Hybrid journal (It can contain Open Access articles) ISSN (Print) 1741-6361 - ISSN (Online) 1741-637X Published by Inderscience Publishers[408 journals]

Authors:S. Dawahra, G. Saba Pages: 1 - 10 Abstract: Owing to the high risk through conducting the experiment of losing the core water shield, the analytical and statistical (MCNP) methods were used to estimate the gamma dose rate at the reactor top when the core water shield level decreases to the value less than the Miniature Neutron Source Reactor (MNSR) safety limit. For normal operation condition, the results showed that the core water shield attenuated the gamma dose rate by 3 × 10<SUP align="right">6</SUP> times. The gamma dose rate values at the reactor top due to the core and the water impurities were 46.8, 89.9 and 52 μSv.h<SUP align="right">−1</SUP> for the statistical, analytical calculations and experimental reference value respectively. For the accidental condition (the core water shield level is 12 cm), the results showed that the maximum gamma dose rate value at the reactor top was 1.4 × 10<SUP align="right">−8</SUP> μSv.h<SUP align="right">−1</SUP>. Therefore, no one should stay at the reactor top if an accident occurs. These results of dose rate can be used as reference values in order to set the radiation protection requirements in the MNSR reactor. Keywords: MNSR; miniature neutron source reactors; core water shield; gamma rays; gamma dose rate; MCNP; low power research reactors; nuclear reactors; nuclear energy; nuclear power; nuclear safety; reactor top; radiation protection Citation: International Journal of Nuclear Energy Science and Technology, Vol. 10, No. 1 (2016) pp. 1 - 10 PubDate: 2016-05-04T23:20:50-05:00 DOI: 10.1504/IJNEST.2016.076341 Issue No:Vol. 10, No. 1 (2016)

Authors:Abdul Razak Kaladgi, A.D. Mohammed Samee, M.K. Ramis Pages: 11 - 27 Abstract: The objective of this work is to numerically investigate the effect of various design parameters on the steady state thermal performance characteristics of liquid sodium which flows as a coolant over a rectangular nuclear fuel element having non-uniform volumetric energy generation. Accordingly, the stream function-vorticity formulation of the full Navier Stokes equations governing flow and thermal fields in the fluid domain (coolant) are solved simultaneously along with energy equation of fuel element, using finite difference schemes. The results obtained are presented in the form of variation of mean coolant temperature at the exit, average Nusselt number and average skin friction coefficient for a wide range of parameters - aspect ratio, A<SUB align="right">r, conduction-convection parameter N<SUB align="right">cc, total energy generation parameter Q<SUB align="right">t, and flow Reynolds number Re<SUB align="right">H. It is concluded that out of the various parameters studied, the flow Reynolds number, Re<SUB align="right">H considerably influences the mean exit temperature, whereas A<SUB align="right">r has the least influence on it. Keywords: conjugate heat transfer; Thomas algorithm; ADI scheme; forced convection; thermal performance; liquid sodium; coolants; rectangular fuel elements; nuclear fuel elements; finite difference method; mean exit temperature; nuclear energy; nuclear power Citation: International Journal of Nuclear Energy Science and Technology, Vol. 10, No. 1 (2016) pp. 11 - 27 PubDate: 2016-05-04T23:20:50-05:00 DOI: 10.1504/IJNEST.2016.076347 Issue No:Vol. 10, No. 1 (2016)

Authors:Rasoul Ghasemi, Gholam Reza Ansarifar Pages: 28 - 58 Abstract: In this paper, an adaptive sliding mode control system is presented to control the Pressurised-Water Nuclear Reactor (PWR) core power. Sliding mode controller is a robust non-linear control but chattering in this method is undesirable; large parameter uncertainties accommodation envisaged by designing adaptive mechanisms for the controller and high chattering authority are inhibited. The reactor core is simulated based on the point kinetics equations and one delayed neutrons group. The stability analysis is given by means of the Lyapunov approach, thus the control system is guaranteed to be stable within a large range. The adaptive sliding mode control exhibits the desired dynamic properties during the entire output-tracking process independent of the perturbations. Keywords: pressurised water reactors; PWR; nuclear reactors; adaptive SMC; sliding mode control; point kinetics equations; xenon concentration; adaptation laws; Lyapunov approach; nonlinear control; reactor power; online parameter adaptation; nuclear energy; n Citation: International Journal of Nuclear Energy Science and Technology, Vol. 10, No. 1 (2016) pp. 28 - 58 PubDate: 2016-05-04T23:20:50-05:00 DOI: 10.1504/IJNEST.2016.076353 Issue No:Vol. 10, No. 1 (2016)

Pages: 59 - 71 Abstract: In this work, we introduce a new Metropolis algorithm, which is an enhancement of the recent Particle Collision Algorithm (PCA), loosely inspired by the neutron interactions in a reactor. This novel method is called the Cross-Section Particle Collision Algorithm (CSPCA), as it incorporates the concept of cross-section from Reactor Physics, in the sense that points in the search space and their respective fitness-function values are analogous to the neutron cross-sections which are used to express the likelihood of interaction between an incident neutron and a target nucleus. CSPCA is compared against the original PCA and two state-of-the-art metaheuristics, differential evolution and big bang-big crunch. These methods are applied to the turbine balancing problem, which is an NP-hard (i.e. non-deterministic polynomial-time hard) combinatorial optimisation problem that can be used to assess the potential of an algorithm to be applied to Fuel Management Optimisation (FMO). CSPCA performs better than its opponents, showing potential to be used not only in FMO, but also in other nuclear science and engineering optimisation problems. Keywords: Metropolis algorithms; particle collision algorithm; combinatorial optimisation; random keys; fuel management optimisation; cross-section; reactor physics; metaheuristics; differential evolution; big bang-big crunch; turbine balancing; nuclea Citation: International Journal of Nuclear Energy Science and Technology, Vol. 10, No. 1 (2016) pp. 59 - 71 PubDate: 2016-05-04T23:20:50-05:00 DOI: 10.1504/IJNEST.2016.076354 Issue No:Vol. 10, No. 1 (2016)

Pages: 72 - 87 Abstract: The use of TRISO fuel is a main feature of the high temperature reactors. In these reactors, the heterogeneity in the internal composition of the TRISO particle and the random arrangement of coated fuel particles inside graphite matrix creates a significant challenge in the modelling tasks. To simulate spherical elements using MCNPX code repetitive structures are usually created, generating a uniform distribution of the coated fuel particles. The use of these repetitive structures introduces two major limitations: the no randomness of the TRISO particles inside the pebbles and the intersection of the pebble surface with the particles. These limitations could affect significantly the multiplicative properties of the core. In order to study the influence of these approximations in the multiplicative properties of pebble-bed reactor cores, four different configurations for the TRISO particles distribution inside pebble were analysed. On pebble scale, the results showed significant differences between the models where the cutting effect is present and the model that treats the cutting effect. However, on core scale there are no differences among the results of the HTR-10 critical height for the studied models. Keywords: TRISO particles; double heterogeneity; HTR-10; MCNPX; VHTR; modelling; coated fuel particles; particle distribution; pebble bed reactors; high temperature reactors; nuclear reactors; nuclear energy; nuclear power; simulation; rector core Citation: International Journal of Nuclear Energy Science and Technology, Vol. 10, No. 1 (2016) pp. 72 - 87 PubDate: 2016-05-04T23:20:50-05:00 DOI: 10.1504/IJNEST.2016.076357 Issue No:Vol. 10, No. 1 (2016)