ML20247P587

Proposed Tech Specs Re Surveillance Testing of Alternate Shutdown Panel

ML20247P587
Person / Time
Site:	Fort Calhoun
Issue date:	07/31/1989
From:	OMAHA PUBLIC POWER DISTRICT
To:
Shared Package
ML20247P568	List: ML20247P566 ML20247P570 ML20247P587
References
NUDOCS 8908040224
Download: ML20247P587 (7)

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d'h 2 .0 LIMITING CONDITIONS FOR OPERATION 2.15 Instrumentation and control systems (Continued)

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ventilation isolation signals available if the containment ventila-W d <

y tion isolation valves are closed. If after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from time of

{,.f } h. -, 3 t/)s g features I (, initiating a hot shutdown procedure the inoperable engineered safety G

3 q jf ddL o or isolation functions channel has not been stored to m -) U y L,2 operable status, the reactor shall be placed in a cold shutdown 0 , c. A' #d a condition within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This specification applies

d d j ] g),c -r-jj N S .5gto the high rate trip-wide W

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above 10 range log channel when the plant is at or power and is operating below 15% of rated power.

,0 In the event the number of channels of a particular system in service d

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'q -Q falls below the limits given in the columns entitled " Minimum Operable

_.y j Channels" or "tiinimum Degree of Redundancy", except as conditioned by 0- d' -

the column entitled " Permissible Bypass Conditions", the reactor shall y j~E be placed in a hot shutdown coi.dition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; however, oper

)dC[h - d yO cp5 C.- tion can continue without containment ventilation isolation signals e, 6 available if the vei.tilation isolation valves are closed. If minimum c # .J L 'If . t ,.g E

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.4 conditions for engineered safety features or isolation functiens are j dy e jf 2 o &

0 v 4 ,g not met within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from time of discovering loss of operability.

gV 2 5 #* the reactor shall be placed in a cold shutdown condition within the Al. ie id 7 5' following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If the number of operable high rate trip-wide W h d 'thGV O O rang '-'

log channels falls below that given in the column entitlel 0 " Minimum 0 g t power and at or below 15% of rated power, reactor critical o q P 4' E g 3j above 10 be gperable Channels" ininTable 2-2 an operation shalt discontinued and the piant piaced an operational

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mode allowing repair of the inoperable channels before startup or j f $ d d v ['v

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,,,d, ,, reactor critical operation may proceed.

~T g l --t'0 f d "3 o 0- CEA d

If, during power operation, position the rod block indication function and system of the secondary rod block ci y g g 9 8 g -F 0 4 00 g.pf, Ndeviation 0

• f r mrealann thanand 24 the hours, of f dWg CEAor the plantfunction sequencing computer more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the CEAs shall be' withdrawn and maintained at PDIL alarm are inoperable fo

~'$ ' j ; fully withdrawn and the control rod drive system mode switch shall jh{J pdgd 4g g ] {~~~~be roupmaintained in the off position except when manual motion of CEA 4 is required to control axial power distribution.

-r -v to y d e S m 4 Basis CC'h During plant operation, the complete instrumentation systems will normally be in service. P.eactor safety is provided by the reactor protection

.- system, which automatically initiates appropriate action to prevent exceed-

$., ing established limits. Safety is not compromised, however, by continuing

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operation with certain instrumental'on channels out of service since provisions were made for this in the plant design. This specification f outlines limiting conditions for operation necessary to preserve the I effectiveness of the reactor control and protection system when any one or more of the channels are out of service.

All reactor protection and almost all engineered safety feature channels are supplied with sufficient redundancy to provide the capability for channel test at power, except for backup channels such as derived Circuits in engineered safeguards control system. ,

1 8908040224 890731 l PDR ADOCK 05000265 2-66 Amendment No. B, 20. N , 65,gg l P PDC.

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N9 k y[\ 2.g 2.0 LIMITING CONDITIONS FOR OPERATION 2.15 Instrumentation and control Systems (Continued) bg O y Basis (Continued) b5

~~d When one of the four channels is taken out of service for maintenance, the s W protective system logic can be c5anged to a two-out-of-three coincidence k'%y'y"Ihg

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'J for a reactor trip by bypassing the removed channel. If the bypass is not Q D effected, the out-of-service channel (Power Removed) assumes a tripped R

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condition (except hi h(rqte-of-change of power, high power level and high pressurizer pressure , le which results in a one-out-of-three channel logic.

30 If in the 2 of 4 logic system of the reactor protective system one channel f 9D Q+K-R is bypassed and a second channel manually placed in a tripped condition, the reniting logic is 1 of 2. At rated power, the minimum operable high-

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newer 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 b Q(h[D N

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70% rated power which protects the reactor from possibly exceeding design pei. king factors due to undetected flux tilts and from exceeding droppea CEA peakinEf actors.

{ {T 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.

q NN The engineered safety features system provides a 2 of 4 logic on the signals y S -( 8 '

used to actuate the equipment connected to each of the two emergency diesel k generator units.

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9*~ N, The rod block system automatically inhibits all CEA motion in the event a Limiting Condition for Operation (* CO) on CEA insertion, CEA deviation, CEA

$%k G overlap or CEA sequencing is approact ed. The installation of the rod block h

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'1 e system ensures that no single fai~ -e in the control element drive control R g system (other than a dropped CEA) can cause the CEAs to move such that the

- g y ~g 9 CEA insertion, deviation, sequencing or overlap limits are exceeded.

D'N q Accordingly, with the rod block system installed, only the dropped CEA dd, L s event is considered an A00 and factored into the derivation of the Limiting I Safety Syst~i Settings and Limiting Conditions for Operation. With the d k9 'qk rod block function out-of-service several additional CEA deviation events Q b[U O

C g) must be considered as A00s. Analysis of these incidents indicates that the single CEA withdrawal incident is the most limiting of these events. An N%I b8 analysis of the at-power single CEA withdrawal incident was performed for

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i I k D g) ' Fort Calhoun for various initial Group 4 insertions, and it has been '

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[Q4 I concluded that the Limiting Conditions for Operation (LCO) and Limiting Safety System Settings (LSSS) are valid for a Group 4 insertion of less than or equal to 15%. )

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L y References (1) FSAR, Section 7.2.7.1 f

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3 2-66a Amendment No, j!, 20, 25, 32, f3,88 1

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?. 0 SURVEILLANCE RE00fEEMENTS 3.1 ' Instrumentation and Control Applicability Applies to the reactor protective system and other critical instru-mentation and controls.

Objective To specify the minimum frequency and type of surveillance to be applied to critical plant instrumentation and ccatrols, Specifications Calibration, testing and checking of instrument channels, reactor protective system and engineered safeguards system logic channels and miscellaneous-instrument systems and controls shall be performed as specified in Tables 3-1 to 3-3ct.

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 many cases, revealed by alarm or annunciator action and a check supple-ments this type of built-in surveillance.

Based on the District's experience in operation of conventional power plants and on reported nuclear plant experience, a checking frequency of once-per-shift is deemed adequate for reactor and steam system instrumentation. Calibrations are performed to i ensure the presentation and acquisition of accurate information.

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

Other channels, subject only to the " drift" errors, can be expected I to remain within acceptable tolerances if recalibration is performed at each refueling shutdown interval.

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ATTACHMENT B l

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Discussion, Justification and No Significant  ;

Hazards Consideration '

Description of Amendment _ Request to provide the Limiting Conditions for j Operation and Surveillance Requirements for the Alternate Shutdown Panel.

The proposed Technical Specification changes shown on page 2-66,2-66a and 3-1 of the Technical . Specifications provide the operability requirements and basis of these requirements for the Alternato Shutdown Panel (ASP).

The ' addition of Table 3-3a provides the Surveillance Requirements for the ASP.

As requested by Reference 2, and committed to in Reference 3, this Facility License Change (FLC) revises the Technical Specifications to x

N provide surveillance and operability requirements that an appropriate for i a Protective or Safeguards System as applied to the Alternate Shutdown

. Panel (ASP).

n No Significant Hazards Consideration This proposed amendment does not involve significant hazards based on the following.information:

1. Will the change involve a significant increase in the probability or consequence of an accident previously evaluated?

Answer No. This Change decreases the consequences and probability of

' accident / event escalation during the forced evacuation of the Control Room. Since the ASP assumes a major protective function for plant control during the loss of the: Control Room in a fire, this FLC is required to ensure the ASP'S continued operability.

2. Will the change create the possibility of a new or different kind of accident from any accident previously evaluated?

Answer

- No. . It has been determined that a new or different type of accident 'is not created because no new or different-modes of operation are proposed for the plant. Additional surveillance provide a higher level of assurance that the ASP will function if required to do so.

3. Will the change involve a significant reduction in the margin of safety?- { .

Answer No. This change results in an increase in the margin of safety associated with the ASP and the Alternate Shutdown capability by assuring that the system will operate properly through  !

surveillance tests, and applying Limiting Conditions for Operation to th's system. i

y. ATTACHMET B

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