Text

Instrumentation Technical Specifications Proposed Change No. 259, Proposed Technical Technical Specifications 3.2.G and 4.2.G - No Significant Hazards Consideration 3.2/4.1 and 3.2/4.2
	ML031200664
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: ML031200664 (170)
	v • d • e

Proposed Technical Specifications 3.2.G and 4.2.G Post-Accident Monitoring Instrumentation

/

VYNPS 3.2 LIMITING CONDITIONS FOR 4.2 SURVEILLANCE REQUIREMENTS OPERATION G. Post-Accident Monitoring G. Post-Accident Monitoring Instrumentation Instrumentation The post-accident monitoring 1. The post-accident instrumentation for each monitoring instrumentation Function in Table 3.2.6 shall shall be checked and be operable in accordance calibrated in accordance with Table 3.2.6. with Table 4.2.6.

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 .

60 Amendment No.

VYNPS Table 3.2.6 (page 1 of 1)

Post-Accident Monitoring Instrumentation REQUIRED ACTIONS WHEN CHANNELS REQUIRED APPLICABLE MODES OR OTHER PER CHANNELS ARE FUNCTION SPECIFIED CONDITIONS FUNCTION INOPERABLE RUN, STARTUP/HOT STANDBY 2 Note 1

1. Drywell Atmospheric Temperature RUN, STARTUP/HOT STANDBY 2 Note 1

2. Drywell Pressure RUN, STARTUP/HOT STANDBY 2 Note 1

3. Torus Pressure RUN, STARTUP/HOT STANDBY 2 Note 1

4. Torus Water Level RUN, STARTUP/HOT STANDBY 2 Note 1

5. Torus Water Temperature RUN, STARTUP/HOT STANDBY 2 Note 1

6. Reactor Pressure RUN, STARTUP/HOT STANDBY 2 Note 1

7. Reactor Vessel Water Level RUN, STARTUP/HOT STANDBY 2 Note 1

8. Torus Air Temperature RUN, STARTUP/HOT STANDBY 2 Note 1

9. Containment Hydrogen/Oxygen Monitor 2 Note 2

10. Containment High RUN, STARTUP/HOT STANDBY Range Radiation Monitor 61 Amendment No.

VYNPS Table 3.2.6 ACTION Notes for

1. With one or more Post-Accident Monitoring instrumentation channels, Functions other than Function 10, inoperable, take all of the applicable Actions in Notes l.a and l.b below.

a. With one or more Functions with one channel inoperable:

1) Restore channel to operable status within 30 days; or

2) Prepare and submit a special report to the Commission within the next 14 days, outlining the Action taken, the cause of the inoperability, and the plans and schedule for restoring the channel to operable status.

b. With one or more Functions with two channels inoperable:

1) Restore one required channel to operable status within 7 days; or

2) Place the reactor in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

10

2. With one or more Post - Accident Monitoring instrumentation Function Actions in Notes 2.a and channels inoperable, take all of the applicable 2.b below.

a. With one channel inoperable:

1) Restore channel to operable status within 30 days; or

2) Prepare and submit a special report to the Commission within the next 14 days, outlining the Action taken, the cause of the inoperability, and the plans and schedule for restoring the channel to operable status.

b. With two channels inoperable:

1) Restore one channel to operable status within 7 days; or

2) Prepare and submit a special report to the Commission within the next 14 days, outlining the Action taken, the cause of the inoperability, and the plans and schedule for restoring the channels to operable status.

62 Amendment No.

VYNPS Table 4.2.6 (page 1 of 1)

Post-Accident Monitoring Instrumentation Tests and Frequencies FUNCTION CHECK CALIBRATION

1. Drywell Atmospheric Once/Day Every 6 Months Temperature

2. Drywell Pressure Once/Day Once/Operating Cycle

3. Torus Pressure Once/Day Once/Operating Cycle

4. Torus Water Level Once/Day once/operating Cycle

5. Torus Water Once/Day Every 6 Months Temperature

6. Reactor Pressure Once/Day once/operating Cycle

7. Reactor Vessel Water Once/Day Once/Operating Level Cycle

8. Torus Air Temperature Once/Day Every 6 Months

9. Containment Once/Day Once/Operating Hydrogen/Oxygen Cycle Monitor

10. Containment High Once/Day Once/Operating Range Radiation Cycle Monitor 63 Amendment No.

Proposed Bases 3.2.G and 4.2.G Post-Accident Monitoring Instrumentation

VYNPS BASES: 3.2.G/4.2.G POST-ACCIDENT MONITORING INSTRUMENTATION BACKGROUND The primary purpose of the post-accident monitoring (PAM) instrumentation is to display, in the control room, plant variables that provide information required by the control room operators during accident situations. This information provides the necessary support for the operator to take the manual actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for Design Basis Events. The instruments that monitor these variables are designated as Type A, Category I, and non-Type A, Category I, in accordance with Regulatory Guide 1.97 (Ref. 1).

The operability of the post-accident monitoring instrumentation ensures that there is sufficient information available on selected plant parameters to monitor and assess plant status and behavior following an accident. This capability is consistent with the recommendations of Reference 1.

APPLICABLE SAFETY ANALYSES The PAM instrumentation Specification ensures the operability of Regulatory Guide 1.97, Type A variables so that the control room operating staff can:

Perform the diagnosis specified in the Emergency Operating Procedures (EOPs). These variables are restricted to preplanned actions for the primary success path of Design Basis Accidents (DBAs), (e.g., loss of coolant accident (LOCA)), and

Take the specified, preplanned, manually controlled actions for which no automatic control is provided, which are required for safety systems to accomplish their safety function.

The PAM instrumentation Specification also ensures operability of most Category I, non-Type A, variables so that the control room operating staff can:

Determine whether systems important to safety are performing their intended functions;

Determine the potential for causing a gross breach of the barriers to radioactivity release;

Determine whether a gross breach of a barrier has occurred; and

Initiate action necessary to protect the public and for an estimate of the magnitude of any impending threat.

The plant specific Regulatory Guide 1.97 analysis (Ref. 2) documents the process that identified Type A and Category I, non-Type A, variables.

Post-accident monitoring instrumentation that satisfies the definition of Type A in Regulatory Guide 1.97 meets Criterion 3 of 10 CFR 50.36(c)(2)(ii).

Category I, non-Type A, instrumentation is retained in Technical Specifications (TS) because they are intended to assist operators in Amendment No. 79

VYNPS BASES: 3.2.G/4.2.G POST-ACCIDENT MONITORING INSTRUMENTATION APPLICABLE SAFETY ANALYSES (continued) minimizing the consequences of accidents. Therefore, these Category I variables are important for reducing public risk.

LCO Specification 3.2.G and Table 3.2.6 require two operable channels for each Function to ensure that no single failure prevents the operators from being presented with the information necessary to determine the status of the plant and to bring the plant to, and maintain it in, a safe condition following an accident. Furthermore, providing two channels allows an Instrument Check during the post accident phase to confirm the validity of displayed information.

The following list is a discussion of the specified instrument Functions listed in Table 3.2.6.

1. Drywell Atmospheric Temperature Drywell atmospheric temperature is a Type A and Category I variable provided to detect a reactor coolant pressure boundary (RCPB) breach and to verify the effectiveness of Emergency Core Cooling System (ECCS) functions that operate to maintain containment integrity. Two redundant temperature signals are transmitted from separate temperature elements for each channel. The output of one of these channels is recorded on a recorder in a control room. The output of the other channel is displayed on an indicator in the control room.

0 0 The drywell atmospheric temperature channels measure from 0 F to 350 F.

Therefore, the PAM Specification deals specifically with this portion of the instrument channels.

2. Drywell Pressure Drywell pressure is a Type A and Category I variable provided to detect breach of the RCPB and to verify ECCS functions that operate to maintain Reactor Coolant System (RCS) integrity. Two drywell pressure signals are transmitted from separate pressure transmitters for each channel. The output of these channels is displayed on two independent indicators in the control room. The pressure channels measure from -15 psig to 260 psig. Therefore, the PAM Specification deals specifically with this portion of the instrument channel.

3. Torus Pressure Torus pressure is a Type A and Category I variable provided to detect a condition that could potentially lead to containment breach and to verify the effectiveness of ECCS actions taken to prevent containment breach. Two torus pressure signals are transmitted from separate pressure transmitters and displayed on two independent indicators in the control room. The range of Amendment No. 79a

VYNPS BASES: 3.2.G/4.2.G POST-ACCIDENT MONITORING INSTRUMENTATION LCO (continued) indication is - 15 psig to 65 psig. Therefore, the PAM Specification deals specifically with this portion of the instrument channel.

4. Torus Water Level Torus water level is a Type A and Category I variable provided to detect a breach in the RCPB. This variable is also used to verify and provide long term surveillance of ECCS function. The Torus Water Level Function provides the operator with sufficient information to assess the status of both the RCPB and the water supply to the ECCS. The Torus Water Level Function channels monitor the torus water level from the bottom of the torus to 5 feet above the normal torus water level. Two torus water level signals are transmitted from separate level transmitters to two independent control room indicators in the control room. Therefore, the PAM Specification deals specifically with this portion of the instrument channel.

5. Torus Water Temperature Torus water temperature is a Type A and Category I variable provided to detect a condition that could potentially lead to containment breach and to verify the effectiveness of ECCS actions taken to prevent containment breach.

Two redundant temperature signals are transmitted from separate temperature elements for each channel. The temperature channels output to two independent control room indicators. The range of the torus water temperature channels is 0OF to 250'F. Therefore, the PAM Specification deals specifically with this portion of the instrument channels.

6. Reactor Pressure Reactor pressure is a Type A and Category I variable provided to support monitoring of RCS integrity and to verify operation of the ECCS. Two independent pressure transmitters, with a range of 0 psig to 1500 psig, monitor pressure and provide pressure indication to the control room. The output from these channels is provided to two independent indicators in the control room. Therefore, the PAM Specification deals specifically with this portion of the instrument channel.

7. Reactor Vessel Water Level Reactor vessel water level is a Type A and Category I variable provided to support monitoring of core cooling and to verify operation of the ECCS.

Water level is measured by independent differential pressure transmitters for each channel. Each channel measures from -200 inches to + 200 inches, referenced to the top of enriched fuel. The output from these channels is provided to two independent indicators in the control room. Therefore, the PAM Specification deals specifically with this portion of the instrument channel.

Amendment No. 79b

VYNPS BASES: 3.2.G/4.2.G POST-ACCIDENT MONITORING INSTRUMENTATION LCO (continued)

8. Torus Air Temperature to detect Torus air temperature is a Type A and Category I variable provided to verify the effectiveness of ECCS functions that operate a RCPB breach and signals are to maintain containment integrity. Two redundant temperature channel. The output transmitted from separate temperature elements for each in a control room with a of one of these channels is recorded on a recorder 0 is displayed on an range of 50 F to 300 F. The output of the other channel 0 0

the indicator in the control room with a range of 0F to 350 F. Therefore, PAM Specification deals specifically with this portion of the instrument channels.

9. Containment Hydrogen/Oxygen Concentration Monitor provided Containment hydrogen and oxygen monitors are Category I instruments represent a to detect high hydrogen or oxygen concentration conditions that in potential for containment breach. This variable is also important verifying the adequacy of mitigating actions. Hydrogen and oxygen output is concentrations are each measured by two redundant monitors whose monitor provided to the control room. The containment hydrogen and oxygen gas PAM instrumentation consists of two independent gas analyzers. Each The analyzer consists of a hydrogen analyzer and an oxygen analyzer.

range of analyzers are capable of determining hydrogen concentration in the There are two 0% to 30% and oxygen concentration in the range of 0% to 25%.

independent recorders in the control room to display the results.

10. Containment High Range Radiation Monitor to monitor Containment high range radiation is a Category 1 variable provided release the potential of significant radiation releases and to provide site assessment for use by operators in determining the need to invoke detectors are mounted in the emergency plans. Two redundant radiation provides a signal to an independent monitor drywell. Each radiation detector which has a range from 100 R/hr to 107 R/hr. The outputs in the control room, indicators of these radiation monitors are displayed on two independent located in the control room. Therefore, the PAM Specification deals specifically with this portion of the instrument channel.

APPLICABILITY The PAM instrumentation Specification is applicable in the RUN and diagnosis and STARTUP/HOT STANDBY Modes. These variables are related to the required to mitigate DBAs. The applicable DBAs are assumed preplanned actions other Modes and to occur in the RUN and STARTUP/HOT STANDBY Modes. In likelihood of an event that conditions, plant conditions are such that the would require PAM instrumentation is extremely low; therefore, PAM or instrumentation is not required to be operable in these other Modes conditions.

79c Amendment No.

VYNPS BASES: 3.2.G/4.2.G POST-ACCIDENT MONITORING INSTRUMENTATION ACTIONS Table 3.2.6 ACTION Note 1 Table 3.2.6 ACTION Note l.a.l) requires that, when one or more Functions the (except Function 10) have one required channel that is inoperable, required inoperable channel must be restored to operable status within and 30 days. The 30 day Completion Time is based on operating experience the passive nature of the takes into account the remaining operable channels, is assumed to occur from these instrument (no critical automatic action instruments), and the low probability of an event requiring PAM each instrumentation during this interval. If the inoperable channel of Function has not been restored to operable status in 30 days, Table affected be submitted to 3.2.6 ACTION Note l.a.2) requires a special written report the NRC within the next 14 days. The report will outline the preplanned the plans alternate method of monitoring, the cause of the inoperability, and instrumentation to operable status. This and schedule for restoring the action is appropriate in lieu of a shutdown requirement, since another operable channel is monitoring the Function, an alternate method of that monitoring is available, and given the likelihood of plant conditions would require information provided by this instrumentation.

Table 3.2.6 ACTION Note l.b.l) requires that, when one or more Functions, (i.e., two except Function 10, have two required channels that are inoperable Function should channels inoperable in the same Function), one channel in the Time of 7 days be restored to operable status within 7 days. The Completion is based on the relatively low probability of an event requiring PAM the instrument operation and the availability of alternate means to obtain required information. Continuous operation with two required channels indications inoperable in a Function is not acceptable because the alternate applied to the may not fully meet all performance qualification requirements PAM instrumentation. Therefore, requiring restoration of one inoperable in a channel of the Function limits the risk that the PAM Function will be occur. If at least one channel of each degraded condition should an accident not been restored to operable status in 7 days, Table affected Function has which 3.2.6 ACTION Note l.b.2) requires the plant to be brought to a Mode in the plant must be brought to the LCO does not apply. To achieve this status, The allowed Completion Times are at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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 3.2.6 ACTION Note 2 Table 3.2.6 ACTION Note 2.a.1) requires that, when Function 10 has one must be required channel that is inoperable, the required inoperable channel 30 day Completion Time is restored to operable status within 30 days. The into account the remaining operable based on operating experience and takes action channels, the passive nature of the instrument (no critical automatic from these instruments), and the low probability of an is assumed to occur instrumentation during this interval. If the inoperable event requiring PAM 3.2.6 channel has not been restored to operable status in 30 days, Table be submitted to the NRC ACTION Note 2.a.2) requires a special written report within the next 14 days. The report will outline the preplanned alternate 79d Amendment No.

VYNPS BASES: 3.2.G/4.2.G POST-ACCIDENT MONITORING INSTRUMENTATION ACTIONS (continued) method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation to operable status. This action is appropriate in lieu of a shutdown requirement, since another operable channel is monitoring the Function, an alternate method of monitoring is available, and given the likelihood of plant conditions that would require information provided by this instrumentation.

Table 3.2.6 ACTION Note 2.b.1) requires that, when Function 10 has two required channels that are inoperable, one channel should be restored to operable status within 7 days. The Completion Time of 7 days is based on the relatively low probability of an event requiring PAM instrument operation and the availability of alternate means to obtain the required information.

Continuous operation with two required channels inoperable in a Function is not acceptable because the alternate indications may not fully meet all performance qualification requirements applied to the PAM instrumentation.

Therefore, requiring restoration of one inoperable channel of the Function limits the risk that the PAM Function will be in a degraded condition should an accident occur.

Since alternate means of monitoring drywell radiation have been developed and tested, the action required by Table 3.2.6 ACTION 2.b.2), if at least one channel has not been restored to operable status within 7 days, is not to shut down the plant, but rather to submit a special written report to the NRC within the next 14 days. The report will outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the normal PAM instrumentation to operable status.

The alternate means of monitoring may be temporarily installed if the normal PAM channel cannot be restored to operable status within the allotted time.

The report provided to the NRC should also describe the degree to which the alternate means are equivalent to the installed PAM channels and justify the areas in which they are not equivalent.

SURVEILLANCE REQUIREMENTS Surveillance Requirement 4.2.G.1 As indicated in Surveillance Requirement 4.2.G.1, post-accident monitoring instrumentation shall be checked and calibrated as indicated in Table 4.2.6.

Table 4.2.6 identifies, for each Function, the applicable Surveillance Requirements.

Surveillance Requirement 4.2.G.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 . 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. The 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months testing allowance is acceptable since it does not significantly reduce the probability of properly monitoring post-accident parameters, when necessary.

Amendment No. 79e

VYNPS BASES: 3.2.G/4.2.G POST-ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS (continued)

Table 4.2.6, Check failure Performance of an Instrument Check once each day ensures that a gross has not occurred. An Instrument Check is normally a of instrumentation comparison of the parameter indicated on one channel to a similar parameter channels on other channels. It is based on the assumption that instrument monitoring the same parameter should read approximately the same value.

of Significant deviations between instrument channels could be an indication more excessive instrument drift in one of the channels or something even it is serious. An Instrument Check will detect gross channel failure; thus, between key to verifying the instrumentation continues to operate properly Agreement criteria are determined by the plant staff based each Calibration.

on a combination of the channel instrument uncertainties, including may be indication and readability. If a channel is outside the criteria, it 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 4.2.6, Calibration the An Instrument Calibration is a complete check of the instrument loop and sensor. This test verifies that the channel responds to the measured parameter within the necessary range and accuracy. The specified Instrument Calibration Frequencies are based on operating experience.

REFERENCES

1. Regulatory Guide 1.97, "Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," Revision 3, May 1983.

to

2. NRC letter, M.B. Fairtile (NRC) to L.A. Temblay (VYNPC), "Conformance Yankee Nuclear Power Station," December Regulatory Guide 1.97 for Vermont 4, 1990.

79f Amendment No.

Current Technical Specifications Markups 3.2.G and 4.2.G Post-Accident Monitoring Instrumentation

CA VYNPS 3.2 LIMITING CONDITIONS FOR 4.2 SURVEILLANCE REQUIREMENTS OPERATION I

2. If the required action and 3. Perform an instrument associated completion time calibration, except for of Specification 3.2.F.1 is the radiation detectors, not met, within the using a current source following 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s: once every three (3) months. The trip setting

a. Isolate the mechanical shall be

3.0 times vacuum pump; or background at rated thermal power.

b. Isolate the main steam lines; or

c. Place the reactor in

4. Perform an instrument calibration using a radiation source once each I

the SHUTDOWN Mode. refueling outage.

5. Perform a logic system functional test, including

<,"# 7F Ire PA a p.4T 4

mechanical vacuum pump isolation valve, once each oeperating cycle.

G. Post-Accident Instrumentation ~

<G. ~ ~ ~ n Ps-cdetInstrumentation

~ring: reactor power operatiog . The post-accident rIe v ; he in rumentation at instrumentation shall be displ s informati in the Cunct naI~y>eeste adC k Co ol Room nec sary calibrated in accordance with

.n diesk{ f the operatow to initia Table 4.2.6.

pig A e-* d control t n4 systems used during nd follow g a postulate accident o v_ abnormal perating ndition &VA shall be operable in accordance with Table 3.2.6.

H. 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 .,1h .fp.

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 nls the intueti 36 Amendment No. iG, S&, 46, 9&, lll, l32, b64,

Zr VYNPS TABLE 3.2.6 POST-ACCIDENT JNSTRUMENTATION Minimum Number of Operable Inst;_umeat Channels o ), Parameter InstrumentRne 2 Drywell Atmospheric 150°F Temperature (Notes 1 and 3)

I 350°F 2 a D Pressure (-13.5) -(+260) psig (Notes 1 and 3) .5) -(+260) p g 2 Torus Pressure (Notes 1 and 3 ( -3 .5) -(+65) psig I

I. (-3 .5) -(+ psig MI Torus Water Level Meter #LI-16-19-12A 25 f

4. 2 (Notes 1 and 3) Meter #LI-1 12B Z5 t.

I 2 Torus Water Temperature Meter #T 19-33A 0 50°F

'I (Notes 1 and 3) Meter I-16-19-33C / -250°F 2 Reactor Pressure Met #PI-2-3-56A

7. I I

2 (Notes 1 and 3)

Reactor Vessel Water Level (Notes 1 and 3) 'I M er #PI-2-3-56B Meter #LI-2-3-91A Meter #LI-2-3-91B/

/

2 Torus Air Temperature \ Recorder #TR- 45 8.I (Notes 1 and 3) (Red)

Meter #TI-1 41

/valve S 9ty/Relief Valve Positi .ghts RV-2-71(A-D Closed - en rom Pressure Switches rom PS-2-71-(1- A-D) tes 1 and 3)

I Amendment No. 54, 46-, 94, 46, 43., 14G, 1S1, a84, 207 53

VYNPS TABLE 3.2.6 (ContSTd)RUMNTAIO POST-ACCIDENT ISRMNAION Minimum Number of Operable Instrument ,

Channels Parameter 'X', =

Inst ment Ran eJ Meter lA/B C2 sed - Oe

~_ Aoustic Mon ~er (Note 5) 2 Containment Hydrogen/Oxygen Recordp SR-VG-6A (SI) / 0-30% hy gen 9'.

/a . 2 Monitor (Notes 1 and 3)

Containment High-Range Radiation Monitor (Note 6)

C Reco -er SR-VG-6B ter RM-16-19-1W (S 0-25% o 1 R/ -07 e

R/hr

( l / Stack Noble Gas Ep luent 0/ er RM-17-155 le- 107 mR/

'_ N Note ) ,

Amendment No. 463, aG, -S, -9, a-64, .8-S, 207 54

VYNPS TABLE 3.2.6 NOTES (Em)-Within 30 days following the loss of one indication, restore the

-poaiF3..6 inoperable channel to an operable status or a special report to the Acr16sJ Commission must be prepared and submitted within the subsequent 14 sI4dT I.a days, outlining the action taken, the cause of the inoperability, and the plans and schedule for restoring the system to operable status.

Within 7 days following the loss of both indications, restore at least Mur 3 .ai one required channel to an operable status or place the reactor in a A4rEGA)hot shutdown condition within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Ooe 5-From and after the date that safety valve position from the acoustic

\ ~monitor is unavailable, reactor operation may continue for 30 days\

\ provided safety valve position can be determined by monitoring safety

\ valve discharge temperature and primary containment pressure. If after 30 days the inoperable channel has not been returned to an operable status, a special report to the Commission must be prepared and submitted within the subsequent 14 days, outlining the action taken, the cause of the inoperability, and the plans and schedule for restoring the system to operable status.

If one or both parameters are not available (i.e., safety valve discharge temperature and primary containment pressure indication) with one or more safety valve position indications from the acoustic monitor unavailable, continued reactor operation is permissible during the next seven days. In this condition, if both secondary parameters are not restored to an operable status within seven days, the reactor shall be laced in a hot shutdown condition within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Within 30 days following the loss of one indication, or seven days r4Ac; 3.o.4. following the loss of both indications, restore the inoperable 4cr,U.uJA AU C channel(s) to an operable status or a special report to the Commission must be prepared and submitted within the subsequent 14 days, outlining the action taken, the cause of the inoperability, and the plans and schedule for restoring the system to operable status.

ote - From and ter the date that ths parameter is unavail le by Control Room i ication, within 72 ho s ensure that local s pling capability is a ilable. If the Contr Room indication is n restored within 7 ys, prepare and submi a special report to t NRC within 14 days

\lowing the event, ou ining the action take , the cause of the inoperability, and th plans and schedule fo restoring the system to

>___gpeabl sttus. /

When a channel is placed in an inoperable status solely for performa 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 .

G6-, -6, 9&, a11, 131, 4-4, G5, 71, 8G6, 244, 207 55 Amendment No. &4,

L73 VYNPS _ __ _

TABLE 4.2.6 OTALIDi TION RE EA a 0- TO I rwtS v POST-ACCIDENTfNTUETTO ~~~~~~

Parameter Calibration Instrument Check

1. Drywell Atmosphere Temperature Every 6 Months Once Each Day

2. Once/Operating Cycle Once Each Day QPressure 3.I Torus Pressure Once/Operating Cycle Once Each Day Torus Water Level Once/Operating Cycle Once Each Day 4.

Torus Water Temperature Every 6 Months Once Each Day

-7. Reactor Pressure Once/Operating Cycle Once Each Day

17. 1 Reactor Vessel Water Level Once/Operating Cycle Once Each Day Torus Air Temperature Every 6 Months Once Each Day go

-Sarety/Relief Valve Positi Every Refueling Outage (N( Unce acn, Functional Test Co performed quar rly) ery Refueling age (MN Once' Day )i (a Functiona PTFest to

/ performed quarterly)

Amendment No. 6M, 6a, 96, m4, 145 10

EL VYNPS TABLE 4.2.6 C,

<6rIZBR~ON ~R~e :Fjy POST-ACCIDENT (Cont'd)

Parameter Cal ibration Instrument Check Containment Hydrogen/Oxygen Monitor Once/Operating Cycle Once each day Ic, Containment High-Range Radiation Once/Operating Cycle Monitor Once each day

'r-t 4-n,-.1. j,..U1 - ,.

Mal t.aaa...

A&LLAUHuI every operating cycle (a Func ional 7

/

bto-be erformed qa ely) eac Amendment No. 96, 98 71

Ain VYNPS TABLE 4.2 NOTES which opens

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

delay relays and

3. Trip system logic calibration shall include only time timers necessary for proper functioning of the trip system.

definition. The

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

1e7eFed. no

. Functional tests and calibrations are not required when systems are

8. required to be operable.

A' The rmocouples as ted with safety ief valves and /afety valv sed for back-up p9 ition indicati shall be po ion, that may b to be ope

le every operati cycle.

instrumentation. The

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

operating cycle will of operation

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

function. If the

12. Trip system logic testing is not applicable to this met, (every Refueling Outage) is not required surveillance frequency function the Reactor Mode Switch-Shutdown Position functional testing of switch is placed shall be initiated within I hour after the reactor mode Refueling Outage.

in Shutdown for the purpose of commencing a scheduled function (i.e., RBM

13. Includes calibration of the RBM Reference Downscale when >30% Rated Thermal Power).

upscale function is not bypassed 44-5, X64, +8X, 2-, 212 74 Amendment No. 46, O&, 4-5,

Safety Assessment Discussion of Changes 3.2.G and 4.2.G Post-Accident Monitoring Instrumentation

SAFETY ASSESSMENT OF CHANGES TS 3.2.G/4.2.G - POST - ACCIDENT MONITORING INSTRUMENTATION ADMINISTRATIVE A.1 Inthe revision of the Vermont Yankee Nuclear Power Station (VYN PS) 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 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 4.2.G includes reference to CTS Table 4.2.6 for functional test and calibration requirements for post-accident monitoring instrumentation. CTS 4.2.G is revised, in proposed Surveillance Requirement (SR) 4.2.G.1, to delete the reference to functional testing and include reference to check requirements consistent with CTS Table 4.2.6. (CTS Table 4.2.6 includes Check requirements, but does not include Functional Test requirements). This change is a presentation preference and does not alter the current requirements for periodic testing of post-accident monitoring instrument Functions.

Therefore, this change is considered administrative innature.

A.3 CTS 3.2.H and 4.2.H provide requirements that apply to drywell to torus AP instrumentation. Changes to these CTS drywell to torus AP instrumentation requirements are addressed in the Safety Assessment of Changes for CTS 3.2.H1/4.2.H, Drywell to Torus AP Instrumentation. Therefore, this change does not involve a technical change, but is only a difference of presentation preference and is considered administrative.

A.4 CTS Table 3.2.6 Note 8 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 post-accident monitoring instrumentation channel inoperable solely for performance of surveillances. This allowance is moved to proposed SR 4.2.G.1. This change does not involve a technical change, but isonly a difference of presentation preference. Therefore, this change isconsidered administrative.

A.5 CTS 3.2.G specifies an Applicability for post-accident monitoring instrumentation of 'During reactor power operation." The CTS definition of reactor power operation states "Reactor power operation is any operation with the mode switch inthe Startup/Hot Standby or Run..." This change provides an explicit Applicability, in proposed Table 3.2.6 for each post-accident monitoring instrumentation Function. The specified Applicabilities, in proposed Table 3.2.6, are consistent with the CTS definition of reactor power operation (i.e., RUN and STARTUP/HOT STANDBY). Therefore, this change does not involve a technical change, but isonly adifference of presentation preference and isconsidered administrative. The change, providing explicit Mode or conditions of Applicability for each trip function, is consistent with the ISTS.

VYNPS 1 Revision 0

SAFETY ASSESSMENT OF CHANGES TS 3.2.G/4.2.G - POST - ACCIDENT MONITORING INSTRUMENTATION ADMINISTRATIVE (continued)

A.6 CTS Table 4.2 Notes 2,3,8,10, and 11 provide requirements that apply to ECCS instrumentation. The ECCS instrumentation is located in proposed Specifications 3.2.A and 4.2.A. Therefore, the requirements of CTS Table 4.2 Notes 2,3,8, 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 is only a difference of presentation preference and isconsidered administrative. CTS 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 Table 4.2 Notes 4, 12, and 13 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.

TECHNICAL CHANGES - MORE RESTRICTIVE None TECHNICAL CHANGES - LESS RESTRICTIVE

'Generic" LA.1 The CTS Table 3.2.6 details relating to system design and operation (i.e., type of indication, specific instrument tag numbers, and instrument range) are unnecessary inthe TS and are proposed to be relocated to the Technical Requirements Manual (TRM). Proposed Specification 3.2.G and Table 3.2.6 require the post-accident monitoring instrument Functions to be operable. Inaddition, the proposed Surveillance Requirements inTable 4.2.6 ensure the required instruments are properly tested. These requirements are adequate for ensuring each of the required post-accident monitoring instrument 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 is consistent with the ISTS.

RELOCATED SPECIFICATIONS R.1 3.2.G/4.2.G POST-ACCIDENT INSTRUMENTATION LCO Statement:

During reactor power operation, the instrumentation that displays information inthe Control Room necessary for the operator to initiate 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.

VYNPS 2 Revision 0

SAFETY ASSESSMENT OF CHANGES TS 3.2.G14.2.G - POST - ACCIDENT MONITORING INSTRUMENTATION RELOCATED SPECIFICATIONS R.1 Discussion:

(continued)

Each individual post-accident monitoring parameter has a specific purpose; however, the general purpose for accident monitoring instrumentation is to provide sufficient information to confirm an accident is proceeding per prediction, i.e. automatic safety systems are performing properly, and deviations from expected accident course are minimal.

Comparison to Deterministic Screening Criteria:

The NRC position on application of the deterministic screening criteria to post-accident monitoring instrumentation is documented in letter dated May 7, 1988 from T.E. Murley (NRC) to R.F. Janecek (BWROG). The position was that the post-accident monitoring instrumentation table list should contain, on a plant specific basis, all Regulatory Guide 1.97 Type A instruments specified in the plant's Safety Evaluation Report (SER) on Regulatory Guide 1.97, and all Regulatory Guide 1.97 Category 1 instruments.

Accordingly, this position has been applied to the VYNPS Regulatory Guide 1.97 instruments. Those instruments meeting these criteria have remained in Technical Specifications. The instruments not meeting these criteria have been relocated from the Technical Specifications to plant controlled documents.

The following summarizes the VYNPS position for those instruments currently in Technical Specifications.

From NRC SER dated December 4, 1990,

Subject:

Conformance to Regulatory Guide 1.97 for Vermont Yankee Nuclear Power Station.

Type A Variables

1. Reactor Pressure

2. Reactor Vessel Level

3. Drywell Pressure

4. Drywell Temperature

5. Torus Pressure

6. Torus Water Temperature

7. Torus Water Level

8. Torus Airspace Temperature From Regulatory Guide 1.97 and VYNPS submittal to the NRC dated October 30, 1984, NUREG-0737, Supplement I - Regulatory Guide 1.97, as modified by VYNPS letters dated October 25, 1985, August 11, 1987, July 28, 1988, September 1, 1989, and March 29, 1996, and NRC letter dated April 29, 1993.

VYNPS 3 Revision 0

SAFETY ASSESSMENT OF CHANGES TS 3.2.G/4.2.G - POST - ACCIDENT MONITORING INSTRUMENTATION RELOCATED SPECIFICATIONS R.1 Other Type. Categorv 1 Variables (continued)

Primary Containment Isolation Valve Position Containment and Drywell Hydrogen Concentration Containment and Drywell Oxygen Concentration Primary Containment High Radiation For other post-accident monitoring instrumentation currently inTechnical Specifications, their loss is not risk-significant since the variable they monitored did not qualify as a Type A or Category 1variable (one that is important to safety and needed by the operator, so that the operator can perform necessary normal actions).

Conclusion Since the screening criteria have not been satisfied for non-Regulatory Guide 1.97 Type A or Category 1variable instruments, their associated LCO, Actions and Surveillances may be relocated to the Technical Requirements Manual. Changes to the Technical Requirements Manual are controlled using 10 CFR 50.59. The instruments to be relocated are as follows:

1. Safety/Relief Valve Position from Pressure Switches

2. Safety Valve Position from Acoustic Monitor

3. Stack Noble Gas Effluent VYNPS 4 Revision 0

No Significant Hazards Consideration 3.2.G and 4.2.G Post-Accident Monitoring Instrumentation

GENERIC NO SIGNIFICANT HAZARDS CONSIDERATION ADMINISTRATIVE CHANGES

("A.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 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 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 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 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 ina margin of safety.

VYNPS 1 Revision 0

GENERIC NO SIGNIFICANT HAZARDS CONSIDERATION NGENERIC LESS RESTRICTIVE CHANGES:

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

('LA.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 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 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 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 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.

VYNPS 4 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 isprovided insupport of this conclusion.

I1. 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 inTechnical 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 in existing 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 is consistent 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

NO SIGNIFICANT HAZARDS CONSIDERATION TS 3.2.G/4.2.G - POST ACCIDENT MONITORING INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE There were no specific less restrictive changes identified for this Specification.

1 Revision 0 VYNPS

References 3.2.G and 4.2.G Post-Accident Monitoring Instrumentation

3.2.G/4.2.G REFERENCES Post-Accident Monitoring Instrumentation

1. Regulatory Guide 1.97, "Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," Revision 3, May 1983.

2. NRC letter, M.B. Fairtile (NRC) to L.A. Temblay (VYNPC), "Conformance to Regulatory Guide 1.97 for Vermont Yankee Nuclear Power Station," December 4, 1990.

Proposed Technical Specifications 3.2.H and 4.2.H Drywell to Torus AP Instrumentation

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

I I. Recirculation Pump Trip I. Recirculation Pump Trip Instrumentation Instrumentation The recirculation pump trip 1. The recirculation pump instrumentation for each Trip trip instrumentation shall Function in Table 3.2.7 shall be checked, functionally be operable in accordance tested and calibrated in with Table 3.2.7. accordance with Table 4.2.7.

When a channel is placed in an inoperable status solely for performance of required surveillances, entry into associated Limiting Conditions for Operations 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 recirculation pump trip capability.

2. Perform a Logic System Functional Test, including recirculation pump trip breaker actuation, of recirculation pump trip instrumentation Trip Functions once every Operating Cycle.

J. Deleted J. Deleted.

Amendment No. 64

Current Technical Specifications Markups 3.2.H and 4.2.H Drywell to Torus AP Instrumentation

VYNPS 3.2 LIMITING CONDITIONS FOR 4.2 SURVEILLANCE REQUIREMENTS OPERATION or

2. If the required action and 3. Perform an instrument associated completion time calibration, except for of Specification 3.2.F.1 is the radiation detectors, not met, within the using a current source following 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s: once every three (3) months. The trip setting

a. Isolate the mechanical shall be < 3.0 times vacuum pump; or background at rated thermal power.

b. Isolate the main steam lines; or 4. Perform an instrument calibration using a

c. Place the reactor in radiation source once each the SHUTDOWN Mode. refueling outage.

5. Perform a logic system functional test, including mechanical vacuum pump isolation valve, once each operating cycle.

G. Post-Accident Instrumentation G. Post-Accident Instrumentation 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 P4^>>

-zdfC TO 5'C6 Dd*4T shall be operable in accordance with Table 3.2.6.

H. Drywell to Torus, AP Drywell to Instrumentation Instrumentation /

1. During gactor power / The Drywell to T uS AP operation, the Drywell to t Instrumentatio /hall be Tor AP Instrumentation calibrated on every six (r order #1-156-3 and months and instrument strument DPI-1-158-6) check will e made once per shall be operable exce shift. /

as specified in 3.2. 2.

2.From and after the date that one of the ywell to Torus AP instr ents is made or found o be inoperable r any reason, reactor op ation is permissi e only during

\ thesuc eding thirty days unless the instrument is ICE. 21236 Amendment No. &4, i&, 96, 9-, 411, 132, --- - - -

E47. VYNPS 3.2 LIMITING CONDITIONS FOR l 4.2 SURVEILLANCE REQUIREMENTS OPERATION soonermadeoerable. If both inst ments are ade or foun to be inop able, and i ication ca ot be res red within six hour p iod, an ord ly utdown sha -- ED be initiate and the reactor s 11 be in a ot shutdown condition i six hours a d a cold s tdown condi on in the ollowing eigh en hours.

I. Recirculation Pump Trip I. Recirculation Pump Tri Instrumentation Instrumentation\

During reactor power The Recirculation Pump Trip \

operation, the Recirculation Instrumentation shall be Pump Trip Instrumentation functionally tested andl shall be operable calibrated in accordance with in accordance with Table 4.2.1.

Table 3.2.1.

J. Deleted J. Deleted K. Degraded Grid Protective K. Degraded Grid Protective System System During reactor power The emergency bus operation, the emergency bus undervoltage instrumentation undervoltage instrumentation shall be functionally tested shall be operable in and calibrated in accordance accordance with Table 3.2.8. with Table 4.2.8.

L. Reactor Core Isolation L. Reactor Core Isolation Cooling System Actuation Cooling System Actuation When the Reactor Core Instrumentation and Logic Isolation Cooling System is Systems shall be required in accordance with functionally tested and Specification 3.5.G, the calibrated as indicated in/

instrumentation which Table 4.2.9./

initiates actuation of this system shall be operable in accordance with Table 3.2.9.

t<Vsea o A4947CR6{

Amendment No. iG, 06, "G, 111, 212 37

Safety Assessment Discussion of Changes 3.2.H and 4.2.H Drywell to Torus AP Instrumentation

SAFETY ASSESSMENT OF CHANGES CTS: 3.2.H/4.2.H - DRYWELL TO TORUS AP 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.

TECHNICAL CHANGES - MORE RESTRICTIVE None TECHNICAL CHANGES - LESS RESTRICTIVE LC.1 CTS 3.2.H and 4.2.H specify requirements for the drywell to torus Ap instrumentation. This monitoring instrumentation does not necessarily relate directly to maintaining the monitored parameter (drywell to torus Ap) within limits. The ISTS do not specify indication-only equipment to be operable to support operability of a system or component or maintaining variables within limits. Control of the availability of, and necessary compensatory activities if not available, for indication and monitoring instruments are addressed by plant procedures and policies. Therefore, these requirements are to be relocated to the Technical Requirements Manual. Inaddition, TS 3.7.A.914.7.A.9, Drywell/Suppression Chamber dip, provides regulatory control over the requirement to maintain drywell to torus Ap within limits. As a result, the details to be relocated are not required to be inthe Technical Specifications to provide adequate protection of the public health and safety.

Changes to the Technical Requirements Manual are controlled using 10 CFR 50.59. Not including these requirements inthe TS is consistent with the ISTS.

RELOCATED SPECIFICATIONS None VYNPS 1 Revision 0

No Significant Hazards Consideration 3.2.H and 4.2.H Drywell to Torus AP Instrumentation

GENERIC NO SIGNIFICANT HAZARDS CONSIDERATION ADMINISTRATIVE CHANGES

(*A.xC 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 existing Technical Specifications. The reformatting, renumbering, and rewording process involves no technical changes to the existing Technical Specifications. As such, this change is administrative 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 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 innature. Therefore, the change does not involve a significant reduction in a margin of safety.

VYNPS 1 Revision 0

GENERIC NO SIGNIFICANT HAZARDS CONSIDERATION

'GENERIC" LESS RESTRICTIVE CHANGES:

RELOCATION OF INSTRUMENTATION ONLY REQUIREMENTS

('LC.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 in the probability or consequences of an accident previously evaluated?

The proposed change relocates instrumentation requirements from the Technical Specifications to the Bases, UFSAR, procedures, or other plant controlled documents. These requirements are not considered inthe safety analysis. 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 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 requirements 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 inthe methods governing normal plant operation. The proposed change will not impose or eliminate any requirements, and adequate control of the 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 in a margin of safety?

The proposed change will not reduce a margin of safety because it has no impact on any safety analysis assumption. Inaddition, the requirements 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 requirements 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 requirements 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 instrumentation requirements ensures no significant reduction inthe margin of safety.

VYNPS 5 Revision 0

NO SIGNIFICANT HAZARDS CONSIDERATION CTS: 3.2.H/4.2.H - DRYWELL TO TORUS AP INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE There were no specific less restrictive changes identified for this Specification.

VYNPS 1 Revision 0

Proposed Technical Specifications 3.2.1 and 4.2.1 Recirculation Pump Trip Instrumentation

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

I H. Deleted. I I. Recirculation Pump Trip I. Recirculation Pump Trip Instrumentation Instrumentation The recirculation pump trip 1. The recirculation pump instrumentation for each Trip trip instrumentation shall Function in Table 3.2.7 shall be checked, functionally be operable in accordance tested and calibrated in with Table 3.2.7. accordance with Table 4.2.7.

When a channel is placed in an inoperable status solely for performance of required surveillances, entry into associated Limiting Conditions for Operations 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 recirculation pump trip capability.

2. Perform a Logic System Functional Test, including recirculation pump trip breaker actuation, of recirculation pump trip instrumentation Trip Functions once every Operating Cycle.

J. Deleted J. Deleted.

64 Amendment No.

VYNPS Table 3.2.7 (page 1 of 1)

Recirculation Pump Trip Instrumentation ACTIONS WHEN APPLICABLE REQUIRED REQUIRED MODES OR OTHER CHANNELS CHANNELS SPECIFIED PER TRIP ARE TRIP SETTING CONDITIONS SYSTEM INOPERABLE TRIP FUNCTION RUN 2 Note 1 2 82.5 inches

1. Low-Low Reactor Vessel Water Level RUN 2 Note 1 5 10 seconds

2. Time Delay RUN 2 Note 1 5 1150 psig

3. High Reactor Pressure 65 Amendment No.

VYNPS Table 3.2.7 ACTION Notes

1. With one or more recirculation pump trip instrumentation channels inoperable, take all of the applicable Actions in Notes l.a, L.b and l.c below.

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

1) Restore the channel to operable status within 14 days; or

2) Place the inoperable channel in trip within 14 days (not applicable for Trip Function 2 channels and not applicable if the inoperable channel is the result of an inoperable recirculation pump trip breaker).

not

b. With Trip Functions 1 and 2 with recirculation pump trip capability maintained or with Trip Function 3 with recirculation pump trip capability not maintained:

1) Restore recirculation pump trip capability within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

c. With Trip Functions 1, 2 and 3 with recirculation pump trip capability not maintained:

1) Restore recirculation pump trip capability for all but one Trip Function 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, l.b or l.c is not met, immediately take the applicable Action of Note 2.a or 2.b.

2. a. Remove affected recirculation pump from service within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; or

b. Place the plant in STARTUP/HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

66 Amendment No.

VYNPS Table 4.2.7 (page 1 of 1)

Recirculation Pump Trip Instrumentation Tests and Frequencies TRIP FUNCTION CHECK FUNCTIONAL TEST CALIBRATION

1. Low-Low Reactor Vessel Once/Day Every 3 Months Every 3 Months~l),

Water Level Once/Operating Cycle

2. Time Delay NA NA Every 3 Months Once/Day Every 3 Months Every 3 Months(l),

3. High Reactor Pressure Once/Operating Cycle (1) Trip unit calibration only.

67 Amendment No.

Proposed Bases 3.2.1 and 4.2.1 Recirculation Pump Trip Instrumentation

VYNPS BASES: 3.2.I/4.2.I RECIRCULATION PUMP TRIP INSTRUMENTATION BACKGROUND System The Anticipated Transient Without Scram (ATWS) Prevention/Mitigation initiates a Recirculation Pump Trip (RPT), adding negative reactivity, the following events in which a scram does not but should occur, to lessen effects of an ATWS event. Tripping the recirculation pumps adds negative flow reactivity from the increase in steam voiding in the core area as core decreases. When Low - Low Reactor Vessel Water Level or High Reactor Pressure setpoint is reached, the reactor recirculation motor generator (RRMG) field breakers trip.

System The RPT Instrumentation (Ref. 1) of the ATWS Prevention/Mitigation switches that are necessary to cause initiation includes sensors, relays, and include electronic equipment (e.g., trip units) that of an RPT. The channels the compares measured input signals with pre-established setpoints. When setpoint is exceeded, the channel output relay actuates, which then outputs an RPT signal to the trip logic.

The RPT Instrumentation consists of two independent and identical trip systems (A and B), with two channels of High Reactor Pressure and two Each channels of Low - Low Reactor Vessel Water Level in each trip system.

RPT Instrumentation trip system is a two-out-of-two logic for each Trip Function. Thus, either two Low - Low Reactor Water Level or two High Reactor one Pressure signals will trip a trip system. In addition, a combination of Low - Low Reactor Vessel Water Level signal and one High Reactor Pressure of signal (in the same trip system) will trip the trip system. The outputs either trip the channels in a trip system are combined in a logic so that RRMG system will trip both recirculation pumps (by tripping the respective Water Level channel output field breakers). Each Low - Low Reactor Vessel must remain below the setpoint for approximately 10 seconds for the channel output to provide an actuation signal to the associated trip system.

There is one RRMG field breaker provided for each of the two recirculation pumps for a total of two breakers. The output of each trip system is provided to both RRMG field breakers.

APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY The RPT Instrumentation is not assumed to mitigate any accident or transient in in the safety analysis. The RPT Instrumentation initiates an RPT to aid of the fuel cladding following events in which a preserving the integrity scram does not, but should, occur. Based on its contribution to the reduction of overall plant risk, however, the instrumentation meets Criterion 4 of 10 CFR 50.36(c) (2)(ii).

of The operability of the RPT Instrumentation is dependent on the operability the individual instrumentation channel Trip Functions. 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 the worst case design basis conditions, the associated trip will occur within

-- -'---. TOTS 80 pimenumenL Diu.

VYNPS BASES: 3.2.1/4.2.1 RECIRCULATION PUMP TRIP INSTRUMENTATION APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) considered Trip Settings specified in Table 3.2.7. As a result, a channel is calculational as-inoperable if its actual trip setpoint is not within the The actual trip setpoint is found tolerances specified in plant procedures.

calibrated consistent with applicable setpoint methodology assumptions.

trip channel operability also includes the associated recirculation pump breakers (i.e., RRMG field breakers).

to The individual Trip Functions are required to be operable in the RUN Mode Reactor Protection System by protect against common mode failures of the ATWS providing a diverse trip to mitigate the consequences of a postulated and Low - Low Reactor Vessel Water Level event. The High Reactor Pressure reactor Trip Functions are required to be operable in the RUN Mode, since the significant power and the recirculation system could be at high is producing increases or low flow. During this Mode, the potential exists for pressure STARTUP/HOT STANDBY Mode, the water level, assuming an ATWS event. In the thus, reactor is at low power and the recirculation system is at low flow; pressure increase or low water level, assuming an the potential is low for a the RPT Instrumentation is not necessary. In HOT ATWS event. Therefore, rods SHUTDOWN and COLD SHUTDOWN, the reactor is shut down with all control the possibility of a inserted; thus, an ATWS event is not significant and is negligible. In the significant pressure increase or low water level remains REFUELING Mode, the one rod out interlock ensures that the reactor subcritical; thus, an ATWS event is not significant.

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

1, 2. Low - Low Reactor Vessel Water Level and Time Delay Low RPV water level indicates the capability to cool the fuel may be could threatened. Should RPV water level decrease too far, fuel damage result. Therefore, RPT is initiated at low-low RPV water level to aid in maintaining level above the top of the active fuel. The reduction of core neutron flux and thermal power and, therefore, the rate of flow reduces the coolant boiloff.

Reactor vessel water level signals are initiated from four level transmitters of that sense the difference between the pressure due to a constant column water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.

in Four channels of Low - Low Reactor Vessel Water Level, with two channels required to be operable to ensure that no each trip system, are available and can preclude an RPT from this Trip Function on a single instrument failure

- Low valid signal. In addition, a time delay is associated with each Low Vessel Reactor Vessel Water Level channel which delays the Low - Low Reactor providing input to the Water Level Trip Function channel output signal from Delay, with two channels in associated trip system. Four channels of Time to be operable to ensure that no each trip system, are available and required single instrument failure can preclude an RPT from the Low - Low Reactor Vessel Water Level Trip Function on a valid signal.

80a Amendment No.

VYNPS BASES: 3.2.I/4.2.1 RECIRCULATION PUMP TRIP INSTRUMENTATION APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

The Low - Low Reactor Vessel Water Level Trip Setting is chosen so that RPT will not interfere with the Reactor Protection System. The Trip Setting is referenced from the top of enriched fuel. The Trip Setting of the Time Delay associated with the Low - Low Reactor Vessel Water Level Trip Function is chosen to avoid making the consequences of a loss of coolant accident more severe while ensuring the delay has an insignificant affect on the ATWS consequences.

3. High Reactor Pressure Excessively high RPV pressure may rupture the RCPB. An increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This increases neutron flux and thermal power, which could potentially result in fuel failure and overpressurization. The High Reactor Pressure Trip Function initiates an RPT for transients that result in a pressure increase, counteracting the pressure increase by rapidly reducing core power generation. For the overpressurization event, the RPT aids in the termination of the ATWS event and, along with the safety valves, limits the peak RPV pressure to within the required limit.

The High Reactor Pressure signals are initiated from four pressure transmitters that monitor reactor pressure. Four channels of High Reactor Vessel Pressure, with two channels in each trip system, are available and are required to be operable to ensure that no single instrument failure can preclude an RPT from this Trip Function on a valid signal. The High Reactor Vessel Pressure Trip Setting is chosen to provide an adequate margin to the maximum allowable Reactor Coolant System pressure.

ACTIONS Table 3.2.7 ACTION Note 1 For Trip Functions 1, 2, and 3, with one or more Trip Function channels inoperable, but with RPT trip capability for each Trip Function maintained (refer to next paragraph), the RPT instrumentation is capable of performing the intended function. However, the reliability and redundancy of the RPT Instrumentation is reduced, such that a single failure in the remaining trip system could result in the inability of the RPT Instrumentation to perform the intended function. Therefore, only a limited time is allowed to restore the inoperable channels to operable status. Because of the diversity of sensors available to provide trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Trip Functions, and the low probability of an event requiring the initiation of RPT, 14 days is provided to restore the inoperable channel (Table 3.2.7 ACTION Note l.a.l)).

Alternately, for Trip Functions 1 and 3, the inoperable channel may be placed in trip (Table 3.2.7 ACTION Note l.a.2)), since this would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. Inoperable channels may be placed in trip using test jacks or other permanently installed circuits. As noted in Table 3.2.7 ACTION Note 1.a.2), placing the channel in trip with no Amendment No. Bob

VYNPS BASES: 3.2.1/4.2.I RECIRCULATION PUMP TRIP INSTRUMENTATION ACTIONS (continued)

Trip further restrictions is not allowed if the inoperable channel is a or is the result of an Function 2 channel (i.e., Time Delay Trip Function) inoperable breaker, since this may not adequately compensate for the breaker inoperable Trip Function 2 channel or inoperable breaker (e.g., the that it will not open), as applicable. If it is not may be inoperable such the channel in trip (e.g., as in the case where placing the desired to place inoperable channel in trip would result in an RPT), or if the inoperable 2

channel is the result of an inoperable breaker, Table 3.2.7 ACTION Note must be entered and its required Actions taken.

actions Table 3.2.7 ACTION Note l.b is intended to ensure that appropriate inoperable, untripped channels within the same Trip are taken if multiple, RPT trip Function result in the Trip Function 1 and 2 not maintaining A Trip capability or Trip Function 3 not maintaining RPT trip capability.

Function is considered to be maintaining RPT trip capability when sufficient channels are operable or in trip such that the RPT Instrumentation will two trip generate a trip signal from the given Trip Function in either of the signal, and both recirculation pumps can be tripped. For systems on a valid Trip Function in Trip Functions 1 and 2, this requires two channels of each the RRMG field breakers to the same trip system to be operable or in trip and 3, this requires two channels in be operable or in trip. For Trip Function to the same trip system to be operable or in trip and the RRMG field breakers The 72 hour Completion Time is sufficient for the be operable or in trip.

of operator to take corrective action (e.g., restoration or tripping channels) and takes into account the likelihood of an event requiring Trip actuation of the RPT instrumentation during this period and that Functions 1 and 2 or Trip Function 3 still maintain RPT trip capability.

Table 3.2.7 ACTION Note l.c is intended to ensure that appropriate Actions are taken if multiple, inoperable, untripped channels within Trip Functions trip 1, 2, and 3 result in Trip Functions 1, 2, and 3 not maintaining RPT capability. The description of a Trip Function maintaining RPT trip Time capability is discussed in the paragraph above. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion of the Trip Functions is sufficient for the for restoring all but one of operator to take corrective action and takes into account the likelihood during this period.

an event requiring actuation of the RPT Instrumentation Table 3.2.7 ACTION Note 2 With any required Action and associated completion time not met, the plant LCO does must be brought to a Mode or other specified condition in which the not apply. To achieve this status, the plant must be brought to at least STARTUP/HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months (Table 3.2.7 ACTION Note 2.b).

service Alternately, the associated recirculation pump may be removed from (Table 3.2.7 since this performs the intended function of the instrumentation ACTION Note 2.a). The allowed Completion Time of 6 hours is reasonable, full based on operating experience, both to reach STARTUP/HOT STANDBY from power conditions and to remove a recirculation pump from service in an orderly manner and without challenging plant systems.

B0c Amendment No.

VYNPS BASES: 3.2.I/4.2.I RECIRCULATION PUMP TRIP INSTRUMENTATION SURVEILLANCE REQUIREMENTS Surveillance Requirement 4.2.I.1 As indicated in Surveillance Requirement 4.2.I.1, RPT Instrumentation shall be checked, functionally tested and calibrated as indicated in Table 4.2.7.

Table 4.2.7 identifies, for each Trip Function, the applicable Surveillance Requirements.

Surveillance Requirement 4.2.I.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 recirculation pump trip 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. 2) 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 recirculation pumps will trip when necessary.

Surveillance Requirement 4.2.I.2 The Logic System Functional Test demonstrates the operability of the required initiation logic and simulated automatic operation for a specific channel. A system functional test of the recirculation pump trip breakers (i.e., RRMG field breakers) is included in this Surveillance to provide complete testing of the assumed safety function. Therefore, if an RRMG field breaker is incapable of operating, the associated instrument channel(s) 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.

Table 4.2.7, Check Performance of an Instrument Check once per day, for Trip Functions 1 and 3, 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 Amendment No. 80d

VYNPS BASES: 3.2.I/4.2.I RECIRCULATION PUMP TRIP INSTRUMENTATION SURVEILLANCE REQUIREMENTS (continued) 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 4.2.7, Functional Test For Trip Functions 1 and 3, 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. For Trip Functions 1 and 3, the Frequency of "Every 3 Months" is based on the reliability -analysis of Reference 2.

Table 4.2.7, Calibration For Trip Functions 1, 2, and 3, 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 3, 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 2 and the time interval assumption for trip unit calibration used in the associated setpoint calculation.

REFERENCES

1. UFSAR, Section 7.18.

2. GENE-770-06-1-A, RBases for Changes To Surveillance Test Intervals and Allowed Out-of-Service Times For Selected Instrumentation Technical Specifications," December 1992.

Amendment No. 80e

Current Technical Specifications Markups 3.2.1 and 4.2.1 Recirculation Pump Trip Instrumentation

,4.f VYNPS 3.2 LIMITING CONDITIONS FOR 4.2 SURVEILLANCE REQUIREMENTS OPERATION sooner made operable. If both instruments are made or found to be inoperable, and indication cannot be restored within a six hour period, an orderly shutdown shall be initiated and the reactor shall be in a hot shutdown condition in six hours and a cold shutdown condition in the following eighteen hours.

I. Recirculation Pump Trip I. Recirculation Pump Trip Instrumentation Instrumentation 1

~~During reactor ow The Xecirculation trump {rip /_

0

-1c' irc u ati on

h e-LZXon-2eeh Xnstrumentation shall be ump grip Lfnstrumentation functionally tested and M _AC shall be operable calibrated in accordance with flgpg J.9c0I4714 in accordance with Table 4.2 cc Ad 1Table 3.2 J. Delee _J. Deleted lK. Degraded Grid Protective K. Degraded Grid Protectv \\

System Syste\\

During reactor power The emergency bus operation, the emergency bus undervoltage instrumentation undervoltage instrumentation shall be functionally tested shall be operable in and calibrated in accordance accoraance wixh TaDle J.2-O. with Table 4.2.8.

L. Reactor Core Isolation L. Reactor Core Isolation Cooling System Actuation Cooling System Actuation When the Reactor Core Instrumentation and Logic Isolation Cooling System is Systems shall be required in accordance with functionally tested and Specification 3.5.G, the calibrated as indicated in instrumentation which Table 4.2.9.

initiates actuation of this system shall be operable in accordance with Table 3.2.9.

bray

<1 Th WeM*ATE De beco4yeC* F#A vp 7V, ep WevL5 sqlv#ibe rdd 4$UC,1CI~tRp AvoJ~iuaj AtIJA.MD$~

,~9e~cS*TO'JpcswP7Ai.fP C4,PM/01uZ Amendment No. &G, "6, 0&, lll, 212 37

[,A. l VYNPS TABLE 3.2 RECIRCULATION PUMP TRIP INSTRUMENTATION I

kTrpaR Setting

> 6' 10 psi Note fh-rjehed ieV-HE -

5 < 1150 psig Note 2- 2 eotj < 10 seconds Note 7 /Trin SvntemseLo i-- ,_Rte 71

-.1 - - --o -

10, 7 - - --

v."

Amendment No. 6#, 4#, 74, -&, a44, 186 43

VYNPS TABLE 3.2.1 /TES o th w Spray>PCI and RPT, subsystems are,*nitiate Core

. < tofhetw / atrip sys .4m. The subsys)1e "B is id ical

< ALto tumenl not av abl If teminimum ru~mber of dperable ins ks orath e the i erable ch~xnel shalV b e tri eest sta ll e ed b the erma ntl 24 hou or statd above, at channe shall be bde oper le with ns LA ax orderly utdown sh4 1 be i *

> and ated t e reacto shall b in the shu>Ldown condiion withii 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 I one-out-of-two taken twice logic.

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

the system is

5. If the minimum number of operable channels are not available, of Specification 3.5 apply.

considered inoperable and the requirements

6. 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 requireflment-of of Specification 3;5.F.2 and 3.5.F.3 shall apply. If the minimum number operable channels is not available in both trip systems, Specifications 3.5.F.3 shall apply.

I

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

performance of

8. When a channel is placed in an inoperable status solely for For -

required surveillances, entry into associated Limiting Conditions prov ed 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 Function FTr-edundant Trip Functlon mantain u3 the associated Trip ecirculation Pump Trip capability.

of

9. When a channel is placed in an inoperable status solely for performance Conditions.For required surveillances, entry into associated Limiting ACTIONS may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Operation and required LPCI:

10. With one or more channels inoperable for Core Spray and/or capability for A. Within one hour from discovery of loss of initiation feature(s) in one division, declare the associatad systems inoperable, and D

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 are not met, immediately declare the associated systems inoperable.

1. With one or more channels inoperable for injection permissive and/or recirculation discharge valve permissive:

for A. Within one hour from discovery of loss of initiation capability systems inoperable, feature(s) in one division, declare the associated and B. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , restore channel to operable status.

A or B C. If required actions and associated completion times of actions are not met, immediately declare the associated systems inoperable.

MJ" Go gaff+ Af" > 44I Amen4iment No. &&, 464, AV 186

VYNPS TABLE 3.2. TES (Cont'd) ~M'44- st--,4 P-. -">

8. With one or more channels inoperable for ADS:

A. Within one hour from discovery of loss of ADS initiation capability in one trip system, declare ADS inoperable, and B. Within 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months from discovery of an inoperable channel concurrent with HPCI or RCIC System inoperable, restore channel to operable status, and C. Within 8 days, restore channel to operable status.

D. If required actions and associated completion times of actions A, B or C are not met, immediately declare ADS inoperable.

With one or more channels inoperable for Recirculation Pump Trip:

Te%'64- i1 S Within one hour from discovery of loss of Recirculation Pump Trip t El capability restore *ne Trip Function r remove the associated LIn recirculation pump from service in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in Startup/Hot Standby Fji in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

S.7.1 Within 14 days from discovery of an inoperable channel, restore channel 0i4tl. I.& to operable status or place in trip, and i.,. @ Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from discovery of 'tr nc ica abiAzit not mektA. maintained, restore trip function o operable status and, If required actions and associated completion times of actions A, B or C are not met, immediately remove th associated recirculation pump from service in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in /Hot Standby in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

A reca;CA ';

\Ol id's,lt{vM is Age"r Ox'm, CE Amendment No. 186 44b

tAil

'emr w. 4L vo fSCA>

(-9(,u( ;.-.'7 '

55a Amendment No. 164

VYNPS TABLE 4.2.45 4 TES .4ND s FREQUEN CIES RG Y TUAX0 INSTRUMENTATION

_Recirroiation P n1 Trin Actwgtion Wse Trip Function Functional Test Calibration Instrument Check

l. l Low-Low Reactor Vessel Every Three Months IOnce/Operating Cycle . Once Each Day Water Level /S

' High Reactor Pressure Every Three Months nce/Operating Cycle Once Each Day SP4',T2 Trip System Logic i

Amendment No. S#, 4-9, 444, 186 63

I Amendment No. 164 71a C4 3 VYNPS TABLE 4.2 NOTES RAzIZ

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

Trip system logic calibration shall include only time delay relays a

(. 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.

5. /De eted Dleted I J (8i Functional tests and calibrations are not required when systems are no required to be operable.

9. Th thermocouples associated with safety/relief valves and safety va v

/ ~position, that may be used for back-up position indication, shall be \

/ ~verified to be operable every operating cycle. .

l10. Separate functional tests are not required for this instrumentation. Thel I t 11.

~calibration and integrated ECCS tests which are performed once perl operating cycle will adequately demonstrate proper equipment operation.

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 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.

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

Amendment No. 63, "#, a45, 45, 464, a46, 244, 212 74

Safety Assessment Discussion of Changes 3.2.1 and 4.2.1 Recirculation Pump Trip Instrumentation

SAFETY ASSESSMENT OF CHANGES TS 3.2.1/4.2.1- RECIRCULATION PUMP TRIP 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 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.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 inthe Safety Assessments of Changes for CTS 3.2.H1/4.2.H, Drywell to Torus AP Instrumentation. Therefore, this change does not involve a technical change, but is only a difference of presentation preference and isconsidered administrative.

A.3 CTS 3.2.1 specifies an Applicability for recirculation pump trip instrumentation of 'During reactor power operation." The CTS definition of reactor power operation states "Reactor power operation isany operation with the mode switch in the Startup/Ilot Standby or Run...' This change provides an explicit Applicability, in proposed Table 3.2.1 for each recirculation pump instrumentation Trip Function. The specified Applicabilities, inproposed Table 3.2.6 are consistent with the CTS definition of reactor power operation as modified by the CTS Table 3.2.1 Note 19 actions to exit the applicability by placing the plant in Startup/Hot Standby (i.e., Run). Therefore, this change does not involve a technical change, but Isonly 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.

AA CTS 4.2.1 includes reference to CTS Table 4.2.1 for functional test and calibration requirements for recirculation pump trip instrumentation. CTS 4.2.1 is revised, inproposed Surveillance Requirement (SR) 4.2.1.1, to also include reference to check requirements consistent with CTS Table 4.2.1. This change is a presentation preference and does not alter the current requirements to periodically perform checks of certain recirculation pump trip instrument trip functions. Therefore, this change is considered administrative in nature.

A.5 CTS Table 3.2.1 Note 8 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 recirculation pump trip instrumentation channel inoperable solely for performance of surveillances. This allowance ismoved to proposed SR 4.2.1.1. This change does not involve atechnical change, but is only a difference of presentation preference. Therefore, this change is considered administrative.

WNPS, 1 Revision 0

SAFETY ASSESSMENT OF CHANGES TS 3.2.1/4.2.1- RECIRCULATION PUMP TRIP INSTRUMENTATION ADMINISTRATIVE (confinued)

A.6 CTS Table 3.2.1 Notes 3,4,5,6,and a portion of Note 8 provide requirements related to Emergency Core Cooling System (ECCS) instrumentation. The ECCS instrumentation is located inproposed Specifications 3.2.A and 4.2A. Therefore, the requirements of CTS Table 3.2.1 Notes 3,4,5,6,and the applicable portion of Note 8 are physically moved and changes addressed inproposed 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.

A.7 CTS Table 4.2.1 includes a requirement to perform a calibration of recirculation pump trip instrumentation Trip System Logic once per Operating Cycle. Similar to other calibration requirements of Trip System Logic inthe VYNPS CTS, the intent of this requirement is to perform a calibration of time delays necessary for proper functioning of the trip system. In proposed Table 4.2.7, this requirement is reflected with explicit requirements to perform periodic calibrations of the required recirculation pump trip instrumentation time delays (i.e.,

proposed Table 4.2.7 Trip Function 2,Time Delay). Therefore, this change does not involve a technical change, but isonly a difference of presentation preference and is considered administrative.

A.8 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.O.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.9 CTS Table 4.2.1 includes Functional Test requirements for the recirculation pump trip Low

- Low Reactor Vessel Water Level and High Reactor Pressure 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 1.0.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 4.2 Note 4 are consistent with the requirements of the Instrument Functional Test definition. The CTS definition of Instrument Functional Test allows the method of testing described inCTS Table 4.2 Note 4 to be used.

Therefore, CTS Table 4.2 Note 4 is unnecessary and isdeleted. This change does not involve a technical change, but isonly a difference of presentation preference and is considered administrative.

VYNPS 2 Revision 0

SAFETY ASSESSMENT OF CHANGES TS 3.2.V4.2.1 - RECIRCULATION PUMP TRIP INSTRUMENTATION ADMINISTRATIVE (continued)

A.10 CTS Table 4.2 Notes 2,10, and 11 provide requirements that apply to ECCS instrumentation. The ECCS instrumentation islocated inproposed Specifications 3.2.A and 4.2A. Therefore, the requirements of CTS Table 4.2 Notes 2, 10, and 11 are physically moved and addressed inthe changes for proposed Specifications 3.2A 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 Islocated 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 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 Table 4.2 Notes 12 and 13 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 isconsidered administrative.

TECHNICAL CHANGES - MORE RESTRICTIVE M.1 CTS Table 3.2.1 Note 19.A provides actions for the condition of a loss of recirculation pump trip capability from all recirculation pump trip instrumentation trip functions, except Trip System Logic. Inthe event all applicable recirculation pump trip instrumentation trip functions lose trip capability, CTS Table 3.2.1 Note 19.A only requires one trip function to be restored within one hour. The intent of this action was to ensure that in the event of a loss of recirculation pump trip capability from both the low reactor vessel water level function and the high reactor pressure function, recirculation pump trip capability for at least one of these two functions would be restored within one hour to support continued operation inthat condition for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . However, CTS Table 3.2.1 includes a Low - Low Reactor Vessel Water Level Trip Function (proposed Table 3.2.1 Trip Function 1); a Time Delay Trip Function (proposed Table 3.2.1 Trip Function 2), which supports the operability of the Low - Low Reactor Vessel Water Level Trip Function; and High Reactor Pressure Trip Function (proposed Trip Function 3)for which CTS Table Note 19 is applicable. As a result, if all three applicable Trip Functions lose trip capability and only one Trip Function is restored, it isstill possible that a loss of recirculation pump trip capability would exist for both the low reactor vessel water level function and the high drywell pressure function (e.g.,

in the case where only trip capability of the Time Delay Trip Function is restored) and operation would be allowed to continue for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inthis condition. Therefore, Inthe event all applicable recirculation pump trip instrumentation trip functions lose trip capability.

CTS Table 3.2.1 Note 19A (proposed Table 3.2.7 Action Note 1.c) is revised to require restoration of trip capability for all but one Trip Function within one hour. A corresponding change Isalso made to CTS Table 3.2.1 Note 19.C. CTS Table 3.2.1 Note 19.C states

'Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from discovery of discovery of one trip function capability not VYNPS 3 Revision 0

SAFETY ASSESSMENT OF CHANGES TS 3.2.1/4.2.1- RECIRCULATION PUMP TRIP INSTRUMENTATION TECHNICAL CHANGES - MORE RESTRICTIVE M.1 maintained ..." The CTS Table 3.2.1 Note 19.C reference to 'one trip function capability not (continued) maintained" is changed to 'Trip Functions I and 2 with trip capability not maintained or Trip Function 3 with trip capability not maintained in proposed Table 3.2.7 Action Note 1.b. This change represents an additional restriction on plant operation to ensure continued operation with a loss of recirculation pump trip capability from both the low reactor vessel water level and high reactor pressure functions is not allowed for longer than one hour.

The change is consistent with the intent of the ISTS (the ISTS does not include a time delay for this type of recirculation pump trip instrumentation).

M.2 CTS Table 3.2.1 Note 19.B requires inoperable recirculation pump trip instrumentation channels to be restored or placed inthe tripped condition. Proposed Table 3.2.7 Action Note l.a provides the same alternative actions as CTS Table 3.2.1 Note 19.B but includes a limitation on the use of the action to place the inoperable channels intrip. Proposed Table 3.2.7 Action Note l.a precludes the use of the action to trip the inoperable channel if the inoperability is associated with Trip Function 2 (i.e., Time Delay) channels if the inoperability is the result of an inoperable recirculation pump trip breaker. These restrictions are added since (1)placing the Time Delay Trip Function channels intrip (as allowed by CTS Table 3.2.1 Note 19.B) does not perform the intended function (providing a time delay for actuation after certain conditions are satisfied) and could make the consequences of a postulated LOCA more severe, and (2)with the channels inoperable due to an inoperable breaker, tripping the affected channels may not adequately compensate for the inoperable breaker (e.g., the breaker may be inoperable such that it will not open). This change represents an additional restriction on plant operation by requiring the channels inthese conditions to be restored to operable status rather than tripped. The change is consistent with the ISTS.

M.3 CTS Table 4.2.1 does not include explicit requirements to calibrate trip units. Proposed Table 4.2.7 requires calibration of the trip units of the following Trip Functions every 3 months: Low - Low Reactor Vessel Water Level (proposed Table 4.2.7 Trip Function 1)and High Reactor Pressure (proposed Table 4.2.7 Trip Function 3). The trip units of these Trip Functions are currently required by CTS Table 4.2.1 to be calibrated with the rest of the associated instrument loops once per operating cycle. Therefore, this change is more restrictive. This change isnecessary to ensure consistency with assumptions regarding trip unit calibration frequency used in the associated setpoint calculations. This change is consistent with the ISTS.

VYNPS 4 Revision 0

SAFETY ASSESSMENT OF CHANGES TS 3.2.1/4.2.1- RECIRCULATION PUMP TRIP INSTRUMENTATION TECHNICAL CHANGES - MORE RESTRICTIVE (continued)

MA CTS Table 3.2.1 requires a Functional Test of the recirculation pump trip instrumentation Trip System Logic. The CTS definition of Logic System Functional Test (CTS 1.0.H) requires where possible for the action during the test to go to completion and actuate the end device (i.e., pumps will be started and valves will be opened). For the recirculation pump trip instrumentation, actuation of the end device would require actuation of the recirculation pump trip breakers. Inproposed SR 4.2.1.2, the Logic System Functional Test of the recirculation pump trip instrumentation Trip Functions explicitly requires the actuation of the recirculation pump trip breakers to be included inthe test. This change represents an additional restriction on plant operation necessary to ensure complete testing of the safety function. This change is consistent with the ISTS.

M.5 CTS Table 4.2.1 includes a requirement to perform a calibration of recirculation pump trip instrumentation Trip System Logic once per Operating Cycle. Similar to other calibration requirements of Trip System Logic inthe VYNPS CTS, the intent of this requirement is to perform a calibration of time delays necessary for proper functioning of the trip system once per Operating Cycle. Inproposed Table 4.2.7, this requirement made more restrictive and is reflected with explicit requirements to perform calibrations of the required recirculation pump trip instrumentation time delays (i.e., proposed Table 4.2.7 Trip Function 2,Time Delay) 'Every 3 Months. This change is necessary to ensure consistency with assumptions regarding the calibration frequency of these time delays used inthe associated setpoint calculations.

TECHNICAL CHANGES - LESS RESTRICTIVE "Generic" LAA The CTS Table 3.2.1 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 3.2.1 and Table 3.2.7 require the recirculation pump trip instrumentation Trip Functions to be operable. In addition, the proposed Surveillance Requirements inTable 4.2.7 ensure the required instruments are properly tested. These requirements are adequate for ensuring each of the required recirculation pump trip instrumentation Trip Functions is 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 is consistent with the ISTS.

LA.2 CTS Table 3.2.1 and associated Note 1 contain design and operational details of the ECCS and RPT instrumentation (i.e., nomenclature for each of the subsystems, that each of the two Core Spray, LPCI and RPT, subsystems are initiated and controlled by a trip system, and that subsystem *Bis identical to subsystem 'A'). These details are not necessary to ensure the operability of ECCS and RPT instrumentation. Therefore, the information inthis note is to be relocated to Specifications 3.2.A and 3.2.1 Bases, as applicable, and reference VYNPS 5 Revision 0

SAFETY ASSESSMENT OF CHANGES TS 3.2.1/4.2.1- RECIRCULATION PUMP TRIP INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE LA.2 to this information isdeleted from WNPS TS. The requirements of Specifications 3.2.A (continued) and 3.2.1 and the associated Surveillance Requirements for the ECCS and RPT 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 in TS is consistent with the ISTS.

LA.3 The Trip Setting associated with reactor vessel water level trip function (proposed Table 3.2.7 Trip Function 1)iscurrently referenced to 'above the top of enriched fuel.' This detail is to be relocated to the Bases. This reference is not necessary to be included in the VYNPS TS to ensure the operability of the associated recirculation pump trip instrumentation. Operability requirements are adequately addressed inproposed Specification 3.2.1, Table 3.2.7 and the specified Trip Setting. As such, this relocated reference is 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 10 CFR 50.59. Not including these details inTS is consistent with the ISTS.

LA.4 The details in the CTS Table 3.2.1 Note 2,relating to the method used for placing channels intrip, are to be relocated to Specification 3.2.1 Bases. The requirements of proposed Table 3.2.7 ACTIONS Notes ensure inoperable channels are placed intrip or the unit is placed in a non-applicable Mode or condition, as appropriate. As a result, the relocated details inthe CTS Table 3.2.1 Note 2 are not necessary for ensuring the appropriate actions are taken Inthe event of inoperable recirculation pump trip instrumentation channels. As such, these relocated details 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 is consistent with the ISTS.

'Specific" L.1 CTS Table 3.2.1 includes requirements for Trip System Logic associated with the recirculation pump trip instrumentation Trip Functions. The Trip System Logic Is considered part of the recirculation pump trip instrumentation Trip Functions and the requirements for the associated Trip System Logic to be operable are encompassed by the definition of operable. Therefore, the CTS Table 3.2.1 listing of Trip System Logic as a separate Trip Function is unnecessary and isdeleted. With the deletion of separate Trip System Logic Trip Function, the actions associated with inoperable Trip System Logic (CTS Table 3.2.1 Note 2) will now be governed by the actions for the individual proposed Table 3.2.7 recirculation pump trip instrumentation Trip Functions. These proposed Table 3.2.7 Action Notes are less restrictive than the CTS Table 3.2.1 Note 2 actions. However, the proposed actions will ensure, inthe event of inoperabilities, that consistent actions are applied to both recirculation pump trip instrumentation Trip Functions and their associated VYNPS 6 Revision 0

SAFETY ASSESSMENT OF CHANGES TS 3.2.1/4.2.1- RECIRCULATION PUMP TRIP INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE L.1 Trip System Logic for the same level of degradation. This change is acceptable, since the (conUnued) allowed outage times of the proposed Table 3.2.7 Action Notes will limit operation to within the bounds of the applicable analysis, i.e., GENE-770-06-1-A, "Bases for Changes to Surveillance Test Intervals and Allowed Outage Times For Selected Instrumentaton Technical Specifications," December 1992. Application of this analysis to the WNPS recirculation pump trip instrumentation Trip Functions, including the associated Trip System Logic, was approved by the NRC in VYNPS 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 3.2.1 and 4.2.1 Recirculation Pump Trip Instrumentation

GENERIC NO SIGNIFICANT HAZARDS CONSIDERATION ADMINISTRATIVE CHANGES Labeled Comments/Discussions) eA.x 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 involves reformatting, renumbering, and rewording. The reformatting, renumbering, and rewording process involves no technical changes to the exising Technical Specifications. As such, this change is administrative 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 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 (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 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 provides more stringent requirements for operation of the facility. These more stringent requirements do not result in operation 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 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 inthe 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 ina margin of safety.

3 Revision 0 VYNPS

GENERIC NO SIGNIFICANT HAZARDS CONSIDERATION OGENERIC' LESS RESTRICTIVE CHANGES:

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

VYNPS 4 Revision 0

NO SIGNIFICANT HAZARDS CONSIDERATION TS 3.2.1/4.2.1- RECIRCULATION PUMP TRIP 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 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 recirculation pump trip instrumentation are inoperable due to inoperable Trip System Logic. The recirculation pump trip instrumentation isnot 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 recirculation pump trip instrumentation channels will continue to be limited in accordance with Technical Specifications. Since the level of degradation allowed inthe proposed actions is 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 is not allowed when an associated recirculation pump trip instrumentation channel is not capable of performing is 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 recirculation pump trip instrumentation are inoperable. No change isbeing 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 in a margin of safety.

VYNPS, 1 Revision 0

References 3.2.1 and 4.2.1 Recirculation Pump Trip Instrumentation

3.2.1/4.

2.1 REFERENCES

Recirculation Pump Trip Instrumentation

1. UFSAR, Section 7.18.

2. GENE-770-06-1-A, "Bases for Changes To Surveillance Test Intervals and Allowed Out-of-Service Times For Selected Instrumentation Technical Specifications,"

December 1992.

Proposed Technical Specifications 3.2.K and 4.2.K Degraded Grid Protective System

VYNPS 3.2 LIMITING CONDITIONS FOR 4.2 SURVEILLANCE REQUIREMENTS OPERATION K. Degraded Grid Protective K. Degraded Grid Protective System System The emergency bus The emergency bus undervoltage instrumentation undervoltage instrumentation for each Trip Function in shall be functionally tested Table 3.2.8 shall be operable and calibrated in accordance in accordance with Table with Table 4.2.8.

3.2.8.

Amendment No. 6B

VYNPS Table 3.2.8 (page 1 of 1)

Degraded Grid Protective System Instrumentation ACTIONS WHEN APPLICABLE REQUIRED MODES OR OTHER REQUIRED CHANNELS SPECIFIED CHANNELS ARE TRIP SETTING TRIP FUNCTION CONDITIONS PER BUS INOPERABLE

1. Degraded Bus Voltage

a. Voltage (1) 2 Note 1 2 3660 volts and

3740 volts

b. Time Delay (1) 1 Note 2 2 9 seconds and 5 11 seconds (1) When the associated diesel generator is required to be operable.

Amendment No. 69

VYNPS Table 3.2.8 ACTION Notes

1. With one or more required Degraded Bus Voltage Trip Function 1 channels inoperable:

a. Place the inoperable channel in trip within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

If the Action and associated completion time of Note l.a are not met, immediately declare the associated diesel generator inoperable.

2. With one or more required Degraded Bus Voltage Trip Function 2 channels inoperable:

a. Restore the inoperable channel to operable status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

If the Action and associated completion time of Note 2.a are not met, immediately declare the associated diesel generator inoperable.

70 Amendment No.

VYNPS Table 4.2.8 (page 1 of 1)

Degraded Grid Protective System Instrumentation Tests and Frequencies TRIP FUNCTION FUNCTIONAL TEST CALIBRATION

1. Degraded Bus Voltage

a. Voltage (1) Once/Operating Cycle

b. Time Delay (1) Once/Operating Cycle (1) Separate Functional Tests are not required for this Trip Function. Trip Function operability is demonstrated during Trip Function Calibration and integrated ECCS tests performed once per Operating Cycle.

Amendment No. 71

Proposed Bases 3.2.K and 432.K Degraded Grid Protective System

VYNPS BASES: 3.2.K/4.2.K DEGRADED GRID PROTECTIVE SYSTEM INSTRUMENTATION BACKGROUND Successful operation of the required safety functions of the Emergency Core Cooling Systems (ECCS) is dependent upon the availability of adequate power sources for energizing the various components such as pump motors, motor operated valves, and the associated control components. The Degraded Grid Protective System instrumentation monitors the 4.16 kV emergency buses.

Offsite power is the preferred source of power for the 4.16 kV emergency buses. If the monitors determine that insufficient voltage is available and an ECCS initiation signal is present, the buses are disconnected from the offsite power sources and connected to the onsite diesel generator (DG) power sources.

Each 4.16 kV emergency bus has its own independent Degraded Grid Protective System instrumentation and associated trip logic. The voltage for each bus is monitored for degraded voltage.

The Degraded Bus Voltage - Voltage Trip Function is monitored by two undervoltage relays for each 4.16 kV emergency bus, whose outputs are arranged in a two-out-of-two logic configuration (Ref. 1). For the Degraded Bus Voltage - Time Delay Trip Function, two channels for each 4.16 kV emergency bus are provided. However, only one Degraded Bus Voltage - Time Delay channel per bus is dedicated to the DG start function. The other Degraded Bus Voltage - Time Delay channel per bus is dedicated to a control room annunciator function from which manual action is taken for degraded grid protection when an accident signal is not present. The Degraded Bus Voltage

- Time Delay Trip Function is nominally adjusted to 10 seconds since this would be indicative of a sustained degraded voltage condition. When a Degraded Bus Voltage - Voltage Trip Function setpoint has been exceeded and persists for nominally ten seconds, either one of the two Degraded Bus Voltage - Voltage Trip Function channels on an associated 4.16 kV emergency bus will actuate a control room annunciator to alert the operator of the degraded voltage condition. (However, the associated control room annunciator is not subject to the requirements of this Technical Specification) If this sustained degraded voltage condition occurs coincident with a loss of coolant accident (LOCA), the 4.16 kV emergency buses are disconnected from the offsite power sources and connected to the DG power sources. If the sustained degraded voltage condition does not exist at the time of a LOCA, 4.16 kV emergency buses are not disconnected from the offsite power sources and the ECCS loads will start immediately from their normal supplies.

APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY The degraded grid protection assures the ECCS loads and other assumed systems powered from the DGs are powered from the offsite power system as long as offsite power system voltage is within an acceptable value and it assures that loads powered from the DGs when bus voltage is insufficient for continuous operation of the connected loads. The Degraded Grid Protective System instrumentation is required for Engineered Safety Features to function in any accident with a degradation or loss of offsite power. The required channels of Degraded Grid Protective System instrumentation ensure that the ECCS and other assumed systems powered from the DGs, provide plant protection in the event of any of the Reference 2 and 3 analyzed accidents in which a loss of offsite power is assumed. The initiation of the DGs on degradation or loss of offsite power, and subsequent initiation of the ECCS, ensure that the requirements of 10 CFR 50.46 are met.

Amendment No. Bof

VYNPS BASES: 3.2.K/4.2.K DEGRADED GRID PROTECTIVE SYSTEM INSTRUMENTATION APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

Accident analyses credit the loading of the DGs based on the loss of offsite power coincident with a loss of coolant accident (LOCA). The diesel starting and loading times have been included in the delay time associated with each safety system component requiring DG supplied power following a loss of offsite power.

3 of The Degraded Grid Protective System instrumentation satisfies Criterion 10 CFR 50.36(c)(2)(ii).

The operability of the Degraded Grid Protective System instrumentation is dependent on the operability of the individual instrumentation channel Trip Functions. 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 3.2.8.

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 specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below for the Degraded Grid Protective System instrumentation Trip Functions.

l.a, 1.b. Degraded Bus Voltage - Voltage and Degraded Bus Voltage - Time Delay A reduced voltage condition on a 4.16 kV emergency bus indicates that, while offsite power may not be completely lost to the respective emergency bus, available power may be insufficient for starting large ECCS motors without risking damage to the motors that could disable the ECCS function.

Therefore, power supply to the bus is transferred from offsite power to onsite DG power when the voltage on the bus drops below the Degraded Bus Voltage - Voltage Trip Function trip setpoint, is sustained in a degraded condition for approximately 10 seconds and a LOCA condition exists (as indicated by ECCS Low - Low Reactor Vessel Water Level or High Drywell Pressure Trip Function signals). This ensures that adequate power will be available to the required equipment.

The Degraded Bus Voltage Trip Settings are low enough to prevent inadvertent power supply transfer, but high enough to ensure that sufficient power is available to the required equipment. The Time Delay Trip Settings are long enough to provide time for voltage on the station emergency bus to recover from transients such as motor starts or fault clearing, but short enough to ensure that the operating equipment is not damaged by low voltage.

Two channels of Degraded Bus Voltage - Voltage Trip Function and one channel of Degraded Bus Voltage - Time Delay Trip Function per associated bus are required to be operable when the associated DG is required to be operable to ensure that no single instrument failure can preclude the DG function.

Amendment No. 80g

VYNPS BASES: 3.2.K/4.2.K DEGRADED GRID PROTECTIVE SYSTEM INSTRUMENTATION ACTIONS Table 3.2.8 ACTION Note 1 With one or more required channels of the Degraded Bus Voltage - Voltage Trip Function inoperable, the Trip Function is not capable of performing the intended function. Therefore, only 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is allowed to restore the inoperable channel to operable status. 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 3.2.8 ACTION Note l.a. The inoperable channel may be tripped using test jacks or other permanently installed circuits. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure (within the Degraded Grid Protective System instrumentation), and allow operation to continue. 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.

If the Action and associated completion time of Table 3.2.8 ACTION Note l.a are not met, the associated Trip Function is not capable of performing the intended function. Therefore, the associated DG(s) is declared inoperable immediately. This requires entry into applicable LCO and required Actions of the DG Technical Specifications, which provide appropriate actions for the inoperable DG(s).

Table 3.2.8 ACTION Note 2 With one or more required channels of the Degraded Bus Voltage - Time Delay Trip Function inoperable, the Trip Function is not capable of performing the intended function. Therefore, only 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is allowed to restore the inoperable channel to operable status (Table 3.2.8 ACTION Note 2.a). Table 3.2.8 ACTION Note 2.a. 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. 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 of channels.

If the Action and associated completion time of Table 3.2.8 ACTION Note 2.a are not met, the associated Trip Function is not capable of performing the intended function. Therefore, the associated DG(s) is declared inoperable immediately. This requires entry into applicable LCO and required Actions of the DG Technical Specifications, which provide appropriate actions for the inoperable DG(s).

SURVEILLANCE REQUIREMENTS Surveillance Requirement 4.2.K.1 As indicated in Surveillance Requirement 4.2.K.1, Degraded Grid Protective System instrumentation shall be functionally tested and calibrated as indicated in Table 4.2.8. Table 4.2.8 identifies, for each Trip Function, the applicable Surveillance Requirements.

Amendment No. 80h

VYNPS BASES: 3.2.K/4.2.K DEGRADED GRID PROTECTIVE SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS (continued)

Table 4.2.8, Functional Test For Trip Functions l.a and l.b, as indicated in Table 4.2.8 Footnote (1),

separate Functional Tests are not required since Trip Function operability is demonstrated during the Trip Function Calibration and integrated ECCS test performed once per Operating Cycle. For the Trip Function Calibration, the "once per Operating Cycle" Frequency is based upon the time interval assumptions for calibration used in the determination of the magnitude of equipment drift in the associated setpoint analyses. For the integrated ECCS test, the "once per Operating Cycle" Frequency 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 integrated ECCS test when performed at the specified Frequency.

Table 4.2.8, Calibration For Trip Functions l.a and l.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.

REFERENCES

1. UFSAR, Section 8.5.3.

2. UFSAR, Section 6.5.

3. UFSAR, Chapter 14.

Amendment No. 80i

Current Technical Specifications Markups 3.2.K and 4.2.K Degraded Grid Protective System

IIT1 VYNPS 3.2 LIMITING CONDITIONS FOR 4.2 SURVEILLANCE REQUIREMENTS OPERATION sooner made operable. If both instruments are made or found to be inoperable, and indication cannot be restored within a six hour period, an orderly shutdown shall be initiated and the reactor shall be in a hot shutdown condition in six hours and a cold shutdown --EI' condition in the following eighteen hours. amars 5P*47TS

% P4ea I

4.

9 I. Recirculation Pump Trip I. Recirculation Pump Trip Instrumentation Instrumentati.on During reactor power The Recirculation Pump Trip operation, the Recirculation Instrumentation shall be Pump Trip Instrumentation functionally tested and shall be operable calibrated in accordance wil in accordance with Table 4.2.1.

I Table 3.2.1.

J. Deleted J. Deleted K. I.

Degraded Grid Protective K. Degraded Grid Protective System System During rieactor powe~ The emergency bus temergency bus undervoltage instrumentation unndvo age instrumentation shall be functionally tested k__,------ _ shall be operable in and calibrated in accordance

/'%?.dock SCP[ accordance with Table 3.2.8. with Table 4.2.8.

USReactor Core Isolation _ L. Reactor Core Isolation Cooling System Actuation Cooling System Actuation When the Reactor Core Instrumentation and Logic Isolation Cooling System is Systems shall be required in accordance with functionally tested and Specification 3.5.G, the calibrated as indicated i instrumentation which Table 4.2.9.

initiates actuation of this system shall be operable in accordance with Table 3.2.9.

+

(ttavc r-0 06AAsCAre P*w6w>

Amendment No. iG, #G, ", 111, 212 37

VYNPS

'Oj 8ff)#F4o )$TABLE 3.2.8 cn0s ItFrPOO466 DRd INSTRUMENTATION t¢*~t f~

yodor&Ae, Ortc(

e/

~. Trip Setting Degraded Bus Voltage_-_Voltage Note 1 2 per bus Degraded Bus Voltage - Time Note 2 per bus TABLE 3.2.8 T inoperable channel shall be

1. If the minimum number of operable instrument channels are not available, the Jac ktor other p emanently insta'ead cieit with one hour. = 1-.

trippedQ!ng 5f the minimum number of operable instrument channels are not a v s u t~risbefor anlv- -7succeasv as1rrs s Me aa 56 QtdA

-1es iAmendmenL WO4. '" ,-

a VYNPS TABLE 4.2.8 EMg~E:RV~AG~ INST MEeATION Tnip System Calibrai Degraded Bus Voltage See Note dW Once/Operating Cycle I1

..S Amendment No. 98 72

"U VYNPS TABLE 4.2 NOTES

. zot ied.

>2. During each refueling outage, simulated automatic actuation which opens all pilot valves shall be performed such that each trip system logic can be verified independent of its redundant counterpart.

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

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

I

8. Functional 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.

I;44T JaJC (I) c@ 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 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. 43, @4, 4G, gr4, 146, 244, 212 14-5, 74

Safety Assessment Discussion of Changes 3.2.K and 4.2.K Degraded Grid Protective System

SAFETY ASSESSMENT OF CHANGES TS 3.2.K/4.2.K - DEGRADED GRID PROTECTIVE SYSTEM 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.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 inthe Safety Assessments of Changes for CTS 3.2.H1/4.2.H, Drywell to Torus AP Instrumentation. Therefore, this change does not involve a technical change, but isonly a difference of presentation preference and is considered administrative.

A.3 CTS 3.2.K specifies an Applicability for Degraded Grid Protective System instrumentation of "During reactor power operation." This change provides an explicit Applicability, in proposed Table 3.2.8 for each Degraded Grid Protective instrumentation Trip Function.

The specified Applicability, inproposed Table 3.2.8, is consistent with the Modes and conditions specified in CTS 3.2.K, except as provided and justified in change M.1 below.

Therefore, this change does not involve a technical change, but is only a difference of presentation preference and is considered administrative. The change, providing explicit Mode or conditions of Applicability for each trip function, is consistent with the ISTS.

A.4 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 in these 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.

A.5 CTS Table 4.2 Notes 2,3 and 11 provide requirements that apply to ECCS instrumentation.

The ECCS instrumentation is located in proposed Specifications 3.2.A and 4.2.A.

Therefore, the requirements of CTS Table 4.2 Notes 2,3 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 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 in proposed Specifications 3.2.G and 4.2.G. Therefore, this change does not involve a VYNPS 1 Revision 0

SAFETY ASSESSMENT OF CHANGES TS 3.2.K/4.2.K - DEGRADED GRID PROTECTIVE SYSTEM INSTRUMENTATION ADMINISTRATIVE (continued)

A.5 technical change, but is only a difference of presentation preference and is considered (continued) administrative. CTS 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 Table 4.2 Notes 4, 12 and 13 are physically moved and changes addressed in 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 isconsidered administrative.

TECHNICAL CHANGES - MORE RESTRICTIVE M.1 CTS 3.2.K requires the Degraded Grid Protective System instrumentation to be operable only during reactor power operation. Inproposed Table 3.2.8, the Applicability of the Degraded Grid Protective instrumentation Trip Functions Isexpanded in Footnote (1)to, "When the associated diesel generator is required to be operable." The VYNPS CTS Applicability for diesel generators, which are supported by the Degraded Grid Protective instrumentation, Includes more than just 'during power operation.' For example, CTS 3.5.H.4 (which provides low pressure Emergency Core Cooling System (ECCS) requirements) requires a diesel generator to be operable during Refuel or Cold Shutdown when operations with the potendal for draining the reactor vessel are being performed and CTS 3.7.B.1.b (which provides Standby Gas Treatment System requirements) requires diesel generators to operable under certain conditions during Refuel or Cold Shutdown when secondary containment Integrity is required. This change represents an additional restriction on plant operation necessary to ensure that the ECCS and other assumed systems powered by the diesel generators remain capable of providing plant protection during the conditions when the diesel generators are required to be operable.

TECHNICAL CHANGES - LESS RESTRICTIVE

'Generic' LA.1 The CTS Table 3.2.8 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.K and Table 3.2.8 require Degraded Grid Protective instrumentation Trip Functions to be operable. In addition, the proposed Surveillance Requirements inTable 4.2.8 ensure the required instruments are properiy tested. These requirements are adequate for ensuring each of the required Degraded Grid Protective Instrumentation Trip Functions ismaintained 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 is consistent with the ISTS.

VYNPS 2 Revision 0

SAFETY ASSESSMENT OF CHANGES TS 3.2.KJ4.2.K - DEGRADED GRID PROTECTIVE SYSTEM INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE (continued)

LA.2 The details inthe CTS Table 3.2.8 Note 1,relating to the method used for placing channels intrip, are to be relocated to Specification 3.2.K Bases. The requirements of proposed Table 3.2.8 Action Notes are adequate to ensure inoperable channels are placed in trip.

As a result, the relocated details inthe CTS Table 3.2.8 Note 1 are not necessary for ensuring the appropriate actions are taken in the event of inoperable Degraded Grid Protective instrumentation channels. 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 TS Bases are controlled by the provisions of 10 CFR 50.59. Not including these details inTS is consistent with the ISTS.

LC.1 CTS Table 3.2.8 specifies requirements for the operability of the Degraded Bus Voltage alarm instrumentation by requiring that relays 62/3Z and 62/4Z be operable. These relays provide an alarm only function. The ISTS do not specify alarm-only equipment to be operable to support operability of a system or component or maintaining variables within limits. Control of the availability of, and necessary compensatory activities if not available, for indication and monitoring instruments are addressed by plant procedures and policies.

Therefore, these requirements are to be relocated to the Technical Requirements Manual.

As such, the details to be relocated are not required to be inthe Technical Specifications to provide adequate protection of the public health and safety. Changes to the Technical Requirements Manual are controlled using 10 CFR 50.59. Not including these requirements in the TS is consistent with the ISTS.

.Specific' L.1 When one or more Degraded Bus Voltage - Time Delay channels are inoperable, CTS Table 3.2.8, Note 2 limits operation to 7 days, but does not include explicit actions to restore the inoperable channels. Inproposed Table 3.2.8 Action Note 2,an explicit requirement is provided to restore the inoperable channels within 1hour prior to requiring the associated diesel generators to be declared inoperable. Since the applicable VYNPS diesel generator TS allows operation for up to 7 days with an inoperable diesel generator, the total time allowed for the plant to remain inreactor power operation with an inoperable Degraded Voltage - Time Delay channel is extended from 7 days to 7 days + 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The 1hour Ume period for this condition isprovided to attempt to evaluate and repair any discovered inoperabilities prior to declaring the associated diesel generator inoperable.

This 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months time period isconsidered to be acceptable because it minimizes risk while providing time for restoration or tripping of channels. For the proposed Table 3.2.8 Action Note 1 action to declare the associated diesel generator inoperable, this instrumentation, in conjunction with an ECCS initiation signal, provides a start signal for the diesel generators (i.e., it supports diesel generator operability). Therefore, when this instrumentation is inoperable and not restored within the required time period, the appropriate action is to declare the diesel generator inoperable. This is acceptable since the VYNPS TS requirements for diesel generators establish appropriate restrictions and compensatory measures for an inoperable diesel generator. The change is consistent with the ISTS.

VYNPS 3 Revision 0

SAFETY ASSESSMENT OF CHANGES TS 3.2.K14.2.K - DEGRADED GRID PROTECTIVE SYSTEM INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE (continued)

L.2 CTS Table 3.2.8 Note 1,inthe event of one or more inoperable Degraded Bus Voltage -

Voltage channels, requires the channel to be tripped within one hour, but does not provide direction regarding actions to take if the associated channels are not tripped. As such, a shutdown of the reactorwould be required inaccordance with 10 CFR 50.36(c)(2). Under these conditions, proposed Table 3.2.8 Action Note 1 provides actions to declare the associated diesel generator inoperable, which results in entering and taking the appropriate actions inthe associated diesel generator TS if a channel isnot tripped or restored within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . Since this instrumentation, inconjunction with an ECCS initiation signal, provides a start signal for the diesel generators (i.e., it supports diesel generator operability), the appropriate action, inthis condition, would be to declare the diesel generator inoperable.

The current requirements of CTS Table 3.2.8 Note 1are overly restrictive, inthat if the diesel were inoperable for other reasons, a 7 day restoration time is provided; yet currently if an instrument is inoperable and not tripped within one hour, but the diesel isotherwise fully operable, an immediate shutdown is required. This change is acceptable since the VYNPS TS requirements for diesel generators establish appropriate restrictions and compensatory measures for an inoperable diesel generator. The change isconsistent with the ISTS.

RELOCATED SPECIFICATIONS None VYNPS 4 Revision 0

No Significant Hazards Consideration 3.2.K and 4.2.K Degraded Grid Protective System

GENERIC NO SIGNIFICANT HAZARDS CONSIDERATION ADMINISTRATIVE CHANGES (fA.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 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 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 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 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 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 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 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 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 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 ina 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

("LA.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 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 issubject 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 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 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 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

GENERIC NO SIGNIFICANT HAZARDS CONSIDERATION "GENERICO LESS RESTRICTIVE CHANGES:

RELOCATION OF INSTRUMENTATION ONLY REQUIREMENTS (5LC.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 relocates instrumentation requirements from the Technical Specifications to the Bases, UFSAR, procedures, or other plant controlled documents. These requirements are not considered in the safety analysis. 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 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 requirements 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 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 in a margin of safety?

The proposed change will not reduce a margin of safety because it has no impact on any safety analysis assumption. Inaddition, the requirements 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 requirements 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 requirements 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 instrumentation requirements ensures no significant reduction inthe margin of safety.

VYNPS 5 Revision 0

NO SIGNIFICANT HAZARDS CONSIDERATION TS 3.2.K/4.2.K - DEGRADED GRID PROTECTIVE SYSTEM 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 relax the actions when one or more channels of Degraded Bus Voltage - Time Delay Trip Function channels are inoperable by providing one additional hour to restore the inoperable channels. The Degraded Bus Voltage - Time Delay instrumentation is not assumed to be an initiator of any analyzed event. Therefore, this change does not significantly increase the probability of a previously analyzed accident. Continued operation with inoperable Degraded Bus Voltage - Time Delay Trip Function instrumentation channels will still be limited inaccordance with Technical Specifications. Since the level of degradation allowed inthe proposed actions is the same as the current actions, the consequences of an accident occurring during the additional 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 is not allowed when a Degraded Bus Voltage - Time Delay 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 relax actions when one or more channels of the Degraded Bus Voltage - Time Delay Trip Function 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 in a margin of safety. -

WYNPS 1 Revision 0

NO SIGNIFICANT HAZARDS CONSIDERATION TS 3.2.K/4.2.K - DEGRADED GRID PROTECTIVE SYSTEM 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.

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

This change provides an additional action to declare the associated diesel generator inoperable and take the appropriate actions when one or more Degraded Bus Voltage - Voltage channels are inoperable, in lieu of a plant shutdown. The Degraded Bus Voltage - Voltage instrumentation is not assumed to be an initiator of any analyzed event. Therefore, this change does not significantly increase the probability of a previously analyzed accident. Continued operation with inoperable Degraded Bus Voltage - Voltage Trip Function instrumentation channels will still be limited in accordance with Technical Specifications. The role of this instrumentation, in conjunction with an Emergency Core Cooling System initiation signal, isto provide a start signal for the diesel generators (i.e., it supports diesel generator operability). The proposed change to the actions will not allow contnuous operation such that a single failure will preclude diesel generator Initiation from mitigating event consequences. 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 a Degraded Bus Voltage - Voltage 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 provides an additional action to declare the associated diesel generator inoperable and take the appropriate actions when one or more Degraded Bus Voltage - Voltage channels are inoperable, in lieu of a plant shutdown. 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 in a margin of safety.

VYNPS 2 Revision 0

References 3.2.K and 4.2.K Degraded Grid Protective System

3.2.K/4.2.K REFERENCES Degraded Grid Protective System

1. UFSAR, Section 8.5.3.

2. UFSAR, Section 6.5

3. UFSAR Chapter 14.

Proposed Technical Specifications 3.2.L and 4.2.L Reactor Core Isolation Cooling (RCIC)

System Actuation

VYNPS 3.2 LIMITING CONDITIONS FOR 4.2 SURVEILLANCE REQUIREMENTS OPERATION L. Reactor Core Isolation L. Reactor Core Isolation Cooling (RCIC) System Cooling (RCIC) System I

Actuation Actuation The RCIC System 1. The RCIC System instrumentation for each Trip instrumentation shall be Function in Table 3.2.9 shall checked, functionally be operable in accordance tested and calibrated as with Table 3.2.9. indicated in Table 4.2.9.

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 as follows: (a) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Trip Function 3; and (b) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Trip Functions 1 and 2 provided the associated Trip Function maintains RCIC initiation capability.

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

72 Amendment No.

VYNPS Table 3.2.9 (page 1 of 1)

Reactor Core Isolation Cooling System Instrumentation ACTIONS WHEN REQUIRED REQUIRED APPLICABLE MODES OR CHANNELS CHANNELS OTHER SPECIFIED PER TRIP ARE TRIP FUNCTION CONDITIONS SYSTEM INOPERABLE TRIP SETTING

1. Low-Low Reactor RUN, STARTUP/HOT 2 Note 1 2 82.5 inches Vessel Water STANDBY4'), HOT Level SHUTDOWNS, REFUELz1 }

2. Low Condensate RUN, STARTUP/HOT 2 Note 2 2 3.81tt2)

Storage Tank STANDBY(2), HOT Water Level SHUTDOWN(l), REFUELU)

3. High Reactor RUN, STARTUP/HOT 2 Note 3 5 177.0 inches Vessel Water STANDBYt 1 }, HOT Level SHUTDOWNW), REFUELU)

(1) With reactor steam pressure > 150 psig.

(2) Percent of instrument span.

Amendment No.. 73

VYNPS Table 3.2.9 ACTION Notes

1. With one or more RCIC System instrumentation Trip Function 1 channels inoperable:

a. Declare the RCIC System inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of RCIC initiation capability; and

b. Place inoperable channel in trip within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

If any applicable Action and associated completion time of Note l.a or l.b is not met, immediately declare the RCIC System inoperable.

2. With one or more RCIC System instrumentation Trip Function 2 channels inoperable:

a. Declare the RCIC System inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of RCIC initiation capability when RCIC System suction is aligned to the Condensate Storage Tank; and

b. Place inoperable channel in trip or align RCIC System suction to the suppression pool within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

If any applicable Action and associated completion time of Note 2.a or 2.b is not met, immediately declare the RCIC System inoperable.

3. With one or more RCIC System instrumentation Trip Function 3 channels inoperable:

a. Restore inoperable channel to operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

If the Action and associated completion time of Note 3.a is not met, immediately declare the RCIC System inoperable.

Amendment No. 74

VYNPS Table 4.2.9 (page 1 of 1)

Reactor Core Isolation Cooling System Instrumentation Tests and Frequencies TRIP FUNCTION CHECK FUNCTIONAL TEST CALIBRATION

1. Low-Low Reactor Vessel Once/Day Every 3 Months Every 3 Months(l),

Water Level Once/Operating Cycle

2. Low Condensate Storage NA Every 3 Months Every 3 Months i),

Tank Water Level Once/Operating Cycle

3. High Reactor Vessel NA Every 3 Months Every 3 Monthszl, Water Level Once/Operating Cycle (1) Trip unit calibration only.

Amendment No. 74a

Proposed Bases 3.2.L and 4.2.L Reactor Core Isolation Cooling (RCIC)

System Actuation

VYNPS BASES: 3.2.L/4.2.L REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM INSTRUMENTATION BACKGROUND The purpose of the RCIC System instrumentation is to initiate actions to ensure adequate core cooling when the reactor vessel is isolated from its primary heat sink (the main condenser) and normal coolant makeup flow from the Reactor Feedwater System is insufficient or unavailable, such that RCIC System initiation occurs and maintains sufficient reactor water level such that initiation of the low pressure Emergency Core Cooling Systems (ECCS) pumps does not occur. A more complete discussion of the RCIC System is provided in UFSAR, Section 4.7 (Ref. 1).

RCIC System automatic initiation occurs for conditions of Low - Low Reactor Vessel Water Level. The variable is monitored by four transmitters that are connected to four trip units. The Low - Low Reactor Vessel Water Level Trip Function is a single trip system with two trip system logics. The outputs of the trip units are connected to relays whose contacts are arranged in a one-out-of-two taken twice logic arrangement.

The RCIC test line isolation valve is closed on a RCIC initiation signal to allow full system flow.

The RCIC System also monitors the water level in the condensate storage tank (CST) since this is the initial source of water for RCIC operation. Reactor grade water in the CST is the normal source. Upon receipt of a RCIC initiation signal, the CST suction valve is automatically signaled to open.

If the water level in the CST falls below a preselected level, the RCIC suppression pool suction valves automatically open. When the suppression pool suction valves are both fully open, the RCIC CST suction valve automatically closes. Two level transmitters are used to detect low water level in the CST. Either transmitter can cause the suppression pool suction valves to open and the CST suction valve to close (one trip system arranged in a one-out-of-two logic).

The RCIC System provides makeup water to the reactor until the reactor vessel water level reaches the high water level trip (one trip system arranged in a two-out-of-two logic), at which time the RCIC steam admission valve closes.

The RCIC System automatically restarts if a Low - Low Reactor Vessel Water Level signal is subsequently received.

APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY The function of the RCIC System to provide makeup coolant to the reactor is used to respond to transient events. The RCIC System is not an Engineered Safety Feature System and no credit is taken in the safety analyses for RCIC System operation. Based on its contribution to the reduction of overall plant risk, however, the system, and therefore its instrumentation, meets Criterion 4 of 10 CFR 50.36(c)(2)(ii).

The operability of the RCIC System Instrumentation is dependent on the operability of the individual instrumentation channel Trip Functions. 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 the actual trip setpoints within the calculational as-found tolerances provides reasonable Amendment No. 80j

VYNPS BASES: 3.2.L/4.2.L REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM INSTRUMENTATION APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) assurance that, under worst case design basis conditions, the associated trip will occur within the Trip Settings specified in Table 3.2.9. 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 individual Trip Functions are required to be operable in the RUN Mode and in the STARTUP/HOT STANDBY, HOT SHUTDOWN, and REFUEL Modes with reactor steam pressure > 150 psig since this is when RCIC is required to be operable.

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

1. Low - Low Reactor Vessel Water Level Low reactor pressure vessel (RPV) water level indicates that normal feedwater flow is insufficient to maintain reactor vessel water level and that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, the RCIC System is initiated on a Low - Low Reactor Vessel Water Level signal to assist in maintaining water level above the top of the enriched fuel.

Low - Low 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.

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). The Trip Setting is referenced from the top of enriched fuel.

Four channels of Low - Low Reactor Vessel Water Level Trip Function are available and are required to be operable when RCIC is required to be operable to ensure that no single instrument failure can preclude RCIC initiation.

2. Low Condensate Storage Tank Water Level Low water level in the CST indicates the unavailability of an adequate supply of makeup water from this normal source. Normally, the suction valve between the RCIC pump and the CST is open and, upon receiving a RCIC initiation signal, water for RCIC injection would be taken from the CST. However, if the water level in the CST falls below a preselected level, the RCIC suppression pool suction valves automatically open. When the suppression pool suction valves are both fully open, the RCIC CST suction valve automatically closes. This ensures that an adequate supply of makeup water is available to the RCIC pump.

Amendment No. 80k

VYNPS BASES: 3.2.L/4.2.L REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM INSTRUMENTATION APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

Two level transmitters are used to detect low water level in the CST. The Low Condensate Storage Tank Water Level Trip Function Trip Setting is set high enough to ensure adequate pump suction head while water is being taken from the CST. The trip setting is presented in terms of percent instrument span.

Two channels of Low Condensate Storage Tank Water Level Trip Function are available and are required to be operable when RCIC is required to be operable to ensure that no single instrument failure can preclude RCIC swap to the suppression pool source.

3. High Reactor Vessel Water Level High RPV water level indicates that sufficient cooling water inventory exists in the reactor vessel such that there is no danger to the fuel. Therefore, the high water level signal is used to close the RCIC steam admission valve to prevent overflow into the main steam lines (MSLs).

High Reactor Vessel Water Level signals for RCIC are initiated from two level transmitters, which 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.

The High Reactor Vessel Water Level Trip Setting is high enough to preclude closing the RCIC steam admission valve during normal operation, yet low enough to trip the RCIC System to prevent reactor vessel overfill. The Trip Setting is referenced from the top of enriched fuel.

Two channels of High Reactor Vessel Water Level Trip Function are available and are required to be operable when RCIC is required to be operable.

ACTIONS Table 3.2.9 ACTION Note 1 Table 3.2.9 ACTION Note l.a is intended to ensure that appropriate actions are in taken if multiple, inoperable, untripped channels of Trip Function 1 result initiation capability for the RCIC System. In a complete loss of automatic this case, automatic initiation capability is lost if two Trip Function 1 In this channels in the same trip system logic are inoperable and untripped.

initiation capability), the 24 hour allowance of situation (loss of automatic is not appropriate, and the RCIC System must be Table 3.2.9 ACTION Note l.b declared inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after discovery of loss of RCIC initiation capability. The completion time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This completion time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock." For Table 3.2.9 ACTION Note l.a, the completion time only begins upon discovery that the RCIC System cannot be automatically Amendment No. 801

VYNPS BASES: 3.2.L/4.2.L REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM INSTRUMENTATION ACTIONS (continued) initiated due to two inoperable, untripped Low - Low Reactor Vessel Water Level channels in the same trip system logic. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months completion time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.

Because of the redundancy of sensors available to provide initiation signals and the fact that the RCIC System is not assumed in any accident or transient analysis, an allowable out of service time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months has been shown to be acceptable (Ref. 2) to permit restoration of any inoperable channel to operable status. 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 3.2.9 ACTION Note l.b. 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 any required Action and associated completion time of Table 3.2.9 ACTION Note l.a or l.b not met, the RCIC System may be incapable of performing the intended function, and the RCIC System must be declared inoperable immediately.

Table 3.2.9 ACTION Note 2 Table 3.2.9 ACTION 2.a is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels of Trip Function 2 result in automatic RCIC initiation (i.e., suction swap) capability being lost. In this case, automatic RCIC suction swap capability is lost if two Trip Function 2 channels are inoperable and untripped. In this situation (loss of automatic suction swap), the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowance of Table 3.2.9 ACTION Note 2.b is not appropriate, and the RCIC System must be declared inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from discovery of loss of RCIC initiation capability when the RCIC System suction is aligned to the CST. Table 3.2.9 ACTION Note l.a is only applicable if the RCIC System suction is not aligned to the suppression pool since, if aligned, the Trip Function is already performed. The completion time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This completion time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock." For Table 3.2.9 ACTION Note 2.a, the completion time only begins upon discovery that the RCIC System cannot be automatically aligned to the suppression pool due to two inoperable, untripped channels in Trip Function

2. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.

Because of the redundancy of sensors available to provide initiation signals and the fact that the RCIC System is not assumed in any accident or transient analysis, an allowable out of service time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months has been shown to be acceptable (Ref. 2) to permit restoration of any inoperable channel to OPERABLE status. If the inoperable channel cannot be restored to operable Amendment No. 80m

VYNPS BASES: 3.2.L/4.2.L REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM INSTRUMENTATION ACTIONS (continued) status within the allowable out of service time, the channel must be placed in the tripped condition per Table 3.2.9 ACTION Note 2.b, which performs the intended function of the channel (shifting the suction source to the suppression pool). Alternatively, Table 3.2.9 ACTION Note 2.b allows the manual alignment of the RCIC System suction to the suppression pool, which also performs the intended function. If either action of Table 3.2.9 ACTION Note 2.b is performed, measures should be taken to ensure that the RCIC System piping remains filled with water.

With any required Action and associated completion time of Table 3.2.9 ACTION Note 2.a or 2.b not met, the RCIC System may be incapable of performing the intended function, and the RCIC System must be declared inoperable immediately.

Table 3.2.9 ACTION Note 3 A risk based analysis was performed and determined that an allowable out of service time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (Ref. 2) is acceptable to permit restoration of any inoperable Trip Function 3 channel to operable status (Table 3.2.9 ACTION Note 3.a). A required Action (similar to Table 3.2.9 ACTION Note l.a) limiting the allowable out of service time, if a loss of automatic RCIC initiation capability (i.e., loss of high water level trip capability) exists, is not required. Table 3.2.9 ACTION Note 3 applies to the High Reactor Vessel Water Level Trip Function whose logic is arranged such that any inoperable channel will result in a loss of automatic RCIC initiation capability. As stated above, this loss of automatic RCIC initiation capability was analyzed and determined to be acceptable. One inoperable channel may result in a loss of high water level trip capability but will not prevent RCIC System automatic start capability. However, the Required Action does not allow placing a channel in trip since this action would not necessarily result in a safe state for the channel in all events (a failure of the remaining channel could prevent a RCIC System start).

With any required Action and associated completion time of Table 3.2.9 ACTION Note 3.a not met, the RCIC System may be incapable of performing the intended function, and the RCIC System must be declared inoperable immediately.

SURVEILLANCE REQUIREMENTS Surveillance Requirement 4.2.L.1 As indicated in Surveillance Requirement 4.2.L.1, RCIC System instrumentation shall be checked, functionally tested and calibrated as indicated in Table 4.2.9. Table 4.2.9 identifies, for each Trip Function, the applicable Surveillance Requirements.

Surveillance Requirement 4.2.L.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 Amendment No. 80n

BASES: 3.2.L/4.2.L REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS (continued) as follows: (a) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Trip Function 3; and (b) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Trip Functions 1 and 2, provided the associated Trip Function maintains RCIC initiation 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. 2) 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 RCIC System will initiate when necessary.

Surveillance Requirement 4.2.L.2 The Logic System Functional Test demonstrates the operability of the required initiation logic for a specific channel and includes simulated automatic actuation of the channel. The system functional testing performed in Surveillance Requirement 4.5.G.1 overlaps this Surveillance to provide complete testing of the 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 4.2.9, Check Performance of an Instrument Check once per day, for Trip Function 1, 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.

Amendment No. 80o

VYNPS BASES: 3.2.L/4.2.L REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS (continued)

Table 4.2.9, Functional Test For Trip Functions 1, 2 and 3, 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. For Trip Functions 1, 2 and 3, the Frequency of "Every 3 Months"f is based on the reliability analysis of Reference 2.

Table 4.2.9, Calibration For Trip Functions 1, 2, and 3, 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 3, 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 2 and the time interval assumption for trip unit calibration used in the associated setpoint calculation.

REFERENCES

1. UFSAR, Section 4.7.

2. GENE-770-06-2P-A, Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications, December 1992.

Amendment No. 80p

Current Technical Specifications Markups 3.2.L and 4.2.L Reactor Core Isolation Cooling (RCIC)

System Actuation

VYNPS 3.2 LIMITING CONDITIONS FOR 4.2 SURVEILLANCE REQUIREMENTS OPERATION sooner made operable. If both instruments are made or found to be inoperable, and indication cannot be restored within a six hour period, an orderly shutdown shall be initiated and the reactor shall be in a hot shutdown condition in six hours and a cold shutdown condition in the following eighteen hours.

I AE/,tiWie rz 16 1 WMT6 P4CV>

I. Recirculation Pump Trip I. Recirculation Pump Trip Instrumentation Instrumentation During reactor power The Recirculation Pump Trip operation, the Recirculation Instrumentation shall be Pump Trip Instrumentation functionally tested and shall be operable calibrated in accordance with in accordance with Table 4.2.1.

Table 3.2.1.

J. Deleted J. Deleted K. Degraded Grid Protective K. Degraded Grid Protective System Runt am L. During reactor power The emergency bus

\operation, the emergency bus undervoltage instrumentation undervoltage instrumentation shall be functionally tested Y' shall be operable in and calibrated in accordance with Table 4.2.8.

I

-___r-' L. Reactor Core Isolation L. ReactorlCore Isolation System Actuation rhnhe Reactor Core acordanceCooling Isolation with TableSystem3.2.9.

s required in accordance with functionally tested and Spc fia io 35. calibrated as indicated in

,_wntrumentatioL7se'._ Table 4.2.9. IAND

- t~ ~ ~ ~ b operable~~~s<io ohin)

/c th Table 3.2.9. W 6A1j

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mfdpemacew sr-otns Safecksx° eX~Cd 'Pae'eQFA10t^le' TIC R-vu'4' 0"'W'"

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Acb ($) e.A up 7z Ija$ F.<R 7TAIP PaIVQAIVs l Jud st F~eAat66b vuff AIVOCs4 T 7AIP FoacT6J sAArWASIRCC rcDXruT r Amendment No. iG, 96, 19, 411, 212 IA.4' 37

VYNPS TABLE 3.2.9 REACTOR CORE ISOLATION COOLING SYSTEM INSTRUMENTATION Trip Function Trip Setting iI- 2 I Low-Low Reactor Vessel Water Level T-, -A-D S Tart i

>82.5" Note z 1 2 Low Condensate Storage Tan Water Level Note 2 High Reactor Vessel Water 17 Note 31 Level n 7

. C-Z- TripHystem Log z' --

,Pj T,eCF 7- 6,r XW% U46A7 =77RAz Amendment No. 411, 444, 186 57

VYNPS TABLE 3.2.9(NOTES Oetrip system w tvnitiating instr ntation arranged in o out-of-two taetwice logic.

2. fn rip system ith initiating iStrumentation arrne in a/

/ ne-out-of-~two oic./

A. One trip syse arranqed in a two-out-of-two loc.

If e minim umber of ope e channels age not avail e, the s Koperable arc the require s of Speci cation 3. a 2 Shen a channel is placed in an inoperable status solely for performance 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 RCIC initiation capability.

d When a channel is placed in an inoperable status solely for performance of required surveillances, entry into associated Limiting Conditions For eration 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 .

3.2.1. With one or more channels inoperable for RCIC:

AcnoiiJ M048 d o Within one hour from discovery of loss of system initiation capability, declare the RCIC system inoperable, and 6 Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , place channel in trip.

If required actions and associated completion times of actions A or B are not met, immediately declare the RCIC system inoperable.

T51c6 With one or more channels inoperable for RCIC:

7a Within one hour from discovery of loss of system initiation 2F.. capability while suction is aligned to the CST, declare the RCIC system inoperable, and b $ Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , place channel in trip or align suction for the RCIC system to the suppression pool.

If-required actions and associated completion times of actions A or B are not met, immediately declare the RCIC system inoperable.

F4Jtc 09 With one or more channels inoperable for RCIC:

3".v

'Vek . Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , restore channel to operable status.

w If required action and associated completion time of action A is not met, immediately declare the RCIC system inoperable.

Amendment No. 144, 186 58

VYNPS IfTAA TABLE 4.2.9

) TES61ND Q FREQUENCIES REACTOR CORE ISOLATION COOLING SYSTEM INSTRUMENTATION Trip Function Functional Testo %-fE D Calibration w E S <q==iCheck I. Low-Low Reactor Vess el Every Three Months Cycle once each day Water Level

2. Low Condensate Stora ge Every Three Months nce/Operating Cycle Tank Water Level High Reactor Vessel Every Three Months e/Operating Cycle 54 Water Level

__E Trip System Logic Once/Operating Cycle X c ~Cy Ope~rang

-dAte 3)/ 7 M Ce 1;'tt & o I J,M Amendment No. 111, 186 73

S VYNPS TABLE 4.2 NOTES Duingeac rfueling outage, simulated automatic actuation whch

/ f ens r~

allpilt vlvs shall be performed such that each trip system'loi can ent of its redundant counterpart.

-s. Trig~sy~tem loqie calibrationohall incluWE only timejelay relay rnece for prop unctionin of the trip 6ystem

-- This instrumentation is excepted from functional test definition. The -

functional test will consist of injecting a simulated electrical signal into the measurement channel.

5. / Deleted.

De eted

{B Fnzeinsltest an calbraionsarenot required when systems are not0F-

{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 requited 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.

l13. Includes calibration of the REM Reference Downscale function (i.e., RBM upscale function is not bypassed when >30t Rated Thermal Power).

___._ toH 06, No. 44, *20C F MeA sen I 74 Amendment *UV, W21r, =70-rp W01W, -

Safety Assessment Discussion of Changes 3.2.L and 4.2.L Reactor Core Isolation Cooling (RCIC)

System Actuation

SAFETY ASSESSMENT OF CHANGES TS 3.2.J4.2.L - REACTOR CORE ISOLATION COOLING SYSTEM 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.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 inthe 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 is only a difference of presentation preference and is considered administrative.

A.3 CTS 3.2.L specifies an Applicability for Reactor Core Isolation Cooling (RCIC) System instrumentation of 'When the Reactor Core Isolation Cooling System is required in accordance with Specification 3.5.G.' Specification 3.5.G includes the requirements for the RCIC System. This change provides an explicit Applicability, in proposed Table 3.2.9 for each RCIC System instrumentation Trip Function. The specified Applicabilities, in proposed Table 3.2.9, are consistent with the Modes and conditions when the RCIC System are required to be operable by Specification 3.5.G. Therefore, this change does not Involve a technical change, but is only a difference of presentation preference and is considered administrative. The change, providing explicit Mode or conditions of Applicability for each trip function, isconsistent with the ISTS.

A.4 CTS 4.2.L specifies that instrumentation and logic systems shall be functionally tested and calibrated as indicated inTable 4.2.9. Inproposed Surveillance Requirement (SR) 4.2.L.1, the reference to "and logic system," is deleted since associated logic systems are considered part of the RCIC System instrumentation Trip Functions inboth proposed and CTS Tables 3.2.9 and 4.2.9. It is not necessary to explicitly identify logic systems in CTS 4.2.L, since proposed SR 4.2.L.2 (CTS Table 4.2.9 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 RCIC System instrumentation Trip Function). Therefore, this change is considered administrative.

A.5 CTS 4.2.L includes reference to CTS Table 4.2.9 for functional test and calibration requirements for RCIC System instrumentation. CTS 4.2.1 is revised, in proposed SR 4.2.L.1, to also include reference to check requirements consistent with CTS Table 4.2.9.

This change is a presentation preference and does not alter the current requirements to periodically perform checks of certain RCIC System instrument Trip Functions. Therefore, this change is considered administrative in nature.

VYNPS 1 Revision 0

SAFETY ASSESSMENT OF CHANGES TS 3.2.L/4.2.L - REACTOR CORE ISOLATION COOLING SYSTEM INSTRUMENTATION ADMINISTRATIVE (continued)

A.6 CTS Table 3.2.9, Notes 5 and 6,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.L.1 and the allowances of these two notes are combined. This change does not involve a technical change, but isonly a difference of presentation preference. Therefore, this change isconsidered administrative.

A.7 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.8 For the Trip System Logic associated with the RCIC System instrumentation, CTS Table 4.2.9 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 RCIC System instrumentation Trip Functions of CTS Table 4.2.9 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 4.2.9, the RCIC System instrumentation (proposed Trip Functions 1,2, and 3)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 is considered administrative.

A.9 CTS Table 4.2 Notes 2, 10, and 11 provide requirements that apply to ECCS instrumentation. The ECCS instrumentation islocated inproposed 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 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 is only a difference of presentation preference and is considered administrative. CTS Table 4.2 Notes 4, 12, and 13 provide 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 Notes 4, 12, and 13 are physically moved and changes addressed in 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 isconsidered administrative.

VYNPS 2 Revision 0

SAFETY ASSESSMENT OF CHANGES TS 3.2.LJ4.2.L - REACTOR CORE ISOLATION COOLING SYSTEM INSTRUMENTATION TECHNICAL CHANGES - MORE RESTRICTIVE MA CTS Table 4.2.9 does not include explicit requirements to calibrate trip units. Proposed Table 4.2.9 requires calibration of the trip units of the following Trip Functions every 3 months: Low-Low Reactor Vessel Water Level (proposed Table 4.2.9 Trip Function 1), Low Condensate Storage Tank Water Level (proposed Table 4.2.9 Trip Function 2), and High Reactor Vessel Water Level (proposed Table 4.2.9 Trip Function 3). The trip units of these Trip Functions are currently required by CTS Table 4.2.9 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.

M.2 CTS Table 3.2.9 specifies for the Low Condensate Storage Tank Water Level Trip Function that the Trip Setting be > 3%. The function of the Low Condensate Storage Tank Water Level is to provide an automatic transfer of the RCIC suction source from the condensate storage tank to the suppression pool when the level inthe condensate storage tank is no longer sufficient to support adequate RCIC pump suction head. The CTS Trip Setting has been determined to be insufficient to ensure that transfer of the RCIC System suction from the condensate storage tank to the suppression pool occurs prior to potential vortex formation at the RCIC suction inlet inthe condensate storage tank. Therefore, in proposed Table 3.2.9, the Trip Setting for the Low Condensate Storage Tank Water Level Trip Function (Trip Function 2)has been increased to > 3.81 %to account for the additional water level needed to preclude the potential for vortex formation. This minimum level corresponds to the Process Limit used inthe associated setpoint calculation. To account for instrument uncertainties, the instrument setpoint and as-found tolerance (i.e., instrument operability limit) were developed using the Vermont Yankee Instrument Uncertainty and Setpoints Design Guide. Footnote (2)in proposed Table 3.2.9 clarifies that the trip setting is specified interms of percent instrument span. The Instrument setpoint and as-found tolerance are located inplant procedures. This change represents an additional restriction on plant operation necessary to ensure that RCIC System operability is maintained when aligned to the condensate storage tank and that RCIC pump suction transfer to the suppression pool occurs prior to the vortex formation.

TECHNICAL CHANGES - LESS RESTRICTIVE Generics LA.1 The CTS Table 3.2.9 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.L and Table 3.2.9 require the RCIC System Instrumentation Trip Functions to be operable. Inaddition, the proposed Surveillance Requirements inTable 4.2.9 ensure the required instruments are properly tested. These requirements are adequate for ensuring each of the required RCIC System 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 VYNPS 3 Revision 0

SAFETY ASSESSMENT OF CHANGES TS 3.2.J4.2.L - REACTOR CORE ISOLATION COOLING SYSTEM INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE (continued)

LA.1 health and safety. Changes to the TRM are controlled by the provisions of 10 CFR 50.59.

(continued) Not including these details inTS isconsistent with the ISTS.

LA.2 CTS Table 3.2.9 Notes 1,2,and 3 contain design details of the RCIC System instrumentation (i.e., one trip system with initiating instrumentation arranged in a one-out-of-two taken twice logic, one trip system with initiating instrumentation arranged ina one-out-of-two logic, and one trip system arranged ina two-out-of-two logic). These details are not necessary to ensure the operability of RCIC System instrumentation. Therefore, the Information inthese notes Isto be relocated to Specification 3.2.L Bases and reference to this information isdeleted from VYNPS TS. The requirements of Specification 3.21 and the associated Surveillance Requirements for the RCIC System instruments 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 inTS is consistent with the ISTS.

LA.3 The Trip Settings associated with reactor vessel water level trip functions (proposed Table 3.2.9 Trip Functions 1 and 3)are currently referenced to Above Top of Enriched Fuel.,

This detail is to be relocated to the Bases. This reference is not necessary to be included in the WNPS TS to ensure the operability of the associated RCIC System instrumentation.

Operability requirements are adequately addressed Inproposed Specification 3.2.L, Table 3.2.9 and the specified Trip Settings. As such, this relocated reference is 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 10 CFR 50.59. Not including these details inTS is consistent with the ISTS.

'Specific' L.1 CTS Table 3.2.9 includes requirements for Trip System Logics associated with the RCIC System Instrumentation Trip Functions. These Trip Systems Logics are considered part of the RCIC System 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 3.2.9 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 3.2.9 Note

4) will now be governed by the actions for the individual proposed Table 3.2.9 RCIC System instrumentation Trip Functions. These proposed Table 3.2.9 Action Notes are less restrictive than the CTS Table 3.2.9 Note 4 actions. However, the proposed actions will ensure, inthe event of inoperabilities, that consistent actions are applied to both RCIC System instrumentation Trip Functions and their associated Trip System Logics for the same level of degradation. This change Isacceptable, since the allowed outage times of the proposed Table 3.2.9 Action Notes will limit operation to within the bounds of the applicable analysis, i.e., GENE-770-06-2-A, "Addendum to Bases for Changes to VYNPS 4 Revision 0

SAFETY ASSESSMENT OF CHANGES TS 3.2.J4.2.L - REACTOR CORE ISOLATION COOLING SYSTEM INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE (continued)

L.1 Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation (continued) Technical Specifications," December 1992. Application of these analyses to the VYNPS RCIC System instrumentation Trip Functions, including the associated Trip System Logics, was approved by the NRC inWNPS License Amendment No. 186 dated April 3,2000.

This change Isconsistent with the ISTS.

RELOCATED SPECIFICATIONS None VYNPS 5 Revision 0

No Significant Hazards Consideration 3.2.L and 4.2.L Reactor Core Isolation Cooling (RCIC)

System Actuation

GENERIC NO SIGNIFICANT HAZARDS CONSIDERATION ADMINISTRATIVE CHANGES VA.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 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 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 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

("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 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 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 in the 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 ina 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/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 in the 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 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 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 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.L/4.2.L - REACTOR CORE ISOLATION COOLING SYSTEM 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 in the probability or consequences of an accident previously evaluated?

This change will relax the actions when one or more channels of Reactor Core Isolation Cooling (RCIC) System Trip Function instrumentation are inoperable due to inoperable Trip System Logic.

The RCIC System Trip Funcfion 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 RCIC System Trip Function instrumentation channels will still 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 is not allowed when an associated RCIC System 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 relax actions when one or more channels of the affected RCIC System 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 ina margin of safety.

VYNPS 1 Revision 0

References 3.2.L and 4.2.L Reactor Core Isolation Cooling (RCIC)

System Actuation

3.2.U4.2.L REFERENCES Reactor Core Isolation Cooling (RCIC) System Actuation

1. UFSAR, Section 4.7.

2. GENE-770-06-2P-A, Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications, December 1992.

Proposed Technical Specifications Bases 3.1/4.1 and 3.2/4.2 Reactor Protection System Bases and Protective Instrument System Bases

VYNPS BASES:

2.1 FUEL CLADDING INTEGRITY A. Trip Settings The bases for individual trip settings of Section 2.1 are discussed in the Bases for Specifications 3.1.A, 3.2.A and 3.2.B.

Amendment No. 44, e-&, 34, 47, 46, 94, Ads, 446 14

VYNPS (This page intentionally blank.)

Amendment No. b4, 74, 84, 94 15

VYNPS (This page intentionally blank.)

Amendment No. 4., As, 44, 6g, 94, by71 16

VYNPS (This page intentionally blank.)

Amendment No. A4, QS', 84, 464, 474, A&78,BW9 0051 17

Current Technical Specifications Bases Markups 3.1/4.1 and 3.2/4.2 Reactor Protection System Bases and Protective Instrument System Bases

EVYNPS BASES:

2.1 FUEL CLADDING INTEGRITY Xp C A. Trip Settings The bases for individual trip settings are discussed in

1. Neutron Flux Trip Settinr

a. APRM Flux Scram /rip Setting (Run Mode)/ \

The average twer range monitoring (APRM) tasfer frco th f (caltri terare m g heat balance data taken durings t ady ntron f condition reads in percent of rated thermal w oter (159w Mbt)e l BeAause a chambers provide the basic i pssion APtsignals, theaPRv du t the system timecosatfthful esponds directly to average neutro herefre 1 During abnoma tux. duin f transe nts, the instantaneous rate of hean ransfer from sthe fuel ha htidctdb h nurn l a h ca ettig (relt or thermal power) is less thanthinstantaneous neutron flu d to the time constant of the fuel. hrefore, during abnormal\

ational transients, the thermal r of the fuel will be less an that indicated by the w cram setting.

seto alyses are performed to demonst e that the APRM flux scram over the range of settings from a ma m of 120% to the minimum flow biased setpoint of 54% provide rotection from the fuel safety e umit for all abnormal operadonal transients including those that may result in a thermal hy ulic instability.

An increase in tri setn wasrip setting would decrease thes margin present befor te fuel cladding integrity Safety Limitet reached. The APRM sc trip setting was determined by an ath Lysis of margins required provide a reasonable range for mane ering during operation.A educing this operating margin would 2.lrease the frequency ofa rious scrams which have an adverse ect on reactor safety causine ri resulting thermal stri tes. Thus, the APRM scttrip setting was selected because ith t nvides adequate marn for the fuel cladding integr ity S Limit yet allowsthegopet core margin res th thepossibity of unnecessary scrams.

APRM FXxScram Trip Setting (Run Mode) /

Thedcam trip setting must be adjusted nsure that the LHGR tro sient peak is not increased for an cmbination of MFLPD and nactor core thermal power. If the am requires a change due to nabnormal peaking condition, it wX1be accomplished by increasing the APRM gain by the ri o in Specification 2.1.A.l.a, thus assuring a reactor scram a tower than design overpower conditions. For single recir tion loop operation, the APRM flux scram trip setting is reducepn accordance with the analysis presented in NEDO-30060, F suary 1983. This adjustment accounts for the difference betwee the single loop and two loop drive flow at the same core flow, Bdensures that the margin of safety is not reduced during single oop operation.

Analyses of the lim ing transients show that no scram adjustment is required to asX e fuel cladding integrity when the transient is iitiated from teoperating limit MCPR defined in the Core/

Oeating LimiU Report.

Amendment No. 4&, ", 44, 4., 44., 44, 444, 146 14

/1.1 VYNPS ei/lo PAG9 oMrFAJ7?W>a4Y b,))

BAES: 2.1 (Cont Id) /

Flux Scram Trip S/ting (Refuel or Startup and Hot Standby Mg For operation the startup mode while the reactor is a o pressure, th / educed APRM scram setting to 15% of rate Mpwr provides ad hate thermal margin between the stp int n h safety li t, 25% of the rated. (During an outage w n it is necessar to check refuel interlocks, the mode swi h must be moved to the tartup position. Since the APRM reduced /cram may be inopeable at that time due to the disconnectio of the LPRMs, it is r quired that the IRM scram and the SRM sc Am in noncoincidence be n effect. This will ensure that adequat thermal margin is m ntained between the setpoint and the sa ty limit.) The margin s adequate to accommodate anticipated ma euvers associated with station startup. Effects of increasing ressure at zero or low void content are minor, cold water fro sources available during startup is not much colder than that lready in the system, temperature coefficients are small and control rod patterns are constrained to be uniform by oper ing procedures backed up by the rod worth minimizer. Worth of dividual rods is very low in a uniform rod pattern. Thus, of all possible sources of reactivity input, uniform control rod w hdrawal is the most probable cause of significant power rise. Be ause the flux distribution associated with uniform rod withdraw s does not involve high local peaks, a because several rods mus be moved to change power by a significt percentage of rated po r, the rate of power rise is very slow Generally, the heat flx is in near equilibrium with the fiss on rate. In an assume funiform rod withdrawal approach to the cram level, the rate of ower rise is no more than 5% of rated ower per minute, and the RM system would be more than adequate o assure a scram before th power could exceed the safety limit. he reduced APRM scram re ins active until the mode switch is pl ced in the RUN position This switch can occur when reactor p ssure is greater tha 800 psig.

The IRM stem consists of 6 chambers, 3 in ea of the reactor protecton system logic channels. The IRM is a 5-decade instr ent, which covers the range of power evel between that cove d by the SRM and the APRM. The 5 de ades are covered by the IRM y means of a range switch and the 5decades are broken down ino 10 ranges, each being one-half of decade in size. The IRM ram trip setting of 120/125 of full cale is active in each range of the IRM. For example, if the ins ument were on range 1, the scram setting would be a 120/125 o full scale for that range; likewise, if the instrument were range 5, the scram would be 120/125 of full scale on that ra ge. Thus, as the IRM is ranged up to accommodate the increase in -ower level, the scram trip setting is also ranged up. The most gnificant sources of reactivity change during the power incr ase are due to control rod withdrawal.

For in-sequence control ro withdrawal, the rate of change of power is slow enough due to the hysical limitation of withdrawing control rods, that heat ux is in equilibrium with the neutron flux and an IRM scram uld result in a reactor shutdown well before any safety lim is exceeded.

Amendment No. 48, L&, 44, 94 15

iiJ VYNPS p4W. /ATAtr44lY BASES: 2. 1 (Cont Id) Or In order / nure that the! IRM provided adequate p Xtec on against hesingle rod withdrawal error, a range rod withdrpal accidents was analyzed. This analsiX included star ng the accident at various power levels. The most sev e case involves an initial condition in ich the reactor i just subcritical and the IRM system is n yet on scale.

is condition exists at quarter rod dens y. Additional conservatism was taken in this analysis y assuming that the IRM channel closest to the withdrawn r d is bypassed. The results of this analysis show that t reactor is scrammed and peak power limited to one percent rated power, thus maintaining MCPR above the fuel c adding integrity safety limit. Based on the above anal is, the IRM provides protection against local contr rod withdrawal errors and continuous withdrawal of con ol rods in sequence.

. R oeleted I C. Reactor Log~ter Level Scram/

The reacor low water level scram is set at a point hich will prevent reacto operation with the steam separators uncov ad, thus limiting carry under to the recirculation loops. In addi ion, the safety limit is sed on a water level below the scram poin and therefore this se ing is provided.

D. eactor Low Water Level ECCS Initiation Tr Point The core standby cooling subsystems are esigned to provide sufficient cooling to the core to dissipate the ergy associated with the loss-of-coolant accident and to limi fuel clad temperature to well below the clad melting temperature and to limit clad metal-water reaction to less than 1%, to assu e that core geometry remains intact.

The design of the ECCS compone s to meet the above criteria was dependent on three previously set parameters: the maximum break size, the low water level scram s point, and the ECCS initiation setpoint.

To lower the ECCS initiati n setpoint would now prevent the ECCS components from meeting air design criteria. To raise the ECCS initiation setpoint would be in a safe direction, but it would reduce the margin establishe to prevent actuation of the ECCS during normal operation or during ormally expected transients.

Amendment No. 44, Q4, 4*, i-i, 94, 4&4, 211 16

E ~~VYNPS t6<

oASES:2. 1 (ContId)

E.Turbine Stop Val eWlure Scram Trip Setting/

The turbine sto valve closure scram trip anticipates he pressure, neutron flux and heat ux increase that could result from pid closure of the turbine stop alves. With a scram trip setting of <10% of valve closure from full o en, the resultant increase in surfac heat flux is limited such that MCPR emains above the fuel cladding inte ity safety limit even during the wors case transient that assumes the tur ne bypass is closed. This scram s gnal may be bypassed at <30% of reac or Rated Thermal Power.

F. Turbne Control Valve Fast Closure Scram T J control valve fast closure scram i provided to limit the rapid incr se n pressure and neutron flux resultin from fast closure of the turbin control valves due to a load reject' n coincident with failure of the ypass system. This transient is less serere than the turbine stop valve osure with failure of the bypass valve and therefore adequate margin exsts. This scram signal may be bypassed at 30% of reactor Rated Thermal Po r.

Main Steam Line Isolation Va e Closure Scram The isolation scram antic ates the pressure and flux trans ents which occur during an isolation even and the loss of inventory durina pipe break.

This action minimizes e effect of this event on the fu and pressure vessel.

H. Reactor Coolant Lo Pressure Initiation of Main Stea Isolation Valve Closure The low pressu isolation of the main steam lin at 800 psig is provided to give prote ion against rapid reactor depres rization and the resulting rapid cooldo n of the vessel. Advantage is ta. n of the scram feature which occurs whe the main steam line isolation val es are closed, to provide the reactor s tdown so that high power operatio at low reactor pressure does not occu . Operation of the reactor at pr sures lower than 800 psig requir that the reactor mode switch be the startup position where prote gion of the fuel cladding integrit safety limit is provided by the IRM aigh neutron flux scram.

Ths, the combination of main steam ne low pressure isolation and

olation valve closure scram assur the availability of neutron scram protection over the entire range o applicability of the fuel cladding integrity safety limit.

Amendment No. 44, 4, .84, 4-64, 4173, A4i, BVY 00-51 17

V i VYNPS 3.1 Reactor Protection Syste The reactor protection sem automatically initiates a ractor scram to: \

1. preserve the int rity of the fuel barrier;

2. preserve the in egrity of the primary system barr er; and

3. minimize the ergy which must be absorbed, and revent criticality following a ss of coolant accident.

This specifica on provides the limiting conditi s for operation necessary to preserve t e ability of the system to tolera e single failures and still perform its intended function even during per ds when instrument channels may be out f service because of maintenance testing, or calibration. The basis for he allowable out-of-service time is provided in GE Topical Report N C-30851P-A, "Technical Specific ion Improvement Analysis for BWR Reactor rotection System," March 1988.

The eactor protection system is of e dual channel type. The system is ma up of two independent logic ch els, each having three subsystems of t pping devices. One of the thr subsystems has inputs from the manual ram push buttons and the react mode switch. Each of the two remaini ubsystems has an input from at east one independent sensor monitoring each of the critical parameter/. The outputs of these subsystems are combined in a 1 out of 2 log ; i.e., an input signal on either one o both of the subsystems will caus a trip system trip. The outputs of the trip systems are arranged so th a trip on both logic channels is requ ed to produce a reactor scram.

The required condition when the minimum instrument logic condi ions are not met are chosen so s to bring station operation promptly such a condition that the p ticular protection instrument is not r uired; or the station is placed i the protection or safe condition that e instrument initiates. This i accomplished in a normal manner witho subjecting the plant to abnormal/operating conditions.

When the minim requirements for the number of operab e or operating trip system and in rumentation channels are satisfied, t effectiveness of the protection s tem is preserved; i.e., the system ca tolerate a single failure and till perform its intended function of cramming the reactor.

Three AP instrument channels are provided for ach protection trip system to provi e for high neutron flux protection. RM's A and E operate contact in a trip subsystem, and APRM's C an operate contacts in the other rip subsystem. APRM's B, D, and F ar arranged similarly in the other protection trip system. Each protec on trip system has one more AP than is necessary to meet the minimu number required. This allows th bypassing of one APRM per protection rip system for maintenance, t ting, or calibration without changi the minimum number of channels quired for inputs to each trip syst . Additional IRM channels have also een provided to allow bypassing of e such channel. For a description of the Neutron Monitoring Systems, see SAR Section 7.5.

The bases for the scram settings or the IRM, APRM, high reactor pressure, reactor low water level, turbin control valve fast closure, and turbine stop valve closure are discuss d in Specification 2.1.

Amendment No. Q1, -64, ads8, BVY 00-78 29

v

- Aid BASES: 3.1 (Cont'd)

Instrumentation is provided to detect loss-of-coolant accident an initiate the core standby cooling e pment. This instrumentation is a backup to the water level instrume ation which is discussed in Specification 3.2.

The Control Rod Drive Scram S stem is designed so that all of the wat that is discharged from the eactor by the scram can be accommodate in the discharge piping. This d charge piping is divided into two sect ns. One section services the co rol rod drives on the north side of th reactor, the other serves the c trol rod drives of the south side. A art of the piping in each secti is an instrument volume which accomm ates in excess of 21 gallons of wa er and is at the low point in the pip g. No credit was taken for thi volume in the design of the discharg iping as concerns the amount of w er which must be accommodated during scram. During normal operati , the discharge volume is empty; how er, should it fill with water, e water discharged to the piping fro the reactor could not be accommod ed, which would result in slow scra times or partial or no control rX insertion. To preclude this occur nce, level instrumentation has been rovided for the instrument volume ich scram the reactor when the vo me of water reaches 21 gallons. A indicated above, there is suffi ient volume in the piping to accom date the scram without impairment of e scram times or amount of inserti of the control rods. This fu ction shuts the reactor down while ufficient volume remains to commodate the discharged water, a precludes the situation in which a cram would be required but not be ble to perform its function adequately.

The present design of the Scram scharge System is in concert with the BWR Owner's Group criteria, which yve previously been endorsed by the NRC in their generic "Safety Evaluaton Report (SER) for Scram Discharge Syste ",

dated December 1, 1980.

Loss of condenser vacuum ccurs when the condenser can no longer dle the heat input. Loss of c denser vacuum initiates a closure of the urbine stop valves and turbi bypass valves which eliminates the hea input to the condenser. C1s re of the turbine stop and bypass valves causes a pressure transient neutron flux rise, and an increase in s face heat flux. To prevent he clad safety limit from being exceed if this occurs, a reactor scram ccurs on turbine stop valve closure. T turbine stop valve closure ram function alone is adequate to prevyt the clad safety limit from be ng exceeded in the event of a turbine t ip transient without bypass.

Turbine s p valve (TSV) closure and turbine control lve (TCV) fast closure scram signals y be bypassed at <30V of reactor Rated Th al Power since, at low therma power levels, the margins to fuel thermal- draulic limits and reactor prima coolant boundary pressure limits are lar and an immediate scram is not nec sary. This bypass function is normally ac omplished automatically by pressure sw9 hes sensing turbine first stage pressure The turbine first stage pressure tpoint controlling the bypass of the scra signals on TCV fast closure and TSV losure is derived from analysis of reacto pressurization transients. Certain operational factors, such as turbine byp s valves open, can influence the relationship between turbine first sta pressure and reactor Rated Thermal Power.

However, above 30% of reactor Rated T ermal Power, these scram functions must be enable ,

Amendment No. i+, *&, a4-, 173 30

ED VYNPS BSS: 3.1 (Cont'd)

The main steam line i lat on valve closure scram is set to rmwhen the\

isolation valves are 0 percent closed from full open in 3- ut-of-4 lines.

This scram anticip es the pressure and flux transient, which would occur when the valves c se. By scramming at this setting, t resultant transient is in gnificant.

A reactor mo switch is provided which actuates o ypasses the various scram funct ns appropriate to the particular pla operating status.

The manu scram function is active in all mod , thus providing for manual means o rapidly inserting control rods durin all modes of reactor operatrn The RN system provides protection agains short reactor periods and, in co unction with the reduced APRM syste provides protection against e cessive power levels in the startup d intermediate power ranges. A ource range monitor (SRM) system is so provided to supply additional neutron level information during st tup and can provide scram function with selected shorting links remov during refueling. Thus, the IRM d the reduced APRM are normally re ired in the startup mode and may b required in the refuel mode. D ing some refueling activities whic require the mode switch in sta up; it is allowable to disconnect e LPRMs to protect them from damage d ring under vessel work. In lieu o the protection provided by the r duced APRM scram, both the IRM scr and the SRM scram in noncoincidenc are used to provide neutron monito ing protection against excess e power levels. In the power ran , the normal APRM system provides re ired protection. Thus, the lEN sy em and 15%

APRM scram are not re red in the run mode.

If an unsafe failur is detected during surveillance te ing, it is desirable to dete ne as soon as possible if other f lures of a similar type have occurre and whether the particular functi involved is still operable or capa le of meeting the single failure c teria. To meet the requirements o Table 3.1.1, it is necessary that 1 instrument channels in one trip s tem be operable to permit testing n the other trip system.

Thus, when ilures are detected in the first ip system tested, they would have/to be repaired before testing of t e other system could begin.

In the ma ority of cases, repairs or replace ent can be accomplished quickly. If repair or replacement cannot b completed in a reasonable time, eration could continue with one t pped system until the surve /lance testing deadline.

Amendment No. #E, #G., 44B, SWY -01 2, 212 31

I J VYNPS lBASES: 3.1 (Cont Id) o The requirement to hvgl scram functions, except thg lsted i Table 3.1.1, operabl/i h "Refuel" mode is to assu~ that shifting to this mode during reco oeation does not diminis te t need for the \

reactor protet / ystem./

The ability t bypass one instrument channel wh necessary to complete surveillanc testing will preclude continued eration with scram functio which may e either unable to meet the singl failure criteria or complete inoperable. It also eliminates e need for an unnecessary shutdo if the remaining channels and s ystems are found to be op able.

The c ditions under which the bypass is ermitted require an immed te det mination that the particular func on is operable. However, uring th time a bypass is applied, the fun ion will not meet the sin e failure iteria; therefore, it is prudent limit the time the bypass/is in effect by requiring that surveilla e testing proceed on a co inuous basis and that the bypass be removed as soon as testing is complet Sluggish indicator response du ng the perturbation test w 11 be indicative of a plugged instrument line r closed instrument valves This test assures the operability of the reac r pressure sensors as well s the reactor level sensors since both p ameters are monitored thro h the same instrument lines.

The independence of th safety system circuitry is etermined by operation of the scram test swi h. Operation of this swit during the refueling outage and following aintenance on these circui s will assure their continued independe ce.

Amendment No. 4+8, 191 32

LLLoLJ VYNPS BASES4.1 REACTOR PROTECTION SYSTE X A. The scram sensor chanl Xited in Tables 4.1.1 and 4.1.2 are X vied into three groups: A adC. Sensors that make up Group A s e the

\on-off type and will b etd and calibrated at the indicatfi l\ ~intervals.//

Group B devices u lize an analog sensor followed by an mplifier and bistable trip ci cuit. This type of equipment incorpo tes control room mounted i icators and annunciator alarms. A f ure in the sensor or amp fier may be detected by an alarm or an operator who observes tha one indicator does not track the oth s in similar channels. he bistable trip circuit failures ar detected by the periodic esting.

Group devices are active only during a giv portion of the operating cycle For example, the IRM is active durin start-up and inactive durig full-power operation. Testing of t ese instruments is only me a ingful within a reasonable period pr r to their use.

e basis for a three-month functional test interval for group (A) and (B) sensors is provided in NEDC-3085 -A, "Technical Specification Improvement Analysis for BWR Reacto Protection Systems," March 1988.

SRM/IRM/APRM overlap Surveillanc are established to ensure that no gaps in neutron flux indication exist from subcritical.to power operation for monitoring core eactivity status.

The overlap between SRMs an IRMs is required to be demonstrate to ensure that reactor power ill not be increased into a neutron flux region without adequate dication. This is required prior t withdrawing SRMs from t fully inserted position since ind ation is being transitioned fro the SRMs to the IRMs.

The overlap between RMs and APRMs is of concern when re cing power into the IRM range On power increases, the system des n will prevent further increases by initiating a rod block) if adequ e overlap is not maintained. Overlap between IRMs and APRMs exist when sufficient IRMs and APRMs oncurrently have onscale readings su that the transition be een the RUN and STARTUP/HOT STANDBY odes can be made without eith APRM downscale rod block, or IRM up cale rod block.

Overlap bet en SRMs and IRMs similarly exists wh n, prior to withdrawin the SRMs from the fully inserted pos tion, IRMs are above mid-scale n range 1 before SRMs have reached t e upscale rod block.

As note ,IRM/APRM overlap is only required to e met during entry into STARTU /HOT STANDBY Mode from the Run Mode. That is, after the overlap requilement has been met and indication has transitioned to the IRMs, main aining overlap is not required (APRMs may be reading downscale onc in the STARTUP/HOT STANDBY Mode).

I overlap for a group of channels is t demonstrated (e.g., IRM/APRM verlap), the reason for the failure o the Surveillance should be determined and the appropriate chann (s) declared inoperable. Only those appropriate channels that are equired in the current condition should be declared inoperable.

Amendment No. &&, 44i, 144, 186 33

VYNPS ABAES:4.1 (Cont'd) /

LPRM gai ettings are determined from he local flux profiles me 1ue by the fiaversing Incore Probe (TIPS System. This establis isth rel tiw local flux profile for appr iriate representative inpu 0oth APR ystem. The 2,000 megawatt-d ys per short ton (MWD/T) requec is ised on operating experience/with LPRM sensitivity chags n th the resulting nodal powe uncertainty, combined ith other i entified uncertainties, remans less than the total uncertainty i.e., 8.7%) allowed by the GE safety limit analysis.

The ratio of MFLPD to FRP sh 1 be checked once per day when operating at >25% Rated Thermal Power o determine if the APRM ins require adjustment. Because few c ntrol rod movements or po r changes occur, checking these parameters daily is adequate. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months allowance after thermal power >25 Rated Thermal Power is ac ieved is acceptable given the large inhere margin to operating limi at low power Amendment No. 46, -8",191 33a

E VYNPS 3.2 PROTECTIVE INSTRUMENTATION/

In addition to reactor prot et on instrumentation which initiates a rcto scram, station protectiv / srumentation has been provided which id iates action to mitigate the c nsequences of accidents which are beyond e reactor operator's abi ty to control, or terminate a single ope tor error before it results in erious consequences. This set of Specif ations provides the limiti conditions of operation for the primaryystem isolation function nd initiation of the core standby cooli and standby gas treatment sys ems. The objectives of the Specificatio are (i) to assure the effe _iveness of any component of such systems even during periods when p Xtions of such systems are out of servic for maintenance, testing, or c ibration, and (ii) to prescribe the tri settings required to assure adequ te performance. This set of Specificatins also provides the limiting c ditions of operation for the control ro block system and surveillac0 e instrumentation.

Isolati valves are installed in those lines t t penetrate the primary conta ment and must be isolated during a loss of-coolant accident so that the diation dose limits are not exceeded d ing an accident condition.

Act tion of these valves is initiated by p tective instrumentation shown in able 3.2.2 which senses the conditions or which isolation is required.

S ch instrumentation must be available w never primary containment ntegrity is required. The objective i to isolate the primary containmen so that the limits of 10 CFR 100 are n exceeded during an accident. T objective of the low turbine condense vacuum trip is to minimize the radioactive effluent releases to as ow as practical in case of a main condenser failure. Subsequent rel ses would continue until operator action was taken to isolate the main con nser unless the main steam line olation valves were closed automatically on low condenser vacuum. The man al bypass is required to permit initial a artup of the reactor during low wer operation.

The instrumentation which i tiates primary system isolation s connected in a dual channel arrangement. Thus, the discussion given in e bases for Specification 3.1 is appli able here.

The low reactor water 1 el instrumentation is set to tr p when reactor water level is 127" ab e the top of the enriched fuel. This trip initiates closure of Group 2 an 3 primary containment isolatio valves. For a trip setting of 127" abov the top of the enriched fuel, e valves will be closed before perfo tion of the clad occurs even f the maximum break and, therefore, the set ng is adequate.

The top of the e iched fuel (351.50 from vessel ottom) is designated as a common referenc level for all reactor water le el instrumentation. The intent is to m' imize the potential for operat r confusion which may result from differen scale references.

The low-low eactor water level instrumenta ion is set to trip when reactor water leve is 82.5" H2 0 indicated on the actor water level instrumentation above the op of the enriched fuel. This trip initiates closure of the Group 1 p imary containment isolation va yes and also activates the ECCS and RCIC Sys em and starts the standby dies generator system. This trip settin level was chosen to be low eno gh to prevent spurious operation, but high ough to initiate ECCS operatio and primary system isolation so that no m ting of the fuel cladding will occur, and so that post-accident cool ng can be accomplished and the imits of 10CFRIOO will not be violated.

Amendment No. 54, i&, 41, BVW 01 '2, 210 75

AEI VYNPS BASS 3.2 (Cont'd) /

For the complete circumfer break of 28-inch recirculationenai line with the trip setting give above, ECCS initiation and primary syste isolation are initiated n time to meet the above criteria. The a a instrumentation also toers the full range of spectrum breaks at s meets the above criteria. h a C The high drywell onsure instrumentation is a backup to t wager levell instrumentation,d d in addition to initiating ECCS, it aises isolation of Group 2, 3, anthe isolation valves. For the com c cumferential break discussed abo , this instrumentation will initepatrte S operation at aboutl the same tim as the low-low water level instrumentation, thus, the results given abovelare applicable here also. Certen ion valves including the TIP b imitin the los ofl ass invet, dfrmell vent, purge and sump valves ai isolated on high drywell pres sure. ever, since high drywell pressur could occur as the result of non-saf f-related causes, such as not ventinmthe drywell during startup, complete isolation is not detem desira le for these conditions and only cer n valves are required to enturisiarevpr viedsi the minlet amd oulnsa m;easomeurn stm Clo4 . The water level instrumentation in iates protection for the full Sp 0 rum of loss of coolant accidents any causes a trip of certain primary smiem isolation valves. t enturis are provided in the main sta ines as a means of measuring ste and also limiting the elow aa of Tssinventory from the vessel during

/ steam line break accident. In ad ion to monitoring steam flow,/

instrumentation is provided whix causes a trip of Group 1 isolation vales.

IThe primary function of the in is l is to detect a break in the etation odin steam line, thus only Group I ialves are closed. For the worst case/

accident, main steam line brsa outside the drywell, this trip sett of 140 percent of rated steam ow in conjunction with the flow limi n urs and l main steam line valve closr limit the mass inventory loss suchA ht fuel dis not uncovered, clad f emperatures remain less than 1295sa d release of radioactivity to etg s irons is well below iOCFRIOO. a Temperature monitori c nserumentation is provided in M he gi n steam line satunnel to detect le ts in this area. Trips are provided Th this instrumentation an en exceeded cause closure M cof Grou isolation valves.

Its setting of ae ient plus 95mF iste lineuadito mieaks of the order of 5 to 10 gpm; eus, it is capable of covering thi ie spectrum of breaks, For lai e breaks, it is a backup to high thral flow instrumentation discussed abov and for small breaksc with the reius t small release of radioactivitru gives isolation before the limit o 1 oCFRcr are exceeded.

Isolation o the condenser mechanical evacuum (MVPr is assumed in the safety a is for the control rod drop acc ent (CRDA). The MVP isolation instrumention initiates closure of the trsuction isolation valve followindevents in which main steam lin diation monitors exceed a predete

ned value. A High Main Steam neRadiation Monitor trip setting for MVr solation of T 3 times backgrou at rated thermal power (RTP) is as low a practicable without considerateo of spurious trips from nitrogen-16 spik A, instrument instabilities and ther operational occurrences.

Ipo teting the condenser MVP limits e release of fission products in the evnent of a CRDA. 2 Pesure instrumentation is pro fdd which trips when main steam line/

Pessure drops below 800 ps ig., / trip of this instrumentation results in /

Cloureof Group 1 isolation vives. In the refuel, shutdown, and startup/

/moesthis trip function is frovided when main steam line flow excees4W f oraedcapacity. This fu tion is provided primarily to provide/

tprtecton against a press e regulator malfunction which would causet Amendment No. .2&, *B, B4, 484, biSBY GI 52, 212 76

I'-;_ VYNPS BASES: 3.2 (Cont'd) control and/or bypass valves to en, resulting in a rapid depressurizatio and cooldown of the reactor ves el. The 800 psig trip setpoint limits the depressurization such that no xcessive vessel thermal stress occurs as a result of a pressure regulat malfunction. This setpoint was selected far enough below normal main st am line pressures to avoid spurious primary containment isolations.

Low condenser vacuum ha been added as a trip of the Group 1 isolation valves to prevent rele e of radioactive gases from the primary coola through condenser. Tie setpoint of 12 inches of mercury absolute w selected to provide ufficient margin to assure retention capabili in the condenser when gas low is stopped and sufficient margin below no al operating values. /

The HPCI and/or CIC high flow and temperature instrumentatio is provided to detect a brdk in the HPCI and/or RCIC piping. The HPCI nd RCIC steam supply pressu e instrumentation is provided to isolate the ystems when pressure may/e too low to continue operation. These iso ations are for equipment p otection. However, they also provide a div se signal to indicate possible system break. These instruments a included in Technica Specifications because of the potential for ossible system initiat n failure if not properly tested. Tripping this instrumentation result in actuation of HPCI and/or RCIC isolation alves, i.e., Group 6 valve{. A time delay has been incorporated into e RCIC steam flow trip logi to prevent the system from inadvertently i+/- ating due to pressure spi es which may occur on startup. The trip se ings are such that core u overing is prevented and fission product re ease is within limits.

he instrumentation which initiates ECCS ac on is arranged in a dual channel system. Permanently installed cir its and equipment may be used to trip instrument channels. In the nonfail afe systems which require energizing the circuitry, tripping an in rument channel may take the f rm of providing the required relay functio by use of permanently instal d circuits. This is accomplished in so cases by closing logic circu s with the aid of the permanently installed est jacks or other circuitry ich would be installed for this purpose The Rod Block Monitor (RBM) contr rod block functions are prov ded to prevent excessive control rod wi drawal so that MCPR does not ecrease below the fuel cladding integriy safety limit. The RBM is cr dited in the Continuous Rod Withdrawal Dur g Power Range Operation transcnt for preventing excessive control od withdrawal before the fuel ladding integrity safety limit (MCP or the fuel rod mechanical o rpower limits are exceeded. The RBM upper limit is clamped to provide otection at greater than 100% rated c re flow. The clamped value is ycle specific; therefore, it is locate in the Core Operating Limits R ort.

For single recirculat n loop operation, the RBM trip setting is reduced in accordance with the alysis presented in NEDO-30060 February 1983. This adjustment accounts for the difference between the ingle loop and two loop drive flow at the ame core flow, and ensures that the margin of safety is not reduced durin single loop operation.

During hot shu own, cold shutdown, and refueli g when the reactor mode switch is re red to be in the shutdown posit'on, the core is assumed to be subcritical ith sufficient shutdown margin; erefore, no positive reactivity nsertion events are analyzed. T e Reactor Mode Switch-Shutdown Position ntrol rod withdrawal block, requ red to be operable with the mod switch i the shutdown position, ensures t at the reactor remains subcrit cal by blocking control rod withd awal, thereby preserving the assump ions of the safety analysis. Two/channels are required to be Amendment No. #, 5, 69, '8.4-, 94, I 0,0, 211 77

ED VYNPS BASES: 3.2 (Cont'd) operable to ensure that no ingle channel failure will preclude a r d block when required. The is no trip setting for this function nce the channels are mechanical actuated based solely on reactor mode witch position. During refu ing with the reactor mode switch in th refueling position, the refuel osition one-rod-out interlock provides e required control rod withdra. 1 blocks.

To prevent exces ye clad temperatures for the small pip break, the HPCI or Automatic De ressurization System must function sin , for these breaks, react pressure does not decrease rapidly en gh to allow either core spray o LPCI to operate in time. For a break r other event occurring o taide the drywell, the Automatic Depre surization System is initiated n low-low reactor water level only af r a time delay. The arrangem K t of the tripping contacts is such as o provide this function when ne essary and minimize spurious operatio . The trip settings given in theSpecification are adequate to ensure e above criteria are met.

The Aecification preserves the effectivene s of the system during periods of iaintenance, testing, or calibration, d also minimizes the risk of i dvertent operation; i.e., only one in rument channel out of service.

he ADS is provided with inhibit swit es to manually prevent automatic initiation during events where actua ion would be undesirable, such as certain ATWS events. The system i also provided with an Appendix R inhibit switch to prevent inadver nt actuation of ADS during a fir which requires evacuation of the Contr Room.

Four radiation monitors are p vided which initiate isolation o the reactor building and operati of the standby gas treatment sy tem. The monitors are located in th reactor building ventilation duc and on the refueling floor. Any one pscale trip or two downscale tri of either set of monitors will cau e the desired action. Trip setti gs for the monitors on the refueling floor are based upon initiatin normal ventilation isolation nd standby gas treatment system eration so that none of the activity eleased during the refueling ac dent leave the Reactor Building vi the normal ventilation stack bu that all activity is processed by the s andby gas treatment system. Tri settings for the monitors in the ntilation duct are based upon i tiation of the normal ventilation isoation and standby gas treatment stem operation at a radiation leve equivalent to the maximum site eundary dose rate of 500 mrem/year as set forth in the Offsite Dose Calculation Manual. The monitoring s stem in the plant stack represe s a backup to this system to limit gros radioactivity releases to the e irons.

The purp e of isolating the mechanical v cuum pump line is to limit release of radioactivity from the main c ndenser. During an accident, fissio products would be transported f om the reactor through the main stea line to the main condenser. Th fission product radioactivity would be nsed by the main steam line rad tion monitors which initiate isK ation./

Amendment No. 4, -Z&, &3s, 9, a045, 4*&, a4, a@}, 211 78

/1.1 VYNPS BASES. 3.2 (Cont Id) - __

Specification 3.2.G requires pathe post-acc ent monitoring fPAMa instrumentation of Table 3. e operable during reactor power operatis PAM instrumentation is not equired to be operable during shutdown and refueling confns whi nohe likelihood of an event that would require PAM instrumentation i T low. The primary purpose of the opremely instrumentation is ts iclay plant variables that provide informatepn required by tme contor a operators during accident situations.

som an information provide the necessary support for the operator to t~ h manual actions f which no automatic control iB provided and t are required for sa ety systems to accomplish their safety funciU or design basis inidents. The operability of the PAM instre ation ensures tha tere is sufficient information available on Se ted plant parameters Xomonitor and assess plant status and behavo flowing an accident. This capability is consistent with the recoi ndations of Regulat Guide 1.97, "Instrumentation for Light Wate Cooled Nuclear Power ants to Assess Plant and Environs Conditions uring and Following an Ac ident.2/

I most cases, Table 3.2.6 requires a minimum o two operable channels to sure that the operators are provided the inf rmation necessary to determine the status of the plant and to bri the plant to, and maintain it in, a safe condition following an accide . For the majority of parameters monitored, when one of the required channels is inoperable, the required inoperable channel must be rest ed to operable status within 30 days. The 30-day completion time is b ed on operating experience and takes into account the remaining oper le channel (or in the case of a parameter that has only one require channel, an alternate means to monitor the parameter), the passiv nature of the instrument (no critical automatic action is assumed to o ur from these instruments), and the 1 probability of an event requiri PAM instrumentation during this interval.

If a PAM instrument channel as not been restored to an operable atus within the specified inte al, the required action is to prepare a written report to be submitted t the NRC within the following 14 days When a special written report required in accordance with the pro isions of Table 3.2.6, the repor will outline the preplanned alterna method of monitoring, the cause of the inoperability, and the plans nd schedule for restoring the inst entation to an operable status. Thi action is appropriate in lie of a shutdown requirement, since al rnative actions are identified be ore loss of functional capability, a given the likelihood of p nt conditions that would require inf rmation provided by this instrumen tion.

For the majo ity of PAM instrumentation, when two equired channels are inoperable or in the case of a parameter that monitored by only one channel, t .e channel and an alternate means ar inoperable), one channel (or the r quired alternate means) should be r stored to an operable status within yen days. The completion time of a yen days is based on the relati ly low probability of an event re ring PAM instrumentation and the n al availability of alternate mean to obtain the required info ation. Where specified, continuou operation with two required cha nels inoperable (or one channel an the required alternate means in perable) is not acceptable after s en days. Therefore, restoration of o e inoperable channel limits the ri that the PAM function will be in a aded condition should an accide occur.

Amendment No. 2047, 212 79

Al VYNPS ASES: 3.2 (Cont'd)

I For the majority of PAM Inst entation in Table 3.2.6, if two of the required channels (one requ ed channel per valve and alternate means or safety valve position ind ation) remain inoperable beyond the allow d interval, actions must b taken to place the reactor in a mode or ndition in which the limiting c ndition for operation does not apply. To/achieve this status, the reac r must be brought to at least hot shutdo within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed ompletion time is reasonable, based on op ating experience, to rea the required plant conditions from full 0ower conditions in an orderly ma er and without challenging plant systems. /It is not necessary to br g the reactor to cold shutdown since plan conditions during hot shutdown ado such that the likelihood of an accident at would require PAM instrumen ation is extremely low.

The Degrad Grid Protective System has been installe to assure that safety-re ated electrical equipment will not be subj cted to sustained degrade voltage. This system incorporates voltag relays on 4160 Volt Emerge y Buses 3 and 4 which are set to actuate the minimum voltage requl d to prevent damage of safety-related equ ment.

If egraded Grid conditions exist for 10 seco s, either relay will actuate a alarm to alert operators of this conditio i Based upon an assessment of ese conditions the operator may choose to anually disconnect the off-site power. In addition, if an ESF signal is i tiated in conjunction with low voltage below the relay setpoint for 10 s conds, the off-site power will be automatically disconnected.

The Reactor Core Isolation Cooling (R C) System provides makeup water t the reactor vessel during shutdown a d isolation to supplement or repl e the normal makeup sources without t e use of the Emergency Core Cooli g Systems. The RCIC System is init ted automatically upon receipt of/a reactor vessel low-low water lev signal. Reactor vessel high wa r level signal results in shutdown of te RCIC System. However, the syst will restart on a subsequent reacto vessel low-low water level sign . The RCIC System is normally lined up t take suction from the condensat storage tank. Suction will automati ally switch over from the conden te storage tank to the suppression poo on low condensate storage tank evel.

Upon receipt of a LOCA i tiation signal, if normal AC pow r is available, all RHR pumps and both re Spray pumps start simultaneou ly with no intentional time delay If normal AC power is not avail le. RHR pumps A and D start immediate y on restoration of power, RHR p ps B and C start within 3 to 5 secon of restoration of power and bot Core Spray pumps start within 8 to seconds of restoration of power. The purpose of these time delays is to tagger the start of the RHR and 96re Spray pumps on the associated Divis n 1 and Division 2 Buses, thus l iting the starting transients on t 4.16 kV emergency buses. The t e delay functions are only necessary hen power is being supplied from he standby power sources (EDGs). The Vme delays remain in the pump sta logic at all times as the time delay r ay contact is in parallel with t Auxiliary Power Monitor relay conta . Either contact closure will i tiate pump start. Thus, the time delays do not affect low pressure ECCS mp operation with normal AC power avalable. With normal AC power not ailable, the pump start relays which woqld have started the B and C RHR p ps and both Core Spray pumps are blocke by the Auxiliary Power Monitor co acts and the pump start time delay elays become the controlling devi s.

Amendment No. 4&, 4&, 411, 113, 132, W4&, 170, 207 79a

VYNPS 4.2 PROTECTIVE INSTRUMENTATEN/~

The Protective Instr entation Systems covered by this Spec'/ cationar \

listed in Table 4.a Mst of these protective systems are/omposed of two or more independen and redundant subsystems which are c dined in al dual-channel arr gement. Each of these subsystems con ins an arrangement of lectrical relays which operate to mitnate the required system prote v action./l The relays n a subsystem are actuated by a number/of means, including manually- erated switches, process-operated swit es (sensors), bistable devices erated by analog sensor signals, time , limit switches, and other r ays. In most cases, final subsystem lay actuation is obtained by sat sfying the logic conditions establishe by a number of these relay conta ts in a logic array. When a subsystem is actuated, the final subs=stem relay(s) can operate protective ipment, such as valves and pu s, and can perform other protective a ions, such as tripping the main t bine generator unit.

ith the dual-channel arrangement of t se subsystems, the single failure of a ready circuit can be tolerated b cause the redundant subsystem or system (in the case of high pressure coolant injection) will then initiate the necessary protective action. I a failure in one of these circuits occurs in such a way that an actio is taken, the operator is immediately alerted to the failure. If the f ilure occurs and causes no action, it could then remain undetected, ca sing a loss of the redundancy in the dual-channel arrangement. Losss in redundancy of this nature are fo by periodically testing the re ay circuits and contacts in the subsys ms to assure that they are opera ing properly.

The surveillance test inte al for the instrumentation channel fun tional tests are once/three month for most instrumentation. The allow le out-of-service times and surv illance interval is based on the foll ing NRC approved licensing topic reports:

1. GE Topical Repor NEDC-30851P-A, "Technical Specificat n Improvement Ana sis for BWR Reactor Protection Syste H March 1988.

2. GE Topical Re rt NEDC-30851P-A, Supplement 1 "Tec cal Specificatio Improvement Analysis for BWR Control od Block Instrumentat on," October 1988.

3. GE Topical eport NEDC-30851P-A, Supplement 2 echnical Specifica on Improvement Analysis for BWR Iso tion Instrumentation Common t RPS and ECCS Instrumentation," Marc 1989.

4. GE Topi al Report NEDC-31677P-A, "Technical pecification Improvyment Analysis for BWR Isolation Act tion Instrumentation, July 990.

5. GE pical Report NEDC-30936P-A, Parts and 2, "BWR Owners Group Te ical Specification Improvement Me odology (with Demonstration fo BWR ECCS Actuation Instrumentatio ," December 1988.

6. Topical Report GENE-770-06-1-A, " ases for Changes to urveillance Test Intervals and All wed Out-Of-Service Times For elected Instrumentation Technical ecifications," December 1992.

\7 /GE Topical Report GENE-770-06 "Addendum to Bases for Changes t

\ / Surveillance Test Intervals and 11lowed Out-Of-Service Times For

~Selected Instrumentation Techn 1 Specifications,"

a December 1992.

Amendment No. -14, 44. 4&, 14G, 44GG 4-G, BVY 00-78 so

- _.- . - - . - - I - - . - - - . - - - I - - -

VYNPS I-iIg 4.2 PROTECTIVE INSTRUMENTATION d) ont'd Since logic circuit t sresult in the actuation of a n equipment testing of this natu e was done while the plant was ut down for refueling. In thisway, the testing of equipment uld not jeopardize plant operation.

This Specifica on is a periodic testing progr which is based upon the overall testi g of protective instrumentation ystems, including logic circuits as ell as sensor circuits. Table .2 outlines the test, calibratio and logic system functional te schedule for the protective instrumen ation systems. The testing of a subsystem includes a functional test of ach relay wherever practicable. The testing of each relay includ all circuitry necessary to mak the relay operate, and also the prope functioning of the relay contac . Testing of the automatic init ation inhibit switches verifies e proper operability of the swi ches and relay contacts. Functi al testing of the inaccessible t tperature switches associated wit the isolation systems is accompl hed motely by application of a heat urce to individual switches.

All subsystems are functionally ested, calibrated, and operated their

/entirety.//

A channel functional test is erformed for the Reactor Mode S itch -

Shutdown Position function ensure that the entire channe will perform the intended function. Th surveillance is only required be performed once per operating cycle ring refueling. The Refueling utage frequeency is based on the need to rform this surveillance under he conditions that apply during a pla outage. Operating experienc has shown that this surveillance fre ncy is adequate to ensure fun ional operability.

Note 12 of Table 4.2. specifies that if the surveil ance frequency of every Refueling Outa is not met, functional test g of the Reactor Mode Switch - Shutdown P ition function shall be init ted within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after the reactor mode sg ch is placed in the Shutdow psition for the purpoe of commencing a s eduled Refueling Outage. T allows entry into the /

\ Sudow mde wh fithe surveillance requirement is not met./

Amendment No. hi, 94, 94, b-", lSC, 211 80a

Safety Assessment Discussion of Changes 3.1/4.1 and 3.2/4.2 Reactor Protection System Bases and Protective Instrument System Bases

SAFETY ASSESSMENT OF CHANGES TS SECTION 3.1/4.1 - REACTOR PROTECTION SYSTEM BASES TS SECTION 3.2/4.2 - PROTECTIVE INSTRUMENT SYSTEM BASES ADMINISTRATIVE A.1 The Bases of the current Technical Specifications for Sections 3.1/4.1 and 3.2/4.2 (pages 29 through 33a for Section 3.1/4.1 and pages 75 through 80a for Section 3.2/4.2) are completely replaced by revised Bases that reflect the format and applicable content of the proposed Vermont Yankee Nuclear Power Station (VYNPS) Technical Specifications in Sections 3.1/4.1 and 3.214.2, consistent with the Boiling Water Reactor Improved Standard Technical Specifications NUREG-1433, Rev. 2. The revised Bases are as shown inthe proposed WNPS Technical Specification Bases. The Bases changes are made for clarity purposes and conformance to the changes being made to the associated Technical Specifications. Inaddition, the information on Trip Settings provided inthe Bases for current Technical Specifications (CTS) Section 2.1 (pages 14 through17) is superceded by the revised Bases for Technical Specification Secton 3.1/4.1 and no longer necessary.

Therefore, this information inthe Bases for CTS Section 2.1 isdeleted and replaced with a statement referring to the Bases for the applicable Technical Specifications (i.e., the bases for the individual trip settings of Section 2.1 are discussed in the Bases for Specifications 3.1.A. 3.2.A, and 3.2.B). The Bases do not establish actual requirements, and as such do not change technical requirements inthe Technical Specifications. Therefore, the changes are administrative innature and have no negative impact on plant safety.

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

No Significant Hazards Consideration 3.1/4.1 and 3.2/4.2 Reactor Protection System Bases and Protective Instrument System Bases

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 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 asignificant 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 innature. Therefore, the change does not involve a significant reduction in a margin of safety.

.VYNPS 1 Revision 0

NO SIGNIFICANT HAZARDS CONSIDERATION TS SECTION 3.114.1 - REACTOR PROTECTION SYSTEM BASES TS SECTION 3.2/4/2 - PROTECTIVE INSTRUMENT SYSTEM BASES There were no specific less restrictive changes identified for these Bases.

VYNPS 1 Revision 0

ML031200664

Text

Subject:

2.1 REFERENCES

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