ML20085L251
| ML20085L251 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 06/20/1995 |
| From: | Mckee P Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20085L253 | List: |
| References | |
| NUDOCS 9506280391 | |
| Download: ML20085L251 (8) | |
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t UNITED STATES j
NUCLEAR REGULATORY COMMISSION waswinorow, o.c. sosewoot VERMONT YANKEE NUCLEAR POWER CORPORATION DOCKET NO. 50-271 VERMONT YANKEE NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.145 License No. DPR-28 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment filed by the Vermont Yankee Nuclear Power Corporation (the licenseo) dated October 28, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions'of the Act, and the rules and regulations of the Commission; 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 Commission's regulations; D.
The issuance of this amendment will not be inimical to the common i
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 Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment, and j
paragraph 3.B of Facility Operating License No. DPR-28 is hereby amended to read as follows:
9506280391 950620 PDR ADOCK 05000271 P
. Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.145, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3..This license amendment is effective as of its date of issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION Phillip F. McKee, Director Project Directorate I-3 Division of Reactor Prcjects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical
)
Specifications Date of Issuance:
June 20, 1995
t ATTACHMENT TO LICENSE AMENDMENT NO.145, FACILITY OPERATING LICENSE NO. DPR-28 DOCKET NO. 50-271 Replace the following pages of the Appendix A, Technical Specifications, with the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
Remove Insert 53 53 55 55 70 70 74 74 78 78 r
VYNPS TABLE 3.2.6 POST-ACCIDENP INSTRUMENTATION Minimum Number of Operable Instrument Channels Parameter Type of Indication Instrument Ranae 2
Drywell Atmospheric Recorder #TR-16-19-45 0-350*F Temperature (Note 1)
(TE-16-19-30A)
Meter GTI-16-19-308 0-350*F 2
Containment Pressure (Note 1)
Meter GPI-16-19-12A
(-15) -(+260) psig Meter GPI-16-19-12B
(-15) -(+260) psig 2
Torus Pressure (Note 1)
Meter #PI-16-19-36A
(-15) -(+65) psig Meter GPI-16-19-36B
(-15) -(+65) psig 2
Torus Water Level (Note 3)
Meter SLI-16-19-12A 0-25 ft.
Meter #LI-16-19-12B 0-25 ft.
2 Torus Water Temperature Meter #16-19-33A 0-250*F (Note 1)
Meter #16-19-33C 0-250*F 2
Reactor Pressure (Note 1)
Meter #PI-2-3-56A 0-1500 peig Meter GPI-2-3-56B 0-1500 psig 2
Reactor Vessel Water Level Meter #2-3-91A
(-200)-0-(+200) "H O 2
(Note 1)
Meter #2-3-91B
(-200)-0-(+200)"H O g
2 2
Torus Air Temperature (Note 1)
Recorder #TR-16-19-45 0-350*F (TE-16-19-34)
Meter #TI-16-19-41 50-300'F 2/ valve Safety / Relief Valve Position Lights (SRV 2-71-1, 2, 3 Closed - Open From Pressure Switches (A thru D))
(Note 4)
Amendment No. M, H, M, M. +a,145 53 m
n.
. ~.
f' t
j VYNPS I
F TABLE 3.2.6 NOTES i
Note 1 - From and after the date that a parameter is reduced to one I
' 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 <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.
l Note 2 - Deleted.
f 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 <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.
Note 4 - From and after the date that safety / relief valve position from j
pressure switches is unavailable, reactor operation may continue
.t provided safety / relief valve position can be determined from Recorder #2-166-(steam temperature in SRVs, 0-600*F) and Meter 16-19-33A or C (torus water temperature, 0-250*F).
If both parameters 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 <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.
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 l
pressure (-15) -(+260) psig).
If both indications are not available, th9 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 <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.
Note 6 - Within 30 days following the loss of one indication, or seven days j
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 j
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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
i Amendment No. M, e, M, M,
+M, 4M,145 55 I
VYNPS TABLE 4.2.6 CALIBRATION REQUIRDfENTS POST-ACCIDENT INSTRUMENTATION Parameter Calibration Instrusent Check Drywell Atmosphere Temperature Every 6 Months once Each Day Containment Pressure Once/ Operating Cycle
-Once Each Day Torus Pressure Once/ Operating Cycle
'Once Each Day Torus Water Level Once/ Operating Cvele Once Each Day 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 g
Torus Air Temperature Every 6 Months Once Bach Day.
Safety / Relief Valve Position Every Refueling Outage (Note 9)
Once Each Day (a Functional Test to be performed quarterly)
Safety Valve Position Every Refueling Outage (Note 9)
Once Each Day (a Functional Test to be
. performed quarterly.1 i
8 O
Amendment No. W, G, M,-fu,145 70-
l VYNPS TABLE 4.2 NOTES 1.
Initially once per month; thereafter, a longer interval as determined by test results on this type of instrumentation.
2.
During each refueling outage, simulated automatic actuation which opens all pilot valves shall be performed such that each trip system logie 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.
l 5.
Deleted.
6.
Functional tests, calibrations, and instrument checks are not required when these instruments are not required to be operable or are tripped.
Punctional tests shall be performed before each startup with a required i
frequency not to exceed once per week. Calibration shall be performed prior to or during each startup or controlled shutdown with a required frequency not to exceed once per week.
Instrument checks shall be performed at least once per day during those periods when instruments are required to be operable.
i 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.
l 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 i
Section 4.5.F.1.
l P
Amendment No. 64, 96, 446, 145 74
VYNPS pi@
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 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 ar.d minimize spurious operation. The trip settings given in the Specification are adequate to ensure the above criteria are met.
The Specification preserves the effectiveness of the system during periods of maintenance, testing, or calibration, and also minimizes the risk of inadvertent operation; i.e.,
only one instrument channel out of service.
The ADS is provided wath inhibit switches to manually prevent automatic initiation during events where actuation would be undesirable, such as certain ATWS events. The system is also provided with an Appendix R inhibit switch to prevent inadvertent actuation of ADS during a fire which requires evacuation of the Control Room.
Four radiation monitors are provided which initiate isolation of the reactor building and operation of the standby 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 nornal 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 1
release of radioactivity from the main condenser. During an accident, i
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 meismically 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 i
following an accident as required by NUREG-0737 by ensuring continuous on-scale indication of the followir.g: containment pressure in the (-15)
-(+260) psig range; torus water level in the O to 25 foot range (i.e.,
the bottom to 5 feet above the normal water level of the torus pool);
containment hydrogen / oxygen concentrations (0 to 30% hydro 9en and 0 to 25% oxygen); and containment radiation in the 1 R/hr to 10 R/hr gamma.