AECM-87-0215, Forwards Addl Info for Proposed Tech Spec Changes Re Exceptions to Tech Spec 3.0.4 for Second Refueling Outage, Per 871030 Request.Proposed Methods for Complying W/Action Statements Consistent W/Methods Used at Other Plants

Forwards Addl Info for Proposed Tech Spec Changes Re Exceptions to Tech Spec 3.0.4 for Second Refueling Outage, Per 871030 Request.Proposed Methods for Complying W/Action Statements Consistent W/Methods Used at Other Plants

ML20236S505
Person / Time
Site:	Grand Gulf
Issue date:	11/19/1987
From:	Kingsley O SYSTEM ENERGY RESOURCES, INC.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
AECM-87-0215, AECM-87-215, NUDOCS 8711250218
Download: ML20236S505 (14)

Text

. --_- _.

l SYSTEM ENERGY  !

l RE50tJRCES, INC.

Omp D KtGMY JR

/EU$bms November 19, 1987 U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Document Control Desk Gentlemen:

SUBJECT:

Grand Gulf Nuclear Station Unit 1 Docket No. 50-416 License No. NPF-29 3.0.4 Exceptions for RF02 Proposed Amendment to the Operating License (PC0L-87/11 AECM-87/0200)

Additional Information AECM-87/0215

References:

1. System Energy Resources, Inc. letter (AECM-87/0128) dated July 6, 1987

2. NRC letter (MAEC-87/0276) dated October 30, 1987

3. System Energy Resources, Inc. letter (AECM-87/0200) dated October 23, 1987

4. NRC letter (MAEC-87/0273) dated October 20, 1987 System Ener Commission (NRC)gy Resources, Inc. (SERI) has met with the Nuclear Regulatory Station (GGNS) Technical Specification regarding selected exceptions to Technical Specification 3.0.4 for the second refueling outage (References 1, 2 and 3).

On October 30, 1987, the NRC staff transmitted a request for additional information (Reference 4) which requested clarification of the term " functional" and which requested specific information on alternate methods of decay heat removal and reactor coolant circulation to be used when no RHR loop is in operation. This letter and its attachments provide a response to that request.

The request for additional information and this response emphasize alternate methods used in complying with the action statements of Technical Specifications 3.9.11.1 and 3.9.11.2. The proposed methods are consistent with previous practice at GGNS and with the methods being utilized by other plants ,

of a similar general design.

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J16AECM87110501 - 1

AECM-87/0215 Page 2

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If additional information is needed to support your review, please contact this office.

Your. ruly,

/

ODK:bms Attachments cc: Mr. T. H. Cloninger (w/a)

Mr. R. B. McGehee (w/a)

Mr. N. S. Reynolds (w/a)

Mr. H. L. Thomas (w/o)

Mr. R. C. Butcher (w/a)

Dr. J. Nelson Grace, Regional Administrator (w/a)

U. S. Nuclear Regulatory Commission Region 11 101 Marietta St., N. W., Suite 2900 Atlanta, Georgia 30323 Mr. L. L. Kintner, Project Manager (w/a)

Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, Maryland 20814 I Dr. Alton B. Cobb (w/a)

State Health Officer State Board of Health Box 1700 Jacksoa, Mississippi 39205 J16AECM87110501 - 2

BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION l

LICENSE N0. NPF-29 l DOCKET NO. 50-416 IN THE MATTER OF MISSISSIPPI POWER & LIGHT COMPANY and SYSTEM ENERGY RESOURCES, INC.

and SOUTH MISSISSIPPI ELECTRIC POWER ASSOCIATION AFFIRMATION I, 0. D. Kingsley, Jr. , being duly sworn, stated that I am Vice President, Nuclear Operations of System Energy Resources, Inc.; that on behalf i of System Energy Resources, Inc., and South Mississippi Electric Power Association I am authorized by System Energy Resources, Inc. to sign and file '

with the Nuclear Regulatory Commission, this application for amendment of the Operating License of the Grand Gulf Nuclear Station; that I signed this 1 application as Vice President, Nuclear Operations of System Energy Resources, i Inc.; and that the statements made and the matters set for therein are true and correct to the best of my knowledge, informt. tion an ief.

STATE OF MISSISSIPPI COUNTY OF HINDS SUBSCRIBED AND SWORN T0 befor County and State above named, thisAay 8//e of me,f/Mem/ec a Notary Public, in, 1907.

and for the 2

(SEAL) .

ah) k). $'fko Notary Public' My commission expires:

D $$lSi[JTI5f3 @s 5 J10 MISC 87111901 - 1

Attachment 1 to .

AECM-87/0215 j NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING SERI SUBMITTAL DATED OCTOBER 23,1987(PCOL-87/11,AECM-87/0200)

Introduction By letter dated October 30, 1987 the NRC requested additional information based on the NRC staff review of exceptions to Technical Specification 3.0.4.

The following lists each NRC request and the SERI response.

1. NRC Request Define " functional" as used in the attachment to the October 23, 1987 submittal. For example, Paragraph C.2 states "...the intent of the SERI '

outage policy is to maintain at least one ECCS system and one Fuel Pool Cooling and Cleanup System functional at all time."

SERI Response The term " functional" involves assurance that the system can perform its intended safety function (i.e., ECCS can inject into the core at rated flow, Shutdown Cooling can maintain average reactor coolant temperature below Technical Specification limits, etc.). Some manual manipulation is j allowed such as closing breakers and realigning valves. Examples of situations where a system is functional but not operable are situations where there is adequate evidence that a system will otherwise perform its intended safety function but:

- the SSW basin contains less than a 30 day water supply.

- snubber surveillance are' incomplete and seismic qualification is not formally assured,

- Technical Specification response times (or other surveillance testing) is not current,

- room cooling is not available for all or part of the system.

2. NRC Request With respect to the alternate methods capable of decay heat removal and  !

reactor coolant circulation:

(a) Identify the specific alternate methods that are. planned to be used in the second refueling outage, including the times when exceptions to Technical Specification 3.0.4 may be used.

SERI Response During the period from November 30 through December 8, 1987, inclusive, Technical Specification 3.9.11.1 is applicable and requires one shutdown cooling system to be operable and in operation. Both shutdown cooling systems will be inoperable, requiring entry into the Action Statement J16AECM87110501 - 4

i Attachment I to I AECM-87/0215 3 l

of Technical Specification 3.9.11.1. This Action Statement requires an j alternate method of decay heat removal. The alternate raethod of decay l heat removal scheduled is the Fuel Pool Cooling and Cleanup (FPCCU) and i Reactor Water Cleanup (RWCU) systems used concurrently.

]

For the period from December 8, 1987 through December 123, 1987, inclusive, s two shutdown cooling systems are required to be operable with one in '

operation (Technical Specification 3.9.11.2). Over tMs period Residual Heat Removal (RHR) B is scheduled to be operable and in operation and RHR A '

is scheduled to be inoperable. The Reactor Water Cleanup and Control Rod Drive systems (concurrently) could be utilized as an alternate decay heat removal method as required by the Action Statement of Technical Specification 3.9.11.2.

Exceptions to Technical Specification 3.0.4 are scheduled to be utilized on December 8 and December 22, 1987. On December 8 SERI has scheduled-draining the cavity water level below the 22 feet 8 inch level to decontaminate the cavity, remove vibration instrument.ation and replace the (

reactor vessel head. This requires movement from Technical Specification 3.9.11.1 to Technical Specification 3.9.11.2. The Limiting Ccoditions for Operation are not met without relying on the Action Statement of Technical Specification 3.9.11.2 requiring an exception to Technical Specification {

3.0.4. '

Technical Specification 3.5.2 in conjunction with 3.0.4, requires two ECCS  ;

systems to be operable in order to drain the reactor cavity below 22 feet 8 inches. LPCI B is required to be made inoperable after the cavity draining in order to perform maintenance on a valve that can not be serviced with the cavity flooded. The proposed Technical Specification 3.0.4 exception to Spec 3.5.2 is required to prevent the evolution of declaring LPCI B operable to allow cavity draining, then declaring LPCI B inoperative to permit valve maintenance. This valve maintenance will not affect Shutdown Cooling B since the valve requiring maintenance can he isolated from the shutdown cooling loop flow path. ,

?

Additionally Water (SSW)y, during the cavity draining on December 8, Standby ServiceA is and inspections. Technical Specification 3.9.11.2 requires two shutdown .

cooling systems be operable which in turn requires two SSW systems be operable by Technical Specification 3.7.1.1. The Limiting Condition for i

Operation of Technical Specification 3.7.1.1 will not be met for the inoperable shutdown cooling loop described above (Specification 3.9.11.'2).

This 3.0.4 exception will be used at this time, since SSW B will be the only operable SSW loop.

A Technical Specification 3.0.4 exception is scheduled to be required on December 22. On this date reactor head studs are scheduled to be tensioned causing entry into Operational Condition 4. Technical Specification 3.4.9.2 requires two shutdown cooling mode loops of tht.

Residual Heat Removal (RHR) system to be operable in Operational Condition 4. The RHR A loop is not scheduled to be operabic at that time.

Note that the dates used above are based on the current schedule for  :

l RF02 and are subject to change.

J16AECM87110501 - 5

Attachment 1 to ,

.AECM-87/0215 l

NRC Request -

b l 2.(b) Describe the results of the ar,alyses to dewctrate adequate cooling of spent fuel while using these alternate methods, including spent fuel in the reactor vessel, upper containment pool and spent fuel pool. include a description of the assumptions and methods of analysis, including decay heat loads, water circulation paths within the reactor vessel and core, and the criterion for adeouate cooling.

SERI Response The assumptions, methods and results of the decay heat generation calculation for the end of Cycle 2 are included as Attachment 2. The ,

assumptions, methods and results of the calculation of alternate decay ,

heat removal capabilities are included as Attachment 3.

Fuel will be located in the Spent Fuel Pool and the Reactor Vessel dring the periods that alternate methods of decay heat removal will be utilized.

When using alternate decay heat removal systems, irradiated fuel is not scheduled to be located in the upper conta W ent pool.

There are two cases that require discussion regarding water circulation flowpaths within the reactor vessel. The first case pertains to the reactor cavity level at or above 22 feet 8 inches. During this period the alternate decay heat removal method is scheduled to be FPCCU in conjunction with RWCU. The FPCCU system utilizes water that overflows from the spent fuel pool, upper containment pools, transfer canal and cask storage pool thmugh skimmer weirs to the Fuel Pool Drain Tank.

Fuel Pool Cooling Pumps take suction from the drain tank and pump the water through the heat exchangers and filters, and discharge through diffusers located below the water surface in the Spent Fuel Pool, Tra m fer Canal, Upper Containment Pools and the Cask Storage Pool.

The RWCU system can act as a mini-recirmalation system (drawing water from

,. the bottom head and recirculation 1 ices and injecting into the feedwater

line) to mitigate stratification (NED0-24708A Revision 1 December 1980).

The RWCU system takes suction and discharges coolant at the same location as Shutdown Cooling. Water is heated in the core region, the core region water then rises. This has the effect of drawing cooled water from RWCU down through the jet pumps and then up through the core region. With the core and shroud area open to the flooded cavity, there will be thermal' mixing between the warmer water rising from the core and cooler water in the upper pools. Relative temperature differences will cause cooler, more dense water to flow down the cavity to the downtomer regions. This cooler' water will mix with RWCU discharge flow, be drawn down the jet pumps to the area below the core, and then be forced up through the core as warmer, less dense water rises out of the core.

The second case regarding water circulation flowpaths within the reactor vessel involves the draining of the reactor cavity to decontaminate the cavity, remove vibration instrumentation and to replace the reactor head.

During this period RHR B Shutdown Cooling is scheduled to be operable and in operation and the R'JCU system and Control Rod Drive (CRD) system are J16AECM87110501 - 6 b .t _ _ _ _ _ _ _ _ _ _ _ _ _ _

?!' f,

, -f

~ - Attachment 1 to

'6 ,i, , AECM-67/0215

,scjldule/foruseasalternatedecayheatremoval. The CRD system will be injecting water near the bottom of the core and t t RWCU system will be utilized as a mini-recirculation system as described in the first case and for removing water injected by CRD. The water level will be maintained

• l

• sufficiently high to support natural circulation. The water will be

'# heated in t,he core, rise through the core, drawing cooled water from RWCU j

down through the det pumps and then up through the core region.

c In discussions with the Nuclear Steara. Supply System vendor for GGNS, it is

• . worthy of note,that adequate long term cooling of the core following normal shutdown or loss of coolant accident can be accomplished by simply i

,s keeping tho core covered with coolant. The heat generation rate after j flow shutdown or scram rapidly drops off to the point where the fuel J' ' temperature will remain very close to that of the surrounding water. No '

fuel damage will result as long as the core remains covered, as discussed in Section III A-5 of NED0-20566.

NRC Requist, s

2. (c) ribe the tests planned to demonstrate the adequacy of these 4 terr < ate methods, including the location and acceptable value for the measurement of those test parameters which will

/~

demonstrate adequacy. Relate acceptance parameters to the s

analyses identified in (b) ubove. Describe"the reactor and

/ f refueling ::onditions when the tests will be run (e.g., reactor head on, reactor flooded, reactor drained, spent fuel in upper

)  ; . containment pool).

t , >

< iERIResponse i

/

Two tests are scheduled for the tup reactor cavity , levels prior to the use

- of alternate decay heat removal methods. The first, test will demonstrate that the FPCCU system with the RWCU system can cool the fuel remaining in 5 I 4 the vessel and the irradiated fuel discharg3d to the spent fuel pool, with the reactor head'off and upper containment pool flooded up to or above

! 22 feet 8 inches. This test will utilize the Systdn Operating Instructions for FPCCU system n d RWCU system. Temperature monitors G41-T1-R605 for the Spent fuel Pool Temperature, G33-TI-R607 for the Reactor Water Cleanup Suction Tem?erature, B33-TR-R604 for multiple inputs from Bottom Head Drain

,, Temperature and Recirculation Loop Temperature G41-TJR-R005 for multiple inputs from Upper Containment Fuel Pool Temperature, Fuel Poel Cooling Drain Tank Temperature and the Fuel Pool Heat Exchanger butret Temperature wilk be utilized for acceptance of thq test results. Acceptance will be based on a detennination that the temperature measured g by all of the above temperatur u onitors is decreasing or stable below 140 F with RHR shutdown cooling secured but operable and available.

i l Thejsicond test will utilize the RWCU and CRD systems as an alternate de 5 heat removal method with the reactor head off and irradiated fuel meyad to the spent fuel pool. This test willlalso utilize the System r

Operatingnstructions for the RWCU and CRD systems to demonstrate V

J16AECM87110501 - 7

Attachment 1 to AECM-87/0215 the capability to remove decay heat from the reactor vessel. Temperature monitors G33-TI-R607 for the Reactor Water Clean-up Suction Temperature, B33-TR-R604 for multiple inputs from Bottom Head Drain Temperature and Recirculating Loop Temperature will be utilized for acceptance of the test results. Acceptance will be based on a determination that the temperature meaguredbythesetemperaturemonitorsaredecreasingorstablebelow 140 F with RHR shutdown cooling secured but operable and available.

Alternate methods of decay heat removal have been demonstrated at GGNS and other domestic nuclear plants of the same general design type as GGNS.

Off-Normal Event Procedure, Inadequate Decay Heat Removal (05-1-02-111-1) provides corrective action should alternate decay heat removal methods prove to be inadequate.

The calculated results regarding alternate decay heat removal should bound the test results of alternate decay heat removal. However, the purpose of the calculation is to indicate the point during the outage at which the proposed alternate should be capable of removing the required decay heat. Satisfactory test results will prove the alternate decay heat removal capacity.

3. NRC Request For each alternate method of decay heat removal, describe the steps that would be taken in the event the alternate method fails during the time it is being used in the outage (e.g., loss of offsite power) assuming the Residual Heat Removal system is not operable.

SERI Response During the period of time from approximately November 30 through December 8, both shutdown cooling loops of RHR are inoperable for common suction work and ECCS testing. The FPCCU and RWCU systems together will be utilized as an alterna+e shutdown cooling method. In case of a loss of offsite power, FPCCU may be restored to service for cooling the spent fuel pool and the reactor when flooded to 22 feet 8 inches or above. SERI has scheduled the Emergency Diesel Generator associated with the one required ECCS and FPCCU subsystem to be functional. If heat loads are such that FPCCUcanrotremgvedecayheatsufficientlytomaintaincoolant temperature <140 F, a " feed and bleed" type decay heat removal method will be utilized, injecting water into the cavity or reactor and draining l excess water. If offsite power is available, normal pool makeup from the Condensate and Refueling Water System or from Makeup Water Treatment system can be used. In case of a loss of offsite power, CRD, ECCS pumps, ECCS jockey pumps, and/or the SSW pumps can be used. In case of a Station Blackout, procedures are established for providing water from the diesel driven fire pump to the pools and/or reactor vessel. Drain paths can be established through normal cavity drains, RWCU, RHR, or other means. In addition, during the period of November 30 through December 8, RHR B will be undergoing tests and surveillance which involve system lineups associated with those tests. These lineups preclude the use of J16AECM87110501 - 8

i Attachment 1 to i AECM-87/0215 RHR B as an operable shutdown cooling subsystem; however, if necessary, RHR B can be lined up for shutdown cooling and put into operation in a functional condition (although until the tests and surveillance are completed, it could not formally be declared operable).

During the period of time from approximately December 8 through December 22 the reactor cavity will be dretned. RHR B will be available as one method of shutdown _ cooling. Requiring two systems to be operable means that should one system be lost, the second can maintain coolant below the applicable limits. However, should the RHR, RWCU and CRD systems be lost, makeup is available through the systems listed previously and blowdown can be accomplished manually or by flooding the reactor and draining through the steam line drain or S/RV's, thus cooling the reactor by " feed-and-bleed".

l l

l J16AECM87110501 - 9

1 Attachment 2 to AECM-87/0215 METHODOLOGY FOR CALCULATING DECAY HEAT System Energy Resources, Incorporated (SERI) utilized the Oak Ridge National Laboratory (ORNL) computer code ORIGEN2 to perform decay heat predictions.

This code as documented in Reference 1 is a point depletion radioactive-decay code used for determining nuclide compositions and the characteristics of the materials contained therein. The accuracy of ORIGEN2 for decay heat calculations has been evaluated by ORNL in Reference 1 which indicates that ORIGEN2 underpredicts decay heat by approximately 2% relative to the 1978 ANS Standard 5.1 for decay times between 1 day and 1 year. Comparisons to actual measurements (Reference 1) indicates that ORIGEN2 tends to slightly overpredict decay heat. Therefore ORIGEN2 is considered an acceptable and realistic method for determining decay heat.

In order to determine the sensitivity of the key input parameters several cases were run using variations in power / history, total exposure and Uranium enrichment. These evaluations show that, for the decay times of interest, variations in exposure are overshadowed by power history effects which are approximately linear. Small variations in enrichment have little effect.

The following major assumptions were made in the decay heat modeling:

1) All unit operation was at 100% power; operating times were adjusted to give the correct number of effective full power days. Actual outage times were used.

2) Bundle average enrichment for each fuel type was used to determine the initial composition. Nominal fuel weights of Uranium 235 and 238 were used. These were the only nuclices modeled.

3) Batch average relative power was used throughout the cycle.

The first assumption is justified since increases in power result in higher decay heat and the power level for the last portion of Cycle 2 was slightly {

1ess than rated. The second assumption is justified since the results show I little sensitivity to enrichment over a range from natural uranium (no )

enrichment) to a greater enrichment than the maximum enrichment used in either l Cycle 1 or Cycle 2 and since the dominant Cycle 2 fuel type has uniform axial f enrichment except for a 6" natural Uranium reflector region at the top and I bottom of the core. The final assumption is justified based upon the linear ('

relationship between the decay heat and the power history.

The results of these calculations are shown in Table 1.

Reference 1: "0RIGEN2, Isotope Generation and Depletion Code - MATRIX EXPONENTIAL METHOD", A. G. Croff, July 1980, ORNL/TM-7175 J16 MISC 87110901 - 1

_ _ _ _ l

l Attachment 2 to j AECM-87/0215  !

Table 1 CYCLE 2 DECAY HEAT (MW) AFTER SHUTDOWN **

DECAY HEAT (MW) -A DECAY HEAT (MW) FROM TOTAL DECAY HEAT-DAYS

• CYCLE 2 DISCHARGED FUEL FUEL REMAINING IN CORE ( MW )

1 6.032 14.002 20.033 2 4.840 '11.185 16.026 4 3.710 8.587 12.297 l 6 3.096 7.178 10.275 8 2.702 6.265 8.967 10 2.426 5.621 8.047 12 2.224 5.142 7.366 14 2.067 4.767 6.834 16 1.941 4.464 6.405 18 1.836 4.210 6.046 20 1.746 3.992 5.739 25 1.567 3.555 5.122 30 1.430 3.218 4.648 35 1.319 2.949 4.268 40 1.229 2.727 3.956 50 1.088 2.383 3.471 60 0.983 2.126 3.109 80 0.833 1.761 2.594 100 0.725 1.502 2.227 120 0.642 1.303 1.945

• Days after cycle 2 shutdown

• • Excluding cycle 1 spent fuel decay heat i

J16 MISC 87110901 - 2

Attachment 3 to AECM-87/0215 Summary of the Alternate Shutdnwn Cooling Methods Analyses Analyses were performed for two proposed alternate shutdown cooling methods (FPCCU), which Reactoruse different Water Cleanup combinations (RWCU), and of the Fuel Control Rod Drive-(CRD Pool Cooling)and systemsCleanup to remove decay heat from the reactor,'the upper containment pool, and the Auxiliary. Building spent. fuel pool. The decay heat-loads will consist of 264 Cycle 1 and 288 Cycle 2 spent fuel assemblies in the Auxiliary Building spent

, fuel' pool and 512 irradiated fuel assemblies in the reactor.

The first of the two proposed alternative methods uses the FPCCU and RWCU systems beginning on day'24 (scheduled November 30, 1987) of RF02. The FPCCU will remove decay heat.from the Auxiliary Building spent fuel pool while assisting RWCU in removing decay heat from the reactor. The second method uses the FPCCU, RWCU, and CRD. systems beginning on day 32 (scheduled December 8, 1987) of RF02. In this mode the FPCCU system will remove decay heat from Auxiliary building spent fuel pool and the CRD system will assist RWCU in, removing decay heat from the. reactor. The analysis calculates the total heat removal capacity for each method and the decay heat loads the systems will be required to remove in each method. The analysis of the heat removal capacities does not take credit for heat losses through the pool surface, walls, and floor or the cooling system piping.

I The results of the analysis is described in more detail below and .is

. summarized on the attached table summary of alternate cooling methods analysis.

i TheheatremovalcapacitiesoftheFPCCUandRWCU(non-regenerative)geat exchangerswerecalculatedassuminginletcoolingwatertempegaturesof95F and 75 F respectively and a hot side inlet temperature of 140 F, which is the maximum allowable temperature for the pools and reactor. The analysis uses normal operating flowrates for the operation of one train of FPCCU and two pump operation of RWCU. The CRD system heat removal capacity is analyzed by calculating the heat transmitted to CRD gater when increasing the gRD inlet water temperature to the reactor from 70 F to a temperature of 140 F at a i flowrate.of 70 gpm. The results of the analysis show that the heat removal capacities of the FPCCU and RWCU' systems are 11.3 and 9.24 million Btu /hr, respectively. The capacity of the CRD system was calculated to be 2.45 million Btu /hr.

Decay heat loads for Cycle 2 spent and irradiated fuel were analyzed  ;

using Table 1 of Attachment 2. Cycle 1 spent fuel decay heat loads were calculated using Branch Technical Position ASB 9-2, Rev. 2. At the end of day  ;

23 of RF02 the heat load associated with the RF02 irradiated fuel in the )

reactor is 12.73 million Btu /hr, and the total heat load for the RF01 and RF02 spent fuel in the Auxiliary Building spent fuel pool is 7.01 million Stu/hr. 1 At the end of day 31 of RF02 the decay heat load associated with the RF02

' irradiated fuel in-the reactor is 10.80 million Btu /hr and the total heat load for the RF01 and RF02 spent fuel in the spent fuel pool is 6.21 million Btu /hr.

I 1

l J16 MISC 87110901 - 3 l i

l Attachment 3 to AECM-87/0215 Based on the stated assumptions the comparison of the heat removal capacities to the heat loads shows that on day 24 the total heat removal capacity of the FPCCU and RWCU systems (20.54 million Btu /hr) exceeds the heat load total for all fuel (19.74 million Btu /hr). On day 32 of RF02 the total heat removal capacity for the RWCU and CRD systems (11.69 million Btu /hr) exceeds the decay heat 1 cad in the reactor (10.80 million Btu /hr) and the FPCCU l heat removal capacity (11.3 million Btu /hr) exceeds the heat load in the spent I fuel pool (6.21 million Btu /hr).

I I

I 1

s J16 MISC 87110901 - 4 1

Attachment 3 to AECM-87/0215 TABLE

SUMMARY

OF ALTERNATE SHUTDOWN COOLING METHODS System Capacity System Heat Removal Capacity (million Btu /hr)

FPCCU 11.3 RWCU 9.24 CRD 2.45 Alternate Method 1 l

System System Ca)acity Day 24 Requirement (million 3tu/hr) (million Btu /hr)

FPCCU + RWCU 20.54 19.74 l Alternate Method 2 l System System Ca)acity Day 32 Requirement (million 3tu/hr) (million Btu /hr)

RWCU + CRD 11.69 10.8 (reactor)

-FPCCU 11.3 6.21 (spent fuel pool)

J16 MISC 871.10901 - 5

_ _ _ . _ _ _ _ _ _ _ _ ,