ML20206S066

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Proposed TS to Change Number 189,proposing Relocation of Fire Protection Requirements from TSs to TRM
ML20206S066
Person / Time
Site: Vermont Yankee Entergy icon.png
Issue date: 01/22/1999
From:
VERMONT YANKEE NUCLEAR POWER CORP.
To:
Shared Package
ML20206S046 List:
References
NUDOCS 9901280057
Download: ML20206S066 (27)
v • d • e


Text

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c. A verification or coding system for emergency messages between Vermont Yankee and the state police headquarters of the respective states and the Commeonwealth.

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14. Vermont Yankee shall furnish advance notification to MDPH, l or to another Commonwealth agency designated by MDPH, of the time, method and proposed route through the Commeonwealth of any shipments of nuclear fuel and wastes to and from the Vermont Yankee facility which will utilise railways or roadways in the Commonwealth.

l W W -

. The lic see proce with d is r quired o comp 1 e the

' modif ations dentif ed in P ragrap 3.1.1 rough 3 .20 of the l

A-43 NRC' Fire P otecti Safet Evalus ion (SE on the acilit dated J ry 13 1978. ese difica ons a 1 be c lated s 1* *78 specifie in Tab 3.1 o the SE In ad tion, e lic ee shal

,dubmit e addi onal i orisati n iden fled in able 3 of thi SE in ace rdance ith the schedu conta ed ther in. I the ey thes dates e subs tal e not be t, the iconse shall s it a i rep rt, exp sining e cir stanc , toget er with/'sj revise a edule.

/ Security Plan 3.G.f The licensee shall fully implement and maintain in effect all

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I / provisions of the Commsission-approved physical security, guard A-107 / training and qualification, and safeguards contingency plans 8.25.88 / including amendments made pursuant to provisions of the 10.20.88 #

[MiscellaneousAmendmentsandSearchRequirementsrev

/ 10CFR73.55 (51FR27817 and 27822) and to the authority of 10CFR50.90 The plans, which contain Safeguards Information jand 10CFR50.54(p).

/

l protected under 10CFR73.21, are entitled: "Verimont Yankee Nuclear Power Station Physical Security Plan," with revisions submitted

/ through March 16, 1988; " Vermont Yankee Nuclear Power Station l Training and Qualification Plan," with revisions submitted through

[ November 10, 1982; and " Vermont Yankee Nuclear Power Station

/ Safeguards Contingency Plan," with revisions submitted through j

p(December 30, 1985. Changes made in accordance with 10CFR73.55 shall p / be implemented in accordance with the schedule set forth therein.

/

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'Y{' '~' **E* '5 9901290057 990122 PDR ADOCK 05000271 ,

PDR 3 P

VYMPS 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.

Norral 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. Ins t rument Punctional Test - An instrument functional test shall be:

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 verify the operability including alarm and/or trip functions.

H. Loo system Punctional Test - A logic system functional test means a test of all relays and contacts of a logic circuit from sensor to activated device to insure all components are operable per design P{ intent. Where possible, action will go to completion, i.e., pumps will be started and valves opened.

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k% I. Minimum Critical Power Ratio - The minimum critical power ratio is\

defined as the ratio of that power in a fuel essembly which is J

b calculated to cause some point in that assembly to experience boiling transition as calculated by application of the appropriate N i NRC-approved critical power correlation to the actual assembly

% operating power.

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

N Amendment No. M, M, M, H4, -H9- 1

VYNPS l

1.0 DEFINITIONS s - ~__

$s I. Minimum Critical Power Ratio - The minimum critical power ratio is q{ defined as the ratio of that power in a fuel assembly which is j calculated to cause some point in that assembly to experience boiling b 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 I

) mode-selector-switch.

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N( K. Operable - A system, subsystem, train, component or device shall be operable or have operability when it is capable of performing its l

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

1 M. O_perating Cycle - Interval between the end of one refueling outage and the end of the next subsequent refueling outage, i

N. Primary Containment Integrity - Primary containment integrity means that the drywell and pressure suppression chamber are intact and all of l

the following conditions are satisfied: j

1. All manual containment isolation valves on lines connecting to the reactor coolant system or containment, which are not required to be open during accident conditions, are closed. Such valves may be opened intermittently under administrative controls.
2. At least one door in each airlock is closed and sealed. I
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 Amendment No. 64, 44, B4, 44& 2

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! l VYNPS j 1

1.0 DEFINITIONS l

or more plant parameters in order to initiate trip system action. l Initiation of protective action may require the tripping of a l

l N single trip system or the coincident tripping of two trip systems.

3.

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

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4. Protective Function - A system protective action which results from  !

the protective action of the channels monitoring a particular plant condition.

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

Rated Thermal Power - Rated thermal power means a steady state power l

Q.  !

level of 1593 thermal megawatts.

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 i 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 l system is energized with IRM neutron monitoring system trips and  ;

control rod withdrawal interlocks in service and APRM neutron I monitoring system operable.

l 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.

l T. Refueling 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 i of testing and surveillance, a refueling outage shall mean a regularly

scheduled refueling outage; however, where such outages occur within l 8 months of the completion of the previous refueling outage, the i 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, j Amendment No. M, M, M6 3 l

VYNPS 1.0 DEFINITIONS The standby gas treatment system is perable.

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

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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 protection system trip systems are de-energized.

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 212*F.
3. Shutdown means conditions as above such that the effective I multiplication factor ( Kerr) 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.

X. Transition'EoIling - Transition boiling means the boiling regime I 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.  !

l Y. Surveillance Frequency - Unless otherwise stated in these l

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. Surveillance 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, 1 component, or system prior to being required to be operable.

AA. Deleted l

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 Amendment No, M , M , 44, M , M, MG, 424, M6 4

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i VYMPS l 4 1.0- DEFINITIONS i

factors used for this. calculation shall be those listed in NRC i

. Regulatory Guide 1.109, Revision 1, October 1977. I DD.. Solidification - Solidification shall be the conversion of wet wastes i into a form that meets shipping and burial ground requirements. l

. Suitable forms' include dewatered resins and filter sludges.  !

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

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'GG. Deleted HH. Deleted l

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 the calculation of gaseous and liquid effluent monitoring alarm / trip ,

.setpoints, and in the conduction of tha environmental radiological monitoring program.

JJ. Process Control Program (PCP) - A process control program shall contain l the sampling, analysis, tests, and determinations by which wet i 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 l installed to reduce radioactive gaseous effluents by collecting primary l coolant system off-gases from the primary system and providing for j delay or holdup for the purpose of reducing the total radioactivity

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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 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 / Purging - 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.

NN. 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.

Amendment No.' 44, 444r M1- 5

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l VYNPS ,

3.13 LIMITING CONDITIONS FOR 4.13 SURVEILLANC1'; REQUIREMENTS OPERATION e

1) The batteries l cell plates nd ,

battery ra s '

show no v ual indicati n of i physic damage {

or ab rmal dete oration,  !

and

2) e battery-to-attery and terminal ';

connections are clean, tight, i free of corrosion and i coated with i anti-corrosion material. l r

i t

C Fire Hose Stations C Fire Hese* Stations- i

1. Except as specified in 1. Each fire hose station  !

3.13.C.2 below, all hose shall be verified to be ,

stations inside the operable: '

Reactor Building, Turbine Building, and a. At least monthly by those inside the . visual inspection of j Administration Building the station to i which provided coverage assure all equipment  ;

of the Control Room is available. i Building shall be  :

operable whenever b. At least once each equipment in the a as 18 months by _;

protected by the ire removing the hose j hose stations i for inspection and required to be perable, replacing degraded oupling gaskets and >

2. With one or ore of the r acking. l fire hose ations  !

specified n 3.13.C.1 c. At ast once each above in erable, route year ]

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-an add ional equivalent hydro- tatically capac y fire hose to testing ach outside the protected area (s) hose at 0 lbs.

fro an operable hose s tion within one hour. d. At least on per 3 years by

- hydro-statica y L testing inside ose

at 150 lbs.

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Amendment No. 44, 64, 164 244

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l VYNPS 3.13 LIMITING CONDITIONS FOR 4.13 SURVEILLANCE REQUIREMENTS l OPERATION l

e. At least once per 3 years, partial

! open hose stati valves to veri i valve operabi ty I and no block ge.

l l D. CO, Systems D. CO, Systems l

[ 1. ept as specified in 1. The CO2 syst ms located l S ification 3.13.D.2, in the cab vault, l

l the 02 systems located east and st

( in th cable vault, switchge rooms, and i east west diesel ire pump day i

switchg r rooms, and tank r om shall be i diesel fi e pump day demo trated operable.

tank room hall be operable, w never a. At least once per l equipment in he area six months by l protected by e system verifying each CO 2 l

is required to e cylinder associated j operable. with the cable vault and diesel

2. From and after the ate fire pump day tank l that the CO2 system room CO 2 systems the cable vault or a does not contain

! switchgear room is less than 90% of l inoperable, within one its initial charge.

hour a fire watch shall

be established to b. At least once per inspect the location at 18 months by I least or .
e every hour, verifying that the provided that the fire system, including detection system is associated l

operable in accordance ventilation j with 3.13.A. If the dampers, will l fire detection syste actuate I

is also inoperable, automatically to a

, within one hour a simulated actuation l continuous fire w< ch signal.

shall be establi ed with backup fir At least once per suppression e ipmer.t . operating cycle a Restore the C 2 system low path test to operable tatus all be performed within 14 ys or to verify flow submit a port within thr ugh each l the next 0 days to the nozz .

Commiss'en as specified in 6.7 C.2 outlining d. At leas once per the e use of 7 days b verifying i ino rability and the the CO2s rage pl s for restoring the tank assoc 1 ted C 2 system to operable with the swi chgear atus. rooms does no contain less t n 50% level and a

, minimum pressure f 270 psig. \

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Amendment No. 4h f4, H4, 154 245

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VYNPS .

3.13 LIMITING CONDITIONS FOR 4.13 SURVEILLANCE REQUIREMENTS OPERATION j G. D oam Systems G. Foam Systems

1. Except as specified in 1. The foam system '

Specification 3.13.G.2 specified in 3.13.

low, the shall be demonst ted i circulation M.G. Set operable.

Fo System shall be ope ble with its foam a. At least nce per conc trate tank full 12 mont by I i

j (150 lions) whenever cyclin each the Re 'rculation M.G. testa e valve in Sets are operating. the ow path thr ugh at least 4

2. From and a ter the date o complete cycle I that the Re rculation full travel. l M.G. Set Fo System is J inoperable, a ire b. At least once per watch shall be 18 months by:

established to i spect the location at 1 ast 1. Cycling each once every hour; a d a valve in the foam nozzle shall b flow path that brought to the React is not Building elevation testable containing the during plant Recirculation M.G. operation

' l Sets. A 150 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 w th system and its foam concentr e equipment to tank full verify (150 gallons). integrity, and

4. From and aft the date 3. A visual that the Tu ine inspection of Building F am System is the inoperabl a portable Recirculation foam no le shall be M.G. Set Foam brough to the Turbine System foam
g Foam System nozzle area to  ;

! Build) locar ion. A 150 gallon erify that  !

fo concentrate supply t e spray s 11 be available pa tern is not 4

-site. obs eted.  !

4. Foam i concen ate

! samples hall be taken nd

, analyzed f r acceptabili y.

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j Amendment No. M , 90, 156 248 l

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'UYNPS 3.13 LIMITING CONDITIONS FOR 4.13 SURVEILLANCE REQUIREMENTS OPERATION

c. At least once p 3 years by

. performing an air flow test t ough.

the Recire ation j i-M.G. Set oam header  ;

and ver ying each foam n zie is unobs ucted.

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Amendment No.' 44, 164 249 l- ,

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VYNPS TABLE 3.13.A.1 FIRE DETECTION SENSORS Minimum No. of Sensors Required to Be Operahie Sensor Location Heat Flame oke

1. Cable reading Room & Station Battery Room - -

23

2. Switchge Room (East) - -

10

3. Switchgear oom (West) - -

10

4. Diesel Gener or Room (A) - -

2

5. Diesel Generato Room (B) - -

2

6. Intake Structure c ervice Water) 1 1 1
7. Recirc Motor Generas r Set Area -

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

14 8.b Control Room Zone 2 (Con rol Room Panels) - -

18 8.c Control Room Zone 3 (Contr 1 Room Panels) - -

25 8.d Control Room Zone 4 (Control oom Panel - -

10 8.e Control Room Zone 5 (Exhaust & upplv - -

2 Ducts) 9.a Rx Bldg. Corner Rm NW 232 - -

1 9.b Rx Bldg. Corner Rm NW 213 (RCIC - -

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

1 9.d Rx Bldg. Corner Rm NE 213 - -

1 9.e Rx Bldg. Corner Rm SE 23 - -

1 9.f Rx Bldg. Corner Rm SE 13 - -

1 9.g Rx Bldg. Corner Rm 232 - -

1

10. HPCI Room - -

8

11. Torus area 12 -

16

12. Rx Bldg. Cab Penetration Area - -

7

13. Refuel Flo . -

13

14. Diesel O Day Tank Room (A) -

1* 1*

15. Diese Oil Day Tank Room (B) -

1* 1*

16. Tur ne Loading Bay (vehicles) -

3 -

  • NOT : The Diesel Day Tank Rooms require only one detector operable (1 f ame or 1 smoke).

. Amendment No. 44, 67, 164 250

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VYNPS BASES:

3.1 & 4.13 FIRE PROTECTION SYSTEMS May 11, 1976, Vermont Yankee received a letter from the NRC  !

r esting that an in-depth evaluation of the existing fire protectio sys ms be performed using Branch Technical Position (BTP) APCSB 9.5 1 as a ide. Concurrent with this evaluation a fire hazards analys of  :

the en re plant complex was required. In an effort to clarify t e BTP an Appen x A was subsequently issued to specifically address o rating plants. Closed with this Appendix the NRC requested that pr osed I Technical S cifications on fire protection also be submitte . The subject sectz n 3.13/4.13 and the following specific bases re those specifications volving from these efforts.

A. The smoke, h t and flame detectors provide the e ly warning fire detection capa 'lity necessary to detect proble in vital areas of the plant. Su illance requirements assure t se sensors and their associated struments to be operable. en the equipment protected by the d ectors is not required be operable, specifications cover .g the sensors and i ,truments do not apply.

B,C, The Vital Fire Suppress n Water Syste CO2 systems, sprinkler D,F, systems and foam systems ecificatio are provided to meet and pre-established levels of stem op ability in the event of a G fire. These systems provide *he n essary protection to assure safe reactor shutdown. Perio 'c urveillance testing provides assurance that vital fire suppr sion systems are operable.

The east and west switchgear ooms ow pressure CO2 storage tank Technical Specification mi mum leve of 50% provides for sufficient CO2 quantity t achieve an maintain design concentration, in accor nee with NFPA (1993), in the east or west switchgear rooms. The Technical Sp ification minimum tank pressure of 270 psig ill provide the mini pressure to meet system design.

E. Vital fire barri penetration fire seals are p ovided to assure that the fire r sistance rating of barriers is n reduced by a penetration. urveillance inspections shall be pe ormed to insure that the in grity of these seals is maintained.

The dies fire pump has a design consumption rate of 1 gallons of fuel pe hour; therefore, 150 gallons provides for great than 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of operation. Additional fuel can be delivered in about one ur and additional fuel is on site. When the equipmen pr ected by the fire protection systems is not required to b o rable, the specifications governing the fire protection sys m o not apply.

Amendment'No. 44, 64, 154 252

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4 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 I
a. The Plant Operation Review Committee shall be I advisory. l

. The Plant Operation Review Committee shall recommend to the Plant Manager approval or '

jf4 [g* disapproval of proposals under Items 6 (a) through l1. b eV/tkl (d) above.

I frofecfloH Me ran4 ## 1 In the event of disagreement betweec the recommendations of the Plant Operation Review j

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jpgf gg,nf,'ff frecca'ureg Committee and the actions contemplated by the Plant Manager, the course determined by the an d ab itfo/ f rrcoruHbf /t/ Plant Manager to be the more conservative will l cgan y db //74 OlMC/f d r~ be followed with immediate notification to the Manager of Operations.

$ q kudifend The Plant Operation Review Committee shall nake g *'"Ig##' tentative determinations as to whether or not proposals considered by the committee involve 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 scientific or technical assistance in carrying out its responsibilities.

l Amendment No. 64, 121 259

Docket No. 50-271 ,

BVY 99-04 ATTACHMENT 2 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No.189 Supplemental Change: Submittal of Revised Pages Retyped License and Technical Specification Pages I

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c. A verification or coding system for emergency messages between Vermont Yankee and the state police headquarters of the respective states and the Commonwealth.

t

14. Vermont Yankee shall furnish advance notification to MDPH, or to another Commonwealth agency designated by MDPH, of the time, method and proposed route through the l Commonwealth of any shipments of nuclear fuel and wastes to and from the Vermont Yankee facility which will utilize railways or roadways in the Commonwealth.

F. Vermont Yankee shall implement and maintain in effect all i provisions of the approved Fire Protection Program as described in the Final Safety Analysis Report for the facility and as approved in the SER dated January 13, 1978, and supplemental SERs, dated 2/20/80, 10/24/80, 1/13/83, 3/25/86, 12/8/89, 6/9/97, 8/12/97, 9/2/98, and , subject to the following provisions:

Vermont Yankee may make changes to the approved Fire Protection Program without prior approval of the Commission only if those ,

changes would not adversely affect the ability to achieve and >

maintain safe shutdown in the event of a fire. l 3.G Security Plan ,

The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans A-107 including amendments made pursuant to provisions of the b 2b 8 Miscellaneous Amendments and Search Requirements revisions to l 10CFR73.55 (51FR27817 and 27822) and to the authority of 10CFR50.90 and 10CFR50.54 (p) . The plans, which contain J Safeguards Information protected under 10CFR73.21, are entitled:

" Vermont Yankee Nuclear Power Station Physical Security Plan," l with revisions submitted through March 16, 1988; " Vermont Yankee '

Nuclear Power Station Training and Qualification Plan," with revisions submitted through November 10, 1982; and " Vermont Yankee Nuclear Power Station Safeguards Contingency Plan," with revisions submitted through December 30, 1985. Changes made in accordance with 10CFR73.55 shall be implemented in accordance with the schedule set forth therein.

3.H This paragraph deleted.

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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. '

I 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 I 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. l l

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.

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i G. Instrument-Functional Test - An instrument functional test shall be:

l. 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 verify the operability including alarm and/or trip functions.  ;

H. Log System Functional Test - A logic system functional test means a test of all relays and contacts of a logic circuit from sensor to activated device to insure all components are operable per design intent. Where possible, action will go to completion, i.e., pumps will be started and valves opened.

Amendment No. 44, 83, G&, 446, 444 1

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VYNPS 1.0 DEFINITIONS I. Minimum Critical Power Ratio - The minimum critical power ratio is defined as the ratio of that power in a fuel assembly which is 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. j K. Operable - A system, subsystem, train, component or device shall be l operable or have operability when it is capable of performing its specified function (s). Implicit in th4s definition shall be the assumption that all necessary attenc ant instrumentation, controls, normal and emergency electrical powe- sources, cooling or seal water,  ;

lubrication or other auxiliary equipment that are required for the j system, subsystem, train, component or device to perform its l function (s) are also capable of performing their related support I function (s).

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

M. Operating Cycle - Interval between the end of one refueling outage and the end of the next subsequent refueling outage.

N. Primary Containment Integrity - 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 I reactor coolant system or containment, which are not required to be l open during accident conditions, are closed. Such valves may be opened intermittently under administrative controls.
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 i to a trip system a single trip signal related to the plant

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parameter monitored by that instrument channel. '

l 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 Amendment No. 64, 74, 83, 44& 2

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l 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. l
4. Protective Function - 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.

i Q. Rated Thermal Power - Rated thermal power means a steady state power I level of 1593 thermal megawatts. i R. Reactor Power Operation - Reactor power operation is any operation with the mode switch in the "Startup/ Hot Standby" or "Run" position with the i reactor critical and above 1% rated thermal power. j

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. Refueling 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.

Amendment No. -70, 64, -le 3

l VYNPS 1.0 DEFINITIONS

2. The standby gas treatment system is operable.
3. All reactor building automatic ventilation system isolation valves l are operable or are secured in the isolated position.

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 perforned. 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 protection system trip systems are de-energized.

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

I 2. Cold Shutdown means conditions as above with reactor coolant temperature equal to or less than 212*F.

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

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

X. Transition Boilin2 - Transition boiling means the boiling regime  :

between nucleate and film boiling. Transition boiling is the regime in l 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 I 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 he tolerance stated j above is applicable. j Z. Surveillance Interval - The surveillance interval is the calendar time l between surveillance tests, checks, calibrations, and examinations to I

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, i component, or system prior to being required to be operable.

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

l CC. Dose Equivalent I-131 - The dose equivalent I-131 shall be that i concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, 1-132, I-133, I-134 and I-135 actually present. The thyroid dose conversion Amendment No. M, M, 44, M, M, M3, M4, M5 4 l

l VYNPS 1.0 DEFINITIONS 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. Deleted FF. Site Boundary - The site boundary is shown in Figure 2.2-5 in the FSAR. i i

GG. Deleted HH. Deleted 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 the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints, and in the conduction of the environmental radiological monitoring program.

JJ. Pr,ocess Control Program (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 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 / Purging - 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.

NN. Core Operating Limits Report - The Core Operating Limits Report is the l 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 l Specification 6.7.A.4. Plant oporation within these operating limits is addressed in individual speci fications.

Amendment No. 43, 44474&l- 5

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Amendment No. 4, M4 - 249

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Amendment No. 4-3, 4-7,454 252  !

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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. '

h. Review of the Fire Protection Program and implementing procedures, and submittal of recommended changes to the Nuclear Safety Audit and Review i Committee.
7. Authority
a. The Plant Operation Review Committee shall be I advisory. l l
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 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 plaat 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 scientific or technical assistance in carrying out its responsibilities.

Amendment No. 66, 4G4, 259

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