Text

Proposed Tech Specs Change Request Relocating cycle-specific Core Operating Parameter Limits to Colr.Proposed Changes Based on Guidance Found in NRC GL 88-16,WOG-90-016, NUREG-1431 & COLR Approved by NRC
	ML20128F882
Person / Time
Site:	Farley
Issue date:	09/30/1996
From:	; SOUTHERN NUCLEAR OPERATING CO.
To:	;
Shared Package
ML20128F871	List: ML20128F868; ML20128F882;
References
	RTR-NUREG-1431 GL-88-16, NUDOCS 9610080264
	Download: ML20128F882 (111)
	v • d • e

{{#Wiki_filter:_ _ _ _ _ . _ ~ _ _ _ _ _ . _ . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ l. I eld 11 DEFINITIONS _z........................................,,_,, SECTION P.AliE 1.0 O UfMITIONS 1.1

              '                         1.2       ACTION .........................................................                                               1-1 1.3        AXIAL CHANNELFLUX                               DIFFERENCE ..........................................1-1 CALIBRAT!0N.........

1.4 CHANNEL CHECK .............. ...... ................................... 1-1 1.5 1.6 CHANNEL FUNCTION TEST ............. ............................ 1-1 1.7 CONTAlletENT INTEGRITY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 CONTROLLED LEAKAGE .......... ............................ 1-2 m1.8 CORE ALTERATION ................................................ 1-2 r 'I . 9 00SE EQUIVALENT I-131 ............. 1-2 f 1.10 E-AVERAGE DISINTEGRATION ENERGY............ 1.11 .......... ..................... 1-2 1-3 ENGINEERED SAFETY FEATURES RESPONSE 1.12 1-3 1.13 FREQUENCY NOTATION ................. TIME .............. 1.14 GASE0WS-RASWASTE-TREATNENT-SYSTEM (Deleted) .................. 1.15 IDENTIFIED LEAKAGE ..................................... ....... 1-3 1-3 1.16 kigulB-RA9 WASTE-TREATNENT-SYSTEM (Deleted) ..................... ........ 1-4 MAJOR-GHANGES-TG-RADIGAGTIVE-WASTE-TREATMENT-SVSTEMS 1.17 0FFSITE 1-4 00SE CALCULATION MANUAL 1.18 OPERABLE - Oh.RA81LITY . . . . . . . . . .(00CM .....

                                                                                                                                             ........... 1-4 1.19 OPERATJ0NAL MODE - MODE .............. ......................... 1-4 1.20 PHYSICS TESTS ......................... ........................ 1-5 1.21 PRESSURE BOUNDARY LEAKAGE ......................................

1.22 ....... 1-5 1-5 PROCESS CONTROL PROGRAM 1.23 PURGE - PURGING ........(PCP) ..................................

                                                                                                                                   ................. 1-5 1.24 QUADRANT POWER TILT RATIO ...................................... 1-5 1.25 1.26 RATED THERMAL POWER ....................                                       ........ ....................... 1-5  1-6 1.27 REACTOR TRIP SYSTEM RESPONSE T REPORTA8LE EVENT                            .............IME ..
                                                                                                                                      ............... 1-6 1.28       SHUTDolel MARGIN ................................................

1.29 SGklOIFIGATION

                                                                                                                .........                                     1-6 1-6 1.30 SOURCE CHECK                                             ...(Deleted) .......................................
                                                                                                                          .......................              1-6 1-6 1.31 STAGGERED TEST BASIS ................. .........................                                                     1-6 1.32 THEMIAL POWER ........................ .........................                                                     1-7 1.33 UNIDENTIFIED LEAKAGE .................

1.34

                                                                                                                        ................                       1-7
                                                                      ...................NT SYSTEM .................. ........

VENTILATION EXHAUST TREATME 1.35 VENTINE ........ 1-7

     .                                                                                                  ....................................- 1-7 TABLE 1.1 OPERATIONAL MODES .........................................                                                     1-8 TABLE 1.2 FREQUENrY NOTATION ...........................,s...........

1-9

                                   - /&           Core opentty Limik hget , ,                                                       .
                                                                                                                                                              /t FARLEY-UNIT 1                                                        I                             AME)EslENT NO. H. 99 9610080264 960930 PDR             ADOCK 05000348 P                                    PDR

INDEX ADMINISTRATIVE CONTROLS SECTION PAgI Review................................................... 6-10 Audits................................................... 6-11 Authority................................................ 6-12 Records.................................................. 6-12 6.5.3 TECHNICAL REVIEW AND CONTROL Activities............................................... 6-12 Records.................................................. 6-13 6.6 REPORTABLE EVENT ACTION..................................... 6-14 6.7 SAFETY LIMIT VIOLATION ..................................... 6-14

6.8 PROCEDURES AND PROGRAMS..................................... 6-14 6.9 REPORTING REOUIREMENTS 6.9.1 ROUTINE REPORTS Startup Report .......................................... 6-15a Annual Report............................................ 6-16 Annual Radiological Enyjronmental Operating Report....... 6-17 Annual Radioactive Effluent Release Report............... 6-17 Monthly Operating Report................................. 6-19 e C4 we. o u%., '. _r.Limi( Report . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . __Li- 6-19 l l Annual Diesel Generator Reliability Data Report.......... 6-19o.*

        &ammal Reactor Coolant System Specific Activity Report... 6-20                                     l Amaual sealed Source Leakage Report...................... 6-20 6.9.2 SPECIAL REPORTS...........................................                                  6-20 6.10 RECORD RETENTIos...........................................                                  6-20 6.11 RADIATION PROTECTION PROGRAM................................ 6-21a 6.12 HIGH RADIATION N ......................................... 6-22 FARLEY-UNIT 1                     III                                           AMENDMENT NO.

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

OEFINITIONS CONTAINMENT INTEGRITY

1.6 CONTAINMENT INTEGRITY sna11 exist wnen

j a. All penetrations required to be closed during accident conditi either: ons are l 1) 1 solation valve system, orlapaole of being closed matic by an OPE ( 2) valves secured in their closed positions, exc Taele 3.6-1 of Specification 3.6.3, ( b. All equipment hatches are closed and sealed, c. Eacn air lock is OPERA 8LE pursuant to Specification 3.6.1.3, d.

The containment 3.ti.1.2, and leakage rates are within the limits of Specification e.

bellows or 0-rings) is OPERA 8LE.The sealing mechanism , welds, assoc 4 CONTROLLED LEAKAGE - 1.7 coolant CONTROLLED pump seals. LEAKAGE shall be that seal water flow supplied 1.o tne re 1 CORE ALTERATION

1.8 CORE ALTERATION shall be the movement or manipulation of any compone within vessel. the reactor pressure vessel with the vessel hsad removed and fuel in the of a component to a safe conservative position. Suspension of CORE

               / MAT 4 00SE EQUIVALENT I-131 1.9 OOSE E@1 VALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and f astspic stature of I-131.1-132,1-133,1-134 and 1-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in Table E.7 of Regulatory tiuide 1.109. Revision 1,1977. J FARLEY-UNIT 1 1-2 M NDMENT NO. 63

I INSERT A P g3 1-2

  *;- ** ;. t     .    ;  ..

1.8a The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with specification 6.9.1.11. Unit operation within these operating limits is addressed in individual specifications. l W l

a 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 00 RAT!0N CONTROL

                                            $ NUT 00WN MARGIN - 7,,i; M F LIMITING CON 0! TION FOR OPERATION 3.1.1.1 The SHUTDOWN MAAGIN shall be greater than or equal toy."_ M*- wfu for 3 loop operation.

TWE usser sessiF,sa #w TWe APPLICA81LITY: N00ES 1, 28, 3, and 4. g q *. rn werorsAmurera w rwa q . With the SHUTDOW MAAGIN,less thany.~~ dt VA tenediately initiate and continue beration at greater than or equal to 30 spa of a solution containing greater than or equal to 7000 ppe beret or equivalent until the twired SHUTDOW MAAGIN is restored. SURVEILLANCE REQUIRDENTS 4.1.1.1.1 The $NUTOO W MAAGIN shall be determined to be greater than or equal rH 5-deMsNa k. tof.ryg 4., asst sfes*Fep w rara go

a. Within one hour after detection of an inoperable control rod (s) and at least once per 12 hours thereafter while the rod (s) is inoperable.

If the inoperable control red is tenevable or untrippable, the above roeiired SWTDOW MARGIN shall be verified acceptable with an incNased allowance for the withdrawn worth of the 1emovable or untrippeble control rod (s).

b. When in pWDE 1 or N00E 2 with K greater than or equal to 1.0, at least once per 12 hours by vlNfying that control bank position is within the insertion limits of Specification 3.1.3.6.

c. %Aten in IEEE 2 with K less than 1.0, within 4 hours prior to achieving reacter criNlality by verifying that the predicted critical central red positten is witHn the limits of Specification 3.1.3.6.

d. Prior to initial operation above 55 RATED THEN W. POWER after each fuel loading, by censideration of the facters of a belev, with the control banks at the manism insertion lisit of Specification 3.1.3.6.

                                             "See Special Test Exception 3.10.1 O

3/4 1-1 AfMCIOT NO. 26 FAALEY-UNIT 1 e

1 j REACTIVITY CONTROL SYSTDtB SMUTD0t#1 MAAGIN - T, 1 200*F l LIMITIN8 CON 0! TION FOR OPERATION i i 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal tog . "_ 2 ^ V i i APPLICA8!!.!TY: MODE 5. "# 3"

l ACTION:

f %?b L-I With the SHUTDOWN MARGIN less than * " ' - '/t, immediately initiate and ! continue beration at greater than d'r etfual to 30 gis of a solution containing j greater than or equal to 7000 pp5eron or aquivalent until the required j SHUTDOWN MARGIN is restored 1 l $URvf!LLANCE RE00!REMD(75 4 i 1 ) 4.1.1.2 SHUTD0hel MARGIN shall be determined to be greater than or equal toj . " 2 ? 2 '

a. Within one hour after detection of an inoperable control rod (s) and at least.once per 12 hours thereafter while the rod (s) is inoperable.

If the inoperable control rod is innovable or untrippele, the ) SHUTDOWI MAAGIN shall be verified acceptable with an increased i allomence for the withdrawn worth of the immovable or untrippable j control rod (s). l b. At least once per 24 hcurs by consideration of the following factors:

1. Reacter coolant system boren concentration, 4

l 2. Control red positten, i 3. Reacter coolant system average teeMrature,

4. Fuel burnup based on gross thermal energy generation, i

j 5. Xenon concentration, and j

8. Samarium concentration.

3/4 1-3 AMEM0 MENT No. 26 FARLEY-UNIT 1

t fithin +he i EeS re of C ci lif ( 8 t L', (;g + }! a.n en + ke E r4 e f C y ~c t e L ihe ,(, S o i REACTIVITY CONTROL SYSTEMS C mif re d in f b cOLR . T he. y fin;/ up p e. r li g t hu be_ mad % ! MODERATOR TEMPERATURE COEFFICIENT ' l LIMITING CONDITION FOR OPERATION i l 3.1.1.3 Themoderatortemperaturecoefficient(MTC)shallbef

                         ' frl Less than or equal to 0.7 x 10** delta k/k/'F fer th: 11 ::d:                                                      .l 1

vithd: r.. 5:;innin;;-of :/ :10 life (30L), :::diti: for power levels up to 70% THERMAL POVER vith a linear ramp to O delta k/k/*F at 100% ' THERMAL POVER.

t. L;;; ;;;;ir; th:. 5.3 : 10~* d:lt: h/h/*? f:: th: :11 ::d: -

vi4hdravu -en4-et-eye r 14-1 H: (EOL) r-RATED-THERMAL P0"!.r. ' ' j :: di:10:. Bot ard { APPLICABILITY: !;::1!!:::10: 3.1.1.3.4 - MODES 1 :n4 2* onlyt l

                                                   .-Specific :i:   3.1.1.3.5A
                                                                                    - MODES 1, 2 aad 3 onlyt
                            '                                            E OL        UO 80L         (SpecifieU In 4he CUL A v

a. Vith the MTC more positive than the limit tf 3 1.1.3 : 05 :r operation in MODES 1 and 2 may proceed provided:

1. Control rod withdrawal limits are established and maintained sufficient to restore the MTC to within its limit within 24 hours or be in HOT STANDBY vithin the next 6 hours. These withdrav'al limits shall be in addition to the insertion limits of Specification 3.1.3.6.

2. The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the IffC has been restored to within its limit for the all rods withdrawn condition.

3. A Special Report is prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days, describing the value of the measured MTC, the interim control rod withdrawai i limits and the predicted average core burnup necessary for restoring the positive ifrC to within its limit for the all rods withdrawn condition. .s fic,d la lhe Cot A, V

b. With the NTC more negative than the limit Of 3.113 b 15: ;, be in 80T SEUTD0VN within 12 hours. l

With K,,, greater than or equal to 1.0 9 See Special Test Exception 3.10.3 FARLIT - UNIT 1 3/4 1-4 AMENDMENT NO. 57, M 92 l

REAmvm ColffROL SYSTEMS SURVEIL 1 m CE REQUIREMENTS 4.1.1.3 The wrc shall be determined to be witnin its limits during each fuel cycle as follows: 396coFoto ist THE CQ

a. The NTC shall be seasured and compared to the BOL limit g6-

                                               !;::1!h -ter 2 1 1.? 2,                      t ::, prior to initial operation above SE of RATED TIERMAL POWER, after each fuel loading.

b.Thewrcsha}1besensuredatanyTEIRMALPOWERandcomparedto 4.'". ".0' O.1.. L'i'"! ' ell ..l. 1 0.4... , " . . .- l

                                              -90UEE-eend646eek within 7 EFFD af ter reaching an equilibrium boron   concentration indicates                      of 300negative the ifrC is more                  ppe. In the thanevent3 '" -this  *^~ gosperison   '-'--

deMa sa 5, the WFC shall be recessured, an[Toepared to the EOL NTC 4fR ;R: 3.1.1.? 2 , at lesst once per 14 EFFD i limit ; duri the remainder of the fuel cycle. (1)

                                                                                                                                             ,                 .ne w igere p.s o ia W s g                                                                             rdEg M88#"E u            .r , m . u ,a r** Sna&-

(1) once the equillbrium boros concentration (all rods withdrawn, RATED TMN&L POWER condition) is 100. ppe or less, further esasurement of the IffC is accordance with 4.1.l'.3.b any be suspended, providing that the measured IffC at an equilibrius boros concentr , equalto100ppeislessnegativethang.0;10~gtionless.thanor d;l;; L'i'"". TWE /00 fPet Q Saddfo6LAMCE. J.sm o 7 sHsoforo ru nE C

4

FARLEY-tREIT 1 3/6 1-5 Amendment 16o. 28, 86

SI i REACTIVITY COIrfROL SYSTEMS Qpeorgo 10 pdw.coeL : Assidiron of s $$6rories

                     $5tffDOWN ROD INSERTICII LIMIT g-{f s yy e g .

LINITING COIIDITICII POR OPERATION v 3.1.3.5

                     '..;L;':;}.      .

All shutdown rods shall A be t".j .';L;r_; =; :: ~~~ ~' '-' APPLICA8ILITY: MODES 1* and 2*t. ACTI0ll usgance Sa w tnt tuss 4sksA.soon a r Spec,p,as su Twa g with a maximus of one shutdown rod :: fully ritt1 except for surveillancetestingpursuanttoSkification4.1.3.1.2,withinonehour either: g,g,,g 7,g 4,, y. .o.ygw.resar sess4saa Asmeer s#src.,sJe w 7ws (f,gg ,

s. :li; *n"! - "; :: , or b.

Declare the rod to be inoperable and apply Specification 3.1.3.1. 1,yv v tes wsadwea Laensi ' 596CoFof0 oN Tdd Q SURVIILLMICI R30UT*"L"f5 4.1.3.5 Each shutdown rod shall be determined to be ) 0.11,

                                                                                                                            ...:J...   ;;;;

??1 et ;r, '--'rrf r?: A

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

b. At least once per 12 hours thereafter.

                   *5ee Special Test Exceptions 3.10.2 and 3.10.3.

DVith K ,, greater than or equal to 1.0

FARLEY-tatIT 1 3/4 1-20 ANENDNENT NO. 83 t nr

                     =                  _-

i l l l

!         REACTIVITY CONTROL SYSTDIS i                                                                                                     !

CONTROL 200 INSERTION LIMi75 j

1 I

j LIMITING Cole! TION FOR OPERATION j l i J 43.1.3.6 The control banks shall be lietted in physical insertion as -"-- '- l 4

           " r n 0.1-1 rd 0.1--     .

i 1

l APPLICA8!LITY: N00E5 la and 2*f. Qs?nicososo su W6 N C.0LA.

l i j ACTION:

l ' With the control banks inserted beyond the above insertion limits, except for ! surveillance testing pursuant to Specification 4.1.3.1.2, either: i

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

b. Reduce THERMAL POWER within two hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the group position using the M M , or ,y ,, ,,,,7, gg,,,,,, ,, y y y l C. Se in at least NOT STAfGSY within 6 hours.

1 l SURVEILLANCE REQUIREMENTS i i f 4.1.3.5 .The position of each control bank shall be determined to be within

the insertion limits at least once per 12 hours ancept during time intervals when the Red Insertien Limit Monitor is inoperable, then verify the individual rod positions at least once per 4 hours.

I l i i l j j l

          "See special Test Exceptions 3.10.2 and 3.10.3 With K,ff greater then or equal to 1.0.

I i 1 1 , d 3/4 1-21 M W O WIT No. 26 FARLEY-UNIT 1

l

.l b

l 6 3wrc 3, l- l ( '5 Ef9 RCO ledenfiorto(/3 (q4+ gq cuere (PULLY WITHDRAWN-225 TO 231 STEPS, INCLUSNE) f 231

-/

225 \ . - . l _ ' / 200 j \ y j ,

                                                                    \                 / % aANK C                                                              j
                                                                                 /                                                                        / .-
                                  ,                                          f                                                                        f.

b 150 - g / / t / - 4 h / / T aANK D g / / g \ O 100 f: - g x/ ' l z f., -

                                                                                                 /               .

f i- -

                                       "                                              /                                  -
                                                                                /                                               .,
                                                                             /

f f .._ : _..

                                                                                               ._...._.g..         .         ..

g. 0

                                                   / A              '

t t \\ 0 / .2 .4 .8 .8 - 1.0 (PULLY NSERTED) N FRACTION OF RATED THERMAL POWER '- F' 3.1 1 Red Group insereon Limhs Verus Thermal Power Three Loop Operadon 9 FARLEY-UNIT 1 3/4 1-22 AMENUMENT NO. 83

I 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERINCE (AFD) LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shal. be maintained within the limits specified in "'%sC r OL$ i APPLICABILITY: MODE 1 above 50% of RATED THERMAL POWER', l ACTION:

a. With the indicated AXIAL FLUX DIFFERENCE outside of the limits 2.0 is specified in 7.Is_c_o;LR, tt

             'l. Either restore the indicated AFD to within the limits within 15 minutes, or

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minuts.s. ,

l SURVEILLANCE REQUIREMENTS i 4.2.1 The indicated AXIAL FLUX D17FERENCE shall be determined to be 1 within its limits by: l

a. Monitoring the indicated AFD for each OPERABLE excore channel:

1. At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and l

2. At least once per hour with the AFD Monitor Alarm inoperable. l

The indicated AFD shall be considered outside of its limits when at least 2 OPERABLE excore channels are indicating the AFD to be outside its limits.
- See Special Test Exception 3.10.2.

l FARLEY-UNIT 1 3/4 2-1 AMENDMENT NO.

( 5MS p a c)c 1nde/rb o,1 cdl g lch h\qn h, L Delete )I TtGVRE J 120 -

l. I 100 -

( 12, 100) (+9,100) Unacceptable acceptable Operation Operation 80 --

                                          /\             Acceptable X

o. \yperaton 60

(-30, 50) t l

                                                      !Y                          i\(+24,   )

N

                                            ;/                 ;

20 i / ! \l l 50 0 -30 -20 -10 0 10 20 30 40 50 Axial Flux Difference (Delta 1)% . 3 6-..;- M;:- m

                                               -"~ ". _*g r ] f m- ar
                                           -....m               . _

M T T '"

                                                                                  . - - - v e

FARLEY-UNIT 1 3/4 2-2 AMENDMENT NO.

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

I C I l POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - 1 (Z) LIMITING CONDITION FOR OPERATION whn the. l!Mb 5pedf.'ed i n + k t. COL- R. . l 3.2.2 Fg(Z) shall be g i bit:d by 't: f ell &' 7 --l-* * ~ h i p- _ i Z) 5 (L.11) (K(E)] for P > 0.5 for VANTAGE 5 fuel '

  \                                                F i

Fg (Z) s (4.9) ) for P 5 0.5 for VANTAGE 5 f' and f Fg(Z) 5 (L22) [K(Z)] P > 0.5 for fuel l P Fg (E) 5 [4.64) [K(E)) for 0.5 LOPAR fuel where P = THE POWER

THERMAL POWER , and ) is the function obtained from Figure 3.2-2 for iven

p re height location.

k hEELICABILITY: MODE *1 ACTION: 4 With Fg(Z) exceeding its limits

a. Reduce THERMAL 16 at least 14 for each in Fg(3) exceeds the limit within 15 minutes and similarly reduce the Power Range 4

Neutron Flux-Ei(th Trip setpoints within the next 4 hours; POWER OPERATION may proceed for up to a total of 72 hours; subsequent j POWER OPERATION may proceed provided the Overpower delta T Trip Setpoints have been reduced at least 14 for each 14 Fg(I) exceeds q the limit.

l 4: b. THERMAL POWER may be increased provided F (3) is demonstrated through incere mapping to be within its 1 it.

! SURVEILLANCE REQUIREME',!S 4.2.2.1 The provisions of specification 4.0.4 are not applicable. 4.2.2.2 Fg(E) shall be evaluated to determine if it is within its limit by: l ! a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER. i I . FARLEY-UNIT 1 3/4 2-3 AMENDMENT NO. i

                                                                                                                                               *i POWER DISTRIBUTION LIMITS
                            ~ . .

SURVEILLANCE REQUIREMENTS (Continued)

b. C Determining follows:

the computed heat flux hot channel factor Fg (gy, ,, Increase the measured Fg(Z) obtained from the power distribution map by 3% to account for manufacturing tolerances and further increase the value by 5% to account for measurement uncertainties. c. Verifying that FQ (Z), obtained in Specification 4.2.2.2b above, satisfies the relationship in Specification 3.2.2.

d. Satisfying the following relationships Fq C (Z ) s F "" x K (Z )

fo r P > 0.5 P x W (Z ) FqC (Z ) s F " " x K '( Z ) fo r P s 0.5 0.5 x W (Z ) Where Fg (2) is obtained in Specification 4.2.2.2b above, Fg is the Fg limit, K(2) is the normalized Fg(Z) as a function of core height, P is the fraction of RATED THERMAL POWER, and W(Z) is the cycle dependent function that accounts for power distribution transients encountered during normal operation. ro ***[%%%D - r .. . s

                                                                               %)M ^R%EC#lCD 3!!"_'[E2--

g,eg,g,gg, lu%E cg A5 I rovided in Figure 3.2-2 W(E) r a is or Limit i Report '

e. Measuring Fg(I) according to the following schedules

1. Upon achieving equilibrium conditions after exceeding by 20%

or more of RATED THERMAL POWER, the THERMAL POWER at which Fg(5) was last determined *, or

2. At least once per 31 Effective Full Power Days, whichever occurs first.

                *During power escalation after each fuel loading, power level may be increased until equilibrium conditions at any power level greater than or                                                        I equal to 50% of RATED THERMAL POWER have been achieved andia power distribution map obtained.

FARLEY-UNIT 1 3/4 2-4 AMENDMENT NO.

l POWER D2STRIBUTION L2MITS SURVE2LLANCE REQUIREMENTS (Continued)

f. With measurements indicating maximum Fo c(Z)'

t l over(Z) s K(Z)j l C has increased since the previous determination of Fg (Z) either of the following actions shall be taken:

1) Increase Fg (Z) by the Fg (Z) penalty factor specified in the cot R 7eeki..g Facter Lir.it 7,;re .t and verify that this value satisfies the relationship in specification 4.2.2.2d, or I C

2) Fg (Z) shall be measured at least once per 7 Effective Full Power Days until two successive maps indicate that maximum

                                        ' Fo *(Z)'       .    .       .

15 not increasing. over(Z) s K(Z)j l

g. With the relationships specified in specification 4.2.2.2d above not being satisfied:

1) Calculate the percent Fg(I) exceeds its limits by the following expressions Fj (Z) x W (Z) m axim um F ,,,

                                                                     -1     x 100 for P > 0.5 over Z           o
                      ,s                . P                  .)      .
                      .e                .

3 Fj (Z) x W (Z) m axim um -1 x 100 for P s 0.5, and over Z Fo,,,

                      ,s                . 0.5                    .s      .

2) The following action shall be taken:

Within 15 minutes, control the AFD to within new AFD limits p which are determined by reducing the AFD limits specified in W

 .t-Q (,ot.g^ y     0 . 0 . '. , " ici 71x Ciff;;::::, by 1% AFD for each percent Fg(Z) exceeds its limits as determined in specification 4.2.2.2g.1, Within 8 hours, reset the AFD alarm setpoints to these modified i

limits. FARLEY-UNIT 1 3/4 2-5 AMENDMENT NO.

i ,.' ( l5 PGtje i AdcEoo oI/g Ich bIgn k e i l CELI = 7f 1 ,, A

                                                                                                                                                                             /              1 1

] o.e,to 1 s.o. t.e

mummmmmmme===================== %

i s ,1 in.e. o. ass i s S :

, -s

! 0.6 - i

q N - ,i \ 0.2 \ 1

                                                                   .                                                                \

N

N\ 0 ' ' ' " i 0 2 4 8 8 10 12 l i MNM

Figure 3.2-2 K(3) leermalised F,(2) as a Function of Core Reig i

l '

)'

i FARLET - 15t!T 1 3/4 2 7 AISISISFr 110. 26, 92 1

5 POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEARENTHALPYHOTCHANNELFACTOR-F5H ' LIMITING CONDITION FOR OPERATION wi%in %e lW% SP*Clfed M W' CD'AO 3.2.3 F$Hshallbeli8 g ted by *he fell ^"4a; raletiaa=h87- y Th C 1.70 'l - O.? 'l - ?;; fer "* = N 5 f_e1.and-. p j Th f 1.20 'l

0.3 'l - ?;l f: _' ^".'_ ," f _: 1 -- p l

( y_

                                        "4C
                                               = :: ~~:~ 7     -

T"ZP2'.'.' T- C . # y APPLICABILITY: MODE 1 A ' ACTION: WithF5Hexceedingitslimit

a. Within 4 hours either:

1. RestoreF{H to within the above limitX(and demonstrate throughin-coremappingthatF5H is within its limit within 24 hours of exceeding the limit, or 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER l and reduce the Power Range Neutron Flux - High Trip 5etpoints to s 55% of RATED THERMAL POWER within the next 4 hours, and

b. Demonstrate through in-core mapping, if not previously performed pera.1above,thatF{H is within its !.init within 24 hours after exceeding the limit or reduce THERMAL 1 m to less than 5% of RATED THERMAL POWER within the next 2 hours, and

c. Identify and correct the cause of the out of limit condition prior l to increasing THERMAL POWER above the reduced limit required by a or b, above; subsegwnt POWER OPERATION may proceed provided that F$5 Lu demonstrated through in-core mapping to be within its limit at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior to exceeding this THERMAL POWER and within 24 hours after attaining 95% or greater RATED THERMAL POWER.

l l { l FARLEY-UNIT 1 3/4 2-8 AMENDMENT NO. l

l y { 3/4.1 REACTIVITY CONTROL SYSTEMS $ BASES

3/4.1.1 80 RATION CONTROL j 3/4.1.1.1 AND 3/4.1.1.2 55VTD0VN MARGIN .

i A safficient SBt/TDOWN MARGIN ensures that 1) the reactor can be made 1 suberitical from all operating conditions, 2) the reactivity transients

!                associated with postulated accident conditions are controllable withia

acceptable limits, and 3) the reactor vill be maintained sufficiently t

suberitical to preclude inadvertent criticality in the shutdown condition. SRtFfDOUN MARGIN requirements vary throughout core life as a function of i fuel depletion, RCS boros concentration, and RCS T . The aset restrictive condittom occurs at ROL, with T o a spac#@ at ao'loed operating su ruedOLA,,J and resulting uncontrol:,ed RCS cooldown. temperature, and is associated with a post In the analysia of this accident, ) a ainamus savauvwn Maisesef 1 "* 't: t is required to control the 4 reactivity transient. Ac$brdingly, the 55ttfDOWN MAacTa requirement is i based upon analysis this liettingWith assumpticas. condition T and is consistent with FSAR safety ' less than 200'F, the reactivity transients resulting from a postulated slela line break cooldown are minimal and a 46-4 owed 4eMe, h4 SEtrfDOWN MARGIN provides ad { i - C.4#dse toi#protection.) WS g 3/4.1.1.3 NODERATOR TENFERATURE COSFFICIENT , i The limitations on moderator temperature coefficient (NTC) are provided to ensure that the value of this coefficient remalas within the limiting condition assumed in the FSAR accident and transiest analyses. ' The NFC values of this specification are applicable to a specific set of plant conditions: accordingly, verification of NTC values at conditions other than those explicitly stated will require entrapolation to those conditions is order to permit an accurate comparises. i The most negative NfC value equivalent to the meet positive moderator density coeffieleet (SC) was obtained by incrementally correcting the MDC used in the FSAR analyses to nominal operating conditions. These corrections involveda (1) a conversion of the WC used in the FSAR safety analyses to its equivalent NFC, based on the rate of chnage of moderator density with temperature at RATED TERNAL PMR conditions, and (2) subtracting from this vains the largest differences la HTC observed betwees SE al7, rede withdrawn, RATED TIER &L POWER conditions, and those most adverse candittees of mo/A rator temperature and pressure, rod inserties, asial power skewtag, and xenos concentration that can occur in normal operation sad lead f o a significantly more negative BOL NTC at RATED TM POWLL Thera corrections transformed the MDC value pod in the FSAR safety analysee into the limiting NfC valua -f E 2 : 1^~ '92

               -b W.18. The surveillance requirement NfC valueg! EIS * !?"* '-"r I

b W A8. represents a conservative WTC value at co condittom of 300 ppe equilibrium boros concentration, and is obtei asking corregtions

                                                                                                   "          for
                                                                                                      ' ^ " -

J burn g and soluble beron to the limiting NTC ue, 1 -

;                                                                            $P6cw.60anJ'rH g .

FARLEY-UNIT 1 3 3/4 1-1 Amendment No. 26, 86 i

i I b l yyg , rest g 5.a&W t REACTIVITY CONTROL SYSTEMS h i um,r stee,Ame W we,jgid-RAsas b MODERATOR TENFERATURE C0EFFICIElff (Continued) l \ l Once the equilibrium boron concentration falls below 100 ppe, NTC 8 sensurementsb may be suspended provided the esasured NTC value at equilibriu9 i . 7 ;; iforon f:1:2concentration i'i'"7. < 100 ppe is less negative than limiting EOL NTC value Of The 'l diffe{ence

if between this value and the the maximum change in IffC between the 100 ppe equilibrium borosf:ix. L'i'"7 c I l

licensed end-of-cycle, including the effects of boros concentration i reduction, fuel depletion, and end-of-cycle coastdown. 1 i ! and near the end of the fuel cycle are adequate to confirm tha i remains within its limits since this coefficient changes slowly due j principally fuel burnup.to the reduction in RCS boros concentration associated with

i 4

1 1 3/4.1.1.4 MININtRf TENFERATURE FOR CRITICALITT 1 This specification ensures that the reactor will not be ande critical with the Reactor Coolant System average temperature less than 541'F. This limitation is within its is analystd required to ensure 1) the temperature ranmoderator temperature coefficient is within its normal operating range, ge, 2) the protective instrumentation 3 3) the F-12 interlock is above its

setpoint, 4) the pressuriser is capable of being in an OPERABLE status with

, a steam RT,,, bubble, and 5) the reactor pressure vessel is above its minious temperature. 4 i 3/4.1.2 BORATICII STSTEMS l' The boron injection system ensures that negative reactivity control is available during each mode of facility operation. The components required

to perfore this function include 1) borated water sources, 2 pumps, 3) separate flow paths, 4) boric acid transfer pumps,)and 5) ancharging emergency power supply fres OPERABLE diesel generators.

1 Vith the RCS average temperature above 200'F, a minimum of two boron 1 injection flow paths are required to ensure single functions) Ability in the event as assumed failure renders one of the flow path . wrable. The { boration capability of either flow path is sufficient to t x,.ade SEttfD0VN i tyncp:caatkfp 1 l 1 i i I . i , i FAR12T-talIT 1 3 3/4 1-2 Ad=="t No. 26, 86 i i i

   .m.,   .._-.-.._...m Ill O

i l REACTIVITY COMTROL SYSTEMS BASES l BORATION SYSTEMS (Continued) MARGIN from expected operating conditions M '..??" and cooldown to 200*F. Tne naximum exoected boration Gl'.; ,,- after zenon decay capability recutrement occurs at EOL from full power ecutlibrium xenon conditions and requires 11.336 gallons of 7000 ppm borated water from the boric acid storage tanks or 44,826 gallons of 2300 ppm borated water from the refueling water storage tank. Witn the RCS temperature below 200*F, one injection system is acceptable without single failure consideration on the basis of tne stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS system becomes inoperable. and positive reactivity changes in the event the single injection The limitation for a maximum of one centrifugal charging pump to be OPERABLE and tne Surveillance Requirement to verify all cnarging punos except i tne required OPERA 8LE pump to be inoperable below 180'F provides assurance that i a mass addition pressure transient cari be relieved by the operation of a single ' RHR relief valve. 33 ,pec;ge.l in %c Cot.g The baron c ability required below 200*F is sufficient to provide a SHUT 00wM MARGIM J '.". 2:^. -T. af ter xenon decay and cooldown from 200*F to 140*F. This conditipo requires either 2,000 gallons of 7000 ppm borated water f rom the boric acid storage tanks ,or 7,750 gallons of 2300 ppm borates water f rom the refueling water storage tank. The contained water volume limits include allowance for water not available ber ase of discharge line location and other physical enaracteristics. The limits on contained water volume and baron concentration of tne RWST also ensure a pH value of between 8.5 and 11.0 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. The OPERA 8!LITY of one boron injection system during REFUELING ensures that this system is ave 11able for reactivity control while in MODE 6. 3/4.1.3 MOVASLE CONTROL ASSEM8 LIES The specifications of this section ensure that (1) acceptat:1e power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) limit the potential effects of rod misalignment on associated accident analyses. OPERA 81LITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion 11mi.ts. FARLEY-UMIT 1 B 3/4 1 3 AMEN 0 MENT NO. 28.68 l

e: 3/4,2 PoWTR DISTR 280T2cN L2M2TS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (a) meeting the DNB design criterion during normal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding mechanical properties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200*F is not exceeded. The definitions of certain hot channel and peaking factors as used in these specifications are as follows: Fg(Z) Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods and measurement uncertainty. Ffg Nuclear Enthalpy Rise Hot C'hannel Factor, is defined as the ratio of the integral of linear power along the rod.with the highest integrated power to the average rod power. l ht Fg umiT set.amEb l 3/4.2.1 AXIAL FLUX DIFFERENCE i IN THE CoCE 6 The limite IAL FLUE DIFFERENCE (AFD) assure that the Fo(E) upped bound envelope ofp .,., ;_. / C Z , 1-2 2.;; ".. "T 2 times * : x _ _ 21_ _ _ l 2 A__^ __. r---- v 2- 1:: is not exceeded during either normal operation or in the event of xenon redistribution following power changes. Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer determines the one minute average of each of the OPERABLE encore detector outputs and provides an alarm message immediately if the AFD for 2 or more OPERABLE escore channels is outside the allowed AI operating space for RAOC o : ; "'- specified in 7

' E

and the THERMAL POWER is greater than 50% RATED T m POWER. j i

MT., C.013. ' m 0%)#4 FARLEY-UNIT 1 B 3/4 2-1 , AMENDMINT NO. a

                                                                                                          ]!i k.

POVED DISTRIBUTION LIMITS $ 4 BASES "

                       -                                                                                     l 3/4.2.2 and 3/4.2.3 CHANNEL FACTOR                HEAT TLUX HOT CHANNEL FACTOR. NUCLEAR ENTMALPY HOT                       <

The limits on heat flux hot channel factor, and nuclear enthalpy rise het ') channel facter ensure that 1) the design limit on peak local power density is not exceeded, 2) the DNB design criterion is met, and 3) in the event of a LOCA ; the peak fuel clad temperature will not exceed the 2200*F ECCS acceptance criteria limit. I Each of these is measurable but will normally only be determined ) periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic I surveillance is sufficient to insure that the limits are maintained provided:

a. Control rods in a single group move together with no individual 1 red insertion differing by more than 2 12 steps, indicated, from j the group demand position.

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

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

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

Th will be maintained within its limits provided conditions a. through

d. above are maintained. The relaxation of Fh as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits.

l When an Fg measurement is taken, an allowance for both experimental error and manufacturing tolerance must be made. An allowance of 5% is i appropriate for a full core map taken with the incore detector flux mapping ' system and a 3% allowance is appropriate for manufacturing tolerance. The heat flux hot channel factor Fg(Z) is measured periodically and increased by a cycle and height dependent power factor appropriate to RAOC operation, w(I), to provide assurance that the limit on the heat flux het channel factor Fg(I) Le met. W(E) accounts for the effects of normal operational transients within the AFD limits and was determined from expected power control maneuvers over the full range of burnup conditions in the core. WE W(Q fowmen Voe. HemM. qdeon Mb W g LImrTS Nhre. ?9tombEb 19 WE c.otR YEK htt.tF)C9mo4 4 .9 . \ . % . FARLEY-UNIT 1 3 3/4 2-2 AMENDMINT NO.

ll ADM8N8STRAT2VE CONTROLS l MONTHLY OPERATING REPORT 6.9.1.10 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORV's or safety valves, shall be submitted on a monthly basis to the Commission, pursuant to 10 CFR 50.4, no later than the 15th of each month following the calendar month covered by the report.

t N SE RT S m l 9.1.11 The cycle dependent function W(E) and the burnup dependent F C pena actors, required for calculation of FgC(E) specified in . .2, " Heat Flux Hot Ch Factor - Fg(E)," shall be documented in the ing Factor Limit i Report in accordan th the methodology in WCAP-102 -A, " Relaxation of '

Constant Axial Offset Con FQ surveillance ical specification," Rev. 1, ! February 1994 (H Proprietary). l The Peaking Factor Limit rt shall be provi the Commission, pursuant to 10 CFR 50.4, upon ance prior to each reload cycle to NODE 2). In the { event that imit would be submitted at some other time du ore life, it ' wil submitted upon issuance, unless otherwise exempted by the on. ANNUAL DIESEL CENERATOR RELIABILITY DATA REPORT 6.9.1.12 The number of tests (valid or invalid) and the number of failures to start on demand for each diesel generator shall be submitted to the NRC annually. This report shall contain the information identified in Regulatory Position C.3.b of NRC Regulatory Guide 1.108, Revision 1, 1977. l FARLEY-UNIT 1 6-19 AMINDKENT NO. l l

INSERT B Page 6-19 CORE OPERATING LIMITS REPORT 6.9.1.11 Core operating limits shall fee established and documented in the CORE OPERATING LIMITS REPORT before eace. reload cycle or any remaining part of a reload cycle for the following:

1. SHUTDOWN MARGIN limit for MODES 1, 2, 3, and 4 for Specification 3/4.1.1.1,

2. SHUTDOWN MARGIN limit for MODE 5 for Specification 3/4.1.1.2,

3. Moderator Temperature coefficient BOL and ROL limits and 300 ppm and 100 ppm surveillance limits for Specification 3/4.1.1.3,

4. Shutdown Bank Insertion Limit for Specification 3/4.1.3.5,

5. Control Bank Insertion Limits for Specification 3/4.1.3.6,

6. Axial Flux Difference limits for Specification 3/4.2.1,

7. Heat Flux Hot Channel Factor Fg limits, K(Z) figure, W(Z) values, and FQ(Z) Penalty Factors for Epocification 3/4.2.2,

8. Nuclear Enthalpy Rise Hot Channel Factor limits, FAH P, and Power Factor Multiplier, PFAH, for Specification 3/4.2.3.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC in , 1. WCAP-9272-P-A, " Westinghouse Reload Safety Evaluation Methodology," July 1985 (H Proprietary). (Methodology for Specifications 3.1.1.1 - Shutdown Margin - Tavg > 200*F, 3.1.1.2 - Shutdown Margin - Tavg s 200*F, 3.1.1.3 - Moderator Temperature coefficient, 3.1.3.5 - Shutdown Bank Insertion Limit, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.) l

2. WCAP-10216-P-A, Rev. lA, " Relaxation Of Constant Axial Offset Control /

Fg Surveillance Technical Specification," February 1994 (H Proprietary). (Methodology for Specifications 3.2.1 - Axial Flux Difference and 3,2.2 - Heat Flux Hot Channel Factor.)

3. WCAP-10266-P-A, Rev. 2, "The 1981 Version of Westinghouse Evaluation Model Using BASH Code," March 1987 (H Proprietary).

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

e-s INDEX DEFINITIONS SECTION EhGK 1.0 DEFINITIONS 1.1 ACTION........................................................ 1-1 , 1.2 AXIAL FLUX DIFFERENCE......................................... 1-1 CALIBRATION........................................... 4 1.3 CHANNEL 1-1 1.4 CHANNEL CHECK................................................. 1-1 1.5 CHANNEL FUNCTION TEST......................................... 1-1

j. 1.6 CONTAINMENT INTEGRITY......................................... 1-2 a 1.7 CONTROLLED LEAKAGE............................................ 1-2 1.8 CORE ALTERATION............................................... 1-2 j

1.8a CORE OPERATING LIMITS REPORT.................................. 1-2 l 1.9 DOSE EQUIVALENT I-131... .................................... 1-2 1.10 Y-AVERAGE DISINTEGRATION ENERGY............................... 1-3 1.11 ENGINEERED SAFETY FEATURES RESPONSE TIME...................... 1-3 1.12 FREQUENCY NOTATION............................................ 1-3 1.13 (Deleted 1-3 5 1.14 1.15 IDENTIFIED LEAKAGE...............J........o)..................

                                                                                   .................            1-3 (Deleted).................... 1-4 1.16 . _ -_--_ _ _ ________
                                       -.-....--...o
                                                                         ..._... ______.._(Deleted).
                                                                                    -. .-..-                    1-4

, 1.17 OFFSITE DOSE CALCULATION MANUAL (ODCM)....................... 1-4 t 1.18 OPERABLE - OPERABILITY........................................ 1-4 1.19 OPERATIONAL MODE - MODE....................................... 1-5 1.20 PHYSICS TESTS...f............................................ 1-5 1.21 PRESSURE BOUNDARY LEAKAGE..................................... 1-5 1.22 PROCESS CONTROL PROGRAM (PCP)................................. 1-5 1.23 PURGE - PURGING............................................... 1-5 1.24 QUADRANT POWER TILT RATIO..................................... 1-5 1.25 RATED THERHAL POWER........................................... 1-6 1.26 REACTOR TRIP SYSTEM RESPONSE TIME............................. 1-6 4 1.27 REPORTABLE EVENT.............................................. 1-6 l 1.28 SHUTDOWN MARGIN............................................... 1-6 1.29 SOtTDTFTCRTTON- ( De leted ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6 1.30 SOURCE CHECK.................................................. 1-6 1.31 STAGGERED TEST BASIS.......................................... 1-6 1.32 THERMAL POWER................................................. 1-7 1.33 UNIDENTIFIED LEAKAGE......................................... 1-7 1.34 VENTILATION EEHAUST TREATMENT SYSTEM.......................... 1-7 1.35 VENTING....................................................... 1-7 TABLE 1.1 OPERATIONAL MODES 1-8 TABLE 1.2 FREQUENCY NOTATION 1-9 l 4 4 FARLEY-UNIT 1 I AMENDMENT NO.

I l INDIX l ADMINISTRATIVE CONTROLS SECTION IAGI Review................................................... 6-10 i Audits................................................... 6-11 Authority................................................ 6-12 Records.................................................. 6-12 6.5.3 TECHNICAL REVIEW AND CONTROL Activities............................................... 6-12 l Records.................................................. 6-13 6.6 REPORTABLE EVENT ACTION..................................... 6-14 6.7 SAFETY LIMIT VIOLATION 6-14 i 6.8 PROCEDURES AND PROGRAMS.........................,........... 6-14 0 9 _.REPORIING REQUIREMENTS 6.9.1 ROUTINE REPORTS

Startup Report .......................................... 6-15a Annual Report............................................ 6-16 1

Annual Radiological Environmental Operating Report....... 6-17 Annual Radioactive Effluent Release Report............... 6-17 Monthly Opierating Report . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-19 Core Operating Limits Report............................. 6-19 l j Annual Diesel Generator Reliability Data Report.......... 6-19a l Annual Reactor Coolant System Specific Activity Report... 6-20 Annual Sealed Source Leakage Report...................... 6-20 6.9.2 SPECIAL REPORTS........................................... 6-20 6.10 RECORD RETENTION............................................ 6-20 6.11 RADIATION PROTECTION PROGRAM................................ 6-21a 6.12 HIGH RADIATION AREA......................................... 6-22 l l . FARLEY-UNIT 1 XIX AMENDMENT NO. 1 j l

1 DEFINITIONS s l CONTAINMENT INTEGRITY j 1.6 CONTAIMMENT INTEGRITY shall exist when: t 4 a. All penetrations required to be closed during accident conditions are j either:

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

i 2) Closed by manual valves, blind flanges or deactivated automatic valves secured in their closed positions, except as provided in Table 3.6-1 of specification 3.6.3, ) l

b. All equipment hatches are closed and sealed, 1

j c. Each air lock is OPERABLE pursuant to specification 3.6.1.3, =t f d. The containment leakage rates are within the limits of specification ] 3.6.1.2, and l t i e. The sealing mechanism associated with_each penetration (e.g., welds, ! 1 bellows or O-rings) ie OPERABLE. l . l j CONTROLLED LEAKAGE i j 1.7 CONTROLLED I m 3?? shall be that seal water flow supplied to the reactor } coolant pump seals. ! l j CORE ALTERATION i 4 ! 1.8 CORE ALTERATION shall be the movement or manipulation of any component j within the reactor pressure vessel with the vessel head removed and fuel in the ; vessel. Suspension of OORE ALTERATION shell not preclude completion of movement of e component to a safe conservative position. J CORE OPERATING LIMITS REPORI l 1.8a The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with specification 6.9.1.11. Unit operation within these operating limits is addressed in individual specifications. DOSE EQUIVALENT I-131 1.9 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134 and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table E-7 of Regulatory Guide 1.109, Revision 1, 1977. FARLEY-UNIT 1 1-2 AMENDMENT NO.

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

i: 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - Tava >200'F LIMITING COWOITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to the limit specified in the COLR for 3 loop operation. APPLICABILITY: MODES 1, 2*, 3, and 4. ACTION: With the SHUTDOWN MARGIN less than the limit specified in the COLR, immediately l initiate and continue boration at greater than or equal to 30 gym of a solution containing greater than or equal to 7000 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS l 4.1.1.1.1 The SHUTDOWN MARGIN shall be' determined to be greater than or equal to the limit specified in the COLRs l

a. Within one hour after detection of an inoperable control rod (s) and I at least once per 12 hours thereafter while the rod (s) is inoperable. I If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the innovable or untrippable control rod (s).

b. When in MODE 1 or MODE 2 with K gg greater than or equal to 1.0, at least once per 12 hours by verifying that control bank position is within the insertion limits of Specification 3.1.3.6.

c. When in MODE 2 with K,gg less than 1.0, within 4 hours 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 St RATED THERMAL POWER after each fuel loading, by consideration of the factors of a below, with the control banks at the maximum insertion limit of Specification 3.1.3.6.

i L i a I

*See Special Test Exception 3.10.1. .

FARLEY-UNIT 1 3/4 1-1 AMENDMENT NO.

REACTIVITY CONTROL SYSTEMS i SHUTDOWN MARGIN - T ava s 200'F LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to the limit specified in the COLR. l l APPLICABILITY: MODE 5. l ACTION: With the SHUTDOWN MARGIN less than the limit specified in the COLR, immediately l I initiate and continue boration at greater than or equal to 30 gym of a solution containing greater than or equal to 7000 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be dstormined to be greater than or equal to the limit specified in the COLR:

a. Within one hour after detection of an inoperable control rod (s) and at least once per 12 hours thereafter while the rod (s) is inoperable.

If the inophrable control rod is iemovable or untrippable, the SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s).

b. At least once per 24 hours by consideration of the following factors:

1. Reactor coolant system boron concentration,

2. Control rod position, l

l

3. Reactor coolant system average t wperature,

4. Fuel burnup based on gross thermal energy generation,

5. Xenon concentration, and

6. Samarium concentration.

FARLEY-UNIT 1 3/4 1-3 AMENDMENT NO.

t REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION i 3.1.1.3 The moderator temperature coefficient (MTC) shall be within the beginning of cycle life (BOL) limit and the end of cycle life (EOL) limit specified in the COLR. The maximum upper limit shall be less than or equal to

                 ~4 0.7 x 10       delta k/k/'F for power levels up to 70% THERMAL POWER with a linear ramp to O delta k/k/*F at 100% THERMAL POWER.

APPLICABILITY: BOL limit - MODES 1 and 2* only#. EOL limit - MODES 1, 2 and 3 only#. ACTION: 1

a. With the MTC more positive than the BOL limit specified in the COLR, l

operation in MODES 1 and 2 may proceed provided:

1. Control rod withdrawal limits are established and maintained sufficient to restore the MTC to within its limit within 24 hours or be in HOT STANDBY within the next 6 hours. These withdrawal limits shall be in addition to the insertion limits j of Specification 3.1.3.6.

I

2. The control rods are maintained within the withdrawal limits l established above until a subs,equent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition.

3. A Special Report is prepared and submitted to the Commission pursuant to Specification 6.9.2 within 1C days, describing the value of the measured MTC, the interim control rod withdrawal limits and the predicted average core burnup necessary for '

restoring the positive MTC to within its limit for the all rods withdrawn condition.

b. With the MTC more negative than the EOL limit specified in the COLR, l be in HOT SHUTDOWN within 12 hours.

With K gg greater than or equal to 1.0.

     # See Special Test Exception 3.10.3.

l l

                                                                             'e FARLEY-UNIT 1                               3/4 1-4                        AMENDMENT NO.

t REACTIVITY CONTROL SYSTEME SURVEILLANCE REQUIREMENTS I 4.1.1.3 The MTC shall be deteiinined to be within its limits during each l fuel cycle as follows: i

a. The MTC shall be measured and compared to the BOL limit specified j in the COLR, prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading.

b. The MTC shall be measured at any THERMAL POWER and compared to the 300 ppe MTC surveillance limit specified in the COLR l within 7 IFPD after reaching an equilibrium boron concentration of 300 ppe. In the event this comparison indicates the MTC is more negative than the 300 ppe MTC surveillance limit specified in the COLR, the MTC shall be remeasured, and compared to the EOL MTC limit specified in the COLR, at least once per 14 EFPD during the !

remainder of the fuel cycle. (1) l (1) Once the equilibrium boron concentration (all rods withdrawn, RATED THERMAL POWER condition) is 100 ppe or less, further measurement of the MTC in accordance with 4.1.1.3.b may be suspended, providing that the measured MTC at i an equilibrium boron concentration less than or equal to 100 ppa is less ' negative than the 100 ppe MTC surveillance limit specified in the COLR. l FAKLEY-UNIT 1 3/4 1-5 AMENDMENT NO.

REACTIVITY CONTROL SYSTEMS j SHUTDOWN ROD INSERTION LIMIT ] LIMITING CCEDITION FOR OPERATION I l 1 3.1.3.5 All shutdown rods shall be limited in physical insertion as specified in the 00LA. APPLICABILITY: MODES 1* and 2*#. ACTION: With a maximum of one shutdown rod inserted beyond the insertion limit specified in the COLR, except for surveillance testing pursuant to Specification 4.1.3.1.2, within one hour either: a Restore the rod to within the insertion limit specified in the COLR, or

b. Declkre the rod to be inoperable and apply Specificttion 3.1.3.1.

SURVEILLANCE REQUIREMENTS mesaur - - ~ ----- - - - - - - -- - --- - 4.1.3.5 Each shutdown rod shall be determined to be within the insertion limit specified in the COLR

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

b. At least once per 12 hours thereafter.

l l

*See Special Test Exceptions 3.10.2 and 3.10.3.

With Keff greater than or equal to 1.0.

l FARLEY-UNIT 1 3/4 1-20 AMENDMENT NO. l l i

i REACTIVITY CONTROL SYSTEMS CONTROL ROD INSERTION LIMITS i LIMITING CONDITION FOR OPERATION i ) 3.1.3.6 The control banks shall be limited in physical insertion as j specified in the COLR. l l APPLICABILITY: MODES 1* and 2*#, i ACTION: 1 With the control banks inserted beyond the above insertion limits, except for surveillance testing pursuant to Specification 4.1.3.1.2, either i a. Restore the control banks to within the limits within two hours, or b. Reduce THERMAL POWER within two hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the group position using the insertion limits specified in the COLR, or l

c. Be in at least HOT STANDBY within 6 hours.

SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours except during time intervals within the Rod Insertion Limit Monitor is inoperable, then verify the individual rod positions at least once per 4 hours.

See Special Test Exceptions 3.10.2 and 3.10.3.

 # With K,gg greater than or equal to 1.0.

FARLEY-UNIT 1 3/4 1 21 AMENDMENT NO. I

FIGURE 3.1-1 I l (This Figure intentionally left blank.) l 1

FARLEY-UNIT 1 3/4 1-22 AMENDMENT NO.

x l 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AZ T M. FLUX DIFFERENCE (AFD) LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the limits specified in the COLR*. g APPLICABILITY: MODE 1 above 50% of RATED THERMAL POWER **. I ACTION: a. With the indicated AXIAL FLUX DIFFERENCE outside of the limits specified in the COLR. l l 1. Either restore the indicated AFD to within the limits within 15 minutes, or 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes. SURVEILLANCE REQUIREMENTS 4.2.1 The indicated AXIAL FLUX DIFFERENCE shall be determined to be l within its limits by: 1

a. Monitoring the indicated AFD for each OPERABLE excore channel:

1. At least once per 7 days when the AFD Monitor Alarm is l OPERABLE, and i

2. At least once per hour with the AFD Monitor Alarm inoperable.

1 i l

The indicated AFD shall be considered outside of its limits when at least 2 OPERABLE excore channels are indicating the AFD to be outside its limits.

             **See Special Test Exception 3.10.2.

4 FARLEY-UNIT 1 3/4 2-1 AMENDMENT NO.

a l l t (This page intentionally left blank.) l l 1 l w l 1 1

l l FARLEY-UNIT 1 3j4 2~2 AMENDMENT NO.

                                                                                     *I
                                                                                     'l l

POWER DISTRIBUTION LIMITS l 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F % LIMITING CONDITION FOR OPERATION 3.2.2 Fg(I) shall be within the limits specified in the OOLE. l APPLICABILITY: MODE 1. ACTION: With F g(E) exceeding its limits a. Reduce THERMAL POWER at least 14 for each 1%gF (I) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours; POWER OPERATION may proceed for up to a total of 72 hours; subsequent POWER OPERATION may proceed provided the Overpower delta T Trip Setpoints have been reduced at least 14 for each 1% Fg(E) exceeds the limit.

b. THERMAL POWER may be increased provided Fg(I) is demonstrated through incore mapping to be within its limit. e SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable. 4.2.2.2 Fg(E) shall be evaluated to determine if it is within its limit by

a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than St of RATED THERMAL POWER. j FARLEY-UNIT 1 3/4 2-3 AMENDMENT NO.

C POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) i b. Determining the computed heat flux hot channel factor Fg (E), as follows: Increase the measured Fg(E) obtained from the power distribution map by 34 to account for manufacturing tolerances and further increase the value by 5% to account for measurement encertainties.

c. Verifying that Fg (E), obtained in Specification 4.2.2.2b above, satisfies the relationship in Specification 3.2.2.

d. Satisfying the following relationships Fj(Z) s F "" x K (Z ) fo r P > 0.5 P x W (Z )

l F8(Z) s F "" x K (2 ) fo r P s 0.5 0.5 x W (Z) l Where Fg (E) is obtained in Specification 4.2.2.2b above, Fg RTP g, l the Fg limit, K(I) is the normalised Fg(E) as a function of core height, P fa the fraction of RATED THERMAL POWER, and W(E) is the cycle dependent function that accounts for power distribution transients encountered during normal operation. Fg , K(E), and W(3) are specified in the COLR as per Specification 6.9.1.11. l

e. Measuring Fg(I) according to the following schedules

1. Upon achieving equilibrium conditions after exceeding by 20%

or more of RATED THERMAL POWER, the THERMAL POWER at which Fg(E) was last determined *, or

2. At least once per 31 Effective Full Power Days, whichever occur; first.

                        *During power escalation after each fuel loading, power level may be increased until equilibrium conditions at any power level greater than or equal to 50% of RATED THERMAL POWER have been achieved and a power distribution map obtained.

t FARLEY-UNIT 1 3/4 2-4 AMENDMENT NO.

POWER DISTRIBUTION LIMITS l SURVEILLANCE REgUIREMENTS (Continued)

f. With measureswnts indicating maximum

                                     'F nC(Z)'

over(Z) ( K(Z)j has increased since the previous determination of Fg (Z) either of the following actions shall be taken:

1) Increase Fg (E) by the Fg (Z) penalty factor specified in tha COLRandverifythatthisvaluesatisfiestherelationshipin!

Specification 4.2.2.2d, or

2) Fg (Z) shall be measured at least once per 7 Iffective Full Power Days until two successive maps indicate that maximum

                                         'F oC(Z)'

is not increasing. over(Z) s K(Z)cj

g. With the relationships specified in Specification 4.2.2.2d above not being satisfiedt

1) Calcylate the percent Fg(E) exceeds its limits by the following expressions m axim um F"(Z) n x W (Z) -1 x 100 for P > 0.5 over Z Fo,1,

                       .\                 . P               .)

e . 3 m axim um Fo (Z ) x W (Z) - 1 x 100 for P s 0.5, and over Z Fn,7,

                          <                  0.5              .)       ,

2) The following action shall be taken:

Within 15 minutes, control the AFD to within new AFD limits which are determined by reducing the AFD limits specLfied in l the CoLR by 1% AFD for each percent Fg(E) exceeds its limits j as determined in Specification 4.2.2.2g.l. Within 8 hours, l reset the AFD alarm setpoints to these modified limits. l l l l FARLEY-UNIT 1 3/4 2-5 AMENDMENT No.

4 i I 1 1 l l l r I (This page inte.stionally left blank)

                                                                     ]
               /                                                     l l

FARLEY-UNIT 1 3/4 2-7 AMENDMENT NO.

1 POW 5R DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR ENTHALPY HOT CHANNEL FACTOR - Fh LIMITING 00NDITION FOR OPERATION 3.2.3 Fh shall be within the limits specified in the COLR. l l l APPLICABILITY: MODE 1. ACTION: With Fh exceeding its limits

a. Within 4 hours either:

1. Restore Ph to within the above limit, and demonstrate through in-core mapping that Fh is within its limit within 24 hours of exceeding lthe limity or 2.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoints to s 55% of RATED THERMAL POWER within the next 4 hours, and

b. Demonstrate through in core mapping, if not previously performed per a.1 above, that Fh is within its limit within 24 hours after exceeding the limit or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours, and

c. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by a or b, above; subsequent POWER OPERATION may proceed provided that Fh is demonstrated through in-core mapping to be within its limit at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior to exceeding this THERMAL POWER and within 24 hours after attaining 95% or greater RATED THERMAL POWER.

1 I ! FARLEY-UNIT 1 3/4 2-8 AMENDMENT NO. l l

3.4.1 REACTIVITY CONTROL SYSTEMA a

mASES j 3/4.1.1 BORATION CONTROL 1

3/4.1.1.1 =a 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made i suberitical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within a acceptable limits, and 3) the reactor will be maintained sufficiently j suberitical to preclude inadvertent criticality in the shutdown condition. ! SHUTDOWN MARGIN requirements vary throughout core life as a function of i fuel depletion, RCS boron concentration, and RCS T avg. The most ! restrictive condition occurs at EOL, with T yg at no load operating , temperature, and is associated with a postulated steam line break accident f and resulting uncontrolled RCS cooldown. In the analysis of this accident, a minimum SHUTDOWN MARGIN as specified in the COLR is required to control the l reactivity transient. Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis j! assumptions. With T.,9 less than 200*F, the reactivity transients resulting from a postulated steam line break cooldown are minimal and a lower SHUTDOWN MARGIN (specified in the COLR) provides adequate protection. i

3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT d

l The limitations on moderator temperature coefficient (MTC) are provided to j ensure that the value of this coefficient remains within the limiting j condition assumed in the FSAR accident and transient analyses. 4 a The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at conditions

other than those explicitly stated will require extrapolation to those l conditions in order to permit an accurate comparison.

The most negative MTC value equivalent to the most positive moderator , density coefficient (MDC) was obtained by incrementally correcting the MDC j used in the FSAR analyses to nominal operating conditions. These

corrections involved: (1) a conversion of the MDC used in the FSAR safety
analyses to its equi"alent MTC, based on the rate of change of moderator

] density with tempercure at RATED THERMAL POWER conditions, and (2) j subtracting from this value the largest differences in MTC observed j between EOL, all rods withdrawn, RATED THERMAL POWER conditions, and those { most adverse cortditions of moderator temperature and pressure, rod l insertion, axial power skewing, and xenon concentration that can occur in } normal operation and lead to a significantly more negative EOL MTC at ! RATED THERMAL POWER. These corrections transformed the MDC value used in j the FSAR safety analyses into the limiting MTC value specified in the COLR. The surveillance requirement MTC value specified in the COLR represents a conservative MTC value at a core condition of 300 ppe equilibrium boron concentration, and is obtained by making corrections for burnup and soluble j boron to the limiting MTC value specified in the COLR. l l t $ FARLEY-UNIT 1 8 3/4 1-1 AMENDMENT NO. 1

_ _ _ _ ___ m ._.- _ _ . . . _ _ ._ __ -. _ . _ _ - . . _ _ _ _ _ _ _ . . - . _ . _ . _ _ - _ _ _ . . . . 3-REACTIVITY CONTROL SYSTEMS BASES MODERATOR T M ERATURE COEFFICIENT (Continued) l Once the equilibrium boron concentration falls below 100 ppe, MTC measurements may be suspended provided the measured MTC value at an j equilibrium boron concentration s 100 ppe is less negative than the 100 ppm MTC surveillance limit specified in the COLR. The difference between this value l and the limiting EOL MTC value conservatively bounds the mar

mum change in MTC j between the 100 ppe equilibrium boron concentration (all rods withdrawn, RATED THERMAL POWER condition) and the licensed end-of-cycle, including the effects of boron concentration reduction, fuel depletion, and end-of-cycle coastdown.

The surveillance requirements for measurement of the MTC at the beginning ! and near the end of the fuel cycle are adequate to confirm that the MTC l remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with l fuel burnup. 3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactcr Coolant System average temperature less than 541*F. This limitation is required to ensure 1) the moderator temperature coefficient is within its analyzed temperature range, 2) the protective instrumentation is within its normal operating range, 3) the P-12 interlock is above its setpoint, 4) the prossgriser is capable of being in an OPERABLE status with a steam bubble, and 5) the reactor pressure vessel is above its minimum RTNDT temperature. 3/4.1.2 BORATION SYSTEMS The boron injection system ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include 1) borated water sources, 2) charging pumps, 3) separate flow paths, 4) boric acid transfer pumps, and 5) an emergency power supply from OPERABLE diesel generators. With the RCS average temperature above 200*F, a minimum of two boron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable. The boration capability of either flow path is aufficient to provide the required SHUTDOWN MARGIN from expected operating conditions after xenon decay and cooldown to 200*F. The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires 11,336 gallons of 7000 ppe borated water from the boric acid storage tanks or 44,826 gallons of 2300 ppe borated water from the refueling water storage tank. l l r FARLEY-UNIT 1 B 3/4 1-2 AMENDMENT NO. l l

                                                              ~             -                                    -

REACTIVITY. CONTROL SYSTE41 BASE 3 BORATION SYSTEMS (Continued) With the RCS temperature below 200*F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single injection l system becomes inoperable. The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable below 180'F provides assurance that a mass addition pressure transient can be relieved by the operation of a single RHR relief valve. The boron capability required below 200*F is sufficient to provide a SHUTDOWN MARGIN as specified in the COLR after menon decay and cooldown from l 200'F to 140*F. This condition requires either 2,000 gallons of 7000 ppe borated water from the boric acid storage tanks or 7,750 gallons of 2300 ppe borated water from the refueling water storage tank. The contained water volume limits include allowance for water not avsilable because of discharge line location and other physical characteristics. The limits on co5tained water volume and boron concentration of the RWST also ensure a pH value of between 8.5 and 11.0 for the solution recirculated within containment after a LOCA. This pH band minimises the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. The OPERABILITY of one boron injection system during REFUELING ensures that this system is available for reactivity control while in MODE 6. 3/4.1.3 MOVA#ff CGuimOL ASSEMBLIES The specifications of this section ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) limit the potential effects of rod misalignment on associated accident analyses. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits. l l 1 FARLEY-UNIT 1 B 3/4 1-3 AMENDMENT NO. ; i

3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (a) meeting the DNB design criterion during normal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding mechanical properties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200'F is not exceeded. The definitions of certain hot channel and peaking factors as used in these specifications are as follows: Fg(E) Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation I divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods and measurement uncertainty. F"g Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power. 3/4.2.1 AITAL FLUX DIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the Fg(E) upper bound envelope of the Fg limit specified in the OOLR times K(E) is not l exceeded during either normal operation or in the event of xenon redistribution following power changes. Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. S a computer determines the one minute average of each of the OPERABLE encore detector outputs and provides an alarm message immediately if the AFD for 2 or more OPERABLE encore channels is outside the allowed AI operating space for RAOC operation specified in the COLR and the THERMAL POWER is greater than l 50% RATED THERMAL POWER. FARLEY-UNIT 1 8 3/4 2-1 AMENDMENT NO.

POWER D2STRIBUTION LIMITS s' l' a BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR, NUnn_M ENTHALPY HOT CHANNEL FACTOR The limits on heat flux hot channel factor, and nuclear enthalpy rise hot channel factor ensure that 1) the design limit on peak local power density is , not exceeded, 2) the DNB design criterion is met, and 3) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200*F ECCS acceptance criteria limit. l Each of these is measurable but will normally only be determined I periodically as specified in specifications 4.2.2 and 4.2.3. This periodic su rveillance is sufficient to insure that the limits are maintained provided:

a. Control rods in a single group move together with no individual I rod insertion differing by more than i 12 steps, indicated, from the group demand position.

3

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

l ~

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

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

F$gwillbemaintainedwithinitslimiteprovidedconditionsa.through

d. above are maintained. The relaxation of Fh as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits.

When an Fg measurement is taken, an allowance for both experimental error and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full core map taken with the incore detector flux mapping system and a 3% allowance is appropriate for manufacturing tolerance. The heat flux hot channel factor Fg(I) is measured periodically and increased by a cycle and height dependent power factor appropriate to RAOC operation, W(5), to provide assurance that the limit on the heat flux hot channel factor Fg(I) is met. W(E) accounts for the effects of normal operational transients within the AFD limits and was determined from expected power control maneuvers over the full range of burnup conditions in the core. The W(E) function for normal operation and the AFD limits are provided in the COLR , per specification 6.9.1.11. j FARLEY-UNIT 1 B 3/4 2-2 AMENDMENT NO. !

ADMINISTRATIVE CONTROLS MONTHLY OPERATING REPORT 6.9.1.10 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORV's or safety valves, shall be submitted on a monthly basis to the Coamaission, pursuant to 10 CFR 50.4, no later than the 15th of each month following the calendar month covered by the report. CORE OPERATING LIMITS REPORT 6.9.1.11 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the followings i ! 1. SHUTDOWN MARGIN limit for MODES 1, 2, 3, and 4 for Specification 3/4.1.1.1, 1 1

2. SHUTDOWN MARGIN limit for MODE 5 for Specification 3/4.1.1.2, l

3. Moderator Temperature coefficient BOL and BOL limits and 300 ppe and I 100 ppe surveillance limits for specification 3/4.1.1.3,

4. Shutdown Bank Insertion Limit for Specification 3/4.1.3.5, j 5. Control Bank Insertion Limits for Specification 3/4.1.3.6,

6. Axial Flux Dif fea;ence limits for Specification 3/4.2.1,

7. Heat Flux Hot Channel Factor Fg limits, K(I) figure, W(I) values, and l

Fg(E) Penalty Factors for Specification 3/4.2.2,

8. Nuclear Enthalpy Rise Hot Channel Factor limits, Fag , and Power l Factor Multiplier, PFAH, for Specification 3/4.2.3.

The analytical methods used to determine the core operating limits shall be l those previously reviewed and approved by NRC inn i I

1. WCAP-9272-P-A, " Westinghouse Reload Safety Evaluation I Methodology," July 1985 (M Proprietary).

l (Methodology for Specifications 3.1.1.1 - Shutdown Margin - Tavg > 200'F, 3.1.1.2 - Shutdown Margin - Tavg 5; 200*F, 3.1.1.3 - Moderator Temperature l l Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limit, 3.1.3.6 - Control Bank I l Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.) l

                                                                                                        'S g

FARLEY-UNIT 1 6-19 AMENDMENT NO.

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

T1

1 ADMINISTRATIVE CONTROLS S

2. WCAP-10216-P-A, Rev. lA, " Relaxation Of Constant Axial Offset Control /

Fg Surveillance Technical Specification," February 1994 (M Proprietary).

,                        (Methodology for Specifications 3.2.1 - Axial Flux Difference and 3

3.2.2 - Heat Flux Hot Channel Factor.)

'l

3. WCAP-10266-P-A, Rev. 2, "The 1981 Version Of Westinghouse Evaluation i Model Using BASH Code," March 1987 (M Proprietary).

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.) ] 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 accidont 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. - 4 ANNUAL DIESEL GENERATOR RELIABILITY DATA REPORT 1 s 6.9.1.12 The number of tests (valid or invalid) and the number of failures to 1 start on demand for each diesel generator shall be submitted to the NRC annually. This report shall contain the information identified in Regulatory Position C.3.b of NRC Regulatory Guide 1.108, Revision 1, 1977. 1 a .i 5 i FARLEY-UNIT 1 6-19a AMENDMENT NO.

                                                                   'l FNP Unit 2 Technical Speci6 canons I

COLR Implementation Channed Paaan

Umt 1 Revision I %b XIX Replace Page 1-2 Replace Page 3/41-1 Replace Page 3/41-3 Replace Page 3/41-4 Replace Page 3/41-5 Replace Page 3/41-20 Replace Page !!41-21 Replace Page 3/41-22 Replace Page 3/4 2-1 Replace Page 3/4 2-2 Replace Page 3/4 2-3 -

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v IHQQ DEFINITIONS E P.AE 1.0 DEFINITIONS 1.1 ACTION ........................... 1.2 AXIAL FLUX DIFFERENCE ............ ........................... , 11 1.3 CHANNEL CALIBRATION............ ............................. 11 1.4 CHANNEL CHECK ................. ................................ 11 1.5 CHANNEL FUNCTION TEST .............. ........................... 11 1.6 CONTAIM ENT INTEGRITY ............ ........................... 11 1.7 CONTROLLED LEAKAGE ............................................. 12 1-2

    - 1. 8    CORE ALTERATION ................................................
    'I . 9                                                   ..

00SE EQUIVALENT I-131 .......................................... 12 1.10 E-AVERAGE DISINTEGRATION ENERGY ................................ 12 1.11 ....... 13 1.12 ENGINEERED SAFETY FEATURES RESPONS FREQUENCY NOTATION TIME ......................

                                    ...............E.......................                          .........          13 1.13                                                                                                             1-3 1.14   GASE0WS      RA9   WASTE       TREATMENT-SVSTEM                             (Deleted)          .............
                                                                                                         ......         1-3 ......

IDENTIFIED LEAKAGE ......s....... ...................... l 1.15 1-3 1.16 kIQWID-RASWASTE-TREATMENT-SYSTEM (041sted) . ........ . . . . . . .14 ........ 1.17 MAJOR 0FFSITE CHANGES-TG-RASIGAGTIVE-WASTE-TREATMENT-SY DOSE CALCULATION MANUAL 00CM . 1-4 1.18 OPERA 8tE - OPERA 8ILITY .........(.....)........................... 1-4 1.19 OPERATIONAL MODE - MODE ........................................ 1-4 1-5 1.20 PHY S I C S T E ST S . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . .1-5 1.21 PRESSURE B0UNDARY LEAKAGE ...................................... 1-5 1.22 PROCESS CONTROL PROGRAM 1.23 PURGE - PURGING ........(PCP)

                                                                                       ................                1-5
                                          .....................                                                        1-5 1.24 QUADRANT POWER TILT RATIO ...................................... 15 1.25 1.26 RATED THERMAL POWER .........................                         ...... ..................           16
                                 ...............E ..............................

REACTOR TRIP SYSTEM RESPONSE TIM ............ 1-6 1.27 REPORTA8LE EVENT .... 1-6 1.28 SHUTD0tel MARGIN .................... 1.29

                                                                  .........  ...........................               1-6 SekiBIFIGATiell(Deleted).......................................                                           1-6 1.30 SOURCE CHECK ...................................................        ..................                  1-6 1.31 STAGGERED TEST SASIS ............                                                                           1-6 1.32 THERMAL POWER ...................                .......................................                    17 1.33 15110ENTIFIED LEAKAGE ...........................................  .....................                    1-7 1.34 VOITILATION EXMAU$T TREATMENT                                                                               1-7

1. 35 VOIT INS . . . . . . . . . . . ..........................

                                                       . . . . . . . . . . SYSTEM . . . .              . .        -
                                                                                                                     . 1-7 TABLE 1.1 OPERATIONAL MODES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-8              ......-

TA8LE 1.2 FREQUENCY NOTATION ........................................ 1-9 -/. &t ConOpek UmhS T od'''''* FARLEY-UNIT 2 I AME M ENT NO. H. H

! 1 i mn ADMINISTRATIVE OONTROLS SECTION E&GX i j Review................................................... 6-10 l Audits................................................... 6-11 - l Authority................................................ 6-12 Records.................................................. 6-12 6.5.3 TECHNICAL REVIEW AND CONTROL l i Activities............................................... 6-12 ! Records.................................................. 6-13 j 6.6 REPORTABLE EVENT ACTION..................................... 6-14 l i 6.7 SAFETY LIMIT VIOLATION ..................................... 6-14 1 6.8 PROCEDURES hun PROGRAMS..................................... 6-14 : I 6.9 REPORTING REQUIREMENTS l

6. 9 .1. ROUTINE REPORTS I

Startup Report .......................................... 6-15a l Annual. Report............................................ 6-16 Annual Radiological Environmental Operating Report....... 6-17 l l Annual Radioactive Effluent Release Report............... 6-17 Monthly. Operating Report...........'...................... 6-19 ( c re 0,p.ere .w..n\._t l_ m_ih. b, _ ,ro c.h.

                                      , _ _ _ _ _ __.. _ m ............................... 6_1 Annual Diesel Generator Reliability Data Report..........                                                        6-19se Annual Reactor Coolant System Specific Activity Report... 6-20 Annual Sealed Source Leakage Report......................                                                        6-20 6.9.2 SPECIAL REPORTS...........................................                                                           6-20                   !

6.10 RECORD RETENTION............................................ 6-20 6.11 RADIATION PROTECTION PROGRAM................................ 6-21a 6.12 HIGH RADIATION AREA......................................... 6-22 FARLEY-UNIT 2 XIX AMENDMEhT NO.

f OEFINITIOMS

CONTA!!pIENT !MTEGR!TY 1.6 CONTAINMENT INTEGRITY shall exist when; a. All p'enetrations required to be closed during accident conditions are eitner:

1) Capable of being closed by an OPERA 8LE containment automatic isolation valve system, or

2) Closed by manual valves, 611nd flanges or deactivated automatic valves secured in their closed positions, except as previose in

                                          ,      Table 3.61 of Specification 3.6.3,

b. All equipment hatches are closed and se414d, c.

Eacn air lock is OPERABLE pursuant to Specification 3.6.1.3, d. The containment 3.6.1.2, and leakage rates are within the limits of Specification s -

e. The sealing mechanism associated with each penetration (e.g., welds, bellows or 0-rings) is OPERA 4LE.

CONTROLLED LEAKAGE

1.7 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

l CORE ALTERAT!0N

1.8 CORE ALTERATION small be tne movement or manipulation of any component witnin the reactor pressure vessel with the vessel head removed and fuel in the ' vessel. Suspension of CORE ALTERATION shall. not preclude completion of movement of a component to a safe conservative position.

 /NJdA?Y M 00$E EQU! VALENT I-131 1.9 00$E EQUIVALENT 1 131 shall be that concentration of I-131 (microcurie / gram) dich alone r uld produce the same thyroid dose as the quantity and isotopic mixture of I-131,1-132, !-133,1 134 and 1-135 retually present.            1he thyroid dose conversion factors used for this calculation shall be those listed in Table E.7 of Regulatory Guide 1.109 Revision 1,1977 l

FARLEY-UNIT 2 12 AMEN 0HENT NO. H i

INSERT A P:ge 1-2 CORE OPERATING LIMITS REPORT 1.8a The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with specification 6.9.1.11. Unit operation within these operating limits is addressed in individual specifications. l l l l [ : l l i 1 i l 4

   = __

1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTOOWN MARGIN - T >200*F LIMITING CONDITION FOR OPERAON l ! 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to ' " e't! t'* for 3 loop operation. # \ r l APPLICABILITY: MODES 1, 2*, 3, and 4. ?rW( 4:.fnefy .s N cf e ! ACTION: L'" w e.o&M ~" With the SHUTOOWN MARGIN less thany "", Alt; %/L, Y immediately initiate and continue boration at greater than or equal to 30 gpa of a solution containing ! greater than or equal to 7000 ppe baron or equivalent until the required l SHUTDOWN MARGIN is restored. - SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal l to[T Lit; k' :"-3 yg g,m,y ggefppy ,y .rYM c ow:

a. Within one hour after detection of an inoperable control rod (s) and at least once per 12 hours thereafter while the rod (s) is inoperable.

If the inoperable control rod is innovable or untrippable, the above l required SHUTDOWN MARGIN shall be verified acceptable with aa 1 increased allowance for the withdrawn worth of the immovable or untrigable control rod (s).

b. When in !400E 1 or MODE 2 with K greater than or equal to 1.0, at least once per 12 hours by vlNfying that control bank position is within the insertion limits of Specification 3.1.3.6.

c. When in 2 0E 2 with K 1ess than 1.0, within 4 hours prior to achieving reactor crifNality 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 55 RATED THERMAL POWER after each fuel loading, by consideration of the factors of a below, with the control banks at the maximum insertion limit of Specification 3.1.3.6.

         "See Spi W iest Exception 3.10.1 i

FARLEY-UNIT 2 3/4 1-1

11 REACTIVITY CONTROL SYSTEMS SHUTOOWN MARGIN - T,yg i 200*F LIMITING CON 0! TION FOR OPERATION 3.1.1.2 The SHUTOOWN MARGIN shall be greater than or equal t3 ' . ^* c ! - ' <' =_ - APPLICA81LITY: MODE 5. (fg gfy,7 gefgg ACTION: L. /H YHel e o t AF With the SHUTDOWN MARGIN less than 2.^" d:5 '/ . isunediately initiate and continue boration at greater than or equal to 30 gpe of a solution containing greater than or equal to 7000 ppe boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.2 The SHUTOOWN MARGIN shall be determined to be greater than or equal tok. LM: L; "-3 yM t-treff f 3Pfey& tD fM 4 =

a. Within one hour after detection of an inoperable control rod (s) and at least once per 12 hours thereafter while the rod (s) is inoperable.

If the inoperable control rod is immovable or untrippable, the SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s).

b. At least once per 24 hours by consideration of the following factors:

1. Reactor coolant system bcron concentration,

2. Control rod position,

3. Reacter coolant system average temperature,

4. Fuel burnup based on gross thermal energy generation, S. Xenon concentration, and

6. Samarium concentration.

4 3/4 1-3 FARLEY-UNIT 2 ,

w itm a of c a n d A v. E nd of Cy d t.the. f$cy n a .n lldif spec //le.d la + 4, ; REACTIVITY COWROL SYSTEMS c et. A. . T h a m asimm uppe r /.% M c4stl ba MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION ! l l 3.1.1.3 Themoderatortemperaturecoefficient(MTC)shallbef l [ui M ---Less than or equal to 0.7 x 10"' delta k/k/'F ft: Se -11 : d- - k :4a-ia: a' cycle life P L) e--fitier for power levels up to 70% TRERMAL F0VER with a linear ramp to O delta k/k/*F at 100% ! THERMAL POWER.

5. E- : :;;;ir: 9--

2.3 : 10"* delt: U U'? !:: $: 211 ::d: I vi ed tr. : d Of :y:1: lif: (!^'-) , M- : . t.;. 77/.%

rfitier.

Bot t%5 APPLICABILITY: 5;;;ift:::10: 3.1.1.*M - MODES 1 and 2* onlyt

                                                                                                !; nifi:::10: 3.1.1,4,4 - MODES 1, 2 and 3 onlyi C $t. tus F 0'

80L 3peeffej in de W

a. With the MTC more positive than the#11ait ^f 3 1 1 3 'dere, operation 4

                                                                                      'in MODES 1 and 2 may proce,ed provided:                                                            )

1. Control rod withdrawal limits are established and maintained sufficient to restore the MTC to within its limit within 24 hours or be in 50T STANDBY vithin the no::t 6 hours. These .

withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6.

2. The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that !

the MTC has been restored to within its limit for the all rods I withdrawn condition.

3. A Special Report is prepared and submitted to the Consission i pursuant to Specification 6.9.2 within 10 days, describing the value of the measured Nrc, the interia control rod withdrawal limits and the predicted average core burnup necessary for restoring the positive MTC to,withis its limit for the all rods withdrava condition. . .g g 4 g g,

b. With the NTC more negative than the limit of 3.1.1.3.5'dr:, be in 50T SEUTDOUN within 12 hours.

Vith K ,, greater than or equal to .0 i See Special Test Exception 3.10.3 FARLEY - UNIT 2 3/4 1-4 AMENDMENT NO. 4, i 85

i RF.ACMTT C0errROL SYSTEMS SURVIILIANCE REQUIREMENTS 6.1.1 3 fuel cycle as follows:The NTC shall be determined to be within its limits during each C.swer/%sie ,4) ynde coe.p

                                                                                                  .sms--

a. -The f* NTC

                   *;::i     shall fi;;;'
                              . :: be sessured 1.1.1 '        and
                                              . :t:1;.

compared to the 80L linked-4 , 5% of RATED TIERMAL POWER. aftar.each fuel loading. prior to in _5L b.TheNTCsha}lbemeasuredatanyTHERMALPOWERandcomparedto

.i," . 10' J.L . E'E'** '-11 -- ;

a "a"""

                         ;nfi:i :                           ..C.d.__... a.i , . - __ti,        i boron concentration of 300 ppe.within 7 EFPD after reaching an equilibriu indicates the IffC is more negativeInth*as J t'i'**,

the event 2.;,", this 10~poeparison

4. .' : - " "

liein. 'the NTC ::irt r;--i" shall be

                                            ? remeasgu 1 1 1--,   ed and compared to the EOL NTC)l at least once per 14 EFFD              }

curin's the remainder of the fuel cycle. (1) Yk//fdp ta)

         " TMd cy      "

( (W 3* Y

                                                       ~
                                                                $UMVdiL L44)ed'

__ Y L i m t f J ft!'C // 4' O O ) ones tae equilibrium beron concentration (all rods withdrawn, RATED TERNAL POUR condition) is 100 ppe or less, further seasurement of the Mtc in asserdance with 4.1.1.3.b esy be suspended, providing that the sensured NTC at an equilibrius baron concentrgtion less than or equal to 100 ppe is lads negative than t0: if f $ 2 U i/"7. 6 (ppttoo)*)*rst SUA!'VdfL&Annled

                                                  't!**/f   5f*d etfsep /A)
                                                   -fAH6 c.o LAl'.

iEEE: FARLIT-tNIT 2 3/4 1-5 . Amendesst No. 80 b

3' I REAwavan COIrrROL SYSTEMS 1 rossesanse ire! PHY.ssc/*L tA!Jefftle s)

                                                                                                                                         ### 8/iL'C/#/88'/d f SEUTD0gp R00 INSERTION LIMIT                                                                                        de C_ 6_ -4#.
                                                                                                                                                                                              ~

LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown rods shall be f '- -d u--- easa _....--n- . . . 4 APPLICABILITY: MODES 1* and 2*t.

                                                                                                                                                      ,,              ,               ,g g g u reer Y S P ft//i' dse os1YMdce4E Vith a anzimum of one shutdown r                         n: 0.!!y .u; ;....Y                 . except for surveillance testing pursuant to Specification 4.1.3.1.2, within one hour eithers      gewepc rnwr gw A 4 trM/ Al YMd /Alsegre D stoor/y fdt//*/dCF fn? YH L4i*Q OR

a. 4 P_ '_ ' ; -i * " --- ' " -" -- C

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

                                                                                                                                                                 &fW/s)           FWRLc /D$df Yted L/ror/ y . SPEC /M"/dx0
                                                                                                                                                         /

SURVEILLANCE R30tfIREMENTS / /a1 M eecS / f^

                                                            ,         4.1.3.5 Each shutdova rod shall be determined to                                           8 "; ri-"xx (n3 %
                                                  -                   w 231 assys, a d ini                                                                    ^

a. Within 15 minutes prior to withdrawal of any rods in control banks A, 3, C or D during an approach to reactor criticality, and

b. At least oeca per 12 hours thereafter.

                                                                      *5ee special Test Exceptions 3.10.2 and 3 10.3.
                                                                      #Uith K,,, greateu than or equal to 1.0; FARLEY-tstIT 2                                       3/4 1-20                                             AMENDMEKr N0. 76

__ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ ._m

1l

i REACT!v!TY CONTROL SYSTEMS CONTROL R00 INSERT!0N LIMITS LIMITING CONDITION FOR CPERATION 3.1. 3. 6 Th o es..... ,,ircontrolbanksshallbelimitedinphysicalinsertion)a

                                    ,        , . -                                           e r. ' %

APPLICA8!LITY: MODES la and 2"#. " M N I N 7784F c e 4 F. ACTION: With the control banks inserted beyond the above insertion limits, except for surveillance testing pursuant to Specification 4.1.3.1.2, either:

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

b. Reduce THERMAL POWER within two hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the group position using th Q n: '4.e.U r,, q g g g c.

8e in at least NOT STAN00'Y within F hours. '# N W SURVE!LLANCE REQUIROENTS l 4.1.3.6 The position of each coatrol bank shall be determined to be within the insertion limits at least once per 12 hours except during time intervals ! l when the Rod Ir:sertion Limit Monitor is inoperable, then verify the individual ) rod positions at least once per 4 hours.

          "See 5pecial Test Exceptions 3.10.2 and 3.10.3.

fWith K,ff greater than or equal to 1.0. FARLEY-UNIT 2 3/4 1-21

t. frsuns 3I-I <4 CTh is -15g a,e in+an homHg fe-Y+ h/en k)

       .                                                   .Ptedytf fr4oW

{ FULLY Wm4 DRAWN 225 TO 231 STEPS, INCLUSIVS Q j . m_ _. _._ _. 4 ..... .: ..... ...- -- . -

                                                                                                                                           .}.__
                           )                                       p                  .           :             .

3. - y 3 / %*EANKC ] )

                                           <       /                         --

C ./ g C n ) - A l 150 / ^ '

g / ( / .:: / . . . :. _ . g j y- g. :....=.. g /

                                                                           ,(                 / ' ' EANK D gg g
                                                           /                         ). c--. . .. .- - c.: . : . :

t [ -

                                                                            . f (:                     !. :.k :p.                    -l - =-
                                                    )                      f                     %:                                   .

f f- . q g ;_- .- ;

                                                          /                                                  %

~

                     -~r ,                         /

j . N J Z \ f f . : h

                              /_ . . .. . - . .. .. .
                                                                                               .=--. . ;- - - -       .
                                                                                                                         -   -- . .     .          \

f. g . . . . -. ( .:

j ::. --( -- 0 .2 .4 .s .s 1.0 l (PULLY INSERTED) . I FRACTION OF RATED THERMAL POWER I _ - u :.: _ : -e u 2.. u ..; .. . _ _ _ n _ _---- _ S. FARLEY-UNIT 2 3/4 1-22 AMEN 0 MENT NO. 76 l

S! 4: l 1 l i 3/4.2 POWER DISTRIBUTION LIMITS i 3/4.2.1 AY f af. FLUX DIFFERENCE f)FDn LIMITING CONDITION FOR OPERATION 3.2.1 the limits The indicated specified AXIAL FLUX DIFFERENCE (AFD) shall be maintained within C ,_,... ;.; 1.

                                                                         -lN inE i                APPLICABILITY:                   MODE 1 above 50% of RAT D THERMAL POWER **,                                         l ACTION:
                                                                                                                                              )

a. With the indicated AXIAL FLUX DIFFERENCE outside of the limits specified '. n ,___ ' " " -

1. NTNE MS l Iither restore the indicated AFD to within the limits within 15 minutes, or h

2. Reduce THERMAL POWER,to less than 50% of RATED THERMAL POWER

                                                                                                ~                                             )

within 30 minutes. SURVEILLANCE REQUIREMENTS I i 4.2.1 The indicated AXIAL FLUE DIFFERENCE shall be determined to be within its limits by: i l

a. Monitoring the indicated AFD for each OPERABLE encore channel

1. At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and

2. , At least once per hour with the AFD Monitor Alarm inoperable. l 1

The indicated AFD shall be considered outside of its limits when at least 2 OPERABLE encore channels are indicating the AFD to be outside its limits.

I

                **see special Test Exception 3.10.2.

FARLEY-UNIT 2 3/4 2-1 AMENDMENT NO. l

9 l DELETE ( Ficuarf

f 4 ') e idenMo%ll) (M b[qq l 12 1

_./ l 1 100

                                      *12,100)                                   (+9,100)

) Unacceptable cceptable Operation Operation 80 -

                                          / t                                     V                   .             :
      !ee                            /.                         4:= / \                       !
      !                           /

( 30, 50) i y "

(+24,

0 u

      $                  i l                           l                                       I g                                                     2                 l 20                                                                               l I

l /

                                                    )                                                           l O

50 -40 -

                                          -20             -10        0       10      20      30      40 Axial Flitx Difference (Delta 1)%                      .omu
                    "'T.Ta 1.2 . ., ;,                uw. I, : . . _.- .:-   .,L. L M .' "'= ? 'C::      #
                                      ^-
                                     .L.-.Z...-"-^-""^"**"                                ^

v FARLEY-UNIT 2 3/4 2-2 AMENDMENT NO.

o POWER DISTefBUTION LIMITS 3/4,2.2 m*T FLUX HOT CHANNEL FACTOR - FgiEl LIMITING CONDITION FOR OPERATION w :4 W a t-h e. b i b s 3.2.2 Fg(I) shall be l' 'tre by "r "ripardie.d in kht CDi-R . r ' ; relit'?-91;re J' (E) 5 (2,,,,11] (K(3)] for P > 0.5 for VANTAGE 5 fuel P Fg(E) 5 (4. K(E)) for P S 0.5 for VANTAGE 5 f and Fg(E) C [2,22) (K(3) or P > 0.5 for fuel i g Fg(E) 5 (4.64) (K(I)) for P O.5 LOPAR fuel where P = Tires pourn _. RATE 6 THERMAL POWER and) tift) is the function obtained from Figure 3.2-2 for iven p eore height location. APPLICABILITY: MODE 1. ACTION: With Fg(E) exceeding its limits i

a. Reduce TERAMAL POWER at least 14 for each 14 Fg(I) exceeds the limit within 15 minutes and similarly reduce clie Power Range Neutron Flus-Eigh Trip Setpoints within the neat 4 hours; POWER OPERATION may proceed for up to a total of 72 hours; subsequent PC4fER OPERATION may proceed provided the Overpower delta T Trip setpoints have been reduced at least 14 for each 1% Fg(E) exceeds the limit.

l b. TEIRMAL PONER may be increased provided Fg(E) is demonstrated through incore mapping to be within its IIait. SURVEILLANCE IREMENTS 4.2.2.1 The provisions of specification 4.0.4 are not applicable. 4.2.2.2 Fg(E) shall be evaluated to determine if it is within its limit by: l

a. Using the novable incore detectors to obtain a power distribution map at any TERANAL POWER greater than 5% of RATED TEIRMAL POWER.

FARLEY-UNIT 2 3/4 2-3 AMENDNENT NO.

l POWER D25TR2BUT20N L2MITS 4 SURVEILLANCE REQUIREMENTS (Continued)

b. C Determining follows:

the computed heat flux hot channel factor Fg (gy, ,, Increase the measured Fg(Z) obtained from the power distribution map by 3% to account for manufacturing tolerances and further increase the value by 5% to account for measurement uncertainties, c. Verifying that Fg (Z), obtained in Specification 4.2.2.2b above, satisfies the relationship in Specification 3.2.2.

d. Satisfying the fe,11owing relationships i

Fj (Z ) s F "" x K (Z ) 4 fo r P > 0.5 P x W (Z ) Fj (Z ) s F " " x K .( Z ) fo r -P s 0.5 1 0.5 x W (Z) Where Fg (2) is obtained in Specification 4.2.2.2b above, Fg is the Fg limit, K(E) is the normalised Fg(E) as a function of core 4 height, P 4s the fraction of RATED THERMAL POWER, and W(E) is the cycle dependent function that accounts for power distribution transiente encountered dur jin normal operation.

                                    %(,k kHD WL*J Ntt 59tt.tFtED IN W $h O i

Fg, )A r 4 . v0 J L ._ - . i m ' ',

                                                                                      % ${EC.tf W.MT10N y 2.;; ;i??? i:M O                     G,0(, { , g,                j IK                      # provi           in Fi      e 3.2-2 3

p vided n the Radi Peakin actor t tW(I _

e. Measuring Fg(E) according to the following schedule j 1. Upon achieving equilibrium conditions after exceeding by 20%

or more of RATED THERMAL POWER, the THERMAL POWER at which l Fg(I) was last deterisined*, or

2. At least once per 31 Effective Full Power Days, whichever occurs first.

4 4

        *During power escalation after each fuel loading, power level may be increased until equilibrium conditions at any power level greater than or
 ;      equal to 50% of RATED THERMAL POWER have been achieved and a power distribution map obtained.

FARLEY-UNIT 2 3/4 2-4 AMENDMENT No.

POWER DISTRIBUTION LIMITS i SURVE!LI.ANCE REQUIRxMENTS (Centinued)

f. With measurements indicating maximum

                                          'F oC(Z)'

over(Z) s K(Z)j C has increased following since actions shall thebeprevious taken: determination of Fg (E) either of the

1) Increase Fg (1) by the Fg (E) penalty factor specified in the COLR r-~- . 01 :- i' '- -

4 ^. and verify that this value satisfies A the relationship in specification 4.2.2.2d, or C

2) Fg (E) shall be measured at least once per 7 Effective Full Power Days until two successive maps indicate that maximum Fo C(Z)'

is not increasing. over(Z) ( K(Z)s g. With the relationships specified in specification 4.2.2.2d above not being satisfied

1) Calculate the percent Fg(E) exceeds its limits by the following expressions Fj (Z) x W (Z) m axim um -

1 x 100 for P > 0.5 over Z Fg, ,,

                          ,r                      .      P                       .)        ,

f 3 Fj (Z) x W (Z) m axim u m - 1 x 100 for P s 0.5, and over Z Fn,1,

                         ,s                       . 0.5                      .)       .

2) The following action shall be taken: I l

Within 15 minutes, control the AFD to within new AFD limits p which are determined by reducing the AFD limits specified in 988 {h COLR ? 2 1, 1-'r' r*-_r "'_"'

                                                                       - ::. by 1% AFD for each percent Fg(Z)
               >      exceeds its limits as determined in specification 4.2.2.2g.l.

Within 8 hours, reset the AFD alarm setpoints to these modified limits. I FARI.EY-UNIT 2 3/4 2-5 AMENDMENT NO.

t6 i f ct 3 e nko vib o Oc4 k D I 4 b i4 VI C pdedf& RGJW 02

                                                  =%               s 0.8              '
                                          >       y I0.8        .

l 0.4 / 0.2 -

                                 /                       N x

i

                            '                                                        1 0                 '         '      "

2 4 6 8 10 i 12 l CORE HEIGHT (FEE 1) I

           -ri,-. 3.4 4 m u noru m             is ;; = : - " -- -f c =;       m  n um FARLEY - UNIT 2                       3/4 2 7                  A MID e rt N0. 85

POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEARENTHALPYHOTCHANNELFACTOR-Fh LIMITING CONDITION FOR OPERATION win.'n he. lNds f pas.;4te J 4 the, c.o A . 3.2.3 F$H shall beg l' 'ted 5; the fell-8 ; elet!---ki--

                  ; a" i l a '  11         0 1 II     III I;;    "?~'C= 5 f2;l =2f 7$n i 1.00 il              0.2 (1     7); f;; _' 0??.- f;;i l
                     === ? -
                                    . J. . _ _ _ _ _1' '_ M?

APPLICABILITY: MODE 1. A ACTION WithF$Hexceedingitslimit , a. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours and reduce the Power Range Neutron Flux - High Trip setpoints to 5 55% of RATED THERMAL POWER within the next 4 hours, b. Demonstrat,ethroughin-coremappingthatF{giswithinitslimit ! within 24" hours after exceeding the limit or reduce TERRMAL POWER to less than 5% of RATED TERAMAL POWER within the next 2 hours, 1 and c. Identify and correct the cause of the out of limit condition prior to increasing TERRMAL POWER above the reduced limit required by a or b, above; subsequent POWER OPERATION may proceed provided that F{gisdemonstratedthroughin-coremappingtobewithinits limit at a nominal 50% of RATED TERRMAL POWER prior to exceeding this TERRMAL POWER, at a nominai 75% of RATED THERMAL POWER prior to exceeding this TERRMAL POWER and within 24 hours after attaining 95% or greater RATED TEIRMAL POWER. i FARLEY-UNIT 2 3/4 2-8 ' AMENDMINT NO. i i l

?!

3/4.1 REh.u n s t CCffrROL SYSTDS BASES 4 ..................... - - 2 _ 3/4.1.1 80 RATION CONTROL

3/4 1.1.1 AND 3/4.1.1.2 55UTD0VN MARGIN A sufficient 58VfDOWN MARGIN ensures that 1) the reactor can b suberttical from all operating conditions, 2) the reactivity traastents ,

i associated with postulated accident conditions are controllable within i suberitical to preclude inadvertent criticality incondition. the shutdo 55UTD0VN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS i i restrictive condittoa occurs at ROL, with T T,,,,. The most g gjgff, i at no load operating i

      # N 0g           l                   and resultina uncontrolled RCS cooldova.-temperature,            In t and is associa analysis of this accident.

4 a sansmus .asuwven reactivity transient. MARGINicf L"* d:l= E' . s required to control the Accordingly, the SEUTDOUN MARGIN requirement is ' based upea analysis this limiting assumptions. Withcondition T and is consistent with PSAR safety 1 8 resulting from a postulated s, tela line break coeldews are minim ti; L L SEUfDOUN MARGINJ rovides adequate protecties. i 3/4.1.1.3 MODERATOR TUFERATt/RS COSFFICIENT% (.-wt/f/dDIN M VAT The ensure limitations that on moderator temperature conifficient (NfC) are provided to j condittom assumed la the FSAR accident and transiest analy The plaatRfC values of this specification are applicable to a specific set of conditions: accordingly, verificaties of NfC values at conditions j other than these explicitly stated will require estrapolation to these conditions in order to permit an accurate temparisen. The most negative NfC value equivalent to the most positive moderator deastty coefficient (ISC) was obtained by incrementally correcting the MDC

used in the FSAR analyses to nestaal operating conditions. These corrections intelveda analyses to its equivalent NTC, based on the rate of chased of I j density with temperature at RATED TERENAL POWER conditions, and (2) l
subtracting free this value the largest differences la NfC observed 4

between E0L, all reds withdrawn, RATED TERNAL POWER conditions and those eost adverse conditions of moderator temperature and pressure,, rod insertion, axial poser skeving, aas nemen concentraties that can occur in neraal operaties and lead to a significantly more negative 50L NfC at RATED TERENAL POWER. ) the FS&R safety analyses into the limiting NFC vali rfThese corrections J

                   ,a E'i'"*.                         The sarve111amee requirement NTC value -f N^

1 17 f:l= O 4

                    #  a t'L'T represents a conservative NTC value at                                                    ore loa?       dition' "',    'ri nppa of 300    "s equilibrium beres concentration, and is obtai                                           by as king burasp and soluble beros to the limiting NfC v                                                                correptions for f              b7.                                                                                      d         O : if f rl" 2 '"._     ,

i ! C 4 sflterftd D ts) 1AT Q. l FARM T-talIT 2 3 3/A 1 1 Amendment No. 80 i j .*

1 RAAcuvuT C00frROL 3757503 (oarinsem y I/N8# s.t me-p' g % /Mdp/g/yg , N00ERATOR TENFERATURE COEFFICIElff (Continued) sessurements may be suspendedvalue

        .,,jL equilibriug n-y provided ppe,  NTC at the m 1

boron n, y;; concentration

                                        ;.            < 100 ppa is less negati ve thaa,g j limiting 20L NTC value ""'The I ~ diffe{ance
                                                "'             between MLis                   value     and the
                                                     *" n i't' the easisus change in NTC between the 100 ppe equ? conservatively b ilibrium beron licensed end-of-cycle, including the effects                                      on) and the-of beton reduction, fuel depletion, and end-of-cycle coastdown          .

oncentration The surveillance requirements .for measurement of the NTC and near the end of the fuel cycle are adequate to ceafire that'tha b remains withia its limits since this coefficient changes principa at the MTC y due slowl k rnup. lly to the reduction in RCS boroa concentration ass 3/4.1.1.4 , MININUM TENF h TURE FOR CRITIcArv?T This specification ensures that the reactor will not be the Reactor Coolaat System average temperature les made criti d with within its analysed temperature ranlimitation s than 341'F. is This required to en within its normal'operattag range, ge, 2) the protective. instrumentation isre coefficie . a steam bubble, and 3) the reactor pressure RT,,, temperature. status with sinimus ve 3/A.1.2 BORATION SYSTINS available during each mode of facility yoperation.The control is bo to perfore this function teclude 1) borated water sourThe components required ces 2 charging emergency power supply free OPERA 8LE diesel , an gene Vith the RCS average temperature above 200'F, a sinimum of two boron the event as assumed failure readers aone of the f t boration capability of either flow path is sufficient to providas The 58ttfD0VN inope f/M/WGOl W FARLET- W 2 5 3/4 1-1

                                                                                         $==8==atNo.30

1l l

      #Ea; !V!** COU 20L SYSTEMS                                                                           l 9ASES                                                                                                '

I l 90RA*104 SYSTEMS (Continued) wAGGIN f rom espected operating conditions " ' "" "'a- af ter xenon decay and cooldown to 200*F. The maximum exoected boration capaDility requirement occurs at EOL fr'om full power equilierium menon conditions and requires 11.336 gallons of 7000 som borated water from tne boric acid storage tanks or 44,826 i gallons of 2300 som borated water from the refueling water storage tant. ! Witn the RCS temperature below 200'F. one injection system is acceptacle witnout single fatlure consideration on tne basis of tne stable reactivity l condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity cnanges in the event the single injection system becomes inoperable. The limitation for a maximum of one centrifugal charging pump to be SPERABLE and tne Surveillince Requirement to verify all charging pumps except tne reauired OPERABLE oump to be inoperable below 180*F provides assurance tnat a mass addition cressure transient can be relieved by the operation of a single RHR relief valve. g, yA CcQ The boron ability recuired bel'ow 200*F ts sufficient to provide a SerJ'MN MARGIN,;' '.; C'L -;4 after zenon decay and cooldown from 200'F to 140*F. This condition recuires either 2,000 gallons of 7000 ppm borated water f rom tne boric acid storage tanks or 7.750 gallons of 2300 ppm borated water i from tne refueling water storage tank. The contained w ter volume limits include allowance for water not l available because of discharge line location and other physical characteristics. The limits on contained water volume and boron concentration of the RWST l also ensure a oH value of between 8.5 and 11.0 for the solution recirculated I within containment after a LOCA. This pH band minimizes the evolution of l l iodine and minimizes the effect of chloride and caustic stress corrosion on I l mecnanical systems and components. - The OPERABILITY of one boron injection system during REFUELING ensures that this system is available for reactivity control while in MODE 6. l 3/4.1.3 M0VA6LE CONTROL ASSEM8 LIES The specifications of this section ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUT 00WN MARGIN is l maintained, and (3) limit the potential effects of rod misalignment on i associated accident analyses. OPERA 61LITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits. l FARLEY UNIT 2 8 3/4 1 3 AMENDMENT NO. 60

1 3/4.2 POWER "JISTP! BUTTON LIMfTS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate l Frequency) events by: (a) meeting the DNB design criterion during normal ! operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding mechanical properties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial I conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200'F is not exceeded. l The def a.nitions of certain hot channel and peaking f actors as used in I these specifications are as follows: l I l Fg(Z) Heat Flux Hot Channel Factor, is defined as the maximum local i l heat flux on the surface of a fuel rod at core elevation Z i l divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods and measurement uncertainty. F"g 3 Nuclear Enthalpy hise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power. v' "M F, umtT STEciFieb 3/4.2.1 AXfAL FLUX DIFFERENCE ( lH VE, $% The limits o IAL FLUX DIFFERENCE (AFD) assure that the FM E) uppep bound envelope ofg .,,, 2.. . 0. ;; 2.. ^ : _7 times _ ;.J....nli :d "

  - "?r rr :..,_;.1;::?in;"rt:risnotexceededduringeithernormalope(rationorin the event of xenon redistribution following power changes.

I Provisions for monitoring the AFD on an'automacic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer determines the one minute average of each of the OPERABLE excore decoctor l outputs and providee an alarm message immediately if the AFD for 2 or more ! OPERAELE encore channels is outside the allowed AI operating space for RAOC l operation specified in %,__ _ ' ? " and the THERMAL POWER is greater than 50% RATED THERMAL . ME CO 1 1 y s j FARLEY-UNIT 2 3 3/4 2-1 AMENDMENT NO. j i

                                                                                                                 $i PowIR DISTRIBUTION LIM 2TS                                                                             i BASES

__, .g 3/4.2.2 and 3/4.2.3 CHANNEL FACTOR HEAT FLUX HOT CHANNEL FACTOR. NUCLEAR ENTHALPY HOT The limits on heat flux hot channel factor, and nuclear enthalpy rise hot channel factor ensure that 1) the design limit on peak local power density is not exceeded, 2) the DN8 design criterion is met, and 3) in the event of a LOCA ene peak fuel clad temperature will not exceed the 2200*F ECCS acceptance criteria limit. Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is sufficient to insure that the limits are maintained provided: a. Control rods in a single group move together with no individual rod insertion differing by more than 2 12 steps, indicated, from I the group demand position. l l b. Control rod banks are sequenced with overlapping groups as described in Specification 3.1.3.6.

c. The control rod insertion limits of'5pecifications 3.1.3.5 and 3.1.3.6 are maintained.

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

Fh will be maintained within its limits provided conditions a. through

d. above are maintained. The relaxation of Fh as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits.

When an Fg measurement is taken, an allowance for both experimental error and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full core map taken with the incore detector flux mapping system and a 3% allowance is appropriate for manufacturing tolerance. The heat flux hot channel factor Fg(I) is measured periodically and increased by a cycle and height dependent power factor appropriate to RAOC operation, W(I), to provide assurance that the limit on the heat flux hot channel factor Fg(I) is set. W(I) accounts for the effects of normal operational transients within the AFD limits and was determined from expected power control maneuvers over the full range of burnup conditions in the core. WE w(4 FwCTson fck Moe.mm 9%%on Mb WE g Wmm5 hhE. ikomotb m 'TWE COLR P r * *:q E c ss o T e d 4.9.1. R . i : f FARLEY-UNIT 2 3 J/4 2-2 AMENDNENT NO. l

T, ADMINISTRATIVE CONTROLS 4 i MONTHLY OPERATING REPORT 6.9.1.10 ! Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the FORV's or safety valves, shall be submitted on a monthly basis to the Comunission, pursuant to 10 CFR 50.4, no later than the 15th of each month following the calendar month covered by the report. gg g PEAKING FAC M LIMIT REPORT 6.9.1.11 The cycle de dont functi , N(3), and burnup pendent Fg ) penalty actors, requir Flux for calcul ion of Fg (3 specified LCO 3.2.2, " Beat t Channel Fact - Fg(E)," s all be doc tpd in the eaking Fact Limit Re t in accordane with the me odology in -10216-F-A "Relaxatio of Co stant Axial of et control Surveillanc schnical 8 cification Rev. 1, bruary 1994 ( Froprietary) The Peaking actor Limit port shall provided to he Ceauniss n, pursuant t 10 CFR 50. , upon issu e prior to e reload cy e (prior t MODE 2). In a event t t the limit ld be submit ed at some o e tiase d ing core life it will submitted u n issuance, up ess otherwi exempted the Commise n.

ANNUAL DIESEL CEN2RATOR RMLYamTLTTY DATA meson? 6.9.1.12 The number of tests (valid or invalid) and the number of failures to start on demand for each diesel generator shall be submitted to the NRC annually. This report shall contain the information identified it. Aegulatory Position c.3.b of NRC Regulatory Guide 1.108, Revision 1, 1977. l I II { FARLEY-UNIT 2 6-19 AMENDMINT NO. i

i CORE OPERATING LIMITS REPORT 6.9.1.11 Core operating limits shall be established and documented in the CORE OPBRATING LIMITS REPORT before each reload cycle or any reaining part of a reload cycle for the following:

1. SHUTDOWN MARGIN limit for MODES 1, 2, 3, and 4 for Specification 3/4.1.1.1,

2. SHUTDOWN MARGIN limit for MODE 5 for Specification 3/4.1.1.h

3. Moderator Temperature Coefficient BOL and BOL limits and 300 pga and 100 ppm surveillance limits for specification 3/4.1.1.3,

4. Shutdown Bank Insertion Limit for specification 3/4.1.3.5, 1 S. Control Bank Insertion Limits for specification 3/4.1.3.6,

6. Axial Flux Difference limits for Specification 3/4.2.1,

7. Heat Flux Not Channel Factor Fg ' limits, K(E) figure, W(E) values, and Fg(E) Penalty Factors for Specification 3/4.2.2,

8. Nuclear Enthalpy Rise Hot Channel,7 actor limits, FAH , and Power Factor Multiplier, PFAH, for Specification 3/4.2.3.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC in

1. WCAP-9272-P-A, " estinghouse Reload Safety Evaluation Methodology," July 1985 (M Proprietary).

(Methodology for specifications 3.1.1.1 - Shutdown Margin - Tavg > 200*F, 3.1.1.2 - Shutdown Margin - Tavg 5 200'F, 3.1.1.3 - Moderator Temperature coefficient, 3.1.3.5 - Shutdown Bank Insertion Limit, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.)

2. WCAP-10216-P-A, Rev. lA, " Relaxation Of Constant Asial Offset Control /

Fg Surveillance Technical specification," February 1994 (M Proprietary). (Methodology for specifications 3.2.1 - Axial Flux Difference and 3.2.2 - Heat Flux Hot Channel Factor.)

3. WCAP-10266-P-A, Rev. 2, "The 1981 Version Of Westinghouse Evaluation Model Using BASH Code," March 1987 (H Proprietary).

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

1, 1 1 1 INDEX 1 DEFINITIONS SECTION Na1 1.0 DEFINITIONS i 1

,      1.1 1.2 ACTION........................................................

AXIAL FLUX DIFFERENCE......................................... 1-1 1-1 1.3 1.4 CHANNEL CHANNEL CALIBRATION........................................... 1-1 l 1.5 CHECK................................................. CHANNEL FUNCTION TE8T......................................... 1-1 ' 1-1 1.6 CONTAINMENT INTEGRITY......................................... 1-2 1.7 1.8 CONTROLLED LEAKAGE............................................ CORE ALTERATION............................................... 1-2 1-2 1 1.8a CORE OPERATING LIMITS REPORT.................................. 1-2 1.9 l DOSE EQUIVALENT I-131........................................ 1-2 1.10 T-AVERAGE DISINTEGRATION ENERGY............................... 1-3 1 1.11 ENGINEERED SAFETY FEATURES RESPONSE TIME...................... 1-3

,      1.12 FREQUENCY NOTATION............................................                                                                           1-3 1

1.13 (Deleted).................. 1-3 l 1.14 IDENTIFIED LEAKAGE................'............................ 1-3 1.15 'w' ^ w - " '**** "" (D.1.t.d).................... 1-4 , 1.16 . " _ _ _ . " , _ " * " ' _ " _ _ . . . " _ _ _ _ . . ~ . . ~_ " . - " . . ' _ . _ . . _ . ' . . . " . " . . _ - " _ _ - _ - " " (Deleted) 1-4 1.17 OFFSITE DOSE CALCULATION MANUAL (ODCM)........................ 1-4 1.18 OPERAELE - OPERARILITY........................................ 1-4 1.19 OPERATIONAL MODE - MODE....................................... 1-5 1.20 PHYSICS TESTS...f............................................. 1-5 1.21 PRESSURE BOUNDARY LEAKAGE..................................... 1-5 1.22 PROCESS CONTROL PROGRAM (PCP)................................. 1-5 1.23 PURGE - PURGING............................................... 1-5

;      1.24 QUADRANT POWER TILT RATIO.....................................                                                                           1-5

1.25 RATED THERMAL PONER.......................................... 1-6 l 1.26 REACTOR TRIP SYSTEM RESPONSE TIME............................. 1-6 1.27 REPORTABLE EVENT.............................................. 1-6 1.28 SHUTDONN MARGIN............................................... 1-6 1.29 seWG5MeAMeN (Deleted)...................................... 1-6 1.30 SOURCE CHECK.................................................. 1-6

- 1.31 STAGGERED TEST RASIS.......................................... 1-6 s 1.32 THERMAL POWER................................................ 1-7 ' 1.33 UNIDENTIFIED LEAEAGE.......................................... 1-7 1.34 VENTILATION E1HAUST TREATMENT SYSTEM.......................... 1-7 1.35 VENTING....................................................... 1-7 TABLE 1.1 OPERATIONAL MODES 1-8 ) l TABLE 1.2 FREQUENCY NOTATION 1-9 d 3 i i 1 i 1 FARLEY-UNIT 2 I AMENDMENT NO. e l' t 1

ADMINISTRATIVE CONTROLS SECTION EAfa1 Review................................................... 6-10 Audits................................................... 6-11 Authority................................................ 6-12 Records.................................................. 6-12 6.5.3 TECHNICAL REVIEW AND CONTROL Activities............................................... 6-12 Records.................................................. 6-13 6.6 RE POR'I ABLE EVENT ACT I ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-14 6.7 SAFETY LIMIT VIOLATION ..................................... 6-14 6.8 PROCEDURES AND PROGRAMS..................................... 6-14 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTSp Startup Report .......................................... 6-15a Annual Report............................................ 6-16 Annual Radiological Environmental Operating Report....... 6-17 Annual Radioactive Erfluent Release Report............... 6-17 Monthly Operating Report................................. 6-19 Core Operating Limits Report............................. 6-19 l Aru.dal Diesel Gene:rator Reliability Data Report. . . . . . . . . . 6-19a l Annual Reactor Co31 ant System Specific Activity Report... 6-20 Annual sealed source Leakage Report...................... 6-20 6.9.2 SPECIAL REPORTS........................................... 6-20 6.10 RECORD RETENTION............................................ 6-20 6.11 RADIATION PROTECTION PROGRAM................................ 6-21a 6.12 HIGH RADIATION AREA......................................... 6-22 FARLEY-UNIT 2 XIX AMENDdkMT NO.

_ . _ - _. . _- - - -. ~ . - _ . _- - -- DEFINITIONS CONTAINMENT INTEGRITY i i 1.6 OONTAINMENT INTEGRITY shall exist when: ;

a. All penetrations required to be closed during accident conditions I are eithers l

( 1) Capable of being closed by an OPERABLE containment automatic i isolation valve system, or 1 l l

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

b. All equipment hatches are closed and sealed, l c. Each air lock is OPERABLE pursuant to specification 3.6.1.3, I

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

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

welds, bellows or 0-rings) is OPERABLE. CONTROLLED LEAKAGE 1.7 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals. CORE ALTERATION l 1.8 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in l the vessel. Suspension of OORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position. CORE OPERATING LIMITS REPORT l 1.8a The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document l that provides core operating limits for the current reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with specification 6.9.1.11. Unit operation within these operating limits is addressed in individual specifications. DOSE EQUIVALENT I-131 1.9 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134 and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table E-7 of Regulatory Guide 1.109, Revision 1, 1977. I J t FARLEY-UNIT 2 1-2 AMENDMINT NO.

T 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - Tava >200'F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to the limit specified in the COLR for 3 loop operation. APPLICABILITY: MODES 1, 2*, 3, and 4. ACTION: I With the SHUTDOWN MARGIN less than the limit specified in the COLR, immediatelyl initiate and continue boration at greater than or equal to 30 gpm of a solution ! containing greater than or equal to 7000 ppe boron or equivalent until the ! required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS . 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or I equal to the limit specified in the COLR l

a. Within one hour after detection of an inoperable control rod (s) and at least once per 12 hours thereafter while the rod (s) is inoperable. If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified 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,gg greater than or equal to 1.0, at least once per 12 hours by verifying that control bank position it within the insertion limits of Specification 3.1.3.6.

c. When in MODE 2 with K,gg less than 1.0, within 4 hours 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 St RATED THERMAL POWER after each fuel loading, by consideration of the factors of a below, with the control banks at the maximum insertion limit of Specification 3.1.3.6.

See Special Test Exception 3.10.1.

l FARLEY-UNIT 2 3/4 1-1 AMENDMENT NO.

?'

REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - Tava 5 200*F LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to the limit specified in the COLR. APPLICABILITY: MODI 5. ACTION: With the SHUTDOWN MARGIN less than the limit specified in the COLR, immediatelyl initiate and continue boration at greater than or equal to 30 gym of a solution containing greater than or equal to 7000 ppe boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to the limit specified in the COLRs 1

a. Within one hour after detection of an inoperable control rod (s) and at least one,e per 42 hours thereafter while the rod (s) is inoperable. If the inoperable control rod is immovable or untrippable, the SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s).

b. At least once per 24 hours by consideration of the following factors:

1

1. Reactor coolant system boron concentration,

2. Control rod position,

3. Reactor coolant system average temperature,

4. Fuel burnup based on gross thermal energy generation,

5. Xenon concentration, and

6. Samarium concentration.

FARLEY-UNIT 2 3/4 1-3 AMENDMENT NO.

t REACTIVITY CONTROL SYSTEM 3 MODERATOR TEMPERATURE COEFFICIENT LIMITING OONDITION FOR OPERATION i 2 3.1.1.3 The moderator temperature coefficient (MTC) shall be within the I beginning of cycle life (BOL) limit and the end of cycle life (BOL) limit i specified in the COLR. The maximum upper limit shall be

              ~                                                                   less than or equal to 1  0.7 x 10           delta k/k/*F for power levels up to 70% THERMAL              POWER with a linear 4

ramp to O delta k/k/*F at 100% THERMAL POWER. 1 4 APPLICABILITY: BOL limit - MODIs 1 and 2* only#. EOL limit - MODES 1, 2 and 3 onlyd. ACTION:

a. With the MTC more positive than the BOL limit specified in the COLR, l operation in MODES 1 and 2 may proceed provided:

1. Control rod withdrawal limits are established and maintained !

sufficient to restore the MTC to within its limit within 24 l hours or be in HOT STAND 8Y within the next 6 hours. These withdrawal limits shall be in addition to the insertion limits j of sgecification 3.1.3.6.

2. The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifles that the MTC has been restored to within its limit for the all rods withdrawn condition.

3. A special Report is prepared and submitted to the Commission pursuant to specification 6.9.2 within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition,

b. With the MTC more negative than the BOL limit specified in the COLR, be in HOT SHUTDOWN within 12 hours.

With K,gg greater than or equal to 1.0.

  # See special Test Exception 3.10.3.

FARLEY-UNIT 2 3/4 1-4 dMENDMENT NO.

I l REACTIVITY CONTROL SYSTEMS ; I SURVEILLANCE REQUIREMENTS I 4.1.1.3 The MTC shall be determined to be within its limits during each fuel cycle as follows:

a. The MTC shall be measured and compared to the BOL limit specified in the COLR, prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading.

b. The MTC shall be measured at any THERMAL POWER and compared to the 300 ppe MTC surveillance limit specified in the COLR l within 7 EFPD after reaching an equilibrium boron concentration of l 300 ppe. In the event this comparison indicates the MTC is more { i negative than the 300 ppe MTC surveillance limit specified in the COLR, the MTC shall be remeasured, and compared to the BOL MTC limit specified in the COLR, at least once per 14 EFPD during the remainder of the fuel cycle. (1) 1

.=

l l I l l (1) Once the equilibrium boron concentration (all rods withdrawn, RATED THERMAL POWER condition) is 100 ppe or less, further measurement of the MTC in accordance with 4.1.1.3.b may be suspended, providing that the measured MT at an equilibrium boron concentration less than or equal to 100 ppm is less negative than the 100 ppe MTC surveillance limit specified in the COLR. l d FARLEY-UNIT 2 3/4 1-5 AMENDMENT NO.

i REACTIVITY CONTROL SYSTEMS SHUTDOWN ROD INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown rods shall be limited in physical insertion as specified in the COLR. APPLICABILITY: MODES 1* and 2*#. ACTION: With a maximum of one shutdown rod inserted beyond the insertion limit specified in the COLR, except for surveillance testing pursuant to Specification 4.1.3.1.2, within one hour either: a Restore the rod to within the insertion limit specified in the COLR, or

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

l SURVEILLANCE REQUIREMENTS 4.1.3.5 Each shutdown rod shall be determined to be within the insertion limit specified in the COLRs

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

b. At least once per 12 hours thereafter.

See Special Test Exceptions 3.10.2 and 3.10.3.

With Keft greater than or equal to 1.0.

FARLEY-UNIT 2 3/4 1-20 AMENDMENT NO. l

1 REACTIVITY CONTROL SYSTitMS CONTROL ROD INSERTION LIMITS LIMITING CONDITION POR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as specified in the COLR. l APPLICABILITY: MODES la and 2*#. ACTION: With the control banks inserted beyond the above insertion limits, except for surveillance testing pureusnt to specification 4.1.3.1.2, either a, Restore the control banks to within the limits within two hours, or

b. Reduce TMERNAL POWER within two hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the group position using the insertion limits epecified in the COLR, or l

c. Be in at least HOT STANDBY within 6 hours.

SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours except during time intervals within the Rod Insertion Limit Monitor is inoperable, then verify the individual rod positions at least once per 4 hours. l I l l l l 1 l l , i t

See Special Test Exceptions 3.10.2 and 3.10.3.

i

             # With K,gg greater than or equal to 1.0.                                      ,.

FARLEY-UNIT 2 3/4 1-21 AMENDMENT NO.

l' FIGURE 3.1-1 (This Figure intentionally left blank.) 6 4 FARLEY-UNIT 2 3/4 1-22 AMENDMENT NO.

3/4.2 POWER DISTRIBUTION LIMITS [ 3/4.2.1 ArrAL FLUX DIFFERENCE (AFD) LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXILL FLUX DIFFERENCE (AFD) shall be maintained within

the limits specified in the COLR*.

l APPLICABILITY: MODE 1 above 50% of RATED THERMAL Pol'ER**. I l l l ACTION: , { s. With the indicated AXIAL FLUX DIFFERENCE outside of the limits specified in the OOLR. l l

1. Either restore the indicated AFD to within the limits within 15 minutes, or l 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes.

l .- SURVEILLANCE REQUIREMENTS 4.2.1 The indicated AXIAL FLUX DIFFERENCE shall be determined to be within its limits by: /

a. Monitoring the indicated AFD for each OPERABLE excore channels l

1. At least once per 7 days when the AFD Monitor Alarm is l

OPERABLE, and

2. At least once per hour with the AFD Monitor Alarm inoperable.

l l i t

The indicated AFD shall be considered outside of its limits when at least 2

, OPERABLE excore channels are indicating the AFD to be outside its limits.

                **see special Test Exception 3.10.2.

i FARLEY-UNIT 2 3/4 2-1 ., AMENDMENT NO. (

1 4' i i I l (This page intentionally left blank.) 6 1 i l l l i l i k i j 4' FARLEY-UNIT 2 3/4 2-2 AMENDMENT NO.

- _ . -, . . _ . - . - - -- . . -. _ ~ - - _ -. - - . .= . . . . . T' POWER DISTRIBUTION LIMITp 3/4.2.2 sat 1T FLUX HOT CHANNEL FACTOR - Fnf Z1 ! LIMITING CONDITION FOR: OPERATION 3.2.2 Fg(E) shall be within the limits specified in the COLR. l l APPLICABILITY: MODE 1. ACTION: With Fg(E) exceeding its limit: a. Reduce THERMAL POWER at least it for each in gF (I) exceeds the limit within IE minutes and similarly reduce the Power Range Neutron Flux High Trip Setpoints within the next 4 hours; POWER OPERATION may proceed for up to a total of 72 hours; subsequent POWER OPERATION may proceed /provided_the Overpower delta T Trip Setpointe have been reduced at least 1% for each it F (E) g exceeds the limit. b. THERMAL POWER may be increased provided Fg(I) is demonstrated through inoore mapping to be within its limit. SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of specification 4.0.4 are not applicable. l l 4.2.2.2 Fg(Z) shall be evaluated to determine if it is within its limit bys

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

I FARLEY-UNIT 2 3/4 2-3 AMENDMENT NO.

P POWER DISTRIBUTION LIMITS SURVEILLANCE REgUIREMENTS (Continued) b. Determining the computed heat flux hot channel factor FQ (I), as follows: Increase the measured Fg(I) obtained frc4 the power distribution map by 34 to account for manufacturing tolerances and further .l increase the value by 5% to account for measurement uncertainties.

c. Verifying that Fg (E), obtained in Specification 4.2.2.2b above, satisfies the relationship in specification 3.2.2.

d. satisfying the following relationships Fj (Z ) s F""" x K (Z ) fo r P > 0.5 P x W (Z )

Fj (Z ) s F'"" x K (Z ) fo r P- s 0.5 0.5 x W (Z ) Where Fg (5) is obtained in Specification 4.2.2.2b above, Fg is the Fg limit, K(5) is the normalized Fg(E) as a function of core height, P 11s the fraction of RATED THERMAL POWER, and W(5) is the cycle dependent function that accounts for power distribution transients encountered during normal operation. Fg , K(E), and W(I) are specified in the COLR as per specification 6.9.1.11. ' I

e. Measuring Fg(E) according to the followir.g schedules

1. Upon achieving equilibrium conditions after exceeding by 20%

or more of RATED THERMAL POWER, the THERMAL POWER at which Fg(E) was last determined *, or

2. At least once per 31 Effective Full Power Days, whichever occurs first.

      *During power escalation after each fuel loading, power level may be increased until equilibrium conditions at any power level greater than or equal to 50% of RATED THERMAL POWER have been achieved and a power                   '

distribution map obtained. FARLEY-UNIT 2 3/4 2-4 AMENDMENT NO. 1

3 POWER DISTRIBUTION LIMITS i SURVEIIIANCE REQUIREMENTS (Continued) l f. With measurements indicating maximum Fn C(Z)' over(Z) ( K(Z)j ' has increased since the previous determination of Fg (Z) either of the following actions shall be taken:

1) Increase Fg (I) by the Fg (1) penalty factor specified in the COLR and verify that this value satisfies the relationship in l t

Specification 4.2.2.2d, or I

2) Fg (1) shall be measured at least once per 7 Effective Full Power Days until two successive maps indicate that t

maximum

                                                'F C n (Z)'     ,

is not increasing. over(Z) ( K(Z).j _

g. With the relationships specified in Specification 4.2.2.2d above l not being satisfied:

( l

1) Calcylate the percent Pg(Z) exceeds its limits by the following expression:

C m axim um Fn (Z) x W (Z) -1 x 100 for P > 0.5 l over Z F ,, , , l

                                ,s               .      P                             .)    ,

l l

                                'r               .                                    .,     <

\ C m axim um Fn (Z) x W (Z) I ' l

                                                                                         -1    x 100 for P s; 0.5, and over Z         Fn,7,
                                .\               . 0.5                                .)    ,

1

2) The following action shall be taken:

Within 15 minutes, control the AFD to within new AFD limits which are determined by reducing the AFD limits specified in l the COLR by 1% AFD for each percent Fg(Z) exceeds its limits as determined in Specification 4.2.2.2g.l. Within 8 hours, reset the AFD alarm setpoints to these modified limits. FARLEY-UNIT 2 3/4 2-5 AMENDMENT NO.

i l i I l (This page intentionally left blank) ' l l v e l 1 i s l FARLEY-UNIT 2 3/4 2-7 ' AMENDMENT NO.

! 1 l l I POWER DISTRIBUTION LIMITS l 3/4.2.3 NUCLEARENTHALPYHOTCHANNELFACTOR-Fh LIMITING CONDITION FOR OPERATION 3.2.3 Fh shall be within the limits specified in the COLR. l , i l l ( l APPLICABILITY: MODE 1. ACTION: l With Fh exceeding its limits a. Reduce THERMAL POWER to leser than 50%.of RATED THERMAL POWER within 2 hours and reduce the Power Range Neutron Flux - High Trip setpoints to 5 55% of RATED THERMAL POWER within the next 4 hours, b. Demonstrate through in-core mapping that Fh is within its limit within 24 )ours after exceeding the limit or reduce THERMAL POWER te less than 5% of RATED THERMAL POWER within the next 2 hours, and

c. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by a or b, above; subsequent POWER OPERATION may proceed provided that' Fh is demonstrated through in-core mapping to be within its limit at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERNAL POWER prior l to exceeding this THERMAL POWER and within 24 hours after attaining 95% or greater RATED THERMAL POWER.

l l i I I d l FARLEY-UNIT 2 3/4 2-8 ( AMENDMENT NO. l

3.4.1 REACTIVITY CONTROL SYSTEMS i BASES i 1 3/4.1.1 aORATION CONTROL 3/4.1.1.1 mun 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that *) the reactor can be made suberitical from all operating conditions, 2) the reactivity transients i associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently j 1 subcritical to preclude inadvertent criticality in the shutdown condition. i i SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel i depletion, RCS boron concentration, and RCS T avg. The most 4 restrictive condition occurs at EOL, with T avg at no load operating i temperature, and is associated with a postulated steam line break accident i and resulting uncontrolled RCS rw idown. In the analysis of this accident, ) a minimum SHUTDOWN MARGIN as specified in the COLR is required to control the l

reactivity transient. Accordingly, the SHUTDOWN MARGIN requirement is based
upon this limiting condition and is consistent with FSAR safety analysis j assumptions. With Tavg less than 200'F, the reactivity transients resulting j from a postulated steam line break cooldown are minimal and a lower SHUTDOWN f MARGIN (specified in the COLR) provides adequate protection.

l l 3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT { The limitations on modgrator temperature coefficient (MTC) are provided to

ensure that the value*of this coefficient remains within the limiting condition i assumed in the FSAR accident and transient analyses.

} The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at conditions , other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison. The most negative MTC value equivalent to the most positive moderator density coefficient (MDC) was obtained by incrementally correcting the MDC used in the FSAR analyses to nominal operating conditions. These corrections involved (1) a conversion of the MDC used in the FSAR safety analyses to its equivalent MTC, based on the rate of change of moderator density with temperature at RATED THERMAL POWER conditions, and (2) subtracting from this value the largest differences in MTC observed between EOL, all rods withdrawn, RATED THERMAL POWER conditions, and those most adverse conditions of moderator temperature and pressure, rod insertion, axial power skewing, and xenon concentration that can occur in normal operation and lead to a significantly more negative EOL MTC at RATED THERMAL POWER. These corrections transformed the MDC value used in the FSAR safety analyses into the limiting MTC value specified in the COLR. The surveillance requirement MTC value specified in the COLR represents a conservative MTC value at a core condition of 300 ppm equilibrium boron concentration, and is obtained by making corrections for burnup and soluble boron to the limiting MTC value specified in the COLR. l

                                                                                                                                       .c FARLEY-UNIT 2                                                        B 3/4 1-1                    AMENDMENT NO.

_n_.-_--- ._ . - - , - - . . . . _ - _ _ = - . 1l REACTIVITY CONTROL SYSTEMS BASIS MODERATOR T m ERATURE COEFFICIENT (Continued) ' Once the equilibrium boron concentration falls below 100 ppe, MTC ' measurements may be suspended provided the measured MTC value at an equilibrium boron concentration s 100 ppm is less negative than the 100 ppe MTC surveillance limit specified in the COLR. The difference between this value and the limiting EOL MTC value conservatively bounds the maximum change in MTC between the 100 ppe equilibrium boron concentration (all rode withdrawn, RATED THERMAL POWER condition) and the licensed end-of-cycle, including the effects of boron concentration reduction, fuel depletion, and end-of-cycle coastdown. The surveillance requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup. 3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 541*F. This limitation is required to ensure 1) the mMerator temperature coefficient is within its analyzed temperature range, 2) the protective instrumentation is within its normal operating range, 3) the P-12 interlock is above its setpoint, 4) the pressupiser is capable of being in an OPERABLE status with a steam bubble, and 5)*the reactor pressure vessel is above its minimum RTNDT temperature. 3/4.1.2 BORATION SYSTEMS l The boron injection system ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include 1) borated water sources, 2) charging pumps, 3) separate flow paths, 4) boric acid transfer pumps, and 5) an emergency power supply from OPERABLE diesel generators. With the RCS average temperature above 200*F, a minimum of two boron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable. The boration capability of either flow path is sufficient to provide the required SHUTDOWN MARGIN from expected operating conditions after xenon decay and cooldown to 200*F. The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires 11,336 gallons of 7000 ppe borated water from the boric acid storage tanks or 44,826 gallons of 2300 ppm borated water from the refueling water storage tank, i i ! I FARLEY-UNIT 2 8 3/4 1-2 AMENDMENT NO. L _ _ -

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

1 REACTIVITY CONTROL SYSTEMS BASES l , , , , , , BORATION SYSTEMS f CL cainued) With the RCS temperature below 200*F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single injection system becomes inoperable. The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumpe except the required OPERABLE pump to be inoperable below 180*F provides assurance that a mass addition pressure transient can be relieved by the operation of a single RHR relief valve. The boron capability required below 200*F is sufficient to provide a SHUTDOWN MARGIN as specified in the COLR after menon decay and cooldown from l 200*F to 140*F. This condition requires either 2,000 gallons of 7000 ppe borated water from the boric acid storage tanks or 7,750 gallons of 2300 ppe borated water from the refueling water etcrage taph. The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics. The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 8.5 and 11.0 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and comptnants. . The OPERABILITY of one boron injection system during REFUELING ensures that this system is available for reactivity control while in MODE 6. 3/4.1.3 MOVan" CONTROL ASSEMBLIES The specifications of this section ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) limit the potential offacts of rod misalignment on associated accident analyses. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits. l FARLEY-UNIT 2 8 3/4 1-3 AMENDMENT NO.

                                                                                                          'l 3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by:           (a) meeting the DN8 design criterion during normal operation and in short term transients, and (b)               limiting the fission gas release, fuel pellet temperature and cladding mechanical properties to within assumed design criteria.       In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 22OO*F is not exceeded.

The definitions of certain hot channel snd peaking factors as used in these specifications are as follows: Fg(E) Heat Flux Hot Channel Factor, is defined as the maximum local heat l flux on the surface of a fuel rod at core elevation I divided by l the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods and measurement uncertainty. ! F$ Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio l of the integral of linear power along.the rod with the highest integrated power to the average rod power. 3/4.2.1 AIIAL FLUX DIFFERENCE j The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the Fg(Z) upper bound snvelope of the Fg limit specified in the COLR times K(E) is not l exceeded during either normal operation or in the event of xenon l redistribution following power changes. , Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer determines the one minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for 2 or more OPERABLE excore channels is outside the allowed AI operating space for RAOC i operation specified in the COLR and the THERMAL POWER is greater than 504 l RATED THERMAL POWER. i ,. FARLEY-UNIT 2 B 3/4 2-1 AMENDMENT NO.

POWER DISTRIBUTION LIMITS BASES

                                                                                                                  .aunumass=usmusa 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR. NUM M8! ENTHALPY HOT           -

CHANNEL FACTOR The limits on heat flux hot channel factor, and nuclear enthalpy rise hot channel factor ensure that 1) the design limit on peak local power density is not exceeded, 2) the DNB design criterion is met, and 3) in the event of a LOCA the peak fuel clad temperature will not exceed the 22OO'F ECCS acceptance criteria limit. Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is sufficient to insure that the limits are maintained provided:

a. Control rods in a single group move together with no individual rod insertion differing by more than i 12 steps, indicated, from the group demand position,

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

c. The control rod insertion lbsits of Specifications 3.1.3.5 and 3.1.3.6 are maintained.

j d. The axial power distribution, expressed in terms of AXIAL FLUX i DIFFERENCE y is maintained within the limits. F%gwillbemaintainedwithinitslimitsprovidedconditionsa.throughd. l above are maintained. TherelaxationofF%gasafunctionofTHERMALPOWER

allows changes in the radial power shape for all permissible rod insertion limits.

When an Fg measurement is taken, an allowance for both experimental error and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full core map taken with the incore detector flux mapping system and a 3% allowance is appropriate for manufacturing tolerance. The heat flux hot channel factor Fg(E) is measured periodically and increased by a cycle and height dependent power factor appropriate to RAOC operation, W(I), to provide assurance that the limit on the heat flux hot channel factor Fg(1) is met. W(5) accounts for the effects of normal operational transients within the AFD limits and was determined from expected power control maneuvers over the full range of burnup conditions in the core. The W(E) function for normal operation and the AFD limits are provided in the COLR per Specification 6.9.1.11. FARLEY-UNIT 2 B 3/4 2-2 AMENDMENT NO. I

4 ADMINISTRATIVE CoffrROLS MONTHLY OPERATING REPORT 6.9.1.10 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORV's or safety valves, shall be submitted on a monthly basis to the Commission, pursuant to 10 CFR 50.4, no later than the 15th of each month following the calendar month covered by the report. l CORE OPERATINO LIMITS REPORT 6.9.1.11 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

1. SHUTDOWN MARGIN limit for MODES 1, 2, 3, and 4 for Specification 3/4.1.1.1,

2. SHUTDOWN MARGIN limit for MODE 5 for Specification 3/4.1.1.2,

3. Moderator Temperature Coefficient BOL and EOL limits and 300 ppm and 100 ppe surveillance limits for Specification 3/4.1.1.3,

4. Shutdown Bank Insertion Limit for Specification 3/4.1.3.5,

5. Control Bank Insertion Limits for Specification 3/4.1.3.6,

6. Axial Flux Differ,ence limits for Specification 3/4.2.1,

7. Heat Flux Hot Channel Factor FgRTP limits, K(I) figure, W(I) values, and l Fg(E) Penalty Factors for Specification 3/4.2.2,

8. Nuclear Enthalpy Rise Hot Channel Factor limits, FAH , and Power Factor Multiplier, PFaH, for Specification 3/4.2.3.

The analytical methods used to determine the core operating limits shall be l those previously reviewed and approved by NRC in: 1

1. WCAP-9272-P-A, " Westinghouse Reload Safety Ivaluation l Methodology,* July 1985 (H Proprietary).

l (Methodology for Specifications 3.1.1.1 - Shutdown Margin - Tavg > 200'F, 3.1.1.2 - Shutdown Margin - Tavg s 200'F, 3.1.1.3 - Moderator Temperature coefficient, 3.1.3.5 - Shutdown Bank Insertion Limit, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.) i l l FARLEY-UNIT 2 6-19 AMENDMENT NO.

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

l ADMINISTRATIVE CONTROLS

2. WCAP-10216-P-A, Rev. lA, " Relaxation Of Constant Axial Offset Control /

                   #g Surveillance Technical Specification," February 1994 (H Proprietary).

(Methodology for Specifications 3.2.1 - Axial Flux Difference and 3.2.2 - Heat Flux Hot Channel Factor.)

3. WCAF-10266-P-A, Rev. 2, "The 1981 Version of Westinghouse Evaluation Model Using BASH Code," March 1987 (H Proprietary).

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.) The core operating limits shall be determined so that all applicable limits l (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident I 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 l the NRC Document Control Desk with copies to the Regional Administrator and l Resident Inspector. .

           &M[ PAL DIESEL CENERATOR RELIABILITY DATA REPORT 6.9.1.12               The number of tests (valid or invalid) and the number of failures to start on demand for eagh diesel generator shall be submitted to the NRC annually.

This report shall conthin the information identified in Regulatory Position C.3.b of NRC Regulatory Guide 1.108, Revision 1, 1977. l i FARLEY-UNIT 2 6-19a AMENDMENT NO.

I l l ATTACHMENT III FARLEY NUCLEAR PLANT

 . CORE OPERATING LIMITS REPORT Unit x Cycle yy (Example) l l

s'

t Unit x Cycle yy l EXAMPLE COLR FOR FNP l 1.0 CORE OPERATING LIMITS REPORT This Core Ogi. hug Limits Report (COLR) for FNP has been prepared in accordance with the requirements of Technical Specificaten 6.9.1.11. The Technical Specifications affected are listed below: l 3/4.1.1.1 Shutdown Margm - Modes 1,2,3, and 4 l 3/4.1.1.2 Shutdown Margm - Mode 5 3/4.1.1.3 Moderator Temperature Coefficient 3/4.1.3.5 Shutdown Rod Insertion Limit 3/4.1.3.6 Control Rod Insertion Li: nits,. , l 3/4.2.1 Axial Flux Difference l 3/4.2.2 Heat Flux Hot Channel Factor- Fo (Z) l l 3/4.2.3 Nuclear Enthalpy Rise Hot Channel Factor - FL l l l l l l l l I i I

Unit x Cycle yy I EXAMPLE COLR FOR FNP i ! 2.0 OPERATING LIMITS The cycle-spec 6c parameter limits for the speci6 canons listed in sectma 1.0 are pr==weri in the followmg subsections. 'these limits have been developed using the NRC-approved methodologpes specified in Technical Speci6catma 6.9.1.11. 2.1 Shutdown Marrin - MM- 1. 2. 3 and 4 (Specificatmo 3/4.1.1.1) 2.1.1 h SHUTDOWN MARGIN shall be greater than or equal to 1.77 percent delta k/k. 2.2 Shutdown Marma - MMa 5 (Speci&a*6 3/4.1.1.2) 2.2.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.0 percent delta k/k. 2.3 Moderator Tsaseture Coefficieat (Specificauon 3/4.1.1.3) 2.3.1 The Moderator Temperature Coefficient (MTC) shall be:

a. Less than or equal to 0.7 x 10" delta k/k/'F for the all rods withdrawn, hot zero power,

                                         '44 of cycle life (BOL) enaAtiaa, for power levels up to 70% THERMAL POWER with a knear ramp to 0 delta k/kf'F at 100% THERMAL POWER.

d

b. Less negapve than 4.3 x 10 delta k/k/*F for the all rods withdrawn, end of cycle life (EOL), RATED THERMAL POWER enaA*iaa 2.3.2 The MTC Surveillance limitis:

a. The MTC shall be measured at any THERMAL POWER and compared to d

                                         -3.65 x 10 delta k/kf'F (all rods withdrawn, RATED THERMAL POWER condition) after reachmg an equilibrium boron enacearraten of 300 ppm.

b. Once the equilibrium boron concentration (all rods withdrawn, RATED 'IRERMAL l POWER condition) is 100 ppm or less, further measuranent of the MTC in accordance with 4.1.1.3.b may be su_W. provxhng that the measured MTC at an ;

equilibrium boron concentration less than or equal to 100 ppm is less negative than ! 4

-4.0 x 10 delta k/k/ F.

2.4 Shutdown Rod fa=~ tion I imit (Specification 3/4.1.3.5) 2.4.1 All shutdown rods shall be fully withdrawn (225 to 231 steps, inclusive). 2.5 Control Rod Insertion I imita (Specification 3/4.1.3.6) 2.5.1 h control rod banks shall be limited in physical inscruon as shown in Figure 1. 2.6 Axial Flux Differt;nce (Specification 3/4.2.1) { Relaxed Axial Offset Control (RAOC) methodology} l .' 2.6.1 The Axial Flux Difference (AFD) acceptable operation limits are provided in Figure 2. 2

l' Unit x Cycle yy EXAMPLE COLR FOR FNP 2.7 Heat Flux Hot Channel Factor - Fo(Z) (Specification 3/4.2.2) (Fo methodology) pan 2.7.1 Fo (Z)s P x K(Z) for P > 0.5 . pan Fo (Z) s 0.5 " x K(Z) for P s 0.5 where: P = RATED THERMAL POWER 2.7.2 F"" n = 2.45 (VANTAGE 5)

                  = 2.32 (LOPAR) 2.7.3  K(Z)is provided in Figure 3 2.7.4   F*(Z) o    s  F"" x K(Z) for PO 0.5 P x W(Z)

Fq (Z)s F"" x K(Z) for P s 0.5 C 0.5 x W(Z) 2.7.5 W(Z) values are provided in Figures 4, 5, 6, and 7. C 2.7.6 The F (Z) penalty factors are provided in Table 1. 2.8 Nuclear Fnthminy Rise Hot Chanel Famr - F5 (Specification 3/4.2.3) 2.8.1 F5 s Fi" x (1+ PF, x (1- P)) p_ THERMAL POWER RATED THERMAL POWER 2.8.2a F"" = 1.30 for LOPAR fuel, and 2.8.2b F"" = 1.70 for VANTAGE 5 fuel 2.8.3 PF, = 0.3 for LOPAR and VANTAGE 5 fuel ,, 3

Unit x Cycle yy EXAMPLE COLR FOR FNP i Table 1 Fj(Z) PENAL'IY FACTOR Cycle Bumup Fj(Z) (MWD,MTU Penalty Factor i l 4820 1.021 4981 1.025 ' 5142

1.029 5303 1.030 5464 1.029 l 5625 1.028 l 5786 1.027 "5947 1.026 6108 1.024 l 6269 1.023 6430 1.021 i Notes: 1 ! 1. 'Ibc Penalty Factor, to be applied to Fj (Z) in accordar.ce with Surveillance 4.2.2.2.f, is the maxunum factor by wiuch Fj (Z) is avW to increase over a 39 EFPD interval (surveillance interval of 31 EFPD plus the maxunum allowable extension not to exceed 25% of the surveillance mterval per Technical Specificata 4.0.2) startmg from the burnup at winch the Fj (Z) was determined.

2. Linear interpolata is W* for intenrMate cycle burnups.

3. For all cycle burnups outside the range of the table, a penalty factor of 1.0200 shall be used.

4

1l Unit x Cycle yy (FULLY WITHDRAWN-225 TO 231 STEPS, INCLUSIVE) 231 225 - f

m /

                                      /
                                   /     \                                  ,
                               /          'BAEC                        j/

l / 1s0 / / [ l

                 /                     i
                                                           /                              1
             /                              -

_ /\ h / \ BANK D E / 8 100 / a '

8 s

0 0 .2 .4 .6 .8 1.0 (FULLED INSERTED) FRACTION OF RATED THERMAL POWER 1 ROD GROUP INSERTION LIMITS VERSUS THERMAL POWER THREE LOOP OPERATION

l l

Figure 1 J i l

1 Unit x Cycle yy 120 100

                                                   /               \

(-12,joo) (+g,100) ump [ L um o ,, / \ w [ /

f 3 f 60 e g

                                      /
                                     <e. .>
                                                                             \

(.24. ., 40 1 .. . a 20

                          -50   -40   -30   -20      -10     0     10      20     30     40    50 Axial Flux DWerence (percent 40 AXIAL FLUX DIFFERENCE LIMITS AS A RJNCTION OF THERMAL POWER FOR RAOC Figure 2 l

Unit x Cycle yy 1.20 0.0,1.0 6.0,1.0 1.00 12.0,0.933 0.80 2 0.60 Z e 0.40 R y . 0.20 0 0 2.0 4.0 6.0 8.0 10.0 12.0 Core Height (ft) K(Z)-Normakzed Fn (Z) AS A FUNCTIDs 0F CORE BEIGET Figure 3

1 f 1.60 Aust Dev.inon ROL ' N uit (fe.o w, , , 1

12.00 Iuun 2 1130 i (s u o 3 1140 i tuvu 4 II.40
lion 5 11 ; 0 1.50
6 i tuu 11 00 lion 7 10J0 4
i uol i

8 1040 i to o 9 10.40 f

1u000 2 10 10;0 1#00) 11 10D0 11737 1

a 12 92 1.1582 13 9M 1.1448 ! 1.40 14 4e 1.1329 1 A 15 92 1.1243 16 48 1.1107 17 8B 1.1049

                          "                                                              18              8A0              1.1127         <

19 8.5 1.1199 l 20 2 1.1258 l _ g 21 SB N 1.1304

       ~1                                                                               22               7B             1.1336 S .30                 a 23               7m             i.i355 24               7.5            1.1361           1 25               7.31           1.135) u                                                         26               75             1.1336 27               62            1.1305            1 28               6D            1.1262 29                Le            1.1208 1.20                   3                                                     30                @             1.1146
                                  ,A

at sm 1.106:

a& 32 13 1.1050 A n 33 5m 1.1139 I A A 34 15 1.1221 i a

                                            "                                         35                 13            1.1308 A       @         &
                                                                  &                   36                 15           1.1406 A                                       37                 W            l.1504 1.10                                  &"              6                     38                  W            l.1593 39                  w            1.1677 40                 G            1.1754 41                  W           l.1826 42                 33           1.1892 43                 3m           1.1957.

44 3.4 12050 45 3m 12223 46 3m 12490 1 00 47 2m i 8:0 0 2 4 6 8 10 12 4s 2m IJi41

                                                                                     #                  2.s         i.3476 CORE HEIGHT (Fest)                                50                 2m          IJ812 Top       51                  23          1.4143 Bottom                                                            .

52 13 1.0000 53 1m 1.0000 54 1.8 1.0000 55 m 1.0000 This figure is referred to 56 13 1.0000 ts by Technical Specifications g gg [g 4.2.2.2d and B 3/4.2.2

59 am 1.a000 60 a3 1.0000 61 48 1.0000 Top and Bottom 15% Excludec per Technical Specification RAOC W(z) at 150 MWD /MTU ,. 4.2.2.2 Figure 4

3, l 1.60 l Azul Elev.inon yg. ; i Pome go,el g 1 12.00 1.0000 2 11.N0 10000 1 3 11.6 0 10000 4 11.40 i.0000 1.50 .' 86 1120 11 00 00N 1.0000 7 10 80 1.0000 8 1040 i0000

9 10.40 1.0000 10 1020 t0000 11 10 00 1.2700 '

12 9 80 1.2469 13 1.227s 1.40 14

                                                                                                                 ,9#60 i .2 42 15

20 1.1971 16 95) 1.1776 17 8E 1.1705 18 840 1.1713 a 19 8.40 1.1740 20 8.20 1.1747

N ^ 2I IE 1.1738

 -1 30                                                                                                 22        7m              1.17 5 3                                                                                                     23 24 7.60 7.40 1.1677 A                                                                                                               1.1625 25        72             1.1559 26        72             1.14al A                                                                                    27        6R             1.1391 A                                         28        6AD            1.1290 A                                      A                                          29        W              l.1175 1.20

30 62 1.1073 y 31 68 1.0973 3 6 32 55 1.0951

                                            /                                                         33 34 5m 18 52 1.1018 1.1000 3"                                                          35                       1.!!37
                                         &                                                            36        55             1.1205
                                        &                                                             37        45             1.1277                     )

1.10

                                       '                                                              38        W              I1339 8'                                                                39        M              1.1395 40        G              1.1445                      ,

41 43 1.1439 i 42 la 1.1523 43 148 1.1564

                            .                                                                         44        18            1.1596 45         la             1.1624 46         15            1.1817 O   2                4      6         8              10              12 d

4' E te lE 1.2725 CORE HEIGHT (Fest) 50 la 1.3039 Bottom Top . Q g jg 53 im i.0000 54 Le 1.0000 55 12 1200 This figure is referred to 56 12 1.0000 {

2 00 by Technical Specifications ,

4.2.2.2d and B 3/4.2.2 59 tm 1.0000 60 43 1.0000

61 tm 1.0000 Top and Bottom 15% Exclude per Technical Specification RAOC W(s) at 3000 MWD /MTU ,

4.2.2.2 Figure 5

linit x Cyclo yy 1 5 1.60

Ami Elevanon stol.

Pome (fusel ,ve n l

L 1 12.00
l um 2 11.80 iU000 3 11a0 10000 4 11.40 1.0000 5 1120 1.0000 1.50
6 11.0 0 1.0000 7 10E Im00 8 1040 1.0000
O 10.40 1.0000 10 1020 10000

11 10D0 12621 l 12 95 12458 13 9A0 1.2288 1.40 14 9.5 12116 15 em i.1965 16 92 1.1974 17 SE I2056 18 EJO I.2105 19 8.5 12150 20 82 1.2175 q 21 8B 1.2182

    -1                                                                                        22                   72             1.2171 3 30                                                          ,.

23 24 7m 7.e 1.2142 12094

                                                                        "                    25                    73             1.2027 a

26 73 1.1954 27 62 1.1871 l & 28 6J0 1.1770 J A 29 6.8 1.1654 1.20 7 30 62 1.1522

                                          ~
                                                   ,a              .=                        31                    us            1.1383
                         &                        &                                          32                   M              l.1216
                                                ,A                                           2                    SS             1.1107
                            &                                                               34                    18            1.1130 l
                              "A                                                            N                     13            1.!!52 1.10
                                %    W%      &

36 37 38 15 W 48 1.1159 l.1163 1.!!58 l 39 te 1.1148 e 43 1.!!$1 4B 41 1.1168 l 42 32 1.1188 43 sm 1.1202 44 18 1.1215 4s sa 1.1246 46 25 1.1313 1.00 47 2m 1.1397 0 2 4 6 8 10 12 4s 2Js 1.1481 i CORE HEGHT(Feet) e 2.e I.i397 50 22 1.1786 Bottom Top , s g g jg

ss im Im00 54 1.4 1.0000 I

This figure is referred to *

                                                                                           !                    IS            I.0000 by Technical Specifications 57                  as             1.0000

58 SJe 1.0000 4.2.2.2d and 5 3/4.2.2 59 ge 1.0000 i
60 82 1.0000 61 45 1.0000
Top and Bottom 157. Excludec per Technical Specification

! RAOC W(z) at 11000 MWD /MTU , 4.2.2.2 Figure 6 I l I - -

1 60 Azul Elevshon EOL l Point p.,, ,

ww 1 12.00 2 ii n 1, ,0000 00 3

4 11 60 1200

11.40 1.0000 1 5 II;o 1.50
ia000 ,

6 11.00 i m00 ' I 7 10 80 1.0000 ' 1

8 10 60 10000 I l

9 10.40 1.0000 10 1020 1.0000 11 10 00 124.a. 12 9JD 12291 13 9A0 1.2174 i ' 1.40 14 '# 12136 i 15 920 1:089 l 16 92 1.2088 17 8.2 1.2124 18 840 1 207 19 8.40 12304 20 8.20 1.2439 21 8.00 12561 51 30 22 7m 12652 3 23 24 7AO 1.2720 gE Aag 7.40 1.2764 A 25 7.2D 1.2775

n 26 72 1.2764 ag A di 27 6m t.2739

                 "                                                        A i                                                                                A                     28         640           1.2692 I                    A                                                                                 N          Le            1.2414 l                                              ..A                           O                         30         @

1.20 A 1.2515 68 AMr 31 1.239' l 32 1m 1.2238 l A A 33 5d0 1.2085 A 34 18 1.2014 a - - 35 13 1.1984 36 1m 1.1944 d 37 m 1.1887 1.10 3s us 1.1811 39 up 1.1720 40 43 1.1616 41 W !.1494 42 la 1.1390 43 1A0 1.1315 44 14 1.1229 45 . M !.1188 46 3m 1.1262 l 1.00 47 2m 1.1476 0 2 4 6 8 10 12 4s 2m 1.1700 CORE HEIGHT (Foot) g y jj$ Bottom %

5152 2m la 1.2363 1.0000 53 la im00 54 lm 1.0000
55 12 1200 This figure is referred to IE by Technical Specifications h gg l$

58 em 1.0000 4.2.2.2d and 5 3/4.2.2 59 44 1.0000 m em 1.0000 i 1 61 W 14000 4 , l Top and Bottom 15% Excludec

 !                                                                                                     per Technical Specification RAOC W(s) at 19000 MWD /MTU                                                              ..         4.2.2.2 l

Figure 7 l}}

ML20128F882

Text

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