ML20113A434
| ML20113A434 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee  |
| Issue date: | 03/27/1985 |
| From: | VERMONT YANKEE NUCLEAR POWER CORP. |
| To: | |
| Shared Package | |
| ML20113A424 | List: |
| References | |
| NUDOCS 8504100361 | |
| Download: ML20113A434 (25) | |
Text
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VYNPS ,
1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING 1.1 FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING TNTEGRITY Applicability Applicability Applies to the interrelated variable associated Applies to trip setting of the instruments and with fuel thermal behavior. devices which are provided to prevent the nuclear system safety limits from being exceeded.
Objective Objective To establish limits below which the integrity of To define the level of the process variable at the fuel cladding is preserved. which automatic nrotective action is initiated.
Specification Specification A. Bundle Safety Limit (Reactor Pressure >800 A. Trip Settings psia and Core Flow >10% of Rated)
The limiting safety system trip settings When the reactor pressure is >800 psia and shall be as specified below:
core flow is >10% of rated, the existence of a minim m critical power ratio (MCpR) less 1. Neutron Flux Trip Settings than 1.07 shall constitute violation of the fuel cladding integrity safety limit. a. APRM Flux Scram Trip Setting (Run Mode)
When the mode switch is in the RUN position, the ApRM flux scram trip setting shall be as shown on Figure 2.1.1 and shall be:
S 1066W + 54%
Amendment No. 64, 5 e
8504100361 850327 PDR ADOCK 05000271 p PDR
YYNPS
- 1. SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING where:
S= setting in percent of rated thermal power (1593 MWt).
W= percent rated drive flow where 100% rated drive flow is that flow equivalent to 48 x 10 6 lbs/hr core flow.
i In the event of operation with the ratio of NFLPD to FRP greater then 1.0, the APRM gain shall be increased by the ratio: NFLPD FRP where: MFLPD = maximum fraction of
- limiting power density where the -
limiting power density is 13.4 i
kW/ft for 8 x 8 fuel.
i FRP = fraction of rated
', power (1593 MWt).
In the event of operation with the ratio of NFLPD to FRP equal to or less than 1.0, the APRM gain shall be equal to or greater than.1.0.
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 thermal power.
Amendment No. 64, Sa I
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VYWPS
- 1. SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING
- b. Flux Scram Trio Settina (Refuel or Startup and Hot Standby 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 of full scale.
a Amendment No. SA. 78 Sb e
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VYWPS 1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING B. Core Thermal Power Limit (Reactor Pressure B. APRM Rod Block Trip Setting g800 psia or Core Flow I of Rated)
- 1. The APRM rod block trip setting shall be When the reactor pressure is 1800 psia or as shown in Figure 2.1.1 and shall be:
core flow 1 1 0% or rated, the core thermal power shall not exceed 25% of rated thermal S RB 1066W + 42%
power.
where:
C. Power Transient SRB = rod block setting in i To ensure that the safety limit established percent of rated thermal 4 in Specification 1.1A and 1.1B is not power (1593 MWt).
exceeded, each required scram shall be i
initiated by its expected scram signal. The W = percent rated drive flow safety limit shall be assumed to be exceeded where 100% rated drive ~
when scram is accomplished by means other flow is that flow than the expected scram signal. equivalent to 48 x 106
- lbs/he core flow.
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 13.4 kW/ft for 8 x 8 fuel.
FRP = fraction of rated power (1593 MWt).
Amendment No. 64 '
6 i
VYWPS 1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING 9
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.
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i Amendtrent No. 64 6a e
VYMPS TABLE 3.1.1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT REOUIREMENTS Modes in Which Functions Must be Minimum Number Required Conditions When Operating Operating Instrument Minimum Conditions For Refuel Startup Run Channels Per Operation Are Not Trip Function Trip Settings (1) (12) Trip System (2) Satisfied (3)
I
- 1. Mode Switch I K K 1 A in Shutdown l
a
- 2. Manual Scram K K K 1 A
- 3. IRM j High Flux s 120/125 I K K(11) 2 A l INOP K K K(11) 2 A -
I
- 4. APRM High Flux s 0.66W+54%(4) K 2 A or B l
j (flow bias) j High Flux s 15% K K 2 A I (reduced)
INOP 1 2(5) A or B j Downscale 1 2/125 K 2 A or b i
l S. High Reactor s 1055 psig K K K 2 A j Pressure 1
4
- 6. High Drywell 1 2.5 psig K K K 2 A l
Pressure
- 7. Reactor Low (6) 1 127.0 inches K K K 2 A Water Level
- 8. Scram Discharge S 21 gallons I K K 2 A J Volume High fpervolume)
Level
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l Amendment No. 39, SS, 78, 78, 79 19 1
- 9. Channel signals for the turbine control valve fast closure I. rip 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 perinitted when the first stage turbine
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pressure is less than 30% of normal (220 psia).
- 11. The IRN 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,
- 5. Scram dircharge 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. 8A, 78
- 21a i
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VYWPS TABLE 3.2.1 (Cont)
ENWRGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION Himh Pressure Coolant Ittiection System Cinimum Number of Operable Instrument Required Action When Minimum Channels per Trip Conditions for Operation are System Trio Function Trio Level Settina Not Satisfied 2 (Note 3) Low-Low Reactor Vessel Same as LPCI Note 5 Water Level 2 (Note 4) Low Condensate Storage > 3% Note 5 Tank 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 $177 inches above top Note 5 Water Level of enriched fuel Amendment No. 68, 85 38 e
VYMPS .
J TABLE 3.2.2 -
PRIMARY CONTAINNENT ISOLATION IBSTRUMENTATION I
tinimaanImmeber of
- operable Instrument Required Action When Minimum
! Channels per Trip Conditions for operations are l
i System -
Trio Function Trip Settina Not Satisfied (Note 2) 2 Low-Low Reactor vessel 182.5" above the A Water Level top of enriched fuel 1
)
) 2 af 4 in each of . liighMainhteamLine 12120F B >
l 2 channels Area Temperature i
2/cteen line High Main Steam Line Flow $140% of rated flow B 2/(Note 1) Low Main Steam Line Pressure 1800pois ,
a i
, 2/(Wote 6) High Main Steam Line Flow 140% of rated flow B
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2 Low Reactor Vessel Same as Reactor A Water Level Protection System 2 High Main Steam Line $3 I background at 5 Radiation (7) (8) rated power (9) 2 High Drywell Pressure Same as Reactor A 1 -
Protection System j 2/(Note 10) Condenser Low Yacuum $12" Hg absolute A l
l 1 Trip System Logic -
A 1
i Amendment No fe[, f(, 86 41 i
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VYNPS TABLE 3.2.5 CONTROL NOD BLOCK INSTRUMENTATION Minimum Number of operable Instrument Modes in Which Function Channels per Trip Must be operable System (Note 1) Trip Function Refuel Startup Run Trip Settina Startup Range Monitor 2 a. Upscale (Note 2) I I i 5 x 105 cps (Note 3)
- b. Detector Not Fully Inserted X X Intermediate Range Monitor (Note 1) 2 a. Upscale K X 1 108/125 Full Scale 2 b. Downscale (Note 4) K K 1 5/125 Full Scale 2 c. Detector Not Fully Inserted I I -
Average Power Range Monitor 2 a. Upscale (Flow Blas) K i 0.66W + 42% (Note 5) 2 b. Downscale 1 1 2/125 Full Scale Rod Block Monitor (Note 6)
(Note 9)
I a. Upscale (Flow Bias)(Note 7) I i 0.66W + N (Note 5) 1 b. Downscale (Note 7) I 1 2/125 Full Scale (Note 8) 1 Scram Discharge Volume I X X 1 12 Gallons (per volume) 1 Trip System Logic X X X e
Amendment No. SS, 72 76 47
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VYWpS TABLE 3.2.5 WOTES
- 1. There shall be two operable or tripped trip systems for each function in 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 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 nus.ber of instrument channels are not available for both trip systems, the systems shall be tripped.
- 2. One of these trips may be bypassed. The SRM function may be bypassed in the higher IRN ranges when the IRN upscale rod block is operable.
- 3. This function may be bypassed when count rate is >100 cys or when all IRN range switches are above position 2.
- 4. IRN downscale may be bypassed when it is on its lowest scale.
- 5. "W** is percent rated drive flow where 100% rated drive flow is that flow equivalent to 48 x 10 6 lbs/hr core flow. Refer to L.C.O. 3.11.C for acceptable values for N. .
- 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 30-day period. ,
- 7. The trip may be bypassed when the reactor power is $30% of rated. An RBN 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 without one hour.
- 9. With one RBN channel inoperable:
- a. Verify that the reactor is not operating on a limiting control rod pattern, and
- b. Restore the inoperable RBN channel to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Otherwise, place the inoperable rod block monitor channel in the tripped condition within the next hour.
Amendment No. 6A, 73, 76, 48
VYWPS TABLE 3.2.6 POST-ACCIDRET INSTRUMENTATION Mininasm Wumber of Operable Instrument Instrument Channels Parameter Type of Indication Range 2 Drywell Atmospheric Temperature Recorder #16-19-45 0-3000 F (Note 1) Recorder #7R-1-149 0-3000F 2 Drywell Pressure (Note 1) Recorder #16-19-44 0-80 psis Torus Pressure (Note 1) 0-80 psia 2 Torus Water Level (Note 3) Meter #16-19-46A 0-3 ft.
Meter #16-19-46B 0-3 fL.
2 Torus Water Temperature (Note 1) Meter #16-19-48 60-1800F 2 Reactor Pressure (Note 1) Recorder #6-97 0-1200 psig Meter #6-90A 0-1200 psig Meter #6-908 0-1200 psig -
2 Reactor Vessel Water Level Meter #2-3-91A (-150)-0-(+150)"H 2O (Note 1) Meter #2-3-91B (-150)-0-(+150)"H 2O 1 Control Rod Position (Note 1.2) Meter 0-4S" RPIS 1 Neutron Monitor (Note 1.2) Meter 0-125% Rated flux 1 Torus Air Temperature (Note 1) Recorder #TR-16-19-45 0-300 0 F 2/ valve Safety / Relief Valve Position Lights (SRV 2-71-A through D) Closed - Open from pressure switches (Note 4) 1/ valve Safety Valve Position from Meter 21-2-1A/B Closed - Open Acoustic Monitor (Note 5)
Note 1 - From and af ter the date that one of these parameters is not indicated in the Control Room, continued reactor
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operation is permissible during the next seven days. If reduced to one indication of a parameter operation is perwissible for 30 days.
Note 2 - Control rod position and neutron monitor instruments are considered to be redundant to each other.
Amendment No. 63 49
VTEPE TABLE 4.2.1 (Cont)
Loer Pressure Coolant Iedection System Trip Punction Punction Test (8) Calibration (8) Instrument Check Loer Reactor Pressure #1 (Note 1) Once/ Operating Cycle --
Nigh Drywell Pressure #1 (Note 1) Once/ Operating Cycle Once Each Day Loer-Low Reactor Vessel Water Level (Note 1) Once/ Operating Cycle Once Each Day Reactor Vessel Shroud Level (Note 1) Every 3 Months -
Lost Reactor Pressure #2 (Note 1) Every 3 Months --
l RNR Pump Discharge Pressure (Note 1) Every 3 Months --
High Drywell Pressure #2 (Note 1) Every 3 Months -
Loer Reactor Pressure #3 (Note 1) Once/ Operating Cycle --
l Auxiliary Power Monitor (Note 1) Every Refueling Outage Once Each Day Pump Bus Power Monitor (Note 1) None Once Each Day LPCI Crosstie Monitor Mone None Once Each Day Trip System Logic Every 6 Months Every 6 Months ---
(Note 2) (Note 3)
Amendment No. 58, 76 51 e
VYWPS 3.6 LIMITING CONDITIONS FOR OPERATION 4.6 SURVEILI.ANCE REQUINEMENTS
- 2. All hydraulic snubbers whose seal materials are other than ethylene propylene or other material that has been demonstrated to be compatible with the operating environment shall be visually inspected for operability every 31 days.
- 3. The initial inspection shall be perforined within 6 months from the date of issuance of these specifications. For the purpose of entering the schedule in Specification 4.6.I.1, it shall be assumed that the facility had been on a 6-month inspection interval.
- 4. Once each refueling cycle, a representative sample of approximately 10% of the snubbers shall be functionally tested for operability including .
verification of proper piston movement, lockup, and bleed. For each unit and subsequent unit found inoperable, an additional 10% shall be so tested until no more failures are found or all units have been tested. Snubbers of rated capacity greater than 50,000 lbs need not be functionally tested.
J. Thermal Hydraulic Stability J. Theriaal Hydraulic Stability When the reactor mode switch is in RUN, the reactor shall not intentionally be operated in a natural circulation mode nor shall an idle recirculation pump be started with the reactor in a natural . circulation mode.
Amendment No. 64 110b A _ _
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VYWPS 3.10 LIMIT 15C COWDITIONS FOR OPERATION 4.10 SURVEILLANCE REQUIRmENTS
- 3. Operation with Inoperable C--_: rnts B. Operation with Inoperable C-__:- -ts Whenever the reactor is in Bun Mode or Startup Mode with the reactor not in the Cold Condition, the requirements of 3.10.A shall be met except:
- 1. Diesel Generators 1. Diesel Generators From anc after the date that one of the When it is determined that one of the diesel diesel generators or its associated buses are generators is inoperable the requirements of made or found to be inoperable for any reason Specification 4.5.H.1 shall be satisfied.
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 date that ventilation Samples of the Battery Room atmosphere shall is lost in the battery room portable be taken daily for hydrogen concentration -
ventilation equipment shall be provided. determination.
- 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.
- c. From and after the date that one of the two 24 volt ECCS Instrumentation battery 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.
Amendment No. 63 176
VYWp3 LIMITING COWDITIOES FOR OPERATION SURVEILLANCE REQUIREMENTS 3.11 REACTOR FUEL ASSEMBLIES 4.11 REACTOR FUEL ASSEMBLIES Applicability: Applicability:
The Limiting Conditions for Operation associated The Surveillance Requirements apply to the with the fuel rods apply to these parameters which parameters which monitor the fuel rod operating monitor the fuel rod op eating conditions. conditions.
Objective: Objective:
The Objective of the Limiting Conditions for The Objective of the Surveillance Requirements is operation is to assure the perfomance of the fuel to specify the type and frequency of survelliance rods. to be applied to the fuel rods.
Speelfications: Specifications: ,
A. Average Planar Linear Heat Generation Rate A. Average Planar Linear Heat Generation Rate -
(APLHCR) (APLHGR) .
During steady state power operation, the The APLHCR for each type of fuel as a APLHOR for each type of fuel as a function of function of average planar exposure shall be average planar exposure shall not exceed the determined daily during reactor operation at limiting values shown in Tables 3.11-1A >25% rated thermal power.
through C. If at any time during steady state operation it is determined by normal surveillance that the limiting value for APLHCR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits.
If the APLHCR is not returned to within Amendment No. 64 '
180-a i __ . _ _ _ __.
r VYWpS LIMITIEC CONDITIONS FOR OPERATION SURVEILLANCE REQUIERHENTS prescribed limits within two (2) hours, the reactor shall be brought to the shutdown conditions within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.
B. Linear Heat Generation Este (LMGE) B. Linear Heat Generation Este (LNGE)
During steady state power operation, the The LHCR as a function of core height shall linear heat generation rate (tJtGR) of any rod be checked daily during reactor operation at in any fuel assembly at any axial location 1 25% rated therinal power.
shall not exceed the maximum alloweble LHGE of 13.4 kW/fL for 8x8, 8x8R, and 78x8R.
If at any time during steady state operation C. Miniansa critical power Ratio -
it is determined by normal surveillance that the limiting value for LHGR is being MCpt shall be deterinined daily during reactor exceeded, action shall be initiated within 15 power operation at 1 25% rated thermal power minutes to restore operation to within the and following any change in power level or prescribed limits. If the LNGR is not distribution that would cause operation with returned to within the prescribed limits with a limiting control rod pattern as described two (2) hours, the reactor shall be brought in the bases for Specification 3.3.B.6.
to shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.
Amendment No. 64, 180-b e
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VTWPS Tabla 3.11-2 NCPR Operating Limits MCPR Operating Limit for Value of "B" in RBM Average Control Rod Cycle Puel Type (2)
Equation (1) Scram Time Exposure Range sus 8x8R P8x8R 42% Equal or better BOC to EOC-2 GWD/T 1.29 1.29 1.29 than L.C.O. EOC-2 CWD/T to EOC-1 GWD/T 1.29 1.29 1.29 3.3 C.I.1 EOC-1 CWD/T to EOC 1.30 1.30 1.30 Equal or better BOC to EOC-2 GWD/T 1.29 1.29 1,29 than L.C.O. EOC-2 CWD/T to EOC-1 CWD/T 1.33 1.31 1.31 3.3 C.1.2 EOC-1 CWD/T to EOC 1.36 1.35 1.35 41% Equal or better BOC to EOC-2 GWD/T 1.25 1.25 1.25 than L.C.O. EOC-2 GWD/T to EOC-1 CWD/T 1.26 1.25 1.25 3.3 C.1.1 EOC-1 CWD/T to EOC 1.30 1.30 1.30 Equal or better BOC to EOC-2 CWD/T 1.25 1.25 1.25 than L.C.O. EOC-2 CWD/T to EOC-1 CWD/T 1.33 1.31 1.31 3.3 C.I.2 EOC-1 CWD/T to EOC 1.36 1.35 1.35 s 40% Equal or better BOC to EOC-2 GWD/T 1.25 1.25 1.25 than L.C.O. EOC-2 CWD/T to EOC-1 GWD/T 1.26 1.25 1.25 -
3.3 C.I.1 EOC-1 CWD/T to EOC 1.30 1.30 1.30 Equal or better BOC to EOC-2 CWD/T 1.25 1.25 1.25 than L.C.O. EOC-2 CWD/T to EOC-1 CWD/T 1.33 1.31 1.31 -
3.3 C.I.2 EOC-1 CWD/T to EOC 1.36 1.35 1.35 I
(1) The Rod Block Monitor (RBM) trip setpoints are deter 1 mined by the equation shown in Table 3.2.5 of the Technical Specifications.
(2) The current analyses for MCPR Operating Limits do not include 7x7 fuel. On this basis, further evaluation of MCPR Operating Limits is required before 7x7 fuel can be used in reactor power operation.
Amendment No. 72, 180-01 e
i
VYMPS 3.13 LIMITI CONDITIONS FOR OPERATION 4.13 SURVEILLANCE REQUIREMENTS 3 ." Except as specified in Specification 1. Cycling each valve in the 3.13.C.4 below, the Turbine Building flow path that is not Foam System shall be operable with its testable during plant foam concentrate tank full (150 gallons). operation through at least one complete cycle of full
- 4. From and after the date that the Turbine travel.
Building Foam System is inoperable a I portable foam nozzle shall be brought to 2. A visual inspection of the the Turbine Building Foam System foam system and equipment to location. A 150 gallon foam concentrate verify integrity, and supply shall be available on-site.
- 3. A visual inspection of the Recirculation M.G. Set Foam System foam nozzle ares to verify that the spray pattern is not obstructed. -
- 4. Foam concentrate samples shall be taken and analyzed for acceptability,
- d. At least once per 3 years by performing an air flow test through the Recirculation M.C. Set foam header and verifying each foam nozzle is unobstructed.
Amendment No. 67, 187-1 e
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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 ORCANIZATION A.
The Plant Manager has on-site responsibility for the safety and efficient operation of the facility.
. Succession to this responsibility during his absence shall be delegated in writing. t B.
The portion of the corporate management Which relates to the operation of this plant is shown in Figure 6.1.1.
C.
In all natters relating to the operation of the plant and to those Technical Specifications, the Plant Manager shall report to and be directly responsible to the Manager of Operations.
D.
Conduct conditions.
minimum of operations of the plant is shown in Figure 6.1.2 and will be in accordance with the following
.I 1.
An individual qualified in radiation protection procedures shall be present on-site at all times when there is fuel in the reactor.
- 2. Minimum shift staffing on-site shall be in accordance with Table 6.1.1.
3.
A dedicated, licensed Senior Operator shall be in charge of any reactor core alteration.
4.
Qualifications with regard to educational background experience, and technical specialities 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 f. Operations Supervisor
- b. Operations Superintendent g. Reactor and Computer Supervisor
- c. Technical Services Superintendent h. Maintenance Supervisor i d. Maintenance Superintendent 1. Instrument and Control Supervisor l e. Chemistry and Health Physics Supervisor J. Shift Supervisors l
1 I
l 190 Amendment No. 88, 75
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- - RESPONSIBLE FOR FIRE PROTECT 10lf A cist OF TutSE F05fTIONS WILL DE PMICNATED
- - ANSE 88.8-4978 RE: LICENSE AS ALTitsIATE TO THE PLANT MANAGER AND MAY
$0 - LICEmSto Statoa OPEaAT0a se mETITLED AS$1sTANT Pl. ANT MAfBACER 0 - LtCamsto OPERATos AUIILIARY , - AmtatSTRATIVE POSITICWS por suonne. IN IIEALTu PNTSICS MATTERS. TE CIsEMISTRY OPERATORS & IIEALTM PNYSICS SWEavt$0a sans DIRECT ACCESS TO TME FIAaff MANACER
- 192 ORGANIZATIONAL CHART FIGURE 6.1.2 Amendment No. 66. 75. 79. 80
VYNPS TABLE 6.1.1 V;rmont Yankee staff positions that shall be filled by personnel holding Senior Operator and Operator licenses are indicated in the following table:
Title License Requirements Operations Supervisor Licensed Senior Operator Shift Supervisor Licensed Senior Operator Supervisory Control Room Operator Licensed Senior Operator Control Room Operator Licensed Operator MINIMUM SHIFT STAFFING ON-SITE Conditions _
Plant Cold .
Startup Shutdown or and Refueling with Normal Operation (Note 1) Fuel in the Reactor (Note 2) '
Shift Supervisor 1 1 -
Supervisory Control Room Operator 1 -
Centrol Room Operator 2 1 Auxiliary Operator 2 1 Shift Engineer 1 -
P.tes (1) At least one Senior Licensed Operator and one Licensed Operator, or two Senior Licensed Operators, shall be in the Control Room.
(2) At least one Licensed Operator, or one Senior Licensad Operator, shall be in the Control Room, e
193 Amendment No. 83, 79 l
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