ML20151T666
| ML20151T666 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 08/09/1988 |
| From: | Wessman R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20151T670 | List: |
| References | |
| NUDOCS 8808160434 | |
| Download: ML20151T666 (16) | |
Text
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>A C tc UNITED STATES NUCLEAR REGULATORY COMMISSION 2
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l W ASM NG TO N. D. C. 20555
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VERMONT YANKEE Nt! CLEAR POWER CORPORATION DOCKET NO 50-771 VERMONT YANKEE MUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendrent No.106 t.icense No. DPR-28 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Vermont Yankee Nuclear Power Corporation (the licensee) dated November 30, 1987 as supplemented on January 20, 1988 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I:
B.
The facility will operate in conformity with the appitcation, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR part 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the Itcense is amended by changes to the Technical Soecifications as indicated in the attachment to this license amendment.
and paragraph 3.B of Facility Ooerating License No. DPR-28 is hereby anended to read as follows:
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8008160434 000 g PDR ADOCK 0 pg P
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2 (R) Technical Specifications The Technical Specifications, contained in Appendix A, as revised through Amendrient Nn.106, are hereby incorporated in the license. The licensee shall operate the facility in i
accordance with the Technical Specifications.
3.
This license amendment is e#fective as of the date of its issuance.
FOR THE NUCLEAR REGULAT v COMMISSION
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^ Richard H. Wessman, Director Cs'4C eg Project Directorate I-3 4
Division of Reactor Projects I/II
Attachment:
Changes to the Technical Specifications Date of Issuance:
August 9, 1989 1
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' ATTACHMENT TO LICENSE AMENOMENT NO.100 FACILITY OPERATING LICENSE NO. DPR-28 DOCKET NO. 50-271 Reolace the following pages o' the Appendix A Technical Specifications with the enclosed pages.
The revised oaces are identi'ied by amendment number and contain vertical lines indicating the areas of change.
The corresponding-overleaf Dages are providad to maintain dccument completeness.
Remove Pages Insert Paaes 50 50 51 51 52 52 53
~53 53a 53a 54 54 55 55 56 56 57 57 58 SR 59 59 64 64 67 67
1 VYNFS TARI.E 4.2.1 Minimum Test and Calibration Frequencies Emergency Core Coolina Actuation lustrumentation i
l Core Spray System Trip Function Functional Test (8)
Calibration (8)
Inst raiment check Migh Drywell Pressure (Note 1)
Once/ Operating Cycle Once Eacia Day 3
low-low Reactor Vessel (hte 1)
Once/ Operating Cycle Once E. sci. Day j
Water level low Reactor Fressure (hte 1)
Once/ Operating Cycle 4
Fump 14-1A. Discisarge Fress (Note 1)
Every Three Months Auxiliary Power k nitor (Note 1)
Every Refuelling once Each Day
{
Fump Rua Power Monitor (Note 1)
None Once E.sch Day Nigh Sparger Pressure (Note 1)
Every Three Months g
Trip System logic Once/ Operating Cycle Once/ Operating Cycle i
(Note 2)
(Note 3) 1 Amendment No. $$. /p,196 so
VYNPS TABLE 4.2.1 (Continued) low Pressure Coolant Injection 5, steam Trip Function Functional Test (8)
Calibratio,,(8) b te m nt check Low Reactor Pressure No. 1 (Note 1)
Once/ Operating Cycle Nigh Drywell Pressure E. 1 (Note 1)
Once/ Operating Cycle Once E.sch Day low-low Reactor Vessel (Note 1)
Once/ Operating Cycle once E.sch Day hter Level Resctor Vessel Shroud level (Note 1)
Every Three h aths l.
Iow Reactor Pressure W. 2 (Note 1)
Every Three Months RNR Fusep Discharge Pressure (Note 1)
Every Three h aths
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Migh Drywell Pressure h. 2 (hte 1)
Every Three haths l
Iow Reactor Pressure b. 3 (Note 1)
Once/ Operating Cycle Auxiliary Power k nitor (Note 1)
Every Refueling; Outage once Each Day Pusmp Bus Power hoitor (hte 1) hoe once Each Day LPCI Crosstle hnitor hoe boe once Each Day Trip Systesa logic Once/ Operating Cycle Once/ Operating Cycle l
(hte 2)
(hte 3)
- h. $$. '16. [, IO6 Asmendament M
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VYMPS TABLE 4.2.1 (Continued)
Ninh Pressure Coolant Injection System Trip runction Functional TestI8)
Calibration (8) trument Check low-law Reactor Vessel (hte 1)
Once/ Operating Cycle Once Each Day Water Level low Condensate Storage Tank (Note 1)
Every Three Months Water Level Nigh Drywell Pressure (Note 1)
Once/ Operating Cycle once Each Day Bus Power k nitor (Note 1)
None Once Each Day Trip System logic Once/ Operating Cycle Oncef0perating Cycle l
(Note 2)
(Note 3)
Amendment No. 38, /d. p,106 y2 m
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VYNFS 1
TABLE 4.2.1 (Continued)
Automatic Depressurization System l
Trip Functios Functional Test (8)
Calltaration(8)
Instrument check tow-Iow Reactor Vessel (Note 1)
Once/ Operating Cycle Once E.acia Day l
Water Level Nigt Drywell Pressure (Note 1)
Once/ Operating Cycle once E. scia Day Bus Power Monitor (Note 1) bee Once Eacia Day Trip System logic Once/ Operating Cycle Once/ Operating Cycle
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(F.scept Solenoids of (Note 2)
(Note 3)
Valves) 1 l
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l Amendment E. 58. N, JM, 106
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VYMPS TABLE 4.2.i (Continued)
Recirculation Pump Trip Actuation System TJ p Function Functional TestI8)
CalibrationI8)
Instrument Check Low-tow Beact r Vessel (Note 1)
Once/ Operating Cycle Once Each Day Water level %)
Nigh Reactor PressureI4)
(Note 1)
Once/ Operating Cycle Once Each Day Trip System logic Once/ Operating Cycle Once/ Operating Cycle l
Amendment No. /I. 1%
VYNFS TABLE 4.2.2 Minimuss Test and Calibration Frequencies Primary Containment Isolation Instrumentation Trip Function Functional Test (8)
Calibration (8) h tsie nt Check tow-Iow Reactor Wessel (Note 1)
Once/ Operating Cycle Once Each Day hter Invel i
Migh Steam Line Area (Note 1)
Each Refueling Outage Temperature Nigh Steam Line Flow (Note 1)
Every Three haths osace Each Day low Main Steam Line Fressure (Note 1)
'v. := Three haths l
l law Reactor Wessel (Note 1)
Once/ Operating Cycle l
Water Invel Nigh h in Steam Line Radiation (Notes 1 and 7)
Each Refueling Outage Once Each Day Nigh Drywell Fressure (Note 1)
Once/ Operating Cycle Once Each Day Condenser tow Vacuum (Note 1)
Every Three haths Trip System logic Once/ Operating Cycle Once/ Operating Cycle (Note 2)
(Ete 3)
I Amendment m. 58. /p, 105 v.
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VYNPS TABLE 4.2.2 (Continued)
Minimum Test and Calibration Frequencies Nimb Pressure Coolant Injection System _ Isolation Instrumentation Trip Function Functional Test (E)
Calibration (8)
Instrumeist Check Nigh Reactor h ter level (Note 1)
Once/ Operating Cycle Nigh Steam Line Space (Note 1)
Each Refueling Outage Temperature Nigh Steam Line D/F (Note 1)
Every Three N nths l
(Steam Line Break)
I low EPCI Steam Supply Pressure (Note 1)
Every Three N oths l
Main Steam Line Tunnel (Note 1)
Each Refueling Outage Temperature Bus Power Monitor (hote 1)
None Once Each Day Trip System Iagic Once/ Operating Cycle Once/ Operating Cycle (Note 2)
(Note 3)
Amendment No. W, 106 ss
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M VYNPS TABLE 4.2.2 (Continued)
Minimum Test and Calibration Frequencies Reactor Core Isolation Cooling System Isolation Instrumentation Trip Function Functional Test (8)
Calibration (8)
Instrument Check Main Steam Line Tunnel (Note 1)
Each Refueling Outage Tamperature High Steam Line Space (Note 1)
Each Refueling Outa, Temperature i
High Steam Line D/P Including (Note 1)
Every Three Months Time Delay Relays (Steam Line l
Break)
High Reactor Water level (Note 1)
Once/ Operating Cycle i
Low ROIC Steam Stpply Pressure (Note 1)
Every Three Months Bus Power Monitor (Note 1)
None Once Each Day Trip System Logic Once/ Operating Cycle Once/ Operating Cycle (Note 2)
(Note 3)
Amendment No. 58 Ay,106
VYNPS TABLE 4.2.3 Minimum Test and Calibration Frequencies Reactor Building Ventilation and Standby Cas Treatment System Isolation Trip Function Functional Test (8)
CalibrationI8)
Ins t runwnt Check Iow Reactor Vessel (Note 1)
Once/ Operating Cycle Water Level High Drywell Pressure (Note 1)
Once/ Operating cycle Reactor Building Vent Monthly Every Three Months Once Each Day l
Exhaust Radlation Refueling Floor Zone Monthly Every Three Months Once Each Day During l
Radiation
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Refueling Reactor Building Vent Once/ Operating Cycle Once/ Operating Cycle l
Trip Systes logic (Note 2)
(Note 3)
Standi ;/ Cas Treatment Once/ Operating Cycle Once/ Operating Cycle l
Trip System logic (Note 2)
(Note 3) lor,1c Bus Power Monitor (Note 1)
None Once Each Day i
i Amendment No. 55,hh,.106 3f 1
VYNPS TABLE 4.2.4 Minimum Test and Calibration Frequencies Off-cas System Isolation Instrumentation Trip Function Functfonal Test (8)
Ca1ibratlon(8)
Iustrument Check l
Augmented Off-Cas Trip System Once/ Operating Cycle Once/ Operating Cycle
- l Logic (AOC)
(Note 2)
(Note 3)
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Amendment No. JJ, 106 Sa
VYNPS i
TABLE 4.2.5 Minimum Test and Calibration Frequencies Control Rod Block Instrumentation Trip Function Functional Test Calibration Startup Range Monitor a.
Upscale Notes 4 and 6 Note 6 b.
Detector Not Fully Inserted Note C Note 6 Intermediate Range Monitor a.
Upscale h tes 4 and 6 Note 6 b.
Downscale Notes 4 and 6 Note 6 c.
Detector Not Fully Inserted hte 6 Note 6 Average Power Range knitor Upscale (Flow Blas) h tes 1 and 4 Every Three Months a.
b.
Downscale Notes 1 and 4 Every Three Months Rod Block Monitor e.
Upscale (Flow Blas)
Notes 1 and 4 Every Three Months b.
Downscale h tes I and 4 Every Three Months Trip System Logic Once/ Operating Cycle once/ Operating Cycle (Nnte 2)
(Note 3)
High hier Level in Scram Every Three Months Refueling Outage l
Discharge Volume 59 Amendme,nt No. 106
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a VYNPS 3.2 (Continued)
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 f rom f ailed fuel due to transient reactor operation.
Pressure instrumentation I* provided which trips when main steam line pressure drops below 8W 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 setpoint limits the depressurization such that no excessive vessel thermal stress occurs as a result of a pressure regulator malfunction.
This setpoint was selected far enough below normal main steam line pressures to avoid spurious primary containment isolations.
Low condenser vacuum has been added as a trip of the Group 1 Isolation valves to prevent release of radioactive gases from the primary coolant through condenser. The setpoint of 12 inches of mercury absolute was selected to provide suf ficient margin to assure retention capability in the condenser when gas flow is stopped and sufficient margin below normal operating valuea.
The HPCI and/or RCIC liigh flow, stese supply pressure, and temperature instrumentation is psovided to detect a break in the HPCI and/or RCIC piping. Tripping of this instrumentation results in actuation of HPCI and/or RCIC isolation
- valves, i.e., Group 6 valves. A time delay has been incorporated into the RCIC steam flow trip logic to prevent the system from inadvertently isolating due to pressure spikes which may occur on startup. The trip settings are such that core uncovering is prevented and fission product releaae is within limits.
The instrumentation which initiates ECCS action is arranged in a dual channel system. Permanently installed circuits
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and equipment may be used to trip instrument channels.
In the nonfall safe systems which require energizing the circuitry, tripping an instrument channel may take the form of providing the required relay function by use of permanently installed circuits. This is accomplished in some cases by closing logic circuits with the aid of the permanently installed test jacks or other circuitry which would be installed for this purpose.
Amendment No. 69, 84 64
VYNPS 4.2 PROTECTIVE INSTRUMENTATION i
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 l
dual-channel arrangement; Each of cliese 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, j
process-operated switches (sensors), histable devices operated by analog sensor signals, timers, limit switches,*
I 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
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actions, such as tripping the main turbine generator unit.
With the dual-channel arrangement of these subsystems, the single failure of a ready circuit can be tolerated because the redundant subsystem or system (in the case of high pressure coolant injection) will then initiate the i
necessary protective action. If a failure in one of these circuits occurs in such a way that an action is taken, the opecator is immediately alerted to the failure. If the failure occurs and causes no action, it could then d
remain undetected, causing a lose 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 l
they are operating properly.
It has been the practice in boiling water reactor plants to functionally test protective instrumentation sensors k
and sensor relays on-line on a monthly frequency. Since logic circuit tests result in the actuation of plant i
i equipment, testing of this nature was done while the plant was shut down for refueling.
In this way, the testing j
of equipment would not jeopardize plant operation.
This Specification is a periodic testing program which is bssed upon the overall testing of protective l
Ir.strumentation systems, including logic circuits as well as sensor circuits. Table 4.2 outlines the test, j
calibration, and logic system functional test schedule for the protective instrumentation systems.
The testing j
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 onerate, and also the proper functioning of the relay contacts, j
Functional testing of the inaccessible temperature switches associated with the isolation systems is accomplished remotely by application of a heat source to individual switches.
l All subsystems are functionally tested, calibrated, and operated in their entirety.
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Amendment No. 22, 98, pp, Jp),106 61 I'
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