ML20091C897
| ML20091C897 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 03/30/1992 |
| From: | Hebdon F Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20091C902 | List: |
| References | |
| GL-88-16, NUDOCS 9204070195 | |
| Download: ML20091C897 (34) | |
Text
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e UNITED STA*ES 3
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.j NUCLEAR REGULATORY COMMISSION WASHINGTON D.c. 20%6 o,
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%, *... + y TENNESSEE VALLEY,AdHORITY DOCKET NO. 50-328 SE000YAli NUCLEAR PLAN 1, UNIT 2 At1EhDMENUCTACIllTYOPERAT]hGL! CENSE Amendment No.146 License No. DPR-79 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated May 24, 1991 and amplified August 23, 1991, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
lhe facility will operate in conformity with the application, the provisiens of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) that the activities authorized by this anendent can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the cocinon defense and security or to the health and sr.fety of the public; and E.
The issuance of this acendment is in accordance with 10 CFR Part 51 of the Cgmission's regulations and all applicable requirements have been satisfied.
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P l
2-2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-79 is hereby amended to read as follows:
(2) J.gchnical Specificati.qnji The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 146, are hereby incorporated in the
'icense.
The licensee shall operate the facility in accordance with Technical Specifications.
3.
T' ise amendment is effective as of its date of issuance; to be ad in conjunction with the Core Operating Limits Report within FOR THE NUCLEAR REGULATORY COPNISSION
,F b y n -.
Frederick J. Hebc%n, Director Project Directorate 11-4 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications e
Date of Issuance:
March 30, 1992 J
l--_----__-_____-_:_-_______--__--_--_---____-----_-_-_---_--__-_--_------------------------------------------A
MTACHMENTTOLICENSEAMENDMENTNO.146 FACILITY OPERATING tlCENSE NO. DPR-79 DOCKET NO. 50-328 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.
Overleaf pages* are provided to treintain docurent completeness.
REMOVE jfSERT i
1 11 li 1-2 1-2 1-3 1-3 1-4 1-4 1-5 1-5 1-6 1-6 1-7 1-7 1-8 1-8 82-1 B2-1 B2-2 B2-2 3/4 1-4 3/4 1-4 3/4 1-5 3/4 1-5 3/4 1-14 3/4 1-14 3/4 1-20 3/4 1-20 3/4 1-21 3/4 1-21 3/4 1-22 3/4 1-22 3/4 1-23 3/4 1-23 3/4 2-1 3/4 2-1 3/4 2-3 3/4 2-3 3/4 2-4 3/4 2-4 3/4 2-5 3/4 2-5 3/4 2-6 3/4 2-6 3/4 2-7 3/4 2-7 3/4 2-8 3/4 2-8 B3/4 1-2 83/4 1-2 B3/4 2-1 83/4 2-1 B3/4 2-2 B3/4 2-1 B3/4 2-4 B3/4 2-4 6-22 6-22 6-22a v
l INDEX DEFIN!110NS SECTION PAGE 1.0 DEFINITIONS 1.1 ACTICN............................................~...............
1-1 1.2 AX1AL FLUX DIFFERENCE............................................
1-1 1.3 RYPASS LEAKAGE PATH...............................................
1-1 1.4 CHANNEL CAllBRAT10N...............................................
1-1 1.5 CHANNEL CHECK............................
1-1 1.6 CHANNEL FUNCTIONAL TEST.........................................,.
1-2 1.7 CONTAINMENT INTEGRITY.............................................
1-2 s
1.8 CONTROLLED LEAKAGE................>...............................
1-2 1.9 CORE ALTERATION.............,.......................
1-2 1.10 CORE OPERATING LIMIT REP 0RT.......................................
1-2 1.11 DOSE EQUIVALENT 1-131............................................
1-3
' 12 E-AVERAGE DISINTEGRATION ENERGY...................................
1-3 1.13 ENGINEERED SAFETY FEATURE RESPONSE TIME...........................
1-3
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1.14 FREQUENCY N0TATION................................................
1-3 1.15 GASECUS RADWASTE TREATNENT SYSTEM.................................
1-3 1.16 IDENTIFIED LEAKAGE................................................
14 1.17 MEMBERS OF THE PUBLIC.............................................
1-4 a
1.16 0FFSITE DOSE CALCyLATION MANUAL...................'................
1-4 1.19 OPERABLE - OPERABILITY............................................
1-4 1.20 OPERATIONAL MODE - M0DE...........................................
1-5 1.21 PHYSICS TESTS....................................................
1-5 i
1.22 PRESSURE BOUNDARY LEAKAGE.........................................
1-5 1.23 PROCESS CONTROL PR0 GRAM...........................................
1-5 SEQUGYAH - UNIT 2 I
Amendment No.146 i
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i INDEX DEFINITIONS SECTION A_GE D
1.0 CEFINITIONS-(Continued) 1.24-PURGE-PURGING..............................
4.....................
15 1.25 QUADRANT POWER TILT RATI0.........................................
1-5
- 1. 2 6 R A T E D T H E RMAL P 0W E R..............................................
1-6 1.27 REACTOR 1 RIP SYSTEM RESPONSE TIME......,..........................
1-6 1.28 REPORTABLE EVENT..................................................
1-6 1.29 SHIELD BUILDING INTEGRITY,........................................
1-6 L
1.30 SHUT 00WN PARGIN...................................................
1-6 1.31 SITE 80VNDARY.....................................................
1-6 1.32 SOLIDIFICATION..............................................,.....
1-7 1.33 SOURCE CHECK......................................................
1-7 1.34 STAGGERED TEST BA5!S..............................................
12/
1.35 THERMAL P0WER................................................
1-7 1.36 UNICENTIFIED LEAKAGE..............................................
1-7 l
l.
1.37 UNRESTRICTED AREA...............................................
1.J i
s l
1.38-VENTILATION.EXHi.UST TREATMENT SYSTEM..............................
1-8
- r.
1.39 VENTING...........................................................-
1-8
.0PERATIONAL N0 DES (TABLE 1.1)..........................................
1-9 FREQUENCY NOTATION (TABLE 1.2).........................................
1-10 L
l' I
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l l-l-
.SEQUOYAH - UNIT 2 II Amendment No. 146 J
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L DEFINITIONS CHANNEL FUNCTIONAL TEST 1.6 A CHANNEL FUNCTIONAL TEST shall be.
a.
Analog char.nels - the injection of a cimulated signal into the channel as close to the sensor as practicable to verify 0PERABILITY F
including alarm and/or trip functions.
b.
Sistable charnels - the injection of a simulateo signal into the sensor to verify OPERABILITY including ala;m and/or trip functicns, c.
Digital channels - the injection of a simulated signal into the
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channel as close to the sensor input to the process racks as practi-cable to verify GPEPABILITY including alarm and/or trip functicas.
CONTAINMENT INTEGRITY
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1.7 CONTAINMENT INTEGRITY shall exist wnen:
o I
a.
All pr.netrations required to be closeo during accident conditions are either:
1)
Capable of being closed by an CPERABLE containment autenatic isolation valve systen, or i
2)
Closed by manual valves, blind flanges, cr deactivated auto-matic valves secured in their clcsed positioid, except as provided in Table 3.6-2 of specification 3.6.3.
b.
All equipment hatches ace closed and sealed, f
Each air lock is in compliance with the requirements of Specification 3.6.1.3, E~
d.
The containment leakage rates are within the limits of Specification g
3.5.1.2, and e.
The sealing mechanism associated with each penetration (e.g., welds, bellows or 0-rings) is OPERABLE.
CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactar
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coolant pump seals.
CORE ALTERATION
- 1. 9 CORE ALTERATION shall be the mcvement or manipulation of any comocnent within the reactor pressure vessel with the vessel head removed and fuel in the vessel.
Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.
CORE OPERATING LIMITS REPORT 1.10 The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document T
that providas core operating limits for the current operating yeload cycle.
EE-These cycle-specific corc operating iimits shall be deterniined for each reload cycle in accordance with Specification 6.9.1.14.
Unit operation within these
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operating limits is addressed in individual specifications.
SEQUOYAH - UNIT 2 1-2 Amendment No. 63, 117, 132, 146 E
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DEFINITIONS DOSE EQUIVALENT I-131 1.11 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie /
l l
gram) which alone would produce the same inyroid dose as the quantity and iso-topic mixture of I-131,1-132, 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."
3 E - AVERAGE DISINTEGRATION ENERGY 1.12 E shall be the average (weighted in proportion to the concentration of I
each radionuclide in the reactor coolant at the time of sampling) cf the sum of the average beta and gamma energies p?r disintegration (ia MeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at
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least 95% of the total non-icdine activity in the coolant.
ENGINEERED SAFETY FEATURE RESPONSE TIME f
1.13 The ENGINEERED SAFETY FEATURE RESPONSE TIMF. shall be that time interval I
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, pua.p discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays'where applicable.
FREQUENCY NOTATION 1.14 The FREQUENCY NOTATION specified for the performance of Surveillance l
Requirements shall correspond to the intervals defined in Table 1.2.
i GASE005 RADWASTE TREATMENT SYSTEM 1.15 A Gi. W '15 RADWASTE TREATMENT SYSTEM is any system designed and ir. stalled l
to reduce 'idioactive gaseous effluents by collecting primary coolarit system offgases f rorr. the primary systen and providing fer delay or holdup for th?
purpose of reducing the total radioactivity prior to release to the environment.
SEQUOYAH - UN!T 2 1-3 Amendment No. 63, 146
DEFIN'TIONS IDENTIFIE_0_ LEAKAGE L'
- 1. 16 IDENTIFIED LEAKAGE shall be:
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a.
Leakage (except CONTROLLED LEAKAGE) into closed systems, eut.o as pump seal or valve packing leaks that are captured and conducted to
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a sump or collecting tank, or
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I.e3kage into the containment atmosphere from sources that are both b.
specifically located and known either not to interfere with the operation of leakoge detection systems or not to be PRESSURE BOUNDARY
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LEAMAGE, or c.
Reactor coolant system leakage through a steam generator to the secondary system.
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MEMBERS OF THE PUBLIC 9
1.17 MEMBERS OF THE PUBLIC shall include all individuals who are not occupa-l tionally associated with the plant.
This category shall include non employees of the licensee who are permitted to use portions of the site for recreational,
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occupational, or other purposes not associated with plant ' unctions.
This category does not include non-amployees such a vending mochine servicemen or postmen who, as part of their formal job function, occas nally enter an area
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that is controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials.
0FFSITE DOSE CALCULATION MANUAL E-1.18 The OFFSITE DOSE CALCULATION MANUAL (0DCM) shall contain the methodology l
t.
and parameters used in the calculation of offsite doses resulting from radio-m active gaseous and liquid effluents, in the calculatien of gaseous and liquid effluent monitoring alarm / trip setpoints and in the conduct of the Radiological Environmental Monitoring Program.
The 00CM shall 61so contain (1) the Radio-e active Ef fluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.5 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Semiannual Radioactive Effluent Release Reports required by Specifications 6.9.1.6 and 6.9.1.8.
OPERABLE - OPERABILITY I
1.19 A system, subsystem, train, or component or device shall be OPERABLE or l
have OPERABILITY when it is capable of performing its specified function (s),
and when all necessary attendant instrumentatica, controls, a normal and an emergency electrical power source, scoling er seal water, lubrication or other auxiliary equipment that are requir 3d for the system, subsystem, train, com-ponent or device to perform its function (s) are alsu capable of performing thair related support function (s).
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SEQUOYAH - UNIT 2 1-4 Amendment No. 63, 134, 146 l
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OEFINITIONS 4.
UPERATIONAL t+00E - MODE 1.20 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive l
combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.3.
P_HYSICS TESis 1.21 hiYSICS TESTS shall be those tests performed to measure the fundamental l
nuclear characterO tics of the reactor core and related instrumentation and 1) described in Chapter 14.0 of the FSAR, 2) authorized under the previsions cf 10 CFR 50.59, or 3) otherwise approved by the Commissicn.
PRESSURE B0UNDARY LEAKAGF 1.22 PRES $URE BOUNDARY LEAKAGE shall be leakage (except steam generator tubo l
leakage) through a non-isolable fadlt in a Reactor Coolant System component body, pipe wall or vessel wall.
.PPOCESS CONTROL PROGRAM (PCP) 1.23 The PROCESS CONTROL PROGRAM shall contain the current formulas, sampling, j
analyses, tests, and determinations to be made to ensure that the processing and packaging of solid 'adioactive wastes basca on demonstrated processing of actual or simulated wet solid wastes wili be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71; State regulations; and other requirements governing the disposal of solid radioactive wastos.
P_ URGE - PURGING' l
1.24 PURGE or PURGING is the controlled procass of discharging
- air or gas
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i from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in~such a tunner that replacement air or gas is raquired t2 purity the confinement.
QUADRANT POWER TILT RATIO 1.25 QUADRANT POWER fit.T RATIO shall be the ratic of the maximum upper excore l
l detector calibrated output to the average of the upper u core detector cali-L' brated outputs, or the ratio of the maximum lower excore detector calibrated i
output to the average of the lower excore detector ralibrated outputs, which-i ever is greater. With one excore detector inoperable, the remaining three j.
detectors shall be used for computing the average.
L i
SEQUOYAH - UNIT 2 1-5 Amendment. No. 63, 134, 146 I
DEFINITIONS RATED THERMAL POWER (RTP) 1.26 RATED THERMAL POWER (RTP) shall be a total reactor core heat transfer i
rate to the reactor coolant of 3411 MWt.
REACTOR TRIP SY5 TEM _ RESPONSE TIME 1.27 The REACTOR TRIP SYSTEM RESPONSE TIME shell be the time interval from l
when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coil voltage.
REPORTABLE EVENT 1.28 A REPORTABLE EVENT shall be any of those conditions specified in Section l
50.73 to 10 CFR Dart 50.
SHIELD BUILDING INTEGRITY 1.29 SHIELD BUILDING INTEGRITY shall exist when:
l a.
The door in each access opening is closed except when the access opening is being used for normal transit entry and exit, b.
The emergency gas treatmerit system is GPERABLE.
c.
The scaling mechanism associated with each penetration (e.g., welds, bellows or 0-rings) is OPERABLE.
SJJUT00WNMARGIN
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1.30 SHUTDOWN MARGIN shall be the instantaneous amount of *eactivity Dy which I
the reactor is subtritical or would be subcritical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.
SITE B0bHDARY 1.31 The SITE BOUNDARY chall be that line beyond which the land is not owned, l
leased, or otherwise controlled by the licensee (see figure 5.1-3).
SEQUOYAH - UNIT 2 1-6 Amendment No. 63, 132, 146
l' DEFINITIONS
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SOLIDIFICATION l.32 Deleted.
I SOURCE CHECK 1.33 Deleted.
l STAGGEJEDTESTBASIS
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1.34 A STAGGERCD TEST BASIS shall consist of:
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a.
A test schedule for n systems, subsystems, trains. or other designated I
components obtained by dividing the specified test interval into n equal subintervals,
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b, The testing of one system, subsystem, train of other designated component at the beginning of each subinterval.
THERMAL POWER 1.35 THERMAL POWER shall be the total reactor core heat transfer rate to the
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1.36 UNIDENTIFIED LEAKAGE shall be :ll leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.
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UNRESTRICTED AREA 1.37 An UNRESTRICTED AREA shall be any area, at or beyond the site boundary to l
which access is not controlled by the licensee for purposes of protection of F
individuals from exposure to radiation and radioactive materials or any 3rea within the site boundary used for residential quarters or industrial, conwer-cial, institutional, and/or recreational purposes.
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SEQUOYAH - UNIT 2 1-7 Amendment No. 63, 134, 146 i
DEFINITIONS
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VENTILATION EXHAUST TREAT!iENT SYSTEM i
1.38 A VENTILATION EXHAUST TREATMENT 515 TEM is any system designed and I
installed to reduce gaseaus radic1cdine or radioactive material in particulate form in effluents by past.ing ventilation or vent e>.haust gases through charcoal 41sorbers anc/or hEPA filters for the purpose of renCving iodines or partic-ulates from ths: gaseous exhaust streem prior to the release to the enyh onrrent (such a system is not consicered to have any effect on noble gas effluerts).
Engineered Safety Feature (ESF) atmospheric cleanup systenis are not considered to be VEhllLATION EXHAUST TREATMENT SYSTEM components.
VENTING 1.39 YENTING is the controlled process of discharging air or gas from a j
confinement to maintain terrperature, pressure, humidity, concentratior, or other operating condition. in such a manner that replacement air or ges is not l
provided or required during VENTING.
Vent, used in system names, does not irrply a VENTING prceess, m
SEQUOYAH - UNIT 2 1-8 Amendment No. 63, 146 i
2.1 SAFE 1f LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel and possible cladd1ng perforation wnich would result in the release of fission products to the reactor coolant.
0\\erheating of the ftel claading is prevented by restricting fuel operation to within tu, nucleate boiline, regime where the heat transfer coefficient is large and the cladding surface 6emperature is slightly above the conlant saturation temperature.
Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nuclecte boiling (DNB) and the resultant sharp reduction in heat transfer cocfficient.
DNB is not a directly measurable parameter during operation and therefore THERMAL PCWER and Reactor Coolant Temperature and Pressure have been related to ONS through the WRB-1 correlction and the W-3 correlation for conditions outside the range of WRB-1 corcelation.
The DNB correlations have been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions.
The local DNB heat-flux ratio, ONBR, defined as the ratio of the heat flux that would cause DNB at. a particular core location to the local heat flux, is indicative of the margin to DNR.
The DNB design basis is as follows:
there must be at least a 95 percent probability that the minimum DNBR of the limiting rod during Condition I and Il events-is greater than or equal to the ONBR limit of the CNB correlation being used (the WRB-1 or W-3 correlation in this application).
The correlation
'0NBR limit is established based on the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence that ONB will not occur when the minimum DNBR is at the DNBR limit.
-The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure ar.a average temperature for which the minimum DNBR is no less than the safety analysis ONBR lim!t, or the average entnalpy at the vessel exit is equal to the enthalpy of saturated liquid.
Thecurvesarebasedonanenthalpyhotchannelfactor,FfH,specifiedin the Core Operating Limit Report (COLR) and a reference cosine with a peak of 1.55 for axial power shape. - An allowance is included for an increase in Fhatreducedpowerbasedontheexpression:
F g,p gy 'pp6K (7.p))
N R
where P =
THERMAL POWER RATED THERMAL POWER N
F
= the F litnit at RATED THERMAL POWER (RTP) specified in the g
COLR, and PF6g = the power factor multiplier for F specified in the COLR.
SEQUOYAH - UNIf 2 B 2-1 Amendment No. 21, 104, 130, 146
b SAFETY LIMITS r
BASES 2.1.1 REACTOR CORE (Continued)
These limiting heat flux conditions are higher tIan those calculated for the range of all control rods fully withdrawn to the aaximun aliowable control rod insertion assuming the axial power imbalance is within the limits of the fy (delta I) function of the Overtemperature Delta T trip.
When the axial power imbalance is not within the tolerance, the axial power imbalance o fect on the Overtemperature delta T trips will reduce rm setpoints to crovide protection consistent with core safety limits.
o 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System frot,. overpressurization and thereby prevents the release of i
radionuclides contained in the reactor coolant from reaching the containment L
atmosphera.
The reactor pressure vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Flant which permits a maximum transient pressure of 110% (2735 psig) of design pressure.
The' Reactor Coolant System piping, valves ana fittings, are designed to ANSI B 31.1 1967 Edition, which permits a maximum transient pressure of 120% (2985 psig) of component design pressure.
The Safety Limit of 2735 psig is therefore consistent with I
the design criteria and associated code requirements.
The entire Reactor Coolant System is hydrotested at 3107 psig, 125% of design pressure, to demonstrate integrity prior to initial operation.
2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS
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The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the values at 3
which the Reactc; Trips are set for each functional unit.
The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents, Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Tsip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.
T SEQUOYAH - UNIT 2 B 2-2 Amendment No. 130, 146
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- - - - - - - - ' ' " - - ' - - ~ ~ - - - -
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REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITlhG CONDITION FOR OPERATION 3.1.1.3 The moderator temperature coefficient (MTC) she.ll bt within the limits specified in the COLR.
The maximum upper limit shall be less than 0 celta K/K/* f.
APPLICABILITY:
Beginning of Cycle Life (BOL) Limit - Modes 1 and 2* onlyr End of Cycle Life (EOL) Limit - Medes 1, 2 and 3 onlyt ACTION:
a.
With the MTC more positive than the BOL limit specified in the COLR operation in Fodes 1 and 2 may preceed provided:
1.
Centrol rod withdrawal limits are established and maintained sufficient to restore the MTC to less positive than the BOL limit specified in the COLR within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in H0T STANDBY within tne next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
These withdrawal limits shall be in addition to tne insertion limits of Specification 3.1.3.6.
2.
The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition.
3.
In lieu of any other report required by Specification 6.6.1, a Special Report is prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition, b.
With the MTC more negative then the EOL limit specified in the COLR l
be in HOT SHUT 00hN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
eff greater than' or equal to 1.0
- With K PSee Special Test Exception 3.10.3 SEQUOYAH - UNIT 2 3/4 1-4 Amendment No. 28,146
REACTIVITY CONTROL SYSTEMS 9
SURVEILLANCE REQUIREMENTS 4.1.1.3 The MTC shall be determined to be within its limits during each fuel cycle as follows:
a.
The MTC shall be measured and compared to the BOL limit specified in the COLR prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading, b.
The MTC shall be measured at any THERMAL POWER and compared to the 300 PPM surveillance limit specified in the COLR (all rods withdrawn, l
RATED THERMAL POWER condition) within 7 EFPD after reaching an equilibrium baron concentration of 300 ppm.
In the event this comparison indicates the MTC is more negative than the 300 PPM surveillance limit specified in the COLR, the MTC shall be remeasured and compared to the E0L MTC limit specified in the COLR at least once per 14 EFPD during the remainder of the fuei cycle,
)
4 SEQUOYAH - UNIT 2 3/4 1-5 Amendment No.146
RfACTIVITY CCATROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES CA00P HEI6HT LIMITING CON 0! TION FOR OPERA 110N 3.1. 3. )
All full Icngth (shutoown anc Control) rods shall be GPERACLE and pcsitioned within i 12 steps (indicateo position) of their group step counter demand posit 1cn.
APPLICABILITY: Modes l' ano E*.
ACTION:
a.
With one or more full 1(noth rods inoperable cue to being immovable as a result of excessise friction or mechanical interference or known to be untrippable, determine that the SHUTOOWN MARGIN require-ment of Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and be in h0T STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b.
With more than one full length rod inoperable er misaligned 'am the Srcup step counter demano position by more than i 12 steps (indicatec position), be in HOT STAN08Y within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
c.
With one full length rod inoperable due to causes other than addressed by ACTION a, above, or misaligneo from its group step counter demand height by more than i 12 steps (indicated position), POWER OPERATION may continue provided that within one Four eitner:
1.
The red is restored to GPERA8LE status within the above alignment requirements, or 2.
The remainder of the rods in the group with the inoperable rod are aligned to within + 12 steps of tha inoperable rod while niaintaining the rod sequence and insertion limit of Specifica-tion 3.1.3.6.
The THERMAL POWER level shall be restricted l
pursuant to Specification 3.1.3.6 during subsequent operation, or 3.
The rod.is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. POWER OPERATION may then continue provided that:
a)
A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall confirm that the previourly analyzed results of these accidents remain valid for the curation of operation under these o
conditiens.
- 5ee 5pecial Test Exceptions 3.10.2 and 3.10.3.
5-QuGYAH - UNIT 2 3/4 1-14 Amendment No. 104, 146
~,
REACTIVITY CONTROL SYSTEMS SHUTD0kN R0D INSERTION LINIT LIMITING CONDIT!En FCR OPERATION 3.?. 3.5 All shutdown roas shall t.e limited in physical insertion as specified in the COLR:
APPLICABILITYs Modes 1* ana 2*t.
ACTION:
With a maximum of one shutdown rod insertea beyond the insertion limit speci-t'.ed in the COLR, exctpt fcr surveillance testing pursuant to Specifica-tion 4.1.3.1.2, within one hour either:
a.
Restore the rod to within the insertion limit specified in the l
COLR, or i
b.
Declare the rod to be inoperable and apply Specificaticn 3.1.3.1.
SpAVEILLANCEREQUIREMENTS 4.1.3.5 Each shutdown rod shall be determined to be within the insertion l
limit specified in the COLR:
a.
-Within 15 minutes prior to withcrawal of any rods in control banks A, B, C or'D during an approach to reactor criticality, and b.
At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.
- 3'ee SpecTal Test Exceptions 3.10.2 anc 3.10.3.
- With Keff greater than or equal to 1.0 SEQUOYAH - UNIT 2 3/4 2-20 An.endment No. 98,146
kEACTIVITY CCNTROL SYSTEMS CONTROL ROD INSER110N LIMITS LIMITlhG CONDITION FOR OPERATION 3.1.3.6 The control banks shali Le limiteo in physica) insertion as specified in the COLR.
APPLICAEILITi Modes 1* and 2**.
ACTION:
With the control beriks inserted beycod the insertion limits, except for sur-seillance testing pursuant to Specification 4.1.3.1.2, either:
a.
Restore the control banks to within the limits within two hours, or b.
Recuce THERMAL POWER within two hours to less than ur equal to that fraction of RATED THERMAL POWER wnich is allowed by the group position using the insertion limits specified in the COLR, or c.
Be in at least h0T STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.1.3.6 The positicn of each control bank shall be determined to be within the insertion limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals when the Rod Insertion Limit Monitor is inoperable, then verify the individual red positions at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
- 5ee Special Tesf Exceptions 3.10.2 and 3.10.3.
FWith X ff gt *'er than or equal to 1.0, SEQUOYAH - UNIT 2 3/4 1 21 Amendment No. 33, 104, 146 1
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m l
e SEQUOYAH - Utin 2 3/4 1-22 Am igment ho. 98, l
This page intentionally deleteo, 9
I SEQUOYAH - UNIT 2 3/4 1-23 Am ndment iio. 33, 9E
3/4.2 POWER DISTRIBUTION LIMITS
^
3/4.2.1 AXIAL FLUX DIFFERENCE (AFD)
LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the limits specified-in the COLR.
l APPLICABILITY:
MODE 1 above 50% of RATED THERMAL F0hER*.
ACTION:
a.
With the indicated AXIAL FLUX DIFFERENCE outside of the limits specified in the COLR; 1.
Either restore the indicated AFD to within the limics within 15 minutes, or 2.
Reduce THERMA! POWER to less than 50% of 1ATED THERMAL PuWER within 30 minutes and reduce the Power Range Neutron-Flux-High Trip setpoints to less than or equal to 55 percent of RATED THERMAL PCKER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
b.
THERMAL F0WER shall not be increased abcve 50% of RATED THERMAL POWER unless the indicated AFD is within the limits specified in the COLR.
SEQUOYAH - UNIT 2 3/4 2-1 Amendment No. 21,146
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SEQUOfAH - UNIT 2 3/4 2-3 Amendment No. 2L 146
POWER DISTRIBUTION LIMITS 3/4.2.2 HEATFLUXHOTCHANNELFACTOR-Fg IMITING CONDITION FOR OPERATION 3.2.2 F (z) shall be limited by the following relationships:
9 F (z) $ [F
] [K(z)] for P > 0.5 q
P F (z) 1 [F
] [K(z)] for P $ 0.5 q
0.5 where F
= the F limit at RATED THERMAL POWER (RTP) q specified in the COLR, THERMAL POWER
, and p _".
RATETTHERMft. POWEl K(z) = the normalized F (z) as a function of core height speci-fied in the COLR.
O APPLICABIL2:
MODE 1 ACTION:
With F (z) exceeding its limit:
g a.
Reduce THERMAL POWER at least 1% for nch 1% F (z) exceeds the limit
~
q within 15 minutes and similarly reds.ce the Power f ange Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsequent POWER OPERATION may proceed provided the Overpower Delta T Trip setpoints (value of K ) have been reduced at least 1% (in AT span) for each 1% F (z) 4 9
exceeds the limit, b.
Identify and correct the cause of the out of limit condition prior to increasing. THERMAL POWER; THERMAL POWER may then be increased provided F (z) is demonstrated through incore mapping to be within its limit.0 SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisicas of Specification 4.0.4 are not applicable.
SEQUOYAH - UNIT 2 3/4 2-4 Amendment No.
21, 95, 131, 146
- - - - - = - - - - - - - - - -. - - - - - - - - - -
POWER DISTRIBUTION LIM _ITS SURVEILLANCE __ REQUIREMENTS (Continued) 4.2.2.2 F (z) shall be evaluated to determine if F (z) is within its q
limit by:
9 Using the movable incore detectors to obtain a power distribu-a.
tion map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.
b.
Increasing tr.e measured F (z) compo' lent of the power distribution q
map by 3 percent to account for manufacturing tolerances and further increasing the value by 5% to accouN. for measurement uncertainties.
Satisfying the following relationship:
c.
M Fg (z) $ F x K(z) for P > 0.5 P x W(z)
N RTP Fq (2) 1 F x K(z) for P 5 0.5 W(z) x 0.5 where F (z) is the measured F (z) increased by the allowances for 9
RTP manufacturing tolerances and measurement uncertainty, F 43 q
the F limit, K(z) is the normalized F (z) as a function of core 9
9 height, P is the relative THERMAL POWER,and W(z) is the cycle dependent function that accounts for power distribution transients encountered during normal operation.
F
, K(z), and W(z) are speci-fied in the COLR as per Specification 6.9.1.14.
MeasuringF[(z)accordingtothefollowingschedule:
d.
1.
Upon achieving equilibrium conditions after exceeding by 10 percent or more of RATED THERMAL POWER, the THERMAL POWER at which F (z) was last determined,* or q
2.
At least.once per 21 effective full power days, whichever occurs first.
- During power escalation at the beginning of eacn cycle: power level may be increased until a power level for extended operation has been achieved and a power distribution map obtained.
SEQUOYAH - UNIT 2 3/4 2-5 Amendment No. 21, 95, 131, 146
POWER DISTRIBb, ION LIMITS-SURVEILLANCE REQUIREMENTS (Continued) e.
With measurements indicating F (z) maximum over z K(z) has increased since the previous determinatin of F N(z) either of the following actions shall be taken:
9 1.
in (2) shall be incre? sed by 2 percent over that specifited in 4.2.2.2.c, or N(z) shall be measured at least once per 7 effective full 2.
Fq power days until 2 successive maps indicate that F" (I) maximum is not increasing, over z-
~k(z) f.
With the relationships specified in 4.2.2.2.c above not being satisfied:
1.
Calculate the percent F (z) exceeds its limit by the following q
expression:
N
]x100 Fg (z) x W(z)
-1 maximum for P > 0.5
(
over z RTP p
x K(z) l l \\
Q l
l
<\\
.J
/
)
IQ (z) x (2) aximum
-1 x 100 for P < 0.5
% over z TP p
x K(z) l
_0. 5 j
2.
Eitner of the following actions shall be taken:
L a.
Place the core in an equilibrium condition where the limit in 4.2.2.2.c is satisfied.
Power level may then be increased provided the AFD limits of Specification 3 are reduced 1% AFD for each percent F (z) exceeded its l
limit, or q
b.
Comply with the requirements of Specification 3.2.2 for F (z) exceeding its limit by the percent calculated above.
9 SEQUOYAH - UNIT 2 3/4 2-6 Amendment No. 21, '35, 131, 146
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A s
acums. uur 2 m 2-7 menemeet m. m. le
POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR ENTHALPY HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION _
N 3.2.3 The Nuclear Enthalpy Hot Channel Factor, F shall be limited by the following relationship:
g F q $, F P (1.0 + PF3g (1.0-P)]
THERHAL PCWER where P = 44TED THERMAL PW T '
(=TheF limit at RATED THERMAL POWER (RTP) specified in the N
COLR, and N
PF3g = The power factor multiplier for F specified in the COLR.
A_PPLICABILITY:
MODE 1 ACTION:
With F exceeding its limit:
H 4,
Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hourt. and reduce the Power Range Neutt on Flux-High Trip 5etpoints to 5 5ta, of RATEJ THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, N
b.
Demonstrate thru in ore mapping that F is within its limit g
within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeoing the limit or reduce THERMAL POWER to less than-5% of RATED THERHAL POWER vithin the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />,-and c.
Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by a.
or b. above; subsequent POWER OPERATION may proceed provided that l
F is demonstrated through in-core mapping to be within its limit g
I at a nominal 50% of RATED THERMAL POWER prior to exceeding this l-THERMAL POWER," at a nominal 75% of RATED THERMAL POWF.R prior to l
-exceeding this THERMAL POWER and within 24 hourt ifter attaining L
95% or greater RATED THERM _AL POWER.
SEQUOYAH - UNIT 2 3/4 2-8 Amendment No. 21, 130, 145
- - -. - - - - -, ~ - - -
_P_EACTIVITY CONTROL SYSTEMS P,ASES 3/4.1.1.3 PCCERATOR _TffPERAlbRE_ COEFFICIENT (Continued) involved subtracting the increnental change in the MDC associated with a core ccr.cition of all rcds inserted (rost positive MDC) to an all rods withorawn cct: cit 1cn and, a conversion f or the rate of chenge of rioderator censity with ternperature at RATED THERMAL P0WER concitioris.
This value of the PLC was then transf ortred into the limiting end of cycle life (ECL) PTC value.
The 300 PfM serveillance limit MIC value represcrts a conservative value (with corrections for burnup and soluble boron) at a core condition cf 300 ppm equilibrium boron concentration ano is cbtained by meking these correcticr.s to the limitir.g EOL a
MTC value.
The surveillance requiren,ents for raersurenent of the M1C at the beginning and rt6r the end of the f uel cycle ar e adecuate to confirni thet the MTC remains within its lin.its since this cccf ficient charges slowly due principally to the reducticn in RCS ooron contentration associated with fuel burnup.
3/4.1.1.4 MlhlMUM TEMPERtTURE FOR CRITICAlli_Y This specification ensures that the reactor will not be made critical with the Reactor Coolant Syst(s average ter.perature less than 541'F.
This linaitation is required to ensure 1) the moderator ten perature coef ficient is within it analyzed temperature range, 2) the protectivt instrumentation is within its normal operating rarge, 3) the F-12 interlock is above its setpoint,
- 4) the pressurizer is capable cf being in a CPERABLE status with a steam tubble, and 5) the reactor pressure vessel is above its minimum RI NDT temperature.
3/4.1.2 _EOPATION SYSTEMS The boron injection system ensures that negative reactivity control is available curing each mode of facility operation.
The components required to perform this function include 1) borated water sources, 2) charging punps, 3) separate flow paths, 4) boric acid transfer pumps, 5) associated heat tracing sys t en.5, and 6) an emergency power supply from OPERACLE diesel generators.
With the RCS average temperature above 350*F, a mininum of two separate and redundant boren injection system are provided to ensure single functional capability in the event an assumeo f ailure renders one of the flow paths inoperabic. The boration capability of either flow path is sufficient to SEQUOYAH - UhlT E B 3/4 1-2 Amendment No.146
4 M4.2P0WERD151RIEUTIONLIMITS W I5 The specifications of this section provide assurance of fuel integrity during Conc 1tton 1 (hcimal Operation) ard 11 llncidents cf Federate Frequency) events t'y:
(c) naintaining the calculated CNER in ti.e core at or aticyc design during nor.tal cptritten and in short term transients, anc (ti) limiting the fissicn gas release, f uti pellCt terrperature and cloading trechanical
- rcperties to within asswec design critcria.
In additicn, limiting the pcat linear power dersity during Ccrdition 1 events provides assurance that the initial cenditions tssumed 1or the LOC A ir elyses are met crid the ECC5 acceptance criteria litnit of 2200*f 15 not exceed (d.
The definitions of certain hot charrti cna peaking tectors as used in these specifications are as f ollcws:
F (z)
Heat flux hot Channel factc-r, is defined as the maxinium local 0
heet flux on the surf ace of a f uel roc at core elevation z civided in the average fuel rod heat flux, allcwing for manufacturing toierances on fuel pellets and rods.
F Nuclear Enthairy Rise Hot Chanr.el f actor, is defincd as the ratio of the integral of linear power alcr:9 the too with the highest integrated power to the tverage red power.
3/4.i. 1 AXIAL FLUX DIFFERENCE (AFD)
The limits on AXIAL FLUX DIFFERENCE (AFO) assure that the F (z) upper g
t, cur.d cnveloce of thc F limit specified in the COLR times the nornalized l
g axial peaking factor is not excetoed during either normal operation cr in the event of xencn redistribution following power changes.
Provisions for monitcring the AfD on an automatic basis are derived from the plant proceis cc-Mter through the AFD Monitor Alarm.
The compuer deter-trnnes the one minute average of each of the CPERABLE excore detector outputs and provides an alarm message inn.ediately if the AfD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are outside the allowed Wl-Power opteating space and the THERMAL POWER is greater than 50 percent of RATED THERMAL POWER.
3/4.2.2 and 3/4.2.3 HEAT FLUX AND NUCLEAR ENTHALPY h0T CHANNEL FACTORS I
Tht limits on heat flux hot channel factor and t.uclear enthalpy hot chan-nel factor ensure that 1) the design limits on peak local pcwer density and minimum DNBR are not exceeded ana 2) it. the event of a LOCA the peak fuel clad e
temperature will not exceed the 2200'F ECCS acceptance criteria limit.
SEQUOYAH, - Oh!T 2 E 3/4 2-1 Attendment No. 21, 130, 131, 146 l
POWER DISTRIBUTION LIMITS BASES Each of these hot channel factors is meesurable but will normally only be deterreir:ed periodically as specif iec in Specifications 4.2. 2 and 4.2.L This periodic surveil W ee is sufficier.t to insure that the limits are maintained provided:
a.
Control rods in a single group move together with no ind'.vidual rod insertion differing by more than + 13 steps from the group demand
- position, b.
Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6.
c.
The control rod insertion limits of specifications 3.1.i.5 and 3.1.3.6 are miintained.
d.
The axial power distribution, expressed in terms of AX1AL FLUX DIFFERENCE, is maintained within the limits.
The F limit as a function of THERMAL POWER allows changes in the radial q
power shape for all permissible rod insertion limits.
F will be maintained q
within its limits provided conditions a thru a above, are maintained.
When an F measurement is taken, both experimental error and manufacturing q
tolerance must ba allowed for.
The 5% is the appropriate allowance for a full core map taken with the in core detector flux mapping system and 3% is the j
appropriate allowance for manufacturing tolerance.
Wnen F is measured, experimental error must be allowed for and 4% is the appropriate allowance for a full core map taken with the in-core detection ThespecifiedlimitforF?g also contains an 8% allowance for uncer-system.
tainties which rnean that normal operation will result in F
.08.
H i i
The 8% allowance is based on the'following considerations.
a.
abnormal perturbations in the radial power shape, such as f rom rod misalignment, effect F more directly than F g
q, b.
although rod movement has a direct influence upon limiting F to q
within its limit, such control is not readily available to limit I g, and c.
errors in prediction for cc ntrol power shape detected during startup physics test can be compensated for in F by restricting axial flux q
distribution.
This compensation for F is less raadily available.
H SEQUOYAH - UNIT 2 B 3/4 2-2 Admendment No. 21, 130, 146 m__
POWER 0151RIBUT10N LIMITS BASES Fuel rod bowing reduces the value of DNB ratio.
Margin has ',een retained between the ONBR value used in the safety analysis (1.38) and the WRB 1 correlation limit (1.17) to completely offset the rod bow penalty.
The applicable value of rod bow penalty is referenced in the FSAR.
Margin in excess of the rod bow penalty is available for plant design flexibility.
The hot channel f acter F N(2) is measured periodically and increased by q
a cycle and height dependent power factor W(z), to provide assurance that the limit on the hot channel f actor, F (Z), is met.
W(2) accounts for the effects 9
of normal operation transients and was determined from expected power control maneuvers over the full range of burnup conditions in the core.
The W(2) l function is specified in the COLR.
3/4.2.4 QUADRANT POWER TI,LT RATIO The 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 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 correction of a dropped ot misaligned rod.
in the event such action does not correct the tilt, the margin for uncertainty on F is reinstated by reducing 9
the power by 3 percent from RATED THERMAL POWER for each percent of tilt in excess of 1.0.
3/4.2.5 DNB PARAMETERS _
The limits on the ONB 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 been analytically cemonstrated adequate to maintain a minimum DNBR of greater than or equal to the safety analysis DNBR limit throughout each analyzed transient.
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.
SEQUOYAi - UNIT 2 0 3/4 2-4 Admendment No. 21, 130,
)
146 l
l ADMINISTRATIVE CONTROLS MONTHLY REACTOR OPERATING REPORT 6.9.1.10 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs or Safety Valves, shall be submitted on a monthly basis no later than the 15th of each rnonth following the calendar month covered by the report.
CORE OPERATING LIMITS REPORT 6.9.1.14 Core operating limits shall be established and dccumented in the.
CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:
g 1.
Moderator Temperature Coefficient BOL and EOL limits and 300 ppm surveillance limit for Specification 3/4.1.1.3, 2.
Shutdown Bank Insertion Limit for Specification 3/4.1.3.5, 3.
Control Bank Insertion limits for Specification 3/4.1.3.6, 4.
Axial Flux Difference timits for Specification 3/4.2.1, 5.
Heat flux Hot Channel fac*.or, K(2), and W(z) for Specification 3/4.2.2, and 6.
Nuclear Enthalpy Hot Channel Factor and Power Factor Multiplier for Specification 3/4.2.3.
6.9.1.14.a The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC in:
1.
WCAP-9272-P-A, " WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOCY", July 1985 (W Proprietary).
(Methodology for 5,tecifications 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5
- Shutdown Bank Insertion Limit, 3.1.3.6 -
Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Hot Channel Factor.)
2.
WCAP-10216-P-A, " RELAXATION OF CONSTANT AX1AL 0FFSET CONTROL F g SURVEILLANCE TECHNICAL SPECIFICATION", JUNE 1983 (W Proprietary).
(Methodology for Specification 3.2.1 - Axial flux Difference (Relaxed Axial Of fset Control) and 3.2.2 - Heat Flux Hot Channel Factor (V(z) surveillance requirements for F Hethodology).)
q 3.
WCAP-10266-P-A Rev. 2, "THE 1981 REVISION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE", March 1987, (W Proprietary).
(Methodology for Specification 3.2.2 - Heat FTux Hot Channel Factor).
6.9.1.14.b The core opcrating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanicai 'i its, core thermal-hydraulic limits.
m ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.
SEQUOYAH - UNIT 2 6-22 Amendment Nos. 44, 50, 64 66, 107, 134, 146
- - - - ^ - - - - - - - - - - ~ - - - - - - - - - - - - - - - - - - " - -
.-.___.._____.__.m t
i ADMIN!$TRAfi'/E CONTROLS r
CORE 0PERATING LIMITS REPORT (Continued)
~
6.9.1.14.c THE CORE OPERAT!NG LIMITS REPORT shall be provided within 30 days after cycle start up (Mode 2) for each reload cycle er within 30 da/s of issuance of any midcycle revision to the NRC Document Control Desk with copies i
to the Regional Administrator and Resident Inspector.
i SPEC!At REPORTS 6,0.2.1 Special reports shall be subnd(teo within the time period specified for each report, in accordance witn 10 CFR 50.4.
6.9.2.2 Diesel Generator Reliability l$krovement Frogram As a minimum the Reliability Irprovement Program report for NRC audit, required by LCD 3.8.1.1, T6ble 4.8-1, shall include:
(a) a summary of al? tests (velio and invalid) that occurred within the time pcriod over which the last 20/100 valid tests were performed (b) analysis of failures and dotevmination of root causes of failures (c) evaluation of eachoof the recommendations of NUREG/CR 0660 " Enhancement rfOnsiteEmergen;yDieselGeneratorReliabilityinOperatlngReactors,"
with respect to their application to the Plant (d) identification of all actions taken or to be taken to 1) correct the root causes nf failares defined in b) abcve and 2) achieve a generai improveme..t of diesel generator reliability (e) the chedule for implementation cf each action fror.; d) above (f) an asse'3sment of the existing reiiability of electric power to engineered-safety-feature equipment t
SE000YAH - UNIT 2 6-22a Amendment Nos. 44, 50, 64, 66, 107, 134, 146 l
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. _,.. _. _ _... _.. _... _,. _,... _. _..... _,.... _,., _ _, _. -. _ -,,.., _. _ ~. _ _.... _. - -,. _. ~
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-