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

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

Comparison of Thorium-Plutonium fuel and MOX fuel for PWRs

Thorium-plutonium (Th,Pu) oxide fuels will provide an evolutionary way tosimultaneously reduce plutonium volumes and capture energy from this material. In this work wecompare the neutronic properties of Th,Pu-fuel and MOX fuel with different Pu isotope vectors.For these studies, burn-up simulations are performed for a regular MOX PWR fuel assembly andfor a thorium-plutonium PWR fuel assembly of the same geometry. The neutronic properties andperformance of the assemblies are investigated by lattice calculations using CASMO-5. Theplutonium content of the two fuel types is chosen so that the same total energy release per fuelassembly is achieved, which demanded a somewhat higher plutonium content in the thoriumplutoniumcase. The assemblies are then analyzed with regards to temperature coefficients,delayed neutron fractions, control rod and boron worths, coolant void reactivity (CVR) and decayheat. Overall, the results show that MOX and Th,Pu-fuel have fairly similar neutronic propertiesin existing PWRs. Th,Pu-fuel offers an advantage over MOX fuel with regards to CVR values andplutonium consumption. The conclusion is therefore that introducing Th,Pu-fuel would improvethese factors without imposing any major hurdles from a reactor physics point of view

Comparison of Thorium-Plutonium fuel and MOX fuel for PWRs

Thorium-plutonium (Th,Pu) oxide fuels will provide an evolutionary way tosimultaneously reduce plutonium volumes and capture energy from this material. In this work wecompare the neutronic properties of Th,Pu-fuel and MOX fuel with different Pu isotope vectors.For these studies, burn-up simulations are performed for a regular MOX PWR fuel assembly andfor a thorium-plutonium PWR fuel assembly of the same geometry. The neutronic properties andperformance of the assemblies are investigated by lattice calculations using CASMO-5. Theplutonium content of the two fuel types is chosen so that the same total energy release per fuelassembly is achieved, which demanded a somewhat higher plutonium content in the thoriumplutoniumcase. The assemblies are then analyzed with regards to temperature coefficients,delayed neutron fractions, control rod and boron worths, coolant void reactivity (CVR) and decayheat. Overall, the results show that MOX and Th,Pu-fuel have fairly similar neutronic propertiesin existing PWRs. Th,Pu-fuel offers an advantage over MOX fuel with regards to CVR values andplutonium consumption. The conclusion is therefore that introducing Th,Pu-fuel would improvethese factors without imposing any major hurdles from a reactor physics point of view

Comparison of thorium-based fuels with different fissile components in existing boiling water reactors

Three different types of thorium based BWR fuel have been developed, in each of which thorium was combined with a different fissile component, the three components being reactor grade plutonium, uranium enriched to 20% in uranium 235 and pure uranium 233. A BWR nuclear bundle design, based on the geometrical fuel assembly design GE14, was developed for each of these fissile components. The properties and performance of the corresponding fuel assemblies were investigated via full core calculations carried out for an existing BWR and compared with the ones of an ordinary Low Enriched Uranium (LEU) fuel, which was developed for reference. The fuel assemblies and cores were designed to meet existing fuel design criteria, and were then analyzed with regards to reactivity coefficients, delayed neutron fractions, control rod worths and shutdown margins. The results show that all three alternatives seem to be feasible, although some difficulties remain with complying with the thermal limits, and with the moderator temperature and coolant void coefficients of the U-233 containing fuel being positive under some circumstances