ML20009B841

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Safety Review of the Design,Operation and Radiation Sections of the General Electric Morris Operation Consolidated Safety Analysis Report
ML20009B841
Person / Time
Site: 07001308
Issue date: 07/31/1981
From: Hammond C, Luk K, Mcbride J
OAK RIDGE NATIONAL LABORATORY
To:
NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
References
CON-FIN-B-0102, CON-FIN-B-102 NUREG-CR-1697, ORNL-TM-439, NUDOCS 8107170235
Download: ML20009B841 (39)
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Y' \\' f 'NW --f QA. A ,fgQY ';.f # ~1 iO\\ ') fU MQ"1& fh,[Q9f M"mQ]'R'n_~hZ g.:l%%'kQ&%% @%ll@&' h. t .y.m Mf. f y' _Q y yW s NUREG/CR-1697 ORNL/TM-439 Dist. Category AN CHEMICAL TECHNOLOGY DIVISION SAFETY REVIEW OF THE DESIGN, OPERATION, AND RADIATION SECTIONS OF THE GENERAL ELECTRIC MORRIS OPERATION CONSOLIDATED SAFETY ANALYSIS REPORT J. P. McBride With an Appendix by K. H. Luk C. R. Hammond Union Carbide Engineering Manuscript Ccmpleted: January 1981 Date Published: July 1981 Prepared for U.S. Nuclear Regulatory Commission Office of Nuclear Material Safety and Safeguards Division of Fuel Cycle and Material Safety Washington, D.C. 20555 NRC FIN No. B0102 OAK RIDGE NATIONAL LABORATORY Oak Ridge, Tennessee 37830 operated by UNION CARBIDE CORPORATION for the DEPARTMENT OF ENERGY iii TABLE OF CONTENTS fang ABSTRACT. 1 1. INTRODUCTION. '1 2. GENERAL DESCRIPTION. 2 3. DESICN CRITERIA AND COMPLIANCE.. 3 4. FACILITY DESIGN AND DESCRIPTION. 7-5. RADIATION PROTECTION. 11 6. ACCIDENT SAFETY ANALYSIS.. 11 7. CONDUCT OF OPERATIONS. 15 8. REFERENCES. 16 APPENDIX. 19 2 y 1 -SAFETY REVIEW OF THE DESIGN, OPERATION, AND RADIATION SECTIONS OF THE GENERAL ELECTRIC MORRIS OPLRATION CONSOLIDATED SAFETY ANALYSIS REPORT-J. P. McBride ABSTRACT A safety review was made of Sections 4 through 9 of the Consolidated Safety Analysis. Report (CSAR) for the GE Morris Operation spent-fuel storage facility. The sections reviewed include Design Criteria and Compliance, Facility Design and Descriptio2, Radiation Protection, Accident Analysis, and Conduct of' Operations. The safety review was performed in accordance with the Code of Federal Regulations, Title 10, Part 72, " Licensing Requirements for the Storage of Spent Fuel-in an Independent Spent Fuel Storage Installation" and contains independent estimations of source terms and dose-commitments from postulated accidents in the storage facility and a struc-tural analysis of the Morris Operation cranes as an appendix. The review confirms that the features of the facility as described in Sections 4 through 9 of the CSAR fulfill the safety requirements of 10 CFR 72, and it is concluded that spent-fuel handling and storage at the Morris Operation do not present significant risks to public health and safety. 1. INTRODUCTION The U.S. Nuclear Regulatory Commission (NRC) in May 1979 requested the Oak Ridge National Laboratory to participate in a safety review of the General Electric Morris Operation as described in the revised Consolidated Safety Analysis Report (CSAR) NED0-21326C (January 1979).1 l The GE Morris Operation is an irradiated-fuel storage facility located near Morris, Illinois. The scope of ORNL's participation in the safety review covered the following sections of the CSAR: 1. Section 4: Design Criteria and Compliance j 2. Section 5: Facility Design and Description ( 3. Section 7: Radiation Protection 4. Section 8: Accident Safety Aaalysis 5. Section 9: Conduct of Operations l 12-The Morris Operation was built as En integral part of the GE Midwest Fuel Recovery Plant (MFRP), Docket No. 50-268, and was licensed for the receipt of spent fuel in December 1971. The MFRP Construction Permit (No. CPCSF-3) was terminated on August 23,1974, but the Materials License. N. SNM-1265 for the receipt and storage of spent fuel containing o up to 100 metric tons (MT) of heavy metal was continued. Spent-fuel storagecagacitywassubsequentlyincreasedfrom100MTofheavymetal to 750 MT. The revised CSAR is a consolidation of all safety analysis -information submitted by the General Electric Company relating to the receipt,' storage, and transfer of irradiated nuclear fuel at the Morris Operation and disregards those features of the MFRP not applicable to fuel storage. The present review was performed in accordance with the Code of Federal Regulations, Title 10, Part 72 (10 CFR 72)3 and includes independent estimations of source terms and dose commitments from postu-lated accidents in.the storage facility. A structural analysis of the Morris Operation cranes is included as an Appendix. A site visit to the Morris Operation was made as a part of the safety review. 2. GtNERAL DESCRIPTION The General Electric facilitiec are located near Morris, Illinois, adjacent to the Dresden Naclear Power Station (DNPS), and consist of the Morris Operation buildings and a training center for nuclear power sta-tion operators - the Boiling-Water Reactor Training Center (BWRTC).* The Morris Opertion is licensed for the receipt, storage, and tranafer of nuclear fuel from boiling-water reactors (BWRs) and pressucized water reactors (PWRs). Operations related to the maintenance of General Electric fuel shipping casks are also conducted at this site. The Morris Operation fuel storage facility includes three intercon-nected water-filled basins with cranes, water treatment systems, and other facilities required to receive irradiated fuel and store it under-water for an indefinite period. The fuel storage equipment in the basins is designed to protect the integrity of the fuel rods during seismic or meteorological events. Special procedures and isolation can be provided for storage of damaged or leaking fuel. Security measures are in effect to protect the facility against unauthorized access. Based on the service life of nonreplaceable components (concrete basin and basin liner), the normal service life of the facility would be more than 100 years, although it is intended for interim storage only. The Morris Operation does not encompass BWRTC activities, although both are General Electric operations. L 1 3 I In December 1975, General Electric was authorized by _ the NRC to increase the fuel storage capacity of the facility from about 100 MI of heavy metal to 750 MT through the installation of a newly designed fuel storage system and appropriate changes in fuel handling and support systems. This project converted the former high-level waste storage basin to a fuel storage basin. The capacity-expansion project was completed in 1976. 3. DESIGN CRITERIA AND COMPLIANCE The Harris Operation is designed to store in a water basin irra-diated, light-water-reactor fuel from nuclear power stations. The design basis fuel in the CSAR was U0, with an initial enrichment of 5% 235U or 2 less, clad in stainless steel, zirconium, or Zircaloy. The fuel was assumed to have been irradiated at power loads up to 40 kW/kg U to an exposure of 44,000 mwd /MTU and cooled 90 days. Fuel, typically received, had exposures of 33,000 mwd /MTU or less and much longer cooling times. As of February 1,1977, the average exposure' of stored fuel was less than 1500 mwd /MTU, and the average cooling time was about 4.7 years. Part 72 stipula-tes that spent fuel cust undergo at least one-year's decay prior to storage in an independent spent fuel storage installation. Hence, independent ana-lyses at ORNL, conducted as part of the safety review, are based on PWR fuel irradiated at 40 kW/kg U to an exposure of 44,000 mwd /MTU and cooled 1 year. Estimates of the radionuclide activity in the model spent fuel, made using the ORIGEN code,4 are given in Table 1. s, 3.1 General Design Criteria Structures, systems, and components important te safety are designed to withstand the effects of natural phenomena (i.e., tornadoes, including tornado winds, hurricanes, earthquakes, etc.), without impairing their capa-bility for safe shutdown, radioace.ive inventory control, and the prevention of significant radiation exposure to operating personnel or the general public. The design criteria presented in the CSAR were evaluated against the requirements set forth in Subpart F, " General Design Criteria," of 10 CFR 72. We found that the design criteria fulfilled those requirements. 4 l Table 1. Radionuclide activity in spent ' fuel af ter one year decay" Conditions: Burnup - 44,000 Mid/NT Specific power - 40 kW/kg Half-Activity Isotope life (Ci/MT) Fission products Kr-85 10.7 y 10,880 Sr-89 50.5 d 5,267 Sr-90 29.1 y 87,430 Y-90 64.0 h 87,450 Y-91 58.5 d 14,220 Zr-95 64.0 d 31,920 Nb-95m 86.6 h 237 Nb-95 35.2 69,330 Tc-99 2.13 x 105 16.6 y Ru-103 39.3 d 3,015 Rh-103m 56.1 m 2,718 Ru-106 1.01 y 407,500 Rh-106 29.9 s 407,500 Ag-110m 249.8 d 3,019 Ag-110 24.6 s 40.2 Cd-113m 14.6 y 87.3 Cd-115m 44.6 d 8.1 Sn-119m 245.0 d 118 Sn-123 129.1 d 642 Sb-124 60.2 d 35.7 Sb-125 2.77 y 15,640 Te-123m 119.6 d 4.5 Te-125m 58 d 3,808 Te-127m 109 d 1,829 Te-127 9.35 h 1,792 Te-129m 33.6 d 30.1 Te-129 69.6 m 19.6 I-129 1.57 x 107 0.0426 y I-131 8.04 d Xe-131m 11.9 d Cs-134 2.06 y 183,600 Cs-137 30.0 y 135,000 Ba-137m 2.55 m 127,700 Ba-140 12.8 d 0.005 La-140 J.2 h 0.005 l t m-5 Table 1.~(continued) Half-Activity Isotope - life (Ci/MT) Fission products Ce-141 32.5 d 721 Ce-144 284.2 d 556,100 Pr-143 13.6 d 0.013 Pr-144m 7.2 m 6,675 Pr-144 17.3 m 556,100 Nd-147 11.06 d Pa-147 2.62 y 101,000 Pm-148m 41.3 d 61.3 Pa-148 5.37 d 3.5 Sm-151 90.0 y 416 Eu-154 8.6 y 16,140 Eu-155 4.96 y 9,345 Eu-156-15.2 d 0.02 Tb-160 72.3 d 58.5 Total activity 2.85 x 106 ,Transuranics Np-239 2.35 d 43.4 Pu-241 14.4 y 155,960 An-241 432 y 383 Am-243 7378 y 43.4 Cm-242 163 d 15,050 cm-243 28.5 y 40.4 Co-244 18.1 y 7,076 Total activity 1.79 x 105 a A. G. Croff et al., Revised Unznium Plutonium Cycle NR and WR Nodele for the ORICEN Computer Code, ORNL/TM-6051 (September 1978). 6 3.2 Tornado Design Criteria Plant structure and components are designed to withstand wind velo-cities of 177 km/h (110 mph) without impairment of safety-relatei functions. They are also designed to withstand the effects of potential wineitorne missi-les and short-term wind velocities (f 483 km/h (300 mph) with pressure dif-ferentials of up to 3 pai without damage to fuel in storage to an extent significantly affecting public health and safety. 3.3 Seismic Design Criteria The main building, including all portions of the structure now used for irradiated fuel storage, was constructed to seismic specifications for a design-basis earthquake (as defined in the CSAR) of 0.1 x g and a maximum (safe shutdown). earthquake of 0.2 x g. The response criteria used in the design calculations are based on the north-south component of the 1940 El Centro, California earthquake normalized to horizontal ground accelerations of 0.1 x g for the design-basis earthquake and 0.2 x g for the maximum earthquake, thus fulfilling the requirements of Appendix A, requirement for seismic design of 10 CFR 72.glants and the licensing 10 CFR 100, for the siting of nuclear power 3.4 Compliance Structural and system resistance to tornado and seismic phenomena 5 and were not were evaluated as part of the safety analyses of the MFRP reevaluated as part of this safety review. However, NRC requested that ORNL perform an independent analysis to determine the effects of a maxi-mum or safe shutdown earthquake on the cask handling crane and support structure and the storage basin crane. The analysis is presented in the Appendix. The results indicate that these systems meet the structural requirements to successfully withstand the stresses induced by a safe shut-down earthquake. The maximum earthquake is equated in Appendix A, 10 CFR 100, with the " Safe Shutdown Earthquake" (SSE) of a nuclear power plant. Unlike the CSAR use of the term and according to Appendix A, the SSE is also commonly referred to as the " Design Basis Earthquake" (DE). An acceptable seismic criterion in 10 CFR 72 for an independent spent-fuel storage installation (ISFSI) evaluated under the criteria of Appendix A, 10 CPR 100, is that the ISFSI-DE be equivalent to the SSE of a nuclear plant (i.e., in the case of the Horris Operation, the maximum earthquake). l l 7 Storage 'of _ spent. nuclear fuel at the Morris Operation was initiated in January 1972. An evaluation of the design and operation of'that faci-Llity, based on experience obtained in spent-fuel storage during the

following 7, years, ar pears in a recent report (July 1979)- prepared for ' the -

NRC by Pacific Northwest Laboratory.6 he following passage is taken from the abstract of that report: The purpose of this report is to provide a description of spent fuel handling activltles and systems, and an analysis of the storage perf ormenos as developed over the seven year operational history of the Morris Operation. Des ign considerations-and perf ormanos are analyzed for both . tha _ basin and key supporting systems. - The bases for this analysis. are the provlsions for containing radioactive by-product materials, for shielding f rom the radiation they emit, and for preventing the formation cf a critical array. These provisions have been met ef fectively over the history of storage at Morris. The release of redloactive materials is minimized by the protection of the cladding Integrity, the contain-ment of the basin water, the removal of radioective and other ; contaminants f rom the water, end by filtering and then displersing the basin air. 4. FACILITY DESIGN AND DESCRIPTION All radioactive material handling related to fuel storage at the Morris Operation is in facilities located within a protected area. No radioactive liquid effluents are released to the environs, nd no burial of radioactive or contaminated material occurs on the tract. The only radioactive or contaminated waste materials leaving the site are effluents 7 vented through the ventilation stack or solid low-level radioactive wastes that are shipped 'offsite. Offsite shipments are made in accordance with applicable U.S. Nuclear Regulatory Commission, U.S. Department of Trans-portation, and other State and Federal regulations. The principal structure at the Morris Operation site is the main building. his safety analysis is concerned only with the use of this o i-I 4 8 t, structure : for fuel receipt, storage and shipment..The fuel storage operations utilize the following portions of the main' building: 1. cask receiving and decontamination areas, 2. . cask unloading basin 3. fuel storage basins, 4. basin support systems (water cooling, filtration, etc.), and 5. control room. Irradiated nuclear fuel is received at the Nbrris Operation in - shielded shipping casks which are designed, loaded, and transported in accordance with regulatory requirements of the U.S. Department of Transportation. Prior to shipping, fuel is inspected for defects; known defective fuel-is not normally accepted by Morris Operation. Prior.to unloading the fuel, the casks are decontaminated and flushed to detect any damaged-fuel and then lowered into the cask unloading basin. Fuel is unloaded under a minimum of 2.74 m (9 ft) of water and placed in stain- . less steel basket assemblies designed to protect the fuel' from physical damage (four PWR elements or nine BWR elements per basket assembly), maintain the fuel in a subcritical configuration, aad permit the trans- - port of the fuel to the storage basins. A doorway guard at the entrance to the storage basin prevents the basket from tipping and discharging the fuel. Baske; aounting provisions in the fuel basins provide seismologi-cal restraints. The basins were constructed below ground with stainless-steel-lined, 1 reinforced-concrete walls about.61 m (2 ft) thick poured in contact with the sides of a bedrock excavation. The south wall of the basin is about 1.22 m (4 f t) thick, because it was intended to stand independent of the surrounding rock to facilitate possible future expansion. A leak detec-tion system and pump-out facilities are provided for interconnected channels spaced at regular intervals between the concrete walls and floor and the stainless steel liner. A ventilation system is provided that is designed so that air passes sequentially from areas of low contamination potential to areas of higher contamination potential and thence through a sand filter and the 91.44-m (300-ft) stack. Special vent hoods are available for fuel bundles con-taining defective fuel rods to collect escaping gases, which are filtered and then vented via the stack. l l l 9 Basin water is circulated through a system that reduces radioactive contamination by ion exchange and filtration.. The system includes a ver-tical pressure shell containing a precoated leaf filter. The filter screens are precoated with a cellulose fiber filter aid and then over-coated with either Powdex anion-cation resin or Zeolon, a synthetic alumino-silicate molecular sieve, for removal of both particulates and soluble ions. A suction system is provided to clean the basin floors and remove floating debris. Radioactive materials are collected and stored in 'the low-activity-waste (LAW) vault. The underground LAW vault [14.0 m (46 ft) in diameter and 23.5 m (77.2 ft) deep] is constructed of steel-lined, reinforced concrete about 0.61 m (2 ft) thick, poured directly against excavated rock. A vault cover of reinforced concrete is provided. An inner steel tank [11.7 m (38.5 ft) 6 L in diameter and 21.0 m (69 f t) deep], having a capacity of 2.27 x 10 (600,000 gal), is the waste receptacle within the vault. A second, under-ground waste-storage vault -- a stainless-steel-lined, reinforced-concrete cylinder, identified in the CSAR as a cladding vault -- which may also be used for low-level-waste storage, is also located onsite. Liquid from this vault can be transferred to the LAW vault. The naterial in the LAW vault can be pumped to an evaporator located in the canyon area of the repro-cessing plant. The overhead vapor from the evaporator is routed through the sand filter and discharged up the steck, while the concentrate is returned to the LAW vault where solida precipitate as the solution tem-perature drops. The safety evaluation report of ref. 2 approved the present configura-tion of the Morris Operation provided certain structural features not then in existence be incorporated in the design. These features included new storage baskets and racks, structural modifications associated with cask handling and unloading, and changes in the basin water cooling and cleanup systems. For the most part these features were proposed by and incor-porated in the present design by the applicant and approved by the NRC. Our review confirms that these and the other features of the facility as described in the CSAR fulfill the safety requirements of 10 CFR 72; hence, we are able to conclude that spent-fuel handling and storage at the Morris Operation do not present significant risks to public health and safety. 4.1 Utility and Support Systems Utflity and support systems provided for the MFRP are utilized for the operation of the spent-fuel storage facility. A deep well is the normal supply for the potable, utility, demineralized, ar.d fire protection 10 water systems. A second well, furnishing 113.6 L/ min (30 gal / min) to the adjoining Boiling-Water Reactor Training Center, is interconnect.4 with the Morris Operation system and is used as a backup supply. Natural gas La supplied for-the steam system by the local ' utility. Electrical power is furnished by Commonwealth ' Edison Company through two separnte 34,000-V transelssion lines connected to two onsite transformer systems.- While loss cf _ electrical power would not result in unsafe conditions, a diesel-driven, standby generator is available to supply power for essential services in the event that both of the incoming power sources from the utility are lost.. A sanitary sewer system is routed to lagoons for treatment prior to chlorina-tion and release to the river. An industrial sewer system, meeting state requirements for release, also discharges to the Kankakee River. Chemical wastes having concentrations above discharge limits are sent to an onsito, earth-diked evaporation pond having no discharge. 4.2 Ventilation System The ventilation system for the Morris Operation has been maintained as installed for the fuel recovery operation. Fresh air supplied to the cack decontamination area and the basin area flows through the " canyon" and the process cells into an exhaust duct, through a sand filter, and is discharged through a 91.4-m (300-f t) stack to the atmosphere. Canopy hoods for placement over possible leaking fuel elements in storage also vent to the canyon. The ventilation system is considered acceptable for the spent fuel storage operation. 4.3 Waste Handling Liquid and slurry wastes generated by the Morris Operation consist largely of cask coolant, decontamination solutions, and ion exchange resins and filter media from the pool water cleanup system. These are transferred to the LAW vault, where the supernatant liquid is ' routinely concentrated for volute reduction by an evaporator and the vapor discharged through the sand filter and up the monitored 91.4-m (300-f t) stack. Periodic flushing of the evaporator boiler with water and/or acid and the return of the solutions and suspended solids to the LAW vault during normal operations prevent the accumulation of significant levels of radioactivity in the evaporator.7 A decision on the ultimate disposal of j the accran'ated solid material in the LAW vault has not been made.7 The low-level solid wastes generated in cask decontamination, laboratory Operations, and other work at the site are packaged in drums and shipped to a commercial waste burial site. Management of these wastes is consistent with plans previously reviewed and appre7ed for the spent fuel recovery operation.8 l l l 11 5. RADIATION PROTECTION During the site visit to Morris Operation, special attention was given to the measures being taken for the confinement of radioactivity and minimit itioa of personnel exposure. Plant managemant provided charto demonstrating the continued ability of the basin water decontaminatica system to quickly remove any unusual release of radioactivity to the storage pool and routinely maintain contamination levels below the occu-pational' maximum permissible concentrations (MPCs) of 10 CFR 20 (ref. 9). We cbserved that radiation alarm and monitoring systems were adequate and well placed and access to areas of potential contamination were controlled. As a result of the site visit and evaluation cf the measures for radiologi-cal protection and effluent control given in the CSAR, it is our opinion that routine operation of the facility does not present a sigaificant radiological health or safety risk and fulfills all the requirements for radiological protection of 10 CFR 72.

6. ACCI'1ENT SAFETY ANALYSIS The applicant has calculated the possible consequences from the following postulated incidents:

1. Cask Drop in Unloading Basin 2. Basin Leakage 3. Loss of Basin Cooling 4. Cooling System Leak 5. Low-Activity-Waste Vault Leckage 6. Missile Impact on Basin Structure 7. Fuel Bundle Drop 8. Fuel Basket Drop 9. Tornado-Generated Missile 10. Criticality 4 1; m - r '12-j' ') ' The cask drop and - basin leakage were previ)usly evaluated -by the NRC. ' staff.2 We-reviewed the CSAR evalcation of the. cask drop, basin

. leakage, loss'of
basin' cooling, cooling systems, and LAW 1eakage and agree

- that. there would be no danger to the health and safety of plant personnel -or: general' public should any of these. incidents _ occur.. We agree with the. conclusion l in the CSAR: that the penetration of the. basin cliner by a tornado- -generated missile is unlikely;-ia any. event, the ability of the Horris Operation :to: detect' and ' expeditiously repair such leaks has been demon-

strated (cf. CSAR, Sect. 8.3,' p. 5).
The radiological consequences of postulated accidents that could. result in> of fsite. radioactive releases were-. independently evaluated as given - below.

The stored fuel is presumed to have a burnup 'of 44,000 mwd per metric ton of ' h a vy metal b.40 kW/kg U and cooled l' year. The ORIGEN-code 4 was used to estimate the radionuclide activity in the spent fuel (Table 1). Our calcu-latiens assumed ground-level releases, a wind speed ' of 1 m/s and Pasquill diffusion category F. Exposures (50 year dose commitments) were estimated using the /*RDOS code 10 at the 150-m (500-f t) site boundary '(i.e., the country road; cf. Fig. 3-3 and Sect. 3.2.2.4,- pp. 3.4 ~and 3.7, respectively, ~of 'the 'CSAR); and 'at 800 m (2600 ft), the nearest nermanent residence. The estimated X/Q's at 150 and 800 m (500 and 2600 ft} were 5.6 x 10~3 ~ and 2.4 x 10-4, respectively, for noble gases and 3.4 x 10-and 5.8 x 10-5, respec- ~ - tively, for iodine, where a deposition velocity of 0.01 m/s was assumed. 1. ) 6.1 Fuel Bundle Drop f It is assumed that all the rods in a dropped PWR fuel bundle were rup-tured, releasing all fission gases in the plenum (30% of the 85Kr and 10% of . the iodine in the fuel bundle).ll Ouries releatad, assuming that none of the 85Kr and 99% of the iodine dissolve in the basin water,ll would ' be about 1510 curies of 85Kr and 1.97 x 10-5 curies of 129 Estimated maximum expo-1 sures at the site boundary,150 m (500 ft), are 4.0 mrem whole body and 0.11 l . arem thyroid. These exposures are fractions of 8.0 x 10-4 and 2.2 x 10-5, respectively, of the 5-rem limit for the dose commitments to offsite indivi-e duals from -design-basis accidents given in 10 CFR 72.- Estimated indivi-dual whole body and thyroid doses at the nearest pe rmanent residence, 800 m 2 l (0.50 mi), would be 0.17 mrem and much less than 0.01 mrem, respectively. 6.2 Fuel Basket Drop The maximum drop of a fuel storage basket in the unloading pit of the storage pool is 6.86 m (22.5 f t) in water. ' It is unlikely that in the I event a basket containing spent fuel (four PWR or nine BWR fuel bundles) is dropped that the fuel liner will be penetrated or the fuel rods in the fuel bundle ruptured. However, for the ' purpose of establishing an upper limit to the radiological consequences,f such an accident, it is assumed that all of the fuel rods in the contained spent fuel (four PWR fuel bundles) i-r m m x-O g 4 m 7 y. w. ee-/,,, .gg., ,ec m_ _..,oe .l, '.;'f.'-. / ~ )* _ __ [ p p... _ t)[ _ _ ~ ~_ _ ; f, ~ 13 rupture and the plenum gases are released to the basin water. The curies released would be four times those assumed for the single-fuel bundle drop discussed above (i.e., 85Kr, 6040 curies and 129, 7.88 x 10-5 curier). I The estimated maximum, individual whole-body and thyroid exposures would be 16.0 and 0.44 mrem, respectively, at the 150-m site (500-ft) boundary and 0.68 and about 0.01 mrem, respectively, at 800 m (2600 ft). 6.3 Tornado-Generated Missile It is postulated that a tornado generated missile has the potential of impacting as many as six BWR or four PWR fuel assemblies and that the rods in these assemblies would fail. Assuming that a basket containing four PWR fuel assemblies was struck, the releases and exposures would be the same as those calculated for the fuel-basket drop discussed above. 6.4 Criticality Criticality accidents which have occurred in the nuclear industry have been associated with chenical reprocessing or assemblies involving plutonium or highly enriched uranium. No criticality accidents have occurred in systems containing low-enriched uranium. We agree with the statements in the CSAR that a criticality event in a spent-fuel storage pool is highly improbable and precluded by many factors including fuel basket design, geometric restraints imposed on fuel storage, operating procedures and low fissile content of the fuel assemblies. However, as a conservative measure, the radiological consequences associated with cri-ticality events at various fission yields were evaluated. It was assumed that all fission products, including fission gases, were contained within the UO2 fuel matrix and not released as a result of the postulated criticalities. Criticality accidents which have oocurred in heterogeneous, water-moderated systems have resulted in total fission yields ranging from 3 x 1016 to 1.2 x 1020 (See ref. 12.) Radiation doses at the surface of the storage pool from promgt gammas produced by nuclear excursions involving 1018, 1019, and 102 fissluns/s were calculated.13 The nuclear excursions were assum<d to occur 4.88 m (16 f t) under the surface of the poo), and calculations were made assuming (1) a point source and (2) a 30-cm (ll.8-in.) diam sphere. Source energies and photons (Table 2) were taken from a distribution curve in ref. 14, and 7.5 prompt gammas per fission were assumed. Surface radiation doses from the postulated criti-calities are given in Table 3. In the event of a criricality excursion, radiation levels at the surf ace of the pool would be low enough to allow ~

g y-

.14 K a Table 2. -Prompt gamma 1 fission spectra Source strength ..(Nev) % of total ~ photon - .3 19.48 1.0 28.14. 1.5 20.30 2.0. 10.82 2.5 ,8.12 .3.0 6.09 4.0 4.33 5.0: 1.62 6.0 0.65 7.0 0.48 aE. K. Hyde, " Fission Phenomena," Vol. 3, The Nuclear Properties of the Heavy hetals, Prentice-Hall, New Jersey,1964. . ) Table 3. Surface radiation doses.from subsurface [4.88 m -(16 ft)] nuclear excursions in spent-fuel storage pool 5 Surface dose Fissions /s. Source geometry (area) 1 x 1018 sphere 0.10 a 1 x '1018 point 1.05 1 x 1019 -spherea 0.97 1 x 1019 point 10.5 s 1 x 1020 spherea 9.66 1 x 1020 point 10.5 i 830-cm (11.8-in.) diam. d 4 ,ww ,m'-,E,,- ,-w ---%.r-, -r-, ,nv. oe,, F 15-for' prospt remedial action. The effects of a criticality event in a spent-fuel storage pool would be similar to those resulting from the short-term operation of. a low power, swimming pool-type reactor commonly used in research. -7. CONDUCT OF OPERATIONS 7.1 Organization and Staff Qualifications ' General Electric Company, the sole awaer and operator of the Mperis Operation, has demonstrated comp (tence in the nuclear industry. Personnel in key positions et the Morris Facility have obtained experience at handling radioactive materf als while employed by General ' Electric or other companies at other nuclear facilities. General Electric has established minimum qualifications for management, supervisory, and technical positions' necessary for the safe and efficient operation of the Morris Operation satisfying the requirements of 10 CFR Part 72.17, " Contents of Application: Technical Qualifications." A plant safety com - mittee composed of the managers of the various plant organizations has jurisdiction over matters having radiological or nuclear safety implica-tions. The Licensing and Radiological Safety Senior Engineer,' the secre-tary of this committee, reports directly to the manager of Morris Oceration and is responsible for coordinating site regulatory matters. Besides participating in general orientation and safety courses, operator tech-nicians are required to participate in onsite training programs prepared by management and engineering personnel qualified in the assigned topical or functional area. Records of activities relating to plant safety are accu-mulated to assist in the application of safety principles and objectives to plant operation. Periodic internal audits are conducted by Morris Operation management in safeguards, criticality and radiation safety. These audits are subject to review by teams from the corporate organization external to the Morris Operation. 7.2 Emergency Plans The applicant has described in the CSAR plans for coping with the following classes of emergencies: (1) criticality incidents, (2) con- -tamination accidents, (3) fire, (4) major equipment failures or opera-tional accidents, and (5) other conditions such as effects from natural phenomena. These plans have been compared with and fulfill the require-ments of Appendix E to 10 CFR Part 50, " Emergency Plans for Production and Utilization Facilities." Details of emergency agreements and assistance ar'rangements with law enforcement, medical, and other local agencies, and services are given in a supporting document, Appendix 1, NED0-21894, o " Radiological Emergency Plan for Morris Operation," which was not available for review by ORNL. {@ g -4 ' 7.3 Decommissioning The decommissioning plar, for the Morris Operation is contained in Appendix A.7 of the CSAR. We have' compared the plan with the technical 1 requirements of 10 CFR Part 72.18, " Decommissioning Plan, Including its Financing'".and find it in compliance, but the financial arrangements for the decommissioning of the Morris Operation are not included in Appendix A.7. 8. REFERENCES

1. -General Electric Company, Consolidated S2facy Analysis Report for

' Morrie Operation, Morrie, Illinois, NEDO-21326C (January 1979). 2. Safety Evaluation Report by the. Division of hel Cycle and M2terial Related to License Amendmont for M1terial Licence No. SNM-1266 Morris Operation Facility, Grundy County, Illinois for Genent Electric Company Docket No. 70-1308, U.S. Nuclear Regulatory Commiasion, Washington, D.C. (Dec. 3, 1975). 3. Code of Federal Regulctione, Title 10, Part 72, " Licensing Require-ments for the Storage of Spent Fuel in an Independent Spent Fuel S*orage Installation," Office of Standards Development, U.S. Nuclear Reguletory Commission, Washington, D.C. (October 1980). 4. A. C. Croff et al., Revised Uranium Plutonium Cycle WR and WR Modele for the ORICEN Computer Code, ORNL/TM-6051 (September 1978). 5. Salety Evaluation of the Midwest het Recovery Plant, General Elec ty'it' Company, Docket No. 50-268, Directorate of Licensing Fuels and Materials, U.S. Atomic Energy Commission, Washington, D.C., Appendix D and E (December 1972). 6. K. J. Eger and C E. Zima, Commentary on Spent hel Storage at Mor**fe Operation, prepared for the U.S. Nuclear Regulatory Commission by Pacific Northwest Laboratory, NUREC/CR-0956, PNL-3065 (July 1979).* 7. Ibid., pp. 64-67. 8. Safety Evaluation of the Midueet het Reco"ery Plant, General Electric Company, Docket No. 50-268,, Directorate of Licensing Fuels and Materials, U.S. Atomic Energy Commission, Washington, D.C., (December 1972), pp. 34-45. J i l l' I j 17 9.. K. J. Eger and G. E. Zima, Commentary on Spent het Storage at' Norris Operation, prepared for the U.S. Nuclear Regulatory Commission .by Pacific Northwest Laboratory, NUREG/CR-0956, PNL-3065 (July 1979), pp. 17-31.* .10. R. E. Moore, The AIRDOS-II Computer Code for Estimating Radiation Dose to Min from Airbornc Radionuclides in Areas Surrounding Nuclear Pteilities, ORNL-5245:(April 1977). 11. Regulatory Guide 1.25, Ass'unptions Used for Evaluating the Potential Radiological Consequences of a het Handling Accident in the hel Handling and Storage Facility for Boiling and Pressurized Water Rnactors, Directorate of Regulatory Standards, U.S. Atomic ' Energy Commission (Mar. 23, 1972). 12. W. R. Stratton, A Review of Criticality Incidents, LA-3611 (January 1967). 13. H. F. Soard, ORNL,. personal communication, July 6,1979. (Calculations made using code in ref. 12). 14. E. K. Hyde, " Fission Phenomena," vol. 3, The Nuclear Properties of the Heavy Elements, Prentice Hall, New Jersey (1964). 15. E. D. Arnold and B. F. Maskewitz, SDC, A Shielding - Design Calculation Code for hel-Handling Facilit.'es, ORNL-3041 (March 1966). 4

  • Available for purch se from the NRC/GPO Sales Program, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and/or the National Technical Information Service, Springfield, VA 22161.

19 APPENDIX: STRUCTURAL ANALYSIS OF MO2RIS 0?ERATION CRANES K. H. Ink

  • C. R. Hammond*

UCC-ND Engineering March 20, 1980 J f i s 20 'A.1 Introduction This study is an independent evaluation of the effects of a safe shutdown earthquake (SSE) on the cask handling crane,- its support struc-ture, and the storage basin crane at-the General Electric Morris e Operation.1 The analysis was performed at the request of the U.S. Nuclear Regulatory Commission. A finite element model was developed for the cask handling crane and its support structure. The dynamic stresses in the support structure in a safe shutdov- -sathquake were calculated ly the response spectrum method. The of the seismically induced lateral forces acting on the crane rat a its accessories were also evaluated. The lateral forces acting on the storage basin-crane rail were-treated as statically applied loads with magnitudes directly proportional-to the seismic ground accelerations. The effects of these forces on the crane rail and tie-down clips were evaluated. The results show the following: 1. The maximum combined stress that would occur in acy component-of the support structure of the cask handling crane during a SSE is 14,000 psi. The combined stress is the sum of the stresses due to the weight of the crane and structure and those induced by the seismic motions. The maximum combined stress found by the analysis is less than the minimum yield strength of structural steel and would not be expected to cause a failure in the support structure. 2. The maximum seismically induced lateral force acting on the cask handling crane from a SSE would be 14.2 std tons (28,400 lb). When this lateral force is transferred to one tie-down clip, it is likely that the clip would yield and become a plastic hinge. However, the adjacent clips would absorb the lateral loadings on tra crane rail. 3. 3 The loading on the tie-down clips of the storage basin crane from a SSE would be minimal. It is unlikely that these ' tie-down clips would fail as the result of a SSE. A.2 Method of Analysis The cask handling crane (CHC) is located on an elevated crane runway in the cask receiving area. The elevation is about 10.1 m (33 ft) above the floor. of the building. Loadings imposed on the support structure i e 21 during a safe shutdown earthquake were determined using the finite ele-ment method. The results were used to evaluate the stresses in the sup-port structure end. the interacting forces between the crane wheels and the supporting rail. The storage basin crane (SBC) is at ground level. The dynamic ampli-fication ef fects on the crane during a SSE were judged to be minimal. Forces acting on' the basic crane rail during a SSE were evaluated by a static analysis. A.2.1 Safe shutdown earthquake The safe shutdown earthquake used in the studies is the maximum earthquake in the design criteria of the Morris Operation facility (ref. 1, p. 4-20). The ' maximum horizontal ground acceleration is 0.2 x g, with the vertical ground accelycation two-thirds that of the horizontal. The horizontal and vertical design response spectra used in the studies are from the NRC Regulatory Guide on the design response spectra for seismic design of nuclear power plants.2 A damping value of 5% of criti-cal,dasping was used. A.2.2 SAP computing code structural analysis.g code is a general purpose finite element program for The SAP computin The program tras originally developed at the University of California, Berkeley. Currently, it is maintained by the Civil Engineering Department at the University of Southern Califernfa. The program is available. t the ORNL Computing Center (SAP V - USC a Version TVo) and was used in this study. A.2.3 Cask handling crane The cask receiving area enclosure is a steel frame building attached to a concrete foundation. The crane rail is elevated and attached to girders supported by inside columns. The "A" frame suppo. cs are provided to one side of the c~ane runway. Purlins are attached to the our. side columns and connected by sag rods. The building is covered with industrial insulation panels. Building dimensions and identities of structural members were obtained from General Electric's drawings num-bered MFR-E-2228A,4 MFR-E-2228B,5 MFR-E-2228D,6 MFR-E-2228E,7 and Fluor Corporation's drawings numbered 5-2104A6 and 5-2103A.9 22 Information on the crane rail and accessories were taken from Fluor Corporation's drawing No. 5-2105A10 and Bethlehem Steel's Booklet 335L.ll Cross-sectional properties of structural members were evaluated in accordance with AISC Manual of Steel Constrauction.12 The cask handling crane is supplied by the Whiting Corporation. It has a rated capacity of 125 std tons (250,000 lb). Informa ti( T on the CHC was provided by Whiting's drawing No. U-59991.13 A finite element model wac developed for a middle section of the cask receiving area enclosure. The CHC was assumed to be located in the center of the section. A sketch of the model is shown in Fig. A.1. The crane rail is parallel to the Y exis. The Z axis is vertical, and the direction transverse to the crane runway is the X axis. The two "A" frames defined the section modeled. All st;uctural members, with the exception of the crane, were modeled by beam elements. The crane was treated as a rigid body. The crane wheels were simulated by short beams. The load (shipping cask) carried by the crane was modeled by a concentrated mass attached to a sof t beam element suspended trom the center of the crane. The sof t beam ele-ment has the same stiffness as a suspend 1d pendulum of the same length. The soft beam element is 109 cm (43 in.) long, and it is also the minimum 1(ngth of the load carrying crane cable. The stiffness of the purlins, sag rods, and insulation panels was not included in the model. However, their masses were distributed ta the appropriate nodes as concentrated The finite element model is one section of the cask receiving masses. area enclosure and the effects of the remaining parts of the building were accounted for by imposing appropriate constraints on the model. The horizontal and vertical ground accelerations were input along the X and Z axes, respective 1v. Motion in the Y direction (along the crane runway) would be small, and, consequently, the translation along the Y axis was eliminated at all nodes. Fixed-end supports were assumed at the Fround level. In order to simulate the support provided by the rest of the enclosure, the rotational degrees of freedom about the X and Z axes were also eliminated at nodes on the sections defined by the two "A" frames. A.2.4 Storage basin crane The storage basin crane is located at ground level. It has a rated capacity of 7.5 std tons (15,000 lb). Because of its low elevation, the ~ dynamic amplification ef fect was assumed to be negligible and the total transverse forces acting on the ciane rail were taken to be 20% of the total weight. These transverse forces were divideo equally among the four wheels. The ef fects of the SSE on the crane rail and its accessories were evaluated by a static analysis. k 23 ORNL-0WG.81-607 f 'M f,^Q ' d / x N Fig. A.i. Finite element model - cask handling crane and support s.ructure. e k 24 A.3 Analytical Results A.3.1 Cank handling crane The dynamic loadings on the support structure of the CHC were eva-luated using the response spectrum analysis option in SAP with the NRC design response spectra.2 The lowest 12 natural frequencies were used in the calculations and are shown in Table A.1. Support structure. Important structural members of the CHC support atructure are identified in Fig. A.2. The structural members and the stress distribution are symmetrical to a midplane between the two "A" frame sections. The total stress in each structural member consists of two components; a static and a dynamic stress. Each component was com-puted separately. The static stresses in the support structure were calculated using the static analysis option in SAP. The dynamic stresses were evaluated using the response spectrum analysis option in SAB and in accordance to the NRC response combination method.14 The total stresses in tht fmportant structural members are summarized in Table A.2. All the calculated stresses are less than the nominal yield stress for structural steel, thus the safe shutdown earthquake is not expected to produce permanent deformation or collapse of the CHC support structure. Crane rail and accesr.ories. The crane rail is held in place by double clips and reversible fillers spaced at a 0.61-m (2-ft) interval. A sec-tion of the crane rail, its accessories, and key dimensions are shown in Fig. A.3. The section is that of a standard 135-lb rail.12 The transverse forces acting on the cranc rail are the computed lateral inertia (shear) forces acting on the runway. The calculated maximum shear was 14.2 std tons (28,400 lb). It was assumed that this transverse shear acted on the top of t he rail. Its ef fects on the rail accessories are discussed below. o Sliding of Rail Section. The free-body diagram fot' the forces under consideration is shown in Fig. A.4. Under the action of the transverse shear V, the rail could slide and impact in the filler plate if the fric-tion force F is not suf ficient to overcoca the transverse shear V. The friction force P is proportional to the net vertical lcad P. The net vertical load P is the static wheel loading (one-fourth of the cotal weight) less the maximum vertical shear produced by the seismic motion. For the CHC, the static wheel loading is 29.6 std tons (59,200 lb). The calculated maximum shear is 8.2 std tons (16,400 lb) and the corresponding net vertical load P i 21.4 std tons (42,800 lb). The static friction coefficient between dry steel surfaces is approximately 0.7 (ref.15) and the maximum friction force that can be developed between the rail and the 25L a Table A.1. Cranc support structure natursl frequencies Frequency Mode No. (Hz) 1-0.62 2 3.61 3 8.92 4 11.07 -5 11.50 6 11.84 / 12.42 8 12.42 9 13.32 10 15.31 11 15.32 12 17.44 aNRC Regulatory Gu1de 1.60, The Design Response Spectra for Seismic Design of Nuctear Pouer PLante (December 1973). 26 ORNL-DWG.81-608 Detail of Front "A" Frame Section not sMwn. 3 h / z7 7 \\_ 10 2 20 ( ~,/w ~ e z 22 I g / 6 Fig. A.2. Numbering system for important structural members - crane support structure. r Table A.2. Stress distribution of crane support structure G Component Reference b c No. Designation drawing No. (s )1 (8 )2 c c 1 6W15.5 GE NFR-E-2228D (Rev. 1) 603 -603 2 6W15.5 GE MPR-E-2228D (Rev.1) 882 -882 3 10W33 GE MFR-E-2228A (Rev. 2) 563 -563 4 8W24 Fluor 5-2104 A (Rev. 4) 446 .-446 5 rail and girder Fluor 5-2105A (Rev. 2) 9,775 -9,774. 6 6W15.5 Fluor 5-2104A (Rev. 4) 447 -447 7 6W15.5 GE NFR-E-22288 (Rev. 2) 664 -664 8 6W15.5 GE MFR-E-2228A (Rev. 2) 823 -823 9 8W24 Fluor 5-2104A (Rev. 4) 342 -342 10 rail and girder Fluor 5-2105A (Rev. 2) 14,139 -14,139 11 14W61 Fluor 5-2103A (Rev. 1) 7,960 -7,799 w 12 21W62 GE MFR-E-2228B (Rev. 2) 12,827 -12,500 13 10W33 GE MFR-E-2228A (Rev. 2) 5,461 -5,008 14 14W61 Fluor 5-2103A (Rev. 1) 9,000 -9,919 15 10W49 GE MFR-E-2228E (Rev. 1) 9,587 -9,040 16 2L 4 x 3 x 5/16 GE MFR-E-2228E (Rev. 1) 6,059 -5,931 17 2L 3 x 2 x 1/2 GE MFR-E-2228E (Rev. 1) 7,745 -7,833 18 2L 3 x 2 1/2 x 5/16 GE MFR-E-2228E (Rev.1) 7,301 -7,365 19 2L 3 x 2 x 1/2 GE MFR-E-2228E (Rev. 1) 10,026 -9,880 20 2L 5 x 3 x 7/16 GE MPR-E-2228E (Rev. 1) 8,36_0 -8,231 21 2L 5 x 3 x 7/16 GE MFR-E-2228E (Rev. 1) 8,559 -9,072 22 10W49 GE MFR-E-2228E (Rev. 1) 7,415 -7,155 aSee Fig. A.2. b(,c)1 - maximum total tensile stress in psi.

  1. (s )2 - maximum total compressive stress in psi.

e 28 ORNL-DWG,81-609 a; % 'k h m-I. t i i n. ++ous in.. ous ton i-sotts ,o. i.etts ,,3 , y,. U V SINGLE CtlPs ANO FEttes 3, Ang HAtf uMGTH g 3 i o O l I h is- ,s-l arveesiste entra i oovett cte l l l ---ian.- in. -- - in. -. e f MND N iv.- J i. v.- q,_ l l l l h* i i i ri-i. i. l Fig. A.3. Crane rail and accessories. 29 ORNL-DWG.81-610 ASCE 40-lb. Rail Section 135-lb. Rail Section: 3 H = 5.15 (in.) H = 3.5 (in.) L = 5.1875 (in.) L = 3.5 (in.) l. [ H ll <C i L-d m Fig. A.4. Free-body diagram for rail sliding case. 30 girder is 15.0 std tons (30,000 lb). Therefore, the friction force can balance the maximum transverse shear of 14.2 std tons (28,400 lb) and slipping of the rail is not anticipated. e Forces on the Clip. The transverse force V might cause an Lapending rotation about one edge of the rail section. The section is assumed to pivot about the Point A, and the free-body diagram of the rail section under loading is shown in Pig. A.5. The reaction force of the slip is represented by the downward force C. The value of R is the ground reac-tion force point A. The net vertical load, determined in the previous subsection is 21.5 std tons (42,900 lb). Using the dimension' sf a standard raill2 and summing moments about the point A, the clip reaction force was found to be.49 std tons (9870 lb). To evaluate the loading on the bolts, it was conservatively assumed that the force C acted on one double clip (Fig. A.3). The axial stress produced in the two 1-in. bolts is approximately 6300 psi. This esti-ma ted bolt tensile stress is much smaller than the normal yield stress for a 1-in. carbon steel bolt. To evaluate the stresses in the clip caused by the rcaction force C, the clip was modeled as a straight cantilever beam with an end force E. A sketch of the beam model is shown in Fig. A.6. The length of 1.3 in. is measured from the center of the bolt to the clip bend. The clip is 0.5 in thick, has a width of 6 in. and is made of carbon steel with an ASTM designation of A663, Grade 60. The yield strength of that grade of carbon steel is 30,000 psi. n The crane has a wheel base of 2.84 m (9 ft 4 in.) and the rail clips are on 0.61-m (2-ft) intervals. On the average, the transverse shear acting on the rail would be transferred to at least two clips. In the two clip case, the end force E would be one-half of the reaction force C as calculated previously. The corresponding maximum clip bending and shear stresses were 23,000 psi and 2470 psi, respectively. These stresses are less than the nominal yield stress for the carbon steel and failure of the clip is not likely. If a wheel is resting directly over a clip during an earthquake, most, if not all, of the wheel transverse shear would be transferred to one clip and the end force E would equal the clip reaction force C. For this case, the maximum clip bending stress would exceed yield and the clip could become a plastic hinge. As soon as one clip began to yield, the transverse load would be transferred to the two adjacent clips. Results from the previous discussion indicated that two clips can absorb the maximum transverse shear without yielding. Therefore, it is con-ceivable that the transverse shear produced by seismic motions could cause yielding in one clip but the adjacent clips would provide more than adequate compensation for such a failure. 31 ORNL-DWG.81-611 135-lb. Rail Section ASCE 40-lb. Rail Sect. ion: H = 5.75 (in.) H = 3.5 (in.) L = 5.1875 (in.) 3 L = 3.5 (in.) V-t> x j C s V A v b k L R Fig. A.S. Free-body diagram for clip reaction force. 1 1 32 ORNL-DWG,81-612 l= 3" d o 5" e \\> J 6 1 i Fig. A.6. Structural model for clip bending analysis, e 33 A.3.2 Storage basin crane The total weight (dead weight and load) of the storage basin crane is 21.5 std tons (43,000 lb). The crane rail is an ASCE 40-lb rail section.12 The runway is held in place by pressed, single clips and rever-sible filler at 0.61-m (2-ft) intervals. Sliding of rail sections. Due to its ground level elevation, dynamic amplification was assumed to be negligible for the SBC. The free-body diagram under consideration is shown in Fig. A.4. The transverse shear V was taken to be 20% of the total weight (i.e., 0.2 x g lateral accelera-tion), equally divided, or 2150 lb per wheel. The corresponding vertical shear is 1075 lb. The net vertical force at each wheel, P, namely, one-fourth of total weight minus the vertical shear, is 9680 lb. Using a static coef ficient of 0.7 (ref.15), the maximum friction force that can be developed between the rail and the steel support plate is 6770 lb. The friction force F is significantly larger than the transverse force V and slipping of the crane rail on the support plate is not expected. Forces on the clip. The free body diagram of the forces under con-sideration is shown in Fig. A.S. The transverse shear V is 2150 lb. The net vertical force P is 9680 lb, R is the ground reaction force at point A, and C is the clip reaction force acting on the rail. Using the dimen-sions of an ASCE 40-lb raill2 and summing moments about the point A, it was found that the moment produced by the force P was more than suf-ficient to counter balance that produced by the trnsverse force V. The rail section cannot rotate and, consequently, there is no reaction force f rom the clip. The loadings on the clips would be negligible and the transverse shear is not expected to cause a failure in the clips of the SBC. 34 REFERENCES 1. Consolidated Safety Analysie Report for Morris Operation, vols. I and II, Report No. NEDO-21326b and NED0-21326c, General Electric Company (January 1979). 2. NRC Regulatory Guide 1.60, The Design Response Spectra for Seismic Deafgn of Nuclear Pocer Plante (December 1973). 3. E. L. Wilson, SAP - A Ceneral Structural Analyefe Progmm, SESM Report 70-20, Department of Civil Engineering, University of California, Berkeley (1970). 4. General Electric Cotapany drawing No. MFR-E-2228A, Rev. 1, October 1976. 5. General Electric Company drawing No. MFR-E-2228B, Rev. 2, October 1976. 6. General Electric Company drawing No. MFR-E-2228D, Rev. 1, October 1976. 7. General Electiic Company drawing No. MFR-E-2228E, Rev. 1, October 1976. 8. Fluor Corporation drawing No. 5-2104A, Rev. 4, November 1975. 9. Fluor Corporation drawing No. 5-2103A, Rev. 1, September 1968. 10. Fluor Corporation drawing No. 5-2105A, Rev. 2, December 1968. 11. Crane Rails. Booklet 3351, Bethlehem Stect (December 1978). 22. Aiznual of Steel Construction, 7th ed. American Institute of Steel Cons truction, New York, 1970. 13. Whiting Corporation drawing No. U 509991, Rev. 5, May 1977. 14. NRC Regulatory Guide 1.92, Rev. 1, Response Combination Method (February 1976). 15. Baumeister, Ava11one, and Baumeister, M2rk's Standard Handbook for Mechanical Engineering, 8th ed., McGraw-Ilill, New York, 1978. i NUREG/CR-1697 ORNL/NUREG/TM-439 INTERNAL DISTRIBUTION 1. J. W. Boyle 30. L. B. Shappert 2. W. D. Burch 31. M. G. Stewart '~' 3. J. M. Chandler 32. V.C.A. Vaughen 4. E. L. Compere 33. J. P. Witherspoon 5. M. J. Peldman 34. R. G. Wymer i 6. B. C. Finney 35. L. Burris, Jr. (consultant) 7-11. E. J. Frederick 36. G. R. Choppin (consultant) 12. R. W. Glass 37. W. H. Corcoran (consultant) 13. C. H. Hammond 38. S. W. Drew (consultant) 14. G. S. Hill 39. A. M. Squires (consultant) 15. R. A. Lorenz 40. M. E. Wadsworth (cansultant) 16. A. L. Lotts 41-42. Laboratory Records = 17. K. H. Luk 43. Laboratory Records RC 18. A. P. Malinauskas 44-45. Central Research Library 19-23. J. P. McBride 46. ORNL-Y-12 Technical ti 24. H. R. Heyer Library Document t 25. H. A. Nelms Reference Section 26. J. W. Roddy 47. ORNL Patent Section 27-29. C. H. Shappert EXTERNAL DISTRIBUTION s 48. A. T. Clark, U.S. Nuclear Regulatory Commission, Office of Nuclear Materials Safety and Safeguards, Washington, D.C. 20555 49. Of fice of Assistant Manager for Energy Research and Development, DOE-ORO 50-150. Given distribution as shown in Category AN (NTIS-10) 151-153. Technical Information Center, DOE-OR ,i U S. NUCLEAR REGULATORV COMMISSION NUREG/CR-1697 BIBLIOGRAPHIC DATA SHEET ORNL/TM-439 4 TITLE AND SUBT8TLE (Add Volume Na. of mprmnate)

2. (Leave btenki Safety Review of the Design, Operation, and Radiation Sections of the General Electric Morris Operation
3. RECi:iENT S ACCESSION nom Consolidated Safety Analysis Report
7. AUTHORIS)
5. D ATE REPORT COMPLETED J.P. McBride, K.H. Luk, C.R. Hammond

"{"T" l '^1981 ua y 9 PE RF ORMING ORG ANIZATION NAME AND M AILING ADDRESS t/nclude I,a Codel DATE REPORT ISSUED EoNTH l Y E A F, ' = Oak Ridge National Laboratory July 1981 Oak Ridge, TN 37830 e- (t e,v, u.n * >

8. (Leave Nank)
12. SPONSORING ORGAN 12AT ON N AME AND MAILING ADDRESS (Inctuar I,a Cod =1
10. PROJE CT/ TASK / WORK UNIT NO.

Office of Nuclear Material Safety and Safeguards Division of Fuel Cycle and Material Safety

11. CO JR ACT NO.

U.S. Nuclear Regulatory Commission Washington, DC 20555 FIN B0102

13. TYPE OF REPORT PE RIOD COVE RE D (/nclusive dates) 15 SUPPLEMENTARY NOTES 14 / Leave o/ma!

is AusTR AcT A safety review was made of Sections 4 through 9 of the Consolidated Safety Analysis Report (CSAR) for the GE Morris Operation spent-fuel storage facility. The sections reviewed include Design Criteria and Compliance, Facility Design a d Description, Radiation Protection, Accident Analysis, and Conduct of Operations. The safety review was performed in ,~ accordance with the Code of Feder'al Regulations, Title 10, Part 72, " Licensing Requirements for the Storage of Spent Fuel in an Independent Spent Fuel Storage Installation" and contains independent estimations of source terms and dose-commitments f rom postulated accidents in the storage facility and a struc-tural ar.alysis of the Morris Operation crancs as an appendix. The review confirms that the features of the facility as described in Sections 4 through 9 of the CSAR fulfill the saf ety requirements of 10 CFR 72, and it is concluded that spent-L fuel handling and storage at the Morris Operation do not present significant risks to public health and safety. I 7 KE Y WORDS AND 00CUMLNT AN ALYSIS 17a DE SC RIPTORS -z 17ts ME N TI F IE HS OPE N E N DE O T E R YS 18 AV AIL ABILITY ST ATEMENT 19 SE CURITY CLASS ITA,s reporff 21 NO CF PAGES Unclassified 2 Unlimited Un'8Es'sT[fM'" '""' S ~ac coav m r m mu immun n e u i mu mmmmi mmmm mium immmmmm u ii m.

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