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Amend 125 to License DPR-40,revising Tech Specs to Place Operability & Surveillance Requirements on Alternate Shutdown Panels
	ML20033F909
Person / Time
Site:	Fort Calhoun
Issue date:	03/19/1990
From:	Hebdon F; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20033F905	List: ML20033F904; ML20033F909; ML20033F912;
References
	NUDOCS 9004040065
	Download: ML20033F909 (11)
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OMAHA PlfBLIC POWER DISTRICT DOCKET NO. 50-785 FORT CALH0liN STATION LtN11 NO._)

i AMENDMENT T0 rACILITY OPERAT_ING LICENSE Amendnent No.125 License No. DPR-40 i

1.

The Nuclear Regulatory Connission (the Connission) has found that A.

The application for amendment by the On.aha Public Power District (thelicensee)datedOctober 27, 1989 as supplenented January 11, 1980, complies with the standards and requirenents of the Atomic Energy Act of 1954, as acended (the Act), and the Connission's rules and regulations set forth in 10 CFR Chapter I;

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B.

The facility will optrate in conformity with the application, as amended, the provisions of the Act, and the rules and regulatient.

of the Connission; C.

There is reasonable assurance:

(i)thattheactivitiesauthorized by this anendnent can be conductsd without endogering the health

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and safety of the public, and (ii) that such activities will be conducted in comp'tionce with the Corrission's regulations;

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D.

The issuance of t515 licenn anensent will not be inimical tu the i

corron defense and security or to the health and safety of the public; and t

i E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Connission's regulations and all applicable requirenents have been satisfied.

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t 9004040065 900319 FDR ADOC K 0500028".

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Accordir. gly, f acility Operating License No. DFR-40 is amended by changes to the Technical $pecifications as indicated in the attachnent to this i

license amendment, and paragraph 3.B. of Facility Operating License No.

DPR-40 is hereby anended to read as follows:

B.

Technical Specifications l

The Technical Specifications contained in Appendix A, as revised through Amendment No. 125, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

The license anendment is effective as of the date prior to going critical l

af ter the Cycle 13 refueling is completed.

FOR THE NUCLEAR REGULATORY COMMIS$10N e? Y

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Frederick J. Hebdon, Director Project Directorate IV Division of F.eactor Projects - !!!,

IV, V end Special Projects Office of Nuclear Reactor Regulation Attachn(nt:

Cherses to the Technical Sp cificatiens Octe of Itsuance: March 19, 1990 F

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r ATTACHMENT TO,J.!C{NS{, AMjNDMENT N0 l 5, ul FACILITY OPERATING LICENSE NO. DPR-40 DOCKET NO. 50-285 Revise Appendix "A' Technical Specifications as indicated below. The revised paps are identified by amendment number and contain vertical lines indicating the area of change.

Remove Pages Insert Pages y

v vi vi 2-66 2-66

2. f>6 a 2-66a 2-66b 3-1 3-1 3-16d 3-16e

TECHNICALSPECJ[JCATIONS_-TABg1 TABLE OF CONTENTS ititE DE$CRIPTION PAtti 3-3 Minimum Frequencies for Checks, Calibrations, and Testing of Miscellaneous Instrumentation and Controls........

3-13 3-14 3-15 3-16 3 16e 3-16b 3-16c 3 3a Minimum frequency for Checks, Calibrations and functional TestingofAlternateShutdownPanels(Al-185andAI-212)and Emergency Auxiliary Feedwater Panel (Al-179) Instrumentation and Control Circuits.....................

3-16d 3-16e 3-4 Mininum frequencies for Sampling Tests............

3-18 3-19 35 Minimum Frequencies for Equipment Tests...........

3-20 3 20a 3-20b 3-20c 3-20d 3-6 Reactor Coolant Pump Surveillance.......,,,.....

3-27 3-7 Capsule Removal Schedule...................

3-27 3-9 Radiological Environmer.t Monitoring Program.........

3-66 3-67 3 11 Radioactive Liquid Waste Sampling and Analysis........

3-72 3-73 3-12 Radioactive Gaseous Waste Sampling and Analysis.......

3-74 3-75 3 13 Steam Generator Tube Inspection...............

3-90 5.2 1 Minimum Shif t Crew Composition................

5-2 y

Amendnent No. 115,125

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I T[CHNICAL $PECIFICATIONS - TABLES i

TABLE OF Cp0,y{N,TS_lALP_HABETICA{, PRE {R)

TAPLE DESCRIPTION PA!E i

3-7 Capsule Removal $chedule...................

3 27 i

l 2-1 ESFS Initiation Instrumentation Setting Limits........

2-64 2-64a 1

2-7 Fire Detection Zones.....................

2-90 2-0 Fire Hose Station Locations.................

2 94 l

2 95 j

27 Halen Area Fire Zones....................

2 90s I

i i

2-4 Instrument Operating Conditions for Isolation functions...

2 69 2 69a 2-2 Instrument Operating Requirements for RPS..........

2 67 2 67a l

2-3 Instrurrent Operating Requirements for En Sefety features............ gineered i

2 68 2-68a t

2 68b i

s 2-5 Instrumentation Operating Requirements for Other safet t

Features functions.................. y 2-70 3-3a

!;inimum frequency for Checks, Calibrations and Functional Testing of Alternate shutdown Panels ('Al-185 and AI-212) and l

Emergency Auxiliary Teedwater Panel (Al-179) Instrumentetion and Control Circuits.....................

3-16d 3-16e l

3-2 t'inir.um Frequencies for Checks, Calibrations and Testing of Engineered safety features, Instrumentation and Controls...

3-7 j

3-8 f

3-9 3-10 3 11 l

3-12 l

3-12a l

3-3 Minimum Frequencies for Checks, Calibrations, and Testing of Miscellaneous Instrumentation and Controls........

3 13 3-14 3-15 i

3-16 3-16a 3-16b 316c 3-1 Minimum frequencies for Checks, Calibrations.

and Testing of RPS......................

3-3 3-4 3-5 3-6 vi Amendment No. 115,125

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2.0 LIMITING CONDITIONS FOR OPERATION 2.15 instrunntat%~n'aWtoW*ol 5ystses (Continued) ventilation isolation signals available if the containment ventila-tion isolation valves are closed.

If after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from time of initiating a hot shutdown procedure the inopernble engineered safety features or isolation functions channel has not been stored to operable status, the reactor shall be placed in a cold shutdown condition within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This specification applies to the hi above10'ghratetripwiderangelogchannelwhentheplantisator 5 power and is operating below ISE of rated power.

(3)

In the event the nuidber of channels of a particular s stem in service falls below the limits given in the columns entitled { Minimum Operable Channels' of

Minimum Degree of Redundancy", except as conditioned by the column entitled ' Permissible Bypass Conditions", the reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ' however, opera-tioncancontinuewithoutcontainmentventilationisolationsignals available if the ventilation isolation valves are closed.

If minimum conditions for engineered safety features or isolation functions are not met within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from time of discovering loss of operability, the reactor shall be placed in a cold shutdown condition within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

If the number of operable high rate trip-wide range log channels falls below that given in the column entitled

Minimum perable Channels

in Table 2 2 and the reactor is at or above10'g%powerandatorbelow155ofratedpower,reactorcritical cperttion shall be discontinued and the

)lant placed in an operational modr allowirg repair of the inoperable ciannels befere startup or reactor critical cperation may proceed.

If, during power operation, the rod block function of the secondary CEA position indication system and rod block circuit are inoperable for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or the plant computer PDIL alarm CEA group deviation alarm and the CEA sequencing function are inoperable for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , the CEAs shall be withdrawn and maintained at fully withdrawn and the control rod drive system mode switch shall be traintained in the off position except when manual motion of CEA i

Group 4 is required to control axial power distribution.

l (4) In the event that any of the following Alternate Shutdown Panel instrurnentation or control circuits become inoperable, either restore theinoperablecomponent(s)tooperablestatuswithinsevendays,or l'

be in hot shutdown within the next twelve hours. This specification is applicable in Modes 1 and 2.

I, WideRangeLogarithmicPower(Al-212)

)

SourceRangePower(Al-212)

ReactorCoolantColdLegTemperature(AI-185)

ReactorCoolantHotLegTemperature(Al-185)

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PressurizerLevel(AI-185)

Volume Control Tank Level (AI-185) 2-66 Amendment No. 8,20,f A,f f,SS.125 i

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1 2.0 Lit!! TING CONDITIONS FOR OPERATION i

2.15 TimiiW'eidation and control systems (Continued)

($) In the event that any of the following Emergency Auxiliary Feedwater Panel instrumentation or control circuits becone inoperable either i

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restoretheinoperablecomponent(s)tooperablestatuswithInseven days or be in hot shutdown within the next twelve hours. This i

specIficationisapplicableinModes1and2.

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SteamGeneratorLevel,WideRange(Al-179)

J SteamGeneratorLevel,HarrowRange(A!-179)

SteamGeneratorPressure(Al-179)

PressurizerPressure(Al-179) j Basis

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During plant operation, the complete instrunentation systems will normally be in service.

Reactor safety is provided by the reactor protection system, which automatically initiates appropriate action to prevent exceed-

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ing established limits. Safety is not compromised, however, by continuing operation with certain instrumentation channels out of service since provisions were made for this in the plant design. This specification cutlines limiting conditions for operation necessary to preserve the effectiveness of the reactor control and protection system when any one or more of the channels are out of service.

)

i All reactor protection and almost all engineered safety feature channels i

are supplied with sufficient redund?ncy to provide the capability for i

cherrel test at power, except for backup channels such as derived circuits I

in engineered safeguards control system.

I l

When one of the four channels is taken out of service for maintenance, the

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protective system logic can be changed to a two-out-of-three coincidence I

for a reactor trip by bypassing the removed channel.

If the bypass is not l

effected,theout-of-servicechannel(PowerRemoved)assunesatripped ecr.dition(excepthigh pressurizerpressure),gte-of-changeofpower,highpowerlevelandhigh which results in a one out-of-three channel logic.

If in the 2 of 4 logic system of the reactor protective system one channel

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is t) passed and a second channel manually placed in a tripped condition, the resulting logic is 1 of 2.

At rated power, the minimum operable high-l power level channel is 3 in order to provide adequate power tilt detection.

If only 2 channels are operable, the reactor power level is reduced to 70% rated power which protects the reactor from possibly exceeding design reaking factors due to undetected flux tilts and from exceeding dropped i

CEA peaking factors.

All engineered safety features are initiated by 2-out-of-4 logic matrices except containment high radiation which operates on a 1-out of-5 basis.

References _

(1) FSAR, Section 7.2.7.1 l

2-66a Amendment No. E,29,25,22,4,SS, 125 1

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i 2.0 L1HITING CONDITIONS FOR OPERATION 2.15 TostrunifiWilon ar# Lontrol systces (Continued)

Basis (Continued).

l The engineered safety features system provides a 2 of 4 logic on the signals used to actuate thi (quipment connected to each of the two emergency diesel generator units.

The rod block system automatically inhibits all CEA motion in the event a Limiting Condition for Operation CLCO) on CEA insertion CEA deviation, CEA overlap or CEA sequencing is approached. TheinstallatIonoftherodblock L

system ensures that ro single failure in the control element drive control system (otherthanadroppedCEA)cancausetheCEAstomovesuchthatthe CEA insertion, deviation,' sequencing or overlap limits are exceeded.

Accordingly, with the rod block system installed, only the dropped CEA event is considered an A00 and factored into the derivation of the Limiting Safety System Settings and Limiting Conditions for Operation. With the rod block function out-of-service several additional CEA deviation events cust be considered as A00s. Analysis of these incidents indicates that the single CEA withdrawal incident is the most limiting of these events. An analysis of the at power single CEA withdrawal incident was performed for Fort Calhoun for various initial Group 4 insertions, and it 4as been concludedthattheLimitingConditionsforOperation(LCO)andLimiting Safety System Settings (LSSS) are valid for a Group 4 insertion of less than or equal to 15f.

TbtcperebilityoftheAlternateShutdownPanel(A1-185),includingWide Range Lcgarithmic Power and Source Range Monitors on Al-212 and Emergency AuxiliaryTeedwaterPanel(Al-179)instrumentandcontrolcIrcuitsensures that sufficient capability is available to permit entry into and maintenance of the Hot Shutdown Mode from locations outside of the Control Room. This carability is rtquired in the event that Control Room habitability is lost due to fire in the cable spreading room or Control Room.

2-66b Amendment No.125

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3.0 SURVE1LLANCE RE0_UiREMENTS j

3.1 1,nsfrumentationandControl Applicability Applies to the reactor protective system and other critical instru-i rentation and controls.

Obje ctiv,e,,

To specify the minimum frequency and type of surveillance to be applied to critical plant instrumentation and controls.

Specifications Calibration, testing and checking of instrument channels, reactor j

protective system and engineered safeguards system logic channels and miscellaneous instrument systems and controls shall be performed as specified in Tables 3-1 to 3-34.

l i

Basis Failures such as blown instrument fuses, defective indicators, and faulted amplifiers which result in " upscale" or "downscale" indica.

tion can be easily recognized by simple observation of the functioning of an instrument or system.

Furthermore, such failures are, in man cases, revealed by alarm or annuncictor action and a chtek supple y l

rtnts this type cf built-in surveillance.

i Eased on the District's experience in operation of conventional i

rower plants and on re i

frequency of once-per. ported nuclear plant experience, a checking shift is deemed adequate for reactor and steam system instrumentation. Calibrations are performed to ensure the presentation and ecquisition of accurate information, i

The power range safety channels are calibrated daily against a calorimetric balance standard to account for errors induced by changing rod patterns and core physics parameters.

i Other channels subject only to the " drift" errors, can be expected toremainwithInacceptabletolerancesifrecalibrationisperformed at each refueling shutdown interval.

3-1 Amendwent No. f.722.125

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ML20033F909

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