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Neutronic analysis of Thorium Nitride (Th, U233)N fuel of 500MWth Gas Cooled Fast Reactor (GFR) has been done. In this study the neutronic analysis use SRAC2006 code both PIJ and CITATION calculation. The data libraries use JENDL 4.0. First calculation is survey parameter with U-233 enrichment variation. From the homogeneous core configuration calculation, when the enrichment of U-233 is 8.2%, the maximum k-eff value is 1,00819 with excess reactivity value 0,812%. The average power density is 63 Watt/cc and the maximum power density 100 Watt/cc. The heterogeneous core configuration calculation has been done to flattening the power of the reactor. The variation fuel of F1:F2:F3 = 7.8%:8%:8.8%. The fraction of fuel : cladding: coolant = 60%:10%:30%. The max k-eff value of heterogeneous core configuration is 1,01229 with excess reactivity value 1.21%. The average power density is 65 Watt/cc and the maximum power density 92 Watt/cc. The power density distribution of heterogeneous core configuration is flatter than homogeneous core configuration.

The modified CANDLE (Constant Axial shape of Neutron flux, nuclide densities, and power shape During Life of Energy production) strategy shuffling in an axial direction has been successfully applied to gas cooled fast reactor. This paper investigated the utilization of natural uranium, enriched nitride and Thorium ( 238 U 15 N and 232 Th) as fuel on 3000MWt reactor power and refuelling every 10 years of burnup. The reactor core is partitioned into ten regions with the same volume in the axial direction. Initially, a fuel input ( 238 U 15 N and 232 Th) is placed in region 1. After ten years of burnup, a fuel in region 1 was moved to region 2, then to region 3, and so on until the fuel in region 9 was moved to region 10. The fuel in 10 th region was removed in the core. The neutronic computations were performed in two different ways by using SRAC 2006 code and JENDL4.0 as a nuclear data library. Firstly, PIJ was used for fuel cell calculations and secondly CITATION for a reactor core calculation. The results show that the effective multiplication factor is greater than one, at the beginning of life, its value is approximately 1.0028 and at the end of life is approximately 1.0279. This indicates that the reactor is capable to operate through a burn-up period by employing 238 U 15 N and 232 Th (10%) as fuel cycle input. At the beginning of life, the integral conversion ratio is approximately 9.9455 and at the end of life is approximately 1.2906. The burnup level at the end life is about 30.29% HM.

The use of mixed fuel in the form of thorium-uranium nitride (ThN-UN) offers the potential to improve thermal efficiency and optimize criticality, especially in fast reactors like Gas-cooled Fast Reactors (GFRs). To maintain its criticality, Th-232 absorbs neutrons and transforms into a new fissile material U-233. The inclusion of fertile material Th-232 necessitates a more precise reactor geometry design to ensure the reactor remains critical state until the end of the burn-up period. This study aims to compare variations in the percentage of U-233 enrichment in heterogeneous core geometries with five fuel variations (F1:F2:F3:F4:F5) using the OpenMC program at a power of 100 MWth. Benchmarking is performed by measuring the effective multiplication factor (k-eff) over 5 years of burn-up using reference data from previous studies. If the error value is less than 2%, further calculations will be conducted for both homogeneous and heterogeneous core configurations. The homogeneous calculations indicate that the U-233 enrichment percentage of 8.5% yields the most optimal results. This homogeneous data then is used for calculations in the heterogeneous core with 5 fuel variations. The heterogeneous core configuration is designed using five types different fuel percentages cases, each with variations in ring geometry. The comparison results show that the case 5 heterogeneous core geometry design, with a percentage distribution of 7%:7.5%:8.5%:9%:10.5%, achieves good optimization in terms of k-eff value, excess reactivity, neutron flux, and extended burn-up over a 10-years period.

Gas-cooled Fast Reactor (GFR) is one of the fourth-generation reactors that is still being developed. The GFR system combines the advantages of a fast spectrum system for long-term sustainability of uranium resources and waste reduction (through fuel reprocessing and long-lived actinide fission) with a high-temperature system (industrial use and high thermal cycle efficiency). Several types of reactors, research methods have been carried out such as the type of output power produced, fuel variations, fuel composition variations, ring geometry variations, and core configuration variations have been carried out. Research has been carried out related to the thermal hydraulic analysis of GFR using various numerical methods such as Computation Fluid Dynamics, Runge Kutta method up to Order 5, a genetic algorithm, and MATLAB code to find some parameters i.e. d. This article discusses the results of several neutronic and thermal hydraulic analyzes from research that has been carried out and discusses things that can be developed from previous GFR research.

This review summarizes the development of the Gas-Cooled Fast Reactor (GFR) concept from the early 1970s until now, focusing specifically on structural materials and advanced fuel cladding materials. Materials for future nuclear energy systems must operate under more extreme conditions than those in the current Gen II or Gen III systems. These conditions include higher temperatures, a higher displacement per atom, and more corrosive environments. This paper reviews previous GFR concepts in light of several promising candidate materials for the GFR system. It also reviews the recent development of nuclear power and its use in the peaceful exploration of space. The final section focuses on the development and testing of new advanced materials such as SiCf/SiC composites and high entropy alloys (HEA) for the construction and development of GFRs.

The paper builds on the results so far within the research project focused on conceptual design of an innovative safety system for a Gas-Cooled Nuclear Reactors. The paper focuses on a description of the design methodology of the gas turbine flow path, which is part of the secondary circuit of the proposed safety system. The aim is to propose the design of the last stage turbine blading and then to extract a special bucket blade profile that will match the expected flow direction during deep off-design modes. The extracted blade profile will then be subsequently assembled into a blade row which will be tested in the suction-type high-speed wind tunnel in the Aerodynamic Laboratory of the Institute of Thermomechanics Czech Academy of Science in Nový Knín.

The development and deployment of a new generation of nuclear reactors necessitates a thorough evaluation of techniques used to characterize nuclear materials for nuclear forensic applications. Advanced fuels proposed for use in these reactors present both challenges and opportunities for the nuclear forensic field. Many efforts in pre-detonation nuclear forensics are currently focused on the analysis of uranium oxides, uranium ore concentrates, and fuel pellets since these materials have historically been found outside of regulatory control. The increasing use of TRISO particles, metal fuels, molten fuel salts, and novel ceramic fuels will require an expansion of the current nuclear forensic suite of signatures to accommodate the different physical dimensions, chemical compositions, and material properties of these advanced fuel forms. In this work, a semi-quantitative priority scoring system is introduced to identify the order in which the nuclear forensics community should pursue research and development on material signatures for advanced reactor designs. This scoring system was applied to propose the following priority ranking of six major advanced reactor categories: (1) molten salt reactor (MSR), (2) liquid metal-cooled reactor (LMR), (3) very-high-temperature reactor (VHTR), (4) fluoride-salt-cooled high-temperature reactor (FHR), (5) gas-cooled fast reactor (GFR), and (6) supercritical water-cooled reactor (SWCR).

As one of the advanced reactors new generation concept, Gas-cooled Fast Reactors can achieve higher electricity efficiency compare to the previous generation. The modular design is perfectly matched with Indonesia region that has many small islands. In this paper will be discussed about the design study of the fuel assembly for modular nuclear power plants using helium as cooling (GFR). The design reactors with hexagonal assembly and square assembly as a comparison to see which assembly configuration in GFR to get optimum results. Calculation for neutronic used Monte Carlo method with parallelization to accelerate computing time. Using OpenMC program code to build full three-dimension core simulation and neutronic calculation with nuclear data ENDF/B-VIII.b5 as a library. The neutronics results show the optimum design can be achieved.

Indonesia has many small regions and islands with low electrification ratio that needs small power plants. Gas-cooled Fast Reactors (GCFR) is one of the fourth generations of an advanced nuclear power plant with an improved safety system and optimum electricity production that match Indonesia's energy needs. The neutronic analysis study and optimization of small modular GCFR has been performed with Monte Carlo method OpenMC code that use computational parallelization to speed up calculation time. The full core reactor design is cylindrical with a radius 100 cm and height is 120 cm. Additional 50 cm of radius and 40 cm for the height for the reflector. The core using several types of fuel composed of natural Uranium mixed with spent fuel Plutonium. The variation in fuel fraction pin and percentages of Plutonium in fuel to achieved optimum core design. The design of core reactor has flattened flux and power distribution.

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.