Roald Wigeland

Separations and Transmutation Criteria to Improve Utilization of a Geologic Repository

Abstract This paper describes the results of a study that uses the thermal performance of the repository to establish chemical separations and transmutation criteria for commercial spent nuclear fuel of benefit to a geologic repository, as measured by the allowable increase in utilization of repository space. The method for determining the chemical elements to be separated is based on the thermal performance of the repository. The important chemical elements are identified, the order of importance of the separated elements is established, and the relationship between the efficiency of the chemical separation and the resulting increase in utilization is determined. The proposed repository at Yucca Mountain is used as an example of a geologic repository for the purposes of illustrating the magnitude of the benefits that are possible and the implications for repository size and operation. This work is being done in support of the U.S. Department of Energy Advanced Fuel Cycle Initiative, where numerous reactor, processing, and recycling strategies are being examined to determine the impact on issues important to the viability of nuclear electricity generation, including the disposal of spent nuclear fuel and nuclear waste.

Performance of Metal and Oxide Fuel Cores During Accidents in Large Liquid-Metal-Cooled Reactors

This paper reports on a cooperative effort among European and U.S. analysts, which is an assessment of the comparative safety performance of metal and oxide fuels during accidents in a 3500-MW (thermal), pool-type, liquid-metal-cooled reactor (LMR) is performed. The study focuses on three accident initiators with failure to scram: the unprotected loss-of-flow (ULOF), the unprotected transient overpower, and the unprotected loss-of-heat-sink (ULOHS). Core designs with a similar power output that have been previously analyzed in Europe under ULOF accident conditions are also included in this comparison. Emphasis is placed on identification of design features that provide passive, self-limiting responses to postulated accident conditions and quantification of relative safety margins. The analyses show that in ULOF and ULOHS sequences, metal-fueled LMRs with pool-type primary systems provide larger temperature margins to coolant boiling than do oxide-fueled reactors of the same design.

Impact on geologic repository usage from limited actinide recycle in pressurized light water reactors.

A project has been conducted as part of the U.S. Department of Energy Advanced Fuel Cycle Initiative to evaluate the impact of limited actinide recycling in light water reactors on the utilization of a geologic repository where loading of the repository is constrained by the decay heat of the emplaced materials. In this study, it was assumed that spent PWR fuel was processed, removing the uranium, plutonium, americium, and neptunium, along with the fission products cesium and strontium. Previous work had demonstrated that these elements were responsible for limiting loading in the repository based on thermal constraints. The plutonium, americium, and neptunium were recycled in a PWR, with process waste and spent recycled fuel being sent to the repository. The cesium and strontium were placed in separate storage for 100–300 years to allow for decay prior to disposal. The study examined the effect of single and mutliple recycles of the recovered plutonium, americium, and neptunium, as well as different processing delay times. The potential benefit to the repository was measured by the increase in utilization of repository space as indicated by the allowable linear loading in the repository drifts (tunnels). The results showed that limited recycling would provide only a small fraction of the benefit that could be achieved with repeated processing and recycling, as is possible in fast neutron reactors.

Economic Analysis of Complex Nuclear Fuel Cycles with NE-COST

Abstract The purpose of this work is to present a new methodology and the associated computational tools developed within the U.S. Department of Energy Fuel Cycle Options Campaign to quantify the economic performance of complex nuclear fuel cycles. The levelized electricity cost at the busbar is generally chosen to quantify and compare the economic performance of different base load—generating technologies, including nuclear; the levelized electricity cost is the cost that renders the risk-adjusted discounted net present value of the investment cash flow equal to zero. The work presented here is focused on the calculation of the levelized cost of electricity of fuel cycles at mass balance equilibrium, which is termed levelized cost of electricity at equilibrium (LCAE). To alleviate the computational issues associated with the calculation of the LCAE for complex fuel cycles, a novel approach has been developed. This approach has been termed the island approach because of its logical structure, in which a generic complex fuel cycle is subdivided into subsets of fuel cycle facilities called islands, each containing one and only one type of reactor or blanket and an arbitrary number of fuel cycle facilities. A nuclear economic software tool, NE-COST, written in the commercial programming software MATLAB©, has been developed to calculate the LCAE of complex fuel cycles with the island computational approach. NE-COST has also been developed with the capability to handle uncertainty: the input parameters (both unit costs and fuel cycle characteristics) can have uncertainty distributions associated with them, and the output can be computed in terms of probability density functions of the LCAE. In this paper, NE-COST will be used to quantify, as examples, the economic performance of (a) once-through systems of current light water reactors (LWRs), (b) continuous plutonium recycling in fast reactors (FRs) with drivers and blankets, and (c) recycling of plutonium bred in FRs into LWRs. For each fuel cycle, the contributions to the total LCAE of the main cost components will be identified.

Thorium Fuel Cycle Option Screening in the United States

Abstract As part of a nuclear fuel cycle evaluation and screening (E&S) study, widely ranging thorium fuel cycle options were evaluated, and their performance characteristics and challenges to implementation were compared to those of other nuclear fuel cycle options based on criteria specified by the Nuclear Energy Office of the U.S. Department of Energy. The evaluated nuclear fuel cycles included the once-through, limited, and continuous recycle options using critical or externally driven nuclear energy systems. The E&S study found that the continuous recycle of 233U/Th in fuel cycles using either thermal or fast reactors is an attractive promising fuel cycle option with high effective fuel resource utilization and low waste generation, but they did not perform quite as well as the continuous recycle of Pu/U using a fast critical system, which was identified as one of the most promising fuel cycle options in the E&S study. This is because compared to their uranium counterparts, the thorium-based systems tended to have higher radioactivity in the short term (~100 years postirradiation), because of differences in the fission product yield curves, and in the long term (100 000 years postirradiation), because of the decay of 233U and daughters, and because of higher mass flow rates due to lower discharge burnups. Some of the thorium-based systems also require enriched uranium support, which tends to be detrimental to resource utilization and waste generation metrics. Finally, similar to the need to develop recycle fuel fabrication, fuels separations, and fast reactors for the most promising options using Pu/U recycle, the future thorium-based fuel cycle options with continuous recycle would also require such capabilities; however, their deployment challenges are expected to be greater since past development of such facilities has not reached a comparable level of maturity.

Reactor-based management of used nuclear fuel: assessment of major options.

This paper discusses the current status of the ongoing Advanced Fuel Cycle Initiative (AFCI) program in the U.S. Department of Energy that is investigating the potential for using the processing and recycling of used nuclear fuel to improve radioactive waste management, including used fuel. A key element of the strategies is to use nuclear reactors for further irradiation of recovered chemical elements to transmute certain long-lived highly-radioactive isotopes into less hazardous isotopes. Both thermal and fast neutron spectrum reactors are being studied as part of integrated nuclear energy systems where separations, transmutation, and disposal are considered. Radiotoxicity is being used as one of the metrics for estimating the hazard of used fuel and the processing of wastes resulting from separations and recycle-fuel fabrication. Decay heat from the used fuel and/or wastes destined for disposal is used as a metric for use of a geologic repository. Results to date indicate that the most promising options appear to be those using fast reactors in a repeated recycle mode to limit buildup of higher actinides, since the transuranic elements are a key contributor to the radiotoxicity and decay heat. Using such an approach, there could be much lower environmental impact from the high-level waste as compared to direct disposal of the used fuel, but there would likely be greater generation of low-level wastes that will also require disposal. An additional potential waste management benefit is having the ability to tailor waste forms and contents to one or more targeted disposal environments (i.e., to be able to put waste in environments best-suited for the waste contents and forms).