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Peggs S.,Brookhaven National Laboratory | Horak W.,Brookhaven National Laboratory | Roser T.,Brookhaven National Laboratory | Parks G.,Cambridge U. | And 14 more authors.
IPAC 2012 - International Particle Accelerator Conference 2012 | Year: 2012

The potential for thorium as an alternative or supplement to uranium in fission power generation has long been recognised, and several reactors, of various types, have already operated using thorium-based fuels. Accelerator Driven Subcritical (ADS) systems have benefits and drawbacks when compared to conventional critical thorium reactors, for both solid and molten salt fuels. None of the four options - liquid or solid, with or without an accelerator - can yet be rated as better or worse than the other three, given today's knowledge. We outline the research that will be necessary to lead to an informed choice. Copyright © 2012 by IEEE.

Insulander Bjork K.,Thor Energy | Insulander Bjork K.,Chalmers University of Technology | Mittag S.,Helmholtz Center Dresden | Nabbi R.,Jülich Research Center | And 3 more authors.
Progress in Nuclear Energy | Year: 2013

A benchmark exercise for thorium-plutonium fuel, based on experimental data, has been carried out. A thorium-plutonium oxide fuel rodlet was irradiated in a PWR for four consecutive cycles, to a burnup of about 37 MWd/kgHM. During the irradiation, the rodlet was inserted into a guide tube of a standard MOX fuel assembly. After the irradiation, the rod was subjected to several PIE measurements, including radiochemical analysis. Element concentrations and radial distributions in the rodlet, multiplication factors and distributions within the carrier assembly of burnup and power were calculated. Four participants in the study simulated the irradiation of the MOX fuel assemblies including the thorium-plutonium rodlet using their respective code systems; MCBurn, HELIOS, CASMO-5 and ECCO/ERANOS combined with TRAIN. The results of the simulations and the measured results of the radiochemical analysis were compared and found to be in fairly good agreement when the calculated results were calibrated to give the same burnup of the thorium-plutonium rodlet as that experimentally measured. Average concentrations of several minor actinides and fission products were well reproduced by all codes, to the extent that can be expected based on known uncertainties in the experimental setup and the cross section libraries. Calculated results which could not be confirmed by experimental measurement were compared and only two significant anomalies were found, which can probably be addressed by limited modifications of the codes. © 2013 Elsevier Ltd. All rights reserved.

Chen C.-F.,Los Alamos National Laboratory | Kelly J.,THOR ENERGY | Asphjell O.,THOR ENERGY | Papin P.A.,Los Alamos National Laboratory | And 4 more authors.
Journal of the American Ceramic Society | Year: 2014

The paper describes an effective procedure for mixing and conditioning ThO2 and CeO2 powders so they are suited for pressing and sintering into high-density (Th0.9,Ce0.1)O2 ceramic pellets - this material being a "pilot" for (Th,Pu)O2 fuels. Wet ball milling with an organic dispersant aided the powder dispersing process by reducing the agglomeration of very small oxide particles. Homogeneous elemental distributions were seen within the calcined powder mixture. Heat treatments were applied to the calcined, mixed ThO2/CeO2 mix to study phase and surface area transformations. Solid solution formation commences at around 1300°C and goes to completion at a temperature of 1500°C. We also report the effect of a granulation strategy that can be applied to the production of high quality, mixed ThO2 nuclear fuel ceramics. Sized granules of blended ThO2/CeO2 powder were produced from precompacted disks of this material that were subsequently heat treated. This had a positive effect on die filling and compaction into green pellets, as well as on final sintered (Th,Ce)O2 pellet density. The microstructure of the sintered (Th,Ce)O2 ceramic was characterized using SEM-based electron back-scatter diffraction from which a uniform density and grain size were readily apparent. XRD results showed that a single phase Th0.9Ce0.1O2, fuel ceramic had been produced. Its density was ∼94% TD. © 2014 The American Ceramic Society.

Insulander Bjork K.,Thor Energy | Insulander Bjork K.,Chalmers University of Technology | Drera S.S.,Thor Energy | Kelly J.F.,Thor Energy | And 10 more authors.
Annals of Nuclear Energy | Year: 2015

Thorium based fuels are being tested in the Halden Research Reactor in Norway with the aim of producing the data necessary for licensing of these fuels in today's light water reactors. The fuel types currently under irradiation are thorium oxide fuel with plutonium as the fissile component, and uranium fuel with thorium as an additive for enhancement of thermo-mechanical and neutronic fuel properties. Fuel temperatures, rod pressures and dimensional changes are monitored on-line for quantification of thermo-mechanical behavior and fission gas release. Preliminary irradiation results show benefits in terms of lower fuel temperatures, mainly caused by improved thermal conductivity of the thorium fuels. In parallel with the irradiation, a manufacturing procedure for thorium-plutonium mixed oxide fuel is developed with the aim to manufacture industrially relevant high-quality fuel pellets for the next phase of the irradiation campaign. © 2014 Elsevier Ltd. All rights reserved.

Insulander Bjork K.,Thor Energy | Insulander Bjork K.,Chalmers University of Technology | Kekkonen L.,Fortum | Kekkonen L.,Institute for Energy Technology of Norway
Journal of Nuclear Materials | Year: 2015

Thorium-plutonium Mixed OXide (Th-MOX) fuel is considered for use in light water reactors fuel due to some inherent benefits over conventional fuel types in terms of neutronic properties. The good material properties of ThO2 also suggest benefits in terms of thermal-mechanical fuel performance, but the use of Th-MOX fuel for commercial power production demands that its thermal-mechanical behavior can be accurately predicted using a well validated fuel performance code. Given the scant operational experience with Th-MOX fuel, no such code is available today. This article describes the first phase of the development of such a code, based on the well-established code FRAPCON 3.4, and in particular the correlations reviewed and chosen for the fuel material properties. The results of fuel temperature calculations with the code in its current state of development are shown and compared with data from a Th-MOX test irradiation campaign which is underway in the Halden research reactor. The results are good for fresh fuel, whereas experimental complications make it difficult to judge the adequacy of the code for simulations of irradiated fuel. © 2015 Elsevier B.V.

Insulander Bjork K.,Thor Energy | Insulander Bjork K.,Chalmers University of Technology | Fhager V.,Thor Energy | Demazire C.,Chalmers University of Technology
Progress in Nuclear Energy | Year: 2011

With the aim of investigating the technical feasibility of fuelling a conventional BWR (Boiling Water Reactor) with thorium-based fuel, computer simulations were carried out in a 2D infinite lattice model using CASMO-5. Four different fissile components were each homogenously combined with thorium to form mixed oxide pellets: Uranium enriched to 20% U-235 (LEU), plutonium recovered from spent LWR fuel (RGPu), pure U-233 and a mixture of RGPu and uranium recovered from spent thorium-based fuel. Based on these fuel types, four BWR nuclear fuel assembly designs were formed, using a conventional assembly geometry (GE14-N). The fissile content was chosen to give a total energy release equivalent to that of a UOX fuel bundle reaching a discharge burnup of about 55 MWd/kgHM. The radial distribution of fissile material was optimized to achieve low bundle internal radial power peaking. Reactor physical parameters were computed, and the results were compared to those of reference LEU and MOX bundle designs. It was concluded that a viable thorium-based BWR nuclear fuel assembly design, based on any of the fissile components, can be achieved. Neutronic parameters that are essential for reactor safety, like reactivity coefficients and control rod worths, are in most cases similar to those of LEU and MOX fuel. This is also true for the decay heat produced in irradiated fuel. However when Th is mixed with U-233, the void coefficient (calculated in 2D) can be positive under some conditions. It was concluded that it is very difficult to make savings of natural uranium by mixing LEU (20% U-235) homogenously with thorium and that mixing RGPu with thorium leads to more efficient consumption of Pu compared to MOX fuel. © 2011 Elsevier Ltd. All rights reserved.

Drera S.S.,Thor Energy | Insulander Bjork K.,Thor Energy | Insulander Bjork K.,Chalmers University of Technology | Kelly J.F.,Thor Energy | Kelly J.F.,Institute for Energy Technology of Norway
Progress in Nuclear Energy | Year: 2014

An evolutionary, rather than a revolutionary approach to thorium fueled reactors is proposed with an introduction of thorium as the fertile component in mixed oxide fuel for conventional light water reactors. The utility of thorium as a component in today's light water reactors offers improved accident tolerance due to the superior material properties of thorium fuels over conventional uranium fuels. A few notable advantages include better thermal conductivity, higher melting point, higher oxide stability and superior spent fuel characteristics. Consequently, Thor Energy along with a consortium of industrial partners has established a fuel production and irradiation program aimed to license thorium fuels for use in today's light water reactors. Due to the morphology and physical properties of calcined thorium oxide powder, pellet fabrication used for standard uranium oxide fuels must be altered slightly for thorium bearing fuels to yield a product with acceptable theoretical densities, microstructure, and material integrity. At beginning of irradiation life fuel temperature data demonstrates improved fuel characteristics over standard uranium oxide fuel. Fuel centerline operating temperatures are 30-40 K less with a thorium mixed oxide fuel as compared to standard uranium fuel. © 2014 Elsevier Ltd. All rights reserved.

Insulander Bjork K.,Thor Energy | Insulander Bjork K.,Chalmers University of Technology
Progress in Nuclear Energy | Year: 2013

The objective of this study is to develop an optimized BWR fuel assembly design for thorium-plutonium fuel. In this work, the optimization goal is to maximize the amount of energy that can be extracted from a certain amount of plutonium, while maintaining acceptable values of the neutronic safety parameters such as reactivity coefficients, shutdown margins and power distribution. The factors having the most significant influence on the neutronic properties are the hydrogen-to-heavy-metal ratio, the distribution of the moderator within the fuel assembly, the initial plutonium fraction in the fuel and the radial distribution of the plutonium in the fuel assembly. The study begins with an investigation of how these factors affect the plutonium requirements and the safety parameters. The gathered knowledge is then used to develop and evaluate a fuel assembly design. The main characteristics of this fuel design are improved Pu efficiency, very high fractional Pu burning and neutronic safety parameters compliant with current demands on UOX fuel. © 2013 Elsevier Ltd. All rights reserved.

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