[en] This review identifies typical features of the design of transportation flasks and considers some of the analytical tools required for the analysis of impact events. Because of the complexity of the physical problem, it is unlikely that a single code will adequately deal with all the aspects of the impact incident. Candidate codes are identified on the basis of current understanding of their strengths and limitations. It is concluded that the HONDO-II, DYNA3D AND ABAQUS codes which ar already mounted on UKAEA computers will be suitable tools for use in the analysis of experiments conducted in the proposed AEEW programme and of general flask impact problems. Initial attention should be directed at the DYNA3D and ABAQUS codes with HONDO-II being reserved for situations where the three-dimensional elements of DYNA3D may provide uneconomic simulations in planar or axisymmetric geometries. Attention is drawn to the importance of access to suitable mesh generators to create the nodal coordinate and element topology data required by these structural analysis codes. (author)

[en] This document presents the results of a study performed to evaluate the applicability of dunnage material for the load securance of TRU containers internal to the TRUPACT-I cavity. Dunnage was investigated for this purpose as were the use of tie-downs and blocking

[en] A computer program POST-CASKETSS has been developed for the purpose of calculation result representation for thermal and structural analysis computer code system CASKETSS (CASKETSS means a modular code system for CASK Evaluation code system for Thermal and Structural Safety). Main features of POST-CASKETSS are as follows; (1) Function of calculation result representation for thermal and structural analysis computer programs is provided in the program. (2) Two and three dimensional graphic representation for finite element and finite difference programs are available in the program. (3) The capacity of graphics of geometry, temperature contor and temperature-time curve are provided for thermal analysis. (4) The capacity of graphics of geometry, deformation, stress contor, displacement-time curve, velocity-time curve, acceleration-time curve, stress-time curve, force-time curve and moment-time curve are provided for structural analysis. (5) This computer program operates both the time shearing system and the batch system. In the paper, brief illustration of calculation method, input data and sample calculations are presented. (author)

[en] We analyzed the impacts of transportation of research reactor spent fuel from US and foreign reactors for the US Department of Energy's (DOE) Office of Defense Programs. Two separate shipment programs were analyzed. The shipment of research reactor spent fuel from Taiwan to the US (Fuel Movement Program), and the return of research reactor spent fuels of US origin from foreign and domestic reactors (Research Reactor Fuel Return Program). To perform these analyses, a comprehensive methodology for analyzing the probabilities and consequences of transportation in coastal waters and port facilities, handling at the port, and shipment by truck to reprocessing facilities was developed. The Taiwanese fuel consists of low-burnup aluminum-clad metallic uranium research reactor spent fuel; the other fuels are primarily aluminum-clad oxide fuels. The Fuel Movement Program is ongoing, while the Fuel Return Program addresses future shipments over a ten-year period. The operational aspects of the Taiwanese shipments have been uniform, but several possible shipping configurations are possible for the Fuel Return Program shipments. The risks of transporting spent nuclear fuel and other radioactive materials by all modes have been analyzed extensively. Comprehensive assessments, which bound the impacts of spent fuel transport, demonstrate that when shipments are made in compliance with applicable regulations, the risks for all such transport are low. For comparison with previously licensed transport activities and to provide continuity with earlier analyses, the results for shipment of 150-day-old commercial pressurized water reactor (PWR) spent fuel are presented as part of this study

[en] A filtered pressure equalization system (FPES) for a waste transportation package has been designed and qualified to withstand the package environments induced by the normal and accident condition requirements specified in 10 CFR 71. The use of an FPES reduces the package structural requirements to withstand pressure variations due to gas generation from the payload and elevation changes, and permits the design of a more weight efficient package. The filter is capable of trapping airborne particles, should any be released from the contents to the containment vessel. Particle containment is important for flow from inside the package to outside. 1 ref., 5 figs

[en] Since 1962 the US Department of Energy (USDOE) and its predecessor organizations have supported and participated in a series of triennial scientific and technical symposia known as PATRAM, an acronym for the Packaging and Transportation of Radioactive Materials. The purpose of these symposia is to bring together, on an international basis, experts from industry, government, and research organizations to exchange technical information relating to all aspects of the packaging, testing and transport of radioactive materials. Through the sustained excellence of their programs, these symposia have become the premier domestic and international forums for this increasingly important technical field. Worldwide, PATRAM is the only symposium that is devoted entirely to this subject. The Symposia began in December 1962 with a US Atomic Energy Commission meeting held in Germantown, MD, with 20 papers and representatives from five countries

[en] In 1985, the Nuclear Energy Agency-Committee on Reactor Physics (NEACRP) established a working group on shielding assessment of transportation packages. Following the initial distribution of a set of six problems, discussions were held at the Organization for Economic Cooperation and Development (OECD) Headquarters in Paris, France, in June/July 1986, May 1988, and February/March 1990. The US contribution to the working group is documented in this report. The results from this effort permit the evaluation of a number of approximations and effects that must be considered in a typical shielding analysis of a transportation cask. Among the effects reported here are the performance of multiple cross-section sets, the comparison of several source generation codes, and multidimensional versus one-dimensional (1-D) analyses. 18 refs., 16 figs., 33 tabs

[en] Experiments have been conducted to determine the applicability of a chemical ion-exchange model to characterize the problem of nuclear fuel transportation cask contamination and release (''weeping''). Surface charge characteristics of Cr₂O₃ and stainless steel (304) powders have been measured to determine the potential for ion exchange at metal oxide -- aqueous interfaces. The solubility of pool contaminant Co and Cs electrolytes at varying pH and the adsorption characteristics of these ions on Cr₂O₃ and stainless steel powders in aqueous slurries have been studied. Experiments show that Co ions do reversibly adsorb on these powder surfaces and, more specifically, that adsorption occurs in the nominal pH range (pH = 4--6) of a boric acid-moderated spent fuel pool. Desorption has been demonstrated to occur at pH ≤ 3. Cs ions also have been shown to have an affinity for these surfaces although the reversibility of Cs⁺ bonding by H⁺ ion exchange has not been fully demonstrated. These results have significant implications for effective decontamination and coating processes used on nuclear fuel transportation casks. 9 refs., 5 figs., 1 tab

[en] This paper presents an overview of the procedures used at the Argonne National Laboratory (ANL) to review Safety Analysis Reports for Packagings (SARPs) submitted to the US Department of Energy (DOE) for issuance of a Certificate of Compliance. Prior to certification and shipment of a packaging for the transport of radioactive materials, a SARP must be prepared describing the design, contents, analyses, testing, and safety features of the packaging. The SARP must be reviewed to ensure that the specific packaging meets all DOE orders and federal regulations for safe transport. The ANL SARP review group provides an independent review and evaluation function for the DOE to ensure that the packaging meets all the prescribed requirements. This review involves many disciplines and includes evaluating the general information, drawings, construction details, operating procedures, maintenance and test programs, and the quality assurance plan for compliance with requirements. 14 refs., 6 figs

[en] Fracture toughness testing of ductile cast iron (DCI) at elevated loading rates has been performed at Sandia. The intent of the test program was to generate rigorously measured dynamic fracture toughness data in order to enhance the DCI material property data base and to evaluate the effect of rate on the upper-shelf fracture toughness. The test requirements outlined in the ASTM Standard Test Method for J/sub Ic/, A Measure of Fracture Toughness (E 813-87) were adapted to elevated loading rates in a manner which allowed the inherent rigor of this test method to be maintained. The actual measurements were performed on compact tension specimens using a closed-loop servo-hydraulic test frame with special fixturing to enhance test control. The DCI alloy examined was a high quality casting of the type used in European-produced casks (i.e., it had a fully ferritic matrix, with only Type I and II graphite nodules). The test results show that there is essentially no effect of loading rate on the upper-shelf fracture toughness over the range tested. The current results are in contrast to other reports which suggest that there is a decrease in the upper-shelf toughness as the loading rate is raised above a certain level. 11 refs., 4 figs