Design Study of 200MWth Gas Cooled Fast Reactor with Nitride (UN-PuN) Fuel Long Life without Refueling

Syarifah, Ratna Dewi; Yulianto, Yacobus; Su’ud, Zaki; Basar, Khairul; Irwanto, Dwi

doi:10.1051/matecconf/20168203008

Cited by 13 publications

(6 citation statements)

References 4 publications

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“…The data refers to the parameters and specifications of the previously optimized SRAC code, which are listed in Table 1. SRAC reference data uses JENDL 3.2 version of cross section library data [22]. Data analysis in this study consisted of 3 stages, i.e determining the minimum number of particles, determining the minimum number of cycles through entropy data, and code validation.…”

Section: Design and Calculation Methodsmentioning

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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Section: Design and Calculation Methodsmentioning

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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“…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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“…The calculation use hexagonal fuel pin calculation and R‐Z geometry core calculation with SRAC2006 code system and JENDL4.0 data libraries. Beforehand, neutronic analysis of GFR with SRAC2006 code system with modified CANDLE burnup scheme has been analyzed . And then the neutronic calculation of thorium fuel has been calculated before .…”

Section: Introductionmentioning

confidence: 99%

“…Beforehand, neutronic analysis of GFR with SRAC2006 code system with modified CANDLE burnup scheme has been analyzed. [2][3][4][5][6][7][8][9] And then the neutronic calculation of thorium fuel has been calculated before. 10 The calculation is for thermal reactors, ie, high temperature gas cooled reactor, pressurized water reactor, and boiling water reactor.…”

Section: Introductionmentioning

confidence: 99%

Comparison of uranium plutonium nitride (U-Pu-N) and thorium nitride (Th-N) fuel for 500 MWth gas-cooled fast reactor (GFR) long life without refueling

et al. 2017

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SummaryComparison of uranium plutonium nitride and thorium nitride fuel for 500 MWth gas-cooled fast reactor has been done. Gas-cooled fast reactor is one type of generation IV reactor that can be operated in high temperature. Due to the high temperature, it can be used in hydrogen production. In this study, we compare the neutronic analysis of two fuel types, ie, uranium nitride fuel (U,Pu)N and thorium nitride fuel (Th,U233)N. The neutronic calculation uses SRAC2006 code system, and the data libraries use JENDL4.0.First, the fuel pin calculation (PIJ calculation) has been done to take the macro data that are used in CITATION calculation. Both uranium and thorium fuel use heterogeneous configuration with 3 variation fuel in the core. F1 is located in the central core, F2 middle core, and F3 outer core.The variation of fuel fraction is 40% until 65%, cladding 10%, and coolant 25% until 40%. For (U,Pu)N fuel, the diameter of the active core is 220 cm, and the height of the active core is 110 cm. And for (Th,U233)N, the diameter of the active core is 250 cm and the height of the active core is 150 cm. The reflector radial-axial width is 50 cm. For uranium plutonium nitride fuel, the type of the fuel in one core is varied; ie, F1 is 8%, F2 is 10%, and F3 is 12%. For thorium nitride fuel, the type of the fuel in one core also varied; ie, F1 is 7.8%, F2 is 8%, and F3 is 8.8%. The optimum value of thorium nitride is when fuel fraction of region F1 = 60%, F2 = 57.5%, F3 = 60%, the burn up time up to 20 years without refueling, max k-eff value is 1.0109, and max excess reactivity value is 1.08%. Neutronic analysis shows that both uranium and thorium fuel have excess reactivity value less than 2% but thorium fuel has excess reactivity less than uranium nitride fuel.Uranium fuel has better breeding capability than thorium fuel. Therefore, it is better to use uranium fuel for fast reactor like GFR, which has high breeding capability.

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Design Study of 200MWth Gas Cooled Fast Reactor with Nitride (UN-PuN) Fuel Long Life without Refueling

Cited by 13 publications

References 4 publications

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

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

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

Comparison of uranium plutonium nitride (U-Pu-N) and thorium nitride (Th-N) fuel for 500 MWth gas-cooled fast reactor (GFR) long life without refueling

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