ML20043H505

Proposed Tech Specs Correcting Typos & Refs to Fsar,Testing to Be Done During Initial Plant Startup & Incorporating Revised Operating Power/Flow Map for Two Reactor Recirculation Loop Operation

ML20043H505
Person / Time
Site:	Clinton
Issue date:	06/19/1990
From:	ILLINOIS POWER CO.
To:
Shared Package
ML20043H503	List: ML20043H505 U-601650, Requests Changes to License NPF-62,correcting Typos, Refs to Fsar,Testing to Be Done During Initial Plant Startup & Incorporating Revised Operating Power/Flow Map for Two Reactor Recirculation Loop Operation
References
NUDOCS 9006260034
Download: ML20043H505 (32)

Text

Att:chment l'

. 4 to U 601650 Page 9 of 38 Table 1 Affected. Bases Pages i

f Affected Section, Associated.Doscribed Table or Finure_

Channe Nn=her~

EA&R 2.2,1, item 11 l'

B 2-9 3/4.1.3 1"

.B 3/4 1 2 3/4.1.4 l'

B 3/4 1 4 Table B 3.2.1 1 l'

'B 3/4 2-3 3/4.2.3

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B 3/4'2 Figure B 3/4,2,3-1 2-B 3/4 2 7 3/4.3,3 6

B 3/4 3-3' 3/4,3,4 1

B 3/4 3-3 3/4.4.1

~1 B'3/4 4-1,.2 and 3 3/4,4,5 3

B 3/4 4 5 3/4,4,6

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B 3/4 4e5 i

3/4.5.1 and 3/4.5.2 6

B 3/4 5 a 3/4.6.1.4 1

B 3/4'6-2 3/4,6,1,8 4=

B 3/4'6 3 3/4,6.2.5 5

B.3/4-6 5 3/4,6.2.7 1_

B'3/4 6-5 3/4,6.3 1

B'3/4:6 6 3/4.6.4 1

B'3/4 6 7 3/4.6.6 1

B.3/4 6-8

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1 B'3/4 7-1 3/4.7.3 6

B 3/4 7-1 3/4.7.6

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_B 3/4.7-3 3/4.8.1, 3/4.8.2, and 3/4,8.3 7

B'3/4 8 1 3/4.8.1, 3/4,8.2, and 3/4,8.3 8

B 3/4 8 2

'3/4.9.12 1

B 3/4.9-3 1

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LIMITING SAFETY SYSTEM SETTINGS

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BASES 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued) i 9.

Scram Discharoe Volume Water Level-High The scram discharge volume receives the wate.r displaced by the motion'of the control rod drive pistons during a reactor scram. Should this volume fill up to a point where there is insufficient volume to accept the displaced water at pressures below 65 psig, control rod insertion would be hindered. The reac-tor is, therefore tripped when the water level has reached'a point high enough to indicate that It is indeed filling up, but the volume.is still great enough to accommodate the water from the movement of the rods at pressures below 65 psig when they are tripped. The trip setpoint for each scram discharge l

volume is equivalent to a contained volume of 19.6 gallons of water.

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10. Turbine Stop Valve-Closure The turbine stop valve closure trip anticipatos the pressure, neutron flux, and heat flux increases that would result from closure of the stop valves.

With a trip setting of 5% of valve closure from full open, the resultant increase in heat flux is. such that adequate thermal. margins :are maintained during the worst case, transient.

11. Turbine Control Valve Fast Closure, Valve Trip System 011 Pressure-Low The turbine control valve fast closure trip anticipates the pressure, neutron-flux, and heat flux increase that could result from fast closure of the turbine control valves due to load rejection with or without coincident failure of the turbine bypass valves. The Reactor Protection System initiates,a trip when fast closure of the control valves is initiated by the fast acting solenoid valves.

and in less than 20 esec after the start of control valve fast closure.. This I

is achieved by the action of the fast acting solenoid valves.in rapidly reducing hydraulic trip oil pressure at the main turbine control valve actuator disc dump valves..This loss of. pressure is sensed by pressure switches whose contacts form'the two-out-of-four logic input to the Reactor-Protection System. This trip setting, a slower closure time, and a different valve characteristic from that of the turbine stop valve combine to produce transients which are very similar to that for the sto valve. Relevant transient analyses are discussed

'j lysis Repor g in Section 15.2.2 of the et A dahe

12. Reactor Mode' Switch Shutdown on

-The reactor mode switch Shutdown position provides additional manual reactor trip capability to the manual scram pushbutton switches.

13. Manual Scram

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The manual scram pushbutton switches provide a diverse means for initiet.ing a reactor shutdown (scram) to the automatic protective instrumentation channels and provides manual reactor trip capability.

CLINTON - UNIT 1 8 2-9 I

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,4 Page 11 of 38 REACTIVITY CONTROL SYSTEMS +

. BASES

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3/4.1.3 CONTROL RODS (Continued)

A limitation on inoperable rods is set-such that thel resultant effect on total rod worth and-scram shape will be kept to a minimum? The requirements for the j

indication of systematic pro-various scram time measurements ensure that any, timely basis.-

blems with rod drives will be investigated on a Damage within the control rod drive mechanism could be a generic problem, there-fore with a control rod immovable because of excessive friction or mechanical i

interference, operation of the reactor is limited =to a time period which is reasonable to determine the cause of the inoperability and at the same time prevent operation with a large number of inoperable control rods.

j Control rods that are inoperable for other reasons:are permitted to be taken 1

out of service provided that those in the nonfully-inserted position are con-sistent with the SHUTDOWN MARGIN requirements.

The number of control rods permitted to:be inoperable could be more than the eight allowed by the specification, but the occurrence of eight inoperable rods 4

could be indicative of a generic problem and the reactor must be shut down for investigation and resolution;of the problem..

I The control rod system is designed to bring the reactor-subcritical at a rate fastenoughtopreventtheMCPRfrombecominglessithanthefue}claddins@ufetv-=Q.

limit.during the limiting power transient analyzed in Section 15.4 of tt.e A M (f5 A L

This analysis shows that-the negative reactivity rates resulting from t.1e scrasr W -

with the average response of all the drives as given-in the specifications, i

provide the required protection and MCPR remains greater.than the fuel cladding.

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safety limit. The occurrence of. scram times longer then~those specified should 1

be viewed as an indication of a systemic problem with the rod drives and there -.

fore the surveillance interval is reduced:inl order to prevent operation of. the i

reactor for long periods of time with a potentially serious problem.

l The scram discharge volume is requirrd to be OPERABLE so' that it will.be avail-able when needed to accept discharre water from the control rods during a

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reactor scram and will isolate tN reactor coolant system from the containment when required.

Control rods with inoperable accumulator,s are declared inoperable and Specifi-L cation 3.1.3.1 then applies. This prevents a pattern of inoperable accumulators l

that would result in less; reactivity insertion on a scram than has been analyzed l-even though control rods with inoperable accumulators may still be inserted with normal drive water pressure. Operability of the accumulator ensures that there is a means available to insert the control rods even under the mesi unfavorable depressurization of the reactor.

Control rod coupling integrity ed to ensure compliance with thj analysis of the rod drop accident in th The overtravel position feature provides l

the only positive means of determining that a rod is properly coupled ~and Ibere-fore this check must be performed prior to achieving criticality af.ter comp +eting

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CLINTON - UNIT 1.

B 3/4 1-2

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Page l'2' of 3'8 REACTIVITY' CONTROL SYSTEMS BASES' 3'4.1.4 CONTROL R00 PROGRAM CONTROL:.' (Continued):

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The analysis of the rod. drop accident is presented in Section 15.4 of the and the techniques of the analysis'are presented in a topical feport,iReference 1, and two-supplements, References 2 and 3.

The RPCS is also designed to automatically prevent fuel damage'in the event'of.

erroneous rod withdrawal from locations of high power density during higher, power operation.

A dual channel system is provided that, above the low power setpoint,. restricts' the withdrawal distances of all non peripheral control rods. ThisLrestriction '

is greatest at highest power levels.-

l 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM The standby liquid control system provides a backup capability' for' bringing the h

reactor from full power to a cold, xenon-free shutdown, assuming'that the with '

drawn control rods remain fixed in the rated power pattern. To meet this objec-'

tive it is necessary to inject a quantity of sodium pentaborate solution which l

produces a concentration of 66.0 ppm in the reactor core and othe.r piping' systems 7 p.

, connected to the reactor vessel. To allow for potential leakage' and imperfect.

mixing this concentration is increased by.25%. The required concentration is

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achieved by having a minimum available quantity of 3574 gallons of4 sodium-pen--

L' taborate solution containing a minimum of 4246 lbs, of sodium pentaborate. ' This F

quantity of solution is a net amount which is above the pump suctiori, thus L

allowing for the portion which cannot be injected. The pumping rate of-41.2 gpm

.-1 per pump provides a negative reactivity injection rate over the pemissible pentaborate solution volume range, which adequately compensates for the positive-reactivity effects due to temperature and xenon during shutdown. ;The tempera- -

l ture requirement is necessary to ensure that the sodium pentaborate remains in-l-

solution.

With redundant pumps and explosive injection valves and with a highly reliable control rod scram system, operation of the reactor is permitted to continue for short periods of time with the system inoperable or for longer periods of time with one of the redundant components inoperable.

l Surveillance requirements are established on a frequency that assures a high reliability of the system. Once the solution is established, sodium;pentaborate-1.

C. J. Paone, R. C. Stirn and J. A. Woolley, " Rod Orop Acciderit Analysis for Large BWR's," G. E. Topical Report HED0-10527, March 1972 2.

C. J. Paone, R. C. Stirn and R. M. Young, Supplement 1 to HEDO-10527c July 1972 3.

J. M. Haun, C. J. Paone and R. C. Stirn, Addendum 2. " Exposed fores," 1 Suppl.ement 2 to HE00-10527, January 1973 CLINTON - UNIT 1 B 3/4 1-4 r

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Pagec13 of 38 BASES TABLE B 3.2.1-1 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-0F-COOLANT-ACCIDENT ANALYSIS *

' Plant Parameters:.

Core THERMAL POWER...........'.......... _ 3015' HWt** which corresponds-to 105% of rated steam flow

~

Vessel Steam Output..'..................jl3.08 x 106 11b,/hr which corresponds to 105% of rated steam flow Vessel. Steam Dome Pressure..............

1060 psia i

Design Basis Recirculation L'ne Break Area for:

a.

Large reaks 2.2 ft,-

2 b.

Small Breaks.

0.09 ft,

2 Fuel Parameters:

PEAK TECHNICAL INITIAL

. SPECIFICATION--

DESIGN MINIMUM

. LINEAR HEAT'

. AXIAL CRITICAL i

FUEL BUNDLE ~

-GENERATION RATE _

_ PEAKING

-POWER-i

)

, FUEL TYPE GE0 METRY (kW/ft)'

FACTOR RATIO Initial and 8'x 8 1;4 1.17***

t Reload. Cores

• A more detailed listing of input of.each model and-resented in Section II of Reference 1 and-Section 6.3 of th

= = _

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• • This power level meets the Appendix K requirement of 102%.- The core heatup.

calculation assumes a bundle power consistent' with' operation 'of the highest powered rod at 102% of its Technical. Specification LINEAR. HEAT GENERATION,

RATE limit.

• • • For single recirculation Idop operation, loss of nucleate boiling is assumed at 0.1 seconds af ter a LOCA, regardless of initial MCPR.

For core flows less than 85% of' rated, the initial MCPR is taken from the MCPR Curve specified in'the CORE OPERATING LIMITS REPORT.

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1. This value is specified in the CORE OPERAflNG LIMITS REPORT.

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CLINTON - UNIT 1

.B 3/4 2-3 Amendment No. 28

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Attechment 1 E

to U-601650 POWER DISTRIBUTION LIMITS Page 14 of 38 i

BASES 3/4.2.3 MINIMUM CRITICAL POWER RATIO (Continued)

I The MCPR s are established.to protect the core from plant transients.other than p

core flow increases, including the localized event such as rod withdrawal error.

The MCPR s.were calculated based upon the most limiting transient at the given-p core power level, including feedwater controller and load rejection l

transients.

For core power below 40% of RATED THERMAL POWER @ ere.the E0C-RPT and reactor scram on turbine 'stop valve closure and turbine'c ;ntrol valve fast closure are bypassed, separate sets of MCPR limits are provided for-p high and low core flows to account for the sensitivity to initial core flows.

For core power above 40% of. RATED THERMAL POWER, bounding MCPR limits were developed.

P At THERMAL POWER levels ;1ess than or' equal to 25% of RATED THERMAL POWER, the i

reactor will be operating at minimum recirculation pump. speed and the moderator-void content will be very small.

For all designated control. rod patterns which may be employed at this point, operating plant. experience indicates that the resulting MCPR value is in excess of requirements by a considerab

1. 6 During initial start-up testing of the plant, a MCPR evaluationigrgin.

5%gf RATED THERMAL _ POWER level with minimum recirculation pumpD_

d l J g K i 4e)demonstra that future MCPR evaluation w

this oower lev e 911T TN5E.

nnecessary. The daily requirement for OS -

calculating MC)R w en H RMA s greater than or equalito-25% of RATED THERMAL POWER is sufficient since power distribution shifts;are very slow when there have not been significant power lor control rod changes. 'The requirement for calculating MCPR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the completion of-a THERMAL. POWER L

increase of at least 15% ~of RATED THERMAL POWER ensures thermal liniits are met l

after power distribution shifts whilestill allotting time for:the power dis-tribution to stabilize.

The requirement for calculating MCPR afterninitially determining'a LIMITING CONTROL R00 PATTERN-exists ensures that MCPR will be known following a change.in THERMAL-~ POWER or power shape that' could place operation exceeding a thermal limit.

3/4.2.4 LINEAR HEAT GENERATION RATE L

This specification assures that the Linear Heat Generation Rate (LHGR) in any l

rod is less than the design linear heat-generation even if: fuel __ pellet densifi-l cation is postulated.

The daily requirement for calculatin ~LHGR when THERMAL POWER is greater than or equal to.25% of RATED THERMAL POW!R is sufficient since power distribution-shifts are very slow when there have not been significant power or control rod changes. The requirement to calculate LHGR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.after-the completion of a THERMAL POWER increase of at least 15% of RATED-THERMAL POWER ensures'ther-l mal limits are met after power distribution shift. Calculating LHGR after I

initially determining a LIMITING CONTROL R0D PATTERN exists ensures that LHGR will be known following a change in THERMAL POWER or power shape that could place operation exceeding a thermal limit.

i CLINTON - UNIT 1 B 3/4 2-5 Amendment No. 18 s

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'Page'1. of 38 i

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A. NATURAL CIRCULATION '

8. low mECIRCULATION PUMP SPEED VALVE MINIMUM POSITIGN

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C. LOW RECIRCULAflON PUMP SPEED V ALVE MAXIMUM PoslTION i

O. R ATED RECIRCULATION PUMP SPEED VALVE MINIMUM POSITION 130 l

l INCR ASE D CORE

, 20 I

FL REGION -

EXTENDEO LOAD

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i LINE RE0lON-

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110

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100nsi (100It07).

APRM STP S AM APRM ROD BLOC 90 f,_

MECOSOUNDARY ROD LINES

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g 105%-

to 100%

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100.8/32.7) g 30 A

30 l

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30 CAVITATK)ef PROTECTION 20 ggg opgnAy; is noy goyaggo WITHOUT A ERENCE TO GE

-SERVICE INF 4AATION LETTER NO. 300. *'Wwm E THERMAL '

10 HYDRAULIC STA LITY" ISEE SPEC 3.4.1 1)

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10 20 30 40 So to 70 00 90 100 110' 120.

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CORE FLOW (%)

1 Bases Figure B 3/4.2.3-1 Reactor Operating Map for Two Recirculation Loop Operation CLINTON - UNIT 1 B 3/4 2 7 Amendment No.

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. A= NATURAL CIRCULATION l

l B = LOW RECTRCULATION PUMP SPEED, VALVE MINIMUM POSITION 140 C = LOW RECIRCULATION PUMP SPEED, VALVE MAXIMUM POSITION l..

Z" D= RATED REC!RCULATION PUMP SPEED, VALVE MINIMUM POSITION IQ' q

130 f g ; {'

} h ~(j INCREASED CORE -

120 EXTENDED LOAD FLOW REGION UNE REGON

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10S%

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BOUNDARY F A A

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(60.8/32.7) r-50 O$

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OPERATION IS NOT ADV! SED Oh

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WITHOUT ADHERENCE TO GE 20 -

PROTECTION SERVICE INF.)RMATION LETTER g y(

LINE NO. 380 W CCRE THERMAL 10 -

HYDRAUUC STAB 1UTY" z

(SEE SPEC 3,4.1-1) 2 ER O

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10 20 30 40-5V c0 170.7 80 90

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Attachmant 11 to U-601650:

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INSTRUMENTATION-1)

BASES-3/4.3.2 CONTAINMENT AND REACTOR VESSEL' ISOLATION CONTROL SYSTEM (cont each case which in turn determines the valve speed in conjunction with the 13 sec-It-follows that checking the valve r,peeds and the 13'second time for

.ond delay.

emergency power establishment will establish the response time for the isolation functions..

i-1 Operation with a tri set less conservative than its Trip setoot'nt but within its specifiedAllowablehalueisacce)tableonthebasisthatthedifferencebetween

(

each Trip Setpoint and=the A11owa)1e Value-is equal to or less'than the drift allowance assumed for each trip in the safety. analyses. The Trip Setpoint'and.

l Allowable Value also contain additional. margin for instrument accuracy and

.i calibration, capability.-

i 3/4.3.3 EMERGENCY' CORE COOLING SYSTEM ACTUATION INSTRUMENTATION l

The emergency core cooling system actuation instrumentation is provided.toLiniti-ate actions to mitigate the consequences of accidents that are beyond the ability of the operator to control. - This. specification provides=the OPERABILITY require-monts, trip setpoints and response times that will ensure effectiveness of the systems to provide the desigt.+ protection.1 Although the instruments aire listed l

by system, in some cases the same instrument may besused to, send the: actuation

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signal to more than~one-system at the'same time..

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TN6ERT Operation with a trip: set less conservative than its Trip Setooint but within its 9

specified Allowable Value is acce> table on the basissthat the' difference between each irip setpoint and the-Allowaale Value is equal; to or less than the drift -

allowance assumed for each trip in the-safety analyses. -The Trip Setpoint and Allowable Value also contain: additional margin for instrument accuracy and calibration capability.

The emergency core cooling system (ECCS) pump minimum flow insthments? are pro.

vided to ensure that ECCS pump minimum flow paths are preserved to prevent pump -

damage in the event that ECCS pumps'are started without reactor or test,line flow The minimum. flow instruments, are-not part of ECCS actuation instrumentation.

paths.

i 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION' The anticipated transient without scram (ATWS) recirculation pump trip system provides a means of limiting the consequences of the unlikely occurrence of a failure to scram during an. anticipated transient. The response of thelplant to this postulated event f alls within the-envelope of study events in General.Elec-tric Company Topical Report NE00-10349 t

c 1971 and NEDO 24222, dated December 1979, and Section 15.8 of th

SAS, The end-of-cycle rectreulation pump trip (EOC-RPT) system is an essential 4

safety supplement to the Reactor Protection System. The purpose of the EOC-RPT i

The is to recover the loss of thermal margin which occurs at the end-of-cyg,le.

)

physical phenomenon involved is that the-void reactivity feedback due to a pressurization transient can add positive mtivity to the reactor systee.'at a

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l Specific ACTION statements are provided which are required'to be satisfied in the event an instrument channel (s) is declared inoperable.

- For-the ADS, this may require declaring 1the associated ADS trip system-inoperable.- - Although the RCIC system may,not be considered OPERABLE per Specification 3.7.3, for the purposes.of satisfying the ACTION i

t requ remen s for one ADS trip system inoperable, the RCIC system may be considered OPERABLE when reactor pressure is'less'than or: equal to 150-psig.-

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BASES t

3/4.4.1 RECIRCULATION SYSTEM i

The impact of-single recirculation loop operation upon plant ofety is ass % sed

' and shows that single-loop operation is ~ permitted if the MCPR fuel cladding-safety limit is increased as noted by Specification 2.1.2, APRM scram and-control rod block setpoints are adjusted as noted in Tables 2;2.1-1 and 3.3.6-2,.

respectively, MAPLHGR limits ire decreased in accordance with the values specified in the CORE OPERATING LIMITS REPORT, and MCPR operating < limits are.

adjusted in accordance with the' values specified in the CORE OPERATING LIMITS j'

REPORT.

Additionally, surveillance on the volumetric flow rate-of the operating: recir-culation loop is imposed to exclude the possibility of excessive core internals vibration.

The surveillance on differential temperatures below (30%)* THERMAL-POWER or (50%)* rated recirculation loop flow'is to mitigate the undue thermal stress on vessel nozzles, recirculation pump,-and vessel bottom head during:the extended operation of the single recirculation loop mode.

An inoperable jet pump is not in itself, a sufficient reason to= declare a re-circulation loop inoperable, but it does, in case of a design-basis-accident, j

increase the blowdown area and reduce the capability,uf reflooding the core; thus, the requirement for shutdown of the facility with a jet pump inoperable.

Jet pump failure can be detected by. monitoring jet pump performance on a pre-3 i

scribed schedule for significant degradation.

Significant degradation is-indicated if more than one of three specified surveillances performed confirms unacceptable deviations from established patterns or relationships.

The surveillances, including the associated acceptance criteria, are. in accordance with General Electric Service Information Letter No. 330, the recommendations according to NUREG/CR-3052, p'Closcout of. IE Bulletin 80-07:of which are co

)

BWR Jet Pump Assembly Failure." Performance of the specified surveillances,. however, is 1

not reouired when thermal power is less than 25% RATED. THERMAL POWER because flowoscillationsandjQnoiseprecludes.thecollectionofrepeatable meaningful data during low flow conditions approaching the threshold: response lj of the associated flow instrumentation.

Recirculation loop flow mismatch limits are in compliance with ECCS LOCA analysis design criteria for two recirct.lation loop operation. The limits-

)

will ensure an adequate core flow coastdtwn from either recirculation loop i

following a LOCA.

In the case where the pismatch limits cannot be maintained during two loop operation, continued operation is permitted in a single, recirculation loop mode.

In order to prevent undue stress on the vessel nozzles and bottom head region,

~f the recirculation loop temperatures shall be within 50 F of each other prior to startup of an idle loop.

The loop temperature must also be within 50 F of-the reactor pressure vessel coolant temperature to prevent thermal shock'to the recirculation pump and recirculation nozzles.

Sudden equilization of a tempera-ture difference > 100 F between the reactor vessel bottom head coolant and the coolant in the upper region of the reactor vessel by increasing core flow rate would cause undue stress in the reactor vessel bottom head.

• Initial Values.

FinalvaluestqbedeterminedduringStartupTes'tingba}edon the threshold THERMAL POWER and recirculation loup flow which wil'1 sweep the cold water from the vessel bottom head preventing stratification.

CL1HTON - UNIT 1 B 3/4 4-1 Amendment No. 28

to U-601650 4

Page 20 of'38 REACTOR COOLANT SYSTEM BASES 3/4.4.1 RECIRCULATION SYSTEM (Continued)

The objective of GE BWR plant and fuel design is to provide stable. operation

~

with margin over the normal operating domain.

However, at the high power / low flow corner of the operating-domain, a small probability of neutron flux limit cycle oscillations exists depending on combinations of operating conditions

[-

(e.g., rod pattern, power shape).

To provide assurance that neutron flux limit cycle oscillations are detected and suppressed, APRM and LPRM neutron flux noise levels should be monitored while operating in this region.

Stability tests at operating BWRs were reviewed to determine a generic region

=

of the power / flow map in which surveillance of neutron flux noise levels should be performed.

A' conservative decay ratio of 0.6 was chosen as the bases for 3

determining the generic region for surveillance to account for the plant to q

plant variability of decay ratio with core and fuel designs.-

This generic region has been determined.to correspond to a core flow of less than or equal to 45% of rated core flow and a THERMAL POWER greater than that specified in Figure 3.4.1.1-1.

Plant specific ca.lculations can be performed to determine an applicable region for monitoring neutron flux noise levels.

In this case the degree of conserva-tism can be reduced since plant to plant variability would-be eliminated.

In this case, adequate margin will be assured by monitoring the region which has a decay ratio greater than or equal to 0.8.

Neutron flux noise limits are also established to ensure early detection of limit cycle neutron flux oscillations.

BWR cores typically operate with' neutron-flux noise caused by random boiling and' flow noise.

Typical-neutron flux noise levels of 1-12%

of low to high r $ rated power.(peak-to peak) have been reported for the range ecirculation loop flow during both single and dual recirculation loop operation.

Neutron flux noise'1.evels which significantly bound these values are considered in the thermal / mechanical design of GE BWR fuel and are found to be of negligible consequence.

In addition,~ stability tests at operat-ing BWRs have demonstrated that when stability related neutron flux limit cycle oscillations occur they result in peak-to peak neutron flux limit cycles of 5-10 times the typical values.

Therefore, actions taken to reduce neutron flux noise levels exceeding three (3) times the typical value are sufficient to ensure early detection of limit cycle neutron flux oscillations.

Typically, neutron flux noise levels show a gradual increase in absolute magni-tude as core flow is increased'(constant control _ rod pattern) with two reactor recirculation loops in operation.

Therefore, the baseline neutron flux noise level obtained at a specific core flow can be applied over a range of core flows.

To maintain a reasonable variation between the low flow and high flow end of the flow range, the range over which a specific beseline is applied should not exceed 20% of rated core flow with two recirculation loops in operation.

Data from tests and operating plants indicate that a range of 20% of rated core flow will result in approximately a 50% increase in neutron flux noise level during operation with two recirculation loops. Baseline data should be taken near the maximum rod line at which the majority of operatiTm will occur.

However, baseline datf taken at low rod lines (i.e. lower power) q-will result in a conservative value since the neutron flux noise level is proportional to the power level at a given core flow.

CLINTON - UNIT 1 8 3/4 4-2 Amendment No. 18

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Page 21 of 38 R_EACTOR COOLANT SYSTEM

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BASES 3/4.4.1 RECIRCULATION SYSTEM (Continued)'

4 The recirculation flow control valves provide regulation of individual recir-culation loop drive flows; which, in turn, will vary the flow' rate of coolant through the reactor core over a range consistent with the rod pattern and re-circulation pump speed.

The recirculation flow contro1' system consists of the electronic and hydraulic components necessary for the positioning of the two hydraulically actuated flow control: valves.

Solid state control-logic will generate a flow control valve " motion inhibit" signal-in response to any one of several hydraulic power unit or analog control circuit failure signals.

The " motion inhibit" signal causes hydraulic power unit shutdown and hydraulic isolation such that the flow cor.crol valve fails "as is."

This design feature insures that the flow control Talves do_not respond to potentially erroneous control signals.

Electronic' limiters exist.in the position control loop of each flow control valve to limit the flow control valve stroking rate to 1011% per _second in opening and closing directi'ons on a control signal failure.

The analysis of the recirculation flow control failures on increasing and decreasing flow are.

presented in Sections -15.3 and 15.4 of the*

tively.

The required surveillance interval is adequate t ure thatithe flow control valves remain OPERA 8LE and not so frequent astto cause excessive-year on the system components., -

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3/4.4.2 SAFETY / RELIEF VALVES The safety valve function of the safety / relief valves (SRV)-operate to prevent the reactor coolant system from being pressurized above the Safety Limit of L

1375 psig in accordance with the ASME Code.

A total of.ll 0PERABLE safety--

relief valves is required to limit reactor pressure to within ASME III allowable values for the worst case upset transient.1 Any combination of'S SRVs-operating in the relief mode and 6 SRVs operating in the safety mode is acceptable.

Demonstration of the safety-relief valve lift settings will occur only during shutdown and will be performed in accordance with the provisions of Specifica-tion 4.0.5.

l The low-low set system ensures that safety / relief valve discharges!are minimized for a second opening of these valves, following any overpressure transient.

This is achieved by automatically lowering the. closing setpoint of 5 valves ~and lowering the opening setpoint of 2 valves following the initial. opening.

In this way, the frequency and magnitude of the containment blowdown duty cycle is substantially reduced.

Sufficient redundancy is provided for the low-low set system such that faiTure of any one valve to open or close at its reduced set-point does not violate the design basis.

3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.3.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from t,he reactor coolant pressure bourtdary. hese i

CLINTON - UNIT 1 8 3/4 4-3 Amendment No.

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h Page 22 of 38 REACTOR COOLANT SYSTEM BASES 3/4.4. 5 SPECIFIC ACTIVITY-The limitations on the specific activity of the primary coolant ensure that the i

2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />' thyroid and whole body doses resulting from a main steam -line. failure out-side the containment during steady state operation will not exceed small-frac-tions of the dose guidelines of 10 CFR 100.

The values for the limits 'on speci-fic activity represent interim limits based upon a parametric evaluation by the NRC 'of typical site locations.

These values are conservative. in that specific-site parameters, such as site boundary location and meteorological conditions, were not considered in this evaluation..

The ACTION statement permitting POWER OPERATION to continue for limited ~ time i

periods with the primary coolant's specific activity greater than 0.2 micro-I curies per gram DOSE EQUIVALENT I-131, but less than or equal to 4.0_microcurtes 1

per gram DOSE EQUIVALENT I-131, accommodates possible lodine _ spiking =.xrdr wh may ocrur followins shanges in_THE_RMAL P0]IER.( w m o.vn wm. l.' k

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.a Closing tTie main steam line isolation valves' prevents the release of: activity-

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to the environs should a steam line rupture occur outside containment.

2 The surveillance requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in. sufficient time to take corrective action.

3/4.4.6 PRESSURE / TEMPERATURE LIMITS

%AR.

All components in the reactor coolant system are d ! signed to withstand the effects of cyclic loads due to system temperature and pressure changes. These-cyclic loads are introduced by normal. load transielts, reactor trips, and start-up and shutdown operations. The various categorie load cycles used for r

design purposes are provided in Section 3.9 of thd During startup and the rates of temperature and pressure changes.are limited so that the-l.

shutdown maximum s,pecified heatup and cooldown rates are consistent with the design assumptions-and satisfy the stress limits for cyclic operation.

The operating limit curves of Figure 3.4.6.1-1 are derived from the fracture toughness requirements of 10 CFR 50 Appendix G and ASME Code Section III, i

Appendix G.

The curves are based on the RTNOT and stress intensity factor o

information for the reactor vessel components. Fract e toughness limits and the basis of compliance are more fully discussed ig subsection 5.3.1.5 l

entitled " Fracture Toughness."

CLINTON - UNil 1 B 3/4 4 5 Amendment No.18 1

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c Attachment-l' to U-601650-Page 23 of 38

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Cor te Min 3 AcTeoM EMERGENCY CORE COOLING SYSTEM NN BASES 3/4.5.1 and 3/4.5.2 ECCS - OPERATING AND SHUTDOWN (Continued)-

The capacity of the system is: selected to provide the required core cooling.

^

The HPCS pump is designed to deliver greater than or equal to 467/1400/5010 gpm at differential pressures of 1177/1147/200 psid.

Initially, water from the reactor core isolation cooling (RCIC) tank.is used instead of, injecting water from the suppression pool into-the reactor, but no credit is taken in the safety.

analyses for the RCIC tank water.

With the HPCS system inoperable, adequate core cooling:is assured by the OPERA-BILITY of the redundant and diversified automatic depressurization system and-both the LPCS and'LPCI systems.

In addition, the reactor core isolation cooling system, a system for which no. credit is taken in the safety analysis, will auto-matica11y provide makeup at reactor operating pressures on a reactor low water level condition. The HPCS out-of-service period of 14 days is based on the n

demonstrated nPERABilfTY of redundant and diversified low pressure core cooling.

systems Q Q E %

The su'rveillance requirements provide adequate as'surance that the HPCS system will be OPERABLE when required. Although all active components-are testcble and full flow can be demonstrated by recirculation through a test' loop during.

reactor' operation,'a complete functional test with reactor vessel injection requires reactor shutdown. The pump discharge piping is maintained full to-prevent water hammer damage.

Upon failure of the HPCS system to function properly after a small break loss-of-coolant accident, the automatic depressurization system (ADS) automatically causes selected safety-relief valves to open, depressurizing the reactor so

~

that flow from the low pressure core cooling systems can enter the core in time to limit fuel cladding temperature to less than 2200*F. ADS is conservatively required to be OPERABLE whenever reactor vessel pressure exceeds 100 psig.-

This pressure is substantially below that for which the low pressure core cool-ing systems can provide adequate core cooling for events. requiring ADS.

ADS automatically controls seven selected safety-relief valves although the safety analysis only takes credit for six valves.

It is therefore appropriate to permit one valve to be out of-service for up to 14 days without materially reducing system reliability.

3/4.5.3 SUPPRESSION POOL The suppression pool is required to be OPERABLE as part of the ECCS to ensure that a sufficient supply of water is available to the HPCS, LPCS and LPCI systems in the event of a LOCA. This limit on suppression pool mininum water volume ensures that sufficient water is available to permit recirculation cool-ing flow to the core. The OPERABILITY of the suppression pool in OPERATIONAL CONDITIONS 1, 2 or 3 is required by Specification 3.6.3.1.

CLINTON - UNIT 1 B 3/4 5-2

i Attach:2nt 1 Lto U 601650-Page 24 of 38 Insert for Pane B 3/4 5-2

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Because of this redundancy and diversity, the RCIC system may-be considered OPERABLE, for the purposes of satisfying the ACTION-requirements-for HPCS, when reactor pressure is. less than or equal to 150 psig, even though the RCIC system may.not;be considered OPERABLE per Specification-3.7.3 1

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i Page 25.of 38 CONTAINMENT SYSTEMS 1

. BASES 3/4.6.1.4 MSIV: EAKAGE CONTROL SYSTEM'

_Calaulated doses resultine from the maximum leakage allowancefor the main Cst /j35)1ineisolation.valvesinthepostulatedLOCA.situationswould.beasmall-j Traction of the 10 CFR 100 guidelines,L provided the main steam line system from j

the isolation valves up to and including the MSIV-LCS motor operated boundary

.i valve: remains intacto, Operating experience has indicated that degradation has l!

occasionally occurred in tne leaktightness of the MSIV's such that the specified. -

i leakage requirements have not always-been maintained-continuously.

The require-l ment for the leakage control system will reduce the untreated leakage from the-

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MSIV's when isolation of the primary system and containment is required.-

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3/4.6.1.5~ CONTAINMENT STRUCTURAL INTEGRITY l

This limitation ensures that the structural-integrity of the containment wil1~

be maintained comparable to the original design standards-for the life of-the i

unit.- Structural integrity is required to ensure that the; containment will withstand the maximum pressure of 15 psig in the event of a. steam line break-accident.

A visual inspection in conjunction with Type:A leakage tests is.

sufficient ~to demonstrate this capability.

3/4.6.1.6 CONTAINMENT INTERNAL PRESSURE, t

3 The limitations on containment to secondary containment differential pressure 2

ensure that the containment peak calculated pressure of 9.0 psig does,not exceed l

the design pressure-of 15.0 psig during design-basis steam line break conditions or that the external pressure differential.does not' exceed the design maximum' external pressure differential of 3.0 psid.- The limit of -0.25 to +0.25. psid I

for initial containment to, secondary containment pressure w'ill limit.the l

containment pressure to 9.0 psid which is lesslthan the design pressure and is l

consistent with the safety analysis for containment design pressure.

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3/4.6.1.7 PRIMARY CONTAINMENT AVERAGE AIR TEMPERATURE The limitation on containment average air temperature ensures that the contain-ment peak air temperature does not exceed the design temperature of 185'F during l

steam line break conditions and is consistent with the safety analysis.

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_ CONTAINMENT SYSTEMS 1

I BASES 3/4.6.1.8 CONTAINMENT-BUILDING VENTILATION AND PURGE SYSTEMS i

The 36-inch containment purge supply and exhaust isolation valves have permanently-l installed blocking devices to restrict their opening to' 50' during plant OPERA-i 110NAL CONDITIONS 1, 2 and-3, since these valves have not been demonstrated:

I capable of closing from the full open position during an accident.

Maintaining

.l these valves blocked during plant operations ensures that excessive.-quantities of radioactive materials will not be released via'the containment purge system.

To provide assurance that the 36-inch valves cannot be inadvertently fully opened, they are blocked in accordance with staf f's recommendations accepted in a

SSER S,, paragraph 6.2.4.1.

q The use of the containment purge lines is restricted to the 12-inch purge supply and exhaust isolation valves since, unlike the 36-inch valves, the 12-inch valves _

i close during accident conditions and therefore the site boundary-dose guidelines of 10 CFR Part 100 would not be exceeded in the event of an accident during purg-ing operations.

The design of the 12-inch purge supply.and exhaust isolation valves meets the requirements of Branch Technical. Position CSB 6-4, " Containment Purging During Normal Plant Operations."

D4erO The use of the 12-ioch cuaT_nment purAe exhaust and suoply lines shall be 'in i

accorda_nce_g; _;;r;{.l O,G,,%,, %G (Or',) "LL,,,7;J a ii..,

T., 0.c 5;ht.t Turg h n" pr;;id:d.in Il'in; b P:;;; (!P) L:tt:r U ^?21, d;t:d f a m e = =_k m. in ihoA TL...

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sanou s Continuous containment purge using the 36-inch containment building ventilation system is limited to only OPERATIONAL CONDITIONS 4:and 5.

Intermittent use of the 36-inch system during OPERATIONAL COND2TIONS 1, 2, and-3 is-permitted only for the purpose of reducing airborne activity levels, or containment pressure, and-atmosphere control (excluding tKalure and humidity), and shall.-i pi '^^ '=-

' = - r ?SS dry:p g,mHe3 io &c Mme,TimW4 den m Specif 4.dienJ.4.l.BD

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eaK8ge~ integrity'teits with a maximum allowable leakage rate for 36-inch supply and exhaust isolation valves will provide early indication of resilient material

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k seal degradation and will allow the-opportunity for repair before gross leakage

-l failures develop.

The,0.60 La leaking limit should not be exceeded when the 1

CLINTON - UNIT 1 B 3/4 6-3 Amendment No. 7 i

Att0 hment 1 to U.601650 Page 27 of 38 1Dagrt for Pare B 3/4 6 3

• Clinton Power Station Report on Containment-Purge Operational Data Cathering and Evaluation Program and Proposed Containment Purge Criteria" provided in Illinois Power (IP) letter U.601410, dated April 4, 1989. Section 6 of the report provides the criteria for governing operation of the containment building ventilation (36 inch) and the continuous purge (12 inch) systems.

The criteria balance ALARA dose to the worker with protection of the health and safety of the public.

Since continuous operation of the 12 inch containment purge system has been shown to be required to support access to the containment to perform Technical Specification surveillances during normal conditions while in OPERATIONAL CONDITIONS 1, 2 and 3, continuous operation of=the 12 inch system is allowed except while venting the drywell for pressure control or while operating the 36 inch system._

l l

Attochment 1 to U0601650 Page 28 of 38 CONTAINMENT SYSTEMS BASES I) 3/4.6.2.4 ORWELL STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the drywell will be maintained comparable to the original design specification for the life of the unit.

AvisualinspectioninconjunctionwithTypeAleakagetestsissuffi-cient to demonstrate this capabil1ty.

3/4.6.2.5 ORWELL INTERNAL PRESSURE Q The limitations on drywell-to-containme1t differential pressure ensure that the drywell peak calculated pressure of 19, 7 psig d6es not exceed the design pressure of 30.0 psig and that the containment p the design pressure of 15.0 psig durinn $ g

@ t a conditioni.eak pressure o Q 0 psig does not ex The maxi q g g 3 mumexternaldrywellpressuredifferenUa s tim'ted to 0.2 psid,nwen Aelow %

thepressureatwhichsuppressionpoolwaterwillbeforcedoverthe@@Ito and into the drywell.

The limit of 1.0 psid for initial positive drywe lwall g yi;, -

containment pressure will limit the drywell pressure to 19.7 psid which is less than the design pressure and is consistent with the safety analysis to limit drywell internal pressure.

3/4.6.2.6 ORWELL AVERAGE AIR TEMPERATURE The limitation on drywell average air temperature ensures that peak drywell I

')

temperature does not exceed the design temperature of 330'F during LOCA condi-tions and is consistent with the safety analysis.

3/4.6.2.7 ORWELL VENT AND PURGE The drywell purge system must be normally mcintained closed to eliminate.a potential challenge to containment structural integrity due to a steam bypass of the suppression pool.

Intermittent vent!q of the drywell is allowed for pressure control durino OPERATIONAL CONDITIONS 1, 2, and 3, but the cumulative time of venting is limited to 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> per 365 days.

Venting of the drywell is prohibited when the 12-inch continuous containment purge system or the 36-inch containment building ventilation system supply or exhaust valves are This eliminates any resultant direct leakage path from the drywell to osen.

tie environment.

In OPERATIONAL CONDITIONS 1 2 and 3 the drywell isolation valves (IVQ002, IVQ003)havepermanentlyins,talledblockingdevicessoasnottoopenmore than 50'.

Inis assures that the valve would be able to close against drywell pressure buildup resulting from a LOCA.

Operati of th

,r3al.

nt and purge 24-inch supply and exhaust valves during plant (ope a condition 4 and 5 is unrestricted, and the cumulative time for ve7n 1 n purgeiTp'ent n is unlimited.

LoilCAPS i

6 CLINTON - UNIT 1 B 3/4 6-5 Amendment No.18 y

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L Attcchment 1 to U-601650 Page 29 of 38 I

CONTAINMENT SYSTEMS l

I BASES 3/4.6.3 DEPRES$URIZATION $YSTEMS-The specifications of th s section ensure that the drywell and containment pres-sure will not exceed the design pressure of 30 psig and 15 psig, respectively, during primary system blowdown from full operating pressure.

The suppression pool water volume must absorb the associated decay and structural sensible heat released during a reactor blowdown from 1040 psia. Using conser-vative parameter inputs, the maximum calculated containment pressure during and following a design basis accident is below the containment design pressure f 15 psig. -Similarly the drywell pressure r.emains below the design pressure of 30 psig. The maximum and minimum water volumes for the suppression pool are 150,230 cubic feet and 146,400 cubic. feet, respectively. These values include the water volume of the containment pool, horizontal vents, and weir annulus.

Testing in the Mark'III Pressure Suppression Test Facility and analysis have assured that the suppression pool temperature will not rise above 185'F for the full range of break sizes.

Should it be necessary to make the suppression pool inoperable, this shall only

/,i be done as specified in Specification 3.5.3.

l Expe'rimental data indicates that effective steam condensation without excessive load on the containment pool walls will occur with a quencher device and pool temperature below 200*F during relief valve operation. Specifications have been '

placed on the envelope of reactor operating conditions to assure the bulk pool temperature does not rise above 185'F in compliance with the containment struc-tural design criteria..-

In addition to the limits on temperature of the suppression pool water, op6 tat-ing procedures define t;.e action to be taken in the event a safety-relief valve inadvertently opens or sticks open. As a minimum this action shall include:

(1) use of all available means to close the valve, (2) initiate suppression pool water cooling, (3) initiate reactor shutdown, and (4) if other safety-relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open safety' relief valve.to assure mixing and uniformity of energy insertion to the pool.

The containment' spray system consists of two 100% capacity trains, each with two. spray rings located at different elevations about the inside circumference of the containment. RHR A pump supplies one train and RHR pump B supplies the RHR pump C cannot supply the spray system. Dispersion of the flow of other.

n5M water is effected by 251 nozzles in each train, enhancing the N M water vapor in the containment volume and preventing overpress$izaY leat rejection is through the RHR heat exchangers. The turbulence caused by the spray system aids in mixing the containment air volume to maintain a itgpogeneous mixture for H control.

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Page 30 of 38

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CONTAINMENT SYSTEMS t

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BASES 3/4.6.3 DEPRESSURIZATION SYSTEMS (Continued)

The suppression pool cooling function is a mode of the RHR syst'em and functions as part of the containment heat removal system. The purpose of the system is to ensure containment integrity-following a LOCA by preventing excessive con-tainment pressures and temperatures. The suppression pool cooling mode is designed to limit the long term bulk temperature of the pool to 185'F consider-i ing all of the post-LOCA energy additions. The suppression pool cooling trains, being an integral part of the RHR system, are redundant, safety-related component systems that are initiated following the recovery of the reactor vessel watar level by ECCS flows from.the RHR system. Heatrejectiontothestandbyservice..

water is accomplished in the RHR heat exchangers.

i The suppression pool make-up system provides water from the upper containment-i pool to the suppression pool by gravity flow through two 100% capacity dump lines following a LOCA. The quantity of water provided is sufficient to account for all conceivable post-accident entrapment volumes, ensuring the long ters energy sink capabilities of the suppression pool and maintaining the' water cover-age over the uppermost drywell vents. The minimum freeboard distance above the suppression pool high water level to the top of the weir wall is adequate to s

preclude flooding of the drywell in the event of an inadvertent dump. During refueling, neither automatic nor manual action can open the make-up dump valves, i (s j

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P 3/4.6.4 CONTAINMENTISOLATIONVALVES The OPERABILITY of the containment isolation valves ensures that the contain-l ment atmosphere will be isolated from the outside environment.in the event of a release of radioactive material to the containment atmosphere or pressuriza,,,

tion of the containment and is consistent with the requirements of GDC 54-through 57 of Appendix A to 10 CFR $0 "and ti quirements of NUREG-0660 as clarified by HUREG-0737 as described in th Appendix 0,' item II.E.4.2 (Containment Isolation Dependability)."

M'essurement o'f the closure time of automatic containment isolation valves is performed for the purpose of demonstrating PRIMARY CONTAINMENT INTEGRITY and system OPERABILITY (Specification 3/4.6.

The Haximum Isolation Times (MIT) for p ntainment automatic isolation valves listed in this specification are r the analytical times used in the accident analysis; described in the or times derived by applying margins to the test data obtained by pe,r ng testing in accordance with.the Inservice Testing program (IST) outlined in Section XI of the ASME Code. For non-analytical automatic primary containment isolation valves, the*MIT is de-i I

rived as follows:

1)

Valves with full stroke times less than or equal to 10 seconds, MIT = Initial Base Line Time X 2 2)

Valves with full stroke tims greater than 10 seconds, MIT = InTtial f

Base Line Time X 1.5.

CLINTON - UNIT 1 B 3/4 6-7 4

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' CONTAINMENT SYSTEMS BASES i

3/4.6.5 DRYWELL POST-LOCA VACUUM RELIEF VALVES Drywell vacuum relief valves are provided on the drywell to pass sufficient quantities of gas from the containment to the drywell to prevent an excess negative pressure from developing in the drywell.

3/4.6.6 SECONDARY CONTAINMENT The secondary containment completely encloses the primary containment, except for the upper personnel hatch.

It consists of the fuel building, gas control boundary, and portions of the auxiliary building enclosed by the extension et the gas control boundary an.d the ECCS cubicles and areas as described irffeAUSA Figure 6.2-132.

The standby gas treatment system (SGTS) is designed to a77.rev and maintain a negative 1/4' W.G. pressure within the secondary containment following a design basis accident.

This design provides for the capture within the secondary containment of the radioactive releases from the primary contain-ment, and their filtration before release to the atmosphere.

Establishing and maintaining a vacuum in the secondary containment with the standby gas treatment system once per 18 months, along with the surveillance of the doors, hatches, dampers and valves, is adequate to ensure that there are no violations of the' integrity of the secondary containment.

The inleakage values are not verified in the surveillances since no credit for dilution was taken in the dose calculation.

As noted however, adequate drawdown is i

verified once per 18 months.

The acceptance criteria specified in Figure 4.6.6.1-1 for the drawdown test is based on a computer model, verified by actual performance of drawdown tests, in which the drawdown time determinedforaccidentconditionsisadjustedtoaccountforperformanceof the test during normal plant conditions. The acceptance criteria indicated per Figure 4.6.6.1-1 is based on conditions corresponding to power operation (with the turbine building ventilation system in operation) and wind speeds-less than or equal to 10 mph.

The acceptance criteria for plant conditions other than those assumed will be adjusted as necessary to reflect the conditions which exist during performance of the surveillance test.

The OPERABILITY of the standby gas treatment systems ensures that sufficient

(

iodine removal capability will be available in the event of a LOCA.

The reduc-tion in containment iodine inventory reduces the resulting site boundary radia-tion doses associated with containment leakage.

The operation of this system and resultant iodine removal capacity are consistent with the assumptions used in the LOCA analyses.

Continuous operation of the system with the heaters OPERABLE for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> during each 31-day period is sufficient to reduce the buildup of moisture on the absorbers and HEPA filters.

3/4.6.7 ATHOSPHERE CONTROL The OPERABILITY of the systems required for the detection and control bf hydr 6 gen gas ensures that these systems will be available to maintain the hydrogen con-centration within the containment below its flammable limit during post-LOCA l

CLINTON - UNIT 1 B 3/4 6-8 Amendment No. 21

4 Attechm2nt 1 1

to U 601650 l

Page 32 of 38 3/4.7 PLANT SYSTEMS i

BASES i

3/4.7.1 SHUTDOWN SERVICE WATER SYSTEM The OPERABILITY of the shutdown service water system ensures that sufficient cooling capacity is available for continued operation of safety-related equip-ment during accident conditions.

The redundant cooling capacity of these sys-tems, assum{ng a_ single failure, is consistent with the assumptions used in the acciden+ t ii W O within_ acceptable limits.

l The ultim e heat s specification ensure that sufficient cooling capa-is available for continued operation of safety #related equipment for at 30 days to permit safe shutdown and cooldown of the reactor.

The surveil-l..,

ca lance requirements ensure that quantities maintained are consi ent with the assumption used in the accident analysis as described in the s

and the guid-l ance provided in Regulatory Guide 1.27, January 1976.

g 3/4.7.2 CONTROL ROOM VENTILATION SYSTEM The OPERA 81LITY of the control room ventilation system ensures that 1) the ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment and instrumentation cooled by this system and i

2) the control room will remain habitable for operations personnel during and following all design basis accident conditions. Continuous operation of the system with the heaters 0PERABLE for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> during each 31 day period is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters.

The OPERABILITY of this system in conjunction with control room design provi-P sions is based on limiting the radiation exposure to personnel occupying the I

control room to 5 rem or less whole body, or its equivalent.

This limitation is consistent with the requirements of General-Design Criterion 19 of Appendix "A",10 CFR 50.

Surveillance testing provides assurance that system and component performances continue to be in accordance with performance speci-fications for Clinton Unit 1, including applicable parts of ANSI N509-1980.

l 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM i

The reactor core isolation cooling (RCIC) system is provided to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel without requiring actuation of any of the Emergency Core Cooling System equipment.

The RCIC system is conserv -

atively required to be OPERABLE whenever reactor pressure exceeds 150 psig.

This pressure is substantially below that for which the. low pressure core cool-ing sys

_cartprovide adequate core cooling for events requiring the RCIC system.

CRT.)

The RCIC system specifications are applicable during OPERATIONAL GONDITIONS 1,

@nd 3 when reactor vessel pressure exceeds 150 psig because RCIC'is the,,pri-l W ry (non-ECCS) source of emergency core cooling when the reactor is pressurized.

With the RCIC system inoperable, adequate core cooling is assured by the OpElf-

.f BILITY of the HPCS system and justifies the specified 14 day out-of-service i

period.

CLINTON - UNIT 1 8 3/4 7-1

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Page 33 of.38 l

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Although the RCIC system suiy not_ be considered OPERABLR per-i Specification 3.7.3, the RCIC system may be considered OPERABLE for the purposes of satisfying the ACTION requirements of other Specifications i

with reactor pressure less than or equal to 150 peig.

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Attechment 1 l

to U-601650 -

Page 34 of 38 PLANT SYSTEMS BASES 3/4.7.4 SNUBBERS (Continued)

The requirement to monitor the snubbet service life is included to etc....

ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions. These records will provide statistical bases for future consideration of snubber service life. The requirements for the maintenance of records and the snubber service life review are not intended to affect plant operation.

3/4.7.5 SEALED SOURCE CONTAMINATION The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium.

This limitation will ensure that leakage from byproduct, source, and special nuclear material sources will not exceed allowable intake values. Sealed sources are classified into three groups according to their use, with surveil-lance requirements commensurate with the probability of damage to a source in that group. Those sources which are frequently handled are required to be tested more often than those which are not. Sealed sources which are continu-ously enclosed within a shielded mechanism, i.e., sealed sources within radia-tion monitoring devices,'are considered to be stored and need not be tested unless they are removed from the shielded mechanism.

3/4.7.6 MAIN TURBINE BYPASS SYSTEM 9

The main turbine bypass system is required to be OPERABLE con i tent with the as6umptions of the feedwater controller failure analysis 1 Chapter 15.

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CLINTON - UNIT 1 8 3/4 7-3

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a tachment 1 SpcCdiCO i o i 38 ctppucabu, AcTeok) SWemenM i

8 3448 ELECTRICAL POWER SYSTEMS BASES,

3/4.8.1.,3/4.8.2. AND 3/4.8.3 AC SOURCES. DC MURCES AND ONSITE POWER -

DISTRIBUTION SY$,T[MS The OPERABILITY of the AC and DC power sources and associated distribution systems during operation ensures that sufficient power will be available to supply the safety related equipment required for (1) the safe shutdown of the facility and (2) the mitigation and control of accident conditions within the facility. The ainimum specified independent and redundant AC and DC power sources and distribution systems satisfy the requirements of General Design Criteria 17 of Appendix "A" to 10 CFR 50.

i The ACTION requirements specified for the levels of degradation of the power i

sourect provide restriction upon continued facility operation commensurate with the level of degradation. The OPERABILITY of the power sources are consistent with t% initial condition assumptions of the accident analyses and are based upon maintaining at least Division I or II of the onsite AC and DC power sources and associated distribution systems OPERABLE during accident conditions coinci-dent with an assumed loss of offsite power and single failure of the other onsite AC source. Division III supplies the high pressure core spray (HPCS) system only."

TThe AC and DC source allowable out-of-service times are based on Regu1 Qr n

L Guide 1.93. "Availabjlity of Electrical Power Sources." December 1974 19When

_ ggt diesel generator 1A or IB is inoperable, there is an additional ACTIOR require-ment to verify that all required systems, subsystems, trains, components and devices, that depend on the remaining OPERABLE diesel generator 1A or 1B as a source of emergency power, are also OPERABLE. This requirement is intended to i

provide assurance that a loss of offsite power event will not result in a com-l plete loss of safety function of critical systems during the period diesel generator 1A or 1B is inoperable. The term verify as used in this context means to administrative 1y check by examining logs or other information to determine if certain components are out-of-service for maintenance or other reasons.

It does not mean to perform the surveillance requirements needed to demonstrate the OPERABILITY of the component, j

The OPERABILITY of the minimus specified AC and DC power sources and asso-ciated distribution systems during shutdown and refueling ensures that (1) the facility can be maintained in the shutdown or refueling condition for extended time periods and (2) sufficient instrumentation and control capability is avail-able for monitoring and maintaining the unit status.

l l

The surveillance requirements for demonstrating the OPERABILITY of the diesel

]

l generators are in accordance with the recommendations of Regulatory Guide 1.9 -

" Selection of Diesel Generator Set capacity for Standby Power Supplies",

March 10, 1971, Regulatory Guide 1.108, " Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants," Revision 1 August 1977 and Regulatory Guide 1.137 " Fuel-011 Systems for Standby Diesel Generators," Revision 1, October 1979.

l l

CLINTON - UNIT 1 8 3/4 6-1 i

Atte:hment 1 ta U.601650 Page 36 of 38 Insert for Page B 3/4 8-1 i

With one diesel generator inoperable, the ACTION statements require the remaining diesel generators to be demonstrated OPERABLE by performing Surveillance Requirements 4.8.1.1.2.a.4 and 4.8.1.1.2.a.5, unless the diesel generator became inoperable due to preplanned preventive maintenance or testing.

In this context, preplanned preventive maintenance is any maintenance which if not performed would not, in itself, render.the diesel generator inoperable.

Preventive maintenance includes the correction of minor leaks that do not otherwise make the diesel generator inoperable.

If a second diesel generator is made inoperable solely to perform OPERABILITY testing as required by the ACTION statements, this second diesel generator need not be considered inoperable during the performance of this OPERABILITY testing.

If, however, a condition is discovered during the preplanned preventive maintenance or testing that affects the diesel generator's OPERABILITY, the diesel generator must be declared inoperable.

f I

Attechmnt 1 e

to U-601650 Page 37 of 38 ELECTRICAL POWER SYSTEMS BASES l

3/4.8.1, 3/4.8.2, and 3/4.8.3 AC. SOURCES,'DC SOURCES. AND ONSITE POWER DISTRIBUTION SYSTEMS (Continued).

I The surveillance requirements for demonstrating the OPERABILITY;of the unit batteries are in accordance with the recommendations of Regulatory Guide 1.129

" Maintenance Testing and Replacement of Large lead Storage Batteries for Nuclear Power Plants," February 1978, Regulatory Guide 1.32, " Criteria for Safety-Related Electric Power Systems for Nuclear Power Plants," February 1977, IEEE Std 450 1975, "IEEE Recommended Practice for Maintenance Testing, and Replace-mentofLargeLeadStorageBatteriesforGeneratingStatIonsandSubstations,"

and IEEE Std 308-1974 "1EEE ',tandard Criteria for Class IE Power Systemo far -

Nuclear Power Generating St:,tions" with exceptions noted in the 0^; TU."

Verifying average elect'olyte temperature above the minimum for which the battery was sized, total battery terminal voltage on float charge, connection resistance values ano the performance of battery service and discharge tests ensures the effectiveness of the charging system, the ability to handle high discharge rstes and compares the battery capacitvlt_.that time with that rat ed ennacity.

Wetu4 MN

@ 8ilLti MEMn er ied 4l.M(MD Table 4.5.2. b 1 specifies tr e nomal T mus Tor eam cengnateu pilot cell f.

each connecte d cell for elet trolyte level float voltage and specific gravit The limits fc r the designata d pilot celW, float voltage and specific gravity.

l 2.13 volts and

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for each connected cell for float voltage and specific gravity, greater than,S 2.13 volts and ret : : i M 020 b:!r: the r: '::tur:r': ';!' chtr;: :;::t O%N *$

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OPERABILITY and canabil_iti of the_balte_ry reusc. MUK o t j(46 r

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Operation with@a u.-y WoleiNeWhito I5h4nd) lia.at Dus wnm" 41 are M.Va tidiA i

r yu..mer cumae w. uvrm :1 facommended the allowable value specified in Table 4.8.2.1-1 is permit;er' for up to 7. days.

During this 7 day period:

(1) the allowable values for eltetrolyte level ensures %4 ma%w%(;

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j no physical damage to the plates wit adequate electron transfer capability:

(2) the allowable value for the avera specific gravity o' all the cellg-met u -

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(4) the allowable value for an individual cell's float vo tage, greater than y ts ensures the battery's capability to perform i'.s desigri function.

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CTRICAL EQUIPMENT PROTECTIVE DEVICES ee mangsackers % enswe, Containment electrical penetrations and penetration conductors are protected - - - -- --

J by demonstrating the OPERABILITY of primary and backup overcurrent protection CLINTON - UNIT 1 B 3/4 8-2 k

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Attcchment 1 to U-601650 Pcg) 38 cf 38 REFUELING OPERATIONS BASES 3/4.9.12 INCLINED FUEL TRANSFER SYSTEM Thepurposeoftheinclinedfueltransfersystemspecification.is4 control personnel access to those potentially high_ radiation areas"@ adjacent to the system and to assure safe operation cf the system. j 4

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CLINTON

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PAst i or 2 SAFETY EVALUATION SCREENING 1.1 Document Evaluated Punenn CHAwnts To Tucuwicht S Ptet ricATION Batti This form is used to document the justification for not performing a full CNP 1.09 safety evaluation for design or document changes, tests, and experiments.

If all the questions can be answered "NO" with documented justification, then a full safety evaluation is not required. This screening form is to be vaulted with the document evaluated. A copy of the form and the document evaluated shall be sent to Licensing and Safety.

I 2.1 1s this a change co the facility as described in the SAR Yss (That is, does it result in any condition (including No i qualifications), operation, analysis result, or function Don't Know contrary to the current SAR descriptions)?

Applies regardless of the safety classification of the item being a.

changed.

b.

Applies whether the specific item being changed is identified in the SAR or not, Applies even if no hardware is being changed, but the plant does not c.

match the SAR description in some way.

(Note: minor clerical corrections to the SAR such as valve number corrections may not require safety evaluations.)

2.2 Is this a change to a procedure as described in the SARI Yes (That is, is any system or component operated or is any No d organization function performed in any way contrary to a Don't Know description in the SAR or assumed in any SAR analysis)?

(Includes changes to acceptance criteria, setpoints or commitments described in the SAR) 2.3 Is this a test or experiment not described in the SARI Yes (That is, is any system or component operated in any way No v' i

contrary to a description in the SAR or assumed in Don't Know 1

any SAR analysis)?

(Past2or2)

3.1 Justification

SEE A MAcutb i

i Note 1: The "SAR" includes all referenced documents addressed in the Clinton SAR, such as the Emergency plan, the Security Plan, and letters on the Clinton docket supporting any analyses.

Originator:

T.B.Etwoon bd).b~

G /490 Name Si, nature Date Director:

b.2i b

Name

'Signatbre _

J//RNO 0

2 if er bate Upon completion, this screening form shall be vaulted with the document evaluated. A copy, with a copy of the document evaluated, shall be forwarded to Supervisor Technical Assessment, Licensing and Safety, V 920.

l FORM SE 1 1-Revision 1 10/89 i

,e S:foty Evaluatien Scre:ning l

Page 2 of 2 j

j Although the proposed changes are to the Bases for the applicable Technical Specifications, the Bases themselves are not part of the Technical Specifications (as stated in 10CFR50.36).

The proposed 4

changes are not therefore subject to the requirements of 10CFR50.90 and therefore do not require a review for sigr.ificant hazards consideration.

With respect to the requirements of 10CFR50.59, it has been determined that the proposed changes are subject to review for a safety evaluation screening.in accordance with the Clinton Power-Station Safety Evaluation Program.- 2ndustry guidance concerning safety consideration that should be given to proposed changes-to t.he Technical Specification Bases is currently under development.

Notwithstanding, the proposed changes to the Bases are described in the attached letter, U.601650. Justification is provided for each l

change.

It should be noted that the proposed changes were prepared such that consistency between the Technical Specification Bases, the CPS Technical Specifications, and the CPS USAR is maintained <

therefore Items 2.1, 2.2 and 2.3 of the Safety Evaluation Screening Form are all checked "No", and no proposed changes to the USAR are associated with these proposed Bases changes.

It is also noted in the attached letter, where applicable, that the proposed changes to the Bases do not constitute a significant reduction in a margin of safety since the margins of safety are preserved by the Technical Specifications themselves, t

3 RFP7:TBE1