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 (6 papers in press)

Regular Issues

  • Design and test of an update of TRAC-BF1 for the modelling of supercritical water reactors   Order a copy of this article
    by Emilio Martinez-Camacho, Jaime Baltazar Morales-Sandoval, J. Manuel Gallardo-Villarreal, Raymundo A. Sanchez-Salazar 
    Abstract: Some designs of nuclear power plants are bringing the working fluid to supercritical conditions (SCW) to improve their efficiency. But the evaluation of nuclear plants response to transient events or postulated accidents is complex. In the academic environment, analysis is even more complex since the codes are generally restricted access. The possibility of improving routines and interfaces with the TRAC-BF1 code, to allow the modelling of complete plants that use SCW during operational events or postulated accidents, was established as the main objective of this investigation. The new tools allow the analysis of models with steam cycles including supercritical regions such as the high-performance light water reactor, in which core nominal conditions were modeled for verification purposes. Also, a benchmark with experimental data was performed as part of the tests for new models. Results showed that our code modification is capable of modelling through the critical point without defining a pseudo-two-phase region.
    Keywords: supercritical water; TRAC-BF1; SCWR; best estimate; modelling; educational tool.

  • VVER-440 fresh core depletion benchmark analysis using MCNPX   Order a copy of this article
    by Sardar Muhammad Shauddin 
    Abstract: The paper presents outcomes obtained by reproducing a VVER-440 lattice physics benchmark using the MCNPX code. The benchmark data are related to the fresh core depletion calculations of the Kozloduy nuclear power plant unit-2. A whole study is performed using different zones (including assembly burnup) and the different numbers of radial and axial regions within the whole core. Neutronics as well as thermal-hydraulics core design basis limits for the first-cycle are studied. The infinite multiplication factor behaviour of a fuel assembly with or without the presence of soluble boron in the moderator/coolant at different full power days (FPD) is studied. Assembly-wise normalised radial power distribution (RPD) at different FPD is performed and compared with the measured data. Core critical boron concentration (CBC) and average axial power distribution (APD) for different FPD are calculated and discussed. In the CBC calculation, the MCNPX results were found to stay closer to the measured value up to 51.3 FPD. In the RPD and APD calculations, the MCNPX results were found to lie within the 10% deviation from the measured and benchmark average values, respectively.
    Keywords: depletion calculation; radial power distribution; axial power distribution; critical boron concentration; VVER-440; water water energetic reactor model 440; MCNPX.

  • Relativistic fictitious forces in uniformly rotating frames   Order a copy of this article
    by Adrian Sfarti 
    Abstract: In the current paper we present a generalisation of the transforms from the frame co-moving with an accelerated particle for uniform circular motion into an inertial frame of reference. The solution is of great interest for real time applications because earth-bound laboratories are inertial only in approximation. The motivation is that the real life applications include accelerating and rotating frames with arbitrary orientations more often than the idealised case of inertial frames; our daily experiments happen in the laboratories attached to the rotating Earth. Our paper is divided into three main sections; the first section deals with the theory of the dynamics, i.e. forces, the second section deals with the application of the theory to the derivation of the relativistic fictitious forces (Coriolis, centrifugal and Euler) occurring in the rotating frame. The third section deals with the relativistic fictitious forces in the quasi-inertial frame of the lab. We will show that there is not only a fictitious force that emerges in the rotating frame but also a fictitious power. The present paper is of interest to researchers working with any type of circular particle accelerator.
    Keywords: general coordinate transformations; uniform rotation; four-vector; relativistic Coriolis force; relativistic Coriolis power; relativistic centrifugal force; relativistic centrifugal power; relativistic Euler force.

  • Study of radiolytic gas bubbles formation and behaviour in an aqueous uranyl sulfate solution using ImageJ   Order a copy of this article
    by Daniel Milian Pérez, Daylen Milian Pérez, Liván Hernández Pardo, Daniel E. Milian Lorenzo, Abel Gámez Rodríguez, Carlos A. Brayner De Oliveira Lira, Antônio Celso Dantas Antonino 
    Abstract: The 99Mo global supply shortages that started in 2009-2010, as well as the advantages of the use of aqueous solutions of uranium salts for the production of 99Mo, has prompted the assessment of new alternatives based on aqueous homogeneous systems for this purpose. However, the use of aqueous solutions of uranium salts poses a variety of challenges for the large scale production of 99Mo. Perhaps the most significant is the radiolytic decomposition of water molecules to form microbubbles. The formation of these radiolytic gas bubbles plays an important role in the neutron and thermohydraulic performance of these systems. This paper delves into the study of the formation of radiolytic gas bubbles in uranium salt solutions using the ImageJ software for the determination of parameters of interest of the gas bubbles, such as their diameter and velocity and the local gas volume fraction for three electron beam power levels, 6, 12 and 15 kW. The results obtained for these parameters were consistent with those reported by other authors in the literature. It was determined that the radiolytic gas bubbles average diameter varies between 200 and 300 m, increasing with the increase in power. A similar behaviour was observed for the bubble velocity, which increases accordingly with the increase in power density. The study of the gas volume fraction indicated that although the radiolytic gas bubble size does not increase significantly with increasing power, the number of bubbles and amount of gas produced increase appreciably.
    Keywords: aqueous solution; uranium salts; aqueous homogeneous systems; radioisotope production; ImageJ.

  • Numerical and experimental estimation of creep-fatigue life of Inconel 625 bellows at 570   Order a copy of this article
    by S. C. S. P. Kumar Krovvidi 
    Abstract: The methodology for creep-fatigue design of bellows is not covered in design codes such as Expansion Joint Manufacturers Association (EJMA). This paper presents the analysis and experimental validation of the creep-fatigue design of Inconel-625 bellows at 570
    Keywords: bellows; creep-fatigue interaction; RCC-MR; Chaboche and Rousselier visco-plastic constitutive model; Inconel-625; high temperature testing.

  • Investigation of thermal-hydraulic transient analysis of hot fuel rod in the pump failure accident   Order a copy of this article
    by Mohammad Bagher Sadeghiazad, Farzad Choobdar Rahim 
    Abstract: This study describes the thermodynamic analysis of the Bushehr water-water energetic reactor subchannels of the reactors core fuel assemblies in Bushehr Nuclear Power Plant. In the transient state, the modified COBRA-EN code was used for the failure of one of the cooling pumps. Among the channels in the reactors core fuel assemblies that are used to cool the core, there is a channel that has a maximum power factor that is known as the hottest channel. It is very important to thermodynamically investigate this channel to determine the cooling features of the reactors core. Finally, the numerical modelling results were compared with the reported safety results of Bushehr plants reactor in final safety analysis reports.
    Keywords: thermal-hydraulics; COBRA-EN code; fuel assemblies; subchannel method; transient state; VVER-1000 reactor.