ML20245D945

Safety Evaluation Supporting Amend 48 to License NPF-18

ML20245D945
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Site:	LaSalle
Issue date:	06/15/1989
From:	Office of Nuclear Reactor Regulation
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION AMENDMENT.NO. 48 TO FACILITY OPERATING LICENSE NO. NPF-18 COMMONWEALTH EDIS0N COMPANY g

'LASALLE COUNTY STATION, UNIT 2 i

DOCKET NO. 50-374 w

1.0' INTRODUCTION By' letter from C. M. Allen, Commonwealth Edison Company (Ceco), to USNRC, dated September 19, 1986 Technical Specification changes were proposed for LaSalle County Station Unit 2 to allow use of high density spent fuel racks. The

' installation of these racks will extend full core discharge capability beyond the 1990 time frame. The LaSalle County Station Unit 2 received a full power operating license on March 23, 1984. At the time of licensing, the spent fuel pool contained sufficient storage locations to accommodate 1080 fuel bundles.

With the existing storage racks, the ability to offload a full core will be lost in 1990. Consequently, the' licensee proposed to re-rack the spent fuel pool in order to expand the spent fuel storage capacity.. The new 3roposed high density storage racks will increase the storaga capacity ~of tie spent fuel pool to 4078 fuel bundles and is projected to provide storage capacity until the year 2000 while.still maintaining the ability to offload a full core. The s

licensee provided additional information in support of the re-rack request in submittals dated August 18 November 17, 24, 1987, May 17, 1988, and June 6, 3

'1S89.

2.0 EVALUATION The licensee's submittals were reviewed in accordance with the requirements of GeneralofDesignCriteria2)44,and61andtheguidelinesofNUREG-0800, "StandardReviewPlan,"(SRP and NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants."

2.1 Decay Heat Generation Rate The licensee stated in the September 19, 1986 submittal that the calculation of

'y7" NUREG-0800, Standard Review Plan, Section 9.1.3 and Branch Technical Position the decay heat generation rate was in accordance with the guidelines of Mg ASB 9-2.

However, the licensee did not assume that the spent fuel pool was a,

R filled with spent fuel. For the normal maximum heat load case, the licensee iF assumed the pool is filled with one third core refuelings every 18 months

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except for 764 empty fue? storage locations which were reserved for the core 7

offload, five empty defective fuel storage locations, and 189 additional empty ry locations. The licensee assumes that no fuel is moved from the reactor for the

'h first 7 days (168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br />) after shutdown and has taken credit for moving fuel

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-from the reactor to the spent fuel pool at a rate of 4 fuel. bundles per hour.

om The abnormal maximum heat load case has the same assumptions as the normal maximum heat load case except the 764 empty fuel storage locations are filled with a full core offload. With these deviations from the guidelines, the

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. licensee calculated a heat generation rate of 13.98 MBtu/Hr for the normal maximum heat load case and 31.61 MBtu/Hr for the abnormal maximum heat loao case.

The staff performed an independent calculation of the heat generation rate in accordance with the guidelines in the Standard Review Plan, Section 9.1.3 and Branch Technical Position ASB 9-2 assuming the anticipated 18-month operating cycle. The staff calculated a normal maximum heat generation rate of 17.6 MBtu/Hr and an abnormal maximum heat generation rate of 41.0 NBtu/Hr. The staff utilized these calculated values when assessing the adequacy of. the spent fuel pool cooling system.

Because of the differences between the staff and licensee calculated heat generation rates, the staff performed additional calculations to verify the licensee's method of calculating the heat generation rate. The current reload batch was used as the verification point. The licensee calculated the heat generation rate for the current reload batch to be 10.02 MBtu/Hr.

Based on the assumption that the entire reload batch enters the spent fuel pool instantaneously at 267 hours0.00309 days <br />0.0742 hours <br />4.414683e-4 weeks <br />1.015935e-4 months <br /> (the licensee's assumed 7 day delay in refueling plus one half of the licensee's anticipated refueling tine), the staff calculated a heat generation rate for the current reload bath of 10.09 MBtu/Hr. Thus, the staff finds that the licensee has used on acceptable method for determining the spent fuel heat generation rate and that the difference between the licensee's values and those calculated by the staff are due solely to the assumptions used by the licensee.

2.2 Spent Fuel Pool Cooling System The spent fuel pool cooling system (SFPCS) consists of two identical trains of equipnent. Each train consists of one 3000 gpm centrifugal pump and one 14.6 MBtu/Hr tube-and-shell heat exchanger. Af ter water from the spent fuel pool is cooled by the heat exchangers, it is purified by the spent fuel pool cleanup system. Neither the SFPCS nor the cleanup system are seismic Category I.

In the event of an excessive heat load, the "B" loop of the Residual Heat Removal (RHR) system can be used to cool the spent fuel pool. The RHR system, including all piping to and from the spent fuel pool, is independent of the SFPCS and is seismic Category I.

2.2.1 Heat Removal Capability Under the normal maximum heat load conditions (17.6 MBtu/Hr) and a single failure of one SFPCS train, the remaining SFPCS train will maintain the spent fuel pool water temperature below 126*F which is less than the 140*F temperature guideline specified in the Standard Review Plan, Section 9.1.3.

For the abnormal maximum heat load condition (41.0 MBtu/Hr), one train of the SFPCS will maintain the spent fuel pool water temperature below 168"F which is below boiling. Thus, the staff finds that the SFPCS meets the requirements of General Design Criterion 44, " Cooling Water" with respect to providing adequate pool cooling.

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, l 2.2.2 Protection Against Natural Phenomena 2.2.2.1 Makeup Water The SFP cooling capability is reviewed with respect to the requirements of General Design Criterion 2. " Design Bases for Protection Against Natural Pheno-mena," which includes protection against earthquakes, hurricanes, tornadoes, or other natural events. The SFPCS is not seismic Category I and it is not powered by a Class IE source (i.e., on-site emergency diesel generator). Under such circumstances, SRP Section 9.1.3 identifies an alternative method for cooling of spent fuel following an earthquake.

Specifically, the SRP discusses use of a seismic Category I spent fuel pool makeup water capability and a seismic Category I ventilation system to process potential radiological releases to the pool building resulting from pool boilino. The LaSalle FSAR identifies the emergency fuel pool makeup system (EFPMS) as the seismic Category I makeup water system for the spent fuel pool.

The EFPMS includes two 300 gpm pumps and is part of the seismic Category I core standby cooling system - equipment cooling water system (CSCS-ECWS).

2.2.2.2 Building Ventilation With regard to qualified ventilation capability when seismic Category I spent fuel pool cooling (SGTS) as the qualified ventilation system.is not provided, the LaSal treatment system The SGTS is designed to seismic Category I criteria and consists of two redundant filter trains. This system is designed to remain operational during design basis events and is protected against natural phenomena.

2.2.4 Loss of Cooling In the event that all SFP cooling is lost, the spent fuel pool temperature will increase until boiling is achieved. The licensee has estimated the time from the loss of pool cooling until the pool boils (from an initial pool temperature of 120 F) for the normal maximum heat load case to be approximately 16.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and for the abnormal heat load case to be approximately 4.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />. The calculated boiloff rates are estimated to be 28.9 gpm and 65.3 gpm, respectively.

The staff finds that the EFPMS capability is in excess of those estimated boiloff rates, and there is reasonable time to take action to provide SFP makeup. The staff further concludes that the seismic Category I EFPMS and SGTS meet the requirements of GDC 2 for ensuring adequate spent fuel pool cooling and prevention of unacceptable radiological releases following an earthquake.

2.3 Heavy Load Handli_ng The new spent fuel storage racks weigh more than a fuel bundle, channel and its handling tool. Thus, the spent fuel storage racks are considered to be heavy loads. The reactor building crane will be used to move the storage racks within the reactor building and the spent fuel pool. As part of the previous staff review of compliance with guidelines of NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants," the staff concluded in NUREG-0519, Supplement No. 5, dated April 1983 that the reactor building crane met the guidelines of NUREG-0612.

ik :In the August 18, 1987 submittal, the licensee provided information which E

identified the path of travel for each of the fuel storage racks within the reactor building. The licensee stated that all Unit 2 fuel will be mcVed into

.the Unit I spent fuel pool prior to the re-racking of the Unit 2 pool.

Therefore, no racks, either existing or new, will be carried over spent fuel or

< racks containing spent fuel.

The licensee specifically indicated the heavy load handling paths and laydown areas for the storage racks. The licensee also indicated that one special lifting device will:be used in the spent fuel pool re-racking procedure.

By submittal dated November 24, 1987, the licensee provided drawings of the US Tool and Die spent fuel rack lifting rig which -

indicates redundancy in the lifting rig thereby satisfying staff guidelines.

The staff finds that heavy load handling will be performed in accordance with the guidelines of NUREG-0612 and that the requirements of General Design Criterion 61, " Fuel Storage and Handling and Radioactivity Control," are met as they relate to proper load handling to ensure against an unacceptable release of radioactivity or a criticality accident as a result of a postulated load drop.

2.4 Criticality. Aspects The change to Section 5.6.1.1.b and 5.6.3 of the Technical Specifications would increase the number of fuel assemblies which could be stored in the spent fuel pool from 1,120 to 4,078 and change the storage cell spacing from 7 to 6.26 inches. Other previously approved specifications of Section 5.6 would remain unchanged. The change.is based on the installation of an array of new fuel racks in the' spent fuel pool which can provide the closer packing of fuel assemblies. Required criticality margins are maintained by incorporation of boren containing material in the rack design. This is a commonly used feature for high density rack design, and a large number of similar designs have been approved by the NRC.

The rack design and safety analyses, including nuclear criticality analyres, were provided by U.S. Tool and Die (USTD). They have previously providec the (similar). design and analysis methodology for the Nine Mile Point Unit 2 (9MP2) spent fuel racks. These were reviewed and approved by the NRC staff for the 9MP2 Final Safety Evaluation Report. The rack design is a rectangular array of stbrsge cells configured so that there is boron, in the form of Boraflex sheets, between each pair of fuel assemblies. This includes Boraflex on the outer edge of racks, which is arranged so that there is boron between asstmblies facing each other across rack to rack gaps. The B-10 loading of the

,Boraflex is 0.020 gm/ cme minimum and the cell pitch is 6.26 inches.

The criticality calculations for the racks wero done using the LEOPARD /PDQ-7 code package. These programs were origir. ally developed in the Naval Reactor Program, were further developed and used by Westinghouse and others, and have been a widely used industry standard methodology (for the analyses of both reactor and fuel storage multiplication factors k-infinity and k-effective) and reactivity changes as a function of component changes. LEOPARD is primarily used to gt.nerate cross sections for use in PDQ-7 diffusion theory calculations.

This code package has been used for the criticality aralysis of severci staff approved fuel storage systems.

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, ' The methodology has been benchmarked against a number of relevant critical experiments covering a range of geometries and material compositions and fuel enrichments, and including poison sheets. These present geometrically representative configurations, many of which match those used to mockup fuel storage racks.

In particular, they have included a series of criticals by Battelle Pacific Northwest Laboratories which are frequently used to check fuel storage calculations.

USTD has used benchmark calculations of these Battelle-experiments to develop analysis methodology bias and uncertainty factors to be added to k-effective calculations for the LaSalle Unit 2 racks.

- USTD has also determined the potential variation of. the rack and fuel parameters which are used in determining the k-effective of the rack-fuel system. These parameters include poison thickness, cell pitch, stainless steel thickness, fuel density and eccentric fuel position.

(B-10 area density was used at the minimum specification value.) The variation of k-effective with these parameters -(taken at a 95/95 probability / confidence level) was determined. These (independent parameters) were statistically combined with the methodology uncertainty to provide a delta-k uncertainty which was added to the base k-effective calculation.

USTD has investigated abnormal conditions which might be associated with the spent fuel pool and has determined that, with the exception of the placement of i

a fuel assembly outside and imediately adjacent to the racks, such events',

e.g., dropped fuel bundle or increased pool water temperature, have negligible or negttive effects on k-effective. The effect of the external fuel assembly was calculated and included as a positive bias in determining the rack k-effective.

For the base case rack calculation the USTD mockup assumed (I) an infinite array of storage cells on a 6.26 inch pitch and with appropriate thicknesses of Boraflex, steel plates and water gaps between fuel assemblies, (2) typical General Electric 8x8 bundle fuel pin arrays with 62 3.416 percent U-235 enriched fuel pins and 2 water rods (3)thebundlewasunitradiatedand containednoburnablepoison,and(4)thepoolwaterwas68'F.

For this base configuration, and adding reactivity value for assuming (conservathely) zirconium channels on the fuel bundle, axial end effects and corrections for cutouts in the Boraflex plates, the k-effective was calculated to be 0.9250. The total calculational bias and statistically combined calculational and mechanical uncertainty reactivity, 0.0093 delta-k, plus the external assembly accident reactivity, 0.0098 delta-k, was added to this to give a total k-effective of 0.9441. This is to be compared with an NRC required limit of 0.95.

USTD also carried out sensitivity calculations (in addition to those used to define mechanical uncertainty) to indicate the extent of the conservatism in the base model. These included reactivity cffects of typical multi-enrichment fuel pin distributions, typical burnable poisons in the fuel pins, and s)acer grids. These calculations indicated a more realistic k-effective would be under 0.90.

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.c 'The submittal also presented some results of calculations using an assembly

'k-infinity rather than the fuel enrichment approach. This can take direct advantage of the enrichment distribution and particularly the burnable poison and burnup reactivity effects, and has been used, e.g., by General Electric, for fuel storage analysis. However, there is insufficient information in the present submittal, particularly on uncertainty analysis, to consider judgments on this approach at this time.

2.4.1. Criticality Evaluation The basic criticality design of the new racks, using boron lined cells to provide the appropriate neutron multiplication level for the closer packed array.of high density racks, is a comonly used concept and has been accepted for mary spent fuel' storage pools. A design very similar in detail (and

' analysis) has been previously reviewed and accepted for Nine Nile Pcint 2.

It Lis an acceptable esign concept for maintaining criticality levels for the LaSalle Unit 2 pool.

The analytical methodology used by USTD to analyze the criticality and reactivity change characteristics of the racks is a standard methodology, commonly used and approved for other utilities for such analyses. The LEOPARD /PDQ-7 code package provides an acceptable methodology for base calculations and for sensitivity calculations. These methods have been benchmarked against an appropriate selection of critical experiments, with results falling within expected ranges of deviations from the experiments. The derivation of the uncertainty of the methodology from this benchmarking follows normal procedures and also' falls within an expected range.

It is bcceptabic.

The examination of uncertainties to be attributed to variances in diuensions and materials in the fuel and racks has ccvered an acceptable range of parameters and has used a suitable, standard methodology for determining reactivity effects and their statistical combination. The examination of the effects of abnormal ecoditions has covered the standard events relating to changes in temperature, movements, misplacement and dropping of assemblies and other equipment, and the results are reasonable and acceptable.

i The model used for the base calculations is generally conservative, particularly in not including burnable poison effects. The base calculation and added factors for uncertainties, giving a total k-effective of 0.9441, are thus acceptable for a maximum average planar U-235 enrichment of 3.416 percent.

(The analysis was for fuel with one axial enriched region with 6 inch natural uranium end regions, but the same limit would also apply, without further analysis, to newer fuel designs with multiple axial enrichment regions.) These calculations indicate a small margin to the staff required Technichi Specification limit of 0.95 for the spent fuel pool, including uncertainties and biases. The results are acceptable.

Aux 111try calculations have indicated that a slightly higher enrichment, a little over 3.50 percent eculd be accorscdated within limits, and that with a k-infinity approach an even higher enrichment could be ecnsidered. However, these increases are not directly part of this review and the above acceptance is limited to a meximum (planar) enrichment of 3.416 percert.

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The only requested Technical Specification changes are to 5.6.1.1.b 'and 5.6.3 L

changing the storage cell pitch to 6.26 inches and increasing the allowed number of fuel assemblies in the spent fuel pool racks ~from 1120 to 4078.-

f These are: acceptable changes.

2.5 Radiological Aspects The plan proposed by the licensee entails the transfer of all spent' fuel sedules from the Unit 2 pool into the Unit 1 pool through the cennecting spent fuel cask pit.. With the fuel removed from Unit 2 SFP, the licensee has two.

options-for the actual reracking of the pool.

Following decontamination and surveying the pool, replacement of the SFP racks can be accomplished under water with the use of divers (wet option) or the rerack can be completed with the pool drained (dry option).

Based on previcus industry' experience, the licenses fias estimated the whole

. body radiation exposure to con.plete the rerack operation at 5 person-rem for the dry option, and 10 person-rem for the het option.

" Although the' dry option would incur less direct radiation exposure, it involves a higher potential for generating cirborne radioactive contamination. The methods proposed by the licensee to minimize the' generation of airborne contamination include 1) decontamination of SFP surfaces, 2) keeping SFP surfaces wetted with pool drained, 3) provide a filtered enclosure for the SFP if necessary to. contain the activity to-the pool.

Based on our review of the LaSalle proposal, we conclude that the projected activities and estimated person-rem doses for this project are reasonable.

CEC intenos to take itLARA considerations into account and to implement reason-able dose-reducing-activities. We conclude that CEC will be able to maintain individual occupational radiation exposures within the limits of 10 CFR Part 20 and maintain doses ALARA,-consistent with the guidelines of Regulatory Guide 8.8.

Therefore, the proposed radiation protection aspects of the SFP rerack are acceptable.

2.5.1 Accident Analysis The staff has reviewed the accidental fission product releases that could occur i

at LaSalle Unit 2 in conjunction with the proposed reracking of the SFP. The

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staff finds that neither the reracking operation nor the increased density of fuel in the pool resulting from the proposed modification pose a new type of

. accident not previously considered nor do they effect the assumptions or results of the previous accident analysis.

2.5.2 Radiological Impact Assessment /Public Radiation Exposure This section contains the staff's estimates of the impacts on the public from the proposed SFP modification. Major sources of radioactivity and principal environmental pathways were considered in preparing this section. Since the licensee is retaining the option of performing this modification with the SFP full or drained, the impact of both is considered.

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, 2.5.2.1 Radioactive Wastes The plant contains radioactive waste treatment systems designed to collt1 and process the gaseous, liquid, and solid waste that might contain radioactive material. The radioactive waste treatment systems are evaluated in the Final Environmental Statement (FES) dated Novenber 1978 (US NRC 1978). There will be no change in the waste treatment systems described in the FES because of the proposed SFP rerack, 2.5.2.2 Radioactive Material Released to the Atmosphere With respect to releases of gaseous materials to the atmosphere, the only radioactive gas of significance which could be attributable to storing addi-tional spent fuel assenblies for a longer period of time would be the noble gas radionuclides Krypton-85 (Kr-85).

Experience has demonstrated that after spent fuel has decayed four to six months, there is no longer a significant release of fission products, including Kr-85, from stored spent fuel containing clad-ding defects. To determine the average annual release of Kr-85, we assume that all of the Kr-85 released from any defective fuel discharged to the SFP will be released prior to the next refueling (typically 12 to 18 months).

Enlarging the storage capacity of the SFP has no effect on the average annual quantities of Kr-85 released to the atmosphere each year.

Iodine-131 releases from spent fuel assemblies to the SFP water will not be significantly increased because of the expansion of the fuel storage capacity since the Iodine-131 inventory in the fuel will decay to negligible levels between refuelings.

Most of the tritium in the SFP water results from activation of baron and lithium in the prfmary coolant which will not be affected by the proposed changes. A relatively small amount of tritium is contributed during reactor operation by fissioning of reactor fuel and subsequent diffusion of tritium through the fuel and fuel cladding. Tritium release from the fuel essentially occurs while the fuel is hot, that is, during operations and, to a limited extent, shortly af ter shutdown. Thus, expanding the SFP capacity will not significantly increase the tritium activity in the SFP since the additional fuel stored in the expanded pool will be aged fuel.

Storing additional spent fuel assemblies is not exsected to increase the bulk water temperature during normal refuelings above t1e value used in the design analysis. Therefore, it is not expected that there will be any significant change in the annual release of tritium or iodine as a result of the proposed modifications from that previously evaluated in the FES. Most airborne releases of tritium and iodine result from evaporation of reactor coolant, which contains tritium and iodine in higher concentrations than the SFP.

Therefore, even if there were a higher evaporation rate from the SFP, the increase in tritium and iodine releases from the plant, as a result of the increase in stored spent fuel, would be small compared to the amount normally released from the plant and that which was previously evaluated in the FES.

Performing this modification with the SFP drained provides an increased potential for the generation of radioactive aerosols (airbone particulate).

Additional controls have been proposed by the licensee to minimize the

9. generation of airborne radioactive particulate.

In addition, the station Radiological Effluent Technical Specifications, which are not being changed by this action, limit the total gaseous and particulate materials released to the atmosphere. The' increase in airborne radioactivity released from the plant due

. to the proposed action is' expected to be less than 1 percent of the radioactivity that is released from the plant by this pathway during normal operation of the plant.

2.5.2.3 Solid Radioactive Wastes The concentration of radionuclides in the pool water is controlled by the SFP

- cleanup system and by decay of short-lived isotopt.s. The activity is highest during refueling operations when reactor coolant water is introduced into the pool, and decreases as the pool water is processed through the SFP cleanup system. The increase of radioactivity, if any, due to the proposed modification, should be minor because of the capability of the cleanup system to continuously remove radioactivity in the SFP water to acceptable levels.

If the present spent fuel racks to be removed from the SFP of LaSalle, Unit 2 are contaminated, they may be disposed of as low level solid waste. Averaged over the lifetime of the station, thic would increase the total waste volume shipped from the station by less than 1 percent. This will not have any significant additional environmental impact.

2.5.2.4 Radioactive Material Released to Receiving Waters There should not be a significant1ncrease in the liquid release of radionuclides from the plant as a result of the proposed modifications.

Draining the pool to perform the reracking operation would be the only source of significant quantities of. liquid radioactive waste. The slightly contaminated water in the SFP would be treated with the liquid radwaste system and released to the environment. After processing in the liquid radwaste system, the concentration of radioactivity would be a fraction of.that allowed by 10 CFR Part 20.

In addition the Radiological Effluent Technical Specifications that limit the release of radioactive material to the environment are not being changed by this action. The total activity released, as liquid waste, to the environment from this rerading operation is expected to be less than 1 percent of the radioactivity released in liquid form from the plant during normal operatic's.

2.5.3 Radiological Impact Assessment / Occupational Exposure This section contains the staff's evaluation of the estimates of the additional radiological impacts on the plant workers from the proposed operation of the modified SRP.

The occupational exposure for the proposed modification of the SFP is estimated

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by the licensee to be less than 10 person-rems. This dose is less than 2 percent of the average annual occupational dose of 735 person-rems per unit per l

year for operating BWRs in the United States (US NRC 1988). The small increase in radiation dose should not affect the licensee's ability to maintain individual t

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. I occupational doses within the limits of 10 CFR Part 20, and is as low as is l

' Regulatory Guide 8.8 (US NRC 1978) should preclude any significant occupational reasonably achievable.

Normal radiation control procedures (US NRC 19810 and radiation exposures.

j Based on present and projected operations in the SFP area, we estimate that the l

proposed operation of the modified SFP should add only a small fraction to the I

total annual occupational radiation dose at this facility.

Thus, we conclude that the proposed storage of spent fuel in the modified SFP will not result in any significant increase in doses received by workers.

2.6 Radioactive Waste Treatment System,s_

From operating experience it is known that radioactivity releases from newly discharged spent fuel drop to insignificant level after the spent fuel has decayed in the SFP 4 to 6 months.

The refueling cycle at LaSalle Unit 2 is 18 months, and during refueling one third of the core is discharged to the SFP.

Therefore, every 18 months, LaSalle Unit 2 will discharge about one third of the core to the SFP which will release approximately the same amount of radio-activity regardless of the size of the SFP or the total spent fuel inventory.

LaSalle Unit 2 has been provided with radioactive waste treatment systems designed to collect and procers the gaseous, liquid and solid wastes that might contain radioactive material.

The radioactive waste treatment systems have been previously evaluated by the staff and found acceptable.

There will be no change in the radioactive waste treatment systems as a result of the proposed installation of the new racks. Therefore, the staff finds that the proposed reracking and modifications to increase the spent fuel storage capacity will cause no significant additional environmental radiological impact.

2.7 Structural Design Aspects This evaluation addresses the adequacy of the structural aspects of the proposed application.

The Brookhaven National Laboratory (BNL) assisted the staff in reviewing various analyses and responses submitted by the licensee, and in auditing the methodology and sample calculations.

Attached Appendix A is the technical evaluation report (TER) developed by BNL.

The staff accepts the findings and conclusior,s of the TER by incorporating the TER as part of this evaluation.

There are two spent fuel storage pools at LaSalle Station.

The proposed application is for rtracking the spent fuel storage pool of Unit 2.

The storage pool is 40 ft. O in. in east-west direction, and is 34 ft.0 in north-south direction with a curved containment wall cut out in the north wall.

All reinforced concrete walls and the floor slab are 6 ft. O in thick.

The pool walls and floor slab ere lined with 1/4 in. thick stainless steel liner to ensure the water thickness of the pool.

Leak chases are provided to collect any potential leakage through the liner.

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, l The proposed high density storage rack is a honeycomb configuration of I

identical stainless steel cells.with sheet Boralfex poison material captured 3

between the side walls of all_ adjacent cells.

The individual cell is 6 inches square and has wall ~ thicknesses of 0.09 in.

The cells'are held-together with fusion welds.

A total of-4073 individual storage spaces are arranged in 20 l

distinct rack modules of various arrays of fuel cells. 'Each rack is reinforced at the top corners walls.

Each rack is supported by five pedestal 1

supports; four corner adjustable pedestals and one in the center being a fixed (not adjustable) support.

The racks are arranged with surface contacts between the reinforcing plates and gaps with walls varying from a minimum of 2 inches to a maximum of 4.25 inches.

The rack modules and their supports are fabricated from ASTM A-240, Type 304 austenitic stainless steel sheet and plate materials.

The proposed application if the storage of a single fuel assembly in each storage location of the high density racks.

2.7.1 Structural Analysis The primary areas of review associated with the proposed application are focused towards assuring the structural integrity of the fuel, fuel cells, rack modules, and the spent fuel pool floor and walls under the postulated

' loads'(Appendix D of SRP 3.8.4) and fuel handling accidents.

The major areas of concern and their resolutions are outlined in the following paragraphs.

a.

Fuel Handling Building and Spent Fuel Storage Pool The Fuel Handling Building analysis and design had been reviewed and accepted during the initial licensing stages.

The pool floor slab and walls were reanalyzed to account for the added load of the fuel, the racks and the associated impact loads.

The flow slab elevation is 804 ft.

9 in.

The original licensee's reanalysis did not adequately consider the irrpact loads resulting from rack movements under a postulated seismic.

event.

The later reanalysis included these additionh1 loads together with the hydrodynamic loads resulting from the rack movements.

The stresses in the cor: rete and reinforcing steel at critical sections are found to be with5: the acceptable criteria.

A detailed evaluation of the l

affected spent fael components is provided in Appendix A.

b.

High Density Racks l

The seismic analysis of the free standing racks in the licensing report (dated September 16, 1986) was based on the two dimensional single rack j

seismic analyses for two horizontal direction and the equivalent static loads obtained from the vertical response spectra.

The resulting codirectional loads and displacements were combined using the square root of the sum of the squares method.

This method may provide bounding loads.

However, it has been shown in prior licensing reviews that it would not be able to simulate the potential displacements (and resulting I

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1 5'

, impact,.if any) of the' free standing rack system.

Later, the licensee performed three dimensional single rack analyses with three components of the postulated earthquake, acting simultaneously, and considering the bounding coefficients of friction of 0.2 and 0.8.

The licensee also performed two dimensional analyses of a row of four racks with simultaneous input of a horizontal and vertical component of earthquake.

The results of.the later analyses have been used in assuring the adequacy i

of the rack system.

Major components of the rack are evaluated for the maximum stresses compared against the stresses allowable by the criteria in Appendix D of Standard Revitw Plant 3.8.4.

A minimum ration of 1.1 against allowable was found for th2 fusion welds holding the cells together.

In order to assure the rs.t Integrity under resulting impact load, a separate finite i

element analy<f5 of a rack subjected to impact load, discretely distributed i

over a portion of the rack, was performed by the licensee.

The results of the analysis indicated that the rack cannot withstand such an impact load without significant deformation of the cell walls.

The licensee decided to protect the potentially vulnerable portions of each rack by means of reinforcing plates.

The fuel rack system was also evaluated for the inadvertent drop of a fuel assembly during_ fuel handling operation.

Two critical cases of fuel assembly _ drop were evaluated; (1) a straight drop of a fuel assembly on the top of the rack structure from 30 inch height, (2) a straight drop of a fuel assembly into the cell of the rack from 30 inch height above the top of the rack.

Energy balance approach W th conservative assumptions indicated that in case (1), the large plastic deformation would be limited to the rack module above the active fuel region, and in case (2),

the liner plant would not be perforated.

Such deformations are acceptable under this type of accident.

A detailed evaluation of the analysis of high density racks is provided in Appendix A.

Based on its evaluation of the licensee's submittal, the supplementary information provided by the licensee, discussion with the licensee at meetings, and information audited by the staff and its consultant, the staff concludes that the licensee's structural analyses and design of the spent fuel rack modules and the spent fuel pool are in compliance with the acceptance criteria set forth in the FSAR and consistent with the current licensing practice and, therefore, are acceptable.

It should be pointed out, however, that the installation of new racks (wet reracking) on the liner plate and additional bridge plates will require utmost care in levelling and spacing the racks in the desired configuration.

A thorough review of the installation procedures and inspection of the installed racks is warranted.

It is recommended that the licensee develop walkdown procedures to be implemented after a seismic event equivalent to or exceeding the Operating Basis Earthquake (OBE).

The walkdown should include the inspection of rack modules, their displacements, and assessment of damage (if any) to adjoining structures and components.

. 3.0 ENVIRONMENTAL CONSIDEP.ATION The Commission prepared an Environmental Assessment of the proposed action, which was noticed in the Federal Register April 3, 1989 (54 FR 13445) and has concluded that an environmental impact statement is not warranted because

'here will be no environmental impact attributable to the action beyond that which has been predicted and described in the Commission's Final Environmental Statement related to the Operation of LaSalle Station, Unit I and 2 dated November 1978.

The Commission published a Notice of Consideration of Issuance of Amendment and Opportunity for Prior Hearing (52 FR 43810) November 16, 1987. No hearing requests were received.

4.0 QNCLUSIONS Based on its review of the proposed expansion of the spent fuel pool at LaSalle Unit 2, the staff concludes that:

1.

The proposed expansion of the LaSalle Unit 2 spent fuel pool complies with the requirements of General Design Criteria 2, 44, and 61 and the guidelines of NUREG-0612, and the Standard Review Plan with respect to the capability to provide adequate spent fuel pool cooling and safely handle heavy loads. The staff, therefore, finds the proposed expansion to be acceptable.

2.

The criticality aspects of the new spent fuel racks are acceptable.

3.

The estimated additional radiation doses to the general public are much less than those incurred during normal operation of LaSalle County Nuclear Station. The licensee has taken appropriate steps to ensure that occupational dose will be maintained as low as is reasonably achievable and within the limits of 10 CFR Part 20. The total occupational dose estimated to be associated with the proposed modification of the expanded fuel pool is less than 10 person-rems, which is less than 2 percent of the average annual total occupational dose at the LaSalle County Station Unit 2.

On the basis of the foregoing evaluation, it is concluded that there would be no significant additional environment radiological impact attributable to the proposed reracking and modification to increase the spent fuel storage capacity at the LaSalle County Station Unit 2.

4.

There will be negligible increase in gaseous, solid and liquid radioactive material as a result of the spent fuel 2001 expansion itself, or the con-tinued storage of additional fuel assemalies. There is no impact on the public since there is no increase in the calculated average annual quan-tities of Kr-85 released to the atmosphere. There will be adequate spent fuel pool cooling and proper heavy load handling to ensure against an l

unacceptable release of radioactivity or a criticality accident as a result of a postulated load drop.

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. 5.

The structural design, material compatibility and chemical stability are acceptable.

'The staff has concluded, based on the considerations discussed above, that (1)

.there is reasonable assurance that the health and safety of the public will not

.be endangered by operation in the proposed manner, and (2) such activities will

, be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and. security or to the health and safety of the public.

5.0 REFERENCES

U.S. Nuclear Regulatory Commission, 1978, " Final Environmental Statement Related to Operation of LaSalle County Nuclear Power Station Units Nos. I and

'2," November 1978.

--- 1977, Regulatory Guide 1.109, Revision 1, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," October 1977.

--- 1978, Regulatory Guide 8.8, revision 3, "Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations Will Be as Low as Is Reasonably Achievable," June 1978.

--- 1981, NUREG-0800, " Radiation Protection," in: " Standard Review Plan,"

Chapter 12, July 1981 (formerly issued as NUREG-75/087).

--- 1988, NUREG-0713, Volume 7,*0 occupational Radiation Exposure at Commercial Nuclear Power Reactors 1985," March 1988.

--- Letter (and attachments) from C. Allen, CECO, to H. Denton, NRC, dated September 19, 1986, "LaSalle County Station Unit 2 Proposed Amendments to Technical Specification...High Density Spent Fuel Racks."

--- PNL-2438, " Critical Separation 8etween Subcritical Clusters of 2.35 Wt%

235-U Enriched 00 Rods in Water with Fixed Neutron Poisons."

2

--- Letter from C. Allen, CECO to NRC, dated August 18, 1987, "LaSalle County Station Unit 2, Proposed Amendments to Technical Specification...High Density Spent Fuel Racks."

--- Letter from C. Allen, Ceco to NRC, dated November 5, 1987, "LaSalle County Station Unit 2 Proposed Amendments to Technical Specification...High Density Spent Fuel Racks."

--- Letter from C. Allen, CECO to NRC, dated November 24, 1987, "LaSalle County Station Unit 2 Proposed Amendments to Technical Specification...High Density Spent Fuel Racks."

-')

'I 15 -

i

--. Letter from C. Allen, CECO to NRC, dated May 17,'1988, "LaSalle County.

Station' Unit _2 Proposed Amendments to Technical Specification...High Density

{

Spent Fuel Racks."

1

--- Letter from Wayne Morgan, Ceco ~to NRC, dated June 6, 1989, "LaSalle County Station Unit 2 Proposed Amendments to Technical. Specification... High Density Spent Fuel Racks."

Principal Cont'ributors:

J. Ridgely, SPLB/NRR H. Richings, SRXB/NRR R. Pedersen, PRPB/NRR A. Chu, SPLB/NRR P. Shemanski PDIII-2/NRR H. Ashar, ESGB/NRR Dated:

June 15, 1989 i

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__________--___---___-___-_-_-_--_-_______---____-_____________--_____-__-__--___-_--_______----A

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i APPENDIX A TECHNICAL EVALUATION REPORT EVALUATION OF THE HIGH DENSITY SPENT. FUEL RACK.

~ STRUCTURAL ANALYSIS FOR

-COMMONWEALTH EDISON COMPANY LASALLE fSUNTY STATION UNIT 2 NRC DOCKET NO.'50-374 By J. Braverman STRUCTURAL ANALYSIS DIVISION DEPARTMENT OF NUCLEAR ENERGY BROOKHAVFN NATIONAL LABORATORY UPTON, NEW YORK June 1989 Prepared for U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation FIN A-3841, Task No. 20, TAC No. 62832 O()

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Executive Summary This report describes and presents the results of the BNL technical evaluation of the ' structural analysis submitted by Commonwealth Edison Company (CECO) in support of their licensing

. submittal on the use of high density spent fuel racks at LaSalle

~

County Station (LSCS) Unit 2.

The review was conducted to ensure that the racks meet all structural requirements as defined in the NRC Standard Review Plan and the NRC OT Position for Review and Acceptance of Spent Fuel Pool Storage and Handling applications.

The proposed high density spent fuel storage rack modification involves the installation of twenty free-standing, self-supporting

. modules of varying sizes arranged next to one another.

Each rack module consists of individual cells of square cross-section, each designed to accommodate one fuel assembly.

Since the racks are neither anchored to the pool floor or walls nor connected to each other, during an earthquake, the racks would be free to slide and' tilt.

Because of the nonlinear nature of this design, a time history analysis was required to characterize the seismic response of the fuel racks.

The BNL review focused primarily on the seismic analysis of the fual rack modules because of the complexity of the analysis method and the number of simplifying assumptions that were required in developing the dynamic models. BNL also reviewed other analyses performed by the Licensee including fuel handling accident analyses, thermal analyses, and spent fuel pool analyses.

During the course of the review, a number of questions were raised regardin the adequacy of the fuel rack dynamic models.

Concerns wer.e caised that single rack models may underpredict seismic forces and displacements that would occur in the real multiple rack fuel pool environment.

The use of a two-dimensional (2-D) model and analysis (E-W/ vertical and then N-S/ vertical to predict the non-linear response due to three perpendicular and simultaneous inputs was another major concern.

Concerns were also raised regarding the adequacy of the fuel racks to sustain the calculated impact load.

To address such concerns, the Licensee provided additional information and performed additional studies, including multiple fuel rack seismic analyses, to demonstrate the adequacy of the high density racks.

The additional studies indicated that the forces from the multiple rack analyses were generally lower, however, forces and displacements from the 3-D single rack study were larger than the single rack design basis results.

In spite of the larger forces and displacements the structural adequacy of the

racks, fuel assemblies and pool structure was demonstrated.

These results coupled with the conservatism present in the analyses demonstrate the adequacy of the fuel rack design.

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Based on5 the BNL review-- of ' the' Licensee's analysis, it is concluded that the proposed LSCS Unit'2 high' density fuel racks and

spent fuel' pool are designed.with sufficient capacity to withstand the:' effects of-the. required environmental and abnormal' loads.

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3.

l TABLE OF CONTENTS l

1.0 INTRODUCTION

1 1.1 Purcose 1

1.2 -Backaround 1

1.3 Scone of Review.

1 2.0 ' ACCEPTANCE CRITERIA 2

3.0 FUEL RACK DESCRIPTION 3

4.0-TECHNICAL EVALUATION.

4 4.1 Fuel Rack Seismic Analysis 4

4.1.1 Dynamic Model 4

4.1.2 Fluid Coupling Effects 7

4.1.3 Friction Effects.

8 4.1.4 Damping 8

4.1.5 _ Seismic Input Motion.

8 4.1.6 Analysis Method 9

4.1.7 Analysis Results..

10 4.1.8 Evaluation of Results 10 4.2 Multi-Rack Seismic Analysis.

11 4.2.1 Multi-Rack Model.

11 4.2.2 Multi-Rack Analysis /Results 12 4.3 Rack Impact Evaluation 13 4.4 Thermal Analysis 14 4.5 overall Evaluation of Seismic Analysis Results 15 4.6 Fuel Handling Accident Analyses.

15 4.7 Spent Fuel Pool Analysis 16 4.7.1 Loads and Load Combinations 17 4.7.2 Spent Fuel Pool Structure Analysis (SWEC) 17 4.7.3 Spent Fuel Pool Structure Analysis (S&L) 18

5.0 CONCLUSION

S 19

6.0 REFERENCES

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1.O INTRODUCTION-1.1' Purpose This technical evaluation report (TER) describes and presents the results of the BNL review of CECO's licensing submittal on the use of high density fuel racks at LSCS Unit 2 with respect to their structural adequacy.

1.2 Backaround Each of the LSCS units cas a separate spent fuel pool to provide storage-for irradiated opent fuel.

This TER addresses LSCS Unit 2 spent fuel racks which are to be installed in the Unit 2 fuel pool located at elevation 804' - 9" of the reactor building.

With the planned installation of the twenty racks, there will be a total capacity of 4073 cells and 5 defective fuel locations.

The proposed racks consist of individual cells of square cross-section, each of which ' accommodates a single BWR fuel assembly. The cells are assembled into distinct modules of varying sizes which are to be arranged within the existing spent fuel pool as shown in Figure 1.

Each module is free-standing and self-supporting.

The Licensee provided a summary of-his safety analysis and evaluation of the proposed racks in a Licensing Report (Ref. 1).

The report described the structural analysis and design of the new fuel racks.

It also gave a description of postulated dropped fuel and jammed fuel accident analyses.

1 The NRC initially reviewed the Licensing Report and generated a list'of information needed to complete the review (Ref. 2).

The Licensee provided the information and responses in a later submittal (Ref. 3a). The NRC then requested additional information q

in Ref. 4 which was responded to by the Licensee in Reference 3b.

j In addition to reviewing all of these submittals, BNL also participated in an audit of the fuel rack analyses at the offices of Sargent and Lundy Engineers, Chicago, Illinois.

In addition, a meeting was held on May 31, 1989 at the NRC to resolve the remaining open items.

4 1.3 ScoDe of Review The objective of the BNL technical review was to evaluate the adequacy of the Licensee's structural analysis and design of the proposed high density spent fuel racks and spent fuel pool.

Due to the complex nature of the fuel rack seismic analysis, the primary focus of the review was on the adequacy of the non-linear i

fuel rack models and their dynamic analysis.

The structural evaluation of fuel racks subjected to the dropped fuel and jammed 1

)

e I

fuel _ handling accidents described in the Licensee's report (Ref.

I

1) were included in this review. However, the definition of these postulated accidents and. their parameters (drop height, uplift force, ~ etc. ) were beyond the scope of this review.

A limited review of the spent fuel pool was conducted to ensure that

]

appropriate

loads, methodology and acceptance criteria were l

applied.

2.O ACCEPTANCE CRITERIA

)

The acceptance criteria for the evaluation of the spent. fuel rack applications are provided in the NRC OT Position for Review I

and Acceptance of Spent Fuel Storage 'and Handling Applications (Ref. 5).

Structural requirements and criteria given in this position paper were updated and included as Appendix D to Standard Review Plan 3.8.4, " Technical Position on Spent Fuel Pool Racks,"

(Ref. 6).

These documents state that the main safety function of the spent fuel pool and fuel racks is to maintain the spent fuel assemblies in a safe configuration through all environmental and abnormal loadings, such as earthquakes, and impact due to spent fuel cask drop, drop of a spent fuel assembly, or drop of any other heavy object during routine spent fuel handling.

Section 2 of SRP 3.8.4, Appendix D gives the applicable Codes, Standards and Specifications.

Construction materials should conform to Section III, Subsection NF of the ASME Code.

Design, fabrication and installation of stainless steel spent fuel racks may be performed based upon the ASME Code Subsection NF require-ments for Class 3 componen,t supports.

Requirements for seismic and impact loads are discussed in Section 3 of Appendix'D.

It states that seismic excitation along three orthogonal directions should be imposed simultaneously for the design of the new rack system.

Submergence in water may be taken into account.

The effects of submergence,are considered on a case-by-case basis.

Impact loads generated by the closing of fuel assembly to fuel rack gaps during a seismic excitation should be considered for local as well as overall effects.

It should also be demonstrated that the consequent loads on the fuel assemblies do nct lead to fuel damage.

Loads generated from other postulated events may be acceptable if sufficient analytical parameters are provided for review.

1 Loads and load combination requirements are provided in l

Section 4.

Specific loads and load combinations are acceptable if they are in conformance with Section 3.8.4-II.3 and Table 1,

Appendix D of the Standard Review Plan. Changes in temperature distribution should be considered in the design of the pool structure.

Temperature gradients across the rack structure due to differential heating effects between a full and an empty cell should be incorporated in the rack design.

Maximum uplift forces from the crane should be considered in the design.

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Section 5 discusses design and analysis procedures. It states that design and analysis procedures in accordance with Section 3.8.4-II.4 of the Standard Review Plan are acceptable. The effects

-of gaps, sloshing water, and increase of effective mass and damping.

due to submergence in water should be quantified.

Details of the mathematical.model including a description of how the important parameters are obtained should be provided.

Structural acceptance criteria are provided'in Section 6.

The acceptance criteria are given in Table 1 of Appendix D.

For impact loading, the ductility ratios utilized to absorb kinetic energy 1

should be quantified.

When considering seismic loads, factors of

. safety against gross sliding and overturning of the racks shall be in accordance with Section 3.8.5-II.5 of the Standard Review Plan unless it can be shown that either (a) sliding motions are minimal, impacts between adjacent racks and between racks and walls are prevented and the factors of safety against tilting are met, or (b) sliding and tilting motions will be contained within geometric constraints and any impact due to the clearances is incorporated.-

3.0 FUEL RACK DESCRIPTION The new high ' density spent fuel storage racks consist of individual cells with 6 inch by 6 inch nominal internal square cross-section, each of which accommodates a single General Electric BWR fuel assembly or equivalent, from either LSCS Unit 1 or Unit 2.

A total of 4073 cells and five defective fuel storage cells are

- arranged in 20 distinct modules of varying sizes. The arrangements of the rack modules in th9 spent fuel pool is shown in Figure 1.

A typical rack elevation is shown in Figure 2.

The modules make nominal surface contact between reinforcing plates at the top corner of each rack.

There are eleven different types of modules in the pool.

Tables 1 and 2 summarize the physical data for each module type.

The rack modules and their supports are fabricated from ASTM A-240, Type 304 austenitic stainless steel sheet and plate material.

The cells are held together with fusion welds.

Boraflex serves as the neutron absorber material.

Each rack module consists of the following components:

o Internal square box o

Neutron absorber material (Boraflex) - between adjacent cells o

Boraflex sheathing-external cells only o

Baseplates o

Corner support assembly o

Center support assembly o

Reinforcement plates at the top for impacts 3

g Figures 3 and 4 show ' a typical plan and elevation view, respectively, of a cell, showing the boraflex poison design.

The adjacent cells ot each module are welded together through fusicn welds.

A 1/2" haseplate is welded to the bottom of each cell with 3/32 inch fillet welds.

Each baseplate has a 4 1/8 inch diameter hole concentrically located with respect to each square tube.

These holes provide the path for coolant flow.

Each module has five support legs.

The corner supports are adjustable in length to enable leveling of the rack.

The variable height support. assembly consists of a flat-footed spindle which rides into an internally-threaded cylindrical member.

The cylindrical member is attached to a 1/2" thick base and gussets to form the corner assembly.

This assembly is attached to the underside of the rack module through fillet welds.

Figure 5 shows a vertical cross-section of the adjustable support assembly.

Figure 6 shows the center support leg which is not adjustable.

The support legs located near liner seams rest on la thick bridge plates on the spent fuel pool floor.

Figure 7 shows these bridge plates which include a 12" span centered on the liner seam to protect the seam from movements of the rack feet.

4.0 TECHNICAL EVALUATION

4.1 Euel Rack Seismic Analysis The spent fuel storage racks are seismic Category I equipment required to remain functional during and after a safe shutdown earthquake (SSE).

As described in Section 3 0, the proposed racks consist of 20 distinct free-standing modules which are neither anchored to the pool floor, attached to the side walls, nor connected to each other.

Any rack may be completely loaded with fuel assemblies, partially loaded, or completely empty.

The fuel assemblies are free to rattle within their storage cells.

Seismic forces are transmitted to the racks through friction l

j at the support leg to pool floor interface.

If seicmic displace-ments are large enough, the racks can slide and the support legs can lift off and impact the pool floor.

Because of these non-linearities, a time history analysis of nonlinear rack models was required to characterize the seismic response of the fuel racks.

BNL's review of the details of the modeling technique and analysis l

method is described in the following sections.

4.1.1 Dynamic Model The design basis analysis is a 2-D single rack analysis described in the Safety Analysis Report (Ref. 1).

During the review of the LSCS licensing report, concerns were raised that a 1

2-D single rack seismic analysis may not adequately predict rack 4

4 movements or forces.

In response to questions regarding 3-D and multi-rack behavior, Ceco provided additional analyses.

Two types of confirmatory seismic analyses were performed for the LSCS high density spent fuel racks. A 3-D single rack analysis was performed that accounts for the effects of three simultaneous earthquake components.

Separately, a 2-D multi-rack analysis was performed to predict the behavior of a row of racks.

The 3-D single rack evaluation is discussed in this section while the 2-D multi-rack analysis is discussed in section 4.2.

The 3-D single rack analysis is discussed in this TER.rather than the 2-D single rack analysis presented in Reference 1, because the generated loads and displacements are larger for the 3-D single rack analysis and because it serves as the most recent design check of the racks, fuel, and pool structure.

The mathematical model of the 3-D single rack module is shown in Figures 8 and 9.

The stiffness of the rack elements are determined by calculation and tests.

The analysis considers the Southwest corner rack number one.

This rack was chosen because it is one of the largest racks.

The fluid coupling for a corner rack is~ smaller than for an interior rack or edge rack.

The smaller fluid coupling will promote larger rack movements and impacts. The rack is considered full of normal weight fuel.

The analysis was performed for a combination of SSE (identi-fled as Design Basis Earthquake in FSAR) and SRV since the rack

. behavior is nonlinear and is controlled by the higher seismic levels.

The rack to wall gaps considered are the nominal gaps shown on the rack layout' drawings (in existence prior to the decision made to install the 1/4" reinforcement plates at the top of each rack). Gaps with adjacent racks are assumed to be 1/4" for fluid coupling calculations.

Gaps with adjacent racks for impact are assumed to be 1/8" for impact calculations.

The 3-D single rack model consists of a center stick repre-senting the rack and four sticks (one in each quadrant) repre-senting the fuel.

There are three levels of masses considered in the model for each of the sticks:

one at the base level, one at rack midheight and one at the top of the rack.

One quarter of the mass is lumped at the base, 1/2 at midheight and 1/4 at the top of the rack for rack, fuel, and fluid mass.

The vertical component

{

of fuel and rack mass is lumped at the base level including its representative rocking inertia.

Impacts between fuel and rack, rack to rack, and rack to wall are considered at each of the mass levels also.

The rack stick models the flexibility of the rack in flexure.

The fuel sticks model the flexibility of the fuel in flexure.

The members which connect the rack stick to the feet model the vertical flexibility of the rack.

The four corner feet and the center feet are represented by springs which model the local vertical flexibility of the rack above the feet and the overall lateral flexibility of the rack.

The feet springs are

]

.g.

attached to the rack stick with flexible elements which model the local flexibility.of the rack.

The sliding-uplifting motions of the lower end of pedestals on the pool floor are modeled by plane contact surfaces.

The contact conditions have the following properties.

o the points of contact are assumed not known apriori o

frictional sliding is modeled using coulomb friction with a coefficient of friction o

repeated contact (impact) and separation (tension release) is permitted in any sequence.

Fluid coupling between rack and fuel assemblies, and between rack and adjacent racks or walls is simulated by including inertial coupling terms in the equations of motion.

This is discussed in detail below.

Fluid damping between rack and fuel assemblies, and between rack and adjacent racks is conservatively neglected in the model.. In addition, the form drag opposing the motion of the fuel assemblies and the racks through the water is neglected.

In order to simulate the motion of adjacent fuel racks, the 3-D single rack model assumes a symmetry plane midway between adjacent racks.

Thus, the model assumes that each adjacent rack moves completely out of phase with the rack being analyzed.

This assumption is intended to predict conservative rack to rack impact forces.

The assumption that adjacent racks are vibrating 180 degrees out of phase seems to maximize the retarding effect of fluid forces l

and reduce the maximum impact velocities of the racks.

However, the out of phase assumption results.in numerous rack to rack impacts and is judged to cause the worst impacts based on the l

results obtained.

If the racks are moving in phase, impacts will be reduced.

Thus for maximizing impacts, this approach is acceptable.

To address the potential for in phase movements of multiple racks, a separate multi-rack analysis for a row of four l

l racks was performed (see section 4.2)

The replacement high density spent fuel racks will rest on bridge plates (Figure 7) on the pool floor.

The bridge baseplates span seam welds on the pool floor.

The bridge plates are not attached to the pool floor but have collars which engage the feet they support.

The baseplates were not included in the rack model but were assumed to move with the rack.

The baseplates are large enough to accom odate the possible slippage of the fuel racks during an earthquake without coming in contact with the liner seams.

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l Elastic coupling between the rattling fuel masses was used for the analyses.

This elastic coupling represents the stiffness of I

the fuel skeleton and its effects on the coupling of the rattling l

masses.

The fuel was modeled as two lumped masses at equally spaced elevations above the rack baseplate.

The fuel-rack impacts q

l would be expected to occur at the spacer grid locations and at the upper end fittings.

The selection of only two impact locations combined with the assumption that all fuel assemblies move in-phase was judged to result in conservative fuel to rack impact loads.

4.1.2 Fluid Coupling Effects The effects of submergence of the fuel racks in a pool of water has a significant offect on their seismic response.

The dynamic rack model incorporated inertial coupling (fluid coupling) terms in the equations of motion to account for this effect.

The rack and fuel model incorporates the inertial fluid coupling effects.

Fluid coupling terms' corresponding to fuel vibrating within the rack and rack vibrating within the adjacent pool walls and adjacent rack boundary are included in the er,uations of motion.

In tha single rack analysis, the fluid coupling terms for fuel bundles within the rack cells were based on the methodology presented in Reference 7.

In the case of the single rack vibrating within the adjacent walls and racks, because of unequal gap sizes and relative width to height ratio of the rack, it is difficult to determine an accurate flow pattern.

Hence, for this case, the hydrodynamic coupling terms for relative motion between the rack and adjacent walls and rack boundary, was computed using a 3-D finite element fluid and rigid rack block model.

ADINA (Ref. 8) was used for this purpose.

The model took into account unequal gap sizes and both horizontal and vertical fluid flow and also took into account coupling between two horizontal directions of motion of the rack.

The use of the methodology presented by Dong in Reference 7 to calculate the fluid coupling for the fuel bundles within the rack cells considers fluid flow through the fuel bundles.

This procedure is considered to be more realistic and is thus accept-able. The calculation for hydrodynamic coupling terms for relative notion between the rack and adjacent walls and rack boundary was questioned for its accuracy.

The Licensee stated that the results of the ADINA 3-D finite element fluid and rigid rack model were compared to a simplified case in Fritz's paper (Reference 9), which is a common approach used by other plants to design their fuel racks.

The results for the two methods were comparable and thus, the methodology used to calculate the hydrodynamic coupling terms is considered acceptable.

7

4.1.3 Friction Effects Friction elements were used at the bottom of rack support leg elements of the model.

The value of the coefficient of friction was based on documented test results given in Reference 10.

The results of 199 tests performed on austenitic stainless steel plates submerged in water showed a mean value of coefficient of friction to be 0.503 with a stancard devj ation of 0.125.

Based on twice the standard deviation, the upper and lower bounds are 0.753 and 0.253, respectively.

Two separate analyses were performed for each load case with values of coefficient of friction equal to 0.2 (lower limit) and 0.8 (upper limit), respectively.

The use of both an upper and lower bounding value for the coefficient of friction is judged to be appropriate.

Previous studies have indicated that low friction results in maximum sliding response of the racks while high friction results in maximum rocking or tilting response.

Consideration of both cases should provide worst case displacements, stresses and impact loads.

4.1.4 Damping Damping of the rack motion would develop from material hysteresis (material damping),

structural deformation of the interconnected components (structural damping) and fluid damping effects.

In the analyses of the LaSalle racks, a maximum of 4%

structural damping was utilized during the SSE.

This 'is within the values specified in the FSAR and USNRC Regulatory Guide 1.61 (Ref.

11).

Other damping contrib,utions are conservatively neglected.

Structural damping was included in all of the structural elements.

The damping matrix [C] was formed by a linear combi-nation of the structural mass matrix [M] and stiffness matrix [K]

as follows:

[C] = c4 [M] + g [K]

where o< and # are related to the damping coefficient ( A = 0.04) and the frequencyur, of the r,ystem by:

d A,,

+ _

2%

2 4

The values of oc. and # are selected such that for a frequency range of interest A is equal or less than 0.04.

The above methodology for treating damping represents an acceptable approach for the seismic analysis of the LSCS racks.

4.1.5 Seismic Input Motion The seismic loads applied to the LSCS fuel rack model were 8

_________m._

o 4

three acceleration time histories corresponding to the three global directions (North-South, East-West and Vertical).

These motions i

were synthetically developed time histories which were based on the f

fuel pool design response spectra.

The three time histories are i

shown in Figures 10 to 12 for the SSE load case.

The artificial time histories were checked by the Licensee for statistical i

independence between the three motions.

J To permit the use of the existing fuel pool design response i

spectra, the Licensee determined the increased mass of the new high density racks with fuel and compared it to the mass of the building j

structure.

Since the increase in mass was small (approximately 1.5%), it was concluded that the increased mass due to the high density racks should not significantly affect the overall dynamic i

response of the building.

The " initial" 3-D single rack analysis conservatively used an envelope of vertical floor spectra which considered floor slab amplifications for all areas at the pool elevation.

Since the pool-slab frequency remains in the rigid range no additional amplifi-cculon of pool responses will occur.

Thus, a "refinedd 3-D single rack analysis was run using the floor spectra at the wall thereby removing the conservatism in the vertical spectra.

For other changes included in the " refined" 3-D model see section 4.3.

A comparison of the pool design response spectra and spectra generated from the synthetic time history was provided and reviewed.

The broadened design spectra were used to make the comparison with the response spectra of the synthetic time histories.

The comparisons for all three directions were very good.

Based on the Licensee's description and the information reviewed, the methodology used to develop the seismic input for the fuel rack seismic analysis is acceptable and consistent with industry practice.

l 4.1.6 Analysis Method l

The analytical model described above was analyzed using ADINA (Ref. 8) computer code.

The rack model was subjected to the loading caused by the simultaneous action of three components of l

earthquake motion at the fuel pool elevation and the dead load of the fuel rack system.

The analysis procedure used the direct integration technique for solving the equation of : notion.

A time j

step increment of 0.002 second was used and deten.ined to be small enough for the. highest frequency of interest in the dynamic response of the fluid rack system.

The solution to the dynamic equilibrium equations is obtained using the full Newton interaction with stiffness matrix reformatted at the beginning of each new time step and each iteration.

l 9

4-4 7-The public domain program ADINA was used to perform the 3-D single rack analysis.

This program was verified by the supplier n"L and by S&L under the requirements of S&L's QA program.

This program has been used and reviewed / accepted by the NRC for fuel 4-rack analysis on the Byron docket, 50-454.

4.1.7 Analysis Results The ADINA program computed displacements and forces at each instant of time during the earthquake.

Stresses for the rack were computed from maximum forces occurring in any one analysis.

These stresses were checked against the design limits.

The load combinations and acceptance limits were reviewed and were found to agree with those presented in Appendix D to Standard Review Plan 3.8.4 (Ref. 6).

Maximum calculated pedestal forces and maximum impact forces (rack to rack, rack to wall, and fuel to ~ rack) are presented in Tabic 3.

Forces, stresses, r.nd safety factors at critical rack components are summarized in Table 4.

Since the Southwest corner rack was analyzed and it impacted against the South wall and West wall, the maximum displacement of this rack was 5 1/4 inches in the West direction.

Information dealing with the potential damage to the fuel resulting from fuel to cell wall impacts was requested.

The Licensee indicated that lateral and vertical impact between the fuel and the rack would not damage the fuel.

The Licensee demonstrated that the fuel in the racks are subjected to smaller accelerations than the fue"1 in the reactor core.

4.1.8 Evaluation of Results The results of the Licensee's seismic analysis indicated that all stresses in the racks would meet their allowables and impact

' loads on fuel assemblies would not damage the fuel.

In addition, with the installation of reinforcement plates at the top corner of each rack (except those near the containment where the wall slopes away),

the seismic analysis demonstrated that the structural integrity of the racks is maintained under the rack to rack and rack to wall impacts.

However, considering the potentially unconservative modeling assumptions regarding multiple rack behavior and out of phase motion between racks, the Licensee performed an additional analysis to address such concerns.

This analysis is summarized in Section 4.2 and the overall assessment of the seismic analysis results is given in Section 4.5.

10

.1

e-4.2 Multi-Rack Seismic Analysis As a result of concerns related to the adequacy of a single rack model in predicting forces and displacements that would occur if multi-rack effects were considered, additional analyses were

(

)

performed.

A description of these additional analyses and their results are provided below.

4.2.1 ' Multi-Rack Model This seismic analysis consists of a row of four racks to i

investigate the adequacy of the deign basis single rack models in predicting the response of fuel racks in the actual multi-rack fuel pool environment.

An issue of particular concern was the possi-bility that in a multi-rack environment, a row of racks may pile up on one side of the pool and hit the wall with large impact forces. Although the walls were originally designed to accommodate seismic loads from the existing fuel racks, impact loads on the wall could damage the walls or liner resulting in unacceptable leakage of water from the pool.

The following provides a description of the modeling para-meters used in the 2-D multi-rack analysis:

The West row of racks that include rack numbers 1 through o

4 were modeled and analyzed.

This is an edge row so rack to rack fluid coupling is less than an interior row and displacements will tend to be larger.

The following fdel loading was considered in the racks o

starting from the North wall:

empty, half full, full, and three quarters full.

This loading was selected to promote out of phase response and maximum rack displace-ment.

Normal unchanneled fuel was considered in this analysis.

The multi-rack analysis was performed for coefficients o

of friction of 0.2 and 0.8 to cover the lower and upper limits.

The rack to wall gaps considered were the nominal gaps o

shown on the rack layout drawing (in existence prior to the decision of installing the reinforcement plates).

Gaps with adjacent racks were assumed to be 1 inch for fluid coupling calculations and 1/4 inch for physical impact calculations between racks.

The larger fluid coupling gap was selected to conservatively reduce fluid coupling effects and conservatively estimate impacts and displacements.

Sketches of the model are provided in Figures 13 and 14, and a description of the model is as follows:

11

b.

4 q;

i The - two-dimensional four rack model' consists of, for each rack,La center stick representing the rack and onesstick repre-senting the fuel. There are six levels of masses considered in the-

.model for each of the sticks, one at the base level and five above t

-the base.

One-tenth of the mass is lumped at the base, 1/5 at the next.four levels, and 1/10 at the top of the rack for rack, fuel,

-i

~

.and fluid mass..The vertical component of fuel and rack mass is lumped.. at the base level including its representative rocking inertia.

' The. fluid hydrodynamic coupling terms. for relative motion l

between the rack and wall or adjacent rack are computed using the H

Fritz model.

The fluid coupling terms for the fuel ~ consider flow through the bundle and is based on Dong.

Potential impacts between fuel and rack, rack to rack and rack

- to wall are considered at each of the mass levels.- Stiffness for the impact springs corresponds to the local flexibil'ity of the rack and'is calculated'from test results.

The rack stick'models the

' flexibility - of the' rack in flexure.-

The fuel sticks model the flexibility of the-fuel.

The members.which connect the rack stick to the base model the vertical flexibility of the rack.

The four corner supports and the center support are represented by springs which ~ model ' the local vertical flexibility of the rack and the overall lateral flexibility of the rack.

The support springs are attached;to the rack stick with flexible elements which model the vertical flexibility. of the rack.

Rack-sliding is modeled by sliding surface at the base of the supports.

The friction force is the concurrent normal -force multiplied by the coefficient of friction at a specific time during the seismic event.

4.2.2 Multi-Rack Analysis /Results The 2-D multi-rack model shown in Figure 14 was analyzed with the.RACKOE computer program, which was developed and verified by U.S. Tool and Die (UST&D).

The analysis was performed for the combination'of SSE,and SRV sines the higher seismic level will control the dynamic response. The North-South time history and the vertical time history were applied simultaneously.

These time histories are the'same as those utilized for the initial 3-D single rack analysis (see section 4.1.5).

The key responses (pedestal

forces, impact

fcrces, and displacements are presented in Table 5.

Comparisons with the initial 3-D single rack and refined 3-D single rack analysis are

. presented in Table 6.

Tne comparison with the initial 3-D single rack model indicates that the forces are generally smaller for the multi-rack model.

Comparison with the refined 3-D model indicates comparable pedestal forces and smaller impact forces for the multi-rack.model..The comparison with the initial 3-D single rack model is more meaningful since they both used the same vertical dynamic l

12 c,-..

input while the refined 3-D model eliminated the conservatism in the vertical amplified floor spectra.

The impact forces for the multi-rack analysis were much smaller than the forces from the initial 3-D analysis and the refined 3-D analysis.

This is probaoly due to having only a 2-D model and 2-D input time history.

The addition of the third input simultaneously, would increase the response, particularly if tilting occurs about both horizontal axes thereby causing torsional rotation about the vertical axis as well.

A review of the displacement time history response did show movement for all four racks in the North direction until impact with the North wall occurred.

In addition, the motion of the three racks having partial and full fuel assemblies were very similar indicating strong hydrodynamic coupling effects while the empty rack motion was more erratic with very little correlation to the other three rack motions.

4.3 Rack Impact Evaluation A major concern which arose during the review of the rack responses was the extremely hich rack to rack and rack to wall impacts. The initial 3-D single rack analysis calculated an impact force of 172 kips at the top of the rack for rack to wall impact.

Since the rack rotated about its vertical axis, this impact force would be applied nat as a uniform pressure across the face of the rack but as a lino load.

The resulting stresses would be quite high possibly exceeding the buckling stresses for the cell wall material.

Thus, the Licensee developed the " refined" 3-D single rack model to eliminate some of the conservatism inherent in the

" initial" 3-D model and thus obtain more realistic impact forces.

The following improvements / revisions were incorporated into the i

" refined" 3-D model:

l The vertical response spectra at the pool wall was used o

as input rather than the envelope of anplified floor slab response spectra (see section 4.1.5).

In addition, 4%

damped spectra were used to develop the synthetic time history rather than the 2% damped spectra (conservatively used in the initial 3-D analysis as if it were 4% damped spectra).

l l

o The gap springs between the rack to wall and rack to adjacent rack were distributed along the width of the rack at the top, middle and bottom elevations.

o The fuel mass was redistributed to account for the fuel being 13 inches below the top of the rack.

13 l

C___._

l l

+

o The horizontal stiffness at the base of the rack was modified to account for the center pedestal (original total stiffness was based on the assumption that there are only four pedestals at the corners).

The results of this analysis are presented in Table 4.

Comparisons with the " initial" 3-D analysis show total rack to rack impacts are about the same, while rack to wall impacts are lower for the " refined" 3-D analysis.

The rack to wall impact at the top mass actually reduced substantially from 172 kips to 119 kips.

The rack to wall impact was judged to be more critical for the rack adequacy than the rack to rack impact since there is more flexibility in impacts between two racks.

A detailed 2-D model representing the cells near the top of the rack was developed to analyze the local effects of the impact force.

This model is shown in Figure 15.

As a result of this analysis, the. compressive stresses in some cell walls exceeded the allowable buckling stress.

Thus, the Licensee proposed to install reinforcement plates as shown in Figure 16, to distribute the impact force over a suffi-cient number of cells where the compressive stresses are low enough.

These reinforcement plates will be installed at the top four corners of each rack except those next to the containment where the containment wall slopes away.

The middle and bottom of the racks will not receive such plates because these areas are less critical.

At the bottom elevation of the racks, there are 1/2 inch thick base plates in each cell which.would transfer the load.

In the middle of the rack, the full available cell length is engaged to rcsist the load with no " free end" effeet which exists at the top.

The continuity of the cell will result in a higher buckling capacity of the cell wall when compared to the top.

In addition, there are a few other conservatism inherent in the analysis which would reduce the calculated impact force (e.g. with the 1/4 inch plates, the angle of impact would be greatly reduced thereby distributing the force over more cells in the horizontal direction).

With the addition of the 1/4" thick reinforcement phtes as shown on Figure 16, the impact capacity of the racks nas been demonstrated to be acceptable.

4.4 Thermal Analysis Weld stresses due to heating of an isolated hot cell were computed.

Ihe analysis assumed that a single cell is heated over its entire length to a temperature above the value asscciated with the adjacent cell. No thermal gradient was assumed in the vertical direction. Using a temperature differential of 36*F, weld stresses near the top of the cell were found to be below the allowable value.

14 L_

I 4.5 Overall Evaluation of Seismic Analysis Results Although the multi-rack analysis did not account for three-dimensional cross coupling effects, it is reasonable to judge that it captured the primary multi-rack response and it did provide the interaction forces and displacement behavior of a row of racks in one direction.

Even with its limitations the multi-rack analysis has some conservatism which include (1) fluid cross-coupling effects were not included, (2) friction effects between racks of adjacent rows during impacts were not included, (3) much larger vertical spectra, (4) all fuel assemblies at a given elevation vibrating in phase, and (5) no fluid damping.

The initial 3-D single rack analysis contains the same conservatism listed above while the refined 3-D analysis for the single rack contains the same conservatism less item 3.

The design adequacy of the fuel racks subjected to the governing SSE load case has been shown to be adequate for the loads from all three models (multi-rack, initial 3-D single rack, and refined 3-D' single rack) with the exception of impact loads.

For consideration of impact loads, the conservatism in the vertical seismic spectra was removed, the single rack model was refined (distributed impact springs) and reinforcement plates at the top corner of each rack had to be added.

This resulted in sufficient rack impact capacity for the calculated loads obtained from the refined 3-D model.

Based on the above discussion and the. ample design margins shown in Table 4, it has been demonstrated that the racks meet the current licensing requirements.

Therefore, it is concluded that during the SSE load case, the fuel racks will maintain their structural integrity and the fuel assemblies will not sustain damage.

4.6 Fuel Handling Accident Analyses The Licensee performed structural analyses and evaluations for four postulated fuel handling accidents.

The four types of fuel handling accidents considered are:

.1 Straight Fuel Drop Onto Top of Rack A 680 pound fuel assembly dropping 30 inches on top of the rack was assumed.

This input energy was used to calculate the plastic Qformation in the cell walls based on actual cel.1 box crush tests. Using this approach, the vertical plastic deformation was calculated at 2.7 inches.

This limits the deformations at the very top away from the active fuel zone.

15

^

q J.,

j a

f y

.2 Inclined' Fuel. Drop Onto Top of Rack j

Since the inclined drop would distribute its impact over more than one cell, the plastic deformation for this case would be 1ess severe.

)

.3 Straight Fuel Drop Through the Cell i

This analysis concluded that the dropped fuel assembly would have sufficient energy to break the welds holding t

the individual cell baseplate to the cell. Although, the accident would render one storage cell location unusable,.

l the Licensee concluded that the physical configuration of the spent fuel storage cell will not be changed.

Therefore, the subcritical array of the rack would be maintained.

As for the pool liner, an analysis was performed to determine if the fuel drop would penetrate the pool liner.

This analysis was conservatively performed since the velocity of impact was calculated neglecting the negating effects of

buoyancy, skin

friction, and stagnation drag force.

In

addition, the energy dissipated due to deformation of the fuel assembly was neglected.

If the liner plate alone was considered, the fuel assembly would penetrate the plate.

However, the formulation took advantage of the' concrete beneath the liner plate to successfully demonstrate that the liner plate would not be penetrated.

a s

l

.4 1200 Pound Uplift Due to Fuel Jamming A 1200 pound uplift force and a 1200 pound plus one fuel assembly weight downward force were each applied to a single cell separately.

The most critical stress was calculated to be 1,231 psi (in the welds) which is well below the allowable value.

Based upon the above discussion and review of the general methodology, the structural adequacy of the racks and pool liner f

under the postulated fuel handling accidents has been adequately demonstrated.

4.7 Spent Fuel Pool Analysis The LSCS pool was initially evaluated for the increased loads from the new fuel racks by Stone & Webster (SWEC).

Sargent and Lundy (S&L) performed an independent evaluation of the pool and evaluated impacts and fluid coupling forces on the pool.

This includes an evaluation of rack impacts on the pool structure.

The SWEC evaluation is described first and then the S&L independent evaluation is described.

16 m_-__m_______._--

l L

4.7.1 Loads and Load Combinations 1

The following design loads were considered in the reanalysis of the spent fuel pool.

o Normal operating and accident temperature loading including thermal gradient loads.

o Dead loads including spent

fuel, fuel

racks, cask

storage, self weight of structure and N drostatic f

pressure.

o Live loads from adjacent slabs.

o Hydrodynamic forces and excitation of dead load for OBE, SSE, SRV and LOCA.

The above loadings were combined using load combinations per LSCS Mark II D.A.R. Table 4.3-1.

Additional loading combinations were also evaluated deleting only the thermal load from any of the above load combinations.

The acceptance criteria and the allowable stresses are shown in LSCS Mark II D.A.R-The resulting stresses were evaluated against the following E:ceptance criteria:

o Maximum allowable steel stress = 54 ksi o

Maximum allowable concrete compressive stress = 3.825 ksi 4.7.2 Spent Fuel Pool Structure Analysis (SWEC)

The pool was analyzed by SWEC by the finite element method of analysis using the ANSYS computer program. The pool slab and walls were modelled using 3-D isoparametric solid elements in three layers. Appropriate boundary conditions were imposed on this model to adequately represent the effects of the unmodelled portions of the structure.

The assumptions made in the analysis were:

1.

The boundary at the containment was considered fixed for the purpose of obtaining maximum loads. at the boundary.

2.

The spent fuel and the rack loads were uniformly distributed over the pool slab.

3.

Maximum stresses in the pool were obtained by enveloping the results of the analysis considering the boundary at the containment as fixed and as hinged.

17 m_m___.__._________

l:

F 4.

A cracked section analysis was performed for thermal loads only. Analysis for mechanical loads conservatively considered an uncracked section.

5..

The model included the entire spent fuel pool and up to the centerline of adjoining new fuel storage pool with appropriate boundary conditions to simulate the continuity of the entire structure.

6.

Following ACI 349-85, App. B, thermal loads were not included where they reduced stresses.

The analysis for the mechanical loads was performed using an uncracked model.

The SWEC analysis concluded that the fuel structure could accommodate the design basis loads including those resulting from the replacement racks.

l 4.7.3-Spent Fuel Pool Structure Analysis (S&L)

S&L performed an independent evaluation of the LSCS pool for the high density spent fuel rack loads including rack fluid coupling forces and wall impacts.

This evaluation considered the critical section identified by SWEC in their evaluation and sections determined to be critical in the initial pool design. The critical sections are shown in Figure 17.

The evaluation con-sidered the load combinations shown in the FSAR and the DAR l

including the load combinations determined to be critical by SWEC in their evaluation.

~

The S&L evaluation ac~ ounted for the redistribution of moment c

that would occur due to cracking of the periphery of the pool slabs where high uoments occur. This effect was conservatively neglected in the SWEC evaluation.

The assumed amount of cracking was confirmed to be appropriate based on the cracked concrete section analysis results.

The fuel pool design basis finite element model was utilized for the S&L evaluation.

The pool slab and walls were modeled utilizing a mesh of quadrilateral plete elements.

The finite element SLSAP model incorporating the cracked concrete section was used for the analysis of dead weight and seismic loads from the rack and the impact load on the floor slab.

The programs used in the, S&L evaluation were SLSAP and TEMCO.

These programs have been reviewed and accepted by the NRC in previous similar applications.

The latest pool structure analysis considered the rack impact loads from the 3-D single rack analysis.

These leads on the wall and slab were applied at critical sections to produce maximum effects on the pool structure.

18

Table 7 summarizes the results of the S&L evaluation. The S&L independent review concluded that the SWEC evaluation of the pool yielded the appropriate critical section and load combination and provided very conservative stress results.

Considering moment redistribution based on cracking yields a more accurate picture of

~

pool stress and shows from Table 7 a factor of safety of 1.15 above allowable stresses.

The stresses and safety factors shown on the table include the effects of any one fuel rack impacting the wall or the slab in combination with the appropriate loads in the critical load combination at a given critical section under consideration.

The results of the evaluations performed by SWEC and S&L provide good assurance that the pool structure is capable of supporting the new high density racks filled with normal weight fuel.

5.O CONCLUSIONS With the addition of reinforcement plates at the four top corners of each rack, the impact capability of the racks has been demonstrated.

All critical stresses in the racks have been shown to be less than the allowable values.

It has also been shown that impact loads generated between the fuel assemblies and cell walls would not lead to damage.

Furthermore, it has been demonstrated that the existing spent fuel pool has adequate capacity to accommodate the loads resulting from the high density racks and fuel assemblies.

Based on the review and evaluation of the Licensing Report, additional analysen and information provided by the Licensee during the course of this review, and the above discussion, it is concluded that the proposed LSCS Unit 2 fuel racks and pool structure have sufficient structural capacity to withstand the effects of all required environmental and abnormal loadings discussed in this report.

l l

19

i,

6.0 REFERENCES

1.

Commonwealth Edison ' Company (CECO) letter, "LaSalle County

' Station Unit '.t Proposed Amendments.to Technical Specification for Facility Operating Licensee NPF-18 High Density Spent Fuel Racks NRC Dockot No. 50-374," dated 9/19/86.

2.

NRC letter to CECO, " Additional Information on Re-Racking of Spent-Fuel Pool LaSalle County -Station, Unit 2,"

.E.G.

Adensam to D.L.

Farrar, dated 11/28/86.

3a.

CECO' letter to NRC, High Density Spent Fuel Racks NRC Docket No. 50-374, C.M. Allen to USNRC, dated 8/18/87.

3b.

Ceco letter to NRC, Response to NRC Request for Additional Information; High Density Spent Fuel Racks NRC Docket Nos. 50-373 and 50-374," W.E. Morgan to USNRC, dated 6/6/89.

4.

NRC letter to CECO, P.C.

Shemanski to H.E.

Bliss, dated 4/18/89.

5.

USNRC letter to all power reactor licensees, from B.K. Grimes, dated April 14, 1978, "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," as amended by the NRC letter dated January 18, 1979.

6.

US Nuclear Regulatory Commission, " Standard Review Plan for the Rev3aw of Safet-y Analysis Reports for Nuclear Power Plants," NOREG-0800, Section 3.8.4, Revision 1, July, 1981.

7.

R.G. Dong, " Effective Mass and Damping of Submerged Struc-tures" Lawrence Livermore Laboratory, UCRL-52342, April 1, 1978.

8.

ADINA Engineering Report AE 84-1 S&L Program No. 09.7.199-3.0.

9.

R.J. Fritz, "The Effects of Liquids on the Dynamic Motions of Immersed Solids,"

Journal of Engineering for

Industry, Transactions of the ASME, February, 1972, pp 167-172.

i 10.

E.

Rabinowicz, " Friction Coefficients for Water Lubricated Stainless Steels for a Spent Fuel Rack Facility," a Report for Boston Edison Company, MIT, 1976.

11.

US Nuclear Regulatory Guide 1.61, " Damping Values for Seismic Design of Nuclear Power Plants," October 1973.

20

--___s

i TABLE 1 TABLE OF RACK DATA Rack Number of Number of Cells Number of Cells Total-Number Number Racks In N-S Direction In E-W Direction Of Cells Per Module 1

1 16 15 234(1) 2,3 2

16 15 240 4

1 16 15 217(2) i 5

1 15 17 235(3) 6

.1 15 15 225 7,10,11,14,15 5

13 15 195 8,16 2

15 15 202(4) 9,12,13 3

12 15 180 17 1

12 15 174(5) 18,19 2

14 15 202(6) 20 1

13.

15 185(7)

(1) Total cells ec;uals 16x15 lese., 3x2 (2) Total ce) s equals 16x15 less 3x5,1x3,1:<1, 2x1, Ix2 (3) Total cells eauals 15x17 leys 10x2 (4) Total cells equals 15x15 less 2x8, 1x3, 1x4 (5) Total cells equals 12x15 less 3x2 (6) Total cells equals 14x15 less 2x4 (7) Total cells equals 13x15 less Ex2

7--,

x--

'),

(.

v.

4.4 h s

t

.s 1:

1 i

TABLE 2-i RACK DIMENSIONS AND WEIGHTS PA1 Approximate-Rack Number Nominal Cross-section Estimated Dry Dimensions (inches)'

Weight (LBS) 1 Per Rack N-S E-W 11-100.3.

94.0 28,890~

2,3' 100.4 94.0 29,570 4.

~100.3 93.9 26,880 5.

93.9

~ 93.9 28,990 l

6.

93.9' 93.8 27,800

. 7,10,11,14,15

.93.9 81.4 24,900 I

,s

'8,16 94.0 93.8 25,110

.9.12,13 75.2 93.9 22,540 17-74.9 94.2 21,830 q

-18,19 87.5 94.0 25,110 l

1 20.

81.3 93.9 23,120 l

r l

N i

i

)

1

_ - - - - - _ - - - _ = _ _

l TABLE 3 REFINED 3-D SINGLE RACK Maximum Pedestal Forces l

Friction Axial Force Shear Force Coefficient (kips)

(kips)

Foot 207 p = 0.2 38 Rack-192' Foot 247 p = 0.8 174 Rack 231 Maximum Impact Forces Friction Location Rack to Rack Rack to Wall Fuel to Rack Coefficient (kips)

(kips)

(kips / Fuel)

Upper 117 119 0.34 p = 0.2 Middle 220 228 0.42 Bottom 209 221 N/A Upper 98 119 0.36 p = 0.8 Middle 166 179 0.45 Bottom 154 163 N/A i

k

[

i TABLE 4 i

REFINED 3-D SINGLE RACK MAXJMllM SSE LOADS, STRCSSES AND SAFETY FACTCNS i

i FOR 240-CELL RACK COMPONENTS WITH STANDARD FUEL

{

_m_ C -MU N ???

i l<

{

I3II jll (7) l I I i ,I i s 8 O ) '.,, (3),(4) = mas (S) \\ REVISED 1-RACK 3-D ALLOW - MIN. COMPCNENT MAX. MAX. ABLE SAFETY 111PER I PTI ON LOAD STEEES STEE,9E T6CTOE (KIPS) (KST) (ESI) I Tl!RPAD HEGION 247.0 5.07 24.22 4.76 2 Fusion WELn' V 214 4 17.an 29. vis 1.63 H 173.7 3 BASEPLATE V 247.0 26.13 29.06 1.11 TO CELL WAl.h N 173.7 4 PED Pl. ATE To N 173 7 16.80 29.06 1.75 BASUPLATE WELD 5 eunsET PLATM V 147.0 20.20 36.0 1.78 H 173 7 6 cENTen PUD To V 3 F.J 13.09 29.06 1.22 cr.LL WALL H 23.+ 7 FUEL To RACK .223 20.37 28.8 },48 AMDACT I i

• considers vertical snear, thornal shear,and inpact shear i

7,. c. TABLE 5 MULTI-RACK SEISMIC ANALYSIS RESULTS '(KIPS & INCHES) i, Component Rack i Rack 2 Rack 3 Rack 4 Axial Force on Foot 191.0~ 236.5-208.8 174.e Shear' Force on Foot 152.5 153.9 191.8 139.1 Rack to Rack Impact 25.8 18.8-12.9 . Rock to Wall Impact 33.8 Displacement

1. 5 1.66 1.55 2.77 Minimum rack' factor of safety - Fusion veld = f,/4 f..:

___m__.___._

m-m,, i ' TABLE 6 COMPARISON'OF RESULTS ~ INITIAL 3-D-REFINED 3-D SINGLE RACK SINGLE RACK MULTI-RACK AXIAL FORCE ON FOOT (KIPS) 555 247' 334-SHEAR FORCE ON FOOT (KIPS)' 82 174 217

• RACK TO RACK IMPACT (KIPS) 520 546 155**
• RACK TO WALL

-IMPACT (KIPS) 680 568 203**

• • • DISPLACEMENT (IN.)

5,25 5.25 2.77

• This represents total impact over entire rack height.
  • This was conservatively calculated as no. of mass levels (=6) times the maximum impact occurring at any one mass level.
    • -In.all three cases the maximum displacements correspond to sliding until impact occurs with adjacent fuel pool walls.

P 4 L 1 TABLE 7 l Spent Fuel Pool Structure Maximum Tensile and Compressive Stress Summary For Design Basis, SRV and LOCA Load Combinations Rebar Concrete Section Max Tensile Stress (ksi) Safety Max. Compressive Safety Horizontal Vertical Factor Stress (psi) Factor 19 23.1 46.9 1.15 2,344 1.63 20 45.0 18.9 1.20 2,651 1.44 21 44.8 39.4 1.20 2,600 1.47

1. Allowable Stress = Concrete : 0.85 f'c = 3825 pai Rebar

0.90 Fy

= 54 kai

2. See FigureI7 for locationg of the sections.

3. Sections 20 and 21 are slab sections where, Horizontal = NS Vertical = EW 4. Safety Factor = Allowable Stress / Actual Stress 5. Minimum safety factor for pool structure for shear = 1.31 6. Minimum saftey factor for Normal Load Combination = 1.53 4 A#%~~

r . s, FIGURE 1 ?- N I i l l 18 CONTROL ROD STORAGE '. : TI '#3[ji-iiT [ j i i"' T,;_,,_, I !I;.. 9 _-l ,y

• 4W4Wi 4-tH + Hi-< - HT

^ t-1

II III I I 1.

I. I I --- ', ; j I [ t i I g y C 234-N 240' -::N 240 21'7 ~ (15x16-6)! t:(15x16g_g'E I C ~ gj }5x16,') J'it '- 15x16-2 3 1 !Il I I ! ! I N' .. l I I. Iy [ y Mf g g g i j i, 1 1.

++

~I ^! T. ^!!. 1. !. !!! M ;_.1

o2 11 1 -

1 1 _ i c m.1 1. 1 3, _ 1 :

+_(15x15 f r 235

g 225 O

E (15x17 p (15x15) g (19515x1.9. 73)

i i

. 1 1 = 20) -->q, 1 _.-.mo44 y :!t

I '

2_; 1 i I i ii ^Jj _ 2 FUEL ';180 ; 195."'$' J95 160 PREP 1(15xliE (15x13) x (15x 13) - :,15x12) v i - i.,iii MACHINE 11 m 1: : 1 1 +- 'i!!j' - i ll: ';;>.; ' ~ 1 i 1 ; . ;j: i 1 l i~ ~ -; J i; '

t :

$ 180 395 it 195 ,dg3h,57 1(15x3 2I; (15x3 3) t(15x13) 22 17 CONTROL -e -- 23) ROD STORAGE CHANNEL I. 1 '^ STORAGE - i: RACK [.;;i;;;, ; ..i

_De, PLATFO I

i [ FUEL - --. 174 _ ;.:. 202

(15gl4-165 5-

:

+ 35 4;:~Z (15x13-::-~

[15x 12.

) 1 1 ' 3o ~~-- 6) ~ PREP L r 1.,, MACHINE g d li:. 1 i: -l i u

i:

5 DEFECTIVE FUEL 4073 NOR ML . SPENT FUEL STORAGE STORAGE 3 CONTROL ROD GUIDE TUEE STORAGE ' PLAN ARRANGEMENT OF LASALLE COUNTY STATION UNIT 2

{' a FIGURE 2 i . V.., I f"* 9 P. P"*'"'*""t t" ,P" M*""*t .r tg 9 P gs"* 7 9*" 9 t'-*3, 7 3 o 0 (> G c 3 i e a s C .. w . G s w &. s is O ) d e'mr,e -' mmmv-- >mmmr%mmmmmm .~i> N l9 N 'N 'N e .N N f4 N N ID N N l N '- N l% 8 s

1. e N

Y t N N D N N N'% N N N~ N-l N N N ~, Q ~_~~- - -- - - i 8 i t

p. :-c..

,'IN). W./- L_ ..s.. o t TYPICAL RACK ELEVATION i I i 1

4 FIGURE 3 w

6. 2 6~5 A.* %... -.O. !..~.; L_' _

l \\ ShJ t $f 4(j',*, i9et A NtDuc T'r'N fc*k& r_c p;/ V'.m TH /39 L ou A e ' **%HO)J( o OE;G 9r/c d 8 ~/0)_ w ~. & %-,// '~ ,,, 's c Usix is ,y . / s' O N'< /, < ..,.x Y ,/,,/, s ' s' //_- s s s s i ( s s a s s a I / -(,07f A'24. R iSONh 8&; . <><s.ww. . c8 2 mar. ~ ~ ~ 4 (090 nom. V//.L L) ~ ~ ( . on u,. . Ce y W.u. m ,Iy ' ^*'12 84': ARE.),C 9=.n'.w_' A-4,

5. !N,41A,K.

' ) ? s n A VL RA S, E, % 1A NL 'r'.~_f C # _1 l..'_; _E__b_c_.(a-v y t' ? s ( 6.180 r. 020 M 'A RE ).Q.t~c r[].r-.- q 'k, % .e _u x V

1. r4 s

v, ,1 hj a b AVERAGE })cRnS1.Rh1 SED. dEEAS A _y mal \\f LTAC77/ RED (62er =.o25 nunne L 4, y 9 T~7 1 V y ~ e N ,h ,b 'Q gf I ~I I 3 \\ i TYPICAL CELI. PLAN

r_ O FIGURE 4 li ~-if hda TT{,,,A ~!DCA L. i t/il.* */ WEL D Y b 51/RS, EAc.14 WAj.L. s g 325 g; a I a w I I l y I, k .se R. s r 7'rP l I I i i i l e n I 'l - jr L i 4'/3 g 54 + 4-1 L): l h N07*CHES l'! N ')d 42 ace s m w 'l M I BORAFL E X k Poisont Y. M A Ti"/>IA L d l' (z%. .o _,02 0 c/t. l' n

• Q l

9 = + = p> I l c q 7 rora mea O I

a. n I

u u 7n = 2 % & sa rents I k i t. .l E t e- - a d 't 1 l l I 'c. (f ~ I 0 l l W k l 0 k 9 I = o, l s N t P I L g i TYPICAL CELL ELEVATION

FIGURE 5 E.'/ "X i8 /8"X l'-6 jg 2 5 S a E.3/4" 8 8-PLACES 5"g 5 /4" N-28... = THD. THRU-it / t \\ h' / / \\/ /////// / ///////1 h [ [ l (/ h k s 6" _ _., I.' t-N 3,- s\\,. v.1,. 7 7;i, q ,q ee q - - c_ i N t. i, .g. i m / *. )[ l 5'/4"- 8 UNC-2 A i TYP\\ f, m /4 / i ~CHRO*1E PL ATE THE /l i 5 g/16 \\5 TYPV5 /5 I' 'l E.1 7 '/2". & / gil*x s's - P CHAMFER._ ld ..s.. ..s Ref. Dwgs. U.S. Tool & Die, Inc. 8601-26 - ~4 8601-27 TYPICAL ADDJUSTABLE SUPPORT LEG t

O. i a. FIGURE 6 CENTER FIXED SUPPORT LEG l /16.\\ /TYP !\\ 3/16 V t,3 3 5 a /8"X Il /8"XO'-ll fg p (TYP. ) / 3/16 t.3/8"(TYR)# 'TYR 3/ M5fg_3 N 16 J.

~

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3/e" STIFF.1.' u (TYP.) PLAN 3 I l1/8" = 88 5 /16"

3 C

s e E 'O J ~ l b 1 Vl /a-2 /2 3/16 3 8 ~ 4'/8"O HOL E l THRU R (TYP.) g /8"X 12"X l'-0" 3 t

I FIGURE 7-l FLOOR BRIDGE PLATE ^ i V \\' TY'p .N 3/4 j~~/t./"xt3"xt.'m" K //6V { x, l r I I i 1: I l 1 j 1 1 l 1 l M 'I i/ /. { x / I g / s l1 t /- (" h r s x / ,i e i - f. / "x 5'"x 0 'lo %.. _ 3 bh6V-l /0 m ~+ U l /3 m. __. - PLAM G */s / ' 0 /' 2 '2 '/4 ft.l"x3 / / ~ /,' 3 e \\. /s yyglN. iI 16 l t_ _

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l 3 3 s = = N /fe N W a / 3,, V =j 1.94N 5 /5xl 0!" 0

4. h."x s h"x /?o" ELEVA 7/ OA/

__.m._ _ _. _._ _ _ _ _ _ _- _. _ _ _ _ _ _ _ _ _ _ _.. _ _ _. _

FIGURE 8 REFINED 3-D SINGLE RACK 8'T h-rvz / s. 1 f4 'Eh ch $*~ !?Q;El 'I I f/ Ir' / s 'O B'a' \\ '85 'gr dl/ / N4 I \\ / 78 \\ m I 7 " ' ' ' - G?. &k, SI rit % Eo I N j j yp 7 / N [ w'3g-(Y () ', M -Q s'. 67 I \\ \\ Y65 I I Go \\ St I / P \\ [' [f SEE FIGURE 9 FOR dr i g DETAILS. r a ~ g N S $I ~ 5A? 3s-Gh,5 ' t 9 s "f(g ,e a s u y D y' 3 ' ' ' Et' S 43 /26, g / 44 S I 2, I U an M l I M t t h 4, 't I % dd 36 2 I l '~ 3 ~ ~ ~ i, ar / - N,i - + n u ^/; x y .. 3 'l i Y E, f l 1 .b

(, a I[:.,1. J. :. 'd FIGURE 9 , RACK TO WALL IMPACT-SPRINGS 4 . Ith 7% m y L/L,5 o Tnf. 11 s p17 27C iS ~

• ro 0

k Q$ V Q Q l 3. 4 e s$ W{s 4 A., i - F 4+r = ED. r ., s ,ss, n H4 H wr wo me e_ pwi w u. m sms n> G. 6 nz Fw--i rut nH re-p ~ n ir* @ un' lh %)o 19 7 qKhf, "h b s .-hi in as g "A h se, iv ,, G v in no ei ez i9s 4,WI b_,c ! HWh A 1 _t 1 1 1 1 b kf ?s .M Y M7 7" T M7 g,9 i 4 _.-.m .th ________________._.__m._.___

h.,r- _ x _ _ _ _ . a. R. y d.t/ Nh d,. : ): dIDGUH I0 C C. M C O. I w e ) tP Q a = ~ m N w q. g ~ a n o. 3 a ~O 1r s _% O 0 ~ m n w O U O. C Z n 3C *"*' n a m 3 q ~~ O 3 0 0 0 N' O hu ~ y -mm:"" p nG o .b y 1 A

====5" 4 M {: 5J 8 \\. ..N i. ~ I o O a 06 0 04 0 02 0 00 0 02.p-04 0-06 0 e G NI NOITARELECCA DC 3 O

r I

OC i Z g -

1.w.2.m W;. [* y L <' SF dIDOH3 II O ' O. O ~ 3d. - -~r i M M c On -^I g( i N m "a-. g; m. 9 1-* O O n = _ - C =

• O

@n = ht n n a ~m o.z ~

?^

~ m m n Oo m 2 n: .N - O n(m i x t-- C s -b, M b = i' n M l o 1 l 0 -N l l ll 1 3c l. i se o g 1 06 0 04 0 02 0 00 0 02 0-04 0-06 0 e Z G NI NOITARELECCA 30 2 0 1 n I im ) 1

i Q O. 1 eno a, a ,.N O .a h _. - j,. N h Q M -G Q o. M -- =_c i = l O m g Q g z ---._n= -a. ^- m 3 .--e o .O, H r-m ..1 a A __t"__ = - = _ C -N M -_.2 ~4 C O. 06 0 0 4..0 0 2..0 00 0 02.p-04.p-06 0 c G NI NOITARELECCA

(( FIGURE 13 e r. LASALLE UAl/T-2 4 RACK SC/SH/C N0DCL N-S DIRECT /06 STANDSRD TU[L VALUCS USCO 70R HVDRO DVMRM/C MR$$ -\\ 2 ff S" / *CRP Q RP Q RP 4.g i ~ h' h#' EA \\ =.--- -- e - -e /'/////// ///////i /////// /////// n n j M / 7r / l$ E / f y / 1 4_ c' / L. i c- / 7 ar y ag e3 nrq 240 CELL 240 CflL 240l*[LL 2/7 C[LL V rutL LORD full LCAO 'It full fMPTY LCAD 4 l i i J t i i a i W L/ WW N/W W \\_/ W W L'W W i /3 -' CClL5 .~ /2 // C[lL S 12 ' C{ttS ~ c(LL S ~ 2D MULTI-RACK MODEL t l i )

FIGURE 14'- 3 l 1 S QWm] "lI =% ~ m Y l u @W9 m. ~2% 5 4 @-se a t i.j :I $ u- -.[

• -* Dj

( ~ te e e - =@ s 3 a 3 m l 7 .) A f. s u G @, E :: @ n No N <E ~a _'G G"5 3 a l 8 d J '-* m __r-M 1 U o-e. J / ~ M,. in @8 E E _ @ s e s w- ? J gg E5 -I D d d D b. u u 5.$ @ E'G M s g _r k U4

• ~*o-a 8

A 8. l ( D$ mag.x= l E E 8 8 8 (Mm.

• • IQ sms.mssm 9

e .x_-__

m FIGURE 15- / 4 / / 4-we dal Luac rm) E i -,/ y 4-r,1 4 g, e- / 4~ / A A A A -Abr -A$F b -b$ Note : / oade opp /d J ab al/ nedes. Spafa/dl.s&;suG consistad Ai; M 6/e 6 a-s/ako e/ and rofafda/ actelaraks of the mastic nof from 3D dynam)c aal sis y 2-D St(ttc. Mode.l For Loca.R hyt k gow,e,

f _' Mb.':y _ . FIGURE 16 REINFORCEMENT PLATE DETAILS o [.- f/ o l~ I ' 06k (TT f'. ) 3 l l l \\ l hlit"< l 0 "x 7 '- ) " - C AloTE )) ) N il it n n gl g jDJuh007wtw (ppg,) % O S TR M fuu YA V \\-ran arsvoeu<L ae-s PLAd 0F TfPICAL 4.cA C0e.Wse. ree:JFagosMcWT DETAIL f-Toe op exa ? l ' T4 P P \\'/ y l/4 3 !$E b/P& / '4 TNs.4.. m .m m .m m m I/ I I I I I rt '/4"3 l l w w y w w m w ~ ,g & e yf ,wg TYP. - T'r 6

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1. 0'"f*W W S

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lz :.!-:. ~..P [ i% FIGURE 17- .o r: FUEL POOL STRUCTURE .O 1, L-E L. S 4 3 '- 6" ~ u 1 EL. 8 2 0'- 6"

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