Temitope A. Taiwo

Variations in nuclear waste management performance of various fuel-cycle options

The variations in the nuclear waste management performance have been assessed for 40 fuel-cycle examples with the calculation of the mass, activity, toxicity and decay heat of spent nuclear fuel and high-level waste at 10, 100 and 100,000 years after reactor discharge. The variation in the 10 years activity is primarily due to the variation in the specific activity of the fission products, which is higher for thorium fuel cycles, and is reduced when the fuel residence time is long. The variation in the 100,000 years activity is primarily explained by the quantity of U-233 and Pu-239 sent to nuclear waste, which is linked to the type of fuel and of reprocessing scheme employed. The difference between the inhalation toxicity and the activity is explained by the variations in the effective dose coefficients since heavy actinides such as Pu, Am and Cm have a predominant effect on the inhalation toxicity. Materials for disposal such as fission products and transuranics are responsible for most of the mass, and radiotoxicity of high-level waste, but the reprocessing/separation losses also have a potentially significant impact on the results.

Evaluation of Long-Life Transuranics Breakeven and Burner Cores for Waste Minimization in a PB-GCFR Fuel Cycle

Transuranics (TRU) breakeven and burner core designs have been studied for the Pebble-Bed Gas-Cooled Fast Reactor (PB-GCFR), which was developed under a 2-yr U.S. Department of Energy Nuclear Energy Research Initiative project. The issues of minimizing waste production, fuel cost, and burnup reactivity swing, and maximizing TRU burning have been investigated primarily from a neutronics viewpoint. For TRU breakeven cores, it was found that for the given core power [300 MW(thermal)] and power density (50 MW/m3), the lowest amount of radiotoxic TRU to be processed is obtained for a long-life (single-batch) core of 30-yr duration. Minimizing the TRU processed results in a minimization of the TRU losses that ultimately will have to be entombed in a geologic repository. The results show that the single-batch, long-life PB-GCFR could be designed to operate over a wide range of cycle lengths and fuel loadings. By modifying the TRU feed to have a higher minor actinide (MA) fraction than contained in light water reactor spent fuel, the burnup reactivity swing for the long-life core can be reduced significantly. With this approach, it is also possible to configure the long-life PB-GCFR core as a TRU burner using nonuranium fuel. A nonuranium fuel PB-GCFR with 24% plutonium and 76% MAs can operate for 17 full-power years and achieve 25% burnup with a reactivity swing of 3%Δk.

Feasibility of Recycling Plutonium and Minor Actinides in Light Water Reactors Using Hydride Fuel

The objective of this DOE NERI program sponsored project was to assess the feasibility of improving the plutonium (Pu) and minor actinide (MA) recycling capabilities of pressurized water reactors (PWRs) by using hydride instead of oxide fuels. There are four general parts to this assessment: 1) Identifying promising hydride fuel assembly designs for recycling Pu and MAs in PWRs 2) Performing a comprehensive systems analysis that compares the fuel cycle characteristics of Pu and MA recycling in PWRs using the promising hydride fuel assembly designs identified in Part 1 versus using oxide fuel assembly designs 3) Conducting a safety analysis to assess the likelihood of licensing hydride fuel assembly designs 4) Assessing the compatibility of hydride fuel with cladding materials and water under typical PWR operating conditions Hydride fuel was found to offer promising transmutation characteristics and is recommended for further examination as a possible preferred option for recycling plutonium in PWRs.

Impacts of Heterogeneous Recycle in Fast Reactors on Overall Fuel Cycle

A study in the United States has evaluated the attributes of the heterogeneous recycle approach for plutonium and minor actinide transmutation in fast reactor fuel cycles, with comparison to the homogeneous recycle approach, where pertinent. The work investigated the characteristics, advantages, and disadvantages of the approach in the overall fuel cycle, including reactor transmutation, systems and safety impacts, fuel separation and fabrication issues, and proliferation risk and transportation impacts. For this evaluation, data from previous and ongoing national studies on heterogeneous recycle were reviewed and synthesized. Where useful, information from international sources was included in the findings. The intent of the work was to provide a comprehensive assessment of the heterogeneous recycle approach at the current time.

Numerical Benchmarks for Very High-Temperature Reactors Based on the CNPS Critical Experiments

An evaluation of the Compact Nuclear Power Source (CNPS) experiments conducted at Los Alamos National Laboratory in the 1980s has been done using information available in the open literature. The MCNP4C Monte Carlo results for critical test configurations are in good agreement with the experimental values; the keff values are generally within 0.5% of the experimental values. The calculated total and differential rod worths and material worths were also found generally close to experimental values. These good results motivated the utilization of the experimental test data for the specification of two- and three-dimensional numerical benchmark cases that could be used for the verification and validation of core physics codes developed for Very High Temperature Reactor (VHTR) analysis, particularly the deterministic lattice and whole-core physics codes. To define the benchmark cases, the irregular arrangement of channels in the actual CNPS core was simplified to a regular Cartesian geometry arrangement in the benchmark cases, while preserving the important neutronics characteristics of the CNPS. The results of deterministic calculations using the HELIOS/DIF3D code package were compared to MCNP4C results to show the usefulness of the numerical benchmark cases.

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.

Development of an Engineered Producet Storage Concept for the UREX+1 Combined Transuraqnic?Lanthanide Product Streams

The U.S. Department of Energy is developing next generation processing methods to recycle uranium and transuranic (TRU) isotopes from spent nuclear fuel. The objective of the 3-year project described in this report was to develop near-term options for storing TRU oxides isolated through the uranium extraction (UREX+) process. More specifically, a Zircaloy matrix cermet was developed as a storage form for transuranics with the understanding that the cermet also has the ability to serve as a inert matrix fuel form for TRU burning after intermediate storage. The goals of this research projects were: 1) to develop the processing steps required to transform the effluent TRU nitrate solutions and the spent Xircaloy cladding into a zireonium matrix cermet sotrage form; and 2) to evaluate the impact of phenomena that govern durability of the storage form, material processing, and TRU utiliztion in fast reactor fuel. This report represents a compilation of the results generated under this program. The information is presented as a brief technical narrative in the following sections with appended papers, presentations and academic theses to provide a detailed review of the projects accomplishments.