Effect of the thorium oxide content on the leaching of a mixed thorium-uranium oxide fuel

Leaching of uranium from uranium oxide fuel in contact with water can be a radiation hazard problem in the case of fuel cladding failure, either during nuclear reactor operation or in an interim storage, as well as in a final repository. One way to mitigate this is to reduce the solubility of the fuel matrix by the mixing uranium oxide with a compound which is less soluble but otherwise of similar properties. In this paper, the effect of thorium oxide content on the leaching of the uranium oxide matrix is investigated. The method was to study the leaching of the uranium oxide fuel matrix as a function of a varying content of thorium oxide, using materials manufactured by powder co-milling. It was found that the substitution of more than 25% UO2 with ThO2 reduces the matrix leaching by more than one order of magnitude in most of the different leaching solutions investigated. The substitution of 7% UO2 with ThO2 results in a reduction of matrix leaching by 10-90%, depending on the concentration of borate and dissolved oxygen in the leaching solution

Experimental determination of concentration factors of Mn, Zn and I in the phytoplankton species Phaeodactylum Tricornutum

Anthropogenic radionuclides released into the environment cause a radiation dose to wildlife and humans which must be quantified, both to assess the effect of normal releases, and to predict the consequences of a larger, unplanned release. To estimate the spread of the radioactive elements, the ecosystem around release points is modelled, and element uptake is usually quantified by concentration factors (CF), which relates the concentration of an element in an organism to the concentration of the same element in a medium under equilibrium conditions. In this work, we experimentally determine some phytoplankton CF that are needed for improved modelling of the marine ecosystems around nuclear facilities and release points. CFs that require better determination have been identified through literature search. Sensitivity studies, using the currently used ecosystem modelling software PREDO, show that for most studied groups, the dose committed by the respective radionuclides is almost proportional to the corresponding phytoplankton CFs. In the present work, CFs are determined through laboratory experiments with cultured phytoplankton and radionuclides of the concerned elements, assessing the element uptake by the phytoplankton through detection of the emitted radiation. The three CF assessed in this work were those for manganese, zinc and iodine in phytoplankton. Conservative estimates of these CF based on the present data are 40 000 L/kg for manganese, 50 000 L/kg for zinc and 180 L/kg for iodine with the phytoplankton masses referring to their dry weight

Scoping Studies of Dopants for Stabilization of Uranium Nitride Fuel

Uranium nitride (UN) is considered as nuclear reactor fuel because of, among other reasons, its high uranium density and its high thermal conductivity. Its main drawback is that it relatively easily dissolves in hot water, which is particularly problematic when it is used in water-cooled reactors. One possible remedy to this is to add some corrosion inhibitor as dopant to the UN matrix. A number of dopants have been identified that have the potential to inhibit the dissolution process, and their respective merits have been investigated both by neutronic simulations and dissolution experiments. It is concluded that chromium is the most promising candidate

Modelling of runaway electron dynamics during argon-induced disruptions in ASDEX Upgrade and JET

Disruptions in tokamak plasmas may lead to the generation of runaway electrons that have the potential to damage plasma-facing components. Improved understanding of the runaway generation process requires interpretative modelling of experiments. In this work we simulate eight discharges in the ASDEX Upgrade and JET tokamaks, where argon gas was injected to trigger the disruption. We use a fluid modelling framework with the capability to model the generation of runaway electrons through the hot-tail, Dreicer and avalanche mechanisms, as well as runaway electron losses. Using experimentally based initial values of plasma current and electron temperature and density, we can reproduce the plasma current evolution using realistic assumptions about temperature evolution and assimilation of the injected argon in the plasma. The assumptions and results are similar for the modelled discharges in ASDEX Upgrade and JET. For the modelled discharges in ASDEX Upgrade, where the initial temperature was comparatively high, we had to assume that a large fraction of the hot-tail runaway electrons were lost in order to reproduce the measured current evolution

Thorium fuels for light water reactors - steps towards commercialization

Thorium-containing nuclear fuel is proposed as a means of gaining a number of benefits in the operation of light water reactors, some related to the nuclear properties of thorium and some related to the material properties of thorium dioxide. This thesis aims to investigate some of these benefits and to widen the knowledge base on thorium fuel behaviour, in order to pave the way for its commercial use.Part of the work is dedicated to finding ways of utilizing thorium in currently operating light water reactors which are beneficial to the reactor operator from a neutronic point of view. The effects of adding different fissile components to the fertile thorium matrix are compared, and the neutronic properties of the preferred alternative (plutonium) are more closely investigated. The possibility to use thorium as a minor component in conventional uranium dioxide fuel is also subject to study.Another part of the work is related to the thermal mechanical behaviour of thorium containing nuclear fuel under irradiation. To assess this behaviour, an irradiation experiment has been designed and is ongoing in the Halden research reactor. Existing software for prediction of thermal-mechanical fuel behaviour has been modified for application to mixed thorium and plutonium oxide fuel, and the preliminary simulation output is compared with irradiation data.The conclusion of the research conducted for this thesis is that the adoption of thorium containing fuel in light water reactors is indeed technically feasible and could also beattractive to reactor operators in a number of different aspects. Some steps have been taken towards a more complete knowledge of the behaviour of such fuel and therewith towards its commercial use

Thorium fuels for light water reactors - steps towards commercialization

Thorium-containing nuclear fuel is proposed as a means of gaining a number of benefits in the operation of light water reactors, some related to the nuclear properties of thorium and some related to the material properties of thorium dioxide. This thesis aims to investigate some of these benefits and to widen the knowledge base on thorium fuel behaviour, in order to pave the way for its commercial use.Part of the work is dedicated to finding ways of utilizing thorium in currently operating light water reactors which are beneficial to the reactor operator from a neutronic point of view. The effects of adding different fissile components to the fertile thorium matrix are compared, and the neutronic properties of the preferred alternative (plutonium) are more closely investigated. The possibility to use thorium as a minor component in conventional uranium dioxide fuel is also subject to study.Another part of the work is related to the thermal mechanical behaviour of thorium containing nuclear fuel under irradiation. To assess this behaviour, an irradiation experiment has been designed and is ongoing in the Halden research reactor. Existing software for prediction of thermal-mechanical fuel behaviour has been modified for application to mixed thorium and plutonium oxide fuel, and the preliminary simulation output is compared with irradiation data.The conclusion of the research conducted for this thesis is that the adoption of thorium containing fuel in light water reactors is indeed technically feasible and could also beattractive to reactor operators in a number of different aspects. Some steps have been taken towards a more complete knowledge of the behaviour of such fuel and therewith towards its commercial use

A BWR fuel assembly design for efficient use of plutonium in thorium–plutonium fuel

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

Thorium-plutonium fuel for long operating cycles in PWRs - preliminary calculations

Preliminary calculations have been carried out to investigate the possibility of extending oper-ating cycle length in PWRs by use of Thorium-Plutonium mixed oxide fuel (Th-MOX). Thecalculations have been carried out in two dimensions, using the fuel assembly burnup simula-tion program CASMO-5. The reload scheme and the operating parameters are modelled on theSwedish PWR Ringhals 3 and a normal UOX fuel assembly designed for this reactor has beenused as a reference. Results show that an extension of the currently employed 12-month oper-ating cycle length is possible, either with a burnable absorber or with a modified fuel assemblydesign, assuming the same 3-batch reload scheme as currently used in Ringhals 3.The initial k∞ of the new Th-MOX fuel design was designed not to exceed that of the refer-ence UOX fuel. The power peaking factor is initially significantly lower than the reference,but slightly higher later in the life of the fuel assembly. All reactivity coefficients are withinacceptable range. The worth of control rods and soluble boron are lower than the reference, asexpected for a plutonium-bearing fuel

Thorium as an additive for improved neutronic properties in boiling water reactor fuel

This article treats the replacement of burnable absorbers with a fertile absorber in boiling water reactor fuel. The target is to improve the fuel economy while meeting the same safety demands as the currently used conventional uranium oxide (UOX) fuel. A candidate fertile absorber is Th-232, and this work investigates the impact of replacing part of the U-238 in UOX fuel with Th-232. Computer simulations have been carried out and comparisons made for fuel assemblies with fertile and burnable absorbers, loaded in the boiling water reactor Oskarshamn 3, using the tools and methods that are normally employed for reload design and safety evaluation for this reactor. The results show that power balance and shutdown margins can be improved at the cost of higher enrichment needs. Alternatively, the fuel can be designed to just fulfil the relevant safety criteria, giving slightly lower uranium needs, which may compensate for the increased enrichment costs

DEVELOPMENT OF A FUEL PERFORMANCE CODE FOR THORIUM-PLUTONIUM FUEL

Thorium-plutonium Mixed OXide fuel (Th-MOX) is considered for use as light water reactor fuel. Both neutronic and material properties show some clear benefits over those of uranium oxide and uranium-plutonium mixed oxide fuel, but for a new fuel type to be licensed for use in commercial reactors, its behaviour must be possible to predict. For the thermomechanical behaviour, this is normally done using a well validated fuel performance code, but given thescarce operation experience with Th-MOX fuel, no such code is available today.In this paper we present the ongoing work with developing a fuel performance code for prediction of the thermomechanical behaviour of Th-MOX for light water reactors. The wellestablished fuel performance code FRAPCON is modified by incorporation of new correlations for the material properties of the thorium-plutonium mixed oxide, and by develoment of a new subroutine for prediction of the radial power profiles within the fuel pellets. This paper lists the correlations chosen for the fuel material properties, describes the methodology for modifying the power profile calculations and shows the results of fuel temperature calculations with the code in its current state of development. The code will ultimately be validated using data from a Th-MOX test irradiation campaign which is currently ongoing in the Halden research reactor