[en] A preliminary analysis of the concrete cask for the intermediate dry storage of the spent fuel of NPP Krsko should include an estimation of the effective multiplication factor. Assuming 16x16 fuel elements, 4.3% initial enrichment, 45 GWd/tU burnup and 10 years cooling time, a concrete spent fuel capacity of 10 spent fuel assemblies is proposed. Fuel assemblies are placed inside inner cavity in a 'basket' - a boron (1%) doped steel structure. Heavy concrete (25% Fe), 45 cm thick, is enclosed in a carbon steel shell. There is also a stainless steel (SS304) lining of the storage cavity. Isotope inventory of the spent fuel after a 10 years cooling time is calculated using ORIGEN-S functional module of the SCALE-4.2 code package. The effective multiplication factor k_eff of dry (helium filled) and wet (water filled) cask for fresh and used fuel is calculated using CSAS4 Monte Carlo method based control module of the same SCALE-4.2 code package. The obtained results of k_eff of the dry cask for fresh and spent fuel are well below the required 0.95 value, but those for the water filled cask are above this value. Therefore, several additional calculations of the k_eff varying the thickness of a boral basket structure which had replaced the stainless steel one were done. It turned out that at least a 1.5 cm thick boral wall was needed to meet the required 0.95 value for k_eff. (author)

[en] Test facilities at an establishment in Belin-Beliet, France, for packages used to transport radioactive materials are described. The capabilities cover all Type A, Type B and Special Form Material requirements. (author)

[en] The facilities available at the Scalbatraio Laboratory, for testing packages used for the transportation of radioactive materials, are described. (author)

[en] Short communication

[en] A practical procedure has not been developed and accepted internationally to quantify the leakage of aerosols in tests of radioactive transport flasks. Hence it is necessary to understand and quantify aerosol penetration through model pathways that have leakage rates close to the limits set by standards of flask integrity. The penetration of particles from 0.5 to 15 μm volume equivalent diameter has been measured through critical orifices in the range from 2 to 100 μm at thickness ranging from 12.7 to 509 μm. The present study is a limiting case where a capillary is reduced in length until it becomes an orifice. In reality, leaks across seals will normally take the form of short capillaries. A common correlation has been found in the variation between air leakage rate and particle penetration for both capillaries and orifices, and this relationship has enabled mass-based particle penetration rates to be estimated. (author)

[en] Type A package (Sample A) and Type IP-2 package (Sample B) are sometimes used for solid radioactive waste transportation and solid and liquid radioactive waste transportation, respectively. Therefore, Samples A and B were tested by using spray-up, fall, compression, and perforation methods. Although a fall test revealed a slightly small breakage in Sample A, no fluorescence was found in the inner surface of the package. For Sample B, none of the inner change was observed. Neither Sample A nor B was found to have radioactive leakage on the outer surface. Nor was there significant change in dose equivalents. Both Type A and Type IP-2 packages were judged to coincide with the regulatory standards. (N.K.)

[en] It is often desirable to have a package for fissile material shipment certificated for specified fissile and nonfissile mass limits but without restrictions on the disposition of the materials inside the containment vessel. Reasoned arguments are presented which indicate that, provided there are no significant quantities of moderating material in the containment vessel, then analyzing a package with a sphere and shell model for the fissile material will bound the multiplication factor. That is, any other disposition of this same material in the same shipping container and for the same array size will be less reactive. Calculational results are presented which support this argument. (Author)

[en] This document details performance test which was conducted on a Pacific Sierra Nuclear VSC-17 ventilated concrete storage cask configured for pressurized-water reactor (PWR) spent fuel. The performance test consisted of loading the VSC-17 cask with 17 canisters of consolidated PWR spent fuel from Virginia Power's Surry and Florida Power ampersand Light Turkey Point reactors. Cask surface, concrete, air channel surfaces, and fuel canister guide tube temperatures were measured, as were cask surface gamma and neutron dose rates. Testing was performed with vacuum, nitrogen, and helium backfill environments in a vertical cask orientation. Data on spent fuel integrity were also obtained

[en] The natural frequencies and collapse strength of radioactive waste transport casks are analyzed by means of finite element method. The calculated casks are those for high, intermediate and low level waste. The natural frequencies are obtained in two cask-positions, vertical and horizontal. The buckling or collapse strength is analyzed under external pressure. MARC is used as a finite element program. Conclusions obtained are as follows. (1) The natural frequencies calculated by the handbook formulas of beams can be considerably higher than those by finite element method in case of bending mode shapes. (2) The vibration modes of the casks for low or intermediate level waste are mainly the deflection of lid and bottom plates, whose natural frequencies can be estimated by the handbook formulas of circular plate. (3) The collapse strength of casks can be obtained by elastic buckling analysis in case of the cask for low level waste with small rigidity, or by non-linear analysis in case of the cask for high or intermediate level waste with large rigidity. (4) The collapse modes in non-linear analysis are same as the first modes in elastic buckling analysis. (author)

[en] The double lid system includes a box lid with an all round lid seal and a barrel lid. The opening of the barrel is also surrounded by a flange type barrel seal, which corresponds to the radial outer area of the lid seal. To smooth out any unevenness on the parts acting together, the radial inner part of the barrel seal covered by the barrel lid edge has a closed annular air cushion of oval cross section. (DG)