Text

Reformatted TSs Current Through Amend 137
	ML20070K418
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	07/19/1994
From:	; VERMONT YANKEE NUCLEAR POWER CORP.
To:	;
Shared Package
ML20070K411	List: ML20070K408; ML20070K418;
References
	NUDOCS 9407260253
	Download: ML20070K418 (298)
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i ENCLOSURE A Vermont Yankee Reformatted Technical Specifications current through Amendment 137 f

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9407260253 940719 PDR ADOCK 05000271 P PDR

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APPENDIX A ,

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.19.

OPERATING LICENSE DPR-28 TECHNICAL SPECIFICATIONS i

AND BASES ,

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VEPRONT Y;01KEE NUCLEAR POWER STATIOli ,

VERNON, VEPRONT ,

VERMONT YAtVEE NUCLEAR POWER CORPORATIOf, ,

DOCKFT NO. 50-271 1

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Reissued by Change Nos. 13, 15, and 17 Dated 1/17/74, 1/28/74, and 4/10/74

i VYNPS TABLE OF CONTENTS GENERAL Pace No.

1.0 DEFINITIONS.................................................... 1 5.0 DESIGN FEATURES................................................ 253 6.0 ADMINISTRATIVE CONTROLS........................................ 255 ;

6.1 ORGANIZATION........................ .......................... 255 i 6.2 REVIEW AND AUDIT............................................... 258 A. Plant Operation Review Committee........................... 258 .

B. Nuclear Safety Audit and Review Committee.............. ... 259 ,

6.3 ACTION TO BE TAKEN IN THE EVENT OF AN ABNORMAL OCCURRENCE IN PLANT OPERATION.................................. 261 6.4 ACTION TO BE TAKEN IF A SAFETY LIMIT IS EXCEEDED............... 262 ,

6.5 PLANT OPERATING PROCEDURES.................. .................. 262 6.6 PLANT OPERATING RECORDS........................................ 265 6.7 PLANT REPORTING REQUIRD4ENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 266 I 6.8 FIRE PROTECTION INSPECTION..................................... 276 6.9 D NIRONMENTAL QUALIFICATION.................................... 276 6.10 INTEGRITY OF SYSTEMS OUTSIDE CONTAINMDJT. . . . . . . . . . . . . . . .

...... 277 6.11 IODINE MONITORINC.............................................. 277 [

6.12 PROCESS CONTROL PROGRAM (PCP).................................. 277 6.13 OFF-SITE DOSE CALCULATION MANUAL (ODCM)........................ 278 l 6.14 MAJOR CHANGES TO RADIOACTIVE LIQUID, GASEOUS, AND SOLID WASTE TREATMENT SYSTEMS... .................................... 278 ,

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f VYNPS TABLE OF CONTENTS (Continued) l LIMITING SAFETY Pace Na SYSTEM SETTING SAFETY I.IMITS 1.1 FUEL CLADDING INTEGRITY....................... 6 ... 2.1 1.2 REACTOR COOLANT SYSTEM........................ 18 ... 2.2 Q ' TING CONDITIONS OF OPERATION Pace No. SURVEILLANCE 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 Instrumentation............. 35 ... G ,

H. Drywell to Torus AP Instrumentation........ 36 ... H I. Recirculation Pump Trip Instrumentation........................... 36 ... I l BASES 75 l

1 3.3 CONTROL ROD SYSTEM............................ 81 ... 4.3 A. Reactivity Limitations.................... 81 ... A B. Control Rods.............................. 82 ... B 1 C. Scram Insertion Times..................... 85 ... C !

D. Control Rod Accumulators.................. 87 ... D E. Reactivity Anomalies...................... 88 ... E l BASES 89 i

3.4 REACTOR STANDBY LIQUID CONTROL SYSTEM......... 92 ... 4.4 i A. Normal Operation.......................... 92 ... A !

B. Operation with Inoperable Components...... 93 ... B l C. Liquid Poison Tank - Boron Concentration............................. 93 ... C j l BASES 97 l t

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l Amendment No. G, 95 -ii-i...-, -

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VYNPS I

TABLE OF CONTENTS (Continued)

LIMITING CONDITIONS OF OPERATION ParJe No . SURVEILLANCE 3.5 CORE AND CONTAINMENT COOLING SYSTEMS.......... 99 ... 4.5 A. Core Spray and Low Pressure Coolant Injection................................. 99 ... A B. Containment Spray Cooling Capability...... 102 ... B C. Residual Heat Removal (RHR) Service Water System.............................. 103 ... C D. Station Service Water and Alternate Cooling Tower Systems .................... 104 ... D E. High Pressure Coolant Injection (HPCI)

System.................................... 105 ... E F. Automatic Depressurization System......... 106 ... F G. Reactor Core Isolation Cooling System (RCIC)..................................... 107 ... G H. Minimum Core and Containment Cooling System Availability....................... 108 ... H I. Maintenance of Filled Discharge Pipe...... 109 ... I 1

110 l BASES 3.6 REACTOR COOLANT SYSTEM........................ 115 . . . . 4.6 A. Pressure and Temperature Limitations...... 115 ... A B. Coolant Chemistry......................... 116 ... B C. Coolant Leakage........................... 119 ... C D. Safety and Relief Valves.................. 120 ... D E. Structural Integrity...................... 120 ... E F. Jet Pumps................................. 121 ... F ,

G. Single Loop Operation..................... 122 ...

H. Recirculation System'...................... 126 ...

I. Shock Suppressors......................... 128 ... I J. Thermal Hydraulic Stability............... 134 ... J l BASES 139 3.7 STATION CONTAINMENT SYSTD4S . . . . . . . . . . . . . . . . . . . 146 ... 4.7 A. Primary Containment....................... 146 ... A B. Standby Gas Treatment..................... 152 ... B C. Secondary Containn ent Sys tem. . . . . . . . . . . . . . 155 ... C D. Primary containment Isolation valves...... 156 ... D ,

l BASES 163 3.8 RADIOACTIVE EFFLUENTS......................... 172 ... 4.8 A. Liquid Effluentst concentration.......... 172 ... A l B. Liquid Effluents: Dose................... 173 ... B C. Liquid Radwaste Treatment................. 173 ... C i' D. Liquid Holdup Tanks....................... 174 ... D E. Gaseous Effluents: Dose Rate............. 174 ... E l

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l Amendment No. 43, 64, B3, 95 -iii- l l

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VYNPS TABLE OF CONTENTS ,

(Continued)

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LIMITING CONDITIONS OF OPERATION Pace No. SURVEILLANCE F. Gaseous Effluents: Dose From Noble cases..................................... 175 ... F G. Gaseous Effluents: Dose From Iodine-131, Iodine-133, Tritium, and Radioactive Materials in Particulate Form............. 176 ... G H. Gaseous Radwaste Treatment................ 176 ... H I. Ventilation Exhaust Treatment............. 176 ... I J. Explosive Gas Mixture..................... 177 ... J K. Steam Jet Air Ejector (SJAE) ............. 177 ... K L. Primary Containment......... ............. 178 ... L M. Total Dose................................ 178 ... M N. Solid Radioactive Waste... ............... 179 ... N l BASES 185 3.9 RADIOACTIVE EFFLUENT MONITORING SYSTEMS....... 190 ... 4.9 A. Liquid Effluent Monitoring System......... 190 ... A B. Gaseous Effluent Instrumentation.......... 190 ... B C. Radiological Environment Monitoring Program................................... 191 ... C D. Land Use Census........................... 191 ... D E. Intercomparison Program................... 192 ... E l BASES 209 3.10 AUXILIARY ELECTRICAL POWER SYSTEMS............ 211 ... 4.10 A. Normal Operation.......................... 211 ... A B. Operation with Inoperable Components...... 215 ... B C. Diesel Fue1............................... 218 ... C l BASES 220 3.11 REACTOR FUEL ASSEMBLIES....................... 224 ... 4.11 A. Average Planar LHGR....................... 224 ... A B. LHGR...................................... 225 ... B C. MCPR...................................... 226 ... C ,

l BASES 227 3.12 REFUELING AND SPENT FUEL HANDLING............. 229 ... 4.12 ,

A. Refueling Inter 1ocks...................... 229 ... A B. Core Monitoring........................... 230 ... B C. Fuel Storage Pool Water Level............. 231 ... C l Amendment No. 44, 67, &b 95 -iv-

VYNPS TABLE OF CONTENTS (Continued) l l

i LIMITING CONDITIONS OF OPERATION Pace No. SURVEILLANCE D. Control Rod and Control Rod Drive Maintenance............................... 232 ... D E. Extended Core Maintenance................. 233 ... E j F. Fuel Movement............................. 235 ... F <

G. Crane Operability......................... 235 ... G ;

H. Spent Fuel Pool Water Temperature......... 236 ... H l BASES 237 3.13 FIRE PROTECTION SYSTEM........................ 240 ... 4.13 l BASES 252 !

5.0 DESIGN FEATURES............................... 253 6.0 ADMINISTRATIVE CONTROLS....................... 255 i

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l Amendment No. M, 44, G, &&, 95 -v- 1 1-I i I

r VYNPS 1.0 DEFINITIONS 1.0 DEFINITIONS The succeeding frequently used terms are explicitly defined so that a uniform interpretation of the specifications may be achieved.

A. Reportable Occurrence - The equivalent of a reportable event which shall be any of the conditions specified in Section 50.73 to 10CFR Part 50.

B. Alteration of the Reactor Core - The act of moving any component affecting reactivity within the reactor vessel in the region above the core support plate, below the upper grid and within the shroud.

Normal movement of control rods or neutron detectors, or the replacement of neutron detectors is not defined as a core alteration.

C. Hot Standby - Hot standby means operation with the reactor critical and the main steam line isolation valves closed.

D. Immediate - Immediate means that the required action will be ,

initiated as soon as practicable considering the safe operation of the unit and the importance of the required action.

E. Instrument Calibration - An instrument calibration means the '

adjustment of an instrument signal output so that it corresponds, within acceptable range and accuracy, to a known value(s) of the parameter which the instrument monitors. Calibration shall encompass the entire instrument including actuation, alarm, or trip. Response time as specified is not part of the routine instrument calibration ;

but will be checked once per operating cycle.

F. Instrument Check - An instrument check is qualitative determination of acceptable operability by observation of instrument behavior during operation. This determination shall include, where possible, comparison of the instrument with other independent instruments measuring the same variable.

G. Instrument Functional Test - An instrument functional test shall bei

1. Analog channels - the injection of a signal into the channel as close to the sensor as practicable to verify operability including alarm and/or trip functions.

2. Bistable channels - the injection of a signal into the sensor to I verify the operability including alarm and/or trip functions. -

l H. Loo System Functional Test - A logic system functional test means a test of all relays and contacts of a logic circuit-from sensor to '

l activated device to insure all components are operable per design >

intent. Where possible, action will go to completion, i.e., pumps l will be started and valves opened.

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I. Minimum Critical Power Ratio - The minimum critical power ratio is defined as the ratio of that power in a fuel assembly which is i calculated to cause some point in that assembly to experience boiling transition as calculated by application of the appropriate NRC-approved critical power correlation to the actual assembly operating power.

J. Mode - The reactor mode is that which is established by the mode-selector-switch.

I Amendment No. W, M, M, 4%, 137 1 1

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e-1 VYNPS 1.0 DEFINITIONS K. Operable - A system, subsystem, train, component or device shall be operable or have operability when it is capable of performing its specified function (s). Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support ;

function (s).

L. Operatina - Operating means that a system or component is performing its intended functions in its required manner.

H. Operatino Cycle - Interval between the end of one refueling outage and the end of the next subsequent refueling outage.

N. Primary Containment Intecrity - Primary containment integrity means that the drywell and pressure suppression chamber are intact and all of the following conditions are satisfied:

1. All manual containment isolation valves on lines connecting to the reactor coolant system or containment which are not required i to be open during accident conditions are closed.

2. At least one door in each airlock is closed and sealed. >

3. All automatic containment isolation valves are operable or deactivated in the isolated position.

4. All blind flanges and manways are closed.

O. Protective Instrumentation Definitions

1. Instrument Channel - An instrument channel means an arrangement of a sensor and auxiliary equipment required to generate and transmit to a trip system a single trip signal related to the plant parameter monitored by that instrument channel.

2. Trip System - A trip system means an arrangement of instrument channel trip signals and auxiliary equipment required to initiate action to accomplish a protective trip function. A trip system may require one or more instrument channel trip signals related to one or more plant parameters in order to initiate trip system

action. Initiation of protective action may require the tripping of a single trip system or the coincident tripping of two trip systems.

3. Protective Action - An action initiated by the protection system when a limit is reached. A protective action can be at a channel or system level. -

4. Protective Punction - A system protective action which results from the protective action of the channels monitoring a particular plant condition.

P. Rated Neutron Flux - Rated neutron flux is the neutron flux that corresponds to a steady state power level of 1593 thermal megawatts.

Q. Rated Thermal Power - Rated thermal power means a steady state power level of 1593 thermal megawatts.

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Amendment No. 64, 74, 83 2 l

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VYNPS 1.0 DEFINITIONS R. Reactor Power Operation - Reactor power operation is any operation with the mode switch in the "Startup/ Hot Standby

or *Run' position with the reactor critical and above 1% rated thermal power.

1. Startup/ Hot Standby Mode - In this mode the low turbine condenser volume trip is bypassed when condenser vacuum is less than 12 inches Hg and both turbine stop valves and bypass valves are closed; the low pressure and the 10 percent closure main steamline isolation valve closure trips are bypassed; the reactor protection system is energized with IRM neutron monitoring system trips and control rod withdrawal interlocks in service and APRM neutron monitoring system operable.

2. Run Mode - In this mode the reactor system pressure is equal to or greater than 800 psig and the reactor protection system is energized with APRM protection and RBM interlocks in service.

S. Reactor Vessel Pressure - Unless otherwise indicated, reactor vessel pressures listed in the Technical Specifications are those measured by the reactor vessel steam space detector. ,

T Refuelino Outage - Refueling outage is the period of time between the shutdown of the unit prior to a refueling and the startup of the plant subsequent to that refueling. For the purpose of designating frequency of testing and surveillance, a refueling outage shall mean a regularly scheduled refueling outage; however, where such outages occur within 8 months of the completion of the previous refueling outage, the required surveillance testing need not be performed until the next regularly scheduled outage.

U. Secondary Containment Integrity - Secondary containment integrity means that the reactor building is intact and the following conditions are met:

1. At least one door in each access opening is closed.

2. The standby gas treatment system is operable.

3. All reactor building automatic ventilation system isolation valves are operable or are secured in the isolated position.

4 V. Shutdown - The reactor is in a shutdown condition when the reactor "

mode switch is in the shutdown mode position and no core alterations are being performed. When the mode switch is placed in the shutdown position a reactor scram is initiated, power to the control rod '

drives is removed, and the reactor proteccion system trip systems are de-energized. t

1. Hot Shutdown means conditions as above with reactor coolant temperature greater that 212*F.

2. Cold Shutdown means conditions as above with reactor coolant temperature equal to or less than 2120F.

3. Shutdown means conditions as above such that the effective multiplication factor (K,gg) of the core shall be less than 0.99.

W. Simulated Automatic Actuation - Simulated automatic actuation means applying a simulated signal to the sensor to actuate circuit in question.

Amendment No. 74, 84 3

i VYNPS 1.0 DEFINITIONS X. Transition Boilino - Transition boiling means the boiling regime between nucleate and film boiling. Transition boiling is the regime in which both nucleate and film boiling occur intermittently with neither type being completely stable.

Y. Surveillance Frequency - Unless otherwise stated in these specifications, periodic surveillance tests, checks, calibrations, and examinations shall be performed within the specified surveillance intervals. These intervals may be adjusted plus 25%. The operating cycle interval is considered to be 18 months and the tolerance stated above is applicable.

Z. Su rveillance Interval - The surveillance interval is the calendar time between surveillance tests, checks, calibrations, and examinations to be performed upon an instrument or component when it is required to be operable. These tests unless otherwise stated in these specifications may be waived when the instrument, component, or system is not required to be operable, but these tests shall be performed on the instrument, component, or system prior to being required to be operable.

AA. Vital Fire Suppression Water System - The vital fire suppression water system is that part of the fire suppression system which protects those instruments, components, and systems required to perform a safe shutdown of the reactor. The vital fire suppression system includes the water supply, pumps, and distribution piping with associated sectionalizing valves, which provide immediate coverage of the Reactor Building, Control Room Building, and Diesel Generator Rooms.

BB. Source Check - The qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

CC. Dose Equivalent I-131 - The dose equivalent I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134 and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in NRC Regulatory Guide 1.109, Revision 1 October 1977.

DD. Solidification - Solidification shall be the conversion of wet wastes into a form that meets shipping and burial ground requirements.

Suitable forms include dewatered resins and filter sludges.

EE. Member (s) of the public - Members of the public shall include all persons who are not occupationally associated with the plant. This category does not include employees of the utility, its contractors or vendors. Also excluded from this category are casual visitors to the plant and persons who enter the site to service equipment or to make deliveries.

FF. Site Boundary - The site boundary is shown in Figure 2.2-5 in the FSAR.

GG. Deleted HH. Deleted Amendment No. 49, G4, 44, 70, 64, 103, 124 4

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VYNPS 1.0 DEFINITIONS II. Off-Site Dose Calculation Manual (ODCM) - A manual containing the current methodology and parameters used in the calculation of off-site doses due to radioactive gaseous and liquid effluents, in i the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints, and in the conduction of the environmental radiological ;

monitoring program.

JJ. Process Control Procram (PCP) - A process control program shall contain the sampling, analysis, tests, and determinations by which wet radioactive waste from liquid systems is assured to be converted to a form suitable for off-site disposal.

KK. Gaseous Radwaste Treatment System - The Augmented Off-Gas System '

(AOG) is the gaseous radwaste treatment system which has been designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system off-gases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

LL. Ventilation Exhaust Treatment System - The Radwaste Building and AOG Building ventilation HEPA filters are ventilation exhaust treatment systems which have been designed and installed to reduce radioactive material in particulate form in gaseous effluents by passing ventilation air through HEPA filters for the purpose of removing l radioactive particulates from the gaseous exhaust stream prior to release to the environment. Engineered safety feature atmospheric ,

cleanup systems, such as the Standby Gas Treatment (SBGT) system, are not considered to be ventilation exhaust treatment system components.

MM. Vent /Purcino - Vent / Purging is the controlled process of discharging '

air or gas from the primary containment to control temperature, pressure, humidity, concentration or other operating conditions.

IRJ. Core Operating Limits Report - The Core Operating Limits Report is the unit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.7.A.4. Plant operation within these operating limits is addressed in individual specifications.

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1 Amendment No. 83, 116 5 l

VYNPS 1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING '

l.1 FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY Applicability: Applicability: ,

Applies to the interrelated Applies to trip setting of the variable associated with fuel instruments and devices which thermal behavior. are provided to prevent the nuclear system safety limits from being exceeded.

Obiective: Obiective:

To establish limits below which To define the level of the the integrity of the fuel process variable at which cladding is preserved. automatic protective action is initiated.

specification: Specification:

A. Bundle Safety Limit (Reactor A. Trip Settinos Pressure >800 psia and Core Flow >10% of Rated) The limiting safety system trip settings shall be as When the reactor pressure is specified below

>800 psia and the core flow is greater than 10% of 1. Neutron Flux Trip rated, either: settinos

1. For a core loading which a. APRM Flux Scram Trip consists of at least two Settina (Run Mode) successive reloads of P8x8R, BP8x8R, GE8x8E, When the mode switch or GE8x8EB fuel with is in the RUN high (31.04) position, the APRM beginning-of-life bundle flux scram trip R-factor (one reload of . Setting shall be as which is fuel in its sh.'wn or first cycle of Figure 2.1.1 and operation), the shall bei existence of a Minimum Critical Power Ratio S$0.66(W-AW)+54%

(MCPR) of less than 1.04 >

(1.05 for Single Loop where:

Operation) shall constitute violation of S = setting in the Fuel Cladding percent of '

Integrity Safety Limit rated (FCISL); thermal power or (1593 MWt)

2. For all other core W = percent loadings, the existence rated two of a MCPR less than 1.07 loop drive (1.08 for Single Loop flow where Operation) shall 100% rated constitute violation of drive flow the FCISL. is that

, flow l equivalent i to 48 x 10 6 lbs/hr core flow ,

Amendment No. 44, 44, 64, 90, 94, 109 6

VYNPS 1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING B. Core Thermal Power Limit AW = difference (Reactor Pressure g 800 psia between two or Core Flow 1 10% of Rated) loop and single loop When the reactor pressure is drive flow

$800 psia or core flow $10% at the same of rated, the core thermal core flow.

power shall not exceed 25% This of rated thermal power. difference must be C. Power Transient accounted for during To ensure that the safety single loop limit established in operation.

Specification 1.1A and 1.1B AW = 0 for is not exceeded, each two loop required scram shall be operation.

initiated by its expected scram signal. The safety In the event of limit shall be assumed to be operation with the exceeded when scram is ratio of MFLPD to accomplished by means other FRP greater than than the expected scram 1.0, the APRM gain signal. shall be increased by the ratio: MFLPD D. Whenever the reactor is FRP shutdown with irradiated fuel in the reactor vessel, where the water level shall not be less than 12 inches above MFLPD = maximum the top of the enriched fuel fraction when it is seated in the of core. limiting power density where the limiting power density is defined in the core Oper-ating Limits Report.

FRP = fraction of rated power (1593 MWt).

In the event of operation with the ratio of MFLPD to FRP equal to or less than 1.0, the APRM gain shall be equal to or greater than 1.0.

Amendment No. 44, 39, 47, 64, 64, 68, 94, 116 7 N

VYNPS 1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING For no combination of loop recirculation flow rate and core thermal power shall the APRM flux scram trip setting be allowed to exceed 120% of rated thenmal power.

b. Flux Scram Trip settina (Refuel or Startup and Hot Standbv Mode)

When the reactor mode switch is in the REFUEL or STARTUP position, average power range monitor (APRM) scram shall be set down to less than or equal to 15% of rated neutron flux (except as allowed by Note 12 of Table 3.1.1). The IRM flux scram setting shall be set at less than or equal to 120/125 .f full scale.

B. APRM Rod Block Trio Settinq

1. The APRM rod block trip setting shall be as shown in Figure 2.1.1 and shall be:

S ag 5 066(W-4W)+42% l where:

S gg = rod block setting in percent of rated thermal power (1593 MWt)

W = percent rated ,

two loop drive flow where 100%

rated drive '

flow is that flow equivalent to 48 x 106 lbs/

hr core flow ,

l Amendment No. 40, 64, 76, 90, 94 8

VYNPS 1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING AW = difference between two loop and single loop drive flow at the same core flow. This difference must be accounted for during single loop operation.

AW = 0 for two loop operation.

In the event of operation with the ratio of MFLPD to FRP greater than 1.0, the APRM gain shall be increased by the ratio:

MFLPD FRP where:

MFLPD = maximum fraction of limiting power density where the limiting power density is defined in the Core Operating Limits Report.

FRP = fraction of rated power (1593 MWt).

In the ever.t of operation with the ratio of MFLPD to FRP equal to or less than 1.0, the APRM gain shall be equal to or greater than 1.0.

C. Reactor low water level scram setting shall be at least 127 inches above the top of the enriched fuel.

Amendment No. 44, 44, 64, 64, 68, 98, 94, 116 9

VYNPS 1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING D. Reactor low-low water level Emergency Core Cooling System (ECCS) initiation shall be at least 82.5 inches above the top of the enriched fuel.

E. Turbine stop valve scram shall be less than or equal to 10% valve closure from full open.

F. Turbine control valve fast closure scram shall, when operating at greater than 30% of full power, trip upon actuation of the turbine control valve fast closure relay.

G. Main steam line isolation valve closure scram chall be less than or equal to 10%

valve closure from full open.

H. Main steam line low pressure initiation of main steam line isolation valve closure shall be at least 800 psig. l l

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Amendment No. GB, 84 10

j VYNPS l 1

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ao 80 100 120 o 30 to RECIRCULATION FLOW (% RATED)

For sir >gle loop operation, the APRM Scrom and Rod Block settings are odjusted occording to Tel.hnical Specifications 2.tA.to and 2.1.B.1 Amendment No. %, 94 11

VYNPS BASES:

1.1 1 TL CLADDING TNTEGRITY Refer to Section S.2 of General Electric Company Licensing Topical Report, " United States Supplement, General Electric Standard Application for Reactor Fuel *, NEDE-240ll-P-A-US (Most Recent Revision).

The MCPR fuel cladding integrity safety limit is increased by 0.01 for single loop operation in order to account for increased core flow measurement and TIP reading uncertainties, as discussed in " Vermont Yankee Nuclear Power Station Single Loop Operation", NEDO-30060, February 1983.

1 B. Core Thermal Power Limit (Reactor Pressure 1 800 psia or Core Flow 510% of Rated)

At pressures below 800 psia, the core elevation pressure drop (0 power, O flow) is greater than 4.56 psi. At low power and all flows this pressure differential is maintained in the bypass region of the core. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and all flows will always be greater than 4.56 psi. Analyses show that with a flow of 28 x 103 lbs/hr bundle flow, bundle pressure drop -

is nearly independent of bundle power and has a value of 3.5 psi.

Thus, the bupdle flow with a 4.56 psi driving head will be greater than 28 x 10 lbs/hr irrespective of total core flow and independent of bundle power for the range of bundle powers of concern. Full ,

scale ATLAS test data taken at pressures from 14.7 psia to 800 psia <

indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors this corresponds to a core thernal power of more than 50%. Thus, a core ,

thermal power limit of 25% for reactor pressures below 800 psia or !

core flow less than 10% is conservative.

C. Power Transient Plant safety analyses have shown that the scrams caused by exceeding ,

any safety setting will assure that the Safety Limit of Specification 1.1.1A or 1.1.lB will not be exceeded. Scram times are checked periodically to assure the insertion times are adequate. The -

thermal power transient resulting when a scram is accomplished other than by the expected scram signal (e.g., scram from neutron flux following closure of the main turbine stop valves) does not necessarily cause fuel damage. However, for this specification a safety Limit violation will be assumed when a scram is only accomplished by means of a backup feature of the plant design. The concept of not approaching a safety Limit provided scram signals are operable is supported by the extensive plant safety analysis.

The computer provided with Vermont Yankee has a sequence annunciation program which will indicate the sequence in which events such as scram, APRM trip initiation, pressure scram initiation, etc. occur.

This program also indicates when the scram setpoint is cleared. This will provide information on how long a scram condition exists and thus provide some measure of the energy added during a transient.

(

Amendment No. 46, 49, 47, 94 12

VYNPS BASES: 1.1 (Cont'd)

D. Reactor Water Level (Shutdown condition)

During periods when the reactor is shutdown, consideration must also be given to water level requirements due to the effect of dacay heat.

l If reactor water level should drop below the top of the enriched fuel during this time, the ability to cool the core is reduced. This reduction in core cooling capability could lead to elevated cladding temperatures and clad perforation. The core can be cooled sufficiently should the water level be reduced to two-thirds the core height. Establishment of the safety limit at 12 inches above the top of the enriched fuel provides adequate margin. This level will be '

continuously monitored.

1 i

I L

1 Amendment No. 44, 68 13

l l

VYNPS BASES.

l l 2.1 FUEL CLADDING INTEGRITY l A. Trip settinos l

The bases for individual trip settings are discussed in the following I paragraphs, l

1. Neutron Flux Trip Settinqs

a. APRM Flux Scram Trio Settino (Run Model The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady state conditions, reads in percent of rated thermal power (1593 MWt). Because fission chambers provide the basic input signals, the ADR'l system responds directly to average neutron flux. During transients, the instantaneous rate of heat transfer from the fuel (reactor thermal power) is less than the instantaneous neutron flux due to the time constant of the fuel. Therefore, during abnormal operational transients, the thermal power of the fuel will be less than that indicated by the neutron flux at the scram setting.

Analyses demonstrate that with a 120 percent scram trip setting, none of the abnormal operational transients analyzed violate the fuel Safety Limit and there is substantial margin from fuel damage. Therefore, the use of flow referenced scram trip provides even additional margin.

An increase in the APRM scram trip setting would decreace the margin present before the fuel cladding integrity Safety Limit is reached. The APRM scram trip setting was determined by an analysis of margins required to provide a reasonable range for maneuvering during operation. Reducing this operating margin would increase the frequency of i spurious scrams which have an adverse effect on reactor safety because of the resulting thermal stresses. Thus, the APRM scram trip setting was selected because it provides adequate margin for the fuel cladding integrity Safety Limit yet allows operating margin that reduces the possibility of unnecessary scrams.

APRM Flux Scram Trio Settina (Run Mode)

The scram trip setting must be adjusted to ensure that the LHGR transient peak is not increased for any combination of MFLPD and reactor core thermal power. If the scram requires a change due to an abnormal peaking condition, it will be accomplished by increasing the APRM gain by the ratio in Specification 2.1.A.l.a, thus assuring a reactor scram at ]

j lower than design overpower conditions. For single i recirculation loop operation, the APRM flux scram trip setting is reduced in accordance with the analysis presented in NEDO-30060, February 1983. This adjustment accounts for the difference between the single loop and two loop drive flow at the same core flow, and ensures that the margin of safety is not reduced during single loop operation.

Analyses of the limiting transients show that no scram adjustment is required to assure fuel cladding integrity when the transient is initiated from the operating limit MCPR defined in the Core Operating Limits Report.

Amendment No. 44, as, 49, 47, 64, 94, 116 14

VYNPS BASES: 2.1 (Cont'd)

Flux Scram Trip Settino (Refuel or Startup and Hot Standby Mode)

For operation in the startup mode while the reactor is at low pressure, the reduced APRM scram setting to 15% of rated power provides adequate thermal margin between the setpoint and the safety limit, 25% of the rated. (During an outage when it is necessary to check refuel interlocks, the mode switch must be moved to the startup position. Since the APRM reduced scram may be inoperable at that time due to the disconnecticn of the LPRMs, it is required that the IRM scram and the SRM scram in noncoincidence be in effect.

This will ensure that adequate thermal margin is maintained between the setpoint and the safety limit.) The margin is adequate to accommodate anticipated maneuvers associated ,

with station startup. Effects of increasing pressure at zero or low void content are minor, cold water from sources available during startup is not much colder than that already in the system, temperature coefficients are small, and control rod patterns are constrained to be uniform by operating procedures backed up by the rod worth minimizer.

Worth of individual rods is very low in a uniform rod pattern. Thus, of all possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power rise. Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to change power by a significant percentage of rated power, the rate of power rise is very slow. Generally, the heat flux is in near equilibrium with the fission rate. In an assumed uniform rod withdrawal approach to the scram level, the rate of power rise is no more than 5% of rated power per minute, and the APRM system would be more than adequate to assure a scram before the power could exceed the safety limit. The reduced APRM scram remains active until the mode switch is placed in the RUN position. This switch can occur when reactor pressure is greater than 800 psig.

The IRM system consists of 6 chambers, 3 in nach of the reactor protection system logic channels. The IRM is a 5-decade instrument, which covers the range of power level between that covered by the SRM and the APRM. The 5 decades are covered by the IRM by means of a range switch and the 5 decades are broken down into 10 ranges, each being one-half of a decade in size. The IRM scram trip setting of 120/125 of full scale is active in each range of the 1RM.

For example, if the instrument were on range 1, the scram setting would be a 120/125 of full scale for that range; likewise, if the instrument were on range 5, the scram would be 120/125 of full scale on that range. Thus, as the IRM is ,

ranged up to accommodate the increase in power level, the scram trip setting is also ranged up. The most significant sources of reactivity change during the power increase are due to control rod withdrawal. For in-sequence control rod withdrawal, the rate of change of power is slow enough due ,

to the physical limitation of withdrawing control rods, that heat flux is in equilibrium with the neutron flux and an IRM s: ram would result in a reactor shutdown well before any safety limit is exceeded.

Amendment No. 46, 48, 84, 94 15

VYNPS BASES: 2.1 (Cont'd)

In order to ensure that the IRM provided adequate protection against the single rod withdrawal error, a range of rod withdrawal accidents was analyzed. This analysis included starting the accident at various power levels. The most severe case involves an initial condition in which the reactor is just subcritical and the IRM system is not yet on scale. This condition exists at quarter rod density.

Additional conservatism was taken in this analysis by assuming that the IRM channel closest to the withdrawn rod is bypassed. The results of this analysis show that the reactor is scrammed and peak power limited to one percent of rated power, thus maintaining MCPR above the fuel cladding integrity safety limit. Based on the above analysis, the IRM provides protection against local control rod withdrawal errors and continuous withdrawal of control rods in sequence.

B. APRM Rod Block Trip Settin,g Reactor power level may be varied by moving control rods or by varying the recirculation flow rate. The APRM system provides a control rod block to prevent rod withdrawal beyond a given point at the fuel cladding integrity safety limit. This rod block trip setting, which is automatically varied with recirculation loop flow rate, prevents an increase in the reactor power level to excessive values due to control rod withdrawal. The flow variable trip setting provides substantial margin from fuel damage, assuming a steady-state operation at the trip setting, over the entire recirculation flow range. The margin to the Safety Limit increases as the flow decreases for the specified trip setting versus flow relationship, therefore the worst case MCPR which could occur during steady-state operation is at 108% of rated thermal power because of the APRM rod block trip setting. The actual power distribution in the core is established by specified control rod sequences and is monitored continuously by the incore LPRM system. As with the APRM scram trip setting, the APRM rod block trip setting must be adjusted downward if the ratio of MFLPD to FRP exceeds the specified value. If the APRM rod block requires a change due to abnormal peaking conditions, it will be accomplished by increasing the APRM gain by the ratio in ,

Specification 2.1B, thus ensuring a rod block at lower than design overpower conditions. As with the APRM flux scram trip setting, the APRM rod block trip setting is reduced for single recirculation loop operation in accordance with the analysis presented in NEDO-30060, February, 1983. This adjustment accounts for the difference between the single loop and two loop drive flow at the same core flow, and ensures that the margin of safety is not reduced during single loop operation.

C. Reactor Low Water Level Scram The reactor low water level scram is set at a point which will prevent reactor operation with the steam separators uncovered, thus limiting carry-under to the recirculation loops. In addition, the safety limit is based on a water level below the scram point and therefore this setting is provided.

D. Reactor Low Water Level ECCS Initiation Trin Point The core standby cooling subsystems are designed to provide sufficient cooling to the core to dissipate the energy associated with the loss-of-coolant accident and to limit fuel clad temperature to well below the clad melting temperature, and to limit clad Amendment No. +B, 36, 47, 64, 94 16

VYNPS BASES: 2.1 (Cont'd) metal-water reaction to less than 1%, 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 Stop Valve Closure Scram Trip 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 l 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 isolation valve closure scram assures the available of neutron scram protection over the entire range of applicability of the fuel ,

eladding integrity safety limit.

l l

l Amendment No. +6, BG, 84 17

1 I

VYNPS I 1.2 SAFETY LIMIT 2.2 LIMITING SAFETY SYSTEM SETTING 1.2 REACTOR COOLANT SYSTE24 2.2 REACTOR COOLANT SYSTEM Applicability: Applicability:

Applies to limits on reactor Applies to trip settings for coolant system pressure. controlling reactor system 1 pressure.

Obiectives Obiective:

To establish a limit below which To provide for protective action the integrity of the reactor in the event that the principle coolant system is not threatened process variable approaches a due to an overpressure safety limit.

condition.

Specification: Specification:

The reactor coolant system pressure shall not exceed 1335 psig at any time when irradiated fuel is prosent in the reactor vessel.

A. Reactor coolant high pressure scram shall be less than or equal to 1055 psig.

B. Primary system relief and safety valve settings shall be as follows:

1 valve at $1080 psig 2 valves at $1090 psig 1 valve at $1100 psig 2 valves at.51240 psig (safety valves) ll -

I Amendment No. 18 18

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

VYNPS BASES:

1.2 REACTOR COOLANT SYSTEM The reactor coolant system is an important barrier in the prevention of uncontrolled release of fission products. It is essential that the >

integrity of this system be protected by establishing a pressure limit to be observed for all operating conditions and whenever there is irradiated fuel in the reactor vessel.

The pressure safety limit of 1335 psig as measured by the vessel steam space pressure indicator is equivalent to 1375 psig at the lowest elevation of the reactor coolant systom. The 1375 psig value is derived from the design pressures of the reactor pressure vessel, and the coolant system piping. The respective design pressures are 1250 psig at 5750F and 1148 psig at 5600F. The pressure safety limit was chosen as the lower of the pressure transients permitted by the applicable design codes: ASME Boiler and Pressure Vessel Code, Section III-A for the pressure vessel, ASME Boiler and Pressure Vessel Code Section III-C for the recirculation pump casing, and USASI B31.1 Code for the reactor coolant system piping. The ASME Boiler and Pressure Vessel Code permits pressure transients up to 10% over design pressure (110% x 1250 = 1375 psig), and the USASI Code permits pressure transients up to 20% over the design pressure (120% x 1148 = 1378 psig).

The safety valves are sized to prevent exceeding the pressure vessel code limit for the worst-case isolation (pressurization) event (MSIV closure) assuming indirect (neutron flux) scram.

2.2 REACTOR COOLANT SYSTEM ;

The settings on the reactor high pressure scram, reactor coolant system l relief and safety valves, have been established to assure never reaching i the reactor coolant system pressure safety limit as well as assuring the system pressure does not exceed the range of the fuel cladding integrity j safety limit. In addition to preventing power operation above 1055 psig, i I

the pressure scram backs up the APRM neutron flux scram for steam line isolation type transients. (See FSAR Section 14.5 and Supplement 2 to .

Proposed Change No. 14, November 12, 1973.) l l

l

'l i

l l

1

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I i

Amendment No. 18 19

VYNPS 3.1 LIMITING CONDITIONS FOR 4.1 SURVEILLANCE REQUIREMENTS OPERATION 3.1 REACTOR PROTECTION SYSTEM 4.1 REACTOR PROTECTION SYSTEM Applicability: Applicability:

Applies to the operbility of Applies to the surveillance of plant instrumentation and the plant instrumenttion and control systems required for control systems required for reactor safety, reactor safety.

Obiectives Obiective:

To specify the limits imposed on To specify the type and plant operation by those frequency of surveillance to be instrument and control systems applied to those instrument and required for reactor safety. control systems required for reactor safety.

Specification: 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, from the opening of the respectively.

sensor contact up to and including the opening of the scram solenoid relay shall l 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 density and fraction of

a. The APRM System gains rated power shall be shall be adjusted by the determined 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 i reduce the ratio of MFLPD to FFP.

Amendment No. 61 20 I

VYNPS TABLE 3.1.1 PEACTOR PROTECTION SYSTEM (SCRAM) INSTPITMENT REOUIREMENTS Required Conditions Minimum Number When Minimum Mcdes in Which Functions Must Operating Conditions For be Operating Instrument Operation Channels Per Are Not Trip Function Trip Settinos 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

3. IBM High Flux 1120/125 X X X(11) 2 A INOP X X X(11) 2 A

4. APRM High Flux 5 066 (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 12/125 X 2 A or B

5. High Reactor $1055 psig X X X 2 A Pressure

6. High Drywell $ 2.5 psig X X X 2 A Pressure

7. Reactor Low (6) >127.0 inches X X X 2 A Water Level

8. Scram Discharge $21 gallons X X X :: A Volume High Level (per volume)

I Amendment No. i% 44, 64, 68, M, M, M, 44, 94 21 l

7 VYNPS TABLE 3.1.1 (Cont *d)

Required Conditions Min. 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 Pefuel (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 rated power (8)

10. Main steamline $10% valve X 4 A or C isolation valve closure closure

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

12. Turbine stop valve $10% valve (10) X 2 A or D closure closure 22

VYNPS TABLE 3.1.1 NOTES

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

a) mode switch in shutdown b) manual scram c) high flux IRM or high flux SRM in coincidenco d) scram discharge volume high water level

2. Whenever an instrument system is found to be inoperable, the instrument system output relay shs11 be tripped immediately. Except for MSIV and Turbine Stop Valve Position, thin action shall result in triP ping the trip system.

3. When the requirements in the column " Minimum Number of Operating Instrument Channels Per Trip System" cannot be met for one system, that system shall be tripped. If the requirements cannot be met for both trip systems, the appropriate actions listed below shall be taken:

a) Initiate insertion of operable rods and complete insertion of all operable rods within four hours, b) Reduce power level to IRM range and place mode switch in the

'Startup/ Hot Standby" position within eight hours.

c) Reduce turbine lead 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 .

4. 'W' is percent rated two loop dgive 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.

This difference must be accounted for during single loop operation.

AW = 0 for two recirculation loop operation.

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 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. 24, GG, 66, 76, 94 23

VYNPS TABLE 3.1.1 NOTES (Cont'd)

9. Channel signals for the turbine control valve fast closure trip shall be derived from the same event or events which cause the control valve fast closure.

10. A turbine stop valve closure and generator load rejection bypass is permitted when the first stage turbine pressure is less than 30% of normal (220 psia).

11. The IRM scram is bypassed when the APRMs are on scale and the mode switch is in the run position.

12. While performing refuel interlock checks which require the mode switch to be in Startup, the reduced APRM high flux scram need not be operable provided:

a. The following trip functions are operable:

1. Mode switch in shutdown,

2. Manual scram,

3. liigh flux IRM scram

4. High flux SRM scram in no..oincidence,

5. Scram discharge volume high water level, and;

b. No more than two (2) control rods withdrawn. The two (2) control rods that can be withdrawn cannot be faced adjacent or diagonally adjacent.

Amendment No. 44, as, 64, 76, 90 24

VYNPS TABLE 4.1.1 SCRAM INSTRUMENTATIC'I AND LOGIC SYSTEMS FUNCTIONAL TESTS MINIMUM FUNCTIONAL TEST FREOUENCIES FOR SAFETY INSTRUMENTATION, LOGIC SYSTEMS AND CONTROL CIRCUITS Instrument Channel Group (3) Functional Test I7I Minimum FrequencyI4I Mode Switch in Shutdown A Place Mode Switch in Shutdown Each Refueling Outage Manual Scram A Trip Channel and Alarm Every 3 Months IRM High Flux C Trip Channel and Alarm (5) Before Each Startu g Weekly During Refueling Inoperative C Trip Channel and Alarm Before Each Startu Weekly During Refueling APRM High Flux B Trip output Relays (5) Once Each Week High Flux (Reduced) B Trip Output Relays (5) Before Each Startu g Weekly During Refueling Inoperative B Trip output Relays Once Each Week Downscale B Trip Output Relays (5) Once Each Week Flow Eias B Trip Output Relays (5) g1)

High Reactor Pressure B Trip Channel and Alarm (5) g1) l High Drywell Pressure B Trip Channel and Alarm (5) g1)

Low Reactor Water Level (2) (8) B Trip Channel and Alarm (5) (1)

High Water Level in Scram Discharge B Trip Channel and Alarm (5) (1)

Volume High Main Steam Line RadiationI2) B Trip Channel and Alarm (5) Once Each Week Main Steam Line Iso. Valve Closure A Trip Channel and Alarm (1)

Turbine Con. Valve Fast Closure A Trip Channel and Alarm (1)

Turbine Stop Valve Closure A Trip Channel and Alarm (1)

Scram Test Switch A Trip Channel and Alarm Each Refueling Outage First Stage Turbine Pressure - A Trip Channel and Alarm Every 6 Months Permissive l Amendment No. H, M, GB, 76 25

l VYNPS TABLE 4.1.1 NOTES

1. Initially once per month; thereafter, with an interval not less than one month nor more than three months. The compilation of instrument failure rate data may include data obtained from other Boiling Water Reactors for which the same design instrument operates in an environment similar to that of Vermont Yankee.

2. 7r. instrument check shall be performed on reactor water level and reactor pressure instrumentation once per day and on steamline radiation monitors once per shift. ,

3. A description of the three groups is included in the basis of this specification.

4. Functional tests are not required when the systems are not required to be operable or are tripped. If tests are missed, they shall be performed prior to returning the systems to an operable status.

5. This instrumentation is exempted from the Instrument Punctional Test Definition (1.G.). This Instrument Punctional Test will consist of injecting a simulated clectrical signal into the measurement channels.

6. Frequency need not exceed weekly.

7. A functional test of the logic of each channel is performed as indicated.

This coupled with placing the mode switch in shutdown each refueling outage constitutes a logic system functional test of the scram system.

8. The water level in the reactor vessel will be perturbed and the corresponding level indicator changes will be monitored. This test will be performed every month af ter the completion of the monthly tests program.

I t

Amendment No. Et, 58 26

VYNPS TABLE 4.1.2 SCRAM INSTRUMDTP CALIBRATION MINIMUM CALIBRATION FREOUENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNELS Instrumant Channel Group III Calibration StandardI4I Minimum FrequancyI2)

High Flux APRM Output Sigaal 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 LPRM B(5) 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 l 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 Approp te Radiation Refueling Outage Source First Stage Turbine Pressure A Pressure Source Every 6 Months and After Permissive Refueling Main Steam Line Isolation Valve A (6) Refueling Outage Closure Amendment No. 44, M, M, M, M, 76 27

VYNPS TABLE 4.1.2 NOTES

1. A description of the three groups is included in the bases of this Specification.

2. Calibration tests are not required when the systems are not required to l be operable or are tripped. If tests are missed, they shall be performed l prior to returning the systems to an operable status. 1

3. A current source provides an instrument channel alignment every 3 months.

4. Response time is not part of the routine instrument check and calibration, but will be checked every operating cycle.

l 5. Does not provide scram function.

l 6. Physical inspection and actuation.

t i

Amendment No. 21 28

VYNPS 1

l BASES: i 3.1 Reactor Protection System

]

The reactor protection system automatically initiates a reactor scram to:

i

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 buttona 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.

l 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 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 preserved; i.e., the system can tolerate a single failure and still perform its intended function of scramming the i reactor.

Three APRM instrument channels are provided for each protection trip system to prov'de for high neutron flux protection. APRM's A and E operate contacts in a trip subsystem, and APRM's C an E operate contacts 4 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 and on FSAR Page 7.5-8.

i; 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.

Instrumentation (pressure switches) 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. 21 29 i

VYNPS BASES: 3.1 (Cont'd)

The Control Rod Drive Scram System is designed so that all of the water that is discharged from the reactor by the scram can be accommodated in the discharge piping. This discharge piping is divided into two sections, one section services the control rod drives on the north side of the reactor, the other serves the control rod drives of the south side. A part of the piping in each section is an instrument volume which accommodates in excess of 21 *silons of water and is at the low point in the piping. No credit was ta.en for this volume in the design of the discharge piping as concerns the amount of water which must be accommodated during a scram. During normal operation, the discharge volume is empty; however, should it fill with water, the water discharged to the piping from the reactor could not be accommodated, which would result in slow scram times or partial or no control rod insertion. To preclude this occurrence, level instrumentation has been provided for the instrument volume which scram the reactor when the volume of water reaches 21 gallons. As indicated above, there is sufficient volume in the piping to accommodate the scram without impairment of the scram times or amount of insertion of the control rods. This function shuts the reactor down while sufficient volume remains to accommodate the discharged water, and precludes the situation in which a scram would be required but not be able to perform its function adequately. The present design of the Scram Discharge system is in concert with the BWR Owner's Group criteria, which have previously been endorsed by the NRC in their generic " Safety Evaluation Report (SER) for Scram Discharge Systems",

dated December 1, 1980.

Loss of condenser vacuum occurs when the condenser can no longer handle the heat input. Loss of condenser vacuum initiates a closure of the r.urbine stop valves and turbine bypass valves wha h eliminates the heat j input to the condenser. Closure of the turbine stop and bypass valves causes a pressure transient, neutron flux rise, and an increase in surface heat flux. To prevent the clad safety limit from being exceeded if this occurs, a reactor scram occurs on turbine stop valve closure.

The tu.bine stop valve closure scram function alone is adequate to prevent the clad safety limit from being exceeded in the event of a turbine trip transient without bypass.

High radiation levels in the main steam line tunnel above that due to the normal nitrogen and oxygen radioactivity is an indication of leaking fuel. A scram is initiated whenever such radiation level exceeds three times nornal background. The purpose of this scram is to reduce the source of such radiation to the extent necessary to prevent release of radioactive materials to the turbine. An alarm is initiated whenever the radiation level exceeds 1.5 times normal background to alert the operator to possible serious radioactivity spikes due to abnormal core behavior.

The air ejector off-gas monitors serve to back up the main steam line monitors to provide further assurance against release of radioactive materials to site environs by isolating the main condenser off-gas line to the main stack.

The main steam line isolation valve closure scram is set to scram when the isolation valves are 10 percent closed from full open in 3-out-of-4 lines. This scram anticipates the pressure and flux transient, which would occur when the valves close. By scramming at this setting, the resultant transient is insignificant.

A reactor mode switch is provided which actuates or bypasses the various scram functions appropriate to the particular plant operating status.

The manual scram function is active in all modes, thus providing for manual means of rapidly inserting control rods during all modes of reactor operation.

Amendment No. B+, 76 30 9

j VYNPS BASES: 3.1 (Cont'd)

The IRM system provides protection against short reactor periods and, in conjunction with the reduced APRM system provides protection against excessive power levels in the startup and intermediate power ranges. A source range monitor (SRM) system is also provided to supply additional neutron level information during startup and can provide scram function with selected shorting links removed during refueling. Thus, the IRM and the reduced APRM are normally required in the startup mode and may be required in the refuel mode. During some refueling activities which require the mode switch in startup; it is allowable to disconnect the LPRMs to protect them from damage during under vessel work. In lieu of the protection provided by the reduced APRM scram, both the IRM scram and the SRM scram in noncoincidence are used to provide neutron monitoring protection against excessive power levels. In the power range, the ,

normal APRM system provides required protection. Thus, the IRM system and 15% APRM scram are not required in the run mode. The requirement that the IRMs be inserted in the core until the APRMs read at least 2/125 of full scale assures that there is proper overlap in the neutron monitoring systems.

If an unsafe failure is detected during surveillance testing, it is desirable to determine as soon as possible if other failures of a similar 1 type have occurred and whether the particular function involved is still operable or capable of meeting the single failure criteria. To meet the requirements of Table 3.1.1, it is necessary that all instrument channels in one trip system be operable to permit testing in the other trip system.

l Thus, when failures are detected in the first trip system tested, they ,

would have to be repaired before testing of the other system could begin. :

In the majority of cases, repairs or replacement can be accomplished l quickly. If repair or replacement cannot be completed in a reasonable time, operation could continue with one tripped system until the surveillance testing deadline.

The requirement to have all scram functions, except those listed in Table 3.1.1, operable in the ' Refuel

mode is to assure that shifting to this mode during reactor operation does not diminish the need for the -

reactor protection system.

The ability to bypass one instrument channel when necessary to complete surveillance testing will preclude continued operation with scram functions which may be either unable to meet the single failure criteria or completely inoperable. It also eliminates the need for an unnecessary shutdown if the remaining channels and subsystems are found to be r operable. The conditions under which the bypass is permitted require an immediate determination that the particular function is operable.

However, during the time a bypass is applied, the function will not meet the single failure criteria; therefore, it is prudent to limit the time the bypass is in effect by requiring that surveillance testing proceed on a continuous basis and that the bypass be removed as soon as testing is completed.

Sluggish indicator response during the perturbation test will be indicative of a plugged instrument line or closed instrument valves.

Testing inmediately af ter functional testing will assure the operability of the instrument lines. This test assures the operability of the reactor pressure sensors as well as the reactor level sensora since both parameters are monitored through the same instrument lines.

1 Amendment No. B1, 78 31

1 VYNPS BASES: 3.1 (Cont'd)

The independence of the safety system circuitry is determined by operation of the scram test switch. Operation of this switch during the refueling outage and following maintenance on these circuits will assure their continued independence.

The calibration frequency, using the TIP system, specified for the LPRMs will provide assurance that the LPRM input to the APRM system will be corrected on a timely basis for LPRM detector depletion characteristics.

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VYNPS BASES:

4.1 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 i indicated intervals. Initially the tests are more frequent than Yankee experience indicates necessary. However, 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.

Group B devices utilize an analog sensor followed by an amplifier and bi-stable trip circuit. This type of equipment incorporates control room mounted indicators and annunciator alarms. A failure in the sensor or amplifier may be detected by an alarm or by an operator who observes that one indicator does not track the others in similar channels. The bistable trip circuit failures are detected by the ;

periodic testing. l 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 i adequate.

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r Amendment No. 68, 61 33

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VYNPS 3.2 LIMITING CONDITIONS FOR 4.2 SURVEILLANCE REQUIREMENTS OPERATION 3.2 PROTECTIVE INSTRUMENT SYSTEMS 4.2 PROTECTIVE INSTRUMENT SYSTEMS Applicability: Applicability:

Applies to the operational Applies to the surveillance status of the plant requirements of the 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 protective instrumentation protective instrumentation systems. systems.

Specification: Specification:

A. Emeroency Core Coolino A. Emeroency Core Coolino System System 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 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 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 Standby Gas Treatment System Initiation _ Treatment System Initiation The instrumentation that Instrumentation and logic initiates the isolation of systems shall be 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. <

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l VYNPS 3.2 LIMITING CONDITIONS FOR 4.2 SURVEILLANCE REQUIREMENTS l OPERATION '

l D. Off-Gas System Isolation D. Off-Gas System Isolation l During reactor power Instrumentation and logic cperation, 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.

E. Control Rod Block Actuation E. Control Rod Block Actuation During reactor power Instrumentation and logic operation the systems shall be instrumentation that functionally tested and initiates control rod block calibrated as indicated in shall be operable in Table 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 are open, the mechanical securing of the mechanical 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 Instrumentation Instrumentation During reactor power The post-accident operation, the instrumentation shall be instrumentation that functionally tested and 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 postulated accident of abnormal operating condition shall be operable in accordance with Table 3.2.6.

Amendment No. 9 35 i

VYNPS 3.2 LIMITING CONDITIONS FOR 4.2 SURVEILLANCE REQUIREMENTS OPERATION H. Drywell to Torus AP H. Drywell to Torus AP Instrumentation Instrumentation '

l. During reactor power The Drywell to Torus AP operation, the Drywell Instrumentation shall be to Torus AP calibrated once every six 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 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 Trip I. Recirculation Pump Trip l Instrumentation Instrumentation During reactor power The Recirculation Pump Trip operation, the Recirculation Instrumentation shall be .

Pwmp Trip Instrumentation functionally testod and j shall be oporative in calibrated in accordance accordance with Table 3.2.1. with Table 4.2.1.

l J. Deleted J. Deleted l ,

1 K. Dooraded Grid Protective K. Decraded Grid Protective l System system During reactor power The emergency bus l operation, the emergency bus undervoltage instrumentation !

undervoltage instrumentation shall be functionally tested shall be operative in and calibrated in accordance accordance with Table 3.2.8. with Table 4.2.8.

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Amendment No. 50, 58, 96, 96, 1++, 132 36 i

VY:!PS 3.2 LIMITING CONDITIONS FOR 4.2 SURVEILLANCE REQUIREMENTS OPERATION L. Reactor Core Isolation L. Reactor Core Isolation Coolinc System Actuation Coolinc System Actuation When the Reactor Core Instrumentation and Logic Isolation Cooling System is Systems shall be required in accordance with functionally tested and Specification 3.5.G, the calibrated as indicated in .

instrumentation which Table 4.2.9.

initiates actuation of this system shall be operable _in ;

accordance with Table 3.2.9.

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i Amendment No. Se, 94, 90, 111 37

VYNPS TABLE 3.2.1 DfEPGENCY 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 Ftmet ion Trio Level Settina are Not Satisfied 2 High Drywell Pressure 5 2 5 psig Note 2 2 Low-Low Reactor Vessel Water 282.5" above top of Note 2 Level enriched fuel 1 Low Reactor Pressure 300 1P $ 350 psig Note 2 (PT-2-3-56C/D(S1))

2 Low Reactor Pressure 300 1P5 350 psig Note 2 (PT-2-3-56A/B(S1) & 52C/D(M))

1 Time Delay (14A-K16A & B) 510 seconds Note 2 2 Pump 14-1A, Discharge Pressure 1100 psig Note 5 1 Auxiliary Power Monitor --

Note 5 1 Pump Bus Power Monitor --

Note 5 l 1 High Sparger Pressure 15 psid Note 5 1 Trip System Logic --

Note 5 Amendment No. 44, 60, 4M, 131 38

VYNPS TABLE 3.2.1 (Cont'd)

EMERGENCY CORE COOLING SYSTEM AC7UATION INSTRUMENTATION Low Pressure Coolant Injection System A & B (Note 1)

Minimum Number of Required Action When Operable Instrument Minimum Conditions Channels per Trip for Cperation System Trip Function Trip Level Settina are Not Satisfied 1 Low Reactor Pressure 300 g p 5 350 psig Note 2 (PT-2-3-56C/D(S1))

2 High Drywell Pressure 5 2 5 psig Note 2

( PT-10-101 A-D (11) )

2 Low-Low Reactor vessel Water 282.5= above top of Note 2 Level enriched fuel 1 Time Delay (10A-K51A & B) 0 seconds Note 5 1 Reactor Vessel Shroud Level 12/3 core height Note 5 1 Time Delay (10A-K72A & B) 560 seconds Note 5 1 Time Delay (10A-K50A & B) 15 seconds Note 5 1 Low Reactor Pressure 100 g p 5 150 psig Note 2 (PS-2-128A & B) 2 per pump RHR Pump A & C Discharge y_100 psig Note 5 Pressure 2 High Drywell Pressure $ 2.5 psig Note 2 (PT-10-101A-D(S1))

Amendment No. M, 44, 68, 110 39

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i VYNPS TABLE 3.2.1 (Cont'd)

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

Minimum Number of Required Action When Operable Instrument Minimum Conditions i Channels per Trip for Operation System Trip Function Trip Level Setting are Not Satisfied 1 Time Delay (10A-K45A & B) 16 minutes Note 5 2 Low Reactor Pressure 300 $ p s 350 psig Note 2 (PT-2-3-56A/B(S1) & 52C/D(M))

1 Auxiliary Power Monitor --

Note 5 1 Pump Bus Power Monitor --

Note 5 1 Trip System Logic --

Note 5 Amendment No. 44, 110 40

l 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 Trip Function Trio Level Setting are Not Satisfied 2 (Note 3) Low-Low Reactor Vessel Water Same as LPCI Note 5 Level 2 (Note 4) Low Condensate Storage Tank 1 3% Note 5 Water Level 2 (Note 3) High Drywell Pressure Same as LPCI Note 5 1 (Note 3) Bus Power Monitor -- Note 5 ,

1 (Note 4) Trip System Logic ~~ Note 5 2 (Note 7) High Reactor Vessel Water <177 inches above top of Note 5 Level enriched fuel Amendment No. 46, 65, 90 41 >

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

VYNPS TABLE 3.2.1 (Cont'd)

EMERGENCY CORE COOLING SYSTEM AC"JATION INSTRUMENTATION Automatic Depressurization Minimum Number of Required Action When Operable Instrument Minimum Conditions Channels per Trip for Operation System (Note 4) Trip Function Trir Level Settina are Not Satisfied 2 Low-Low Reactor Vessel Water Same as Core Spray Note 6 Level 2 High Drywell Pressure $ 2.5 psig Note 6 1 Time Delay (2E-KSA and B) 5120 seconds Note 6 1 Bus Power Monitor -- Note 6 1 Trip System Logic -- Note 6 2 Time Delay 58 minutes Note 6 (2E-K16A and B, 2E-K17A and B)

Amendment No. 44, 105 42

VYNPS TABLE 3.2.1 (Cont'd)

RECIRCULATION TUMP TRIP ACTUATION INSTRUMENTATION Recirculation Pump Trip - 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 Low-Low Reactor Vessel Water > 6' 10.5" above top of Note 2 Level enriched fuel 2 High Reactor Pressure $ 1150 psig Note 2 2 Time Delays s 10 seconds Note 2 1 Trip Systems Logic --

Note 2 Amendment No. GB, 68, -74, 75 43

VYNPS TABLE 3.2.1 NOTES l 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 permanently installed circuits. If the channel cannot be tripped by the means stated above, that channel shall be made operable within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or an orderly shutdown shall be initiated and the reactor shall be in the '

cold shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . l

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

4. One trip system with initiating instrumentation arranged in a 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 requirements of Specification 3.5.F.2 and 3.5.F.3 shall apply. If the minimum number of operable channels is not available in both trip systems, Specifications 3.5.F.3 shall apply.

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

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Amendment No. 58 44

VYNPS TABLE 3.2.2 PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION Minimum Nuzrber of Required Action When Operable Instrument Minimum Conditions Channels per Trip for Operation are System Trip Function Trip Setting Not Satisfied (Note 2) 2 Low-Low Reactor Vessel Water 182.5" above the top of A Level enriched fuel 2 of 4 in each of High Main Steam Line Area $212oF B 2 channels Temperature 2/ steam line High Main Steam Line Flow $140% of rated flow B 2/(Note 1) Low Main Steam Line Pressure 1800 psig B 2/(Note 6) High Main Steam Line Flow <40% of rated flow B 2 Low Reactor Vessel Water Level Same as Reactor A Protection System 2 High Main Steam Line Radiation $3 x background at rated B (7) (8) power (9) 2 High Drywell Pressure Same as Reactor A Protection System l 2/(Note 10) Condenser Low Vacuum 512" Hg absolute A 1 Trip System Logic --

A Amendment No. 9. 68, M, M, 90 45

VYNPS TABLE 3.2.2 (Cont'd)

HIGH PRESSURE COOLANT INJEC'" ION SYSTEM ISOLATION INSTRUMENTATION Minimum Number of Required Action When Operable Instrument Minimum Conditions channels per Trip for Operation are System Trip Function Trip Level Setting not Satisfied 2 per set of 4 High Steam Line Space $212oF Note 3 Temperature 1 High steam Line d/p (Steam 1195 inches of water Note 3 Line Break) 4 (Note 5) Low HPCI Steam Supply Pressure >70 psig Note 3 2 Main Steam Line Tunnel 1212 F Note 3 Temperature 1 Time Delay (23A-K48) 535 minutes Note 3 (23A-K49) 1 Bus Power Monitor -- --

1 Trip System Logic -- --

Amendment No. 66, 111 46

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 Trip Function Trip Level Setting not Satisfied (Note 2) 2 Main Steam Line Tunnel $212oF Note 3 Temperature 1 Time Delay (13A-K41) 535 minutes Note 3 (13A-K42) 2 per set of 4 High Steam Line Space $2120F Note 3 Temperature 1 High Steam Line d/p (Steam $195 inches of water Note 3 Line Break) l 4 (Note 5) Low Steam Supply Pressure >50 psig Note 3 1 Bus Power Monitor --

Note 3 1 Trip System Logic --

Note 3 1 Time Delay (13A-K7) 31t$7 seconds Note 3 (13A-K31) i Amendment No. O , 111 47

VYNPS TABLE 3.2.2 NOTES

1. The main steam line low pressure need be available only in the 'Run' mode.

2. If the minimum number of operable instrument channels are not available for one trip system, that trip system shall be tripped. If the minimum number of operable instrument channels are not available for both trip systems, the appropriate actions listed below shall be taken:

A. Initiate an orderly shutdown and have reactor in the cold shutdown condition in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

B. Initiate an orderly load reduction and have reactor in ' Hot Standby' within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

3. Close isolation valves in system and comply with Specification 3.5.

l 4. Deleted.

5. One trip system arranged in a one-out-of-two twice logic.

6. The main steam line high flow is available only in the ' Refuel,'

' Shutdown,' and 'Startup* modea.

7. This signal also automatically closes the mechanical vacuum pump suction line isolation valves.

8. Channel shared by the Reactor Protection and Primary Containment Isolation Systems.

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

10. A key lock switch is provided to permit the bypass of this trip function to enable plant startup and shutdown when the condenser vacuum is greater than 12 inches Hg absolute provided that both turbine stop and bypass valves are closed.

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Amendment No. 9, 69, 111 48

VYNPS TABLE J.2.3 REACTVR BUILDING VENTILATION ISOLATION & STANDBY GAS TREATMENT SYSTEM INITI ATION Minimum Number of Required Action hten Operable Instrument Minimum conditions Channels per Trip for Operation System Trio Function Trip Settino are not Met 2 Low Reactor Vessel Water oevel Same as PCIS Note 1 2 High Drywell Pressure Same as PCIS Note 1 1 Reactor Building Vent $14 mr/hr Note 1 1 Refueling Floor Zone Radiation $100 mr/hr Note 1 1 Reactor Building Vent Trip --

Note 1 System Logic 1 Standby Gas Treatment Trip --

Note 1 System Logic 1 Logic Bus Power Monitor --

Note 1 i

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.

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VYNPS TABLE 3.2.4 OFF-GAS SYSTD4 ISOLATION INSTRUMDJTATION Minimum Number of Required Action When 4

Operable Instrument Minimum Condition

Channels per Trip for Operation Svstem Trip Function Trip Setting Are Not Met 1 Time Delay (Stack Off-Gas 1 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 rieck monitoring systems is operable and off-gas system temperature and pressure are measured continucusly.

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Amendment No. 9, 83 50

VYNPS TABLE 3.2.5 CONTROL ROD BLOCK INSTPIYMENTATION Minimum Number of Modes in which Function Operable Instrument Must be operable Channels per Trip System Trip Function Refuel Startup Run Trip Settino l

Startup Range Monitor 2 a. Upscale (Note 2) X X 55 x 105 cps (Note 3) 2 b. Detector Not Fully Inserted X X l Intermediate Range Monitor (Note 1) 2 a. Upscale X X $108/125 Full scale 2 b. Downscale (Note 4) X X 15/125 Full Scale 2 c. Detector Not Fully Inserted X X Average Power Range Monitor 2 a. Upscale (Flow Bias) X $ 0.66(W-AW)+42% (Note 5) 2 b. Downscale X g2/125 Full Scale l Rod Block Monitor (Note 6)

(Note 9) 1 a. Upscale (Flow Bias) (Note 7) X $ 0.66(W-AW)+N (Note 5) 1 b. Downscale (Note 7) X 12/125 Full Scale l 1 Scram Discharge Volume X X X $12 Gallons (Note 8) (per volume) 1 Trip System Logic X X X l

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l Amendment No. 49, 45, 64, 66, M, 4 , M , 94, 131 51

VYNPS TABLE 3.2.5 NOTES

1. There shall be two operable or tripped trip systems for each function in i the required operating mode. If the minimum number of operable instruments are not available for one of the two trip systems, this condition may exist for up to seven days provided that during the time i the operable system is functionally tested immediately and daily thereafter; if the condition lasts longer than seven days, the system shall be tripped. If the minimum number of instrument channels are not available for both trip systems, the systems shall be tripped. i

2. One of these trips may be bypassed. The SRM function may be bypassed in the higher IRM ranges when the IRM upscale rod block is operable.

3. This function may be bypassed when count rate is y,,100 cps or when all IRM :

range switches are above Position 2. I

4. IRM downscale may be bypassed when it is on its lowest scale.

5. "W" is percent rated two loop dgive flow where 100% rated drive flow is that flow equivalent to 48 x 10 lbs/hr core flow. Refer to the Core Operating Limits Report for acceptable values for N. AW is the difference between the two loop and single loop drive flow at the same core flow. This difference must be accounted for during single loop operation. AW = 0 for two recirculation loop operation.

6. The minimum number of operable instrument channels may be reduced by one for maintenance and/or testing for periods not in excess of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in a

any 30-day period.

< 7. The trip may be bypassed when the reactor power is $30% of rated. An RBM ;

channel will be considered inoperable if there are less than half the ;

total number of normal inputs from any LPRM level.

8. With the number of operable channels less than required by the minimum operable channels per trip function requirement, place the inoperable channel in the tripped condition within one hour.

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9. With one RBM channel inoperable:

a. Verify that the reactor is not operating on a limiting control rod pattern, and ,

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

Otherwise, place the inoperable rod block monitor channel in the tripped )

condition within the next hour. l i

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Amendment No. M , M , M, %, H, 116 52 l i

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VYNPS TABLE 3.2.6 POST-ACCIDDJT INSTRUMDJT1.2 2ON Minimum Number of Operable Instrument l Channels Parameter Tvr>e of Indication Instrument Rance 2 Drywell Atmospheric Recorder #TR-16-19-45 0-3500F Temperature (Note 1) (TE-16-19-30A)

Meter *TI-16-19-30B 0-3500F 2 Containment Pressure (Note 1) Meter #PI-16-19-12A 0-275 psia Meter *PI-16-19-12B 0-275 psia 2 Torus Pressure (Note 1) Meter #PI-16-19-36A 0-80 psia Meter #PI-16-19-36B 0-80 psia 2 Torus Water Level (Note 3) Meter #LI-16-19-12A 0-25 ft.

Meter *LI-16-19-12B 0-25 ft.

2 Torus Water Temperature Meter #16-19-33A 0-250oF (Note 1) Meter #16-19-33C 0-250 F 2 Reactor Pressure (Note 1) Meter #PI-2-3-56A 0-1500 psig Meter #PI-2-3-56B 0-1500 psig 2 Reactor Vessel Water Level Meter #2-3-91A (-200)-0-(+200)"H 2O (Note 1) Meter 42-3-91B (-200)-0-(+200)"H O 2 1 Control Rod Position Meter 0-48* RPIS (Notes 1, 2) 1 Neutron Monitor (Notes 1, 2) Meter 0-125% Rated Flux 2 Torus Air Temperature (Note 1) Recorder *TR-16-19-45 0-3500F (TE-16-19-34)

Meter #TI-16-19-41 50-300oF 2/ valve Safety / Relief Valve Position Lights (SRV 2-71-A thru D) Closed - Open From Pressure Switches (Note 4)

Amendment No. Se, e, 90, 96, 113 53

VYNPS TABLE 3.2.6 (Cont'd)

POST-ACCIDEPTT INSTRUMDJTATION j

Minimum Number of Operable Instrument Channels Parameter Type of Indication Instrument Rance 1/ valve Safety Valve Position From Meter Z1-2-1A/B Closed - Open Acoustic Monitor (Note 5) 2 Containment Hydrogen / Oxygen Meter SR-VG-6A 0-30% hydrogen Monitor (Note 1) Meter SR-VG-6B 0-25% oxygen 2 Containment High-Range Meter RM-16-19-1A/B 1 R/hr-107 R/hr Radiation Monitor (Note 6) 1 1 Stack Noble Gas Effluent Meter RM-17-155 0.1 - 107 mR/hr (Note 7) t i

i i

Amendment No. (A , M, %, 98 54

VYNPS TABLE 3.2.6 NOTES Note 1 - From and after the date that a parameter is reduced to one indication, operation is permissible for 30 days. If a parameter is not indicated in the Control Room, continued operation is permissible during the next seven days. If indication cannot be restored within the next six hours, 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 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months .

Note 2 - Control rod position and neutron monitor instruments are considered to be redundant to each other.

Note 3 - From and after the date that this parameter is reduced to one indication in the Control Room, continued reactor operation is permissible during the next 30 days. If both channels are inoperable and indication cannot be restored in six hours, 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 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months .

Note 4 - From and after the date that safety / relief valve position from pressure switches is unavailable, reactor operation may continue provided safety / relief valve position can be determined from Recorder #2-166 (steam temperature in SRVs, 0-6000F) and Meter 16-19-33A or C (torus water temperature, 0-250 F). If both rsrameters are not available, the reactor shall be in a hot shutdown condition in six hours and a cold shutdown condition in the following 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months .

Note 5 - From and after the date that safety valve position from the acoustic monitor is unavailable, reactor operation may continue provided safety valve position can be determined from Recorder #2-166 (thermocouple, 0-600 F) and Meter #16-19-12A or B (containment pressure 0-275 psia). If both indications are not available, the reactor shall be in a hot shutdown condition in six hours and in a cold shutdown condition in the following 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months .

Note 6 - Within 30 days following the loss of one indication, or seven days following the loss of both indications, restore the inoperable channel (s) to an operable status or a special report to the Commission pursuant to Specification 6.7 must be prepared and submitted within the subsequent 14 days, outlining the action taken, the cause of the inoperability, and the plans and schedule for restoring the system to operable status.

Note 7 - From and after the date that this parameter is unavailable by Control Room indication, and cannot be restored within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , continued reactor operation is permissible for the next 30 days provided that local sampling capacity is available. If the Control Room indication cannot be restored within 30 days, the reactor shall be in hot shutdown within six hours and in cold shutdown within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Amendment No. 50, 64, 94, 96, +44, 131 55

l i

i i VYNPS 1

TABLE 3.2.8 f

EMEPGENCY BUS UNDERVOLTAGE INSTRUMENTATION Minimum l

Number of i Operable Inst rurnent s Parameter Trip Setting Requirad 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 1 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 circuits within one hour.

2. If the minimum number of operable instrument channels are not available, reactor power operation is permissible for only 7 successive days unless the system is sooner made operable.

l Amendment No. 98 56

VYNPS .

TABLE 3.2.9 REAC'IVR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION Minimum Number of Required Action When Operable Instrument Minimum Conditions Channels per Trip for Operation System Trip Function Trip Level Settino are not Satisfied 2 (Note 1) Low-Low Reactor Vessel Water >82.5" Above Top of Note 4 Level Enriched Fuel 2 (Note 2) Low Condensate Storage Tank >3% Note 4 Water Level 2 (Note 3) High Reactor Vessel Water $177* Above Top of Note 4 Level Enriched Fuel 1 Bus Power Monitor --

Note 4 i 1 Trip System Logic --

Note 4 l

l i

t Amendment No. 111 57 i

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

VYNPS TABLE 3.2.9 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. l l

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

i P

l l

I t

i t

f i

Amendment No. 111 58

VYNPS TABLE 4.2.1 MINIMUM TEST AND CALIPPATION FREOUENCIES EMERGENCY CORE COOLING ACTUATION INSTRUMENTATION Core Spray System Trio Function Functional Test (8) Calibration (8) Instrument Check High Drywell Pressure (Note 1) Once/ Operating Cycle Once Each Day Low-Low Reactor Vessel (Note 1) Once/ Operating Cycle Once Each Day Water Level Low Reactor Pressure (Note 1) Once/ Operating Cycle --

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

Low Reactor Pressure (Note 1) Once/ Operating Cycle --

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

52C/D(M))

Pump 14-1A, Discharge (Note 1) Every Three Months --

Pressure Auxiliary Power Monitor (Note 1) Every Refueling Once Each Day Pump Bus Power Monitor (Note 1) None Once Each Day High Sparger Pressure (Note 1) Every Three Months --

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

(Note 3) f Amendment No. GB, M, -146, 110 59

. . . + . - . - _ _____ _______m , -

VYNPS TABLE 4.2.1 (Cont'd)

MINIMUM TEST AND CALIPPATION FREQUE'!CIES EMERGENCY CORE COOLING ACTUATION INSTRUME?TTATION Low Pressure Coolant Injection System Trip Function Functional Test (8) Calibration (8) Instrument Check Low Reactor Pressure (Note 1) Once/ Operating Cycle --

(PT-2-3-56C/D(S1))

High Drywell Pressure (Note 1) Once/ Operating Cycle Once Each Day (PT-10-101A-D(M))

Low-Low Reactor Vessel (Note 1) Once/ Operating Cycle Once Each Day Water Level Reactor Vessel Shroud (Note 1) Once/ Operating Cycle --

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

(PT-2-128A/B)

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

Pressure High Drywell Pressure (Note 1) Once/ Operating Cycle --

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

Low Reactor Pressure (Note 1) Once/ Operating Cycle --

( PT-2-3-56A/B) (S1) &

52C/D(M))

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) 1 Amendment No, M, 50, M, M M6, 4M, 129 60

VYNPS TABLE 4.2.1 (Cont'd)

MINIMUM TEST AND CALIBRATION FREOUENCIES EMERGENCY CORE COOLING ACTUATION INSTRUMENTATION High Pressure Coolant Injection System

( Trio Function Functlona1 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) Every three months --

Tank Water Level High Drywell Pressure (Note 1) Once/ operating cycle Once each day Bus Power Monitor (Note 1) None Once each day Trip System Logic Once/ operating cycle Once/ Operating cycle --

(Note 3)

High Reactor Vessel (Note 1) Once/ operating cycle --

Water Level i

i' Amendment No. 44, GB, M, 66, f G6, 444, 111 61

-.----__---. _.___----.-___._-_ - - - __ - --.__ - -----.. .a - --- - -_--- < - - - - _ _ _ - --m -- - --_---

I I VYNPS TABLE 4.2.1 (Cont'd)

MINIMUM TEST AND CALIBRATION FREOUENCIES EMERGENCY CORE COOLING ACTUATION INSTRUMENTATION Automatic Depressurization System Trip Function Funetiona1 Test (8) Calibration (8) Instrument Check Low-Low Reactor Vessel (Note 1) Once/ Operating Cycle Once Each Day Water Level High Drywell Pressure (Note 1) 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 (Not=s 2 and 11) (Note 3)

Valves) 4 Amendment No. M, M, M5, 106 62 i

VYNPS TABLE 4.2.1 (Cont'd)

MINIMUM TEST AND CALIBRATION FPEOUENCIES EMERGENCY CORE COOLING ACTUATION INSTRUMENTATION Recirculation Pump Trip Actuation System Trio Function _ Functional Test (8) Calibration (8) Instrument Check Low-Low Reactor Vessel (Note 1) Once/ Operating Cycle Once Each Day Water Level (4)

IIigh Reactor Pressure (Note 1) Once/ Operating Cycle Once Each Day ;

(4) l Trip System Logic Once/ Operating Cycle Once/ Operating Cycle --

}

l l

I J

Amendment No. G8, 106 63

I VYNPS TABLE 4.2.2 j MINIMUM TEST AND CALIBRATION FREQUENCIES PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION ,

Trio Function Functional Test (8) Calibration (8) Instrument Check Low-Low Reactor Vessel (Note 1) Once/ Operating Cycle Once Each Day Water Level High Steam Line Area (Note 1) Each Refueling Outage --

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

Pressure t

Low Reactor Vessel Water (Note 1) Once/ Operating Cycle --

Level High Main Steam Line (Notes 1 and 7) Each Refueling Outage Once Each Day Radiation High Drywell Pressure (Note 1) 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)

L I

i Amendment No. 9,58, %,4M, 133 64

VYNPS TABLE 4.2.2 (Cont'd)

MINIMUM TEST AND CALIBRATION FREOUENCIES HIGH PRESSURE COOLANT INJECTION SYSTDf ISOIATION INSTRUMENTATION Trip Function Nnctional Test (8) Calibration (8) Instrument Check l High Steam Line Space (Note 1) Each refueling outage --

Temperature High Steam Line D/P (Note 1) Every three months --

(Steam Line Break)

Low HPCI Steam Supply (Note 1) Every three months --

Pressure Main Steam Line Tunnel (Note 1) Each refueling outage --

Temperature Bus Power Monitor (Note 1) None Once each day Trip System Logic Once/ operating cycle Once/ operating cycle --

(Note 3) i y

l ,

Amendment No. GB, M6, 4M, 111 65

VYNPS TABLE 4.2.2 (Cont'd)

MINIMUM TEST AND CAL 1BRATION FREOUENCIES REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION INSTRUMENTATION Trip Function Functional Test (8) Ca1ibration(B) Instrument Check Main Steam Line Tunnel (Note 1) Each refueling outage --

Temperature High Steam Line Space (Note 1) Each refueling outage --

Temperature High Steam Line d/p (Note 1) Every three months --

including time delay relays (Steam Line Break) l Low RCIC Steam Supply (Note 1) Every three months --

Pressure Bus Power Monitor (Note 1) None Once each day Trip System Logic Once/ operating cycle Once/ operating cycle --

(Note 3) s Amendment No. 58, O , 446, +M, 111 66 m _ _ _ _ _ _ _ _ _ _ _ _ __ _ - - . _ , . - ~ - -- _ . _ _ . . _ . -- - . ~ . - __ - - . . . . - . _ - , _ _ _ _ _ . _ _

VYNPS TABLE 4.2.3 MINIMUM TEST AND CALIPRATION FPEOUENCIES REACTOR BUILDING VENTILATION AND STANDBY GAS TREA'INE?TT SYSTEM ISOLATION Trip Function Functional Test (8) Calibration (8) Instrument Check Low Reactor Vessel Water (Note 1) Once/ Operating Cycle --

Level High Drywell Pressure (Note 1) 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 --

Trip System Logic (Note 3)

Standby Gas Treatment Once/ Operating Cycle Once/ Operating Cycle --

Trip System Logic (Note 3)

Logic Bus Power Monitor (Note 1) None Once Each Day 1

i Amend:nent No. 58, %, 406, 110 67

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VYNPS ,

I TABLE 4.2.5 MINIMUM TEST AND CALIBRATION FREOUENCIES CONTROL FOD BLOCK INSTRUMENTATION Trio Function Functional Test Calibration Startup Range Monitor

a. Upscale Notes 4 and 6 Note 6 l b. Detector Not Fully Inserted' Note 6 NA Inte. mediate Range Monitor

a. Upscale Notes 4 and 6 Note 6

b. Downscale Notes 4 and 6 Note 6 l c. Detector Not Fully Inserted Note 6 NA Average Power Range Monitor

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

b. Downscale Notes 1 and 4 Every Three Months Rod Block Monitor

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

b. Downscale Notes 1 and 4 Every Three Months Trip System Logic Once/ Operating Cycle Once/ Operating Cycle (Note 3)

High Water Level in Scram Discharge Every Three Months Refueling Outage Volume 1

Amendment No. 446, 444, 136 69

3-4 VYNPS TABLE 4.2.6 CALIBRATION REQUIREMEFES POST-ACCIDENT INSTRUMENTATION Parameter Calibration Instrument Check Drywell Atmosphere Temperature Every 6 Months Once Each Day Containment Pressure Once/ Operating Cycle Once Each Day

, l Torus Pressure Once/ Operating Cycle Once Each Day Torus Water Level Once/ Operating Cycle Once Each Day i Torus Water Temperature Every 6 Months Once Each Day Reactor Pressure Once/ Operating Cycle Once Each Day Reactor Vessel Water Level once/ Operating Cycle Once Each Day Control Rod Position (Note 5) Once Each Day Neutron Monitor Same as Reactor Protection Systems Once Each Day Torus Air Temperature Every 6 Months once Each Day Safety / Relief Valve Position Every Refueling Outage (Note 9) Once Each Day (a Functional Test to be performed quarterly) i Safety Valve Position Every Refueling Outage (Note 9) Once Each Day (a Functional Test to be

performed quarterly) 1 l

l i

t Amendment No. M, G, M, 113 70

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

_ _ = .

VYNPS TABLE 4.2.6 CALIBRATION PEOUIRD4DTTS POST-ACCIDENT INSTRUMD7TATION (Cont'd)

Parameter Calibration Instrument Check Containment Hydrogen / Oxygen Monitor Once/ Operating Cycle Once each day Containment High-Range Radiation Once/ Operating Cycle Once each day Monitor Stack Noble Gas Effluent Every Operating Cycle (a Functional Once each day Test to be performed quarterly)

Amendment No. 96, 98 71

VYNPS L

TABLE 4.2.8 EMERGENCY BUS UNDERVOLTAGE INSTRUMENTATION Trip System Functional Test Calibration (8) '

Degraded Bus Voltage See Note 10 Once/ Operating Cycle i

l l

Amendment No. 98 72'

VYNPS TABLE 4.2.9 MINIF E TEST AND CALIPRATION FREOUENCIES REACTOR CORE ISOLATION COOLIPC SYSTEM ACTUATION INSTRUMENTATION Trip 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)

L Amendment No. 111 73

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

VYNPS TABLE 4.2 NOTES

1. Initially once per months thereafter, a longer interval as determined by test results on this type of instrumentation.

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

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

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

5. Check control rod position indication while performing the surveillance requirement of Section 3.3.

1

6. Functional tests, calibrations, and instrument checks are not required when these instruments are not required to be operable or are tripped.

Functional tests shall be performed before each startup with a required 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 ins truments are required to be operable.

7. This instrumentation is excepted from the functional test definitions and shall be calibrated using simulated electrical signals once 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. 64, 96, 105 74

VYNPS BASES:

3.2 PROTECTIVE INSTRUMENTATION In addition to reactor protection instrumentation which initiates a reactor scram, station protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the reactor operator's ability to control, or terminate a single operator error before it results in serious consequences. This set of Specifications provides the limiting conditions of operation for the primary system isolation function and initiation of the core standby cooling and standby gas treatment systems. The objectives of the Specifications are (i) to assure the effectiveness of any component of such systems even during periods when portions of such systems are out of service for maintenance, testing, or calibration, and (ii) to prescribo the trip settings required to assure adequate performance. This set of Specifications also provides the limiting conditions of operation for the control rod block system and surveil. lance instrumentation.

Isolation valves (Note 1) are installed in those lines that penetrate the primary containment and must be isolated during a loss-of-coolant accident so that the radiation dose limits are not exceeded during an accident. condition. Actuation of these valves is initiated by protective instrumentation shown in Table 3.2.2 which senses the conditions for which isolation is required. Such instrumentation must be available whenever primary containment integrity is required. The objective is to isolate the primary containment so that the limits of 10 CFR 100 are not exceeded during an accident. The objective of the low turbine condenser vacuum trip is to minimize the radioactive effluent releases to as low as practical in case of a main condenser failure. Subsequent releases would continue until operator action was taken to isolate the main condenser unless the main steam line isolation valves were closed automatically on low condenser vacuum. The manual bypass is required to permit initial startup of the reactor during low power operation.

The instrumentation which initiates primary system isolation is connected in a dual channel arrangement. Thus, the discussion given in the bases for Specification 3.1 is applicable here.

The low reactor water level instrun s:tation is set to trip when reactor water level is 127' above the top of the enriched fuel. This trip initiates closure of Group 2 and 3 primary containment isolation valves.

For a trip setting of 127' above the top of the enriched fuel, the valves will be closed before perforation of the clad occurs even for the maximum break and, therefore, the setting is adequate.

The top of the enriched fuel (351.5" f rom vessel bottom) is designated as a common reference level for all reactor water level instrumentation.

The intent is to minimize the potential for operator confusion which may result from different scale references.

The low-low reactor water level instrumentation is set to trip when reactor water level is 82.5" H O 2 indicated on the reactor water level instrumentation above the top of the enriched fuel. This trip initiates closure of the Group 1 primary containment isolation valves and also l activates the ECCS and RCIC System and starts the standby diesel generator system. This trip setting level was chosen to be low enough to prevent spurious operation, but high enough to initiate ECCS operation and primary system isolation so that no melting of the fuel cladding will occur, and so that post-accident cooling can be accomplished and the limits of 10CFR100 will not be violated. For the complete circumferential break of 28-inch recirculation line and with the trip Note 1 - Isolation valves are grouped as listed in Table 3.7.1.

Amendment No. GB, 66, 111 75

VYNPS BASES: 3.2 (Cont'd) setting given above, ECCS initiation and primary system isolation are initiated in time to meet the above criteria. The instrumentation also covers the full range of spectrum breaks and meets the above criteria.

The high drywell pressure instrumentation is a backup to the water level instrumentation, and in addition to initiating ECCS, it causes isolation of Group 2, 3, and 4 isolation valves. For the complete circumferential break discussed above, this instrumentation will initiate ECCS operation at about the same time as the low-low water level instrumentation, thus, the results given above are applicable here also. Group 2 isolation valves include the drywell vent, purge, and sump isolation valves. High drywell pressure activates only these valves because high drywell pressure could occur as the result of nonsafety-related causes such as

) not purging the drywell air during startup. Total system isolation is not desirable for these conditions and only the valves in Group 2 are required to close. The water level instrumentation initiates protectica l for the full spectrum of loss-of-coolant accidents and causes a trip vt !

! all primary system isolation valves.

Venturis are provided in the main steam lines as a means of measuring )

steam flow and also limiting the loss of mass inventory from the vessel '

during a steam line break accident. In addition to monitoring steam l flow, instrumentation is provided which causes a trip of Group 1 isolation valves. The primary function of the instrumentation is to detect a break in the main steam line, thus only Group 1 valves are closed. For the worst case accident, main steam line break outside the drywell, this trip setting of 140 percent of rated steam flow in conjunction with the flow limiters and main steam line valve cloaure limit the mass inventory loss such that fuel is not uncovered, cladding temperatures remain less than 12950F and release of radioactivity to the l

environs is well below 10CFR100.

Temperature monitoring instrumentation is provided in the main steam line tunnel to detect leaks in this area. Trips are provided on this instrumentation and when exceeded cause closure of Group 1 isolation valves. Its setting of ambient plus 950F is low enough to detect leaks of the order of 5 to 10 gpm; thus, it is capable of covering the entire spectrum of breaks. For large breaks, it is a backup to high steam flow instrumentation discussed above, and for small breaks, with the resultant small release of radioactivity, gives isolation before the limits of 10CFR100 are exceeded.

High radiation monitors in the main steam line tunnel have been provided to detect gross fuel failure resulting from a control rod drop accident.

This instrumentation causes closure of Group 1 valves, the only valves required to close for this accident. With the established setting of 3 times normal background and main steam line isolation valve closure, fission product release is limited so that 10CFR100 limits are not exceeded for the control rod drop accident, and 10CFR20 limits are not exceeded for gross fuel failure during reactor operations. With an alarm setting of 1.5 times normal background, the operator is alerted to possible gross fuel failure or abnormal fission product releases from failed fuel due to transient reactor operation.

Pressure instrumentation is provided which trips when main steam line pressure drops below 800 psig. A trip of this instrumentation results in closure of Group 1 isolation valves. In the refuel, shutdown, and startup modes, this trip function is provided when main steam line flow exceeds 40% of rated capacity. This function is provided primarily to provide protection against a pressure regulator malfunction which would cause the control and/or bypass valves to open, resulting in a rapid depressurization and cooldown of the reactor vessel. The 800 psig trip Amendment No. GG, 66, 68, 64, 86 76

VYNPS PASES: 3.2 (Cont'd) setpoint limits the depressurization such that no excessive vessel thermal This stresswas setpoint occurs as a result of a pressure regulator malfunction pressures selected far enough below normal main steam line .

to avoid spurious primary containment isolations.

Low condenser vacuum has been added as a trip of the Group 1 isolation valves through to prevent release of radioactive gases from the prinary coolant condenser.

selected to provide sufficient margin to assureThe setpoint of 12 inches of mercury absolute was the condenser when gas flow is stopped and sufficientretention operating values. margincapability in below normal The HPCI and/or RCIC high flow, steam supply pressure, and temperature instrumentation is provided to detect a break in the HPCI and/or RCIC piping.

and/or RCIC Tripping of this isolation instrumentation valves, i.e.,

results in actuation of HPCI Group 6 valves. A time delay has beaa incorporated into the RCIC steam flow trip logic to prevent the system occur onfrom inadvertently isolating due to pressure spikes which may startup.

prevanted and fissien productThe trip settings are such that core uncovering is release is within limits.

The chan, tstrumentation system. which initiates ECCS action is arranged in a dual to trip instrumentPermanentlychannels. installed circuits and equipment may be used energizing the circuitry, tripping In thean nonfail safe systems which require instrument form of providing the channel may take the installed circuits. This required relay function by use of permanently circuits with the aid of the permanently installed testis accomplished in some cases by closing jacks or other circuitry which would be installed for this purpose.

The control rod block functions are provided to prevent excessive control rod withdrawal so that MCPR does not decrease below the fuel cladding integrity e.g.,

any safety limit.

trip on one Thesixtrip of the logic for this function is 1 out of n; APRMs, in a rod block. six IRMs or four SRMs will result may be reduced by The one minimum instrument channel requirements for the IRM maintenance, testing orfor a short period of time to allow for calibration.

aid o,.y and is not needed for rod withdrawal.The RBM is an operational guide and For m.ngle recirculation loop operation, the RBM trip setting is reduced in accordance with the analysis presented in HEDo-30060, February 1983.

This two loopadjustment drive flow accounts for thecore at the same difference flow, between the single loop and safety is not reduced during single loop operation.and ensures that the margin of The APRM rod block trip is flow referenced and prevents a significant reductiongross provides in MCPR coreespecially protection;during operation at reduced flow. The APRM i.e., limits the gross core power ir s.- ease.nce. from The withdrawal of control rods in the normal withdrawal fuei cladding integrity safety limit. trips are set so that MCPR is maintained greater than the operation, For single recirculation loop the APRM rod block the analysis presented in NEDE-30060, February trip setting is reduced in accordance with 1983. This adjustment accounts for the difference between the flow at the same core flow, and ensures that the margin single loop and two-loop drive reduced during single loop operation. of safety is not The IRM rod block function provides protection. The scaling arrangementlocal as well as gross core is such that trip setting is less than a factor worst-case of 10 above the indicated level. Analysis of the accident j the fuel cladding integrity safety limit.results in rod block action before MCPR approaches Amendment No. 9, M, 69, 84, 94

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VYNPS BASES: 3.2 (Cont'd)

A downscale indication on an APRM or IRM is an indication the instrument has failed or the instrument is not sensitive enough. In either case, the instrument will not respond to changes in control rod motion and thus control rod motion is prevented.

To prevent excessive clad temperatures for the small pipe break, the HPCI or Automatic Depressurization System must function since, for these breaks, reactor pressure does not decrease rapidly enough to allow either I core spray or LPCI to operate in time. For a break or other event

( occurring outside the drywell, the Automatic Depressurization System is initiated on low-low reactor water level only after a time delay. The arrangement of the tripping contacts is such as to provide this function when necessary and minimize spurious operation. The trip settings given ,

in the Specificaticn are adequate to ensure the above criteria are met. l The Specification preserves the effectiveness of the system during i periods of maintenance, testing, or calibration, and also minimizes the l risk of inadvertent operation; i.e., only one instrument channel out of service.

The ADS is provided with inhibit switches to manually prevent automatic i initiation during events where actuation would be undesirable, such as !

certain ATWS events. The system is also provided with an Appendix R l inhibit switch to prevent inadvertent actuation of ADS during a fire I which requires evacuation of the Control Room, i l l Four radiation monitors are provided which initiate isolation of the reactor building and operation of the standoy gas treatment system. The monitors are located in the reactor building ventilation duct and on the refueling floor. Any one upscale trip or two downscale trips of either set of monitors will cause the desired action. Trip settings for the monitors on the refueling floor are based upon initiating normal ventilation isolation and standby gas treatment system operation so that none of the activity released during the refueling accident leave the Reactor Building via the normal ventilation stack but that all activity is processed by the standby gas treatment system. Trip settings for the monitors in the ventilation duct are based upon initiation of the normal ventilation isolation and standby gas treatment system operation at a radiation level equivalent to the maximum site boundary dose rate of 500 mrem / year as given in Specification 3.8.E.1.a. The monitoring system in the plant stack represents a backup to this system to limit gross radioactivity releases to the environs.

The purpose of isolating the mechanical vacuum pump line is to limit release of radioactivity from the main condenser. During an accident, fission products would be transported from the reactor through the main steam line to the main condenser. The fission product radioactivity would be sensed by the main steam line radiation monitors which initiate ;

isolation.

Post-accident instrumentation parameters for Containment Pressure, Torus Water Level, Containment Hydrogen / Oxygen Monitor, and Containment High-Range Radiation Monitor, are redundant, environmentally and I seismically qualified instruments provided to enhance the operators' ability to follow the course of an event. The purpose of each of these instruments is to provide detection and measurement capability during and following an accident as required by NUREG-0737 by ensuring continuous on-scale indication of the following: containment pressure in the O to 275 paia range; torus water level in the O to 25 foot range (i.e., the bottom to 5 feet above the nornal water level of the torus pool);

containment hydrogen / oxygen concentrations (0 to 30% hydrogen and 0 to 25% oxygen); and containment radiation in the 1 R/hr to 10 R/hr gamma.

Amendment No. 9, 84, 96, 144, 145, 444, 132 78

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

VYNPS BASES: 3.2 (Cont'd) t The Degraded Grid Protective System has been installed to assure that safety-related electrical equipment will not be subjected to sustained degraded voltage. This system incorporates voltage relays on 4160 Volt Emergency Buses 3 and 4 which are set to actuate at the minimum voltage required to prevent damage of safety-related equipment.

If Degraded Grid conditions exist for 10 seconds, either relay will actuate an alarm to alert operators of this condition. Based upon an assessment of these conditions the operator may choose to manually disconnect the off-site power. In addition, if an ESF signal is initiated in conjunction with low voltage below the relay setpoint for 10 seconds, the off-site power will be automatically disconnected. ;

The Reactor Core Isolation Cooling (RCIC) System provides makeup water to the reactor vessel during shutdown and isolation to supplement or replace the normal makeup sources without the use of the Emergency Core cooling Systems. The RCIC System is initiated automatically upon receipt of a reactor vessel low-low water level signal. Reactor vessel high water level signal results in shutdown of the RCIC System. However, the system will restart on a subsequent reactor vessel low-low water level signal.

The RCIC System is normally lined up to take suction from the condensate i storage tank. Suction will automatically switch over from the condensate i storage tank to the suppression pool on low condensate storage tank level.

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Amendment No. 98, 111 79 i 1

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VYNPS BASES:

4.2 PROTECTIVE INSTRUMENTATION The Protective Instrumentation Systems covered by this Specification are listed in Table 4.2. Most of these protective systems are composed of two or more independent and redundant subsystems which are combined in a dual-channel arrangement. Each of these subsystems contains an arrangement of electrical relays which operate to initiate the required system protective action.

The relays in a subsystem are actuated by a number of means, including manually-operated switches, process-operated switches (sensors), bistable devices operated by analog sensor signals, timerse limit switches, and other relays. In most cases, final subsystem relay actuation is obtained by satisfying the logic conditions established by a number of these relay contacts in a logic array. When a subsystem is actuated, the final subsystem relay (s) can operate protective equipment, such as valves and pumps, and can perform other protective actions, such as tripping the main turbine generator unit.

With the dual-chann91 arrangement of these subsystems, the single failure of a ready circuit can be tolerated because the redundant subsystem or system (in the care of high pressure coolant injection) will then initiate the necessary protective action. If a failure in one of these circuits occurs in such a way that an action is taken, the operater is immediately alerted to the failure. If the failure occurs and causes no action, it could then remain undetected, causing a loss of the redundancy in the dual-channel arrangement. Losses in redundancy of this nature are found by periodically testing the relay circuits and contacts in the subsystems to assure that they are operating properly.

It has been the practice in boiling water reactor plants to functionally test protective instrumentation sensors and sensor relays on-line on a monthly frequency. Since logic circuit tests result in the actuation of plant equipment, testing of this nature was done while the plant was shut down for refueling. In this way, the testing of equipment would not l jeopardize plant operation.

This Specification is a periodic testing program which is based upon the overall testing of protective instrumentation systems, including logic circuits as well as sensor circuits. Table 4.2 outlines the test, calibration, and logic system functional test schedule for the protective instrumentation systems. The testing of a subsystem includes a functional test of each relay wherever practicable. The testing of each relay includes all circuitry necessary to make the relay operate, and also the proper functioning of the relay contacts. Testing of the automatic initiation inhibit switches verifies the proper operability of the switches and relay contacts. Functional testing of the inaccessible temperature switches associated with the isolation systems is accomplished remotely by application of a heat source to individual switches.

All subsystems are functionally tested, calibrated, and operated in their entirety.

Amendment No. ++, 96, 96, 146, 106 80

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1 VYNPS 3.3 LIMITING CONDITIONS FOR l 4.3 SURVEILLANCE REQUIREMENTS l OPERATION 3.3 CONTROL ROD SYSTEM 4.3 CONTROL ROD SYSTEM Applicability: Applicability:

Applies to the operational Applies to the surveillance i status of the control rod requirements of the control rod

)

system. system.

Obiective Obiective:

To assure the ability of the To verify the ability of the control rod system to control control rod system to control reactivity. reactivity.

Specification: Specification:

A. Reactivity Limitations A. Reactivity Limitations

1. Reactivity Marcin - Core 1. Reactivity Marcin - Core Loadino Loading The core loading shall Control rods shall be I be limited to that which withdrawn following a can be made subcritical refueling outage when 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 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.

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2. Reactivity Marcin - 2. Reactivity Marcin -

Inoperable Control Rods Inoperable Control Rods Control rod driven which Each partially or fully cannot be moved with withdrawn operable control rod drive control rod shall be l 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 3 not due to a failed cannot be moved and for l control rod drive which control rod drive i mechanism collet mechanism damage has not I housing. The control been ruled out. The i 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 i I

i Amendment No. 20 81

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VYNPS I 3.3 LIMITING CONDITIONS FOR 4.3 SURVEILLANCE REQUIREMENTS OPERATION disarmed electrically of inoperable rods has except for control rods been reduced to less which are inoperable than two and if it has because of scram times been demonstrated that greater than those control rod drive specified in mechanism collet housing Specification 3.3.C. In failure is not the cause no case shall the number of an immovable control of inoperable rods which rod, are not fully inserted be greater than six during power operation.

B. Control Rods B. Control Rods

1. Each control rod shall 1. The coupling integrity be either coupled to its shall be verified:

drive or placed in the inserted position and (a) When a rod is its directional valves withdrawn the first >

disarmed electrically, time subsequent to hben removing up to one each refueling cottrol rod drive per outage or after quadrant for inspection maintenance, and the reactor is in observe discernable the refueling mode, this response of the requirement does not nuclear apply. instrumentation; however, for initial rods when response is not discernable, subsequent exercising of these rods after the reactor is critical shall be performed to verify instrumentation response; and (b) When a rod is fully withdrawn, observe that the rod does not go to the over-travel position. Prior to startup following a refueling outage, each rod shall be fully withdrawn I continuously to l observe that the rate of withdrawal is proper and that the rod does not go to the over-travel position.

Following uncoupling, each control rod drive and blade shall be tested to verify

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1 Amendment No. 20 82 i

VYNPS 3.3 LIMITING CONDITIONS FOR 4.3 SURVEILLANCE REQUIREMENTS OPERATION I' positive coupling and the results of each test shall be i recorded. This test shall consist of checking the operability of the over-travel indicator circuit prior to coupling ,

by withdrawing the

( drive snd observing the over-travel i light. The drive '

and blade shall then be immediately coupled and fully withdrawn. The position and over-travel lights shall be observed.

2. The Control Rod Drive 2. The Control Rod Drive Housing Support System Housing Support System shall be in place when shall be inspected after the Reactor Coolant reassembly and the J System is pressurized results of the above atmospheric inspection recorded.

pressure with fuel in the reactor vessel unless all operable control rods are fully inserted.

3. While the reactor is 3. Prior to control rod below 20% power, the Rod withdrawal for startup Worth Minimizer (RWM) the Rod Worth Minimizer shall be operating while (RWM) shall be verified moving control rods as operable by except that: performing the followings (a) If after withdrawal (a) The Reactor of at least 12 Engineer shall control rods during verify that the a startup, the RWM control rod fails, the startup withdrawal sequence may continue for the Rod Worth provided a second Minimizer computer licensed operator is correct.

verifies that the operator at the reactor console is following the ,

control rod j program; or l l

(b) If all rods, except (b) The Rod Worth )

those that cannot Minimizer i be moved with diagnostic test l control rod drive shall be performed.

Amendment No. 39 83 l

1 t-------------- - - - - - - - - _ . _ _ _ _ _ _ _ _ - . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _________ _ _ _ _ _ _ _

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

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

(d) An out-of-sequence control rod shall be withdrawn no more than three l

notches and the rod l block function verified.

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

280 cal /g is not exceeded.

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

l three counts per second. rate of at least three counts per second.

6. During operation with 6. When a limiting control limiting control rod rod pattern exists, an patterns either instrument functional i test of the RBM shall be 1

(a) Both RBM channels performed prior to shall be operable; withdrawal of the or designated rod (s) and daily thereafter.

l (b) Control rod withdrawal shall be blocked; or Amendment No. 19, 61 84

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VYNPS '

i 3.3 LIMITING CONDITIONS FOR 4.3 SURVEII. LANCE REQUIREMENTS OPERATION (c) The operating pcuer 7. The scram discharge level shall be volume drain and vent limited so that the valves shall be verified MCPR will remain open at least once per above the fuel month. These valves may cladding integrity be closed intermittently safety limit for testing under assuming a single administrative control, error that results in complete withdrawal of any single operable control rod.

C. Scram Insertion Timen C. Scram Insertion Times 1.1 The average scram time, 1. After refueling outage based on the and prior to operation de-energization of the above 30% power with scram pilot valve reactor pressure above solenoids of all 800 psig all control operable control rods in rods shall be subject to the reactor power scram-time measurements operation condition from the fully withdrawn shall be no greater position. The scram thans times for single rod scram testing shall be .

Drop-Out % Inserted Avg. Scram measured without of From Fully Inserticn reliance on the control Position withdrawn Time (sec) rod drive pumps.

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

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

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

Amendment No. 44, 25, 49, 74, 73 85 i

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VYNPS 3.3 LIMITING CONDITIONS FOR 4.3 SURVEILLANCE REQUIREMENTS OPERATION 1.2 If Specification ,

3.3.C.1.1 cannot be met, l the average scram time, based on the de-energization of the scram pilot valve solenoids of all operable control rods in the reactor power operation condition shall be no greater than:

Drop-Out % Inserted Avg. Scram of From Fully Insertion Position Withdrawn Time (sec) 46 4.51 .358 36 25.34 1.096 26 46.18 1.860 06 87.84 3.419 The average of the scram insertion times for the three fastest control rods of all groups of four control rods in a two by two array shall be no greater than Drop-Out % Inserted Avg. Scram )

of From Fully Insertion Position Withdrawn Time (sec) 46 4.51 .379 36 25.34 1.164 26 46.18 1.971 06 87.84 3.624

2. The maximum scram insertion time for 90%

insertion of any operable control rod shall not exceed 7.00 seconds.

l Amendment No. 34, 70 86

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VYNPS 1 l

l 3.3 LIMITING CONDITIONS FOR 4.3 SURVEILLANCE REQUIREMENTS l OPERATION )

)

3. If Specification ,

3.3.C.1.2 cannot be met, i the reactor shall not be made supercritical; if operating, the reactor shall be shut down immediately upon determination that average scram time is deficient.

4. If Specification 3.3.C.2 cannot be met, the deficient control rod shall be considered inoperable, fully inserted into the core, and electrically disarmed.

D. Control Rod Accumulators D. Control Rod Accumulators At all reactor operating Once a shift check the <

pressures, a rod accumulator status of the pressure and may be inoperable provided level alarms for each that no other control rod in accumulator.

the nine-rod square array around this rod has a

1. Inoperable accumulator. I f

2. Directional control valve electrically disarmed while in a non-fully inserted i position. '

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3. Scram insertion greater '

than maximum permissible insertion time.

If a control rod with an inoperable accumulator is inserted " full-in' and its directional control valves are electrically disarmed, it shall not be considered to have an inoperable l accumulator. '

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l Amendment No. 70 87 l i

VYNPS 3.3 LIMITING CONDITIONS FOR 4.3 SURVEILLANCE REQUIREMENTS OPERATION E. Reactivity Anomallem 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 1% 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 l appropriate. power operating conditions, the critical rod F. If Specification 3.3A configuration will be through D 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.

Amendment No. 39 88

m ._.

l VYNPS l I

l BASES: 1 3.3 & 4.3 CONTROL ROD SYSTEM A. Reactivity Limitations

1. Reactivity Marcin - Core Loading i

The core reactivity limitation is a restriction to be applied l principally to the design of new fuel which may be loaded in the core or into a particular refueling pattern. Satisfaction of the limitation can only be demonstrated at the time of loading and must be such that it will apply to the entire subsequent fuel cycle. At each refueling the reactivity of the core loading will be limited so the core can be made suberitical by at least R + 0.25% Ak with the highest worth control rod fully withdrawn and all others inserted. The value of R in % Ak is the amount by which the calculated core reactivity, at any time in the operating cycle, exceeds the reactivity at the time of the demonstration. R must be a positive quantity or zero. The value of R shall include the potential shutdown margin loss assuming full B4 C settling in all inverted poison tubes present in the core. The 0.25% Ak is provided as a finite, demonstrable, sub-criticality margin.

t

2. Reactivity Marcin - Inoperable control Rods Specification 3.3.A.2 requires that a rod be taken out of service if it cannot be moved with drive pressure. If a rod is disarmed electrically, its position shall be consistent with the shutdown reactivity limitation stated in Specification 3.3.A.1. This assures that the core can be shutdown at all times with the remaining control rods, assuming the highest worth, operable control rod does rod insert. An allowable pattern for control ,

rods valved out of service will be available to the reactor operator. The number of rods permitted to be inoperable could be nany more than the six allowed by the Specification, particularly late in the operation cycle; however, the occurrence of more than six could be indicative of a generic control rod drive problem and the reactor will be shutdown. Also if damage within the control rod drive mechanism and in particular, cracks in drive internal housing, cannot be ruled out, then a generic problem affecting a number of drives cannot be ruled out.

Circumferential cracks resulting from stress assisted intergranular corrosion have occurred in the collet housing of drives at several BWRs. This type of cracking could occur in a number of drives and if the cracks propagated until severance of the collet housing occurred, scram could be prevented in the affected rods. Limiting the period of operation with a potentially severed collet housing and requiring increased surveillance after detecting one stuck rod will assure that the reactor will not be operated with a large number of rods with failed collet housings.

B. Control Rods ,

t

1. Control rod dropout accidents as discussed in the FSAR can lead to significant core damage. If coupling integrity is maintained, the possibility of a rod dropout accident is eliminated. The overtravel position feature provides a positive check as only uncoupled drives may reach this position. Neutron instrumentation response to rod movement provides a verification that the rod is following its drive.

Amendment No. GO, NVY 87-131 89

VYNPS BASES: 3.3 & 4.3 (Cont'd)

2. The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the extremely remote event of a housing failure. The amount of reactivity which could be added by this small amount of rod withdrawal, which is less than a normal single withdrawal increment, will not contribute to any l damage of the primary coolant system. The design basis is given j in Subsection 3.5.2 of the FSAR, and the design evaluation is l 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. l l

l 3. In the course of performing nornal startup and shutdown procedures, a pre-specified sequence for the withdrawal or l insertion of control rods is followed. Control rod dropout l l 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 I insertions to those listed in the pre-specified sequence and j provides an additional check that the reactor operator is j following prescribed sequence. Although beginning a reactor ,

statiup without having the RWM operable would entail unnecessary risk, continuing to withdraw rods if the RWM fails subsequently J 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 RNM inoperable, if the reactor is subcritical and all I other rods are fully inserted. 1.bove 20% power, the RWM is not I needed since even with a single arror an operator cannot withdraw a rod with sufficient worth, which if dropped, would result in anything but minor consequences.

l 4. Refer to the Vermont Yankee Core Performance Analysis report.

5. The Source Range Monitor (SRM) system has no scram functions. It does provide the operator with a visual indication of neutron level. The ccnsequences of reactivity accidents are a function of the initial neutron flux. The requirement of at least three counts per second assures that any transien should it occur, begins at or above the initial value of 10~g,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 l 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 occurrence of inoperable control rods.

Amendment No. as, 39, 61, 70 90

VYNPS BASES: 3.3 & 4.3 (Cont'd)

I

7. Periodic verification that the Scram Discharge Volume (SDV) drain and vent valves are maintained in the open position provides assurance that the SDV will be available to accept the water displaced from the control rod drives in the event of a scram.

C. Scram Insertion Times The Control Rod System is designed to bring the reactor suberitical at a rate fast enougu to prevent fuel damage. The limiting power transient is that resulting from a turbine stop valve closure with a failure of the Turbine Bypass System. Analysis of tl.is transient shows that the negative reactivity rates resulting from the scram with the average response of all the drives ao given in the above specification, provide the required protection, and MCPR remains ,

greater than the fuel cladding integrity safety limit.

The scram times for all control rods shall be determined during each operating cycle. The weekly control rod exercise tent serves as a periodic check against deterioration of the Control Rod System and also verifies the ability of the control rod drive to scram. The frequency of exercising the control rods under the conditions of two or more control rods valved out of service provides even further assurance of the reliability of the remaining control rods.

D. Control Rod Accumulators Requiring no more than one inoperable accumulator in any nine-rod (3x3) square array is based on a series of XY PDQ-4 quarter core calculations of a cold, clean core. The worst case in a nine-rod withdrawal sequence resulted in a K,gg $ 1 .0. Other repeating rod sequences with more rods withdrawn resulted in Kogf 2,1.0. At reactor pressures in excess of 800 psig, even those control rods with inoperable accumulators will be able to meet required scram insertion ~

times due to the action of reactor pressure. In addition, they may be normally inserted using the Control-Rod-Drive Hydraulic System.

Procedural control will assure that control rods with inoperable accumulators will be spaced in a one-in-nine array rather than grouped together.

E. Reactivity Anomalies During each fuel cycle, excess operating teactivity varies as fuel depletes and as any burnable poison in supplementary control is burned. The magnitude of this excess reactivity may be inferred from the critical rod configuration. As fuel burnup progresses, anomalous behavior in the excess reactivity may be detected by comparison of the critical rod pattern selected base states to the predicted rod inventory at that state. Power operation base conditions provide the most sensitive and directly interpretable data relative to core reactivity. Furthermore, using power operating base conditions permits frequent reactivity comparisons. Requiring a reactivity comparison at the specified frequency assures that a comparison will ,

be made before the core reactivity change exceeds 1% Ak. Deviations !

in core reactivity greater than 1% Ak are not expected and require ,

thorough evaluation. One percent reactivity limit is considered safe j since an insertion of the reactivity into the core would not lead to transients exceeding design conditions of the Reactor System.

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Amendment No. 45, 73 91

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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 Applicability: Applicability:

Applies to the operating status Applies to the periodic testing of the Reactor Standby Liquid requirement for the Reactor ,l Control System. Standby Liquid Control System. j Obiective: Obiective:

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

specification: Specification:

A. Normal Operation A. Normal Operation l Except as specified in 3.4.B The Standby Liquid Control below, the Standby Liquid System shall be verified i

control System shall be operable by:

l operable during periods when fuel is in the reactor unless:

1. The reactor is in cold 1. Testing pumps and valves shutdown in accordance with Specification 4.6.E. A and minimum flow rate of 35 gpm at 1275 psig shall be verified for each pump by recirculating demineralized 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:

1. Testing that the setting of the pressure relief valves is between 1400 and 1490 psig.

2. Initiating one of the standby liquid control loops, excluding the primer chamber and inlet fitting, and verifying that a flow path from a pump to the reactor Amendment No. 142, 128 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.

3. Testing the new trigger assemblies by installing one of the assemblies in the test block and firing it using the installed circuitry. i Install the unfired assemblies, taken from the same batch as the ,

fired one, into the '

explosion valves.

4. Recirculating the 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 i 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 is sooner made operable.

C. Liquid Poison Tank - Boron C. Liquid 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. +G2, 114 93

I VYNPS 3.4 LIMITING CONDITIONS FOR 4.4 SURVEILLANCE REQUIREMENTS OPERATION

2. The solution 2. Sodium pentaborate 1 temperature, including concentration shall be that in the pump suction determined at least once piping, shall be a month and within maintained above the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the curve shown in addition of water or Figure 3.4.2. boron, or if the solution temperature drops below the limits ,

specified by Figure 3.4.2.

D. If Specification 3.4.A or B is not met, an orderly ,

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

E. If Specification 3.4.C is not met, action shall be immediately initiated to correct the deficiency. If at the end of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months the system has not been restored to full operability, then a shutdown shall be initiated with the reactor in cold shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of

initial discovery, i

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l Amendment No. 75, 76, 102 94

i vytJPs l FIGURE 3.4.1 STANDBY LIOUID COtTTROL SYSTEM BORON SOLUTION REOUIREMENTS e . : i '

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l I NET V01.LM: OF S OLUTIO: IN TAN 4 (Callons) 95

VYNPS FIGURE 3.4.2 SODIUM PENTABORATE SOLUTION TEMPERATURE REOUIREMENTS

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VYNPS l l

BASES:

3.4 & 4.4 REACTOR STANDBY LTOUID CONTROL SYSTEM J

A. Normal Operation e The design objective of the Reactor Standby Liquid Control System is to provide the capability of bringing the reactor from full power to a cold, xenon-free shutdown assuming that none of the withdrawn control rods can be inserted. To meet this objective, the Liquid Control System is designed to inject a quantity of boron which produces a concentration of 800 ppm of boron in the reactor core in less than 138 minutes. An 800 ppm boron concentration in the reactor '

core is required to bring the reactor from full power to a 5% Ak suberitical condition. An additional margin (25% of boron) is added for possible imperfect mixing of the chemical solution in the reactor water. A minimum quantity of 3850 gallons of solution having a 10.1%

sodium pentaborate concentration is required to meet this shutdown requirement.

The time requirement (138 minutes) for insertion of the boron solution was selected to override the rate of reactivity insertion due to cooldown of the reactor following the xenon poison peak. For a required minimum pumping rate of 35 gallons per minute, the maximum not storage volume of the boron solution is established as 4830 gallons.

Boron concentration, solution temperature, and volume are checked on a frequency to assure a high reliability of operation of the system should it ever by required. Pump operability testing in accordance with Sracification 4.6.E is adequate to detect if failures have occurred. Flow, relief valve, circuitry, and trigger assembly testing at the prescribed intervals assures a high reliability of system operation capability. Recirculation of the borated solution is done during each operating cycle to ensure one suction line from the boron tank is clear.

B. Operation With Inoperable Components only one of the two standby liquid control pumping circuits is needed for proper operation of the system. If one pumping circuit is found to be inoperable, there is no immediate threat to shutdown capability, and reactor operation may continue while repairs are being made. Assurance that the system will perform its intended function is obtained from the results of the pump and valve testing performed in accordance with ASME Section XI requirements. Whenever one redundant component is inoperable, the potential for extended operation with two subsystems inoperable is reduced by requiring that the redundant component be tested within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

C. Liquid Poison Tank - Boron Concentration The solution saturation temperature varies with the concentration of sodium pentaborate. The solution shall be kept at least 100F above the saturation temperature to guard against boron precipitation. The 10*F margin is included in Figure 3.8.3 of the FSAR. Temperature and liquid level alarms for the system are annunciated in the control Room.

Amendment No. 440, 114, 128 97

i l

VYNPS BASES: 3.4 & 4.4 (Cont'd) i l \

Once the solution has been made up, boron concentration will not vary l unless more boron or more water is added. Level indication and alarm indicate whether the solution volume has changed which might indicate a possible solution concentration change. Considering these factors, the test interval has been established. -

Sodium pentaborate concentration is determined within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the addition of water or boron, or if the solution temperature drops below specified limits. The 24-hour limit allows for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of mixing, subsequent testing, and notification of shift.

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1 l

Amendment No. 4 M , 128 98

- - _ . - - . _ _ - . - . = _ . - . - -

VYNPS 3.5 LIMITING CONDITION FOR 4.5 SURVEILLANCE REQUIREMENT OPERATION 3.5 CORE AND CONTAINMENT COOLING 4.5 CORE AND CONTAINMENT COOLING 1 SYSTEMS SYSTEMS Applicability: Applicability:

Applies to the operational Applied to periodic testing of status of the Emergency Cooling the emergency cooling Subsystems. subsystems.

Obiective Obiective:

To assure adequate cooling To verify the operability of the capability for heat removal in core containment cooling the event of a loss-of-coolant subsystems, accident or isolation from the -

normal reactor heat sink.

Specification: Specification: I A. Core Spray and Low Pressure A. Core Spray and Low Pressure ;

coolant Iniection Coolino Iniection ,

t

1. Except as specified in Surveillance of the core '

Specifications 3.5.A.2 Spray and LPCI subsystems through 3.5.A.4 below shall be performed as and 3.5.H.3 and 3.5.H.4, follows:

both Core Spray and the LPCI Subsystems shall be 1. General Testino operable whenever irradiated fuel is in Item Frequency the reactor vessel and prior to a reactor a. Simulated Each re-startup from the cold Automatic fueling shutdown condition. Actuation outage J Test l

b. Operability I testing of I pumps and valves shall be in '

accordance !

with <

Specification I 4.6.E.

- c. Flow Rate Each re-Test-Core fueling j Spray pumps outage i shall deliver i at least 3000 gpm (torus to torus) against a system head of 120 psig.

Each LPCI pump shall deliver 7450 i 150 gpm (vessel to vessel).

Amendment No. 46, 47, 42, 128 99

1 l

VYNPS 3.5 LIMITING CONDITION FOR 4.5 SURVEILLANCE REQUIREMENT OPERATION

2. From and after the date 2. When one core Spray that one of the Core Subsystem is made or Spray Subsystems is made found to be inoperable, or found to be the active components of inoperable for any the redundant Core Spray reason, reactor Subsystem shall have operation is permissible been or shall be only during the demonstrated to be succeeding seven days operable within 24 unless such subsystem is hours.

sooner made operable, provided that during such seven days, all active components of the other Core Spray Subsystem, the LPCI Subsystems, and the diesel generators required for operation of such components if no external source of power were available, shall be operable.

3. From and after the date 3. When one of the LPCI ,

that one of the LPCI pumps is made or found pumps is made or found to be inoperable, the to be inoperable for any remaining operable LPCI reason, reactor pumps shall have been or operation is permissible shall be demonstrated to only during the be operable within succeeding seven days 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

unless such pump is sooner made operable, provided that during such seven days, the remaining active components of the LPCI Containment Cooling Subsystem and all active components of both Core Spray subsystems and the diesel generators required for operation of such components if no external source of power were available, shall be operable.

l l

1 Amendment No. B4, 47, ++4, 128 100

VYNPS 3.5 LIMITING CONDITION FOR 4.5 SURVEILLANCE REQUIREMENT OPERATION

4. a. From and after the 4. When a LPCI Subsystem is date that a LPCI made or found to be Subsystem is made inoperable, the active I or found to be components of the i inoperable due to redundant LPCI Subsystem failure of the shall have been or shall l associated UPS, be demonstrated to be reactor operation operable within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is permissible only (except the during the Recirculation System succeeding thirty discharge valves).

days, for the 1989/90 operating cycle, unless it is sooner made operable, provided that during that time the associated motor control center (89A or 89B) is powered from its respective maintenance tie, all active components of the other LPCI and the Containment Cooling ,

Subsystem, the Core i Spray subsystems, at,d the emergency l diesel generators i shall be operable, the requirements of Specification 3.10.A.4 are met, and the 4160 volt i tie line to the j Vernon Hydro is the '

operable delayed access power sourCO.

b. From and after the date that a LPCI Subsystem is made or found to be

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inoperable for any .

reason, other than ,

failure of the UPS during the 1989/90 operating cycle, or i Specification j 3.5.A.4.a is not !

met, reactor i operation is l permissible only during the succeeding seven days unless it is sooner made operable, provided Amendnent No. G4, 444, 118 101

i

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VYNPS 3.5 LIMITING CONDITION FOR 4.5 SURVEILLANCE REQUIREMENT OPERATION that during that time all active components of the other LPCI and the Containment Cooling Subsystem, the Core Spray Subsystems, and the diesel generators required for operation of such ccmponents if no external source of power were available, shall be operable.

5. All recirculation pump 5. Recirculation pump discharge valves and discharge valves shall bypass valves shall be be tested to verify full operable or closed prior open to full closed in to reactor startup. 27 < t < 33 seconds and bypass valves shall be tested for operability in accordance with Specification 4.6.E.

6. If the requirements of Specifications 3.5.A cannot be met, an orderly shutdown of the reactor shall be initiated and the reactor shall be in a cold shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

B. Containment spray coolino B. Containment Spray coolino capability capability

1. Both containment cooling 1. Surveillance of the '

spray loops are required drywell spray loops ,

to be operable when the shall be performed as I reactor water follows. During each temperature is greater five-year period, an air than 212*F except that a test shall be performed containment Cooling on the drywell spray Subsystem may be headers and nozzles, inoperable for thirty days.

2. If this requirement 2. When a Containment l cannot be met, an Cooling Subsystem is :

orderly shutdown shall made or found to be I be initiated and the inoperable, the active reactor shall be in the components of the :

cold shutdown condition redundant Containment !

within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Cooling Subsystem shall have been or shall be demonstrated to be-operable within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Amendment No. G7, as, ++4, 144, 128 102

VYNPS 3.5 LIMITING CONDITION FOR 4.5 SURVEILLANCE REQUIREMENT OPERATION C. Residual Heat Removal (RHR) C. Residual Heat Removal (RHRL service Watar system Service Water System Surveillance of the RHR Service Water System shall be performed as follows:

1. Except as specified in 1. RHR Service Water cpecifications 3.5.C.2, subsystem testing:

and 3.5.C.3 below, both RHR Service Water Operability testing of Subsystem loops shall be pumps and valves shall operable whenever be in accordance with irradiated fuel is in Specification 4.6.E.

the reactor vessel and prior to reactor startup Each RHR service water from a cold condition. pump shall deliver at least 2700 gpm and a pressure of at least 70 psia shall be maintained at the RHR heat exchanger service water outlet when the corresponding pairs of RHR service water pumps and station service water pumps are operating.

2. From and after the date 2. When one of the RHR that one of the RHR service water pumps is service water pumps is made or found to be made or found to be inoperable, the operable inoperable for any RHR service water pumps reason, reactor shall have been or shall operation is permissible be demonstrated to be only during the operable within succeeding thirty days 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

unless such pump is sooner made operable, provided that during such thirty days all other active components of the RHR Service Water Subsystem are operable.

3. From and after the date that one RHR Service 3. When one RHR Service Water Subsystem is made Water Subsystem is made or found to be or found to be inoperable for any inoperable, the active reason, reactor components of the operation is permissible redundant RHR Service only during the Water Subsystem shall succeeding seven days have been or shall be unless such subsystem is demonstrated to be t

sooner made operable, operable within provided that all active 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , components of the other RHR Service Water Amendment No. 144, 128 103

1 1

VYNPS I

3.5 LIMITING CONDITION FOR 4.5 SURVEILLANCE REQUIREMENT OPERATION Subsystem, both Core Spray Subsystems, and both diesel generators required for operation of such components if no external source of power were available, shall be operable.

4. If the requirements of Specification 3.5.C cannot be met, an orderly shutdown shall be initiated and the '

l reactor shall be in a cold shutdown condition '

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

)

D. Station Service Water and D. Station Servico Water and Alternate Coolina Tower Alternate Coolino Tower Systems Systems Surveillance of the Station Service Water and Alternate Cooling Tower Systems shall be performed as follows:

1. Except as specified in 1. Operability testing of ,

Specifications 3.5.D.2 pumps and valves shall and 3.5.D.3, both )

be in accordance with 1 Station Service Water Specifiestion 4.6.E.

Subsystem loops and the Each pump shall deliver alternate cooling tower at least 2700 gpm shall be operable against a TDH of whenever irradiated fuel 250 feet.

is in the reactor vessel and reactor coolant temperature is greater than 2120F.

2. From and after the date 2. When one Station Service that one of the Station Water Subsystem is made Service Water Subsystems or found to be is made or found inope rehle , the active inoperable for any components of the reason, reactor redundant Station operation is permissible Service Water Subsystem only during the and the alternate succeeding 15 days cooling tower fan shall unless such subsystem is have been or shall be made operable, provided demonstrated to be that during such 15 days operable within all other active 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , components of the Station Service Water and Alternate Cooling Tower Systems are operable.

Amendment No. 144, 128 104

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 shall have operation is permissible been or shall be only during the demonstrated to be succeeding seven days operable within unless such subsystems 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

are nade operable, provided that during such seven days all other active components of the other subsystem are operable.

4. If the requirements of Specification 3.5.D cannot be met, an orderly shutdown shall be initiated and the reactor shall be in a cold shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

E. Hich Pressure Coolino E. Hioh Pressure Coolant Iniection (HPCI) System Iniection (HPCI) System Surveillance of HPCI System shall be performed as follows:

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

Operability testing of

a. The HPCI System the pump and valves shall be operable. shall be in accordance i with

b. The condensate Specification 4.6.E.

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

I Amendment No. e7, +44, 128 105

I i

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 mads or found to be I

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 I succeeding seven days operable within 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 Depressurization be demonstrated Subsystems, the Core by performing a Spray Subsystems, the functional test LPCI Subsystems, and the of the trip l 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 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. Except as apecified in 1. Operability testing of Specification 3.5.F.2 the relief valves shall below, the entire be in accordance with Automatic Specification 4.6.E.

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. From and after the date 2. When one relief valve of that one of the four the Automatic Pressure relief valves of the Relief Subsystem is made Automatic or found to be Depressurization inoperable, the HPCI Subsystem are made or Subsystem shall have found to be inoperable been or shall be due to malfunction of demonstrated to be the electrical portion operable within of the valve when the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Amendment No. G4, 444, 128 106

i VYNPS I

3.5 LIMITING CONDITION FOR 4.5 SURVEILLANCE REQUIREMENT j 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 sooner made operable, provided that during j such seven days both the l

remaining Automatic j Relief System valves and the HPCI System are operable. !

3. If the requirements of Specification 3.5.F cannot be met, an orderly shutdown shall be initiated and the reactor pressure shall i

be reduced to 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 l Coolina System (RCIC) Coolina 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 Frequency shall be operable l whenever the reactor Simulated Each re-pressure is greater than automatic fueling 150 psig and irradiated actuation outage fuel is in the reactor test 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 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 operating that during such 7 days pressure when all active components of recirculating to the I the HPCI System are Condensate Storage Tank.

operable.

Amendment No. 44, 27, 444, 128 107 l l

_ _ . _ _ _ . _ _ - . . . . . . - . - - - - - _ _ _ . - - . )

VYNPS 3.5 LIMITING CONDITION FOR 4.5 SURVr:1. LANCE REQUIREMENT OPERATION , i

3. If the requirements of Specification 3.5.G cannot be met, an orderly shutdown shall l be initiated and the :

reactor pressure shall l be reduced to 120 psig [

within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . L H. Minimum Core and Containment H. Minimum Core and Containment l Coolino System Availability Coolina System Availability !

i

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

permissible only during generator shall have ,

the succeeding seven been or shall be '

days, provided that all demonstrated to be I of the Low Pressure Core operable within l Cooling and Containment 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . j l Cooling Subsystems connecting to the l' operable diesel

{

generator shall be operable. If this requirement cannot be 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 . ,

2. Any combination of inoperable components in the Core and Containment i l Cooling Systems shall l not defeat the '

capability of the remaining operable ,

components to fulfill '

the more and containment cooling functions.

3. When irradiated fuel is in the reactor vessel )

and the reactor is in the cold shutdown condition, all Core and Containment Cooling Subsystems may b7 inoperable provided no work is permitted which has the potential for draining the reactor vessel. i I

l l

Amendment Nc. B7, 114 108 l

i i

VYNPS 3.5 LIMITING CONDITION FOR 4.5 SURVEILLANCE REQUIREMENT 1 OPERATION l I

4. When irradiated fuel is I

in the reactor vessel and the reactor is in the refueling condition, i both LPCI subsystems, or ,

both Core Spray systems, '

or one diesel generator may be inoperable provided that a source of water of greater than 300,000 gal. is available to the operable core cooling subsystem.

I. Maintenance of Filled I. Maintenance of Filled Discharce Pipe Discharce Pipe Whenever core spray The following surveillance subsystems, LPCI subsystem, requirements shall be ,

HPCI, or RCIC are required adhered to to assure that ;

to be operable, the the discharge piping of the i discharge piping from the core spray subsystems, LPCI '

pump discharge of these subsystem, HPCI and RCIC are systems to the last block filled:

valve shall be filled. I

1. Every month and prior to i the testing of the LPCI subsystem and core spray (

subsystem, the discharge piping of these systems shall be vented from the high point and water flow observed.

2. Following any period where the LPCI subsystem or core spray subsystems have not been required to be operable, the i discharge piping of the 1 inoperable system shall l be vented from the high i point prior to the return of the system to service. ;

)

3. Whenever the HPCI or RCIC system is lined up to take suction from the torus, the discharge piping of the HPCI and RCIC shall be vented from the high point of the system and water

, flow observed on a monthly basis.

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l Amendment No. 128 109 I

VYNPS PASES: ,

l 3.5 CORE AND CONTAINMENT COOLANT SYSTEMS A. Core Spray Coolino System and Low Pressure Coolant Iniection System This Specification assures that adequate standby cooling capability is available whenever irradiated fuel is in the Reactor Vessel.

Based on the loss-of-coolant analyses, the Core Spray and LPCI systems provide sufficient cooling to the core to dissipate the energy associated with the loss-of-coolant accident and to limit the ['

accident-caused core conditions as specified in 10CFR50, Appendix K.

The analyses consider appropriate combinations of the two Core Spray Subsystems and the two LPCI Subsystems associated with various break locations and equipment availability in accordance with required single failure assumptions. (Each LPCI Subsystem consists of the LPCI pumps, the recirculation pump discharge valve, and the LPCI i injection valve which combine to inject torus water into a recirculation loop.)

The LPCI System is designed to provide emergency cooling to the core e by flooding in the event of a lose-of-coolant accident. This system is completely independent of the Core Spray System; however, it does function in combination with the Core Spray System to prevent excessive fuel clad temperature. The LPCI and the Core Spray Systems provide adequate cooling for break areas up to and including the double-ended recirculation line break without assistance from the high pressure emergency Core cooling Subsystems.

The intent of these specifications is to prevent startup from the [

cold condition without all associated equipment being operable.

However, during operation, certain components may be out of service t for the specified allowable repair times. Assurance that the systems will perform their intended function is obtained from the results of I the pump and valve testing performed in accordance with ASME l Section XI requirements referenced in Specification 4.6.E. Whenever one redundant system is inoperable, the potential for extended i operation with two subsystems inoperable is reduced by requiring that I the redundant subsystem be tested within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . i B. and C. Containment Sprt" Coolino Capability and RHR Service Water ,

System -

The containment heat removal portion of the RHR System is provided to remove heat energy from the containment in the event of a loss-of-coolant accident. For the flow specified, the containment long-term pressure is limited to less than 5 psig and, therefore, the flow is more than ample to provide the required heat removal capability.

Reference:

Section 14.6.3.3.2 FSAR.

I Each Containment Cooling Subsystem consists of two RHR service water I pumps, I heat exchanger, and 2 RHR (LPCI) pumps. Either set of equipment is capable of performing the containment coo. ' f'umtion.

In fact, an analysis in Section 14.6 of the FSAR shows t cos

( subsystem consisting of 1 RHR service water pump, I heat exenanger, I and 1 RHR pump has sufficient capacity to perform the cooling function. Assurance that the systems will perform their intended function is obtained from the results of the pump and valve t'ating performed in accordance with ASME Section XI requirements referenceo in Specification 4.6.E. Whenever one redundant system is inoperable, the potential for extended operation with two subsystems inoperable is reduced by requiring that the redundant subsystem be tested within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

I 1

Amendment No. G4, 144, 128 110 l

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VYNPS l

i BASES: 3.5 (Cont'd)

D. Station Service Water and Alternate Coolino Tower System The Station Service Water Subsystems and the Alternate Cooling Tower System provide alternate heat sinks to dissipate residual heat after a shutdown or accident. Each Station Service Water Subsystem and the l Alternate Cooling Tower System provides sufficient heat sink capacity f

I to perform the required heat dissipation. The Alternate Cooling Tower System will provide the necessary heat sink in the event both Station Service Water Subsystems become incapacitated due to a loss of the Vernon Dam with subsequent loss of the Vernon Pond.

E. Hich Pressure Coolant Iniection System The High Pressure Coolant Injection System (HPCIs) is provided to adequately cool the core for all pipe breaks smaller than those for which the LPCI or Core Spray Cooling Subsystems can protect the core.

The HPCIs meets this requirement without the use of outside power.

For the pipe breaks for which the HPCIs is intended to function the core never uncovers and is continuously cooled; thus, no clad damage occurs and clad temperatures remain near normal throughout the transient.

Reference:

Subsection 6.5.2.2 of the FSAR.

F. Automatic Depressurization System l The relief valves of the Automatic Depressurization System are a j backup to the HPCIs. They enable the Core Spray Cooling System or LPCI Subsystem to provide protection against the small pipe break in

! the event of HPCI failure by depressurizing the reactor vessel l

rapidly enough to actuate the Core Sprays or LPCI Subsystem. Either of the two Core Spray Cooling Systems or LPCIs provides sufficient I flow of coolant to prevent clad melting. All four relief valves are l included in the Automatic Pressure Relief System. (See VYNPS, FSAR Vol. 4, Appendix B.)

l G. Reactor Core Isolation Coolino System l

The Reactor Core Isolation Cooling System (RCIC) is provided to maintain the water inventory of the reactor vessel in the event of a main steam line isolation and complete loss of outside power without the use of the emergency core cooling systems. The RCIC meets this requirement. Reference Section 14.5.4.4 FSAR. The HPCIS provides an incidental backup to the RCIC system such that in the event the RCIC-should be inoperable no loss of function would occur if the HPCIS is operable.

H. Minimum core and Containment Coolino System Availability The core cooling and the containment cooling subsystems provide a method of transferring the residual heat following a shutdown or accident to a heat sink. Based on analyses, this specification assures that adequate cooling capacity is available by precluding any combination of inoperable components from fulfilling the core and containment cooling function. It is permissible, based upon the low heat load and other methods available to remove the residual heat, to t disable all core and containment cooling systems for maintenance if l the reactor is cold and shutdown and there is no potential for draining the reactor vessel. However, if refueling operations are in progress, one coolant injection system, one diesel and a residual of at least 300,000 gallons is required to assure core flooding capability.

Amendment No. 47, 114 111 .

VYMPS BASES: 3.5 (Cont'd)

I. Maintenance of Filled Discharce Pipe Full discharge lines are required when the core spray subsystems, HPCI and RCIC are required to be operable to preclude the possibility of damage to the discharge piping due to water hammer action upon a 4 pump start.

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Amendment No. 18 112

VYNPS BASES:

4.5 CORE AND CONTAINMENT COOLANT SYSTEMS A. Core Spray and LPCI During normal plant operation, manual tests of operable pumps and valves shall be conducted in accordance with Specification 4.6.E to demonstrate operability.

During each refueling shutdown, tests (as summarized below) shall be i conducted to demonstrate proper automatic operation and system f performance.

l Periodic testing as described in Specification 4.6.E will demonstrate that all components which do not operate during normal conditions will operate properly if required.

The automatic actuation test will be performed by simulation of high drywell pressure or low-low water level. The starting of the pump and actuation of valves will be checked. The normal power supply will be used during the test. Testing of the sequencing of the pumps when the diesel generator is the source of power will be checked during the testing of the diesel. Following the automatic actuation test, the flow rate will be checked by recirculation to the suppression chamber. The pump and valve operability checks will be performed by manually starting the pump or activating the valve. For the pumps, the pump motors will be run long enough for them to reach operating temperatures.

B. and C. Containment Spray Coolino Capability and RHP "orvice Water systems The periodic testing requirements specified in Specifications 4.5.B and C will demonstrate that all components will operate properly if required. Since this is a manually actuated system, no automatic actuation test is required. The system will be activated manually and the flow checked by an indicator in the control room.

Once every five years air tests will be performed to assure that the containment spray header nozzles are operable. l D., E., and F. Station Service Water and Alternate Coolino Tower Systems and High Pressure Coolant Iniection and Automatic j Depressurization System 1 l

l The testing described in Specification 4.6.E for the HPCI System will !

demonstrate that the system will operate if required. The Automatic ne pressurization System is tested during refueling outages to avoid an undesirable blowdown of the Reactor Coolant System.

The HPCI Automatic Actuation Test will be performed by simulation of the accident signal. The test is normally performed in conjunction l with the automatic actuation of all Core Standby Cooling Systems.

G. Reactor Core Isolation Coolino System ;

i Frequency of testing of the RCIC System is the same as the HPCI, per Specification 4.6.E, and demonstrates that the system is operable if needed.

1 Amendment No. e4, 444, 128 113

VYNPS l l

BASES: 4.5 (Cont'd) I l

H. Minimum core and containment Coolino System Availability Assurance that the diesels will perform their intended function is obtained by the periodic surveillance test and the results obtained from the pump and valve testing performed in accordance with ASME l Section XI requirements described in Specification 4.6.E. Whenever a diesel is inoperable, the potential for extended operation with two j diesels inoperable is reduced by requiring that the redundant diesel 1 be tested within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . l I. Maintenance of Filled Discharae Pipe observation of water flowing from the discharge line high point vent as discussed in Section I assures that the Core Cooling Subsystems will not experience water hammer damage when any of the pumps are started. Core Spray Subsystems and LPCI Subsystems will also be vented through the discharge line high point vent following a return from an inoperable status to assure that the system is " solid

and ready for operation.

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Amendment No. 44, 34, 444, 128 114

. :r,h.A 4 - eL l

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i VYNPS 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS OPERATION 3.6 REACTOR COOLANT SYSTEM 4.6 REACTOR COOLANT SYSTEM Applicability: Applicability:

j Applies to the operating status Applies to the periodic of the reactor coolant system. examination and testing requirements for the reactor coolant system.

Obiective: Obiective:

To assure the integrity and safe To determine the condition of operation of the reactor coolant the reactor coolant system and system. the operation of the safety devices related to it.

Specification: Specification:

A. Pressure and Temoerature A. Pressure and Temperature Limitations Limitations

1. The reactor coolant 1. The reactor coolant system temperature and temperature and pressure pressure shall be shall be recorded at limited in accordance least once per .or with the limit lines during system heatup, shown on Figure 3.6.1 cooldown and inservice during heatup, cooldown, leak and hydrostatic criticality (except for testing operations.

the purposes of low power physics testing),

and inservice leak and hydrostatic testing.

2. The maximum heatup or 2. The reactor coolant cooldown rate is 1000F temperature and pressure when averaged over any shall be recorded at the one hour period, time of reactor criticality.

3. The reactor vessel head 3. When the reactor vessel bolting shall not be head bolting is being tensioned unless the tightened or loosened, temperature of the the reactor vessel shell vessel head flange and temperature immediately the head is greater than below the vessel flange 700F. shall be permanently recorded.

4. The pump in an idle 4. Prior to and after recirculation loop shall startup of an idle not be started unless recirculation loop, the the temperatures of the temperature of the ,

coolant within the idle reactor coolant in the and operating operating and idle loops recirculation loops are shall be recorded.

within 500F of each other.

Amendment No. 33 115

l VYNPS 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS OPERATION

5. The reactor vessel irradiation surveillance specimens shall be removed and examined to determine changes in material properties in accordance with the following schedule:

CAPSULE REMOVAL YEAR 1 10 2 30 3 Standby The results shall be used to update Figures 3 6.2 and 3.6.3.

The removal times shall be referenced to the refueling outage following the year specified, referenced to the date of commercial operation.

B. Coolant Chemis t ry B. Coolant Chemi s t ry

1. a. During reactor 1. a. A sample of reactor power operation, coolant shall be the radioiodine taken at least ,

concentration in every 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months and the reactor coolant analyzed for shall not exceed radioactive iodines 1.1 microcuries of of I-131 through l I-131 dose I-135 during power l equivalent per gram operation. In '

of water, except as addition, when allowed in steam jet air Specification ejector monitors 3.6.B.l.b. indicate an increase in radioactive gaseous effluents of 25 percent or 5000 pCi/sec, whichever is greater, during steady state reactor operation a reactor coolant sample shall be taken and analyzed I for radioactive iodines.

Amendment No. 34, ,91 116

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l VYNPS 3.6 LIMITING CONDITIONS FCR 4.6 SURVEILLANCE REQUIREMENTS 1 OPERATION i

b. The radiciodine b. An isotopic concentration in analysis of a )

the reactor coolant reactor coolant l shall not exceed sample shall be :

1.1 microcuries of made at least once I-131 dose per month.

equivalent per gram of water, for greater than 24 consecutive hours. ,

c. The radioiodine c. Whenever the concentration in radiciodine the reactor coolant concentration of ,

shall not exceed prior steady-state 4.0 microcuries of reactor operation I-131 dose is greater than equivalent per gram 0.011 pCi/gm but of water. less than 0.11 pCi/gm, a sample of reactor coolant shall be taken within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of the next reactor startup and !

analyzed for radioactive iodines of I-131 through I-135.

d. Whenever the radiciodine concentration of prior steady-state reactor operation is greater than 0.11 pCi/gm, a sample of reactor coolant shall be taken prior to the next reactor startup snd analyzed for ;

radioactive iodines of I-131 through I-135, as well as within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following a reactor startup.

4 4

Amendment No. 91 117

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VYNPS 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS )

OPERATION

e. With the ;

radioiodine l concentration in the reactor coolant greater than 1.1 microcuries/

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 and analyzed for radioactive iodines of I-131 through I-135, until the specific activity of the reactor coolant is restored below 1.1 microcuries/

gram dose equivalent I-131.

2. The reactor coolant 2. During startups and at 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 Chloride ion 0.01 ppm

3. For reactor startups the 3. a. With steaming rates maximum value for greater than or conductivity shall not equal to exceed 10 umho/cm and 100,000 pounds per the maximum value for hour, a reactor 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 j analyzed for conductivity and chloride ion content.

I Amendment No. 91 118

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3.6 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 i specification 3.6.B.3 i 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

5. If Specification 3.6.B.

is not met, an orderly ,

shutdown shall be j 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 Leakace

1. Any time irradiated fuel Reactor coolant system is in the reactor vessel leakage shall be checked and and reactor coolant logged at least once per temperature is above day.

212oF, reactor coolant 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.

2. Both the sump and air sampling systems shall I be operable during power operation. From and after the date that one of these systems is made or found inoperable for any reason, reactor operation is permissible only during succeeding seven days.

119 !

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VYNPS l l

3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS OPERATION

3. If these conditions cannot be met, initiate an orderly shutdown 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 .

D. Safety and Relief Valves D. Safety and Relief Valves

1. During reactor power 1. Operability testing of operating conditions and Safety and Relief Valves whenever the reactor shall be in accordance coolant pressure is with Specification greater than 120 psig 4.6.E. The lift point and temperature greater of the safety and relief than 3500F, both safety valves shall be set as valves shall be specified in operable. The relief Specification 2.2.B. ,

valves shall be operable, except that if one relief valve is inoperable, reactor power shall be immediately reduced to '

and maintained at or below 95% of rated power.

2. If Specification 3.6.D.1 is not met, initiate an orderly shutdown and the reactor coolant pressure shall be below 120 psig and 3500F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

E. Structural Intecrity and E. Structural Intecrity and operability Testino operability Testino e The structural integrity and 1. Inservice inspection of I the operability of the safety-related safety-related systems and components shall be components shall be performed in accordance maintained at the level with Section XI of the required by the original ASME Boiler and pressure -

acceptance standards Vessel Code and 1 throughout the life of the applicable Addenda as p} ant. required by 10 CFR 50, Section 50.55a(g),

except where specific written relief has been granted by the NRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).

Amendment No. +4, +B, 46, 99, 128 120 l I

VYNPS 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS OPERATION

2. Operability testing of safety-related pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g),

except where specific written relief has been granted by the NRC pursuant to 10 CFR 50, Section 50.55a(g) (6) (i) .

F. Jet Pumps F. Jet Pumps

1. Whenever the reactor is 1. Whenever there is in the startup/ hot recirculation flow with standby or run modes, the reactor in the all jet pumps shall be startup/ hot standby or intact and all operating run modes, jet pump jet pumps shall be integrity and operable. If it is operability shall be determined that a jet checked daily by pump is inoperable, an verifying that the orderly shutdown shall following two conditions be initiated and the do not occur reactor shall be in a simultaneously:

cold shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , a. The recirculation pump flow differs by more than 10%

from the established speed-flow characteristics.

b. The indicated total core flow is more than 10% greater than the core flow value derived from established power-core flow relationships.

2. Flow indication from 2. Additionally, when each of the twenty jet operating with one pumps shall be verified recirculation pump, the prior to initiation of diffuser to lower plenum reactor startup from a differential pressure cold shutdown condition. shall be checked daily, and the differential pressure of any jet pump in the idle loop shall not vary by more than 10% from established patterns.

Amendment No. 92, 99 121

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VYNPS i 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS I OPERATION

3. The indicated core flow 3. The baseline data is the sum of the flow required to evaluate the indication from each of conditions in the twenty jet pumps. Specifications 4.6.F.1 If flow indication and 4.6.F.2 shall be failure occurs for two acquired each operating or more jet pumps, cycle.

immediate corrective action shall be taken.

If flow indication for ,

all but one jet pump cannot be obtained within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months an orderly shutdown shall 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 .

G. Sincle Loop Operation G. Sinole Loop Operation

1. The reactor may be 1. With one recirculation started and operated or pump not in operation, operation may continue core flow between 34% ,

with a single and 45% of rated, and i recirculation loop core thermal power provided thatt greater than the limit specified in

a. The designated Figure 3.6.4 (Region 2),

adjustments for establis APRM flux scram and andLPRMy1paselineAPRM neutron flux rod block trip noise levels prior to settings (Specifi- entering this region, :

cations 2.1.A.1.a provided that baseline r and 2.1.B.1, values have not been Table 3.1.1 and established since the Table 3.2.5), rod last core refueling.

block monitor trip '

Baseline values shall be setting established with one (Table 3.2.5), MCPR recirculation pump not ;

fuel cladding in operation and core integrity safety thermal power less than limit (Specifi- or equal to the limit '

cation 1.1.A), and specified in MCPR operating Figure 3.6.4.

limits and MAPLHGR ;

limits, provided in '

the Core Operating Limits Report, are initiated within .

8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . During the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , either these adjustments must be completed or the reactor brought to Hot Shutdown.

(1) Detector Levols A and C of one LPRM string per core octant plus detector Levels A and C of one LPRM string in the center of the core shall be monitored.

Amendment No. 44, 94, 116 122

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VYNPS 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS OPERATION l 1

b. With one l recirculation pump '

not in operation, l core thermal power )

greater than the -

limit specified in !

Figure 3.6.4, and core flow between ;

34% and 45% of '

rated (Region 2 of Figure 3.6.4):

(1) If baseline APRM d l LPRMI neutron flux noise levels have been established .

since the last core refueling, initiate action within 15 minutes such that theg{RM and LFRM I !

neutron flux noise levels are determined within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , ands (i) If the A andLPRMy{y neutron '

i flux noise levels are less than or equal to '

3 times their '

established baseline levels, I continue to determine the noise levels at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and within l 30 minutes after the completion I

(1) Detector Levels A and C of one LPRM string per core octant plus detector Levels A and C of one LPRM string in the center of the core shall be I monitored.

Amendment No. 94 123

VYNPS 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS OPERATION 1 of a core thermal power increase greater !

than 5% of rated core thermal power, or (ii) If the APPR and/g)

LPRM neutron flux noise levels are greater than 3 times their established baseline levels, initiate action within '

15 minutes such that .

the noise i levels are restored to within the required limits ;

within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months by increasing core flow i and/or by initiating ,

an orderly reduction of core '

thermal power by ;

inserting control '

rods.

P l

(1) Detector Levela A and C of one LPRM string per core octant plus detector Levels A and C of one LPRM string in the center of the core shall be '

monitored.

Amendment No. 94 124

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VYNPS 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS OPERATION (2) If baseline APRM d LPRMI neutron flux noise levels have not i been ,

established '

since the last core refueling, initiate action within 15 minutes of entering this region (Region 2 of Figure 3.6.4) such that operation is outside this region within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

c. With one recirculation pump not in operation, core thermal power gr. eater than the limit specified in Figure 3.6.4, and core flow less than 34% of rated (Region 1 of Figure 3.6.4),

initiate action within 15 minutes of entering this region such that operation is outside this region within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

d. The idle loop is isolated by elec-trically disarming the breaker to the recirculation pump motor generator set drive motor prior to startup cr. if .

uisabled during l reactor operation, l within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , l and until such time '

as the inactive recirculation loop is to be returned j to service.

(1) Detector Levels A and C of one LPRM string per core octant plus detector Lovels A and C of one LPRM string in the center of the core shall be monitored.

Amendment No. 94 125 !

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VYNPS 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS OPERATION i

e. The recirculation system controls will be placed in the manual flow control mode.

l H. Recirculation System H. Recirculation System '

1. With two recirculation 1. With two recirculation pumps in operation, with pumps in operation, total core flow less total core flow less than 45% of rated, and than 45% of rated, and core thermal power core thermal power greater than the limit greater than the limit specified in specified in Figure 3.6.4 (Regions 1 Figure 3.6.4 (Regions 1 and 2): and 2), establish ,

base

a. If basel LPRM((peAPRMand

' neutron flux andLPRM{gyAPRMneutron noise levels prior to flux noise levels entering these regions, have been provided that baseline established since values have not been the last core established since the refueling, initiate last core refueling.

action within Baseline values shall be 15 minutes such established with core that thermal power less than LPRM{peAPRMand I neutron or equal to the limit flux noise levels specified in i are determined Figure 3.6.4.

within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , '

and:

i (1) If t APRM and LPRMgaII neutron flux noise levels are less than or equal to 3 times their established baseline levels, continue to determine the noise levels at least once per #

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

30 minutes after the completion of a core thermal I

t (1) Detector Levels A and C of one LPRM string per core octant plus detector Levels A and C of one LPRM string in the center of the core shall be ,

monitored. i Amendment No. te, 92, 94 126 ,

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VYNPS I 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS OPERATION !

l power increase ,

greater than 5%

of rated core thermal power, or .,

(2) If the APRM and/or LPRMII) i neutron flux +

noise levels are greater than 3 times

their established baseline levels, initiate action within ;

15 minutes such <

that the noise levels are restored to '

within the required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months by increasing core flow ;

and/or by initiating an orderly reduction of core thermal ,

power by inserting control rods. 1

b. If base 1 andLPRM{gyAPRM neutron flux noise levels -

have not been established since ;

the last core refueling, initiate ,

action within l 15 minutes of entering these regions (Regions 1 <

and 2 of !

Figure 3.6.4) such that operation is outside these 4 regions within L 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . .

r l

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(1) Detector Levels A and C of one LPRM string per core octant plus detector Levels A and C of one LPRM string in the center of the core shall be monitored.

Amendment No. 94 127

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VYNPS 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS i OPERATION

2. Operation with one recirculation loop is permitted according to Specification 3.6.G.l.

3. With no reactor coolant system recirculation loops in operation, immediately initiate an orderly reduction in i

core thermal power to l less than or equal to the limit specified in Figure 3.6.4 (Region 3),

l and initiate measures I

such that the unit is in Hot Shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

I. Shock Suppressors (Snubbers) I. Shock Suppressors (Snubbers)

1. Except as noted in 1. Each snubber shall be 3.6.I.2 and 3.6.I.3 demonstrated operable by below, all required performance of the safety-related snubbers following inspection shall be operable program, whenever its supported system is required to be a. Visual Inspections operable.

Visual inspections

2. With one or more shall be performed required snubbers in accordance with inoperable, within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , replace or schedules restore the snubber to operable status and perform an engineering No.

evaluation per Inoperable Specification 4.6.I.lb Snubbers Per Next Required and c, on the supported Inspection Inspection component. In all Period Intervals cases, the required snubbers shall be made 0 18 months 125%

operable or replaced 1 12 months 125%

prior to reactor 2 6 months 125%

startup. 3, 4 124 days 125%

5, 6, 7 62 days 125%

l 3. If the requirements of 8 or more 31 days 125%

3.6.I.1 and 3.6.I.2 cannot be met, the The snubbers may be supported system shall categorized into be declared inoperable two groups: the and the appropriate accessible and action statement for those inaccessible that system shall be during reactor followed. operation. Each group may be inspected independently in accordance with the above schedule.

l l Amendment No. 24, GG, 94 128 I

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VYNPS 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS OPERATION The inspection interval shall not be lengthened more than one step at a time. Inaccessible snubbers are required to be inspected only if the period of time in which they become accessible is greater than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ,

b. Visual Inspection Acceptance Criteria Visual inspections shall vorify (1) that there are no visible indications of damage or impaired operability, and (2) that the snubber installation exhibits no visual indications of detachment from foundations or supporting structures.

Snubbers which appear inoperable as a result of visual inspections may be determined operable for the purpose of establishing the next visual inspection interval, providing that (1) the cause of the rejection is clearly established and remedied for that particular snubber and for other snubbers that may be generically susceptible; and (2) the affected snubber is functionally tested in the as-found condition and determined operable per Specification 4.6.I.c, as applicable.

Amendment No. 94, B9, 94 129 l

VYNPS 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS OPERATION When the fluid port of a hydraulic snubber is found to be uncovered, the snubber shall be determined inoperable unless it can be determined operab.e via functional testing for the purpose of establishing the next visual inspection interval. The functional test, in this case, shall be started with the piston in the as-found condition, extending the piston rod in the tension mode direction.

c. Functional Tests At least once per 18 months during shutdown, a representative sample of 10% of the snubbers in use in the plant shall be functionally tested either in place or in a bench test. For each snubber that does not meet the functional test acceptance criteria of Specification 4.6.I.l.d, an additional 10% of the snubbers shall be functionally tested until no more failures are found or until all snubbers have been functionally tested.

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Amendment No. G4, 69, 94 130 l

VYNPS 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS OPERATION Snubbers of a rated capacity greater than the capability of the testing machine shall be functionally tested as follows: (1) the lock up and bleed velocity of the snubber valve shall be verified by testing it on a cylinder that is within the capability of the testing machina, (2) the free stroke of the cylinder shall be checked, and (3) the pressure retaining capability of the cylinder shall be checked.

Snubbers identified as especially difficult to remove or in high radiaticn areas shall also be included in the representative sample.

In addition to the regular sample, snubbers which failed the previous functional test shall be retested during the next test period unless the root cause for the problem has l been determined and I corrective actions implemented. If a spare snubber has been installed in place of a failed

, snubber, than both l the failed snubber I

(if it is repaired l and installed in another position) and the spare snubber shall be retested during the next test period.

l l

Amendment No. 69, 94 131 )

1 i

VYNPS j l

3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS I OPERATION l l

Failure of these anubbers shall not entail functional testing of additional snubbers.

If any snubber selected for i functional testing either fails to lock up or fails to move, i.e., frozen in place, the cause will be evaluated and if caused by manufacturer or ;

design deficiency, l all generically susceptible snubbers of the same design subject to the same defect shall be functionally tested. This testing requirement ,

shall be :

independent of the l requirements stated above for snubbers >

not meeting the ,

functional test acceptance criteria. ,

For the snubber (s) found inoperable, a documented engineering evaluation shall be ;

performed on the <

component (s) which

are supported by the snubber (s).

The scope of the ;

evaluation shall be based on engineering judgement and may ,

be limited to a .

visual inspection of the supported component (s). The purpose of this engineering evaluation shall be l to determine if the component (s) supported by the Amendment No. 89, 94 132

VYNPS 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS OPERATION snubber (s) were adversely affected by the inoperability of the snubber (s) in order to ensure that the supported component remains capable of meeting the designed service.

d. Hydraulic Snubbers Punctional Test Acceptance Criteria

, The hydraulic l snubber functional test shall verify that:

1. Activation (restraining action) is achieved within the specified range of velocity cr acceleration in both tension and compression.

2. Snubber bleed, or release rate, where i required, is within the specified range in compression or tension.

For snubbers specifically required to not displace under continuous load, the l ability of the snubber to withstand load without displacement shall be verified.

Amendment No. 24, 69, 94 133 l

VYNPS 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIRDIENTS OPERATION J. Thermal Hydraulic Stability J. Thermal Hydraulic Stability

1. When the reactor mode switch is in RUN, the reactor shall not intentionally be operated in a natural circulation mode, except as permitted by Specification 3.6.H.3, '

nor shall an idle recirculation pump be started with the reactor in a natural circulation mode.

Amendment No. 39, 64, 94, 94 134

E e

5 k Valiu ihru 4.45EB kTr!(t) (31 EfPf]

E Curvesareinoccordcaceilth

& A usted Reference Imperature at 1/4T = 89r 10CfR50 App G ond RG 1.99 Rev 2 A usted Refererce Imperature etS/4T = 737

.u A usted Reference Imperature at flonge = 207 j ? YLCD I 5

I I >

+ . ,

-l- l l

? sT, ..

Leak Test j Upper Limit m - x 7 QCQ.- h.

V *n0 NY O'* t/ ,

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? ^

a

- 3 B00 - -

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=

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5 S$ *N1 UprLimitj ve eO 400 - operation Suberitical.

50 E/hr '8 to0 r/he x

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i

< ,,0s 44 -50 g/hr ,

x 100 r/hr J I

critical t*

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l I l I Celticality 4 0 ' . Region d

i 60 E0 100 110 1f'0 o

15'0 14'O 5'0 '

~

200 220 240 25'0 280 300 E REACTOR COOLANT TEMPERATURE - 0F L

w

VYNPS FIGURE 3.6.2 FAST ITEUTRON FLUENCE (E>1 MEV) AS A FUNCTION OF THERMAL ENERGY ,

AND FULL POWER YEARS 1018 g,

1017 _ ed ,

E *

{ ,

9 !

e

~

h ev 3/4T b

c .

. t

- t s

1016 -

~

l t

t 1 1015 e 10 20 30 TIMF (FULL POWER YEARS) i 4' l l l 0 1.39 2.8 4.18 W WR (th) x 10 8 - '

REFERENCE:

L. H. Iowry et al. "Damination. Testing, and Evaluation of !

Irradiated Pressure Vessel Surveillance Specimens f rom Vermont Yankee Nuclear Power Station. l l

Bate 11e Columbus Laboratories Report dBCL-585-84-3, May 15, 1984 l Amendment No. M, 93 136 ;

l i

1

---w,n- -- - , - , . . .

VYNPS FIGURE 3.6.3 FLUENCE FACTOR FOR USE IN REGULA'IVRY CUIDE 1.99 REVISION 2

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30 10

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(E>l MeV)

Amendment No. -34 , 93 137

1 E

E 70 , , , i , , , . .. ..i.4.i.i

'+-~~"-

.,_L__ __ g

= ._____ _

.:::: :::::: ~ -- --~~-"

si

? RECf0N 1 REG 1J12 _.

~

E ~bbb5 APRM/LPRM ._

APRM/l.PRM _ .___._. _

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_ .: monitoring :: .___4_ _ _-.__. o 60 - :: required ::::::

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required g

~::::

~~~~--- '~~

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m T .::::~ : _ _ single loop n a _t_ Single loop - ___ _. o g operation.

____. __ x OPerollon not ____- ._.

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20 30 40 50 60 70 CORE FLOW (% RATED) ..

- 1 -

VYNPS BASES:

3.6 and 4.6 REACTOR COOLANT SYSTEM A. Pressure and Temperature Limitations All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 4.2 of the FSAR. During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall. Thece thermal induced compressive stresses tend to alleviate the tensile stresses induced by their internal pressure. Therefore, a pressure-temperature curve based on steady-state conditions (i.e., no thermal stresses) represents a lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing locations.

The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the controlling location. The thermal gradients established during heatup produce tensile stresses at the outer wall of the vessel. These stresses are additive to the pressure induced tensile atresses which are already present. The ,

thermal induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined. Subsequently, for the cases in which the outer wall of the vessel becomes the 7 stress controlling location, each heatup rate of interest must be analyzed on an individual basis.

In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures should be within 50'F of each other prior to startup of an idle loop.

The reactor vessel materials have been tested to determine their "

initial reference temperature nil-ductility transition temperature (RTNDT) of 40'F maximum. Reactor operation and resultant fast neutron (E > 1 Mev) irradiation will cause an increase in the RT UDT-Therefore, an adjusted reference temperature can be predicted using current industry practices and Vermont Yankee Surveillance Program data. (Regulatory Guide 1.99, Revision 2, and Battelle Columbus Laboratory Report BCL 585-84-3, dated May 15, 1984. The pressure / temperature limit curve, Figure 3.6.1, includes predicted adjustmengsforthisshift in RT gg7 for operation through 4.46 x 10 MWH(t), as well as ad]ustments for possible errors in the pressure and temperature sensing instruments.

The reference temperature of the closure flange material wra determined by material testing and Branch Technical Position MTEB 5-2, " Fracture Toughness Requirements for Older Plants". The closure flange is located in a low neutron fluence area and therefore no measurable RTNDT shift is expected over the life of the plant.

Amendment No. 33, 69, 61, 93, 120 139

VYNPS BASES: 3.6 and 4.6 (Cont'd)

The actual shift in RTNDT f 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 RT NM is determined in accordance with Regulatory Guide 1.99, Revision 2. The heatup and cooldown curves must be recalculated when the ART NDT determined from the surveillance capsule is different from the calculated ART NDT f0Y 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 to assure compliance with the requirements of Appendix H to CFR Part 50. '

Coolant Chemistry A steady-state radioiodine concentration limit of 1.1 pCi 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 .

near the limit, as set forth in Specification 3.8.C.la, or there is a t 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 meters at the nearest site boundary (190 m) for a X/O = 3.9 x 10~3 sec/m 3 (Pasquill D and 0.33 m/sec 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 consequences of a postulate 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 radiciodine concentration '

would not be expected to change rapidly during steady-state operation I 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.

1 kwendment No. 43, 69, 94, 93 140 l

VYNPS BASES: 3.6 and 4.6 (Cont'd)

Whenever an isotopic analysis is performed, a reasonable effort will be made to determine a significant percentage of those contributors representing the total radioactivity in the reactor coolant sample.

Usually at least 80 percent of the total gamma radioactivity can be identified by the isotopic analysis.

It has been observed that radioiodine concentration can change rapidly in the reactor coolant during transient reactor operations, such as reactor shutdown, reactor power changes, and reactor startup if failed fuel is present. As specified, additional reactor coolant samples shall be taken and analyzed for reactor operations in which steady-state radioiodine concentrations in the reactor coolant indicate various levels of iodine releases from the fuel. Since the radiciodine concentration in the reactor coolant is not continuously measured, reactor coolant sampling would be ineffectivo as a means to rapidly detect gross fuel element failures. However, aume capability to detect gross fuel element failures is inherent in the radiation monitors in the off-gas system on the main steam line.

Materials in the primary system are primarily 304 stainless steel and Zircaloy. The reactor water chemistry limits are established to prevent damage to these materials. The limit placed on chloride concentration is to prevent stress corrosion cracking of the stainless steel.

When conductivity is in its proper normal range (approximately l 10 pmho/cm during reactor startup and 5 mho/cm during power operation), pH and chloride and other impurities affecting conductivity must also be within their normal range. When and if conductivity becomes abnormal, then chloride measurements are made to determine whether or not they are also out of their normal operating values. This would not necessarily be the case. Conductivity could be high due to the presence of a neutral salt, e.g., Na2SO 4 , which would not have an effect on pH or chloride. In such a case, high conductivity alone is not a cause for shutdown. In some types of water-cooled reactors, conductivities are in fact high due to purposeful addition of additives. In the case of BWRs, however, no additives are used and where neutral pH is maintained, conductivity provides a very good measure of the quality of the reactor water.

Significant changes therein provide the operator with a warning mechanism so he can investigate and remedy the condition causing the change before limiting conditions, with respect to variables affecting the boundaries of the reactor coolant, are exceeded.

Methods available to the operator for correcting the off-standard condition include operation of the reactor cleanup system reducing the input of impurities and placing the reactor in the cold shutdown condition. The major benefit of cold shutdown is to reduce the temperature dependent corrosion rates and provide time for the cleanup system to reestablish the purity of the reactor coolant.

During startup periods, which are in the category of less than 100,000 pounds per hour, conductivity may exceed 5 mho/cm because of the initial evolution of gases and the initial addition of dissolved metals. During this period of time when the conductivity exceed 5 mho (other than short term spikes), samples will be taken to assure the chloride concentration is less than 0.1 ppm.

The conductivity of the reactor coolant is continuously monitored.

The samples of the coolant which are taken every 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months will serve as a reference for calibration of these monitors and is considered adequate to assure accurate readings of the monitors. If conductivity is within its normal range, chlorides and other Amendment No. 91, Change 142 (errata) 141

VYNPS BASES: 3.6 and 4.6 (cont'd) impurities will also be within their nornal 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 required by Specification 4.6.B.2 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 domineralizer breakthrough and subsequent conductivity levels in the reactor vessel water.

C. Coolant Leakane 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 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 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 four relief valves are required to provide a pressure margin greater than the recommended 25 psi below the safety valve actuation settings as 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 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 j 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 11% 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.

E. Structural Integrity and Operability Testino l

A pre-service inspection of the components listed in Table 4.2-4 of I the FSAR was conducted after site erection to assure freedom from l defects greater than code allowance; in addition, this serves as a !

reference base for further inspections. Prior to operation, the l l reactor primary system was free of gross defects. In addition, the facility has been designed such that gross defects should not occur Change 16/ March 28, 197 4, +3, 44, 128 142 l

VYNPS BASES: 3.6 and 4.6 (Cont' throughout plant life. The inser/ ice inspection and testing programs are performed in accordance with 10CFR50, Section 50.55a(g) except where specific relief has been granted by the NRC. These inspection and testing programs provide further assurance that gross defects are not occurring and ensure that safety-related components remain operable.

The type of inspection planned for each component depends on location, accessibility, and type of expected defect. Direct visual examination is proposed wherever possible since it is sensitive, fast, and reliable. Magnetic particle and liquid penetrant inspections are planned where practical, and where added sensitivity is required. Ultrasonic testing and radiography shall be used where defects can occur on concealed surfaces.

The in-service inspection and testing programs presented at this time are based on a thorough evaluation of present technology and state-of-the-art inspection and testing techniques.

F. Jet Pumps Failure of a jet pump nozzle assembly hold down mechanism, nozzle assembly and/or riser, would increase the cross-sectional flow area for blowdown following the design basis double-ended line break.

Therefore, if a failure occurred, repairs must be made.

The detection technique is as follows. With the two recirculation pumps balanced in speed to within 15%, the flow rates in both recirculation loops will be verified by main Control Room monitoring instruments. If the two flow rate values do not differ by more than 10%, riser and nozzle assembly integrity has been verified. If they do differ by 10% or more the core flow rate measured by the jet pump diffuser differential pressure system must be checked against the core flow rate derived from the measured value of loop flow to core flow correlation. If the difference between measured and derived core flow rate is 10% or more (with the measured value higher) diffuser measurements will be taken to define the location within the vessel of failed jet pump nozzle (or riser) and the plant shut & wn for repairs. If the potential blowdown flow area is increased, the system resistance to the recirculation pump is also reduced; hence, the affected drive pump will 'run out" to a substantially higher flow rate (approximately 115% to 120% for a single nozzle failure). If the two loops are balanced in flow at the same pump speed, the resistance characteristics cannot have changed. Any imbalance between drive loop flow rates would be indicated by the plant process instrumentation. In addition, the affected jet pump would provide a leakage path past the core thus reducing the core flow rate. The reverse flow through the inactive jet pump would still be indicated by a positive differential pressure but the net effect would be a slight decrease (3% to 6%) in the total core flow measure. This decrease, together with the loop flow increase, would result in a leak of correlation between measured and derived core flow rate.

l The following factors form the basis for the surveillance requirements:

. A break in a jet pump decreases the flow resistance i characteristic of the external piping loop causing the recirculation pump to operate at a higher flow condition when compared to previous operation.

1 Amendment No 45, 128 143 i

VYNPS BASES: 3.6 and 4.6 (Cont'd)

. The change in flow rate of the failed jet pump produces a change in the indicated flow rate of that pump relative to the other pumps in that loop. Comparison of the data with a normal relationship or pattern provides the indication necessary to detect a failed jet pump.

. The jet pump flow deviation pattern derived from the diffuser to lower plenum differential pressure readings will be used to further evaluate jet pump operability in the event that the jet pumps fail the tests in Specifications 4.6.F.1 and 2.

Agreement of indicated core flow with established power-core flow relationships provides the most assurance that recirculation flow is not bypassing the core through inactive or broken jet pumps. This bypass flow is reverse with respect to normal jet pump flow. The indicated total core flow is a summation of the flow indications for the twenty individual jet pumps. The total core flow measuring instrumentation sums reverse jet pump flow as though it were forward flow. Thus, the indicated flow is higher than actual core flow by at least twice the normal flow through any backflowing pump. Reactivity inventory is known to a high degree of confidence so that even if a jet pump failure occurred during a shutdown period, subsequent power ascension would promptly demonstrate abnormal control rod withdrawal for any power-flow operating map point.

A nozzle-riser system failure could also generate the coincident failure of a jet pump body; however, the converse is not true. The lack of any substantial stress in the jet pump body makes failure impossible without an initial nozzle-riser system failure.

G. Sirale Loop Operation Continuous operation with one recirculation loop was justified in

" Vermont Yankee Nuclear Power Station Single Loop Operation",

NEDO-30060, February 1983, with the adjustments specified in Technical Specification 3.6.G.l.a.

APRM and/or LPRM oscillations in excess of those specified in Section 3.6.G.l.b could be an indication that a condition of thermal hydraulic /neutronic instability exists and that appropriate remedial I action should be taken. By restricting core flow to greater than or ;

equal to 34% of rated, which corresponds to the core flow at the 80% l rod line with 2 recirculation pumps running at minimum speed, the region of the power / flow map where these oscillations are most likely to occur is avoided (Region 1 of Figure 3.6.4). These specifications are based upon the guidance of CE SIL #380, Revision 1, dated February 10, 1984.

During single loop op+ ration, the idle recirculation loop is isolated by electrically disa. ming the recirculation pump motor generator set drive motor, until ready to resume two loop operation. This is done to prevent a cold water injection transient caused by an inadvertent pump startup.

Under single loop operation, the flow control is placed in the nanual mode to avoid control oscillations which may occur in the recirculation flow control system under these conditions.

H. Recirculation System The largest recirculation break area assumed in the ECCS evaluation was 4.14 square feet.

Amendment No. M, M, 94 144

t VYNPS I

BASES: 3.6 and 4.6 (Cont'd)

APRM and/or LPRM oscillations in excess of those specified in Section 3.6.H.1 could be an indication that a condition of thermal hydraulic instability exists and that appropriate remedial action should be taken. These specifications are based upon the guidance of CE sib #380, Revision 1, dated February 10, 1984.

Specification 3.6.H.3 restricts reactor operation under natural circulation conditions in order to avoid potential thermal hydraulic /neutronic instabilities.

I. Shock Suppressors (Snubbers)

All snubbers are required operable to ensure that the structural integrity of the Reactor Coolant System and all other safety-related systems is maintained during and following a seismic or other event initiating dynamic loads.

The visual inspection frequency is based upon maintaining a constant level of snubber protection to systems. Therefore, the required inspection interval varies inversely with the observed snubber f ailures and is determined by the number of inoperable snubbers found during an inspection. Inspections performed before that interval has elapsed may be used as a new reference point to determine the next inspeccion. However, the results of such early inspections performed ,

before the original required time interval has elapsed (nominal time less 25%) may not be used to lengthen the required inspection interval. Any inspection whose results require a shorter inspection interval will override the previous schedule.

When the cause of the rejection of a snubber is clearly established and remedied for that snubber and for any other snubbers that may be generically susceptible, and verified by functional testing, that snubber may be exempted from being counted as inoperable.

Generically susceptible snubbers are those which are (1) of a specific make or model, (2) of the same design, and (3) similarly ,

located or exposed to the same environmental conditions such as temperature, radiation, and vibration. These characteristics of the snubber installation shall be evaluated to determine if further functional testing of similar snubber installations is warranted, i When a snubber is found inoperable, an engineering evaluation is performed, in addition to the determination of the snubber mode of failure, in order to determine if any safety-related component or system has been adversely affected by the inoperability of the snubber. The engineering evaluation shall determine whether or not the snubber mode of failure has imparted a significant effect or degradation on the supported compoaent or system.

To provide assurance of snubber functional reliability, a representative sample of the installed snubbers will be functionally tested once each operating cycle. Observed failures of these sample snubbers shall require functional testing of additional units.

J. Thermal Hydraulic Stability Not allowing startup of an idle recirculation pump from natural circulation conditions prevents the rea:tivity insertion transient that would occur.

Amendment No. 24, 49, 69, 94 145

l I

1 l

VYNPS 3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS OPERATION 3.7 STATION CONTAINMENT SYSTEMS 4.7 STATION CONTAINMENT SYSTEMS Applicability: Applicability: I Applies to the operating status Applies to the primary and of the primary and secondary secondary containment system containment systems. integrity.

Obiective: , Obiective:

To assure the integrity of the To verify the integrity of the primary and secondary primary and secondary containment systems. containments.

j Specification: Specification:

1 A. Primary Containment A. Primary Containment

1. Whenever primary 1. The suppression chamber containment is required, water level and the volume and temperature shall be temperature of the water checked once per shift.

in the suppression A visual inspection of chamber shall be the suppression chamber maintained within the interior including water following limits: line regions and the interior painted

a. Maximum Water surfaces above the water Temperature during line shall be made at normal operation - each refueling outage.

1000F. Whenever there is indication of relief

b. Maximum Water valve operation which Temperature during adds heat to the any test operation suppression pool, the which adds heat to pool temperature shall the suppression be continually monitored pool - 1000F. and also observed and logged every 5 minutes

c. If Torus Water until the heat addition Temperature exceeds is terminated. Whenever ,

1100F, initiate an there is indication of immediate scram of relief valve operation the reactor. Power with the temperature of operation shall not the suppression pool be resumed until reaching 1600F or more j the pool and the primary coolant temperature is system pressure greater reduced below tha.n 200 psig, an l 1000P. external visual examination of the

d. During reactor suppression chamber isolation shall be conducr.ed conditions, the before resuming power reactor prensure operation.

vessel shall be depressurized to less than 200 psig at normal cooldown rates if the torus water temperature er.ceeds 1200F.

Amenoment No. M , %, 88 146 l _

VYNPS 3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS OPERATION

e. Minimum Water Volume -

68,000 cubic feet

f. Maximum Water Volume -

l 70,000 cubic feet

2. Primary containment 2. The primary containment integrity shall be integrity shall be maintained at all times demonstrated as required when the reactor is by Appendix J to 10 CFR critical or when the Part 50. The primary reactor water containment shall meet temperature is above the containment 212oF and fuel is in the acceptance requirements reactor vessel except set forth in that while performing low appendix, power physics tests at atmospheric pressure at a. Penetrations and power levels not to seals listed in exceed 5 Mw(t). Table 4.7.1 shall be leak tested at

3. Whenever primary 44 psig (Pa).

containment is required, the total primary b. Type C tests shall >

containment leakage rate be performed on the shall not exceed isolation valves 0.8 weight percent per listed in ,

day (L a) at a pressure Table 4.7.2.a.

of 44 psig (P a)- '

3. Prior to violating the

4. Whenever primary integrity of a system containment is required, outside the primary the leakage from any one containment, which is isolation valve shall connected to any valve not exceed 5 percent of listed in Table 4.7.2b, the maximum allowable the isolation valves leak rate (La ) at peak bounding the opening accident pressure (P a) shall have Type C tests and the leakage from any performed. If the one main steam line opening cannot be isolation valve shall isolated from the not exceed 15.5 scf/hr containment by two at 44 psig (P )a . isolation valves which meet the acceptance criteria of Appendix J (10CFR Part 50), a blank flange shall be ,

installed on the !

opening.

4. The leakage from any one isolation valve shall not exceed 5% of Ltm.

The leakage from any one main steam line isolation valve shall not exceed 11.5 scf/hr t

at 24 psig (Pt). Repair and retest shall be conducted to insure compliance.

Amendment No. 50 147

VYNPS 3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS OPERATION

5. Pressure suppression 5. Pressure suppression Chamber-Reactor Buildinc Chamber - Reactor vacuum Breakers Buildina vacuum Breakers

a. Two of two pressure a. The pressure suppression suppression chamber-reactor chamber-Reactor building vacuum Building vacuum breaker systems breaker shall be operable instrumentation at all times when including setpoint the primary shall be checked containment for proper integrity is operation every required. The three months.

setpoint of the differential b. Operability testing pressure of the vacuum instrumentation breakers shall be which actuates the in accordance with pressure Specification suppression 4.6.E. Each vacuum chamber-reactor breaker shall be building tested to determine air-operated vacuum that the force breakers shall be required to open 50 5 psid. The the vacuum breaker self actuating does not exceed the vacuum breakers force specified in shall open fully Specification when subjected to a 3.7.A.5.a and each force equivalent to vacuum breaker or less than shall be inspected 0.5 psid acting on and verified to the valve disk. meet design requirements,

b. With one Reactor Building -

Suppression chamber vacuum breaker inoperable for opening but known to be closed, restore the inoperable vacuum breaker to OPERABLE status within seven (7) days or then be in cold shutdown within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c. With one Reactor Building -

Suppression chamber vacuum breaker failed open - power operation may continue provided the other vacuum breaker in that Amendment No. 128 148 I

1 i VYNPS 1

3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS l OPERATION line is verified to be closed and conditions required I

by 3.7.D.2 are met.

l l 6. Pressure Suppression 6. Pressure Suppression Chamber - Drywell Vacuum Chamber - Drywell Vacuum Breakers Breakers

a. When primary a. Periodic containment is Operability Tests required, all suppression chamber Operability testing j

- drywell vacuum of the vacuum l breakers shall be breakers shall be operable except in accordance with l during testing and Specification 4.6.E I

as stated in and following any Specifications release of energy 3.7.A.6.b and c, to the suppression ,

below. Suppression chamber. I chamber - drywell Operability of the vacuum breakers corresponding shall be considered position switches operable if: and position l

indicators and (1) The valve is alarms shall be demonstrated verified monthly to open fully and following any with the maintenance.

applied force at all valve b. _Refuelino Outace positions not Tests exceeding I that (1) All .l I

equivalent to suppression 0.5 psi chamber - I acting on the drywell i suppression vacuum i chamber face breaker of the valve position disk. indication and alarm (2) The valve can systems shall be closed by be calibrated gravity, when and released functionally after being tested.

opened by remote or (2) At least two manual means, (2) of the i to within not suppression greater than chamber -

the drywell f equivalent of vacuum 0.05 inch at

breakers all points shall be i along the inspected.

seal surface of the disk.

Amendment No. 128 149

VYNPS 3.7 LIMITING CONDITIONS FOR 4 . '" SURVEILLANCE REQUIREMENTS OPERATION (3) The position If alarm system deficiencies will are found i annunciate in such that the control Specification room if the 3.7.A.6 could valve opening not be met, exceeds the all vacuum equivalent of breakers 0.05 inch at shall be all points inspected and along the deficiencies seal surface corrected.

of the disk.

(3) A dryws11 to

b. Up to two (2) of suppression the ten (10) chamber leak suppression chamber rate test

- drywell vacuum shall breakers may be demonstrate determined to be that with an inoperable provided initial that they are differential secured, or known pressure of to be, in the not less than closed position. 1.0 psi, the differential

c. Reactor operation pressure may continue for decay rate fifteen (15) days shall not provided that at exceed the least one position equivalent of alarm circuit for the leakage each vacuum breaker rate through is operable and a 1-inch each suppression orifice, chamber - drywell vacuum breaker is physically verified to be closed immediately and daily thereafter.

7. Oxvoen Concentration 7. Oxycen Concentration

a. The primary The primary containment containment oxygen concentration atmosphere shall be shall be measured and reduced to less recorded on a weekly

than 4 percent basis.

oxygen with nitrogen gas during reactor power i operation with reactor coolant pressure above 90 psig, except as specified in Specification 3.7.A.7.b.

l Amendment No. 149, 128 150

i I

I i

VYNPS 3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS i OPERATION l

b. Within the 24-hour period subsequent to placing the reactor in the Run mode following a '

shutdown, the containment atmosphere oxygen concentration shall be reduced to less than 4 percent and 1

maintained in this condition.

Deinerting may commence 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to a shutdown.

8. If Specification 3.7.A.1 through 3.7.A.7 cannot be met, an orderly shutdown shall be initiated immediately and the reactor shall be in a cold shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

9. Drywell/ Suppression 9. Drywell/ Suppression Chamber d/p Chamber d/p

a. Differential a. The differential pressure between pressure between the drywell and the drywell and suppression chamber suppression chamber shall be maintained shall be recorded 2,1.7 psi except as once per shift. ,

specified in t 3.7.A.9.b and 3.7.A.9.c below,

b. The 2,1.7 psi b. The operability of differential the low pressure shall be differential ,

established within pressure alarm 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of shall be verified achieving operating once per week.

pressure and ;

temperature. The differential pressure may be reduced to <1.7 psi 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to commencing a cold shutdown.

c. The differential pressure may be reduced to <l.7 psi for a maximum of four hours (period to begin when the l Amendment No. 50 151

I l

l 4

)

VYNPS l I

3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS ;

OPERATION l AP is reduced to l

<l.7) during required operability testing of the HPCI system pump, the RCIC system pump, the drywell-suppression chamber vacuum breakers, and the suppression chamber-reactor building vacuum breakers, and SBGTS testing.

I d. If the specifications of '

3.7.A.9.a cannot be met, and the differential pressure cannot be restored within the subsequent six (6) hour period, an orderly shutdown shall be initiated and the reactor shall be in a Hot Shutdown condition in six (6) hours and a Cold Shutdown condition in the following eighteen (18) hours.

B. Standby Gas Treatment System B. Standby Gas Treatment System

1. Except as specified in 1. At least once per Specification 3.7.B.3 operating cycle, not to below, both circuits of exceed 18 months, the the standby gas following conditions treatment system and the shall be demonstrated.

diesel generators required for operation a. Pressure drop of such circuits shall across the combined be operable at all times HEPA and charcoal when secondary filter banks is containment integrity is less than 6 inches required. of water at 1500 cfm 110%.

b. Inlet heater input is at least 9 kW.

Amendment No. +6, 49, 50 152

VYNPS 3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS OPERATION )

2. a. The results of the 2. a. The tests and in-place cold DOP sample analysis of and halogenated Specification hydrocarbon tests 3.7.B.2 shall be at design flows on performed initially HEPA and charcoal and at least once filter banks shall per operating cycle show >99% DOP not to exceed removal and 2,99% 18 months, and halogenated following painting, hydrocarbon fire or chemical removal. release in any ;

ventilation zone

b. The results of communicating with laboratory carbon the system, while '

sample analysis the system is shall show 2,95% operating, that I

radioactive methyl could contaminate iodide removal. the HEPA filters or (1300C, 95% RH) charcoal adsorbers.

c. System fans shall b. Cold DOP testing be shown to operate shall be performed within 110% of aftur each complete design flow. or partial replacement of the HEPA filter bank.

c. Halogenated '

hydrocarbon testing shall be performed after each complete or partial replacement of the charcoal filter bank.

In addition, the '

sample analysis of Specification i 3.7.B.2.b and the halogenated hydrocarbon test shall be performed after overy 720 hcurs of system operation,

d. Each circuit shall be operated with the heaters on at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months every month.

o. An ultrasonic loak test shall be performed on the gaskets sealing the housing panels downstream of t.he '

l HEPA filters and adsorbers at least 1 Amendment No. 16, 49 153

VYNPS 3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS OPERATION

3. From and after the date once per operating that one circuit of the cycle not to exceed standby gas treatment 18 months. If the system is made or found ultrasonic test to be inoperable for any indicates the reason, reactor presence of a leak, operation is permissible the condition will only during the be evaluated and succeeding seven days the gasket repaired unless such circuit is or replaced as sooner made operable, necessary.

provided that during such seven days all f. DOP and halogenated active components of the hydrocarbon test other stam25y gas shall be performed treatment circuit shall following any be operable. design modificat..on to the Standby Gus

4. If this condition cannot Treatment System be met, procedures shall housing that could be initiated immediately have an effect on to establish the the fi?.ter conditions listed in efficiency.

Specifications 3.7.C.1(a) through (d), g. An air distribution and compliance shall be test demonstrating completed within uniformity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, t20% across the HEPA filters and ,

charcoal adsorbers shall be performed if the SBGTS housing is modified such that air distribution could be affected.

3. a. At least once per operating cycle automatic initiation of each branch of the Standby Gas Treatment System shall be demonstrated.

b. Operability testing of valves shall be in accordance with Specification j 4.6.E.

l l c. When one circuit of I ths Standby Gas Treatment System is made or found to be inoperable, the other circuit shall l have been or shall i

be demonstrated to l be operable within l 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Amendmen? No. 46, 49, 444, 128 154

i VYNPS 3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS OPERATION C. Secondary Containment System C. Secondary Containment System

1. Integrity of the 1. Surveillance of secondary containment secondary containment system shall be shall be performed as maintained during all follows:

modes of plant operation except when all of the a. A preoperational following conditions are secondary mets containment capability test

a. The reactor is shall be conducted suberitical and after isolating the Specification 3.3.A Reactor Building is met and and placing either Standby Gas

b. The reactor water Treatment System temperature is fi.'.ter train in below 2120F and the operation. Such reactor coolant teatr shall system is vented. demonstrate the capability to

c. No activity is maintain a being performed 0.15 inch of water which can reduce vacuum under calm the shutdown margin wind ,

below that (2 < u < 5 mph) specified in condition with a specification filter train flow 3.3.A. rate of not more than 1500 cfm.

d. The fuel cask or irradiated fuel is b. Additional tests not being moved in shall be performed the Reactor during the first Building. operating cycle under an adequate number of different environmental wind conditions to enable valid extrapolation of the test results. .

1

c. Secondary l containment j capability to l maintain a i 0.15 inch of water I vacuum under calm J wind (2<G<5 mph) conditions with a filter train flow rate of not more than 1,500 cfm, shall be demonstrated at least quarterly and at each refueling outage prior to refueling.

Amendment No. 114 155

l 1

l VYNPS l

3.7 LIMITING CONDITIONS FOR 1 4.7 SURVEILLANCE REQUIREMENTS OPERATION

2. Core spray and LPCI pump 2. The core spray and LPCI lower compartment door lower compartment openings shall be closed openings shall be at all times except checked closed daily.

during passage or when reactor coolant temperature is less than 2120F.

D. Primary Containment D. Primary Containment Isolation Valves Isolation Valves

1. During reactor power 1. Surveillance of the operating conditions all primary containment isolation valves listed isolation valves should in Table 4.7.2 and all be performed as follows:

instrument line flow check valves shall be a. The operable operable except as isolation valves specified in that are power Specification 3.7.D.2. operated and automatically initiated shall be tested for automatic initiation and the closure times specified in Table 4.7.2 at least or.ce per operating cycle. ,

b. Operability testing of the primary containment isolation valves shall be performed in accordance with Specification '

4.6.E.

c. At least once per quarter, with the reactor power less than 75 percent of rated, trip all main steam isolation valves (one at a time) and verify closure.

time.

d. At least twice per week, the main steam line isolation valves shall be exercised ,

by partial closure '

and subsequent l reopening. l I

l 1

Amendment No. 54, 66, +28, 134 156 l

VYNPS 3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS OPERATION

2. In the event any 2. Whenever an isolation isolation valve valve listed in 4.7.2 is specified in Table 4.7.2 inoperable, the position becomes inoperable, of at least one other reactor power operation valve in each line may continue provided at having an inoperable least one valve in each valve shall be logged line having an daily.

inoperable valve is in the mode corresponding

to the isolated condition.

3. If Specifications 3.7.D.1 and 3.7.D.2 cannot be 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 .

1 157

- . . . . . . . .- .. _. _ . _ . . _ _ = . . _ - . - _ - - - . . . . .

VYNPS ;

1 1

TABLE 4.7.1 PENETRATIONS AND SEALS SUBJECT TO TYPE B TESTING 1

l Penetration Number of Number Identification Penetrations X-7A, D Main Steam Line A, D 4 X-9A, B Feedwater Line A, B 2 X-11 HPCI Steam Line 1 X-12 Shutdown Cooling Supply 1 X-13A, B RHR Return to Reactor 2 X-14 Supply to Reactor Water Cleanup 1 X-16A, B Core Spray to Reactor 2 X-1 Equipment Access Hatch 1 X-3 Drywell Head Flange 1 X-4 Drywell Head Access Hatch 2 X-6 CRD Removal Hatch 1 SLH-A, H Shear Lug Access Covero 8 X-202A, H&J, K Vacuum Relief Access Covers 10 X-213A, B Torus Drains 2 X-200A, B Torus Manways 2 158

t VYNPS I

^

TABLE 4.7.2.a PRIMARY CONTAIFBfE?TP 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 Sional 1 Main Steam Line Isolation (2-80A, D& 4 4 5 (Note 2) Open GC 2-86A, D) l 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 2 RHR Discharge to Radwaste -( 10- 57, 10-66) 2 25 Closed SG 2 Drywell Floor Drain (20-82, 20-83) 2 20 Open GC 2 Drywell Equipment Drain (20-94, 20-95) 2 20 Open GC 4 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 i 10 Closed

SC (16-19-7) i 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)

J i

Valves 16-19-7 and 16-19-7A shall have stops installed to limit valve opening to 500 or less.

Amendment No. M, M, 131 159

-r VYNPS TABLE 4.7.2.a

(Cont'd)

PRIMARY CONTAitMENT ISOLATION VALVES

! VALVES St.1A7ECT 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 3 Exhaust to Standby Gas Treatment System 1 10 Open GC (16-19-6) 3 Containment Purge Supply (16-19-23) 1 10 Open GC 3 Containment Purge Makeup (16-20-20, 3 NA Closed SC 16-20-22A, 16-20-22B) 5 Reactor Cleanup System (12-15, 12-18) 1 1 25 Open GC l 6 HPCI (23-15, 23-16) 1 1 55 Open GC 6 RCIC (13-15, 13-16) 1 1 20 Open GC Primary / Secondary Vacuum Relief (16-19-11A, 2 NA Closed SC 16-19-11B)

Primary / Secondary Vacuum Relief (16-19-12A, 2 NA Closed Process 16-19-12B) i

' l 3 Containment Air Sampling (VG 23, VG 26, 4 5 Open GC 109-76A&B)

Feedwater Check Valves (V2-27A, -96A, -28A, MA Open Process

-28B)

Amendment No 5B, (rl , 74, 91, 122 160

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

VYNPS TABLE 4.7.2.b PRIMARY CONTAINMENT ISOLATION VALVES VALVES NOT SUBJECT TO TYPE C LEAKAGE TESTS

! Number of Power Or+ rated Valves Maximum Action on Isolation Operating Normal Initiating Group (1) Valve Identification Inboard Outboard Time (sec) Position Signal 2 RHR Return to Suppression Pool (10-39A, B) 2 70 Closed SC 2 RRR Return to Suppression Pool (10-34A, B) 2 120 Closed SC 2 RHR Drywell Spray (10-26A, B& 10-31A, B) 4 70 Closed SC 2 RHR Suppression Chamber Spray (10-38A, B) 2 45 Closed SC 3 Containment Air Compressor Suction (72-38A, 2 20 Open GC B) 4 RHR Shutdown Cooling Supply (10-18, 10-17) 1 1 28 Closed SC Standby Liquid Control Check Valves (11-16, 1 1 NA Closed Proc.

11-17) i

Hydrogen Monitoring (109-75 A, 1-4; 10 NA NA NA l

109-75 B-D, 1-2)

Sampling Valves - Inlet S

Hydrogen Monitoring (VG-24, 25, 33, 34) 4 NA NA la l t l

I l

j

These valves are remote manual sampling valves which do not receive an isolation signal. Only one valve in each line is required to be operable.

l t

Amendment No. GB, 6b . -146, 122 161

VYNPS TABLE 4.7.2 NOTES

1. Isolation signals are as follows:

Group 1: The valves in Group 1 are closed upon any one of the following conditions:

1. Low-low reactor water level

2. High main steam line radiation

3. High main steam line flow

4. High main steam line tunnel temperature

5. Low main steam line pressure (run mode only) l 6. Condenser low vacuum Group 2: The valves in Group 2 are closed upon any one of the following conditions: ,

1. Low reactor water level '

2. High drywell pressure Group 3: The valves in Group 3 are closed upon any one of the following conditions:

1. Low reactor watvar level

2. High drywell proscure

3. High/ low radiation - reactor building ventilation exhaust plenum or refueling floor -

Group 4: The valves in Group 4 are closed upon any one of the following conditions:

1. Low reactor water level ,

2. High drywell pressure

3. High reactor pressure Group 5: The valves in Group 5 are closed upon low reactor water level.

Group 6: The valves in Group 6 are closed upon any signal representing a steam line break in the HPCI system's or RCIC system's ,

respective steam line. The signals indicating a steam line !

break for the respective steam line are as follows:

1. High steam line space temperature

2. High steam line flow

3. Low steam line pressure

4. High temperature in the main steam line tunnel (30 minute ,

delay for the HPCI and the hCIC)

2. The closure time shall not be less than 3 seconds. :

l t

Amendment No. 9 162

VYNPS BASES:

3.7 STATION CONTAINMENT SYSTEMS A. Primary Containment The integrity of the primary containment and operation of the core standby cooling systems in combination limit the off-site doses to ,

values less than to those suggested in 10 CFR 100 in the event of a break in the primary system piping. Thus, containment integrity is specified whenever the potential for violation of the primary reactor system integrity exists. Concern about such a violation exists whenever the reactor is critical, above atmospheric pressure and i temperature above 212oF. An exception is made to this requirement i during initial core loading and while a low power test program is being conducted and ready access to the reactor vessel is required.

The reactor may be taken critical during the period; however, restrictive operating procedures will be in effect again to minimize the probability of an accident occurring. Procedures and the Rod f Worth Minimizer would limit control worth to less than 1.30% delta k.

The pressure suppression pool water provides the heat sink for the reactor primary system energy release following postulated rupture of the system. The pressure suppression chamber water volume must absorb the associated decay and structural sensible heat released >

during primary system blowdown from 1000 psig.

Since all the gases in the drywell are purged into the pressure suppression chamber air space during a loss-of-coolant accident, the pressure resulting from isothermal compression plus the vapor ;

pressure of the liquid must not exceed 62 psig, the allowable pressure suppression chamber pressure. The design volume of the suppression chamber (water and air) was obtained by considering that the total volume of reactor coolant to be condensed is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber (Reference Section 5.2 PSAR).

1 Using the minimum or maximum water volumes given in the specification, containment pressure during the design basis acc isapproximately44psig,which1sbelowthedesignof56psig.{gynt The minimum volume of 68,000 ft 3 results in a submergency approximately four feet. ThemajorityoftheBodegatestsyjIwere run with a submerged length of four feet and with complete condensation. Thus, with respect to downcomer submergence, this specification is adequate. ,

Themaximum]ymperatureattheendofblowdowntestedduringthe Humbolt Bay and Bodega Bay tests was 170*F and this is conservatively taken to be'the limit for complete condensation of the .

reactor coolant, although condensation would occur for temperature above 170'F.

l i

f (1) Robbins, C. H., " Tests on a Full Scale 1/48 Segment of the Humbolt Bay (

Pressure Suppression Containment", GEAP-3596, November 17, 1960.

(2) Bodega Bay Preliminary Hazards Summary Report, Appendix 1, Docket 50-205, December 28, 1962.

(3) Code Allowable peak accident pressure is 62 psig.

Amendment No. Ltr dtd 7/1/85 163 l i

l

VYNPS BASES: 3.7 (Cont'd)

In conjunction with the Mark I Containment Long-Term Program, a plant unique analysis was performed (see Vermont Yankee letter, dated April 27, 1984, transmitting Teledyne Engineering Services Company Reports, TR-5319-1, Revision 2, dated November 30, 1983 and ,

TR-5319-2, Revision 0) which demonstrated that all stresses in the !

suppression chamber structure, including shell, external supports, j vent system, internal structures, and attached piping meet the !

structural acceptance criteria of NUREG-0661. The maintenance of a 1 drywell-suppression chamber differential pressure of 1.7 psid and a suppression chamber water level corresponding to a downcomer submergence range of 4.29 to 4.54 ft. will assure the integrity of i the suppression chamber when subjected to post-LOCA suppression pool '

hydrodynami. forces.

Using a 50*F rise (Section 5.2.4 FSAR) in the suppression chamber ;

water temperature and a minimum water volume of 68,000 ft3, the 1700F '

temperature which is used for complete condensation would be approached only if the suppression pool temperature is 120*F prior to the DBA-LOCA. Maintaining a pool temperature cf 1000F will assure that the 1700F limit is not approached. i Experimental data indicate that excessive steam condensing loads can be avoided if the peak temperature of the suppression pool is l maintained below 1600F during any period of relief valve operation with sonic conditions at the discharge exit. Specifications have been placed on the envelope of reactor operating conditions.so that -

the reactor can be depressurized in a timely manner to avoid the regime of potentially high supprossion chamber loadings.

In addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event a relief valve inadvertently opens or sticks open. This action would includes (1) use of all available means to close the valve, (2) initiate suppression pool water cooling heat exchangers, (3) initiate reactor shutdown, and (4) if other relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open relief valve to assure mixing and uniformity of energy insertion to the pool.

Double isolation valves are provided on lines which penetrate the primary containment and open to the free space of the containment. '

closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system. Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a loss-of-coolant accident. Details of the isolation valves are discussed in Section 5.2 of the FSAR.

~

The purpose of the vacuum relief valves is to equalize the pressure between the drywell and suppression chamber and suppression chamber .

and reactor building so that the structural integrity of the containment is maintained.

Technical Specification 3.7.A.9.c is based on the assumption that the operability tecting of the pressure suppression chamber-reactor I building vaci.um breaker, when required, will normally be performed during the rame four hour testing interval as the pressure suppression chamber-drywell vacuum breakers in order to minimize operation with <l.7 psi, differential pressure.

Amendment No. 44, 64, GB, Ltr dtd 7/1/85 164

i l

VYNPS BASES: 3.7 (Cont'd)

The vacuum relief system from the pressure suppression chamber to Reactor Building consists of twc 100% vacuum relief breakers (2 parallel sets of 2 valves in series). Operation of either system will maintain the pressure differential less than 2 psig; the external design pressure is 2 psig. With one vacuum breaker out of service there is no immediate threat to accident mitigation or primary containment and, therefore, reactor operation can be continued for 7 days while repairs are being made.

The capacity of the ten (10) drywell vacuum zelief valves is sized to 1 limit the pressure differential between the suppression chamber and drywell during post-accident drywell cooling operations to the design limit of 2 psig. They are sized on the basis of the Bodega Bay pressure suppression tests. The ASME Boiler and Pressure Vessel Code, Sect.on III, Subsection B, for this vessel allows eight (8) operable valves, therefore, with two (2) valves secured, containment integrity is not impaired.

Each drywell-suppression chamber vacuum breaker is fitted with a redundant pair of limit switches to provide fail-safe signals to panel mounted indicators in the Reactor Building and alarms in the Control Room when the disks are open more than 0.050" at all points along the seal surface of the disk. These switches are capable of transmitting the disk closed to open signal with 0.01' movement of the switch plunger. Continued reactor operation with failed components is justified because of the redundance of components and circuits and, most importantly, the accessibility of the valve lever arm and position reference external to the valve. The fail safe feature of the alarm circuits assures operator attention if a line fault occurs.

The requirement to inert the containment is based on the recommendation of the Advisory Committee on Reactor Safeguards. This recommendation, in turn, is based on the assumption that several percent of the zirconium in the core will undergo a reaction with steam during the loss-of-coolant accident. This reaction would release sufficient hydrogen to result in a flammable concentration in the primary containment building. The oxygen concentration is therefore kept below 4% to minimize the possibility of hydrogen combustion. ,

General Electric has estimated that less than 0.1% of the zirconium would react with steam following a loss-of-coolant due to operation of emergency core cooling equipment. This quar'.ity of zirconium would not liberate enough hydrogen to form a combustible mixture.

B. and C. Standby Gas Treatment System and Secondary Containment Svstam The secondary containment is designed to minimize any ground level release of radioactive materials which might result from a serious accident. The Reactor Building provides secondary containment during reactor operation, when the drywell is sealed and in service; the Reactor Building provides primary containment when the reactor is shutdown and the drywell is open, as during refueling. Because the secondary containment is an integral part of the complete containment system, secondary containment is required at all times that primary containment is required except, however, for initial fuel loading and low power physics testing.

Amendment No. 49, Bases Change, 119 165 1

VYNPS l

BASES: 3.7 (Cont'd) ,

l The standby gas treatment system is designed to filter and exhaust the Reactor Building atmosphere to the stack during secondary '

containment isolation conditions, with a minimum release of radioactive materials from the Reactor Building to the environs. To insure that the standby gas treatment system will be effective in removing radioactive contaminates from the Reactor Building air, the system is tested periodically to meet the intent of ANSI N510-1975.

Both standby gas treatment fans are designed to automatically start upon containment isolation and to maintain the Reactor Building >

pressure to approximately a negative 0.15 inch water gauge pressure; all leakage should be in-leakage. Should the fan fail to start, the redundant alternate fan and filter system is designed to start automatically. Each of the two fans has 100% capacity. This substantiates the availability of the operable circuit and results in no added risk; thus, reactor operation or refueling operation can continue. If neither circuit is operable, the plant is brought to a condition where the system is not required.

D. Pr ima ry Containment Isolation Valves Double isolation valves are provided on lines that penetrate the primary containment and communicate directly with the reactor vessel and on lines that penetrate the primary containment and communicate with the primary containment free space. Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system. Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a loss-of-coolant accident.

4.7 STATION CONTAINMENT SYSTEMS A. Primary Containment System The water in the suppression chamber is used only for cooling in the event of an accident, i.e., it is not used for normal operation; therefore, a weekly check of the temperature and volume is adequate to assure that adequate heat removal capability is present.

The interiors of the drywell and suppresalon chamber are painted to prevent rusting. The inspection of the paint during each major refueling outage, approximately once per year, assures the paint is intact. Experience with this type of paint at fossil fueled

, generating stations indicates that the inspection interval is adequate.

Because of the large volume and thermal capacity of the suppression pool, the volume and temperature normally changes very slowly and monitoring these parameters daily is sufficient to establish any temperature trends. By requiring the suppression pool temperature to be continually monitored and frequently logged during periods of significant heat addition, the temperature trends will be closely followed so that appropriate action can be taken. The requirement for an external visual examination following any event where potentially high loadings could occur provides assurance that no significant damage was encountered. Particular attention should be focused on structural discontinuities in the vicinity of the relief valve discharge since these are expected to be the points of highest stress. Visual inspection of the suppression chamber including water line regions each refueling outage is adequate to detect any changes in the suppression chamber structures.

Amendment No. +6, 49 166

l VYNPS l

BASES: 4.7 (Cont'd)

The primary containment preoperational test pressures are based upon l the calculated primary containment pressure response in the event of I a loss-of-coolant accident. The peak drywell pressure would be about ,

44 psig which would rapidly reduce to 27 psig within 10 seconds following the pipe break. Following the pipe break, the suppression chamber pressure rises to 27 psig within 10 seconds, equalizes with drywellpressuryyndthereforerapidlydecayswiththedrywell pressure decay.

I ThedesiggIpressure 56 psig.

of the drywell and absorption chamber is The design leak rate is 0.5%/ day at a pressure of 62 psig. As pointed out above, the pressure response of the drywell and suppression chamber following an accident would be the same after

( about 10 seconds. Based on the primary containment pressure response and the fact that the drywell and suppression chamber function as a j

, unit, the primary containment will be tested as a unit rather than the individual components separately.

The design basis loss-of-coolant accident was evaluated at the primary containment maximum allowable accident leak rate of 1.5%/ day at 44 psig. The analysis showed that with this leak rate and a standby gas treatment system filter efficiency of 90% for halogcns, 95% for particulates, and assuming the fission product release fractions stated in TID-14844, the maximum total whole body passing j cloud dose is about 1.65 rem and the maximum total thyroid dose is I about 280 rem at the site boundary over an exposure duration of two hours. The resultant dose that would occur for the duration of the I I

accident at the low population distance of 5 miles is lower than those stated due to the variability of meteorological conditions that l would be expected to occur over a 30-day period. Thus, these doses l are the maximum that would be expected in the unlikely event of a 1 design basis loss-of-coolant accident. These doses are also based on the assumption of no holdup in the cecondary containment, resulting in a direct release of fission products from the primary containment q through the filters and stack to the environs. Therefore, the i specified primary containment leak rate and filter efficiency are l conservative and provide margin between expected off-site doses and i 10 CFR 100 guidelines. An additional factor of two for conservatism is added to the above doses by limiting the test leak rate (L a) to a '

value of 0.80%/ day. ,

(1) Section 5.2 of the FSAR.

(2) 62 psig is the maximum allowable peak accident pressure for this design (56 psig) pressure.

Amendment No. 131 167

VYNPS i

BASES: 4.7 (Cont'd)

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 value will be 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 tent intervals, the maximum allowable operational leak rate (Ltm), which 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 rate test program of such penetration is conducted at the peak accident pressure of 44 psig to insure not only that the leakage remains acceptably low but also that tha sealing materials 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 conjunction with Specification 4.6.E. Inspections and calibrations 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 psid 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 allowablg bypass area between the drywell and suppression chamber as 0.12 ft l

This is equivalent to one vacuum breaker open by three-eighths of an l inch (3/8*) as measured at all points around the circumference of the I disk or three-fourths of an inch (3/4*) as measured at the bottom of l 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.

l Amendment No. 50, 128 168 l

I

VYNPS BASES: 4.7 (Cont'd)

At the end of each refueling cycle, a leak rate test shall be performed to verify that significant leakage flow paths do not exist between the drywell and suppression chamber. The drywell pressure will be increased by at least 1 psi with respect to the suppression chamber pressure and held constant. The 2 psig set point will not be exceeded. The subsequent suppression chamber pressure transient (if !

any) will be monitored with a sensitive pressure guage. If the l drywell pressure cannot be increased by 1 psi over the suppression chamber pressure it would be because a significant leakage path exists; in this event the leakage source will be identified and eliminated before power operation is resumed. If the drywell pressure can be increased by 1 psi over the suppression cramber the rate of change of the suppression chamber pressure must not exceed a rate equivalent to the rate of leakage from the drywell through a i

1-inch orifice. In the event the rate of change exceeds this value I then the source of leakage will be identified and eliminated before haver eperation is resumed.

The drywell-suppression chamber vacuum breakers ate exercised in accordance with Specification 4.6.E and immediately following i termination of discharge of steam into the suppression chamber. This monitoring of valve operability is intended to assure that valve )

operability and position indication system performance does not 4 degrade between refueling inspections. When a vacuum breaker valve I is exercised through an opening-closing cycle, the position indicating lights are designed to function as follows:

Full Closed 2 White - On (Closed to 50 050' open)

Open 2 White - Off

(>0.050' open to full open)

During each refueling outage, two drywell-suppression chamber vacuum breakers will be inspected to assure sealing surfaces and components have not deteriorated. Since valve internals are designed for a 40-year lifetime, an inspection program which cycles through all valves in one-eighth of the design lifetime is extremely conservative. .

Experience has shown that a weekly measurement of the oxygen concentration in the primary containment assures adequate surveillance of the primary containment atmosphere.

B. and C. Standbv Cas Treatment System and Secondary containment System Initiating reactor building isolation and operation of the standby gas treatment system to maintain at least a 0.15 inch of water vacuum within the secondary containment provides an adequate test of the operation of the reactor building isolation valves, leakage tightness of the reactor building, and performance of the standby gas treatment system. Functionally testing of initiating sensors and associated trip channels demonstrates the capability for automatic actuation.

Periodic testing gives sufficient confidence of reactor building integrity and standby gas treatment system performance capability.

d Amendment No. 128 169

VYNPS PASES: 4.7 (Cont'd)

The test frequencies are adequate to detect equipment deterioration prior to significant defects, but the tests are not frequent enough to load the filters, thus reducing their reserve capacity too quickly. That the testing frequency is adequate to detect deterioration was demonstrated by the tests which showed no loss of filter efficiency after 2 years of operation in the rugged shipboard environment on the NS Savannah (ORNL 3726). Pressure drop tests across filter sections are performed to detect gross plugging of the filter media. Considering the relatively short time that the fans may be run for test purposes, plugging is unlikely, and the test interval is reasonable Such heater tests will be conducted once during each operating cycle. Considering the simplicity of the heating circuit, the teat frequency is sufficient. Air distribution tests will be conducted ance during each operating cycle.

The in-place testing of charcoal filters is performed using a l halogenated hydrocarbon, which is injected into the system upstream of the charcoal filters. Measurements of the challenge gas concentration upstream and downstream of the charcoal filters is made. The ratio of the inlet and outlet concentrations gives an overall indication of the leak tightness of the system. Although this is basically a leak test, since the filters have charcoal of known efficiency and holding capacity for elemental iodine and/or methyl iodine, the test also gives an indication of the relative efficiency of the installed system.

High-efficiency particulate air filters are installed before and after the charcoal filter to minimize potential release of particulates to the environment and to prevent clogging of the iodine filters. An efficiency of 99% is adequate to retain particulates that may be released to the Reactor Building following an accident.

This will be demonstrated by testing with DOP as testing medium.

The efficiencies of the particulate and charcoal filters are sufficient to prevent exceeding 10CFR100 limits for the accidents analyzed. The analysis of post-accident hydrogen purge assumed a charcoal filter efficiency of 95%. Hence requiring in-place test efficiencies of 99% for these filters provides adequate margin. The laboratory methyl iodide removal test is porformed at 95% relative humidity to assure adequate margin over the design relative humidity of 70%.

The test interval for filter efficiency was selected to minimize plugging of the filters. In addition, testing for methyl iodide removal efficiency will be demonstrated. This will be done either by removal of a charcoal sample cartridge which contains charcoal equivalent to the bed thickness or removing one adsorber tray from ;

the system and using the charcoal therein, after mixing, to obtain at l least two samples equivalent to the bed thickness. Any HEPA filters found defective should be replaced with filters qualified according to Regulatory Position C.3.d of Regulatory Guide 1.52. If laboratory 1 test results are unacceptable, all charcoal adsorbent in the system I should be replaced with charcoal adsorbent qualified according to Regulatory Guide 1.52.

Amendment No. 15 170 I

_ _ _ _ _ _ _ _ _ _ _ _ _ ____-__A

l J

VYNPS BASES: 4.7 (Cont'd)

D. Primary Containmont Isolation Valves l Those large pipes comprising a portion of the reactor coolant system whose failure could result in uncovering the reactor core are !

supplied with automatic isolation valves (except those lines needed for emergency core cooling system operation or containment cooling). l The closure times specified herein and per Specification 4.6.E are '

adequate to prevent loss of more cooling from the circumferential rupture of any of these lines outside the containment than from a steam line rupture. Therefore, the isolation valve closure times are sufficient to prevent uncovering the core.

Purge and vent valve testing performed by Allis-Chalmers has demonstrated that all butterfly purge and vent valves installed at Vermont Yankee can close from full open conditions at design basis containment pressure. However, as an additional conservative measure, limit stops have been added to valves 16-19-7/7A, limiting the opening of these valves to 50* open while operating, as requested by NRC in their letter of May 22, 1984. (NVY 84-108)

In order to assure that the doses that may result from a steam line break do not exceed the 10CFR100 guidelines, it is necessary that no fuel rod perforation resulting from the accident occur prior to closure of the main steam line isolation valves. Analyses indicate the fuel rod cladding perforations would be avoided for the main steam valve closure times, including instrument delay, as long as 10.5 seconds. The test closure time limit of five seconds for these main steam isolation valves provides sufficient margin to assure that cladding perforations are avoided and 10CFR100 limits are not exceeded. Redundant valves in each line ensure that isolation will be effected applying the single failure criteria.

The main steam line isolation valves are functionally tested on a more frequent interval to establish a high degree of reliability.

The containment is penetrated by a large number of small diameter instrument lines. The flow check valves in these lines are tested for operability in accordance with Specification 4.6.E.

l l

1 l

Amendment No. 91, 128 171

1 I

VYNPS j 3.8 LIMITING CONDITIONS FOR 4.8 SURVEILLANCE REQUIREMENTS OPERATION 1 l'

3.8 RADIOACTIVE EFFLUENTS 4.8 RADIOACTIVE EFFLUENTS Applicability: ApplicabilitV:

Applies to the release of all Applies to the required radioactive effluents from the surveillance of all radioactive plant. effluents released from the plant.

Obiective Obiective:

To assure that radioactive To ascertain that all i

effluents are kept 'as low as is radioactive effluents released reasonably achievable" in from the plant are kept 'as low accordance with 10CFR50, as is reasonably achievable" in Appendix I and, in any event, accordance with 10CFR50, are within the limits specified Appendix I and, in any event, I l in 10CFR20. are within the limits specified l in 10CFR20.

Specification: Specification A. Liquid Effluents: A. Liquid Effluents:

Concentration concentration

1. The concentration of 1. Radioactive material in radioactive matcrial in liquid waste shall be liquid effluents sampled and analyzed in released from the site accordance with shall be limited to the requirements of concentrations specified Table 4.8.1. The in 10CFR Part 20, results of the analyses Appendix B, Table II, shall be used in Column 2 for accordance with the radionuclides other than methods in the ODCM to noble gases and assure that the j 2x10'4 uci/ml total concentrations at the '

activity concentration point of release are for all dissolved or limited to the values in entrained noble gases. Specification 3.8.A.1.

2. With the concentration of radioactive material in liquid effluents released from the site exceeding the limits of Specification 3.8.A.1, immediately take action to decrease the release rate of radioactive materials and/or increase the dilution flow rate to restore the concentration to within the above limits.

Amendment No. 83 172

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

J VYNPS 3.8 LIMITING CONDITIONS FOR 4.8 SURVEILLANCE REQUIREMENTS 1 OPERATION B. Liquid Effluents: Dose B. Liquid Effluents: Dose

1. The dose or dose 1. Cumulative dose commitment to a member contributions shall be of the public from determined in accordance radioactive materials in with the methods in the liquid effluents ODCM at least once per released from the site month if releases during shall be limited to the the period have followings occurred.

a. During any calendar quarters less than or equal to 1.5 mrom to the total body, and less than or equal to 5 mrem to any organ, and

b. During any calendar year:

less than or equal to 3 mrem to the total body, and less than or equal to 10 mrem to any organ.

C. Liquid Radwaste Treatment C. Liquid Radwaste Treatment

1. The liquid radwaste 1. See Specification treatment system shall 4.8.B.1.

be used in its designed modes of operation to reduce the radioactive naterials in the liquid waste prior to its discharge when the !

estimated doses due to the liquid effluent from the site, when averaged with all other liquid release over the last month, would exceed .

0.06 mrem to the total body, or 0.2 mrem to any '

organ.

1 I

1 Amendment No. 83 173

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

VYNPS 3.8 LIMITING CONDITIONS FOR 4.8 SURVEILLANCE REQUIREMENTS OPEFATICN D. Eicruid Holdup Tanks D. Liquid Holdup Tanks

1. The quantity of 1. The quantity of !

radioactive material radioactive material ;

contained in any outside contained in each of the tank

shall be limited liquid holdup tanks * ;

to less than or equal to shall be determined to i 10 curies, excluding be within the limits of tritium and dissolved or Specification 3.8.D.1 by i entrained noble gases. analyzing a ,

representative sample of i With the quantity of the tank's contents ;

2.

radioactive material in within one week any outaide tank

following the addition exceeding the limit of of radioactive materials i specifice. tion 3.8.D 1, to the tank. One sample immediately take action may cover multiple to susper.d all additions additions.

of radioa tive material ,

to the tar.k. Within !

48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , reduce the tank contents to within the limit.

E. Caseous Effluents Dose E. Gaseous Effluents: Dose Rate Rate

1. The dose rate due to 1. The dose rate due to radioactive materials noble gases in gaseous 4 released in gaseous effluents shall be effluents from the site determined to be within to areas at and beyond the limits of '

the site boundary shall Specification 3.8.E.1 in be limited to the accordance with the followings methods in the ODCM.

a. For noble gases; 2. The dose rate due to i less than or equal Iodine-131, Iodine-133, ,

to 500 mrem /yr to tritium and the total body and radionuclides in less than or equal particulate form with .

to 3,000 mrem /yr to half-lives greater than '

the skin, and 8 days in gaseous effluents shall be determined to be within the limits of

Specification 3.8.E.1 in >

i

NOTE: Tanks included in this Specification are only those outdoor tanks that are not surrounded by liners, dikes, or walls capable of holding l the tank's contents, or that do not have tank overflows and l surrounding area drains connected to the liquid radwaste treatment i system. j Amendment No. 63, 95 174

i e

l VYNPS l

4 3.8 LIMITING CONDITIONS FOR 4.8 SURVEILLANCE REQUIREMENTS i OPERATION

b. For Iodine-131, accordance with the !

Iodine-133, tritium methods in the ODCM by and radionuclides obtaining representative in particulate form samples and performing with half-lives analyses in accordance greater than with the sampling and 8 days; less than analysis program ,

or equal to specified in 1,500 mrem /yr to Table 4.8.2. '

any organ.

2. With the dose rate (s) exceeding the above ,

limits, immediately take {

action to decrease the ;

release rate to within the limits of Specification 3.8.E.1.

F. Gaseous Effluents: Dose from Noble Cases F. Gaseous Effluents: Dose from Noble Cases

1. The air dose due to l noble gases released in 1. Cumulative dose

, gaseous effluents from contributions for the the site to areas at and total time period shall !

beyond the site boundary be determined in shall be limited to the accordance with the followings methods in the ODCM at least once every month,

a. During any calendar ,

quarter:

less than or equal .

to 5 mrad for gamma ,

radiation, and less than or equal r

to 10 mrad for beta radiation, and

b. During any calendar ;

year. ,

less than or equal l to 10 mrad for gamma radiation, and l 1ess than or equal ,

to 20 mrad for beta l radiation. )

l l

Amendment No. 83 175 l

l VYNPS 3.8 LIMITING CONDITIONS FOR 4.8 SURVEILLANCE REQUIREMENTS '

OPEPATION G. Gaseous Effluents: Dose G. Gaseous Effluents: Dose from Iodine-131, Iodine-133, from Iodine-131, Iodine-133, Tritium, and Radionuclides Tritium, and Radionuclides i in Particulate Form in Particulate Form

1. The dose to a member of 1. Cumulative dose the public from contributions for the Iodine-131, Iodine-133, total time period shall .

tritium, and be determined in i radionuclides in accordance with the particulate form with methods in the ODCM at half-lives greater than least once every month.

8 days in gaseous effluents released from the site to areas at and beyond the site boundary shall be limited to the following: ;

a. During any calendar quarter:

less than or equal to 7.5 mrem to any organ, and

b. During any calendar year:

less than or equal [

to 15 mrem to any !

organ.

9. Gaseous Radwaste Treatment H. Gaseous Radwaste Treatment

1. The Augmented Off-Gas 1. The readings of the System (AOG) shall be relevant instrument used in its designed shall be checked every mode of operation to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the main reduce noble gases in condenser SJAE is in use gaseous waste prior to to ensure that the AOG ,

their discharge whenever is functioning.

the main condenser steam '

jet air ejector (SJAE) is in operation.

I. Ventilation Exhaust I. Ventilation Exhaust Treatment Treatment t

1. The AOG and Radwaste 1. See Specification Building Ventilation 4.8.F.1 for surveillance Filter (HEPA) Systems related to AOG and shall be used to reduce Radwaste Building particulate materials in ventilation filter ,

gaseous waste prior to system operation.

their discharge from those buildings when the i estimated doses due to l

gaseous effluent l releases from the site l

! I Amendment No. 83 176 l

t :

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

VYNPS 3.8 LIMITING CONDITIONS FOR 4.8 SURVEILLANCE REQUIREMENTS OPERATION to areas at and beyond the site boundary would exceed 0.3 mrem to any organ over one month.

J. Explosive cas Mixture J. Explosive Gas Mixture

1. If the hydrogen 1. The concentration of concentration in the hydrogen in the off-gas off-gas downstream of system downstream of the the operating recombiner recombiners shall be reaches four percent, continuously monitored take appropriate action by the hydrogen monitor that will restore the required operable by concentration to within Table 3.9.2. ,

the limit within i 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . I K. Steam Jet Air Eiector K. Steam Jet Air Eiector (SJAE) (SJAE)

1. Cross radioactivity 1. The gross radioactivity release rate from release rate shall be the SJAE shall be continuously monitored limited to less in accordance with than or equal to Specification 3.9.B.

0.16 C1/sec (after 30 minutes decay).

2. With the gross 2. The gross radioactivity radioactivity release rate of noble release rate at the gases from the SJAE SJAE exceeding the shall be determined to above limit, be within the limit of restore the gross Specification 3.8.K.1 at radioactivity the following release rate to frequencies by within its limit performing an isotopic within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or analysis (for Xe-138, i be in at least Hot Xe-135, Xe-133, Kr-88, Standby within the Kr-85m, Kr-87) on a subsequent representative sample of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . gases taken at the discharge.

3. With the gross radioactivity release a. Once per week.

rate at the SJAE greater b. Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months than or equal to following an 1.5 Ci/sec (after increase of 25% or 30-minute decay), 5000 restore the gross microcuries/sec, radioactivity release whichover is rate to less than greater, in 1.5 Ci/sec (after steady-state t 30-minute decay), or be activity levels in Hot Standby within during steady-state 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , reactor operation, 7 as indicated by the SJAE monitor.

Amendment No. 83 177

VYNPS l

3.8 LIMITING CONDITIONS FOR 4.8 SURVEILLANCE REQUIREMENTS OPERATION L. Primary Containment L. Primary Containment i

1. If the primary 1. The primary containment containment is to be shall be sampled prior Vented / Purged, it shall to venting / purging per be Vented / Purged through Table 4.8.2, and if the the Standby Gas results indicate Treatment System radioactivity levels in ,

whenever the airborne excess of the limits of ;

radioactivity levels in Specification 3.8.L.1, containment exceed the the containment shall be levels specified in aligned for 10CFR20, Appendix B, venting / purging through Table I, Column 1 and the Standby Gas notes 1-5 thereto. Treatment System. No sampling shall be

2. With the requirements of required if the Specification 3.8.L.1 venting / purging is not satisfied, through the Standby Gas immediately suspend all Treatment (SBGT) System.

Venting / Purging of the containment.

3. During normal refueling and maintenance outages when primary containment is no longer required, then Specification 3.8.G shall supersede Specifications 3.8.L.1 and 2.

M. Total Dose M. Total Dose

1. The dose or done 1. Cumulative dose commitment to a member contributions from of the public* from all liquid and gaseous ,

station sources is effluents shall be limited to less than or determined in accordance equal to 25 mrem to the with Specifications total body or any organ 4.8.B.1, 4.8.F.1, and (except the thyroid, 4.8.G.1.

which is limited to less than or equal to 75 mrem) over a calendar ,

year. '

i F

NOTE: For this Specification a member of the public may be taken as a real individual accounting for his actual activities.

Amendment No. 83 178 ,

1 VYNPS I

3.8 LIMITING CONDITIONS FOR 4.8 SURVEILLANCE REQUIREMENTS OPERATION

2. With the calculated dose 2. Cumulative dose i from the release of contributions from radioactive materials in direct radiation from liquid or gaseous plant sources shall be effluents exceeding determined in accordance twice the limits of with the methods in the Specifications ODCM. This requirement 3.8.B.1.a, 3.8.B.1.b, is applicable only under 3.8.F.1.a, 3.8.F.1.b, conditions set forth in 3.8.G.1.a, or 3.8.G.1.b, Specification 3.8.M.2.

calculations should be made, including direct radiation contributions from the station to determine whether the above limits of Specification 3.8.M.1 have been exceeded.

N. Solid Radioactive Waste N. Solid Radioactive Waste

1. The solid radwaste 1. Verification of system shall be used in solidification of wet accordance with a waste shall be performed Process Control Program as required and in as described in accordance with the Section 6.12 to process Process Control Program.

wet radioactive waste (spent resins / filter sludges) to meet shipping and burial ground requirements.

2. With the provisions of Specification 3.8.N.1 not satisfied, suspend shipments of defectively processed or defectively packaged solidified wet radioactive wastes from the site.

l l

1 l

Amendment No. 83 179 1

l 1

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

l VYNPS TABLE 4.8.1 RADIOACTIVE LIOUID WASTE SAMPLING AND ANALYSIS PROGRAM i

Lower Limit Minimum Type of of Detection Liquid Sampling Analysis Activity (LLD)

Release Type Frequency Frequency Analysis (uci/ml)a ,

Batch Waste Prior to each Prior to Principal 5 x 10~7 Releage release Each each release Gamma Tanks Batch Each Batch Emitterad I-131 1 x 10-6 One Batch per Once per Dissolved and 1 x 10-5 month sampled month Dtrained prior to a Gates (Camma release Emitters)

Prior to each Once per H-3 1 x 10-5 release Each month Composite c Batch Gross Alpha 1 x 10~7 Prior to each once per Sr-89, Sr-90 5 x 10-8 release Each quarter Batch Composite c Fe-55 1 x 10-6 i

i i

l i

b Amendment No. 83 180

VYNPS TABLE 4.8.1 NOTATION:

a. The LLD is the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be -

detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system (which may include radiochemical separation):

LLD = 4.66

Sb l

E

V

K

Y

e ~ ^

d' where:

LLD = the lower limit of detection as defined above (microcuries or picoeuries/ unit mass or volume)

Sb = the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (counts / minute)

E = the counting efficiency (counts / disintegration) r V = the sample size (units of mass or volume)

K = 2.22 x 106 disintegrations / minute / microcurie or 2.22 disintegration / minute /picoeurie as applicable Y = the fractional radiochemical yield (when applicable)

A = the radioactive decay constant for the particular radionuclide

(/ minute)

At = the elapsed time between sample collection and analysis (minutes)

Typical values of E, V, Y and At can be used in the calculation. In calculating the LLD for a radionuclide determined by gamma-ray spectrometry, the background shall include the typical contributions of other radionuclides normally present in the samples. ,

Analysis shall be performed in such a manner that the stated LLDs will be '

achieved under routine conditions. Occasionally, background ,

fluctuations, unavoidably small sample sizes, the presence of interfering '

nuclides, or other uncontrollable circumstances may render these LLos-unavailable.

It should be recognized that the LLD is defined as a "before the fact' limit representing the capability of a measurement system and not as an "after the fact' limit for a particular measurement. This does not preclude the calculation of an "after the fact" LLD for a particular measurement based upon the actual parameters for the sample in question and appropriate decay correction parameters such as decay while sampling and during analysis,

b. A batch release is the discharge of liquid wastes of a discrete volume.

Prior to sampling for analysis, each batch shall be isolated and then thoroughly mixed to assure representative sampling.

Amendment No. 83 181

VYNPS TABLE 4.8.1 NOTATION: (Cont'd) r

c. A composite sample is one in shich the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is representative of the liquids released. Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent release. ,

d. The principal gamma emitters for which the LLD specification will apply are exclusively the following radionuclides: Mn-54, Fe-59, co-58, Coabu, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are '

below the LLD for the analyses should not be reported as being present at the LLD level, but as "not detected". When unusual circumstances result in LLDs higher than required, the reasono shall be documented in the Semiannual Effluent Release Report, r

b I

4 Amendment No. 83 182

VYNPS j

TABLE 4.8.2 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM i

Lower Limit Minimum Type of of Detection i Gaseous Sampling Analysis Activity (LLD)

Release Type Frequency Frequency Analysis (uCi/ml)a A. Steam, Jet Once per week once per Xe-138, 1 x 10-4 Air E)ector Grab Sample week Xe-135, .

Xe-133, 1 Kr-88, l Kr-8',

Kr-85M B. Containment Prior to each Prior to Principal 1 x 10-4 Purge release Each each release Gamma Purge Grab Each Purge Emitters d Sample C. Main Plant Once ger Once Principal 1 x 10-4 Stack month Grab monthger Gamma ;

Sample Emittersd H-3 1 x 10-6 Continuous

Once per I-131 f 1 x 10-12 weekb Charcoal ,

Sample [

Continuous

Once per Principal 1 x 10~11 ,

week Gamma Particulate Emitters d Sample (I-131)

Continuous' Once per Gross Alpha 1 x 10-11 ,

month <

Composite Particulate Sample Continuoui.* Once per Sr-89, Sr-90 1 x 10-11 quarter Composite >

Particulate i Sample Continuous Noble Gas Noble Gases 1 x 10-5 [

Monitor Gross Beta !

or Gamma l

l l

i Amendment No. 83 183

VYNPS TABLE 4.8.2 NOTATION:

I

a. See footnote a. of Table 4.8.1.

b. Samples shall be changed at least once per 7 days and analyses shall be l completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after removal from samplers. Sampling shall also be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for at least 7 days following each shutdown, startup or thermal power change exceeding 25% of rated thermal power in one hour, and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of changing the samples. When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding LLDs may be increased by a factor of 10.

This requirement to sample at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for 7 days applies only if: (1) analysis shows that the dose equivalent I-131 concontration in the primary coolant has increasedmoretganafactorof3andthe resultant concentration is at least 1 x 10~ pCi/ml; and (2) the noble gas monitor shows that effluent activity has increased more than a factor of 3.

c. Sampling and analyses shall also be performed following shutdown, startup, or a thermal power change exceeding 25% of rated thermal power per hour unless: (a) analysis shows that the dose equivalent I-131 concentration in the primary coolant has not increased mgre than a factor of 3 and the resultant concentration is at least 1 x 10- pCi/ml; and (2) the noble gas monitor shows that effluent activity has not increased more than a factor of 3.

d. The principal gamma emitters for which the LLD specification will apply are exclusively the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135 and Xe-138 for gaseous emissions, and Mn-54, Fe-59, Co-58, co-60, 2n-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions. This list does not mean that on1' these nuclides are to be detected and reported. Other peaks which ar ha'surable and identifiable, together with the above nuclides, shall c1so be identified and reported. Nuclides which are below LLD for the anais.es should not '

be reported as being present at the LLD level for that nuclide, but as "not detected". When unusual circumstances result in LLDs higher than required, the reasons shall be documented in the Semiannual Ef fli ent Release Report.

e. The ratio of the sample flow rate to the sampled stream flow rata chall be known for the time period covered by each dose or dose rato calculation made in accordance with Specifications 3.8.E.1, 3.6.F.1 and 3.8.G.I.

f. The gaseous waste sampling and analysis program does not explicitly require sampling and analysis at a specified LLD to determine the I-133 release. Estimates of I-133 releases shall be determined by counting the weekly charcoal sample for I-133 (as well as I-131) and assume a constant release rate for the release period.

Amendment No. 83 184

. _ _ . .___ _ . ~ _

VYNPS BASES:

3.8 RADIOACTIVE EFFLUENTS A. Liquid Effluents: Concentration This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site above background (at the point of discharge from the plant discharge into Connecticut River) will be less than the concentration levels specified in 10CFR Part 20, Appendix B, Table II, Column 2.

This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will result in exposure within (1) the Section II.A design objectives of Appendix I, 10CFR Part 50, to a member of the public, and (2) the limits of 10CFR Part 20.106 (e) to the population.

The concentration limit for noble gases is based upon the assumption that Xe-135 is the controlling radionuclide and its MPC in air (submersion) was converted to an equivalent concentration in water using the International Commission on Radiological Protection (ICRP)

Publication 2.

B. Liquid Effluents: Dose This specification is provided to implement the requirements of Sections II.A, III.A, and IV.A of Appendix I, 10CFR Part 50. The Limining Condition for Operation implements the guides set forth in Sect.on II.A of Appendix I. The requirements provide operating d

ficxibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid 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 an individual through appropriate pathways is unlikely to be substantially underestimated. In addition, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in potable drinking water that are in excess of the requirements of 40CFR 141. No drinking water supplies drawn from the Connecticut River below the plant have been identified. The appropriate dose equations for implementation through requirements of the Specification sre described in the Vermont Yankee Off-Site Dose Calculation Manual. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents were developed from the methodology provided in Regulatory Guide 1.109,

" Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR '

Part 50, Appendix I", Revision 1, October 1977 and Regulatory Guide 1.113, " Estimating Aqustic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I", Revision 1, April 1977.

C. Liquid Radwaste Treatment The requirement that the appropriate reiLions of this system as indicated in the ODCM be used, when specified, provides assurar.ce j that the releases of radioactive materials in liquid effluent < will be kept "as low as is reasonably achievable". This specification ,

implements the requirements of 10CFR Part 50.36a and the der.ign objective given in F-ntion II.D of Appendix I to 10CFR Part 50. The l specified limits e s..ing the use of appropriate portions of the l Amendment No. 83 185

VYNPS BASES: 3.8 (Cont'd) 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 Holdur) Tanks The tanks listed in this Specification include all outdoor tanks that contain radioactivity that are not surrounded by liners, dikes, or walls capable of holding the tank contents, or that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system.

Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10CFR Part 20, Appendix B, Table II, Column 2, at the nearest portable 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 limits 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 assurance that radioactive material discharged in gaseous effluents will not result in the exposure of member (s) of the public either within or outside the site boundary, to annual average concentrations 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 *.o 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 radioactive 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 i

VYNPS BASES: 3.8 (Cont'd) specified in the ODCM for calculating the doses due to the actual releases of radioactive noble gases in gaseous effluents. The ODCM also provides for determining the air doses at the site boundary based upon the historical average atmospheric conditions.

The equations specified in the ODCM for calculating the doses due to the actuai release rates of radioactive noble gases in gaseous effluentr. were developed from the methodology provided in Regulatory Guide 1.109, ' Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR Part 50, Appendix I", Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors,' Revision 1, July 1977.

G. Gaseous Effluents: Dose from Iodine-131, Iodine-133, Tritium, and Radionuclides in Particulate Form This specification is provided to implement the requirements of Section II.C, III.A, and IV.A of Appendix I, 10CFR Part 50. The Limiting condition for Operation are the guides set forth in Section II.C 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 radioactive materials 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 that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a member of the public through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of the subject materials were also developed using the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR Part 50, Appendix I', Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors,' Revision 1, July 1977. These equations also provide for determining the actual doses based upen the historical average atmospheric conditions. The release rate specifications for Iodine 131, Iodine-133, tritium, and radionuclidas in particulate form with half-lives greater than 8 days are depender,t on the existing radionuclide pathways to man, in areas at and beyond its site boundary. The pathways which were examined in the development of these specifications were: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, 3) depcsition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man.

H. Gaseous Radwaste Treatment The requirement that the appropriate portions of the Augmented Off-Gas (AOG) System be used whenever the SJAE is in operation provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "a low as is reagenably achievable". This specification implements the requirements of 10CFR Part 50.36a and the design objectives of Appendix I to 10CFR Part 50.

Amendment No. 83 187

l VYNPS BASES: 3.8 (Cont'd)

I. Ventilation Exhaust Treatment The requirement that the AOG Building and Radwaste Building HEPA filters be used when specified provides reasonable assurance that the release of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable *. This specification implements the requirements of 10CFR Part 50.36a and the design objective of Appendix I to 10CFR Part 50. The requirements governing the use of the appropriate portions of the gaseous radwaste filter systems were specified by the NRC in NUREG-0473, Revision 2 (July 1979) ae a '

suitable fraction of the guide set forth in Sections II.B and II.C of '

Appendix I, 10CFR Part 50, for gaseous effluents.

J. Explosi"y G 4_ Mixture The hy6u Jen ronitors are used to detect possible hydrogen buildups which could result in a possible hydrogen explosion. Automatic isolation of the off-gas flow would prevent the hydrogen explosion and possible damage to the augmented off-gas system. Maintaining the concentration of hydrogen below its flammability limit provides assurance that the releases of radioactive materials will be controlled.

K. Steam Jet Air Eiector (S3AE)

Restricting the gross radioactivity release rate of gases from the main condenser SJAE provides reasonable assurance that the total body exposure to an individual at the exclusion area boundary will not exceed a small fraction of the limits of 10CFR Part 100 in the event this effluent is inadvertently discharged directly to the environment without treatment. This specification implements the requirements of General Design Criteria 60 and 64 of Appendix A to 10CFR Part 50.

b. Primary Containment (MARK I)

This specification provides reasonable assurance that releases from containment purging / venting operations will be filtered through the Standby Gas Treatment System so that the annual dose limits of 10CFR Part 20 at the site boundary will not be exceeded. The dose objectives of Specification 3.8.G restrict purge / venting operations when the Standby Gas Treatment System is not in use and gives ,

reasonable assucance that all releases from the plant will be kept *

as low as is reasonably achievable".

M. Total Dese This specification is provided to meet the done limitations of 40CFR Part 190 that have been incorporated into 10CFR Part 20 by 46 FR 18525. The specification requires the preparation and submittal of a Specific Report whenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix I. For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a member of the public will exceed the dose limits of 40C7R Part 190 if the individual reactors remain within the reporting requirement level. The Special Report will describe a course of action that should result in the limitation of the annual dose to a member of the public to within the 40CFR Part 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the member of the public is estinated to exceed the requirements of 40CFR Part 190, the Special >

Report with a request for a variance (provided the release conditions resulting in violation of 40CFR Part 190 have not already been Amendment No. 83 188

i VYNPS BASES: 3.8 (Cont'd) corrected), in accordance with the provisions of 40CFR Part 190.11 and 10CFR Part 20.405c, is considered to be a timely request and fulfills the requirements of 40CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40CFR Part 190, and does r-* spply in any way to the other requirements for dose limitation of .'Or.d Part 20, as addressed in Specification 3.8.A and 3.8.E. An individual is not considered a member of the public during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.

N. Solid Radioactive Waste This specification implements the requirements of 10CFR Part 50.36a with respect to the handling of solid radioactive waste (spent resin and filter sludges only). The establishment and implementation of a Process Control Program (PCP), provides the operational guidelines by which proper dewatering of filter media and spent resins in preparation for off-site disposal is assured.

l Amendment No. 83 189

VYNPS l

3.9 LIMITI!JG CONDITIONS FOR 4.9 SURVEILLANCE REQUIREMENTS l OPERATION 3.9 RADIOACTIVE EFFLUENT MONITORING 4.9 _ RADIOACTIVE EFFLUENT MONITORING SYSTEMS SYSTEMS Applicability: Applicability:

Applies to the monitoring Applies to the required systems or programs which surveillance of the monitoring perform a surveillance, systems or programs which protective or controlling perform a surveillance, function on the release of protective or controlling radioactive effluents from the function on the release of plant and their identification radioactive effluents from the in the environment. plant and their identification in the environment.

Obiectives Obiective:

To assure the operability of the To specify the type and radioactive effluent monitoring frequency of surveillance to be systems and environmental applied to the radioactive programs, effluent monitoring system and environmental programs.

Specifications: Specifications:

A. Liquid Effluent A. Liquid Effluent Instrumentation Instrumentation

1. During periods of 1. Each radioactive liquid release through the effluent monitoring monitored pathway, the instrumentation channel radioactive liquid shall be tested and effluent monitoring calibrated as indicated instrumentation channel in Table 4.9.1.

shall be operable in accordance with Table 3.9.1 with their alarm setpoints set to ensure that the limits of Specification 3.8.A.1 are not exceeded.

B. Gaseous Effluent B. Gaseous Effluent Instrumentation Instrumentation

1. The gaseous process and 1. Each gaseous process or effluent monitoring effluent monitoring instrumentation channels instrumentation channel shall be operable in shall be tested and accordance with calibrated as indicated Table 3.9.2 with their in Table 4.9.2.

alarm / trip setpoints set to ensure that the i limits of Specifications 3.8.E.1.a, 3.8.J.1, and 3.8.K.1 are not exceeded, i

l Amendment No. 83 190 !

VYNPS 3.9 LIMITING CONDITIONS FOR 4.9 SURVEILLANCE REQUIREMENTS OPERATION C. Radioloqical Environmental C. Radiolocical Environmental Monitorino Proqram Monitorina Procram

1. The radiological 1. The radiological environmental monitoring environmental monitoring program shall be samples shall be conducted as specified collected pursuant to in Table 3.9.3. Table 3.9.3 from the locations given in the ODCM and shall be analyzed pursuant to the requirements of Table 3.9.3 and the ,

detection capabilities required by Table 4.9.3.

?

D. Land Use Census D. Land Use Census

1. A land use census shall 1. The land use census be conducted to identify shall be conducted at the location of the least once per year nearest milk animal and between the dates of the nearest residence in June 1 and October 1 by each of the either a door-to-door 16 meteorological survey, aerial survey, ;

sectors within a or by consulting local distance of five miles. agricultural The survey shall also authorities. The 1 identify the nearest results of the land use milk animal (within census shall be included 3 miles of the plant) to in the annual-the point of predicted Radiological highest annual average Environmental D/Q value in each of the Surveillance Report three major pursuant to ,

meteorological sectors Specification 6.7.C.3. ;

due to elevated releases '

from the plant stack.

2. With a land use census identifying one or more i

locations which yield a ;

calculated dose or dose i commitment (via the same exposure pathway) at least 20 percent greater than at a location from which samples are currently being obtained in accordance with Specification 3.9.C.1, l

add the new location (s) to the radiological environmental monitoring program within 30 days if permission from the owner to collect samples can be obtained, and sufficient sample volume is available. The sampling location (s), l excluding the control Amendment No. 83 191

VYNPS 3.9 LIMITING CONDITIONS FOR 4.9 SURVEILLANCE REQUIREMENTS OPERATION station location, having the lowest calculated dose or dose commitment (via the same exposure pathway) may be deleted '

from this monitoring program after October 31 of the year in which this land use census was conducted.

E. Intercomparison Procram E. Intercomparison Procram

1. Analyses shall be 1. A summary of the results performed on referenced of analyses performed as radioactive materials part of the above supplied as part of an required Intercomparison Intercomparison Program Program shall be which has been approved included in the Annual by NRC. Radiological Environmental Surveillance Report.

The identification of the NRC approved Intercomparison Program which is being participated in shall be stated in the ODCM.

t I

Amendment No. 83 192

VYNPS TABLE 3.9.1 RADIOACTIVE LIOUID EFFLUENT MONITORING INSTRUMENTATION i

i Minimum !

Channels i Operable Notes !

l

1. Gross Radioactivity Monitors not i Providing Automatic Termination of Release

a. Liquid Radwaste Discharge Monitor 1* 1,4,5

b. Service Water Discharge Monitor 1 2,4,5

2. Flow Rate Measurement Devices

a. Liquid Radwaste Discharge Flow 1* 3,4 Rate Monitor i

During releases via this pathway.

l l

Amendmont No. 83 193

(

VYNPS TABLE 3.9.1 NOTATION NOTE 1 - With the number of channels operable less 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 r

o. At least two technically qualified members of the Facility Staff ind.ipendently verify the release rate calculations and discharge lina valving.

Otherwise, suspind release of radioactive effluents via this pathway.

I NOTE 2 - With the number of channels operable less 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 samplesarecollectedandanalyzedforgrossradioacgivity (beta or gamma) at a lower limit of detection of at least 10- microcurie /ml.

NOTE 3 - With the number of channels operable less than required by the minimum channels operable requirement, effluent releases via this pathway may continue 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 be used to estimate flow.

NOTE 4 - With the number of channels operable less than required by the minimum channels operable requirement, exert reasonabic efforts to return the instrument (s) to operable status prior to the next release.

NOTE 5 - The alarm setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the Off-Site Dose Calculation Manual (ODCM). With a radioactive liquid -

effluent monitoring instrumentation channel alarm setpoint less ,

conservative than a value which will ensure that the limits of e.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 change the !

setpo'.nt so it is acceptably conservative.

l 1

Amendment No. 83 194 1

VYNPS I TABLE 3.9.2 CASEOUS EFFLUENT MONITORING INSTRUMENTATION Minimum Channels ,

Instrument Operable Notes !

1. Steam Jet Air Ejector (SJAE)

a. Noble Gas Activity Monitor 1 7, 8, 9 ,

2. Augmented Off-Gas System

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

b. Flow Rate Monitor 1 1, 5, 6

c. Hydrogen Monitor 1 3, 5, 6

3. Plant Stack l a. Noble Gas Activity Monitor 1 5, 7, 10

b. Iodine Sampler Cartridge 1 4, 5

c. Particulate Sampler Filter 1 4, 5

d. Sampler Flow Integrator 1 1, 5 ;

e. Smack Flow Rate Monitor 1 1, 5 i

?

1 l

Amendment No. 84, 103 195 1

VYNPS TABLE 3.9.2 NOTATION NOTE 1 - With the number of channels operable less than required by the minimum channels operable requirement, effluent releases via this pathway may continue 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 .

NOTE 2 - With the number of channels operable less than required by the ;

minimum channels operable requirement, effluent releases via this pathway may continue 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.

NOTE 3 - With the number of channels operable less than required by the ,

minimum channels operable requirement, operation of the AOG System l may continue provided gas samples are collected at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and analyzed within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or an orderly ,

transfer of the off-gas effluents from the operating recombiner to the standby recombiner shall be made.

NOTE 4 - With the number of channels operable less than required by the minimum channels operable requirement, effluent releases via the affected pathway may continue provided samples are continuously collected with auxiliary sampling equipment.

NOTE 5 - 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 within 30 days.

NOTE 6 - During releases via this pathway.

NOTE 7 - The alarm / trip setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the Off-Site Dose Calculation Manual (ODCM). With a gaseous process or effluent monitoring instrumentation channel alarm / trip setpoint less '

conservative than a value which will ensure that the limits of 3.8.E.1.a and 3.8.K.1 are met, immediately take actions to suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable, or chango the i setpoint so it is acceptably conservative.

NOTE 8 - Minimum channels operable required only during operation of the Steam Jet Air Ejector.

i NOTE 9 - With the number of channels operable less than required by the minimum channels operable requirement, gases from the SJAE may be ;

released to the environment for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided: ,

l

1. The AOG System is not bypassed; and

2. The AOG System noble 9as activity monitor is operable.

NOTE 10 - With the number of channels operable less than required by the minimum channels operable requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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Amendment No. 64, 103 196

1 VYNPS TABLE 3.9.3 RADIOLOGICAL ErNIPONMENTAL MONITORING PROGRAM l

Exposure Pathway Number of Sample Sampling and Collection Type and Frequency and/or Sample Locations ^ Frequency of Analysis

1. AIRBOPJIE

a. Radioiodine and Samples from 5 locations: Continuous operation of Radioiodine canister:

Particulates sampler with sample Analyze each sarple for 1 sample from up valley, collection semimonthly or I-131.

within 4 miles of Site more frequently as Particulate sampler:

Boundary. (major wind required by dust loading Gross beta radioactivity direction) or plant gffluent releases.

analysis on each sample following filter change.c 1 sample from down valley, Composite (bylocagion) within 4 miles of Site for gamma isotopic at l Boundary. (major wind least once per quarter.

f direction) i

! I 1 sample each from the -

vicinity of two nearby communities, within 10 miles of Site Boundary.

1 sample from a control location.

Amendment No. 64, 103 197

VYNPS l TABLE 3.9.3 (Cont'ds PADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM i

Exposure Pathway Number of Sample Sampling and Collection Type and Frequency and/or Sample Locations" Frequency of Analysis

2. DIRECT RADIATION D 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 meteorlogical sector) locations, de-dose only within a range of 0 to quarterly unless gaseous 4 km9; release LCO was exceeded in period.

16 incident response stations (one in each meteorlogical sector) within a range of 2 to 8 km9; the balance of the stations to be placed in special interest areas and control station areas.

Amendment No. 83 198

VYNPS TABLE 3.9.3 (Cont'd)

FADIOLOGICAL E?NIROtNENTAL MONITORING PROGRAM

! Exposure Pathway Number of Sample Sampling and Collection Type and Frequency and/or Sample Locations ^ Frequency of Analysis ,

3. WATEREORNE

a. Surface

1 sample upstream. Monthly grab sample. Gamma isotopic analysisd of each sample. Tritium analysis of composite 1 sample downstream. Composite sample collected sample at least once per over quarter.

monthg period of one

b. Ground 1 sample from within 8 km Quarterly. Gamma isotopied and distance. tritium analyses of each 1 sample from a control Quarterly. ,

location.

c. Sediment from 1 sample from downstream Semiannually. Gamma isotopic analysis d i Shoreline area with existing or of each sample. j potential recreational value.

1 sample from north storm As specified in the ODCM.

drain outfall.

I Amendment No. 83 199

VYNPS TABLE 3.9.3 (Cont'd) 4 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Ntunber of Sample Sampling and Collection Type and Frequency and/or Sample Locations

Frequency of Analysis

4. INGESTION

a. Milk Samples from milking Semimonthly if milking Gamma isotepied and I-131 animals in 3 locations animals are identified on analysis of each sample.

within 5 km distance pasture; at least once per i having the highest dose month at other times.

potential. If there are less than 3 primary locations available then 1 i or more secondary sample from milking animals in each of 3 areas between 5 to 8 km distance where doses are calculated to be greater than 1 mrem per year.

1 sample from milking animals in a control location.

b. Fish 1 sample of two Semiannually. Gamma isotopic analysisd on edible portions. I recreationally important I species in vicinity of plant discharge area. t 1 sample (preferably of same species) in areas not influenced by plant .

i discharge. .

c. Vegetation 1 grass sample at each air Quarterly when available. Gamma isotopic analysisd sampling station. of each sample.

1 silage sample at each At time of harvest. Gamma isotopic analysisd milk sampling station (as of each sample.

l available). ,

Amendment No. 83 200

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

VYNPS TABLE 3.9.3 NOTATION a Specific parameters of distance and direction sector from the centerline of the reactor and additional descriptions where pertinent, shall be provided for each and every sample location in Table 3.9.3 in a table and figure (s) in the ODCM. Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, malfunction of automatic sampling equipment and other legitimate reasons. If specimens are unobtainable due to sampling equipment malfunction, every reasonable effort shall be made to complete corrective action prior to the end of the next sampling period. All deviations from the sampling schedule shall be documented in the annual Radiological Environmental Surveillance Report pursuant to Specification 6.7.C.3. It is recognized that, at times, it may not be possible or practicable to continue to obtain samples of the media of choice at the most desired location or time. In these instances, suitable alternative media and locations may be chosen for the particular pathway in question and appropriate substitutions made within 30 days in ,

the radiological environmental monitoring program. In lieu of a Licensee Event Report and pursuant to Specification 6.7.C.1, identify the cause of the unavailability of samples for that pathway and identify the new location (s) for obtaining replacement samples in the next Semiannual Radioactive Effluent Release Report and also include in the report a revised figure (s) and table for the ODCM reflecting the new location (s),

b one or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate continuously may be used in place of, or in addition to, integrating dosimeters. For the purposes of this table, a Thermoluminescent Dosimeter (TLD) is considered to be one phosphor; two or more phosphors in a packet are considered as two or more dosimeters.

Film badges shall not be used as dosimeters for measuring direct radiation. The 40 stations is not an absolute number. The frequency of analysis or readout for TLD systems will depend upon the characteristics of the specific system used and should be selected to obtain optimum dose information with minimal fading.

C Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or more after sampling to allow for radon and thoron daughter decay. If gross beta activity in air particulate samples is greater than ten times the yearly mean of control samples, ganna isotopic analysis shall be performed on the individual samples.

d Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility.

The ' upstream sample' shall be taken at a distance beyond significant influence of the discharge. The " downstream" sample shall be taken in an area buyond but near the mixing zone.

f Composite sample aliquota shall be collected at time intervals that are very short (e.g., hourly) relative to the compositing period (e.g.,

monthly) in order to assure obtaining a representative sample.

9 Each meteorological sector shall have an established " inner" and an

" outer" monitoring location based on ease of recovery (i.e., response time) and year-round accessibility.

h Sample collection will be performed weekly whenever the main plant stack effluent release rate of I-131, as determined by the sampling and analysig program ofSample 1 x 10- uCi/sec.

4.8.2, is equal Table collection to or greater than will revert back to semimonthly no sooner than at least two weeks after the plan rateofI-131fallsandremainsbelow1x10gstackeffluentrelease uCi/sec.

Amendment No. 83 201 l

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VYNPS TABLE 3.9.4 I

PEPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN EtNIFONMENTAL SAMPLES *I Reporting Levels Airborne Particulate or Fish Milk Vegetation Sediment Analysis Water (pCi/1) Gases (pci/m3) (pCi/Kg, wet) (pCi/l) (pCi/Kg, wet) (pCi/Kg, dry)

H-3 2 x 104(b)

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 10 2 1 x 10 4 3 x 103(C)

Zn-65 3 x 10 2 2 x 10 4 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 1 x 10 3 Cs-137 50 20 2 x 10 3 70 2 x 10 3 Ba-La-140 2 x 10 2 3 x 10 2 (a) Reporting levels way 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.

4 (b) Reporting level for drinking water pathways. For nondrinking water pathways, a value of 3 x 10 pCi/l may be used.

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

Amendment No. G3, 103 202

i VYNPS TABLE 4.9.1 RADIOACTIVE LTOUID EFFLUENT MONI'IORING INSTRUMENTATION SURVEILLANCE REOUIRDfEPTrs Instrument Instrument Instrument Instrument Check Source Check Calibration Functional Test

1. Gross Radioactivity Monitors not Providing Automatic Termination of Release

a. Liquid Radwaste Discharge Once each day

Prior to each once each Once each Monitor (3) release, but no 18 months (1) quarter (2) more than once each month

b. Service Water Discharge Monitor Once each day once each month once each once each (3) 18 months (1) quarter (2)

2. Flow Rate Measurement Devices

a. Liquid Radwaste Discharge Flow Once each day Not Applicable Not Applicable Once each Rate Monitor quarter *

During releases via this pathway.

Amendment No. 83 203

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

VYNPS TABLE 4.9.1 NOTATION (1) The Instrument Calibration for radioactivity measurement instrumentation shall include the use of a known (traceable to National Bureau of Standards) liquid radioactive source positioned in a reproducible i geometry with respect to the sensor. These standards shall permit calibrating the system over its normal operating range of energy and rate.

(2) The Instrument Functional Test shall also demonstrate the control Room alarm annunciation occurs if any of the following conditions exists:

L (a) Instrument indicate measured levels above the alarm setpoint.

(b) Circuit failure.

(c) Instrument indicates a downscale failure.

(d) Instrument controls not set in operate mode.

(3) The alarm setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the off-Site Dose Calculation Manual (ODCM). ,

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Amendment No. 83 204 !

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VYNPS TABLE 4.9.2 GASEOUS EFFLUENT MONI% RING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Instrument Instrument Instrument Instrument Check Source Check Calibration Functional Test

1. Steam Jet Air Ejector (SJAE)

a. Noble Gas Activity Monitor Once each day ** Once each month once each once each 18 months (3) quarter (2)

2. Augmented Off-Gas System

a. Noble Gas Activity Monitor Once each day

Once each month once each once each 18 months (3) quarter (1)

b. Flow Rate Monitor Once each day

Not Applicable Once each Not Applicable 18 months

c. Hydrogen Monitor Once each day

Not Applicable Once each Once each month quarter (4)

3. Plant Stack

a. Noble Gas Activity Monitor Once each day once each month once each once each 18 months (3) quarter (2)

b. Sampler Flow Integrator Once each week Not Applicable Once each Not Applicable 18 months

c. System Flow Rate Monitor Once each day Not Applicable Not Applicable Not Applicable

During releases via this pathway.

- During operation of main condenser SJAE.

Amendment No. 83 205

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

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 exists:

(a) Instrument indicate measured levels above the alarm setpoint, r (b) Circuit failure.

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

(b) Circuit failure.

(c) Instrument indicates a downscale failure.

(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.

(4) The Instrument Calibration shall include the use of standard gas samples (high range and low range) containing suitable concentrations, hydrogen balance nitrogen, for the detection range of interest per Specification 3.8.J.l.

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Amendment No. 83 206 l

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VYNPS TABLE 4.9.3 NOTATION (a) See Footnote (a) of Table 4.8.1.

(b) Parent only.

(c) If the measured concentration minus the 5 sigma counting statistics is found to exceed the specified LLD, the sample does not have to be analyzed to meet the specified LLD.

(d) This list does not mean that only these nuclides are to be considered.

Other peaks that are identifiable, together with those of the listed nuclides, shall also be analyzed and reported in the Annual Radiological Environmental Surveillance Report pursuant to Specification 6.7.C.3.

(e) The Ba-140 LLD and concentration can be determined by the analysis of its short-lived daughter product La-140 subsequent to an 8 day period following collection. The calculation shall be predicted on the normal ingrowth equations for a parent-daughter situation and the assumption that any unsupported La-140 in the sample would have decayed to an ,

insignificant amount (at least 3.6 percent of its original value). The ingrowth equations will assume that the supported La-140 activity at the time of the collection is zero. '

(f) Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD, but as "not detected". For purposes of averaging, the LLD will be assumed to be zero, e

i 1

i

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j Amendment No. 83 208

VYNPS j i

BASES:

3.9 RADIOACTIVE EFFLUENT MONITORING SYSTEMS A. Liquid Effluent Instrumentation The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The alarm setpoints for these instruments are to ensure that the alarm will occur prior to exceeding the limits of 10CFR Part 20.

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 of exceeding the 10CFR Part 20 concentration limits are low, and the potential off-site dose consequences are also low.

B. Gaseous Effluent Instrumentation The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive !

materials in gaseous effluents during actual or potential releases of gaseous effluents. The alarm / trip setpoints for these instruments i are provided to ensure that the alarm / trip will occur prior to exceeding the limits of 10CFR Part 20. This instrumentation also includes provisions for monitoring (and controlling) the concentrations of potentially explosive gas mixtures in the waste gas holdup system. r C. Radioloaical Environmental Monitoring Procram The radiological monitoring program required by this specification provides measurements of radiation and of radioactive materials in those exposure pathways 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 levels of radiation are not higher than expected on the basis of the .'

effluent measurements and modeling of the environmental exposure pathways.

l 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 i ALARA and environmental reporting levels. This radiological environmental monitoring program has therefore been significantly modified as provided for by Regulatory Guides 4.3 (C.2.a) and 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 teg years of operation. An I-131 release rate trigger value of 1 x 10- uCi/sec from the plant stack will require that air sample collection be increased to weekly. The i

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Amendment No. 83 209 !

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VYNPS BASES: 3.9 (Cont'd) 1 x 10-1 uCi/gec I-131 value corresponds to the LLD air concentration [

of 0.07 pci/m at the maximum predicted air monit which exhibitsamaximumquarterlyX/Qvalueof2x10-9ringagation, sec/m . A factor of 3.5 below the LLD value has also been included in the stack release rate value to account for meteorological fluctuations in X/0 Due to the large local population of cows and the ready availability of milk samples, food product sampling has been eliminated from the program in lieu of milk sampling. Since milking cows in the area ;

~

spend very little time on pasture, silage and grass sampling have been instituted as an indicator of radionuclide deposition.

The detection capabilities required by Table 4.9.3 are considered optimum for routine environmental measurements in industrial laboratories. It should be recognized that the LLD is defined as 4 before-the-fact limit representing the capability of a measurement system and not as an after-the-fact limit for a particular measurement. This does not preclude the calculation of an after-the-fact LLD for a particular measurement based upon the actual parameters for the sample in question.

D. Land Use Census This specification is provided to ensure that changes in the use of areas at and beyond the site boundaries are identified and that modifications to the monitoring program are made if required by the results of this census. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50. The requirement of a '

garden census has been eliminated along with the food product monitoring requirement due to the substantial and widespread !

occurrence of dairy farming in the surrounding area which dominates the food uptake pathway.

The addition of new sampling locations to Specification 3.9.C, based ,

on the land use census, is limited to those locations which yield a calculated dose or dose commitment greater than 20 percent of the calculated dose or dose commitment at any location currently being sampled. This eliminates the unnecessary changing of the environmental radiation monitoring program for new locations which, within the accuracy of the calculation, contributes essentially the same to the dose or dose commitment as the location already sampled.

The substitution of a new sampling point for one already sampled when the calculated difference in dose is less than 20 percent, would not be expected to result in a significant increase in the ability to detect plant effluent related nuclides.

E. Intercomparison Procram The requirement for participation in an intercomparison program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of a quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid for the purposes of Section IV.B.2 of Appendix I to 10 CFR Part 50.

1 Amendment No. 83 210

VYNPS 3.10 LIMITING CONDITIONS FOR 4.10 SURVEILLANCE REQUIREMENTS OPERATION 3.10 AUXILIARY ELECTRICAL POWER 4.10 AUXILIARY ELECTRICAL POWER SYSTEMS SYSTEMS Applicability: Anplicability:

Applies to the auxiliary Applies to the periodic testing electrical power systems. requirements of the auxiliary electrical power systems.

Obiectivo: Obiective:

To assure an adequate supply of To verify the operability of electrical power for operation the auxiliary electrical power of those systems required for systems.

reactor safety.

Specification: Specification:

A. Normal Operation A. Normal Operation -

The reactor shall not be 1. Diesel Generators made critical unless all of the following conditions a. Each diesel are satisfied. generator shall be started and loaded

1. Diesel Generators once a month to.

demonstrate Both emergency diesel operational generators shall be readiness. The operable and capable of test shall starting and reaching continue until the rated voltage and diesel engine and f requency in not more the generator are than 13 seconds. at equilibrium temperature at

, expected maximum emergency loading not to exceed the continuous rating.

During this test, the diesel starting time to reach rated voltage and frequency shall be logged and the air compressor shall be checked for operation and its ability to recharge air ,

receivers. The l diesel fuel oil l transfer pumps l shall be tested in i accordance with Specification 4.6.E.

Amendment No. 128 211 i

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l VYNPS 3.10 LIMITING CONDITIONS FOR 4.10 SURVEILLANCE REQUIREMENTS OPERATION

b. The undervoltage automatic starting circuit of each diesel generator shall be tested once a month,

c. Once per operating i eycle, the actual conditions under which the diesel generators are required to start automatically will be simulated and a test conducted to demonstrate that they will start within 13 seconds and accept the emergency load and start each load within the specified starting time. The results shall be logged.

2. Battery systems 2. Battery Systems The following battery a. Every veek the systems shall be specific gravity, operable: temperat.tre, level, ar.d voltage

a. The four Neutron of the pilot cell Monitoring and . and overa'.1 Process Radiation battery vcitage Batteries, shall be miasured associated and logged.

chargers, and 24 VDC b. Every three months Distribution the voltage, Panels. temperature, level, and

b. The two main specific gravity station battery of each cell, and systems consisting overall battery of: voltage shall be measured and

1. Battery A1, logged.

Battery Charger A or Spare Charger AB and Bus DC-1.

2. Battery B1, Battery Charger B or Spare Charger AB and Bus DC-2.

Amendment No. GB, 125 212

VYNPS 3.10 LIMITING CONDITIONS FOR 4.10 SURVEILLANCE REQUIREMENTS l OPERATION i

c. Two Switchyard c. Once per operating Batteries each with cycle each ECCS one associated battery, Alternate charger and its Shutdown AS-2 associated DC battery, and Main distribution panel. Station battery shall be subjected

d. Both ECCS to a Service (Load Instrumentation Profile) discharge batteries, test. The specific associated gravity and voltage chargers, and of each cell shall distribution be measured after panels. the recharge at the end of the

e. The Alternate discharge test and Shutdown AS-2 logged, battery, one of the two associated d. Once every five chargers, and DC years, each ECCS, Distribution panel UPS, AS-2, and Main DC-2AS. Station Battery shall be subjected

f. Both UPS batteries, to a Performance associated (capacity)

Uninterruptible Discharge Test.

Power Supplies and This test will be MCC 89A and B. performed in lieu of the Service Test requirements of 4,10.A.2.c above,

e. Each 480 V Uninterruptible Power System shall be checked daily.

f. 480 V Motor Control Centers 89A and 89B shall be checked daily.

g. Once per operating cycle, the actual conditions under which the 480 V Uninterruptible Power Systems are required will be simulated and a test conducted to demonstrate equipment performance.

Amendment No. 68, 125 213

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l VYNPS 3.10 LIMITING CONDITIONS FOR 4.10 SURVEILLANCE REQUIREMENTS OPERATION

3. Emoraency Buses 3. Emeroency Buses The emergency 4160 volt The emergency 4160 volt Buses 3 and 4, and buses and 480 volt 480 volt Buses 8 and 9 buses shall be checked shall be energized and daily, operable.

4. Off-Site Power 4. Off-Site Power

a. At least one The status of the off-site off-site power sources

, transmission line shall be checked daily.

and at least one start-up transformer in service.

b. One of the following additional sources of delayed access power:

The main stepup transformer and unit auxiliary transformer available and capable of supplying power to the emergency 4160 volt buses or, The 4160 volt tie line to Vernon Hydro-Electric station capable of supplying power to either of the two emergency 4160 volt buses.

5. Reactor Protection System Power Protection 5. Reactor Protection System Power Protection Two RPS power protection panels for once per operating each inservice RPS MG cycle, the operability set or alternate power of each overvoltage, source shall be undervoltage, and operable. underfrequency protective device shall be demonstrated by the performance of an instrument channel calibration test.

Settings shall be ,

verified to be in i accordance with l

Table 4.10.1.

Amendment No. e6, 144. 125 214

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

VYNPS 3.10 LIMITING CONDITIONS FOR 4.10 SURVEILLANCE REQUIREMENTS ,

OPERATION j B. Operation With Inoperable B. Operation With Inoperable !

Components Components Whenever the reactor is in Rur. Mode or Startup Mode with the reactor not in the i Cold condition, the l requirements of 3.10.A 1 shall be met except l

1. Diesel Generators 1. Diesel Generator From and after the date When one of the diesel that one of the diesel generators is made or generators or its found to be inoperable, associated buses are the requirements of made or found to be Specification 4.5.H.1 inoperable for any shall be satisfied.

reason and the remaining diesel generator is operable, the requirements of Specification 3.5.H.1 shall be satisfied. '

2. Batteries 2. Batteries

a. From and after the Samples of the Battery date that Room atmosphere shall ventilation is lost be taken daily for in the Battery Room hydrogen concentration portable determination, ventilation equipment shall be provided.

b. From and after the date that one of the two 125 volt Station Battery Systems is made or found to be inoperable for any reasons, continued reactor operation is permissible only during the succeeding three days provided Specification 3.5.H is met unless such Battery System is sooner made operable.

Amendment No. 99, 14G, 114 215

VYNPS 3.10 LIMITING CONDITIONS FOR 4.10 SURVEILLANCE REQUIREMENTS i OPERATION

c. From and after the date that one of the two 24 volt ECCS Instrumentation Buttery Systems is made or found to be inoperable for any reason, continued reactor operation ,

is permissible only >

during the succeeding three days unless such Battery System is sooner made operable. 1

d. From and after the date tha*. the AS-2 125 Volt battery system is made or found to be inoperable for any reason, continued reactor operation is permissible i provided Diesel Generator DG-1-1A control power is transferred to i Station Battery B1 and a fire watch is ;

established to inspect the cable vault a minimum of every two hours.

e. From and after the date that one of the two 24 Volt Neutron Monitoring and Process Radiation Monitoring battery systems is found or made to be inoperable for any reason, continued reactor operation is permissible providing the minimum channel requirements of Sections 3.1 and 3.2 for the Neutron Monitoring and Process Radiation !

Monitoring systems are met.

Amendment No. GB, ++4, 125 216 l

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VYNPS l

3.10 LIMITING CONDITIONS FOR 4.10 SURVEILLANCE REQUIREME!TTS OPERATION ;

f. From and after the date that one of the two 125 volt Switchyard battery >

systems is found or made to be inoperable for any reason, continued reactor operation is permissible provided that the j other 125 volt Switchyard battery system is operable.

3. Off-Site Power

a. From and after the date that both a. When one of the startup diesel generators transformers and or associated buses one diesel is made or found to generator or be inoperable, the associated buses requirements of are made or found Specification to be inoperable 4.5.H.1 shall be for any reason, satisfied.

reactor operation may continue provided the requirements of Specification 3.5.H.1 are satisfied.

b. From and after the b. When both delayed date that both access off-site delayed access power sources are off-site power unavailable, both sources become diesel generators y unavailable, and associated reactor operation buses shall have may continue for been or shall be ,

seven days provided demonstrated to be both emergency operable within diesel generators, 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

associated buses, and all Low Pressure Core and Containment Cooling Systems are operable.

Amendment No. B6, ++4, 125 217

VYNPS 3.10 LIMITING CONDITIONS FOR 4.10 SURVEILLANCE REQUIREMENTS '

OPERATION

4. 480 V Uninterruptible 4. 480 V Uninterruptible Power Systems Power Systems From and after the date When it is determined that one that one Uninterruptible Power Uninterruptible Power System or its System or its associated Motor associated Motor Control Center are made Control Center is or found to be inoperable, the inoperable for any requirements of reason, the Specification 4.5.A.4 requirements of shall be satisfied.

Specification 3.5.A.4 shall be satisfied.

5. RPS Power Protection From and after the date that one of the two redundant RPS power protection panels on an in-service RPS MG set or alternate power supply is made or found to be inoperable, the associated RPS MG set or alternate supply will be taken out of service until the panel is restored to operable i status. -

C. Diesel Fuel C. Diesel Fuel There shall be a minimum of 1. The quantity of diesel 25,000 usable gallons of generator fuel shall be diesel fuel in the diesel logged weekly and after fuel oil storage tank. each operation of the unit.

2. Once a month a sample of diesel fuel shall be taken and checked for quality. The quality shall be within the applicable limits specified on Table I of ASTM D975-68 and logged.

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Amendment No. G6, 112 218 '

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VYNPS f

TABLE 4.10.1 [

REACTOR PROTECTION SYSTEM POWER PROTECTION l

Setpoints for !

Parameter Panels A1, A2, B1, B2, C1, C2 Overvoltage $ 125.5 volts .

Overvoltage Time Delay $ 0.35 seconds Undervoltage 2 111 volts ;

Undervoltage Time Delay $ 0.35 seconds Underfrequency 2 56.5 Hz Underfrequency Time Delay 5 0.35 seconds l

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Amendment No. 112 219

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

VYNPS BASES:

3.10 AUXILIARY ELECTRIC POWER SYSTEMS A. The objective of this Specification is to assure that adequate power will be available to operate the emergency safeguards equipment.

Adequate power can be provided by any one of the following sources:

either of the startup transformers, backfeed through the main transformer, the 4160 volt line from the Vernon Hydroelectric Station or either of the two diesel generators. The backfeed through the main transformer and 4160 volt Vernon line are both ,

delayed access off-site power sources. Backfeeding through the main transformer can be accomplished by disconnecting the main generator from the main transformer and energizing the auxiliary transformer from the 345 kV switchyard through the main transformer. The time required to perform this disconnection is approximately six hours.

The 4160 volt line from the Vernon Hydroelectric Station can be connected to either of the two emergency buses within seconds by simple manual switching operation in the Main Control Room.

Two 480 V Uninterruptible Power Systems supply power to the LPCIS valves via designated Motor Control Centers. The 480 V Uninterruptible Power Systems are redundant and independent of any on-site power sources.

This Specification assures that at least two off-site and two on-site power sources, and both 480 V Uninterruptible Power Systems will be available before the reactor is taken beyond "just critical

testing. In addition to assuring power source availability, all of the associated switchgear must be operable as specified to assure that the emergency cooling equipment can be operated, if required, from the power sources.

Station service power is supplied to the station through either the unit auxiliary transformer or the startup transformers. In order to start up the station, at least one startup transformer is required to supply the station auxiliary load. After the unit is synchronized to the system, the unit auxiliary transformer carries the station auxiliary load, except for the station cooling tower loads which are always supplied by one of the startup transformers.

The station cooling tower loads are not required to perform an engineered safety feature function in the event of an accident; therefore, an alternate source of power is not essential. Normally one startup transformer supplies 4160 volt Buses 1 and 3, and the other supplies Buses 2 and 4; however, the two startup transformers are designed with adequate capacity such that, should or.e become or be made inoperable, temporary connections can be made to supply the total station load (less the cooling towers) from the other startup trans f o rmer .

A battery charger is supplied for each battery. In addition, the two 125 volt station batteries have a spare charger available.

Since one spare 125 volt station charger is available, one station battery charger can be allowed out of service for maintenance and repairs.

Power for the Reactor Protection System is supplied by 120 V ac motor generators with an alternato supply from MCC-88. Two redundant, Class lE, seismically qualified pcwer protection panels are connected in series with each ac power source. These panels provide overvoltage, undervoltage, and underfrequency protection for the system. Setpoints are chosen to be consistent with the input !

power requirements of the equipment connected to the bus.

l Amendment No. G6, 58, 44G, 144, 125 220 l 1

l.

VYNPS l

BASES: 3.10 (Cont'd)

B. Adequate power is available to operate the emergency safeguards equipment from either startup transformer or for minimum engineered safety features from either of the emergency diesel generators. 1 Therefore, reactor operation is permitted for up to seven days with i both delayed-access off-site power sources lost.

Each of the diesel generator units is capable of supplying 100 percent of the minimum emergency loads required under postulated design basis accident conditions. Each unit is physically and i

electrically independent of the other and of any off-site power source. Therefore, one diesel generator can be allowed out of service for a period of seven days without jeopardizing the safety of the station.

In the event that both startup transformers are lost, adequate power l

is available to operate the eme.gency safeguards equipment from either of the emergency diesel generators or from either of the l

i delayed-access off-site power sources. Also, in the event that both emergency diesel generators are lost, adeyjate power is available immediately to operate the emergency safeguu*ds equipment from at least one of the startup transformers or from either of the delayed-access off-site power sources within six hours. The plant is designed to accept one hundred percent load rejection without adverse effects to the plant or the transmission system. Network stability analysis studies indicate that the loss of the Vermont f Yankee unit will not cause instability and consequent tripping of the connecting 345 kV and 115 kV lines. The Vernon feed is an independent source. Thus, the availability of the delayed-access off-site power sources is assured in the event of a turbine trip.

Therefore, reactor operation is permitted with the startup transformers out of service and with one diesel generator out of service provided the NRC is notified immediately of the event and I restoration plans.

Either of the two station batteries has enough capacity to energize l the vital buses and supply d-c power to the other emergency '

equipment for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months without being recharged. In addition, two i 24 volt ECCS Instrumentation batteries supply power to instruments I that provide automatic initiation of the EFCS and some reactor pressure and level indication in the Control Room.

Due to the high reliability of battery systems, one of the two batteries may be out of service for up to three dayo. This ;

minimizes the probability of unwarranted shutdown by providing adequate time for reasonable repairs. A station battery, ECCS Instrumentation battery, or an Uninterruptible Power System battery is considered inoperable if more than one cell is out of service. A i cell will be considered out of service if its float voltage is below l 2.13 volts and the specific gravity is below 1.190 at 77 F.

The Battery Room is ventilated to prevent accumulation of hydrogen gas. With a complete loss of the ventilation system, the accumulation of hydrogen would not exceed 4 percent concentration in 16 days. Therefore, on loss of Battery Room ventilstion, the use of portable ventilation equipment and daily sampling provide assurance that potentially hazardous quantities of hydrogen gao will not accumulate.

C. The minimum diesel fuel supply of 25,000 gallons will supply one diesel generator for a minimum of seven days of operation satisfying the load requirements for the operation of the safeguards equipment.

Additional fuel can be obtained and delivered to the site from nearby sources within the seven-day period.

Amendment No. E6, 56, 61 221

.~

VYNPS BASES:

4.10 AUXILIARY ELECTRICAL POWER SYSTEMS A. The monthly tests of the diesel generators are conducted to check for equipment failures and deterioration. The test of the undervoltage automatic otarting circuits will prove that each diesel will receive a start signal if a loss of voltage should occur on its emergency bus. The loading of each diesel generator is conducted to demonstrate proper operation at less than the continuous rating and at equilibrium operating conditions. Generator experience at other generator stations indicates that the testing frequency is adequate to assure a high reliability of operation should the system be required.

Both diesel generators have air compressors and air receivers tanks for starting. It is expected that the air compressors will run only infrequently. During the monthly check of the units, each receiver ,

will be drawn down below the point at which the compressor automatically starts to check operation and the ability of the compressors to recharge the receivers. i Following the tests of the units and at least weekly, the fuel volume remaining will be checked. At the end of the monthly load test of the diesel generators, the fuel oil transfer pump will be operated to refill the day tank. The day tank level indicator and alarm switches will be checked at this time. Puel oil transfer pump operability testing is in accordance with Specification 4.6.E.

The test of the diesels and Uninterruptible Power Systems during each refueling interval will be more comprehensive in that it will functionally test the system; i.e., it will check starting and closure of breakers and sequencing of loads. The units will be started by simulation of a loss of coolant accident. In addition, a loss of normal power condition will be imposed to simulate a loss of

off-site power. The timing sequence will be checked to assure proper loading in the time req"lred. Periodic tests between refueling intervals check the capability of the diesels to start in the required time and to deliver the expected emergency load requirements. Periodic testing of the various components plus a functional test at a refueling interval are sufficient to maintain adequate reliability.

B. Although the Main Station, ECCS, AS-2, and UPS batteries will '

deteriorate with time, utility experience indicates there is almost no possibility of precipitous failure. The type of surveillance described in this specification is that which has been demonstrated over the years to provide an indication of a cell becoming irregular ;

or unserviceable long before it becomes a failure.

The performance discharge test provides adequate indication and assurance that the batteries have the specified ampere hour capacity. The rate of discharge during this test shall be in accordance with the manufacturer's discharge characteristic curves for the associated batteries. The results of these tests will be logged and compared with the manufacturer's recommendations of ,

acceptability.

The service discharge rust provides a test of the batteries ability to satisfy the design requirements (battery duty cycle) of the ;

associated de system. This test will be performed using simulated or actual loads at the rates and for the durations specified in the design load profile.

i Amendment No. 26, ta5, 128 222

VYNPS I

BASES: 4.10 (Cont'd)

C. Logging the diesel fuel supply weekly and after each operation assures that the minimum fuel supply requirements will be !

maintained. During the monthly test for quality of the diesel fuel oil, a viscosity test and water and sediment test will be performed as described in ASTM D975-68. The quality of the diesel fuv1 oil will be acceptable if the results of the tests are within the 1 limiting requirements for diesel fuel oils shown on Table 1 of ASM D975-68.

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223 :

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-r -*----r- +- -- e-T-- - -.e-- .<-

VYNPS 3.11 LIMITING CONDITIONS FOR 4.11 SURVEILLANCE REQUIREMENTS OPERATION 3.11 REACTOR FUEL ASSEMBLIES 4.11 REACTOR FUEL ASSEMBLIES Applicability: Applicability:

The Limiting Conditions for The Surveillance Requirements operation associated with the apply to the parameters which fuel rods apply to these monitor the fuel rod operating parameters which monitor the conditions.

fuel rod operating conditions.

Obiective: Obiective:

The objective of the Limiting The Objective of the Conditions for Operation is to Surveillance Requirements is to assure the performance of the specify the type and frequency fuel rods, of surveillance to be applied to the fuel rods.

Specifications: Specifications:

A. Averace Planar Linear Heat A. Average Planar Linear Heat Generation Rate (APLHGR) Generation Rate (APLHGR)

During steady state power The APLHGR for each type of operation, the APLHGR for fuel as a function of each type of fuel as a average planar exposure function of average planar shall be determined daily exposure shall not exceed during reactor operation at the limiting values 225% rated thermal power.

provided in the Core Operating Limits Report.

For single recirculation loop operation, the limiting values shall be the values provided in the Core Operating Limits Report listed under the heading " Single Loop Operation.' If at any time during steady-state operation it is determined by normal surveillance that the limiting value for APLHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the APLHGR is not returned to within prescribed limits within two (2) hours, the reactor shall be brought to the shutdown conditions within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

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

Amendment No. 46, 47, 64, 94, 94, 444, 406, 116 224

l VYNPS 3.11 LI!4ITING CONDITIONS FOR 4.11 SURVEILLANCE REQUIRE!4ENTS OPERATION B. Linear Heat Generation Rate B. Linear Heat Generation Rate (LHGR) (LHOR)

During steady state power The LHGR as a function of operation, the linear heat core height shall be generation rate (LHOR) of checked daily during any rod in any fuel reactor operation at 125%

assembly at any axial rated thermal power.

location shall not exceed the maximum allowable LHGR provided in the Core Operating Limits Report.

If at any time during

{

steady state operation it is determined by normal I' surveillance that the limiting value for LHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the LHGR is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to shutdown condition within l 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . Surveillance and I corresponding action shall continue until reactor operation is within the prescribed limits.

I l

l Amendment No. 44, 47, 64, 106, 116 225

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - . _ _ _ _ _ _ _ _ _ -___-__________--_______________-_-._____-.__________-..______-___-_u

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

1. During steady state MCPR shall be determined power operation the daily during reactor power MCPR operating value operation at 2,25% rated shall be equal to or thermal power and following greater than the MCPR any change in power level limits provided in the or distribution that would Core Operating LimitO cause operation with a Report. For single limiting control rod recirculation loop pattern as described in the operation, the MCPR bases for Limits at rated flow Specification 3.3.B.6.

are also provided in the Core Operating Limits Report. For core flows other than rated, the Operating MCPR Limit shall be the above value multiplied by Kg where Kf is provided in the Core Operating Limits Report. If at any time during steady-state operation it is determined by normal surveillance that the limiting value for MCPR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the steady-state MCPR is not returned to within the prescribed limits within two (2) hours, the reactor power shall be brought to shutdown condition, I within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

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

l Amendment No. te, 44, &&, 94, 116 226

VYNPS BASES:

3.11 FURL RODS A. Averace Planar Linear Heat Generation Rule (APLHGR)

Refer to the appropriate topical reports listed in Specification 6.7.A.4 for analyses methods.

(Note: All exposure increments in this Technical Specification section are expressed in terms of megawatt-days per short ton.)

l The MAPLHGR reduction factor of 0.83 for single recirculation loop operation is based on the assumption that the coastdown flow from the unbroken recirculation loop would not be available during a postulated large break in the active recirculation loop, as discussed in NEDO-30060, " Vermont Yankee Nuclear Power Station Single Loop Operation." February 1983.

B. Linear Heat Generation Rate (LHGR)

Refer to the appropriate topical reports listed in Specification 6.7.A.4 for analyses methods.

C. Minimum Critical Power Ratio (MCPR)

Operatino Limit MCPR

1. The MCPR operating limit is a cycle-dependent parameter which can be determined for a number of different combinations of operating modes, initial conditions, and cycle exposures in order to provide reasonable assurance against exceeding the Fuel Cladding Integrity Safety Limit (FCISL) for potential abnormal occurrences. The MCPR operating limits are justified by the analyses, the results of which are presented in the current cycle's Core Performance Analysis Report. Refer to the appropriate topical reports listed in Specification 6.7.A.4 for analysis methods, The 0.01 increase in MCPR operating limits for single loop operation accounts for increased core flow measurement and TIP reading uncertainties, as discussed in NEDO-30060, " Vermont Yankee Nuclear Power Station Single Loop Operation, " February 1983.

Amendment No. 46, 47, 74, 63, 94, 95, 444, 116 227

l VYNPS ,

i BASES:

4.11 FUEL RODS A. The APLHGR, LHGR and MCPR shall be checked daily to determine if fuel burnup, or control rod movement has caused changes in power distribution. Since changes due to burnup are slow, and only a few control rods are removed daily, a daily check of power distribution is adequate. For a limiting value to occur below 25% of rated thermal power, an unreasonably large peaking factor would be required, which is not the case for operating control rod sequences.

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

C. Minimum Critical Power Ratio (MCPR) - Surveillance Requirement At core thermal power levels less than or equal to 25%, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience indicated that the resulting MCPR value is in excess of requirements by a considerable margin. With this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR. During initial start-up testing of the plant, a MCPR evaluation will be made at 25% thermal power level with minimum recirculation pump -

speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary. The daily requirement for calculating MCPR above 25%

rated thermal power is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The requirement for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in power or power shape (regardless of magnitude) that could place operation at a thermal limit.

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Amendment No. 18 228 l

VYNPS 3.12 LIMITING CONDITIONS FOR 4.12 SURVEILLANCE REQUIREMENTS ;

OPERATION 3.12 REFUELING AND SPENT FUEL 4.12 REFUELING AND SPENT FUEL HANDLING HANDLING Applicability: Applicability:

Applies to fuel handling, core Applies to the periodic testing reactivity limitations, and of those interlocks and spent fuel handling. instruments used during -

refueling and to the testing of the reactor building crane.

Obiective: Obiective ;

To assure core reactivity is To verify the operability of within capability of the instrumentation and interlocks control rods, to prevent used in refueling and the criticality during refueling, operability of the reactor ,

and to assure safe handling of building crane. '

spent fuel casks. !

Specification: Specification:

A. Refuelino Interlocks A. Refuelino Interlocks The reactor mode switch Prior to any fuel handling, shall be locked in the with the Head off the

" Refuel

position during reactor vessel, the core alterations and the refueling interlocks shall refueling interlocks, be functionally tested.

listed below, shall be They shall also be tested operable except as at weekly intervals specified in thereafter until no longer Specifications 3.12.D and required and following any !

3.12.E. repair work associated with

the interlocks.

1. Control Rod Blocks '

a. Mode switch in Startup/ Hot '

Standby and refueling platform over the reactor. ;

b. Puel on any i refueling hoist -

and refueling platform over the reactor. l

c. Mode switch in i Refuel with one i

control rod withdrawal permit.

2. Refuelino Platform ,

Reverse Motion (toward l reactor vessel) Block !

a. Mode switch in Startup/ Hot Standby.

Amendment No. 44, 29 229 l

1

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

i VYNPS 3.12 LIMITING CONDITIONS FOR 4.12 SURVEILLANCE REQUIREMENTS l OPERATION

b. Any control rod out and fuel on any refueling hoist. !

3. Refuelino Platform Hoists Blocks j

a. Any control rod out and fuel on any refueling hoist over the vessel.

b. Hoist overload.

c. High position limitation.

B. Core Monitorino B. Core Monitorina During core alterations two Prior to making any SRM's shall be operable, alterations to the core the one in the core quadrant SRM's shall be functionally where fuel or control rods tested and checked for s are being moved and one in neutron response. ,

Thereafter, the SRMs shall an adjacent quadrant. For an SRM to be considered be checked daily for operable the following response, conditions shall be satisfied:

1. The SRM shall be inserted to the normal operating level. (Use of special movable, dunking type detectors during initial fuel loading and major core alternations in place ;

of normal detector is permissible as long as the detectors is i connected into the proper circuitry which contain the required ,

rod blocks). !

2. The SRM shall have a minimum of 3 cps with 1 all rods fully inserted .

in the core. !

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Amendment No. 18 230 5

VYNPS 3.12 LIMITING CONDITIONS FOR 4.12 SURVEILLANCE REQUIREMENTS OPERATION

3. Prior to spiral Prior to spiral unloading unloading, the SRMs or reloading, the SRMs ,

shall be proven shall be functionally operable as stated in tested. Prior to spiral Sections 3.12.B.1 and reloading, the SRMs shall 3.12.B.2 above, be checked for neutron however, during spiral response.

unloading the count rate may drop below 3 cps. ,

4. Prior to spiral reloading, two ,

diagonally adjacent fuel assemblies, which have previously accumulated exposure in the reactor, shall be loaded into their designated core positions next to each of the 4 SRMs to obtain the required 3 cps.

Until these eight !

bundles have been loaded, the 3 cps requirement is not necessary.

C. Puel Storage Pool Water C. Puel Storace Pool Water Level Level Whenever irradiated fuel is Whenever irradiated fuel is stored in the fuel storage stored in the fuel storage pool the pool water level pool, the pool level shall shall be maintained at a be recorded daily. ,

level of at least 36 feet.

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1 Amendment No. 59, 77 231 l l

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VYllPS 3.12 LIMITItJG CONDITIOtJS FOR 4.12 SURVEILLANCE REQUIREMEtJTS l OPERATIOff D. Control Rod and Control Rod D. Control Rod and control Rod Drive Maintenance Drive Maintenance A maximum of two 1. Sufficient control rods non-adjacent control rods shall be withdrawn separated by more than two prior to performing control cells in any this maintenance to direction, may be withdrawn demonstrate with a from the core for the margin of 0.25 percent purpose of performing Ak that the core can be control rod and/or control made subcritical at any rod drive maintenance time during the provided the following maintenance with the conditions are satisfied strongest operable control rod fully

1. The reactor mode switch withdrawn and all other shall be locked in the operable rods fully

" Refuel

position. The inserted.

refueling interlock which prevents more 2. Alternately, if a than one control rod minimum of eight from being withdrawn control rods may be bypassed for one surrounding each of the control rods on control rod out of which maintenance is service for maintenance '

being performed. All are to be fully other refueling inserted and have their interlocks shall be directional control operable. valves electrically disarmed, the

2. Specification 3.3.A.1 0.25 percent Ak margin shall be met, or the shall be met with the control rod directional strongest control rod control valves for a remaining in service minimum of eight during the maintenance centrol rods period fully withdrawn.

surrounding each drive i out of service for maintenance shall be disarmed electrically and sufficient margin to criticality demonstrated.

3. SRMs shall be operable ;

(a) in each core quadrant containing a control rod on which maintenance is being performed, and (b) in a quadrant adjacent to one of the quadrants r specified in Specification 3.12.D.3.(a) above.

Requirements for an SRM to be considered r operable are given in ,

l Specification 3.12.B.

Amendment No. 18 232 P

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VYNPS 1

3.12 LIMITING CONDITIONS FOR 4.12 SURVEILLANCE REQUIREMENTS 1 OPERATION E. Extended Core Maintenance E. Extended Core Maintenance More than two control rods Prior to control rod ,

may be withdrawn from the withdrawal for extended reactor core provided the core maintenance, that following conditions are control rods control cell satisfied shall be verified to contain no fuel assemblies.

1. The reactor modo switch 1. This surveillance 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 being withdrawn may be bypassed on a withdrawn control rod after the fuel assemblies in the cell containing (controlled by) that control rod have been removed from the reactor core. All .

other refueling interlocks shall be operable.

2. SRMs shall be operable 2. This surveillance in the core quadrant requirement is the same where fuel or control as that given in rods are being moved, Specification 4.12.B.

and in an adjacent quadrant. The requirements for an SRM '

to be considered ,

operable are given in Specification 3.12.B.

3. If the spiral unload / reload method of core alteration is to be used, the following conditions shall be ;

met:

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.

l Amendment No. +4, 59, 77- 233 i

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VYNPS 3.12 LIMITING CONDITIONS FOR 4.12 SURVEILLANCE REQUIREMENTS OPERATION

b. The core may be spirally reloaded to either the original configuration or a different configuration in the reverse sequence of that used to unload, -

with the exception a that two (2) -

diagonally i adjacent fuel assemblies, which have previously accumulated exposure in the reactor, shall be loaded into their designated core positions next to each of the four (4) SRMs to obtain the required 3 cps. Until these eight (8) bundles have been loaded, the 3 cps requirement is not necessary.

Following insertion of the initial eight (8) bundles, the reactor will be spirally reloaded from the center cell outwards, until the core is fully loaded.

c. At least 50% of the fuel assemblies to be reloaded into the core shall have previously accumulated a minimum exposure of 1000 Mwd /T.

Amendment No. 59, 77 234

VYNPS 3.12 LIMITING CONDITIONS FOR 4.12 SURVEILLANCE REQUIREMENTS OPERATTON F. Fuel Movement F. Puel Movement Puel shall not be moved or Prior to any fuel handling handled in the reactor core or movement in the reactor for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following core, the licensed operator reactor shutdown to cold shall verify that the shutdown conditions. reactor has been in the .

cold shutdown condition for a minimum of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

G. Crane OperabiliLZ G. Crane Operability l l

1. The Reactor Building 1. a. Within one month crane shall be operable prior to spent when the crane is used fuel cask handling for handling of a spent operations, an fuel cask. inspection of crane cables, sheaves, hook, yoke, and cask lifting trunnions will be made.

These inspections shall meet the requirements of ANSI Standard B30.2, 1967. A crane rope shall be '

replaced if any of the replacement '

criteria given in ANSI B30.2.0-1967 are met.

b. No-load mechanical and electrical tests will be conducted prior to i lifting the empty ,

cask from its !

transport vehicle ;

to verify proper operation of crane controls, brakes and lifting speeds. A i functional test of the crane brakes -

will be conducted >

each time an empty 1 cask is lifted clear of its ;

transport vehicle. !

l Amendment No. 29 235 l

i VYNPS 3.12 LIMITING CONDITIONS FOR 4.12 SURVEILLANCE REQUIREMENTS l OPERATION )

2. Crane Travel 2. Crane Travel Spent fuel casks shall crane travel limiting be prohibited from mechanical stops shall travel over irradiated be installed on the fuel assemblies. crane trolley rails prior to cask handling operations to prohibit cask travel over irradiated fuel assemblies. ,

1 H. Spent Puol Pool Water H. Spent Puel Pool Water Temperature Temperature Whenever irradiated fuel is Whenever irradiated fuel is stored in the spent fuel in the spent fuel pool, the pool, the pool water pool water temperature temperature shall be shall be recorded daily.

maintained below 1500F. If the pool water temperature reaches 1500F, all refueling operations tending to raise the pool water temperature shall ,

cease and measures taken i' immediately to reduce the pool water temperature below 1500F.

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Amendment No. 37 236 l

I VYNPS l

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 I 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 i reactivity limitation of Specification 3.2 limits the core alterations to assure 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 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 tuel or control rods are being moved assures adequate monitoring of that quadrant during such alterations. The requirement of 3 counts per second provides assurance that neutron 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 cps. ;

Prior to spiral reload, two diagonally adjacent fuel assemblies, which have previously accumulated exposure in the reactor, will be loaded into their designated core positions next to each of the 4 SRMs to obtain the required 3 eps. Exposed fuel continuously produces neutrons by spontaneous fission of certain plutonium isotopes, photo fission, and photo disintegration of deuterium in t the moderator. This neutron production is normally great enough to i meet the 3 eps minimum SRM requirement, thereby providing a means by 4 which SRM response may be demonstrated before the spiral reload begins. 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. ,

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Amendment No. 44, 59, 77 237

. - _ _ - - ~~

VYNPS BASES: 3.12 & 4.12 (Cont'd)

C. To assure that there is adequate water to shield and cool the irradiated fuel assemblies stored in the pool, a minimam povi water level is established. This minimum water level of 36 feet is established because it would be a significant change from the normal ,

level, well above a level to assure adequate cooling (just above active fuel).

D. During certain periods, it is desirable to perform maintenance on two control rods and/or control rod drives at the same time. This specification provides assurance that inadvertent criticality does not occur during such maintenance.

The maintenance is performed with the mode switch in the " Refuel

position to provide the refueling interlocks normally available during refueling operations as explained in Part A of these Bases.

In order to withdraw a second control rod after withdrawal of the first rod, it is necessary to bypass the refueling interlock on the first control rod which prevents more than one control rod from being withdrawn at the same time. The requirement that an adequate shutdown margin be demonstrated with the control rods remaining in service ensures that inadvertent criticality cannot occur during this maintenance. The shutdown margin is verified by demonstrating that the core is shut down even if the strongest control rod remaining in service is fully withdrawn. Disarming the directional control valves does not inhibit control rod scram capability.

E. The intent of this specification is to permit the unloading of a significant portion of the reactor core for such purposes as '

inservice inspection requirements, examination of the core support plate, etc. This specification provides assurance that inadvertent criticality does not occur during such operation.

This operation is performed with the mode switch in the " Refuel' position to provide the refueling interlocks normally available during refueling as explained in the Bases for Specification 3.12.A.

In order to witndraw more than one control rod, it is necessary to bypass the refueling interlock on each withdrawn control rod which prevents more than one control rod from being withdrawn at a time.

The requirement that the fuel assemblies in the cell controlled by the control rod be removed from the reactor core before the interlock can be bypassed ensures that withdrawal of another control rod does not result in inadvertent criticality. Each control rod essentially provides reactivity control for the fuel assemblies in the cell associated with that control rod. Thus, removal of an entire cell (fuel assemblies plus control rod) results in a lower '

reactivity potential of the core.

one method available for unloading or reloading the core is the spiral unload / reload. A spiral unloading pattern is one by which the fuel in the outermost cells (four fuel bundles surrounding a control rod) is removed first. Unloading continues by unloading the remaining outermost fuel by cell spiralling inward towards the center cell which is the last cell removed. Spiral reloading is reverse of unloading, with the exception that two (2) diagonally adjacent bundles, which have previously accumulated exposure in-core, are placed next to each of the 4 SRMs before the actual spiral reloading begins. The spiral reload then begins in the center cell and spirals outward until the core is fully loaded.

i Amendment No. 77 238 l

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VYNPS BASES: 3.12 & 4.12 (Cont'd) l Additionally, at least 50% of the fuel assemblies to be reloaded into the core shall have previously accumulated a minimum exposure of 1000 Mwd /T to ensure the presence of a minimum neutron flux as described in Bases Section 3.12.B.

F. The intent of this specification is to assure that the reactor core has been in the cold shutdown condition for at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following power operation and prior to fuel handling or movement.

The safety analysis for the postulated refueling accident assumed that the reactor had been shut down for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for fission product decay prior to any fuel handling which could result in dropping of a fuel assembly.

G. The operability requirements of the reactor building crane ensures that the redundant features of the crane have been adequately inspected just prior to using it for handling of a spent fuel cask.

The redundant hoist system ensures that a load will not be dropped for any postulated credible single component failures. Details of the design of the redundant features of the crane and specific testing requirements for the crane are delineated in the Vermont Yankee document entitled ' Reactor Building Crane Modification" (December 1975).

H. The Spent Puel Pool Cooling System is designed to maintain the pool water temperature below 1250F during normal refueling operations.

If the reactor core is completely discharged, the temperature of the pool water may increase to greater than 1250P. The RHR System supplemental fuel pool cooling may be used under these conditions to maintain the pool water temperature less than 1500F.

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Amendment No. 29, &7, 69, 77 239

VYNPS 3.13 LIMITING CONDITIONS FOR 4.13 SURVEILLANCE REQUIREMENTS OPERATION 3.13 FIPE PROTECTION SYSTEM 4.14 FIRE PROTECTION SYSTEM Applicability: Applicability:

Applies to the operational Applies to the surveillance status of the fire protection requirements of the fire systems. protection systems.

Obiectives Obiective:

To assure adequate capability To verify the operability of to detect and suppress a fire the fire protection systems.

which could affect the safe shutdown of the reactor.

Specification: Specification:

A. Fire Detection A. Fire Detection

1. Except as specified in 1. Each of the sensors Specification 3.13.A.2 specified in 3.13.A.1 below, the minimum and their associated 4 number of fire instruments including detection sensors and the supervisory their associated circuitry shall be instrument for each demonstrated operable location shall be at least once per operable in accordance 6 months.

with Table 3.13.A.1, whenever the equipment ,

it protects is required to be operable.

2. From and after the date that less than the minimum number of sensors or their associated instruments are found to be operable, a fire watch >

shall be established to inspect the location with the inoperable sensor or instruments at least once every hour. Restore the required number of sensors and instruments to operable status within 14 days or submit a report within the next 30 days to the Commissicn as specified in 6.7.C.2 outlining the cause of malfunction and the plans for restoring the instrument (s) to operable status.

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l Amendment No. 43 240 I l

l VYNPS 3.13 LIMITING CONDITIONS FOR 4.13 SURVEILLANCE REQUIREMENTS I OPEMATION B. Vital Fire Suppression B. Vital Fire Suppression l Water System Water System 1

1. Except as specified in 1. The Vital Fire .

Specification 3.13.B.2 Suppression Water l and 3.13.B.3 below, the System shall be j Vital Fire Suppression demonstrated operable:

Water System shall be operable with: a. At least once per month by starting

a. Two fire pumps each pump and operable and lined operating it for up to the fire 15 minutes, suppression loop. 1

b. At least once each 1

b. Water available month by verifying from the each valve in the Connecticut River. flow path is in its correct position.

c. An operable flow (For electrically path capable of supervised valves, taking suction from adequate the Connecticut verification is a River and visual check of transferring the electrical water through the indication.)

distribution piping with operable c. At least once each sectionalizing year by performance control or of a system flush isolation valves to of the yard fire !

the yard hydrant loop.

curb valves and the j hose station d. At least once per i isolation valves. 12 months by 1 cycling each l

2. From and after the date testable valve in !

that less than the the flow path ;

above required through at least !

equipment is operable, one complete cycle restore the component of full travel.

to operable status within 7 days or submit e. At least once each a report within the 18 months: j next 30 days to the i Commission as specified 1) By performing l in 6.7.C.2 outlining a system the plans and functional .i procedures to be used test by to provide for the loss simulating of redundancy in this sequential system. automatic start of the

3. With the fire suppression fire pumps as water supply system applicable to inoperable; maintain the Vital Fire

a. Establish a backup Suppression fire suppression Water System water system within pressure of at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; and least 125 psig.

Amendment No. 43 241

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l VYNPS 3.13 LIMITING CONDITIONS FOR 4.13 SURVEILLANCE REQUIREMENTS OPERATION

b. Submit a Special 2) By verifying Report as specified that each pump in 6.7.C.2; will develop a flow of at

1) By telephone least 2500 gpm within at a discharge 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , pressure of at least 115 psig

2) Confirmed by corrected for telegraph, river water mailgram, or level.

facsimile transmission 3) Cycling each no later than valve in the the first flow path that working day is not following the testable event, and during plant operation

3) In writing through at within 14 days least one following the complete cycle event, of full outlining the travel, action taken, the cause of f. At least once per the 3 years by inoperability performing a flow and the plans test in accordance and schedule with Chapter 5, for restoring Section II, of the the system to Fire Protection operable Handbook, status, or 14th Edition, published by the

c. If a. above cannot National Fire be fulfilled, place Protection the reactor in hot Association, standby within the next six (6) hours 2. The fire pump diesel ,

and in cold engine shall be l shutdown with the demonstrated OPERABLE:

following thirty (30) hours. a. At least once per month by verifying;

1) The fuel storage tank contains at least 150 gallons of fuel, and

2) The diesel starts from ambient conditions and operates for at least 20 minutes.

Amendment No. 43 242 a

VYNPS

'3.13 LIMITING CONDITIONS FOR 4.13 SURVEILLANCE REQUIR9!ENTS OPERATION

b. At least once per quarter by verifying that a sample of diesel fuel from the fuel storage tank is within the acceptable limits specified in Table 1 of ASTM D975-68 with respect to viscosity, water content, and sediment.

c. At least once per 18 months by verifying the diesel starts from ambient conditions on the auto-start signal and operates for > 20 minutes while loaded with the fire pumn.

3. The fire pump uiesei starting 24-volt battery bank and charger shall b<,

demonstrated OPERABLE:

a. At least once per week by verifying that:

1) The electrolyte level of each battery is above the plates, and

2) The overall battery voltage is 1 24 volts.

b. At least once per quarter by verifying that the specific gravity is appropriate for continued service of the battery.

c. At least once per 18 months by verifying that:

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( Amendment No. 43 243

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 damage or abnormal deterioration, and

2) The battery-to-battery and terminal connections are clean, tight, free of corrosion and coated with anti-corrosion material.

C. Fire Hose Stations C. Fire Hose Stations

1. Except as rpecified in 1. Each fire hose station 3.12.C.2 below, all shall be verified to be hose stations inside operable the Reactor Building, Turbine Building, and a. At least monthly by those inside the visual inspection Administration Building of the station to which provided coverage assure all of the Control Room equipment is Building shall be available.

operable whenever equipment in the areas b. At 1. east once each protected by the fire lo months by hose stations is removing the hose required to be for incpection and operable. replacing degraded coupling gaskets

2. With one or more of the and reracking.

fire hose stations specified in 3.13.C.1 c. At least once each above inoperable, route year by an additional hydro-statically equivalent capacity testing each fire hose to the outside hose at unprotected area (s) 250 lbs.

from an operable hose station within one d. At least once per hour. 3 years by hydro-statically testing inside hose at 150 lbs.

Amendment No. 44, 67 244

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l VYNPS 3.13 LIMITING CONDITIONS FOR 4.13 SURVEILLANCE REQUIREMENTS OPERATION

e. At least once per 3 years, partially open hose station valves to verify valve operability and no blockage.

D. Hiqh Pressure C0 7 System D. Hiqh Pressure CO 3 System

1. Except as specified in 1. The CO2 system located Specification 3.13.D.2, in the cable vault, the CO2 systems located switchgear room, and in the cable vault, diesel fire pump day switchgear room, and tank room shall be diesel fire pump day demonstrated operable.

tank room shall be operable, whenever a. At least once per ,

equipment in the area six months by protected by the system verifying each CO2 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, c. At least once per within one hour a operating cycle a continuous fire watch flow path test shall be established shall be performed with backup fire to verify flow suppression equipment, through each Restore the CO2 system nozzle.

to operable 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.

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Amendment No. 43, 64, 117 245 I

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VYNPS 3.13 LIMITING CONDITIONS FOR 4.13 SURVEILLANCE REQUIREMENTS OPERATION l 3. From and after the date that the CO2 system in the diesel tire pump day tank room is inoperable, within one hour a fire watch shall be established to inspect the location at least once every hour.

Restore the system to operable 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 system to operable status E. Vital Fire Barrier E. Vital Fire Barrier Penetration Fire Seals Penetration Fire Seals

1. Except as specified in 1. Vital fire barrier Specification 3.13.E.2 penetration seals shall below, vital fire be verified to be barrier penetration functional by visual seals protecting the inspection at least Reactor Building, once per operating Control Room Building, cycle and following any and Diesel Generator repair.

Rooms shall be intact.

2. From and after the date a vital fire barrier penetration fire seal is not intact, a continuous fire watch shall be established on at least one side of the affected penetration within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

F. Sprinkler Systems F. Sprinkler Systems

1. Except as specified in 1. Each of the sprinkler Specification 3.13.F.2 systems specified in below, those sprinkler Table 3.13.F.1 shall be systems listed in demonstrated operable:

Table 3.13.F.1 shall be operable whenever a. At least once per equipment in the area 12 months by protected by those cycling each sprinklers is required testable valve in to be operable. the flow path through at least one complete cycle of full travel.

Anendment No. 44, 67 246 t

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VYNPS 3.13 LIMITING CONDITIONS FOR 4.13 SURVEILLANCE REQUIREMENTS 1 OPERATION l

2. From and after the date b. At least once each that one of the month by verifying sprinkler systems each valve in the l specified in flow path is in its 1 Table 3.13.F.1 is correct position. ,

inoperable, a fire (For electrically '

watch shall be supervised valves, established within one adequate i hour to inspect the verification is a location with the visual check of inoperable sprinkler electrical system at least once indication.)

every hour. Restore the sprinkler system to c. At least once per operable status within 18 months by; 14 days or submit a report within the next 1. Cycling each

30 days to the valve in the Commission as specified flow path that L in 6.7.C.2 outlining is not the cause of the testable malfunction and the during plant r plans for restoring the operation system to operable through at status. least one complate cycle .

of full travel. ,

2. A visual inspection of the sprinkler. L headers to verify their integrity, and

3. A visual inspection of each nozzle's spray area to verify that the spray pattern is not obstructed.

4. Verifying that automatic valves actuate r to their Correct '

position from a test signal.

d. At least once per .

3 years by performing a flow test through each open head sprinkler header and verifying each open head sprinkler nozzle is unobstructed.

Amendment No. 67 247 i

l VYNPS l

3.13 LIMITING CONDITIONS FOR 4.13 SURVEILLANCE REQUIREMENTS I OPERATION G. Foam Systems G. Foam Systems

1. Except as specified in 1. The foam system Specification 3.13.G.2 specified in 3.13.G below, the shall be demonstrated Recirculation M.G. Set operable.

Foam System shall be operable with its foam a. At least once per l concentrate tank full 12 months by (100 gallons) whenever cycling each the Recirculation M.G. testable valve in Sets are operating. the flow path l through at least ;

2. From and after the date one complete cycle that the Recirculation of full travel.

M.G. Set Foam System is inoperable, a fire b. At least once per watch shall be 18 months by:

established to inspect i the location at least 1. Cycling each once every hour; and a valve in the ,

foam nozzle shall be flow path that brought to the Reactor is not Building elevation testable containing the during plant Recirculation M.G. operation Sets. A 100 gallon through at foam concentrate supply least one shall be available on complete cycle site, of full travel. '

3. Except as specified in Specification 3.13.G.4 2. A visual below, the Turbine inspection of Building Foam System the foam shall be operable with system and its foam concentrate equipment to tank full verify (150 gallons). integrity, and ,

4. From and after the date 3. A visual that the Turbine inspection of Building Foam System is the inoperable a portable Recirculation foam nozzle shall be M.G. Set Foam brought to the Turbine System foam -

Building Foara System nozzle area to ,

location. A 150 gallon verify that foam concentre.te supply the spray shall be avr.11able pattern is not on-site. obstructed. ,

4. Foam concentrate samples shall be taken and analyzed for acceptability.

Amendment No. 64, 90 248

VYNPS 3.13 LIMITING CONDITIONS FOR 4.13 SURVEILLANCE REQUIRD4ENTS OPERATION

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d. At least once per 3 years by performing an air flow test through the Recirculation M.G. Set foam !

header and verifying each foam -

nozzle is '

unobstructed.

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Amendment No. 67 249

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l VYNPS TABLE 3.13.A.1 FIRE DETECTION SENSORS Minimum No. of Sensors Roguired to Be Operable Sensor Location Heat Flame Smoke

1. Cable Spreading Room & Station Battery Room - -

23

2. Switchgear Room - -

20

3. Diesel Generator Room (A) - -

2

4. Diesel Generator Room (B) - -

2

5. Intake Structure (Service Water) 1 1 1 l

6. Recirc Motor Generator Set Area 3 -

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

14 h 7.b Control Room Zone 2 (Control Room Panels) - -

18 !

7.c Control Room Zone 3 (Control Room Panels) - -

25 7.d Control Room Zone 4 (Control Room Panels) - -

10 7.e Control Room Zone 5 (Exhaust & Supply - -

2 i Ducts) 8.a Rx Bldg. Corner Rm NW 232 - -

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8.b Rx Bldg. Corner Rm NW 213 (RCIC) - -

1 8.c Rx Bldg. Corner Rm NE 232 - -

1 8.d Rx Bldg. Corner Rm NE 213 - -

1 8.e Rx Bldg. Corner Rm SE 232 - -

1 8.f Rx Bldg. Corner Rm SE 213 - -

1 8.g Rx Bldg. Corner Rm SW 232 - -

1

9. HPCI Room - -

8

10. Torus area 12 -

16

11. Rx Bldg. Cable Penetration Area - -

7

12. Refuel Floor - -

13

13. Diesel Oil Day Tank Room (A) -

1* 1*

14. Diesel Oil Day Tank Room (B) -

1* 1*

15. Turbine Loading Bay (vehicles) -

3 -

NOTE: The Diesel Day Tank Rooms require only one detector operable (1 flame oyr 1 smoke).

Amendment No. 44, 67 250

f VYNPS TABLE 3.13.F.1 SPRINKLER SYSTE'MS

1. Reactor Building Penetration Area Preaction System

2. Diesel Generator Room A System i

3. Diesel Generator Room B System (

4. Turbine Loading Bay System

5. Diesel-driven Fire Pump System I

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Amendment N' 07 251

l VYNPS BASES: )

3.13 & 4.13 FIRE PROTECTION SYSTEMS On May 11, 1976, Vermont Yankee received a letter from the NRC requasting that an in-depth evaluation of the existing fire protection systems be i.erformed using Branch Technical Position (BTP) APCSB 9.5-1 as a guide. Concurrent with this evaluation a fire hazards analysis of the entire plant complex was required. In an effort to clarify the BTP an Appendix A was subsequently issued to specifically address operating plants. Enclosed with this Appendix the NRC requested that proposed Technical Specifications on fire protection also be submitted. The subject section 3.13/4.13 and the following specific bases are those specifications evolving from these efforts.

A. The smoke, heat and flame detectors provide the early warning fire detection capability necessary to detect problems in vital areas of the plant. Surveillance requirements assure these sensors and their associated instruments to be operable. When the equipment protected by the detectors is not required to be operable, specifications cevering the sensors and instruments do not apply.

l B,C, The Vital Fire Suppression Water System, CO2 systems, sprinkler D, F, systems and foam systems specifications are provided to meet and pre-established levels of system operability in the event of a G fire. These systems provide the necessary protection to assure safe reactor shutdown. Periodic surveillance testing provides assurance that vital fire suppression systems are operable.

E. Vital fire barrier penetration fire seals are provided to assure that the fire resistance rating of barriers is not reduced by a penetration. Surveillsnce inspections shall be performed to insure that the integri*y of these seals is maintained.

The diesel fire pump has a design consumption rate of 18 gallons of fuel per hour; therefore, 150 gallons provides for greater than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of operation. Additional fuel can be delivered in about ,

one hour and additional fuel is on site. When the equipment protected by the fire protection systems is not required to be operable, the specifications governing the fire protection system do not apply.

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Amendment No. 44, 67 252 ,

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1 VYNPS 5.0 DESIGN FEATURES 5.1 Site The station is located on the property on the west bank of the l Connecticut River in the Town of Vernon, Vermont, which the :

Vermont Yankee Nuclear Power Corporation either owns or to which it has perpetual rights and easements. The site plan showing the exclusion area boundary, boundary for gaseous ef fluents and boundary for liquid effluents is on Figure 2.2-5 in the FSAR. The minimum distance to the boundary of the exclusion area as defined in 10CFR100.3 is 910 feet.

No part of the site shall be sold or leased and no structure shall be located on the site except structures owned by the Vermont Yankee Nuclear Power Corporation or related utility companies and used in conjunction with normal utility operations.

5.2 Reactor ;

A. The core shall consist of not more than 368 fuel assemblies.

B. The reactor core shall contain 89 cruciform-shaped control rods. The control material shall be boron carbide powder (B4 C) or hafnium, or a combination of the two.

5.3 Reactor Vessel The reactor vessel shall be as described in Table 4.2-3 of the ,

FSAR. The applicable design codes shall be as described in subsection 4.2 of the FSAR.

5.4 containment A. The principal design parameters and applicable design codes for the primary containment shall be as given in Table 5.2.1 of the FSAR.

B. The secondary containment shall be as described in subsection 5.3 of the FSAR and the applicable codes shall be as described in Section 12.0 of the FSAR.

C. Penetrations to the primary containment and piping passing through such penetrations shall be designed in accordance with standards set forth in subsection 5.2 of the FSAR.

5.5 Spent and New Puel Storace A. The new fuel storage facility shall be such that the effective is less multiplication than 0.90 and factor when(K,gg)less floodedthan is 0.95.of the fuel when dry ,

B. The K,fg of the fuel in the spent fuel storage pool shall be ;

less than or equal to 0.95.

C. Spent fuel storage racks may be moved (only) in accordance with written procedures which ensure that no rack modules are moved over fuel assemblies.

Amendment No. 37, 84, 123 253 r

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VYNPS D. The number of spent fuel assemblies stored in the spent fuel l pool shall not exceed 2870.

E. The maximum core geometry infinite lattice multiplication factor of any segment of the fuel assembly stored in the spent fuel storage pool or the new fuel storage facility i shall be less than or equal to 1.31 at 20oc.

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l i Amendment No. M, Me, 130 254 l

VYNPS 6.0 ADMINISTRATIVE CONTROLS 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 quality of operation and maintenance of a nuclear power reactor. These controls shall be adhered to.

6.1 ORGANIZATION A. Lines of authority, responsibility, and communication shall be established and defined for the highest management levels through intermediate levels to and including all operating 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 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 i 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 l

b. Superintendent (s) I

c. Chemistry Supervisor !

d. Radiation Protection Supervisor l

e. Operations Supervisor (See Item D.7, Page 190a) I

f. Reactor and Computer Supervisor j

g. Maintenance Supervisor

h. Instrument and Control Supervisor

i. Shift Supervisors Amendment No. 64, 76, 64, 90, 144, 121 255

VYNPS

5. The Radiation Protection Supervisor or Plant Health Physicist shall meet or exceed the qualifications of Regulatory Guide 1.8, Revision 1 (September 1975).

6. The Shift Engineer shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents.

7. If the Operations Supervisor does not possess a Senior Operator License, then an Assistant Operations Supervisor shall be designated that does possess a Senior Operator License. All instructions to the shift crews involving licensed activities shall then be approved by designated Assistant Operations Supervisor.

8. The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate on-site manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.

E. A Fire Brigade of at least 5 members shall be maintained on-site at all times.# This excludes 2 members of the minimum shift crew necessary for safe shutdown of the plant l and any personnel required for other essential functions during a fire emergency.

Fire Brigade composition may be less than the minimum requirements for a i period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected '

absence of Fire Brigade members provided immediate action is taken to i restore the Fire Brigade to within the minimum requirements.

Amendment No. 44, 64, 75, 79, 67, 444, 141, 131 256

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

VYNPS TABLE 6.1.1 Vermont Yankee staff positions that shall be filled by personnel holding Senior Operator and Operator licenses are indicated in the following tables Title Meense Requirements Operations Supervisor Licensed Senior Operator (Except as specified in Specification 6.1.D.4)

Shift Supervisor Licensed Senior Operator Supervisory Control Room Operator Licensed Senior Operator Control Room Operator Licensed Operator CONDITIONS Cold Shutdown plant Startup or Refueling and Normal With Fuel in the MINIMUM SHIFT STAFFING ON-SITE Operation (Note 1) Reactor (Note 2)

Shift Supervisor 1 1 Supervisory Control Room Operator 1 -

Control Room Operator 2 1 Auxiliary Operator 2 1 Shift Engineer 1 -

NOTES:

(1) At least one Senior Licensed Operator and one Licensed Operator, E two Senior Licensed Operators, shall be in the Control Room.

(2) At least one Licensed Operator, E one Senior Licensed Operator, shall be in the Control Room.

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Amendment No. G , 79, 67, 90 257

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VYNPS 6.2 REVIEW AND AUDIT organizational units for the review and audit of plant operations shall be constituted and have the responsibilities and authorities I outlined below: 1 A. Plant Operations Review Committee

1. Membership

a. Chairman: Plant Manager

b. Vice-Chairman: Superintendent (s)

c. Engineering Support Supervisor

d. Operations Supervisor

e. Maintenance Supervisor

f. Reactor and Computer Supervisor

g. Chemistry Supervisor

h. Instrument and Control Supervisor

i. Radiation Protection Supervisor

2. Oualifications 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 or higher than as described in Specification 6.1.

3. Meetino Frequency: Monthly, and as required, on call of the Chairman.

4. Ouorum: 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 by the Plant Manager; however, there shall be no more than three (3) alternates serving on the committee at any one time.

6. Responsibilities ,

a. Review proposed normal, abnormal, and emergency 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. l I

I

c. Review proposed changes to Technical ;

Specifications. 1

d. Review proposed changes or modifications to plant systems or equipment, which changes would require a ,

change in procedures in (a) above. l

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

Amendment No. 66, 45, 14+, 121 258 i

VYNPS

f. Investigate reported instances of violations of Technical Specifications, such investigations to include reporting, evaluation, and recommendations to prevent recurrence, to the Manager of Operations.

g. Perform special reviews and investigations and render reports thereon as requested by the Chairman of the Nuclear Safety Audit and Review Committee.

7. Authority

a. The Plant Operation Review Committee shall be adviso ry .

b. The Plant Operation Review Committee shall recommend to the Plant Manager approval or .

disapproval of proposals under Items 6 (a) through (d) above.

1. In the event of disagreement between the recommendations of the Plant Operation Review Committee and the actions contemplated by the '

Plant Manager, the course determined by the Plant Manager to be the more conservative will be followed with immediate notification to the Manager of Operations,

c. The Plant Operation Review Committee shall make tentative determinations as to whether or not >

proposals considered by the Committee involve t unreviewed safety questions. This determination shall be subject to review by the Nuclear Safety Audit and Review Committee.

8. Records Minutes shall be kept at the plant of all meetings of the Plant Operation Review Committee and copies shall be sent to the Manager of Operations and the Nuclear Safety Audit and Review Committee.

B. Nuclear Safety Audit and Review Committee

1. The Committee shall consist of at least six (6) persons:

a. Chairman

b. Vice Chairman

c. Four technically qualified persons who are not members of the plant staff.

d. No more than three members shall be selected from the organization reporting to the Manager of Operations,

e. The Committee will obtain advice and counsel from '

scientific or technical personnel employed by the Company or other organizations whenever the ,

committee considers it necessary to obtain further i scientific or technical assistance in carrying out its responsibilities.

Amendment No. 66, 121 259

I VYNPS

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

2. Oualifications The Committee shall consist of a minimum of six (6) members of designated alternates who as a group employ expertise in the following areas:

a. Nuclear Power Plant Technology

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 Frequenev: Semi-annually and as required on '

call of the Chairman.

4. Ouorums Chairman or Vice Chairman plus four members or designated alternates.

5. Responsibilities:

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 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. 44, 64, 74, 64, 121 260

VYNPS

v. the Process Control Program and implementing procedures for processing and packaging of radioactive waste at least once per 24 months; vi. the performance of activities required by the Quality Assurance Program to meet the provisions of Regulatory Guide 1.21, Revision 1, June 1974, and Regulatory Guide 4.1, Revision 1, April 1975, at least once per 12 months.

e. Investigate all reported instances of violations of Technical Specifications, reporting findings and recommendations to prevent recurrence to the Manager of Operations.

f. Perform special reviews and investigations and render reports thereon as requested by the Manager of Operations.

g. Review proposed tests and experiments and results thereof when applicable.

h. Paview abnormal performance of plant equipment and anomalies. ,

1

1. Review unusual occurrences and incidents which are reportable under the provisions of 10 CFR Part 20 and 10 CFR Part 50.

j. Review of occurrences if safety limits are exceeded.

6. Authority

a. Review proposed changes to the operating license including Technical Specifications and revised bases for submittal to the NRC. ,

b. Review proposed changes or modifications to plant systems or equipment, provided that such changes or modifications do not involve unreviewed safety questions. !

c. Recommend to the Manager of Operations appropriate action to prevent recurrence of any violations of <

Technical Specifications.

d. Evaluate actions taken by the Plant Operation Review Committee.

7. Records Minutes of all meetings of this committee shall be recorded. Copies of the minutes shall be forwarded to the Manager of Operations, the Vice President -

Operations, the Plant Manager and any others that the Chairman may designate.

6.3 ACTION TO BE TAKEN IN THE EVENT OF A REPORTABLE OCCURRENCE IN PLANT OPERATION Applies to administrative action to be followed in the event of a reportaole occurrence in plant operation.

Amendment No. 64, 83 261

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VYNPS Any reportable occurrence shall be reported to the Manager of Operations and shall be reviewed by the Plant Operations Review Committee. This Committee shall prepare a separate, sequentially numbered, report for each reportable occurrence. Each report shall describe the circumstances leading up to and resulting from the occurrence, the corrective action taken by the shift, an attempt to define the cause of the occurrence, and shall recommend appropriate action to prevent or reduce the probability of a repetition of the occurrence.

Copies of all such reports shall be submitted to the Chairman of the Nuclear Safety Audit and Review Committee for review and to the Manager of Operations for review and approval of any recommendations.

6.4 ACTION TO BE TAKEN IF A SAFETY LIMIT IS EXCEEDED Applies to administrative action to be followed in the event a safety limit is exceeded.

If a safety limit is exceeded, the reactor shall be shutdown immediately. An inmediate report shall be mado to the Manager of Operations. A complete analysis of the circ- 6tances leading up to and resulting from the situation together .tth recommendations by the Plant Operations Review Committee shall also be prepared.

This report shall be submitted to the Manager of Operations and the Chairman of the Nuclear Safety Audit and Review Committee.

Reactor operation shall not be resumed until authorized by the U.S. Nuclear Regulatory Commission.

l 6.5 PLANT OPERATING PROCEDURES A. Detailed written procedures, involving both nuclear and non-nuclear safety, including applicable check-off lists and instructions, covering areas listed below shall be prepared and approved.

All procedures shall be adhered to.

1. Normal startup, operation and shutdown of systems and components of the facility.

2. Refueling operations.

3. Actions to be taken to correct specific and foreseen potential malfunctions of systems or components, suspected Primary System leaks and abnormal reactivity changes.

4. Emergency conditions involving potential or actual release of radioactivity.

5. Preventive and corrective maintenance operations which could have an effect on the safety of the reactor.

6. Surveillance and testing requirements.

1

7. Fire protection program implementation including minimum fire brigade requirements and training. The training program shall meet or exceed the requirements of Section 27 of the NFPA Code 1976. Training sessions will be scheduled as plant operations permit but will be completed in specified subjects annually. Initial fire brigade training shall be completed by March 13, 1978.

Amendment No. 44, 45, 83 262 I

1 VYNPS )

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) 8. Process Control Program in-plant implementation.

9. Off-Site Dose Calculation Manual in-plant implementation.

B. Radiation control standards and procedures shall be prepared, approved and maintained and made available to all station personnel. These procedures shall show permissible radiation exposure, and shall be consistent with the requirements of 10 CFR Part 20. This radiation protection program shall be organized to meet the requirements of 10 CFR Part 20.

1. Paragraph 20.203, " Caution signs, labels, signals and controls". In lieu of the ' control device" or

alarm signal" required by paragraph 20.203 (c) (2), each high radiation area in which the intensity of radiation is 1000 mrem /hr or less shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit.* Any individual or group of individuals permitted to enter such areas shall be provided with one or more of the following:

a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.

b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received.

Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them.

c. A Health Physics qualified individual (i.e.,

qualified in radiation protection procedures) with a radiation dose rate monitoring device who is responsible for providing positive control over the activities within the area and who will perform periodic radiation surveillance at the frequency i specified in the RWP. The surveillance frequency will be established by the Plant Health Physicist. ,

l The above procedure shall also apply to each high radiation area in which the intensity of radiation is greater than 1000 mrem /hr. In addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the l administrative control of the Shift Supervisor on duty l and/or the Plant Health Physicist.

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Health Physics personnel shall be exempt from the RWP issuance requirement during the performance of their assigned radiation protection duties, i providing they are following plant radiation protection procedures for entry I into high radiation areas.

Amendment No. 36, 49, 83 263

r VYNPS C. Procedures prepared for A and B above shall be reviewed and approved by the applicable Department Supervisor and the Plant Manager, or his designee (superintendent level).

D. Temporary changes to procedures described in Specification 6.5.A above which do not change the intent of the original procedure, may be made with the concurrence of two individuals holding senior operator licenses. Such changes shall be documented and subsequently reviewed by the PORC and approved by the Plant Manager or his designee.

E. Temporary changes to procedures described in Specification 6.5.B may be made with the concurrence of an individual holding a senior operator license and the health physicist on duty.

F. Licensed radioactive sealed sources shall be leak tested for contamination. Tests for leakage and/or contamination shall be performed by the licensee or by other persons specifically authorized by the Commission or an agreement state as follows:

1. Each licensed sealed source, except startup sources previously subjected to core flux, containing radioactive materials, other than Hydrogen 3, with half-life greater than thirty days and in any form, other than gas, shall be tested for leakage and/or contamination at intervals not to exceed six months.

2. The periodic leak test required does not apply to sealed sources that are stored and are not being used. The sources exempted from this test shall be tested for leakage prior to any use or transfer to another user unless they have been leak tested within six months prior to the date of use or transfer. In the absence of a certificate from a transferrer indicating that a leak test has been made within six months prior to the transfer, sealed sources shall not be put into use until tested.

3. Each sealed startup source shall be tested within 31 days prior to being subjected to core flux and following repair or maintenance to the source.

The leakage test shall be capable of detecting the presence of 0.005 microcurie of radioactive material on the test sample. If the test reveals the presence of 0.005 microcurie or more of removable contamination, it shall immediately be withdrawn from use, decontaminatad, and repaired, or be disposed of in accordance with Commission regulations.

Notwithstanding the periodic leak tests required by this Technical Specification, any licensed sealed source is exempt from such leak test when the source contains 100 microcuries or less of beta and/or gamma emitting material or 5 microcuries or less of alpha emitting material.

A special report shall be prepared and submitted to the Commission within 90 days if source leakage tests reveal the l

presence of >0.005 microcuries of removable contamination.

Amendment No. 44, 66, 76, 60, 84, 404, 121 264 i

I' VYNPS i

6.6 PLANT OPERATING RECORDS A. Records and/or logs relative to the following items shall be kept in a manner convenient for review and'shall be retained for at least five years:

1. Records of normal plant operation, including power levels and periods of operation at each power level.

2. Records of principal maintenance activities, including inspection and repair or principal items of equipment pertaining to nuclear safety.

3. Records of reportable occurrences.

4. Records of periodic checks, inspection and/or calibrations performed to verify that surveillance requirements are being met.

5. Records of any special reactor test or experiments.

6. Records of changes made in the Operating Procedures.

7. Records of radioactive shipments.

8. Test results, in units of microcuries, for leak tests performed on licensed sealed sources.

9. Results of annual physical inventory verifying accountability of licensed sources on record.

B. Records and/or logs relative to the following items shall be ,

recorded in a manner convenient for review and shall be i retained for the life of the plant:

1. Records of substitution or replacement of principal items of equipment pertaining to nuclear safety.

2. Records of changes and drawing changes made to the plant as it is described in the safety Analysis Report.

3. Records of plant radiation and contamination surveys.

4. Records of new and spent fuel inventory, transfers of fuel, and assembly histories.

5. Records of radioactivity in liquid and gaseous wastes released to the environment.

6. Records of radiation exposure for all plant personnel, including all contractors and visitors to the plant in accordance with 10 CFR 20.

7. Records of transient or operational cycling for those '

plant components that have been designed to operate safely for a limited number of transients or operational cycles.

8. Records of inservice inspections of the reactor coolant system.

9. Minutes of meetings of the Plant Operation Review Committee and the Nuclear Safety Audit and Review Board.

Amendment No. 31 265

l VYNPS l

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 REOUIREMENTS In addition to the applicable reporting requirements of Title 10 Code of Federal Regulations, the following identified reports shall be submitted to the Director of the appropriate Regional Office of Inspection and Enforcement unless otherwise noted.

A. Routine Reports

1. Startup Report 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 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 predictions and specifications. Any corrective actions that were required to obtain satisfactory operation 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 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 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.

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

Amendment No. 40, 83 266

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VYNPS The dose assigriment to various duty functions may be i estimates based on pocket dosimeter, TLD or film badge l measurement. Small exposures totaling less than 20% of j the individual total dose need not be accounted for. In i the aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to ,

specific major work functions. !

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3. Monthly Statistical Report Routine reports of operating statistics and shutdown j experience shall be submitted on a monthly basis to the Office of Management Information and program Control, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the appropriate Regional Office, to arrive no later than the fifteenth of each month following the calendar month covered by the report. These reports shall include a narrative summary of operating experience during the report period which describes the operation of the facility.

4. Core Operatino Limits Report The core operating limits shall be established and documented in the Core Operating Limits Report (COLR) before each reload cycle or any remaining part of a reload cycle for the followings (a) The Average Planar Linear Heat Generation Rates (APLHGR) for Specifications 3.11.A and 3.6.G.la, (b) The Kg core flow adjustment factor for Specification 3.11.C., (c) The Minimum Critical Power Ratio (MCPR) for Specifications 3.11.C and 3.6.G.la, and (d) The Linear Heat Generation Rates (LHGR) for Specifications 2.1.A.la, 2.1.B.1, and 3.11.B. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in:

Report, E. E. Pilat, " Methods for the Analysis of Boiling Water Reactors Lattice Physics," YAEC-1232, December 1980 (Approved by NRC SER, dated September 15, 1982).

Report, D. M. VerPlanck, ' Methods for the Analysis of Boiling Water Reactors Steady State Core Physics,"

YAEC-1238, March 1981 (Approved by NRC, SER, dated j September 15, 1982). J Report, J. M. Holzer, " Methods for the Analysis cf Boiling Water Reactors Transient Core Physics," 3 YAEC-1239P, August 1981 (Approved by NRC SER, dated ]

September 15, 1982).

Report, S. P. Schultz and K. E. St. John, " Methods for the Analysis of Guide Puel Rod Steady-State Thermal Effects (FROSSTEY): Code /Model Description Manual,'

YAEC-1249P, April 1981 (Approved by NRC SER, dated i September 27, 1985). l 1

1 l

i Amendment No. 49, 43, 84, 95, 116 267

r VYNPS Report, A. A. F. Ansari, " Methods for the Analysis of Boiling Water Reactors: Steady-State Core Flow Distribution Code (FIBWR)," YAEC-1234, December 1980 (Approved by URC SER, dated September 15, 1982).

Report, S. P. Schultz and K. E. St. John, " Methods for the Analysis of Oxide Fuel Rod Steady-State Thermal Effects (FROSSTEY): Code Qualification and Application," YAEC-1265P, June 1981 (Approved by NRC SER, dated September 27, 1985).

Report, A. A. F. Ansari and J. T. Cronin, " Methods for the Analysis of Boiling Water Reactors: A System Transient Analysis Model (RETRAN)," YAEC-1233, April 1981. (Approved by NRC SERs, dated November 27, 1981 and September 4, 1984).

Report, A. A. F. Ansari, K. J. Burns and D. K. Beller,

" Methods for the Analysis of Boiling Water Reactors:

Transient Critical Power Ratio Analysis (RETRAN-TCPYA01)," YAEC-1299P, March 1982 (Approved by NRC SER, dated September 15, 1982).

Report, A. S. Diciovine, et al., 'CASMO-3G Validation,"

YAEC-1363-A, April 1988.

Report, A. S. DiGiovine, J. P. Gorski, and M. A. Tremblay, ' SIMULATE-3 Validation and Verification," YAEC-1659-A, September 1988.

Report, R. A. Woehlke, et al.,

'MICBURN-3/CASMO-3/ TABLES-3/ SIMULATE-3 Benchmarking of Vermont Yankee Cycles 9 through 13," YAEC-1683-A, March 1989.

Report, J. T. Cronin, " Method for Generation of One-Dimensional Kinetics Data for RETRAN-02,*

YAEC-1694-A, June 1989.

Report, V. Chandola, M. P. LeFrancois, and J. D. Robichaud, ' Application of One-Dimensional Kinetics to Boiling Water Reactor Transient Analysis Methods,' YAEC-1693-A, Revision 1, November 1989.

Report, L. H. Steves, et. al, 'HUXY: A Generalized Multirod Heatup Code with 10CFR50, Appendix K Heatup Option: User's Manual,' XN-CC-33(A), Revision 1, dated November 14, 1975 (Approved by NRC SER, dated March 6, 1975).

Report, 'RELAPSYA, A Computer Program for Light-Water Reactor System Thermal-Hydraulic Analysis," YAEC-1300P, '

October 1982 (Approved by NRC SERs, dated August 25, ;

1987 and October 21, 1992). '

Report, R. T. Fernandez and H. C. daSilva, Jr . , " Vermont Yankee BWR Loss-of-Coolant Accident Licensing Analysis ,

Method," YAEC-1547, June 1986 (Approved by NRC SER, j dated October 21, 1992). !

Letter from R. W. Capstick (VYNPC) to USNRC, "HUXY Computer Code Information for the Vermont Yankee BWR LOCA Licensing Analysis Method,' FVY 87-63, dated ,

June 4, 1987 (Approved by NRC SER, dated '

February 27, 1991).

1 Amendment No. 144, 4G4, 135 268 1

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VYNPS Letter from R. W. Capstick (VYNPC) to USNRC, ' Request for Supplemental Safety Evaluation Report Supporting the Use of RELAP5YA for Vermont Yankee Nuclear Power Station," FVY 88-006, dated January 26, 1988 (Approved by NRC SERs, dated February 27, 1991 and October 21, 1992).