International Journal of Nuclear Energy Science and Technology
These articles have been peer-reviewed and accepted for publication but are pending final changes, are not yet published and may not appear here in their final order of publication until they are assigned to issues. Therefore, the content conforms to our standards but the presentation (e.g. typesetting and proof-reading) is not necessarily up to the Inderscience standard. Additionally, titles, authors, abstracts and keywords may change before publication. Articles will not be published until the final proofs are validated by their authors.
Forthcoming articles must be purchased for the purposes of research, teaching and private study only. These articles can be cited using the expression "in press". For example: Smith, J. (in press). Article Title. Journal Title.
Articles marked with this shopping trolley icon are available for purchase - click on the icon to send an email request to purchase.
International Journal of Nuclear Energy Science and Technology (5 papers in press)
Burn/breed: a wavelet-based nuclear fuel burnup and decay heat code 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.
Differential probabilistic space-temporal model for real-time power prognosis in failures in a nuclear reactor by Alejandro Nuñez-Carrera Abstract: The aim of this paper is the neutronic flux prognosis in a nuclear reactor for faults in the measurement of local power reactor monitors (LPRMs) in real time using a differential probabilistic space-temporal model (DPSTM). The LPRMs do provide inputs to the average power range monitor (APRM). The LPRM house a fission chamber and their associated signal cables. The failure of one or more chains of LPRMs is common during the operational cycle. The circuit average only LPRM signals that are operational and the output from the averaging circuit for each APRM channel is the route to the process computer. The DPSTM allows a reliable reconstruction of the real time signals of those LPRMs that are out of order. The DPSTM is evaluated in terms of predictive accuracy for different time horizons and compared to a time series. The DPSTM based prognosis methodology was developed and validated with real signals of Ringhals stability benchmarks. Keywords: BWR; LPRM; APRM; Ringhals NPP; Bayesian network; neutron flux; spatial-temporal model; Markov random; prognosis process; real time.
Proton-induced spallation of mercury and xenon targets for neutron flux produced between CEM03.01 and INCL4-ABLA spallation packages by Abdessamad Didi, Hassane Dekhissi Abstract: Accelerator-driven subcritical reactors (ADS) are one of the pathways considered for the incineration of minor actinides. They use a proton beam accelerator up to a GeV energy, bombarding a spallation target, usually made of a material of high atomic number (W, Pb, Ta or U). The neutrons generated during the spallation reactions are then multiplied. The main goal of this study is to make a comparison between two models of physical packages via the Monte Carlo code MCNP6. The first package is the INCL4-ABLA and the second one is the CEM03.01. We have deduced the best package according to the neutron production rate. In this study, we are interested in two cylindrical targets of natural mercury and xenon, 20 cm in height and 20 cm in diameter, and a proton beam varying from 0.1 to 1.5 GeV. Keywords: spallation; neutron flux; yield neutron; accelerator; high energy; MCNP.
Enhanced calculations of fusion barrier distribution for heavy-ion fusion reactions using the Wong Ffrmula by Fouad A. Majeed, Fatima M. Hussain, Yousif A. Abdul-Hussien Abstract: The effect of coupled channels in heavy ion fusion reaction for the systems 40Ca + 192Os, 40Ca + 194Pt, and 48Ca + 197Au are discussed. The fusion cross-section, σ_fus, the fusion barrier distribution, D_fus, and fusion probability P_fus are investigated. The fusion barrier distribution is calculated using numerical three point, five point and Wong methods. Full quantum coupled channels calculations are performed using CCFULL code with all order coupling to compare with available experimental data. The χ^2 values for the cases of no coupling and coupling effects included shows clearly that the present calculations are in good agreement with the experimental data. Keywords: Wong formula; coupled channel; CCFULL code; heavy ion fusion.
Small modular reactor core neutronic evaluation via Monte Carlo method by Reza Akbari, Dariush Rezaei Ochbelagh, Ahmad Gharib Abstract: The Small Modular Reactors (SMRs) have been getting a lot of positive attention recently, because of their features such as lower initial costs, better safety features than older power reactors, district heating, co-generation, energy storage, desalination, and hydrogen production. Most of the SMRs are in their different design stages, so demonstrating the differences and similarities of the different aspects (economic, safety, neutronic, etc.) of these new integral reactors in comparison to the conventional large reactors is valuable and useful in the design procedure. The main purpose of this study is an evaluation of the SMART reactor core (an SMR with certified design) neutronic parameters via the Monte Carlo method using the MCNPX code. The SMART neutronic parameters, such as axial and radial distributions of neutron fluxes, Power Peaking Factors (PPFs), effective delayed neutron fraction, xenon and samarium effects, burnup calculation and neutron flux energy spectrum, have been assessed. Daily load following operation in soluble-boron-free with control regulating banks is one of the best SMART reactor core advantages. Accordingly, the effects of main regulating bank insertion in SMART core have been evaluated. For developed model verification, some of the neutronic parameters have been compared with the SMART Standard Safety Analysis (SSAR) and show proper match. Then other neutronic parameters of the SMART core as a pioneer SMR have been calculated and evaluated. Keywords: small modular reactor; SMART; neutronic assessment; Monte Carlo method.