ML20005D616

Proposed Tech Spec Sections 2.15 & 3.1,providing Surveillance Testing Requirements of Alternate Shutdown Panel

ML20005D616
Person / Time
Site:	Fort Calhoun
Issue date:	10/27/1989
From:	OMAHA PUBLIC POWER DISTRICT
To:
Shared Package
ML20005D614	List: ML20005D613 ML20005D615 ML20005D616
References
NUDOCS 8911030126
Download: ML20005D616 (11)

Text

.

i TECHNICAL SPECIFICATIONS -. TA,8gS TABLE OF CONTENT _S, ,

TAttE . DESCRIPTION g  ;

3-3 Minimum frequencies for Checks, Calibrations, and Testing ,

, .. ,ci, Miscellaneous Instrumentation,and y trog ...... -

'3-3a

- Minimum Frequency for Checks, Calibrations and Functional Testing of Alternate Shutdown Panel M 3-15 3-16 i

(AI-185) and Emergency Auxiliary Feedwater Panel 3-16a

't

'\

(Al-179) Instrumentation and Control Circuits . . . . 3-16d 3-16b LL v.s. _.. % y 3-1(c; 3-4 Minimum Frequencies for Sampling Tests. . . . . . . . . . . .

3-19 3-5 Minimum Frequencies for Equipment Tests . . . . . . . . . . . 3-20 3-20a 3-20b 3-?Oc 3-20d 3-6 Reactor Coolant Pump Surveillance . . . . . . . . . . . . . . 3 27 3-7 Capsule Removal Schedule. . . . . . . . . . . . . . . . . . . 3-27 3-9 Radiological Environ' ment Monitoring Progrem'.. . . . . . . . . 3-66 3-67 3-11 Radioactive Liquid Waste Sampling and Analysis. . . . . . . . 3-72 3-73 3-12 Radicactive Gaseous Waste Sampling and Analysis . . . . . . . 3-74 3-75 3-13 Steert. Ger.erator Tube Inspection . . . ... ...... ... 3-90 5.?-1 Minimum Shif t Crew Composition. . . . ... ...... ... 5-2

(

)

v Arnendment No.116 89110'30126 891027

bc T,[C,H,N,1,C,AL, SPECIFICATIONS . TABLES .

TACLE CF CONTENTS (ALPHABETICAL ORDER)

TABLE DESCRIPTION PA,G,E 3-7 Capsule Removal Schedule. . . . . . . . . . . . . . . . . . . 3-27 21 ESFS Initiation Instrumentation Setting Limits. . . . . . . . 2-64 2-64a 2-7 Fire Detection Zones. . . . . . . . . . . . . . . . . . . . . 2 00 2-8 Fire Hose Station Locations . . . . . . . . . . . . . . . . .  ?-94  ;

2-96 27 Ha lon Area Fire Zones . . . . . . . . . . . . . . . . . . . .  ?-90a 24 Instrument Operating Conditions for Isolation Functinns . . .  ?-69 2-69a E-0 Instrun.ent Operating Requirements for RPS . . . . . . . . . . 0 67 T 674 ,

2-3 Instrut.er.t Operating Requirements fer En Safety Features . . . . . . . . . . . . gineerec . . .........  ?-68 2-68a 2-0P.b 2-5 Instruc4ntatiori Crerating Requirecients (cr Other Safet Features Functions. . . . . . . . . . . . . . . . . . y ...,  ?.70

, _- ~. , ' -

., . . ~ -

3-3a Minimum Frequency for Checks, Calibrations and - % . .,

Functional Testing of Alternate Shutdown Panel '

(Al-185) and Emergency Auxiliary Feedwater Panel (AI-179) Instrumentation and Controi Circuits . . . . 3-16d 3-10e-32 f*inicium Freputncies for Checks, Calibrations ant; Testing of Engineerec Safety Features. Instrumentation arc Controis. . . 37 3-8 39 3-10 3 11 31P 3-1?a 3-3 Finimum Frequencies for Checks. Calibrations, and Testing of Miscellaneous Instrumentation and Controls . . . . . . . . 3-13 3-la 3-15 3-16 3 16a 3 16b 3 16c 3-1 Minimum Frecuencies for Checks. Calibrations, and Testing of RPS. . . . . . . . . . . . . . . . . . . . . . 3-3 3-a i 3-5 3-6 v1 Ainenceent fio, 116 l-

l

, i g ,

2.0 LIMITING CONDITIONS FOR OPERATION

?.15 Instrumentation and Coprol Systems (Continued) ventilation isolation signals available if the containment ventila-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 initiating a hot shutdown procedure the inoperable 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This specification applies to the hi above 10*gh rate trip-wide

% power range log and is operating channel below 15% ofwhen ratedthe plant is at or power.

l (3) In the event the number of channels of a particular system in service falls below the limits given in the columns entitled " Minimum Operable ,

Channels" or " Minimum Degree of Redundancy", except as conditioned by the column entitled "Pennissible Bypass Conditions", the reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; however, opera-tion can continue without containment ventilation isolation signals 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, if the nuraber of operable high rate triprwide ,

rang log channels falls below that given in the column entitled

" Minimum perable Channels" in Table 2-2 and the reactor is at or above 10 g% power and at or below 15% of rated power, reactor cri

) operation shall be discontinued and the plant placed in an operational mode allowing repair of the inoperable channels before startup or reactor critical operation 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 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 <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the CEAs shall be' withdrawn and maintained at

, fully withdrawn and the control rod drive system mode switch shall

-u be maintained in the off position except when manual motion of CEA r A Gr up 4 is required to control axial power distribution.

~

Bass '

During plant operation, the complete instrumentation systems will normally be in service. Reactor safety is provided by the reactor protection system, which automatically initiates appropriate ac'/on to prevent exceed-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 outlines 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.

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.

2-66 Amandman+ m e w n nM l

INSERT A

'(4) In the event that any of the following Alternate Shutdown Panel instrumentation becomes inoperable, either restore the inoperable component (s) to operable status within seven days, or be in hot shutdown within the next twelve hours. This specification is applicable in Modes 1 and 2.

Wide Range Logarithmic Power (AI-212)

Source Range Power (AI-212)

Reactor Coolant Cold Leg Temperature (Al-185)

Reactor Coolant Hot Leg Temperature (AI-185)

Pressurizer Level (AI-185)

Volume Control Tank Level (AI-185)

(5) In the event that any of the following Emergency Auxiliary Feedwater Panel instrumentation becomes inoperable, either restore the l inoperable component (s) to operable status within seven days, or be in hot shutdown within the next twelve hours. This specification is applicable in Modes 1 and 2.

Steam Generator Level, Wide Range (AI-179)

~~~

Steam Generator level, Narrow Range (AI-179)

Steam Generator Pressure (AI-179) ,_, -

3

._, 1 Pressurizer Pressure (Al-179)

INSERT A b.

/ ,,  %'

i l

l l

l

2.0 LIMITING CONDITIONS FOR OPERATION 2.15 Instrumentation and control Systeims (Continued) t Basis (Continued)

When one of the four channels is taken out of service for maintenance, the protective system logic can be changed to a two-out-of-three coincidence for a reactor trip by bypassing the removed channel. If the bypass is not  :

effected, the out-of-service channel (Power Removed) assumes a tripped condition (excepthigh te-of-change of power, high power level and high pressurizer pressure),(p' which results in a one-out-of-three channel logi; r If in the 2 of 4 logic system of the reactor protective system one channel is bypassed and a second channel manually placed in a tripped condition, the resulting logic is 1 of 2. At rated power, the minimum operable high-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 peaking factors due to undetected flux tilts and from exceeding dropped 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.

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

N The rod block system automatically inhibits all CEA motion in the event a

/

Limiting Condition for Operation (LCO) on CEA insertion CEA deviation CEA overlap or CEA sequencing is approached. The 111stallation of the tvd block system ensures that no single failure in the control element drive control ,

system (other than a dropped CEA) can cause the CEAs to move such that the 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 must-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 has been concluded that the Limiting Conditions for Operation (LCO) and Limiting

• Safety System Settings (LSSS) are v.alid for a Group 4 insertion of less than or equal to 15%.

TussRr S 5'

~

-. - ~ lteferences (1) FSAR, Section 7.2.7.1 I 2-66a Amendment No. 8, 20, 25, 32. A3,88 i

l INSERT B r

The operability of the Alternate Shutdown Panel (Al-185), including Wide Range Logarithmic Power and Source Range Monitors on AI-212, and Emergency Auxiliary Feedwater Panel (Al-179) instrument and control circuits ensures that sufficient capability is available to permit entry into and

• maintenance of the Hot Shutdown Mode from locations outside of the Control Room. This capability is required in the event that Control Room habitability is lost due to fire in the cable spreading room or Control

i. , Room.

1

_. ,- INSERT B ,

M =g e ,

_- na

b .

3.0 SURVEILLANCE REOUlF.EMENTS 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 controls.

Specifications ,

Calibration, testing and checking of instrument channels, reactor protective system and engineered safeguards system logic channels and miscellanecus instrument systems-and controls shall be performed as specified in Tables 3-1 tQct, 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 alann 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-shif t is deeeed adequate for reactor and steam system instrumentation. Calibratiens are performed to ensure the presentation ar.d acquisition of accurate information.

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

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

3-1 Amendment No. 9,122

(!! ' 'I i , ' .!!  !! i

L ' I!!! '! [il l H N T O -

I IH -

D W D TC S E E AT -

N I L I RI O G L E L EW I E . G . P V P PS CT . LD E ED E H P E P O IAN T R C LR C T A L A L NCO S DA R A R I EA NII TDT E

T LO OB U

O TO OB O U W E K N

E R

VU LN

=

S T

INA RIC L C

L S N L

S L .

EN W

S A .

TN U

S AA VM

! A I A TO T TO T VO S O S u GED N NR N NR N EI E LI E CN OGN R AT E AT E LT R OT NG K

I LNI A

E T

LN OO R

R LN OO M A P RA R

P AU RC TC O C ER0 N OC O OC U EI L NI L PR G R I C C C C Z A OD A MH G

L O

NEA ACO G RT N

N RT N

N IM I T

CN I 9T P

N I

R T

RRB U

N I

OI*

TW W

O OI TW O W MI 5S N

E E

NS I T N

E RO N-T N EOL 5 C N C N 5R R UR R EI S O SSO 5 AE . K AE . K EA E LA E PT -

E C I R E N E N RO F OO T E WMN T E .

TS RWO TI O

T RWO TI O

T PS FR VB F .

FR OA .

RCN IO L W K EAO AT EAT EAT EL DS EL IO DS PIO I

A N

A AO LO R

AR C

MU I RRA AEC 5 T RRA AEC T

S RO AR NE N RO AR NE N YWT FIA

~

O N LN EN IC PPI 8 PPI W T WS PT WS I

T O

I IT EE WWT OOI LR AI MMD OEN J

D MMD OEN J

8 WN O O~

O NO MN OO O

NO I

RDE ENP R -

C T VN CPW CC CTI A CTI A CC KT CC KT VAO N)A U5T FSN R

S a a

a IE b b b a b a b r. -

N MI5 AAR (T

S S NLN OEI IN TA)

A AP9 1 '

3 R 7 -

3 BW1 I

L0I

- EE 0 N R N R N R N R N R R E

L A0A CT(

0 -

E B ,5 R A NL F T S5E K N .

CEA -

TP EA CNR RE _

RET _

OTA E E E E E E FLW C T T T T T .

A0 NN A A A A A Y E AO R R R R R .

FE KOF E

LI LT K

C B

I K

C B

I K

C B

I K

C B

I K

C B

I T U

Q R Y EE I

VU E

N C

L A

C E

H C

L A

C E

H C

L A

C E

H C

L A

E H

L A

S E

E A RF C C C T _

R I U . . . . . . . . . . .

F L I

S a b a b a b a b a b a M X U U N A DS _

I N

NR D Y AO L T _

I O O M K E

RI EN T

C N _

G WO TE TE R OM N NR _

E N

P E

AH LT AU LT L

O E R CG OA OA R T L EIN OR OR E NL O D GNA CE CE Z OE R N N NHR) P) P) I ) CV) TS)

A O AT 2 RM5 RM5 R 5 E5 NT5 LI RIE1 OE8 OE8 U 8 EL8 OI8 ET RC2 TT1 TT1 5L1 M 1 CU1 MIP AR EAR -

DGUI IOOA C

AGI EEA

- C AGI EEA

- 5E -

EVI REA UK -

LNI OAA PRI C -

HC WLS( RL( SIA RL( PL( VT( AC(

CS E . . . . . .

D 1 2 3 4 5 6 y._E .

,Ii  ;

TABLE 3-3A - (Continued)

NININUM FREQUENCY FOR CNECKS, CALIBRATIONS AM FUNCTIONAL TESTIN 0F ALTERNATE SMTDOW PANEL (AI-185)

AN ENERGENCY AUXILIARY FEEDWATER PANEL (AI-179) INSTRONENTATION AM CONTROL CIRCNITS CHANNEL SURVEILLANCE SURVEILLANCE DESCRIPTION FUNCTION FRE00ENCY NETN00

7. STEAN GENERATOR a. CHECK N a. COMPARE INDEPEWENT LEVEL LEVEL, WIDE RANGE I NICATIONS.

(AI-179)

b. CALIBRATE R b. KNOWN DIFFERENTIAL APPLIED TO SENSOR 5.

8. STEAN GENERATOR a. CHECK N c COMPARE INDEPERENT LEVEL LEVEL, NARROW RANGE IMIICATIONS.

(AI-179) w b. CALIBRATE R b. KNOWN DIFFERENTIA. APPLIED TO j SENSOR 5. -

9. STEAM GENERATOR a. CHECK M a. COMPARE I NEP010ENT PRES $W E PRESSURE INICATIONS. .

(AI-179) i'

b. CALIBRATE R b. KNOWN PRES $URE APPLIED TO SENSOR.

10. PRESSURIZER a. CNECK N a. COIFARE INEPE WENT PRE 55NRE PRESSURE INDICATIONS. i (AI-179)

b. CALIBRATE R b. KNOW PRESSURE APPLIED TO SENSOR. i

11. EAFW CONTROL a. TEST R a. VERIFY PROPER VALVE OPERATICE AND CIRCUITS INDICATION TINIOUGN MANGAL SWITCW (AI-179) OPERATION.

l t

t t

h 1

l-

+

i ATTACHMENT B ,

t o

?

Discussion, '

Justification and No Significant Hazards Consideration i

Description of Amendment Request to provide the Limiting Conditions for Operation and Surveillance Requirements for alternate and auxiliary alternate shutdown panels. ,

The proposed Technical Specification changes shown on page 2-66,2-06a and 3 1 of the Technical Specifications provide the operability requirements and basis of these requirements for alternate and auxiliary alternate shutdown panels. The addition of Table 3-3a provides the Surveillance .

Requirements for alternate and auxiliary alternate shutdown panels.

As requested by Reference 2, and committed to in Reference 3, this Facility License Change (FLC) revises the Technical Specifications to provide surveillance and operability requirements that are appropriate for a Protective or Safeguards System as applied to alternate and auxiliary '

alternate shutdown panels.

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

1. involve a significant increase in the probability or consequence of an accident previously evaluated. This Change decreases the consequences and probability of accident / event escalation during the forced evacuation of the Control Room. Since the alternate and auxiliary alternate shutdown panels assumes a major protective function for plant control during the loss of the Control Room in a fire, this FLC is required to ensure continued .

operability of alternhte and auxiliary alternate shutdown panels.

2. create the possibility of a new or different kind of accident from any accident previously evaluated. 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 surveillances provide a higher level of assurance that alternate and auxiliary alternate shutdown panels will function if required to do so.

3. involve a significant reduction in the margin of safety. This change results in an increase in the margin of safety associated with alternate and auxiliary alternate shutdown panels and the Alternate Shutdown capability by assuring that the system will operate properly through surveillance tests, and applying Limiting Conditions for Operation to this system.

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

1 ATTACHMENT B

-_