Ruth F. Weiner

RadCat 2.0 User Guide.

This document provides a detailed discussion and a guide for the use of the RADCAT 2.3 Graphical User Interface input file generator for the RADTRAN code. The differences between RADCAT 2.3 and RADCAT 2.2 can be attributed to the addition of the graphical outputs, and the revisions within RADTRAN 5.6. As of this writing, the RADTRAN version in use is RADTRAN 5.6.

RADTRAN 6 Technical Manual

This Technical Manual contains descriptions of the calculation models and mathematical and numerical methods used in the RADTRAN 6 computer code for transportation risk and consequence assessment. The RADTRAN 6 code combines user-supplied input data with values from an internal library of physical and radiological data to calculate the expected radiological consequences and risks associated with the transportation of radioactive material. Radiological consequences and risks are estimated with numerical models of exposure pathways, receptor populations, package behavior in accidents, and accident severity and probability.

RADTRAN 6/RadCat 6 user guide.

This document provides a detailed discussion and a guide for the use of the RadCat 6.0 Graphical User Interface input file generator for the RADTRAN code, Version 6. RadCat 6.0 integrates the newest analysis capabilities of RADTRAN 6.0, including an economic model, updated loss-of-lead shielding model, a new ingestion dose model, and unit conversion. As of this writing, the RADTRAN version in use is RADTRAN 6.02.

Thermal analysis of proposed transport cask for three advanced burner reactor used fuel assemblies

Abstract Preliminary studies of used fuel generated in the US Department of Energys Advanced Fuel Cycle Initiative have indicated that current used fuel transport casks may be insufficient for the transportation of said fuel. This work considers transport of three 5-year-cooled oxide advanced burner reactor used fuel assemblies with a burn-up of 160 MWD kg–1. A transport cask designed to carry these assemblies is proposed. This design employs a 7-cm-thick lead gamma shield and a 20-cm-thick NS-4-FR composite neutron shield. The temperature profile within the cask, from its centre to its exterior surface, is determined by two-dimensional computational fluid dynamics simulations of conduction, convection and radiation within the cask. Simulations are performed for a cask with a smooth external surface and various neutron shield thicknesses. Separate simulations are performed for a cask with a corrugated external surface and a neutron shield thickness that satisfies shielding constraints. Resulting temperature profiles indicate that a three-assembly cask with a smooth external surface will meet fuel cladding temperature requirements but will cause outer surface temperatures to exceed the regulatory limit. A cask with a corrugated external surface will not exceed the limits for both the fuel cladding and outer surface temperatures.

Evaluating Igneous Activity at Yucca Mountain

The U.S. Department of Energy (DOE) plans to submit a license application in 2008 to the U.S. Nuclear Regulatory Commission (NRC) to construct a repository for high-level radioactive waste and spent nuclear fuel at Yucca Mountain, Nevada. One challenge of the NRCs licensing decision is the evaluation of the potential risk from release of radioactive material by igneous activity at Yucca Mountain during the approximately 1 million year lifetime of the repository. A volcano such as the nearby Lathrop Wells volcano (Figure 1) could erupt through the repository (extrusive scenario), or an igneous dike could intersect it (intrusive scenario). Although the likelihood of either at Yucca Mountain is very low, each is being evaluated.

Ingestion dose model for transportation risk assessment

Abstract Risks of transporting radioactive materials can be estimated using the programme and code RADTRAN. Potential radiation doses to various receptors are calculated by RADTRAN, including doses from routine, incident free transportation and from transportation accidents. If radioactive material is released from a transportation vehicle in an accident, agricultural products in the plume footprint could be contaminated. This paper discusses a method for calculating radiation doses from ingestion of such radioactively contaminated food stuffs. Transportation of radioactive materials occurs throughout the USA, so that agricultural products along many transportation corridors could be affected. However, doses from ingesting agricultural crops contaminated from a traffic accident would be very small compared to natural background radiation.

RadCat 3.0 user guide.

RADTRAN is an internationally accepted program and code for calculating the risks of transporting radioactive materials. The first versions of the program, RADTRAN I and II, were developed for NUREG-0170 (USNRC, 1977), the first environmental statement on transportation of radioactive materials. RADTRAN and its associated software have undergone a number of improvements and advances consistent with improvements in both available data and computer technology. The version of RADTRAN currently bundled with RadCat is RADTRAN 6.0. This document provides a detailed discussion and a guide for the use of the RadCat 3.0 Graphical User Interface input file generator for the RADTRAN code. RadCat 3.0 integrates the newest analysis capabilities of RADTRAN 6.0 which includes an economic model, updated loss-of-lead shielding model, and unit conversion. As of this writing, the RADTRAN version in use is RADTRAN 6.0.

RADTRAN/RADCAT user guide.

RADTRAN is a program and code for calculating the risks of transporting radioactive materials. The first versions of the program, RADTRAN I and II, were developed for NUREG-0170 (USNRC, 1977), the first environmental impact statement on transportation of radioactive materials. RADTRAN and its associated software have undergone a number of improvements and advances consistent with improvements in computer technology.

Safety case for transporting spent nuclear fuel

Abstract The transportation safety case for transporting spent nuclear fuel is a requirement for licensing. It has both qualitative and semiquantitative aspects. The qualitative aspects include transportation regulations, radiation dose limits, role of the transportation package in transportation, transportation package certification process, training, emergency response, the performance of the transportation package in accidents and the evaluation of past transportation accidents. The quantitative aspects support the qualitative descriptions. Radiation doses accrued by members of the public and by workers are calculated using the code RADTRAN. Dose from both routine, incident free highway transportation and from highway transportation accidents are part of the safety case and will be compared with both background doses and the regulatory safety criteria. The radiation doses from routine transportation are calculated for the following: the maximally exposed member of the public, doses to vehicle escorts and doses to vehicle crew. Collective doses to populations are calculated for representative routes. Collective dose depends on the number of people affected as well as on the extent of the radiation from the source to which reference groups are exposed. Accidents involving loss of gamma shielding and loss of confinement integrity are discussed, as are accidents in which there is no impact on the cargo.

Spent fuel transportation risk assessment: overview

Abstract The US Nuclear Regulatory Commission (NRC) is responsible for issuing regulations for the packaging of spent fuel (and other large quantities of radioactive material) for transport that provide for public health and safety during transport [Title 10 of the Code of Federal Regulations (10 CFR) Part 71, ‘Packaging and transportation of radioactive waste’, dated 26 January 2004]. In September 1977, the NRC published NUREG 0170, ‘Final environmental statement on the transportation of radioactive material by air and other modes’, which assessed the adequacy of those regulations to provide safety assurance. In that assessment, the measure of safety was the risk of radiation doses to the public under routine and accident transport conditions, and the risk was found to be acceptable. Since that time, there have been two affirmations of this conclusion for spent nuclear fuel (SNF) transportation, each using improved tools and information that supported the earlier studies. This report presents the results of a fourth investigation into the safety of SNF transportation. The risks associated with SNF transportation come from the radiation that the spent fuel gives off, which is attenuated, but not eliminate, by the transportation casks shielding, and the possibility of the release of some quantity of radioactive material during a severe accident. This investigation shows that the risk from the radiation emitted from the casks is a small fraction of naturally occurring background radiation and the risk from accidental release of radioactive material is several orders of magnitude less. Because there have been only minor changes to the radioactive material transportation regulations between NUREG 0170 and this risk assessment, the calculated dose due to the external radiation from the cask under routine transport conditions is similar to what was found in earlier studies. The improved analysis tools and techniques, improved data availability, and a reduction in the number of conservative assumptions has made the estimate of accident risk from the release of radioactive material in this study approximately five orders of magnitude less than what was estimated in NUREG 0170. The results demonstrate that NRC regulations continue to provide adequate protection of public health and safety during the transportation of SNF.