Fissile utilization of uranium, thorium, and plutonium fuels for a 60
            <scp>MWt</scp>
            block‐type
            <scp>high‐temperature</scp>
            gas‐cooled reactor

Irwanto, Dwi; Pratama, Ayu Lia

doi:10.1002/er.7506

Cited by 2 publications

(1 citation statement)

References 20 publications

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“…Two fourth-generation nuclear reactors are the Gas-cooled Fast Reactor (GFR) and the Very High-Temperature Gas Reactor (VHTR). Several studies on developing the neutronic aspect of generation IV reactors, especially GFR and VHTR reactors, have been carried out previously [4][5][6][7]. Neutronics analysis is the foundation for creating and designing improvements to the Gas-cooled Reactor, with one important factor in neutronic analysis being the nuclear library.…”

Section: Introductionmentioning

confidence: 99%

Effective Multiplication Factor Analysis of Gas-cooled Reactor using OpenMC code with ENDF/B-VII.1, ENDF/B-VIII.0 and JENDL-5.0 Nuclear Data

Rachmat,

Irwanto

2024

J. Phys.: Conf. Ser.

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As the demand for nuclear energy keeps rising, Generation IV reactor designs continuously improve. One such potential reactor design is the Gas-cooled Reactor. Neutronic analysis is the foundation for ongoing design modifications and alternatives for the Helium Gas-cooled Reactor. An important aspect of neutronic analysis is the nuclear data library. This research compares how different nuclear data libraries impact a Helium Gas-cooled Reactor design by comparing the neutron effective multiplication factor (k-eff) value. This research was conducted by the Monte Carlo method using OpenMC code. ENDF/B-VII.1, ENDF/B-VIII.0 and JENDL-5.0 were used as nuclear data libraries to simulate the reactor design for ten years of depletion time with 1-year timesteps. The reactor is designed in two different spectrums: the fast neutron spectrum design and the thermal neutron spectrum design. The differences between the two designs are in the cladding and reflector materials where the thermal neutron spectrum uses graphite moderator. The results for thermal spectrum design show that while each nuclear data library’s average k-eff values vary, they all produce the same results at timestep zero, with k-eff values around 1.02500. The average k-eff value from simulation using the ENDF/B-VII.1 library is 1.01477, the average k-eff value from simulation using the ENDF/B-VIII.0 library is 1.01122, and the average k-eff value from simulation using the JENDL-5.0 library is 1.01047. Each simulation result has an average uncertainty of 0.00040. The difference is due to the different amounts of data in each library. JENDL-5.0 has the highest amount of data inside its nuclear libraries, followed by ENDF/B-VIII.0 and ENDF/B-VII.1. Meanwhile, the average difference between the different libraries is insignificant for the thermal gas-cooled reactor simulation results. ENDF/B-VIII.0 produces an effective k-average of 1.04437, and JENDL-5.0 produces an effective k-average of 1.04273. Mass change over time results also shows that the transmutation of Uranium-238 into Plutonium-239 occurred. The fast spectrum reactor design shows a greater increase in plutonium-239 than the thermal spectrum design. The three libraries used did not show large differences in results in calculating changes in the mass of Uranium-238 over time due to Uranium-238‘s large mass, but they still showed differences in Plutonium-239‘s mass change over time. The average value difference between libraries in the fast neutron spectrum reactor design is 50 kg, and in the thermal spectrum it is 30 kg.

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Section: Introductionmentioning

confidence: 99%

Effective Multiplication Factor Analysis of Gas-cooled Reactor using OpenMC code with ENDF/B-VII.1, ENDF/B-VIII.0 and JENDL-5.0 Nuclear Data

Rachmat,

Irwanto

2024

J. Phys.: Conf. Ser.

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Effect of Different Temperature and Nuclear Data Libraries in Criticality Calculations of 300 MWt Molten Salt Reactor

Variastuti,

Irwanto

2024

J. Phys.: Conf. Ser.

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The Molten Salt Reactor (MSR) is a type of reactor in which coolant and fuel are blended with liquid salts. Liquid fuel enhances heat transfer and temperature control, establishing the MSR as one of the Generation IV reactor types. This study aims to analyze the impact of temperature variations on MSR reactor criticality and evaluate the effect of different nuclear data libraries, namely JENDL 3.2, JENDL 3.3, and JENDL 4.0, in neutron analysis. The observed temperature variations are 600 K, 700 K, 833 K, 903 K, and 1000 K. The analysis spans low to operational temperatures to comprehend reactor performance under differing thermal conditions. The study is conducted on a 300 MWTh MSR design, utilizing SRAC2006 with PIJ and CITATION modules for calculations. The results reveal the impact of temperature variations and nuclear data library disparities on reactor criticality. The outcomes demonstrate that higher temperatures correspond to lower values of the effective multiplication factor. At lower temperatures, neutrons experience greater moderation compared to higher temperatures. As a result, a higher number of thermal neutrons influences an increased probability of fission reactions within the reactor. Different nuclear data libraries also yield varied criticality values due to differing cross-sectional areas and quantities of data within each JENDL library. JENDL 4.0 generates the highest criticality value, attributed to elevated cross-sectional regions of each respective nuclear data entry and a greater quantity of nuclear data entries than JENDL 3.2 and JENDL 3.3.

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Fissile utilization of uranium, thorium, and plutonium fuels for a 60 MWt block‐type high‐temperature gas‐cooled reactor

Cited by 2 publications

References 20 publications

Effective Multiplication Factor Analysis of Gas-cooled Reactor using OpenMC code with ENDF/B-VII.1, ENDF/B-VIII.0 and JENDL-5.0 Nuclear Data

Effective Multiplication Factor Analysis of Gas-cooled Reactor using OpenMC code with ENDF/B-VII.1, ENDF/B-VIII.0 and JENDL-5.0 Nuclear Data

Effect of Different Temperature and Nuclear Data Libraries in Criticality Calculations of 300 MWt Molten Salt Reactor

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