ML20039B795
| ML20039B795 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 12/15/1981 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20039B792 | List: |
| References | |
| NUDOCS 8112230563 | |
| Download: ML20039B795 (119) | |
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~I l ENCLOSURE 1 l .f i '.' PROPOSED TECHNICAL SPECIFICATION CHANGES
.l FOR SEQUOYAH NUCLEAR PLANT UNIT 1
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) - 1. '
1.0 DEFINITIONS DEFINED TERMS The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications. ACTION 1.1 ACTION shall be that part of a Specffication which prescribes remedial measures required under designated conditions. AXIAL FLUX DIFFERENCE 1.2 AXIAL FLUX DIFFERENCE shall be the difference in normali1ad flux signals between the top and bottom halves of a two section excore neutron detectors. CHANNEL CALIBRATION 1.3 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the cha nel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALI-BRATION shall encompass the entire channel. including the' sensor and alarm (<6f and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. . The
. CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.
CHANNEL CHECK 1.4 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same_ parameter. I CHANNEL' FUNCTIONAL TEST
- 1. 5 A CHANNEL FUNCTIONAL TEST shall be: -
- a. Analog channels - the injection of a simulated signal into the channel as close to the sensor as' practicable to verify OPERABILITY including alarm and/or trip functions.
i b. Bistable channels - the injection of a simulated signal into the
; sensor to verify OPERABILITY including alarm and/or trip functions.
SEQUOYAH - UNIT 1 I 1-1 m
1 , DEFINITIONS CONTAINMENT INTEGRITY
}
1.6 CONTAINMENT INTEGRITY shall exist when:
- a. All penetrations required to be closed during accident conditions are either:
- 1) Capable of being closed by an OPERA 6LE containment automatic isolation valve system, or 1
- 2) Closed by manual valves, blind flanges, or deactivated auto- j matic valves secured in their closed positions, except as '
provided in Table 3.6-2 of Specification 3.6.3.
- b. All equipment hatches are closed and sealed, !
- c. Each air lock is OPERABLE pursuant to Specification 3.6.1.3, l
- d. The containment leakage rates are witt' '.he limits of Specification 3.6.1.2, and
- e. The sealing mechanism associated with each penetration:(e.g., welds, bellows or 0-rings) is OPERABLE.
CONTROLLED LEAXAGE .
- 1. 7 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.
CORE ALTERATION -
- 1.8 CORE ALTERATION shall be the movement or manipulation of any component j within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of j movement of a component to a safe conservative position.
! DOSE EQUIVALENT I-131 i l.9 00SE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie /' - l l 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."
.c;z (hei SEQUOYAH - UNIT 1 1-2 J'
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1 DEFINITIONS 1 l - AVERAGE DISINTEGRATION ENERGY [ 1.10 E shall be the average (weighted in proportion to the concentration of r 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 RESPONSE TIME 1.11 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. FREOUENCY NOTATION 1.12 The FREQUENCY NOTATION specified for the performance of Surveillance
.h -Requirements shall correspond to the intervals defined in Table-1.2.
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GASEOUS RADWASTE TREATMENT SYSTEM 1.13 A GASEOUS RA0 WASTE TREATMENT SYSTEM is any cysten designert and installed to reduce radioactive gaseous effluents by collecting primary toolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment. IDENTIFIED LEAKAGE 1.14 IDENTIFIED LEAKAGE shall be: l'
- a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve _ packing leaks that are captured and conducted to a sump or collecting tank, or -
- b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the i operation of leahge detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or
- c. Reactor coolant system leakage through a steam generator to the.
secondary system. SEQUOYAH - UNIT 1 1-3 1 l'
1 . c OEFINITIONS
} OFFSITE DOSE CALCULATION MANUAL
. l.15 .The OFFSITE DOSE CALCULATION MANUAL shall contain the methodology and.
parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints. It shall also contain the radiological environmental monitoring program. OPERABLE - OPERABILITY 1.16 A systr. subsystem, train, or component or device shall be OPERABLE or have OPERAILITY when it is capable of performing its specified function (s), and when all necessary attendant instrumentation, controls, a normal and an. emerger.cy electrical power source, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, com-ponent or device to perform its function (s) are also capable of performing their related support function (s). OPERATIONAL MODE - MODE 1.17 An OPERATIONAL MODE (i.e., MODE) shall correspond-to any one inclusive ccmbination of core reactivity condition, power level and average reactor coolant temperature specified in-Table 1.1. - PHYSICS TESTS 1.18 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14.0 of the FSAR, 2) authorized under the provisions of
'10 CFR 50.59, or 3) otherwise approved by the Commission.
PRESSURE BOUNDARY LEAKAGE
'..19 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) 1.20 The PROCESS CONTROL PROGRAM shall contain the sampling, analysis, and formulation determination by which SOLIDIFICATION of radioactive wastes from liquid systems is assured. 1 l s SEQUOYAH - UNIT 1 , 1-4 e v y m u , y
-1 ~
DEFINITIONS
.!if PURGE -PURGING ,
i 1.21 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration
.., or other operating condition, .in sucr. a manner that replacement air or gas is required to purify the confinement.
QUADRANT POWER TILT RATIO .~ 1.22 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper exec.v detector cali-brated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, which-ever is greater. With one excore detector inoperable, i.he remaining three detectors shall be used for computing the average. RATED THERMAL POWER 1.23 RATED THERMAL POWER ~; hall be a total reactor core heat transfer rate to the reactor coolant of 3411 MWt.
.f@ = REACTOR TRIP SYSTEM RESPONSE TIME 1.24 The REACTOR TRIP SYSTEM RESPONSE TIME shall 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 OCCURRENCE 1.25 A REPORTABLE OCCURRENCE.shall be any. of those conditions specified in Speci fications 6.9.1.12 and 6.1.1.13. SHIELO BUILDING INTEGRITY 1.26 SHIELD BUILDING INTEGRITY shall aist when: l a. The door in each access openirg is closed except when the access opening is being used for norn.al 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) ic 9PERABLE.
a SEQUOYAH - UNIT 1
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DEFINITIONS h SHUTCOWN MARGIN I 5 1.27 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which 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. - SOLIDIFICATION 1.28 SOLIDIFICATION shall be the conversion of radioac'tive wastes from liquid systems to a uniformly distributed, monolithic, immobilized solid with definite volume and shape, bacnded by a stable surface of distinct outline on all sides 2 (free-standing). SOURCE CHECK 1.28 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is ec osed to a radioactive source. STAGGERED TEST BASIS -
- ~A r 1.29 A STAGGERED TEST BASIS shall consist of: -
- a. . A test schedule for n systems, subsystems, trains or other designated compon'nts e obtained by dividing the specified test interval into n equal subintervals,
- b. The testing of one system, subsystem, train or other designated component at the beginning of each subinterval.
9 THERMAL POWER 1.31 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. UNIDENTIFIED LEAKAGE l 1.32 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE l or CONTROLLED LEAKAGE. l b[
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! SEQUOYAH - UNIT 1 , 1-6 4 l i tc l lI . lt
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DEFINITIONS C :'
) VENTILATION EXHAUST TREATMENT SYSTEM 1.33 A VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or partic-ulates from the gasecus exhaust stream prior to the release to the environment, engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYST.EM components.
VENTING 1.34 VENTING is the controlled process of dischargirg air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other cperating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process. 4 0 SEQUOYAH - UNIT 1 1-7 L
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0.0 0.1 f2 0.3 0.4 0.5 0.6 0.7 0.8 0. 9 1.0 1.1 1.2 { FRACTION OF RATED THERMAL POWER 1 F jurs 2.1 1. Reactor Core Safety Limit-Four Loops irt Creration l l l l l e
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SEQUOYAH - UNIT 1 2-2 e t 1 k
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 1 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS
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' 2. 2.1 The reactor trip system instrumentation and interlocks setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.
APPLICABILITY: As shown for each channel in Table 3.3-1. ACTION: 4
) With a reactor trip system instrumentation or interlock setpoint less'conserva-tive than the value shown in the Allowable Values column of Table 2.2-1,' declare the channel inoperable and apply the applicable ACTION statement requirement of.
4 Specification 3.3.1 until the channel is restored to OPERABLE status with its ] trip setpoint adjusted consistent with the Trip Setpoint value. i 4 a l . l-i 4 SEQUOYAH - UNIT 1 2-4 l
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g ' TABLE 2.2-1 (Continued) E g REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINy N FUNCTIONAL UNIT TRIP SETPOINT z ALLOWABLE VALUES U 21. Turbine Impulse Chamber Pressure - ,$ 10% Turbine Impulse
- (P-13) Input to Low Power Reactor Trips i 11% Turbine Imruise Pressure Equivalent Pressure Equivalent Block P-7
- 22. Power Range Neutron Flux (P-8) Low < 35% of RATED < 36% of RATED Reactor Coolant Loop Flow, and THERMAL POWER THERMAL POWER Reactor Trip _
- 23. ~Poiler Range'Nsutr6n F16FF(P-10) ~ i
> 10% of RATED > 9% of RATED Enable block of Source, Intermediate, THERMAL POWER THERMAL POWER and Power Ran0e (low setpoint) reactor Trips ,
'? 24. Reactor Trip P-4 Not Applicable Not Applicable
- 25. ; Power Range Neutron Flux - (P-9) - < 50% of RATED < 51% of RATED
.. Blocks Reactor Trip for Turbine THERMAL POWEE '
THERMAL POWER Trjp Below 50% Rated Power ' NOTATION NOTE 1: Overtemperature AT ( I
- 5 I l+tEj ) 5 AT, { K) - K2 (1 + 1 362 )[T(I*I6~ )-T'] + K3 (P-P') - f j(AI)}
4 R I where:
,l#g I+I l- = Lag compensator on measured'AT
- 3 g.. '
'"E I j = Time csnstants utilized in the lag compensator for AT3 'l = 2 secs.
AT, = Indicated AT at RATED THERMAL POWER K y $ 1.14. K 2, = 0.009 - .-
I i REACTIVITY CONTROL SYSTEMS 2 I SHUTDOWN MARGIN - T,yg Less Than or Equal to 200'F 3 , l ' LIMITING CONDITION FOR OPERATION _ 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to 1.0% delta k/k. APPLICABILITY: MODE 5. t ACTION: l i I With the SHUTDOWN MARGIN less than 1.0% delta k/k, immediately initiate and t continue boration at greater than or equal to 10 gpm of a solution containing
'j greater than or equal to 20,00 ppm or equivalent until the. required l
. SHUTDOWN MARGIN is restored. {
s SURVEILLANCE REQUIREMENTS i to i b d ak l l a. Within one hour after detection of an inoperable control rod (s) and j~ at least once per 12 hours thereafter while the rod (s) is inoperable.
; If the inoperable control rod is immovable or untrippable, the
+ SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s).
- b. At least once per 24 hours by consideration of the following factors: ,
- 1. Reactor coolant system boron concentration,
- 2. Control rod position, .
- 3. Reactor coolant system average temperature,
- 4. Fuel burnup based on gross thermal energy generation, --
- 5. Xenon concentration, and
- 6. Samaritz concentration.
SEQUOYAH - UNIT 1 3/4 1-3
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REAL NVITY.
-. CONTROL SYSTEMS FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the following three boron injection flow paths shall be OPERABLE:
- a. The flow path from tne boric acid tanks via a boric acid transfer pump and a charging pump to the Reactor Coolant System.
- b. Two flow paths from the refueling water storage tank via charging
' pumps to the Reactor Coolant System. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With only one of the above required boron injection flow paths to the Reactor Coolant System OPERABLE, restore at least two boron injection flow paths to the Reactor Coolant System to OPERABLE status within 72 hours or be in at least HOT STANDBY and borated to a SHUTOOWN MARGIN equivalent to at least 1% delta k/k at 200*F within the next 6 hours; restore at least two flow paths to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours. SURVEILLANCE REQUIREMENTS i 4.1.2.2 At least two of the above required flow paths shall be demonstrated OPERABLE:
- a. At least once per 7 days by verifying that the temperature of the best traced portion of the flow path from the' boric acid tanks is greater than or equal to 145*F when it is a required water source.
I
- b. At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path that,is not locked, sealed, ~~
or otherwise secured in position, is in its correct position.
- c. At least once per 18 months during shutdown by verifying that each automatic valve in the flow path actuates to its correct position on a safety injection test signal.
! d. At least once per 18 months by verifying that the flow path i required by specification 3.1.2.2a delivers at least 10 gpm to the Reactor Coolant System.
SEQUOYAH - UNIT 1 3/4 1-8
POWER DISTRIBUTION LIMITS ACTION: (Continued).
- b. THERMAL POWER shall not be increased above 90% of RATED THERMAL POWER unless the indicated AFD is within the 15% target band dnd ACTION a.2.a)l), above has been satisfied.
- c. THERMAL POWER shall not be increased above 50% of P \TED THERMAI.
POWER unless the indicated AFD has not been outside of the + 5%- target band for more than I hour penalty deviation cumulative during the previous 24 hours. Power increases'above 50% of rated THERMAL POWER do not require being within the target band provided the accumulative penalty deviation is not violated. SURVEILLANCE REOUIREMENTS 4.2.1.1 The indicated AXIAL FLUX DIFFERENCE shall be determined to be within its limits during POWER OPERATION above 15% of RATED THERMAL POWER by:
- a. Monitoring the indicated AFD for each OPERABLE excore channel:
- 1. At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and
- 2. At least once per hour for the first 24 hours after restoring the AFD Monitor Alarm to OPERABLE status.
- b. Monitoring and logging the indicated AXIAL FLUX DIFFERENCE for each OPERABLE excore channel at least once per hour for the first 24 hours and at least once per 30 minutes thereafter, when the AXIAL FLUX DIFFERENCE Monitor Alarm is inoperable. The logged values of the indicated AXIAL FLUX DIFFERENCE shall be assumed to exist during the interval preceding each logging.
4.2.1.2 The indicated AFD shall be considered outside of its 1 5% target band when at least 2 OPERABLE excore channels. are indicating the AFD to be outside the target band. Penalty deviation outside of the 1 5% target band snall be accumulated on a tir.e basis of: ' a. One minute penalty deviation for each one minute of POWER OPERATION outside of the target band at THERMAL POWER levels equal to or above ! 50% of RATED THERMAL POWER, and . I
- b. One-half minute penalty deviation for each one minute of POWER OPERATION outside of the target band at THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER.
4.2.1.3 The target flux difference of each OPERABLE excore channel shall be determined by measurement at least once per 92 Effective Full Power Days. .The provisions of Specification 4.0.4 are not app!! cable. SEQUOYAH - UNIT 1 3/4 2-2 f~ I L
I POWER DISTRIBUTION LIMITS 1 SURVEILLANCE RE';JIREMENTS (Continued) At least once per 31 EFPD, whichever occurs firsts
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b) RTP
, 2. When the F is less than or equal to the F limit for the appropriate measured core plane, additional power distribution maps shall be taken and F compared to F RTP and F at least x
~
once per 31 EFPD.
*- The F limit for RATED THERM 5L POWER (F P) shall be provided for all core planes containing bank "D" control rods and all unrodded core planes in a Radial Peaking Factor Limit .hport per Specification 6.9.1.14.
i
- f. The F limits of e, above, are not applicable in the following core plane regions as measured in percent of core height from the bottom of the fuel:
- 1. Lower. core region from 0 to 15%, inclusive.
- 2. Upper core region from 85 to 100%, inclusive.
- 3. Grid plane regions at 17.8 1 2%, 32.1 1 2%, 46.4 1 2%, 60.6 1 ,
2% and 74.9 1 2%, inclusive.
- 4. Core plane regions within 1 2% of core height (1 2.88 inches) about the bank demand position of the bank "D" control rods. - -
- g. With F exceeding F , the effects of.F on Fq (Z) shall be evaluated to determine if F q (Z) i's within its limit.
e SEQUOYAH - UNIT 1 3/4 2 ' e
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l l 1 i t i SEQUOYAH - UNIT 1 3/4 2-9 ) l
POWER DISTRIBUTION LIMITS I, 3/4.2.3 RCS FLOWRATE Ah! R LIMITING CONDITION FOR OPERATION ' 3.2.3 The combination of indicated Reactor Coolant System (RCS) total flow rate and R;, R2 shall be maintained within the regions of allowable operation shown on Figure 3.2-3 for 4 loop operation: Where: N F
- a. = AH ,
R) 1.49 [1.0 + 0.2 (1.0 - P)]
- b. R
- R-2 [1 - R8P (Bu)] ,
THERMAL POWER
- c. P =
RATED THERMAL POWER '
- d. F =
Measureo values of F H obtained by using the movable incore detectors to obtain a power distribution map. l The measured values of F H shall be used to calculate R since Figure 3.2-3 includes me6surement uncertainties of 3.5% for flow ~and 4% for incore measurement of F H l
- e. !
RBP (Bu) =
' Rod Bow Penalty as a function of region average burnup as shown in Figure 3.2-4, where a region is defined as those assembifes with the same loading date (reloads) or enrichment (first core),
and APPLICABILITY: MODE 1 ACTION: With the combination of RCS total flow rate ar.d R), R 2 ptside the regions l of acceptable operation shown on Figure 3.2-3: *
- a. Within 2 hours:
- 1. Either restore the combination of RCS total flow rate !
and R j ,R2 to within the above limits, or. !
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High trip setpoint to less than or equal.to 55%'of RATED THERMAL POWER within the next 4 hours.
SEQUOYAH - UNIT 1 3/4 2-10
l I
- 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION i -
! LIMITING CONDITION FOR OPERATION i 3.3.1.1 As a minimum, the reactor trip system instrumentation channels and ! interlocks o' Table 3.3-1 shall be OPERABLE with RESPONSE TIMES as shown in l Table 3.3-2. APPLICABILITY: As shown in Table 3.3-1. ACTION: I
; 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 CHANNEL CHECK, CHANNEL 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 shall 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 demonstrated to be within its 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 are tested 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. SEQUOYAH - UNIT 1 3/4 3-1
e - m TABLE 3.3-1 E g REACTOR TRIP SYSTEM INSTRUMENTATION
. ;;E z
- MINIMUM E TOTAL NO. CilANNELS CHANNELS APPLICABLE q FilNCTIONAL UNIT OF CilANNELS TO TRIP OPERABLE MODES ACTION
~
- 1. Manual Reactor Trip 2 1 2 1, 2, and
- 1
- 2. Power Range, Neutron Flux 4 2 3 1, 2 2
- 3. Power Range, Neutron Flux 4 2 3 1, 2 2
lligh Positive Rate'
- 4. Power Range, Neutron Flux, 4 #
2 3 1, 2. 2 High Negative Rate RA 5. Intermediate Range, Neutron Flux 2 2 w 1 1, 2, and
- 3 E 6. Source Range, Neutron Flux A. Startup 2 1 2 2,,, and
- 4 B. Shutdown 2 0 1 3, 4 and 5 >
5 .
- 7. Overtemperature Delta T .
Four Loop Operation 4 2 3 1, 2 6 Three Loop Operation 4 1** 3 1, 2 9
- 8. Overpower Delta T Four Loop Operation 4 2 3 1, 2 6 Three Loop Operation 4 1** 3 1, 2 9
- 9. Pres,surizer Pressure-Low 4 2 3 1, 2 6
- 10. Pressurizer Pressure--High 4 2
, 1, 2 6
- 11. Pressurizer Water Level--High 3 2
~
2 1, 2 7* e
TABLE 3.3-1 (Continued) M 2 REACTOR TRIP SYSTEM ItiSTRUMEtlTATI0tt S! E MINIMUM i TOTAL NO. CHAtitlELS CHAtitlELS APPLICABLE cE MODES ACTION FUNCTI0tlAL UNIT OF CHANNELS TO TRIP OPERABLE 5
- 19. Safety Injection Input 2 1, 2 12 from ESF 2 1 2 1, 2, and
- 12
- 20. Reactor Trip Breakers 2 1 2 1, 2, and
- 12
- 21. Automatic Trip Logic 2 1 Reactor Trip System Interlocks
~
22. A. Intermediate Range 2, and* 8a 2 2 Neutron Flux P-6 1 M B. Power Range Neutron 2 3 1 8b Flux - P-7 4 Y .C. ~ Power Range'Noutron 2 3 1 8e Flux - P-8 4 D. Power Range Neutron 1, 2 8d 4 2 3 ,
'I_ Flux - P-10 E. Turbine 1mpulse Chamber 2 1 2 1 Bb Pressure - P ,
~
F. Power Range Neutron 4 2 3 1 8e Flux - P-9 G. keactor Trip - P-4 2 ~1 2 1, 2, ar.d* 14
'INSTRUMENTATTON TABLE 3.3-1 (Continued)
TABLE NOTATION With the reactor trip system breakers in the closed position and the control , AA rod drive system capable of rod withdrawal, and fuel in the reactor vessel. [ The channel (s) associated with the protective functions derived from the out of service Reactor Crolant Loop shall be placed in the tripped condition. . The provisions of S' p ecification 3.0.4 are not applicable. I High voltage to detector may be de-energized above the P-6 (Block of Source
- - Range Reactor Trip) setpoint.
~I L
ACTION STATEMENTS s ACTION 1 - With the number of channels OPERABLE one less than required by the Minimum' Channels OPERABLE requirement, restore the inoperable.' channel to OPERABLE status within 48 hours or be in HOT STANDBY within the next 6 hours and/or open the reactor trip breakers. ACTION 2--
'With the number of OPERABLE channels one less than the Total
Number of Channels, STARTUP and POWER OPERATION may proceed proVided the following conditions are satisfied:
- a. The inoperable channel is placed in the tripped condition within 1 hour.
~
- b. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.1.1.1.
- c. Either, THERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL and the Power Range, Neutron Flux high trip reduced to less than or equal to 85% of RATED THERMAL POWER within 4 hours; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours.
d. The QUADRANT POWER TILT RATIO, as indicated by the remaining three detectors is verified consistent with the normalized symmetric power distribution obtained by using the movable ~ incore detectors in the four pairs of symmetric thimble I lccations at least once per 12 hours when THERMAL POWER is greater than 75% of RATED THERMAL POWER.
~
I l l i SEQUOYAH - UNIT 1 3/4 3-5 s I t a'-W E
o . [ INSTRUMENTATION TABLE 3.3-1 (Continued)
, .Y
. ACTION 8 - With less than the flinimum Number of Channels OPERABLE, declare.
the interlock inoperable and verify-that all af fected channels-of the functions listed below are OPERABLE or apply the appro-priate ACTION statement (s) for those functions. Functions to i be evaluated are:
- a. Source Range Reactor Trip.
l. l' b. Reactor Trip , Low Reactor Coolant loop Flow (2 loops) Undervoltage Underfrequency-Pressurizer Low Pressure Pressurize'r High Level
- c. Reactor Trip Low Reactor Coolant Loop Flow (1 loop)
- d. ' Reactor Trip Intermediate Range Low Power Range Source Range
- e. Reactor Trip Turbine Trip ACTION 9 - With a channel associated with an operating loop inoperable, restore the inoperable channel to OPERABLE status within 2 hours or be in HOT STANDBY within the next'6 hours; however, one channel
- associated with an operating loop may be bypassed for up to
- 2. hours for surveillance testing per Specification 4.3.1.~l.1.
ACTION 10 - With one channel inoperable, restore the inoperable channel to-OPERABLE status within 2 hours or reduce THERMAL POWER to below. the P-8 (Block Low Reactor Coolant Pump Flow) setpoint bresker within the next 2 hours. Operation below the P-8 (Block of Low Reactor Coolant Pump Flow) setpoint breaker may continue pursuant to ACTION ~11.
-ACTION'l'1 - With less than the Minimum Number of Channels OPERABLE, operation may continue provided'the inoperable channel is'placed in the
~
tripped condition within 1 hour. - ACTION 12 - With the number of channels OPERABLE one'less than required by
~
i the Minimum Channels OPERABLE requirement, be in at least-HOT STANDBY within 6 hours; however, one channel may be bypassed. for.up to 1 hour for surveillance testing per Specification 4.3.1.1.1 provided the other channel'is OPERABLE. t SEQUOYAH - UNIT 1 3/4 3-7 ,
d TABLE 3.3-2 (Continued) REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES S! E-FUNCTIONAL UNIT RESPONSE TIME
!! -13. Loss of Flow - Two ?. oops
-4 (Above P-7 and below P-8) $ 1.0 seconds e
- 14. Fair. Steam Generator Water Level--
Law-Low $ 2.0 seconds
- 15. Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level NOT APPLICABLE
- 16. Undervoltage-Reactor Coolant Pumps $ 1.2 s'econds
,, 17. Underfrequency-Reactor Coolant Pumps 5 0.6 seconds .s
((
- 18. Turbine Trip U$ . A. Low Fluid Oil Pressure NOT APPLICABLE B. Turbine stop Valve NOT APPLICABLE
- 19. Safety Injection Input from ESF NOT APPLICABLE
- 20. Reactor Trip Breakers NOT APPLICABLE
- 21. Automatic Trip Logic 'NOT APPLICABLE
- 22. Reactor Trip System Interlocks NOT APPLICABLE a
9 e a
vi , TABLE 4.3-1 ES . E5 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 5
- c
' CHANNEL H0 DES 'IN WiiICH Ei .
CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CllECK CALIBRA110f] TEST REQUIRED
- 1. Manual Reactor Trip N.A. N.A 5/U(1) 1, 2, and *
- 2. Power Range, Neutron Flux S D(2),.M(3) M 1, 2 and Q(6) .
1
- 3. Power Range, Neutron Flux, N.A. R(6) M 1, 2 High Positive Rate
- 4. Power Range, Neutron Flux, 0.A. R(6) M 1, 2 High Negative Rate ,
c, 5. Intermediate Range, S R(6) S/U(1) 1, 2, and *
]; Neutron Flux u
2, 6. . Source Range, Neutron' Flux 'S(7) .R(6) M and S/U(1) . 2, 3, 4, 5, and * - a
- 7. Overtemperature Delta T S R M ,
1, 2
- 8. Overpower Delta T S R H 1, 2
- 9. Pressurizer Pressure--Low S R H 1, 2
- 10. Pressurizer Pressure--High S R H 1, 2
- 11. Pressurizer' Water Level--High S R H 1, 2
- 12. Loss of Flow - Single Loop S R M 1
- 13. Loss of Flow - Two Loops S R N.A. 1
m TABLE 4.3-l'(Continued) REACTOR TRIP SYSTEM INSTRUMENTATID'N 50RVEII.LAtiCE REQUIREMENTS g , x
' CHANNEL~ H0 DES IN WHICH E
CHANNEL CHANNEL FUNCTIONAL ~ SURVEILLANCE - 3 FUNCTIO?!AL UNIT CHECK CALIBRATION TEST ' REQUIRED
~
- 14. Main. Steam Generator Water S R H 1, 2 Level--Low-Low
~15. Steam /Feedwater Flow Mismatch and S- R H 1, 2 Low Steam Generator Water Level
- 16. Undervoltage Reactor Coolant N.A. R H 1 Pumps g 17. Underfrequency - Reactor Coolant N.A. R M 1
- Pumps w
1 18. Turbine Trip A. Low Fluid Oil Pressure N.A. N.A. S/U(1) 1 B. Turbine Stop Valve Closure N.A. N.A. S/U(1) 1
- 19. Safety Injection Input t'om ESF
. N.A. N.A. M(4) 1, 2
~20. Reactor Trip Breaker N.A. N.A. -M(5) and S/U(1) 1, 2, and
- l 21. Automatic Trip Logic N.A. N.A. M(5) 1, 2, and
- l
- 22. Reactor Trip System Interlocks A. ' Intermediate Range N.A R S/U (8) 2, and
- Neutron Flux, P-6 B. -Power Range Neutron N.A. R S/U (8) 1-Flux, P-7 C. Power Range Neutron .N.A. R S/U (8) 1 Flux, P-8 D. Power Range Neutron N.A. R S/U (8) 1, 2 Flux,'P-10 E. Turbine Impulse Chamber -N.A. R S/U (8) 1. j Pressure, P-13 F. Power. Range Neutron Flux, P-9 N.A. R S/U (8) 1 G. . Reactor Trip, P-4 N.A. _
.R __
S/U (8) 1, 2, and
- INSTRUMENTATION TABLE 4.3-1 (Continued)
NOTATION . With -the reactor trip system breakers closed and the control rod' drive system capable of rod withdrawal. L(1) _- If not performed in previous 7 days. (2) -
~ Heatbalanceon,1y,above15%ofkkTEDTHERMALPOWER. Adjust channel .
t if absolute difference greater than 2 percent. - (3) '- Compare incore to excore axial flux difference above 15% of RATED THERMAL' POWER. Recalibrat.e if the absolute difference greater than , or equal to 3 percent. , (4) - Manual ESF. functional input check every 18 months. (5) - Each train or logic channel shall be tested at least e ery 62' ay.: .
-on a STAGGERED TEST BASIS.
(6) -
-Neutron detectors may be excluded from CHANNEL CALIBRATION.
(7) - Below P-6 (Block of Source Range Reactor Trip) setpoint. ., (8) - Logic only, each startup or when requirsd sith the reactor tri? system breakers closed and.the. control rod drive system capable of rod withdrawal if not performed ~in previous 92 days. I T-a k e e SEQUOYAH - UNII.1 3/4 3-13'
. . . . . . . . . . . - = * * ** = = + + - - - - - * " - -
TABLE 3.3-3 (Continued) l v, ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION E , Es MINIMUM g TOTAL N0. CHANNELS CHANNELS APPLICABLE I FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION I 95 c. Containment Ventilation Ej Isolation
- 1) Manual 2 1 2 1, 2, 3, 4 19
- 2) Autonatic Isclation 1 2 1, 2, 3, 4 15 l Lo3i c _
__ _ 2_ l 3) Containment Gas 2 1 1 1,2,3,4 19 Monitor Radioactivity-High
- 4) Containmen. Purge 2 1 1 1, 2, 3, 4 19
)
Air Exhaust Monitor Radioactivity-High u, 5) Containment Particu- 2 1 1 1,2,3,4 19 3; late Activity High s>
- 2. 4. STEAM LINE ISOLATION 03
- a. Manual 1/ steam line 1/ steam line 1/ operating 1,2,3 25 steam line -
- b. Automatic 2 1 2 1, 2,' 3 23 Actuation Logic
- c. Containment Pressure-- 4 2 3 1,2,3 18 High-High
- d. Steam Flow in Two 1, 2, 3 Steam Lines--High Four Loops 2/ steam line 1/ steam line 1/ steam line 16*
Operating any 2 steam lines . 1 Three Loops . 2/ operating l ### /any 1/ operating 17 l Operating steam line operating steam line I steam line ' l l
- m. - .. . _ . . _
' TABLE'3.3-3 (Continued)'
vs. . . . jj'
-ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION ,
S'
- r,
- MINIMUM
- TOTAL NO. CHANNELS- CilANNELS APPLICABLE- ,'
Si ~ FUNCTIONAL UNIT OF CilANNELS' TO TRIP OPERABLE MODES . ACTION A _. 6. AUXILIARY FEE 0 WATER
- a. Manual. Initiation 2 1 2 ' 1, 2, 3 24
- b. Automatic Actuation 2- 1 -2 1,2,3 23 Logic .
I
- c. Main Stm. Geri. Water
, level-Low-Low
- i. Start Motor u, Driven. Pumps. 3/stm.-gen. 2/stm. gen. 2/stm. gen. , 1, 2, 3 16.
2 any stm gen. us A, . ii. Start Turbine-c) Driven Pump 3/stm. gen. 2/stm.' gen.- 2/stm. gen' 1, 2, 3 16 any 2 stm. gen. .
- d. S . I .'
' Start Motor-Driven Pumps and Turbine
.. Driven Pump -See 1 above (all S.I. . initiating functions and requirements)
- e. ' Station Blackout Start Motor-Driven '
Rump associated .2/ shutdown -1/ shutdown -2/ shutdown with the shutdown . board board board 1, 2, 3 20 board and Turbine Driven Pump .
- f. Trip of. Main Feedwater mps Start Moto.' riven '
LPumps and Turbine Driven Pump <1/ pump 1/ pump 1/ pump; 1, 2 . -20" '
- g. Auxiliary Feedwater . . -
Suction Pressure-tow 3/ pump 2/ pump. 2/ pump 1,2,3 20^
INSTRUMENTATION I i a .
.. l TABLE 3.3-3 (Continued) i e .
( ACTION 21 - With the number of OPERABLE Channels one less than the. Total I Number of Channels, STARTUP and/or P0'#ER OPERATION may, proceed l provided the following conditions are satisfied:
- a. The inoperable channel is placed in the tripped condition within I hour.
- b. The Minimum Channels OPERABLE requirements is met; however,-
one additional channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.2.1.1. ACTION 22 - With less than the Minimum Number of Channels OPERABLE, declare the interlock inoperable and verify that all affected channels of the functions listed below are OPERABLE or apply the appropriate ACTION statement (s) for those functions. Functions to be evaluated are:
- a. Safety Injection Pressurizer Pressure
- b. Safety Injection High St im Line Flow Steam Line Isolation High Steam Line Flow Steam Dump
- c. Turbine Trip Steam Generator Level High-High Feedwater Isolation Steam Generator Level High-High ACTION 23 - With the number of OPERABLE channels one less than the Total Number. of Channels, be in at least HOT STANDBY within 6 hours and in at least HOT SHUTDOWN within the following 6 hours; however, one channel may be bypassed for up to 1 hour
. for surveillance testing per Specification 4.3.2.1.
ACTION 24 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within 6 hours and in at least HOT SHUTDOWN within the following 6 hours. I ACTION 25 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE , status within 48 hours or declare the associated valve inoperable and take the ACTION required by Specification 3.7.1.5. SEQUOYAH - UNIT 1 3/4 3-23 i . l [
TABLE 3.3-4 (Continued)
' ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS -
a 2, TRIP SETPOINT ALLOWABLE V LUES 7 FUNCTIONAL UNIT
- 2. CONTAINMENT SPRAY Manual Initiation Not Applicable Not Applicatile
- e. a.
Not Applicable j
- b. Automatic Actuation Logic Not Applicable l
- c. Containment Pressure--High-High 5 2.81 psig 5 2.97 psig
- 3. CONTAINMENT ISOLATION
~
- a. Phase "A" Isolation
- 1. Manual Not Applicable Not Applicable
- 2. From Safety Injection Not Applicable Not Applicable ,
[ m Automatic Actuation logic s
- b. Phase "B" Isolation .
- 1. Manual Not Applicable Not Applicable
- 2. Automatic Actuation Logic -Not Applicable Not Applicable
- 3. Containment Pressure--High-High 5 2.81 psig $ 2.97 psig
- c. Containment Ventilation Isolation
- 1. Manual Not Applicable .Not Applicable Not Applicable Not Applicable *
~2 .' Automatic.Tsolation Lonic_ __
TABLE 3.3-4 (Continued) N Ef1GIf1EERED SAFETY FEATURE ACTUATION SYSTEM Ifl5TRUllENTATION TRIP SETPOINTS
- E , -
5 . 7FUNCTIONALUNIT TRIP SETPOINT ALLOWABLE VALUES E 6. AUXILIARY FEEDWATER a Manual Not Applicable Not Applicable
- b. Automatic Actuation logic Not Applicable Not Applicable
- c. Main Steam Generator Water Level-low-low > 21% of narrow range > 20% of narrow rpnge Instrument span each Instrument span each
, steam generator steam generator
- d. S.I. See 1 above (all SI Setpoints)
(" e. Station Blackout 0 volts with a 5.0 second' 0 volts with a 5.0 1.0 second time. delay time delay Y . O f. Trip of Main Feedwater . N.A. N.A. > Pumps _ , _ _ , _,
. . . g. Auxiliary Feedwater Suction 1 2 psig (motor driven pump) 1 1 psig (motor driven pump)
Pressure-Low 1-6.5 psig (turbine driven pump) 1 5.5 (turbine driven pump)
- 7. LOSS OF POWER
- a. 6.9 kv Shutdown Board Undervoltage
- 1. Loss of Voltage 0 volts with a 0 volts with a 1.5 second time 1.5 1 0.5 second time delay delay
- 2. Load Shedding 0 volts with a 0 volts with a 5.0 second time delay 5.0 1 1.0 second time delay
- 8. ENGINEERED' SAFETY FEATURE ACTUATION SYSTEM INTERLOCKS
'a. Pressurizer Pressure Manual Block of Safety Injection P-Il 5 1970 psig i 1980 psig
TABLE 3.3-4 (Continued) ~ j@ ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 5
- E
, FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES
!! 8. . ENGINEERED SAFETY FEATURE ACTUATION-
-{ SYSTEM INTERLOCKS (Continued) avprevents Manual Block of Safety Injection P-12 5 540 F $ 512* F
- c. T,yg .
Manual Block of Safety Injection, Steam Line Isolation, Block Steam ' Oump Sb 540*F i538*F gg . -.
- d. Steam Generator Level i> Turbine Trip, Feedwater Isolation gg P-14 (See 5. above)
- 9. AUTOMATIC SWITCH 0VER TO CONTAINMENT SUMP *
- a. RWST Level - Low 130" from tank base 130" 1 4" from tank base COINCIDENT WITH Containment Sump Level - High 30"'above elev. 680' 30" i 2.5" above elev. 680' AND Safety Injection (See 1 above for all Safety Injection Setpoints/ Allowable Valves)
=
3 6
TABLE 3.3-5 (Continued) i ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS e
- 6. Steam Flow in Two Steam Lines-High Coincident with Steam Line Pressure-Low
- a. Safety Injection (ECCS) i 13.0(7)/23.0(1)
- b. Reactor Trip (from SI)
~
$ 3.0
- c. Feedwater Isolation < 8.0(2)
- d. Containment Isolation-Phase "A"(3) 18.0(8)/28.0( )
- e. Containment Ventilation Isolation Not Applicable
- f. Auxiliary Feedwater Pumps 1 60
- g. Essential Raw Cooling Water System 1 65.0(8)/75.0( ) -
s
- h. Steam Line Isolation < 8.0
- 1. Emergency Gas Treatment System 38.0(9)
- 7. Containment Pressure--High-High
- a. Containment Spray 1 58.00(9)
- b. Containment Isolation-Phase "B" 1 65(8)/75(9)
- c. Steam Line Isolation 1 7.0
- d. Containment Air Return Fan 2:540.0 and sE660
- 8. Steam Generator Water Level--High-High
- a. Turbine Trip-Reactor Trip i 2.5
- b. Feedwater Isolation i 11.0(2)
- 9. Main Steam Generator Water Level -
Low-Low
- a. Motor-driven Auxiliary 1 60.0 g.
Feedwater Pumps ( )
- b. Turbine-driven Auxiliary i 60.0 Feedwater Pumps (5) m SEQUOYAH - UNIT 1 3/4 3-31
'i 3
TABLE 4.3-2 (Continued) E$ i '- f3 ' ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SE SURVEILLANCE REQUIREMENTS 3E c: CHANNEL MODES IN WHICH 55 CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE CHECK CALIBRATION TEST REQUIRED (( FUNCTIONAL UNIT .
- 3. CONTAINMENT ISOLATION
- a. Phase "A" Isolation .
- 1) Manual N.A. N.A. M(1) 1, 2, 3, 4 i i 2) From Safety Injection N.A. N.A. M(2) 1, 2, 3, 4 Automatic Actuation Logic ,
- b. Phase "B" Isolation i
- 1) Manual N.A. N.A. M(1) 1, 2, 3, 4 ,
}{
i' 2) Automatic Actuation N.A. N.A. M(2) 1, 2, 3, 4 , i e IN Logic ,
- 3) Containment Pressure-- S R M. 1, 2, 3 i
High-H10h . 1- ! c. Containment Ventilation _ Isolation l l , 1) Manual N.A. N.A. M(1) 1 , 2 , 3 ,- 4 l l
- 2) Automatic Isolation Logic N.A. N.A. M(2) 1, 2, 3, 4 l .
- 3) Containment Gas Monitor 15 R H 1,2,3,4 f
i Radioactivity-High .. ,
TABLE 4.3-2 (Contir.ued)- Yn ~' ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION f!-< SURVEILLANCE REQUIREMENTS
~ 35 ,
CHANNEL MODES IN WHICil E CliANNEL FUNCTIONAL SURVEILLANCE CHANNEL'
$ FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED g,
R .M 1, 2, 3
- c. Main Steam Generator Water S Level-Low-Low See 1 above (all SI surveillance requirements)
~
- d. S.I.
- e. Station Blackout N.A. R N.A. 1, 2, '3 Trip of Main Feedwater N.A. N.A. R '1, 2 f.
- - ~
Pumps ---
- g. Auxiliary feedwater Surtion N.A. R M 1, 2, 3 .
~y Pressure-Low 0
- 7. LOSS OF POWER ,
- a. 6.9 kv Shutdown Board Undervoltage
- 1. Loss'of Voltage S R- M i , 2, 3, 4
- 2. Load Shedding S R N.A. !, 2,'3, 4
- 8. ENGINEERED SAFETY FEATURE-
- ACTUATION SYSTEM INTERLOCKS
- a. Pressurizer Pressure, N.A. R (4) N.A. ,1, 2, 3 P-11 .
- b. N.A. R (4) .N.A. 1,2,3
- T,yg, P-12
- c. Steam Generator N.A. R (4) N.A. 1, 2 +
Level, P-14 i ' .__ __-__. _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ . . . . _ _ _ _ _ _ .m ._m.__ _ - _ _ . . _ _ _ _ - . - ._ .m _ m -m . _ _ _ _ _ _ _ . _ _ _ _ . _ _ .-
m
=: ' %,
. { ;; .) y*
, , TABLE 4.3-2 (Continued) .
m E ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION h x SURVEILLANCE REQUIREMENIS , i e ' CHANNEL MODES FOR milch 5 CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE IS-FUNCTIONAL UNIT CllECK CALIBRATION TEST REQUIRED
. 9. AUTOMATIC SWITCH 0VER TO
' CONTAINMENT SUMP
- a. RSWT Level.- Low S. R H 1, 2, ' 3, 4 -
C0lNCIDENT WITH 4 Containment Sump Level - High S .R H 1,2,3,4 ' AND .
, Safety Injection (See 1 above for all Safety Injection Surveillance Requirements)
D Y Z m k i-i 4
M g TABLE 4.3-2 (Continued)
' . 'd. .
> TABLE NOTATION (1) Manual actuation switches shall be tested at least once per 18 months during shutdown. All other circuitry associated with manual safeguards
, actuation shall receive a CHANNEL FUNCTIONAL TEST at.least once per 31 days.
-(2) Each train or logic channel shall be tested at least every. 62 days on a STAGGERED TEST BASIS. ,
b
-.- ^(3) The CHANNEL FUNCTIONAL TEST shall include exercising the transmitter by 1 applying either a vacuum or pressure to the appropriate side of the transmitter.
(4) The total interlock function shall be demonstrated OPERABLE during CHANNEL
- CALIBRATION testing of each channel affected by interlock operation.
f
+
e
.i.
t. l l i l { SEQUOYAH - UNIT 1 3/4 3-36 4 j,. < l
lk, 'D ' f, ,e, f u; :- l' v, lABLE 3.3-6 m
- E' RADIAI10N MONITORING INSTRUMENTATION S
M MINIMllM CilANNELS APPLICA0ll' ALARM /IRIP MEASUREMENT INSTRUMENT OPERABLE MODES SEIPulNI- RANGE ACIl0N g q 1. AREA MONITOR $
~I 4 e a. _ Fuel Storage Pool Area 1
$ 15 mR/hr 10 --10 mR/hr 26
- b. Containment Area l 1, 2, 3 & 4 N/A 1- 10 R/hr*** 30
- 2. PROCESS MONITORS
~3
- a. Containment Purge Air 1 1, 2, 3, 4 & 6 $ 8.5 x 10 pCi/cc 10 - 10 cpm 28
- b. Containment
- i. Gaseous Activity
-3 pCi/cc 10 - 107 cpm 28 d)Ventildlion Isolation 1 ALL MODES '~< 8.5 x 10 b)RCS Leakaue Detection 1 1, 2, 3 & 4 N/A 10 - 10 cpm 27 g ii. Particulate Activity
-5 pCi/cc 10 - 107 cpm A a) Ventilation isolation 1 ALL MODES ~< l.5 x 10 28 y b)RCS Leakage Detection 1 1, 2, 3 & 4 N/A 10 - 10 cpm 27
^ o c. Control Room Isolation 1 ALL MODES 1 400 cpm ** 10 - 10 cpm 29
- d. Noble Gas Effluent Monitors
- With fuel in the storagg pool or building
^^ Equivalent .to 1.0 x 10 pCi/cc. *** Heasurement range by extrapolation.
INSTRUMENTATION. TABLE 3.3-6 (Continued) + ACTION STATEMENTS ACTION 26 - With the numoer of OPERABLE channels less than' required by the Minimum Channels (PERABLE requirement, perform area surveys of
'the monitored-are with portable monitoring instrumentation at.
least once per 24 hours. ACTION 27 - With tne numoer of OPERABLE channels less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.4.6.1.
' ACTION 28 - With the number of'0PERABLE channels less than required by the-Minimum Channels OPERABLE requirement, comply with the ACTION requirements if Specification 3.9.9.
ACTION 29 - With the number of OPERABLE channels less than required by the + Minimum Channels OPERABLE requirement, within 1 hour initiate-and maintain operation of the control room emergency ventilation system in the recirculation made of operation. 1' ACTION 30 - With the numoer of OPERABLE Channels less than the Minimum Channels OPERABLE requirement, restore the inoperable Channel (s) to OPERABLE Status within 7 days, or be in at least HOT STANDBY within the next 6 hours, and in at least HOT SHUTDOWN within the following 6 hours and in COLD SHUTOOWN within the subsequent 24 hours. I I f P l t-i o i' ' SEQUOYAH - UNIT 1 3/4 3-41 . t f , , --- - e
~.-- -. .
f Sh " EE TABLE 4.3-3 9L II RADIAfl0N MONITORING INSTRUMENTATION SURVEftlANCE REQUIREMENTS CllANNEL MODES FOR WillCH c: CilANNEL CilANNEL FUNCTIONAL SURVEILLANCE IS
' 35 '
-' CllECK CAllBRATION TEST REQUIRED INSTRUMENT w
- 1. AREA MONITORS
^
-a. Fuel Storage Pool Area S R M
- b. Containment Area S R H 1, 2, 3 & 4
- 2. PROCESS MONITORS
- d. Containment Purge Air Exhaust 5 R H 1, 2, 3, 4 & 6
- b. Containment .
- i. Gaseous Activity 5 R H ALL MODES.
a) Ventilation Isolation H 1, 2, 3, & 4 ! b)RCS Leakage Detection S R If ii. Particulate Activity
" H ALL MODES ,
a) Ventilation Isolation S R H 1, 2, 3 & 4 l
-b)RCS Leakage Detection S R .
{[
** M ALL MODES l c. Control Room Isolation S R
- d. Noble Gas Effluent Monitors
' With fuel.in the storage pool or buildinsl.
INSTRUMENTATION MOVABLE INCORE DETECTORS LIMITING CONDITION FOR OPERATION . .__ _ .. 3.3.3.2 The movable incore detection system shall be OPERABLE with:
- a. At least 75% of the detector thimbles,
- b. A minimum of 2 detector thimbles per core quadrant, and
- c. Sufficient movable detectors, drive, and readout equipment to map these thimbles.
APPLICABILITY: When the movable incare detection system is used for:
- a. Recalibration of the excore neutron flux detection system,
- b. Monitoring the QUADRANT POWER TILT RATIO, or N
- c. Measurement of F g, F q(Z) and Fxy*
ACTION: With the movable incore detection system inoperable, do not use the system for the above applicable monitoring or calibration functions. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 1' 4.3.3.2 The movable incore detection , system shall be demonstrated OPERABLE by normalizing each detector output when required for: -
- a. Recalibration of the excore neutron flux detection system, or ,,
- b. Monitoring the QUADRANT POWER TILT RATIO, or
- c. Measurement of F H, Fq(Z) and Fxy.
b SEQUOYAH - UNIT 1 3/4 3-43
TABLE 3.3-7 SEISMICMONITORINGINSIRUMENTATION .
~
-MINIMUM
. MEASUREMENT INSTRUMENTS
- INSTRUMENTS AND SENSOR LOCATIONS RANGE OPERABLE
- 1. Triaxial Time-History Accelerographs
- a. 0-XT-52-75A, Containment Elev. 734 0-1.0g i b.-0-XT-52-758, Annulus Elev 680
.0-1.0g 1*
- c. 0-XR-52-77, Diesel Building Elev. 722 0-1.0g 1
- 2. Triaxial Peak Accelerographs
- a. 0-XR-52-82, Auxiliary Building Elev.
689 0-5.0g 1
- b. 0-XR-52-83, Auxiliary B'uilding Elev.
736 0-5.0g 1
- c. 0-XR-52-84, Control Building Elev.
732 0-5.0g 1-
~ 3. Biaxial Seismic Switches 4
, .a. 0-XS-52-79, Annulus Elev. 680 0.025-0.25g 1*
- b. 0-XS-52-80, Annulus Elev. 680 0.025-0.25g 1*
l ,
- c. 0-XS-52-81, Annulus Elev. 680 0.025-0.25g 1*
f ., 4. Triaxial Response-Spectrum Recorders t.- l a. 0-XR-52-86, Annulus Elev. 680 2-25.4 Hz, 0.003-90g 1* l b.' 0-XR-52-87, Reactor Containment 2-25.4 Hz, 0.003-90g i Bldg. Elev. 734
- c. 0-XR-52-88, Aux. CR Elev. 734 2-25.4 Hz, 0.003-90g 1
- d. 0-XR-52-89, DB Bldg. 2A Elev. 713 2-25.4 Hz, 0.003-90g 1
{ "With reactor control room indication SEQUOYAH - UNIT 1 3/4 3-45 d 4 L
l TABLE 4.3-4
^
~
SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.s . . . . . CHANNEL CHANNEL CHANNEL FUNCTIONAL INSTRUMENTS AND SENSOR LOCATIONS CHECK- CALIBRATION TEST
- 1. Triaxial Time-History Accelerographs
- a. 0-XT-52-75A, Containment Elev. 734 M* R SA
- b. 0-XT-52-758, Annulus Elev. 680** M* -R SA
- c. 0-XR-52-77, Diesel Building Eley. 722 M* R SA-
- 2. Triaxial Peak Accelerographs
- a. 0-XR-52-82, Auxiliary Building Elev. 689 NA R NA
- b. 0-XR-52-83, Auxiliary Building Elev. 736 NA R NA
- c. 0-XR-52-84, Control Building Elev. 732 NA R NA i
- 3. Biaxial jSeismicSwitches
-a. 0-XS-52-79, Annulus Elev. 680** M R SA
- b. 0-XS-52-80, Annulus Elev. 680** M R SA -
- c. 0-XS-52-B1,' Annulus Elev. 680** M R SA
- 4. Triaxial Response-Spectrum Recorders
- a. 0-XR-52-86**, Annulus Elev. 680 M R NA
- b. 0-XR-52-87, Reactor Containment NA R NA Bldg. Elev. 734
- c. 0-XR-52--08, Aux. CR Elev. 734 NA R NA
, d. 0-XR-52-69, OB Bldg. 2A Elev. 713 NA R NA "Except seismic trigger
**With reactor control-room indications. '
e SEQUOYAH UNIT 1 3/4 3-46
INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION I LIMITING CONDITION FOR OPERATION 3.3.3.7 The accident monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE. APPLICABILITY: MODES 1, 2 and 3.
~
ACTION:
- a. With the number of OPERABLE accident monitoring instrumentation channels less than the Required Number of Channels shown in Table 3.3-10, restore the inoperable channel (s) to OPERABLE status within 7 days, or be in at least HOT SHUTDOW4 within the next 12 hours.
- b. With the number of OPERABLE accident monitoring instrumentation channels less than the MINIMUM CHANNELS OPERABLE requirements of Table 3.3-10, restore the inoperable channel (s) to OPERABLE status within 48 hours or be in at least HOT SHUTDOWN within the next 12 hours.
SURVEILLANCE REOUIREMENTS 4.3.3.7 Each accident monitoring instrumentation channel shall be demon-strated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION
, operations at the frequencies shown in Table 4.3-7.
SEQUOYAH - UNIT 1 3/4 3-55 l l l I
n: a- . TAllLE 3.3-10 E ACCIDENI MONITORING INSTRUMENIA110N f- @ S.
.g. MINIMUM
,. CilANNELS REQUIRED NO.
c- OF CllANNELS OPLRA8l.E 5e INSTRUMENT 1 2 H 1. Reactor Coolant Tilot (Wide Range) 1 2
- 2. Reactor Coolant. TCold ( de Range) 1 2
- 3. Containment Pressure 1
2
.4. Refueling Water Storage Tank Level 1
2
- 5. Reactor Coolant Pressure ..
1 Pressurizer Level (Wide'Ranga) 2 6. R 2/ steam line - 1/ steam line
- 7. Steam Line Pressure 1/ steam generator 1/ steam generator
- 8. ' Steam Generator Level . - (Wide Range) o 1/ steam generator Steam Generator Level - (Narrow Range) 'l/ steam generator 9.
1/ pump 1/ pump
- 10. Auxiliary feedwater Flow Rate' ,
0 1
- 11. Reactor Coolant System Subcooling Margin Monitor 2/ valve 1/ valve
- 12. Pressurizer PORV Position Indicator
- 2/ valve 1/ valve
- 13. Preshurizer PORV Block Valve Position Indicator **
2/ valve 1/ valve
- 14. Safety Valve Position Indicator 1
Containment Water Level (Wide. Range)' 2 15. 4/ core quadrant 2/ core quadrant
- 16. In Core Thermocouples
*Not applicable if the associated block valve is in the closed position.
**Not applicable if the block valve is verif ied .in.the closed. position with power to the valve operator removed.
M a *
. E! . TABLE 4.3-7 ,
I ACCIDENT MONITORING lNSTRUMENTA110N SURVEllLANCE REQUIREMENTS-
^
SE CilANNEL CilANNEL
- ) CAllBRAIl0N g INSTRUMENT CllECK
- 1. Reactor Coolant Tgot (Wide Range) M R
- 2. Reactor Coolant TCold (Wide Range) M R
- 3. Containment Pressure M R
- 4. Refueling Water Storage Tank Level M R
- 5. Reactor Coolant Pressure M R ca 6. Pressurizer Level M R-3
; 7. Steam Line Pressure M R u, -
j -4 8. Steam Generator Level - (Wide) M R
- 9. Steam Generator Level - (Narrow) M R I
- 10. Auxiliary Feedwater Flowrate .M R
, 11. Reactor Coolant System Subcooling M -R Margin Monitor
- 12. Pressurizer PORV Position Indicator .M R
! 13. Pressurizer PORV Block Valve M R
-l P,osition Indicator t 14. Safety Valve Position Indicator M R l '
- 15. Containment Water Level (Wide Range) M .R I
; 16. In Core Thermocouples M R
} TABLE 3.3-11 (Continued) m .
~. FIRE DETECTION INSTRUMENTS i .
I h' - Fire Minimum Instruments Operable j Zone Instrument Location Ionization Photoelectric Thermal Infrared-161 SG Blwdn. Rm. El. 734 4 162 EGTS Rm.'El. 734 3 163 EGTS Rm. El. 734 3 164 EGTS Fitr. A El. 734 1 L 165 EGIS Fitr. A El. 734 1 4 166 EGTS Fltr. B El. 734' 1 167 EGTS Fitr. B El. 734 1 j- 172' Mech. Eqpt. Rm. El. 734 1 173 Mech. Eqpt. Rc. El. 734 1 176 480-V Shtdn. Bd. Rm. lAl El. 734 2 1 188 480-V Shtdn. Bd. Rm. 2Al El. 734 2 177 480-V Shtdn; Bd. Rm. lAl El. 734 2
; 189 480-V!Shtdn. Bd. Rm. 2A1 El. 734 2 L 178 480-V Shtdn. Bd. Rm. lA2 El.-734' 2
-190 480-V Shtdn. Bd. Rm. 2A2-E1. 734 3
!, 179 480-V Shtdn. Bd. Rm. lA2 El. 734- 2
- I 191~: 480-V Shtdn. Bd. Rm. 2A2 EL. 734' 3 180 480-V Shtdn. Bd. Rm. 181 El. 734 2 5 192 480-V.Shtdn. Bd. Rm. 2Bl El. 734 2 ,,
181 480-V Shtdn. Bd. Rm. 181 El. 734 - 2. ! 193 480-V Shtdn. Bd. Rm. 2B1 El. 734 2 \ 1182 480-V Shtdn. Bd. Rm. 182-El. 734 3 194 480-V Shtdn. Bd. Rm. 282 El. 734 2 , 183 480-V Shtdn. Bd. Rm. 182 El. 734 3
~
195 480-V Shtdn. Bd. Rm. 282 El. 734 2 184 6.9-KV Shtdn. Bd. Rm. A El. 734 6 4-
-SEQUOYAH - UNIT 1- 3/4 3-60 -f.
- e. .'
. 4 F
~,4 - - , p ,, r ----.. -
r4 , ,-s,, ,e,. ,- m y - - , , , - - , aw,w> w- - - - , ,,e-,-4--, w
i.
~ . _.
{ TABLE 3.3-11 (Continued)
$ FIRE DETECTION INSTRUMENTS
. . . . . .. \ '. -
s Fire Minimum Instruments Operable Zone. Instrument Location Ionization Photoelectric Thermal -Infrared 69 Mech. Equip. Rm. El. 669 2 s 70 Aux. Bldg. A5-A11, Col. W-X, El. 669 5 - g 71 Aux. Bldg. AS-All, . Col. W-X, E1. 669 - 5 i- 72 - Aux. FW Pump Turbine 1 A-5, El. 669 1 73 Aux. FW Pump Turbine 1A-S, El. 669 1 76 S.I. & Charging Pump Rms. El. 669 5
- 77. S.I. Pump Rm. lA, El. 669 1 78 S.I. Pump Rm. 1B,.E1. 669 1 79 Charging Pump Rm.-1C, El. 669. 1 80 Charging Pump Rm. 18, El. 669 1 81 Charging Pump Rm. lA, El. 669 1 j 88 Aux. Bldg. Corridor Al-A8, El. 669 8 g 89 Aux. Bldg. Corridor Al-A8, E1. 669 8 1
90 Aux. Bldg. Corridor A8-A15, El. 669 8 91 Aux. Bldg. Corridor A8-A15, El. 669 8 92 Aux. Bldg. Corridor Col. U-W, El. 669 4
, 93. Aux. Bldg. Corridor Col. U-W, El. 669 4 94 Ya?ve Galley, El. 669 2 95 valve Galley, E1. 669 2 ,
39 Cont. Spray Pump 1A-A, El. 653 2 40 Cont. Spray Pump 18-B, E1. 653 2 ,, 43 RHR Pump 1A-A, E1. 653 2 44 RHR Pump 18-B, El. 653 2 47 Aux. Bldg. Corridor, E1. 653 10 SEQUOYAH - UNIT 1 3/4 3-65
, ,; - t.-
/
p 1 1 1 TABLE 4.3-8~(Continued) TABLE N0fATION a During l'iquid additions to the tank. - (1) I'i'C'A4NEL n H UNCTIONAL TEST shall also demonstrate that automatic isolation 2f viis pathway and control room alarm annunciation occurs if any of the foitowing conditions exists:
- 1. Instrument indicates measured leve es above the alarm / trip setpoint.
- 2. Circuit failure.
- i. .
1 (2)- The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:
- 1. Instrument indicates measured levels above the alarm setpoint.
- 2. Circuit failure.
Y
,, .i (3)r The initial CHANNEL CALIBRATION shall be performed usirg one or more of the reference standards certified by the National Bureau of Standards or using standards that have been obtained .from suppliers that participate
. in measurement' assurance activities with NBS. These. standards shall
. permit! calibrating the system over its intended' range of energy and measurement range. For subsequent' CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be'used.
1 (4) CHANNEL CHECK shall consist of verifying indication' of ' flow during periods
.cf release. ' CHANNEL CHECK shall be made at least-once per 24 hours on b days on which. continuous, periodic, or. batch releases are made.
i 3
^
4 -~ i . r - SEQUOYAH - UNIT 1 3/4 3-73' h 4-, -
~
A ' m
l { TABLE 4.3-9 (Continued) TABLE NOTATION e At all times. During waste gas disposal system operation. During shield building exhaust system operation.
**** During waste gas releases.
(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exists:
- 1. Instrument indicates measured levels above the alarm / trip setpoint.
- 2. Circuit failure. ,
(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:
- l. Instrument indicates measured levels above the alarm setpoint.
- 2. Circuit failure. s 1
(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards or using standards that have been obtained from suppliers that participate
, in measurement assurance activities with NSS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.
(1) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal: .
- 1. One volume percent hydrogen, balance nitrogen, and
- 2. Four volume percent hydrogen, balance nitrogen.
(5) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal: . I,
- 1. One vo'ame percent oxygen, balance nitrogen, and
- 2. Four volume percent oxygen, balance nitrogen.
SEQUOYAH - UNIT 1 3/4 3-86 i (
REACTOR COOLANT SYSTEM HOT STANDBY LIMITING CONDITION FOR OPERATION
~
3.4.1.2 a. At least two of the reactor coolant loops listed below shall be OPERABLE.
- 1. Reactor Coolant Loop A and its associated steam generator and Reactor Coolant pump,
- 2. Reactor Coolant Loop B and its associated steam generator and Reactor Coolant pump,
- 3. Reactor Coolant Loop C and its associated steam generator and Reactor Coolant pump,
- 4. Reactor Coolant Loop D and its associated steam generator and Reactor Coolant pump.
- b. At least one of the above coolant loops shall be in operation.*
APPLICABILITY: MODE 3
~~
ACTIOi-
- a. With less than the above required Reactor Coolant loops OPERABLE, restore the required loops to OPERABLE status within 72 hours or be in HOT SHUTDOWN within the next 12 hours.
- b. With no Reactor Coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required Reactor Coolant loop to operation.
SURVEILLANCE REQUIREMENTS 4.4.1.2.1 At least the above required Reactor Coolant pumps, if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability. 4.4.1.2.2 The required steam generators shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to 21 percent at least once per 12 hours. 4.4.1.2.3 At least one Reactor Coolant loop shall be verified to be in operation and circulating reactor coolant at least once per 12 hours.
"All Reactor Coolant pumps may be de-energized for up to I hour prov'
/ (1) no operations are permitted that would cause dilution of the react coolant system boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.
, SEQUOYAH - UNIT 1 3/4 4- la -
/
l ,
/ ,' 4 g
m'
1 l REACTOR CCOLANT SYSTEM . '7 I HOT SHUT 00'aN
}
LIMITING CONDITION FOR OPERATION 3.4.1.3 a. At least two of the reactor coolant and/or Resid"al heat removal (RHR) loops listed below shall be OPERABLE:
- 1. Reactor Coolant Loop A and its associated steam generator and reactor coolant cump,
- 2. Reactor Coolant Loc? B and its associated steam generator and reactor coolant pump,
- 3. Reactor Coolant Loop C and its associated steam generator and reactor coolent cump,
- 4. Reactor Coolant Loop 3 and its associated staam generator and reactor ccolant _Tc,
- 5. Residual Heat Remora! 000 A,
- 6. Residual Heat Remova! cop 8.
. ('; b. At least one of tne abc;e reactcr coolant and/or RHR loops shall
'J be in operatien.*'
APPLICABILITY: MODE 4. ACTION:
- a. With less than the above required locps OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status.as soon as possible; be in COLD SHUTDOWN within 20 hours.
- b. With no reactor coolant or RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to cperation.
**Asi reactor coolant pumps and residual heat rcmoval pumps may be de energized for up to I hour provided 1) no operations are permitted that ould cause dilution of the Reactor Coolant System 'uoron concentration, ano 2) core outlet temperature is maintained at least 10 F below saturation temperature.
fi.h
'A!
SEQUOYAH - UNIT 1 3/4 4-2 .
\,
l
REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS 4.4.1.3.1 The recuired reactor coolant pump (s), if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker , alignments and indicated power availability. 4.4.1.3.2 The required steam generator (s) shall be determined OPERABLE by serifjing secondary side water level to be greater than or equal to 10 percent (wide-range indication) at least once per 12 hours. 4.4.1.3.3 At least one reactor coolant or RHR loop shall be verified to be in operation and circulating reactor coolant at least once per 12 hours. O e SEQUOYAH - UNIT 1 3/4 4- 2a
- REACTOR COOLANT SYSTEM s COLD SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.~1.4..Two# residual heat removal (RHR) loops shall be OPERABLE
- and at least one RHR loop shall be in operation.**
- APPLICABILITY: MODE 5.
ACTION:
- a. With less than the above required RHR/ reactor coolant loops OPERABLE,
- immediately. initiate corrective action to return the required RHR/
- reactor coolant loops to OPERABLE status as soon as possible.
- b. With no RHR loop in cperation, suspend.all operations involving a
. reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the. required RHR loop to operation.
i SURVEILLANCE REQUIREMENTS 4.4.1.4 .The residual heat removal loop shall be determined to be~in operation and circulating reactor coolant at least once per 12 hours. J
#0ne RHR loop may be inoperable for up to 2 hours for surveillance testing provided the other RHR loop is OPERABLE and in operation. Four filled reactor coolant loops with at least 2 steam generators having levels greater than or
. equal to 10 percent (wide-range indication) may be substituted for one RHR loop.
*The normal or emergency power source may be. inoperable.
**The RHR pumps may be de-energized for up to 1 hour provided 1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and 2) core outlet temperature is maintained at'least 10 F below saturation temperature. -
3/4 4-2b j
.SEQUOYAH - UNIT 1
e REACTOR COOLANT SYSTEM RELIEF VALVES - OPERATING h (.. LIMITING CONDITION FOR OPERATION 3.'4.3.2 All~ power operated relief valves (PORVs) and their associated block valves shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. - ACTION:
- a. With one or more PORV(s) inoperable, .within 1 hour eithei restore the PORV(s) to OPERABLE status'or close the associated block valve (s) and remove power from the block valve (s); otherwise, be in at least HOT. STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With one or more block valve (s) inoperable, within 1 hour either restore the block valve (s) to OPERABLE status or close the block valve (s) and remove power from the block valve (s); otherwise, ce in at least HOT STANDBY within the. next 6 hours and in COLD SHUT 00WN within the following 30 hours.
- c. The provisions of Specification 3.0.4 are not applicable, f.
SURVEILLANCE REQUIREMENTS _ 4.4.3.2.1 In addition to the requirements of Specification .4.0.5, each PORV - shall be' demonstrated OPERABLE at least once' per 18 months by:
- a. Performance of a CHANNEL CALIBRATION, and
- b. Operating the valve through one complete ' cycle of full travel.
4.4.3.2.2 Each block valve shall be demonstrated OPERABLE at least once per
-92 days by operating the valve through one complete cycle of full travel.
-4.4.3.2.3 The emergency power supply-for the PORVs and block valves shall be .
demonstrated OPERABLE at least once per 18 months by:
- a. Transferring motive and control power from the normal to the-emergency power supply, and
-b. Operating the valves through-a complete cycle of full travel.
SEQUOYAH UNIT 1 3/4 4-4a . I i
REACTOR COOLANT SYSTEM
; ' c- 3/4.4.4 PRES 5URIZER LIMITING CCNDITION FOR OPERATION 3.4.4 The pressurizer shall be OPERABLE with a water volume of less than or equal to 1656 cubic feet (equivalent to an indicated level of less-than or equal to 92% on the narrow range instrumentation), and at least two groups of' presurizer heaters each having a capacity of at least 150 kw.
APPLICABILITY: . MODES 1, 2 and 3 ACTION:
- a. With one group of pressurizer heaters inoperable, restore at least-two groups to OPERABLE status within 72 hours or be in at least HOT STANDBY wi.h the next 6 hours and in HOT SHUTDOWN within the following 6 hou's.
- b. With the pressurizer otherwise inoperable, be in at least HOT STAN0BY' with the reactor trip breakers cpen within 6 hours and in HOT 5F2T00WN within the following 6 hours.
SURVEILLANCE REQUIREMENTS 4.4.4.1 The pressurizer water volume shall be determined to be within its limit at.least once per 12 hours. 4.4.4.2 The capacity of each of the above required groups of pressurizer heaters shall be verified by measuring circuit current at least once per 92 days. 4.4.4.3 The emergency power supply for the pressurizer heaters shall be demon-strated OPERABLE at least once per 18 months by transferring power from the normal to the emergency power supply and energizing the heaters. SEQUOYAH - UNIT 1 3/4 4-5 4
REACTOR CCOLANT SYSTEM ( LD 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.5.1 The following Reactor Coolant System ic kage detection systems snall be OPERABLE: -
- a. The lower containment atmosphere particulate radioactivity monitoring system,
- b. The containment pocket sump level monitoring system, and
- c. The lower containment atmosphere gaseous radioactivity monitoring system.
APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With only two of the above required leakage detection systems OPERABLE, opera-tion may continue for.up to 30 days provided grab samples of the containment
. atmosphere are obtained and analyzed at least once per 24 hours when the
? required gaseous or particulate radioactive monitoring system is inoperable; otherwise, be in at least HOT' STANDBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.4.6.1 The leakage detection systems shall be demonstrated OPERABLE by:
- a. The lower containment atmosphere gaseous and particulate monitoring system performance of CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the frequencies specified in Table 4.3-3, and
- b. Containment pocket sump level monitoring system-performance of CHANNEL -
CALIBRATION at least once per 18 months. O \1 SEQUOYAH - UNIT 1 3/4 4-13
REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE f G LIMITING CONDITION FOR OPERATION , 3.4.6.2 Reactor Coolant System leakage shall te limited to:
- a. No PRES $URE BOUNDARY LEAKAGE,
- b. 1 GPM UNIDENTIFIED LEMAGE,
- c. 1 GPM total primary +rsecondary leakage through all steam generators and 500 gallons per day through any one steam generator,
- d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and
- e. 40 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of T235 20 psig. .
- f. 1 GP'4 leakage at a Reactor Coolant System pressure of 2235 1 20 psig frt ,any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1.
APPLICABILITY: MODES 1, 2, 3 and 4 : ACTION: ., pas
- a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours and in COLD SHUTOOWN within the following 30 hours.
- b. With any Reactor Coc! ant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, and leakage frem Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours or be in' at least HOT STANDBY within the next 6 hours and in COLD SHUTCOWN within the following 30 hours.
- c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours by use of at least two closed manual or deactivated automatic valves, j
or be in at least HOT STANDBY within the next 6 hours and in COLD t SHUT 00WN within the fol!cwing 30 hours. SURVEILLANCE REOUIREMENTS 4.4.6.2.1 Reactor Ccolant System leakages shall be demonstrated to be within each of the above limits by: SEQUOYAH - UNIT 1 3/4 4-14 , i i
[ . . REACTOR COOLANT $^ille-G. SURVEILLANCE 3,EfJIREMENTS (Cortinus e dl , a. Monitoring the lower containment atmosphere particulate rafficactivity r monitor at least once per 12 hours.
- b. Monitoring the containment pocket sump inventory and discharge at le nt once per 12 hours.
- c. Meest"em? ' r,f the CONTROLLED LEAKAGE to the reactor coolant pump seals when the Reactor Coolant System pressure is 2235 2 20 psig at least once per 31 days with the modulating valve fully open. The provisions of Specification 4.0.4 are not applicable for entry into Mode 3 or 4.
- d. Performance of a Reactor Coolant System water inventory balance at least once per 72 hours.
- e. Monitoring the reactor head flange leakoff system at least orca per 24 hours.
4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specif%d in Table 3.4-1 shall be demonstrated OPERABLE pursuant-to Specification 4.0.5,
. except that in lieu of any leakage testing requirements required by Specifica-ci' tion 4.0.5, each valve shall be demonstrated OPERABLE by verifying leakage to be within its limit:
- a. At least once per 18 months.
- b. Prior to entering MODE 2 whenever the plant has been in COLD 5FUT00'aN for 72 hours or more-and if leakage testing has not been performed in the previous 9 months.
- c. Prior to returning the valve to service following maintenance, repair or replacement work on the valve.
- d. Within 24 hours following valve actuation due to automatic or manual action or flow tnrough the valve.
The provisions of Specification 4.0.4 are not applicable for entry into MODE 3
- or 4.
SEQUOYAH - UNIT 1 3/4 4-15
TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE NUMBER FUNCTION 63-560 Accumulator Discharge . 63-561 Accumulator Discnarge 63-562 Accumulator Discharge 63-563 Accumulator Discharge 63-622 Accumulator Discharge 63-623 , Accumulator Discharge 63-624 Accumulator Discharge 63-625 Accumulator Discharge 63-551 -Safety Injection (Cold Leg) 63-553 Safety Injection (Cold leg) 63-557 Safety Injection (Cold Leg) 63-555 Safety Injection (Cold Leg) 63-632 Residual Heat Removal (Cold Leg) 63-633 Residual Heat Removal (Cold Leg) 63-634 Residual Heat Removal (Cold Leg) 63-635 Residual Heat Removal (Cold Leg) 63-641 Residual Heat Removal / Safety Injection (Hot Leg) 63-644 Residual Heat Removal / Safety Injection (Hot Leg) 63-558 Safety Injection (Hot Leg) 63-559 Safety Injection (Hot leg) 63-543 Safety Injection (Hot Leg) 63-545 Safety Injection (Hot Leg) 63-547 Safety Injection (Hot Leg) 63-549 Safety Injection (Hot Leg) 63-640 Residual Heat Removal (Hot Leg) 63-643 Residual Heat Removal (Hot Leg) 87-558 Upper Head Injection 87-559 Upper Head Injection 87-560 Upper Head Injection 87-561 Upper Head Injection 87-562 Upper Head Injection 87-563 Upper Head Injection FCV-74-1* Residual Heat Aemoval FCV-74-2* Residual Heat Removal
\
\
*These valves do not have to be leak tested following manual or automatic actuatico or flow through the valve.
SEQUOYAH - UNIT 1 3/4 4- 15a i
REACTOR COOLANT SYSTEM i 3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and 3.4-3 curing heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:
- a. A maximum heatup of 100*F in any one hour period.
- b. A maximurn cooldown of 100*F in any one hour period.
- c. A maximum temperature change of less than or equal to 5 F in any one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.
APPLICABILITY: At all times. ACTION: With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operations or be in at least HOT STANDBY within the next 6 hours and reduce the RCS T,yg and pressure to less than 200 F and 500 psig, respectively, within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations. 4.l.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals required by 10 CFR 50, Appendix H in accordance with the schedule in Table 4.4-5. The results of these examinations shall be used to update Figures 3.4-2 and 3.4-3. SEQUOYAH - UNIT 1 3/4 4-23
g,a . eD . ,b'f:.' , V"
, C~ ' TABLE 4.4-5 m
- j. REACTOR VESSEL' MATERIAL SURVEILLANCE PROGRAM -'WITHDRAWL SCHEDULE 5
CAPSULE VESSEL LEAD-g NUMBER LOCATION FACTOR WITHDRAWAL TIME (EFPY) T. 4* 3.73 IN REFUELING U 140 3.73 3 X 220 3.73 5 Y 320 3.73 9 5 40 1.09 E0L
, V. 176* 1.09 STBY s
W 184* , 1.09 STBY i
=
U Z- 356* 1.09 STBY O .
=
/
~ _ - .
l l 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS COLD LEG INJECTION ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.5.1.1 Each cold leg injection accumulator shall be OPERABLE with:
- a. The isolation valve open,
- b. A ,ontained borated water volume of between 7857 and 8071 gallons of borated water,
- c. Between 1900 and 2100 ppm of boron, and
- d. A nitrogen' cover pressure of.between 385 and 447 psig.
APPLICABILITY: MODES 1, 2 and 3.* ACTION: ?
- a. With one cold leg injection accumulator inoperable, except as a i result of a closed isolation valve, restore the inoperable !
I accumulator to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours-and in HOT SHUTDOWN within the follnwing 6 hours. .
- b. With one cold leg injection accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in HOT STANDBY within one hour and be in HOT' SHUTDOWN l within the next 12 hours.
~
SURVEILLANCE REQUIREMENTS 1 4.5.1.1.1 Each cold leg injection accumulator shall be demonstrated OPERABLE:
- a. At least once per 12 hours by:
- 1. Verifying, by the absence of alarms or by measurement of levels and pressures, the contained borated water volume and nitrogen cover pressure in the tanks,'and .
- 2. Verifying that each cold leg injection accumulator isolation valve is open.
" Pressurizer pressure above 1000 psig.
- 'SEQUOYAH - UNIT 1- 3/4 5-1 ll i m m__-._ _ _ __
EHERGENCY CORE COOLING SYSTEMS 3/4.5.5 REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION-3.5.5 The refueling water storage tank (RWST) shall be OPERABLE with:
- a. A contained borated water volume of between 370,000 and 375,000 gallons,
- b. A baron concentration of between 2000 and 2100 ppm of- boron,
- c. A miaimum solution temperature of 60 F, and
- d. A maximum solution temperature of 105 F.
APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With the RWST inoperable, restore the tank to OPERABLE status within 1 hour or be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.5.5 The RWST shall be demonstrated OPERABLE: , a. At least once per 7 days by: i
- 1. Verifying the contained borated water volume in the' tank, and
, 2. Verifying the boron concentration of the water. l b. At least once per~ 24 hours by verifying the 'RWST temperature. t SEQUOYAH - UNIT 1 3/4 5-13' 4
3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained. APPLICABILITY: MODES 1, 2, 3 cnd 4. ACTION: Without ttimary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within one hour orwithin SHUTDOWN be intheatfollowing least HOT STANDBY 30 hours. within the next 6 hours and in COLD SURVEILLANCE REQUIREMENTS
,4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:
a. At least once per 31 days by verifying that all penetrations
- not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their Specification positions, except as provided in Table 3.6-2 of 3.6.3. -
b. By verifying that Specification each containment air lock is OPERABLE per 3.6.1.3. c. After each closing of each penetration subject to Type B testing, except the containment air locks, if opened following a Type A or B test, by leak rate testing the seal with gas at P , 12 psig, and verifying that when the measured leakage rate.for^these seals is added to the leakage rates _ determined pursuant to Specification 4.6.1.2.d for all other Type B and C penetrations, the combined leakage rate is less than or equal to 0.60 L ' a
*Except valves, blind flanges, and deactivated automatic valves which are located inside secured in the the annulus closed or containment and are locked, sealed or otherwise position. These penetrations shall be verified closed during each COLD SHUTDOWN more often than once per 92 days.
except that such verification need not be performed i', SEQUOYAH - UNIT 1 3/4 6-1 e i
l ,
, CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS 4.6.1.3 Each containment air lock shall-be demonstrated OPERABLE:
- a. After each opening, except when the air lock is being used for multiple entries, then at least once per 72 hours, by verifying seal leakage less than or equal to 0.01 L when a the volume between the door seals is pressurized to greater than or equal to 6 psig for at least 15 minutes,
- b. By conducting an overall air lock leakage test at not less.than P a (12 psig) and by within its limit:gerifying the overall air lock leakage rate is
- 1. At least once per six months, and
- 2. Prior to establishing CONTAINMENT INTEGRITY if opened when CONTAINMENT INTEGRITY was not required when maintenance has been performed on the air lock that could affect the air lock sealing capability.*
- c. At least once per 6 months by verifying that only one door in each air lock can be opened at a time.
l i I l The provisions of Specification 4.0.2 are not applicable.
- Exemption to Appendix "J" of 10 CFR 50.
i l I S:
~
i SEQUOYAH - UNIT 1 3/4.6-8 w m,w y - - - + - --m, - p
o ,
' JNTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT 59 RAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.1 Two independent containment spray systems shall be OPERABLE with each spray system capable of taking suction from the RWST and transferring suction to the containment sump.
APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With one containment spray system inoperable, restore the inoperable spray system to OPERABLE status within 72 hours or be in HOT STANDBY within the next 6 hours restore the inoperable spray system to OPERABLE status within the next 48 hours or be in COLD SHUTDOWN within the next 30 hours. SURVEILLANCE REOUIREMENTS 4.6.2.1 Each containment spray system shall be demonstrated OPERABLE:
- a. At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path that is nut locked, sealed, or otherwise secured in position, is in its correct position.
- b. By verifying, that on recirculation flow, each pump develops a discharge pressure of greater than or equal to 140 psig when tested pursuant to Specification 4.0.5.
- c. At least once per 18 months du' ring shutdown, by:
- 1. Verifying that each automatic valve in the flow path actuates to its correct position on a Containment, Pressure--High-High test signal. ,,
- 2. Verifying that each spray pump starts automatically on a Containment Pressure--High-High test signal.
- d. At least once per 5 years by performing an air or smoke flow test through each spray header and verifying each spray nozzle is unobstructed. .
i 1
; SEQUOYAH - UNIT 1 3/4 6-16
CONTAINMENT SYSTEMS
} 3/4.6.3 CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.3 The containment isolation valves soecified in Table 3.6-2 shall be OPERABLE with isolation times as.shown in Table 3.6-2.
APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With one or more of the isolation valve (s) specified in Table 3.6-2 inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and either:
- a. Restore the inoperable valve (s) to OPERABLE status within 4 ho'..s, or
- b. Isolate each affected penetration within 4 hours by use of at least one deactivated automatic valve secured in the isolation position, or
- c. Isolate each affected penetration within 4 hours by use of at least one closed manual valve or blind flange; or
- d. Be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.6.3.1 The isolation valves specified in Table 3.6-2 shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by performance of a cycling test'and verification of isolation time. 4.6.3.2 Each isolation valve specified in Table 3.6-2 shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE at least once per 18 months by:
- a. Verifying that on a enase A containment isolation test signal, each Phase A isolation valve acti l'.es to its isolation a sition.
- b. Verifying that on a Phase B containment isolation test signal, each Phase B isolation valve actuates to its isolation position.
SEQUOYAh - UNIT 1 3/4 6-17
CONTAINMENT SYSTEMS 1 SURVEILLANCE PEOUIREMENTS (Continued)
\ -
- c. Verifying that on a Containment Ventilation isolation test signal, each Containment Ventilation Isolation valve actuates to its isolation position.
4.6.3 3 The isolation time of each power operated or automatic valve of Table 3.6-2 shall be determined to be within its limit when tested pursuant to Specification 4.0.5. 4.6.3.4 Each Containment Purga isolation valve shall be demonstrated OPERABLE within 24 hours after each closing of the valve, except when the valve is being used for multiple cyclings, then at least once per 72 hours, by verifying that when the measure leakage rate of these valves is added to the leakage rates determined pursuant to Specification 4.6.1.2.d for all other Type B and C penetrations, the combined leakage rate is < 0.60 L,. 6 O SEQUOYAH - UNIT 1 3/4 6-18
4 CONIAINMLN1 SYS1' EMS
- 3/4.6.4 COMBUSTIBLE gas CONTROL HYDROGEN MONITORS I.
LIMITING CONDITION FOR OPERATION 1 3.6.4.1 .Two independent containment hydrogen monitors shall be OPERABLE.
- APPLICABILITY: MODES 1 and 2.
ACTION: With.one hydrogen monitor inoperable, restore the inoperable monitor to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours. 4 i 4 l SURVEILLANCE RiQUIREMENTS. 1 4.6.4.1 Each hydrogen monitor;shall be demonstrated OPERABL'E by the performance of a CHANNEL CHECK at least once per 12- hours, a CHANNEL FUNCTIONAL TEST at least once per 31 days, and at least once per 92 days on a STAGGERED TEST BASIS by performing a CHANNEL CALIBRATION using sample gas containing: l .
- a. One volume percent hydrogen, balance nitrogen.
- _ b .- Four volume percent hydrogen, balance nitrogen, i
i i L 1 SEQUOYAH - UNIT 1 -3/ 4 6-24 4
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t Tills PAGE DELETED
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il l I 1 SEQUCYAH - UNIT 1 3/4 525b
- ~_ _ _ "ee 7
CONTAINMENT SYSTEMS CONTAINMENT AIR RETURN FANS .,,. LIMITING CONDITION FOR OPERATION 3.6.5.6 Two independent containment air return fans shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. . ACTION: With one containment air return fan inoperable, restore the inoperable fan to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUT 00WN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.6.5.6 Each containment air return fan shall be demonstrated OPERABLE:
- a. At least once per 92 days on a STAGGERED TEST BASIS by: ,
- 1. Verifying that the fan motor current is 28 i 7.5 amps with the backdraft dampers closed, and
- 2. Verifying that with the fan off, the air return fan damper opens when a torque of less than or equal to 68.1 inch-pounds is appied to the counterweight.
- b. At least once per 18 months by verifying that the air return fan starts on an auto-start signal after a 10 1 1 minute delay and operates for at least 15 minutes.
G SEQUOYAH - UNIT 1 3/4 6-33 t
, PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION
- 3. 7.1. 2 At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:
- a. Two motor-driven auxiliary feedwater pumps, each capable of being powered from separate shutdown boards, and
- b. One turbine-driven auxiliary feedwater pump capable of being powered from an OPERABLE steam supply system.
APPLICABILITY: MODES 1, 2 and 3. ACTION:
- a. With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours,
- b. With two auxiliary feedwater pumps inoperable, be in at least HOT STANDBY within 6 hours and in HOT SHUTDOWN within the following 6 hours.
- c. With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible.
SURVEILLANCE REQUIREMENTS
- 4. 7.1. 2 In addition to the requirements of Specification 4.0.5 each auxiliary feedwater pump shall be demonstrated OPERABLE by:
- a. Verifying that:
- 1. each motor-driven pump develops a differential pressure of greater than or equal to 1397 psid on recirculation flow.
- 2. the steam-turbine driven puap develops a differential pressure of greater than or equal to 1183 psid on recirculation flow when the secondary steam supply pressure is greater than 842 psig.
The. provisions of Specification 4.0.4 are not applicable for entry into MODE 3. - SEQUOYAH - UNIT 1 3/4 7-5 \
. .-. . - - , _ _ . - _ . .. .. . - - , ..=
l j PLANT SYSTEMS .,,
\:Y2r SURVEILLANCE REQUIREMENTS (Co3tinued) .
- 3. cach automatic control valve in the flow path is OPERABLE.
, whenever the auxiliary feedwater system is placed in automatic control or when above 10% of RATED THERMAL POWER.- '
- b. At least once per 18 months during shutdown by:
- 1. Verifying that each automatic valve in the f. low path actuates to its correct position upon receipt of an auxiliary feedwater actuation test signal and a low auxiliary feedwater pump suction
, pressure test signal.
- 2. Verifying that each auxiliary feedwater pump starts as designed' automatically upon receipt of each auxiliary feedwater actuation test ~ signal,
- c. At least once per 7 days by verifying that each non-automatic valve i in the auxiliary feedwater system flowpath is in its correct position.
' IRE.
i i I i i t 9 s SEQUOYAH - UNIT 1 ~ 3/4 7-6 l l t
PLANT SYSTEMS (- 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION LIMITING CONDITION FOR OPERATION _ ,, 3.7.2 The temperatures of both the primary and secondary coolants in the steam generators shall ce greater than 70*F when the pressure of either i coolant in the steam generator is greater than 200 nsig. l APPLICABILITY: At all times. ACTION: 1 With the rcquirements of the above specification not satisfied: 2 a. Reduce the stGae generator pressure of the applicable side to less than or equal te 200 psic within 30 minutes, and .l b. Perform an engineering evaluation t0 determine the effect of the overpressurization on the structural integrity of the steam generator. Determine that the steam generator remains acceptable for continued operation prior to increasing its temperatures above 200*F. []- SURVEILLA!:CE RE0VIREMENTS f 4.7.2 The pressure in each side of-the steam generator shall be determined , to be less than 200 psig at least once per hour when the temperature of either the primary or secondary coolant is less than 70*F. J i 5
. I t
, SEQUOYAh - UNIT 1 3/4 7-11 1- I g g- y r' -v v -- + T . .+% , a-- -. 4 - g
.+ -
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- PLANT SYSTEMS
- 3/4.7.3 COMPONENT COOLING WATER SYSTEM ,
LIMITING CONDITION FOR OPERATION '/ ' 7a _ i j 7. 3.7.3 At least two independent component co6fing water loops shall be OPERABLE. e . x APPLICABILITY: MODES 1, 2, 3 and 4, ;
, . rr
, ACTION: f '- . y "(f With only one component cooling water icop'0PERABLE, re' store at^1 east two .
loops to OPERABLE status within 72 hours or be in at~least HOT STANDBY within the next 6 hours and in COLD SHUTDWN within the folicwing 30 hours.
.1 SURVEILLANCE REQUIREMENTS
. 4.7.3 At least two component cooling water loops shall be demonstrated OPERABLE:
- a. At least once per 31 days on a _ STAGGERED TEST BASIS by verifying that each valve (manual, power operated or autonatic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position, is in its correct position. y '
- }
- b. ' At least once per 18 months, durir.g shutdohn, by verifying that each component cooling system pump starts automatically on a Safety Injection tact signal. ,
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l PLANT SYSTEMS
'. 3/4.'7 4 ESSENTIAL RAW COOLING WATER SYSTEM LIMITIhG CLNDITION FOR OPERATION 3.7.4 'At least two independent essential raw cooling water (ERCW) loops ,
shall be OPERABLE. ; APPLICABILITY _: MODES 1, 2, 3 and 4. ACTION: With only one ERCW loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours or be in'at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REOUIREMENTS 4.7.4 'Ati least two ERCW loops shall be demonstrated OPERABLE:
- a. At least once per 31 days by verifying that each valve (manual, power operated or automatic) servicing safety related equipment that is not locked, sealed, or otherwise secured in position, is in its correct position.
s
- b. At least cace per 18 months, during shutdown, by:
' i. Verifying that each automatic valve servicing safei.y related equipment' actuates to its correct position on a Safety Injecticn
. .- c -
' .- . test signal.
- 2. -
V r:fying that each ERCW purn starts automatically on a Safety Injection test signal. c . . m
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t-SEQUOYAH - UNIT 1 3/4 7-14 - 'y?. "
6 PLANT SYSTEMS 3/4.7.6 FLOOD PROTECTION PLAN I t I LIMITING CONDITION FOR OPERATION 3.7.6 The flood protection plan shall be ready for implementation to maintain the plant in a safe condition. APPLICABILITY: When one or more of the.following conditions exist:
- a. heavy rainfali conditions in the east Tennessee watershed,
- b. an early warning or alert that a critical combination of flood and/or high headwater levels may or have developed,
- c. .an early warning or alert involving Fontana Dam, or
- d. recognizable seismic activity in the east Tennessee region. ,
ACTION:
- a. With a Stage I flood warning issued initiate and conplete within 10 hours the Stage I flood protection plan which shall include being in at least HOT STANDBY within 6 hours, with a SHUTDOWN MARGIN of at least 5% delta k/k and T/ avg .less than or equal to 350'F within the C.f.~
following 4 hours. If within 10 hours following the issuance of a Stage I flood warning communications between the TVA Division of Water Resources and the Sequoyah Nuclear. Plant cannot be verified, initiate and complete the Stage II flood protection procedure within the following 17 hours. With a Stage II flood warning issued initiate
- the Stage II flood protection plan in time to ensure completion.
before the predicted flooding of the site and no later than ~.7 hours prior to the predicted arrival time of the initial critical flood level (697 ft msl winter and 703 ft ms1 -summer).
- b. With a seismic event occurring after a critical combination of flood and/or headwater alerts are issued verify and maintain communications between TVA Power Control Center and the Sequoyah Nuclear Plant within 6 hours or initiate and complete the-Stage I flood protection plan within the following 10 hours. If communications have not been established upon completion of the Stage I flood protection plan -
initiate and complete the Stage II flood protection plan within the foilowing 17 hours. SEQUOYAH - UNIT 1 3/47-1 6 I'
. . .=
PLANT SYSTEMS a 1 ACTION: (Continued)
- c. With a Fontana Dam Alert issued verify and maintain communications between Fontana Dam and the Sequoyah Nuclear Plant with I hour or initiate and complete the Stage I flood protection plan within 10 hours. If communications have not been established upon completion of the Stage I flood protection plan initiate and complete the Stage II flood protection. plan within the following 17 hours. l
- d. With either the Norris, Cherokee, Douglas, Fort Loudon, Fontana, Hiwassee, Apalachia, Blue Ridge or Tellico dam failed seistically after a critical combination of flood and/or headwater alerts is l issued initiate and complete the Stage I. flood protection plan I within 10 hours. Upon completion nf the Stage I flood protection plan initiate and complete the Stage II flood protection plan within the following 17 hours. Both the Stage I and the Stage II flood i protection plans will be terminted.if it is determined that the l i
potential for flooding the site does not exist. l 1 SURVEILLANCE REQUIREMENTS 4.7.6.1 The water level in the forebay shall be determined at least once per ((1 8 hours when the water level is less than or equal to 697 feet Mean Sea Level L' USGS datum during October 1 through April 15, or 703 feet Mean Sea Level USGS datum during April 16 through September 30; and at least once per 2 hours when the water level is above these limits. 4.7.6.2 Communications between Sequoyah Nuclear Plant:
- a. and TVA Division of Water Resources shall be maintained every l 3 hours during heavy rainfall condition in the east Tennessee ;
watershed. !
- b. and TVA Power Control Center shall be maintained every 3 hours following a recognizable seismic event that has occurred when a critical combination of flood and/or headwater alert is issued.
Communications shall be maintained until it has been determined 1 ( that the potential for flooding the site does not exist, - l
- c. and Fontana Dam shall be maintained every hour when an alert involving-Fontana Dem has been issued by TVA Division of Water i Resources. l l
l l n l SEQUOYAH - UNIT 1 3/4 7-17 1 i 1
PLANT SYSTEMS 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM
- l.
LIMITING CONDITION FOR OPERATION -- s
-i 3.7.7 Two independent control room emergency ventilation systems shall be OPERABLE.
AP P LICABI LIT'.': ALL MODES ) I ACTION: MODES 1, 2, 3 and 4 With one control room emergency ventilation' system inoperable, restore the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. MODES 5 and 6
- a. With one control room. emergency ventilation system inoperable, restore the inoperable system to OPERABLE status within 7 days or initiate and maintain operation cf the control room emergency ventilation system in the recirculation mode.
4
- b. With both control room emergency air ventilation systems inoperable, suspend all operations involving CORE ALTERATIONS or positive
, . reactivity changes. i t I c. , The provisions of Specification 3.0.3 are not applicable in MODE 6. , SURVEILLANCE RE0UIREMENTS - I. 4.7.7 Each control room emergency ventilation system shall be demonstrated OPERABLE: l a. At least once per 12 hours by verifying that the control room i air temperature is less than or equal to 104 F. . l b. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and
; charcoal adsorbers and verifying that the system operates for at least 15 minutes. .
l .
- c. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone
- i. communicating with the system by:
SEQUOYAH - UNIT-1 3/4 7-18 4 L
PLANT SYSTEMS 1, 3/4.7.8 AUXILIARY BUILDING GAS TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION . . 3.7.8 Two independent auxiliary building gas treatment filter trains shall be OPERABLE. PPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With one auxiliary building gas treatment filter train inoperable, restore the inoperable train to OPERABLE status within 7 days or be'in at least HOT STANDBY within the next 6. hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REOUIREMENTS 4.7.8 Each auxiliary building gas treatment filter train shall be demonstrateo OPERABLE:
- a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through~the HEPA filter and charcoal adsorber train and verifying that the system operates for at least 10 hours with the heaters on.
. b. At least once per 18 months or (1) afte. any structural maintenance.
on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system by:
- 1. Verifying that the cleanup system satisfies the in place testing acceptance criteria and uses the test procedures of..
Regulatory Positions C.S.a, C.S.c and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978 (except for the provisions
- of ANSI N510 Sections 8 and 9), and the system flow rate is j 9000 cfm + 10%. .
i l 2. Verifying within 31 days after removal that a laboratory analysis-of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978.
- 3. . Verifying a system flow rate of 9000 cfm + 10% during system
'cperation wt.en tested in accordance with XNSI N510-1975.
SEQUOYAH - UNIT 1 3/4 7-20 l
PL}NT SYSTEMS 3/4.7.9 SNUBRERS LIMITING CONDITION FOR OPERATION 3.7.9 All safety-related snubbers saall be OPERABLE. The snubbers are shown in Tables 4.7.9.a and 4.7.9.h and are listed in Surveillance Instruction SNP SI-162., Any exemptions to the surveillance program are shown in Table 4.7.9.c and in SNP SI-162. APPLTCABILITY: Modes 1, 2, 3, and 4. (Modes 5 and 6 for snubbers located on systems or partial systems required OPERABLE in these Modes.) ACfl0N: With one or more snubbers inoperable, within 72 hours replace or restore the inoperable snubber (s) to OPERABLE status and perform an engineering evaluation on the attached component or declare the attached system inoperable and follow the appropriate ACTION statement for that system. SURVEILLANCE REQUIREMENTS 4.7.9 Each safety-related snubber shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5. These snubbers are shown in Tables 4.7.9.a and 4.7.9.b, and are listed in Surveillance Instruction SNP SI-162. Table 4.7.9.b is a detailed tabulation of the hydraulic snubbers which are also shown in Table 4.7.9.a. Any exemption to any portion of the surveillance program for any snubber is shown in Table 4.7.9.c.
- a. Inspection Groups The snubbers may be categorized into two major groups based on whether the snubbers are accessible or inaccessible during react or operation. These major groups may be further subdivided into subgroups based on design, environment, or other features which may be expected to alicct the OPERABILITY of the snubbers within the subgroup. Each subgroup or group may be inspected independently in accordance with 4.7.9.b through 4.7.9.h.
- b. Visual Inspection Schedule and Lot Size The first inservice visual inspection of snubbers shall be completed by October 31, 1981, and shall include all snubbers on safety-related systems. If less than two (2) snubbers are found SEQUOYAlf - UNIT 1 3/4 7-22
4 i.
/' ' pLhNT' SYSTEMS-I SURVEILLANCE REQUIREMENTS (Cen' II""'I) '
b: Visual Inspection He ltlule and Lot Size (Cont'd) . inoperable durinp* 8 3". I rst insgrvice visual inspection, the : { second inservice vi"",,I inspection shall be performed 12 months -1 i 1 25% . from the dat" "a t he t i rs t inspeg ts. on. Othawa.se, subsequent ' visual inspections hh,,11 be' performed in accordance with the following schedule: Subsequent. Visual Number of Inoperabl" Inspection Period'* # Snubbers per Inspe" Jon Period 0- 18 months i 25% 1 12 months i 25% 2 6 months i 25% 3, 4 124 days i 25% 62 days 1 25%
. 5, 6, 7 31 days i 25%
8 or mor,.
- c. Visual Inspection'P"Il'!#*#"'
Visual-inspectionn ng,,,t 1 verify (1) . that there are no ' visible indications of dami!P' # impaired OPERABILITY,--(2) bolts attaching the snubber to the I"'nula i n r supputing structure am secure, . 5 and (3) snubbera allis,in to sectkns of safetMelaW systems that have experict:r '"l unexpected potentially damaging transients since'the last innp",Iion Period shall be evaluated for the possibility o f . coin *"'I'"' #"*E" "" '""C ' "* " Y "" " ' applicable, to cuniil8'"P "^ II'Y' in pu e as,a usu R of visual inspections Snubbers which appe"It ! may be determined 881'lIUUlLE for the purpose of establishing the next visual inspet I l'"' . in av 1, pr vi ing that (1) tiie cause of the rejection in , )carly established and remedied for that - i particular snubber Hol f r ther snubbers that may be generically , susceptible; and (?; t he aUntM snutilin is functionally tested, I if applicable, in sh,, an-found condition and determined OPERABLE per Specification 4. / iye. lydraulic snubbers with inoperable single or common f lu g,l reservoirs which have uncovered fluid ports shall be derid8'"I jn perable. When hydraulic snubbers
' E 8 are tested,'..the tests shall be
~ .
which have uncove <"I I b.'th the piston at the as-found setting si I' _ performed by start inh and extending the pl"g.ni r d in the extenston mode direction.
- The inspection interval shall Hul be kngtkned mon dian one step at a time.
# The provisions of SpecificatI"H 4 0.2 are not . applicable.
SEQ 110YAll y tJNIT 1 1/4 7-23 s
- ~ _ .
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) s'
- c. Visual Inspection Performance and Evaluation (Cont'd)
' Also, snubbers which have been made inoperable as the, result of unexpected transients, isolated damage or other such random events, when the provisions of 4.7.9.g and 4.7.9.h have been met and.any other appropriate corrective action implemented, shall not be counted in determining the next visual inspection interval.
- d. Functional Test Schedule, Lot Size, and Composition During each refueling outage, a representative sample of 10% of the total of the safety related snubbers in use in the plant shall be functionally tested either in place or in a bench test.
The representative sample selected for functional testing shall include the various configurations, operating environments, and the range of size and cepacity of snJbbers within the groups or subgroups. The representative sample should be weighted to include more snubbers from severe service areas such as near heavy equipment. Unless a failure analysis as required by 4.7.9.f indicates otherwise, the sample shall be a composite based on the ratio of each group to the _s total number of snubbers installed in the plant. Snubbers placed in a the same location as snubbers which failed the previous functional (NJJ test shall be included in the next test lot if the failure analysis shows that failure was due to location. The security ti fcsteners for attachment of the snubbers to the component and to the snubber anchorage shall be verified on snubbers selected for functional tests.
- e. Functional Test Acceptance Criteria The snubber functional test shall verify that:
- 1. Activation (restraining action) is achieved within the specified range in both tension and compression, except that inertia dependent, acceleration limiting mechanical snubbers, may be tested to verify only that activation takes place in both directions of travel.
I SEQUOYAH - UNIT 1 3/4 7-23
PI: ANT SYSTEMS SURVEILI.ANCE REQ 111REMENTS (Continued)
- 2. Snubber bleed, or release where required, is present in both tension and compression, within the specified range.
- 3. The force required to initiate or maintain motion of the snubber is within the specified range in both directions of travcl. Also, the increase in the force required shall not exceed 50 percent of the amount required at the last surveillance test of that snubber, provided that the force required is at Icast 25 pounds.
- 4. For snubbers specifically required not to displace under continuous l'ad, the ability of the snubber to withstand load without displacement shall be w;rified.
- 5. Testing methods may be used to measure parameters indirectly or parameters other than those specified if those results can be correlated to the sp2cified parameters through established methods.
- f. Functional Test Failure Analysis and Additional Test Lots If any snubber selected for functional testing either fails to lock up or fails to move due to manufacture or design deficiency, all snubbers of the same design subject to the same defect sb-ll be functionally tested.
If more than two snubbers do not meet the functional test acceptance criteria, ar* additional lot equal to one-half the original lot size shall be functionally tested for each failed snubber in excess of the two allowed failures. An engineering evaluation shall be made of each failure to meet the functional test acceptance criteria to determine the cause of the failure. The result of this analysis shall be used, if applicable, in selecting snubbers to be tested in the subsequent lot in an effort to determine the operability of other snubbers which may be subject to the same failure mode. (Selection of snubbers for future testing may also be based on the failurt analysis.) Testing shall continue until not more than one additional inoperable snubber is found within a subsequent required lot SEQUOYAH -ITNIT 1 3/4 7-25
, PLANT SYSTEMS SURVEILLANCE REQllIREMENTS Gontinued) f.
Functional Test Failure Analysis and Additional Test Lots (Cont'il) or all snubbers of the origina: .nspection group have been tested, or all' suspect snubbers identified by the failure analysis'have been tested, as applicable. The discovery of loose or missing at tachment fasteners will be evaluated to determine whether the caase may be localized or generic. The result of the evaluation will be used to select other suspect snubbers for verifying the attachment fasteners, as applicable. Snubbers shall not he subjected to prior maintenance specifically for the purpose of meeting functional test requirements.
- g. Functional Test Failure - Attached Component Analysis For snubber (s) found inoperable, an engineering evaluation shall be performed on the components which are restrained by the snubber (s).
The purpose of this engineering evaluation shall be to determine if the components restrained by the snubber (s) were adversely affected by the inoperability of the snubbers (s), and in order te ensure s that the re' trained component remains capable of meeting the designed service.
- h. Functional Testing of Repaired and Spare Snubbers Snubbers which fail the visual inspection or the functional test acceptance criteria shall be repaired or replaced. Replacement snubbers and snubbers which have repairs which might affect the functional test results shall be tested to meet the functional test criteria before installation in the unit. These snubbers shall have met the acceptance criteria subsequent to their most recent service, and the functional test must have been performed within 12 months before being installed in the unit.
- i. Snubber Service Li fe prol; ram The seal service li fe of hydraulic snubbers shall be monitored to ensure that the seals do not-fail between surveillance inspections. The mayimum expected service life for the various seals, seal materials, and applications shall be estimated based on engineering information, and the seals shall be replaced so that the maximum expected service life does not expire during a period when the snubber is required to be operable. The seal replacements shall be documented and the documentation shall be retained in accordance with 6.10.2.n.
SEQUOYAll " NIT 1 3/4 7-26
/
i PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) i . S,nubber Service Life Program (Cent'd) Mechanical snubber drag force increases greater than 50 percent of previously measpred values shall be evalua*.ed as an indi-cation of imperuling failure of the tennbber. The:;c evaluations and any associated corrective action shall be documented and the documentation shall be retained in accordance with 6.10.2.n. j . Exemption From Visual Inspection or Functional Tests Permanent or other exemptions from the surveillance program for individual snubbers may be granted by the Commission if a justifiable basis for exemption is presented and if applicable snubber life destructive testing was performed to qualify snubber operability for the applicable design conditions at either the completion of their fabrication er at a subsequent date. Snubbers so exempted shall be listed in Table 4.7.9.c and shall continue to be listed in the plant instruction SNP SI-162 indicating the extent of the exemptions. O SEQUOYA!! - UNIT 1 3/4 7-27
Tabic 4.7.9a 3afety Related Snubbers
- ACCESSIBLE INACCESSTELE Small Medium & Large Small Medium & Large PSA llyd. Paui PSA PSA PSA iunrde}{yd!
Size 1/4 1/2 1 l3 10 35 l 100 1/4l1/2 1 3 10 l 35 100l l l MS 22; 9 3 9 9 7 12 1 20 l 16 i AMS 1 2 3 l l AFD 4 1 1 4 5 1 2 2 l 2l FD 2 6 1 1 2 1l _CC 8 10 8 10 21 5 4 2[ l l ' 4! ST 3 1 2l 37 12 2 9 ! 15 1! I-4l CS 3 4 2 3 1 l15 1l l l CVC 7 5 1 4 24 l 7 3 8l 1 l l l RC 15 16 29 40 19 8l l ! UI'I 1 4 7 20 24 l 5 l 1l SCB 1 1 1 7 8 l5 l i ! l l FPC 2 4 3 1l l l l g ERCW 2 5 4 22l19 23 15 l l l l RilR S 2 2 6 2 2 1l l l 3I TC 8 6 5 l ) ! l l WD 9 l l l l DW 1 1 1 l l l I SA 1 1 1 l l l k PW l 2 1 1l l l l l AC511 12 l ' 7 l l l l l s l l (
! l l
5 $d, 44 24 _19 31 30 11 7 154 86 87l106 77 15 l ro t a l 68 98 47 l 240 285 20 l31
*Saubb'rs may be added to safety related systems without prior License Amendment
'to Table 4.7.9a provided that a revision to Tabic 4.7.9a is included with the next Licence Amendment request. Any exemptionn to the provisions of the surveillance program for any snubber is indicated in Table 3.7.9.c.
SEQUOYAH UNIT 1 3/4 7-28
v1 Jc Tabic 4.7.9b ~ Safety Related Hydraulic Snubbers * *
- O ACCESSIBLE D INACCESSIBLr
[ . Sub l Sub' - c Size 1 113 2 215 3k 4 5 6 8 , Total Size 1 115 [2 - 215 3k 4 '5 6 8 a . Total - Q !s l 1 5 5 1 l 12 MS
- l l 8 I 8 5 -
~
A:ts 3 l l 3 AMS l l I i 0 AFD 2 3 l 5 AFD 2 l l 2 FD 3 2 1 l 6 FD 1 l 1 CC 1 2 3 3 1 l 10 cc 2 2 l l l 4 ST 1 1 2 ST ! 1 2 1 f l 4 CS 1 l 1l1 3 CS I I l 0 rt'r 4 l l 4 CVC ! l 0 u RE O RC ! l 0 tml 0 UHI 1 l l 1 + w h e SG3 0 SGB l ! l l 0 FPC 0 FPC l l l 0 E RHR 2 g 3 2 RHR 2 1 , l 3 TF 0 IC l l 0 _KD 0 wn l l O_
, DW 0 Du l l 0
$A . O SA l 0 PW ! O PW l 'O arsy 0 AC&H 0 ERCW ,
0 _ERCW 0 0 Tatal 47 Total 31
- Snubbers nay be added to safety related systems without prior License T.mendment to Table 4.7.9b provided that a revisfor. to Table 4.7.9b is included with the next License Amendment request. Any exemptions to the provisions of the surveillance program for any snubber is-indicated in Tabic 3.7.9::.
= -
Table 4.7.9c Safety Related Snubbers - Exemptions to the Surveillance-Program
-4
.s (EXD11'TED SNUBBERS TO IIE ADDED LATER.)
SEQUOYAli - UNIT .1 3/4 7-30
o Pages. 3/4 7-31 through .3/4 7-36a deleted 4 m 4 f SEQUOYAli - UNIT I
,j PLANT SYSTEMS 3/4.7.10 SEALED SOURCE CONTAMINATION
\
LIMITING CONDITION FOR OPERATION 3.7.10 Each sealed source containing radioactive material either in excess of 100 microcuries of beta and/or gamma emitting material or 5 microcuries of alpha emitting material shall be free of greater than or equal to 0.005 micro-curies of removable contamination. APPLICABILITY: At all times. ACTION:
- a. With a sealed source having removable contamination in excess of the above limits, immediately withdraw the sealed source from use and:
- 1. Either decontaminate and repair the sealed source, or
- 2. Oispose of the sealed source in accordance with Commission Regulations.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.7.10.1 Test Recuirements - Each sealed source shall be tested for leakage j and/or contamination by: l a. The licensee, or l l b. Other persons specifically authorized by the Commission or an I Agreement State. The test method shall have a detection sensitivity of at least 0.005 microcuries per test sample. l 4.7.10. 2 Test Frecuencies - Each category of sealed sources (excluding startup sources and fission detectors previously subjected to core flux) shall
; be tested at the frequency described below.
I
- a. Sources in use - At least once per six months for all sealed sources containing radioactive materials:
- 1. With a half-life greater than 30 days (excluding Hydrogen 3),
and
, 2. In any form other than gas.
s SEQUOYAH - UNIT 1 3/4 7-37 i
a . PLANT SYSTEMS
\
SURVEILLANCE REOUIREPENTS (Continued)
- h. Stored sources not in use - Each' sealed source and fission detector shall be tested prior to use or transfer to another licensee unless tested within the previous six months. Sealed sources and fission detectors transferred without a certificate indicating the last test date shall be tested prior to being placed into use.
- c. Startuo sources and 'ission detectors - Each sealed startup source and fission detector shall be tested within 31 days prior'to being subjected to core flux or installed in the core and following repair or maintenance to the source.
d.7.10. 3 Recorts - A report shall be prepared and submitted to the Conmission on an annual basis if sealed source or fission detector leakage tests reveal the presence of greater than or ecual to 0.005 microcuries of. removable contamination. l e 4 SE000YAH - UNIT 1- 3/4 7-38 i -
PLANT SYSTEMS
-i FIRE HOSE STATIONS L: .;
LIMITING CONDITION FOR OPERATION 3.7.11.4 The fire hose stations shown in Table 3.7-5 shall be OPERABLE. , APPt1CABILITY: Whenever equipment in the areas protected by the fire hose stations is required to be OPERABLE. 1 ACTION:
- a. With one or more of the fire hose stations shown in Table 3.7-5 inoperable, route an additional equivalent capacity fire hose to the unprotected area (s) from an OPERABLE hose station within I hour if the inoperable fire hose is the primary means of fire suppression; other-wise, route the additional hose within 24 hours. Restore the fire hose station to OPERABLE status within 14 days or, in lieu of any other report required by Specification 6.9.1, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days outlining the action taken, the cause of the inoperability, and plans and schedule for restoring the station to OPERABLE status.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS O 4.7.11.4 Each of the fire hose stations shown in Table 3.7-5 shall be demonstrated OPERABLE:
~
- a. At least once per 31 days by a visual inspection of the fire hose stations accessible during plant operations to assure all required equipment is at the station. ;
, b. At least once per 18 months by:
- 1. Visual inspection of all the stations not accessible during plant operations to assure all required equipment is at the station.
- 2. Removing the hose for inspection and re-racking, and
- 3. Inspecting all gaskets and replacing any degraded gaskets in the couplings.
- c. At least once per 3 years by:
- 1. Partially opening each hose station valve to verify valve OPERABILITY and no flow blockage.
- 2. Conducting a hose hydrostatic test at a pressure of 300 psig or tg at least 50 psig above maximum fire main operating pressure, In/A whichever is greater.
SEQUOYAH - UNIT 1 3/47-45 9
Table 3.7-5'(Continued) FIRE HOSE STATIONS LOCATION . ELEVATION HOSE RACK #
- g. CCW Intake Pumping Station 690
- 0-26-866 690 0-26-867-690 0-26-868 690 0-26-869 1690 0-26-870
- h. ERCW Pumping Station 688 0-26-927-688 26-926-688 0-26-930-704- 26-931 704 0-26-925 704 0-26-923 720 0-26-929.
-Ah 720 0-26-924
'Y - 720 0-26-932 o
(.c{N SEQUOYAH --UNIT'1 3/4 7-48
.i
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- 4. Verifying the diesel starts from ambient condition and accelerates to at least 900 rpm in less than or equal to .
10 seconds. The generator voltage and frequency shall be 6900 1 690 volts and 60 1 1.2 Hz within 10 seconds after the start signal. The diesel generator shall be started for this test by using one of the following signals with startup on each signal verified at least once per 124 days:
] a) Manual.
b) Simulated loss of offsite power by itself.
' c) Simulated loss of offsite power in conjunction with an ESF actuation test signal.
i d) An ESF actuation test signal by itself.
- 5. Verifying the generator is synchronized, loaded to greater than -
or equal to 4000 kw in less than or equal to 60 seconds, and operates for greater than or equal to 60 minutes, and
- 6. Verifying the diesel generator is aligned to provide standby power to the associated shutdown boards.
- b. At least once per 31 days and after each operation of the diesel where the period of operation was greater than or equal to 1 hour by checking for- and removing accumulated water from the engine-mounted fuel tanks.
, c. At least once per 92 days and from new fuel oil prior to addition to the 7-day tanks by verifying that a sample obtained in accordance with ASTM-0270-1975 has a water and sediment content of less than or equal to .05 volume percent and a kinematic viscosity @ 100 F of l
greater than or equal to 1.8 but less than or equal to 5.8 centi-stokes when tested in accordance with ASTM-D975-77, and an impurity level of less than 2 mg. of insolubles per 100 ml. when tested in accordance with ASTM-02274-70.
- d. At least once per 18 months, during shutdown, by:
- 1. Subjecting the diesel to an inspection in accordance with procedures prepared in conjun:: tion with its manufacturer's c recommendations for this class.of standby service,
- 2. Verifying the generator capability to reject a load of greater than or equal to 600 kw while maintaining vol_tage at 6900 1 690 volts and frequency at 60 1 1.2 Hz.
- 3. Verifying the generator capability to reject a load of'4000 kw '
without tripping. The generator voltage shall not exceed 7866 volts during and following the load rejection.
- 4. Simulating a loss of offsite power by itself, and:
a) Verifying de-energization of the shutdown boards and load shedding from the shutdown boards. SEQUOYAH - UNIT 1 3/4 8-3 I
i l'
! ELECTRICAL POWER SYSTEMS 4 '
, SURVEILLANCE REQUIREMENTS-(Continued)
Within 5 minutes after completing this 24 hour test, perform : Specification 4.8.1.1.2.c.4. The generator' voltage and fre-
..quency shall be 6900 690 volts and 60 1.2 Hz within 10 seconds after the start signal; the steady state ~ generator voltage and frequency shall be maintained within these limits during this test.
- 9. Verifying that the auto-connected loads to each diesel generator do not exceed the 2000 hour rating of 4000 kw.
- 10. Verifying the diesel generator's capability to:
a)- Synchronize with the offsite power source'while the generator is loaded with'its emergency loads upon a simulated 4 restoration of offsite power. l- b) Transfer its loads to the offsite power source, and c) Be restored to its shutdown status.
- 11. Verifying that.with the diesel generator operating in a test mode (connected to its bus), a simulated safety injection signal overrides the test mode by (1) returning the diesel generator to standby operation and (2)' automatically energizing the emergency loads with offsite power.
- 12. Verifying that the automatic load sequence timers are OPERABLE with the setpoint for each sequence timer within + 5 percent of its design setpoint. _
- 13. Verifying that the following diesel generator lockout features prevent diesel generator starting only when required:
a) Engine overspeed b) 86 7 lockout relay
- e. At least once per 10 years or after any modifications which could affect diesel generator interdependence by starting the diesel _ .,
generators simultaneously,_ during shutdown, and verifying that.the diesel generators accelerate to at least 900 rpm in less than or equal to 10 seconds.
- f. At least once per 10 years
- by: -
- 1. Draining each fuel oil storage tank, removing the accumulated sediment and cleaning the tank using a sodium hypoclorite solution, and SEQUOYAH - UNIT 1 3/4 8-5
*These requirements are waived for the initial surveillance. -
EllCTRICAL' POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- 2. Performing a pressure test of those portions of the diesel fuel oil system designed to Section III, subsection ND of the ASME Code at a test pressure equal to 110 percent of the system
! design pressure. 4.8.1.1.3 The 125-volt D.C. distribution panel, 125-volt 0.C. battery bank 4 ! and associated charger for each diesel generator shall be demonstrated {' OPERABLE: .
- a. At least once per 7 days by verifying:
- 1. That the parameters in Table 4.8-la meet the Category A limits. l
- 2. That the total battery terminal voltage is greater than or
- equal to 129-volts on float charge.
- b. At least once per 92 days by:
- 1. Verifying that the parameters in Table 4.8-la meet the Category B l limits, !
- 2. Verifying there is no visible corrosion at either terminals or connectors, or the cell to terminal connection resistance of these items is less than 150 x 10 8 ohms, and i'
- 3. Verifying that the average electrolyte temperature of 6 connected cells is above 60 F.
- c. At least once per 18 months by verifying that: l
- 1. The cells, cell plates and battery racks show no visual l indication of physical damage or abnormal deterioration. ,
- 2. The battery to battery and terminal connecticm. are clean, tigot and coated with anti-corrosion material.
- 3. The resistance of each ce}l to terminal connection is less than or equal to 150 x 10 8 ohms. ,
4.8.1.1.4 Reports - All diesel generator failures, valid or non-valid, shall be reported to the Commission pursuant to Specification 6.9.1. Reports of diesel generator failures shall include the information recommended in Regula-tory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 1977. If the number of failures in the last 100 valid tests (on a per nuclear unit .. basis) is greater than or equal to 7, the report shall be supplemented to include the additional information recommended in Regulatory Position C.3.b of l Regulatory Guide 1.108, Revision 1, August 1977. SEQUOYAH - UNIT 1 3/4 8-6
1-TABLE 4.8-la 1
-, . DIESEL GENERATOR BATTERY SURVEILLANCE REQUIREMENTS __
g
-CATEGORY A(1) CATEGORY B(2),
Parameter Limits for each Limits for each Allowable (3) designated pilot connected cell value for each cell connected cell Electrolyte >Minhum level > Minimum level Above top of Level
~
indication mark,' indication mark, plates, and < h" above and <'\" above and not
-maxisum level maxi 5um level overflowing indication mark indication mark Float Voltage > 2.13 volts
> 2.13 volts (c) > 2.07 volts?
Not more than
.020 below the average of all
> 1.190 connected cells N Specifi # 1.195(b)
Gravity [a) Average of all Average of all-connected cells connectg) cells
> 1.200 > 1.190 (a) Corrected for electrolyte temperature and level.
(b) Or battery charging current is less than 2 amps. (c) Corrected for average electrolyte temperature. (1) For any Category A parameter (s) outside the limit (s) shown, the battery
. nay be considered OPERABLE provided that within 24 hours all the Category 8 measurements are taken and found to be within their allowable values, and provided all parameter (s) are restored to within limits within the next 6 days.
(2) For any Category B parameter (s) outside the limit (s) shown, the battery , may be considered OPERABLE provided that they are within their allowable values and provided the parameter (s) are restored to within limits'within 7 days. (3) Any Category B parameter not within its allowable value' indicates an inoperable battery. -
' t SEQUOYAH - UNIT 1 3/4 8-7a 4
r . -
-,- -- c.., -, . ~ _ - ,_ _ .-
ELECTRICAL POWER SYSTEMS y,; (@{$P- L3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS A.C. DISTRIBUTION - OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.1 The following A.C. electrical boards and inverters shall bs OPERABLE and energized with tie breakers open between redundant boards: 6900 Volt Shutdown Board 1A-A 6900 Volt Shutdown Board IB-B 6900 Volt Shutdown Board 2A-A 6900 Volt Shutdown Board 28-B 480 Volt Shutdown Board 1Al-T 480 Volt Shutdown Board lA2-A 480 Volt Shutdown Board 181-B 480 Volt Shutdown Board 182-B 480 Volt Shutdown Board 2Al-A 480 Volt Shutdown Board 2A2-A 480 Volt Shutdown Board 281-B 480 Volt Shutdown Board 2B2-B 120 Volt A.C. Vital Instrument Power Board Channels 1-I and 2-I energized from inverters 1-I and 2-I connected to D.C. Channel I*. 120 Volt A.C. Vital Instrument Power Board Channels 1-II and 2-II energized from inverter 1-II and 2-II connected to D.C. Channel II*. 120 Volt A.C. Vital Instrument Power Board Channels 1-III and 2-III
$$$5- energized from inverter 1-III and 2-III connected to D.C.' Channel II:
- P 120 Volt A.C. Vital Instrument Power Board Channels 1-IV and 2-IV energized from inverter 1-IV and 2-IV connected to D.C. Channel IV".
APPLICABILITY: MODES 1, 2, 3 and 4. ACTION:
- a. With less than the above complement of A.C. boards GPERABLE and energized, restore the inoperable boards-to OPERABLE status within 8 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With one inverter inoperable, energize the associated Vital Instrument Power Board within 8 hours; restore the inoperable inverter to OPERABLE status within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.8.2.1 The specified A.C. boards and inverters shall be determined OPERABLE and energized with tie breakers open between redundant boards at least once per 7 days by verifying correct breaker alignment and indicated voltage on the busses.
~~'
Two invertes may be disconnected from their D.C. source for up to 24 hours for th. purpose of performing an equalizing charge on their associated battery bank provid, f2L<di i ,. (1) the vital instrument power board is OPERABLE and energized, and (2) the vital - instrument power boards associated with the other battery banks are OPERA 8LE and energized from their respective inverters connected to their respective O.C. sourc-SEQUOYAH - UNIT 1 3/4 8-10 I u
._.m. _ _ ; ._m. . ~ _ _ .
D f 4 L
4 w., . ELECTRICAL POWER SYSTEMS A.C. DISTRIBUTION - SHUTDOWN Y.e - LIMITING CONDITION FOR OPERATION 3.8.2.2 As a minimum, the following A.C. electrical boards and inverters shall be OPERABLE and energized: 2- 6900 volt shutdown boards, either l A-A and 2A-A or 1B-B and 28-8, t- 4- 480 volt shutdown boards associated with the required OPERABLE 6900 volt shutdown boards, 2- 120 volt A.C. vital instrument power boards either Channels I and III or Channels II and IV energized from their respective inverters connected to their respective D.C. battery banks, and 480 volt shut-down boards. APPLICABILITY: - MODES 5 and 6. I ACTION: With less than the above complement of A.C. boards and inverters OPERABLE and energized, establish CONTAINMENT INTEGRITY within 8 hours. g.:: o SURVEILLANCE REOUIREMENTS 4.8.2.2 The specified A.C. boards and inverters shall be determined OPERABLE and energized at least once per 7 days by verifying correct breaker alignment
- and indicated voltage on the bus.
t t , SEQUOYAH - UNIT 1 ! 3/.4 8-11 n,,*mz
. .,. y -) v .: -...?
- - - , - , , -.. , , - - - n ---
f o a , . l ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS 4.8.2.3.1 Each D.C. bus train shall be determined OPERABLE and energized with tie breakers open between redundant busses at least once per 7 days by verifying correct breaker alignment, indicated power availability from the charger. and battery, and voltage on the bus of greater than or cqual to 125 volts. 4.8.2.3.2 Each 125-volt battery bank and charger shall be demonstrated OPERABLE:
- a. At least once per 7 days by:
- 1. Verifying that the parameters in Table 4.8-2 meet the Category A limits, and
- 2. Verifying total battery terminal voltage is greater than or equal to 129-volts on float charge. -
- b. At least once per 92 days and within 7 days after a battery discharge (battery terminal voltage below 110-volts), or battery overcharge (battery terminal voltage above 150-volts), by:
- 1. Verifying that the parameters in Table 4.8-2 meet the Category B limits,
- 2. Verifying there is no visible corrosion at either terminals or connectors, or the connection resistance of these items is less than 150 x 10 8 ohms, and
- 3. Verifying that the average electrolyte temperature of 6 connected cells is above 60 F.
! c. At least once per 18 months by verifying that:
- 1. The cells, cell plates, and battery racks show no visual indication of physical damage or abnormal deterioration,
- 2. The cell-to-cell and terminal connections are clean, tight, and '
coated with anti-corrosion material,
- 3. The resistance of each cell-to-terminal connection is less than
~
or equal to 150 x 10 8 ohms, and
- 4. The battery charger will supply at least 150 amperes at 125-volts
! for at least 4 hours.
l - SEQUOYAH - UNIT 1 M S4 2
I
' \
; +- ,e ,
7,,_ _- n. ,' ,
. ELECTRICAL POWER SYSTEMS-p p .. - .
^
m ' n . u SURVEILLANCE REQUIREMENTS (Continued)i J-t '
.1 e M, ,
/' - , ,[ X
- d. At least once per 18 months by verifying that th,e b;atterp capacity _, .
is adequate to supply and maintairifjn OPERABLE status al1 of the 5 is . -
'^
actual or simulated emergency loads for;2, hours whensthe battery iv subjected to a battery servicestesti , j % , m, D~ c , .
- e. At least once per 60 months by verifying'that the battery capacity ^ -q
- s. < ..
is at least 80% of the manufacturers; rating?when ' subjected to a _ ;, performance' discharge test. Once per' 60 month interval, this {~2,, performance discharge tu t may be performed in lieu of the: battery % service test. '
- f. Annual performance discharge tests _ battery capacity shall e; . } ' ~
s ,
.given to any battery that shows sighs of 'degr:adation or. has reached.; -
85% of the service life expected for.the spplication. ' Degfaf h tion ,4 1 . ,f is indicated when the battery capacity drops"more t,han 10% of. rated: L _ 1 -r capacityfromitsaverageonpreviousperform.3ncetests,orisb,elow^1-90% of the manufacturer's rating.~ ~ Q "' '
.g ~ w e! '
r m ,ny , , ,
't.
g 4
/
'e Y p?
4 .. 4 ( _ 1,
,% [ %'. '*'g a
- n. s, - ,p."
# d- N N
4 N ' [,7 ^g W
~
j" % * ' pr s
}, 4 ,-a y* _,_
E.- -.. + -
%s3 "
s.. A w O-x 4 g >
* ~,
y
+
'f
' # . *^
r, ,
* * - e .
,\
V . g "g _. x .
*=
9g
,~
/
e 9 m = #
~3 N
. g ,
'. s. .,,,. +
w '47 fu.
; SEQUOYAH - UNIT 1 3/4 8-13 , ; y > s
- s. -
~
.g o.'
~
-d 'w h
- j sI
- a~y; '7- ; .
.m 1 - -c
* ' % x 1,_ .=. .
,. wa , . - .-
\
4 A - w m.,..
'r, m A ./ .
x q . s . .,z e t. .- ..
.. . '1'/-
.M~ ,%
TABLE 4.8-2 4
- e. .
k( BATTERY SURVEILLANCE REOUIREMENTS y~ , -.c.c / , si a w - N CATEGORY B(2) TP
~
m - *
- v. CATEGORY A(1) .
w ,< - s .Paralieter
'./s Limits for each~ Limits for each Allowable (3) value for each
'\ designated pilot connected cell "li
- .\ -
' cell connected cell Els:trolyte- / >fiinimum level- > Minimum level Above tcp of tevel x ir,dication mark, indication mark, plates, H.g ~:; and i k" above and i " above and not
' 1'
-3 overflowing
~ maximum level maximum level sy ? indication mark indication mark
"}
- ' :2 ~~ , ss - -
JJ 'Floa't Voltage '~~ > 2.13 volts > 2.13 volts (C) > 2.07 volts
/ . ...
Not more than
.020 below the s.'
J average of all
> 1.195 connected cells e .,
b . s 0)~ & , k. ' Specjfig)i Gravity $^\~>1.20'0 Average of all Average of all
- n. vb s
'-2
A- m s-
, , ' ': connected cells
> 1.205 .
connecteg) _ 1.195
\
cells 34, ^ " , x J s' .x s. - W, A s ,
'\ e' ,
** g.
(O. C cted for-klectro)yte temperature and level.
- (b)?. orreOr, battery charging current is less than 2 amps. 'f . "'3 ,
(c)\ Corrected fortaverage~ electrolyte temperature. . (1) For any Category A parameter (s) outside the limit (s) shown, the battery
~ , .
7y . Oc' may be considered 0?fRABLE provided t_ hat within 24 hours all the Category B measurements are-taken and found-to be within their allowable values, and ~ N provided all paraw;er(s),are restored to within limits within the next
. 1 ;';x e ~ m 6 day @ 's Y. .'.I ? As 4.s i,. u (1) '. For a3y Category -B parameter (s) outside the limit (s) shown, the battery -
may be scosicered CPERABLE provided that they are within their allowable
'A 4 values dad provide 5 tl:e'p3rameter(s) are restored to within limits within
~
"' A days.~.a ; ?
(.
'4 S(3) Any Category 'B parameter 71ot within its allowable value indicates an
. s, 4 y . a "iroperable battery.
,Z s ' if 7 , ^y.
+ Q 4 ~' 4 . ,i , (" t. \ 3;csy ,7 <- y v.'.. ,5 , 2.
-% p. -
T,w O, . g .- - .
- 1. -
, 'SEQt!0VMi [ NIT 1 s, s\' s] - . 3/.4 8- 13a
- - - . . xa q
x y N . s
*t .,
. l ., ( v,.
-' . ,' s * .
) N D. ., i -#
,<r up:% .x u.
6- .
- v. %
s \ % u 7 ' , ~. .
- un s
*lq. * ; _ . , ~.
s'
/ -s J'< .j ' .
1 . TABLE 3.8-2 MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION AND/0R BYPASS DEVICES
~
" " Valve No. Function . . . . ._ Bypass Device 1-FCV-62-63 -s Isolation for Seal Water Filter No 1- FCV 138 Safe Shatdown Redundancy (CVCS) No 1- FCV-62-98 ECCS Operation No 1-FCV-62-99 ECCS Operation No 1-FCV-62-90 ECCS Operation No 1- FCV-62-91 ECCS Operation No 1- FCV-62-61 Cont. Isolation No 1- LCV-62-132 ECCS Operation No 1- LCV 133 ECCS Operation No 1-LCV-62-135 ECCS Operation No 1- LCV-62-136 ECCS Operation No 1-FCV-74-1 Open for Normal Plant Cooldown No 1-FCV- 74-2 Open for Normal Plant Cooldown No 1-FCV-74-3 ECCS Operation No 1-FCV-74-21 ECCS Operation No 1-FCV-74-12 RHR Pump, Mini-flow Protects Pump No 1-FCV-74-24 RHR Pump, Mini-flow Protects Pump No 1-FCV-74-33 ECCS Operation No
. 1-FCV-74-35 ECCS Operation No ECCS Operation No W' 1-FCV-74-7 No 1-FCV-74-6 ECCS Operation 1-FCV-63-156 ECCS Flow Path No 1-FCV-63-157 ECCS Flow Path No 1-FCV-63-39 BIT Injection No 1-FCV-63-40 GIT Injection No 1-FCV-63-25 BIT Injection No 1 FCV-63-26 e T Injection No 1 FCV-63-il8 RCS Pressure Boundary No 1-FCV-63-98 RCS Pressure Bourdary No 1-FCV-63-80 RCS Pressure Boundary No 1-FCV-63-67 RCS Pressure Boundary No 1-FCV-63-1 ECCS Operation No 1-FCV-63-72 ECCS Flow Path from Cont. Sump No 1-FCV-63-73 ECCS Flow Path from Cont. Sump No 1-FCV-63-8 ECCS Flow Path No 1-FCV-63-11 ECCS Flow Path No ECCS Cooldown Flow Path No 1-FCV-63-93 ECCS Cooldown Flow Path No 1-FCV-63-94 1-FCV-63-172 ECCS Elow Path 'No 1-FCV-63-5 ECCS Flow Path No 1-FCV-63-47 Train Isolation No 1-FCV-63-48 Train Isolation . No 1-FCV-63-4 SI Pump Mini-flow No 1-FCV-63-175 SI Pump Mini-flow No C3 3/4 8-35 SEQUOYAH - UNIT 1
'l ,
lN
f TABLE 3.8-2 (Continued) f MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION AND/0R BYPASS DEVICES Valve No. Function Bypass Device 1- FCV-63-3 SI Pump Mini-flow No 1- FCV-63-152 ECCS Recirc 'No 1- FCV 153 ECCS Fectre No 1-FCV-63-22 ECC5 Recirc No 1-FCV-3-33 Quick Closing Isolation No 1-FCV-3-47 Quick Closing Isolation No 1-FCV-3-87 Quick Closing Isolation No 1-FCV-3-100 Quick Closing Isolation No 1- FCV-1-15 Stm Supply to Aux FWP turbine No 1-FCV-1-16 Stm Supply to Aux FWP turbine No 1-FCV-3-179A ERCW Sys Supply to Pump No 1-FCV-3-179B ERCW Sys Supply to Pump No 1-FCV-3-136A ERCW Sys Supply to Pump No 1-FCV-3-136B ERCW Sys Supply to Pump No 1- FCV-3-il 6A. ERCW Sys Supply to Pump No 1-FCV-3-il 68 ERCW Sys Supply to Pump No 1-FCV-3-126A ERCW Sys Supply to Pump No 1-FCV-3-1268 ERCW Sys Supply to Pump No . jn 1-FCV-70-133 Isolation for. RCP Oil Coolers & Therm B No :1 1-FCV-70-139 Isolation for RCP Oil Coolers & Therm B No 1-FCV-70-4 Isolation for Non-Essential Loads No 1-FCV-70-143 Isolation for Excess Letdown Ht Xchngr No 1-FCV-70-92 Isolation for RCP Oil Coolers & Therm B No 1-FCV-70-90 Isolation for RCP Oil Coolers & Therm B No 1-FCV-70-87 Isolation for RCP Oil Coolers & Therm B No 1-FCV-70-89 Isolation for RCP Oil Coolers & Therm B No 1-FCV-70-140 Isolation for RCP Oil Coolers & Therm B No 1-FCV-70-134 Isolation for RCP Oil Coolers & Therm B No 1-FCV-67-67* OG Ht Ex No 2-FCV-67-65* OG Ht-Ex No 1-FCV-67-66* OG Ht Ex No 2-FCV-67-68* DG Ht Ex No 1-FCV-67-123 CSS Ht Ex Supply No 1-FCV-67-125 - CSS Ht Ex Supply No 1-FCV-67-124. CSS Ht Ex Discharge No 1-FCV-67-126 CSS Ht Ex Discharge No - 0 -FCV-67-151
- CCW Ht Ex Throttling No 0-FCV-67-152* CCW Ht Ex Throttling No 2-FCV-67-146 CCW Ht Ex Throttling No 2-FCV-67-223 Isolation of IB/2A HOR's' No 1-FCV-67-83 Cont. Isol. Lower .No 1-FCV-67-88 Cont. Isol. Lower No 1-FCV-67-87 Cor.t. Isol. Lower No 1-TCV-67-424* CCW Ht Ex Isolation NO 1-FCV-67-478* Isolation of.1B ERCU HDR NO (.r3{p SEQUOYAH - UNIT 1. 3/4 8-36 I
- Common to Units 1&2 I
,e e f . ..
(.j TABLE 3.8-2 (Continued) MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION AND/0R BYPASS DEVICES
- Valve No. Function Bypass Device 1-FCV-67-95 Cont. Isol. Lower No 1-FCV-67-96 Cont. Isol. Lower No 1-FCV-57-91 Cont. Isol. Lower No 1-FCV-67-103 Cont. Isol. Lower No 1-FCV-67-104 Cont. Isol. Lower No 1-FCV-67-99 Cont. Isol. Lower No 1-FCV-67-111 Cont. .Isol. Lower No 1-FCV-67-112 Cont. Isol. Lower No 1-FCV-67-107 Cont. Isol. Lower No ;
1-FCV-67-130 Cont. Isol. Upper No 1-FCV-67-131 Cont. Isol. Upper No 1-FCV-67-295 Cont. Isol. Upper No 1-FCV-67-134 Cont. Isol. Upper No 1-FCV-67-296 Cont. Isol. Upper No 1-FCV-67-133 Cont. Isol. Upper No 1-FCV-67-139 Cont. Isol. Upper No 1-FCV-67-297 Cont. Isol. Upper No (:n 1-FCV-67-138 Cont. Isol. Upper Cont. Isol. Upper No Q;g 1-FCV-67-142 No 1-FCV-67-?o8 Cont. Isol. Upper No 1-FCV-67-141 Cont. Isol. Upper No 1-FCV-72-21 Cont. Spray Pump Sv Mon No 1-FCV-72-22 Cont. Spray Pump Su ion No 1-FCV-72-44 Cont. Spray ? ump Suction No 1-FCV-72-45 Cont. Spray Pump Suction No 1-FCV-72-2 Cont. Spray Isol. No 1-FCV-72-39 Cont. Spray Isol. No 1-FCV-72-40 RHR Cont. Spray Isol. No 1-FCV-72-41 RHR Cont. Spray Isol. No u j , SEQUOYAH - UNIT 1 3/4 8-37 i
L 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3 9.1 With the reactor vessel head closure bolts less than fully tensioned or witt the head removed, the boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the following reactivity conditions is cet:
- a. Either a K,ff of 0.95 or less, which includes a 1% delta k/k conser-vative allowance for uncertainties, or
- b. A boron concentration of greater than or equal to 2000 ppm, which includes a 50 ppm conservative allowance for uncertainties.
APPLICABILITY: MODE 6* ACTION: With the requirements of the above specification not satisfied, _immediately i suspend all operations involving CORE ALTERATIONS or positive reactivity - changes and initiate and continue boration at greater than or equal to 10 gpm of a solution containing 20,000 ppm boron or its equivalent until Keff is reduced to less than or equal to 0.95 or the boron concentration is restored tc greater than or equal to 2000 ppm, whichever is the more restrictive. The-f provisions of Specification 3.0.3 are not applicable. SURVEILLANCE RE0VIREMENTS 4.9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to:
- a. Removing or unbolting the reactor vessel head, and
- b. Withdrawal of any full length control rod in excess of 3 feet from
its fully it..erted position within the reactor pressure vessel. I l
-1
^The reactor shall be maintained in MODE 6 whenever fuel is in the reactor vessel with the vessel head closure bolts less than fully tensioned or with-the head removed. ,
6 SEQUOYAH - UNIT 1 3/4 9-1 l i 1
1 REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued) e . 4.9.1.2 The boron concentration of the reactor coolant system and the refueling canal shall be determined by chemical analysis at least once per 72 hours. 4.9.1.3 One of the following valve combinations shall be verified closed under administrative control at least once per 72 hours: Combination B Combination C Combination 0 Combination A
- a. 2-81-536 a. 2-81-536 a. 2-81-536
- a. 2-81-536 b. 2-62-907
- b. 2-62-922 b. 2-62-922 b. 2-62-907
- c. 2-62-916 c. 2-62-914 c. 2-62-914
- c. 2-62-916 d. 2-62-921
- d. 2-62-933 d. 2-62-940 d. 2-62-921
- e. 2-62-696 e. 2-62-933 e. 2-62-940
- f. 2-62-929 f. 2-62-929
- g. 2-62-932 g. 2-62-932 2-FCV-62-128 h, 2-62-696 h.
- i. 2-FCV-62-128 9
SEQUOYAH - UNIT 1 3'/4 9- la
REFUELING OPERATIONS 1 3/4.9.2 INSTRUMENTATION LIMITING CONDITION FOR OPERATION . 3.9.2 As a minimum, two source range neutron flux monitors shall be OPERABLE and operating, each with continuous visual indication in the control room and one with audible indication in the containment and control rocm. APPLICABILITY: MODE 6. ACTION:
- a. With one of the above required monitors inoperable or not operating, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes. ,
- b. With both of the above required monitors inoperable or not operating,
, determine the boron concentration of the reactor coolant system at least once per 12 hours.
- c. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.9.2 Each source. range neutron flux monitor shall be demonstrated OPERABLE by performance of: o a. A CHANNEL CHECK at least once per 12 hours,
- b. A CHANNEL FUNCTIONAL TEST at least once per 7 days, and
- c. A CHANNEL FUNCTIONAL TEST within 8 hours prior to the initial start of CORE ALTERATIONS.
SEQUOYAH - UNIT 1 3/4 9-2 a i e
REFUELING OPERATIONS ', 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS LIMITING CONDITION FOR OPERATION 3.9.4 The containment building penetrations shall be in the following status:
- a. The equipment door closed and held in place by a minimum of four bolts,
- b. A minimun of one door in each airlock is closed, and
- c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:
- 1. Closed by an isolation valve, blind flange, or manual valve, or
- 2. Be capable of being closed by an OPERABLE automatic Containment Ventilation isolation valve.
APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment. ACTION: With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel in the containment building. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE REOUIREMENTS 4.9.4 Each of the above required containment building penetrations shall be determined to be either in its closed / isolated condition or capable of being closed by an OPERABLE automatic Containment Ventilation isolation valve within 100 hours prior to the start of and at least once per 7 days during CORE ALTERATIONS or movement of irradiated fuel in the containment building by: ,
- a. Verifying the penetrations are in their closed / isolated condition, or
- b. Testing the Containment Ventilation isolation valves per the applicable portions of Specification 4.6.3. 2 ,
SEQUOYAH - UNIT 1 3/4 9-4 1__
REFUELING OPERATIONS
.?:::.
7#' 3/4.9.6 MANIPULATOR CRANE LIMITING CONDITION FOR OPERATION 3.9.6 The manipulator crane and auxiliary hoist shall be used for movement of drive rods or fuel assemblies and shall be OPERABLE with:
- a. The manipulator crane used for movement of fuel assemblies having:
- 1. A minimum capacity of 2750 pounds, and
- 2. An electrical overload cut of f limit less than or equal to 2700 pounds.
- b. The auxiliary hoist used for latching and unlatching drive rods having:
- 1. A minimum capacity of 610 pounds, and
- 2. A loao indicator which shall be used to prevent lifting loads ...,
in excess of 600 pounds. *f' APPLICABILITY: During movement of drive rods or fuel assemblies within the reactor pressure vessel. ACTION: With the requirements for crane and/or hoist OPERABILITY not satisfied, suspend use of any inoperable mainipulator crane and/or auxiliary hoist from operations involving the movement of drive rods and f ael assemblies within the reactor pressure vessel. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE REQUIREMENTS 4.9.6.1 Each manipulator crane used for movement of fuel assemblies within the reactor pressure vessel shall be demonstrated OPERABLE within 100 hours prior to the start of such operations by perferming a load test of at least 2750 pounds and demonstrating an. automatic electrical load cut off when the crane load exceeds 2700 pounds. 4.9.6.2 Each auxiliary hoist and associated load indicator used for movement of drive rods within the reactor pressure vessel shall be demonstrated OPERABLE ,, within 100 hours prior to the start of such operations by performing a load 1. test of at least 610 pounds. SEQUOYAH - UNIT 1 3/4 9-6
t REFUELING OPERATIONS LOW WATER LEVEL LIMITING CONDITION FOR OPERATION t 3.9.8.2 Two independent Residual Heat Removal (RHR) loops shall be OPERABLE.* A?PLICABILITY: MODE 6_when the water level above the top of the reactor
-pressure vessel flange is less than 23 feet.**
t ACTION:
- ~
- a. With less than the required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERAFLE status as soon as possible.
, b. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REOUIREMENTS 4.9.8.2 The required Residual Heat Removal loops shall be determined OPERABLE' per Specification 4.0.5. ? I "The normal or emergency power source may be inoperable for each RHR leo,0.
** Prior to initial criticality only one independent RHR loop shall be required OPERABLE.
SEQUOYAH - UNIT 1 3/49-8a t. l l
s i 3/4.10 SPECIAL TEST FXCEPTIONS 3/4.10.1 SHUTDOWN MARGI.i LIMITING CONDITION FOR OPERATION ,_ 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.! may be suspended for measurement of control rod worth and shutdown margin provided reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from OPERABLE control rod (s). AFPLICABILITY: MODE 2. ACTION: I
- a. With any full length control rod not fully inserted and with less than the above reactivity equivalent available for trip insertion, immediately ir.itiate and continue baration at greater than or equal to 10 gpm of a solution containing 20,000 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.
- b. With all full length control rods inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and continue boration at greater than or equal to 10 gpm of a solution containing 20,000 ppm boron or its equivalent until the SHUTOOWN MARGIN required by Specification 3.1.1.1 is restored.
~^
l SURVEILLANCE REQUIREMENTS 4.10.1.1 The position of each full length rod either partially or fully withdrawn shall be determined at least once per 2 hours. 4.10.1.2 Each full length rod not fully inserted shall be demonstrated capable of full insertion when tripped from at least 50% withdrawn position within 24 hours prior to reducing the SHUTDOWN MARGIN to less than the limits of -- Specification 3.1.1.1. SEQUOYAH - UNIT 1 3/4 10-1
TABLE 4.11-2 (Continued)
. TABLE NOTATION l
- b. Analyses shall also be performed following shutdown from 115% RATED
- THERMAL POWER, startup to 215% RATED THERMAL POWER or a THERMAL
. POWER change exceeding 15 percent of RATED THERMAL POWER within a one hour period.
- c. Tritium grab samples shall be taken at least once per 24 hours when '
the refueling canal is flooded,
- d. Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours after changing sor after removal from sampler). Sampling shall also be performed at least.once per 24 hours for at least 2 days following each shutdown from 215% RATED THERMAL POWER, startup to 315% RATED THERMAL TWER or THERMAL POWER change exceeding 15% of RATED THERMAL PCWER in one hour and analyses shall be completed within 48 hours of changing. When samplas collected for 24 hours are analyzed, '.he corresponding LLD's may Se increased by a factor of 10.
- e. Tritium grab samples shall be taken at' least once per 7 cys from the ventilation exhaust,from the spent fuel pool area, whermver spent fuel is in the spent fuel pool.
- f. The ratio of the sample fkr rate to tb sampled stream flow rate shall be kne n for the time period covered by each dose or dose rate calculation made in accordance with Specifications 3.11.2.1, 3.11.2.2 and 3.11.2.3.
~
- g. The principal gamma emit.ters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Hn-54, Fe-59,
. Co-58. Co-60, 2n-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions. This list does not mean that oniy these - ~
nuclides are to be detected and reported. Other peaks which are measureable and identifiable, together with the above nuclides, shall also be identified and report. l l
- h. During releases via this exhaust system.
- i. In MODES 1, 2, 3 and 4, the upper and lower compartments of the con-tainment shall be sampled prior to VENTING or PURGING. Prior to entering MODE 5, the upper and lower compartments of the containment shall be sampled. The incore instrument room purge sample shall be obtained at the shield building exhaust between 5 and 10 mir.utes
. ft,llowing initiation of the incore instrument room purge..
SEQUOYAH - UNIT 1 3/4 11-11
i l 1 ', RADI0 ACTIVE EFFLUENTS EXPLOSIVE CAS MIXTURE . IIMITING CONDITION FOR OPERATION 3.11.2.5 The concentration of oxygen in the waste gas holdup system shall be limited to less than or equal to 2% by volume whenever the hydrogen concentration exceeds 4% by volume. APPLICABILITY: At all times. ACTION:
- a. With the concentration of oxygen in a waste gas holdup tank greater than 2% by volume but less than or equal to 4% by volume, reduce the oxygen concentration to the above limits within 48 hours.
- b. With the concentration of oxygen in a waste gas holdup tank greater than 4% by volume and the hydrogen concentration greater than 2% by vohme, immediately suspend all additions of waste gases to the affected waste gas h.oldup tank and reduce the concentration of oxygen to less than or equa1 to 2% by volume within one hour. !
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
. SURVEILLANCE REOUIREMENTS 4.11.2.5 The concentration of hydrogen and oxygen in the waste gas holdup systen shall te determined to be within the above limits by monitoring the waste gas additions to the waste gas holdup system with the hydrogen and oxygen monitors required OPERABLE by Table 3.3-13 of Specification 3.3.3.10.
O SEQUOYAH - UNIT 1 3/4 11-15
v, TABLE 4.11-2 E RADI0 ACTIVE GASE0US WASTE MONITORItJG SAMPLING Afl0 AtlALYSIS PROGRAM-8 E Minimum Lower Limit of
, Sampling Analysis Type of Detection (pCi/ml)((LD) g Gaseous Release Type Frequency Frequency Activity Analysis .
G P P 4 A. Waste Gas Storage Fach Tank Each Tank Principal Gamma Emitters 9 1x10 Tank Grab Sample Pl Di B. Containment Purge Each Purge Each Purge Principal Gamma Emitters 9 lx10 -I Grab Sample
-6 H-3 1x10
~4 C. Noble Gases and M M Principal Gamma Emitters 9 lx10 Tritium Grab 1.CondensgrVacuum Sample -6 H-3 1x10 Exhaust 2.AuxiliakeBuilding s Exnaust a
- 3. Service Building 5 Exhaust a 4.Shieldgujljing Exhaust I d -12 D. Iodine and Parti- Continuous W I-131 lx10 ,
culates Sampler Charcoal
- 1. Auxiliary Building Sample I-133 lx10 -10 Exhaust d -Il Continuous W Principal Gamma Emitters 9 lx 10
- 2. Shield Building Sampler Particulate (!-131,Others)
Exhaust Sample
-II Continuous # M Gross Alpha lx10 Composite Sampler Particulate Sample Continuous I Sr-89, Sr-90 NO I Comp 0 site Sampler Particulate Sample -
-6 E. Noble Gases all Continuous # Noble Gas Noble Gases 1x10 Releases types as Monitor Monitor Gross Beta or Gamma listed in A, B, '
and C above ,
, RADI0 ACTIVE EFFLUENTS 3/4.11.4 TOTAL OOSE .
- . . . . . . - *v .
LIMITING CONDITION FOR OPERATION
~v . ..
3.11.4 The dose cr dose commitment to any member of the public, due to releases of radioactivity from uranium fuel cycle sources, shall be limited to less than or equal to 25 mrem to the total body or any organ (except the thyroid, which shall be limited to less than or equal to 75 mrem) over 12 consecutive months. APPLICABILITY: At all times. ACTION:
- a. With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice tne limits of Specifica-tions 3.ll.l.2.a 3.11.1.2.b, 3.ll.2.2.a. 3.ll.2.2.b, 3.ll.2.3.a. or 3.ll.2.3.b, in lieu of any other report required by Specification 6.9.1, prepare and submit a Special Report to the Director, Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, D.C.
20555, within 30 days, which defines the corrective action to be taken to reduce subsequent releases ts prevent recurrence of exceeding the limits of Specification 3.11.4. This Special Report shall include an analysis which estimates the radiation exposure (dose) to a member of the public from uranium fuel cycle sources (including all effluent pathways and direct radiation) for a 12 consecutive month period that includes the release (s) covered by this report. If the estimated dose (s) exceeds the limits of Specification 3.11.4, and if the release condition resulting in violation of 40 CFR 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR 190 and including the specified information of 9 190.11(b). The variance only relates to the limits of 40 CFR 190, and does not apply in any way to the requirements for dose limitation of 10 CFR Part 20, as addressed in other sections of this technical specification. .
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
l SURVt'ILLANCE REOUIREMENTS , 4.11.4 Dose Calculations Cumulative cose contributions from liquid and gaseous effluents shall be determined in accordance with Specifications 4.11.1.2, 4.11.2.2, and 4.11.2.3, and in accordance with the ODCM. SEQUOYAH - UNIT 1 3/4 11-19 i
- - + -
POWER OISTRIBUTRON LXMITS BASES 4
- When an qF measurement is taken, an allowance for both experimental error
; and manufacturing tolerance. must be made. An allowance of 5% is appropriate for a full core map taken with the incore detector flux mapping system and a 3% allowance is appropriate for manufacturing tolerance.
1 The F limit for RATED THERMAL POWER (FRTP) as provided in the Radial Peaking xy x
-l Factor limit report per Specification 6.9.1.14 was determined from expected power control maneuvers over the full range of burnup conditions in the core.
When RCS flow rate and Fh are measured, no additional allowances are necessary prior to comparison with the limits of Figures 3.2-3 and 3.2-4. Measurement errors of 3.5% for PSS total flow rate and 4% for F have been allowed for in determination of the design DNBR value. l i The 12 hour periodic surveillance of indicated RCS flow is sufficient to 1 l detect only flow degradation which could lead to operation outside the acceptable region of operation shown on Figure 3.2-3. [ 3/4.2.4 OUADRANT POWER, TILT RATIO I Ibe quadrant power tilt ratio limit assures that the radial power distri-bution satisfies' the design values used in the power capability analysis. Radial power distribution measurements are made during startup testing and periodically during power operation. The limit of 1.02 at which corrective action is required provides DNB and linear heat generation rate protection with x y plane power tilts. i l . The two hour time allowance for operation with a tilt condition greater - than 1.02 but less than 1.09 is provided to allow identification and cor-rection of a dropped or misalicned rod. In the event such action does not correct the tilt, the margin for uncertainty on F qis reinstated by reducing the power by 3 percent for each percent of tilt in ocess of 1.0. l 3/4.2.5 DNB PARAMEIERS i . Ihe limits on the DNB related parameters assure that each of the para-meters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have Men analytically demonstrated adequate to maintain a minimum DNBR of 1.30 throughout each analyzed transient. The 12 hour periodic surveillance of t1ese parameters thru instrument readout is sufficient to ensure that the parameters are restored within thei'r limits following load changes and other expected transient operation. L. -SEQUOYAH - UNIT 1- B 3/4 2-5 'I c ._ .-
g . . o . -i il 3/4.4 REACTOR COOLANT SYSTEM i. l BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in operation, and maintain DNBR above 1.30.during all norral operations and anticipated transients. In MODES 1 and 2 with one reactor coolant loop not in operation this specification requires that the plant be in at least HOT STANDBY within 1 hour. In MODE 3, a single reactor coolant loop provides sufficient heat removal c apability for removing decay heat; t.awever, single failure considerations-require that two loops be OPERABLE. In MODE 4, a single reactor coolant loop or residual heat removal (RHR) loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops be OPERABLE. Thus, if t'ie reactor cLalant loops are nt.t OPERABLE, this specification requires two RHR loop? to be OPERABLE. In MODE 5 single failure considerations require that two RHR loops be OPERABLE. y The operation of one Reactor' Coolant Pump or one RPR pump provides adequate flow to ensure v.xing, prevent stratification and produce gradual reactivity changes' dJring boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be witnin the capability of operator recognition and control. 3/4.4.2 and 3/4.4.3 SAFETY AND RELIEF VALVES The pressurizer code safety valves operate to prevent the RCS Each safety valvefrom being is designed pressurized above its Safety Limit of 2735 psig.lbs per hcur of saturated s to relieve 420,000
.The relief capacity of a single safety valve is adequate to relieve any over-pressure condition which could occur during shutdown. In the event that no
\
\
Q SEQUOYAH - UNIT 1 B 3/4 4-1 f_
REACTOR COOLANT SYSTEM i 8 BASES safety valves are OPERABLE, an operating RHR loop, connected to the RCS,*
.provides overpressure relief capability and will prevent RCS overpressurization.
During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2735 psig. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss of load assuming no reactor trip until the first Reactor Protective System trip set point is reached (i.e. , no . credit is taken for a direct reactor trip on the loss of load) and also assuming no operation of the power operated relief valves or steam dump valves. Demonstration of the safety valves' lift settings will occur only during
~~
shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code. The power operated relief valves (PORVs) and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump. Operation of the PORVs minimizes the undesirable opening cf the spring-loaded pressurizer code safety valves. Each PORV has a remotely operated block valve to provide positive shutoff capability should a relief valve becca,e inoperable. 3/4.4.4 PRESSURIZER The limit on the maximum water volume in the pressurizer assures that the parameter is maintained within the normal steady state envelope of operation assumed in the SAR. The limit is consistent with the initial SAR assumptions. The 12 hour pericdic surveillance is sufficient to ensure that the parameter is restored to within its limit fellowing expected transient operation. The maximum water volume also ensures that a steam bubble is formed and thus the RCS is not a hydraulically solid system. The requirement that 150 kw of
- pressurizer heaters and their associated controls be capable of being supplied electrical power from an emergency bus provides assurance that the plant will I be able to control reactor coolant pressure and establish natural circulation conditions. -
l 3/4.4.5 STEAM GENERATORS .. The Surveillance Requirements for inspection of the steam generator tubes ! ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator
- Jing is essential in order to maintain surveillance of the conditions of the tubes in the event that there-is evidence
[ of mechanical damage or progressive degradation due to design, manufacturing
; errors, or inservice conditions that lead to corrosion. Inservice inspection h of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.
i l SEQUOYAH - UNIT 1 B 3/4 4-2
F-PLANT SYSTEM 5
\'
BASES
~~~ ~ ~' '
3/4.7.1.4 ACTIVITY The limitations on secondary system specific activity ensure that the resultant off-site radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture. 'This dose also includes the effects of a coincident 1.0 GPM primary to secondary tube leak in the steam generator of the affected steam line. These values are consistent with the assumptions used in the accident analyses. 3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blowdown in the event of a steam line rupture. .This restriction is required to 1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and 2) limit the pressure _ rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the main steam isolation valves within the closure times of the surveillance requirements are consistent with the assumptions used in the accident analyses. 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION The limitation on steam generator pressure and temperature ensures that the pressure induced stresses in the steam generators do not exceed the maximum allowable fracture toughness stress limita. The limitations of 70 F and
- 200 psig are based on a steam generator RTNDT f 60*F and are sufficient to prevent brittle fracture 1
3/4.7.3 COMPONENT COOLING WATER SYSTEM 4 The OPERABILITY of the component cooling water system ensures that sufficient cooling capacity is available for continued operation of safety related equipment ' during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the accident analyses. .- 3/4.7.4 ESSENTIAL RAW COOLING WATER SYSTEM The OPERABILITY of the essential raw cooling water system ensures that sufficient cooling capacity is available for continued' operation of safety related equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the accident conditions within acceptable limits. SEQUOYAH - UNIT 1 8 3/4 7-3
PLANT SYSTEMS BASES r I
$ 3/4.7.8 AUXILIARY BUILDING GAS TREATMENT SYSTEM The OPERABILITY of the auxiliary building gas treatment system ensures that radioactive materials leaking from the ECCS equipment following a LOCA are filtered prior to reaching the environment. The operation of this system and the resultant effect on offsite dosage calculations was assumed in the accident I analyses. ANSI N510-1975 will be used as a procedural guide for surveillance testing. Cumulative operation of the system with the heaters on for 10 hours over a 31 day period is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters.
3/4.7.9 SNUBBERS Snubbers are designed to prevent unrestrained pipe or component motion under dynamic loads as might occur during an earthquake or severe transient, while allowing normal thermal motion during startup and shutdown. The consequence of an inoperable snubber is an increase in the probability of structural damage to piping or components as a result of a seismic or other event initiating dynamic loads. It.is therefore required that all snubbers required to protect the primary coolant system or any other safety system or. component be operable during reactor operation. n., 1' Because the snubber protection is required only during relatively low. probability events, a period of 72 hours is allowed to replace or restore the inoperable snubber (s) to operable status and perform an engineering evaluation cn the supported component or declare the, supported system inoperable.and follow the appropriate limiting condition for operation statement for that system. The engineering evaluation is performed to determine whether the mode of failure of the snubber has adversely affected any safety related component or system. Safety-related snubbers are visually inspected for overall integrity and operability. The inspection will include verification of proper orientation, 4 adequate fluid nevel if applicable, and attachment of the snubber to its anchorage. The removal of insulation or the verification of torque values for threaded fasteners is not required for visual inspections. The inspection frequency is based upon maintaining a constant level of snubber '* protection. Thus, the required inspection interval varies inversely with the observed snubber failures. The number of inoperable snubbers found during' a required inspection determines the time interval for the next required inspection. Inspections performed before that interval has elapsed may be used as a new reference point to determine the next inspection. However, the results.of such early inspections performed before the original required time-interval has elapsed (nominal time less 25 percent) may not be used to lengthen the required inspection interval. Any inspection whose results require a. o shorter inspection interval will override the previous schedule. SEQUOYAH-UNIT 1! B 3/4 7-5
PLANT SYSTEMS BASES
- 5. .
3/4.7.9 SNUBBFoS (cont'd) When the cause of the rejection of a snubber in a visual inspection is clearly established and remediN for that snubber and for any other snubbers that ray be generically susceptible and operability verified by inservice functional testing, if anplicable, that snubber may be exempted from being counted as inoperable. Generically susceptible snubbers are those which are of a specific make or model and have th 5ame design features directly related to rejection of the snubber, or are similarly located or exposed to the same environmental conditions such as temperature, radiation, and vibration. Inspection groups may be established based on design features and installed conditions which may be expected to be generic. Each of these inspection groups are inspected and tested separately unless an engineering analysis indicates the inspection group is improperly constituted. All suspect snubbers are subject to inspection and testing regardless of inspection groupings. To further increase the assurance of snubber reliability, functional tests shall be performed during each refueling outage. These tests will include stroking of the snubbers to verify proper movement, activation, and bleed or release. The performance of hydraulic snubbers generally depends on a clean, deaerated fluid contained within variable pressure chambers, flowing at closely controlled rates. Since these characteristics are subject to changa with exposure to the reactor environment, time, and other factors, their performance within the specified range should be verified. Mechanical snubbers which depend upon overcoming the inertia of a mass and the braking action of a capstan spring contained within the snubber for limiting the acceleration of the attached component (within the load rating of the snubber) are not subject to changes in performance in the same manner as hydraulic snubbers. Pending the development of information regarding the change during the service of the snubber of the acceleration / resistance relationship and the optimum method for detecting this change, these mechanical snubbers may be tested to verify that when subjected to a large change in velocity the resistance to movement increases greatly. The performance change information is to be developed in order outage. to establish test methods to be used during and after the first refueling Ten percent of the total population of approximately 700 snubbers is an adequate sample for functional tests. The initial sample is to be proportioned among the groups in order to obtain a representative sample. Observed failures of more than two snubbers in the initial lot will require an engineering analysis and testing of additional snubbers selected from snubbers.likely to have the same defect. A thorough inspection of the snubber threaded attachments to the pipe or components.and the anchorage will be made in conjunction wit:h all required functional tests. , SEQUOYAH - UNIT 1 B 3/4 7-6
a PLANT SYSTEMS 8'ASES {., b. 3/4.7.10 SEALED SOURCE CONTAMINATION The limitations on removable contamination for sources requiring-leak testing, including alpha emitters, based on 10 CFR 70.39(c) limits for plutonium. This limitation will ensure that leakage from byproduct, source, and special nuclear material sources will not exceed allowable intake values. Sealed sources are classified into three groups according to their use, with-surveillance requirements commensurate with the probability of damage to a
-source in that group.
Those sources which are-frequently handled are required to be tested more often that, those which are not. Sealed sources which are-continuously enclosed within a shielded mechanism (i.e., sealed sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism. t
)
, i
. -l l
SEQUOYAH - UNIT-1 , B 3/4 7-6a i
-r
1 [ 3/4.8 ELECTRICAL POWER SYSTEMS i
- BASES 3/4.8.1 AND 3/4.8.2 A.C. SOURCES AND ONSITE POWER DISTRIBUTION SYSTEMS The OPERABILITY of the A.C. and 0.C power sources and associated distribution systems during operation ensures that sufficient power will be available to supply the safety related equipment required for 1) the safe shutdown of the facility and 2) the mitigation and control of accident condi-tions within the facility. The minimum specified independent and redundant A.C. and 0.C. power sources and distribution systems satisfy the requirements of General Design Criterion 17 of Appendix "A" to 10 CFR 50.
The ACTION requirements specified for the levels of degradation of the power sources provide restriction upon continued facility operation commensurate ' with the level of degradation. The OPERABILITY of the power sources are consistent with the initial condition assumptions of the safety analyses and are based upon maintaining at least one redundant set of onsite A.C. and D.C. power sources and associated distribution systems OPERABLE during accident conditions coincident with an assumed loss of offsite power and single failure of the other onsite A.C. source. The OPERABILITY of the minimum specified A.C. and 0.C. power sources and C associated distribution systems during shutdown and refueling ensures that
- 1) the facility can be maintained in the shutdown or refueling condition for extended time periods and 2) sufficient instrumentation and control capability is available for monitoring and maintaining the unit status.
The Surveillance Requirements for demonstrating the OPERABILITY of the diesel generators are in accordance with the recommendations of Regulatory Guides 1.9 " Selection of Diesel Generator Set Capacity for Standby Power Supplies", March 10,1971,1.108 " Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants," Revision 1, August 1977, and 1.137 " Fuel-Oil Systems.for Standby Diesel Generators," Revisien 1, October 1979. The Surveillance Requirement for demonstrating the OPERABILITY of the Station batteries are are based on the recommendations of Regulatory Guide 1.129
" Maintenance Testing and Replacement of Large Lead450-1980, Storage "IEEE Batteries for Recommended Nuclear Power Plants," February 1978, and IEEE Std Practice for Maintenance, Testing and Replacement of Large Lead Storage Batteries for Generating Stations and Substations."
Verifying average electrolyte temperature above the minimum for whicn the battery was sized, total battery terminal voltage onfloat charge, connection resistance values and the performance of battery service and discharge tests ensures the effectiveness of the charging system, the ability to handle high discharge rates and compares the battery capacity at that time with the rated capacity. SEQUOYAH - UNIT 1 B 3/4 8-1 I h
, - ~
g
?
BASES i.
. A.C. SOURCES AND ONSITE POWER DISTRIBUTION SYSTEMS (Continued)
Table 4.8-2 specifies the normal limits for each designated pilot cell and each connected cell for electrolyte level,- float voltage and specific gravity. The limits for the designated pilot cells float voltage and specific gravity, greater than 2.13 volts and .015 below the manufacturer's full charge specific gravity or a battery charger current that had stabilized at a . low value, is characteristic of a charged cell with adequate capacity. The normal i limits for each connected cell for float voltage and specific gravity, greater-than 2.13 volts and not more tha .020 below the manufacturer's full charge specific gravity with an average specific not more than .010 below the manufacturer' gravity s full of allspecific' charge the connected gravity,cells ensures the OPERABILITY and capability of the battery. Operation with a battery cell's parameter outside the normal limit but within the allowable value specified in Table 4.8-2 is permitted for up to 7 days. During this 7 day period: (1) the allowable values for electrolyte level ensures no physical damage to the plates with an adequate electron transfer capability; (2) the allowable value for the average specific gravity' of all the cells, not more than .020 below the manufacturer's recommended full charge specific gravity, ersures that the decrease in rating will be less than the safety margin provided in sizing; (3) the allowable value for an individual- - cell's specific gravity, ensures that an individual cell's specific gravity . will not be more than .040 below the manufacturer's full charge specific gravity and that the overall capability of the battery will be maintained within an acceptable limit; and (4) the allowable value for an individual cell's float voltage, greater than 2.07 volts, ensures the battery's capability to perfccm its design function. 3/4.8.3 ELECTRICAL EOUIPMENT PROTECTIVE DEVICES Containment electrical penetrations and penetration conductors are protected l by either deenergizing circuits not required during reactor operation or by demonstrating the OPERABILITY of primary and backup overcurrent protection circuit breakers during periodic survaillance. The surveillance requirements applicable to lower voltage circuit breakers and fuses provides assurance of breaker and fuse reliability by testing at-least one representative sample of each manufacturers brand of circuit breaker and/or fuse. Each manufacturer's molded case and metal case circuit breakers , and/or fuses are grouped into representative samples which are then tested on 1 a rotating basis to ensure that all breakers and/or fuses are tested. If a wide variety exists within any manufacturer's brand of circuit breakers and/or ; fuses, it is necessary to divide that manufacturer's breakers and/or fuses into groups and treat each group as a separate type of breaker or fuses for surveillance purposes. ' ( SEQUOYAH - UNIT 1 i B 3/4 8-2 1-5
+ '
+ w BASES {Ajp I~ El.ECTRICAL EQUIPMENT PROTECTIVE DEVICES (Continued) . The OPERABILITY of f.he motor operated-valves thermal overload protection and/or bypass devices ensures that these devices will not prevent safety related valves from perftrming their function. The Surveillance Requirements for demendrating the OPERABILITY of these devices are in accordance with Regula*.ory Guide 1.106 "Teermal Overload Protection for Electric Motors on
' Motor Operated Valves", Revision 1, March 1977.
Circuit breakers actuated by fault currents are used as isolation devices in this plant, The OPERABILITY of these circuit breakers ensures that the IE ' busses will be protected in the event of faults in non qualified loads powered , by the busses. J O O
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SEQUOYAH - UNIT 1 B 3/4 8-3
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RADIOACTIVE EFFLUENTS BASES 3/4.11.2.6 GAS DECAY TANKS Restricting the quantity of radioactivity contained in each gas decay tank provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting total body exposure to an individual at the nearest exclusion area boundary will not exceed 0.5 rem. This is consistent with Standard Review Plan 15.7.1, " Waste Gas System Failure". 3/4.11.3 SOL _ID RADIOACTIVE WASTE The OPERABILITY of the solid radwaste system ensures that the system will be available for use whenever solid radwastes require processing and packaging p.rior to being shipped offsite. This specification implements the requirements of 10 CFR P'et 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50. 'he. process parameters included in establishing the PROCESS CONTROL PROGRAM may include, but are not limited to waste type, waste pH, waste / liquid / solidificaticn agent / catalyst ratios, waste oil content, waste principal chemical constituents, mixing and curing times. 3/4.11.4 TOTAL DOSE This specification is provided to meet the dose limitations of 40 CFR 190. The specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix I. For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a member of the public will exceed the dose limits of 40 CFR 190 if the individual reactors remain within the reporting requirement level. The Special Report will describe a course of action which should result in the limitation of dose to a member of the public il for 12 consecutive months to within the 40 CFR 190 limits. For the purposes of tne Special Report, it may be assumed that the dose commitment to the member of the public from other uranium fuel cycle sources is negligible, with the exception that dose c3rtributions from other nuclear fuel cycle facilities at ,- the same site or within a radius of 5 miles must be considered. , j SEQUOYAH - UNIT 1 8 3/4 11-5 1
. . o .
DESIGN FEATURES
- 1
- 5.6 FUEL STORAGE CRITICALITY - SPENT FUEL 5.6.1.1 The spent fuel storage racks are designed and shall be maintained with:
- a. Ak eff equivalent to less than 0.95 when flooded with unborated water, which includes a conservative allowance of 1.78% delta k/k for uncertainties as described in Section 4.3 of the FSAR.
- b. A nominal 10.375. inch center-to-center distance between fuel assemblies placed in the storage racks.
CRITICAUTY - NEW FUEL
- 5. 6.1. 2 The new fuel pit storage racks are designed and shall be maintained l with a nominal 21.0 inch center-to-center distance between new fuel- assemblies such that k,77 will not exceed 0.98 when fuel having a maximum enrichment of 3.5 weight percent U-235 is in place and aqueous foam moderation is assumed.
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DRAINAGE 5.6.2 The spent fuel storage pool-is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 722 ft.
. CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1386 fuel assemblies.
5.7 COMPONENTCYCLICORTRANSIENTLIME 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits.of Table 5'.7-1. SEQUOYAH - UNIT 1 i
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e POSITION NUMBER OF INDIVI'007,LS REQUIs J T) FILL POSITION
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MODES 1, 2, 3 & 4 MODES 5 & 6 -
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c' WITil UNIT.1 IN MODES 1, 2, 3 OR 4 POSITION NUMBER F.INDIVIOUALS REQUIRED TO FILL POSITION C.-
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A0 2 a None a - STA l i a/ Individual may fill the same position on' Unit 1 h/ One of the two required individuals may hill the same position o.gunit 1.
~st SEQUOYAH - UNIT 1 6-6 l
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TA5LENOTATION eg' y L yF
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_s S! Shif t Supervisor with a Se;ii'or Reactor Operators License on Unit 2 W x30% MO t-- Individual pith a Senior Reactor Operators License on Unit 2 Individual with a Reactor Operators License on Unit 2 N A0 - Awiliar; Operator i e - N STA - Shift Tee.hnical Advisor Wl? ,-
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y - g , A v29xcept-for n .tnan the wifjeum the Shift Supervisor,ofTable 6.2-1 for"a ceriod of time not tothe Shift Crew Comp requirements N;
,- +. i exceed;2thours in order to accommodate unexpected absence of on-duty shift crew eN#becers provided immed? ate action is taken to restore the Shif t Crew Composition
" N iS to within the minitii6m requirements = of Table 6.2-1. This provision does not W QV peQ2Jt any shift crew position to be unmanned upon shift change due to an g
n. g o':.comingjshif t crewman being late or absent.
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~O 'c Occing dny absence of the Shift Supervisor from the Control Room while the 3 4,7'- gun.}t is in MODES 1, 2,. 3 or 4, an individual (otner thart the Shift Technical A A q 'Adv$sor) with a valid SR0 license shall be designated to assume the Control V ~ RucKerydnd ft.nction. During an absence of the Shift Supervisor from the s Contro! Room while the unit'is in MODE 5 or 6, an individual with a valid SRO 3' ' ,
^ "'or RViidsnse (other'thhn the shif t Technical Advisor) shall be designated to .
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W W * .t-17 ' . 1 ADMINISTRATIVE CONTROLS
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"(c/ 6.2.3 INDEPENDENT SAFETY' ENGINEERING GROUP (ISEG)
FUNCTION 6.2.3.1 The ISEG shall function to examine plant operating characteristics,
- NRC issuances, industry advisories, Licensing Event Reports and other sources
'which may indicate areas for improving plant safety.
COMPOSITION
~
The ISEG'shall be composed of at least five dedicated full-time
~
'6.2.3.2
' engineers located onsite. RESPONSIBILITIES 6.2.3.3 Tne ISEG shall be responsible for maintaining surveillance of plant activities to provide ' independent verification
- that these activities are performed correctly and that human errors are reduced as much as practical.
AUTHORITY 6.2.3.4 The ISEG shall make detailed recommendations for revised procedures, equipment modifications, or other means of improving plant safety to the Assistant Director for Maintenance and Engieering of the Division of Nuclear Power. 6.2.4 SHIFT TECHNICAL ADVISOR (STA) 6.2.4.1 The STA shall serve in an advisory capacity to the shift supervisor on matters pertaining'to the engineering aspects of assuring safe operation of the ' unit. 6.3 L' NIT STAFF QUALIFICATIONS 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions and the supplemental requirements specified in Section'A and C of Enclosure 1 of March 28, 1980 NRC nj ' ' letter to all licensees, except for the Health Physicist who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975. i "Not responsible for sign-off function. [ 6. t e SEQUOYAH - UNIT 1 6- 7 i !
ADMINISTRATIVE CONTROLS
~
i 1 (iv) Procedures for the recording and management of data, (v) Procedures defining corrective actions for off-control point chemistry conditions, (vi) Procedures identifying (a) the authority responsible for the interpretation of the data; and (b) the sequence and timing of administrative events required to initiate corrective action, and (vii) Monitoring of the condensate at the discharge of the condensate pumps for evidence of condenser in-leakage. Whe- condenser in-leakage is confirmed, the leak shall be repaired, plugged, or isolated
. -- d. Backuo Method for Determining Subcooling Margin A program which will ensure the capability to accurately monitor the Reactor Coolant System Subcooling Margin. This program shall include the following:
(i) Training of personnel, and (ii) Procedures for monitoring.
- e. Postaccident Samoling A program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions. The program shall include the following:
(i) Training of personnel, (ii) Procedures for sampling and analysis, (iii) Provisions for maintenance of sampling and analysis equipment.
~
h I I SEQUOYAH - UNIT 1 6-15b
ADMINISTRATIVE CONTROLS
- e. An unplanned offsite release of 1) more than 1 curie of radioactive r material in liquid effluents, 2) more than 150 curies of noble gas in gaseous effluents, or 3) more than 0.05 curies of radiciodine in gaseous effluents. The report of an unplanned offsite release of radioactive material shall include the following information:
- 1. A description of the event and equipment involved.
- 2. Cause(s) for the unplanned release.
- 3. Actions taken to prevent ~ recurrence.
- 4. Consequences of the unplanned release.
- f. Measured levels of radioactivity in an environmental sampling medium j
determined to exceed the recorting level values of Table 3.12-2 when averaged over any calendar quarter sampling period. RADIAL PEAKING FACTOR LIMIT REPORT 6.9.1.14 The F limit for Rated Thermal Power (FRTP) x shall be provided to the Director of the Regional Office of Inspection and Enforcement, with a copy to the Director, Nuclear Reactor Regulation, Attention, Chief of the Core Performance Branch, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555 for all core planes containing bank "0" control rods and all unrodded core 3 planes at least 60 days prior to cycle initial criticality. In the event
;) that the limit would be submitted at some other time during core life, it will be submitted 60 days prior to the date the limit would become effective unless otherwise exempted by the Ccsnission.
RTP Any information needed to support F will be by request from the NRC and neednotbeincludedinthisreport? SPECIAL REPORTS i - 6.9.2 Special reports shall be submitted to the Director of the Office of Inspection and Enforcement Regional Office within the time period specified for each report. 6.10 RECORD RETENTION l in addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated. 6.10.1 The folicwing records shall be retained for at least five years:
- a. Records and logs of unit coeration covering time interval at each power level.
I '
- b. Records and logs of principal maintenance activities, inspections, j repair and replacement of principal iteme of equipment related to nuclear safety.
SEQUOYAH - UNIT 1 6-22
l I . . ADMINISTRATIVE' CONTROLS ' 6.10.1 (continued)
^
- c. 'All REPORTABLE OCCURRENCES-submitted to the Commission. :
..d. Records of surveillance. activities,' inspections-and calibrations-require'd by these Technical Specifications,
- e. -Records of changes made to the procedures required by Specification 6.8.1 and 6.8.4.
f. Records offradioactive shipme6ts.
- g. Records of sealed source and fission detector leak cests and results.
- h. ' Records of annual physical inventory of all sealed source material of record, 6-2 2a
'SEQUOYAH - UNIT 1-
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; ADMINISTRATIVE CONTR01 ,
). 6.10.2 ~ TheLfollowing records shall be retained for,the duration of the Unit e
Operating License: - a, ' Records and drawing changes reflecting" unit design modifications . made to systems'and equipment described in the Final Safety Analysis-
~
*- Report.
- b. Records of new and irradiated fuel inventory, fuel -transfers and assembly barnup histories.
- c. Records of radiation exposure for all. individuals entering radiation control areas. '
- d. Records of. gaseous and liquid radioactive material released to the environs,
- e. Records of transient or~ operational cycles,for those unit' components" ~
- identified in Table 5.7-1.
- f. Records o,f reacter tests and experiments.
- g. Records of training and qualification ~for current members of the
- unit staff.
, h'. Records of in-service inspections per' formed pursuant to these-
, Technical Specifications. -
- i. Records of ' Quality Assurance activities, required by Lthe 0perationai Quality 1 Assurance Manual. '
- j. Records of.. reviews performed for changes made to procedures or 3 equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.
f .
- k. Records of meetings of the PORC,.RARC, and the NSRB.
- - . 1. Records of' analyse's required byfthe radiological environmental
- s. ; monitoring program.
Records of secondary water sampling and water quality. m. l n. Records.of the service live monitoring of-all hydraulic and mechanical l snubbers listed on Tables 3.7-4a'and 3.7-4b including the maintenance performed to renew the service life.
- o. Records for Environmental Qualification which are covered under .
, the provisions of Paragraph 2.C.(12)(b) of License No. DPR-77.
- 6.11 RADIATION PROTECTION PROGRAM
~
L ' Procedures 'for. personnel radiation protection shall be prepared . consistent p with .the requirements of 10 CFR Part 20 and shall be approved,, maintained"and
. adhered to fer all operations involving personnel radiation exposure.
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- SEQUOYAH - UNIT 1 6-23 f1"f,.:
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ENCLOSURE 2
-JUSTIFICATION This change in the unitil technical specifications (T/S) involves updati5g -
the unit 1 T/S to be consistent with the unit 2 T/S. 'All' proposed changes-have been'previously reviewed and approved by-.the NRC for unit 2...These changes clarify the T/S and are-being submitted at: the request of the NRC. t 1 t
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