ML18139B504

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Tech Spec Changes Re Auxiliary Feedwater Flow Rate Instrumentation Surveillance Requirements & Containment Isolation Valves
ML18139B504
Person / Time
Site: Surry  Dominion icon.png
Issue date: 08/21/1981
From:
VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML18139B503 List:
References
NUDOCS 8109010278
Download: ML18139B504 (59)
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Text

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.
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b.

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  • TS 3.1-2 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 The shutdown rods may remain withdrawn.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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It TS 3.1-3

e. 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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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 <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least hot shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and the reactor coolant system temperature and pressure less than 350°F and 450 psig, respectively, within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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c. With the pressurizer otherwise inoperabl~, be in at least hot shutdown with the reactor trip breakers open within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and the reactor coolant system temperature and pressure less than 350°F and 450 psig, respectively, within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
6. 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 <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 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 in at least hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold 0

shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

c. With one or more block valve(s) inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Basis 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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  • 1 e e 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 t'

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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e e 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 e TS 3.7-1 3.7 INSTRUMENTATION SYSTEMS Operational Safety Instrumentation Applicability:

Applies to reactor and safety features instrumentation systems.

Objectives:

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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
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 <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

e e 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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e e 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

e e 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.

e e TS 3.7-7 Auxiliary Feedwater System Actuation

.J 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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  • TS 3.7-8
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 (3) accident as shown in the safety analysis.
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 Tavg. setting limit for SIS and 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 )

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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 4

information on the Process Radiation Monitoring System is available in the FSAR. ( )

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

monitor indication.

  • TS 3.7-9 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

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i TABLE 3. 7-1 REACTOR TRIP INSTRUMENT OPERATING CONDITIONS 1 2 3 4 OPERATOR ACTION IF CONDITIONS OF DEGREE COLUMN 1 OR 2 MIN. OF EXCEPT AS CONDI-OPERABLE REDUN- PERMISSIBLE BYPASS TIONED BY COLUMN 3 FUNCTIONAL UNIT CHANNELS DANCY CONDITIONS CANNOT BE MET

1. Manual 1 Maintain hot e shutdown
2. Nuclear Flux Power Range 3 2 Low trip setting when 2 *Maintain hot of 4 power channels greater shutdown than 10% of full power *
3. Nuclear Flux Intermediate 1 2 of 4 power channels greater Maintain hot Range than 10% full power shutdown
4. Nuclear Flux Source Range 1 1 of 2 intermediate rang~

10 Maintain.hot channels greater than 10 shutdown amps

s. Overtemperature aT 2 1 Maintain hot shutdown
  • 6. Overpower aT 2 1 Maintain hot e

shutdown

7. Low Pressurizer Pressure 2 1 3 of 4 nuclear power channels Maintain hot and 2 of 2 turbine load shutdown channels less than 10% of -l rated *power (/)

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8. Hi Pressurizer Pressure 2 1 Same as Item 7 above Maintain hot -....J I

shutdown 0

TABLE 3.7-1 REACTOR TRIP INSTRUMENT OPERATING CONDITIONS 1 2 3 4 OPERATOR ACTION IF CONDITIONS OF DEGREE COLUMN 1 OR 2 MIN. OF EXCEPT AS CONDI-OPERABLE REDUN- PERMISSIBLE BYPASS TIONED BY COLUMN 3 I FUNCTIONAL UNIT CHANNELS DANCY CONDITIONS CANNOT BE MET

9. Pressurizer-Hi Water Level
  • 2 1 3 of 4 nuclear power channels and 2 of 2 Maintain hot shutdown e:

turbine load channels '.I

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less than 10%*of rated power

10. Low Flow 2/ operable If inoperable loop Maintain hot loop channels are not in service shutdown  :

they must be placed in the tripped mode

11. Turbine Trip 2 1 Maintain less than 10% rated power
12. Lo-Lo Steam Generator 2/non-iso- I/non- Maintain hot Water Level lated loop isolated loop shutdown
13. Underfrequency 4KV Bus 2 1 Maintain hot ~

shutdown

14. Undervoltage 4KV Bus 2 1 Maintain hot shutdown

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TABLE 3.7-1 REACTOR TRIP INSTRUMENT OPERATING CONDITIONS 1 2 3 4 OPERATOR ACTION IF CONDITIONS OF DEGREE COLUMN 1 OR 2 MIN. OF EXCEPT AS CONDI- .

  • OPERABLE REDUN- PERMISSIBLE BYPASS TIONED BY COLUMN 3 .

FUNCTIONAL UNIT CHANNELS DANCY CONDITIONS CANNOT BE MET

15. Control rod misalignment Monitor**

a) rod position deviation 1 Log individual rod positions once/hour, and after a load change

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

b) quadrant power tilt 1 Log individual upper monitor (upper and upper and lower ion lower excore neutron chamber currents once/

detectors) hour and after a load change> 10% or after

> 30 inches of control rod motion.

16. Safety Injection See Item 1 of TS Table 3~7-2 e

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TABLE 3.7-1 REACTOR TRIP INSTRUMENT OPERATING CONDITIONS i.

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OPERATOR ACTION IF CONDITIONS OF DEGREE *coLUMN 1 OR 2 MIN. OF EXCEPT AS CONDI-OPERABLE REDUN- PERMISSIBLE BYPASS TIONED BY COLUMN 3 FUNCTIONAL UNIT CHANNELS DANCY CONDITIONS CANNOT BE MET

17. Low steam generator water level with steam/feedwater I/non-iso-lated loop I/non-iso-Maintain hot shutdown e mismatch flow lated loop
    • If both rod misalignment monitors (a and b) inoperable for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or more, the nuclear overpower trip shall be reset to 93 percent of rated power in addition to the increased surveillance noted.

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TABLE 3.7-2 ENGINEERED SAFEGUARDS ACTION 1 2 3 4 OPERATOR ACTION IF CONDITIONS OF DEGREE COLUMN 1 OR 2 MIN. OF .EXCEPT AS CONDI-OPERABLE REDUN- PERMISSIBLE BYPASS TIONED BY COLUMN 3 FUNCTIONAL UNIT CHANNELS DANCY CONDITIONS CANNOT BE MET

1. SAFETY INJECTION
a. Manual 1 0 Cold shutdown e
b. High Containment Press. 3 1 Cold shutdown
c. High Differential Press. 2/non-iso- 1/non- Primary Pressure Cold shutdown between any Steam Line and lated loop isolated less than 2000 psig the Steam Line Header loop except when reactor is critical
d. Pressurizer Low-Low Press. 2 1 Primary Pressure Cold shutdown less than 2000 psig except when reactor is critical
e. High Steam Flow in 2/3 1/steamline *** .Reactor Coolant aver- Cold shutdown Steam Lines with Low T 2 T signals 1 age temperature less or Low Steam Line Pres~~g 2 sti§m Press. 1 than 543°F (nominal)

Signals during heatup and

.cooldown

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

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TABLE 3.7-2 ENGINEERED SAFEGUARDS ACTION INSTRUMENT OPERATING CONDITIONS 1 2 3 4 OPERATOR ACTION IF CONDITIONS OF DEGREE COLUMN 1 OR 2 MIN. OF EXCEPT AS CONDI-OPERABLE REDUN- PERMISSIBLE BYPASS TIONED BY COLUMN 3 FUNCTIONAL UNIT CHANNELS DANCY CONDITIONS CANNOT BE MET

2. CONTAINMENT SPRAY Manual Cold shutdown e
a. 2 **
b. High Containment Press. 3 1 Cold shutdown (Hi-Hi Setpoint)
3. AUXILIARY FEEDWATER
a. Steam Generator Water Level Low-Low
i. Start Motor 2/Stm. Gen. 1 Loop Stop Valve in res- Place inoperable Driven*Pumps pective loop closed channel in Tripped II. Start Turbine 2/Stm. Gen. 1 condition within Driven Pumps one hour b.

c.

RCP Undervoltage Start Turbine Driven Pump Safety Injection 2 1 (All safety injection initiating functions and requirements)

Place inoperable channel in Tripped condition within one hour -

Start Motor Driven Pumps

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TABLE 3.7-2 ENGINEERED SAFEGUARDS ACTION INSTRUMENT OPERATING CONDITONS 1 2 3 4 OPERATOR ACTION IF CONDITIONS OF DEGREE COLUMN 1 OR 2 MIN. OF EXCEPT AS CONDI-i' OPERABLE REDUN- PERMISSIBLE BYPASS TIONED BY COLUMN 3 d.

FUNCTIONAL UNIT Station Blackout Start Motor* Driven Pump CHANNELS 2

DANCY 0

CONDITIONS CANNOT BE MET Restore inoperable channel within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in hot shutdown within

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next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the follow-ing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

e. Trip of Main Feedwater Pumps 1/Pump I/Pump Restore inoperable Start Motor Pumps channel within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in hot shutdown within next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the follow-ing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

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TABLE 3.7-3 INSTRUMENT OPERATING CONDITIONS FOR ISOLATION FUNCTIONS INSTRUMENT OPERATING CONDITIONS 1 2 3 4 OPERATOR ACTION IF CONDITIONS OF DEGREE COLUMN 1 OR 2 MIN. OF EXCEPT AS CONDI-OPERABLE REDUN- PERMISSIBLE BYPASS TIONED BY COLUMN 3 FUNCTIONAL UNIT CHANNELS DANCY CONDITIONS CANNOT BE MET

1. CONTAINMENT ISOLATION e
a. Safety Injection See Item No. 1 of Table 3.7-2 Cold shutdown
b. Manual 1 Hot shutdown
c. High Containment Press. 3 1 Cold shutdown (Hi Setpoint)

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d. High Containment Press. 3 1 Cold shutdown
2. STEAM LINE ISOLATION
a. High Steam Flow in 2/3 lines 1/steamline *** Cold shutdown and 2/3 Low Tavg or 2/3 2/T 1 av2 Low Steam Pressure signars 2 Stm. Press. 1 signals
b. High Contaimnent Pres.s. 3 1 Cold shutdown (Hi-Hi Level)
c. Manual 1/line Hot shutdown

-I

(./)

3. FEEDWATER LINE ISOLATION w

-....J

a. Safety Injection See Item No. 1 of Table 3.7-2 Cold shutdown I

-....J

      • With the specJfied minimum operable channels the 2/3 high steam flow is already in the trip mode

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

TABLE 3.7-4 ENGINEERED SAFETY FEATURE SYSTEM INITIATION LIMITS INSTRUMENT SETTING No. FUNCTIONAL UNIT CHANNEL ACTION SETTING LIMIT 1 High Containment P_ressure (High Contain-. a) Safety Injection ~5 psig ment Pressure Signal) b) Containment Vacuum Pump Trip c) High Press. Containment Iso.

d) Safety Injec.tion Contain. Iso.

e) F.W. Line Isolation 2 High High Containment Pressure (High High a) Containment Spray ~25 psig Containment Pressure Signals) b) Recirculation Spray c) Steam Line Isolation d) High High Press. Contain. !so.

3 Pressurizer Low Low Pressure a) Safety Injection ~l, 700 psig  :.*':l b) Safety Injection Cont. !so.*

c) Feedwater Line Isolation 4 High Differential Pressure Between a) Safety Injection ~150 psi Steam Line and the Steam Line Header b) Safety Injection Contain. !so.,

c) F.W. Line Isolation 5 High Steam Flow in 2/3 Steam Lines a) Safety Injection ~40% (at zero load) of full steam flow

~40% (at 20% load) of full steam flow b) Steam Line Isolation ~110% (at full load) of c) Safety Injection Contain. !so. full steam flow f .

Coincident with Low T

. Line Pressure

. a\7g or Low Steam d) F.W. Line Isolation

~541°F T

~500 psig steam line pressure avg -

-I

(/)

w

_,I CX)

TABLE 3.7-4 ENGINEERED SAFETY FEATURE SYSTEM INITIATION LIMITS INSTRUMENT SETTING No. FUNCTIONAL UNIT CHANNEL ACTION SETTING LIMIT 6 AUXILIARY FEEDWATER

a. Steam Generator Water Level Low-Low Aux. Feedwater Initiation ~5% narrow range S/G Blowdown Isolation
b. RCP Undervoltage Aux. Feedwater Initiation ~70% nominal
c. Safety Injection Aux. Feedwater Initiation All S.I. setpoints d.

e.

Station Blackout Main Feedwater Pump Trip Aux. Feedwater Initiation Aux. Feedwater Initiation

~46.7% nominal N.A.

e

-I

(/)

j" w

I U)

TABLE 3.7-5  :

I

\

AUTOMATIC FUNCTIONS OPERATED FROM RADIATION MONITORS ALARM AUTOMATIC FUNCTION MONITORING ALARM SETPOINT MONITOR CHANNEL AT ALARM CONDITlONS REQUIREMENTS µCI/cc

1. Pr.6cess vent particulate Stops discharge from contain. See Specifications -8 Particula!~ ~4x10 and gas monitor~. . vacuum systems and waste 3.11 and 4.9 Gas ~9x10

. (RM,-GW-101 & RM-GW-102) gas decay tanks (shuts Valve Nos. RCV-GW-160, FCV-GW-260, FCV-GW-101)

2. Component cooling water Shuts surge tank vent valve See Specifications ~Twice Backgroun~

radiation monitors HCV-CC-100 3.13 and 4.9 (RM-CC-105 & RM-CC-106) d 3 . . Liquid waste disposal Shuts effluent discharge See Specifications ~1.SxlO -3 radiation monitors valves FCV-LW-104A and 3.11 and 4.9

-(RM-LW-108) FCV-LW-104B

4. Condenser air ejector Diverts flow to the contain- See Specification ~1.3 radiation monitors ment of the affected unit 3.11 and 4.9 (RM-SV-111 & RM-SV-211) (Opens TV-SV-102 and shuts TV-SV-103 or opens TV-SV-202 and shuts. TV-SV-203)
5. Containment particulte Trips affected unit's purge See Specifications Particula!s ~9xl0- 9 and gas monitors (RM-RMS-159 & RM-RMS-160, RM-RMS-259 & RM-RMS-260) supply and exhaust fans, closes affected unit's purge air butterfly valves (MOV-VS-lOOA, B, C & Dor MOV-VS-200A, B, C & D) 3.10 and 4.0 Gas ~lxlO
6. Manipulator crane area Trips affected unit's purge See Specifications ~50 mr:em/hr monitors (RM-RMS-162 & supply and exhaust fans, 3.10 and 4.9 -i RM-RMS-262) closes affected unit's <.n '.'

purge air butterfly valves .w (MOV-VS-lOOA, B, C & Dor -....J I

MOV-VS-200A, B, C & D ...~ -

N 0

!.~ t*

TABLE 3.7-6 ACCIDENT MONITORING INSTRUMENTATION TOTAL NO. MINIMUM CHANNELS I*

I INSTRUMENT OF CHANNELS OPERABLE

1. Auxiliary Feedwater Flow Rate 1 per S/G 1 per S/G
2. Reactor Coolant System Subcooling Margin Monitor 2 1
3. PORV Position Indicator (Primary Detector) I/valve I/valve 4.

s.

PORV Position Indicator (Backup Detector)

PORV Block Valve Position Indicator I/valve 0 e

I/valve I/valve

6. Safety Valve Position Indicator (Primary Detector) I/valve I/valve
7. Safety Valve Position Indicator (Backup Detector) I/valve 0

--i

(/)

........w I

N

--1

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

3.8 CONTAINMENT e

  • TS 3.8-1 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 .

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.

B. 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.

Basis 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 <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, 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 Reactor Coolant Pump FSAR Section 5.2 Containment Isolation FSAR Section 5.2.1 Design Bases FSAR Section 5.5.2 Isolation Design

e TS 3.8-5 TABLE 3.8-1**

UNIT NO. 1 CONTAINMENT ISOLATION VALVES VALVE NUMBER FUNCTION A. 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 N Relief Line 2
10. TV-SI-100 Accumulator N Relief Line 2
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
  • ~***-***-- *-*---. *- - ..... _' *~* -

e ** TS 3.8-6 TABLE 3.8-1**

UNIT NO. 1 CONTAINMENT ISOLATION VALVES (Continued)

VALVE NUMBER FUNCTION

. i

! 21. TV-SS-103 Residual Heat Removal System Sample

22. TV-SS-106A Reactor Coolant Hot Leg Sample
23. TV-SS-106B Reactor Coolant Hot Leg Sample
24. TV-SS-102A Reactor Coolant Cold Leg Sample
25. TV-SS-102B Reactor Coolant Cold Leg Sample
26. TV-SS-104A Pressurizer Relief.Tank Vapor Sample
27. TV-SS-i04B Pressurizer Relief Tank Vapor Sample-
28. TV-CH-1204 Letdown Isolation Valve
  • 29. TV-PG-1519A Primary Grade Water to Pressurizer Relief Tank
30. TV-BD-lOOA* Steam Generator Blowdown Valve
31. TV-BD-lOOB* Steam Generator Blowdown Valve
32. TV-BD-lOOC* Steam Generator Blowdown Valve
33. TV-BD-lOOD* Steam Generator Blowdown Valve
34. TV-BD-lOOE* Steam Generator Blowdown Valve
35. TV-BD-lOOF* Steam Generator Blowdown Valve
36. TV-DA-lOOA Containment Sump Pump Isolation

. 37. TV-DA-lOOB Containment Sump Pump Isolation

38.
39. TV-MS-110* Main Steam Drain Trip Valve
40. TV-LM-lOOA Containment Isolatio~ Monitoring
41. TV-LM-lOOB Containment IsoJation Monitoring
42. TV-LM-lOOC Containment Isolation Monitoring
    • --- ¥**- **-*.

e TS 3.8-7 TABLE 3.8-1**

UNIT NO. 1 CONTAINMENT ISOLATION VALVES (Continued)

VALVE NmJBER FUNCTION

43. TV-,LM-lOOD Containment Isolation Monitoring
44. TV-LM-lOOE Containment Isolation Monitoring
45. TV-,LM-lOOF Containment Isolation Monitoring
46. TV-LM-lOOG Containment Isolation Monitoring
47. TV-LM-lOOH Containment Isolation Monitoring
48. TV-:CV-150A Containment Vacuum Suction Valve
49. TV-CV-150B Containment Vacuum Suction Valve
50. -TV-LM-101A Leakage Monitoring Sealed Reference
51. TV-LM..-lOlB Leakage Monitoring Sealed Reference
52. TV-CV-150C Containment Vacuum Suction Valve
53. TV-CV-150D Containment Vacuum Suction Valve
54. TV-SV-102A Condenser Air Ejector Vent Trip Valve B. PHASE II CONTAINMENT ISOLATION (HI CLS SIGNAL)
1. TV-RM-lOOA Containment Air & Particulate Rad. Mon. TV's
2. TV-RM-lOOB Containment Air & Particulate Rad. Mon. TV's
3. TV-RM-lOOC Containment Air & Particulate Rad. Mon. TV's

. 4. TV-IA-lOlA Containment Instr. Air Compressor Suction

5. TV-IA-101B Containment Instr. Air Compressor Suction

e TABLE 3.8-1**

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

VALVE NUMBER FUNCTION C. 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 D. CONTAINMENT PURGE &EXHAUST
1. MOV-VS-lOOC R.C. Purge Exhaust MOV's 2 *. MOV-VS-lOOD R.C. Purge Exhaust MOV's I
3. MOV-VS-101 R.C. Purge Exhaust Bypass MOV I

I .

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 I  :

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

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

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

  • r e TS 3.8-9 TABLE 3.8-1**
  • .1 UNIT NO. 1 CONTAINMENT ISOLATION VALVES (Continued) i VALVE NUMBER FUNCTION E. REMOTE MANUAL VALVES
1. MOV-CS-101A Containment Spray Discharge Valve
2. MOV-CS-101B Containment Spray Discharge Valve
3. MOV"'.'CS-101C Containment Spray Discharge Valve
4. MOV-CS-lOlD Containment Spray Discharge Valve
s. MOV-RS-lSSA Outside Recirc. Spray Suction Valve
6. MOV-RS-lSSB Outside Recirc. Spray Suction Valve
7. MOV-RS-156A Outside Recirc. Discharge Valve
8. MOV-RS-156B Outside Recirc. Discharge Valve
9. MOV-1842 Bypasses Boron Injec. Tank to Cold Leg Injec.
10. MOV-RH-100 Resi. Heat Remov. to RWST
11. FCV-1160 Loop Fill Header Flow Valve
12. MOV-1890A Lo Header S. I. Pump Disch. from Hot Leg
13. MOV-1890B Lo Header S. I. Pump Disch. from Hot Leg I
14. MOV-1890C Lo Header S. I. Pump Disch. from Cold Leg
15. MOV-1869A !so. from Hot Leg to Hi Header S. I. Line A
16. MOV-1869B !so. from Hot Leg to Hi Header S. I. Line B
17. MOV-1860A Iso. from Sump to Lo Header S. I.
18. MOV-1860B !so. Valve from Sump to Lo Header S. I.
19. MOV-SW-I04A* SW to "A" HX' s
20. MOV-SW-104B* SW to "B" HX' s
21. MOV-SW-104C* SW to "C" HX' s

e e TS 3.8-10 TABLE 3.8-1**

UNIT NO. 1 CONTAINMENT ISOLATION VALVES (Continued)

VALVE NUMBER FUNCTION

22. MOV-SW-104D* SW to "D" HX's
23. MOV-SW-105A* SW from "A" HX's
24. MOV-SW-105B* SW from "B" HX's
25. MOV~SW-105C* SW from "C" HX's
26. MOV-SW-105D* SW from "D" HX' s
27. HCV-CV-100 Cont. Vacuum Isolation F. MANUAL VALVES
1. 1-SI-150 Boron Injection Tanlc 1" line
2. 1-SI-32 Accumulator Fill Valve
3. 1-GW-182 Discharge from Hydrogen Analyzer
4. 1-GW-183 Discharge from Hydrogen Analyzer
5. 1-SA-60 Service Air to Containment
6. 1-SA-62 Service Air to Containment
7. 1-IA-446 Instrument Air to Containment
8. 1-VA-1 Outside Isolation from Primary Vent Pot 9.* 1-VA-6 Inside Isolation from Primary Vent Pot
10. 2-IA-446 Cross Tie from #2 Instrument Air Header
11. 1-GW-175 Suction from Containment to H2 Analyzer
12. 1-GW-166 Suction from Containment to H2 Analyzer
13. 1-GW-.174 Inlet to Cont. from H Analyzer Outside Cont.

2

' Outside Iso. Vlv for Cont. Fire Protection

14. 1-FP-151
15. 1-FP-152 Outside Iso. Vlv for Cont. Fire Protection

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

e e TS 3.8-11 TABLE 3.8-1**

UNIT NO. 1 CONTAINMENT ISOLATION VALVES (Continued)

VALVE NUMBER

  • FUNCTION
16. l-RL-3 Inlet Vlv to Cavity from RCS Outside Cont.
17. 1-RL-S Inlet Vlv to *cavity from RCS Inside Cont.
18. l-RL-13 Suction Vlv to 1-RL-P-IA Inside Containment
19. 1-RL-15 Suction Vlv to 1-RL-P-IA Outside Containment
20. 1-SI-73 Accumulator N Fill Vlv Outside Containment 2
21. 1-SI-174 Bypasses MOV-1869A
22. 1-SW-208 RS HX SW Drain
23. l-SW-106 RS HX SW Drain
24. 1-cv-2 Cont. Vacuum Isolation G. CONTAINMENT CHECK VALVES
1. 1-FP-153 Inside Cont. - Fire Protection Header
2. 1-VP-12 Inside Cont. - Air Eject Disch to Cont.
3. 1-RS-17 Inside Cont. - RS Disch to Cont. A
4. 1-RS-11 Inside Cont. - RS Disch to Cont. B
5. l-CS-13 Inside Cont. - Discharge of 1-CS-P-IA
6. l-CS-24 Inside Cont. - Discharge of 1-CS-P-IB
7. 1-IA-938 Inside Cont. - Disch of Cont. IA Component
8. 2-IA-446 Manuai Valve - Disch. of IA Component Unit #2.
9. 1-SI-234 Check Inside Cont. - N2 to Accumulator
10. 1-IA-939 Check Inside Cont.~ Disch. of Cont. IA Component Unit #1
11. 1-IA-446 Manual Vlv - Disch. of Unit 1 Instr. Air Comp.

e e TS 3.8-12 TABLE 3.8-1**

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 AII rum HX
17. 1-CC-176* cc to 11BII rum HX
18. 1-SI-225 IIlISI from BIT
19. 1-CC-242* CC to "A" Air Recirc.
20. 1-CC-233* cc to 11 B11 Air Recirc.
21. 1-CC-224* cc to 11 C" 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 e TS 3.8-13 TABLE 3.8-2**

UNIT NO. *2 CONTAINMENT ISOLATION VALVES VALVE NUMBER FUNCTION A.. PHASE I CONTAINMENT ISOLATION (SAFETY INJECTION SIGNAL)

1. MOV-2867C Boron Injection Tank Outlet
2. MOV-2867D Boron Injection Tank Outlet
3. MOV-2289A Charging Line
4. MOV-2381 Reactor Coolant Pump Seal Water Return
5. HCV-2200A Letdown Orifice Isolation
6. HCV-2200B Letdown Orifice Isolation
7. HCV-2200C Letdown Orifice Isolation
8. TV-SI-201A Accumulator N2 Relief Line
9. TV-SI-201B Accumulator N Relief Line 2
10. TV-SI-200 Accumulator N2 Relief Line
11. TV-VG-209A Primary Drain Transfer Tank Vent
12. TV-VG-209B Primary Drain Transfer Tank Vent

. 13. TV-DG-208A Primary Drain Transfer Pump Discharge

14. TV-DG-208B Primary Drain Transfer Pump Discharge
15. TV-CC-209A* Component Cooling from RHR's
16. TV-CC-209B* Component Cooling from RHR's
17. TV-SS-200A Pressurizer Liquid Sample
18. TV-SS-200B Pressurizer Liquid Sample
19. TV-SS-201A Pressurizer Vapor Sample
20. TV-SS-201B Pressurizer Vapor Sample

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

  • -~--- - . -*--*--** ..,, ...*.. ~-- - :..~. *-* .::,

e e TS 3.8-14 TABLE 3 . 8.-2**

UNIT NO. 2 CONTAINMENT ISOLATION VALVES (Continued)

VALVE NUMBER FUNCTION

21. TV-SS-203 Residual Heat Removal System Sample
22. TV"."'SS-206A Reactor Coolant Hot Leg Sample
23. TV-SS-206B Reactor Coolant Hot Leg Sample 24 .. TV-ss~202A Reactor Coolant Cold Leg Sample
25. TV'.'"SS-202B Reactor Coolant Cold Leg Sample
26. TV-SS-204A Pressurizer Relief Tank Vapor Sample
27. TV-SS-204B Pressurizer Relief Tank Vapor Sample
28. TV-CH-2204 Letdown Isolation Valve
29. TV~PG-2519A Primary Grade Water to Pressurizer Relief Tank
30. *TV-BD-200A* Steam Generator Blowdown Valve
31. TV-BD-200B* Steam Generator Blowdown Valve
32. TV-BD-200C* Steam Generator Blowdown Valve 33 . TV-BD-200D* Steam Generator Blowdown Valve
34. TV-BD-200E* Steam Generator Blowdown Valve
35. TV-BD-200F* Steam Generator Blowdown Valve
36. TV-DA-200A Containment Sump Pump Isolation
37. TV-DA-200B Containment Sump Pump Isolation
38. TV-MS-209* Main Steam Drain Trip Valve
39. TV-MS-210* Main Steam Drain Trip Valve
40. TV-LM-200A Containment Isolation Monitoring
41. TV-LM-200B Containment Isolation Monitoring.
42. TV-LM-200C Containment Isolation Monitoring

e e TS 3.8-15 TABLE 3.8-2**

UNIT NO. 2 CONTAINMENT ISOLATION VALVES (Continued)

VALVE NUMBER FUNCTION

43. TV-LM-200D Containment Isolation Monitoring
44. TV-LM-200E Containment Isolation Monitoring
45. TV..:LM-200F Containment Isolation Monitoring
46. .TV-LM-200G Containment Isolation Monitoring
47. TV-LM-200H Containment Isolation Monitoring
48. TV-CV-250A Containment Vacuum Suction Valve
49. TV-CV-250B Containment Vacuum Suction Valve so. TV-LM-201A Leakage Monitoring Sealed Reference
51. TV-LM-20IB Leakage Monitoring Sealed Reference
52. TV-CV-250C Containment Vacuum Suction Valve
53. TV-CV-250D Containment Vacuum Suction Valve
54. TV-SV-202A Conde~ser Air Ejector Vent Trip Valve B. PHASE II CONTAINMENT ISOLATION (HI CLS SIGNAL)
1. TV-RM-200A Containment Air & Particulate Rad. Mon. TV's
2. TV-RM-200B Containment Air & Particulate Rad. Mon. TV's
3. *TV-RM-200C Containment Air & Particulate Rad. Mon. TV's I
4. TV-IA-201A. Containment Instr. Air Compressor Suction
s. TV-IA-20IB Containment Instr. Air Compressor Suction

-- TABLE 3.8-2**

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

VALVE NUMBER FUNCTION C. 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~ Vlv.
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 D. 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
6. MOV-VS-202 Contain. Vacuum Breaker Atmos. Supply MOV

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

'. --~ ....... - . . *-* .... *... =*~.: . . l,,* ** -.  :*cc.*-.-* ** * -.",.,.

  • e e 1* ...

TS 3.8-17 I

I .

I

  • i I

I I .*

TABLE 3.8-2**

UNIT NO. 2 CONTAINMENT ISOLATION VALVES (Continued)

I '

'.1 VALVE NUMBER FUNCTION E. REMOTE MANUAL VALVES

1. MOV-CS-201A Containment Spray Discharge Valve
2. MOV-CS-201B Containment Spray Discharge Valve
3. MOV-CS-201C Containment Spray Discharge Valve
4. MOV-CS-201D Containment Spray Discharge Valve
5. MOV-RS-255A Outside Recirculation Spray Suction Valve
6. MOV-RS-255B Outside Recirc. Spray Suction Valve
7. MOV-RS-256A Outside Recirc. Discharge Valve
8. MOV-RS-256B Outside Recirc. Discharge Valve
9. MOV-2842 Bypasses Boron Injec. Tank to Cold Leg Injec.
10. MOV-RH-200 Resi. Heat Remov. to RWST
11. FCV-2160 Loop Fill Header Flow Valve
12. MOV-2890A Lo Header S.I. Pump Disch. from Hot Leg
13. MOV-2890B Lo Header S.I. Pump Disch. from Hot Leg
14. MOV-2890C Lo Header S.I. Pump Disch. from Cold Leg
15. MOV-2869A !so. from Hot Leg to Hi Header S. I. Line A
16. MOV..;2869B !so. from Hot Leg to Hi Header S. I. Line B
17. MOV-2860A Iso. from Sump to Lo Header S. I.
18. MOV-2860B !so. Valve from Sump to Lo Header S. I.
19. MOV-SW-204A* SW to "A" HX's
20. MOV-SW-204B* SW to "B" HX's
21. MOV-SW-204C* SW to "C" HX's

e e TS 3.8-18 TABLE 3.8-2**

UNIT NO. 2 CONTAINMENT ISOLATION VALVES (Continued)

VALVE NUMBER FUNCTION

22. MOV-SW-204D* SW to IID" HX' s
23. MOV-SW-205A* SW from "A" HX's
24. MOV-SW-205B* SW from "B" HX's
25. MOV-SW-205C* SW from "C" HX's
26. MOV-SW-205D* SW from "D" HX's
27. HCV-CV-200 Cont. Vacuum Isolation F. MANUAL VALVES
1. 2-SI-150 Boron Injection Tank 1" line
2. 2-SI-3Z Accumulator Fill Valve
3. 2-GW-182 Discharge from Hydrogen Analyzer
4. 2-GW.l.183 Discharge from Hydrogen Analyzer
5. 2-SA-60 Service Air
6. 2-SA-62 Service Air
7. 2-IA-446 Instrument Air to Containment
8. 2-VA-1 Outside Isolation from Primary Vent Pot
9. 2-VA-6 Inside Isolation from Primary Vent Pot
10. 2-IA-446 Cross Tie from #1 Instrument Air Header
11. 2-GW-175 Suction from Cont. to H2 Analyzer
12. 2-GW-166 Suction from Cont. to H2 Analyzer
13. 2-GW-174 Inlet to Cont. from H Analyzer Outside Cont.

2

14. 2-FP-151 Outside Iso. Vlv for Cont. Fire Protection
15. 2-FP-152 Outside Iso. Vlv for Cont. Fire Protection

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

- ~ -'*"** ~ ..

e e TS 3.8-19 TABLE 3.8-2**

UNIT NO. 2 CONTAINMENT ISOLATION VALVES (Continued)

VALVE NUMBER FUNCTION

16. 2-RL-3 Inlet Vlv to Cavity from RCS Outside Cont .

' 17. 2-RL-5 Inlet Vlv to Cavity from RCS*Inside Cont.

18. 2-RL-13 Suction Vlv to 2-RL-P-lA Inside Containment
19. 2-RL-15 Suction Vlv to 2-RL-P-lA Outside Containment
20. 2-SI-73 Accumulator N Fill Vlv Outside Containment 2
21. 2-SI-174 Bypasses MOV-1869A
22. 2-SW-208 RS HX SW Drain
23. 2-SW-106 RS HX SW Drain
24. 2-cv-2 Cont. Vacuum Isolation G. CONTAINMENT CHECK VALVES
1. 2-FP-153 Inside Cont. - Fire Protection Header
2. 2-VP-12 Inside Cont. - Air Eject Disch to Cont.
3. 2-RS-17 Inside Cont. - RS Disch to Cont. A
4. 2-RS-11 Inside Cont. - RS Disch to Cont. B
5. 2-CS-13 Inside Cont. - Discharge of 2-CS-P-lA
6. 2-CS-24 Inside Cont. - Discharge of 2-CS-P-lB
7. 2-IA-938 Inside Cont. - Disch of Cont. IA Component
8. 2-IA-446 . Manual Valve - Disch. of IA Component Unit #2
9. 2-SI-234 Check Inside Cont. - N2 to Accumulator
10. 2-IA-939 Check Inside Cont. - Disch. of Cont. IA Component Unit #2
11. 2-IA-446 Manual Vlv - Disch. of Unit 2 Instr. Air Comp.

... ; :.: .. ,:: ~:*,**.:"..'. .....

e e TS 3.8-20 TABLE 3.8-2**

UNIT NO. 2 CONTAINMENT ISOLATION VALVES (Continued)

VALVE NUMBER FUNCTION

  • , 12. 2-RC-160 Check Valve Inside Contain. from PG Supply

.I

  • 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.

-~~

e TS 4.1-1 4.1 OPERATIONAL SAFETY REVIEW Applicability

  • i l

Applies to items directly related to safety limits and limiting conditions for operation.

1 ***

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.

2.

e Each Pressurizer PORV block valve shall be demonstrated TS 4.1-la 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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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.

TABLE 4.1-1 MINIMUM FREQUENCIES FOR CHECK, CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS CHANNEL DESCRIPTION CHECK CALIBRATE TEST REMARKS

1. Nuclear Power Range s D (1) BW(2) 1) Against a.heat balance standard M(3) Q (3) 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.
2. Nuclear Intermediate Range *S(l) N.A. P(2) 1) Once/shift when in service
2) Log level; bistable action (permissive, rod stop, trip)
3. Nuclear Source Range *S(l) N.A. P(2) 1) Once/Shift when in service
2) Bistable action (alarm, trip)
4. Reactor Coolant Temperature *S R BW(l) 1) Overtemperature - AT BW(2) 2) Overpower - AT
5. Reactor Coolant Flow s R M
6. Pressurizer Water Level s s

R M e

7. Pressurizer Pressure (High & R M Low)
8. 4 Kv Voltage and Frequency s R M Reactor protection circuit only

--l 9 *I Analog Rod Position *S(l ,2) R M(3) 1) With step counters (/)

(4) 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/

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TABLE 4.1-1 (Continued) t l

Ii fl

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CHANNEL DESCRIPTION CHECK CALIBRATE TEST REMARKS

10. Rod Position Bank Counters S(l,2) N.A. 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. 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
20. Boric Acid Control N.A. ]l N.A. e '
21. Containment Sump Level N.A. R N.A.
22. Accumulator Level and Pressure s R N.A.

-I

23. Containment Pressure-Vacuum Pump s R N.A. (/)

System .__.

.i:,,.

I

24. Steam Line Pressure s R M

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

l TABLE 4.1-1 'I 11 Ir'.

CHANNEL \!,:

DESCRIPTION CHECK CALIBRATE TEST REMARKS ,.

25. Turbine First Stage Pressure s R M
26. Emergency Plan Radiation Instr. *M R M I I
27. Environmental Radiation Monitors *M N~A. N.A. TLD Dosimeters .:
28. Logic Channel Testing N.A. N.A. M
29. Turbine Overspeed Protection N.A. R R Trip Channel (Electrical) 3Q. Turbine Trip Setpoint N.A. R R Stop valve closure or low EH fluid pressure e
31. Seismic Instrumentation M SA M
32. Reactor Trip Breaker N.A. N.A. M
33. Reactor Coolant Pressure (Low) N.,A. R N.A.
34. Auxiliary Feedwater
a. Steam Generator Water s R M Level Low-Low
b. RCP Undervoltage s R M c.

d.

e.

S.I.

Station Blackout Main Feedwater Pump Trip (All Safety Injection surveillance requirements)

N.A.

N.A.

R N.A.

N.A.

R S - Each shift M - Monthly

--i D - Daily P - Prior to each startup if not done previous week (./)

W - Weekly R - Each Refueling Shutdown ._.

NA - Not applicable BW - Every two week~ I SA - Semiannually AP - After each startup if not done previous week 00 Q - Every 90 effective full power days

  • See Specification 4.lD

TABLE 4.1-2 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL

  • CHANNEL CHECK INSTRUMENT CALIBRATION .
1. Auxiliary Feedwater Flow Rate p R
2. Reactor Coolant System Subcooling Margin Monitor M R
3. PORV Position Indicator (Primary Detector) M R
4. PORV Position Indicator (Backup Detector) M R
s. PORV Block Valve Position Indicator M R
6. Safety Valve Position Indicator M. R
7. Safety Valve Position Indicator (Backup Detector) M R

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\ i I i TABLE 4.1-2A , i 1 i'I MINIMUM FREQUENCY FOR EQUIPMENT TESTS II /l FSAR SECTION .i DESCRIPTION TEST "FREQUENCY REFERENCE Ii i/1*.

1. Control Rod Assemblies Rod drop times of all fuli length Each refueling shutdown .or after 7 rods at hot and cold conditions disassembly or maintenance re- 1:

1*.'

quiring the breech of the Reactor I*

Coolant System integrity /

2. Control Room Assemblies Partial movement of all rods Every 2 weeks 7 Ii
3. Refueling Water Chemical Functional Each refueling shutdown 6 .:

4.

5.

6.

Addition Tank Pressurizer Safety Valves Main Steam Safety Valves Containment Isolation Trip Setpoint Setpoint

  • Functional Each refueling shutdown Each refueling shutdown Each refueling shutdown 4

10 5

I I

7 .. Refueling System Interlocks *Functional Prior to refueling 9.12

8. Service Water System *Functional Each refueling shutdown 9.9 9, Fire Protection Pump and Functional Monthly 9.10 Power Supply
10. Primary System Leakage *Evaluate Daily 4
11. Diesel Fuel Supply *Fuel Inventory .5 days/week 8.5
12. Boric Acid Piping Heat *Operational Monthly 9.1 9:

Tracing Circuits

13. Main Steam Line Trip Functional 10 (1) Full closure (1) Each cold shutdown (2) Partial closure (2) Before each startup

--i

(/)

.~-

I I.D C"

TABLE 4.l-2A (CONTINUED)

. MINIMUM FREQUENCY FOR EQUIPMENT TESTS FSAR SECTION DESCRIPTION TEST FREQUENCY REFERENCE

14. Service Water System Valves Functional Each refueling 9.9 in Line Supplying Recircu-lation Spray Heat Exchangers i' I
15. Control Room Ventilation *Ability to maintain positive pres- Each refueling interval 9.13 System sure for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> using a volume of (approx. every 12-18 months) air equivalent to or less than stored in the bottled air supply
16. Reactor Vessel Overpressure Functional & Setpoint Prior to decreasing RCS None Mitigating System (except temperature below 3S0°F backup air supply) and monthly while the RCS is <350°F and the Reactor Vessel Head is bolted
17. Reactor Vessel Overpressure Setpoint Refueling None.

Mitigating System Backup*

Air Supply

-I

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

i*

TABLE 4.1-2A (CONTINUED)

MINIMUM FREQUENCY FOR EQUIPMENT TESTS FSAR SECTION DESCRIPTION TEST FREQUENCY 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 condition for refueling, after each time the. .

plant is placed in cold shutdown condition '11111111111' for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 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.

(a)

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.

(b)

Minimum differential test pressure shall not be below 150 psid.

  • See Specification 4.1.D.

e

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I I.O a.

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