Text

Proposed Tech Specs,Removing cycle-specific Operating Limits & Instituting Core Operating Limits Rept,Per Generic Ltr 88-16
	ML20086G647
Person / Time
Site:	Fort Calhoun
Issue date:	11/27/1991
From:	; OMAHA PUBLIC POWER DISTRICT
To:	;
Shared Package
ML20086G466	List: ML20086G462; ML20086G474; ML20086G647;
References
	GL-88-16, NUDOCS 9112050187
	Download: ML20086G647 (80)
	v • d • e

,

TABLEOFCONTENTS_(Continued)

Pace 4.3 Nuclear Steam Supply System (NSSS)....................... 4-3 ,

i 4.3.1 Reactor Coolant Sys, m............................ 4-3 4.3.2 Reactor Core and Contro1. . . . . . . . . . . . . . . . . . . . . . . . . . 4-3 l 4.3.3 Emergency Core Cooling............................ 4-3 4.4 Fuel Storage............................................. 4-4 ;

i 4.4.1 New Fuel Storage.................................. 4-4 4.4.2 Spent Fuel Storage................................ 4-4 4.5 Seismic Design for Class 1 Systems....................... 4-5 5.0 ADMI N I ST RAT I VE CONT R0L S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5- 1 5.1 Responsibility........................................... 5-1 1 5.2 Organization...t......................................... 5-1 5.3 Facili ty Sta f f Quali fica tions . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-la .

5.4 Training................................................. 5-3 l 5.5 Review and Audit......................................... 5-3 ,

-5.5.1 Plant Review Committee (PRC)...................... 5-3 5.5.2 Safety Audit and Review Conmittee (SARC). . . . . . . . . . 5-5 5.5.3 . Fire Protection Inspection........................ 5-8a 5.6 Reportabl e Event Act1on. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-9 l l 5.7 Safety Limit Violation................................... 5-9 5.8 Procedures............................................... 5-9 5.9 Reporting Requirements................................... 5-10 5.9.1 Routine Reports................................... 5-10 5.9.2 Reportable Events................................. 5-12 1 5.9'3 Special Reports................................... 5-15 5.9.4 Unique Reporting Requirements..................... 5-15 5.'! 5 Ca v CMd. in > nth CyrT Puh

. 5.10 Records Retention.......................................

9 5-18 5.11 Radiation Protection Program............................. 5-19 5.12 DELETED .

5.13 Seconda ry Wa ter Chemi s try. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-20 5.14 Sy s t ems I n t e g ri ty . . . . . . . . . . . . .. . . . . . . . . . . . . . . , . . . . . . . . . . . . 5- 21 5.15 Post-Accident Radiological Sampling and Monitoring....... 5 21 6.0 INTERIM SPECIAL TECHNICAL SPECIFICATIONS....................... 6-1 6.1 Limits on Reactor Coolant Pump Operation................. 6-1 6.2 Use of a Spent Fuel Shipping Cask. . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.3 Auxiliary Feedwater Automatic-Initiation Setpoint........ 6-1 6.4 Operation With less Than 75% of Incore Detector Strings 0perable....................................... 6-1 iii Amendment No. 32,$ A, A3,Ef EE,57,

/3,gg,gg,d,gg 9112050187 ?11127 FDR ADOCK 05000295 P FDR !

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TECHNICAL SPECIFICATIONS FIGURES 1

TABLE Or CONTENJT1 FIGURE PAGE WHICH DESCRIPTION '

FIGURE FOLLOWS 11 TMLP Safety limits 4 Pump Operations. . . . . . . . . . 13 1-2 i Axial Power Distribution LSSS for 4 Pump Operation. . .

13 l-3 . " !.! 0 0 7..,, Gen . Hun. '.b p Wr,T r . . . . . . . . .

1-3 2 1A RCS Press-Temp Limits Heatup. . . . . . . . . . . . . . 26 2 1B RCS F-Temp Limits Cooldown. . . . . . . . . . . . . 26 2-3 Predicted Radiation Induced NOTT Shift. . . . . . . . . 26 2 11 MIN BAST Level vs Stored BAST Concentration . . . . . 2-19 l 2-12' Boric Acid Solubility in Water. . . . . . . . . . .-. .

2 19 l 2 10 Spent Fuel Pool Region 2-Storage Criteria . . . . . . . 2 38 24 Pett.. P P W 4 .-.--. .-. . . . . . . . . . . . . . . . 2 53 25 e'r '- o a i4a. . u. + o + - a.

mm .t9t* IE. . . 2 53 2 6- 1,GO- N Ex::r:

";,, iter'rg-of tuq, pqt.gr,s. . . . . . . 2-53 2-7 K0 Te r OG hu n i i a r i n g .

9 %E70 . . . . . . . . . . . 2 53 28 Flux Peaking Augmentation factors . . . . . . . . . . . 2 53 29 l,F[y.r.d-Cerm rear Lir,itniens. 9.etE.TP . . . . , _

F g 2-53 i

1 1

l viii Amendment No. J15,126.131 r l

l- j L -

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l DEFINITIONS

REACTOR OPERATING CONDIT10 % (Continued) f.gid Shutdown Condition (Operating Mode 4)

The reactor conlant Tcold is less than 210*F and the reactor coolant is at shutdown boron concentration.

Refuelina Shutdown Condition (Operating Mode 5)

The reactor coolant is at refueling boron concentration and T cold is less than 210*F.

Refuelina coeration Any operation involving the shuffling, removal, or replacement of nuclear fuel, CEA's, or startup sources.

The Refuelina Boron Concentration -thd .5pecdWd in i he

( Ccce Cp55 Lov ct n 94PwT.

A reactor coolant boron concentration of at least -1900-pom, which corresponds I to a shutdown margin of not less than 5'/. with all CEA's withdrawn.

Shutdown Bor M qncentration The boron concentration required to make the reactor subtritical by the amount defined in. paragraph 2.10.

bedhco Refuelino Outaae or Refuelina Shutdown A plant outage or shutdown to perform refueling operations upon reaching the planned fuel depletion for a specific core.

Plant Ooeratina Cvele The time period from a Refueling Shutdown to the next Refueling Shutdown.

Amendment No. 2!,32,/1,/3,Jp3,133 2

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

?ETUl!"'C'!S Asinutr.a1 Fever ? tit -

Asimuthal Power Tilt shall be the maxirum difference between the power generated in any core quadrant (upper er lever) and the average pcVer of all quadrants in that axial half (upper or icver) of the core :11vided by the average power of all quadrants in that ' axial half ( upper or icver) of the core.

Unrodded Planar Padial Fankine Tsetor - F 9 The Unrodded Planar Radial Feaking Factor is the =axt=u ratio of the peak to average power density of the individual fuel rods in any of the unrodded hori: ental planes , excluding ati=uthalg tilt , T . The. unw2m F,3 lina le Unredded ? .tearnted Fadial Paskina Fsetor - F3 in iki C<ve gm bm Limiti S p1 h 0

"he Unrodded Integrated Radial Peaking Factor is the ratto of the pea., pin power to the average pin power in an un '^dded core, excluding a:1muthal tilt T . 'Ift g modrnton p (;y; j. b . ,pyylcd ,n & Ogf_ O ryutn, [!mih [r Ay f .

g p Fire Suteression Water System The fire suppression vater system consists of fire ] z.mps and distribution piping with associated sectionalizing control or isolation valves. Such valves include yard hydrant curb va.1ves , and the first . valve ahead of the water flov alars device on each sprinkler, hese standpipe or spray system riser.

' Prceess centrol P*orrns ( PCP)

A manual or. set of operating procedures detailing the program of sa=pling, analysis , and evaluation.

Done Fouivalent T-111 L

hat concentration of :-131 (uci!sm) which alene vould produce the same thyroid dose as the quantity and isotopic mixture of I-131, :-132, :-133 , .

I-13L and :-135 actually present.

( In other verds .

7 Amendment :lo. 32 , .13, 67, %

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se Equivalent -111 s s Ci / c /

U C1/ A o f I- 131

+ 0.0361 x aci/ m cf I-1:0

+ 0.270 x aci/c of I .~ 33 i 1

+ 0.0169 x LCi/ c of *.-iTw

+ 0. 08 36 x u C1/ g. o f I .' 3 5 2 - Avarnee M sinterrit ten Enore 5 in the average (veighted in pretortlen to the e:ncentration of eacn rsdionuclide in the te At: tor coolant at the ti:::e of sampling) of the sum of the average beta and gacca energies per disintegratien, in Mrl, for isotopes, Other than iodines , vith half lives greater than li minutes takir.g up at le as t 9!, of the total non-lodine radicactivity in the C001 ant.

Of fs it e Pet. a hieulat ten Meurtl ( cPC.M .~ - )

A manus,1 containing the methodology u2d paranaters to be used in thet

1) calculation of doses in the unrestricted area due to radioactive liquid and pscous effluents, ) calculatien or liquid and gaseous effluent monitoring instrumentation setpoints, and 3) specific details pertinent to the radiological environmental monitoring program.

N r c - Pu r ri le, A teans for the re= ova.1 and :eplseement Of gasee within the centainment building.

< ant i n e

A teans for the reductien of pressure gester than atecopneric within tne

entainment structure.

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Amendment :io. 67, f6

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l.0 9f(TY tim 115 tJ49 L]MJlll@ StJ[TY SYS1[M $[TT]fMS 1.1 iLfit y l i mi t s - P ect t or ( p re (C on t i nued )

would cause Df48 at a particular core locl. tion to the actual heat flux at that location, is indicative of the margin to DfiB. The minimum value of the DlJBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.18. A Df4BR of 1.18 corresponds to a 95% probability at a 95% confidente level that DdB will not occur, which is considered an appropriate margin to Of4B for all operating conditions.(1)

The curves of figure 1 1 represent the loci of points or reactor thermal power (either neutron flux instruments or a T instruments), reactor coolant system pressure, and cold leg temrarature for which the DfdBR is 1.18. The area of safe operation is below these lines.

The reactor core safety limits are bated on radial peaks limited by the CEA insertion limits in Section 2.10 and axial shi. pes within the axial I power distributign trip limits in figure 12 and a total unrodded planar radial peak (f 3he4&SS-in469ure44 Js basedgn tha I anumptlon4behtW)-u.oMJ5 n rodde d-i n t eg ra t ed_to ta Lra d i a l-pea k- g tr70 M h b-peakin t-factoMs-tlight4Phigher-4more-conser-v(T -) is at4ve)_than I the-maximum-predie' ed-unrodded-t ot til-r a d i a bp ea k-d u r4 ng-c o re-l i f e r excluding measuremont._unce rtaint y ,

flow maldistributi >n effects for operation under less than full reactor coolant flow have 3een evaluated via model test.(2) The flow model data established the ma ldistribution factors and hot channel inlet temperature for the thermal an alyses that were used to establish the safe operating envelopes presente i in figure 1-1. The reactor protective system is designed to preven t any anticipated combination of transient conditions for reactor that would coolant result in system temperature, pressure, and thern.a1 power level l a Dt4BR of less than 1.10.(1)

References (1) USAR, Sectiol 3.6.6 i (2) USAR. Sectiom 1.4.6 1

as spciGed m +he OcuR. Th e The, .wd Nn c$ i o /

tcw b mre irip movements 9 wit be uxthm the h.n m pro wci d in ihe ccLR, The The e.,wt

~, nwqm/ Ln 3 PNmce vp h bued en an uncodded AbynTed -tc%\ rac%\ pmk (\Q )

Syd il predided in %e OctR, l-2 Amendment flo. E,32,;f 3, (

A7,70,77, 97,117,126

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60 70 SO 90 100 110 120 CORE POWER (". OF RATED POWER)

P ygg = M.W@($ + 18.W - 112 W ig 1

PF(B) = 1.0 01100"..

= - 0088 + 1,8 50%<B<100%

o = 1.4 i Gs50%

[ . A1(Y) = -0.35294Yi + 1.08824 Y, <_ . 2 5

=

i- 0.57143Y; + 0.875 Yj > .25 Thermal Marcin/ Low Pressure LSSS Omono Public Power District I Figure 4 Pomo Ooerc. tion fort Calhoun Station-Unit No. i 1-3 Amendment No. 8,20.47,70,77.92, 109, 117, 126

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

N 9

i 1

1.0 SAFETY LlHITS AND LIMIT!!(G SAFETY SYSTEM SETTINGS 1.3 Limitino Safety System Settinas, Reactor Protective System (Continued)

(3) High Pressurizer Pressure - A reactor trip for high pressurizer pressure is provided in conjunction with the reactor and steam system safety valves to prevent reactor coolant system overpressure (Specification 2.1.6). In the event of loss of load without reactor trip, the temperature and pressure of the reactor coolant system would-increase due to the reduction in the heat removed from the coolant via the steam gener-ators. The power-operated relief valves are set to operate concurrently with the high pressurizer pressure reactor trip. This setting is 100 psi below the nominal safety valve setting (2500 psia) to avoid un-necessary operation of the safety valves. This setting is consistent with.the trip point assumed in the ac-cidentanalysis.(1)

(4) _T__hermal Margin / Low Pressure Trip - The thermal margin /

low pressure trip is provioco to prevent operation when the DNBR is less than 1.18, including allowance for j measurement error. The thermal and hydraulic limits

' shown .cn.54 sum 3 define the limiting values of re-actor coolant pressure, reactor inlet temperature, axial shape index, and reactor power level which ensure that f the thermal criteria (e) are not exceeded. The low set point of a 1750 psia trips the reactor in the unlikely event of a loss-of-coolant accident. The thermal margin /

low pressure trip set points shall be set according to the formttla-given-on4fgure-lv3.- The variebiet-irr-the 4ermule-ered efir1Ed 3T: d -

B-

-a Jiigh-auctioncor4d-thermal-(tT3-oMtitieer powee- [

in4-o(-r4ted power.

T.I C = rnra inlet-temperaturev "f.

7 eseter pressurer psiav- )

P Y p- Ax.ial-Shepe-frideti etiv 3

L in +Ile (' essure 4 Amp Ojxat: hen rp[re~UWnwl omh4 Ohgh LoN in -h cot.R -

pTgyn cpwn in -ihe CdR

[W 1enm( Dh Apr[Lc"O

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Amendment No. E.20,12. 1-8

7,70,71. 92 h

TABLE 1-1 P

R_PS LIMITING SAFETY SYSTEM SETTitlGS No. Reactor Trio Trio Setooints 1 High Power Level (A) 4-Pump Operation 3107.0% of Rated Power 2

Low Reactor Coolant Flow (B)(F) 4 Pump Operatioa 1955 --f 4 Pump Flow 3

Low Steam Generator Water Level 31.2% of Scale (Top of feedwater ring; 4'10" below normal water level) 4 Lo's Steam Generator Pressure (C) 3500 p:ia 5 High Pressurizer Pressure 32400 psia 6 Thermal Margin / Low Pressure (B)(F) 1750 psia to 2400 psia (depending on the re-actor coolant temper-ature as showr in F49 ttM-3-)

7 High containment Pressure (D) 15 psig 8

Axial Pcwer Distribution (E) (Fijure 1-2) ;

9 Steam Generator Differential Pressure 3135 psid

_-Hw T}wenwl Okut;n1[ LOC 4 Emp opcmT<m Ryre P "' #,

in 4ht CCLR Amendment No. 7,M,/7,77,92 1-10 1

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

2.0 LIMITING CONDITIONS FOR OPERATION 2,3 Emeraency Core Coolino System AQol icabil_it.y Applies to the operating status of the emergency core cooling system.

Ob.iective To assure operability of equipment required to ra a decay heat from the core.

Specifications (1) Minimum Reauirements The reactor shall not be made critical unless all of ti following conditions are met:

a. The SIRW tank contains not less than 283,000 gallons of water with a boron concentration of at-least 00-ppm-at a temperature l not less than 50*F. +he a b bw fxem covtenTreebed b.- One means of temperature indication (local) of the SIRW tank is operable.

c, All four safety injection tanks are operable and pressurized to at least 240 psig with a tank liquid of at least 116.2 inches (67%) and a maximum level of 128.1 inches (74%) with refueling baron concentration,

d. One level and one pressure instrument is operable on each safety injection tank.

e. One low-pressure safety injection pump is operable on each bus,

f. One high-pressure safety injection pump is operable on each bus.

g. Both shutdown heat exchangers and three of four component cooling heat exchangers are operable,

h. Piping and valves shall be operable to provide two flow paths from the SIRW tank to the reactor coolant system.

i. All valves,- piping and interlocks associated with the above components and required to function during accident conditions are operable. HCV-2914, 2934, 2974, and 2954 shall have power removed from the motor operators by locking open the circuit breakers in the power supply lines to the valve motor operators.

FCV-326 shall be locked open.

2-20 Amendment No. J7,JJ,/J,Jp),JJ7,JJ9 133

2.0 LIMITING CONDITIONS FOR OPERATION -

2.3 Emeraency Core Coolina System (Continued)

(3) Protection Aaainst low Temperature Overoressurization The following limiting conditions shall be applied during scheduled heatups and cooldowns. Disabling of the HPSI pumps need not be required if the reactor vessel head, a pressurizer safety valve, or a PORV is removed.

Whenever ti.a reactor coolant system cold leg temperature is below 320'F, at least one (1) HPSI pump shall be disabled.

Whenever the reactor coolant system cold leg temperature is below 312'F, at least two (2) HPSI pumps shall be disabled.

Whenever the reactor coolant system cold leg temperature is below 271*F, all three (3) HPSI pumps shall be disabled.

In the event that no charging pumps are operable, a single H'"I pump may be made operable and utilized for boric acid injection to the Core.

Basis The normal procedure for starting the reactor is to first heat the reactor coolant to near operating temperature by running the reactor coolant pumps. The reactor is then made critical by withdrawing CEA's and diluting boron in the reactor coolant. With this mode of start-up, the energy stored in the reactor coolant during the approach to criticality is

-substantially equal to that during power operation and therefore all engineered safety features and auxiliary cooling systems are required to be fully operable. - During low power physics tests at low temperatures, there is a negligible amount of stored energy in the reactor coolant; therefore, an accident comparable in severity to the design basis accident is not possible and the engineered safeguards systems are not required.

The SIRW tank contains a minimum of ?8

! containing at4 east-1900-ppm 4oronN.3,000 This gallons of usable is sufficient boronwater i concentration to provide-a shutdown margin of SY., including allowances for_ uncertainties, w temperature of 60*F.g a'l control rods withdrawn and a new core at a

a. A beon emcedredw) c4 ct't tenst 4he re h G.v barco cov1ccmweltm.

The limits for the safety injection tlnk pressure and volume assure the required amount of water injection during an accident and are based on values used for the accident analgses. The minimum 116.2 inch level

! corresponds to a volume of 825 ft gndthemaximum128.1inchlevel l corresponds to a volume of 895.5 ft

Prior to the time the reactor is brought critical, the valving of the l safety injection system must be checked for correct alignment and l

appropriate valves locked. Since the system is used for shutdown cooling, i the valving will be changed and must be properly aligned prior to start-up l of the reactor.

, 2-22 Amendment No. J7,JS,/3,f7,hf,7/,

17,199,197, n3

2.0 llHITING CONQli10NJ FOR OPERATION 2.8 Refuelina Operations (Continued) incident could occur during the refueling operations that would result in a hazard to public health and safety.(1) Whenever changes are not being made in core geometry one flux monitor is sufficient. This permits maintenance of the instrumentation. Continuous monitoring of radiation levels and neutron flux provides immediate indication of an unsafe condition. The shutdown cooling pump is used to maintain a uniform boron concentration.

The shutdown margin as indicated will keep the core subcritical even if all CEA's were withdrawn from the core. During refueling operations, the reactor refueling cavity is filled with approximately 250,000 gallons of borated water. The boron concentration of this water (4t least 900 ppm- l boron) is sufficient to maintain the reactor subcritical by mor than Sr.

including allowance for untertainties, in the cold condition wi h all rods withdrawn.(2) ncentration ensure the proper shutdown Periodic checks margin. of refueling Communication water boron [ allow requirement the control room operator to inform the refueling machine op.erater of any impending unsafe condition detected from ,the-main mentrol board indicators during fuel movement. \ of <Ct least 4he-boccm ccvcest%%n. 'T bh n c)

In addition to the above engineered safety features, interlocks are utilized during refueling operations to ensure safe handling. An excess weight interlock is provided on the lifting hoist to prevent movement of more than one fuel assembly at a time. In addition, interlocks on the auxiliary building crane will prevent the trolley from being moved over storage racks containing irradiated fuel, except as necessary for the handling of fuel. The restriction of not moving fuel in the reactor for a period of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after the power has been removed from the core takes advantage of the decay of the short half-life fission products and allows for any failed fuel to purge itself of fission gases, thus reducing the consequences of fuel handling accident.

The ventilation air for both the containment and the spent fuel pool area flows through absolute particulate filters and radiation monitors before discharge at the ventilation discharge duct. In the event the stack discharge should indicate a release in excess of the limits in the technical specifications, the containment ventilation flow paths will be closed automatically and the auxiliary building ventilation flow paths will be closed manually. In addition, the exhaust ventilation ductwork from the spent fuel storage area is equipped with a charcoal filter which will be manually put into operation whenever irradiated fuel is being handled.(1)

References (1) USAR, Section 9.5 (2) USAR, Section 9.5.1.2 2-39 Amendment No. Jf,7E,JN.JJ7, 133

2. 0 LIMITING CONDITIONS FOR OPERATION 2.10 Renetor Core (Continued)

~ 2.10.2 Reactivity Control Systems and Core physica Parameters Li nits (Continued)

Control Element Assemblies (b) Full Ieneth CEA Position Durine Power Operation All full length (shutdown and regulating) CEA's shall be oper-able with each CEA of a given group positioned within 12 inches (actual position) of all other CEA's in its group. If one or more of the CEA's is inoperable or misaligned, determine the cause and comply v h one of the following s

a. If one or more full length CEA's are inoperable due tot

1) being immovable as a result of excessive friction or mechanical interference, or 2) known to be untrippable, determine that the shutdown margin requirement of Speci-fication 2.10.2(1) is satisfied within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in at least hot shutdovn within 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months ,

b. With one full length CEA inoperable due to causes other than addressed in item a. above, and inserted beyond the Long Tern Steady State Insertion Limits but within its above specified alignment requirements,. power operation may continue for up to 7 EFPD's per occurrence with a total accumulated time of <14 EFPD per calendar year.

c. With one full length CEA inoperable due to causes other than addressed in item a. above, but within its above specified alignment requirements and either fully with-drawn or above the Long Term Steady State Insertion Limits if in CEA group h, power operation may continue,

d. With one or more full length CEA's misaligned from any other CEA's in its group by more than 12 inches but less than 18 inches (actual position) vithin one hour either:

i) Restore the misaligned CEA(s) to within 12

,dev& Insect W inches (actual position) of any other CEA's

,7 f u,y provid Kk in its group (realignment shall be made while 9 q maintaining the allowable CEA sequence and N Od -

CEA insertion limits m-. , iir m M ); or 1u (ii) Declare the CEA's inoperable. Power operation may continue provided all of the following conditions are met:

1. The power level shall be reduced to <70".

of the maximum allovable power level for f the existing Reactor Coolant Pump com-bination within an additional one hour; l' if negative reactivity insertion is re-i quired to reduce power, boration shall I

be used.

AmendmentNo./. , 32 2-50a

/

2.0 LIMITING CONDITIONS FOR OPERATION h',i, 2.10 Reactor Core (Continued) 2.10.2 Reactivity Control Systems and Core Physics Parameters Limits (Continued)

2. Within one hour after reducing the power as required by 1., above, the remainder of the CEA's in the group with the inoperable CEA shall be aligned to within 12 inches (actual position) of the inoperable CEA of Yb bef while maintaining the allovable CEA sequence ThpcicM hecho and in_sertion limits chpvn=on-Mgure 2- b ;

the power level shaBle restricted pur-M M'he6)u eCdJ h, mvtde suant to the applicable section of b. or M A c. above during subsequent operation.

or (iii) Be in hot shutdown within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

e. With one full length CEA misaligned from any other CEA's in its group by 18 inches or more (actual position), re-duce power to <70% of the maximum allowable power level for the existing Eeactor Coolant Pump combination within one hour. If negative reactivity insertion is required to reduce power, boration shall be used. Within one hour after reducing power as required above, either:

_(j ) Restore the CEA(s) to within 12 inches (actual db Otoer position) of any other CEA's in its group g ,yl g t 'Ls.trTwvi (realignment shall be made while maintaining the allovable CEA sequence and CEA insertion Le g7yce Pmv3ef m the C4LR ~

units shoymp); or (ii) Declare the CEA(s) inoperable and determine that the shutdown margin requ!rement of Specification 2.10.2(1) is satisfied and align the remainder of the CEA's in the group with the inoperable CEA to within 12 inches (actual positic,n) of the inoperable CEA while maintaining the allow-of iht ,bM '~ able CEA sequence and insertion limits aho w gggg h,2,4e1 -e r-Figure-2--4 The shutdown margin require-ment of Specification 2.10.2(1) shall be deter-Li dT E)un+ 04vd(g I mined at least once per shift. The power level in ~kh CCd d shall be restricted pursut.nt to the applicable section of b. or c. above during subsequent operation; or (iii) Be in hot shutdown within an additional 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

f. With more than one full length CEA inoperable or mis-aligned from any other CEA in its group by 18 inches (actual position) or more , be in at least hot shutdown g within 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months .

l \

l Amendment No. 32 2-50b

2.0 LIMITING CONDITIONS FOR OPERATION 2.10 Reactor Core (Continued) 2.10.2 Reactivity Control Systems and Core Physics Parameters Limits (Continued)

(5) Non-trippable CEA Position Durina Power Operation i All non-trippable CEA's (NTCEA) shall be withdrawn to at least 114 inches (actual position). If one or more NTCEA's becomes misaligned from other NTCEA's by more than 12 inches (actual position) either:

I

a. Restore the NTCEA to within the specified alignment requirements within one hour, or

b. Be in at least hot shutdown within an additional 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

(6) Shutdown CEA Insertion Limit During Power Operation All shutdown CEA's shall be withdrawn to at least 114 inches as a condition 'for reactor criticality, or with one or more' shutdown CEA's inserted to more than 114 inches withdrawn, except for surveillance testing, within one hour, either:

a. Withdraw the CEA's to at least 114 inches, or

b. Declare the CEA's inoperable and apply Specification y 2.10. 2( 4 ) .

(1)_ Regulating CEA Insertion Limits During Hot Standby and Power siTicviec( uithsi Operation 4lw ttccep%Ie opMiig regulating CEA groups shall bekimi-ted-to-the inser44on.

ruuje Oc ree/AT"9 sec uence and-to-theinsertion-limits 1hown-on-F-igure 2-4-except rio PT'm o f durcing CEA exercises above 114 inches. With all CEA's operable, th Pmde DPM CEA insertion beyond the Long Term Insertion Limits are restricted 1,eerTton - unur to-Q my - prcVvit 1. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval,

%e COLR -

- 2. 4 EFPD per 30 EFPD interval, and

3. 14 EFPD per calendar year,

a. When the regulating CEA groups are inserted beyond the Transient Insertion Limits within two hours, either:

(i) Restore the regulating CEA groups to above the Transient Insertion Limits, or (ii) Reduce reactor power to the allowed power of f -Hgure-B-4 which pemits continued operation

{ above the Transient Insertion Limit using the existing CEA group position.

ihe Poive c Dry, dent herTic-n Le T INfft

} of %e CULR Amendment No. 32, 109 2-50c

s * '

f I I I l

- 12 8 10 0.8 LONG TERM STEADY STATE (94.5 INCHES WITHDRAWN)

INSERTION L'MIT

, 7S.6 h SHORT TERM 3 -50.4 -626 ------- -- -"'

IN $t.i T10N LiusT 2S.2 80Q8

-0 716 7

M

,50.4 O

.* $ Y R A N $1( N T INSE 9 TION

$ .-62 6 212 #

a l

2 -nOQ8 - 0 S - -

O g" -75.6 6* 50.4

-t S -

d3

-212-126

-0 10 0.0

-716

, -50.4 8

i

-212

-0 ' I I i -- I 0 20 40 60 AEACTOM POW E R, ( % Of ALLOWE D POWER )

to co POWER DEPENDENT _

OMAHA PUBUC POWEft DISTPICT INSERTION LIMIT FIGURE FORT CALHOUN STA110N tlNTT NQ l 2*4 Arnendmen t No. 3. 20, TJ , 70

[4Lb1L - -((<pm , r es e pt.

16 I l i i i C

t b

'4 m 15 -

H -

< UNACCEPTABLE OPERATION 2

H

.j 14.4 KW/FT

=

x j 14 ~

ACCEPTABLE OPERATION -

'4 m

c m

I Q 13 -

9 .

~

c 12 ' ' ' ' '

0 2500 5000 7500 10000 12500 15000 CYCLE AVERAGE BURNUP (hnVD/MTU) l l

Allowable Peak Linear Heat Rate - Omaha Public Power Distnct Figure vs. Burnuo Fort Calhoun Station-Unit No. I 2-5 Amendment No. 8,20,32.J7,II, 92, 109, IJ7, 126

k, u =~

~'

k. $ # -

[m, .

i10 i i i , ,

100 - -

90 - - i b 80 -

0.08, 80) (0.10,80)

(-0.2. 75. 0) (0. 2, 75. 0)

B 70 - -

E a

d 60 - -

es a

50 -

5 x AD -

E E 30 -

8 l

20 -

l 10 -

0 ' ' '

l -0.3 -0.2 -0.1 0.0 0.1 0.2 0.3 AXIAL SHAPE INDEX Y I

LimitingCanditionforOperationfor OmahaPublicP0werDistrict Figure Exc0re Monitoring of UiB FortCalhounStation-UnitNo,i 2-6 -

Amendment No. 8,20,32,f 3,f 7 JD,78,92,JO9,Il7

' ' ^

-Q Ac. g s _,

(' g Q 110 , , , , ,

100 -

(-0. 057,100)j (0.098,100) -

/

90 -

/

/ -

E w /

g 80 -

/ _

(-0. 2. 75. 0) (0.2.75.0 8 70 -

C; 60 -

c5

!! 50 -

x W 40 -

E g 30 -

8 20 -

10- -

i 0 I t

-0.3 -0.2 -0.1 0.0 0.1 0.2 0.3 AXIAL SHAPE INDEX Y 1 l.imitingCanditionforOperation OmahaPublicP0werDistrict Figure.

f0rDNBHonitoring FortCalh0unStation-UnitNO.1 2-7 Amendment No. $,20.32 A7,79,77,121 .

.\ jQ)],]>bJ ' l+-.;gets :- (~'oLR 110 , -

r , , , , , ,

i Fa LIMIT ?$L: HIT 100 -

=

-w 2

C

' 90 ~

\ ~

5 a \'N N

E 80 -

u.

=

- (1.36, 75) (1.92. 75)

=

W i. 1 l

M 70 - I l

w I i I I i

l i 60 -

1 I I t-t- i 0 ' ' ' ' '

1.65. -1.70 1.75 1.80 - 1.85 1.90 1.95 2.00 2.05 Frn AND Frn Fh,%anuCorePcWer Figure OmanaPublicPowerDistr:::

Limitations FortCalhounStation-UnitHa.1 2-9

- Amendment No. E,20,32,43,#7,/0,II, 92,109, JJ7,126

. . . _-. _ _ _ . _ _ _ _m .

i 2.0 '::4!T:::3 C0!TDIT:0"G FOR OPE *ATION 2.10 Reacter Ccre (Continued)

\

2.10.3 In-Ocre !nstrumentatien Applies to the operability and alarm values of the rhedium detector in-core instruments system.

Objective To specify the functional requirements on the use of the rhodium in-ccre instrumentation system for (1) the recalibration of the ex-ecre detecter inputs to the axial power distributien trip calculaters and (2) scnitcring of kv/ft and power distributien.

Specificatien (1) A minimum of four in-core locaticns at each detector level (cr a total of 16 detectors) vith at least ene location in the center seven revs of fuel assemblies and at least one location outside the center seven r0vs of fuel assemblies shall be cpers"le during recalibration cf the ex-ccre detectors.

(2) The in-ccre detecter system shall be operable with:

(a) At least 75% of all in-ccre detector strings, and

,f] _

(b) A nini=Im: of_tvo in core detector strings per full axial length quadrant whenever the in-core system is used to monitor F xyT , 7 3* ,

the radial power distribution - and the peak linear heat rate. An operable in-core detector string shall consist of three or more operabic rhodium detectors. '41th the in-core detector system inoperable, do not use the system for the-above applicable monitoring functions.

(3) If the recalibration of the ex-core detectors has not been acecmplished within the previous 30 equivalent full-power days , .ceduce-the-exiel-pover-distributien .m.n=inc-D OC _;pddmits-and-tripnetpointe, PisuresMATT-and-l+,-br orG3-A61 untr. If the recalibration of the ex-core de-tectors has not been accomplished within the previous 200 equivalent full power days, the power shall be limited to less than_that corresponding to 75% of the peak linear heat rate permitted by Specification 2.10.k. (1) .

(k) After each fuel loading,-the incere detector system shall be operable with at least 75% of the incore dete; tor strings operable and a minimum of two quadrant syn =etric incore detector string locations cer core quadrant for the- initial measurement of the linear heat rate, FRm ,

s , Fxy' and the acicuthal ~ cover tilt.

')

A eniment No. .' , 4d, LT 2-5h

ADD then:

(a) reduce the axial power distribution monitoring trip setpoints (Figure 1-2) by 0,03 ASI units; and (b) reduce the axial power distribution monitoring Limiting Condition for Operations (LC0 for Excore Monitoring of LHR and LC0 for DNB Honitoring figures provided in the COLR) by 0.03 ASI units.

9. . O 7..,....,.._.a m ,,-.u -. _m.

- ,,4]I . w- . , .:.:.. .m. . . u .n . -w m.T 2.13 Reset:r C:re iC:ntinued!

/, 2.10.3 In-Cere :nstrumentatien

(

(a) An operable incere detector string shall censi:t of three operable rhodium detect 0rs.

(b) A quadrant sy--a+ '- ncere detecter string locat10n i

shall censist of a location with a symmetric counter-part in any other quadrant.

(c) The initial measurement of the linear heat rate, F 3,FT of the'y,,I firstand full asinuthal

Ore power pcVer tilt shall consist distribution calcu-laticn based en incere detector signals mad < st a power level greater than h0 percent of rated power following each fuel loading.

4 If an initial ceasurement of the linear heat rate, 7 3T, F..,-

s and asi=uthal . cover tilt cannot be made with an cu.er-abie incere detector system as defined above, reacter pcVer shall be restricted to less than 75 percent of the peak allevable heat rate.

Basis The in-ccre instrument system is used to conitor core perfor ance and to insure :peration within the limits used as initial condi-f~*- tiens f:r the safety analysis in three ways:

op m m

-(1) to verify thst the radial punk:.ng facters (Fxy* and ?g-)

are ;ess than the limits .apeoCicd in S"ecificatiens 2.10. ,(2) and 2.10.k( 3), frevded m 4he GM (2) to actuate alarms set on each individual instrument to insure operaticn within specified kv/ft limits of igur:e-

-G.-h -snd Sw Allcm.bble bk lowAr Nht 6de W2 Burauf are providecA m +he CCL%covl (3) to metermine the axial shape index for periodic verifi-cati:n of the calibration of the ex-core detecter system.

The specifi:ation requires a minimum number of detectors and pr per distribution to perform these functicns. In-ccre rhodium detect rs in acnjunction with analytical ecmputer codes calculate pcVer listributione from which Fxy and F3 are determined. Alarm

' limits are set en each in-core instru=ent to insure operation within specified kv/ft 1;.mits.

f _. 0211bratica of the ex-core detector input to the AFD calculator L is required to eliminate ASI uncertainties _ due te instrument drift and axially nenunif:rn detector exposure. If the re-

alibrati:n is not performed in the period specified, the pre-scribed steps vill assure safe operation of the reactor.

Peferences l

l ill Ivaluation of Uncertainty in the :iuclear F rm Facter Measured by Self-?cvered Fined In-Core Detectcr Systems -

CI:I?D-153, August, lo7L.

( a*"-art "o. 14 , 72 , L- ' _ 2-5L _

2. 0 LIMITING CONDITIONS FOR OPERATf0N 2.10 Reactor Core (Continued) 2.10.4 Power Distribution Limits Applicability Applies to power operation conditions.

Ob.iective To ensure that peak linear heat rates, DNB margins, and radial peaking factors are maintained within acceptable l_imits during power operation.

Specification of ihe All*O' M"E L"#

Linear Heat Rate 1

ggs m. L< >m{3 EpuY pM A'd '" f (1) ggg;g ,

The linear heat rate shall not' exceed th.4 limits -shovn-on-Figure-2-5 when the following factors are appropriately

. included:

1. Flux peaking augmentation factors are shown in Figure 2-8.

2. A measurement-calculational uncertainty factor of 1.062, y -

3. An engineering uncertainty factor of 1.03,

4. A linear heat rate uncertainty factor of 1.002 due to axial fuel densification and thermal expansion, and

5. A power measurement uncertainty factor of 1.02.

The linear heat rate shall be monitored by the incore l detector system in accordance with' specifications 2.10.4(1)(a) or 2.10.4(1)(b), or maintain the Axial Shape Index,- Y,, within the limits of-Figure-2-6-irt-Oe LmiTn .c accordance-wi th- speci fi ca t i on_2J0.4(1)(-c) . 2 (kgmmler Op(er.de a) When Ov-theheere linear%rrev%y heat rate is ofcont Prcutded LH nuously in %d OM -

monitored by the incere detectors, and the. linear heat rate is. exceeding its-

~

limits as indicated by four or more valid coincident inc, ore

-detector alarms, either:

(i) Restore the linear heat rate to within its limits within one hour, or (iil Be in at least hot standby within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

l

~} 2-56 Amendment No. 5,20,24,32,

3,47,77,117

[. 2.0 LIMITING CONDITIONS FOR OPERATION

~

2.10 Reactor Corg (Continued) 2.10.4 Power Distribution limits (Continued)

(b) If while operating under the provisions of part (a), the plant computer incore detector alarms become inoperable, operation may be continued for seven days from the date of the last valid core power distribution without reducing power provided each of the following conditions is satisfied:

(i) A core power distribution was obtained utilizing incore detectors within 7 days prior to the incore detector alarm outage and the measured peak linear heat rate was no greater than 90% of the value allowed by (1) above.

(ii) The Axial Shape Index as measured by excore detectors remains within i .05 of the value obtained at the time of the last measured incore power distribution.

' (iii) Power is not increased nor has it been increased since the time of the last incore power distribution.

ih- LmoL. C(ad ib re- w(%ia fer Cper nT]wn Oy(c) When the linear heat rate is continuou ly monitored by the g h ibnTWmg excore detectors, withdraw the full le long term insertion limits of Specification 2.10.2.7 and 9th CEA's beyond the

(

b b-prew3td "7'"sh*.

i , wiith the limits of Mmaintain igure-2-6.theIfAxial Shape the linear Index, heat ra Y;te is exceeding its limits gg -

as determined by the Axial Shape Index, Y , being outside the 1imits of _ Figure 46r.whete_.100_percen;Lof_the allowable power-repres ts-the-aaximum-power-allowed by-the-foM owing

-expression:

-We Lookr;" Cond M' N fer CPMk -14 r4-r-H- !

@ ve I b & n v. j c.p gq h a ce W re-p,wicted )W D -1. L45-the-maximun-allowable Unear-heat-ratens Cod ~-

dete mined-from-Figure-t=5n nd-ts-basid on the co re-a verage- burnup- e t-t he-t ime-of-t he-l a tes t--

incore-powor-sap.-

h -M-is-the-maximum-aMowable fraction _of7rated th e rma l-powe r- a s-d e t e rmi ned-by-t he-F HmM Curve of4fgure-2-9-when-monitoring bf_excoce detector &r-N-1-when-monitoring 4w/ft U51ng Cosdi%v incore-detectors.

% 1.imchq

-Oc Opmd e Occ (i) Restore the reactor power and Axial Shape Index, Y ,

{ wye _ Obuta-e} to within the limits of.Eigtge-24 within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ,J or ofll4Rf)n'("Y f ro vycif d i 0M \ 2-57 Amendment No. 5, 29, 32,

---~J f3, il, Ill,126 a

I

l 2.0 LIMITING CONDITIONS FOR OPiRATION

} 2.10 Bractor Core (Continued) 2.1C.4 Power Distribution limits (Continued)

(ii) Be in at lecst hot st Andby within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

(2) Total Intearated Radial Peakino factor I

The calculated value of F defined by F T - Fg (ieT shall be4fmited-to1170R.F9 isdeterm!nedfromaho)wer i distribution map with na non Rtrippable CEA's inserted and with all full length CEA's at or above the Long Term Steady State Insertion Limit for the existing teactor Coolant Pump combination. The azimuthal tilt, T , is ;he measured value of T q F is determined. 9

- wthm 4he limg provaed at the time de CcLR, i,3 With FR I

7 e 21[0with4n grg 4g g, l Q( j i (a) Reduce / power to bring power andIFR within the limits of 549ure+9; withdraw the full length CEA's to or beyond the Long Term S teady State Insertion Limits of Specification 2.10.2(7),

Bi a and fu or lly withdyaw the NTCEA's,N(' AumvTsevor v)we provicled bg dM -the Q Q a.d Cem BeinatleastIotltandby.

(b) m %e Ccd.

Iy'")I Total Planar Radial Peakina Factor (3) u3cthm ye- 1.mJ provicted m 4he Cod .

H The 'ealculated v alue of F defined as F T=F (1+T fjQ .shall be 4 tatted-tu._145*2 F shallbeN5termiNEdfrok)apower p,a distributionmapwithnonon-hippableCEA'sinsertedandwithall KN

$7 1 full length CEA's at or above the Long Term Steady State Insertion Limit for the existing _ Reactor Coolant Pump combination. This M j determination shall be limited to core planes between 15% and 85% of WM full' core heightiinclesive and shall exclude. regions. influence by grid MPii effects. The azimuthal tilt, T ,9 is the measured value of T at 9

l R L' j the time F xy is' determined.

6p 4

!"k; ' 3 With FxYT 1 5 within 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months s: l

' % e timit provtclect m the cco,R.

.e (a) Reduce power to bring power and F x I to within the limits of l'

F49 ure-24, withdraw the full leng{h CEA's to or beyond the Long

,c Term Steady State Insertion Limits of Specification 2.10.2(7),

,f and fully withdraw the NTCEA's, or v-c (b) Be in at least hot standby.

cd be S"C '""M**

g Aede'ETx e

F

~ cm

@R

_ 2-57a Amendment No. E , /J, fi, e ie, ii, n.

JEF, ni,126 l@J F

TM]g

1

. .. I o-2.0 LIMITING CONDITIONS FOR OPERATION-

f- 2.10- _ Reactor Core (Continued)

-( 0 2.10.4 PowerDistributionLimits(Continued)

(4) Azimuthal Power Tilt (Tq)

When operating-above 70% of rated power, -

(a) The azimuthal power tilt (Tq) shall not exceed 0.10 1 t

whenever Mini CECOR/BASSS is operable,_ the CEA's are at or above the Long Term Insertion Limit and Mini !

CFCOR/BASSS is being utilized to monitor F T and FTp . 1 xy (b) The azimuthal power tilt (Tq)- shall not exceed 0.03 whenever the provisions of 2.10.4(4)(a) do NOT allow Hjni F

CECOR/BASSS to be utilized to monitor Q and I tb. be With the indicated

>0.03 but <0.10,azimuthal power correct the powertilttilt determined within two hours or determine within the next.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and at least once per subsequent 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , } hat the total integrated radial. peaking factor, F is within the limit of Specification 2.10.4(2) anb,that the total planar radial peaking factor, F T, is within the limit of 2.10.4(3), or redece power tdyless -than 70% of rated power within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of confirming Tq ->0.03.

(c) With the indicated power tilt'detennined to be >.10,

.[. -

p wer operation may proceed up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months proviHed F and F T do not exceed the power limits offigure-

-2 ,orEEinatleasthotstandbywithin6 hours.

Subsequent operation for-the-purpose of measurement to identify the cause of the tilt is allowable provided-the power level is restricted to 20% of the maximum -

allowable thermal power level for the existing reactor.

coolant pump combination.

T ihe F r -g g cuyj 0% Aw k.mim< +4 a fejart pevidecl ut 4% CCLR, Amendment No. 22, /3, 47, 92 2-57b 1.

2.0 LIMITING COND7TIONS FOR OPERATION 2.10 Reactor Core (Continued)

( 2.10.4 Power Distribution Limits (Continued)

(5) DNBR Margin During Power Operation Above 15% or Rated Power (a) The following DNB related parameters shall be maintained w ithin the limits shown:

-b Lmuh Cowlth

{cr CFn1 (9 i) Cold Leg Temperature 5543*F* C ,

11) Pressurizer Pressure 22075 psia scuethrumgd tii) Reactor Coolant Flow 2197,000 gpm" L)N l'vp m premdn{ iv) Axial Shape Index, Y y 5 Figure-2-7*** (2 )

in the Ccut gg ggg g gg ,g the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce power to less than 15% of rated power within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

Basis -hw Q G %d core T6v AomMb0 Fwk wvM in

Ccd /

The limitation on linear heat rate ensures that'in the event of a LOCA,I the peak temperature of the fuel cladding will not exceed 2200'F.

Either of the two core powar distribution monitoring systems, the Excore Detec-tor Monitoring System, or the Incore Detector Monitoring System, provide adequate monitoring of the core power distribution and are capable of ve{rifying that the ' linear heat rate does not exceed its limit. The Excore Detector

( Monitoring System performs this function by continuously monitoring the axial shape index with the operable quadrant symmetric excore neutron flux detectors i and_yerifying that the axial shape index is maintained within the allowable 91mits of-FTgure-2-6 as adjusted by Specification 2.10.4(1)(c) for the a' lowed linear heat rate of Figur 5; RC Pump configuration, and F T of -Figur 2-9r In conjunction with the u e of the excore monitoring system Ehd in establishing the axial shape index limits, the following assumptions are made: (1) the CEA insertion limits of Specification 2.10.1(6) and long term insertion limits of Specification 2.10.1.(7) are satisfied, (2) the flux peaking augmentation factors are as shown in Figure 2-8, and (3) the total planar radial peaking factor does not exceed the limits of Specification 2.10.4(3).

-ik AtkeoaWe R4 k.wu Had Me A M-7 Mcg ce- pruided i ,s the CcLR.

" Limit not applicable during either a thermal power ramp in excess of 5% e-rated thermal power per minute or a thermal power step of greater than 10%

of rated thermal power.

- This number is an actual limit and corresponds to an indicated flow rate of 202,500 gpm. All other values in this-listing are indicated values and include an allowance for measurement uncertainty (e.g., 543'F, indicated, allows for an actual T of 545'F.

c

1RThe-AXIAL. SHAPE-INDEL Core-power-sha1Lbe ma.intained_within-the-limits .

estaMished-by the Batter Axlal-Shape-Selection Systas-(BASSS)-for-CEA insertion-of-the-lead-bank-of+65% when-BASSS-is7perablir; orMthin the-limits-of-F4gure 2-1 d') whw n %e In e Or cme Trite t Timydee prouded in the Cod Card mm 2-57c Amendment No. 32,43,57,79, Q) pas y o Su p erdLunTu]

u m & c D2 77,92,209,117 yw,wq F um $ pecaed m %e W .

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

2.0 LI"! TING COUC~TIONS FOR OPERATICU 7 2.10 Reacter Core (Continued)

- \)x 2.10.4 Power Distribution Li=its (Continued)

The Incore Detector Monitoring Syste= provides a direct =easure of the peaking factors and the alar =s which have been established for the individual incore detector segments ensure that the peak linear heat rates vill be continuously maintained within the allovable li=its of -F4g ae-2-5 The setpoints for these alar =s include allowances, set in the conserv tive directions, for the factors listed in 2.10.k.(ll, the Stre LA N A Ln d Hed M-

\!S . i%co u f +Q u re pmvMed h3 ihe. CcLR .

Total Planar and Integrated Radial Penking Factors IF.yT X and FoT) and A inuthal Pover Tilt (T q___

)

m The limitations of F and T are provided to ensure that the assu=ptions used in the analysis k,or establishing the Linear Heat Rate and Local Power Density - High LCO's and LSSS setpoints re=ain valid during operation at the various allovable CEA group insertion limits. The limitations of F T R and T q are provided to ensure that the assumptions used in the analysis establishing the DNS Margin LCO, and Thermal Margin /Lov Pressure LSSS setpoints remain valid durine operation at the various allovable CEA group insertion li=its.

It'Fxy T, FR or T exceed q their basic limitations, operation may continue under the additional restrictions i= posed by the action statements since these additional restrictions provide adequate assurance that the assumptions used in establishing the' Linear Heat Rate, Ther=al Margin /Lov Pressure and Local Power Density - High LCO's and LSSS setpoints remain valid. An animuthal power tilt >0.10 is not expected and if it should occur, subsequent operation vould-be restricted to caly those cperations required to identify the cause of this unexpected tilt.

m -

The value of T, that =uct be used in the equation Fxy - =Fxy(1 + Tq) and Fa - =

FR (1 + Tq ) is she censured tilt.

m -

The surveillance require =ents for verifying that Fxy , Fa', and Tq are within their limits provide assurance that the actual

-exceed the assu=ed values. Verifying F xy

and Fp m values m of F.f+

a and .Q do not a

after each fuel .oading prior-to exceeding TOT. of rated power provides additional assurance that the core was properly loaded.

DUSR Margin Durinc Power Operation Above 157 of Rated Power The selection of li=iting safety systen settings and reactor operating limits is such that:

1. No specified acceptable fuel design limits vill be e: seded as a result of the design basis anticipated operational occurren as, and

2. The consequences of the design basic postulated accidents will be no more severe than the predicted acceptable consequences of the accident analysis i in Secticn lb.

-FCET W 2-57d Amend =ent No. 32, 57

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

2.0 LT!!ITING C0!DITIC!!S FOR OPERATIO" fh, 4 ho ih b)13 g

[?,T f> [;Y[ q,yl Cc re Qaw('

2.10 Reactos Core (Continued) Figa m pi^ m ini "T

M 2.10.h Pover Distribution Li tita (Conti r.ued )

i In order for these objectives to be met, ,the reactor cust be operated consistent with the operating limits specified for gargin to DI3.

The para:eter limits given in (5) and -Fig lure-3-9-along with the parameter linits on quadrant tilt and control element assembly position (Figure- )

provide a high degree of assurance that the D!!B overpower margin vill be maintained during steady state operation. {pQ g , -g I,w, mi, Li,@

The actions specified assure that the reactort NabtrouhXeveect

^

In 4ht* COLN ht to a safe condition.

The reactor coolant pu=p differential pressure monitoring system that vill be used to measure flow provides an accurate method of determining reactor coolant flow.

The procedure for determining individual pump and reactor vessel flow vill be as follows:

1. Obtain a pu=p casing AP, using the precision resistor and high accuracy digital voltmeter and converting to pressure.

2. Obtain cold leg temperature and pressurizer pressure.

3. Correct the reading to the curve specific gravity.

k. Obtain pu=p flows fro: individual pu=p casing vs. flow curves.

5 Add the individual punp flows to obtain the best estimate reactor vessel flow.

-POR"'t\LF.CU2- 2-57e i.=endment 'io. 32, 57

l 2.0 LIMITING CONDITIONS FOR OPERATIONS 2.14 Enaineered Safety Features System initiation Instrumentation Settinal (Continued)

(3) Containment Hiah Radiation (Air Monitoring) (Continued)

The setpoints for the isolation function will be calculated in accordance with the ODCM, Each channel is supplied from a separate instrument A.C. bus and each auxiliary relay requires power to operate. On failure of a single A.C. supply, the A and B matrices will assume a one-out-of-two logic.

(4) Low Steam Generator pressure A signal is provided upon sensing a low pressure in a steam generator to close the main steam isolation valves in order to minimize the temperature reduction in the reactor coolant system with resultant loss of water level and possible addition of reactivity. The setting of500psiaincludes(g)22psiuncertaintyandwasthesettingusedin the safety analysis.

Closure of the MSIVs (and the bypass valves, along with main feedwater isolation and bypass valves) is accomplished by the steam generator isolation signal which is a logical coabination of low steam generator pressure or high containment pressure.

As part of the AFW actuation logic, a separate signal is provided to terminate flow to a steam generator upon sensing a low pressure in that steam generator if the other steam generator pressure is greater than the pressure setting. This is done to minimize the temperature reduction in the reactor coolant system in the event of a main steamline break. The setting of 466.7 psia includes a +31.7 psi uncertainty; therefore, a setting of 435 psia was used in the safety analysis.

(5) SIRW Tank Low level Level switches are provided on the SIRW tank to actuate the valves in the safety injection pump suction lines in such a manner so as to switch the water supply from the SIRW tank to the containment sump for a recirculation mode of operation after a period of approximately 24 minutes following s safety injection signal. The switchover point of 16 inches &b;ve tank bottom is set to prevent the pumps from running dry during the 10 seconds required to stroke the valves and to hold in reserve approximately 28,000 gallons ofat-4eaC900-ppm borated l

-wa ter. The FSAR loss of coolant accident analysis W assumed the recirculation started when the minimum usaale volume of 283,000 gallons had been pumped from the tank.

_ ptg gp at lemst' ihe cebehng * "C#'

2-62 Amendment No. E,32,f3,EE,EE,J03,J0E 133

4 3.0 SURVEILLANCE REQUIREMENTS 3.10 Reactor Core Parameters (Continued)

(2) Moderator. Temperature Coefficient The MTC shall be determined at the following frequencies and power conditions during each fuel cycle:

1. Prior to initial operation above 5% of rated power, after each fuel loading.

2. At any power level within 500 MWD /T of initial operation after each refueling.

3. At any power level within i 14 EFPD of reaching a l rated power equilibrium boron concentration of 300 ppm.

(3) Regulating CEA Insertion Limits

a. The position of each regulating CEA group shall be detemined to be above the Transient insertion Limits at least once per shift.

b. The accumulated times during which the regulating CEA groups are inserted beyond the Steady State Insertion Limits but above the Transient Insertion Limits shall be detemined once per day. .

f (4) Linear Heat Rate Monitoring Systems

a. The incore detector monitoring system may be used for monitoring the core power distribution provided that at least once per 31

_ days of accumulated power operation the incore detector alarm;

& Lum %vj Om&Tw)9enerated satisfy the requirements by the plant of the computer are map.

core distribution verified to be valid and d,er Ogudwe {cc The excore detector monitoring system may be used for monitoring Enm e (YDUWnb1.

,, - the core power distribution by:

or lM R t w ee prevded in 'th 1. Verifying at least once per 31 days of accumulated power 7 operation that the axial shape index, YI , monitoring gg g I limit setpoints are maintained within the allowable limits

~ ~" of.Ei,9ureM as adjusted by Specification 2.10.4(1).

.-- w (5) Total Integrated and Total Planar Radial Peaking Factors (FaT and FxyT )

~

FRT and FxyT shall be detemined to be within the limits of Specification 2.10.4 at the following intervals:

a. After each refueling and prior to operation above 70 percent "

of rated power. .

b. At least once per 31 EFPD's of accumulated power operation.

Amendment No. 32,109 3-63a

l l

4.0 DESIGN FEATURES I

4.4 Fuel Storace 4.4.1 New Fuel Storace The new unirradiated fuel bundles will normally be stored in the dry new fuel storage rack with an effective multiplication factor of less than 0.9. The open grating floor below the rack and the covers above the racks, along with generous provision for drainage, precludes flooding of the new fuel storage rack.

New fuel may also be stored in shipping containers or in the spent fuel pool racks which have a maximum effective multiplication factor of 0.95 with Fort Calhoun Type C fuel and unborated water.

The new fuel storage racks are designed as a Class I structure.

4.4.2 Soent Fuel Storace Irradiated fuel bundles will be stored prior to off-site shipmcnt in the stainless steel lined spent fuel pool. The spent fuel pool is normally filled with borated water with a concentration of at least

-1900-ppm ihc rekeMc; bore n ecmcenttuTcen . l The spent fuel racks are designed as a Class I structure.

Normally the spent fuel pool cooling system will maintain the bulk water temperature of the pool below 120*F. Under other conditions of fuel discharge, the fuel pool water temperature is maintained below 140'F.

The spent funi r3cks are designed and will be maintained such that the calculated effective multiplication factor is no greater than 0.95 (including all known uncertainties) assuming the pool is flooded with unborated water. The racks are divided into 2 regions. Region 1 racks are surrounded by Boraflex; Region 2 racks have no poison. Acceptance criteria for fuel storage in Regions 1 and 2 are delineated in Section 2.8 of these Technical Specifications.

4-4 Amendment No. JJ,f7,73,Jp) ,133

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

5.0 ADPINISTRATIVE CONTR0!.S Responsibilities 5.5.1.6 The Plant Review Comittee shall be responsible for:

a. Review of 1) all procedures required by Specification 5.8 and changes thereto, 2) any other proposed procedures or changes thereto as determined by the Manager - Fort Calhoun Station to affect nuclear safety,

b. Review of all proposed tests and experiments that affect nuclear safety, c ., Review of all proposed changes to the Technical Specifications.

d. RM o & WI Fapud chmes % 4N Cne c p .m tu m L or.,Ts Rep +rt g/ Revirw of all proposed changhs or redifications to plaht systems or equiptrent that af fect nuclear sa'ety.

.p p'. Investigation of all violations of the Technical Specifications and shall prepare and forvard a report <.overino evaluation and recomendations to prevent recurrence to the Division Manacer -

Nuclear Operations and to the Chaiman of the Safety Audit and l Review Comittee, ej /'. Review of facility operations to detect potential safety ha;ards.

ijg. Performance of special r!. views and investigations and reports thereon as requested by the Chairman of the Safety Audit and Review Comittee.

lfh'. Review of the Site Security Plan and implementing procedures and shall submit recomended changes to the Chaiman of the Safety Audit and Review Comittee.

j f'.' Review of the Site Emergency Plan and implementing procedures and d shall subr'it recomended chances to the Chairman of the Safety iudit and Review Committee.

p icview of all Re;'ortable Events.

Authority 5.5.1.7 The Plant Review Comittee shall:

a. Recomend in writing to the Manager - Fort Calhoun Station approval or disapproval of items considered under 5.5.1.6(a) through (g) above.

.e 5-4 Amendment No. 9 Is, M , 99,115

i

-l 5.0 ADMINISTRATIVE CONTROLS ,.p l

/

5.5.1.7.

b. Render determinationt in writing with regard to/w !

each item considered "mier 5.5.1.6(a) through above Jb)jether or not constitutes en unreviewed safety question,

c. Provide immediate written notification to the Division Manager Nuclear Operations and the Safety Audit . 1 Review Committee of disagreement between the Plant Review Committee and the

-Manager Fort Calhoun Station; however, the Manager Fort Calhoun Station shall have responsibility for resolution of such disagreements pursuant to 5.1.1 above.

Records 5.5.1.8 The Plant Review Committee shall maintain written minutes of each meeting and copies shall be provided to the Division Manager -

Nuclear Operations and Chairman of the Safety Audit and Review Comini t t ee .

5.5.2 Safety Audit and Reviaw Committee (SARCl Eunction 5.5.2.1 The Safety Audit and Review Committee f. hall fu,pction to provide the indepenaent review and audit of designated activities in the areas of;

a. nuclear power plant operation

b. nuclear engineering

c. chemistry anc raciochemistry

d. metallurgy

e. instrumentation and control

f. raciological safety

g. mechanical and electrical engineering h .. quality assurance Comoosition i

L 1.

5.5.2.2 The Safety Audit ana Review Committee shall be composed of:

[ Chairman: Division Manager - Nuclear Services i Member: Senior Vice President l Member: Division Manager - Nuclear Operations Member: Division Manager - Production Engineering Member: Manager Fort Calhoun Station n Member: Manager - Radiological Services

.Memoer: Qualified Consultants as Reauirea ana as Determinea ,

by SARC Chairman 5-5 amencment No. 26.93.39.101 !!3.119. 22

ADD ~

5.9.5 . Core Operating Limits Report

a. Core Operating Limits shall be established and documented in the Core Operating Limits Report before cach reload cycle or any remaining part of a reload cycle, b.- The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC as follows:

1. OPPD NA-8301-P-A, Rev. 3 " Reload Core Analysis Methodology Overview" dated April 1988.

2. OPPD NA-836. P-A, Rev. 2 "Neutronics Design Methods and Verincation" dated April 1988.

3. OPPD-NA-8303-P-A, Rev. 2 " Transient and Accident Methods and Verincation" dated April 1988.

c. The core operating limits shall be determined so that all applicable 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 Dok with copies to the Region IV Administrator and Senior Resident Inspector.

1 l :'

5-17a Amendment .No.

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

IMLE OF CONIEMfS (Continued)

Pdat 4.3 Nuclear Steam Supply System (NS$5) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-3 4.3.1 Reactor Coolant $yStem . . . . . . . . . . . . . . . ...................43 4.3.2 Reactor Core and Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 3 4.3.3 Emergency Core Cooling . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-3 4.4 Fael Storage ..................... . . . . . . ...................44 4.4.1 New Fuel Storage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-4 4.4.2 Spent Fuel Storage ............ . . . . . ..... . . . . . . . . . . . . . 4 -4 4.5 Seinmb Design for Class 1 Systems ................ 4 ...............45 5.0 ADMINISTRATlVE CONTRO> 6 51 5.1 Responsibili t y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.2 Organiration ................................ .... ... ...51 5.3 Facility $taff Qualifications .. ..... .... .. ........... ..... ..... ... 5.la 5.4 Tnd n i n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 - 3 5.5 Review od Audit ............................................5.1 5.5.1 Plant Review Committne (PRC) . . . . . . . . . . ...................53 5.5.2 Safety Audit and Revitw Committeo (SARC) . . . . ....... . . . . . . . . . 5 -5 5.5.3 Fire Protection inspection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 Ba 5.6 Reportable Event Action . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-9 5.7 Safety Limit Violation . . . . . . . . .. . . . . . . . .. . . . . . . . . .. . . . . . ,. . . . . . 5-9 5.8 Procalures ................................................59 5.9 Reporting Requirements ....................................... 5 10 5.9.1 Routine Reports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 10 5.9.2 R eportable Es ents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 12 5.9.3 Special Reports .......................................515 5.9.4 Unique Reparting Requiremer.2s . . . . . . . . . . . . . . ..............,5-15 Core Operating Limitt laport . . . . . . . . . . . . .

5.9.5

............... 5 17a l 5.10 Records Retention . . .. ... ......... .... .... .. . .. .......... ... 5-18 5.11 Radiation Protection Ptqtam . . . . . . . . . . . . . . . . ...................$.19 5.12 DELETED 5.13 Secondary Water Chemistry . . . . . . . . . . . . . . . . . . . . . . ..............5-20 5.14 Systems Integrity . . . . .......................... . . . . . . . . . . . . 5 -21 5.15 Post Accident Radiological Sampling and Monitoring . . . . . . . . . . . . . . . . . . . . . 5-21 6.0 INTERIM SPECIAL TECilNICAL SPECIFICATIONS 61 6.1 Limits on Reactor Coolant Pump Operation . . . . , . . . .........,....,....61 6.2 Use of a Spent Fuel Shipping Cask . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 1 6.3 Auxinary Fen! water Automatic initiation Setpomt . . . . . .... . ......... . .... 6-1 6.4 Operation With Less Than 75 9( of incoce Detector Strings Operable ......... ..................,........... . 6-1 iii Amendment No. P,34,43,64,65,67

'*1,80,06,93,'

IECllNICAIJil'EClllCAllONS - FIG 1!lifS IAllLILOf CONTEN"?,'

1 1%GE WillCll i hd S Ill3CHil110ti iIGURIR01LOES 1 4-1 TMLP Safety Limits 4 Pump Operations . . . . . . . . . . . . . . . . . . . . 1-3 1-2 Axial Power Distribution LSSS for 4 Pump Operations . . . . . . . . . . . 1-3 2-1A ItCS Press Temp Limits IIcatup .........................26 2 111 RCS Press-Temp Limits Cooldown ,,.. .. ..... .. .... ...... 2-6 23 Predictc/ Radiation Induced NDTI' Shift . . . . . . . . . . . . . . . . . . . . 2 6 2 11 MIN llAST Level vs Stored IIAST Concentration ............. 2-19 2-12 Iloric Acid Solubility in Water . . . . . . . . . . . . . . . . . . . . . . . . 2-19 2-10 Si x nt Fuel Pool Region 2 Storage Criteria . . . . . . . . . . . . . . . . . . 2-38 28 Flux Peaking Augmentation Factors . . . - . . . . . . . . . . . . . . . . . . 2-53 l

. g - g +

1 l -- - _ _ , . . - . --

DEFINITIONS REACTOILOl' ERA 11'iG_CONIllIl0NS (Continued)

Cold Shutds!LCondition (Operating Mode 4)

The reactor coolant Tuu is less than 210'F and the reactor coolant is at shutdown baron concentration.

Rduding shutdown Condillen (Operating Mode 5)

The reactor coolant is at refueling boron concentration and Tuu is let.s than 210"F.

Eduding_ Operation Any operation inw..ving the shuffling, removal, or replacement of nuclear fuel, CEA's, or startup sources.

' llc _Btfacf TJoron Concentralinn A reactor coolant boron concertration of at least that specined in the Core Operating Limits Report which corresponds to a shutdown margin of not less than 5% with all CEA's withdrawn.

I ShuldemLiloron Concentration The boron concentration required to make the reactor suberitical by the amount defined in

\

Section 2.10. l _

Refueling Outage or Refuding3hutdomi A plant outage or shutdown to perform refueling operations upon reachin;; the phnned fuel depletion for a specific core.

Plant _ON Ialing Cycle The time period from a Refueling Shutdown to the next Refueling Shutdown.

2 Amendment No. 24,32,41,43,103,133 l

____j

DEFINIT 10ES Arimuthat Power Tilt - T, Azimuthal Power Tilt shall be the maximum difference bet'veen the imwer generated in any core quadrant (upper or lower) and the average [wwer of all quadrants in that axial half (upper or ,

lower) of the core divided by the average power of all quadrants in that axial half (up;w or !

lower) of the core, Untedded PlancLRadialfcaking FacioL-J,3 The Unrodded Planar Radial Peaking Factor is the maximum ratio of the peak to average power density of the individual fuel rods in any of the unrodded horizontal plancs, excluding azimuthal tilt, T,. The maximum F,y limit is provided in the Core Operating Limits Report. l Unrodded_lategrated_ Radial Peaking Facier_-la The Unrodded Integrated Rr. dial Peaking Factor is the ratio of the peak pin power to the average pin power in an unrodded core, excluding azimuthal tilt, T,. The maximum F, limit is provided l in the Core Operating Limits Report, Eiic_ Suppression WaleLSyllen1 The fire suppression water system consists of fire pumps and distribution piping with associated sectionalizing control or isolation valves Such valves include yard hydrant curb valves, and the first valve thead of the water flow alarm device on each sprinkler, hose standpip or spray system riser.

Erocess Control Program (PCP)

A manual or so of operating procedures detailing the program of sampling, analysis, and evaluation.

Dmcfaulyalent 1-131 That concentration of 1 131 ( Cl/gm) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131,1-132,1-133,1-134 and 1-135 ac'aaDy present, in other words, 7 Amendment No. 32,38,67,86 i

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

DEFINITIONS Dose Ikl uivalent 1-131 (pCilgm) = pCi/gm of I 131

+ 0.0361 x pCi/gm of 1132

+ 0.270 x pCi/gm ofI 133

+ 0.0169 x pCi/gm of I 134

+ 0.0838 x Ci/gm of 1-135 S - Average Disintegration Enercy S is the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration, in MEV, for isotopes, other than iodines, with half lives greater than 15 minutes making up at least 95% of the total non-iodine radioactivity in the coolant.

Offsite Dose Calculation Manual (ODChD A manual containing the methodology and parameters to be used in the: 1) calculation of doses in the unrestricted area due to radioactive liquid and gaseous ef0uents,2) calculation of liquid and gaseous efnuent monitoring instrument:. tion setpoints, and 3) specific details pertinent to the radiological environmental monitoring program.

Eurge-Purging A means for the removal and replacement of gases within the containment building.

YcnliDI A means for the reduction of pressure greater than atmospheric within the containment structure.

Core Operating Limits Report (COLR)

The Corc Operating Limits Repon (COLR) is a Fort Calhoun Station Unit No. I specific document that provides core operating limits for the current operating cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Section 5.9.5. Plant operation within these operating limits is addressed in the individual specifications.

References (1) USAR, Section 7.2 (2) USAR, Section 7.3 8 Amendment No. 67,86

l 1.0 SAEED' LIS11TS AND LISilTING_SMETY SYSTEN1 SETTINGS 1.1 Safety Limht _ Reactor Core (continued) would cause DNil at a particular core location to the actual heat flux at that location, is indicative of the margin to DNil. The minimum value of the DNIlR during steady state operation, normal operational transients, and anticipated transients is limited to 1.18. A DNilR of 1.18 corresponds to a 95% probability at a 95% con 0dence level that DNil will not occur, which is considered an appropriate margin to DNil for all operating conditions.*

The curves of Figure 1-1 represent the loci of points or reactor thermal power 0 ther neutron Dux instruments or AT instruments), reactor coolant system pressure, and cold leg temperature for which the DNilR is 1.18. The area of safe operation is below these lines.

The reactor core safety limits are based on radial peaks limited by the CEA insertion limits in Section 2.10 and axial shapes within the axial power distribution trip limits in Figure 1-2 and a total unrodded planar radial peak (F,/) as specified in the COLR. The Thermal hf argin/lmw Pressure trip requirements shall be within the limits providcJ in the COLR. The Thermal hiargin/ Low Pressure trip is based on an unrodded integrated total radial peak (F/) that is provided in the COLR.

Flow maldistribution effects for operation under less than full reactor coolant flow have been evaluatcd via model test.* The flow model data established the maldistribution factors and hot channel inlet temperature for the thermal analyses that were used to establish the safe operatinE envelopes presented in Figure 1-1. The reactor protective system is designed to prevent any anticipated combination of transient conditions for reactor coolant system temperature, pressure, and thermal power level that would result in a DNilR of less than 1.18.m BIfcIcacts (1) USAR, Section 3.6.6 (2) USAR, Section 1.4.6 1-2 Amendment No. 8,12,4M7 70,77,92,417,126 !

1.0 SAFETY _ljAllTS__AND LIMITING SAIID' SYSTEM SE'ITINGS 1.3 Limiline Safety SystCHLStiling3JIBflor Protectlye Sysics (continued)

(3) liigh Pre 11urizer Presstlic - A reactor trip for high pressurizer pressure is provided in conjunction with the reactor and steam system safety valves to prevent reactor coolant system overpressure (Specincation 2.1.6). In the event of loss of load without reactor trip, the temperature and pressure of the reactor coolant system would increase due to the reduction in the heat removed from the coolant via the steam generators. The power-operatul relief valves are set to operate .oncurrently with the high pressurizer pressure reactor trip. This setting is 109 psi below the nominal safety valve setting (2500 psia) to avoid unnecessary operacion of the safety valves. This setting is consistent with the trip point assumed in the accident analysis.("

(4) Thermal Margin /imw Pressurg_ Trip - The thermal margin / low pressure trip is provided to prevent operation when the DNilR is less than 1.18, including allowance for measurement crror. The thermal and hydraulic limits shown in the Thermal Margin /imw Pressurc 4 Pump Opciation Figure, contained in the COLR, denne the limiting values of reactor coolant pressure, reactor inlet temperature, axial shape index, and reactor power level which ensure that the thermal criterla* are not exceeded. The low set point of a 1750 psia trips the reactor in the unlikely event of a loss-of-coolant accident. The thermal margin / low pressure trip set points shall be set according to the equation given in the COLR for the Thermal Margin / low Pressure Limit.

4 e

1-8 Amendment No. 8,20,32 47,70,77,92 w g. .. . ..-, _. _-r ,y - , _ , y

TAllLE 1 1 R11Llh11 TING SAFETY SYSTEM SEHINGS Nnu Reac19tldo Idvltinv11ts I liigh Power 1.evel (A) 1107.0% of Rated Power 4-Pump Operation 2 lxw Reactor Coolant Flow (ll)(F) 4 Pump Operation A95% of 4 Pump Flow 3 low Steam Generator Water l_cvel 31.2% of Scale (Top of feedwater ring; 4'10" below normal water level) 4 Low Steam Generator Pressure (C) 1500 psia 5 Iligh Pressuriter Pressure 12400 psia 6 Thermal Margin /lmw Pressure (ll)(F) 1750 psia to 2400 psia (depending on the reactor coolant temperature as shown in the Thermal Margin / Low Pressure 4 Pump Operation Figure provided in the COLR) 7 High Containment Pressure (D) 15 psig 8 Axial Power Distribution (E) (Figure 1-2) 9 Steam Generator Differential Pressure 1135 psid 1-10 Amendment No. 7,32,47,77,92

2.0 LIMiTlNG_ COED 1110ES_EOJt OPIMATION 2.3 IlmcIgency Core Cooling SyMem Applicability Applies to the operating status of the emergency core cooling system.

Oldeclire To assure operability of equipment required to remove decay heat from the core.

SpecificallRns (1) ldiniinum Requiremtats The reactor shall not be made critical unless all of the following conditions are met:

5

a. The SIRW tank contains not less than 283,000 gallons of water with a boron concentration of at least the refueling boron concentration at a l temperature not less than 507,

b. One means of temperature indication (local) of the SIRW tank is operable,

c. All four safety injection tanks are operable and pressurired to at least 240 psig vcith a tank liquid of at least 116.2 inches (67%) and a maximum level of 128.1 inches (74%) with refueling boron concentration.

l

d. One level and one pressure instrument is operable on cach safety injection tank.

i

c. One low-pressure safety injection pump is operable on each bus,

f. One high-pressure safety injection pump is operable on each bus.

g. Both shutdown heat exchangers and three of four component cooling heat exchangers arc operable.

l l h. Piping and valves shall be operable to provide two flow paths from the l SIRW tank to the reactor coolant system.

i. All valves, piping and interlocks associated with the above components and required to function durind accident conditions are operable.

HCV-2914,2934, 2974, and 2954 shall have power removed from the motor operators by locking open the circuit breakers in the power supply

! lines to the valve motor operators. FCV-326 shall be locked open.

L 2 20 Amendment No. 11,1-7,32,43,403, 117,419:133 1

2.0 IJAILTING CONDIllONS FOR OPERATIO.N 2.3 Etnergency Core Cooling System (Contim,4) l (3) helecdon Against low Temocrature Overpic11urization ,

The following limiting conditions shall be applied during scheduled heatups and cooldowns. Disabling of the llPSI pumps need not be required if the reactor vessel head, a pressurlier safety valve, or a PORV is removed. l 1

Whenever the reactor coolant system cold leg temperature is below 32WF, at least j one (1) IIPSI pump shall be disabled. j Whenever the reactor coolant system cold leg temperature is below 312"F, at least two (2) llPSI pumps shall be disabled.

Whenever the reactor coolant system cold leg temperature is below 27P'F, all threc (3) IIPSI pumps shall be disabled, in the event that no charging pumps are operable, a single llPSI pump may be made operable and utilized for beric acid injection to the core.

Basis The normal procedure for starting the reactor is to first heat the reactor coolant to near operating temperature by running the reactor coolant pumps. The reactor is then made critical by withdrawing CEA's and diluting boron in the reactor coolant. With this mode of start-up, the energy stored in the reactor coolant during the approach to criticality is subsiantially equal to that during power operation and therefore all engineered safety features and auxiliary cooling systems are required to be fully operable. During low .

power physics tests at low temperatures, there is a negligible amount of stored energy in the reactor coolant; therefore, an accident comparable in severity to the design basis accident is not possible and the engineered safeguards systems are not required.

The SIRW tank contains a minimum of 283,000 gallons of usable water conta'.ning a boron concentration of at least the refueling boron concentration. This is sufDcient boron concentration to provide a shutdown margin of 5%, including allowances for uncertainties, with all control rods withdrawn and a new core at a temperature of 6&F.*

The limits for the safety injection tank pressure and volume assure the required amount of water inJcction during an accident and are based on values used for the accident analyses. The minimum i16.2 inch level corresponds to a volume of 825 ft' and the maximum 128.1 inch level corresponds to a volume of 895.5 ft'. Prior to the time the reactor is brought critical, the valving of the safety injection system must be checked for correct alignment and appropriate valves locked. Since the system is used for shutdown i cooling, the valving will be changed and must be properly aligned prior to start-up of the reactor.

2-22 Amendment No. I1,1-7,39,43,47,64, 74,77,400,403,133

2.0 1,lMITING_ CONDITIONS FOR OfflRT10N 2.8 Rchteling Operations (Continued) incident could occur during the refueling operations that would result in a hazard to public health and safety.(" Whenever changes are not being made in core geometry one flux monitor is sufficient. This permits maintenance of the instrumentation. Continuous monitoring of radiation levels and neutron flux provides immediate indication of an unsafe condition. The shutdown cooling pump is used to maintain a uniform boron concentration.

The shutdown margin as indicated will keep the core scritical even if all CEA's were withdrawn from the core. During refueling operations, the reactor refueling cavity is filled with approximately 250,000 gallons of borated water. The boron concentration of this watcr (of at least the refueling boron concentration) is suf0cient to maintain the l reactor suberitical by more than 5%, including allowance for uncertainties, in the cold condition with all rods withdrawn.m Periodic checks of refueling water boron concentration ensures the proper shutdown margin. Communication requirements allow the control room oper? tor to inform the refueling machine operator of any impending unsafe condition detected from the main control room board indicators during fuel movement, in addition to the above engineered safety features, interlocks are utilized during refueling operations to ensure safe handling. An excess weight interlock is provided on the lifting hoist to prevent movement of more than one fuel assembly at a time. In addition, interlocks on the auxiliary building crane will prevent the trolley from being moved over storage racks containing irradiated fuel, except as necessary for the handling of fuel. The restriction of not moving fuel in the reactor for a period of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after the power has been removed from the core takes advantage of the decay of the short half-life fission products and allows for any failed fuel to purge itself of fission gases, thus reducing the consequences of fuel handling accident.'

The venti!ation air for both the containment and the spent fuel pool area Dows through absohte particulate filters and radiation monitors before discharge at the ventilation discharge duct. In the event the stack discharge should indicate a release in excess of the Ihnits in the technical specifications, the containment ventilation flow paths will be closed automatically and the auxiliary building ventilation How paths will be closed manually. In addition, the exhaust ventilation ductwork from the spent fuel storage area is equipped with a charcoal filter which will be manually put into operation whenever irradiated fuel is being handled ("

Refereness (1) USAR, Section 9.5 (2) USAR, Section 9.5.1.2 i

i l 2-39 Amendment No. 24,75,403d4hl33 L

l l

2.0 LIMITING CONDITIONS FOR OPERATION 2.10 Reaslor Core (Continued) 2.10.2 Reactivity CentroLSy11 ems and Core Physics Parameters Limits (Continued)

Control Element Assemblics (4) Full Length CEA Posillon DuriDg Power Oncration All full length (shutdown and regulating) CEA's shall be operable with each CEA of a given group positioned within 12 inches (actual position) of all other CEA's in its group. If one or more of the CEA's is inoperable or misaligned, determine the cause and comply with one of the following:

a. If one or more full length CEA's are inoperable due to: 1) being immovable as a result of excessive friction or mechanical interference, or

2) known to be untrippable, determine that the shutdown margin requirement of Specification 2.10.2(1)is satisfied within I hour and be in at least hot shuthwn within 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months ,

b. With one full length CEA inoperable due to causes other than addressed in item a. above, and inserted beyond the Long Term Steady State Insertion Limits but within its above specified alignment requirements, power operation may continue for up to 7 EFPD's per occurrence with a total accumulated time of < 14 EFPD per calendar year.

c. With one full length CEA inoperable due to causes other than addressed in item a. above, but within its above specified alignment requirements and either fully withdrawn or above the Long Term Steady State Insertion Limits if in CEA group 4, power operation may continue.

d. With one or more fulllength CEA's misaligned from any other CEA's in its group by more than 12 inches but less than 18 inches (actual position) within one hear either:

(i) Restore the misaligned CEA(s) to within 12 inches (actual position) of any other CEA's in its own group (realignment shall be made while maintaining the allowable CEA sequence and CEA insertion limits of the Power Dependent Insertion Limits Figure provided in the COLR; or (ii) Declase the CEA's inoperable. Power operation may continue provided all of the following conditions are met:

1. The power level shall be reduced to .5;.70% of the maximum allowable power level for the existing Reactor Coolant Pump combination within an additional one hour; if negative reactivity insertion is required to reduce power, boration shall be used.

2-50a Amendment No. 8,20,32

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

2.0 LIMITING CONDITIONS FOR OPER ATION 2.10 Reactor Cors (Continued) 2.10.2 Reactivity CentmLSyllems and Core Physisslarameters Limits (Continued)

2. Within one hour after reducing the power as required by 1., above, the remainder of the CEA's in the group with the inoperable CEA shall be aligned to within 12 inches (actual position) of the inoperable CalA while maintaining the allowable CEA sequence and insertion limits for Power Dependent insertion Limit Figure provided in the COLR, the power level shall be restricted pursuant to the applicable section oi b. or c above during subsequent operation.

or (iii) Be in hot shutdown within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

e. With one fulllength CEA mbaligned from any other CEA's in its group by 18 inches or more (actual position), reduce power to 170% of the maximum allowable power level for the cxisting Reactor Coolant Pump combination within one hour, if negative reactivity insertion is required to reduce power, boration shall be used. Within one hour after reducing power as required above, either:

(i) Restore the CEA(s) to within 12 inches (actual position) of any other CEA's in its group (realignment shall be made while maintaining the allowable CEA sequence and CEA insertion limits of the Power Dependent Insertion Limit Figure in the COLR); or (ii) Declare the CEA(s) inoperable and determine thM the shutdown margin requirement of Specification 2.10.2(1)is y satisfied and align the remainder of the CEA's in the grog

with the inoperable CEA to within 12 inches (actual position) of the inoperable CEA while maintaining the allowable CEA sequence and insertion limits shown ir, the Power Dependent insertion Limit Figure provided in the

,. COLR. The shutdown margin requirement of Specification 2.10.2(1) shall be determined at least once per shift. The l

power level shall be restricted pursuant to the applicable section of b. or c. above during subsequent operation; or l- (iii) Be in hot shutdown within an additional 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

l f. With more than one full length CEA inoperable or misaligned from any other CEA in its group by 18 inches (actual position) er more, be in at least hot shutdown within 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months .

2-50b Amendment No. 32 l

2.0 IJh1111NG_ CONDITIONS FOR_DECHAHON 2.10 ReacicLCOIr (Continued) 2.10.2 Reactivity Control SynemtanLCore_Pbysics Paratnciers 1.imils (Contined)

(5) Nonicippable CEA Position During Power Operalian All non-trippable CEA's (NTCEA) shall be withdrawn to at least 114 inches (actual position), if one or more NTCEA's beconn misaligned from other NTCEA's by more than 12 inches (actual position) citku

a. Restore the NTCEA to within the specified alignment requirements within one hour, or

b. Be in at least hot shutdown within an additional 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

(6) Shutdgwn CEA Insertion _LhniLDEiDB PowcLDDcEt1100 All shutdown CEA's shall be withdrawn to at least 114 inches as a condition for reactor critica!ity, or with one or more shutdown CEA's inserted to more than 114 inches withdrawn, except for surveillance testing, within one hour ither:

a. Withdraw the CEA's to at least 114 inches, or

b. Declare the CEA's inoperable and apply Specification 2.10.2(4).

(7) Regulating CEA inicItion Limits During l{ot Standby and Power Occration The regulating CEA groups shall be positiond within the acceptable operating range for regulating rod position of the Power Dependent insertion Limits Figure prosided in the COLR except during CEA exercises above 114 inches. With all CEA's operable, CEA insertion beyond the long Term insertion Limits are restricted to:

1. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval, 2, 4 EFPD per 30 EFPD interval, and

3. 14 EFPD per calendar year.

a. When the regulating CEA groups are inserted beyond the Transient Insertion Limits within two hours, either:

(i) Restore the regulating CEA groups to above the Transient Insertion Limits, or (ii) Reduce reactor power to the allowed power of the Power Dependent insertion Limit Figure of the COLR which permits continued operation above the Tiansient Insertion Limit using the existing CEA group imsition.

2-50c Amendment No. ^42,109

)

2.0 LIMIIING_COliDHiQSS10E OFimA110N 2.10 lkattor Corc (Continued) 2.10.3 In-Cere Instrutucntalien Applies to the operability and alarm values of the rhoc...; a detector in-core instruments system.

OldctliYC To specify the functional requirements on the use of the rhodium in-core instrumentation system for (1) the recalibration of the ex-core detectot inputs to the axial power di'tribution trip calculators and (2) monitoring of kw/ft and power distribution.

Specintatica (1) A minimum of four in-core locations at each detector level (or a total of 16 detectors) with at least one kicat!on in the center seven rows of fuel assemblics and at least one location outside the center seven rows of fuel assemblics shall be operable during recalibration of the ex-core detectors.

(2) The in core detector system shall be operable with:

(a) At least 75% of all in-core detector strings, and (b) A minimum of two in core detector strings per full axial length quadrant whenever the in-core system is used to monitor F,,T, F,T, the radial power distribution and the peak linear heat rate. An operable in-core detector string t. hall consist of three or more operable rhodium detectors. With the in-core detector system inoperable, do not use the system for the above applicable monitoring functions.

(3) If the recalibration of the ex-core detectors has not been accomplished within the previous 30 equivalent full power days, then:

(a) reduce the axial power distribution monitoring trip setpoints (Figure 1-2) by 0.03 ASI units; and (b) reduce the axial power distribution monitoring 1.imiting Condition for Operations (LCO for Excore Monitoring of LilR and LCO for DNil Monitoring Figures provided in the COLR) by 0.03 ASI units, if the recalibration of the ex core detectors has not been accomplished within the previous 200 equivalent full power days, the power shall be limited to less than that corresponding to 75%

of the peak linear heat rate permitted by Specincation 2.10.4.(1).

(4) After each fuel loading, the incore detector system shall be opemble with at least 73% of the incore detector strings operable and a minimum of two quadrant symmetric incore detector string locations per core quadrant for the initial measurement of the linear heat rate, F,T, F,,T and the azimuthal power tilt.

2 54 Amendment No. 14,32,47 1

2.0 LIMITING CONI)1flONS FOR OPERATION 2.10 Reactor core (Continued) 2.10.3 In-Corc_lmtrumentation '

(a) An operable incore detector string shall consist of three operable rhodium detectors. ,

(b) A quadrant symmetric incore detector string location shall consist of a location with a symmetric counterpart in any other quadrant.

(c) The initial measurement of the linear heat rate, Fat , FxyTand azimuthal power ,

tilt shall consist or the first full core power distribution calculation based on incore detector signals made at a power level greater then 40 percent of rated power following each fuel loading.

If an initial measurement of the linear heat rate, Fni , FxyTand azimuthal power tilt cannot be made with an operable incore detector system x defm' ed above, reactor power shall be restricted to less than 75 percent of the peak allowable heat rate.

1 Mis The in-core instrument system is used to monitor core performance and to insure operation within the limits used as initial conditions for the safety analysis in three ways:

(1) to verify that the radial peaking factors (FnT and Far) are less than the limits of Specifications 2.10.4(2) and 2.10.4(3), provided in the COI.R. l

(" 'o v.r. ate alarms set on each individual instrument to insure operation within specified

1. # limits as provided in the Allowable Peak Linear licat Rate vs. Burnup Figure l pc ided in the COLR, and (3) to determine the axial shape index for periodic verification of the calibration of the ex core detector system.

The specification requires a minimum number of detectors and proper distribution to perform these functions. In-core rhodium detectors in conjunction with analytical computer codes calculate power distribations from which F,, and F, are determined. Alarm limits me set on each in-core instrument to insure operation within specified kw/ft limits.

Calibration of the ex-core detector input to the APD calculator is required to climinate ASI uncertainties due to instrument drift and axially nonuniform detector exposure. If the recalibration is not performed in the period specified, the prescribed steps will assure safe operation of the reactor.

References (1) Evaluation of Uncertainty in the Nuclear Form Factor Measured by Self-Powered Fixed in-Core Detector Systems - CENPD-153, August,1974 2 55 Amendment No. 14,32,47 t

2.0 1,lMITING COND1110NS FOR OPERATION 2,10 Reactor Core (Continued) 2.10.4 Power Distribution Limits Applicability Applies to power operation conditions.

Oldestire ;

To ensure that peak linear heat rates, DNil margins, and radial peaking factors are maintained within acceptable limits during power operation.

SacslRcation (1) 1,lticaLikaLRats The linear heat rate shall not exceed the limits of the Allowable Peak Linear lleat Rate vs. Ilurnup Figure provided in the COLR when the following factors are appropriately !ncluded:

1. Flux peaking augmentation factors are shown in Figure 2-8.

2. A measurement-calculational uncertainty factor of 1.062,

3. An engineering uneettainty factor of 1.03,

4. A linear heat rate uncertainty factor of 1.002 due to axial fuel densificatior and thermal expansion, and

5. A power measurement uncertainty factor of 1.02, The linear heat rate shall be monitored by the incore detector system in accordance with specifications 2.10.4(1)(a) or >

2.10.4(1)(b), or maintain the Axial Shape Index, Yi within the limits of the Limiting. Condition for Operation for Excore Monitoring of LIIR Figure provided in the COLR.

(a) When the linear heat rate is continuously monitored by the incore detectors, and the linear heat rate is exceeding its limits as indicated by four or more valid coincident incore de ector alarms, either:

(i) Restore the linear heat rate to within its limits within one hour, or l

(ii) Bc in at least hot standby within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

2-56 Amendment No. 5,20,24,32 l

43,47,77,1l'7 l

2.0 Llh11 TING COND1110NSE)]LQl'EHATION 2.10 EracleLQue (Continued) 2.10.4 Power Distribution Limits (Continued)

(b) If while operating under the provisions of part (a), the plant computer incore detector alarms become inoperable, o;x ration may be continued for seven days from the date of the last valid core power distribution without reducing power provided each of the following conditions is satisfied:

(i) A core power distribution was obtained utilizing incore detectors within 7 days prior to the incore detector alarm outage and the measured peak linear heat rate was no greater than 90% of the value allowed by (1) above.

(ii) The Axial Shape Index as measured by excore detectors remains within i.05 of the value obtained at the time of the last measured incore power distribution.

(iii) Power is not increased nor has it been increased since the time of the last incore power distribution.

(c) When the linear heat rate is continuously monitored by the excore detectors, withdraw the full length CilA's beyond the long term insertion limits of Specification 2.10.2.7 and maintain the Axial Shape Index, Yi within the limits of Limiting Condition for Operations for the lixcore hionitoring of LilR Figure provided in the COLR. If the linear heat rate is exceeding its limits as determined by the Axial Shape Index, Yi , being outside the limits of the Limiting Condition for Operation for fixcore Monitoring of LilR Figure provided in the COLR:

(i) Restore the reactor power and Axial Shape Index, Yi , to within the limits of the Limiting Condition for Operations for Excore Monitoring of LilR Figure provided in the COLR within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or 1

2-57 Amendment No. 5,20,32,43, 47,117,126 l

2.0 LLJ1 TING CONDITIONS FOR OPERATION 2.10 Reactor Core (Continued) 2.10.4 Incr Distribution Limits (Continued)

(ii) Ik in at least hot standby within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

(2) Inlal Integrated Radial Peakmg FaClet The calculated value of F,7 denned by F,7 = F, (1 +T,) shall be within the limit provided in the COLR. Fa is determined from a power distribution map with no non-trippable CEA's inserted and with all full length CEA's at or above the long Term Steady State Insertion Limit for the existing Reactor Coolant Pump combination. The azimuthal tilt, T , is the measured value of T, at the time Fa is determined.

With F,7 A the limit provided in the COLR within 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months s: l T

(a) Reduce power to bring power and Fa within the limits of the F,T, Fx/

and Core Power Limitations Figure provided in the COLR, withdraw the full length CEA's to or beyond the Long Term Steady State Insertion Limits of Specification 2.10.2(7), and fully withdraw the NTCEA's, or (b) Ik in at least hot standby.

(3) Total Planar Radial Peaking Factor The calculated value of F,yT defined as F,yT = F,,(1 +T,) shall be within the limit provided in the COLR. F,, shall be determined from a power distribution map with no non-trippabic CEA's inserted and with all fulllength CEA's at or above the Long Term Steady State Insertion Limit for the existing Reactor Coolant Pump combination. This determination shall be limited to core planes between 15% and 85% cf full core height inclusive and shall exclude regions influenced by grid effects. The azimuthal tilt, T,, is the measured value of T, at the time F,, is determined.

With F,,T A the limit provided in the COLR within 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months s: l (a) Reduce power to bring power and F,,7 to within the limits of the FnT, F,,T and Core Power Limitations Figure provided in the COLR, withdraw the full leagth CEA's to or beyond the long Term Steady State Insertion Limits of Specification 2.10.2(7), and fully withdraw the NTCEA's or (b) lle in at least hot standby.

I 2-57a Amendment No. 32,43,47,70,77, 92,109,447,126

2.0 LD11 TING CONDITIONS FOR OPERATION 2.10 Reactor core (continued) 2.10.4 Power Distribution Limits (Continued)

(4) Arjmuthal Powcr Tilt (T.)

When operating above 70% of rated power, (a) The azimuthal power tilt (T,) shall not execed 0.10 whenever Mini CECOR/IIASSS is operable, the CEA's are at or above the long Term insertion Limit and Mini CECOR/IIASSS is being utilized to monitor F,,'

and F,T.

(b) 'the azimuthal power till (T,) shall not execed 0.03 whenever the provisions of 2.10.4(4)(a) do NQI allow Mini CECOR/BASSS to be utilized to monitor F,,' and Fai. With the indicated azimuthal power tilt determined to be >0.03 but <0.10, correct the power tilt within two hours or determine within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and at least once per

subsequent 8 hoars, that the total integrated radial peaking factor, Fa', is within the limit of Specification 2.10.4(2) and that the total planar radial peaking factor, F.,7, is within the limit of 2.10.4(3), or reduce power to Ic.is than 70% of rated power within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of confirming '1, >0.03.

(c) With the indicated power till determined to be 2.10, power operation may proceed up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months provided Fa7and F,yT do not exceed the power limits of the Fat, F,,T and the Core Power Limitations Figure provided in the COLR, or be in at least hot standby within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Subsequent l operation for the purpose of measurement to identify the cause of the tilt is allowable provided the power level is restricted to 20% of thu maximum allowable thermal power icvel for the existing reactor coolant pump combination, l

l l

2-57b Amendment No. 32,13,4-7,92

2.0 LIMJIING COEnlILONUDdi'13(ATION 2.10 Reactor Core (Continued) 2.10.4 l'enIJistribution Lintits (Continued)

(5) DN11R Matgin During Pout _ Operation Abog 15% of Rated _Powc1 l

(a) The following DNil related oarameters shall be maintained within the limits, shown:

(i) Cold leg Temperature (1)* l (ii) Pressurizer Pressure 22075 psia *

(iii) Reactor Coolant Flow A197,000 gpm" (iv) Axial Shape Index, Yi (2) l (b) With any of the above parameters exceeding the limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce power to less than 15% of rated power within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , llasis The limitalicn on linear heat rate ensures that in the event of a LOCA, the peak temperature of the fuel cladding will not execed 220(TF.

Either of the two core power distribution monitoring systems, the Excore Detector Monitoring System, or the incore Detector Monitoring System, provide adequate monitoring of the core power distribution and are capable of verifying that the linear heat rate does not exceed its limit. The Excore Detector Monitoring System performs this function by continuously monitoring the axial shape index with the operable quadrant symmetric excore neutron flux detectors and verifying that the axial shape index is maintained within the allowable limits of the Limiting Condition for Operation for 11xcore Monitoring of LilR Figure provided in the COLR as adjusted by Specl0 cation 2.10.4(1)(c) for the allowed linear heat rate of the Allowable Peak Linear Heat Rate vs, llurnup Figure provided in the COLR, RC pump con 0guration, and F,,T of the FJ, F,,7 and Core Power Limitations Figure provided in the COLR.

In conjunction with the use of the excore monitoring system and in establishing the axial shape index limits, the following assumptions are made: (1) the CEA insertion limits of Specincation 2.10.l(6) and long term insertion limits of Speci0 cation 2.10.l(7) are satis 0cd, (2) the flux peaking augmentation factors are as shown in Figure 2 8, and (3) the total planar radial peaking factor does not execed the limits of Speci0 cation 2.10.4(3).

Limit not applicable during either a thermal power ramp in excess of 5 % of rated thermal power per minute or a thermal power step of greater than 10% of rated thermai power.

" This number is an actual limit and corresponds to an indicated flow rate of 202,500 gpm. All other values in this listing are indicated values and include an allowance for measurement uncertainty (e.g.,543'F, indicated, allows for an actual T, of 545"F.

(1) Within the limit for Core inlet Temperature provided in the COLR.

(2) Within the Limiting Condition for Operation for DNil Monitoring Figure provided in the COLR.

2-57c Amendment No. 42,43,5-7,70, 77,92,409,117

2.0 blillIING COND11[ONS FOlt OPEllATION 2.10 Reactor core (continued) 2.10.4 j'ower Distribution 1.imits (Continued)

The incore Detector hionitoring system provides a direct meast..e of the peaking factors and the alaims which have been established for the individualincore detector segments ensure that the peak linear heat rates will be continuously maintained within the allpw,ble limits of the Allowable Peak Linear lleat Itate vs. Ilurnup Figure provided in the COLit. The setpoints for these alarms include allowances, set in the conservative directions, for the factors listed in 2.10.4.(1).

'.fotal Planar and IntegratriRadiaLPeakinc Factert(F.,T_and_E,')_and A7imultal Power Tilt (T.3 The limitaticas of F,,7 and T, are provided to ensure that the assumptiens used in the analysis for establishing the Linear lleat llate and local Power Density - liigh LCO's and LSSS setpoints remain valid during operation at the various allowable CEA group insertion limits. The limitations of F,7 and T, are provided to ensure that the assumptions used in the analysis establishing the DNB hiargin LCO, and Thermal hiargin/ Low Pressure LSSS setpoints remain valid during operation at the various allowabic CEA group insertion limits if F,,T, F,7 or T, exceed their basic limitations, operation may continue under the additional restrictions imposed by the action statements since these additional testrictions provide adequate assurance that the assumptions used in establishing the Linear lleat Itate, Thermal hiargin/ Low Pressure and local Power Density - liigh LCO's and LSSS setpoints remain valid. An azimuthal power tilt >0.10 is not expected and if it should occur, subsequent operation would be restricted to only those operations required to identify the cause of this unexpected tilt.

The value of T, that must be used in the equation F,,7 = F,,(1 + T,) and F,T = F,(1 + T,) is the measured tilt.

T i The surveillance requirements for verifying that F,y , Fa and T, are within their limits provide t

assurance that the actual values of F,y and T, do not exceed the assumed values. Verifying F,y i and F/ after each fuel loading prior to exceeding 70% of rated power provides additional assurance that the core was properly loaded.

DNillt hiargin During Power Operation dove 15% of Itakd Fown The selection of limiting safety system settings and reactor operating limits is such that:

1. No specified acceptable fuel design limits will be exceeded as a result of the design basis anticipated operational occurrences, and

2. The consequences of the design basis postulated accidents will be no more severe than the predicted acceptable consequences of the accident analysis in Section 14.

2-57d Amendment No. M,57 l

= _

_--a-----___------ - - _ - _ - _ _ _ --_w_ _- - - _ _ - - - _ - _ _ - - m

1 2.0 1AllTING COND1110NS FOR OPERATION i 2.10 Reactor Cole (Continued) 2.10.4 Power Distribution Litniu (Continued) in order for these objectives to be met, the reactor must be operated consistent with the operating limits specified for margin to DNil The parameter limits given in (5) and the F,', F,,' and Core Power Limitations Figure provided in the COLR along with the parameter limits on quadrant tilt and control element assembly position (Power Dependent insertion Limit Figure provided in the COLR) provide a high depee of assurance that the DNil overpower margin will be maintained during steady state operation.

The actions specified assure that the reactor is brought to a safe condition.

The reactor coolant pump differential pressure monitoring system that will be used to measure llow provides an accurate method of determining reactor coolant flow.

The Trocedure for determining individual pump and reactor vessel now will be as follows:

1. Obtain a pump causing AP, using the precision resistor and high accuracy digital voltmeter and converting to pressure.

2. Obtain cold leg temperature and pressurinr pressure.

3. Correct the reading to the curve specific gravity.

4. Obtain pump flows from individual pump casing vs. How curves.

5. Add the individual pump flows to obtain the best estimate reactor vessel flow.

2-57e Amendment No. 32,57

2.0 LIAllTING COElllllRNS FOR DPERATION 2.14 IkginccIcdlafr'is_EcaturcLSystem initiation Instnimentation Sellings (Continued)

(3) C9ntainmcalHigh RadialiDn (Air Monitoring) (Continued)

The setpoints for the isolation function will be calculated in accordance with the ODCM.

Each channel in supplied from a separate instrument A.C. bus and each auxiliary relay requires power to operate. On failure of a single A.C. supply, the A and B matrices will assume a one-out-of-two logic.

(4) 12LESicam Generator Prenutc A signal is provided upon sensing a low pressure in a steam generator to close the main steara isolation valves in order to minimize the temperature reduction in the reactor coolant system with resultant loss of water level and possibic addition of reactivity. The setting of 500 psia includes a 122 psl uncertainty and was the setting used in the safety analysis.*

Closure of the MSIVs (and the bypass valves, along with main feedwater isolation and bypass valves) is accomplished by the steam generator isolation signal which is a logical combination of low steam generator pressure or high containment pressure.

As part of the AFW actuation logic, a separate signal is 9tovided to terminate flow to a steam generator upon sensing a low pressure in that steam generator if the other steam genciator pressure is greater than the pressure setting. This is done to minimite the temperature reduction in the reactor coolant system in the event of a main steam line break. The setting of 466.7 psia includes 1. +31,7 psi uncertainty; therefore, a setting of 435 psia was used in the safety analysis.

(5) SIRW Tank Low Level .

l.cvel switches are provided on the SIRW tank to actuate the valves in the safety injectici pump suction lines in such a manner so as to switch the water supply from the SIRW tank n the containment sump for a recirculation mode of operation after a period of approximately 24 minutes following a safety injection signal. The switch-over point of 16 inches above tank bottom is set to prevent the pumps from running dry daring the 10 seconds required to streke the valves and to hold in reserve approximately 28,000 gallons of water of at least the refueling boron concentration. The FSAR loss of coolant accident analysis

assumed the recirculation started when the minitr.um usable volume of 283,000 gallons had been pumped from the tank.

2-62 Amendment No. 5,32,43,65,86,403,108,133

3.0 SURVEll.LANCidliQUIREMiiNTS 3.10 Reactor Core Parameters (Continued)

(2) Moderator Temper.ture Coef0cient The MTC shall be determined at the following frequencies and power conditions during each fuel cycle:

1. Prior to initial operation above 5% of rated power, after each fuel loading, i

2. At any power level within 500 MWD /T of initial operation after each refueling.

3. At any power level within i 14 EFPD of reaching a rated power equilibrium baron concentration of 300 ppm.

(3) Regulating CEA Insertion Limits t

a. The position of each regulating CEA group shall be determined to be above the Transient Insertion Limits at least once per shift.

b. The accumulated times during which the regulating CEA groups are inserted beyond the Steady State Insertion Limits but above the Tnmsient insertion Limits shall be determined once per day.

(4) Linear lieat Rate Monitoring Systes.1s

a. The incore detector monitoring system may be used for monitaring the core power distributior, provided that at least once per 31 days of accumulated power operation the incore detector alarms generated by the plant computer are verified to be valid and satisfy the requirements of the core distribution map.

b. The excore detector monitoring system may be used for monitoring the core power distribution by:

1. Ver;iying at least once per 31 days of accumulated power operation that the axial shape index, Yi , monitoring limit setpoints are maintained within the allowable limits of the Limiting Condition for Operations for Excore LiiR Monitoring Figure provided in_ the COLR, as adjusted by ;

Speci0 cation 2.10.4(1).

l 1

(5) Total Integrated and Total Planar Radial Peaking Factors (FxT and F,,T)

F,7 and F,,7 shall be determined to be within the limits of Speci0 cation 2.10.4 at the following l intervals

a. After each refueling and prior to operation above 70 percent of rated power.

b. At least once per 31 EFPD's of accumulated power operation.

3-63a Amendment No. 3h109 i

. i 4.0 DESIGLEEAH!Ris 4.4 EucLSicIagc 4.4.1 New l oci sicags ,

The new unitradiated fuel bundles will normally be stored in the dry new fuel storage rack with an effective multiplication factor of less than 0.9. The open grating Door below the rack and the covers above the racks, along with generous provision for drainage, precludes no(xling of the new fuel storage rack.

New fuel may also be stored in shipping containers or in the spent fuell xxil racks which have a maximum effective n'ultip!! cation factor of 0.95 with Fort Calnoun Type C fuel and unborated water.

The new fuel storage racks are designed as a Class I structure.

4.4.2 Sgnt Fuel S10Iacc trradiated fuel bundles will be stored prior to off site shipment in the stainless steel lined spent fuel pool. The spent fuel pool is normally Olled with borated water with a ,

concentration of at least the refueling boron concentration. ,

The spent fuel racles are designed as a Class I structure.

Normally the spent fuel pool cooling system will maintain the bulk water temperature of the pool below 120"F. Under other conditions of fuel discharge, the fuel pool water temperature is maintained below 140'F.

l The spent fuel racks are designed and will be maintained such that the calculated l effective multiplication factor is no greater than 0.95 (including all known uncertainties) assuming the lxx>l is Gooded with tmborated water. The racks are divided into 2 regions. Region I racks are surrounded by Ik)ranex; Region 2 racks have no poison.

Acceptance criteria for fuel storage in Regions I and 2 are delineated in Section 2.8 of l these Technical Speci0 cations.

1 1

l l

4-4 Amendment No. 43,43,75,103,133

~ .- .- --

5.0 AllMINISTMTIVILCMI'ROLS Rupelnibililin 5.5.1.6 The Plant Review Committee shall be responsible for: '

a. Review of 1) all procedures required by Specincatlun 5.8 and changes thereto, 2; any other proposed procedures or changes thereto as determined by the Manager - Fort Calhoun Station to affect ne:Icar safety,

b. Review of all proposed tests and experiments that affect nuclear safety.

c. Review of all proposed ci'anges to the Technical Specincations,

d. Review of all proposed changes to the Core Operating Limits Report.

l

c. Review of idl proposed chang:s or modifications to plant systems or equipment l that affect nuc' car safety.

f. Investigation of all violations of the Technical Specifications and shall prepare and l

. .rward a report covering evaluation and recommendations to prevent recurrence to the Division Manager Nuclear Operations and to the Chairman of the Safety Audit and Review Committee.

g. Review of facility operations to detect potential safety hazards. l

h. Performance of special reviews and investigations and reports thereor. as l requested by the Chairman of the Safety Audit and Review Committee,

i. Review of the Site Security Plan and implememing procedures and shall submit recommended changes to the Chairman of the Safety Audit and Rev Committee.

J. Review of the Site Emergency P!an and implementing procedures and shall submit recommended changes to the Chairman of the Safety Aud Committee,

k. Review of all Reportable Events. l Authority 5.5.1.7 The Plant Review Committee shall:

a. Recommend in writing to the Manager - Fort Calhoun Station approval or

! disapproval of items considered under 5.5.1.6(a) through (c) above. l t

5-4 Amendment Nc. 9,19,84,99,115 l

5.0 bDMINISTRATIVE CONTROLS 5.5.1.7 b. Render determinations in writing with regard to whether or not each item considered under 5.5.1.6(a) through (f) above constitutes an unreviewed safety l question,

c. Provide immediate written notification to the Division Manager Nuclear Operations and the Safety Audit and Review Committee of disagreement between the Plant Review Committee and the Manager - Fort Calhoun Station; however, the Manager - Fort Calhoun Station shall have responsibility for resolution of such disagreements pursuant to 5.1.1 above.

Recmh 5.5 1.8 The Plant Review Committee shall maintain written minutes of each meeting and copies shall be provided to the Division Manager - Nuclear Operations and Chairman of the Safety Audit and Review Committee.

5.5.2 hfety Audit and Review Committee (SARC

' BElim 5.5.2.1 The Safety Audit and Review Committee shall function to provide the independent review and audit of designated activities in the areas of:

a. nue: car power plant operation

b. nuclear engineering

e. chemistry and radiochemistry

d. metallurgy

c. instrumentation and control

f. radiological safety

g. mechanical and electrical engineering

h. quality assurance

.CoJupD1illE 5.5.2.2 The Safety Audit and Review Committee shall be composed of:

Chairman: Division Manager - Nuclear Services Member: Senior Vice President Member: Division Manager - Nuclear Operations Member: Division Manager - Production Engineering Member: Manager - Fort Calhoun Station l Member: Manager - Radiological Services Member: Qualified Consultants as Required and r.s Determined by SARC Chairman l

5-5 Amendment No. 86,93,99,401,115,4-19,132

E9.5 CI v. A e 1. .tLReson 1 & >< a..ag L.mits shall be established and documented in the Core Operating nc " , befort each reload cycle or any remaining part of a reload cycle, g3 utalytical methods used to deterin..te the core operating limits shall be those

' / reviewed and approved by the NRC as follows:

s.

1 OPPD-:. "^1-P-A, Rev. 3 " Reload Core Analysis Methodology li Overvies- n ; April 1988.

' PPD-MA-8302-P-A, Rev. 2 "Neutronics Design Methods and Verification' dated Apri! 1988.

k 3. OPrD NA *F6-P-a, Rev. 2 " Transient and Accident Methods and Verification" dated April 1988.

c. The core operath.g limits shall be detern' iced so thst all applicable limits of the s,.i 'v analysis are met. The Core Operating Limits Report, including any mid-cycx avisions or supplemems thereto, shall be provided upon issuance, for each reload cycs to the NRC Docuinent Control Desk with copies to the Region IV Administrator and Senior Residere Inspector.

3 J

i

?

m x

5-17a Amend lent No.

~~ _ .

ATTACHMENT D f.xamp.ie CDIfLC. Ating Limits Report

Fort Calhoun Station, Unit 1 Cyclej a_C ore _O p eratinglimita_B e port Due to the critical aspects of the safety analysis inputs contained in this reaart, changos rnay not be roado to this report without concurrence of the Production Engineering Division, Nuclear Engineering Department.

IABLE_OECONTENIS lem Dcacilplion Eago 1.0 Introduction 1 2.0 Coro Operating Limits 1 3,0 TM/LP Limit 1 4.0 Maximum Coro inlet Temperature 2 S,0 Power Dependent insertion Limit 2 6.0 Linear Heat Rate 2 7,0 Excore Monitoring of LHR 2 8.0 Peaking Factor (FxyT , FnT) Limits 2 9.0 DNB Monitoring 3 10.0 FRT, F, T and Coro Power Limitations 3 11,0- . Refueling Boron Concontration 3 i

1 I

e

l LIST _DEIABLES Inblello. Ilito Eago 1 TM/LP Coefficients 1 2 Maximum Fn and Fxy Limits 3 t

puu

_c L[ST OF FIGUBES Figuto_tio. Ilite Eage 1 Thermal Margin / Low Pressuro 4 Pump Operation 4 2 Power Dependent Insertion Limit 5 3 Allowance Peak Linear Heat Rate vs. Burnup 6 4 Excore Monitoring Of LHR 7 5 DNB Monitoring 8 6 FxyT , pgT and Core Power Limitations 9

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

1.0 Initoducilon This report provides the cycle-specific limits for operation of the Fort Calhoun Station Unit 1 for Cycle 13 operation. It includes limits for:

1.) TM/LP LSSS 4 Pump Operation (PVAR) 2.) Core inlet Temperature (Tw) 3.) Power Dependent insertion Limit (PDIL) 4.) Allowable Peak Linear Heat Rate l 5.) Excore Monitoring cl LHR T

. 7.)

6.)integrated Planar Radial RadialPeaking Peaking Factor Factor (Fn (Fxy ) T) 8.) DNB Monitoring 9.-) FnT/p ,yTversus Power Trade off curve 10.) Refueling Boron Concentration These limits are applicable for the duration of Cycle 13. For subsequent cycles the limits will be reviewed and revised as necessary. In addition, this report includes a number of cycle-specific coefficients used in the generation of certain reactor protective system trip setpoints or at awable increases in radial peaking factors.

This report has been prepared in accordance with the requirements of Technical Specification 5.9.5. The core operating limits havo been developed using the NRC methodologies listed in Tech-nical Specification 5.9.5.b.

2.0 Core Operating Limitu The values and limits presented within this section have been derived using the NRC-approved .

methodologies listed in Technical Specification 5.9.5. All values and limits in this section apply to Cycle 13 operation. Cycle 13 must be opoated within the bounds of these limits and all othe's specified in the Technical Specifications.

3.0 IM/LP Limit The TM/LP coefficients for Cycle 13 are shown below:

Table 1 Coafficient Valua a 29.73 6 18.44 y 11240 1

The TivulP setpoint is calculated by the PVAg equation, shown in Figure 1; PVAR = 29.73 PF(B) A1(Y)B + 18.44 TIN - 11240 l PF(B) = 1.0 B.>_100%

= -0.008B + 1.8 50% < B < 100%

= 1.4 8 5 50%

A1(Y) = -0735294Yi + 1.08824 Yi s 0.25

= 0.57143Y; + 0.875 Yi > 0.25 l l B = High Auctioneered thermal (AT) or Nuclear Power, % of rated power Y = Axial Shape Index, asiu i Tni = Core inlet Temperature, F i PVAR = Reactor Coolant System Pressure, psia 4.0 Maximum Core inlet Temp _erature The maximum core inlet temperature (flN) for Cycle 13 shall not exceed 543*F.

This limit is not applicable during either a thermal power ramp in excess of 5% of rated thermal power per minute or a thermal power step greater thc. .10% of rated thermal power.

l 5.0 Power Dependent insertion Limit i

The power dependent insertion limit is oefined in Figure 2 for Cycle 13 operation.

6.0 Lineatthat. Bate l The allowable peak linear heat rate vs. burnup is 14,4 kw/ft for Cycle 13 for all burnup points as

! shown on Figure 3.

1 l

l 7.0 Excare Monitorina of LHR j The allowable operation for power versus axial shape index for monitoring of LHR with excore de-tectors for Cycle 13 is shown in Figure 4.

8.0 Peaking Eastor Limits The maximum full power values for the unrodded panar radial peaking factor (FxyT ) and integrated radial peakirig factor (FRT) are shown in Table 2.

1 l

2 l

Table 2 Maxjmum Fu]LPower F 9and Fxy LJ1mila Ecahina Factor 1imith FT 1.75 Fh 1.70 9.0 DND_ Mon!1orjng The limits for DNB as a function of axial s:. ape index and core power is shown in Figuro 5.

Coro power shall be maintained within the limits established by the Botter Axial Shapo Selection System (BASSS) for CEA insertion of the lead bank of <65% (i.e., greater than or equal to 44.1 inches withdrawn) when BASSS is operablo, or within the limits of Figuro 5.

10.0 Fyl F 9T and. Core PoWctilmliallons Coro power limitations versus Fn Tand FxyT are shown in Figure G. ,

110 Helueling Boron _ Concentration The refueling boron concentration must be maintained with a boron concentration of at least 1900 ppm in the reactor coolant system to ensure a shutdown margin of not less tha'15% with all CEAs withdrawn.

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510 1750 psic 500 60 .. 70 80 90 i 100 110 120 CORE POWER (". OF RATED POWER)

P yAg =M.W @ @$+2.W - 11240 ig PF(B) =10 81100%.

= - Ur,88 -- 1.8 50%<B<100%

= 1.4 Bs50%

l A1(Y) = -0.35294Yi + 1.08824 Y; s .25

=

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fort C0thoun Station-Unit No.1 i 1

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INSENTION LIMIT

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i i i 73 i i 0 2500 5000 7500 10000 12500 15000 CYCLE AVERAGE BURNUP (MWD /MTU) i Allowable Peak Linear Heat Rate Omaha Public Power Distnct Figure I vs. Burnuo l Fort Calhoun Station-Unit No.1 3 s

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Limitaticas Fort Calhoun Statiand.b.it No. ! 6 I

9

_-_-___-__-________---____-_-_______________-______-_--____-_-_____-____-_-_____-_-_-___A

ML20086G647

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