Gas Cooled Fast Reactor Research and Development in the European Union

Gas cooled fast reactor research in Europe

Authors

It is a given that fast reactors are sustainable nuclear energy sources, for both utilization of fissile material and minimization of nuclear waste, due to the hard neutron spectrum and the strategies for recycling the nuclear fuel materials. The goal of the gas-cooled fast reactor (GFR) is to convert it into an economic electricity generator, with good sustainability and safety characteristics, but also capable of minimizing nuclear waste via transmutation of minor actinides. This work presents a contribution to the neutronic analysis of the GFR as a transmutation facility of minor actinides. In this study, the fuel assembly is a hexagonal lattice of fuel pins. The materials used are mixes of uranium and plutonium carbide or oxide as fuel in pins, silicon carbide as cladding, and helium gas as coolant. The Monte Carlo code SERPENT was used to perform the criticality calculations during the fuel depletion. Two different fuel mixes of uranium, plutonium and minor actinides in the pins of the assembly were compared during a burnup of 1200 days of irradiation (equivalent to 50 GWd/t). The evolution of the atomic densities and the mass inventory, that of consumption versus production, and that of different fissile, minor actinides, fission products and transuranic nuclides in the fuel, as well as the k-effective multiplication factor during the irradiation time, were tracked. The results confirmed that the radiotoxicity of the nuclear waste of LWRs can be reduced using GFRs. One of the fuel mixes studied came from nuclear fuel discharged of a typical PWR with a burnup of 48 GWd/t and five years of cooling post-discharge. This mix was compared with another resulting from a second recycling. Results for several nuclides are presented and an assessment in terms of advantages for breeding and/or transmutation capabilities of each mix is discussed in the paper.

Gas Cooled Fast Reactor System (GFR)

Improvement of the Decay Heat Removal Characteristics of the Generation IV Gas-cooled Fast Reactor

A system dynamics model is applied to study the impact of spent fuel recycling on the US nuclear energy system and the implications of a bilateral hypothetical collaboration between the US and Brazil. In particular, the impact of different options for advanced fuel cycle facilities needed for the US nuclear energy market alone and under a Brazil partnership is studied. Different recycling technology options are considered: (1) thermal recycling of transuranics (TRUs) in Light Water Reactors (LWRs) using Combined Non-Fertile and UO2 Fuel (CONFU); (2) recycling of TRU in fertile-free metallic fuel in fast Actinide Burner Reactors (ABRs); and (3) fast recycling of TRU with UO2 in self-sustaining Gas-cooled Fast Reactors (GFRs). These recycling options are truly advanced and will require significant development, but they allow the evolution of nuclear energy systems with minimum transuranic elements in the nuclear waste, hence reduce the long term environmental burden of the waste. Case studies for different advanced technology introduction dates are examined under a prescribed rate of growth in demand for nuclear electricity. The timing of introduction of recycling is important for proper technology development. In the model, the rate of deployment is restrained by the industrial capacity to build TRU separation plants, as well as the desire for high utilization factor of the deployed facilities over their life time. It is notable that all recycling options result in only modest reductions in total uranium consumption at the end of the century. Due to its unity fissile conversion ratio, early introduction of the self-sustaining GFR recycling scheme leads to the largest reduction in uranium consumption and enrichment requirements, about 20%. On the other hand, the CONFU recycling scheme keeps the TRU inventory in the entire system well below other schemes, and guarantees equilibrium between the generation and consumption of TRU without investments in fast reactors. The three schemes reduce the TRU sent to the repository for disposal by two orders of magnitude, but the ABR and the GFR schemes require the introduction of the more costly fast reactors. The rate of deployment of fast reactors becomes limited by the availability of TRU from LWRs in the last quarter of the century. Moreover, an assessment of the impact on the US nuclear energy system of a major commitment to supply the nuclear fuel cycle needs of other countries, with Brazil as the example, is made. The partnership with Brazil implies that the US recycling facilities would be used at higher utilization factor, and more TRU will be available to start fast reactors in the US. Economic analysis for the US nuclear energy market indicates that the CONFU technology is more attractive at current uranium price level, and that fast recycling becomes as attractive as thermal recycling at much higher uranium prices.

Development of the control assembly pattern and dynamic analysis of the generation IV large gas-cooled fast reactor (GFR)

GFR equilibrium cycle analysis with the EQL3D procedure

Core neutronics characterization of the GFR2400 Gas Cooled Fast Reactor

Modular Gas-cooled Fast Reactor (GFR) is one of six advanced reactor concepts set by the generation IV international forum. Modular GFR has the potential for use actinide recycling and closed fuel cycle as well as applying fast reactor, using helium gas as the main coolant, high working temperature and low void reactivity effect. The neutronic analysis of nuclear reactor means behavior study of subatomic particles that interact with matter. In this paper, the feasibility of plutonium fuel in modular Gas-cooled Fast Reactor (GFR) was investigated. The Monte Carlo method has advantages in full-scale and heterogeneous three-dimensional (3D) geometry modeling using Evaluated Nuclear Data File (ENDF/B-VIII.b5) nuclear data but requires a highly computation time. Since the progress of high performance computing, the reactor physicist community began proposing to use Monte Carlo method for nuclear reactor simulation through the parallelization of calculations. The GFR feasibility design study will carried out with plutonium fuel as fuel cycle inputs with 5-25% of fissile contain. The most important neutronic parameters characterizing of GFR core are determined for beginning of life (BOL) and during burnup calculation conditions. The results of calculation series in parallel computing give the good agreement will be faster of calculation time when more threads. Materials of (U-Pu)C and (U-Pu)N fuel are the good candidates to be chosen in GFR research that give keff more than 1.2 in fissile contain 20%Pu. The variation fissile contain gives the linearity with keff. In depletion simulation, the core reactor still in critical during 20 years operation, burn up values linear with operation time and mass evolution of plutonium and uranium from start-up core to equilibrium core.

A comparative study on recycling spent fuels in gas-cooled fast reactors

The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection. Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with on outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom and Switzerland), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report outlines the current design status of the GFR, and includes work done in the areas mentioned above for this fiscal year. In addition, this report fulfills the Level 2 milestones, ''Complete annual status report on GFR reactor design'', and ''Complete annual status report on pre-conceptual GFR reactor designs'' in work package GI0401K01. GFR funding for FY05 included FY04 carryover funds, and was comprised of multiple tasks. These tasks involved a consortium of national laboratories and universities, including the Idaho National Laboratory (INL), Argonne National Laboratory (ANL), Brookhaven National Laboratory (BNL), Oak Ridge National Laboratory (ORNL), Auburn University (AU), Idaho State University (ISU), and the University of Wisconsin-Madison (UW-M). The total funding for FY05 was $1000K, with FY04 carryover of $174K. The cost breakdown can be seen in Table 1.

The large 2400 MWth Gas-cooled Fast Reactor (GFR) is one of the most promising advanced reactor concepts currently being investigated within Generation IV. The current work presents the three-dimensional simulation and analysis of GFR core behavior related, primarily, to control assembly (CA) fast movements and/or to accidental withdrawals/ejections. These two categories of events differ in the speed at which the CA is removed. The transient analysis is focused on two different GFR core designs: (1) the reference design (“2004-Core”) and (2) a variant design (“2007-Core”), the main difference being in the core geometry, i.e., in the height-to-diameter (H/D) ratio (2.35 m for the latter core, compared to 1.55 m for the reference design). The reported work firstly concerns the development and benchmarking of coupled, full-core 3D neutron kinetics (NK) and 1D thermal-hydraulics (TH) models for both GFR designs. The models were developed using the TRACE and PARCS codes, which form part of the FAST code system applied at PSI for the generic analysis of advanced fast-spectrum systems. The neutronics is simulated in the nodal diffusion approximation, in conjunction with spatial kinetics. In the TH model, each of the fuel subassemblies (SAs) is modeled individually, so that one can accurately track changes in the thermal-hydraulic-channels surrounding the control assemblies. Consistent nodalization schemes are used in the TH and NK models, in order to provide a 1:1 mapping. Using the developed coupled models, spatial effects have been studied for a variety of hypothetical CA driven transients. Asymmetric CA ejections (speed ∼ 4 cm/s) and withdrawals (speed ∼ 2 mm/s) were analyzed, and large spatial deformations of the power map (up to +21%, relative to average fission power in the TH channels adjacent to the ejected CA) were observed. The analysis has clearly brought out the importance of 3D spatial effects, in the context of the correlation between CA-worths and the loose neutronic coupling of core regions in the case, for example, of a low-H/D core. A sensitivity study of the spatial effects to different parameters (e.g., number of assemblies being ejected and the CA speed) has been conducted, in order to fully characterize the GFR dynamic response. Furthermore, comparisons between the GFR core variants were made. Lower temperatures (fuel, clad and coolant) and lower spatial effects were obtained for the second core. Based upon the results obtained, no major safety-related problem has been found as regard to the CA pattern developed for the GFR. In order to limit the spatial effects, the CAs have to be moved either independently or in symmetric configurations, e.g., three CAs with 120°C symmetry.

Selection of sodium coolant for fast reactors in the US, France and Japan

Fast reactor is the main reactor type of the fourth generation reactors. It has the advantages of good fuel breeding characteristics, strong transmutation ability, high thermal efficiency and good safety. The research of fast reactor is the focus of current advanced reactor. According to the characteristics of fast reactors, the differences between fast reactors and pressurized water reactors in the fuel breeding, transmutation and safety are analyzed in this paper. On this basis, the characteristics of sodium-cooled fast reactors, lead and lead-bismuth eutectic cooled fast reactors and gas-cooled fast reactors are analyzed and their feature differences are compared. According to the basic characteristics of reactor physics, thermal-hydraulics and safety, the characteristics of these fast reactors are analyzed and compared, their respective advantages are given, the technical problems and difficulties are pointed out. The advantages and suitable application fields of these fast reactors are illustrated, the development opportunities and technical challenges are discussed. Some analysis results can provide reference for the research, design and application of advanced nuclear reactors.