ML022410359
| ML022410359 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 08/27/2002 |
| From: | NRC/NRR/DLPM |
| To: | |
| References | |
| TAC MB2227 | |
| Download: ML022410359 (15) | |
Text
1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING For no combination of loop recirculation flow rate and core thermal power shall the APRM flux scram trip setting be allowed to exceed 120% of rated thermal power.
- b.
Flux Scram Trip Setting (Refuel or Startup and Hot Standby Mode)
When the reactor mode switch is in the REFUEL or STARTUP position, average power range monitor (APRM) scram shall be set down to less than or equal to 15% of rated neutron flux (except as allowed by Nqte 12 of Table 3.1.1).
The IRM flux scram setting shall be set at less than or equal to 120/125 of full scale.
B.
Deleted C.
Reactor low water level scram setting shall be at least 127 inches above the top of the enriched fuel.
Amendment No.
4,,
44, 4, -4, Z1 8
2.1 LIM:TING SAFETY SYSTEEM SETTI*;r knendment No.
- i44, 44, 64-,
64, 464, "Q, 44, 4444, 48-,
211 1.1 SAFETY LIMIT 9
FICURU.E -. --
APRM FLOW REFERENCE SCRAM SETTING 1307 110 100 90-APRM Flow Biased Scram
- Setpoints shall be < values shown on the graph.
-1' 40 s0 I
100 RECIRCULATION FLOW(% RATED)
For single loop operalion.
the APRM Scram setting is adjusted according to Technical Specilcation 2.1.A. I.a 90 70 eo-I.
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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 subcritical 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Additional conservatism was taken in this analysis by assuming that the IRM channel closest to the withdrawn rod is bypassed.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 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.
Deleted 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 water level below the scram point and therefore this setting is provided.
D.
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.
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3.1 LIMITING CONDITIONS FOR OPERATION 3.1 REACTOR PROTECTION SYSTEM Applicability:
Applies to the operability of plant instrumentation and control systems required for reactor safety.
Objective:
To specify the limits imposed on plant operation by those instrument and control systems required for reactor safety.
Specification:
A.
Plant operation at any power level shall be permitted in accordance with Table 3.1.1.
The system response time from the opening of the sensor contact up to and including the opening of the scram solenoid relay shall not exceed 50 milliseconds.
B. During operation at >25%
Rated Thermal Power with the ratio of MFLPD to FRP greater than 1.0 either:
- a.
The APRM System gains shall be adjusted by the ratio given in Technical Specification 2.1.A.1, or
- b.
The power distribution shall be changed to reduce the ratio of MFLPD to FRP.
Amendment No.
4-i, 444, 4-"8 211 4.1 SURVEILLANCE REQUIREMEENTS 4.1 REACTOR PROTECTION SYSTEM Aoplicability:
Applies to the surveillance of the plant instrumentation and control systems required for reactor safety.
Objective:
To specify the type and frequency of surveillance to be applied to those instrument and control systems required for reactor safety.
Specification:
A.
Instrumentation systems shall be functionally tested and calibrated as indicated in Tables 4.1.1 and 4.1.2, respectively B.
Once within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after
>25% Rated Thermal Power and once a day during operation at >25% Rated Thermal Power thereafter, the maximum fraction of limiting power density and fraction of rated power shall be determined and the APRM system gains shall be adjusted by the ratio given in Technical Specification 2.l.A.l.a.
20
VYNPS TABLE 3.2.5 CONTROL ROD BLOCK INSTRUMENTATION Requ i red Channels Trip Funct4on Modes in Which Function Must be Operable Refuel Startup Run Trip Setting (Notes 9 and 10) 2 2 2 Rod Block Monitor (RBM A/B)
- a.
Upscale (Flow Bias)
(Note 7)
- b.
C.
Downscale (Note 7)
Inop (Note 7)
X
<0.66(W-AW)+N with a maximum as defined in the COLR (Note 5)
X
>2/125 Full Scale X
(Noteo 13) 2 Reactor Mode Switch -
Shutdown Position (Note 12)
Amendment No.
4-,
Q-,
64, 4-a, --6, ",
-4, -34,
-64, 4-14, 4---
211 51
TABLE 3.2.5 NOTES
- 1.
Deleted.
- 2.
Deleted.
- 3.
Deleted.
- 4.
Deleted.
- 5.
"W" is percent rated two loop drive flow where 100% rated drive flow is that flow equivalent to 48 x 106 lbs/hr core flow.
Refer to the Core Operating Limits Report for acceptable values for N.
AW is the difference between the two loop and single loop drive flow at the same core flow.
This difference must be accounted for during single loop operation.
AW = 0 for two recirculation loop operation.
- 6.
Not used.
- 7.
The trip may be bypassed when the reactor power is
<30% of Rated Thermal Power.
An RBM channel will be considered inoperable if there are less than half the total number of normal inputs from any LPRM level.
- 8.
With the number of operable channels less than the required number, place the inoperable channel in the tripped condition within one hour.
- 9.
With one or two RBM channels inoperable:
- a.
Verify that the reactor is not operating on a limiting control rod pattern (as described in the Bases for Specification 3.3.B.6); and
- b.
If one RBM channel is inoperable, restore the inoperable channel to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; and
- c.
If the required actions and associated completion times of Notes 9.a and 9.b above are not met, or if two RBM channels are inoperable, place one RBM channel in the tripped condition within the next hour.
- 10.
When a channel is placed in an inoperable status solely for performance of required surveillances, entry into associated Limiting Conditions for Operation and required action notes may be delayed for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provided the associated Trip Function maintains Control Rod Block initiation capability.
- 11.
Deleted.
- 12.
Required to be operable when the reactor mode switch is in the shutdown position.
- 13.
With one or more Reactor Mode Switch -
Shutdown Position channels inoperable, immediately suspend control rod withdrawal and immediately initiate action to fully insert all insertable control rods in core cells containing one or more fuel assemblies.
Amendment No. 64,,
6, 4 94, 04, 4-,86 211 52
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TABLIE 4. 2. 5 MINIMUM T_.3'_
AND CALIBRATION FREQUENCIES (CONTROL ROD BLOCK INSTRUMENTATION
.unrctional Test Eveiy Three Months (Note 4)
Fvery Three Months (Note 4)
Every Three Months Uri,'o/Operatiing Cycle
-',,r y Rftl i1l. ing Outage (Note 12)
Calibration Every Three Months (Note 13)
Every Three Months Once/Operating Cycle (Note 3)
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.1
- 44., - I, 1
1 4ý 4-1", 211
TABLE 4.2 NOTES
- 1.
Not used.
- 2.
During each refueling outage, simulated automatic actuation which opens all pilot valves shall be performed such that each trip system logic can be verified independent of its redundant counterpart.
- 3.
Trip system logic calibration shall include only time delay relays and timers necessary for proper functioning of the trip system.
- 4.
This instrumentation is excepted from functional test definition.
The functional test will consist of injecting a simulated electrical signal into the measurement channel.
- 5.
Deleted.
- 6.
Deleted.
- 7.
This instrumentation is excepted from the functional test definitions and shall be calibrated using simulated electrical signals once every three months.
- 8.
Functional tests and calibrations are not required when systems are not required to be operable.
- 9.
The thermocouples associated with safety/relief valves and safety valve position, that may be used for back-up position indication, shall be verified to be operable every operating cycle.
- 10.
Separate functional tests are not required for this instrumentation.
The calibration and integrated ECCS tests which are performed once per operating cycle will adequately demonstrate proper equipment operation.
- 11.
Trip system logic functional tests will include verification of operation of all automatic initiation inhibit switches by monitoring relay contact movement.
Verification that the manual inhibit switches prevent opening all relief valves will be accomplished in conjunction with Section 4.5.F.1.
- 12.
Trip system logic testing is not applicable to this function.
If the required surveillance frequency (every Refueling Outage) is not met, functional testing of the Reactor Mode Switch-Shutdown Position function shall be initiated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after the reactor mode switch is placed in Shutdown for the purpose of commencing a scheduled Refueling Outage.
- 13.
Includes calibration of the RBM Reference Downscale function (i.e.,
RBM upscale function is not bypassed when >30% Rated Thermal Power).
Amendment No.
4Q
-44,
,4,
, 4-"4, 211 74
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 i isolaticn 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 and temperature instrumentation is provided to detect a break in the HPCI and/or RCIC piping.
The HPCI and RCIC steam supply pressure instrumentation is provided to isolate the systems when pressure may be too low to continue operation.
These isolations are for equipment protection.
However, they also provide a diverse signal to indicate a possible system break.
These instruments are included in Technical Specifications because of the potential for possible system initiation failure if not properly tested. 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 aresuch 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 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 Rod Block Monitor (RBM) control rod block functions are provided to prevent excessive control rod withdrawal so that MCPR does not decrease below the fuel cladding integrity safety limit.
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.
During hot shutdown, cold shutdown, and refueling when the reactor mode switch is required to be in the shutdown position, the core is assumed to be subcritical with sufficient shutdown margin; therefore, no positive reactivity insertion events are analyzed.
The Reactor Mode Switch-Shutdown Position control rod withdrawal block, required to be operable with the mode switch in the shutdown position, ensures that the reactor remains subcritical by blocking control rod withdrawal, thereby preserving the assumptions of the safety analysis.
Two channels are required to be Amendment No.
.9, 2-59,"
84,
, 4-Q
, 211
BASES:
3.2 (Cont'd) operable to ensure that no single channel failure will preclude a rod block when required.
There is no trip setting for this function since the channels are mechanically actuated based solely on reactor mode switch position.
During refueling with the reactor mode switch in the refueling position, the refuel position one-rod-out interlock provides the required control rod withdrawal blocks.
To prevent excessive clad temperatures for the small pipe break, the HPCI or Automatic Depressurization System must function 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 operation.
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 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 set fcrth in the Offsite Dose Calculation Manual.
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, 24 4*,
O443, *4*,
&, 4-84, 4*9-3 211 78
SES:
2 PROTECTIVE INSTRUMENTATION (Cont'd)
Since logic circuit tests result in the actuation of plant equipment, testing of this nature was done while the plant was shut down for refueling.
In this way, the testing cf equipment would not jeopardize plant operation.
This Specification is a periodic testing program which is based upon the overall testing of protective instrumentation systems, including logic circuits as well as sensor circuits.
Table 4.2 outlines the test, calibration, and logic system functional instrumentation systems.
The testing of test of each relay wherever practicable.
includes all circuitry necessary to make proper functioning of the relay contacts initiation inhibit switches verifies the switches and relay contacts.
Functional temperature switches associated with the remotely by application of a heat source test schedule for the protective a subsystem includes a functional The testing of each relay the relay operate, and also the Testing of the automatic proper operability of the testing of the inaccessible isolation systems is accomplished to individual switches.
All subsystems are functionally tested, calibrated, and operated in their entirety.
A channel functional test is performed for the Reactor Mode Switch Shutdown Position function to ensure that the entire channel will perform the intended function.
The surveillance is only required to be performed once per operating cycle during refueling. The Refuelin-Outage frequeency is based on the need to perform this surveillance under the conditions that apply during a plant outage.
Operating experience has shown that this surveillance frequency is adequate to ensure functional operability.
Note 12 of Table 4.2.5 specifies that if the surveillance frequency of every Refueling Outage is not met, functional testing of the Reactor Mode Switch -
Shutdown Position function shall be initiated within I hour after the reactor mode switch is placed in the Shutdown position for the purpose of commencing a scheduled Refueling Outage.
This allows entry into the Shutdown mode when the surveillance requirement is not met.
Amendment No.
4.-, 46, -4, 445,4 211486 BA 4.
80a
BASES:
3.3 & 4.3 (Cont'd)
- 2.
The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the extremely remote event of a housing failure.
The amount of reactivity which could be added by this small amount of rod withdrawal, which is less than a normal single withdrawal increment, will not contribute to any damage of the primary coolant system.
The design basis is given in Subsection 3.5.2 of the FSAR, and the design evaluation is given in Subsection 3.5.4.
This support is not required if the reactor coolant system is at atmospheric pressure since there would then be no driving force to rapidly eject a drive housing.
- 3.
In the course of performing normal startup and shutdown procedures, a pre-specified sequence for the withdrawal or insertion of control rods is followed.
Control rod dropout accidents which might lead to significant core damage, cannot occur if this sequence of rod withdrawals or insertions is followed.
The Rod Worth Minimizer restricts withdrawals and insertions to those listed in the pre-specified sequence and provides an additional check that the reactor operator is following prescribed sequence.
Although beginning a reactor startup without having the RWM operable would entail unnecessary risk, continuing to withdraw rods if the RWM fails subsequently is acceptable if a second licensed operator verifies the withdrawal sequence.
Continuing the startup increases core power, reduces the rod worth and reduces the consequences of dropping any rod.
Withdrawal of rods for testing is permitted with the RWM inoperable, if the reactor is subcritical and all other rods are fully inserted.
Above 20% power, the RWM is not needed since even with a single error an operator cannot withdraw a rod with sufficient worth, which if dropped, would result in anything but minor consequences.
- 4.
Refer to the "General Electric Standard Application for Reactor Fuel (GESTAR II)," NEDE-24011-P-A, (the latest NRC-approved version will be listed in the COLR).
- 5.
The Source Range Monitor (SRM) system provides a scram function in noncoincident configuration.
It does provide the operator with a visual indication of neutrc-level.
The consequences of reactivity accidents are a function of the initial neutron flux.
The requirement of at least three counts per second assures that any transient, should it occur, begins at or above the initial value of 10-9 of rated power used in the analyses of transients from cold conditions.
One operable SRM channel is adequate to monitor the approach to criticality, therefore, two operable SRM's are specified for added conservatism.
- 6.
The Rod Block Monitor (RBM) is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power level operation. During reactor operation with certain limiting control rod patterns, the continuous withdrawal of a designated single control rod could result in a violation of the MCPR safety limit or the 1% plastic strain limit.
A limiting control rod pattern is a pattern which results in the core being on a thermal limit (i.e.,
operating on a limiting value for APLHGR, LHGR, or MCPR.
During use of such patterns, it is judged that testing of the RBM system prior to withdrawal of such rods will provide added assurance that improper withdrawal does not occur.
It is the responsibility of the Nuclear Engineer to identify these limiting patterns and the designated rods either when the patterns are initially established or as they develop due to the occurrence of inoperable control rods.
Amendment No. -24, -.3,
-4r,.44,
-4, VY 99 fig )11 90
3.6 LIMITING CONDITIONS FOR OPERATION
- 3.
The indicated core flow is the sum of the flow indication from each of the twenty jet pumps.
If flow indication failure occurs for two or more jet pumps, immediate corrective action shall be taken.
If flow indication for all but one jet pump cannot be obtained within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> an orderly shutdown shall be initiated 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 />.
G.
Single Loop Operation
- 1. The reactor may be started and operated or operation may continue with a single recirculation loop provided that:
- a.
The designated adjustments for APRM flux scram setting (Specification l.A.l.a and Table 3.1.1),
rod block monitor trip setting (Table 3.2.5),
MCPR fuel cladding integrity safety limit (Specifi cation 1.1.A), and MCPR operating limits and MAPLHGR limits, provided in the Core Operating Limits Report, are initiated within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
During the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, either these adjustments must be completed or the reactor brought to Hot Shutdown.
4.6 SUR'.EILLANCE REQU:REM'_ENNTS
- 2.
The surveilla.,ce requirements of 4.6.F.1 and 4.6.F.2 do not apply to the idle loop and associated jet pumps when in single loop operation.
- 4.
The baseline data required to evaluate the conditions in Specifications 4.6.F.1 and 4.6.F.2 shall be acquired each operating cycle.
Baseline data for evaluating 4.6.F.2 while in single loop operation shall be updated as soon as practical after entering single loop operation.
N' Amendment No. 44, 44, 44-6, 4-44, 44-,
4.", 4-M6, 211 122
include a tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem/yr and their associated man-rem exposure according to work and job functions, 1/ e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance),
waste processing, and refueling.
The dose assignment to various duty functions may be estimates based on Self-Reading Dosimeter (SRD),
TLD or film badge measurement.
Small exposures totaling less than 20% of the individual total dose need not be accounted for.
In the aggregate, at least 80% of the total whole body dose received from external sources should be assigned to specific major work functions.
B.
Monthly Operating Reports Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis no later than the fifteenth of each month following the calendar month covered by the report.
These reports shall include a narrative summary of operating experience during the report period which describes'the operation of the facility.
C.
Core Operating Limits Report The core operating limits shall be established and documented in the Core Operating Limits Report (COLR) before each reload cycle or any remaining part of a reload cycle for the following:
- 1.
The Average Planar Linear Heat Generation Rates (APLHGR) for Specifications 3.11.A and 3.6.G.la,
- 2.
The K, core flow adjustment factor for Specification 3.11.C.,
- 3.
The Minimum Critical Power Ratio (MCPR) for Specifications 3.11.C and 3.6.G.]a,
- 4.
The Linear Heat Generation Rates (LHGR) for Specifications 2.1.A.la and 3.1l.B, and
- 5.
The Power/Flow Exclusion Region for Specifications 3.6.J.la and 3.6.J.lb.
The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in:
- Report, E.
E. Pilat, "Methods for the Analysis of Boiling Water Reactors Lattice Physics," YAEC-1232, December 1980 (Approved by NRC SER, dated September 15, 1982).
1/ This tabulation supplements the requirements of 20.2206 of 10 CFR Part 20.
Amendment No.
4-a, 44.,
- 444, 4-544, 444, 2112 259