International Journal of Nuclear Energy The latest articles from Hindawi Publishing Corporation © 2016 , Hindawi Publishing Corporation . All rights reserved. Probabilistic Structural Integrity Analysis of Boiling Water Reactor Pressure Vessel under Low Temperature Overpressure Event Wed, 25 Nov 2015 09:10:24 +0000 The probabilistic structural integrity of a Taiwan domestic boiling water reactor pressure vessel has been evaluated by the probabilistic fracture mechanics analysis. First, the analysis model was built for the beltline region of the reactor pressure vessel considering the plant specific data. Meanwhile, the flaw models which comprehensively simulate all kinds of preexisting flaws along the vessel wall were employed here. The low temperature overpressure transient which has been concluded to be the severest accident for a boiling water reactor pressure vessel was considered as the loading condition. It is indicated that the fracture mostly happens near the fusion-line area of axial welds but with negligible failure risk. The calculated results indicate that the domestic reactor pressure vessel has sufficient structural integrity until doubling of the present end-of-license operation. Hsoung-Wei Chou and Chin-Cheng Huang Copyright © 2015 Hsoung-Wei Chou and Chin-Cheng Huang. All rights reserved. Simulation of the Westinghouse AP1000 Response to SBLOCA Using RELAP/SCDAPSIM Tue, 16 Dec 2014 07:54:30 +0000 Great interest is given now to advanced nuclear reactors especially those using passive safety components. The Westinghouse AP1000 Advanced Passive pressurized water reactor (PWR) is an 1117 MWe PWR designed to achieve a high safety and performance record. The AP1000 safety system uses natural driving forces, such as pressurized gas, gravity flow, natural circulation flow, and convection. In this paper, the safety performance of the AP1000 during a small break loss of coolant accident (SBLOCA) is investigated. This was done by modelling the AP1000 and the passive safety systems employed using RELAP/SCDAPSIM code. RELAP/SCDAPSIM is designed to describe the overall reactor coolant system (RCS) thermal hydraulic response and core behaviour under normal operating conditions or under design basis or severe accident conditions. Passive safety components in the AP1000 showed a clear improvement in accident mitigation. It was found that RELAP/SCDAPSIM is capable of modelling a LOCA in an AP1000 and it enables the investigation of each safety system component response separately during the accident. The model is also capable of simulating natural circulation and other relevant phenomena. The results of the model were compared to that of the NOTRUMP code and found to be in a good agreement. Ayah Elshahat, Timothy Abram, Judith Hohorst, and Chris Allison Copyright © 2014 Ayah Elshahat et al. All rights reserved. Modeling of SPERT IV Reactivity Initiated Transient Tests in EUREKA-2/RR Code Tue, 09 Dec 2014 00:10:07 +0000 EUREKA-2/RR code has been used for SPERT IV reactor benchmark calculations against the experimental results provided by IAEA (International Atomic Energy Agency) obtained for a series of transient tests initiated by step insertion of different magnitudes of positive reactivity with varying degrees of different controlled parameters such as reactor initial power, coolant temperature and coolant flow condition. 20 out of 39 tests that fall under forced convection mode have been considered for the present simulation provided the reactor scram system is disabled. Peak power and peak clad temperature due to transient have been calculated and it was found that although peak clad temperature values agreed, the peak power values seem to underestimate the experimental values. Further study appears to be needed to identify the limitations in modeling or examining the effect of input parameters during modeling to obtain the better simulation results. N. H. Badrun, M. H. Altaf, M. A. Motalab, M. S. Mahmood, and M. J. H. Khan Copyright © 2014 N. H. Badrun et al. All rights reserved. Molecular Dynamics Study of Hydrogen in α-Zirconium Sun, 09 Nov 2014 00:00:00 +0000 Molecular dynamics approach is used to simulate hydrogen (H) diffusion in zirconium. Zirconium alloys are used in fuel channels of many nuclear reactors. Previously developed embedded atom method (EAM) and modified embedded atom method (MEAM) are tested and a good agreement with experimental data for lattice parameters, cohesive energy, and mechanical properties is obtained. Both EAM and MEAM are used to calculate hydrogen diffusion in zirconium. At higher temperatures and in the presence of hydrogen, MEAM calculation predicts an unstable zirconium structure and low diffusion coefficients. Mean square displacement (MSD) of hydrogen in bulk zirconium is calculated at a temperature range of 500–1200 K with diffusion coefficient at 500 K equals 1.92 10−7 cm2/sec and at 1200 K has a value 1.47 10−4 cm2/sec. Activation energy of hydrogen diffusion calculated using Arrhenius plot was found to be 11.3 kcal/mol which is in agreement with published experimental results. Hydrogen diffusion is the highest along basal planes of hexagonal close packed zirconium. Ravi Kiran Siripurapu, Barbara Szpunar, and Jerzy A. Szpunar Copyright © 2014 Ravi Kiran Siripurapu et al. All rights reserved. Improving Nuclear Safety of Fast Reactors by Slowing Down Fission Chain Reaction Thu, 16 Oct 2014 13:01:12 +0000 Light materials with small atomic mass (light or heavy water, graphite, and so on) are usually used as a neutron reflector and moderator. The present paper proposes using a new, heavy element as neutron moderator and reflector, namely, “radiogenic lead” with dominant content of isotope 208Pb. Radiogenic lead is a stable natural lead. This isotope is characterized by extremely low micro cross-section of radiative neutron capture (~0.23 mb) for thermal neutrons, which is smaller than graphite and deuterium cross-sections. The reflector-converter for a fast reactor core is the structure capable of transforming some part of prompt neutrons leaked from the core into the reflected neutrons with properties similar to those of delayed neutrons, that is, sufficiently large contribution to reactivity at the level of effective fraction of delayed neutrons and relatively long lifetime, comparable with lifetimes of radionuclides-emitters of delayed neutrons. It is evaluated that the use of radiogenic lead makes it possible to slow down the chain fission reaction on prompt neutrons in the fast reactor. This can improve the fast reactor safety and reduce some requirements to the technologies used to fabricate fuel for the fast reactor. G. G. Kulikov, A. N. Shmelev, and V. A. Apse Copyright © 2014 G. G. Kulikov et al. All rights reserved. Predicting and Preventing Flow Accelerated Corrosion in Nuclear Power Plant Mon, 13 Oct 2014 07:05:57 +0000 Flow accelerated corrosion (FAC) of carbon steels in water has been a concern in nuclear power production for over 40 years. Many theoretical models or empirical approaches have been developed to predict the possible occurrence, position, and rate of FAC. There are a number of parameters, which need to be incorporated into any model. Firstly there is a measure defining the hydrodynamic severity of the flow; this is usually the mass transfer rate. The development of roughness due to FAC and its effect on mass transfer need to be considered. Then most critically there is the derived or assumed functional relationship between the chosen hydrodynamic parameter and the rate of FAC. Environmental parameters that are required, at the relevant temperature and pH, are the solubility of magnetite and the diffusion coefficient of the relevant iron species. The chromium content of the steel is the most important material factor. Bryan Poulson Copyright © 2014 Bryan Poulson. All rights reserved. Thermal Analysis of ZPPR High Pu Content Stored Fuel Wed, 17 Sep 2014 11:36:28 +0000 The Zero Power Physics Reactor (ZPPR) operated from April 18, 1969, until 1990. ZPPR operated at low power for testing nuclear reactor designs. This paper examines the temperature of Pu content ZPPR fuel while it is in storage. Heat is generated in the fuel due to Pu and Am decay and is a concern for possible cladding damage. Damage to the cladding could lead to fuel hydriding and oxidizing. A series of computer simulations were made to determine the range of temperatures potentially occuring in the ZPPR fuel. The maximum calculated fuel temperature is 292°C (558°F). Conservative assumptions in the model intentionally overestimate temperatures. The stored fuel temperatures are dependent on the distribution of fuel in the surrounding storage compartments, the heat generation rate of the fuel, and the orientation of fuel. Direct fuel temperatures could not be measured but storage bin doors, storage sleeve doors, and storage canister temperatures were measured. Comparison of these three temperatures to the calculations indicates that the temperatures calculated with conservative assumptions are, as expected, higher than the actual temperatures. The maximum calculated fuel temperature with the most conservative assumptions is significantly below the fuel failure criterion of 600°C (1,112°F). Charles W. Solbrig, Chad L. Pope, and Jason P. Andrus Copyright © 2014 Charles W. Solbrig et al. All rights reserved. Computational Model for the Neutronic Simulation of Pebble Bed Reactor’s Core Using MCNPX Wed, 17 Sep 2014 06:17:39 +0000 Very high temperature reactor (VHTR) designs offer promising performance characteristics; they can provide sustainable energy, improved proliferation resistance, inherent safety, and high temperature heat supply. These designs also promise operation to high burnup and large margins to fuel failure with excellent fission product retention via the TRISO fuel design. The pebble bed reactor (PBR) is a design of gas cooled high temperature reactor, candidate for Generation IV of Nuclear Energy Systems. This paper describes the features of a detailed geometric computational model for PBR whole core analysis using the MCNPX code. The validation of the model was carried out using the HTR-10 benchmark. Results were compared with experimental data and calculations of other authors. In addition, sensitivity analysis of several parameters that could have influenced the results and the accuracy of model was made. J. Rosales, A. Muñoz, C. García, L. García, C. Brayner, J. Pérez, and A. Abánades Copyright © 2014 J. Rosales et al. All rights reserved. Sensitivity Analysis of the TRIGA IPR-R1 Reactor Models Using the MCNP Code Tue, 09 Sep 2014 09:08:14 +0000 In the process of verification and validation of code modelling, the sensitivity analysis including systematic variations in code input variables must be used to help identifying the relevant parameters necessary for a determined type of analysis. The aim of this work is to identify how much the code results are affected by two different types of the TRIGA IPR-R1 reactor modelling processes performed using the MCNP (Monte Carlo N-Particle Transport) code. The sensitivity analyses included small differences of the core and the rods dimensions and different levels of model detailing. Four models were simulated and neutronic parameters such as effective multiplication factor (), reactivity (), and thermal and total neutron flux in central thimble in some different conditions of the reactor operation were analysed. The simulated models presented good agreement between them, as well as in comparison with available experimental data. In this way, the sensitivity analyses demonstrated that simulations of the TRIGA IPR-R1 reactor can be performed using any one of the four investigated MCNP models to obtain the referenced neutronic parameters. C. A. M. Silva, J. A. D. Salomé, B. T. Guerra, C. Pereira, A. L. Costa, M. A. F. Veloso, M. A. B. C. Menezes, and H. M. Dalle Copyright © 2014 C. A. M. Silva et al. All rights reserved. Thermal Conductivity of Uranium Nitride and Carbide Mon, 01 Sep 2014 08:37:54 +0000 We investigate the electronic thermal conductivity of alternative fuels like uranium nitride and uranium carbide. We evaluate the electronic contribution to the thermal conductivity, by combining first-principles quantum-mechanical calculations with semiclassical correlations. The electronic structure of UN and UC was calculated using Quantum Espresso code. The spin polarized calculations were performed for a ferromagnetic and antiferromagnetic ordering of magnetic moments on uranium lattice and magnetic moment in UC was lower than in UN due to stronger hybridization between 2p electrons of carbon and 5f electrons of uranium. The nonmagnetic electronic structure calculations were used as an input to BolzTrap code that was used to evaluate the electronic thermal conductivity. It is predicted that the thermal conductivity should increase with the temperature increase, but to get a quantitative agreement with the experiment at higher temperatures the interaction of electrons with phonons (and electron-electron scattering) needs to be included. B. Szpunar and J. A. Szpunar Copyright © 2014 B. Szpunar and J. A. Szpunar. All rights reserved. RELAP5 Analyses of ROSA/LSTF Experiments on AM Measures during PWR Vessel Bottom Small-Break LOCAs with Gas Inflow Mon, 01 Sep 2014 00:00:00 +0000 RELAP5 code posttest analyses were performed on ROSA/LSTF experiments that simulated PWR 0.2% vessel bottom small-break loss-of-coolant accidents with different accident management (AM) measures under assumptions of noncondensable gas inflow and total failure of high-pressure injection system. Depressurization of and auxiliary feedwater (AFW) injection into the secondary-side of both steam generators (SGs) as the AM measures were taken 10 min after a safety injection signal. The primary depressurization rate of 55 K/h caused rather slow primary depressurization being obstructed by the gas accumulation in the SG U-tubes after the completion of accumulator coolant injection. Core temperature excursion thus took place by core boil-off before the actuation of low-pressure injection (LPI) system. The fast primary depressurization by fully opening the relief valves in both SGs with continuous AFW injection led to long-term core cooling by the LPI actuation even under the gas accumulation in the SG U-tubes. The code indicated remaining problems in the predictions of break flow rate during two-phase flow discharge period and primary pressure after the gas inflow. Influences of the primary depressurization rate with continuous AFW injection onto the long-term core cooling were clarified by the sensitivity analyses. Takeshi Takeda Copyright © 2014 Takeshi Takeda. All rights reserved. A Parametric Study of the Impact of the Cooling Water Site Specific Conditions on the Efficiency of a Pressurized Water Reactor Nuclear Power Plant Thu, 28 Aug 2014 00:00:00 +0000 In this study, the thermal analysis for the impact of the cooling seawater site specific conditions on the thermal efficiency of a conceptual pressurized water reactor nuclear power plant (PWR NPP) is presented. The PWR NPP thermal performance depends upon the heat transfer analysis of steam surface condenser accounting for the key parameters such as the cooling seawater salinity and temperature that affect the condenser overall heat transfer coefficient and fouling factor. The study has two aspects: the first one is the impact of the temperature and salinity within a range of (290 K–310 K and 0.00–60000 ppm) on the seawater thermophysical properties such as density, specific heat, viscosity, and thermal conductivity that reflect a reduction in the condenser overall heat transfer coefficient from 2.25 kW/m2 K to 1.265 kW/m2 K at temperature and salinity of 290 K and 0.00 ppm and also from 2.35 kW/m2 K to 1.365 kW/m2 K at temperature and salinity of 310 K and 60000 ppm, whereas the second aspect is the fouling factor variations due to the seawater salinity. The analysis showed that the two aspects have a significant impact on the computation of the condenser overall heat transfer coefficient, whereas the increase of seawater salinity leads to a reduction in the condenser overall heat transfer coefficient. Mohamed M. A. Ibrahim and Mohamed R. Badawy Copyright © 2014 Mohamed M. A. Ibrahim and Mohamed R. Badawy. All rights reserved. Development of Thermal Models and Analysis of UO2-BeO Fuel during a Loss of Coolant Accident Tue, 26 Aug 2014 05:40:16 +0000 Small fraction of high conductivity BeO in UO2 fuel significantly improves thermal conductivity and also affects the overall performance of the fuel during steady state operation and during transients. In this study, performance of UO2-BeO composite under transient conditions such as loss of coolant accident (LOCA), using FRAPTRAN (fuel rod analysis program transient), was carried out. The subroutines in FRAPTRAN code that calculate key thermophysical properties such as thermal conductivity, specific heat capacity, and specific enthalpy were modified to account for the presence of the BeO in UO2. The fuel performance parameters like gas gap pressure, energy stored in fuel, and temperature profiles were studied. The simulation results showed reductions in fuel centerline temperatures and lower temperature drop across fuel rod cross-section under normal fuel operations. It was observed that there was reduction in gas gap pressure and energy stored in fuel. Transient conditions involving cladding rupture were investigated and important performance parameters such as cladding strain were studied. During these transients, the addition of BeO to UO2 fuel seems beneficiary. Deepthi Chandramouli and Shripad T. Revankar Copyright © 2014 Deepthi Chandramouli and Shripad T. Revankar. 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.