ML20246M638
| ML20246M638 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 01/25/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| CON-#189-8341 OLA-S-001, OLA-S-1, NUDOCS 8903270215 | |
| Download: ML20246M638 (87) | |
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. 4 m iip 4 % r SArETY EVAltlATION pv THE OFFTCE OF NUCLEAR PEACTOR PEGULATION.
RELATING TO TPE RERACKING OF.THE SPENT FUEL POOL AT THF ST. LUCIE PI_ ANT, UNIT NO. 1 AS PFLATED TO AMENnMENT NO. 91 TO UNIT 1 FACILITY OPERATING LICENSE NO. OPR-67 FLORIDA POWFP AND LIGHT CfWPANY DOCKET NO. 50-335 I
I NUCLEAR REGULATORY COMMISSION l Cocket No. 50 ~DU 4d Official Exh. No. 50/ Ei, /
in the matter of IkiE da '/ do M - 5/, Q l V i Staff i- IDENTiflED - !
Applicant RECEIVED ' -
j intervenor ~~ REJECTED 1 Con r c r DATE !M N Other Witness Reporter b' Ih" '
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( TABLE OF CONTENTS PACE
- 1. INTRODUCTION 1
?.1 Licensee Submittal end Staff Review 1 1.2 Summary neseription of Rerecking ?
P. CRITICALITY CONSIDERATIONS 2
?.1 Criticality Analysis 2
?.? Technical Specification Changes 4 P.3 Conclusions 4
- 3. MATERIAL COMPATIBILITY AND CHEMICAL STABILITY 4
- 4. STRUCTI' PAL DESIGN 6
- 5. SPENT FUFL P00L COOLING AND LOAD HfNDLING 7 5.1 Decay Heat Generation Rate 7 5.2 Spent Fuel Pool Cooling System 8 5.3 Heavy Load Handling 9 5.4 Light Lead Handlino 11 5.5 Conclusions 11
- 6. SPENT FUEL POOL CLEANUP SYSTEM 11
- 7. RADIATION PROTECTION AND ALARA CONSIDERATIONS 12
- 8. ACCIDENT ANALYSES 17
- 9. RADIOACTIVE WASTE TREATMENT 13
- 10. SIGNIFICANT HAZARDS CONSIDERATION COMMENTS 13 4
- 11. FINAL NO SIGNIFICANT HAZARDS CONSIDERATION 18
- 12. ENVIRONMENTAL CONSIDERATIONS 21 1
- 13. CONCLUSIONS 22 14 REFERENCES 22 APPENDIX A: Technical Evaluation Report by Brookhaven National Laboratory
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1 1.. INTRODUCTION 1.1 Licensee Submittal and $taff Review This report presents the NRC staff safety evaluation for the reracking of the spent fuel pool at the St. Lucie Plant, Unit No. 1. By letter dated June 12, 1087, the Florida Power erd Light Company submitted an application to increase the storage capacity of the spent. fuel pool, including the appropriate and recessary charges to the Technical Specifications. The licensee requested the increase in storace capacity because the pool lost full core reserve capability following a refueling outage completed in Apri1~1987.
The June 12, 1987 reouest for the amendment, including the staff's proposed "No Significant Hazards Consideration," was noticed in the Federal _ Register on August 31, 1987 (52 FR 32852). Further details are addressed in Section 10 cf
! this report.
The application is based on the licensee's " Spent Fuel Storage Facility Modifi- {
cation Safety Analysis Report" which was submitted as an enclosure to the. I June 12, 19P,7 application. During its review of the application, the staff re-quested additional information from the licensee; the additional information was provided by letters dated September 8, 1987, October 20, 1987 (three letters) December 21, 1987 December 22, 1987, and December 73, 1987 (three letters 1 In addition, the staff met with the licensee on a number of occa-sions as reported in meeting minutes dated September 11, 1087, October 21, 1987 December 4, 19P7, and December 9, 1987 Ry latter dated January P9, 1988, the licensee submitted Revision 1 to the " Spent Fuel Storage Facility Modifica-tion Safety Analysis Report." The purpose of the revision was to incorporate changes to certain sections resulting from the FPAL and NRC correspondence and the meetings with the staff. The additional submittals supplemented and clarified the amendment request and did not alter the action noticed in the Federal Register or affect the staff's initial determination concerning the amendment request (See Section 11).
This report was prepared by the staff of the Office of Nuclear Reactor Regula-tion. Technical assistance for the structural evaluation of. the spent fuel racks and pool was provided by the Brookhaven National Laboratory, Upton, New York. The principal contributors to this report are:
H. Ashar Structural and Geosciences Branch L. Kopp Reactor Systems. Branch G. DeGrassi Brookhaven National Laboratory (Consultant)
J. Minns Radiation Protection Branch
- 1. Spickler Radiation Protection Franch J. Ridgely Plant Systems Branch F. Witt Chemical Engineering Branch P. Wu Chemical Engineering Branch E. Tourigny Pro.iect Directorate II-2
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1.2 Sumary Description of Reracking The amendment would authorire the licensee to increase the spent fuel peel-storage capacity from 728 to 1706 fuel assemblies. The preposed expansion is ,
to be achieved by reracking t!,e spent fuel pool into two discrete regions. l New, high-density storage racks (free-standing) will be used. The existing- i storace racks (free-standira) will be removed, cleered of loose contamination,-
packaged and shipped off-site.
Region 1 of the spent fuel pool includes '4 modules (racksi having a total of 342 storage cells. The nominal center-to-center spacing is_10.1? inches. All ,
i cells can be utilized for storage and each cell car accept new fuel assemblies with enrichments up to 4.5 weight percent U-235 or spent fuel assenh11es'that.
have not achieved adeouste burnup for Region 2. Region 2 includes 13 modules (racks) having a totel of 1364 storage cells. The nominal center-to-canter specing is 8.86 inches. All cells can be utilized for storage and each cell can accept spent fuel assemblies with varicus initial enrichments that heve 1 accumulated minimum burnups. Each cell in each region is designed to accom-modate a single Cerbustion Engineering or Advanced Nuclost Fuels Corporation '
(fomerly Exxon) PWR fuel assembly or equivalent, from either St. Lucie Unit.
The hipb-density spent fuel storage rack cells are fabricated from 0.080 inch thick type 30AL stainless steel plates. In Region 1, strips"of Boraflex 1eutron absorber material are sandwiched between the cell walls end a stainless steel coverplate. In Region ?, the Boraflex strips are sandwiched between the 1 adjacent cell walls. The cells, which fom a medule, are welded to a base plate, and a top gird'e bar is welded to the top of the module.
The new racks are not doubled-tiered and all racks will sit on the spent fuel pool fi er.
t The amendment application does not involve rod consolidation.
The proposed expansion of the spent fuel pool storage capacity to 1706 fuel assemblies should provide adequate storage urtil the year 2008, assuming full core offload capability. In addition, the expansion should he adequate until 1
a federal repository is available for spent fuel. I I
?. CP.ITICALITY CONSIDERATIONS 2.1 Criticality Anelysis The calculation of the effective multiplication factor, k CASMO-2E two-dimensional multigroup transport theory compcode. 6Nr, makes In addi-use of the tion, for independent verification, criticality calculations were also perfomed with the KEND-IV Ponte Carlo code, as well es the EPRI-CELL and NULIF codes. These independent verification calculations substantiate the CASMO-?E calculations and resulted.in a calculational bias of 0.0013 and a 95/95 probability / confidence uncertainty cf 0.0018.
In order to talculate the criterion for accepteble burnup for storage in Region 2, calculations were made for fuel of several different initial enrich-ments. At each enrichment, a limiting reactivity value, which included an additionel factor for uncertainty in the burnup analysis, was established.
Purnup values that yielded the limiting reactivity values were then detemined
-for each enrichmert from which the acceptable burnup domain for storage in Region 2, as shown in proposed technical specification' Figure 5.6-1, was obtained. The staff finds this procedure acceptable.
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For the Recion 1 analysis, the total uncertainty is the statistical combination of the calculated bias uncertainty and manufacturinti and mechanical uncer-tairties due to variations in boron loading in the Beraflex absorber sheets.
Boraflex width tolerance, Boraflex thickness, inner stainless steel storage box dimension, flux trap water gap thickness, steinless steel thickness, fuel enrichment and density, end fuel pin pitch. Other uncertainties due to tem-perature variations and eccentric positioning of the fuel assembly in the storage rack are ercounted for by assuming worst-case conditions; i.e., condi-tirns which result in the highest calculated reactivity.
In the Region 2 analysis, the same uncertainties are considered, except there is no water opp and, hence, no cap thickness uncertainty. In addition, an uncertainty due to the burnup analysis is estimated and treeted as an additive terri in determining the burrup versus enrichment limiting reactivity values in Fioure 5.6-1, rather than being combined statistically with the other l uncertair. ties.
The staff concludes that the appropriate uncertainties have been considered and have been calculated in an acceptable manner. In addition, these uncertainties were determined with at least a 95% probability and 95% confidence level, thereby meetinn the NRC requirements, end are acceptable.
For Region 1, the rack multiplication factor is calculated to be 0.9409, including uncertainties at the 95/95 prohebility/ confidence level, where fuel having en enrichment of 4.5 weight percent U-235 is stored therein. Fuel cf l either the Combustion Engineering (CE) or Advanced Puclear Fuels (ANF) type l
from St. Lucie Urit 3 or linit 9 may be stored.
For Region 2 the rack multiplication factor is calculated to be 0.9435 for the most reactive irradiated fuel permitted to be stored in the racks; i.e.,
fuel with the minimum burnup permitted for each initial enrichment as shown in Figure 5.6-1. The design will eccept #uel of 4.5 weight percent U-235 initial enrichment burned to 36.5 MWD /kgU of either the CE or ANF type from Units 1 and 2.
Therefore, the results of the criticality analyses meet the staff's acceptance criterion of k no greeter than 0.95, including all uncertainties at-the 95/95 probability /co8Ndencelevel.
Most abnomal storage conditions will not result in an increase in .the k of the racks. For example, loss of a cooling system will result in an incrINe in pool temperature, but this causes a decrease in the k,ff value.
It is possible to postulate events, such as an inadvertent misplacement of a fresh fuel assembly either into a ee gion ? storage cell or outside.and adjacent to a rack redule, which could lead to an increase in pool reactivity. However, for such events, credit may be taken for the Technical Specifications minimum requirement of 1720 ppm of boron in the pool water. The reduction in the l k value caused by the boron (approximately 0.24) more than offsets the
! rINtivityadditioncausedbycredibleaccidents.
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4 2.2 Technical Specifications Changes The following Technical Specifications (TS) changes have been proposed as a result of the replacenert of the existino spent fuel pool racks at Unit 1. The staff finds these channes acceptable.
- 1. TS 5.6.1.a.1 is revised to correrpond to the Standard Technical Specifications for Combustion Engineering PWRs (NUREG-0212, Rev. 2).
2.
TS 5.6.1.a.? is revised to show the nominal center-to-center spacing for the new storage racks.
- 3. TS 5.6.1.a.3 is edited to discuss the boron concentration in the pool water only.
- a. TS 5.6.1.a.4 is added to indicate the presence of Boraflex in the storage cells.
- 5. TS 5.6.1.b and accompanying fioure 5.6-1 are added to r.how the increased spent fuel enrichment permitted in the pool.
- 6. TS 5.6.1.c is editorially charned from "b" to "c".
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TS 5.6.3 is charged to show the capacity of the high-capacity spent fuel storage racks.
2.3 Conclusions Based on the review described above, the staff finc's the criticality aspects of the design of the St. Lucie Unit I spent fuel racks to be acceptable. The staff concludes the.t CE or ANF fuel from Unit 1 or Unit 2 may be safely stnred in Region 1. provided that the enrichment does not exceed 4.5 weight percent U-235.
Any of these fuel types may also be stored in Region 2 provided they neet the burnup and enrichment limits specified in Figure 5.6-1 of the St. Lucie Unit 1 TS.
3.
MATERIAL COMPATAPILITY AND CHEMICAL STAPILITY The staff reviewed the compatibility and chemical stability of the high density spent fuel storage rack materials wetted by the pon1 water. The proposed racks are fabricated from ASME SA-240-3036 austenitic stainless steel sheet and plate material, SA-331-CF3 casting material and SA-564-630 precipitation-hardened stainless steel (to 1100*F) for supports only. The weld filler materie.1 utili7ed in body welds is ASMF SFA-5.9, classification ER 308L. The neutron absorger material is Boraflex with a minimum B-10 areal degsity of 0.0238 gm/cm for the 342 Region I storage cells and 0.0098 gm/cm' for the 136a Region 2 storage cells. Boraflex is a silicone-based polymer containing fine particles of boron carbide in a homogeneous, stable natrix.
The annulus spaces that contain the Rorafley. in the high density racks are vented to the spent fuel pool. Verting of the annuli will allow gas generated by the chemicel and radiolytic decomposition of the silicone polymer binder, when exposed to the themal and radiation environment, to escape. This will prevent pressure buildup and possible bulging or swelling of the stainless steel e.bsorber sheathino.
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4 The austenitic stainless steel (304Li used in the spent fuel storage racks is not susceptible to stress corrosion cracking and thus, corrosion in the spent fuel storage pool environment should be of little significance during the 11'e of the plant. .The spent fuel pool water is processed by filtration ard demineralization to maintain water purity and clarity. Dissimilar metal-contact corrosion (galvanic attack between the stainless steel rack assemblies and Zircal.oy in the fuel assemblies) should rot be significant because the materials are protected by highly passivating oxide films and are, there'ere, at similar galvanic potentials.
Qualification tests have shown that Boraflex~ does not possess 1eachable
' halogens that could be released into the spent fuel pool water in the presence of radiation. Similar conclusions have been made- regardino the leachire of boror from the Boraflex.
Although Boraflex bas undergere extensive qualification testino to study the-effects of gama irradiation in various environments and to verify its structural integrity and suitability as a neutron absorbing material, recent annrialies have been identified in the Quad Cities and Point Beach high density spent fuel racks due to Boraflex shrinkage caused by irradiation.: To preclude similar problems at St. Lucie Unit No. 1 the specification for the handling and installation of- the Boraflex recuires that it not be installed in a stretched condition. The use of adhesives in the attachment of the Poraflex to the rack cell is not pemitted. In addition the manufacturing process avoids techniques that could pinch the Boreflex. Therefore, the St. Lucie Plant Unit '
No. I rack design and fabrication process allows expected shrinkage without-cre.cking and gap formation. Furthermore, the spent fuel rack design recuires that oversized Boraflex sheets be used to provide a four-inch shrinkroe allow-ance and that allowances for the elastic rebound of the Boraflex naterial be-made before installation should the material be stretched during shipment or handling.
To provide added assurance for detection of degradation of the Boraflex, the licensee has comitted to conduct a long-tem and accelerated surveillance test program. Each surveillance coupon-(5 inches by 15 inches) containing Peraflex of a thickness similar to that used in the racks, is encased in a stainless steel jacket, the alloy of which is identi'ied to that used in the racks. The coupon jacket permitt wetting and ventino of the specimen to the spent fuel pool water similar to that of the rack. The long-tem coupon examination frequency occurs after irradiation times of 90 days,1@ days,1 year, 5 years,10 years,15 years, 25 years and 35 years. The accelerated test coupon examination frecuency is after each discharge from the second to ninth discharge rack utilization. Acceptance criteria for cortinued use are dimensional changes of no more than 2.5% from the original, hardness not less than 90% of the original, and minimal areal density of boron not less than the original.
The staff has reviewed the proposed surveillance program for monitorire the Beraflex in the St. Lucie Plant Unit No. I spent fuel storage' cool and concludes that the procram can reveal deterioration that may. lead to loss of neutron absorbing capability during the life of the spent fuel racks. In the unlikely event o' Boraflex deterioration, the monitoring program will detect such deterioration and the licenset will have sufficient time to take corrective action. In the event o# unanticipated degraded coupons, the storage racks will be inspected and then NPC will.be informed if the inspection reveals Boraflex decred: tion in the storage m ks.
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Rased on the above discussion, the staf# concludes that corrosion of the high density racks due to the spent fuel pool environment should be of little significance during the life of the facility. The staff finds that implementation of the proposed surveillance program and the selection of appropriate materials of construction by the licensee, meet the requirements of 10 CFR 50, Appendix A, General Design Criterion (GDC) 61 (regarding the capability to pemit appropriate periodic irspection and testing of components) and GDC 6' (regarding preventing criticality by maintaining structural integrity of components and of boron absorber material) and are, therefore, acceptable.
- 4. STRUCTURAL DESIGN This evaluation addresses the adecuacy of the structural aspects of the pro-posed amendment. The Brookhaven National Laboratory (RNL) assisted the staff in reviewing various analyses and responses submitted by the licensee.
Attached is the technical evaluation repnrt (TER) developed by PNL (Appendix A).
The staff accepts the findines and conclusions of the TED by incorporating the TER as a part of thir safety evaluation.
The spent fuel storage pool is located in the fuel handling building, which is a Seismic Category I structure. The pool is 33 feet by 37 feet in plan and is 40 feet, 6 inches deep. The reinforced concrete foundation mat, which is g I) 9'-6" thick except in the spent fuel cask storage area wherecit is 6'-0" thick, q provides flone space for the spent fuel racks. The reinforced concrete walls enclosino the spent fuel storage area vary in thickness from ?'-0" to 5'-0".
The real walls are lined with 3/16 inch stainless steel plates and the pool
'loor is lined with i inch stainless steel plates.
The proposed high-density storage racks consist of individual cells with B.65 inches by P.65 inches square cross-section, each of which would accommodate a single combustion Engineering or ANF PWR fuel assembly. A total of 1706 cells are arranged in 17 distinct rack modules of various arrays of fuel cells.
Fech rack nodule is equipped with 3/4 inch thick by 31 inch high girdle bars at the upper end designed to withstand the impact loads under the postulated seismic conditions. The rack modules are free-standing, and they make surface contact at the girdle bar locations providing a nominal 11 inch gap between adjacent module cell walls.
The primary areas of review associated with the proposed application are focussed towards assuring the structural integrity of the fuel, fuel cells, rack modules, and the spent fuel pool floor and walls under the pnstulated (Appendix D of SRP 3.B.4) loads and fuel handling accidents. The major areas of concern and their resolutions are outlined in the following paragraphs.
The fuel handling building analysis and design had been reviewed and accepted during the initial licensing stages. Sinct the effect of the additional fuel rack load on the pool floor is limited to the mat in the pool area, the licensee reanained the lower portion of the walls, the pool floor, and the effects on the underlying soil. The design-analysis results satisfy the acceptance criteria.
Details of the enalysis, design and adequacy of the pool, pool liner and its anchorages are discussed in Section 4.5 of Apperdix A.
The plant is located on potentially liquefiable soil. During the operating license review, the licensee provided sufficient data and analyses to demon-strate that the factor of safety against liovefaction under a Safe Shutdown
L ! 7 Earthquake (SSF) is more than 2. Durino this review, the staff expressed a concern about the effects of added weight on liquefaction potential under the ;
postulated seismic condition (i.e., SSE).- Rased on the research work published by Seed Idriss and other researchers in the publication " Liquefaction of Soils During Earthquake (National Academy Press,1985)," where it was shown that soils 1 sub.iected to static shear stresses prior to an earthquake have hioher resistance ;
to liquefaction, the licensee concluded that the added weight would maintain or ]
improve the resisterte to liquefaction. The licensee's report also indicated
- that the maximum bearing pressure on the soil under the combined effects of 1
dead load (including the added fuel weight) and an SSE is less than the allowable bearing capacity of the soil. The staff accepts the licensee's conclusion and considers the concern as resolved.
The adequacy of considering a single rack model in the seismic analysis was nuestioned. The seismic motion of a single rack is coupled to the motion of adjacent racks throuch impact forces and fluid coupling forces. .The single rack model constrains the motion of a rack within an imaginary boundary, i
Maximum displacements cannot exceed one-half the gap to the ad,4acent racks.
For sufficiently strong seismic motion, sliding and tilting motions of the racks could be larger than these predicted by a constrained single rack model resulting in higher impact velecities than would be predicted by a single rack-model. Under worst conditions, rows of racks could slide together in one direction and pile up against a pool wall. The additional mass of racks l involved in the impact could generate larcer loads on the racks and the pool l walls. This concern may be more critical for the pool walls, since they'are I not designed to acconodate seisnic impact loads from the' fuel racks. To resolve the concern, the licensee perfomed a two-dimensional multiple rack analysis of a single row of fuel racks to determine the extent of displacement under an SSE. The limited multiple rack analysis indicated that the correspond-ing displacements are small (less than or equal to 1/2 inch) compared to the I
minimum clearance provided (3 1/2 inches) between the edge racks.and the walls. j A detailed discussion of the other concerns, the. comparative results of various analyses and conclusions thereof are provided in Section 4 ? of Appendix A.
Based on its evaluation of the licensee's submittal, the supplementary informa-tion provided by the licensee. discussions with the licensee at meetings, and infomation audited by the staff and its consultant, the staf# concludes that the licensee's structural analyses of the spent fuel rack modules and the 4
spent fuel pool are in compliarce with the acceptance criteria set forth in the FSAR and consistent with the current licensing practice and, therefore, are acceptable.
- 5. SPENT FUEL POOL COOLING AT LOAD HANDLING 5.1 Decay Heat Generation Rate In the June 1?, 1987 submittal, the licensee stated that the calculation of the decay heat generating rate was in accordance with the guidelines of Standard Review Plan (SPP) Section 9.1.3 and Branch Technical Position ASB 9-2. For the normal moximum heat load condition, the licensee assumed the pool was filled with one-third core refuelings every 18 months # rom the St. Lucie Unit I reactor and calculated a heat generation rate of 16.42 MBTU/Fr. The abnormal maximum beat load condition had the same assumptions as the norma 1' maximum heat
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, e load condition, except that the ?)7 empty feel storage locations were filled with a full core offload. For this condition, the licensee calculated a heat generation rate of 33.70 MBTU/Pr at 169 hours0.00196 days <br />0.0469 hours <br />2.794312e-4 weeks <br />6.43045e-5 months <br /> into the refueling outage in lieu of the 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> identified.in the SRP.
The staff perfonned an independent calculation of the heat generation rate in accordance with the guidelines in SRP Section 9.1.3 and Branch Technical Position ASB 9-? assuming the anticipated 18-month operatinc cycle. The staff calculated a norral maximum heat generation rate of 16.84 MBTU/Pr and an abnormal maximum heat generation rate of 33.56 MBTU/Pr at 169 hours0.00196 days <br />0.0469 hours <br />2.794312e-4 weeks <br />6.43045e-5 months <br /> into the refueling outage and 34.96 PRTU/Hr at 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br />. The licensee's calculation of the normal maximum heat load is not significantly different from the staff's calculated value, and thus, the staff concludes that the licensee has properly calculated the heat generation rate in accordance with the guidelines of the SRP.
5.? Spent Fuel Pool Coolino System The spent fuel pool cooling system (SFPCSi consists of one train of equipment, including two 3560 opm centrifugal pumps and one tube-and-shell heat exchanger with a heat transfer capability of approximately 34 MBTU/Hr. as indicated in the FSAR. After water from the spent fuel pool is cooled by the heat exchanger, it is puri'ied by the spent fuel pool cleanup system. Neither the SFPCS nor the cleanup system are seismic Category I. In the event of a less of SFPCS, a seismic Category I salt water makeup supply'to the spent fuel pool is availabic from the intake cooling water intertie.
The SPFCS heat exchanger is a low pressure, low temperature component.
Maintenance of the Feat exchanger, such as tube cleaning or plugging, can be scheduled to be performed when the heat being generated by the spent fuel is low, such as immediately prior to entering a refueling outage when the time until the spent fuel rool reaches boiling will be significantly longer than the 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> calculated for the normal maximum heat load case. Thus, the staff concludes that having a single heat exchanger is acceptable.
5.2.1 Heat Removal Capability Under the normal maximum heat load conditions (16.84 METU/Hr) using one SFPCS pump, the SFPCS heat exchanger will maintain the spent fuel pool water temperature below 134*F, which is less than the la0"F temperature guideline specified in SRP Section 9.1.3. For the abnormal maximum heat load condition (33.56 MBTU/Pr) using one SFPCS pump, the heat exchanger will maintain the spent fuel pool water temperature below 167"F, which is well below boiling.
Thus, the staff finds that the SFPCS meets the requirements of GDC 44, " Cooling Water" with respect to providing adequate pool cooline under normal heat load conditions following a single failure.
5.2.2 Protection Against Natural Phenomena The SFP cooling capability was reviewed with respect to the re GDC 2. " Design Bases for Protection Against Natural Phenomena,quirements
" which includes of protection against earthquakes, hurricanes, tornadoes, or other natural events.
The SFPCS is not seismic Cateanrv I. Under such circumstances, SRP Section 9.1.3 identifies an alternative method for cooling of spent fuel following an earthquake.
o, Specifically, the SRP discusses use of a seismic Category I spent fuel pool makeup water capability and a seismic Category I ventilation system to process potential radiological releases to the pool building resulting from pool boiling.
F.2.2.1 Makeup Water The St. Lucie t' nit 1 FSAR identifics several makeup water sources. The refueling water storage tank and the primary water tank are seistaic Category .I sources of water. In addition. salt water can be provided to the spent fuel i
pool from the intake structure via the seismic Category I intake cooling water system at the rate of 150 gpm.
5.2.2.2 Buildirg Ventilation I The licensee has not taken credit for any ventilation system to mitigate the offsite releases due to boiling of the spent fuel pool water. The licensee has provided the results of the offsite dose consequence analysis in their submittal da+ed December 23, 1987, which indicates that the maximum calcula I edult absorbed thyroid dose is 0.I'/3 rem, the whole body dose is 1.82 x 10'ged I rem, and the skin dose is ?.18 x ig 5 rem at the low population rone'. Since the thyroid dose is less than 1" o# 10 CFP 100 limits (300 rem) and the whole body and skin doses are insignificant, the staff concludes that not using any ventilation system to mitipate the release of radioactivity when the water in the spent fuel pool is boiling meets the requirements of GDC 60, " Control of Releases of Radioactive Materials to the Environment."
5.2.? Loss of Conling In the event that all SFP cooling is lost, the spent fuel pool temperature will increase urtil boiling is achieved. The licensee has estimated the time from the loss of pool cooling until the pool water boils for the normal maximum heat load condition to be approximately 16.79 hours9.143519e-4 days <br />0.0219 hours <br />1.306217e-4 weeks <br />3.00595e-5 months <br /> and for the abnormal heat load condition to be approximately 7.47 hours5.439815e-4 days <br />0.0131 hours <br />7.771164e-5 weeks <br />1.78835e-5 months <br />. The calculated boil-off rates are estimated to be 33.9 gpm and 60.5 gpm, respectively. The staff finds that the intake cooling water system capability is in excess of those estimated boil-off rates and there is reasonable time to take action to provide SFP
- makeup. The staff further concludes that the makeup water system, without any ventilation system to mitigate the release of radioactive materials, meets the requirements of GDC 2 " Design Bases for Protection Against Natural Phenomena,"
for ensuring adequate spent fuel pool cooling and prevention of unacceptable radiological releases following an earthquake. ,
5.3 Heavy load Handling The new spent #uel storaga racks weich more than a fuel assembly and its handling tool. Thus, the spent #uel storage racks are consic'ered to be heavy .
loads. The cask handling crane will be used tn move the new storage racks into the fuel handling building and into the cask area within the spent fuel pool, and to remove the existing storage racks from the cask area to the cask decontamination area outside of the fuel handling building. The movement of the cask handling crare is physically limited by the opening in the side wall and the roof of the fuel handling building. This opening is nomally closed by a L-shaped door. The cask handlirq crane, due to this limitation, tannot carry heavy loads over spent fuel. In the previous review of compliance with 1
the cuidelines of NUREG-0612, " Control of Heavy Loads at Nuclear Power Pier.ts," the staff concluded in a Safety Evaluation Peport dated March 4,1985, that the cask handlirp crane met the guidelines of NUREG-0612.
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10 Due to the physical limitations of the lifting capability of the cask handling crane, a new, temporary crane will be installed in the fuel hendling building.
The temporary crane will be used to move the new storage racks to their appropriate positions in the pool frem the cask area and the existing racks from their present locations to the cask area. The temporary crane will also be used as a platform for re-rigging of the new and existing storage racks on the cask handling crane.
In the December 22, 1987 submittal, the licensee provided installation 6 tails for the temporary crane. It will be brought into the fuel handling building l
as five separate pieces (two truck units, two girder piecer. and the hoist unit). The information provided in the December 23, 1987 submittal.
i-demonstrated that no piece of the tersporary crane will be carried over spent fuel or over racks containing spent fuel. To provide assurance that no part.
i of the temporary crane will be carried over spent fuel, the licensee committed to nerk the fuel hendling machine over the nearest spent fuel assembly as a physical barrier to movement.
In the December 22, 1987 submittal, the licensee provided the results of evaluations of three potential load drop accidents: (1) the temporaryy crane I dropping a spent fuel storage rack in the spent fuel pool; (21 the cask handling crane drepping)a spent fuel storage rack into the cask aren of the spert fuel pool; and (3 .the cesk handling crane dropping a spent fuel storage rack onto the temporary crane. In all cases, the radiological consequences of the load drop are less than tnat for the cask drop accident identified in the FSAR. The rack drop accidents involving the cask handling crane would require the L-shaped door in the fuel handling building to be open. Thus, no credit was taken for retention of the radioactivity by the building. The licensee !
committed to remove the temporary crane and to perform a load test on it if i a heavy load is dropped on+o it. When the cask handling crane is moving a rack into or out of the #uel building, the tenpnrary crane will be located nex+ to the north wall of the spent fuel pool. Three new racks will be placed 1
I along the north wt.11 beside the cask area. The three new racks will, as part '
of the fuel shuffling program, contain spent fuel; however, no spent fuel will l
be placed under the temporary crane parkinc location. Thus, if the temporary crane were to fail as the result of a load drop from the cask handling crane, J ro spent 'uel would be impacted.
The temporary crane is not single-failure proof. The licensee stated in the i December 22, 1987, submittal that the tafety factor!. for all lead-hearing components of the temporary crane meet or exceed the safety factors identified in NUREG-0612. All welds will be inspected using either liquid penetrant or magnetic particle methods. The crane hoist will be load-tested to 150% of the rated load. With these measures, the rtaff finds that the temporary crane meets the guidelines of NUREG-0612.
The licensee provided drawings of the special lifting devices for the new and existing spent fuel storage racks. These drawings demonstrate that the lifting devices are single-failure proof and thus meet the guidelines of NUREG-0610.
I In the December 23, 1987 submittal, the licensee provided drawings that show the order in which the existing racks will be removed and the new racks wil' be installed. These drawings also identify those storage locations that will i
contain spent #uel and verify that the racks will not be transported over spent l fuel or over racks containing spent fuel.
l
r 11 From the above review, the staff finds that handling heavy loads during the reracking procedures is in accordance with the guidelines of NUREG-0612 and, therefore, the requirements of GDC 61, " Fuel Storage and Handling and Radio-activity Control," are met as they relate to proper load handling to ensure against an unacceptable release of radioactivity er a criticality accident as a result of a postulated load drop.
5.4 Light Lead Handlinn A light load is defined as any load that weighs less than a fuel assembly and. .
its handling tool. In a submittal dated December 22, 1987, the licensee provided the results of an evaluation of light load drops for St. Lucie Unit 1.
The licensee reviewed the light loed analysis that was performed for St. Lucie Unit 2 at the time of licensing, which was approved by the staff in NUREG-0843, Supplement 3, dated April 1983. The licensee verified that those light loads evaluated for Unit 2 are applicable for Unit 1. From that review, the licensee rencluded that the consequences to spent fuel from a. light load drop would be less than that for a design basis fuel handling accident, namely the failure o' all fuel pins in one fuel assembly.
5.5 Ocnclusions Based on the above, the staff concludes that the proposed expansion of the St.
Lucie Unit I spent fuel pool complies with the requirements 'of General Design Criteria ?, 44, 60, and 61 and the guidelines of NUREG-0610, and Regulatory Guide 8.8 with respect to the capability to provide adequate spent fuel pool cooling, safe loading hardling, and to maintain offsite and onsite radiological releases withir acceptable limits. The staff, therefore, finds the proposed expansion to be acceptable.
- 6. SPENT FUEL POOL CLEANUP SYSTEM The spent fuel pon1 (SFP) cleanup or purification system maintains pool water clarity and purity. It consists of a 150 gpm purification oump, a cartridge filter, a mixed bed deminerali7er, and the required piping, valves, and instrumentation. The pump draws water from the SFP and discharges through the cartridge filter and the demineralized. The water is then returned to the pool. It is possible to operate the system with either the filter or demineralized bypassed.
radioactivity and impurity 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 St. Lucie Plant, Unit No. I will not 'ncrease the refueling frequency and fraction of the core repleced after each fuel cycle.
Therefore, the frequency of operating the spent fuel pool cleanup system is not expected to increase. Similarly, the chemical and radionuclides composi-tion of the spent fuel pool water 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 'uel stored in pools after the fuel has cooled for several months. Thus, the increased quantity of spent fuel to be stored in the St. Lucie Plant, Unit No. I fuel pool will not increase significantly the total fission product activity in the spent fuel pool water during the operation of the pool.
( 12 The staff has-evaluated the information provided by the licensee. Based on this evaluation and its experience with other high-density spent fuel storage i facilities, including evaluation of operating data, the staff has determined !
that the proposed reracking of the spent fuel pool at St. Lucie Plant. Unit l
flo. I will not adversely affect the performance capability or capacity o' the !
spent fuel poel cleanup system. The radioactivity and impurities in the pool '
water are not expected to increase as a result o' the reracking. Replacement of filters or demineralizers would offset any unanticipated increase of the radioactivity and impurity level of the water in the event of a reduction of the decontaminetton effectiveness.
On the basis of the above discussion, the spent fuel pool rerack is acceptable.
- 7. RADIATION PROTECTION AND ALARA CONSIDERATIONS I 1
The additional occupational radiation exposure associated with the actual rerackira o' the pool is estimated by the licensee to be less than 15 person-ren. { ;
In a letter dated October 20, 1987 FPL provided additional information I i
describing action to be taken during SFP modification. Some of the ALARA '
activities directed to the reduction of occupational radiation include: (a) vacuum cleanirg of SFP floors will be performed remotely from the surface; (b) maximutr water shielding to reduce dose rates to divers, if they are used ; (c) underwater radiation surveys; (d) calibrated alarmine dosimet'ers and personnel nonitoring dosimeters for divers, if they are used; (e) hydrolysing and cleaning of old spent fuel racks; (f) the use of remote operations for rack removal and replacement operations; and (g) SFP purification system augmented by urderwater vacuum system to maintain radioactive contamination ALARA and maintain SFP clarity.
The licensee has also provided a description of contained and airborne radioactivity sources related to the SFP water, which may become airborne as a result of failed fuel and evaporation. The staff has reviewed these source terms and finds them acceptable.
Rased on our review of the St. Lucie's submittals, we conclude that the pro,4ected activities and estinated person-rem doses for this pro,iect appeer reasonable. FPL intends to take ALARA considerations into account, and to implement reasonable dose-reducing activities. We conclude that FPL will be able to maintain individual occupational radiation exposures within the applicable limits of 10 CFR Part 20, and maintain doses ALARA, consistent with the guidelines of Reculatory Guide R 8, Therefore, the proposed radiation protection aspect of the SFP rerack is acceptable.
- 8. ACCIDENT ANALYSES The staff has reviewed the accidental fission product releases that could occur at the St. Lucie Unit I facility in con.iunction with the proposed rerackina of the SFP. The only potential releases that have not been previously aralyzed by the staff as part of the original SER are the potential offsite consequences of the dropping of a cask into the reracked full SFP and release of fission products from the spent fuel resulting from the boiling of i
the pool water. The consequences of these accidents have been reviewed by the licensee and the staff.
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. With regard to cask drop accident, the most conservative case occurs with the cask being dropped into the SFP 1490 hours0.0172 days <br />0.414 hours <br />0.00246 weeks <br />5.66945e-4 months <br /> after the following fuel cycle history:
l One-third of a core is placed in the.SFP each year during refueling for the next 20 years. Following the 21st year of operation, the entire core is removed from the reactor and placed into the pool, which fills the pool. The number of essemblies damaged is equal to a full-core offload plus the remainder of the pool filled with discharged assemblies from previous refuelings..
The 1490 hour0.0172 days <br />0.414 hours <br />0.00246 weeks <br />5.66945e-4 months <br /> figure is the earliest that a cask could be moved into the SFP i area with a full pool based on the TS. It is assumed that all the spent fuel l in the pool (8 full coresi is damaged with the release of 10% of the noble i gases (except Kr-85) and the iodines and 30% of the Kr-P5 to the pool water, )
I and with 99% of the released iodine remaining in the pool water. The remainder of the released fissinn products is released to the environment. The resulting dose to an individual at the exclusion zone boundary would be 21 thyroid-rem and less than 0.1 rem to the whole body.- !
In their December 23, 1987 submittal, the licensee presented a conservative enelysis of the radiological consequences of boiling of the SFP water. The sta'f has reviewed the licensee's analysis. The staff analys.is differs onl ,
slightly (staff utilization of slightly more conservative dilution factors)y.
The potential doses at the exclusion boundary (0.97 miles ) and low population zone boundary (1.0 miles) are approximately 0.1 thyroid-rem and 10'g rem to the whole body.
The potential doses resulting from the cask drop and spent fuel pool water boiling accidents are well below the allowable 10 CFR 100 guidelines doses of 300 rem to thyroid and 25 rem to the whole body. Therefore, the accident analysis aspect of the SFP rereck is acceptable.
- 9. RADIOACTIVE WASTE TREATMENT The plant contains radioactive waste treatment systems designed to collect and process the gaseous, liauid, and solid waste that might contain radioactive material. The radioactive waste treatment systems are evaluated in the Final i
Environmental Statement (FES) dated June 1973 (US NRC 1973). There will be no change in the waste treatment systems described in the FES because of the proposed SFP rerack.
- 10. SIGNIFICANT HAZARDS CONSIDERATION COMMENTS The licensee's request for amendment was noticed on August 31, 1987 (59 FR
?78521, followed by a biweekly notice on September 23, 1987 (52 FR 35813). By letter dated September 30, 1987, Mr. Campbell Rich requested a public hearing.
An Atomic Safety and Licensing Board was established on 0:tober 22, 1087 to consider the reouest. In pleadings filed November 4 and 9, 1987, both the licensee and the NRC staff pointed out that the letter failed to meet the requirements of 10 CFR 2.714 and that, therefore, the request should be denied.
By Memorandum and Order of November 13, 1987, the Board directed the licensee and Mr. Rich to seek infonnal resolution of Mr. Rich's concerns and set January 15, 1982 as the deadline for filing an amended petition. Pr. Rich met with the licensee and subsequently filed an amended petition which proffered 16 contentions. The licensee and the staff responded to the contentions by
14
- r pleadings dated February 1,1988 and February 4,1988, respectively. The Licensing Board has not yet ruled on the contentions but has scheduled oral arguments on intervention and the contentions for March 79, 1988 (53 FR 5661).
The propcsed contentions and the staff comments are contained below. ;
Contention 1: "That the expansion of the spent fuel pool at St. Lucie, Unit No. 1 is a significant hazards consideration and requires that a public hearing be held before issuance of the license amerdnents fsic)."
The staff may issue and make imediately effective an amendment to an operating license pursuant to the Comission's regulations. A public /
hearing need not be held before issuance of the amendment. The staf*
hes followed the Commission's regulations in the ifcensing action.
A Final No Significant Harards Consideration Determination is included in this safety evaluation.
Contention ? " Expansion of the spent fuel pool at the St. Lucie facility, Unit No. I constitutes a major Federal action and requires 1
that the Conmission prepare an environmental im:iact statement in accordance with the National Environmer.tal Policy Act of 1969 (MEPA) and 10 CFR Part 51."
1 i
The staf' prepared an Environmental Assessment related to this licensing action. Based en the Assessment, the sta#f made a finding of no signif-icant impact pursuant to 10 CFR 51.3P (53 FR 7065). Therefore, no envi-ronmental impact statement need be prepared.
Contention 3: "That the calculation of radiological consequences resulting ' rom a cask drop accident are (sic) not conservative, and the radiation releases in such an accident will no [ sic) be ALARA, and will not meet with the 10 CFR Part 100 criteria."
As Low As Is Reasonably Achievable (ALARA1 applies to normal plant operations. ALARA is not a consideration in accident analysis consequences determination. The licensee addressed the cask drop accident in the licensing submittal. The staff reviewed the licensee's analysis (including input assumptions) and agrees with the licensee's conclusions. Section 8 of this evaluation contains the details of the staff's independent evaluation. l
! l Contention 4: "That the consequences of a cask drop accident or an l accident similar in nature and effect are greatly increased due to the l presence of a large crane to be built inside the spent fuel pool building in order to facilitate the reracking."
The large crane that will be " built" in the fuel handling building is considered a tenporary construction crane. The crane will be used to remove the existing racks and install the new racks. The crane will be in the fuel handling building for only a few months. Once the rerack modification is completed, the crane will be removed from the building. The spent fuel cask and the temporary construction crane will never be in the building at the same time. Thus, there is no possible accident as a result of the temporary construction crane and cask being in the building at the same time.
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15 The contention also refers to "an accident similar in nature." This was also evaluated by the staff as follows. The staff evaluated the use of the temporary construction crane to be used during the rerack modifice-tion. The staff postulated various load drop accidents, such as the drop I of a rack during the rerack modification, in spite of the fact that no i heavy load will be carried over spent fuel or. over any rack which contains spent fuel. The staff concluded that in all cases, the radiological '
consequences of the load drop accident are less than that for the cask drop eccident evaluated. Section 5.3 of this evaluation contains the details of the staff's evaluation.
Contention 5: "That FPAL has not provided a site specific radiological analysis of a spent fuel boiling event that proves that off-site dose limits and personal isic? exposure limitt will not be exceeded in allowin the pool to boil with makeup water from only seismic Category 1 sources."g The licensee and the staff used the Standard Review Plan (SRP) as guidance in the spent fuel pool cooling analysis. The SRP specifies that the pool water temperature should not exceed 140*F (a single active. failure to the system is assumed 1 under normal refueling condi-tions and not exceed boiling (a single active failure need not be considered 1 under full core discharge conditions. Independent calcula-tions performed by the staff and licensee concluded that the SRP acceptance criteria are met. Nevertheless, the licensee and the staff, as a further precaution, postulated the site-specific pool boiling event and evaluated the radiological consequences and makeup water sources. The staff concluded that the radiological consequences were well within the guidelines of 10 CFR Part 100 and seismic Category I makeup water sources were available to supply makeup water to the pool.
Section 5.P.1 contains the staff's evaluation of heat removal capabil-ity. Section 5.2.2 contains the staff's evaluation of makeup water sources. Section 8 contains the staff's evaluation of the radiological consequences as a result of pool boiling. ~
Contention 6: "The Licensee and Staff have not adequately considered or analyzed materials deterioration or failure in materials integrity resulting from the increased generation of heat and radioactivity as as Isic] result of increased capacity and long-tem storage in the spent fuel pool."
The staff reviewed materials integrity of all materials used in the spent fuel pool. The corrosion of the high density racks due to the '
spent fuel pool environment should be of little significance during the life of the facility. The long-term durability of Boraflex is ensured by the proposed surveillance program. Section 3 of this document contains the staff evaluetion to support these conclusions.
Contention 7: "That there is no assurance that the health and safety of the workers will be protected during spent fuel pool expansion, ard that the NRC estimates of between 80-130 rem / person will not meet ALARA requirements, in particular, those in 10 CFR Part 20."
The staff evaluated the occupational radiation doses to workers involved with reracking the spent fuel pool. The staff concludes that the occupational radiation exposure is less than 15 person-rem, within
16 the applicable limits to 10 CFR Part 20. and is ALARA. Section 7 contains the staff's evaluation of doses to workers. In addition, Section 3.2 of the Environmental Assessment also addresses doses to workers.
Contention 8: "That the hinb density design of the fuel storage racks will cause higher heat loads and increases in water temperature which could cause a loss-of-cooling accident and/or challenge the reliability and testability of the systems designed for decay heat and other residual heat removal, which could, in turn, cause a major release of radioactivity into the environment."
The staff's comments ere the same as those in response to Contention F.
Contention 9: "That the cooling system will be unable to accommodate the increased heat load in the pool resulting from the high-density storage system and a full core discharge in the event of a sirgle failure of any of the pumps or the electrical power supply to the pumps on the shell side of the coolinn system and/or in the case of a single failure of the electrical power supply to the pumps on the pool side of the spent pool cooling system. This inability will, therefore, create a greater potential for an accidental release of radioactivity into the environment."
The staff's coments are the same as those in response tn Contention 5.
Contention 10: "That in calculating tirse to boil after loss of cooling after completion of full core discharge with the preserce of the proposed 1706 assemblies, Fp4L utilized a different set of assumptions than in determining the original fioures for time to boil as indicated in the Finel Safety Analysis Report for the St. Lucie plant Unit No.1.
(9.1-49. Table 0.1-3)."
The staff's comments are the same as those in response to Contention 5.
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- Contention 11: "That the proposed use of high-density storage racks designed and fabricated by the Joseph Oats Corporation is utilization.
of an essentially new and unproven technology."
The staff does not agree that the proposed use of high density storage racks designed and fabricated by Joseph Oats Corporation is utilization of an essentially new and unproven technology. A large number of high-density storage racks, which utilize Boraflex, have been fabricated by J. Oats for other utilities. These racks have been installed and are currently storing spent fuel. Similar statements can be made of other fabricators of spent fuel storage racks. Section 3 of this i
evaluation contains the staff's rack materials evaluation.
Contention IC: "That the presence of depraded Boraflex specimens or absorber sheets on the floor of the pool will pose an increased hazard I
I in promoting the propagation of cladding fire to low power bundles and thus promote a far larger spent fuel pool accident."
Boraflex specimens or absorber sheets will not be located on the floor of the pool. The Boraflex will be installed as part of the racks, within the rack structure. See Section 3.
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, 17 Contention 13: "The Licensee has cet analyzed the effect that a hurricane or tornado could have on the spent fuel storage facility or its contents, and that the SEP reglects certain accidents that could be caused by such natural disasters."
The staff evaluated the-fuel handling building structure under natural phenomena conditions when the unit was originally licensed. The staff's SER evaluation. dated November 8,1974, SER, Supplement 1 dated ~
May 9,1975. and SER, Supplement ? dated March 1,1976, served as the licensing basis to approve the St. Lucie 1 safety-related structures, including the fuel handling buildire, and considered natural phenomena.
The rerack itself will not involve any changes to the fuel handling building / spent fuel pool; thus, natural phenomena need not he reanalyzed as part of this review.
Contention 14: "That FP8L has not properly considered or evaluated the radiological consequences to the environment and surrounding,-
human population of an accident in the spent fuel pool."
Secticn 8 of this document contains the staff's evaluation of pnstulated accidents. The consequences of the postulated accidents are within the guidelines of 10 CFR Part 100. ,
Contention 15: "That the increase of the spent fuel' pool capacity, which includes ' fuel rods which have experienced fuel failure and fuel rods that are more highly enric'1ed, will cause the requirements. of ANSI-N16-1975 not to be met ant. will increase the probability that a criticality accident will occur in the spent fuel pool and will exceed 10 CFR Part 50, A 62 criterion."
The steff used the Standard Review Plan to evaluate the criticality aspects of the spent fuel pool rerack. The results showed that the rerack is acceptable from a criticality perspective. The staff's l criticality evaluation is contained in Sectier 2 of this document.
Contention 16: "That FPRL has not responded to the concerns as presented by the NRC by outlining a loading schedule for the spent
'uel pool detailing how the most recently discharged spent fuel will be isolated from other recently discharged fuel and/or a full core discharge in order to mitigate potential risks from fires in the i spent fuel pools fsic] resultino in releases of radioactivity into the environment in excess of the 10 CFR 100 criteria."
l' The staff did not express a concern in regard to a loading and storage configuration for discharged fuel in connection with this rerack appli-cation. The licensee proposed limiting the spent fuel assemblies having einimum burnup per proposed Technical Specification Figure 5.6-1. The I
sta'f finds the proposed controls for placement of spent fuel assemblies in Region 1 and Region ? acceptable, and concludes that no other loading and storage controls are necessary. See Section 2.0 of this document.
In addition, the staff has generally addressed the potential for cladding fires in Section 5 of the Environmental Assessment.
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- 11. FINAL NO SIGNIFICANT HAZARDS CONSIDERATION The licensee's request for amendment to the operating license for the St. Lucie Plant, Unit No. 1, including a proposed determination by the staff of no significant hazards consideration, was individually noticed in the Federal Renister on August 31, 1987, followed by e biweekly notice on September 23, i 1987. This is the staff's final deteminatien of no significant hazarris consideration. f '
The Comission's regulations in 10 CFR 50.92 include three standards used by the NRC staff to arrive at a determination that a reouest for amendment involves no significtrt hazards considerations. These regulations state that the Comission may make such a final determination if operation of a facility -
in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (21 create the possibility of a new nr different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.
The proposed spent fuel pool expansion amendment is similar to more than 100 earlier requests from other utilities for spent fuel pool expansinns. The me.fority of these requests have already been gre.nted by the NRC; others are under staff reviest. The knowledge and experience osined by the NRC staff in reviewing and evaluating these similar reouests were utilize'd in this j evaluation. The licensee's request does not use any new or unproven technology in either the analytical techniques necessary to support the expansion or in the construction process.
The staff has determined that the licensee's request for amendment to expand the spent fuel pool storage capacity for the St. Lucie Plant, Unit No. I by reracking to allow closer spacing of spent fuel assemblies does not i significantly increase the probability or consequences of accidents previously {
t evaluated; does not create new accidents not previously evaluated; and does not result in any significant reduction in the margins of safety with respect to criticality, cooling or structural considerations.
The following staff evaluation in relation to the three standards demonstrates that the proposed amendment for the SFP expansion does not involve a significant hazards consideration.
First Standard
" Involve a significant increase in the probability or consequences of an accident previously evaluated."
The following postulated accidents and events involving spent fuel storage have been identified and evaluated by the licensee. The staff likewise evaluated the same accidents and events.
- 1. A spent fuel assembly drop in the spent fuel pool.
- 2. Loss of spent fuel pool cooling system flow.
- 3. A seismic event.
19
- 4. A spent fuel cask drop.
- 5. A construction accident.
)
The probability of any of the first four accidents is not affected by the racks themselves; thus the modification cannot increase the probability of these accidents. As for the construction accident, the licensee will not carry any rack c'irectly over the stored spent fuel assemblies. All work in the spent 'uel pool area will be controlled and perfonned in strict accordance with specific written procedures. The crane that will be used to bring the racks into the Fuel Handlier Building has been evaluated and found acceptable.
In addition,.the temporary construction crane, which will be used to move racks within the spent 'uel pool area, has been evaluated and fourd. acceptable.
Section 5.0 of this safety evaluation contains the details of the staff's analysis. Thus, the probability of a construction accident is not signifi-cantly increased as a result of reracking. Accordingly, the proposed modi-fication does not invnive a significant increase in the probability of. an accident previously evaluated.
As noted in Section 2.0 of this safety evaluation, the consequences of a spent i
' fuel assembly drop in the spent fuel pool (scenario 1) was evaluated and it was found that the criticality acceptance criterion, k less than or equal to 0.95, is not violated. In addition, the radiologi$ consequences of a fuel assembly drop are not changed from the previous analysis. The staff also conducted an evaluation of the potential consequences of a fuel handling accident. The. staff analysis found that the calculated doses are less than 10 CFR Part 100 guidelires. The results of the analysis show that a dropped 'j spent fuel assembly on the racks will not distort the racks such that they would not perform their safety ' unction. Section 8.0 contains the details n' the staff's accident analysis. Thus, the consequences of this type accident are not changed from the previously evaluated spent fuel assembly drops which have been found acceptable.
The consequences of a loss of spent fuel pool cooling system flow (scenario 2) have been evaluated and it was found that sufficient time is available to provide an alternate means for cooling N.e., the fire hose stations) in the event of a failure in the soling system (see Section 5.0 of this safety evaluation). Thus, the w.lequencel of this type of accident are not signif-icantly increased from previously evaluated loss of cooling system flow accidents.
The consequences of a seismic event (scenario 3) have been evaluated and are acceptable. The new racks will be designed and fabricated to meet the requirements of applicable portions of the NDC Regulatory Guides and published standards. The new free-standing racks are designed, as are the existing free-standing racks, so that the floor loading from racks completely filled with spent fuel assemblies, partially filled, or empty at the time of the incident, does not exceed the structural capability of the spent fuel pool. The Fuel Handling Building and spent fuel pool structure have been evaluated for the increased loading from the spent fuel racks in accordance with the criteria previously evaluated by the staff and found acceptable. Section 5.0 contains the details of the staff's analysis. Thus, the consequences of a seismic event are not significantly increased from previously evaluated events.
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?O The (see consequences Section 8.0 of thisofsafety a spent fuel cask evaluation drop)(scenario
. The 4) have been evaluated radiological consequences.o*
the cask drop are well within'the guidelines of 10 CFR 10R and the doses are not increased as compared to the doses analyzed for the presently installed racks. The cask drop analysis is based on administrative and Technical Specifi-cations controls which ensure that minimum requirements for decay of irradiated fuel assemblies in the entire spent fuel pool are met prior to movement of the cask into'the cask area of the spent fuel-pool. Analyses also demonstrate that' k will always be IPss than the NRC acceptarce criterion.. In addition..
1 Nape from a cask drop will not exceed the makeup capabilities of the spent fuel pool. Thus, the consequences of a cask drop accident will not increase from previously evaluated accident analyses.
The consequences of a construction accident (scenario 5) are enveloped bv the spent fuel cask drop analysis. No rack (old or new) weighs more than a single 25 ton cask. In addition, all movements of heavy loads handled during the rerack operation will comply with. the NRC guidelines presented in NUREG-061P.,
" Control of Heavy Loads at Nuclear Power Plants." The consequences of a !
construction accident are not increased from previously evaluated accident l analyses.
Therefore, it is concluded that the proposed amendment to replace the spent fuel racks in the spent fuel pool will not involve a significant increase in the probability or consequences of an accident previously evaluated.
Second Standard
" Create the possibility of a new or different kind of accident from any accident previously evaluated."
As noted in various sections of this safety evaluation and the consultant's !
Technical Evaluation Report description of acceptance criteria (Section ?.0),
the staff evaluated the proposed modification in accordance with the guidance of the NRC position paper entitled, "0T Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," appropriate NRC Regulatory Guides, appropriate NRC Standard Review Plans, and appropriate industry codes and standards. In addition, the staff has reviewed several previous NRC Safety Evaluations for rerack applications similar to this proposal. No unproven techniques and methodologies were utilized in the analysis and design of the proposed high density racks. No unproven technology will be utilized in the fabrication and installation process of.the new racks. The basic reracking technology in this case has been developed and demonstrated in numerous applications for a fuel pool capacity increase which have already received NRC staff approval.
Third Standard
" Involve a significant reduction in a margin of safety."
The staff Safety Evaluation review process has established that the issue of margin of safety, when applied to a reracking modification, should address the following areas:
- 1. Nuclear criticality considerations
- 2. Thernal-hydraulic considerations
- 3. Mechanical, material and structural considerations.
21 i
The established acceptance criterion for criticality is that the neutron multi- '
plication factor in spent fuel pools shall be less than or equal to 0.95, including all uncertainties, under all conditions. This margin of safety has been adhered to in the criticality analysis methods for the new rack design.
The methods used in the criticality analysis conform with the applicable pertions of the appropriate staff guidance and industry codes, standards, and specifications.
In meeting the acceptance criteria for criticality in the spent fuel pool, such that k is always less than 0.95, including uncer-
.taintiesata?5f/95%probabTffty/confidencelevel,theproposedamendmentto 1 rerack the spent fuel pool does not involve a significant reduction in a margin !
of safety for nuclear criticality. Section ?.0 contains the deta1 9 of the staff's analysis, r
Conservative methods were used to calculate the maximum fuel temperature and the increase in temperature of the water in the spent fuel pool. The themal-hydraulic evaluation used the methods used for evaluations of the present spent fuel reeks in demonstrating the temperature margins of safety are maintained.
The proposed modification will increase the heat load in the spent fuel pool.
The evaluation shews that the spent fuel will be adequately cooled. Section 5.0 contains the details of the staff's analysis. Thus, there is no significant reduction in the margin of safety for thermal-hydraulic or spent fuel coolina concerns.
The main safe' ty function of the spent fuel pool and the racks is to maintain the spent fuel assemblies in a safe configuration through all normal or ,
abnormal loadings, such as an earthquake, impact due to a spent fuel cask )
drop, drop of a spent fuel assembly, or drop of any other heavy object. The mechanical, material, and structural design of the new spent fuel racks is in accordarce with applicable portions of the "NRC Position for Review and Acceptance of Spent Fuel Storage and Hendling Applications," dated April 14, 1978, as modified January 18, 1979; Standard Review Plan 3.8.4; and other applicable NRC guidance and industry codes. The rack materials used are i
1 corpatible with the spent fuel pool and tha spent fuel assemblies (see Section 3.0 of this safety evaluation). The structural considerations of the new racks L
address margins of safety against tilting and deflection or movement, such that the racks are not damaged durino impact (see Sectier 4.0 of this safety evalua-l tion). In addition, the spent fuel assemblies remain intact and no criticality concerns exist. Thus, the margins of safety are not significantly reduced by the proposed rerack.
Summary Based on the forepoing and the fact that the reracking technolocy in this instance has been well-developed and demonstrated, the Comission has concluded that the standards of 10 CFP 50.99 are satisfied. Therefore, the Commission has made a final determination thet the proposed amendment for spent fuel pool expansion does not involve a significant hazards consideration.
- 12. ENVIRONMENTAL CONSIDERATIONS l
A separate Environmental Assessment has been prepared pursuant to 10 CFR Part 51.
The Notice of issuance of Environmental Assessment and Finding of No Significant Impact was published in the Federal Register on March 4,1988(53FR7065).
1 i 22
- 13. CONCLUSIONS The staff has reviewed and evaluated the licensee's request for amendment *or.
the St. Lucie Plant, Unit I regarding the' expansion of the spent fuel pool.
Based on the considerations discussed in this safety evaluation, the staff ,
concludes that: I (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 from any accident previously evaluated. (c) significantly reduce a margin of safety; and therefore, the amendment does not involve significant hazards considerations; ;
(2) there is reasonable assurance that the health er.d safety of the public will not be endangered by operation in the proposed manner, and (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 oublic.
l la. REFERENCES - FP&L
- 1. FP&L letter No. L-87-2A5,.1une 12, 1987 from C. O. Woody (FPAL) to US NRC, Sub.iect: Proposed License Amendment Spent Fuel Pool Rerack.
Subject:
Spent Fuel Rerack. )
Subject:
Spent Fuel Pool Rerack-Design and Analysis.
Subject:
Spent Fuel Pool Rerack - Boraflex and Pool Cleanup.
Subject:
Spent Fuel Pool Perack - Radioactive Sources Dose Rate, and Dose Assessment.
- 6. FP&L letter No. L-87-519, December 21, 1987, from C. O. Woody (FP&L) to US NRC, Sub,iect: Environmental Effects of Transportation of Fuel and Waste.
Subject:
Spent Fuel Pool Rerack.
Subject:
Spent Fuel Pool Rerack - Design and Analysis. j
US NRC,
Subject:
Spent Fuel Pool Rerack - Design and Analysis.
l
Subject:
Spent Fuel Rerack.
1 11. FP&L letter No. L-88-38, January 29, 1988, from C. O. Woody, (FP&L) to I
US NRC,
Subject:
Spent Fuel Rerack.
i
o ??-
REFERENCES - NRC
- 12. U.S. Nuclear Regulatory Comission, letter dated July 16, 1987 from E. G. Tourigny (NRC) to C. O. Woody, (FP&L),
Subject:
Recuest for l 1
Additional Information.
i
- 13. U.S. Nuclear Regulatory Comission, letter dated Aunust ?0,1987 from E. G. Tourigny (NRC) to C. O. Woody, (FP&L),
Subject:
Recuest for Additional Information.
- 14. U.S. Nuclear Regulatory Commission, letter dated Auoust 25, 1987, from !
H. N. Berkow (NRC) to C. O. Woody (FP&L),
Subject:
Spent Fuel Pool Expansion. Also: Federal Register Notice, 52 FR 32852, August 31, 1087.
15.
U.S. Nuclear Reg (NRC)'to C. O. Woody (FP&L1, E. G. Tourigny . Request
Subject:
ulatory for Comissio!
Additional Information. '
- 16. U.S. Nuclear Regulatory Commission, meeting minutes dated September 11, l 1987 from E. G. Tourigny (NRC1,
Subject:
Sumary of September 2,1987 l
Meeting with FP&L and NRC Staff Regarding the Reracking of the Spent '
l Fuel Pool.
- 17. U.S. Nuclear Regulatory Commission, letter dated Septeribe' r 21, 1987, from E. G. Tourigny (NRC) to C. O. Woody (FP&L),
Subject:
Request for Additional Information. I 1P. U.S. Nuclear Regulatory Commission, meeting minutes dated October 21, 1987 ;
from E. G. Tourigny (NRC), Subfect: Sumary of October 2, .1987 Meeting j with FP&L and NRC Staff Regarding the Reracking of the Spent Fuel Pool.
- 19. U.S. Fuclear Regulatory Comission, letter dated October 23, 1987 from E. G. Tourigny (NPC) to C. O. Woody (FP&L),
Subject:
Request for Additional Information.
- 20. U.S. Nuclear Regulatory Comission, letter dated November 25, 1987 from 1 E. G. Tourigny (NRC) to C. O. Woody (FP&L), Sub.iect: Recuest for Additional Information.
- 21. U.S. Nuclear Regulatory Comission, meeting minutes dated December 4,1887 from E. G. Tourigny (NRC),
Subject:
Sumary of October 29 and 30,1987 Audit of J. Oats and HOLTEC in Support of Reracking of the Unit 1 Spent Fuel Pool.
- 22. U.S. Nuclear Regulatory Comission, meeting minutes dated December 9,1987 from E. G. Tourigny (NRC),
Subject:
Sumary of November 24, 1987 Meeting between FP&L and NRC Staff Regarding the Reracking of the Spent Fuel Pool.
- 23. U.S. Nuclear Regulatory Comission, letter dated February 29, 1988 from E. G. Tourigny (NRC) to C. O. Woody (FP&L),
Subject:
Environmental Assessment and Finding of No Significant Impact - Spent Fuel Pool Expansion, St. Lucie Plant, Unit No. 1. Also: Federal Register Notice, 53 FR 7065, March 4, 1988.
L
- 1
-s
?4
. REFERENCES - OTHER PA. Campbell Rich to U.S. Nuclear Regulatory Comission Secretary to the -
Comission, letter dated September 30, 1987 l l
- 25. Campbell Rich to U.S. Nuclear Regulatory Comission, undated letter, enveloped postmarked January 15, 1988.
Dated: March 11,1988
Attachment:
Appendix A 1
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7590-01 UNITED STATES NUCLEAR REGULATORY C06911SSION
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FLORIDA P(YER AND LIGHT COMPANY l l
DOCKET NO. 50-335
]
NOTICE OF ISSUANCE OF AMFNDMENT TO FACILITY OPERATING LICENSE AND FINAL DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION The U. S. Nuclear Regulatory Commission (the Commission) has issued Amendment No. 91 to Facility Operating License No. DRP-67, issued to the Florida Pcwer and Light Company, (the licensee), which revised the Technical Specifications for operation of the St. Lucie Plant, Unit NoJ 1, located in St. Lucie County, Florida. The amendment was effective as 'of the date of its issuance.
The amendment allows the expansion of the spent fuel pool storage capacity from the current 728 fuel assemblies to the proposed 1706 fuel assemblies. The expansion is to be achieved by removing the existing racks and installing new, higher density ones.
The application for the amendment complies with the standards and reouire-ments of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations. The Commission has made appropriate findings as reouired by the Act and the Comission's rules and regulations in 10 CFP Chapter 1, which are set forth in the license amendment.
1
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- p. 1 l
Notice of Consideration of Issuance of Amen #ent and Proposed No Significant '
i Hazards Consideration Determination and Opportunity for Hearing in connection !
with this action was published in the FEDERAL REGISTER on August 31 1987 (59 FR 3?852). A reeuest for a hearing was filed on September 30,1987 by Mr. Campbell Rich.
l
-Under its regulations, the Commission may issue and make an amendment immediately effective, notwithstanding the pendency before it of a request for a bearing from any person, in advance of the holding and completion of any required hearing, where it has detemined that no significant hazards'considera-tions are involved. i The Commission has applied the standards of 10 CFR 50.92 and has made a final determination that the amendment involves no significant hazards considerations. The basis for this determination is contained in the Safety Evaluation related to this action. Accordingly, as described above. the amendment has been issued and made imediately effective and any hearing will be held after issuance.
The Commission has prepared an Environmental Assessment (March 4, 19e8, 53 FR 7065) related to the action and has concluded that an environmental impact statement is not warranted because there will be no environnantal impact attributable to the action beyond that which has been predicted and. described in the Commission's Final Environmental Statement for St. Lucie Unit 1 dated June 1973.
For further details with respect to the action, see (1) the application for amendment dated June 12, 1987, as supplemented by letters dated September 8, lop 7, October 20,1987 (three letters), December 21, 1987 December 2?, 1987,
y,v T
1 December 23, 1987 (three letters), and January 29, 1988 (?) Amendment No. 91 to Facility Operating License No. DPR-67, (3) the Commission's related Sefety Evaluation, er.d (4) the Commission's related Environmental Assessment. All of these items are available for public inspection at the Commission's Public
~
Document Room, 1717 H Street, N.W., Washington, D.C., and at th' e Indian River Junior College Library, 3208 Virginia Avenue, Fort Pierce, Florida 33450. A l copy nf items (2), (3), and (4) may be obtained upon request addressed to the !
U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, Attention:
Director, Division of Reactor Projects I/II. -
Dated at Rnckv111e, Maryland this lith
- day of March 1988.
FOR THE NUCLEAR REGULATORY COMMISSION
[k
- W E. G. Tourigny, adect Manager Project Direc orage II-2 Division of React'or Projects-T/II Office of Nuclear Reactor Regulation
i APPENDIX A
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TECHNICAL EVALUATION REPORT EVALUATION OF THE HIGH DENSITY SPENT FUEL RACK STRUCTURAL ANALYSIS FOR FLORIDA POWER AND LIGHT COMPANY ST. LUCIE PLANT - UNIT NO. 1 1
By 1
G. DeGrassi 1
i STRUCTURAL ANALYSIS DIVISION DEPARTMENT OF NUCLEAR ENERGY BROOKHAVEN NATIONAL LABORATORY l
UPTON, NEW YORK l
l February 1988 Prepared for U.S. Nucelar Regulatory Commission Office of Nuclear Reactor Regulation l Fin A-3841
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, k Executive Summary This report describes and presents the results of the BNL technical evaluation of the structural analysis submitted by Florida Power and Light-Company in support of their licensins submittal on the use of high density j fuel racks at St. Lucie Unit No. 1. The review was conducted' to ensure chat i the racks meet all structural requirements as defined in the NRC Standard '
Review Plan and the NRC OT Position for Review and Acceptance of Spent Fuel Pool Storage and Handling applications.
The. proposed raracking of the spent fuel pool involves the installation of seventeen free-standing, self-supporting modules of varying sizes arranged within close proximity to each other and the pool walls. Each rack module -
consists of 1ndividual cells of square cross-section, each designed to 3
I accommodate one fuel assembly. . Since the racks are neither anchored to the pool floor or walls nor connected to each other, during an earthquake, the racks would be free to slide and tilt. ~For an earthquake of sufficient j intensity, the racks could impact each other and the pool walls. Because of '
the nonlinear nature of this design, a time history analysis was required to characterize the seismic response of the fuel rscks.
I The BNL review focused primarily on the seismic analysi's of the fuel rack '
modules because of the complexity of the analysis method and the number of simplifying assumptions that were required in developing the dynamic models.
BNL also reviewed other analyses performed by the Licensee ' including fuel-handling accident analyses, thermal analyses,' and spent fuel pool and liner analyses.
During the course of the review, a number of questions were raised regarding the adequacy of the fuel rack dynamic models. Concerns were raised that single rack models may underpredict seismic forces and displacements that ,
would occur in the real multiple rack fuel pool environment (Section 4.1.1). 1 Concerns were also raised regarding the adequacy of fluid coupling _ assumptions used in the models (section 4.1.2). In response to these questions, the Licensee provided additional information and performed additional studies, including multiple fuel rack seismic analyses, to demonstrate the~ adequacy of the design basis results.
The additional studies indicated that the design basis models predict conservative seismic loads and displacements. It was noted, however, that the most significant factor contributing to the ennservatism was the use of twice the fuel assembly weight in the design basis models. Nevertheless, the results of these studies coupled with the significant safety factors in the results provided a high level of confidence to conclude that there is sufficient conservatism in the results to compensate for analytical uncertainties. '
Based on the BNL review of the Licensee's analyses, it was concluded thst the proposed St. Lucie Unit I high density fuel racks.and' spent fuel pool are demigned with sufficient capacity to withstand the effects of the required environmental and abnormal loads.
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.. N TABLE OF CONTENTS Page
1.0 INTRODUCTION
1.1 Purpose 1 1.2 Background 1 1.3 Scope of Review 1 .]
3 1
2.0 ACCEPTANCE CRITERIA 2 3.0 FUEL RACK DESCRIPTION 3 4
4.0 TECHNICAL EVALUATION
4 4.1 Fuel Rack Seismic Analysis 4 4.1.1 Dynamic Model 5 4.1.2 {
Fluid Coupling Effects 7 4.1.3 Friction Effects 10 l 4.1.4 Damping -
11 4-1.5 Seismic Loads 11 4.1.6 Lead Cases 11 4.1.7 Analysis Method 12 4.1.8 Analysis Results 13 i 4.1.9 Evaluation of Results 14 l 4.2 Additional Fuel Rack Seismic Studies 14 4.2.1 Single Rack Studies 14 4.2.2 Multiple Rack Studies 15
! 4.2.3 Overall Evaluation of Seismic Analysis Results 16 j
4.3 Thermal Analysis 16 )
4.4 Fuel Handling Accident Analysis 17 4.5 Spent Fuel Pool Analysis 19 4.5.1 Load and Load Combinations 19 4.5.2 Spent Fuel Pool Structure /.nalysis 20 4.5.3 Pool Liner and Anchorage analysis 20,
5.0 CONCLUSION
S 21
6.0 REFERENCES
23 y
S k
LIST OF TABLES TABLE TITLE PAGE 1 MODULE DATA 25 2 MODULE DIMENSIONS AND WEIGHTS 26 1
3 RACK MODEL PARAMETERS 27 l 4 RACK SEISMIC ANALYSIS RESULTS j IMPACT LOADS AND STRESS FACTORS 28 5 RACK SEISMIC ANALYSIS RESULTS
SUMMARY
DISPLACEMENTS AND FLOOR LOADS 29 6
SUMMARY
OF SAFETY FACTORS IN' CRITICAL 30 FUEL RACK LOCATIONS 7
RESULTS OF SINGLE RACK STUDIES-FULLY LOADED G1 RACK WITH COF = 0.8 31-8 RESULTS OF MULTIPLE RACK STUDIES-FULLY-LOADED 1A , A2e Bs i 32RACKS WITH COF = 0.2 32 9
RESULTS OF MULTIPLE RACK STUDIES-FULLY LOADED 1A , A2 , B 1, B2 RACKS WITH-
- COF = 0.8 33 10 RESULTS OF MULTIPLE RACK STUDIES-FULLY LOADED1A , A2 , B 1, B2 RACKS 34 11 SPENT FUEL POOL STRUCTURE MAXIMUM STRESS
SUMMARY
35- t 8
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- y. _e s i
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LIST OF.TIGURES FIGURE TITLE PAGE 1
SPENT FUEL POOL LAYOUT 36 2
TYPICAL RACK ELEVATION-REGION 1 37 ,
3 TYPICAL RACK ELEVATION-REGION 2 38 4
TYPICAL CELL ELEVATION-REGION 1 39 5
3 x 3 TYPICAL ARRAY-REGION 1 40 6
TYPICAL CELL ELEVATION REGION 2 41 7
3 x 3 TYPICAL ARRAY-REGIGN 2 42 8
ADJUSTABLE SUPPORT LEG 43 9 l SCHEMATIC MODEL OF FUEL RACK 44 j 10 FUEL HACK MODEL SHOWING RACK-TO-RACK l
IMPACT SPRINGS 45 l 11 IMPACT SPRING ARRANGEMENT AT NODE i 46 12 SPRING MASS SIMULATION FOR TWO-DIMENSIONAL MOTION f 47 13 NORTH-SOUTH SSE 48 I 14 il EAST-WEST SSE 49 i 15 . VERTICAL SSE 50 I 1
16 SPENT PUEL POOL MAT PLAN AND SECTION 51 17 SPENT FUEL POOL MODEL OVERALL VIEW 52 vii
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1.0 INTRODUCTION
1.1 Purpose i This technical evaluation report (TER) describes and presents the results
'of the BNL review of Florida Power and Light. Company's licensing submittal on the use of high density fuel racks at St. Lucie Unit No. I with respect to their structural adequacy.
1.2 Background
The existing racks in the-spent fuel pool have 728 total storage cells.
With the presently available storage cells,.St. Lucie, Unit No. 1 lost the full-core reserve storage capacity af ter the seventh refueling which was completed in the spring of 1987. To correct this situation and provide sufficient capacity to store discharged fuel assemblies, the Licensee has requested NRC to issue a License Amendment to replace the existing storage racks with new high density spent fuel storage racks.- The.new racks will allow for more dense storage of spent fuel, thus enabling the existing pool'to store more fuel. The new high density racks have.a usable storage capacity of 1706 cells, extending the full-core reserve storage capability until the year 2009. -
The proposed racks consist of individual cells of square cross-section, each of which accommodates a single PWR fuel assembly. The cells are assembled into distinct modules of varying sizes which are to be arranged within the existing spent fuel pool. Each module is free-standing and self-supporting.
The Licensee provided a summary of his safety analysis and evaluation of the proposed racks in a Safety Analysis Report (Ref. 1). The report described the structural analysis of the new fuel racks and the existing fuel pool. It also gave a description of postulated dropped fuel and jammed fuel accident analyses.
BNL reviewed the Safety Analysis Report and generated a list of additional information needed to complete the review (Ref. 2). . The Licensee provided the additional information in later submittals (Ref. 3a, b, c). In addition, BNL' ,
participated in a limited audit of the fuel rack analysis and fabrication in the offices of Holtec International, the fuel rack designer, and at Joseph Oat Corporation, the fuel rack fabricator.
1.3 Scope of Review The objective of the BNL technical review was to evaluate the adequacy of the Licensee's structural analysis and design of the proposed high density spent fuel racks and spent fuel pool. Due to the complex nature of the fuel rack seismic analysis, the primary focus of the review was on the adequacy of the non-linear fuel rack models and their dynamic analysis. The structural evaluation of fuel racks subjected to the dropped fuel and jammed fuel 1 !
e i
handling accidents described in the Licensee's report (Ref. 1) were included in this review. However, the definition of these postulated accidents and !
their parameters (drop height, uplif t force, etc.) were beyond the scope of this review. A limited review of the spent fuel pool was conducted to insure that appropriate loads, methodology and acceptance criteria were applied.
1 2.0 ACCEPTANCE CRITERIA The acceptance criteria for the evaluation of the spent fuel rack appli- y cations are provided in the NRC OT Position for Review and Acceptance of Spent l Fuel Storage and Handling Applications (Ref. 4). Structural requirements and !
criteria given in this position paper were updated and included as Appendix D to Standard Review Plan 3.8.4, " Technical Position on Spent Fuel Pool Racks,"
(Ref. 5). These documents state that the main safety function of the spent 4
I fuel pool and fuel racks is to maintain the spent fuel assemblies in a safe '
configuration through all environmental and abnormal loadings, such as earth-quakes, and impact due to spent fuel cask drop, drop of a spent fuel assembly, i or drop of any other heavy object during routine spent fuel handling. '
Section 2 of SRP 3.8.4, Appendix D gives the applicable Codes, Standards I and Specifications. Construction materials should conform 'to Section III, 1 Subsection NF of the ASKE Codes. Design, fabrication and installation of ;
stainless steel spent fuel racks may be performed based upon the ASME Code !
Subsection NF requirements for Class 3 component supports. I I
Requirements for seismic and impact loads are discussed in Section 3 of l Appendix D. It states that seismic excitation along three orthogonal !
directions should be imposed simultaneously for the design of the new rack system. Submergence in water may be taken into account. The effects of sub- I j
mergence are considered on a case-by-case basis. Impact Loads generated by i the closing of fuel assembly to fuel rack gaps during a seismic excitation I
should be considered for local as well as overall effects. It should also be !
demonstrated that the consequent loads on the fuel assemblies do not lead to fuel damage. Loads generated from other postulated events may be acceptable if sufficient analytical parameters are provided for review.
Load and load combination requirements are provided in Section 4.
Specific loads and load combinations are acceptable if they are in conformance with Section 3.8.4-II.3 and Table 1 Appendix D of the Standard Review Plan.
Changes in temperature distribution should be considered in the design of the pool struwcure. Temperature gradients across the rack structure due to differential heating effect between a full and an empty cell should be incorp-orated in the rack design. Maximum uplif t forces from the crane should be considered in the design.
Section 5 discusses design and analysis procedures. It states that design and analysis procedures in accordance with Section 3.8.4-II.4 of the Standard Review Plan are acceptable. The effects of gaps, sloshing water, and increase of effective mass and damping due to submergence in water should be quantified. Details of the mathematical model including a description of how 2
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e the important parameters are obtained should be provided.
Structural acceptance criteria are provided-in Section 6. The acceptance criteria are given in Table 1 of Appendix D. For impact loading, the ductility ratios utilized to absorb kinetic energ) should be quantified. When considering seismic loads, factors of safety against gross sliding and over-turning of the racks shall be in accordance with Section 3.8 5-II.5 of the ;
Standard Review Plan unless it can be shown that either (a) sliding motions are minimal, impacts between adjacent racks and between racks and walls are prevented and the factors of safety against tilting are met, or (b) sliding and tilting motions will be contained within geometric constraints and any impact due to the clearances is incorporated.
3.0 FUEL RACK DESCRIPTION The new high density spent fuel storage racks consist of individual cells with B.65 inch by 8.65 inch nominal square cross-section, each of which accom-modates a single Combustion Engineering or Exxon PWR fuel assembly or equiva-lent, from either St. Lucie Unit 1 or Unit 2. A total of 1706 cells are arranged in 17 distinct modules of varying sizes in two regions. Region 1 is designed for storage of new fuel assemblies with enrichments up to 4.5 weight , i percent U-235 or for fuel assemblies with the same maximum enrichment which '
have not achieved adequate burnup for Region 2. The Region 2 cells are l capable of accommodating fuel assemblies with various initial enrichments which i
have accumulated minimum burnups within an acceptable bound as discussed in the Licensee's Safety Analysis Report (Ref. 1). The arrangement of the rack modules in the spent fuel pool is shown in Figure 1. Typical Region 1 and Region 2 racks are shown in Figure 2 and 3. Each rack module is equipped with girdle bars at the upper end, 3/4-inch thick by 31/2 inches high. The modules make surface contact between their contiguous walls at the girdle bar locations and thus maintain a nominal 1 1/2 inch gap between adjacent module cell walls.
The modules in the two regions are of eight different types.
Tables 1 and 2 summarize the physical data for each modu? e type.
The rack modules are fabricated from ASME SA-240-304L austenitic stainless steel sheet and plate material, and SA-351-CF3 casting material and SA-564-630 precipitation hardened stainless steel for supports. The weld filler material utilized in body welds is ASME SFA-5.9, Classification ER 308L. Boraflex serves as the neutron absorber material. Boraflex is a silicone-based polymer containing fine particles or boron carbide in a homogeneous, stable matrix.
Each rack module consists of the following components:
Internal square tube l
Neutron absorber material (Boraflex)
Poison sheathing (Region 1 only)
Cap element (Region 1 only)
Baseplate 1
3 1
e . '
(
. 1 Support assembly
- i Top lead-in (Region 1 only) l i
Figures 4 and 5 show a typical Region 1 cell elevation and a typical 3x3 i array horizontal cross-section. Figures 6 and 7 show the same views of a I typical Region 2 rack module. The figures show that the major difference between the Region 1 and Region 2 module designs is the larger pitch between cells in Region 1. Channel shaped gap elements are welded between the Region 1 cell tubes to maintain the minimum flux trap required between adjacent internal cells. Region 1 modules use poison; sheathing (cover sheets) to position and retain the Boraflex absorber. material around each cell wall. Io Region 2 modules, the Boraflex absorber material is placed between the walls of interior cells and kept in place by stainless steel connecting strips. The Region 1 modules also provide lead-ins at the top of each cell wall to facilitate fuel assembly insertion.
i The adjacent cells of each module are welded together either through gap elements (Region 1) or. side connecting strips (Region 2) to form a honeycomb structure. The honeycomb is welded to a 3/4 inch thick baseplate with 3/32 inch fillet welds. The baseplate has 6-inch diameter holes concentrically located with respect to each square tube, except at support leg locations, where the hole size is 5 inches in diameter. These holes provide the primary path for coolant flow.
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Each module has at least four support legs. All supports are adjustable in length to enable leveling of the rack. The variable height support -)
q assembly consists of a flat-footed spindle which rides into an internally-threaded cylindrical member. The cylindrical member is attached to the under-side of the baseplate through fillet and partial penetration welds. Figure 8 shows a vertical cross-section of the adjustable support assembly. !
l The support legs will rest on 1-1/4 inch thick plates on the spent fuel pool floor.
Additional plates will be provided for those areas of the pool floor plates.
where the rack support legs are located which do not already have f The new plates will not be attached to the pool floor. Aside from i the addition of these plates, the Licensee has indicated that no other spent fuel pool modifications are needed to accomodate the new racks.
4.0 TECHNICAL EVALUATION
4.1 Fuel Rack Seismic Analysis The spent fuel storage racks are seismic Category I equipment required to remain functional during and af ter a safe shutdown earthquake. .As described i in Section 3.0, the proposed racks consist of 17 distinct free-standing modules which are neither anchored to the pool floor, attached to the side walls, nor connected to each other. Any rack may be completely loaded with 4
fuel assemblies, partially loaded, or completely empty. The fuel assemblies are free to rattle within their storage cells. !
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4 Seismic forces are transmitted to the racks through friction at the support legJt o pool floor interface. If seismic displacements are large-enough, the racks can impact against each other or the pool walls and the support legs can lif t off and impact the pool floor. Because of these non-linearities, .a time history analysis of nonlinear rack acdels was required to characterize the seismic response of the fuel racks. BNL's review of the details of the modeling technique and analysis method is described in the following sections.
4.1.1 Dynamic Model l The Licensee's mathematical model of a spent fuel racks module is shown in Figurer 9, 10, 11, and 12.
The Licensee indicated that the rack structure _ is very rigid and that its motion can be characterized in terms of six degrees of -
freedom at the rack base. Figure 9 shows the rack as a rigid stick on a rigid base with three transnational (qi, q2 93) and three rotational (q4, q5 96) degrees of freedom. The fuel assemblies are treated as five lumped masses located at.different elevations (nodes 1* to 5*). All fuel assemblies in a rack module are assumed to vibrate in phase. Each lumped mass is assumed to rattle independently within the rack cell gaps and is repre-sented by two horizontal transnational degrees-of freedos. . Impacts between the rack and fuel assembly lumped masses are accounted for by the use of compression-only gap elements as shown in Figure 11. The support legs are modeled as compression-only springs (S1 to S4 in Figure 9) which consider the local verrical flexibility of the rack-support interface. Friction elements' are used at the bottom of the support legs. Figure 10 shows the impact springs acting through gap elements to simulate the interface with adjacent rack modules or pool walls. Five impact. springs per side' are used at both the girdle bar and baseplate elevations. Figure 12 shows a two dimensional repre-sentation of the model with only one " rattling" fuel mass to clarify the overall model concept.
Fluid coupling between rack and fuel asemblies, and between rack and adjacent racks or walls is simulated by including inertial coupling terms in the equations of motion. This is discussed in detail below. Fluid damping between rack and fuel assemblies, and between rack and' adjacent racks is neglected in the model.
In order to simulate the motion of adjacent fuel racks, the model assumes a symmetry plane midway between adjacent racks. Thus the model assumes that each adjacent rack moves completely out of phase with the rack being analyzed. This assumption is intended to predict conservative rack to rack impect forces.
To complete the review of the adequacy of.the model, the Licensee was requested to provide typical fuel rack and fuel assembly design drawings and a list of key modeling parameters. The Licensee provided typical drawings (Ref. 6-9) and a list of model parameters shown in Table 3. Impact spring values were based on local stiffness of the rack at the support foot to pool 5
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liner interface and at the fuel assembly to rack cell interface. Rack to rack impact spring values were set at 1 x 100 lb/in. This value is reasonable for base plate-locations but significantly larger then would.be expected at the girdle bar locations. However, the use of a high spring constanc at this location should be conservative and overestimate peak impact loads at the girdle bars.
The Licensee indicated that the new racks will rest on 1-1/4 inch base plates on the pool floor. The plate material is 304 stainless steel which is the same material as the pool liner. Most of the baseplates were added to the )
pool at the time at the last rerack. Additional baseplates will be added to accomodate the support configuration of the new racks. The existing plates were welded to the pool liner. The new plates will not be arcached to the pool floor. The baseplates were not included in the rack model.but were assumed to move in the same manner as the rack floor. The Licensee indicated that this assumption is reasonable because the friction coefficient between-the baseplates and liner should be greater then the friction coefficient between the baseplate and rack support feet.because~of the differences in b;terials.
A review of the drawings indicates that the baseplates are large enough to accomodate a reasonable amount of slippage of the fuel racks during an earthquake. Overall, the use of baseplates is a desirable design feature since they will serve to distribute fuel rack loads over a large-area and will protect the pool liner from local punching or tearinc.at.the rack la; interfaces.
The weight of the fuel included in the model was based on 2500 pounds per fuel assembly which is about twice the design weight. The Licensee used the higher weight to account for possible use of consolidated fuel in the future.
For this application, the Licensee stated that the higher weight should pro-vide conserystive results. The results of additional analytical studies were provided to support this position as discussed in Section 4 2.1. Since the Licensee's proposed Licensee amendment did not involve the use of consolidated fuel, the higher weight was considered a conservative modeling assumption in 3 this review.
If the Licensee intends to use consolidated fuel at a later date, further evaluation would be required to reassess the safety margins and to consider other factors which may affect the seismic design.
The Licensee was asked to provide justification for treating the fuel assemblies as five independent rattling masses. The Licensee stated that the fuel assemblies have a natural frequency much lower than _the rack and sub-mitted additional studies to demonstrate that the effects of coupling the masses are not significant when compared to the overall conservatism of the model. This is discussed in Section 4.2.1. The fuel was modeled as five lumped masses at equally spaced elevations above the baseplate. In reality, fuel-rack impacts would be expected to occur at the nine spacer grid locations and at the upper and lower end fittings. The selection of only five impact locations combined with the assumption that all fuel assemblies move in-phase should result in conservative fuel-to-rack impact loads.
6 r
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The Licensee was asked to provide justification for the assumption that the motion of the rack can be represented by a rigid six degree of freedom structure. The Licensee indicated that for a typical rack, the lowest natural frequency of the rack gridwork vibrating in' water is 32 Hz. For seismic analysis, it is appropriate to consider this as a rigid body whose motion can be described by a six degree of freedom model.
The adequacy of analyzing only a single rack model in the seismic analysis was questioned. The seismic motion of a single rack is coupled to the motion of adjacent racks through impact forces and fluid coupling forces. The single rack model constrains the motion of a rack within an imaginary boundary.
Maximum displacements cannot exceed one-half the gap to the adjacent; racks.
For sufficiently strong seismic motion, sliding and tilting motsans of the racks could be larger than those predicted by a constrained single rack model resulting in higher impact velocities than would be predicted by a single rack model. Under worst conditions, rows of racks could slide together in one direction and pile up against a pool wall. The additional mass of racks involved in the impact could generate larger loads on the racks and the pool walls.
This concern may be more critical for the pool walls, since they are not designed to accomodate seismic impact loads from the fuel racks. In response to these concerns, the Licensee committed to perform a two {
dimensional multiple rack analysis of a single row of fuel racks to determine the extent of rack displacement under an SSE. The results and evaluation of the multiple rack analysis is discussed in Section 4.2.2.
4.1.2 Fluid Coupling Effects The effect of submergence of the fuel racks in a pool of water has a j significant effect on their seismic response. The dynamic rack model incorporated inertial coupling (fluid coupling) terms in the equations of motion to account for this effect. For two bodies (mass mi and m2) i sdjacent to each other in a frictionless fluid Ledium, Newtons equations of motion have the form:
(mi + Mit) N1 - M 12 k2= applied forces on mass mi
-M21 *X i + (82 + H22)X'2 = applied forces on mass m2
.. .. i X,X2 1
respectively. denote absolute accelerations of mass at and "2 M11, M12 M21 and M22 are fluid coupling coefficients '
which depend on the shape of the bodies and their relative disposition. The basic theory is summarized in a paper by Fritz (Ref. 10). The equations indicate that the effect of the fluid is to add a certain amount of mass to the body (M11 to body 1), and an external force which is proportional to the acceleration of the adjacent body. Thus the acceleration of one body affects the force on the adjacent body. The force is a strong function of the interbody gap, reaching large values for very small gaps. It should be noted that fluid coupling is based on fluid inertial effects and does not constitute damping. Fluid damping was not included in the model.
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Fluid coupling terms were included in the equations of action for fuel masses vibrating within the racks and for racks vibrating adjacent to other racks or the pool wall. The coupling terms modeling the effects of fluid flowing between adjacent racks were computed by assuming that all adjacent racks are vibrating 180 degrees out of phase with the rack being analysed.
Therefore, only one rack was considered surrounded by a hydrodynamic mass com-puted as if there were a plane of symmetry in the middle of the gap region.
Fluid virtual mass was included in the vertical direction vibration equations of the rack. Virtual inertia was also added to the governing ,
l equations corresponding to the rotational degrees of freedom. The effect of sloshing of water in the pool was neglected. This effect was shown to be negligible at the bottom of the pool. )
4 Several questions were raised regarding the conservatism of the fluid coupling parameters used in the analysis: (1) The Frits approach makes the assumption that the vibratory deflections are small relative to the. size of the gaps. This assumption does not correspond to the conditions that would prevail during an earthquake where the rack-to-fuel and rack-to-rack displace-ments would be as large as the gaps. Fluid coupling coefficients are cal-culated on the basis of a constant gap assumption. As fuel racks move away l from each other, the coupling coefficients should decrease;.resulting in lower fluid coupling forces and possibly higher. velocities. (2) The assumption that adjacent racks are vibrating 180 degrees out of phase seems to maximise the retarding effect of fluid forces and reduce the maximum impact velocities of the racks. This can result in unconservative rack-to-rack impact forces.
(3)
The rack-to-fuel fluid coupling terms were ' calculated based on the assumption that the fuel assemblies are solid square cross-sectional bodies, and that all of the surrounding water flows in the fuel assembly / cell wall space around the periphery of the fuel.
rods with gaps between rods. In reality, the fuel assemblies are arrays of fuel can flow both around and through As athefuel assembly vibrates within a cell, water.
fuel. The resulting fluid coupling forces can then be much lower than predicted by this model.
1 The Licensee provided additional information to justify the conservatism of the fluid coupling aerumptions. Frevious studies by Singh and Soler (Ref. 11) have shown that for large deflections, the contribution of the fluid I leads to terms non-linear springs. in the mass matrix and to terms which can be considered as For the small deflection assumption, the non-linear spring terms disappear the equations above. and only the mass matrix terms-are included as shown in The referenced paper provided the results of a study .
which considered the effects of the non-linear spring terms in a fuel- !
assembly / cell model. I The Licensee stated that the study demonstrated that the inclusion of these terms leads to lowering of the structural response. In response to the question regarding the consideration of flow area thro, ugh the I fuel assemblies, the Licensee indicated that the flow of water through a fuel )
assembly array of rods involves repeated changes in the flow cross-sectional area which would result in significant hydraulic pressure losses. The hydraulic pressure loss due to flow through the narrow convergent / divergent 8
.t.
k e
channels is an important mechanism for energy loss from the vibrating rack system.
The referenced paper was reviewed for applicability. The study involved the non-linear seismic analysis of a simplified two degree of freedom model of-a single fuel' assembly / rack cell system. The fuel assembly was modeled as an )
unperforated square cross-section to simulate a channeled BWR fuel assembly.
Equations of motion were written to incorporate large deflection inertial ;
coupling and fluid damping due to frictional losses. A time history analysis was performed by applying a sinusoidal ground acceleration to the model.
Cases which were analyzed included the following considerations: 1) No fluid effects, 2) Small deflection fluid coupling, 3) Large deflection fluid l
couplins, no fluid damping,<4) Large deflection fluid coupling with damping, )
- 5) Large deflection fluid coupling with reduced damping. In case 5, the fluid damping was taken as 1% of the values used in case 4 in an attempt to simulate the possible damping effect of unchannelled fuel assemblies. The authors recognized the differences in fluid effects between unchannelled fuel such in as the St. Lucie PWR fuel assemblies and channelled' fuel assemblies used BWR's. Channelled fuel assemblies can be appropriately represented as solid squara cross-sectional bodies. The authors stated, "It is clear that the dampir.g and virtual mass effects from an unchannelled fuel assembly should be substantially less since the confined fluid has more unobstructed area in which to flow as the fuel assembly moves relative to the cell wall. .In addition, there are substantial differences in the flow field which should be considered in any analysis of unchannelled fuel. Nevertheless, case 5 may give some indication of what might be expected if only unchannelled fuel assemblies are in the rack".
forcesTheand results rack of the study spring were presented in terms of fuel-to-rack impact forces.
level and pool floor loads. The latter forces are a measure of rack stress The results showed that the forces predicted by the small deflection model (case 2) exceeded the forces predicted by the large displacement models with damping (cases 4 and 5). 'A comparison between the results of the small deflection (case 2) model and the large deflection model with no damping (case 3) showed that the small deflection model predicted -
higher rack spring forces but lower fuel to rack impact forces.
The referenced study does not resolve all of the concerns related to the fluid coupling model assumptions. It provides evidence that large deflection inertial ef fects combined with damping tend to predict lower forces then a small deflection model. However, none of the models properly modeled fluid inertial effects for unchannelled fuel as is used in St. Lucie. The reduced damping used in case 5 was only meant to give an indication of trends which might be seen for unchannelled versus channelled fuel response. There was no analytical or experimental evidence to demonstrate the equivalence of case 5 '
parameters to unchannelled fuel parameters.
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on the other hand, the case'l results (vibration in air, no fluid effects) may be interpreted as an upper bound case. When compared with the. case 2
- small deflection model results, case 1 predicted 25% higher fuel to rack ,
impact forces and 20% higher rack spring forces than case 2.. When viewed I together, the results of the 5 cases provide a measure of the sensitivity of -)'
variations of fluid coupling parameters in predicting seismic response forces. This together with safety margins can be used to assess the adequacy of the design. .i I
4.1.3 Friction Effects Friction elements were used at the bottom of' rack support leg elements of :
the test results The model. givenvalue of the coefficient in Reference 12. of friction was based on documented The results of 199 tests performed on i austenitic stainless steel plates submerged in water showed a mean value of I coefficient'of friction to be 0.503 with a standard deviation of 0.125. Based ,
on twice the standard deviation, the upper and lower bounds are 0.753 and '
O.253, respectively. Two separate analyses were performed for each load case with values of coefficient of friction equal to 0 2 (lower' limit) and 0.8 '
(upper limit), respectively.
1 The Licensee was asked to provide justification for using the same ,
friction coefficient for both static and sliding rack conditions. He indi- .l '
cated that there is only a small difference between the static and sliding values. ;
The use of both an upper and lower bounding value is judged to be appropriate. Previous studies have indicated that low friction results in- l maximum sliding response of the racks while high friction results in maximum i rocking or tilting response. Consideration of both cases should prt vide worst case displacements, stresses and impact loads.
4.1.4 Damping Since the model assumed that the fuel rack' gridwork and baseplate are rigid, and the fuel assemblies can be treated as independent lumped masses, no damping resulting from structural deformations of the components was assumed.
Structural damping was included in all of the impact spring elements. For SSE load conditions, 2% structural damping was used. This value is in accordance with the FSAR and represents an acceptable, conservative value for impact damping.
4.1 5 Seismic Loads Seismic floor response spectra for the spent fuel pool floor were developed using the methods described in the FSAR. The parameters of the original lumped mass model of the Fuel Handling Building were adjusted to reflect the increased mass corresponding to the new high density spent fuel storage racks. New response spectra curves were generated using the same method which was used in the original dynamic analysis. Minimum and maximum fuel rack weights were considered in the analysis, corresponding to the empty 10
(
e and full conditions of the racks. Three ground motion acceleration records . ;
(as used in the original plant design) were used as input. These six combi- )
nations of parameters resulted in six response spectra curves, which were then 1 broadened + 20% and enveloped into one curve which envelopes the full spectrum -
of rack loading conditions. Six such curves were developed, two (OBE and SSE). ;
for each direction (NS, EW, vertical). The response spectrum curves are included in the Licensee's Safety Analysis Report (Reference 1)
The revised response spe:tra were used to generate statistically inde-pendent acceleration time histories, one for each of the three orthogonal directions. A computer program was used to generate the artificial time histories as a sum of sinusoidal waves. The program.used s' iteration approach whereby the calculated response provided by the simulated seismic excitation was compared with the " target" design response spectrum. The amplitudes of the sine waves were modified at each iteration' step so as to obtain the best agreement at certain control frequencies specified by the user. The resulting time histories used in the fuel rack analysis are shown in Figures 13 to 15.
The Licensee was asked to provide a comparison of the design response spectra with the artificial time history response spectra. This comparison was provided in terms of velocity response spectra plots'in Reference'3a. The plots showed reasonable agreement between the calculated curves and the design curves. ,
l Based on the Licensee's description, the methodology used to develop the seismic input for fuel rack analysis is acceptable and consistent with industry practice.
4.1.6 Load Cases Rack modules B2, C1 and H1 (see Figure 1) were analyzed to show that structural integrity is maintained during a seismic event. The Licensee was asked to provide the basis for seicetion of these specific racks and gave the following information:
i Module B2 is representative of a region 1 rack. It is the largest region 1 rack and is located in a corner of the pool.
Module G1 is a large region 2 rack located in a corner of the' pool wall and the cask area wall. This rack has six feet, two of which have an initial gap and are designed to come into contact with the floor only when rocking is sufficient to close the gap. The eccentric placement of its main support legs causes this rack to be relatively more prone to rocking, thus resulting in potentially higher displacements and stresses than a more conventional region 2 rack.
Module H1 is a region 2 rack with a cut-out and one additional support foot. For conservatism, the rack was considered to have' 104 cells loaded with fuel but used a planform for analysis that was less stable than the planform actually present.
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( i For each rack module, several analyses were performed to investigate the variations (fully loaded,in friction half fullcoefficient and empty).(C0F = 0.2 and 0.8) and fuel load condition The Licensee's choice of modules does not cover every configuration but the selection was based on reasonable conservative considerations such as large weight and tendency to rock. All of~the rack modules analyzed are located next to a pool wa13 or corner. Modules in this area would have less.
significant fluid coupling forces. The variation in friction coefficient and fuel load cover a reasonable range of conditions.
- . a 4.1.7 Analysis Method Once the rack seismic models were assembled, equations of motion of the system were written and solved using the DYNARACK computer program. The analysis method is based on the component element method of analysis described in Reference 13. The solution of the problem involves the following steps
- 1. Development of a mathematical model of the rack structure in terms of lumped masses, non-linear springs, fluid,' coupling elements, and provisions for three dimensional kinetic degrees of freedom.
2.
Developmentofequationsforthekinetic,energiesoftbrack,thefuel assemblies, and the entrained and.coupli~ng fluid energies.
3.
Application of Lagrange's formula' tion to assemble the displacement coupled secondorderdifferentialequationsintheprescribedgeneralized coordinates. The set of equations are then numerically solved by the DYNARACK computer program. "
The Licensee was asked to provide additi.onal information on the DYNARACK program and its verification. This program' l.
program which has been utsed and accepted.by'$s NRC inaprevious refine version of the DYNAHIS fuel rack analyses. Both programs provide the numeric)1 solution for non-linear models i of structures under time history inputs. Th$ Licensee stated that verifi-cation of the DYNARACK program was carried out in accordance with Quality AssuranceProceduresfollowing10CFR50, App (ndixB. Validation of DYNARACK results involves: (1) comparison with analyfical solutions and with numerical solutions obtained from other computer codes's and (2) manual calculations of mass atrix terms and comparison with results DYNARACK. #
determined internally by Based on the information provided, the a element method and use of the DYNARACKam progj.pp11 cation to analyze of the component the non-linear lumped mass models of the fuel racks is acc#ptable.
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4 1.8 Analysis Results The DYNARACK program computed displacements and element forces at each '
instant of time during the earthquake.
were computed based on the nodal forces. Stresses at critical-rack locations These stresses were ':hecked against the design limits. Stresses were presented in ' terms of highest stress factors for each load case. Stress factors R1 through R6 were defined as the max-imum computed stress to its allowable value. The stress limits were derived i
from the ASME Code,Section III, Subsection NF, in conjunction with material properties from the Section III appendices and supplier's catalog. The f aulted condition (Level D) limits from Section III, Appendix F, were used for the SSE allowables.
The stress factors were defined as follows:
R1- Ratio of direct teasile or compressive stress on a net section to its allowable value (note support feet only support compression)
R2- Ratio of gross shear on a net section to its allowable value R3- Ratio of maximum bending stress due te bending about ..se x-axis to its allowable value for the section -
R4- Ratio of maximum bending stress due to bending about the y-axis to its allowable value R5- Combined flexure and compressive factor (as defined in ASME Code Section III, Appendix XVII)
R6- Combined flexure and tension (or compression) factor (as defined in ASME Code,Section III, Appendix KVII)
The limiting value of each stress factor is 1.0 for OBE conditions. For SSE conditions, the limit is 2.0 for the rack material and upper part of the support feet, and 1.53 for the lower support feet. i Maximum stress factors for the rack base and support feet for each load case are presented in Table 4.
The Licensee stated that the critical stress factors reported for the support feet were all for the upper segment of the feet and should be compared to a limiting value of 2 0. Table 4 also presents maximum fuel assembly-to-cell impact loads, rack-to-rack impact loads and rack-to-wall impact loads. Table 5 presents maximum rack displacements and floor loads.
In addition to determining stress. factors, the Licensee performed additional calculations to evaluate the adequacy of welds, the effects'of rack-to-rack and rack-to-fuel impact loads, and other local effects. These calculations were not included in the Safety Analysis Report (Ref.1). During the audit at Holtec International, sample calculations were reviewed. Table 6 summarizes the safety factors in critical rack locations.
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4.1.9 Evaluation of Results The results of the Licensee's seismic analysis indicated that all stresses in the racks would meet their allowables, impact loads on fuel assemblies would not damage the fuel, and rack displacements would not be large enough to result in impacts with the pool wall. However, considering the potentially unconservative modeling assumptions discussed in Sections 4.1.1 and 4.1.2 regarding multiple rack effects, rattling fuel mass representation, and fluid l
coupling considerations, it was judged prudent to have the Licensee perform additional studies to address the questions raised. An issue of particular concern was the possibility that the single rack models would underpredict displacements of racks adjacent to pool walls and that rack-to-wall impacts would occur. The walls were not designed to accommodate seismic impact loads from the fuel racks and damage to the walls or liner could result in unacceptable leakage of water from the pool.
The additional studies performed by the Licensee are discussed in Section 4.2.
The evaluation of their results and the overall assessment of the seismic analysis results is given in Section 4.2.3.
4.2 Additional Fuel Rack Seismic Studies As a result of questions raised during the review of the fuel rack dynamic analysis model (Section 4.1.1), the Licensee performed additional analyses.
Single rack model studies were carried out to address questions regarding the l I
adequacy of treating the fuel assemblies as five independent rattling masses and using twice the fuel weight in the models. Multiple rack studies were performed in response to questions regarding the adequacy of a single rack model in predicting forces and displacements that would occur if multiple rack effects were considered. A description of these additional analyses and their results is discussed below.
4.2.1 Single Rack Studies Two additional seismic analyses of single rack models were performed for a fully loaded G1 rack with coefficient of friction equal to 0.8 and a fuel weight that per cell equal to 1300 lbs. The design basis analysis had indicated overallthis load case was the most critical case which predicted the highest response. In the first additional analysis (Case 1), the fuel was modeled in the same manner as the design basis analysis, i.e. as five inde-pendent rattling masses. In the second run (Case 2), the five fuel marses were connected by springs, thus providing a beam representation of the fuel assemblies. The springs did not represent the actual flexural rigidity of a fuel assembly but were based on the properties of a fictitious channel around the assembly.
This flexural rigidity appears to be of the same order of mag-nitude study. as the actual flexural rigidity and is judged to be reasonable foc this 14
e , , ,
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The results of the single rack model studies are presented in Table 7.
Key forces, stresses and displacements are compared. The Case 1 versus Case 2 comparison indicates generally comparable results. .The elastically coupled mass model (Case -2) results do not exceed the independent mass model (Case 1) results by more than 15%. However, the results of both cases are clearly enveloped by the design basis case by significant margins. Therefore, this study demonstrates that modeling the fuel with twice its actual weight provides a significant level of conservatism which adequately compensates for the smaller potential unconservatism of modeling the fuel 'as independent rattling masses.
4.2.2 Multiple Rack Studies The Licensee performed additional seismic analyses of a row of four racks to investigate the adequacy of the design basis single rack models in pre-dicting the response of fuel racks in the actual multiple rack fuel pool environment. An issue of particular concern was the possibility that in a .q multiple rack environment, the peripheral racks may hit and damage the walls of the pool.
The four racks studied were racks A1, A2, B1 and 52, next to the south pool wall shown in Figure 1. A simplified planar two dimen'sional model of'the row of racks was developed. Each rack was represented by a four degree of freedom model representing horizontal.and vertical translations of the rack, planar rotation (rocking) of the rack, and horizontal translation (rattling) of the fuel assemblies. The racks were assumed to be fully loaded with fuel using the nominal fuel weight (1250 lb/ assembly) which is half the weight used in the single rack design basis models. Support spring constants, impact spring constants and gaps were consistent with the design basis models. Fuel to cell fluid coupling coefficients were reduced to '50% of the " blunt body" )
value in an attempt to compensate for the potential' overprediction of fluid coupling forces predicted by the design basis models as discussed in Section 4.1.2.
Runs were made for both the 0.2 and 0.8 coefficients of friction. The i model. loading of the E-W and vertical SSE were applied simultaneously to the seismic 'l I
1 The key responses were corpared with the corresponding responses from the single rack design basis analysis of the B2 rack. These results are presented in Tables 8 and 9. The Licensee stated that the results support the conserva-tism of the design basis model.
Both displacements and impact loads were pre-dicted to be lower by the multiple rack model. The smaller displacements .
supported the conclusion that the peripheral racks would not hit the pool walls.
During the course of the analysis, the Licensee decided to modify the side gap spacing between the pool wall and the peripheral racks from 4.5 inches to 5.5 inches. The multiple rack analysis was rerun to reflect the revised spacing. A comparison of responses between the two multiple rack studies is.
presented in Table 10.
The results showed slight increases in responses but 1
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the loads and displacements were still enveloped by the single rack design basis results by significant margins.
In evaluating the results of this study, several factors had to be weighed. The row of racks that was selected for analysis was a representative-row and was not expected to have the highest response. The Licensee therefore i made an appropriate comparison when he compared the results of this study to the results of the B2 rack design basis model. It was important to evaluate the results on a comparative basis and recognize that they are not worst case results.
It should also be recognized that the results may be somewhat unconserva-tive because the model assumed planar two-dimensional motion. In this type of.
model, only one horizontal component of the earthquake could be applied.
Three dimensional cross-coupling effects could not be accounted for. Never-theless, it is reasonable to expect this 2-D multiple rack model to capture the primary response and potential interaction effects of a row of fuel racks in one direction.
4.2.3 Overall Evaluation of Seismic Analysis Resnits The results of the additional studies presented in Tables 7 through 10 support the adequacy of the design basis (single rack model) results. Both the single and multiple rack models used in these studies utilized actual fuel weight instead of twice the fuel weight as used in the design basis models.
It appears that the high fuel weight was the most significant contributor to )'
the conservatism of the design basis model results. Further studies would be required to prove that single rack models using actual fuel s 'ights would i always give conservative results. However, for this applicati n, these '
studies have provided a reasonably high level of confidence in che adequacy of the results. A review of the safety factors predicted by the design basis models (Tables 4 through 6) provide further assurance that the racks are designed with sufficient safety margin to compensate for uncertainties in the seismic analysis.
Based on the results of the Licensee's seismic analyses, it is concluded that during an SSE, the fuel racks will maintain their structural integrity, fuel assemblies will not sustain damage, and rack displacement will not be large enough to result in pool wall impacts.
4.3 Thermal Analysis Weld stresses due to heating of an isolated hot cell were computed. The analysis assumed that a single cell is heated over its entire length to a temperature above the value associated with all surrounding cells. No thermal gradient was assumed in the vertical direction. Using the temperatures associated with this unit, weld stresses along the entire cell length were found to be below the allowable value with a safety factor of 2.2 as indicated in Table 6.
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4.4 Fuel Handling Accident Analyses l
The Licensee performed structural analyses and evaluations for three post-ule.ted fuel handling accidents. The accidents and the analysis summaries were described in the Safety Analysis Report as follows:
- 1. Dropped Fuel Accident I A fuel assembly is dropped from 36 inches above the module, falls into a cell, and impacts the base. The final velocity and total energy at impact was calculated. To study the baseplate integrity, it was assumed that the energy was directed toward punching of the baseplate in shear and thus was transformed into work done by the shear stresses. It was determined that shearing deformation of the l baseplate was less than the thickness of the baseplate so it was j concluded that local piercing of the baseplate will not occur. j Direct impact with the pool liner would not occur. The 4 suberiticality of the adjacent fuel assemblies would not be violated. ;
j
- 2. Dropped Fuel Accident II '
One fuel assembly drops from 36 inches above the rack and hits the 1 I
top of the rack. By applying an energy balance approach, it was l determined that permanent deformation of the rack would be limited to j the top region such that the rack cross-sectional geometry at the l level of the top of the active fuel and below is not altered. The region of local permanent deformation was shown not to extend below six inches from the rack top. f
- 3. Jammed Fuel Handling Equipment A 4000 pound uplift force was applied at the top of the rack at the weakest storage location. The force was applied on one wall ,of a storage cell as an upward shear force. The plastic deformation was found to be ;
limited to the region well above the top of the active fuel.
The Licensee concluded that these analyses proved that the rack modules are engineered to provide maximum safety against all postulated abnormal and accident conditions. During the audit, the Licensee was asked to provide the calculations for the dropped fuel accident I for a detailed review. The review and evaluation of this calculation is discussed below.
A key element of the dropped fuel accident analysis was the calulation of impact velocity. The model for predicting velocity treated the fuel assembly as a free body falling through a channel. The model considered gravit'y and fluid forces and accounted for virtual mass effects. Fluid forces were determined by applying basic fluid mechanics laws of. continuity and energy.
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k However, the model was found to be unconservative in calculating the pressure I build-up within a cell. The model assumed that as a fuel assembly falls J through a rack cell, .all of the vater in 'the cell is forced out. through the baseplate holes at the bottom. Flow through the-fuel assembly was neglected.
Since that flow area is significant, the model may have overpredicted the i fluid retarding force .and underpredicted ' the impact velocity and kinetic '
energy of the fuel assembly as it hits the baseplate.
In evaluating the potential penetration of the baseplate, the kinetic energy of the fuel assembly was set equal to the work performed as a slug is punched out of the baseplate. The calculation showed that the depth of pene-tration is less than the plate thickness and concluded that penetration would not occur. Furthermore, the Licensee stated that the purpose of the calculation was only to show that'there is no danger to the pool liner. In the event of a dropped fuel assembly, the base plate could be expected to plastica 11y deform and separate from the rack cells but this would not affect center-to-center spacing. The Licensee stated that the baseplate, even with plastic bending occuring, would not touch the liner floor.
A number of weaknesses were noted in the evaluation: 1) The equation for work required to penetrate the plate lacked sufficient experimental verification. The use of empirical penetration formulas would have been more ]
appropriate. 2) The shear area was underpredicted. This area was based on the solid square cross-sectional area.of the fuel' assembly. In reality, the fuel assembly rests on four legs with a much smaller contact area. 3) The conclusion that the baseplate would not contact the floor was not substantiated by any calculation for plastic deformation or ductility ratio.
Although a number of weaknesses were identified, there were also a number of conservatism which must be considered. The fuel weight used in the velocity and energy calculations was 2500 pounds which is nearly twice the i actual weight. The fuel was assumed rigid for impact stress calculations.
All of the impact kinetic energy was assumed to be directed toward punching of the baseplate in shear. None of this energy was directed toward bending of j the plate or compressing the fuel.
i I
To evaluate the final conclusions of this analysis, BNL performed some simplified calculations using bounding assumptions. Kinetic energy at impact was calculated on the basis of the actual fuel weight and velocity of a free-f alling body in air. The resulting kinetic energy was approximately twice that used in the Licensee's calculations. Baseplate penetration was evaluated by empirical penetration formulas for steel targets commonly used for missile penetration analysis in the nuclear industry. Both the Ballistic Research Laboratory (BRL) formula and the Stanford Equations were applied (Ref. 14). The missile contact area was based on the fuel assembly leg area rather than the total cross-sectional area. The results of this analysis concurred with the Licensee's conclusion that baseplate penetration would not occur.
18 l .. . . .
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e The Licensee's conclusion that the baseplate would not contact the pool floor could not be verified by a simplified analysis due to the complex nature of the structure. However, it should be noted that since the pool floor had been shown capable of withstanding a fuel cask drop, it would be reasonable to conclude that the floor has sufficient strength to withstand the impact load resulting from the drop of a single fuel assembly.
4.5 Spent Fuel Pool Analysis 4.5.1 Loads and Load Combinations The reanalysis of the spent fuel pool considered the following design loads:
Structural Dead Load (D)
Live Load (L)
Normal Operating Thermal Loads (T)
Accident (Loss of Fuel Pool Cooling) Thermal Load (TA )
Fuel Cask Drop Load (M)
The following load combinations, from the St. Lucie, Unit No. 1, Updated FSAR, Section 3.8.1.5, were considered:
a) Normal Operation 1.5 (D+T) + 1.8L b) OBE Condition 1.25 (D+T+0BE+0.2L) c) SSE Condition 1.05 (D+T+0.2L) + 1.0 SSE d) Accident and Cask Drop 1.05 (D+Tg+0.2L)
For the evaluation of the liner and liner' anchors, the above load combinations 1.0. were applied except that load factors for all cases were equal to Linear analyses were performed initially to determine the critical load combinations. As a result, the following loading cases were selected for the non-linear concrete cracking analysis:
- 1. 1 5D + 1.BL
- 2. 1.05 (D + Twinter + 0.2L) + 1.0 SSE
- 3. 1.05 (D + Tsummer + 0.2L) + 1.0 SSE 19
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- 4. !
1.05 (D + 0 2L) + 1.0 SSE l
- 5. 1.05 (D + TA + 0.2L)
- 6. -1.05 (D + Twinter + 0.2L) + 1.0M
- 7. 1.05 (D + 0 2L) + 1.0M !
4.5.2 Spent Fuel Pool Structure Analysis !
A finite element model of the lower portion of the spent fuel pool structure was developed. Since the effect of the additional. fuel rack load on the pool floor is limited to the mat in the pool area, the upper. portion of the pool walls was not reevaluated. The model included the lower portion of the walls up to elevation 45.25 ft, the pool floor and the underlying soil.
The structural components included in the model are shown in Figure 16. A computer plot of the finite element model is shown in Figure 17 which shows ,
the overall view of the model indicating the composite of the four exterior and one interior walls.
In this analysis, the EBS/NASTRAN prog ram was used. The Licenaea was asked to provide additional information on this computer program. This was provided in Reference 3c.
EBS/NASTRAN is an enhanced NASTRAN program developed by Ebasco. It has all of the NASTRAN capabilities plus additional features. One of the additional features is the ability to perform concrete cracking analysis. This feature incorporates a special plate element which consists of a user-specified number of layers, each having~a different proportion of steel to concrete area, representing'the presence of reinforcing steel. Each layer will crack or re-close according to the stress-strain relationships of the concrete and steel. .Thus, a cracking pattern and stress redistribution can be determined. A verification problem was submitted which.
demonstrated good agreement of analytical results with experimental data. 3 The maximum stress results in the concrete and rebars from the nonlinear-analysis of the seven load cases are presented in Table 11. The design stress limits described in the St. Lucie Unit 1 FSAR were used in the evaluation.
The capacity of all sections was computed.in accordance with ACI 318-63 Part IV-B, Ultimate Strength Design. Table 11 indicates minimum safety factors for each loading case. Safety factor is defined as allowable stress divided by maximum actual stress including load factors.- The smallest safety factors are 1.10 for reinforcement bar tension, 2.65 for concrete compression, and 1.05 for concrete shear. Based on these results, it can be concluded that the spent fuel pool structure can accomodate the revised loads.
4.5.3 Pool Liner and Anchorage Analysis The liner and its anchors were evaluated for the temperature load, the strain induced load due to the deformation of the floor, and the horizontal seismic load. The POTSUKF computer program was used for the liner buckling 20
m k '
analysis due to the temperature and strain induced loads, The Licensee was asked to provide additional information on this computer program. -This was provided in Reference 3c. POSBUKF is a program developed by Ebasco to examine the elastic post-buckling behavior of a flat plate subjected to thermal and lateral loading using an energy method approach. The program' determines the deflected shape of a buckled plate by minimization of potential energy, and from this calculates plate stresses utilizing strain-displacement and stress-strain relationships for the particular case under study. The program was verified by comparison of test problem results. to hand calculation results.
The liner anchors were evaluated for the unbalanced liner in-plane force due to the temperature and strain induced loads, as well as horizontal seismic in-plane shear force.
The acceptance criteria for the liner and anchors was in accordance with the requirements of ACI-ASME Section'III, Division 2, Subsection CC for containment liners. The critical loading case for the liner was the case which included accident thermal load. The analysis showed that the maximum j calculated strain was below the Code Strain allowable with a safety factor of 5.2. The buckling analysis indicated that the liner plate would not buckle. l' Two loading conditions were considered in the liner anchor evaluation; one was the strain-induced load which produced the unbalanced in-plane force at the edge of the pool area, and the other was the horizontal seismic load transmitted through friction between the rack support and the liner. The analysis indicated that Code allowables were met with minimum safety factors of 2.5 for the strain-induced load case, and 1.33 for the seismic load case.
Based on these results, it can be concluded that the fuel pool liner and anchorage can accommodate the revised loads.
5.0 CONCLUSION
S Based on the review and evaluation of the Licensee's Safety Evaluation Report and additional information provided by the Licensee during the course of this review, it is concluded that the proposed St. Lucie Unit 1 fuel racks have sufficient structural capacity to withstand the effects of all required environmental and abnormal loadings discussed in this report. Impact loads generated by the closing of fuel assembly to fuel rack cell gaps during the SSE would not lead to damage. Furthermore, the existing spent fuel pool should have adequate capacity to accommodate the increased loads resulting from the storage of more fuel assemblies in the pool.
All concerns related to the adequacy of the dynamic single rack design basis models including multiple rack effects (Section 4.1.1), rattling fuel mass representation (Section 4.1.1), and fluid coupling considerations (Section 4 1 2) were resolved by additional studies performed by the '
Licensee. These studies (Section 4.2.) investigated multiple rack effects and the sensitivity of model variations. They demonstrated that the single rack design basis models predict conservative seismic loade and displacements.
21 I
4 Although the studies were limited in scope, they provided evidence which indicated that the most significant contributor to the conservatism of the design basis models was the use of twice the fuel assembly design weight in the models. An additional analysis of a single rack model which used the actual fuel. weight predicted displacements and impact loads which were approx-imately half of the corresponding design basis model results. Analysis of multiple rack models which also used actual fuel weights showed similar trends in the results. Thus it was judged that the design basis models have suffi-cient conservatism to compensate for potential underprediction of response due to the modeling concerns discussed in this report.
i I
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l l
22 l
l
l 4
6.0 REFERENCES
- 1. Florida Power and Light Company, St. Lucie Plant - Unit No. 1 Spent Fuel Storage Facility Modification Safety Analysis Report, Docket No. 50-335.
- 2. NRC letter, G. Bagchi to E. Tourigny, " Request for Additional Information
- Proposed License Amendment - Spent-Fuel Rerack St. Lucie Unit 1 Docket No. 50-335 TAC #65587", dated August 7,1987.
3a. FP&L letter L-87-422, C.O. Woody to USNRC, "St. Lucie Unit 1 Docket No.
50-335, Spent Fuel Pool Rerack - Design and Analysis", dated October 20, 1987.
3b. FP&L letter L-87-535, C.O. Woody to USNRC, "St. Lucie Unit 1. Docket No. l 50-335, Spent Fuel Rerack-Design and Analysis," dated December 23, 1987. I 1
3c. FP&L letter L-37-536, C.O. Woody to USNRC, "St. Lucie Unit 1, Docket No.
50-335, Spent Fuel Rerack-Design and Analysis," dated December 23, 1987 4.
USNRC letter to all power reactor licensees, from B.K. Grimes, dated April 14, 1978, "0T Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", as amended by the NRC letter dated January 18, 1979.
- 5. US Nuclear Regulatory Commission, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," NUREG-0800, Revision 1 July, 1981.
- 6. Ebasco Drawing 8770-G-830, " Fuel Handling Bldg Spent Fuel Pit Liner",
Sheet 1, Rev. 4; Sheet 2, Rev. 2; Sheet 3, Rev. 4; Sheet 4, Rev. O.
- 7. Joseph Oat Drawing D-8286, Rev.1, " Details Region I, Spent Fuel Storage Racks".
- 8. Joseph Oat Drawing D-8288, Rev.1, " Plan Diagram of Can to Gap Element Joints, Region I, Spent Fuel Storage Racks".
- 9. Combustion Engineering Drawing E-13172-161-101, Rev. 5, Sheet 1 of 2,
" Fuel Bundle Assembly".
- 10. R.~. Fritz, "The Effects of Liquids of the Dynamic Motions of Immersed Solids", Journal of Engineering for Industry. Transactions of the ASME, February, 1972, pp 167-172.
- 11. K.P. Singh and A.I. Soler, " Dynamic Coupling in a Closely Spaced Two-Body System Vibrating in a Liquid Medium: The Case of Fuel Racks", 3rd International Conference oa Nuclear Power Safety, Keswick, England, May 1982.
23
t 1
1 4,
- 12. E. Rabinowicz, " Friction Coefficients for Water Lubricated Stainless Steels for a Spent Fuel Rack Facility", a report for Boston Edison Company, MIT, 1976.
- 13. S. Levy and J.P.D. Wilkinson, "The Component Element Method in Dynamics",
McGraw-Hill, 1976.
- 14. R.C. Gwaltney, " Missile Generation and Protection in Light-Water-Cooled I Power Reactor Plants", ORNL-NSIC-22, September 1968.
l l
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L___---____--___-_--_-______-_---_---- - _ _ - - -
, s g
^
TABLE 1 TABLE OF MODULE DATA NO. 0F N0. 0F . ;
CELLS CELLS TOTAL NO. l NO. OF IN N-S =IN E-W OF CELLS MODULE I.D. MODULES DIRECTION DIRECTION PER MODULE Region 1 2 9 9 81 Al to A2 ,
l Region 1 2 9 10 90 B1 to B2 Region 2 4 13 9 117 C1 to C4 l
Region 2 3 13 8 104 D1 to D3 Region 2 2 11 8 88 El to E2 Region 2 1 12 8 96 F1 l
Region 2 2 12 9 108 G1 to G2 Region 2 1 13 8 96 Hi '
25
- j
. . <, +
1 a -
TABLE 2 i
MODULE DIMENSIONS AND WEIGHTS NOMINAL CROSS-SECTION -ESTIMATED DRY DIMENSIONS
! . WEIGHT (1bs)
MODULE I.D. N-S E-W ,PER MODULE i
Region 1 90-1/4" 90-1/4" 26,700 Al to A2 s Region 1 90-1/4" 100-7/16"- 29,800 B1 to B2 Region 2 115-11/16" 1/6" '24,100 C1 to C4 Region 2 115-11/16" 71-3/16" 21,500 D1 to D3 Region 2 97-7/8" 71-3/16" 18,200 El to E2 i
Region 2 106-3/4" F1 71-3/16" 19,800 Region 2 106-3/4" 80-1/16" 22,300 C1 to 02 Region 2 115-11/16" 71-3/16" H1 19,800 1
26 E_-______..-----__------_--
- - - - - - - - - - -- - - - ~ ~
c TABLE 3 RACK MODEL PARAMETERS-Rack Module H1 B2 Gl*
KI (#/in) .359 x 106 .310 x 106 .372 x 106*** .
Kw (#/in) .1 x 10 7 .1 x 107 .1 x 107 **
I Kf (#/in) .221 x 1010 .221 x 1010 .221 x 1010 Kd (#/in) .112 x 107 .109 x 107 .123 x 107 l
KR (#in) .567 x 108 .567 x 108 .567 x 108' rad .
I 1
h (in) 6.125 6 125 6.125 l 1
1 H (in) 169 169 169 W4 (1b) 19800 29800 22300 Wf (1b) 260000 225000 270000.
Lx (in) 71 90 80 Ly (in) 116 100 107
- 6 support feet (.1875" initial gap on.2 of 6 supports)
- Where 2 racks are adjacent, gap between base plates =
.625"; gap between girdle bars = .375"
- Nominal gap between cell vall and fuel assembly = .125" l
' KI - fuel assembly-to-cell wall impact spring rate Kw - rack-to-rack or rack-to-wall impact spring rate Kf - friction spring rate (active prior to sliding)
KR - 8pring rate representive of rotational resistance between liner and support leg Kd - support leg axial spring rate h - length of support leg H - height of rack above base plate ,
Wr - Weight of rack without fuel Wf - Weight of fuel Lx - Planform dimension (X-direction)
Ly - Planform dimension (Y-direction) 27
. 7 7 i 9 2 6 7 1 1 0 6 2 2 6 2 8 3 .
9 6 1 2 2 4 0 5 5 4 2 7 0 2 3 7 1 1 2 6 1 1 _
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) 6
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O cp R 3 0 1 2 5 3 5 2 3 _
T au C RS =
A 4 7 9 8 9 5 6 _
F rr 2 7 4 2 9 3 7 9
5 1 7 6 2 5 S
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6 4
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0 0 2 T
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UL / / / / / / / / / /
9 2 1 0 9 8 7 6 3 S ( 1 9 9 2 5 3 1 0 7 1 T S 7 7 5 2 2 9 1 R 2 3 3 1 2 0 0 1 3 2 L
U S
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3 0 2 E C 5 7 6 7 9 3 R A 2 8 3 3 1 2 7 6 9 _
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S 0 0 8 2 3 0 1 I S / / / / / / / /
S Y
L S
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N S -
A t D R 7 1 7 9 6 C N 6 2 1 1 0 6 4 8 8 2 2 6 5 9 I A 2 4 4 1 2 4 1 4 2 4 6 1 _
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0 0 0 0 L cl *4 5 5
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A RtB 1 1 1 1 0 0 0 0 4 4 T /c/ xx x x 1 1 0 1 1 1 1 1 0 0 k aB 43 2 9 x x / x x x x x 1 1 cpC 96 2 5 1
5 4 0 1 2 0 0 5 x x am 97 0 3 9 6
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x 1
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es 6 3 1 4 9 0 6 6 6 4 0 0 7 P usomo FAtIL 9 4 3 0 4 0 9
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9 9 1 9 9 2 7 8 8 7 9 8 1 R A
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/ n 8 8 8 2 8 e = = =o = = = = = =
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FI 5 1 1 2 1 0 2 2 1 1 T 0 x 1 2 9 6 3 XR AE 1 2 /. /. /. /. /. /. ~/. /. /.
x 9 3 5 3 /. /. 1 3 7 7 7 MV 7 4 7 8 4 3 8 3 8 4 5 1 7
2 7 5 6 6 8 5 1 8 0 2 4 2 8 m 8
7 9 0 7 3 4 7 3 4 6 7 2 _
7 1 6 0 0 3 2 4 2 0 4 1 2 2 1 1 3 6 1 2 1 2 1 1 Y _
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A R 5 5 5 5 5 4 4 5 5 5 5 5 M O) T 0 0 0 0 0 0 0 0 5
M Of E 1 1 1 1 1 1 0 0 0 0 0 1 1 U
S L( E F F x x x x x x x x 1
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S XA4 4 8 3 9 T AO 1, 9 8 3 9 3 9 7 5 5 1 2 0 L S ML 4 5 5 2 3 7 8 3 U D 4 2 5 2 3 S A E O R L S R .)
I O TN 1 1 1 1 S O 1 1 1 1 2 1 RI - - - - - -
Y L E( 0 0 - - - -
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_ AS 3 9 2 2 1 5 4 C MI 7 0 2 5 3 7 3 4 8 3 _
_ I S D 9 9 1 1 9 4 M T 9 1 8 4 2 3 3 6 4 S N m I E E M .
m S E P _
C S _
K A I) _
C L DN 5 0 7 1 4 8 A P I 0 1 7 8 8 6 8 2 1 R S .( 4 7 4 9 8 8 3 8 4 3 3 1 1 0 7 8 2 4 _
I X 6 6 6 0 2 7 5 1 2 D 4 3 0 2 2 AY 4 2 2 2 2 _
_ - MD _
u 5 '
E P L S B I) 4 7 7 A DN 8 9 0 6 7 7 T I 8 1 0 4 7 2 2 7 1
.( 4 6 2 1 2 6 1 0 9 0 3 8 7 5 4 5 1 3 X 4 3 8 0 1 7 AX 3 8 1 1 2 5 3 5 2 2 1 1 MD -
e c
n d e l l l l l a
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/
1
/ i o C 1 tL .
r D 2 8 8 2 8 2 8 2 8 8 8 er A 2 8 Va O = = = = = = = = = = =
e s L U p p p U U U p p p p p
= = =h U S -
L A
C=
I R TA RE E EH .
L 1 1 1 1 1 2 2 Z 2 2 1 VS _
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-TABLE 6
SUMMARY
OF SAFETY FACTORS IN CRITICAL FUEL RACK LOCATIONS SAFETY -
ITEM / LOCATION FACTOR COMMENTS i
Support foot to baseplate 2.44
- veld stress I-Cell'to baseplate weld stress 3.15 Cell to gap channel weld 2.94
~
Stress due to ' seismic loads il stress cell to gap channel weld 2.20- Thermal stress due to effects stress of isolated hot cell
. l 1
Impact load on girdle bar 2 17 i 1
1 Cirdle bar shear stress 1.70 1
Cell wall stress due to h 2 54 '
girdle bar impact load <
l
\
Impact load between fuel 3.58 assembly and cell wall Based on cell well limit load !
i Impact load between fuel 1 51 Based on. plastic deformation ^
assembly and cell wall of fuel spacer grids * ,
Shear load on baseplate 3.0 near a support foot Compressive stress in cell 4.56 wall Based on local buckling considerations Rack to wall impact loads -
No Impacts with pool walls occur at any location 30 Sa
- y factor on fuel rod crushing is significantly higher !
i l
l
I TABLE 7 i
RESULTS OF SINGLE RACK STUDIES I FULLY LOADED G1 RACK WITH COF = 0.8 >
i INDEPENDENT ELASTICALLY COUFLED ' DESIGN BASIS ITEM FUEL MASSES FUEL MASSES MODEL (1300 lb/ fuel) .(1300 lb/ fuel) (2500#/ Fuel)
Fuel / Rack 453.3 514.4 1221.3 Impact (#/ cell) l l E sck/ Rack 7.133x104 /0. 6.249x104 /0. 1.359x10 /o.o 5
Impact jl (BP/GB) (#) ..
l Rack / Wall 0./0. 0./0. 0./0. -
Impact .
(BP/GB) (#)
R6 Stress .401/.736 .421/.795 .576/1.273 Factors (Rack Base /
Support)
Max. Disp. .5717 .5709 1.7407 DX (in.)
i Max. Disp. .3230 .3479 .6147 DY (in.)
l Max. Vert. .0823 .0802 .0909 Disp. (in.)
Max. Floor 3.934x105 3.800x105 5.877x105 Load (4 Feet) (#)
Max. Floor 180237./ 190218/ 279673/
Load (#) 108454. 110134 186242.
Vertical /
Shear 31
^
1
. k TABLE 8 RESULTS OF HULTIPLE RACK STUDIES FULLY LOADED1 A , 2A , B1 , B 2, RACKS WITH C0F = 0.2
, SINGLE RACK B2 ITEM
' MULTI-RACK MODEL Design Basis-Model Rack / wall at girdle Of Of bar - impact ' load Rack / Rack at girdle 0#
bar - impact load .5312.x 105 Rack / wall at base- Oi plate - impact load Of Rack cell wall to '613. 891.
fuel assembly i
(per cell - impact load) l Vertical load on .6425 x 105 lb 1'
pool floor from 1.378 x 105 lb one foot .,
1 Rack / Rack at Of baseplate - impact Of load ,
- i Nax. E-W rack .126 inch displacement at .2088 inch !
top of rack i
32
v
)
t
.. i TABLE 9 i
RESULTS OF MULTIPLE RACK STUDIES FULLY LOADED1 A , A 2 , B 1, B2 RACKS WITH COF = 0.8 ;
SINGLE' RACK B2 ITEM MULTI-RACK MODEL Design Basis Model !
i Rack / wall at girdle Of Of 1
I bar - impact. load l
Rack / Rack at girdle 0#
bar - impact load 1.17 x 105 Rack / wall at base- . Of Of plate - impact load i
l Rack cell wall to 612.
fuel assembly 974. lb' l (per et11 - impact i load)
Vertical load on .715 x 105 lb pool floor fro,n 2.231 x 105 13 I. one foot Rack / Rack at Of baseplate - tapact Of l load Max. E-W rack .091 inch displacement at .4238 inch top of rack 33
+- ,
(
TABLE 10 RESULTS OF MULTIPLE RACK STUDIES SIDE CAPS (SG) = 4.5", 5.5" FULLY LOADED1 A , A2 , B 1, B2 RACKS COF = .8 C0F = .2 ITEM SC = 4.5" SG = 5.5" SC = 4.5" SG = 5.5" Rack / Fuel 612. 604. 613.
Impact Load 618.
(per cell)
Rack / Wall 0. O. O.
Impace at O.
Girdle Bar -
Rack / Wall 0. O.
Impact at O. 0.
Baseplate Rack / Rack O. O. O. O.
l Impact at l Girdle Bar l
Rack / Rack O. O. O. O.
Impact at Baseplate Max. Support 71500. 78400. 64250. 65500.
Foot Load (1 foot)
Max. Horiz. .0911 .1196 .126 Disp. at .1491 Top of Rack (in.)
l l
l 34
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S c 8 7 5 4 6 5 2 nF 4 0 1 5 8 0 6 sos uC 1 1 1 1 1 1 1 m
if xo a
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C A v L 8 3 3 4 0 2 U M i eL 3 0 5 4 9 4
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=
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.i FIGURE 1
. SPENT MIEL POOL LAYOUT l
36 1
m ._
~.
t.05" y to 12.- m GIRDLE BAR
' PITCH I l 1 .
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