International Journal of Nuclear Energy The latest articles from Hindawi Publishing Corporation © 2014 , Hindawi Publishing Corporation . All rights reserved. Experimental and Theoretical Investigation of Three Alloy 690 Mockup Components: Base Metal and Welding Induced Changes Sun, 03 Aug 2014 08:50:27 +0000 The stress corrosion cracking (SCC) resistance of cold deformed thermally treated (TT) Alloy 690 has been questioned in recent years. As a step towards understanding its relevancy for weld deformed Alloy 690 in operating plants, Alloy 690 base metal and heat affected zone (HAZ) microstructures of three mockup components have been studied. All mockups were manufactured using commercial heats and welding procedures in order to attain results relevant to the materials in the field. Thermodynamic calculations were performed to add confidence in phase identification as well as understanding of the evolution of the microstructure with temperature. Ti(C,N) banding was found in all materials. Bands with few large Ti(C,N) precipitates had negligible effect on the microstructure, whereas bands consisting of numerous small precipitates were associated with locally finer grains and coarser grain boundary carbides. The Ti(C,N) remained unaffected in the HAZ while the carbides were fully dissolved close to the fusion line. Cold deformed solution annealed Alloy 690 is believed to be a better representation of this region than cold deformed TT Alloy 690. Rickard R. Shen, Bartek Kaplan, and Pål Efsing Copyright © 2014 Rickard R. Shen et al. All rights reserved. Analysis of Loss of Flow Events on Brazilian Multipurpose Reactor Using the Relap5 Code Tue, 03 Jun 2014 08:26:00 +0000 This work presents the thermal hydraulic simulation of the Brazilian multipurpose reactor (RMB) using a RELAP5/MOD3.3 model. Beyond steady state calculations, three transient cases of loss of flow accident (LOFA) in the primary cooling system have been simulated. The RELAP5 simulations demonstrate that after all initiating events, the reactor reaches a safe new steady state keeping the integrity and safety of the core. Moreover, a sensitivity study was performed to verify the nodalization behavior due to the variation of the thermal hydraulic channels in the reactor core. Transient calculations demonstrate that both nodalizations follow approximately the same behavior. Humberto V. Soares, Ivan D. Aronne, Antonella L. Costa, Claubia Pereira, and Maria Auxiliadora F. Veloso Copyright © 2014 Humberto V. Soares et al. All rights reserved. The Impact of Climate Changes on the Thermal Performance of a Proposed Pressurized Water Reactor: Nuclear-Power Plant Thu, 10 Apr 2014 08:33:05 +0000 This paper presents a methodology for studying the impact of the cooling water temperature on the thermal performance of a proposed pressurized water reactor nuclear power plant (PWR NPP) through the thermodynamic analysis based on the thermodynamic laws to gain some new aspects into the plant performance. The main findings of this study are that an increase of one degree Celsius in temperature of the coolant extracted from environment is forecasted to decrease by 0.39293 and 0.16% in the power output and the thermal efficiency of the nuclear-power plant considered, respectively. Said M. A. Ibrahim, Mohamed M. A. Ibrahim, and Sami. I. Attia Copyright © 2014 Said M. A. Ibrahim et al. All rights reserved. Assessment of Tritium Activity in Groundwater at the Nuclear Objects Sites in Lithuania Wed, 12 Mar 2014 11:49:55 +0000 The assessment of nuclear objects sites in Lithuania, including groundwater characterization, took place in the last few years. Tritium activity in groundwater is a very useful tool for determining how groundwater systems function. Natural and artificial tritium was measured in 8 wells in different layers (from 1.5 to 15 meters depth). The results were compared with other regions of Lithuania also. The evaluated tritium activities varied from 1.8 to 6.4 Bq/L at nuclear objects sites in Lithuania and they are coming to background level (1.83 Bq/L) and other places in Lithuania. The data show, that groundwater at the nuclear power objects sites is not contaminated with artificial tritium. In this work, the vertical tritium transfer from soil water to the groundwater well at nuclear objects site was estimated. The data show that the main factor for vertical tritium transfer to the well depends on the depth of wells. Vigilija Cidzikienė, Vaidotė Jakimavičiūtė-Maselienė, Raselė Girgždienė, Jonas Mažeika, and Rimantas Petrošius Copyright © 2014 Vigilija Cidzikienė et al. All rights reserved. Oxidation/Corrosion Behaviour of ODS Ferritic/Martensitic Steels in Pb Melt at Elevated Temperature Mon, 03 Mar 2014 16:34:50 +0000 Lead-based melts (Pb, Pb-Bi) are considered as candidate coolants and spallation neutron targets due to their excellent thermophysical and nuclear properties. However, the corrosion of structural materials remains a major issue. Oxide dispersion strengthened (ODS) ferritic/martensitic steels are considered for high temperature application for both fission and fusion reactor concepts. The oxidation/corrosion kinetics in a static oxygen-saturated Pb melt at temperature of 550°C as well as the morphology and composition of scales formed on ferritic/martensitic Fe-9Cr-1.5W and ferritic Fe-14Cr-1.5W ODS steels have been investigated. Both materials showed homogeneous multiple, dense scales that consisted of typical combination of Fe3O4 as outer sublayer and (Fe,Cr)3O4 as inner sublayer. A nonuniform growth of inner oxide sublayers into the metal matrix as well as a good adhesion to the metal substrate is observed. With the prolongation of exposure from 240 to 1000 h, observed scales grow from 35 µm to 45 µm for ODS Fe-9Cr steel and from 40 µm to 60 µm for ODS Fe-14Cr steel with the thinning rates of 0,22 and 0,31 mm/year correspondingly. The mechanism of scales formation is discussed. O. I. Yaskiv and V. M. Fedirko Copyright © 2014 O. I. Yaskiv and V. M. Fedirko. All rights reserved. A Small-Sized HTGR System Design for Multiple Heat Applications for Developing Countries Thu, 26 Dec 2013 19:21:56 +0000 Japan Atomic Energy Agency has conducted a conceptual design of a 50 MWt small-sized high temperature gas cooled reactor (HTGR) for multiple heat applications, named HTR50S, with the reactor outlet coolant temperature of 750°C and 900°C. It is first-of-a-kind of the commercial plant or a demonstration plant of a small-sized HTGR system to be deployed in developing countries in the 2020s. The design concept of HTR50S is to satisfy the user requirements for multipurpose heat applications such as the district heating and process heat supply based on the steam turbine system and the demonstration of the power generation by helium gas turbine and the hydrogen production using the water splitting iodine-sulfur process, to upgrade its performance compared to that of HTTR without significant R&D utilizing the knowledge obtained by the HTTR design and operation, and to fulfill the high level of safety by utilizing the inherent features of HTGR and a passive decay heat removal system. The evaluation of technical feasibility shows that all design targets were satisfied by the design of each system and the preliminary safety analysis. This paper describes the conceptual design and the preliminary safety analysis of HTR50S. Hirofumi Ohashi, Hiroyuki Sato, Minoru Goto, Xing Yan, Junya Sumita, Yujiro Tazawa, Yasunobu Nomoto, Jun Aihara, Yoshitomo Inaba, Yuji Fukaya, Hiroki Noguchi, Yoshiyuki Imai, and Yukio Tachibana Copyright © 2013 Hirofumi Ohashi et al. All rights reserved. Study of Thorium Fuel Cycles for Light Water Reactor VBER-150 Mon, 23 Dec 2013 09:41:59 +0000 The main objective of this paper is to examine the use of thorium-based fuel cycle for the transportable reactors or transportable nuclear power plants (TNPP) VBER-150 concept, in particular the neutronic behavior. The thorium-based fuel cycles included Th232+Pu239, Th232+U233, and Th232+U and the standard design fuel UOX. Parameters related to the neutronic behavior such as burnup, nuclear fuel breeding, MA stockpile, and Pu isotopes production (among others) were used to compare the fuel cycles. The Pu transmutation rate and accumulation of Pu with MA in the spent fuel were compared mutually and with an UOX open cycle. The Th232+U233 fuel cycle proved to be the best cycle for minimizing the production of Pu and MA. The neutronic calculations have been performed with the well-known MCNPX computational code, which was verified for this type of fuel performing calculation of the IAEA benchmark announced by IAEA-TECDOC-1349. Daniel Evelio Milian Lorenzo, Daniel Milian Pérez, Lorena Pilar Rodríguez García, Jesús Salomón Llanes, Carlos Alberto Brayner de Oliveira Lira, Manuel Cadavid Rodríguez, and Carlos Rafael García Hernández Copyright © 2013 Daniel Evelio Milian Lorenzo et al. All rights reserved. Economic Assessment of Russian Nuclear Strategies on the Basis of Fast Breeder Reactors Wed, 09 Oct 2013 18:17:07 +0000 The paper assesses the economic risk caused by the delay in commissioning innovative nuclear power plants with fast breeder reactors in Russia. The risk is quantitatively measured by the excessive costs for energy development and the possibility of implementing the considered variants that differ in power consumption, technical and economic indices of the reactors, and constraints on CO2 emissions. The probability distribution functions of economic losses for different strategies of nuclear energy development are constructed. O. V. Marchenko and S. V. Solomin Copyright © 2013 O. V. Marchenko and S. V. Solomin. All rights reserved. Generalized and Stability Rational Functions for Dynamic Systems of Reactor Kinetics Tue, 13 Aug 2013 13:04:33 +0000 The base of reactor kinetics dynamic systems is a set of coupled stiff ordinary differential equations known as the point reactor kinetics equations. These equations which express the time dependence of the neutron density and the decay of the delayed neutron precursors within a reactor are first order nonlinear and essentially describe the change in neutron density within the reactor due to a change in reactivity. Outstanding the particular structure of the point kinetic matrix, a semianalytical inversion is performed and generalized for each elementary step resulting eventually in substantial time saving. Also, the factorization techniques based on using temporarily the complex plane with the analytical inversion is applied. The theory is of general validity and involves no approximations. In addition, the stability of rational function approximations is discussed and applied to the solution of the point kinetics equations of nuclear reactor with different types of reactivity. From the results of various benchmark tests with different types of reactivity insertions, the developed generalized Padé approximation (GPA) method shows high accuracy, high efficiency, and stable character of the solution. Ahmed E. Aboanber Copyright © 2013 Ahmed E. Aboanber. All rights reserved. Experimental Measurements of Drop Size Distributions in 30 mm Diameter Annular Centrifugal Contactor with 30% TBP-Nitric Acid Biphasic System Sun, 23 Jun 2013 11:12:44 +0000 For design and development of liquid-liquid extraction systems, it is essential to have an accurate estimation of hydrodynamic and mass transfer characteristics of the employed contactor. In the present study, experimental evaluations consisted primarily of determining the maximum solution throughput that could be processed without cross-phase contamination at a given rotor speed, O/A flow ratio, and organic-aqueous solution pair in a 30 mm bowl diameter centrifugal contactor. In addition, analysis included experimental drop size determinations as well as holdup determination. The experimental drop size distributions are expected to be helpful for modeling work. Shekhar Kumar and U. Kamachi Mudali Copyright © 2013 Shekhar Kumar and U. Kamachi Mudali. All rights reserved. Detection of the Departure from Nucleate Boiling in Nuclear Fuel Rod Simulators Sun, 02 Jun 2013 15:27:55 +0000 In the thermal hydraulic experiments to determin parameters of heat transfer where fuel rod simulators are heated by electric current, the preservation of the simulators is essential when the heat flux goes to the critical point. One of the most important limits in the design of cooling water reactors is the condition in which the heat transfer coefficient by boiling in the core deteriorates itself. The heat flux just before deterioration is denominated critical heat flux (CHF). At this time, the small increase in heat flux or in the refrigerant inlet temperature at the core, or the small decrease in the inlet flux of cooling, results in changes in the heat transfer mechanism. This causes increases in the surface temperature of the fuel elements causing failures at the fuel (burnout). This paper describes the experiments conducted to detect critical heat flux in nuclear fuel element simulators carried out in the thermal-hydraulic laboratory of Nuclear Technology Development Centre (CDTN). It is concluded that the use of displacement transducer is the most efficient technique for detecting critical heat flux in nuclear simulators heated by electric current in open pool. Amir Zacarias Mesquita and Rogério Rivail Rodrigues Copyright © 2013 Amir Zacarias Mesquita and Rogério Rivail Rodrigues. All rights reserved.