ML20134L926
| ML20134L926 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 11/08/1996 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20134L925 | List: |
| References | |
| NUDOCS 9611220096 | |
| Download: ML20134L926 (5) | |
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ENCLOSURE TECHNICAL SPECIFICATION (TS) BASES CHANGE SEQUOYAH NUCLEAR PLANT (SON) UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 LIST OF AFFECTED PAGES Unit 1 B 3/4 3-2 8 3/4 7-4 l
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Unit 2 I
I' B 3/4 3-2 l
8 3/4 7 4 i
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s t
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9611220096 961108 PDR ADOCK 05000327 p
INSTRUMENTATION
' BASES The measurement of response time at the specified frequencies provides assurance that the protective and ESF action function associated with each channel is completed within the time limit assumed in the accident analyses.
i LNo credit was taken in the analyses for those channels with response times j
indicated as not applicable in the updated final safety analysis report.
lR194 l
Response time may be demonstrated by any series of sequential,
)
overlapping or total channel test measurements provided that such tests i
demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test j
measurements or 2) utilizing replacement sensors with certified response times.
Action 15 of Table 3.3-1, Reactor Trip System Instrumentation, allows the breaker to be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for the purpose of performing maintenance. The 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is based on a Westinghouse analysis performed in R58 WCAP-10271, Supplement 1, which determines bypass breaker availability.
The placing of a channel in the trip condition provides the safety BR-9 function of the channel.
If the. channel is tripped for testing and no other condition would have indicated inoperability, the channel should not be I
declared inoperable.
l 3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that 1) the radiation levels are continually measured in the areas served by the individual channels and 2) the alarm or automatic action is initiated when the radiation l
level trip setpoint is exceeded.
l l
3 /4. 3. 3. 2 MOVABLE INCORE DETECTORS The OPERABILITY of the movable incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the reactor core. The OPERABILITY of this system is demonstrated by irradiating each detector used and determining the acceptability of its voltage curve.
For the purpose of measuring F Z) or F AH a full incore flux map is used. Quarter-corefluxmaps,asdefkn(edinWCAP-8648, June 1976, may be used in recalibration of the excore neutron flux detection system, and full incore flux maps or symmetric incore thimbles may be used for monitoring the QUADRANT POWER TILT RATIO when one Power Range Channel is inoperable.
3/4.3.3.3 SEISMIC INSTRUMENTATION j
The OPERABILITY of the seismic instrumentation ensures that sofficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety. This capability is required to permit comparison of the measured response to that used in the J
l February 15, 1996 i
SEQUOYAH - UNIT 1 B 3/4 3-2 Amendment No. 54, 190 1
I 1
1 a
i
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- DLAN SYSTEMS BASES 3/4.7.5 ULTIMATE HEAT SINK (UHS)
The limitations on UHS water level and temperature ensure that suf ficient R83 cooling capacity is available to either 1) provide normal cooldown of the facility, or 2) to mitigate the effects of accident conditions within acceptable limits.
The limitations on the maximum temperature are based on providing a 30 R12 day cooling water supply to safety related equipment without exceeding their design basis temperature and is consistent with the recommendations of Regulatory Guide 1.27, " Ultimate Heat Sink for Nuclear Plants", March 1974.
The limitations on minimum water level are based on providing sufficient flow to the ERCW serviced heat loads after a postulated event assuming a time.
R83 dependent drawdown of reservoir level.
Flow to the major transient heat loads (CCS and CS heat exchangers) is balanced assuming a reservoir level of elevation 670.
The time-independent heat loads (ESF room coolers, etc.) are balanced assuming a reservoir level of elevation 639.
BR-10 3/4.7.6 FLOOD PROTECTION The requirements for flood protection ensures that facility protective actions will be taken and operation will be terminated in the event of flood conditions. A Stage 1 flood warning is issued when the water in the forebay is predicted to exceed 697 feet Mean Sea Level USGS datum during October 1 through April 15, or 703 Feet Mean Sea Level USGS datum during April 15 through September 30.
A Stage II flood warning is issued when the water in the forebay is predicted to exceed 703 feet Mean Sea Level USGS datum. A maximum allowed water level of 703 Mean Sea Level USGS datum provides sufficient margin to ensure waves due to high winds cannot disrupt the flood mode preparation. A Stage I or Stage II flood warning requires the imple-mentation of procedures which include plant shutdown.
Fureher, in the event of a loss of
(
communications simultaneous with a critical combination flood, headwaters, and/or seismically induced dam failure the plant will be shutdown and flood protection measures implemented.
l 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM The OPERABILITY of the control room ventilation system ensures that 1) the ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment and instrumentation cooled by this system and 2) the control room will remain habitable for operations personnel l
during and following all credible accident conditions. The OPERABILITY of this I
system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rem or less whole body, or its equivalent. This limitation is consistent with the requirements of General Design Criteria 19 of Appendix "A",
ANSI N510-1975 will be used as a procedural guide for surveillance testing.
SEQUOYAH - UNIT 1 B 3/4 7-9 Amendment No. 79 September 13, 1996
INSTRUMENTATION BASES REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION (Continued)
)
The measurement of response time at the specified frequencies provides assurance that the protective and the engineered safety feature actuation associated with each channel is completed within the time limit assumed in the accident analyses. No credit was taken in the analyses for those channels with response times indicated as not applicable in the updated final safety analysis report.
R182 1
I I
Response time may be demonstrated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined.
Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test l
measurements or 2) utilizing replacement sensors with certified response times.
Action 15 of Table 3.3-1, Reactor Trip System Instrumentation, allows the breaker to be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for the purpose of performing R46 maintenance.
The 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is based on a Westinghouse analysis performed in WCAP-10271, Supplement 1, which determines bypass breaker availability.
The placing of a channel in the trip condition provides the safety function of the channel.
If the channel is tripped for testing and no other BR-10 j
l condition would have indicated inoperability, the channel should not be declared inoperable.
3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that 1) the radiation levels are continually measured in the areas served by the individual channels and 2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded.
3/4.3.3.2 MOVABLE INCORE DETECTORS The OPERABILITY of the movable incere detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the reactor core. The OPERABILITY of this system is demonstrated by irradiating each detector used and determining the acceptability of its voltage curve.
For the purpose of measuring F (Z) or F AH a full incore flux map is Quarter-corefluxmaps,asdekinedinWCAP-8648, June 1976, may be used used.
in recalibration of the excore neutron flux detection system, and full incore flux maps or symmetric incore thimbles may be used for monitoring the QUADRANT POWER TILT RATIO when one Power Range Channel is inoperable.
3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event i
and evaluate the response of those features important to safety. This i
capability is required to permit comparison of the measured response to that used in the R72 a
February 15, 1996 2
SEQUOYAH - UNIT 2 B 3/4 3-2 Amendment Nos. 46, 72, 182 i
=.
- PLANT SYSTEMS l
BASES 3/4.7.S tTLTIMATE HEAT SINK The limitations on the ultimate heat sink water level and temperature R70 ensure that sufficient cooling capacity is available to either 1) provide normal cooldown of the facility, or 2) to mitigate the effects of accident conditions within acceptable limits.
The limitation on maximum temperature is based on providing a 30 day cooling water supply to safety related equipment without execeding their design basis temperature and is consistent with the recommendations of Regulatory Guide 1.27,
" Ultimate Heat Sink for Nuclear Plants", March 1974.
The limitations on minimum water level are based on providing sufficient flow to the ERCW serviced heat loads after a postulated event assuming a time dependent drawdown of reservoir level. Flow to the major transient heat loads R70 (CCS and CS heat exchangers) is balanced assuming a reservoir level of el. 670.
The time independent heat loads (ESF room coolers, etc.) are balanced assuming a reservoir level of el. 639.
lBR-11 3/4.7.6 FLOOD PROTECTION The requirements for flood protection ensures that facility protective actions will be taken and operation will be terminated in the event of flood conditions.
A Stage 1 flood warning is issued when the water in the forebay is predicted to exceed 697 feet Mean Sea Level USGS datum during October 1 through April 15, or 703 Feet Mean Sea Level USGS datum during April 15 through September 30.
A Stage II flood warning is issued when the water in the forebay is predicted to exceed 703 feet Mean Sea Level USGS datum. A maximum allowed water level of 703 Mean Sea Level USGS datum provides sufficient margin to ensure waves due to high winds cannot disrupt the flood mode preparation. A Stage I or Stage II flood warning requires the implementation of procedures which include plant shutdown.
Further, in the event of a loss of communications simultaneous with a critical combination flood, headwaters, and/or seismically induced dam failure the plant will be shutdown and flood protection measures implemented.
3/4.7.7 CONTROL ROCM EMERGENCY VENTILATION SYSTEM The OPERABILITY of the control room ventilation system en? res that 1) the ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment and instrumentation cooled by this system and 2) the control room will remain habitable for operations personnel during and following all credible accident conditions.
The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rem or less whole body, or its equivalent.
This limitation is consistent with the requirements of General Design Criteria 19 of Appendix "A",
ANSI N510-1975 will be used as a procedural guide for surveillance testing.
l SEQUOYAH - UNIT 2 B 3/4 7-4 Amendment No. 70 September 13, 1996
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