Andrew Worrall

The Reemergence of the Thorium Fuel Cycle: A Special Issue of Nuclear Technology

Thorium was intensively studied from the 1960s to the 1980s in the United States and internationally as a potential basis for future nuclear fuel cycles. After demonstration of feasible thorium fuel cycle concepts, the United States decided instead to pursue liquid metal, fast breeder reactors using uranium and plutonium. Worldwide interest in thorium fuel cycle development continued at a reduced level, with India having invested the most resources. Recently, the thorium fuel cycle has been the subject of renewed interest, partly due to speculated growth in nuclear energy worldwide (hence putting potential strain on uranium reserves) and partly due to the pursuit of advanced reactor concepts designed to enhance safety and economics—which also have the potential to use thorium to further improve fuel cycle performance. This renewed interest often addresses new possibilities for using thorium in the modern era; however, it can be difficult to discern between actual, unique characteristics of the new thorium concepts and misconceptions disseminated by advocates and detractors. Therefore, right now is a good time to discuss experience with the thorium fuel cycle to date, provide an even-handed description of its inherent attributes, and identify some of the data gaps that have yet to be resolved. The renewed interest in thorium is supported, in part, by a resurgence of national and industrial programs focused on thorium-based nuclear fuel cycles. India has described plans for a three-stage nuclear energy strategy that integrates thorium-based fuels: Stage 1 involves natural uranium–burning pressurized heavy water reactors, to produce plutonium and stockpile it for future use; Stage 2 uses this stockpiled plutonium in fast breeder reactors with thorium blankets to produce 233U (and additional plutonium) and recycles plutonium back to the fast reactor; finally, Stage 3 uses recovered 233U (from Stage 2) in advanced heavy water–moderated, light water–cooled reactors. Currently, Stage 1 is operational, Stage 2 is in advanced testing, and Stage 3 is in advanced design.1 China is planning to build two experimental molten salt reactors: The first, which is to commence operation in 2017, will use low-enrichment uranium, spherical pebble fuel, and LiF-BeF2 molten salt as the coolant; this is intended to set the stage for a second molten salt reactor (scheduled to commence operation in 2020), which will use thorium-based fluid fuel and include fuel salt processing.2 China is also considering the use of Canadian-designed fuels in pressurized heavy water reactors, which have the potential to incorporate thorium.3 Also, Thor Energy is conducting experiments focused on demonstrating fuel manufacturing, materials, and nuclear performance of PuO2-ThO2 and UO2-ThO2 ceramic fuels. Test pins composed of thorium-uranium and thoriumplutonium oxide mixtures are currently being irradiated in the Halden test reactor, and additional testing of thoriumbased oxide fuel pins is planned.4 Differences in the major technical features of the thorium-233U fuel cycle and the currently implemented fuel cycle, based on 235U and plutonium, present implications for facility design and operation, along with waste disposal. To set the context for this special issue of Nuclear Technology, we will discuss some of the major technical features and characteristics that help to frame the dialogue regarding the use of thorium (a more comprehensive summary is found in Ref. 5). Thorium is fertile but does not contain natural fissile isotopes, so external fissile material is required to produce 233U at the onset of fuel cycle implementation. Thorium-based fuels offer the potential for higher conversion ratios than uranium-based fuels in thermal reactors, since (a) 233U has a relatively low neutron capture (nonfission) cross section compared NUCLEAR TECHNOLOGY · VOLUME 194 · iii–iv · MAY 2016

Safeguards Considerations for Thorium Fuel Cycles

Abstract By around 2025, thorium-based fuel cycles are likely to be deployed internationally. States such as China and India are pursuing research, development, and deployment pathways toward a number of commercial-scale thorium fuel cycles, and they are already building test reactors and the associated fuel cycle infrastructure. In the future, the potential exists for these emerging programs to sell, export, and deploy thorium fuel cycle technology in other states. Without technically adequate international safeguards protocols and measures in place, any future potential clandestine misuse of these fuel cycles could go undetected, compromising the deterrent value of these protocols and measures. The development of safeguards approaches for thorium-based fuel cycles is therefore a matter of some urgency. Yet, the focus of the international safeguards community remains mainly on safeguarding conventional 235U- and 239Pu-based fuel cycles while the safeguards challenges of thorium-uranium fuel cycles remain largely uninvestigated. This raises the following question: Is the International Atomic Energy Agency and international safeguards system ready for thorium fuel cycles? Furthermore, is the safeguards technology of today sufficiently mature to meet the verification challenges posed by thorium-based fuel cycles? In defining these and other related research questions, the objectives of this paper are to identify key safeguards considerations for thorium-based fuel cycles and to call for an early dialogue between the international safeguards and the nuclear fuel cycle communities to prepare for the potential safeguards challenges associated with these fuel cycles. In this paper, it is concluded that directed research and development programs are required to meet the identified safeguards challenges and to take timely action in preparation for the international deployment of thorium fuel cycles.

Two-Dimensional Neutronic and Fuel Cycle Analysis of the Transatomic Power Molten Salt Reactor

This status report presents the results from the first phase of the collaboration between Transatomic Power Corporation (TAP) and Oak Ridge National Laboratory (ORNL) to provide neutronic and fuel cycle analysis of the TAP core design through the Department of Energy Gateway for Accelerated Innovation in Nuclear, Nuclear Energy Voucher program. The TAP design is a molten salt reactor using movable moderator rods to shift the neutron spectrum in the core from mostly epithermal at beginning of life to thermal at end of life. Additional developments in the ChemTriton modeling and simulation tool provide the critical moderator-to-fuel ratio searches and time-dependent parameters necessary to simulate the continuously changing physics in this complex system. Results from simulations with these tools show agreement with TAP-calculated performance metrics for core lifetime, discharge burnup, and salt volume fraction, verifying the viability of reducing actinide waste production with this design. Additional analyses of time step sizes, mass feed rates and enrichments, and isotopic removals provide additional information to make informed design decisions. This work further demonstrates capabilities of ORNL modeling and simulation tools for analysis of molten salt reactor designs and strongly positions this effort for the upcoming three-dimensional core analysis.

Analysis of key safety metrics of thorium utilization in LWRs

Abstract Thorium has great potential to stretch nuclear fuel reserves because of its natural abundance and because it is possible to breed the 232Th isotope into a fissile fuel (233U). Various scenarios exist for utilization of thorium in the nuclear fuel cycle, including use in different nuclear reactor types (e.g., light water, high-temperature gas-cooled, fast spectrum sodium, and molten salt reactors), along with use in advanced accelerator-driven systems and even in fission-fusion hybrid systems. The most likely near-term application of thorium in the United States is in currently operating light water reactors (LWRs). This use is primarily based on concepts that mix thorium with uranium (UO2 + ThO2) or that add fertile thorium (ThO2) fuel pins to typical LWR fuel assemblies. Utilization of mixed fuel assemblies (PuO2 + ThO2) is also possible. The addition of thorium to currently operating LWRs would result in a number of different phenomenological impacts to the nuclear fuel. Thorium and its irradiation products have different nuclear characteristics from those of uranium and its irradiation products. ThO2, alone or mixed with UO2 fuel, leads to different chemical and physical properties of the fuel. These key reactor safety–related issues have been studied at Oak Ridge National Laboratory and documented in “Safety and Regulatory Issues of the Thorium Fuel Cycle” (NUREG/CR-7176, U.S. Nuclear Regulatory Commission, 2014). Various reactor analyses were performed using the SCALE code system for comparison of key performance parameters of both ThO2 + UO2 and ThO2 + PuO2 against those of UO2 and typical UO2 + PuO2 mixed oxide fuels, including reactivity coefficients and power sharing between surrounding UO2 assemblies and the assembly of interest. The decay heat and radiological source terms for spent fuel after its discharge from the reactor are also presented. Based on this evaluation, potential impacts on safety requirements and identification of knowledge gaps that require additional analysis or research to develop a technical basis for the licensing of thorium fuel are identified.

Effect of highly enriched/highly burnt UO2 fuels on fuel cycle costs, radiotoxicity, and nuclear design parameters

Abstract A study of high-burnup pressurized water reactor (PWR) fuel management schemes extending to 100 GWd/tonne is presented. The Studsvik Scandpower code suite was used to model a Westinghouse three-loop PWR core, and realistic loading patterns were derived. The loading patterns were optimized for minimum power peaking and maximum cycle length using engineering judgment and automated binary shuffles. Gadolinia was found to control power peaking to within current FΔH design limits up to 70 GWd/tonne, with only a slight deterioration thereafter. The moderator temperature coefficient, boron coefficient, and control rod worth were calculated and shown to fall within existing design limits. An economic analysis was carried out to determine the most economic discharge burnup based on fuel cycle costs only. It was found that the lowest fuel cycle costs were obtained with average discharge burnups between 70 to 75 GWd/tonne (initial enrichments between 6 to 7 wt%). The energy generated per tonne of uranium ore used was calculated and shown to peak between 40 to 60 GWd/tonne. Also, the radiotoxicity generated per GW·yr(electric) was calculated for each fuel management scheme and found to reduce considerably with burnup between 100 and 100 000 yr.

Applications for Thorium in Multistage Fuel Cycles with Heavy Water Reactors

Abstract 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.

Report on Reactor Physics Assessment of Candidate Accident Tolerant Fuel Cladding Materials in LWRs

This work focuses on ATF concepts being researched at Oak Ridge National Laboratory (ORNL), expanding on previous studies of using alternate cladding materials in pressurized water reactors (PWRs). The neutronic performance of two leading alternate cladding materials were assessed in boiling water reactors (BWRs): iron-chromium-aluminum (FeCrAl) cladding, and silicon carbide (SiC)-based composite cladding. This report fulfills ORNL Milestone M3FT-15OR0202332 within the fiscal year 2015 (FY15)

Technology Implimentation Plan - ATF FeCrAl Cladding for LWR Application

Technology implimentation plan for FeCrAl development under the FCRD Advanced Fuel program. The document describes the activities required to get FeCrAl clad ready for LTR testing