Neutronic analysis on molten salt reactor FUJI-12 using 235U as fissile material in LiF-BeF2-UF4 fuel

Mabruri, Ahmad Muzaki; Syarifah, Ratna Dewi; Aji, Indarta Kuncoro; Hanifah, Zein; Arkundato, Artoto; Jatisukamto, Gaguk

doi:10.15587/1729-4061.2022.265798

Cited by 3 publications

(3 citation statements)

References 5 publications

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“…The plutonium material consists of several isotopes, including Pu-238, Pu-239, Pu-240, Pu-241, and Pu-242. The percentage of plutonium isotopes respectively is 1.8 %, 58.7 %, 24.2 %, 11.4 %, and 3.9 % [18]. The reactor core is coated by the reflector and absorber materials axially and radially.…”

Section: Methodsmentioning

confidence: 99%

Analysis of variation minor actinide pin configurations Np-237, AM-241, and Cm-244 in UN-PuN fueled pressurized water reactor

Syarifah,

Nasrullah,

Prasetya

et al. 2024

Eureka: PE

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Actinide minor is a reactor waste with high toxicity and a long half-life. Minor actinides can be reduced by reusing them as fuel mixtures in reactors. This research uses PWR reactors with the primary fuel UN-PuN or Uranium Plutonium Nitride with a burning time of 5 years. The fuel consists of enriched Uranium, reactor-grade Plutonium from LWR waste, and minor actinides including Neptunium-237, Americium-241, and Curium-244. The purpose of this study was to find a design that is effective in reducing minor actinide waste. There are six designs or cases used in the addition of minor actinides. Each case has six minor actinide pins in each assembly. The addition of minor actinides is arranged in heterogeneous cores. The analysis was carried out by observing the values of k-eff, excess reactivity, and mass of minor actinides obtained from simulations using OpenMC code 0.13.2 and the ENDF/B-VIII library. The homogeneous core obtained an excess reactivity of 9.7 % with a percentage of plutonium of 8 %. The results of the homogeneous core are used as a reference for preparing a heterogeneous core. The heterogeneous core obtained an excess reactivity of 9.9 % with a percentage of plutonium F1: 5.5 %, F2: 8 %, and F3: 10.5 %. Np-237 can be reduced by 53 kg, and Am-241 can be reduced by 61 kg with minor actinide pins in case 1. Cm-244 can be reduced by 363 kilograms with minor actinide pins in case 6. Excess reactivity in the addition of Np-237 and Am-241 decreased to 5.3 %, while the accumulation of Cm-244 increased to 12.1 %.

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Section: Methodsmentioning

confidence: 99%

Analysis of variation minor actinide pin configurations Np-237, AM-241, and Cm-244 in UN-PuN fueled pressurized water reactor

Syarifah,

Nasrullah,

Prasetya

et al. 2024

Eureka: PE

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“…Meanwhile, the absorber comes from boron carbide (B4C) material, which has high hardness properties, stability at high temperatures, and good neutron absorption ability [26]. Reflector and absorber materials have a density of 3.210 gr/cm 3 and 2.52 gr/cm 3 [27].…”

Section: Gas-cooled Fast Reactor Designmentioning

confidence: 99%

A comparative analysis of gas-cooled fast reactor using heterogeneous core configurations with three and five fuel variations

Prasetya,

Syarifah,

Karomah

et al. 2024

EEJET

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GFR or Gas-cooled Fast Reactor is one type of fast generation-IV that uses a very high cooling temperature. Thus, it is necessary to have the right reactor core design so that the power distribution of neutrons produced reaches a safe and even limit point. The use of a uniform (homogeneous) reactor core can produce peaking power. This is very avoidable because it will cause a reactor accident. In this study, researchers tried to compare the results of the analysis for two heterogeneous reactor core designs including the configuration of 3 fuel variations and 5 fuel variations using UN-PuN fuel. This study aims to determine the keff value produced by both types of fuel variations during 5 years of burn-up and determine the characteristics of neutron flux, fission rate, and fission product during 15 years of burn-up. This study was started by calculating the homogeneous and heterogeneous core of 3 and 5 fuel variations with neutron transport simulation involving OpenMC. The calculation results show that the heterogeneous core configuration of 5 fuel variations for the keff value is more optimal than 3 fuel variations, because it has the smallest excess reactivity value. The neutron flux and fission rate characteristics for 5 fuel variations are more evenly distributed when compared to 3 fuel variations to maintain neutron lifetime and reactor life in operation. Burn-up residual plutonium material and minor actinide waste for 5 fuel variations have less mass than 3 fuel variations. The results of neutronic analysis of GFR reactors with heterogeneous reactor core designs for 5 fuel variations are better in terms of reactor criticality, neutron power distribution, and waste produced. Finally, optimization of the UN-PuN fuel volume fraction of 60 % provides the optimal keff value

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“…GFR is classified as a nuclear power plant which has the opportunity to replace the role of coal in producing energy in the future [1]. GFR or Gas-cooled Fast Reactor is one type of IV generation reactor that prioritizes four main objectives, i.e sustainability, safety and reliability, economics, and proliferation resistance and physical protection [2].…”

Section: Introductionmentioning

confidence: 99%

Validation of OpenMC Code Criticality Value Calculation for GFR Reactor with UN-PuN Fuel

Prasetya,

Mabruri,

Karomah

et al. 2024

J. Phys.: Conf. Ser.

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The use of OpenMC code needs to be validated with other code, so that an accurate and valid criticality value is generated. The validation value is affected by the minimum number of particles used. Determination of the minimum number of particles is carried out by varying the number of particles so that they have a convergent value, which is in the range of 100-50000 particles. The results of determining the minimum number of particles are then used to find the minimum number of cycles consisting of active and inactive batches. This study uses the criticality value parameter in the form of an effective multiplication factor (k-eff) as a benchmark for code accuracy. The k-eff values generated by OpenMC and SRAC are then compared and the validation error value is searched. The error value is found by calculating the k-eff (Δk-eff) difference between the two codes. OpenMC code can be said to be validated if it has an error value of less than 1% against the calculation results of the SRAC code. The calculation results of determining the minimum number of particles show k-eff values in the range of 35000 - 50000 minimal changes significantly or not fluctuating. The calculation of k-eff and entropy in cycle 500 shows a convergent value at the active batch value of >30. Based on the data cycle, researchers used 100 active batches and 30 inactive batches to perform validation calculations. The results of the validation calculation using UN-PuN fuel with a plutonium material composition of 10% showed a maximum k-eff error value of 0.819%. The Δk-eff obtained is 0.008.

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Neutronic analysis on molten salt reactor FUJI-12 using 235U as fissile material in LiF-BeF2-UF4 fuel

Cited by 3 publications

References 5 publications

Analysis of variation minor actinide pin configurations Np-237, AM-241, and Cm-244 in UN-PuN fueled pressurized water reactor

Analysis of variation minor actinide pin configurations Np-237, AM-241, and Cm-244 in UN-PuN fueled pressurized water reactor

A comparative analysis of gas-cooled fast reactor using heterogeneous core configurations with three and five fuel variations

Validation of OpenMC Code Criticality Value Calculation for GFR Reactor with UN-PuN Fuel

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