Text

Proposed Tech Specs Re Core Operating Limit Rept
	ML20091B251
Person / Time
Site:	Sequoyah
Issue date:	05/24/1991
From:	; TENNESSEE VALLEY AUTHORITY
To:	;
Shared Package
ML20091B244	List: ML20091B240; ML20091B251;
References
	NUDOCS 9105310054
	Download: ML20091B251 (82)
	v • d • e

_ . _ _ _ _ _ _ __-._ _ _ _ . _ _ _ _ . _ ._______... _ _ _ _ _ .___ _ ._

l i i ;

FJJCLOSURE 1 PROPOSED TECilNICAL SPECIFICATION CilANGE l

SEQUOYAll NUCLEAR PIANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328

'l (WA-SQN-TS-91-08 )

LIST OF AFFECTED PAGES UniL1 UniL2 .

I 1 II II 1-2 1 1-3 1-3 1-4 1-4 1-5 ,

1-5 1-6 1-6 1-7 1-7 3/4 1-4 1-8 3/4 1-5 3/4 1-4 3/4 1-14 3/4 1-5 3/4 1-20 3/4 1-14 3/4 1-21 3/4 1-20 3/4 1-22 3/4 1-21 3/4 1-23 3/4 1-22 3/4 2-1 3/4 1-23 3/4 2-4 3/4 2-1 ,

3/4 2-5 3/4 2-3 3/4 2-6 3/4 2-4 3/4 2-7 3/4 2-5 3/4 2-9 3/4 2-6 3/4 2-10 3/4 2-7 B3/4 1-2 3/4 2-8 B3/4 2-1 B3/4 1-2 B3/4 2-2 B3/4 2-1 B3/4 2-4 B3/4 2-2 6-21 B3/4 2-4 6-22 9105310054 910524 7 PDR ADOCK 050 P

- . - . - - - - ....-__. _ . - . ._. . - . . . _ . . . . - - - _ , - . _ - ~ . - . . . - . . - . , , . . _ . . . - . _ - . - , _ . . _ _ . - , .

t ) g INDEX DEFINITIONS SECTION

1. 0 DEFINITIONS 1.1 ACT10N.................. ......................................... 1-1 <

1. 2 AXIAL FLUX DIFFERENCE.......................................... .. 1-1

1. 3 BYPASS LEAKAGE PATH............................................... 1-1 1.4 CHANNEL CALIBRATION...< ........................................... 1-1 1.5 C HAN N E L C H E C K . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1-1.

1.6 CHANNEL FUNCTIONAL TEST........................................... 1-2

1. 7 CONTAINHENT INTEGRITY............................................. 1-2

1. 8 CONTROLLED LEAKAGE................................................ 1-2

/

10 lancALTERATION...................................................

1. 9 CORE 1-2 R O/+0tArrs/G t.os,ars ifulver'

/. / / -l-10 DO S E EQU I VA L E NT I- 131. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 N

. /./2. -h-1+ E- AVERAGE DI S INTEGRATION ENERGY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3

/. /3 1.14 ENGINEERED SAFETY FEATURE RESPONSE TIME........................... 1-3

/ s 1 1.13- F R EQ U E N CY NO T AT ION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 ......

l. /s 14- GASE005 RA0 WASTE TREATMENT SYSTEM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 l

1 /. //. 15- I D E NT I F I E D L EA KAG E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 l.17 16 M EMB E R S O F T H E PUB LI C . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4

l. /g it O F F S I TE DO S E C ALCU LAT I ON MANU AL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 I. I 9 -la 0 P E RAB L E - O P E RAB I L I TY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4

/.zo 1,19 0PERATIONAL H00E - H00E. . . . . . . . . . . . . . . . ........ .............. 1-4

/. J ! 20 P HY S I C S T E S T S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . '. . . . . . . . . 1-4

/ 11 1.21-PRESSURE BOUNDARY LEAKAGE........ . ...... ........... ........... 1-5

/.13 -Me P ROC E S S CO NTRO L P R0G RAM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , , , . . . . . . . . 1-5

\

l Anendment No. 71 -

SEQUOYAH - UNIT 1 1 Pay 18, 1988 l

- ---,.-,w. ,., , r--ww- .,

. s INDEX DEFINITIONS ,

SECT 10N PAGE 1.0 DEflNITIONS (Continued) l ,2A .M4- PU RG E - P U RG I NG . . . . . . . . . . . . . . . . . . . .. .............. .. ........... 1-5

'-E l , zc444-QUAD RANT POWE R T I LT RAT 10. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

I Z M J S RATED THERMAL POWER....... ... ............... ..... .......

1, z7 .JJ6. RE ACTOR TRIP SYSTEM RESPONSE TIME. . . . . . . . . . . . . . . . . . . . . . . . . I I, 2 t7-147 R E PO R T AB L E EV ENT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

/, z.y .4J8. SHI E LD BUI LDI NG INT EGRITY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

/, 3o 449 S HU T DOWN MARG I N . . . . . . . . . . . . . . . . . . . . . . . . . ...... ............

6 I l .J l -1J0 S I T E B0VND ARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . .. 1-6

/ .12 -141 50 L 101 F I C AT 10 N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6 P75

/, 3 3 -18S SOURC E CHEC K. . . . . . . . . . . . . . . . . . ... .... ...... .. .... . .... ... 1-6

_ ] , ff --l-B S T AGG E R E D T E S T B A S I S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1-6

/, K -144 T H E RMA L P0WE R . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6

/,3r 4d5 UN I D E N T I F I ED L E AKAG E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-7 f,37 lJ G UNRESTRICTED AREA................. ............... .............. 1-7

/JS 't-7 VENTILATION EXHAUST TREATMENT SYSTEM. . . . . . . . . . . . . . . . . . . . . . . . . . . 1-7

/.194-38 VENTING.. . ....... . . ... .............................. ...... 1- 7 OPERATIONAL MODES (TABLE 1.1)...... . . . ...... , . ....... 1-8 FREQUENCY NOTATION (TABLE 1.2). ........ .. .. . ... . . .. .... 1-9 Amendment No. 71 May 18, 1988 SEQUOYAH - UNIT 1 11

, s DEFINITION 5

,g CHANNEL FUNCTIONAL TEST

1. 6 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 9 includi,ng alarm and/or trip functions.

b.

Sistable channels - the injection of a simulated signal into the [

sensor to verify OPERABILITY including alarm ana/or trip functions.

c.

Digital channels - the injection of a simulated signal into the

, channel as close to the sensor input to the process racks as R145 practicable functions.

to verify OPERASILITY including alarm and/or trio CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:

a. All are Tenetrations either: required to be closed durirg accident conditions 1)

Capable of being closed by an OPERABLE containment automatic isolation valve system, or fjp3h 2)

Closed by manual valves, blind flanges, or ceactivated automatic 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.

ach air lock is in compliance with the requirements of w' fication 3. 5.1. 3, d.

The containment leakage rates are within the limits of Specification 3.6.1.2, and e.

The sealing mechansim associati:d with each penetration (e.g.,

we'ds, bellows, or 0-rings? is CPERABLE.

CONTROLLED LEAKAGE 1.3 coolant 'ONTROLLEO pump seals. LEArv\GE shall be tnat seal water flew suppliec to the reac*or CORE ALTERATION 1.9 CCRE ALTERATION shall be the movement or manipulation of any component within the vessel. tne reactor pressure vesd with the vessel head removec anc fuel in Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position. '

eORA OMAAnt% LI w 1T- 12ttPa RY

/. to L IN.pcererreremw a) y 399

~-

SEQUOYAH - UNIT 1 1-2 Amencment No. 17., 71, 130, 141

_ _ _ _ -__ ____-_ _ _ - _ - - - - - - - - - ~~

Ar7atgg,7~r 73,

. i

/, /0 ,The CORE OPERATING LIMITS REPORT (COLR) is the junit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be

, determined for each reload cycle in accordance with

! Specificationqb fxm Unit operation within these ioperating limits is addressed in individual specifications.

G R .\.\H t

i e

4 O e

, i DOSE-EQUIVALENT I-131 l

/,// -1r10. DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / l R; gram)-which_alone would produce the came thyroid dose as the quantity and isotopic mixture of I-131, I-132, 1-133, I-134, and 1-135 actually present.

The thyroid dose conversion f actors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

5 - AVERAGE DISINTEGRAlION ENERGY

_f,/2. "1713 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) fnr 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

/,/3 -4r42 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval R:

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,

%.s.r FREQUENCY NOTATION R'

f, j c/j -tT13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

- GASEOUS RADWASTE TREATPENT SYSTEM R'

/,/6I~tr14 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce. radioactive gaseous effluents by collecting primary coolant system of fgases 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

- f ./6 -4r45- IDENTIFIED LEAKAGE shall be: R

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 SEQUOYAH - UNIT 1 1-3 Amendment No. X12 71 May 18, 1988

. ~. . . . - - _ - - --_ - .- .-. . - .- . _

b. Leakage into the containment atmosphere from s0urces that are both specifically located and known either not to interfere with the i operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or

c. Reactor coolant system leakage through a steam generator to the secondary system.

MEMBER (S) 0F THE PUBLIC

/,q -1d6 MEMBERS OF THE PUBLIC shall include all individuals who are not occupationally associated with the plant. This category shall include non-employees of the licensee who are permitted to use portions of the site R75 for recreational, occupational, or other purposes not associated with plant functions. This category does not include non-employees such as vending a machine servicemen or postmen who, as part of their formal job function, occasionally enter an area that is controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials.

OFFSITE DOSE CALCULATION MANUAL (00CM)

/,lS 17 The 0FFSITE DOSE CALCULATION MANUAL (00CM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radio-active gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints, and in the conduct of the Radiological g3 Environmental Monitoring Program. The 00CM shall also contain (1) the Radioactive Effluent Contrnis and RMiological Environmental Monitoring Programs required by Section 6.8.5 and (2) descriptions of the information that should be included in we the Annual Radiological Environmental Operating and Semiannual Radioactive Ef fluent Release Reports required by Specifications 6.9.1.6 and 6.9.1.8.

OPERABLE - OPERABILITY R75 J.; 1fr A system, subsystem, train, or component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s), and when all necessary attendant instrumentation, controls, a normal and an emergency electrical power snurce, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s).

OPERATIONAL MODE - MODE R75

/. bo 19 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power lavel and average reactor coolant temperature specified in Table 1.1.

PHYSICS TESTS R75 f,21 1de 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.

1-4 Amendment No. 12, 71, 148 SEQUOYAH - UNIT 1 NOV 161980

PRESSURE BOUNDARY LEAKAGE

/,77~M1- 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]

j,23. l'~22 The PROCESS CONTROL PROGRAM shall contain the current formulas, sampling, p analyses, tests, and determinations to be made to ensure that the processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way at to assure compliance with 10 CFR Parts 20, 61, and 71; State regulations; and other g requirements governing the disposal of solid radioactive wastes.

PURGE - PURGING

/,gf -Id3 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 such a manner that replacement air or gas is required to purify the confinement.

QUADRANT POWER TILT RATIO A3dir24- QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated (g/ outputs, or the ratio of the maximum lower excore @tector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

[ With one excore detector inoperable, the remaining three detectors shall be l used for computing the avera;e.

RHEDTHERMALPOWER(RTP) R145

/, g(, M& RATED THERMAL POWER (RTP) shall be a total reactor core heat transfer rate to the reactor coolant of 3411 MWt.

l I

REACTOR TRIP SYSTEM RESPONSE TIME 16 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from f,27 +when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coil voltage.

REPORTABLE EVENT A REPORTABLE EVFNT shall be any of those conditions specified in

/,23 Section 50.73 to 10 CFR Part 50.1 +7' l

l SEQUOYAH - UNIT 1 1-5 Amendment No. 12, 71, 141 148 NOV 16 '.090

SHIELD BUILDING INTEGRITY pgM M 8' SilIELD BUILDING INTEGRITY shall exist when;

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 treatment system is OPERABLE.

c. The sealing mechanism associated with each penetration (e.g., ,

welds, bellows or 0 rings) is OPERABLE.

SilUTDOWN MARGIN M 9 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which R75

/' g 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 witiirawn.

SITE BOUNDARY

/,3/.-180'TheSITEBOUNDARYshallbethatlinebeyondwhichthelandisnotowned, R75 leased, or otherwise controlled by the licensee (see Figure 5.1-1).

SOLIDIFICATION ./

/,J2 1d1- Deleted SOURCE CHECK j,J3.1dt Deleled R

STAGGERED TEST BASIS R75 f,y[ -143- A STAGGERED TEST BASIS shall consist of:

a. A test schedule for n systems, subsystems, trains or other designated components 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.

THERMAL POWER

/,36 Ad4* THERMAL POWER shall be the total reactor core heat transfer rate to the R75 reactor coolant.

SEQUOYAH - UNIT 1 1-6 Amendment No. 12, 7L 148 w

N OV l e., < wJ

1 UNIDENilflfD LEAKAGE

/, ,g (, -id'r UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE lR75 or CONTROLLED LEAKAGE.

UNRESTRICTED AREA f, 3 7 J A fr An UNRESTRICTED AREA shall be any area, at or beyond the site boundary R75 to which access is not controlled by the licensee for purposes of protection

of individuals from exposure to radiation and radioactive materials or any area within the site boundary used'for residential quarters or industrial, commerical, institutional, and/or recreational purposes.

VENTILATION EXHAUST TREATMENT SYSTEM Ad7 A VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and f, 3 3 installed to reduce gaseous radiciodine 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 parti-culates from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents).

Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING R75

!' 4 48-" VENTING is the contro d process of discharging air or gas from a

$dy) confinement to maintain temperature, pressure, humidity, concentration or other operating 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.

SEQUOYAH - UNIT 1 1-7 Amendment No. Q 71 May 18, 1988

- __ - .- . . . - - ~ . . - . . _

REACTIVITY CONTROL SYSTEMS .

MODERATOR TEMPERATURE COEFFIClfNT LIMITING CONDITION FOR OPERATION 3.1.1.3 Themoderatortempe5aturecoefficient(MTC])

rne m A x e o ,um a e,*a. c shall u m , r- bet utT,4:s swn u. p umruir i<

t smr o5 Dc< rn KjKl,.

yran seceu oxo in rwe cos- .

-a Uss pM i ti"^ than -de4ta--kA/4--fee--the-aM-cods--w&thdeawnr-beginntw}-

-ef--eyc4e--44fe-4801--} r-botc-s eco-4NERMAL-90WER-cond4t4ent

-4

-b. Lcr; negat4ve-then -?.0 x 10 h'M%MbhWW

-deawar -end-ef-eyele-44fe-fEOL-)-RATED-THERHAL P0U0 condi t4en-ija ,~,a un c, or c ye u c 4 isx (dor ) L t**

t~

APPLICABILITY: Specification 3.1.1. 3.a - MODES 1 and 2* only#

Specification 3.I 1.3.!> - MODES 1, 2 and 3 only#

GNO oC tytn< L.tr< ( Go s. ) t. s en e r-ACTION:

gou Secc ofoco tN rms couA.

a. With the MTC more positive than the4 11mit c' 3. ' '.3.c obove operation in MODES 1 and 2 may proceed provided:

1. Control rod withdrawal limits are established and maintained suf ficient to restore the MTC to less positive than 0 delts rW Bo' ddl'J#

L i m i r- s Weicearo k/k/ f within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY within the next 6

'"

C "#

hours. These withdrawal limits shall be in addition to the -

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 MEC 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 R40 Special F Jort is prepared and submitted to the Corr. mission pursuant ,o 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.

g o t. S K csf,40 in rrtl dou R.

1 2 b-abo n, be in

b. With the MTC more negative than the4 11 mit 2.'

HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l *With Keff greater than or equal to 1.0 l #See Special Test Exception 3.10.3 ,

November 23, 1984 %

SEQUOYAH - UNIT 1 3/4 1-4 Amendment No. 36

i REACTIVITY CONTROL SYSTEMS w

SURVEILLANCE REQUIREMENTS 4.1.1.3 The MTC shall be determined to be within its limits during each fuel cycle as follows:

sfru FsG O IN Y t4 E COLk

a. The MTC shall be measured and compared to the BOL limit ++-6 pee 444ee-'

tier 3_' '.3.a, above, prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading, TM ? o ppm 5**V'-'*'< ' ' ' ' T' 'M ' ' F ' * '" "" ' * *

b. The MTC shall be measured at any THERM L POWER and compared to w -3.1 x 10'4 dcita k/k/ I (all rods withdrawn, RATED THERMAL POWER condition) within 7 EFP0 after reaching an equilibrium boron concentration of 300 ppm. In the event this comparison indicates that MTC is mo.e negative than -h-1 x 10 -4 de4ta k/k/ I, the MTC 4'

shall be remeasured, and compare to the E0L MTC limit of Spcci44ee- sFeciF' AO m ra rs cot /A4er. 3. ' 1. 3. b, at least once er 14 EFP0 during the remainder of the fuel cycle, k%w

""* T tf *C 3aa l's G LA*l fri s L G A~ C6 L s vu t~

bfG c. omo in 7,n cm-. A'.

SEQUOYAH - UNIT 1 3/4 1-5

e 8 REACTIV11Y CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES i

GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3.1.3.1 All full length (shutdown and control) rods shall be OPERABLE and positioned within i 12 steps (indicated position) of their group step counter demand position.

APPLICABILITY: MODES l' and 2*

ACTION:

a. With one or more full length rods inoperable due to being immovable as a result of excessive f riction or mechanical interference or known to be untrippable, determine that the SHUT 00WN MARGIN requirement of Specification 3.1.1.1 is satisfied within I hour and be in HOT STANOBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With more than one full length rod inoperable or misaligned from the group step counter demand position by more than i 12 steps (indicated position), bs. in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c. With one full length rod inoperable due to causes other than addressed by ACTION a, above, or misaligned from its group step counter demand , _ _ . .

height by more than d 12 steps (indicated position), POWER OPERATION m4*w) may continue provided that within one hour either: -

1. The rod is restored to OPERABLE status within the above alignment requirements,

2. The remainder of the rods in the group with the inoperable rod are aligned to within i 12 steps of S N '{.,"4 "' ", g g, '

the inoperable rad within one hour while maintaining / P '

the rod sequence and insertion limit of f4 m: 3 s

' 1-lx R118 T,the THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or

3. The rod is declared inoperable and the SHUT [)0WN MARGIN requi-rement of Speci fica tion 3.1.1.1 is sa t i s f ied. POWER OPERATION may then continue provided that:

5eeSpeci[Ilesttxceptions3102 and 3.10.3. . .

SEQUOYAH - UNIT 1 3/4 1-14 Amendment No.114 May 5, 1989 j l

RfACTIV!1Y CONTROL SYS1 EMS SHU100WN ROD INSERTION LIMIT r

(

LIMITING CONDITION FOR OPERATION L e ns ur ro su f+t yv e m- i.~ sx. crow 14VGke r i eno 3.1.3.5 All shutdown rods shall be -4 les rwi caest Rll? l

. .y ' i tMeaw++ 3 A- l APPLICABILITY: MODES 1* and 2*#

4 ACTION: ,,,7g ,ypo y,z ,, g,,,,,, oa L , n , e Sg e , ,,,g o

~

in reo.x <%e With a maximum of one shutdown rod D u!!y eithdrawn, except for surveillance testing pursuant to Specification 4.1.3.1.2, within one hour either:

1371. rvac *rrtr k'a0 7b w o rr< m we in s <im0N L i n , r- 5Me i A6D t^2 a, -Ftr+-ly w i t hd raw-We-w&ce * '"'b D A

b. Declare the red to be inoperable and apply Specification 3.1.3.1.

SURVEILLANCE REQUIREMENTS iNM m a Lemer Ape:cer=in.o in rn c GOLI 4.1.3.5 Each shutdown rod shall be determined to be ;ully --ithdrawr.

a. Within 15 minutes prior to withdrawal of any rods in control banks A, B, C or 0 during an approach to reactor criticality, and

b. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

'See Special Test Exceptions 3.10.2 and 3.10.3.

With Kgf7 greater than or equal to 1.0.

EEu14y+ it hd twwn-s he H -be-t he -co nd i t ion-whe re-s hu tdowera nd-een t ex>1- hank te tre-tt,rptei-t-f on-withi-n-the--intervel-of .-~ 2Pe--entt ~ 231--ttepni4:hdr. - 'r

--im.4ttt4ve-- R112 SEQUOYAH - UNIT ] 3/4 1-20 Amendment No. 108 March 2&, 1989

REACTIVITY CONTROL SYSTEMS i-CONTROL R00 INSERTION LIMITS LIMITING-CONDITION FOR OPERATION 3.1. 3. 6 The control banks shall be limited in physical insertion as chea" ';

-Figu e '.1-1. specu cmo - 7 w e c o u g . .. .R45 APPLICABILITY: MODES 1* and 2*#,

ACTION:

With the control banks inserted beyond the *4*ner insertion limits, except for surveillance testing pursuant to Specification 4.1.3.1.2, either:

a. Restore the control banks to within the limits within two hours, or

b. Reduce THERMAL POWER within two hours to less than or equal to that fraction of RATED THERMAL POWER which is allowea by the group position using the-ebovc figun, cc Rll8 A ms e,ers aa t.s .s~ s n 5Menfo&o s u r,+g Cas_ g ,o g,

c. Be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

$NE SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the Rod Insertion Limit Monitor is inoperable, then verify the individual rod positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

I

See Special Test Exceptions 3.10.2 and 3.10.3.

With Keff greater than or equal to 1.0.

1 SEQUOYAH - UNIT 1 3/4 1-21 Amendment No. 41,114 -

May 5, 1989

Df( L fi T*C (Fu Withdr wn)"

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SEQUOYAH - UNIT 1 3/4 1-22 Amendment No. 108 March 28 1989

l 1Xi L W TN.

l 7 ,/' l R

IIVITY fahTROL SYSTE)) ,-

I

/

FIGUffE 3.1-1 N.0M'i 10N [ / ,

/ fully ' /

hdrawn shaj)/eb the co

/ /

ition where tfutdown and c itrol banks are at a,p ition wi in the into val of > 22 d < 231 ste withdrawn, i ciusive. /

~ ~

,/ p Rll2 There no rod in ertion limi hen the shutJdwn and control ba ks are at -

a p M ion withi he interva 222 and < 231 steps withdrawn,,A clusive.

I fully withdrawn positipa hall be s Mfied in a reload dTety evalua :on or each qyde of opera On and, onc ecified, shall no ed ess such af ange is spept 1callyevapa d. j/ echang/

Tf2 SEQUOYAH - UNIT 1 3/4 1-23 Amendment No. 41,100 March 28, 1989

l-l l

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE (AFDJ LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the aMowed-opacc L in o r-s roent4onal--space-4eHned-bf.

aic o ia roc c ou, gute 3. 2-1.

APPLICABILITY: MODE 1 above 50% RATED THERMAL POWER

ACTION: '

a. With the indicated AXIAL FLUX DIFFERENCE outside of the Figur 3.2-1 limitsx srec F<ao <~ rue C4 "4)

1. Either restore the indicated AFD to within the-Agurc 3.2 limits within 15 minutes, or

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux-High Trip setpoints'to less than or equal to 55 percent of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ,

b. THERMAL POWER shall not be increased above 50% of RATED THERMAL dd POWER unless the indicated AFD is within the Figurc 3.2 limits x sen p s a o <~ ru co w a. . .

December 23, 1982 SEQUOYAH - UNIT 1 3/4 2-1 Amendment No. 19 1

\

' D E W T 'E Revision 23

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i December 23, 19H2 Amendment No. 19 SEQUOYAH UNIT 1 3/4 2 t*

.. 4

POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR-F {Z)

LlHITING CONDITION FOR OPERATION 3.2.2 Fq (Z) shall be limited by the following relationships:

F9 (Z) < +[G -[K(G) for P"> 0.5 [F7 ][ KN] at44 P

[ Fo*] [ K (2 )]

FQ (Z) < [2. 2?] h(?)] for P < 0.5 o,s R144 WH4Ac Ferg 7gg g L, % or arso Twwt Pw4 (^'r d 9 ##

" THERMAL P WER "*C'#>

RATED THERMAL POWER 3 K D:W TM N o tt ete.c o Fy le.) As d A mr< w o"

-end-n(B-i+-the-func-t-ien-obt.e4eee.F-fm--f4 9 ure 3. 2" f+r ' given ere height-lecatien. co4E #vis a r sM(c i c,( o <<v ru ir cod .

APPLICABILITY: MODE 1 ACTION:

With F (Z) exceeding its limit:

9 W.w

a. Reduce THERMAL POWER at least 1% for each 1% n F (Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed provided the Overpower Delta T Trip Setpoints (value of K4 ) have been reduced at least 1% (in oT span) for each 1% F (Z) exceeds the limit, Q

b. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; THERMAL POWER ,ay then be increased provided Fq (Z) is demonstrated through incore mapping to be within its limit SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

R144 SEQUOYAH - UNIT 1 3/4 2-5 Amendment No. 19, 95, 100 1 MAY 111990

= .____ _ - - . - _ . _ _ _ .

POWER DISTRIBUTION tlMITS SURVEILLANCE REQUIREMENTS (Continued) f 4.2.2.2 limit by: F9 (3) shall be evaluated to determine if QF (Z) is within its a.

Using the movable incore detectors to obtain a power distribu-tion map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.

b. Increasing the measured F g) component of the power distribution map by 3 percent to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties. '

c. Satisfying the following relationship:

F 2q4 "x K(z)

FQ "(z) 0.5 R144 F

g hw(z) x gg) K(w) f 3 meaowuwm q (z) $ W(z) x 0.'S f r P $ 0.5 F 9(p) /13 A Fu-xr'o R144 oc N esa.c fiO4 H G where Fg (z) is the measured Fn (z) creased by the allowances p pCP manufacturing tolerances and me rement uncertainty,4 9

'ws is the F limit, M(z) in given ,i,,igure3r2-2,Pisther91ative THERM 3LPOWER,andW(z)isthecycledependent function that Aw,a acct'unts for power distribution transients encountered during normal operation, -T h i s func t-i c a i s g Wen-br-th e i'e a k i ng F a c te r

-L4m4-Repoet-es--pa r Spec i f ica t ic a c. 0.1.14, d.

7 n e.ct=vio i" ress c o urt n ' s k a ci'ce ocieAoa(a) Ff x c, .g . t,. ma iq. d* ) d" Measuring qF "(z) according to the following schedule:

1. Upon achieving equilibrium conditions af ter exceeding by 10 percent or more of RATED THERMAL POWER, the THERMAL POWER at which Fq (z) was last determined,* or

2. At 'mt once per 31 effective full power days, whichever occurs first.

During power escalation at the beginning of each cycle, power level may be incre'ased until a power level for extended operation has been achieved and a power distribution map obtained.

R144 SEQUCfAH - UNIT 1 3/4 2-6 Amendment No.' 19, 95, 100 MAY 111.090

t POWER DISTRIBUTION LIMITS i

SURVEILLAt4CE REOUIREMENTS (Continued)

e. With measurements indicating M

maximum IO (#)

, over z d(z) has increased since the previous determinatin of f M(z) eitter of the following actions shall be taken:

1. F q

N(7) shall be increased by 2 percent over that specified in 4.2.2.2.c, or

2. F q

"(z) shall be measured at least once per 7 effective full power days until 2 successive maps indicate that maximum M Q

(I) is not increasing, over z K(z)

f. 7 ith the relationships specified in 4.2.2.2.c above not being satisfied:

gga 1. Calculate the percent F 9 (z) exceeds its limit by the following expression:

( - -, h M

F maximum 9 (z) x W(z) i -1 x 100 for P > 0.5

$ over z

~

I FW'2. 32 x K(2) V

) Rl44 l (f -

f 9

maximum Q (Z)

Wl'd -1 ,,x 100 for P < 0.5 over z y M x K(z) gl44 l Frp R /- 0. 5 ,

9 a s

2. Either of the following actions shall be taken: y yy g

a. Place the core inanequilibriumconditionwherethe[

limi t in 4. 2. 2. 2.c is satis fied. Power level may Jnen be increased provided the AFD limits of T'gus 3.c=1- are reduced 1% AFD f or each percentq F (z) exceeded its Iimit, or t, . Comply with the requirements of Specification 3.2.2 for Fq (z) exceeding its limit by the percent calculated above.

R144 l

. \

SEQUOYAH - Ut4IT 1 3/4 2-7 Amendment No. 19, 95, laa j MAY 111990

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POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR ENTHALPY HOT CHANNEL FACTOR f LIMITING CONDITION FOR OPERATION --

N RI' 3.2.3 The Nuclear Enthalpy Hot Channel Factor, F g , shall be limited by the following relationship: ,

---Where+--

RrP i

Fu A PFm 4 F3g 1 M- [1. 0 + -0. 3 (l 0- P )] ,

THERMAL POWER (aatte-b p =

RATED THERMAL POWER '

_ cre hhs = T et s F E n t.i n ,- ,z y g n,x, yr,r,inu A~ r ,z WWj spec u ra n o t ^t T r+ w c e, s ,q ,q p o APPLICABILITY: MODE 1 P F4 9 : Tw Musa 7mru en -r,a,a M 9" ' " # '~ 7ur Coua.

ACTION:

With F H xceeding its limit:

a. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to i 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ,

b. Demonstrate thru in-core mapping that F H is within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and R14:

c. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWEh above the reduced limit required by a.

or b. above; subsequent POWEP OPERATION may proceed provided that g is demonstrated through in-core mapping to be within its limit F{N at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior to exceeding this THERMAL POWER and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months af ter attai.1ing 95% or greater RATED THERMAL POWER.

SEQUOYAH - UNIT 1 3/4 2-10 Amendment No. 19, 138

.. ! 0 ..0 l 1

REACTIVITY CONTROL SYSTEMS BASES END OF CY 6' < L ' #'C (fo'-)

condition of all rods inserted most positive MOC) to an all rods withdrawn condition and, a conversion fo the rate of change of moderator density with temperature at RATED THERMAL P WER conditions. This value of the MDC was then 'g o o ppm transformed into the limiting gMTC value. -?.0-x-le'4 defte-k/kPf- The'[MTC y,y*

value +f -3.1 x 10 delta k/k/ f-- represents a conservative value (with correc-tions for burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making these corrections to the limiting EoL-

~# VHW 6' FDC nl = ^.0 10 MW/*% M T't The surveillance requirements for measurement of the MTC at the beginning and near the end of each fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RC5 boron concentration associated with fuel burnup.

3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY s

YW$

This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 541*F. This limitation is required to ensure 1) the moderator temperature coefficient is within its analyzed temperature range, 2) the protective instrumentation is within its normal operating range, 3) the P-12 interlock is above its setpoint,

4) the pressurizer is capable of being in a OPERABLE status with a steam bubble, and 5) the reactor pressure vessel is above its minimum RT t"*P""*t"I'*

NOT 3/4.1.2 BORATION SYSTEMS The boron infection system ensures that negative reactivity control is available during each mode of facility operation. The e.cmponents required to perform this function include 1) borated water sources, 2) charging pumps, 3) separate flow paths, 4) boric acid transfer pumps, 5) associated heat tracing systems, and 6) an emergency power supply from OPERABLE diesel generators.

l l With the RCS average temperature above 200*F, a minimum of two separate and redundant boron injection systems are provided to ensure single functional l capability in the event an assumed failure renders one of the systems inoperable.

The boration capability of either flow path is suf ficient to provide a SHUTDOWN MARGIN from expected operating conditions of 1.6% delta k/k af ter xenon decay and cooldown to 200 F. The maximum expected boration capability requirtment occurs at EOL from full power equilibrium xenon condition, and requires a SEQUOYAH - UNIT I B 3/4 1-2

l l

l 3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (a) maintaining the calculated DNBR in the core at or above design during normal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding mechanical properties to within assumed design criteria. In addition, limiting the ceak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200*F is not exceeded.

The definitions of certain hot channel and peaking f actors as used in these specifications are as follows:

F0 (Z) Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing fer 'nanuf acturing tolerances on fuel pellets and rods.

Fh Nuclear Enthalpy Rise Hot Channel Factor is uefinec as the ratio of the a integral of linear power along the rod with the highest integrated power to

~~~

the average roc power.

nu f Lq t w r- 1.Patsp x o in VM Coud 3/4.2.1 AXIALFLUXOIEPbNCE(AFD)

The limits

/

envelope of-fH!I$oti AXIAL times the FLUX 01FFERENCE normalized assure axial peaking that is factor thenot F (Z) upper boundR144 exceedec during either normal operation or in tne event of xenon redistribution follow-ing power changes.

Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer deter-mines the one minute average of each of the OPERABLE excere cetector outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are outside the allowec aI-?cwer coerating space and the THERMAL POWER is greater than 50 percent of RATED THERMAL POWER.

3/4.2.2 end 3/4.2.3 HEAT ; LUX AND NUCLEAR ENTHALPY HOT CHANNEL FACTORS Rl42 The limits on the heat flux hot channel f ac'.or and tne nuclear enthalpy rise het channel tactor ensure that 1) the design limits on ceak local power density and minimum DNBR are not exceeded anc 2) in the event of a LOCA tne peak fuel clad temperature will not exceed tne 22000F ECCS acceptance criteria limit.

.N l "'.y SEQUOYAH - UNIT 1 B 3/4 2-1 Amendment No. 19, 138, 140 Correction Letter of 5-16-90

POWER OISTRIBUTION LIMITS -

BASES (

Each of these hot channel factors is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic provided: surveillance is sufficient to insure that the limits are maintained a.

Control rods in a single group move together with no individual red insertion differing by more than + 13 steps from the group. demand position.

b.

Control rod groups in Specification are sequenced with overlapping groups as described 3.1.3.6.

c. The control red insertion limits nf Specifications 3,1.3.5 and 3.1.3.6 are maintained.

d.

The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.

The Fh limit as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits. Ffg will be maintained within its limits provided conditions a thru d above, are maintained.

When an F %M tolerance must9be allowed for. measurement is taken, both experimental error and manufacturin The 5% is the approcriate allowance for a full core map taken with the incere detector flux mappin l

appropriate allowance for manufacturing tolerance. g system and 3% is the R142 When an F g is measured, experimental error must be allowed for and 4% is the appropriate allowance for a full core map taken with the incore detection system. The specified limit for F H also contains an 8% allowance for MP uncertainties which mean that normal operation will result in F W<

h

~

Fag 1.55/1.08. The e% allowance is based on the following considerations.

t a.

abnormal perturbatios in the radial power shape, such as from rod misalignment, effect F g more directly than F .

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 F

g , and c.

errors in prediction for control power shape detected during startup physics test can be compensated for in9 F by restricting axial flux distributton. ThiscompensationforFbH 9 is less readily available.

W 0 8 N90 SEQUOYAH - UNIT 1 8 3/4 2-2 l Amendment No. 19,-138 Correction letter of 5-16-90

~

.rs

POWER DI5TMPUTION LDiiTs V,'t s

BASES / 5

( l s

Fuel rod bowing reduas the value of DM ratic. Margin has been retained betwcen the CNH v:iue used in tne saf a:y analysis (1.33) and the WRB-1 .{

corralation limit (1.17) to completely offsat tha red bow penalty.

p- R142 The applicable value w rod bew pensity is rafstenced in the FSAR.

Margin in excen of the rod bow penalty is availabic for plant design flexibility.

The hot channal factor F I is measurad periodically and increased by a cycleandheightdeoendentpcha(r) factor,W(z),toprovideassurancethatthe i

lirait on the het channel facter, F p (t), is cet. W(2) accounts for the effects of no- esi cparatien tn .ienu :nd ' eas warmined from axpected power control inneuvers over t.ne f ui! range of burnu: conditions in the core. The W(z) function h ne r" c. in 2: providee in tha Ndn; 7actr List 4errt -

-par-Si>e c i !i c : t ' ; ; E . 2. '. 1 '.- 1.5 s Pir c oM o iH rH<r C 0 t- 2 3/4.2.a GUADRANT ?CW!R i:LT RATIO The quadran. 73ar tilr. ratic limit anures that the radial power distri-bution satisfies the design valuac used in the power capability analysis.

Radial power distribution measuramants are made during startup testing and &

periedically during powar opcration.

The two hour tim 3 alicvance for cparation with a tilt condition greater than 1.02 but less then 1.03 is proviced to allow identification and cor-rection of a droeped er misalignad red. In the avsnt such action does not correct the tilt, tha m.r;in fer uncertainty en F is reinstated by reducing the power by 3 parcsnt from MTED THERNAL POWER fhr each percent of tilt in excess of 1.0.

3/4.2.5 DrG PARAMETEi!

The limi s on tha _M raletad paraat;rs assure that each of the para-reters an naiq uir e 3i ain *.na nory:i study state envele:e of operation assumed in tha t a nti w tnj accident analyses. The limits are consistent with tne initill Fir -

wuw; ions anc have been analytically demonstrated adeCJate t ' :3 4 '. , i a ai:,'.um CN3E Cf greatur inan or equal to the safety R142 anciysis ONO lic, t tc:cughout saca anelyzed transiert.

The 12 Muc nric 'i: wrvaill:nce of thest par m ters through instrument read 0ut is :ui",' ct ta m:u,: tbc. t h parame:=.rs are restored within their limi ts follem g ice..i cb :ss and otNr expected transient cperation.

$} . *y <c c r, ,

SEQUOYAK - UET 1 6 3/4 2-4 Amerdment No. 19, 133

_ - - --- - . . - - - - -. . - . . ~ . - - . - . - _ -

ADMINISTRATIVE CONTROLS MONTHLY REACTOR OPERATING REPORT 6.9.1.10 Routine reports of operating stat istics and shutt.,wn experience, including documentation of all challenges to the PORVs or Safety Valves, shall R76 be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report, Any changes to the OFFSITE DOSE CALCULATION MANUAL shall be submitted with the Monthly Operating Report within 90 days in which the change (s) was made effective.

In addition, a report of any major changes to the radioactive waste treatment systems shall be submitted with the Monthly Operating Report for the period in which the evaluation was reviewed and accepted by the PORC, Gordc OffM rn4 /imor; SAM/ oat'

-RA01 AL-PEAKING 4AC-TOR 44 H H-RENJ pm g gf;fw a v,m f,vs e r C 6.9.1.14 [TW(z) ~ nction norma! pthation h dl be provi d at lea 60 Aays nor to ycle ini al critic ty. In e event th t hese va es

,wtfu l subm 76 60 sprior)tf.edats to the atethevalc)eswoulde other ecome t me effect duringfrore ve unlesslife, therwis i will be st)

/ itted j mpted y the C ission.

An nforma ' n needed to suport - will be b request fropt the NR and ne

,po be i edinth)s' report, j' ,/

SPECIAL REPORTS -

%&J 6.9.2.1 Special reports shall'be 'submitiad within the time pericd specified for each report, in accordance with 10 CFR 50.4. R76 6.9.2.2 Diesel Generator Reliabilit1/ Improvement Proqram As a minimum the Reliability Improvement Program report for NRC audit, required by LCO 3.8.1.1, Table 4.8-1, shall include:

Rs6 (a) a summary of all tests (valid and invalid) that occurred within the time period over which the last 20/100 valid tests were performed (b) analysis of f ailures and determination of root causes of failures (c) evaluation of each of the recommendations of NUREG/CR-0660, " Enhancement of Onsite Emergency Diesel Generator Reliability in Operating Reactors,"

with respect to their application to the Plant (d) identification of all actions taken or to be taken to 1) correct the root causes of failures defined in b) above and 2) achieve a general improvement of diesel generator reliability (e) the schedule for implementation of each action f rom d) above (f) an assessment of the existing reliability of electric power to engineered-safety-feature equipment al2tl SEQUOYAH - UNIT 1 6-21 Amendment Nos. 52,58,72,74,117 June 19, 1989

. ._____.___.__...__..____.__.____-____m_.

INSERT C CORE OPERATING LIMITS REPORT 6.9.1.14 Core operating limits shall be established and documented in the

CORE OPERATING LIMITS REPORT before each reload cycle or any remaining Part of a reload cycle for the following:

1. Moderator Temperature Coef ficient 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 limits for Specification 3/4.2.1,

5. Heat Flux Hot Channel Factor, K(Z), and W(Z) for Specification 3/4.2.2, and

6. Nuclea* 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 METHODOLOGY", July 1985 (W Proprietary).

(Methodology for Specifications. 3.1.l.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 AXIAL OFFSET CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION", JUNE 1983 (W Proprietary), i (Methodology for Specification 3.2.1 - Axial Flux difference (Relaxed Axial Offset Control) and 3.2.2 -

Heat Flux Hot Channel Factor (W(2) surveillance requirements for Fq Methodology).)

3. WCAP-10266-P-A Rev.2, "THE 1981 REVISION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE", March l987, (W Proprietary).

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor).

l l

. . . _ , _ . _ - _ . . . _ _ _ . . . -_ _ _ . . . . . . _ _ _ _, - . . - . _ . _ . = _ . _ . - _ . . _ . . .4_.m___ _ . - - . ,

n

_2 CORE OPERATING I.IMITS REPORT (continued) 6.9.1.14.b The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

6.9.1.14.c Ti!E. CORE OPERATING 1.1MITS REPORT shall be provided within 30 days af ter cycle start-up (Mode 2) for each reload cycle or within 30 days of issuance of any midcycle revision to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

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.I.N._D_E_X.

OfflNil!0NS SECTION PAGE 1.0 DEFINITIONS 1.1 ACT10N......'...................................................... 1-1 1.2 AX1AL FLUX 01FFERENCE.................... ........................ 1- 1

1. 3 BYPASS LEAKAGE PATH........................... . . . . . ............. 1-1

1. 4 C H AN N E L C A '.10 R AT 10 N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1

1. 5 CHANNEL CHICK..................................................... 1-1 (63 1.6 CHANNEL FUNCTIONAL TEC1............... .............. ............ 1-2

1. 7 CONTAINMENT INTEGRITY... .. ......................... .... .-2

1. 8 CONTROLLED LEAKAGE.... .......................................... . 1-2

1. 9 CORE ....... . 1-2 Mo Cotti ALTERATION...............k........................

0/k o M 4 Ls n o r- / m- 1-3 f , o +.-M 00 5 C E QU IV A L E NT 1 - 131. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

f, t ?,-h-ti E- AV E RACE D I SI NT EG R AT I ON EN E RG f. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 f,,3 4-if ENGINEERED SAFETY FEATURE RESPONSE TIME. . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 f, r y .ad-3 F R EQ U E N CY N O T AT 10 N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

13 1-3 f, /f+-14 GASEOUS RA0 WASTE TREATMENT SYSTEM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1-4

/ , / (. MS- ! D E N T I F I E D L E A A AG E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

................................... 1-4 f, /7 4d& MEMBERS OF THE PUBLIC. . . . . . . .

/, I G .2d7 0 F F S I T E DO S E C A LCU LAT ION MANU A L . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4

/,I9 M & OPERABLE - OPERABILITY.............................. .... . . .... 1-4

/, 7o .his O P E RAT I O N A L MO D E - M0 0 E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . 1-5 l /,2/ E0 P H Y S I C S T E S T S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5

/,2 2 .M1-f RESSURE BOUNDARY LEAKAGE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..... 1-5

/, 2 3 -M2 P ROC E S S CONT RO L P R0G R AM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 e

I Amendment No. 63 SEQUOYAH - UNIT 2 May 18, 1988 I

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I INDEX I DEFINITIONS SECTION PAGE 1.0 DEflNITl_0NS (r 'inued)

/, 2 */ .1 r 2 31U RG E - PU RG I NG . . . . . . . . . . . . . . . . . . . . .... .............. ....... 1-5

/* 2 I.1-24 QUADRANT POWER TILT RATIO. . . ........ ......... .. .. .... ..... 1-5

/, 2 4 H5 R AT E D T H E RMAL POWE R . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .... ..... .... 1-6 f,zy .L 26 REACTOR TRIP SYSTEM P.ESPONSE TIME.......... .......... .. ........ 1-6

/, 7 8 J.27. R E P O R T AB L E E V E N T . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6 n63

...... ...... . 1-6

/. 2 9 k/ 28 SHIELD BUILDING INTEGRITY.... ... ...... . ......

f, 3 6 ),29 SHUTDOWN MARGIN................ ................... .............. 1-6

/, 3/ ,LJO S I T E B 0 V N D A R Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1- ri j,j2,LJ1' SOLIDIFICATION. . ................................. ............... 1-7

/,13 Jaf SOURC E CHEC K. . . . . . . . . . . . . . . . . . . . . . ............ .................. 1-7 1-7

}yo.lc33STAGGEREDTESTBASIS............ .............. .. .............

6. .~. m. .. ..t f,J[.lA+THERMALP0WER.................................................... 1-7 -

7,f(, ,Jr3S UNIDENTIFIED LEAKAGE...... ... . ................. ... ........ 1-7 r,J 7 A 3& UNRESTRICTED AREA.......... ..... ................ ... .... ...... 1-7

/,/171r37 VENTILATION EXHAUSl TREATMENT SYSTEM. . . . . . . . ...... ..... ... ... 1-8

/. 3 7 4 .* V E N T I NG . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . ..... .................. 1-8 OPERATIONAL MODES (TABLE 1.1).. .... . .. ....... . . . . ........ 1-9 FREQUENCY NOTATION (TABLE 1.2).. .. ...... .... .... . .. ...... 1-10 Amendment No. 63 Hny 18, 1988 SEQUOYAH - UNIT 2 11

OffIN!TIONS

(

CHANNEL FUNCTIONAL TEST 463

1. 6 A CHANNEL ~UNCTIONAL TEST shall be:

a. Analog channels - the injection of a simulated signal into the channel as close to the seasor as practicable to verify OPERABILITY including alarm and/or tM p functions.

b. Bistable channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.

c. Digital channel', - the injection cf a simulated signal into the channel as close to the sensor input to the process racks as practi. P.132 cable to verify OPERABILITY including alarm and/or trip functions.

CONTAINMENT INTER. pity '

1. 7 CONTAINMENT INTEGRITY shall exist when: 163

a. All penetrations required to be closed during accident conditions are either:

1) Capable of being closed b,, an OPERABLE containment automatic isolation valve system, or

$WL%

2) Closed by manual valves, blind flanges, or daactivated auto-matic valves secured in their closed positio . except as provided in Table 3.6-2 of Specification 3.6.3.

b. All equipment hatches are closed and sea'..d,

c. Each air lock is in c..;mpliance with the requirement:, of Ril7 Specification 3.6.1.3,

d. The containment leakage rates are within the limits of Specification 3.6. .

ind

e. The sec11ng 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 reactor R63 coolant pump seals.

CORE ALTERATION 1.9 CORE ALTERATION shall be the movement or manipulation of any component R63 within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Cuspension of C0f!E ALTERATION shall not preclude completion of movement of a component to a safe conservative position. '

gpqpe one Aws L- t m o r s (2,cso t r~

/. /O ( ins et t r A rtMet o ,w r C} ,

SEQUOYAH - UNIT 2 1-2 Amendment No. 63, 117 , 132 00T 311980

, . /fr7eleititway Q

/, /o iThe CORE OPERATING LIMITS REPORT (COLR) is the l unit-specificdocumentthatprovidescoreoperating

' limits for the current operating reload cycle. These cycle specific core operating limits shall be detennined for each reload cycle in accordance with SpecificationGE9 Unit operation within these

., operatinglimits[_E27 is addressed in individual specifications.

C G,9.\.14 e

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. , - -c- , . _ ,-.-_ ,_- ,,. . , , , , . . , , , , , - , - ,

OEFIN1110NS f

DOSE EQUIVALENT I-131

/,// & W DOSE EQUIVALENT l-131 shall be that concentration of I-131 (microcurie / lR6:

gram) which alone would produce the same thyroid dose as the quantity and iso-topic mixture of I-131,1-112,1-133, I-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table 111 of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

E - AVERAGE DISINTEGRATION ENERGY lR6;

/,/p,1Al* E shall be the average (weighted in proportion to the concentration of 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-lodine activity in the coolant.

ENGINEERED SAFETY FEATURE RESPONSE TIME lR6:

/,/ 3 Mr The ENGINEERED SAFETY FEATORE 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 ty 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.

I FREQUENCY NOTATION f, //-

MT The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

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I GASE0US RADWASTE TREATMENT SYSTEM

/,/6 .&ie A GASEOUS RADWA$TE TREATMENT SYSTEM is any system designed and installed h6:

to reduce radioactive gaseous effluents by collecting primary coolant 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.

SEQUOYAH - UNIT 2 1-3 Amendrent No. 63 May IF, 1988

DEFINIT 10NS IDENTIFIED LEAKAGE A/(, A1S-IDENTIFIED LEAKAGE shall be: K63

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 operation of leakage detection systems or not to be PRESSURE B0UNDARY LEAKAGE, or

c. Reactor coolant system leakage through a steam generator to the secondary system.

MEMBERS OF THE PUBLIC

/,/7 -1d6- MEMBERS OF THE PUBLIC shall include all individuals who are not occupa-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, R63 occupational, or otner purposes not associated with plant functions. This category does not include non employees such a vending machine servicemen or ,

1 postmen who, as part of their formal job function, occasionally enter an area Nag /

that is controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials.

OFFSITE DOSE CALCULATION MANUAL

/,/9 It The OFFSITE 00SE CALCULATION MANUAL (03CM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radio- R136 active gaseous and liquid effluents, in the calculation of gaseous and liquid offluent monitoring alarm / trip setpoints and in the conduct of the Radiological Environmental Monitoring Program. The 00CM shall also contain (1) the Radio-active Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.5 and (2) descriptions of the information that should RW 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.

OPERABl.E - OPERABILITY

/,/7 lelt- A system, subsystem, train, or component or device shall be OPERABLE or R63

have OPERABILITY when it is capable of performing its specified function (s),

! and when all necessary attendant instrumentation, controls, a normal and an I emergency electrical power source, cooling or seal water, lubrication or other i

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).

1-4 Amendment No. 63, 134 SEQUOYAH - UNIT 2 p- mg

._ _ _ _ _ _ _ _ _ _ . . _ _ ~ _ _ _ .. . _ ___ _ _ _ _ _ . _ _ _ _ _ _ ._ _ ._ _ _ __

OfflNITIONS OPERATIONAL MODE - MODE

/,74 ,td'I An OrERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive gg3 combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.1.

PHYSICS TESTS R63

/,7 / , Mnuclear O PHYSICS TESTS shall characteristics of bethethose reactor tests performed core and related to measure the fundamental 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 R63 f'7 p .E 21 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) througn a non-isolable fault in a Reac'or 'colant System component body, pipe wall or vessel wall.

PROCESS CONTROL PROGRAM (PCP)

- left The PROCESS CONTROL PROGRAM shall contain the current formulas, sampling,

-/,2 D _ analyses, tests, and determinations to be made to ensure that the processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to gg34 assure compliance with 10 CFR Parts 20, 61, and 71; State regulations; and other requirements governing the disposal of solid radioactive wastes.

PURGE - PURGING

3 7, yg .M3 from PURGE or PURGING a confinement to maintainis the temperature, controlled process pre.sure, of discharging air or gas humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

QUADRANT POWER TILT RATIO QUADRANT POWER TILT RATIO shall bi the ratio of the maximum upper excore R63 j,g( .ir24 detector calibated output to the average of the upper excore detector cali-brated outputs, or the ratio of the m&ximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, which-ever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.

4 1-5 Amendment No. 63 134 SEQUOYAH - UNIT 2

{l *; l0

DEFINITIONS 4

RATED THERMAL POWER (RTP) ,

7,74 JM RATED THERMAL POWER (RTP) shall be a total reactor core heat transfer h63 rate to the reactor coolant of 3411 Mdt.

REACTOR TRIP SYSTEM RESPONSE TIME j,7,7 ,1M The REACTOR 1 RIP SYSTEM RESPONSE TIME shall be the time interval from k63 when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coil voltage.

REPORTABLE EVENT f,/ 6,JrfT A REPORTABLE EVENT shall be any of those conditions specified in Section h63 50.73 to 10 CFR Part 50.

SHIELD BUILDING INTEGRITY

/,2 9 ,Jdff SHIELD BUILDING INTEGRITY shall exist when: 03

a. The door in each access opening is closed except when the access opening is being used for normal transit entry and exit. yr

b. The emergency gas treatment system is OPERABLE.

c. The sealing mechanism associated with each penetration (e.g., welds, bellows or 0-rings) is OPERABLE.

SHUTDOWN MARGIN 63 f,3o ,1d9-' SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subtritical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

SITE BOUNDARY R63

/,3 / ,1,-30' The SITE B0VHDARY shall be that line beyond which the land is not owned, leased, or otherwise controlled by the licensee (see figure 5.1-1).

SEQUOYAH - UNIT 2 1-6 Amendment No.'63, in 00T 311E

DEFINITIONS - ..___

SOLIDIFICATION

/,12. JrST Deleted. R134 SOURCE CHECK

/,3 3 4f57 Deleted.

kl34 STAGGERED TEST BASIS R63 f, ygl Jedf A STAGGERED TEST DASIS shall censist of:

a. A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval int.o n equal suointervals,

b. The testing of one system, subsystem, train or other designated component at the beginning of each subinterval.

THERMAL POWER u63 hga g ff.Jt34- THERMAL POWER shall be the total reactor core heat transfer rate to tnts reactor coolant.

UNIDENTIFIED LEAKAGE Jr55F UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE R63 f,34 or CONTROLLED LEAKAGE, ,

f UNRESTRICTED AREA l,37 J<aG' An UNRESTRICTED AREA shall be any area, at or beyond the site boundary to which access is not controlled by the licensee for purposes of protection of R63 individuals from exposure to radiation and radioactive materials or any area within the site boundary used for residential quarters or industrial, commer-cial, institutional, and/or recreational purposes, i

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1-7 Amendment No. 63, 134 SEQUOYAH - UNIT 2 NC:: .]

t perINiitoNS VENillATION EXHAUST TREATMENT SYSTEM

/,18 Jr37 A VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and lR6 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 gaseous exhaust stream prior to the release to the environment !

(such a system is not considered to have any effect on noble gas effluents).

Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

4 VENTING I,39 A dtf VENTING is the controlled process of discharging air or gas from a lR6 confinement to maintain temperature, pressure, humidity, concentration or other operating 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.

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l SEQU0YAH - UNIT 2 1-8 Amendment No. 63 May 18, 1988

1 i

REACTIVITY CONTROL SYSTEMS <

MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION w'o "

3.1.1.3 The moderator temperatura coef ficient (MIC) shall beWirma ra te~r5o ocs.rA K/K/9 ou ruc cu"R. TH c rn on u m w wn t. i n. o r- s um ne i.<s s run

-a. Len--positive-than 0 d !t: i'l/ r f er4hc- 31' rodw44*awa r-4eg &tming-o f -eye 4 e44-f eH40 L-)r40 t-r eco-THERMAL-POWER-eend444en,-

-b,---L+66-nega 44 ve-4,ha n--4-*40' 4 delta 4A/EF-foe--the al1 reds--w4-thdeewm

-end-e f--cyc4e44fe-(4 GL-)r-R AT EO-T H E RMAL--Pi)WER- cond 444ent

&sinnim; on Gyet< Lirc (90L.) L '"' ' 1'~

APPLICABILITY: 4peeff4eet4cn 3.1.1-3--e - MODES 1 and 2^ only#

5;;ccif teet4en 3.1.1. 3.b - MODES 1, 2 and 3 only#

d"O of C yc t-k t iM (<t o t ) 1.i m a r~

yg S/3ecofico /N rad O'K

a. With the MTC more positive than theAlimit cf 3.1. '. 3ra-obover operation in MODES I and 2 may proceed provided:

1. Control rod withdrawal limits are established and maintained suf ficient to restore the MTC to less positive than de4te-- 7 H5 ' Lot.

L t ~iv s#(c'M0-4hM within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY within the next e

~'

tu re co'4 /6 hours. These withdrawal limits shall be in addition to the 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 HTC 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 lR28 Special Report is prepared and sut;mitted to the Comission 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, gng spreifoEO IN THE C.0LA

b. With the MTC more negative than theglimit-of 3.1.'.3. W hev% be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With Kg. greater than or equal to 1.0

See Special Test Exception 3.10.3 November 23, 1984 3/4 1-4 Amendment No.28 ;

SEQUOYAH - UNIT 2 i

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REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS [

4.1.1.3 The MTC shall be determined to be within its limits during each fuel cycle as follows:

s m esso on vrw Co

a. The HTC shall be measured and compared to the BOL limit +f--GpeeH&

cat 44n-3. ' ',--3. a . above, prior to initial optration above 5% of RATED THERMAL POWER, after each fuel loading.

b. TheMfCshallbemeasuredatanyTHERMALPOWERandcomparedto 3.1 Twt Ja. /W wumw%_.toMMA/4 (all rods withdrawn, RATED THERMAL POWER L%.r vmmaa #~ e condition) within 7 EFr0 after reaching an equilibrium boron concen-W' -

tration of 300 ppm. In the evep this comparison indicates the MTC i

is more-negative than g.1 w-10--} 1:fehs-t/kMf, the MTC shall "be'"

remeasured, and compared to the EOL MTC limit +(--mec+fua-- MP'

-thn -3.1.1. 3.b, at.A east once per 14 EFPD during the remainder o[

the fuel cycle.,-

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O 3/4 1-5 SEQUOYAH - UNIT 2 T

- , - - . . ~ , . -- . , _ - - _ . - - - - - . _ - - . - , - _ _ . ~ . , . . - _ _ _ , . . , . - . - - . .

- . . _ . .-. . _-_ - -_ - . _ - - - - - - _ - . _ _ - _ _ - _ _ = _ - _ _ _ - - _ - . - . _ -

Rf AC1]V11Y CON 1ROL SYSTfMS 3/4.1.3 MOVA0lE CONTROL ASSIMBt11.5 GROUPHEIGjH LIMITING CONDITION FOR OPERATION 3.1.3.1 All full length (shutdown and control) rods shall be OPERABLE and positioned within i 12 steps (indicated position) of their group step counter demand position.

APPLICABILITY: Modes l' and 2*.

ACfl0N:

a. With one or more full length rods inoperable due to being immovable as a result of excessive f riction or mechanical ird erference or known to be t,ntrippable, determine that the SHUT 0dWN MARGIN require-ment of Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

b. With more than one full length rod inoperable or misaligned from the group step counter demand position by morc than i 12 steps (indicated position), be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

c. With one full length rod inoperable due to causes other than addressed by ACTION a, above, or misaligned from its group step counter demand height by more than i 12 steps (indicated position), POWER OPERATION may continue provided that within one hour either:

1 The rod is restored to OPERABLE status within the above alignmant requirements, or ,

c ge,raMio~

2. The remainder of the rods in the group with the inoperable rod 3'b 3'b

are aligned to within + 12 steps of the inoperable rod while maintaining the rod seluence and insertion limit of Ekyn-,3M , R104 Tihe THERMAL POWER level shall be restricted pursuant to Specifi-cation 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:

d) A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these conditions.

^5ee Special Test Exceptions 3.10.2 and 3.10.3.

SEQUOYAH - UNIT 2 3/4 1-14 Amendment No.104 O

May 5, 19H

REACTIVITY CONTROL SYSTEMS .-

SHUTDOWN R00 INSERTION LIMIT LIMITING CONDITION FOR OPERATION R9 3.1.3.5 All shutdown ro'ds shall be fu14y-wi-thdrawarAL c , e i rer o in PHysee a i n :,as ro on ns siv e o ,c co is rn< ca u st . a

/.PPLIC ABILI fY: Modes 1* and 2*#.

ACTION:

K,m ec p ge yoy, n,e < ~ s era ries tine m.y rou st e ,re ro ecut With a maximum of one shutdown rod 75t..J.uLly 4thdrawth except for surveillance testing pursuant to Specification 4.1.3.1.2, within one hour either:

&s rvas inc Nov n w a rn ,c,1 ran: savva rrow t. < ,u . r

a. Fully AtMntw-the-vodr-w W " u Go w 1" rne co u<c , ose.

b. Declare the rnd to be inoperable and apply Specification 3.1.3.1.

YZW iURVEILLA,NCE REQUIREMENTS w,w,a vm insuriva u m se 4.1.3.5 Each shutdown rod shall be determined to be fully stMeewn: {iTri Ca (6. c8 o4Iwe O (W

d. Within 15 minutes prior to withdrawal 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

'See Special Test. Exceptions 3.10.2 and 3.10.3.

With Keff grcater than or equal to 1.0 afuHy-w i t hd rawn-s ha H-be- the sond i t-i on-whe re-s hu tdownwnd < on t rol-ba nk s-aee-

-a t-a-po s4t4 on-w i-t hi n-t he -i nt e rv el-o f i 222-snd@l steps 9(itttdrawn, R98

--.inclus ius,-

l l

SEQUOYAH - UNIT 2 3/4 1-20 Amendment No. 98 March 28, 1989

_ _ _ _ _ _ . - - , _ . . . - _ . _ , - . . - _ . . _ . _ . . _ . _ . _ . _ _ _ _ . _ . _ _ _ . _ _ . _ ~ _ _ _ _ - _ _

. _ _ - . _ _ _ _ _ _ _ _ _ _ _ . . . - ._ - _ _ _ _ _ _ . _ _ . _ _ ._ - . _ ~ _ _ _ _ _ _ _

RIACilVliY CON 1ROL SYETIMS CONTROL ROD IN$lRT10N l!M1iS LIP 111NG CONDITION FOR OPERATION ,

i 3.1. 3. 6 The control banks shall be limited in physical insertion as down-b- f t33

' LigurM,44, 3 vu , p . 5 o fa rea co'<.

APPllCABILITY: Modes 1* and 2*#. !

ACTION:

With the control banks inserted beyond the h insertion limits, except for surveillance testing pursuant to Specification 4.3.3.1.2, either:

a. Restore the control banks to within the limits within two hours, or

b. Reduce THERMAL POWER within two hours to less than or equal to that j fraction of RATED THERMAL POWER which is allowed by the group ;

position using the -above ficee-oe fumano~ L 6 - . vm '

en TM& C s s A' , o k D*c 8 #4+d K104 l

c. Be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the Rod Insertion Limit Monitor is inoperable, then verify the individual rod positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

See Special Test Exceptions 3.10.2 and 3.10.3.

With Kgff greater than or equal to 1.0.

V SEQUOYAH - UNIT 2 3/4 1-21 Amendment No. 33, 104 May 5, 1989

, , REACTIVITY CONTROL SYST[Hs

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SEQUOYAH - UNIT 2 3/4 1-22 ,4mendment No. 98 March 28, 1989

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. i REACTIVITY CONTR0t SYSTEMS f

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FJ RE 3.1-1, NOTATION / -'

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FullyW1thdrawnshal[betheconditjo[whert shutdowfIand control,b$nks are at / R98

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/ interval oVj, 222 and </ 23Vsteps withdr, / awn, inclusive.

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There are nb rod inserti'onfllimits whcii the sfIutdown and cont'rol banks position gwithin the interijal > 222 and v231 steps withd/ayn, inclusive / The fully' withdrawn posit.Mn shalT be specified in a reinad safety evaluatfon for I cadi cycle of operat' ion and, once specified, shall,5fot be changed unless such j !

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SEQUOYAH - UNIT 2 3/4 1-23 Amendment No. 33, 98 March 28, 1989

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l 3/4.2 POWER DISTRIBUTION LIMITS '

3/4.2.1 AXIAL FLUX OlFFERENCF (AFD)

LIMITING __ CONDITION FOR OPERATION 3.2.1 The indicated AX1AL FLUX O!FFERENCE (AFO) shall be* maintained Uithin the alleezed 9fmrationa! qwe *fi wd by figues 3.2- 1.

Lim i15 NMc scito 84 ritt cout APPLICABILITY: MODE 1 above 50% RATED THERMAL POWER ACTION:

a. With the indicated AXIAL FLUX OlFFERENCE outside of the rijurc t 3.2-1 limitsy srec ersco /u rw co ufa .

1. Either restore the indicated AFD to within the-figure 3.2-1 R 21 limits within 15 minutes, or

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes.and reduce the Power Range Neutron Flux-High Trip setpoints to less than or equal to 55 percent of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Q b. THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD is within the -F-igurc 3.21 limits s spe e ws t o in rue cou a . '

I~ -

i SEP 2 91983 SEQUOYAH - UNIT 2 3/4 2-1 Amendment No. 21

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0PERATION '\E OPEKATION

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,.41AL FLUX 01FFE3ENCE LJEITS AS A FUNCTICN OF RAJID' THERMAL POWER SEP 2 91983 SEQUOYAH UNIT 2 3/4 2-3 Amendment No. 21

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POWER DISTRIBUTION LIMITS 3/4.2.2 HEATFLUXHOTCHANNELFACTOR-FfD LIMITING CONDITION FOR OPERATION 3.2.2 Fq (Z) shall be limited by the following relationships:

-- 'Frl"][ KW)]

, F9 (Z) 1 [2.32] [K(Z)] for P > 0.5 -

P g331

+-

g _ [Fgre][g(;2)]

Fq (Z) $ [2.32) [K(Z)] for P 5 0.5 o,s gi31 waste Fcf T*HL Fq Lw Ar<!*

W a~~%* P ~ 'ea h THERMAL POWER m ' ' #)

W ee P = RATED THERMAL POWER I X(2)z:. mt doe- Au Pxo Vq (2) A *" A P"" ' Y"' 4 "

-end- K(Z) i the--funct4en-et>ta4ned-4eee-f4 9 uce 3. 2-2 for a gi ven

-esve--he4ght-4esa%4en, eona ttA t G H r* spa e- f < i; o <~ rtu c et2.

APPLICABILITY: MODE 1 ACTION:

R21 With F (Z) exceeding its limit: ,,

9 m%m

a. Reduce THERMAL POWER at least 1% for each 1% q F (Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; 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 thre.w.h incore mapping to be within 9

ts limit.

SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

Rl H SEQUOYAH - UNIT 2 3/4 2-4 Amendment No. 21, 95, 131 l

00T F 20

l POWlR DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Cor,tinued) l 4.2.2.2 F9 (z) shall be evaluated to determine if9F (Z) is within its limit i by:

a. Using the movable incore detectors to obtain a power distribution ,

map at any THERMAL POWER greater than 5% of RATED THERMAL POWER. i

b. Increasing the measured Fq (z) component of the power distribution R21 ,

map by 3 percent to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties.

c. Satisfying the following relationship:

pv tre M

xK(z} for P > 0.5 R13 F (z) 1 @P x W(z) p /W M

F (z) s 1 x K(z] for P 1 0.5 g)i c nN pohnu M db R13 W(z) x 0.5 n., a sw c.r s es c c, a W K '" A rP

\

h where F (z) is measured F 9 (z) in jer sed by the allowances for manufacturing tolerances and 5surement uncertainty, is the F limit,4(:) !c givs :n Figurc 3.2-2, P is the relative q

THERMAL POWER, and W(z) is the cycle dependent function that accounts for power distribution transients encountered during normal operation.

-4Amit--Repoet-as-peMpeM estion C. 3.1.14. " . K G2 up e) pg c,pe s p< n o -Th+s-fttnei.-ion-is-

<~ Tw; given in the-Peeking

d. Measuring MF g (z) according to the following schedule:

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 31 effective full power days, whichever occurs first.

0uring power escalation at the beginning of each cycle, power level may be increased until a power level for extended operation has been achieved and a power distribution map obtained.

R13 i

SEQUOYAH - UNIT 2 3/4 2-5 Amendment No. 21, 95, 131-Correction Letter: 04/19/89 00T 20 20

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

e. With measurements indicating maximum F9 (z) over z

, K(z) .

has increased since the previous determination of FIq(z) either i

of the following actions shall be taken: l l

1. F"(z) shall be increased by 2 percent over that specified in 4.2.2.2.c, or I

2. F"(z) shall be measured at least once per 7 effective full

I power days until 2 successive maps indicate that maximum F"(z) 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 perr.ent F (z) exceeds its limit by the following expression: 9

~ ~

F M maximum F9 (z) x WI) -1 x 100 for P > 0.5 over z x K(z)

_ f4'r W 9

@P

. . j R131

~ ~

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\

'" (max imum I Q(z) x W(z) -1 x 100 f or P <: 0.5 q

R131 over x K(z) l 2. Either of the following actions shall be taken:

a. Place the core in an equilibrium condition where the limit in 4.2.2.2.c is satisfied. Power level maj l then be increased provided the AFD limits of R21 sro u:,ccoo t.2.1 Figure 3.2-1 are reduced 1% AFD for each percent F (z) exceeded its 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 R131 SEQUOYAH - UNIT 2 3/4 2-6 Amendment No. 21, 95, 131 Correction Letter: 04/19/89 00T 20 ii.00

POWER O!STRIBUTION LIMITS p g L rs W K(2 AS A FUNCTION OVCORE HEIGHT

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POWER DISTRI_00 TION LIMITS e

3/4.2.3 NUCLEAR ENTHALPY HOT CHANNEL FACTOR I "

i LIMITING CONDITION FOR OPERATION 3.2.3 The Nuclear Enthalpy Hot Channel Factor, F g,shall be limited by the N

1 following relationship: _,

-.a._

ermr 4,

N Fem PF.m F

AH $ 455- [ 1. 0 + -04 (1. 0 - P ) )

(unAd 4r P= THERMAL POWER ,

ATED IHERMAL POWER F$}n= 77M F}$ Limor Arskrw Trw w P*""d W W" ' #' '

APPLICABILITY: MODE 1. m rtW C o wR , M o ACTION: ff D '2' 8"" ^ ' ' N" * # ' ' ' " ' ^ ' " A '4 P0u Gi4e a f <eo in r+ + sc c o t. < e. ,

With F g exceeding its limit:

R.

a. Reduce THERMAL POWER to less than 50% of RATED THER*iAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip 5etpoints to 1 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , ggg

b. Demonstrate through in-core mapping that F is within its limit AH within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , 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 F

AH is demons trated through in-core mapping to be within its limit at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 751, of RAlED THERMAL POWER prior to exceeding this THERMAL POWER and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after attaining 95%

or greater RATED THERMAL POWER.

s.

SEQUOYAH - UNIT 2 3/4 2-8 Amendment No. 21, 130 00T M C I

. . _ _ _ -_ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ - . _ _ _ _ ____..____________.______________-___.-_____.___-_____.J

REAC11VITY CONTROL SYSTEMS I

BASES

_. reu o oc cycM W4 (60 9 3/4.1.1.3 MODERATORTEMPERATURECOEFFIC/ENT(Continued) involved subtracting the incremental hange in the MOC associated with a core '

condition of all rods inserted (mo pcs,itive MOC) to an all rods withdrawn condition and, a conversion for e rate of change of moderator density with temperature at RATED THERMAL P0 ER conditiora. This value of the MDC was then Soo PPM transformed !nto the limitinga TC value 4.0 x lO 4 den k/k/?. The[MTC WL a. v4 :#

t ~ r-

-4 delte-Vtfaf represents a conservative value (with v a l ue -of-444-40 corrections for burnup and soluble boron) at a' core condition of 300 ppm equilibrium boron concentration and is obtained by making these corrections to EOL. .

the l i mi t i ng MTC va l ue, W4-4-x-40 4-WW8F.

4 The surveillance requirements for measurement of the MTC at the beginning and near the end of the fuel cycle arc adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

Ik.%

3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALI1Y This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 541 F. This limitation is required to ensure 1) the moderator temperature coefficient is within it analyzed temperature range, 2) the protective instrumentation is within its normal operating range, 3) the P-12 interlock is above its setpoint,

4) the pressurizer is capable of being in a OPERABLE status with a steam bubble, and 5) the reactor pressure vessel is above its minimum RTNOT temperature.

3/4.1.2 80 RATION SYSTEMS The boron injection system ensures that negative reactivity control is available during each mode of facility operation, The components required to ,

perform this function include 1) borated water sources, 2) charging pumps, 3) separate flow paths, 4) boric acid transfer pumps, 5) associated heat tracing systems, and 6) an emergency power supply from OPERA 8LE diesel generators.

With the RCS average temperature above 200*F, a minimum of two separate and redundant boron injection system are provided to ensure single functional capability in the event an assumed f ailure renders one of the flow paths inoperable. The boration capability of either flow path is sufficient to SEQUOYAH - UNIT 2 B 3/4 1-2

l 3/4.2 POWER DISTRIBUll0N LIMITS BASES The specifications of this section provide assurance of fuci integrity during Condition I (Normal Operation) and 11 (Incidents of Moderate Frequency) events by: (a) m'aintaining the calculated DNBR in the core at or above design during normal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding mechanical properties to within assumed design criteria. In addition, limiting the peak linear power density during Condition 1 events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200 F is not exceeded.

The. definitions of certain hot channel and peaking factors as used in these specifications are as follows: R21 F (Z) Heat Flux Hot Channel Factor, is defined as the maximum local 9 heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods.

F"H Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the :nd with the highest integrated power to the average rod power.

N '" ' "

3/4.2.1 AXIAL FLUX DIFFERENCE (AFD) 9 ThelimitsonAXIAL[fLUXDIhERENCE(AFO)assurethattheF(Z) 9 upper R13 bound envelope of @ Mimes the normalized axial peaking f actor is not exceeded during either normal operation or in the event of xenon redistribution following power changes. .

Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The compuer deter-mines the one minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are outside the allowed al-Power operating space and the THERMAL POWER is greater than 50 percent of RATED THERMAL POWER.

R130 3/4.2.2 and 3/4.2.3 HEAT FLUX AND NUCLEAR ENTHALPY HOT CHANNEL FACTORS The limits on heat flux hot channel factor and nuclear enthalpy hot chan-nel factor ensure that 1) the design limits on peak local power density and R21 minimum DNBR are not exceeded and 2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200 F ECCS acceptance criteria limit.

B 3/4 2-1 Amendment No. 21, 130,131 SEQUOYAH - UNIT 2 00T 29 30

POWER DISTRIBUTION LIMITS BASES Each of these hot channel factors is measurable but will normally only be gi; determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is sufficient to insure that the limits are maintained R21 provided:

a. Control rods in a single group move together with no individual rod insertion dif fering by more than + 13 steps from the group demand position.

b. Control rod groups are sequenced with overlapping groups as described in Specification 3.L 3.6.

c. The control rod insertion limits of specifications 3.1.3.5 and 3.1.3.6 are maintained.

d. The axial power distribution, expressed in terms of AXIAL FLUX OIFFERENCE, is maintained within the Mmits.

N limit as a function of THERMAL POWER allows changes in the radial The F g 31 N

power shape for all permissible rod insertion limits. F eH will be maintained 9,%

within its limits provided conditions a thru d above, are maintained.

When an F measurement is taken, both experimental error and manufacturing q

tolerance must be 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 appropriate allowance for manufacturing tolerance.

N When F g 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 N frP system. The specified limit for F aH also contains an 8% allowance forg fu uncertainties which mean that normal operation will result in c AH *

The 8% allowance is based on the following considerations.

a. abnormal perturhationsN in the radial power shape, such as frcm rod misalignment, effect F g more directly than F q,

b. although rod movement has a direct influence upon limiting qF to wfthinitslimit,suchcontrolisnotreadilyavailabletolimit F H, and

c. errors in prediction for control power shape detected during startup physics test can be compensated for in Fq by restricting axial flux distribution. This compensation for F"g is less readily available.

B 3/4 2-2 Admendment No. 21,130 SEQUOYAH - UNIT 2 OCT 02 50

POWER DISTRIBUTION LIMITS (

BASES Fuel rod bowing reduces the value of DNB ratio. Margin has been retained between the DNBR value used in the safety analysis (1.38) and the WRB-1 R130 correlation limit (1.17) to completely offset the rod bow penalty.

The applicable valoc of rod bow penal;y is referenced in the FSAR.

Margin in excess of the rod bow penalty is available for plant design flexibility.

The hot channel factor F N(z) 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 factor, F (z), is met.

9 W(z) accounts for the effects ut of normal operation transients and was determined from expected power control maneuvers over the full range of burnup conditions in the core. The W(z) f unction fee-norme4-operation-is1wevided-in-the-Peeking-Feeter44mi4;--Hepoet-

-pee-6pec4 f4 cat 4on 5. 9.1.14~ / ~- L F" c W"f D IN DW Co w k_ .

3/4.2.4 QUADRANT POWER TILT 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 or misaligned rod. in the event such action does not correct the tilt, the margin for uncertainty on F q

is reinstated by reducing 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 DNB related parameters assure that each of the para-meters are maintained within the normal steady state envelope o' cr3 ration assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of greater than or equal to the safety 4t30 analysis DNBR limit throughout each analyzed transient.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months periodic surveillance of these parameters through instrument readout is suf ficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

B 3/4 2-4 Amendment No. 21,130 SEQUOYAH - UNIT 2 '

00T 02 J^0

ADMINISTRATIVE CONTR01.5 '

f 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 R64 be submitted on.a monthly basis no later than the 15th of each month following the calendar month covered by the report.

Any changes to the 0FFSITE DOSE CALCULATION MANUAL shall be submitted with the Monthly Operating Report within 90 days in which the change (s) was inade effective.

In addition, a report of any major changes to the radioactive waste treatment systems shall be submitted with the Monthly Operating Report for the period in which the evaluation was reviewed and accepted by the PORC.

Coin or % m p n Lem er.s /2xAa r-4AMAL-PEAK 4NG-FAC-TOM-If4IT REPORT f.9.1.14[The (2) f ction F normal paration rrtfall be rovideg at lea 5.

M" 160 tfay / pci to c le in' iai .rit' ality. I he ev . that ese va es l/ou' be bmit ed at eme other me durino cre lif4., it w Las o itted/

6. day prior o the ta the y ues would 'ecome f effectiv uni 4.cherwig R64 yg 'exerppted by he 1ssion.

' ny ipf matJs needed)6 suport W(x will be by reyfest frop the NRC'and ed -

notAe incl 0ded in ttWs report. /

/ / / /

SPECIAL REPORTS .. .

MM$

6.9.2.1 Spccial reports shall be submitted within the time period specified for each report, in accordance with 10 CFR 50.4. R64 6.9.2.2 Diesel Generator Reliability Improvament Program R44 As a minimum the Reliability Improvement Program report for NRC audit, required by LC0 3.8.1.1, Table 4.8-1, shall include:

(a) a summary of all tests (valid and invalid) that occurred within the time period over which the lest 20/100 valid tests were performed l (b) analysis of fa: lures and determination of root causes of failures (c) evaluation of each of the recommendations of NUREG/CR-0660, " Enhancement of Onsite Emergency Diesel Generator Reliability in Operating Reactors,"

with respect to their application to the Plant (d) identification of all actions taken or to be taken to 1) correct the oot coures of failures defined in b) above and 2) achieve a general improvement of diesel generator reliability (e) the scheduie for implementation of each action from d) above l

(f) an assessment of the exis, ting reli.=bility of electric power to engineered-safet.y-feature equipment R1 SEQUOYAH - UNIT 2 6-22 Amendment Nos. 44, 50, 64, 66, 107, 134 NOV 1C '

INSERT C CORF OPERATING LIMITS REPORT 6.9.1.14 Core operating limits shall be established and documented in the CORE OPERATING LIMITS-REPCRT before each reload cycle or any.

remaining part of a reload cycle for the following:

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 Lirits for Specification 3/4.1.3.6,

4. Axial Flux Difference limits for Specification 3/4.2.1, S. Ileat Flux Hot Channel Factor, K(Z), and W(Z) f or Specification 3/4.2.2, and

> 6. Nuc1 car Enthalpy llot Channel Fae. tor and Power Factor Multiplier for Specification 3/4.2.3.

6. 9.1.1 ' . 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 METil000 LOGY", July 1985 (W Proprietary).

(Methodology for Specifications 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 AX1 AL OFFSET CONTROL FQ_SUEVEILLANCE TECHNICAL SPECIFICATION", JUNE 1983 (W Proprietary).

(Methodology for Specification 3.2.1 - Axial Flux dif f erence (kelaxed Axial Of f set Control) and 3.2.2 -

Heat Flux Hot Channel Factor (W(2) surveillance requi rements for Fq Methodology). )

3. WCAP-10266-P-A Rev.2, "THE 1981 REVISION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE", March 1987,- ,

(W Propriatary).

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Fac tor).

a

- . . . . _ .. _ .._ m ._ _ . . _ _ _ _. _ m _ . _ _ . _ _ _ _ _ - _ _ _ . . . . _ _ _ _ . ~ -. . _ -.. - .

_2-C0_RE CPFRATING LIMITS REPORT (continued) 6.9.1.14.b The core operating limits shall be determined sa that all applicable limits (e.g., fuel thermal-mechar.ical limits, core thermal-hydraulic limits. ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

~

6.9.1.14.c THE CORE OPERATING LIMITS REPORT shall be provic'ed within 30 days after cycle start-up (Mode 2) for each reload cycle or within 30 days of issuance of any midcycle revision to the NRC Documeret Control Cesk with copies to the Regional Administrator and Resident Inspector.

5 9

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. . . , - . . , . ~ . .

ENCLOSUhE 2 PROPOSED TECifNICAL SPECIFICATION CilANGE SEQUOYAli NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SQN-TS-91-08)

DESCRIPTION AND JUSTIFICATION F i CREATING Tile CORE OPERATING LIMITS REPORT t> -

l DesE lJLt 1on of . ChanE The proposed technical specification (TS) changes concern the relocation of several cycle-specific core' operating limits f or Sequoyah Nuclear Plant The impacted TSs i from the TSs to the Core Operating Limits Report (COLR).

will be amended to note that the limit has been relocated to the COLR, and '

a COLR paragraph will be added to the Administrative The COLR Controls Sectiontotobe will be required replace the Radial Peaking Factor Report.

submitted to NRC within 30 days after cycle start-up (Mode '2) or upon issuance of any midcycle revision to allow for continued trending of the cycle-specific parameters.

The proposed changes will reference the COLR for specific parameters and will ensure that cycle-specific parameters are maintained within the limits of the COLR. The cycle-specific parameterinclude: limits proposed for relocation to the COLR as part of this TS change

1. Moderator Temperature Coefficient

2. Shutdown Bod Insertion Limits

3. Control Rod Insertion Limits

4. Axial Flux Difference

5. Ileat Flux Ilot Channel Factor

6. Nuclear Enthalpy l{ot Channel Factor Note: A listing of each revised TS is provided in Attachment A.

Reason for Change In Generic Letter (CL) 88-16, " Removal of Cycle-Specific Parameter Limits From Technical Specifications," NRC encouraged licensees to remove certain cycle-specific parameters from the TS provided that these parameters are determined by NRC-approved metnodologies. Presently, the parameters' described above can change from cycle to cycle and would require a TS revision each cycle. By removing these certain parameters from the TS and creating a separate report (COLR) The that contains these specific values. TS COLR will replace the Radial revisions are no. longer required. This change will Peaking Factor Limit Report required by TS 6.9.1.14 result in a resource savings for both NRC-and SQN.

Justification for Change-The current TS method of controlling the above reactor physics parameters to ensure conformance to 10 CFR 50.36 (which requires the lowest functional levels acceptable for continued safe _ operation) is to specify the values determined to be w! thin the acceptance criteria using an NRC-approved calculation methodology.

The methodologies for calculating these parameters have oeen approved-by NRC.

The_ removal of cycle-dependent variables from the TS has no impact upon plant oper+. on or safety. No safety-related equipment, safety function, or plant operatin s will be altered as a . result of this proposed change.

limits will be Since applicable Updated Final Safety Analysis Report maintained, and the TSs will continue to require operation within the core-operating limits calculated by the approved methodologies, this proposed change is administrative in nature and does not aifect the purpose of the TS-Involved. Appropriate actions to be taken if the limits are violated will remain in the TSs.

.. - - ~_ ._ _ _ ~- _ __ __._.___._.-_ __ _ _- . _ _ _ ___ _ _ _

_ . . . _ . __ _ _ _ _ . . . _ _ _ . _ . . _ _ _ _ _ _ _ . .. _ - . . . . . _ _ . _ _ _ _ _ _ _ . . ~ . _

j This proposed change will control the cycle-specific parameters within the acceptance criteria and ensure conformance to 10 CFR 50.36 by using the approved methodology instead of specifying TS values. The COLR will document the specific parameter limits resulting from NRC-approved calculations, inciviing midcycle or other revisions to parameter values.

Therefore, the proposed change is in conformance with the requirements of '

10 CFR 50.36.

Any changes to the COLR will be mad.e in accordance with the requirements of 10 CFR 50.59, with a copy of the revised COLR sent to NRC as required in Section 6.9.1.14 of the TSs. From cycle to cycle, the COLR will be revised such that the appropriate core operating limits for the applicas e unit and cycle will apply. Therefore, the need to continually revise TSs for every reload is eliminated.

Environmental Ih. pact Evaluation The proposed change request does not involve an unreviewed environmental question because operation of SQN Units 1 and 2 in accordance with this change would not:

1. Result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Sta' ement (FES) as modified by the Staff's testimony to the Atomic Safety and Licensing Board, supplements -to the FES, environmental impact appraisals, or decisions of the Atomic Safety and Licensing Board.

4 2. Result in a significant change in effluents or power levels.

3. Result in matters not previously reviewed in the licensing basis for SQN that may have a significant environmental impact.

+

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ATTACilMENT A TECHNICAL SPECIFICATIONS AFFECTED IW PROPOSED MIENDMENT AND BRIEF DESCRIPT10N OF CllANGE PACE

TECilNICAL SPECIFICATION CHANGE DESCRIPTION Relocates MTC limi ; to the 3/4 1-4 3.1.1.3 Moderator Temp-crature Coefficient (MTC) Core Operating Limit Report (COLR) and corrects references.

3/4 1-5 4.1.1.3 MTC Surveillance Relocates MTC limits to the Requirements COLR and corrects references.

B3/4 1-2 3/4.1.1.3 Moderator Change deletes specific MTC Temperature Coefficient values.

Bases 3/4 1-14 3.1.3.1.c.2 Movable Change replaces Figure 3.1-1 Control Assemblies- references with Group ileight Speci f icat ion 3.1. 3.6.

3/4 1-20 3.1.3.5 Shutdawn Rad Change clarifies that the fully Insertion Limit withdrawn position for shutu annhanks is specified in the COLR, 3/4 1-21 3.1.3.6 Control Rod Change removes reference to Insertion Limit Figure 3.1-1 and relocates to COLR.

3/4 1-22 Figure 3.1-1 Rod Eank Insertion Relocates figure to COLR.

Limits Versus Thermal Power Four Loop Operation 3/4 1-23 Figure 3.1-1 Notation Revised and relocated with figure to CCLR.

3/4 2-l 3.2.1 Axial Flux Rer; aces references to Figure Dif ' erence 3.2-1 that are moved to the COLR.

3/4 2-4 Figure 3.2-1 Axial Flux Figure moved to the COLR.

Difference Limits as a Function of Rated Thermal Power 33/4 2-1 3/4.2.1 Axial Flux The Fq limit is removed and Difference Bases reference to the COLR adued.

B3/4 2-2 3/4.2.2 and 3/4.2.3 Relocates the numerical value Heat Flux and Nuclear of F6H to the COLR.

Enthalpy llot Chann' L Factors Bases

..m.___._. . _ . ~ . . . _ - . . . __ _ .__ _ _ . . . . . . . _ _ . _ . . . . _ _ _ . . . _ _ _ . . . ~ . . . _ _ _ _ _ _ _ - . _ .

l

- TijcHJil_ CAL SPEC I FICATIONS_AFFECTED BY ,PR0l ogLD AMENDMENT AND BRIEF DESCRIPTION OF Cl!ANGE PAGE* TEC11NICAL SPI:CIFICATION CilANGE DESCRIPT1nN 3/4 2-5 3.2.2 llcal Flux Ilot Relocates the Fq-and K(Z)

Channel Factor to the COLR. The numerical Fq limit isge functionF(f,placedwitha that is to be specifled 2n the COLR.

3/4 2-6L7 4.2.2.2 lleat Flux llot Same as 3.2.2 above.

Channel Factor Surveillance Requirements 3/4 2-9 Figure 3.2-2 K(Z)-Normalized Figure movea to the COLR.

Eq(Z) as a Function of Core IIelght 3/4 2-10 3.2.3 Nuclear Enthalpy llot Relocagesthenumerical value Channel Factor of FManddefines PFaj[ in the COLR.

B3/4 2-4 3/4 E.2 and 3/4.2.3 llent The Radial Peaking Factor Flux and Nuclear Enthalpy Limits Report is replaced liot Channel Factor Bases by the COLR.

6-21 6.9.1,14 Radial Peaking Replace the Radial Peaking Factor Limit Report Factor Report with the COLR.

Unit 1 pages listed only, Unit 2 will be similar.

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ENCLOSURE 3 :

.I

! PROPOSED TECIINICAL SPECIFICATION CHANGE '

SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 :

i' DOCKET NOS. 50-327 AND 50-328 l

(TVA-SQN-TS-91-05)

DETERMINATION OF HO SIG IFICANI HAZARDS CONSIDERATION 1

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-. . - - - - - - - - - - . - - - - - - - - - - . - . - - - ~

Significcnt llavards Evaluation TVA has evaluated the proposed technical upecificat ion (TS) change and has deterniined that it does not represent a rignificant hazards consideration baseo on criteria established in 10 CFR 50.92(c). Operation of sequoyah Nuclear Plant (SQN) in accordance with the proposed amendraent will not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated.

The removal of cycle-specific core operating limits from the SQN TSs has no influence or impact on the probability or consequences of any accident previously evaluated. Although not in the TSs, the core operating limits will be followed in the operation of SQN. The proposed amendment does not affect the actions to be taken when or if limits are exceeded. Each accident analysis addressed in the SQN Updated Final Safety Analysis Report will be examined with respect to changes in cycle-dependent parameters, which are obtained from the use of NRC-approved reload design methodologies. This will ensure that the transient evaluation of new reloads is bounded by previously accepted analysis. This examination, which will be performed in accordance with the requirements of 10 CFR 50.59, ensures that future reloads will not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Create the possibility of a new or-different kind of accident from any previously analyzed.

Operating SQN in accordance with the proposed change will not create the possibility of a new or different kind of accident from any previously analyzed. The removal of the specific core operating limits from the TSs does not modify safety-related equipment or ,

systems, nor does it change any safety-related setpoints used to prevent or mitigate previously analyzed accidents. The core operating licits will be defined in a separate document (COLR) from the TS and l

L will be adhered to during plant operation. Also, the limiting l

condition of operation requirements remain in effect and appropriate actions will be taken if any limits are exceeded. Therefore, the proposed amendment does not in any way create the possibility of a new or different kind of accident from any accident previously evaluated.

L 3. Involve a significant reduction in a margin of safety.

ll The margin of safety is not affected by the removal of cycle-specific l

core operating limits from the TSs. The margin of safety presently

_provided by current TSs remains unchanged. Appropriate measures exist to control the values of these cycle-specific limit _. The proposed amendment continues to require operation within the core limits as obtained from the NRC-approved reload design methodologies and appropriate actions to be taken when or if limits are viclated remain unchanged.

. ~ . . - . . . . - - _ . . . . . . - - . - . .- . -. . - . - . - - - . . - - - - . - . . _ - . - . ~. .~ . _

... .-~_ . _ .

The development of the limits for future reloads will continue to conform to those methods described in NRC-approved documentation. In addition, each future reload will involve a 10 CFR 50.59 safety review

-to ensure that operation of the unit within the cycle-specific limits will act involve a significant reduction in a margin of safety.

Therefore, the proposed changes will only move the pertinent parameters from one document to another and do not impact the operation of SQN in a manner that involves a reduction in the margin of safety.

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I ENCLUSURE 4 I

PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAll NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SQN-TS-91 -Ofs )

SAMPLE CORE OPERATING LIMITS REPORT (COLR) 8 G

1 l'

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.R-

. _ . . . . . . .. ._ _ _.__ .__ _._______ ._ _ - . ._ _ . ._.._. ....__m._ ___.__m._-- .m.

r SAMPLE SEQUOYAll NUCLEAR PLANT CORE OPERATING LIMITS REPORT REVISION A APRIL 27, 1991 Not To Be Used For Operation.

For Illustration Only Reviewed:

/

Reactor Engineering Supervisor Date

-Approved:

/

Technical Support Manager Date I

PORC Chairman Date

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SAMPLE COLR FOR SEQUOYAH UNIT ( } CYCLE { l 1.0 -CORE OPERATING LIMITS REPORT This Core Operating Limits Report (COLR) for Sequoyah Unit [ ] Cycle {]

has been prepared in accordance with the requirements of Technical Specification (TS) 6.9.1.14.

The TSs affected by this report are listed below:

3/4.1.1.3 Moderator Temperature Coefficient 3/4.1.3.5 Shutdown _' Rod Insertion Limit 3/4.1.3.6 Control: Rod Insertion Limits 3/4.2.1 Axial Flux Difference 3/4.2.2 Heat Flux Hot Channel Factor 3/4.2.3 Nuclear Enthalpy Hot Channel Factor 2.0 OPERATING LIMITS The cycle-specific parameter limits for the specifications listed in Section 1.0 are presented in the following subscutions. These limits have been developed using the NRC-approved methodologies specified in TS 6.9.1.14.

2.1 Moderator Temperature Coefficient (Specification 3/4.1.1.3)

{3/4.1.1.3] ,

2.1.1 The moderator temperature coefficient (MTC) limits are:

I The BOL/ARO/HZP-MTC shall be less positive than -0.5 Ak/k/*F.

The EOL/AR0/RTP-MTC shall be less negative than -4.0x10-4 Ak/k/'F.

2.1.2 :The 300 ppm surveillance limit is:

The measured-300 ppm /ARO/RTP-MTC should be less negative than or equal to -3.1x10-4 Ak/k/*F.

where: BOL stands for Beginning of-Cycle Life ARO stands for ALL Rods Out HZP stands for Hot Zero THERMAL POWER.

EOL stands -for- End of Cycle Lif e RTP stands for RATED THERMAL POWER Page 1 of 10 l

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SAMPLE COLR FOR SEQUOYAH UNIT [ ] CYCLE [ }

2.2 Shutdown Rod Insertion Limit (Specification 3/4.1.3.5)

[3/4.1.3.5) 2.2.1 The shutdown rods shall be withdrawn to a position as defined below:

Cycle Burnup (FND/MTU) Steps Withdrawn 1 2,000 > 226 to 1 231

> 2,000 to < 14,000 > 222 to 1 231

> 14,000 > 226 to 1 231 2.3 Control Rod Insertion Limit (Specification 3/4.1.3.6)

[3/4.1.3.6}

2.3.1 The control rod banks shall be limited in physical insettion as shown in Figure'1.

2.4 Axial Flux-Difference (Specification 3/4.2.1)

.[3/4.2.1) 2.4.1 The axial flux difference (AFD) limits are provided in Figure 2.

2.5 Heat Flux Hot Channel Factor - 9F (Z) (Specification 3/4.2.2)

[3/4.2.2]

RTP Fg .

Fg(Z) 1-

K(Z)

. for P > 0.5

P f.

I RTP Fg Fg(Z) 1

K(Z) for P 1 0.5 0.5 THERMAL POWER where P =

RATED THERMAL POWER RTP 2.5.1 Eg = 2.32 2.5.2 K(Z) is provided in Figure 3.

Page 2-of 10 .

. - - . ~ _ - . . - . ~ . . . . . . - _ . - . - - - - - - .

SAMPLE COLF FOR SEQUOYAH UNIT [ ] CYCLE [ ]

2.5.3 Note that the W(Z) values required by TS SR 4.2.2.2 are provided-in Figures 4 through 7.

N 2.6 Huc1rar_EnthalpnRisclaL.ChanacLIactor - F AH (SP ecification 3/4.2.3)

{3/4.2.3]

N RTP -

FAH I FAH * (1+PFAH * [1-P} )

THERMAL POWER where P =

RATED THERMAL POWER RTP-2.6.1 FAH = 1.55 2.6.2 -PFAH = 0.3 4

Page 3 of 10

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J._ .

m

>.e 1 A

v 2 100 - ~

.c ,

C2

% O .. BANKO' 2 $(0, 35) :

.c- <

$0 I

u-C' t _ --

t--- ~

m t- __.

tE - (0. 9.01f T 0-0 32 14 0.5 13 1.]

(FWly pq ACT:CN OF ;I A TED THC;1MA1 PCWE.9 len.ca il 56 aN / -

R00 SANK [NSERiiON LIMITS VERCUS THEAMAL POWER FCUR LOOP OPERA T Ott pa ns sa rl f Fully withdrwn sn.,ll g be the <:ondition where shutdown ant: contro l -bd.3 s are it a POSi tion wi thin the interval of 3 222 and g 231 gtep5 ,,i thdr lwn , i nc l u s i ve

/ v u. y M r o p ,C , o ra O e + ^ u- 6~ T a,'L #o :; , n * ~ 41 b ^ '~ / ~ 4 o el f uJ'J )

e , c r. a 4, .ana ,* (< Dv'

.u s p / m v. . [ ~ 's,L%ud.

6 a ca o a i- 2.1 C T' O f 2 3; 5 2 coa T O 2, /J, coa g ;,. 1_ t T' o S L31 s

( /4,aaa z 2 ' (.. TO 62.3i Pac 2 4 0 '^ '<'

, . ..-. . ~ .. - - . . . - - - . _ - . - . . . - _ - , . - . . - . . . . . . . _ - - -

. col (L Poa 5'eqas3 6a (10

r [ ] e y c <x C]

. - :: = ,= : =. ._; . . .. n .. . . . .

.._L . . . . . . . .

.-..:==--~--. _. ,. .=. : = . t_ -_. .= : . . =: .=. .

. .- . ;; ; _._---_ . . . __ . . _ . . . : n -- _ . m u . :- __ . . . r _ . - - --- . m . _ . - -

. . . . _ . _ .... g ' !K -

_g.

. - -. . . - ..___.r..--_=._=_=.=.-.m. ., -_=.=..._:__=_t.___--_=._:_.__:_.

=.. - - -

-< a - - - -

== m. _:

a. .. ,;p-o; :a

_ . . y_ _.

- - - (-1L,:00). . . . _ _:*r ~

" : ( 6 .100 ) _._

100 - I- -

- - - - I $ -\

[ bi}[CCIPTIBLE ! ~ 'UNNCC5PTABLE E'-

y/

~

l OPERATICN OPERATION

? \E -

- I V

- : 1-80

.j -s

___=_

1 s

f 1: \

q, k __.-

/.s. ACCEPTABLE

,f - OPERATION'-

~

~ \ s

% 60 ; .\

g ___ / - 's 0 -" \

. . . .--. i s

=_ =. =:i (- 31,50 )

(20,50)--

s . 6% -

l 4m' ._

t 9Q .

i- _. _

ae. N

~ ene

. e .

T..._-

20 I .g.'___ _

,c ..g

..._____.c

-=: . _ _ _ _ .

T; . ,;5. ._._ .:= =_=. . =.=.:.._._3. _ _

.. . _ _ . 3._...._. z.:._..

.---- _-. _. ._ _...... . . = 5. . .:;;:= .._ _,

0

-30 -20 -10 0 10. 20- 30 40 50

-50' -40 Flux Difference (al):

4 a

h dd d .,,,

-I AXIAL FLUX OIFFERENCE LIMITS As A FUNCT'CN OF RATED THERMAL POWER f/)c;c _ 5 of /0 .

1.20 1.10 -

1.00 - -

g 0.90 - r LU E S 0.80 - U h

h a

O O_

ta h

g-

0.70 -- >

O Total Peaking Factor 2 o O 0.60 - . 2.32 h [ '

. 0.50 -- Core Height K(Z)

N 0 2 0'40 -

0.000 1.000 '

j T 6 000 1.000 }

O A 1 z 0.30 -

' U' 8

z 12.000 0.925 y R 0.20 -- *i 3

1 0.10 -- q; 0.00 - . , i 1 i i i i i i i D' O.0 2.0 4.0 6.0 8.0 10.0 12 0 9 x '

CORE HEIGHT (FEET) 9 A

5604< 3 K(Z) - flormalized Eq(Z) as a Function of Core lleight LJ

. - ~.

d o' R. FoA M9 u-opt um r [ - ] _ e yew [ J I

1.50 i . . . . . au iso

. t i e . ; ; a i

. . i a a i j e HeI9h1 M AX e e , s .

~

_a e : i i e e bD -WTl

00 1.0000 I*45

i e . e e e i e

O.0000 1,00o0

. i i e . , ,

0,2000 3,0p90 e i i e

.e i. ,~~,g , t , , ,

O.4000 1,0000 i i i i

0,6000 1,0000

, , ~7* , , , , , , , , , , ,

0,8000 1,00o0

e i a e

I a e i e 'a"i i i , '

1.0000 1,0000

- * * * *

a a e i i e i e i e

1.2000 1,0000 e e i 4 e s l # t t i l g i i

1.4000 1.0000

, , i , , , , , , j , , , , ,

1.0000 1.0000 1.8000 1.3231 J ' s * *

i I i i 1 I I t 2,0000 1.2980

* * ' ' ' ' I 3 8 8 2.20CD 1,2724 g*33 . .

i i e 4 4 e i i e i 2,4000 1.2511 e a i i s t ; i t i t : 2.6000 1.2321

. . e i ; i e i i e 2.8000 1.2121 F4 -- --

I ' '

3.0000 1.1927

' ' 8 I I I I ! 3.2000 1,1785 ss

.' i +

i e s. I e i i e t ' 3.4000 1.1729

yl*30 s . e i e i 1 i e i , e i . 3.6000' 1.1724 ,

i , , , 3,8000 1 1728 y< ,

e

, a i , l l 4,0000 1.1719

.g

. *, . . . i . , e i l 4,2000 1.1700 ,

- * * -i *

i i t i 4 e e 4,4000 i,qg73

. - 4

e ie i i a 4- e *

4.6000 1.1637 l'1.25 a . , , , , , e , , , , 4.8000 1.i5s8 1.1537 C) , , , , , , , ; ; , , 3 , , 5.0000

, . 5.2000 1,1467

+-

I 1 ion

' i* I '

5,4000 1,1510

>- +

Q:

' * .' I 6 L's s'* 8 I i s

' ' ' 5,G000 5,8000 1.1634 1.1754

< i ' e t i + i i '

-sl.20 ., , , i. . . .. . ^ , -

, , e,0000 1.1863

S , , .

, , 9 , , * , ( , , 6.2000 1.1362

-) ' '

6.4000 1.2047 y) ~' .***o ' ' ' I 6.6000 1.2117

4*e t

e i 1 * ' I 8 8 8.8000 1.2172

. . e i . I e 7.0000 1,2209 1.15 , ... ..i. . i , 7.2000 1.2227

, , j,, , , 7.4000 1.2224

-1.2137 7,6000

8 I ' ' ' '

7.8000 1.2144 y l

'l 4 i i

i i 8 ' 8,0000 1.2070 i<i.i : e a e e 8.2000 1.19G8 1.10 , , , .

8,4000 1,1809

' ' ' 8.6000 1.1688

'. ' ' ' ' ' ' ' ' 8.8000 1.1703

' ' ' ' ' ' ' ' ' ' ' ' 9.0000 1.1801 e i i e i a i s i i a 8 1 i

9.2000 1.1905 6 i e s e 6 e . i a e 9.4000 1.1992 1.05 , , , , , , , . , , , , s.6000 i 2078 i e e i 9.8000 1.2249

. 4 e i s-i i i 6 j__, t 10.000 1,25G7 i t e . 4 e i l i i i i e i * ' 10.200 1,2903

< 1 l 6 . i e i i i I i I 4 I t *

10,400 1.0000

, . ,. .,,i i i a e .

10,600 1.0000 1.00

10.800 1.0000 0 2 4 5 8 K) 12

it,000 i.0000

": '1, 0000 Crrr" " 'H'ClGHT

- ' I F""~

\ T 1i TOP

=

11.400 BO T'TO M e 11.600 '1.0000 11.800 1.0000

12.000 1.0000

FIGURE y i

RAOC

SUMMARY

OF MAX W(Z) AT 150 MWD /MTU

TCP AND 00iTOM 1* *. E.XCt.UDED AS PER TECH SPEC 4.2.2.2.G i-PSC 7 0" /0 i

- . . . ~ . . -- . . . .-. .. . . .. - - _ _ . _ . ~ . . .

+ Ok k Fo(2 @ AH _-()n t r- [' ] C yceq (, j l

< e a 1

1.M - . . i i . i au 4000

a i I a # # ._f .

i #

l

. . i j t i , , , bet 9ht M Ax j : i i U2' s i

e e t e t 4 l 8 i ' ' ' i 6 8 ' 8

O.0000 1,oooo

{g ,i 1 e i . . . . i

O.2000 1,oooo i . i i i i + i e i

O.4000 1 oooo 6 i e

O.6000 1,00o0 i , j g e , , [--

l

O.8000 1,oooo

* * *

I I i i i a i . 1,0000 8 i 1.0000 1

i i a l *

i i e

1.2000 i.ooOO 1 40 t 6 i a a .. s a e i , e i i

1.4000 1,0000

1.6000 1,0000 l e j i e i i ii I l l 6 . l l 1.8000 1.233g t i I e i i e a i } l e i a 2.0000 1.2787 I I i 8 i s i e e a e t I i t i 6 2.2000 1.2570 i i e i i e i e is i i 4 e a . e 2.4000 1.2355

,1.35 , , , . , , , , , , , , , i 2.6 COO 1.2130 t a l I l a a 4 2.8000 1.1906 i l 1 e #

3.0000 1.1750 es i i i i I I I i 1 3.2000 1.1672 gj i t i 8 8 e i e i i e i s i a i i t i e a 3,4000 1,1o43 g,

. e i i . '

i 4 a i 6 i e e 1 a e 3.6000 1.1623 gla I 3.8000* 1.1596 i i e' . . s.e i e s e i . e 4,0000 1.1559 x , t , , , , , , , l ; ; ~{ 1- , , #

4.2000 1.1508 I 1 i e i I ' i I 4.4000 1.1469 a

i i i i 4 l 4

i

. i 8 4 e i :

1 I le**, i i

i e 4.6000 1.1458 a . 8 i a i i t iei i 9 . 6 i e 4.8000 1.1426 L1.25 , , , . , , .. , , ,, , , , , 5.0000 i.i418 O , , , , , , , , , , , , , , , 5.2000 1.1431

, , , 5.4000 1,1623

.e i i i t i i i i 5.6000 1.1782

). e i i I i '

e t 8

i s i

8 i 1 ' 8 ** i 8 5.8000 1.1945 Ct e 8

<C . + # . > a . a # e . i e e i 6.0000 1.2096 '

e.2OOO i.2232 s1.20 , , , , , , , , , , , . , , , ,

6.4000 1.2353 M

' ' ' ' ' ' 1.2457 6.5000 5

y) 8 > ' ' * * '.

8 ' 8 6

* ' 8.8000 1.2543 e i 6 .io_ i I a t I 1 e i i e 7.0000 1.2608

.

i ' s i e i . . i 7.2000 1.2851 t .i.**

e a 1.15 . , ,

, , , , , , , 7.4000 1.266s 7.6000 1.2661 i i I I I + 1 1.2624 8 I i i t . i 7.8000 i t a B i 8 8 8 ' ' 8,0000 1.2558 a , a s :

i : 4 i i t 6 8.2000 1.2460

, . : i . . i i i e i i 8.4000 1.2332 8.6000 1.216s 1.10 , , , , , , , , '

8.8000 1.1999

I a i i + i e i i e I a i i i I 9.0000 1.1932 I i 4 e i

i i e i a 4 i i

i t i g,2000 1.1965 t i t i i 6 i i i I e ' i i 9,4000 1.2044 i e 1.2154

, e . . , , , , e e i , , e , , 9.6000 1.05 , , , , , , . , , s.8000 1.2276 10.000 1.2333 i t 4

i i t i 1.2512 e e i I e i i i e t 10.200 I 6 e t ' ' I i i i e i e ! ' ' 8

10.400 1.0000 t I I i i i i i e 10.600 1,0000 t i 1 6 e i e i i t

, , , , . , , , , , , , , i i , , i

10.800 1.0000

,

11.000 1.0000

!.00 3 10 12

21.200 1,0000 0 2 4 6 C0^"i "'lGHT (FEET) Toe lP!88 E8888 acrrrou

12.000 1,0000 FK3URE {s j P e

RAOC

SUMMARY

CF MAX W(Z) AT 4000 MWO/MTU

TCP Af40 BOTTOM 15*. EXCLUCED AS PER TECH SPEC 4.2.2.2.G gwc 8 ef m

- -- --- . ., _ ._ . _ _ _ - ._~s. . - _ _ _ _ _ _ _ ,,

U N YC0 WE G LA O} $ $+

(h g g- j ( l e .

l 130 e . . s . l e i , a a i : , , g i . BU 10000

'

1 6

i a i i . I 6 '

Height WAy 8 8 i + i i

4 i e i e 8 i (Feett _ wy; i i *

I . i s i a e g*g e

O.0000 1.00o0 I ) i i i t i i , ,

O.2000

' i 1,oooo t e i i e i e : i , i , , ; e i

O.4000 1,oooo

' i i i 8 i 6 I t I i I I e i i i

O.0000 1,Oooo i I i i i < t I i e i e s i e i

O.8000 1,ooon

i i i e i e i e i i , ,

1.0000 1,ovoo i e i e

1.2000 1,oooo i . i , e i e i , , , ,

e i i e

- : .

1.4000 1,0o00 I ' ' ' ' ' ' ' ! ' I I ' 3 8 8 8 8

1.6000 1.oo00 i ' 8 1 i e I ! i ( i i i t i I i 1.8000 1.2376

' i t i t i e i i ; e , , , , 2.0000 1.2255

. . . i i ; i , , , , , 2.2000 1,2140 1.204s .r a 4 e t

-f 1.35 i

2.4000 .

. , i , , , , , 2.6000 1,1373

8 ' ' ' 8 ' I I I ' i i i 8 ! L_ 2.3000 1.18ts fs 8 8 8 8 1 ' 8 i i 8 I I I I i ! l 340000 1.17gO pq e i i 8 I 6 I I t ) : 8 3.300C 1.17*7 1.1750 8

ss i a i . . i i, , , , , , , 3.4000 y1.30 a e 3.6000 1.1804

. ' , , , ' ' 8 13' I e I I I ' 3.8000 .. f.1862 i i i a 4,0000 1,1933 X i ' i i 8 I i l i. 1 6 4 i t i i 8 ' . I ! I t i e a i e i . (.2000 1.2001 i i j , '

6 i j g , 4.4000 1,2070

e +

e i i a i l 4.6000 1.2145

. .. e. A i a i , , i i , ,

4.8000 i.22cs L1.25 ' ' , '

__ _ _ ' i e i i i f i i i e i 5.0000 1.223.:

C) ' *

i i i e i je i I e i e a 5.2000 1.2254

' 1 ,I I t *

i .e* I e i e I t i i i e i 5.4000 1.238:

>- a 6 .

, , , 5,6000 1.2566 i a le . . . i a e i 6 1 , ,

Q: e -

5.8000 1,2703 o , : , , . * , ,

. 4 . e 6.0000 t.28i6 s

<1.20

-e

' ' ' ' ' ' ' ' ' ' i ' 6.2000 1,2911 4 . ' ' i

l, i i e i i l' i ") i 6.4000 3,;g79

6.6000 1.3023 y *

,* . I

, i e

e' e i i

i e.,*

i e i l

}

6 , , 6.8000 1.3040

. . i i i

, e r

i e , ( , , ,

7.0000 1.3023

. i g ,

7.2000 1.2390 I.l$

I ' ' i ' l i

6 i i 7,4000 1,;g;O 8 8 I i i 8 6 I i

I e i 7.6000 1.2820

: I i e ( i i e i e 7.8000 1.2688

. 6

, , , j # , . 8.0000 1.2528

. I ;~ g .

. i

, , , g , , j ,

8.2000 1.2339

, . g .

8 4000

. i.2iO1 1.10 ' . I e . . i e , e i e 8.6000 1.1880

4

i ' i I . I i t i i i i + 8.8000 1.1753 i 4 . . i p i i e s e t . . i , i 9.0000 1.17C6 i 3.2000 1.1721 i . i i e e i ; i i i i i 1 e i e s 9.4000 1.1774 e . .

i . i i i i i I i i e i i . 9.6000 1.1832 z$ i . i I . e i I . . i i i i i i i e i e . 9.8000 1.1882 i e i I i e i 1 i i 6 I i i 3 e i i 10.000 1.1938 i a e i i i e i a 10.200 1.2034 e i a i i i i e i e

10.400 1.0000

~~

, , , , , , ,

10.600 1.0000

' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' '10.800 1.0000 00 p' = 11.000 1.0o00 0 2 4 6 8 10

11.200 1.0000 co i i Of.1 CORE HEIGHT (c-i) rtt .OP i

ii.400

It.sOO

11.8C0 1.0000 1.0000 1.0000

12.000 1.0000 FIGURE 6 L

RACC

SUMMARY

OF MAX W(Z) AT 10000 MWO/MTU

TOP ANO BOTTOM 15% EXCt.tJCED AS Pc3 TECH SPEC 4 -' ? G

& &E 9 O/= /O

. - - , . . _ . _ - - - - -. . , - . - . - =.--~ - . . - ~ - c- .. . - -- - . -

. V0 tV U L j U U' 6 >+~ G U O j /k g.] G ( C L et {

1.50 '

u_L '

i i .

eu i4ooo I i i e e t : i e e i s s s e belght wg t # s e a e i i e

t 4 i i 1 , (Feett _g,x7g 1 i i . e i I ' ' ' ' ' I ' ' ' ' ' ' e 0.0000 j,4$ ' '

1.0000 8 * ' i * '

I '

O.2000 g,oooo i 4 e i _. I e .

t i e a i e

O.4000 1 oooo t I i l 4 ,

O.0000 1,oooo 1 i ~

i i e 1

O.8000 1,oooo

< e 1 ,

j i i i . ( . t l r # #

1.0000 1.oooo

. i ( , i -i j ,

i , i g , i

2 2ooo 1.0000 I*40 i I i 1 1 ' t ! ' ' ' '

1.4000 1.oooo

! 8 8 '

I i l i e f ij i _a i i 1 e e i ! l

1 0000 1,oooo 6 6 I e a e il e _ i i i e i i i 1.8000 1.25G1 g , , 2.0000 1,2419

, , i ; i j s " s'3 { g 2.2000 1,2275 d ooo - < 1 2129 1*35 f i t - i i e i

a i, I 8 I e s 2.6000' 1.1980 e i e i i i i eo e i I, i i e t i i 2.8000 1.1C37

e s i e e i . i l l t s # t i 3.0000 1.1728 s i 3,2000 1.1717 P4 , , a . g i , , ; g , , , , ,

se 3.4000 1.1825

' ' ' ' ' ' I ' ' ' ' ' ' ' ' .6 t.1938

1*30 e # 6 6 i i e t * . s i 4 i

I t i 3.8000 1.2c=2

>< a e e i e i l 6 a

i I *i 4 , a e a '

4.0000 1.1)( 1 i +4 e i 6 e i e i l t 4.2000 1.2311

<C i e i i 4.4000 1.2409 22 ; i e i i e e i e i t i i i i i 4.6000 1.2488 I + i i

e. # 4 4 a 4.8000 1.2526

. . e i .T 4

- i i . I e I t i i i . ' e i 5.0000 1.2590 O i , e i I . t 6 6 . t i e 5.2000 1.2752

e

i r

. i

. I i i s' i e i a i 5.4000 1.2965

>" i e i ,

i ' ,' ' 5.6000 1.3175 I si ' ' ' I l ' I 8 5.8000 1.3349 Cf e e'

i e e 6 e *i

4 a 6,0000 1.3494

<,20 3i ' .

a i

, , , ,, , , , , , , . . , , , e,2 coa 1,3si2 Zi .

.. g . , , , , .* ,

. , , 8.4000 1.3G97 i 6,6000 1.3750

!) . , , , , , , * , , ,

, , , , '^ 6.8000 1.3769 W i e i : e i i e i e i a e i 7.0000 1.3753 i a e 8 e a i 4 a 1 e i e i- ' 7,2cco 1,37o0 lN . . i e i + 6 . i e i e i 7.4000 1.3G09

. I e i e 6 i e i i . 7.4000 1.3480 i e 7.8000 1.3310 i , , , , ,

i i . i i . , i 8,0000 1.3105 i e . 1 - i I .* I e i 8.2000 1.28G4

i * ' i ' '

8.4000 1.2571 1.10 8

,' _j , , , ,' .' . . 8.s000 1.2284 e i . , , . # . . a a r i , , 8.8000 1.2076 i e =

9.0000 f.19G1 t t i i e i e i 6 .

I a 9.2000 1.1809 i i i i i e i e a i i i

9,4 con j,j7gg e

s,

4 i > e i e a i . e i 6

' ' 9.6000 1.1815 1.05 , , , , , , , , ,

,' , , , s.8000 i.19se i e i e i i t i i e i i 10.000 1.20G8 s e i e 10.200 1.2180 i . i i t i # I e i i i i i t

10.4C0 1.0000 i e i t i e i =

t i i i i t i i j i ' = 10.000 1,0000 e i i i .

I e i a a i i ' '

10.800 1.0000 1,00

'. .20a

'. 4 6 8 10 P. 31 i, coco 0

"- 1 oooo c0RE

HEIGHT (FEET) TOP

1 1. s'co30 1.0000 EOTTOM

11,800 1.0000

12.000 1.0000 FIGURE 7 RACC

SUMMARY

OF MAX W(Z) AT 14000 MWO/MTU i'

l-

TCP ANO BOTTOM 15% EXCLUCCD AS PE.R TECH SPEC 422JLG l

l S7W /0 & /0

~ ~ ~ , .__

ML20091B251

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= :=:_::.

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ML20091B251

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