Publications

Abstract not provided.

This report summarizes the discussion and conclusions reached during a table top exercise held at Sandia National Laboratories, Albuquerque on September 3, 2014 regarding a recently described approach for nuclear warhead verification based on the cryptographic concept of a zero-knowledge protocol (ZKP) presented in a recent paper authored by Glaser, Barak, and Goldston. A panel of Sandia National Laboratories researchers, whose expertise includes radiation instrumentation design and development, cryptography, and arms control verification implementation, jointly reviewed the paper and identified specific challenges to implementing the approach as well as some opportunities. It was noted that ZKP as used in cryptography is a useful model for the arms control verification problem, but the direct analogy to arms control breaks down quickly. The ZKP methodology for warhead verification fits within the general class of template-based verification techniques, where a reference measurement is used to confirm that a given object is like another object that has already been accepted as a warhead by some other means. This can be a powerful verification approach, but requires independent means to trust the authenticity of the reference warhead - a standard that may be difficult to achieve, which the ZKP authors do not directly address. Despite some technical challenges, the concept of last-minute selection of the pre-loads and equipment could be a valuable component of a verification regime.

Abstract not provided.

Abstract not provided.

Abstract not provided.

Abstract not provided.

Abstract not provided.

Abstract not provided.

Abstract not provided.

Approximately 13% by volume of the US Department of Energy (DOE) current backlog of radioactive waste is characterized as high-level waste. Transportation of these wastes requires that the waste package have adequate shielding against gamma radiation. This project investigates the radiation shielding performance of titanium and depleted uranium, which have been proposed for use as gamma shielding materials in DOE transportation packages, by experimentally determining their buildup factors. Buildup factors are important in shield heating and radiation damage calculations. A point-isotropic-source type of buildup factor is the most useful for application in the point-kernal approach utilized in many simple shielding codes. The point-kernal method provides reasonable results for cases in which the shield is made of one solid material and the source can be approximated as one homogeneous material. The point-kernal method has been incorporated into a large number of shielding codes treating three-dimensional geometry using buildup factor data in some form. Buildup factors vary with a number of parameters such as the distance of penetration through the attenuating medium; the geometric configuration of the attenuating medium, source and detector position; the composition of the medium; the detector response function; and the energy and direction of emission of the source photons, ideally taken to be monoenergetic and isotropic.

This paper discusses the development of the software for Source Term Analyses for Containment Evaluations (STACE). This software is being developed for the Source Term Technical Issue Resolution Program at Sandia National Laboratories (SNL) in support of the Cask Systems Development Program (CSDP) that is sponsored by the US Department of Energy`s Office of Civilian Radioactive Waste Management (OCRWM). STACE is a system of computer codes operating under a graphics-based controller that performs source term analysis of spent fuel transport casks. Output from STACE includes the cladding breach probability, the releasable radionuclide concentrations, and maximum permissible gas flow rates past the closure seals. STACE is anticipated being used for on- and off-site situations related to the handling and transport of spent fuel casks.

This paper presents a methodology for determining the response of spent fuel assembly spacer grids subjected to transport cask impact loading. The spacer grids and their interaction with rod-to-rod loading are the most critical components governing the structural response of spent fuel assemblies. The purpose of calculating the assembly response is to determine the resistance to failure of spent fuel during regulatory transport. The failure frequency computed from these analyses is used in calculating category B spent fuel cask containment source term leakage rates for licensing calculations. Without defensible fuel rod failure frequency prediction calculations, assumptions of 100% fuel failure must be made, leading to leak tight cask design requirements.

A methodology for determining the probability spent-fuel cladding breach due to normal and accident class B cask transport conditions is introduced. This technique uses deterministic stress analysis results as well as probabilistic cladding material properties, initial flaws, and breach criteria. Best estimates are presented for the probability distributions of irradiated Zircaloy properties such as ductility and fracture toughness, and for fuel rod initial conditions such as manufacturing flaws and PCI part-wall cracks. Example analyses are used to illustrate the implementation of this methodology for a BWR (GE 7 {times} 7) and a PWR (B&W 15 {times} 15) assembly. The cladding breach probabilities for each assembly are tabulated for regulatory normal and accident transport conditions including fire.