Text

Proposed Tech Specs,Consisting of Proposed Change 180, Correcting Typos & Text Inconsistencies
	ML20097C049
Person / Time
Site:	Vermont Yankee
Issue date:	02/05/1996
From:	; VERMONT YANKEE NUCLEAR POWER CORP.
To:	;
Shared Package
ML20097C048	List: ML20097C045; ML20097C049;
References
	NUDOCS 9602080074
	Download: ML20097C049 (79)
	v • d • e

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- Affected Technical Specification Pages 4

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( VYNPS 1

TABLE OP CONTENTS (Continced) l 1

LIMITING SAFETY j Pace No. SYSTEM SETTING l SAFETY LIMITS l, l

1.1 FUEL CLADDING INTEGRITY....................... 6 ... 2.1 j

, i 1.2. REACTOR COOLANT SYSTEM........................ 18 ' 2.2

... l LIMITING CONDITIONS OF OPERATION Pace No. SURVEILLANCE !

1 3.1 REACTOR PROTECTION SYSTEM..................... 20 ... 4.1 BASES 29 3.2 PROTECTIVE INSTRUMENT SYSTEMS................. 34. ... 4.2 A. Emergency core Cooling System............. 34 ... A !

B. Primary Containment Isolation............. 34 ... B C. Reactor Building Ventilation Isolation and Standby Gas Treatment System Initiation................................ 34 ... C D. Air Ejector Off-Gas System Isolation...... 35 ... D E. Control Rod Block Actuation............... 35 ... E F. Mechanical Vacuum Pump Isolation.......... 35 ... F

G. Post-Accident Instrmnentation............. 35 ... G H. -Drywell~to Torus AP Instrumentation........ 36 ... H 1 I. Recirculation Pump Trip 1 Instrumentation........................... 36 ... I l J. Deleted................................... 36 ... J 1 K. Degraded Grid Protective System........... 36 ... K-L. Reactor Core Isolation Cooling System Actuation................................. 37 ... L BASES 75 3.3- CONTROL ROD SYSTEM............................ 81 ... 4.3 A. Reactivity Limitations.................... 81 ... A I B. Control Rods.............................. 82 ... B l C. Scram Insertion Times..................... 85 ... C D. Control Rod Accumulators.................. 87 ... D E. Reactivity Anomalies...................... 88 ... E BASES 89 ,

1 3.4. REACTOR STANDBY LIQUID CONTROL SYSTEM......... 92 ... 4.4 A. Normal Operation.......................... 92 ... A B. Operation with Inoperable Components...... 93 ... B ,

C. Liquid Poison Tank - Boron Concentration............................. 93 ... C BASES 97 Amendment No. G7, 95, -ii-

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VYNPS '~

BASES: 2.1 (Cont'd) metal-water reaction to less than it, to assure that core geometry remains intact.

The design of the ECCS components to meet the above criteria was dependent on three previously set parameters: the maximum break size, the low water level scram setpoint, and the ECCS initiation setpoint. To lower the ECCS initiation setpoint would now prevent the ECCS components from meeting their design criteria. To raise the ECCS initiation setpoint would be in a safe direction, but it would reduce the margin established to prevent actuation of the ECCS during normal operation or during normally expected transients.

E. Turbine Stoo Valve Closure Scram Trio Settino The turbine stop valve closure scram trip anticipates the pressure, neutron flux and heat flux increase that could result from rapid closure of the turbine stop valves. With a scram trip setting of

<10% of valve closure from full open, the resultant increase in surface heat flux is limited such that MCPR remains above the fuel cladding integrity safety limit even during the worst case transient that assumes the turbine bypass is closed. This scram is bypassed when turbine steam flow is below 30% of rated, as measured by turbine first stage pressure.

F. Turbine Control Valve Fast Closure Scram The control valve fast closure scram is provided to limit the rapid -

increase in pressure and neutron flux resulting from fast closure of the turbine control valves due to a load rejection coincident with failure of the bypass system. This transient is less severe than the turbine stop valve closure with failure of the bypass valves and therefore adequate margin exists.

G. Main Steam Line Isolation Valve closure Scram The isolation valve closure scram anticipates the pressure and flux transients which occur during normal or inadvertent isolation valve closure. With the scram setpoint at 10% of valve closure, there is no increase in neutron flux.

H. Reactor Coolant Low pressure Initiation of Main Steam Isolation valve Closure The low pressure isolation of the main steam lines at 800 psig is provided to give protection against rapid reactor depressurization and the resulting rapid cooldown of the vessel. Advantage is taken of the scram feature which occurs when the main steam line isolation valves are closed, to provide the reactor shutdown so that high power operation at low reactor pressure does not occur. Operation of the reactor at pressures lower than 800 psig requires that the reactor mode switch be in the startup position where protection of the fuel cladding integrity safety limit is provided by the IRM high neutron flux scram.

Thus, the combination of main steam line low pressure isolation and l isolation valve closure scram assures the availability of neutron scram protection over the entire range of applicability of the fuel cladding integrity safety limit.

Amendment No. 18, E5, 64, 17

o s VYNPS 3.1 LIMITING CONDITIONS FOR 4.1 ~ SURVEILLANCE REQUIREMENTS OPERATION 3.1 REACTOR PROT 2CTION SYSTEM 4.1 REACTOR PROTECTION SYSTEM Applicability: Aeplicability:

l Applies to the operability of Applies to the surveillance of plant instrumentation and the plant instrumentation and l control systems required for control systems required for reactor safety. reactor safety.

i Obiective Obiective:

To specify the limits imposed on To specify the type 4.nd plant operation by those frequency of surveil hnee to be instrument and control systems applied to those instrument and required for reactor safety. control systems required for reactor safety, coecification: Specification:

A. Plant operation at any power A. Instrumentation systems level shall be permitted in shall be functionally tested accordance with Table 3.1.1. and calibrated as indicated .

The system response time in Tables 4.1.1 and 4.1.2, l from the opening of the .

respectively. '

sensor contact up to and ,

including the opening of the scram solenoid relay shall not exceed 50 milliseconds.

B. During operation with the B. Once a day during reactor ratio of MFLPD to FRP power operation the maximum greater than 1.0 either: fraction of limiting power j density and fraction of

a. The APRM System gains rated power shall be shall be adjusted by the detemined and the APRM ratios given in system gains shall be Technical adjusted by the ratios given Specifications 2.1.A.1 in Technical and 2.1.B or Specifications 2.1.A.1.a and 2.1.B.

b. The power distribution shall be changed to reduce the ratio of MFLPD to FRP.

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Amendment No. 64,. 20 l l

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O VYNPS TABLE 3.1.1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT REQUIREMENTS Required Conditions Minimum Number When Minimum Modes in Which Functions Must Operating Conditions For be Operating Instrument Operation Channels Per Are Not Trip Function Trio Settings Refuel (1) Startup (12) Run Trip System (2) Satisfied (3)

1. Mode Switch in X X X 1 A Shutdown

2. Manual Scram X X X 1 A l 3. IRM (7-41(A-F))

High Flux 5120/125 X X X(11) 2 A INOP X X X(11) 2 A l 4. APRM (APRM A-F)

High Flux 1 066 (W-AW)+54% X 2 A or B (flow bias) (4)

High Flux (reduced) 515% X X 2 A INOP X 2(5) A or B Downscale >2/125 X 2 A or B

5. High Reactor $1055 psig X X X 2 A Pressure l (PT-2-3-55(A-D)(M))

6. High Drywell $ 2.5 psig X X X 2 A Pressure l (PT-5-12(A-D)(M))

7. Reactor Low (6) >127.0 inches X X X 2 A Water Level (LT-2-3-57A/B(M))

(LT-2-3-58A/B(M))

8. Scram Discharge 521 gallons X X X 2 A Volume High Level ; (per volume) l (LT-3-231(A-H)(M))

Amendment No. M, M, 44, 64, 68, %, M, M, 94, 94, ,

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.-i VYNPS TABLE 3.1.1 (Cont'd) l REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT REQUIREMENTS Required Conditions l Minimum No. When Minimum Modes in which Functions Must Operating Conditions For-Trip Settings be Operating Instrument Operation And Allowable Channels Per Are Not Trio Function Deviations Refuel (1) Startup Run Trip System (2) Satisfied (3)

9. Main steamline high 3x normal X X X 2 A or C radiation (7) background at (RD-17-230(A-D)/ rated power (8) '

RM-17-251(A-D)/

RR-17-252)

10. Main steamline <10% valve X 4 A or C isolation valve closure closure (POS-2-80A/

86A-A1/B1, POS-2-80B/

86B-Al/B2, POS-2-80C/

86C-A2/B1, I

POS-2-80D/

86D-A2/B2)

11. Turbine control (9)(10) X 2 A or D valve fast closure l (PS-(37-40))

12. Turbine stop valve <10% valve (10)

X 2 A or D closure closure l (SOVS-5-(1-4))

Amendment No. 22

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4 VYNPS TABLE 3.1.1 NOTES

1. When the reactor is subcritical and the reactor water temperature is less than 212*F, only the following trip functions need'to be operable:

a) Emode switch in shutdown b) . manual scram c) high flux IRM or high flux SRM in coincidence d) scram discharge volume high water level

2. Whenever an instrument system is found to be inoperable, the instrument-system output relay shall be tripped immediately. Except 'for MSIV and .

Turbine Stop Valve Position, this action shall result in tripping the trip system.

3. When the requirements in the column " Minimum Number of Operating Instrument Channels Per Trip System" cannot be met for one syctem, that system shall be tripped. If the requirements cannot be met for both trip ;

systems, the appropriate actions listed below shall be taken:

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a)- Initiate insertion of operable rods and complete insertion o?' all operable rods within four hours. !

b) Reduce power level to IRM range and place mode switch in the J

Startup/ Hot Standby" position within eight hours.

j c) Reduce turbine load and close main steam line isolation valves within ;

8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . ,

d) Reduce reactor power to less than 30% of rated within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

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4. 'W" is percent rated two loop drive flow where 100% rated drive flow is that flow equivalent to 48 x 10 lbs/hr core flow. AW is the difference )

between the two loop and single loop drive flow at the same core flow. I This difference must be accounted for during single loop operation. l AW = 0 for two recirculation loop operation, j l

5. To be considered operable an APRM must have at least 2 LPRM inputs per level and at least a total of 13 LPRM inputs, except that channels A, C, D, and F may lose all LPRM inputs from the companion APRM Cabinet plus one additional LPRM input and'still be considered operable. !

6. The top of the enriched fuel has been designated as 0 inches and provides l common reference level for all vessel water level instrumentation.

7. Channel shared by the Reactor protection and Primary Containment Isolation Systems.

8. An alarm setting of 1.5 times normal background at rated power shall be established to alert the operator to abnormal radiation levels in primary coolant.

Amendment No. B+, 92, 66, 76, 94, 23

1 VYNPS TABL 4.1.1 (Cont'd)

SCRAM INSTRUMENTATION AND LOGIC SYSTEMS FUNCTIONAL TESTS MINIMUM FUNCTIONAL TEST FREQUENCIES FOR SAFETY INSTRUMENTATION, LOGIC SYSTEMS AND CONTROL CIRCUITS Instrument Channel GrouoI31 Functional TestI73 &nimum Frequency 848 l Scram Test Switch (5A-S2(A-D)) A Trip Channel and Alarm .rch Refualing Outage First Stage Turbine Pressure - A Trip Channel and Alarm ~Every 6 Months Permissive (PS-5-14(A-D))

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k Amendment No. . 25a i

. - - _ . _ _ _ . . _ . - . _ . _ . - . _ . , _ _ _ - _ - . _ _ _ - . _ _ _ _ . . _ . _ _ _ - . _ _ _ _ - - _ _ _ _ - . _ _ _ _ - _ . _ - _ . _ - _ _ _ _ _ _ _ _ - . . . _ _ _ - - - ,_ - . - ,- . - . , ~ - - -

VYNPS TABLE 4.1.2' SCRAM INSTRUMENT CALIBRATION MINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNELS Instrument Channel Groupill Calibration StandardHI Minimum Frecuencyt2) l liigh Flux APRM (APRM A-F)

Output Signal B Heat Balance Once Every 7 Days Output Signal (Reduced) B Heat Balance Once Every-7 Days Flow Bias B Standard Pressure and Voltage Refueling Outage Source l LPRM (LPRM ND-2-1-104(801) B(5) o Using TIP System Every 1000 Equivalent Full Power Hours High Reactor Pressure B Standard Pressure Source Once/ Operating Cycle Turbine Control Valve Fast Closure A Standard Pressure Source Every 3 Months High Drywell Pressure B Standard Pressure Source Once/ Operating Cycle High Water Level in Scram Discharge B Water Level Once/ Operating Cycle Volume Low Reactor Water Level B Standard Pressure Source Once/ Operating Cycle Turbine Stop Valve Closure A (6) Refueling Outage High Main Steam Line Radiation B Appropriate Radiation Refueling Outage Source (33 First Stage Turbine Pressure A Pressure Source Every 6 Months and After l Permissive (PS-5-14(A-D)) Refueling Main Steam Line Isolation Valve A (6) Refueling Outage Closure Amendment No. 44, M, M , 68, 61, %, -

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9 9 VYNPS BASES:

3.1 Reactor Protection system The reactor protection system automatically initiates a reactor scram to:

1. preserve the integrity of the fuel barrier

2. preserve the integrity of the primary system barrier; and

3. minimize the energy which must be absorbed, and prevent criticality following a loss of coolant accident.

This specification provides the limiting conditions for operation necessary to preserve the ability of the system to tolerate single failures and still perform its intended function even during periods when instrument channels may be out of service because of maintenance, testing, or calibration.

The reactor protection system is of the dual channel type. The system is made up of two independent logic channels, each having three subsystems of tripping devices. One of the three subsystems has inputs from the manual scram push buttons and the reactor mode switch. Each of the two remaining subsystems has an input from at least one independent sensor monitoring each of the critical parameters. The outputs of these subsystems are combined in a 1 out of 2 logic; i.e., an input signal on either one or both of the subsystems will cause a trip system trip. The outputs of the trip systems are arranged so that a trip on both logic channels is required to produce a reactor scram. -

The required conditions when the minimum instrument logic conditions are .

not met are chosen so as to bring station operation promptly to such a condition that the particular protection instrument is not required; or the station is placed in the protection or safe condition that the instrument initiates. This is accomplished in a normal manner without i subjecting the plant to abnormal operating conditions. I When the minimum requirements for the number of operable or operating trip system and instrumentation channels are satisfied, the effectiveness of the protection system is preser<ed; i.e., the system can tolerate a single failure and still perform its intended function of scramming the reactor.

Three APRM instrument channels are provided for each protection trip system to provide for high neutron flux protection. APRM's A and E operate contacts in a trip subsystem, and APRM's C an E operate contacts in the other trip subsystem. APRM's B, D, and F are arranged similarly in the other protection trip system. Each protection trip system has one more APRM than is necessary to meet the minimum number required. This allows the bypassing of one APRM per protection trip system for maintenance, testing, or calibration without changing the minimum number of channels required for inputs to each trip system. Additional IRM channels have also been provided to allow bypassing of one such channel.

IRM assignment to the bypass switches is described on FSAR Figure 7.5-9 l and in FSAR Section 7.5.5.4.

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

l Instrumentation is provided to detect a loss-of-coolant accident and initiate the core standby cooling equipment. This instrumentation is a backup to the water level instrumentation which is discussed in Specification 3.2.

Amendment No. 41, 29

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VYNPS BASES:

4.1l REACTOR PROTECTION SYSTEM

'A. The scram sensor channels listed in Tables 4.1.1 and 4.1.2 are divided into three groups: A, B and C. Sensors that make up Group A are the on-off type and will be. tested and calibrated at the indicated intervals. Initially the tests are more frequent than Yankee experience indicates necessary. Howevers by testing more frequently, the confidence level with this instrumentation will increase and testing will provide data to justify extending the test intervals as experience is accrued. i j Group B devices utilize an analog sensor followed by an amplifier and I

l bistable trip circuit. This type of equipment incorporates control room mounted indicators and annunciator alarms. A failure in the I sensor or amplifier may be detected by an alarm or by.an operator who I observes that one indicator does not track the others in similar i channels. The bistable trip circuit failures are detected by the l periodic testing.

-Group C devices are active only during a given portion of the operating cycle. For example, the IRM is active during start-up and inactive during full-power operation. Testing of these instruments is only meaningful within a reasonable period prior to their use.

B. The ratio of MFLPD to FRP. shall be checked once per day to determine if the APRM gains require adjustment. Because few control rod movements or power changes occur, checking these parameters daily is .

adequate.

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Amendment No. 58, 61, 33

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l VYNPS 1 3.2 LIMITING CONDITIONS FOR 4.2 SURVEILLANCE REQUIREMENTS OPERATION 3.2 PROTECTIVE INSTRUMENT' SYSTEMS 4.2 PROTECTIVE INSTRUMENT SYSTEMS Applicability: Aoplicability:

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

protective function.

Obiective: Obiective:

To assure the operability of To verify the operability of i i

protective instrumentation protective instrumentation systems. systems.

Specification: Specification:

A. Emeroency Core Coolino A. Emeroency Core Coolino l System System i

When the system (s) it Instrumentation and logic !

initiates or controls is systems shall be required in accordance with functionally tested and ,

Specification 3.5, the calibrated as indicated in l

instrumentation which Table 4.2.1. -

initiates the emergency core cooling system (s) shall be operable in accordance with _

Table 3.2.1.

B. primary Containment B. primary Containment Isolation Isolation When primary containment Instrumentation and logic integrity is required, in systems shall be accordance with functionally tested and ,

specification 3.7, the calibrated as indicated in l instrumentation that Table 4.2.2.

initiates primary containment isolation shall be operable in accordance with Table 3.2.2.

C. Reactor Buildino Ventilation C. Reactor Buildino Ventilation Isolation and Standby Gas Isolation and Stenflbv Gas Treatment System Initiation Treatment System Initiation The instrumentation that Instrumentation and logic initiates the isolation of systems shall be g the reactor building functionally tested and ventilation system and the calibrated as indicated in actuation of the standby gas Table 4.2.3.

treatment system shall be operable in accordance with Table 3.2.3. 1 1

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1 Amendment No. 34 !

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6 o_ ..-

VYNPS '

3.2 LIMITING CONDITIONS FOR 4.2 SURVEILLANCE REQUIREMENTS OPERATION D. Off-Gas System Isolation D. Off-Gas System Isolation During reactor power . Instrumentation and logic operation, the systems shall be

. instrumentation that . functionally tested and initiates isolation of the calibrated as indicated in -

off-gas system shall be Table 4.2.4.

operable in accordance with Table 3.2.4. i E. Control Rod Block Actuation E. Control Rod Block Actfuation ;

During reactor power Instrumentation and logic operation the systems shall be instrumentation that functionally tested and calibrated as indicated in ;

initiates control rod block

shall be operable in- 'IAble 4.2.5.

accordance with Table 3.2.5.

F. . Mechanical vacuum Pump F. Mechanical Vacuum Pump Isolation . Isolation

1. Whenever the main steam During each operating cycle, line isolation valves automatic isolation and I are opent the mechanical securing of the mechanical l vacuum pump shall be- vacuum pump shall be -

capable of being verified while the reactor automatically isolated is shutdown.

and secured by a signal -

of high radiation in the

' main steam line tunnel or shall be manually ,

isolated and' secured.

2. If Specification 3.2.F.1 is not met following a

. routine surveillance check, the reactor shall be in the cold shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

G. Post-Accident G. Post-Accident Instrtunentation Instrumentation a During reactor power The post-accident l operation, the instrumentation shall be a instrumentation that functionally tested and/or l displays information in the calibrated in accordance control Room necessary for with Table 4.2.6.

the operator to initiate and

control the systems used during and following a l postulated accident or abnormal operating condition shall be operable in accordance with Table 3.2.6.

Amendment No. 9, 35

VYNPS --

3.2 LIMITING CONDITIONS FOR 4.2 SURVEILLANCE REQUIREMENTS ,

OPERATION '

H. Drvwell to Torus AP H. Drvwell to Torus AP

, Instrumentation Instrumentation

'I

1. During reactor power The Drywell to Torus AP_

operation, the Drywell Instrumentation shall be ;

to Torus AP. calibrated once every six- I Instrumentation months and an. instrument ;

(recorder #1-156-3 and check will be made once per :

instrument DPI-1-158-6) shift. , !

shall be operable except

.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 _l reason,' reactor '

operation is. permissible only during the '

succeeding thirty days unless the instrument is 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 Trio I. Recirculation' Pump Trip Instrumentation Instrumentation During reactor power The Recirculation Pump Trip i operation, the Recirculation Instrumentation shall be Pump Trip Instrumentation functionally tested and l shall be operable in calibrated in accordance accordance with Table 3.2.1. with Table 4.2.1.

J. Deleted J. Deleted K. Decraded Grid Protective K. Decraded 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 3.2.7. with Table 4.2.7. l Amendment No. SG, 58, 94, 94, 4M, 4-34, 36

. _ . . . . . .._..-.._..m _.m._._._...- _ .._ _.___ .. . _ . _ . . . . _ _ . _ _. , . , . _ _ _ _

VYNPS 1

3.2 LIMITING CONDITIONS FOR 4.2- SURVEILLANCE REQUIREMENTS.

OPERATION l t

L. Reactor Core Isolation L. Reactor Core Isolation Coolino system Actuation; Coolino System Actuation j

, 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 i '

instrumentation which- -Table 4.2.8. s l

- initiates actuation of this i system shall be operable in l accordance with Table 3.2.8.

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Amendment No. M , % , M , -1 M , 37

VYNPS TABLE 3.2.1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION Core Spray - A & B (Note 1)

Minimum Number of Required Action When Operable Instrument Minimum Conditions Channels per Trip For Operation

' System Trip Function .

Trip Level Setting Are Not Satisfied 2 High Drywell Pressure $2.5 psig Note 2 l ( PT-10-101 ( A-D) (M) )

2 Low-Low Reactor Vessel Water >R2.5* above top of Note 2 l Level (LT-2-3-72 ( A-D) (M) ) enriched fuel 1 Low Reactor Pressure 300 1P5 350 psig Note 2 (PT-2-3-56C/D(M))

2 Low Reactor Pressure 300 1 P 5 350 psig Note 2

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

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

~

1 Time Delay (14A-K16A &-B) 510 seconds Note 2 2 Pump (P-46-1A/B) Discharge >100 psig Note 5 l Pressure (PS-14-44(A-D))

1 Auxiliary Power Monitor --

Note 5 l (LNPX C/D) 1 Pump Bus Power Monitor --

Note 5 l (27/3A/B, 27/4A/B) 1 Trip System Logic --

Note 5 Amendment No. 44, 6B, 440, 4M, 4-40, 443, 38

a VYNPS 2

. . k TABLE 3.2.1 (Cont'd) ,

EMERGENCY CORE COOLING SYSTEM ACTUATION _ INSTRUMENTATION ;

Low Pressure Coolant Injection System A & B (Note 1)

~ '

-Minimum Number of Required Action When- - '

Operable Instrument Minimum Conditions Channels per Trip For Operation System Trio Function . Trio Level Setting Are Not Satisfied 1 Low Reactor Pressure 300.5 p 1 350 psig Note 2 t

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

2. High Drywell Pressure '

1 2.5 psig ' Note 2 3

l. (PT-10-101(A-D)(M)) ;

2 Low-Low Reaptor-Vessel Water >82.5" above top of Note 2l 'i

.] . Level (LT-2-3-72(A-D)(S1)) enriched fuel , ;

1 Time Delay (10A-K51A & B) c0 seconds- Note 5

1 Reactor Vessel' Shroud Level . >2/3 core height Note 5' l (LT-2-3-73A/B(M))

1 Time Delay (10A-K72A"& B) '160 seconds Note 5 1 Time Delay (10A-K50A'& B) 55 seconds'. Note 5 ,

1 Low Reactor Pressure' 100 $ p i 150 psig Note 2~

(PS-2-128A & B) 2 per pump RHR Pump Discharge Pressure >100 psig Note 5-(PS-10-105(A-H)).

2 High Drywell Pressure 12.5 psig Note 2 ,

l (PT-10-101(A-D)(S1)) .

b i

Amendment No. -14, 44, 68, -140, -143, 39 i .

. _ . _ _ _.m_ _ _ _ _ . _ . _ . _ _ . - - . _ . . _ _._______,______m_-_._mm_m_____.,__m _.-_ _ _ - _ _ _ _ _ _

a .. __._..,s. ..,,,s -mwe,,,e<-.- n r . m m. , e- e. -

c. ,-,m.,m-.+ , a-rw- * -i --w-

a VYNPS TABLE 3.2.1 (Cont'd)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION Low Pressure Coolant Injection System A & B (Note 1)

Minimum Number of Required Action When Operable Instrument Minimum Conditions Channels per Trip For Operation System Trio Function Trio Level Setting Are Not Satisfied 1 Time Delay (10A-K45A & B) <6 minutes Note 5 2 Low Reactor Pressure 300 < p < 350 psig Note 2 (PT-2-3-56A/B(S1) &

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

1 Auxiliary Power Monitor --

Note 5 l (LNPX C/D) 1 Pump Bus Power Monitor --

Note 5 .

l (27/3A/B, 27/4A/B) 1 Trip System Logic -~

Note 5 Amendment No. 11, 110, 14G, . 40

VYNPS TABLE 3.2.1 (Cont'd)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION High. Pressure Coolant Injection System Minimum Number of Required Action When Operable Instrument '

Minimum Conditions Channels per Trip . For Operation System Trio Function Trip Level Setting Are Not Satisfied 2 (Note 3) Low-Low Reactor Vessel Water Same as LPCI Note 5 l Level (LT-2-3-72(A-D)(S1))

2 (Note 4) Low Condensate Storage Tank- > 3% Note 5 l Water Level (LSL-107-5A/B) 2 (Note 3) High Drywell Pressure Same as LPCI Note 5 l (PT-10-101(A-D)(M))

l 1 (Note 3) Bus Power Monitor (23A-K41) --

Note 5 1 (Note 4) Trip System Logic --

Note 5 2 (Note 7) High Reactor Vessel Water <l77 inches above top of Note 5 l +

Level (LT-2-3-72A/B) (S4 ) enriched fuel Amendment No. 66, 86, 90, ,

41

VYNPS TABLE 3.2.1 (Cont'd)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION i

Automatic Depressurization Minimum Number of Required Action When Operable Instrument Minimum Conditions Channels per Trip .

For Operation-System (Note 4) Trio Function Trin Level Setting Are Not Satisfied 2 Low-Low Reactor Vessel Water Same as Core Spray Note 6 l Level (LT-2-3-72(A-D)(M))

2 High Drywell Pressure 5 2 5 psig Note 6 l (PT-10-101(A-D)(SI))

l 1 Time Delay (2E-K5A/B) 1120 seconds Note 6 l 1 Bus Power Monitor (2E-KlA/B) --

Note 6 1 Trip System Logic. --

Note 6 2 Time Delay 18 minutes Note 6 l (2E-K16A/B, 2E-K17A/B) k b

l Amendment No. 44, -145, ,

42

VYNPS TABLE 3.2.1 (Cont'd)

RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION I

RecirculationPumpTrip-A&BkNote1)

Minimum Number of Required Action When Operable Instrument Minimum Conditions Channels per Trip For Operation System Trio Function Trio Level Settino Are Not Satisfied 2 Low-Low Reactor Vessel Water > 6' 10.5" above top of Note 2

'l Level (LM-2-3-68(A-D)) enriched fuel 2 High Reactor Pressure 5 1150 psig Note 2 l (PM-2-3-54(A-D))

l 2 '/ime Delays (2-3-68 ( A-D) (X) ) $ 10 seconds Note 2 1 Trip Systems Logic --

Note 2 Amendment No. SB, 68, -70, -75 , -

43

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

i VYNPS ]

l TABLE 3.2.1 NOTES

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

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

or an orderly shutdown shall be initiated and the reactor shall be in the l cold shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . i i

3. One trip system with initiating instrumentation arranged in a one-out-of-two taken twice logic. l 1

4. One trip system with initiating instrumentation arranged in a-

l. one-out-of-two logic.

5. If the minimum number of operable channels are not available, the system

. is considered inoperable and the requirements of Specification 3.5 apply.

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 j requirements of Specification 3.5.F.2 and 3.5.F.3 shall apply. If the i minimum number of operable channels is not available in both trip l systems, specifications 3.5.F.3 shall apply.

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

l 1

l i

1 i

i l

Amendment No. GB, 44

VYNPS TABLE 3.2.2 PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION Minimum Number of Required Action When Operable Instrument Minimum Conditions l Channels per Trip For Operation Are System Trio Function Trio Setting Not Satisfied (Note 2) 2 Low-Low Reactor Vessel Water >82.5" above the top of A Level (LT-2-3-57A/B(S2), enriched fuel LT-2-3-58A/B(S2))

2 of 4 in each of High Main Steam Line Area 12120F B 2 channels Temperature

] (TS (121-124) ( A-D) )

2/ steam line High Main Steam Line Flow <140% of rated flow B j (DPT (116-119 ) ( A-D) (M) ) .

2/(Note 1) Low Main Steam Line Pressure >800 psig B l (PS-2-134(A-D))

2/(Note 6) High Main Steam Line Flow 140% of rated flow B (DPT-2-116A, 117B. 118C, 119D (S1))

2 Low Reactor Vessel Water Level Same as Reactor A (LT-2-3-57A/B(M), Protection System LT-2-3-58A/B(M))

2 High Main Steam Line Radiation 53 x background at rated B (RD-17-230(A-D)/ power (9)

RM-17-251(A-D)/RR-17-252) (7)

(8) 2 + High Drywell Pressure Same as Reactor .A l (PT-5-12(A-D)(M)) Protection. System 2/(Note 10) Condenser Low Vacuum $12" Hg absolute A l (PS-2-11(A-D))

1 Trip System Logic --

A Amendment No. 9, 66, 84, 66, 90, ,

45

VYNPS-TABLE 3.2.2 (Cont'd)

HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION INSTRUMENTATION Minimum Number of Required Action When-Operable Instrument Minimum Conditions Channels per Trip For Operation.Are System Trio Function Trio Level Setting Not Sa t is fi ed 2 per set of 4 High Steam Line Space 1212 F Note 3 Temperature l (TS (101-104) (B-D) )

1 High Steam Line d/p (Steam' 5195 inches of water . Note.3 l Line Break) (DPIS-23-77/78) 4 (Note 5)- Low HPCI Steam Supply. Pressure >70 psig _

' Note 3 l (PS-23-68(A-D))

2 Main Steam Line Tunnel. ~<2120F Note 3 l Temperature (TS-23-(101-104)A) 1 Time Delay (23A-K48) 1 35 minutes Note 3 (23A-K49) l 1 Bus Power Monitor (23A-K38)'- -- --

1 Trip System Logic -- --

Amendment No. 64, 144, 46

_ . - _ .. , . . . . . _ - . . _ _ . _ = . _ , . _ . - . . . . . . _ -......_.-.,-_._-.-_-_.-_,_._-a..-.; . . . . . _ . _ . . . _ . . . . . . _ _ _ . _ . . , _ _ __-__.2.:.

_ _ _ . _ .~

VYNPS TABLE 3.2.2 (Cont'd)

REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION INSTRUMENTATION Minimum Number of Required Action When Operable Instrument Minimum Conditions

' Channels per Trip For Operation Are System Trio Function Trip Level Setting Not Satisfied (Note 2) 2 Main Steam Line Tunnel ~"<212 F Note 3 l Temperature (TS-13-(79-82)A) 1 Time Delay (13A-K41) 5,35 minutes Note 3 (13A-K42) 2 per set of 4 High Steam Line Space 5.212 F Note 3 Temperature l (TS (79-82 ) (B, C, D) )

1 High Steam Line d/p (Steam 3,195 inches of water Note 3 l Line Break) (DPIS-13-83/84) 4 (Note 5) Low Steam Supply Pressure >50 psig Note 3 l (PS-13-87(A-D))

l 1 Bus Power Monitor (13A-K36) --

Note 3 ,

1 Trip System Logic- --

Note 3 1 Time Delay (13A-K7) 3< t 5,7 seconds . Note 3 (13A-K31)

Amendment No. 63, 4-14,

47

VYNPS TABLE 3.2.3 REACTOR BUILDING VENTILATION ISOLATION & STANDBY GAS TREATMENT SYSTEM INITIATION Minimum Number of Required Action When Operable Instrument Minimum Conditions Channels per Trip For Operation System Trio Function Trio Setting Are Not Satisfied 2 Low Reactor Vessel Water Level Same as PCIS Note 1 (LT-2-3-57A/B(M),

LT-2-3-58A/B(M))

2 High Drywell Pressure Same as PCIS Note 1 l ( PT-5-12 ( A-D) (M) )

1 Reactor Building Vent 5,14 mr/hr Note 1 l (RD-17-430A/B, RM-17-452A/B) 1 Refueling Floor Zone Radiation 5.100 mr/hr Note 1 l (RD-17-431A/B, RM-17-453A/B) 1 Reactor Building Vent Trip --

Note 1 System Logic 1 Standby Gas Treatment Trip --

Note 1 System Logic 1 Logic Bus Power Monitor I --

Note 1 l (16A-K52/53)

Note 1 - If the minimum number of operable instrument channels is not available in either trip system for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the reactor building ventilation system shall be isolated and the standby gas treatment system operated until the instrumentation is repaired.

Amendment No. ,

49

~

VYNPS TABLE 3.2.4 OFF-GAS SYSTEM ISOLATION INSTRUMENTATION 1

Minimum Number of Required Action When Operable Instrument Minimum Conditions Channels per Trip For Operation System Trio Function Trio Settino Are Not Satisfied 1 Time Delay (Stack Off-Gas s 2 minutes.- Note 1-Valve Isolation) (15TD & 16TD) $ 30 minutes 1 Trip System Logic- , Note 1 Note 1 - At least one of the radiation monitors between the charcoal bed system and the plant stack shall be operable during operation of the augmented off-gas system. .If this condition cannot be met, continued operation of the augmented off-gas system'is permissible for a period of up to 7 days provided that at least one of the stack monitoring systems is operable and off-gas system temperature and pressure are measured continuously.

L s

i Amendment No. 9, 84, ,

50

. ~ - ,-,--s,. 2 . , - - ,--_ . - . . r - .---- ,. - _ ,- . . - , ,,,s- ,, ---. ..- , -- , ,m. , - - , . - -,r--

VYNPS TA$1LE 3.2.5 CONTROL ROD BLOCK INSTRUMENTATION I

Minimum Number of Modes in Which Function Operable Instrument Must be Operable Channels per Trip System Trio Function Refuel Startup Run Trio Setting Startup Range Monitor 2 a. Upscale (Note 2) X X 15 x 105 cps (Note 3) l 2 (7-40(A-D)) X X

b. Detector Not Fully Inserted l (7-11(A-D)(LS-4))

Intermediate Range Monitor (Note 1) 2 a. Upscale (7-41(A-F)) X X $108/125 Full Scale 2 b. Downscale (Note 4) (7-41(A-F)) X X >5/125 Full Scale 2 c. Detector Not Fully Inserted ; X X l (7-11 (E-K) (LS-4 ) )

Average Power Range Monitor l (APRM A-F) 2 a. Upscale (Flow Bias) X 1 066(W-AW)+42% (Note 5) 2 b. Downscale X 12/125 Full Scale l Rod Block Monitor (RBM A/B)

(Note 6)

(Note 9) 1 a. Upscale (Flow Bias) (Note 7) X 1 066(W-AW)+N (Note 5) 1 b. Downscale (Note 7) X 12/125 Full Scale 1 Scram Discharge Volume X X X $12 Gallons l (Note 8) (per (LT-3-231 A/G) (S1) volume) .

1 Trip System Logic X X X Amendment No. -14, M , 25, 64, 66, 7-1, -76, 90, 94, -1M , 51

VYNPS TABLE 3.2.6 POST-ACCIDENT INSTRUMENTATION Minimum Number of Operable Instrument Channels Parameter Type of Indication Instrument Range i

2 Drywell Atmospheric Recorder 8TR-16-19 0-350"F Temperature (Note 1) Meter STI-16-19-30B (TE-16-19-30A/B) 0-350 F 2 Containment Pressure (Note 1) Meter BPI-16-19-12A (-15) -(+260) psig (PT-16-29A/B) Meter SPI-16-19-12B (-15) -(+260) psig 2 Torus Pressure (Note 1) Meter BPI-16-19-36A (-15) -(+65) psig (PT-16-19-36A/B) Meter #PI-16-19-36B (-15) -(+65) psig 2 Torus Water Level (Note 3) Meter 8LI-16-19-12A 0-25 ft.

l (LT-16-19-10A/B) Meter #LI-16-19-12B 0-25 ft.

2 Torus Water Temperature Meter BTI-16-19-33A 0-250 F (Note 1) (TE-16-19-33A/C) Meter 8TI-16-19-33C 0-2500F 2 Reactor Pressure (Note 1) Meter #PI-2-3-56A 0-1500 psig l (PT-2-3-56A/B) Meter 8PI-2-3-56B 0-1500 psig 2 Reactor Vessel Water Level- Meter 8LI-2-3-91A (-200)-0-(+200) 11 2 0 (Note 1) (LT-2-3-73A/B) Meter'#LI-2-3-91B (-200)-0-(+200) H 2 O 2 Torus Air Temperature (Note 1) Recorder BTR-16-19-45 0-3500F l (TE-16-19-39/41) (TE-16-19-34)

Meter 8TI-16-19-41 50-300 F 2/ valve Safety / Relief Valve Position Lights RV-2-71 (A-D) Closed - Open From Pressure Switches (Note 4) ( PS-2 (1-3 ) ( A-D) )

Amendment No. 50, 63, 90, 96, 4-14 , 145, -

53

. .. . . --= . .

i VYNPS

.I

-TABLE"3.2.6 (Cont'd)

- POST-ACCIDENT INSTRUMENTATION .

1 Minimum Number of .

Operable Instrument Channels Parameter Type of-Indication- Instrument Range l 1/ valve Safety Valve Position From ' Meter ZI-2-1A/B . Closed - Open Acoustic Monitor (Note 5) t

l. (ZE-2-1A/B) !

2 Containment Hydrogen / Oxygen Recorder SR-VG-6A (SI) 0-30% hydrogen Monitor-(Note 1)'(SAH-VG-5A/B) Recorder SR-VG-6B (SII) 0-25%_ oxygen 2 Containment.High-Range .

-Meter RM-16-19-1A/D 1 R/hr-107 R/hr_ I

' Radiation Monitor-(Note 6) l_ (RD-16-19-1A/B) l 1 Stack Noble Gas' Effluent Meter RM-17-155 0.1 - 107 mR/hr 'f l- (Note 7) (RD-17-155) t i

~

.t I ,

J -)

t

! .i l

t

-4 I

i . -1 Amendment No. fr3, 90, 96, 98, 54 )

~

VYNPS

l TABLE 3.2.7 EMERGENCY BUS UNDERVOLTAGE INSTRUMENTATION Minimum Number of Operable .

Instruments Parameter Trio -Settine Required Action 2 per bus Degraded Bus Voltage - Voltage 3,700 volts i 40 volts- Note 1 (27/3Z, 27/3W,.27/4Z, 27/4W) 2 per bus Degraded Bus Voltage'- Time 10 seconds i i second Note 2 Delay (62/3W, 62/3Z, 62/4W, 62/4Z)

TABLE 3.2.7 NOTES

1. If the minimum number of operable instrument channels are not'available, the inoperable channel shall be l tripped using test jacks or other permanently installed cirmaits within one hour.

j

2. If the minimum number of operable instrument channels are not availe.ble, reactor power operation is

permissible for only 7 successive days unless the system is sooner made operable.

Amendment No. 94, 56

e VYNPS l , TABLE 3.2.8 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION Minimum Number of Required Action When Operable Instrument Minimum CoTditions Channels per Trip for Operation System Trip Function Trio Level Setting are not Satisfied 2 (Note 1) Low-Low Reactor Vessel Water >B2.5" Above Top of Note 4 l Level (LT-2-3-72A-D)(M)) Enriched Fuel 2 (Note 2) Low Condensate Storage Tank >3% Note 4 l Water Level (LT-107-12A/B)(M)) ,

i 2 (Note 3) High Reactor Vessel Water <177" Above Top of Note 4 l Level (LT-2-3-72A/B) (S3 ) ) Enriched Fuel.

l 1 Bus Power Monitor (13A-K36) --

Note 4 1 Trip System Logic --

Note 4 r

Amendment No. 4-lh . 57

_ - _ _ , _ . . . _ _ . . _ . _ _ _ _ _ _ _ _ - . . . _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ . . _ _ _________.________________.__.m_-. - - - - _ ._ _ .. __ . - . -

o e l VYNPS l TABLE 3. 2. 8 NOTES

1. One trip system with initiating. instrumentation arranged in a.

one-out-of-two taken twice logic.

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

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

4. If the minimum number of operable channels are not available, the system )

is considered inoperable and the requirements of Specification 3.5 apply. '

i l

l l

l I

i

-)

l l

i l

Amendment'No. 4+1, 58

~

VYNPS TABLE 4.2.1 MINIMUM TEST AND CALIBRATION FREOUENCIES EMERGENCY CORE COOLING ACTUATION INSTRUMENTATION i

Core Spray System Trio Function Functional Test (8) Calibration (8) Instrument Check l High Drywell Pressure (Notes 1 and 4) Once/ Operating Cycle Once Each Day l Low-Low Reactor Vessel (Notes 1 and 4) Once/ Operating Cycle Once Each Day Water Level Low Reactor Pressure (Notes 1 and 4) Once/ Operating Cycle --

Low Reactor Pressure (Notes 1 and 4) Once/ Operating Cycle --

Pump (P-46-1A/B) (Note 1) Every Three Months --

Discharge Pressure ,

Auxiliary Power Monitor (Note 1) Every Refueling Once Each Day Pump Bus Power Monitor (Note 1) None Once Each Day Trip System Logic Once/ Operating Cycle Once/ Operating Cycle --

(Note 3)

Amendment No. 56, % , 106, 140, -140, 443, . 59

VYNPS TABLE 4.2.1 (Cont'd)

MINIMUM TEST AND CALIBRATION FREOUENCIES c

EMERGENCY CORE COOLING ACTUATION INSTRUMENTATION Low Pressure Coolant Injection System Trio Function Functional Test (8) Calibration (8) Instrument Check Low Reactor Pressure (Notes 1 and 4) Once/ Operating Cycle --

High Drywell Pressure (Notes 1 and 4) Once/ Operating Cycle Once Each Day I

l Low-Low Reactor Vessel (Notes 1'and 4) Once/ Operating Cycle Once Each Day Water Level l Reactor Vessel Shroud' (Notes 1 and 4) Once/ Operating Cycle --

Level Low Reactor Pressure (Note 1) Every Three Months --

l RHR Pump Discharge (Note 1) Every Three Months --

Pressure High Drywell Pressure (Notes 1 and 4) Once/ Operating Cycle --

Low Reactor Pressure (Notes 1 and 4) Once/ Operating Cycle --

Auxiliary Power Monitor (Note 1) Every Refueling Outage Once Each Day Pump Bus Power Monitor (Note 1) None Once Each Day Trip System Logic Once/ Operating Cycle Once/ Operating Cycle --

(Note 3)

Amendment No. -14, 58, 46, 90, 146, 144, 449, 442, '

60

- _ _ _ _ . _ _ _ . - _ _ _ _ _ . _ _ _ _ . _ . - _ _ _ _ _ _ _ _ _ _ . _ . _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _- - -. r_ - - _ >m

M VYNPS TABLE 4.2.1 (Cont'd)

MINIMUM TEST AND CALIBRATION FREQUENCIES EMERGENCY CORE COOLING ACTUATION INSTRUMENTATION High Pressure Coolant Injection System Trio Function Functional Test (8)- -Calibration (8) Instrument Check

'l Low-Low Reactor Vessel (Notes 1 and 4) Once/ operating cycle

.Once each day Water Level l Low Condensate Storage (Notes 1 and 4) Every three months '

Tank Water Level l Ifigh Drywell Pressure (Notes 1 and 4) Once/ operating cycle Once"each day

$us Power Monitor (Note 1) None- Once each day Trip System Logic Once/ operating cycle Once/ Operating cycle- --

(Note 3) l High. Reactor Vessel (Notes 1 and 4) Once/ operating cycle .--

Water Level Amendment.No. -lE 58, %, 95, 406, -140, 1-1, . 61

o VYNPS TABLE 4.2.1 (Cont'd)

MINIMUM TEST AND CALIBRATION FREQUENCIES EMERGENCY CORE COOLING ACTUATION INSTRUMENTATION Automatic Depressurization System Trin Function Functional Test (8) Calibration (8)- Instrument Check l Low-Low Reactor Vessel (Notes 1 and 4) Once/ Operating Cycle Once Each Day Water Level l Iligh Drywell Pressure (Notes 1 and 4) Once/ Operating Cycle Once Each Day Bus Power Monitor (Note 1) None Once Each Day Trip System Logic Once/ Operating Cycle Once/ Operating Cycle --

(Except Solenoids of (Notes.2 and 11) (Note 3)

Valves)

Amendment No. 58, M. M5, 106,

62 r

___ ~.

4

- .- e-VYNPS TABLE 4.2.1 (Cont'd)

MINIMUM TEST AND CALIBRATION FREOUENCIES EMERGENCY CORE COOLING ACTUATION INSTRUMENTATION Recirculation Pump Trip Actuation ~ System Trip Function Functional Test (8) Calibration (8) Instrument Check l Low-Low Reactor Vessel (Notes 1 and 4) Once/ Operating Cycle.

Once Each Day' Water Level l~ High Reactor Pressure (Notes 1 and 4) Once/ Operating Cycle- -Once Each Day; Trip System Logic Once/ Operating Cycle Once/ Operating Cycle --

4 s

1 k .

Amendment No. 68, -146, ,

63 i

___,,_w-c---_ - m - . &_ -s-h-

_ - - w- -uw,m 3e-ei= -w 6

VYNPS TABLE 4.2.2 MINIMUM TEST AND CALIBRATION FREOUENCIES PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION Trio Function Functional Test (8) Calibration (8) Instrument Check l Low-Low Reactor Vessel (Notes 1 and 4) Once/ Operating Cycle Once Each Day Water Level High Steam Line Area (Note 1) Each Refueling Outage --

Temperature High Steam Line Flow (Note 1) Once/ Operating Cycle Once Each Day Low Main Steam Line (Note 1) Every Three Months --

Pressure l Low Reactor Vessel Water (Notes 1 and 4) Once/ Operating Cycle --

Level High Main Steam Line (Notes 1 and 7) Each Refueling Outage Once Each Day-Radiation l High Drywell Pressure (Notes 1 and 4) Once/ Operating Cycle Once Each Day Condenser Low Vacuum (Note 1) Every Three Months --

Trip System Logic Once/ Operating Cycle Once/ Operating Cycle --

(Note 2) (Note 3)

Amendment No. 9, 58, %, 446, 3-3, . 64

VYNPS-TABLE 4.2.3 MINIMUM TES" AND CALIBRATION FREOUENCIES REACTOR BUILDING VENTILAT'JON AND STANDBY GAS TREATMENT SYSTEM ISOIATION Trip Function Functional Test (8) Calibration (8) Instrurnent Check l Low Reactor Vessel Water (Notes 1 and 4) Once/ Operating Cycle --

Level l High Drywell Pressure (Notes 1 and 4) Once/ Operating Cycle --

Reactor Building Vent Monthly Every Three Months Once Each Day Exhaust Radiation Refueling Floor Zone Monthly Every Three Months Once Each Day During Radiation Refueling Reactor Building Vent Once/ Operating Cycle Once/ Operating Cycle --

(Note 3)

Trip System Logic Standby Gas Treatment Once/Operatinc Cycle Once/ Operating Cycle --

Trip System Logic (Note 3)

Logic Bus Power Monitor sNote 1) None Once Each Day I

Amendment No. 58, % , 446, MO , 67

, . _ . . . m . .-m _ ,

VYNPS 5

TABLE 4.2.5 MINIMUM TEST AND CALIBRATION FREOUENCIES CONTROL' ROD BLOCK INSTRUMENTATION i Trio Function Functional Test Calibration

-Startup' Range Monitor

!~ a. Upscale Notes 4 and 6 Note'6 i b. Detector Not Fully Inserted Note 6 NA !

j Intermediate Range Monitor i

a. Upscale Notes 4 and 6 Note-6 ,

+ b. Downscale Notes 4 and 6 Note 6

c. Detector Not Fully Inserted Note 6 NA ,

Average Power Range Monitor [

l (APRM A-F)

a. Upscale (Flow Bias) Notes 1 and 4 Every Three Months !

b. Downscale Notes 1 and'4 Every Three Months t

l Rod Block Monitor (RBM A/B)

a. Upscale (Flow Bias) Notes k and 4- Every Three Months

b. Downscale Notes 1 and 4 Every Three Months i

- Trip System Logic 'Once/ Operating Cycle Once/ Operating Cycle i

(Note 3)

~

! High Water Level in Scram Discharge Every Three Months Refueling' Outage Volume t I i i

i i

Amendment No. M 6, 440, H6, 69

. . , _ _ . . _ . . . . . , .. _ - , _ . _ , - , . . . . . . . . _ . - _ . - _ . _ _ _ . . . . ._;. - . . . . . . _ . -...._.m , . _ , , _ . . . . _ . . . . _ .

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p t i l ,

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e V pPS l ' TABLE 4.2.8 MINIMUM TEST AND CALIBRATION FREQUENCIES REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION Trio Function Functional Test (8) Calibration (8) Instrument Check Low-Low Reactor Vessel (Note 1) 'Once/ Operating Cycle- Once each day Water. Level Low Condensate Storage (Note 1) Once/ Operating Cycle Tank Water Level High Reactor' Vessel -(Note 1) Once/ Operating Cycle --

Water Level .

Bus Power Monitor- (Note 1) None Once each day Trip System Logic Once/ Operating Cycle ' Once/ Operating Cycle- --

(Note 3)

Q Amendment No. 4-1-1, . 73-

.. , -- . _ . ~ . _ . . . - ~ . - . . . ... . - - . . ~ . - . . _ - . _ _ ~ . _ . _ . -

VYNPS

']

4 TABLE 4.2 NOTES

1. Initially once per month; thereafter, a longer interval as. determined by.

test results on this type of instrumentation.

. 2. During each refueling outage, simulated automatic. actuation which opens 1 all pilot valves shall be performed such that each trip system logic can l 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 shall be functionally tested in accordance with Definition G.I.

5. Deleted.

6. Functional tests, calibrations,'and instrument checks are not required when these instruments are not. required to be operable or are tripped. 1 Functional tests shall be performed before each startup with a required 'j frequency not to exceed once per week. Calibration shall be performed prior to or-during each startup or controlled shutdown with a required frequency not to exceed once per week. Instrument checks shall be performed at least once per day during those periods when. instruments are required to be operable.

7. This instrumentation is' excepted from the functional test definitions and shall be calibrated using simulated electrical signals onca every three months. .

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.

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.

Amendment No. O , 96, 445, 446, 74

- - n n , . - - , i. e , ., , , , - - -

e . 1 VYNPS 3.3 LIMITING CONDITIONS FOR 4.3 SURVEILLANCE REQUIREMENTS OPERATION 3.3 CONTROL ROD SYSTEM 4.3 CONTROL POD SYSTEM Aeolicability: Aeolicability:

Applies to the operational Applies to the surveillance status of the control rod requirements of the control rod system. sys te:a.

Obiective: Obiective: i To assure the ability of the To verify the ability of the control rod system to control control rod system to control reactivity. reactivity. !

1 Specification: Specification: 1 A. Reactivity Limitations _ A. Reactivity Limitations

1. . Reactivity Marcin - Core 1. Reactivity Marcin - Core Loadino Loadinc The core loading shall control rods shall be l be limited to that which withdrawn following a !

can be made subcritical refueling outage when 1 in the most reactive core alterations were condition during the performed to demonstrate -

operation cycle with the a shutdown margin of {

highest worth, operable 0.25 per cent Ak at any ;

control rod in its fully time in the subsequent I withdrawn position and fuel cycle with the all other operable rods highest worth operable inserted. control rod fully withdrawn and all other operable rods inserted.

2. Reactivity Marcin - 2. Reactivity Marcin -

Inoperable control Rods Inoperable Control Rods l Control rod drives which Each partially or fully cannot be moved with withdrawn operable control rod drive control rod shall be pressure shall be exercised one notch at considered inoperable. least once each week.

If a partially or fully This test shall be withdrawn control rod performed at least once drive cannot be moved per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in the with drive or scram event power operation is pressure, the reactor continuing with two or shall be brought to a more inoperable control shutdown condition rods or in the event within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months unless power operation is investigation continuing with one demonstrates that the fully or partially cause of the failure is withdrawn rod which not due to a failed cannot be moved and for control rod drive which control rod drive mechanism collet mechanism damage has not housing. The control been ruled out. The rod directional control surveillance need not be valves for inoperable completed within control rods shall be 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months if the number Amendment No. GG, 81

.- o VYNPS 3.3= LIMITING CONDITIONS FOR 4.3 SURVEILLANCE REQUIREMENTS OPERATION E. Reactivity Anomalies E. Reactivity Anomalies The' reactivity equivalent of -During the startup test

'the difference between the- program and startups actual critical rod following refueling outages, configuration and the the critical rod expected configuration configurations will be during power operation shall compared to the expected not exceed it Ak. If this configurations at selected limit is exceeded, the operating conditions. These reactor will be shut down comparisons will be used as until'the cause has been base data for reactivity

. determined and corrective- monitoring during subsequent actions have been taken if power operation throughout such actions are the fuel cycle. At specific appropriate, power operating conditions, the critical rod F. If Specification-3.3A configuration will be-l through E above are not met, compared to the an. orderly shutdown'shall be configuration expected based initiated and the reactor upon appropriately corrected shall be in the cold past data. This comparison shutdown condition within will be made at least'every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . equivalent full power month.

4 i

Amendment No. M, 88

- -- ~. _ __ . _. .

VYNPS -

BASES: 3.3 & 4.3 (Cont'd)

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

3. In the course of performing normal startup and shutdown procedures, a pre-specified sequence for the withdrawal or insertion of control rods is followed. Control rod dropout accidents which might lead to significant core damage, cannot occur if this sequence of rod withdrawals or insertions is followed. The Rod Worth Minimizer restricts withdrawals and insertions to those listed in the pre-specified sequence and provides an additional check that the reactor operator is following prescribed sequence. Although beginning a reacto::-

startup without having the RWM operable would entail unnecessary risk, continuing to withdraw rods if the RWM fails subsequently is acceptable if a second licensed operator verifies the withdrawal sequence. Continuing the startup increases core

. power, reduces the rod worth and reduces the consequences of dropping any rod. Withdrawal of rods for testing is permitted with the RWM inoperable, if the' reactor is suberitical and all .

other rods are fully inserted. Above 20% power, the RWM is not needed since even with a single error an operator cannot withdraw a rod with sufficient worth, which if dropped, would result in anything but minor consequences.

4. Refer to the Vermont Yankee Core Performance Analysis Report.

l*

5. The source Range Monitor (SRM) system has no scram functions. It does provide the operator with a visual indication of neutron level. The consequences of reactivity accidents are a function of the initial neutron flux. The requirement of at least three counts per second assures that any transient should it occur, begins at or above the initial value of 10-8,of rated power used in the analyses of transients from cold conditions. One operable SRM channel is adequate to monitor the approach to criticality, therefore, two operable SRM's are specified for added conservatism. >

6 '. The Rod Block Monitor (RBM') is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power level operation. During reactor operation with certain limiting control rod patterns, the withdrawal of a designated single control rod could result in one or more fuel rods with MCPR less than the fuel cladding integrity safety limit. During use of such patterns, it is judged that testing of the RBM system prior

. to withdrawal of such rods will provide added assurance that improper withdrawal does not occur. It is the responsibility of the Nuclear Engineer to identify these limiting patterns and the designated rods either when the patterns are initially established or as they develop due to the occurren.a of inoperable control rods.

Amendment No. M, M, 64, M, 90

VYNPS 3.4 LIMITING CONDITIONS FOR 4.4 SURVEILLANCE REQUIREMENTS OPERATION 3.4 REACTOR STANDBY LIOUID CONTROL 4.4 REACTOR STANDBY LIOUID CONTROL SYSTEM SYSTEM Apolicability: Applicability:

Applies to the. operating status Applies to the periodic testing of the Reactor Standby Liquid requirement ~for the Reactor Control System. Standby Liquid Control System, i

Obiective: Obiective:

)

To assure the availability of an To verify the operability of the independent reactivity control Standby Liquid Control System.

mechanism.

Specification: Soecification:

A. Normal Operation A. Normal Operation Except as specified ir 3.4.B The Standby Liquid Control below, the Standby Liquid System shall be verified control System shall be operable by:

operable during periods when fuel is in the reactor j unless: -

1

1. The reactor is in cold 1. Testing pumps and valves shutdown in accordance with ,

_ Specification 4.6.E. A j and minimum flow rate of 35 gpm at 1275 psig shall be verified for each pump by recirculating domineralized water to the test tank.

2. Control rods are fully 2'. Verifying the continuity inserted and of the explosive charges Specification 3.3.A is at least monthly.

met.

In addition, at least once during each operating cycle, the Standby Liquid Control System shall be verified operable by:

3. Testing"that the setting j of the pressure relief valves is between 1400 '

l and 1490 psig.

4. Initiating one of the l

, standby liquid control loops, excluding the i primer chamber and inlet fitting, and verifying that a flow path from a l pump to the reactor

/

Amendment No. -144, 4-28, 92

VYNPS 3.4 LIMITING CONDITIONS FOR 4.4 SURVEILLANCE REQUIREMENTS OPERATION vessel is available by pumping demineralized water into the reactor vessel. Both loops shall be tested over the course of two operating cycles.

5. Testing the new trigger l assemblies by installing one of the assemblies in the test block and firing it using the installed circuitry.

Install the unfired assemblies, taken from the same batch as the fired one, into the explosion valves.

6. Recirculating.the l borated solution.

B. Operation with Inoperable B. Operation with Inoperable Components Components From and after the date that When a component becomes a redundant component is inoperable, its redundant .

made or_found to be component shall be or shall inoperable, reactor have been demonstrated to be operation is permissible operable within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

during the succeeding seven days unless such component l's sooner made operable.

C. Liquid Poison Tank - Boron C. Licuid Poison Tank - Boron Concentration Concentration At all times when the Standby Liquid Control system is required to be operable, the following conditions shall be met:

1. The net volume versus 1. The solution volume in concentration of the the tank and temperature sodium pentaborate in the tank and suction solution in the standby piping shall be checked liquid control tank at least daily.

shall meet the requirements of Figure 3.4.1.

Amendment No. 442, 4-14, 93

o s VYNPS 3.5 LIMITING CONDITION FOR 4.5 SURVEILLANCE REQUIREMENT OPERATION

3. From and after the date 3. When the Alternate that the Alternate Cooling Subsystem or Cooling Tower Subsystem both Station Service or both Station Service Water Subsystems are Water Subsystems are made or found to be made or found inoperable inoperable, the operable for any reason, reactor subsystem (s) shall have l operation is permissible been or shall be only during the demonstra.ted to be succeeding seven days operable within l unless such subsystem (s) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

are made operable, provided that during such seven days all other active components of the other l subsystem (s) are operable.

4. If the requirements of i specification 3.5.D cannot be. met, an -

orderly shutdown shall I be initiated and the ;

reactor shall'be in a i

cold shutdown condition -

within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . l E. Hioh Pressure Coolant ;

- E. Hich Pressure Coolino Iniection (HPCI) System i Iniection'(HPCI) System Surveillance of HPCI System shall be performed as

. follows:

1. Testino

1. Except as specified in specification 3.5.E.2, Item Frecuency whenever irradiated fuel is in the reactor vessel Simulated Each re-and reactor pressure is Automatic fueling greater than 150 psig Actuation outage and prior to reactor Test startup from a cold condition: Operability testing of the pump and valves

a. The HPCI System shall be in accordance shall be operable. with Specification 4.6.E.

b. The condensate The HPCI System shall storage tank shall deliver at least contain at least 4250 gpm at normal 75,000 gallons of reactor operating condensate water, pressure when recirculating to the Condensate Storage Tank.

Amendment No. M, 444, 4M, 105

VYNPS 3.5 LIMITING CONDITION FOR 4.5 SURVEILLANCE REQUIREMENT OPERATION

2. From and after the date 2. When the HPCI Subsystem that the HPCI Subsystem is made or found to be 1 is made or found to be inoperable, the inoperable for any Automatic reason, reactor Depressurization System operation is permissible shall have been or shall ,

only during the be demonstrated to be succeeding seven days operable within 1 unless such subsystem is 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

sooner made operable, provided that during NOTE: Automatic ;

such seven days all Depressurization active components of the System Automatic operability shall i Depressurization be demonstrated -

Subsystems, the Core by performing a '

Spray Subsystems, the functional test ,

LPCI Subsystems, and the of the trip RCIC System are system logic, operable.

3. If the requirements of ;

Specification 3.5.E cannot be met, an orderly shutdown shall be initiated and the reactor pressure shall be reduced to less than 120 psig within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

F. Automatic Depressurization F. Automatic Depressurization System System surveillance of the Automatic Depressurization System shall be performed as follows:

1. Operability testing of

1. Except as specified in the relief valves shall Specification 3.5.F.2 be in accordance with below, the entire Specification 4.6.E.

Automatic Depressurization Relief System shall be operable at any time the reactor pressure is above 100 psig and irradiated fuel is in the reactor vessel.

2. When one relief valve of

2. From and after the date the Automatic Pressure that one of the four Relief Subsystem is made relief valves of the or found to be Automatic inoperable, the HPCI Depressurization Subsystem shall have Subsystem are made or been or shall be found to be inoperable demonstrated to be ,

due to malfunction of operable within the electrical portion 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

of the valve when the Amendment No. M, 144, 448, 106

l J

e .

VYNPS 3.5 LIMITING CONDITION FOR 4.5 SURVEILLANCE REQUIREMENT OPERATION reactor is pressurized above 100 psig with irradiated fuel in the reactor vessel, continued reactor operation is permissible only during the succeeding seven days unless such a valve is i sooner made operable, provided that during such seven days both the remaining Automatic Relief System valves and the HPCI System are operable.

3. If the requirements of Specification 3.5.F .

cannot be met, an i orderly shutdown shall be initiated and the reactor pressure shall l be reduced to less than 100 psig within ~

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

G. Reactor Core Isolation G. Reactor Core Isolation Coolino System (RCIC) Coolino System (RCIC)

Surveillance of the RCIC System shall be performed as follows:

1. Except as specified in 1. Testino ,

Specification 3.5.G.2 ,

below, the RCIC System Item Frecuency shall be operable whenever the reactor Simulated ^ Each re-pressure is greater than automatic fuelin'g 150 psig and irradiated actuation outage fuel is in the reactor test i vessel. (testing valve

2. From and after the date operability) ,

that the RCIC System is made or found to be Operability testing of inoperable for any the pump and valves j reason, reactor shall be in accordance operation is permissible with only during the specification 4.6.E.

succeeding 7 days unless The RCIC System shall such system is sooner deliver at least 400 gpm made operable, provided at normal reactor j that during such 7 days operating pressure when all active components of recirculating to the the HPCI System are Condensate Storage Tank.

operable.

Amendment No. M. M, 4-14, as, 107

VYNPS 3.5 LIMITING CONDITION FOR 4.5 SURVEILLANCE REQUIREMENT OPERATION

3. If the requirements of Specification 3.5.G cannot be met, an orderly shutdown shall be initiated and the reactor pressure shall l be reduced to less than 120 psig within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , i H. Minimum core and Containment H. Minimum Core and Containment Coolino System Availability Coolino System Availability

1. During any period when 1. When one of the l- one of the emergency emergency diesel l diesel generators is generators is made or inoperable, continued found to be inoperable, reactor operation is .the remaining diesel

. permissible only during generator shall have the succeeding'seven been or shall be days, provided that all- demonstrated to be l of the LPCI and CS and operable within Containment Cooling 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Subsystems connecting to the operable diesel '

generator shall be operable. If this requirement cannot be met, an orderly shutdown _

shall be initiated and the reactor shall be in i the cold shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2. Any combination.of I

-inoperable components in the Core and Containment Cooling Systems shall not defeat the capability of the remaining operable - l i

components to fulfill the core and containment cooling functions. l

3. When irradiated fuel is in the reactor vessel and the reactor is in the cold shutdown condition, all Core and l Containment Cooling )

Subsystems may be inoperable provided no work is permitted which has the potential for draining the reactor 1 vessel. l i

l Amendment No. M, 444, 108 l

. _ _ . . . . _ . _ _ . . . . _ - . . . _ _ - _ . __ _ . _.. _ _ ..-.m.._.-m...._ _ _ . _ _ .

1 VYNPS I 1

, 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE' REQUIREMENTS !

OPERATION

{

~

e. -With the !

-radioiodine concentration in

. the reactor coolant i

, greater than i 1.1 microcuries/ t gram dose equivalent.I-131, a.

sample of reactor coolant shall'be !

taken every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months j and analyzed for radioactive iodines of I-131 through j I-135, until the r specific activity ;

of the reactor

, coolant is restoredi l f

below-1.1 microcuries/ ;

gram dose.

equivalent I-131.

2. The reactor coolant 2. DuriNg'startups and at l water shall not exceed. -steaming rates below !

the following. limits 100,000 pounds per hour, * !

with steaming rates less a sample of reactor- ,

than 100,000 pounds per coolant shall be taken .

hour except as specified every four hours and !

in specification- -analyzed for- ,

3.6.B.3: conductivity and !

chloride content. '

. conductivity Sumho/cm l, Chloride ion 0.1 ppm 3 .- For reactor startups the 3. a. With steaming rates maximum value for- greater than or .

conductivity shall not equal to

. exceed'10 unho/cm and 100,000 pounds per. l the maximum value for hour, a reactor i chloride ion. coolant sample concentration shall not shall be taken at exceed 0.1 ppm, in the least.every reactor coolant water 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months and when for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months the continuous after placing the conductivity reactor in the power monitors indicate operating condition, abnormal conductivity (other than'short-term spikes), and analyzed for conductivity and chloride ion content.

Amendment No. %, ,

118

.. e WNPS - -

3 l. 6 LIMITING CONDITIONS FOR- 4.6 SURVEILLANCE REQUIREMENTS OPERATION -

b. When the continuous conductivity monitor is inoperable, a reactor coolant.

sample shall be taken'every four

- hours and analyzed for conductivity and chloride ion content.

4. Except as specified'in Specification 3.6.B.3 above, the reactor coolant water shall.not exceed the following limits with steaming rates greater than or equal to 100,000 pounds per hours.

Conductivity 5 uhmo/cm-Chloride ion- 0.5 ppm l S. If Specification 3.6.B -

is not met, an orderly shutdown shall be initiated and the __.

reactor shall be in the cold shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

C. Coolant Leakace C. Coolant Leakaae'

1. a. Any time irradiated 1. . Reactor coolant system fuel is in the leakage, for the purpose-reactor vessel and of satisfying reactor coolant Specification 3.6.C.1, temperature is shall be checked and above 212*F, logged once per shift, reactor coolant not to exceed 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

leakage into the primary containment from unidentified sources shall not exceed 5.gpm. In addition, the total reactor coolant system leakage into the primary containment shall not. exceed 25 gpm.

b. While in the run mode, reactor coolant leakage into the primary I containment from l unidentified i sources shall not l

Amendment No. 4M, 119

+ --n , ---

~

= a VYNPS BASES: 3.6 and 4.6 (Cont'd)

The actus1 shift in RTyyp of the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASTM E185 reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area. Since the neutron spectra at the irradiation samples and vessel inside radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel.

Battelle Columbus Laboratory Report BCL-585-84-3, dated May 15, 1984, provides this information for the ten-year surveillance capsule. In order to estimate the material properties at the 1/4 and 3/4 T positions in the vessel plate, the shift in RTyyp is determined in accordance with Regulatory Guide 1.99, Revision 2. The heatup and cooldown curves must be recalculated when the ARTer determined from the surveillance capsule is different from the calculated ARTyp7 for the equivalent capsule radiation exposure.

The pressure-temperature limit lines, shown on Figure 3.6.1, for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10CFR50 for reactor criticality and for inservice leak and hydrostatic testing. -

The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided j to assure compliance with the requirements of Appendix H to 10CFR l

. 1 Part 50.

l l B. Coolant Chemistrv l A steady-state radiciodine concentration Ibnit of 1.1 Ci of I-131 dose equivalent per gram of water in the Reactor Coolant System can be reached if the gross radioactivity in the gaseous effluents is l

near the limit, as set forth in Specification 3.8.E.1, or there is a failure or prolonged shutdown of the cleanup demineralizer. In the event of a steam line rupture outside the drywell, the NRC staff calculations show the resultant radiological dose at the site boundary to be less than 30 Rem to the thyroid. This dose was calculated on the basis of the radiciodine concentration limit of 1.1 pCi of I-131 dose equivalent per gram of water, atmospheric diffusion from an equivalent elevated release of 10 meters3 at the nearest site boundary (190 m) for a X/Q = 3.9 x 10 4 sec/m (Pasquill D and 0.33 m/see equivalent), and a steam line isolation valve closure time of five seconds with a steam / water mass release of 30,000 pounds.

The iodine spike limit of four (4) microcuries of I-131 dose equivalent per gram of water provides an iodine peak or spike limit for the reactor coolant concentration to assure that the radiological l consequences of a postulated LOCA are within 10CFR Part 100 dose guidelines.

The reactor coolant sample will be used to assure that the limit of Specification 3.6.B.1 is not exceeded. The radioiodine. concentration would not be expected to change rapidly during steady-state operation over a period of 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months . In addition, the trend of the radioactive gaseous effluents, which is continuously monitored, is a good indicator of the trend of the radiciodine concentration in the reactor coolant. When a significant increase in radioactive gaseous effluents is indicated, as specified, an additional reactor coolant sample shall be taken and analyzed for radioactive iodine.

Amendment No. 44, 6B, 91, 43, 140

. s VYNPS BASES: 3.6 and 4.6 (Cont'd) impurities will also be within their normal ranges. The reactor cooling samples will also be used to determine the chlorides.

Therefore, the sampling frequency is considered adequate to detect long-term changes in the chloride ion content. Isotopic analyses l required by Specification 4.6.B.l.b may be performed by a gamma scan and gross. beta and alpha determination.

The conductivity of the feedwater is continuously monitored and alarm set points consistent with Regulatory requirements given in Regulatory Guide 1.56, " Maintenance of Water Purity in Boiling Water Reactors," have been determined. The results from the conductivity monitors on the feedwater can be correlated with the results from the conductivity monitors on the reactor coolant water to indicate demineralizer breakthrough and subsequent conductivity levels in the reactor vessel water.

C. Coolant Leakace The 5 gpm limit for unidentified leaks was established assuming such leakage was coming from the reactor coolant system. Tests have been conducted which demonstrate that a relationship exists between the size of a crack and the probability that the crack will propagate.

These tests suggest that for leakage somewhat greater than the limit l specified for unidentified leakage, the probability is small that imperfections or cracks associated with such leakage would grow rapidly. Leakage less than the limit specified can be detected within a few hours utilizing the,available leakage detection systems.

If the limit is exceeded and the origin cannot be determined in a reasonably short time the plant should be shutdown to allow further investigation-and corrective action.

The 2 gpm increase limit in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period for unidentified leaks was established as an additional requirement to the 5 gpm limit by Generic Letter 88-01, "NRC Position on Intergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping."

The removal capacity from the drywell floor drain sump and the equivalent drain sump is 50 gpm each. Removal of 50 gpm from either of these sumps can be accomplished with considerable margin.

D. Safety and Relief Valves Parametric evaluations have shown that only three of the fot relief valves are required to provide a pressure margin greater than the recommended 25 psi below the safety valve actuation settings aa well as a MCPR > 1.06 for the limiting overpressure transient below 98%

power. Consequently, 95% power has been selected as a limiting power level for three valve operation. For the purposes of this limiting condition a relief valve that is unable to actuate within tolerance l of its set pressure is considered to be as inoperable as a mechanically malfunctioning valve.

Experience in safety valve operation shows that a testing of 50% of the safety valves per refueling outage is adequate to detect failures or deterioration. The tolerance value is specified in Section III of the ASME Boiler and Pressure Vessel Code as 1% of design pressure.

An analysis has been performed which shows that with all safety valves set 1% higher the reactor coolant pressure safety limit of 1375 psig is not exceeded.

Change 16/ March 28, 1974, 44, 44, 4G4, 449, 142

. t VYNPS BASES: 3. 6 and 4.6 (Cont'd)

E. Structural Inteority and coerability Testino l- A pre-service inspection of the components listed in Table 4.2-3 of.

the FSAR was conducted after site erection to assure freedom from defects greater than code allowance; in addition, this serves as a reference base for further inspections. Prior to operation, the reactor primary system was free of gross defects. In addition, the facility has been designed such that gross defects should not occur i

1 4

Amendment No. 449, 142a W,., ,. -

.--e, -m , -- - . - - - - -- , , - - ~ . . , , - ,

VYNPS "

~

TABLE 4.7.2.a PRIMARY COffrAINMENT ISOLATION VALVES VALVES SUBJECT TO TYPE C LEAKAGE TESTS Number of Power Operated Valves ' Maximum Action on Isolation Operating Normal . Initiating Group (1) Valve Identification Inboard Outboard Time (sec) Position Sicnal 1 Main Steam Line Isolation (2-80A, D& 4 4 5 (Note 2) Open GC 2-86A, D) 1 Main Steam Line Drain (2-74, 2-77) . 1 1 35 Closed- SC 1 Recirculation Loop Sample Line (2-39, 2-40) 1 1 5 Closed SC l 2 RHR Discharge to Radwaste (10-57, 10-66) 2 25 Closed SC

'2 Drywell Floor Drain (20-82, 20-83) 2 20 Open .GC 2 Drywell Equipment Drain (20-94, 20-95)' 2 20 Open GC 3 Drywell Air Purge Inlet (16-19-9) 1 10 ' closed SC 3 Drywell Air Purge Inlet (16-19-8) 1 10 Open GC 3 Drywell Purge & Vent Outlet (16-19-7A) 1 10_ Closed

SC 3 Drywell Purge & Vent Outlet Bypass 1 10 Closed SC (16-19-6A) 3 Drywell & Suppression Chamber Main Exhaust 1 10 closed

SC (16-19-7)

~

3 Suppression Chamber Purge Supply (16-19-10) 1 10 Closed SC 3 Suppression Chamber Purge & Vent Outlet 1 10 Closed SC (16-19-7B) 3 Suppression Chamber Purge & Vent Outlet 1 10 Open GC Bypass (16-19-6B)

Valves 16-19-7 and 16-19-7A shall have stops installed to limit valve opening to 50* or less.

Amendment No. 50, &h 4-3+, 159 5

. : )

i l

VYNPS i BASES: 4.7 (Cont'd) i The maximum allowable test leak rate at the peak accident pressure of 44 psig (La) is 0.80 weight % per day. The maximum allowable test leak rate at the retest pressure of 24 psig (Lt) has been conservatively determined to be 0.59 weight percent per day. This j value was' verified to be conservative by actual primary containment leak rate measurements at both 44 psig and 24 psig upon completion of the containment structure.

To allow a margin for possible leakage deterioration between test j

' intervals, the maximum allowable operational leak rate (Ltm), which I will be met to remain on the normal test schedule, is 0.75 Lt.

As most leakage and deterioration of integrity is expected to occur through penetrations, especially those with resilient seals, a periodic leak

ate test program of such penetration is conducted at the peak aceit nt pressure of 44 psig to insure not only that the leakage remains acceptably low but also that the sealing net: rials can withstand the accident pressure.

The leak rate testing program is based on AEC guidelines for development of leak rate testing and surveillance schedules for reactor containment vessels.

Surveillance of the suppression Chamber-Reactor Building vacuum breakers consists of operability checks and leakage tests (conducted as part of the containment leak-tightness tests). These vacuum breakers are normally in the closed position and open only during -

tests or an accident condition. Operability testing is performed in l conjunction with Specification 4.6.E. Inspections and calibrations j are performed during the refueling outages; this frequency being based on equipment quality, experience, and engineering judgment.

The ten (10) drywell-suppression vacuum relief valves are designed to open to the full open position (the position that curtain area is equivalent to valve bore) with a force equivalent to a 0.5 psi-differential acting on the suppression chamber face of the valve disk. This opening specification assures that the design limit of 2.0 paid between the drywell and external environment is not exceeded. Once.each refueling outage each valve is tested to assure that it will open fully in response to a force less than that specified. Also it is inspected to assure that it closes freely and operates properly.

The containment design has been examined to establish the allowable bypass area between the drywell and suppression chamber as 0.12 ft 2, This is equivalent to one vacuum breaker open by three-eighths of an inch (3/8') as measured at all points around the circumference of the disk or three-fourths of an inch (3/4") as measured at the bottom of the disk when the top of the disk is on the seat. Since these valves open in a manner that is purely neither mode, a conservative allowance of one-half inch (1/2') has been selected as the maximum permissible valve opening. Assuming that permissible valve opening could be evenly divided among all ten vacuum breakers at once, valve open position assumed to indication for an individual valve must be activated less than fifty-thousandths of an inch (0.050") at all points along the seal surface of the disk. Valve closure within this limit may be determined by light indication from two independent position detection and indication systems. Either system provides a control room alarm for a nonseated valve.

Amendment No.- % , +24, 168

e

< h l

VYNPS BASES: 3.8 (Cont'd) .

1 liquid radwaste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I, 10CFR Part 50, for liquid effluents.

D. Liquid Holdup Tanks The tanks licted in this Specification include all outdoor tanks that contain radioactivity tha': are not surrounded by liners, dikes, or walls capable of holding che tank contents, or that do not have tank overflows and surroundinf. area drains connected to the liquid !

radwaste treatment systed.

Restricting the quantity ai radioactive material contaired in the specified tanks provides assurance that in the event of an l uncontrolled release of the F.nks' contents, the resulting l concentrations would be less than the limits of 10CFR Part 20,

-l Appendix B, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area.

E. Gaseous Effluents: ' Dose Rate This specification is provided to ensure that the dose at any time at and beyond the site boundary from gaseous effluents will be within the annual dose Ibnits of 10CFR Part 20. The annual dose limits are the doses associated with the concentrations of 10CFR Part 20, !

Appendix B,: Table II, Column 1. These limits provide reasonable l assurance that radioactive material discharged in gaseous effluents . '

l will not result in the exposure of member (s) of.the public either l

within or outside the site boundary, to annual average concentrations l

exceeding the limits specified in Appendix B, Table II of 10CFR _-

Part 20 [10 CFR Part 20.106(b)]. For member (s) of the public who may at times be within the site boundary, the occupancy of the individual will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the site boundary. The specified limits as determined by the methodology in the ODCM, restrict, at all times, the corresponding gamma and beta dose rates above background to a member of the public at or beyond the site boundary to -(500) mrem / year to the total body or to (3,000) mrem / year to the skin.

Specification 3.8.E.b also restricts, at all times, comparable with the length of the sampling periods of Table 4.8.2 the corresponding thyroid dose rate above background to an infant via the cow-milk-infant pathway to 1500 mrem / year for the nearest cow to the plant.

F. Gaseous Effluents: Dose from Noble Cases This specification is provided to implement the requirements of Sections II.B, III.A, and IV.A of Appendix I, 10CFR Part 50. The Limiting Condition for operation implements the guides set forth in Section II.B of Appendix I. The requirements provide operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that.the releases of radioactir-material in gaseous effluents will be kept 'as low as is reasonably achievable". The Surveillance Requirements implement the requirements in Section III.A of Appendix I, i.e., that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of any member of the public through appropriate pathways is unlikely to be substantially underestimated. The appropriate dose equations are

. Amendment No. 83, 186

~

VYNPS TABLE 3.9.1 RADIOACTIVE LIOUID EFFLUENT MONITORING INSTRUMENTATION Minimum Channels Operable Notes

1. Gross Radioactivity Monitors not Providing Automatic Termination of Release

a. Liquid Radwaste Discharge Monitor 1* 1,4,5 l (RD-17-330/PP-17-331/RM-17-350)

b. Service Water Discharge Monitor 1 2,4,5 l (RD-17-332/PP-17-333/RM-17-351)

2. Flow Rate Measurement Devices

a. 3' Liquid Radwaste Discharge Flow 1* 3,4 Rate Monitor l (FIT-20-485/442/FR-20-441)

'During releases via this pathway.

Amendment No, M , 193

. .) l

-i VYNPS.

. TABLE 3.9.1 NOTATION NOTE 1.- With the number-of channels operable les,a than required by the' . :

minimum channels operable requirement, effluent releases may continue ;

provided that prior to initiating a release:

[ a. 'At'least two independent samples are analyzed in accordance with !

Specification 4.8.A.1, and ,

b. At least two. technically qualified members of the Facility Staff j independently verify the release rate calculations and discharge line valving. i

{

.otherwise, suspend release of radioactive effluents'via this pathway. ;

i NOTE 2 - With the number of channels operable l'ess than required by the minimum channels operable requirement, effluent releases via this :

pathway may continue provided that, at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , grab !

samples are collected and analyzed for gross radioactivity (beta or '{

gamma) at'a lower-limit of detection of at least 10-7 microcurie /ml. i NOTE 3 - With .the number of cw.nels operable less than required by the minimum channels opercele requirement, effluent releases via this :

pathway may continua provided the flow rate is estimated at least !

.* once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases. Pump. performance curves may i be used to estimate flow. ';

NOTE 4 - With the number of' channels operable less than required by the. +

minimum channels operable requirement, exert reasonable efforts to .

return the instrument (s) to operable status prior to the next release. 1 NOTE 5 - The alarm setpoints of these channels shall be determined'andf ;

adjusted in accordance with the methodology'and parameters in the off-Site Dose Calculation Manual (oDCM). With a radioactive liquid- 1 effluent monitoring instrumentation channel alarm setpoint less conservative than a value which will ensure that the limits of

'l 3.8.A.1 are met during periods of release, immediately take action to suspend the release of radioactive liquid effluents monitored by the affected channel or declare the channel inoperable; or chan'ge'the setpoint so it is' acceptably conservative.

I 1

Amendment No. B&, 194 1

9 VYNPS TABLE 3.9.2 GASEOUS EFFLUENT MONITORING INSTRUMENTATION Minimum Channels Instrument Operable Notes

1. Steam Jet Air Ejector (SJAE)

a. Noble Gas Activity Monitor 1 7, 8, 9 (RD-17-130A/RM-17-150A, RD-17-130B/RM-17-150B) !

2. Augmented Off-Gas System

a. Noble Gas Activity Monitor Between 1 2, 5,6,7 the Charcoal Ped System and the Plant Stack (Providing Alarm and Automatic Termination of Release)

(RE OG-3107/RAN OG-3127, RE OG-3108/RAN OG-3128)

b. Flow Rate Monitor 1 1, 5, 6 (FE OG-1802/FT OG-1902/FI-OG-2002, FE OG-1804/FT OG-1904/FI-OG-2004, FE OG-1805/FT OG-1905/FI-OG-2008)

c. Hydrogen Monitor 1 3, 5, 6 ;

(H2E OG-2901A/H2AN OG-2921A, l H2E OG-2901B/H2AN OG-2921B, '

H2E OG-2902A/H2AN OG-2922A, H2E OG-2902B/H2AN OG-2922B)

3. Plant Stack

a. Noble Gas Activity Monitor 1 5, 7, 10 l (RD/RM-17-156, RD/RM-17-157)

b. Iodine Sampler Cartridge 1 4, 5

c. Particulate Sampler Filter 1 4, 5

d. Sampler Flow Integrator 1 1, 5 l (FI-17-156/157) ,

e. Stack Flow Rate Monitor 1 1, 5 )

l (FE-108-22A&B/FA-108-22/FI-108-22) 1 l

Amendment No. 64, MG, 195

i I

VYNPS .

. TABLE 3.9.3 (Cont'd) ,

RADIOLOGICAL ENVIRONME!!TAL MONITORING PROGRAM Exposure Pathway Number of Sample Sampling and Collection Type and Frequency.

and/or Sample Locations

Frequency of Analysis

2. # DIRECT-RADIATION b 40 routine monitoring Quarterly. Gamma dose, at least once stations as follows: per quarter.

16 incident response '

Incident response TLDs in stations (one in each the outer monitoring l meteorological sector) locations, de-dose only within a range of 0 to quarterly unless gaseous 4 kmS; release LCO was exceeded in period.

16 incident response stations (one in each l meteorological sector) within a range c-f 2 to 8 km9s the balance of the stations to be placed in special interest areas and control station areas.

Amendment No. 83, 198 i

-VYNPS TABLE 3.9.4 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLESI *I Reporting Levels Airborne Particulate or Fish Milk Vegetation Sediment Analysis Water (pCi/1) Cases (pCi/m3) (pci/Kg,-wet) (pCi/l) (pCi/Kg, wet) (pCi/Kg, dry)-

H-3 2 x 104N Mn-54 1 x 10 3 3 x 10 4 Fe-59 4 x 10 2 1 x.10 4 Co-58 1 x 10 3 3 x 10 4 Co-60 3 x 102 1 x 10 4 3 x 10 (C) 3 2n-65 3 x 10 2 2 x 104 Zr-Nb-95 4 x 10 2 I-131 0.9 3 1 x 10 2 Cs-134 30 -10 1 x 10 3 60 'l x 10 3 Cs-137 50 20 2 x 103 70 2 x 103 Ba-La-140 2 x 10 2 3 x 102 (a) Reporting levels may be averaged over a calendar quarter. When more than one of the radionuclides in Table 3.9.4 are detected in the sampling medium, the unique reporting requirements are not exercised if the following condition holds:

concentration (1) , concentration (2) + ... < 1.0.

~

reporting level (1) reporting level (2)

When radionuclides other than those in Table 3.9.4 are detected and are the result of plant effluents, the potential annual dose to a member of the public must be less than or equal to the calendar year limits of Specifications 3.8.B, 3.8.E and 3.8.F.

(b) Reporting level for drinking water pathways. For nondrinking water pathways, a value of 3 x 10 pCi/1

~

8 may'be' used.

(c) Reporting level for individual grab samples taken at North Storm Drain Outfall only.

Amendment No. &&, 444, .

202

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

t i'

VYNPS TABLE 4.9.2 NOTATION (1) The Instrument Functional Test shall also demonstrate that automatic ,

isolation of this pathway and the Control Room alarm annunciation occurs if-any of the following conditions existsi j l (a)' Instrument indicates measured levels above the alarm setpoine. '

(b) Circuit. failure.

o (c) Instrument indicates a'downscale failure. !

(d) Instrument controls not set in operate mode. !

(2) The Instrument Functional Test shall also demonstrate that Control Room .

alarm annunciation occurs when any of the following conditions exist:

(a)-Instrument indicates measured levels;above the alarm setpoint. .;

-i

-(b) Circuit-failure, j

(c)= Instrument indice.tes a downscale failure. I (d) . Instrument controls are not set in operate mode.

(3) The Instrument Calibration for radioactivity measurement' instrumentation shall include the use of a known (traceable to National Bureau of ,

Standards) radioactive source positioned in a reproducible geometry with , {

respect to the sensor. These standards should permit calibrating the }

system over its normal operating range of rate capabilities, j (4) The Instrument Calibration shall include the use of standard gas samples

-]

(high range and low range) containing suitable concentrations,. hydrogen l balance air, for the detection range of interest per Specification 3.8.J.l.

}

Amendment.No. en , 206

j VYNPS l BASES:

3.9 RADIOACTIVE EFFLUENT MONITORING SYSTEMS A. Licuid Effluent Instrumentation The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive j

! materials in liquid effluents during actual or potential releases of i liquid effluents. The alarm setpoints for these instruments are to I ensure that the alarm will occur prior to exceeding the limits of 10CFR Part 20. i Automatic isolation function is not provided on the liquid radwaste discharge line due to the infrequent nature of batch, discrete volume, liquid discharges (on the order of once per year or less),

and the administrative controls provided to ensure that conservative discharge flow rates / dilution flows are set such that the probability i of exceeding the 10CFR Part 20 concentration limits are low, and the !

potential off-site dose consequences are also low. I B. Gaseous Effluent Instrumentation The radioactive gaseous effluent instrumentation is provided to !

monitor and control, as applicable, the releases of radioactive i materials in gaseous effluents during actual or potential releases of j gaseous effluents. The alarm / trip setpoints for these instruments l are provided to ensure that the alarm / trip will occur prior to j exceeding the limits of 10CFR Part 20. This instrumentation also * '

includes provisions for monitoring (and controlling) the concentrations of potentially explosive gas mixtures in tha waste gas holdup system. -

C. Radiolocical Environmental Monitorino procram The radiological monitoring program required by this specification provides measurements of radiation and of radioactive materials in those exposure pa+hways and for those radionuclides which lead to the

, highest potential radiation exposures of member (s) of the public ,

resulting from the station operation. This monitoring program )

implementsSection IV.B.2 of Appendix I to 10 CFR Part 50 and thereby supplements the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials and j levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of the environmental exposure pathways.

Ten years of plant operation, including the years prior to the implementation of the Augmented Off-Gas System, have amply demonstrated via routine effluent and environmental reports that plant effluent measurements and modeling of environmental pathways are adequately conservative. In all cases, environmental sample results have been two to three orders of magnitude less than expected by the model employed, thereby representing small percentages of the AIlRA and environmental reporting levels. This radiological environmental monitoring program has therefore been significantly l modified as provided for by Regulatory Guide 4.1 (C.2.b), Revision 1, April 1975. Specifically, the air particulate and radiciodine air sampling periods have been increased to semimonthly, based on plant effluent and environmental air sampling data for the previous ten years of operation. An 1-131 release rate trigger value of 1 x 10'1 uCi/sec from the plant stack will require that air sample collection be increased to weekly. The Amendment No. 64, 209 l

e .

T;

.)

VYNPS 3.12 LIMITING CONDITIONS FOR 4.12 SURVEILLANCE REQUIREMENTS OPEFATION I

E. Extended' core Maintenance E. Extended Core Maintenance l More than two control rods Prior-to control rod may be withdrawn from the withdrawal for extended core, i reactor core provided the maintenance, that control following: conditions are rod's control cell _shall'be l satisfied:- verified to contain no fuel assemblies, j

1. The reactor mode switch 1. This surveillance j shall be locked in the requirement is the same

_" Refuel" position. The as that given in

' refueling interlock .

Specification 4.12.A.

which prevents more than one control rod from I being withdrawn may be )

bypassed on a vithdrawn '

i control rod after the _l fuel assemblies in the '

cell containing '

(controlled by).that !

control rod have been !

removed from the reactor '

l core.. All other refueling interlocks I shall be operable. -

2. SRMs shall.be operable in the core quadrant _2. _This surveillance ,

where fuel or control- requirement is the same rods are being moved, as that given in and in an adjacent Specification 4.12.B. l quadrant. The J requirements'for an SRM. 1 to be considered operable are given in- )

Specification 3.12.B. .)

3. If the spiral unload / reload method of l core alteration is_to be J used, the following 1 conditions shall be mets

a. Prior to spiral unload and reload, the SRMs shall be proven operable as stated in Specification 3.12.B1 and 3.12.B2. However, during spiral unloading, the count rate may drop below 3 cps.

Amendment No. M, W,' M, 233

- - . _ . . _ , - .-- - -- . . . _ - . .~ .

. e VYNPS BASES:

3.12 & 4.12 REFUELING A. During refueling operations, the reactivity potential of the core is being altered. It is necessary to require certain interlocks and restrict certain refueling procedures such that there is assurance that inadvertent criticality does not occur.

To minimize the possibility of loading fuel into a cell containing no control. rod, it is required that all control rods are fully inserted when fuel is being loaded into the reactor core. This requirement assures that during refueling the refueling interlocks, as designed, will prevent inadvertent criticality. The core reactivity lbmitation l of Specification 3.3 limits the core alterations to assur< that the resulting core loading can be. controlled with the Reactivity Control System and interlocks at any time during shutdown or the following operating cycle.

The addition of large amounts of reactivity to the core is prevented r by operating. procedures, which are in turn backed up by refueling interlocks on rod withdrawal and movement of the refueling platform.

When the mode switch is in the " Refuel" position, interlocks prevent the refueling platform from being moved over the core if a control rod is withdrawn and fuel is on a hoist.

Likewise, if the refueling platform is over the core with fuel on a hoist, control rod motion is blocked by the interlocks. With the mode switch in the refuel position, only one control rod can be -

withdrawn.

B. The SRMs are provided to monitor the core during periods of station shutdown and to guide the operator during refueling operations and

, station startup. Requiring two operable SRMs in or adjacent to any core quadrant where fuel or control rods are being moved assures !

adequate monitoring of that quadrant during such alterations. The l requirement of 3 counts per second provides assurance that neutron l flux is being monitored. Under the special condition of complete spiral core unloading, it is expected that the count rate of the SRMs will drop below 3 cps before all the fuel is unloaded. Since there will be no reactivity additions, a lower number of counts will not present a hazard. When all of the fuel has been removed to the spent fuel storage pool, the SRMs will no longer be required. Requiring the SRMs to be operational prior to fuel removal assures that the SRMs are operable and can be relied on even when the count rate may go below 3 eps.

Prior to spiral reload, two diagonally adjacent fuel assemblies, i which have previously accumulated exposure in the reactor, will be l loaded into their designated core positions next to each of the i 4 SRMs to obtain the required 3 cps. Exposed fuel continuously l produces neutrons by spontaneous fission of certain plutonium isotopes, photo fission, and photo disintegration of deuterium in the ;

moderator. This neutron production is normally great enough to meet the 3 cps minbswn SRM requirement, thereby providing a means by which SRM response may be demonstrated before~the spiral reload begins. 1 During the spiral reload, the fuel will be loaded in the reverse sequence that it was unloaded with the exception of the initial eight (8) fuel assemblies which are loaded next to the SRMs to provide a means of SRM response, j Amendment No. 16, 59, 44, 237

VYNPS -

3.13 LIMITING CONDITIONS FOR 4.13 SURVEILLANCE REQUIREMENTS OPERATION

1) The batteries, cell plates

and battery racks show no visual indication of physical s damage or abnormal deterioration

, and

2) The battery-to-battery and terminal connections i are clean, i tight,' free 1 of corrosion and coated ,

with l anti-corrosio l C. Fire Hose Stations ~ n material. i

1. Except as specified in C. Fire Hose Stations -

l 3.13.C.2 below, all-hose stations inside the 1. 'Each fire hose stacion Reactor Building, shall be verified to be Turbine Building, and operable those inside the Administration Building a. At least monthly by which provided coverage . visual inspection of the Control Room of the station to Building shall be assure all operable whenever equipment is equipment in the areas available.

protected by the . fire hose stations is b. At least once each required to be operable. 18 months by removing the hose

2. With one or more of the for inspection and fire hose stations replacing degraded specified in 3.13.C.1 coupling gaskets above inoperable, route and reracking.

an additional equivalent capacity fire hose to c. At least once each the unprotected area (s) year by from an operable hose hydro-statically station within one hour, testing each outside hose at 250 lbs.

d. At least once per 3 years by hydro-statically testing inside hose at 150 lbs.

Amendment No. 44, W,' 244

. t

-VYNPS i

3.13 LIMITING CONDITIONS..*OR 4.13 SURVEILLANCE REQUIREMENTS OPERATION

e. At least once per

--3 years,' partially open hose station valves to verify valve operability and no blockage.

l D. 'Hich Pressure CO, Systems 'D . Hich Pressure ' CO, Systems -l.

1. Except as specified in 1. The CO2 systems located l Specification 3.13.D.2, in the cable vault,.-

the CO2 systems located switchgear rooms, and l in-the cable vault, diesel fire pump day l- switchgear rooms, and . tank room shall be >

diesel fire pump day demonstrated operable.

tank room shall be operable, whenever a. At least once per i equipment in the area- six. months by verifying each CO2 protected by the system is required to be cylinder does not operable. contain less than 90% of its. initial

2. From and after the date charge.

that the CO2 system in the cable vault or a b. At least once per switchgear room is 18 months by .

inoperable, within'one: ' verifying that the ,

hour a fire watch shall system, including

be established to associated-

inspect the location at ventilation least once every hour, dampers, will provided that the fire actuate detection system is automatically to a operable in accordance simulated actuation-with 3.13.A. If the . signal.

fire detection system is also inoperable, within c. At least once per one hour a continuous operating cycle a fire watch shall be flow path test i established with backup shall be performed fire suppression- .

to verify flow equipment. Restore the through each CO2 system to operable nozzle.

status within 14 days or submit a report within

.the next 30 days to the Commission as specified in 6.7.C.2 outlining the cause of inoperability and the plans for restoring the CO2 system

., to operable status.

_ Amendment No. 44, f4, M4, 245 5

- ~ . ,

~ . . _

WNPS -

3.13 LIMITING CONDITIONS FOR 4.13 SURVEILLANCE REQUIREMENTS-OPFRATION

c. At least once per l

3. years by performing an air-flow test through the Recirculation- +

M.G. Set foam-header and verifying each foam-nozzle is unobstructed.

i I

l l

l i

1 l

l 1

i i

l

' Amendment No. W , ,

249 ;

,\

. - ~ - - - - - ~ . . . . _-._ - ---. - . . ~ . = , - . . _ . . -- - . . . _ _ -

VYNPS '

TABLE 3.13.A.1 FIRE DETECTION' SENSORS I

i Minimum No. of Sensors' 4 Recuired to Be Ooerable' !

Sensor Location Heat ' Flame Smoke

'1. Cable. Spreading. Room & Station Battery Room ~ - -

23: '

-l =2. Switchgear Room-(East) - -

10 l '3. 'Switchgear Room (West). - -

10 Diesel Generator Room-(A)

-l ~4. - -

~2 (

.l 5. Diese1LGenerator Room (B) - -

2 j

l 6. Intake Structure (Service Water) 1 1 1 :

L i

.l 17. Recirc Motor Generator Set. Area 3 <

8 -

l 8.a .-Control Room Zone 1 (Control. Room Ceiling) - -

14 !

.l - 8.b: Control Room Zone'2 (Control Room. Panels) - -

18-i l 8.c ~~ Control Room Zone 3 (Control Room Panels). - -

25

.l 8.d. Control Room Zone 4 (Control Room Panels) - -

104 .

f l 8.e Control Room Zone 5 (Exhaust & Supply. Ducts) - -

2

{

-l 9.a Rx Bldg. Corner Rm NW 232 -

~1 )

l :9.b Rx Bldg.' Corner Rm NW 213 (RCIC) - -

1 (

l 9.c Rx Bldg. Corner Rm NE 232 - -

1 ;

j ~9.d Rx Bldg. Corner Rm NE 213 - -

1 l 9.e 'Rx Bldg. Corner Rm SE 232 - -

1 l

' j ' . 9.f Rx Bldg. ' Corner Rm SE 213 - -

1 ]

l. 9.g . Rx Bldg. Corner Rm SW 232 1 l

l 10. HPCI Room - ' -

8 l' 11. Torus area 12 -

16 l' 12. -Rx Bldg. Cable Penetration Area - - 7 l 13. Refuel Floor - -

13

'l 14. Diesel Oil Day Tank Room (A) -

1* 1*

l 15. Diesel Oil Day Tank Room (B) -

1* 1* j l 16. Turbine Loading Bay (vehicles) -

3 -

)

i

NOTE: The Diesel' Day Tank Rooms require only one detector operable (1 flame !

or 1 smoke).

Amendment No. 43, G@, 250

l VYNPS 6.0 ADMINISTRATIVE CONTROLS I l

l Administrative controls are the written rules, orders, instructions, procedures, policies, practices, and the designation of authorities and ]

responsibilities by the management to obtain assurance of safety and l quality of operation and maintenance of a nuclear power reactor. These controls shall be adhered to.

1 6.1 ORGANIZATION l I

A. Lines of authority, responsibility, and communication shall ]

be established and defined for the highest management levels l through intermediate levels to and including all operating 1 organizational positions. These relationships shall be !

documented and updated, as appropriate, in the form of '

organizational charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be documented in the Yankee Operational Quality Assurance Manual. ;

B. The Plant Manager shall be responsible for overall unit safe operation and shall have control over those on-site activities necessary for safe operation and maintenance of the plant. Succession to this responsibility during his absence shall be delegated in writing.

C. The Manager of Operations shall have corporate responsibility a for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety.

D. Conduct of operations of the plant will be in accordance with the following minimum conditions.

1. An individual qualified in radiation protection procedures shall be present on-site at all times when there is fuel in the' reactor.

2. Minimum shift staffing on-site shall be in accordance with Table 6.1.1.

3. A dedicated, licensed. Senior Operator shall be in charge of any reactor core alteration.

4. Qualifications with regard to educational background experience, and technical specialties of the key supervisory personnel listed below shall apply and be maintained in accordance with the levels described in the American National Standards Institute N18.1-1971,

" Selection and Training of Personnel for Nuclear Power Plants".

a. Plant Manager

b. Superintendent (s)

c. Chemistry Manager

d. Radiation Protection Manager

e. Operations Manager (See Item 6.1.D.7)

f. Reactor Engineering Manager

g. Maintenance Manager

h. Instrument and Control Manager

i. Shift Supervisors Amendment No. M, M, M, M, M, m, 255

.. . _ _ _ _ . _ . - . _ _ . . . . _ _ . . _.m _

o .. ,

I

-VYNPS I i

6.2 REVIEW AND' AUDIT

~

organizational' units:for the review and audit of plant operations j shall be constituted and have.the responsibilities and. authorities i outlined below:

'A. Plant Operations Review Committee

1. Membership T

a. " Chairman: Plant Manager f

b. Vice-Chairman: Superintendent (s). l

c. Engineering Manager

d. Operations Manager I

e. Maintenance Manager _

t

f. Reactor Engineering Manager i

g. Chemistry Manager .i

h. Instrument and Control Manager ;

i. Radiation Protection Manager

2. Qualifications .i

~

The qualifications of the regular members.of the Plant: ;

Operations Review Committee with regard to the combined.

experience and technical specialties of the individual.

members shall be maintained at a level at least equal to i or higher than as described in Specification 6.1. ?

3. Meetina'Frecuenev: Monthly, and as required,:on call of

~ '

the Chairman.

4. Quorum: Chairman or Vice-Chairman:plus four members or

'their designated alternates.

NOTE: For purposes of satisfying a quorum, a Vice-Chairman may be considered a member providing that Vice-Chairman is not presiding.

over the meeting.

5. Designated alternates shall be from other plant ]

personnel in the appropriate disciplines,or as selected i by the Plant Manager; however, there shall be no more- -!

than three (3) alternates serving on the committee at i any one time.

6. Responsibilities

a. Review proposed normal, abnormal, and emergency j operating procedures. Review all proposed ]

maintenance procedures and proposed changes to those procedures; and any other proposed procedures or changes thereto as determined by the Plant Manager to affect nuclear safety.

b. Review proposed tests and experiments.

c. Review proposed changes to Technical Specifications.

d. Review proposed chan'ges or modifications to plant I systems or equipment, which changes would require a ;

change in procedures in (a) above. I

e. Review plant operations to detect any potential safety hazards.

Amendment.No. M , M , 4M , 4 M ,. 258

o e VYNPS

f. -The Committee membership and-its Chairman and Vice Chairman shall be appointed as specified in the Yankee Quality Assurance Manual.

2. Qualifications The Committee shall consist of a minimum of six (6)

'l -members plus designated alternates who as a group employ expertise in the following areas:

a. Nuclear Power Plant Technology i

b .' Reactor Operations

c. . Utility Operations

d. Power Plant Design

e. Reactor Engineering

f. Radiation Safety

g. Safety Analysis'

h. Instrumentation and Control

i. Metallurgy

3. Meetino Frequency: .Sami-annually and as required on call of the Chairman.

Ouorum: Chairman or Vice Chainnan plus four members or ;

4.

, designated alternates.

5. Responsibilities: j

a. Review proposed changes to the operating license including Technical Specifications. .

b. Review minutes of meetings of the Plant Operation Review Committee to determine if matters considered by that committee involve unreviewed or unresolved safety questions.

c. Review the safety evaluations for changes to equipment or systems completed under the provisions of Section 50.59 10 CFR to verify that such actions ,

did not constitute an unreviewed safety question.

d. Periodic audits of implementing procedures, shall be performed under cognizance of the Committee.

Included in these audits, but not limited to, are the following specific activities:

1. plant operations; ii. facility fire protection program; iii. the radiological environmental monitoring i program and the results thereof at least once per 12 months; iv, the off-Site Dose Calculation Manual and implementing procedures at least once per 24 months; Amendment No. G, - 66, M, M, 4M, 260

O e ,

VYNPS

10. Records for Environmental Qualification which are covered under the provisions of paragraph 6.9.

11. Records of analysis required by the Radiological Environmental Monitoring Program. ,

6.7 REPORTING REOUTREMENTS In addition to the applicable reporting requirements of Title 10' .

Code of Federal Regulations, the following identified reports '

l shall be submitted to the appropriate Regional Office unless otherwise noted. $

A. Routine Reports

1. Startup Report i

A summary report of plant startup and power escalation '

testing shall be submitted following (1) receipt of an operating license, (2) amendment to the license l involving a planned increase in power level, j (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and '

(4) modifications that may have significantly altered _

the nuclear, thermal or hydraulic performance of the '

plant. The report shall address each of the tests i identified in the FSAR and shall, in general, include a !

description of the measured values of the operating J conditions or characteristics obtained during the test -

program and a comparison of these values with design I predictions and specifications. Any corrective actions j that were required to obtain satisfactory operation i shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report.

Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption of commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the j startup report does not cover all three events (i.e.,

initial criticality, completion of startup test program, and resumption or commencement of commercial power operation), supplementary reports shall be submitted at least every three months until all three events have been completed.

2. Annual Report I 1

An annual report covering the previous calendar year shall be submitted prior to March 1 of each year. The annual report shall include a tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated man-rem exposure according to work and job functions, 1/ e.g.,

reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling.

, 1/ This tabulation supplements the requirements of 20.407 of 10CFR Part 20.

Amendment No. 4B, &&, 266

0 e VYNPS

10. Records for Environmental Qualification which are covered under the provisions of paragraph 6.9.

11. Records of analysis required by the Radiological Environmental. Monitoring Program.

6.7 REPORTING REOUIREMEtrrS In addition to the applicable reporting requirements of Title 10 Code of Federal Regulations, the following identified reports l

shall be submitted to the appropriate Regional Office unless otherwise noted. i A. Routine Reoorts

1. Startup Reoort A summary report of plant startup aid power escalation testing shall be submitted following gl) receipt of an ,

operating license, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal or hydraulic performance of the plant. The report shall address each of the tests identified in the FSAR and shall, in general, include a .)

description of the measured values of the operating conditions or characteristics obtained during the test :

program and a comparison of these values with design i predictions and specifications. Any corrective actions l that were required to obtain satisfactory operation ,

shall also be described. Any additional specific !

details required in license conditions based on other 1 commitments shall be included in this report. _l Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption of commencement of commercial power operation, or (3) 9 months following ;

l initial criticality, whichever is earliest. If the startup report does not cover all three events (i.e.,

initial criticality, completion of startup test program, and resumption or commencement of commercial power operation), supplementary reports shall be submitted at least every three months until all three events have been completed.

2. Annual Report i An annual report covering the previous calendar year shall be submitted prior to March 1 of each year. The annual report shall include a tabulation on an annual ,

basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated man-rem exposure according to work and job functions, 1/ e.g.,

reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling.

1/ This tabulation supplements the requirements of 20.407 of 10CFR Part 20.

Amendment No. 44, G&, 266

r WNPS Letter'from L. A. Tremblay, Jr. (WNPC) to USNRC,

" Supplemental Information to WNPC April 19, 1990 Response Regarding FROSSTEY-2 Fuel Performance Code,*

BW 90-054, dated May 10, 1990 (Approved by NRC SER, dated September 24, 1992).

Letter from L. A'. Tremblay, Jr. (WNPC) t o USNRC,

" Responses to Request for Additional Information on FROSSTEY-2 Fuel Performance Code,

BW 91-024, dated March 6, 1991 (Approved by NRC SER, dated September 24, 1992).

i Letter from L. A. Tremblay, Jr. (WNPC) to USNRC, "LOCA-Related Responses to Open-Issues on FROSSTEY-2 Fuel Perfomance Code,

BW 92-39, dated March 27, 1992 (Approved by NRC SER, dated September 24, 1992).

l Letter from L. A. Tremblay, Jr. (WNPC) to USNRC, "FROSSTEY-2 Fuel Performance Code - Vermont Yankee Response to Remaining Concerns,

BW 92-54, dated May.

15, 1992 (Approved by NRC SER, dated September 24, 1992).

Report,

Loss-of-Coolant Accident Analysis for Vermont Yankee Nuclear Power Station," NEDO-21697, August 1977, as amended (Approved by NRC SER, dated November 30, 1977).

Report, " General Electric Standard Application for~

Reactor Fuel (GESTARII),* NEDE-240ll-P-A, GE Company Proprietary (the latest NRC-approved version.will be listed in the COLR).

The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical-limits, core thermal-hydraulic limits, ECCS limits, ,

nuclear limits such as shutdown margin, and transient !

and accident analysis limits).of the safety analysis are '

met. The C(.J R, including any mid-cycle revisions or supplements thereto/ shall be provided upon issuance, .

for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident !

Inspector.

B. Reportable Occurrences ,

i This section deleted. l C. Unique Reportino Requirements ,

i

1. Annual Radioactive Effluent Release Report '

a. Within 90 days after January 1 of each year, a report shall be submitted covering the radioactive content of effluents released to unrestricted areas during the previous calendar year of operation.

)

1 l

)

I l

I

~ Amendment'No. M , 83, 95, 4%, 4M, 4M, 444, 4M, 270 !

pay ,

VYNPS 6.13 OFF-SITE DOSE CALCULATION MANUAL (ODCM)

An Off-Site Dose Calculation Manual shall contain the current methodology and parameters used in the calculation of off-site doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints, and in the conduct of the environmerital radiological monitoring program.

A. Licensee initiated changes to the ODCM:

1. Shall be submitted to the Commission in the Annual Radioactive Effluent Release Report for the period in which the change (s) was made effective. This submittal shall contain:

a. Sufficiently detailed information to support the rationale for the change without benefit of additional or supplemental information.

Information submitted should consist of a package of those pages of the ODCM which were changed with each page numbered and provided with the revision number, together with appropriate analyses or evaluations ju.-tifying the change (s).

b. A determination that the change will not reduce the accuracy or reliability of dose ~

calculations or setpoint determinations.

c. Documentation of the fact that the change has been reviewed by PORC and approved by the Manager of Operations (MOO).

2. Shall become effective upon review by PORC and approved by the Manager of Operations (MOO).

6.14 MAJOR CHANGES TO RADIOACTIVE LIOUID, GASEOUS, AND SOLID WASTE TREATMENT SYSTEMS

Licensee-initiated major changes to the radioactive waste systems (liquid, gaseous, and solid):

A. Shall be reported to the Commission in the Annual Radioactive Effluent Release Report for the period in which the evaluation was reviewed by the PORC. The discussion of each change shall contain:

1. A summary of the evaluation that led to the determination that the change could be made in accordance with 10CFR Part 50.59;

2. Sufficient detailed information to support the reason for the change without benefit of additicnal or supplemental information;

3. A detailed description of the equipment, components, and processes involved and the interfaces with other plant systems;

Licensee may choose to submit the information called for in this l Specification as part of the periodic FSAR update.

Amendment No. 64, 96, 404, 444, 278

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ML20097C049

Text

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