ML20100Q400
| ML20100Q400 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 12/04/1984 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Vermont Yankee |
| Shared Package | |
| ML20100Q404 | List: |
| References | |
| DPR-28-A-084 NUDOCS 8412170100 | |
| Download: ML20100Q400 (10) | |
Text
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o UNITED STATES g
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g NUCLEAR REGULATORY COMMISSION 3 **
'Q blr E WASHINGTON, D. C. 20555
.....l VERMONT YANKEE NUCLEAR POWER CORPORATION DOCKET NO. 50-271
-VERMONT YANKEE NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 84
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License No. DPR-28 1.
The Nuclear Regulatory Commission (the Comission) has found that:
A.
The application for amendment by Vermont Yankee Nuclear Power Corporation (thelicensee)datedJanuary 23, 1984 complies with the standards and re as amended (the Act)quirements of the Atomic Energy Act of 1954, i
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 (1) 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 I
public; and E.
The issuance of this amendm6nt is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirenents have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-28 is hereby anended to read as follows:
i 8412170100 841204 PDR ADOCK 05000271 P
PDR m.-.-
.. (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 84, 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 C0ttMISSION omenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing Attach:.ient:
Char;es to the Technical Specifications Date of Issuance:
December 4,1984 e
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ATTACHMENT TO LICENSE AMEllDMEllT NO. 84 FACILITY OPERATING LICENSE 110. DPR-28 DOCKET NO. 50-271 Revise the Technical Specifications as follows:
Remove Insert 3
3 7
7 14a,,
14a 15 15 15a 15a 41 41 64 64 i-
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i VYNPS
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3.
Proteettve-eetion - An action initiated by 2.
Run Mode - In this mode the reactor system the protection system when a limit is pressure is equal to or' greater than 800 psig reached. A protective action can be at a and the reactor protection system is channel or system level, energized with APRM protection and RBM interlocks in service.
4.
Protective Function - A system protective action which results from the protective S.
Reactor Vessel Pressure - Unless otherwise action of the channels monitoring a indicated, reactor vessel pressures listed in the particular plant condition.
Technical Specifications are those measured by the reactor vessel steam space detector.
P.
Rated Neutron Flux - Rated neutron flux is the neutron flux that orresponds to a steady state T.
Refueltag Outage - Refueling outage is the period power level of.l?
thermal megawatts.
of time between the shutdown of the unit prior to a refueling and the startup of the plant Q.
Rated Thermal Power - Rated thermal power means a subsequent to that refueling. For the purpose of steady state power level of 1593 thermal megawatts.
designating frequency of testing and surveillance, a refueling outage shall mean a regularly R.
Reactor Power Operation - Reactor power operation scheduled refueling outage; however, where such is any oper ition with the mode switch in the outages occur within 8 months,of the completion of "Startup/ Rot Standby" or "Run" position with the the previous refueling outage, the required
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reactor critical and above 1% rated thermal power.
surveillance cesting need not be performed until i
the next regularly scheduled outage.
j 1.
Startup/ Hot Standby Mode - In this mode the low turbine condenser volume trip is bypassed U.
Secondary Containment Integrity - Secondary when condenser vacuum is less than' 12 inches containment integrity means that the reactor Hg and both turbine stop valves and bypass butiding is intact and the following conditions valves are closed; the low pressure and the are met:
10 percent closure main steamline isolation valve closure trips are bypassed; the reactor 1.
At least one door in each access opening is protection system is energized with IRM
- closed, neutron monitoring system trips and control rod withdrawal interlocks in service and APRM 2.
The standby gas treatment system is operable.
neutron monitoring system operable.
3.
All reactor building automatic ventilation system isolation valves are operable or are secured in the isolated position.
Amendment No. 70, 84 3
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VYZPS 1.1 SAFETY LIMIT-- -
2.1 LIMITING SAFETY SYSTEM SETTING D.
Whenever the reactor is shutdown with irradiated C.
Reactor low water level scram setting shall be at fuel in the reactor vessel, the water level shall least 127 inches above the top of the enriched not be less than 12 inches above the top of the
' fuel.
enriched fuel when it is seated in the core.
D.
Reactor low-low water level Emergency Core Cooling System (ECCS) initiation shall be at least 82.5 inches above the top of the enriched fuel.
E.
Turbine stop valve scram shall be less than or equal to 10% valve closure from full open.
F.
Turbine control valve fast closure scram shall, when operating at greater than 30% of full power, trip upon actuation of the turbine control valve fast closure relay.
C.
Main steam line isolation v'alve closure scram shall be less than or equal to 10% valve closure f rom full open.
H.
Main steam line low pressure initiation of main steam line isolation valve closure shall be at least 800 psig.
l Amendment No. 68. 84 7
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VYNPS i
i APRM Flux Scram Trip Setting (Run Mode) l j
The scram trip setting must be adjusted to ensure that the LHCR transient peak is not increas' d for any combination of e
l MFLPD and reactor core thermal power.
If the scram requires a change due to an abnormal peaking condition, it will be cecomplished by increasing the APRM gain by the ratio in Specification 2.1.A.1.a.,thus assuring a reactor scram at lever than design overpower conditions.
l I
Analyses of the limit.ing transients show that no scram adjustment is required to assure fuel cladding integrity when I
the transient is initiated from the operating limit MCPR (Specification 3.11C).
Flux Screa Trip Setting (Refuel or Startup and Hot Standby Mode) l Fcr operation in the startup mode while the reactor is at low pressure, the reducedAPRM scraa setting to 15% of rated i
power provides adequate thermal margin between the setpoint and the safety limit, 25% of the rated. (During'an outage
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when it is necessary to check refuel interlocks, the mode switch must be moved to the startup position. Since the l
APRM reduced scram may be inoperable at that time due to the disconnection of the LPRMs, it is required that the IRN j
j cerca and the SRM scram in noncoincidence be in effect. This will ensure that adequate thermal margin is maintained j
between the setpoint and the safety limit.) The margin is adequate to accommodate anticipated maneuvers associated 4
with station startup. Effects of increasing pressure at zero or low void content are minor, cold water from sources i
cvailable during startup is not auch colder than that already in the system, temperature coefficients are small, and control rod patterns are constrained to be uniform by operating procedures backed up by the ro'd worth minimizer.
I Worth of individual rode is very low in a uniform rod pattern. Thus, of all possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power rise. Because the flux distribution 4
l escociated with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to l
j change power by a significant percentage of. rated power, the rate of power rise is very slow. Generally, the heat 1
j flux is in near equilibrium with the fissiott rate. In an assumed uniform rod withdrawal approach to the scram level, j
the rate of power rise is no more than 5% of rated power per minute, and the APRM system would be more than adequate to assure a scram before the power could exceed the safety limit. The reduced APRM scram remains active until the 1
moda switch is placed in the RUN position. This switch can occur when reactor pressure is greater than 800 psig.
The IRM system consists of 6 chambers, 3 in each of the reactor protection system logic channels. The IRM is_a 5-dreade instrument, which covers the range of power level between that covered by the SRM and the APRM. The 5 decades are covered by the IRM by means of a range switch and the 5 decades are broken down into 10 ranges, each being 4
i one-half of a decade in size. The IRM scram trip setting of 120/125 of full scale is active in each range of the
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IRM. For example, if the instrument were on range 1, the scram setting would be a 120/125 of full scale for that t
j range; likewise, if the instrument were on range 5, the scram would be 120/125 of full scale on that range. Thus, as i
j the IRM is ranged up to accommodate the increase in power level, the scram trip setting is also ranged up.
The most significant sources of reactivity change during the power increase are due to control rod withdrawal. For in sequence l
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centrol rod withdrawal, the rate of change of power is slow enough due to the physical limitation of withdrawing l
control rods, that heat flux is in equilibrium with the neutron flux and an IRM scram would result in a reactor f
shutdown well before any safety limit is exceeded.
i f
sendment No. 78. 84 14a 1
VYNpS 2.1 (cont.)
D.
Reactor I.dwNter Level ECCS Initiation Trip Point The core standby cooling subsystems are designed to provide sufficient cooling to the core to dissipate the _
energy associated with the loss-of-coolant accident and to limit fuel clad temperature to well below the clad melting temperature..and to limit clad metal-water reaction to less than 1%, to assure that core geometry remains intact.
The design of the ECCS components to meet the above criteria was dependent on three previously set parameters:
the maximum break size, the low water level scram setpoint, and the ECCS initiation setpoint.. To lower the ECCS initiation setpoint would now prevent the ECCS components from meeting their design criteria. To raise the ECCS initiation setpoint would be in a safe direction, but it would reduce the margin established to prevent actuation of the ECCS during normal operation or during normally expected transients.
E.
Terbine Stop Valve Closure Scram Trip Setting The turbine stop valve closure scram trip anticipates the pressure, neutron flux and heat flux increase that could result from rapid closare of the turbine stop valves. With a scram trip setting of <10% of valve closure from full open, the resultant increase in surface heat flux is limited such that MCPR remains above the fuel cladding integrity safety limit even during the worst case transient that assumes the, turbine bypass is closed.
This scram is bypassed when turbine steam flow is below 30% of rated, as measured by turbine first stage pressu re.
F.
Turbine Control Valve Fast Closure Scram The control valve fast closure scram is provided to limit the rapid increase in pressure and neutron flux resulting from fast closure of the turbine control valves due to a load rejection coincident with failure of the bypass system. This transient is less severe than the turbine stop valve closure with failure of the bypass valves and therefore adequate margin exists.
C.
Main Steam Line Isolation Valve Closure Scram The isolation valve closure scram anticipates the pressure and flux transients which occur during normal or inadvertent isolation valve closure. - With the scram setpoint at 10% of valve closure, there is n'o increase in neutron flux.
H.
Reactor Coolant Low Pressure Initiation of Main Steam Isolation Valve Closure The low pressure isolation of the main steam lines at 800 psig is provided to give protection against rapid reactor depressurization and the resulting rapid cooldown of the vessel. Advantage is taken of the scram feature which occurs when the main steam line isolation valves are closed, to provide the reactor shutdown so that high power operation' at low reactor pressure does not Amendment No. 25, 84' 15
4 VYNPS.
2.1 (cont.)
occur. Opdrawof the reactor.at pressures lower than 800 psig requires that the reactor mode switch be in the startup position where protection of the fuel cladding integrity safety limit is provided by the IRM high neutron flux scraa.
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Thu2, the combination of main steam line low pressure isolation and' isolation val'e closure scram assures the v
cvailable of neutron scram protection over the entire range of applicability of the fuel cladding integrity safety licit.
Amendment No. 84 15a
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TABLE 3.2.2 PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION Minimum Number of Operable Instrument Required Action When Minimum Channels per Trip Conditions for Operation are System Trip Function Trip Setting Not Satisfied (Note 2) 2 Low-Low Reactor Vessel
> 82.5" above the A
Water Level top of enriched fuel 2 of 4 in each of High Main Steam Line f,2120F B
2 channels Area Temperature 2/ steam line High Main Steam Line
$120% of rated flow B
Flow 2/(Note 1)'
Low Main Steam Line
>800 psig B
Pressure 2/(Note 6)
High Main Stean Line
<40% of rated flow B
Flow 2
Low Reactor Vessel Same as Reactor Protection A
Water Level Systen 2
High Main Steam Line
<3 X Background at rated B
Radiation (7) (8) power (9) i 2
High Drywell Pressure Same as Reactor Protection A
System 2/(Note 10)
Condenser Low Vacuum
> 12" Hg absolute A
1 Trip System fogic A
Amendment No. 68, 84 41
-3.2 (Continued) ---- E.
i High radiation monitors in the main steam line tunnel have been provided to detect gross fuel failure resulting f rom 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 anid main steam line isolation valve i
cicoure, fission product release is limited so that 10CFR100 limits are not exceeded for the control rod drop accident i
cnd 10CFR20 limits are not exceeded for gross fuel failure during reactor operations. With an alarm setting of 1.5 j
times normal background, the operator is alerted to possible gross fuel failure or abnormal fission product releases i
f rom failed fuel due to transient reactor operation.
Prassure instrumentation is provided which trips when main steam line pressure drops below 800 psig. A trip of this j
instrumentation results in closure of C: ap 1 isolation valves. In the refuel, shutdown, and startup modes, this trip j
function is provided when main steam line flow exceeds 40% of rated capacity. This function is provided primarily to j
provide protection against a pressure regulator malfunction which would cause the control and/or bypass valves to cpen, resulting in a rapid depressurization and cooldown of the reactor vessel. The 800 psig trip setpoint limits the l
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 j
f rom the primary coolant through condenser. The setpoint of 12 inches of mercury ebsolute 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 values.
The HPC1 and/or RCIC high flow, steam supply pressure, and temperature instrumentation is provided to detect a break in the HPC1 and/or RCIC piping. Tripping of this instrumentation results in actuation of HPC1 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 l
system from inadvertently isolating due to pressure spikes which may occur on startup. The trip settings are such l
that core uncovering is prevented and fission product release is within limits, i
The instrumentation which initiates ECCS action is arranged in a dual chsnnel system. As for other vital instrumentation arranged in this fashion, the specification preserves the effectiveness of the system even during l
periods when maintenance or testing is being performed. Permanently installed circuits and equipment may be used to I
trip instrument channels. Ir the non-fail safe systems which require energizing the circuitry, tripping an instrument l
channel may take the form of aviding the required relay function by use of permanently installed circuits. This is I
accomplished in some cases by closing logic circuits with the aid of the permanently inatalled test jacks or other circuitry which would be installed for this purpose.
, Amendment No. 69,84 64 t
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