ML19270F287

Safety Evaluation of Nuclear Fuel Svc Inc Model NFS-4 Spent Fuel Shipping Cask.Concludes Cask Meets 10CFR71 Requirements,Subj to Conditions Detailed

ML19270F287
Person / Time
Site:	07002623, 07106698
Issue date:	11/14/1972
From:	NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
To:
Shared Package
ML19270F270	List: ML19270F269 ML19270F280 ML19270F283 ML19270F287 ML19270F292
References
NUDOCS 7902060122
Download: ML19270F287 (16)

Text

{{#Wiki_filter:J, ,3 S ,. s 1 9-l'N 14 79,2 7 C UNITED STATES ATOMIC ENERGY COMMISSION SAFETY EVALUATION BY THE TRXiSPORTATION 3RA'iCH DIRECTORATE OF LICENSING NUCLEAR TUEL SERVICES, INC., MODEL NFS-4 SPENT FUEL SHIPPING CASK Summarv By application dated January 3,1972, Nuclear Fuel Services, Inc. (NFS), requested approval to deliver a large quantity of byproduct material and special nuclear material in the form of. irradiated fuel elements to a carrier for transport in the Model N7S 4 shipping cask. Subsequent sub-mittals, indica ted below, transmitted additional information and revisions to the original application. A consolidated application was transmitted by NFS letter dated October 6,1972. Based on the statements and repre-sentations as contained in the consolidated application, we have con-cluded that the contents of one (1) PWR or two (2) BWR fuel elements within. the Model NFS-4 shipping cask meet the requirements of 10 CFR Part 71 subject to the conditions stated below. S_ubmit tals 1. Application dated January 3,1972, with the Safety Analysis Report (SAR) and drawing;s for the Model NFS-4 shipping cask. 2. Supplement dated Yebruary 1,' 1972, providing revised drawings a'nd brochure along with a test report concerning the Miser fire valve. 3. Supplement dated March 1, 1972, providing additional information regarding the structural analysis as Addendum No.1 to the application. 4 Supplement dated May 4,1972, providing additional information regarding the thermal analysis and the overall design. 5. Supplement dated July 10, 1972, providing information regarding thermal testing, and operational and general procedures. 6. Supplement dated September 19, 1972, providing information on NFS and Contractor Organizational Chart, revisions to the Quality Assurance Program, and proposed initial and routine acceptance tests on the cask. O 790206012A

.s .l. .') ^ NOV 14 1972 C 7. Consolidated application dated October 6, 1972, incorporating the original application snd all additional submittals. 8. Supplement dated November 9,1972, providing NFS Drawing No. E 10080, Rev. 12 (Sheets 1 through 4). Drawines E 10080, Rev. 12 NFS-4 Spent Fuel Shipping Cask 1-PWR/2-BWR (Sheets 1 through 4). Figure 2.1.3 Fuel Canis ters Package Description The NFS-4 shipping cask is primarily designed for legal weight shipment of one (1) PWR or cuo (2) BWR fuel assemblies by motor vehicle under sole use assignment. The contents are shipped wet, with the cask designed to retain the coolant and the loaded fuel canisters under both the normal and accident damage tests of 10 CFR Part 71. The calculated pressure, design pressure and temperature for the neutron shield tank and cask cavity for the normal conditions of transport are as ( follows: Item Neutron Shield Tank Cask Cavitf Calculated Pressure 85 psig 167 psig esign Pressure 100 psig 170 psig Temperature 304*F 357*F The calculated pressure, design pressure and temperature for the cask cavity following the accident damage test which considers the loss of water from the neutron shield tank are as follows: Item Neutron Shield Tank Cask Cavity Calcula:ed Pressure NA 1006 psig Design Pressure NA 1100 psig Temperature NA 534*F

m d MOV 11 1972 C Figure 1 provides a view of the cask in perspective. It is a right circu-lar cylinder with upper and lower impact limiters located at the ends of the cylindrical surf ace and a lid impact limiter. The other main components are the neutron shield tank, surge tank, lid, gamma shield and support plates. Figure 2.1.1 disclose.s the longitudinal cross section, the PWR and Bb'R configurations, and two transverse cross sections of the cask. The gross weight of the ' cask is approximately 50,000 pounds. The overall dimensions are 214 inches in length with the lid impact limiter, 202 inches without the lid impact limiter, and 50 inches in diameter. The inner cavity is 178 inches long and 13.5 inches in diameter. This cavity is penetrated by one vent, one relief, and two drain lines. All penetrations are through the lower shield disc or the cavity flange. Gamma shielding is provided by lead located between the inner and outer concentric stainless steel shells. The inner and outer shells are, respectively, fabricated of 5/16-and 1-1/4-inch thick stainless steel. Axial copper fins are imbedded into the lead to transfer the heat across the lead / steel interf ace of the inner and outer shell with a minimum temperature gradient. The annulus of lead shielding has a nominal thickness of 6-5/8 inches. The annulus is, shaped such that, for 5 inches from the bottom and for 25 inches from the top, its thickness is reduced to 5-1/2 inches. The re-duction in thickcass of the annulus at the bottom is caused by the. presence of a 5-inch long by 1-1/4-inch thick void to allow for any lead expansion during fire, postulated in the accident damage test. The smaller lead thickness at the upper end of the annulus is accomolished by reducing the diameter of.the outer shell by 2.5 inches over the 25 inches. Gamma shielding at the bottom of the cask is provided by an 8-inch thick s tainless steel plug. Gamma shielding at the top is provided by a 7.5-inch thick stainless steel lid. The cavitf flange is a stainless steel ring with a 30-inch outside diameter, a 17-inch inside diameter and 7 inches in thickness. Six l-1/4-inch diameter holes with heli-coil threaded inserts are provided for bolting the cask lid t.o the flange. The upper face of the flange is machined flat to serve as the sealing surface for the lid gaskets. Three counterbores in the exterior of the flange house the vent valve, rupture disc-pressure relief valve ecmbination and a head closure gasket leak check valve.

l 3 NOV 14 372 C The drain valves are Worchester Valve Co. 316 stainless steel 1/2-inch F466T SE Miser ball type with teflon seats and a secondary metal seal. The ball of the valve may have a bleed hole to equalize the pressure between the cask cavity and the ball passage in the closed position. Valve access is pre-rided by a 5-inch inside diameter tubing extendir.g from the outer shell of the cask to the outer perimeter of the lower i= pact limiter. The tubing is sealed by a 5-inch diameter threaded port cover which captures an elastomer 0-ring seal. The vent valve and the head closure gasket leak-check valve are the same as the drain valves and are provided with simil r access, closure and seals. The cask cavity is equipped with a 1/2-inch diameter rupture disc rated at 1100 psig at s 600* F. The rupture disc is mounted in a counterbore on the lid flange. A safety relief valve rated at 200 psig at N 350*F is installed downstream of the cavity rupture disc. The lid is 7.5 inches thick with a maximum diameter of 26.5 inches. It is a stainless steel frustum of a cone with a flange 2 inches in thickness. It has six counterbored holes for the closure bolts and four 1-inch diameter blind-threaded holes for attaching the lid impact limiter or the lid lifting device. The bottom of the lid's flange has two grooves to ( accommodate the two polytetrafluoroethylene 0-rings. Tha lid is bolted to the cavity flange by six ASTM-A-320, Grade L43,1-1/4-inch diameter steel bolts. Neutron shielding is provided by a (borated) water-ethylene glycol mixture contained in the neutron shield tank surrounding the outer shell. The outer wall of the tank is about 39 inches in diameter.is constructed of 11-gauge stainless steel, and provides a 4.5-inch thick annular cavity sub-divided into four individually sealed compartments. Each compartment is

e. quipped with a l'A psig rupture disc.

Four surge tanks, located at the upper end of the cask beyond the " active fuel region, are provided to accocmodate the expansion of the liquid neutron shielding mixture. The lid impact limiter is made up of 12-inch thick balsa cylinder, over-laps the outer lid circle, is enclosed within a 12-gauge stainless steel container, is separated from the lid by a 1/8-inch sheet of asbestos, and is attached to the cask lid by four 1-inch diameter bolts. The stainless steel lower impact limiter ring surrounds the cask lower casting, contains a balsa wood dise placed directly under the cask bottom, and has 3/3-inch thick stainless steel gusset plates extending radially from the center of the cask, but starting from the cask outer shell. The limiter acts as pedestal for the vertical support of the cask.

~~~3 7 ~ fl0V 14 1972 C The upper impact limiter consis ts of a stainless s teel-sheathed, b21sa-filled ring which surrounds the cask cavity flange. The upper imoact limiter has a 1/8-inch thick asbestes sheeting between the balsa nnd the cask outer shell. Gusset plates, similar to the ones in the lower impact limiter, are provided within the limiter. The limiter acts as a support member during transport. Two lifting flanged trunnions are attached to the perineter of the upper i= pact limiter. Two other trunnions, located on either side of the lower impact limiter, are pro-vided for rotating'the cask to and from the horizontal position on the trailer. The four trunnions are used for cask tiedown. The fuel assemblies are positioned by the respective fuel canisters shown in Figure 2.1.3. Spacers are used to reduce movement of shorter fuel shipments. Packace contents The contents include the coolant plus one of the following: (1) One (1) NA fuel assembly (2) Two (2) B'.a fuel assemblies. The fuel assenblies are positioned within the appropriate fuel canisters shown in Figure 2.1.3. The maximum weight of the contents is 1500 pounds, excluding the internal coolant and fuel canisters. The maximum amount of decay heat is 22.4 kw. The following table shows other limiting data for the two types of fuel assemblies: Fuel Assem'alv Data P!G BWR Envelope, inches 8.60 x 8.60 x 150 5.44 x 5.44 x 144 ~ Enrichment, w/o U-235 3.6 3.0 Weight of Ura:tiu=, Kg 470 197 5.51 H/U Atomic Ratio Containment The NFS-4 cask is designed to provide containment of the fuel canisters and the coolant under the normal and damage test conditions. Positive closure has been provided at the following points:

3 ) fiOV 14 '972 1. The bolted $otnt between the cask flange and the lid. 2. The drain valves located in the lower shield disc. 3. The vent valve located in the lid flange. 4. The ruptura disc-pressure relief valve system within the cavity flange. Structural Analysis A. Lifting and Tiedcun The upper trunnions function as the cask's lifting devices. A lid lifting mechanism is secured to the lid by four bolts. The trunnions and the bolts have been designed to lift three times the respective lifting load without exceeding the maximum allowable stress of any of their components. The upper and lower trunnions are used as the tiedown system and have been designed to withstand without yielding the 10-g, 5-g and 2-g forces in the respective longitudinal, transverse and vertical direccions. The Applicant has demonstrated that the design meets the above tiedown requirements without exceeding the allowable stresses. We agree with the Applicant that the stresses imposed. on the trunnions during the lif ting or rotation of the cask are less severe than the tiedown lohds. For tiedown the most stressed components of the upper and lower trunnions were found to be the tube and the mounting plate. B. Normal Conditions The main shell, the neutron shield tank and connecting ribs between the two shells have been analyzed individually to determine their ability to support five times the weight of the cask when loaded in a single beam configuration. The maximum stress occurs in the main shell, but all are below the allowable. The two shells are found capable of with-standing an external pressure of 25 psig or 0.5 times the standard atmospheric pressure. We agree that' the side drop was determined to be the most damaging orientation for the one foot drop test. The Applicant considered i= pact on the limiters and on the trunnions. The results of the one foot side O

NOV 14 1972 ( drop on the limiters indicate that they remain functional and that they will not significantly affect any other vital cask component. The one foot drop on the trunnions indicated that the trunnions will deform but remain functional. The drop of the IJ-pound,1-1/4-inch diameter steel cylinder on the neutron shield tank does not produce penetration since the kinetic energy required to shear the shell with the above cylinder is larger than the kinetic e'ergy of the cylinder when dropped from a height of 40 inches. C. Da= ace Cenditions The Applicant considered the mest damaging orientation of impact and determined the respective deformation and g-loads as follows: Mode of Impact Deformation, inches G-load Top 8.06 44.67 Top Corner 12.23 45.50 UFper L1= iter 7.5 96.0 ~ Side Lower Limiter 9.0 96.0 .~ Bottom 4.68 78.8 Bottom Corner 7.95 The Applicant has shown that the impact limiters are capable of absorbing the high energy of impact for the various configurations of drop. The energy is absorbed by the deformation of the impact limiters, which act as sacrificial members. Their deformation produces the g-forces on the cask as indicated. b

( NOV 14 1972 ( The Applicant analyzed the lid bolts and found that they can withstand the maximum tensile loads resulting f rom the corner drop without ex-ceeding their yield stress. Our independent analysis confirmed the Applicant's findings. The Applicant calculated the slump of the lead shield, for a bottom end impact, to be 0.315 inch. Our independent analysis ccnfirmed that the shielding would not be significantly reduced due to the lead alemping. The following orientations were considered in the puncture analysis: (a) On the weaker region of the main outer shell. (b) On the side of the cask, so that its center of gravity is directly above the bar. (c) On the side of the cask lid. (d) On the vent and drain valves. The results of the analysis indicate that no puncture will result on the main shell, bending stress will be less than allowable, there will be a radial displacecent of 0.0044 inch of the lid (contain=ent will be main-tained), and there will be no damage to the vent and drain valves. Upon loss of th.e neutron shield fluid and exposure of the cask to 1475'F for a 1/2-hour fire, the inner shell, the cask lid flange, lid and closure ~ bolts were found to provide containment. Thermal Analysis The cask is designed to contain the central cavity fluid (primary coolant) under both normal and accident damage test conditions. The neutron shield fluid is retained under normal and cold weather conditions. ~ The Applicant presented calculations to determine the pressure in these fluid regions for the stated conditions. Also, an analysis was presented to deternine the magnitude of the central cavity pressure surge which would occur if the exposed fuel were suddenly quenched in the cavity water, thus generating stems (as might occur if the cask were moved following an accident). In cenjunction with the fire cond,. ion analysis, the Applicant presented a parmneter study to determine, within the range of expected emissivities for tne opposing surfaces of the dry neutron shield tank, the maximum cavity water temperature. At the icwcr limit of effective emissivity, N

.s NOV 14 1972 C the equilibrium coolaat temperature following the fire was shown to be controlling, while at the upper limit the peak coolant temperature follow-ing the fire was shown to be controlling. In all cases the Applicant used a one-dimensional radial model and the TAP ccmputer program to solve both steady state and transient problems associated with his analysis. In addition, the Applicant wrote two special programs: CATECA, to calculate the temperatures of the exposed (out of the cavity water) fuel rods, and UNG, which was used in con-junction with CATECA to determine the position of the water level in the central cavity as a function of cask angle away from horizontal. Based on his overall thermal analysis, the Applicant concluded that the cask would comply with the thermal requirements of 10 CFR Part 71. Our analysis of the thermal behavior of the NFS-4 cask was basically in two parts. The first part was a series of HEATING 4 calculations to determine the temperature distributions in the cask for the nomal and fire accident conditions specified in 10 CFR Part 71. The second part was a group of calculations to detemine the pressure in the central cavity for the normal and fire accident conditions. This pressure is of great i=portance in determining the ability of the cask to perfom as predicted by the Applicant and to satisfy the regulations. We identified three parameters whose values might be subject to question and analyzed the effect of changes in these parameters using HEATING 4. The parameters were (1) the effective emissivity of the dry neutron-shield tank walls, (2) the equivalent conductivity of the neutron shield tank fluid, and (3) the film resistance at the central cavity wall. Our independent analysis showed that, even if the critical parameters were varied over some reasonable range, the cavity pressure would remain well below the rupture disc set-point of the cask, and the quenching will probably not produce a significant pressure surge. With regard to the overall thermal review, we note the following: 1. The Applicant's analysis is acceptable. 2. The central cavity pressure is the significant parameter in deter-mining whether the NFS-4 cask can maintain the integrity of its central cavity under the environmental conditions specified in 10 CFB Part 71. We believe we have identified and investigated the aspects of the analysis which significantly affect this pressure-G

] ) NO'l 1 1 1972 3. Our analysis of the pressure surge problem assumes the correctness of the Applicant's C;G and CATECA programs. The Applicant's refer-ence to experimental evidence which verifies his predictions with CATECA supports our assumption. 4. The possible central cavity pressure surge has not been precisely predicted. Such a prediction would almost certainly require a sophisticated, transient thermal-h'/draulic analysis. Our independent analysis shows that, for the expected cavity temperatures, the pressure surge will not cause the cavity pressure to exceed the relief pressure. 5. Based on the Applicant's fire testing of the ball valves that will be used for the drain and vent lines, and our examination of a fire tested valve and a new sample valve, we believe the valves are adequate and should provide contain=ent over the range of temperatures and pressures calculated by the Applicant. Shielding Analysis The Applicant used a 27-20 group neutron gamma ray coupled cross section set (E. D. Arnold, ORNL) in conjunction with the ANISN one-dimensional ( transport program to perform shielding calculations. The gamma source term was stated to be that of mixed fission products for LWR fuels. based on 120 day cooling time. The neutron source term was taken to be 9.2 x 109 neutrons per second. The results of the shielding analysis by the Applicant imilcated that the gamma dose rate would not be affected by loss of the shielding water, but the neutron dose rate would increase by a factor of 27. We independently calculated the gexia ray and neutron dose rates using the AN1SN program and the same gecretric model as the Applicant (see Figure 3.3.1 in the SAR). Our gamn ray source strength was based on an ORIGEN calculation for LWR fuel with a burnup of 33,000 FMD/lfr and 120 days cooling time. We used the same neutron source strength as the Applicant with a newer 21-18 group neutron-gamma ray coupled cross section set (SNAP Library, ORNL). Our celculations indicated that the ga=ma ray dose rate would increase by a factor of 1.4 and the neutron dose rate would increase by a factor of 27 when the shielding water is lost as a result of the accident test series. (These calculations assumed that the water contains a maximum of 1.0 weight percent boron.)

., TT, 'T NOV 14 1972 (.. Criticality Analysis The cask is capable of accoc=odating one (1) PWR type fuel assembly, or two (2) BWR type fuel assemblies. Criticality studies by the Applicant indicated that the BWR and PWR fuels can be shipped as Fissile Class III with one (1) package per shipment. The critica._ y of the proposed shipments was evaluated using the KENO Monte Carlo computer program. A Knight-modified 16-group Hansen-Roach cross section set was used in the calculations. value calculated for the BWR fuel shipment was 0.965 + The maxi =um k.(he PWR fuel shipment 0.009, and foI* the maxi =um k was 0.985 + 0.007. gf Benchmark calculations were performed to assure that the neutron cross section data used in the analyses would be conservative in determining the keff values. The benchmark calculations were correlated with data taken from Yankee critical experiments. Because of the conservative method the Applicant used in determining the op for the U-238 cross section, the benchmark calculations showed the keff value calculated to be 7* conservative. Thus, the kef f values calculated by the Appli-cant for the shipments should be conservative. The Applicant's calculations were spot-checked and found to be accurate. In addition, we are in agreement with the Applicant's methods and ( assumptions. Thus, no independent calculations have been performed. Quality Assurance NFS has established a quality assurance program covering items prel sented in Appendix B of 10 CFR Part 50, including acceptance tests to be conducted prior to first use of the packaging. The package design, coupled with the quality assurance program and the acceptance tests, should ensure that the NFS-4 will perform as designed under the normal and accident conditions of transport. Conditions The safety of the cask was confir=ed by the Transportation Branch on the basis of the following conditions: 1. The maxi =um quantity of material per package shall be limited to one (1) PWR fuel asse=bly or two (2) BWR fuel assemblies in the respective canisters. 2. The maxi =um deca < o(. c generation per package will not exceed 12.4 kw. 3. The maxie ~ c cis

he of the cask shall not exceed 50,000 pounds.

4 The cask shall be shipv6d with coolant. The operating procedures Q.i

S. ] NOV 14 1972 for loading the cask shall require that 24 gallons of coolant shall be drained of f f rom the water filled cask cavity. 5. The coolant is considered part of the package contents. The radio-active contamination limits specified in 71.35(a)(4) do not apply. 6. The water-ethylene glycol mixture in the neutron shield tanks may contain up to 1.0 weight percent boron. This mixture shall not freeze or pescipitate in a temperature range from -40*F to 330*F. 7. When needed, sufficient antifreeze'in the cask cavity shall be used to prevent damage of the package by freezing. 8. The cask contents shall be so limited under normal conditions of transport that the sum of 27 tices the neutron dose rata plus 1.4 times the ga=ma dose rate will not exceed 1,000 millirems per hour at three (3) feet f rom the external surface of the package. 9. The vent and drain valves shall be 1/2" F466T SE Miser ball valves (Worchester Valve Company, Inc.). The ball of the valve may have a bleed hole to equalize the pressure between the cask cavity and the ball passage in the closed position. 10. Periodic maintenance and testing of 0-rings, drain and vent ball valves, relief valves, and rupture discs of the cask shall be as follows: ~ Cask Co=conent Period Test / Action Ball Valve ~ Quarterly Hydro test to 167 psig* Ball Valve Annually Replace seats and seals 0-rings Each shipment Test to 80 psig* . 0-rings Quarterly Test to 167 psig* 0-rings Annually Test to 1006 psig* C vity Relief Valve Quarterly Manually actuated ** C y Relief Valve Annually Test at set point t ty Rupture Disc Quarterly Inspection ty Rupture Disc Annually Replacement t. fleutron Shield Tank Annually Replacement Rupture Discs There shall be no visual (pressure gauge) indications of pressure drop for the component under test during a 10-minute test peried. Otherwise, corrective action shall be taken and the test repeated until such time as the component meets the specified test.

• • Any detectable pressure buildup between the valve and the cavity rupture disc shall reeult in the replacement of the cavity rupture disc.

-) NOV 14 1972 (~ 11. In addition 'o the requirements of Subpart D of 10 CFR Part 71, each package prior to first use shall meet all of the acceptance tests and criteria specified on pages A-21 and A-22 of the final appli-cation dated October 6,1972. The results of the tests shall be documented and retained for the life of the cask. 12. At periodic ratervals not to exceed three (3) years, the thermal performance of the cask shall be analyzed to verify that the cash operation has not degraded below that which is licensed. Following the initial acceptance tests, the heat source may be that provided by the decay heat from the loading of the package, provided that the heat source is equal to at least 25% of the design heat load for the package. Each cask that fails to meet the thermal acceptance criteria given on page A-21 shall be withdrawn from service until corrective action can be completed or the license amended to limit the package to lower heat load. [ Charles E. MacDonald, Chief Transportation Branch { Directorate of Licensing W O

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