ML20211M734
| ML20211M734 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 08/30/1999 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20211M732 | List: |
| References | |
| NUDOCS 9909090194 | |
| Download: ML20211M734 (26) | |
Text
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L IXEE EQUIggg 7_Ig1 1.11 DOSE EQUIVALENT I-131 shall be.that concentration of I-131 (microcurie /
R159 gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133 I-134, and I-135 actually present.
The thyroid dose conversion factors used, for this calculation shall be those listed in Table III-of TID-14844,
Calcul+ tion of Distance Factors for Power and Test Peactor Sites.
E - AVEPAGE DISINTEGRATION ENERGY 1.12 E shall be the average (weighted in proportion to the concentration of R159 each radionuclide in the reactor coolant at the time of sanpling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-lodine activity in the coolant.
(ESF)
ENGINEERED SAFETY FEATURE APESPWSE TIME lR159
- 1.13 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.
FREQUENCY NOI' TION A
1.14 The FREQUENCY NOTAL' ION specified for the performance of Surveillance R159 Requirements shall correspond to the intervals defined in Table 1.2.
GASEOUS RAIEASTE TREA7MEtU SYSTEM 1.15 A GASEOUS RADWASTE TREAIMENT SYSTEM is any system designed and installed l R159 to reduce radioactive gaseous effluents by collecting primary coolant system offgases fram the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.
This area is affected by Technical Specification IDENTIFIED LEAKAGE Change 98-10 1.16 IDENTIFIED IEAKAGE shall be:
/
l R159 i
G Isakage (exctpt O'NIPOUFD IEAKP&) into Closed syStsS, sudi as pump seal or valve packing leaks that are captured and conducted to s y =or-collecting tank, or m
\\
x m
y-Tic resporse tine nuy be neasured by neans of any series of sequernial, omrlapping, or total steps so that the entire response time is measured. In lieu of measurement, response tine nny be verifed for selected componeras provided dat de conponents and tic netledology for verification love been previously reviewed and approved by tic NRC.
C#" &o99g9 W 90830
'-3 Amendment No. 12 71, 155 l
99 October 23,1991 l
PDR ADOCK 05000327 i
P PDR
f O
I PRESSURE BOUNDARY LEAV, AGE 1.22 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.
PROCESS CONTROL PROGRAM (PCP) 1.23 DELETED R237 PURGE - PURGING 1.24 PURGE or PURGING is the controlled process of discharging air or gas from 1
a confinement to maintain temperature, pressure, humidity, concentration or j
other operating condition, in such a manner that replacement air or gas is j
required to purify the confinement.
)
QUADRANT POWER TILT RATIO 1.25 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated R205 outputs, whichever is greater.
RATED THERMAL POWER (RTP)
R145 1.26 RATED THERMAL POWER (RTP) shall be a total reactor core heat transfer rate to the reactor coolant of 3411 MWt.
(RTS) l REACTOR TRIP SYSTEM RESPONSE TIME (RTS) 1.27 The REACTOR TRIP SYSTEM RESPONSE TIME d all be the time interval from when the monitored parameter exceeds its+ trip setpoint at the channel sensor until loss of stationary gripper coil voltage.
REPORTABLE EVENT 1.28 A REPORTABLE EVf21T shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.
The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may_be verified for selected components provided that the components and the methodology verification have been previously reviewed and approved by the NRC.
l July 1, 1998 l
SEQUoYAH - UNIT I 15 Amendment No.
12, 7 1, 141, 148, 155, 201, 233
1 l
3 / 4. 3 INSTRUMENTATION e
3/4.3.1 Reactor Trip System Instrumentation LIMITING CONDITION FOROPERATION 3.3.1.1 As a minimum, the reactor trip system instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE.
l APPLICABILITY:
As shown in Table 3.3-1.
R194 ICTION:
As shown in Table 3.3-1.
SURVEILLANCE REOUIREMENTS 4.3.1.1.1 Each reactor trip system instrumentation channel and interlock shall be demonstrated OPERABLE by the performance of the GW@EL QECK, QWWEL R16 CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies-shown in Table 4.3-1.
4.3.1.1.2 The logic for the interlocks shall be denenstrated OPERABLE prior-to each reactor startup unless nerformed during the oreceeding 92 days. The total interlock function shall be demonstrated OP"-RABLE at leas,t once per 8 months during CHANNEL CALIBRATION testing of each channel affected by interlock operation.
I verified l 4.3.1.1.
REACICR TRI TSTEM RESPCNSE TIME of each reactor trip function verification l shall be denunctrated to be within its limit at least once per 18 months.
Neutron detectors are exempt from response time testing. Each 4est*
include at least one 1 ic train such that both logic trains are tcctci at R194 least once are tchl*per 36 mor is and one channel per function such that all channelsN at least once every N times 18 months where N is the total number of kerifiedI redundant channels in a specific reactor trip function as shown in the " Total No. of Channels" column of Table 3.3.1.
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SEQUOYAH. UNIT I 3/4 3-1 Amendment No. 12, 190 Novertiber 9 1994
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INSTRWENTATIOJ 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRWENTATION
' LIMITING CONDITION FOR OPERATION 3.3.2.1 The Engineered Safety Feature Actuation System (ESFAS) instrunentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4.
APPLIGDBILTIY:
As shown in Table 3.3-3.
ACTIW:
Rl94 l
l a.
With an ESFAS instrumentation channel or interlock trip setpoint J
i
'less conservative than the value shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the i
applicable ACTIGi requirement of Table 3.3-3 until the channel is l
restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint value.
b.
With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3-3.
SURVEILIANCE REQUIRDB71S 4.3.2.1.1 Each ESFAS instrumentation channel shall be demonstrated OPERABLE-by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL i
l FUNCTIONAL TEST ' operations for the MODES and at the frequencies shown in Table 4.3-2.
4.3.2.1.2 The logic for the interlocks shall be demonstrated OPERABLE during the automatic actuation logic test. The total interlock function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by interlock operation.
verified 4' 3.2.1.3 THELENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function l
shall be dc:nonstrated to be within the limit at least once per 18 months. Each est shall include at least one logic train such that both logic trains are tested at least once per 36 months and one channel per function such that all
/
channels ar tcoted at least once per N times 18 months where N is the total l
number of redundant channels in a specific ESEAS function as shown in the
" Total No. of Channels" Column of Table 3.3-3.
l verification l l verifiedi ly_eri fied l I
l SEQUOYAH - WIT 1 3/4 3-14 Amendment No. 190
p
-3/4.3 - INSTRWENTATICN BASTS --
3/4.3.1'and 3/4.3.2 PROTECTIVE AND ENGINEERED SAFETY FFATURES (ESF)
INSTRWENTATICN The OPERABILITY of the protective and ESF instrumentation systems and interlocks: ensure that 1) the associated ESF action and/or reactor trip will be. initiated when the parameter monitored by each channel or cambination
.thereof reaches its setpoint, 2) the specified coincidence logic is maintained, 3)- sufficient redundancy is maintained to permit a channel to be out of service for testing.or maintenance, and 4) sufficient system functional capability is available for protective and ESF purposes from l
. diverse parameters.
The OPERABILITY of these systems is required to provide the overall l-reliabilitykectionandmitigationofaccidentandtransient-conditions.Theredundancy and diversity as i
for the pro integrated operation of each of these systems is consistent with the assumptions used in the accident analyses.
The Engineered Safety Features System interlocks perform the functions indicated below on increasing the required parameter, consistent with the l
setpoints listed in Table 3.3-4:
l P-11 Defeats the manual block of safety injection actuation on low-pressurizer pressure.
L P-14 Trip of all feedwater pumps, turbine trip, closure of feedwater R145 l
isolation valves and inhibits.feedwater control valve modulation.
on decreasing the required parameter the opposite function is performed at reset'setpoints.
R145 l
The surveillance requirements specified for these systems ensure that the overall system functional capability is maintained camparable to the original design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability.
R203 The surveillance for the camparison of the incore to the excore Axial Flux
]
Difference is required only when reactor power is > 15 percent. The 96 hour0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> delay in the first performance of the surveillance after reaching 15 percent reactor thermal power (RTP), following a refueling outage, is to achieve a
' higher power level and approach Xenon stability. The surveillance is typically
]
performed when RTP is > 30 percent to ensure the results of the evaluation are 1
more accurate and the adjustments more reliable. The fregaency of 31 EFPD is 1
to allow slow changes in neutron flux tc be better detected during the fuel cele.
Info Only f
May 30 1995 SEQUOYAH UNIT 1 B 3/4 3-1 Amendment No. 141, 199
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l r
l INSTRUMENTATION l verification
[ actuation l j
l safety l l
tripG g
BASES
/
reactor
>==
The ::::= I:nt of the response time at the specified frequencies provides assurance that the p::tectit.and ESF ccti_n function assoc ated with each channel is completed within the time limit assumed in the c uident 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.
1 Action 15 of Table 3.3-1, Reactor Trip System Instrumentation, allows the breaker to he 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 RS8 WCAP-10271, Supplement 1, which determines bypass breaker availability.
The placing of a channel in the trip condition provides the safety R242 i
function of the channel, if the channel is tripped for testing and no other j
condition would have indicated inoperability, the channel should not be declared inoperable.
'The Auxiliary Feedwater (AEM) Suction Pressure-Low function must be OPERABLE in MODES 1, 2, and 3 to ensure a safety grade supply of water for the R242 I AFW System to maintain the steam generators as the heat sink for the reactor.
This function does not have to be OPERA 13LE in MODES 5 and 6 because heat being generated in the reactor is removed via the Residual Heat Removal (RHR) System and does not require the steam generators as a heat sink. In MODE 4, AEW automatic suction transfer does not need to be OPERABLE because RHR will already be in operation, or sufficient time is available to place RHR in
- operation to remove decay heat.
This area is affected by Techn.ical Specification 3/4.3.3 MONITORING INSTRUMENTATION Change 99-03 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION p
1
~r 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.
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3/4.3.3.2 MOVABLE INCORE DETECTORS The OPERA 31LITY 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 OPE N ILITY of this system is demonstrated by
- irradiating each detector used and determining the acceptability of its voltage curve.
For the purpose of measuring Fe(X,
,Y Z) or Fu(X,Y) a full incore flux cap l R227 l
.is used. Quarter-care flux maps, as defined in WCAP-8G48, June 1976, may be j
used in recalibration of the excore neutron flux detection system, and full I
incore flux maps or symmetric incore thimbles may be used for monitoring the QUADRANT POWER TILT RATIO when one Power Range Channel is inoperable.
November 19, 1998 SEQUOYAH _ UNIT I B 3/4 3-2 Amendment No. 54, 190, 223, 238 i
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e
+
INSERTI
' Response time may be verified by actual response time tests in any series o f sequential, overlapping or total channel measurements, or by the summation of allocated sensor, signal processing and actuation logic response times with actual response time tests on the remainder of the channel. Allocations for sensor response times may be derived from: (1) historical records based on acceptable re'sponse time tests (hydraulic, noise, or power interrupt-tests), (2) inplace, onsite, or offsite (e.g.
vendor) test measurements, or (3) utilizing vendor engineering specifications. WCAP-13632-P-A Revision 2 (January 1996)
" Elimination of Pressure Sensing Response Time Testing Requirements,"
provides the basis and methodology for using allocated sensor response times in the overall verification of the channel response time for specific sensors identified in the WCAP. Response time verification for other sensor types must be demonstrated by test. TVA has verified that the selected components at sequoyah are the same Manufacturer and Model No. as evaluated in WCAPs 13632-P-A and 14036-P-A. WCAP-14036-P-A Revision 1 (October, 1998) " Elimination of Periodic Protection Channel Response Time Tests," provides the basis and methodology for using allocated signal processing and actuation logic response times in the overall verification of the protection system channel response time.
The allocations for sensors, signal conditioning, and actuation logic response times must be verified prior to placing the component in operational service and reverified following maintenance that may adversely affect response time. Periodically, sensors, signal conditioning and logic components are functionally tested.
i 1
b EEINITIQE D& E BJUIVAIlNT I-131 1.11 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie /
gram) which alone would produce the same thyroid dose as the quantity and iso-l R146 topic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, Calculation of Distance Factors for Power and Test Reactor Sites.
E - REENE DISINIFIFATICN ENEPGf 1.12 E" shall be the average (weighted in proportion to the concentration of R146 each radionuclide in the reactor coolant at the time of sampling) of the sum I
of the average beta and gantaa energies per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.
l(ESF)l BGINiEED SVETI FUEFE A ESEOEE TIbE 1.13 The EtEINEERED SAFEIY FEATURE PESKNSE TIME shall be that time interval R146 fram when the monitored parameter exceeds its ESF actuation setpoint at the l
channel sensor until the ESF equignent is capable of perfonning its safety function (i.e., the valves travel to their required positians, punp discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.
FEUJE1CI FDIATICN 1.14 The FPEQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.
lR146 GPSEUB PADPSIE 'IRAlbHJT SYSIEM 1.15 A GASEOUS PADHDSTE TPEATMENT SYSTEM is any system designed and installed R146 to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the j
purpose of reducing the total radioactivity prior to release to the envirorrnent.
The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously revimed and approved by the NRC.
EBl OfAH - UNIT 2 1-3 Amendment No. 63, 146 March 30, 1992 e
p
- DEFINITICNS -
y PATED THERMAL POWER (RTP) 11.26 BATED THERMAL EGER (RTP) shall be ~a total reactor core heat transfer i R14 rate to the reactor coolant of 3411 MWt.
' l (RTS)l l
REACTOR TRIP SYSTE24 ^ RESPONSE TIME l (RTS) l 1.27 The REACTOR TRIP SYSTEM RESPONSETgshall be the thne interval from l R14 when the monitored parameter exceeds.its trip setpoint at the channel sensor until-loss of' stationary gripper coil voltage.
REPORTABLE EVENT 1.28 A REPORTABLE EVENT shall be any of those conditions specified in S l R14 50.73 to-10 CER Part 50.
SHIELD BUILDING INTEGRITY l'29 SHIELD BUILDING INTEGRITY shall exist when:
lR14 a.
The door in each access opening is closed except when the access opening is being used for normal trarait entry and exit, b.
The. emergency gas treatment-system is OPERABLE.
j
. c.
The sealing mechanism associated with each penetration (e.g., welds, bellows or O-rings) is OPERABLE.
l SHUTDOWN MARGIN 1.30 SHUTDOWN MARGIN shall be the instantaneous amount of. reactivity by,, which.
lR14 the' reactor is subcritical or_would be subcritical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of-highest reactivity worth which is assumed to be fully withdrawn.
SITE BOUNDARY-1.31'The SITE BOUNDARY shall be'that line beyond which the land is not owned, lRl4 leased,- or otherwise controlled by the licensee (see figure 5.1-1).
l-
, The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by tle NRC.
-SEQUOYAH - UNIT 2 1-6 Anendment No. 63, 132, 146 l'
March 30,1992
v:
..; ; 2 f
43/4.3 INSTUMENTATION F
3/4.3.1 IUCICR 'IBIP SHEM IN3IRMNMIIN i
4 LIMIT 1NG CONDITICN EDR OPEPATIQ4-3.3.1 As - a mininun, the' reactor trip system instrumentation. channels and -
interlocks of Table 3.3-1 shall be OPERABLE.
l t
R182-APPLICABILITY:' 'As'shown'in Table 3.3-1.
L
.PCTIQ1:-
L As shown in Table 3.3-1.
SURVEILIANCE REQUIFB4ENTS l
l 4.3.1 1 1 Each'remtor trip system instrumentation channel and interlock shall-be demonstrated CrZPABLE by the perfonnance of the CHAtNEL CHECK, CHANNEL CALIBPATION and CHANNEL EllNCTIONAL TEST operations for the KODES and at the l
.. frequencies.shown in Table.4.3-1.-
4.3.1.1.2 The logic for the interlocks shall be demonstrated OPEPABLE prior to each reactor startup unless perfonned.during the preceeding 92 days. The total
. interlock function shall be demonstrated OPEPABLE at least once per 18 months
'during QNNEL CALIBRATION testing of each channel affected by interlock operation.
i.
verified l 4.3.1.1.3 PBCKR TRIP SYSTEM PESEGE TIME of each reactor trip function
- shall be & r;nctrated to be within its limit at least once per 18 months,) verification l Neutron d ectors are exempt from response time testing. Each test tian j
j include - least one logic train such that both logic trains are tectc" t R182 least o ce per 36 months and one channel per function such that all channe o are t :ted at least once every N times 18 months where N is the total number of verified redundant channels in a specific reactor trip function as shown in the " Total
.No. of-Channels" column of Table 3.3.1.
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SEG USH - UET.2 3/4 3-1 Atutitelt tb.182 November,9,19M l
7
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INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3 2 The Engineered Safety Feature Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPEPABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4.
l APPLICABILITY:
As shown in Table 3.3-3.
ACTICN:
a.
With an ESFAS instrumentation channel or interlock trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPEPABLE status with the trip setpoint adjusted consistent with the Trip Setpoint value.
b.
With an ESFAS instrumentation channel or interlock inoperable, take the ACTICN shown in Table 3.3-3.
j SURVEIILANCE REOVIREMENTS 4.3.2.1.1 Each ESFAS instrumentation channel and interlock shall be demon-
.strated OPEPABLE by the perfonnance of the CHANNEL CHECK, CHANNEL CALIBPATION and CHANNEL FUNCTIONAL TEST operations for the FDDES and at the freglencies shown in Table 4.3-2.
4.3.2.1.2 The logic for the interlocks shall be demonstrated OPEPABLE during i
the automatic actuation logic test. The total interlock function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBPATION testing of each channel affected by interlock operation.
Ivorifiedl 4.3.2.1.3 THE ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function m
shall be demenstrated to be within the limit at least once per 18 months. Each test shall include at least one logic train such that both logic trains are tested at least once per 36 months and one channel per function such that all channels a r tested at least once per N times 18 months where N is the total number of edundant channels in a specific ESFAS function as shown in the
" Total No.
f Channels" Column of Table 3.3-3.
lverificationiiverifiedI verifiedI SHD(AH - UNIT 2 3/4 3-14 Amendment No.
182 November 9,1994
=
6 INSTRUMENTATION l
BASES REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATIO!C l verification l l reactor trip l l ESF l
function l The
-'c = Ece of the r[sponse the[ measured frequencies provides /
time at assurance that the p:ctcet wc and the cr.gircered cafety icat= c actuation v
associated with each channel is complete within the time limit assumed in the 1
accidert analyses. No credit was taken in the analyses for those channels with i Micated as not applicable in the updated final safety analysis response e
report.
safety R182 p yes S ys'eepe.ng g ygsge Q pp W M M R*8 Pops epi @elff empns yMed@Whpl@hatj auch dNt sjdemons tge} thehto td@
Replaced ManeQeskmeasurementii3 rovidsd6t hannels by lnsert 1 b$efWer 1hd$dd1#1hdd resp 6nseg(time $rd; pie 6enne teT ffsit4%tymendnsenO""$5Ndi:i~$$Aj$dhs52MMad UEllisinsi$sh!s@$$$ !/ f.#PMbhNNM$N85NikhbNN k dkEk hENISNk ?[dk$$M lI Action 15 of Table 3.3-1, Reactor Trip System Instrumentation, allows the breaker to be bypassed for up to 4 hours for the purpose of performing R46 maintenance. The A hours 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 condition would have indicated inoperability, the channel should not be declared inoperable. The Auxiliary Feedwater (AFW) Suction Pressure-Low function must he OPERABLE in MODES L 2, and 3 to ensure a safety grade supply of water for the AEW System to maintain the steam generators as the heat sink for the reactor. R228 This function does not have to be OPEPABLE in MODES 5 and 6 because heat being generated in the reactor is removed via the Residual Heat Pemoval (RHR) System and does not require the steam generators as a heat sink. In MODE 4, AFW automatic suction transfer does not need to he OPEPABLE because PJi_,- will already he in operation, or sufficient time is available to place RFIR in operation to remove decay heat. This area is affected by Technical Specification 3/4.3 3 MONITORING INSTRUMENTATION Change 99-03 3/4.3.3.1 RADIATION MONITORING INSTRbnw AA11w Q P' The OPERABILITY of the radiation monitoring channels ensures that 1) e 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. c 3/4.3.3.2 MOVABLE INCCORE DETECTORS The OPEPABILITY 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 the reactor core. The OPEPABILITY of this system is demonstrated by irradiating each detector used and determining the acceptability of its voltage curve. For the purpose of measuring ED(X,, Y Z) or F (X, Y) a full incore flux map ! p214 ( is used. Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be l used in recalibraticn of the excore neutron flux, detection system, and full i incore flux maps or synnetric incore thimbles may be used for monitoring the ) QUADRANT POWER TILT PATIO when one Power Range Channel is inoperable. I l November 19, 1998 SEQUOYAH UNIT 2 B 3/4 3-2 Amendment Nos. 46, 72,
- 182, 214, 228 l
l
V l I a l ! s INSFRTI Response time may be verified by actual response time tests in any series of sequential, overlapping or total channel measurements, or by the summation of allocated sensor, signal processing and actuation logic response times with actual response time tests on the remainder of the channel. Allocations for sensor response times may be derived from: (1) historical records based on acceptable response time tests (hydraulic, noise, or power interrupt tests), (2) inplace, onsite, or offsite (e.g. vendor) test measurements, or (3) utilizing vendor engineering specifications. WCAP-13632-P-A Revision 2 (January 1996) " Elimination of Pressure Sensing Response Time Testing Requirements," provides the basis and methodology for using allocated sensor response times in the overall verification of the channel response time for specific sensors identified in the WCAP. Response time verification for other sensor types must be demonstrated by test. TVA has verified that the selected components at Sequoyah are the same Manufacturer and Model No. as evaluated in WCAPs 13632-P-A and 14036-P-A. WCAP-14036-P-A Revision 1 (October, 1998) " Elimination of Periodic Protection Channel Response Time Tests," provides the basis and methodology for using allocated signal processing and actuation logic response times in the overall verification of the protection system channel response time. The allocations for sensors, signal conditioning, and actuation logic response times must be verified prior to placing the component in operational service and reverified following maintenance that may adversely affect response time. Periodically, sensors, signal I conditioning and logic components are functionally tested. j l 1 l l l i l t I
c. ENCLOSURE 3 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT (SQN) UNITS 1 AND 2 PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE TS 99-08 REVISED PAGES I. AFFECTED PAGE LIST Unit 1 1-3 1-5 3/4 3-1 3/4 3-14 B 3/4 3-2 Unit 2 1-3 1-6 3/4 3-1 3/4 3-14 B 3/4 3-2 II. REVISED PAGES See attached. 1
4 4 DOSE EOUIVALENT I-131 1.11 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / lR159 gram) which alone would produce the same thyroid dose as the quantity and i isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites." E-AVERAGE DISINTEGRATION ENERGY 1.12 E shall be the average (weighted in proportion to the concentration of lR159 each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant. ENGINEERED SAFETY FEATURE (ESP) RESPONSE TIME l 1.13 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval lR159 from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously i l reviewed and approved by NRC. l FREOUENCY NOTATION 1.14 The FREQUENCY NOTATION specified for the performance of Surveillance lR159 Requirements shall correspond to the intervals defined in Table 1.2. GASEOUS RADWASTE TREATMENT SYSTEM 1.15 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed lR159 to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system an. providing for delay or holdup for the d purpose of reducing the total radioactivity prior to release to the j environment. IDENTIFIED LEAKAGE 1.16 IDENTIFIED LEAKAGE shall be: lR159 a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captvred and conducted to a sump or collecting tank, or l SEQUOYAH - UNIT 1 1-3 Amendment No. 12, 71, 155,
6 PRESSURE BOUNDARY LEAKAGE i !1.22 PRESSURE BOUNDARY LEAKAGE shall be leakage (ev:ept steam generator tube leakage) through a non-isolable fault in a Reactor Coo! ant System component body, pipe wall or vessel wall. PROCESS CONTROL' PROGRAM (PCEL i 1.23 DELETED R237 PURGE - PURGING 1.24 PUROE or PURGING.is the controlled process of discharging air or gas from a confinem<ent to maintain temperature, pressure, humidity, concentration or other operating condition,.in such a manner that replacement air or gas is required to purify the confinement. ) J OUADPANT POWER TILT RATIO 1.25 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower e.xcore detector calibrated R205 outputs, whichever is greater. RATED THERMAL POWER (RTP) 1.26 " RATED THERMAL POWER (RTP) shall be,a total reactor core heat transfer rate to the reactor coolant of 3411 MWt. l REACTOR TRIP SYSTEM (RTS) RESPONSE TIME' j i 1.27 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its (RTS) trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC. REPORTABLE EVENT 1.28 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50. 1 l SEQUOYAH - UNIT 1 1-5 Amendmer.t No. 12, 71, 141, 148, 155, 201, 233, 1 t.-
4 Y.' X 3/4.3-INSTRUMENTATION 3 .3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION ~ 3.3.1.1 As a minimum, the reactor trip system instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE. lR194 APPLICABILITY: As shown.in Table 3.3-1. ACTION: As shown in Table 3.3-1. SURVEILLANCE REQUIREMENTS 4.3.1 1.1 Each reactor trip system instrumentation channel and interlock shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL R16 CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Table 4.3-1. 4.3.1.1.2 'The. logic for the interlocks shall be demonstrated OPERABLE prior to ) each reactor-startup unless performed during the preceeding 92 days. The totel i interlock function shall be demonstrated OPERABLE at least once per 18 mon'.hs j during CHANNEL CALIBRATION testing of each channel affected by interlock operation. 4.3.1.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall be verified to be within its limit at least once per 18 months. Neutron detectors are exempt from response time testing. Each verification shall include at least one logic train such that both logic trains are verified at least once per 36-months and one channel per function such that all channels are verified at least once every N times 18 months where N is the total number of redundant channels jn a specific reactor trip function as shown in the " Total No. of Channels" column of. Table 3.3.1. i i i I l S2 QUO'JAH - be *1 3/4 3-1 Amendment Nos. 12, 190, l I
F l 6 l INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION l-L 3.3.2.1 The Engineered Safety Feature Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4. lR194 APPLICABILITY: As shown in Table 3.3-3. ACTION: a. With an ESEAS instrumentation channel or interlock trip setpoint less conservative than the value shown in the Allowable values column of Table 3.3-4, declare the channel inoperable and apply the applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint value. b. With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3-3. SURVEILLANCE REQUIREMENTS i 4.3.2.1.1 Each ESFAS instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL 1 FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Table 4.3-2. 4.3.2.1.2 The logic for the interlocks shall be demonstrated OPERABLE during the automatic actuation logic test. The total interlock function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by interlock operation. 4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be verified to be within the' limit at least once per 18 months. Each verification shall include at least one logic train such that both logic trains are verified at least once per 36 months and one channel per function such that all channels are verified at least once per N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the " Total No. of Channels" Column of Table 3.3-3. l SEQUOYAH - UNIT 1 3/4 3-14 Amendment No. 190, i I i
[i i j l INSTRUMENTATION i BASES The verification of response time at the specified frequencies provides assurance that the reactor trip and ESF actuation function associated with each channel is completed within the time limit assumed in the safety 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. lR194 Response time may be verified by actual response time tests in any series of sequential, overlapping or total channel measurements, or by the summation of allocated sensor, signal processing and actuation logic response times with actual response time tests on the remainder of the channel. Allocations for. 1 sensor response times may be derived from: (1) historical records based on acceptable response time tests (hydraulic, noise, or power interrupt tests), (2) inplace, onsite, or offsite (e.g., vendor) test measurements, or (3) utilizing vendor engineering specifications. WCAP-13632-P-A Revision 2 (January 1996), " Elimination of Pressure Sensing Response Time Testing Requirements," provides the basis and methodology for using allocated sensor i response times in the overall verification of the channel response time for specific sensors identified in the WCAP. Response time verification for other sensor types must be demonstrated by test. TVA has verified that the selected components at Sequoyah are the same Manufacturer and Model No. as evaluated in WCAPs 13632-P-A and 14036-P-A. WCAP-14036-P-A Revision 1 (October 1998), " Elimination of Periodic Protection Channel Response Time Tests," provides the basis and methodology for using allocated signal processing and actuation logic response times in the overall verification of the protection system channel response time. The allocations for sensors, signal conditioning, and actuation i logic response times must be verified prior to placing the component in operational service and reverified following maintenance that may adversely affect response time. Periodically, sensors, signal conditioning and logic components are functionally tested. Action 15 of Table 3.3-1, Reactor Trip System Instrumentation, allows the breaker to be bypassed for up to 4 hours for the purpose of performing 4 maintenance. The 4 hours 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 function of the channel. If the channel is tripped for testing and no other condition would have indicated inoperability, the channel should not be BR-9 declared inoperable. The Auxiliary Feedwater (AFW) Suction Pressure-Low function must be OPERABLE in MODES 1, 2, and 3 to ensure a safety grade supply of water for the AFW System to maintain the steam generators as the heat sink for the reactor. This function does not have to be OPERABLE in MODES 5 and 6 because heat being R242 generated in the reactor is removed via the Residual Heat Removal (RHR) System and does not require the steam generators as a heat sink. In MODE 4, AFW automatic suction transfer does not need to be OPERABLE because RHR will already be in operation, or sufficient time is available to place RHR in operation to remove decay heat. SEQUOYAH - UNIT 1 B 3/4 3-2 Amendment No. 54, 190, 223,
- 238,
1 a n:,* i -i INSIEUMENTATION BASES 3/4.3.3 MONITORINO INSTRUMENTATION j l 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of-the radiation monitoring channels ensures that 1) the rediation levels are continually measured in the areas served by the individual 1 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 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 curves For the purpose of measuring F (X,Y,Z) or Fa (X,Y) a full incore flux map lR227 n is used. Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in'recalibration of the excore neutron flux detection system, and full incere flux maps or symmetric incore thimbles may be used for monitoring the QUADRANT POWER TILT RATIO when one Power Range Channel is inoperable. ) 1 I 1 Novembcr 19, 1998 SEQUOYAH UNIT 1 B 3/4 3-2a Amendment No, 54, 190, 223, 238 i
I 4 o DEFINITIONS DOSE EOUIVALENT I-131 1.11 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / lR146 gram) which alone would produce the same thyroid dose as the quantity and iso-topic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites." 5 - AVERAGE DISINTEGRATION ENERGY 1.12 E shall be the average (weighted in proportion to the concentration of lR146 each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant. ENGINEERED SAFETY FEATURE (ESP) RESPONSE TIME l 1.13 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval lR146 from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge I pressures reach their required values, etc.). Times shall include diesel { generator starting and sequence loading delays where applicable. The response time may be measured by means of any series of sequential, overlapping, or j total steps so that the entire response time is measured. In lieu of ) measurement, response time may be verified for selected components provided J that the components and the methodology for verification have been previously reviewed and approved by the NRC. nm FREOUENCY NOTATION 1.14 The FREQUENCY NOTATION specified for the performance of Surveillance lR146 Requirements shall correspond to the intervals defined in Table 1.2. GASEOUS RADWASTE TREATMENT SYSTEM 1.15 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed lR146 to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the 3 purpose of reducing the total radioactivity prior to release to the environment. l l l l l l i f SEQUOYAH - UNIT 2 1-3 Amendment Nos. 63, 146,
1 .'+ . DEFINITIONS RATED THERMAL POWER (RTP) lR132 1.26 RATED THERMAL POWER (RTP) shall be a total reactor core heat transfer lR146 rate to the reactor coolant of 3411 MWt. REACTOR TRIP SYSTEM (RTS) RESPONSE TIME l 1.27 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from lR146 when the monitored parameter exceeds its (RTS) trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by NRC. REPORTABLE EVENT 1.28 A REPORTABLE EVENT shall be any of those conditions specified in Section lR146 50.73 to 10 CFR Part 50. SHIELD BUILDING INTEGRIH 1.29 SHIELD BUILDING INTEGRITY shall exist when: lR146 a. The door in each access cpening is closed except when the access j opening is being used for normal transit entry and exit. b. The emergency gas treatment system is OPERABLE. c. The sealing mechanism associated with each penetration (e.g., welds, bellows or 0-rings) is OPERABLE. SHUTDOWN MARGIN 1.30 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which lR146 the reactor is suberitical or would be suberitical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn. SITE BOUNDARY 1.31 The SITE BOUNDARY shall be that line beyond which the land is not owned, lR146 leased, or otherwise controlled by the licensee (see figure 5.1-1). SEQUOYAH - UNIT 2 1-6 Amendment Nos. 63, 132, 146,
J i + O I 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor trip system instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE. lR182 j APPLICABILITY: As shown in Table 3.3-1. ACTION: As shown in Table 3.3-1. SURVEILLANCE REQUIREMENTS 4.3.1.1.1 Each reactor trip system instrumentation channel and interlock shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL 3 CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Table 4.3-1. 4.3.1.1.2 The logic for the interlocks shall be demonstrated OPERABLE prior to each reactor startup unless performed during the preceeding 92 days. The total interlock function chall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by interlock operation. 4.3.1.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall be verified to be within its limit at least once per 18 months. Neutron detectors are exempt from response time testing. Each verification shall include at least one logic train such that both logic trains are verified at least once per 36 months and one channel per function such that all channels are verified at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the " Total No. of Channels" column of Table 3.3.1. i l l i 4 \\ SEQUOYAH - UNIT 2 3/4 3-1 Amendment No. 182, l
n b i INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION i j 3.3.2 The Engineered Safety Feature Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their trip i setpoints set consistent with the values shown in the Trip setpoint column of Table 3.3-4. lR182 j i APPLICABILITY: As shown in Table 3.3-3. ACTIQN: a. With an ESFAS instrumentation channel or interlock trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint value. b. With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3-3. SURVEILLANCE REQUIREMENTS 4.3.2.1.1 Each ESFAS instrumentation channel and interlock shall be demon-strated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Table 4.3-2. 4.3.2.1.2 The logic for the interlocks shall be demonstrated OPERABLE during the automatic actuation logic test. The total interlock function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by interlock operation. 4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be verified to be within the limit at least once per 18 months. Each verification shall include at least one logic train such that both logic trains are verified at least once per 36 months and one channel per function such that all channels are verified at least once per N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the " Total No. of Channels" Column of Table 3.3-3. I l SEQUOYAH - UNIT 2 3/4 3-14 Amendment No. 182,
p a e INSTRUMENTATION BASES REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION The verification of response time at the specified frequencies provides assurance that the reactor trip and the ESF actuation function associated with each channel is completed within the time limit assumed in the safety 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. lR182 Response time may be verified by actual responce time tests in any series of sequential, overlapping or total channel measurements, or by the summation of allocated sensor, signal processing and actuation logic response times with actual response time tests on the remainder of the channel. Allocations for sensor response times may be derived from: (1) historical records based on acceptable response time tests (hydraulic, noise, or power i..terrupt tests), (2) inplace, onsite, or offsite (e.g., vendor) test measurements, or (3) utilizing vendor engineering specifications. WCAP-13632-P-A Revision 2 (January 1996), " Elimination of Pressure Sensing Response Time Testing Requirements," provides the basis and methodology for using allocated sensor response times in the overall verification of the channel response time for specific sensors identified in the WCAP. Response time verification for other sensor types must be demonstrated by test. TVA has verified that the selected components at Sequoyah are the same Manufacturer and Model No. as evaluated in WCAPs 13632-P-A and 14036-P-A. WCAP-14036-P-A Revision 1 (October 1998), ) " Elimination of Periodic Protection Channel Response Time Tests," provides the basis and methodology for using allocated signal processing and actuation logic response times in the overall verification of the protection system channel response time. The allocations for sensors, signal conditioning, and actuation logic response times must be verified prior to placing the component in operational service and reverified following maintenance that may adversely i affect response time. Periodically, sensors, signal conditioning and logic components are functionally tested. Action 15 of Table 3.3-1, Reactor Trip System Instrumentation, allows the breaker to be bypassed for up to 4 hours for the purpose of performing maintenance. The 4 hotrs is based on a Westinghouse analysis performed in R46 WCAP-10271, Supplement 1, which determines bypass breaker availability. The placing of a cha.'nel in the trip condition provides the safety function of the channel. If the channel is tripped for testing and no other condition would have indicated inoperability, the channel should not be BR-10 declared inoperable. The Auxiliary Feedwater (AFW) Suction Pressure-Low function must be OPERABLE in MODES 1, 2, and 3 to ensure a safety grade supply of water for the AFW System to maintain the steam genarators as the heat sink for the reactor. Thie function does not.have to be OPERABLE in MODES 5 and 6 because heat being R228 generated in the reactor is removed via the Residual Heat Removal (RHR) System and does not require the steam generators as a heat sink. In MODE 4, AFW automatic suction transfer does not need to be OPERABLE because RHR will already be in operation, or sufficient time is available to place RHR in operation to remove decay heat. SEQUOYAH - UNIT 2 B 3/4 3-2 Amendment No. 46, 72, 182, 214,
- 228, f
n- _y ! {f t O INSTRUMENTATION BASES-3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATIQH-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 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 (X,Y,Z) or Fan (X, Y) a full incore flux map lR214 n is used. Quarter-core flux maps, as defined in WCAP-8G48, 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. t l l ] l \\ l SEQUOYAH - UNIT 2 B 3/4 3-2a Amendment No. 46, 72, 182, 214, 228 j I}}