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This study aims to design a fuel cell based on Natural Uranium Metal (U-10%wtZr) with the addition of minor actinides, such as Am-241, Np-237, and Cm-244. The coolant used was Helium, and the cladding was carried out with Stainless Steel (SS316). This fuel cell is pin-shaped with a diameter of 1.4 cm and was then tested using 5% enrichment with and without minor actinide. Calculation was then carried out based on the neutron transport equation using SRAC. The results showed that the value of the infinite multiplication factor (KInf) after the addition of 5% Am-241 and Np-237 with 5% enrichment has not reached a critical state and it decreased at the beginning of the burn-up time. However, as more percent of Cm-244 was added, the KInf increased and reached a critical state at a concentration of 4% with a value of 1.001198. [ABSTRACT FROM AUTHOR]

This study aims to analyze the fuel cells in a Gas-Cooled Fast Reactor (GFR) after the addition of different Uranium Nitride (UN) concentrations and minor actinide using OpenMC. Variations in fuel composition were carried out by changing the percentage of U-235 content from 0% to 10%. The results showed that the greater the composition of U-235, the higher the value of the Effective Multiplication Factor (Keff) and reaction rate. Meanwhile, the flux distribution in the center showed the distribution of the most neutrons, which reduced as they move away from the center. Minor actinides were added in the form of Am241 and Cm244 when the U-235 content was at 5%. The addition of Am241 led to a decrease in the Keff value, but it increased along with burn-up time after adding Cm244. The minor actinides were added to reduce the use of uranium and increase the utilization of nuclear waste. [ABSTRACT FROM AUTHOR]

This study analyzes pin-shaped fuel cells in two types of fuel mixtures, namely Uranium-Plutonium-Nitride and Uranium-Plutonium-Zirconium. Each fuel is varied by changing the percentage of Uranium-235 content as much as 1% to 10%. Using the Monte Carlo method in the OpenMC program code, calculations were performed to obtain the effective multiplication factor, rate of fission reaction, and distribution of neutron flux for two years of burn-up. The calculation results for the effective multiplication factor and the rate of fission reaction show that the greater the enrichment of uranium, the greater the value of these parameters. In the neutron flux distribution calculation fuel cells have the highest value in the middle area of the fuel, and its value gets smaller toward the edge of the cell. By doing this, fuel cell analysis so that later it can be used as a reference in the preparation of fuel cells in the reactor core that is safe and efficient. [ABSTRACT FROM AUTHOR]

The objective of the present research is to assess the feasibility design of small long-life Gas Cooled Fast Reactor with helium as coolant. GCFR included in the Generation-IV reactor systems are being developed to provide sustainable energy resources that meet future energy demand in a reliable, safe, and proliferation-resistant manner. This reactor can be operated without enrichment and reprocessing forever, once it starts. To obtain the capability of consuming natural uranium as fuel cycle input modified CANDLE burn-up scheme was adopted in this system with different core design. This study has compared the core with three designs of core reactors with the same thermal power 600 MWth. The fuel composition each design was arranged by divided core into several parts of equal volume axially i.e. 6, 8 and 10 parts related to material burn-up history. The fresh natural uranium is initially put in region 1, after one cycle of 10 years of burn-up it is shifted to region 2 and the region 1 is filled by fresh natural uranium fuel. This concept is basically applied to all regions, i.e. shifted the core of the region (i) into region (i+1) region after the end of 10 years burn-up cycle. The calculation results shows that for the burn-up strategy on "Region-8" and "Region-10" core designs, after the reactors start-up the operation furthermore they only needs natural uranium supply to the next life operation until one period of refueling (10 years). [ABSTRACT FROM AUTHOR]

A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8% HM. From the neutronic point of view, this design is in compliance with good performance. [ABSTRACT FROM AUTHOR]

A design study of small Gas Cooled Fast Reactor with two region fuel has been performed. In this study, design GCFR with Helium coolant which can be continuously operated by supplying mixed Natural Uranium without fuel enrichment plant or fuel reprocessing plant. The active reactor cores are divided into two region fuel i.e. 60% fuel fraction of Natural Uranium as inner core and 65% fuel fraction of Natural Uranium as outer core. Each fuel core regions are subdivided into ten parts (region-1 until region-10) with the same volume in the axial direction. The fresh Natural Uranium initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 filled by fresh Natural Uranium. This concept is basically applied to all regions in both cores area, i.e. shifted the core of ith region into i+1 region after the end of 10 years burn-up cycle. For the next cycles, we will add only Natural Uranium on each region-1. The burn-up calculation is performed using collision probability method PIJ (cell burn-up calculation) in SRAC code which then given eight energy group macroscopic cross section data to be used in two dimensional R-Z geometry multi groups diffusion calculation in CITATION code. This reactor can results power thermal 600 MWth with average power density i.e. 80 watt/cc. After reactor start-up the operation, furthermore reactor only needs Natural Uranium supply for continue operation along 100 years. This calculation result then compared with one region fuel design i.e. 60% and 65% fuel fraction. This core design with two region fuel fraction can be an option for fuel optimization. [ABSTRACT FROM AUTHOR]

A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE burn-up scheme has been performed. In this study, design GCFR with Helium coolant which can be continuously operated by supplying mixed Natural Uranium/Thorium without fuel enrichment plant or fuel reprocessing plant. The active reactor cores are divided into two region, Thorium fuel region and Uranium fuel region. Each fuel core regions are subdivided into ten parts (region-1 until region-10) with the same volume in the axial direction. The fresh Natural Uranium and Thorium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh natural Uranium/Thorium fuel. This concept is basically applied to all regions in both cores area, i.e. shifted the core of ith region into i+1 region after the end of 10 years burn-up cycle. For the next cycles, we will add only Natural Uranium and Thorium on each region-1. The calculation results show the reactivity reached by mixed Natural Uranium/Thorium with volume ratio is 4.7:1. This reactor can results power thermal 550 MWth. After reactor start-up the operation, furthermore reactor only needs Natural Uranium/Thorium supply for continue operation along 100 years. [ABSTRACT FROM AUTHOR]

A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE Burn-up has been performed. The objective of this research is to get optimal design parameters of such type reactors. The parameters of nuclear design including the critical condition, conversion ratio, and burn-up level were compared. These parameters are calculated by variation in the fuel fraction 47.5% up to 70%. Two dimensional full core multi groups diffusion calculations was performed by CITATION code. Group constant preparations are performed by using SRAC code system with JENDL-3.2 nuclear data library. In this design the reactor cores with cylindrical cell two dimensional R-Z core models are subdivided into several parts with the same volume in the axial directions. The placement of fuel in core arranged so that the result of plutonium from natural uranium can be utilized optimally for 10 years reactor operation. Modified CANDLE burn-up was established successfully in a core radial width 1.4 m. Total thermal power output for reference core is 550 MW. Study on the effect of fuel to coolant ratio shows that effective multiplication factor (keff) is in almost linear relations with the change of the fuel volume to coolant ratio. [ABSTRACT FROM AUTHOR]

Nuclear reactor design and analysis of next-generation reactors require a comprehensive computing which is better to be executed in a high performance computing. Flat flux (FF) approach is a common approach in solving an integral transport equation with collision probability (CP) method. In fact, the neutron flux distribution is not flat, even though the neutron cross section is assumed to be equal in all regions and the neutron source is uniform throughout the nuclear fuel cell. In non-flat flux (NFF) approach, the distribution of neutrons in each region will be different depending on the desired interpolation model selection. In this study, the linear interpolation using Finite Element Method (FEM) has been carried out to be treated the neutron distribution. The CP method is compatible to solve the neutron transport equation for cylindrical geometry, because the angle integration can be done analytically. Distribution of neutrons in each region of can be explained by the NFF approach with FEM and the calculation results are in a good agreement with the result from the SRAC code. In this study, the effects of the mesh on the keff and other parameters are investigated. [ABSTRACT FROM AUTHOR]