ML20205P438

Safety Evaluation Supporting Amend 2 to License NPF-76

ML20205P438
Person / Time
Site:	South Texas
Issue date:	11/01/1988
From:	Office of Nuclear Reactor Regulation
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ML20205P435	List: ML20205P431 ML20205P438 ML20205P454
References
NUDOCS 8811080131
Download: ML20205P438 (32)

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a, NUCLEAR REGULATORY COMMISSION 3

i WASHING TON, D. C. 20H4 9*...*e SAFETY EVALUATION BY THE Off!CE OF NUCLEAR REACTOR _ REGULATION SUPPORTING AMENDHENT NO. 2 TO

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TABLE OF CONTENTS PAGE 1.

INTRODUCTION 1

1.1 Licensee Submittal and Staff Review 1

1.2 Sumary Description of Reracking 2

2.

CRITICALITY CONSIDERATIONS 2

2.1 Criticality Analysis 2

2.2 Technical Specification Changes 5

2.3 Conclusions 5

3.

MATERIAL COMPATIBILITY AND CPEtt! CAL STABILITY 6

l 3.1 Discussion 6

3.F Evaluation 6

3.3 Conclusions 8

4 STRUCTURAL DESIGt:

9

.1 Fuel Handling Euilding and Spent Fuel Pool 9

4.2 Rack Analysis and Design 10 4.3 Fuel Handling Accident Consideration 11 4.4 Conclusions 12 5.

SPENT FUEL PCOL COOLING AhD LOAD hat:DLING 12 5.1 Decay Heat Generation Pete 12 5.2 Spent Fuel Fool Cooling Systen 13 5.3 Loss of Cooling 14 5.4 Evilding Ventilation 15 5.5 Heavy Load Handlins 15 5.6 Conclusions 15 6.

SPENT FUEL POOL CLEANUP SYSTEM IE 7.

AADIATION PROTECTION AND ALARA CONSIDERATIONS 16 8.

ACCIDENT ANALYSES 17 l

9.

RAD 10 ACTIVE WASTE TREATMENT 17 l

10.

SLH1'ARY OF STAFF EVALUAT10h 18 l

11. ENVIR0kttENTAL CONSIDERATIONS 18

12. C0hCLUS10NS 16 i

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UNITED 8TATES

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NUCLEAR REGULATORY COMMIS$10N 5

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SAFETY EVALUATION BY iHE OFFICE OF NUCLEAR REACTOR REGULATION St!MORTING AMENDMENT NO. 2 TO FACILITY OfERAT,1NG,t.! CENSE,NO,,NPF,-76 HOUSTON l1GHTING & p0WER COMPANY CITY PUBLIC SER',1CE BOAR 0 0F SAN ANTONIO CENTRAL POWER AND LIGHT COMPANY CITYOFAUSTIN,,,TQAS DOCKET NO. 50-498 SOUTH TE(AS,f R,00,EC,T,,,0N,IT,1 1.

INTRODUCTION l'1 L i C ' D 5 " 5.ulm,1,t,t,a,1,,ap,d, 5,t,a,f,f,,R,ey,1,ey This report presents the NRC staff safety evaluation for the reracking of the spent fuel pool at the South Texas project. Unit No. 1.

By letter dated March 8, 1988 as supplerented March 26, 1988 Houston Lightirg & Power Corpany (HL&P, the licensee) submitted an application to increase the storage capacity of the spent fuel pool, including the appropriate and necessary changes to the Technical Specifications. The licensee requested the increast, in storage capacity because the spent fuel pool contained only 196 total storage cells.

That storage capacity was adequate only for Unit 1 initial fueling and testing, and the initial part of fuel cycle 1, which is currently in progress.

The March 8,1988 request for the atendment was neticed in the Federal Rejister on June 23,1988 (53 FR 23707) as a Consideration of Issuance of Amendrint tF Facility Operating License and Opportunity for Hearing. The notice was supple-cented on Septerber 14, 1988 (53 FR 35570).

The application is based on the licensee's *High Density Spent Fuel Racks Safety Analysis Report

• which was subnitted as an enclosure to the March 8, 1988 application. During its review, the staff requested additional inforration from the licensee. The additional information was provided by letters dated August 9, 10, 19, 30, and Septerter 21, 22, and 29,1988.

This report was prepared by the staff of the Office of Nuclear Reactor Regulation.

The principal contributors to this report are:

H. Ashar Structural and Geosciences Eranch L. Kopp Reactor Systers Eranch J. Martin Radiation Protection French

2-W. LeFave plant Systens Branch J. Wing Cherical Enegineering Branch G. Dick Project Directorate IV 1.2 Sureary Dese,rjp,tjon,o,f R,erac, king The amendrent would authorize the licensee to increase the spent fuel pool storage capacity fron 196 to 1969 fuel asserblies.

The proposed expansion is to be achieved by reracking the s New, high-density storage racks (pent fuel pool into two discrete regions.

free-standing) will be used. The existing storage rocks (free-standing) will be reroved, packaged and stored on-site.

The spent fuel pool is a stainless-steel lined reinforced ccncrete pool and is an integral part of the fuel Handling Building (FHB). The pool walls are 5 feet 3 inches to 7, feet, 9 inches thick and the basemat is 6 feet, O inches thick. The walls and floor are lined with a 1-inch thick stainless steel liner to ensure the leaktight integrity of the pool. The liner plate welds are bacled with fabricated tenbers to collect water lealage from the puol. Any leakage entering the formed channels is directed to the Liquid Waste Processirs Systcm via the FHB purp.

Region I of the spent fuel pool includes 6 modules (racks) having a total of 288 storage cells. The nominal center-to-center spacing is 10.95 inches.

All cells can be utilized for storage and each cell can accept new fuel asser-blies with enrictrents up to 4.5 weight percent U-235 or spent fuel asserblies that have not echieved burnup adequate for storage in Region 2.

Region 2 includes 14 modules (racks) havirg a total of 1(81 storage cells. The nor.inal center-to-center spacing is 9.15 inches. All cells can be utilized for storage and each cell can accept spent fuel asserblies with various initial enrichments that have achi*sved n.inirvn burnups. Each cell in each region is designed to accorodate a single pWR fuel asserbly, or equivalent.

The high-density spent fuel storage rack cells are fabricated fron ASTM A240 Type 304L stainless steel plates. The Regicn 1 racks are a welded honeycerb array of square boxes separated by narrow rectangular water bones. Strips of Beraflex neutron absorber are affixed on the outside face of the long sides.

Stainless steel sheets are welded over the Eoraflex sheets to hold them in a fixed position on the box.

In Region 2, the Boraflex strips are located between adjacent walls. To provide space for the Boraflex strips between the cells, a double row of ratchirg flat round raised areas are starped into the side walls of each cell. The cells are welded together at the raised areas to hold the Boraflex in place. The cells are welded to individual asserbly bases and to one arother. The final rack arrangerent is shcwn in Figure 1.

Figures 2 and 3 show the cell design for the Region 1 and Region 2 racks, respectively.

The fuel rack rodule asserbly consists of the storage cells (and integrally welded base plates) welded together and c.ounted on the support pedestals.

The pedestals (four per rack rotule) are provided with holes and passages for flow to holes in the storage cell botter, plates.

Figure 4 illustrates

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,' support arrangerents. The tops of the support plates are welded to the fuel cell base plates. The leveling screws transmit the loads to the pool floor embedments, provide a sliding contact and provide for the leveling adjustrent of the rack.

The new racks are not doubled-tiered and all racks will sit on the spent fuel pool floor.

The proposed expansion of the spent fuel pool storege capacity to 1969 fuel assemblies will provide adequate storage until the year 2020, while maintaining full core offload capability.

In addition, it is expected that the expansion will be adequate untti a federal repository is available for spent fuel.

The proposed request is for the storage of a single fuel assently in each storage locatiun of the high density racks. However, raost of the analyses have been perfortied with the cunsolidated fuel weight in the storage locations. For the sake of analysis, the conservative assuraptions have been Nde to simulate gaps and spring constants. The staff finds the approach acceptable for evalua-ting the proposed reracking.

However, this safety evaluation approves NLt.P's request, that is the storage of non-consolidated fuel.

2.

CRITICALITY C0h51 DERAT 10h5 2.1 C,r,1,t i ca,1,i,ty Ana ly si s 2.1.1 C a l cule t t on, f,e,t,h,od,s The calculaticn of the effective rultiplication factor, K rakes use of the PDQ-7 two-dirensional four-grcup diffusion theory computef [c,de with neutron f

cross secticns generated by the LEOPARD code. These codes were benchmarked against a series of critical experirents with characteristics similar to the South Texas spent fuel pool racks. These corparisons resulted in a rodel bias of + 0.0067 ar.d a 95/95 prcbability/ confidence uncertainty of e 0.0027 for the Region I racks and a ocdel bias of + 0.0057 and a 95/95 uncertainty of e 0.0086 for the Region 2 racks.

In order to calculate the criterion for acceptable burnup for storage in Region 2, calculations were made for fuel of several different initial enrichnents and, at each enrichment, a limiting reactivity value, which included an addi-tional factor for uncertainty in the burnup analysis, was established. Burnup values

• which yielded the limiting reactivity values were then deterriined for each enrichrent from which the acceptable burnup dorain for storage in Region 2 as shown in proposed technical specification Figure 5.6 1 (Figure 5 of this SE),wasobtained. The staff finds this procedure acceptable.

2.1 ? Treatrent of,y,n,c,e,r,t,a,inties A correction for the reactivity effect of pool ter'perature is included as well as a geoe:etric sodeling effect bias to account for resh spacing and sreared stainless steel-water corposition effects.

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4 For the Region 1 analysis, the total uncertainty is the statistical contination i

of the calculational uncertainty and rancfacturing and inechanical uncertainties j

due to variations in Boraflex thickness, inner stainless steel storage box dimension, stainless steel thickness, and fuel enrichinent and density.

i In the Region 2 analysis, the sare uncertainties are considered.

In addition, an uncertainty due to the burnup analysis is estinated and combined statistically j

with the other uncertainties.

7 The staff concludes that the appropriate uncertainties have been considered and l

have been calculated in an acceptable ranner.

In addition, these uncertainties I

were deterrained at least at a 95% probability 951 confidence level, thereby setting the NRC requirerents, end are acceptable.

2.1.3 Results of Analy, sis For Region 1, the rack rultiplication factor is calculated to be 0.9250, including uncertainties at the 95/95 probability / confidence level, when fuel having an enrichment of 4.5 weight percent U-235 is stored therein. Although the pool is norr. ally flooded with water borated to 2500 por, unberated water was assured in the analysis.

For Region 2, the rack rultiplication factor is calculated to be 0.9478 for the most reactive irradiated fuel pernitted to be stored in the racks, i.e., fuel with the f.:inir.on burnup perraitted for each initial enrichnent as shchn in Figure 5.

The design will accept fuel of 4.5 weight percent U-235 initial enrichnent burned to 40.0 VWD/kgu. The analysis cf the Region 2 racks also assured full flooding by unborated water.

Therefore, the results of the criticality analyses reet the staff's acceptance criterion of K no greater than 0.95 including all uncertainties at the 95/95 probability /cebdence level, i

Host abnorral storage conditions will not result in an increase in the K of the racks.

For exarple, loss of a ccoling system will result in a decreb in the K,ff value since reactivity decreases with decreasing water density.

It is possible to postulate events, such as an inadvertent siisplacement of a fresh fuel assembly either into a Region 2 storage cell or outside and adjacent to a rack todule, which could lead to an increase in pool reactivity. Hcwever, for such events credit ray be taken for the Technical Specification requirerent of at least 2500 ppm of boron in the refueling canal during refueling operations.

The reduction in the K value caused by the boron :: ore than offsets the reactivity addition cabd by credible accidents.

j, The staff considered the possibility cf irradiation induced axial shrinkage of the Forafle; patiels as decurented in NRC Inforration Notice No. 87-43. Eased l

on this, the licensee has perforced analyses te determine the reactivity l

effects of potential Boraflex shrinkage on the South Texas spent fuel pool.

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Several scenarios were evaluated ranging frorn shrinkage of the top and bottom of each Boraflex panel with a corresponding exposure of active fuel at each end, to a single tear in each panel at the active fuel mid-pla..e.

The results indicate that sufficient inargin is available in both the Region 1 and Region 2 rack design to accorrmodate at least 8 inches of shrinkage at each end.

In addition, for Region 1, a mid-plane gap of up to 4.5 inches in every panel would not prevent the rnaintenance of a k less than 0.95.

For Region 2 mid-plane gaps in every panel of up to 3Nfinches could be accorcodated.

If the mid-plane gaps are assurred to occur in only two of the four panels in each Region 2 cell, gaps as large as 10 inches could be accortnodated without pre-venting the traintenance of a k less than 0.95. Therefore, although it is notlikelythatsignificantga$formationwilloccurintheBoraflexpanels, f

the staff believes that there would be sufficient tt.Te to detect such ancmalies and provide appropriate actions before any significant adverse reactivity effects occur.

2.2 T e c h n i c a,1, Sp,e,c,1,f,1,c a,t,i e n,,C,h,a,ng,e s The following Technical Specification (TS) changes have been reade as a result of the proposed spent fuel pool storage podifications. The staff finds thesc changes acceptable.

1.

TS 5.6.1.1 and 5.6.1.2 are ctrbined into one Specification (5.6.1).

The new TS 5.6.1 correctly accounts for the uncertainties and tolerances assured in the criticality analyses as well as the nominal center-to-center fuel asserbly spacing, the maxirun allowable l'-235 enrichtent, and the installation of Borafley between spent fuel asserblies.

2.

Figure 5.61 has been added to specify the initial enrichrent vs. burnup requirerents to be ret prior to storage of fuel asstrblies into Region 2.

3.

The spent fuel pool storage capacity has been ircreased fron 195 to 1969 fuel asserblies in TS 5.6.3.

The TS charses are effective as of the date of this Safety Evaluation.

2.3 Conclusion Based on the review described above, the staff finds that the criticality aspects of the design of the South Texas Unit I spent fuel racks are acceptable and reet the requirements of General Design Criterion (GDC) 62 for the prevention of criticality in fuel storage and handling.

The staff concludes that fuel fron Unit I riey be safely stored in Region 1 provided that the enrichrent does not exceed 4.5 weight percent U-235. Any of these fuel asserblies ray also be stored in Region 2 provided they meet the burnup and enrichrent limits specified in Figure 5.6-1 of the South Texas Technical Specifications.

-s.

3.0 MATERIAL COMPATABILITY AND CHEMICAL STABILITY 3.1 Discussion The staff has reviewed the compatability and chemical stability of the materials (encept the fuel assemblies) wetted by the pool water, in accordance with Section 9.1.2 of the Standard Review Plan (NUREG-0800 July 1981). The STP-1 pool contains oxygen-satu ated demineralized water which has 2500 parts per million of boron as boric acid.. The pool is lined with stainless steel and has two adjacent regions of storage. The principal construction eaterials for the proposed new racks in the spent fuel storage pools are ASTM A-240 Type 304L austenitic stainless steel for structure and Beraflex for neutron absorption.

The racks are welded horeycosb arrays of square stain 1tss steel boxes forming individual cells for fuel storage. Each of the four sides of a given storage cell has a Boraflex asserbly, except those sides that are nearest to the storage pool walls.

I In Region 1, the Boraflex assesbly consist, of a thin rectangular stainless steel water box with a Ecraflex sheet affixed on one side of the box and another Coraflex sheet on the other. A thin stainless steel plate is welded over each of the two Foraflex sheets on the water box. The entire assesbly is rerovable frori the storage cell.

In Region 2, a Beraflex sheet is positioned bctween two adjacent walls of the square storage cells. The Boraflex Sheets in Region 2 are not rerovable.

In both regions, single sheets of Boraflex are used, and the Boraflex sheets are not mechanically fastened to any surface or structure.

The licensee proposed an inservice surveillarce program for the Eoraflex raterial, using serple coupons that are a4de of the sare e4terial corpost-tion, fabricated by the same rethod, certified to the same criteria, cut to the sare physical direnstens, and encased in the same r.aterial as the reroyable i'oraflex asstrbites. A s,inisum of one such coupon and a string cf foot-lor:g sarples of the same raterial will be prov;ded for the storage pool in L' nit 1.

Evaluaticn of the coupon perfor:4nce will irclude visual inspectio. and seasure-sents of the neutron attenuation, hardness, and physical dirensions.

Initial surveillance of the specissens will be perforsed af ter five years of exposure to the storage pool environr+nt.

Based on the results of the initial surveillance, the Itcensee will determine the schedule and extent of additional surveillance.

The licensee, however, has provided no corrective actions to take if degradation af k raflex asserblies is found, but will evaluate available plant data ex parfor:4nce from the nuclear industry to redify the surveillauce 6

<en warranted ar.d justified.

3..,

Jation The stainless steel in the storage nool liners und rack asserblies is compatible with the oxygen saturated borated water and radiation envirortent of the spent fuel pool.

in this envirorcent, corrosion of Type 304L stainless steel is r.ot i

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expected to exceed a rate of 6 x 10 inch per year. This corrosion rate is negligible for e.'en the thinnest stainless steel walls of 0.03 inch in the rack assemblies. Contact corrosion or galvanic attack between the stainless steel in the pool liners or rack assemblies and the Inconel/Zircaloy in the fuel asserblies to be stored will not be significant, because all these raterials are protected by passivating oxide films.

Bordflex is conposed of non-retallic r,aterials and, therefore, will not develop a galvanic potential with the retal coreponents.

Space is available to allow escape of any gas which may be generated from the polymer binders in the Boraflex during heating and irradiation, thus preventing pcssible bulging or swelling cf the Boraflex asserblies. Soraflex, an elastor,er of rethylated polysiloxane filled with boron carbide powder, is used as a neutron absorber (poison) in the spent fuel storage facilities of trary ruclear power plants.

It has undergone u tensive testing to determine the effects of garra irradiation in various environrents and to verify its structural integrity and suitability as a neutron absorbing raterial. The evaluation tests have shown that Boraflex is unaffected by the pool water environrent and will net be degraded by corrosion. TesthwereperformedattheUniversityof Michigan, exposing Borafle> to 1.03 x 10 rads of ganta radiation with sub-stantial concurrent neutron flux in berated water. These tests indicated that Boraflex naintains its neutron attenuation capabilities af ter being subjected l

l to an environrent of borated water and gar.ra irradiation.

Irradiation will cause sorte loss of flexibility and shrinkage of the Boraflex.

l Long-term borated water scab tests at high tesperatures were also conducted.

The tests show that Ecrafle> withstood a berated water irrersion at 240'T for 251 days without visible distortion or softening.

The Boraflex shcwed no evidence of swelling or loss of aliility to paintain a uniform distribution of bcren carbide. The spent fuel pool water terperature under norral operating i

conditicns will be approximately 105'T which is well below the 240'T test terperature.

In general, the rate of a chendcal reaction decreases openentiall,,

with decru. sing terperature. Therefore, the staff does not anticipate any significant deterioration of the Boraflex at the nors.a1 peol operating ccoditions over the design life of the spent fuel racks.

The tests have shown that neither irradiation, environrent, nor Boraflex cosposition have a discernible effect on the neutron transsiission of the Boraflex material. The tests also have shown that Boraflex does not possess leachable halogens that might be released into the pool environment in the presence of radiation. Similar conclusions are reached regarding the leaching of eierental boron from the Boraflex.

Boron carbide of the grade nore, ally l

present in the Boraflex will typically contain 0.1 weight percent of soluble bcron. The test results have confirmed the encapsulation capability of the J

silicone polyrer matrix to prevent the leaching of soluble species from the boron carbide, i

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Recently, ancralies ranging f rom minor physical changes in color, size, hardness, and brittleness to gap formation of up to four inches wide were observed in Boraflex panels that have been used in three nuclear power plants. The exact cechanisms that caused the observed physical degradations of Boraflex hav? oct been confirmed.

But the staff can postulate that garna radiation from the spent fuel initially induced crosslinking of the polymer in Boraflex, producing Shrinkage of the Boraflex naterial. When crosslinking became saturated, scissioning (a process in which bonds between atons are broken) of the polyrer predominated as the accurtlated radiation dose increased.

Scissioning produced porosity which allowed the spent fuel pool water to perr.eate the Boraficx raterial. $cissioning and water perreation could embrittle the Eoraflex raterial.

In short, garca radiation fron spent fuel is the most probable cause of the obser,ed physical degradations, such as changes in color, size, hardness, i

and brittleness. The stoff does not have sufficient inforoation to deterrire conclusively what caused the gap ferration in sore Coraflex parels. Powever, it is conceivable the.t if the two ends of a full. length Boraflex panel are physically restrained, then shrinkage caused by garr.a radiationi can break up the panel and lead to gap forcation.

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The staff determined that ressor,able assurance exists that physical restraints are absent in the Boraflex pas.els of the South Texas project, because the Boraflex sheets are oct rechar.ically fastened to any structure.

It is not likely that gaps will fore: to any significant extent in the Ecraflex panels during the projected life of the Boraflex asstrblies. However, niner physical degradatiens can take place in the Boraflex fron irtadiation.

i In the unlibely event of gap forration in the Coraflex parels that would lead I

to loss of neutren absorbing capatility, the ronitoring prograti will detect I

i such degraded Boraflex parels, and the licensee would have sufficient tire to l

replace theri in Region I cf the storage poolf.

In Region 2 where the Buraflex l

l sheets are t.ot renovable, the licensee can either place new Boraflex stetts in the offected empty storage cells or restrict the use of the affected cells fer i

fuel storage if degraded Boreflex is fcund, 3.3

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t Based on the above discussion, the staff concludes that the cerrosion of the

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l trent fuel pool corponents due to the pool environrent will be of little i

significance during the life of the facility. Corponents in the spent fuel l

1 storage pools are constructed of alloys which have a 1cw differential galvanic potential between them and have a high resistance to general corrosion, locali:ed i

corrosion, and galvanic corrosion.

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The staff further concludes that the environtental ccrpatibility cf the materials used in the spent fuel storage pools is adequate based on the test data cited in Section 3.2 and actual service experience at operating reactor facilities, i

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The staff has reviewed the proposed surycillance program for nionitoring the Boraflex in the spent fuel storage pool and concludes that the progran can reveal deterioration that might lead to loss of neutron absorbing capability during the life of the spent fuel racks. However, if a significant loss of neutron absorbing capability is found in any Boraflex panel, the licensee should take corrective actions such as replacement of the degraded Boraflex panel, insertion of a new Boraflex sheet in the affected storage cell, or restriction of use of the affected cell for fuel storage.

The staff finds that the proposed monitoring program and the selitction of appropriate materials of construction by the licensee teet the requirements of 10 CFp,Part 50. Appendix A. GDC 61 regarding the capability to permit appropriate oc,todic inspection and testing of conponents, and GDC 62 regardirg preventior of criticality by the use of boren poison and by r,atntainisig structural integrity of corponents, and are, therefore, acceptable.

4 STRUCTURAL DESIGN 2

s For this portion of the review, the prinary focus was assuring the structural integrity of the fuel, the fuel cells rack rredules, and the spent fuel pcol floorandwallscrderthepostulatedloads(Appendix 0ofSRP3.8.4),andfuel handling accidents. The najor areas of concern end their resolution are discussed in the iollowing paragraphs.

1 41 F,u,e,1,H,ap,d,1,ips,,E u,1,1,d,ips, a p,d, 5 p,ep,t, f u e 1 P o o l The fuel Handling tuildir4 analysis end design was reviewed and accepted during the Operating License stage. FL&P, however, perforred the seistic arialysis of the fuel hardling tuilding incorporating the revised rass of the proposed reracking (tensolidated fuel). The soil structure interaction analysis and the irput rotion were consittred in the sare way as in the original aralysis. A ccrrerative review of the output tables indicated less than 57 differences in pedal frequer.cies. / 150, a corparison of acceleration response spectra indicated negligibic differences in spectral accelerations at the spent fuel pool floor level. The analysis thus :enfirred the validity of the basic input criteria for the seismic analysis of the high density racks. HL&P also retalculated tts differences in soil bearing pressure and the facter of sefety against liquifica-j tion due to the added mass. The soil bearing pressure increased from 20.2 ksf to 22 ksf against the allewable of 32 ksf under dead load and Safe Shutdown 1

Earthquake ($$E) ccmbination. The minirum safety factor of 1.4 against liquifi-

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cation retained unaffected.

HL&P 4'150 perforred a detailed seismic analysit ci spent fuel pool areas affected by the proposed reracking. HL&p deranstrated that the pintrum safety factors at various critical sections of the pool walls and floor slab were higher than 1.0 for all conditions of loading considering the consolidated fuel l

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10-weight. Hcwever, the design margin for transverse shear in tt< 7 pent fuel peol floor slab was trarginally above 1.0 for standard fuel.

For consclidated fuel, HL&P derronstrated a similar margin when the confirmatory basii response spectra described in the FSAR Section 3.7.2.4 was used. This evaluaticn pertains to the use of standard (single fuel asserbly per storage location) fuel, for which the staff considers the design to be adequate.

The staff had expressed a concern regarding the ittegrity of pool floor liner plate before the rack pedestals start sliding under the postulated earthquale loading. The license perforr.ed a rigorous analysis of the liner-cencrete interface and its anchorage at the erbedded plates and leak chases. With conservatively estimated sheer load, and considering a very low coefficient of friction of 0.15 between the liner and the concrete, the licensee der.onstrated that r.argins of safety in excess of 1.0 exist for stresses in liner plate, crbedded plates, anchor studs, anchor bolts and connecting M cs. The staff finds the licensee's conclusior,recardir.g the integrity of liner plate ut. der postulated loadings to be acceptable.

8 a c,k,,A.a,1y,s,1,s, a,nd, D e s i o n 4.2 f.

Tables 1, 2 and 3 provide the rack redu'e data, dimensions and pertinent t'cdeling parar,eters. IIL&P's analysis is based on one set of synthetic tirt-histories. The tietf expressed concern regarding the adequacy of energy content at the frequencies of interest, when used for nco-linear rack analysis.

NLLP generated tre Pcwer Spcctral Censity (PSD) functions for the floos input I

riotier, used in the racn aralysis ard ccrpared ther, with the target FSD cbtained I

by the r ethod in fWREG/CR-3509, "Power Spactral Otosity functions Coppatible with Regulatory Guide 1.60. Fesponse Spectra." A tytical corparison is shewn in Figure 6.

In gereral, and particularly in the low frequency range of interest, the corputed FSO catteded the target P3D by a good r.argin. I dip at 6.3 Fr was indicative of the characteristics of the design response spectra.

At frequencies higher then 12 Hz, the dips are e>pected when pSD5 for in-ftructure r>otia are cerpared to the target pSD for ground rotion. On an overall besis, the ccrperisen irdicated attquate energy ccntent for time. histories beitig used for the rack aralysis.

Requirerents for seisric and ispect loads are discussed in Section 3 of Appendix 0 of SRP Section 3.8.4 There it is stated that seiseic escitaticn along three orthogonal directier.s should be irposed sir,ultaneously for the design of the new rack system.

HL&F's origir,a1 rac6 ardlyses were based on the square root of the sum of the squares (SR55) ccabination for the rack resporses due to the three cosponents of the earthquake to be considered.

The rack responses (displacerent, forces) were separate?y calculated for two horizontal directicns using a proprietary corputer code *RACF0E". The respor.ses due to the vertical ccrporent were calculated using an equivalent static sethod 1

, with a dynamic load factor of 1.5.

This procedure appeared to provide bcuoding calculations for forces at the pedestal, but the staff expressed cont.ern regarding the ability of the procedure to provide a realistic assessment of the displacerents under the three cor ponents of an earthquake.

"RACK 0E" as used by HL&P is a two-dirensional code, capable of performing two-dirensional dynsmc analysis with sir.ultaneous seismic input in two directions. HL&P perforced eulti-rack analyses of 3-racks in a rew (E-W direction) with varying leading conditions for each rack with simultaneous application of E-W and vertical tire

(

histories. The rulti-rack r.ot.'el is shown in Figure 6.

The gar used between the racks in these analyses was 1.0 inch. The results indicated that the raxirut relative displacemer.t between the two adjacent racks was 0.73 inches l

l and 0.41 inches for the coefficients of friction of 0.2 and 0.8 respectively.

None of the cases showed rack-to-rack or reck-t dwall interactions. The cross ccupling effects were ignored in these alculations resulting in calculated l

displactments that are larger than the actual displacer,ents.

Eased on the conservatively corputed maxitoum loadirgs, fur load corbinations stresses in varicus l

I criticel components of the rack r'odules were ccrputed I

recorrended in Table 1 of Appendix 0 of SRP 3.8.4 The stresses were corpared against tre requirenents of Subsection t;F of ASME Code and rinirnur. saf ety facters as ratios of the allcwable divided by the actust stresses were corputed.

Table 4 is a surrary of the st.fety factors at critical rack locations.

The staff was concerned that after a seisric ever,t the racks could reve, creating gaps between the racks different fron the 1 inch initial spacing such that the ccnfiguration of the racks would no lorger be bounded by the seismic analysis.

In resconse, the lictrsec corttitted tu rodify the plar,t procedures to inclutt a requirtee,t to perferr.,a walldewn of the spent fuel pool to checl the ract configuration af ter a seisteit event (i.e., Operating Easis Earthquale or greater).

Based on the results of the HL&P's seisriic analysis, the staff concludes that during an SSE the fuel racks sill raintain treir structural integrity, fuel asserblies will rct sustain doage, and rack displacerents will r et be large trough tc result in racl tu-rack or rack-te wall irpact.

4.3 F u e 1 P a n d 1 i 03, A c,c,1,d e n,t,,C,o,n s 1,d,e,r,a,t,i o,n FL&P perforred structural analyses and evaluations of fcur postulated fuel handling accidents:

A.

Dropped Fuel Accident 1 A censolidated fuel can uter was assured to have drcpped from 14 inches above the tcp of a rack redule ard directly irpacted the bottom plate of a fuel cell. The final velocity and tocal energy were considered assunirs j

no energy dissipation in the canister.

Based on this consideration the botton plate weld could fail, thus allowing the botton plate and the fuel canister to itpact the ptol liner. Pcwever FL&P deronstrated that the

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1

• . impact energy would not perforate tie liner. HL&P used Ballistic Research Laboratory (BRL) formula for predicting the liner penetration / perforation.

Considering the conservative approach used by HL&P and that the staff's evaluation concerns single fuel assemblies, the analysis results are acceptable to the staff.

B.

Dropped Fuel Accidents 2 and 3 In Accident 2 HL&P considered a drop of a consolidated fuel canister from 14 inches above the top of a rack. The top portion of the rack could sustain son.e plastic (pern.anent) deformation. However, HL&P's calculations confirmed that the safe, subcritical configuration of the stored fuel wculd tot be compromised due to such an accident. The staff accepts the HL&P's findings for the purpose of this evaluation.

Accident 3, which postulates an inclined drop of a fuel canister, would not be as severe as Accident 2, as the impact energy would be distributed over a large area of a rack module.

C.

Jcmed Fuel Assembly HL&P considered the rack stresses when 4000 lbs. of force was applied to unjam a fuel assambly in a storage location. This force was considered at any height of the fuel storage cell.

H'.dF's calculations indicated the stresses resulting from application cf auch a force to be within the acceptoble criteria. The staff finds the postulation of the force to be reasonable and the final conclusion 5 to be acceptable.

{

In any of the postulated accidents, damage to the dropped or jamed fuel assembly is possible. According to HL&P, the consequences of such damage are bounded by the desigt. basis fuel-handling accidents described iri the licensee's FSAR Section 15.7.4 4.4 Conclusions Tne steff concludes that the structural design, rack design and fuel handling accident considerations ate acceptable.

j 5.

SPENT FUEL POOL COOLING AND LOAD HANDLING The licensee's submittal was reviewed in accordance with the requiremen's c

of GDC 2, 44, and 61, and the guidelines of NUREG-0800, "Standard Review Plan" (SRP) and NUREG-0612. "Control of Hecvy Loads at Nuclear Power Plants."

5.1 Decay Heat, Gen, erat,jon_ba,te a

The licensee stated in the March 8, 1988 submittel that the calculation of the de ay heat generation rate was in accordance with the guidelines of HUREG-0800, t

l SRr Section g.1.3 and Branch Technical Position ASB 9-2.

For the nornal l

maximum heat load case the licensee assumed the pool was filled with one-third core refuelings every 12 months (nintaining a full core discharge capabilit with the final one-third core being placed in the pool at 140 hours0.00162 days <br />0.0389 hours <br />2.314815e-4 weeks <br />5.327e-5 months <br /> (Case A)y) l

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- i and at 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> (Case B) after shutdown. The two cases of 140 hours0.00162 days <br />0.0389 hours <br />2.314815e-4 weeks <br />5.327e-5 months <br /> and 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> were calculated because the South Texas plant has a fast refueling option which has the capabi

.y to offload one-third of a core in 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />. The specific recommendation in SRP Section 9.1.3 is 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> (Case C). The licensee calculated heat loads and fuel pcol temperatures (one pool cooling train and two pool cooling train operation) for both the 140 hour0.00162 days <br />0.0389 hours <br />2.314815e-4 weeks <br />5.327e-5 months <br /> and 80 hour9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> cases and for the SRP Section 9.1.3 assumptions of one-third core after 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br />, one-third core at one year, plus one-third core af ter 400 days. The maximum calculated pool temperatures with one and two trains operating are:

1 Cooling T,rai,n 2 Cooling Trains Case A 145.7'F 126.0*F Case B 150.7'F 129.2'F 1

Case C 131.2*F 118.7'F i

For the abnormal naximum heat load case (Case 0), the licensee assumed the same conditions as 10 Cases A and B except that the last one-third core offload had been in the pool for 36 days plus a full core offload 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> after shutdown.

I The recornendetions of SRP Section 9.1.3 are one-third core in the pool for 400 days, one-third for 36 days and one full core at 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> af ter shutdown. The calculated pool water temperature for Case D is 155.4*F with two pool cooling trains operating.

To verify the licensee's calculated spent fuel heat loads, the staff perforned an independent calcJ13 tion for the maxir4un abnormal storage condition of Case C using STP ASB 9-2 guidelines. The staff calculated a heat load of 58.03 MBtu/hr as corpared with 63.15 MBtu/hr.

Because the licensee's calculated valuewasbasedonconservativeassumptionsascomparedwiththestaff's(the licensee assumed last refueling was grr;ter than 1/3 core leeving no eupty storage spaces) and nct appreciably different based on the high rate of decay heat energy, the staff finds that the licensee has properly calculated the heat generation rate in accordance with the SRP.

5.2

,Spe,n,t, Fuej_,Pooj,,Cpol,ing System i

The spent fuel pool cooling system (SFPCS) consists of two seismic Category 1, i

Quality Group C cooling water trains each with one pump and one heat exchanger.

After the spent fuel pool water is cooled in the heat exchangers, it is purified by the non-seismic Category I cleanup system.

In th: event of a loss of th's SFPCS, there are several sources of pool makeup water available including a seismic Category I source from the low-head safety injection pumns.

l In its April 1986 Safety Evaluation Report (SER), NUREG-0781, for South Texas Units 1 and 2, the staff concluded that the SFPCS met the acceptance criteria of SRP Section 9.1.3 including 000 2 and was acceptable. The bases for this J

a conclusion have not changed as a result of the proposed reracking, except with regard to the requirements of GDC 44, "Cooling Water". The change in the basis for GDC 45 is due to the new decay heat loads which are higher for the increased storage capacity.

As indicated in Section 5.1, the design of the SFPCS still meets the 140*F fuel pool water temperature reconnendation of SRP Section 9.1.3 when calculating the maximum normal heat load using the assumptions identified in the SRP.

Under the higher heat load conditions identified using the licensee's more conservative assumptions for South Texas, the recommended pool temperature of 140'F for single train operation is acceptable because:

a.

The assumptions used in the calculations are more conservative than staff guidelines; b.

The SFPCS is a. safety-related system; c.

For the worst case (Case A) the 140*F could be exceeded for only 11.5 days; d.

With two trains operating, the pool temperatures for Cases A and B are well below 140'F; e.

The 140*F is a recommended limit and the likelihood of exceeding that recomendation is low given the conditions and conservatiscis assures in the calculation; and f.

The effect of pool water temperature slightly above 140*F on spent fuel storage safety is ne9,.gible.

For the abnormal maximum heat load (Csse D), the SFPCS will maintain pool water temperature at er below 155.4*F with two trains of cooling which is well below the recorrnended no boiling lirait of SRP Section 9.1.3 under these conditions.

As a result of its review, the staff finds that the SFPCS still meets the requirements of GCC 44 with respect to providing adequate pool cooling under maximum normal heat load conditions following a single failure.

5.3 1.os,s,of Cooling In the event that all SFP cooling is lost, the spent fuel pool temperature will increase until boiling occurs. The licensee has estimated the time from the loss of pool coohng until the pool boils for the four cases identified above. The times for the various cases including the boil-off rates are:

a.

Case A - 8.29 hours3.356481e-4 days <br />0.00806 hours <br />4.794974e-5 weeks <br />1.10345e-5 months <br /> at 54 gpm b.

Case B - 6.67 hours7.75463e-4 days <br />0.0186 hours <br />1.107804e-4 weeks <br />2.54935e-5 months <br /> at 62 gpm c.

Case C - 15.49 hours5.671296e-4 days <br />0.0136 hours <br />8.101852e-5 weeks <br />1.86445e-5 months <br /> at 35 gpm d.

Case D - 2.86 hours9.953704e-4 days <br />0.0239 hours <br />1.421958e-4 weeks <br />3.2723e-5 months <br /> at 135 gpm The assured (seismic Category I) pool makeup water source from the low pressure injection system has a capability in excess of the above boil-off rates. This seismic Category I makeup source is adequate to provide water for the higher boil-off rate of the expanded storage capacity and, therefore, still meets the requirements of GDC 2 "Design Bases for Protection Against Natural Phenorena."

s

• 5.4 BuildigYentilation j

The seismic Category I fuel handling building (FHB) ventilation exhaust system is designed to limit offsite doses in the event of a fuel handling accident, The staff's evaluation and conclusions regarding the consequences of a fuel handling accident identified in Section 15 of NUREG-0781 have not changed as a result of the proposed increased storage capacity because the accident analysis is based on the activity released from the last one third of a core placed in the pool. Thus, the FHB ventilation exhaust system continues to be acceptable.

5.5 H,eavy,j.gdHandling The new and old spent fuel storage racks are considered to be heavy loads and will be moved by the FHB overhead crane. The FHB overhead crane is a single-failure-proof crane which meets the guidelines of tiUREG-0554, "Single Failure-Proof Cranes for Nuclear Power Plants" and flVREG-0612 "Control of Heavy Loads at fluclear Power Plants," as indicated by the staff's acceptance in NUREG-0781.

Because the reracking will take place prior to the Unit 1 first refueling, heavy loads will not be carried over spent fuel during the reracking operation.

The rnethods and equipment used for the reracking vill be in accordance with Section 5.1.1 of itVREG-0612, which includes the identification of sate load paths for heavy loads, procedures for load handling, and operator training.

The spent fuel shipping cask cannot be carried over the spent fuel pool due to crane travel limitations, so a cask drop accident will not affect spent fuel or the spent fuel pool cooling system, as previcusly deterrnined by the staff.

Therefore, storage of spent fuel in the new proposed high density storage racks will not affect the staff's previous acceptance of the spent fuel cask drop analysis as contained in NUREG-0781. As a result of its review, the staff finds that heavy loads handling will be perforrred in accordance with the guidelines of itVREG-0612 and there'ere the requirements of GDC 61, Fuel Storage and Handling and Radioactivity Control," are ret 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 heavy load drop.

i 5.6 Conclusion Based on the above, the staff cone.ludes that the proposed expansion of the South Texas, Unit I spent fuel pool storage capacity complies with the require-ments.of General Design Criteria 2, 44 and 61, the guidelines of flVREG-0612 and applicable acceptance criteria of the Standard Review Plan with respect to the capability to provide adequate spent fuel pool cooling and the safe handling of heavy loads. The staff, therefore, concludes that the proposed spent fuel pool i

expansion is acceptable with respect to spent fuel pool cooling and load handling.

l 1

. a L

6.

SPElli FUEL POOL CLEANUP SYSTEM The spent fuel pool cleanup system at STP-1 is an integral out of the spent fuel pool cooling system. The system is designed to maintain water quality and clarity and to remove decay heat generated by the spent fuei asser";1ies in the spent fuel pool and in the temporary in-containment storage area. The chanup system is also designed to purify water in the refueling cavity and the refueling water storage t6nk.

The system includes all components and piping from inlet to exit from the spent fuel pool, in-containment storage area, refueling cavity, and piping used for fuel pool makeup, from the refueling water stocsge tank and the cicanup filters /demineralizers to the point of discharge to ths radwaste system. The spent fuel pool cooling and cleanup system consists of two maximum normal heat load full-capacity fuel pool cooling trains (each wita a pump and heat exchanger), two demineralizer purification trains, a spent fuel pool surface skimmer loop, and a reactor cavity filtratiun systen. The spent fuel pool cooling puraps can be powered from the Class IE emergency sources.

Radioactivity and irrpurity levels in the water of a spent fuel pool increase primarily during the refueling operations as a result of fission product leakage from defective fuel elements being discharged into the pool and to a lesser degree during other spent fuel handling operations. The reracking of the spent fuel pool at the South Texas Project, Unit I will not increase the refueling frequency and fraction of the core replaced af ter each fuel cycle.

Therefore, the frequency of operating the spent fuel pool cleanup system is net expected to increase.

Sinil6rly, the chemical ar.d radionuclide composition of the spent fuel pool Wdter will not Change as a result of the proposed reracking.

Following the discharge of spent fuel from the reactor into the pool, the fission product inventory in the spent fuel and in the pool water will decrease by radioactive decay.

Furthermore, experience also shows that there is no significant leakage of fission products from spent fuel stored in pools after the fuel has cooled for several months.

Thus, the increased quantity of spent fuel to be stored in the South Texas Project, Unit I fuel pool will not increase significantly the total fission product activity in the spent fuel pool water during,the operation of the pool.

7.

RADIAT10ft PROTECT 10fi AhD ALARA C0tiSIDERAT10lls In as much as the new spent fuel racks will be installed in the SFP before the pool is used for storage of spent fuel, there will be no additional occupational radiation exposure associated with the reracking of the spent fuel pool.

HL&P has considered any increases in exposure from spent fuel storage, airborne radiation, solid radioactive waste (resins, filters, and corrosion product crude)andconcludedthatnosignificantincreasesareexpected.

The staff has reviewed HL&P's analysis and finds it acceptable.

.e The radiological protection of workers during fuel iandling operations will not change because the spent fuel will remain covered by 23 feet of water, as before, and spent fuel will be covered by at least :.0 feet of water during spent fuel handling operations.

Based on the review of the HL&P's submittal, the staff concludes that the projected activities and estimated person-rem doses for this project are reasonable. HL&P intends to take ALARA considerations into account, and to implement reasonable dose-reducing activities. The staff concludes that HL&P will be able to maintain individual occupational radiation exposures within the applicable limits of 10 CFR Part 20, and maintain ALARA doses, consistent with the guidelines of Regulatory Guide 8.8.

Therefore, the proposed radiation protection aspect of the SFP rerack is acceptable.

8.

ACCIDENT ANALYSES The staff has reviewed the accident analysis that could occur at STP-1 in conjunction with the proposed reracking. The applicable accidents were cask drop, loads over the spent fuel, and spent fuel pool boiling.

The proposed changes do not affect the previcusly approved cask drop analysis.

Crane design and building arrangement prevent movement of the cask over the fuel pool and prevent interference of the cask crane bridge, trolley, and hoist with fuel racks or building structures.

The rail for the cask handling crane stcps at the edge of the cask loading pool, which is more than 25 ft. from the spent fuel pool boundary. Building arrangement, crane control, end lifting rig design restrict vertical lift of the cask to an elevation such that the cask will not be higher than 30 feet above the floor in the Fuel Handling Building.

In accordance with 10 CFR Part 71, the spent fuel shipping cask is d"signed to sustain a free-fall of 30 feet onto an unyielding surface followed by a specified puncture, fire, and imersion in water with the release of no raore then a specified small quantity of radioactivity.

The spent fuel cask crane is not capable of traveling over the spent fuel l

pool.

i in the spent fuel pool boiling accident, it was assumed that a loss of spent fuel pool cooling occurred after a refueling where 1/3 of the core had been removed and placed in the spent fuel pool. As a result of boiling it was assumed that the greatest contribution to iodine leakage was from the off loaded 1/3 core. The dose consequences at the exclusion zone boundary (0-2 hours) and low population zone boundary (0-30 days) were 0.0002 and 0.54 thyroid-rem l

respectively, The potential doses resulting from the accidents considered were well below the allowable 10 CFR Part 100 guidelines. Therefore, the accident analysis aspect of the spent fuel pool rerack is acceptable.

9.

RADI0 ACTIVE WASTE TREATMENT j

l The plant contains a radioactive waste r-anagerent system designed to provide for the controlled handling and tre.:tment of liquid, gaseous, and solid wastes.

l

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18-The radioactive waste management system was evaluated in staff Safety Evaluation Report (SER)datedApril1986(l:UREG-0781). There will be no changes in the system described in the SER because of the proposed SFP rerack.

10. SUMARY OF STAFF EVALUATI0ll The staff has reviewed and evaluated HL&P's request for the expansion of the spent fuel pool capacity. Based on the considerations discussed in this safety evaluation, the staff concludes that the analyses 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.

The approval is based on the storage of non-consolidated fuel and the installa-tion of all racks prior to storage of any spent fuel in tne spent fuel pool.

If IIL&P should change its plans and decide to store spent fuel in the pool before completing the installation of the new racks, it should subnit documen-tation to the staff for prior review addressing all significant changes from the request the staff is now approving.

It was ncted during the staff review that while the proposed surveillance program for monitoring Beraflex in the spent fuel pool was acceptable, no corrective action was proposed in the event that Eoraflex degradation was observed.

It is recorrended that a plan of corrective actions be developed and irnplemented.

11.

EilVIP0tWEi4TAL C0liS10 ERAT 101:S 1he i< arch 8,198P request for arendn'ent was noticed in the Federal Register on June 23,1988 (53 FR 23707) as a Consideration of Issuance of ArtendmenY to Facility Operating Licenne and Opportunity for Hearing.

It was suppleraented on September 14, 1988 (53 FR 35570). }lo hearing requests were received.

A separate Environmental Assessnent has been prepared pursuant to 10 CFR Part

51. The flotice of Issuance of Environtrental Assessment and Finding of io Significant Irnpact was published in the Federal Register on October 28, 1988 (53 FR 43788).

12. C0tiCLUS10ils The staff has reviewed and evaluated the licensee's request for amendtrent for the South Texas Project, Unit I regarding the expansion of the spent fuel poul.

Based on the considerations discussed in this safety evaluation, the staff concludes that:

(1) this amendment will not, (a) significantly increase the probability or consequences of accidents previously evaluated, (b) create the possibility of a new or different accident frorn any accident previously evaluated, (c) significently reduce a margin of safety; and therefore, the amendment does not involve significant hazards considerations; (E) there is reescnable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and

.. '; 9 (3) such activities will be conducted in compliance with the Comission's regulations and the issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public.

The amendment is in effect as of the date of the staff's Safety Evaluation.

Dated: November 1,1988 u

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TAB 12 1 SPENT FUEL RACK DATA Rack Module Storage Cells Array Region Number Per Module Size

• 1 1

48 8x6 1

2 46 8x6 1

3 48 8x6 1

4 48 8x6 1

5 48 8x6 1

6 48 8x6 2

7 110 10 x 11 2

8 110 10 x 11 2

9 110 10 x 11 2

10 121 11 x 11 2

11 132 12 x 11 2

12 132 12 x 11 2

13 121 11 x 11 2

14 121 11 x 11 2

15 132 12 x 11 2

16 132 12 x 11 2

17 110 11 x 10 2

18 110 11 x 10 2

19 120 12 x 10 2

20 120 12 x 10 The array size indicates the number of storage cells in the N.S

• direction a the number of cells in the E.V direction.

Note: This is the same table as Table 3.2 in Reference 1.

L4/NRC/cj

• ' s s

tASLE 2 RACK MODU 12 DIMERSIONS AND WEIGHTS Rack Estimated Dry Veight (1bs)per Module Nominal Cross Section Estimated Dry Module with Number Dimensions (inches)

Weight (1bs)

Single Density NS E.V For Module ruel 1

88 66 26.100 114.516 2

88 66 26,100 114,516 3

88 66 26,100 114,516 4

88 66 26,100 114.516 5

88 66 26.100 114.516 6

88 66 26,100 114,516 7

91 101 23.040 225,660 8

91 101 23.040 225 o40 9

91 101 23,040 225,660 10 101 101 25,220 248,102 11 110 101 27.400 270,544 12 110 101 27,220 270,364 13 101 101 25.400 248.282 14 101 101 25,400 248,282 15 110 101 27.600 270,744 16 110 101 27,420 270,564 17 101 91 23,200 225,820 18 101 91 23.200 225,820 19 110 91 25,200 246,240 20 110 91 25.040 246,080 e

L4/NRC/cj

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TABLE 3 i

RACK MODEL FARAMETERS

• Rack 11 x 10 12 x 11 8x6 Module K

(1b/in)

E.V.099 x 10'

.0871 x 10

.0527 x 10 6

6 y

NS

.099 x 10

.0873 x 10

.0788 x 10 1.77x10f 1.84xlof K

(1b/in)

E.V 7

d 1.40 x 10 N.S 1.87 x 10 1.75 x 10 1.40 x 10 h

(in) 13.12 13.12 13.12 H

(in) 201.31 201.31 201.31 W

(1b) 22938 27366 26047 W[((1b) 386936 463094 88416 L

in) 91.50 100.65 65.70 1

x L

(in) 100.65 109.80 87.60 y

Rack model parameters are for the consolidated fuel except the 8 x 6 size which only will store single density spent fuel.

Nominal gap between cell wall & fuel assembly - 0.185" Nominal gap between cell wall & fuel assembly - 0.237" Kg - Fuel assembly to. cell vall impact spring rate Kd Vertical axial spring rate for concrete, pedestal, base plate and cell deformation.

h - 14ngth of support leg l

1 H-Height of rack above base plate V

Veight of rack without fuel Vf Weight of fuel

'l L, - Mam m h W u (x-W W m-N O L

Platfora dimension (y Direction - North) y l

L4/NRC/cj

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TABLE 4

SUMMARY

OF SATITT FACTORS IN CRITICAL FUEL RACK IDCATIONS Item / Location Safe ty Comments Factor *

• Support Footing 1.97 Table 6.6*

(Pedestal) to Baseplate Veld Stress cell to Baseplate Veld 1.06 Table 6.6*

Stress Cell to Cell Veld 1.12 Thermal Plus Seismic stress Stress Due to Effects of Isolated Hot Cell.

Impact Load Between Fuel 1.20 Standard Fuel Assembly and Cell Vall Shear Load on Baseplate 1.07 Table 6.6*

Near a Support Footing Compressive Stress in 4.23 Based on Local Cell Vall Buckling Considerations (Standard Fuel)

Rack to Vall lupact Leads No Impact with Fool Valls occur at any Location Table 6.6,1.icensing submittal. 37 ML AZ 2417; see Table 6.6 for other related Safety Factors.

All safety Factors are for consolidated fuel unless otherwise noted.

LA/NRC/cj