Gas cooled fast reactor research in Europe

Gas‐cooled fast reactor (GFR) research is directed towards fulfilling the ambitious goals of Generation IV (Gen IV), that is, to develop a safe, sustainable, reliable, proliferation‐resistant and economic nuclear energy system. The research is directed towards developing the GFR as an economic electricity generator, with good safety and sustainability characteristics. Fast reactors maximise the usefulness of uranium resources by breeding plutonium and can contribute to minimising both the quantity and radiotoxicity nuclear waste by actinide transmutation in a closed fuel cycle. Transmutation is particularly effective in the GFR core owing to its inherently hard neutron spectrum. Further, GFR is suitable for hydrogen production and process heat applications through its high core outlet temperature. As such GFR can inherit the non‐electricity applications that will be developed for thermal high temperature reactors in a sustainable manner. The Euratom organisation provides a route by which researchers in all European states, and other non‐European affiliates, can contribute to the Gen IV GFR system. This paper summarises the achievements of Euratom′s research into the GFR system, starting with the 5th Framework programme (FP5) GCFR project in 2000, through FP6 (2005 to 2009) and looking ahead to the proposed activities within the 7th Framework Programme (FP7).

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

GFR equilibrium cycle analysis with the EQL3D procedure

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

Authors

The Gas-Cooled Fast Reactor (GFR) is one of the reactor concepts selected by the Generation IV International Forum (GIF) for the next generation of innovative nuclear energy systems. It was selected among a group of more than 100 prototypes and his commercial availability is expected by 2030. GFR has common goals as the rest GIF advanced reactor types: economy, safety, proliferation resistance, availability and sustainability. Several GFR fuel design concepts such as plates, rod pins and pebbles are currently being investigated in order to meet the high temperature constraints characteristic for a GFR working environment. In the previous study we have compared the fuel depletion results for heterogeneous GFR fuel assembly (FA), obtained with TRITON6 sequence of SCALE6.0 with the results of the MCNPX-CINDER90 and TRIPOLI-4-D codes. Present work is a continuation of neutronic criticality analysis of heterogeneous FA and full core configurations of a GFR concept using 3-D Monte Carlo codes KENO-VI/SCALE6.0 and MCNP5. The FA is based on a hexagonal mesh of fuel rods (uranium and plutonium carbide fuel, silicon carbide clad, helium gas coolant) with axial reflector thickness being varied for the purpose of optimization. Three reflector materials were analyzed: zirconium carbide (ZrC), silicon carbide (SiC) and natural uranium. ZrC has been selected as a reflector material, having the best contribution to the neutron economy and to the reactivity of the core. The core safety parameters were also analysed: a negative temperature coefficient of reactivity was verified for the heavy metal fuel and coolant density loss. Criticality calculations of different FA active heights were performed and the reflector thickness was also adjusted. Finally, GFR full core criticality calculations using different active fuel rod heights and fixed ZrC reflector height were done to find the optimal height of the core. The Shannon entropy of the GFR core fission distribution was proved to be useful technique to monitor both fission source convergence (stationarity) and core eigenvalue convergence (keff) to fundamental eigenmode with MCNP5. All calculations were done with ENDF/B-VII.0 library. The obtained results showed high similarity with reference results.

Neutronic analysis of comparation UN-PuN fuel and ThN fuel for 300MWth Gas Cooled Fast Reactor long life without refueling has been done. Gas Cooled Fast Reactor is a Generation IV reactor with gas coolant (i.e. helium) and using fast spectrum neutron. The neutronic calculation was carried out using SRAC (Standard Reactor Analysis Code) version 2006 under the Linux Operating System with nuclear data library JENDL4.0. The first calculation is fuel pin cell calculation (PIJ-method) by using a hexagonal cell and then followed by the calculation of the core reactor (CITATION-method). The calculation of the core reactor used homogeneous and heterogeneous core configuration. The UN-PuN fuel use plutonium as a fissile material and natural uranium as a fertile material and the ThN fuel use U233 as a fissile material and natural thorium as a fertile material. The percentages of fissile material are varied in heterogeneous core configuration. It is used to decrease the peaking power in the center of the core. The heterogeneous core configuration contains of Fuel 1 (F1) 8% fissile materials, Fuel 2 (F2) 10% fissile materials, and Fuel 3 (F3) 12% fissile materials. F1 is located in the central core, F2 middle core and F3 outer core. The diameter and height active core are 240 cm and 100 cm. The reflector radial-axial width is 50 cm. All of the calculations can reach burn up time more than 20 years with excess reactivity less than 1 percent (Δk/k <1%) both UN-PuN fuel and ThN fuel. It means that the reactor stable in 20 years. The average of power density both of UN-PuN fuel and ThN fuel are around 66 Watt/cc. The maximum power density of UN-PuN fuel is 94Watt/cc and ThN fule is 129Watt/cc. The UN-PuN fuel has lower maximum power density value than ThN fuel. So, for fast neutron spectrum reactor especially Gas Cooled Fast Reactor type, it is better used UN-PuN fuel than ThN fuel.

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)

The paper reviews the sol–gel methods used for the preparation of nuclear fuel materials in the form of microspheres. It also discusses how these microspheres can be fabricated into nuclear fuels for reactors such as High Temperature Gas Cooled Reactors and Fast Reactors. The performance of these microsphere-based fuels is reviewed. More recent applications, such as the transmutation of minor actinides, (Np, Am and Cm) and hydrogen production, are also briefly covered.

The design study of GFR concepts comprises neutronic analysis of fuel pin and fuel assembly. The Monte Carlo method has advantages in three-dimensional (3D) geometry modeling but requires a high computation time. In this research, the comparative study of Gas-cooled Fast Reactor (GFR) using the Monte Carlo code. The GFR feasibility design study will be carried out with natural uranium with plutonium as fuel cycle inputs. The Monte Carlo method simulates GFR model at full-scale and heterogeneous three-dimensional (3D) using Evaluated Nuclear Data File (ENDF/B-VIII.b5) nuclear data. The code of Monte Carlo methods will be used in this research are the Monte Carlo N - Particle (MCNP) and OpenMC. The comparison of the GFR fuel assembly calculation simulation results is made between the MCNP and OpenMC code. The equilibrium cycle configuration is used as the basis model for the comparisons. The comparison of MCNP and OpenMC code gives a good agreement in criticality calculation of GFR that achieves delta kinf less than 1%. The (U-Pu)N fuel is a good candidate to be chosen in GFR research that gives kinf more than 1.1 in fissile contain 10%Pu and has the highest thermal conductivity. The Zircaloy-4 is the best candidate for material cladding in GFR design that provides the highest kinf.

In this study gas cooled reactor system are combined with modified CANDLE burn-up scheme to create small long life fast reactors with natural circulation as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. Therefore using this type of nuclear power plants optimum nuclear energy utilization including in developing countries can be easily conducted without the problem of nuclear proliferation. In this paper, optimization of Small and Medium Long-life Gas Cooled Fast Reactors with Natural Uranium as Fuel Cycle Input has been performed. The optimization processes include adjustment of fuel region movement scheme, volume fraction adjustment, core dimension, etc. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 75 W/cc. With such condition we investigated small and medium sized cores from 300 MWt to 600 MWt with all being operated for 10 years without refueling and fuel shuffling and just need natural Uranium as fuel cycle input. The average discharge burn-up is about in the range of 23-30% HM.

Core neutronics characterization of the GFR2400 Gas Cooled Fast Reactor