Neutronics Assessment of the Use of Thorium Fuels in Current Pressurized Water Reactors

The use of thorium fuel in current PWRs in a once-through fuel cycle is an attractive option dueto potential advantages such as high conversion ratio and low minor actinidegeneration. Thecurrent neutronics assessments indicate that the thorium fuel cycle could supplement the currenturanium-plutonium fuel cycle to improve operational performance and spent fuel considerationin current PWRs without core and subassembly modifications. Neutronics safety parameters inthe PWR cores with the thorium fuels are within the range of current PWRs. The PWR cores with thorium fuels have significantly higher conversation ratios which couldenable efficient fuel utilization. Further, it is shown that the use of thorium as a fertile materialcan reduce minor actinide generation and the radio-toxicity of spent fuels. In considerationsrelated to proliferation resistance, the results of the current analyses show no significantdifference between the studied thorium fuels and the standard oxide fuel for the assumedcharacteristics and burnup levels.JRC.F.4-Safety of future nuclear reactor

Neutronics Assessment of the Use of Thorium Fuels in Current Pressurized Water Reactors

The use of thorium fuel in current PWRs in a once-through fuel cycle is an attractive option due to potential advantages such as high conversion ratio and low minor actinide generation. The current neutronics assessments indicate that the thorium fuel cycle could supplement the current uranium eplutonium fuel cycle to improve operational performance and spent fuel consideration in current PWRs without core and subassembly modifications. Neutronics safety parameters in the PWR cores with the thorium fuels are within the range of current PWRs. The PWR cores with thorium fuels have significantly higher conversion ratios which could enable efficient fuel utilization. Further, it is shown that the use of thorium as a fertile material can reduce minor actinide generation and the radio-toxicity of spent fuels. In considerations related to proliferation resistance, the results of the current analyses show no significant difference between the studied thorium fuels and the standard oxide fuel for the assumed characteristics and burnup levels.JRC.F.4-Safety of future nuclear reactor

Study on the Use of Hydride Fuel in High-Performance Light Water Reactor Concept

Hydride fuels have features which could make their use attractive in future advanced power reactors. The potential benefit of use of hydride fuel in HPLWR without introducing significant modification in the current core design concept of the High-Performance Light Water Reactor (HPLWR) has been evaluated. Neutronics and thermal hydraulic analyses were performed for a single assembly model of HPLWR with oxide and hydride fuels. The hydride assembly shows higher moderation with softer neutron spectrum and slightly more uniform axial power distribution. It achieves a cycle length of 18 months with sufficient excess reactivity. At Beginning of Cycle the fuel temperature coefficient of the hydride assembly is higher whereas the moderator and void coefficients are lower. The thermal hydraulic results show that the achievable fuel temperature in the hydride assembly is well below the design limits. The potential benefits of the use of hydride fuel in the current design of the HPLWR with the achieved improvements in the core neutronics characteristics are not sufficient to justify the replacement of the oxide fuel. Therefore for a final evaluation of the use of hydride fuels in HPLWR concepts additional studies which include modification of subassembly and core layout designs are required.JRC.F.5-Nuclear Reactor Safety Assessmen

Upgrade of the Sub-channel Thermal-hydraulic Analysis Code COBRA-EN for Super-critical Water Reactors

Super-Critical Water Reactors (SCWRs) have been outlined as a possible option for next generation nuclear reactor technology. Fuel assembly design represents a crucial aspect for the success of this concept. Large density variations, low moderation, heat transfer enhancement and deterioration, all play a key role in the core design parameters. Current qualified sub-channel thermal-hydraulic analysis codes are not able to run under super-critical water conditions. At JRC-IE the sub-channel thermal-hydraulic analysis code COBRA-EN has been upgraded to work above the critical pressure. The water properties package of the code is based on the IAPWS Industrial Formulation 1997 for the Thermodynamic Properties of Water and Steam. The heat transfer and pressure drop correlations for the super-critical region have also been integrated. As part of the code assessment, both a hexagonal and a square fuel assembly configuration have been analysed. The code has also been applied, coupled with MCNP, to investigate the impact of the use of hydride fuel in super-critical fuel assembly. Analyses performed included steady-state density distribution, pressure drop, axial and radial coolant and fuel rod temperatures. COBRA-EN effectively captured the trends seen in similar studies with acceptable accuracy. Future activities will include the implementation of new features (wall, wire wrap model) and the validation of the code against experimental data.JRC.F.4-Safety of future nuclear reactor

Current Status of the CAD Interface Program McCad for MC Particle Transport Calculations

The program McCad provides a CAD interface for the Monte Carlo transport code MCNP. It is able to convert CAD data into MCNP input geometry description and provides GUI components for visualization, data exchange and modeling. It performs sequences of tests on CAD data to check its suitability for neutronics analysises. McCad has been validated in several applications including the generation of a 40º torus sector of ITER. It is coded for Linux platforms in the programming language C++. The implemented conversion algorithm uses the Open CASCADE CAD kernel for its geometric computations and the GUI is based on the Qt4 libraries. The paper provides an insight in the current status of McCad and its functionality. The implemented algorithms for conversion and void completion are described in detail.JRC.F.4-Safety of future nuclear reactor

Pre-conceptual thermal–hydraulics and neutronics studies on sodium-cooled oxide and carbide cores

The Generation IV International Forum (GIF) has among its main goals the excellence in safety and reliability of the proposed innovative nuclear systems. The development of computational tools that are able to assist the design and safety analysis of these innovative reactor concepts is crucial. In this line, the JRC-IET is developing a static and dynamic integrated safety analysis platform with the objective to perform an integrated core and safety analysis of nuclear rector systems. The first application of this platform has been made in the framework of the Collaborative Project on European Sodium Fast Reactor (CP-ESFR) that is also part of EURATOM contribution to GIF. Two core designs have been currently proposed for the 3600 MWth sodium-cooled reactor concept based on oxide and carbide fuel respectively. Using the integrated safety analysis platform, static calculation on neutronics (incl. reactivity coefficients) and thermal-hydraulic characteristics for both oxide and carbide reference cores have been conducted. The quantities evaluated include: keff, coolant heat-up, void, and Doppler reactivity coefficients, axial and radial expansion reactivity coefficients, pin-by-pin calculated power profiles, average and peak channel temperatures. This paper presents the tools and the models applied in the study together with the relevant results for the oxide and carbide core.JRC.F.5-Nuclear Reactor Safety Assessmen

On Use of Hydride Fuel in HPLWR

Hydride fuels have features which could make their use attractive in future advanced power reactors. The feasibility of the use of the metallic hydride fuel U-ZrH1.6 in the current concept of the High-Performance Light Water Reactor (HPLWR) has been assessed. Neutronics and thermal hydraulic analyses were performed for a single assembly model of HPLWR with oxide and hydride fuels. The hydride assembly shows higher moderation with softer neutron spectrum and more uniform axial power distribution. It achieves a cycle length of 18 months with sufficient excess reactivity. The burnup in the hydride assembly is about a factor 2 higher than in the oxide assembly. The Doppler coefficient of the hydride assembly is higher whereas the moderator and void coefficients are lower. The thermal hydraulic results show that the achievable cladding temperature in the hydride assembly is well below the design limits. Due to the high thermal conductivity of hydride fuel, the pellet peak temperature remains below the hydride fuel temperature normal operation limit.JRC.F.4-Safety of future nuclear reactor