ML20064D692
| ML20064D692 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 02/25/1994 |
| From: | HOUSTON LIGHTING & POWER CO. |
| To: | |
| Shared Package | |
| ML20064D688 | List: |
| References | |
| NUDOCS 9403110359 | |
| Download: ML20064D692 (3) | |
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4-3/4.3 INSTRUMENTATION BMES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGIf4EERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION The OPERABILITY of the Reactor Trip System and the Engineered Safety Features Actuation System instrumentation and interlocks ensures that:
(1) the associated ACTION and/or Reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its Setpoint, (2) the specified coincidence logic is maintained, (3) sufficient redundancy is main-tained to permit a channel to be out-of-service for testing or maintenance, and (4) suf ficient system functional capability is available from diverse parameters.
The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility 4
design for the protection and mitigation of accident and transient conditions.
The integrated operation of each of these systems is consistent with the assumptions used in the safety analyses.
The Surveillance Requirements speci-fied for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveil-lance tests performed at the minimum frequencies are sufficient to demonstrate this capability.
Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with WCAP-10271,
" Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System,",,anddupplements to that report //o su r n)
Surveillance intervals and out of service times were determined based on maintaining an appropriate level of reliability of the Reactor Protection System instrumentation.
The Engineered Safety Features Actuation System Instrumentation Trip Setpoints specified in Table 3.3-4 are the nominal values at which the bistables are set for each functional unit.
A Setpoint is considered to be adjusted consistent with the nominal value when the "as measured" Setpoint is within the band allowed for calibration accuracy.
To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which Setpoints can be measured and calibrated, Allowable Values for the Setpoints have been specified in Table 3.3-4.
Opera-tion with Setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error. An optional provision has been included for determining the OPERABILITY of a channel when its Trip Setpoint is found to exceed the Allowable Value.
The methodology of this option utilizes the "as measured" deviation from the specified calibration point for rack and sensor components in conjunction with a statistical combination of the other uncertainties of the instrumentation to measure the process variable and the uncertainties in calibrating the instrumentation.
In Equation 2.2-1, Z + R + 5 < TA, the interactive effects of the errors in the rack and the sensor, anii the "as measured" values of the errors are considered.
Z, as specified in Table 3.3-4, in percent span, is the statistical summation of err. ors assumed in the analysis excluding those associated with the sensor and rack drif t and the accuracy of their measurement.
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,4 3/4'.5 EMERGENCY CORE COOLING SYSTEMS DASES 3/4.5.1 ACCUMULATORS The OPERABILITY of each Reactor Coolant System (RCS) accumulator ensures that a sufficient volume of borated water will be immediately forced into the reactor core through three cold legs in the event the RCS pressure falls below the pressure of the accumulators. This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures.
The limits on accumulator volume represent a spread about an average value used in the safety analysis and have been demonstrated by sensitivity studies to vary the peak clad temperature by less than 20 F.
The limit on accumulator pressure ensures that the assumptions used for accumulator injec-tion in the safety analysis are met.
The accumulator power operated isolation valves are considered to be
" operating bypasses" in the context of IEEE Std. 279-1971, which requires that bypasses of a protective function be removed automatically whenever permissive conditions are not met.
In addition, as these accumulator isolation valves fail to meet single failure criteria, removal of power to the valves is required.
The limits for operation with an accumulator inoperable for any reason except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator which u
If a closed isolation may result in unacceptable peah cla(dj?g temperatures.the full capability of one accumu valvecannotbeopenedwithinode-fiour is not available and prompt action is required to place the reactor in a mode where this capability is not required.
3/4.5.2'and 3/4.5.3 ECCS SUBSYSTEMS The OPERABILITY of three independent ECCS subsystems ensures that suffi-cient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single failure considera-tion.
Each subsystem operating in conjunction with the accumulators is cap-able of supplying sufficient core cooling to limit the peak cladding tempera-tures within acceptable limits for all postulated break sizes ranging from the double ended break of the largest RCS cold leg pipe downward. One ECCS is assumed to discharge completely through the postulated break in the RCS loop.
Thus, three trains are required to satisfy the single failure criterion. Note that the centrifugal charging pumps are not part of ECCS and that the RHR pumps are not used in the injection phase of the ECCS.
Each ECCS subsystem and the RHR pumps and heat exchanges provide long-term core cooling capability in the recirculation mode during the accident recovery period.
When the RCS temperature is below 350 F, the ECCS requirements are balanced between the limitations imposed by the low temperature overpressure protection and the requirements necessary to mitigate the consequences of a LOCA below 350*F. At these temperatures, single failure considerations are not required because of the stable reactivity condition of the reactor and the limited core cooling requirements. Only a single Low Head Safety Injection pump is required to mitigate the effects of a large-break LOCA in this mode.
However, two are SOUTH TEXAS - UNITS 1 & 2 B 3/4 5-1 l
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