Fuel Cycle Options Research Articles

The development of future options for aqueous recycling of spent nuclear fuels

Since 2017, the UK Government has invested in research and technology development of advanced nuclear technologies including options for the future recycling of spent nuclear fuels. Whilst the UK has previously reprocessed and recycled uranium and plutonium from spent fuels, industrial scale reprocessing has finished and future fuels are being stored ready for disposal in a deep geological repository around the end of the century (i.e.transition to the open fuel cycle). Nevertheless, it is possible that an expansion of the use of nuclear energy to help meet Net Zero carbon emission targets and changing public perceptions or geopolitical situations over this period may necessitate a return to some form of closed fuel cycle. Consequently, the Advanced Fuel Cycle Programme (AFCP) has been developing advanced recycling options using both aqueous and pyro-chemical methods that would reduce wastes, costs and environmental impacts but with enhanced flexibility and barriers against nuclear proliferation risks. As well as technology development, AFCP has developed approaches for evaluating the benefits and disbenefits of future fuel cycle options under the framework of sustainability. This article provides an overview of the progress made under AFCP on the aqueous recycling of spent fuels, including the management of secondary wastes arising from the reprocessing schemes.

Comparative Evaluation of Direct Disposal and Pyro-SFR Nuclear Fuel Cycle Alternatives Using Multi Criteria Decision Making in Korea

The Korean government is currently evaluating two alternatives, direct disposal and pyroprocessing, for the disposal of spent nuclear fuel. This paper presents the ranking results of comparing and evaluating direct disposal and pyro-SFR fuel cycle alternatives using multi-criteria decision-making methods such as AHP, TOPSIS, and PROMETHEE. In considering the various evaluation criteria involved in these two alternatives, we aimed to determine the optimal choice in terms of the economic and social conditions of Korea. The evaluation criteria considered were safety, resource availability, environmental impact, economics, nuclear proliferation resistance, and public acceptance. The results show that the pyro-SFR fuel cycle alternative is more advantageous than direct disposal in the AHP and TOPSIS methods, whereas direct disposal is more advantageous in the PROMETHEE method because the ranking is reversed. TOPSIS assigns the ideal value and the most negative value among the input values to each criterion as a parameter reflecting the concept of distance between the best alternative and the worst alternative. In contrast, the PROMETHEE method first selects the preference function including the preference threshold, and calculates the preferred outflow and the preferred inflow for the detailed evaluation indicators. Therefore, differences exist in the methodologies of multi-criteria decision making. Nonetheless, the analysis results of the back-end fuel cycle option can greatly contribute to establishing a nuclear policy for the back-end nuclear fuel cycle, and these efforts will enable sustainable nuclear power generation.

Scenarios of nuclear energy for countries with different options of nuclear fuel cycle: Utilization and perspective

Countries around the world have made an international commitment to decrease the amount of greenhouse gas emissions by 2030 according to the Paris Agreement. Among several power stations options of low carbon emissions, nuclear energy has been proved the most reliable source for electricity production. According to the sustainable development scenario targets, nuclear energy should play an important role in the next years for the decarbonization of the energy system. Countries with consolidated nuclear programmes such as Canada and the UK have similar nuclear fuel technology development, but they have chosen different nuclear fuel cycle options: open fuel cycle and closed fuel cycle, respectively. Both of them have future plans to continue with their nuclear programme. Therefore, this works evaluates the nuclear energy systems for three different scenarios for high, basic and low production of nuclear energy for both countries. The results compare the resources needs, the economy of the nuclear fuel cycle options, the policy of each country and the future of the nuclear power plants to achieve the reduction of the greenhouse gases. The main findings are that reprocessing is a good option for the UK for the whole or partial new nuclear fleet and becomes viable if the price of uranium increases in the market. For Canada, it is worth continuing to increase its nuclear energy up to 1.5 times its current energy capacity. The main difference between the programmes of the two countries is the higher amount of spent fuel generated in the case of Canada due to the choice of reactor technology.

Development of experimental irradiation strategies to evaluate the robustness of TODGA and water-soluble BTP extraction systems for advanced nuclear fuel recycling

More sustainable nuclear fuel cycle options will entail the implementation of efficient separation processes of actinides from lanthanides, which currently continues being a critical milestone. One of the limiting points to ensure a safe and stable long-term operation is the resistance against radiation of the extraction systems under development. During the latest years, processes like i-SANEX and EURO-GANEX have been developed and tested to separate actinides from lanthanides, addressing the heterogeneous and the homogeneous recycling strategy respectively. Both of them are the result of the combination of an organic phase containing TODGA, and an aqueous phase containing the sulfonated 2,6-bis(5,6-di(sulphophenyl)-1,2,4-triazin-3-yl)pyridine. In this work, we report an extended stability study of the water-soluble SO3-Ph-BTP as part of the full system considering both phases, the organic and aqueous phase. A systematic set of gamma irradiation experiments were carried out in order to explore the effects of each phase and the influence of the experimental irradiation methodology chosen in the global system resistance. For that, different phase composition, oxygen content and contact between phases have been considered to achieve a better simulation of the most relevant process conditions. The current results corroborate not only that the stability of SO3-Ph-BTP has been sub-estimated, but also that the experimental irradiations test for process demonstration should involve the contact between both phases, the corresponding organic and aqueous phase.

Demand-Driven Deployment Capabilities in Cyclus, a Fuel Cycle Simulator

The present U.S. nuclear fuel cycle faces challenges that hinder the expansion of nuclear energy technology. The U.S. Department of Energy identified four nuclear fuel cycle options that make nuclear energy technology more desirable. Successfully analyzing the transitions from the current fuel cycle to these promising fuel cycles requires a nuclear fuel cycle simulator that can predictively and automatically deploy fuel cycle facilities to meet user-defined power demand. This work introduces and demonstrates the demand-driven deployment capabilities in Cyclus, an open-source nuclear fuel cycle simulator framework. User-controlled capabilities such as time-series forecasting algorithms, supply buffers, and facility preferences were introduced to give users tools to minimize power undersupply in a transition scenario simulation. The demand-driven deployment capabilities are referred to as d3ploy. We demonstrate the capability of d3ploy to predict future commodities’ supply and demand, and automatically deploy fuel cycle facilities to meet the predicted demand in four transition scenarios. Using d3ploy to set up transition scenarios saves the user simulation setup time compared to previous efforts that required a user to manually calculate and use trial and error to set up the deployment scheme for the supporting fuel cycle facilities.

Extension of lifespan of graphite in fuel blocks of high-temperature gas-cooled reactors as the resource for ensuring design values of nuclear fuel burn-up

High-temperature gas-cooled reactor (HTGR) is one of promising candidates for new generation of nuclear power reactors. This type of nuclear reactor is characterized with the following principal features: highly efficient generation of electricity (thermal efficiency of about 50%); the use of high-temperature heat in different production processes; reactor core self-protection properties; practical exclusion of reactor core meltdown in case of accidents; the possibility of implementation of various nuclear fuel cycle options; reduced radiation and thermal effects on the environment, forecasted acceptability of financial performance with respect to cost of electricity as compared with alternative energy sources. The range of output coolant temperatures in high-temperature reactors within the limits of 750–950 °C predetermines the use of graphite as the structural material of the reactor core and helium as the inert coolant. Application of graphite ensures higher heat capacity of the reactor core and its practical non-meltability. Residence time of reactor graphite depends on the critical value of fluence of damaging neutrons (neutrons with energies above 180 keV). In its turn, the value of critical neutron fluence is determined by the irradiation temperature and flux density of accompanying gamma-radiation. The values of critical fluence for graphite decrease within high-temperature region of 800–1000 °C to 1·1022 – 2·1021 cm–2, respectively. The compactness of the core results in the increase of the fracture of damaging neutrons in the total flux. These circumstances predetermine relatively low values of lifespan of graphite structures in high-temperature reactors. Design features and operational parameters of GT-MHR high-temperature gas-cooled reactor are described in the present paper. Results of neutronics calculations allowing determining the values of damaging neutron flux, nuclear fuel burnup and expired lifespan of graphite of fuel blocks were obtained. The mismatch between positions of the maxima in the dependences of fuel burnup and exhausted lifespan of graphite in fuel blocks along the core height is demonstrated. The map and methodology for re-shuffling fuel blocks of the GT-MHR reactor core were developed as the result of analysis of the calculated data for ensuring the matching between the design value of the fuel burnup and expected total graphite lifespan.

Modeling a fast spectrum molten salt reactor in a systems dynamics fuel cycles code

Molten salt reactor (MSR) concepts are being actively developed by several private companies, with some touting the fuel cycle benefits of this technology. Simulating the deployment of MSRs and their effect on fuel cycle transitions is the role of a system dynamics fuel cycles tool. However, many current fuel cycles tools have difficulty modeling MSR isotopic variations because they were designed for modeling existing reactors based on discrete fuel elements with batch fueling schemes (i.e., charge and discharge). This work presents ORION fuel cycle modeling results using a specific high power density fast spectrum MSR design. For code verification, a single-stage MSR fuel cycle was set up in ORION and results were compared with the results from the SCALE-based reactor physics model (ChemTriton). A transition scenario was set up in ORION from the current light water reactor fleet to a future fleet of MSRs to study the characteristics of an MSR that affect its performance during transition. Several transition scenarios were set with holds and delays in place that are representative of a sodium fast reactor as modeled in Evaluation Group 23 for the Fuel Cycle Options Campaign Evaluation and Screening Study. Several MSR characteristics lend to its high performance and ease in transition from the current fuel cycle: fissile material is only loaded at initial startup; all fissile material is available as it is bred due to online fuel processing; and fissile material recirculating through the core is briefly held in tanks for processing and has very minimal out-of-core time (on the order of a minutes to a few days). These characteristics are in contrast to a solid-fueled reactor, which requires discharge, cooling, and fabrication of fuel all of which could take a few months to a few years and repeated loading of fresh fuel assemblies. A solid-fueled reactor only discharges a fraction of a core every year.

The economic competitiveness of promising nuclear energy system: A closer look at the input uncertainties in LCOE analysis

In this study, we aimed to provide important information about the potential economic benefits and risks of nuclear electricity generation associated with existing and prevailing nuclear technologies and to examine the economic effects of nuclear fuel cycle strategies in Korea. An economic analysis model that evaluates the overall life-cycle costs of nuclear energy systems coupled with multiple fuel cycle options was specially developed by using the levelized cost of electricity (LCOE) as the fundamental methodology. This model is capable of identifying a range of techno-economic uncertainties underlying each individual nuclear energy system taking into account the state of the art in fuel cycle technologies. It can also quantify and incorporate the resulting impacts into a system-wide LCOE distribution for each fuel cycle option based on Monte Carlo probabilistic simulation. We analyzed and discussed examples of the economic performance of 13 promising candidates for nuclear energy systems integrated with extensive fuel cycle technologies (including one direct disposal and 12 specific reprocessing and recycling fuel cycle options). We also conducted a sensitivity analysis to investigate the major sensitivity factors of the system component cost in each fuel cycle option and their impacts on individual economic performances. Furthermore, a closer look at the techno-economic uncertainties of advanced fuel cycle technologies in a break-even analysis offers evidence of the potential economic feasibility and cost-reduction opportunities in the reprocessing and recycling options relative to the direct disposal of spent nuclear fuel.

Applications for Thorium in Multistage Fuel Cycles with Heavy Water Reactors

Certain characteristics of heavy water reactors (HWRs), such as a more flexible neutron economy compared to light water (due to reduced absorptions in hydrogen), online refueling capability, and having a thermal neutron spectrum, make them potentially attractive for use with a thorium fuel cycle. Three options that combine HWRs with thorium-based fuels are considered in this paper: a Near-Term option with minimal advanced technology requirements, an Actinide Management option that incorporates the recycle of minor actinides (MAs), and a Thorium-Only option that uses two reactor stages to breed and consume 233U, respectively. Simplified, steady-state simulations and corresponding material flow analyses are used to elucidate the properties of these fuel cycle options. The Near-Term option begins with a low-enriched uranium oxide pressurized water reactor (PWR) that discharges spent nuclear fuel, from which uranium and plutonium are recovered to fabricate the driver fuel for an HWR that uses thorium oxide as a blanket fuel. This option uses 28% less natural uranium (NU) and sends 33% less plutonium to disposal than the conventional once-through uranium fuel cycle on an energy-normalized basis. The Actinide Management option also uses spent nuclear fuel from a PWR using enriched uranium oxide fuel (both a low- and high-enrichment variant are considered), but the uranium is recycled for reuse in the PWR while the plutonium and MAs are recycled and used in conjunction with thorium in an HWR with full recycle. Both enrichment variants of this option achieve a more than 95% reduction in transuranic actinide disposal rates compared to the once-through option and a more than 60% reduction compared to closed transuranic recycle in a uranium-plutonium–fueled sodium fast reactor. The Thorium-Only option breeds a surplus of 233U in a thorium-based HWR to supply fissile material to a high-temperature gas-cooled reactor, both of which recycle uranium and thorium. This option requires no NU and produces few transuranic actinides at steady state, although it would require a greater technology maturation effort than the other options studied. Collectively, the options considered in this study are intended to illustrate the range of operational missions that could be supported by fleets that integrate thorium and HWRs.

Integrated system evaluation of nuclear fuel cycle options in China combined with an analytical MCDM framework

Our work is a first attempt to evaluate comprehensively the sustainability of future nuclear fuel cycle options in China based on dynamic system modeling combined with a Multi-Criteria Decision Making (MCDM) analysis. Considering six essential indicators, we developed an MCDM framework to evaluate and rank China-specific sustainability in line with the quantitative performance of four candidate options of fuel cycle transition scenarios from the existing to advanced nuclear energy systems through 2100. Four transition scenarios identified are 1) the direct disposal of PWR (Pressurized Water Reactor) spent fuel, 2) the mono-recycling of PWR spent fuel through PWR-MOX (Mixed Oxides), 3) the PWR-MOX followed by Fast Reactor (FR), and 4) the full recycling of PWR spent fuel using FR. We also further analyzed the roles of the different representatives of decision makers and their impacts on the overall ranking by applying AHP, TOPSIS, and PROMETHEE II methods from a weighting perspective. The results showed that Scenario 1 is the least sustainable and attractive fuel cycle option. Instead, Scenario 4 substantially satisfies the requirements underlying the current definition of nuclear sustainability, while remaining in line with China's closed nuclear fuel cycle policy. This strategic MCDM framework also provides flexibility to cope with the evolution and inherent uncertainties of technological development, which is conducive to identifying the trade-offs of future nuclear energy systems considered at a national level.

National Options for a Sustainable Nuclear Energy System: MCDM Evaluation Using an Improved Integrated Weighting Approach

While the prospects look bright for nuclear energy development in China, no consensus about an optimum transitional path towards sustainability of the nuclear fuel cycle has been achieved. Herein, we present a preliminary study of decision making for China’s future nuclear energy systems, combined with a dynamic analysis model. In terms of sustainability assessment based on environmental, economic, and social considerations, we compared and ranked the four candidate options of nuclear fuel cycles combined with an integrated evaluation analysis using the Multi-Criteria Decision Making (MCDM) method. An improved integrated weighting method was first applied in the nuclear fuel cycle evaluation study. This method synthesizes diverse subjective/objective weighting methods to evaluate conflicting criteria among the competing decision makers at different levels of expertise and experience. The results suggest that the fuel cycle option of direct recycling of spent fuel through fast reactors is the most competitive candidate, while the fuel cycle option of direct disposal of all spent fuel without recycling is the least attractive for China, from a sustainability perspective. In summary, this study provided a well-informed decision-making tool to support the development of national nuclear energy strategies.

An Advanced Sodium-Cooled Fast Reactor Core Concept Using Uranium-Free Metallic Fuels for Maximizing TRU Burning Rate

In this paper, we designed and analyzed advanced sodium-cooled fast reactor cores using uranium-free metallic fuels for maximizing burning rate of transuranics (TRU) nuclides from PWR spent fuels. It is well known that the removal of fertile nuclides such as 238U from fuels in liquid metal cooled fast reactor leads to the degradation of important safety parameters such as the Doppler coefficient, coolant void worth, and delayed neutron fraction. To resolve the degradation of the Doppler coefficient, we considered adding resonant nuclides to the uranium-free metallic fuels. The analysis results showed that the cores using uranium-free fuels loaded with tungsten instead of uranium have a significantly lower burnup reactivity swing and more negative Doppler coefficients than the core using uranium-free fuels without resonant nuclides. In addition, we considered the use of axially central B4C absorber region and moderator rods to further improve safety parameters such as sodium void worth, burnup reactivity swing, and the Doppler coefficient. The results of the analysis showed that the final design core can consume ~353 kg per cycle and satisfies self-controllability under unprotected accidents. The fuel cycle analysis showed that the PWR–SFR coupling fuel cycle option drastically reduces the amount of waste going to repository and the SFR burner can consume the amount of TRUs discharged from 3.72 PWRs generating the same electricity.

Comparison and Screening of Nuclear Fuel Cycle Options in View of Sustainable Performance and Waste Management

Is it true that a nuclear technology approach to generate electric energy offers a clean, safe, reliable and affordable, i.e., sustainable option? In principle yes, however a technology impact on the environment strongly depends on the actual implementation bearing residual risks due to technical failures, human factors, or natural catastrophes. A full response is thus difficult and can be given first when the wicked multi-disciplinary issues get well formulated and “resolved”. These problems are lying at the interface between: the necessary R&D effort, the industrial deployment and the technology impact in view of the environmental sustainability including the management of produced hazardous waste. As such, this problem is clearly of multi-dimensional nature. This enormous complexity indicates that just a description of the problem might cause a dilemma. The paper proposes a novel holistic approach applying Multi-Criteria Decision Analysis to assess the potential of nuclear energy systems with respect to a sustainable performance. It shows how to establish a multi-level criteria structure tree and examines the trading-off techniques for scoring and ranking of options. The presented framework allows multi-criteria and multi-group treatment. The methodology can be applied to support any pre-decisional process launched in a country to find the best nuclear and/or non-nuclear option according to national preferences and priorities. The approach addresses major aspects of the environmental footprint of nuclear energy systems. As a case study, advanced nuclear fuel cycles are analyzed, which were previously investigated by the Nuclear Energy Agency (NEA/OECD) expert group WASTEMAN. Sustainability facets of waste management, resource utilization and economics are in focus.

Additional Capability Being Developed from UNF-ST&DARDS Unified Database: System Analysis and Fuel Cycle Option Analysis

The Used Nuclear Fuel Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS) is being developed for the U.S. Department of Energy’s Office of Nuclear Energy by the national laboratories. An important part of UNF-ST&DARDS is the Unified Database (UDB), which contains information that can support a variety of activities including fuel storage, fuel transportation, and disposal-related system analysis. Currently, the main application of the UDB is to support evaluation of the characteristics of discharged spent nuclear fuel (SNF) from the U.S. commercial reactors. However, because of the extensive amount of data that has been collected and analyzed for UNF-ST&DARDS, there are many more applications that can utilize the UDB including system analysis with the Next-Generation System Analysis Model (NGSAM) and fuel cycle analysis with fuel cycle simulation codes such as ORION. Going forward, NGSAM and fuel cycle transition analysis with ORION integrate UDB data wherever possible in the UDB’s development plan. These advances in NGSAM and fuel cycle analysis can be used in conjunction with the UDB to help answer more complex questions about the optimization, utilization, storage, and eventual disposal of SNF.

Break-Even Point Analysis of Sodium-Cooled Fast Reactor Capital Investment Cost Comparing the Direct Disposal Option and Pyro-Sodium-Cooled Fast Reactor Nuclear Fuel Cycle Option in Korea

The purpose of this paper is to recommend a break-even point for the capital investment cost for a Sodium-cooled Fast Reactor (SFR) when choosing between a Pyro-SFR nuclear fuel cycle (recycling option via Pyro-processing) and a direct disposal option. This is because the selection of an alternative cannot be justified without a guarantee of economic feasibility. The calculation of a break-even point is necessary because SFR capital investment cost makes up the largest share of the cost for electricity generation. In other words, the cost of capital investment is an important cost driver, and the one that exerts the greatest effect on Pyro-SFR nuclear fuel cycle economics. In the end, the break-even point of the SFR capital investment cost between the Pyro-SFR nuclear fuel cycle and the direct disposal was calculated to be 4284 US$/kWe. In other words, it is possible to claim that the potential for the economic viability of the Pyro-SFR nuclear fuel cycle is greater (compared to investing in direct disposal) when the SFR capital investment cost is 4284 US$/kWe or less. In addition, Pyro-SFR technology will enable sustainable nuclear power generation.

Fuel Sustainability And Actinide Production Of Doping Minor Actinide In Water-Cooled Thorium Reactor

Fuel sustainability of nuclear energy is coming from an optimum fuel utilization of the reactor and fuel breeding program. Fuel cycle option becomes more important for fuel cycle utilization as well as fuel sustainability capability of the reactor. One of the important issues for recycle fuel option is nuclear proliferation resistance issue due to production plutonium. To reduce the proliferation resistance level, some barriers were used such as matrial barrier of nuclear fuel based on isotopic composition of even mass number of plutonium isotope. Analysis on nuclear fuel sustainability and actinide production composition based on water-cooled thorium reactor system has been done and all actinide composition are recycled into the reactor as a basic fuel cycle scheme. Some important parameters are evaluated such as doping composition of minor actinide (MA) and volume ratio of moderator to fuel (MFR). Some feasible parameters of breeding gains have been obtained by additional MA doping and some less moderation to fuel ratios (MFR). The system shows that plutonium and MA are obtained low compositions and it obtains some higher productions of even mass plutonium, which is mainly Pu-238 composition, as a control material to protect plutonium to be used as explosive devices.

Illuminating Fuel Cycle Decision Drivers Using a Decision Analysis Framework

Nuclear fuel cycle studies have provided a wealth of information on the potential impacts of advanced recycling systems. Deciding on fuel cycle implementation pathways, however, requires synthesizing volumes of data and navigating trade-offs between fuel cycle options. This research presents a framework intended to aid fuel cycle decision makers by focusing on the cost reduction/waste mitigation trade-off as a lens for choosing a near-term strategy. The framework consists of a fuel cycle simulation coupled to a decision tree model that maps evolution scenarios. System scenarios are constructed by considering the technological options for fuel cycle evolution and key uncertainties expected to affect the desirability of those options. For this study, the once-through fuel cycle is compared to a self-sustaining fast reactor (FR) fuel cycle. Scenarios are compared using a value function that incorporates cost and waste metrics. The results indicate that uranium costs and the attainable level of reprocessing efficiency may not significantly impact the suite of desirable decisions. On the other hand, the pattern and timing of nuclear builds as well as the extent to which FRs provide true waste mitigation more significantly impact the attractiveness of closing the fuel cycle.

Pressurized Heavy Water Reactor Technology: Its Relevance Today

The pressurized heavy water reactor (PHWR) technology was conceived in Canada and has moved to several nations for commercial production of electricity. Currently, 49 power reactors operate with PHWR technology producing nearly 25 GWe. The technology is flexible for adopting different fuel cycle options which include natural uranium, different mixed oxide (MOX) fuel, and thorium. The technology has made substantial improvement in materials, construction, and safety since its inception. PHWRs have demonstrated excellent performance historically. Their safety statistics are excellent. Indian PHWRs also have shown economic competitiveness even in small sizes, thus providing an ideal design for new entrants. While the technology features of PHWRs are available even in textbooks, the objective of this paper is to highlight the historical development and salient features, and innovations for further improvement in operation, safety and economics. Thus, this paper shall serve as a curtain raiser for the special issue “Pressurized Heavy Water Reactors (PHWRs) Safety: Post Fukushima.”

Core Design Studies of a Fast Reactor and Accelerator-Driven System for Two-Stage Fast Spectrum Fuel Cycle Option

This paper presents the core design studies of a sodium-cooled fast reactor (SFR) and a sodium-cooled accelerator-driven system (ADS) for a two-stage fast-spectrum fuel cycle to enhance uranium resource utilization and reduce nuclear waste generation. The first-stage SFR starts with low-enriched uranium (LEU) fuel and operates with the recovered uranium and plutonium from the discharged fuels and natural uranium at equilibrium. The recovered minor actinides (MAs) are sent to the second-stage ADS, where they are burned in an inert matrix fuel form. Reference core designs were developed for a 1000-MW(thermal) LEU-fueled breakeven fast reactor (LEUBFR) and an 840-MW(thermal) MA-fueled ADS blanket. The SFR starts with uranium fuel with a 235U enrichment of 13.6% and reaches a fuel-breakeven core after 14 cycles with an 18-month cycle length. At the equilibrium state, one ADS supports 37 fast reactors. Using the performance parameters of SFR and ADS, the proposed two-stage fuel cycle was evaluated. The results of the equilibrium cycle analysis showed that the two-stage fuel cycle option could achieve a high reduction in waste generation because of the continuous recycling of the plutonium and MAs. In addition, the mass flow data showed that this two-stage fuel cycle option increases the efficiency of natural uranium utilization and reduces the nuclear waste generation compared to the conventional two-stage fuel cycle options based on thermal and fast-spectrum systems.

Impact of thermal spectrum small modular reactors on performance of once-through nuclear fuel cycles with low-enriched uranium

Small modular reactors (SMRs) may offer potential benefits relative to large light water reactors, such as enhanced flexibility in deployment and operation. However, it is vital to understand the holistic impact of SMRs on nuclear fuel cycle performance. The focus of this paper is the fuel cycle impacts of light water SMRs in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary neutronics and fuel cycle analyses conducted in support of the US Department of Energy, Office of Nuclear Energy, Fuel Cycle Options Campaign. The hypothetical light water SMR example case considered in these preliminary scoping studies is a “cartridge type” one-batch core with slightly less than 5.0% enrichment.The high-level issues identified and preliminary scoping calculations in this paper are intended to inform decision makers regarding potential fuel cycle impacts of one-batch thermal-spectrum SMRs. In particular, this paper highlights the impact of increased neutron leakage and a reduced number of batches on the achievable burnup of the reactor. Fuel cycle performance metrics for the simplified example SMR analyzed herein are compared with those for a conventional three-batch light water reactor (LWR) in the following areas: nuclear waste management, environmental impact, and resource utilization. The metrics performance for such an SMR is degraded for the mass of spent nuclear fuel and high-level waste disposed of per energy generated, mass of depleted uranium disposed of per energy generated, land use per energy generated, and carbon emissions per energy generated.Finally, it is noted that the features of some SMR designs impact three main aspects of fuel cycle performance: (1) small cores, which mean high leakage (there is a radial and an axial component); (2) a heterogeneous core and extensive use of control rods and burnable poisons; and (3) single-batch cores. But not all SMR designs have all of these traits. The approach used in this study is an example bounding case, and not all SMRs may be impacted to the same extent.