bluuebunny/crossref_metadata_2025_split huggingface dataset:
Design optimization of shell-and-tube heat exchangers using single objective and multiobjective particle swarm optimization
Authors: Elsays M. A. , Aly M. Naguib , Badawi A. A. 
Published: December 2013

Abstract
The Particle Swarm Optimization (PSO) algorithm is used to optimize the design of shell-and-tube heat exchangers and determine the optimal feasible solutions so as to eliminate trial-and-error during the design process. The design formulation takes into account the area and the total annual cost of heat exchangers as two objective functions together with operating as well as geometrical constraints. The Nonlinear Constrained Single Objective Particle Swarm Optimization (NCSOPSO) algorithm is used to minimize and find the optimal feasible solution for each of the nonlinear constrained objective functions alone, respectively. Then, a novel Nonlinear Constrained Multobjective Particle Swarm Optimization (NCMOPSO) algorithm is used to minimize and find the Pareto optimal solutions for both of the nonlinear constrained objective functions together. The experimental results show that the two algorithms are very efficient, fast and can find the accurate optimal feasible solutions of the shell and tube heat exchangers design optimization problem.
Neutron transmutation doping conceptual design
Authors: Osman M. M. , Agamy S. A. , Nagy M. S. , Sultan M. 
Published: December 2013

Abstract
This paper describes and investigates upgrading of the silicon doping facility at the ETRR-2 research reactor to accommodate larger size silicon ingots in the two positions available for irradiation in the thermal column of the reactor. Two methods of irradiation are proposed. Both are based on irradiating two ingots simultaneously in one position. After a certain irradiation time, the silicon ingots are reshuffled inside the irradiation channel. The results obtained show that the proposed irradiation strategies of doping silicon ingots are simple to handle and result in acceptable values of resistivity variation.
Chebyshev acceleration of criticality calculations in boundary element applications to neutron diffusion
Authors: Engin O. , Ozgener B. 
Published: December 2013

Abstract
In this study, Chebyshev polynomial method has been applied in the acceleration of the outer iterations for the numerical solution of the two dimensional neutron diffusion equation by the boundary element method. Boundary element discretization of the diffusion equation results in a full and nonsymmetric coefficient matrix which is quite different than the sparse and symmetrical matrices of the finite element and finite difference methods to which Chebyshev acceleration has been classically applied. To assess the merit of the Chebyshev method in the case of boundary element discretization of the multigroup diffusion equation constitutes the main objective of the study. Numerical experimentation has established that the Chebyshev acceleration is effective also in case of boundary element discretization despite the dissimilarity of the coefficient matrices compared to the more conventional methods.
Analysis of the instability of OTSG of movable nuclear power devices by using the multivariable frequency domain method
Authors: Hou S.-X. , Zhang Q.-H. , Luo J.-J. , Xu J. , Tian T. 
Published: December 2013

Abstract
A new method for analyzing the problem of the thermal-hydrodynamic instability of OTSG (Once-Through Steam Generator) of movable nuclear power devices is presented in this paper. The method is based on multivariable frequency domain theory and is more efficient for analyzing the instability of OTSG with coupling interactions and complicated boundary conditions than the usually used single variable method. A mathematical model for the system is derived from the fundamental equations by using perturbation, Laplace-transform and the nodalization techniques. The stable boundary and parametric effects on the stability of the system are evaluated with a computer code. Numerical examples are given and the predictions of the model are in good agreement with the experimental results.
In-reactor measurements of fuel centerline temperature variation during power change
Authors: Horhoianu G. , Ionescu D. V. , Pauna E. I. 
Published: December 2013

Abstract
The thermal response of the experimental fuel elements is examined by thermocouples placed at the centre of the fuel. Such measurements allow the determination of thermal feedback effects induced by simultaneous liberation of fission gases and pellet thermal expansion densification, swelling and relocation of fuel fragments. This paper describes briefly measuring techniques developed and currently used in INR Pitesti, and presents and discusses selected experimental results obtained during in-pile experiments performed in the TRIGA SS reactor of INR Pitesti in order to measure the fuel centerline temperature variation during power change.
Kritikalitätssicherheit der Endlagerung ausgedienter Kernbrennstoffe
Authors: Neuber J. C. 
Published: December 2013

Kurzfassung
In dem nachfolgenden Beitrag werden die Grundsätze der Kritikalitätssicherheitsauslegung eines Endlagers ausgedienter Kernbrennstoffe erläutert. Diese Erläuterungen umfassen:

Sicherheitsprinzipien der Kritikalitätssicherheitsauslegung eines Endlagers

Konsistenz der Sicherheitsprinzipien beim Übergang von der Betriebsphase in die Nachbetriebsphase des Endlagers

Die prinzipielle Abhängigkeit der Länge erforderlicher Nachweiszeiträume von den eingelagerten Nuklidinventaren und deren zukünftigen Entwicklungen, die durch den radioaktiven Zerfall sowie durch mögliche langzeitliche Entwicklungen des Endlagers bedingte Einwirkungen bestimmt werden

Wahrscheinlichkeit des Eintretens von Kritikalität hinsichtlich bisher erwogener Einlagerungskonzepte und die Notwendigkeit des Nachweises der Sicherheit des Endlagers gegen potentielle Kritikalität in der Nachbetriebsphase

Die sich zur Beantwortung der Frage der Sicherheit des Endlagers gegen potentielle Kritikalität für die Kritikalitätsanalyse ergebenden Aufgaben werden skizziert. Die Bearbeitung dieser Aufgaben macht die Modellierung komplexer Mechanismen (Szenarien), die zu Kritikalität führen können, die Modellierung der komplexen und verschiedenartigen Kritikalitätsexkursionen und die Modellierung der Auswirkungen solcher Exkursionen auf die Sicherheit des Endlagers erforderlich. Die Entwicklung entsprechender Modelle erfordert die konzertierte Zusammenarbeit vieler Fachdisziplinen. Es wird angeregt, die bereits bestehenden Kooperationen unter Einbeziehung der Kritikalitätsanalyse in einer nationalen Initiative zum Projekt „Endlagerung“ noch näher miteinander zu vernetzen. Aufgabe einer solchen Kooperation wird es auch sein, die Risiken aller Tätigkeiten, Einrichtungen und Randbedingungen, die mit der Abfallkonditionierung sowie der Beladung, dem Betrieb und der Nachbetriebsphase eines Endlagers verbunden sind, zu bewerten und so optimal auszubilanzieren, dass das Gesamtrisiko dem Minimierungsgebot genügt. Dies schließt eine Bilanzierung des Risikos einer potentiellen Kritikalität in der Nachbetriebsphase gegen die Risiken ein, die mit der Durchführung von Maßnahmen verbunden sind, die zur physikalisch sicheren Verhinderung des Eintretens von Kritikalität in der Nachbetriebsphase vor und während der Einladung in das Endlages erforderlich wären.
Uncertainty analysis for fission products transport in CANDU primary heat transport during a severe accident
Authors: Apostol M. , Constantin M. , Leca A. 
Published: December 2013

Abstract
The work realized under the Severe Accident Research Network of excellence (SARNET) project has shown that the SOPHAEROS module, part of Accident Source Term Evaluation Code (ASTEC) can be fully used to simulate the fission products transport and deposition phenomena in the CANDU Primary Heat Transport (PHT) system. This paper presents an uncertainty analysis for the fission products transport in the CANDU PHT system during a severe accident to obtain the domains of the output parameters, for this study masses of Caesium, Strontium and Iodine deposited in the PHT system and its nodes, taking into account the associated input parameters uncertainties. Five uncertain parameters, the starting time for the releasing process, the duration of the releasing process, the releasing fractions for Cs, Sr and I have been chosen. To generate aleatory values for the uncertain parameters, a method and software have been developed and Monte Carlo simulations to determine uncertainties propagation through the SOPHAEROS module has been carried out.
Axial-power-shape induced frequency shifts in BWR stability measurements
Authors: Nylén H. 
Published: December 2013

Abstract
Higher reactor-core-resonance frequency than expected was recognised at the 1800 MWt BWR Barsebäck 2 in Sweden during stability measurements in 2004, and has subsequently been analysed in some detail. Estimation of the void velocity was made by analysing the phase shift between two levels in the LPRM-detector strings, and compared to calculations. Discrepancies in calculated frequency response between simulated perturbation and results from noise analysis were observed, whereas good agreement between calculated and measured void transport was found. The observed deviations in frequency calculations are most likely affected by the pronounced downward shift in power during the stability measurements, and in such circumstances, the shift in reactor-core-resonance frequency can be estimated from void transport calculations.
Numerical simulation on a HPLWR fuel assembly flow with one revolution of wrapped wire spacers
Authors: Kiss A. , Laurien E. , Aszódi A. , Zhu Y. 
Published: December 2013

Abstract
Three dimensional computational-fluid-dynamics simulations are performed for the fluid flow within a 40 rod fuel bundle in a square arrangement with a central moderator channel. To ensure spacing between the rods, the design of the bundle uses thin wires wrapped counter-clockwise around each rod. This geometry is presently investigated in the framework of the European High-Performance Light-Water Reactor (HPLWR), which operates at supercritical pressure of 25 MPa. A section with one revolution located in the evaporator region of the HPLWR core is investigated using hydraulic (to ensure fully developed flow inlet boundary conditions and reference for heated cases) and thermal-hydraulic boundary conditions. The geometry of wrapped wires gives rise to additional mixing and a circulating or ‘sweeping’ flow near the outer and inner regions of the fuel element next to the wall of the so called fuel assembly and moderator box. Some interesting flow features associated with the complex three-dimensional flow with significant transverse velocity components are visualized as the first evaluated result of this diversified investigation.
Neutronic and burn-up calculations of heterogeneous Thorium/Uranium fuel in pressurized water reactors
Authors: Foad B. , Mansour H. , Aziz M. 
Published: December 2013

Abstract
The purpose of this work is to study the feasibility of the Thorium/Uranium fuel cycle in heterogeneous Pressurized Water Reactors (PWR) core design. This paper focuses on the neutronic and burn-up analysis of the Thorium/Uranium fuel using the computer codes MCNPX and WIMS. The design is based on the Whole Assembly Seed and Blanket (WASB) concept, in which the individual seed (Uranium) and blanket (Thorium-Uranium) units occupy one full-size PWR assembly in a checkerboard core configuration. The results of the present models were compared with the solution of benchmark problems and satisfactory agreement was found.
Steam drum process dynamics and level control of a pressure tube BWR
Authors: Gaikwad A. J. , Vijayan P. K. , Bhartiya S. , Kumar R. , Lele H. G. , Sinha R. K. 
Published: December 2013

Abstract
The Advanced Heavy Water Reactor (AHWR) is a pressure tube type light water cooled heavy water moderated Boiling Water Reactor (BWR) with natural circulation at all power levels. It has parallel inter-connected loops with 452 boiling channels in the Main Heat Transport (MHT) system configuration. These multiple (four) interconnected loops influence the Steam Drum (SD) level control adversely. Such a behavior has not been reported in the open literature. The MHT configuration has been chosen based on comprehensive overall design requirements and certain Postulated Initiated Events (PIEs). This does not allow the partitioning of the Common Reactor Inlet Header (CRIH). If partitioning of the CRIH into four segments is allowed then, it will make each loop independent. Then the SD level control problems subside as the unaccounted interaction among the loops is eliminated. It has also been observed that the open loop response is stable, non-oscillatory and non-diverging for a step change in the feed flow rates. A conventional individual 3-element SD level controller cannot account for the highly coupled and interacting behaviors, of the four loops and SD levels. To overcome these interactions it is proposed to interconnect all the four steam drums in the liquid and vapor regions respectively. The influence of the interconnect configuration and the level controller are studied in detail to find a robust solution. The response obtained for unsymmetrical core power disturbances shows that the SD levels do not diverge and quickly settle to the set points assigned with SD interconnect. The proposed scheme also works well for most of the PIEs considered.
Rapid preparation of Uranium and Thorium alpha sources by electroplating technique
Authors: Hao Le Cong , Van Tao Chau , Van Dong Nguyen , Nghi Ho Nha , Dung Pham Thuy , Van Thong Luong 
Published: December 2013

Abstract
In this study, a simple procedure for the radiochemical determination of thorium isotopes and a robust technique for rapid preparation of uranium and thorium alpha source by electro-deposition using organic buffer solution is described. From the experiments, experience has shown that (1) TOPO is one of the best solvent for the extraction of thorium, (2) isopropyl alcohol is a suitable buffer solution for the electrodepositing of uranium; while acetone is the best choice in case of thorium, (3) the possibility of using of organic phase; containing thorium in the final chemical separation step for electrodepositing, (4) the use of only 3–5 ml of isopropyl alcohol or acetone; 3–5 mm of anode-cathode distance and relatively current 4–15 mA for electro-deposition procedure, (5) the actinides can be deposited quantitatively on a stainless steel disk within 20 to 40 min with nearly 100% of electro-deposition efficiency and the energy resolution of alpha-ray spectra of the electrode-posited sources obtained from alpha spectrometry are within the range of 37–50 KeV (FWHM).
Prediction, analysis and solution of flow inversion phenomenon in a typical MTR reactor with upward core cooling
Authors: El-Morshedy S. El-Din 
Published: December 2013

Abstract
Research reactors of power greater than 20 MW are usually designed to be cooled with upward coolant flow direction inside the reactor core. This is mainly to prevent flow inversion problems following a pump coast down. However, in some designs and under certain operating conditions, flow inversion phenomenon is predicted. In the present work, the best-estimate Material Testing Reactors Thermal-Hydraulic Analysis program (MTRTHA) is used to simulate a typical MTR reactor behavior with upward cooling under a hypothetical case of loss of off-site power. The flow inversion phenomenon is predicted under certain decay heat and/or pool temperature values below the design values. The reactor simulation under loss of off-site power is performed for two cases namely; two-flap valves open and one flap-valve fails to open. The model results for the flow inversion phenomenon prediction is analyzed and a solution of the problem is suggested.
Preliminary evaluation of a radioactive waste repository safety performance by a Monte Carlo simulation-based reliability model
Authors: Cadini F. , Avram D. , Zio E. , Luce A. , Taglioni A. 
Published: December 2013

Abstract
Disposal facilities are designed to provide long-term isolation of the wastes from the human environment by means of a system of barriers, both engineered and natural. In this regard, it is required that the containment function of the multi-barrier system be evaluated by a quantitative risk analysis procedure, also called performance assessment. This entails the implementation of a Probabilistic Safety Assessment model for representing the stochastic failure behavior of the repository barriers, estimating the release rates of radionuclides into the groundwater stream and the doses from subsequent ingestion of drinking water. To this aim, a Monte Carlo simulation scheme is provided within a reliability modelling approach for the safety performance analysis of a repository, accounting also for barrier degradation processes. An application to a case study of the literature is presented to validate the approach and demonstrate its simplicity and flexibility.
Inverse problems using Artificial Neural Networks in long range atmospheric dispersion
Authors: Sharma P. K. , Gera B. , Ghosh A. K. 
Published: December 2013

Abstract
Scalar dispersion in the atmosphere is an important area wherein different approaches are followed in development of good analytical models. The analyses based on Computational Fluid Dynamics (CFD) codes offer an opportunity of model development based on first principles of physics and hence such models have an edge over the existing models. Both forward and backward calculation methods are being developed for atmospheric dispersion around NPPs at BARC. Forward modeling methods, which describe the atmospheric transport from sources to receptors, use forward-running transport and dispersion models or computational fluid dynamics models which are run many times, and the resulting dispersion field is compared to observations from multiple sensors. Backward or inverse modeling methods use only one model run in the reverse direction from the receptors to estimate the upwind sources. Inverse modeling methods include adjoint and tangent linear models, Kalman filters, and variational data assimilation, and neural network. The present paper is aimed at developing a new approach where the identified specific signatures at receptor points form the basis for source estimation or inversions. This approach is expected to reduce the large transient data sets to reduced and meaningful data sets. In fact this reduces the inherently transient data set into a time independent mean data set. Forward computations were carried out with CFD code for various cases to generate a large set of data to train the Artificial Neural Network (ANN). Specific signature analysis was carried out to find the parameters of interest for ANN training like peak concentration, time to reach peak concentration and time to fall. The ANN was trained with data and source strength and locations were predicted from ANN. The inverse problem was performed using the ANN approach in long range atmospheric dispersion. An illustration of application of CFD code for atmospheric dispersion studies for a hypothetical case is also included in the paper.
Application of the TN
method to critical slab problem for one-speed neutrons with forward and backward scattering and efficiency of reflection coefficient
Authors: Bülbül A. , Kara A. , Anli F. 
Published: December 2013

Abstract
The one-speed neutron transport equation which includes isotropic forward and backward scattering has been studied using the TN approximation. Critical thicknesses for different values of the reflection coefficient R and c values have been calculated. The study shows that the TN method gives nearly the same results as the PN approximation in one-dimensional geometry. The results obtained in this study are presented in tables and are compared with PN approximation results.
LOCAs with release of insulation material: generic aspects and international solutions
Authors: Bröcker A. , Pointner W. 
Published: December 2013

Abstract
In case of a LOCA within the containment the released water/steam jet can fragment insulation material from the leaking pipe and from pipes in the neighborhood. The released insulation material and other materials like e.g. dust and paint chips are washed into the sump. A part of this debris reaches the sump strainers after starting ECCS pumps. Some material will be retained by the strainers depending on the mesh size or hole diameter of the strainers. The residual material penetrates the strainers and reaches downstream components and especially fuel elements. Therefore it has to be demonstrated that fuel elements are sufficiently cooled after LOCAs with release of insulation material and other materials. Several events since 1992 have shown the possibility of strainer clogging. This article discusses generic aspects and some international approaches to sump clogging issue and related tasks. An overview of the status quo and comparison of different approaches and background is given.
Multi-scale coupled code systems: from coarse-mesh to high-fidelity LWR core calculations
Authors: Périn Y. , Klein M. , Pasichnyk I. , Velkov K. , Seubert A. , Pautz A. 
Published: December 2013

Abstract
In order to perform safety oriented transient analyses, the entire nuclear power plant has to be simulated in most cases. However, a high level of resolution is only needed where the safety parameters may reach the safety criteria. In this paper, three scale resolutions are defined (NPP level, core level, fuel pin level). Thermal-hydraulic and neutron kinetic codes are already available for each scale. By coupling them, it is possible to develop a complete simulation tool giving access to the required level of precision at each scale level. For this purpose the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH has developed a number of coupled thermal-hydraulics/neutron-kinetics code systems which can solve the defined goal on assembly-wise or pin-by-pin level. Three types of couplings are applied depending on the utilized codes and application purposes: internal, external and parallel. An overview of the existing and future coupled systems available at GRS is given in this paper.
Development of multi-physics code systems based on the reactor dynamics code DYN3D
Authors: Kliem S. , Gommlich A. , Grahn A. , Rohde U. , Schütze J. , Frank T. , Gomez A. , Sanchez V. 
Published: December 2013

Abstract
The reactor dynamics code DYN3D has been coupled with the CFD code ANSYS CFX and the 3D thermal hydraulic core model FLICA4. In the coupling with ANSYS CFX, DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the coupling with FLICA4 only the neutron kinetics module of DYN3D is used. Fluid dynamics and related transport phenomena in the reactor's coolant and fuel behavior is calculated by FLICA4.
The correctness of the coupling of DYN3D with both thermal hydraulic codes was verified by the calculation of different test problems. These test problems were set-up in such a way that comparison with the DYN3D stand-alone code was possible. This included steady-state and transient calculations of a mini-core consisting of nine real-size PWR fuel assemblies with ANSYS CFX/DYN3D as well as mini-core and a full core steady-state calculation using FLICA4/DYN3D.