ML20079C735

Proposed Tech Specs,Replacing MAPLHGR Limit Factor 0.85 for Single Recirculation Loop Operation W/Ref to Core Operating Limits Rept & Correcting Number of Nozzles in RHR Containment Spray Train a

ML20079C735
Person / Time
Site:	Clinton
Issue date:	06/17/1991
From:	ILLINOIS POWER CO.
To:
Shared Package
ML20079C731	List: ML20079C735 U-601848, Requests Change to Tech Spec Bases to Replace Maplhrg Limit Factor 0.85 for Single Recirculation Loop Operation W/ Ref to Core Operating Limits Rept & Correcting Number of Nozzles in RHR Containment Spray Train a
References
NUDOCS 9106260007
Download: ML20079C735 (4)

Text

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l Attachment 1 j to U-601848 Page 3 of f, POWER DISTRIBUT,I.O_N._t lM115 BASES _

3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (Continued)

b. Model Change

1. Core CCFL pressure differential - 1 psi - Incorporate the assumption that flow from the bypass to lower plenum must overcome a 1 psi pressure drop in core.

2. Incorporate NRC pressure transfer assumption - The assumption used in the SAFE-REFLOOD pressure transfer when the pressure is increasing was changed.

A few of the changes af fcct the accident calculation irrespective of CCFL.

These changes are listed below,

a. Input Change

1. B eak Areas - The bMA break area was calculated more accurately.

b. Model Change

1. Improved Radiation and Conduction Calculation - Incorporation of CHASTE 05 for heatup calculation.

A list of the significant plant input parameters to the loss-of coolant accident analysis is presented in Bases Table B 3.2.1-1.

For plant operation with a single recirculation loop, the MAPLHGR limits specified in the CORE O.PERAIlm_UMITS _PEPORT ara multiplied hvQsmallest of MAPFAC f , MAPf*C p opQ~?t 0: 'ic U . 1 .='. nt ':. tm E 6 oE,)is derived from LOCA analyses initiated from single loop operation to account for earlier boiling transition at the limiting fuel node compared to standard LnCA evaluations. '

3/4.2.2 APRM SETPOINTS [0ELETED]

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CLINTON - UNIT 1 B 3/4 2-2 Amendment No. 28 l

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Attachment 1 to U-601848 Page 4 of 6 j CONTAINMENTSYSTEM.$

8ASES 3/4.6.3 OEPRES$URIZATION SYSTEMS .

The specifications of this 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 of 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 be done as specified in Specification 3.5.3.

Expe'rimental data indicates that effective steam condensation without excessive load on the containment pool walls will occur with-a quencher device and pool 3 -

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, operat-ing procedures defind the 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 al_1 available means to close the valve, (2) initiate suppression pool i 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 8 supplies the other. RHR pump C cannot supply the_sprar system. Dispersion of the flow of

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water is er rectea ogE.sm" (Ur:O] enhancing the condensaton of Heat l'

water vapor in the c36tatnrant volume an'd preventing overpressurif ation.

The turbulence caused by the rejection is through the RHR heat exchangers.

spray system' aids-in mixing the containment air volume to maintain a homogeneous mixture for. H control. m-

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s Attachm2nt I to U-601848 i

. Page 5 of 6 I CONTAINMENT SYSTEMS BASES 3/4.6.5 ORYWELL POST-LOCA VACUUM RELIEF VALVES Drywell vacuum relief valves are provided on the dryell k 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 for the secondary containment upper personnel hatch. completely encloses the primary containment, except It consists of the. fuel building, gas control boundary, and portions of the auxiliary building enclosed by the extension of the gas6.2-132.

Figure control boundary and the ECCS cubicles and areas as described in FSAR and maintain a negative 1/4 Thestandbggastreatmentsystem(SGTS)isdesignedtoachieve following a design basis accident.W.G. pressure within the secondary containment 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 secondery containment with the standby gas treatment system once per 18 months, along with the surveillance of the docrs, 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 verified once per 18 months. The acceptance criteria specif t,td 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 determined for accident conditions is adjusted to account for performance of 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 (withthan less the turbine or equalbuilding to 10 mph. ventilation system in operation) and wind speeds 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 in theresultant iodine removal capacity are consistent with the assumptions used LOCA analyses.

CPERABLE 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 sufficiatContinuous operatio ta o duce t y G M5cd eb iMVLtuM of 'M6Vols. We%fWH4tbuildun

%,.,. mea we.iixv. nf moisturt Aw.hi-i 01 th 3/4.6.7 AIMOSFHERE CONTRO etchs h u W.ad p ad to Ac ac,g g g q g, The OPERABILITY of the systems required for the detection and control of hydrogen gas ensures that these systems will'be available to maintain the hydrogen con-centration within the containment below its flammable limit during post-LOCA CLINTON - UNIT 1 8 3/4 6-8 Amendment No. 21

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' Attachment 1

, to U-601848 Page 6 of 6 3/$.7 PLANT SYSTEMS BASES 3/4.7.1 SHUT 00WN 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 acciuent conditions. The redundant cooling capacity of these sys-tems, assuming a single failure, is consistent with the assumptions used in the accident concit: rs '*.hin acceptable limits.

[ The ultimate hest sir % (Jh5) specification ensure that sufficient cooling capa-city is available for continued operation of safety-related equipment for at leat'30 days to permit safe shutdown and cooldown of the reactor. The surveil-lance requirements ensure that quantities maintained are consistent with the assumption used in the accident analysis as described in the FSAR and the guid-ance provided in Regulatory Guide 1.27, January 1976

( 3/4.7.2 CONTROL ROOM VENTILATION SYSTEM i

The OPERABILITY uf the control room ventilation system ensures that 1) the ambient air temperature does not exceed the allowable temperature for continuous t duty rating for the equipment and instrumentation cooled by this system and

2) the control room will ramain habitable for operations rersonnel durina ant

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follcwing all desi$ basis _ accident _condQions, FCont1nuouc operation 6f thQ j syitem witW the heaters OPEIRABLE for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> du' ring each 31 day period is t, sufficient to red _uc_e the_ buildup of moisture on the_adsorbers and HEPA filters

70FERA6ILITY of this syst~em irccrnfunction~with control room design provi "

. sions is based on limiting the radiation exposure to personnel occupying the .

control room to 5 rem or less whole body, or its equivalent. Thiv, limitation i is consistent with the requirements of General Design Criterica U uf -

- Appendix "A", 10 CFR 50. Surveillance testing provides m urance that system and component performances continue to be in accordance with performance speci-ficat. ions for Clinton Unit 1, including applicable pa-ts of ANSI H509-1980.A 314,,.7.3REACTORCOREISOLATIONCOOLINGSYSTEM e reactor core isolation cooling (RCIC) system is provided to assure adequate s e cooling in the event of reactor isolation from its primary heat sink and

'ne 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-y ing systems can provide adequate core cooling for events requiring the RCIC system.

The RCIC system soecifications are applicable during OPERATIONAL CONDITIONS 1, 2 and 3 when reactor vessel pressure exceeds 150 psig because RCIC is the pri-mary (non-ECCS) source of errergency core cooling when the reactor is pressurized.

With the RCIC system inoperable, adequate core cooling is assured by the OPERA-BILITY of the HPCS system and justifies the specifled 14 day out-of-service I period.

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