ML20079Q650
| ML20079Q650 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 01/23/1984 |
| From: | VERMONT YANKEE NUCLEAR POWER CORP. |
| To: | |
| Shared Package | |
| ML20079Q644 | List: |
| References | |
| NUDOCS 8402010285 | |
| Download: ML20079Q650 (7) | |
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- P- 1 57 3.. ' Protective Actioc( ;A3 action. initiated by 4 ~ 7 y '2, sRun Mode .In this mode 7the reactor system the protect (on syst.emlwhe 6 a' limit is -
. '~e d ; pressure is equal to or greater than 800 psig ~
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and the reactor prote3 tion system is . 2 -
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% 5 4. - Protective Function - A system protective W e%;Os 1 '
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~; ' action which results from the protectiv D 3' '_ Sb React 5r2 W asel' Pressure - Unless.otherwise
\ _,,' actioa of the channels moni;orire a .T "~ 9 '
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, '* g M t indicated, reactor vessel pressures ' listed inbythe y\ y y, ,; mTechnical Specifications are 'those measured the;
.part{cular plant condition. .
, N .g 2s< reactor vessel steam space detector.
QPg RgdjNdet?ron~ Dax - Rated neutron flux is 'the-
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% s neutrott fluxsthar acorrecconds to a steady septej , T ... REfuell'ag Outage .- Refueling outlage is the period
%' irdrer*%l'elel'of 13,93 thermal ' megawatts. '*
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s of time between the' shutdown of the' unit prior to 4
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y' - .a refueling and the startup of the plant Q. s' Rated' Thermal Power " Rated thermal power means a ., subsequent to that refueling. For the purpose of steaf.y state. power. level of 1593 thermal negr.vatts..,.. designating frequency of testing and surveillance,
- a. refueling outage shall mean a regularly R! Reactor Power Operation , Reactor powe,r operation,- scheduled refueling outage; however, where such is any operation with,the modelsititch in the outages occur within 8 months of the completion of f "
"Startup/ Hot Standby" or "Run~'pos,itidn'ulth the' the previous refueling. outage, the required' reactor criiical and above 1% rated' thermal power. surveillance testing need not be performed until the next regularly scheduled outage, l.. 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 building is intact and the following conditions valves are closed; the Iow 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 mystem is operable.
neutron monitoring system operable.
- 3. All reactor building automatic ventilation system isolation valves are operable or are secured in the isolated position.
Aatndment No. 70 3 a
VYNPS 1.1 - SLFETY _ LIMIT - 2.1 LIMITING SAFETY SYSTEM SETTING D. Whenever the reactor is shutdewn 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 scraa shall be less than or equal to 10% valve closure.from full open.
i 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.
G. Main steam line isolation valve closure scram-shall be less than or equal to 10% valve closure l from full open.
H. Main steam line low pressure initiation of mai1 j steam line isolation valve closure shall be at j least 600 psig. l l
l Amendment No. 68 7
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a VYNPS -
'APdM Flux Scram Trip Settina (Run Mode).
.The scrani trip setting must be adjusted.to ensure that the LHGR transient peak is not increased for any combination of 4-MPLPD and reactor core thermal power. If.the scram requires a change due to an abnormal peaking condition, it will.be secomplished by increasing the APRM gain by the ratio in Specification 2.1.A.l.a. thus assuring a reactor scram at-icwer than design overpower conditions.
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.' Analyses 'of the limiting transients show that no scram edjustment is required to assure fuel cladding integrity when
'the transientLisLinitiated from.the operating limit MCPR (Specification 3.11C).
Flux Scras' Trip Setting (Refuel or Startup'and Hot Standby Mode) l.
'For operation in thk startup-mode while the' reactor is at low pressure, the reduced APRM scram setting to 15% of rated
[ puwer provides adequate thermal margin between the setpoint and the safety limit, 25% of the rated.. (During an outage whsn it is'necessary to check refuel interlocks, the mode switch must be moved to the startup position. Since the APRM reduced . scram may. be inoperable at that. time due to the disconnection of the LPRMs, it is required that the IRM ,
scram and the.SRM scram in noncoincidence be in effect. This will ensure that adequate thermal usrgin is maintained I
between the setpoint and the safety limit.) The margin is adequate to accommodate anticipated maneuvers associated-with station startup. . Effects of increasing pressure at zero or low void content are minor, cold water from sources .!
available'during startup is t:ot auch colder than that already in the system, temperature coefficients are small, and centrol rod patterns are constrained to be uniform by operating procedures backed up by the rod worth minimizer.
' Worth of ind!viduel rods is very low in a uniform rod pattern. Thus, of all possible sources of reactivity input,
] uniform control rod wi:hdrawal is_the most probable cause of significant power rise. Because the flux distribution F associated with uniform rod withdrawals does not involve high local peaks, and because'several rods must be moved to change power by a significant percentageLof rated. power, the rate of power rise is very slow. Generally, the heat flux is in near equilibrium with the fission-rate. In an assumed aniform 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 uatil the l lmodeswitchisplaced-intheRUNposition. This switch can occur when reactor pressure is greater than 800 psig. ,
j Tha IRM system consists of 6 chambers, 3 in each of the reactor protection system logic channels. The IRM is a i 5-decade 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
, ons-half of a decade in size. The IRM scram trip setting of 120/125 of full scale is active in each range of the j IRM. For example, if the instrument were on range 1, the scram setting would'be a 120/125 of full scale for that j range; likewise, if the. instrument were on range 5, the scram would be 120/125 of full scale on that range. Thus, as ths IRM is-ranged up to accommodate the-increase in power level, the scram trip setting is also ranged up. The most i significant sources of reactivity change during the power increase are due to control rod withdrawal. For in sequence l control rod withdrawal, the rate of change of power is slow enough due to the physical limitation of withdrawing 1 4 . control rods, that heat flux-is in equilibrium with the neutron flux and an IRM scram would result in a reactor i ohutdown well before any safety limit is exceeded.
4 Amendment No. 78 14a i , _ _ _ _ _. - - -- .
7YNPS 2.1 (cont.)-
D. Reactor Low Water Level ECCS Initiation Trip Point' *
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-The core standby' cooling subsystems are designed to . provide suf ficient cooling to the core. to dissipate; the .
energy associated with the loss-of-coolant accident ~and to limit fuel clad temperature to well below e.he 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 3 Jnitiation 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. Turbine Stop Valve Closure Scram Trip Setting The turbine stop valve closure scram trip anticipates the preseure, neutron flux and heat flux increase that could result from rapid closure 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' auch 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 pressure.
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 r f 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.
G. Main Steam Line Isolation Valve Closure Scram The isolation valve closure scram anticipates the pressure and flux transients which occur during normal nr inadvertent isolation valve closure. With the scram setpoint at 10% of valve closure, there is no 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 15
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-- VYNPS .
2.1-(cont.).
occur. Operation of the' reactor at pressures lower than'800 psig requires that the reactor mode switch be in the startup position where protection of the-ft.el cladding integrity safety limit is provided by the IRM high - ,
neutron flux' scram.
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- Thus, the combination of main steam line low pressure isolation and=.ieoistian valve closure scram assures the evallable of neutron scram protection over the entire range.of applicability of the fuel cladding integrity safety.
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VYNPS 4
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 Vescel > 82.5" above the -A Water Level top of enriched fuel 2 of 4 in each of High Main Steam Line i 2120F B 2 channels Area Temperature 2/ steam line High Main Steam Line 1120%,of rated flow B Flow 2/(Note 1) Low Main Steam Line >800 psig B Pressure 2/(Note 6) High Main Steam Line 140% of rated flow B j Flow 2 Low Reactor Vessel Same es Reactor Protection A Water Level System 2 High Main Steam Line' 1." X Background at rated B Radiation (7) (8) power (9) 2 High.Drywell Pressure Same as Reactor Protection A System 2/(Note 10) Condenser Low Vacuum > 12" Hg absolute A 1 Trip System-Logic. --
A i
Amendment No. 68 41 4
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. VYNPS 3.2 ; (Continued) ~ -
High radiation" monitors 'in the main Lateam 'line tunnel havel been provided (to detect gross fuel- failure resulting f rom a csntrol' rod drop. accident._ This instrumentation causes closure of Group 1 valves,'the only valves require'd to close-fer ' 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 reactorfoperations. With an alarm setting of 1.5..
times . normal 1 background. the operator is alerted ,to possible gross fuel failure or abnormal fission product- releases
~
. f rem. failed fuel' due to transient reactor operation.
. Pressure instrumentation is provided which trips when' main' steam line pressure drops below 800 psig. A-trip offthis 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 againstLa pressure regulator malfunction which would cause the control and/or bypass valves to cpan, . resulting in a rapid depressurization and cooldown of the reactor vessel. The 800 psig trip setpoint limits the
' dspressurization such that no. excessive . vessel thermal stress occurs as a result of a pressure . regulator' malfunction.
This setpoint wcs 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
- f rom the primary coolant through condenser. The setpoint of 12 inches of mercury absolute was aelected to provide suf ficient margin' to assure retention capability in the condenser when gas flow is stopped and sufficient margin below normal oporating values.
The HPCI and/or RCI' high flow, steam supply pressure, and temperature instrumentation is provided 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 PCIC steam flow trip logic to prevent the system f rom inadvertently isolating due to pressure spikes which may occur en startup. The trip settings are such that core uncovering is prevented and fission product release is within limits.
Thi instrumentation which' initiates ECCS action is arranged in a dual channel system. As for other vital instrumentation arranged in this fashion, the specification preserves the ' effectiveness of the system even during periods when maintenance or testing is being performed. Permanently installed circuits'and equipment may be used to trip instrument channels. In the non-fail 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 installad test jacks or other circuitry which would be installed for this purpose.
Amendment No. 69 64 i