ML20199L403
| ML20199L403 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 01/19/1999 |
| From: | Bill Dean NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20199L408 | List: |
| References | |
| NUDOCS 9901270220 | |
| Download: ML20199L403 (9) | |
Text
.
'pM%
//
1 UNITED STATES j
j NUCLEAR REGULATORY COMMISSION o'
WASHINGTON, D.C. 20066 4 001
...../
I VERMONT YANKEE NUCLEAR POWER CORPORATION DOCKET NO. 50-271 VERMONT YANKEE NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE i
Amendment No.165 i
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 licensee) dated December 11,1998, complies with the standards and requirements i
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 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.
l l
9901270220 990119 PDR ADOCK 05000271 P
PDR l
2-
- 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 Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 165, 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, to be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION William M. Dean, Director Project Directorate 1-2 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications j
Date of issuance: January 19, 1999 1
(
l DATED: aannarv 19-togg AMENDMENT NO. 165 TO FACILITY OPERATING LICENSE NO. DPR-28 VERMONT i
YANKEE NUCLEAR POWER STATION 1
b l
PUBLIC i
J. Zwolinski W. Dean
)
R. Croteau i
T. Clark C. Berlinger OGC l
G. Hill (2) l W. Beckner ACRS T. Harris (TLH3)
C. Cowgill, RI l
i l
l l
l ATTACHMENT TO LICENSE AMENDMENT NO.165 l
l l
FACILITY OPERATING LICENSE NO. DPR-28 DOCKETllo. 50-271
)
l Replace the following pages of Appendix A Technical Specification with the attached pages.
The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
Remove Insert 2
2 3
3 4
4 164 164 164a 164a 4
I
m._. _. _ __ _ _ _ _ _ _. _ _ _ _.
e e
VYNPS l.0 DEFINITIONS v
K.
Operable - A system, subsystem, train, component or device shall L9 operable or have operability when it is capable of performing its 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, 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 funecions in its required manner.
M.
Operating Cycle - Interval between the end of one refueling outage and the end of l
the next subsequent refueling outage.
N.
Primary Containment Intecrity - 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 reactor coolant system or containment, which are not required to be open during accident conditions, are closed. Such valves may be opaned 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 1
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 und 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.
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 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.
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 po*:er level of 1593 thermal megawatts, 1
Q.
Rated Therwal Power - Rated thermal power means a steady state power level of 1593 thermal megawatts.
Amendment No. u, w, w,165 2
m --
VYNPS l.
3.0 DEFINITIONS 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 14 rated thermal power.
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 l
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 l
i l
i 2.
Run Mode - In this modJ the reactor system preFsure is equal to or greater than 800 psig and the reactor protection system is energized with APRM protection and i
j l
RBM interlocks in service.
l 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.
i T.
Refueling Outage - Refueling outage is the period of time between the shutdown of i
the unit prior to a refueling and the startup of the plant subsequent to that l
refueling. For the purpose of designating frequency of testing and surveillance, a refueling outage shall mean a regularly scheduled refueling outages however, where such outages occur within B months of the completion of the previous refueling outage, the required surveillance testing need not be performed until the next 4
regularly scheduled outage, l
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.
2.
The standby gas treatment system is operable.
3.
All reactor building automatic ventilation system isolation valves are operable l
l or are secured in the isolated position.
i 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 i
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 j
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.
1 4
3.
Shutdown means conditions as above such that the effective multiplication factor (IGer) of the core shall be less than 0.99.
l Amendment No. w, % 165 3
WNPS 1.0 DEFINITIONS l
l W.
Simulated Automatic Actuation - Simulated automatic actuation means applying a simulated I
signal to the sensor to actuate circuit in question.
X.
Transition Boilino - Transition boiling means the boiling regime 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.
Y.
Surveillance Frequency - Unless otherwise stated in these 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, component, or system prior to being required to be operable.
AA. Vital Fire Suppression Water System - The vital fire suppression water system is that part of the fire suppression system which protects those instruments, components, and systems required to perform a safe shutdown of the reactor. The vital fire suppression system includes the water supply, pumps, and distribution piping with associated sectionalizing valves, which provide immediate coverage cf the Reactor Building, Control Room Building, and Diesel Generator Rooms.
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 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.
GG.
Deleted RH.
Deleted a
Amendment No. M, M, M, M, M, M4, M4, uh 165 4
e VYNPS BASES:
3.7 (Cont'd)
In conjunction with the Mark I Containment Long-Term Program, a plant unique analysis wrs performed (see Vermont Yankee letter, dated April 27, 1984, transmitting Teledyne Engineering Services Company Reports, TR-5319-1, Revision 2, dated November 30, 1983 and TR-5319-2, Revision Of which demonstrated that all stresses in the suppression chamber structure, including shell, external supports, vent system, internal structures, and attached piping meet the structural acceptance criteria of NOREG-0661. 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 ft. will assure the integrity of the suppression chamber when subjected to post-LOCA suppression pool hydrodynamic forces.
I 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 only if the suppression pool temperature is 120*F prior to the DBA-LOCA. Maintaining a pool temperature of 100'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 the discharge exit.
Specifications have been placed on the envelope of reactor operating conditions so that the 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 action to be taken in the event a relief valve inadvertently opens or sticks open. This action would include:
(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 sand 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 the containment.
Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system. Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a loss-of-coolant accident.
Details of the isolation valves are discussed in Section 5.2 of the FSAR.
Manual primary containment isolation valvos that are required to be closed by the definition of Primary Containment Integrity may be opened intermittently under administrative controls.
These controls consist.of stationing a dedicated operator, with whom Control Room communication is immediately available, in the immediate vicinity of the valve controls.
In this way, the penetration can be rapidly isolated when a need for primary containment isolation is indicated.
Amendment No. 44, 60, 44, Ltr dtd '/1/95, 165 164
O VYNPS The purpose of the vacuum relief valves is to equalize the pressure between the drywell and suppression chamber and suppression chamber and reactor building so that the structural integrity of the containment is maintained.
Technical Specification 3.7.A.9.c is based on the assumption that the l
operability testing of the pressure suppression chamber-reactor l
building vacuum breaker, when required, will normally be performed l
during the same four hour testing interval as the pressure suppression chamber-drywell vacuum breakers in order to minimize operation with <1.7 psi, differential pressure.
1 l
l L
l 4
l Amendment No. M, H, SS, Ltr dtd "'/1/et 165 164a
_, _ _