17.

2-CC-176*

cc to "B II RllR HX

18.

2-SI-225 HHS! from BIT

19.

2-CC-242*

CC to "A" Air Recirc.

I 20..

2-CC-233*

CC to "B" Air Recirc.

21.

2-CC-224*

CC to "C" Air Recirc.

22.

2-CH-309 Normal Chg. Hdr

23.

2-CC-1*

CC to "A" RCP

24.

2-CC-58*

'cc to "B" RCP

25.

2-CC-59*

cc to "C" RCP

26.

2-SI-224 HHSI BIT Bypass

27.

2-SI-226 HHS! to Hot Legs

28.

2-SI-228 LHSI Pp Discharge

29.

2-SI-229 LHSI Pp Discharge

30.

2-SI-227 LHSI to Hot Leg

- Not subject to Type "C" Testing.

- - Modifications to this table should be submitted to the NRC as part of the next license amendment.

i l

1

-~~ - **---...... --*.

e TS 4.1-1 4.1 OPERATIONAL SAFETY REVIEW Applicability Applies to items directly related to safety limits and limiting conditions for operation.

Objective To specify the minimum frequency and type of surveillance to be applied to unit equipment and conditions.

Specification A.

Calibration, testing, and checking of instrumentation channels shall be performed as detailed in Table 4.1-1.

B.

Equipment tests shall be conducted as detailed below and in Table 4.1-2A.

1.

Each Pressurizer PORV shall be demonstrated operable:

a.

At least once per 31 days by performance of a channel functional test, excluding valve operation, and

b.

At least once per 18 months by performance of a channel calibration.

e TS 4.1-la

2.

Each Pressurizer PORV block valve shall be demonstrated operable at least once per 92 days by operating the valve through one complete cycle of full travel.

3.

The pressurizer water volume shall be determined to be within its limit as-defined in Specification 2.3.A.3.a at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months whenever the reactor is not subcritical by at least 1% ak/k.

C.

Sampling tests shall be conducted as detailed in Table 4.1-213.

D.

Whenever containment integrity is not required, only the asterisked items in Table 4.1-1 and 4.1-2A and 4.1-2B are applicable.

E.

Flushing of sensitized stainless steel pipe sections shall be conducted as detailed in TS Table 4.1-3A and 4.1-3B.

1.

2.

3.

4.

5.

6.

7.

8.

9

I CHANNEL DESCRIPTION Nuclear Power Range Nuclear Intermediate Range Nuclear Source Range Reactor Coolant Temperature Reactor Coolant Flow Pressurizer Water Level Pressurizer Pressure (High &

Low) 4 Kv Voltage and Frequency Analog Rod Position TABLE 4.1-1 MINIMUM FREQUENCIES FOR CHECK, CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS CHECK s

M(3)

S(l)

S(l)

S s

s s

s

S(l,2)

(4)

CALIBRATE D (1)

Q (3)

N.A.

R R

TEST BW(2)

P(2)

BW(l)

BW(2)

M M

M(3)

REMARKS

1) Against a.heat balance standard

2) Signal of AT; bistable action (permissive, rod stop, strips)

A

3) Upper and lower chambers for syme~c offset by means of the moveable incore detector system.

1) Once/shift when in service

2) Log level; bistable action (permissive, rod stop, trip)

1) Once/Shift when in service

2) Bistable action (alarm, trip)

1) Overtemperature - AT

2) Overpower - AT e

Reactor protection circuit only

1) With step counters

--l

(/)

2) Each six inches of rod motion

.i::,.

when data logger is out of I

service

0)

3) Rod bottom bistable action

4) NA when reactor is in cold shut-down

.... ~*

r

/

[

TABLE 4.1-1 (Continued) t l Ii fl

/,

CHANNEL DESCRIPTION CHECK CALIBRATE TEST REMARKS

10.

Rod Position Bank Counters S(l,2)

N.A.

1) Each six inches of rod motion ii when data logger is out of service

2) With analog rod position

11.

Steam Generator Level s

R*

M

12.

Charging Flow N.A.

R N.A.

13. Residual Heat Removal Pump Flow N.A R

N.A.

e

14.

Boric Acid-Tank Level

D R

N.A.

15. Refueling Water Storage s

R M

Tank Level

16. Boron Injection Tank Level w

N.A.

17.

Volume Control Tank Level N.A.

R N.A.

1~.

Reactor Containment Pressure-CLS

D R

M(l)

1) Isolation Valve signal and spray signal

19.

Processing and Area Radi~tion

D R

M Monitoring Systems e

20.

Boric Acid Control N.A.

]l N.A.

21.

Containment Sump Level N.A.

R N.A.

22.

Accumulator Level and Pressure s

R N.A.

23.

Containment Pressure-Vacuum Pump s

R

-I N.A.

(/)

System

.i:,,..

24.

Steam Line Pressure s

I R

M

/

CHANNEL DESCRIPTION

25.

Turbine First Stage Pressure

26.

Emergency Plan Radiation Instr.

27.

Environmental Radiation Monitors

28.

Logic Channel Testing

29.

Turbine Overspeed Protection Trip Channel (Electrical) 3Q.

Turbine Trip Setpoint

31.

Seismic Instrumentation

32. Reactor Trip Breaker

33. Reactor Coolant Pressure (Low)

34.

Auxiliary Feedwater

a.

Steam Generator Water Level Low-Low

b.

RCP Undervoltage TABLE CHECK s

M

M N.A.

N.A.

M N.A.

N.,A.

s s

4.1-1 CALIBRATE R

R N~A.

N.A.

R R

SA N.A.

R R

R

',.... '. ~-.

TEST M

M N.A.

M R

R M

M N.A.

M M

c.

S.I.

(All Safety Injection surveillance requirements)

d.

Station Blackout N.A.

R

e.

Main Feedwater Pump Trip N.A.

N.A.

S - Each shift D - Daily W - Weekly NA - Not applicable SA - Semiannually Q - Every 90 effective M - Monthly P - Prior to each startup if not done previous week R - Each Refueling Shutdown BW - Every two week~

AP - After each startup if not done previous week full power days

See Specification 4.lD N.A.

R

~

REMARKS TLD Dosimeters Stop valve closure or low EH fluid pressure e

--i

(./)

I 00 l

11

'I Ir'.

\\!,:,.

I I

TABLE 4.1-2 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL INSTRUMENT CHECK

1.

Auxiliary Feedwater Flow Rate p

2.

Reactor Coolant System Subcooling Margin Monitor M

3.

PORV Position Indicator (Primary Detector)

M

4.

PORV Position Indicator (Backup Detector)

M

s.

PORV Block Valve Position Indicator M

6.

Safety Valve Position Indicator M.

7.

Safety Valve Position Indicator (Backup Detector)

M

CHANNEL CALIBRATION R

R R

-I ti)

I I.O Ill

DESCRIPTION

1.

Control Rod Assemblies

2.

Control Room Assemblies

3.

Refueling Water Chemical Addition Tank

4.

Pressurizer Safety Valves

5.

Main Steam Safety Valves

6.

Containment Isolation Trip 7.. Refueling System Interlocks

8.

Service Water System 9,

Fire Protection Pump and Power Supply

10.

Primary System Leakage

11. Diesel Fuel Supply

12.

Boric Acid Piping Heat Tracing Circuits

13.

Main Steam Line Trip

'* :..,. *--* -**~..... *.. ;....:.... : :*. _

~*, *:.. _*.. :; *- *--.,.* *--......

,,'.. *~**.'. '.~..... ':.. _.,

. i..

TABLE 4.1-2A MINIMUM FREQUENCY FOR EQUIPMENT TESTS TEST Rod drop times of all fuli length rods at hot and cold conditions Partial movement of all rods Functional Setpoint Setpoint

Functional

Functional

Functional Functional

Evaluate

Fuel Inventory

Operational Functional (1) Full closure (2) Partial closure "FREQUENCY Each refueling shutdown.or after disassembly or maintenance re-quiring the breech of the Reactor Coolant System integrity Every 2 weeks Each refueling shutdown Each refueling shutdown Each refueling shutdown Each refueling shutdown Prior to refueling Each refueling shutdown Monthly Daily

.5 days/week Monthly (1) Each cold shutdown (2) Before each startup FSAR SECTION REFERENCE 7

7 6

\\ i '

I i

, i 1 i' I I l I /

. i Ii i/

1*.

1:

1*.'

I*

/

Ii 4 --

I 10 I

5 9.12 9.9 9.10 4

8.5 9.1 9:

10

--i

(/)

~-.

I I.D C"

TABLE 4.l-2A (CONTINUED)

. MINIMUM FREQUENCY FOR EQUIPMENT TESTS

14.

15.

DESCRIPTION Service Water System Valves in Line Supplying Recircu-lation Spray Heat Exchangers Control Room Ventilation System TEST Functional

Ability to maintain positive pres-sure for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months using a volume of air equivalent to or less than stored in the bottled air supply

16.

Reactor Vessel Overpressure Functional & Setpoint Mitigating System (except backup air supply)

17.

Reactor Vessel Overpressure Setpoint Mitigating System Backup*

Air Supply FREQUENCY Each refueling Each refueling interval (approx. every 12-18 months)

Prior to decreasing RCS temperature below 3S0°F and monthly while the RCS is <350°F and the Reactor Vessel Head is bolted Refueling FSAR SECTION REFERENCE 9.9 9.13 None None.

-I

(/)

~

I I.O n

i' I

DESCRIPTION TABLE 4.1-2A (CONTINUED)

MINIMUM FREQUENCY FOR EQUIPMENT TESTS TEST FREQUENCY

. :.. -**-*-.. '***:--*-~...... *.. :.. -~-- -... : *.

FSAR SECTION REFERENCE

18.

Primary Coolant System Functional

1. Periodic leakage (a) on each valve listed in Specification 3.1.C.7a shall be accomplished prior to entering power operation condition after every time the plant is placed in the cold shutdown (a)

(b) condition for refueling, after each time the..

plant is placed in cold shutdown condition

'11111111111' for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months if testing has not been accomp-lished in the preceeding 9 months, and prior to returning the valve to service after maintenance, repair or replace-ment work is performed.

To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved procedures ~nd supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.

Minimum differential test pressure shall not be below 150 psid.

See Specification 4.1.D.

e

--f

(/)

I I.O

a.

i* n

ML18139B504

Text

References:

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