ML20210D337
| ML20210D337 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 07/20/1999 |
| From: | VERMONT YANKEE NUCLEAR POWER CORP. |
| To: | |
| Shared Package | |
| ML20210D326 | List: |
| References | |
| NUDOCS 9907270203 | |
| Download: ML20210D337 (20) | |
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VenuoNT YAMEE Nect. tan Powut Coswouxuos Docket No. 50-271 BVY 99-82 Attachment 3 3 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 218 Enhancements to Support Implementation ofIncreased Core Flow 1
Marked-up Version of the Current Technical Specifications and Bases 9907270203 990720 PDR ADOCK 05000271 p PDR
VYNPS FIGURE 2.1-1 APPM FLOW REFEPENCE SCPN4 AND APRN ROD BLOCK SETTINGS v
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t the APRM Serwn and Rod Block settings are odjusted occordirq to Tebical Speelfications 2.1.A.to and 2.tB.1 Amendment Nc. H. 94
}&ggs,: 2.1 (Cont'd)
In order to ensure that the IRM provided adequate protection against the single rod withdrawal error, a range of rod withdrawal accidents was analyzed. This analysis included starting the accident at various power levels. The most severe case involves an initial condition in which the reactor is just suberitical and the IRM system is not yet on scale. This condition exists at quarter rod density.
Additional conservatism was taken in this analysis by assuming that the IRM channel closest to the withdrawn rod is bypassed. The results of this analysis show that the reactor is scrammed and peak power 10mited to one percent of rated power, thus maintaining MCPR above the fuel cladding integrity safety limit. Based on the above analysis, the IRM provides protection against local control rod withdrawal errors and continuous withdrawal of control rods in sequence.
B. APRM Rod Block Trio Settinc
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u, A fReactorpo%recirculationflowrate.
I varying t e e APRM system provides a contrcl od block to prevent red wit rawal beyond a given point ..e y"C"4. { the fue cladding integrity safety setti ,
mit. This rod block trip which is automatically va led with recirculatica loop flw
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race prevents an increase in th reactor power level to excess 4e val es due to control rod withd- wal. The flow variable trip etting pr vides substantial margin fr fuel damage, assuming a see y-state o eration at the trip setting over the entire recirculatio flow ange. The margin to the S ety Limit increases as the fl decreases for the specifie trip setting versus flow rol ionship, therefore the worst case PR which could occur during y eady-state operation is at 108% of ated thennal power because of/ he t APRM rod block trip setting. actual power distribution i fthe core is established by specif d control rod sequences and e APRM monitored ;
continuously by the core LPRM system. As with scram trip ~
setting, the APRM r block trip setting must be djusted downward if the ratio of MFLPD o FRP exceeds the specifie value. If the APRM rod block require a change due to abnormal p king conditions, it will be accompi had by ine--asing the APRM kE in by the ratio in specification .1B, t(us < ., ting a rod blodk at lower than design overoower con .tions/ As with the APRM flux scram trip setting, the APRM rod block trip setting is reduced for single recirculation loop operation in acecrdance with the analysis presented in NEDo-30060, February, 1983. This adjustment accounts for the difference between the single loop and two loop drive ,.flow me at the same core .,m#.
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C. Reactor Low Water Level Scram The reactor low water level scram is set at a point which will prevent reactor operation with the steam separatorsInuncovered, additicn, the thus limiting carry-under to the recirculation loops.
safety limit is based on a water level below the scram point and therefore this setting is provided.
D. Reactor Lew Water Level ECOS Initiation Trio 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 16 Amendment No. te, GG, en, 64, 94
Insert 1, Bases page 16 The purpose of the APRM rod block function is to avoid conditions that would require Reactor Protection System action if allowed to proceed. The APRM upscale rod block alarm setting is selected to initiate a rod block before the APRM high neutron flux scram setting is reached. The APRM upscale rod block trip setpoint is varied as a function of reactor recirculation flow. The slope of the rod block trip response curve with recirculation flow is adjustable to allow tracking of the required trip setpoint with recirculation flow changes. This provides an effective rod block if core average power is increased above the power level specified at any flow rate.
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VYNPS 3 333,: 3.2 (Cont'd) setpoint limits the 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 from the primary coolant through condenser. The setpoint of 12 inches of mercury absolute was selected to provide sufficient margin to assure retention capability in the condenser when gas flow is stopped and sufficient margin below normal operating values.
%e HPCI and/or RCIC 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 RCIC steam flow trip logic to prevent the system from inadvertently isolating due to pressure spikes which may occur on startup. The trip settings are such that core uncovering is prevented and fission product release is within limits.
The instrumentation which initiates ECCS action is arranged in a dual channel system. Permanently installed circuits and equipment may be used to trip instrument channels. In the nontail 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 installed test jacks or other 1 circuitry which would be installed for this purpose.
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Ny*h w'4 rod block f tions are provi ed to pfevent ex saive control !
fThecent _ awal so that PR does not de ease below the el cladding 6#41 rod wit intog cy safety limi' The trip logi for this functi is 1 out of n-e.g. any trip on on of the six AP , six IRMs or f r SRMs will res e in rod block, minimum inst nt channel requ ements for the RM be redued by ne for a short eriod of time t allow for I intenance, tes ing or calibra on. The RBM is operational de and ,"
aid only and g not needed for od withdrawal. 1 For single roeirculation loop operation, the RBM trip setting is reduced J in accordance with the analysis presented in NEDO-30060, February 1983. 4 This adjustment accounts for the difference between the single loop and two loop drive flow at the same core flow, and ensures that the margin of safety is not reduced during single loop operation, pJ *~ .s AP rod block t,fip is flod ferench rev nIfIcD N ow. The AP l
l T - reduc on in MCPR pecially du .ng operatio the at reduced M 'prov* des gross e a protectio , i.e., limi gross ore power t .y c.,A- 3 in *ase from w~ hdrawal of ontrol rodsgn the norma withdrawal a Pha s ence. Th trips are s so that MC in mainta ed crogter t_
uel_c1=dM n intaeity fety limit / For single recirculation loop operation, the APRM rod block trip setting is reduced in accordance with the analysis presented in NEDE-30060, February 1983. This adjustment accounts for the dif ference between the sinale lo<p and two-loco drive ;
flow at the same core flow rana ensures ~that thi marcih of safatv in a9~
@1.4 1.. b.w .iuvly lwwy operanon. . - -
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- The IRH rod block function provides local as well as gross core J protection. The scaling arrangement is such that trip setting is less than a factor of 10 above the indicated level. Analysis of the worst-case accident results in rod block action before MCPR approaches the fuel cladding integrity safety limit, M, 94 77 Amendment No. 4, M, W
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l l Insert 2, Bases page 77 The trip logic for the nuclear instrumentation control rod block logic is 1 out of n; i.e., any trip on one of the six APRMs, six IRMs or four SRMs will result in a rod block. The minimum instrument channel requirements for the IRMs may be reduced by one for a short period of time to allow for maintenance, testing or calibration. The RBM is credited in the Continuous Rod Withdrawal During Power Range Operation transient for preventing excessive control rod withdrawal before the fuel cladding integrity safety limit (MCPR) or the fuel rod mechanical overpower limits are exceeded. The RBM upper limit is clamped to provide protection at greater than 100% rated core flow. The clamped value is cycle specific; therefore, it is located in the l
Core Operating Limits Report.
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! Insert 3, Bases page 77 -
The purpose of the APRM rod block function is to avoid conditions that would require Reactor Protection System action if allowed to proceed. The APRM upscale rod block alarm setting is selected to initiate a rod block before the APRM high neutron flux scram setting is reached. The !
APRM upscale rod block trip setpoint is varied as a function of reactor recirculation flow. The f
slope of the rod block trip response curve with recirculation flow is adjustable to allow tracking
{
of the required trip setpoint with recirculation flow changes. This provides an effective rod block if core average power is increased above the power level specified at any flow rate.
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vuuuiN1 Ymu:. Necu:.ut Powat Cosmon u tos Docket No. 50-271 !
l BVY 99-82 l
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Attachment 4 Vermont Yankee Nuclear Power Station l
l Proposed Technical Specification Change No. 218 Enhancements to Support Implementation ofIncreased Core Flow Retyped Technical Specification and Bases Pages I
1 1
, FIGURE 2.1-1 )
APRM FLOW REi'c.RENCE SCRAM AND APRM ROD BLOCK SETTINGE l
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For single loop operation, the APRM Scram and Rod Block settings are adjusted according to Technical Specifications 2.1.A.1.a and 2.1.B.1 Amendment No. +0, 94 11
4 .
VYNPS BASES: 2.1 (Cont'd)
In order to ensure that the IRM provided adequate protection against the single rod withdrawal error, a range of rod withdrawal accidents was analyzed. This analysis included starting the accident at various power levels. The most severe case involves an initial condition in which the reactor is just suberitical and the IRM system is not yet on scale. This condition exists at quarter rod density. Additional conservatism was taken in this analysis by assuming that the IRM channel closest to the withdrawn rod is bypassed. The results of this analysis show that the reactor is scrammed and peak power limited to one percent of zated power, thus maintaining MCPR above the fuel cladding integrity safety limit.
Based on the above analysis, the IRM provides protection against ,
local control rod withdrawal errors and continuous withdrawal of i control rods in sequence.
B. APRM Rod Block Trip Setting The purpose of the APRM rod block function is to avoid conditions that would require Reactor Protection System action if allowed to proceed. The APRM upscale rod block alarm setting is selected to initiate a rod block before the APRM high neutron flux scram setting is reached. The APRM upscale rod block trip setpoint is varied as a function of reactor recirculation flow.
The slope of the rod block trip response curve with recirculation flow is adjustable to allow tracking of the required trip setpoint with recirculation flow changes. This provides an effective rod block if core average power is increased above the power level specified at any flow rate.
As with the APRM flux scram trip setting, the APRM rod block trip setting is reduced for single recirculation loop operation in accordance with the analysis presented in NEDO-30060, February, 1983. This adjustment accounts for the difference between the single loop and two loop drive flow at the same core flow.
C. Reactor Low Water Level Scram The reactor low water level scram is set at a point which will prevent reactor operation with the steam separators uncovered, thus limiting carry-under to the recirculation loops. In addition, the safety limit is based on a v'ter level below the scram point and therefore this setting is provided.
_. Reactor Low Water 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.
Amenament No. M, M, 44, 4, G4 16
i i VYNPS
(
BASES: 2.1 (Cont'd)
E. Turbine Stop Valve Clogure Scram Trip Setting The turbine stop valve closure scram trip anticipates the pressure, 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 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 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 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.
G. 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 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 i 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 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 fuel cladding integrity safety limit is provided by the IRM high neutron flux scram.
Thus, the combination of main steam line low pressure isolation and isolation valve closure scram assures the availability of neutron scram l protection over the entire range of applicability of the fuel cladding integrity safety limit.
Amendment No. 44, G&, G4, 164 17
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a VYNPS BASES:
3.2
- PROTECTIVE INSTRUMENTATION In addition to reactor protection instrumentation which initiates a reactor j scram, station protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the reactor operator's ability to control, or terminate a single operator error before it results in serious consequences. This set of Specifications provides the limiting conditions of operation for the primary system isolation function and initiation of the core standby cooling and standby gas treatment systems. The objectives of the Specifications are (i) to assure the effectiveness of any component of such systems even during j periods when portions of such systems are out of service for maintenance, testing, or calibration, and (ii) to prescribe the trip settings required to ;
assure adequate performance. This set of Specifications also provides the l
limiting conditions of operation for the control rod block system and surveillance instrumentation.
Isolation valves (Note 1) are installed in those lines that penetrate the primary containment and must be isolated during a loss-of-coolant accident so that the radiation dose limits are not exceeded during an accident condition. Actuation of these valves is initiated by protective j instrumentation shown in Table 3.2.2 which senses the conditions for which !
isolation is required. Such instrumentation must be available whenever l primary containment integrity is required. The objective is to isolate the l primary containment : i that the limits of 10 CFR 100 are not exceeded during )
an accident. The ob 'tive of the low turbine condenser vacuum trip is to i minimize the radioa ve effluent releases to as low as practical in case of a main condenser fai.;re. Subsequent releases would continue until operator action was taken tc solate the main condenser unless the main steam line j isolation valves were closed automatically on low condenser vacuum. The manual bypass is required to permit initial startup of the reactor during low power operation.
The instrumentation which initiates primary system isolation is connected in a dual channel arrangement-. Thus, the discussion given in the bases for Specification 3.1 is applicable here.
The low reactor water level instrumentation is set to trip when reactor water level is 127" above the top of the enriched fuel. This trip initiates closure of Group 2 and 3 primary containment isolation valves. For a trip setting of 127" above the top of the enriched fuel, the valves will be closed before perforation of the clad occurs even for the maximum break and, therefore, the setting is adequate.
The top of the enriched fuel (351.5" from vessel bottom) is designated as a common reference level for all reactor water level instrumentation. The intent is to minimize the potential for operator confusion which may result from different scale references.
The low-low 2 actor water level instrumentation is set to trip when reactor
- ater level is 82.5" H2 O indicated on the reactor water level instrumentation above the top of the enriched fuel. This trip initiates closure of the Group 1 primary containment isolation valves and also activates the ECCS and RCIC System and starts the standby diesel generator system. This trip setting level was chosen to be low enough to prevent spurious operation, but high enough to initiate ECCS operation and primary system isolation so that no melting of the fuel cladding will occur, and so that post-accident cooling can be accomplished and the limits of 10CER100 will not be violated.
Note 1 - Isolation valves are grouped as listed in Table 3.7.1.
Amendment No. GB, 68, 444 75 l
= ,
VYNPS BASES: 3.2 (Cont'd)
For the complete circumferential break of 28-inch recirculation line and with the trip setting given above, ECCS initiation and primary system isolation are initiated in time to meet the above criteria. The instrumentation also covers the full range of spectrum breaks and meets the above criteria.
The high drywell pressure instrumentation is a backup to the water level instrumentation, and in addition to initiating ECCS, it causes isolation of Group 2, 3, and 4 isolation valves. For the complete circumferential break discussed above, this instrumentation will initiate ECCS operation at about the same time as the low-low water level instrumentation, thus, the results given above are applicable here also. Group 2 isolation valves include the drywell vent, purge, and sump isolation valves. High drywell pressure activates only these valves because high drywell pressure could occur as the result of nonsafety-related causes such as not purging the drywell air during startup. Total system isolation is not desirable for these conditions and only the valves in Group 2 are required to close. The water level instrumentation initiates protection for the full spectrum of loss-of-coolant accidents and causes a trip of all primary system isolation valves.
Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident. In addition to monitoring steam flow, instrumentation is provided which causes a trip of Group 1 isolation valves.
The primary function of the instrumentation is to detect a break in the main steam line, thus only Group 1 velves are closed. For the worst case accident, main steam line break outside the drywell, this trip setting of 140 percent of rated steam flow in conjunction with the flow limiters and main steam line valve closure limit the mass inventory loss such that fuel is not uncovered, cladding temperatures remain less than 1295'F and release of radioactivity to the environs is well below 10CFR100.
Temperature monitoring instrumentation is provided in the main steam line tunnel to detect leaks in this area. Trips are provided on this instrumentation and when exceeded cause closure of Group 1 isolation valves.
Its setting of ambient plus 95'F is low enough to detect leaks of the order of 5 to 10 gpm; thus, it is capable of covering the entire spectrum of breaks. For large breaks, it is a backup to high steam flow instrumentation discussed above, and for small breaks, with the resultant small release of radioactivity, gives isolation befora the limits of 10CFR100 are exceeded.
High radiation monitors in the main steam line tunnel have been provided to detect gross fuel failure resulting from 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 and main steam line isolation valve closure, fission product release is limited so that 10CFR100 limits are not exceeded for the control I rod drop accident. With an alarm setting of 1.5 times normal background, the operator is alerted to possible gross fuel failure or abnormal fission product releases from failed fuel due to transient reactor operation.
Pressure instrumentation is provided which trips when main steam line pressure drops below 800 psig. A trip of this instrumentation results in closure of Group 1 isolation valves. In the refuel, snutdown, 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 against a pressure regulator malfunction which would cause the Amendment No. 36, 68, &4, &6, 151 76
J VYNPS BASES: 3.2 (Cont'd) control and/or bypass valves to open, resulting in a rapid depressurization and cooldown of the reactor vessel. The 800 psig trip setpoint limits the 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 from the primary coolant through condenser. The setpoint of 12 inches of mercury absolute was selected to provide sufficient margin to assure retention capability in the condenser when gas flow is stopped and sufficient margin below normal operating values.
The HPCI and/or RCIC 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 RCIC steam flow trip logic to prevent the system from inadvertently isolating due to pressure spikes which may occur on startup.
The trip settings are such that core uncovering is prevented and fission product release is within limits.
The instrumentation which initiates ECCS action is srranged in a dual channel system. Permanently installed circuits and equipment may be used to trip instrument channels. In the nonfail 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 installed test jacks or other circuitry which would be installed for this purpose.
The trip logic for the nuclear instrumentation control rod block logic is 1 out of n; i.e., any trip on one of the six APRMs, six IRMs or four SRMs will result in a rod block. The minimum instrument channel requirements for the IRMs may be reduced by one for a short period of time to allow for maintenance, testing or calibration. The RBM is credited in the Continuous Rod Withdrawal During Power Range Operation transient for preventing excessive control rod withdrawal before the fuel cladding integrity safety limit (MCPR) or the fuel rod mechanical overpower limits are exceeded. The RBM upper limit is clamped to provide protection at greater than 100% rated core flow. The clamped value is cycle specific; therefore, it is located in the Core Operating Limits Report.
For single recirculation loop operation, the RBM trip setting is reduced in accordance with the analysis presented in NEDO-30060, February 1983. This adjustment accounts for the difference between the single loop and two loop drive flow at the same core flow, and ensures that the margin of safety is not reduced during single loop operation.
The purpose of the APRM rod block function is to avoid conditions that would require Reactor Protection System action if allowed to proceed. The APRM upscale rod block alarm setting is selected to initiate a rod block before the APRM high neutron flux scram setting is reached. The APRM upscale rod block trip setpoint is varied as a function of reactor recirculation flow.
The slope of the rod block trip response curve with recirculation flow is adjustable to allow tracking of the required trip setpoint with Amendment No. 9, G4, 64, 64, G4 77
VYNPS BASES: 3.2 (Cont'd) l ' recirculation flow changes. This provides an effective rod block if core I average power is increased above the power level specified at any flow rate.
For single recirculation loop operation, the APRM rod block trip setting is reduced in accordance with the analysis presented in NEDE-30060, February 1983. This adjustment accounts for the difference between the single loop and two-loop drive flow at the same core flow.
The IRM rod block function provides local as well as gross core protection.
The scaling arrangement is such that trip setting is less than a factor of 10 above the indicated level. Analysis of the worst-case accident results in rod block action before MCPR approaches the fuel cladding integrity safety limit.
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 fanction 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 and minimize spurious eperation. 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 with inhibit switches to manually prevent automatic i 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 normal 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 release of radioactivity from the main condenser. During an accident, 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.
Amendment No. 9, M, M, M, M3, 4M 78
l VYNPS j
BASES: ,3 . 2 (Cont'd)
Post-accident instrumentation parameters for Containment Pressure, Torus Water Level, Containment Hydrogen / Oxygen Monitor, and Containment High-Range Radiation Monitor, are redundant, environmentally and seismically 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 following an accident as required by NUREG-0737 by ensuring continuous on-scale indication of the following: 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 I normal water level of the torus pool); containment hydrogen / oxygen concentrations (0 to 30% hydrogen and 0 to 25% oxygen); and containment radiation in the 1 R/hr to 10' R/hr gamma.
The Degraded Grid Protective System has been installed to assure that I safety-related electrical equipment will not be subjected to sustained l degraded voltage. This system incorporates voltage relays on 4160 Volt Emergency Buses 3 and 4 which are set to actuate at the minimum voltage required to prevent damage of safety-related equipment.
If Degraded Grid conditions exist for 10 seconds, either relay will actuate l an alarm to alert operators of this condition. Based upon an assessment of these conditions the operator may choose to manually disconnect the off-site power. In addition, if an ESF signal is initiated in conjunction with low voltage below the relay setpoint for 10 seconds, the off-site power will be j automatically disconnected.
The Reactor Core Isolation Cooling (RCIC) System provides makeup water to the reactor vessel during shutdown and isolation to supplement or replace the normal makeup sources without the use of the Emergency Core Cooling l Systems. The RCIC System is initiated automatically upon receipt of a j l reactor vessel low-low water level signal. Reactor vessel high water level
{
i signal results in shutdown of the RCIC System. However, the system will j The RCIC restart on a subsequent reactor vessel low-low water level signal.
i System is normally lined up to take suction from the condensate storage tank. Suction will automatically switch over from the condensate storage tank to the suppression pool on low condensate storage tank level.
Upon receipt of a LOCA initiation signal, if normal AC power is available, all RHR pumps and both Core Spray pumps start simultaneously with no intentional time delay. If normal AC power is not available, RHR pumps A and D start immediately on restoration of power, RHR pumps B and C start within 3 to 5 seconds of restoration of power and both Core Spray pumps i start within 8 to 10 seconds of restoration of power. The purpose of these l
time delays is to stagger the start of the RHR and Core Spray pumps on the associated Division 1 and Division 2 Buses, thus limiting the starting transients on the 4.16 kV emergency buses. The time delay functions are only necessary when power is being supplied from the standby power sources
! (EDGs). The time delays remain in the pump start logic at all times as the time delay relay contact is in parallel with the Auxiliary Power Monitor l relay contact. Either contact closure will initiate pump start. Thus, the time delays do not affect low pressure ECCS pump operation with normal AC j l power available. With normal AC power not available, the pump start relays j which would have started the B and C RHR pumps and both Core Spray pumps are ;
blocked by the Auxiliary Power Monitor contacts and the pump start time j delay relays become the controlling devices.
Amendment No. %, M, 444, 44-3, 4-32, 446, 170 79