Publications

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A slight modification of a package to transport solid metal contents requires inclusion of a thin titanium liner to protect against possible eutectic formation in 10 CFR 71.74 regulatory fire accident conditions. Under severe transport regulatory impact conditions, the package contents could impart high localized loading of the liner, momentarily pinching it between the contents and the thick containment vessel, and inducing some plasticity near the contact point. Actuator and drop table testing of simulated contents impacts against liner/containment vessel structures nearly bounded the potential plastic strain and stress triaxiality conditions, without any ductile tearing of the eutectic barrier. Additional bounding was necessary in some cases beyond the capability of the actuator and drop table tests, and in these cases a stress-modified evolution integral over the plastic strain history was successfully used as a failure criterion to demonstrate that structural integrity was maintained. The Heaviside brackets only allow the evolution integral to accumulate value when the maximum principal stress is positive, since failure is never observed under pure hydrostatic pressure, where the maximum principal stress is negative. Detailed finite element analyses of myriad possible impact orientations and locations between package contents and the thin eutectic barrier under regulatory impact conditions have shown that not even the initiation of a ductile tear occurs. Although localized plasticity does occur in the eutectic barrier, it is not the primary containment boundary and is thus not subject to ASME stress allowables from NRC Regulatory Guide 7.6. These analyses were used to successfully demonstrate that structural integrity of the eutectic barrier was maintained in all 10 CFR 71.73 and 71.74 regulatory accident conditions. The NRC is currently reviewing the Safety Analysis Report.

The Plutonium Air Transportable Package, Model PAT-1, is certified under Title 10, Code of Federal Regulations Part 71 by the U.S. Nuclear Regulatory Commission (NRC) per Certificate of Compliance (CoC) USA/0361B(U)F-96 (currently Revision 9). The National Nuclear Security Administration (NNSA) submitted SAND Report SAND2009-5822 to NRC that documented the incorporation of plutonium (Pu) metal as a new payload for the PAT-1 package. NRC responded with a Request for Additional Information (RAI), identifying information needed in connection with its review of the application. The purpose of this SAND report is to provide the authors responses to each RAI. SAND Report SAND2010-6106 containing the proposed changes to the Addendum is provided separately.

The Plutonium Air Transportable Package, Model PAT-1, is certified under Title 10, Code of Federal Regulations Part 71 by the U.S. Nuclear Regulatory Commission (NRC) per Certificate of Compliance (CoC) USA/0361B(U)F-96 (currently Revision 9). The purpose of this SAR Addendum is to incorporate plutonium (Pu) metal as a new payload for the PAT-1 package. The Pu metal is packed in an inner container (designated the T-Ampoule) that replaces the PC-1 inner container. The documentation and results from analysis contained in this addendum demonstrate that the replacement of the PC-1 and associated packaging material with the T-Ampoule and associated packaging with the addition of the plutonium metal content are not significant with respect to the design, operating characteristics, or safe performance of the containment system and prevention of criticality when the package is subjected to the tests specified in 10 CFR 71.71, 71.73 and 71.74.

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The National Nuclear Security Administration (NNSA) has submitted an application to the Nuclear Regulatory Commission (NRC) for the air shipment of plutonium metal within the Plutonium Air Transportable (PAT-1) packaging. The PAT-1 packaging is currently authorized for the air transport of plutonium oxide in solid form only. The INMM presentation will provide a limited overview of the scope of the plutonium metal initiative and provide a status of the NNSA application to the NRC.

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The Waste Management Symposium 2007 is the most recent in a long series that has been held at Tucson, Arizona. The meeting has become extremely popular as a venue for technical exchange, marketing, and networking involving upward of 1800 persons involved with various aspects of radioactive waste management. However, in a break with tradition, the symposium organizers reported that next year’s Waste Management Symposium would be held at the Phoenix, AZ convention center. Additionally, most of the WM07 sessions dealt with the technical and institutional issues relating to the resolution of waste disposal and processing challenges, including a number of sessions dealing with related transport activities.

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This multinational test program is quantifying the aerosol particulates produced when a high energy density device (HEDD) impacts surrogate material and actual spent fuel test rodlets. The experimental work, performed in four consecutive test phases, has been in progress for several years. The overall program provides needed data that are relevant to some sabotage scenarios in relation to spent fuel transport and storage casks, and associated risk assessments. This program also provides significant political benefits in international cooperation for nuclear security related evaluations. The spent fuel sabotage--aerosol test program is coordinated with the international Working Group for Sabotage Concerns of Transport and Storage Casks (WGSTSC), and supported by both the U.S. Department of Energy and Nuclear Regulatory Commission. This report summarizes the preliminary, Phase 1 work performed in 2001 and 2002 at Sandia National Laboratories and the Fraunhofer Institute, Germany, and documents the experimental results obtained, observations, and preliminary interpretations. Phase 1 testing included: performance quantifications of the HEDD devices; characterization of the HEDD or conical shaped charge (CSC) jet properties with multiple tests; refinement of the aerosol particle collection apparatus being used; and, CSC jet-aerosol tests using leaded glass plates and glass pellets, serving as representative brittle materials. Phase 1 testing was quite important for the design and performance of the following Phase 2 test program and test apparatus.

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Effective management of DOE`s transportation operations requires better data than are currently available, a more integrated management structure for making transportation decisions, and decision support tools to provide needed analysis capabilities. This paper describes a vision of an advanced logistics management system for DOE, and the rationale for developing improved modeling and simulation capability as an integral part of that system. The authors illustrate useful types of models through four examples, addressing issues of transportation package allocation, fleet sizing, routing/scheduling, and emergency responder location. The overall vision for the advanced logistics management system, and the specific examples of potential capabilities, provide the basis for a conclusion that such a system would meet a critical DOE need in the area of radioactive material and waste transportation.

The US DOE has amassed over 555,000 metric tons of depleted uranium from its uranium enrichment operations. Rather than dispose of this depleted uranium as waste, this study explores a beneficial use of depleted uranium as metal shielding in casks designed to contain canisters of vitrified high-level waste. Two high-level waste storage, transport, and disposal shielded cask systems are analyzed. The first system employs a shielded storage and disposal cask having a separate reusable transportation overpack. The second system employs a shielded combined storage, transport, and disposal cask. Conceptual cask designs that hold 1, 3, 4 and 7 high-level waste canisters are described for both systems. In all cases, cask design feasibility was established and analyses indicate that these casks meet applicable thermal, structural, shielding, and contact-handled requirements. Depleted uranium metal casting, fabrication, environmental, and radiation compatibility considerations are discussed and found to pose no serious implementation problems. About one-fourth of the depleted uranium inventory would be used to produce the casks required to store and dispose of the nearly 15,400 high-level waste canisters that would be produced. This study estimates the total-system cost for the preferred 7-canister storage and disposal configuration having a separate transportation overpack would be $6.3 billion. When credits are taken for depleted uranium disposal cost, a cost that would be avoided if depleted uranium were used as cask shielding material rather than disposed of as waste, total system net costs are between $3.8 billion and $5.5 billion.

In the early 1980s, the US Department of Energy (DOE) implemented a program to encourage beneficial uses of nuclear byproduct materials, such as cesium-137 and strontium-90, created during the production of defense materials. Potential uses of the cesium-137 ({sup 137}CS) isotope included sterilizing medical products, maintaining the quality of certain food products, and disinfecting municipal sewage sludge. Strontium-90 ({sup 90}Sr) is a good heat source and has been used in thermoelectric generators and other products that require a constant supply of heat. During that same period, a proposed facility in Albuquerque, New Mexico, was designed to use cesium-137 to sterilize sewage sludge. To support the sewage sludge treatment facility, Sandia National Laboratories was funded by the DOE to develop a Nuclear Regulatory Commission (NRC)-certified Type B shipping container to transport cesium chloride (CsCl) or strontium fluoride (SrF{sub 2}) capsules produced by the Hanford Waste Encapsulation and Storage Facility (WESF) in the State of Washington. The primary purpose of the Beneficial Uses Shipping System (BUSS) cask is to provide shielding and confinement, as well as impact, puncture, and thermal protection for certified, special form contents during transport under normal and hypothetical accident conditions. The BUSS cask was designed to meet dimensional and weight constraints of the WESF and user facilities. Attaining as-low-as-reasonably-achievable (ALARA) radiation exposures in the design and operation of the transport system was a major design goal. Another goal was to obtain regulatory approval of the design by preparing a safety analysis report for packaging (SARP) (Yoshimura et al. 1993).

The underlying report for this paper evaluates options for using depleted uranium as shielding materials for transport systems for disposal of vitrified high-level waste (VHLW). In addition, economic analyses are presented to compare costs associated with these options to costs, associated with existing and proposed storage, transport, and diposal capabilities. A more detailed evaluation is provided elsewhere. (Yoshimura et al. 1995.)

The Beneficial Uses Shipping System cask is a Type B packaging developed by Sandia National Laboratories for the U.S. Department of Energy. The cask is designed to transport special form radioactive source capsules (cesium chloride and strontium fluoride) produced by the Department of Energy`s Hanford Waste Encapsulation and Storage Facility. This paper describes the cask system and the analyses performed to predict the response of the cask in impact, puncture, and fire accident conditions as specified in the regulations. The cask prototype has been fabricated and Certificates of Compliance have been obtained.

The Shippable Storage Cask Demonstration Project is intended to demonstrate casks which can be used for both shipping and storing spent nuclear fuel assemblies. The demonstration included the requirement that the casks be certified for shipping by the US Nuclear Regulatory Commission (NRC). After a lengthy review process which resulted in the resolution of several important technical issues, designs for two similar casks have been certified. This paper describes the certification phase of the demonstration. Based on experience gained during certification phase of the demonstration. Based on experience gained during certification, observations and recommendations have been developed which can benefit others seeking NRC approval of transportation cask designs.

Static and dynamic analyses of an impact limiter for a spent fuel cask have been performed using the finite element analysis code PRONTO2D (Taylor and Flanagan, 1987). The impact limiter contained wood as the energy absorbing material, with the wood confined by a cylindrical metal outer skin and sixteen metal stiffeners (gussets). The object of these analyses was to determine how the wood interacts with the metal stiffeners and to determine if the impact limiter would behave differently under static versus dynamic loading conditions. Originally, the metal gusset strength was assumed to be limited by the elastic buckling load. Further analysis showed that the gusset strength was not limited to the elastic buckling load and that each gusset contributed significantly to the impact limiter's strength. The current analyses investigated the strength of a flat plate or gusset used in impact limiter systems. 3 refs., 6 figs.

Rigid, closed-cell, polyurethane foam is used in impact limiters in nuclear waste transport containers. During a hypothetical nuclear waste transport accident, the foam is expected to absorb a significant amount of impact energy by undergoing large inelastic volume reductions. Consequently, the crushing of polyurethane foams must be well characterized and accurately modeled to properly analyze a transport container accident. At the request of Sandia National Laboratories, a series of uniaxial, hydrostatic and triaxial compression tests on polyurethane foams were performed by the New Mexico Engineering Research Institute (NMERI). The combination of hydrostatic and triaxial tests was chosen to provide sufficient data to characterize both the volumetric and deviatoric behaviors of the foams and the coupling between the two responses. Typical results from the NMERI tests are included in this paper. A complete description of these tests can be found in Neilsen et al., 1987. Constitutive models that have been used in the past to model foam did not capture some important foam behaviors observed in the NMERI tests. Therefore, a new constitutive model for rigid, closed-cell, polyurethane foams was developed and implemented in two finite element codes. Development of the new model is discussed in this paper. Also, results from analyses with the new model and other constitutive models are presented to demonstrate differences between the various models. 4 refs., 6 figs., 1 tab.

Wood is often used as the energy absorbing material in impact limiters, because it begins to crush at low strains, then maintains a near constant crush stress up to nearly 60 percent volume reduction, and there ''locks up.'' Hill has performed tests that show how wood is one of the best absorbers of energy per pound. However, wood's orthotropic behavior for large crush is difficult to model. In the past, analysts have used isotropic foam-like material models for modeling wood. A new finite element technique is presented in this paper that gives a better model of wood crush than the model currently in use. The orthotropic technique is based on locally isotropic, but globally orthotropic (LIGO) assumptions in which alternating layers of hard and soft crushable material are used. Each layer is isotropic; however, by alternating hard and soft thin layer, the resulting global behavior is orthotropic. In the remainder of this paper, the new technique for modeling orthotropic wood crush will be presented. The model is used to predict the crush behavior for different grain orientations of a 5 /times/ 5 inch sample of balsa wood. As an example problem, an impact limiter containing balsa wood as the crushable material is analyzed using both an isotropic model and the alternating layer model. 9 refs., 7 figs.