Text

Instrumentation Technical Specifications Proposed Change No. 259, Proposed Technical Specifications - References
	ML031200259
Person / Time
Site:	Vermont Yankee
Issue date:	04/25/2003
From:	Balduzzi M; Entergy Nuclear Vermont Yankee, Entergy Operations
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	BVY 03-40
	Download: ML031200259 (179)
	v • d • e

Proposed Technical Specifications 3.2.B and 4.2.B Primary Containment Isolation

VYNPS 3.2 LIMITING CONDITIONS FOR 4.2 SURVEILLANCE REQUIREMENTS OPERATION 3.2 PROTECTIVE INSTRUMENT SYSTEMS 4.2 PROTECTIVE INSTRUMENT SYSTEMS B. Primary Containment Isolation B. Primary Containment Isolation The primary containment 1. The primary containment isolation instrumentation for isolation instrumentation each Trip Function in Table shall be checked, 3.2.2 shall be operable in functionally tested and accordance with Table 3.2.2. calibrated as indicated in Table 4.2.2.

When a primary containment isolation channel, and/or the affected primary containment isolation valve, is placed in an inoperable status solely for performance of required instrumentation surveillances, entry into associated Limiting Conditions for Operation and required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Trip Function maintains isolation capability.

2. Perform a Logic System Functional Test of Primary Containment isolation instrumentation Trip Functions once every Operating Cycle.

Amendment No. 43

VYNPS Table 3.2.2 (page 1 of 3)

Primary Containment Isolation Instrumentation ACTIONS WHEN ACTIONS APPLICABLE MODES REQUIRED REQUIRED REFERENCED OR OTHER CHANNELS CHANNELS FROM SPECIFIED PER TRIP ARE ACTION TRIP TRIP FUNCTION CONDITIONS SYSTEM INOPERABLE NOTE 1 SETTING

1. Main Steam Line Isolation a.Low-Low Reactor RUN, STARTUP/HOT 2 Note 1 Note 2.a 2 82.5 Vessel Water STANDBY, HOT inches 1

Level SHUTDOWN, REFUEL (

b.High Main Steam RUN, STARTUP/HOT 8 Note 1 Note 2.a S 196 OF for Line Area STANDBY, HOT channels 13 monitoring Temperature SHUTDOWN, REFUEL4 outside steam tunnel and 5 200 OF for channels monitoring inside steam tunnel c.High Main Steam RUN, STARTUP/HOT 2 per Note 1 Note 2.a S 140% of STANDBY, HOT main rated flow Line Flow 1 SHUTDOWN, REPUEL' ) steam line d.Low Main Steam RUN 2 Note 1 Note 2.c 2 800 psig Line Pressure e.High Main Steam STARTUP/HOT 2 Note 1 Note 2.a S 40% of STANDBY, HOT rated flow Line Flow - Not 41 3 in RUN SHUTDOWN, REFUEL f.Condenser Low RUN, STARTUP/HOT 2 Note 1 Note 2.a 5 12 2 inches Hg Vacuum STANDBY' ), HOT SHUTDOWN , absolute 41 REFUEL ad 2)

2. Primary Containment Isolation a.Low Reactor RUN, STARTUP/HOT 2 Note 1 Note 2.b 2 127.0 Vessel Water STANDBY, HOT inches 4

Level SHUTDOWN, REFUEL 1) b.High Drywell RUN, STARTUP/HOT 2 Note 1 Note 2.b S 2.5 psig Pressure STANDBY, HOT 41 3 SHUTDOWN, REFUEL (1) With reactor coolant temperature > 212 OF.

(2) With any turbine stop valve or turbine bypass valve not closed.

44 Amendment No.

VYNPS Table 3.2.2 (page 2 of 3)

Primary Containment Isolation Instrumentation ACTIONS WHEN ACTIONS APPLICABLE MODES REQUIRED REQUIRED REFERENCED OR OTHER CHANNELS CHANNELS FROM SPECIFIED PER TRIP ARE ACTION TRIP TRIP FUNCTION CONDITIONS SYSTEM INOPERABLE NOTE 1 SETTING

3. High Pressure Coolant Injection (HPCI) System Isolation a.High Steam Line RUN, STARTUP/HOT 6 Note 1 Note 2.d S 196 OF Space STANDBY, HOT Temperature SHUTDOWN, REFUELM b.High Steam Line RUN, STARTUP/HOT 1 Note 1 Note 2.d 5 195 d/p (Steam Line STANDBY, HOT inches of 1

Break) SHUTDOWN, REFUEL ( water c.Low Steam RUN, STARTUP/HOT 4 Note 1 Note 2.d 2 70 psig 13 Supply Pressure STANDBY ), HOT SHUTDOWN 3 ),

REFUEL(3) d.High Main Steam RUN, STARTUP/HOT 2 Note 1 Note 2.d 5 200 °F Line Tunnel STANDBY, HOT 1

Temperature SHUTDOWN, REFUEL 2 e.High Main Steam RUN, STARTUP/HOT 1 Note 1 Note 2.d 5 35 Line Tunnel STANDBY, HOT minutes 1

Temperature Time SHUTDOWN, REFUEL' )

Delay

4. Reactor Core Isolation Cooling (RCIC) System Isolation a.High Main Steam RUN, STARTUP/HOT 2 Note 1 Note 2.d 5 200 °F Line Tunnel STANDBY, HOT Temperature SHUTDOWN, REPUEL 1) b.High Main Steam RUN, STARTUP/HOT 1 Note 1 Note 2.d !5 35 Line Tunnel STANDBY, HOT minutes Temperature Time SHUTDOWN, REFUEL 1)

Delay c.High Steam Line RUN, STARTUP/HOT 6 Note 1 Note 2.d 5 196 °F Space STANDBY, HOT Temperature SHUTDOWN, REFUEL'1)

(1) With reactor coolant temperature > 212 °F.

(3) With reactor steam pressure > 150 psig.

Amendment No. 45

VYNPS Table 3.2.2 (page 3 of 3)

Primary Containment Isolation Instrumentation ACTIONS WHEN ACTIONS APPLICABLE MODES REQUIRED REQUIRED REFERENCED OR OTHER CHANNELS CHANNELS FROM SPECIFIED PER TRIP ARE ACTION TRIP TRIP FUNCTION CONDITIONS SYSTEM INOPERABLE NOTE 1 SETTING

4. RCIC System Isolation (Continued) d.High Steam Line RUN, STARTUP/HOT 1 Note 1 Note 2.d S 195 d/p (Steam Line STANDBY, HOT inches of 1 water Break) SHUTDOWN, REFUELM e.High Steam Line RUN, STARTUP/HOT 1 Note 1 Note 2.d 2 3 seconds d/p Time Delay STANDBY, HOT and SHUTDOWN, REFUEL(1) 5 7 seconds f.Low Steam RUN, STARTUP/HOT 4 Note 1 Note 2.d 2 50 psig Supply Pressure STANDBY{"M, HOT SHUTDOWN 43 ),

REFUELM3 }

5. Residual Heat Removal Shutdown Cooling Isolation a.High Reactor RUN, STARTUP/HOT 1 Note 1 Note 2.d 5 150 psig Pressure STANDBY, HOT SHUTDOWN' REFUEL 13 (1) With reactor coolant temperature > 212 "F.

(3) With reactor steam pressure > 150 psig.

Amendment No. 46

VYNPS Table 3.2.2 ACTION Notes

1. With one or more required Primary Containment Isolation Instrumentation channels inoperable, take all of the applicable Actions in Notes l.a and 1.b below.

a. With one or more Trip Functions with one or more required channels inoperable:

1) For Trip Functions 2.a and 2.b, place inoperable channel in trip within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; and

2) For Trip Functions 3.e, 4.b, and 4.e, restore inoperable channel to operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; and

3) For all other Trip Functions, place inoperable channel in trip within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. With one or more Trip Functions with isolation capability not maintained:

1) Restore isolation capability within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

Penetration flow paths, isolated as a result of complying with the above Actions, may be unisolated intermittently under administrative controls.

If any applicable and associated completion time of Note l.a or l.b is not met, take the applicable Actions of Note 2 below and referenced in Table 3.2.2 for the channel.

2. a. Isolate the associated Main Steam Line within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (penetration flow paths may be unisolated intermittently under administrative control); or Place the reactor in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and place the reactor in COLD SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. Place the reactor in COLD SHUTDOWN within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c. Place the reactor in STARTUP/HOT STANDBY within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

d. Isolate the affected penetration flow path within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (penetration flow paths may be unisolated intermittently under administrative control).

Amendment No. 47

VYNPS Table 4.2.2 (page 1 of 2)

Primary Containment Isolation Instrumentation Tests and Frequencies TRIP FUNCTION CHECK FUNCTIONAL TEST CALIBRATION

1. Main Steam Line Isolation

a. Low-Low Reactor Once/Day Every 3 Months Every 3 Months<1 ,

Vessel Water Level Once/Operating Cycle

b. High Main Steam Line NA Every 3 Months Each Refueling Outage Area Temperature

c. High Main Steam Line Once/Day Every 3 Months Every 3 Monthse13 ,

Flow once/Operating Cycle

d. Low Main Steam Line NA Every 3 Months Every 3 Months Pressure

e. High Main Steam Line Once/Day Every 3 Months Every 3 Months~1 ),

Flow - Not in RUN Once/Operating Cycle

f. Condenser Low Vacuum NA Every 3 Months Every 3 Months Months

2. Primary Containment Isolation

a. Low Reactor Vessel NA Every 3 Months Every 3 Months^1 ),

Water Level Once/Operating Cycle

b. High Drywell Once/Day Every 3 Months Every 3 Months<'>,

Pressure once/Operating Cycle

3. High Pressure Coolant Injection (HPCI) System Isolation

a. High Steam Line NA Every 3 Months Each Refueling Outage Space Temperature

b. High Steam Line d/p NA Every 3 Months Months Every 3 Months (Steam Line Break)

c. Low Steam Supply NA Every 3 Months Every 3 Months Pressure

d. High Main Steam Line NA Every 33 Months Each Refueling Outage Tunnel Temperature

e. High Main Steam Line NA NA Once/Operating Cycle Tunnel Temperature Time Delay (1) Trip unit calibration only.

48 Amendment No.

VYNPS Table 4.2.2 (page 2 of 2)

Primary Containment Isolation Instrumentation Tests and Frequencies TRIP FUNCTION CHECK FUNCTIONAL TEST CALIBRATION

4. Reactor Core Isolation Cooling (RCIC) System Isolation

a. High Main Steam Line NA Every 3 Months Each Refueling Outage Tunnel Temperature

b. High Main Steam Line NA NA Once/Operating Cycle Tunnel Temperature Time Delay

c. High Steam Line NA Every 3 Months Each Refueling Outage Space Temperature

d. High Steam Line d/p NA Every 3 Months Every 3 Months (Steam Line Break)

e. High Steam Line d/p NA Every 3 Months Every 3 Months (Steam Line Break)

Time Delay

f. Low Steam Supply NA Every 3 Months Every 3 Months Pressure

5. Residual Heat Removal Shutdown Cooling Isolation

a. High Reactor NA Every 3 Months Every 3 Months Pressure Amendment No. 49

Proposed Bases 3.2.3 and 4.2.B Primary Containment Isolation

VYNPS BASES: 3.2.B/4.2.B PRIMARY CONTAINMENT ISOLATION BACKGROUND The primary containment isolation instrumentation automatically initiates closure of appropriate primary containment isolation valves (PCIVs). The function of the PCIVs, in combination with other accident mitigation systems, is to limit fission product release during and following postulated Design Basis Accidents (DBAs). Primary containment isolation within the time limits specified for those isolation valves designed to close automatically ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a DRA.

The isolation instrumentation includes the sensors, relays, and switches that are necessary to cause initiation of primary containment and reactor coolant pressure boundary (RCPB) isolation. Most channels include electronic equipment (e.g., trip units) that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs a primary containment isolation signal to the isolation logic. Functional diversity is provided by monitoring a wide range of independent parameters. The input parameters to the isolation logics are (a) reactor vessel water level, (b) area ambient temperatures, (c) main steam line (MSL) flow, (d) main steam line pressure, (e) condenser vacuum, (f) drywell pressure, (g) high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) steam line d/p, (h) HPCI and RCIC steam line pressure, and (i) reactor vessel pressure. Redundant sensor input signals from each parameter are provided for initiation of isolation.

Primary containment isolation instrumentation has inputs to the trip logic of the isolation functions listed below.

1. Main Steam Line Isolation The Low - Low Reactor Vessel Water Level, Low Main Steam Line Pressure, High Main Steam Line Flow - Not in RUN, and Condenser Low Vacuum Trip Functions each receive inputs from four channels. The outputs of these channels are combined in a one-out-of-two taken twice logic to initiate isolation of all main steam isolation valves (MSIVs), MSL drain valves, and recirculation loop sample isolation valves.

The High Main Steam Line Flow Trip Function uses 16 flow channels, four for each steam line. One channel from each steam line inputs to one of the four trip strings. Two trip strings make up each trip system and both trip systems must trip to cause an isolation of all MSIVs, MSL drain valves, and recirculation sample isolation valves. Each trip string has four inputs (one per MSL), any one of which will trip the trip string. The trip strings are arranged in a one-out-of-two taken twice logic. This is effectively a one-out-of-eight taken twice logic arrangement to initiate isolation.

The High Main Steam Line Area Temperature Trip Function receives input from 16 channels, four for each of four main steam line areas. The logic is arranged similar to the High Main Steam Line Flow Trip Function. One channel from each steam tunnel area inputs to one of four trip strings. Two trip strings make up a trip system and both trip systems must trip to cause isolation.

MSL Isolation Trip Functions isolate the Group 1 valves.

Amendment No. 76

VYNPS BASES: 3.2.B/4.2.B PRIMARY CONTAINMENT ISOLATION BACKGROUND (continued)

2. Primary Containment Isolation The Low Reactor Vessel Water Level and High Drywell Pressure Trip Functions each receive inputs from four channels. For each Trip Function, the outputs of these channels are combined in a one-out-of-two taken twice logic to initiate isolation of the PCIVs identified in Reference 1.

Primary Containment Isolation Trip Functions isolate the Groups 2, 3, and 4 valves. Group S valves are also isolated by the Low Reactor Vessel Water Level Trip Function.

3, 4. High Pressure Coolant Injection System Isolation and Reactor Core Isolation Cooling System Isolation The HPCI High Steam Line d/p, RCIC High Steam Line d/p, and RCIC High Steam Line d/p Time Delay Trip Functions each receive input from two channels, with each channel in one trip system using a one-out-of-one logic. The trip systems are arranged in a one-out-of-two logic. Each of the two trip systems is connected to both valves on the associated penetration.

The HPCI and RCIC Low Steam Supply Line Pressure Trip Functions each receive input from four steam supply pressure channels. The outputs from the associated steam supply pressure channels are connected in a one-out-of-two-twice trip system logic arrangement. There are two trip system logics which provide input to one trip system. The trip system must trip to initiate isolation of both valves on the associated penetration.

The HPCI and RCIC High Main Steam Line Tunnel Temperature Trip Functions each receive input from 4 channels. Four channels, each with an associated temperature switch, are connected in a one-out-of-two-twice arrangement which provides input to two trip systems. Both trip systems must trip to initiate isolation of both valves on the associated penetration. In addition, the HPCI and RCIC High Main Steam Line Tunnel Temperature Trip Functions each have time delays. These Time Delay Trip Functions each receive input from two channels, with each channel in one of the trip system using a one-out-of-one logic. The trip systems are arranged in a one-out-of-two logic.

The HPCI and RCIC High Steam Line Space Temperature Trip Functions each receive input from 12 channels. There are three steam line areas each monitored by one set of four channels. One channel from each of the three steam line areas inputs to one of the four trip strings. Two trip strings make up each trip system and both trip systems must trip to cause an isolation of both valves on the associated penetration. The trip strings are arranged in a one-out-of-two taken twice logic. This is effectively a one-out-of-six taken twice logic arrangement to initiate isolation.

HPCI System and RCIC System Isolation Trip Functions isolate the Group 6 valves, as appropriate.

Amendment No. 76a

VYNPS BASES: 3.2.B/4.2.B PRIMARY CONTAINMENT ISOLATION BACKGROUND (continued)

5. Residual Heat Removal Shutdown Cooling Isolation The High Reactor Pressure Trip Function receives input from two channels. The outputs from these channels are arranged in a one-out-of-two logic to initiate isolation of the Shutdown Cooling (SDC) supply isolation valves.

The Residual Heat Removal Shutdown Cooling Isolation Trip Function isolates the Group 4 SDC supply isolation valves.

APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY The isolation signals generated by the primary containment isolation instrumentation are implicitly assumed in the safety analyses of Reference 2 to initiate closure of valves to limit offsite doses.

Primary containment isolation instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii). Certain instrumentation Trip Functions are retained for other reasons and are described below in the individual Trip Functions discussion.

The operability of the primary containment isolation instrumentation is dependent on the operability of the individual instrumentation channel Trip Functions specified in Table 3.2.2. Each Trip Function must have the required number of operable channels in each trip system, with their trip setpoints within the calculational as-found tolerances specified in plant procedures.

Operation with actual trip setpoints within calculational as-found tolerances provides reasonable assurance that, under worst case design basis conditions, the associated trip will occur within the Trip Settings specified in Table 3.2.2. As a result, a channel is considered inoperable if its actual trip setpoint is not within the calculational as-found tolerances specified in plant procedures. The actual trip setpoint is calibrated consistent with applicable setpoint methodology assumptions.

Certain Emergency Core Cooling Systems (ECCS) valves (e.g., containment spray isolation valves) also serve the dual function of automatic PCIVs. The signals that isolate these valves are also associated with the automatic initiation of the ECCS. Some instrumentation requirements and Actions associated with these signals are addressed in Specification 3.2.A, Emergency Core Cooling Systems (ECCS)," and are not included in this specification.

In general, the individual Trip Functions are required to be operable in RUN, STARTUP/HOT STANDBY, HOT SHUTDOWN, and REFUEL (with reactor coolant temperature

> 212 F) consistent with the Applicability for Primary Containment 0 Integrity requirements in Specification 3.7.A.2. Trip Functions that have different Applicabilities are discussed below in the individual Trip Functions discussion.

The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Trip Function by Trip Function basis.

Main Steam Line Isolation l.a. Low - Low Reactor Vessel Water Level Low reactor pressure vessel (RPV) water level indicates that the capability to Amendment No. 76b

VYNPS BASES: 3.2.B/4.2.B PRIMARY CONTAINMENT ISOLATION APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, isolation of the MSIVs and other interfaces with the reactor vessel occurs to prevent offsite dose limits from being exceeded. The Low - Low Reactor Vessel Water Level Trip Function is one of the many Trip Functions assumed to be operable and capable of providing isolation signals. The Low - Low Reactor Vessel Water Level Trip Function associated with isolation is assumed in the analysis of the recirculation line break (Ref. 3). The isolation of the MSLs supports actions to ensure that offsite dose limits are not exceeded for a DBA.

Reactor vessel water level signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Low - Low Reactor Vessel Water Level Trip Function are available and are required to be operable to ensure that no single instrument failure can preclude the isolation function.

The Low - Low Reactor Vessel Water Level Trip Setting is chosen to be the same as the ECCS Low - Low Reactor Vessel Water Level Trip Setting (Specification 3.2.A) to ensure that the MSLs isolate on a potential loss of coolant accident (LOCA) to prevent offsite doses from exceeding 10 CFR 100 limits. The Trip Setting is referenced from the top of enriched fuel.

This Function isolates the Group 1 valves.

1.b. High Main Steam Line Area Temperature Main steam line tunnel temperature is provided to detect a leak in the RCPB in the steam tunnel and provides diversity to the high flow instrumentation.

Temperature is sensed in four different areas of the steam tunnel in the vicinity of the main steam lines. The isolation occurs when a very small leak has occurred in any one of the four areas. If the small leak is allowed to continue without isolation, offsite dose limits may be reached. However, credit for these instruments is not taken in any transient or accident analysis in the UFSAR, since bounding analyses are performed for large breaks, such as MSLBs.

Main steam line area temperature signals are initiated from a total of sixteen temperature switches located in the four areas being monitored.

Sixteen channels of High Main Steam Line Area Temperature Trip Function are available and are required to be operable to ensure that no single instrument failure can preclude the isolation function.

The High Main Steam Line Area Temperature Trip Setting is chosen to provide early indication of a steam line break.

These Functions isolate the Group 1 valves.

1.c. High Main Steam Line Flow High Main Steam Line Flow is provided to detect a break of the MSL and to initiate closure of the MSIVs. If the steam were allowed to continue flowing out of the break, the reactor would depressurize and the core could uncover.

If the RPV water level decreases too far, fuel damage could occur. Therefore, Amendment No. 76c

VYNPS BASES: 3.2.B/4.2.B PRIMARY CONTAINMENT ISOLATION APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) the isolation is initiated on high flow to prevent or minimize core damage.

The High Main Steam Line Flow Trip Function is directly assumed in the analysis of the main steam line break (MSLB) (Ref. 4). The isolation action, along with the scram function of the Reactor Protection System (RPS), ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46 and offsite doses do not exceed the 10 CFR 100 limits.

The MSL flow signals are initiated from 16 differential pressure transmitters that are connected to the four MSLs (the differential pressure transmitters sense differential pressure across a flow restrictor). The differential pressure transmitters are arranged such that, even though physically separated from each other, all four connected to one MSL would be able to detect the high flow. Four channels of High Main Steam Line Flow Trip Function for each MSL (two channels per trip system) are available and are required to be operable so that no single instrument failure will preclude detecting a break in any individual MSL.

The Trip Setting is chosen to ensure that fuel peak cladding temperature and offsite dose limits are not exceeded due to the break.

This Trip Function isolates the Group 1 valves.

1.d. Low Main Steam Line Pressure Low MSL pressure indicates that there may be a problem with the turbine pressure regulation, which could result in a low reactor vessel water level condition and the RPV cooling down more than 100'F/hr if the pressure loss is allowed to continue. The Low Main Steam Line Pressure Trip Function is directly assumed in the analysis of the pressure regulator failure (Ref. 5).

For this event, the closure of the MSIVs ensures that the RPV temperature change limit (1000 F/hr) is not reached. In addition, this Trip Function supports actions to ensure that Safety Limit l.l.B is not exceeded. (This Trip Function closes the MSIVs prior to pressure decreasing below 785 psig, which results in a scram due to MSIV closure, thus reducing reactor power to

< 25% RATED THERMAL POWER.)

The MSL low pressure signals are initiated from four pressure switches that are connected to the MSL header. The switches are arranged such that, even though physically separated from each other, each pressure switch is able to detect low MSL pressure. Four channels of Low Main Steam Line Pressure Trip Function are available and are required to be operable to ensure that no single instrument failure can preclude the isolation function.

The Trip Setting was selected to be high enough to prevent excessive RPV depressurization.

The Low Main Steam Line Pressure Trip Function is only required to be operable in the RUN Mode since this is when the assumed transient can occur (Ref. 5).

This Trip Function isolates the Group 1 valves.

Amendment No. 76d

VYNPS BASES: 3.2.B/4.2.B PRIMARY CONTAINMENT ISOLATION APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) i.e. High Main Steam Line Flow - Not in RUN provides High Main Steam Line Flow when the reactor mode switch is not in RUN causes the protection for a turbine pressure regulator malfunction which open or protection for a turbine control valves and turbine bypass valves to main steam line break. These events would result in a rapid depressurization and cooldown of the RPV. The High Main Steam Line Flow - Not in RUN Trip Function was credited in the MSLB at low power analysis.

The MSL flow signals are initiated from 4 differential pressure transmitters, one connected to each of the four MSLs (the differential pressure switches sense differential pressure across a flow restrictor). Four channels of High Main Steam Line Flow - Not in RUN Trip Function (two channels per trip system) are available and are required to be operable so that no single instrument failure will preclude providing protection against a turbine pressure regulator malfunction or a break in any individual MSL.

break.

The Trip Setting is chosen to provide early indication of a steam line to The High Main Steam Line Flow - Not in RUN Trip Function is only required be operable in the STARTUP/HOT STANDBY, HOT SHUTDOWN, and REFUEL (with 0 for reactor coolant temperature > 212 F) Modes. In the RUN Mode, protection the depressurization resulting from a turbine pressure regulator malfunction is provided by the Low Main Steam Line Pressure Trip Function and protection by for depressurization resulting from a main steam line break is provided the High Main Steam Line Flow Trip Function.

This Trip Function isolates the Group 1 valves.

l.f. Low Condenser Vacuum The Low Condenser Vacuum Trip Function is provided to prevent main overpressurization of the main condenser in the event of a loss of the Since the integrity of the condenser is an assumption in condenser vacuum.

offsite dose calculations, the Low Condenser Vacuum Trip Function is assumed to be operable and capable of initiating closure of the MSIVs. The closure lead to of the MSIVs is initiated to prevent the addition of steam that would additional condenser pressurization and possible rupture, thereby preventing a potential radiation leakage path following an accident.

Condenser vacuum pressure signals are derived from four pressure switches that sense the pressure in the condenser. Four channels of Low Condenser Vacuum Trip Function are available and are required to be operable to ensure that no single instrument failure can preclude the isolation function.

The Trip Setting is chosen to prevent damage to the condenser due to As pressurization, thereby ensuring its integrity for offsite dose analysis.

the channels are not required to be indicated in Footnote (2) to Table 3.2.2, operable in the STARTUP/HOT STANDBY, HOT SHUTDOWN, and REFUEL (with reactor and coolant temperature > 212 F) Modes when all turbine stop valves (TSVs) 0 turbine bypass valves (TBVs) are closed, since the potential for condenser overpressurization is minimized. A key lock switch is provided to manually bypass the Low Condenser Vacuum Trip Function channels to enable plant 76e JmkUmLUF "RMeLLUC AL L'irn w.%

A

VYNPS BASES: 3.2.B/4.2.B PRIMARY CONTAINMENT ISOLATION APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

Hg startup and shutdown when condenser vacuum is greater than 12 inches absolute and all TSVs and TBVs are closed.

This Trip Function isolates the Group 1 valves Primary Containment Isolation 2.a. Low Reactor Vessel Water Level be Low RPV water level indicates that the capability to cool the fuel may threatened. The valves whose penetrations communicate with the primary The containment are isolated to limit the release of fission products.

containment on low RPV water level supports actions isolation of the primary The Low to ensure that offsite dose limits of 10 CFR 100 are not exceeded.

Reactor Vessel Water Level Trip Function associated with isolation is to implicitly assumed in the UFSAR analysis as these leakage paths are assumed be isolated post LOCA.

Low Reactor Vessel Water Level signals are initiated from level transmitters of that sense the difference between the pressure due to a constant column and the pressure due to the actual water level (variable water (reference leg)

Trip leg) in the vessel. Four channels of Low Reactor Vessel Water Level Function are available and are required to be operable to ensure that no single instrument failure can preclude the isolation function.

same as The Low Reactor Vessel Water Level Trip Setting was chosen to be the the RPS Low Reactor Vessel Water Level scram Trip Setting to (Specification 3.1.A), since isolation of these valves is not critical The Trip Setting is referenced from the top of orderly plant shutdown.

enriched fuel.

This Trip Function isolates the Groups 2, 3, 4, and 5 valves.

2.b. High Drywell Pressure High drywell pressure can indicate a break in the RCPB inside the primary containment. The isolation of some of the primary containment isolation dose valves on high drywell pressure supports actions to ensure that offsite limits of 10 CFR 100 are not exceeded. The High Drywell Pressure Trip Function, associated with isolation of the primary containment, is implicitly to assumed in the UFSAR accident analysis as these leakage paths are assumed be isolated post LOCA.

that High drywell pressure signals are initiated from pressure transmitters the pressure in the drywell. Four channels of High Drywell Pressure sense no Trip Function are available and are required to be operable to ensure that single instrument failure can preclude the isolation function.

Pressure The Trip Setting was selected to be the same as the ECCS High Drywell Trip (Specification 3.2.A) and RPS High Drywell Pressure (Specification 3.1.A)

Settings, since this may be indicative of a LOCA inside primary containment.

This Trip Function isolates the Groups 2, 3 and 4 valves.

76f Amendment No.

VYNPS BASES: 3.2.B/4.2.B PRIMARY CONTAINMENT ISOLATION APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

High Pressure Coolant Injection System Isolation and Reactor Core Isolation Cooling System Isolation 3.a, 4.c HPCI and RCIC High Steam Line Space Temperature High Steam Line Space Temperature Trip Functions are provided to detect a leak from the associated system steam piping. The isolation occurs when a very small leak has occurred and is diverse to the high flow instrumentation.

If the small leak is allowed to continue without isolation, offsite dose limits may be reached. These Trip Functions are not assumed in any UFSAR transient or accident analysis, since bounding analyses are performed for large breaks such as recirculation or MSL breaks. However, these instruments prevent the RCIC or HPCI steam line breaks from becoming bounding.

High Steam Line Space Temperature signals are initiated from temperature switches that are appropriately located to detect a leak from the system piping that is being monitored. For each Trip Function, there are four instruments that monitor each of three locations. Twelve channels for HPCI High Steam Line Space Temperature are available and are required to be operable to ensure that no single instrument failure can preclude the isolation function. Twelve channels for RCIC High Steam Line Space Temperature are available and are required to be operable to ensure that no single instrument failure can preclude the isolation function.

The Trip Settings are set high enough above anticipated normal operating levels to avoid spurious isolation, yet low enough to provide timely detection of a HPCI or RCIC steam line break.

These Trip Functions isolate the associated Group 6 valves.

3.b., 4.d. HPCI and RCIC High Steam Line d/p (Steam Line Break)

High Steam Line d/p (Steam Line Break) Trip Functions are provided to detect a break of the RCIC or HPCI steam lines and initiate closure of the steam line isolation valves of the appropriate system. If the steam is allowed to continue flowing out of the break, the reactor will depressurize and the core can uncover. Therefore, the isolations are initiated on high d/p to prevent or minimize core damage. The isolation action, along with the scram function of the RPS, ensures that the requirements of 10 CFR 50.46 are met. Specific credit for these Trip Functions is not assumed in any UFSAR accident analyses since the bounding analysis is performed for large breaks such as recirculation and MSL breaks. However, these instruments prevent the RCIC or HPCI steam line breaks from becoming bounding.

The HPCI and RCIC High Steam Line d/p (Steam Line Break) signals are initiated from differential pressure switches (two for HPCI and two for RCIC) that are connected to the associated system steam lines. Two channels of both HPCI and RCIC High Steam Line d/p (Steam Line Break) Trip Functions are available and are required to be operable to ensure that no single instrument failure can preclude the isolation function.

Amendment No. 76g

VYNPS BASES: 3.2.B/4.2.B PRIMARY CONTAINMENT ISOLATION APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

The Trip Settings are set high enough above anticipated normal operating levels to avoid spurious isolation, yet low enough to provide timely detection of a HPCI or RCIC steam line break.

These Trip Functions isolate the associated Group 6 valves.

3.c., 4.f. HPCI and RCIC Low Steam Supply Pressure Low steam supply pressure indicates that the pressure of the steam in the HPCI or RCIC turbine may be too low to continue operation of the associated system's turbine. These isolations are for equipment protection. However, they also provide a diverse signal to indicate a possible system break.

These instruments are included in Technical Specifications because of the potential for possible system initiation failure resulting from these instruments.

The HPCI and RCIC Low Steam Supply Pressure signals are initiated from pressure switches (four for HPCI and four for RCIC) that are connected to the associated system steam line. Four channels of both HPCI and RCIC Low Steam Supply Pressure Trip Functions are available and are required to be operable to ensure that no single instrument failure can preclude the isolation function.

The Trip Settings are selected to be below the pressure at which the system's turbine can effectively operate.

Since these Trip Functions are provided for equipment protection, they are only required to be operable when the HPCI and RCIC System are required to be operable. Therefore, as indicated in Footnote (3) to Table 3.2.2, in the STARTUP/HOT STANDBY, HOT SHUTDOWN, and REFUEL Modes, the channels are only required to be operable when reactor steam pressure is > 150 psig.

These Trip Functions isolate the associated Group 6 valves.

3.d., 3.e., 4.a., 4.b. HPCI and RCIC High Main Steam Line Tunnel Temperature and Time Delay HPCI and RCIC High Main Steam Line Tunnel Temperature Trip Functions are provided to detect a leak from the associated system steam piping. The isolation occurs when a very small leak has occurred and is diverse to the high flow instrumentation. If the small leak is allowed to continue without isolation, offsite dose limits may be reached. These Trip Functions are not assumed in any UFSAR transient or accident analysis., since bounding analyses are performed for large breaks such as recirculation or MSL breaks. However, these instruments prevent the RCIC or HPCI steam line breaks from becoming bounding.

HPCI and RCIC High Main Steam Line Tunnel Temperature signals are initiated from temperature switches that are appropriately located to detect a leak from the associated system piping that is being monitored. For each Trip Function, there are four instruments that monitor the area. Four channels for HPCI High Main Steam Line Tunnel Temperature are available and are NrJ WUImUALT. 76h "UCenULUCesnLA w.

VYNPS BASES: 3.2.B/4.2.B PRIMARY CONTAINMENT ISOLATION APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) required to be operable to ensure that no single instrument failure can preclude the isolation function. Four channels for RCIC High Main Steam Line Tunnel Temperature are available and are required to be operable to ensure that no single instrument failure can preclude the isolation function.

The Trip Settings are set high enough above anticipated normal operating levels to avoid spurious isolation, yet low enough to provide timely detection of a HPCI or RCIC steam line break.

These Trip Functions isolate the associated Group 6 valves.

4.e RCIC High Steam Line d/p Time Delay The RCIC High Steam Line d/p Time Delay is provided to prevent false isolations on RCIC High Steam Line d/p during system startup transients and therefore improves system reliability. This Trip Function is not assumed in any UFSAR transient or accident analyses.

The RCIC High Steam Line d/p Time Delay Trip Function delays the RCIC High Steam Line d/p (Steam Line Break) signal by use of time delay relays. When a RCIC High Steam Line d/p (Steam Line Break) signal is generated, the time delay relays delay the tripping of the associated RCIC isolation trip system for a short time. Two channels of RCIC High Steam Line d/p Time Delay Trip Function are available and are required to be operable to ensure that no single instrument failure can preclude the isolation function.

The Trip Setting is chosen to be long enough to prevent false isolations due to system starts but not so long as to impact compliance with 10CFR50.46 requirements.

This Trip Function, in conjunction with the RCIC High Steam Line d/p (Steam Line Break) Trip Function, isolates the RCIC System Group 6 valves.

Residual Heat Removal Shutdown Cooling Isolation 5.a. High Reactor Pressure The High Reactor Pressure Trip Function is provided to isolate the shutdown cooling portion of the Residual Heat Removal (RHR) System. This interlock is provided only for equipment protection to prevent an intersystem LOCA scenario, and credit for the interlock is not assumed in the accident or transient analysis in the UFSAR.

The High Reactor Pressure signals are initiated from two pressure switches.

Two channels of High Reactor Pressure Trip Function are available and are required to be operable to ensure that no single instrument failure can preclude the isolation function. The Trip Function is only required to be operable in the RUN, STARTUP/HOT STANDBY, HOT SHUTDOWN, and REFUEL (with 0

reactor coolant temperature > 212 F) Modes, since these are the only Modes in which the reactor can be pressurized; thus, equipment protection is needed.

The Trip Setting was chosen to be low enough to protect the system equipment from overpressurization.

This Trip Function isolates the Group 4 SDC supply isolation valves.

Amendment No. 76i

VYNPS BASES: 3.2.B/4.2.B PRIMARY CONTAINMENT ISOLATION ACTIONS Table 3.2.2 ACTION Note 1 Because of the diversity of sensors available to provide isolation signals and the redundancy of the isolation design, an allowable out of service time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , depending on the Trip Function (12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for those Trip Functions that have channel components common to RPS instrumentation, i.e., Trip Functions 2.a and 2.b, and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for those Trip Functions that do not have channel components common to RPS instrumentation, i.e., all other Trip Functions), has been shown to be acceptable (Refs. 6 and 7) to permit restoration of any inoperable channel to operable status. This out of service time is only acceptable provided the associated Trip Function is still maintaining isolation capability (refer to the next paragraph). For all Trip Functions except for Trip Functions 3.e, 4.b, and 4.e, if the inoperable channel cannot be restored to operable status within the allowable out of service time, the channel must be placed in the tripped condition per Table 3.2.2 ACTION Note 1.a.l) or l.a.3), as applicable. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue with no further restrictions. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an isolation), the applicable actions of Table 3.2.2 ACTION Note 2 must be taken. For Trip Functions 3.e, 4.b, and 4.e, Table 3.2.2 ACTION Note l.a.2) requires the channel to be restored to operable status. Table 3.2.2 ACTION Note l.a.2) does not allow placing the channel in trip since this action would not necessarily result in a safe state for the channel in all events.

Table 3.2.2 ACTION Note l.b is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Trip Function result in redundant automatic isolation capability being lost for the associated penetration flow path(s). The Trip Functions are considered to be maintaining isolation capability when sufficient channels are operable or in trip, such that both trip systems will generate a trip signal from the given Trip Function on a valid signal. For Trip Functions l.a, l.d, i.e, 1.f, 2.a, 2.b, 3.b, 3.d, 4.a, 4.d, and 5.a, this would require both trip systems to have one channel operable or in trip. For Trip Function l.c, this would require both trip systems to have one channel, associated with each MSL, operable or in trip. Trip Functions l.b, 3.a and 4.c, consist of channels that monitor several locations within a given area (e.g., different locations within the main steam tunnel area). Therefore, this would require both trip systems to have one channel per location operable or in trip. For Trip Functions 3.e, 4.b and 4.e, this would require both trip systems to have one channel operable. For Trip Functions 3.c and 4.f (which only have one trip system for each Trip Function), this would require one trip system to have one channel in each trip system logic operable or in trip.

The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.

Amendment No. 76j

VYNPS BASES: 3.2.B/4.2.B PRIMARY CONTAINMENT ISOLATION ACTIONS (continued)

Table 3.2.2 ACTION Note 1 also allows penetration flow path(s) to be unisolated intermittently under administrative controls. These administrative controls consist of stationing a dedicated operator at the controls of the valve, who is in continuous communication with the control room. In this way, the penetration can be rapidly isolated when a need for primary containment isolation is indicated.

Table 3.2.2 ACTION Notes 2.a, 2.b, 2.c and 2.d If any applicable Action and associated completion time of Table 3.2.2 ACTION Note l.a or 1.b are not met, the applicable Actions of Table 3.2.2 ACTION Note 2 and referenced in Table 3.2.2 (as identified for each Trip Function in the Table 3.2.2 "ACTIONS REFERENCED FROM ACTION NOTE 1" column) must be immediately entered and taken. The applicable Action specified in Table 3.2.2 is Trip Function and Mode or other specified condition dependent.

For Table 3.2.2 ACTION Note 2.a, if the channel is not restored to operable status or placed in trip within the allowed Completion Time the associated MSLs may be isolated, and, if allowed (i.e., plant safety analysis allows operation with an MSL isolated), operation with that MSL isolated may continue. Isolating the affected MSL accomplishes the safety function of the inoperable channel. This action will generally only be used if a Trip Function 1.c channel is inoperable and untripped. The associated MSL(s) to be isolated are those whose High Main Steam Line Flow Trip Function channel(s) are inoperable. Alternately, the plant must be placed in a Mode or other specified condition in which the LCO does not apply. This is done by placing the plant in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. Table 3.2.2 ACTION Note 2.a also allows penetration flow path(s) to be unisolated intermittently under administrative controls. These administrative controls consist of stationing a dedicated operator at the controls of the valve, who is in continuous communication with the control room. In this way, the penetration can be rapidly isolated when a need for primary containment isolation is indicated.

For Table 3.2.2 ACTION Note 2.b, if the channel is not restored to operable status or placed in trip within the allowed Completion Time, the plant must be placed in a Mode or other specified condition in which the LCO does not apply. This is done by placing the plant in COLD SHUTDOWN within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

For Table 3.2.2 ACTION Note 2.c, if the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, the plant must be placed in a Mode or other specified condition in which the LCO does not apply. This is done by placing the plant in at least STARTUP/HOT STANDBY within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The allowed Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is reasonable, based on operating experience, to reach STARTUP/HOT STANDBY from full power conditions in an orderly manner and without challenging plant systems.

Amendment No. 76k

VYNPS BASES: 3.2.B/4.2.B PRIMARY CONTAINMENT ISOLATION ACTIONS (continued)

For Table 3.2.2 ACTION Note 2.d, if the channel is not restored to operable status or placed in trip within the allowed Completion Time, plant operations may continue if the affected penetration flow path(s) is isolated. Isolating the affected penetration flow path(s) accomplishes the safety function of the inoperable channel. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time is acceptable because it minimizes risk while allowing sufficient time for plant operations personnel to isolate the affected penetration flow path(s). Table 3.2.2 ACTION Note 2.d also allows penetration flow path(s) to be unisolated intermittently under administrative controls. These administrative controls consist of stationing a dedicated operator at the controls of the valve, who is in continuous communication with the control room. In this way, the penetration can be rapidly isolated when a need for primary containment isolation is indicated.

SURVEILLANCE REQUIREMENTS Surveillance Requirement 4.2.B.1 As indicated in Surveillance Requirement 4.2.B.1, primary containment isolation instrumentation shall be checked, functionally tested and calibrated as indicated in Table 4.2.2. Table 4.2.2 identifies, for each Trip Function, the applicable Surveillance Requirements.

Surveillance Requirement 4.2.B.1 also indicates that when a channel (and/or the affected PCIV) is placed in an inoperable status solely for performance of required instrumentation Surveillances, entry into associated LCO and required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Trip Function maintains isolation capability. Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance, the channel must be returned to operable status or the applicable LCO entered and required Actions taken. This allowance is based on the reliability analysis (Refs. 6 and 7) assumption of the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months testing allowance does not significantly reduce the probability that the PCIVs will isolate the penetration flow path(s) when necessary.

Surveillance Requirement 4.2.B.2 The Logic System Functional Test demonstrates the operability of the required initiation logic and simulated automatic operation for a specific channel.

The automatic initiation testing required by the PCIV Technical Specifications overlaps this Surveillance to provide testing of the assumed safety function. For Main Steam Line Isolation Trip Functions, a simulated automatic actuation, which opens all pilot valves of the main steam line isolation valves, shall be performed such that each trip system logic can be verified independent of its redundant counterpart. The Frequency of "once every Operating Cycle, is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has demonstrated that these components will usually pass the Surveillance when performed at the specified Frequency.

No WUIlmULILLL 761 JUm=enUmeLL=AL rl.

1

VYNPS BASES: 3.2.B/4.2.B PRIMARY CONTAINMENT ISOLATION SURVEILLANCE REQUIREMENTS (continued)

Table 4.2.2, Check Performance of an Instrument Check once per day for Trip Functions l.a, l.b, i.e, l.h, and 2.b, ensures that a gross failure of instrumentation has not occurred. An Instrument Check is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. An Instrument Check will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each Calibration.

Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit. The Frequency is based upon operating experience that demonstrates channel failure is rare. The Instrument Check supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.

Table 4.2.2, Functional Test For Trip Functions l.a, l.b, l.c, l.d, l.e, l.f, 2.a, 2.b, 3.a, 3.b, 3.c, 3.d, 4.a, 4.c, 4.d, 4.e, 4.f, and 5.a, a Functional Test is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of "Every 3 Months" is based on the reliability analysis of References 6 and 7.

Table 4.2.2, Calibration For Trip Functions l.a, l.b, 1.c, l.d, 1.e, l.f, 2.a, 2.b, 3.a, 3.b, 3.c, 3.d, 3.e, 4.a, 4.b, 4.c, 4.d, 4.e, 4.f, and 5.a, an Instrument Calibration is a complete check of the instrument loop and the sensor. This test verifies that the channel responds to the measured parameter within the necessary range and accuracy. An Instrument Calibration leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. The specified Instrument Calibration Frequencies are based upon the time interval assumptions for calibration used in the determination of the magnitude of equipment drift in the associated setpoint analyses.

Amen=1ulienst ST-N Lnu.

76m

VYNPS BASES: 3.2.B/4.2.B PRIMARY CONTAINMENT ISOLATION SURVEILLANCE REQUIREMENTS (continued)

For Trip Functions l.a, l.c, l.e, 2.a, and 2.b, a calibration of the trip units is required (Footnote (1)) once every 3 months. Calibration of the trip units provides a check of the actual setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the calculational as-found tolerances specified in plant procedures.

The Frequency of every 3 months is based on the reliability analysis of References 6 and 7 and the time interval assumption for trip unit calibration used in the associated setpoint calculation.

REFERENCES

1. Technical Requirements Manual.

2. UFSAR, Chapter 14.

3. UFSAR, Table 6.5.3.

4. UFSAR, Section 14.6.5.

5. UFSAR, Section 14.5.4.1.

6. NEDC-31677P-A, "Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation," July 1990.

7. NEDC-30851P-A Supplement 2, "Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation," March 1989.

AhI L amnenumet:11 E76n "U.

Current Technical Specifications Markups 3.2.B and 4.2.B Primary Containment Isolation

VYNPS 3.2 LIMITING CONDITIONS FOR 4.2 suRVEILLAmcE REQUIREMENTS OPERATION 3.2 PROTECTIVE INSTRUMENT SYSTEMS 4.2 PROTECTIVE INSTRUMENT SYSTEMS I -

I- -34p-llcabilit:

Applies to the operational status Applies to the surveillance of the plant instrumentation requirements of the systems which initiate and instrumentation systems which control a protective function. initiate and control a protective function.

tbective: Objective:

ro assure the operability of To verify the operability of rotective instrumentation protective instrumentation ystems. systems.

pecification: Specification:

. Emergency Core Cooling System A. Emergency Core Cooling System When the system(s) it initiates Instrumentation and logic r controls is required in systems shall be functionally ccordance with Specification tested and calibrated as I .5, the instrumentation which indicated in Table 4.2.1 nitiates the emergency core

'I. ooling system(s) shall be perable in accordance with able 3.2.1. v C- -* ~ v- 4 s4?&,rk fL C.isc B. Primay ContXwainment Isolation lB. Primary Containment Isolation ovef k-61e TJce\,al Aw >

say11 Kfw3^&,co( Adi^ lsok-400 >

¢;7V^ 'ALc e is s iSAA' & '° S eS LI o 1A c P XT I

ij14s#A44:k 1.eL. ' an 4- 1 ni

, 'T- "I- L.

I Amendment No. 164 .6,tv we 4v G 1^" rvckC

Sl>^ r6;

- rE;&

VYNPS A.'1 A.

P1t AftfJ Coor~lvIA . TABLE 3.2. (

&-oA1IO 4 (Cont'd) u_&@ LXAC.-

EMERGE Y CORE COOLING SYSTEM ACTUATO INSTRUMENTATION CMoift 5..

014 ui-S Minimum Number of Operable Instrument B'bw Pressure Coolant Injection Systemok & B Ufote 1) )

Lo_

C 6quir Min ACTION WhU urn Cogitiov I

Channels per Trip or Oratio K System Trip Function Trip b yStting reNo Satisfied (Note 8) Low Reactor Pressure 300 < p < 350 psig Note 1 (PT-2-3-56C/D(M))

2 (Note 8) High Drywell Pressure <2.5 psig Note 10 (PT-10-101(A-D)(M))

2 (Note 8) Low-Low Reactor Vessel Water >82.5" above top of Note 10 Level (LT-2-3-72(A-D)(S1)) enriched fuel 1 (Note 9) Reactor Vessel Shroud Level >2/3 core height Note 13 (LT-2-3-73A/B(M))

1 (Note 9) Time Delay (10A-K72A & B) <60 seconds Note 12 1 (Note 9) Pump Start Time Delay < t < 5 seconds Note 1 1 qpe)

,Pump (A-D) Discharge Lg (Note 8) Pressure (PS-10-105(A-H))

2 (Note 8) High Drywell Pressure <2.5 psig Note 13 (PT-10-101(A-D)(S1))

A Amendment No. 44, 44, 48, 110, 6-44, 4.64, -4w, 186 39

VYNPS ElI ?I-so-o DA coaoilAwr TABLE 3.2 4 (Cont'd)

EMEREW CORE 40OLING SY4TEM ACUTO INSTRUMENTATION T44.:,

_row Presdure Coola Injectio System A & 4(Note 1) . ,

I .- 1A Minimum Number of kequ ed ACT NWh I Operable Instrument M wimum Cgditio #

Channels per Trip 7For eratio System Trip Function Trip 5 e<SettinX Are N Satisfied 1 (Note 9) Time Delay (1OA-K45A & B) <6 minutes Note 2 (Note 8) Low Reactor Pressure 300 < p < 350 psig Note 11 (PT-2-3-56A/B(M) &

PT-2-3-52C/D(M))

1 (Note 8) Auxiliary Power Monitor Note 10 (LNPX C/D)

\Q (Note 8) Pump Bus Power Monitor Note 10 7.6 Trip S v em Logic Note)

-- - - I- - --- I I N---

L.

Amendment No. 44, Q11, 4, 464, 186 40

TABLE 3.2. s

1. Each of the two Core Spray, LPCI and RPT, subsystems are initiated and a controlled by a trip system. The subsystem 'B' is identical to the subsystem "A".

2. If the minimum number of operable instrument channels are not available, the inoperable channel shall be tripped using test jacks or other permanently installed circuits. If the channel cannot be tripped by the - e means stated above, that channel shall be made operable within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or an orderly shutdown shall be initiated and the reactor shall be in the cold shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3. One trip system with initiating instrumentation arranged in a one-out-of-two taken twice logic.

4. One trip system with initiating instrumentation arranged in a one-out-f /

-( 7. the m unuter of oppdble cha=nnSs are not vailable, she sysIpm L Vbleand in~e d

qonsid requroftsSf t~

Any one of the two trip systems will initiate ADS. If the minimum number of operable channels in one trip system is not available, the requirements of Specification 3.5.F.2 and 3.5.F.3 shall apply. If the minimum number of operable channels is not available in both trip systems, Specifications 3.5.F.3 shall apply.

7. One trip system arranged in a two-out-of-two logic.

Opeatin ad rquired ACTIONS may be delayed for up to 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months spriiviaed required surveillances, entry into associated Limiting Conditions For / -

Operation and required ACTIONS may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months poied/;,

When a channel is placed in an inoperable status solely for performance o A ~_ rpcblit i~~~~~~~~~~~4n12Etiaion cap-abilit rRcruainPm

^..

With one or more channels inoperable Trp Fnd/toniLPC-:

& Within one hourbmefrom discovery of helayed A.'.

met, noTIN

.LOeatinadreqie loss of i or- sptem n hus fea ureas) s one diision boo. it oe or more channels inoperable for ijction Spermis B. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , place channel in trip.

C. If required actions and associated completion times of actions A or B -

and are not met, immediately declare the associated systems inoperabl Amendment No. i9, 4.64, 186 44

lAI VYNPS 4 4cfi6iIS 65tJAJGb p',QM ,'IcT.@95 /

TABLE 3.2.2 Minimum Number of Operable Instrument PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION

/

_idATOfhf~

Mini m Cond ions FoVOperatifn Are Channels per Trip Not atisfi (ot 2)

System Trip Function Trip Setting

>82. 5J(bc )v.he A

_ _t Low-Low Reactor Vessel Water 3-5cAb(6, Leve <g K di 1.k High Main Steam Line Area j9*Ib ZoF /o'f

'" eqA#W" raited flow High Main Steam Line Flow c140% of irated flow bc TD-PV2~- ( 16 - ) (9(M)

Low Main Steam Line Pressure >800 psig g-Hiqh Main Steam Line Flo c40% of riFted flow r(DPY 2 A,1Z ~R V11 PC,ll/qb(S~t E3 LowRea r Ve-ssel Water Leve r(LT-3-5/AB___ a Shtem 1 High Drywell Pressure P te io C_4EE

.7. k Condenser Low Vacuum c12" Hg a bsolute

-401,.01A (1 ./ Trip Stem Logic / !10 "4, BS, 94, 446, 212 45 Amendment No. #, 64,

VYNPS TABLE 3.2.2 (Cont' d)

HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION INSTRUMENTATIC)N Minimum Number of eirfd ACTI9N When operable Instrument Minrium Co itions/

Channels per Trip F; oe ion Ar

- %,n 4n+;

nn" TrIip Settin Not Satisfied Oiyuslm A&L. v CUISEV 0 -. High Steam Line Space 1 Note (Notes I 5 Temperature i LA,(I 34)(B-D 1 High Steam Line d/p (' <195 inches of water Note

-3.6 (Notes (El- I Line Break) KDPTSP777 4_ -, LowLHPCI Steam Supply >70 psig Note S c I (Notes %i a, E

-()l _C-Note

<35 minutes Note 202 46 Amendment No. 48, 44g, 464, 484,

VYNPS Ei]

TABLE 3.2.2 (Cont'd)

REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION INSTRUMENTAT:ION se-*AcOuft e44 1 equie ACIO Whe Mini Con tion Fo Operaion f e Trip Settin, \ ;_atisfied Note

<35 minutes Note 4,b Note 4,C

<195 inches of water Note

>50 psig Note I -_ 7 Z 3< t <7 seconds Note <

P-4.e (Notest 1R:

vIcd" 6MSd rp Amendment No. 44, 444, 144, 484, 202 47

~VYNPS

/TBE 3.2.21tNOTIES main steam line low pressure need be available only in the R m

2. If tie minimum np er of operable ins ument channels are available for one trip syst that

. trip system P 11lbe tripped. If fe minimum number of operabl d nstrument channels e not available for) th trip systems, the appr;riate actions listed 6elow shall be taken:

AT0iE an o e have reactor in the cold shutdown 2.b ion in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

dtIOA mit e o load r uctIon an have reactor in "Hot Standby" ahn 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

cves In E nd co <ywit pecific

/-f-w

@d t w`1 6.

7 T-he main steam line high flow is ay in the "Refuel, e(

Wen c e A Atha ffected pia c ml eD s eed. L C 12htdinche Hgd abSol ruteprovdedshtbt.ubn tp n yasvle r

-r A

10. A y lock switch is provided to permit the bypass of this trip function to solelorspromne frqie place plant enable inoperable in an startup andstatus shutdown when the condenser vacuum is greater than 12 inches Hg absolute provided that both turbine stop and bypass va for Opteraio an reuie ACIN/a edlye o pt or c d x a lWhen a c ormoel, and/or the affected primary containment isolation vave, os placed in an inoperable status solely for performance of required instrumentation surveillances, entry into associated Limiting Conditions/

for Operation and required ACTIONS may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months prov~maintainedheastore ioTrion capabiityains isolation capability.IO W vy tainment by i S3ecification byrequired 4t 37.B~athereshallbetW operable or t A< ped trip systet'efor eac xcept as provided for be-low: L.

Sr With one or more tions with isolation capability not bo're,I-b maintained restore isolation capability in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or take the ACTION required by Table 3.2.2.

AcTioj t. With one or more channels inoperable, place the inoperable channels (s) t ain the tripped condition within:

1) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for trip functions common to RPS instrumentation, and

2) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for trip functions not common to RPS instrumentation, or, initiate the ACTION required by Table 3.2.2.

peE-A77o4J = .oipJePAIIOJS j~*~A p^T9So /1>od"7'b Ar 4 kesouvr ofwe eof~ofCL)Ijf it~ &.jiTA at 771 a

A&,{

Mos e cmsj MAY 5( 014PSOkCA71b ,Mwett77ejTrLY ev AU.8Jbe4 J Co"0ro45.

Amendment No. #, -64,111, +8X, 212 48

VYNPS TABLE 3.2.2 NOTES (Cont'd)

Whenever the High Pressure Cooling Injection System and Reactor Core Isolation Cooling System are required to be operable in accordance with AO-"* Specification 3.5, the low steam supply pressure automatic isolation trip 17t system shall be operable except as providd below:

A.Josr Z With the iolation trip function not maintained, restore

{.k isolation cap in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or take the ACTION required by Table 3.2.2.

AtC7'01 B. With one or more required channels inoperable, place the inoperable

)00F channel(s) in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or take the ACTION}..M5) required by Table 3.2.2.

r EE*TO LW PA-714 % Al ArCS'7 4a7 of

{ OPLY16 Pi7rq ,,L BAY EBB Ma76 eTjC 7C

/Pv17c ,Tre-JrW VN ,J$7*sr1* j ,

)'

Amendment No. 202 48a

____1 VYNPS TABLE 2. E (Cont Id)

(riT TE&D Qf iDFREQUENCIES A MRGEfiCORE COOpdING ACTUATIO2 INSTRUMENTATION Low Pre sure Cool nt Iniection 2ystem.)

Trip Function Functional Test (0- EA. i Calibration Instrument Check o eactor ressure Every Three Months Once/operatng Cycle --

(PT-2-3-56C/D(M))

High Drywell Pressure Every Three Months Once/Operating Cycle Once Each Day (PT-10-lOlA-D(M))

-Low Reactor Vessel Every Three Months Once/Operating Cycle Once Each Day ater Level/

eactor Vessel Shroud Every Three Months Once/Operating Cycle -

Auxilary Poesr MonEvery Three Months Three Months Every R umDischarge Every Three MonVtLs-1 EUCT y Th-e ots-- -

Pressure High Drywell Pressure Every Three Months Once/Operating Cycle (PT-10-iStA-D(Sl)) O ow Reactor Pressure Every Three Months Once/Operating Cycle -

(PT-2-3-56A/B)(M) &

52C/D(M))

uiliary Power Monitor Every Three Months None Once Each Day/

pBsPwrMnnr erThee on+-h NP Once Each DS

.5e,4.jgX STrip System Logic Once/Operating Cycle rnc ratin ct-A.2.oJ Amendment No. A-, .&, *4, O4, Q4, i4l, 24, 4-4, 4A6i, 186 60

VYNPS TABLE 4.2.2

( ff TEt~~ Sgg* FREQUENCIES PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION Trip Function Functional Tes Calibratio~no Instrument Check

.a Low-Low Reactor Vessel Every Three Months Once/Operating Cycle Once Each Day Watehvel High Steam Line Area Every Three Months Each Refueling a Temperature

,.e High -am Line Flow Every Three Months Once/Operating Cycle Once Each Day I. / 9 9* _, ,A - re d 1.d Low Main Steam Line Every Three Months Every Three Months Pressure Low Reactor Vessel Water Every Three Months Once/Operating Cycle Level High Drywell Pressure Every Three Months Once/Operating Cycle Once Each Day Condenser Low Vacuum Every Three Months Every Three Months

_ _A~

S .2.8t Trip System Logic Once/Operating Cycle

)-L#ArldJ) )j 4J Amendment No. #, &S, A4, 44S, a34, 446, 212 64

VYNPS TABLE 4.2.2 (Cont'd)

<i0 ) TE ND Q e R FREQUENCIES HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION INSTRUMENTATION Trip Function Functional Test Calibration Instrument Check High Steam Line Space Every Three Months Each refueling outage --

1,66 Temperature 3 6 High Steam Line D/P Every Three Months Every three months --

(Steam Line Break) 3.c Low HPCI Steam Supply Every Three Months Every three months --

Pressure 31 Main Steazft Line Tunnel Temperature Every Three Months Each refueling outage --

S.LTrip System Logic Once/operating cycle ating cycle 3e 3T.epesvM¢, asT^<

Amendment No. i&, 496, 41.0, 4-11, 186 65

VYNPS TABLE 4.2.2 (Cont'd)

Q TE i ND FREQUENCIES REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION INSTRUMENTATION Trip Function Functional Testwt E ~ C Instrument Check 4Ramin 4

Steam Line Tunnel Every Three Months Each refueling outage k Temperature 4 c High Steam Line Space Every Three Months Each refueling outage Temperature 44A High Steam Line d/p Every Three Months Every three months including time delay

4. e relays (Steam Line Break)

Low RCIC Steam Supply Every Three Months Every three months Pressure Trip System Logic Once/operating cycle Amendment No. i&, 44, 44G, 440, 144, 186 66

- D VYNPS TABLE 4.2 NOTES

_.__ _.e_

2. During eag refueling outage,,?Imulated automatic act ation which opens

\all piJ valves shall be formed such that eachip system logic can be S ified independent its redundant count <art.

3. Trip system on shall include only time delay relays and timers necessary for proper functioning of the trip system.

r4. This instrumentation is excepted from functional test definition. The functional test will consist of injecting a simulated electrical signal into the measurement channel.

td ete21

8. Funct ' tests and calibratio re not required systems are no.

re red to be operable. _

9. The thermocouples associated with safety/relief valves and safety valve position, that may be used for back-up position indication, shall be verified to be operable every operating cycle.

10. Separate functional tests are not required for this instrumentation. The calibration and integrated ECCS tests which are performed once per operating cycle will adequately demonstrate proper equipment operation.

11. Trip system logic functional tests will include verification of operation of all automatic initiation inhibit switches by monitoring relay contact movement. Verification that the manual inhibit switches prevent opening all relief valves will be accomplished in conjunction with Section 4.5.F.I.

12. Trip system logic testing is not applicable to this function. If the required surveillance frequency (every Refueling Outage) is not met, functional testing of the Reactor Mode Switch-Shutdown Position function shall be initiated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after the reactor mode switch is placed in Shutdown for the purpose of commencing a scheduled Refueling Outage.

13. Includes calibration of the RBM Reference Downscale function (i.e., RBM upscale function is not bypassed when >30% Rated Thermal Power).

Amendment No. 46i, 9., a r, 464, 8&, 211, 212 74

Safety Assessment Discussion of Changes 3.2.B and 4.2.B Primary Containment Isolation

SAFETY ASSESSMENT OF CHANGES TS: 3.2.B1/4.2.B - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION ADMINISTRATIVE A.1 Inthe revision of the Vermont Yankee Nuclear Power Station (VYNPS) current Technical Specifications (CTS), certain wording preferences or conventions are adopted which do not result intechnical changes (either actual or interpretational). Editorial changes, reformatting, and revised numbering are adopted to make the VYNPS Technical Specifications (TS) more consistent with human factor principles used inthe Boiling Water Reactor Improved Standard Technical Specifications (ISTS), NUREG-1433, Rev. 2. These format and presentation changes are being made to improve usability and clarity. The changes are considered administrative.

A.2 CTS 3.2.B specifies an Applicability for primary containment isolation instrumentation of "When primary containment is required, in accordance with Specification 3.7." CTS Table 3.2.2 Note 12 specifies an Applicability of 'Whenever Primary Containment integrity is required by Specification 3.7.A.2.' CTS Table 3.2.2 Note 13 specifies an Applicability of

'Whenever the High Pressure Cooling Injection System and Reactor Core Isolation Cooling System are required to be operable in accordance with Specification 3.5' for the High Pressure Coolant Injection (HPCI) System and Reactor Core Isolation Cooling (RCIC)

System Low Steam Supply Pressure Trip Functions. Specification 3.5 provides requirements for the HPCI and RCIC Systems. Specification 3.7 includes the requirements for the primary containment. This change provides an explicit Applicability, inproposed Table 3.2.2 for each primary containment isolation instrumentation trip function. The specified Applicabilities, in proposed Table 3.2.2, are consistent with the Modes and conditions when primary containment integrity is required to be operable by Specification 3.7, and, for the HPCI System and RCIC System Low Steam Supply Pressure Trip Functions, when HPCI and RCIC are required to be operable by Specification 3.5; except as provided and justified inchange L.1 below. Therefore, this change does not involve a technical change, but is only a difference of presentation preference and isconsidered administrative. The change, providing explicit Mode or conditions of Applicability for each trip function, is consistent with the ISTS.

A.3 CTS 4.2.B specifies that instrumentation and logic systems shall be functionally tested and calibrated as indicated inTable 4.2.2. Inproposed Surveillance Requirement (SR) 4.2.B.1, the reference to "and logic system," is deleted since associated logic systems are considered part of the primary containment isolation instrumentation Trip Functions in both proposed and CTS Tables 3.2.2 and 4.2.2. It is not necessary to explicitly identify logic systems in CTS 4.2.B, since proposed SR 4.2.B.2 (CTS Table 4.2.2 requirements to perform Functional Tests of Trip System Logic) continues to require performance of surveillance testing of Trip System Logic (i.e., performance of Logic System Functional Tests for each primary containment isolation instrumentation Trip Function). Therefore, this change is considered administrative.

VYNPS 1 Revision 0

SAFETY ASSESSMENT OF CHANGES TS: 3.2.B14.2.B - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION ADMINISTRATIVE (continued)

A.4 CTS 4.2.B includes reference to CTS Table 4.2.2 for functional test and calibration requirements for primary containment isolation instrumentation. CTS 4.2.B isrevised, in proposed SR 4.2.B.1, to also include reference to check requirements consistent with CTS Table 4.2.2. This change is a presentation preference and does not alter the current requirements to periodically perform checks of certain primary containment isolation instrument trip functions. Therefore, this change isconsidered administrative in nature.

A.5 CTS Table 3.2.1, Note 9 and CTS Table 3.2.2 Note 11, provide allowances to delay entry into actions for 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for the situation of a channel inoperable solely for performance of surveillances. These allowances are moved to proposed SR 4.2.B.1 and the allowances of these two notes are combined. This change does not involve atechnical change, but is only a difference of presentation preference. Therefore, this change is considered administrative.

A.6 CTS Table 3.2.1 includes a Low Pressure Coolant Injection System - Low Reactor Pressure Trip Function (PS-2-128A &B). The purpose of this trip function isto provide isolation of the shutdown cooling portion of the Residual Heat Removal (RHR) System to protect that system from overpressurization due to high reactor pressure. This isolation provides for equipment protection to prevent an intersystem LOCA scenario. As a result, the name of this trip function is revised to 'Residual Heat Removal Shutdown Cooling Isolation - High Reactor Pressure" inproposed Table 4.2.2 to more accurately reflect its function. The design and operation of the actual instrumentation is unchanged. Therefore, this change isconsidered administrative.

A.7 CTS Tables 3.2.1 and 4.2.1 include requirements for the Low Pressure Coolant Injection System - Low Reactor Pressure Trip Function (PS-2-128A & B)and associated Trip System Logic. The purpose of this Trip Function and associated Trip System Logic isto provide isolation of the shutdown cooling portion of the Residual Heat Removal (RHR)

System to protect that system from overpressurization due to high reactor pressure. This isolation provides for equipment protection to prevent an intersystem LOCA scenario.

These requirements are to be moved to the primary containment isolation instrumentation TS. Given the isolation function of this Trip Function and associated Trip System Logic, they are more appropriately located inproposed Tables 3.2.2 and 4.2.2. As a corresponding change, the reference to initiation capability in CTS Table 3.2.1 Note 10 (proposed Table 3.2.2 Action Note 1)iseditorially changed to isolation capability. This change does not involve a technical change, but is only a difference of presentation preference and isconsidered administrative. The change, including the requirements this Trip Function and associated Trip System Logic in the primary containment isolation instrumentation TS, isconsistent with the ISTS.

VYNPS 2 Revision 0

SAFETY ASSESSMENT OF CHANGES TS: 3.2.B14.2.B - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION ADMINISTRATIVE (continued)

A.8 Notes 10 and 5 to CTS Table 3.2.1 provide actions when the minimum number of channels per trip system requirement is not met. These requirements are divided and identified in two separate columns in proposed Table 3.2.2 titied, 'ACTIONS WHEN REQUIRED CHANNELS ARE INOPERABLE' and 'ACTIONS REFERENCED FROM ACTION NOTE 1.' This change is a presentation preference and does not alter the current action requirements when the required channels are inoperable. Therefore, this change is considered administrative in nature.

A.9 CTS 3.2.1 and 4.2.1 and associated Notes provide requirements related to Recirculation Pump Trip instrumentation. The requirements applicable to the Recirculation Pump Trip instrumentation are physically moved and changes addressed inproposed Specifications 3.2.1 and 4.2.1. Therefore, this change does not involve a technical change, but isonly a difference of presentation preference and is considered administrative. CTS 3.2.1 and 4.2.1 and associated Notes also provide requirements for ECCS Instrumentation. The ECCS instrumentation requirements are included in proposed Tables 3.2.1 and 4.2.1.

Changes to the ECCS instrumentation requirements are addressed in the safety assessment of changes for TS 3.2.A/4.2.A, ECCS Instrumentation. Therefore, this change does not involve a technical change, but isonly a difference of presentation preference and isconsidered administrative. CTS Table 4.2 Notes 4, 12, and 13 provides requirements that apply to control rod block instrumentation. The control rod block instrumentation is located inproposed Specifications 3.2.E and 4.2.E. Therefore, the requirements of CTS Table 4.2 Note 4 are physically moved and changes addressed inproposed Specifications 3.2.E and 4.2.E. Therefore, this change does not involve a technical change, but is only a difference of presentation preference and is considered administrative. CTS Table 4.2 Note 9 provides requirements that apply to post-accident monitoring instrumentation. The post-accident monitoring instrumentation is located inproposed Specifications 3.2.G and 4.2.G. Therefore, the requirements of CTS Table 4.2 Note 9 are physically moved and changes addressed inproposed Specifications 3.2.G and 4.2.G. Therefore, this change does not involve a technical change, but isonly adifference of presentation preference and isconsidered administrative.

A.10 The Primary Containment Isolation Instrumentation portion of CTS Tables 3.2.2 and 4.2.2 isdivided into two sections, Main Steam Line Isolation (Trip Function 1), and Primary Containment Isolation (Trip Function 2) inproposed Tables 3.2.2 and 4.2.2. The appropriate individual trip functions are placed with the proper isolation. Since the current requirements are maintained (except as addressed inthe changes below), the change is considered to be administrative in nature. This change isconsistent with the ISTS.

A.1 1 Note 12 to CTS Table 3.2.2 provides actions when the minimum number of channels per trip system requirement is not met. These requirements are identified ina separate column in proposed Table 3.2.2 titled, 'ACTIONS WHEN REQUIRED CHANNELS ARE INOPERABLE.' This change is a presentation preference and does not alter the current action requirements when required primary containment isolation instrumentation channels are inoperable. Therefore, this change is considered administrative in nature.

VYNPS 3 Revision 0

SAFETY ASSESSMENT OF CHANGES TS: 3.2.B1/4.2.B - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION ADMINISTRATIVE (continued)

A.12 The name of the CTS Table 3.2.2 High Main Steam Line Flow Trip Function (DPT-2-116A, 117B, 118C, 119D (S1)) is revised to reflect the condition when this trip function is available. This trip function is available only inthe Refuel, Shutdown, and Startup Modes (i.e., not available in Run). Therefore, the name of proposed Tables 3.2.2 and 4.2.2 Trip Function I.e is "High Main Steam Line Flow - Not in RUN.' The design and operation of the actual instrumentation is unchanged. Therefore, this change is considered administrative.

A.13 CTS 3.2.B requires that the instrumentation inCTS Table 3.2.2 be operable when primary containment integrity is required in accordance with Specification 3.7. CTS 3.7 requires primary containment integrity when reactor water temperature is above 2120 F and fuel is in the reactor vessel. The Primary Containment Isolation Instrumentation Low Main Steam Line Pressure Trip Function requirements of CTS Table 3.2.2 are modified by CTS Table 3.2.2 Note 1. CTS Table 3.2.2 Note 1 states that the main steam line low pressure need be available only inthe Run Mode. The intent of this note is to waive the operability requirements of the Primary Containment Isolation Instrumentation Low Main Steam Line Pressure Trip Function when the reactor is not in the Run Mode. The 'APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS' column inproposed Table 3.2.2 requires the Low Main Steam Line Pressure Trip Function (proposed Table 3.2.2, Trip Function 1.d) to be operable inthe Run Mode which isequivalent to CTS requirements. As such, this change is considered administrative innature. This change is consistent with the ISTS.

A.14 CTS 3.2.B requires that the instrumentation in CTS Table 3.2.2 be operable when primary containment integrity is required inaccordance with Specification 3.7. CTS 3.7 requires primary containment integrity when reactor water temperature is above 212 F and fuel is in the reactor vessel. The Primary Containment Isolation Instrumentation High Main Steam Line Flow Trip Function (DPT-2-116A, 117B, 118C, 119D (S1)) requirements of CTS Table 3.2.2 are modified by CTS Table 3.2.2 Note 6. CTS Table 3.2.2 Note 6 states that the main steam line high flow is available only inthe Refuel, Shutdown, and Startup Modes. The intent of this note is to waive the operability requirements of the Primary Containment Isolation Instrumentation High Main Steam Line Flow Trip Function (DPT-2-116A, 117B, 118C, 119D (S1)) when the reactor is inthe Run Mode. The 'APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS' column in proposed Table 3.2.2 requires the High Main Steam Line Flow Trip Function (proposed Table 3.2.2, Trip Function 1.e) to be operable in the Startup/Hot Standby, Hot Shutdown, and Refuel (with reactor coolant water temperature > 2120F) Modes which isequivalent to CTS requirements. As such, this change is considered administrative in nature.

VYNPS 4 Revision 0

SAFETY ASSESSMENT OF CHANGES TS: 3.2.B14.2.B - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION ADMINISTRATIVE (continued)

A.15 CTS 3.2.B requires that the instrumentation in CTS Table 3.2.2 be operable when primary containment integrity is required in accordance with Specification 3.7. CTS 3.7 requires primary containment integrity when reactor water temperature is above 212 0F and fuel is in the reactor vessel. The Primary Containment Isolation Instrumentation Condenser Low Vacuum Trip Function requirements of CTS Table 3.2.2 are modified by CTS Table 3.2.2 Note 10. CTS Table 3.2.2 Note 10 states *Akey lock switch is provided to permit the bypass of this trip function to enable plant startup and shutdown when condenser vacuum is greater than 12 inches Hg absolute provided that both turbine stop and bypass valves are closed.' The intent of this note is to waive the operability requirements of the Primary Containment Isolation Instrumentation Condenser Low Vacuum Trip Function when the reactor is not inthe Run Mode and all turbine stop and bypass valves are closed. The "APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS" column in proposed Table 3.2.2 requires the Condenser Low Vacuum Trip Function (proposed Table 3.2.2, Trip Function 1.1) to be operable in the Run, Startup/Hot Standby (with any turbine stop valve or turbine bypass valve not closed), Hot Shutdown (with any turbine stop valve or turbine bypass valve not closed), and Refuel (with reactor coolant water temperature > 212 0F and with any turbine stop valve or turbine bypass valve not closed) Modes which isequivalent to CTS requirements. As such, this change isconsidered administrative innature.

A.16 All HPCI System Isolation High Steam Line Space channels are required to be operable to assure isolation with the worst single failure. CTS Table requires a minimum of 2 per set of 4 channels per trip system of the HPCI System Isolation High Steam Line Space Trip Function (proposed Table 3.2.2, Trip Function 3.a) to be operable. There are three locations (i.e., 3 sets), each monitored by one set of 4 channels. As a result, there are a total of 12 channels for this trip function, with 6 channels per trip system. Therefore, the minimum number of channels per trip system required to be operable for this trip function is specified as "6"inproposed Tablq 3.2.2. Since this change involves no design change but is only a difference of nomenclature and presentation preference, this change is considered administrative.

A.17 The CTS Table 3.2.2 Note 3 action requirement to "comply with Specification 3.5" is an unnecessary reminder that other Technical Specifications may be affected when isolation valves inthe HPCI System or RCIC System are closed. This is essentially a cross reference" between Technical Specifications that has been determined to be adequately provided through training. Therefore, the deletion isconsidered to be administrative. This change isconsistent with the ISTS.

A.18 CTS Table 3.2.2 Notes 12 and 13 refer to automatic isolation trip functions. However, CTS Table 3.2.2 (proposed Table 3.2.2) includes only automatic isolation trip functions. Manual isolation instrumentation trip functions are not included in the VYNPS CTS. Therefore, it is unnecessary to use the word 'automatic" when referring to CTS Table 3.2.2 isolation trip functions in proposed Table 3.2.2 Action Note 1. This change isa presentation preference VYNPS 5 Revision 0

SAFETY ASSESSMENT OF CHANGES TS: 3.2.B14.2.B - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION ADMINISTRATIVE A.18 and does not alter the current action requirements when required primary containment (continued) isolation instrumentation channels are inoperable. Therefore, this change isconsidered administrative innature.

A.19 CTS Table 3.2.2 Note 2 states, inthe first paragraph, 'If the minimum number of operable instrument channels are not available for one trip system, that trip system shall be tripped.

If the minimum number of operable instrument channels are not available for both trip systems, the appropriate actions listed below shall be taken...' However, due to the presentation of the Notes inCTS Table 3.2.2, Note 2 actions are only taken after actions associated with CTS Table 3.2.2 Note 12 or 13, as applicable, are taken. Since the CTS Table 3.2.2 Note 12 and 13 actions (proposed Table 3.2.2 ACTION Note 1)already provide the appropriate NRC approved actions for each of the conditions addressed in the first paragraph of CTS Table 3.2.2 Note 2,the first paragraph of CTS Table 3.2.2 Note 2 is unnecessary and is deleted. Since actions when required primary containment isolation instrumentation channels are inoperable will confinue to be taken inthe same manner and in the same time period, the deletion isconsidered administrative in nature.

A.20 For the Trip System Logic associated with the Low Pressure Coolant Injection System -

Low Reactor Pressure Trip Function (PS-2-128A &B), CTS Table 4.2.1 includes a requirement to perform a calibration of Trip System Logic once per Operating Cycle. For the Trip System Logic associated with the Primary Containment Isolation Instrumentation, CTS Table 4.2.2 includes arequirement to perform a calibration of Trip System Logic once per Operating Cycle. These requirements are modified by Table 4.2 Note 3. Note 3 states,

'Trip system logic calibration shall include only time delay relays and timers necessary for proper functioning of the trip system.' The Low Pressure Coolant Injection System - Low Reactor Pressure Trip Function instrumentation of CTS 4.2.1 and the Primary Containment Isolation Instrumentation Trip Functions of CTS 4.2.2 do not include any time delay relays or timers necessary for proper functioning of the trip systems. Therefore, this Note is deleted and, inproposed Table 4.2.2, the Residual Heat Removal Shutdown Cooling Isolation instrumentation (proposed Trip Function 5.a), Main Steam Line Isolation instrumentation (proposed Trip Functions l.a through 1.f) and Primary Containment Isolation instrumentation (proposed Trip Functions 2.a and 2.b) do not include calibration requirements for time delay relays or timers. As a result, this change does not involve a technical change, but isonly a difference of presentation preference and is considered administrative.

A.21 CTS Table 4.2 Note 8 states that functional tests and calibrations are not required when systems are not required to operable. The requirements of this Note are duplicated inthe CTS definition 1.0.Z, 'Surveillance Interval,' which states that these tests unless otherwise stated inthese specifications may be waived when the instrument, component, or system is not required-to be operable, but that these tests shall be performed on the instrument, component, or system prior to being required to be operable. Therefore, CTS Table 4.2 Note 8 is unnecessary and its deletion is considered to be administrative. The change is consistent with the ISTS.

VYNPS 6 Revision 0

SAFETY ASSESSMENT OF CHANGES TS: 3.2.B14.2.B - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION ADMINISTRATIVE (continued)

A.22 CTS Table 4.2.2 includes a requirement to perform a calibration of Trip System Logic once per Operating Cycle. This requirement is modified by Table 4.2 Note 3. Note 3 states, Trip system logic calibration shall include only time delay relays and timers necessary for proper functioning of the trip system." Inproposed Table 4.2.2, this requirement is reflected with explicit requirements to perform calibrations of the required HPCI System Isolation and RCIC System Isolation instrumentation time delay relays and timers (i.e., proposed Table 4.2.2 Trip Function 3.e., HPCI System Isolation - High Main Steam Tunnel Temperature Time Delay, and Trip Function 4.b, RCIC System Isolation - High Main Steam Tunnel Temperature Time Delay) once per Operating Cycle. Therefore, this change does not involve a technical change, but is only a difference of presentation preference and is considered administrative.

TECHNICAL CHANGES - MORE RESTRICTIVE M.1 CTS Table 3.2.1 includes a Low Pressure Coolant Injection System - Low Reactor Pressure Trip Function (PS-2-128A &B). The purpose of this trip function is to provide isolation of the shutdown cooling portion of the Residual Heat Removal (RHR) System to protect that system from overpressurization due to high reactor pressure. This isolation provides for equipment protection to prevent an intersystem LOCA scenario. The associated action for this Trip Function, inthe event the instrumentation is inoperable and not restored within the allowed time period, inCTS Table 3.2.1 Note 10 requires that the associated systems to be declared inoperable. Continued operation, in this condition, is then allowed for a limited duration in accordance the associated system TS. However, the TS actions for Low Pressure Coolant Injection System inoperabilities do not require that overpressure protection of the RHR System be provided. Therefore, in the same condition, proposed Table 3.2.2 Action Note 2.d will require the associated penetration to be isolated within one hour, thus restoring overpressure protection for the RHR System. This change represents an additional restriction on plant operation necessary to provide overpressure protection to prevent an intersystem LOCA scenario. The change isconsistent with the ISTS.

M.2 CTS Table 3.2.2 provides requirements for the Primary Containment Isolation Instrumentation - High Main Steam Line Area Temperature Trip Function. CTS Table 3.2.2 specifies that, for this trip function, the minimum number of operable instrumentation channels per trip system is 2 of 4 in each of 2 channels. There are 2 trip systems for this trip function, each with 2 channels. As such, the CTS require that only 2 temperature sensor inputs (out of 4) per channel be operable per trip system. Therefore, the total number of temperature sensors required to be operable by CTS Table 3.2.2 is8 (i.e.,

2/channel x 2 channels/trip system x 2 trip systems). The temperature sensor arrangement inthe channels is such that each of the 4 sensors in a channel monitors temperature of a different area of the main steam lines. This arrangement, when combined with the CTS allowance for any two sensors inone or more logic channels to be inoperable without affecting compliance with the Limiting Conditions for Operation, results in the possibility for a loss of isolation capability for a leak in a specific main steam line area. Therefore, the VYNPS 7 Revision 0

SAFETY ASSESSMENT OF CHANGES TS: 3.2.B1/4.2.B - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION TECHNICAL CHANGES - MORE RESTRICTIVE M.2 required number of number of operable channels per tip system for the High Main Steam (continued) Line Area Temperature Trip Function (proposed Table 3.2.2 Trip Function 1.b) is revised to 8 per trip system (for a total number of temperature sensors required to be operable of 16).

This change represents an additional restriction on plant operation to ensure adequate temperature monitoring of the main steam line areas is provided.

M.3 CTS 3.2.B requires that the Primary Containment Isolation Instrumentation High Main Steam Line Area Temperature Trip Function (proposed Table 3.2.2 Trip Function 1.b) and the High Main Steam Line Flow (DPT-2-(116-119) (A-D) (M)) Trip Function (proposed Table 3.2.2 Trip Function I.c) be operable inRun, Startup/Hot Standby, Hot Shutdown, and Refuel (with reactor coolant water temperature > 212 0F) Modes and that the Primary Containment Isolation Instrumentation High Main Steam Line Flow (DPT-2-116A, 117B, 118C, 119D (SI)) Trip Function (proposed Table 3.2.2 Trip Function 1.e) be operable in Startup/Hot Standby, Hot Shutdown, and Refuel (with reactor coolant water temperature >

212 0F)Modes. Inthe event minimum conditions for operation are not satisfied for these trip functions, CTS Table 3.2.2 requires compliance with CTS Table 3.2.2 Note 2.B. CTS Table 3.2.2 Note 2.B requires an orderly load reduction to be initiated and to the have the reactor InHot Standby within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . However, this action does not result in satisfying the isolation function of the inoperable instrumentation. Nor does this action result in placing the reactor ina Mode Inwhich the Instrumentation isnot required to be operable to provided isolation. As a result, CTS would allow continued operation with the isolation function of this Instrumentation not maintained. Therefore, proposed Table 3.2.2 Action Note 2.a will require, Inthe same condition, the associated main steam line to be isolated within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (which satisfies the isolation function of the inoperable instrumentation) or that the reactor be placed InHot Shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and InCold Shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (which places the reactor ina Mode inwhich the instrumentation is not required to be operable). The time period specified for exiting the Modes of Applicability are consistent with times to reach the same Modes provided inother VYNPS TS. This change represents an additional restriction on plant operation necessary to ensure that either the isolation function of the inoperable instrumentation issatisfied or that the reactor is placed Ina Mode inwhich the instrumentation Isnot required to function to provide isolation. The change is consistent with the ISTS.

M.4 CTS Table 3.2.2 Note 3 effectively specifies actions to isolate the affected penetration flow path (i.e., close isolation valves in the system) but does not state the time period inwhich this action isto be completed. Proposed Table 3.2.2 Action Note 2.d provides a one hour time period inwhich to complete the action of isolating the affected penetration flow.

Placing a limitation on the time allowed to complete the associated actions represents an additional restriction on plant operation since the time period allowed to complete the isolation will be controlled through TS. This change is consistent with the ISTS.

VYNPS 8 Revision 0

SAFETY ASSESSMENT OF CHANGES TS: 3.2.B/4.2.B - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION TECHNICAL CHANGES - MORE RESTRICTIVE (continued)

M.5 CTS Table 3.2.2 Note 13.B provides actions for inoperable time delay channels for HPCI and RCIC Isolation Instrumentation. These actions allow the inoperable channel to be tripped, rather than requiring the channel to be restored to operable status. The subject channels are associated with Time Delay Trip Functions. As such, placing these Time Delay Trip Functions channels intrip (as allowed by Note 13.B) does not perform the intended function (providing a time delay for Isolation after certain conditions are satisfied) and could adversely impact the capability of the HPCI or RCIC Systems to perform their intended functions. Therefore, the actions for inoperable channels of these Time Delay Trip Functions (proposed Table 3.2.2 Trip Functions 3.e, HPCI System Isolation - High Main Steam Line Tunnel Temperature Time Delay, 4.b, RCIC System Isolation - High Main Steam Line Tunnel Temperature Time Delay, and RCIC System Isolation - High Steam Line d/p Time Delay) are revised, inproposed Table 3.2.2 Action Note 1.a.2), to require that they be restored to operable status rather than being placed in trip. This change represents an additional restriction on plant operation and isconsistent with the ISTS (i.e.,

inoperable time delays are required to be restored to operable status rather than placed in trip).

M.6 CTS Table 4.2.2 does not include explicit requirements to calibrate trip units. Proposed Table 4.2.2 requires calibration of the trip units of the following Trip Functions every 3 months: Main Steam Line Isolation - Low-Low Reactor Vessel Water Level (proposed Table 4.2.2 Trip Function I.a); Main Steam Line Isolation - High Steam Line Flow (proposed Table 4.2.2 Trip Function 1.c); Main Steam Line Isolation - High Steam Line Flow-Not in Run (proposed Table 4.2.2 Trip Function I.e); and Primary Containment Isolation - High Drywell Pressure (proposed Table 4.2.2 Trip Function 2.b). The trip units of these Trip Functions are currently required by CTS Table 4.2.2 to be calibrated with the rest of the associated instrument loops once per operating cycle. Therefore, this change is more restrictive. This change is necessary to ensure consistency with assumptions regarding trip unit calibration frequency used inthe associated setpoint calculations. This change isconsistent with the ISTS.

M.7 CTS Table 3.2.2 specifies for the Primary Containment Isolation High Main Steam Line Area Temperature, the HPCI System Isolation High Steam Line Space Temperature, the HPCI System Isolation Main Steam Line Tunnel Temperature, the RCIC System Isolation Main Steam Line Tunnel Temperature, and the RCIC System High Steam Line Space Temperature Trip Functions that the Trip Settings be < 212 0F. The function of these instruments Isto provide Isolation inthe event of breaks in the associated steam lines. The CTS Trip Settings have been determined to be insufficient to ensure isolation occurs as assumed in the high energy line break and Equipment Qualification (EQ) Program analyses. Therefore, in proposed Table 3.2.2, the Trip Setting for the Primary Containment Isolation High Main Steam Line Temperature Area Temperature Trip Function (Trip Function I.b) has been decreased to < 196 0F for channels monitoring outside the steam tunnel and < 2000 Ffor channels monitoring Inside the steam tunnel; the Trip Settings for the HPCI System Isolation and RCIC System Isolation High Steam Line Space Temperature Trip Functions (Trip Functions 3.a and 4.c, respectively) have been VYNPS 9 Revision 0

SAFETY ASSESSMENT OF CHANGES TS: 3.2.B14.2.B - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION TECHNICAL CHANGES - MORE RESTRICTIVE M.7 decreased to < 1960F; and the Trip Settings for the HPCI System Isolation and RCIC (continued) System Isolation Main Steam Line Tunnel Temperature Trip Functions (Trip Functions 3.d and 4.a,' respectively) have been decreased to < 2000F. These revised Trip Settings are consistent the assumptions of the high energy line break and EQ Program analyses and correspond to the Analytical Limit used inthe associated setpoint calculations. To account for instrument uncertainties, the instrument setpoints and as-found tolerances (i.e.,

instrument operability limits) were developed using the Vermont Yankee Instrument Uncertainty and Setpoints Design Guide. The Instrument setpoints and as-found tolerances are located inplant procedures. This change represents an additional restriction on plant operation necessary to ensure that isolation of the associated steam lines occurs as assumed inanalyses.

TECHNICAL CHANGES - LESS RESTRICTIVE

'Generic LA.1 The CTS Tables 3.2.1, 3.2.2, 4.2.1 and 4.2.2 details relating to system design and operation (i.e., the specific instrument tag numbers) are unnecessary in the TS and are proposed to be relocated to the Technical Requirements Manual (TRM). Proposed Specification 3.2.B and Table 3.2.2 require the Primary Containment Isolation Instrumentation Trip Functions to be operable. Inaddition, the proposed Surveillance Requirements InTable 4.2.2 ensure the required Instruments are properly tested. These requirements are adequate for ensuring each of the required Primary Containment Isolation Instrumentation Trip Functions are maintained operable. As such, the relocated details are not required to be in the VYNPS TS to provide adequate protection of the public health and safety. Changes to the TRM are controlled by the provisions of 10 CFR 50.59. Not including these details inTS Isconsistent with the ISTS.

LA.2 CTS Table 3.2.1 includes requirements for the Low Pressure Coolant Injection System -

Low Reactor Pressure Trip Function (PS-2-128A & B). The purpose of this trip function is to provide isolation of the shutdown cooling portion of the Residual Heat Removal (RHR)

System to protect that system from overpressurization due to high reactor pressure. This isolation provides for equipment protection to prevent an Intersystem LOCA scenario. CTS Table 3.2.1 includes Trip Settings of 100 < p < 150 psig. The upper Trip Setting ensures the RHR System isisolated from the Reactor Coolant System prior to being overpressurized due to high reactor pressure. The lower Trip Setting isan operational detail that isnot directly related to the operability of the associated instrumentation. This detail is to be relocated to plant procedures. The upper Trip Setting Isthe required limitation for the parameter and this value isretained inthe WNPS TS. As such, the lower Trip Setting for this trip function Isnot required to be inTS to provide adequate protection of the public health and safety. Changes to the relocated lower Trip Setting inthe plant procedures will be controlled by the provisions of 10 CFR 50.59. Not including these details inTS isconsistent with the ISTS.

VYNPS 10 Revision 0

SAFETY ASSESSMENT OF CHANGES TS: 3.2.B14.2.B - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE (continued)

LA.3 The Trip Setting associated with reactor vessel water level trip function (proposed Table 3.2.2 Trip Function 1.a)iscurrently referenced to 'above the top of enriched fuel., This detail isto be relocated to the Bases. This reference is not necessary to be included in the VYNPS TS to ensure the operability of the associated primary containment isolation instrumentation. Operability requirements are adequately addressed in proposed Specification 3.2.B, Table 3.2.2 and the specified Trip Setting. As such, this relocated reference is not required to be inthe VYNPS TS to provide adequate protection of the public health and safety. Changes to the TS Bases are controlled by 10 CFR 50.59. Not including these details in TS is consistent with the ISTS.

LA.4 Details of the methods for performing CTS Table 3.2.2 Notes 2.A and 2.B (proposed Table 3.2.2 Action Notes 2.b and 2.c), associated with placing the reactor inCold Shutdown (i.e.,

initiating an orderly shutdown) or Hot Standby (i.e., initiating an orderly load reduction), are to be relocated to plant procedures. These details are not necessary to ensure the actions of placing the reactor inCold Shutdown mode or Hot Standby and exiting the applicable Mode of the associated primary containment isolation instrumentation is accomplished.

The requirements of proposed Table 3.2.2 and Table 3.2.2 Action Notes 2.b and 2.c are adequate to ensure this action is accomplished. As such, these relocated details are not required to be in the VYNPS TS to provide adequate protection of the public health and safety. Changes to the plant procedures are controlled by the provisions of 10 CFR 50.59.

Not including these details in TS isconsistent with the ISTS.

LA.5 For the Primary Containment Isolation Instrumentation - Condenser Low Vacuum Trip Function, CTS Table 3.2.2 Note 5 states that 'A key lock switch is provided to permit the bypass of this trip function to enable plant startup and shutdown when condenser vacuum isgreater than 12 inches Hg absolute provided that both turbine stop and bypass valves are closed.' The system design details in CTS Table 3.2.2 Note 5 are to be relocated to the Bases and the reference to this information is deleted from the WNPS TS. These design details are not necessary to be included inthe TS to ensure the operability of the Condenser Low Vacuum Trip Function instrumentation since operability requirements are adequately addressed in proposed Specification 3.2.B and Table 3.2.2. Therefore, these relocated details are not required to be inthe TS to provide adequate protection of the public health and safety. Changes to the Bases are controlled by the provisions of 10 CFR 50.59. Not including these details inTS is consistent with the ISTS.

VYNPS 11 Revision 0

SAFETY ASSESSMENT OF CHANGES TS: 3.2.B1/4.2.B - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE (continued)

LA.6 CTS Table 3.2.2 Note 12 contains design and operational details of the primary containment isolation instrumentation (i.e., there shall be two operable or tripped trip systems for each Trip Function). These details are not necessary to ensure the operability of associated primary containment isolation instrumentation. Therefore, the information in this note isto be relocated to Specification 3.2.B Bases and reference to this information is deleted from WNPS TS. The requirements of Specification 3.2.B and the associated Surveillance Requirements for the associated primary containment isolation instruments are adequate to ensure the instruments are maintained operable. As such, these relocated requirements are not required to be inthe VYNPS TS to provide adequate protection of the public health and safety. Changes to the TS Bases are controlled by the provisions of 10 CFR 50.59. Not including these details inTS isconsistent with the ISTS.

LA.7 CTS Table 3.2.2 Note 5 contains design details of the HPCI System Isolation - Low Steam Supply Pressure Trip Function and RCIC System - Low Steam Supply Pressure Trip Function instrumentation (i.e., one trip system arranged in a one-out-of-two twice logic).

These details are not necessary to ensure the operability of associated isolation instrumentation. Therefore, the information inthese notes is to be relocated to Specification 3.2.B Bases and reference to this information is deleted from WNPS TS.

The requirements of Specification 3.2.B and the associated Surveillance Requirements for these isolation instruments are adequate to ensure the instruments are maintained operable. As such, these relocated requirements are not required to be in the VYNPS TS to provide adequate protection of the public health and safety. Changes to the TS Bases are controlled by the provisions of 10 CFR 50.59. Not including these details inTS is consistent with the ISTS.

LA.8 CTS Table 4.2.2 and associated Note 2 describe details of the performance of the Functional Test of the Trip System Logic associated with the Main Steam Line Isolation Trip Functions. These details are to be relocated to Bases. These details are not necessary to ensure the operability of the associated Trip System Logic instrumentation. The VYNPS TS definition of Logic System Functional Tests, the requirements of proposed Specification 3.2.B, and the associated Surveillance Requirements (including the requirements to periodically perform Logic System Functional Tests) are adequate to ensure the associated Trip System Logic is maintained operable. As such, these relocated details are not required to be inthe VYNPS TS to provide adequate protection of the public health and safety. Changes to the Bases are controlled by the provisions of 10 CFR 50.59. Not including these details inTS isconsistent with the ISTS.

"Specific" L.1 CTS Tables 3.2.1 and 4.2.1 include requirements for the Low Pressure Coolant Injection System - Low Reactor Pressure Trip Function (PS-2-128A &B)and associated Trip System Logic. The purpose of this Trip Function and associated Trip System Logic is to provide isolation of the shutdown cooling portion of the Residual Heat Removal (RHR)

System to protect that system from overpressurization due to high reactor pressure. This VYNPS 12 Revision 0

SAFETY ASSESSMENT OF CHANGES TS: 3.2.B1/4.2.B - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE L.l isolation provides for equipment protection to prevent an intersystem LOCA scenario. CTS (continued) 3.2.A specifies an Applicability for this Trip Function and associated Trip System Logic of "When the system(s) it initiates or controls is required inaccordance with Specification 3.5.'

Specification 3.5 requires the Low Pressure Coolant InjecUon (LPCI) System to be operable whenever irradiated fuel isin the reactor vessel and prior to reactor startup from cold shutdown. Specification 3.5 also allows ECCS (including LPCI) to be inoperable in Cold Shutdown and Refueling provided no operations with the potential for draining the reactor vessel are inprogress. The 'APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS' column inproposed Table 3.2.2 requires the Residual Heat Removal Shutdown Cooling Isolation Trip Function (i.e., Trip Function 5.a) to be operable inthe Run, Startup/Hot Standby, Hot Shutdown, and Refuel (with reactor coolant water temperature >

212 0F) Modes since these are the only Modes inwhich the reactor can be pressurized such that isolation of the RHR System is required to prevent overpressurization. Therefore, the requirements for the Residual Heat Removal Shutdown Cooling Isolation Trip Function inCold Shutdown or Refueling (with reactor coolant water temperature < 212 0F)are not necessary to protect the RHR System from overpressurization due to high reactor pressure and are deleted. This change is consistent with the ISTS.

L.2 CTS Table 3.2.1 Note 1O.A requires that associated systems be declared inoperable within 1hour of discovery of loss of initiation capability for feature(s) inone division when Low Pressure Coolant Injection System - Low Reactor Pressure Trip Function channels are inoperable. This Note was intended to provide requirements to ensure that a complete loss of function (inthis case, loss of capability to isolate the penetration with at least one isolation valve) does not exist (for more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ) due to more than one instrument channel of an individual Trip Function being inoperable. However, the subject action requirements were written to require the implementation of the more restrictive allowed outage Umes associated with a loss of function (i.e., 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ) even for conditions for which safety function was maintained. As an example, for a Trip Function with two trip systems each providing isolation signals to two isolation valves in a penetration flow path, if the isolation capability associated with one of the two isolation valves is inoperable, the instrumentation inoperability can be such that the automatic isolation of the penetration by the second isolation valve can still be accomplished by the Trip Function. Therefore, a complete loss of function has not occurred and it isnot appropriate to apply the more restrictive loss of function allowed outage Ume of 1hour for this condition. Consistent with the intent of these 'loss of function' requirements, proposed Table 3.2.2 Action Note I .bis revised to require, with isolation capability not maintained, that isolation capability be restored within one hour. This change is acceptable, since if sufficient instrument channels are operable or in trip such that a loss of isolation capability has not occurred, the allowed outage times of proposed Table 3.2.2 Action Note l.a will limit operation inthis condition to within the bounds of the applicable analysis, i.e., NEDC-31677-P-A, 'Technical Specification Improvement Analyses for BWR Isolation Actuation Instrumentation,' July 1990. This change isconsistent with the ISTS.

VYNPS 13 Revision 0

SAFETY ASSESSMENT OF CHANGES TS: 3.2.B1/4.2.B - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE (continued)

L.3 CTS 3.2.B provides requirements for the Primary Containment Isolation Instrumentation Low-Low Reactor Vessel Water Level Trip Function (proposed Table 3.2.2 Trip Function 1.a) and the Primary Containment Isolation Instrumentation Condenser Low Vacuum Trip Function (proposed Table 3.2.2 Trip Function I.f). Inthe event minimum conditions for operation are not satisfied for these trip functions, CTS Table 3.2.2 requires compliance with CTS Table 3.2.2 Note 2.A. CTS Table 3.2.2 Note 2.A requires an orderly load reduction to be Initiated and to the have the reactor in Cold Shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The function of this instrumentation Isto provide main steam line isolation. Therefore, proposed Table 3.2.2 Action Note 2.a will require, inthe same condition, the associated main steam line to be Isolated within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (which satisfies the Isolation function of the inoperable instrumentation) or that the reactor be placed in Hot Shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and inCold Shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (which places the reactor ina Mode inwhich the instrumentation Isnot required to be operable). The total time period specified for exting the Modes of Applicability isconsistent with times to reach the same Modes provided in CTS Table 3.2.2 Note 2.A. The change associated with allowing Isolation of the affected main steam line isconsidered acceptable since manual isolation of the affected main steam line accomplishes the same action of the actuation instrumentation and operation with a main steam line isolated has been analyzed and shown to be acceptable as part of a safety evaluation. Some conditions may affect the isolation logic for only one main steam line. In these cases, it isnot necessary to require a shutdown of the unit; rather, isolation of the affected line returns the system to astatus where it can perform the remainder of its isolation function, and continued operation is allowed (although it may be at a reduced power level) consistent with the plant specific reload analyses which support operation with one main steam line out of service. The change is consistent with the ISTS.

L.4 CTS Tables 3.2.1 and 3.2.2 include requirements for Trip System Logics associated with the primary containment isolation instrumentation Trip Functions. These Trip Systems Logics are considered part of the primary containment Isolation Instrumentation Trip Functions and the requirements for these associated Trip System Logics to be operable are encompassed by the definition of operable. Therefore, the CTS Table 3.2.1 and 3.2.2 listing of Trip System Logics as separate Trip Functions is unnecessary and isdeleted.

With the deletion of separate Trip System Logic Trip Functions, the actions associated with inoperable Trip System Logic (CTS Table 3.2.1 Note 5 and CTS Table 3.2.2 Notes 2.A and 3)will now be governed by the actions for the individual proposed Table 3.2.2 primary containment isolation instrumentation Trip Functions. These proposed Table 3.2.2 Action Notes are less restrictive than the CTS Table 3.2.1 Note 5 and CTS Table 3.2.2 Notes 2.A and 3 actions. However, the proposed actions will ensure, in the event of inoperabilities, that consistent actions are applied to both primary containment isolation instrumentation Trip Functions and their associated Trip System Logics for the same level of degradation.

This change is acceptable, since the allowed outage times of the proposed Table 3.2.2 Action Notes will limit operation to within the bounds of the applicable analysis, i.e., NEDC-31677-P-A, 'Technical Specifications Improvement Analyses for BWR Isolation Actuation WNPS 14 Revision 0

No Significant Hazards Consideration 3.2.B and 4.2.1B Primary Containment Isolation

GENERIC NO SIGNIFICANT HAZARDS CONSIDERATION ADMINISTRATIVE CHANGES

("A.x Labeled Comments/Discussions)

Inaccordance with the criteria set forth in10 CFR 50.92, Vermont Yankee Nuclear Power Station has evaluated this proposed Technical Specifications change and determined it does not represent a significant hazards consideration. The following is provided Insupport of this conclusion.

1. Does the change involve a significant Increase Inthe probability or consequences of an accident previously evaluated?

The proposed change involves reformatting, renumbering, and rewording. The reformatting, renumbering, and rewording process involves no technical changes to the existing Technical Specifications. As such, this change isadministrative innature and does not impact initiators of analyzed events or assumed mitigation of accident or transient events. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes Inmethods governing normal plant operation. The proposed change will not impose any new or eliminate any old requirements. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change Involve a significant reduction ina margin of safety?

The proposed change will not reduce a margin of safety because it has no Impact on any safety analyses assumptions. This change isadministrative innature. Therefore, the change does not involve a significant reduction ina margin of safety.

VYNPS 1 Revision 0

GENERIC NO SIGNIFICANT HAZARDS CONSIDERATION TECHNICAL CHANGES - MORE RESTRICTIVE

("M.x Labeled CommentslDiscussions)

Inaccordance with the criteria set forth in 10 CFR 50.92, Vermont Yankee Nuclear Power Station has evaluated this proposed Technical Specifications change and determined it does not represent asignificant hazards consideration. The following is provided Insupport of this conclusion.

1. Does the change involve a significant increase inthe probability or consequences of an accident previously evaluated?

The proposed change provides more stringent requirements for operation of the facility. These more stringent requirements do not result Inoperation that will Increase the probability of initiating an analyzed event and do not alter assumptions relative to mitigation of an accident or transient event. The more restrictive requirements continue to ensure process variables, structures, systems, and components are maintained consistent with the safety analyses and licensing basis. Therefore, this change does not involve a significant increase inthe probability or consequences of an accident previously evaluated.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not Involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in the methods governing normal plant operation. The proposed change does impose different requirements. However, these changes are consistent with the assumptions inthe safety analyses and licensing basis. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction ina margin of safety?

The imposition of more restrictive requirements either has no Impact on or increases the margin of plant safety. As provided Inthe discussion of the change, each change inthis category is by definition, providing additional restrictions to enhance plant safety. The change maintains requirements within the safety analyses and licensing basis. Therefore, this change does not involve asignificant reduction in a margin of safety.

VYNPS 3 Revision 0

GENERIC NO SIGNIFICANT HAZARDS CONSIDERATION

'GENERIC' LESS RESTRICTIVE CHANGES:

RELOCATING DETAILS TO TECHNICAL SPECIFICATION BASES, UFSAR, PROCEDURES, OR OTHER PLANT CONTROLLED DOCUMENTS

("LA.x' Labeled Comments/Discussionsj Inaccordance with the criteria set forth in 10 CFR 50.92, Vermont Yankee Nuclear Power Station has evaluated this proposed Technical Specifications change and determined Itdoes not represent a significant hazards consideration. The following is provided insupport of this conclusion.

1. Does the change involve a significant increase Inthe probability or consequences of an accident previously evaluated?

The proposed change relocates certain details from the Technical Specifications to the Bases, UFSAR, procedures, or other plant controlled documents. The Bases, UFSAR, procedures, and other plant controlled documents containing the relocated information will be maintained in accordance with 10 CFR 50.59. The UFSAR is subject to the change control provisions of 10 CFR 50.71 (e), and the plant procedures and other plant controlled documents are subject to controls Imposed by plant administrative procedures, which endorse applicable regulations and standards. Since any changes to the relocated details inthe Bases, UFSAR, procedures, or other plant controlled documents will be evaluated per the requirements of 10 CFR 50.59, no significant increase inthe probability or consequences of an accident previously evaluated will be allowed. Therefore, this change does not involve a significant increase inthe probability or consequences of an accident previously evaluated.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be Installed) or a change inthe methods governing normal plant operation. The proposed change will not impose or eliminate any requirements, and adequate control of the information will be maintained. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction ina margin of safety?

The proposed change will not reduce a margin of safety because ithas no impact on any safety analysis assumptions. Inaddition, the details to be transposed from the Technical Specifications to the Bases, UFSAR, procedures, or other plant controlled documents are the same as the existing Technical Specifications. Since any future changes to these details in the Bases, UFSAR, procedures, or other plant controlled documents will be evaluated per the requirements of 10 CFR 50.59, no significant reduction ina margin of safety will be allowed. Based on 10 CFR 50.92, the existing requirement for NRC review and approval of revisions, to these details proposed for relocation, does not have a specific margin of safety upon which to evaluate. However, since the proposed change isconsistent with the BWR Standard Technical Specifications, NUREG-1433, approved by the NRC Staff, revising the Technical Specifications to reflect the approved level of detail ensures no significant reduction inthe margin of safety.

WNPS 4 Revision 0

NO SIGNIFICANT HAZARDS CONSIDERATION TS 3.2.B1/4.2.B - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION L.1 CHANGE Inaccordance with the criteria set forth in10 CFR 50.92, Vermont Yankee Nuclear Power Station has evaluated this proposed Technical Specifications change and determined Itdoes not represent a significant hazards consideration. The following is provided insupport of this conclusion.

1. Does the change involve a significant increase inthe probability or consequences of an accident previously evaluated?

The proposed change deletes the requirements inCold Shutdown or Refueling (with reactor coolant water temperature < 212 0F) for the Residual Heat Removal (RHR) Shutdown Cooling Isolation Trip Functions associated with RHR Shutdown Cooling isolation valves. This change will not result in any hardware or operating procedure changes. The RHR Shutdown Cooling Isolation instrumentation is not assumed to be an initiator of any analyzed event. The role of the Instrumentation Isto protect that RHR System from overpressurization due to high reactor pressure. InCold Shutdown or Refueling (with reactor coolant water temperature < 212 0 F), the RHR Shutdown Cooling isolation instrumentation is not needed to protect that RHR System from overpressurization due to high reactor pressure. Therefore, this proposed change will not involve a significant increase Inthe probability or consequences of an accident previously evaluated.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not necessitate a physical alteration of the plant (no new or different type of equipment will be Installed) or changes inparameters governing normal plant operation.

The proposed change still ensures the affected instrumentation is required to be operable when It is necessary to perform its function. Therefore, it does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction ina margin of safety?

The proposed change deletes the requirements in Cold Shutdown or Refueling (with reactor coolant water temperature < 212 0F)for the Residual Heat Removal (RHR) Shutdown Cooling Isolation Trip Functions associated with RHR Shutdown Cooling isolation valves. Due to the pressure and temperature limitations of Cold Shutdown or Refueling (with reactor coolant water temperature < 2112 0F), the RHR Shutdown Cooling isolation instrumerntation is not needed in these Modes to protect that RHR System from overpressurization due to high reactor pressure. As such, this change does not involve a significant reduction in a margin of safety.

VYNPS 1 Revision 0

NO SIGNIFICANT HAZARDS CONSIDERATION TS 3.2.B1/4.2.B - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION L.2 CHANGE Inaccordance with the criteria set forth in 10 CFR 50.92, Vermont Yankee Nuclear Power Station has evaluated this proposed Technical Specifications change and determined Itdoes not represent a significant hazards consideration. The following is provided in support of this conclusion.

1. Does the change Involve a significant increase inthe probability or consequences of an accident previously evaluated?

This change will extend operation when one or more channels of the affected Primary Containment Isolation Trip Function Instrumentation are Inoperable and a loss of function has not occurred. The affected Primary Containment Isolation Trip Function instrumentation is not considered to be an initiator of any accidents previously analyzed. Therefore, this change does not significantly increase the probability of a previously analyzed accident. Continued operation with inoperable Primary Containment Isolation Trip Function instrumentation channels will continue to be limited inaccordance with Technical Specifications. Since the capability to perform the safety function will still be maintained inthe proposed condition, the consequences of an accident occurring during the time allowed by proposed change are the same as the consequences currentiy allowed. Therefore, this change does not significantly increase the consequences of a previously analyzed accident.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not necessitate a physical alteration of the plant (no new or different type of equipment will be installed) or changes in parameters governing normal plant operation.

The proposed change still ensures that continued operation inthe applicable condition is not allowed when an associated Primary Containment Isolation Trip Function instrumentation channel is not capable of performing its required safety function. Therefore, it does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

This change will extend operation when one or more channels of the affected Primary Containment Isolation Trip Function instrumentation are inoperable and a loss of function has not occurred. No change is being made inthe manner Inwhich systems relied upon inthe safety analyses provide plant protection. The capability to perform the required safety function will still be maintained inthe proposed condition. Plant safety margins continue to be maintained through the limitations established Inthe Technical Specifications. This change does not impact plant equipment design or operation, and there are no changes being made to safety limits or limiting safety system settings. The proposed change does not impact safety. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

VYNPS 2 Revision 0

NO SIGNIFICANT HAZARDS CONSIDERATION TS 3.2.B1/4.2.B - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION L.3 CHANGE Inaccordance with the criteria set forth in 10 CFR 50.92, Vermont Yankee Nuclear Power Station has evaluated this proposed Technical Specifications change and determined It does not represent a significant hazards consideration. The following isprovided insupport of this conclusion.

I1. Does the change involve a significant increase inthe probability or consequences of an accident previously evaluated?

This change will allow continued operation with inoperable channels if the affected main steam line penetration is isolated. Isolated penetrations are not considered as an initiator for any accidents previously analyzed. Therefore, this change does not significantly increase the probability of a previously analyzed accident. Further, Isolating the penetration fulfills the post accident function of the isolation logic. Therefore, this change does not significantly increase the consequences of a previously analyzed accident.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not introduce a new mode of plant operation and does not involve physical modification to the plant (an evaluation has already been performed for the condition of one main steam line isolated). Therefore, it does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction ina margin of safety?

When minimum conditions are not met for the Main Steam Line Isolation Trip Functions associated with Low-Low Reactor Vessel Water Level or Condenser Low Vacuum, the action is revised to allow isolation of the affected main steam line. Manual isolation of the affected main steam line accomplishes the same action of the actuation instrumentation and operation with a main steam line isolated has been analyzed and shown to be acceptable. The proposed change provides the benefit of avoiding a potential shutdown transient when the capability to satisfy the safety function of the affected instrumentation exists. Therefore, this change does not involve a significant reduction in a margin of safety since the required safety function of the inoperable channels will be fulfilled.

VYNPS 3 Revision 0

NO SIGNIFICANT HAZARDS CONSIDERATION TS 3.2.B1/4.2.B - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION L.4 CHANGE Inaccordance with the criteria set forth in 10 CFR 50.92, Vermont Yankee Nuclear Power Station has evaluated this proposed Technical Specifications change and determined it does not represent a significant hazards consideration. The following is provided insupport of this conclusion.

1. Does the change involve a significant increase inthe probability or consequences of an accident previously evaluated?

This change will relax the actions when one or more channels of primary containment isolation Trip Function instrumentation are inoperable due to inoperable Trip System Logic. The primary containment isolation Trip Function instrumentation is not considered to be an initiator of any accidents previously analyzed. Therefore, this change does not significantly increase the probability of a previously analyzed accident. Continued operation with inoperable primary containment isolation Trip Function instrumentation channels will continue to be limited in accordance with Technical Specifications. Since the level of degradation allowed inthe proposed actions are the same as the current actions, the consequences of an accident occurring during the time allowed by proposed change are the same as the consequences currently allowed.

Therefore, this change does not significantly increase the consequences of a previously analyzed accident.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not necessitate a physical alteration of the plant (no new or different type of equipment will be installed) or changes in parameters governing normal plant operation.

The proposed change still ensures that continued operation inthe applicable condition isnot allowed when an associated primary containment isolation Trip Function instrumentation channel isnot capable of performing its required safety function. Therefore, it does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

This change will relax actions when one or more channels of the affected primary containment isolation Trip Function instrumentation are inoperable. No change is being made inthe manner in which systems relied upon inthe safety analyses provide plant protection. Plant safety margins continue to be maintained through the limitations established inthe Technical Specifications. This change does not impact plant equipment design or operation, and there are no changes being made to safety limits or limiting safety system settings. The proposed change does not impact safety. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

VYNPS 4 Revision 0

NO SIGNIFICANT HAZARDS CONSIDERATION TS 3.2.B14.2.B - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION L.5 CHANGE Inaccordance with the criteria set forth in 10 CFR 50.92, Vermont Yankee Nuclear Power Station has evaluated this proposed Technical Specifications change and determined it does not represent a significant hazards consideration. The following isprovided insupport of this conclusion.

I1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

This change allows an isolated primary containment penetration to be opened under administrative controls. Primary containment isolation is not considered an initiator of any previously analyzed accident. Therefore, this change does not significantly increase the probability of such accidents.

Although primary containment isolation is considered inthe mitigation of the consequences of an accident, the administrative controls provide acceptable compensatory actions to assure the penetration is isolated inthe event of an accident. Therefore, the consequences of a previously analyzed event that may occur during the opening of the isolated line are not significantly increased.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

This change provides acceptable compensatory actions following failure of other equipment. The current requirements are based on providing a single active failure proof boundary to compensate for the loss of one of the two active isolation boundaries. This change provides an alternative which meets the original criteria of a single active failure proof boundary and iscapable of returning the system to its original configuration (i.e., configuration which can provide a single active failure proof boundary.) Therefore, this change does not create the possibility of a new or different kind of accident from any previously analyzed accident.

3. Does this change involve a significant reduction ina margin of safety?

The margin of safety considered indetermining the required compensatory action is based on providing the single active failure proof boundary. Opening of primary containment penetrations on an intermittent basis is required for performing surveillances, repairs, routine evolutions, etc. which minimizes the possibility of a transient due to a required plant shutdown. The administrative controls provide acompensatory boundary inthis condition. Since the compensatory boundary is capable of essentially meeting the same criteria of a single active failure proof boundary, the change does not involve a significant reduction inthe margin of safety.

VYNPS 5 Revision 0

References 3.2.B and 4.2.B Primary Containment Isolation

3.2.B14.2.B REFERENCES Primary Containment Isolation

1. Technical Requirements Manual.

2. UFSAR, Chapter 14.

3. UFSAR, Table 6.5.3.

4. UFSAR, Section 14.6.5.

5. UFSAR, Section 14.5.4.1.

6. NEDC-31677P-A, Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation," July 1990.

7. NEDC-30851 P-A Supplement 2, 'Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation,"

March 1989.

Proposed Technical Specifications 3.2.C and 4.2.C Reactor Building Ventilation Isolation and Standby Gas Treatment System Initiation

VYNPS 3.2 LIMITING CONDITIONS FOR 4.2 SURVEILLANCE REQUIREMENTS OPERATION 3.2 PROTECTIVE INSTRUMENT SYSTEMS 4.2 PROTECTIVE INSTRUMENT SYSTEMS C. Reactor Building Ventilation C. Reactor Building Ventilation Isolation and Standby Gas Isolation and Standby Gas Treatment System Initiation Treatment System Initiation The reactor building 1. The reactor building ventilation isolation and ventilation isolation and Standby Gas Treatment System Standby Gas Treatment initiation instrumentation System initiation for each Trip Function in instrumentation shall be Table 3.2.3 shall be operable checked, functionally in accordance with tested and calibrated as Table 3.2.3. indicated in Table 4.2.3.

When a channel is placed in an inoperable status solely for performance of required instrumentation surveillances, entry into the associated Limiting Conditions for Operation and required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Trip Function maintains reactor building ventilation isolation capability and Standby Gas Treatment System initiation capability.

2. Perform a Logic System Functional Test of reactor building ventilation isolation and Standby Gas Treatment System initiation A4nstrumentation Trip Functions once every Operating Cycle.

Amendment No. so

VYNPS Table 3.2.3 (page 1 of 1)

System Reactor Building Ventilation Isolation and Standby Gas Treatment Initiation Instrumentation ACTIONS WHEN REQUIRED REQUIRED APPLICABLE MODES OR CHANNELS CHANNELS OTHER SPECIFIED PER TRIP ARE CONDITIONS SYSTEM INOPERABLE TRIP SETTING TRIP FUNCTION RUN, STARTUP/HOT 2 Note 1 2 127.0 inches

1. Low Reactor Vessel Water STANDBY, HOT SHUTDOWN, Level REFUELS, (2)

RUN, STARTUP/HOT 2 Note 1 5 2.5 psig

2. High Drywell Pressure STANDBY, HOT SHUTDOWN, REFUELU1 RUN, STARTUP/HOT 1 Note 1 S 14 mR/hr

3. High Reactor Building STANDBY, HOT SHUTDOWN, Ventilation REFUEL"1 ', (2), (3),

Radiation (4)

RUN, STARTUP/HOT 1 Note 1 5 100 mR/hr

4. High Refueling Floor Zone STANDBY, HOT SHUTDOWN, Radiation REFUEL"', (2), (3),

(4)

(1) With reactor coolant temperature > 212 OF.

vessel.

(2) During operations with potential for draining the reactor cask in secondary (3) During movement of irradiated fuel assemblies or fuel containment.

(4) During Alteration of the Reactor Core.

51 Amendment No.

VYNPS Table 3.2.3 ACTION Note

1. With one or more required Reactor Building Ventilation Isolation and Standby Gas Treatment System Initiation Instrumentation channels inoperable, take all of the applicable Actions in Notes l.a and l.b below.

a. With one or more Trip Functions with one or more required channels inoperable:

1) For Trip Functions 1 and 2, place inoperable channel in trip within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; and

2) For Trip Functions 3 and 4, place inoperable channel in trip within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. With one or more Trip Functions with isolation or initiation capability not maintained:

1) Restore isolation and initiation capability within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

If any applicable Action and associated completion time of Note l.a or l.b is not met, isolate the Reactor Building Ventilation System and place the Standby Gas Treatment System in operation within I hour.

Amendment No. 52

VYNPS Table 4.2.3 (page 1 of 1)

Reactor Building Ventilation Isolation and Standby Gas Treatment System Initiation Instrumentation Tests and Frequencies TRIP FUNCTION CHECK FUNCTIONAL TEST CALIBRATION

1. Low Reactor Vessel NA Every 3 Months Every 3 Monthsti),

Water Level Once/Operating Cycle

2. High Drywell Pressure NA Every 3 Months Every 3 Montholl, once/operating Cycle

3. High Reactor Building Once/Day Every 3 Months Every 3 Months Ventilation Radiation

4. High Refueling Floor Once/Day Every 3 Months Every 3 Months Zone Radiation During REFUELING (1) Trip unit calibration only.

53 Amendment No.

Proposed Bases 3.2.C and 4.2.C Reactor Building Ventilation Isolation and Standby Gas Treatment System Initiation

VYNPS BASES: 3.2.C/4.2.C REACTOR BUILDING VENTILATION ISOLATION AND STANDBY GAS TREATMENT SYSTEM INITIATION BACKGROUND The reactor building ventilation isolation and Standby Gas Treatment System initiation instrumentation automatically initiates closure of the Reactor Building Automatic Ventilation System Isolation Valves (RBAVSIVs) and starts the Standby Gas Treatment (SGT) System. The function of these components and systems, in combination with other accident mitigation systems, is to limit fission product release during and following postulated Design Basis Accidents (DBAs) (Ref. 1). Reactor Building (i.e., secondary containment) isolation and establishment of vacuum with the SGT System ensures that fission products that leak from primary containment following a DBA, or are released outside primary containment, or are released during certain operations when primary containment is not required to be operable, are maintained within applicable limits.

The isolation instrumentation includes the sensors, relays, and switches that are necessary to cause initiation of reactor building ventilation isolation and Standby Gas Treatment System operation. Most channels include electronic equipment (e.g., trip units) that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs a reactor building ventilation isolation and Standby Gas Treatment System initiation signal to the isolation and initation logic. Functional diversity is provided by monitoring a wide range of independent parameters. The input parameters to the isolation and initiation logic are (1) reactor vessel water level, (2) drywell pressure, (3) reactor building ventilation radiation, and (4) refueling floor zone radiation.

Redundant sensor input signals from each parameter are provided for initiation and isolation.

For both the Low Reactor Vessel Water Level and High Drywell Pressure Trip Functions, the reactor building ventilation isolation and Standby Gas Treatment System initiation logic receives input from four channels. The outputs of the channels are arranged in one-out-of-two taken twice logics.

For the High Reactor Building Ventilation Radiation and High Refueling Floor Zone Radiation Trip Functions, two radiation detectors and monitors are provided for each Trip Function. Each channel includes a radiation detector and associated monitor. The outputs of the channels are arranged in a one-out-of-two logic. In addition, the outputs of each channel are provided to both Trip Systems A and B. As such, any High Reactor Building Ventilation Radiation or High Refueling Floor Zone Radiation Trip Function channel will initiate reactor building ventilation isolation and Standby Gas Treatment System operation. (For the purposes of the Technical Specifications, the A radiation detectors and monitors should be considered to be associated with the Trip System A and the B radiation detectors and monitors should be considered to be associated with Trip System B.) Trip System A initiates startup of SGT subsystem A and initiates isolation of the reactor building supply and exhaust outboard isolation valves. Trip System B initiates startup of SGT subsystem B and initiates isolation of the reactor building supply and exhaust inboard isolation valves. As such, either Trip System isolates the secondary containment and provides the necessary filtration of fission products.

APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY The isolation and initiation signals generated by the reactor building ventilation isolation and Standby Gas Treatment System initiation Amendment No. 760

VYNPS BASES: 3.2.C/4.2.C REACTOR BUILDING VENTILATION ISOLATION AND STANDBY GAS TREATMENT SYSTEM INITIATION APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) instrumentation are implicitly assumed in the safety analyses of References 2, 3, and 4, to initiate closure of the RBAVSIVs and start the SGT System to limit offsite doses.

Reactor building ventilation isolation and Standby Gas Treatment System initiation instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

The operability of the reactor building ventilation isolation and Standby Gas Treatment System initiation instrumentation is dependent on the operability of the individual instrumentation channel Trip Functions specified in Table 3.2.3.

Each Trip Function must have the required number of operable channels in each trip system, with their trip setpoints within the calculational as-found tolerances specified in plant procedures. Operation with actual trip setpoints within calculational as-found tolerances provides reasonable assurance that, under worst case design basis conditions, the associated trip will occur within the Trip Settings specified in Table 3.2.3. As a result, a channel is considered inoperable if its actual trip setpoint is not within the calculational as-found tolerances specified in plant procedures. The actual trip setpoint is calibrated consistent with applicable setpoint methodology assumptions.

In general, the individual Trip Functions are required to be OPERABLE in RUN, STARTUP/HOT STANDBY, HOT SHUTDOWN, REFUEL (with reactor coolant temperature >

212 0 F), during operations with the potential for draining the reactor vessel (OPDRVs), during movement of irradiated fuel assemblies or fuel cask in secondary containment, and during Alteration of the Reactor Core; consistent with the Applicability for the SGT System and secondary containment requirements in Specifications 3.7.B and 3.7.C. Trip Functions that have different Applicabilities are discussed below in the individual Trip Functions discussion.

The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Trip Function by Trip Function basis.

1. Low Reactor Vessel Water Level Low reactor pressure vessel (RPV) water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. An isolation of the secondary containment and actuation of the SGT System are initiated in order to minimize the potential of an offsite release. The Low Reactor Vessel Water Level Trip Function is one of the Trip Functions assumed to be operable and capable of providing isolation and initiation signals. The isolation and initiation of systems on Low Reactor Vessel Water Level support actions to ensure that any offsite releases are within the limits calculated in the safety analysis.

Low Reactor Vessel Water Level signals are initiated from level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Low Reactor Vessel Water Level Trip Function are available and are required to be operable to ensure that no single instrument failure can preclude the isolation and initiation function.

Amendment No. 76p

VYNPS BASES: 3.2.C/4.2.C REACTOR BUILDING VENTILATION ISOLATION AND STANDBY GAS TREATMENT SYSTEM INITIATION APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

The Low Reactor Vessel Water Level Trip Setting was chosen to be the same as the Reactor Protection System (RPS) Low Reactor Vessel Water Level Trip Setting (Specification 3.1.A), since this could indicate that the capability to cool the fuel is being threatened. The Trip Setting is referenced from the top of enriched fuel.

in The Low Reactor Vessel Water Level Trip Function is required to be operable RUN, STARTUP/HOT STANDBY, HOT SHUTDOWN, REFUEL (with reactor coolant temperature > 212'F) where considerable energy exists in the Reactor Coolant System (RCS); thus, there is a possibility of pipe breaks resulting in significant releases of radioactive steam and gas. In COLD SHUTDOWN and REFUELING (with reactor coolant temperature < 212'F), the probability and consequences of these events are low due to the RCS pressure and temperature In limitations of these Modes; thus, this Trip Function is not required.

addition, the Trip Function is also required to be operable during OPDRVs to ensure that offsite dose limits are not exceeded if core damage occurs.

2. High Drywell Pressure High drywell pressure can indicate a break in the reactor coolant pressure of boundary (RCPB). An isolation of the secondary containment and actuation the SGT System are initiated in order to minimize the potential of an offsite release. The isolation and initiation of systems on High Drywell Pressure supports actions to ensure that any offsite releases are within the limits calculated in the safety analysis.

High drywell pressure signals are initiated from pressure transmitters that Trip sense the pressure in the drywell. Four channels of High Drywell Pressure are required to be operable to ensure that no single Function are available and instrument failure can preclude performance of the isolation and initiation function.

The Trip Setting was chosen to be the same as the RPS High Drywell Pressure Trip Setting (Specification 3.1.A) since this is indicative of a loss of coolant accident (LOCA).

The High Drywell Pressure Trip Function is required to be operable in RUN, STARTUP/HOT STANDBY, HOT SHUTDOWN, REFUEL (with reactor coolant temperature 212 F) where considerable energy exists in the RCS; thus, there 0 is a possibility gas.

of pipe breaks resulting in significant releases of radioactive steam and COLD SHUTDOWN and REFUELING (with reactor This Trip Function is not required in coolant temperature < 212'F) because the probability and consequences of these these events are low due to the RCS pressure and temperature limitations of Modes.

3, 4. High Reactor Building Ventilation Radiation and High Refueling Floor Zone Radiation High reactor building ventilation radiation or refuel floor zone radiation is an indication of possible gross failure of the fuel cladding. The release may 76q AmenAment No.

AllE

VYNPS GAS BASES: 3.2.C/4.2.C REACTOR BUILDING VENTILATION ISOLATION AND STANDBY TREATMENT SYSTEM INITIATION APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

RCPB or the have originated from the primary containment due to a break in the accident. When High Reactor Building refueling floor due to a fuel handling Refueling Floor Zone Radiation is detected, Ventilation Radiation or High are initiated secondary containment isolation and actuation of the SGT System actions to limit the release of fission products as assumed in the to support UFSAR safety analyses (Ref. 4).

Floor Zone The High Reactor Building Ventilation Radiation and High Refueling from radiation detectors that are located on Radiation signals are initiated duct coming from the reactor building and the refueling the ventilation exhaust floor zones, respectively. Two channels of High Reactor Building Ventilation Radiation Trip Radiation Trip Function and two channels of High Refueling Floor operable to ensure that no single Function are available and are required to be and initiation function.

instrument failure can preclude the isolation of the fuel The Trip Settings are chosen to promptly detect gross failure cladding.

Floor Zone The High Reactor Building Ventilation Radiation and High Refueling to be operable in RUN, STARTUP/HOT Radiation Trip Functions are required 0

> 212 F)where STANDBY, HOT SHUTDOWN, REFUEL (with reactor coolant temperature of pipe considerable energy exists in the RCS; thus, there is a possibility of radioactive steam and gas. In COLD breaks resulting in significant releases 0 c 212 F), the SHUTDOWN and REFUELING (with reactor coolant temperature RCS pressure probability and consequences of these events are low due to the of these Modes; thus, these Trip Functions are not and temperature limitations In addition, the Trip Functions are also required to be operable required.

fuel cask in during OPDRVs, during movement of irradiated fuel assemblies or the Reactor Core, because the secondary containment, and during Alteration of due to fuel failures (due to the capability of detecting radiation releases must be provided to ensure that fuel uncovery or dropped fuel assemblies) offsite dose limits are not exceeded.

ACTIONS Table 3.2.3 ACTION Note 1 signals and Because of the diversity of sensors available to provide isolation out of service time of the redundancy of the isolation design, an allowable Function (12 hours for those Trip 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months depending on the Trip common to RPS instrumentation, i.e.,

Functions that have channel components that do not have Trip Functions 1 and 2, and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for those Trip Functions common to RPS instrumentation, i.e., all other Trip channel components has been shown to be acceptable (Refs. S and 6) to permit Functions),

restoration of any inoperable channel to operable status. This out of service associated Trip Function is still time is only acceptable provided the (refer to next paragraph). If the inoperable maintaining isolation capability to operable status within the allowable out of channel cannot be restored the channel must be placed in the tripped condition per Table service time, Placing the inoperable channel in 3.2.3 Note l.a.1) or l.a.2), as applicable.

restore capability trip would conservatively compensate for the inoperability, Alternately, to accommodate a single failure, and allow operation to continue.

trj

%mrLTm-nF 76r AmenZ1LU11=L%-l Wou.

VYNPS BASES: 3.2.C/4.2.C REACTOR BUILDING VENTILATION ISOLATION AND STANDBY GAS TREATMENT SYSTEM INITIATION ACTIONS (continued) where if it is not desired to place the channel in trip (e.g., as in the case placing the inoperable channel in trip would result in an isolation or and the initiation), the Reactor Building Ventilation System must be isolated one hour. Isolating the SGT System must be placed in operation within the next Reactor Building Ventilation System and placing the SGT System in operation to performs the intended function of the instrumentation and allows operation continue.

are taken Table 3.2.3 Note l.b is intended to ensure that appropriate actions if multiple, inoperable, untripped channels within the same Trip Function result in a complete loss of isolation capability for the associated for the penetration flow path(s) or a complete loss of initiation capability is considered to be maintaining isolation and SGT System. A Trip Function when sufficient channels are operable or in trip in both initiation capability the given Trip trip systems, such that a trip signal will be generated from RBAVSIVs in Function on a valid signal. This ensures that isolation of the two the operation of the SGT System can be the associated penetration flow path and initiation signal from the given Trip Function.

initiated on an isolation and to have one For the Trip Functions 1 and 2, this would require each trip system require channel operable or in trip. For Trip Functions 3 and 4, this would Time is one channel to be operable or in trip. The one hour Completion intended to allow the operator time to evaluate and repair any discovered it minimizes inoperabilities. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time is acceptable because risk while allowing time for restoration or tripping of channels.

ACTION If any applicable Action and associated Completion Time of Table 3.2.3 secondary containment Note l.a or l.b are not met, the ability to isolate the must be and start the SGT System cannot be ensured. Therefore, further actions the secondary containment isolation performed to ensure the ability to maintain flow and SGT System initiation function. Isolating the associated penetration path(s) and starting the associated SGT System within the next one hour to performs the intended function of the instrumentation and allows operation personnel to establish continue. One hour is sufficient for plant operations required plant conditions or to declare the associated components inoperable without unnecessarily challenging plant systems.

SURVEILLANCE REQUIREMENTS Surveillance Requirement 4.2.C.1 As indicated in Surveillance Requirement 4.2.C.1, reactor building ventilation be isolation and Standby Gas Treatment System initiation instrumentation shall in Table 4.2.3. Table checked, functionally tested and calibrated as indicated 4.2.3 identifies, for each Trip Function, the applicable Surveillance Requirements.

Surveillance Requirement 4.2.C.1 also indicates that when a channel is placed in an inoperable status solely for performance of required instrumentation Surveillances, entry into associated LCO and required Actions may be delayed and for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Trip Function maintains isolation of the Surveillance, or expiration of initiation capability. Upon completion or the the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance, the channel must be returned to operable status XmT' 76s

.Amitna1MCIenLx N. %

VYNPS BASES: 3.2.C/4.2.C REACTOR BUILDING VENTILATION ISOLATION AND STANDBY GAS TREATMENT SYSTEM INITIATION SURVEILLANCE REQUIREMENTS (continued) applicable LCO entered and required Actions taken. This allowance is based on the reliability analysis (Refs. 5 and 6) assumption of the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months testing allowance does not significantly reduce the probability that the RBAVSIVs will isolate the penetration flow path(s) and that the SGT System will initiate when necessary.

Surveillance Requirement 4.2.C.2 The Logic System Functional Test demonstrates the operability of the required initiation logic and simulated automatic operation for a specific channel. The testing required by the SGT System and RBAVSIVs Technical Specifications overlaps this Surveillance to provide testing of the assumed safety function.

The Frequency of "once every Operating Cycle" is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has demonstrated that these components will usually pass the Surveillance when performed at the specified Frequency.

Table 4.2.3, Check Performance of an Instrument Check once per day, for Trip Function 3, and once per day during REFUELING, for Trip Function 4, ensures that a gross failure of instrumentation has not occurred. An Instrument Check is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. An Instrument Check will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each Calibration. Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit. The Frequency is based upon operating experience that demonstrates channel failure is rare. The Instrument Check supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.

Table 4.2.3, Functional Test For Trip Functions 1, 2, 3, and 4, a Functional Test is performed on each required channel to ensure that the channel will perform the intended function.

Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of *Every 3 Months" is based on the reliability analysis of References 5 and 6.

Amendment No. 76t

VYNPS BASES: 3.2.C/4.2.C REACTOR BUILDING VENTILATION ISOLATION AND STANDBY GAS TREATMENT SYSTEM INITIATION SURVEILLANCE REQUIREMENTS (continued)

Table 4.2.3, Calibration For Trip Functions 1, 2, 3, and 4, an Instrument Calibration is a complete check of the instrument loop and the sensor. This test verifies that the channel responds to the measured parameter within the necessary range and accuracy. An Instrument Calibration leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. The specified Instrument Calibration Frequencies are based upon the time interval assumptions for calibration used in the determination of the magnitude of equipment drift in the associated setpoint analyses.

For Trip Functions 1 and 2, a calibration of the trip units is required (Footnote (1)) once every 3 months. Calibration of the trip units provides a check of the actual setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the calculational as-found tolerances specified in plant procedures. The Frequency of every 3 months is based on the reliability analysis of Reference 6 and the time interval assumption for trip unit calibration used in the associated setpoint calculation.

REFERENCES

1. UFSAR, Section 5.3.

2. UFSAR, Section 7.17.2.

3. UFSAR, Section 14.6.3.6.

4. UFSAR, Section 14.6.4.4.

5. NEDC-31677P-A, "Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation," July 1990.

6. NEDC-30851P-A Supplement 2, "Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation,"

March 1989.

Amendment No. 76u

Current Technical Specifications Markups 3.2.C and 4.2.C Reactor Building Ventilation Isolation and Standby Gas Treatment System Initiation

V`YPS 3.2 LIMITING CONDITIONS FOR 4.2 SURVEILLANCE REQUIREMENTS OPERATION 3.2 PROTECTIVE INSTRUMENT SYSTEMS 4.2 PROTECTIVE INSTRUMENT SYSTEMS APplicability: Applicability:

Applies to the operational status Applies to the surveillance of the plant instrumentation requirements of the \

systems which initiate and instrumentation systems which control a protective function. initiate and control a protective function.

Objective: Objective:

To assure the operability of To verify the operability of protective instrumentation protective instrumentation systems. systems.

Specification: Specification:

A. Emergency Core Cooling System A. Emergency Core Cooling System When the system(s) it initiates Instrumentation and looic or controls is required in systems shall be functionally accordance with Specification tested and calibrated as l 3.5, the instrumentation which initiates the emergency core indicated in Table 4.2.1.

'J cooling system(s)shall be operable in accordance with Table 3.2.1.

. Primary Containment Isolation B. Primary Containment Isolation When primary containment Instrumentation and logic integrity is required, in systems shall be functionally accordance with tested and calibrated as Specification 3.7, the indicated in Table 4.2.2.

instrumentation that initiates primary containment isolation shall be operable in accordance with/ s ft es..

Table 3.2.2.

a-ciV i

C. Reactor Building Ventilation Reacto Bukl~dina Ventilation Isolation and Standby Gas Isolation and Standby Gas Treatment System Initiation Treatment System Initiation tA actuation of the standby gas i

I Ze El=t~~sus t allb operable in accordance with Table 3.2.3.

Ame n N uents v.J16'4 ;S I4o Sf t p'4'E 'La4- 4

/~r vv .AlL t e JF 4

%,;e ag~ - r~ 3.1,3 /

i Amendment No. 164

El VYNPS TABLE 3.2.3 E Requir d ACTI9 WheiiN Mi For Co ition ratio I

System Trip Function Trip Setting 1re Not Satisfied

.1 I.

-4, I

Amendment No. 444, 186 49

VYNPS TABLE 3.2.3 /4OTES I.f the minimum number of operable instrument channels is not available in either trip system, the reactor building ventilation system shall be isolate W~k I and the standby gas treatment system operated NxcEIv e insAr.menta Q5Ahen a channel, e'dor teaaectei primar contimn 1 1 U is placed in an inoperable status solely for performance of required 4 instrumentation surveillances, entry into associated Limiting Conditions for operation and required ACTIONS may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Trip Function maintains Reactor Building Ventilation Isolation and tandb Gas Treatment System Initiationr enever Reactor Building Ventilation Isolation and Standby Gas TrE System Initiation are required by Specification 3-7.B-and 3.7.C theesha twoer r oreach re F orin ce provided for be ow: - A2 L With one or more automatic functions with isolation capability not maintained restore isolation and initiation capability in 1 hourA',

bye heACICNreur y Table s. . . --- . > ;f p... 09 With one or more channels inoperable, place the inoperable channefris) so..* in the tripped condition within:

1) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for trip functions common to RPS instrumentation, and

2) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months foritrip functions not common to RPS instrumentation, or, initiate the ACTION required by Table 3.2.

.,A .11 Amendment No. 186 49a

VYNPS TABLE 4.2.3

( ~TESJ&ND FREQUENCIES REACTOR BUILDING VENTILATION AND STANDBY GAS TREATMENT SYSTEM ISOLATION Trip Function Functional Test h B Calibration - Instrument Check Low Reactor Vessel W ater Every Three Months Cycle --

Level

2. High Drywell Pressuree Every Three Months- Cyce 3 Reactor Building Ven Every Three Months Every Three Months Once Each Day

. xhaust Radiation 4 4Reueling Floor Zone Every Three Months Every Three Months Once Each Day During Radiation Refueling Reactor Building Ven-t Once/Operating Cycle O Ope ing le --

L Trip System Logic I :TA A.2.Ca.. Standby Gas Treatmen t Once/Operating Cycle Trip System Logic

_,, LIE Amendment No. &8, 44, 404, 414, 186 67

VYNPS TABLE 4.2 NOTES

62. During each refueling outage, simulated automatic actuation which opens all pilot valves shall be performed such that each trip system logic can e verified independent of its redundant counterpart.

( 3. Trip ~y1em logic calibratio; ihall include only t e delay relays a.10 p system.

t tim1nsnecessary for proErfunctioning of the C. This instrumentation is excepted from functional test definition.

functional test will consist of injecting a simulated electrical signal The A 7jl into the measurement channel.

t5. g eted

6. /De et l/ De egg B. F tests and calibrat ns are not required w n systems are not req ed to be operable.

9. The thermocouples associated with safety/relief valves and safety valve position, that may be used for back-up position indication, shall be verified to be operable every operating cycle.

10. Separate functional tests are not required for this instrumentation. The calibration and integrated ECCS tests which are performed once per operating cycle will adequately demonstrate proper equipment operation.

11. Trip system logic functional tests will include verification of operation of all automatic initiation inhibit switches by monitoring relay contact movement. Verification that the manual inhibit switches prevent opening all relief valves will be accomplished in conjunction with Section 4.5.F.1.

12. Trip system logic testing is not applicable to this function. If the not met, I

required surveillance frequency (every Refueling Outage)Dr~ais -4t rin firn-t"inni

- . . - - - -- I -- -f -I--

.L. "-Am functional testing or the shall be initiated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after the reactor mode switch is placed in Shutdown for the purpcise of commencing a scheduled Refueling Outage.

13. Includes calibration of t,he RBM Reference Downscale function (i.e., RBM upscale function is not bBypassed when >30t Rated Thermal Power).

AJ Amendment No. 46, 9s, b-e&, 45, a64, a--, 24, 212 74

Safety Assessment Discussion of Changes 3.2.C and 4.2.C Reactor Building Ventilation Isolation and Standby Gas Treatment System Initiation

SAFETY ASSESSMENT OF CHANGES TS: 3.2.C/4.2.C - REACTOR BUILDING VENTILATION ISOLATION AND STANDBY GAS TREATMENT SYSTEM INITIATION INSTRUMENTATION ADMINISTRATIVE A.1 Inthe revision of the Vermont Yankee Nuclear Power Station (VYNPS) current Technical Specifications (CTS), certain wording preferences or conventions are adopted which do not result in technical changes (either actual or interpretational). Editorial changes, reformatting, and revised numbering are adopted to make the VYNPS Technical Specifications (TS) more consistent with human factor principles used inthe Boiling Water Reactor Improved Standard Technical Specifications (ISTS), NUREG-1433, Rev. 2. These format and presentation changes are being made to improve usability and clarity. The changes are considered administrative.

A.2 CTS 4.2.C specifies that instrumentation and logic systems shall be functionally tested and calibrated as indicated inTable 4.2.3. Inproposed Surveillance Requirement (SR) 4.2.C.1, the reference to 'and logic system,' isdeleted since associated logic systems are considered part of the reactor building ventilation isolation and Standby Gas Treatment System initiation instrumentation Trip Functions inboth proposed and CTS Tables 3.2.3 and 4.2.3. It is not necessary to explicitly identify logic systems inCTS 4.2.C, since proposed SR 4.2.C.2 (CTS Table 4.2.3 requirements to perform Functional Tests of Trip System Logic) continues to require performance of surveillance testing of Trip System Logic (i.e., performance of Logic System Functional Tests for each reactor building ventilation isolation and Standby Gas Treatment System initiation instrumentation Trip Function). Therefore, this change isconsidered administrative.

A.3 CTS 4.2.C includes reference to CTS Table 4.2.3 for functional test and calibration requirements for reactor building ventilation isolation and Standby Gas Treatment System initiation instrumentation. CTS 4.2.C is revised, inproposed SR 4.2.C.1, to also include reference to check requirements consistent with CTS Table 4.2.3. This change is a presentation preference and does not alter the current requirements to periodically perform checks of certain reactor building ventilation isolation and Standby Gas Treatment System initiation instrumentation trip functions. Therefore, this change is considered administrative innature.

A.4 CTS Table 3.2.3 Note 2 provides allowances to delay entry into actions for 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for the situation of a channel inoperable solely for performance of surveillances. These allowances are moved to proposed SR 4.2.C.1. This change does not involve a technical change, but isonly a difference of presentation preference. Therefore, this change is considered administrative.

A.5 Note 3 to CTS Table 3.2.3 provides actions when the minimum number of channels per trip system requirement isnot met. These requirements are identified in a separate column in proposed Table 3.2.3 titled, 'ACTIONS WHEN REQUIRED CHANNELS ARE INOPERABLE.' This change is a presentation preference and does not alter the current action requirements when required reactor building ventilation isolation and Standby Gas Treatment System initiation instrumentation channels are inoperable. Therefore, this change is considered administrative in nature.

VYNPS 1 Revision 0

SAFETY ASSESSMENT OF CHANGES TS: 3.2.C14.2.C - REACTOR BUILDING VENTILATION ISOLATION AND STANDBY GAS TREATMENT SYSTEM INITIATION INSTRUMENTATION ADMINISTRATIVE (continued)

A.6 CTS Table 3.2.3 includes a Reactor Building Vent Trip Function and a Refueling Floor Zone Radiation Trip Function. The purpose of these trip functions is to provide isolation of the reactor building and initiation of the Standby Gas Treatment System on receipt of a valid high radiation signal. As a result, the name of these trip functions are revised to 'High Reactor Building Ventilation Radiation" and 'High Refueling Floor Zone Radiation,"

respectively, inproposed Table 4.2.3 to more accurately reflect their functions. The design and operation of the actual instrumentation is unchanged. Therefore, this change is considered administrative.

A.7 CTS Table 3.2.3 Note 3 provides actions to be taken when one or more reactor building ventilation isolation and Standby Gas Treatment System initiation instrumentation channels are inoperable. Inthe event these actions are not completed within the specified time period, CTS Table 3.2.3, Notes 3.A and 3.B (as applicable) require the action required by Table 3.2.3 to be taken. For each of the reactor building ventilation isolation and Standby Gas Treatment System initiation instrumentation trip functions, CTS Table 3.2.3 specifies that CTS Table 3.2.3 Note 1is required. Therefore, as a human factors improvement, the actions of CTS Table 3.2.3 Notes 1 and 3 have been combined into one action (proposed Table 3.2.3 Action Note 1). Therefore, this change does not involve a technical change, but is only a difference of presentation preference and is considered administrative.

A.8 CTS Table 3.2.3 Note 3 specifies an Applicability for reactor building ventilation isolation and Standby Gas Treatment System initiation instrumentation of 'When Reactor Building Ventilation Isolation and Standby Gas Treatment System Initiation are required by Specification 3.7.B and 3.7.C.' Specifications 3.7.B and 3.7.C include the requirements for the Standby Gas Treatment System and the Secondary Containment System (which requires reactor building isolation). This change provides an explicit Applicability, in proposed Table 3.2.3 for each reactor building ventilation isolation and Standby Gas Treatment System initiation instrumentation trip function. The specified Applicabilities, in proposed Table 3.2.3, are consistent with the Modes and conditions when the Standby Gas Treatment System and the Secondary Containment System are required to be operable by Specifications 3.7.B and 3.7.C, respectively, except as provided and justified inchanges L.2 and L.3 below. Therefore, this change does not involve a technical change, but is only a difference of presentation preference and isconsidered administrative. The change, providing explicit Mode or conditions of Applicability for each trip function, isconsistent with the ISTS.

W VNPS 2 Revision 0

SAFETY ASSESSMENT OF CHANGES TS: 3.2.C/4.2.C - REACTOR BUILDING VENTILATION ISOLATION AND STANDBY GAS TREATMENT SYSTEM INITIATION INSTRUMENTATION ADMINISTRATIVE (continued)

A.9 CTS Table 4.2 Note 8 states that functional tests and calibrations are not required when systems are not required to operable. The requirements of this Note are duplicated inthe CTS definition 1.0.Z, Surveillance Interval, which states that these tests unless otherwise stated inthese specifications may be waived when the instrument, component, or system is not required to be operable, but that these tests shall be performed on the instrument, component, or system prior to being required to be operable. Therefore, CTS Table 4.2 Note 8 is unnecessary and its deletion isconsidered to be administrative. The change is consistent with the ISTS.

A.10 For the Trip System Logic associated with the reactor building ventilation and the Standby Gas Treatment System instrumentation, CTS Table 4.2.3 includes requirements to perform a calibration of Trip System Logics once per Operating Cycle. These requirements are modified by Table 4.2 Note 3. Note 3 states, MTrip system logic calibration shall include only time delay relays and timers necessary for proper functioning of the trip system.' The reactor building ventilation isolation and the Standby Gas Treatment System initiation Trip Functions of CTS Table 4.2.3 do not include any time delay relays or timers necessary for proper functioning of the trip systems. Therefore, this Note isdeleted and, in proposed Table 4.2.3, the reactor building ventilation isolation and the Standby Gas Treatment System initiation instrumentation (proposed Trip Functions 1 through 4) do not include calibration requirements for time delay relays or timers. As a result, this change does not involve a technical change, but isonly adifference of presentation preference and is considered administrative.

A.1 1 CTS Table 4.2 Notes 2, 10, and 11 provide requirements that apply to ECCS instrumentation. The ECCS instrumentation islocated in proposed Specifications 3.2.A and 4.2.A. Therefore, the requirements of CTS Table 4.2 Notes 2, 10, and 11 are physically moved and addressed inthe changes for proposed Specifications 3.2.A and 4.2.A.

Therefore, this change does not involve a technical change, but is only a difference of presentation preference and isconsidered administrative. CTS Table 4.2 Notes 4, 12, and 13 provides requirements that apply to control rod block instrumentation. The control rod block instrumentation islocated Inproposed Specifications 3.2.E and 4.2.E. Therefore, the requirements of CTS Table 4.2 Notes 4,12, and 13 are physically moved and addressed in the changes to proposed Specifications 3.2.E and 4.2.E. Therefore, this change does not involve a technical change, but is only a difference of presentation preference and is considered administrative. CTS Table 4.2 Note 9 provides requirements that apply to post-accident monitoring instrumentation. The post-accident monitoring instrumentation is located in proposed Specifications 3.2.G and 4.2.G. Therefore, the requirements of CTS Table 4.2 Note 9 are physically moved and changes addressed inproposed Specifications 3.2.G and 4.2.G. Therefore, this change does not involve a technical change, but is only a difference of presentation preference and is considered administrative.

VYNPS 3 Revision 0

SAFETY ASSESSMENT OF CHANGES TS: 3.2.C/4.2.C - REACTOR BUILDING VENTILATION ISOLATION AND STANDBY GAS TREATMENT SYSTEM INITIATION INSTRUMENTATION TECHNICAL CHANGES - MORE RESTRICTIVE M.1 CTS Table 3.2.3 Note I provides actions to isolate the Reactor Building Ventilation System and operate the Standby Gas Treatment System but does not state the time period in which these actions are to be completed. Proposed Table 3.2.3 Action Note 1provides a one hour time period inwhich to complete the action of isolating the Reactor Building Ventilation System and placing the Standby Gas Treatment System inoperation. Placing a limitation on the time allowed to complete the associated actions represents an additional restriction on plant operation since the time period allowed to complete the isolation will be controlled through TS. This change Isconsistent with the ISTS.

M.2 CTS Table 4.2.3 does not include explicit requirements to calibrate trip units. Proposed Table 4.2.3 requires calibration of the trip units of the following Trip Functions every 3 months: Low Reactor Vessel Water Level (proposed Table 4.2.3 Trip Function 1)and High Drywell Pressure (proposed Table 4.2.3 Trip Function 2). The trip units of these Trip Functions are currently required by CTS Table 4.2.3 to be calibrated with the rest of the associated instrument loops once per operating cycle. Therefore, this change is more restrictive. This change is necessary to ensure consistency with assumptions regarding trip unit calibration frequency used inthe associated setpoint calculations. This change is consistent with the ISTS.

TECHNICAL CHANGES - LESS RESTRICTIVE

'Generic' LA.1 The CTS Table 3.2.3 details relating to system design and operation (i.e., the specific Instrument tag numbers) are unnecessary Inthe TS and are proposed to be relocated to the Technical Requirements Manual (TRM). Proposed Specification 3.2.C and Table 3.2.3 require the reactor building ventilation isolation and the Standby Gas Treatment System initiation instrumentation Trip Functions to be operable. Inaddition, the proposed Surveillance Requirements inTable 4.2.3 ensure the required instruments are properly tested. These requirements are adequate for ensuring each of the required reactor building ventilation isolation and the Standby Gas Treatment System Initiation Trip Functions are maintained operable. As such, the relocated details are not required to be inthe VYNPS TS to provide adequate protection of the public health and safety. Changes to the TRM are controlled by the provisions of 10 CFR 50.59. Not including these details inTS Isconsistent with the ISTS.

LA.2 CTS Table 3.2.3 Note 3 contains design and operational details of the reactor building ventilation isolation and the Standby Gas Treatment System initiation instrumentation (i.e.,

there shall be two operable or tripped trip systems for each Trip Function). These details are not necessary to ensure the operability of associated reactor building ventilation isolation and the Standby Gas Treatment System initiation instrumentation. Therefore, the information inthis note isto be relocated to Specification 3.2.C Bases and reference to this information is deleted from WNPS TS. The requirements of Specification 3.2.C and the VYNPS 4 Revision 0

SAFETY ASSESSMENT OF CHANGES TS: 3.2.C/4.2.C - REACTOR BUILDING VENTILATION ISOLATION AND STANDBY GAS TREATMENT SYSTEM INITIATION INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE LA.2 associated Surveillance Requirements for the associated reactor building ventilation (continued) isolation and the Standby Gas Treatment System initiation instrumentation are adequate to ensure the instruments are maintained operable. As such, these relocated requirements are not required to be inthe WNPS TS to provide adequate protection of the public health and safety. Changes to the TS Bases are controlled by the provisions of 10 CFR 50.59.

Not including these details in TS isconsistent with the ISTS.

Specific' L.1 The Applicability of the Low Reactor Vessel Water Level Trip Function of CTS Table 3.2.3 ismWhen Reactor Building Ventilation Isolation and Standby Gas Treatment System Initiation are required by Specification 3.7.B and 3.7.C.' The requirements of Specifications 3.7.B and 3.7.C, and as a result the requirements for this Trip Function, are applicable in the Run, Startup/Hot Standby, Hot Shutdown, and Refuel (with reactor coolant water temperature > 212 0F) Modes and during operations with a potential for draining the reactor vessel, during movement of irradiated fuel assemblies or fuel cask insecondary containment, and during alteration of the reactor core. The Low Reactor Vessel Water Level Trip Function is required to support the operability of the Secondary Containment System and the Standby Gas Treatment System to ensure fission products entrapped within secondary containment are treated prior to discharge to the environment. When the plant isin Cold Shutdown or Refuel (with reactor coolant water temperature < 212 0F), the probability and consequences of adesign basis accident that is postulated to leak fission products into secondary containment are reduced due to the temperature and pressure limitations inthese Modes and conditions. However, inCold Shutdown or Refuel (with reactor coolant water temperature < 212 0F), activities are conducted for which significant releases of radioactivity are postulated due to reductions inreactor vessel water level. As a result, the Low Reactor Vessel Water Level Trip Function is required to be operable inCold Shutdown or Refuel (with reactor coolant water temperature < 2121F), when activities are inprogress which could result inreactor vessel water level reductions (i.e., during operations with the potential for draining the reactor vessel (OPDRVs)). Low Reactor Vessel Water Level actuation of the reactor building ventilation isolation and the Standby Gas Treatment System initiation isnot assumed to mitigate the consequences of postulated events inthe Cold Shutdown or Refuel (with reactor coolant water temperature <

212 0F)Mode when OPDRVs are not being conducted. Inthe Cold Shutdown or Refuel (with reactor coolant water temperature < 212 0F) Mode, when other activities are conducted for which significant releases of radioactivity are postulated and OPDRVs are not being conducted, other reactor building ventilation isolation and the Standby Gas Treatment System initiation Trip Functions are required to be operable to generate the required isolation and initiation signals if required. As such, proposed Table 3.2.3 Trip Function 2 is revised to only be required inthe Run, Startup/Hot Standby, Hot Shutdown and Refuel (with reactor coolant temperature > 212 0F) Modes and during OPDRVs. This change isconsistent with the ISTS.

VYNPS 5 Revision 0

SAFETY ASSESSMENT OF CHANGES TS: 3.2.C/4.2.C - REACTOR BUILDING VENTILATION ISOLATION AND STANDBY GAS TREATMENT SYSTEM INITIATION INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE (continued)

L.2 The Applicability of the High Drywell Trip Function of CTS Table 3.2.3 is'When Reactor Building Ventilation Isolation and Standby Gas Treatment System Initiation are required by Specification 3.7.B and 3.7.C.' The requirements of Specifications 3.7.B and 3.7.C, and as a result the requirements for this Trip Function, are applicable inthe Run, Startup/Hot Standby, Hot Shutdown, and Refuel (with reactor coolant water temperature > 212 0F)

Modes and during operations with a potential for draining the reactor vessel, during movement of irradiated fuel assemblies or fuel cask insecondary containment, and during alteration of the reactor core. The High Drywell Pressure Trip Function is required to support the operability of the Secondary Containment System and the Standby Gas Treatment System to ensure fission products entrapped within secondary containment are treated prior to discharge to the environment. When the plant is inCold Shutdown or Refuel (with reactor coolant water temperature < 212 0 F), the probability and consequences of a design basis accident that is postulated to leak fission products into secondary containment are reduced due to the temperature and pressure limitations in these Modes and conditions. Inaddition, inthese Modes or conditions, there is insufficient energy inthe reactor vessel to pressurize the primary containment and the primary containment is not required to be operable. As a result, High Drywell Pressure actuation of the reactor building ventilation isolation and the Standby Gas Treatment System initiation is not assumed to mitigate the consequences of postulated events inthe Cold Shutdown or Refuel (with reactor coolant water temperature < 212 0F) Mode. Inthe Cold Shutdown or Refuel (with reactor coolant water temperature < 212 0F)Mode, when other activities are conducted for which significant releases of radioactivity are postulated, other reactor building ventilation isolation and the Standby Gas Treatment System initiation Trip ,

Functions are required to be operable to generate the required isolation and initiation signals if required. As such, proposed Table 3.2.3 Trip Function 2 is revised to only be required in the Run, Startup/Hot Standby, Hot Shutdown and Refuel (with reactor coolant temperature> 212 0F)Modes. This change is consistentwith the ISTS.

L.3 CTS Table 3.2.3 includes requirements for Trip System Logics associated with the reactor building ventilation isolation and the Standby Gas Treatment System initiation instrumentation Trip Functions. These Trip Systems Logics are considered part of the reactor building ventilation isolation and the Standby Gas Treatment System initiation instrumentation Trip Functions and the requirements for these associated Trip System Logics to be operable are encompassed by the definition of operable. Therefore, the CTS Table 3.2.3 listing of Trip System Logics as separate Trip Functions is unnecessary and is deleted. With the deletion of separate Trip System Logic Trip Functions, the actions associated with inoperable Trip System Logic (CTS Table 3.2.3 Note 1)will now be governed by the actions for the individual proposed Table 3.2.3 reactor building ventilation isolation and the Standby Gas Treatment System initiation instrumentation Trip Functions.

These proposed Table 3.2.3 Action Notes are less restrictive than the CTS Table 3.2.3 Note 1 actions. However, the proposed actions will ensure, inthe event of inoperabilities, VYNPS 6 Revision 0

SAFETY ASSESSMENT OF CHANGES TS: 3.2.C/4.2.C - REACTOR BUILDING VENTILATION ISOLATION AND STANDBY GAS TREATMENT SYSTEM INITIATION INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE L.3 that consistent actions are applied to both reactor building ventilation isolation and the (continued) Standby Gas Treatment System initiation Instrumentation Trip Functions and their associated Trip System Logics for the same level of degradation. This change is acceptable, since the allowed outage times of the proposed Table 3.2.3 AcUon Notes will limit operation to within the bounds of the applicable analysis, i.e., NEDC-31677-P-A,

'Technical Specifications Improvement Analyses for BWR Isolation Actuation Instrumentation, Part 2,' July 1990, and NEDC-30851 -P-A Supplement 2, Technical Specifications Improvement Analyses for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation,' March 1989. Application of these analyses to the WNPS reactor building ventilation Isolation and the Standby Gas Treatment System initiation instrumentation Trip Functions, including the associated Trip System Logics, was approved by the NRC inVYNPS License Amendment No. 186 dated April 3,2000. This change is consistent with the ISTS.

RELOCATED SPECIFICATIONS None VYNPS 7 Revision 0

No Significant Hazards Consideration 3.2.C and 4.2.C Reactor Building Ventilation Isolation and Standby Gas Treatment System Initiation

GENERIC NO SIGNIFICANT HAZARDS CONSIDERATION ADMINISTRATIVE CHANGES (WAex Labeled Comments/Discussions)

Inaccordance with the criteria set forth in10 CFR 50.92, Vermont Yankee Nuclear Power Station has evaluated this proposed Technical Specifications change and determined it does not represent a significant hazards consideration. The following is provided insupport of this conclusion.

1. Does the change involve a significant increase inthe probability or consequences of an accident previously evaluated?

The proposed change involves reformatting, renumbering, and rewording the existing Technical Specifications. The reformatting, renumbering, and rewording process involves no technical changes to the existing Technical Specifications. As such, this change isadministrative in nature and does not impact initiators of analyzed events or assumed mitigation of accident or transient events. Therefore, this change does not involve a significant increase inthe probability or consequences of an accident previously evaluated.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes inmethods governing normal plant operation. The proposed change will not impose any new or eliminate any old requirements. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction ina margin of safety?

The proposed change will not reduce a margin of safety because it has no impact on any safety analyses assumptions. This change is administrative innature. Therefore, the change does not involve a significant reduction in a margin of safety.

VYNPS 1 Revision 0

GENERIC NO SIGNIFICANT HAZARDS CONSIDERATION TECHNICAL CHANGES - MORE RESTRICTIVE rM.xo Labeled Comments/Discussions)

Inaccordance with the criteria set forth in 10 CFR 50.92, Vermont Yankee Nuclear Power Station has evaluated this proposed Technical Specifications change and determined it does not represent a significant hazards consideration. The following isprovided insupport of this conclusion.

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed change provides more stringent requirements for operation of the facility. These more stringent requirements do not result inoperation that will increase the probability of initiafing an analyzed event and do not alter assumptions relative to mitigation of an accident or transient event. The more restrictive requirements continue to ensure process variables, structures, systems, and components are maintained consistent with the safety analyses and licensing basis. Therefore, this change does not involve asignificant Increase inthe probability or consequences of an accident previously evaluated.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes inthe methods governing normal plant operation. The proposed change does impose different requirements. However, these changes are consistent with the assumptions in the safety analyses and licensing basis. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction ina margin of safety?

The imposition of more restrictive requirements either has no impact on or increases the margin of plant safety. As provided inthe discussion of the change, each change Inthis category is by definition, providing additional restrictions to enhance plant safety. The change maintains requirements within the safety analyses and licensing basis. Therefore, this change does not involve a significant reduction in a margin of safety.

VYNPS 3 Revision 0

GENERIC NO SIGNIFICANT HAZARDS CONSIDERATION OGENERIC' LESS RESTRICTIVE CHANGES:

RELOCATING DETAILS TO TECHNICAL SPECIFICATION BASES, UFSAR, PROCEDURES, OR OTHER PLANT CONTROLLED DOCUMENTS rLA.x' Labeled Comments/Discussions)

Inaccordance with the criteria set forth In10 CFR 50.92, Vermont Yankee Nuclear Power Station has evaluated this proposed Technical Specifications change and determined it does not represent a significant hazards consideration. The following is provided Insupport of this conclusion.

1. Does the change involve a significant increase Inthe probability or consequences of an accident previously evaluated?

The proposed change relocates certain details from the Technical Specifications to the Bases, UFSAR, procedures, or other plant controlled documents. The Bases, UFSAR, procedures, and other plant controlled documents containing the relocated information will be maintained Inaccordance with 10 CFR 50.59. The UFSAR is subject to the change control provisions of 10 CFR 50.71(e), and the plant procedures and other plant controlled documents are subject to controls imposed by plant administrative procedures, which endorse applicable regulations and standards. Since any changes to the relocated details inthe Bases, UFSAR, procedures, or other plant controlled documents will be evaluated per the requirements of 10 CFR 50.59, no significant increase in the probability or consequences of an accident previously evaluated will be allowed. Therefore, this change does not involve a significant increase inthe probability or consequences of an accident previously evaluated.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change inthe methods governing normal plant operation. The proposed change will not impose or eliminate any requirements, and adequate control of the information will be maintained. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The proposed change will not reduce a margin of safety because it has no impact on any safety analysis assumptions. Inaddition, the details to be transposed from the Technical Specifications to the Bases, UFSAR, procedures, or other plant controlled documents are the same as the existing Technical Specifications. Since any future changes to these details inthe Bases, UFSAR, procedures, or other plant controlled documents will be evaluated per the requirements of 10 CFR 50.59, no significant reduction ina margin of safety will be allowed. Based on 10 CFR 50.92, the existing requirement for NRC review and approval of revisions, to these details proposed for relocation, does not have a specific margin of safety upon which to evaluate. However, since the proposed change is consistent with the BWR Standard Technical Specifications, NUREG-1433, approved by the NRC Staff, revising the Technical Specifications to reflect the approved level of detail ensures no significant reduction inthe margin of safety.

WNPS 4 Revision 0

NO SIGNIFICANT HAZARDS CONSIDERATION TS 3.2.C/4.2.C. - REACTOR BUILDING VENTILATION ISOLATION AND STANDBY GAS TREATMENT SYSTEM INIATION INSTRUMENTATION L.1 CHANGE Inaccordance with the criteria set forth in 10 CFR 50.92, Vermont Yankee Nuclear Power Station has evaluated this proposed Technical Specifications change and determined it does not represent a significant hazards consideration. The following is provided insupport of this conclusion.

1. Does the change involve a significant increase Inthe probability or consequences of an accident previously evaluated?

This change will eliminate the Current Technical Specification 3.2.C Applicability requirements of the Low Reactor Vessel Water Level Trip Function for Modes or conditions other than Run, 0

Startup/Hot Standby, Hot Shutdown, Refuel (with reactor coolant temperature > 212 F) and during operations with the potential for draining the reactor vessel (OPDRVs). This reactor building ventilation isolation and the Standby Gas Treatment System initiation Trip Function is not assumed to be an initiator of any analyzed accident. Therefore, this change does not significantly increase the probability of a previously analyzed accident. Further, this Trip Function is not credited for mitigation of any accident or transient InModes or conditions other than Run, Startup/Hot Standby, Hot Shutdown, Refuel (with reactor coolant temperature > 212 0F) and during OPDRVs. As such, the consequences of an accident occurring with the proposed change are the same as the consequences of an accident occurring with current requirements. Therefore, this change does not significantly increase the consequences of a previously analyzed accident.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not necessitate a physical alteration of the plant (no new or different type of equipment will be Installed) or changes inparameters governing normal plant operation.

The proposed change still ensures the affected instrumentation is required to be operable when it is necessary to perform its function. Therefore, it does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction ina margin of safety?

This change is acceptable because it does not Impact the ability of the affected reactor building ventilation isolation and the Standby Gas Treatment System initiation Instrumentation to perform its intended function which is to support the Secondary Containment System and the Standby Gas Treatment System inthe performance of their safety functions. Inaddition, the change is consistent with the safety analysis since the reactor building ventilation isolation and the Standby Gas Treatment System initiation on low reactor vessel water level is a not assumed to mitigate postulated events in Modes or conditions other than Run, Startup/Hot Standby, Hot Shutdown, Refuel (with reactor coolant temperature > 2120F) and during OPDRVs. Since the change has no effect on any safety analysis assumptions or Initial conditions, the margins of safety continue to be maintained. Therefore, this proposed change does not involve a significant reduction in a margin of safety.

VYNPS 1 Revision 0

NO SIGNIFICANT HAZARDS CONSIDERATION TS 3.2.C/4.2.C. - REACTOR BUILDING VENTILATION ISOLATION AND STANDBY GAS TREATMENT SYSTEM INIATION INSTRUMENTATION L.2 CHANGE Inaccordance with the criteria set forth in 10 CFR 50.92, Vermont Yankee Nuclear Power Station has evaluated this proposed Technical Specifications change and determined Itdoes not represent a significant hazards consideration. The following Isprovided insupport of this conclusion.

I1. Does the change involve a significant increase inthe probability or consequences of an accident previously evaluated?

This change will eliminate the Current Technical Specification 3.2.C Applicability requirements of the High Drywell Pressure Trip Function for Modes or conditions other than Run, Startup/Hot Standby, Hot Shutdown and Refuel (with reactor coolant temperature > 2120F). This reactor building ventilation isolation and the Standby Gas Treatment System initiation Trip Function is not assumed to be an initiator of any analyzed accident. Therefore, this change does not significantly increase the probability of a previously analyzed accident. Further, this Trip Function is not credited for mitigation of any accident or transient InModes or conditions other than Run, 0

Startup/Hot Standby, Hot Shutdown and Refuel (with reactor coolant temperature > 212 F). As such, the consequences of an accident occurring with the proposed change are the same as the consequences of an accident occurring with current requirements. Therefore, this change does not significantly increase the consequences of a previously analyzed accident.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not necessitate a physical alteration of the plant (no new or different type of equipment will be installed) or changes Inparameters governing normal plant operation.

The proposed change continues to ensure the affected instrumentation iscapable of performing its function as assumed in the safety analyses. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction ina margin of safety?

This change is acceptable because it does not Impact the ability of the affected reactor building ventilation isolation and the Standby Gas Treatment System initiation Instrumentation to perform its intended function which is to support the Secondary Containment System and the Standby Gas Treatment System inthe performance of their safety functions. Inaddition, the change is consistent with the safety analysis since reactor building ventilation isolation and the Standby Gas Treatment System initiation on high drywell pressure is a not assumed to mitigate postulated events in Modes or conditions other than Run, Startup/Hot Standby, Hot Shutdown and Refuel (with reactor coolant temperature > 212 0F). Since the change has no effect on any safety analysis assumptions or initial conditions, the margins of safety continue to be maintained. Therefore, this proposed change does not Involve a significant reduction in a margin of safety.

VYNPS 2 Revision 0

NO SIGNIFICANT HAZARDS CONSIDERATION TS 3.2.C/4.2.C. - REACTOR BUILDING VENTILATION ISOLATION AND STANDBY GAS TREATMENT SYSTEM INIATION INSTRUMENTATION L.3 CHANGE Inaccordance with the criteria set forth In10 CFR 50.92, Vermont Yankee Nuclear Power Station has evaluated this proposed Technical Specifications change and determined it does not represent a significant hazards consideration. The following is provided insupport of this conclusion.

I1. Does the change involve a significant increase inthe probability or consequences of an accident previously evaluated?

This change will relax the actions when one or more channels of reactor building ventilation isolation and the Standby Gas Treatment System initiation Trip Function instrumentation are Inoperable due to Inoperable Trip System Logic. The reactor building ventilation Isolation and the Standby Gas Treatment System Initiation Trip Function Instrumentation is not considered to be an initiator of any accidents previously analyzed. Therefore, this change does not significantly increase the probability of a previously analyzed accident. Continued operation with inoperable reactor building ventilation isolation and the Standby Gas Treatment System initiation Trip Function instrumentation channels will continue to be limited Inaccordance with Technical Specifications. Since the level of degradation allowed Inthe proposed actions isthe same as the current actions, the consequences of an accident occurring during the time allowed by proposed change are the same as the consequences currently allowed. Therefore, this change does not significantly increase the consequences of a previously analyzed accident.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not necessitate a physical alteration of the plant (no new or different type of equipment will be installed) or changes in parameters governing normal plant operation.

The proposed change still ensures that continued operation inthe applicable condition Isnot allowed when an associated reactor building ventilation Isolation and the Standby Gas Treatment System initiation Trip Function instrumentation channel is not capable of performing its required safety function. Therefore, it does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction ina margin of safety?

This change will relax actions when one or more channels of the affected reactor building.

ventilation isolation and the Standby Gas Treatment System Initiation Trip Function Instrumentation are Inoperable. No change is being made inthe manner inwhich systems relied upon inthe safety analyses provide plant protection. Plant safety margins continue to be maintained through the limitations established inthe Technical Specifications. This change does not impact plant equipment design or operation, and there are no changes being made to safety limits or limiting safety system settings. The proposed change does not Impact safety. Therefore, the proposed change does not involve a significant reduction Ina margin of safety.

VYNPS 3 Revision 0

References 3.2.C and 4.2.C Reactor Building Ventilation Isolation and Standby Gas Treatment System Initiation

3.2.C/4.2.C REFERENCES Reactor Building Ventilation Isolation and Standby Gas Treatment System Initiation

1. UFSAR, Section 5.3.

2. UFSAR, Section 7.17.2.

3. UFSAR, Section 14.6.3.6.

4. UFSAR, Section 14.6.4.4.

5. NEDC-31677P-A, Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation," July 1990.

6. NEDC-30851 P-A Supplement 2, "Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation,"

March 1989.

Proposed Technical Specifications 3.2.D and 4.2.D Off-Gas System Isolation

VYNPS 3.2 LIMITING CONDITIONS FOR 4.2 SURVEILLANCE REQUIREMENTS OPERATION D. Deleted. D. Deleted.

I I E. Control Rod Block Actuation E. Control Rod Block Actuation The control rod block 1. The control rod block instrumentation for each instrumentation shall be I Trip Function in Table functionally tested and 3.2.5 shall be operable in calibrated as indicated accordance with Table in Table 4.2.5.

3.2.5.

When a Rod Block Monitor channel is placed in an inoperable status solely for performance of required surveillances, entry into associated Limiting Conditions for Operation and required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Trip Function maintains control rod block initiation capability.

Amendment No. 54

Current Technical Specifications Markups 3.2.D and 4.2.D Off-Gas System Isolation

EIJ VYNPS SURVEILLANCE REQUIREMENTS 3.2 LIMITING CONDITIONS FOR OPERATION _

D. Off-Gas System Isolati Dur g reactor wer Inst Xmntation zfd logic o ation, the that/

sy es shall A

_r_- r )strumentat'~ fctionally eted an initiates olation a the Ca ibrated /8indic in off-a sytemcshal 4e Table 4.2/4./

operal/i acrd nce

< wihTble3.2./ _v _,ezoVe TZ VCPAR~rC 1P464f >

E. Control Rod Block Actuation E. Control Rod Block Actuation During reactor power Instrumentation and logic operation the systems shall be instrumentation that functionally tested and initiates control rod block calibrated as indicated in shall be operable Table 4.2.5.

in accordance with Table 3.2.5.

F.Mcaia aumPm ====t F. Mechanical Vacuum Pump Isolation Instrumentation Isolation Instrumentation When the reactor is in the The High Main Steam Line RUN or STARTUP/HOT STANDBY Radiation Trip Function for Mode and the mechanical mechanical vacuum pump vacuum pump is in service, isolation shall be checked, four (4) channels of the functionally tested and High Main Steam Line calibrated as indicated in Radiation Trip Function for Surveillance Requirements mechanical vacuum pump 4.2.F.1, 2, 3, 4 and 5.

isolation shall be operable, except as provided below. When a channel is placed in an inoperable status solely

1. With one or more channels for performance of required inoperable, within 12 surveillances, entry into hours: associated Limiting Conditions for Operation and

a. Restore the inoperable required actions may be channel(s) to operable delayed for up to six (6) status; or hours provided the associated trip function maintains

b. Place the inoperable mechanical vacuum pump channel(s) or isolation capability.

associated trip system in the trip condition 1. Perform an instrument (not applicable if the check once each day.

inoperable channel is the result of an 2. Perform an instrument inoperable mechanical functional test once every vacuum pump isolation three (3) months.

valve).

< dd:TD5 t~f# 6 X6>

Amendment No. A, i64, 212 35

/ ABE 3.2

//OFF-GAS STMISOLATION INSTRUMENTATION Number of Requited Action When Instrument M um Conditions per Trip X s For Operation l_/_System / Trip Function Trip Setting Are Not Satisfied 1 Time Delay (Stack Off-Ga < 2 minutes Note 1 Valve Isolation) (15TD 16TD) < 30 minutes 1 Trip System Logic -- Note 1 Note 1 - t least one of the radiat n monitors between the cha oal bed system and the pla stack shall be opera e during operation of the adgmented off-gas system. I this condition cannot be m , continued oeraion of the augmented off tem is permissible period of up to 7 days ided that at least

\ oeof the stack monitoring sWtems is operable and off-ga sstem temperature and p sure are measured /

\ ontinuously. ,

Amendment No. 4, Rib, 164 50

Amendment No. ., O', 106 68 1.I VYNPS TABLE 4.2 NOTES Ce No used.

2. During each efueling outage, simulate automatic actuation wh opens all pilot alves shall be performed uch that each trip syst logic can R be ver ed independent of its re dant counterpart.

3. Tr system logic calibratio 0shall include only time elay relays and mers necessary for prop functioning of the trip ystem.

{ 4. This instrumentation is excepted from functional test definition. The 5 functional test will consist of injecting a simulated electrical signal into the measurement channel.

16.,/'Delet I tests and cali tions are not requi when systems are no re ed to be operable t 4R (9. The thermocouples associated with sarety/reliet valves and satety valve position, that may be used for back-up position indication, shall be verified to be operable every operating cycle.

10 Separate functional tests are not required for this instrumentation. Th calibration and integrated ECCS tests which are performed once per operating cycle will adequately demonstrate proper equipment operation.

11. Trip system logic functional tests will include verification of operation of all automatic initiation inhibit switches by monitoring relay contact movement. Verification that the manual inhibit switches prevent opening I all relief valves will be accomplished in conjunction with Section 4.5.F.I.

If the

12. Trip system logic testing is not applicable to this function.

required surveillance frequency (every Refueling Outage) is not met, functional testing of the Reactor Mode Switch-Shutdown Position function shall be initiated within 1 hour after the reactor mode switch is placed in Shutdown for the purpose of commencing a scheduled Refueling Outage. I

13. Includes calibration of the RBM Reference Downscale function (i.e., RBM upscale function is not bypassed when >30% Rated Thermal Power).

11------------------------

9&, 45, 45, i64, 81-, Al, 212 74 Amendment No. 64,

Safety Assessment Discussion of Changes 3.2.D and 4.2.D Off-Gas System Isolation

SAFETY ASSESSMENT OF CHANGES CTS: 3.2.D/4.2.D - OFF-GAS SYSTEM ISOLATION INSTRUMENTATION ADMINISTRATIVE A.1 Inthe revision of the Vermont Yankee Nuclear Power Station (VYNPS) current Technical Specifications (CTS), certain wording preferences or conventions are adopted which do not result intechnical changes (either actual or interpretational). Editorial changes, reformatting, and revised numbering are adopted to make the VYNPS Technical Specifications (TS) more consistent with human factor principles used inthe Boiling Water Reactor Improved Standard Technical Specifications (ISTS), NUREG-1433, Rev. 2. These format and presentation changes are being made to improve usability and clarity. The changes are considered administrative.

A.2 CTS 3.2.E, 4.2.E, and Table 4.2 Notes 4, 12, and 13 provide requirements that apply to control rod block instrumentation. The control rod block instrumentation is located in proposed Specifications 3.2.E and 4.2.E. Therefore, the requirements of CTS 3.2.E, 4.2.E, and Table 4.2 Notes 4,12, and 13 are physically moved and addressed inthe changes to proposed Specifications 3.2.E and 4.2.E. Therefore, this change does not involve a technical change, but isonly a difference of presentation preference and is considered administrative. CTS Table 4.2 Note 9 provides requirements that apply to post-accident monitoring instrumentation. The post-accident monitoring instrumentation is located in proposed Specifications 3.2.G and 4.2.G. Therefore, the requirements of CTS Table 4.2 Note 9 are physically moved and changes addressed in proposed Specifications 3.2.G and 4.2.G. Therefore, this change does not involve a technical change, but is only a difference of presentation preference and is considered administrative. CTS Table 4.2 Notes 10 and 11 provide requirements that apply to ECCS instrumentation. The ECCS instrumentation is located inproposed Specifications 3.2.A and 4.2.A. Therefore, the requirements of CTS Table 4.2 Notes 10 and 11 are physically moved and addressed inthe changes for proposed Specifications 3.2.A and 4.2.A. Therefore, this change does not involve a technical change, but isonly a difference of presentation preference and is considered administrative.

TECHNICAL CHANGES - MORE RESTRICTIVE None TECHNICAL CHANGES - LESS RESTRICTIVE None VYNPS 1 Revision 0

SAFETY ASSESSMENT OF CHANGES CTS: 3.2.D/4.2.D - OFF-GAS SYSTEM ISOLATION INSTRUMENTATION RELOCATED SPECIFICATIONS R.1 3.2.D/4.2.D OFF-GAS SYSTEM ISOLATION INSTRUMENTATION LCO Statement:

During reactor power operation, the instrumentation that initiates isolation of the off-gas system shall be operable inaccordance with Table 3.2.4.

Discussion:

The radioactive off-gas processing system is not a safety system and is not connected to the primary coolant piping. The off-gas isolation instrumentation is used to ensure conformance with the discharge limits of 10 CFR 20. There isanother Specification (Specification 3.8.K14.8.K, Steam Jet Air Ejector) that ensures 10 CFR 100 limits are not exceeded inthe event of a failure of the radioactive off-gas processing system. Information provided by this instrumentation on radiation levels would be limited or no use in identifying/assessing core damage inthe event of an accident and they are not installed to detect excessive reactor coolant leakage.

Comparison to Screening Criteria:

1. Off-gas system isolation instrumentation isnot used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a Design Basis Accident (DBA).

2. Although off-gas activity is an initial condition of a DBA, this process variable is addressed by another Technical Specification. Criterion 2 is satisfied for the process variable (steam jet air ejector radioactivity) that is addressed by another Technical Specification. However, Criterion 2 is not satisfied for off-gas system isolation instrumentation, since this instrumentation is not a process variable that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

3. Off-gas system isolation instrumentation is not part of the primary success path that functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

4. As discussed in Sections 3.5 and 6 and summarized inTable 4-1 (item 145) of NEDO-31466, 'Technical Specification Screening Criteria Application and Risk Assessment," dated November 1987, the loss of the off-gas system isolation instrumentation was found to be a non-significant risk contributor to core damage frequency and offsite releases. VYNPS has reviewed this evaluation, considers it applicable to VYNPS, and concurs with the assessment.

VYNPS 2 Revision 0

SAFETY ASSESSMENT OF CHANGES CTS: 3.2.D/4.2.D - OFF-GAS SYSTEM ISOLATION INSTRUMENTATION RELOCATED SPECIFICATIONS R.I (continued)

Conclusion:

Since the screening criteria have not been satisfied, the Off-Gas System Isolation Instrumentation LCO, Actions, and Surveillances will be relocated to the Technical Requirements Manual.

Changes to the Technical Requirements Manual are controlled using 10 CFR 50.59.

VYNPS 3 Revision 0

No Significant Hazards Consideration 3.2.D and 4.2.D Off-Gas System Isolation

GENERIC NO SIGNIFICANT HAZARDS CONSIDERATION ADMINISTRATIVE CHANGES (WA.xU Labeled Comments/Discussions)

Inaccordance with the criteria set forth In10 CFR 50.92, Vermont Yankee Nuclear Power Station has evaluated this proposed Technical Specifications change and determined it does not represent a significant hazards consideration. The following Isprovided insupport of this conclusion.

1. Does the change involve a significant Increase inthe probability or consequences of an accident previously evaluated?

The proposed change involves reformatting, renumbering, and rewording the existing Technical Specifications. The reformatting, renumbering, and rewording process involves no technical changes to the existing Technical Specifications. As such, this change is administrative in nature and does not impact initiators of analyzed events or assumed mitigation of accident or transient events. Therefore, this change does not involve asignificant increase Inthe probability or consequences of an accident previously evaluated.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be Installed) or changes in methods governing normal plant operation. The proposed change will not impose any new or eliminate any old requirements. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction ina margin of safety?

The proposed change will not reduce a margin of safety because it has no impact on any safety analyses assumptions. This change is administrative in nature. Therefore, the change does not Involve a significant reduction in a margin of safety.

VYNPS 1 Revision 0

GENERIC NO SIGNIFICANT HAZARDS CONSIDERATION RELOCATED SPECIFICATIONS

("R.x' Labeled Comments/Discussions)

Inaccordance with the criteria set forth in 10 CFR 50.92, Vermont Yankee Nuclear Power Station has evaluated this proposed Technical Specifications change and determined it does not represent a significant hazards consideration. The following is provided Insupport of this conclusion.

1. Does the change involve a significant increase inthe probability or consequences of an accident previously evaluated?

The proposed change relocates requirements and surveillances for structures, systems, components or variables that do not meet the criteria for inclusion in Technical Specifications as identified in 10 CFR 50.36 (c)(2)(ii). The affected structures, systems, components or variables are not assumed to be initiators of analyzed events and are not assumed to mitigate accident or transient events. The requirements and surveillances for these affected structures, systems, components or variables will be relocated from the Technical Specifications to an appropriate administratively controlled document which will be maintained pursuant to 10 CFR 50.59. Inaddition, the affected structures, systems, components or variables are addressed inexisting surveillance procedures which are also controlled by 10 CFR 50.59 and subject to the change control provisions imposed by plant administrative procedures, which endorse applicable regulations and standards. Therefore, this change does not Involve a significant Increase inthe probability or consequences of an accident previously evaluated.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be Installed) or a change inthe methods governing normal plant operation. The proposed change will not impose or eliminate any requirements and adequate control of existing requirements will be maintained. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction ina margin of safety?

The proposed change will not reduce a margin of safety because it has no impact on any safety analysis assumptions. Inaddition, the relocated requirements and surveillances for the affected structure, system, component or variable remain the same as the existing Technical Specifications. Since any future changes to these requirements or the surveillance procedures will be evaluated per the requirements of 10 CFR 50.59, no significant reduction ina margin of safety will be permitted.

The existing requirement for NRC review and approval of revisions, Inaccordance with 10 CFR 50.92, to these details proposed for relocation does not have a specific margin of safety upon which to evaluate.

However, since the proposed change isconsistent with the BWR Standard Technical Specification, NUREG-1433 approved by the NRC Staff, revising the Technical Specifications to reflect the approved level of detail ensures no significant reduction inthe margin of safety.

VYNPS 2 Revision 0

HAZARDS CONSIDERATION NO SIGNIFICANT NO SIGNIFICANT HAZARDS CONSIDERATION CTS: 3.2.D14.2.D - OFF - GAS ISOLATION INSTRUMENTATION There were no specific less restrictive changes identified for this Specification.

VYNPS 1 Revision 0

Proposed Technical Specifications 3.2.E and 4.2.E Control Rod Block Actuation And related Technical Specifications:

3.3 - Control Rod System 3.11 - Reactor Fuel Assemblies

VYNPS 3.2 LIMITING CONDITIONS FOR 4.2 SURVEILLANCE REQUIREMENTS OPERATION D. Deleted.

I D. Deleted. I E. Control Rod Block Actuation E. Control Rod Block Actuation The control rod block 1. The control rod block instrumentation for each instrumentation shall be I Trip Function in Table functionally tested and 3.2.5 shall be operable in calibrated as indicated accordance with Table in Table 4.2.5.

3.2.5.

When a Rod Block Monitor channel is placed in an inoperable status solely for performance of required surveillances, entry into associated Limiting Conditions for Operation and required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Trip Function maintains control rod block initiation capability.

Amendment No. 54

VYNPS Table 3.2.5 (page 1 of 1)

Control Rod Block Instrumentation ACTIONS WHEN APPLICABLE REQUIRED REQUIRED MODES OR OTHER CHANNELS CHANNELS SPECIFIED PER TRIP ARE TRIP SETTING TRIP FUNCTION CONDITIONS FUNCTION INOPERABLE

1. Rod Block Monitor 2

a. Upscale (Flow Bias) > 30% RATED 2 Note 1 S 0.66(W)+N 1 THERMAL POWER with a maximum as defined in the COLR

b. Downscale > 30% RATED 2 Note 1 2 2/125 full THERMAL POWER scale

c. Inop > 30% RATED 2 Note 1 NA THERMAL POWER

2. Reactor Mode Switch- (1) 2 Note 2 NA Shutdown Position (1) When reactor mode switch is in the shutdown position.

(2) Trip Setting 5 0.66 (W-AW)+N for single loop operation.

Amendment No. 55

VYNPS Table 3.2.5 ACTION Notes

1. With one or two RBM channels inoperable, take all of the applicable Actions in Notes l.a and l.b below.

a. If one RBM channel is inoperable, restore the inoperable channel to operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. If the required Action and associated completion time of Note l.a above is not met, or if two RBM channels are inoperable, place one RBM channel in trip within the next hour.

2. With one or more Reactor Mode Switch - Shutdown Position channels inoperable, immediately suspend control rod withdrawal and immediately initiate Actions to fully insert all insertable control rods in core cells containing one or more fuel assemblies.

Amendment No. 56

VYNPS Table 4.2.5 (page 1 of 1)

Control Rod Block Instrumentation Tests and Frequencies TRIP FUNCTION FUNCTIONAL TEST CALIBRATION

1. Rod Block Monitor (RBM)

a. Upscale (Flow Bias) Every 3 Months Every 3 Months(2)

b. Downscale Every 3 Months Every 3 Months'2 )

c. Inop Every 3 Months NA

2. Reactor Mode Switch - Every Refueling NA Shutdown Position Outage"l)

(1) Not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after the reactor mode switch is placed in the shutdown position.

(2) Neutron detectors are excluded.

Amendment No. 57

VYNPS 3.3 LIMITING CONDITIONS FOR 4.3 SURVEILLANCE REQUIREMENTS OPERATION pressure, are fully inserted, no more than two rods may be moved.

(c) Out-of-sequence control rods in each distinct RWM group shall be selected and the annunciator of the selection errors verified.

(d) An out-of-sequence control rod shall be withdrawn no more than three notches and the rod block function verified.

4. Control rod patterns and 4. The control rod pattern the sequence of and sequence of withdrawal or insertion withdrawal or insertion shall be established shall be verified to such that the rod drop comply with accident limit of Specification 3.3.B.4.

280 cal/g is not exceeded.

5. Control rods shall not 5. Prior to control rod be withdrawn for startup withdrawal for startup or refueling unless at or during refueling, least two source range verification shall be channels have an made that at least two observed count rate source range channels greater than or equal to have an observed count three counts per second. rate of at least three counts per second.

6. Deleted.

I 6. Deleted. I Amendment No. A4, 6h8 84

VYNPS 3.3 LIMITING CONDITIONS FOR 4.3 SURVEILLANCE REQUIREMENTS OPERATION

7. The scram discharge volume drain and vent valves shall be verified open at least once per month. These valves may be closed intermittently for testing under administrative control.

C. Scram Insertion Times C. Scram Insertion Times 1.1 The average scram time, 1. After refueling outage based on the and prior to operation de-energization of the above 30% power with scram pilot valve reactor pressure above solenoids of all 800 psig all control operable control rods in rods shall be subject to the reactor power scram-time measurements operation condition from the fully withdrawn shall be no greater position. The scram than: times for single rod scram testing shall be Drop-Out % Inserted Avg. Scram measured without of From Fully Insertion reliance on the control Position Withdrawn Time (sec) rod drive pumps.

46 4.51 0.358 2. During or following a 36 25.34 0.912 controlled shutdown of 26 46.18 1.468 the reactor, but not 06 87.84 2.686 more frequently than 16 weeks nor less The average of the scram frequently than 32 weeks insertion times for the intervals, 50% control three fastest control rod drives in each rods of all groups of quadrant of the reactor four control rods in a core shall be measured two by two array shall for scram times be no greater than: specified in Specification 3.3.C.

Drop-Out % Inserted Avg. Scram All control rod drives of From Fully Insertion shall have experienced Position Withdrawn Time (sec) scram-time measurements each year. Whenever 50%

46 4.51 0.379 of the control rod 36 25.34 0.967 drives scram times have 26 46.18 1.556 been measured, an 06 87.84 2.848 evaluation shall be made to provide reasonable assurance that proper control rod drives performance is being maintained. The results of measurements performed on the control rod drives shall be submitted in the start up test report.

Amendment No. -4, 4&, 34, 4, b8 85

VYNPS 3.11 LIMITING CONDITIONS FOR 4.11 SURVEILLANCE REQUIREMENTS OPERATION C. Minimum Critical Power Ratio C. Minimum Critical Power Ratio (MCPR) (MCPR)

1. During operation at >25% MCPR shall be determined Rated Thermal Power the once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after MCPR operating value >25% Rated Thermal Power, shall be equal to or daily during operation at greater than the MCPR >25% Rated Thermal Power limits provided in the thereafter.

Core Operating Limits Report. For single recirculation loop operation, the MCPR Limits at rated flow are also provided in the Core Operating Limits Report. For core flows other than rated, the Operating MCPR Limit shall be the above value multiplied by Kf where Kf is provided in the Core Operating Limits Report. If at any time during operation at >25t Rated Thermal Power it is determined by normal surveillance that the limiting value for MCPR is being exceeded, MCPR(s) shall be returned to within the prescribed limits within two (2) hours; otherwise, the reactor power shall be brought to <25t Rated Thermal Power within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.

Amendment No. a-, 4-, 8A, 04, GA6, aS8&2 226

Proposed Bases 3.2.E and 4.2.E Control Rod Block Actuation And related Technical Specification Bases:

3.3 - Control Rod System 3.11 - Reactor Fuel Assemblies

VYNPS BASES: 3.2.E/4.2.E CONTROL ROD BLOCK ACTUATION BACKGROUND Control rods provide the primary means for control of reactivity changes.

Control rod block instrumentation includes channel sensors, logic circuitry, switches, and relays that are designed to ensure that specified fuel design limits are not exceeded for postulated transients and accidents. During high power operation, the rod block monitor (RBM) provides protection for control rod withdrawal error events. During shutdown conditions, control rod blocks from the Reactor Mode Switch-Shutdown Position Function ensure that all control rods remain inserted to prevent inadvertent criticalities.

The purpose of the RBM (Ref.l) is to limit control rod withdrawal if localized neutron flux exceeds a predetermined setpoint during control rod manipulations. It is assumed to function to block further control rod withdrawal to preclude a Fuel Cladding Integrity Safety Limit violation. The RBM supplies a trip signal to the Reactor Manual Control System (RMCS) to appropriately inhibit control rod withdrawal during power operation above the 30% RATED THERMAL POWER setpoint when a non-peripheral control rod (except control rod 35-34) is selected. The RBM has two channels, either of which can initiate a control rod block when the channel output exceeds the control rod block setpoint. One RBM channel inputs into one RMCS rod block circuit and the other RBM channel inputs into the second RMCS rod block circuit. A rod block in either RMCS circuit will provide a control rod block to all control rods. The RBM channel signal is generated by averaging a set of local power range monitor (LPRM) signals. One RBM channel averages the signals from LPRM detectors at the A and C positions in the assigned LPRM assemblies, while the other RBM channel averages the signals from LPRM detectors at the B and D positions. Assignment of LPRMs to be used in RBM averaging is controlled by the selection of control rods. The RBM is automatically bypassed and the output set to zero if a peripheral rod (or control rod 35-34) is selected or the APRM used to normalize the RBM reading is at < 30% RATED THERMAL POWER. If any LPRM detector assigned to an RBM is bypassed, the computed average signal is automatically adjusted to compensate for the number of LPRM input signals. The minimum number of LPRM inputs required for each RBM channel to prevent an instrument inoperative trip is four when using four LPRM strings, three when using three LPRM strings, and two when using two LPRM strings. Each RBM also receives a recirculation loop flow signal from the associated flow converter.

When a control rod is selected, the gain of each RBM channel output is normalized to a reference APRM. The gain setting is held constant during the movement of that particular control rod to provide an indication of the change in the relative local power level. If the indicated power increases above the preset limit, a rod block will occur. In addition, to preclude rod movement with an inoperable RBM, a downscale trip and an inoperable trip are provided.

With the reactor mode switch in the shutdown position, a control rod withdrawal block is applied to all control rods to ensure that the shutdown condition is maintained (Ref. 2). This Trip Function prevents inadvertent criticality as the result of a control rod withdrawal during COLD SHUTDOWN and HOT SHUTDOWN or during a Refueling Outage when the reactor mode switch is required to be in the shutdown position. The reactor mode switch has two channels, each inputting into a separate RMCS rod block circuit. A rod block in either RMCS circuit will provide a control rod block to all control rods.

Amendment No.

77

VYNPS BASES: 3.2.E/4.2.E CONTROL ROD BLOCK ACTUATION APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY l.a, l.b, l.c Rod Block Monitor The RBM is designed to prevent violation of the Fuel Cladding Integrity Safety Limit and the cladding 1% plastic strain fuel design limit that may result from a single control rod withdrawal error (RWE) event. The analytical methods and assumptions used in evaluating the RWE event are summarized in Reference 3. The cycle-specific analysis considers the continuous withdrawal of the maximum worth control rod at its maximum drive speed from the reactor, which is operating at rated power with a control rod pattern that results in the core being placed on thermal design limits. The condition is analyzed to ensure that the results obtained are conservative; the approach also serves to demonstrate the functions of the RBM.

The REM Trip Functions satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

Two channels of each of the RBM Trip Functions are required to be operable, with their trip setpoints within the calculational as-found tolerances specified in plant procedures, as applicable, to ensure that no single instrument failure can preclude a rod block from these Trip Functions. In addition, to provide adequate coverage of the entire core, LPRM inputs for each RBM channel are required from greater than or equal to half the total number of inputs from any LPRM level. The upper limit of the RBM Upscale (Flow Bias) Trip Function is clamped to provide protection at greater than 100% rated core flow. This clamped value is cycle-specific and is included in the Core Operating Limits Report. Trip Settings are specified for RBM Upscale (Flow Bias) and RBM Downscale Trip Functions. The terms for the Trip Setting of the RBM Upscale (Flow Bias) Trip Function are defined as follows:

W is percent of rated two loop drive flow where 100% rated drive flow is that flow equivalent to 48 X 106 lbs/hr core flow; and AW is the difference between two loop and single loop drive flow a the same core flow (this difference must be accounted for during single loop operation). AW = 0 for two loop operation.

Operation with actual trip setpoints within calculational as-found tolerances provides reasonable assurance that, under worst case design basis conditions, the associated trip will occur within the Trip Settings specified in Table 3.2.5. As a result, a channel is considered inoperable if its actual trip setpoint is not within the calculational as-found tolerances specified in plant procedures. The actual trip setpoint is calibrated consistent with applicable setpoint methodology assumptions.

The RBM is assumed to mitigate the consequences of an RWE event when operating > 30% RATED POWER THERMAL and a non-peripheral control rod is selected. Below this power level, or if a peripheral control rod is selected, the consequences of an RWE event will not exceed the Fuel Cladding Integrity Safety Limit and, therefore, the RBM is not required to be operable.

^Aue A. JNO.

Amenumentf L77a

VYNPS BASES: 3.2.E/4.2.E CONTROL ROD BLOCK ACTUATION APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

2. Reactor Mode Switch - Shutdown Position During HOT SHUTDOWN and COLD SHUTDOWN, and during Refueling Outages when the reactor mode switch is in the shutdown position, the core is assumed to be subcritical; therefore, no positive reactivity insertion events are analyzed.

The Reactor Mode Switch-Shutdown Position control rod withdrawal block ensures that the reactor remains subcritical by blocking control rod withdrawal, thereby preserving the assumptions of the safety analysis.

The Reactor Mode Switch-Shutdown Position Trip Function satisfies Criterion 3 of 10 CFR 50.36(c) (2)(ii).

Two channels are required to be operable to ensure that no single channel failure will preclude a rod block when required. There is no Trip Setting for this Trip Function since the channels are mechanically actuated based solely on reactor mode switch position. During shutdown conditions (HOT SHUTDOWN and COLD SHUTDOWN, and Refueling Outages when the reactor mode switch is in the shutdown position), no positive reactivity insertion events are analyzed because assumptions are that control rod withdrawal blocks are provided to prevent criticality. Therefore, when the reactor mode switch is in the shutdown position, the control rod withdrawal block is required to be operable. With the reactor mode switch in the refueling position, the refuel position one-rod-out interlock provides the required control rod withdrawal blocks.

ACTIONS Table 3.2.5 ACTION Note 1 With one RBM Trip Function l.a, l.b, or l.c channel inoperable, the remaining operable channel is adequate to perform the control rod block function; however, overall reliability is reduced because a single failure in the remaining operable RBM channel can result in no control rod block capability for the RBM. For this reason, Table 3.2.5 ACTION Note l.a requires restoration of the inoperable channel to operable status. The Completion Time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is based on the low probability of an event occurring coincident with a failure in the remaining operable channel.

If the Table 3.2.5 ACTION Note l.a required action is not met and the associated Completion Time has expired, the inoperable channel must be placed in trip within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . If both RBM Trip Function l.a, l.b, or l.c channels are inoperable, the RBM is not capable of performing its intended function; thus, one channel must also be placed in trip. This initiates a control rod withdrawal block, thereby ensuring that the RBM function is met. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities and is acceptable because it minimizes risk while allowing time for restoration or tripping of inoperable channels.

Table 3.2.5 ACTION Note 2 With one Reactor Mode Switch-Shutdown Position control rod withdrawal block channel inoperable, the remaining operable channel is adequate to perform the Amendment No. 77b

VYNPS BASES: 3.2.E/4.2.E CONTROL ROD BLOCK ACTUATION ACTIONS (continued) control rod withdrawal block function. However, since the required actions of Table 3.2.5 ACTION Note 2 are consistent with the normal action of an operable Reactor Mode Switch-Shutdown Position Trip Function (i.e.,

maintaining all control rods inserted), there is no distinction between having one or two channels inoperable.

In both cases (one or both channels inoperable), suspending all control rod withdrawal and initiating action to fully insert all insertable control rods in core cells containing one or more fuel assemblies will ensure that the core is subcritical with adequate Shutdown Margin ensured by Specification 3.3.A.I. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and are therefore not required to be inserted. Action must continue until all insertable control rods in core cells containing one or more fuel assemblies are fully inserted.

SURVEILLANCE REQUIREMENTS Surveillance Requirement 4.2.E.1 As indicated in Surveillance Requirement 4.2.E.1, control rod block instrumentation shall be functionally tested and calibrated as indicated in Table 4.2.5. Table 4.2.5 identifies, for each Trip Function, the applicable Surveillance Requirements.

Surveillance Requirement 4.2.E.1 also indicates that when an RBM channel is placed in an inoperable status solely for performance of required instrumentation Surveillances, entry into associated LCO and required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Trip Function maintains control rod block capability. Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance, the channel must be returned to operable status or the applicable LCO entered and required Actions taken.

This allowance is based on the reliability analysis (Ref. 4) assumption of the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months testing allowance does not significantly reduce the probability that a control rod block will be initiated when necessary.

Table 4.2.5, Functional Test For Trip Functions l.a, l.b, and l.c, a Functional Test is performed on each required channel to ensure that the channel will perform the intended function. The Functional Test of the RBM channels includes the Reactor Manual Control "Select Relay MatrixN System input. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of "Every 3 Months" is based on the reliability analysis of Reference 5.

For Trip Function 2, a Functional Test is performed to ensure that the entire channel will perform the intended function. The Functional Test for the Reactor Mode Switch-Shutdown Position Trip Function is performed by attempting to withdraw any control rod with the reactor mode switch in the Amendment No. 77c

VYNPS BASES: 3.2.E/4.2.E CONTROL ROD BLOCK ACTUATION SURVEILLANCE REQUIREMENTS (continued) shutdown position and verifying a control rod block occurs. As noted in Table 4.2.5 Footnote (1), the Surveillance is not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after the reactor mode switch is in the shutdown position, since testing of this interlock with the reactor mode switch in any other position cannot be performed without using jumpers, lifted leads, or movable links.

This allows entry into the HOT SHUTDOWN and COLD SHUTDOWN Modes if the "Every Refueling Outage" Frequency is not met. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months allowance is based on operating experience and in consideration of providing a reasonable time in which to complete the Surveillance Requirement. The Frequency of "Every Refueling Outage" is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has demonstrated that these components will usually pass the Surveillance when performed at the specified Frequency.

Table 4.2.5, Calibration For Trip Functions l.a and 1.b, an Instrument Calibration is a complete check of the instrument loop and the sensor. This test verifies that the channel responds to the measured parameter within the necessary range and accuracy.

An Instrument Calibration leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. The specified Instrument Calibration Frequencies are based upon the time interval assumptions for calibration used in the determination of the magnitude of equipment drift in the associated setpoint analyses.

The RBM is automatically bypassed when power is below a specified value or if a peripheral control rod is selected. The power level is determined from the APRM signals input to each RBM channel. The automatic bypass setpoint must be verified periodically to be < 30% RATED THERMAL POWER. As a result, the Instrument Calibration of Trip Function l.a must also include calibration of the RBM Reference Downscale function (i.e., RBM Upscale (Flow Bias) Trip Function is not bypassed when > 30% RATED THERMAL POWER). In addition, it must also be verified that the RBM is not bypassed when a control rod that is not a peripheral control rod is selected (only one non-peripheral control rod is required to be verified). If any bypass setpoint is nonconservative, then the affected RBM channel is considered inoperable. Alternatively, the APRM channel can be placed in the conservative condition to enable the RBM.

If placed in this condition, the Surveillance Requirement is met and the RBM channel is not considered inoperable.

As noted in Footnote (2), neutron detectors are excluded from the Surveillance because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal. Changes in neutron detector sensitivity are compensated for by performing the 7 day heat balance calibration and the 2000 MWD/T LPRM calibration against the Traversing Incore Probe System of the Reactor Protection System Technical Specification.

Amendment No. 77d

VYNPS BASES: 3.2.E/4.2.E CONTROL ROD BLOCK ACTUATION REFERENCES

1. UFSAR, Section 7.5.8.

2. UFSAR, Section 7.7.4.3.2.

3. UFSAR, Section 14.5.3.1.

4. GENE-770-06-1-A, "Addendum to Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," December 1992.

5. NEDC-30851-P-A, Supplement 1, "Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation," October 1988.

Amendment No. 77e

VYNPS BASES: 3.3 & 4.3 (Cont'd)

2. The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the extremely remote event of a housing failure. The amount of reactivity which could be added by this small amount of rod withdrawal, which is less than a normal single withdrawal increment, will not contribute to any damage of the primary coolant system. The design basis is given in Subsection 3.5.2 of the FSAR, and the design evaluation is given in Subsection 3.5.4. This support is not required if the reactor coolant system is at atmospheric pressure since there would then be no driving force to rapidly eject a drive housing.

3. In the course of performing normal startup and shutdown procedures, a pre-specified sequence for the withdrawal or insertion of control rods is followed. Control rod dropout accidents which might lead to significant core damage, cannot occur if this sequence of rod withdrawals or insertions is followed. The Rod Worth Minimizer restricts withdrawals and insertions to those listed in the pre-specified sequence and provides an additional check that the reactor operator is following prescribed sequence. Although beginning a reactor startup without having the RWM operable would entail unnecessary risk, continuing to withdraw rods if the RWM fails subsequently is acceptable if a second licensed operator verifies the withdrawal sequence. Continuing the startup increases core power, reduces the rod worth and reduces the consequences of dropping any rod. Withdrawal of rods for testing is permitted with the RWM inoperable, if the reactor is subcritical and all other rods are fully inserted. Above 20% power, the RWM is not needed since even with a single error an operator cannot withdraw a rod with sufficient worth, which if dropped, would result in anything but minor consequences.

4. Refer to the 4General Electric Standard Application for Reactor Fuel (GESTAR II),- NEDE-24011-P-A, (the latest NRC-approved version will be listed in the COLR).

5. The Source Range Monitor (SRM) system provides a scram function in noncoincident configuration. It does provide the operator with a visual indication of neutron level. The consequences of reactivity accidents are a function of the initial neutron flux. The requirement of at least three counts per second assures that any transient, should it occur, begins at or above the initial value of l0o- of rated power used in the analyses of transients from cold conditions. One operable SRM channel is adequate to monitor the approach to criticality, therefore, two operable SRM's are specified for added conservatism.

6. The Rod Block Monitor (REM) is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power level operation. During reactor operation with certain limiting control rod patterns, the continuous withdrawal of a designated single control rod could result in a violation of the MCPR safety limit or the 1t plastic strain limit.

Amendment No. b-r,, 34, 6., 4GI, .64, BWA 99 55 90

VYNPS BASES:

4.11 FUEL RODS A. The APLHGR, LHGR and MCPR shall be checked daily when operating at

>25% Rated Thermal Power to determine if fuel burnup, or control rod movement has caused changes in power distribution. Since changes due to burnup are slow, and only a few control rods are removed daily, a daily check of power distribution is adequate. For a limiting value to occur below 25% of rated thermal power, an unreasonably large peaking factor would be required, which is not the case for operating control rod sequences. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance after thermal power >25%

Rated Thermal Power is achieved is acceptable given the large inherent margin to operating limits at low power levels.

B. At certain times during plant startups and power changes the plant technical staff may determine that surveillance of APLHGR, LHGR and/or MCPR is necessary more frequently than daily. Because the necessity for such an augmented surveillance program is a function of a number of interrelated parameters, a reasonable program can only be determined on a case-by-case basis by the plant technical staff. The check of APLHGR, LHGR and MCPR will normally be done using the plant process computer. In the event that the computer is unavailable, the check will consist of either a manual calculation or a comparison of existing core conditions to those existing at the time of a previous check to determine if a significant change has occurred.

If a reactor power distribution limit is exceeded, an assumption regarding an initial condition of the DBA analysis, transient analyses, or the fuel design analysis may not be met. Therefore, prompt action should be taken to restore the APLHGR, LHGR or MCPR to within the required limits such that the plant operates within analyzed conditions and within design limits of the fuel rods. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months completion time is sufficient to restore the APLHGR, LHGR, or MCPR to within its limits and is acceptable based on the low probability of a transient or DBA occurring simultaneously with the APLHGR, LHGR, or MCPR out of specification.

C. Minimum Critical Power Ratio (MCPR) - Surveillance Requirement At core thermal power levels less than or equal to 25%, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience indicated that the resulting MCPR value is in excess of requirements by a considerable margin. With this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR. During initial start-up testing of the plant, a MCPR evaluation will be made at 25t thermal power level with minimum recirculation pump speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary. The daily requirement for calculating MCPR above 25% rated thermal power is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes.

Amendment No. 4&, 42& 228

Current Technical Specifications Markups 3.2.E and 4.2.E Control Rod Block Actuation And related Technical Specifications:

3.3 - Control Rod System 3.11 - Reactor Fuel Assemblies

VYNPS 3.2 LIMITING CONDITIONS FOR 4.2 SURVEILLANCE REQUIREMENTS OPERATION L

r D. Off-Gas System Isolation D. Off-Gas System Isolation During reactor power Instrumentation and logic operation, the systems shall be instrumentation that functionally tested and _0 1 initiates isolation of the calibrated as indicated i off-gas system shall be Table 4.2.4.

operable in accordance with Table 3.2.4.

--- y At, tab Cd^"* A*4 6loc E. Control Rod Block Actuation E. Control Rod Block Actuation Ajnstrumentatio functionally tested and calibrated as indicated in Table 4.2.5.

allJ) in accordance with P/Table 3.2.5.

. Mechanical Vacuum Pump F. Mechanical Vacuum Pump I Isolation Instrumentation Isolation Instrumentation When the reactor is in the The High Main Steam Line RUN or STARTUP/HOT STANDBY Radiation Trip Function for Mode and the mechanical mechanical vacuum pump vacuum pump is in service, isolation shall be checked, four (4) channels of the functionally tested and High Main Steam Line calibrated as indicated in Radiation Trip Function for Surveillance Requirements mechanical vacuum pump 4.2.F.1, 2, 3, 4 and 5.

isolation shall be operable, except as provided below. When a channel is placed in an inoperable status solely

1. With one or more channels for performance of required inoperable, within 12 surveillances, entry into hours: associated Limiting Conditions for Operation and

a. Restore the inoperable required actions may be channel(s) to operable delayed for up to six (6) status; or hours provided the associated trip function maintains

b. Place the inoperable mechanical vacuum pump channel(s) or isolation capability.

associated trip system in the trip condition 1. Perform an instrument (not applicable if the check once each day.

inoperable channel is the result of an 2. Perform an instrument inoperable mechanical functional test once every vacuum pump isolation three (3) months.

valve).

IC we __C##. #+reAe C" .4 94o.CA: A4 ,PJ,1V4 CJ/4PACan6L /a &t1Ceb i't i 4

/jNPJ64464Z xlrT* %J 04" 4-Yj~d Pd~jrA 4cAq"jC'd4 .Eowdbo up ToJ#6A PAOVI&cowrebTwe r~l FVJT,*

Amend rnent No. .5, 164, 21 dT4E b &8LOCj 1 4 ~*jflO7*d CAPAO'u-

'RoAAJ

VYNPS TABLE 3.2.5 CONTROL ROD BLOCK INSTRUMENTATION Modes in Which I Required Must be Oper Channels Trip Function Refuel Startup 71*I.. Trip Setting Rod Block Monitor l0 (Notes 2 a. Upscale (Flow Bias) <0.66(W-AW)+N with a maximum as defined in the COLRNo 5 1.4 2 b. >2/125 ii:Scale

c. Inop

,a (Note 2 Reactor Mode Switch - Shutdown Position ZooR Trip Sys W x

Amendment No. 44, a&, 64, 66, 7s, 6, g4, 4, 1-3-1, 464, 48e, 487, 211 51

VYNPS TABLE 3.2.5 NOTES 4 _+/-leted. /

0W" is perce rated two ioop drive ow where 100% rated dr e flow is that flow uivalent to 48 x 106 lb )/hr core flow. Refer the Core Operating/imits Report for accep)able values for N. AW s the differenc 3* betweeBr he two loop and single oop drive flow at the ame core flow.

F'oJmferri 1,* This Xfference must be accou ed for during single op operatio W = 0 rP or two recirculation loop operation.

lT tp b assed when the reactor power is <30% of Rated Therma PowerMca w r be considered inoprab theotal number oVnormal inputs from X y LPRM level.

ha (8. the~n~ber of operable c inels less than the 'quired number, placet With l

(8. the i erable channel in stripped condition , hin one hour.

MT X With one or two RBM channels inoperable X4. Ab 1.6 #d?.D VerifyatUa. the reactor is i operating on a fli0ting control ro patt5Kn (as described in_oe Bases for S ecipfecation 3.3.B.6); anv Mdrs I-L.

EsG If one RBM channel is inoperable, restore the inoperable channel to

'J6T6 e operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; and Aj,1; a3 If the requir dActior4 and associated completion times of Noteo.a tgjj? above not met, or if two RBM channels are inoperable, place one RBM channel in the tripped condition within the next hour.

c When a channel is placed in an inoperable status solely for performance of required surveillances, entry into associated Limiting Conditions for Operation and required action notes may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Trip Function maintains Control Rod Block initiation capability.

1i2. Required to be operable when the reactor mode switch is in the shutdown position. I' O With one or more Reactor Mode Switch - Shutdown Position channels ic.Id; inoperable immediately suspend control rod withdrawal and immediately A d1F 4t initiate Actio Ano fully insert all insertable control rods in core cells containing one or more fuel assemblies.

Amendment No. 64, *, As, -g, 94, a4, , 211 52

VYNPS TABLE 4.2.5 CCIP ) TE)9 BR m FREQUENCIES CONTROL ROD BLOCK INSTRUMENTATION Trip Function Functional Test Calibration Rod Block Monitor (RB")

Every Three Months Every Three Months

.4 1 a. Upscale (Flow Bias)

Every Three Months

b. Downscale Every Three Months F.c I c. Inop Every Three Months System Logic Once/Operating Cycle 129 Reactor Mode Switch -

Position Shutdown Every Refueling Outage (Note 12)

\

(A) FS3v-Ud4J )Cxc(U6b 4>eT-e*Sde A Amendment No. 146, a10, 36, .81-, 211 69

VYNPS

'B!

TABLE 4.2 NOTES Ca /se~sd.

fueling outage, simulated automatic actuation all pilot valves shall be performed such that each trip system logic can be verified independent of its redundant counterpart.

3 tim necessary for proper f ctioning of the tr system.

4. This inst mentation is except from functional te definition. The functioW 1 test will consist f injecting a simul ed electrical signal into > e measurement chan K .

I5\. tededt nctional tests and calibrations are not required when systems are not required to be operable.

9. The thermocouples associated with safety/relief valves and safety valve position, that may be used for back-up position indication, shall be verified to be operable every operating cycle.

10. Separate functional tests are not required for this instrumentation. The calibration and integrated ECCS tests which are performed once per operating cycle will adequately demonstrate proper equipment operation.

11. Trip system logic functional tests will include verification of operation of all automatic initiation inhibit switches by monitoring relay contact movement. Verification that the manual inhibit switches prevent opening all relief valves will be accomplished in conjunction with Section 4.5.F.l. ~f 0 TpB m o s not licale to s funct If the required surveillance frequency (every Refueling Outage) is not met, functional testin of the Reactor Mode Switch-Shutdown Position function shall te within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after the reactor mode switch is placed ienfueling Outage #

Qt~Iupscale inicludes calibr ion of the RBM Refe func Downscale funct n (i.e., RBM) chce n is not bypassed w n >30% Rated Therm Power).

Amendment No. 6&, 46, SaG, 445, a64, e84, 211, 212 74

VYNPS 3.3 LIMITING CONDITIONS FOR 4.3 SURVEILLANCE REQUIREMENTS OPERATION pressure, are fully inserted, no more than two rods may be moved.

(c) Out-of-sequence control rods in each distinct RWM group shall be selected and the annunciator of the selection errors verified.

(d) An out-of-sequence control rod shall be withdrawn no more than three notches and the rod block function verified.

4. Control rod patterns and 4. The control rod pattern the sequence of and sequence of withdrawal or insertion withdrawal or insertion shall be established shall be verified to such that the rod drop comply with accident limit of Specification 3.3.B.4.

280 cal/g is not exceeded.

5. Control rods shall not 5. Prior to control rod be withdrawn for startup withdrawal for startup or refueling unless at or during refueling, least two source range verification shall be channels have an made that at least two observed count rate source range channels greater than or equal to have an observed count three counts per second. rate of at least three counts per second.

n a imiting control rod pa ern ex dts, an inst ent f ctional tea of the.?&M shalom formed ror tort withdraw of the/

designaied rod(s) an Ldaily thereafter.

Amendment No. PG, 61 84

VYNPS 3.3 LIMITING CONDITIONS FOR 4.3 SURVEILLANCE REQUIREMENTS

7. The scram discharge volume drain and vent valves shall be verified open at least once per month. These valves may be closed intermittently for testing under administrative control.

C. Scram Insertion Times 1.1 The average scram time, 1. After refueling outage based on the and prior to operation de-energization of the above 30% power with scram pilot valve reactor pressure above solenoids of all 800 psig all control operable control rods in rods shall be subject to the reactor power scram-time measurements operation condition from the fully withdrawn shall be no greater position. The scram than: times for single rod scram testing shall be Drop-Out % Inserted Avg. Scram measured without of From Fully Insertion reliance on the control Position Withdrawn Time (sec) rod drive pumps.

46 4.51 0.358 2. During or following a 36 25.34 0.912 controlled shutdown of 26 46.18 1.468 the reactor, but not 06 87.84 2.686 more frequently than 16 weeks nor less The average of the scram frequently than 32 weeks insertion times for the intervals, 50% control three fastest control rod drives in each rods of all groups of quadrant of the reactor four control rods in a core shall be measured two by two array shall for scram times be no greater than: specified in Specification 3.3.C.

Drop-Out % Inserted Avg. Scram All control rod drives of From Fully Insertion shall have experienced Position Withdrawn Time (sec) scram-time measurements each year. Whenever 50%

46 4.51 0.379 of the control rod 36 25.34 0.967 drives scram times have 26 46.18 1.556 been measured, an 06 87.84 2.848 evaluation shall be made to provide reasonable assurance that proper control rod drives performance is being maintained. The results of measurements performed on the control rod drives shall be submitted in the start up test report.

Amendment No. *4, aS, -3, 4, 73 85

VYNPS BASES: 3.3 & 4.3 (Cont'd)

2. The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the extremely remote event of a housing failure. The amount of reactivity which could be added by this small amount of rod withdrawal, which is less than a normal single withdrawal increment, will not contribute to any damage of the primary coolant system. The design basis is given in Subsection 3.5.2 of the FSAR, and the design evaluation is given in Subsection 3.5.4. This support is not required if the reactor coolant system is at atmospheric pressure since there would then be no driving force to rapidly eject a drive housing.

3. In the course of performing normal startup and shutdown procedures, a pre-specified sequence for the withdrawal or insertion of control rods is followed. Control rod dropout accidents which might lead to significant core damage, cannot occur if this sequence of rod withdrawals or insertions is followed. The Rod Worth Minimizer restricts withdrawals and insertions to those listed in the pre-specified sequence and provides an additional check that the reactor operator is following prescribed sequence. Although beginning a reactor startup without having the RWM operable would entail unnecessary risk, continuing to withdraw rods if the RWM fails subsequently is acceptable if a second licensed operator verifies the withdrawal sequence. Continuing the startup increases core power, reduces the rod worth and reduces the consequences of dropping any rod. Withdrawal of rods for testing is permitted with the RWM inoperable, if the reactor is subcritical and all other rods are fully inserted. Above 20% power, the RWM is not needed since even with a single error an operator cannot withdraw a rod with sufficient worth, which if dropped, would result in anything but minor consequences.

4. Refer to the "General Electric Standard Application for Reactor Fuel (GESTAR II)," NEDE-24011-P-A, (the latest NRC-approved version will be listed in the COLR).

5. The Source Range Monitor (SRM) system provides a scram function in noncoincident configuration. It does provide the operator with a visual indication of neutron level. The consequences of reactivity accidents are a function of the initial neutron flux. The requirement of at least three counts per second assures that any transient, should it occur, begins at or above the initial value of 10-8 of rated power used in the analyses of transients from cold conditions. One operable SRM channel is adequate to monitor the approach to criticality, therefore, two operable SRM's are specified for added conservatism.

6. The Rod Block Monitor (RBM) is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power level operation. During reactor operation with certain limiting control rod patterns, the continuous withdrawal of a designated single control rod could result in a violation of the MCPR safety limit or the 1i plastic strain li A- 13,M3.tl control roa pattern is a pattern whic rodusei ein hen te patteg on a thermal limyt(i.e., operating th limiting value four -L , LHGR, or MCPR During use of such patterns, it is ,Ridgedthat testing oj!the RBM system priorK withdral of Rich rods will provi eaded assurance that dmpoer withdrawa~l OX asnot occur. It i6h responsibityo/h Nuclear Eninet detfy these. lim'ing patterns an thXeinted ros itir he he pattern are initially estalse or as they develo due to the occurre e of inoperable con ol rods.

Amendment No. a&, 3-, i6, ;4, ;64-, EW 99 -S, 211 90

VYNPS 3.11 LIMITING CONDITIONS FOR 4.11 SURVEILLANCE REQUIREMENTS OPERATION C. Minimum Critical Power Ratio C. Minimum Critical Power Ratio (MCPR) (MCPR)

1. During operation at >25% MCPR shall be determined Rated Thermal Power the once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after I MCPR operating value >25% Rated Thermal Power, shall be equal to or aaily during operation at greater than the MCPR >25% Rated Thermal Power limits provided in the Thereafter a lowing Core Operating Limits ___nge i power evel or Report. For single distributi that Gduld recirculation loop cause ope tion wih a operation, the MCPR limitin contro> rod pattrn Limits at rated flow are as des ribed id the basds also provided in the Core Operating Limits Report. For core flows other than rated, the Operating MCPR Limit shall be the above value multiplied by K where Kf is provided fn the Core Operating Limits Report. If at any time during operation at >25%

I Rated Thermal Power It is determined by normal surveillance that the limiting value for MCPR is being exceeded, MCPR(s) shall be returned to within the prescribed limits within two (2) hours; otherwise, the reactor power shall be brought to <25% Rated Thermal Power within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.

Amendment No. 4A, 44, -3, 04, 444, 188 226

VYNPS BASES:

4.11 FUEL RODS A. The APLHGR, LHGR and MCPR shall be checked daily when operating at

>25% Rated Thermal Power to determine if fuel burnup, or control rod movement has caused changes in power distribution. Since changes due to burnup are slow, and only a few control rods are removed daily, a daily check of power distribution is adequate. For a limiting value to occur below 25% of rated thermal power, an unreasonably large peaking factor would be required, which is not the case for operating control rod sequences. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance after thermal power >25%

Rated Thermal Power is achieved is acceptable given the large inherent margin to operating limits at low power levels.

B. At certain times during plant startups and power changes the plant technical staff may determine that surveillance of APLHGR, LHGR and/or MCPR is necessary more frequently than daily. Because the necessity for such an augmented surveillance program is a function of a number of interrelated parameters, a reasonable program can only be determined on a case-by-case basis by the plant technical staff. The check of APLHGR, LHGR and MCPR will normally be done using the plant process computer. In the event that the computer is unavailable, the check will consist of either a manual calculation or a comparison of existing core conditions to those existing at the time of a previous check to determine if a significant change has occurred.

If a reactor power distribution limit is exceeded, an assumption regarding an initial condition of the DBA analysis, transient analyses, or the fuel design analysis may not be met. Therefore, prompt action should be taken to restore the APLHGR, LHGR or MCPR to within the required limits such that the plant operates within analyzed conditions and within design limits of the fuel rods. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months completion time is sufficient to restore the APLHGR, LHGR, or MCPR to within its limits and is acceptable based on the low probability of a transient or DBA occurring simultaneously with the APLHGR, LHGR, or MCPR out of specification.

C. Minimum Critical Power Ratio (MCPR) - Surveillance Requirement At core thermal power levels less than or equal to 25%, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience indicated that the resulting MCPR value is in excess of requirements by a considerable margin. With this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR. During initial start-up testing of the plant, a MCPR evaluation will be made at 25% thermal power level with minimum recirculation pump speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary. The daily requirement for calculating MCPR above 25% rated thermal power is sufficient since power distribution shifts are very slow when there have not been significant power or rod ^hnnmes. T t._

u ating MCPR when a imiting contr rod pattern s approacd ensures Wat MCPR wil e known fol ing a change power o ower shape {egardless o magnitude) th could lace eration atta therml limit.

Amendment No. 48, 188 228

Safety Assessment Discussion of Changes 3.2.E and 4.2.E Control Rod Block Actuation

SAFETY ASSESSMENT OF CHANGES TS 3.2.E/4.2.E - CONTROL ROD BLOCK INSTRUMENTATION ADMINISTRATIVE A.1 Inthe revision of the Vermont Yankee Nuclear Power Station (VYNPS) current Technical Specifications (CTS), certain wording preferences or conventions are adopted which do not result intechnical changes (either actual or interpretational). Editorial changes, reformatting, and revised numbering are adopted to make the WNPS Technical Specifications (TS) more consistent with human factor principles used inthe Boiling Water Reactor Improved Standard Technical Specifications (ISTS), NUREG-1433, Rev. 2. These format and presentation changes are being made to improve usability and clarity. The changes are considered administrative.

A.2 CTS 3.2.D, 4.2.D, Table 3.2.5, Table 4.2.5 and associated Notes provide requirements that apply to off-gas isolation instrumentation. Changes these CTS off-gas isolation instrumentation requirements are addressed inthe Safety Assessments of Changes for CTS 3.2.D14.2.D, Off-Gas Isolafion Instrumentation. Therefore, this change does not involve a technical change, but isonly a difference of presentation preference and is considered administrative.

A.3 CTS 3.2.E indicates that, during power operation, the control rod block instrumentation shall be operable in accordance with Table 3.2.5. The Rod Block Monitor (RBM) Trip Functions (proposed Table 3.2.5, Trip Functions I.a, 1.band 1.c)are required by CTS Table 3.2.5 to be operable in the Run Mode. CTS Table 3.2.5 Note 7 (first sentence) applies to these Trip Functions and indicates that these Trip Functions may be bypassed when reactor power is < 30% of Rated Thermal Power. The intent of this note is to waive the operability requirements of the RBM Trip FuncUons when reactor power Is< 30%. The

'APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS' column inproposed Table 3.2.5 requires the RBM Trip Functions to be operable when reactor power Is> 30%

RATED THERMAL POWER which is equivalent to CTS requirements. For the Reactor Mode Switch - Shutdown Position Trip Function, CTS Table 3.2.5 Note 12 modifies the requirements and indicates that this Trip Function is required to be operable when the reactor mode switch is inthe shutdown position. The 'APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS' column in proposed Table 3.2.5 requires the Reactor Mode Switch - Shutdown Position to be operable as idenUfied in proposed Table 3.2.5 Footnote

1. This Footnote states, 'When the reactor mode switch is Inthe shutdown position. As such, these changes are considered administrative in nature.

A.4 CTS 4.2.E specifies that instrumentation and logic systems shall be functionally tested and calibrated as Indicated InTable 4.2.5. Inproposed Surveillance Requirement (SR) 4.2.E.1, the reference to 'and logic system,' is deleted since associated logic systems are considered part of the control rod block instrumentation Trip Functions included in proposed Tables 3.2.5 and 4.2.5. Inaddition, proposed Table 3.2.5 will delete explicit reference to 'Trip System Logic' as a separate Trip Function. The requirement that the

'Trip System Logic' be operable is required by both the definition of Operable and the TS requirements for the control rod block instrumentation Trip FuncUons to be operable, without explicit reference to 'Trip System Logic.' The control rod block instrumentation design at WNPS only includes one 'Trip System Logic.' Therefore, when the 'Trip System VYNPS 1 Revision 0

SAFETY ASSESSMENT OF CHANGES TS 3.2.E/4.2.E - CONTROL ROD BLOCK INSTRUMENTATION ADMINISTRATIVE A.4 Logic' is inoperable, both RBM channels would be inoperable and proposed Table 3.2.5 (continued) Action Note 1.b, would require one channel to be trip in one hour, which is equivalent to the actions inCTS Table 3.2.5 Note 8 (which is also proposed to be deleted as part of this change). For the purpose of testing, it isnot necessary to explicitly identify 'Trip System Logic" in CTS Table 4.2.5 for this design, since proposed Table 4.2.5 continues to require performance of surveillance testing of the 'Trip System Logic through the requirements for performance of Instrument Functional Tests. Therefore, this change does not involve a technical change, but is only a difference of presentation preference and is considered administrative.

A.5 CTS Table 3.2.5 Note 10 provides an allowance to delay entry into actions for 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for the situation of a RBM channel inoperable solely for performance of surveillances. This allowance is moved to proposed SR 4.2.E.1. This change does not involve a technical change, but is only a difference of presentation preference. Therefore, this change Is considered administrative.

A.6 Notes 9 and 13 to CTS Table 3.2.5 provide actions when the required channels requirement isnot met for the associated Trip Functions. These requirements are identified ina separate column inproposed Table 3.2.5 titled, 'ACTIONS WHEN REQUIRED CHANNELS ARE INOPERABLE," for each of the associated Trip Functions This change is a presentation preference and does not alter the current action requirements when required control rod block instrumentation channels are inoperable. Therefore, this change is considered administrative innature.

A.7 For the Trip System Logic associated with the RBM control rod block instrumentation, CTS Table 4.2.5 includes requirements to perform a calibration of Trip System Logics once per Operating Cycle. These requirements are modified by Table 4.2 Note 3. Note 3 states,

'Trip system logic calibration shall include only time delay relays and timers necessary for proper functioning of the trip system." The control rod block instrumentation Trip Functions of CTS Table 4.2.5 do not include any time delay relays or timers necessary for proper functioning of the trip systems. Therefore, this Note isdeleted and, in proposed Table 4.2.5, the control rod block Instrumentation (proposed Trip Functions l.a, .b,1.c, and 2)do not include calibration requirements for time delay relays or timers. As a result, this change does not Involve a technical change, but is only a difference of presentation preference and isconsidered administrative.

A.8 CTS Table 4.2.5 includes Functional Test requirements for the RBM Upscale and Downscale Trip Functions. These requirements are modified by CTS Table 4.2 Note 4.

Note 4 states, 'This instrumentation is excepted from functional test definition. The functional test will consist of injecting a simulated electrical signal into the measurement channel.' The definition of Instrument Functional Test for this type of instrumentation (CTS I.0.G.1) is,'the injection of a signal into the channel as close to the sensor as practicable to verify operability including alarm and/or trip functions.' The requirements of CTS Table VYNPS 2 Revision 0

SAFETY ASSESSMENT OF CHANGES TS 3.2.E/4.2.E - CONTROL ROD BLOCK INSTRUMENTATION ADMINISTRATIVE A.8 4.2 Note 4 are consistent with the requirements of the Instrument Functional Test definition.

(continued) The CTS definition of Instrument Functional Test allows the method of testing described in CTS Table 4.2 Note 4 to be used. Therefore, CTS Table 4.2 Note 4 is unnecessary and is deleted. This change does not involve a technical change, but is only a difference of presentation preference and Isconsidered administrative.

A.9 CTS Table 4.2 Notes 2,8, 10, and 11 provide requirements that apply to ECCS instrumentation. The ECCS instrumentation is located inproposed Specifications 3.2.A and 4.2A. Therefore, the requirements of CTS Table 4.2 Notes 2, 8,10, and 11 are physically moved and addressed inthe changes for proposed Specifications 3.2.A and 4.2.A. Therefore, this change does not involve a technical change, but isonly a difference of presentation preference and is considered administrative. CTS Table 4.2 Note 9 provides requirements that apply to post-accident monitoring instrumentation. The post-accident monitoring Instrumentation is located inproposed Specifications 3.2.G and 4.2.G.

Therefore, the requirements of CTS Table 4.2 Note 9 are physically moved and changes addressed inproposed Specifications 3.2.G and 4.2.G. Therefore, this change does not involve a technical change, but isonly a difference of presentation preference and is considered administrative.

TECHNICAL CHANGES - MORE RESTRICTIVE M.1 CTS Table 4.2 Note 12 applies to the Reactor Mode Switch - Shutdown Position Trip Function. This note allows the Functional Test of the Reactor Mode Switch - Shutdown Position Trip Function to be initiated with 1hour after the reactor mode switch inplaced in shutdown. This note does include a time limit on completion of the Functional Test. Since testing of this Trip Function with the reactor mode switch in any other position cannot be performed without using jumpers, lifted leads, or movable links, the intent 6f this allowance is to ensure the required Functional Test iscompleted ina timely manner as soon as plant conditions exist to perform the test. Therefore, in proposed Table 4.2.5 Footnote (1), the allowance is revised by requiring the Functional Test of this Trip Function to be performed (i.e., completed) within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after the reactor mode switch is placed inthe shutdown position. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months allowance continues to provide a reasonable time inwhich to complete the required Functional Test. This change represents an additional restriction on plant operation necessary to ensure that the Functional Test issatisfactorily completed ina timely manner. The change isconsistent with the ISTS.

VYNPS 3 Revision 0

SAFETY ASSESSMENT OF CHANGES TS 3.2.E/4.2.E - CONTROL ROD BLOCK INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE

'Generic' LA.1 The CTS Table 3.2.5 details relating to system design and operation (i.e., the specific instrument tag numbers) are unnecessary in the TS and are proposed to be relocated to the Technical Requirements Manual (TRM). Proposed Specification 3.2.E and Table 3.2.5 require the control rod block instrumentation Trip Functions to be operable. Inaddition, the proposed Surveillance Requirements inTable 4.2.5 ensure the required instruments are properly tested. These requirements are adequate for ensuring each of the required control rod block instrumentation Trip Functions are maintained operable. As such, the relocated details are not required to be inthe WNPS TS to provide adequate protection of the public health and safety. Changes to the TRM are controlled by the provisions of 10 CFR 50.59. Not including these details inTS isconsistent with the ISTS.

LA.2 The LPRM inputs for operability of the RBM are relocated to Specification 3.2.E Bases.

The Specification 3.2.E Bases indicates that if sufficient LPRMs are not available (the same requirements as specified inCTS Table 3.2.5, Note 7,second sentence), then the associated RBM is inoperable. As such, CTS Table 3.2.5 Note 7,second sentence, is not necessary in WNPS TS control rod block instrumentation Table 3.2.5. The definition of operability suffices. Therefore, the relocated details of the note are not required to be In the WNPS TS to provide adequate protection of the public health and safety. Changes to the TS Bases are controlled by the provisions of 10 CFR 50.59. Not including these details inTS isconsistent with the ISTS.

LA.3 CTS Table 3.2.5 Note 5 contains design and operational details of the RBM Upscale (Flow Bias) Trip Function Trip Setting for two recirculation loop and single recirculation loop operation. These details are not necessary to ensure the operability of the RBM Upscale (Flow Bias) Trip Function. Therefore, the information inthis note isto be relocated to Specification 3.2.E Bases and reference to this Information is deleted from WNPS TS.

The requirements of Specification 3.2.E and associated Table 3.2.5 which includes RBM Upscale (Flow Bias) Trip Function Trip Settings for both two recirculation loop and single recirculation loop operation are adequate to ensure the RBM Upscale (Flow Bias) Trip Function is maintained operable. As such, these relocated requirements are not required to be inthe WNPS TS to provide adequate protection of the public health and safety.

Changes to the TS Bases are controlled by the provisions of 10 CFR 50.59. Not including these details inTS is consistent with the ISTS.

LA.4 CTS Table 4.2.5 and associated Note 13 describe details of the performance of Instrument Calibrations of the RBM Upscale (Flow Bias) Trip Function. These details are to be relocated to Bases. These details are not necessary to ensure the operability of the RBM Upscale (Flow Bias) Trip Function instrumentation. The VYNPS TS definition of Instrument Calibration, the requirements of proposed Specification 3.2.E, and the associated Surveillance Requirements (including the requirements to periodically perform Instrument Calibrations) are adequate to ensure the RBM Upscale (Flow Bias) Trip Function instrumentation is maintained operable. As such, these relocated details are not required VYNPS 4 Revision 0

SAFETY ASSESSMENT OF CHANGES TS 3.2.E/4.2.E - CONTROL ROD BLOCK INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE LA.4 to be inthe VYNPS TS to provide adequate protection of the public health and safety.

(continued) Changes to the Bases are controlled by the provisions of 10 CFR 50.59.

'Specific" L.1 CTS Table 3.2.5 Note 9.a, which requires verification that the reactor is not operating on a limiting control rod pattern when one or two RBM channels are inoperable, CTS 3.3.B.6, which provides additional RBM requirements when operating on a limiting control rod pattem, CTS 4.3.B.6, which requires a Functional Test of the RBM prior to control rod withdrawal and daily thereafter when the reactor is operaUng on a limiting control rod pattern and the CTS 4.11 .Crequirement to determine Minimum Critical Power Ratio (MCPR) following any change inpower level or distribution that would cause operation on a limiting control rod pattem, have been deleted. Corresponding changes are also made to the associated Bases. Since a limiting control rod pattern isoperation on a power distribution limit (such as Average Planar Unear Heat Generation Rate or MCPR), the condition is extremely unlikely. It Isalso highly unlikely that, when operating on a power distribution limit, the operator would withdraw a control rod. The status of power distribution limits does not affect the operability of the RBM and therefore, no additional requirements on the RBM System are necessary (e.g., that it be tripped within one hour with a channel inoperable while on a limiting control rod pattern or that both RBM channels be operable while on a limiting control rod pattem). Adequate requirements on power distribution limits are specified inthe Specifications inTS 3.11/4.1 1,Reactor Fuel Assemblies, without the need for additional determination of MCPR when operating on a limiting control rod pattem.

Furthermore, due to the improbability of operating exactly on a thermal limit, the CTS Actions and Surveillance Requirements would almost never be required. Inaddition, since the Surveillance Requirements inCTS 4.3.B.6 and CTS 4.11 .Care not specific as to when

'prior to' or 'following' operation on a limiting control rod pattem, respectively, they are required to be performed, these two frequency requirements could be satisfied by the normal periodic Surveillances. The additional requirements associated with operation on a limiting control rod pattern are not necessary to ensure that the RBM channels are maintained operable or that power distribution parameters are maintained within limits.

Therefore, the deletion of the CTS requirements associated with operation on a limiting control rod pattern is not safety significant. This change is consistent with the ISTS.

L.2 This change adds a note to the RBM Instrument Calibrations in CTS Table 4.2.5 excluding the neutron detectors from this Surveillance (proposed Table 4.2.5 Footnote (2)). The Instrument Calibration is a complete check of the instrument loop and the sensor. The test verifies that the channel responds to the measured parameter within the necessary range and accuracy. The neutron detectors are excluded from the RBM Instrument Calibration because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal. Inaddition, changes inneutron detector sensitivity are compensated for by performance of the 7 day heat balance calibration and the 2000 MWD/T Local Power Range Monitor calibration using the Traversing Incore Probe System of the VYNPS Reactor Protection System TS. This change is consistent with the ISTS.

VYNPS 5 Revision 0

SAFETY ASSESSMENT OF CHANGES TS 3.2.E/4.2.E - CONTROL ROD BLOCK INSTRUMENTATION RELOCATED SPECIFICATIONS None VYNPS 6 Revision 0

No Significant Hazards Consideration 3.2.E and 4.2.E Control Rod Block Actuation

GENERIC NO SIGNIFICANT HAZARDS CONSIDERATION ADMINISTRATIVE CHANGES

('A.x' Labeled Comments/Discussions)

Inaccordance with the criteria set forth in 10 CFR 50.92, Vermont Yankee Nuclear Power Station has evaluated this proposed Technical Specifications change and determined it does not represent a significant hazards consideration. The following is provided insupport of this conclusion.

1. Does the change Involve a significant increase inthe probability or consequences of an accident previously evaluated?

The proposed change involves reformatting, renumbering, and rewording. The reformatting, renumbering, and rewording process involves no technical changes to the existing Technical Specifications. As such, this change is administrative in nature and does not impact initiators of analyzed events or assumed mitigation of accident or transient events. Therefore, this change does not involve a significant Increase In the probability or consequences of an accident previously evaluated.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be Installed) or changes inmethods governing normal plant operation. The proposed change will not impose any new or eliminate any old requirements. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction ina margin of safety?

The proposed change will not reduce a margin of safety because it has no impact on any safety analyses assumptions. This change is administrative innature. Therefore, the change does not involve a significant reduction ina margin of safety.

VYNPS 1 Revision 0

GENERIC NO SIGNIFICANT HAZARDS CONSIDERATION TECHNICAL CHANGES - MORE RESTRICTIVE

('M.x' Labeled Comments/Discussions)

Inaccordance with the criteria set forth in 10 CFR 50.92, Vermont Yankee Nuclear Power Station has evaluated this proposed Technical Specifications change and determined it does not represent a significant hazards consideration. The following is provided in support of this conclusion.

1. Does the change involve a significant increase inthe probability or consequences of an accident previously evaluated?

The proposed change provides more stringent requirements for operation of the facility. These more stringent requirements do not result inoperation that will increase the probability of initiaUng an analyzed event and do not alter assumptions relative to mitigation of an accident or transient event. The more restrictive requirements continue to ensure process variables, structures, systems, and components are maintained consistent with the safety analyses and licensing basis. Therefore, this change does not involve a significant increase inthe probability or consequences of an accident previously evaluated.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in the methods governing normal plant operation. The proposed change does impose different requirements. However, these changes are consistent with the assumptions Inthe safety analyses and licensing basis. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction ina margin of safety?

The imposition of more restrictive requirements either has no impact on or increases the margin of plant safety. As provided inthe discussion of the change, each change inthis category is by definition, providing additional restrictions to enhance plant safety. The change maintains requirements within the safety analyses and licensing basis. Therefore, this change does not involve a significant reduction in a margin of safety.

VYNPS 3 Revision 0

GENERIC NO SIGNIFICANT HAZARDS CONSIDERATION "GENERIC LESS RESTRICTIVE CHANGES:

RELOCATING DETAILS TO TECHNICAL SPECIFICATION BASES, UFSAR, PROCEDURES, OR OTHER PLANT CONTROLLED DOCUMENTS FLA.x Labeled Comments/Discussions)

Inaccordance with the criteria set forth in 10 CFR 50.92, Vermont Yankee Nuclear Power Station has evaluated this proposed Technical SpecificaUons change and determined it does not represent a significant hazards consideraUon. The following Isprovided insupport of this conclusion.

1. Does the change Involve asignificant increase inthe probability or consequences of an accident previously evaluated?

The proposed change relocates certain details from the Technical Specifications to the Bases, UFSAR, procedures, or other plant controlled documents. The Bases, UFSAR, procedures, and other plant controlled documents containing the relocated information will be maintained in accordance with 10 CFR 50.59. The UFSAR is subject to the change control provisions of 10 CFR 50.71(e), and the plant procedures and other plant controlled documents are subject to controls imposed by plant administrative procedures, which endorse applicable regulations and standards. Since any changes to the relocated details in the Bases, UFSAR, procedures, or other plant controlled documents will be evaluated per the requirements of 10 CFR 50.59, no significant increase inthe probability or consequences of an accident previously evaluated will be allowed. Therefore, this change does not involve a significant Increase in the probability or consequences of an accident previously evaluated.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The proposed change will not impose or eliminate any requirements, and adequate control of the information will be maintained. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The proposed change will not reduce a margin of safety because it has no impact on any safety analysis assumptions. Inaddition, the details to be transposed from the Technical Specifications to the Bases, UFSAR, procedures, or other plant controlled documents are the same as the existing Technical Specifications. Since any future changes to these details inthe Bases, UFSAR, procedures, or other plant controlled documents will be evaluated per the requirements of 10 CFR 50.59, no significant reduction in a margin of safety will be allowed. Based on 10 CFR 50.92, the existing requirement for NRC review and approval of revisions, to these details proposed for relocation, does not have a specific margin of safety upon which to evaluate. However, since the proposed change is consistent with the BWR Standard Technical Specifications, NUREG-1433, approved by the NRC Staff, revising the Technical Specifications to reflect the approved level of detail ensures no significant reduction inthe margin of safety.

VYNPS 4 Revision 0

NO SIGNIFICANT HAZARDS CONSIDERATION TS 3.2.E/4.2.E - CONTROL ROD BLOCK INSTRUMENTATION L.1 CHANGE Inaccordance with the criteria set forth in 10 CFR 50.92, Vermont Yankee Nuclear Power Station has evaluated this proposed Technical Specifications change and determined it does not represent a significant hazards consideration. The following is provided insupport of this conclusion.

I1. Does the change involve a significant increase Inthe probability or consequences of an accident previously evaluated?

The proposed change deletes the additional Rod Block Monitor (RBM) and Minimum Critical Power Ratio (MCPR) action and surveillance requirements associated with operating on a limiting control rod pattem. The RBMs and MCPR are not assumed to be initiators of any analyzed event.

During operation on a limiting control rod pattem, power distribution parameters continue to meet acceptable required limits. Inaddition, the status of power distribution limits (i.e., operation on a limiting control rod pattem) does not affect the operability or reliability of the RBMs. Technical Specifications continue to provide assurance that power distribution parameters are maintained within limits and the RBMs are maintained capable of performing their required function. As a result, the consequences of an accident occurring with the proposed change are the same as the consequences occurring with the current requirements. Therefore, the proposed change does not involve asignificant increase inthe probability or consequences of an accident previously evaluated.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not necessitate a physical alteration of the plant (no new or different type of equipment will be installed) or changes inparameters governing normal plant operation.

The proposed change continues to ensure the affected instrumentation iscapable of performing its function as assumed in the safety analyses and that power distribution parameters are maintained within required limits. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

Deleting the additional Rod Block Monitor (RBM) and Minimum Critical Power Ratio (MCPR) action and surveillance requirements associated with operating on a limiting control rod pattern does not involve a significant reduction ina margin of safety. This change is acceptable because it does not impact the ability of the RBMs to perform their intended safety function or the ability to maintain power distribution parameters within required limits. Since Technical Specifications continue to provide adequate assurance of RBM operability and maintenance of power distribution parameters within required limits, the ability to satisfy safety analysis assumptions and initial conditions Is unchanged. Inaddition, the probability of actually operating on a limiting control rod pattern is very low. Since the change has no effect on any safety analysis assumptions or initial conditions, the margins of safety continue to be maintained. Therefore, this proposed change does not involve a significant reduction ina margin of safety.

VYNPS 1 Revision 0

NO SIGNIFICANT HAZARDS CONSIDERATION TS 3.2.E/4.2.E - CONTROL ROD BLOCK INSTRUMENTATION L.2 CHANGE Inaccordance with the criteria set forth in 10 CFR 50.92, Vermont Yankee Nuclear Power Station has evaluated this proposed Technical Specifications change and determined it does not represent a significant hazards consideration. The following is provided insupport of this conclusion.

I1. Does the change involve a significant increase inthe probability or consequences of an accident previously evaluated?

The proposed change excludes neutron detectors from the Rod Block Monitor (RBM) Instrument Calibration Surveillance Requirements. The RBM Instrumentation and associated Surveillance Requirements are not assumed to be initiators of any analyzed event. The Technical Specifications will continue to ensure that changes inthe associated neutron detectors are compensated for in other periodic Surveillance Requirements. As such, the proposed change will not affect the ability of the RBM to perform its safety function of preventing a continuous rod withdrawal error event and accident consequences are not changed. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not necessitate a physical alteration of the plant (no new or different type of equipment will be Installed) or changes in parameters governing normal plant operation.

The proposed change continues to ensure the affected instrumentation is capable of performing its function as assumed inthe safety analyses. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction ina margin of safety?

Excluding the neutron detectors from the RBM Instrument Calibration does not involve a significant reduction ina margin of safety. The Technical Specifications will continue to ensure the RBM is capable of performing its safety function. Changes inthe neutron detectors, excluded from the RBM Instrument Calibration by this proposed change, are compensated for in other periodic Technical Specification Surveillance Requirements. As such, the proposed change will not affect the ability of the RBM to perform its safety function of prevenUng a continuous rod withdrawal error event. Since the change has no effect on any safety analysis assumptions or initial conditions, the margins of safety continue to be maintained. Therefore, this proposed change does not involve a significant reduction in a margin of safety.

WNPS 2 Revision 0

References 3.2.E and 4.2.E Control Rod Block Actuation

3.2.E/4.2.E REFERENCES Control Rod Block Actuation

1. UFSAR, Section 7.5.8.

2. UFSAR, Section 7.7.4.3.2.

3. UFSAR, Section 14.5.3.1.

4. GENE-770-06-1 -A, "Addendum to Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," December 1992.

5. NEDC-30851 P-A Supplement 1, "Technical Specifications Improvement Analysis for BWR Control Rod Block Instrumentation," October 1988.

Proposed Technical Specifications 3.2.F and 4.2.F Mechanical Vacuum Pump Isolation Instrumentation

VYNPS 3.2 LIMITING CONDITIONS FOR 4.2 SURVEILLANCE REQUIREMENTS OPERATION F. Mechanical Vacuum Pump F. Mechanical Vacuum Pump Isolation Instrumentation Isolation Instrumentation

1. When the reactor is in the 1. The High Main Steam Line I I RUN or STARTUP/HOT STANDBY Radiation Trip Function Mode and the mechanical for mechanical vacuum pump vacuum pump is in service, isolation shall be 4 channels of the High checked, functionally I Main Steam Line Radiation tested and calibrated as Trip Function for indicated in Surveillance mechanical vacuum pump Requirements 4.2.F.l.a, b, isolation shall be c, d and e. I operable.

I When a channel is placed in an inoperable status solely for performance of required surveillances, entry into associated Limiting Conditions for Operation and required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Trip Function I maintains mechanical vacuum pump isolation capability.

a. Perform an Instrument Check once each day.

b. Perform an Instrument Functional Test once every 3 months.

c. Perform an Instrument Calibration, except for radiation detectors, using a current source once every 3 months.

The Trip Setting shall be < 3.0 X background at rated thermal power.

d. Perform an Instrument Calibration using a radiation source once each Refueling Outage.

e. Perform a Logic System Functional Test, including mechanical vacuum pump isolation valve, once every Operating Cycle.

Amendment No. 58

VYNPS 3.2 LIMITING CONDITIONS FOR 4.2 SURVEILLANCE REQUIREMENTS OPERATION

2. If Specification 3.2.F.1 is not met, take all of the applicable Actions in Specifications 3.2.F.2.a and 2.b below.

a. With one or more channels inoperable:

1) Restore the inoperable channel(s) to operable status within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; or

2) Place the inoperable channel(s) or associated trip system in the trip condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (not applicable if the inoperable channel is the result of an inoperable mechanical vacuum pump isolation valve).

b. If the required Action and associated completion time of Specification 3.2.F.2.a above is not met, or if mechanical vacuum pump isolation capability is not maintained:

1) Isolate the mechanical vacuum pump within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; or

2) Isolate the main steam lines within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; or

3) Place the reactor in the SHUTDOWN Mode within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Amendment No. 59

Proposed Bases 3.2.F and 4.2.F Mechanical Vacuum Pump Isolation Instrumentation

VYNPS BASES: 3.2.F/4.2.F MECHANICAL VACUUM PUMP ISOLATION BACKGROUND The mechanical vacuum pump isolation instrumentation initiates an isolation of the mechanical vacuum pump following events in which main steam radiation monitors exceed a predetermined value. Tripping and isolating the mechanical vacuum pumps limits control room and offsite doses in the event of a control rod drop accident (CRDA).

The mechanical vacuum pump isolation instrumentation includes sensors, relays and switches that are necessary to cause initiation of mechanical vacuum pump isolation. The channels include electronic equipment that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs an isolation signal to the mechanical vacuum pump isolation logic.

The isolation logic consists of two independent trip systems, with two channels of the High Main Steam Line Radiation Trip Function in each trip system. Each trip system is a one-out-of-two logic for this Trip Function.

Thus, either channel of the High Main Steam Line Radiation Trip Function in a trip system is needed to trip the trip system. The outputs of the channels in a trip system are arranged in a logic so that both trip systems must trip to result in an isolation signal.

The mechanical vacuum pump isolation valve is also associated with this Trip Function.

APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY The mechanical vacuum pump isolation is assumed in the safety analysis for the CRDA. The mechanical vacuum pump isolation instrumentation initiates an isolation of the mechanical vacuum pump to limit control room and offsite doses resulting from fuel cladding failure in a CRDA.

The mechanical vacuum pump isolation instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

The operability of the mechanical vacuum pump isolation instrumentation is dependent on the operability of the four High Main Steam Line Radiation Trip Function instrumentation channels with their trip setpoints within the calculational as-found tolerances specified in plant procedures. Operation with actual trip setpoints within calculational as-found tolerances provides reasonable assurance that, under worst case design basis conditions, the associated trip will occur within the Trip Settings specified in Surveillance Requirement 4.2.F.l.c as required by the CRDA analysis. As a result, a channel is considered inoperable if its actual trip setpoint is not within the calculational as-found tolerances specified in plant procedures. The actual trip setpoint is calibrated consistent with applicable setpoint methodology assumptions. The High Main Steam Line Radiation Trip Setting was chosen to be as low enough to ensure that control room and offsite dose limits are not exceeded in the event of a CRDA, but high enough to avoid spurious isolation due to nitrogen-16 spikes, instrument instabilities, and other operational occurrences. Channel operability also includes the mechanical vacuum pump isolation valve.

Amendment No. 78

VYNPS BASES: 3.2.F/4.2.F MECHANICAL VACUUM PUMP ISOLATION APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

The mechanical vacuum pump isolation is required to be operable in RUN and STARTUP/HOT STANDBY when the mechanical vacuum pump is in service to mitigate the consequences of a postulated CRDA. In this condition, fission products released during a CRDA could be discharged directly to the environment.

Therefore, the mechanical vacuum pump isolation is necessary to assure conformance with the radiological evaluation of the CRDA. In other Modes or conditions, the consequences of a control rod drop are insignificant, and are not expected to result in any fuel damage or fission product releases. When the mechanical vacuum pump is not in operation in RUN and STARTUP/HOT STANDBY, fission product releases via this pathway would not occur.

ACTIONS Specification 3.2.F.2.a With one or more High Main Steam Line Radiation Trip Function channels inoperable, but with mechanical vacuum pump isolation capability maintained (refer to Specification 3.2.F.2.b Bases), the mechanical vacuum pump isolation instrumentation is capable of performing the intended function.

However, the reliability and redundancy of the mechanical vacuum pump isolation instrumentation is reduced, such that a single failure in one of the remaining channels could result in the inability of the mechanical vacuum pump isolation instrumentation to perform the intended function. Therefore, only a limited time is allowed to restore the inoperable channels to operable status. Because of the low probability of an extensive number of inoperabilities affecting multiple channels, and the low probability of an event requiring the initiation of mechanical vacuum pump isolation, 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months has been shown to be acceptable (Ref. 1) to permit restoration of any inoperable channel to operable status. Alternately, the inoperable channel or associated trip system may be placed in trip, since this would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. As noted, placing the channel in trip with no further restrictions is not allowed if the inoperable channel is the result of an inoperable mechanical vacuum pump isolation valve, since this may not adequately compensate for the inoperable mechanical vacuum pump isolation valve (e.g., the isolation valve may be inoperable such that it will not close). If it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel would result in loss of condenser vacuum), or if the inoperable channel is the result of an inoperable mechanical vacuum pump isolation valve, Specification 3.2.F.2.b must be entered and its required actions taken.

Specification 3.2.F.2.b With any required Action and associated completion time of Specification 3.2.F.2.a not met, the plant must be brought to a Mode or other specified condition in which the LCO does not apply. To achieve this status, the plant must be brought to at least SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Alternately, the mechanical vacuum pump may be isolated since this performs the intended function of the instrumentation. An additional option is provided to isolate Amendment No. 78a

VYNPS BASES: 3.2.F/4.2.F MECHANICAL VACUUM PUMP ISOLATION ACTIONS (continued) the main steam lines, which may allow operation to continue. Isolating the main steam lines effectively provides an equivalent level of protection by precluding fission product transport to the condenser. This isolation is accomplished by isolation of all main steam lines and main steam line drains which bypass the main steam isolation valves.

Specification 3.2.F.2.b is also intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels result in the High Main Steam Line Radiation Trip Function not maintaining mechanical vacuum pump isolation capability. The High Main Steam Line Radiation Trip Function is considered to be maintaining mechanical vacuum pump isolation capability when sufficient channels are operable or in trip such that the mechanical vacuum pump isolation instruments will generate a trip signal from a valid High Main Steam Line Radiation signal, and the mechanical vacuum pump will be isolated.

This requires one channel of the High Main Steam Line Radiation Trip Function in each trip system to be operable or in trip, and the mechanical vacuum pump isolation valve to be operable.

SURVEILLANCE REQUIREMENTS Surveillance Requirement 4.2.F.1 As indicated in Surveillance Requirement 4.2.F.1, the High Main Steam Line Radiation Trip Function for mechanical vacuum pump isolation shall be checked, functionally tested and calibrated as indicated Surveillance Requirements 4.2.F.l.a, b, c, d, and e.

Surveillance Requirement 4.2.F.1 also indicates that when a channel is placed in an inoperable status solely for performance of required instrumentation Surveillances, entry into associated LCO and required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Trip Function maintains mechanical vacuum pump isolation capability. Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance, the channel must be returned to operable status or the applicable LCO entered and required Actions taken. This allowance is based on the reliability analysis (Ref. 1) assumption of the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months testing allowance does not significantly reduce the probability that a mechanical vacuum pump will isolate when necessary.

Surveillance Requirement 4.2.F.l.a, Instrument Check Performance of an Instrument Check once each day ensures that a gross failure of instrumentation has not occurred. An Instrument Check is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.

Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. An Instrument Check will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each Calibration. Agreement criteria are determined by the plant staff based Amendment No. 78b

VYNPS BASES: 3.2.F/4.2.F MECHANICAL VACUUM PUMP ISOLATION SURVEILLANCE REQUIREMENTS (continued) on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit. The Frequency is based upon operating experience that demonstrates channel failure is rare. The Instrument Check supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.

Surveillance Requirement 4.2.F.l.b, Instrument Functional Test An Instrument Functional Test is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of "once every 3 months" is based on the reliability analysis of Reference 1.

Surveillance Requirements 4.2.F.l.c and 4.2.F.l.d, Instrument Calibrations An Instrument Calibration is a complete check of the instrument loop and the sensor. This test verifies that the channel responds to the measured parameter within the necessary range and accuracy. An Instrument Calibration leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. Surveillance Requirement 4.2.F.l.c requires a calibration to be performed once every 3 months using a current source. This current source is provided downstream of the radiation detectors. As such, the radiation detectors are excluded from the 3 month calibration. Surveillance Requirement 4.2.F.l.d requires a calibration to be performed once each Refueling Outage using a radiation source. The radiation detectors are included in the once each Refueling Outage calibration. The specified Instrument Calibration Frequencies are based upon the time interval assumptions for calibration used in the determination of the magnitude of equipment drift in the associated setpoint analyses.

Surveillance Requirement 4.2.F.l.e, Logic System Functional Test The Logic System Functional Test demonstrates the operability of the required trip logic for a specific channel. Actuation of the mechanical vacuum pump isolation valve is included as part of this Surveillance to provide complete testing of the assumed safety function. Therefore, if the isolation valve is incapable of actuating, the instrument channel would be inoperable. The Frequency of "once every Operating Cycle" is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has demonstrated that these components will usually pass the Surveillance when performed at the specified Frequency.

Amendment No. 78c

VYNPS BASES: 3.2.F/4.2.F MECHANICAL VACUUM PUMP ISOLATION REFERENCES

1. NEDC-30851P-A, Supplement 2, Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation, March 1989.

Amendment No. 78d

Current Technical Specifications Markups 3.2.F and 4.2.F Mechanical Vacuum Pump Isolation Instrumentation

VYNPS 3.2 LIMITING CONDITIONS FOR 4.2 SURVEILLANCE REQUIREMENTS OPERATION

. Off-Gas System Isolation D. Off-Gas System isolation During reactor power Instrumentation and logic operation, the systems shall be instrumentation that functionally tested and initiates isolation of the calibrated as indicated in off-gas system shall be Table 4.2.4.

operable in accordance with Table 3.2.4. MV1 A4I4 P4 >

1. _

E. Control Rod Block Actuation E. Control Rod Block Actuatio During reactor power Instrumentation and logic operation the systems shall be instrumentation that functionally tested and initiates control rod block calibrated as indicated in shall be operable Table 4.2.5.

in accordance with Table 3.2.5.

F. Mechanical Vacuum Pump F. Mechanical Vacuum Pump Isolation Instrumentation Isolation Instrumentation When the reactor is in the /. The High Main Steam Line RUN or STARTUP/HOT STANDBY Radiation Trip Function for Mode and the mechanical mechanical vacuum pump vacuum-blmp is in service, isolation shall be checked, Archannels of the functionally tested and High Main Steam Line calibrated as indicated in Radiation Trip Function for Surveillance Requirements mechanical vacuum pump 4.2.Fa and isolation shall be operabl When a channel is placed in an inoperable status solely dL./-With one or more channels for performance of required

{ inopera surveillances, entry into associated Limiting Conditions for Operation and Restore the inoperable required ections may be channel(s) to operable delayed for up to iB r statuspl-o~r hours provided the associated Arip Function maintains an0Place the inoperable l mechanical vacuum pump isolation capability.

associated trip system/

in the trip conditio o^ @. Perform an /nstrument (not applicable if the .1h eck once each day.

inoperable channel is the result of an 4 @. Perform an/nstrument inoperable mechanical /unctional &est once every vacuum pump isolation Qo Gf3drmonths.

valve).

X' ecirflCAT R.Q.J. is 146T O$81 r b 5PaCwI oe Amendment No. 4, 44s, 212 35

EJl VYNPS 3.2 LIMITING CONDITIONS FOR 4.2 SURVEILLANCE REQUIREMENTS OPERATION A, X. If the required , tion and C @. Perform an estrument associated completion time alibration, except for of Specification 3.2.F is the radiation detectors, not met.J nt Fh',_- using a current source

_ a o- _v e--nHe 2 g rsJ.t once every4gioavC months. The AripBetting

/ Isolate the mechanica.1 shall be 5 3.0 vacuum pumpA or g, background at rated 1-1 thermal power.

/ Isolate the main stea 7~w lines .~ Perform an Anstrument A libration using a S)4 Place the reactor in radiation source once each the SHUTDOWN Mod%:__. /efuelingiutage.

oR ,i"CCq4,VI441C VACOUMA POM Perform a q ic,,ystem I #a.l4ri@

1oV SF œrecV c coped,~,1

,&P

'V.C

'GS irs Afzr functional Yest, including mechanical vacuum pump ED- isolation valve, once Aperating/jycle.

Post-Accident Instrumentation G. Post-Accident Instrumentat During reactor power operation, The post-accident the instrumentation that instrumentation shall be displays information in the functionally tested and Control Room necessary calibrated in accordance with for the operator to initiate Table 4.2.6. /

and control the systems used during and following a postulated accident or abnormal operating condition shall be operable in accordance with Table 3.2.6. __ ~ ~<a4Z ra VCP.4k~rE Aild oH. Drywell to Torus AP H. Drywell to Torus AP Instrumentation Instrumentation

1. During reactor power The Drywell to Torus AP\

operation, the Drywell to Instrumentation shall be\

Torus AP Instrumentation calibrated once every six (recorder #1-156-3 and months and an instrument instrument DPI-1-158-6) check will be made once per shall be operable except shift.

as specified in 3.2.H.2.

2.From and after the date that one of the Drywell to Torus AP instruments is made or found to be inoperable for any reason, reactor operation is permissible only during the succeeding thirty days nless the instrument is Amendment No. 64, "&, 06, 0&, 4i1, i34, a-s, 212 36

Safety Assessment Discussion of Changes 3.2.F and 4.2.F Mechanical Vacuum Pump Isolation Instrumentation

SAFETY ASSESSMENT OF CHANGES TS 3.2.F/4.2.F - MECHANICAL VACUUM PUMP ISOLATION INSTRUMENTATION ADMINISTRATIVE A.1 Inthe revision of the Vermont Yankee Nuclear Power Station (VYNPS) current Technical Specifications (CTS), certain wording preferences or conventions are adopted which do not result intechnical changes (either actual or interpretational). Editorial changes, reformatting, and revised numbering are adopted to make the WNPS Technical Specifications (TS) more consistent with human factor principles used in the Boiling Water Reactor Improved Standard Technical Specifications (ISTS), NUREG-1433, Rev. 2. These format and presentation changes are being made to improve usability and clarity. The changes are considered administrative.

A.2 CTS 3.2.D and 4.2.D provide requirements that apply to off-gas isolation instrumentation.

Changes these CTS off-gas isolation instrumentation requirements are addressed inthe Safety Assessments of Changes for CTS 3.2.D/4.2.D, Off-Gas Isolation Instrumentation.

Therefore, this change does not involve a technical change, but is only a difference of presentation preference and isconsidered administrative. CTS 3.2.H and 4.2.H provide requirements that apply to drywell to torus AP instrumentation. Changes these CTS drywell to torus AP instrumentation requirements are addressed in the Safety Assessments of Changes for CTS 3.2.H/4.2.H, Drywell to Torus AP Instrumentation. Therefore, this change does not involve a technical change, but isonly a difference of presentation preference and isconsidered administrative.

TECHNICAL CHANGES - MORE RESTRICTIVE MA CTS 3.2.F.1 provides actions for one or more mechanical vacuum pump isolation instrumentation channels inoperable. These actions provide 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to restore the inoperable channel(s) or place them or their associated trip system inthe trip condition.

Proposed TS 3.2.F.2 isrevised to address the condition of multiple, inoperable, untripped mechanical vacuum pump isolation instrumentation channels that result in the High Main Steam Line Radiation Trip Function not maintaining mechanical vacuum pump isolation capability. The proposed change will eliminate the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowed outage time for this condition. Inthis condition (i.e., loss of mechanical vacuum pump isolation capability),

continued operation in accordance with the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowed outage time is not appropriate, nor consistent with the applicable analyses of NEDC-30851 P-A, Supplement 2,'Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation.' This change is an additional restriction on plant operation necessary to achieve consistency with applicable analyses.

TECHNICAL CHANGES - LESS RESTRICTIVE None RELOCATED SPECIFICATIONS None VYNPS 1 Revision 0

No Significant Hazards Consideration 3.2.F and 4.2.F Mechanical Vacuum Pump Isolation Instrumentation

GENERIC NO SIGNIFICANT HAZARDS CONSIDERATION ADMINISTRATIVE CHANGES (A.x' Labeled Comments/Discussions)

Inaccordance with the criteria set forth in 10 CFR 50.92, Vermont Yankee Nuclear Power Station has evaluated this proposed Technical Specifications change and determined it does not represent a significant hazards consideration. The following Isprovided Insupport of this conclusion.

1. Does the change involve a significant increase inthe probability or consequences of an accident previously evaluated?

The proposed change involves reformatting, renumbering, and rewording the existing Technical Specifications. The reformatting, renumbering, and rewording process involves no technical changes to the existing Technical Specifications. As such, this change isadministrative Innature and does not Impact initiators of analyzed events or assumed mitigation of accident or transient events. Therefore, this change does not involve a significant Increase inthe probability or consequences of an accident previously evaluated.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be Installed) or changes inmethods governing normal plant operation. The proposed change will not impose any new or eliminate any old requirements. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The proposed change will not reduce a margin of safety because it has no impact on any safety analyses assumptions. This change is administrative in nature. Therefore, the change does not involve a significant reduction in a margin of safety.

VYNPS 1 Revision 0

GENERIC NO SIGNIFICANT HAZARDS CONSIDERATION TECHNICAL CHANGES - MORE RESTRICTIVE rM.x' Labeled Comments/Discussions)

Inaccordance with the criteria set forth In10 CFR 50.92, Vermont Yankee Nuclear Power Station has evaluated this proposed Technical Specifications change and determined it does not represent a significant hazards consideration. The following is provided in support of this conclusion.

1. Does the change involve a significant increase inthe probability or consequences of an accident previously evaluated?

The proposed change provides more stringent requirements for operation of the facility. These more stringent requirements do not result inoperation that will increase the probability of initiating an analyzed event and do not alter assumptions relative to mitigation of an accident or transient event. The more restrictive requirements continue to ensure process variables, structures, systems, and components are maintained consistent with the safety analyses and licensing basis. Therefore, this change does not involve a significant increase inthe probability or consequences of an accident previously evaluated.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes inthe methods governing normal plant operation. The proposed change does impose different requirements. However, these changes are consistent with the assumptions in the safety analyses and licensing basis. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The imposition of more restrictive requirements either has no impact on or increases the margin of plant safety. As provided inthe discussion of the change, each change inthis category is by definition, providing additional restrictions to enhance plant safety. The change maintains requirements within the safety analyses and licensing basis. Therefore, this change does not involve a significant reduction in a margin of safety.

VYNPS 3 Revision 0

NO SIGNIFICANT HAZARDS CONSIDERATION TS 3.2.F14.2.F - MECHANICAL VACUUM PUMP ISOLATION INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE There were no specific less restrictive changes identified for this Specification.

VYNPS 1 Revision 0

References 3.2.F and 4.2.1F Mechanical Vacuum Pump Isolation Instrumentation

3.2.F/4.2.F REFERENCE Mechanical Vacuum Pump Isolation Instrumentation

1. NEDC-30851 P-A, Supplement 2, Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation, March 1989.

ML031200259

Text

Conclusion:

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