Text

Proposed Tech Spec Supporting Use of Vantage 5H Fuel W/ Intermediate Flow Mixers Beginning in Cycle 7
	ML20116A868
Person / Time
Site:	Wolf Creek
Issue date:	10/28/1992
From:	; WOLF CREEK NUCLEAR OPERATING CORP.
To:	;
Shared Package
ML20116A856	List: ML20116A853; ML20116A859; ML20116A868; ML20116A873;
References
	NUDOCS 9210300190
	Download: ML20116A868 (72)
	v • d • e

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Appendix A 1

Technical Specification Changes / Markups I

9210300190 921028 j PDR ADOCK 05000482

, P PDR

' 5- ** -

CEFIN!t!0ns E LMI 1.0 DEFINITIONS i

1.1 ACT!04........................................................ 11 1.2 ACTUATION LOGIC TEST.......................................... 11 1.3 ANALOG CHANNEL OPEAATIONAL T[57...............................

1-1

1. 4 Ax ! A L F L U X 0 ! F F E MMC E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 .

1.5 CHANN E L CA LI B RATI ON. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 '

1. 6 C HANN E L CM C K. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11

1. 7 CONTAINMENT INTEGRITY......................................... 12

1. 8 CONTROLLED LEAKAGE............................................

12

1. 9,p CORE ALTERATION........... g .g p. g ,.p.g.g fg . 12 I.it -die- DOS E EQUIVALINT ! 131. . . . . . . . . . . . . . . . . . . . . . . .......... . . . . . . . gg .

12 I.12. -Iru- l- AVERAGE O!$!NTEGRATION ENERGY. . . . . . . . . . . . . . . . . . . . . . .12 .......

f .I3 4rit- ENGINEE RED SAFETY FEATURES RESPONSE TIME. . . . . . . . . . . . . . . . . . . . . 1-3 Ny 4r14- FREQUENCY NOTAT!0N...........................................

1-3 ur 4rl+ IDENTIFIED LEAKAGE........................................... 1-3

( 4 -1r15- MAS T E R RE LAY Tt37. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1 17 4rie MDSE R( 5) 0F THE PUBLIC. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 IM -irlf- 0FF5 !TE DOSE CALCULATION M4NUAL. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 I p; -trie- OP ERA 8 LE - OPERA 8 ! LITY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14

/.20 4r14- O P E RATIONAL MDDE - N00(. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . < 14 i il 4r40 Pb s!CS TEST 5................................................ 1-4

/ M -irti- PRES $URE 300MBARY LEARAGE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4

(.7 14rae P ROCES$ CONTROL P908Aldt. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 l.t4 -ir H PURGE PO48!MS....................... ...................... 1-4

/.2i 4,44. QUADRANT POWtt TI LT RAT!0. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 g 6 4-f5- RATED THf W#L P0WER. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15

/.27 4r46 REACTOR TRIP $YSTEM RisP0NSE TIME............................ 15 l.M -irf7- R EPottMLI EYENT. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 1114r44 SHUTDAC MRAt1N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 l .3o 4rt9- $ 1 TE 80LNGARY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 t.3l 4r36- S LAvi RE LAY Tt5T. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 b

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

&[t h0LF CREEK UNIT 1 1 Amerwhent No. 42 i

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-SAF@TV LIMITS AND L!MITING SAFETY SYSTEM SETTINGS SECT!O'd pact 21 SAFETY LIMITS 0.1.1 REACTOR C0RE................ ... ........................... 21 2.1.2 REACTOR COOLANT SYSTEM PRESSURE............................. 2-1 l

c[gj$I ~!;UC 2.1 1 0:ACT00 C00 CAICTY LIMIT "000 LOOPS l" OPERATIO" . 2- 0

2. 2 LIMITING SArETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0!NTS............... 2-3 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP 5ETPOINTS.... 2-4 BASES SECTION PAGE ,

2.1 SAFETY LIMITS -

2.1.1 REACTOR C0RE................................................ B 2-1 2.1.2 REACTOR COO LANT SYSTEM PRES $URE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM IN5TRUMENTATION SETP0lNTS............... B 2 o i -

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

L9f'!NG CON 0!T!CNS roe opt At!ON AND SURVE!LLANCE RECUIRffsNTS i ECT!C1 pace ,

1/,4,0 apDLICA8!(!TY............................................... 3/4 0 - , -!

1(4;1 REACTIVITY CONTROL $YSTEMS e

3/4.1.1 80 RATION CCNTROL I Shutdown Margin - T avg

> 200'F........................... 3/4 1-1 Shutdown Margin - T,yg i 200'F...........................

3/4 1 3 Moderator Temperature Coefficient........................ 3/4 1-4 Minimum Temperature for Criticality...................... 3/4 1-6 3/4.1.2 BORATION SYSTEMS i Flow Path - Shutdown..................................... 3/4 1 7 Flow Paths Operating................................... 3/4 1 8 Charging Pump - Shutdown................................. 3/4 1-9 '

Charging Pumps - operating............................... 3/4 1-10 Scrated Water-Source - Shutdown.......................... 3/4 1 11 Borated Water Sources - Operating........................ 3/4 1-12 +

./4.1.3 MOVABLE CONTROL ASSEM8 LIES Group Height............................................. 3/4 1-14 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-LENGTH 3 R00......................................... 3/4 1-15 :

Position Indication Systems - Operating.................. 3/4.1 17 Position Indication Systen - Shutdown.. ................. 3/4 1-13 Rod Orop Time............................................ 3/4-1-19 Shutdown Rod Insertion Limit............................. 3/4 1-20 Control Rod Insertion Limits........................ .... 3 /4 1 .".

FIGURE 3.1-1 R00.8ANK INSERTION LIMITS VERSUS THERMAL .

POWER-FOUR LOCP 0PERATION............. .. 3/4 1-201 .

ba I<+<.

l aC.? CREEK - UNIT 1 IV

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i!N79tNG Cao!9?oa5 r0s 0* ERAT!ON ANO sylvt!LLAMCE REQU!aEMEwf 5

$ECT!0N - F, (g y,2) eact 3/4.7 E CI5TRIBUTION LIMITS l 3/_4. 2.1 AX1AL FLUX O!FFERENCE..... ...u ........................ 3/4 2 1 a og /I!

- JURE 3.21 AXIALFLUXO!FFERENCELINITf'ASAFUNCTION0[

a TED TMERMat RowtR... a . ............... u. a i

3/4.2.2 HEATFLUXNOTCHANNELFACT0R..$.......................... 3/4 2 4

~

/il.dl dIGURE 3.2 2 ((Z) N0RM.LIZED F;(Z) A5 A FUNCTION 0F CORE %. _

3/4 2 5 3/4.2.3 /E ge ,5 rLOW RATE AM(PNUCLEAR ENTHALPY R!$E NOT CHA u4 2..

in 12" ' '9&'rMuM"?". ': 7. . . . . . . . . . . . . . . . . u4 2.,

3/4.2.4 QUADRANT POWEit T!LT T10................................ 3/4 2 11 t 3/4.2.5 DNS PAMMETERS. . . . . . . . . . ............................... 3/4 2-14 '

TABLE 3.2 1 DNS PAAANETER5........... ........................... 3/4 2-15 3/4.3 !NSTRUMENTATION -F69 (x, y')

3/4.3.1 REACTOR TRIP $YSTEM INSTRUMENTATION...................... 3/4 3-1 TA8LE 3.fl REACTOR TRIP $YSTEM INSTRUNENTAT!0N................... 3/43-2 Te4LE 3,3-2 REACTOR TRIP SYSTEM INSTRL8ENTATION RESPONSE TIME 5.... 3/4 3-7 fA8LE 4.3-1 REACTOR TRIP SYSTEM INSTRLMENTATION SUWI!LUW8CE REQUIREMNT5........................................ 3/4 3 9 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION........................................ 3/4313 TA4LE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION $YSTEM INSTRUMENTATION..................................... 3/4 3 34 TA8LE 3.3-4 ENGINEERG SAFETY FEATURES ACTUATION SYSTEM INSTRtaWNTATION TRIP SETP0!NTS...................... 3/4 3-22 TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE T11E5............. 3/4 3-29 TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRLDENTATION SURVE!LLAMCE REQUIREENTS. . . . . . . . . . . 3/4 3-34 l

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WLF CREEK UNIT 1 Y

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SECTION paGE 6.5.2 NUCLIAR SAFFTY REV!EV Costi1TTEE (NSRC)

Function.................................... ................. 69 Composition................................................... 6-10 Alternates.................................................... 6 10 Consultants................................................... 5 10 Meeting Frequency..... 4

...................................... 6 10 l

Quorum........................................................ 6-10 i Review........................................................ 6-11 1

Audits........................................................ 6-11 '

rec 0rds....................................................... 6*12 6.6 REPORTABLE EVENT ACT!0N............................................ 6 13 6.7 SAFETY LIMIT V!0LATION............................................. 6-13 t 6.8 PROCE00kES AND Pe0 GRAMS............................................ 6-13 6.)

--- REPORTI.NG. AE00!RF__, N,,T,[

0.9.1 ROUTINE RLPU4T5............................................... 6-18 S t a r t up R epo rt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 19 [

Annual Reports................................................ 6-19 Annual Radiological Environmental Operating Report............ 6-20 ;

Semiannual Radiocctive Eff1 cant Reler.se Peport................ 6-20 Monthly Operating Report...................... 1.............. 6 20

, M ' 9 "d W-W ' t " ; : -t . . . . . . . . .. . . . . . . . . . . . . . . . . . . , 6-21 6.9.2 SPECIAL REPORT 5............................................... 6-21 6.10 REC,0R0 RETDrT10N............... .................................. 6-21 Co re ogevdiq LimdT Report WOLF CREEK - UNIT 1 XXI Amendment No. 42

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S.E! NIT!CNS.

CONTAIhM?,NT INTEGRITY.

. 1. 7 CONTAINMENT INTEGRITY shall exist when:

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

1) Capaele of being closed by an OPERABLE containment automatic isolation valve system, or

2) Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed cositions, except as orovicec in Table 3.6-1 of Specification 3.6.3.

b. All equipment hatches are closed and sealed,

c. Each air lock is in compliance witt the requirements of Specification 3.6.1.3,

d. The containment leakage rates are within the liMts of Specification

3.6.1.2, and

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

bellows, or 0-rings) is OPERA 8LE.

CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow from the reactor coolar:

pump seals. ,

CCRE ALTERATION, 1.9 CORE ALTERATION shall be the movement or manipulation of any component-within the reactor vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

/AMtf 00SE EQUIVALENT I-131 Il 1.)6 00SE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / -s,)

wnich alone would produce the same thyroid dose as the quantity and isotopic mixture of I-1X . I-132, I-133, I-134, and I-135 4ctually present. The tnyroi:

-dose conversios' factors used for this calculation shall be those listed in Table III of TfD-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

? - AVERAGE DISINTEGRATION ENERGY

1. I shall be the average (weighted in proportion to the concentratien :f eacn radionuclice in the reactor coolant at the time of sampling) of the 5-7 of tne average beta and gamma energies per disintegration (in MeV) far iso:::es, other than iocines, with half-lives greater tnan 15 minutes, making up at 235:

95'.' of the total noniodine activity in the coolar.t.

WOLF CREEK - UNIT 1 1-2 L

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INSERT (Add to page 1-2)

CORE OPERATING LIMITS REPORT 1.10 The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determi.ed for each reload cycle in accordance with Specification 6.9.1.9. Plant operation within these operating limits is addressed in individual Specifications.

l

OEF7N!??6NS ENGINEERED SAFETY FEATURES RFSPONSE TIME 1.h The ENGIMERED SAFETY FEATURES (ESF) RESPONSE TIMC shall be t interval free when the monitored parameter exceeds its ESF Actuation Setpoint at the channel senser until the ESF equipment is capable of performing-its- -

safety function (i.e. , the valves travel-to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.

FREQUENCY NOTATION

1. h The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

10ENTIFIEL LEAKAGE f

1.)(( IDENTIFIED LEAKAGE shall oe:

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 800NOARY LEAKAGE, or

c. Reactor Coolant Systes leakage through a steam generator to the Secondary Coolant System.

MASTER RELAY TEST ob

1. g A MASTER RELAY TEST shall be the energization of each anster relay and verification of OPERASILITY of each relay. The MASTER RELAY TEST shall include a continuity check of each associated slave relay.

MEMBER (S) 0F THE PUBLIC 1.h MEMBER (S) 0F THE PUBLIC shall include all persons who are not-l occupationally associated with the plant. This category does-not incluoe

employees of the licensee, its contractors, or vendors. Also excluded from-this category are persens who entar the site to service equipment or to make

[

deliveries, This category does include persons who use portions of the site t for recreattenal, occup4tional, or other purposes not associated with the plant.

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WOLF CREEK - UNIT 1 1-3 l

CEFfMITIONS i

0FFSITE 005E CALCULATION ." "JAL and parameters used in the calculation of offsite doses re active gaseous and Ifquid. effluents, in the calculation of gaseous and 11wid effluent Radiological monitoring Monitoring Alana/ Trip 5etpoints, and in the conduct of the. Environmental Progras.

The 00CM shall also contain (1) the Radioac- l tive Ef fluent Controls and Radiological Environmental Monitoring Progrees - i required by Section 6.4.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Sesiannual Radio- ,

active Effluent _ Release Reports required by Specifications 6.9.1.5 and 6.9.1.7.

OPERA 8LE - OPERA 8!LITY 1..M A system, subsystem, train, cocoonent or device shall be OPERA 8LE or have OPERASILITY when it is capable of performing its specified function (s) and wnen all necessary attendant instrumentation, controls, _ electrical power..

cooling or seal water, lubrication or other auxiliary equipment that are - ,

required for the systas, subsystas, train, component, or device to perfom its function (s) are also capable of performing their related support function (s).

OPERATICKAL 2 0E - N00E L 1.if An OptRATIONAL M00E (i.e., MODE) shall correspond-to any one inclusive combination of core reactivity coMition, power level, and average reactor coolant temperature specified in Table 1.2.

PHYSICS TESTS

1. d PHYSICS TESTS shall be those tests performed to esasure the fundamental nuclear characteristics of the core and related instrumentation:

Chapter 14.0 of the F5AR, or (2) authorized under the previsions o(1) f 10described in CFR 50.59,.

or (3) otherwise approved by the Commission.

PRESSURE BOUNDARY L m me 1.)[ PRES $URESOUNDANYLEAKAGEshallbelaakage(exceptsteamgeneratortube leakage) body, pipe through wall, ora vessel nonisolable ws11.fault in a Reacter Coelant System component PROCESS CONTROL pana m 1.jNThePROCESSCONTROLPROGRAM(PCP)shallcontainthecurrentformulas, sasoling, analyses, test, and determinations.to be sede to ensureLthat processing and packaging of solid radioactive wastes based on demonstrated processing of actual er_ simulated wet solid westes will be accomplished in such a way as te assure compItance with 10 CFR Parta 20, 61, and 71,'Stata regulations, disposal burtalredithetive of solid ground requirementa, wasta. and other requirementa governing the PutGE - PURGING

1. 5 PURGE or PURGING shall be any controlled process.of discharging air or-

gas. from a confinement to maintain- temperature, pressure, humidity, concentration, or_ other operating l condition, in such a manner that replacement air or gas is required to purify the confinement. "

'4LF CREEK - UNIT 1 1-4 Amendment No. 42 w -

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DU2 NIT 20NS QUADRANT POWER TILT RATIO 14 1.4+ QUADRANT POWER TILT RATIO shall be the ratio of the maxima upper encore detector calibrated output to the average of the upper excore detector cali-brated outputs, or the ratio of the maximum lower excore detector calibrated output to the nerage of the lower excore detector calibrated outputs, whichevar is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.

RATED THERMAL POWER 24 1.45- RATED THERMAL POWER shall be a total core heat transfer rate to the reactor coolant of 3411 Wt.

REACTOR TRIP $_YSTEM RESPONSE TIME 27 1.4(r- The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval free when the monitored parameter exceeds its Trip setpoint at the channel sensor until' loss of stationary gripper coil voltage.

REPORTA8LE EVENT l 22 1.e A REPORTA4LE EVENT shall be any of those cond'tions ' Mc.ifled in Section 50.73 to 10 CFR Part 50.

SHUTOOWN MARGIN 1.h SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be- subcritical free 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 30 1.49 The $!TE SOUNDAAY shall be that line beyond which the land is neither owned, nor leased, nor othenvise controlled by the licensee.

SLAVE RELAY TEST

1. A SLAVE RELAY TEST shall be the energization of each slave relay and verification of OPEAA4!LITY of each relay. - The SLAVE RELAY TEST shall include a c.- inuity check, as a ninimum, of associated testable actuation devices.

l WOLF CREEK - UNIT 1 1-5 knen@mnt No. 42 1

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2.0 SAFETY-LIMfTS AND LfMITfNG SAFETY SYSTEM SETTfNGS-2.1 SAFETY-LIMITS REACTOR CORE-2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest ia operating loop coolant temperature (T,yg) shall not exceed the limits 5 + a r4" 2.1-1 for four loop operation.

N APPLICABILITY: MODES 1 and 2. ff*c'ft04ic +ke CollE CP# N P "A tovert e r Pc otr ( cet.a.) .

ACTION:

Whenever the point defined by the combination of the highest operating loop ,

average temperature and THERMAL POWER has exceeded the appropriate pres-surizer pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the requirements of Specification 6.7.1.

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

APPLICABILITY: MODES 1, 2, 3, 4, and 5.

ACTION:

H0 DES 1 and 2:

be Whenever the Reactor in HOT STAN0BY with Coolant the Reactor Systes pressure R olant Systemhaspressure exceeded 2735its-l within psig,imit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the requirements of Specification 6.7.1, MODES 3, 4,'and 5:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its li it within 5 minutes, and comply with the requirements of Specification 6.7.1.

l l-l WOLF CREEK - UNIT 1 2-1 l ~

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O.0 0.1 .2 0.3 0.4 0.5 0.6 0.7 0.5 .9 1.0 1.1 1.2 FRACTION OF RATED THERWAL l

1 FIGURE 1.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION 2-2 Amendment No. 51

/0LFCREEK-UNIT 1

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TABLE 2.2-1 g

5 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SEIPOINTS 9

g; SENSOR

^ TOTAL ERROR FUNCTIONAL' UNIT ALLOWANCE (TA) I (5) TRIP SETPOINT ALLOWABLE VALUE ,

c '

g 1. Manual Reactor Trip N.A. N.A. N.A. N.A. N.A. ,

~ 2. Power Range, Neutron Flux

!a. High Setpoint 7.5 4.56 0 1109% of RTP* 1112.3% of RTP*

b. Low Setpoint 8. 3 4.56 0 $25% of RTP* 128.3% of RTP*

3. Power Range, Neutron Flux, 2.4 0.5 0 <4% of RTP* with <6.3% of RfP* with-High Positive Rate i time constant i time. constant 12 seconds 12 seconds

4. ' Power Range, Neutron Flux, 2.4 0.5 0 <4% of.RTP* with <6.3% of RTP* with m 5tigh Negative Rate . i time constant i time constant a 12 seconds 12 seconds ,

5. Intermediate Range, 17.0 8.41 0 125% of RTP* $35.3% of RTP*

Neutron Flux

6. Source Range, Neutron Flux 17.0 10.01 0 $105 cps $1.6 x 105 cps.

1, 3 r // M1

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7. Overtemperature AT ,h t- 5&- A-77 See Note 1 See Note 2 '

2.93 o.M

&S- MI,0

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8. Overpower AT 4:e3 .eM See Note 3 See Note 4 :l_

9. Pressurizer Pressure-Low 3.7 0.71 2.49 11915 psig 11906 psig I

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10. Pressurizer Pressure-High 7.5 0.71 2.49 12385 psig <2400 psig

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11. ' Pressurizer Water Level-High 8.0. 2.18 1.96 192% of instrument <93.9% of instrument-span- span

. *RTP = RATED THERMAL. POWER m

- Loop design flow = 93,75 gpa 93,/,cic>

.w N

^

g TABLE 2.2-1 (Continued) r-

TABLE NOTATIONS O

.A NOTE 1: DVERIEMPERATURE AT

. I' II * b) I '

3 c-x AT (1 +1)1:II * **S) (1 + 135)2<(1ATo + 165 (Ki-K

) [T (1 + 1a5) - T'] + K (P - P') - f 1(AI))

-4

= Measured AT; Where: AT 1 I = Lead-lag compensator on measured AT; 3 t 4

ti, ta = Time constants utilized in lead-lag compensator for AT, ti = gs ,

12 = 3 s;

'I = Lag compensator on measured AT; 7 5 2

~

4 13

= Time constant utilized in the lag compensator for AT, 13 =//s;

= Indicated AT at RATED THERMAL POWER; AT, K-i = 1.10; K2

= 0.0137/*F;

[

I{ = The function generated by the lead-lag compensator for T,,9 dynamic compensation; g 14, is = Time constants utilized in the lead-lag compensator for T,,,,1. =gs, is = 4 s; a

z P T = Average temperature. *F;

=

-1

165 ag'c g ensator..on measured I,,g; ti. = Iise constant utilized in;the measured I,,g lag compensator, is = 0 s;

,1-4 TABLE 2.2-1 (Continued)

E IA8tE NOTAT10NS'(ContinuedJ m NOTE.1: (Continued) n I' $ .588.5'F (Nominal T,,,!at RATED.lHERNAL POWER);

$ 'Ka .

= 0.000671;

" = Pressurizer pressure, psig;.

P'

- P' = '2235 psig (Nominal RCS operating pressure);

5- = Laplace transfem operator, s 1;

.--and3f (AI) is a function'of-the.. indicated difforence between top and bottem detectors:of.the'.

power-range neutron-lon chambers; with gains to be selected based on measured instrument response'during plant STARTUP tests such that:

-4S%

y ;(i) for qtol % -between K a M .+. R ,.f (AI) = 0, where qt and g are percent

DIED THEiglAL POWER in'the top and bottee halves.of the core respectively, and qt * % IS

- total THERMAL POWER in. percent of; RATED THEIBIAL POWER; 25%

(ii)- .for each percent that the magnltudeg g g' exceeds --NE; the. AT Trip' Setpoint -

shall be automatically reduced by-1-59N of its'value at RATED TERIAL PohER; and -

(iii) for each percent-that the magnitgggg .g exceeds +M, the AT-Trip Setpoint' k.

o

= shall be automatically reduced by4.G of its value at RATED THENIAL POWER.

k R .NDIE 2i- ' The channel's maximum Trip. Setpoint.shall. not exceed its. computed Trip.Setpoint by more than I' g -2 M of AT span.

1.4 g g. ~.

-,ao- ng r ~ ,e w , _ e ,

_rs:- -+w- + m

IABLE 2.2-1 (Continued) 6 G TABLE NOTATIONS (Continued) n HOTE 3: OVERPOWER AT I - Ks [T f l f r as -I AT + 1 fI 5) $ AT,(K 4 - Ksy f 'ft25 1 1e5

- f(AU) 2 z

Z

" Where: AT = Measured AT; I

f* = Lead-lag compensator on wasured AT; (o

ti, 12 = Time constants utilized in lead-lag compensator for AT, 1 =[s.52 = 3 5;

= Lag compensator on measured AT; I + T35 1

13

= Ilse constant utilized in the lag compensator for AT, t2=/s; 7

e ,

AT, = Indicated AT at RATED TEERMAL POWER; K4 =M

Ks

0.02/*F for increasing average temperature and 0 for decreasing everage temperature; y[7 125

The function generated by the rate-lag compensator for T,yg dynamic k compensation; 17

= Time constant utilized in the rate-lag compensator for Tavg' = l0 'A 1

1 =

1+156 Lag c spensator on measured T,yg; C is

=

Time constant utilized in the measured I,,g lag compensator, is = 0 s; i

1

TABLE 2.2-1 (Continued)

E TABLE NOTATIONS (Continted)

O p NOTE 3: (Continued) x ,

K. = 0.00128/*F for T > I" and K4 = 0 for T < T";

E q T = Average tesserature, *F; I" =

Indicated I,,g at RATED INEIMAL POWER (Calibration temperature for AT instrumentation, < 588.5'F);

5 = Laplace transform operator, s 1; and f (al) = 0 for all Al.

-NOTE 4: The channel's maximum Trip Setpo'nt i shall not exceed its computed Trip Setpoint by more than y 4.7% ef AT span. l E' /.47%

n -

.E e

as SpeiW t & COM 2.1 -SAFETY LINITS o Pda mr %rs Noe r Nou.)

BASES 2.- l .1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of t fuel and possible cladding perforation which would result in the release of ission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regi where the heat transfer coefficient is large and the cladding surface tempe ature is slightly above the coolant saturation temperature.-

Operation above the upper boundary of the nucleate boiling egime could result in excessive cladding temperatures because of the onset f departure from nucleate boiling (DN8) and.the resultant. sharp reduction- 1 heat transfer-coefficient. DN8 is not a directly measurable parameter durin operation-and therefore THERMAL POWER and Reactor Coolant Temperature and P ssure have been-related to DN8 through DNBR ccrrelations. DNBR correlations ave been developed-to predict the DNB flux and the.)ocation of DN8 for axially nifore and-nonuniform heat flux distributions. . The local DNS heat flu ratio (DNBR) is defined as the ratio of the heat flux that would cause DN8 t a particular core location to the local heat flux, and is-indicative of the rgin to DN8.

The DN8 design basis is as follows: there must be a least-a 95 percent probability that the minimus DN8R of the limiting rod dur ng Condition I and-II events is greater than or equal to the DN8R limit of DN8 correlation 5:f:; 2 :d (th: "- -I :: :!: tie: '- S!: -!!::tten).

limit is established based on the entire ap;plicable experimental data . set such-The correlation DN8R that there is a 95 percent probability with 95 percent confidence that DN8-will not occur when the minimum-DN8R is at the'DN8R limit (1.17 ':r th: "" M

rr;!:thn). For plant conditions which fall outsida the range of applicability of theg g correlation the W-3 correlation is used.

In addi ion, DN8 margin is maintained by performing safety analyses to a higher val than the-correlation limit, called the-safety analysis limit-DNBR. The margin between the safety analysisilimit DNBR and the correlation L limit DN8R is used to cover known DN8R penalties and provide margin for design flexibility.ne saMY aulpii h1 bNddt / 5 94CY'd IM*4# C81N*

The curves '$I;b.1-1 show the loci of points of THERML POWER,

- Reactor Coolant 5ystes pressure ^$'and average temperature for which.the minimum DNBR is no less than tho' applicable safety analysis lisit DN8R, or the average [ ,

enthalpy at the vessel exit is equal to'the enthalpy of saturated 1_iquid.:

4e deq Fu <rece0*J idt These curves are based on :: ::th_! g y t d ::::! *::ter, F ,s *c' 1.55 c t-and a reference cosine with a peak of 1.55 for axial power shape. t ! hi:n;;

( 2. . <. n..a.a

. ~.

a

_-- ... .. a..

.. . . . ........ ...-s. eN. . y .

. ..__ x ..>. ,- .. u-- .> . . n . ,- . . .- >. .- . . :

T" 4 1.55 [1: ^ 1 (! "M .

i "h:x " i: it: ' ::li: . ' "".T:" T:::""t/. ."~ "".

l L faendment No. 23, 51-

_W OLF CREEK - UNIT 1 -B 2-1 l l

_ . = _ _ _ _ _ _

LIMITING SAFETY SYSTEM SETTINGS BhSES Intermediate and Source Range. Neutron Flux The Intermediate and Source Range, Neutron Flux trips provide core protection curing reactor startup to sitigate the consequences of an un-controlled condition.

red cluster control assembly bank withdrawal from a subcritical These trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron Flux channels. The Source Range channels will initiate a Reactor trip at about 105 counts per second unless manually blocked when P-6 becomes active. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active.

Overtemoerature AT The Overtemperature si trip for all combinations of pressum,provides core protection to prevent DN8 power, coolant temperature, and axiel power distribution, provided that the tr&nsient is slow with respect to piping transit delays from the core to the temperature detectors and prsssure is within the range between the Pressurizer High and Low Pressure trips. The Setpoint is automatically varied with: (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water and includes cynamic compensation for piping dela cetectors, (2) pressurizar pressure,ysand from(3) the corepower axial to the distribution.

loop temperature With nomal axial power distribution, core Scfety Lfait as shown i this Reactor trip limit is always below the gm 2.1-1, If axial peaks are greater than design, as indicated by the di erence between top and bottom power range nuclear detectors, the acacto trip is automatically reduced according to the notations in Table 2.2:1. + g c, o g g, Overoowe AT The Overpower ai Reactor trip provides assurance of fuel integrity l (e.g. , no fuel pellet salting and less than 1% cladding strain) under all possible overpowr conditions, limits the required range for Overtemperatura AT trip, and providas a backup to the High Neutron Flux trip. The Setpoint is automatically varied with: (1) coolant temperature to correct for tescerature induced changoes.in density and heat capacity of water, and (2) rate of change of temperatures.for dynamic compensation for piping delays from the core to the loco temperature detectors, to ensure that the allowable heat generation rate (kW/ft) is not' exceeded. The Overpower ai trip provides protection to mitigate the consequeness of various size steam breaks as reported in WCAP-9226, " Reactor Core Response to Excessive Secondary Steam Releases."

l WOLF CREEK - UNIT 1 3 2-5 Amendment No. 52 l

a: E:.: : 0:*, R L S v 5' EMS-3a11 E :;':0N CON'ROL

.: . O NN s' A R G I N

-av;

.:W -~.23 :0NDITION :CR OPERATION 3.1.1.1 e SWT004N MAR 3IN shall be greater tnan or equal to 1.3*, sk/t fer

<:.r 'ceo c:eration.

.::'E!.:~#: MODES 1, 2", 3, d4, udf

.:"!ON:

W1:n the SHUT 00WN MARGIN less than 1.3% ok/k, immediately initiate and continue

ration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or equivalent until the required SHUTDOWN MARG lN is restored.

SURVEILLANCE REOUIREMENTS 4.1.1.1.1 The SHUTCOWN MARGIN shall be determined to be greater than or equal to 1.3*,sk/k:

a. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after detection of an inopeP3ble Control rod (s) and t; least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod (s) is inoperable.

If the inoperable control rod is immovable or untrippable, the- acove required SHUTDOWN MARGIN shall be ve-ified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s);

b.

When in MODE 1 or MODE 2 with K,ff greater than or equal to 1 at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6;

c. When in MODE 2 with K gf less than 1, within 4 hou:'s prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6;

d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of Specification 4.1.1.1.le. below, with the control banks at the maximum insertion limit of Specification 3.1.3.6; and "See Special Test Exception Specification 3.10.1.

WOLF CREEK-- UNIT 1 3/4 1-1

REACTIVITY CCNTROL SYSTEMS SURVEILLANCE REOUIREMENTS (Continued) i e, wheninMODE3,r/4,[atleastonceper24hoursbyconsicerationof the following factors:

1) Reactor Coolant System boron concentration,

2) Control rod position,

3) Reactor Coolant System average tempxrature,

4) Fuel burnup based on gross thermal energy generation,

5) Xenon concentration, and

6) Samarium concentration.

4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement witnin 2 1% ak/k at least once per 31 Effective Full Power Days (EFPD). This comparison shall consider at least those factors stated in Specification 4.1.1.1.le. above. The predicted reactivity values snall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 EFP0 af ter each fuel loading.

OL CREEA - UNIT 1 3/4 1-2

E::-* : - ::.*::. ? S?!"5 5 ,, ; <!'. UtRON- < 200':

g- }c

'. : e * *. * * :. :0N::': N :; c RATION '

s i

1" *-a t ';* :WN MAR 3 /

N shall be greater than or equal to 1% rp[k.

. .. e...: . . , ng.:. ..:

,/ o

'

:'. / -

/

a't t e 5%;T.SCaW MARGIN less than 1% ak/k, immediately int (iate anc cor.;inue

-ati:n at gre'ater inan or eaval to 30 gpm of a solutioq/containing greater ina er e:.al to 7000 ppm boron or equivalent until thefrequired SHUTDOWN MA: 2:N is restored. /

/

SU vE:.'.ANCE REQUIREMENTS /

\

A. l. l. 2 The SHUTDOWN MARGIN shall be deter 'ned to be greater than or equal to 1% ak/k: ',

3 a.

Within I hour af ter detectiony tf an inoperable control rnd(s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months sthereafter while the rod (s) is inoperacle, If the inoperable control p'd is immovable or untrippable, the SHUTDOWN MARGIN shall be 4erified acceptable with an increased allowance for the withd awn worth of the immovable or untrippable control rec (s); and s

0.

At least once per 2- hours by consideration of the following actors:

/

1) Reactor Coolant System boron coacentration,

2) Control .r/od position, !

s

3) Reactor Coolant System average temperature, A) Fuel burnup based on gross thermal energy generation,

5) Xenon concentration, and

6) Samarium concentration. .

\

/

/ s

\

-Oil CREEK - UNIT 1 3/4 1-3

42ACTIV!TY CONTROL SYSTEMS (CDERATOR TEMPERATURE COEFFICIENT A E*L U#if*P'cifi'd & % C*t' O f tfA11 n G Mu ur IteEPottT (COLf)-

LIMITING CONDITION FOR OPERATION I

3.1.1. 3 The moderator temperature coefficient (MTC) shall be:

a. Less positive than ^ 2E'r for the all rods withdrawn, ::eginning of cycle life (BOL), 'et re e THERMAL POWER condition..,w.-

b. Less negative than l'tke. Gol limit specibed in ne coLK

~

1: 10 ' ik/S/** for the all rods witnerawn, end of cycle life (EOL), RATED THERMAL POWER condition.

,.- BOL Lim a APPLICABILITY: 4 5;::f 'f::ti;r, 0.1.1. 2e. - MODES 1 and 2#" .

< p::ifi::ti;r. 3.1.1.05. - MODES 1, 2, and 3#.

ACTION: Ob bi" ROL- g pecifred in St C00

a. With the MTC more positive than the limitg e4--Speci fi::W- 3.1.1. h-above, operation in MODES 1 and 2 m proceed proviced:

wifkin 1.

Control red withdrawal limits are established and maintair.ed sufficient to restore the MTC to ?::: ;;;iu s: :nr _

" the 20L 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 nours. sq These withdrawal limits shall-be in addition to tne insertion limits of Specification 3.1.3.6; s pe d 4 ..

Ik A m8

2. The control rods are maintained within the witacrawal 'mu -

established above until a subsequent calculation vert fies tnat the MTC has been restored to within its limit for tne s r::s withdrawn condition; and

3. A Special Report is prepared and submitted to the Commission pursuant to. Specification 6.9.2 within 10 days, .descrf::ing tne value of the measured MTC, the interim control od witncrawal limits, and the predicted average core burnup necessary for restoring the positive MTC to within its limit for une all rec 3 withdrawn condition.

60L.. SPaci('00 \n %. C&Lk b.

With the MTC more above,beinHOT' SHUTnegative than the limit)c' SedW:'

00hNwithin12iours ": . K k.

"Witn Xg f greater than or equal-to 1.

See Special Test Exceptien Specification 3.10.3.

WOLF CREEK - UNIT 1 3/4 1-4

E A:T:v:n C r ;;L 5<5TEMS i

SUtvE!LLANCE RE00: 0.E.v E NT S 4.1.1.3 The MTC shall be determinec to be witnin its limits : urin; each fuel cycle as fcllows:

het zen > awk 3gg ,s g, g

a. The MTC shall be measurec and compared to the BOL timit ef 3 ::i'i::-

t r ? _1.1. I1, a5cve, prior to initial operation aSove AiA of RATED i

THERMAL POWER, after each fuel loading; anc

. ine MTC shall be measured at any THERMAL POWER an: ::-:are: to y 2.0 - 10 ' 14/k/ F (all rocs witharawn, RATED THERNAL POWER conci-

/ tion) within 7 EFPD af ter reaching an equilibrium boron concen-

[ tration of 300 ppm. In the event this comparison incicates tne MTC i s more negative than -0. 2 - 10 ' ak,'k/ F, the MTC shall be remeasure:,

anc comoarec to the EOL MTC limit,cf Op;;i'ic;ti;.- 2.1.1. 0 0. - a t -

least once per 14 F"PD uring the remainder of the fuel cycle.

\

-f ke 3 00 PPM S u rveilla nce Lr mif S pecihed i n M e Co LR WOLF CREEK - UNIT 1 3/4 1-5

REACTIVITY CONTROL M TO'S 3/4.1.3 MOVAILE CONTROL ASSEMBLICS GROUP HEIGHT LIMITI M C0WOITION FOR OPERATION 3.1. 3.1 All full-length shutdown and control rods shall be OPERABLE and positioned within 2 12 steps (indicated position) of their group step counter demand position.

APPLICABILIH: MODES la and 2",

ACTION: The ACTION to be taken is based on the cause of inoperability of control rods as follows:

ACTION More Than CAUSE OF IN0PERASILITY 3ne Rod One Rod a) Innovable as a result of excessive (1) (1) friction or nochanical interference or known to be untrippable.

b) Misaligned froe its group step (3) (2) counter demand height or from any i other rod in its group by more than t 12 steps (indicated position).

c) Inoperable due to a rod control urgent (4) (4) failure alars or other electrical probles in the rod control system, _

but trippable.

ACTION 1 - Determine that the SHUTDOWN MAK3!N requirement 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 STAM06Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 2 - Se in HOT stale 8Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

> ACTION 3 .794R OPERATION say continue provided that within 1 hour:

I. The rod is restored to OPERA 8LE status within the above alignment requiresents, or

2. The rod is declared inoperable and the remainder of the rods in the group with the inoperable rod are aligned to within

12 steps of the inoperable rod while maintaining the rod sequence and insertion limits o N yr; :.1-1. The THERf%L POWER 1evel shall be re ricted pursuant to Specifica-tion 3.1.3.6 during subsequent ration, or Setci(icrkon 3.l.3. fo .

"See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

WOLF CREEK - UNIT 1 3/4 1-14 Amendment No. 27, M l

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

MAC*IVfrY C0 $ROL SYSTEMS Sf.iT00WN R00'fNSER?f0N LIM 2T I LIMITING CONDITION FOR_0PERATIch h m iei se pi p a esl <+sertow a s spreono,is % e. conc 3.1.3.5 All shutdown rods shall be fully * *~ ra or n m ut, p u nS gefocT (cotR).

A APPLICABILITY: MODES-1* and 2*#.

{,.u er;} keyond 4e inserlior' U$ s Pe CIVtb in 4 e COLRJ aith a maximum of one shutdown rod n t fe!!y eithd' un, except for surveillance testing pursuant to Specification 2.1.3.1.2, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either:

Pencre 4e. cod n nitkin ,

4. UllyrithdPS th0"Od, 6 M c If.St riio r lim d 5fe?l ifd 'e Of C.0LA,OY~

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

SURVEILLANCE REQUIREMENTS

% b e. w i th s -+ k e (+ s e.r% I i n '. f s f e c i$td s e % CoLK :

4.1.3.5 Each shutdown rod shall be determined fully =f ther-A

a. Within 15 minutes prior to withdrawal of any rods _in Control Bank 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 Specifications 3.10.2-and 3.10.3.

With K,ff greater than or equal to 1.

WOLF CREEK - UNIT 1 3/4 1-20 l.

REACTIVITY CONTROL SYSTEMS CONTA0L 200 INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as & - da M ;= ? 1-1.

speci&ed in %s cms opsunn Lon Ws REPoa r (cos.2) .

APPLICA8ILITY: MODES la and 2*#.

ACTION:

d d 4 % COL 8.

With the control banks inserted beyond the above-insertion limits, except for surveillance testing pursuant to Specification 4.1.3.1.2: A

a. Restore the control banks to within the limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or

b. Reduce THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank position using theGiert-' - "[o'[-IIEts 5 Pe C ifit d .'* M E COLb 0F

c. Be in at least HOT STAN08Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVE!LLANCE 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 Specifications 3.10.2 and 3.10.3.

With K,ff greater than or equal to 1.

l' l

WOLF CREEK - UNIT 1- 3/4 1-21

. DELETED (FULLY WTTNDRAWNO )

ESA [(38.Y%'12l1 I (74.7 % 122.) _

too ,

/, ./, '

N g>W 4X 3 - r

/

Y /

j f i

[l \

. (o% 14th 2

[ /

(100% 141)j 40

\ / '

/

" 18 / / /

\ /.Au O / /

tto

\ gr / -

/

"i - r

/

/' l }

/ N / /

/ \ / /

/ \ / / .ANx o

- / \ //

/ N'/

' (Q% 44) [j\

i( '8

\

/ \

// \

(30.2% 0)(( \

0 to 40 60 80 00 0 (FULLY INSERTED) l I

'\

FIGURE 3.1-1 \

\

ROO SANK !NSERTION LIMIT VERSUS \

THERMAL POWER-FOUR LOOP OPERATION \

+F lly Withdrawn shall be the condition where conttd rods are at a position Mithin the interval of n 222 and :s 231 steps withdrawn,

/ Amendment No.32 WOLF CREEK - UNIT I 3/4 1-22 l

l

3/4 h Dr~Ei O:s?5:su*fcN Ltw!TS s,pe e4 e{ is A e '.ea r 3/4.2.1 3 RIAL rLUX OIFFEAENCE-(AFD) oppg4ri,sk M HtT1 R E fevi (COLR) .-

t:u! TING CONDITION-FOR' OPERATION- "

3.2.1 The indt'ated- AXIAL- FLUX OIFFERENCE (AFD)be-maintainec shall witnin tne allowed operational space ref4 t ty ri;u : 3.2-;.y

oLICABIL:Tv:

M00E'1 above 50 PERCENT RATED- THE".:'.AL POWER".

ACT:CN:

With the indicated AFD outside of the E!;ur: 2.2-1spe cife'el in +Le cout [ .

a.

limits,-

1.

Either restore the indicated AFD'to within the rig. : 2.21*- , l limits within

'N

15. minutes -or-Sg.ci6ed iw &ke cot R i 2.

Reduce THERMAL POWER to less- than' 50% of RATED TH POWER within 30 minutes and reduce the Power Range Neut*:n Flux - High Trio Setootnts to less than or equal to 5E% :f 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: RATE unlesstheindicatedAFDiswithintheFf;;r;2.2-1-limitp.p i

srecificj- in the coLR l-t I

l-1 1-1 l'

l "See-50eclai Test Exception 3.10.2.

a I WOLF CREEK - UNIT 1- 3/4 2 '

Amencment No. -

_-__-_--_---.__--__--m_

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u. .:..= .:: : _- =, ..u . ; " : _ .--. :_="--: ::.t ~__ =. ~. . . _. _. ..; .. .= .. :. :..;,. .....

:t :=-

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

... (- 15, t 0 0) .- ( 6,10 0 :. :.: . .). . .. . . .

1C0 ....f=._=-

m. . _ . . .

. + . . . _ . _ _ . . .- . . . t . .,r_. : . . ._ -=-

- .: ..a. =__ =.._. __ _ .

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_ _ . . . - :c. .... ... .a.: _.:-.. _-

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g _ . .

i_ -

e . . - . . .. :.: = v- . . .

._. _...t.. .. . _ . ::~-.: .-:=..-. - -

C* * ..__

._. . _ _. _ - . . . _:" :$f = 1

a

_ .T * ="3. *

= __ .._. ...- ... i..

2 .

.. . - : r r .- .:- .

==m c - -

i ...._

w 60 _v __

_. .. ACCEPTABLE o pg R A T10 N ...un: =. . =.z; .... -: =s p y . _. _ _.. -.g 2

c ..

_=

W w

(-30,80T ~~

4 -

-- (28.80) _ __ 'Jr c

g, 40 .

_-_ . = :.: .= _ _ ..,.,,.m. ,3.jjg:

O -

=

_ ==t .

~~

vt

_ ====4

.e 0 -

I

_ =

2 50

-30 -10 0 10 3 80 FLUX DIFFERENCE ( A1) S 4

s FIGURE 3.2-1 WOLF CREEK UNIT 1 L

AX1AL FLUX D FERENCE LIMITS AS A FUNCTION OF MATED THEhMA POWER s.

N Ns

. w.: C?_f.- W: 1 3/c : 3 A." SM .en e.1

.wn

_ . . . . . . . . . . . . - - + -

~

PCw!R D:574!!UI!CN Lfut75 -

3/4.2"?- HEAT FLUX HOT CHANNEL CACIOR - A N v*---

~

L:4! TING-CON 07 TION FOR OPERATION F(y,Y,1) z 9

3.2.2 p shall be limit:d by the following relationships:

RTP Fg ^(x,gr.) J @ i ( ) CKCZ)) for P > 0.5, and F 4

([4f,Z) ~g(2 }- 1 ( ) [K(Z)] for P 5 0.5.- F -

Whe re:

05 nyst,tr THERMAL POWER i.

p , RATED THERMAL power ' and K(Z) = the fu eti-nor=ltze) Fetri(xj s) limit as a furche olr ed ' e- H ge e ? ?-? #"

core heignt-!cratiaa as ? C'>eS

) s p e cif(ed (n th e C ol.R .

ADSLICABILITY: HODE 1.

AC'!CN:

Fq (x,%I)

With p exceeding its limit: ~

mag

a. Reduce THERMAL POWER a least 1% for eacn n p exceeds :.e (i :: ,M '

within-15 minutes and similarly reduce the Power nee Neut--.

4g Flux-Hich Trio Setooints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; OkER OPERAT:t

~

3 _ may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsecuent POWER OPERATION-may proceed provided the Overpower AT Trip 5et:oints have been reduced at least 1% for each 1% fq-f24- exceeds the li-ity C.

4 ' Y' d #. Identify and correct the cause of the out-of-limit condition :ri:r to increasing ACTION a., THERMAL POWER above the reduced limit requirec :y aoove; THERMAL POYER may then ce increased provice:

is comonstrated through incere mapping to be within its_ limit. W W [vgz) s'aC'5 CREEK - us 7 ;

3/4 2 4

~_

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

t

" INSERT 1 (Add to page 3/4 2r.)

FgMA(X,Y,Z)

= thg(measured Fg X,Y,Z), increased heat-fluxby-3%

hot-channel factor, to account for manufacturing-tolerances and further increased by 5%

to account for measurement uncertainty, RTP Fq = the Fg' Limit at RATED THERMAL POWER (RTP) ,

as specified in the CORE OPERATING LIMITS REPORT (COLR),

INSERT 2 (Add to page 3/4 2-4)

b. Control the - AFD to within new AFD limits which are determined by reducing the allowable THERMAL POWER at each point along the AFD ligtt lines of Specification 3.2.1 at least 1% for each'1% Fo (X,Y,Z) exc3eds the limit:within: 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and declare the AFD monitor' alarm inoperable until the AFD alarm' setpoints are reset to the modified limits; and.

l b

l l

p l

i I

l 1

l

' C L->t o

i ,i , i

_! . . . . . i _ .L_. .

i . l .i. t. . i i i !

, i

, , i -

j l t l , -

i I

. t, , .i .

l

. . . - . . . . _ _ . - l- - ' . . ,'... ..

4 i i } l 1 i

I ., .e ,

l . l' .. . . . . . . ....

1

/.'.;.. . ;

, a I- - I 2 5 l-1

,i , l i -

. . . . . . . .1 . . . . .

1 -

o i

~

.- : i I' ;-

l 0

5 I .o v l

l [ l ~

e .

_ _ .- I, .J l

'" 3 i t t

l . - e i w v

. .. %, .e 1' t i : C l i . N.

i . l . o_ m e w

, = ,

o i -

i - w c: e o.

+

I

- l t

. yo e

a-I i . -

.i

. Ni i i ~

i

=

1.

f . . .l -

, d

,I i ,

1 y i ,I l . ..i ..

+

, ,=

. I

. =

t 3 . .

i i . . . .

+

' ..-...-.....L..

' . ]> -

. 8 l l

\

~ ~

l

.. L 4 .

x

.-.--..u' i i t ' '

c. . o

- Sib< , o , c. o en o o' ei (Z)Dj g37,5yggog _gggy

[

l l

l i

! WOLF CRED - UNIT 1 3/4 2-5 l

! i s

l POWER ')!5it!!UTION ' !w!?$ l SURVE!L_ LANCE REQUIREMENT $

hX,Y, Z) h l/,(, Z) 4 2.2.1 The ovisions of Specification 4.0.4 are t applicable. !

r

y- shall be evaluated to determine if QF (!) is within its Ifm1t Oy:

4.2.2.2 !

a. Using the mova010 incore detaci; ors to obtain a power distribu' ion !

map at any THERMAL POWER greater than 5% of RATED THERMAL PC'ER !

t. :ncreeefag the-mou ured e Ay re-acaeat c' the ce'er distairet4:a - -

20 -

by 3% to account for manufacturing tolerances anc furtner increasirg 1 ne value by 5% to account for measurement uncertainties; 7

. Com ring the F,y computed (F0 ) obtained in specification 4.2,2 .. l acove o:

1) The limits for RATED THERMAL POWER (FRTP) 7,7 gn, ,,,7,,,,,t, j xy core planes given in Specifications 4.-2 2.2e. a,c f. .

measure below, and Q 2) The relationshi ,

b L W F oy = FRTP xy (1 .2(1-P)] l 4 ,

D Where F l

i s the limit r Practionti TPU MAL P0'aER ocertt'c- i A "Y expressed as a function o Rip me., P ' t ce fractic of :M E:

I i

THERMAL POWER at .hich F xy wa measw*tc.

d. Remeasuring F xy according to the follow g schecule: l g

i

1) When F C

xy-is greater than the F Axy limit r tne appropriite c.

measured core plane but less than the F ( r ationsnio, ac:itt: a l >

xy power distribution maps shall be taken and F,y empareo ts :9 3 i -

L and F xy either:

L a) Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ifter exceeding by 20% of RATES ,

i fFERMALPOWERorgreP,er,theTHERMALPCWERaton:n~l was last determined, or ItHtaf ri; :.:r :::er: ";5- - -% -.

O At ':::t :::: p r 21 PO, -

b

'a0LF CREE ( _ UNIT 1 3/4-2 6

-e*-Es- r--=-- ..%w.r,- ..-,m.%-u...,w..--m,e-w e.,- ,,.e,_..,a, .em,.. % --, nee n --%,e,-,w.-w,,,,w,---.-%-.. . , , ,%gm, y..-.,--..,y, .y y 3 wrem W,%, ,-4

1 INSERT p.1 of 3 (Add to page 3/4 2-6)

M

b. Measuring FQ (X,Y,Z) at the earliest of:

1. At least once per 31 Effective Full Power Days, or

2. of RATED THERMAL POWER After exceeding the THERMAp POWERby 20% or morg(X,Y,Z) at which Pg was last determined ;

c. satisfying the relationship presented in Specification 3.2.2;

d. Satisfying the following relationship:

FgM (X,Y,Z) 1 (Fg(X,Y,Z))HOM where, (Fo(X,Y,Z)) NOM represents the nominal design power distribution increased by an allowance for the expected deviation between the nominal design power distribution and the msasurement and is specified in the COLR.

If the above relationship is not satisfied, then for that location perform the following:

1. Calculate the % margin to the maximum allowable design as follows:

FgM (X,t,Z)

100

% Operational Margin = ( 1 - -----b(X,Y,Z))OP (Fg )

% Reactor Protection Fp'M ( X , Y , Z )

Setpoint (RPS)

Margin

=( 1 - ----(Fg b(X,Y,Z))RPS) 100 l

l l

During power escalation at the beginning of each cycle, THFRMAL POWER may be increated until a power level for extended operation has been achieved and a power distribution may be obtained.

INSERT p.7 of 3 (Add to page 3/4 2-6) where, [Fgb(X,Y,Z?)0P and (Fgb(X,Y,Z))RPS are the operational and RFS design paaking limits and are specified in the COLR.

2. Find the minimum operational Margin of all locations examined in 4.2.2.2.d.1, above. If the minimum margin is less than 0, EITHER of the following actions shall be taken:

a. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , control the AFD to within new '

AFD limits that are determined by:

Reduced negative AFD Limit =

The negative AFD Limit in Specification 3.2.1 plus the absolute value of the quantity (op Mar NSLOPE

Minimum Operational Margin),

Reduced positive AFD Limit =

The positive AFD Limit in Specification 3.2.1 minus the absolute value of the quantity (op Mar PSLOPE

Minimum Operational Margin),

where, the op Mar NSLOPE and op Mar PSLOPE are specified in the COLR, and declare the AFD monitor alarm inoperable until the AFD alarm setpoints are modified to the limits of 4.2.2.2.d.2.a, or

b. Comply with the ACTION requirements of Specification 3.2.2, treatingthemarginviolagonin4.2.2.2.d.1, above, as the amount by which Fg (X,Y,Z) is exceeding its limit.

3. Find the minimum RPS margin of all locations examined in 4.2.2.2.d.1, above. If the minimum margin in-less than 0, the following action shall be taken:

Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , reduce the negative f i (AI) limit and the positive ft (AI) limit of the OTAT as follows:

INSERT p.3 of 3 (Add to page 3/4 2-6)

Reduced negative ft(AI) Limit =

ft(AI) of Table 2.2-1 plus the absolute value of the quantity (the RPS Mar NSLOPE

Minimum RPS Margin),

Reduced positive f 1 (AI) Limit a f t(AI) of Table 2.2-1 minus the absolute value of the quantity (the RPS Mar PSLOPE

Minimum RPS Margin),

where, RPS Mar NSLOPE and RPS Mar PSLOPE are specified in the COLR.

e. The limita in Specification 4.2.2.2.d are not applicable in the following core plane regions as measured in porcent of corc height from the bottom of the fuelt

1. Lower core region from 0 to 15%, inclusive,

2. Upper core region from 85 to 100%, inclusive,

3. Crid Plane Regions, and

4. Core plane regions within +/- 2% of core height

(+/- 2.88 inchee) about the bank demand position of the Bank "D" control rods.

~

D E LETE D ;

NPOWER DISTRIBUTION t!MITS FILLANCE REQUIREMENTS (Continued) 1 N

limit to the

) WhentheFfy is less tnan or equal to the F, appropriate measured et *e plane, additional power d ibution N

ps shall be taken r.nd F compared to F, and F,y at least on per 31 EFPO.

e. The F,y Its s for RATED THERKAL POWER (F P) sh I be provided for all cor's plane containing Bank "0" control s and all unrodded core planes in a adial Peaking Factor Lim Report per Specification 6.9.1 ;

.2.2e., above, are not applicable f.

The F,y limits of Specif tion 4.

in the following core plane re ons as meuured in percent of core height from the bottom of the uel:

1) Lower core region f 0 to , inclusive,

2) Upper core regio rom 85 to 10 inclusive,

3) Grid plane ions at 17.8

2%, 32. t 2%, 44.4 1 2%, 60.6 2 2%

and 74.9 i , inclusive, and

4) Core p ne regions within i 2% of core he t (* 2.88 inches) abou he bank demand position of the Bank " control rods.

g.

C '

With F y exceeding F,y, the effects of F,y on Fg (Z) all be evaluated to termine if Fq (Z) is within its limits.

4.2.2.3 hen Fg (Z) is measured for other than F,y determinations, an erall mens ed F (Z) shall be obtained from a power distribution map and increa d q

by s.

to account for manufacturing tolerences and further increased by 5% to count for measurement uncertainty. .

WOLF CREEK - UNIT 1 3/4 2-7

POWER DIS?RIBU?!ON LIMITS 3/4.2.3 ^0: fl^U ""; .'"" NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - F"p (4y) l.!MITING C01@! TION FOR OPERATION I!)Cf =

2 1 2.3 The ::rNtb ef !demd Scur 51=t !" t -(AC4)-tetel-How '

rateandRshallbemaintainedwithintheregionofai:ow=ableoperation l

's hey on Agure 3.2-3 for four loop operation.

I W r H

\e: F g ,

k" 1.49 (1.0 + 0.3 (1.0 P)] l j

N N THERMAL POWER , and b* P l = RATg INUUML PUWER N

c. Fh=Waa'iu values of F obtained by using the movable incore ,

detectors obtain a power distribution map. The sensured i values of shall be used to calculate R since Figure 3.2-3 :

includes esasu nt uncertainties of 2.55 for flow and 4% l l

for incure esasu nt of F"g.

aPPLICA81LITY: MODE 1.* l 4CTION:

ith the combination of RCS total flow ra nd R outside the region of acceptable operation shown on Figure 3.2-3:

a. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either: \

o

1. Restore the combination of RCS tota flow rate and R

[ to within the above Itaits, or -

(g 2. Reduce THEIMAL POWR to less than 50K of and reduce the Power Range Neutron Flux -

TED THERMAL POWER h Trip Setpoint td I

less than or equal to 55% of RATED THERMAL R within the !

-next 4 hnurs,

b. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of initially being outside the above 1 its, verify through incore flux mapping and RCS total- flow rate comp ison that' the combination of R and RCS total flow rate are restored (o within the above limits, or reduce THERMAL POWER to less than 5% oryRATED.

THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ; and N N

4c; Op;;ia' Test En;;tica %;;i'i;atica :.10,t --

WOLF CREEK - UN!I 1 3/4 2 3 Amencment No :

z__________ _ _

l

I INSERT p. 1 of 2 (Add to page 3/4 2-8) ;

3.2.3 FAH(X,Y) shall be limited by the following relationship FAHR M (X,Y) i FAHRb(X,Y) where, FAHRM(X,Y) = the maximum measured radial peak ratio defined in the Core Operating Limits Report (COLR) .

FAHRL(X,Y) = the maximum allowable radial peak ratio defined and specified in the CO LR .

APPLICABILITY: MODE 1 ACTION With FAH(X,Y) exceeding its limits

a. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , reduce the allowagle THERMAL POWER from for each 1% that-RATEg(THERMAL FAHR X,Y) exceeds POWER at least the limit, andRRH%

b. Within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months either:

1. Restore FAHR M(X,Y) to within-the limit for RATED THERMAL POWER, or

~

2. Reduce the Power-Range Neutron Flux - High Tgip Setpoint at least RRH4-for-each 1% that FAHR (X,Y) exceeds that-limit, and 4

l l

l RRH compensate is the amount for each of it THERMAL POWp(X,Y)reductionrequirgdto..

that' FAHR exceeds FAHR (X,Y) ,

and is specified in the COLR.

\

~ +

,.,,.,.ync. p 2,+. = e -e---se- . .* ,-w,.f < y. r y

_ . _ . _ _ . _ - . _ . _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ . . _ .._. _ _ _ _ _... _>~. -- ___ _ __ _,

INSERT p. 2 of 2 (Add to page 3/4 2-8)

c. Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of initially being outside the limit, either:

1. Restore FAMR M(X,Y) to within the limit for RATED THERMAL POWER, or ;

2. Perform the-following actionsi l

a. term by at least TRH** for each ReducetheOp(TK 1% that FAHR X,Y exceeds the limit, and-

b. Verify through incore mapping that FAHR M(X,Y) is _

restored to within the limit for the THERMAL POWER allowed by ACTION.a, 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 i

)

i t

- TRH_is the amour't of OTAT-K1 setpoint reduction required to-compensate for each 1% that FAHRM(X,Y) exceeds the limit and.is specified in the COLR.

DEEL6TED EASUREMENT UNCERTAINTIES OF 2.5% FOR FLOW ,

AND 4.0% FOR INCORE MEASUREMENT OF FJ

/

ARE INCLUDED IN THIS FIGURE m -

/

\ /

{u \ g /

. 3 [

l" $/

1. /s

.. d

\

8 f 0m k

\

h /

/ \.

\

,[ l I \

o.so o.es tm im 1.

R = RJ/1.49(1.0 + 0.3(1.0 P)]

! FIGURE 3.2-3 RCS TOTAL FLOW RATE VER$85 R '

FOUR LOOPS IN OPERATION \

s 3/4 2-9 Amendment No.23, 51 WOLF CREEK - UNIT 1

i POWER 0!S?R280T!0N LICIT 3 ,

i L!HITING CONDITION FOR OPERAT!0N ACTION (Continued) a, n 3/o:- c . 2. . b dp Identify and correct he cause of the out of-limit condition prior ,

to increasin THE L POWER above the reduced THERMAL POWER limit 1 required by CTION .2. :nd/:r b;, above; subsequent POWER OPERATION ,4 :

may proceed provided that 4.% c"-eti-P ef * :nd ind!::ted "CS- P/.sr (X,Y);

is reteL'S rete e : demonstrated, throu h incore flux mappingM "C! te* !' d'- ete tec deer, to be w thin the regi;r. ;f ;;;;;;;;'; // mil Tl 5080' M A epecetien e 6 n enTHERMAL POWER levels:r. 3.24 prior to exceedin in d e cout

1. A nominal 50% of RATED THERMAL POWER,

2. A nominal 75% of RATED THERMAL POWER, and

3. Within 24 nours of attaining greater than or equal to 95% of RATED THERMAL POWER.

SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

1.2 2 N: ::d= tier :f 'Mic:ted 805 totif1= r;t; =d 't ; Mil be- -

determined to be within the region of acceptable operation o' Cigure 3.2 3:

/NTd f a. Phar to operation above 75% of RATED THERMAL POWER after each fuel load'tng and

b. At least on er 31 Effective Full Power Days, t.2.3.3 The indicated RCS 1 flow rate shall be verified to be within the r egion of acceptable operation Fi ure 3.2-3 at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months wnen l

1 he most recently obtained value o obtained per Specification 4.2.3.2, is a slumed to exist.

.2.3.4 The E5 total flow rate indicators s I be subjected to a CHANNEL ;

Ad ALIBRATION at,Teest once per 18 months, WC t-d .2.3.5 The RCS total flow rate shall be determined by cision heat balance n easurement at least once ;or 18 months. Within 7 days pr to performing l-D 1 he precision heat balance, the instrumentation used for date ationofsteaml l Gg ressure, feedwater pressure, feedwater temperature, and feedwatchenturi-

.:P in the calorimetric calculations shall be calibrated. N l N. ,

d.2.3.6 The feedwater venturi shall be inspected for-fouling and cleaned as -

rececee j et ' eet :=: ;:r le :: nth:. --

WOLF CREEK - UNIT 1 3/4 2-10

INSERT (Add to page 3/4 2-10) 4.2.3.2 FAHRM(X,Y) shall be evaluated to determine whether Fag (X,Y) is within its limit by

a. Measuring FAHRM(X,Y) according to the following schedules

1. Prior to operation above 75% of RATED THERMAL POWER at the beginning of each cycle, and

2. At least once per 31 Effective Full Power Days.

b. Satisfying the following relationships FAHRM(X,Y) $ FAHRNOM ( X , ~ ')

where, FAHRNOM(X,Y) represents the nominal design power distribution increased by an allowance for the expected deviation between the nominal design power distribution and the measurement and is specified in the COLR.

If the above relationship is not satisfied, then for that location perform the following:

1. Calculate-the % margin to the maximum allowable design as follows:

FAHRM(X,Y) g yAH Hargin = (1 - -------)

L(X,Y) 100 FAHR

2. Find the minimum margin for all locations examined in 4.2.3.2.b.1, above. If the minimum margin is less than 0, comply with the ACTION requirements of Specification 3.2.3.

- ~

POWER O!$Ta!BUTICN LIMITS 3/4.2.5 CNB PARAMETERS ._

LIMITING CON 0! TION FOR OPERATION

3. 2. 5 limits shown on Table 3.2 1:The following DNB related parameters shall be mai a.

Reactor Coolant System 7, g, e e-b.

+ Pressurizer Press..e3 a.4 ApoLICABILITY: MODE 1.* c. Pudor Gelost Spfre (RC3) Flw Rede ACTION:

(L. With '; I or 2 of TAIAr ?.1~l within its limit within RATED THERMAL POWER within the next 4 h*urs, 2 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months or reduce THERMAL n SP. o f POWER to r

' If)$ff T L.

SURVEILLANCE REOUIREMENTS

.1 4.2.54 J their limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .Each of the parameters of Table 3.

INLE/tT 2 e s IN;Ft T 3

$u $po c <o l T<:f Exc e p he n 6 p,u henfron 3 ,10. 4 h r-

~

3. J. b . C .

WOLF CREEK - UNIT 1 3/4 2-14 l

i

INSERT I (Ad eo po$c 3/4 2-14) 1

h. With the comb 4netta.+d. RCS total flow rata +ad-A- outside the region of '

acceptabl6 operation chr = f!p : 3. 3- 3:

N m

! a. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either: W O. .* i 4 I. Restore the :Mf--if r Of "Cf total flow rate and_A.

to within the abow-lisit/,, or 6,2. Reduce THERMAL POWER to less than 50% of RATED THERML POWER.

2. 4 dir to k vu r s ej 4e r ;

c. bsbc Mc. Mal f/N rde 4 0 Nifkik +bt b specihed in +h e CO L A , o v' R

6. * / educe the Power Range Neutron Flux - High Trip 5etpoint to less than or equal to Su of RATED THERMAL POWER.

72 3#.

Within

. _ _"8- .e 4&. ,hours , sof_initia._lly m _ ..ebeing oesoutside

...,, , t,he

,....9:= limit #T verify

,_ge. tn,t the tf r-the above limit " " :t RES total flow ratebe restored to within THERMAL POWER ithin w/, the or reduce next THERML 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ; and POWER t/less than 5% of RATED y 4 ,3 sVe 4 *a is +keCoLR 4 f. Identify and corrock the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION

r9-' , above; subse say proceed provided that the em' et!-- -'

quent POWER OPERATION rd indicated RCS total flow rate'5 demonstrated.r *trO t- tm e -"; eM nct *-* ei *1- ** r 7 -'te , to be w' thin the region of acceptable f operation :hr = f t

(

THERMAL POWER 1evals: p 3.2-3 prior to exceeding the following a E. A nominal 5CE of RATED THERMAL POWER, 6 2. A nominal 75% of MTED THERMAL POWER, and

.c J. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of attaining greater than or equal to 95% of RATEC THERMAL POWER.

--- - - - . ,- , - ,-.r- -

USEa7 2 (Add +v ps y 3/4 : - w)

,e n o 5 pw &ca %

!.:. T. C 4.2.J.1 The provisions of Specification 4.0.4 are not applicable.

A DJ s E R. T 2 [ A44 % p c g e. 3/l/ 2 - I +)

e. 2. 0. 2 Tra c;c4h, tier, e? kdi::::d 000 t:ul = r:t: :nd " :h:M i:

determined to be within the region of acceptable operation of Figure 3. -

a. Prior to operation above 75% of RATED THERMAL P loading, and ter each fuel 4 I b b. At least once per 31 Ef a Full Power Days.

W

@ 4.2.3.3 region of The indic S total flow rate shall be verified to be within the a th a le operation of Figure 3.2 3 at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when recently obtained value of R obtained per Specification 4.2.3.2, is

- :: = d t: ::ht.

53 4.2.S The RCS total flow rate indicators shall be subjected to a CHANNEL CAllBRATION-at least once per 18 months, f.s 4.2.s-5 The RCS total flow rate shall be determined by precision heat balance measurement at least once per 18 months. Within 7 days prior to performing the precision heat balance, the instrumentation used for determination of steam pressure, feedwater pressure, feedwater temperature, and feedwater venturi AP in the calorimetric calculations shall be calibrated.

E. C 4.2.Mr The feedwater venturi shall be inspected for fouling and cleaned as necessary at least once per 18 months.

W Li CRE:: "tui 1 3/4 2 10 t

IA9L! 3. 21 CNB :cauETERS t.!MITS Feur L: Ops in D1Da"E*lo 0:eration

,1, In::':sted Rea:t:r ::olant System T,yg 1 $92.$'F y, :n:':stec D essuri:e* Pressure > 2200 psig a

3. fl os uh r Coole ek yS sir ~ F(w hie. As sr>ecl(ied is tke Corts CVE: 01 Hk LIM ITS R f 90KT (COLR)

" Limit not applicable during either a THERMAL POWER ramo in excess of 5*; of RATED THERKh PCNER per minute or a THERMAL POWER step in excess of 10% of RATED THERMAL POWER.

~.

WOLF CREEK - UNIT 1 3/4 2-15

t INSTRUugyTat;0N; _

"0VBLE !NCORE CETECTCR,5 L:'A:?!NG CONDITION FCR OPERATION 3.3.3.2 !$e Mcvaole Incore Cetection System shall be OPERABLE with:

a. At least 75% of the detector thimbles. *

' I

o. A minimum of two detector thimbles per core ouadrant, and F

c. Sufficient movable detectors, drive, and readout equipment to mao these thimbles.

APDLICABILITY: When the Movable Incore Detection System is used for:

a. Recalibration of the Excore Neutron Flux Detection System,

b. Monitoring the QUADRANT POWER TILT RATIO, or

c. Measurement of r . r (7) us r

,, p [ y,g 7,,) m f p [y, s()

ACTION: *

a. With the Movable Incore Detection System inoperable do not use tne system for the above applicable monitoring or calibration functions.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

Y SURVEILLANCE REQUIREMENTS , 4.3.3.2 The Movable Incore Detection System shall be demonstrate 1 OPERABLE at 9 least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by normalizing each detector output when=requirea for:

a. Reeelibration of the Excore Neutron Flux Detection System, or-

b. Monitoring:the QUADRANT POWER TILT RATIO. or Measurement of e ,

eg E :-z ",y. (x,y, z[ d (X,Y)

! wi '

,- UNIT 1 3/4 3-43 H

.i,,, , ,,. ..:..,..._ ._..m._.,,-,.., . . _ . . . _ - . . . . . , . , _...-4-... .,m,

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

title i 7-2 5?Em Lim 5 :, t E t, ;;..V

. s PER3 00c ytLVE vp5ER 13f;:

LIri sETTINftx-) a n :r:.:t s::t Lcon 1 Loco 2 Leon 3 Loco 4 iC 55 VC65 V075 V045 1185 psig 16.0 so. in.

V056 s'066 V076 V046 1197 psig 16.0 sq. in.

057 V067 V077 \047 1210 psig 16.0 so, in.

V;5S V068 V078 V048 1222 psig 16.0 so. in.

V059 VC69 V079 V049 1234 psig 16,0 so, in.

4 The lift setting pressure shall correspond to ambient conditione of t he valve at nominal operating temperature and pressure.

. , g(tes i. e st i w 3, t h e valv e s will be lest d 1l % ,

r WOLF CREEK - UNIT 1 3/4 7-3 *

{.- !:, E . " :* .

. . !:r:. :: : :;-':' .

. . . . . . .: .4

.: '. ~ t :-:n ::9:e trati:n of all filled portiens of the React :::'a

,ste e : t e -e'.e: 1; canal sna11 be maintainec uniform an: sofftt'ent ;;

+ s. e 1 3: t*e : e est-1:tive of the following reactivity ::n:ite: s s r.

1. A t,y, :' :.!5 er less, or Me lid s pe ciht e d is %e 2 - ::r:- :: :ent ation of greater than or eaval to 200:- ::-

A

r. ::E:;: <- q;;g g. C64 E O PER 4 n n a LI M I Ps R E/oR T (COL R),

icitn the re:vire ents of the above specification not satisfied, irteciately s s:e c all c:e-ations involving CORE ALTERATIONS or positive reactivity

nan;es an: int:1 ate and continue boration at greater than or eaual to 30 gom

f n 5: 1stion ::ntain'rg greater than or equal to 7000 com coron or ats ecui -

aient untti K,77 is reducea to less than or equal to 0.95 or tne Doron

n:eatration is restored to greater than or equal to 2000 00 , whi:never 's tne mere restrictive.

A(

SURVE!L'. ANCE E!OU!REUENTS ilt h a d 5 P i f f' d * 'k' COL R' 4.9,1.1 The more restrictive of the above two reactivity concitions shall be cetervined prior to:

a. Removing or unbolting the reactor vessel head, and

b. Withdrawal of any full-length control rod in excess of 3 feet from its fully inserted po:ition within the reactor vessel.

4.9.1.2 Th*e boren concentration of the Reactor Coolant System anc the refueling canal shall be determined by chemical analysis at least once per 72 nours.

4. 9.1. 3 Valves BG-V178 and SG-V601 shall be verified locked closed-and se:urec in position at least once per 31 drys.

'Tne react:r sna11 oe maintained in MODE 6 whenever fuel is in tne rea: tor l

vessel ith tre vessel head closure bolts less than fully tensionec or witn the nea: remevec.

I a;;' UN 7 '.

REEK 3/4 9-1 1

N l O E. L E T E !

l k t.@. C u es F.u.v,v6 pase s

y -. --

N::" ..

40 ." . -

, .. Acasptante 3 ::

e x \

s 3 0 .".

ss ,

4, .

a ..

t D "

z N ct: 20 ..- -___

D a m .

\N

.. N

.. Unsceeptable 1 0 :: '\ N N ,

Ol /[

ENRICHMENT (w/o)

/ F!sunt 3.91 WOLF CRitt MINIM E0pitts FULL MN

/

/ aunhur M A FURCT!0s er it!TIAL (NatcatNT

/ FOR $f04 AGE la Rit!ON 2 WOLF CREEK - UNIT 1 3/4 M g Amendment No.16

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

50

/

40 -

30 --

g m

Acceptatde e

(1.

Unacceptatdo

)

20 -

10 -

Ennchment Burnup 2.1 10 572 26 18 % 9 3.1 2514 38 34 601 45 43.357

' ! ' ' I '

0 1.5 2 2.5 3 . 3.5 4 4.5 5 ENRICHMENT (w/o)

FIGURE 3.91 WOLF CREEK MINIMUM REQUIRED FUEL ASSEMBLY BURNUP AS A FUNCTION OF INITIAL ENRICHMENT l FOR STORAGE IN REGION 2

5:!::JL ~!5* E'( !**:0Ns 3/4 *.0.2 0:000 WE,'ONT, !NSER*::N. W. **WE: 0:5*;:!U*:0N . :~5

!v!*!NG COND!'! N r00 CDERAT: N

3. ' 0. 2 ine Orcup height, inser' on, arc power distribution limits :'

Specifications 3.1.3.1, 3.1.3.5, 3.1,3.6, 3.2.1, and 3.2.4 .may :e s.s:er:ac curing the performance of PHYS!CS TESTS provicec:

a. The THERMAL PCWER is maintained loss tnan or eoual to 85.'. ' Eri:

THERMAL POWER, and ad 3 2.r.c

t. The limits of Specifications 3.2.2 444 3.2.3 3 maietai e:

anc cetermined at the frecuencies specifiec in 5:eci:st :-

4.10.2.2, below.

3:L: 04BIL!*v: MOC E 1.

1~ ,.~n.

<. ,o r 3 2. f. C.

P With any of the limits of Specifications 3.2.2 e 3.2.3 teing exceece: .t'e

ite requirements of Specifications 3.1.3.1,3.1.3.5,3.1.3.6,3.2.1.!.r: 3.2.4 are suscencec, either:

a. Recuce THERMAL POWER suf ficient to satis fy the ACTION rec,. rt e-,2 o f Speci fications 3. 2. 2'**e-3. 2. 3, o r

b. Se in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . N gg 7,y, g,g SL '/E!LLANCE REOU!;EMENTS 4.10.2.1 The THE; MAL POWER shall be cetermined to be less tnan er e: - s' ::

35. of RATED THER.'tAL POWER at least once per hour curing PHYS:03 T!575.

4.10.2.2 ine reautrements of the below listec s:ecifications sna11 :a :s :- a:

at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months curing PHYS!;5 TESTS:

a. Specification [4.2.2.2,-c':.2.3.::

5:ecification 4.2.3.2) and C. $ pac.ific a Yi , e 4. 2.i. 2 ,

>:. ::EE .1:* . 3/.'r 10 2

SPECIAt TEST EXCEPTIONS 3/4.10.4 REACTOR COOLANT LOOPS LIMITING CONDITION FOR OPERATION ,

/

3.10.4 The limitations of the lowing requirteents may be suspended: ;

j

a. Specification 3.2.3 nd 3.4.1.1 - During the performance of startup and PHYSICS TESTS in H00E 1 or 2 provided: i The THERMAL POWER does not exceed the P-10 Interlock Setpoint, l 1) and

2) The Reactor Trip Setpoints on the OPERABLE Intemediate and Power Range channels are set less than or equal to 25% of RATED THERMAL POWER. l

b. Specification 3.4.1.2 - During the performance of hot rod drop time measurements in MODE 3 provided at least three reactor coolant loops as listed in Specification 3.4.1.2 are OPERABLE.

l APPLICABILITY: During operation below the P-10 Interlock Setpoint or performance of not roo crop time measurements.

ACTION:

,2 a. With the THERMAL POWER greater than the P-10 Interlock Setpoint during

- the performance of startup and PHYSICS TESTS, insediately open the Reactor trip breakers,

b. With less than the above required reactor coolant loops OPERABLE during performance of hot rod drop time measurements, immediately place two reactor coolant loops in operation.

SURVEILLANCE REQUIREMENTS l

4. 10. 4. 1 The THERMAL POWER shall be detemined to be less than P-10 Interlock Se(point at least once per hour during startup and PHYSICS TESTS.

4.10.4.2 Each Intermediate and Power Range channel, and P-10 Interlock shall I be subjected to an ANALOG CHANNEL OPERATIONAL TEST witnin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to initiating startup and PHYSICS TESTS.

4.10.4.3 At least the above required reactor coolant loops shall be detersined 0)"RABLE within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to initiation of the hot rod drop time measure-Gds and at least or.ce per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during the hot rod drop time measurements

+rifying correct breaker alignments and indicated power availability.

3/4 10-4 Amendment No. 23. 36 -

WOLF CREEK - UNIT 1

, , ,- ,- N , ,,-r -. -l c- , ,,--me , ,,,A.,,..~.- -- , - - - - , . . , , , . . . -, - ,- ----: . . , , , -

s fed ch Me ( 6 0L) s EACTIVITY CONTROL SiSTEMS

/ speifel em He coU oHGw'd BASES / , m ars ne w (co W

5**

H00ERATOR TEMPERATURE COEFFICIENT (Continued)

The most negative MTC value equivalent to e most positive moderator density coefficient (2C), was obtained try inc ntally correcting the MOC used in the FSAR analyses to nominal operati conditions. These corrections involved subtracting the incremental change n the l@C associated with a core condition of all rods inserted (mostjositi e MC) to an all rods withdrawn condition and, a conversion for thejtate change of moderator density with ,

i temperatureatRATEDTHERMALPOWERecondip'ons. This value of the _MQC,_was then transfomed into the limiting MTCNalue, .1:10' E'h/4. TherMTC value e- ,

3.2 ;,10 ' E'V represents a conservative vala (wie corrections for burnt,p :

and soluble boron) at a core condition of 300 ppe equilibrium boron concentration and is obtained b making these corrections to the limitin KTC value.e+.

' 1= = 10 * ^1'h/g" yec4ed mfle E cc Lg The Surveillance Requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confim 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.

3/4.1.1.4 MINIMM TDFERATURE FOR CRTTICALITY This specification ensures that the reactor will not be nede critical with the Reacter Coolant Systes average temperature less than 551*F. This limitation is required to ensure:, (1) the moderator temperature coefficient is within it analyzed temperature range, (2), the trip instrumentation is within its normal operating range, (3) the pressurizer is capable of being in an OPERA 8LE status with a steam bubble, and (4) the reactor vessel is above its-minimum RT t"Pt"

ET 3/4.1.2 80 RATION SYSTEMS The Boration Systems 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) centrifugal charging pumps, _ .

(3) separate flow paths, (4) boric acid transfer pumps, and (5) an emergency power supply from OPERABLE diesel generators.

With the RCS average temperature equal to or greater than 350'F a minimum -

of two boron injection flow paths are requ bed to ensure single functional-capability __in the event an assumed failure . enders one of the flow paths inoper-aole. The boration capability of either flow path is sufficient to provice a SHUTDOWN MARGIN from expected operatin? conditions of 1,3% ak/k af ter menon decay and cooldown to 200*F. - The mani: 10 expected boration capability require-ment occurs at EOL from full power eouilibrium xenon conditions ana. requires i 17,658 gallons of 7000 ppm burated ater from the boric acid storage tanu or B3,754 gallons of 2400 ppm boratea water from the RwST. With tne RCS aver m temperature less than 350*F, only one ooron injection flow path is rHiuc W0tf CREEK - UN!T 1 0 3/4 1 2 Are c oat M :'

e - -- b* g-ere, w= 4+s-W.?-w e-ew--rawsevi,a-ww--93e'+ wucw w ttww w e --yy yi .de-4-w-~, ~,wwy -we.p. gr r yww qv. t v4vyw whyy-wky tv)@ e-v'N-+,--" "

y-q---y-t-w-g w -w eq e

g g --

fe DLIBf lov1+ k k *-P9: S*% -

"A' "' ' '

f a m ,t r ( c oa) _

/ (

/ '

?

/ 3: : *0w!: ::s : n u ::s .:v: s

/, :_.: :..f.t I t

I

e ::e:i:stiens of this section Orovice assurarce :' fuel tate;*'t e'.

: : :iti:n I ($ rmal Operation) anc II (Incicents of wecerate : :.ea:

eats y:

(a) maintaining tne minimum Ch8R in the care ;"Gater then. : P e:.a'

%-44-:JPing non-al oceration anc in short term transients, anc (:) 3- t ;

t*e "ssien ;as celease, fuel pellet temoerature, anc claccing me:Sanita'

::e-stes t: -i tnin assumec casign criteria.

In accition. Itmitte; tre :m irea* =c er censity curing Concision I events provices assurance tnat t's

'r'ttal :encitions assumec for the LOCA analyses are met anc the E: 3 a::e u :e

-iter's limit of 2200'F is not exceeced.

re cefinitions of certain hot channel and peaking f actors as usec 'n trest 5:ecifications are as follows:

r

,

Heat Flux Het Channel Factor, is cefined as the : ' f local

( heat flux gn the surface of a fuel roc at core elevati:n ! ctvice*

Fq (X,Y,Z) oy the average fuel rod heat flux, allowing for manuf acturing tolerances on fuel pellets and rocs; -

'g % CJ M s* A 1(4 Y)

Nuclear Entnalpy Rise Hot Channel Factor, is cefinec as tne aat':

(F3(4Y) c:he integral of if near power along the red with tne hignest 'ete; tti:

t i

e a_

er to tne average roc powep-:af ctsseJv ((,Y). -

A::t:1 '::d ; ':: : , i: c:" :c :: tM:

tie e' re: ::-t- n ;.

': a n ---- A e* -- L

:.; :;; ::_:r ::r:tt3 th: ':rt::rt:'

U: :. . x: At twx 0!rr!RENCE '" 1 '

(afb) V2 [46Zb N %D>b :

ine limits on AXIAL -

5 FLUX OIFFERENCE' assure that the r en m.-.. -----

i

..m .$--. . A W i

<r . limit :: " -

t ' d : M ir 4 2 ' ;- "

W A d a 0 * * * * ^ * .4net *a re -

ex:ee:e: L L 4

uris; either normal operation or in the event of xenon reciptritutien- foll:-ir; cewer cringes. % Afb hemifs have been ad pated (.or nemsbed usertMnty.-

A Daevisions for tonitoring the AFD on an automatic basis are ceaive: :- '

( ne clant process computer through the AFD Monitor Alarm.- The c:mouter cete--

l

,ines the 1-minute average of each of the OPERABLE excore detectcr Out uts are ocevices an alarm message immeciately if the AFD for at least 2 of 4 or- 2 o' *: i 03ERABLE greater inan 50% of RATED THERMAL POWER. excore channels are outsics the AFD limits and the THERJ'.AL 00' !R is -

U 2.2 at: 1/4.2.3 w! AT *WX HO? OwaNNEi. a:T*R. are *:: ".*****"I a..: .iea N -e. h USE -07 CnAt.NEL fag;R

- e I1mith on neat flux not cnannel fa *3 e**,*a';/ *ise net *nannel f acter easure tna*.: (*.)N#'?"*^ . a

c %'

I e l '

'0*si ::=t* Oensity anc minimum 0. SIR a'e *0% en*eece , ar* (0) ir t*e twe*:

tSe celiq' II*1ts *n Oe3L 1

, - 0.:- ::E!' > us i ; 5 3/ :;

! amen: est s: e -

i l

--9 =-wvw y- g ig-rpp,p.--y,g.f9 mg g-.*-em-- rgym=.-,y,wv-7 yy.e-gg----,-wr- g ,-gr.wyys.7+.3.- - - - . . *ny- sw r*'.r+rg

1 2 .r: ::! e:!u :09 i.:w! s

!a5ES 3/4 2.2 and 3/A.2.3 w! A- F L'.'x 20' Owa NN!'. ::C: R. : : :46.~i:' '* : LT> ele te s...i.; isr Atav i;5E -07 -assE. rec ~:; (;:ntte e:;

LOCA tSe Osak fuel clad tem traturt will not ex:eed the 2000'~ il:2 4::t::2 :t

rite ta limit.

Ea:n of these is measurable but will normally only te :ste-mire

eM:ci: ally as scacified in Scecifications 4.2.2 anc 4.2.3. This :e 4::':

5.rve111ance is sufficient to insure that the limits are maintaire: : :< : :

4. Centrol rods in a single group move together with no incivicati a::

insertien differing by more than : 12 steps, incicatec, f r:m t. e greco comand positten,

e. Control red groues are secuenced witn ever apping grou:s as cas:- :e:

in Sceci fication 3.1. 3.6, CL? Oi!!A UN!T 1 2 3/a : acer- e*: N:. .

POWER O!5TR!8VTION t!M!?S BASES HEAT FLUX H0T CHANNEL FACTOR and "CI ".0" "JTE ^ NUCLEAR ENTHALPY RISE D ANNEL F.'+CIOR (Continued)

c. The control rod ineartion limits of Specification 3.1.3.6 are maint;ined, and

d. The 4 4 41 power distribution, expressed in terns of AX!AL FLUX 01Fi .INCE, is saintained within the limits.

t C* l' '/)h hwill be maintained within its limits provided Conditions a. throu

d. above are maintained. ^: :ted : ' t -- 3 2-3, "!! '! = ::t: rd( . ;

=> 5: J't :::d :'" :;;iMt e another (i.e. , a low measured RCS flow

% N

/yyt rate is acceptable if the seasu F j is also low) to ensure that t alcu-s i lated DNBR will not be below the 54 ty analysis DMR value. T relaxation 4

4 of F N

as a function of THERMAL POWER ows changes in t adial power shape w g 6.-

for all permissible rod insertion limits. 4 '

~J 13 R as calculated in Specification 3.7.3 d in Figure 3.2-3, accounts a forFhlessthe or equal to 1.49. T value is u in the various accident analyses where Fg influences pa ters other than DM , e.g. , peak clad tem-perature, and.thus is the sua "as measured" value all Fuel rod bowin duces the value of 0 4 rstio. Credit is 11able to offset this reau . on in the generic margin. The generic eargins, taling 11.4% DN8R, lately offset any rod bow penalties. This is the ser i between correlation DMR limit (1.17) and the safety-analysis lie t BR (1.32

..: :;;1' nt'; .;1.n ;f M tr_ sc,;; tie; ere .efe..c M ic, tr T L .

i ""- -- S ==;;r .t 44:b , e !!! rtr^ fe- bet' 3-dr-teLct-and manufacturing tolerance sEii. L "- ^ ;1L__,.e of 55 is appropriate y

.t for a full-- -

' A m n the Incore Detector r s " ;;' 0 M tes, and a

4'1 ;; ; r_r- 'O ;;;;;;S'- f:r ==f=ter'z t:1:r==.

4 l

l 8 3/4 2-4 Amendment No, 51 WOLF CREEK - UNIT 1

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

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

i- ,

INSERT 1 (Add to page B 3/4 2-4)

The if91tu on the nuclear enthalpy rise hot channel factor, ,

Pag (X,Y), are specified in the COLR as Maximum Allowable Radial Peak Ratio limits, obtained by dividing the Maximum Allowable Peak (KAP) limit by the axial peak for assembly location (X,Y). By definition, the Maximum Allowable Radial Peak Ratio limits will result in a DNBR for the limiting transient that is equivalent to the DNBR calculated with the design Fay (X,Y) value specified in the COLR and a limiting reference axial power shape.

i

! INSERT 2 (Add to page B 3/4 2-4) l FgM (X,Y,Z) and FaHRM(X,Y) are measured perk .cally to provide assurance that they remain within their 13 A peaking anargin calculation is performed, when necessary, ts arovide the basis for reducing THERMAL POWER, for reducing the widt.t of the AFD limits, and for reducing the f1(AI) limits of the OTAT trip setpoints.

l-l l

l l

l

{

POWER O!STRISLITION LIMITS BASES

~

7 EIT FLUX HOT CHANNEL FACTOR. and RCS FLOW RA1E ANO NUCLEAR ENTHALPY RISE h0 KCHANNEL FACTOR (Continued) adial Peaking Factor, F,y(Z), is sensured periodically to provide assurance t the Hot Channel Factor, F (2), remains within its limit. The F,y limit for RAT THERMAL POWER (FRTP)qas provided in the Radial Peak Factor Limit Report pecification 6.9.1.9 was determined from expected .

power control manuevers r the full range of burnup conditions in the core. '

q When RCS flow rate and F are seasurad, no additional allowances are

' $ necessary prior to comparisen with' limits of Figure 3.2-3. Measurement

] errors of 2.5% for RCS total flow rata d4%forFhhavebeenallowedforin

.c

. determination of the design DN8R value. NN The n~ asurement error for RCS total flow ratsJs based upon perfornicg a precision heat balance and using the result to calilty to the RCS flow rato indicators. Potential fouling of the feedwater ventu ich sight not be detected could bias the result from the precision heat be e in a nun-conservative manner. Therefore, an inspection is performed the feedwater venture each refueling outage.

The 12-hour periodic surveislance of indicated RCS flow is suff ant to ptable-detectonlyflowdegradationwhichcouldleadtooperationoutsidethea%es rep %n of operation shown on Figure 3.2-3. This surveillance also provid adetwate monitoring to detect any core crud buildup. x' %

3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distri-butian satisfies the design values used in the power capability analysis.

Radial power distribution esasurements are made during STARTUP testing and periodically during power operation.

l The 1 1.02, at which corrective ACTION is required, provides DNB 1 and linear ration rata protection with x-y plane power Oilts. A limit of 1.02 va to provide an allowance for the uncertainty associated with the indica power tilt.

The 2-hour time allowance for operation with a tilt condition greater

! than 1.02 but less than 1.09 is provided to allow identification and correc-tion of a dropped or sisaligned control rod. In the event such ACTION does not correct the tilt, the margin for uncertainty ongis reinstated by reducing the maximum allowed power by 3% for each percent of tilt in excess of 1. !

l

> qF (x,Y,2.)

WOLF CREEK - UNIT 1 B 3/4 2-5 Amendment No. 23 1

POWER O!STRIBUTION LIMITS 4

BASES

_y QUADRANTPOWdRTILTRATIO(Continued)

For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the noveabic incore detectors are used to confirm that the normalized syimmetric power distr;bation is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is de n with a full incore f 1 ta. map or two sets of four symmetric thimbles. The > '9ts of foer symmetric thimbles is a unique set of e'ght detector locations. a locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, N-3, 3/4.2.5 DNB PARMETERS pucke Coelut Spfr. T uf .4,e per ssa ci2 e c pro:u re.

The limits on the - elatM "arte 4-assure that each of the pa. noters are maintained within the nonsal steady-state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial,5&ML assumptions and have been analytically demonstrated adequate to ufAR "rstintain a ONBR above the safety analysis 1.ait DMR (1.%' throughout each I analyzed transient. The indicated T,yg valve of 592.5'F a the indicated I pressurizer pressure value of 2220 psig correspond to anal ical limits of 59E*F and 2205 psig respectively, with allowance for seas resent uncertainty.

The 12-hour periodic surveillance of these paramete s through instrurrent readout is sufficient to ensure that the parameters are estored within their limits following load changas and other expected transi t operation, h5 w

spee s (s el M H a coes tra w ne LI M I T.S R E Pb f T ( C O L.8)

WOLF CREEK - UNIT 1 B 3/4 2-6 Amendment No. 51

TrJ : 69 T I (Add 4 (>

8 e 6 3/v 2-6) l Fuel rod bowing reduces the value of DNS ratio. Credit is available to offsts this reduction in the generic sargin. The generic margins, '-' ;

11 it 0""", completely offset any rod tow penalties. This is the margin betwen the correlation DNBR limit (1,17) and the safety analysis limit DN8R (1. :"'., % e s e. limih are s p se.t k ed in H e c ot l .

The applicable talues of rod bow penalties are referenced hi the /$AR, a

Im EAT' 2- (t.M po y 5 3/+ 2N c

(g'y))pe r Spnk ca% ?.2.3, h 44e Cod.

When RCS flow rate

.are measured, andh['

no additional all necessary prior to comparison with the limits Of f f;;r; 3.2-[ Me:surement are

""YSN["of 2.5% for RCS total fNw rate W 4% for have been allowed for in determination of the design DN8R value. (p (y, y) unuchwy The measurement 4*aer for RCS total flow rate is based upon performing a precision heat balance and using the result tc calibrate the RCS flow rate indicaters. Potential fouling of the feedwater venture which might not be detected could bias the result from the precision heat balance in a non-conservative manner. Therefore, an inspection is performed of the feedwater venture each refueling outage.

The 12-hour periodic surveilltM of indicated RCS flow is sufficient to detect only flow degradation which could lead to operatior outside the acceptable region of operation :5:e :r ~f;;r: 3.20. This surveillance also provices adequate monitoring to %etect any core crud buildup.

speuped in +Le c o t.R. .

INSTRUMENTATION BASES 3/4 3.3 %NITORING INSTRUMENTATION 3/4 3.3.1 RADI ATION MONITORING FOR PLANT OPERATIONS The OPERABILITY of the radiation monitoring instrumentation for plant coerations ensures that: (1) the associated ACTION will be initiated when the '

radiation level monitored by each channel or combination thereof reaches its '

5etooint, (2) the specif f ec coincidence logic is maintained, and (3) suf N-cient redundancy is maintained to permit a channel to be out-of-service for testing or maintenance. The radiation monitors for plant operations senses radiation levels in selected plant systems and locations and determines whether or not predetermined limits are being exceeded. If they are, the signals are combined into logic matrices sensitive to comoinations indicative of various accidents and abnormal conditions. Once the required logic combination is complettd, the system sends actuation signals to initiate alarms or automatic isolatio1 action and actuation of Emergency Exhaust or Control Room Emergency ventilation Systems.

3/4.3.3.2 MOVABLE INCORE DETECTORS '

/

< bU Of g i The OPERABILITY of the movable incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatiay neutron flux distribution of the core. The OPERABILITY of this system is demonstrated by irradiating each cetector used and determining the accep ability of its voltage curve.

For the purpose of measuring gT (U :r a full incore flux map is used.

Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in recalibration of the Excore Neutron Flux atection System, and full incore fla maps or symmetric incore thimbles may be used for monitoring the QUADRANT POWER TILT RATIO when one Power Range Neutron Flux channel is inoperable.

3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of t.he seismic instrumentation ensures that suf ficient capability is available to promptly determine the magnitude of a seismic event and evaluate the respaast of those features important to safety. This capability is requ N to permit comparison of the measured response to that used in tPe desigt m is for the facility to determine if plant shutdown is 4N ed a pursuant *.o Appeno1x A of 10 CFR Part 100. The instrumentation is consiste :

witn the recomandations of Regulatory Guide 1.12, " Instrumentation for ,

Earthquakes," April 1974.

WOLF CREEK - UNIT 1 B 3/4 3-4

OESIGN FEATURES

5. 3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The core shall contain 193 fuel assemblies with each fuel assembly nor-mally containing 264 fuel rods claa with Zircaloy-4 except that limiteo sub-stitution of fuel rods by filler rods consisting of Zircaloy-4 or stainless steel or by vacancies may be made if justified by a cycle-specific reload-analysis. Each fuel rod shall have a nominal active fael length of 144 inenes.

The initial core loading shall nave a maximum nominal enrichment of 3.10 =eignt percent U-235. Reload fuel shall be similar in physical design to the initial core loading. ant-seen-have-a-raxhe enekhmenta f ". 50 veigtit7errumMR Delete CONTROL R00 ASSEMBLIES 5.3.2 The core shall contain 53 full-length and no part-length control rod assemblies. The full-length control rod assemblies shall contain a nominal 142 inches of absorber material. All control rod assemblies shall be hafnium, silver-indium-cacmium, or a mixture of both types, All control rods shall be !

Clad with stainless steel tubing.

5.4 REACTOR COOLANT SYSTEM OESIGN PRES 5URE AND TFMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:

a. In accordance with the Code requirements specified in Section 5.2 of the FSAR, with allowanc6 for nortsal degradation pursuant to the applicable Surveillance Requirements,

b. For a pressure of 2485 psig, and

c. For a temperature of 650*F, except for the pressurizer which is 080*F.

VOLUME 5.4.2 The total volume of the Reactor Coolent Systca, including pressurizer and surge line, is 12,135 2 100 cubic feet at a nwinal T,yg of 557'F.

5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The a teorological tower shali be located as shown on Figure 5.1-1.

5-5 Amendment No. 3, If, 19, 25 WOLF CREEK - UNIT 1

l DESIGN FEATURES 56 FUEL STORAGE

_ CRITICALITY 5.6.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. Ak gf equivalent to less than or equal to 0.05 when flooded with unbor ated water. -wh4cA4*c4eoe,-e-eense+ vet 4ve-sttewence cf 2. 0% P=Me 4kAf o unc4*444*t4es an Um -itret-in4ect+cn-4 ihrf-ttie-FSAR-Dett Thi s i s ba sed on new f uel w4.th-an-en-Went+f--4tSO-weight pettent-D**t I ij-23?rin Region 1 and on spent fuel with combination of initial enrichment and discharge exposures, shown in Figure Gr6-t, in Region 2, and 3.9-1 b.

A nominal 9.236 inch center-tc-center distance between fuel assemblies l placed in the storage eacks.

5. 6.1. 2 The k for new fuel for the first core loading stored dry in the eff spent fuel storage racks shall not exceed 0.98 when aqueous foam moderation is assumed.

DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 2040 feet.

CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1344 fuel assemblies.

5. 7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.

WOLF CREEK - UNIT 1 5-7 Amendment No. 16 1

OCLC -E. .-

q .,, ,me -c:.Ww& . O i-N

't i .

50 . .... _

.' ." \ s Q ' .l .

Acceptable /

3,

/

x 30." /

s .

v.

/

v / x c, ..

3 .. ,

@ 20 " j' ',

a .. e s c ..

.. ,/ 3

." / Unacceptable 10 .- N

.. s y

? \

/: :

N~

N y

0 " ' :.:::l. ::: :::l . :. :.:l ::

,e 1.0 2.0 3.0 4.0 - 5.0 ENRICHMENT (w/o)

FIGutt 5.6-1 talLF CRitt RIN!mst et0UlttD Furt. Asmv OURNUP A5 A FUNCT!ss y 3337!AL (NRIMWT FOR 570RAst IN ateI0R t l

l WOLF CREEK - UNIT 1

~

53 Amendment No. 16 i

l ADMIN!sTRATIVE CONTROLS I

-4AGIAL-Ali4MNG4 ACTOR-k!MIT AipOAT-Wag 9 The F,y limits for RATED THERML POWER (F ) shall be provided to thir NRC RegTQAdministrator with a copy to Director of Nuclear ReacDP-Regulation, Attention: CMet Core Performance Branch, U.S. Nuclear RegpatcFy Commission, Washington, D.C. for all core planes containt Barr 0" control rods and Ru!a u all unrodded core planes an plot of prodi F q .Pg ,j) vs Axial Core

'M N Height with the limit envelope at le days prior to each cycle initial n., <, g g crfticality unless otherwise appe by the omaission by letter. In addition, in the event that the limp i uld change requi a new submittal or an amended submittal topPeaking Factor Limit Report, Q hall be subaltted 60 days prior t fra date the limit would become effective %1 ass otherwise a owed b e Cossaission by letter. Any information needed t h spo F 1 be by request from the NRC and need not be included in this repoA

PECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the NRC Regional Office within the time period specified for each report.

6.10 RECORO RETENTION In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

6.10.1 The following records shall be retained for at least 5 years:

a. Records and logs of unit operation covering time interval at each power level;

b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal itses of equipment related to nuclear safety;

c. All REPORTA8LE EVENTS;

d. Records of surveillance activities, inspections, and calibrations required by these Technical Specifications;

e. Secards of changes made to the procedures required by 5pecif1 cation 6.8.1;

f. Iecords of radioactive shipments;

g. Records of sealed source and fission detector leak tasts and results; and

h. Records of annual physical inventory of all sealed source material of record.

WOLF CREEK UNIT 1 6-21 Amendment No. 42

INSERT (Add to page 6-21)

QW E OPERATING LIMITS REPORT (COLR) 6.9.1.9 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT (COLR) before each reload cycle or any remaining part of a reload cycle, for the following:

1. Specification 2.1.1 : Safety Limits - Reactor Core

2. Specification 3.1.1.3 : Moderator Temperature Coefficient (MTC) BOL and EOL limits

3. Specification 3.1.3.5 : Shutdown Rod Insertion Limit

4. Specification 3.1.3.6 : Control Rod Insertion Limits

5. Specification 3.2.1 : Axial Flux Difference (AFD)

6. Specification 3.2.2 : Heat Flux Hot Channel Factor

- Fg(X,Y,Z)

7. Specification 3.2.3 : Nuclear Enthalpy Rise Hot Channel Factor - FAH(X,Y)

8. Specification 3.2.5.c : Reactor Coolant System (RCS) Flow Rate

9. Specification 3.9.1.b : Refueling Boron Concentration The analytical methods used to determino the core operating limits shall be those previously reviewed and approved by the NRC.

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.

The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

.___-_---_--l

ML20116A868

Text

0.02/*F for increasing average temperature and 0 for decreasing everage temperature; y[7 125

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