Forthcoming articles

International Journal of Nuclear Energy Science and Technology

International Journal of Nuclear Energy Science and Technology (IJNEST)

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International Journal of Nuclear Energy Science and Technology (3 papers in press)

Regular Issues

  • Burn/breed: a wavelet-based nuclear fuel burnup and decay heat code   Order a copy of this article
    by Hesham Nasif 
    Abstract: Burn/breed is a code designed to aid in the analysis, prediction and optimisation of fuel burnup performance and decay heat calculation in a nuclear reactor. The code uses the output parameters generated by the Monte Carlo neutronics codes to determine the isotopic inventory as a function of time and power density. Burn/breed directly uses the neutron absorption tally/reaction information generated by Monte Carlo code for each nuclide of interest to determine the nuclide inventory. When the isotope inventories have been calculated for a specific reactor operation and cooling period, the decay heat can be derived. Beside the U-235 and Pu-239 decay chains, the code includes the U-233 decay chain to manage the possible scenarios for using thorium in a nuclear fuel cycle. This paper describes the theoretical basis of this code, and shows the results of the code for two test cases. The results show good agreement with other codes for the first test case and with the experimental results for the second test case.
    Keywords: burn/breed; burnup; decay heat; nuclear reactor; Monte Carlo; isotopic inventory; wavelet; U-233 decay chain.

  • Accuracy of Batan-3DIFF and MCNP6 codes for thermal neutron flux distribution at the irradiation position of the RSG-GAS reactor   Order a copy of this article
    by Surian Pinem, Tagor Malem Sembiring 
    Abstract: This paper presents the accuracy of the neutron diffusion method and Monte Carlo method codes for determining the axial thermal neutron flux in the irradiation position of RSG-GAS reactor. This research will be used for the optimisation of the target in the reactor. The thermal neutron flux is determined by the Au foils activation method. The Au foils are inserted in the very sensitive area and very complex condition since the location is in the central of core and the foils are near to three targets, two low enrichment uranium (LEU) electroplating targets and the TeO2 target, respectively. Effects of the blackness coefficient (for control rod) and the cell model of the target on the core parameters are investigated in this research work. Two cell models of the very thin layer LEU electroplating, the homogenised and multi-zone homogenised cells, are proposed. The MCNP6 calculation results are very good agreement with the experimental results, such as excess reactivity and total control rod worth. However, the neutron diffusion method code, Batan-3DIFF, has a quite higher relative difference of 10.14%. The proposed multi-zone homogenised can improve the accuracy of criticality calculation by 124% compared with the homogenised cell. For the axial thermal neutron distribution, the MCNP gives very satisfactory results since those are within the standard deviation of experimental results. However, the Batan-3DIFF code has an average relative difference of 7%.
    Keywords: RSG-GAS reactor; Au foils; thermal neutron flux; Batan-3DIFF; MCNP6.1.

  • Calculations of current and yield of neutrons using several targets and different proton beam energies   Order a copy of this article
    by Abdessamad Didi, Hassane Dekhissi 
    Abstract: Spallation reactions provide the neutrons which are useful for fundamental research and for several desired applications. It is specifically dedicated to applied research with a neutron flux for driving the subcritical nuclear reactors, also for the transmutation of nuclear wastes. The choice of spallation target remains as a nuclear research activity that is more efficient, more adapted and more coherent to the parameters related to theory and practice through the MC simulation. In this research, we investigated cylindrical spallation targets: beryllium, tin, lead, tungsten, and uranium, which are bombarded by high-energy proton beams. In this work, we focused on the target type of accelerator-driven systems by calculating important parameters, such as yield, current, and neutron spectrum, using the MCNP code and comparing the results with theoretical and experimental results, to approve the proper application of an ADS project.
    Keywords: neutron yield; spallation; high-energy protons; accelerator-driven systems; Monte Carlo transmutation; MCNP.