ML19332E792

Proposed Tech Specs,Making Typos & Administrative Corrections

ML19332E792
Person / Time
Site:	Fort Calhoun
Issue date:	12/01/1989
From:	OMAHA PUBLIC POWER DISTRICT
To:
Shared Package
ML19332E790	List: ML19332E789 ML19332E791 ML19332E792
References
NUDOCS 8912120151
Download: ML19332E792 (43)

Text

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TECHNICAL SPECIFICATIONS - FIGURES ..

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.T.A.B.L.E. .O.F..C. .O.NTE NTS El

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m DESCRIPTION PAGE tlHICH FictL)1LE~4ABt4- FOLlDWS 1-1 TMLP Safety Limits 4 Pump Operations. . . . . . . . . . . . . 1-3 g 1-2

- Axial Power Distribution LSSS for 4 Pump Operation. . . . . . i 1-3 1-3 l TMLP LSSS 4 Pump Operation. . . . . t 2-1A RCS Press-Temp Limits '

........... /-6

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. . .... ...... 7a  ! t 2-1B RCS Press-Temp ' '

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...........  ?-7a i

2-3 Predicted R66 .. ' 7 ;hift. }

........... 2-7a i

2-10 Spent r- i , '

L ar .ege Cruer- ..

......  ?-38a' 2-4 P0ll. .

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..... 2-57e 2-5 A110wabi. # L ee e rMat Ratc vs Bur, p.......... c

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-?-6 LCO for Excore Moni wr ..is ci LHP. . . . . . . . . . ....  ?-57e -

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LCO for D'!D Poni toring. . . . . . . . . . . . . . . . . . . .  ?-57e 2-8 t.c Flux Feaking Augmentation Factors . . . . . . . . . . . . . .

t T T  ?-57e l 2-9 F A, FXY ,

and Core Power Limitations. . . . . . . . . . . . .  ?-57e t L

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P viii Amendment tio.116 f

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. 1. SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS -

- 1.1 Safety Limits Reactor Core (Continuea) ,

L would cause_DNB at-a particular core location to the actual heat- flux at that location, is indicative of the margin to l

• 1 l '

DNB. The minimum value of the DNBR during steady state opera-tion, normal operational transients, and anticipated tran- l sients is limited to 1.18. A DN8R of 1.18 corresponds to a j 95% probability at a 95% confidence level that DNB will not I occur, which is considered an appropriate margin to DNB for l all: operating conditions.(1). l l

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. The curves of Figure-1-1 represent the loci' of points or re-  ;

actor thermal power (either neutron flux instruments or AT in- l struments), reactor coolant system pressure, and cold leg l temperature for which the ONBR is 1.18. The area of safe, opera- l tion is below these lines.

The reactor core safety limits are based on radial' peaks limit-ed by the CEA insertion limits in Section 2-10 and axial

• shapes within the axial power _ distribution trip limits in i Figure 1-2 'and a total unrodded planar radial peak of 1.80. l. l The LSSS in Figure 1-3 is based on the assumption that the un-  !

rodded integrated total radial peak'(F ) is 1.80. .This peak- i ing factor is slightly higher (more ce servative) that the maximum predicted unrodded total radial peak during core life, excluding measurement uncertainty. i t

i l Flow maldistribution effects for operation under less than full = reactor coolant flow have been evaluated via-model

• L At --+.. tea t (2) The flow model data established the maldistribution

. factors and hot channel inlet temperature for the thermal-L analyses that were used to establish the safe operating enve-l lopes presented in Figure 1-1. -The-reactor protective system. 'l is designed to prevent any anticipated combination of tran - '

i sient conditions for reactor coolant system temperature, pres- l l- . sure, and thermal power level that would result in a DNBR of less than 1.18.(3)

References 1

(1) USAR, Section 3.6.7 L 2 . % E. -

(2) USAR, Section 1.4.6

(3) USAR, Secticn 4r6

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l l.> l 1-2 Amendment No. 8,32,43,47, 76,77,92,11.7 3 l

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s 2.'O- LIMITING CONDITIONS FOR OPERATION 2.l' Reactor Coolant System (Continued) 2.1.2 Heatup and Cooldown Rate (Continued)

' ~(a) The curve in. Figure 2-3 shall be used to predict the increase in transition: temperature based on integrated fast neutron flux. If measurements on the irradiation a specimens indicate a deviation from this curve, a new >

curve shall be constructed..

(b) The; limit line on the figures shall be updated for_a new-integrated power period as follows: the total integrated ,

reactor thermal power from startup to the end of the new period shall be converted to an equivalent integrated fast neutron-exposure (Ell MeV), For this plant, based upon

-surveillance materials tests, weld chemical composition data,. and the effect of a reduced vessel fluence rate provided by core load designs beginning with fuel Cycle 8, i

the predicted surface fluence at the initial reactor vessel l

beltline weld material for 40 years at 1500 MWt and an 80%

load factor is 2.55x10 W n'/cm2 The flux reduction applied '

to the fluence calculations was based on Cycle 1-7 average and Cycle 8 average-azimuthal flux distribution plots generated used 00T 4.3. The predicted transition temperature l shift to the end of the new period shall then be'obtained from Figure 2-3.

(c) The limit-lines in Figures 2-1A and 2-1B'shall be moved parallel _to=the temperature axis (horizontal) in the direction of increasing temperature a distance' equivalent to the transition temperature shift during the period since the curves were last constructed. The hf Mp temperature limit line shall remain at 82*F as it is set by the NDTT of the reactor vessel flange and not subject to fast neutron flux. The lowest service temperature shall remain at 182 F because components related to this temperature are also not subject to fast neutron flux.

(d) The Technical Specification 2.3(3) shall be revised each time the curves of Figures 2-1A and 2-18 are revised.

Basis bl tup-All components in the reactor coolant system are designed to withstand the effects of cyclic andpressurechanges.dgadsduetoreactorcoolantsystemtemperature These cyclic loads are introduced by normal unit load transients, reactor trips and startup and shutdown operation.

During' unit startup and shutdown, the rate'sof temperature and pressure changes are-limited. The design nun.ber of cycles for heatup and cool-e down is based upon a rate of 100 F in any one hour period and for cyclic

. operation.

2-4 Amendment No. 22,47,64,74,77,200, n4 g-a __d

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2.0- ' LIMITING CONDITIONS FOR OPERATION 7

2.5 Steam and Feedvater Systems Applicability LApplies to the operating status of the steam and feedvater systems. _j l'

, Objective r To define certain conditions for the steam and feedvater system j necessary to assure adequate decay heat removal. l i

Specifications a bo v e, The reactor coolant shall not be heated obeet 3000F unless the following conditions are met:

~

.l (1) Ecth auxiliary feedvater pumps are. operable. One of the auxiliary feedvater pu=ps mcy be inoperable f or 2h - hours provided that the redundant component shall be tested to  ;

demonstrate operability.

(2) A minimum of 55,000 gallons of water in the emergency feedvater storage tank and a backup water supply to the emergency feedvater storage tank from the Missouri River '

r- by the fire water system.

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(3) All. valves, inter 1ceks and piping associated with the ]

i above components required to function during accident l conditions are operable. Manual valves that could inter-rupt auxiliary feedvater flow to the steam generators chall be locked in the required position to ensure a flev path to the steam generators. -

( h )' ' 2.e main steam stop valves are operable and capable o; .;

closing in four seconds or less under no-flow conditions. i I

Easis

)i A reactor shutdown from power requires a removal of core decay f heat. Immediate decay heat removal requirements are ner= ally satisfied by the steam bypass to the condenser. Therefore,  ;

core decay heat can'be continuously dissipated via the steam q bypass to the condenser as' long as feedvater to the steam <

generator ic available. Normally, the capability to supply feedvater to the steam generators is provided by operation  ;

of the turbine cycle feedvater system. In thc unlikely event 5 of complete loss of electrical power to the station, decay  ;

heat remova). is by steam discharge to the atmosphere via the main ster. safety and atmospheric dump valvea. Either auxi'-

liary feedvater pump can supply sufficient feedvater for re-moval of decay heat from the plant. The minimum amount of water in the emergency feedvater storage tank is the amount needed far 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of such operation. The tank can be re-supplied with water from the fire protection system. (1)

Amendment No. 49- 2-28

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?i 2.0 LIMITING CONDITIONS FOR OPERATION'  !

.a d 2.5. steem and Feedvater Systems (continued) l 0

A closure time of k seconds for the main steam stop valves is considered '

adequate and was selected as-being consistent with expected response-p.

'timefor{nt analysis. 2 .- entation as detailed in the steam line-break incident n

L References u .

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. (1) ,ESAR , Section 9. h . 6 gA :

(2) JSAR,'Section10.3

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's U (3);yAR,Section14.12  !

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[ 2.0 . LIMITING CONDITIONS FOR OPERATION 2.1 Electrics 1 3 utems (Continuec) f rom either one of two diesel generators and of f-site standby power via the unit auxiliary transform 1rs.(1)

The two emergency diesel generators on site do not require outside power for start up or operation.

Upon loss of normal and standby power sources, the 4.16 kV buses lA3 and 1A4 are energited from the diesel generators. . Bus load shedding, trans-fer to the-diesel generator and pickup of critical loads are carried out automatically.(2)

When the turbine generator is out of service for an extended period, the

' generator can-be isolated by opening motor operated disconnect switch DS-T1 in the. bus between the generator and the main transformer,' allow-ing the main transformer and the unit auxiliary power transformers .co be returned to service. (3)

Equipment served by 4' 16 kV and 480 V auxiliary buses and MCC's is '

arranged so that loss of an entire 4.16 kV bus does not compromise safety of the plant during DBA conditions. For example, if 4.16 kV bus LA3 is lost, two raw water pumps, one low pressure safety injection pump, one -t high pressure safety injection pump, one auxiliary feedwater pump, two.

I component cooling water pumps, two containment spray pumps and two con-tainment air fans are lost. This leaves two raw water pumps, one low pressure safety injection pump, two high pressure safety injection pumps, one component cooling water pump, one containment spray pump and two con-tainment air fans which is mora than' sufficient to control cestainment pressure below the design value during the DBA.

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The total uel oil engine base tank capaci:v of 550 gallons on m .h j diesel 4; a .sidcra 'more than cd:pe n ;i.:: .s: ^ ~'" M: 4 rumdng time (worst case loading) is available before transfer o f C.el c:2 from the 18,000 gallon underground storage tanks is mandatort. Two 13 gpm diesel oil. traasfer pumps per diesel, with each being fed from tne diesel it is associated with, are available for transferring fuel oil from the storage-cank to th gdav tanks. The 16,000 gallons in the-storage tank in addition to the eeybpynks will provide diesel operation under the required loading condiciens for a minimum period of M should only one diesel be in operation. It is considered incredible not to be able to *** f 1 gil from one of several sources in the vicinity Of Omana :n less than davsungertheworst of weather conditions, pre e.u re cMA cach ve.t-One battery charger on each battery shall be operating so that the batteries will always be at full charge; this ensures that adequate d-c power will be avilable for all emerger.cy uses. Each battery has one battery charger per=anently connected with a third charger capable of being connected to either bactery bus. The enargers are each rated Acencment No. 7o 2-35

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2.O' . LIMITING COffDITIONS FOR OPERATION

2.7 Flectrical Systems (Continued) 400 for 490 amperes at 130 volts. Except for the first minute following a i DBA;during which the batteries and the charger accommodate al.L the load, l the capacity of the battery charger vill handle all 2-equired ' loads. - Each I of the reactor protective system channels instrumentation channels is supplied by one of the a-c -instrument buses. The removal of one of the a-c instrument buses is permitted as the.2-of-b logic may be manually changed to a 2-of-3 logic without compromising safety.

The engineered safeguards instrument channels use a-c instrument buses (one redundo.nt bus for each channel) and d-c buses (one redundant bus for each logic. circuit).

The removal of one of the a-c ' instrument buses is permitted as the two of four logic automatically becomes a two of three lecie.

Re feren ces

' .M (1) gSAR, Section 8.3.1.2 V, ' 7 . 3 .'t. . *2 (2) /SAR, Section & 471 LA (3) /SAR, Section 8.2.2

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^ l 2.01 LIMITING CONDITIONS FOR OPERATION -

2.8 ~ RefueHno coerations (Continued)  !

1 L 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 maintensnce of the. instrumentation. . Continuous monitoring of radiation l 1evels-and neutron flux prnvides immediate indication of an unsafe condi-L tion. The shutdown cooling pumo is used to maintain a unifera boron concentration.

The shutdown margin as indicated will keep the coreu s'beritical even if all CEA's were withdrawn from the core. Ooring refueling operations, the l reactor refue11ng cavity is filled with approximately 250.000 gallons of.

borated water. . Theboronconcentral%bnofthiswater (at least 1800 ppm  ;

boron) is sufficient to maintain the reactor suberitical by more than 5%, 1

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including allowance for uncertainties, in the cold condition with all rods withdrawa.(2) Periodic checks of refueling water baron concentration ensure the proper shutdown margin. Communication requirements allow the o

control room operator to inform the refueling machine operator of any impending unsafe conditiun detected from the main control board indicators o during fuel movecent.

'In addition to the above engineered safety featuras,,interlocxs are utilized during refueling operations to ensure safe handling.- An excess

• Wi ght interlock is provided on the lifting hoist to prevent movement of more tha one fuel assembly at a time. In addition, interlocks on the auxiliary building crane will prevent the trolley from being moved over storage rccks containing irradiated fuel, except as necessary for the handling of fuel. The restriction of'not moving fuel in the reactor for i a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after the power. has been removea 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 closeo 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 wnenever irradiated fuel is being handled.(1)

References u

(1) JSAR, Section 9.5 (2) 15AR, Section 0. 5.1. 2 6:" 2. 3 . l l 4

2-39 Amendment No. 24,75,103,II7 P

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-2.0L LIM.TTING CONDITIONS FOR OPERATION.  !

2.30 Reactor Core (Continuca) ,

'2.10.4' Power Distribution Limits (Continued) -

(5) DNBR Marcin Durino-Power Oceration Above 15% Rated Power (a) The following DNB related parameters shall be maintained

? 4 within the limits shown:

(i) '

ald Leg Temperature 5543*F* 4 (ii) A sssurizer Pressure 22075 psia"

-(fii) neactor Coo? int Flow R 2197,000 gpm"*

(iv) Axial Shape Index, Y g 5 Figure 2-7"**

(b) With any of the aoove parameters ' exceeding the limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce power to less than 15% of rated power within 'the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Basis-

'The. limitation on linear heat rate ensures that in the event of a LOCA, the

' peak temperature of the fuel cladding will not exceed 2200*F.

Either of the two core power distribution monitoring systems, the Excore Detec-ter Monitoring System, or the Incore Detector Monitoring System, provide b

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 y

Monitoring System performs,this function by continuously m itoring the axial E shape index with the operable quadrant symmetric excore neutron flua detectors and verifying that;the axial shape index is maintained within the aliowable limits of Figure 2-6 as adjusted by Specification 2.10.4(1)(c)-for the allowed.

linear heat rate of Figure 2-5, RC Pump configuratton, and F T of Figure 2-9.

l In conjunction with the use of the excore monitoring system Ed in establishing L the axial shape index limits, the following assumptions are made: (1) the CEA L insertion limits of Specification 2.10.1(6) and long term insertion limits of L

Specification 2.10.1.(7) are satisfied, (2) the- flux peaking augmentation factors are as shown in Figure 2-3, and (3) the total planar radial. peaking

. factor does not exceed the limits of Specification 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 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 actral T of 545*F.

'***The AXIAL SHAPE INDEX.C Core power shall be maintainea within the limits .

established by the Better Axial Shape Selection System (BASSS) for CEA incertion of the lead bank of < 65% when BASSS is operable, or within the limits of Figure 2-7.

2-57c Amendment No. 32,43,57,70, 77,92,I09,117

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2.O LIMITING- CONDITIONS FOR OPERATION 2.12 Control Room Systems

• Aeolicability Applies to the control room air conditioning and filtering systems.

Objective To limit the environmental conditions in the contrel room, under normal and post DBA conditions.

Seecifications (1) If the control room air temperature reaches 1200F, immediate

• action shall be taken t o reduce this temperature. If the temperature cannot be reduced to belev~1200F in four hours, the reaator vill be placed in a hot shutdown condition. ,

(2) A thermometer must be in the control room at all times.

(3) All areas of the plant which hrve safety related instrumentation '

vill be observed during hot functional testin6 to determine local temperatures and monitored during operation if normal plant ventilation is not available,

.(h)

-- x.i e. sa t h; d e^. i t hat i ..m w.. m ' . eu eu :. ;e; v

~~ system . < de or found to be inoperable for an -

ocn, t -

reactor operation .,

'able on . the succeedint; LJ' seven days unless such d- -

- 2. s perable. I:

1hese condi -

inot be met, the reactor sha b -

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'"he reactor protective system and the engineered safeguards syst 3m were designed for'and the instrumentation was tested at 12007.

l Therefore , if the temperature of the control room exceeds 120 0F, the reactor vill be shutdown and the conditions corrected to precluda ,

j

! failure of componenr.s in an untested environment.-

I  !

If the control room air treatment system is found to be. inoperable ,- -

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there is no immediate threat to the control room and reattor opera-  ;

tion may continue for a limited period of time while repairs are being made.

If the system cannot be repaired within seven (7) days, the 4

reactor is shutdown and brought to cold shutdown within 2h hours.

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o 2-59 Amendment No. 15 j

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TABLE 2-10 Post-Accident Monitorino Instrumentation Operatine Li...o s l Minimum Operable Instrument Ctanneh Action

'1. Containment Wide Range Radiation 2 (a)

Monitors-(RM-091A & B)

2. Wide Range koble Gas Stack Monitor l RM-063L Noble Gas Portion Only 1 (a l RM-063M Noble Gas Portion Only 1 (e {

RM-063H Noble Gas Portion Only 1 (a,  !

-3. Main Steam Line Radiation Monitor 1 (a)

.(RM-064)

4. Containment Hydrogen Monitor (VA-81A & B) ,2 (b)(c)

5. Containment Water Leye_L.__.----

Narrow Range (LT-59) & LT-600) 1 l (d)(c)

Wide Range (LT-387 & LT-388) 2 (b)

. 6. Containment Wide Range Pressure 2 (b)(c)

7. Reactor Coolant System Subcooled Margin Monitor 2 (e)(f)

8. Core Exit Thermocouples (1) 2/ Core-Quadrant (g)(h)  ;

9. Reactor Vessel Level (HJTC) (j) 2 (k)(1) l (a) With the number of OPERABLE cnannels less than required by the minimum-l channels operable requirements, initiate the pre-i method of monitoring the appropriate parameter within(s) 72 planned alternate hours, and

1. either restore the inoperable channel (s) to OPERABLE status within '

l 7 days of the event, or

2. prepare and submit a special report to the Consnission pursuant to specification 5.9.3 within 14 days following the event outlining the action taken, the cause of the inoperability, and the plans u

and schedules for restoring the system to OPERABLE status.

(b) With one channel inoperable, restore the inoperable monitor to OPERA 8LE status within 30 days or be in at least H0T SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

l 2-98 Amendment No. BJ, B2, E3, 87,110 l

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3. Illilititi!! FHiMtlEllClici. Folt citi:CE:5. CA!.iltilATIolls AtlD TESTIllG OF REACTOR PR(YTECTIVE SYSTIJi g:

.3- Survei1 lance M t'imunel Description Paric ti on Fresumney Surveillance tietnod

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p 1. P,uer Range Safety. a. Check S u. C<* liarison of feur power

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.tf Channels channel readinds, for both neutron flut and Lliertal power.

.E' b. Adjustment IIII' b b. Channel adjustment. to

$ agree with heut balance calcialation.

O c. Ca<ibrate  :-i(2) c. Internal test signal to end Test. veri fy trips, alarms , per -

niis.sives anil une'.loneer w c1rcult.s.

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2. Vidu-Range Logaritlunic a. Citeck 3 a. Compar_ son of four wide-rande .

Ileutron I-loni tors readings.

b. Tes t,( 3) P- b. I:it.ernal test signals to verify SUR indication and trip, power level pertaissives, instrument, accuracy.

3  !<eactor coolunt. Plov a. Check S a. Comparison or rour separnte total flow indicatior.s.

b. Calibrate R b. Known di 'erential pressure applled to sensors to cali-brute all loop devices.

c. Test M(2) c. Bistable trip tester.(~1)

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s 3.0 SURVEILLANCE' REOUIREMENTS 3.9 Auxiliary Feedwater System (Continued) (

Basis 6

^

The valve position verifications performed monthly and following auxiliary feedwater rystem maintenance will confirm the avail-ability of an auxiliary feedwater flow path to the steam generators.

The testing of-the au..iliary feedwater pumps avery month and af ter enld shutdowns will verify their operability by recirculeting water to the emergency feedwater storage tank. '

Operating:the regulating valves (HCV-1107A, hCV-1107B, HCV-1108A and HCV-11088) one.at a time every three months and after cold shutdowns '

- will confirm a flow path to the steam generators and operability of the valves.

Proper functioning.of the steam turbine admission valve and s*Trting of the feedwcter pump will demonstrate che integrity of the s eam driven pump. Verification of ccrrect operation-will be made both "

from instrumentation within the main control room and direct visual observation of the pumps.  ;

The operability of the auxiliary feedwater system ensure that the 3 0 c

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~ reactor coolant system can be cooled down to less than - F from normal operating condition the event of a total loss of off-site power. N References 6 a T* w L c.b h . th e. S kurde* coelEs Sqsh ws m=9 b e, pisc e.1 iw ro openD$w' ;.

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l 3-62a Amendment No. H , M , 90 l

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5.0' ADMINISTRATIVE CONTROLS .

5.4 Trainino 5.4.1 A' retraining and replacement training program for thit plant staff shall be maintained under thc direction of the Managtr - Training and l  ;

shall meet or exceed the requirements of Section 5.5 ef ANSI l

N18.1-1971 and 10 CFR Part 55. { 1 x

5.4.2 A training program for. the fire brigade shall be maintained under the ,

Manager - Training and shall meet or exceed the requirements of l Section' 27 of NFPA Code-1975, except that the meeting frequently may be quarterly.

5.5 Review and Audit 5.5.1 Plant Revier Conmittee (PRC)

Function 5.5.1.1 The Plant Review Committee shall function to advise the Manager - Fort Calhcun-Station on all matters related to nuclear safety.

Composition 5.5.1.2 - The official Plant Review Committee shall be composed of the:

Chairman : Manager - Fort Calhoun Station Member: Supervisor - Operations Member: Manager - Training Member: Supervisor - Maintenance l.

Member: Supervisor - System Engineering l Member: Manager - Safety Review Group Member: Supervisor - Radiation Protection .

j 4 Member: A r t Chri:t Svo=cus .c - Che m rW'i '

Member: Manager - Quality Assurance and Qualitv Control Aiternates '

5.5.1.3 Alternate members shall be appointed in writing by the Plant Review Committee Chairman sto serve on a temporary basis; however, no more than two alternates shall participate in Plant Review Committee activities at any one time.

Meetino Frecuency 5.5.1.4 The Plant Review Committee shall meet at least once per calendar montn and as convened by the Plant Review Committee Chairman.  ;

Ouorum 5.5.1.5 A quorum of the Plant Review Comnittee shall consist of the Chairman

' and four members including alternates.

5-3 Amendment No. 9, I?. 28, 84,115

x. ,

n U-5.9.1 Continued-work and job functions,2# e.g. , reactor operations and surveil-lance,' inservice inspection, routine maintenance, special' ,

maintenance. (describe maintenance), waste processing, and refueling outages. The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements. Small exposures totalling less than 20% of the individual total dose need not be accounted for. In the aggregate,=at least 80% of the total whole body dose received from external sources shall be assigned to specific major work functions.

c. Monthly Doeratino Reoort. Routine reports of operating statis-tics ano snutoown experience shall be submitted on a monthly basis ' to the U. S. Nuclear Regulatory Commission, Document Control Desk, Mail Station P1-137, Washington, D. C. 20555, with a copy to the appropriate Regional Office, to eeerve no later than the '

fifteenth of each month following the calendar month covered by the report. This monthly report shall/also include a statement regarding any challenges or failures to the- pressurizer power j

i operated relief valves or safety valves / occurring during the subject month. j( be-- pe s + n,= c Kec\

5.9.2 Reportable Event l A Licensee Event Report (LER) shall be submitted to the U.S. Nuclear-Regulatory Commission, Document Control Desk, Mail Station P1-137, l Washington, D. C. 20555 with a copy to Region IV of the NRC, within e

30 days after discovery of any event meeting the requirements of

- 10 CFR 50.73.

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3/ This tabulation supplements the requirements of S 20.407 of 10 CFR

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Part 20.

I 5-12 Amenoment No. 9,2#,35,85,99,1.'M L

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ATTACHMENT B 4

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..* Discussion, Justification and No Significant Hazards Consideration Description of Typographical and Administrative Corrections This request for amendment of the Technical Specifications addresses several administrative and typographical errors. A brief discussion of each follows:

Paae viii, TECHNICAL SPECIFICATIONS - FIGURES, TABLE OF CONTENTS The change corrects two column labels by changing'the word " TABLE" to

" FIGURE". The location reference for Figure "l-3" is corrected to read Page "1-6".

Paae 1-2, 1.1, Safety Limits - Reactor Core (continued)

The word " test" in the last paragraph is corrected to read " tests". '

References (1) and (3) are changed to reflect a section of the current version of the Updated Safety Analysis Report that is a more applicable reference.

Paae 2-4, 2.1.2, Heatup and Cooldown Rate (continued), Paragraph (c).

The word "boilup" is changed to read "boltup".

Pace 2-28, 2.5, Steam and Feedwater Systems; The word "about" is changed to read "above".

Paae 2-29, References.

The references to "FSAR" is changed to read "USAR".

Paae 2-34, Electrical Systems

- Analysis of the diesel generator loading scenario determined that the l

16,000 gallons of fuel- oil in the storage tank and the 550 gallons in the '

base tanks was sufficient under the required loading conditions to provide fuel for diesel operation for a minimum period of four days in lieu of the _

seven days that was stated in the Updated Safety Analysis Report (USAR),

l An evaluation in accordance with 10 CFR 50.59 was performed and the USAR l

. was changed to incorporate the new analysis. The changes ,to the bases section of Paragraph 2.7 of the Technical Specifications will insure consistency.between the Technical Specifications and the information given in the.USAR.

Pace 2-36, 2.7, Electrical Systems (continued)

This change is to update the Basis of Section 2.7 of the Technical Specifications to reflect the increased capacityrof the battery chargers  !

that has resulted from system modifications. The battery charger rating l

was changed from "200" amperes to "400" amperes. The change insures that  :

the basis of the Technical Specifications is consistent with information in the USAR.

The references to "FSAR" are changed to read "USAR' and reference (2) is changed to refer to a more applicable section of the USAR.

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%i Pace 2-39, Refueling Operations (continued)

This change corrects a typographical error in the word " concentration".

- References to "FSAR" are changed to "USAR" and reference (2) is changed to refer to a more applicable section of the USAR.

Paae 2-57c, 2.10.4(5) Power Distribution Limits Shculd read "... Above 15Y. pf Rated Power" - typographical error.

Pace 2-59, 2.12(4), Control Room Systems, Control Room Air Treatment This administrative change clarifies the control room air treatment operability requirements by replacing the legalese statement with the words from the standard technical specifications.

= Paae 2-98, Table 2-10, Post-Accident Monitoring Instrumentation Operating Limits This change corrects a typographical error, form "LT-559" to "LT-599", in item 5 of the Table, i Fiaure 2-8,-Flux Peaking Augmentation Factor This change adds the Amendment number which is not legible.

Table 3-1, Minimum Frequency for Checks, Calibrations and Testing of Reactor Protective Systems.

This change adds " Amendment 60", which was omitted from the page.

Paae 3-62a, 3.9 Basis The temperature and operating conditions were changed to agree with the USAR, Section 9.3.1, and accurately reflect the design basis of the shutdown cooling system.

Paae 5-3, 5.5.1.2 i

' The title " Plant Chemist" is changed to " Supervisor - Chemistry" to conform with present station organization.

Paae 5-12, 5.9.2(c)

The word " arrive" is. changed to "be postmarked" to provide a more definitive requirement for submittal of the Monthly Operating Report and i provide sufficient time for preparation and processing of the Monthly Operating Report.

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ATTACHMENT B Page 2 of 3 l

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I Justification:

2 The Commission has provided guidance concerning the application of the standards for determining whether a si  !

exists by providing certain examples (gnificant hazards 48 FR 14870) consideration of amendments that are considered not likely to involve significant hazards i consideration.- Example (i) relates to a purely administrative change to the Technical Specifications: "for example, a change to achieve consistency throughout the Technical Specifications, correction of an

-error, or change in nomenclature." The proposed chan to Example (i) and are purely administrative changes.ges are similar g Basis for No Significant Hazards Consideration:

This proposed change does not involve significant hazards i consideration because operation of Fort Calhoun Station in accordance l with this change would not:

1. Involve a significant increase in the probability or

- consequence of an accident previously evaluated, This change contains only administrative and typographical corrections.

This change should eliminate confusion and thus decrease the probability of human errors associated with accidents previously evaluated.

2. Create the possibility of a new or different kind of accident from any previously evaluated. This change contains only administrative and typographical corrections. No new or different modes of operation are proposed for the ' plant.

3. Involve a significant reduction in the margin of safety. -

This change contains only administrative and typographical corrections and, as such, does not result in a decrease in the margin of- safety.

Therefore, based on the above considerations, Omaha Public Power District has determined that this change does not involve a significant hazards consideration.

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UNITED STATES NUCLEAR REGULATORY COMMISSION L ,

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October 14, 1988 35'-553 Occkt-t f:o. E0-E85

' Mr. Kenneth J. Morris Divisien Manager - fluclear Operations -j Omaha Public Power Distrir.t i 1623 Harney Street i Omaha,1:ebraska 6810C 1 '

Ocar ftr. Itorris:

SUEJECT: REQUEST FOR CHAliGE T0 BASIS OF FOR1 CALH0Utt STATION TECHf41 CAL i SPECIF1 CAT 10t! ?.7, ELECTRICAL SYSTEMS j

Pursuant to your application for an acendn:ent to modify the Fort Calheur; _

Station Technical Specifications (TS), the Conmiission has determined that an antencment is not needed for this change- to the Basis of TS E.7. Changes only te a' Easis can be made without prior f!EC appreval af ter the completion of c 10 CFR 50.59 Safety Evaluation by the -licensee. The actual correction to the page in the TS'can be included in a subsequent amendment application.  !

- The amount of the applicaticn fee will be credited to your account.

Sincerely, 13 ,

i Patrick D. Milano, Project flanager Project Directerate - IV Divisien of Reactor Projects - III,  ;

IV V and Special Projects '

Office of Nuclear Reactor Regulation cc: See next page l

d ' WCJ, M0G, RLA, KJM, KCH, RKS, RCK, JJF, JMW, BRH, FCS, JBK, DJM, FILE, JRG, RB, OKD,  ;

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. TECHNICAL SPECIFICATIONS - FIGURES

. .T.A.B.L.E. .O.F C.O.NTENTS I 3 Gt O RJE

~TAOLE- PAGE WHICH DESCRIPTION FlciutEhnste. F_otl.bwS 1-1

.THLP Safety Limits 4 Pump Operations. . . . . . . . . . . . . 1-3  : i 1-2 Axial Power Distribution LSSS for 4 Pump Operation. . . . . . 1-3 i i 1-3 i TMLP LSSS 4 Pump Operation. . . . . ..

<f'-6 2-1A RCS Press-Temp Limits Heatup . . . . . . . . . . . . . . . . 2-7a j 2-1B RCS Press-Temp Limits Cooldown. . . . . . . . . . . . . . . .  ?-7a 1  !

2-3 Predicted Radiation Induced NDTT Shift. .......... . 2-7a 2-10 Spent Fuel Pool Region 2 Storage Criteria . . . . . . . . . .  ?-38a j 2 PDIL. . . . . . . . . . . . . . . . . . . . . . . . . . . . .

2-57e 2-5 Allcwable Peak Linear Heat Rate vs Burnup . . . . . . . . . .  ?-57e 1

2-6 LCO for Excore Monitoring of LHP. . . . . . . . . . . . . . .  ?-57e 2-7_

LCO for Dil0 Monitoring. . . . . . . . . . . . . . . . . . . . 2-57e t:

2-8 Flux Peaking Augmentation Factors . . . . . . . . . . . . . .  ?-57e T l 2 i F'TA, Fyy anc Core Power Limitations. . . . . . . . . . . . .  ?-57e I

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C 1.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS -

1.1 Safety Limits - Reactor Core (Continued)  ;

would cause DNB at a particular core location to the actual heat flux at that location, is indicative of the margin to DNB. The minimum value of the DNBR during steady state opera-tion, normal operational transients, and anticipated tran-sients is limited to 1.18. A DNBR of 1.18 corresponds to a 95% probability at a 95% confidence level that DNB will not  ;

occur, which is considered an appropriate margin to DNB for all operating conditions.(1)

The curves of Figure 1-1 represent the loci of points or re-actor thermal power (either neutron flux instruments or AT in-struments), reactor coolant system pressure, and cold leg temperature for which the CNBR is 1.18. The area of safe opera-  !

tion is below these lines.

The reactor core safety limits are based on radial peaks limit-ed 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 of 1.80. l The LSSS in Figure 1-3 is based on the assumption that the un-rodded integrated total radial peak-(F ) is 1.80. This peak-

, ing factor is slightly higher (more co servative) that the maximum predicted unrodded total radial peak during core life, excluding measurement uncertainty.

Flow maldistribution effects for operation under less than full reactor coolant flow have oeen evaluated via model fo m 7 ;t::t.(2) Yhe flow model data established the maldistribution

• factors and hot channel inlet temperature for the thermal l

analyses that were used to establish the safe operating enve--

lopes presented in Figure 1-1. The reactor protective system is designed to prevent any ~an;icipated combination of tran-sient conditions for reactor :colant system temperature, pres-sure, and thermal power leve', that would result in a DNBR of less than 1.18.(3)

References (1) USAR, Seetion S.6.7 'i!> . 2 . 'E . E (2) USAR, Section 1.4.6 (3) USAR, Section 4r6 d B.1.1 5 1

1-2 Amendment No. 8,32,43,47, 70,77,92,117 e

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-2.0 LIMITING CONDITIONS FOR OPERATION. '

2.1 Reactor Coolant System (Continued) 2.1.2 Heatup and Cooldown Rate (Continued)

(a) The curve in Figure 2-3 shall be used to predict the increase in transition temperature based on integrated fast neutron flux. If measurements on the irradiation specimens indicate a deviation from this curve, a new  ;

curve shall be constructed.

(b) The limit line on the figures shall be updated for a new integrated power period as follows: the total integrated reactor thermal power from startup to the end of the new period shall be converted to an equivalent integrated fast neutron exposure (Ei:1 MeV), For this plant, based upon surveiilance materials tests, weld chemical composition data, and the effect of a reduced vessel fluence rate provided by core load designs beginning with fuel Cycle 8, the predicted surface fluence at the initial reactor vessel beltline weld material for 40 years at 1500 MWt and an 80%

load factor is 2.55x1019 n7cm2 The flux reduction applied to the fluence calculations was base'd on Cycle 1-7 average and Cycle 8 average azimuthal flux distribution plots generated used 00T 4.3. The predicted transition temperature shift to the end of the new period shall then be obtained from Figure 2-3.

(c) The limit lines in Figures 2-1A and 2-1B shall be moved parallel to the temperature axis (horizontal) in the ,

direction of increasing temperature a distance equivalent to the transition temperature shift during the period since the curves were last constructed. The teiM;' temperature limit line shall remain at 82 F as it is set by the NDTT of the reactor vessel flange and not subject to fast neutron flux. The lowest service temperature shall remain at 182 F because components related to this temperature are also not subject to fast neutron flux.

(d) The Technical Specification 2.3(3) shall be revised each time the curves of Figures 2-1A and 2-18 are revised.

Basis bdlY. tap-All components in the reactor coolant system are designed to withstand the effects of cyclicgads due to reactor coolant system temperature and pressure changes These cyci'ic loads are introduced by normal unit load transients, reactor trips and startup and shutdown operation.

During unit startup and shutdown, the rates of temperature and pressure changes are limited. The design number of cycles for heatup and cool-down is based upon a rate of 100 F in any one hour period and for cyclic operation.

2-4 Amendment No. 22,47,64,74,77,200, na

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25 Steam and Feedvater Systems AJplicability Applics to the operating status of the steam and feedvater systems.

Objective To define certain conditions for the steam and feedvater system ,

necessary to assure adequate decay heat removal.

l

[~ Specifications

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a bo v e The reactor coolant shall not be heated about 300 F unless the following conditions are met:

(1) Eoth auxiliary feedvater pumps are operable. One of the i auxiliary feedvater pumps may be inoperable for 2h hours provided that the redundant component shall be tested to l demonstrate operability, t (2) A minimum of 55,000 gallons of water in the emergency I feedvater storage tank and a backup water supply to the emergency feedvater storage tank from the Missouri River by the fire water system.

(3)

All valves, interlocks and piping associated with the i above components required to function during accident j conditions are operable. Manual valves that could inter- l rupt auxiliary feedvater flow to the steam generators  !

shall be locked in the required position to ensure a  !

flev path to the steam generators.

(L) 3.e main steam stop valves are operable and capable of closing in four seconds or less under no-flow conditions. 3 Easis A reactor shutdown from power requires a removal of core decay heat. Immediate decay heat removal requirements are normally satisfied by the steam bypass to the condenser. Therefore, core decay heat can be continuously dissipated via the steam bypass to the condenser as long as feedvater to the steam generator is available. Normally, the capability to supply feedvater to the steam generators is provided by operation of the turbine cycle feedvater system. In the unlikely event of complete loss of electrical power to the station, decay heat removal is by steam discharge to the atmosphere via the main steam safety and atmospheric dump valves. Either auxi-liary feedvater pump can supply sufficient feedvater for re-moval of decay heat from the plant. The minimum amount of water in the emergency feedvater storage tank is the amount needed for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of such operation. The tank can be re-supplied with water from the fire protection system.(1)

Amen:i~.ent No. 49 2-28 h

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,(. - 2.0 LIMITING CONDITIONS-FOR OPERATION

\:- 2.5 steam and reedvater Systems .(continued)

A closure time. of h~ seconds for the main steam stop valves is considered adequate and was selected-as being consistent with expected-response time -for. 2{njtg. entation as detailed in the steam line break incident analysis.

gferences u

(1) / SAR, Section 9.k.6

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(2) /SAR, Section 10.3 U

(3) /SAR,Section14.12 l

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f Lt. 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 E!ectrical Svstems (Continuec)-

from either one of two diesel generators and off-site standby power via tne unit auxiliary transformers.(1)

The two emergency diesel generators on site do not require outside power for start up or operation.

. Upon loss of normal and standby power sources, the 4.16 kV buses 1A3 and 1A4 are energi:ed from the diesel generators. Bus load shedding, trans-fer to the diesel generator and pickup of critical loads are carried out automatically.(2)

When the turbine generator is.out of service for an extended period, the

' generator can be isolated by opening motor operated disconnect switch ,

-DS-T1 in the bus between the generator and the main transformer, allow- i ing the main transformer and the unit auxiliary power transformers to be returned to service.(3)

Equip =ent served by 4.16 kV and 480 V auxiliary buses and MCC's is arranged so that loss of an entire 4.16 kV bus does not compromise safety of the plant during DBA conditions. For example, if 4.16 kV bus 1A3 is  ;

lost, two raw water pumps, one low pressure safety injection pump, one -'

high pressure safety injection pump, one auxiliary feedwater pump, two f- component cooling water pumps, two containment spray pumps and two con-tainment air fans are lost. This leaves two raw water pumps, one low pressure safety injection pump, two high pressure safety injection pumps, one component cooling water pump, one containment spray pump and two con-tainment air fans which is more than sufficient to control containment 3 pressure below the design value during the DBA.

assu res I(OnE) hear j Thetotalfueloilenginebasetankcapacityof550gallonsoneach I diesel :; I;nsidcr;d'more than cc:quc.c ;in:2 :pp r _ " -' r ' ' - * "^ure running time (worst case loading) is available before transfer of fuel oil from the 18,000 gallon underground storage tanks is mandatory. Two 13 gpm diesel oil transfer pumps per diesel, with each being fed f rom the diesel it is associated with, are available for transferring fuel  ;

oil from the storage tank to the day tanks. The 16,000 gallons in the storage tank in addition to the SS35YInks will provide diesel operation under the required loading conditions for a minimum period of M s%

snould only one diesel be in operation. It is considered incredible not to be able to  :  :: f {,gilfromoneofseveralsources in the vicinity of Omana in less than davs under the worst of weather conditions.

pro cu re, cNA che.M uer One battery charger on each battery shall be operating so that the batteries will always be at full charge; this ensures that adequate d-c power will be avilable for all emergency uses. Each battery has one battery charger permanently connected with a third charger capable of being connected to either battery bus. The cnargers are each rated Amenc=ent No. 76 2-35

s 2.0 I,IMITING CONDITIONS FOR OPERATION

( 2.7 Electrical Systems (Continued)

Aco for49& amperes at 130 volts. Except for the first minute following a DBA during which the batteries and the charger accommodate all the load, the capacity of the battery charger vill handle all required loads. .Each of the reactor protective system channels instrumentation channels is supplied by one of the a-c instrument buses. The removal of one of the '

a-c instrument buses is permitted as the 2-of b logic may be manually changed to a 2-of-3 losic vithout compromising safety.

The engineered safeguards instrument channels use a-c instrument buses (one redundant bus for each channel) and d-c buses (one redundant bus for each logic circuit). The removal of one of the a-c instrument buses is permitted as the two of four logic automatically becomes a two of three logic.

Eeferences id (1) /SAR, Section 8.3.1.2 0 7. 3 .'4,7 (2) J'SAR, Section W CA (3) /SAR, Section 8.2.2 l

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2.0 LIMITING ~ CONDITIONS FOR OPERATION - l 2.8 Refuelino coerations (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 l made in core geometry one flux monitor is sufficient. This permits i

maintenance of the instrumentation. . Continuous monitoring of radiation '

levels and neutron flux provides immediate indication of an unsafe condi- <

tion. The shutdown cooling pump is used to maintain a unifom boron '

concentration. .

1 The shutdown margin as indicated will keep the core suberitical even if 1 all CEA's were withdrawn from the core. During refueling operations, the reactor refueling cavity is filled with approximately 250,000 gallons of l borated water. The boron concentraltion of this water (at least 1800 ppm boron) is sufficient to maintain the reactor suberitical by more than 5%,

, including allowance for uncertainties, in the cold condition with all rods withdrawn.(2) Periodic checks of refueling water boron concentration ensure the proper shutdown margin. Communication requirements allow the control room operator to inform the refueling machine operator of any impending unsafe condition detected from the main control 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 <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after the power has been removed from the core takes advantage of the decay-of the short half-life fission products and allows l

for any failed fuel to purge itself of fission gases, thus reducing the consequences of fuel handling accident.

l_ 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 l will be closed manually. In addition, the exhaust ventilation ductwork I 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 u

l (1) JSAR, Section 9.5 (2) ISAR, Section 0. 5.1. 2 6:" 7 ' 3 . l l I4 i

2-39 Amendment No. 2#,75,203,117 l

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2. 0 LIMITING CONDITIONS FOR OPERATION 2.10 Reactor Core (Continueo) ,

2.10.4 Power Distribution Limits (Continued) -

(5) DNBR Marcin Ouring Power Ooeration Above IS% Rated Power ,

(a) The following DNS related parameters shall be maintained within the. limits shown:

(i) Cold Leg Temperature 5543'F*

(ii) Pressurizer Pressure 22075 psia *

(iii) Reactor Coolant Flow 2197,000 gpm**

(iv) Axial Shape Index, Y y 5 Figure 2-7""

(b) With any of the above parameters exceeding the limit, restore

-the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce

. power to less than IS% of rated power within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Basis The limitation on linear heat rate ensures that in the event of a LOCA, the ,

peak temperature of the fuel cladding will not exceed 2200*F.

i Either of the two core power distribution monitoring systems, the Excore Detec-i ter 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 l and verifying that the axial shape 'index is maintained within the allowable limits of Figure 2-6 'as adjusted by Specification 2.10.4(1)(c) for the allowed l linear heat rate of Figure 2-5, RC Pump configuration, and F T of Figure 2-9.

i In conjunction with the use of the excore monitoring system Ed in establishing l the axial shape index limits, the following assumptions are made: (1) the CEA

l. 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).

" 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 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.

• • • The AXIAL SHAPE INDEX.C Core power shall be maintained within the limits .

l: established by the Better Axial Shape Selection System (BASSS) for CEA l- insertion of the lead bank of < 6S% when BASSS is operable, or within L the limits of Figure 2-7.

2-57c Amendment No. 32,43,57,75, 77,92,I09,117

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LIMITING CONDITIONS FOR OPERATION 2.12 Control Room Systems Atelicability Applies to the control room air conditioning and filterin6 systems.

Objective

-- To limit the environmental conditions in the contrcl room, under normal and post DBA conditicos.

Soecifications (1) I:' the control room air temperature reachen 1200F, immediate

!J action shall be taken to reduce this temperature. If the temperature cannot be reduced to below 1200F in four hours ,

the reactor vill be placed in a hot shutdown condition.

(2) A thermemeter must be in the control roce at all times.

(3) All areas of the plant which have safety related instrumentation vill be observed during het functional testing to determine local temperatures and monitored during operatien if normal plent ventilatien is not available.

(h) 4 - 2. . af te; Lc da c that A m wl . cc . au u c at:w'

~~~ system .. -ade or found to be inoperable for an' - eca, m Yeactor cperation . , , . -

able on ' ..g-the succeedin,;

P even days unless such 4* b 1s o -

e operable. If hece condi" r v nnot'be met, the reactor sha - 'aceo 4 e + c;nditica .fitM , gh v -

3 asis The reactor protective system and the engineered safeguards system vere designed for and the instrumentation was tested at 1200F.

Therefore , if the temperature of the control r00m exceeds 120 0F, the reactor vill be shutdown and the conditions corrected to preclude i

failure of ecmponents in an untested environment, If the centrol recm air treatment system is found to be inoperable ,

there is no immediate threat to the control room and reactor opera-tion may continue for a limited period of time while repairs are being l made. If the system cannot be repaired within seven (7) days, the reactor is shutdown and brought to cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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i TABLE 2-10 Post-Accident Monitoring Instrumentation Operatino Li...iw Minimum Operable Instrument Channels Action

1. Containment Wide Range Radiation 2 (a)

Monitors (RM-091A & B)

2. Wide Range Noble Gas Stack Monitor RM-063L Noble Gas Portion Only 1 a' -

RM-063M Noble Gas Portion Only 1 a >

RM-063H Noble Gas Portion Only 1 a

3. Main Steam Line Radiation Monitor 1 (a)

(RM-064)-

4 Centainment Hydrogen Monitor (VA-81A & B) ,2 (b)(c)

5. Containment, Water Lev _e.L --

Narrow Range (LT-9Pf & LT-600) 1 Wide Range (LT-387 & LT-388) 2 (d)(c)

(b)  ;

6. Containment Wide Range Pressure 2 (b)(c)

(e)(f)

7. Reactor Coolant System Subcooled Margin Monitor 2

8. -Core Exit Thermocouples (1) 2/ Core Quadrant (g)(h)

9. Reactor Yessel Level (HJTC) (j) 2 (k)(1)

(a) With the number of OPERABLE channels less than required by the minimum channels operable requirements, initiate the pre-planned alternate method of monitoring the appropriate parameter (s) within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and

1. either restore the inoperable channel (s) to OPERABLE status within 7 days of the event, or

2. prepare and submit a special report to the Connission pursuant to

- specification 5.9.3 within 14 days following the event outlining the action taken, the cause of the inoperability, and the plans and schedules for restoring the system to OPERABLE status.

(b) With one channel inoperable, restore the inoperable monitor to OPERABLE status within 30 days or be in at least H0T SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

2-98 Amendment No. SJ, E2, E3, 87,110 i

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I!eutron Ilonitors readings.

b. ' Test,( 3) .P -b. - Int,ernal test signals to

-veriry SUR indication:and' trip,. power level permissives, instrument accuracy.'

3. henetor Coolunt Plow- u. Check .S a. Comparison of four . separate total flow indications.

b. Calibrate ; R . b.' . Known 'di fferential prassure

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3.0' SURVEILLANCE REOUIREMENTS 3.9 Auxiliary Feedwater System (Continued)-

Basis

, The valve position verifications performed monthly and following auxiliary feedwater system maintenance will confirm the avail-ability of an auxiliary feedwater flow path to the steam generators.

The testing of tne auxiliary feedwater pumps every month and after cold shutdowns will-verify their operability by recirculating water to the emergency feedwater storage tank.

Operating the regulating valves (HCV-1107A, HCV-1107B, HCV-1108A and HCV-110SB) one at a time every three months and af ter cold shutdowns will confirm a flow path to the steam generators and operability of the valves. ,

Proper functioning of the steam turbine admission valve and starting of the feedwater pump will demonstrate the integrity of the steam i driven pump. Verification of correct operation will be made both j from instrumentation within the main centrol room and direct visual  !

observation of the pumps, j

-- g o g o f; i The operability of the auxiliary feedwater system ensures

• hat the ,

reactor coolant system can be cooled down to less than L F from l

~-normal operating conditione the event of a total loss of off-site l power. 9 'i References k a T w L c.h im . th = 6 kurdow n c. cole3-s oTem m a y b e. plse ek mTo opers'Teau ;_

(1) FSAR, Section 9.4.

(2) Technical Specification 2.5 ,

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f3 5.0 ADMINISTRATIVE CONTROLS .-

5.4 Training-5.4.1 A retrainir.g and replacement training program for the plant staff shall be maintained under the direction of the Manager - Training and l shall meet or exceed the requirements of Section 5.5 of ANSI N18.1-1971 and 10 CFR Part 55. {

l 5.4.2 A training program for the fire brigade shall be maintained under the Manager - Training and shall meet or exceed the requirements of l Section 27 of NFPA Code-1975, except that the meeting frequently may be quarterly.

5.5 Review and Audit 5.5.1 Plant Revier Committee (PRC)

. Function 5.5.1.1 The Plant Review Comittee shall function to advise the Manager - Fort Calhcun Station on all matters related to nuclear safety.

Composition 5.5.1.2 The official Plant Review Committee shall be composed of the:

Chairman: Manager - Fort Calhoun Station Member: Supervisor - Operations Member: Manager - Training l Member: Supervisor - Maintenance Member: Supervisor - System Engineering l Member: Manager - Safety Review Group j Member: Supervisor - Radiation Protection ,

i Member: l = t Chc- St- S v ee e-m e.o r - C hew s SW'i '

Member: Manager - Quality' Assurance and Qualitv Control  :

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Alternates 5.5.1.3 Alternate members shall be appointed in writing by the Plant Review Comittee Chairman to serve on a temporary basis; however, no more than two alternates shall participate in Plant Review Committee activities at any one time.

Meeting Frecuency 5.5.1.4 The Plant Review Committee shall meet at least once per calendar month and as convened by the Plant Review Comittee Chainnan.

Ouorum 5.5.1.5 ~ A quorum of the Plant Review Comittee shall consist of the Chairman and four members including alternates.

5-3 Amendment Mo, 9, 19, 28, U ,115

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5.9.1 Continued work and job functions,S e.g., reactor operations and surveil-lance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling outages. The dose assignment to various -duty functions may be estimates based on pocket dosimeter. TLD, or film badge measurements. Small exposures totalling less than 20% of the individual total dose need not be accounted for. In tho aggregate, at least 80% of the total whole body dose .*eceived from external sources shall be assigned to specific major work functions, i

c. Monthly Ooeratino Report. Routine reports of operating statis-tics ano snutcown experience shell be submitted on a monthly basis to the U. S. Nuclear Regulatory Commission, Documant Control Desk, Mail Station P1-137, Washington, D. C. 20555, witi; a copy totheappropriateRegionalOffice,toeeefvenolaterthanthe i fifteenth of each month following the cal ndar month covered by the report. This monthly report shal also include a statement  ;

regarding any challenges or failures to he pressurizer power  !

operated relief valves or safety valves occurring during the subject month, b e. y s Fm a" Vec\

'5.9.2 Reportable Event A Licensee Event Report (LER) shall be submitted to the U.S. Nuclear

' Regulatory Commission, Document Control Desk, Mail Station P1-137, l Washington, D. C. 20555 with a copy to Region IV of the NRC, within e 30 days after discovery of any event meeting the requirements of 10 CFR 50.73.

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3/ This tabulation supplements the requirements of S 20.407 of 10 CFR Part 20.

1 1

5-12 Amendment No. 9,24,35,85,99,119 (Next page is 5-15) 1

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ATTACHMENT B l'

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e o Discussion, Justification and No Significant Hazards Consideration Description of Typographical and Administrative Corrections This request for amendment of the Technical Specifications addresses several administrative and typographical errors. A brief discussion of each follows:

Paae viii, TECHNICAL SPECIFICATIONS - FIGURES, TABLE OF CONTENTS The change corrects two column labels by changing'the word " TABLE" to <

" FIGURE". The location reference for Figure "1-3" is corrected to read Page "1-6".

Pace 1-2, 1.1, Safety Limits - Reactor Core (continued)

The word " test" in the last paragraph is corrected to read " tests", i References (1) and (3) are changed to reflect a section of the current version of the Updated Safety Analysis Report that is a more applicable reference.

Pace 2-4, 2.1.2, Heatup and Cooldown Rate (continued), Paragraph (c).

The word "boilup" is changed to read "boltup".

Pace 2-28, 2.5, Steam and Feedwater Systems; The word "about" is changed to read "above".

Paae 2-29, References.

The references to "FSAR" is changed to read "USAR".

Paae'2-34, Electrical Systems Analysis of the diesel generator loading scenario determined that the 16,000 gallons of fuel oil in the storage tank and the 550 gallons in the base tanks was sufficient under the required loading conditions to provide fuel for diesel- operation for a minimum period of four days in lieu of the seven days that was stated in the Updated Safety Analysis Report (USAR).

An evaluation in accordance with 10 CFR 50.59 was performed and the USAR was changed to incorporate the new analysis. The changes.to the bases section of Paragraph 2.7 of the Technical Specifications will insure consistency between .the Technical Specifications and the information given in the USAR.

Paae 2-36, 2.7, Electrical Systems (continued)

This change is to update the Basis of Section 2.7 of the Technical Specifications to reflect the increased capacity of the battery chargers that has resulted from system modifications. The battery charger rating was changed from "200" amperos to "400" amperes. The change insures that '

the basis of the Technical Specifications is consistent with information in the USAR.

The references to "FSAR" are changed to read "USAR' and reference (2) is changed to refer to a more applicable section of the USAR.

ATTACHMENT B Page 1 0F 3

V =c.

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Paae 2-39, Refueling Operations (continued)

This change corrects a typographical error in the word " concentration".

References to "FSAR" are changed to "USAR" and reference (2) is changed to refer.to a more applicable section of the USAR.

Paae 2-57c, 2.10.4(5) Power Distribution Limits Should read ". . . Above 15% sf Rated Power" - typographical error.

Paae 2-59, 2.12(4), Control Room Systems, Control Room Air Treatment This administrative change clarifies the control room air treatment operability requirements by replacing the legalese statement with the words from the standard technical specifications.

Paae 2-98, Table 2-10, Post-Accident Monitoring Instrumentation Operating Limits This change corrects a typographical error, form "LT-559" to "LT-599", in item 5 of the Table.

Fiaure 2-8, Flux Peaking Augmentation Factor This change adds the Amendment number which is not legible.

Table 3-1, Minimum Frequency for Checks, Calibrations and Testing of Reactor Protective Systems.

This change adds " Amendment 60", which was omitted from the page.

! Pace 3-62a, 3.9 Basis The temperature and operating conditions were changed to agree with the USAR, Section 9.3.1, and accurately reflect the design basis of the l shutdown cooling system, i

l~ Pace 5-3, 5.5.1.2 The title " Plant Chemist" is changed to " Supervisor - Chemistry" to conform with present station organization.

Pace 5-12, 5.9.2(c)

The word " arrive" is changed to "be postmarked" to provide a more definitive requirement for submittal of the Monthly Operating Report and provide sufficient time for preparation and processing of the Monthly Operating Report.

ATTACHMENT B Page 2 of 3

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Justification:.

The Commission has provided guidance concerning the application of the standards for determining whether a significant hazards consideration-exists by providing certain examples (48 FR 14870) of amendments that are considered not likely to involve significant hazards consideration. Example (i) relates to a purely administrative change i to the Technical Specifications: "for example, a change to achieve consistency throughout the Technical Specifications, correction of an error, or change in nomenclature." The proposed changes are similar j to Example (i) and are purely administrative changes.

Basis for No Significant Hazards. Consideration: l This proposed change does not involve significant hazards consideration because operation of Fort Calhoun Station in accordance  ;

with this change would not: l

1. Involve a significant increase in the probability or consequence of an accident previously evaluated. This change contains only administrative and typographical corrections.

This change should eliminate confusion and thus decrease the probability of human errors associated with accidents previously evaluated.

2.. Create the possibility of a new or different kind of accident from any previously evaluated. This change contains only administrative and typographical corrections. No new or different modes of operation are proposed for the plant.

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

This change contains only administrative and typographical corrections and, as such, does not result in a decrease in the margin of safety.

Therefore,. based on the above considerations, Omaha Public Pcwer District has determined that this change does not involve a significant hazards consideration.

ATTACHMENT B Page 3 of 3

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UNITED STATES e

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- NUCLEAR REGULATORY COMMISSION

j WASHINGTON. D C. 20555

%**..*./ October 14,1988 T B'- 35 3 ,

Dockt-t f o. E0-785 Mr. Kenneth J. Morris Divisien Manager - flucicar Operations Omaha Public Power District 1623 Harney Street Omaha,f:ebraska 68100

Dear Mr. Morris:

q.--

SUNECT: REQUEST FOR CHAliGE TO BASIS OF FOR1 CALH00ft STATION TECHNICAL SPECIFICATI0t! ?.7,- ELECTRICAL SYSTEMS Pursuant to your application for an amendment to modify the Fort Calhoun Station Technical Specifications (TS), the Coramission has determined that an amendment is not needed for this change to the Basis of TS 2.7. Changes only to a Easis can be made without prior !!DC 6pproval af ter the completion of a 10 CFR 50.59 Safety Evaluation by the licensee. The actual ~ correction to the page.in the TS can be included in a subscquent amendment application.

The amount of the applicatien fee will be credited to your account.

Sincerely, 3 --

Patrick D. Milano, Project flanager Project Directorate - IV Divisien of Reactor Projects - III, "

IV. V and Special Projects-Office of Nuclear Reactor Regulation cc: See next page

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