ML20207M406

Safety Evaluation Granting Util 880308 & 26 Requests to Increase Spent Fuel Pool Capacity Through Use of High Density Storage Racks

ML20207M406
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Site:	South Texas
Issue date:	10/05/1988
From:	Office of Nuclear Reactor Regulation
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NUDOCS 8810180283
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.. 5 :E WASHINGTON, D. C. 2J M 4, . . . . . ,8 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO THE INCREASE IN THE SPENT FUEL POOL CAPACITY THROUGH THE USE OF HIGH DENSITY STORAGE RACKS HOUSTON LIGHTING & POWER COMPANY CITY PUBLIC SERVICE BOARD OF SAN ANTONIO CENTRAL POWER AND LIGHT COMPANY CITY OF AUSTIN, TEXAS DOCKET NO. 50-499 SCUTH TEXAS PROJECT, UNIT 2 i

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TABLE OF CONTENTS-PAGE

1. INTRODUCTION 1 1.1 Applicant Submittal and Staff Review 1 1.2 Summary Description of Reracking 2 P. CRITICAllTY CONSIDERATIONS 3

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2.1 Criticality Analysis 3 2.? Conclusions 5

3. MATERIAL COMPATIBILITY AND CHEMICAL STABILITY 5 3.1 Discussion 5 3.2 Evaluation 6 3.3 Conclusions 8 4 STRUCTURAL DESIGN 9 4.1 Fuel Handling Euilding and Spent Fuel Pool 9 4.2 Rack Analysis and Design 9 4.3 Fuel Handling Accident Consideration 11 4.4 Other Phases of Installation 12 4.5 Conclusions 12

5. SPENT FUEL POOL COOLING AhD LOAD HANDLING 12 l 6.1 Decay Heat Generation Rate 12 5.2 Spent Fuel Pool Cooling System 13 5.3 Loss of Cooling 14 5.4 Fuilding Ventilation 15 ,

5.5 Heavy Load Handling 15 ,

5.6 Conclusions 16

6. SPENT FUEL POOL CLEANUP SYSTEM 16

7. RADIATION PROTECTION AhD ALARA CONSIDERATIONS 17

8. ACCIDENT ANALYSES 17

9. RADIOACTIVE WASTE TREATMENT 18  !

10 ENVIRONMENTAL C0hSIDERAT10NS 18 r

11. CONCLUSIONS 19 I l

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O clou o UNITE D tiTATES 3" .

g y g NUCLEAR REGULATORY COMMISSION a ij WASHINGTON, D. C. 20555

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR PEGULATION RELATED TO THE INCREASE IN 1HE SPENT FUEL POOL CAPACITY THROUGH THE USE OF HIGH DENSITY STORAGE RACKS HOUSTON _ LIGHTING & POWER COMPANY CITY PUBLIC SERVICE BOARD Of SAN ANTONIO CENTRAL POWER AND LIGHT COMPANY CITY OF AUSH L _H MS DOCKET NO._50-499 50llTH TEXAS PROJEC1 f UNU,2

1. INTRODUCTION 1.1 Submittal and Staff Review This report presents the NRC staff safety evaluation for the reracking cf the spent fuel pool at the Scuth Texas Project. Unit No. 2. By letter 1988 as supplemented Parch 26, 1988, Houston Lighting &

dated March 8,(HL8P) submitted a request to increase the storage capacity Power Company of the spent fuel pool. HLIP re';uested to increase its storage capacity at this titre ir, order to reduce the putential for personnel exposure that would be a consideration if the reracking w6re to occur after spent fuel was stored in the pool.

The request is based on HLt<P's "High Density Spent Fuel Racks Safety Analysi, Report," which was submitted as an enclosure to the March 8, 1968 letter. During its review, the staff requested additional information from HL F.P . The additional infon.ation was provided by letters dated August 9, 10, 19, 30, and September 21, 02, 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 Branch L. Kopp Reactor Systems Branch J. Martin Radiation Protection Branch W. LeFaye Plant Systems Branch J. Wing Chemical Enegineering Branch G. Dick Project Directorate IV

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1.2 Sumary Descripgon_ of Reracking The spent fuel pool is a stainless-steel lined reinforced concrete pool and is an integral part of the Fuel Handling Building (FHR). The pool walls are 5 feet, 3 inches to 7 ft, 9 inches thick and the basemat is 6 feet, 0 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 backed with fabricated members to collect water leakage from the pool. Any leakage entering the fortned channels is directed to the Liquid Vaste Processing System via the FHB pump.

HL&P requested approval to increase the spent fuel pool storage capacity from the previously approved number of If1 to 1969 fuel assemblies.

The proposed expansion is to be achieved by reracking the spent fuel pool into two discrete regions. New, high-density storage racks (free-standing) will te used.

Region I of the spent fuel pool includes 6 rodules (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 assemblies with enrichnents up to 4.5 weight percent U-235 or spent fuel assenblies that have not achieved burnup adequate for storage in Region 2. Region 2 includes 14 modules (racks) having a total of 1681 storage cells. The nominal center-to-center spacing is 9.15 inches. All cells can be utilized for storage and each cell can accept spent fuel assenblies with various initial enrichments that have achieved minimum burnups. Each cell in each region is designed to acconcdate a single PVP fuel asscably, or equivalent.

The high-density spent fuel storage rack cells are fabricated from ASTM A240 Type 304L stainless steel plates. The Region 1 racks ere a welded honeycumb array of square boxes separated by narrow rectangular water boxes. Strips of Poraflex neutron absorber are affixed on the outside face of the long sides. Stainless steel sheets are welded over the Foraflex sheets to hold them in a fixed position on the box. In Region 7, the Boraflex strips are located between adjacent walls. The cells are welded to individual assembly bases and to one another. The final rack arrangement is shown in Figure 1. Figures ? and 3 show the cell design for the Region 1 and Region 2 racks, respectively.

The fuel rack module assembly consists-of tht storage cells (and integrally welded base plates) welded together and mounted on the support pedestals.

The pedestals (four per rack module) are provided with holes and passages for flow to holes in the storage cell bottom plates. Figure 4 illustrates 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 embedtrents, 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 spu t fuel pool floor.

The proposed expansion of the spent fuel pool storage capacity to 1969 fuel assemblies should provide adequate storage until the year 2020, wnile maintaining full core offload capability. In addition, the expansion should be adequate until a federal repository is available for spent fuel.

The proposed request is for the storage of a single fuel assembly in each storage location of the high density racks. However, most of the analyses have been perfortned with the consolidated fuel weight in the storage locations. For the sake of analysis, however, the conservative assumptions have been made to simulate gaps and spring constants. The steff finds the approach acceptable for evaluating the preposed reracking.

However, this safety evaluation approves HLlP's request, tht.t is the storage of non-consolidated fuel.

2. CPITICAL11Y CONSIDERATIONS 2.1 Criticality Analysis

?.1.1 Calculationfegods Tne calculation of the effective multiplication factor, K of the PDQ-7 two-dinensicnal four-group diffusion theory ENp utermakes code use with neutron cross sections generated by the LEOPARD code. These codes were benchmarked against a series of critical experinents with chara:ttristics sinilar to the South Texas spent fuel pool racks. These conparisons resulted in a nodel bias of + 0.0067 and a 95/95 probability /

confidence uncertainty of t 0.0027 for the Region 1 racks and a rodel bias et + 0.0057 and a 95/95 uncertainty of 0.0086 for the Region 2 racks.

In order to calculate the criterion for acceptable burnup for storage in '

Region 2, calculations were nade for fuel of several different initial enrichments and, at each enrichment, a limiting reactivity value, which .

included an additional factor for uncertainty in the burnup analysis, was i established. Burnup values which yielded the limiting reactivity values were then determined for each enrichment from which the acceptable burnup  !

domain for storage in Region 2, as shown in proposed technical specifica-tion Figure 5.6-1 (Figure 5 of this SE), was obtained. The staff finds '

this procedure acceptable.

?.1.? Treatment _of Uncertainties A correction for the reactivity effect of pool temperature is included as well as a georietric modeling effect bias to account for nesh spacing and sneared stainless steel-water composition effects.

For the Region 1 analysis, the total uncertainty is the statistical con,bination of the calculational uncertainty and nanufacturing and

mechanical uncertainties due to variations in Boraflex thickness, inner stcinless steel storage box dimension, stainless steel thickness, and  ;

fuel enrichirent and density.

In the Regan ? analysis, the sarne uncertainties are considered. In j addition, at,oncertainty due to the burnup analysis is estimated and t conbirad statistically with the other uncertaintles.

The staff concludes that the appropriate uncertainties have been con-  ;

sidered and have been calculated in an acceptable manner. Ier addition, these uncertainties were determined at least at a 95% probability 95% i confidence' level, thereby meeting the NRC requirements, and are acceptable.

2.1.3 Results.of Analysis .

For Region 1, the rack multiplication factor is calculated to be 0.9750,  ;

including uncertainties at the 95/95 probability / confidence level, when fuel having an enrichment of 4.5 weight percent U-?35 is stored therein.

Although the pool is ncrmally ficoded with water borated to 2500 ppm, -

unberated water was assured in the analysis.

For Region ?, the rack multiplication factor is calculated to be 0.9 0P for the most reactive irradiated fuel pemitted to be stored in the racks, i.e., fuel with the minimum burnup pemitted for each initial enrichment as shown in Figure 5.6-1. The design will accept fuel of 4.5 weight percent U-735 initici enrichment burned to 40.0 FWD /kgu. The analysis of the Region F racks aise assumes full flooding by unborated water. ,

Therefore, the results of the criticality analysis meet the staff's  !

acceptance criterion cf K nu greater than 0.95 including all l uncertaintiesatthe95/95%ohability/confidencelevel.  :

Post abnormal storage conditions will not result in on increase in the K of the racks. For example, lost of a cooling system will result in a'NcreaseintheV f value since reactivity decreases with decreasing f water density. ThiI$nalysisalsoallowsnew(unirradiated)fueltobe .

stored in the Region 1 racks in a dry condition prior to initial core <

loading of Unit 2 since inad.ertent partial or full flooding would main-tain K,ff less than 0.95.

u lt is possible to postulate events, such as an inadvertent misplacerent '

of a fresh fuel asserbly either into a Region 2 storage cell or outside 6nd adjacent to a rack module, which could lead to an increase in pool  !

reactivity. However, for such events credit may be taken for the  !

Technical Specification requirement of at least 2500 ppm of boron in the i refueling canal during refueling operations. The reduction in the K

• value caused by the boron rore than offsets the reactivity addition ,ff i caused by credible accidents.

The staff confidered the possibility of irradiation induced axial shrinksge of the Boraflex panels as documented in NRC Information Notice I I

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l. No. 87 43. Based on this, the applicar t has perforned analyses to determine the reactivity effects of potential Eoraflex shrinkage on the South Texas spent fuel pool.

Several scenarios were evaluated ranging from shrinkage of the top and bottom of each Boraflex panel with . corresponding exposure of active fuel at each end, to a single tear in each panel at the active fuel mid-plane. The results indicate that sufficient margin is available in both the Region 1 and Region 2 rack design to acconnodate at least E inches of shrinkagt 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 traintenance of a k f less than 0.95. For Region 2, mid-plane gaps in every panel of upto3.<g inches could be acconcodated. If the trid-plane gaps are assencd to occur in only two of the four panels in each Region ? cell, gaps as large as 10 inches could be accom odated without preventing the raintenance of a k less than 0.95. Thereforc, although it is not likely that sig-nificaNfgap formation will occur in the Boraflex panels, the staff believes that there would be sufficient tine to detect such anonrelies and provide appropriate actions before any significant adverse reactivity effects occur.

4 E.7 Conclusion Based on the reviev .,cr' bed atove, the staff finds the criticality aspects of the desi p ^^ the South Texas Unit 2 spent fuel racks to be acceptable and to ree . e ree,uirements of General Design Criterion (GDC) 62 for the prevention , .riticality in fuel storage and handling. The staff concludes that fuel from Unit 2 may be safely stored in Region 1 provided that the enrichnent dces not exceed 4.E weight percent U-735.

Any of these fuel assenblies nay also be stored in Region 2 provided they neet the burnup and enrichnent limits specified in Figure 5.

1 3.0 PATERIAL COMPATABILITY AhD CPEPICAL STABILITY 9

3.1 Discussion The staff has revicwed the conpatability ard chemical stability of the raterials (except the fuel assenblics) wetted by the pool water, in 4

accordance with Section 9.1.P of the Standard Review Plan (NUREG-0800, July 1981). The STP-? pool contains oxygen-sat"-

• ineralized water which has 2500 parts per million of boro'n as bo c

.m. The pool is lined with stainless steel and has two adjacent a ,f storage.

The principal construction materials for the propose' 'x racks in the spent fuel storage pools are ASTM A-740 Type 304L austenitic stainless steel for structure and Boraflex for neutron absorption. The racks are welded honeycomb arrays of square stainless steel boxes fortring individual cells for fuel storage. Each of the four sides of a given storage cell has a Boraflex assenbly, except those sides that are nearest to the storage pool walls.

In Pegion 1, the Borafiex assenbly consists of a thin rectangular stain 1 css steel water box with a Boraflex sheet affixed on ort side of 5

the box and another Boraflex sheet on the other. A thin stainless steel plete is welded over each of the two Poraflex sheets on the water box.

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, The entire assenbly is removable from the storage cell. In Region ?, a Boraflex sheet is positioned between two adjacent walls of the square storage 211s. The Boraflex sheets in Region ? are not removable. In both regions, single sheets of Boraflex are used, and the Boraflex sheets are not physically fastened to or permanently glued onto any surface or ,

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HLt.P proposed an inservice surveillance program for the Boraflex materiel, using sample coupons that are made of the same material con 1 position, fabricated by the same method, certified to the same criteria, cut to the sane physical dinensions, and encased in the same material as the removable Boraflex assenblies. A minirum of one such coupon and a string of foot-long sarples of the same raterial will be provided for the storage puoi in Unit ?. Evaluation of the coupon performance will include visual inspec-tion and reasurenents of the neutron attenuation, h6rdness, and physica!

din >ersions. Initial surveillance of the specimens will be perfornied efter ,

five ye6rs of exposure to the storage pool environrent. Based on the '

results of the initial surveillance, HLt.P will deternine the schedule and extent of f.dditional surveillance. HLtP, however, has provided no corrective actions to take when degradation of the Boraflex assenblies is found, but will evaluate available plant data on Foraflex perforvance fron.

the nuclear industry to modify the surveillance progran when warranted and justified.

3. Evaluation ,

The stainless steel in the storage pool liners and rack asserblies is cctpatible with the oxygen-saturated borated water and radiaticn ,

environment of the spent fuel pools. In this environment, corrosion of Type 304L stainless steel is not expected to exceed a rate of 6 x 10"  ;

inch per year. This corrosion rate is negligible for even the thinnest stainless steel walls of 0.03 inch in the rack assemblies. Contact corrostun or galvanic attack between the stainless stael in the pool liners or rack assemblies and the Inccnol/Zircaloy in the fuel assenblies i to be stored will not be significant, because all these raterials are '

protected by passivating cxide filrs. Boraflex is corposed of n",e* " *!lic raterials with the retaland, therefore, will not develop a galvanic e.omponents. '

SpGt 4 vailable to allow escape of any gas that may be generated fron the polyrer binders in the Eoraflex during heating and irradiation, thus preventing possible bulging or swelling of the Boraflex assenblies.

Boraflex, an elastorer of nethylated polysiloxane filled with boron t carbide powder, is used as a neutron aLsorber (poison) in the spent fuel storage facilities of rany nuclear power plants. It has undergene extensive testing to deterriine the effects of ganea irradiation in i variou; environrents and to verify its structural integrity and suitability i

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7 as a neutron absorbing rraterial. The evaluation tests have shown that

'oraflex is unoffected by the pool water environrrent and will not be degraded by corrosion. Tests were perfog'ed at the University of Michigan, exposing Boraflex to 1.03 y 10 rads of ganna radiation with substantial concurrent neutron flux in borated water. These tests inoicated that Borafiex mahtains its neutron attenuation capabilities after being subjected to an environtrent of borated water and garra irradiation. Irradiaticn will e.ause sone loss of flexibility and shrinkage of the Boraflex.

Long-term borated water soak tests at high terrperatures were also conducted. The tests showed that Boraflex withstood a borated water irriersion at F40*F for 251 cays without visible distortion or softening.

The Ecraflex showed no evidence of swelling or loss of ability to traintain a uniform distribution of boron carbide. The spent fuel pcol water terperature under norral operating ccnditicns will be apprcxiraately 10 P F, which is well telcw the 240*F test terperaturt, Ir gere.ral, the rete of a chemical reaction decreases expcnentially with decreasing ttnperature. Therefore, the staff does not anticipate any significant deterioration of the Ecraflex at the pool norral operating condi' ions over the design life of the spent fuel racks.

The tests have shown that ncither irradiation, environment, nor Boraflex ccrposition have a discernible effect on the neutron transmissicn of the Forafitx raterial. The tests also have shown that Foraflex does not possess leachable halogens that rnight te released into the pool environrent in the presence of radiatien. Similar conclusions were reacted regarding the leaching of elenental borci from the Beraflex.

Poron carbide of the grade normally present in the Ecraflex typically contains 0.1 weight rercent of soluble boron. The test results have confiroed the encapsulation capability of tte silicone polyrtr matrix to prevent ;be leaching of soluble species fron the boron carbide.

Recently, arcralies ranging from rninoi physical changes in color, sire, hardness, end brittlentss to gap forr.ation cf up to four inchas in width were observed in Forafley panels used in three nuclear power plants. The exact rechanisrs that causad the observed physical degradations of Boraflex have not been confirred. But the stLff can postulate that gantra radiatien frer the spent fuel inittelly induced crosslinking of the polymer in Boraflex, producing shrinkage of the For'aflex raterial. When crosslinking becar.e saturated, scissioning (a process in which bonds betwetn ators are broken) of the polyrer predominated as the accumulated radiation dose increased. Scissioning prod 9ced porosity, which allowed the spent fuel pool water to perrneate the Foraflex material. Sc' ioning and water perreation could ernbrittle the Boraflex material. .n short, garrea radi-Dy {. ation from spent fuel is tha most probable cause of the observed physical degradation., such as change > in color, sire, hardness, and brittleness.

}j r.[. i q l'ne staff does not have sufficient infortation to deterrnine conclusively A.M what caused the gcp foraation in sone Boraflex panels. Hewever, it is 4[ conceivable thct t' the two ends of a full-lengtt Foraflex panel art J physically restrained, then shrinkage caused by garra radiation can breal a" up the panel and lead to gap forFatien.

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The staff determined that reasonable assurance exists that physical restraints are absent in the Coraflex panet of the South Texas Project, because the Boraflex sheets are not physicaily fastened to or permanently glued onto any structure. It is not likely that gaps will forn, to any significant extent in the Boraflex panels during the projected life of the Boraflex assenblies. However, minor physical degradations can take place in the Boraflex from irradiation.

In the unlikely event of gap forvation in the Boraflex panels that would lead to loss of neutron absorbing capability, the monitoring program will ,

detect such degraded Foraflex panels, and FL&P would have sufficient tire to replace them in Region 16f the storage pools. In Pegion 2 where the Boraflex sheets are not removable. HL8P can either place new Boraflex sheets in the affected empty storage cells or testrict the use of the affected cells for fuel storage if degraded Boraflex is fcund, 3.3 Conclusions Based on the above discussion, the staff concludes that the corrosion of the spent fuel pool cocponents due to the pool environnent should be of little sienificance during the life of the facility. Corponents in ite spent fuei storage pools are constructed of alloys which have a low differential galvanic potentiel tetween them and have a high resistarte '

to gereral corrosion, 1ccalized corrosion, and galvanic corrosion.

The staff further concludes that the er.viiontental corpatibility of the i rattrials used in tne spent fuel storage pcols is aderivate based on the l test dbte cited in Section 3.2 and actual service experience at operatine reactor facilities.

1 l The staff has reviewed the proposed surveillarce program for nonitoring l the Foraflex in the spent fuel storage pool and concludes that the progran can reveal deterioration that might lead to loss of neutron absorbing capetility during the life of the spent fuci racks. However, ,

if a significant loss of neutron absorbing capability is found in any Foraflex panel, HL&P should take corrective actions such as replacnent cf the tjegraded Foraflex par.el, insertion of a new Foraflex 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 selection of appropriate raterials of construction by HLEP meet the recluirements of 10 CFR Part 50, Appendix A, GDC 61 regarding the capability to perinit appropriate periodic inspection and testing of corponents, and GDC 6?

regarding prevention of criticality by the use of boron poison and by reintaining structural integrity of cocponents, and are, therefore, acceptable.

. a. STRUCTURAL DESIGN For this portion of the review, the primary focus was assuring the structural integrity of the fuel, the fuel cells, rack modules, and the spent fuel pool floor and walls under the postulated loads (Appendix D of SRP3.8.4),andfuelhandlingaccidents. The major areas of concern and their resolution are discussed in the following paragraphs.

4.1 Fuel Hand _ ling Puild h y L Spent Fuel Pool  ;

The Fuel Handling Building analysis and design 4 s reviewed and accepted during the Operating License stages. HLAP, however, performed the seismic analysis of the fuel handling building incorporating the revised rass of the proposed reracking (consolidated fuel). The soil structure inter-actiun analysis and the input rotion kere considered in the sane way as in the original analysis. A corparative review of the output tables indicated less than 57 differentes in modal frequencies. Aho, a cenpa"ison of acceleration response spectra indicated ns 111gitle differences in spectral accelerations at the spent fuel poo. floor level. The analysis thus confirted the validity of the basic input criteria for the seismic analysis of the hich density rocks. HLLP also recalculated the differences 13soiTbearingpressureandthefactorofsafetyagainst lic,uification due to the added mass. The soil bearing pressure increased from 70.? ksf to ?? ksf against the allowable of 37 ksi under dead lead and Safe Shutdown Earthquake (SSE) conbination. The minirum safety fetter of 1.4 against lic,uificaticn remained unaffected.

HLtP also performed a detailed seisn'ic analysis of spent fuel pool areas 6ffected by the prepostd retacking. HLAP dencnstrated that the minint.m saftty factors at various critical sections of the post walls and floor slab were higher than 1.0 for all conditions of loading considering the 4 consolidated fuel weight. However, the design rargin for transverse shear in the spent fuel peol floor slab was marginally above 1.0 for standard fuel. For consolidated fuel HLIP demonstrat(d a sinilar margin when the confirmatory-basis response spectra descrited in the FSAR Section 3.7.E.4 was used. This evaluation pertains to the use of standard (single fuel assenbly per stors.ge location) fuel, for which the staff considers the design to be adequate.

4.? Rack Analys_is_and Design.

Tables 1, ? and 3 provide the rack module data, dirensions and pertinent modeling paraneters. HLLP's analysis is based on one set of synthetic tire-histories. The staff expressed concern rega* ding the adequacy of energy content at the frequencies of interest, when used for non-lintar rack analysis. HL&P generated the Power Spectral Density (PSD) functions for the floor input motion used in the rack . analysis and compared them with the target PSD obtained by the method in NUREG/CR-3509, "Power Spectral Censity Functio'ns Conpatible with Regulatory Guide 1.60. Response Spectra " A typical conparison is shown in Figure C. In general, and particularly in the low frequency range of interest, the computed PSD

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cxceeded the target PSD by a good margin. A dip at 6.3 Hr was indicative of the characteristics of the desigt, response spectra. At frequencies.

higher than 12 Hr, the dips are expteted when PSDs for instructive motion are compared to the target PSD for ground motion. On an overall basis, the comparison indicated adequate energy content for time-histories being used for the rack analysis.

Requirenents for seistric and inpact loads are discussed in Section 3 of Appendix 0 of SRP Section 3.8.4 There it is stated that seismic excitation along three or orthogonal directions should be imposed simul-4' taneously for the design of the new rack system. HLAP's original rack analyses were based on the square root of the sum of the sqvares (SRSS)

, conbination for the rack responses due to the three components of the 1

earthquake tn be considered.

The rack responses (displacenent, forces) were separately calculated for two horizontal directions using a prcprietary cornputer code "RACK 0E".

The responses due to ine sertical cenponent were calculated using an equivalent static cetted with a dynariic load factor of 1.5. This procedure appeared to provide bcunding calculations for forcas at the pedestal, but the staff expressed concern regarding the ability of the procedure to provide a realistic assessment of the displacen'ents under the three conponents of an earthquake. "RAckOE" as used by PLIP is a two-dirensional ecde, capable of performing two-dimensional dynamic analysis with sirrultaneous 1 stismic triput in two directwns. PLtP perforned multi-rack analyses of j 3-racks in a rcw (E-W direction) with varying Icading conditions for each a recl with simultaneous applicaticn of E-l' an( vertical tine histories.

The culti-rack rodel is shckn in Figure 6. The gap used between the racls I

in these analyses was 1.0 inches. The results indicated that the raxiruni i relatise displacenent between the two adjacent racks vere 0.73 inches and 1 0.41 for the coefficients of friction of 0.? and 0.8 respectively. Ncne of the cases showed rack-to-rack cr rack-to-wall interactions. The cross couplir:9 eff(cts were ist.ored in these caltblations resulting in calculated displectnents that are leT.r than the actual displacenents.

Pased on the conservatively conputed reximurn loadings, stresses in varicus critical cenponents of the rack nodules were corputed for load corbinations reconcended in Table 1 of Appendix D of SRP 3.8.4. The stresses were conpated against the requirements of Subsection KF of ASME Code and minitrum safety factors es ratios of the lillowable divided by the actual stresses were computed. Table 4 is a sumnary of the-safety factors at critical rack locations.

Fased on the results of the FL&P's seismic analysis, the staff concludes that during an SSE, the fuel racks will maintain their structural integrity, fuel asserblies will not sustain darrage, and rack displacenents will nct be large enough to result in rock to rack or rack to wall impact.

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4.3 Fuel Pandlindecjdent_ Consideration 9

HL&P performed structural analyses and evaluations of four postulated fuel handling accidents:

A. Dropped Fuel Accident 1 A consolidated fuel canister was assumed to have dropped f rom 14 inches above the top of a rack module and directly inpnted the bottom plate of a fuel cell. The final velocity and totai energy '

were considered assuming no energy dissipation in the canister. -

Based en this consideration the bottoni plate weld could fail, thus al!cwing the bottom plate and the fuel canister to impact the pool li.er. However. HLIP derunstrated that the impact energy will nct perforate the liner. HL&P used Fallistic Pesearch Laboratory  ;

(BRL) formula for predicting the liner penetration / perforation.  ;

Considering the conservative approach used by HL8P and that  :

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the staff's evaluation concerns single fuel asscrblies, the analysis l results are acceptable to the staff. '

B. Dropped Fuel Accidents ? and 3  !

In Accident ?. FLlP considered a drop of a consolidated fuel canister 1

frcm 14 inches above the top of a rack. The top po-tien of the rack could sustain sont plastic (pernanent) deforration. Powever. HLtP's ,

calculations confirned that the safe, suberitical configuration of the i stored fuel would not be compromised due to such an accident. The staff accepts the FL&P's findings for the purpose of this evaluation. i Accident 3, which postulates an inclined drop of a fuel canister, would not t:e as severe as Accidcnt P. as the impact energy will be distributed over a large area of a rack nodule.

C. Jarced Fuel Asserbly FLAP cons.idered the rack stresses when 4000 lbs, of force was I applied to unjam a fuel assembly in a storage location. This  !

force was considered at any height of the fuel storage cell.  !

HL&P's calculations indicated the stresses resulting from appli-  !

cation of such a force to be withirr the acceptable criteria. The i staff finds the postulation of the force to be reasonable and the final conclusions to be acceptable.  !

In any of the postulated accidents damage to the dropped or jamed fuel assenbly is possible. According to FL8P. the consequences of such damage  ;

are bounded by the design basis fuel-handling accidents described in the  :

licensee's FSAR Section 15.7.4 j i

i

4.4 Other Phases of Installation The above evaluation is based on the wet storage of fuel in the fuel pool. In this request HL&P proposes to retain certain flexibility in the installation of the high density racks.

In Phase I, HLLP plans to arrange two rows of 8x6 (Region 1) rack irodules with three trodules in each row. The minimum distance between the rack rrodules will be 24 inches in either direction. The minirrum distance to the Nalls will be 29 inches. In this configuration, HLAP proposes to have dry storage of new fuel asserrblies. As there is no water in the pool during this tint, the resistanct provided by the water to the moverrents of the racks under a postulated selsric event is non-existent. The racks could slide and rotate on their pedestal. HL8P's analysis of dry storage conditions indicates that the maxinurn sliding that can occur for one rack could be 6.44 in. If the adjacent rack slid the ser e amount, the iraxirruni relative displacement would be less ther. 13 in. As the rninint.m distance proviced is 24 in., there is no likelihocd of a rack module irrpacting another rack or the nearest wall, l' hen the high coefficient of fricticn of 0.8 is cor'sidered, the calculations indicate that the rack rrodules J

would not slide but could rotate on the pedestals. In order for a rack to overturn, the instantaneous center of gravity of the tack would have to trove 11 1 ;hes horizcntally which would result in the required uplif t of over 4 inches at pedtstal. PLIP's analysis indicates a maxirruir cortbined

. Uplift to be less than 1 in. Thus, the staff concludes that the racks and the stored fut1 will trair.tain their integrity under Phase I configuration r:f rack rredules.

4.5 Cenclusion,s The proposed Phare II installation is sirnilar to that t,eing evaluated (forPhast III), except that the three scuth side racks will not be installed during Phase II. Fased on the evaluations of Phase I and PFASE !!! configurations, the staff considers the Phase !! installa-tion to be acceptable.

5. SPENT FUEL POOL COOLING AND LOAD HANDLING The sutoittal 44, and 61, andwas reviewedofinNUREG-0800, the guicelines accordance "Standard with the require ents (SRP Review Plan" of GDC ?, )

and NUREG-0612. "Control of Heavy Loads at Nuclear Power Plants."

5.1 Decy , Heat Gettration Rate HL&P stated in the Parch 8,1988 submittal that the calculation of the decay beat generation rate was in accordance with the guidelines of NUREG-0800, SR -' tion 9.1.3 and Branch Technical Position (BTP) ASB 9-2.

For the nonral terun heat load case the HL&P assurned the pool was filled

i

,. 4 with one-third core refuelings every I? months (maintaining a full ecre discharge capability) with the final one-third core beir.g placed in the poul at 140 hours0.00162 days <br />0.0389 hours <br />2.314815e-4 weeks <br />5.327e-5 months <br /> (Case A) 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.

4 The twu 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 t.opebility 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 />. Both numbers are conservative with respect to the specific reconcendatio.'s in SRP Section 9.1.3 ethich is 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> l (Case C). HL8P calculated heat losds and fuel pool 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 assurptions of one-third 1 core af ter 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 maxinun. calculated pool torperatures with one and two i trains operating are:

1 Cooling, Train 2 Cooling 3ain I Case A 145.7'F 126.0*F Case B 150.7'F 129.P'F Case C 131.2'F 118.7'F For the abnorral rnazirun heat load case (Case D) FLt.P assured the ure conditions as in Cases A and B except that the last cne-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. These assumptions are also censervative corpartd to the reccm-rrendations of SRP Section 9.1.3 which are one-third ccre in the pool for 400 days, one-third for 30 days and one full core at 150 heurs after shutduwn. The calculated pool water terperature for Cat,e D is 15F.a*F with two pool cooling trains operating.

To serify HLlP's calculated spent fuel heat loads, the staff perforced an independent calculation for the rnarirtuni abnorvel storage condition of Case D using BTP ASB 9-2 guidel'.nen The staff calculated a heat load of 58.03 tTtu/hr corLared to the licensee's calculated value of 63.15 FBtu/hr.

Because the calculated value is conservative corpared to the staff's (HLEP assured last refueling was greater than 1/3 core leaving no erpty storage spacvs) and not appreciahly different based en the high rate of decay heat energy the steff finds that HLt.P has properly calculated the heat generation rate in accordance w'itn the SRP.

5.? Spent Fuel Pool Cooling System The spent fuel pool cooling systern (SFPCS) consists of two seismic Category 1. Cuality Group C cooling water trains each with one purp and one heat exchanger. After the spent fuel pool water is cooled in the heat exchangers, it is purified by the non seismic Cetegory I cleanup

. system. In the event of a loss of the SFPCS, there are several sources cf pool trakeup water available including a seismic Category I source from the low-head safety injection purops, t 6

In its April 19b6 Safety Evaluation Report (SER), NUREG-0781, for South  :

Texas Units i and 2, the staff concluded th6t the SFPCS tret the acceptance crit **% of SRP Section 9.1.3 including GDC 2 and was acceptable. The bases for this conclusion have not changed as a result of the proposed ,

reracking, except with regard to the requirements of CDC 44, "Cooling  :'

Water". The change in the basis for GDC 44 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 terperature recunrendation of SRP Section 9.1.3 when  ;

calculating the maximum norrral heat load using the assunptions identified t in the SRP. Under the higher heat load ecoditions identified using HL&P's fr. ore conservetive assunpiirns for South Texas, the recorrended pool e terperature of 140*F for single train operation is ecceptable because:

a. The assurptions used in the calculations are rore censervative than  ;

staff guidelines;  !

h. The SFFCS is a safety-related systeri;

c. For the worst cese (Case A) the 140'F could be exceeded for only 11.5 days;

d. Vith two trains operating, the pool teriperature for Cases A and R  ;

ye well below 14C*F; ,

c. The 140'F is a recorcended limit and the litelihood of exceeding  ;

that reconnendation is low given the ccr.ditions and conservatisrs '

assuned in the calculation; and

! f. The effect of pool watter terperature slightly above 1ACF on s$ent  :

l fuel ste-age safety is ne.gligible.

For the abnormal maxinun heat load (Case D), the SFPCS will maintain pool watt.r terptrature at cr t'elow 155.4*F with two trains cf couling which is ,

well telow the recorr: ended no boiling limit of SRP Sectien 9.1.3 under these conditions.

As a result of its review, the staff finds that the SFPCS still retts the .

requirements of GDC 44 with respect to providing adequate pool cooling i i

under maxinun norwal hett load conditions following a single failure.  ;

i 5.3 1.o g of Cooling )

I in the event that all SFP cooling is lost, the spent fuel pool terperature .

l will inctease until boiling occurs. HL&P has estimated the tirre from the ,

i i

I i

loss of pool cooling until the pool boils for the four cases idtntified earlier. 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 gper

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 pgr

d. Case 0 - 2.86 hours9.953704e-4 days <br />0.0239 hours <br />1.421958e-4 weeks <br />3.2723e-5 months <br /> at 135 pgm The assured (seismic Category I) pool makeup water source from the low oressure injection system has a capability in ex:ess of the above 11-off rates. This seismic Category I riakeup source is adequate to govide water for the higher boil-off rate of the expanded storage capacity and, thei3 fore, still meets the requirerrents of GDC ?. "Design Bases for Protection Against Natural Phenomena."

5.4 Building 3Mjlgjen 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 conclus;ons regarding the consequences of a fuel handling accident identified in Section 15 of NUPEG-0781 have not changed as a result of the proposed increased storage capcity hecuse the accident analysis is based on the activity released frori the last one third of a core placed in the pool. Thus. the FHB ventilation exhaust syster continues to be acceptable.

5.5 Heavy load Handling The new spent fuel storage racks are considered to be heavy loads and will be noved by the FHB overhead crane. The FHB overhead crane is a singit.

failure-prcof crane which meets the guidelines cf NUREG-0554. "Singic Failure-Proof Cranes for Nuclear Power Plants" ano NUREG-0612 "Control of Heavy Loads at Nuclear Power Plants." as indicated by the staff's acceptance in NUREG-07E1.

Because the reracking will ta6e place prior to the Unit 2 first refueling, heavy loads will not be carried over spent fuel during the revacking operation. The rethods and equiperent used for the veracking will be in accordance with Section 5.1.1 of NUREG-O'612, which includes the identification of safe load paths for heavy loads, procedures for load handlino and operator training.

The spent fuel shipping cask cannot be carried over the spent fuel pool due to crane travel limithtions, and a cask drop accident will not affect spent fuel or the spent fuel pool cooling system as previously detertnined by the staff. Therefere. storage of spent fuel in the new proposed high density storage racks will nct affect the staff's previous acceptance of

. . t 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 performed in accordance with the guidelines of NUREG-0612 and ther6 fore the requirements of GDC 61. Fuel Storage and Handling and Radioactivity Control," are net as they relate to proper load handling to ensure r i against an unacceptable release of radioactivity or a criticality '

accident as a result of a postulated heavy load drop.

! 5.6 Conclusion  !

L

.l Based on the above, the staff concludes that the proposed expansion of I the Scuth Texas, Unit 2 spent fuel pool storage capacity complies with the requirements of General Design Criteria 2, 44 and 61, the guidelines of NUREG-061P and applicable acceptance criteria of the Standard Review Plan with respect to the capability to provide adequate j spent fuel pool coolint and safe handling of heavy loads. The staff, j therefore, concludes that the proposed spent fuel pcol expansion is acceptable with respect to spent fuel pool cooling and load handling.

6 SPENT FUEL FOOL CLEAT:UP SYSTEM l The spent fuel pool cleanup system at STP-2 1s an integral part of the i spent fuel pool cooling system. The system is designed to reintain water quality and clarity and to renove decay beat generated by the spent fut.1 assu blies in the spent fuel pool and in the terporary in-containnt:nt

! storace crea. The cleanup systcm is also designe. .o purify water in the j

refueYingcavitysndtherefuelingwaterstoragetank. The system includes all corponents and piping from inlet to exit from the spent fuel pool, in-ccntainment storage area, refueling cavity, and piping used for fuel pool raleup, from the re.fuelirs water storage tank and the cleanup filters /demintralizers to the point of distb6tge to the radweste system.

Tbt spent fuel pool cooling and cleanur system consists of two raximum nom 61 teat lead full-capacity fuel pool ecoling trains (each with a punp

) and heat o changer), two 6 mineraliter purific6 tion trains, a spent fuel poci surface skirrer loop, and a reactor cavity filtration systcm. The a spent fuel pcol cooling punps can be powered frcm the Class IE energency sources.

Radioactivity ard impurity levels in the, water of a spent fuel pool incr<9se primarily during the refue'.ing opf. rations as a result of fission I produs.t leakage from defective fuel eierents being discharged into the pool and to a lesser degree during other spent fuel handling operations.

The rerackirg of the spent fuel pool at the South Texas Project Unit 0 5111 not increase the refueling frequency and fraction o' the core replaced after each fuel cycle. Therefore, the frequency of operating the spent fuel pool cleanup system is not expected to increase.

J M

Sirrilarly, the chemical and radionuclide cenposition of the spent fuel '

pcol water will oct change as a result of the proposed reracking.  ;

following the discharge of spent fuel from the reactor into the pool, the j fission product inventory in the spent fuel and in the pool water will  !

decrease by radioactive decay. Furthermore, experie 12 also shows that there is no significant leakage of fissien products frorn spent fuel stored in pools af ter the fuel has cooled for several roonths. Thus, the ,

increased quantity of spent fuel to be stored in the South Texas Project, ,

Unit ? fuel pool will not increase significantly the total fission product activity in the spent fuel pool water during the cperation of the pool.

7. RADIATION PROTECTION AhD ALARA CONSIDERATIONS 1

In as noch 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 ne l additional occupational radiation exposure associ6ted with the reracking l cf the spent fuel pcol.

HL&P has considered any increases in expcsure frorn spent fuel storage. [

airborne radiaticn, solid radioactive waste (resins, filters, and  :

corrosion product crude) ar d concluded that no significant increases  ;

are e>pected. The staff hes rt; viewed HL&P's analysis and finds therr  ;

acceptable.

The radiological protection of workers during fuel b6ndlins operations l will not change because the spent fuc1 will renain covered by E3 feet of l rater, as before, and spent fuel will be covered by at least 10 feet of  :

wcter during spent fuel hardling operations.

l Pased en the review of the HL&P's cubnitt61, the staff concludes that the

• projected activities and estir.ated person-ren doses for this project are reascoable. HLtp intends to tale ALARA conside rations into account,  ;

and to inplerent reescnable dose-reducing activities. The staff concludes s that HLFP will be able to reintain individual occupational radiation exposures within the applicable lirrits of 10 CFR Part 70, and tr.aintain >

doses LLARA, consistent with the guidelines of Regulatory Guide P 8. ,

There, re, the proposed radiation protection aspect of the SFP rerack is

acceptable.

8. ACCIDENT ANALYSES i

t lhe staff has reviewed the accident analysis that could occur at STP-? in  !

] conjunction with the preposed reracking. The applicable accidents were  :

I l

cask drop, loads over the spent fuel, and spent fuel pcol boiling.

r I The proposed changes do not affect the previously approved cask drop l l analysis. Crane design and building arrangenent prevent trovecent of the  !

i cast over the fuel pool and prevent interference of the cask crar.e i j bfidge, trolley, and hoist with fuel racks or building structures. The l reil for the cask handling crane sttis at the edge of the cask loading *

]

< i i

I

, }

pool, which is incre than 75 ft. from the spent fuel pool boundary. ,

Building arrangenent, crane control, and lifting rig design restrict l vertical lift of the cask to an elevation such that the cask will not te i 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 designed  ;

to sustain a free-fall of 30 feet onto an unyielding surface followed by a specified puncture, fire, and inversion in water with the release of no trore than a specified small quantity of radioactivity. l The spent fuel cask crane is not capable of traveling over the spent fuel pool.  !

t in the spent fuel pool boiling accident it was assuned that a loss of f spent fuel pool cooling occurred af ter a refueling where 1/3 of the core i had been rer.oved and placed in the spent fuel pocl. As a result of  ;

boiling it was assuned that the greatest contribution to iodine Icatage i was fron the off-loaded 1/3 core. The dose consequences at thc exclusion t rene boundary (0-? hours) arid low population rcne boundary (0-30 cays)  ;

were 0.000? and 0.54 thyroid-rerr. respectively. [

The pottntial doses resulting from the accidents censidered were well  !

below the allowable 10 CFR Part 100 guidelines. Therefore, the accident i aralysis aspect of the spent fut.1 pool rerack is acceptable.  !

?. RAD 10AC11VE WSTE TREA1FENT I 1he plant contains a radioactive waste ranagerrent systerr designed to previde for the controlled handling and treatrent of licuid, gaseous, and l solid wastes. The radioactive waste managerent systert was evaluated in staff Safety Evaluation Report (SER) datec April 1986 (htJREG-0781).

1bere will be oc changes in the syster described in the SER because of the proposed SFP rerack.

10. ENVIRONFEhTAL C0hS10 IRA 110h5  ;

The staff has reviewed the proposed spent fuel pool rodification to the i Scuth Texac Froject Unit ? relative to the requirenents set forth in l 10 CFR Part 51 and has cencluded that there are no significant >

radiological or non-radiolejical impacts associated with the proposed [

action. Further, any irrpact asseriated Vith this action are cortprehended l by the Environtrental Final Statercnt--Construction Phase (March 1975).  !

11. CONCLUSIONS The staf f 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 cf the spent fuel rack rrodules and the spent fuel pool are in corpliance with the acceptance criteria set forth in the FSAR and consistent with the current licensing practice, and therefore are acceptable.

. t The approval is based on the storage of non-consolidated fuel and the installation of all racks prior to r.tcrage of any spent fue's in the spent  ;

fuel pool. If PLtP would change its plans and decide to store spent fuel in the pool before completing the installatien of the new rods it should suteit docuner.tation to the staff for review addressing all significant changes frorn the request the staff is new approving.

Dated: October 5, 1988 k

1 l

High Dernity Spent Twel 5torage Racks

South Temas Eleork Generating Statum -

Houston bghting & Powyer Coenpany

624-(52-)  :

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i TABLE 1 i

SPENT FUEL RACK DATA i Rack Module Storage Cells Array  ;

Region Number Per Module Size

• 1 48 1 8x6 I 1 2 48 8x6 1 3 48 8x6 1 4 48 8x6 l 1 5 48 8x6 l 1 6 48 8x6 l 2 7 110 10 x 11 l 2 8 110 10 x 11 '

2 9 110 10 x 11 2 10 121 11 x 11  :

2 11 132 12 x 11 l 2 12 132 12 x 11 l 2 13 121 11 x 11 2 14 121 11 x 11 i 2 15 132 12 x 11 i 2 16 132 12 x 11 t 2 17 110 11 x 10 '

2 18 110 11 x 10 l 2 19 120 12 x 10  !

2 20 120 12 x 10 l

[

• The array sine indicates the number' of storage cells in the N S I direction a the number of cells in the E.V direction. ,

Note: This is the came table as Table 3.2 in Reference 1,

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TABLE 2 RA M N0DULE DIMENSIONS AND WEIGHTS Estimated Dry Rack Weight (1bs)per Module Nominal Cross.Section Estimated Dry Module with Number Dimensions (inches) Weight (1bs') Single Density NS E.V For Module Fuel 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 23,040 225,660 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 l' 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 #

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. l TABLE 3 RACK MODEL PARAMETERS

• Rack 11 x 10 12 x 11 8x6 Module 6

Ky (1b/in) EW .099 x 10 .0871 x 10 6 .057' x 10 6 NS .099 x 10 .0873 x 10 5 96 x 10 7

K (l'/in) EV 1.77 x 10 d

N.S 1. 7 x 10 1.84x10f 1.75 x 10 1.40x10f 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 Vf L

((1b)

(in) 386936 91.50 100.65 463094 88416 65.70 L* (in) 100.65 109.80 87.60

• 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 vall & fuel assembly - 0.237"

, I Kg Tuel assembly to cell wall impact spring rate  :

i K

d Vertical axial spring rate for concrete, pedestal, base plate and cell i deformation. I h Length of suppr.rt les H- Height of rack Obove base place ,

l V - Weight of rack v.?hout fuel Weight of fuel Wf f L -

Platforv dimension (x - Direction s. East) l L* - Platform dimension (y - Direction - North) -

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TABLE 4

SUMMARY

OF SAFETT FACTOILS IN CRITICAL FUEL RACK IDCATIONS Item / Location Safety Comments Factor *

• Support Footing 1.97 Table 6.6*

(Pedestal) to Baseplate Veld Stress Cell to Baseplate Wald 1.06 Table 6.6*  ;

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

i Impact Load Between Fuel 1.20 Standard Fuel Assembly and Call Wall i Shear Load on Baseplate 1.07 Table 6.6*

Near a Support Footing I

Compressive Stress in 4.23 Based on Local l Cell Wall Buckling Considerations  !

(Standard Fuel)

Rack to Wall Impact .

, No Impact with Pool leads Wal,1s occur at any Location

• Table 6.6 Licensing Submittal, ST.NL-AE.2417; see Table 6.6 for other related Safety Factors.

• • All Safety Factors.are for consolidated fuel unless otherwise noted.

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