ML19276E052
| ML19276E052 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 01/31/1979 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19276E053 | List: |
| References | |
| NUDOCS 7903020242 | |
| Download: ML19276E052 (10) | |
Text
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UNITED STATES y"
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NUCLEAR REGULATORY COMMISSION
.j WASHINGTON,0. C. 20555
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VERMONT YANKEE NUCLEAR POWER CORPORATION DOCKET NO. 50-271 VERMONT YANKEE NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 50 License No. DPR-28 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The applications for amendment by Vermont Yankee Nuclear Power Corporation (the licensee) dated December 10, 1976, April 14, 1977, and May 16, 1978, comply 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 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.
7003030242.
. 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B. of Facility Operating License No. DPR-28 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 50, 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 the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION hY.
Thomas A.
ppolito, Chief Operating Reactors Branch #3 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance:
January 31, 1979
ATTACHMENT TO LICENSE AMENDMENT NO. 50 FACILITY OPERATING LICENSE NO. DPR-28 DOCKET NO. 50-271 Revise Appendix A Technical Specifications as follows:
Remove Insert 34 34 49 49 60 60 129 129 129a 138 138 139 139 Changes on the revised pages are shown by marginal lines.
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c.
4.2 SURVEILLANCE REQUIREMENTS 3.2 LIMITING CONDITIONS FOR OPERATION F.
Mechanical Vacuum Pump Isolation F.
Mechanical Vacuum Pump Isolation During each operating cycle, auto-
- 1. Whenever the main steam line matic isolation and securing of isolation valves are open, the the mechanical vacuum pump shall be mechanical vacuum pump shall be verified while the reactor is capable of being automatically isclated and secured by a signal shutdown.
of high radiation in the main steam line tunnel or shall be manually isolated and secured.
- 2. If Specification 3.2.F.1 is not met following a routine surveil-lance check, the reactor shall be in the cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
G.
Post-Accident Instrumentation G.
Post-Accident Instrumentation During the reactor power operation, the The post-accident instrumentation instrumentation that displays information shall be functionally tested and in the control room necessary for the calibrated in accordance with Table 4.2.6.
operator to initiate and control the systems used during and following a postulated accident of abnormal operating condition shall be operable in accordance with Table 3.2.6.
H.
Drywell to Torus aP Instrumentation H.
Drywell to Torus AP Instrumentation 1.
During reactor power operation, the The Drywell to Torus AP Instrumenta-Drywell to Torus AP Instrumentation tion shall be calibrated once every (recorder #1-156-3 and instrument six months and an instrument check DPl-1-158-6) shall be operable will be made once per shift.
except as specified in 3.2.H.2.
2.
From and after the date that one of the drywell to torus AP instruments is made or found to be inoperable for any reason, reactor operation is permissible only during the succeeding thirty days unless the instrument is sooner made operable.
If both instruments are made or found to be inoperable, and indica-tion cannot be restored within a six hour period, an orderly shutdown shall be initiated and the reactor shall be in a hot shutdown condition in six hours and a cold shutdown condition in the following eighteen hours.
34 Amendment No. 50
VYNPS TABLE 3.2.6 POST-ACCIDENT INSTRUMENTATION Minimum Number Instrument of Operable Instrument Channels Parameter Type of Indication Range 2
Drywell Atmospheric Recorder #16-19-45 0-300*F Temperature (Note 1)
Recorder #TR-1-149 0-300 F 2
Drywell Pressure (Note 1)
Recorder #16-19-44 0-80 psia 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 ft.
2 Torus Water Temperature Meter #16-19-48 60-180 F (Note 1) 2 Reactor Pressure (Note 1)
Recorder #6-97 0-1200 psig Meter #6-90A 0-1200 psig Meter #6-90B 0-1200 psig 2
Reactor Vessel Water Level Meter #2-3-91A
(-150)-0-(+150)"HO 2
(Note 1)
Meter #2-3-91B
(-150)-0-(+150)"H 0 2
1 Control Rod Position Meter 0-48" RPIS (Note 1,2) l Neutron Honitor (Note 1,2)
Meter 0-125i Rated Flux 1
Torus Air Temperature (Note Recorder #TR-16-19-45 0-300*F 1)
Note 1 - From and after the date that one of these parameters is not indicated in the control room, continued reactor operation is permissible during the next seven days.
If reduced to one indication of a parameter operation is permissible for 30 days.
Note 2 - Control rod position and neutron monitor instruments are considered to be redundant to each other.
Note 3 - From and after the date that this parameter is reduced to one indication in the control room, continued reactor operation is permissible during the next thirty 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 />.
49 Amendment No. 50
VYNPS TABLE 4.2.6 CALIBRATION FREQUENCIES POST-ACCIDENT INSTRUMENTATION Parameter Calibration Instrument Check Drywell Atmosphere Temperature every 6 months once each day Drywell and Torus Pressure every 6 months once each day Torus Water Level every 6 months once each shift Torus Water Temperature every 6 months once each day Reactor Pressure every 6 months once each day Reactor Vessel Water Level every 6 months once each day Control Rod Position (Note 5) once each day Neutron Monitor Same as reactor once each day protection systems Torus Air Temperature every 6 months once each day 60 Amendment No. 50
4.2 SURVEILLANCE REQ!!IREMENTS 3.7 LIMITING CONDITIONS FOR OPERATION
- c. Reactor operation may continue for (4) A drywell to suppression chamber leak rate test shall demonstrate fifteen (15) days provided that at that with an initial differential least one position alarm circuit pressure of not less than 1.0 psi, for each vacuum breaker is operable the differential pressure decay and each suppression chamber -
rate shall not exceed the drywell vacuum breaker is physically equivalent of the leakage rate verified to be closed imediately through a 1-inch orifice.
and daily thereafter.
- 7. Oxygen Concentration 7.
Oxygen Concentration a.
The primary containment atmosphere The primary containment oxygen shall be reduced to less than 4 concentration shall be measured and recorded on a weekly basis, percent oxygen with nitrogen gas during reactor power operation with reactor coolant pressure above 90 psig, except as specified in Specification 3.7.A.7.b.
b.
Within the 24-hour period subsequent to placing the reactor in the Run mode following a shutdown, the containment atmosphere oxygen concen-tration shall be reduced to less than 4 percent and maintained in this condition. Deinerting may comence 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to a shutdown.
8.
If Specification 3.7. A.1 through 3.7.A.7 cannot be met, an orderly shutdown shall be initiated imediately and the reactor shall be in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
129 Amendment No. 50
i 3.7 LIMITING CONDITIONS FOR OPERATION 4.7 SURVEILLANCE REQUIREMENTS
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9.
Drywell/ Suppression Chamber d/p 9.
Drywell/ Suppression Chamber d/p a.
Differential pressure between the a.
The differential pressure drywell and suppression chamber between the drywell and shall be maintained >l.7 psi suppression chamber shall be except as specified in 3.7.A.9.b recorded once per shift.
and 3.7.A.9.c below, b.
The operability of the low b.
The >1.7 psi differential pressure differential pressure alarm shalT be established within 24 shall be verified once per hours of achieving operating week.
pressure and temperature. The differential pressure may be reduced to <l.7 psi 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to comencing a cold shutdown.
c.
The differential pressure may be reduced to <l.7 psi for a maximum of four hours (period to begin when the AP is reduced to <l.7) during required operability testing of the HPCI system pump, the RCIC system pump, the drywell-suppression chamber vacuum breakers, and the suppression chamber-reactor building vacuum breakers, and SBGTS testing.
d.
If the specifications of 3.7.A.9.a cannot be met, and the differential pressure cannot be restored within the subsequent six (6) hour period, an orderly shutdown shall be initiated and the reactor shall be in a Hot Shutdown condition in six (6) hours and a Cold Shutdown condition in the following eighteen (18) hours.
129a Amendment No. 50
VYNPS Bases:
3.7 STATION CONTAINMENT SYSTEMS A.
Primary Containment The integrity of the primary containment and operation of the core standby cooling systems in combination limit the off-site doses to values less than those suggested in 10 CFR 100 in the event of a break in the primary system piping. Thus, containment integrity is specified whenever the potential for violation of the primary reactor system integrity exists. Concern about such a violation exists whenever the reactor is critical, above atmospheric pressure and temperature above 212 F.
An exception is made to this require-ment during initial core loading and while a low power test program is being conducted and ready access to the reactor vessel is required. The reactor may be taken critical during this period; however, restrictive Procedures operating procedures will be in effect again to minimize the probability of an accident occurring.
and the Rod Worth Minimizer would limit control worth to less than 1.30% delta k.
The pressure suppression pool water provides the heat sink for the reactor primary system energy release following a postulated rupture of the system. The pressure suppression chamber water volume must absorb the associated decay and structural sensible heat released during primary system blowdown from 1000 psig.
Since all of the gases in the drywell are purged into the pressure suppression chamber air space during a loss-of-coolant accident, the pressure resulting from isothermal compression plus the vapor pressure of the liquid must not exceed 62 psig, the allowable pressure suppression chamber pressure. The design volume of the suppression chamber (water and air) was obtained by considering that the total volume of reactor coolant to be condensed is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber (Reference Section 5.2 FSAR).
Using the minimum or maximum water volumes given in the specification, contain Qt pressure during the design 1 The minimum vo ( e of basis accjdent is approximately 44 psig, which is below the design of S6 psig.The majority of the Bodega testsh were run 68,000 ft results in a submergency of approximately four feet.
with a submerged length of four feet and with complete condensation. Thus, with respect to downcomer submergence, this specification is adequate.
The maximum temperature at the end of blowdown tested during the Humboldt BayU) and Bodega Bay tests was 170 F and this is conservatively taken to be the limit for complete condensation of the reactor coolant, although condensation would occur for temperature above 170 F.
In conjunction with the Mark I Containment Short Term Program, a plant unique analysis was performed (see Vermont Yankee letter dated September 13, 1976) which demonstrated a factor of safety of at least two for the weakest element in the suppression chamber support system and attached piping. The maintenance of a drywell-suppression chamber differential pressure of 1.7 psid and a suppression chamber water level corresponding to a downcomer submergence range of 4.29 to 4.54 feet will assure the integrity of the suppression chamber when subjecte:! to post-LOCA suppression pool hydrodynamic forces.
138
VYNPS 3.7.A (cont'd)
Using a 50 F rise (Section 5.2.4 FSAR) in the suppression chamber water temperature and a minimum water volume of 68,000 ft, the 170 F temperature which is used for complete condensation would be approached 3
Maintaining a pool temperature only if the suppression pool temperature is 210 F prior to the DBA-LOCA.
of 90"F will assure that the 170 F limit is not approached.
Experimental data indicate that excessive steam condensing loads can be avoided if the peak temperature of the suppression pool is maintained below 160 F during any period of relief valve operation with sonic conditions at Specifications have been placed on the envelope of reactor operating conditions so that the the discharge exit.
reactor can be depressurized in a timely manner to avoid the regime of potentially high suppression chamber loadings.
In addition to the limits on temperature of the suppression chamber pool water, operating procedures define the This action would include:
action to be taken in the event a relief valve inadvertently opens or sticks open.
(1) use of all available means to close the valve, (2) initiate suppression pool water cooling heat exchangers, (3) initiate reactor shutdown, and (4) if other relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open relief valve to assure mixing and uniformity of energy insertion to the pool.
Double isolation valves are provided on lines which penetrate the primary containment and open to the free space of Closure of one of the valves in each line would be sufficient to maintain the integrity of the the containment.
Automatic initiation is required to minimize the potential leakage paths frem the pressure suppression system.
containment in the event of a loss-of-coolant accident. Details of the isolation valves are discussed in Section 5.2 of the FSAR.
The purpose of the vacuum relief valves is to equalize the pressure between the drywell and suppression chambe suppression chamber and reactor building so that the structural integrity of the containment is maintained.
l Technical Specification 3.7.A.7.c is based on the assumption that the operability testing of the pressure suppression chamber-reactor building vacuum breaker, when required, will normally be performed during the same four hour testing interval as the pressure suppression chamber-drywell vacuum breakers in order to minimize operation with <l.7 psi, differential pressure.
The vacuum relief system from the pressure suppression chamber to reactor building consists of two 100% vacuum relief breakers (2 parallel sets of 2 valves in series). Operation of either system will maintain the pressure differential less than 2 psig; the external design pressure is 2 psig.
The capacity of the ten (10) drywell vacuum relief valves is sized to limit the pressure differential between the suppression chamber and drywell during post-accident drywell cooling operations to the design limit The ASME Boiler They are sized on the basis of the Bodega Bay pressure suppression tests.
of 2 psig.
and Pressure Vessel Code,Section III, Subsection B, for this vessel allows eight (8) operable 139 Amendment No. 50