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 (10 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.

  • Selection of shielding materials for gamma/X-ray and neutron radiations among the commonly used polymers   Order a copy of this article
    by H.C. Manjunatha, N. Nagaraja, L. Seenappa, K.N. Sridhar, H.B. Ramalingam 
    Abstract: We have studied the X-ray and gamma shielding parameters like mass attenuation coefficient, mean free path, half value layer, tenth value layer, effective atomic numbers, electron density, exposure buildup factors, and specific gamma ray constant in commonly used polymers such as polystyrene, polypropylene, polytetrafluoroethylene (PTFE), poly(vinyl chloride) (PVC), and polychlorotetrafluoroethylene (PCTFE). We have also measured X-ray and gamma shielding parameters at different energies such as 170Tm (0.084 MeV), 137Cs (0.662 MeV) and 60Co (1.170, 1.330 MeV) in the same polymers. The measured values agree well with the theory. Neutron shielding parameters such as coherent neutron scattering length, incoherent neutron scattering length, coherent neutron scattering cross section, incoherent neutron scattering cross section, total neutron scattering cross section and neutron absorption cross section were determined for the same polymers. PCTFE is found to be good shielding material for the gamma/X-ray and neutron radiation.
    Keywords: shielding material; gamma radiation.

  • Estimation of plutonium produced in nuclear power reactors from the electricity generated   Order a copy of this article
    by K.L. Ramakumar 
    Abstract: Estimation of burnup of nuclear fuel, amount of plutonium produced and its Pu-240 content in a typical nuclear power reactor are required for different nuclear information management purposes. Although time-consuming analytical methodologies are capable of giving accurate values (for example, mass spectrometry for post-irradiation examination), empirical computations do contribute for understanding of that information. Electricity generated in a nuclear reactor could be a good indicator for this purpose. It is shown that this empirical computation is better than taking recourse to name-plate capacity. The operating history of Calder Hall and Chapelcross nuclear power reactors in the UK was used to validate the empirical calculations. Reasonably reliable information on total quantity of plutonium, its Pu-240 content and the burnup could be estimated. Based on these empirical calculations, an attempt has been made to deduce the Pu-240 content in the nuclear weapon test carried out by the USA in 1962 using reactor grade plutonium.
    Keywords: computations; empirical calculations; plutonium production; electricity generation; burnup; Pu-240 content; Calder Hall; Chapelcross reactors; The 1962 US nuclear weapons test.

  • Neutronics design study of an advanced lead-cooled modular nuclear reactor   Order a copy of this article
    by Donny Hartanto, Safa Alhamad, Khadijah Mahmoud, Nirmin Kurdi, Muhammad Zubair 
    Abstract: A design of an advanced small modular reactor with long-life core is studied in this paper. The core is designed to produce 45 MWth power with a lifetime of 20 years without refuelling. In order to achieve a compact design and have a good neutron economy, lead is considered as the coolant owing to its excellent neutronics and thermophysical properties. However, the lead coolant speed in the core is limited to 2 m/s to minimise the corrosion and erosion of the structural materials. On the other hand, U15N is used as the fuel which reflects excellent thermophysical properties and compatibility with lead. In this study, the neutronics properties of the core including the reactivity evolution during its lifetime, the control rod worth, and the fuel and coolant reactivity feedbacks are evaluated. It was found that ALMANAR could achieve a long-life core of about 22 effective full power years with excellent inherent safety features. Monte Carlo Serpent code is used to perform the calculations in conjunction with the latest nuclear data library ENDF/B-VIII.0.
    Keywords: ALMANAR; long-life core; lead coolant; U15N; Serpent; ENDF/B-VIII.0.

  • Computational investigation of Tehran Research Reactor graphite reflector replacement with Be, BeO or D2O and its impact on thermal neutron flux enhancement   Order a copy of this article
    by Zohreh Gholamzadeh, Farrokh Khoshahval, Masoud Amin Mozafari, Atieh Joze-Vaziri 
    Abstract: Thermal neutron flux enhancement is a essential objective to achieve more radioisotope production in research reactors. Here, the effect of graphite reflector replacement with other customary reflectors on thermal neutron flux enhancement of the Tehran Research Reactor (TRR) core has been investigated. The core is modelled using Monte Carlo-based code with the graphite routine reflector and with Be, BeO and D2O reflectors. The calculations showed that the reflector replacement could result in a small enhancement of thermal neutron flux while beryllium oxide is suggested as the most efficient material (about 2.78% on average per channel). In addition, instead of total reflector conversion, application of some dimensionally optimised BeO blocks located around the irradiation boxes increases the thermal neutron flux up to 2.62%. Moreover, this study demonstrates that compacting the core could significantly enhance the thermal neutron flux up to 58%, which is noticeably more than the reflector conversion effect.
    Keywords: Tehran Research Reactor; reflector conversion; thermal neutron flux enhancement; Monte Carlo-based calculations; compact core.

  • Empirical formula for bremsstrahlung cross sections in actinides   Order a copy of this article
    by H.C. Manjunatha, L. Seenappa, K.N. Sridhar 
    Abstract: We have formulated the empirical formula for bremsstrahlung cross sections in actinides. This formula produces bremsstrahlung cross section for electrons with energies from 1 keV to 10 GeV incident on atoms with atomic numbers Z = 89 to 103. This formula produces bremsstrahlung cross sections with simple inputs of energy of incident electrons, energy of the emitted bremsstrahlung photons and the atomic number of the target.
    Keywords: bremsstrahlung cross section.

  • Radioactivity Levels of 238U, 234Th, 40K and 137C in the surface soil of selected regions from Baghdad Governorate   Order a copy of this article
    by Noor Adil, Sameera Ahmed Ebrahiem 
    Abstract: Eighteen soil samples were taken from different locations of Baghdad, the capital of Iraq, with a depth of (0-15) cm, and analysed using HPGe detector gamma-ray spectroscopy to determine the specific activity of 238U, 232Th, 40K, and 137Cs as well as the hazard indicators of gamma radiation. According to the values on UNSCEAR, the results presented that the average of specific activity for 238U and 232Th was (15.692
    Keywords: exposure; HPGe detector; hazard indicators.

  • Thermal state of ventilated storage container with spent nuclear fuel under normal operation   Order a copy of this article
    by Svitlana Alyokhina 
    Abstract: Thermal state is an important part of the safety assessment of any nuclear facility. Numerical studies of the thermal state of a container with spent nuclear fuel (SNF) are carried out. The problem was considered in conjugate quasi-static formulation as a part of the multistage approach to the simulation of the thermal processes at the dry SNF storage. Mathematical model and calculation approach to the problem solving were verified by comparison of calculated and measured temperatures of ventilated air in the outlet vents. The storage container, which is operated on Zaporizhska NPP (Ukraine), was modelled under normal operation. The structure of cooling airflow inside the container and the distribution of temperature and heat flux density on the surface of the storage cask were detected. The maximum temperatures inside the storage container at the average monthly temperatures during the year were calculated, and the temperature fluctuation is about 23 degrees. The operation of the ventilated storage containers with SNF at Zaporizhska NPP is safe. Obtained results must be used at the development of the ageing management program for the Dry Spent Nuclear Fuel Storage Facility on Zaporizhska NPP.
    Keywords: safety; spent nuclear fuel; dry storage; thermal processes; convection modeling.