Comparison of 'system thermal-hydraulics-3 dimensional reactor kinetics' coupled calculations using the MARS 1D and 3D modules and the MASTER code

International Nuclear Information System (INIS)

Jung, J. J.; Joo, H. K.; Lee, W. J.; Ji, S. K.; Jung, B. D.

2002-01-01

KAERI has developed the coupled 'system thermal-hydraulics - 3 dimensional reactor kinetics' code, MARS/MASTER since 1998. However, there is a limitation in the existing MARS/MASTER code; that is, to perform the coupled calculations using MARS/MASTER, we have to utilize the hydrodynamic model and the heat structure model of the MARS '3D module'. In some transients, reactor kinetics behavior is strongly multi-dimensional, but core thermal-hydraulic behavior remains in one-dimensional manner. For efficient analysis of such transients, we coupled the MARS 1D module with MASTER. The new feature has been assessed by the 'OECD NEA Main Steam Line Break (MSLB) benchmark exercise III' simulations

TITAN: an advanced three-dimensional coupled neutronic/thermal-hydraulics code for light water nuclear reactor core analysis

International Nuclear Information System (INIS)

Griggs, D.P.; Kazimi, M.S.; Henry, A.F.

1984-06-01

The three-dimensional nodal neutronics code QUANDRY and the three-dimensional two-fluid thermal-hydraulics code THERMIT are combined into TITAN. Steady-state and transient coupling methodologies based upon a tandem structure were devised and implemented. Additional models for nuclear feedback, equilibrium xenon and direct moderator heating were added. TITAN was tested using a boiling water two channel problem and the coupling methodologies were shown to be effective. Simulated turbine trip transients and several control rod withdrawal transients were analyzed with good results. Sensitivity studies indicated that the time-step size can affect transient results significantly. TITAN was also applied to a quarter core PWR problem based on a real reactor geometry. The steady-state results were compared to a solution produced by MEKIN-B and poor agreement between the horizontal power shapes was found. Calculations with various mesh spacings showed that the mesh spacings in the MEKIN-B analysis were too large to produce accurate results with a finite difference method. The TITAN results were shown to be reasonable. A pair of control rod ejection accidents were also analyzed with TITAN. A comparison of the TITAN PWR control rod ejection results with results from coupled point kinetics/thermal-hydraulics analyses showed that the point kinetics method used (adiabatic method for control rod reactivities, steady-state flux shape for core-averaged reactivity feedback) underpredicted the power excursion in one case and overpredicted it in the other. It was therefore concluded that point kinetics methods should be used with caution and that three-dimensional codes like TITAN are superior for analyzing PWR control rod ejection transients

A Three-Dimensional Kinematic and Kinetic Study of the College-Level Female Softball Swing

Directory of Open Access Journals (Sweden)

Monica Milanovich, Steven M. Nesbit

2014-03-01

Full Text Available This paper quantifies and discusses the three-dimensional kinematic and kinetic characteristics of the female softball swing as performed by fourteen female collegiate amateur subjects. The analyses were performed using a three-dimensional computer model. The model was driven kinematically from subject swings data that were recorded with a multi-camera motion analysis system. Each subject used two distinct bats with significantly different inertial properties. Model output included bat trajectories, subject/bat interaction forces and torques, work, and power. These data formed the basis for a detailed analysis and description of fundamental swing kinematic and kinetic quantities. The analyses revealed that the softball swing is a highly coordinated and individual three-dimensional motion and subject-to-subject variations were significant in all kinematic and kinetic quantities. In addition, the potential effects of bat properties on swing mechanics are discussed. The paths of the hands and the centre-of-curvature of the bat relative to the horizontal plane appear to be important trajectory characteristics of the swing. Descriptions of the swing mechanics and practical implications are offered based upon these findings.

Removal of toxic Cr(VI) ions from tannery industrial wastewater using a newly designed three-phase three-dimensional electrode reactor

Science.gov (United States)

Grace Pavithra, K.; Senthil Kumar, P.; Carolin Christopher, Femina; Saravanan, A.

2017-11-01

In this research, the wastewater samples were collected from leather tanning industry at different time intervals. The parameters like pH, electrical conductivity, temperature, turbidity, chromium and chemical oxygen demand (COD) of the samples were analyzed. A three-phase three-dimensional fluidized type electrode reactor (FTER) was newly designed for the effective removal of toxic pollutants from wastewater. The influencing parameters were optimized for the maximum removal of toxic pollutants from wastewater. The optimum condition for the present system was calculated as: contact time of 30 min, applied voltage of 3 V and the particle electrodes of 15 g. The particle electrode was characterized by using FT-IR analysis. Langmuir-Hinshelwood and pseudo-second order kinetic models were fits well with the experimental data. The results showed that the FTER can be successfully employed for the treatment of industrial wastewater.

Three-dimensional space-time kinetic analysis with CORETRAN and RETRAN-3D of the NEACRP PWR rod ejection benchmark

Energy Technology Data Exchange (ETDEWEB)

Ferroukhi, H.; Coddington, P

2001-03-01

One of the activities within the STARS project, in the Laboratory for Reactor Physics and System Behaviour; is the development of a coupling methodology between the three-dimensional, space-time kinetics codes CORETRAN and RETRAN-3D in order to perform core and plant transient analyses of the Swiss LWRs. The CORETRAN code is a 3-D full-core simulator, intended to be used for core-related analyses, while RETRAN-3D is the three-dimensional kinetics version of the plant system code RETRAN, and can therefore be used for best-estimate analyses of a wide range of transients in both PWRs and BWRs. Because the neutronics solver in both codes is based on the same kinetics model, one important advantage is that the codes can be coupled so that the initial conditions for a RETRAN-3D plant analysis are generated by a detailed-core, steady-state calculation using CORETRAN. As a first step towards using CORETRAN and RETRAN-3D for kinetic applications, the NEACRP PWR rod ejection benchmark has been analyzed with both codes, and is presented in this paper. The first objective is to verify the consistency between the static and kinetic solutions of the two codes, and so gain confidence in the coupling methodology. The second objective is to assess the CORETRAN and RETRAN-3D solutions for a well-defined RIA transient, comparing with previously published results. In parallel, several sensitivity studies have been performed in an attempt to identify models and calculational options important for a correct analysis of an RIA event in a LWR using these two codes. (author)

POST: a postprocessor computer code for producing three-dimensional movies of two-phase flow in a reactor vessel

International Nuclear Information System (INIS)

Taggart, K.A.; Liles, D.R.

1977-08-01

The development of the TRAC computer code for analysis of LOCAs in light-water reactors involves the use of a three-dimensional (r-theta-z), two-fluid hydrodynamics model to describe the two-phase flow of steam and water through the reactor vessel. One of the major problems involved in interpreting results from this code is the presentation of three-dimensional flow patterns. The purpose of the report is to present a partial solution to this data display problem. A first version of a code which produces three-dimensional movies of flow in the reactor vessel has been written and debugged. This code (POST) is used as a postprocessor in conjunction with a stand alone three-dimensional two-phase hydrodynamics code (CYLTF) which is a test bed for the three-dimensional algorithms to be used in TRAC

A Three-Dimensional Kinematic and Kinetic Study of the College-Level Female Softball Swing

Science.gov (United States)

Milanovich, Monica; Nesbit, Steven M.

2014-01-01

This paper quantifies and discusses the three-dimensional kinematic and kinetic characteristics of the female softball swing as performed by fourteen female collegiate amateur subjects. The analyses were performed using a three-dimensional computer model. The model was driven kinematically from subject swings data that were recorded with a multi-camera motion analysis system. Each subject used two distinct bats with significantly different inertial properties. Model output included bat trajectories, subject/bat interaction forces and torques, work, and power. These data formed the basis for a detailed analysis and description of fundamental swing kinematic and kinetic quantities. The analyses revealed that the softball swing is a highly coordinated and individual three-dimensional motion and subject-to-subject variations were significant in all kinematic and kinetic quantities. In addition, the potential effects of bat properties on swing mechanics are discussed. The paths of the hands and the centre-of-curvature of the bat relative to the horizontal plane appear to be important trajectory characteristics of the swing. Descriptions of the swing mechanics and practical implications are offered based upon these findings. Key Points The female softball swing is a highly coordinated and individual three-dimensional motion and subject-to-subject variations were significant in all kinematic and kinetic quantities. The paths of the grip point, bat centre-of-curvature, CG, and COP are complex yet reveal consistent patterns among subjects indicating that these patterns are fundamental components of the swing. The most important mechanical quantity relative to generating bat speed is the total work applied to the bat from the batter. Computer modeling of the softball swing is a viable means for study of the fundamental mechanics of the swing motion, the interactions between the batter and the bat, and the energy transfers between the two. PMID:24570623

Three-dimensional impact kinetics with foot-strike manipulations during running

OpenAIRE

Andrew D. Nordin; Janet S. Dufek; John A. Mercer

2017-01-01

Background: Lack of an observable vertical impact peak in fore/mid-foot running has been suggested as a means of reducing lower extremity impact forces, although it is unclear if impact characteristics exist in other axes. The purpose of the investigation was to compare three-dimensional (3D) impact kinetics among foot-strike conditions in over-ground running using instantaneous loading rate–time profiles. Methods: Impact characteristics were assessed by identifying peak loading rates in e...

RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - user's manual

International Nuclear Information System (INIS)

Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.

1998-03-01

This document is the User's Manual for the Boiling Water Reactor (BWR), and Simplified Boiling Water Reactor (SBWR) systems transient code RAMONA-4B. The code uses a three-dimensional neutron-kinetics model coupled with a multichannel, nonequilibrium, drift-flux, phase-flow model of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients. Chapter 1 gives an overview of the code's capabilities and limitations; Chapter 2 describes the code's structure, lists major subroutines, and discusses the computer requirements. Chapter 3 is on code, auxillary codes, and instructions for running RAMONA-4B on Sun SPARC and IBM Workstations. Chapter 4 contains component descriptions and detailed card-by-card input instructions. Chapter 5 provides samples of the tabulated output for the steady-state and transient calculations and discusses the plotting procedures for the steady-state and transient calculations. Three appendices contain important user and programmer information: lists of plot variables (Appendix A) listings of input deck for sample problem (Appendix B), and a description of the plotting program PAD (Appendix C). 24 refs., 18 figs., 11 tabs

A benchmark for coupled thermohydraulics system/three-dimensional neutron kinetics core models

International Nuclear Information System (INIS)

Kliem, S.

1999-01-01

During the last years 3D neutron kinetics core models have been coupled to advanced thermohydraulics system codes. These coupled codes can be used for the analysis of the whole reactor system. Although the stand-alone versions of the 3D neutron kinetics core models and of the thermohydraulics system codes generally have a good verification and validation basis, there is a need for additional validation work. This especially concerns the interaction between the reactor core and the other components of a nuclear power plant (NPP). In the framework of the international 'Atomic Energy Research' (AER) association on VVER Reactor Physics and Reactor Safety, a benchmark for these code systems was defined. (orig.)

RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - models and correlations

Energy Technology Data Exchange (ETDEWEB)

Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.

1998-03-01

This document describes the major modifications and improvements made to the modeling of the RAMONA-3B/MOD0 code since 1981, when the code description and assessment report was completed. The new version of the code is RAMONA-4B. RAMONA-4B is a systems transient code for application to different versions of Boiling Water Reactors (BWR) such as the current BWR, the Advanced Boiling Water Reactor (ABWR), and the Simplified Boiling Water Reactor (SBWR). This code uses a three-dimensional neutron kinetics model coupled with a multichannel, non-equilibrium, drift-flux, two-phase flow formulation of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients and instability issues. Chapter 1 is an overview of the code`s capabilities and limitations; Chapter 2 discusses the neutron kinetics modeling and the implementation of reactivity edits. Chapter 3 is an overview of the heat conduction calculations. Chapter 4 presents modifications to the thermal-hydraulics model of the vessel, recirculation loop, steam separators, boron transport, and SBWR specific components. Chapter 5 describes modeling of the plant control and safety systems. Chapter 6 presents and modeling of Balance of Plant (BOP). Chapter 7 describes the mechanistic containment model in the code. The content of this report is complementary to the RAMONA-3B code description and assessment document. 53 refs., 81 figs., 13 tabs.

RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - user`s manual

Energy Technology Data Exchange (ETDEWEB)

Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.

1998-03-01

This document is the User`s Manual for the Boiling Water Reactor (BWR), and Simplified Boiling Water Reactor (SBWR) systems transient code RAMONA-4B. The code uses a three-dimensional neutron-kinetics model coupled with a multichannel, nonequilibrium, drift-flux, phase-flow model of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients. Chapter 1 gives an overview of the code`s capabilities and limitations; Chapter 2 describes the code`s structure, lists major subroutines, and discusses the computer requirements. Chapter 3 is on code, auxillary codes, and instructions for running RAMONA-4B on Sun SPARC and IBM Workstations. Chapter 4 contains component descriptions and detailed card-by-card input instructions. Chapter 5 provides samples of the tabulated output for the steady-state and transient calculations and discusses the plotting procedures for the steady-state and transient calculations. Three appendices contain important user and programmer information: lists of plot variables (Appendix A) listings of input deck for sample problem (Appendix B), and a description of the plotting program PAD (Appendix C). 24 refs., 18 figs., 11 tabs.

Development and validation of three-dimensional CFD techniques for reactor safety applications. Final report

International Nuclear Information System (INIS)

Buchholz, Sebastian; Palazzo, Simone; Papukchiev, Angel; Scheurer Martina

2016-12-01

The overall goal of the project RS 1506 ''Development and Validation of Three Dimensional CFD Methods for Reactor Safety Applications'' is the validation of Computational Fluid Dynamics (CFD) software for the simulation of three -dimensional thermo-hydraulic heat and fluid flow phenomena in nuclear reactors. For this purpose a wide spectrum of validation and test cases was selected covering fluid flow and heat transfer phenomena in the downcomer and in the core of pressurized water reactors. In addition, the coupling of the system code ATHLET with the CFD code ANSYS CFX was further developed and validated. The first choice were UPTF experiments where turbulent single- and two-phase flows were investigated in a 1:1 scaled model of a German KONVOI reactor. The scope of the CFD calculations covers thermal mixing and stratification including condensation in single- and two-phase flows. In the complex core region, the flow in a fuel assembly with spacer grid was simulated as defined in the OECD/NEA Benchmark MATIS-H. Good agreement are achieved when the geometrical and physical boundary conditions were reproduced as realistic as possible. This includes, in particular, the consideration of heat transfer to walls. The influence of wall modelling on CFD results was investigated on the TALL-3D T01 experiment. In this case, the dynamic three dimensional fluid flow and heat transfer phenomena were simulated in a Generation IV liquid metal cooled reactor. Concurrently to the validation work, the coupling of the system code ATHLET with the ANSYS CFX software was optimized and expanded for two-phase flows. Different coupling approaches were investigated, in order to overcome the large difference between CPU-time requirements of system and CFD codes. Finally, the coupled simulation system was validated by applying it to the simulation of the PSI double T-junction experiment, the LBE-flow in the MYRRA Spallation experiment and a demonstration test case simulating a pump trip

Point kinetics improvements to evaluate three-dimensional effects in transients calculation

International Nuclear Information System (INIS)

Castellotti, U.

1987-01-01

A calculation method, which considers the flux axial perturbations in the parameters related to the reactivity within a point kinetics model, is described. The method considered uses axial factors of consideration which act on the thermohydraulic variables included in the reactivity calculation. The PUMA three-dimensional code as reference model for the comparisons, is used. The limitations inherent to the reactivity balance of the point models used in the transients calculation, are given. (Author)

Core followup studies of the Tarapur Reactors with the three dimensional BWR simulator COMTEG

Energy Technology Data Exchange (ETDEWEB)

Dwivedi, S. R.; Jagannathan, V.; Mohanakrishnan, P.; Srinivasan, K. R.; Rastogi, B. P.

1976-07-01

Both the units of the Tarapur Atomic Power Station started operation in the year 1969. Since then, these units have completed three cycles. For efficient operation and fuel management of these reactors, a three dimensional BWR simulator COMETG has been developed. The reactors are closely being followed using the simulator. The detailed analyses for cycle 3/4 operation of both the units are described in the paper. The results show very good agreement between calculated and measured values. It is concluded that reactor core behaviour could be predicted in a satisfactory manner with the core simulator COMETG.

Three-dimensional fabric reinforced concrete finds first use in reactor building

International Nuclear Information System (INIS)

Akihama, S.; Nakagava, H.

1989-01-01

It is reported about creation of concrete reinforced with synthetic fibers by Japanese firm Kadzima. Synthetic material with three-dimensional orientation of fibers is produced of roving impreganted with synthetic resin. The reinforcement produced is submerged into the concrete matrix. The compression strength of such a material makes up 58 MPa. The new material is used for constructing the nuclear reactor shielding containers

Three-dimensional fluid-structure interaction dynamics of a pool-reactor in-tank component

International Nuclear Information System (INIS)

Kulak, R.F.

1979-01-01

The safety evaluation of reactor-components often involves the analysis of various types of fluid/structural components interacting in three-dimensional space. For example, in the design of a pool-type reactor several vital in-tank components such as the primary pumps and the intermediate heat exchangers are contained within the primary tank. Typically, these components are suspended from the deck structure and largely submersed in the sodium pool. Because of this positioning these components are vulnerable to structural damage due to pressure wave propagation in the tank during a CDA. In order to assess the structural integrity of these components it is necessary to perform a dynamic analysis in three-dimensional space which accounts for the fluid-structure coupling. A model is developed which has many of the salient features of this fluid-structural component system

Space-time reactor kinetics for heterogeneous reactor structure

Energy Technology Data Exchange (ETDEWEB)

Raisic, N [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

1969-11-15

An attempt is made to formulate time dependent diffusion equation based on Feinberg-Galanin theory in the from analogue to the classical reactor kinetic equation. Parameters of these equations could be calculated using the existing codes for static reactor calculation based on the heterogeneous reactor theory. The obtained kinetic equation could be analogues in form to the nodal kinetic equation. Space-time distribution of neutron flux in the reactor can be obtained by solving these equations using standard methods.

Preliminary Development of the MARS/FREK Spatial Kinetics Coupled System Code for Square Fueled Fast Reactor Applications

International Nuclear Information System (INIS)

Bae, Moo Hoon; Joo, Han Gyu

2009-01-01

Incorporation of a three-dimensional (3-D) reactor kinetics model into a system thermal-hydraulic (T/H) code enhances the capability to perform realistic analyses of the core neutronic behavior and the plant system dynamics which are coupled each other. For this advantage, several coupled system T/H and spatial kinetics codes, such as RELAP/PARCS, RELAP5/ PANBOX, and MARS/MASTER have been developed. These codes, however, so far limited to LWR applications. The objective of this work is to develop such a coupled code for fast reactor applications. Particularly, applications to lead-bismuth eutectic (LBE) cooled fast reactor are of interest which employ open square lattices. A fast reactor kinetics code applicable to square fueled cores called FREK is coupled the LBE version of the MARS code. The MARS/MASTER coupled code is used as the reference for the integration. The coupled code MARS/FREK is examined for a conceptual reactor called P-DEMO which is being developed by NUTRECK. In order to check the validity of the coupled code, however, the OECD MSLB benchmark exercise III calculation is solved first

Reactor kinetics revisited: a coefficient based model (CBM)

International Nuclear Information System (INIS)

Ratemi, W.M.

2011-01-01

In this paper, a nuclear reactor kinetics model based on Guelph expansion coefficients calculation ( Coefficients Based Model, CBM), for n groups of delayed neutrons is developed. The accompanying characteristic equation is a polynomial form of the Inhour equation with the same coefficients of the CBM- kinetics model. Those coefficients depend on Universal abc- values which are dependent on the type of the fuel fueling a nuclear reactor. Furthermore, such coefficients are linearly dependent on the inserted reactivity. In this paper, the Universal abc- values have been presented symbolically, for the first time, as well as with their numerical values for U-235 fueled reactors for one, two, three, and six groups of delayed neutrons. Simulation studies for constant and variable reactivity insertions are made for the CBM kinetics model, and a comparison of results, with numerical solutions of classical kinetics models for one, two, three, and six groups of delayed neutrons are presented. The results show good agreements, especially for single step insertion of reactivity, with the advantage of the CBM- solution of not encountering the stiffness problem accompanying the numerical solutions of the classical kinetics model. (author)

Inverse modeling approach for evaluation of kinetic parameters of a biofilm reactor using tabu search.

Science.gov (United States)

Kumar, B Shiva; Venkateswarlu, Ch

2014-08-01

The complex nature of biological reactions in biofilm reactors often poses difficulties in analyzing such reactors experimentally. Mathematical models could be very useful for their design and analysis. However, application of biofilm reactor models to practical problems proves somewhat ineffective due to the lack of knowledge of accurate kinetic models and uncertainty in model parameters. In this work, we propose an inverse modeling approach based on tabu search (TS) to estimate the parameters of kinetic and film thickness models. TS is used to estimate these parameters as a consequence of the validation of the mathematical models of the process with the aid of measured data obtained from an experimental fixed-bed anaerobic biofilm reactor involving the treatment of pharmaceutical industry wastewater. The results evaluated for different modeling configurations of varying degrees of complexity illustrate the effectiveness of TS for accurate estimation of kinetic and film thickness model parameters of the biofilm process. The results show that the two-dimensional mathematical model with Edward kinetics (with its optimum parameters as mu(max)rho(s)/Y = 24.57, Ks = 1.352 and Ki = 102.36) and three-parameter film thickness expression (with its estimated parameters as a = 0.289 x 10(-5), b = 1.55 x 10(-4) and c = 15.2 x 10(-6)) better describes the biofilm reactor treating the industry wastewater.

A development of three-dimensional seismic isolation for advanced reactor systems in Japan: Pt.2

International Nuclear Information System (INIS)

Kenji Takahashi; Kazuhiko Inoue; Asao Kato; Masaki Morishita; Takafumi Fujita

2005-01-01

Two types of three-dimensional seismic isolation systems were developed for the fast breeder reactor (FBR). One is the three-dimensional entire building base isolation system It was developed by collecting concepts Japanese companies from which a combination system with air springs and hydraulic rocking suppression devices was selected. The other is the vertically isolated system for main components with horizontally entire building base isolation, which was developed by adopting coned disk spring devices. In the study, seismic condition was assumed based on a strict reference ground motion. Design data of the building and components are referred to FBR being developed as the 'Commercialized Fast Reactor Cycle System'. Analysis based on these assumed conditions showed suitable combinations of natural frequencies and damping ratios for isolation. Devices were developed to satisfy the combinations. In five years research and development, several verification tests were performed including shake table tests with scaled models. Finally it is found that the two types of seismic isolation systems are available for FBR. The result is reflected in the preliminary design guideline for the three-dimensional isolation system. (authors)

Three-dimensional two-fluid numerical treatment of a reactor vessel in TRAC

International Nuclear Information System (INIS)

Liles, D.R.

1979-01-01

A three-dimensional two-fluid finite difference model has been used in TRAC (Transient Reactor Analysis Code) to represent a pressurized water reactor vessel. Mesh cells may be blocked off completely to represent large flow obstructions such as downcomer walls. The hydrodynamic volumes and flow areas may also be reduced in order to provide a porous matrix simulation of smaller scale strucuture. The finite difference equations are semi-implicit so that stability time scales are associated with material movement and not wave propagation. The block matrix structure is reduced during the implicit pass to a single element seven stripe system which is easily solved iteratively. This procedure has successfully performed numerous simulations of both full sized reactor accidents and smaller scale experments. It has proven to be a useful feature of the TRAC effort

Application of synthesis methods to two-dimensional fast reactor transient study

International Nuclear Information System (INIS)

Izutsu, Sadayuki; Hirakawa, Naohiro

1978-01-01

Space time synthesis and time synthesis codes were developed and applied to the space-dependent kinetics benchmark problem of a two-dimensional fast reactor model, and it was found both methods are accurate and economical for the fast reactor kinetics study. Comparison between the space time synthesis and the time synthesis was made. Also, in space time synthesis, the influence of the number of trial functions on the error and on the computing time and the effect of degeneration of expansion coefficients are investigated. The matrix factorization method is applied to the inversion of the matrix equation derived from the synthesis equation, and it is indicated that by the use of this scheme space-dependent kinetics problem of a fast reactor can be solved efficiently by space time synthesis. (auth.)

Present status on numerical algorithms and benchmark tests for point kinetics and quasi-static approximate kinetics

International Nuclear Information System (INIS)

Ise, Takeharu

1976-12-01

Review studies have been made on algorithms of numerical analysis and benchmark tests on point kinetics and quasistatic approximate kinetics computer codes to perform efficiently benchmark tests on space-dependent neutron kinetics codes. Point kinetics methods have now been improved since they can be directly applied to the factorization procedures. Methods based on Pade rational function give numerically stable solutions and methods on matrix-splitting are interested in the fact that they are applicable to the direct integration methods. An improved quasistatic (IQ) approximation is the best and the most practical method; it is numerically shown that the IQ method has a high stability and precision and the computation time which is about one tenth of that of the direct method. IQ method is applicable to thermal reactors as well as fast reactors and especially fitted for fast reactors to which many time steps are necessary. Two-dimensional diffusion kinetics codes are most practicable though there exist also three-dimensional diffusion kinetics code as well as two-dimensional transport kinetics code. On developing a space-dependent kinetics code, in any case, it is desirable to improve the method so as to have a high computing speed for solving static diffusion and transport equations. (auth.)

Three-dimensional static and dynamic reactor calculations by the nodal expansion method

International Nuclear Information System (INIS)

Christensen, B.

1985-05-01

This report reviews various method for the calculation of the neutron-flux- and power distribution in an nuclear reactor. The nodal expansion method (NEM) is especially described in much detail. The nodal expansion method solves the diffusion equation. In this method the reactor core is divided into nodes, typically 10 to 20 cm in each direction, and the average flux in each node is calculated. To obtain the coupling between the nodes the local flux inside each node is expressed by use of a polynomial expansion. The expansion is one-dimensional, so inside each node such three expansions occur. To calculate the expansion coefficients it is necessary that the polynomial expansion is a solution to the one-dimensional diffusion equation. When the one-dimensional diffusion equation is established a term with the transversal leakage occur, and this term is expanded after the same polynomials. The resulting equation system with the expansion coefficients as the unknowns is solved with weigthed residual technique. The nodal expansion method is built into a computer program (also called NEM), which is divided into two parts, one part for steady-state calculations and one part for dynamic calculations. It is possible to take advantage of symmetry properties of the reactor core. The program is very flexible with regard to the number of energy groups, the node size, the flux expansion order and the transverse leakage expansion order. The boundary of the core is described by albedos. The program and input to it are described. The program is tested on a number of examples extending from small theoretical one up to realistic reactor cores. Many calculations are done on the wellknown IAEA benchmark case. The calculations have tested the accuracy and the computing time for various node sizes and polynomial expansions. In the dynamic examples various strategies for variation of the time step-length have been tested. (author)

Analysis of the three dimensional core kinetics NESTLE code coupling with the advanced thermo-hydraulic code systems, RELAP5/SCDAPSIM and its application to the Laguna Verde Central reactor

International Nuclear Information System (INIS)

Salazar C, J.H.; Nunez C, A.; Chavez M, C.

2004-01-01

The objective of the written present is to propose a methodology for the joining of the codes RELAP5/SCDAPSIM and NESTLE. The development of this joining will be carried out inside a doctoral program of Engineering in Energy with nuclear profile of the Ability of Engineering of the UNAM together with the National Commission of Nuclear Security and Safeguards (CNSNS). The general purpose of this type of developments, is to have tools that are implemented by multiple programs or codes such a that systems or models of the three-dimensional kinetics of the core can be simulated and those of the dynamics of the reactor (water heater-hydraulics). In the past, by limitations for the calculation of the complete answer of both systems, the developed models they were carried out for separate, putting a lot of emphasis in one but neglecting the other one. These methodologies, calls of better estimate, will be good to the nuclear industry to evaluate, with more high grades of detail, the designs of the nuclear power plant (for modifications to those already existent or for new concepts in the designs of advanced reactors), besides analysing events (transitory and have an accident), among other applications. The coupled system was applied to design studies and investigation of the Laguna Verde Nuclear power plant (CNLV). (Author)

Implementation into a CFD code of neutron kinetics and fuel pin models for nuclear reactor transient analyses

International Nuclear Information System (INIS)

Chen Zhao; Chen, Xue-Nong; Rineiski, Andrei; Zhao Pengcheng; Chen Hongli

2014-01-01

Safety analysis is an important tool for justifying the safety of nuclear reactors. The traditional method for nuclear reactor safety analysis is performed by means of system codes, which use one-dimensional lumped-parameter method to model real reactor systems. However, there are many multi-dimensional thermal-hydraulic phenomena cannot be predicated using traditional one-dimensional system codes. This problem is extremely important for pool-type nuclear systems. Computational fluid dynamics (CFD) codes are powerful numerical simulation tools to solve multi-dimensional thermal-hydraulics problems, which are widely used in industrial applications for single phase flows. In order to use general CFD codes to solve nuclear reactor transient problems, some additional models beyond general ones are required. Neutron kinetics model for power calculation and fuel pin model for fuel pin temperature calculation are two important models of these additional models. The motivation of this work is to develop an advance numerical simulation method for nuclear reactor safety analysis by implementing neutron kinetics model and fuel pin model into general CFD codes. In this paper, the Point Kinetics Model (PKM) and Fuel Pin Model (FPM) are implemented into a general CFD code FLUENT. The improved FLUENT was called as FLUENT/PK. The mathematical models and implementary method of FLUENT/PK are descripted and two demonstration application cases, e.g. the unprotected transient overpower (UTOP) accident of a Liquid Metal cooled Fast Reactor (LMFR) and the unprotected beam overpower (UBOP) accident of an Accelerator Driven System (ADS), are presented. (author)

Three-dimensional neutron kinetics-thermal-hydraulics VVER 1000 main steam line break analysis by RELAP5-3D code

International Nuclear Information System (INIS)

Frisani, A.; Parisi, C.; D'Auria, F.

2007-01-01

After the development and the assessment of Three-Dimensional (3D) Neutron Kinetics (NK) - 1D Thermal-Hydraulics (TH) coupled codes analyses methods, deterministic nuclear safety technology is nowadays producing noticeable efforts for the validation of 3D NK - 3D TH coupled codes analyses methods too. Thus, the purpose of this work was to address the capability of the RELAP5-3D 3D NK-3D TH code to reproduce VVER 1000 Nuclear Power Plant (NPP) core dynamic in simulating the mixing effects that could happen in the vessel downcomer and lower plenum during some scenarios. The work was developed in three steps. The first step dealt with the 3D TH modeling of the Kozloduy-6 VVER 1000 reactor pressure vessel. Then this model was validated following a Steam Generator Isolation transient. The second step has been the development of a 3D NK nodalization for the reactor core region. Then the 3D NK model was directly coupled with the previously developed 3D TH model. The third step was the calculation of a Main Steam Line Break (MSLB) transient. The 3D NK global nuclear parameters were then compared with the 0-D results showing a good agreement; nevertheless only the 3D NK- 3D TH model allowed the calculation of each single assembly power trend for this strong NK-TH asymmetric transient. (author)

Three dimensional contact/impact methodology

International Nuclear Information System (INIS)

Kulak, R.F.

1987-01-01

The simulation of three-dimensional interface mechanics between reactor components and structures during static contact or dynamic impact is necessary to realistically evaluate their structural integrity to off-normal loads. In our studies of postulated core energy release events, we have found that significant structure-structure interactions occur in some reactor vessel head closure designs and that fluid-structure interactions occur within the reactor vessel. Other examples in which three-dimensional interface mechanics play an important role are: (1) impact response of shipping casks containing spent fuel, (2) whipping pipe impact on reinforced concrete panels or pipe-to-pipe impact after a pipe break, (3) aircraft crash on secondary containment structures, (4) missiles generated by turbine failures or tornados, and (5) drops of heavy components due to lifting accidents. The above is a partial list of reactor safety problems that require adequate treatment of interface mechanics and are discussed in this paper

Three species one-dimensional kinetic model for weakly ionized plasmas

Energy Technology Data Exchange (ETDEWEB)

Gonzalez, J., E-mail: jorge.gonzalez@upm.es; Donoso, J. M.; Tierno, S. P. [Department of Applied Physics, Escuela Técnica Superior de Ingeniería Aeronáutica y del Espacio, Universidad Politécnica de Madrid, 28040 Madrid (Spain)

2016-06-15

A three species one-dimensional kinetic model is presented for a spatially homogeneous weakly ionized plasma subjected to the action of a time varying electric field. Planar geometry is assumed, which means that the plasma evolves in the privileged direction of the field. The energy transmitted to the electric charges is channelized to the neutrals thanks to collisions, a mechanism that influences the plasma dynamics. Charge-charge interactions have been designed as a one-dimensional collision term equivalent to the Landau operator used for fully ionized plasmas. Charge-neutral collisions are modelled by a conservative drift-diffusion operator in the Dougherty's form. The resulting set of coupled integro-differential equations is solved with the stable and robust propagator integral method. This semi–analytical method feasibility accounts for non–linear effects without appealing to linearisation or simplifications, providing conservative physically meaningful solutions even for initial or emerging sharp velocity distribution function profiles. It is found that charge-neutral collisions exert a significant effect since a quite different plasma evolution arises if compared to the collisionless limit. In addition, substantial differences in the system motion are found for constant and temperature dependent collision frequencies cases.

Point kinetics model with one-dimensional (radial) heat conduction formalism

International Nuclear Information System (INIS)

Jain, V.K.

1989-01-01

A point-kinetics model with one-dimensional (radial) heat conduction formalism has been developed. The heat conduction formalism is based on corner-mesh finite difference method. To get average temperatures in various conducting regions, a novel weighting scheme has been devised. The heat conduction model has been incorporated in the point-kinetics code MRTF-FUEL. The point-kinetics equations are solved using the method of real integrating factors. It has been shown by analysing the simulation of hypothetical loss of regulation accident in NAPP reactor that the model is superior to the conventional one in accuracy and speed of computation. (author). 3 refs., 3 tabs

Three-dimensional Core Design of a Super Fast Reactor with a High Power Density

International Nuclear Information System (INIS)

Cao, Liangzhi; Oka, Yoshiaki; Ishiwatari, Yuki; Ikejiri, Satoshi; Ju, Haitao

2010-01-01

The SuperCritical Water-cooled Reactor (SCWR) pursues high power density to reduce its capital cost. The fast spectrum SCWR, called a super fast reactor, can be designed with a higher power density than thermal spectrum SCWR. The mechanism of increasing the average power density of the super fast reactor is studied theoretically and numerically. Some key parameters affecting the average power density, including fuel pin outer diameter, fuel pitch, power peaking factor, and the fraction of seed assemblies, are analyzed and optimized to achieve a more compact core. Based on those sensitivity analyses, a compact super fast reactor is successfully designed with an average power density of 294.8 W/cm 3 . The core characteristics are analyzed by using three-dimensional neutronics/thermal-hydraulics coupling method. Numerical results show that all of the design criteria and goals are satisfied

Introduction to the neutron kinetics of nuclear power reactors

CERN Document Server

Tyror, J G; Grant, P J

2013-01-01

An Introduction to the Neutron Kinetics of Nuclear Power Reactors introduces the reader to the neutron kinetics of nuclear power reactors. Topics covered include the neutron physics of reactor kinetics, feedback effects, water-moderated reactors, fast reactors, and methods of plant control. The reactor transients following faults are also discussed, along with the use of computers in the study of power reactor kinetics. This book is comprised of eight chapters and begins with an overview of the reactor physics characteristics of a nuclear power reactor and their influence on system design and

Comparison of the results of the fifth dynamic AER benchmark-a benchmark for coupled thermohydraulic system/three-dimensional hexagonal kinetic core models

International Nuclear Information System (INIS)

Kliem, S.

1998-01-01

The fifth dynamic benchmark was defined at seventh AER-Symposium, held in Hoernitz, Germany in 1997. It is the first benchmark for coupled thermohydraulic system/three-dimensional hexagonal neutron kinetic core models. In this benchmark the interaction between the components of a WWER-440 NPP with the reactor core has been investigated. The initiating event is a symmetrical break of the main steam header at the end of the first fuel cycle and hot shutdown conditions with one control rod group stucking. This break causes an overcooling of the primary circuit. During this overcooling the scram reactivity is compensated and the scrammed reactor becomes re critical. The calculation was continued until the highly-borated water from the high pressure injection system terminated the power excursion. Each participant used own best-estimate nuclear cross section data. Only the initial subcriticality at the beginning of the transient was given. Solutions were received from Kurchatov Institute Russia with the code BIPR8/ATHLET, VTT Energy Finland with HEXTRAN/SMABRE, NRI Rez Czech Republic with DYN3/ATHLET, KFKI Budapest Hungary with KIKO3D/ATHLET and from FZR Germany with the code DYN3D/ATHLET.In this paper the results are compared. Beside the comparison of global results, the behaviour of several thermohydraulic and neutron kinetic parameters is presented to discuss the revealed differences between the solutions.(Authors)

One dimensional neutron kinetics in the TRAC-BF1 code

International Nuclear Information System (INIS)

Weaver, W.L. III; Wagner, K.C.

1987-01-01

The TRAC-BWR code development program at the Idaho National Engineering Laboratory is developing a version of the TRAC code for the U.S. Nuclear Regulatory Commission (USNRC) to provide a best-estimate analysis capability for the simulation of postulated accidents in boiling water reactor (BWR) power systems and related experimental facilities. Recent development efforts in the TRAC-BWR program have focused on improving the computational efficiency through the incorporation of a hybrid Courant- limit-violating numerical solution scheme in the one-dimensional component models and on improving code accuracy through the development of a one-dimensional neutron kinetics model. Many other improvements have been incorporated into TRAC-BWR to improve code portability, accuracy, efficiency, and maintainability. This paper will describe the one- dimensional neutron kinetics model, the generation of the required input data for this model, and present results of the first calculations using the model

Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

International Nuclear Information System (INIS)

Henry, A.F.

1980-01-01

Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

FX2-TH: a two-dimensional nuclear reactor kinetics code with thermal-hydraulic feedback

International Nuclear Information System (INIS)

Shober, R.A.; Daly, T.A.; Ferguson, D.R.

1978-10-01

FX2-TH is a two-dimensional, time-dependent nuclear reactor kinetics program with thermal and hydraulic feedback. The neutronics model used is multigroup neutron diffusion theory. The following geometry options are available: x, r, x-y, r-z, theta-r, and triangular. FX2-TH contains two basic thermal and hydraulic models: a simple adiabatic fuel temperature calculation, and a more detailed model consisting of an explicit representation of a fuel pin, gap, clad, and coolant. FX2-TH allows feedback effects from both fuel temperature (Doppler) and coolant temperature (density) changes. FX2-TH will calculate a consistent set of steady state conditions by iterating between the neutronics and thermal-hydraulics until convergence is reached. The time-dependent calculation is performed by the use of the improved quasistatic method. A disk editing capability is available. FX2-TH is operational on IBM system 360 or 370 computers and on the CDC 7600

STEADY-SHIP: a computer code for three-dimensional nuclear and thermal-hydraulic analyses of marine reactors

International Nuclear Information System (INIS)

Itagaki, Masafumi; Naito, Yoshitaka; Tokuno, Yukio; Matsui, Yasushi.

1988-01-01

A code STEADY-SHIP has been developed to calculate three-dimensional distributions of neutron flux, power and coolant temperature in the reactor core of the nuclear ship MUTSU. The code consists of two parts, that is, a few-group three-dimensional neutron diffusion module DIFFUSION-SHIP and a thermal-hydraulic module HYDRO-SHIP: In the DIFFUSION-SHIP the leakage iteration method is used for solving the three-dimensional neutron diffusion equation with small computer core memory and short computing time; The HYDRO-SHIP performs the general thermal-hydraulic calculation for evaluating feedbacks required in the neutronic calculation by the DIFFUSION-SHIP. The macroscopic nuclear constants are generated by a module CROSS-SHIP as functions of xenon poison, fuel temperature, moderator temperature and moderator density. A module LOCAL-FINE has the capability of computing a detailed rod power distribution for each local node in the core, using the boundary conditions on the surface of the node which were supplied by the STEADY-SHIP whole-core calculation. The applicability of this code to marine reactors has been demonstrated by comparing the computed results with the data measured during the MUTSU land-loaded core critical experiments and with the data obtained during the hot-zero-power tests performed for the actual MUTSU plant. (author)

AUS diffusion module POW checkout - 1- and 2-dimensional kinetics calculations

International Nuclear Information System (INIS)

Pollard, J.P.

1977-01-01

POW is the diffusion module 'workhorse' of the AUS reactor neutronics modular code system; its steady state calculations have been checked out against other diffusion codes (particularly CRAM and GOG). Checkout of kinetic aspects, however, is difficult as kinetic codes are not freely available. In this report POW has been checked against three benchmark calculations as well as a calculation on the 100 KW Argonaut reactor Moata. (author)

Survey of the results of a two- and three-dimensional kinetics benchmark problem typical for a thermal reactor

International Nuclear Information System (INIS)

Werner, W.

1975-01-01

In 1973, NEACRP and CSNI posed a number of kinetic benchmark problems intended to be solved by different groups. Comparison of the submitted results should lead to estimates on the accuracy and efficiency of the employed codes. This was felt to be of great value since the codes involved become more and more important in the field of reactor safety. In this paper the results of the 2d and 3d benchmark problem for a BWR are presented. The specification of the problem is included in the appendix of this survey. For the 2d benchmark problem, 5 contributions have been obtained, while for the 3d benchmark problem 2 contributions have been submitted. (orig./RW) [de

Three-dimensional core analysis on a super fast reactor with negative local void reactivity

International Nuclear Information System (INIS)

Cao Liangzhi; Oka, Yoshiaki; Ishiwatari, Yuki; Ikejiri, Satoshi

2009-01-01

Keeping negative void reactivity throughout the cycle life is one of the most important requirements for the design of a supercritical water-cooled fast reactor (super fast reactor). Previous conceptual design has negative overall void reactivity. But the local void reactivity, which is defined as the reactivity change when the coolant of one fuel assembly disappears, also needs to be kept negative throughout the cycle life because the super fast reactor is designed with closed fuel assemblies. The mechanism of the local void reactivity is theoretically analyzed from the neutrons balance point of view. Three-dimensional neutronics/thermal-hydraulic coupling calculation is employed to analyze the characteristics of the super fast reactor including the local void reactivity. Some configurations of the core are optimized to decrease the local void reactivity. A reference core is successfully designed with keeping both overall and local void reactivity negative. The maximum local void reactivity is less than -30 pcm

Evaluation of the rod ejection accident in Westinghouse Pressurized Water Reactors using spatial kinetics methods

International Nuclear Information System (INIS)

Risher, D.H. Jr.

1975-01-01

The consequences of a rod ejection accident are investigated in relation to the latest, high power density Westinghouse reactors. Limiting criteria are presented, based on experimental evidence, and if not exceeded these criteria will ensure that there will be no interference with core cooling capability, and radiation releases, if any, will be within the guidelines of 10CFR100. A basis is presented for the conservative selection of plant parameters to be used in the analysis, such that the analysis is applicable to a wide range of past, present, and future reactors. The calculational method employs a one-dimensional spatial kinetics computer code and a transient fuel heat transfer computer code to determine the hot spot fuel temperature versus time following a rod ejection. Using these computer codes, the most limiting hot channel factor (which does not cause the fuel damage limit criteria to be exceeded) has been determined as a function of the ejected rod worth. By this means, the limit criteria have been translated into ejected rod worths and hot channel factors which can be used effectively by the nuclear designer and safety analyst. The calculational method is shown to be conservative, compared to the results of a three-dimensional spatial kinetics analysis

Three-dimensional calculations of neutron streaming in the beam tubes of the ORNL HFIR [High Flux Isotope Reactor] Reactor

International Nuclear Information System (INIS)

Childs, R.L.; Rhoades, W.A.; Williams, L.R.

1988-01-01

The streaming of neutrons through the beam tubes in High Flux Isotope Reactor at Oak Ridge National Laboratory has resulted in a reduction of the fracture toughness of the reactor vessel. As a result, an evaluation of vessel integrity was undertaken in order to determine if the reactor can be operated again. As a part of this evaluation, three-dimensional neutron transport calculations were performed to obtain fluxes at points of interest in the wall of the vessel. By comparing the calculated and measured activation of dosimetry specimens from the vessel surveillance program, it was determined that the calculated flux shape was satisfactory to transpose the surveillance data to the locations in the vessel. A bias factor was applied to correct for the average C/E ratio of 0.69. 8 refs., 7 figs., 3 tabs

SHOVAV-JUEL. A one dimensional space-time kinetic code for pebble-bed high-temperature reactors with temperature and Xenon feedback

International Nuclear Information System (INIS)

Nabbi, R.; Meister, G.; Finken, R.; Haben, M.

1982-09-01

The present report describes the modelling basis and the structure of the neutron kinetics-code SHOVAV-Juel. Information for users is given regarding the application of the code and the generation of the input data. SHOVAV-Juel is a one-dimensional space-time-code based on a multigroup diffusion approach for four energy groups and six groups of delayed neutrons. It has been developed for the analysis of the transient behaviour of high temperature reactors with pebble-bed core. The reactor core is modelled by horizontal segments to which different materials compositions can be assigned. The temperature dependence of the reactivity is taken into account by using temperature dependent neutron cross sections. For the simulation of transients in an extended time range the time dependence of the reactivity absorption by Xenon-135 is taken into account. (orig./RW)

Physics and kinetics of TRIGA reactor

International Nuclear Information System (INIS)

Boeck, H.; Villa, M.

2007-01-01

This training module is written as an introduction to reactor physics for reactor operators. It assumes the reader has a basic, fundamental knowledge of physics, materials and mathematics. The objective is to provide enough reactor theory knowledge to safely operate a typical research reactor. At this level, it does not necessarily provide enough information to evaluate the safety aspects of experiment or non-standard operation reviews. The material provides a survey of basic reactor physics and kinetics of TRIGA type reactors. Subjects such as the multiplication factor, reactivity, temperature coefficients, poisoning, delayed neutrons and criticality are discussed in such a manner that even someone not familiar with reactor physics and kinetics can easily follow. A minimum of equations are used and several tables and graphs illustrate the text. (author)

Electrochemical pretreatment of heavy oil refinery wastewater using a three-dimensional electrode reactor

Energy Technology Data Exchange (ETDEWEB)

Wei Lingyong [State Key Laboratory of Heavy Oil Processing, China University of Petroleum, Beijing 102249 (China); Guo Shaohui, E-mail: cupgsh@163.co [State Key Laboratory of Heavy Oil Processing, China University of Petroleum, Beijing 102249 (China); Yan Guangxu; Chen Chunmao [State Key Laboratory of Heavy Oil Processing, China University of Petroleum, Beijing 102249 (China); Jiang Xiaoyan [Liaohe Petrochemical Branch Company, PetroChina, Panjin 124022 (China)

2010-12-01

The pretreatment of heavy oil refinery wastewater (HORW) was experimentally investigated using a three-dimensional electrode reactor (TDER) with granular activated carbon (GAC) and porous ceramsite particle (PCP) as the combination particle electrode and DSA type anodes as the anode. The results showed that higher chemical oxygen demand (COD) removal was obtained in TDER comparing with the two-dimensional electrode reactor (without particle electrodes packed), and combination particle electrode was favorable to improve the COD removal efficiency and reduce the energy consumption. The treated HORW under the optimal experimental condition (GAC percentage = 75%, current density = 30 mA/cm{sup 2}, pH not adjusted and treatment time = 100 min) presented that the removal efficiencies of COD, total organic carbon and toxicity units were 45.5%, 43.3% and 67.2%, respectively, and the ratio of 5-day biochemical oxygen demand to COD was increased from 0.10 to 0.29, which is beneficial for further biological treatment. Furthermore, the application of electrospray ionization Fourier transform ion cyclotron resonance mass spectrometry to characterize polar compounds in HORW and their oxidation products was well demonstrated to reveal the composition variation.

Electrochemical pretreatment of heavy oil refinery wastewater using a three-dimensional electrode reactor

International Nuclear Information System (INIS)

Wei Lingyong; Guo Shaohui; Yan Guangxu; Chen Chunmao; Jiang Xiaoyan

2010-01-01

The pretreatment of heavy oil refinery wastewater (HORW) was experimentally investigated using a three-dimensional electrode reactor (TDER) with granular activated carbon (GAC) and porous ceramsite particle (PCP) as the combination particle electrode and DSA type anodes as the anode. The results showed that higher chemical oxygen demand (COD) removal was obtained in TDER comparing with the two-dimensional electrode reactor (without particle electrodes packed), and combination particle electrode was favorable to improve the COD removal efficiency and reduce the energy consumption. The treated HORW under the optimal experimental condition (GAC percentage = 75%, current density = 30 mA/cm 2 , pH not adjusted and treatment time = 100 min) presented that the removal efficiencies of COD, total organic carbon and toxicity units were 45.5%, 43.3% and 67.2%, respectively, and the ratio of 5-day biochemical oxygen demand to COD was increased from 0.10 to 0.29, which is beneficial for further biological treatment. Furthermore, the application of electrospray ionization Fourier transform ion cyclotron resonance mass spectrometry to characterize polar compounds in HORW and their oxidation products was well demonstrated to reveal the composition variation.

Development of a Reduced-Order Three-Dimensional Flow Model for Thermal Mixing and Stratification Simulation during Reactor Transients

Energy Technology Data Exchange (ETDEWEB)

Hu, Rui

2017-09-03

Mixing, thermal-stratification, and mass transport phenomena in large pools or enclosures play major roles for the safety of reactor systems. Depending on the fidelity requirement and computational resources, various modeling methods, from the 0-D perfect mixing model to 3-D Computational Fluid Dynamics (CFD) models, are available. Each is associated with its own advantages and shortcomings. It is very desirable to develop an advanced and efficient thermal mixing and stratification modeling capability embedded in a modern system analysis code to improve the accuracy of reactor safety analyses and to reduce modeling uncertainties. An advanced system analysis tool, SAM, is being developed at Argonne National Laboratory for advanced non-LWR reactor safety analysis. While SAM is being developed as a system-level modeling and simulation tool, a reduced-order three-dimensional module is under development to model the multi-dimensional flow and thermal mixing and stratification in large enclosures of reactor systems. This paper provides an overview of the three-dimensional finite element flow model in SAM, including the governing equations, stabilization scheme, and solution methods. Additionally, several verification and validation tests are presented, including lid-driven cavity flow, natural convection inside a cavity, laminar flow in a channel of parallel plates. Based on the comparisons with the analytical solutions and experimental results, it is demonstrated that the developed 3-D fluid model can perform very well for a wide range of flow problems.

Three-dimensional tsunami analysis for the plot plan of a sodium-cooled fast reactor plant

International Nuclear Information System (INIS)

Hayakawa, Satoshi; Watanabe, Osamu; Itoh, Kei; Yamamoto, Tomohiko

2013-01-01

As the practical evaluation method of the effect of tsunami on buildings, the formula of tsunami force has been used. However, it cannot be applied to complex geometry of buildings. In this study, to analyze the effect of tsunami on the buildings of sodium-cooled fast reactor plant more accurately, three-dimensional tsunami analysis was performed. In the analysis, VOF (Volume of Fluid) method was used to capture free surface of tsunami. At the beginning, it was confirmed that the tsunami experiment results was reproduced by VOF method accurately. Next, the three-dimensional tsunami analysis was performed with VOF method to evaluate the flow field around the buildings of the plant from the beginning of the tsunami until the backwash of that. (author)

Nuclear reactor kinetics and plant control

CERN Document Server

Oka, Yoshiaki

2013-01-01

Understanding time-dependent behaviors of nuclear reactors and the methods of their control is essential to the operation and safety of nuclear power plants. This book provides graduate students, researchers, and engineers in nuclear engineering comprehensive information on both the fundamental theory of nuclear reactor kinetics and control and the state-of-the-art practice in actual plants, as well as the idea of how to bridge the two. The first part focuses on understanding fundamental nuclear kinetics. It introduces delayed neutrons, fission chain reactions, point kinetics theory, reactivit

Sulfide toxicity kinetics of a uasb reactor

Directory of Open Access Journals (Sweden)

D. R. Paula Jr.

2009-12-01

Full Text Available The effect of sulfide toxicity on kinetic parameters of anaerobic organic matter removal in a UASB (up-flow anaerobic sludge blanket reactor is presented. Two lab-scale UASB reactors (10.5 L were operated continuously during 12 months. The reactors were fed with synthetic wastes prepared daily using glucose, ammonium acetate, methanol and nutrient solution. One of the reactors also received increasing concentrations of sodium sulfide. For both reactors, the flow rate of 16 L.d-1 was held constant throughout the experiment, corresponding to a hydraulic retention time of 15.6 hours. The classic model for non-competitive sulfide inhibition was applied to the experimental data for determining the overall kinetic parameter of specific substrate utilization (q and the sulfide inhibition coefficient (Ki. The application of the kinetic parameters determined allows prediction of methanogenesis inhibition and thus the adoption of operating parameters to minimize sulfide toxicity in UASB reactors.

Three-dimensional Mesoscale Simulations of Detonation Initiation in Energetic Materials with Density-based Kinetics

Science.gov (United States)

Jackson, Thomas; Jost, A. M.; Zhang, Ju; Sridharan, P.; Amadio, G.

2017-06-01

In this work we present three-dimensional mesoscale simulations of detonation initiation in energetic materials. We solve the reactive Euler equations, with the energy equation augmented by a power deposition term. The reaction rate at the mesoscale is modelled using a density-based kinetics scheme, adapted from standard Ignition and Growth models. The deposition term is based on previous results of simulations of pore collapse at the microscale, modelled at the mesoscale as hot-spots. We carry out three-dimensional mesoscale simulations of random packs of HMX crystals in a binder, and show that the transition between no-detonation and detonation depends on the number density of the hot-spots, the initial radius of the hot-spot, the post-shock pressure of an imposed shock, and the amplitude of the power deposition term. The trends of transition at lower pressure of the imposed shock for larger number density of pore observed in experiments is reproduced. Initial attempts to improve the agreement between the simulation and experiments through calibration of various parameters will also be made.

Use of principal components analysis and three-dimensional atmospheric-transport models for reactor-consequence evaluation

International Nuclear Information System (INIS)

Gudiksen, P.H.; Walton, J.J.; Alpert, D.J.; Johnson, J.D.

1982-01-01

This work explores the use of principal components analysis coupled to three-dimensional atmospheric transport and dispersion models for evaluating the environmental consequences of reactor accidents. This permits the inclusion of meteorological data from multiple sites and the effects of topography in the consequence evaluation; features not normally included in such analyses. The technique identifies prevailing regional wind patterns and their frequencies for use in the transport and dispersion calculations. Analysis of a hypothetical accident scenario involving a release of radioactivity from a reactor situated in a river valley indicated the technique is quite useful whenever recurring wind patterns exist, as is often the case in complex terrain situations. Considerable differences were revealed in a comparison with results obtained from a more conventional Gaussian plume model using only the reactor site meteorology and no topographic effects

Analysis of Kinetic Parameter Effect on Reactor Operation Stability of the RSG-GAS Reactor

International Nuclear Information System (INIS)

Rokhmadi

2007-01-01

Kinetic parameter has influence to behaviour on RSG-GAS reactor operation. In this paper done is the calculation of reactivity curve, period-reactivity relation and low power transfer function in silicide fuel. This parameters is necessary and useful for reactivity characteristic analysis and reactor stability. To know the reactivity response, it was done reactivity insertion at power 1 watt using POKDYN code because at this level of power no feedback reactivity so important for reactor operation safety. The result of calculation showed that there is no change of significant a period-reactivity relation and transfer function at low power for 2.96 gU/cc, 3.55 gU/cc and 4.8 gU/cc density of silicide fuels. The result of the transfer function at low power showed that the reactor is critical stability with no feedback. The result of calculation also showed that reactivity response no change among three kinds of fuel densities. It can be concluded that from kinetic parameter point of view period-reactivity relation, transfer function at low power, and reactivity response are no change reactor operation from reactivity effect when fuel exchanged. (author)

A three-dimensional transient calculation for the reactor model RAMONA using the COMMIX-2(V) code

International Nuclear Information System (INIS)

Weinberg, D.; Frey, H.H.; Tschoeke, H.

1993-01-01

The safety graded decay heat removal system of the European Fast Reactor needs a high availability. This system operates entirely under natural convection. To guarantee a proper design, experiments are carried out to verify thermal-hydraulic computer codes able to predict precisely local temperature loadings of the components and the reactor tank in the transition region from nominal operation under forced convection to the decay heat removal operation. - With the COMMIX-2 (V) code three-dimensional transient calculations have been performed to simulate experiments in the 360 deg. reactor model RAMONA, scaled 1:20 to the reality with water as simulant fluid for sodium. The computed average and local temperatures as well as the velocity distributions show a good agreement with the experimental results. Further efforts are necessary to reduce the computation time. (orig.)

Effects of applying three-dimensional seismic isolation system on the seismic design of FBR

International Nuclear Information System (INIS)

Hirata, Kazuta; Yabana, Shuichi; Kanazawa, Kenji; Matsuda, Akihiro

1997-01-01

In this study conceptional three-dimensional seismic isolation system for fast breeder reactor (FBR) is proposed. Effects of applying three-dimensional seismic isolation system on the seismic design for the FBR equipment are evaluated quantitatively. From the evaluation, it is concluded following effects are expected by applying the three-dimensional seismic isolation system to the FBR and the effects are evaluated quantitatively. (1) Reduction of membrane thickness of the reactor vessel (2) Suppression of uplift of fuels by reducing vertical seismic response of the core (3) Reduction of the supports for the piping system (4) Three-dimensional base isolation system for the whole reactor building is advantageous to the combined isolation system of horizontal base isolation for the reactor building and vertical isolation for the equipment. (author)

Hamiltonian formalism of two-dimensional Vlasov kinetic equation.

Science.gov (United States)

Pavlov, Maxim V

2014-12-08

In this paper, the two-dimensional Benney system describing long wave propagation of a finite depth fluid motion and the multi-dimensional Russo-Smereka kinetic equation describing a bubbly flow are considered. The Hamiltonian approach established by J. Gibbons for the one-dimensional Vlasov kinetic equation is extended to a multi-dimensional case. A local Hamiltonian structure associated with the hydrodynamic lattice of moments derived by D. J. Benney is constructed. A relationship between this hydrodynamic lattice of moments and the two-dimensional Vlasov kinetic equation is found. In the two-dimensional case, a Hamiltonian hydrodynamic lattice for the Russo-Smereka kinetic model is constructed. Simple hydrodynamic reductions are presented.

Development of three-dimensional nuclear design program for large fast breeder reactor

International Nuclear Information System (INIS)

Inoue, Kohtaro

1987-01-01

The report describes a calculation program for core design, called HICOM, and its calculation accuracy. HICOM is designed for three-dimensional neutron diffusion calculation and combustion calculation for large fast breeder reactors to be conducted according to a control rod plan and fuel replacement plan. The improved coarse mesh technique is applied to neutron diffusion calculation. It is demostrated that HICOM permits rapid and accurate operation. For the evaluation of the applicability of HICOM, three-dimensional six-group neutron diffusion calculation is conducted for a 1,000 MWe axial heterogeneous FBR core. Results demonstrate that the program can perform numerical calculation in a time period shorter than 1-40 that for calculation by CITATION (triangle mesh method). This is achieved by using the improved coarse mesh method and carrying out the operation by a vectorial procedure. For the evaluation of the nuclear calculation accuracy of HICOM, analysis is made of reactivity, output distribution and B 4 C control rod worth emasured in an FCA criticality experiment carried out by the Japan Atomic Energy Research Institute. Calculations are found to agree with measurements within a permissible error. The same level of calculation accuracy is obtained for homogneous core, axial heterogeneous core and cores with internal blankets with different forms. (Nogami, K.)

Simulation and calculation of three-reactor system of catalytic reforming

International Nuclear Information System (INIS)

Rikalovska, Tatjana; Markovska, Liljana; Meshko, Vera; Poposka, Filimena

1999-01-01

The process of catalytic reforming has been operated for quite a long time, one can not always find real data for the kinetics and thermodynamics of the reactions that take place during the catalytic reforming process in order to facilitate the designing of reactor system or its simulation in a wide:ran e of process parameters. Kinetic and thermodynamic data have been collected for the reactions that take place during the catalytic reforming process. The stress has been pointed on four major reactions: dehydrogenation of naphthenes (aromatization), dehydrocyclization of paraffins and hydrocracking of naphthenes and paraffins. On the base of such a kinetic model, the reforming process has been described with a system of differential equations. For the purpose of solving these equations computer programs for simulation of a three-reactor system for adiabatic operation of the reactors. The computer simulation of the mathematical model of this three-reactor system has been accomplished by use of the ISIM-dynamic simulator. The results obtained out of the simulation agree very good with the data of the real process of catalytic reforming in OKTA Crude Oil Refinery in Skopje, Macedonia. (Author)

Kinetic Monte Carlo simulations of three-dimensional self-assembled quantum dot islands

International Nuclear Information System (INIS)

Song Xin; Feng Hao; Liu Yu-Min; Yu Zhong-Yuan; Yin Hao-Zhi

2014-01-01

By three-dimensional kinetic Monte Carlo simulations, the effects of the temperature, the flux rate, the total coverage and the interruption time on the distribution and the number of self-assembled InAs/GaAs (001) quantum dot (QD) islands are studied, which shows that a higher temperature, a lower flux rate and a longer growth time correspond to a better island distribution. The relations between the number of islands and the temperature and the flux rate are also successfully simulated. It is observed that for the total coverage lower than 0.5 ML, the number of islands decreases with the temperature increasing and other growth parameters fixed and the number of islands increases with the flux rate increasing when the deposition is lower than 0.6 ML and the other parameters are fixed. (condensed matter: structural, mechanical, and thermal properties)

A three-dimensional self-learning kinetic Monte Carlo model: application to Ag(111)

International Nuclear Information System (INIS)

Latz, Andreas; Brendel, Lothar; Wolf, Dietrich E

2012-01-01

The reliability of kinetic Monte Carlo (KMC) simulations depends on accurate transition rates. The self-learning KMC method (Trushin et al 2005 Phys. Rev. B 72 115401) combines the accuracy of rates calculated from a realistic potential with the efficiency of a rate catalog, using a pattern recognition scheme. This work expands the original two-dimensional method to three dimensions. The concomitant huge increase in the number of rate calculations on the fly needed can be avoided by setting up an initial database, containing exact activation energies calculated for processes gathered from a simpler KMC model. To provide two representative examples, the model is applied to the diffusion of Ag monolayer islands on Ag(111), and the homoepitaxial growth of Ag on Ag(111) at low temperatures.

A three-dimensional methodology for the assessment of neutron damage and nuclear energy deposition in graphite components of advanced gas-cooled reactors

Energy Technology Data Exchange (ETDEWEB)

Morgan, D.O.; Robinson, A.T.; Allen, D.A.; Picton, D.J.; Thornton, D.A. [TCS, Serco, Rutherford House, Olympus Park, Quedgeley, Gloucester, Gloucestershire GL2 4NF (United Kingdom); Shaw, S.E. [EDF Energy, Barnet Way, Barnwood, Gloucester GL4 3RS (United Kingdom)

2011-07-01

This paper describes the development of a three-dimensional methodology for the assessment of neutron damage and nuclear energy deposition (or nuclear heating) throughout the graphite cores of the UK's Advanced Gas-cooled Reactors. Advances in the development of the Monte Carlo radiation transport code MCBEND have enabled the efficient production of detailed fully three-dimensional models that utilise three-dimensional source distributions obtained from Core Follow data supplied by the reactor physics code PANTHER. The calculational approach can be simplified to reduce both the requisite number of intensive radiation transport calculations, as well as the quantity of data output. These simplifications have been qualified by comparison with explicit calculations and they have been shown not to introduce significant systematic uncertainties. Simple calculational approaches are described that allow users of the data to address the effects on neutron damage and nuclear energy deposition predictions of the feedback resulting from the mutual dependencies of graphite weight loss and nuclear energy deposition. (authors)

Three-dimensional calculation of the flow in the cold plenum of the Fast Breeder Reactor 1500 MW

International Nuclear Information System (INIS)

Chabard, J.P.; Daubert, O.; Gregoire, J.P.; Hemmerich, P.

1987-01-01

To solve thermalhydraulics problems which are rising for example on the various parts of nuclear reactors, several departments of the Direction des Etudes et Recherches are developing the N3S code, three-dimensional code using the finite element method. First, this paper presents the basic equations (Navies-Stokes with turbulence modelling and coupled with the thermal equation) and well suited algorithms to solve them. The industrial adequacy of the code is clearly demonstrated through the application to the computation of the flow in the cold plenum of the Fast Breeder Reactor 1500 MW on a mesh of about 20000 velocity nodes [fr

Reactor kinetics - pulse and steady state

Energy Technology Data Exchange (ETDEWEB)

Estes, B F; Morris, F M [Sandia Laboratories (United States)

1974-07-01

An analytical model has been developed which couples the nuclear and thermal characteristics of the Annular Core Pulse Reactor (ACPR) into a solution which describes both the neutron kinetics of the reactor and the temperature behavior of a fuel-moderator element. The model describes both pulse and steady state operations. This paper describes the important aspects of the reactor, the fuel- moderator elements, the neutron kinetic equations of the reactor, and the time-temperature behavior of a fuel-moderator element that is being subjected to the maximum power density in the core. The parameters which are utilized in the equations are divided into two classes, those that can be measured directly and those that are assumed to be known (each is described briefly). Some of the solutions which demonstrate the versatility of the analytical model are described. (author)

Kinetic and dynamic Delaunay tetrahedralizations in three dimensions

Science.gov (United States)

Schaller, Gernot; Meyer-Hermann, Michael

2004-09-01

We describe algorithms to implement fully dynamic and kinetic three-dimensional unconstrained Delaunay triangulations, where the time evolution of the triangulation is not only governed by moving vertices but also by a changing number of vertices. We use three-dimensional simplex flip algorithms, a stochastic visibility walk algorithm for point location and in addition, we propose a new simple method of deleting vertices from an existing three-dimensional Delaunay triangulation while maintaining the Delaunay property. As an example, we analyse the performance in various cases of practical relevance. The dual Dirichlet tessellation can be used to solve differential equations on an irregular grid, to define partitions in cell tissue simulations, for collision detection etc.

Numerical nodal simulation of the axial power distribution within nuclear reactors using a kinetics diffusion model. I

International Nuclear Information System (INIS)

Barros, R.C. de.

1992-05-01

Presented here is a new numerical nodal method for the simulation of the axial power distribution within nuclear reactors using the one-dimensional one speed kinetics diffusion model with one group of delayed neutron precursors. Our method is based on a spectral analysis of the nodal kinetics equations. These equations are obtained by integrating the original kinetics equations separately over a time step and over a spatial node, and then considering flat approximations for the forward difference terms. These flat approximations are the only approximations that are considered in the method. As a result, the spectral nodal method for space - time reactor kinetics generates numerical solutions for space independent problems or for time independent problems that are completely free from truncation errors. We show numerical results to illustrate the method's accuracy for coarse mesh calculations. (author)

The spatial kinetic analysis of accelerator-driven subcritical reactor

International Nuclear Information System (INIS)

Takahashi, H.; An, Y.; Chen, X.

1998-02-01

The operation of the accelerator driven reactor with subcritical condition provides a more flexible choice of the reactor materials and of design parameters. A deep subcriticality is chosen sometime from the analysis of point kinetics. When a large reactor is operated in deep subcritical condition by using a localized spallation source, the power distribution has strong spatial dependence, and point kinetics does not provide proper analysis for reactor safety. In order to analyze the spatial and energy dependent kinetic behavior in the subcritical reactor, the authors developed a computation code which is composed of two parts, the first one is for creating the group cross section and the second part solves the multi-group kinetic diffusion equations. The reactor parameters such as the cross section of fission, scattering, and energy transfer among the several energy groups and regions are calculated by using a code modified from the Monte Carlo codes MCNPA and LAHET instead of the usual analytical method of ANISN, TWOTRAN codes. Thus the complicated geometry of the accelerator driven reactor core can be precisely taken into account. The authors analyzed the subcritical minor actinide transmutor studied by Japan Atomic Energy Research Institute (JAERI) using the code

Kinetic analysis of sub-prompt-critical reactor assemblies

International Nuclear Information System (INIS)

Das, S.

1992-01-01

Neutronic analysis of safety-related kinetics problems in experimental neutron multiplying assemblies has been carried out using a sub-prompt-critical reactor model. The model is based on the concept of a sub-prompt-critical nuclear reactor and the concept of instantaneous neutron multiplication in a reactor system. Computations of reactor power, period and reactivity using the model show excellent agreement with results obtained from exact kinetics method. Analytic expressions for the energy released in a controlled nuclear power excursion are derived. Application of the model to a Pulsed Fast Reactor gives its sensitivity between 4 and 5. (author). 6 refs., 4 figs., 1 tab

A three-dimensional operational transient simulation of the CANDU core with typical reactor regulating system

Energy Technology Data Exchange (ETDEWEB)

Yeom, Choong Sub; Kim, Hyun Dae; Park, Kyung Seok; Park, Jong Woon [Institute for Advanced Engineering, Taejon (Korea, Republic of)

1995-07-01

This paper describes the results of simulation of a CANDU operational transient problem (re-startup after short shutdown) using the Coupled Reactor Kinetics(CRKIN) code developed previously with CANDU Reactor Regulating System (RRS) logic. The performance in the simulation is focused on investigating the behaviours of neutron power and regulating devices in accordance with the changes of xenon concentration following the operation of the RRS.

CINESP - computational program of spatial kinetics for nuclear reactors in the one-two dimension multigroup diffusion theory

International Nuclear Information System (INIS)

Santos, R.S. dos

1993-01-01

This paper presents a computational program to solve numerically the reactor kinetics equations in the multigroup diffusion theory. One or two-dimensional problems in cylindrical or Cartesian geometries, with any number of energy and delayed-neutron precursors groups are dealt with. The main input and output of the program are briefly discussed. Various results demonstrate the accuracy and versatility of the program, when compared with other kinetics programs. (author)

UV reactor flow visualization and mixing quantification using three-dimensional laser-induced fluorescence.

Science.gov (United States)

Gandhi, Varun; Roberts, Philip J W; Stoesser, Thorsten; Wright, Harold; Kim, Jae-Hong

2011-07-01

Three-dimensional laser-induced fluorescence (3DLIF) was applied to visualize and quantitatively analyze mixing in a lab-scale UV reactor consisting of one lamp sleeve placed perpendicular to flow. The recirculation zone and the von Karman vortex shedding that commonly occur in flows around bluff bodies were successfully visualized. Multiple flow paths were analyzed by injecting the dye at various heights with respect to the lamp sleeve. A major difference in these pathways was the amount of dye that traveled close to the sleeve, i.e., a zone of higher residence time and higher UV exposure. Paths away from the center height had higher velocities and hence minimal influence by the presence of sleeve. Approach length was also characterized in order to increase the probability of microbes entering the region around the UV lamp. The 3DLIF technique developed in this study is expected to provide new insight on UV dose delivery useful for the design and optimization of UV reactors. Copyright © 2011 Elsevier Ltd. All rights reserved.

Three-Dimensional Analysis of the Hot-Spot Fuel Temperature in Pebble Bed and Prismatic Modular Reactors

International Nuclear Information System (INIS)

In, W. K.; Lee, S. W.; Lim, H. S.; Lee, W. J.

2006-01-01

High temperature gas-cooled reactors(HTGR) have been reviewed as potential sources for future energy needs, particularly for a hydrogen production. Among the HTGRs, the pebble bed reactor(PBR) and a prismatic modular reactor(PMR) are considered as the nuclear heat source in Korea's nuclear hydrogen development and demonstration project. PBR uses coated fuel particles embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the core during an operation. PMR uses graphite fuel blocks which contain cylindrical fuel compacts consisting of the fuel particles. The fuel blocks also contain coolant passages and locations for absorber and control material. The maximum fuel temperature in the core hot spot is one of the important design parameters for both a PBR and a PMR. The objective of this study is to predict the hot-spot fuel temperature distributions in a PBR and a PMR at a steady state. The computational fluid dynamics(CFD) code, CFX-10 is used to perform the three-dimensional analysis. The latest design data was used here based on the reference reactor designs, PBMR400 and GTMHR60

Reflected kinetics model for nuclear space reactor kinetics and control scoping calculations

Energy Technology Data Exchange (ETDEWEB)

Washington, K.E.

1986-05-01

The objective of this research is to develop a model that offers an alternative to the point kinetics (PK) modelling approach in the analysis of space reactor kinetics and control studies. Modelling effort will focus on the explicit treatment of control drums as reactivity input devices so that the transition to automatic control can be smoothly done. The proposed model is developed for the specific integration of automatic control and the solution of the servo mechanism problem. The integration of the kinetics model with an automatic controller will provide a useful tool for performing space reactor scoping studies for different designs and configurations. Such a tool should prove to be invaluable in the design phase of a space nuclear system from the point of view of kinetics and control limitations.

Reflected kinetics model for nuclear space reactor kinetics and control scoping calculations

International Nuclear Information System (INIS)

Washington, K.E.

1986-05-01

The objective of this research is to develop a model that offers an alternative to the point kinetics (PK) modelling approach in the analysis of space reactor kinetics and control studies. Modelling effort will focus on the explicit treatment of control drums as reactivity input devices so that the transition to automatic control can be smoothly done. The proposed model is developed for the specific integration of automatic control and the solution of the servo mechanism problem. The integration of the kinetics model with an automatic controller will provide a useful tool for performing space reactor scoping studies for different designs and configurations. Such a tool should prove to be invaluable in the design phase of a space nuclear system from the point of view of kinetics and control limitations

Three-dimensional numerical investigation of a Molten Salt reactor concept with the code CFX-5.5

International Nuclear Information System (INIS)

Yamaji, B.; Csom, G.; Aszodi, A.

2002-01-01

Partitioning and transmutation of actinides and long-lived fission products is a promising option to extend the possibilities and enhance the environmentally acceptable capabilities of nuclear energy. Also the possible implementation of the thorium cycle is considered as a way to reduce the problem of energy resources in the future. For both objectives different molten salt reactor concepts were proposed mainly based on the Molten Salt Reactor Experiment of the Oak Ridge National Laboratory. Not only critical reactors but also accelerator-driven subcritical systems (ADSs) have advantages worth considering for those aims, especially those ones with liquid fuel, such as molten salts. By using liquid fuel which is the coolant medium, too, a basically different thermalhydraulic behavior is expected than in the case of solid fuel and water coolant. In this work our purpose is to present the possible use of Computational Fluid Dynamics (CFD) technology in molten salt thermal hydraulics. The simulations were performed with the three-dimensional code CFX-5.5.(author)

Product Characterization and Kinetics of Biomass Pyrolysis in a Three-Zone Free-Fall Reactor

Directory of Open Access Journals (Sweden)

Natthaya Punsuwan

2014-01-01

Full Text Available Pyrolysis of biomass including palm shell, palm kernel, and cassava pulp residue was studied in a laboratory free-fall reactor with three separated hot zones. The effects of pyrolysis temperature (250–1050°C and particle size (0.18–1.55 mm on the distribution and properties of pyrolysis products were investigated. A higher pyrolysis temperature and smaller particle size increased the gas yield but decreased the char yield. Cassava pulp residue gave more volatiles and less char than those of palm kernel and palm shell. The derived solid product (char gave a high calorific value of 29.87 MJ/kg and a reasonably high BET surface area of 200 m2/g. The biooil from palm shell is less attractive to use as a direct fuel, due to its high water contents, low calorific value, and high acidity. On gas composition, carbon monoxide was the dominant component in the gas product. A pyrolysis model for biomass pyrolysis in the free-fall reactor was developed, based on solving the proposed two-parallel reactions kinetic model and equations of particle motion, which gave excellent prediction of char yields for all biomass precursors under all pyrolysis conditions studied.

SNAP - a three dimensional neutron diffusion code

International Nuclear Information System (INIS)

McCallien, C.W.J.

1993-02-01

This report describes a one- two- three-dimensional multi-group diffusion code, SNAP, which is primarily intended for neutron diffusion calculations but can also carry out gamma calculations if the diffusion approximation is accurate enough. It is suitable for fast and thermal reactor core calculations and for shield calculations. SNAP can solve the multi-group neutron diffusion equations using finite difference methods. The one-dimensional slab, cylindrical and spherical geometries and the two-dimensional case are all treated as simple special cases of three-dimensional geometries. Numerous reflective and periodic symmetry options are available and may be used to reduce the number of mesh points necessary to represent the system. Extrapolation lengths can be specified at internal and external boundaries. (Author)

A DETERMINISTIC METHOD FOR TRANSIENT, THREE-DIMENSIONAL NUETRON TRANSPORT

International Nuclear Information System (INIS)

S. GOLUOGLU, C. BENTLEY, R. DEMEGLIO, M. DUNN, K. NORTON, R. PEVEY I.SUSLOV AND H.L. DODDS

1998-01-01

A deterministic method for solving the time-dependent, three-dimensional Boltzmam transport equation with explicit representation of delayed neutrons has been developed and evaluated. The methodology used in this study for the time variable of the neutron flux is known as the improved quasi-static (IQS) method. The position, energy, and angle-dependent neutron flux is computed deterministically by using the three-dimensional discrete ordinates code TORT. This paper briefly describes the methodology and selected results. The code developed at the University of Tennessee based on this methodology is called TDTORT. TDTORT can be used to model transients involving voided and/or strongly absorbing regions that require transport theory for accuracy. This code can also be used to model either small high-leakage systems, such as space reactors, or asymmetric control rod movements. TDTORT can model step, ramp, step followed by another step, and step followed by ramp type perturbations. It can also model columnwise rod movement can also be modeled. A special case of columnwise rod movement in a three-dimensional model of a boiling water reactor (BWR) with simple adiabatic feedback is also included. TDTORT is verified through several transient one-dimensional, two-dimensional, and three-dimensional benchmark problems. The results show that the transport methodology and corresponding code developed in this work have sufficient accuracy and speed for computing the dynamic behavior of complex multidimensional neutronic systems

Comparison of three combined sequencing batch reactor followed by enhanced Fenton process for an azo dye degradation: Bio-decolorization kinetics study

Energy Technology Data Exchange (ETDEWEB)

Azizi, A., E-mail: armina_84@yahoo.com [Civil and Environmental Engineering Department, Amirkabir University of Technology, Hafez Ave., Tehran15875-4413 (Iran, Islamic Republic of); Alavi Moghaddam, M.R., E-mail: alavim@yahoo.com [Civil and Environmental Engineering Department, Amirkabir University of Technology, Hafez Ave., Tehran15875-4413 (Iran, Islamic Republic of); Maknoon, R., E-mail: rmaknoon@yahoo.com [Civil and Environmental Engineering Department, Amirkabir University of Technology, Hafez Ave., Tehran15875-4413 (Iran, Islamic Republic of); Kowsari, E., E-mail: kowsarie@aut.ac.ir [Department of Chemistry, Amirkabir University of Technology, Hafez Ave., Tehran 15875-4413 (Iran, Islamic Republic of)

2015-12-15

Highlights: • Three combined advanced SBR and enhanced Fenton process as post treatment was compared. • Higher biomass concentration, dye, COD and metabolites removal was presented together. • Pseudo zero and pseudo first-order bio-decolorization kinetics were observed in all SBRs. • High reduction of AR18 to intermediate metabolites was monitored by HPLC. - Abstract: The purpose of this research was to compare three combined sequencing batch reactor (SBR) – Fenton processes as post-treatment for the treatment of azo dye Acid Red 18 (AR18). Three combined treatment systems (CTS1, CTS2 and CTS3) were operated to investigate the biomass concentration, COD removal, AR18 dye decolorization and kinetics study. The MLSS concentration of CTS2 reached 7200 mg/L due to the use of external feeding in the SBR reactor of CTS2. The COD concentration remained 273 mg/L and 95 mg/L (initial COD = 3270 mg/L) at the end of alternating anaerobic–aerobic SBR with external feeding (An-A MSBR) and CTS2, respectively, resulting in almost 65% of Fenton process efficiency. The dye concentration of 500 mg/L was finally reduced to less than 10 mg/L in all systems indicating almost complete AR18 decolorization, which was also confirmed by UV–vis analysis. The dye was removed following two successive parts as parts 1 and 2 with pseudo zero-order and pseudo first-order kinetics, respectively, in all CTSs. Higher intermediate metabolites degradation was obtained using HPLC analysis in CTS2. Accordingly, a combined treatment system can be proposed as an appropriate and environmentally-friendly system for the treatment of the azo dye AR18 in wastewater.

Recent results of three-dimensional CFD simulations of coolant mixing in VVER-440/213 reactor pressure vessel

International Nuclear Information System (INIS)

Kiss, B.; Boros, I.; Aszodi, A.

2008-01-01

The Budapest University of Technology and Economics, Institute of Nuclear Techniques has been working since 2001 on the three-dimensional CFD model of the reactor pressure vessel of the VVER-440 type reactor. During this time period - due to the development of the available computational capacity - a very complex and detailed model of the RPV has been developed. The aim of the construction of the new model is to describe further internal structures of the RPV (e.g. correct modeling of brake tubes, or internals in the upper mixing chamber) and to perform an extensive sensitivity analysis on the different modeling and calculation parameters (e.g. porous region models vs. detailed modeling, or n different turbulence models). The new model can be applied for steady state calculation during normal operational condition and for different transient analyses as well. One interesting application is the participation in a planned benchmark exercise on the start-up of the sixth main coolant pump, which is aimed to compare the capabilities of mixing models of one-dimensional system codes with the results of CFD simulation. (authors)

Handwriting: three-dimensional kinetic synergies in circle drawing movements.

Science.gov (United States)

Hooke, Alexander W; Karol, Sohit; Park, Jaebum; Kim, Yoon Hyuk; Shim, Jae Kun

2012-07-01

The purpose of this study was to investigate central nervous system (CNS) strategies for controlling multifinger forces during a circle-drawing task. Subjects drew 30 concentric, discontinuous clockwise and counter clockwise circles, at self and experimenter-set paces. The three-dimensional trajectory of the pen's center of mass and the three-dimensional forces and moments of force at each contact between the hand and the pen were recorded. Uncontrolled Manifold Analysis was used to quantify the synergies between pen-hand contact forces in radial, tangential and vertical directions. Results showed that synergies in the radial and tangential components were significantly stronger than in the vertical component. Synergies in the clockwise direction were significantly stronger than the counterclockwise direction in the radial and vertical components. Pace was found to be insignificant under any condition.

Thermal hydraulic and neutron kinetic simulation of the Angra 2 reactor using a RELAP5/PARCS coupled model

Energy Technology Data Exchange (ETDEWEB)

Reis, Patricia A.L.; Costa, Antonella L.; Hamers, Adolfo R.; Pereira, Claubia; Rodrigues, Thiago D.A.; Mantecon, Javier G.; Veloso, Maria A.F., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: adolforomerohamers@hotmail.com, E-mail: claubia@nuclear.ufmg.br, E-mail: thiagodanielbh@gmail.com, E-mail: mantecon1987@gmail.com, E-mail: dora@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq), Belo Horizonte (Brazil); Miro, Rafael; Verdu, Gumersindo, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Universidad Politecnica de Valencia (Spain). Departamento de Ingenieria Quimica y Nuclear

2015-07-01

The computational advances observed in the last two decades have been provided direct impact on the researches related to nuclear simulations, which use several types of computer codes, including coupled between them, allowing representing with very accuracy the behavior of nuclear plants. Studies of complex scenarios in nuclear reactors have been improved by the use of thermal-hydraulic (TH) and neutron kinetics (NK) coupled codes. This technique consists in incorporating three-dimensional (3D) neutron modeling of the reactor core into codes, mainly to simulate transients that involve asymmetric core spatial power distributions and strong feedback effects between neutronics and reactor thermal-hydraulics. Therefore, this work presents preliminary results of TH RELAP5 and the NK PARCS calculations applied to model of the Angra 2 reactor. The WIMSD-5B code has been used to generate the macroscopic cross sections used in the NK code. The results obtained are satisfactory and represent important part of the development of this methodology. The next step is to couple the codes. (author)

Determination of two dimensional axisymmetric finite element model for reactor coolant piping nozzles

International Nuclear Information System (INIS)

Choi, S. N.; Kim, H. N.; Jang, K. S.; Kim, H. J.

2000-01-01

The purpose of this paper is to determine a two dimensional axisymmetric model through a comparative study between a three dimensional and an axisymmetric finite element analysis of the reactor coolant piping nozzle subject to internal pressure. The finite element analysis results show that the stress adopting the axisymmetric model with the radius of equivalent spherical vessel are well agree with that adopting the three dimensional model. The radii of equivalent spherical vessel are 3.5 times and 7.3 times of the radius of the reactor coolant piping for the safety injection nozzle and for the residual heat removal nozzle, respectively

Visualizing and quantifying dose distribution in a UV reactor using three-dimensional laser-induced fluorescence.

Science.gov (United States)

Gandhi, Varun N; Roberts, Philip J W; Kim, Jae-Hong

2012-12-18

Evaluating the performance of typical water treatment UV reactors is challenging due to the complexity in assessing spatial and temporal variation of UV fluence, resulting from highly unsteady, turbulent nature of flow and variation in UV intensity. In this study, three-dimensional laser-induced fluorescence (3DLIF) was applied to visualize and quantitatively analyze a lab-scale UV reactor consisting of one lamp sleeve placed perpendicular to flow. Mapping the spatial and temporal fluence delivery and MS2 inactivation revealed the highest local fluence in the wake zone due to longer residence time and higher UV exposure, while the lowest local fluence occurred in a region near the walls due to short-circuiting flow and lower UV fluence rate. Comparing the tracer based decomposition between hydrodynamics and IT revealed similar coherent structures showing the dependency of fluence delivery on the reactor flow. The location of tracer injection, varying the height and upstream distance from the lamp center, was found to significantly affect the UV fluence received by the tracer. A Lagrangian-based analysis was also employed to predict the fluence along specific paths of travel, which agreed with the experiments. The 3DLIF technique developed in this study provides new insight on dose delivery that fluctuates both spatially and temporally and is expected to aid design and optimization of UV reactors as well as validate computational fluid dynamics models that are widely used to simulate UV reactor performances.

An Innovative Three-Dimensional Heterogeneous Coarse-Mesh Transport Method for Advanced and Generation IV Reactor Core Analysis and Design

International Nuclear Information System (INIS)

Rahnema, Farzad

2009-01-01

This project has resulted in a highly efficient method that has been shown to provide accurate solutions to a variety of 2D and 3D reactor problems. The goal of this project was to develop (1) an accurate and efficient three-dimensional whole-core neutronics method with the following features: based solely on transport theory, does not require the use of cross-section homogenization, contains a highly accurate and self-consistent global flux reconstruction procedure, and is applicable to large, heterogeneous reactor models, and to (2) create new numerical benchmark problems for code cross-comparison.

Nuclear reactor kinetics and control

International Nuclear Information System (INIS)

Lewins, J.

1978-01-01

A consistent, integrated account of modern developments in the study of nuclear reactor kinetics and the problem of their efficient and safe control. It aims to prepare the student for advanced study and research or practical work in the field. Special features include treatments of noise theory, reliability theory and safety related studies. It covers all aspects of the operation and control of nuclear reactors, power and research and is complete in providing physical data methods of calculation and solution including questions of equipment reliability. The work uses illustrations of the main types of reactors in use in the UK, USA and Europe. Each chapter contains problems and worked examples suitable for course work and study. The subject is covered in chapters, entitled: introductory review; neutron and precursor equations; elementary solutions at low power; linear reactor process dynamics with feedback; power reactor control systems; fluctuations and reactor noise; safety and reliability; nonlinear systems (safety and control); analogue computing. (author)

Ethanol steam reforming kinetics of a Pd-Ag membrane reactor

Energy Technology Data Exchange (ETDEWEB)

Tosti, Silvano; Borelli, Rodolfo; Borgognoni, Fabio [ENEA, Dipartimento FPN, C.R. ENEA Frascati, Via E. Fermi 45, Frascati (RM) I-00044 (Italy); Basile, Angelo [Institute on Membrane Technology, ITM-CNR, c/o Univ. of Calabria, via P. Bucci, Cubo 17/C, 87030 Rende (CS) (Italy); Castelli, Stefano [ENEA, Dipartimento ACS, C.R. ENEA Casaccia, Via Anguillarese 301, Roma I-00123 (Italy); Fabbricino, Massimiliano; Licusati, Celeste [Dept. of Hydraulic and Environmental Engineering, Univ. of Naples Federico II, Via Claudio 21, Naples 80125 (Italy); Gallucci, Fausto [Fundamentals of Chemical Reaction Engineering Group, Faculty of Science and Technology, University of Twente, Enschede (Netherlands)

2009-06-15

The ethanol steam reforming reaction carried out in a Pd-based tubular membrane reactor has been modelled via a finite element code. The model considers the membrane tube divided into finite volume elements where the mass balances for both lumen and shell sides are carried out accordingly to the reaction and permeation kinetics. Especially, a simplified ''power law'' has been applied for the reaction kinetics: the comparison with experimental data obtained by using three different kinds of catalyst (Ru, Pt and Ni based) permitted defining the coefficients of the kinetics expression as well as to validate the model. Based on the Damkohler-Peclet analysis, the optimization of the membrane reformer has been also approached. (author)

Solution of the reactor point kinetics equations by MATLAB computing

Directory of Open Access Journals (Sweden)

Singh Sudhansu S.

2015-01-01

Full Text Available The numerical solution of the point kinetics equations in the presence of Newtonian temperature feedback has been a challenging issue for analyzing the reactor transients. Reactor point kinetics equations are a system of stiff ordinary differential equations which need special numerical treatments. Although a plethora of numerical intricacies have been introduced to solve the point kinetics equations over the years, some of the simple and straightforward methods still work very efficiently with extraordinary accuracy. As an example, it has been shown recently that the fundamental backward Euler finite difference algorithm with its simplicity has proven to be one of the most effective legacy methods. Complementing the back-ward Euler finite difference scheme, the present work demonstrates the application of ordinary differential equation suite available in the MATLAB software package to solve the stiff reactor point kinetics equations with Newtonian temperature feedback effects very effectively by analyzing various classic benchmark cases. Fair accuracy of the results implies the efficient application of MATLAB ordinary differential equation suite for solving the reactor point kinetics equations as an alternate method for future applications.

Application of hexagonal element scheme in finite element method to three-dimensional diffusion problem of fast reactors

International Nuclear Information System (INIS)

Ishiguro, Misako; Higuchi, Kenji

1983-01-01

The finite element method is applied in Galerkin-type approximation to three-dimensional neutron diffusion equations of fast reactors. A hexagonal element scheme is adopted for treating the hexagonal lattice which is typical for fast reactors. The validity of the scheme is verified by applying the scheme as well as alternative schemes to the neutron diffusion calculation of a gas-cooled fast reactor of actual scale. The computed results are compared with corresponding values obtained using the currently applied triangular-element and also with conventional finite difference schemes. The hexagonal finite element scheme is found to yield a reasonable solution to the problem taken up here, with some merit in terms of saving in computing time, but the resulting multiplication factor differs by 1% and the flux by 9% compared with the triangular mesh finite difference scheme. The finite element method, even in triangular element scheme, would appear to incur error in inadmissible amount and which could not be easily eliminated by refining the nodes. (author)

The study of two, three and four dimensional nonlinear dynamics of nuclear fission reactors and effective parameters on its behaviour

International Nuclear Information System (INIS)

Tajik, M.; Ghasemizad, A.

2008-01-01

In this research, new physical fission reactor parameters which have very sensitive effects on the qualitative behavior of a reactor, are introduced. Therefore, the two, the nonlinear dynamics of two, three and four dimensional, considering almost the effective parameters are formulated for describing nuclear fission reactor systems. Using both analytical and numerical methods, the stability and instability of the given dynamical equations and the conditions of stability are studied in these systems. We have shown that the two parameters of the mean energy residence time in fuel and coolant and also their ratios have the most qualitative effects on the dynamical behaviour of a typical nuclear fission reactor. Increasing or decreasing of these parameters from a captain limit can lead to stability or un stability in a given system

Ozone disintegration kinetics in the reactor for tyres decomposition

International Nuclear Information System (INIS)

Golota, V.I.; Manujlenko, O.V.; Taran, G.V.; Pis'menetskij, A.S.; Zamuriev, A.A.

2010-01-01

The results of theoretical and experimental research of ozone disintegration kinetics in the chemical reactor which is developed for decomposition of tyres in the ozone-air environment are presented. Analytical expression for dependence of ozone concentration in the reactor from time and from parameters of the task, such as volume speed of ozone-air mixture feed on a reactor input, concentration of ozone on the input to the reactor, volume speed of output of the used mixture, reactor size, and square of its internal surface is obtained. It is shown that at the same speed of ozone-air mixture pro rolling through the reactor, with growth of ozone concentration on the input, value of stationary concentration in the reactor grows, remaining always less than concentration on the input. It is also shown that at the same ozone concentration on the input, with growth of speed of ozone-air mixture pro rolling through the reactor, value of stationary ozone concentration in the reactor also grows, remaining always less than ozone concentration on the input. The ozone disintegration kinetics in the reactor in a wide range of speed of ozone-air mixture pro rolling through the reactor (0.15, 0.30, 0.45, 0.60 m3/hour) and various ozone concentration on the input (5, 10, 15, 20 g/m3) is experimentally studied. It is shown that experimental results with good accuracy coincide with the theoretical. Direct experiment showed the essential influence of the internal surface of the reactor on the ozone disintegration kinetics.

Hybrid simulation of reactor kinetics in CANDU reactors using a modal approach

International Nuclear Information System (INIS)

Monaghan, B.M.; McDonnell, F.N.; Hinds, H.W.T.; m.

1980-01-01

A hybrid computer model for simulating the behaviour of large CANDU (Canada Deuterium Uranium) reactor cores is presented. The main dynamic variables are expressed in terms of weighted sums of a base set of spatial natural-mode functions with time-varying co-efficients. This technique, known as the modal or synthesis approach, permits good three-dimensional representation of reactor dynamics and is well suited to hybrid simulation. The hybrid model provides improved man-machine interaction and real-time capability. The model was used in two applications. The first studies the transient that follows a loss of primary coolant and reactor shutdown; the second is a simulation of the dynamics of xenon, a fission product which has a high absorption cross-section for neutrons and thus has an important effect on reactor behaviour. Comparison of the results of the hybrid computer simulation with those of an all-digital one is good, within 1% to 2%

Simulation of A Main Steam Line Break Accident Using the Coupled 'System Thermal-Hydraulics, 3D reactor Kinetics, and Hot Channel' Analysis Capability of MARS 3.0

International Nuclear Information System (INIS)

Jeong, Jae Jun; Chung, Bub Dong

2005-09-01

For realistic analysis of thermal-hydraulics (T-H) transients in light water reactors, KAERI has developed the best-estimate T-H system code, MARS. The code has been improved from the consolidated version of the RELAP5/MOD3 and COBRA-TF codes. Then, the MARS code was coupled with a three-dimensional (3-D) reactor kinetics code, MASTER. This coupled calculation feature, in conjunction with the existing hot channel analysis capabilities of the MARS and MASTER codes, allows for more realistic simulations of nuclear system transients. In this work, a main steam line break (MSLB) accident is simulated using the coupled 'system T-H, 3-D reactor kinetics, and hot channel analysis' feature of the MARS code. Two coupled calculations are performed for demonstration. First, a coupled calculation of the 'system T-H and 3-D reactor kinetics' with a refined core T-H nodalization is carried out to obtain global core power and local departure from nucleate boiling (DNB) ratio (DNBR) behaviors. Next, for a more accurate DNBR prediction, another coupled calculation with subchannel meshes for the hot channels is performed. The results of the coupled calculations are very reasonable and consistent so that these can be used to remove the excessive conservatism in the conventional safety analysis

Electrochemical in situ regeneration of granular activated carbon using a three-dimensional reactor.

Science.gov (United States)

Sun, Hong; Liu, Zhigang; Wang, Ying; Li, Yansheng

2013-12-01

Electrochemical in situ regeneration of granular activated carbon (GAC) saturated with phenol was experimentally investigated using a three-dimensional electrode reactor with titanium filter electrode arrays. The feasibility of the electrochemical regeneration has been assessed by monitoring the regeneration efficiency and chemical oxygen demand (COD). The influence of the applied current, the effluent flow rate, and the effluent path of the electrochemical cell have been systematically studied. Under the optimum conditions, the regeneration efficiency of GAC could reach 94% in 2 hr, and no significant declination was observed after five-time continuous adsorption-regeneration cycles. The adsorption of organic pollutants was almost completely mineralized due to electrochemical oxidation, indicating that this regeneration process is much more potentially cost-effective for application. Copyright © 2013 The Research Centre for Eco-Environmental Sciences, Chinese Academy of Sciences. Published by Elsevier B.V. All rights reserved.

Three-dimensional cooling of muons

CERN Document Server

Vsevolozhskaya, T A

2000-01-01

The simultaneous ionization cooling of muon beams in all three - the longitudinal and two transverse - directions is considered in a scheme, based on bent lithium lenses with dipole constituent of magnetic field in them, created by a special configuration of current-carrying rod. An analysis of three-dimensional cooling is performed with the use of kinetic equation method. Results of numerical calculation for a specific beam line configuration are presented together with results of computer simulation using the Moliere distribution to describe the Coulomb scattering and the Vavilov distribution used to describe the ionization loss of energy.

Generalized Runge-Kutta method for two- and three-dimensional space-time diffusion equations with a variable time step

International Nuclear Information System (INIS)

Aboanber, A.E.; Hamada, Y.M.

2008-01-01

An extensive knowledge of the spatial power distribution is required for the design and analysis of different types of current-generation reactors, and that requires the development of more sophisticated theoretical methods. Therefore, the need to develop new methods for multidimensional transient reactor analysis still exists. The objective of this paper is to develop a computationally efficient numerical method for solving the multigroup, multidimensional, static and transient neutron diffusion kinetics equations. A generalized Runge-Kutta method has been developed for the numerical integration of the stiff space-time diffusion equations. The method is fourth-order accurate, using an embedded third-order solution to arrive at an estimate of the truncation error for automatic time step control. In addition, the A(α)-stability properties of the method are investigated. The analyses of two- and three-dimensional benchmark problems as well as static and transient problems, demonstrate that very accurate solutions can be obtained with assembly-sized spatial meshes. Preliminary numerical evaluations using two- and three-dimensional finite difference codes showed that the presented generalized Runge-Kutta method is highly accurate and efficient when compared with other optimized iterative numerical and conventional finite difference methods

Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in transient state condition

International Nuclear Information System (INIS)

Zhou, Jianjun; Zhang, Daling; Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei

2015-01-01

Highlights: • Developed a three dimensional neutronic/thermal-hydraulic coupled transient analysis code for MSR. • Investigated the neutron distribution and thermal-hydraulic characters of the core under transient condition. • Analyzed three different transient conditions of inlet temperature drop, reactivity jump and pump coastdown. - Abstract: MSR (molten salt reactor) use liquid molten salt as coolant and fuel solvent, which was the only one liquid reactor of six Generation IV reactor types. As a liquid reactor the physical property of reactor was significantly influenced by fuel salt flow and the conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code investigated the neutronics and thermo-hydraulics characteristics of the core in transient condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The code was separately validated by neutron benchmark and flow and heat transfer benchmark. Three different transient conditions was analyzed with inlet temperature drop, reactivity jump and pump coastdown. The results provide some valuable information in design and research this kind of reactor

Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in transient state condition

Energy Technology Data Exchange (ETDEWEB)

Zhou, Jianjun [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China); College of Mechanical and Power Engineering, China Three Gorges University, No 8, Daxue road, Yichang, Hubei 443002 (China); Zhang, Daling, E-mail: dlzhang@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China); Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China)

2015-02-15

Highlights: • Developed a three dimensional neutronic/thermal-hydraulic coupled transient analysis code for MSR. • Investigated the neutron distribution and thermal-hydraulic characters of the core under transient condition. • Analyzed three different transient conditions of inlet temperature drop, reactivity jump and pump coastdown. - Abstract: MSR (molten salt reactor) use liquid molten salt as coolant and fuel solvent, which was the only one liquid reactor of six Generation IV reactor types. As a liquid reactor the physical property of reactor was significantly influenced by fuel salt flow and the conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code investigated the neutronics and thermo-hydraulics characteristics of the core in transient condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The code was separately validated by neutron benchmark and flow and heat transfer benchmark. Three different transient conditions was analyzed with inlet temperature drop, reactivity jump and pump coastdown. The results provide some valuable information in design and research this kind of reactor.

Development and validation of three-dimensional CFD techniques for reactor safety applications. Final report; Entwicklung und Validierung dreidimensionaler CFD Verfahren fuer Anwendungen in der Reaktorsicherheit. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Buchholz, Sebastian; Palazzo, Simone; Papukchiev, Angel; Scheurer Martina

2016-12-15

The overall goal of the project RS 1506 ''Development and Validation of Three Dimensional CFD Methods for Reactor Safety Applications'' is the validation of Computational Fluid Dynamics (CFD) software for the simulation of three -dimensional thermo-hydraulic heat and fluid flow phenomena in nuclear reactors. For this purpose a wide spectrum of validation and test cases was selected covering fluid flow and heat transfer phenomena in the downcomer and in the core of pressurized water reactors. In addition, the coupling of the system code ATHLET with the CFD code ANSYS CFX was further developed and validated. The first choice were UPTF experiments where turbulent single- and two-phase flows were investigated in a 1:1 scaled model of a German KONVOI reactor. The scope of the CFD calculations covers thermal mixing and stratification including condensation in single- and two-phase flows. In the complex core region, the flow in a fuel assembly with spacer grid was simulated as defined in the OECD/NEA Benchmark MATIS-H. Good agreement are achieved when the geometrical and physical boundary conditions were reproduced as realistic as possible. This includes, in particular, the consideration of heat transfer to walls. The influence of wall modelling on CFD results was investigated on the TALL-3D T01 experiment. In this case, the dynamic three dimensional fluid flow and heat transfer phenomena were simulated in a Generation IV liquid metal cooled reactor. Concurrently to the validation work, the coupling of the system code ATHLET with the ANSYS CFX software was optimized and expanded for two-phase flows. Different coupling approaches were investigated, in order to overcome the large difference between CPU-time requirements of system and CFD codes. Finally, the coupled simulation system was validated by applying it to the simulation of the PSI double T-junction experiment, the LBE-flow in the MYRRA Spallation experiment and a demonstration test case

Numerical simulations of subcritical reactor kinetics in thermal hydraulic transient phases

Energy Technology Data Exchange (ETDEWEB)

Yoo, J; Park, W S [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1999-12-31

A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute (KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons, external neutrons form spallation reactions are essentially required for operating the reactor in its steady state. Furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance to the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases. 5 refs., 8 figs. (Author)

Numerical simulations of subcritical reactor kinetics in thermal hydraulic transient phases

Energy Technology Data Exchange (ETDEWEB)

Yoo, J.; Park, W. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute (KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons, external neutrons form spallation reactions are essentially required for operating the reactor in its steady state. Furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance to the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases. 5 refs., 8 figs. (Author)

FLICA-4 (version 1). A computer code for three dimensional thermal analysis of nuclear reactor cores

International Nuclear Information System (INIS)

Raymond, P.; Allaire, G.; Boudsocq, G.; Caruge, D.; Gramont, T. de; Toumi, I.

1995-01-01

FLICA-4 is a thermal-hydraulic computer code, developed at the French Atomic Energy Commission (CEA) for three-dimensional steady-state or transient two-phase flow, and aimed at design and safety thermal analysis of nuclear reactor cores. It is available for various UNIX workstations and CRAY computers under UNICOS.It is based on four balance equations which include three balance equations for the mixture and a mass balance equation for the less concentrated phase which allows for the calculation of non equilibrium flows such as sub-cooled boiling and superheated steam. A drift velocity model takes into account the velocity unbalance between phases. The equations are solved using a finite volume numerical scheme. Typical running time, specific features (coupling with other codes) and auxiliary programs are presented. 1 tab., 9 refs

Liquid phase electro epitaxy growth kinetics of GaAs-A three-dimensional numerical simulation study

International Nuclear Information System (INIS)

Mouleeswaran, D.; Dhanasekaran, R.

2006-01-01

A three-dimensional numerical simulation study for the liquid phase electro epitaxial growth kinetic of GaAs is presented. The kinetic model is constructed considering (i) the diffusive and convective mass transport, (ii) the heat transfer due to thermoelectric effects such as Peltier effect, Joule effect and Thomson effect, (iii) the electric current distribution with electromigration and (iv) the fluid flow coupled with concentration and temperature fields. The simulations are performed for two configurations namely (i) epitaxial growth from the arsenic saturated gallium rich growth solution, i.e., limited solution model and (ii) epitaxial growth from the arsenic saturated gallium rich growth solution with polycrystalline GaAs feed. The governing equations of liquid phase electro epitaxy are solved numerically with appropriate initial and boundary conditions using the central difference method. Simulations are performed to determine the following, a concentration profiles of solute atoms (As) in the Ga-rich growth solution, shape of the substrate evolution, the growth rate of the GaAs epitaxial film, the contributions of Peltier effect and electromigration of solute atoms to the growth with various experimental growth conditions. The growth rate is found to increase with increasing growth temperature and applied current density. The results are discussed in detail

Scaled-up electrochemical reactor with a fixed bed three-dimensional cathode for electro-Fenton process: Application to the treatment of bisphenol A

International Nuclear Information System (INIS)

Chmayssem, Ayman; Taha, Samir; Hauchard, Didier

2017-01-01

In this study, we report on the development of an open undivided electrochemical reactor with a compact fixed bed of glassy carbon pellets as three-dimensional cathode for the application of electro-Fenton process. Bisphenol A (BPA) was chosen as model molecule in order to improve its efficiency to the treatment of persistent pollutants. The study of the BPA removal efficiency in function of the applied current intensity was investigated in order to determine the limiting current of O 2 reduction (optimal conditions of H 2 O 2 production at flow rate of 0.36 m 3 .h −1 ) which was 0.8 A (0.5 A/100 g of glassy carbon pellets). Many parameters have been carried out using this electro-Fenton reactor namely degradation kinetics, influence of anodic reactions on DSA, effect of initial pollutant concentration. In the optimal current condition, the global production rate of H 2 O 2 and ·OH was investigated. The yield of electro-Fenton reaction (conversion of H 2 O 2 to ·OH) was very high (> 90%). The absolute rate of BPA degradation was determined as 4.3 × 10 9 M −1 s −1 . COD, TOC and BOD 5 measurements indicated that only few minutes of treatment by electro-Fenton process were needed to eliminate BPA for dilute solutions (10 and 25 mg.L −1 ). In this case, the biodegradability of the treated solutions occurred rapidly. For higher concentration levels, an efficient removal of BPA appeared for treatment time higher than 1 hour and more than 90 minutes were necessary to obtain the biodegradability of BPA solutions. In optimum conditions, the scale-up of the electrochemical reactor applied to electro-Fenton process was suggested and depended on the concentration level of the pollutant. The operating parameters of the scaled-up reactor might be deduced from the new section of each fixed bed exposed to the flow, from values of liquid flow velocity and from the corresponding limiting current density obtained with the reactor at laboratory scale. The compact fixed bed

Simulation of A Main Steam Line Break Accident Using the Coupled 'System Thermal-Hydraulics, 3D reactor Kinetics, and Hot Channel' Analysis Capability of MARS 3.0

Energy Technology Data Exchange (ETDEWEB)

Jeong, Jae Jun; Chung, Bub Dong

2005-09-15

For realistic analysis of thermal-hydraulics (T-H) transients in light water reactors, KAERI has developed the best-estimate T-H system code, MARS. The code has been improved from the consolidated version of the RELAP5/MOD3 and COBRA-TF codes. Then, the MARS code was coupled with a three-dimensional (3-D) reactor kinetics code, MASTER. This coupled calculation feature, in conjunction with the existing hot channel analysis capabilities of the MARS and MASTER codes, allows for more realistic simulations of nuclear system transients. In this work, a main steam line break (MSLB) accident is simulated using the coupled 'system T-H, 3-D reactor kinetics, and hot channel analysis' feature of the MARS code. Two coupled calculations are performed for demonstration. First, a coupled calculation of the 'system T-H and 3-D reactor kinetics' with a refined core T-H nodalization is carried out to obtain global core power and local departure from nucleate boiling (DNB) ratio (DNBR) behaviors. Next, for a more accurate DNBR prediction, another coupled calculation with subchannel meshes for the hot channels is performed. The results of the coupled calculations are very reasonable and consistent so that these can be used to remove the excessive conservatism in the conventional safety analysis.

Application of the reactor kinetics equations to the reactor safety analysis

International Nuclear Information System (INIS)

Sdouz, G.

1976-01-01

The reactor kinetics equations which can be solved by the computer program AIREK-III are used to describe the behavior of fast reactivity transients. By supplementing this computer program it was possible to solve additional safety problems, e.g. the course of reactor excursions induced by any form of reactivity input, the control of reactivity input as a function of a threshold-energy and the computation of produced energy. (author)

STAB: A kinetic, three-dimensional, one-group digital computer program

International Nuclear Information System (INIS)

Curtis, A.R.; Tyror, J.G.; Wrigley, H.E.

1961-10-01

A computer program for solving the one-group, time dependent, three-dimensional diffusion equation together with auxiliary equations representing heat transfer, xenon production and control rod movements, is presented. The equations and the methods of solution are discussed. (author)

Recursive solutions for multi-group neutron kinetics diffusion equations in homogeneous three-dimensional rectangular domains with time dependent perturbations

Energy Technology Data Exchange (ETDEWEB)

Petersen, Claudio Z. [Universidade Federal de Pelotas, Capao do Leao (Brazil). Programa de Pos Graduacao em Modelagem Matematica; Bodmann, Bardo E.J.; Vilhena, Marco T. [Universidade Federal do Rio Grande do Sul, Porto Alegre, RS (Brazil). Programa de Pos-graduacao em Engenharia Mecanica; Barros, Ricardo C. [Universidade do Estado do Rio de Janeiro, Nova Friburgo, RJ (Brazil). Inst. Politecnico

2014-12-15

In the present work we solve in analytical representation the three dimensional neutron kinetic diffusion problem in rectangular Cartesian geometry for homogeneous and bounded domains for any number of energy groups and precursor concentrations. The solution in analytical representation is constructed using a hierarchical procedure, i.e. the original problem is reduced to a problem previously solved by the authors making use of a combination of the spectral method and a recursive decomposition approach. Time dependent absorption cross sections of the thermal energy group are considered with step, ramp and Chebyshev polynomial variations. For these three cases, we present numerical results and discuss convergence properties and compare our results to those available in the literature.

Kinetics of propionate conversion in anaerobic continuously stirred tank reactors

DEFF Research Database (Denmark)

Bangsø Nielsen, Henrik; Mladenovska, Zuzana; Ahring, Birgitte Kiær

2008-01-01

The kinetic parameters of anaerobic propionate degradation by biomass from 7 continuously stirred tank reactors differing in temperature, hydraulic retention time and substrate composition were investigated. In substrate-depletion experiments (batch) the maximum propionate degradation rate, A......-m, was estimated. The results demonstrate that the rate of endogenous substrate (propionate) production should be taken into account when estimating kinetic parameters in biomass from manure-based anaerobic reactors....

A barrier on the public communication of nuclear technology. How to interpret reactor kinetics

International Nuclear Information System (INIS)

Yamamoto, Akio

2007-01-01

Reactor kinetics is very important to explain the safety of nuclear reactors. However, its description is somewhat complicated and not intuitive. In order to give more intuitive explanation for reactor kinetics, some metaphors that try to capture the feature of reactor behavior are discussed. (author)

Application of the exact distribution pjk in the determination of kinetic parameters in a reactor

International Nuclear Information System (INIS)

Alcala Ruiz, F.

1982-01-01

In this report one distribution of neutron counts obtained by a detector placed in a reactor is studied in order to be used in the determination of reactor kinetic parameters such as β/Λ and reactivities. The parameters accuracy from this new method is compared with the Feynman and Mogilner method, based too in Reactor Neutron Noise Analysis. These three methods have been applied to JEN-2 reactor and the better accuracy and faster collection of experimental data give some interest to the new method which only requires a good footing code. (Author) 68 refs

Continuous electrochemical oxidation of methyl orange waste water using a three-dimensional electrode reactor.

Science.gov (United States)

Liu, Zhigang; Wang, Feifei; Li, Yansheng; Xu, Tianlong; Zhu, Shaomin

2011-06-01

The removal of methyl orange wastewater was experimentally investigated using a three-dimensional electrode reactor with granular activated carbon and titanium filter electrodes arrays. The effects of the electric current, the residence time and the initial dye concentration on the methyl orange removal were evaluated. For the initial concentration of 1150 mg/L, the COD removal was obtained as 90% under the conditions of electric current 2 A, residence time 40 min. The effluent path of the electrochemical cell was optimized, using the anode effluent instead of the top effluent, where the COD removal was increased to 93% and the corresponding energy consumption was decreased from 15.5 to 14.6 kW-hr/kg COD. Copyright © 2011 The Research Centre for Eco-Environmental Sciences, Chinese Academy of Sciences. Published by Elsevier B.V. All rights reserved.

FLICA-4 (version 1) a computer code for three dimensional thermal analysis of nuclear reactor cores

International Nuclear Information System (INIS)

Raymond, P.; Allaire, G.; Boudsocq, G.

1995-01-01

FLICA-4 is a thermal-hydraulic computer code developed at the French Energy Atomic Commission (CEA) for three dimensional steady state or transient two phase flow for design and safety thermal analysis of nuclear reactor cores. The two phase flow model of FLICA-4 is based on four balance equations for the fluid which includes: three balance equations for the mixture and a mass balance equation for the less concentrated phase which permits the calculation of non-equilibrium flows as sub cooled boiling and superheated steam. A drift velocity model takes into account the velocity disequilibrium between phases. The thermal behaviour of fuel elements can be computed by a one dimensional heat conduction equation in plane, cylindrical or spherical geometries and coupled to the fluid flow calculation. Convection and diffusion of solution products which are transported either by the liquid or by the gas, can be evaluated by solving specific mass conservation equations. A one dimensional two phase flow model can also be used to compute 1-D flow in pipes, guide tubes, BWR assemblies or RBMK channels. The FLICA-4 computer code uses fast running time steam-water functions. Phasic and saturation physical properties are computed by using bi-cubic spline functions. Polynomial coefficients are tabulated from 0.1 to 22 MPa and 0 to 800 degrees C. Specific modules can be utilised in order to generate the spline coefficients for any other fluid properties

Three dimensional numerical simulation of a full scale CANDU reactor moderator to study temperature fluctuations

International Nuclear Information System (INIS)

Sarchami, Araz; Ashgriz, Nasser; Kwee, Marc

2014-01-01

Highlights: • 3D model of a Candu reactor is modeled to investigate flow distribution. • The results show the temperature distribution is not symmetrical. • Temperature contours show the hot regions at the top left-hand side of the tank. • Interactions of momentum flows and buoyancy flows create circulation zones. • The results indicate that the moderator tank operates in the buoyancy driven mode. -- Abstract: Three dimensional numerical simulations are conducted on a full scale CANDU Moderator and transient variations of the temperature and velocity distributions inside the tank are determined. The results show that the flow and temperature distributions inside the moderator tank are three dimensional and no symmetry plane can be identified. Competition between the upward moving buoyancy driven flows and the downward moving momentum driven flows in the center region of the tank, results in the formation of circulation zones. The moderator tank operates in the buoyancy driven mode and any small disturbances in the flow or temperature makes the system unstable and asymmetric. Different types of temperature fluctuations are noted inside the tank: (i) large amplitude are at the boundaries between the hot and cold; (ii) low amplitude are in the core of the tank; (iii) high frequency fluctuations are in the regions with high velocities and (iv) low frequency fluctuations are in the regions with lower velocities

A Multi-Physics simulation of the Reactor Core using CUPID/MASTER

International Nuclear Information System (INIS)

Lee, Jae Ryong; Cho, Hyoung Kyu; Yoon, Han Young; Cho, Jin Young; Jeong, Jae Jun

2011-01-01

KAERI has been developing a component-scale thermal hydraulics code, CUPID. The aim of the code is for multi-dimensional, multi-physics and multi-scale thermal hydraulics analysis. In our previous papers, the CUPID code has proved to be able to reproduce multidimensional thermal hydraulic analysis by validated with various conceptual problems and experimental data. For the numerical closure, it adopts a three dimensional, transient, two-phase and three-field model, and includes physical models and correlations of the interfacial mass, momentum, and energy transfer. For the multi-scale analysis, the CUPID is on progress to merge into system-scale thermal hydraulic code, MARS. In the present paper, a multi-physics simulation was performed by coupling the CUPID with three dimensional neutron kinetics code, MASTER. The MASTER is merged into the CUPID as a dynamic link library (DLL). The APR1400 reactor core during control rod drop/ejection accident was simulated as an example by adopting a porous media approach to employ fuel assembly. The following sections present the numerical modeling for the reactor core, coupling of the kinetics code, and the simulation results

A two-point kinetic model for the PROTEUS reactor

International Nuclear Information System (INIS)

Dam, H. van.

1995-03-01

A two-point reactor kinetic model for the PROTEUS-reactor is developed and the results are described in terms of frequency dependent reactivity transfer functions for the core and the reflector. It is shown that at higher frequencies space-dependent effects occur which imply failure of the one-point kinetic model. In the modulus of the transfer functions these effects become apparent above a radian frequency of about 100 s -1 , whereas for the phase behaviour the deviation from a point model already starts at a radian frequency of 10 s -1 . (orig.)

Reactor kinetics methods development. Final report

International Nuclear Information System (INIS)

Hansen, K.F.; Henry, A.F.

1978-01-01

This report is a qualitative summary of research conducted at MIT from 1967 to 1977 in the area of reactor kinetics methods. The objectives of the research were to find methods of integration of various mathematical models of nuclear reactor transients. From the beginning the work was aimed at numerical integration methods. Specific areas of research, discussed in more detail following, included: integration of multigroup diffusion theory models by finite difference and finite element methods; response matrix and nodal methods; coarse-mesh homogenization; and special treatment of boundary conditions

A high-order method for the integration of the Galerkin semi-discretized nuclear reactor kinetics equations

International Nuclear Information System (INIS)

Vargas, L.

1988-01-01

The numerical approximate solution of the space-time nuclear reactor kinetics equation is investigated using a finite-element discretization of the space variable and a high order integration scheme for the resulting semi-discretized parabolic equation. The Galerkin method with spatial piecewise polynomial Lagrange basis functions are used to obtained a continuous time semi-discretized form of the space-time reactor kinetics equation. A temporal discretization is then carried out with a numerical scheme based on the Iterated Defect Correction (IDC) method using piecewise quadratic polynomials or exponential functions. The kinetics equations are thus solved with in a general finite element framework with respect to space as well as time variables in which the order of convergence of the spatial and temporal discretizations is consistently high. A computer code GALFEM/IDC is developed, to implement the numerical schemes described above. This issued to solve a one space dimensional benchmark problem. The results of the numerical experiments confirm the theoretical arguments and show that the convergence is very fast and the overall procedure is quite efficient. This is due to the good asymptotic properties of the numerical scheme which is of third order in the time interval

Study of carbon dioxide gas treatment based on equations of kinetics in plasma discharge reactor

Science.gov (United States)

Abedi-Varaki, Mehdi

2017-08-01

Carbon dioxide (CO2) as the primary greenhouse gas, is the main pollutant that is warming earth. CO2 is widely emitted through the cars, planes, power plants and other human activities that involve the burning of fossil fuels (coal, natural gas and oil). Thus, there is a need to develop some method to reduce CO2 emission. To this end, this study investigates the behavior of CO2 in dielectric barrier discharge (DBD) plasma reactor. The behavior of different species and their reaction rates are studied using a zero-dimensional model based on equations of kinetics inside plasma reactor. The results show that the plasma reactor has an effective reduction on the CO2 density inside the reactor. As a result of reduction in the temporal variations of reaction rate, the speed of chemical reactions for CO2 decreases and very low concentration of CO2 molecules inside the plasma reactor is generated. The obtained results are compared with the existing experimental and simulation findings in the literature.

COMPUTATIONAL AND EXPERIMENTAL MODELING OF THREE-PHASE SLURRY-BUBBLE COLUMN REACTOR

Energy Technology Data Exchange (ETDEWEB)

Isaac K. Gamwo; Dimitri Gidaspow

1999-09-01

Considerable progress has been achieved in understanding three-phase reactors from the point of view of kinetic theory. In a paper in press for publication in Chemical Engineering Science (Wu and Gidaspow, 1999) we have obtained a complete numerical solution of bubble column reactors. In view of the complexity of the simulation a better understanding of the processes using simplified analytical solutions is required. Such analytical solutions are presented in the attached paper, Large Scale Oscillations or Gravity Waves in Risers and Bubbling Beds. This paper presents analytical solutions for bubbling frequencies and standing wave flow patterns. The flow patterns in operating slurry bubble column reactors are not optimum. They involve upflow in the center and downflow at the walls. It may be possible to control flow patterns by proper redistribution of heat exchangers in slurry bubble column reactors. We also believe that the catalyst size in operating slurry bubble column reactors is not optimum. To obtain an optimum size we are following up on the observation of George Cody of Exxon who reported a maximum granular temperature (random particle kinetic energy) for a particle size of 90 microns. The attached paper, Turbulence of Particles in a CFB and Slurry Bubble Columns Using Kinetic Theory, supports George Cody's observations. However, our explanation for the existence of the maximum in granular temperature differs from that proposed by George Cody. Further computer simulations and experiments involving measurements of granular temperature are needed to obtain a sound theoretical explanation for the possible existence of an optimum catalyst size.

A simplified, coarse-mesh, three-dimensional diffusion scheme for calculating the gross power distribution in a boiling water reactor

International Nuclear Information System (INIS)

Borresen, S.

1995-01-01

A simplified, finite-difference diffusion scheme for a three-dimensional calculation of the gross power distribution in the core of a boiling water reactor (BWR) is presented. Results obtained in a series of one- and two-dimensional test cases indicate that this method may be of sufficient accuracy and simplicity for implementation in BWR-simulator computer programs. Computer requirements are very modest; thus, only 3N memory locations are required for in-core treatment of the inner iteration in the solution of a problem with N mesh points. The mesh width may be chosen equal to the fuel assembly pitch. Input data are in the form of conventional 2-group diffusion parameters. It is concluded that the method presented has definite advantages in comparison with the nodal coupling method. (author)

Three dimensional illustrating - three-dimensional vision and deception of sensibility

Directory of Open Access Journals (Sweden)

Anita Gánóczy

2009-03-01

Full Text Available The wide-spread digital photography and computer use gave the opportunity for everyone to make three-dimensional pictures and to make them public. The new opportunities with three-dimensional techniques give chance for the birth of new artistic photographs. We present in detail the biological roots of three-dimensional visualization, the phenomena of movement parallax, which can be used efficiently in making three-dimensional graphics, the Zöllner- and Corridor-illusion. There are present in this paper the visual elements, which contribute to define a plane two-dimensional image in three-dimension: coherent lines, the covering, the measurement changes, the relative altitude state, the abatement of detail profusion, the shadings and the perspective effects of colors.

Three-dimensional Kinetic Pulsar Magnetosphere Models: Connecting to Gamma-Ray Observations

Science.gov (United States)

Kalapotharakos, Constantinos; Brambilla, Gabriele; Timokhin, Andrey; Harding, Alice K.; Kazanas, Demosthenes

2018-04-01

We present three-dimensional (3D) global kinetic pulsar magnetosphere models, where the charged particle trajectories and the corresponding electromagnetic fields are treated self-consistently. For our study, we have developed a Cartesian 3D relativistic particle-in-cell code that incorporates radiation reaction forces. We describe our code and discuss the related technical issues, treatments, and assumptions. Injecting particles up to large distances in the magnetosphere, we apply arbitrarily low to high particle injection rates, and obtain an entire spectrum of solutions from close to the vacuum-retarded dipole to close to the force-free (FF) solution, respectively. For high particle injection rates (close to FF solutions), significant accelerating electric field components are confined only near the equatorial current sheet outside the light cylinder. A judicious interpretation of our models allows the particle emission to be calculated, and consequently, the corresponding realistic high-energy sky maps and spectra to be derived. Using model parameters that cover the entire range of spin-down powers of Fermi young and millisecond pulsars, we compare the corresponding model γ-ray light curves, cutoff energies, and total γ-ray luminosities with those observed by Fermi to discover a dependence of the particle injection rate, { \\mathcal F }, on the spin-down power, \\dot{{ \\mathcal E }}, indicating an increase of { \\mathcal F } with \\dot{{ \\mathcal E }}. Our models, guided by Fermi observations, provide field structures and particle distributions that are not only consistent with each other but also able to reproduce a broad range of the observed γ-ray phenomenologies of both young and millisecond pulsars.

Approximation model of three-dimensional power distribution in boiling water reactor using neural networks

International Nuclear Information System (INIS)

Kobayashi, Yoko; Aiyoshi, Eitaro

2001-01-01

Fast and accurate prediction of three-dimensional (3D) power distribution is essential in a boiling water reactor (BWR). The prediction method of 3D power distribution in BWR is developed using the neural network. Application of the neural network starts with selecting the learning algorithm. In the proposed method, we use the learning algorithms based on a class of Quasi-Newton optimization techniques called Self-Scaling Variable Metric (SSVM) methods. Prediction studies were done for a core of actual BWR plant with octant symmetry. Compared to classical Quasi-Newton methods, it is shown that the SSVM method reduces the number of iterations in the learning mode. The results of prediction demonstrate that the neural network can predict 3D power distribution of BWR reasonably well. The proposed method will be very useful for BWR loading pattern optimization problems where 3D power distribution for a huge number of loading patterns (LPs) must be performed. (author)

Nuclear characteristic simulation device for reactor core

International Nuclear Information System (INIS)

Arakawa, Akio; Kobayashi, Yuji.

1994-01-01

In a simulation device for nuclear characteristic of a PWR type reactor, there are provided a one-dimensional reactor core dynamic characteristic model for simulating one-dimensional neutron flux distribution in the axial direction of the reactor core and average reactor power based on each of inputted signals of control rod pattern, a reactor core flow rate, reactor core pressure and reactor core inlet enthalphy, and a three-dimensional reactor core dynamic characteristic mode for simulating three-dimensional power distribution of the reactor core, and a nuclear instrumentation model for calculating read value of the nuclear instrumentation disposed in the reactor based on the average reactor core power and the reactor core three-dimensional power distribution. A one-dimensional neutron flux distribution in the axial direction of the reactor core, a reactor core average power, a reactor core three-dimensional power distribution and a nuclear instrumentation read value are calculated. As a result, the three-dimensional power distribution and the power level are continuously calculated. Further, since the transient change of the three-dimensional neutron flux distribution is calculated accurately on real time, more actual response relative to a power monitoring device of the reactor core and operation performance can be simulated. (N.H.)

A Study on the Kinetic Characteristics of Transmutation Process Reactor

Energy Technology Data Exchange (ETDEWEB)

Chung, Chang Hyun; You, Young Woo; Cho, Jae seon; Huh, Chang Wook; Kim, Doh Hyung [Seoul National University, Seoul (Korea, Republic of)

1997-07-01

The purpose of this study is to examine the transient heat transfer characteristics of liquid mental as the coolant used in accelerator-driven transmutation process reactor which is related the disposal of high-level radioactive nuclide. At current stage, the accelerator-driven transmutation process is investigated as the most appropriate method among many transmutation process methods. In this study, previous research works are investigated especially about the thermal hydraulics and kinetic behavior of coolant material including heat transfer of coolant in transmutation process reactor. A study on the heat transfer characteristics of liquid metal is performed based on the thermal hydraulic kinetic characteristics of liquid metal reactor which uses liquid metal coolant. Based on this study, the most appropriate material for the coolant of transmutation reactor will be recommended. 53 refs., 15 tabs., 33 figs. (author)

An accurate solution of point reactor neutron kinetics equations of multi-group of delayed neutrons

International Nuclear Information System (INIS)

Yamoah, S.; Akaho, E.H.K.; Nyarko, B.J.B.

2013-01-01

Highlights: ► Analytical solution is proposed to solve the point reactor kinetics equations (PRKE). ► The method is based on formulating a coefficient matrix of the PRKE. ► The method was applied to solve the PRKE for six groups of delayed neutrons. ► Results shows good agreement with other traditional methods in literature. ► The method is accurate and efficient for solving the point reactor kinetics equations. - Abstract: The understanding of the time-dependent behaviour of the neutron population in a nuclear reactor in response to either a planned or unplanned change in the reactor conditions is of great importance to the safe and reliable operation of the reactor. In this study, an accurate analytical solution of point reactor kinetics equations with multi-group of delayed neutrons for specified reactivity changes is proposed to calculate the change in neutron density. The method is based on formulating a coefficient matrix of the homogenous differential equations of the point reactor kinetics equations and calculating the eigenvalues and the corresponding eigenvectors of the coefficient matrix. A small time interval is chosen within which reactivity relatively stays constant. The analytical method was applied to solve the point reactor kinetics equations with six-groups delayed neutrons for a representative thermal reactor. The problems of step, ramp and temperature feedback reactivities are computed and the results compared with other traditional methods. The comparison shows that the method presented in this study is accurate and efficient for solving the point reactor kinetics equations of multi-group of delayed neutrons

SNAP-3D: a three-dimensional neutron diffusion code

International Nuclear Information System (INIS)

McCallien, C.W.J.

1975-10-01

A preliminary report is presented describing the data requirements of a one- two- or three-dimensional multi-group diffusion code, SNAP-3D. This code is primarily intended for neutron diffusion calculations but it can also carry out gamma calculations if the diffuse approximation is accurate enough. It is suitable for fast and thermal reactor core calculations and for shield calculations. It is assumed the reader is familiar with the older, two-dimensional code SNAP and can refer to the report [TRG-Report-1990], describing it. The present report concentrates on the enhancements to SNAP that have been made to produce the three-dimensional version, SNAP-3D, and is intended to act a a guide on data preparation until a single, comprehensive report can be published. (author)

Coupled 3D neutron kinetics and thermalhydraulic characteristics of the Canadian supercritical water reactor

Energy Technology Data Exchange (ETDEWEB)

Hummel, David William, E-mail: hummeld@mcmaster.ca; Novog, David Raymond

2016-03-15

Highlights: • A coupled spatial kinetics and thermalhydraulics model of the PT-SCWR was created. • Positive power excursions were demonstrated during accident-like transients. • The reactor will inherently self-shutdown in such transients with some delay. • A fast-acting shutdown system would limit the consequences of the power pulse. - Abstract: The Canadian Supercritical Water-cooled Reactor concept, as an evolution of the CANada Deuterium Uranium (CANDU) reactor, includes both pressure tubes and a low temperature heavy water moderator. The current Pressure Tube type SCWR (PT-SCWR) concept features 64-element fuel assemblies placed within High Efficiency Re-entrant Channels (HERCs) that connect to core inlet and outlet plena. Among current SCWR concepts the PT-SCWR is unique in that the HERC separates multiple coolant and moderator regions, giving rise to coupled neutronic-thermalhydraulic feedbacks beyond those present in CANDU or contemporary Light Water Reactors. The objective of this work was thus to model the coupled neutronic-thermal hydraulic properties of the PT-SCWR to establish the impact of these multiple regions on the core's transient behavior. To that end, the features of the PT-SCWR were first modeled with the neutron transport code DRAGON to create a database of homogenized and condensed cross-sections and thermalhydraulic feedback coefficients. These were used as input to a core-level neutron diffusion model created with the code DONJON. The behavior of the primary heat transport system was modeled with the thermalhydraulic system code CATHENA. A procedure was developed to couple the outputs of DONJON and CATHENA, facilitating three-dimensional spatial neutron kinetics and coupled thermalhydraulic analysis of the PT-SCWR core. Several postulated transients were initiated within the coupled model by changing the core inlet and outlet boundary conditions. Decreasing coolant density around the fuel was demonstrated to produce positive

Novel swirl-flow reactor for kinetic studies of semiconductor photocatalysis

NARCIS (Netherlands)

Ray, A.K; Beenackers, A.A C M

1997-01-01

A new two-phase swirl-flow monolithic-type reactor was designed to study the kinetics of heterogeneous photocatalytic processes on immobilized semiconductor catalysts. True kinetic rate constants for destruction of a textile dye were measured as a function of wavelength of light intensity and angle

Three-dimensional all-speed CFD code for safety analysis of nuclear reactor containment: Status of GASFLOW parallelization, model development, validation and application

Energy Technology Data Exchange (ETDEWEB)

Xiao, Jianjun, E-mail: jianjun.xiao@kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Travis, John R., E-mail: jack_travis@comcast.com [Engineering and Scientific Software Inc., 3010 Old Pecos Trail, Santa Fe, NM 87505 (United States); Royl, Peter, E-mail: peter.royl@partner.kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Necker, Gottfried, E-mail: gottfried.necker@partner.kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Svishchev, Anatoly, E-mail: anatoly.svishchev@kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Jordan, Thomas, E-mail: thomas.jordan@kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany)

2016-05-15

Highlights: • 3-D scalable semi-implicit pressure-based CFD code for containment safety analysis. • Robust solution algorithm valid for all-speed flows. • Well validated and widely used CFD code for hydrogen safety analysis. • Code applied in various types of nuclear reactor containments. • Parallelization enables high-fidelity models in large scale containment simulations. - Abstract: GASFLOW is a three dimensional semi-implicit all-speed CFD code which can be used to predict fluid dynamics, chemical kinetics, heat and mass transfer, aerosol transportation and other related phenomena involved in postulated accidents in nuclear reactor containments. The main purpose of the paper is to give a brief review on recent GASFLOW code development, validations and applications in the field of nuclear safety. GASFLOW code has been well validated by international experimental benchmarks, and has been widely applied to hydrogen safety analysis in various types of nuclear power plants in European and Asian countries, which have been summarized in this paper. Furthermore, four benchmark tests of a lid-driven cavity flow, low Mach number jet flow, 1-D shock tube and supersonic flow over a forward-facing step are presented in order to demonstrate the accuracy and wide-ranging capability of ICE’d ALE solution algorithm for all-speed flows. GASFLOW has been successfully parallelized using the paradigms of Message Passing Interface (MPI) and domain decomposition. The parallel version, GASFLOW-MPI, adds great value to large scale containment simulations by enabling high-fidelity models, including more geometric details and more complex physics. It will be helpful for the nuclear safety engineers to better understand the hydrogen safety related physical phenomena during the severe accident, to optimize the design of the hydrogen risk mitigation systems and to fulfill the licensing requirements by the nuclear regulatory authorities. GASFLOW-MPI is targeting a high

Development of 3-dimensional neutronics kinetics analysis code for CANDU-PHWR

International Nuclear Information System (INIS)

Kim, M. W.; Kim, C. H.; Hong, I. S.

2005-02-01

The followings are the major contents and scope of the research : development of kinetics power calculation module, formulation of space-dependent neutron transient analysis - implementation of 3-D and 2-G unified nodal method, verification of the kinetics module by benchmark problem - 3-D PHWR kinetics benchmark problem suggested by AECL, reactor trip simulation by shutdown system 1 in Wolsong unit 2. Development of a dynamic linked library code, SCAN D LL, for the coupled calculation with RELAP-CANDU : modeling of shutdown system 1, development of automatic shutdown module - automatic trip module based on rate log power control logic, automatic insertion of shutdown system 1. Development of a link code for coupled calculation - development of SCAN D LL(windows version), verification of coupled code by - 40% reactor inlet header break LOCA power pulse, 100% reactor outlet header break LOCA power pulse, 50% pump suction break LOCA power pulse

Kinetics of Pressurized Water Reactors with Hot or Cold Moderators

Energy Technology Data Exchange (ETDEWEB)

Norinder, O

1960-11-15

The set of neutron kinetic equations developed in this report permits the use of long integration steps during stepwise integration. Thermal relations which describe the transfer of heat from fuel to coolant are derived. The influence upon the kinetic behavior of the reactor of a number of parameters is studied. A comparison of the kinetic properties of the hot and cold moderators is given.

Three-dimensional RAMA fluence methodology benchmarking

International Nuclear Information System (INIS)

Baker, S. P.; Carter, R. G.; Watkins, K. E.; Jones, D. B.

2004-01-01

This paper describes the benchmarking of the RAMA Fluence Methodology software, that has been performed in accordance with U. S. Nuclear Regulatory Commission Regulatory Guide 1.190. The RAMA Fluence Methodology has been developed by TransWare Enterprises Inc. through funding provided by the Electric Power Research Inst., Inc. (EPRI) and the Boiling Water Reactor Vessel and Internals Project (BWRVIP). The purpose of the software is to provide an accurate method for calculating neutron fluence in BWR pressure vessels and internal components. The Methodology incorporates a three-dimensional deterministic transport solution with flexible arbitrary geometry representation of reactor system components, previously available only with Monte Carlo solution techniques. Benchmarking was performed on measurements obtained from three standard benchmark problems which include the Pool Criticality Assembly (PCA), VENUS-3, and H. B. Robinson Unit 2 benchmarks, and on flux wire measurements obtained from two BWR nuclear plants. The calculated to measured (C/M) ratios range from 0.93 to 1.04 demonstrating the accuracy of the RAMA Fluence Methodology in predicting neutron flux, fluence, and dosimetry activation. (authors)

Analytic solutions of the multigroup space-time reactor kinetics equations

International Nuclear Information System (INIS)

Lee, C.E.; Rottler, S.

1986-01-01

The development of analytical and numerical solutions to the reactor kinetics equations is reviewed. Analytic solutions of the multigroup space-time reactor kinetics equations are developed for bare and reflected slabs and spherical reactors for zero flux, zero current and extrapolated endpoint boundary conditions. The material properties of the reactors are assumed constant in space and time, but spatially-dependent source terms and initial conditions are investigated. The system of partial differential equations is reduced to a set of linear ordinary differential equations by the Laplace transform method. These equations are solved by matrix Green's functions yielding a general matrix solution for the neutron flux and precursor concentration in the Laplace transform space. The detailed pole structure of the Laplace transform matrix solutions is investigated. The temporally- and spatially-dependent solutions are determined from the inverse Laplace transform using the Cauchy residue theorem, the theorem of Frobenius, a knowledge of the detailed pole structure and matrix operators. (author)

IBIS, FBR 3-D Steady-State and Kinetics with Thermohydraulic Feedback

International Nuclear Information System (INIS)

Konomura, Mamoru; Tada, Nobuo; Oka, Yoshiaki; An, Shigehiro

1987-01-01

1 - Description of program or function: The IBIS code performs steady state and kinetics calculations based on a three-dimensional nuclear diffusion kinetics with thermal hydraulic feedback. It can calculate the following values in hexagonal-Z geometry of a fast breeder reactor core through the progress of transient: (1) Net reactivity; (2) Total and group-wise delayed neutron fraction; (3) Group-wise delayed neutron precursor concentration; (4) Total power and energy; (5) Space dependent neutron flux in each energy group; (6) Space dependent temperature of each material; (7) Maximum temperature of each material and its location. 2 - Method of solution: The quasi-static method is adopted to solve the three-dimensional nuclear diffusion kinetics problem. The method is the same as employed in the code QX1. The shape function equation is solved with the finite difference treatment as used in the codes CITATION and HONEYCOMB. One-dimensional thermo-hydraulics is solved with a model similar to that given in the code SASLA. Sodium boiling can be taken into account. 3 - Restrictions on the complexity of the problem: The number of neutron energy groups is fixed to 3 groups in the present version of the code

Electro-recovery of gold and silver from a cyanide leaching solution using a three-dimensional reactor

Energy Technology Data Exchange (ETDEWEB)

Reyes-Cruz, V.; Gonzalez, I.; Oropeza, M.T

2004-10-01

The selective electro-recovery of gold and silver values from cyanide leaching solutions containing copper was accomplished in a three-dimensional (3D) electrochemical reactor. This case let to contrast three different points of view when dealing with a composed metallic solution: First, the thermodynamic predictions; second, the microelectrolysis approach and finally, the macroelectrolysis experiments. Standard electrode potentials for the study solution would indicate a tendency for gold to deposit first. However, microelectrolysis studies of the three-metallic solution indicated that gold and silver are co-deposited onto a Vitreous carbon (VC) electrode without copper interference in a narrow potential range. Mass balances during the macroelectrolysis experiments (batch model assuming mass transfer control) indicated a preferential deposition of silver during the first ten minutes, even if gold deposition also occurred. On the other hand, values of Stanton (St) for different linear flow velocity corroborated that metals concentration gradients may establish a limit to make profitable the fluid velocity increase in an electrochemical flow cell. Electrolysis experiments were carried out under potentiostatic (at -1400 mV versus SCE) and galvanostatic (at -3.9 Am{sup -2}) conditions in the FM-01 LC flow cell.

Electro-recovery of gold and silver from a cyanide leaching solution using a three-dimensional reactor

International Nuclear Information System (INIS)

Reyes-Cruz, V.; Gonzalez, I.; Oropeza, M.T.

2004-01-01

The selective electro-recovery of gold and silver values from cyanide leaching solutions containing copper was accomplished in a three-dimensional (3D) electrochemical reactor. This case let to contrast three different points of view when dealing with a composed metallic solution: First, the thermodynamic predictions; second, the microelectrolysis approach and finally, the macroelectrolysis experiments. Standard electrode potentials for the study solution would indicate a tendency for gold to deposit first. However, microelectrolysis studies of the three-metallic solution indicated that gold and silver are co-deposited onto a Vitreous carbon (VC) electrode without copper interference in a narrow potential range. Mass balances during the macroelectrolysis experiments (batch model assuming mass transfer control) indicated a preferential deposition of silver during the first ten minutes, even if gold deposition also occurred. On the other hand, values of Stanton (St) for different linear flow velocity corroborated that metals concentration gradients may establish a limit to make profitable the fluid velocity increase in an electrochemical flow cell. Electrolysis experiments were carried out under potentiostatic (at -1400 mV versus SCE) and galvanostatic (at -3.9 Am -2 ) conditions in the FM-01 LC flow cell

Three-dimensional discrete ordinates reactor assembly calculations on GPUs

Energy Technology Data Exchange (ETDEWEB)

Evans, Thomas M [ORNL; Joubert, Wayne [ORNL; Hamilton, Steven P [ORNL; Johnson, Seth R [ORNL; Turner, John A [ORNL; Davidson, Gregory G [ORNL; Pandya, Tara M [ORNL

2015-01-01

In this paper we describe and demonstrate a discrete ordinates sweep algorithm on GPUs. This sweep algorithm is nested within a multilevel comunication-based decomposition based on energy. We demonstrated the effectiveness of this algorithm on detailed three-dimensional critical experiments and PWR lattice problems. For these problems we show improvement factors of 4 6 over conventional communication-based, CPU-only sweeps. These sweep kernel speedups resulted in a factor of 2 total time-to-solution improvement.

Fully integrated analysis of reactor kinetics, thermalhydraulics and the reactor control system in the MAPLE-X10 research reactor

International Nuclear Information System (INIS)

Shim, S.Y.; Carlson, P.A.; Baxter, D.K.

1992-01-01

A prototype research reactor, designated MAPLE-X10 (Multipurpose Applied Physics Lattice Experimental - X 10MW), is currently being built at AECL's Chalk River Laboratories. The CATHENA (Canadian Algorithm for Thermalhydraulic Network Analysis) two-fluid code was used in the safety analysis of the reactor to determine the adequacy of core cooling during postulated reactivity and loss-of-forced-flow transients. The system responses to a postulated transient are predicted including the feedback between reactor kinetics, thermalhydrauilcs and the reactor control systems. This paper describes the MAPLE-X10 reactor and the modelling methodology used. Sample simulations of postulated loss-of-heat-sink and loss-of-regulation transients are presented. (author)

Kinetic Modeling of Synthetic Wastewater Treatment by the Moving-bed Sequential Continuous-inflow Reactor (MSCR

Directory of Open Access Journals (Sweden)

Mohammadreza Khani

2016-11-01

Full Text Available It was the objective of the present study to conduct a kinetic modeling of a Moving-bed Sequential Continuous-inflow Reactor (MSCR and to develop its best prediction model. For this purpose, a MSCR consisting of an aerobic-anoxic pilot 50 l in volume and an anaerobic pilot of 20 l were prepared. The MSCR was fed a variety of organic loads and operated at different hydraulic retention times (HRT using synthetic wastewater at input COD concentrations of 300 to 1000 mg/L with HRTs of 2 to 5 h. Based on the results and the best system operation conditions, the highest COD removal (98.6% was obtained at COD=500 mg/L. The three well-known first order, second order, and Stover-Kincannon models were utilized for the kinetic modeling of the reactor. Based on the kinetic analysis of organic removal, the Stover-Kincannon model was chosen for the kinetic modeling of the moving bed biofilm. Given its advantageous properties in the statisfactory prediction of organic removal at different organic loads, this model is recommended for the design and operation of MSCR systems.

Fast three-dimensional core optimization based on modified one-group model

Energy Technology Data Exchange (ETDEWEB)

Freire, Fernando S. [ELETROBRAS Termonuclear S.A. - ELETRONUCLEAR, Rio de Janeiro, RJ (Brazil). Dept. GCN-T], e-mail: freire@eletronuclear.gov.br; Martinez, Aquilino S.; Silva, Fernando C. da [Coordenacao dos Programas de Pos-graduacao de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear], e-mail: aquilino@con.ufrj.br, e-mail: fernando@con.ufrj.br

2009-07-01

The optimization of any nuclear reactor core is an extremely complex process that consumes a large amount of computer time. Fortunately, the nuclear designer can rely on a variety of methodologies able to approximate the analysis of each available core loading pattern. Two-dimensional codes are usually used to analyze the loading scheme. However, when particular axial effects are present in the core, two-dimensional analysis cannot produce good results and three-dimensional analysis can be required at all time. Basically, in this paper are presented the major advantages that can be found when one use the modified one-group diffusion theory coupled with a buckling correction model in optimization process. The results of the proposed model are very accurate when compared to benchmark results obtained from detailed calculations using three-dimensional nodal codes (author)

Fast three-dimensional core optimization based on modified one-group model

International Nuclear Information System (INIS)

Freire, Fernando S.; Martinez, Aquilino S.; Silva, Fernando C. da

2009-01-01

The optimization of any nuclear reactor core is an extremely complex process that consumes a large amount of computer time. Fortunately, the nuclear designer can rely on a variety of methodologies able to approximate the analysis of each available core loading pattern. Two-dimensional codes are usually used to analyze the loading scheme. However, when particular axial effects are present in the core, two-dimensional analysis cannot produce good results and three-dimensional analysis can be required at all time. Basically, in this paper are presented the major advantages that can be found when one use the modified one-group diffusion theory coupled with a buckling correction model in optimization process. The results of the proposed model are very accurate when compared to benchmark results obtained from detailed calculations using three-dimensional nodal codes (author)

Reaction kinetic analysis of reactor surveillance data

Energy Technology Data Exchange (ETDEWEB)

Yoshiie, T., E-mail: yoshiie@rri.kyoto-u.ac.jp [Research Reactor Institute, Kyoto University, Kumatori-cho, Sennan-gun, Osaka-fu 590-0494 (Japan); Kinomura, A. [Research Reactor Institute, Kyoto University, Kumatori-cho, Sennan-gun, Osaka-fu 590-0494 (Japan); Nagai, Y. [The Oarai Center, Institute for Materials Research, Tohoku University, Oarai, Ibaraki 311-1313 (Japan)

2017-02-15

In the reactor pressure vessel surveillance data of a European-type pressurized water reactor (low-Cu steel), it was found that the concentration of matrix defects was very high, and a large number of precipitates existed. In this study, defect structure evolution obtained from surveillance data was simulated by reaction kinetic analysis using 15 rate equations. The saturation of precipitation and the growth of loops were simulated, but it was not possible to explain the increase in DBTT on the basis of the defect structures. The sub-grain boundary segregation of solutes was discussed for the origin of the DBTT increase.

Kinetic Monte Carlo simulation of three-dimensional shape evolution with void formation using Solid-by-Solid model: Application to via and trench filling

International Nuclear Information System (INIS)

Kaneko, Yutaka; Hiwatari, Yasuaki; Ohara, Katsuhiko; Asa, Fujio

2013-01-01

In this paper we present the Kinetic Monte Carlo simulation system for the simulation of three-dimensional shape evolution with void formation as a model for electrodeposition. The basic system is the Solid-by-Solid model which is an extension of the conventional Solid-on-Solid model for crystal growth to include void formation. The advantage of the Solid-by-Solid model is that complex three-dimensional shape evolution accompanying void formation (from point defects to macro voids) can be simulated without the difficulty of treating moving boundaries. This model has been extended to include the solution part in which the migration of ions is simulated by the coarse-grained random walk. A multi-scale method is employed to generate the concentration gradient in the diffusion layer. The extended model is applied to the simulation of via and trench fillings by copper electrodeposition. Three kinds of additives are included: suppressors, accelerators and chloride ions. The mechanism of void formation, effects of additives and their influence on the bottom-up filling are discussed within the framework of this model

Effect of Utilization of Silicide Fuel with the Density 4.8 gU/cc on the Kinetic Parameters of RSG-GAS Reactor

International Nuclear Information System (INIS)

Setiyanto; Sembiring, Tagor M.; Pinem, Surian

2007-01-01

Presently, the RSG-GAS reactor using silicide fuel element of 2.96 gU/cc. For increasing reactor operation time, its planning to change to higher density fuel. The kinetic calculation of silicide core with density 4.8 gU/cc has been carried out, since it has an influence on the reactor operation safety. The calculated kinetic parameters are the effective delayed neutron fraction, the delayed neutron decay constant, prompt neutron lifetime and feedback reactivity coefficient very important for reactor operation safety. the calculation is performed in 2-dimensional neutron diffusion-perturbation method using modified Batan-2DIFF code. The calculation showed that the effective delayed neutron fraction is 7. 03256x10 -03 , total delay neutron time constant is 7.85820x10 -02 s -1 and the prompt neutron lifetime is 55.4900 μs. The result of prompt neutron lifetime smaller 10 % compare with silicide fuel of 4.8 gU/cc. The calculated results showed that all of the feedback reactivity coefficient silicide core 4.8 gU/cc is negative. Totally, the feedback reactivity coefficient of silicide fuel of 4.8 gU/cc is 10% less than that of silicide fuel of 2.96 gU/cc. The results shown that kinetic parameters result decrease compared with the silicide core with density 2.96 gU/cc, but no significant influence in the RSG-GAS reactor operation. (author)

Comparison of zero-dimensional and one-dimensional thermonuclear burn computations for the reversed-field pinch reactor (RFPR)

International Nuclear Information System (INIS)

Nebel, R.A.; Hagenson, R.L.; Moses, R.W.; Krakowski, R.A.

1980-01-01

Conceptual fusion reactor designs of the Reversed-Field Pinch Reactor (RFPR) have been based on profile-averaged zero-dimensional (point) plasma models. The plasma response/performance that has been predicted by the point plasma model is re-examined by a comprehensive one-dimensional (radial) burn code that has been developed and parametrically evaluated for the RFPR. Agreement is good between the zero-dimensional and one-dimensional models, giving more confidence in the RFPR design point reported previously from the zero-dimensional analysis

Three-Dimensional Simulation of Ultrasound-Induced Microalgal Cell Disruption.

Science.gov (United States)

Wang, M; Yuan, W; Hale, Andy

2016-03-01

The three-dimensional distribution (x, y, and z) of ultrasound-induced microalgal cell disruption in a sonochemical reactor was predicted by solving the Helmholtz equation using a three-dimensional acoustic module in the COMSOL Multiphysics software. The simulated local ultrasound pressure at any given location (x, y, and z) was found to correlate with cell disruption of a freshwater alga, Scenedesmus dimorphus, represented by the change of algal cell particle/debris concentration, chlorophyll-a fluorescence density (CAFD), and Nile red stained lipid fluorescence density (LFD), which was also validated by the model reaction of potassium iodide oxidation (the Weissler reaction). Furthermore, the effect of ultrasound power intensity and processing duration on algal cell disruption was examined to address the limitation of the model.

Cassandre : a two-dimensional multigroup diffusion code for reactor transient analysis

International Nuclear Information System (INIS)

Arien, B.; Daniels, J.

1986-12-01

CASSANDRE is a two-dimensional (x-y or r-z) finite element neutronics code with thermohydraulics feedback for reactor dynamics prior to the disassembly phase. It uses the multigroup neutron diffusion theory. Its main characteristics are the use of a generalized quasistatic model, the use of a flexible multigroup point-kinetics algorithm allowing for spectral matching and the use of a finite element description. The code was conceived in order to be coupled with any thermohydraulics module, although thermohydraulics feedback is only considered in r-z geometry. In steady state criticality search is possible either by control rod insertion or by homogeneous poisoning of the coolant. This report describes the main characterstics of the code structure and provides all the information needed to use the code. (Author)

Ordering kinetics in quasi-one-dimensional Ising-like systems

International Nuclear Information System (INIS)

Mueller, M.; Paul, W.

1993-01-01

Results are presented of a Monte Carlo simulation of the kinetics of ordering in the two-dimensional nearest-neighbor Ising model in an L x M geometry with two free boundaries of length M much-gt L. This model can be viewed as representing an adsorbant on a stepped surface with mean terrace width L. The authors follow the ordering kinetics after quenches to temperatures 0.25 ≤T/T c ≤1 starting from a random initial configuration at a coverage of Θ=0.5 in the corresponding lattice gas picture. The systems evolve in time according to a Glauber kinetics with nonconserved order parameter. The equilibrium structure is given by a one-dimensional sequence of ordered domains. The ordering process evolves from a short initial two-dimensional ordering process through a crossover region to a quasi-one-dimensional behavior. The whole process is diffusive (inverse half-width of the structure factor peak 1/Δq parallel ∝ √t), in contrast to a model proposed by Kawasaki et al., where an intermediate logarithmic growth law is expected. All results are completely describable in the picture of an annihilating random walk (ARW) of domain walls. 36 refs., 16 figs

Direct-coupled-ray method for design-oriented three-dimensional transport analysis

International Nuclear Information System (INIS)

Bucholz, J.A.; Poncelet, C.G.

1977-01-01

A fast three-dimensional design-oriented transport method has been developed for the solution of both neutron and gamma transport problems. It combines a nodal approach with analytic integral transport to achieve relative speed and accuracy. An analytic solution is obtained for the angular flux in each of the 14 directions defined by the six faces and eight corners of a cubic mesh block. The scheme used to accommodate high-order anisotropic scattering is based on the formulation of ray-to-ray scattering probabilities in an integral sense. A variable mesh approximation has also been introduced to provide greater flexibility. The details of a direct-coupled-ray (DCR) → P 1 conversion technique have been developed but not yet implemented. The DCR method, as implemented in the TRANS3 code, has been used in a number of liquid-metal fast breeder reactor shielding applications. These included a one-dimensional deep penetration configuration and one-, two-, and three dimensional representations of the lower axial shield of the Clinch River Breeder Reactor. Comparisons with ANISN and DOT-III solutions indicated good to excellent agreement in most situations

Kinetic characteristics of the Dalat Nuclear Research Reactor

Energy Technology Data Exchange (ETDEWEB)

An, Tran Khac; Dien, Nguyen Nhi; Hien, Pham Duy [Nuclear Research Inst., Da Lat (Viet Nam); and others

1994-10-01

Kinetic characteristics of the reconstructed nuclear reactor in Dalat is investigated. Experimental parameters measured consist of: temperature coefficient of reactivity for water moderator, xenon poisoning, contribution of delayed photoneutrons induced by Be({gamma}, n) reactions and positive reactivity insertion behavior. (author). 6 refs. 4 figs.

Kinetic characteristics of the Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Tran Khac An; Nguyen Nhi Dien; Pham Duy Hien

1994-01-01

Kinetic characteristics of the reconstructed nuclear reactor in Dalat is investigated. Experimental parameters measured consist of: temperature coefficient of reactivity for water moderator, xenon poisoning, contribution of delayed photoneutrons induced by Be(γ, n) reactions and positive reactivity insertion behavior. (author). 6 refs. 4 figs

Kinetics of two-dimensional electron plasma, interacting with fluctuating potential

International Nuclear Information System (INIS)

Boiko, I.I.; Sirenko, Y.M.

1990-01-01

In this paper, from the first principles, after the fashion of Klimontovich, the authors derive quantum kinetic equation for electron gas, inhomogeneous in z-direction and homogeneous in XY-plane. Special attention is given to the systems with quasi-two-dimensional electron gas (2 DEG), which are widely explored now. Both interaction between the particles of 2 DEG (in general, of several sorts), and interaction with an external system (phonons, impurities, after change carries etc.) are considered. General theory is used to obtain energy and momentum balance equations and relaxation frequencies for 2 DEG in the basis of plane waves. The case of crossed electric and magnetic fields is also treated. As an illustration the problems of 2 DEG scattering on semibounded three-dimensional electron gas and on two-dimensional hole gas are considered; transverse conductivity of nondegenerate 2 DEG, scattered by impurities in ultraquantum magnetic field, is calculated

Reactor thermal behaviors under kinetics parameters variations in fast reactivity insertion

Energy Technology Data Exchange (ETDEWEB)

Abou-El-Maaty, Talal [Reactors Department, Atomic Energy Authority, Cairo 13759 (Egypt)], E-mail: talal22969@yahoo.com; Abdelhady, Amr [Reactors Department, Atomic Energy Authority, Cairo 13759 (Egypt)

2009-03-15

The influences of variations in some of the kinetics parameters affecting the reactivity insertion are considered in this study, it has been accomplished in order to acquire knowledge about the role that kinetic parameters play in prompt critical transients from the safety point of view. The kinetics parameters variations are limited to the effective delayed neutron fraction ({beta}{sub eff}) and the prompt neutron generation time ({lambda}). The reactor thermal behaviors under the variations in effective delayed neutron fraction and prompt neutron generation time included, the reactor power, maximum fuel temperature, maximum clad temperature, maximum coolant temperature and the mass flux variations at the hot channel. The analysis is done for a typical swimming pool, plate type research reactor with low enriched uranium. The scram system is disabled during the accidents simulations. Calculations were done using PARET code. As a result of simulations, it is concluded that, the reactor (ETRR2) thermal behavior is considerably more sensitive to the variation in the effective delayed neutron fraction than to the variation in prompt neutron generation time and the fast reactivity insertion in both cases causes a flow expansion and contraction at the hot channel exit. The amplitude of the oscillated flow is a qualitatively increases with the decrease in both {beta}{sub eff} and {lambda}.

Fractional neutron point kinetics equations for nuclear reactor dynamics

International Nuclear Information System (INIS)

Espinosa-Paredes, Gilberto; Polo-Labarrios, Marco-A.; Espinosa-Martinez, Erick-G.; Valle-Gallegos, Edmundo del

2011-01-01

The fractional point-neutron kinetics model for the dynamic behavior in a nuclear reactor is derived and analyzed in this paper. The fractional model retains the main dynamic characteristics of the neutron motion in which the relaxation time associated with a rapid variation in the neutron flux contains a fractional order, acting as exponent of the relaxation time, to obtain the best representation of a nuclear reactor dynamics. The physical interpretation of the fractional order is related with non-Fickian effects from the neutron diffusion equation point of view. The numerical approximation to the solution of the fractional neutron point kinetics model, which can be represented as a multi-term high-order linear fractional differential equation, is calculated by reducing the problem to a system of ordinary and fractional differential equations. The numerical stability of the fractional scheme is investigated in this work. Results for neutron dynamic behavior for both positive and negative reactivity and for different values of fractional order are shown and compared with the classic neutron point kinetic equations. Additionally, a related review with the neutron point kinetics equations is presented, which encompasses papers written in English about this research topic (as well as some books and technical reports) published since 1940 up to 2010.

Biomedical applications of two- and three-dimensional deterministic radiation transport methods

International Nuclear Information System (INIS)

Nigg, D.W.

1992-01-01

Multidimensional deterministic radiation transport methods are routinely used in support of the Boron Neutron Capture Therapy (BNCT) Program at the Idaho National Engineering Laboratory (INEL). Typical applications of two-dimensional discrete-ordinates methods include neutron filter design, as well as phantom dosimetry. The epithermal-neutron filter for BNCT that is currently available at the Brookhaven Medical Research Reactor (BMRR) was designed using such methods. Good agreement between calculated and measured neutron fluxes was observed for this filter. Three-dimensional discrete-ordinates calculations are used routinely for dose-distribution calculations in three-dimensional phantoms placed in the BMRR beam, as well as for treatment planning verification for live canine subjects. Again, good agreement between calculated and measured neutron fluxes and dose levels is obtained

RA reactor kinetic parameters - Progress report; Kineticki parametri reaktora RA - Izvestaj o napredovanju -

Energy Technology Data Exchange (ETDEWEB)

Petrovic, M; Obradovic, D; Jevtovic, V; Velickovic, Lj [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1965-11-15

The objective of nuclear reactor kinetics study is to analyze the stability of reactor operation in practice. The obtained parameters should define the needed properties of automatic control system relevant for the stability of the designed reactor system. Refining the analytical models is done by using the analysis and interpretation of experimental data. Results of measured the reactor response obtained by using the reactor oscillator ROB-1 are explained by using the space independent model of the zero power reactor, by power reactor model with one feedback circuit, and by a complex model. It was assumed that the perturbations of the system are small and that linearized kinetic equations could be used. Linearized kinetic equation of the reactor system are transformed into the frequency region in order to analyze the measured values directly. The objective of this paper is to measure the RA reactor kinetics parameters, and analyze the stability of reactor operation at power levels high than nominal. Istrazivanja u oblasti kinetike nuklearnih reaktora imaju za cilj da dovedu analizu stabilnosti rada reaktora na nivo 'radne tehnologije'. Dobijeni pararametri treba da specificiraju potrebne karakteristike sistema automatske kontrole za odgovarajucu stabilnost projektovanog reaktorskog sistema. Doterivanjem analitickih modela do takvog nivoa da se zapazeni fenomeni mogu anailitcki predvideti ide preko analize i interpretacije eksperimentalnih podataka. Eksperimentalni rezultati merenja odziva reaktora, izvedeni reaktorskim oscilatorom ROB-1, interpretirani su na osnovu prostorno nezavisnog modela za reaktor nulte snage, modelom reaktora snage sa jednim kolom povratne sprege, kao i kompleksnim modelom. U ovom radu se poslo od toga da su perturbacije parametara sistema male, pa se mogu upotrebiti linearizovane kineticke jednacine. Linearizovane kineticke jednacine reaktorskog sistema transformirane su u frekventno podrucje s ciljem direktne analize mernih rezultata

The Impact of Three-Dimensional Effects on the Simulation of Turbulence Kinetic Energy in a Major Alpine Valley

Science.gov (United States)

Goger, Brigitta; Rotach, Mathias W.; Gohm, Alexander; Fuhrer, Oliver; Stiperski, Ivana; Holtslag, Albert A. M.

2018-07-01

The correct simulation of the atmospheric boundary layer (ABL) is crucial for reliable weather forecasts in truly complex terrain. However, common assumptions for model parametrizations are only valid for horizontally homogeneous and flat terrain. Here, we evaluate the turbulence parametrization of the numerical weather prediction model COSMO with a horizontal grid spacing of Δ x = 1.1 km for the Inn Valley, Austria. The long-term, high-resolution turbulence measurements of the i-Box measurement sites provide a useful data pool of the ABL structure in the valley and on slopes. We focus on days and nights when ABL processes dominate and a thermally-driven circulation is present. Simulations are performed for case studies with both a one-dimensional turbulence parametrization, which only considers the vertical turbulent exchange, and a hybrid turbulence parametrization, also including horizontal shear production and advection in the budget of turbulence kinetic energy (TKE). We find a general underestimation of TKE by the model with the one-dimensional turbulence parametrization. In the simulations with the hybrid turbulence parametrization, the modelled TKE has a more realistic structure, especially in situations when the TKE production is dominated by shear related to the afternoon up-valley flow, and during nights, when a stable ABL is present. The model performance also improves for stations on the slopes. An estimation of the horizontal shear production from the observation network suggests that three-dimensional effects are a relevant part of TKE production in the valley.

Non-linear punctual kinetics applied to PWR reactors simulation

International Nuclear Information System (INIS)

Cysne, F.S.

1978-11-01

In order to study some kinds of nuclear reactor accidents, a simulation is made using the punctual kinetics model for the reactor core. The following integration methods are used: Hansen's method in which a linearization is made and CSMP using a variable interval fourth-order Runge Kutta method. The results were good and were compared with those obtained by the code Dinamica I which uses a finite difference integration method of backward kind. (Author) [pt

Reliability of three-dimensional gait analysis in cervical spondylotic myelopathy.

LENUS (Irish Health Repository)

McDermott, Ailish

2010-10-01

Gait impairment is one of the primary symptoms of cervical spondylotic myelopathy (CSM). Detailed assessment is possible using three-dimensional gait analysis (3DGA), however the reliability of 3DGA for this population has not been established. The aim of this study was to evaluate the test-retest reliability of temporal-spatial, kinematic and kinetic parameters in a CSM population.

Experimental methods of investigation of kinetics and dynamics of nuclear reactors

International Nuclear Information System (INIS)

Costa Oliveira, Jaime M.

1969-03-01

The author presents experimental methods used to study kinetic and dynamic properties of nuclear reactors. Kinetic methods aim at determining characteristic parameters of the behaviour in time of neutrons. Dynamic methods aim at establishing the relationships between the reactor behaviour and its internal and external causes (notably the measurement of transfer functions). The author proposes a classification with respect to the excitation type: periodic excitation (reactivity sinusoidal modulation, source sinusoidal modulation, periodic pulse excitation), non periodic excitation (reactivity monitoring, reactivity linear variation, reactivity variation according to any given law, removal of starting source), random excitation (random reactivity or source excitation), natural fluctuations (alpha-Rossi method, methods of reduced variance, probabilistic methods, correlation methods, spectral analysis method). He also addresses space and energy effects. Applications are reported for low power and power reactors

A methodology for modeling photocatalytic reactors for indoor pollution control using previously estimated kinetic parameters

Energy Technology Data Exchange (ETDEWEB)

Passalia, Claudio; Alfano, Orlando M. [INTEC - Instituto de Desarrollo Tecnologico para la Industria Quimica, CONICET - UNL, Gueemes 3450, 3000 Santa Fe (Argentina); FICH - Departamento de Medio Ambiente, Facultad de Ingenieria y Ciencias Hidricas, Universidad Nacional del Litoral, Ciudad Universitaria, 3000 Santa Fe (Argentina); Brandi, Rodolfo J., E-mail: rbrandi@santafe-conicet.gov.ar [INTEC - Instituto de Desarrollo Tecnologico para la Industria Quimica, CONICET - UNL, Gueemes 3450, 3000 Santa Fe (Argentina); FICH - Departamento de Medio Ambiente, Facultad de Ingenieria y Ciencias Hidricas, Universidad Nacional del Litoral, Ciudad Universitaria, 3000 Santa Fe (Argentina)

2012-04-15

Highlights: Black-Right-Pointing-Pointer Indoor pollution control via photocatalytic reactors. Black-Right-Pointing-Pointer Scaling-up methodology based on previously determined mechanistic kinetics. Black-Right-Pointing-Pointer Radiation interchange model between catalytic walls using configuration factors. Black-Right-Pointing-Pointer Modeling and experimental validation of a complex geometry photocatalytic reactor. - Abstract: A methodology for modeling photocatalytic reactors for their application in indoor air pollution control is carried out. The methodology implies, firstly, the determination of intrinsic reaction kinetics for the removal of formaldehyde. This is achieved by means of a simple geometry, continuous reactor operating under kinetic control regime and steady state. The kinetic parameters were estimated from experimental data by means of a nonlinear optimization algorithm. The second step was the application of the obtained kinetic parameters to a very different photoreactor configuration. In this case, the reactor is a corrugated wall type using nanosize TiO{sub 2} as catalyst irradiated by UV lamps that provided a spatially uniform radiation field. The radiative transfer within the reactor was modeled through a superficial emission model for the lamps, the ray tracing method and the computation of view factors. The velocity and concentration fields were evaluated by means of a commercial CFD tool (Fluent 12) where the radiation model was introduced externally. The results of the model were compared experimentally in a corrugated wall, bench scale reactor constructed in the laboratory. The overall pollutant conversion showed good agreement between model predictions and experiments, with a root mean square error less than 4%.

Multi-dimensional fluid-structure interactions in a pressurized water reactor

International Nuclear Information System (INIS)

Dienes, J.K.; Hirt, C.W.; Stein, L.R.

1977-01-01

Sudden loss of coolant in a pressurized water reactor due to failure of a coolant pipe would result in flashing of the coolant accompanied by the propagation of a rarefaction wave into the downcomer. A computer program that simultaneously calculates the behavior of the coolant and the accompanying motion of the core support barrel which is considered as a three-dimensional shell with both membrane and bending stresses is discussed

Kinetics of anaerobic digestion of labaneh whey in a batch reactor

African Journals Online (AJOL)

SAM

2014-04-16

Apr 16, 2014 ... kinetic constants were determined for labaneh whey and for diluted whey .... reactor has a pH and temperature control system. ... Variable power electric heater was used to heat the reactor. ..... by gas chromatography, Annual book of ASTM Standard, Vol. ... Thesis, The University of Jordan, Amman, Jordan.

Enhancement of safety analysis reliability for a CANDU-6 reactor using RELAP-CANDU/SCAN coupled code system

International Nuclear Information System (INIS)

Kim, Man Woong; Choi, Yong Seog; Sin, Chul; Kim, Hyun Koon; Kim, Hho Jung; Hwang, Su Hyun; Hong, In Seob; Kim, Chang Hyo

2005-01-01

In LOCA analysis of the CANDU reactor, the system thermal-hydraulic code, RELAP-CANDU, alone cannot predict the transient behavior accurately. Therefore, the best estimate neutronics and system thermal-hydraulic coupled code system is necessary to describe the transient behavior with higher accuracy and reliability. To perform on-line calculation of safety analysis for CANDU reactor, a coupled thermal hydraulics-neutronics code system was developed in such a way that the best-estimate thermal-hydraulic system code for CANDU reactor, RELAP-CANDU, is coupled with the full three-dimensional reactor core kinetic code

Analysis of the three dimensional core kinetics NESTLE code coupling with the advanced thermo-hydraulic code systems, RELAP5/SCDAPSIM and its application to the Laguna Verde Central reactor; Analisis para el acoplamiento del codigo NESTLE para la cinetica tridimensional del nucleo al codigo avanzado de sistemas termo-hidraulicos, RELAP5/SCDAPSIM y su aplicacion al reactor de la CNLV

Energy Technology Data Exchange (ETDEWEB)

Salazar C, J H; Nunez C, A [CNSNS, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D.F. (Mexico); Chavez M, C [UNAM, Facultad de Ingenieria, DEPFI Campus Morelos (Mexico)

2004-07-01

The objective of the written present is to propose a methodology for the joining of the codes RELAP5/SCDAPSIM and NESTLE. The development of this joining will be carried out inside a doctoral program of Engineering in Energy with nuclear profile of the Ability of Engineering of the UNAM together with the National Commission of Nuclear Security and Safeguards (CNSNS). The general purpose of this type of developments, is to have tools that are implemented by multiple programs or codes such a that systems or models of the three-dimensional kinetics of the core can be simulated and those of the dynamics of the reactor (water heater-hydraulics). In the past, by limitations for the calculation of the complete answer of both systems, the developed models they were carried out for separate, putting a lot of emphasis in one but neglecting the other one. These methodologies, calls of better estimate, will be good to the nuclear industry to evaluate, with more high grades of detail, the designs of the nuclear power plant (for modifications to those already existent or for new concepts in the designs of advanced reactors), besides analysing events (transitory and have an accident), among other applications. The coupled system was applied to design studies and investigation of the Laguna Verde Nuclear power plant (CNLV). (Author)

Three-dimensional finite-element analysis of the cellular convection phenomena in the Clinch River Breeder Reactor Plant prototype pump

International Nuclear Information System (INIS)

Silver, A.H.; Lee, J.Y.

1983-01-01

Cellular convection was studied rigorously during the development of the Clinch River Breeder Reactor Plant (CRBRP) Program Pumps. This paper presents the development of a three-dimensional finite-element heat transfer model which accounts for the cellular convection phenomena. A buoyancy driven cellular convection flow pattern is introduced in the annulus region between the upper inner structure and the pump tank. Steady-state thermal data were obtained for several test conditions for argon gas pressures up to 93 psig (741 kPa) and sodium operating temperatures to 1000 0 F (811 0 K). Test temperature distributions on the pump tank and inner structure were correlated with numerical results and excellent agreement was obtained

Modeling and control of a large nuclear reactor. A three-time-scale approach

Energy Technology Data Exchange (ETDEWEB)

Shimjith, S.R. [Indian Institute of Technology Bombay, Mumbai (India); Bhabha Atomic Research Centre, Mumbai (India); Tiwari, A.P. [Bhabha Atomic Research Centre, Mumbai (India); Bandyopadhyay, B. [Indian Institute of Technology Bombay, Mumbai (India). IDP in Systems and Control Engineering

2013-07-01

Recent research on Modeling and Control of a Large Nuclear Reactor. Presents a three-time-scale approach. Written by leading experts in the field. Control analysis and design of large nuclear reactors requires a suitable mathematical model representing the steady state and dynamic behavior of the reactor with reasonable accuracy. This task is, however, quite challenging because of several complex dynamic phenomena existing in a reactor. Quite often, the models developed would be of prohibitively large order, non-linear and of complex structure not readily amenable for control studies. Moreover, the existence of simultaneously occurring dynamic variations at different speeds makes the mathematical model susceptible to numerical ill-conditioning, inhibiting direct application of standard control techniques. This monograph introduces a technique for mathematical modeling of large nuclear reactors in the framework of multi-point kinetics, to obtain a comparatively smaller order model in standard state space form thus overcoming these difficulties. It further brings in innovative methods for controller design for systems exhibiting multi-time-scale property, with emphasis on three-time-scale systems.

Two-detector cross-correlation noise technique and its application in measuring reactor kinetic parameters

International Nuclear Information System (INIS)

Lu Guiping; Peng Feng; Yi Jieyi

1988-01-01

The two-detector cross-correlation noise technique is a new method of measuring reactor kinetic parameters developed in the sixties. It has the advantages of non-perturbation in core, high signal to noise ratio, low space dependent effect, and simple and reliable in measurement. A special set of cross-correlation analyzer has been prepared for measuring kinetic parameters of several reactor assemblies, such as the High Flux Engineering Test Reactor, its zero power mock up facility and a low enriched uranium light water lattice zero power facility

Knee joint kinetics in response to multiple three-dimensional printed, customised foot orthoses for the treatment of medial compartment knee osteoarthritis.

Science.gov (United States)

Allan, Richard; Woodburn, James; Telfer, Scott; Abbott, Mandy; Steultjens, Martijn Pm

2017-06-01

The knee adduction moment is consistently used as a surrogate measure of medial compartment loading. Foot orthoses are designed to reduce knee adduction moment via lateral wedging. The 'dose' of wedging required to optimally unload the affected compartment is unknown and variable between individuals. This study explores a personalised approach via three-dimensional printed foot orthotics to assess the biomechanical response when two design variables are altered: orthotic length and lateral wedging. Foot orthoses were created for 10 individuals with symptomatic medial knee osteoarthritis and 10 controls. Computer-aided design software was used to design four full and four three-quarter-length foot orthoses per participant each with lateral posting of 0° 'neutral', 5° rearfoot, 10° rearfoot and 5° forefoot/10° rearfoot. Three-dimensional printers were used to manufacture all foot orthoses. Three-dimensional gait analyses were performed and selected knee kinetics were analysed: first peak knee adduction moment, second peak knee adduction moment, first knee flexion moment and knee adduction moment impulse. Full-length foot orthoses provided greater reductions in first peak knee adduction moment (p = 0.038), second peak knee adduction moment (p = 0.018) and knee adduction moment impulse (p = 0.022) compared to three-quarter-length foot orthoses. Dose effect of lateral wedging was found for first peak knee adduction moment (p knee adduction moment (p knee adduction moment impulse (p knee adduction moment (p = 0.028) and knee adduction moment impulse (p = 0.036). Significant interaction effects were found between orthotic length and wedging condition for second peak knee adduction moment (p = 0.002). No significant changes in first knee flexion moment were found. Individual heterogeneous responses to foot orthosis conditions were observed for first peak knee adduction moment, second peak knee adduction moment and knee adduction moment impulse. Biomechanical response

Study on three dimensional seismic isolation system

International Nuclear Information System (INIS)

Morishita, Masaki; Kitamura, Seiji

2003-01-01

Japan Nuclear Cycle Development Institute (JNC) and Japan Atomic Power Company (JAPC) launched joint research programs on structural design and three-dimensional seismic isolation technologies, as part of the supporting R and D activities for the feasibility studies on commercialized fast breeder reactor cycle systems. A research project by JAPC under the auspices of the Ministry of Economy, Trade, and Industry (METI) with technical support by JNC is included in this joint study. This report contains the results of the research on the three-dimensional seismic isolation technologies, and the results of this year's study are summarized in the following five aspects. (1) Study on Earthquake Condition for Developing 3-dimensional Base Isolation System. The case study S2 is one of the maximum ground motions, of which the records were investigated up to this time. But a few observed near the fault exceed the case study S2 in the long period domain, depending on the fault length and conditions. Generally it is appropriate that the response spectra ratio (vertical/horizontal) is 0.6. (2) Performance Requirement for 3-dimensional Base Isolation System and Devices. Although the integrity map of main equipment/piping dominate the design criteria for the 3-dimensional base isolation system, the combined integrity map is the same as those of FY 2000, which are under fv=1Hz and over hv=20%. (3) Developing Targets and Schedule for 3-dimensional Isolation Technology. The target items for 3-dimensional base isolation system were rearranged into a table, and developing items to be examined concerning the device were also adjusted. A development plan until FY 2009 was made from the viewpoint of realization and establishment of a design guideline on 3-dimensional base isolation system. (4) Study on 3-dimensional Entire Building Base Isolation System. Three ideas among six ideas that had been proposed in FY2001, i.e., '3-dimensional base isolation system incorporating hydraulic

Study on transient of fluidized bed nuclear reactor

International Nuclear Information System (INIS)

Streck, E.E.

1988-01-01

The point kinetic equations for a Fluidized-Bed Nuclear Reactor are solved by the method of Hansen. Due to the time varying nature of the reactor volume, the equations have a non-conventional formulation (moving boundary problem), but the method of solution preserves its asymptotic convergence and efficiency characteristics under this formulation. A one dimensional and linearized thermal hydraulics feedback model was coupled to the point kinetic equations in order to obtain a more realistic representation of the reactor power. The resulting equations are solved by the Euler explicit method. (author)

A three-dimensional pin-wise analysis for CEA ejection accident

Energy Technology Data Exchange (ETDEWEB)

Park, Guen-Tae; Park, Min-Ho; Park, Jin-Woo; Um, Kil-Sup; Choi, Tong-Soo [KEPCO NF, Daejeon (Korea, Republic of)

2016-10-15

The ejection of a control element assembly (CEA) with high reactivity worth causes the sudden insertion of reactivity into the core. Immediately after the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the doppler effect becomes important and turns the reactivity balance and power down to lower levels. The 3-D CEA ejection analysis methodology has been developed using the multi-dimensional code coupling system, CHASER, which couples three dimensional core neutron kinetics code ASTRA, subchannel analysis code THALES, and fuel performance analysis code FROST using message passing interface (MPI). This paper presents the pin-by-pin level analysis result with the 3-D CEA ejection analysis methodology using the CHASER. The pin-by-pin level analysis consists of DNBR, enthalpy and Pellet/Clad Mechanical Interaction (PCMI) analysis. All the evaluations are simulated for APR1400 plant loaded with PLUS7 fuel. In this paper, the pin-by-pin analysis using the multidimensional core transient code, CHASER, is presented with respect to enthalpy, DNBR and PCMI for APR1400 plant loaded with PLUS7 fuel. For the pin-by-pin enthalpy and DNBR analysis, the quarter core for HFP case or 15 - 20 assemblies around the most severe assembly for part powers or HZP cases are selected. And PCMI calculation is performed for all the rods in the whole core during a conservative time period. The pin-by-pin analysis results show that the regulatory guidelines of CEA ejection accident are satisfied.

Modified kinetic-hydraulic UASB reactor model for treatment of wastewater containing biodegradable organic substrates.

Science.gov (United States)

El-Seddik, Mostafa M; Galal, Mona M; Radwan, A G; Abdel-Halim, Hisham S

2016-01-01

This paper addresses a modified kinetic-hydraulic model for up-flow anaerobic sludge blanket (UASB) reactor aimed to treat wastewater of biodegradable organic substrates as acetic acid based on Van der Meer model incorporated with biological granules inclusion. This dynamic model illustrates the biomass kinetic reaction rate for both direct and indirect growth of microorganisms coupled with the amount of biogas produced by methanogenic bacteria in bed and blanket zones of reactor. Moreover, the pH value required for substrate degradation at the peak specific growth rate of bacteria is discussed for Andrews' kinetics. The sensitivity analyses of biomass concentration with respect to fraction of volume of reactor occupied by granules and up-flow velocity are also demonstrated. Furthermore, the modified mass balance equations of reactor are applied during steady state using Newton Raphson technique to obtain a suitable degree of freedom for the modified model matching with the measured results of UASB Sanhour wastewater treatment plant in Fayoum, Egypt.

Three-dimensional ICT reconstruction

International Nuclear Information System (INIS)

Zhang Aidong; Li Ju; Chen Fa; Sun Lingxia

2005-01-01

The three-dimensional ICT reconstruction method is the hot topic of recent ICT technology research. In the context, qualified visual three-dimensional ICT pictures are achieved through multi-piece two-dimensional images accumulation by, combining with thresholding method and linear interpolation. Different direction and different position images of the reconstructed pictures are got by rotation and interception respectively. The convenient and quick method is significantly instructive to more complicated three-dimensional reconstruction of ICT images. (authors)

Three-dimensional ICT reconstruction

International Nuclear Information System (INIS)

Zhang Aidong; Li Ju; Chen Fa; Sun Lingxia

2004-01-01

The three-dimensional ICT reconstruction method is the hot topic of recent ICT technology research. In the context qualified visual three-dimensional ICT pictures are achieved through multi-piece two-dimensional images accumulation by order, combining with thresholding method and linear interpolation. Different direction and different position images of the reconstructed pictures are got by rotation and interception respectively. The convenient and quick method is significantly instructive to more complicated three-dimensional reconstruction of ICT images. (authors)

Numerical Solution of Fractional Neutron Point Kinetics Model in Nuclear Reactor

Directory of Open Access Journals (Sweden)

Nowak Tomasz Karol

2014-06-01

Full Text Available This paper presents results concerning solutions of the fractional neutron point kinetics model for a nuclear reactor. Proposed model consists of a bilinear system of fractional and ordinary differential equations. Three methods to solve the model are presented and compared. The first one entails application of discrete Grünwald-Letnikov definition of the fractional derivative in the model. Second involves building an analog scheme in the FOMCON Toolbox in MATLAB environment. Third is the method proposed by Edwards. The impact of selected parameters on the model’s response was examined. The results for typical input were discussed and compared.

Three-dimensional analysis of craniofacial bones using three-dimensional computer tomography

Energy Technology Data Exchange (ETDEWEB)

Ono, Ichiro; Ohura, Takehiko; Kimura, Chu (Hokkaido Univ., Sapporo (Japan). School of Medicine) (and others)

1989-08-01

Three-dimensional computer tomography (3DCT) was performed in patients with various diseases to visualize stereoscopically the deformity of the craniofacial bones. The data obtained were analyzed by the 3DCT analyzing system. A new coordinate system was established using the median sagittal plane of the face (a plane passing through sella, nasion and basion) on the three-dimensional image. Three-dimensional profilograms were prepared for detailed analysis of the deformation of craniofacial bones for cleft lip and palate, mandibular prognathia and hemifacial microsomia. For patients, asymmetry in the frontal view and twist-formed complicated deformities were observed, as well as deformity of profiles in the anteroposterior and up-and-down directions. A newly developed technique allows three-dimensional visualization of changes in craniofacial deformity. It would aid in determining surgical strategy, including crani-facial surgery and maxillo-facial surgery, and in evaluating surgical outcome. (N.K.).

Three-dimensional analysis of craniofacial bones using three-dimensional computer tomography

International Nuclear Information System (INIS)

Ono, Ichiro; Ohura, Takehiko; Kimura, Chu

1989-01-01

Three-dimensional computer tomography (3DCT) was performed in patients with various diseases to visualize stereoscopically the deformity of the craniofacial bones. The data obtained were analyzed by the 3DCT analyzing system. A new coordinate system was established using the median sagittal plane of the face (a plane passing through sella, nasion and basion) on the three-dimensional image. Three-dimensional profilograms were prepared for detailed analysis of the deformation of craniofacial bones for cleft lip and palate, mandibular prognathia and hemifacial microsomia. For patients, asymmetry in the frontal view and twist-formed complicated deformities were observed, as well as deformity of profiles in the anteroposterior and up-and-down directions. A newly developed technique allows three-dimensional visualization of changes in craniofacial deformity. It would aid in determining surgical strategy, including crani-facial surgery and maxillo-facial surgery, and in evaluating surgical outcome. (N.K.)

Measurements for kinetic parameters estimation in the RA-0 research reactor

International Nuclear Information System (INIS)

Gomez, A; Bellino, P A

2012-01-01

In the present work, measurements based on the neutron noise technique and the inverse kinetic method were performed to estimate the different kinetic parameters of the reactor in its critical state. By means of the neutron noise technique, we obtained the current calibration factor of the ionization chamber M6 belonging to the power range channels of the reactor instrumentation. The maximum current allowed compatible with the maximum power authorized by the operation license was also obtained. Using the neutron noise technique, the reduced mean reproduction time (Λ*) was estimated. This parameter plays a fundamental role in the deterministic analysis of criticality accidents. Comparison with previous values justified performing new measurements to study systematic trends in the value of Λ*. Using the inverse kinetics method, the reactivity worth of the control rods was estimated, confirming the existence of spatial effects and trends previously observed (author)

Development of a calculation method for one dimensional kinetic analysis in fission reactors, with feedback effects

International Nuclear Information System (INIS)

Paixao, S.B.

1985-01-01

The methodology used in the WIGLE3 computer code is studied. This methodology has been applied for the steady-state and transient solutions of the one-dimensional, two-group, diffusion equations in slab geometry, in axial type probelm analysis. It's also studied, based in a WIGLE3 computer code, reactor representative models, considering non-boiling heat transfer. A steady-state program for control rod bank position search- CITER 1D- has been developed. Some criticality research on the proposed system has been done using different control rod bank initial positions, time steps and convergence parameters. (E.G.) [pt

Numerical solution of the point reactor kinetics equations with fuel burn-up and temperature feedback

International Nuclear Information System (INIS)

Tashakor, S.; Jahanfarnia, G.; Hashemi-Tilehnoee, M.

2010-01-01

Point reactor kinetics equations are solved numerically using one group of delayed neutrons and with fuel burn-up and temperature feedback included. To calculate the fraction of one-group delayed neutrons, a group of differential equations are solved by an implicit time method. Using point reactor kinetics equations, changes in mean neutrons density, temperature, and reactivity are calculated in different times during the reactor operation. The variation of reactivity, temperature, and maximum power with time are compared with the predictions by other methods.

Performance, Accuracy and Efficiency Evaluation of a Three-Dimensional Whole-Core Neutron Transport Code AGENT

International Nuclear Information System (INIS)

Jevremovic, Tatjana; Hursin, Mathieu; Satvat, Nader; Hopkins, John; Xiao, Shanjie; Gert, Godfree

2006-01-01

The AGENT (Arbitrary Geometry Neutron Transport) an open-architecture reactor modeling tool is deterministic neutron transport code for two or three-dimensional heterogeneous neutronic design and analysis of the whole reactor cores regardless of geometry types and material configurations. The AGENT neutron transport methodology is applicable to all generations of nuclear power and research reactors. It combines three theories: (1) the theory of R-functions used to generate real three-dimensional whole-cores of square, hexagonal or triangular cross sections, (2) the planar method of characteristics used to solve isotropic neutron transport in non-homogenized 2D) reactor slices, and (3) the one-dimensional diffusion theory used to couple the planar and axial neutron tracks through the transverse leakage and angular mesh-wise flux values. The R-function-geometrical module allows a sequential building of the layers of geometry and automatic sub-meshing based on the network of domain functions. The simplicity of geometry description and selection of parameters for accurate treatment of neutron propagation is achieved through the Boolean algebraic hierarchically organized simple primitives into complex domains (both being represented with corresponding domain functions). The accuracy is comparable to Monte Carlo codes and is obtained by following neutron propagation through real geometrical domains that does not require homogenization or simplifications. The efficiency is maintained through a set of acceleration techniques introduced at all important calculation levels. The flux solution incorporates power iteration with two different acceleration techniques: Coarse Mesh Re-balancing (CMR) and Coarse Mesh Finite Difference (CMFD). The stand-alone originally developed graphical user interface of the AGENT code design environment allows the user to view and verify input data by displaying the geometry and material distribution. The user can also view the output data such

Kinetics, dynamics and neutron noise in Molten Salt Reactors

International Nuclear Information System (INIS)

Pazsit, Imre

2013-01-01

Reactor kinetic and dynamic properties of Molten Salt Reactors (MSR) are investigated in a simple model, which allows closed compact analytical solutions to be obtained. The goal is to gain insight, rather than to produce high-quality quantitative data. Through an interpretation of the different terms in the basic equations, and by means of analytical solutions, various approximations are introduced and their validity discussed. The dynamical behaviour of MSRs and their response to small stationary perturbations is described and discussed in comparison with traditional systems. (author)

Effect of process operating conditions in the biomass torrefaction: A simulation study using one-dimensional reactor and process model

International Nuclear Information System (INIS)

Park, Chansaem; Zahid, Umer; Lee, Sangho; Han, Chonghun

2015-01-01

Torrefaction reactor model is required for the development of reactor and process design for biomass torrefaction. In this study, a one-dimensional reactor model is developed based on the kinetic model describing volatiles components and solid evolution and the existing thermochemical model considering the heat and mass balance. The developed reactor model used the temperature and flow rate of the recycled gas as the practical manipulated variables instead of the torrefaction temperature. The temperature profiles of the gas and solid phase were generated, depending on the practical thermal conditions, using developed model. Moreover, the effect of each selected operating variables on the parameters of the torrefaction process and the effect of whole operating variables with particular energy yield were analyzed. Through the results of sensitivity analysis, it is shown that the residence time insignificantly influenced the energy yield when the flow rate of recycled gas is low. Moreover, higher temperature of recycled gas with low flow rate and residence time produces the attractive properties, including HHV and grindability, of torrefied biomass when the energy yield is specified. - Highlights: • A one-dimensional reactor model for biomass torrefaction is developed considering the heat and mass balance. • The developed reactor model uses the temperature and flow rate of the recycled gas as the practical manipulated variables. • The effect of operating variables on the parameters of the torrefaction process is analyzed. • The results of sensitivity analysis represent notable discussions which were not done by the previous researches

FORE-2, Thermohydraulics and Space-Independent Reactor Kinetics for Transients

International Nuclear Information System (INIS)

Fox, J.N.; Lawler, B.E.; Butz, H.R.; Heames, T.J.

1984-01-01

1 - Description of problem or function: FORE2 is a coupled thermal hydraulics-point kinetics digital computer code designed to calculate significant reactor parameters under steady-state conditions, or as functions of time during transients. The transients may result from a programmed reactivity insertion or a power change. Variable inlet coolant flow rate and temperature are considered. The code calculates the reactor power, the individual reactivity feedbacks, and the temperature of coolant, cladding, fuel, structure, and additional material for up to seven axial positions in three channel types which represent radial zones of the reactor. The heat of fusion, accompanying fuel melting, the liquid metal voiding reactivity, and the spatial and the time variation of the fuel cladding gap coefficient due to changes in gap size are considered. 2 - Method of solution: FORE2 input consists of property data, geometry, power and flow distribution factors, external time varying functions, experimental coefficients, and termination data. The differential equations for fluid flow, heat transfer, and point neutronics are solved by explicit finite-difference procedures. 3 - Restrictions on the complexity of the problem: Reactor excursions which can be calculated are restricted to those transients in which the reactor is not substantially destroyed. As a general rule, changes in reactor geometry and composition during an excursion are limited to those cases in which the reactivity effects of the changes may be considered as small perturbations of the initial system. Thus, accidents involving large-scale disassembly and bulk meltdown of a core are not covered by FORE2. FORE2 is valid only while the core retains its initial geometry

Three-dimensional simulation of viscous-flow agglomerate sintering.

Science.gov (United States)

Kirchhof, M J; Schmid, H -J; Peukert, W

2009-08-01

The viscous-flow sintering of different agglomerate particle morphologies is studied by three-dimensional computer simulations based on the concept of fractional volume of fluid. For a fundamental understanding of particle sintering characteristics, the neck growth kinetics in agglomerate chains and in doublets consisting of differently sized primary particles is investigated. Results show that different sintering contacts in agglomerates even during the first stages are not completely independent from each other, even though differences are small. The neck growth kinetics of differently sized primary particles is determined by the smaller one up to a size difference by a factor of approximately 2, whereas for larger size differences, the kinetics becomes faster. In particular, the agglomerate sintering kinetics is investigated for particle chains of different lengths and for different particle morphologies each having ten primary particles and nine initial sintering contacts. For agglomerate chains, the kinetics approximately can be normalized by using the radius of the fully coalesced sphere. In general, different agglomerate morphologies show equal kinetics during the first sintering stages, whereas during advanced stages, compact morphologies show significantly faster sintering progress than more open morphologies. Hence, the overall kinetics cannot be described by simply using constant morphology correction factors such as fractal dimension or mean coordination number which are used in common sintering models. However, for the first stages of viscous-flow agglomerate sintering, which are the most important for many particle processes, a sintering equation is presented. Although we use agglomerates consisting of spherical primary particles, our methodology can be applied to other aggregate geometries as well.

Three dimensional strained semiconductors

Science.gov (United States)

Voss, Lars; Conway, Adam; Nikolic, Rebecca J.; Leao, Cedric Rocha; Shao, Qinghui

2016-11-08

In one embodiment, an apparatus includes a three dimensional structure comprising a semiconductor material, and at least one thin film in contact with at least one exterior surface of the three dimensional structure for inducing a strain in the structure, the thin film being characterized as providing at least one of: an induced strain of at least 0.05%, and an induced strain in at least 5% of a volume of the three dimensional structure. In another embodiment, a method includes forming a three dimensional structure comprising a semiconductor material, and depositing at least one thin film on at least one surface of the three dimensional structure for inducing a strain in the structure, the thin film being characterized as providing at least one of: an induced strain of at least 0.05%, and an induced strain in at least 5% of a volume of the structure.

Three-dimensional effects of curved plasma actuators in quiescent air

International Nuclear Information System (INIS)

Wang Chincheng; Durscher, Ryan; Roy, Subrata

2011-01-01

This paper presents results on a new class of curved plasma actuators for the inducement of three-dimensional vortical structures. The nature of the fluid flow inducement on a flat plate, in quiescent conditions, due to four different shapes of dielectric barrier discharge (DBD) plasma actuators is numerically investigated. The three-dimensional plasma kinetic equations are solved using our in-house, finite element based, multiscale ionized gas (MIG) flow code. Numerical results show electron temperature and three dimensional plasma force vectors for four shapes, which include linear, triangular, serpentine, and square actuators. Three-dimensional effects such as pinching and spreading the neighboring fluid are observed for serpentine and square actuators. The mechanisms of vorticity generation for DBD actuators are discussed. Also the influence of geometric wavelength (λ) and amplitude (Λ) of the serpentine and square actuators on vectored thrust inducement is predicted. This results in these actuators producing significantly better flow mixing downstream as compared to the standard linear actuator. Increasing the wavelengths of serpentine and square actuators in the spanwise direction is shown to enhance the pinching effect giving a much higher vertical velocity. On the contrary, changing the amplitude of the curved actuator varies the streamwise velocity significantly influencing the near wall jet. Experimental data for a serpentine actuator are also reported for validation purpose.

Study of the stochastic point reactor kinetic equation

International Nuclear Information System (INIS)

Gotoh, Yorio

1980-01-01

Diagrammatic technique is used to solve the stochastic point reactor kinetic equation. The method gives exact results which are derived from Fokker-Plank theory. A Green's function dressed with the clouds of noise is defined, which is a transfer function of point reactor with fluctuating reactivity. An integral equation for the correlation function of neutron power is derived using the following assumptions: 1) Green's funntion should be dressed with noise, 2) The ladder type diagrams only contributes to the correlation function. For a white noise and the one delayed neutron group approximation, the norm of the integral equation and the variance to mean-squared ratio are analytically obtained. (author)

Nanostructural evolution in surveillance test specimens of a commercial nuclear reactor pressure vessel studied by three-dimensional atom probe and positron annihilation

International Nuclear Information System (INIS)

Toyama, T.; Nagai, Y.; Tang, Z.; Hasegawa, M.; Almazouzi, A.; Walle, E. van; Gerard, R.

2007-01-01

The nanostructural evolution of irradiation-induced Cu-rich nanoprecipitates (CRNPs) and vacancy clusters in surveillance test specimens of in-service commercial nuclear reactor pressure vessel steel welds of Doel-1 and Doel-2 are revealed by combining the three-dimensional local electrode atom probe and positron annihilation techniques. In both medium (0.13 wt.%) and high (0.30 wt.%) Cu welds, the CRNPs are found to form readily at the very beginning of the reactor lifetime. Thereafter, during the subsequent 30 years of operation, the residual Cu concentration in the matrix shows a slight decrease while the CRNPs coarsen. On the other hand, small vacancy clusters of V 3 -V 4 start appearing after the initial Cu precipitation and accumulate steadily with increasing neutron dose. The observed nanostructural evolution is shown to provide unique and fundamental information about the mechanisms of the irradiation-induced embrittlement of these specific materials

Study on the numerical analysis of nuclear reactor kinetics equations

International Nuclear Information System (INIS)

Yang, J.C.

1980-01-01

A two-step alternating direction explict method is proposed for the solution of the space-and time-dependent diffusion theory reactor kinetics equations in two space dimensions as a special case of the general class of alternating direction implicit method and the truncation error of this method is estimated. To test the validity of this method it is applied to the Pressurized Water Reactor and CANDU-PHW reactor which have been operating and underconstructing in Korea. The time dependent neutron flux of the PWR reactor during control rod insertion and time dependent neutronic power of CANDU-PHW reactor in the case of postulated loss of coolant accident are obtained from the numerical calculation results. The results of the PWR reactor problem are shown the close agreement between implicit-difference method used in the TWIGL program and this method, and the results of the CANDU-PHW reactor are compared with the results of improved quasistic method and modal method. (Author)

Installation of a three-dimensional simulation method for core-physical description of pebble bed reactors with multiple recycling process at the example of the AVR

International Nuclear Information System (INIS)

Grotkamp, T.

1984-01-01

To describe the core-physical behaviour of pebble bed reactors simulation models are used, which reproduce the burn-up/recycling and - resulting - calculate criticality and neutron spectrum as well as neutron flux and temperature distribution. Modelling the AVR-reactor requires a three-dimensional treating for detailed considerations because of the graphite noses extending into the core. Such a system is built up in the present work and compared with the results of the two-dimensional model standardizated from operational side. The agreement is so good that the latter one is sufficient for the calculations accompanying the operation. The comparison with results of measurement is very satisfying in regard to fuel element distribution and temperature coefficient. As in all theoretical investigations there stays a discrepance of a little more than 1 nile against the measurement at the reactivity equivalence of the AVR rod-bank. On the other hand it is possible to reproduce the rod-bank curve resulting of the calibration very exactly with the present model. (orig.) [de

Application of the three-dimensional transport code to analysis of the neutron streaming experiment

International Nuclear Information System (INIS)

Chatani, K.; Slater, C.O.

1990-01-01

The neutron streaming through an experimental mock-up of a Clinch River Breeder Reactor (CRBR) prototypic coolant pipe chaseway was recalculated with a three-dimensional discrete ordinates code. The experiment was conducted at the Tower Shielding Facility at Oak Ridge National Laboratory in 1976 and 1977. The measurement of the neutron flux, using Bonner ball detectors, indicated nine orders of attenuation in the empty pipeway, which contained two 90-deg bends and was surrounded by concrete walls. The measurement data were originally analyzed using the DOT3.5 two-dimensional discrete ordinates radiation transport code. However, the results did not agree with measurement data at the bend because of the difficulties in modeling the three-dimensional configurations using two-dimensional methods. The two-dimensional calculations used a three-step procedure in which each of the three legs making the two 90-deg bends was a separate calculation. The experiment was recently analyzed with the TORT three-dimensional discrete ordinates radiation transport code, not only to compare the calculational results with the experimental results, but also to compare with results obtained from analyses in Japan using DOT3.5, MORSE, and ENSEMBLE, which is a three-dimensional discrete ordinates radiation transport code developed in Japan

Effects of Electrical Stimulation on the Degradation of Azo Dye in Three-Dimensional Biofilm Electrode Reactors

Directory of Open Access Journals (Sweden)

Xian Cao

2017-04-01

Full Text Available Three-dimensional biofilm electrode reactors (3D-BERs were constructed to degrade the azo dye Reactive Brilliant Red (RBR X-3B. The 3D-BERs with different influent concentrations and external voltages were individually studied to investigate their influence on the removal of X-3B. Experimental results showed that 3D-BERs have good X-3B removal efficiency; even when the influent concentration was 800 mg/L, removal efficiency of 73.4% was still achieved. In addition, the X-3B removal efficiency stabilized shortly after the influent concentration increased. In 3D-BERs, the average X-3B removal efficiency increased from 52.8% to 85.4% when the external voltage rose from 0 to 2 V. We further identified the intermediate products via UV-Vis and gas chromatography-mass spectrometry (GC-MS analyses, and discussed the potential mechanism of degradation. After the conjugate structure of X-3B was destroyed, all of the substances generated mainly consisted of lower-molecular-weight organics.

Theory and application of a three-dimensional code SHAPS to complex piping systems

International Nuclear Information System (INIS)

Wang, C.Y.

1983-01-01

This paper describes the theory and application of a three-dimensional computer code SHAPS to the complex piping systems. The code utilizes a two-dimensional implicit Eulerian method for the hydrodynamic analysis together with a three-dimensional elastic-plastic finite-element program for the structural calculation. A three-dimensional pipe element with eight degrees of freedom is employed to account for the hoop, flexural, axial, and the torsional mode of the piping system. In the SHAPS analysis the hydrodynamic equations are modified to include the global piping motion. Coupling between fluid and structure is achieved by enforcing the free-slip boundary conditions. Also, the response of the piping network generated by the seismic excitation can be included. A thermal transient capability is also provided in SHAPS. To illustrate the methodology, many sample problems dealing with the hydrodynamic, structural, and thermal analyses of reactor-piping systems are given. Validation of the SHAPS code with experimental data is also presented

Space-time neutronic analysis of postulated LOCA's in CANDU reactors

International Nuclear Information System (INIS)

Luxat, J.C.; Frescura, G.M.

1978-01-01

Space-time neutronic behaviour of CANDU reactors is of importance in the analysis and design of reactor safety systems. A methodology has been developed for simulating CANDU space-time neutronics with application to the analysis of postulated LOCA'S. The approach involves the efficient use of a set of computer codes which provide a capability to perform simulations ranging from detailed, accurate 3-dimensional space-time to low-cost survey calculations using point kinetics with some ''effective'' spatial content. A new, space-time kinetics code based upon a modal expansion approach is described. This code provides an inexpensive and relatively accurate scoping tool for detailed 3-dimensional space-time simulations. (author)

Improvement of multi-dimensional realistic thermal-hydraulic system analysis code, MARS 1.3

International Nuclear Information System (INIS)

Lee, Won Jae; Chung, Bub Dong; Jeong, Jae Jun; Ha, Kwi Seok

1998-09-01

The MARS (Multi-dimensional Analysis of Reactor Safety) code is a multi-dimensional, best-estimate thermal-hydraulic system analysis code. This report describes the new features that have been improved in the MARS 1.3 code since the release of MARS 1.3 in July 1998. The new features include: - implementation of point kinetics model into the 3D module - unification of the heat structure model - extension of the control function to the 3D module variables - improvement of the 3D module input check function. Each of the items has been implemented in the developmental version of the MARS 1.3.1 code and, then, independently verified and assessed. The effectiveness of the new features is well verified and it is shown that these improvements greatly extend the code capability and enhance the user friendliness. Relevant input data changes are also described. In addition to the improvements, this report briefly summarizes the future code developmental activities that are being carried out or planned, such as coupling of MARS 1.3 with the containment code CONTEMPT and the three-dimensional reactor kinetics code MASTER 2.0. (author). 8 refs

Improvement of multi-dimensional realistic thermal-hydraulic system analysis code, MARS 1.3

Energy Technology Data Exchange (ETDEWEB)

Lee, Won Jae; Chung, Bub Dong; Jeong, Jae Jun; Ha, Kwi Seok

1998-09-01

The MARS (Multi-dimensional Analysis of Reactor Safety) code is a multi-dimensional, best-estimate thermal-hydraulic system analysis code. This report describes the new features that have been improved in the MARS 1.3 code since the release of MARS 1.3 in July 1998. The new features include: - implementation of point kinetics model into the 3D module - unification of the heat structure model - extension of the control function to the 3D module variables - improvement of the 3D module input check function. Each of the items has been implemented in the developmental version of the MARS 1.3.1 code and, then, independently verified and assessed. The effectiveness of the new features is well verified and it is shown that these improvements greatly extend the code capability and enhance the user friendliness. Relevant input data changes are also described. In addition to the improvements, this report briefly summarizes the future code developmental activities that are being carried out or planned, such as coupling of MARS 1.3 with the containment code CONTEMPT and the three-dimensional reactor kinetics code MASTER 2.0. (author). 8 refs.

Three-dimensional thermal hydraulic best estimate code BAGIRA: new results of verification

International Nuclear Information System (INIS)

Peter Kohut; Sergey D Kalinichenko; Alexander E Kroshilin; Vladimir E Kroshilin; Alexander V Smirnov

2005-01-01

Full text of publication follows: BAGIRA is a three-dimensional inhomogeneous two-velocity two-temperature thermal hydraulic code of best estimate, elaborated in VNIIAES for modeling two-phase flows in the primary circuit and steam generators of VVER-type nuclear reactors under various accident, transient or normal operation conditions. In this talk we present verification results of the BAGIRA code, obtained on the basis of different experiments performed on special and integral thermohydraulic experimental facilities as well as on real NPPs. Special attention is paid to the verification of three-dimensional flow models. Besides that we expose new results of the code benchmark analysis made on the basis of two recent LOCA-type experiments - 'Leak 2 x 25% from the hot leg double-side rupture' and 'Leak 3% from the cold leg' - performed on the PSB-VVER integral test facility (Electrogorsk Research and Engineering Center, Electrogorsk, Russia) - the most up-to-date Russian large-scale four-loop unit which has been designed for modelling the primary circuit of VVER-1000 type reactors. (authors)

Non-destructive characterization of recrystallization kinetics using three-dimensional X-ray diffraction microscopy

DEFF Research Database (Denmark)

Lauridsen, E.M.; Schmidt, Søren; Fæster Nielsen, Søren

2006-01-01

Three-dimensional X-ray diffraction (3DXRD) is used to characterize the nucleation and early growth of individual bulk nuclei in situ during recrystallization of 92% cold-rolled copper. It is found that some cube nuclei, but not all, have a significantly faster initial growth than the average...

New method for the determination of precipitation kinetics using a laminar jet reactor

NARCIS (Netherlands)

Al Tarazi, M.Y.M.; Heesink, Albertus B.M.; Versteeg, Geert

2005-01-01

In this paper a new experimental method for determining the kinetics of fast precipitation reactions is introduced. Use is made of a laminar jet reactor, which is also frequently applied to determine the kinetics of homogeneous gas–liquid reactions. The liquid containing one or more of the

New method for the determination of precipitation kinetics using a laminar jet reactor

NARCIS (Netherlands)

Al-Tarazi, Mousa; Heesink, A. Bert M.; Versteeg, Geert F.

2005-01-01

In this paper a new experimental method for determining the kinetics of fast precipitation reactions is introduced. Use is made of a laminar jet reactor, which is also frequently applied to determine the kinetics of homogeneous gas-liquid reactions. The liquid containing one or more of the

Three Mile Island Unit 1 Main Steam Line Break Three-Dimensional Neutronics/Thermal-Hydraulics Analysis: Application of Different Coupled Codes

International Nuclear Information System (INIS)

D'Auria, Francesco; Moreno, Jose Luis Gago; Galassi, Giorgio Maria; Grgic, Davor; Spadoni, Antonino

2003-01-01

A comprehensive analysis of the double ended main steam line break (MSLB) accident assumed to occur in the Babcock and Wilcox Three Mile Island Unit 1 (TMI-1) has been carried out at the Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione of the University of Pisa, Italy, in cooperation with the University of Zagreb, Croatia. The overall activity has been completed within the framework of the participation in the Organization for Economic Cooperation and Development-Committee on the Safety of Nuclear Installations-Nuclear Science Committee pressurized water reactor MSLB benchmark.Thermal-hydraulic system codes (various versions of Relap5), three-dimensional (3-D) neutronics codes (Parcs, Quabbox, and Nestle), and one subchannel code (Cobra) have been adopted for the analysis. Results from the following codes (or code versions) are assumed as reference:1. Relap5/mod3.2.2, beta version, coupled with the 3-D neutron kinetics Parcs code parallel virtual machine (PVM) coupling2. Relap5/mod3.2.2, gamma version, coupled with the 3-D neutron kinetics Quabbox code (direct coupling)3. Relap5/3D code coupled with the 3-D neutron kinetics Nestle code.The influence of PVM and of direct coupling is also discussed.Boundary and initial conditions of the system, including those relevant to the fuel status, have been supplied by Pennsylvania State University in cooperation with GPU Nuclear Corporation (the utility, owner of TMI) and the U.S. Nuclear Regulatory Commission. The comparison among the results obtained by adopting the same thermal-hydraulic nodalization and the coupled code version is discussed in this paper.The capability of the control rods to recover the accident has been demonstrated in all the cases as well as the capability of all the codes to predict the time evolution of the assigned transient. However, one stuck control rod caused some 'recriticality' or 'return to power' whose magnitude is largely affected by boundary and initial conditions

Three-dimensional, three-component wall-PIV

Science.gov (United States)

Berthe, André; Kondermann, Daniel; Christensen, Carolyn; Goubergrits, Leonid; Garbe, Christoph; Affeld, Klaus; Kertzscher, Ulrich

2010-06-01

This paper describes a new time-resolved three-dimensional, three-component (3D-3C) measurement technique called wall-PIV. It was developed to assess near wall flow fields and shear rates near non-planar surfaces. The method is based on light absorption according to Beer-Lambert’s law. The fluid containing a molecular dye and seeded with buoyant particles is illuminated by a monochromatic, diffuse light. Due to the dye, the depth of view is limited to the near wall layer. The three-dimensional particle positions can be reconstructed by the intensities of the particle’s projection on an image sensor. The flow estimation is performed by a new algorithm, based on learned particle trajectories. Possible sources of measurement errors related to the wall-PIV technique are analyzed. The accuracy analysis was based on single particle experiments and a three-dimensional artificial data set simulating a rotating sphere.

A Derivation of Source-based Kinetics Equation with Time Dependent Fission Kernel for Reactor Transient Analyses

International Nuclear Information System (INIS)

Kim, Song Hyun; Woo, Myeong Hyun; Shin, Chang Ho; Pyeon, Cheol Ho

2015-01-01

In this study, a new balance equation to overcome the problems generated by the previous methods is proposed using source-based balance equation. And then, a simple problem is analyzed with the proposed method. In this study, a source-based balance equation with the time dependent fission kernel was derived to simplify the kinetics equation. To analyze the partial variations of reactor characteristics, two representative methods were introduced in previous studies; (1) quasi-statics method and (2) multipoint technique. The main idea of quasistatics method is to use a low-order approximation for large integration times. To realize the quasi-statics method, first, time dependent flux is separated into the shape and amplitude functions, and shape function is calculated. It is noted that the method has a good accuracy; however, it can be expensive as a calculation cost aspect because the shape function should be fully recalculated to obtain accurate results. To improve the calculation efficiency, multipoint method was proposed. The multipoint method is based on the classic kinetics equation with using Green's function to analyze the flight probability from region r' to r. Those previous methods have been used to analyze the reactor kinetics analysis; however, the previous methods can have some limitations. First, three group variables (r g , E g , t g ) should be considered to solve the time dependent balance equation. This leads a big limitation to apply large system problem with good accuracy. Second, the energy group neutrons should be used to analyze reactor kinetics problems. In time dependent problem, neutron energy distribution can be changed at different time. It can affect the change of the group cross section; therefore, it can lead the accuracy problem. Third, the neutrons in a space-time region continually affect the other space-time regions; however, it is not properly considered in the previous method. Using birth history of the neutron sources

Process and kinetics of the fundamental radiation-electrochemical reactions in the primary coolant loop of nuclear reactors

International Nuclear Information System (INIS)

Kozomara-Maic, S.

1987-06-01

In spite of the rather broad title of this report, its major part is devoted to the corrosion problems at the RA reactor, i.e. causes and consequences of the reactor shutdown in 1979 and 1982. Some problems of reactor chemistry are pointed out because they are significant for future reactor operation. The final conclusion of this report is that corrosion processes in the primary coolant circuit of the nuclear reactor are specific and that radiation effects cannot be excluded when processes and reaction kinetics are investigated. Knowledge about the kinetics of all the chemical reactions occurring in the primary coolant loop are of crucial significance for safe and economical reactor operation [sr

Numerical simulation of stochastic point kinetic equation in the dynamical system of nuclear reactor

International Nuclear Information System (INIS)

Saha Ray, S.

2012-01-01

Highlights: ► In this paper stochastic neutron point kinetic equations have been analyzed. ► Euler–Maruyama method and Strong Taylor 1.5 order method have been discussed. ► These methods are applied for the solution of stochastic point kinetic equations. ► Comparison between the results of these methods and others are presented in tables. ► Graphs for neutron and precursor sample paths are also presented. -- Abstract: In the present paper, the numerical approximation methods, applied to efficiently calculate the solution for stochastic point kinetic equations () in nuclear reactor dynamics, are investigated. A system of Itô stochastic differential equations has been analyzed to model the neutron density and the delayed neutron precursors in a point nuclear reactor. The resulting system of Itô stochastic differential equations are solved over each time-step size. The methods are verified by considering different initial conditions, experimental data and over constant reactivities. The computational results indicate that the methods are simple and suitable for solving stochastic point kinetic equations. In this article, a numerical investigation is made in order to observe the random oscillations in neutron and precursor population dynamics in subcritical and critical reactors.

Application of the exact distribution pj{sub k} in the determination of kinetic parameters in a reactor; Aplicacion de la distribucion exacta p{sub k} a la determinacion de parametros cineticos de un reactor

Energy Technology Data Exchange (ETDEWEB)

Alca Ruiz, F

1982-07-01

In this report one distribution of neutron counts obtained by a detector placed in a reactor is studied in order to be used in the determination of reactor kinetic parameters such as {beta}/{lambda} and reactivities. The parameters accuracy from this new method is compared with the Feynman and Mogilner method, based too in Reactor Neutron Noise Analysis. These three methods have been applied to JEN-2 reactor and the better accuracy and faster collection of experimental data give some interest to the new method which only requires a good footing code. (Author) 68 refs.

Modeling of a three-phase reactor for bitumen-derived gas oil hydrotreating

International Nuclear Information System (INIS)

Chacon, R.; Canale, A.; Bouza, A.; Sanchez, Y.

2012-01-01

A three-phase reactor model for describing the hydrotreating reactions of bitumen-derived gas oil was developed. The model incorporates the mass-transfer resistance at the gas-liquid and liquid-solid interfaces and a kinetic rate expression based on a Langmuir-Hinshelwood-type model. We derived three correlations for determining the solubility of hydrogen (H 2 ), hydrogen sulfide (H 2 S) and ammonia (NH 3 ) in hydrocarbon mixtures and the calculation of the catalyst effectiveness factor was included. Experimental data taken from the literature were used to determine the kinetic parameters (stoichiometric coefficients, reaction orders, reaction rate and adsorption constants for hydrodesulfuration (HDS) and hydrodenitrogenation (HDN)) and to validate the model under various operating conditions. Finally, we studied the effect of operating conditions such as pressure, temperature, LHSV, H 2 /feed ratio and the inhibiting effect of H 2 S on HDS and NH 3 on HDN. (author)

The analysis of RPV fast neutron flux calculation for PWR with three-dimensional SN method

International Nuclear Information System (INIS)

Yang Shouhai; Chen Yixue; Wang Weijin; Shi Shengchun; Lu Daogang

2011-01-01

Discrete ordinates (S N ) method is one of the most widely used method for reactor pressure vessel (RPV) design. As the fast development of computer CPU speed and memory capacity and consummation of three-dimensional discrete-ordinates method, it is mature for 3-D S N method to be used to engineering design for nuclear facilities. This work was done specifically for PWR model, with the results of 3-D core neutron transport calculation by 3-D core calculation, 3-D RPV fast neutron flux distribution obtain by 3-D S N method were compared with gained by 1-D and 2-D S N method and the 3-D Monte Carlo (MC) method. In this paper, the application of three-dimensional S N method in calculating RPV fast neutron flux distribution for pressurized water reactor (PWR) is presented and discussed. (authors)

Variational methods in the kinetic modeling of nuclear reactors: Recent advances

International Nuclear Information System (INIS)

Dulla, S.; Picca, P.; Ravetto, P.

2009-01-01

The variational approach can be very useful in the study of approximate methods, giving a sound mathematical background to numerical algorithms and computational techniques. The variational approach has been applied to nuclear reactor kinetic equations, to obtain a formulation of standard methods such as point kinetics and quasi-statics. more recently, the multipoint method has also been proposed for the efficient simulation of space-energy transients in nuclear reactors and in source-driven subcritical systems. The method is now founded on a variational basis that allows a consistent definition of integral parameters. The mathematical structure of multipoint and modal methods is also investigated, evidencing merits and shortcomings of both techniques. Some numerical results for simple systems are presented and the errors with respect to reference calculations are reported and discussed. (authors)

Predicting transition in two- and three-dimensional separated flows

International Nuclear Information System (INIS)

Cutrone, L.; De Palma, P.; Pascazio, G.; Napolitano, M.

2008-01-01

This paper is concerned with the numerical prediction of two- and three-dimensional transitional separated flows of turbomachinery interest. The recently proposed single-point transition model based on the use of a laminar kinetic energy transport equation is considered, insofar as it does not require to evaluate any integral parameter, such as boundary-layer thickness, and is thus directly applicable to three-dimensional flows. A well established model, combining a transition-onset correlation with an intermittency transport equation, is also used for comparison. Both models are implemented within a Reynolds-averaged Navier-Stokes solver employing a low-Reynolds-number k-ω turbulence model. The performance of the transition models have been evaluated and tested versus well-documented incompressible flows past a flat plate with semi-circular leading edge, namely: tests T3L2, T3L3, T3L5, and T3LA1 of ERCOFTAC, with different Reynolds numbers and free-stream conditions, the last one being characterized by a non-zero pressure gradient. In all computations, the first model has proven as adequate as or superior to the second one and has been then applied with success to two more complex test cases, for which detailed experimental data are available in the literature, namely: the two- and three-dimensional flows through the T106 linear turbine cascade

A WIMS-NESTLE reactor physics model for an RBMK reactor

International Nuclear Information System (INIS)

Perry, R.T.; Meriwether, G.H.

1996-01-01

This work describes the static neutronic calculations made for a three-dimensional model of an RBMK (Russian) reactor. Future work will involve the use of this neutronic model and a thermal-hydraulic model in coupled calculations. The lattice code, WIMS-D, was used to obtain the cross sections for the static neutronic calculations. The static reactor neutronic calculations were made with NESTLE, a three-dimensional nodal diffusion code. The methods used to establish an RBMK reactor model for use in these codes are discussed, and the cross sections calculated are given

A WIMS-NESTLE reactor physics model for an RBMK reactor

International Nuclear Information System (INIS)

Perry, R.T.; Meriwether, G.H.

1996-01-01

This work describes the static neutronic calculations made for a three-dimensional model of an RBMK (Russian) reactor. Future work will involve the use of this neutronic model and a thermal-hydraulic model in coupled calculations. The lattice code, WIMS-D, was used to obtain the cross sections for the static neutronic calculations. The static reactor neutronic calculations were made with NESTLE, a three-dimensional nodal diffusion code. The methods used to establish an RBMK reactor model for use in these codes are discussed, and the cross sections calculated are given. (author)

Isotopic exchange reactions. Kinetics and efficiency of the reactors using them in isotopic separation

International Nuclear Information System (INIS)

Ravoire, Jean

1979-11-01

In the first part, some definitions and the thermodynamic and kinetic isotopic effect concepts are recalled. In the second part the kinetic laws are established, in homogeneous and heterogeneous medium (one component being on occasions present in both phases), without and with isotopic effects. Emphasis is put on application to separation of isotopes, the separation factor α being close to 1, one isotope being in large excess with respect to the other one. Isotopic transfer is then given by: J = Ka (x - y/α) where x and y are the (isotopic) mole fractions in both phases, Ka may be either the rate of exchange or a transfer coefficient which can be considered as the 'same in both ways' if α-1 is small compared to the relative error on the measure of Ka. The third part is devoted to isotopic exchange reactors. Relationships between their efficiency and kinetics are established in some simple cases: plug cocurrent flow reactors, perfectly mixed reactors, countercurrent reactors without axial mixing. We treat only cases where α and the up flow to down flow ratio is close to 1 so that Murphee efficiency approximately overall efficiency (discrete stage contactors). HTU (phase 1) approximately HTU (phase 2) approximately HETP (columns). In a fourth part, an expression of the isotopic separative power of reactors is proposed and discussed [fr

R-102, 1 Group Space-Independent Inverse Reactor Kinetics

International Nuclear Information System (INIS)

Kaganove, J.J.

1966-01-01

1 - Description of problem or function: Given the space-independent, one energy group reactor kinetics equations and the initial conditions, this program determines the time variation of reactivity required to produce the given input of flux-time data. 2 - Method of solution: Time derivatives of neutron density are obtained by application of (a) five-point quartic, (b) three-point parabolic, (c) five-point least-mean-square cubic, (d) five-point least-mean-square parabolic, or (e) five-point least-mean-square linear formulae to the neutron density or to the natural logarithm of the neutron density. Between each data point the neutron density is assumed to be (a) exponential*(third-order polynomial), (b) exponential, or (c) linear. Changes in reactivity between data points are obtained algebraically from the kinetics equations, neutron density derivatives, and the algebraic representation of neutron density. First and second time derivatives of the reactivity are obtained by use of any of the formulae applicable to the neutron density. 3 - Restrictions on the complexity of the problem: Maxima of - 50 delay groups; 1000 data points; 99 data blocks (A data block is a sequence of input points characterized by a fixed time-interval between points, a smoothing option, and a number of repetitions of the smoothing option)

Three-dimensional effects in fracture mechanics

International Nuclear Information System (INIS)

Benitez, F.G.

1991-01-01

An overall view of the pioneering theories and works, which enlighten the three-dimensional nature of fracture mechanics during the last years is given. the main aim is not an exhaustive reviewing but the displaying of the last developments on this scientific field in a natural way. This work attempts to envisage the limits of disregarding the three-dimensional behaviour in theories, analyses and experiments. Moreover, it tries to draw attention on the scant fervour, although increasing, this three-dimensional nature of fracture has among the scientific community. Finally, a constructive discussion is presented on the use of two-dimensional solutions in the analysis of geometries which bear a three-dimensional configuration. the static two-dimensional solutions and its applications fields are reviewed. also, the static three-dimensional solutions, wherein a comparative analysis with elastoplastic and elastostatic solutions are presented. to end up, the dynamic three-dimensional solutions are compared to the asymptotic two-dimensional ones under the practical applications point of view. (author)

Study on coupling of three-dimension space time neutron kinetics model and RELAP5 and improvement of RELAP5

International Nuclear Information System (INIS)

Gui Xuewen; Cai Qi; Luo Bangqi

2007-01-01

A two-group three-dimension space-time neutron kinetics model is applied to the RELAP5 code, which replaces the point reactor kinetics model. A visual operation interface is designed to convenience interactive operation between operator and computer. The calculation results and practical applications indicate that the functions and precision of improved RELAP5 are enhanced and can be easily used. The improved RELAP5 has a good application perspective in nuclear power plant simulation. (authors)

One dimensional reactor core model

International Nuclear Information System (INIS)

Kostadinov, V.; Stritar, A.; Radovo, M.; Mavko, B.

1984-01-01

The one dimensional model of neutron dynamic in reactor core was developed. The core was divided in several axial nodes. The one group neutron diffusion equation for each node is solved. Feedback affects of fuel and water temperatures is calculated. The influence of xenon, boron and control rods is included in cross section calculations for each node. The system of equations is solved implicitly. The model is used in basic principle Training Simulator of NPP Krsko. (author)

Kinetic Monte Carlo studies of the reaction kinetics of crystal defects that diffuse one-dimensionally with occasional transverse migration

DEFF Research Database (Denmark)

Heinisch, H.L.; Trinkaus, H.; Singh, Bachu Narain

2007-01-01

The reaction kinetics of the various species of mobile defects in irradiated materials are crucially dependent on the dimensionality of their migration. Sink strengths for one-dimensionally (1D) gliding interstitial loops undergoing occasional direction changes have been described analytically...

Electron tomography, three-dimensional Fourier analysis and colour prediction of a three-dimensional amorphous biophotonic nanostructure

Science.gov (United States)

Shawkey, Matthew D.; Saranathan, Vinodkumar; Pálsdóttir, Hildur; Crum, John; Ellisman, Mark H.; Auer, Manfred; Prum, Richard O.

2009-01-01

Organismal colour can be created by selective absorption of light by pigments or light scattering by photonic nanostructures. Photonic nanostructures may vary in refractive index over one, two or three dimensions and may be periodic over large spatial scales or amorphous with short-range order. Theoretical optical analysis of three-dimensional amorphous nanostructures has been challenging because these structures are difficult to describe accurately from conventional two-dimensional electron microscopy alone. Intermediate voltage electron microscopy (IVEM) with tomographic reconstruction adds three-dimensional data by using a high-power electron beam to penetrate and image sections of material sufficiently thick to contain a significant portion of the structure. Here, we use IVEM tomography to characterize a non-iridescent, three-dimensional biophotonic nanostructure: the spongy medullary layer from eastern bluebird Sialia sialis feather barbs. Tomography and three-dimensional Fourier analysis reveal that it is an amorphous, interconnected bicontinuous matrix that is appropriately ordered at local spatial scales in all three dimensions to coherently scatter light. The predicted reflectance spectra from the three-dimensional Fourier analysis are more precise than those predicted by previous two-dimensional Fourier analysis of transmission electron microscopy sections. These results highlight the usefulness, and obstacles, of tomography in the description and analysis of three-dimensional photonic structures. PMID:19158016

Measurements of kinetic parameters by noise techniques on the MINERVE reactor

International Nuclear Information System (INIS)

Carre, J.C.; Da Costa Oliveira, J.

1975-01-01

Noise measurements were determined on ERMINE a fast thermal coupled reactor built in MINERVE. A reactor without feedback, and a reactor with an automatic control rod were both considered. The first case concerned the measurements of auto and cross power spectral density obtained with one or two neutron detectors, and the determination of: neutron lifetime; efficiency for one ion chamber; power level of the reactor; maximal speed and acceleration of the control rod for the design of an automatic reactor control actuator. The second case was concerned with measurements of the auto power spectral density in reactivity for the control rod, and the estimation of: the transfer function of the automatic pilot; the neutron lifetime; and the standard error affecting the results obtained by the oscillation method. The results proved that the pile noise theory with a point kinetic model is sufficient for application on zero power reactors. (U.K.)

Three-dimensional Monte Carlo calculation of some nuclear parameters

Science.gov (United States)

Günay, Mehtap; Şeker, Gökmen

2017-09-01

In this study, a fusion-fission hybrid reactor system was designed by using 9Cr2WVTa Ferritic steel structural material and the molten salt-heavy metal mixtures 99-95% Li20Sn80 + 1-5% RG-Pu, 99-95% Li20Sn80 + 1-5% RG-PuF4, and 99-95% Li20Sn80 + 1-5% RG-PuO2, as fluids. The fluids were used in the liquid first wall, blanket and shield zones of a fusion-fission hybrid reactor system. Beryllium (Be) zone with the width of 3 cm was used for the neutron multiplication between the liquid first wall and blanket. This study analyzes the nuclear parameters such as tritium breeding ratio (TBR), energy multiplication factor (M), heat deposition rate, fission reaction rate in liquid first wall, blanket and shield zones and investigates effects of reactor grade Pu content in the designed system on these nuclear parameters. Three-dimensional analyses were performed by using the Monte Carlo code MCNPX-2.7.0 and nuclear data library ENDF/B-VII.0.

Three-dimensional Monte Carlo calculation of some nuclear parameters

Directory of Open Access Journals (Sweden)

Günay Mehtap

2017-01-01

Full Text Available In this study, a fusion-fission hybrid reactor system was designed by using 9Cr2WVTa Ferritic steel structural material and the molten salt-heavy metal mixtures 99–95% Li20Sn80 + 1-5% RG-Pu, 99–95% Li20Sn80 + 1-5% RG-PuF4, and 99–95% Li20Sn80 + 1-5% RG-PuO2, as fluids. The fluids were used in the liquid first wall, blanket and shield zones of a fusion–fission hybrid reactor system. Beryllium (Be zone with the width of 3 cm was used for the neutron multiplication between the liquid first wall and blanket. This study analyzes the nuclear parameters such as tritium breeding ratio (TBR, energy multiplication factor (M, heat deposition rate, fission reaction rate in liquid first wall, blanket and shield zones and investigates effects of reactor grade Pu content in the designed system on these nuclear parameters. Three-dimensional analyses were performed by using the Monte Carlo code MCNPX-2.7.0 and nuclear data library ENDF/B-VII.0.

Development of 2-D/1-D fusion method for three-dimensional whole-core heterogeneous neutron transport calculations

International Nuclear Information System (INIS)

Lee, Gil Soo

2006-02-01

To describe power distribution and multiplication factor of a reactor core accurately, it is necessary to perform calculations based on neutron transport equation considering heterogeneous geometry and scattering angles. These calculations require very heavy calculations and were nearly impossible with computers of old days. From the limitation of computing power, traditional approach of reactor core design consists of heterogeneous transport calculation in fuel assembly level and whole core diffusion nodal calculation with assembly homogenized properties, resulting from fuel assembly transport calculation. This approach may be effective in computation time, but it gives less accurate results for highly heterogeneous problems. As potential for whole core heterogeneous transport calculation became more feasible owing to rapid development of computing power during last several years, the interests in two and three dimensional whole core heterogeneous transport calculations by deterministic method are increased. For two dimensional calculation, there were several successful approaches using even parity transport equation with triangular meshes, S N method with refined rectangular meshes, the method of characteristics (MOC) with unstructured meshes, and so on. The work in this thesis originally started from the two dimensional whole core heterogeneous transport calculation by using MOC. After successful achievement in two dimensional calculation, there were efforts in three-dimensional whole-core heterogeneous transport calculation using MOC. Since direct extension to three dimensional calculation of MOC requires too much computing power, indirect approach to three dimensional calculation was considered.Thus, 2D/1D fusion method for three dimensional heterogeneous transport calculation was developed and successfully implemented in a computer code. The 2D/1D fusion method is synergistic combination of the MOC for radial 2-D calculation and S N -like methods for axial 1

Secondary motion in three-dimensional branching networks

Science.gov (United States)

Guha, Abhijit; Pradhan, Kaustav

2017-06-01

A major aim of the present work is to understand and thoroughly document the generation, the three-dimensional distribution, and the evolution of the secondary motion as the fluid progresses downstream through a branched network. Six generations (G0-G5) of branches (involving 63 straight portions and 31 bifurcation modules) are computed in one go; such computational challenges are rarely taken in the literature. More than 30 × 106 computational elements are employed for high precision of computed results and fine quality of the flow visualization diagrams. The study of co-planar vis-à-vis non-planar space-filling configurations establishes a quantitative evaluation of the dependence of the fluid dynamics on the three-dimensional arrangement of the same individual branches. As compared to the secondary motion in a simple curved pipe, three distinctive features, viz., the change of shape and size of the flow-cross-section, the division of non-uniform primary flow in a bifurcation module, and repeated switchover from clockwise to anticlockwise curvature and vice versa in the flow path, make the present situation more complex. It is shown that the straight portions in the network, in general, attenuate the secondary motion, while the three-dimensionally complex bifurcation modules generate secondary motion and may alter the number, arrangement, and structure of vortices. A comprehensive picture of the evolution of quantitative flow visualizations of the secondary motion is achieved by constructing contours of secondary velocity | v → S | , streamwise vorticity ω S , and λ 2 iso-surfaces. It is demonstrated, for example, that for in-plane configuration, the vortices on any plane appear in pair (i.e., for each clockwise rotating vortex, there is an otherwise identical anticlockwise vortex), whereas the vortices on a plane for the out-of-plane configuration may be dissimilar, and there may even be an odd number of vortices. We have formulated three new parameters

Thermal decomposition kinetics of ammonium uranyl carbonate

International Nuclear Information System (INIS)

Kim, E.H.; Park, J.J.; Park, J.H.; Chang, I.S.; Choi, C.S.; Kim, S.D.

1994-01-01

The thermal decomposition kinetics of AUC [ammonium uranyl carbonate; (NH 4 ) 4 UO 2 (CO 3 ) 3 [ in an isothermal thermogravimetric (TG) reactor under N 2 atmosphere has been determined. The kinetic data can be represented by the two-dimensional nucleation and growth model. The reaction rate increases and activation energy decreases with increasing particle size and precipitation time which appears in the particle size larger than 30 μm in the mechano-chemical phenomena. (orig.)

Modeling of a three-phase reactor for bitumen-derived gas oil hydrotreating

Energy Technology Data Exchange (ETDEWEB)

Chacon, R.; Canale, A.; Bouza, A. [Departamento de Termodinamica y Fenomenos de Transporte. Universidad Simon Bolivar, Caracas (Venezuela, Bolivarian Republic of); Sanchez, Y. [Departamento de Procesos y Sistemas. Universidad Simon Bolivar (Venezuela, Bolivarian Republic of)

2012-01-15

A three-phase reactor model for describing the hydrotreating reactions of bitumen-derived gas oil was developed. The model incorporates the mass-transfer resistance at the gas-liquid and liquid-solid interfaces and a kinetic rate expression based on a Langmuir-Hinshelwood-type model. We derived three correlations for determining the solubility of hydrogen (H{sub 2}), hydrogen sulfide (H{sub 2}S) and ammonia (NH{sub 3}) in hydrocarbon mixtures and the calculation of the catalyst effectiveness factor was included. Experimental data taken from the literature were used to determine the kinetic parameters (stoichiometric coefficients, reaction orders, reaction rate and adsorption constants for hydrodesulfuration (HDS) and hydrodenitrogenation (HDN)) and to validate the model under various operating conditions. Finally, we studied the effect of operating conditions such as pressure, temperature, LHSV, H{sub 2}/feed ratio and the inhibiting effect of H{sub 2}S on HDS and NH{sub 3} on HDN. (author)

Three-dimensional biomedical imaging

International Nuclear Information System (INIS)

Robb, R.A.

1985-01-01

Scientists in biomedical imaging provide researchers, physicians, and academicians with an understanding of the fundamental theories and practical applications of three-dimensional biomedical imaging methodologies. Succinct descriptions of each imaging modality are supported by numerous diagrams and illustrations which clarify important concepts and demonstrate system performance in a variety of applications. Comparison of the different functional attributes, relative advantages and limitations, complementary capabilities, and future directions of three-dimensional biomedical imaging modalities are given. Volume 1: Introductions to Three-Dimensional Biomedical Imaging Photoelectronic-Digital Imaging for Diagnostic Radiology. X-Ray Computed Tomography - Basic Principles. X-Ray Computed Tomography - Implementation and Applications. X-Ray Computed Tomography: Advanced Systems and Applications in Biomedical Research and Diagnosis. Volume II: Single Photon Emission Computed Tomography. Position Emission Tomography (PET). Computerized Ultrasound Tomography. Fundamentals of NMR Imaging. Display of Multi-Dimensional Biomedical Image Information. Summary and Prognostications

Three-dimensional neuroimaging

International Nuclear Information System (INIS)

Toga, A.W.

1990-01-01

This book reports on new neuroimaging technologies that are revolutionizing the study of the brain be enabling investigators to visualize its structure and entire pattern of functional activity in three dimensions. The book provides a theoretical and practical explanation of the new science of creating three-dimensional computer images of the brain. The coverage includes a review of the technology and methodology of neuroimaging, the instrumentation and procedures, issues of quantification, analytic protocols, and descriptions of neuroimaging systems. Examples are given to illustrate the use of three-dimensional enuroimaging to quantitate spatial measurements, perform analysis of autoradiographic and histological studies, and study the relationship between brain structure and function

Review of Kaganove's solution for the reactor point kinetics equations

International Nuclear Information System (INIS)

Couto, R.T.; Santo, A.C.F. de.

1993-09-01

A review of Kaganove's method for the reactor point kinetics equations solution is performed. This was method chosen to calculate the power in ATR, a computer program for the analysis of reactivity transients. The reasons for this choice and the adaptation of the method to the purposes of ATR are presented. (author)

A Derivation of Source-based Kinetics Equation with Time Dependent Fission Kernel for Reactor Transient Analyses

Energy Technology Data Exchange (ETDEWEB)

Kim, Song Hyun; Woo, Myeong Hyun; Shin, Chang Ho [Hanyang University, Seoul (Korea, Republic of); Pyeon, Cheol Ho [Kyoto University, Osaka (Japan)

2015-10-15

In this study, a new balance equation to overcome the problems generated by the previous methods is proposed using source-based balance equation. And then, a simple problem is analyzed with the proposed method. In this study, a source-based balance equation with the time dependent fission kernel was derived to simplify the kinetics equation. To analyze the partial variations of reactor characteristics, two representative methods were introduced in previous studies; (1) quasi-statics method and (2) multipoint technique. The main idea of quasistatics method is to use a low-order approximation for large integration times. To realize the quasi-statics method, first, time dependent flux is separated into the shape and amplitude functions, and shape function is calculated. It is noted that the method has a good accuracy; however, it can be expensive as a calculation cost aspect because the shape function should be fully recalculated to obtain accurate results. To improve the calculation efficiency, multipoint method was proposed. The multipoint method is based on the classic kinetics equation with using Green's function to analyze the flight probability from region r' to r. Those previous methods have been used to analyze the reactor kinetics analysis; however, the previous methods can have some limitations. First, three group variables (r{sub g}, E{sub g}, t{sub g}) should be considered to solve the time dependent balance equation. This leads a big limitation to apply large system problem with good accuracy. Second, the energy group neutrons should be used to analyze reactor kinetics problems. In time dependent problem, neutron energy distribution can be changed at different time. It can affect the change of the group cross section; therefore, it can lead the accuracy problem. Third, the neutrons in a space-time region continually affect the other space-time regions; however, it is not properly considered in the previous method. Using birth history of the

Kinetics Analysis of Synthesis Reaction of Struvite With Air-Flow Continous Vertical Reactors

Science.gov (United States)

Edahwati, L.; Sutiyono, S.; Muryanto, S.; Jamari, J.; Bayuseno, dan A. P.

2018-01-01

Kinetics reaction is a knowledge about a rate of chemical reaction. The differential of the reaction rate can be determined from the reactant material or the formed material. The reaction mechanism of a reactor may include a stage of reaction occurring sequentially during the process of converting the reactants into products. In the determination of reaction kinetics, the order of reaction and the rate constant reaction must be recognized. This study was carried out using air as a stirrer as a medium in the vertical reactor for crystallization of struvite. Stirring is one of the important aspects in struvite crystallization process. Struvite crystals or magnesium ammonium phosphate hexahydrates (MgNH4PO4·6H2O) is commonly formed in reversible reactions and can be generated as an orthorhombic crystal. Air is selected as a stirrer on the existing flow pattern in the reactor determining the reaction kinetics of the crystal from the solution. The experimental study was conducted by mixing an equimolar solution of 0.03 M NH4OH, MgCl2 and H3PO4 with a ratio of 1: 1: 1. The crystallization process of the mixed solution was observed in an inside reactor at the flow rate ranges of 16-38 ml/min and the temperature of 30°C was selected in the study. The air inlet rate was kept constant at 0.25 liters/min. The pH solution was adjusted to be 8, 9 and 10 by dropping wisely of 1 N KOH solution. The crystallization kinetics was examined until the steady state of the reaction was reached. The precipitates were filtered and dried at a temperature for subsequent material characterization, including Scanning Electron Microscope (SEM) and XRD (X-Ray diffraction) method. The results show that higher flow rate leads to less mass of struvite.

Annual progress report for 1982 of Theoretical Reactor Physics Section

International Nuclear Information System (INIS)

Rastogi, B.P.; Kumar, Vinod

1983-01-01

The progress of work done in the Theoretical Reactor Physics Section of the Bhabha Atomic Research Centre, Bombay, during the calendar year 1982 is reported in the form of write-ups and summaries. The main thrust of the work has been to master the neutronic design technology of four different types of nuclear reactor types, namely, pressurized heavy water reactors, boiling light water reactors, pressurized light water reactors and fast breeder reactors. The development work for the neutronic analysis, fuel design, and fuel management of the BWR type reactors of the Tarapur Atomic Power Station has been completed. A new reactor simulator system for PHWR design analysis and core follow-up was completed. Three dimensional static analysis codes based on nodal and finite element methods for the design work of larger size (500-750 MWe) reactors have been developed. Space link kinetics codes in one, two and three dimensions for above-mentioned reactor systems have been written and validated. Fast reactor core disruptive analysis codes have been developed. In the course of R and D work concerning various types of reactor projects, investigations were also carried in the allied areas of Monte Carlo techniques, integral transform methods, path integral methods, high spin states in heavy nuclei and a hydrodynamics model for a laser driven fusion system. (M.G.B.)

Study of structural attachments of a pool type LMFBR vessel through seismic analysis of a simplified three dimensional finite element model

International Nuclear Information System (INIS)

Ahmed, H.; Ma, D.

1979-01-01

A simplified three dimensional finite element model of a pool type LMFBR in conjunction with the computer program ANSYS is developed and scoping results of seismic analysis are produced. Through this study various structural attachments of a pool type LMFBR like the reactor vessel skirt support, the pump support and reactor shell-support structure interfaces are studied. This study also provides some useful results on equivalent viscous damping approach and some improvements to the treatment of equivalent viscous damping are recommended. This study also sets forth pertinent guidelines for detailed three dimensional finite element seismic analysis of pool type LMFBR

Three-dimensional velocity map imaging: Setup and resolution improvement compared to three-dimensional ion imaging

International Nuclear Information System (INIS)

Kauczok, S.; Goedecke, N.; Veckenstedt, M.; Maul, C.; Gericke, K.-H.; Chichinin, A. I.

2009-01-01

For many years the three-dimensional (3D) ion imaging technique has not benefited from the introduction of ion optics into the field of imaging in molecular dynamics. Thus, a lower resolution of kinetic energy as in comparable techniques making use of inhomogeneous electric fields was inevitable. This was basically due to the fact that a homogeneous electric field was needed in order to obtain the velocity component in the direction of the time of flight spectrometer axis. In our approach we superimpose an Einzel lens field with the homogeneous field. We use a simulation based technique to account for the distortion of the ion cloud caused by the inhomogeneous field. In order to demonstrate the gain in kinetic energy resolution compared to conventional 3D Ion Imaging, we use the spatial distribution of H + ions emerging from the photodissociation of HCl following the two photon excitation to the V 1 Σ + state. So far a figure of merit of approximately four has been achieved, which means in absolute numbers Δv/v=0.022 compared to 0.086 at v≅17 000 m/s. However, this is not a theoretical limit of the technique, but due to our rather short TOF spectrometer (15 cm). The photodissociation of HBr near 243 nm has been used to recognize and eliminate systematic deviations between the simulation and the experimentally observed distribution. The technique has also proven to be essential for the precise measurement of translationally cold distributions.

A highly accurate benchmark for reactor point kinetics with feedback

International Nuclear Information System (INIS)

Ganapol, B. D.; Picca, P.

2010-10-01

This work apply the concept of convergence acceleration, also known as extrapolation, to find the solution to the reactor kinetics equations describing nuclear reactor transients. The method features simplicity in that an approximate finite difference formulation is constructed and converged to high accuracy from knowledge of how the error term behaves. Through Rom berg extrapolation, we demonstrate its high accuracy for a variety of imposed reactivity insertions found in the literature as well as nonlinear temperature and fission product feedback. A unique feature of the proposed method, called RKE/R(om berg) algorithm, is interval bisection to ensure high accuracy. (Author)

Rod drop in the LR-0 reactor core comprising 55 fuel assemblies

International Nuclear Information System (INIS)

Hadek, J.; Grundmann, U.

1989-09-01

Data from the third stage of kinetic measurements on the LR-0 reactor, performed in 1988, were employed for additional calculations using the 3-dimensional neutron kinetics code HEXDYN3D. The reactor consists of subassemblies similar to those in the WWER-1000 (PWR) reactor. The theoretical and experimental results are compared for the time behavior of the neutron flux caused by drop of the control rod cluster in various subassemblies of the reactor. The results demonstrate that the HEXDYN3D code is well suited to the treatment of the space-time behavior of the neutron flux. (author). 21 figs., 2 tabs., 16 refs

Track 3: growth of nuclear technology and research numerical and computational aspects of the coupled three-dimensional core/plant simulations: organization for economic cooperation and development/U.S. nuclear regulatory commission pressurized water reactor main-steam-line-break benchmark-I. 5. Analyses of the OECD MSLB Benchmark with the Codes DYN3D and DYN3D/ATHLET

International Nuclear Information System (INIS)

Grundmann, U.; Kliem, S.

2001-01-01

The code DYN3D coupled with ATHLET was used for the analysis of the OECD Main-Steam-Line-Break (MSLB) Benchmark, which is based on real plant design and operational data of the TMI-1 pressurized water reactor (PWR). Like the codes RELAP or TRAC,ATHLET is a thermal-hydraulic system code with point or one-dimensional neutron kinetic models. ATHLET, developed by the Gesellschaft for Anlagen- und Reaktorsicherheit, is widely used in Germany for safety analyses of nuclear power plants. DYN3D consists of three-dimensional nodal kinetic models and a thermal-hydraulic part with parallel coolant channels of the reactor core. DYN3D was coupled with ATHLET for analyzing more complex transients with interactions between coolant flow conditions and core behavior. It can be applied to the whole spectrum of operational transients and accidents, from small and intermediate leaks to large breaks of coolant loops or steam lines at PWRs and boiling water reactors. The so-called external coupling is used for the benchmark, where the thermal hydraulics is split into two parts: DYN3D describes the thermal hydraulics of the core, while ATHLET models the coolant system. Three exercises of the benchmark were simulated: Exercise 1: point kinetics plant simulation (ATHLET) Exercise 2: coupled three-dimensional neutronics/core thermal-hydraulics evaluation of the core response for given core thermal-hydraulic boundary conditions (DYN3D) Exercise 3: best-estimate coupled core-plant transient analysis (DYN3D/ATHLET). Considering the best-estimate cases (scenarios 1 of exercises 2 and 3), the reactor does not reach criticality after the reactor trip. Defining more serious tests for the codes, the efficiency of the control rods was decreased (scenarios 2 of exercises 2 and 3) to obtain recriticality during the transient. Besides the standard simulation given by the specification, modifications are introduced for sensitivity studies. The results presented here show (a) the influence of a reduced

Neutron flux shape effects in large fast reactor safety calculations

International Nuclear Information System (INIS)

Galati, A.; Loizzo, P.; Musco, A.

1978-01-01

Three classes of accidents in a large fast reactor were studied by the two-dimensional core dynamics code NADYP-2. A Modified version of the code, including a point kinetics module, allowed comparison between 2D and 0D power, reactivity and temperature histories. A strong shape effect was evidenced by these calculations in the boiling phase of LOF accidents as well as in the accident generated by control rod removal. Some future possibilities of by passing the consequences of this effect are indicated

Three dimensional canonical transformations

International Nuclear Information System (INIS)

Tegmen, A.

2010-01-01

A generic construction of canonical transformations is given in three-dimensional phase spaces on which Nambu bracket is imposed. First, the canonical transformations are defined as based on cannonade transformations. Second, it is shown that determination of the generating functions and the transformation itself for given generating function is possible by solving correspondent Pfaffian differential equations. Generating functions of type are introduced and all of them are listed. Infinitesimal canonical transformations are also discussed as the complementary subject. Finally, it is shown that decomposition of canonical transformations is also possible in three-dimensional phase spaces as in the usual two-dimensional ones.

Phenolic Wastewater Treatment using Activated Carbon in a Three Phase Fluidized-Bed Reactor

Directory of Open Access Journals (Sweden)

Pornsiri Tongprem

2009-11-01

Full Text Available Phenolic wastewater treatment was investigated using activated carbon in a lab scale three phase fluidized-bed reactor. The reactor with effective volume of 272 ml, 300 mm in height and 40 mm in diameter was made from transparent acrylic that allowed to observe the phenomena occurring inside. Phenol 10 mg/l and air were used as representative agents that were continuously fed to the reactor at a constant flow rate of 1 and 2 l/min with co-current and up-flow, respectively. Comparison of the phenolic adsorption under five different conditions: (a fresh Acs, (b 1st reused Acs, (c fresh Fe/Acs, (d 1st reused Fe/Acs, and (e 2nd reused Fe/Acs, have been carried out. The phenolic wastewater was re-circulated through the reactor and its concentration was measured with respect to time. The experimental adsorption results revealed that both fresh Acs and Fe/Acs gave the better results than reused Acs and reused Fe/Acs, respectively. The adsorption in all cases of Acs and Fe/Acs would follow Pseudo-second order kinetic.

Summary of the LLNL one-dimensional transport-kinetics model of the troposphere and stratosphere: 1981

International Nuclear Information System (INIS)

Wuebbles, D.J.

1981-09-01

Since the LLNL one-dimensional coupled transport and chemical kinetics model of the troposphere and stratosphere was originally developed in 1972 (Chang et al., 1974), there have been many changes to the model's representation of atmospheric physical and chemical processes. A brief description is given of the current LLNL one-dimensional coupled transport and chemical kinetics model of the troposphere and stratosphere

Comparison of three ICF reactor designs

International Nuclear Information System (INIS)

Hogan, W.J.

1984-01-01

Three concepts for inertial confinement fusion (ICF) reactors are described and compared with each other, and with magnetic fusion and fission reactors on the basis of environmental impact, safety and efficiency. The critical technical developments of each concept are described. The three concepts represent alternative development paths for inertial fusion

HMS-burn: a model for hydrogen distribution and combustion in nuclear reactor containments

International Nuclear Information System (INIS)

Travis, J.R.

1985-01-01

It is now possible to analyze the time-dependent, fully three-dimensional behavior of hydrogen combustion in nuclear reactor containments. This analysis involves coupling the full Navier-Stokes equations with multi-species transport to the global chemical kinetics of hydrogen combustion. A transport equation for the subgrid scale turbulent kinetic energy density is solved to produce the time and space dependent turbulent transport coefficients. The heat transfer coefficient governing the exchange of heat between fluid computational cells adjacent to wall cells is calculated by a modified Reynolds analogy formulation. The analysis of a MARK-III containment indicates very complex flow patterns that greatly influence fluid and wall temperatures and heat fluxes

The influence of pH adjustment on kinetics parameters in tapioca wastewater treatment using aerobic sequencing batch reactor system

Science.gov (United States)

Mulyani, Happy; Budianto, Gregorius Prima Indra; Margono, Kaavessina, Mujtahid

2018-02-01

The present investigation deals with the aerobic sequencing batch reactor system of tapioca wastewater treatment with varying pH influent conditions. This project was carried out to evaluate the effect of pH on kinetics parameters of system. It was done by operating aerobic sequencing batch reactor system during 8 hours in many tapioca wastewater conditions (pH 4.91, pH 7, pH 8). The Chemical Oxygen Demand (COD) and Mixed Liquor Volatile Suspended Solids (MLVSS) of the aerobic sequencing batch reactor system effluent at steady state condition were determined at interval time of two hours to generate data for substrate inhibition kinetics parameters. Values of the kinetics constants were determined using Monod and Andrews models. There was no inhibition constant (Ki) detected in all process variation of aerobic sequencing batch reactor system for tapioca wastewater treatment in this study. Furthermore, pH 8 was selected as the preferred aerobic sequencing batch reactor system condition in those ranging pH investigated due to its achievement of values of kinetics parameters such µmax = 0.010457/hour and Ks = 255.0664 mg/L COD.

VENUS-2, Reactor Kinetics with Feedback, 2-D LMFBR Disassembly Excursions

International Nuclear Information System (INIS)

Jackson, J.F.; Nicholson, R.B.; Weber, D.P.

1980-01-01

1 - Description of problem or function: VENUS-2 is an improved edition of the VENUS fast-reactor disassembly program. It is a two- dimensional (r-z) coupled neutronics-hydrodynamics code that calculates the dynamic behavior of an LMFBR during a prompt-critical disassembly excursion. It calculates the power history and fission energy release as well as the space-time histories of the fuel temperatures, core material pressures, and core material motions. Reactivity feedback effects due to Doppler broadening and reactor material motion are taken into account. 2 - Method of solution: The power and energy release are calculated using a point-kinetics formulation with up to six delayed neutron groups. The reactivity is a combination of an input driving function and feedback effects due to Doppler broadening and material motion. An adiabatic model is used to calculate the temperature increase throughout the reactor based on an initial temperature distribution and power profile provided as input data. These temperatures are, in turn, converted to fuel pressures through one of several equation of state options provided. The material motion that results from the pressure buildup is calculated by a direct finite difference solution of a set of two-dimensional (r-z) hydrodynamics equations. This is done in Lagrangian coordinates. The reactivity change associated with this motion is calculated by first-order perturbation theory. The displacements are also used to adjust the fuel densities as required for the density dependent equation-of- state option. An automatic time-step-size selection scheme is provided. 3 - Restrictions on the complexity of the problem: VENUS-2 is written so that the dimensions of the storage arrays can be readily changed to accommodate a broad range of problem sizes. In the base version, the total number of mesh intervals is restricted such that (NR+3)*(NZ+3) is less than 700, where NR and NZ are the total number of mesh intervals in the r and z

Application of preconditioned conjugate gradient-like methods to reactor kinetics

International Nuclear Information System (INIS)

Yang, D.Y.; Chen, G.S.; Chou, H.P.

1993-01-01

Several conjugate gradient-like (CG-like) methods are applied to solve the nonsymmetric linear systems of equations derived from the time-dependent two-dimensional two-energy-group neutron diffusion equations by a finite difference approximation. The methods are: the generalized conjugate residual method; the generalized conjugate gradient least-square method; the generalized minimal residual method (GMRES); the conjugate gradient square method; and a variant of bi-conjugate gradient method (Bi-CGSTAB). In order to accelerate these methods, six preconditioning techniques are investigated. Two are based on pointwise incomplete factorization: the incomplete LU (ILU) and the modified incomplete LU (MILU) decompositions; two, based on the block tridiagonal structure of the coefficient matrix, are blockwise and modified blockwise incomplete factorizations, BILU and MBILU; two are the alternating-direction implicit and symmetric successive overrelaxation (SSOR) preconditioners, derived from the basic iterative schemes. Comparisons are made by using CG-like methods combined with different preconditioners to solve a sequence of time-step reactor transient problems. Numerical tests indicate that preconditioned BI-CGSTAB with the preconditioner MBILU requires less CPU time and fewer iterations than other methods. The preconditioned CG-like methods are less sensitive to the time-step size used than the Chebyshev semi-iteration method and line SOR method. The indication is that the CGS, Bi-CGSTAB and GMRES methods are, on average, better than the other methods in reactor kinetics computation and that a good preconditioner is more important than the choice of CG-like methods. The MILU decomposition based on the conventional row-sum criterion has difficulty yielding a convergent solution and an improved version is introduced. (author)

An efficient technique for the point reactor kinetics equations with Newtonian temperature feedback effects

International Nuclear Information System (INIS)

Nahla, Abdallah A.

2011-01-01

Highlights: → An efficient technique for the nonlinear reactor kinetics equations is presented. → This method is based on Backward Euler or Crank Nicholson and fundamental matrix. → Stability of efficient technique is defined and discussed. → This method is applied to point kinetics equations of six-groups of delayed neutrons. → Step, ramp, sinusoidal and temperature feedback reactivities are discussed. - Abstract: The point reactor kinetics equations of multi-group of delayed neutrons in the presence Newtonian temperature feedback effects are a system of stiff nonlinear ordinary differential equations which have not any exact analytical solution. The efficient technique for this nonlinear system is based on changing this nonlinear system to a linear system by the predicted value of reactivity and solving this linear system using the fundamental matrix of the homogenous linear differential equations. The nonlinear point reactor kinetics equations are rewritten in the matrix form. The solution of this matrix form is introduced. This solution contains the exponential function of a variable coefficient matrix. This coefficient matrix contains the unknown variable, reactivity. The predicted values of reactivity in the explicit form are determined replacing the exponential function of the coefficient matrix by two kinds, Backward Euler and Crank Nicholson, of the rational approximations. The nonlinear point kinetics equations changed to a linear system of the homogenous differential equations. The fundamental matrix of this linear system is calculated using the eigenvalues and the corresponding eigenvectors of the coefficient matrix. Stability of the efficient technique is defined and discussed. The efficient technique is applied to the point kinetics equations of six-groups of delayed neutrons with step, ramp, sinusoidal and the temperature feedback reactivities. The results of these efficient techniques are compared with the traditional methods.

Validation of the reactor dynamics code HEXTRAN

International Nuclear Information System (INIS)

Kyrki-Rajamaeki, R.

1994-05-01

HEXTRAN is a new three-dimensional, hexagonal reactor dynamics code developed in the Technical Research Centre of Finland (VTT) for VVER type reactors. This report describes the validation work of HEXTRAN. The work has been made with the financing of the Finnish Centre for Radiation and Nuclear Safety (STUK). HEXTRAN is particularly intended for calculation of such accidents, in which radially asymmetric phenomena are included and both good neutron dynamics and two-phase thermal hydraulics are important. HEXTRAN is based on already validated codes. The models of these codes have been shown to function correctly also within the HEXTRAN code. The main new model of HEXTRAN, the spatial neutron kinetics model has been successfully validated against LR-0 test reactor and Loviisa plant measurements. Connected with SMABRE, HEXTRAN can be reliably used for calculation of transients including effects of the whole cooling system of VVERs. Further validation plans are also introduced in the report. (orig.). (23 refs., 16 figs., 2 tabs.)

Three-dimensional microbubble streaming flows

Science.gov (United States)

Rallabandi, Bhargav; Marin, Alvaro; Rossi, Massimiliano; Kaehler, Christian; Hilgenfeldt, Sascha

2014-11-01

Streaming due to acoustically excited bubbles has been used successfully for applications such as size-sorting, trapping and focusing of particles, as well as fluid mixing. Many of these applications involve the precise control of particle trajectories, typically achieved using cylindrical bubbles, which establish planar flows. Using astigmatic particle tracking velocimetry (APTV), we show that, while this two-dimensional picture is a useful description of the flow over short times, a systematic three-dimensional flow structure is evident over long time scales. We demonstrate that this long-time three-dimensional fluid motion can be understood through asymptotic theory, superimposing secondary axial flows (induced by boundary conditions at the device walls) onto the two-dimensional description. This leads to a general framework that describes three-dimensional flows in confined microstreaming systems, guiding the design of applications that profit from minimizing or maximizing these effects.

Neutronics methods for transient and safety analysis of fast reactors

Energy Technology Data Exchange (ETDEWEB)

Marchetti, Marco

2017-07-01

neutron flux. Recently, there has been a tendency to abandon twodimensional reactor models in favor of full three-dimensional ones. Only the latter models are adequate to represent phenomena that are inherently three-dimensional. Because of the computational costs attached, three-dimensional accident simulation would become practical only if all the different modules applied for the safety analysis, and therefore the neutronics module as well, are sufficiently fast. In this work, fast neutron solvers are found to reduce the total computational time even by 60%, paving thus the way to three-dimensional modeling, and to a more accurate description of the neutron field and its influence on the results of safety analyses.

Studies on the molten salt reactor. Code development and neutronics analysis of MSRE-type design

International Nuclear Information System (INIS)

Zhuang Kun; Cao Liangzhi; Zheng Youqi; Wu Hongchun

2015-01-01

The molten salt reactor is characterized by its use of the fluid-fuel, which serves both as a fuel and as a coolant simultaneously. The position of delayed neutron precursors continuously changes both in the core and in the external loop due to the fuel circulation, and the fission products are extracted by an online fuel reprocessing unit, which all lead to the modeling methods for the conventional reactors using solid fuel not applicable. This study establishes suitable calculation models for the neutronics analysis of the molten salt reactor and develops a new code named MOREL based on the three-dimensional diffusion steady and transient calculations. Some numerical tests are chosen to verify the code and the numerical results indicate that MOREL can be used for the analysis of the molten salt reactor. After verification, it is applied to analyze the characteristics of a typical molten salt reactor, including the steady characteristics, the influence of fuel circulation on the kinetic behaviors. Besides, the influence of online fuel reprocessing simulation is also examined. The results show that inherent safety is the character of the molten salt reactor from the aspect of reactivity feedback and the fuel circulation has great influence on the kinetic characteristics of molten salt reactor. (author)

A method for three-dimensional structural analysis of reinforced concrete containment

International Nuclear Information System (INIS)

Kulak, R.F.; Fiala, C.

1989-01-01

A finite element method designed to assist reactor safety analysts in the three-dimensional numerical simulation of reinforced concrete containments to normal and off-normal mechanical loadings is presented. The development of a lined reinforced concrete plate element is described in detail, and the implementation of an empirical transverse shear failure criteria is discussed. The method is applied to the analysis of a 1/6th scale reinforced concrete containment model subjected to static internal pressurization. 11 refs., 14 figs., 1 tab

Development of a computer code for Dalat research reactor transient analysis

International Nuclear Information System (INIS)

Le Vinh Vinh; Nguyen Thai Sinh; Huynh Ton Nghiem; Luong Ba Vien; Pham Van Lam; Nguyen Kien Cuong

2003-01-01

DRSIM (Dalat Reactor SIMulation) computer code has been developed for Dalat reactor transient analysis. It is basically a coupled neutronics-hydrodynamics-heat transfer code employing point kinetics, one dimensional hydrodynamics and one dimensional heat transfer. The work was financed by VAEC and DNRI in the framework of institutional R and D programme. Some transient problems related to reactivity and loss of coolant flow was carried out by DRSIM using temperature and void coefficients calculated by WIMS and HEXNOD2D codes. (author)

Dimensional control and check of field machining parts for reactor internals installation

International Nuclear Information System (INIS)

Zhang Caifang

2010-01-01

Some key issues of dimensional control for reactor internals installation are analyzed, and important technical requirements of crucial quality control elements on the measurement, machining, and checking of reactor internals filed machining parts are discussed. Moreover, provisions on quality control and risk prevention of reactor internals filed machining parts are presented in this paper. (author)

Kinetic studies on a repetitively pulsed fast reactor

International Nuclear Information System (INIS)

Das, S.

1982-01-01

Neutronic analysis of an earlier proposed periodically pulsed fast reactor at Kalpakkam (KPFR) has been carried out numerically under equilibrium and transient conditions using the one-point model of reactor kinetics and the experimentally measured total worth of reactivity modulator, the parabolic coefficient of reactivity of the movable reflector and the mean prompt neutron lifetime. Results of steady-state calculations - treated on the basis of delayed neutron precursor and energy balances during a period of operation - have been compared with the analytical formulae of Larrimore for a parabolic reactivity input. Empirical relations for half-width of the fast neutron pulse, the peak pulse power and the power at first crossing of prompt criticality have been obtained and shown to be accurate enough for predicting steady-state power pulse characteristics of a periodically pulsed fast reactor. The concept of a subprompt-critical reactor has been used to calculate the fictitious delayed neutron fraction, β of the KPFR through a numerical experiment. Relative pulse height stability and pulse shape sensitivity to changes of maximum reactivity is discussed. With the aid of new safety concepts, the Power Amplification Factor (PAF) and the Pulse Growth Factor (Rsub(p)), the dynamics KPFR under accidental conditions has been studied for step and ramp reactivity perturbations. All the analysis has been done without taking account of reactivity feedback. (orig.)

Detailed flow analysis for the Three Mile Island unit 2 reactor accident

International Nuclear Information System (INIS)

Lillington, J.N.; Lyons, A.J.

1990-01-01

Some particular characteristics of the steam flow in the accident at the Three Mile Island unit 2 pressurized water reactor are investigated using the AEA Technology Flow3D code. Natural circulation flows with heat removal from the core and deposition in the upper plenum are predicted during the primary heating phase. The structure of the upper plenum cylinder and core blockage, owing to material relocation, are shown to force the flow into a complex three-dimensional pattern. The flows and temperature distributions from the calculations are shown to be consistent with the observed damage pattern above the core. Despite high core temperatures, damage was limited by the operation of one of the pumps at the end of the initial heating phase. Flow3D calculations are also carried out to demonstrate that the three-dimensional buoyancy driven flows are completely destroyed by the high steam generation rates arising from the pump operation. (author)

Evolution of three-dimensional relativistic current sheets and development of self-generated turbulence

Science.gov (United States)

Takamoto, M.

2018-05-01

In this paper, the temporal evolution of three-dimensional relativistic current sheets in Poynting-dominated plasma is studied for the first time. Over the past few decades, a lot of efforts have been conducted on studying the evolution of current sheets in two-dimensional space, and concluded that sufficiently long current sheets always evolve into the so-called plasmoid chain, which provides a fast reconnection rate independent of its resistivity. However, it is suspected that plasmoid chain can exist only in the case of two-dimensional approximation, and would show transition to turbulence in three-dimensional space. We performed three-dimensional numerical simulation of relativistic current sheet using resistive relativistic magnetohydrodynamic approximation. The results showed that the three-dimensional current sheets evolve not into plasmoid chain but turbulence. The resulting reconnection rate is 0.004, which is much smaller than that of plasmoid chain. The energy conversion from magnetic field to kinetic energy of turbulence is just 0.01 per cent, which is much smaller than typical non-relativistic cases. Using the energy principle, we also showed that the plasmoid is always unstable for a displacement in the opposite direction to its acceleration, probably interchange-type instability, and this always results in seeds of turbulence behind the plasmoids. Finally, the temperature distribution along the sheet is discussed, and it is found that the sheet is less active than plasmoid chain. Our finding can be applied for many high-energy astrophysical phenomena, and can provide a basic model of the general current sheet in Poynting-dominated plasma.

A central European training course on reactor physics and kinetics - the 'Eugene Wigner Course' - Organisers view

International Nuclear Information System (INIS)

Boeck, H.; Villa, M.; Matejka, K.; Sklenka, L.; Miglierini, M.; Sukods, C.

2004-01-01

Initiated by the 5th Framework Program of the European Commission, the European Nuclear Engineering Network (ENEN) is preparing the future European Nuclear Education schemes, degrees and requirements. To fully utilize the benefits of international cooperation and to promote the knowledge of students in nuclear engineering a 2.5 weeks course has been held, both in spring 2003 and 2004. The main emphasis of the course is to perform reactor physics and kinetics experiments on three different research- and training reactors in three different locations (Vienna, Prague, Budapest). The experimental work is preceded by theoretical lectures aiming to prepare the students for the experiments (Bratislava). The students' work will be evaluated, and upon success the students will get a certificate. The finally accepted credit (ECTS) value will be determined by the students' home university. The ENEN-recommended value is between 6 and 8 ECTS. The more detailed description of the course will be given in the full paper. (author)

Real-time simulation of response to load variation for a ship reactor based on point-reactor double regions and lumped parameter model

Energy Technology Data Exchange (ETDEWEB)

Wang Qiao; Zhang De [Department of Nuclear Energy Science and Engineering, Naval University of Engineering, Wuhan 430033 (China); Chen Wenzhen, E-mail: Cwz2@21cn.com [Department of Nuclear Energy Science and Engineering, Naval University of Engineering, Wuhan 430033 (China); Chen Zhiyun [Department of Nuclear Energy Science and Engineering, Naval University of Engineering, Wuhan 430033 (China)

2011-05-15

Research highlights: > We calculate the variation of main parameters of the reactor core by the Simulink. > The Simulink calculation software (SCS) can deal well with the stiff problem. > The high calculation precision is reached with less time, and the results can be easily displayed. > The quick calculation of ship reactor transient can be achieved by this method. - Abstract: Based on the point-reactor double regions and lumped parameter model, while the nuclear power plant second loop load is increased or decreased quickly, the Simulink calculation software (SCS) is adopted to calculate the variation of main physical and thermal-hydraulic parameters of the reactor core. The calculation results are compared with those of three-dimensional simulation program. It is indicated that the SCS can deal well with the stiff problem of the point-reactor kinetics equations and the coupled problem of neutronics and thermal-hydraulics. The high calculation precision can be reached with less time, and the quick calculation of parameters of response to load disturbance for the ship reactor can be achieved. The clear image of the calculation results can also be displayed quickly by the SCS, which is very significant and important to guarantee the reactor safety operation.

Improving fuel quality by whole crude oil hydrotreating: A kinetic model for hydrodeasphaltenization in a trickle bed reactor

International Nuclear Information System (INIS)

Jarullah, A.T.; Mujtaba, I.M.; Wood, A.S.

2012-01-01

Highlights: ► Asphaltene contaminant must be removed to a large extent from the fuel to meet the regulatory demand. ► Kinetics for hydrodeasphaltenization are estimated via experimentation and modeling. ► Using the kinetic parameters, a full process model for the trickle bed reactor (TBR) is developed. ► The model is used for simulating the behavior of the TBR to get further insight of the process. ► The influences of operating conditions in the hydrodeasphaltenization process are reported. -- Abstract: Fossil fuel is still a predominant source of the global energy requirement. Hydrotreating of whole crude oil has the ability to increase the productivity of middle distillate fractions and improve the fuel quality by simultaneously reducing contaminants such as sulfur, nitrogen, vanadium, nickel and asphaltene to the levels required by the regulatory bodies. Hydrotreating is usually carried out in a trickle bed reactor (TBR) where hydrodesulfurization (HDS), hydrodenitrogenation (HDN), hydrodemetallization (HDM) and hydrodeasphaltenization (HDAs) reactions take place simultaneously. To develop a detailed and a validated TBR process model which can be used for design and optimization of the hydrotreating process, it is essential to develop kinetic models for each of these reactions. Most recently, the authors have developed kinetic models for all of these chemical reactions except that of HDAs. In this work, a kinetic model (in terms of kinetic parameters) for the HDAs reaction in the TBR is developed. A three phase TBR process model incorporating the HDAs reactions with unknown kinetic parameters is developed. Also, a series of experiments has been conducted in an isothermal TBR under different operating conditions affecting the removal of asphaltene. The unknown kinetic parameters are then obtained by applying a parameter estimation technique based on minimization of the sum of square errors (SSEs) between the experimental and predicted concentrations of

Application of the fractional neutron point kinetic equation: Start-up of a nuclear reactor

International Nuclear Information System (INIS)

Polo-Labarrios, M.-A.; Espinosa-Paredes, G.

2012-01-01

Highlights: ► Neutron density behavior at reactor start up with fractional neutron point kinetics. ► There is a relaxation time associated with a rapid variation in the neutron flux. ► Physical interpretation of the fractional order is related with non-Fickian effects. ► Effect of the anomalous diffusion coefficient and the relaxation time is analyzed. ► Neutron density is related with speed and duration of the control rods lifting. - Abstract: In this paper we present the behavior of the variation of neutron density when the nuclear reactor power is increased using the fractional neutron point kinetic (FNPK) equation with a single-group of delayed neutron precursor. It is considered that there is a relaxation time associated with a rapid variation in the neutron flux and its physical interpretation of the fractional order is related with non-Fickian effects from the neutron diffusion equation point of view. We analyzed the case of increase the nuclear reactor power when reactor is cold start-up which is a process of inserting reactivity by lifting control rods discontinuously. The results show that for short time scales of the start-up the neutronic density behavior with FNPK shows sub-diffusive effects whose absorption are government by control rods velocity. For large times scale, the results shows that the classical equation of the neutron point kinetics over predicted the neutron density regarding to FNPK.

Influence of dissolved oxygen on the nitrification kinetics in a circulating bed biofilm reactor

Energy Technology Data Exchange (ETDEWEB)

Nogueira, R.; Melo, L.F. [University of Minho, Braga (Portugal). Dept. Bioengineering; Lazarova, V.; Manem, J. [Centre of International Research for Water and Environment (CIRSEE), Lyonnaise des Eaux, Le Pecq (France)

1998-12-01

The influence of dissolved oxygen concentration on the nitrification kinetics was studied in the circulating bed reactor (CBR). The study was partly performed at laboratory scale with synthetic water, and partly at pilot scale with secondary effluent as feed water. The nitrification kinetics of the laboratory CBR as a function of the oxygen concentration can be described according to the half order and zero order rate equations of the diffusion-reaction model applied to porous catalysts. When oxygen was the rate limiting substrate, the nitrification rate was close to a half order function of the oxygen concentration. The average oxygen diffusion coefficient estimated by fitting the diffusion-reaction model to the experimental results was around 66% of the respective value in water. The experimental results showed that either the ammonia or the oxygen concentration could be limiting for the nitrification kinetics. The latter occurred for an oxygen to ammonia concentration ratio below 1.5-2 gO{sub 2}/gN-NH{sub 4}{sup +} for both laboratory and pilot scale reactors. The volumetric oxygen mass transfer coefficient (k{sub L}a) determined in the laboratory scale reactor was 0.017 s{sup -1} for a superficial air velocity of 0.02 m s{sup -1}, and the one determined in the pilot scale reactor was 0.040 s{sup -1} for a superficial air velocity of 0.031 m s{sup -1}. The k{sub L}a for the pilot scale reactor did not change significantly after biofilm development, compared to the value measured without biofilm. (orig.) With 7 figs., 5 tabs., 24 refs.

Tag gas burnup based on three-dimensional FTR analysis

International Nuclear Information System (INIS)

Kidman, R.B.

1976-01-01

Flux spectra from a three-dimensional diffusion theory analysis of the Fast Test Reactor (FTR) are used to predict gas tag ratio changes, as a function of exposure, for each FTR fuel and absorber subassembly plenum. These flux spectra are also used to predict Xe-125 equilibrium activities in absorber plena in order to assess the feasibility of using Xe-125 gamma rays to detect and distinguish control rod failures from fuel rod failures. Worst case tag burnup changes are used in conjunction with burnup and mass spectrometer uncertainties to establish the minimum spacing of tags which allows the tags to be unambiguously identified

Calculation of three-dimensional fluid flow with multiple free surfaces

International Nuclear Information System (INIS)

Vander Vorst, M.J.; Chan, R.K.C.

1978-01-01

This paper presents a method for computing incompressible fluid flows with multiple free surfaces which are not restricted in their orientation. The method is presented in the context of the three-dimensional flow in a Mark I reactor pressure suppression system immediately following a postulated loss of coolant accident. The assumption of potential flow is made. The numerical method is a mixed Eulerian-Lagrangian formulation with the interior treated as Eulerian and the free surfaces as Lagrangian. The accuracy of solution hinges on the careful treatment of two important aspects. First, the Laplace equation for the potential is solved at interior points of the Eulerian finite difference mesh using a three-dimensional ''irregular star'' so that boundary conditions can be imposed at the exact position of the free surface. Second, the Lagrangian free surfaces are composed of triangular elements, upon each vertex of which is applied the fully nonlinear Bernoulli equation. One result of these calculations is the transient load on the suppression vessel during the vent clearing and bubble formation events of a loss of coolant accident

An atmospheric pressure flow reactor: Gas phase kinetics and mechanism in tropospheric conditions without wall effects

Science.gov (United States)

Koontz, Steven L.; Davis, Dennis D.; Hansen, Merrill

1988-01-01

A new type of gas phase flow reactor, designed to permit the study of gas phase reactions near 1 atm of pressure, is described. A general solution to the flow/diffusion/reaction equations describing reactor performance under pseudo-first-order kinetic conditions is presented along with a discussion of critical reactor parameters and reactor limitations. The results of numerical simulations of the reactions of ozone with monomethylhydrazine and hydrazine are discussed, and performance data from a prototype flow reactor are presented.

A new integral method for solving the point reactor neutron kinetics equations

International Nuclear Information System (INIS)

Li Haofeng; Chen Wenzhen; Luo Lei; Zhu Qian

2009-01-01

A numerical integral method that efficiently provides the solution of the point kinetics equations by using the better basis function (BBF) for the approximation of the neutron density in one time step integrations is described and investigated. The approach is based on an exact analytic integration of the neutron density equation, where the stiffness of the equations is overcome by the fully implicit formulation. The procedure is tested by using a variety of reactivity functions, including step reactivity insertion, ramp input and oscillatory reactivity changes. The solution of the better basis function method is compared to other analytical and numerical solutions of the point reactor kinetics equations. The results show that selecting a better basis function can improve the efficiency and accuracy of this integral method. The better basis function method can be used in real time forecasting for power reactors in order to prevent reactivity accidents.

Preparation of three-dimensional shaped aluminum alloy foam by two-step foaming

International Nuclear Information System (INIS)

Shang, J.T.; Xuming, Chu; Deping, He

2008-01-01

A novel method, named two-step foaming, was investigated to prepare three-dimensional shaped aluminum alloy foam used in car industry, spaceflight, packaging and related areas. Calculations of thermal decomposition kinetics of titanium hydride showed that there is a considerable amount of hydrogen releasing when the titanium hydride is heated at a relatively high temperature after heated at a lower temperature. The hydrogen mass to sustain aluminum alloy foam, having a high porosity, was also estimated by calculations. Calculations indicated that as-received titanium hydride without any pre-treatment can be used as foaming agents in two-step foaming. The processes of two-step foaming, including preparing precursors and baking, were also studied by experiments. Results showed that, low titanium hydride dispersion temperature, long titanium hydride dispersion time and low precursors porosity are beneficial to prepare three-dimensional shaped aluminum alloy foams with uniform pores

Physics of nuclear reactors

International Nuclear Information System (INIS)

Baeten, Peter

2006-01-01

This course gives an introduction to Nuclear Reactor Physics. The first chapter explains the most important parameters and concepts in nuclear reactor physics such as fission, cross sections and the effective multiplication factor. Further on, in the second chapter, the flux distributions in a stationary reactor are derived from the diffusion equation. Reactor kinetics, reactor control and reactor dynamics (feedback effects) are described in the following three chapters. The course concludes with a short description of the different types of existing and future reactors. (author)

Method and program for complex calculation of heterogeneous reactor

International Nuclear Information System (INIS)

Kalashnikov, A.G.; Glebov, A.P.; Elovskaya, L.F.; Kuznetsova, L.I.

1988-01-01

An algorithm and the GITA program for complex one-dimensional calculation of a heterogeneous reactor which permits to conduct calculations for the reactor and its cell simultaneously using the same algorithm are described. Multigroup macrocross sections for reactor zones in the thermal energy range are determined according to the technique for calculating a cell with complicate structure and then the continuous multi group calculation of the reactor in the thermal energy range and in the range of neutron thermalization is made. The kinetic equation is solved using the Pi- and DSn- approximations [fr

Inverse kinetics technique for reactor shutdown measurement: an experimental assessment. [AGR

Energy Technology Data Exchange (ETDEWEB)

Lewis, T. A.; McDonald, D.

1975-09-15

It is proposed to use the Inverse Kinetics Technique to measure the subcritical reactivity as a function of time during the testing of the nitrogen injection systems on AGRs. A description is given of an experimental assessment of the technique by investigating known transients created by control rod movements on a small experimental reactor, (2m high, 1m radius). Spatial effects were observed close to the moving rods but otherwise derived reactivities were independent of detector position and agreed well with the existing calibrations. This prompted the suggestion that data from installed reactor instrumentation could be used to calibrate CAGR control rods.

UNICIN - an one-dimensional computer code for reactor kinetics

International Nuclear Information System (INIS)

Rosa, M.A.P.; Alcantara, H.G. de; Nair, R.P.K.

1984-01-01

A program for the solution of the time- and space-dependent multigroup diffusion equations and the delayed-neutron precursors concentration equations in one dimensional geometries by the weighted residual method is described. The discretized equations are solved through an iterative procedure with convergence accelerated by the over-relaxation method. The system is perturbed through the variation of the nuclide concentrations in specified regions. Two feedback effects are included, namely, the temperature and the burnup. (Author) [pt

Development of a three-dimensional computer code for reconstructing power distributions by means of side reflector instrumentation and determination of the capabilities and limitations of this method

International Nuclear Information System (INIS)

Knob, P.J.

1982-07-01

This work is concerned with the detection of flux disturbances in pebble bed high temperature reactors by means of flux measurements in the side reflector. Included among the disturbances studied are xenon oscillations, rod group insertions, and individual rod insertions. Using the three-dimensional diffusion code CITATION, core calculations for both a very small reactor (KAHTER) and a large reactor (PNP-3000) were carried out to determine the neutron fluxes at the detector positions. These flux values were then used in flux mapping codes for reconstructing the flux distribution in the core. As an extension of the already existing two-dimensional MOFA code, which maps azimuthal disturbances, a new three-dimensional flux mapping code ZELT was developed for handling axial disturbances as well. It was found that both flux mapping programs give satisfactory results for small and large pebble bed reactors alike. (orig.) [de

Multigroup perturbation model for kinetic analysis of nuclear reactors

International Nuclear Information System (INIS)

Souza, G.M.

1989-01-01

The scope of this work is the development of a multigroup perturbation theory for the purpose of Kinetic and dynamic analysis of nuclear reactors. The equations that describe the reactor behavior were presented in all generality and written in the shorthand notation of matrices and vectors. In the derivation of those equations indetermined operators and discretizing factors were introduced and then determined by comparision with conventional equations. Fick's Law was developed in higher orders for neutron and importance current density. The solution of the direct and adjoint fields were represented by combination of the eigenfunctions of the B and B* operators and the eigenvalue modulus equality was established mathematically. In the derivation of the reactivity expression the B operator perturbation was split in two non coupled to the flux form and level. The prompt neutrons effective mean life was derived from reactor equations and importance conservation. The establishment of the Nordheim's equation, although modified, was based on Gandini. Finally, a mathematical interpretation of the flux-trap region was avented. (author)

AIREKMOD-RR, Reactivity Transients in Nuclear Research Reactors

International Nuclear Information System (INIS)

Baggoura, B.; Mazrou, H.

2001-01-01

1 - Description of program or function: AIREMOD-RR is a point kinetics code which can simulate fast transients in nuclear research reactor cores. It can also be used for theoretical reactor dynamics studies. It is used for research reactor kinetic analysis and provides a point neutron kinetic capability. The thermal hydraulic behavior is governed by a one-dimensional heat balance equation. The calculations are restricted to a single equivalent unit cell which consists of fuel, clad and coolant. 2 - Method of solution: For transient reactor kinetic calculations a modified Runge Kutta numerical method is used. The external reactivity insertion, specified as a function of time, is converted in dollar ($) unit. The neutron density, energy release and feedback variables are given at each time step. The two types of reactivity feedback considered are: Doppler effect and moderator effect. A new expression for the reactivity dependence on the feedback variables has been introduced in the present version of the code. The feedback reactivities are fitted in power series expression. 3 - Restrictions on the complexity of the problem: The number of delayed neutron groups and the total number of equations are limited only by computer storage capabilities. - Coolant is always in liquid phase. - Void reactivity feedback is not considered

Three-dimensional magnetospheric equilibrium with isotropic pressure

International Nuclear Information System (INIS)

Cheng, C.Z.

1995-05-01

In the absence of the toroidal flux, two coupled quasi two-dimensional elliptic equilibrium equations have been derived to describe self-consistent three-dimensional static magnetospheric equilibria with isotropic pressure in an optimal (Ψ,α,χ) flux coordinate system, where Ψ is the magnetic flux function, χ is a generalized poloidal angle, α is the toroidal angle, α = φ - δ(Ψ,φ,χ) is the toroidal angle, δ(Ψ,φ,χ) is periodic in φ, and the magnetic field is represented as rvec B = ∇Ψ x ∇α. A three-dimensional magnetospheric equilibrium code, the MAG-3D code, has been developed by employing an iterative metric method. The main difference between the three-dimensional and the two-dimensional axisymmetric solutions is that the field-aligned current and the toroidal magnetic field are finite for the three-dimensional case, but vanish for the two-dimensional axisymmetric case. With the same boundary flux surface shape, the two-dimensional axisymmetric results are similar to the three-dimensional magnetosphere at each local time cross section

Three dimensional visualization of medical images

International Nuclear Information System (INIS)

Suto, Yasuzo

1992-01-01

Three dimensional visualization is a stereoscopic technique that allows the diagnosis and treatment of complicated anatomy site of the bone and organ. In this article, the current status and technical application of three dimensional visualization are introduced with special reference to X-ray CT and MRI. The surface display technique is the most common for three dimensional visualization, consisting of geometric model, voxel element, and stereographic composition techniques. Recent attention has been paid to display method of the content of the subject called as volume rendering, whereby information on the living body is provided accurately. The application of three dimensional visualization is described in terms of diagnostic imaging and surgical simulation. (N.K.)

Three-dimensional simulation of flow, salinity, sediment, and radionuclide movements in the Hudson River estuary

International Nuclear Information System (INIS)

Onishi, Y.; Trent, D.S.

1985-04-01

The three-dimensional, finite difference model, FLESCOT simulates time-varying movements of flow, turbulent kinetic energy, salinity, water temperature, sediment, and contaminants in estuarine, coastal, and ocean waters. The model was applied to a 106-km (66-mi) reach of the Hudson River estuary in New York between Chelsea and the mouth of the river. It predicted the time-varying, three-dimensional distributions of tidal flow, salinity, three separate groups of sediments (i.e., sand, silt, and clay), and a radionuclide ( 137 Cs) in both dissolved and particulate (those sorbed by sediments) forms for over 40 days. The model also calculated riverbed elevation changes caused by sediment deposition and bed erosion, bed sediment size distribution and armoring, and distributions of the particulate 137 Cs sorbed by sand, silt, and clay in the bed

(Weakly) three-dimensional caseology

International Nuclear Information System (INIS)

Pomraning, G.C.

1996-01-01

The singular eigenfunction technique of Case for solving one-dimensional planar symmetry linear transport problems is extended to a restricted class of three-dimensional problems. This class involves planar geometry, but with forcing terms (either boundary conditions or internal sources) which are weakly dependent upon the transverse spatial variables. Our analysis involves a singular perturbation about the classic planar analysis, and leads to the usual Case discrete and continuum modes, but modulated by weakly dependent three-dimensional spatial functions. These functions satisfy parabolic differential equations, with a different diffusion coefficient for each mode. Representative one-speed time-independent transport problems are solved in terms of these generalised Case eigenfunctions. Our treatment is very heuristic, but may provide an impetus for more rigorous analysis. (author)

A new coupling kernel for the three-dimensional simulation of a boiling water reactor core by the nodal coupling method

International Nuclear Information System (INIS)

Gupta, N.K.

1981-01-01

A new coupling kernel is developed for the three-dimensional (3-D) simulation of Boiling Water Reactors (BWR's) by the nodal coupling method. The new kernel depends not only on the properties of the node under consideration but also on the properties of its neighbouring nodes. This makes the kernel more useful in particular for fuel bundles lying in a surrounding of different nuclear characteristics, e.g. for a controlled bundle in the surrounding of uncontrolled bundles or vice-versa. The main parameter in the new kernel is a space-dependent factor obtained from the ratio of thermal-to-fast flux. The average value of the above ratio for each node is evaluated analytically. The kernel is incorporated in a 3-D BWR core simulation program MOGS. As an experimental verification of the model, the cycle-6 operations of the two units of the Tarapur Atomic Power Station (TAPS) are simulated and the result of the simulation are compared with Travelling Incore Probe (TIP) data. (orig.)

Kinetic study on the effect of temperature on biogas production using a lab scale batch reactor.

Science.gov (United States)

Deepanraj, B; Sivasubramanian, V; Jayaraj, S

2015-11-01

In the present study, biogas production from food waste through anaerobic digestion was carried out in a 2l laboratory-scale batch reactor operating at different temperatures with a hydraulic retention time of 30 days. The reactors were operated with a solid concentration of 7.5% of total solids and pH 7. The food wastes used in this experiment were subjected to characterization studies before and after digestion. Modified Gompertz model and Logistic model were used for kinetic study of biogas production. The kinetic parameters, biogas yield potential of the substrate (B), the maximum biogas production rate (Rb) and the duration of lag phase (λ), coefficient of determination (R(2)) and root mean square error (RMSE) were estimated in each case. The effect of temperature on biogas production was evaluated experimentally and compared with the results of kinetic study. The results demonstrated that the reactor with operating temperature of 50°C achieved maximum cumulative biogas production of 7556ml with better biodegradation efficiency. Copyright © 2015 Elsevier Inc. All rights reserved.

Experimental and Kinetic Modeling Study of Ethyl Levulinate Oxidation in a Jet-Stirred Reactor

KAUST Repository

Wang, Jui-Yang

2017-06-01

A jet-stirred reactor was designed and constructed in the Clean Combustion Research Center (CCRC) at King Abdullah University of Science and Technology (KAUST); was validated with n-heptane, iso-octane oxidation and cyclohexene pyrolysis. Different configurations of the setup have been tested to achieve good agreement with results from the literature. Test results of the reactor indicated that installation of a pumping system at the downstream side in the experimental apparatus was necessary to avoid the reoccurrence of reactions in the sampling probe. Experiments in ethyl levulinate oxidation were conducted in the reactor under several equivalence ratios, from 600 to 1000 K, 1 bar and 2 s residence time. Oxygenated species detected included methyl vinyl ketone, levulinic acid and ethyl acrylate. Ethylene, methane, carbon monoxide, hydrogen, oxygen and carbon dioxide were further quantified with a gas chromatography, coupled with a flame ionization detector and a thermal conductivity detector. The ethyl levulinate chemical kinetic model was first developed by Dr. Stephen Dooley, Trinity College Dublin, and simulated under the same conditions, using the Perfect-Stirred Reactor code in Chemkin software. In comparing the simulation results with experimental data, some discrepancies were noted; predictions of ethylene production were not well matched. The kinetic model was improved by updating several classes of reactions: unimolecular decomposition, H-abstraction, C-C and C-O beta-scissions of fuel radicals. The updated model was then compared again with experimental results and good agreement was achieved, proving that the concerted eliminated reaction is crucial for the kinetic mechanism formulation of ethyl levulinate. In addition, primary reaction pathways and sensitivity analysis were performed to describe the role of molecular structure in combustion (800 and 1000 K for ethyl levulinate oxidation in the jet-stirred reactor).

Measures of the zero power nuclear reactor's kinetic parameters with application of noise analysis

International Nuclear Information System (INIS)

Martins, F.R.

1992-01-01

The purpose of this work was to establish an experimental technique based on noise analysis for measuring the ratio of kinetic parameters β/ Λ and the power of the Zero Power Nuclear Reactor IPEN-MB 01. A through study of the microscopic and macroscopic noise analysis techniques has been carried out. The Langevin technique and the point kinetic model were chosen to describe the stochastic phenomena that occur in the zero power reactor. Measurements have been made using two compensated ionization chambers localized in the water reflector at symmetric positions in order to minimize spatial effects on the neutron flux fluctuation. Power calibrations based on the low frequency plateau of the cross-power spectral density has also been carried out. (author)

Comparison of reactor RA-4 kinetics with simulations with Matlab-Simulink for one group and six groups of delayed neutrons

International Nuclear Information System (INIS)

Orso, J A

2012-01-01

The critical state of a nuclear reactor is an unstable equilibrium. The nuclear reactor can go from critical to subcritical state or can go from critical to hypercritical state. Although the evolution of the system in these cases is slow, it requires the intervention of an operator to correct deviations. For this reason an automatic control technique was designed, based on the kinetic point to a group of delayed neutrons, which corrects deviations automatically. In this paper we study the point kinetics models in a group and six groups of delayed neutrons for different values of reactivity using the simulations software MATLAB, Simulink. A comparison of two models with the reactor kinetic behavior is made (author)

Initial value problem for the equations of reactor kinetics

International Nuclear Information System (INIS)

Kyncl, J.

1987-08-01

The initial value problem for the equations of reactor kinetics is solved while taking temperature feedback into account. The space where the problem is solved is chosen such as to correspond to the mathematical properties of cross-section models. The local solution is found by the iterative method, its uniqueness is proved and it is also shown that the existence of global solution is ensured in most cases. Finally, the problem of a weak solution is discussed. (author). 5 refs

SPQR: a Monte Carlo reactor kinetics code

International Nuclear Information System (INIS)

Cramer, S.N.; Dodds, H.L.

1980-02-01

The SPQR Monte Carlo code has been developed to analyze fast reactor core accident problems where conventional methods are considered inadequate. The code is based on the adiabatic approximation of the quasi-static method. This initial version contains no automatic material motion or feedback. An existing Monte Carlo code is used to calculate the shape functions and the integral quantities needed in the kinetics module. Several sample problems have been devised and analyzed. Due to the large statistical uncertainty associated with the calculation of reactivity in accident simulations, the results, especially at later times, differ greatly from deterministic methods. It was also found that in large uncoupled systems, the Monte Carlo method has difficulty in handling asymmetric perturbations

Numerical Simulation of Three-Dimensional Flow Through Full Passage and Performance Prediction of Nuclear Reactor Coolant Pump

International Nuclear Information System (INIS)

Li Ying; Zhou Wenxia; Zhang Jige; Wang Dezhong

2009-01-01

In order to achieve the level of self-design and domestic manufacture of the reactor coolant pump (nuclear main pump), the software FLUENT was used to simulate the three-dimensional flow through full passage of one nuclear main pump basing on RNG κ-ε turbulence model and SIMPLE algorithm. The distribution of pressure and velocity of the flow in the impeller's surface was analyzed in different working conditions. Moreover, the performance of the pump was predicted based on the simulation results. The results show that the distributions of pressure and velocity are reasonable in both the working and back face of the blade in the steady working condition. The pressure of the flow is increased from the inlet to the outlet of the pump, and shows the maximal value in the impeller region. Comparatively satisfactory efficiency and head value were obtained in the condition of the pump design. The shaft power of the nuclear main pump is gradually increased with the increase of the flow flux. These results are helpful in understanding the change of the internal flow field in the nuclear main pump, which is of some importance for the pre-exploration and theoretical research on the domestic manufacture of the nuclear main pump. (authors)

Development and verification of a three-dimensional core model for WWR type reactors and its coupling with the accident code ATHLET. Final report

International Nuclear Information System (INIS)

Grundmann, U.; Lucas, D.; Mittag, S.; Rohde, U.

1995-04-01

The main goal of the project was the coupling of the 3D core model DYN3D for Russian VVER-type reactors, which has been developed in the RCR, with the thermohydraulic code ATHLET. The coupling has been realized on two basically different ways: - The implementation of only the neutron kinetics model of DYN3D into ATHLET (internal coupling), - the connection of the complete DYN3D core model including neutron kinetics, thermohydraulics and fuel rod model via data interfaces at the core top and bottom (external coupling). For the test of the coupling, comparative calculations between internal and external coupling versions have been carried out for a LOCA and a reactivity transient. Complementary goals of the project were: - The development of a DYN3D version for burn-up calculations, - the verification of DYN3D on benchmark tasks and experimental data on fuel rod behaviour, - a study on the extension of the neutron-physical data base. The project contributed to the development of advanced tools for the safety analysis of VVER-type reactors. Future work is aimed to the verification of the coupled code complex DYN3D-ATHLET. (orig.) [de

Deciphering robust reactor kinetic data using mutual information

International Nuclear Information System (INIS)

Kumar, P.T. Krishna

2007-01-01

Experimentalists use Chauvenets's criterion to check the quality of any measured data. Based on this criterion they rejected data having high degree of correlation. Multivariate techniques like principal component analysis used for analysis of these correlated data, does not provide any scope to minimize the effect of correlation. We propose a novel method using information theory and the technique of determinant inequalities developed by us to reduce the effect of correlation among these data without summarily rejecting them. We demonstrate the utility of our technique in transient measurements of kinetic parameters performed on the commercially advanced gas cooled reactor (CAGCR)

Calculation of accurate albedo boundary conditions for three-dimensional nodal diffusion codes by the method of characteristics

International Nuclear Information System (INIS)

Petkov, Petko T.

2000-01-01

Most of the few-group three-dimensional nodal diffusion codes used for neutronics calculations of the WWER reactors use albedo type boundary conditions on the core-reflector boundary. The conventional albedo are group-to-group reflection probabilities, defined on each outer node face. The method of characteristics is used to calculate accurate albedo by the following procedure. A many-group two-dimensional heterogeneous core-reflector problem, including a sufficient part of the core and detailed description of the adjacent reflector, is solved first. From this solution the angular flux on the core-reflector boundary is calculated in all groups for all traced neutron directions. Accurate boundary conditions can be calculated for the radial, top and bottom reflectors as well as for the absorber part of the WWER-440 reactor control assemblies. The algorithm can be used to estimate also albedo, coupling outer node faces on the radial reflector in the axial direction. Numerical results for the WWER-440 reactor are presented. (Authors)

Simulation of radiation effects on three-dimensional computer optical memories

Science.gov (United States)

Moscovitch, M.; Emfietzoglou, D.

1997-01-01

A model was developed to simulate the effects of heavy charged-particle (HCP) radiation on the information stored in three-dimensional computer optical memories. The model is based on (i) the HCP track radial dose distribution, (ii) the spatial and temporal distribution of temperature in the track, (iii) the matrix-specific radiation-induced changes that will affect the response, and (iv) the kinetics of transition of photochromic molecules from the colored to the colorless isomeric form (bit flip). It is shown that information stored in a volume of several nanometers radius around the particle's track axis may be lost. The magnitude of the effect is dependent on the particle's track structure.

Three-dimensional laser-induced fluorescence measurements of turbulent chemical plumes

Science.gov (United States)

True, Aaron; Crimaldi, John

2017-11-01

In order to find prey, mates, and suitable habitat, many organisms must navigate through complex chemical plume structures in turbulent flow environments. In this context, we investigate the spatial and temporal structure of chemical plumes released isokinetically into fractal-grid-generated turbulence in an open channel flow. We first utilized particle image velocimetry (PIV) to characterize flow conditions (mean free stream velocities, turbulence intensities, turbulent kinetic energy dissipation rates, Taylor Reynolds numbers). We then implemented a newly developed high-resolution, high-speed, volumetric scanning laser-induced fluorescence (LIF) system for near time-resolved measurements of three-dimensional chemical plume structures. We investigated cases with and without a cylinder wake, and compare statistical (mean, variance, intermittency, probability density functions) and spectral (power spectrum of concentration fluctuations) characteristics of the chemical plume structure. Stretching and folding of complex three-dimensional filament structures during chaotic turbulent mixing is greatly enhanced in the cylinder wake case. In future experiments, we will implement simultaneous PIV and LIF, enabling computation of the covariance of the velocity and chemical concentration fluctuations and thus estimation of turbulent eddy diffusivities. NSF PHY 1555862.

Heat, mass, and momentum transport model for hydrogen diffusion flames in nuclear reactor containments

International Nuclear Information System (INIS)

Travis, J.R.

1985-01-01

It is now possible to analyze the time-dependent, fully three-dimensional behavior of hydrogen diffusion flames in nuclear reactor containments. This analysis involves coupling the full Navier-Stokes equations with multi-species transport to the global chemical kinetics of hydrogen combustion. A transport equation for the subgrid scale turbulent kinetic energy density is solved to produce the time and space dependent turbulent transport coefficients. The heat transfer coefficient governing the exchange of heat between fluid computational cells adjacent to wall cells is calculated by a modified Reynolds analogy formulation. The analysis of a MARK-III containment indicates very complex flow patterns that greatly influence fluid and wall temperatures and heat fluxes. 18 refs., 24 figs

Comparison of finite element and influence function methods for three-dimensional elastic analysis of boiling water reactor feedwater nozzle cracks. Phase report

International Nuclear Information System (INIS)

Besuner, P.M.; Caughey, W.R.

1976-11-01

The finite element (FE) and influence function (IF) methods are compared for a three-dimensional elastic analysis of postulated circular-shaped surface cracks in the feedwater nozzle of a typical boiling water reactor (BWR). These are two of the possible methods for determining stress intensity factors for nozzle corner cracks. The FE method is incorporated in a direct manner. The IF method is used to compute stress intensity factors only when the uncracked stress field (i.e., the stress in the uncracked solid at the locus of the crack to be eventually considered) has been computed previously. Both the IF and FE methods are described in detail and are applied to several test cases chosen for their similarity to the nozzle crack problem and for the availablility of an accurate published result obtained from some recognized third method of solution

The application of polynomial chaos methods to a point kinetics model of MIPR: An Aqueous Homogeneous Reactor

International Nuclear Information System (INIS)

Cooling, C.M.; Williams, M.M.R.; Nygaard, E.T.; Eaton, M.D.

2013-01-01

Highlights: • A point kinetics model for the Medical Isotope Production Reactor is formulated. • Reactivity insertions are simulated using this model. • Polynomial chaos is used to simulate uncertainty in reactor parameters. • The computational efficiency of polynomial chaos is compared to that of Monte Carlo. -- Abstract: This paper models a conceptual Medical Isotope Production Reactor (MIPR) using a point kinetics model which is used to explore power excursions in the event of a reactivity insertion. The effect of uncertainty of key parameters is modelled using intrusive polynomial chaos. It is found that the system is stable against reactivity insertions and power excursions are all bounded and tend towards a new equilibrium state due to the negative feedbacks inherent in Aqueous Homogeneous Reactors (AHRs). The Polynomial Chaos Expansion (PCE) method is found to be much more computationally efficient than that of Monte Carlo simulation in this application

Kinetic characterization for hemicellulose hydrolysis of corn stover in a dilute acid cycle spray flow-through reactor at moderate conditions

International Nuclear Information System (INIS)

Jin, Qiang; Zhang, Hongman; Yan, Lishi; Qu, Liang; Huang, He

2011-01-01

The kinetic characterization of hemicellulose hydrolysis of corn stover was investigated using a new reactor of dilute acid cycle spray flow-through (DCF) pretreatment. The primary purpose was to obtain kinetic data for hemicellulose hydrolysis with sulfuric acid concentrations (10-30 kg m -3 ) at relatively low temperatures (90-100 o C). A simplified kinetic model was used to describe its performance at moderate conditions. The results indicate that the rates of xylose formation and degradation are sensitive to flow rate, temperature and acid concentration. Moreover, the kinetic data of hemicellulose hydrolysis fit a first-order reaction model and the experimental data with actual acid concentration after accounting for the neutralization effect of the substrates at different temperatures. Over 90% of the xylose monomer yield and below 5.5% of degradation product (furfural) yield were observed in this reactor. Kinetic constants for hemicellulose hydrolysis models were analyzed by an Arrhenius-type equation, and the activation energy of xylose formation were 111.6 kJ mol -1 , and 95.7 kJ mol -1 for xylose degradation, respectively. -- Highlights: → Investigating a novel pretreatment reactor of dilute acid cycle spray flow-through. → Xylose yield is sensitive to flow rate, temperature and acid concentration. → Obtaining relatively higher xylose monomer yield and lower fermentation inhibitor. → Lumping hemicellulose and xylan oligmers together in the model is a valid way. → The kinetic model as a guide for reactor design, and operation strategy optimization.

Application of point kinetic model in the study of fluidized bed reactor dynamic

International Nuclear Information System (INIS)

Borges, Volnei; Vilhena, Marco Tullio de; Streck, Elaine E.

1995-01-01

In this work the dynamical behavior of the fluidized bed nuclear reactor is analysed. The main goal consist to study the effect of the acceleration term in the point kinetic equations. Numerical simulations are reported considering constant acceleration. (author). 7 refs, 4 figs

G4-STORK: A Geant4-based Monte Carlo reactor kinetics simulation code

International Nuclear Information System (INIS)

Russell, Liam; Buijs, Adriaan; Jonkmans, Guy

2014-01-01

Highlights: • G4-STORK is a new, time-dependent, Monte Carlo code for reactor physics applications. • G4-STORK was built by adapting and expanding on the Geant4 Monte Carlo toolkit. • G4-STORK was designed to simulate short-term fluctuations in reactor cores. • G4-STORK is well suited for simulating sub- and supercritical assemblies. • G4-STORK was verified through comparisons with DRAGON and MCNP. - Abstract: In this paper we introduce G4-STORK (Geant4 STOchastic Reactor Kinetics), a new, time-dependent, Monte Carlo particle tracking code for reactor physics applications. G4-STORK was built by adapting and expanding on the Geant4 Monte Carlo toolkit. The toolkit provides the fundamental physics models and particle tracking algorithms that track each particle in space and time. It is a framework for further development (e.g. for projects such as G4-STORK). G4-STORK derives reactor physics parameters (e.g. k eff ) from the continuous evolution of a population of neutrons in space and time in the given simulation geometry. In this paper we detail the major additions to the Geant4 toolkit that were necessary to create G4-STORK. These include a renormalization process that maintains a manageable number of neutrons in the simulation even in very sub- or supercritical systems, scoring processes (e.g. recording fission locations, total neutrons produced and lost, etc.) that allow G4-STORK to calculate the reactor physics parameters, and dynamic simulation geometries that can change over the course of simulation to illicit reactor kinetics responses (e.g. fuel temperature reactivity feedback). The additions are verified through simple simulations and code-to-code comparisons with established reactor physics codes such as DRAGON and MCNP. Additionally, G4-STORK was developed to run a single simulation in parallel over many processors using MPI (Message Passing Interface) pipes

Thermal hydraulic and neutron kinetic coupled simulation of the IPR-R1 Triga reactor

Energy Technology Data Exchange (ETDEWEB)

Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia; Silva, Clarysson A.M. da; Veloso, Maria Auxiliadora F.; Soares, Humbero V., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: clarysson@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: betovitor@ig.com.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq Rede), Rio de Janeiro, RJ (Brazil)

2013-07-01

The nuclear industry and the scientific community have turned the attention for the development of coupled 3D neutron kinetics (NK) and thermal-hydraulic (TH) system codes to investigate specific nuclear reactor transients. Improving in theoretical investigations of complex phenomena in nuclear reactor technology have been increased thanks to numerical methods and computational resources incorporated in nuclear codes. This paper presents a model for the IPR-R1 TRIGA research reactor using the RELAP5-3D 3.0 code. The development and the assessment of the thermal-hydraulic RELAP5 code model for the IPR-R1 have been validated for steady state and transient situations and the results were published in preceding works. Results of RELAP5-3D steady state and a transient case presented in this paper show good agreement with experimental data, validating then this model for point kinetic calculations. To supply adequate cross sections to the NK code, the WIMSD5 is being used. First results of steady state calculation using the 3D neutron modeling are being presented in this paper. (author)

A new improvement on a chemical kinetic model of primary reference fuel for multi-dimensional CFD simulation

International Nuclear Information System (INIS)

Zhen, Xudong; Wang, Yang; Liu, Daming

2016-01-01

Highlights: • A new optimized chemical kinetic mechanism for PRF is developed. • New mechanism optimization is performed based on the CHEMKIN simulations. • More reactions of C_0–C_1 oxidation are added in the present mechanism. • Good performance is achieved of mechanism by validating various reactors and operating conditions. - Abstract: In the present study, for the multi-dimensional CFD (computational fluid dynamics) combustion simulations of internal combustion engines, a new optimized chemical kinetic reaction mechanism for the oxidation of PRF (primary reference fuel) instead of gasoline has been developed. In order to carry out the in-depth research for combustion phenomenon of internal combustion engines, an optimized reduced PRF mechanism including more intermediate species and radicals was developed. The developed mechanism contains of iso-octane (C_8H_1_8) and n-heptane (C_7H_1_6) surrogates, which contains of 51-species and 193 reactions. Compared with many other mechanisms of PRF, more reactions of C_0–C_1 oxidation (100 reactions) are added in the present mechanism. In order to improve the performances of the model, the developed mechanism focused on the improvement through the prediction of the ignition delay time. The developed mechanism has been validated against various experimental and simulation data including shock tube data, laminar flame speed data and HCCI (homogeneous charge compression ignition) engine data. The results showed that the developed PRF mechanism was agreements with the experimental data and other approved reduced mechanisms, and it could be applied to the multi-dimensional CFD simulations for internal combustion engines.

Inverse kinetics method with source term for subcriticality measurements during criticality approach in the IPEN/MB-01 research reactor

International Nuclear Information System (INIS)

Loureiro, Cesar Augusto Domingues; Santos, Adimir dos

2009-01-01

In reactor physics tests which are performed at the startup after refueling the commercial PWRs, it is important to monitor subcriticality continuously during criticality approach. Reactivity measurements by the inverse kinetics method are widely used during the operation of a nuclear reactor and it is possible to perform an online reactivity measurement based on the point reactor kinetics equations. This technique is successful applied at sufficiently high power level or to a core without an external neutron source where the neutron source term in point reactor kinetics equations may be neglected. For operation at low power levels, the contribution of the neutron source must be taken into account and this implies the knowledge of a quantity proportional to the source strength, and then it should be determined. Some experiments have been performed in the IPEN/MB-01 Research Reactor for the determination of the Source Term, using the Least Square Inverse Kinetics Method (LSIKM). A digital reactivity meter which neglects the source term is used to calculate the reactivity and then the source term can be determined by the LSIKM. After determining the source term, its value can be added to the algorithm and the reactivity can be determined again, considering the source term. The new digital reactivity meter can be used now to monitor reactivity during the criticality approach and the measured value for the reactivity is more precise than the meter which neglects the source term. (author)

A two dimensional code (R,Z) for nuclear reactor analysis and its application to the UAR-RI reactor

International Nuclear Information System (INIS)

Bishay, S.T.; Mikhail, I.F.I.; Gaafar, M.A.; Michaiel, M.L.; Nassar, S.F.

1988-01-01

A detailed study is given of fuel consumption in completely reflected cylindrical reactors. A two group, two dimensional (r,z) code is developed to carry out the required procedure. The code is applied to the UAR-RI reactor and the results are found to be in complete agreement with the experimental observations and with the theoretical interpretations. 7 fig., 12 tab

Kinetic modeling of the photocatalytic degradation of clofibric acid in a slurry reactor.

Science.gov (United States)

Manassero, Agustina; Satuf, María Lucila; Alfano, Orlando Mario

2015-01-01

A kinetic study of the photocatalytic degradation of the pharmaceutical clofibric acid is presented. Experiments were carried out under UV radiation employing titanium dioxide in water suspension. The main reaction intermediates were identified and quantified. Intrinsic expressions to represent the kinetics of clofibric acid and the main intermediates were derived. The modeling of the radiation field in the reactor was carried out by Monte Carlo simulation. Experimental runs were performed by varying the catalyst concentration and the incident radiation. Kinetic parameters were estimated from the experiments by applying a non-linear regression procedure. Good agreement was obtained between model predictions and experimental data, with an error of 5.9 % in the estimations of the primary pollutant concentration.

Three-dimensional particle image velocimetry measurement technique

International Nuclear Information System (INIS)

Hassan, Y.A.; Seeley, C.H.; Henderson, J.A.; Schmidl, W.D.

2004-01-01

The experimental flow visualization tool, Particle Image Velocimetry (PIV), is being used to determine the velocity field in two-dimensional fluid flows. In the past few years, the technique has been improved to allow the capture of flow fields in three dimensions. This paper describes changes which were made to two existing two-dimensional tracking algorithms to enable them to track three-dimensional PIV data. Results of the tests performed on these three-dimensional routines with synthetic data are presented. Experimental data was also used to test the tracking algorithms. The test setup which was used to acquire the three-dimensional experimental data is described, along with the results from both of the tracking routines which were used to analyze the experimental data. (author)

Prospects in deterministic three dimensional whole-core transport calculations

International Nuclear Information System (INIS)

Sanchez, Richard

2012-01-01

The point we made in this paper is that, although detailed and precise three-dimensional (3D) whole-core transport calculations may be obtained in the future with massively parallel computers, they would have an application to only some of the problems of the nuclear industry, more precisely those regarding multiphysics or for methodology validation or nuclear safety calculations. On the other hand, typical design reactor cycle calculations comprising many one-point core calculations can have very strict constraints in computing time and will not directly benefit from the advances in computations in large scale computers. Consequently, in this paper we review some of the deterministic 3D transport methods which in the very near future may have potential for industrial applications and, even with low-order approximations such as a low resolution in energy, might represent an advantage as compared with present industrial methodology, for which one of the main approximations is due to power reconstruction. These methods comprise the response-matrix method and methods based on the two-dimensional (2D) method of characteristics, such as the fusion method.

Experimental and Kinetic Modeling Study of Ethyl Levulinate Oxidation in a Jet-Stirred Reactor

KAUST Repository

Wang, Jui-Yang

2017-01-01

levulinate chemical kinetic model was first developed by Dr. Stephen Dooley, Trinity College Dublin, and simulated under the same conditions, using the Perfect-Stirred Reactor code in Chemkin software. In comparing the simulation results with experimental

AUS, Neutron Transport and Gamma Transport System for Fission Reactors and Fusion Reactors

International Nuclear Information System (INIS)

1990-01-01

1 - Description of program or function: AUS is a neutronics code system which may be used for calculations of a wide range of fission reactors, fusion blankets and other neutron applications. The present version, AUS98, has a nuclear cross section library based on ENDF/B-VI and includes modules which provide for reactor lattice calculations, one-dimensional transport calculations, multi-dimensional diffusion calculations, cell and whole reactor burnup calculations, and flexible editing of results. Calculations of multi-region resonance shielding, coupled neutron and photon transport, energy deposition, fission product inventory and neutron diffusion are combined within the one code system. The major changes from the previous release, AUS87, are the inclusion of a cross-section library based on ENDF/B-VI, the addition of the POW3D multi-dimensional diffusion module, the addition of the MICBURN module for controlling whole reactor burnup calculations, and changes to the system as a consequence of moving from IBM mainframe computers to UNIX workstations. 2 - Method of solution: AUS98 is a modular system in which the modules are complete programs linked by a path given in the input stream. A simple path is simply a sequence of modules, but the path is actually pre-processed and compiled using the Fortran 77 compiler. This provides for complex module linking if required. Some of the modules included in AUS98 are: MIRANDA Cross-section generation in a multi-region resonance subgroup calculation and preliminary group condensation. ANAUSN One-dimensional discrete ordinates calculation. ICPP Isotropic collision probability calculation in one dimension and for rod clusters. POW3D Multi-dimensional neutron diffusion calculation including feedback-free kinetics. AUSIDD One-dimensional diffusion calculation. EDITAR Reaction-rate editing and group collapsing following a transport calculation. CHAR Lattice and global burnup calculation. MICBURN Control of global burnup

TEMPEST: A three-dimensional time-dependent computer program for hydrothermal analysis: Volume 1, Numerical methods and input instructions

International Nuclear Information System (INIS)

Trent, D.S.; Eyler, L.L.; Budden, M.J.

1983-09-01

This document describes the numerical methods, current capabilities, and the use of the TEMPEST (Version L, MOD 2) computer program. TEMPEST is a transient, three-dimensional, hydrothermal computer program that is designed to analyze a broad range of coupled fluid dynamic and heat transfer systems of particular interest to the Fast Breeder Reactor thermal-hydraulic design community. The full three-dimensional, time-dependent equations of motion, continuity, and heat transport are solved for either laminar or turbulent fluid flow, including heat diffusion and generation in both solid and liquid materials. 10 refs., 22 figs., 2 tabs

The validation of neutron kinetic calculations of CEGB reactors

International Nuclear Information System (INIS)

Emmett, J.C.A.; Hutt, P.K.; Nunn, D.L.; Waterson, R.H.

1982-01-01

Reactor kinetic calculations are required by the CEGB to predict space and time varying neutron fluxes through the course of various hypothesized core transients. These transients arise through flow or reactivity perturbations occurring in a part of the core. A description is given of the results of dual programmes of work undertaken at BNL to validate such calculations. Firstly, analyses have been carried out to establish how data for these calculations should best be derived. Secondly, experimental measurements have been compared against the predictions of such calculations with data derived in the recommended way. (author)

Structures of two-dimensional three-body systems

International Nuclear Information System (INIS)

Ruan, W.Y.; Liu, Y.Y.; Bao, C.G.

1996-01-01

Features of the structure of L = 0 states of a two-dimensional three-body model system have been investigated. Three types of permutation symmetry of the spatial part, namely symmetric, antisymmetric, and mixed, have been considered. A comparison has been made between the two-dimensional system and the corresponding three-dimensional one. The effect of symmetry on microscopic structures is emphasized. (author)

AIREK-PUL, Periodic Kinetics Problems of Pulsed Reactors

International Nuclear Information System (INIS)

Inzaghi, A.; Misenta, R.

1984-01-01

1 - Nature of physical problem solved: Solves periodic problems about the kinetics of pulsed reactors or problems where the reactivity has rapid variations. The program uses two constant steps for the integration of the system of differential equations, the first step during the first half-period and the second step during the second half-period. Available for either single or double precision. 2 - Method of solution: The differential equations are integrated using the fourth-order Runge-Kutta method as modified by E.R. Cohen (Geneva Conference, 1958). 3 - Restrictions on the complexity of the problem: The maximum number of differential equations that can be solved simultaneously is 50

Three-Dimensional FIB/EBSD Characterization of Irradiated HfAl3-Al Composite

Energy Technology Data Exchange (ETDEWEB)

Hua, Zilong; Guillen, Donna Post; Harris, William; Ban, Heng

2016-09-01

A thermal neutron absorbing material, comprised of 28.4 vol% HfAl3 in an Al matrix, was developed to serve as a conductively cooled thermal neutron filter to enable fast flux materials and fuels testing in a pressurized water reactor. In order to observe the microstructural change of the HfAl3-Al composite due to neutron irradiation, an EBSD-FIB characterization approach is developed and presented in this paper. Using the focused ion beam (FIB), the sample was fabricated to 25µm × 25µm × 20 µm and mounted on the grid. A series of operations were carried out repetitively on the sample top surface to prepare it for scanning electron microscopy (SEM). First, a ~100-nm layer was removed by high voltage FIB milling. Then, several cleaning passes were performed on the newly exposed surface using low voltage FIB milling to improve the SEM image quality. Last, the surface was scanned by Electron Backscattering Diffraction (EBSD) to obtain the two-dimensional image. After 50 to 100 two-dimensional images were collected, the images were stacked to reconstruct a three-dimensional model using DREAM.3D software. Two such reconstructed three-dimensional models were obtained from samples of the original and post-irradiation HfAl3-Al composite respectively, from which the most significant microstructural change caused by neutron irradiation apparently is the size reduction of both HfAl3 and Al grains. The possible reason is the thermal expansion and related thermal strain from the thermal neutron absorption. This technique can be applied to three-dimensional microstructure characterization of irradiated materials.

Elastocapillary fabrication of three-dimensional microstructures

NARCIS (Netherlands)

van Honschoten, J.W.; Berenschot, Johan W.; Ondarcuhu, T.; Sanders, Remco G.P.; Sundaram, J.; Elwenspoek, Michael Curt; Tas, Niels Roelof

2010-01-01

We describe the fabrication of three-dimensional microstructures by means of capillary forces. Using an origami-like technique, planar silicon nitride structures of various geometries are folded to produce three-dimensional objects of 50–100 m. Capillarity is a particularly effective mechanism since

Numerical solution of multigroup diffuse equations of one-dimensional geometry

International Nuclear Information System (INIS)

Pavelesku, M.; Adam, S.

1975-01-01

The one-dimensional diffuse theory is used for reactor physics calculations of fast reactors. Computer program based on the one-dimensional diffuse theory is speedy and not memory consuming. The algorithm is described for the three-zone fast reactor criticality computation in one-dimensional diffusion approximation. This algorithm is realised on IBM 370/135 computer. (I.T.)

Kinetic modeling of hydrocracking reaction in a trickle-bed reactor with Pt/Y-zeolite catalysts

Energy Technology Data Exchange (ETDEWEB)

Lee, BalSang; Park, Myung-June; Kim, Young-A; Park, Eun Duck [Ajou University, Suwon (Korea, Republic of); Han, Jeongsik [Agency for Defense Development, Daejeon (Korea, Republic of); Jeong, Kwang-Eun; Kim, Chul-Ung; Jeong, Soon-Yong [Korea Research Institute of Chemical Technology (KRICT), Daejeon (Korea, Republic of)

2014-03-15

A kinetic model is developed to predict the entire distribution of hydrocarbon products for the hydrocracking reaction with Pt/Y-zeolite catalysts in a trickle-bed reactor. Operating conditions, such as temperature, pressure, and wax and H{sub 2} flow rates were varied to evaluate their effects on conversion and distribution, and kinetic parameters were estimated using the experimental data that covers the window of operating conditions. The comparison between experimental data and simulated results corroborated the validity of the developed model, and the quantitative prediction of the reactor performance was clearly demonstrated. To make evident the usefulness of the model, an optimization method, genetic algorithm (GA), was applied, and the optimal condition for the maximum production of C{sub 10}-C{sub 17} was successfully calculated.

Kinetic modeling of hydrocracking reaction in a trickle-bed reactor with Pt/Y-zeolite catalysts

International Nuclear Information System (INIS)

Lee, BalSang; Park, Myung-June; Kim, Young-A; Park, Eun Duck; Han, Jeongsik; Jeong, Kwang-Eun; Kim, Chul-Ung; Jeong, Soon-Yong

2014-01-01

A kinetic model is developed to predict the entire distribution of hydrocarbon products for the hydrocracking reaction with Pt/Y-zeolite catalysts in a trickle-bed reactor. Operating conditions, such as temperature, pressure, and wax and H 2 flow rates were varied to evaluate their effects on conversion and distribution, and kinetic parameters were estimated using the experimental data that covers the window of operating conditions. The comparison between experimental data and simulated results corroborated the validity of the developed model, and the quantitative prediction of the reactor performance was clearly demonstrated. To make evident the usefulness of the model, an optimization method, genetic algorithm (GA), was applied, and the optimal condition for the maximum production of C 10 -C 17 was successfully calculated

Automated Determination of Oxygen-Dependent Enzyme Kinetics in a Tube-in-Tube Flow Reactor.

Science.gov (United States)

Ringborg, Rolf H; Toftgaard Pedersen, Asbjørn; Woodley, John M

2017-09-08

Enzyme-mediated oxidation is of particular interest to synthetic organic chemists. However, the implementation of such systems demands knowledge of enzyme kinetics. Conventionally collecting kinetic data for biocatalytic oxidations is fraught with difficulties such as low oxygen solubility in water and limited oxygen supply. Here, we present a novel method for the collection of such kinetic data using a pressurized tube-in-tube reactor, operated in the low-dispersed flow regime to generate time-series data, with minimal material consumption. Experimental development and validation of the instrument revealed not only the high degree of accuracy of the kinetic data obtained, but also the necessity of making measurements in this way to enable the accurate evaluation of high K MO enzyme systems. For the first time, this paves the way to integrate kinetic data into the protein engineering cycle.

Quantification of plant cell coupling with three-dimensional photoactivation microscopy.

Science.gov (United States)

Liesche, J; Schulz, A

2012-07-01

Plant cells are directly connected by plasmodesmata that form channels through the cell wall and enable the intercellular movement of cytosolic solutes, membrane lipids and signalling molecules. Transport through plasmodesmata is regulated not only by a fixed size-exclusion limit, but also by physiological and pathological adaptation. To understand plant cell communication, carbon allocation and pathogen attack, the capacities for a specific molecule to pass a specific cell-wall interface is an essential parameter. So far, the degree of cell coupling was derived from frequency and diameter of plasmodesmata in relevant tissues as assessed by electron microscopy of fixed material. However, plasmodesmata functionality and capacity can only be determined in live material, not from electron microscopy, which is static and prone to fixation artefacts. Plasmodesmata functionality was a few times assessed using fluorescent tracers with diffusion properties similar to cytosolic solutes. Here, we used three-dimensional photoactivation microscopy to quantify plasmodesmata-mediated cell-wall permeability between living Cucurbita maxima leaf mesophyll cells with caged fluorescein as tracer. For the first time, all necessary functional and anatomical data were gathered for each individual cell from three-dimensional time series. This approach utilized a confocal microscope equipped with resonant scanner, which provides the high acquisition speed necessary to record optical sections of whole cells and offers time resolution high enough to follow the kinetics of photoactivation. The results were compared to two-dimensional measurements, which are shown to give a good estimate of cell coupling adequate for homogenous tissues. The two-dimensional approach is limited whenever tissues interfaces are studied that couple different cell types with diverse cell geometries. © 2011 The Authors Journal of Microscopy © 2011 Royal Microscopical Society.

The study of the irradiation-induced embrittlement of reactor pressure vessels. Analysis of surveillance test specimens of a commercial nuclear reactor pressure vessel studied by three-dimensional atom probe and positron annihilation

International Nuclear Information System (INIS)

Nagai, Yasuyoshi; Toyama, Takeshi; Hasegawa, Masayuki

2007-01-01

The study of embrittlement of nuclear power reactor pressure vessels (RPVs) is of critical importance for the safety assessment in the nuclear industry. Some origins of embrittlement are attributed to fine Cu precipitates, matrix defects, grain boundary segregation of P and late blooming phase. This review article described nanostructural observation by three-dimensional atom probe (3DAP) and positron annihilation spectroscopy (PAS). The density and sizes of Cu-rich nanoprecipitates and grain boundary segregation are sensitively detected by 3DAP, and vacancies are probed by PAS. Element analysis around vacancies and fine microstructural Cu precipitates not containing vacancies are successfully observed by a coincidence doppler broadening method. The nanostructural evolution of irradiation-induced Cu-rich nanoprecipitates (CRNPs) and vacancy clusters in surveillance test specimens of commercial nuclear reactor pressure vessel steel welds of Doel-2 in Belgium were revealed by combining 3DAP and PAS. In both medium (0.13 wt%) and high (0.30 wt%) Cu welds, the CRNPs were found to form readily at the very beginning of the reactor lifetime. On the other hand, small vacancy clusters start appearing after the initial Cu precipitates and accumulate steadily with increasing neutron dose. The CRNPs were also observed at very low dose rate of neutrons in the test specimen of Calder Hall Reactor of Japan Atomic Power Company. The significant enhancement of these Cu precipitates results in the embrittlement in practical RPVs. At very high dose of 2.2x10 18 n/cm 2 by JMTR, the Cu precipitates were scarcely observed, and the irradiation-induced embrittlement was primarily caused from vacancy-impurity complexes and dislocation loops. (author)

Simplest simulation model for three-dimensional xenon oscillations in large PWRs

International Nuclear Information System (INIS)

Shimazu, Yoichiro

2004-01-01

Xenon oscillations in large PWRs are well understood and there have been no operational problems remained. However, in order to suppress the oscillations effectively, optimal control strategy is preferable. Generally speaking in such optimality search based on the modern control theory, a large volume of transient core analyses is required. For example, three dimensional core calculations are inevitable for the analyses of radial oscillations. From this point of view, a very simple 3-D model is proposed, which is based on a reactor model of only four points. As in the actual reactor operation, the magnitude of xenon oscillations should be limited from the view point of safety, the model further assumes that the neutron leakage can be also small or even constant. It can explicitly use reactor parameters such as reactivity coefficients and control rod worth directly. The model is so simplified as described above that it can predict oscillation behavior in a very short calculation time even on a PC. However the prediction result is good. The validity of the model in comparison with measured data and the applications are discussed. (author)

Application of a two-region kinetic model for reflected reactors to experimental data

International Nuclear Information System (INIS)

Busch, R.D.; Spriggs, G.D.; Williams, J.G.

1996-01-01

Reflected reactors constitute one of the most important classes of nuclear reactors. Yet, during the past 50 yr, a plethora of experimental data involving reflected systems has been reported in the literature that cannot be satisfactorily explained using the open-quotes standardclose quotes (i.e., one-region) point-kinetic model. In particular, many have observed that the prompt-decay a curves obtained from Rossi-α and pulsed-neutron experiments can exhibit multiple decay modes in the vicinity near delayed critical in some types of reflected systems. When analyzed using theories based on the standard point-kinetic model, these data yielded system lifetimes that do not always agree well with the lifetimes predicted by numerical solutions of the multigroup, multidimensional diffusion or transport equations. In several cases, when the longest lived decay mode (i.e., the dominant root) was plotted as a function of reactivity, the a curve intercepted the reactivity axis at a reactivity significantly greater than 1$. Brunson dubbed this seemingly inexplicable behavior as the open-quotes dollar discrepancy.close quotes Furthermore, it has also been observed that the kinetic behavior of some reflected, fast-burst assemblies exhibits a very pronounced nonlinear relationship between reactivity and the initial inverse period for reactivity insertions > 1 $

Three-dimensional printing and pediatric liver disease.

Science.gov (United States)

Alkhouri, Naim; Zein, Nizar N

2016-10-01

Enthusiastic physicians and medical researchers are investigating the role of three-dimensional printing in medicine. The purpose of the current review is to provide a concise summary of the role of three-dimensional printing technology as it relates to the field of pediatric hepatology and liver transplantation. Our group and others have recently demonstrated the feasibility of printing three-dimensional livers with identical anatomical and geometrical landmarks to the native liver to facilitate presurgical planning of complex liver surgeries. Medical educators are exploring the use of three-dimensional printed organs in anatomy classes and surgical residencies. Moreover, mini-livers are being developed by regenerative medicine scientist as a way to test new drugs and, eventually, whole livers will be grown in the laboratory to replace organs with end-stage disease solving the organ shortage problem. From presurgical planning to medical education to ultimately the bioprinting of whole organs for transplantation, three-dimensional printing will change medicine as we know in the next few years.

Three-dimensional studies of the 700 MWe steam generator design

International Nuclear Information System (INIS)

John, B.; Pietralik, J.

2006-01-01

The next stage in the Indian nuclear power programme envisions building 700 MWe Indian Pressurized Heavy Water Reactor (IPHWR) units. This involves up-rating of all the plant equipment including the reactor, steam generators (SGs), turbo-generator, major pumps, etc. The SG used in the current generation of 540 MWe IPHWRs, is a mushroom type, inverted U-tube, natural-circulation SG. The 700 MWe SG is of the same type and has the same tube bundle design and the same heat transfer area. The tube diameter, tube pitch, and outer diameter of the SG sections are the same as for the 540 MWe SG. The geometry of the feedwater header, the flow restrictor in the downcomer and the flow distribution plate are different in the two designs. The changes were required due to a 26% increase in steam flow rate while maintaining the same circulation ratio. This paper describes the design of the 700 MWe SG and a thermalhydraulic analysis using a one-dimensional, in-house code and a three-dimensional code called THIRST developed by AECL. The codes were validated against the 540 MWe SG data. The analysis was made for the 700 MWe SG for two versions: with and without integral preheater. The results of the THIRST runs were used for a flow-induced vibration analysis. The results of the flow-induced vibration analysis show that the vibrations are not excessive. (author)

Visual Interpretation with Three-Dimensional Annotations (VITA): Three-Dimensional Image Interpretation Tool for Radiological Reporting

OpenAIRE

Roy, Sharmili; Brown, Michael S.; Shih, George L.

2013-01-01

This paper introduces a software framework called Visual Interpretation with Three-Dimensional Annotations (VITA) that is able to automatically generate three-dimensional (3D) visual summaries based on radiological annotations made during routine exam reporting. VITA summaries are in the form of rotating 3D volumes where radiological annotations are highlighted to place important clinical observations into a 3D context. The rendered volume is produced as a Digital Imaging and Communications i...

Three-dimensional reconstruction of functional brain images

International Nuclear Information System (INIS)

Inoue, Masato; Shoji, Kazuhiko; Kojima, Hisayoshi; Hirano, Shigeru; Naito, Yasushi; Honjo, Iwao

1999-01-01

We consider PET (positron emission tomography) measurement with SPM (Statistical Parametric Mapping) analysis to be one of the most useful methods to identify activated areas of the brain involved in language processing. SPM is an effective analytical method that detects markedly activated areas over the whole brain. However, with the conventional presentations of these functional brain images, such as horizontal slices, three directional projection, or brain surface coloring, makes understanding and interpreting the positional relationships among various brain areas difficult. Therefore, we developed three-dimensionally reconstructed images from these functional brain images to improve the interpretation. The subjects were 12 normal volunteers. The following three types of images were constructed: routine images by SPM, three-dimensional static images, and three-dimensional dynamic images, after PET images were analyzed by SPM during daily dialog listening. The creation of images of both the three-dimensional static and dynamic types employed the volume rendering method by VTK (The Visualization Toolkit). Since the functional brain images did not include original brain images, we synthesized SPM and MRI brain images by self-made C++ programs. The three-dimensional dynamic images were made by sequencing static images with available software. Images of both the three-dimensional static and dynamic types were processed by a personal computer system. Our newly created images showed clearer positional relationships among activated brain areas compared to the conventional method. To date, functional brain images have been employed in fields such as neurology or neurosurgery, however, these images may be useful even in the field of otorhinolaryngology, to assess hearing and speech. Exact three-dimensional images based on functional brain images are important for exact and intuitive interpretation, and may lead to new developments in brain science. Currently, the surface

Three-dimensional reconstruction of functional brain images

Energy Technology Data Exchange (ETDEWEB)

Inoue, Masato; Shoji, Kazuhiko; Kojima, Hisayoshi; Hirano, Shigeru; Naito, Yasushi; Honjo, Iwao [Kyoto Univ. (Japan)

1999-08-01

We consider PET (positron emission tomography) measurement with SPM (Statistical Parametric Mapping) analysis to be one of the most useful methods to identify activated areas of the brain involved in language processing. SPM is an effective analytical method that detects markedly activated areas over the whole brain. However, with the conventional presentations of these functional brain images, such as horizontal slices, three directional projection, or brain surface coloring, makes understanding and interpreting the positional relationships among various brain areas difficult. Therefore, we developed three-dimensionally reconstructed images from these functional brain images to improve the interpretation. The subjects were 12 normal volunteers. The following three types of images were constructed: routine images by SPM, three-dimensional static images, and three-dimensional dynamic images, after PET images were analyzed by SPM during daily dialog listening. The creation of images of both the three-dimensional static and dynamic types employed the volume rendering method by VTK (The Visualization Toolkit). Since the functional brain images did not include original brain images, we synthesized SPM and MRI brain images by self-made C++ programs. The three-dimensional dynamic images were made by sequencing static images with available software. Images of both the three-dimensional static and dynamic types were processed by a personal computer system. Our newly created images showed clearer positional relationships among activated brain areas compared to the conventional method. To date, functional brain images have been employed in fields such as neurology or neurosurgery, however, these images may be useful even in the field of otorhinolaryngology, to assess hearing and speech. Exact three-dimensional images based on functional brain images are important for exact and intuitive interpretation, and may lead to new developments in brain science. Currently, the surface

Code Coupling for Multi-Dimensional Core Transient Analysis

International Nuclear Information System (INIS)

Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il

2015-01-01

After the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the Doppler effect becomes important and turns the reactivity balance and power down to lower levels. Although this happens in a very short period of time, only few seconds, the energy generated can be very significant and cause fuel failures. The current safety analysis methodology which is based on overly conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KEPCO Nuclear Fuel(KNF) is developing the multi-dimensional safety analysis methodology to mitigate the consequences of the single CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST, which have transient calculation performance, were coupled using message passing interface (MPI). This paper presents the methodology used for code coupling and the preliminary simulation results with the coupled code system (CHASER). Multi-dimensional core transient analysis code system, CHASER, has been developed and it was applied to simulate a single CEA ejection accident. CHASER gave a good prediction of multi-dimensional core transient behaviors during transient. In the near future, the multi-dimension CEA ejection analysis methodology using CHASER is planning to be developed. CHASER is expected to be a useful tool to gain safety margin for reactivity initiated accidents (RIAs), such as a single CEA ejection accident

Code Coupling for Multi-Dimensional Core Transient Analysis

Energy Technology Data Exchange (ETDEWEB)

Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il [KEPCO NF, Daejeon (Korea, Republic of)

2015-05-15

After the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the Doppler effect becomes important and turns the reactivity balance and power down to lower levels. Although this happens in a very short period of time, only few seconds, the energy generated can be very significant and cause fuel failures. The current safety analysis methodology which is based on overly conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KEPCO Nuclear Fuel(KNF) is developing the multi-dimensional safety analysis methodology to mitigate the consequences of the single CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST, which have transient calculation performance, were coupled using message passing interface (MPI). This paper presents the methodology used for code coupling and the preliminary simulation results with the coupled code system (CHASER). Multi-dimensional core transient analysis code system, CHASER, has been developed and it was applied to simulate a single CEA ejection accident. CHASER gave a good prediction of multi-dimensional core transient behaviors during transient. In the near future, the multi-dimension CEA ejection analysis methodology using CHASER is planning to be developed. CHASER is expected to be a useful tool to gain safety margin for reactivity initiated accidents (RIAs), such as a single CEA ejection accident.

A new formulation for the importance function in the kinetics of subcritical reactors

International Nuclear Information System (INIS)

Silva, Cristiano da; Senra Martinez, Aquilino; Carvalho da Silva, Fernando

2012-01-01

Highlights: ► In this paper we propose a new formulation for the importance function in the kinetics of subcritical systems. ► We analyze the relevance of an external neutron source for the subcritical interval 0.95 eff eff is the multiplication factor according to the physical properties of the nuclear reactor. For the purposes of validation of the proposed method we will use, as a reference method, the expansion in modes of the time-dependent neutron flux for the solution of the onedimensional diffusion equation. It will be presented results that demonstrate the precision of the proposed method when compared to the conventional point kinetic equations. The results show that the new point kinetic equations are rather precise in the subcriticality range considered.

Towards Free-Form Kinetic Structures

DEFF Research Database (Denmark)

Parigi, Dario; Kirkegaard, Poul Henning

2012-01-01

of pin-slot paths starting from the local displacements of element [2] [3]. In the design of kinetic structures, in particular when complex three dimensional and non regular configurations are involved, the functionality is frequently related to a global displacement capability of the assembly rather...... for the generation of free-form kinetic structures....

Advanced particle-in-cell simulation techniques for modeling the Lockheed Martin Compact Fusion Reactor

Science.gov (United States)

Welch, Dale; Font, Gabriel; Mitchell, Robert; Rose, David

2017-10-01

We report on particle-in-cell developments of the study of the Compact Fusion Reactor. Millisecond, two and three-dimensional simulations (cubic meter volume) of confinement and neutral beam heating of the magnetic confinement device requires accurate representation of the complex orbits, near perfect energy conservation, and significant computational power. In order to determine initial plasma fill and neutral beam heating, these simulations include ionization, elastic and charge exchange hydrogen reactions. To this end, we are pursuing fast electromagnetic kinetic modeling algorithms including a two implicit techniques and a hybrid quasi-neutral algorithm with kinetic ions. The kinetic modeling includes use of the Poisson-corrected direct implicit, magnetic implicit, as well as second-order cloud-in-cell techniques. The hybrid algorithm, ignoring electron inertial effects, is two orders of magnitude faster than kinetic but not as accurate with respect to confinement. The advantages and disadvantages of these techniques will be presented. Funded by Lockheed Martin.

Application of Simulated Three Dimensional CT Image in Orthognathic Surgery

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyun Don; Park, Chang Seo [Dept. of Dental Radiology, College of Dentistry, Yensei University, Seoul (Korea, Republic of); Yoo, Sun Kook; Lee, Kyoung Sang [Dept. of Medical Engineering, College of Medicine, Yensei University, Seoul (Korea, Republic of)

1998-08-15

In orthodontics and orthognathic surgery, cephalogram has been routine practice in diagnosis and treatment evaluation of craniofacial deformity. But its inherent distortion of actual length and angles during projecting three dimensional object to two dimensional plane might cause errors in quantitative analysis of shape and size. Therefore, it is desirable that three dimensional object is diagnosed and evaluated three dimensionally and three dimensional CT image is best for three dimensional analysis. Development of clinic necessitates evaluation of result of treatment and comparison before and after surgery. It is desirable that patient that was diagnosed and planned by three dimensional computed tomography before surgery is evaluated by three dimensional computed tomography after surgery, too. But Because there is no standardized normal values in three dimension now and three dimensional Computed Tomography needs expensive equipment and because of its expenses and amount of exposure to radiation, limitations still remain to be solved in its application to routine practice. If postoperative three dimensional image is constructed by pre and postoperative lateral and postero-anterior cephalograms and preoperative three dimensional computed tomogram, pre and postoperative image will be compared and evaluated three dimensionally without three dimensional computed tomography after surgery and that will contribute to standardize normal values in three dimension. This study introduced new method that computer-simulated three dimensional image was constructed by preoperative three dimensional computed tomogram and pre and postoperative lateral and postero-anterior cephalograms, and for validation of new method, in four cases of dry skull that position of mandible was displaced and four patients of orthognathic surgery, computer-simulated three dimensional image and actual postoperative three dimensional image were compared. The results were as follows. 1. In four cases of

Application of Simulated Three Dimensional CT Image in Orthognathic Surgery

International Nuclear Information System (INIS)

Kim, Hyun Don; Park, Chang Seo; Yoo, Sun Kook; Lee, Kyoung Sang

1998-01-01

In orthodontics and orthognathic surgery, cephalogram has been routine practice in diagnosis and treatment evaluation of craniofacial deformity. But its inherent distortion of actual length and angles during projecting three dimensional object to two dimensional plane might cause errors in quantitative analysis of shape and size. Therefore, it is desirable that three dimensional object is diagnosed and evaluated three dimensionally and three dimensional CT image is best for three dimensional analysis. Development of clinic necessitates evaluation of result of treatment and comparison before and after surgery. It is desirable that patient that was diagnosed and planned by three dimensional computed tomography before surgery is evaluated by three dimensional computed tomography after surgery, too. But Because there is no standardized normal values in three dimension now and three dimensional Computed Tomography needs expensive equipment and because of its expenses and amount of exposure to radiation, limitations still remain to be solved in its application to routine practice. If postoperative three dimensional image is constructed by pre and postoperative lateral and postero-anterior cephalograms and preoperative three dimensional computed tomogram, pre and postoperative image will be compared and evaluated three dimensionally without three dimensional computed tomography after surgery and that will contribute to standardize normal values in three dimension. This study introduced new method that computer-simulated three dimensional image was constructed by preoperative three dimensional computed tomogram and pre and postoperative lateral and postero-anterior cephalograms, and for validation of new method, in four cases of dry skull that position of mandible was displaced and four patients of orthognathic surgery, computer-simulated three dimensional image and actual postoperative three dimensional image were compared. The results were as follows. 1. In four cases of

Magnetic field generation by pointwise zero-helicity three-dimensional steady flow of an incompressible electrically conducting fluid

Science.gov (United States)

Rasskazov, Andrey; Chertovskih, Roman; Zheligovsky, Vladislav

2018-04-01

We introduce six families of three-dimensional space-periodic steady solenoidal flows, whose kinetic helicity density is zero at any point. Four families are analytically defined. Flows in four families have zero helicity spectrum. Sample flows from five families are used to demonstrate numerically that neither zero kinetic helicity density nor zero helicity spectrum prohibit generation of large-scale magnetic field by the two most prominent dynamo mechanisms: the magnetic α -effect and negative eddy diffusivity. Our computations also attest that such flows often generate small-scale field for sufficiently small magnetic molecular diffusivity. These findings indicate that kinetic helicity and helicity spectrum are not the quantities controlling the dynamo properties of a flow regardless of whether scale separation is present or not.

Application of finite-element method to three-dimensional nuclear reactor analysis

International Nuclear Information System (INIS)

Cheung, K.Y.

1985-01-01

The application of the finite element method to solve a realistic one-or-two energy group, multiregion, three-dimensional static neutron diffusion problem is studied. Linear, quadratic, and cubic serendipity box-shape elements are used. The resulting sets of simultaneous algebraic equations with thousands of unknowns are solved by the conjugate gradient method, without forming the large coefficient matrix explicitly. This avoids the complicated data management schemes to store such a large coefficient matrix. Three finite-element computer programs: FEM-LINEAR, FEM-QUADRATIC and FEM-CUBIC were developed, using the linear, quadratic, and cubic box-shape elements respectively. They are self-contained, using simple nodal labeling schemes, without the need for separate finite element mesh generating routines. The efficiency and accuracy of these computer programs are then compared among themselves, and with other computer codes. The cubic element model is not recommended for practical usage because it gives almost identical results as the quadratic model, but it requires considerably longer computation time. The linear model is less accurate than the quadratic model, but it requires much shorter computation time. For a large 3-D problem, the linear model is to be preferred since it gives acceptable accuracy. The quadratic model may be used if improved accuracy is desired

Reaction kinetics in open reactors and serial transfers between closed reactors

Science.gov (United States)

Blokhuis, Alex; Lacoste, David; Gaspard, Pierre

2018-04-01

Kinetic theory and thermodynamics of reaction networks are extended to the out-of-equilibrium dynamics of continuous-flow stirred tank reactors (CSTR) and serial transfers. On the basis of their stoichiometry matrix, the conservation laws and the cycles of the network are determined for both dynamics. It is shown that the CSTR and serial transfer dynamics are equivalent in the limit where the time interval between the transfers tends to zero proportionally to the ratio of the fractions of fresh to transferred solutions. These results are illustrated with a finite cross-catalytic reaction network and an infinite reaction network describing mass exchange between polymers. Serial transfer dynamics is typically used in molecular evolution experiments in the context of research on the origins of life. The present study is shedding a new light on the role played by serial transfer parameters in these experiments.

Unsteady single-phase natural circulation flow mixing prediction using CATHARE three-dimensional capabilities

Energy Technology Data Exchange (ETDEWEB)

Salah, Anis Bousbia; Vlassenbroeck, Jacques [Bel V - Subsidiary of the Belgian Federal Agency for Nuclear Contro, Brussels (Belize)

2017-04-15

Coolant mixing under natural circulation flow regime constitutes a key parameter that may play a role in the course of an accidental transient in a nuclear pressurized water reactor. This issue has motivated some experimental investigations carried out within the OECD/NEA PKL projects. The aim was to assess the coolant mixing phenomenon in the reactor pressure vessel downcomer and the core lower plenum under several asymmetric steady and unsteady flow conditions, and to provide experimental data for code validations. Former studies addressed the mixing phenomenon using, on the one hand, one-dimensional computational approaches with cross flows that are not fully validated under transient conditions and, on the other hand, expensive computational fluid dynamic tools that are not always justified for large-scale macroscopic phenomena. In the current framework, an unsteady coolant mixing experiment carried out in the Rossendorf coolant mixing test facility is simulated using the three-dimensional porous media capabilities of the thermal–hydraulic system CATHARE code. The current study allows highlighting the current capabilities of these codes and their suitability for reproducing the main phenomena occurring during asymmetric transient natural circulation mixing conditions.

Effect of ultrasound on flotation kinetics in the reactor-separator

International Nuclear Information System (INIS)

Filippov, L O; Matinin, A S; Samiguin, V D; Filippova, I V

2013-01-01

Effect of the ultrasound on flotation kinetics in reactor-separator has been studied for chalcopyrite/quartz mix mineral system. Under ultrasound treatment, recovery of chalcopyrite into bulk concentrate is higher than that at reagent-only treatment. It can be explained by increased of flotation rate for slow fraction as defined by Kelsall model. The slow fraction flotation rate increase multiplied by 6 vs. ultrasound treatment. Additional effect of the ultrasound treatment has been noticed under conditions when gangue minerals detachment from bubbles can be controlled. Reactor-separator has advantages over other types of flotation cells for this purpose providing a special zone for the ultrasound treatment that can be easily designed in this impeller less machine. The ultrasound influence on particles collision probability is able to explain of chalcopyrite recovery increase in the concentrate and activation chalcopyrite particles flotation.

Effect of ultrasound on flotation kinetics in the reactor-separator

Science.gov (United States)

Filippov, L. O.; Matinin, A. S.; Samiguin, V. D.; Filippova, I. V.

2013-03-01

Effect of the ultrasound on flotation kinetics in reactor-separator has been studied for chalcopyrite/quartz mix mineral system. Under ultrasound treatment, recovery of chalcopyrite into bulk concentrate is higher than that at reagent-only treatment. It can be explained by increased of flotation rate for slow fraction as defined by Kelsall model. The slow fraction flotation rate increase multiplied by 6 vs. ultrasound treatment. Additional effect of the ultrasound treatment has been noticed under conditions when gangue minerals detachment from bubbles can be controlled. Reactor-separator has advantages over other types of flotation cells for this purpose providing a special zone for the ultrasound treatment that can be easily designed in this impeller less machine. The ultrasound influence on particles collision probability is able to explain of chalcopyrite recovery increase in the concentrate and activation chalcopyrite particles flotation.

Comparison of BWR-6 pressurization transients with one-dimensional and point kinetics

International Nuclear Information System (INIS)

Serra, J.M.; Mata, P.; Cronin, J.T.

1992-01-01

This paper focuses on the differences between the results of core reload licensing calculations for the BWR-6 plant when performed with a one-dimensional (1-D) versus a point kinetics model. More specifically, the improvement in critical power ratio which would be expected from a change in methods from a point to a 1-D kinetics core wide transient calculation for pressurization transients is investigated. To qualitatively assess critical power ratio (CPR) improvement, core wide transient and hot channel calculations of a generator load rejection with failure of the steam by-pass system and a feedwater controller failure of maximum demand are performed with both, point and 1-D kinetics models in the core wide simulation. Additionally, a sensitivity study on the frequency of power shape function updating in the 1-D kinetics calculation is performed

THE MATHEMATICAL MODEL DEVELOPMENT OF THE ETHYLBENZENE DEHYDROGENATION PROCESS KINETICS IN A TWO-STAGE ADIABATIC CONTINUOUS REACTOR

Directory of Open Access Journals (Sweden)

V. K. Bityukov

2015-01-01

Full Text Available The article is devoted to the mathematical modeling of the kinetics of ethyl benzene dehydrogenation in a two-stage adiabatic reactor with a catalytic bed functioning on continuous technology. The analysis of chemical reactions taking place parallel to the main reaction of styrene formation has been carried out on the basis of which a number of assumptions were made proceeding from which a kinetic scheme describing the mechanism of the chemical reactions during the dehydrogenation process was developed. A mathematical model of the dehydrogenation process, describing the dynamics of chemical reactions taking place in each of the two stages of the reactor block at a constant temperature is developed. The estimation of the rate constants of direct and reverse reactions of each component, formation and exhaustion of the reacted mixture was made. The dynamics of the starting material concentration variations (ethyl benzene batch was obtained as well as styrene formation dynamics and all byproducts of dehydrogenation (benzene, toluene, ethylene, carbon, hydrogen, ect.. The calculated the variations of the component composition of the reaction mixture during its passage through the first and second stages of the reactor showed that the proposed mathematical description adequately reproduces the kinetics of the process under investigation. This demonstrates the advantage of the developed model, as well as loyalty to the values found for the rate constants of reactions, which enable the use of models for calculating the kinetics of ethyl benzene dehydrogenation under nonisothermal mode in order to determine the optimal temperature trajectory of the reactor operation. In the future, it will reduce energy and resource consumption, increase the volume of produced styrene and improve the economic indexes of the process.

A three-dimensional analyses of fluid flow and heat transfer for moderator integrity assessment in PHWR

International Nuclear Information System (INIS)

Bang, K. H.; Lee, J. Y.; Yoo, S. O.; Kim, M. W.; Kim, H. J.

2002-01-01

Three-dimensional analyses of fluid flow and heat transfer has been performed in this study. The simulation of SPEL experimental work and comparison with experimental data has been carried out to verify the analyses models. Moreover, to verify the CANDU-6 reactor type, analyses of fluid flow and heat transfer in the calandria under the condition of steady state has been performed using FLUENT code, which is the conventional code for a three-dimensional analyses of fluid flow and heat transfer for moderator integrity assessment in PHWR thermal-hydraulics. It is found that the maximum temperature in the moderator is 347K (74 ), so that the moderator has the enough subcoolability to ensure the integrity of pressure tube during LOCA conditions

Three-dimensional low-energy topological invariants

International Nuclear Information System (INIS)

Bakalarska, M.; Broda, B.

2000-01-01

A description of the one-loop approximation formula for the partition function of a three-dimensional abelian version of the Donaldson-Witten theory is proposed. The one-loop expression is shown to contain such topological invariants of a three-dimensional manifold M like the Reidemeister-Ray-Singer torsion τ R and Betti numbers. (orig.)

Physical modeling and numerical simulation of subcooled boiling in one- and three-dimensional representation of bundle geometry

International Nuclear Information System (INIS)

Bottoni, M.; Lyczkowski, R.; Ahuja, S.

1995-01-01

Numerical simulation of subcooled boiling in one-dimensional geometry with the Homogeneous Equilibrium Model (HEM) may yield difficulties related to the very low sonic velocity associated with the HEM. These difficulties do not arise with subcritical flow. Possible solutions of the problem include introducing a relaxation of the vapor production rate. Three-dimensional simulations of subcooled boiling in bundle geometry typical of fast reactors can be performed by using two systems of conservation equations, one for the HEM and the other for a Separated Phases Model (SPM), with a smooth transition between the two models

A Monte Carlo Green's function method for three-dimensional neutron transport

International Nuclear Information System (INIS)

Gamino, R.G.; Brown, F.B.; Mendelson, M.R.

1992-01-01

This paper describes a Monte Carlo transport kernel capability, which has recently been incorporated into the RACER continuous-energy Monte Carlo code. The kernels represent a Green's function method for neutron transport from a fixed-source volume out to a particular volume of interest. This method is very powerful transport technique. Also, since kernels are evaluated numerically by Monte Carlo, the problem geometry can be arbitrarily complex, yet exact. This method is intended for problems where an ex-core neutron response must be determined for a variety of reactor conditions. Two examples are ex-core neutron detector response and vessel critical weld fast flux. The response is expressed in terms of neutron transport kernels weighted by a core fission source distribution. In these types of calculations, the response must be computed for hundreds of source distributions, but the kernels only need to be calculated once. The advance described in this paper is that the kernels are generated with a highly accurate three-dimensional Monte Carlo transport calculation instead of an approximate method such as line-of-sight attenuation theory or a synthesized three-dimensional discrete ordinates solution

Three-dimensional reactor model for the Paks NPP full-scope simulator

International Nuclear Information System (INIS)

Gyori, C.; Hegyi, G.; Hozer, Z.; Kereszturi, A.; Maraczy, C.

1993-01-01

The reactor model includes thermohydraulic and neutron-physical components. The thermohydraulic model is based on the SMABRE code developed at the Technical Research Centre of Finland for the analysis of loss-of-coolant transients in PWRs. The fuel rod model will be replaced by a new software module providing a comprehensive description of the behavior of fuel rods during reactor transients and hypothetical accidents. The calculation is performed in four individual models: fuel rod temperature model, fuel rod internal pressure model, fuel rod deformation model and fuel rod failure model. In the neutron-physical model the core is calculated with nodes for all of the 349 fuel assemblies, and each assembly is calculated in ten layers. (Z.S.) 1 fig., 5 refs

Development of MARS for multi-dimensional and multi-purpose thermal-hydraulic system analysis

Energy Technology Data Exchange (ETDEWEB)

Lee, Won Jae; Chung, Bub Dong; Kim, Kyung Doo; Hwang, Moon Kyu; Jeong, Jae Jun; Ha, Kwi Seok; Joo, Han Gyu [Korea Atomic Energy Research Institute, T/H Safety Research Team, Yusung, Daejeon (Korea)

2000-10-01

MARS (Multi-dimensional Analysis of Reactor Safety) code is being developed by KAERI for the realistic thermal-hydraulic simulation of light water reactor system transients. MARS 1.4 has been developed as a final version of basic code frame for the multi-dimensional analysis of system thermal-hydraulics. Since MARS 1.3, MARS 1.4 has been improved to have the enhanced code capability and user friendliness through the unification of input/output features, code models and code functions, and through the code modernization. Further improvements of thermal-hydraulic models, numerical method and user friendliness are being carried out for the enhanced code accuracy. As a multi-purpose safety analysis code system, a coupled analysis system, MARS/MASTER/CONTEMPT, has been developed using multiple DLL (Dynamic Link Library) techniques of Windows system. This code system enables the coupled, that is, more realistic analysis of multi-dimensional thermal-hydraulics (MARS 2.0), three-dimensional core kinetics (MASTER) and containment thermal-hydraulics (CONTEMPT). This paper discusses the MARS development program, and the developmental progress of the MARS 1.4 and the MARS/MASTER/CONTEMPT focusing on major features of the codes and their verification. It also discusses thermal hydraulic models and new code features under development. (author)

Development of MARS for multi-dimensional and multi-purpose thermal-hydraulic system analysis

International Nuclear Information System (INIS)

Lee, Won Jae; Chung, Bub Dong; Kim, Kyung Doo; Hwang, Moon Kyu; Jeong, Jae Jun; Ha, Kwi Seok; Joo, Han Gyu

2000-01-01

MARS (Multi-dimensional Analysis of Reactor Safety) code is being developed by KAERI for the realistic thermal-hydraulic simulation of light water reactor system transients. MARS 1.4 has been developed as a final version of basic code frame for the multi-dimensional analysis of system thermal-hydraulics. Since MARS 1.3, MARS 1.4 has been improved to have the enhanced code capability and user friendliness through the unification of input/output features, code models and code functions, and through the code modernization. Further improvements of thermal-hydraulic models, numerical method and user friendliness are being carried out for the enhanced code accuracy. As a multi-purpose safety analysis code system, a coupled analysis system, MARS/MASTER/CONTEMPT, has been developed using multiple DLL (Dynamic Link Library) techniques of Windows system. This code system enables the coupled, that is, more realistic analysis of multi-dimensional thermal-hydraulics (MARS 2.0), three-dimensional core kinetics (MASTER) and containment thermal-hydraulics (CONTEMPT). This paper discusses the MARS development program, and the developmental progress of the MARS 1.4 and the MARS/MASTER/CONTEMPT focusing on major features of the codes and their verification. It also discusses thermal hydraulic models and new code features under development. (author)

Analytical solution of point kinetics equations for linear reactivity variation during the start-up of a nuclear reactor

International Nuclear Information System (INIS)

Palma, Daniel A.P.; Martinez, Aquilino S.; Goncalves, Alessandro C.

2009-01-01

The analytical solution of point kinetics equations with a group of delayed neutrons is useful in predicting the variation of neutron density during the start-up of a nuclear reactor. In the practical case of an increase of nuclear reactor power resulting from the linear insertion of reactivity, the exact analytical solution cannot be obtained. Approximate solutions have been obtained in previous articles, based on considerations that need to be verifiable in practice. In the present article, an alternative analytic solution is presented for point kinetics equations in which the only approximation consists of disregarding the term of the second derivative for neutron density in relation to time. The results proved satisfactory when applied to practical situations in the start-up of a nuclear reactor through the control rods withdraw.

Analytical solution of point kinetics equations for linear reactivity variation during the start-up of a nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Palma, Daniel A.P. [CEFET QUIMICA de Nilopolis/RJ, 21941-914 Rio de Janeiro (Brazil)], E-mail: agoncalves@con.ufrj.br; Martinez, Aquilino S.; Goncalves, Alessandro C. [COPPE/UFRJ - Programa de Engenharia Nuclear, Rio de Janeiro (Brazil)

2009-09-15

The analytical solution of point kinetics equations with a group of delayed neutrons is useful in predicting the variation of neutron density during the start-up of a nuclear reactor. In the practical case of an increase of nuclear reactor power resulting from the linear insertion of reactivity, the exact analytical solution cannot be obtained. Approximate solutions have been obtained in previous articles, based on considerations that need to be verifiable in practice. In the present article, an alternative analytic solution is presented for point kinetics equations in which the only approximation consists of disregarding the term of the second derivative for neutron density in relation to time. The results proved satisfactory when applied to practical situations in the start-up of a nuclear reactor through the control rods withdraw.

Modeling of kinetics of Cr(VI) sorption onto grape stalk waste in a stirred batch reactor

International Nuclear Information System (INIS)

Escudero, Carlos; Fiol, Nuria; Poch, Jordi; Villaescusa, Isabel

2009-01-01

Recently, Cr(VI) removal by grape stalks has been postulated to follow two mechanisms, adsorption and reduction to trivalent chromium. Nevertheless, the rate at which both processes take place and the possible simultaneity of both processes has not been investigated. In this work, kinetics of Cr(VI) sorption onto grape stalk waste has been studied. Experiments were carried out at different temperatures but at a constant pH (3 ± 0.1) in a stirred batch reactor. Results showed that three steps take place in the process of Cr(VI) sorption onto grape stalk waste: Cr(VI) sorption, Cr(VI) reduction to Cr(III) and the adsorption of the formed Cr(III). Taking into account the evidences above mentioned, a model has been developed to predict Cr(VI) sorption on grape stalks on the basis of (i) irreversible reduction of Cr(VI) to Cr(III) reaction, whose reaction rate is assumed to be proportional to the Cr(VI) concentration in solution and (ii) adsorption and desorption of Cr(VI) and formed Cr(III) assuming that all the processes follow Langmuir type kinetics. The proposed model fits successfully the kinetic data obtained at different temperatures and describes the kinetics profile of total, hexavalent and trivalent chromium. The proposed model would be helpful for researchers in the field of Cr(VI) biosorption to design and predict the performance of sorption processes.

[Bone drilling simulation by three-dimensional imaging].

Science.gov (United States)

Suto, Y; Furuhata, K; Kojima, T; Kurokawa, T; Kobayashi, M

1989-06-01

The three-dimensional display technique has a wide range of medical applications. Pre-operative planning is one typical application: in orthopedic surgery, three-dimensional image processing has been used very successfully. We have employed this technique in pre-operative planning for orthopedic surgery, and have developed a simulation system for bone-drilling. Positive results were obtained by pre-operative rehearsal; when a region of interest is indicated by means of a mouse on the three-dimensional image displayed on the CRT, the corresponding region appears on the slice image which is displayed simultaneously. Consequently, the status of the bone-drilling is constantly monitored. In developing this system, we have placed emphasis on the quality of the reconstructed three-dimensional images, on fast processing, and on the easy operation of the surgical planning simulation.

Three-Dimensional Printing Surgical Applications.

Science.gov (United States)

AlAli, Ahmad B; Griffin, Michelle F; Butler, Peter E

2015-01-01

Three-dimensional printing, a technology used for decades in the industrial field, gains a lot of attention in the medical field for its potential benefits. With advancement of desktop printers, this technology is accessible and a lot of research is going on in the medical field. To evaluate its application in surgical field, which may include but not limited to surgical planning, surgical education, implants, and prosthesis, which are the focus of this review. Research was conducted by searching PubMed, Web of science, and other reliable sources. We included original articles and excluded articles based on animals, those more than 10 years old, and those not in English. These articles were evaluated, and relevant studies were included in this review. Three-dimensional printing shows a potential benefit in surgical application. Printed implants were used in patient in a few cases and show successful results; however, longer follow-up and more trials are needed. Surgical and medical education is believed to be more efficient with this technology than the current practice. Printed surgical instrument and surgical planning are also believed to improve with three-dimensional printing. Three-dimensional printing can be a very powerful tool in the near future, which can aid the medical field that is facing a lot of challenges and obstacles. However, despite the reported results, further research on larger samples and analytical measurements should be conducted to ensure this technology's impact on the practice.

Simulation of radiation effects on three-dimensional computer optical memories

International Nuclear Information System (INIS)

Moscovitch, M.; Emfietzoglou, D.

1997-01-01

A model was developed to simulate the effects of heavy charged-particle (HCP) radiation on the information stored in three-dimensional computer optical memories. The model is based on (i) the HCP track radial dose distribution, (ii) the spatial and temporal distribution of temperature in the track, (iii) the matrix-specific radiation-induced changes that will affect the response, and (iv) the kinetics of transition of photochromic molecules from the colored to the colorless isomeric form (bit flip). It is shown that information stored in a volume of several nanometers radius around the particle close-quote s track axis may be lost. The magnitude of the effect is dependent on the particle close-quote s track structure. copyright 1997 American Institute of Physics

A two-dimensional kinetic model of the scrape-off layer

International Nuclear Information System (INIS)

Catto, P.J.; Hazeltine, R.D.

1993-09-01

A two-dimensional (radius and poloidal angle), analytically tractable kinetic model of the ion (or energetic electron) behavior in the scrape-off layer of a limiter or divertor plasma in a tokamak is presented. The model determines the boundary conditions on the core ion density and ion temperature gradients, the power load on the limiter or divertor plates, the energy carried per particle to the walls, and the effective flux limit. The self-consistent electrostatic potential in the quasi-neutral scrape-off layer is determined by using the ion kinetic model of the layer along with a Maxwell-Boltzmann electron response that occurs because most electrons are reflected by the Debye sheaths (assumed to be infinitely thin) at the limiter or divertor plates

The Three-dimensional Digital Factory for Shipbuilding Technology Research

Directory of Open Access Journals (Sweden)

Xu Wei

2016-01-01

Full Text Available The three-dimensional digital factory technology research is the hotspot in shipbuilding recently. The three-dimensional digital factory technology not only focus on design the components of the product, but also discuss on the simulation and analyses of the production process.Based on the three-dimensional model, the basic data layer, application control layer and the presentation layer of hierarchical structure are established in the three-dimensional digital factory of shipbuilding in this paper. And the key technologies of three-dimensional digital factory of shipbuilding are analysed. Finally, a case study is applied and the results show that the three-dimensional digital factory will play an important role in the future.

HIDENEK: an implicit particle simulation of kinetic-MHD phenomena in three-dimensional plasmas

International Nuclear Information System (INIS)

Tanaka, Motohiko.

1993-05-01

An advanced 'kinetic-MHD' simulation method and its applications to plasma physics are given in this lecture. This method is quite suitable for studying strong nonlinear, kinetic processes associated with large space-scale, low-frequency electromagnetic phenomena of plasmas. A full set of the Maxwell equations, and the Newton-Lorentz equations of motion for particle ions and guiding-center electrons are adopted. In order to retain only the low-frequency waves and instabilities, implicit particle-field equations are derived. The present implicit-particle method is proved to reproduce the MHD eigenmodes such as Alfven, magnetosonic and kinetic Alfven waves in a thermally near-equilibrium plasma. In the second part of the lecture, several physics applications are shown. These include not only the growth of the instabilities of beam ions against the background plasmas and helical kink of the current, but they also demonstrate nonlinear results such as pitch-angle scattering of the ions. Recent progress in the simulation of the Kelvin-Helmholtz instability is also presented with a special emphasis on the mixing of plasma particles. (author)

Breeder reactors

International Nuclear Information System (INIS)

Gollion, H.

1977-01-01

The reasons for the development of fast reactors are briefly reviewed (a propitious neutron balance oriented towards a maximum uranium burnup) and its special requirements (cooling, fissile material density and reprocessing) discussed. The three stages in the French program of fast reactor development are outlined with Rapsodie at Cadarache, Phenix at Marcoule, and Super Phenix at Creys-Malville. The more specific features of the program of research and development are emphasized: kinetics and the core, the fuel and the components [fr

Towards three-dimensional optical metamaterials

Science.gov (United States)

Tanaka, Takuo; Ishikawa, Atsushi

2017-12-01

Metamaterials have opened up the possibility of unprecedented and fascinating concepts and applications in optics and photonics. Examples include negative refraction, perfect lenses, cloaking, perfect absorbers, and so on. Since these metamaterials are man-made materials composed of sub-wavelength structures, their development strongly depends on the advancement of micro- and nano-fabrication technologies. In particular, the realization of three-dimensional metamaterials is one of the big challenges in this research field. In this review, we describe recent progress in the fabrication technologies for three-dimensional metamaterials, as well as proposed applications.

Photocatalytic mineralization of commercial herbicides in a pilot-scale solar CPC reactor: photoreactor modeling and reaction kinetics constants independent of radiation field.

Science.gov (United States)

Colina-Márquez, Jose; Machuca-Martínez, Fiderman; Li Puma, Gianluca

2009-12-01

The six-flux absorption-scattering model (SFM) of the radiation field in the photoreactor, combined with reaction kinetics and fluid-dynamic models, has proved to be suitable to describe the degradation of water pollutants in heterogeneous photocatalytic reactors, combining simplicity and accuracy. In this study, the above approach was extended to model the photocatalytic mineralization of a commercial herbicides mixture (2,4-D, diuron, and ametryne used in Colombian sugar cane crops) in a solar, pilot-scale, compound parabolic collector (CPC) photoreactor using a slurry suspension of TiO(2). The ray-tracing technique was used jointly with the SFM to determine the direction of both the direct and diffuse solar photon fluxes and the spatial profile of the local volumetric rate of photon absorption (LVRPA) in the CPC reactor. Herbicides mineralization kinetics with explicit photon absorption effects were utilized to remove the dependence of the observed rate constants from the reactor geometry and radiation field in the photoreactor. The results showed that the overall model fitted the experimental data of herbicides mineralization in the solar CPC reactor satisfactorily for both cloudy and sunny days. Using the above approach kinetic parameters independent of the radiation field in the reactor can be estimated directly from the results of experiments carried out in a solar CPC reactor. The SFM combined with reaction kinetics and fluid-dynamic models proved to be a simple, but reliable model, for solar photocatalytic applications.

Three-dimensional imaging modalities in endodontics

Science.gov (United States)

Mao, Teresa

2014-01-01

Recent research in endodontics has highlighted the need for three-dimensional imaging in the clinical arena as well as in research. Three-dimensional imaging using computed tomography (CT) has been used in endodontics over the past decade. Three types of CT scans have been studied in endodontics, namely cone-beam CT, spiral CT, and peripheral quantitative CT. Contemporary endodontics places an emphasis on the use of cone-beam CT for an accurate diagnosis of parameters that cannot be visualized on a two-dimensional image. This review discusses the role of CT in endodontics, pertaining to its importance in the diagnosis of root canal anatomy, detection of peri-radicular lesions, diagnosis of trauma and resorption, presurgical assessment, and evaluation of the treatment outcome. PMID:25279337

Three-dimensional imaging modalities in endodontics

Energy Technology Data Exchange (ETDEWEB)

Mao, Teresa; Neelakantan, Prasanna [Dept. of Conservative Dentistry and Endodontics, Saveetha Dental College and Hospitals, Saveetha University, Chennai (India)

2014-09-15

Recent research in endodontics has highlighted the need for three-dimensional imaging in the clinical arena as well as in research. Three-dimensional imaging using computed tomography (CT) has been used in endodontics over the past decade. Three types of CT scans have been studied in endodontics, namely cone-beam CT, spiral CT, and peripheral quantitative CT. Contemporary endodontics places an emphasis on the use of cone-beam CT for an accurate diagnosis of parameters that cannot be visualized on a two-dimensional image. This review discusses the role of CT in endodontics, pertaining to its importance in the diagnosis of root canal anatomy, detection of peri-radicular lesions, diagnosis of trauma and resorption, presurgical assessment, and evaluation of the treatment outcome.

Three-dimensional imaging modalities in endodontics

International Nuclear Information System (INIS)

Mao, Teresa; Neelakantan, Prasanna

2014-01-01

Recent research in endodontics has highlighted the need for three-dimensional imaging in the clinical arena as well as in research. Three-dimensional imaging using computed tomography (CT) has been used in endodontics over the past decade. Three types of CT scans have been studied in endodontics, namely cone-beam CT, spiral CT, and peripheral quantitative CT. Contemporary endodontics places an emphasis on the use of cone-beam CT for an accurate diagnosis of parameters that cannot be visualized on a two-dimensional image. This review discusses the role of CT in endodontics, pertaining to its importance in the diagnosis of root canal anatomy, detection of peri-radicular lesions, diagnosis of trauma and resorption, presurgical assessment, and evaluation of the treatment outcome

A three-dimensional field solutions of Halbach

International Nuclear Information System (INIS)

Chen Jizhong; Xiao Jijun; Zhang Yiming; Xu Chunyan

2008-01-01

A three-dimensional field solutions are presented for Halback cylinder magnet. Based on Ampere equivalent current methods, the permanent magnets are taken as distributing of current density. For getting the three-dimensional field solution of ideal polarized permanent magnets, the solution method entails the use of the vector potential and involves the closed-form integration of the free-space Green's function. The programmed field solution are ideal for performing rapid parametric studies of the dipole Halback cylinder magnets made from rare earth materials. The field solutions are verified by both an analytical two-dimensional algorithm and three-dimensional finite element software. A rapid method is presented for extensive analyzing and optimizing Halbach cylinder magnet. (authors)

Catalytic wet oxidation of phenol in a trickle bed reactor over a Pt/TiO2 catalyst.

Science.gov (United States)

Maugans, Clayton B; Akgerman, Aydin

2003-01-01

Catalytic wet oxidation of phenol was studied in a batch and a trickle bed reactor using 4.45% Pt/TiO2 catalyst in the temperature range 150-205 degrees C. Kinetic data were obtained from batch reactor studies and used to model the reaction kinetics for phenol disappearance and for total organic carbon disappearance. Trickle bed experiments were then performed to generate data from a heterogeneous flow reactor. Catalyst deactivation was observed in the trickle bed reactor, although the exact cause was not determined. Deactivation was observed to linearly increase with the cumulative amount of phenol that had passed over the catalyst bed. Trickle bed reactor modeling was performed using a three-phase heterogeneous model. Model parameters were determined from literature correlations, batch derived kinetic data, and trickle bed derived catalyst deactivation data. The model equations were solved using orthogonal collocations on finite elements. Trickle bed performance was successfully predicted using the batch derived kinetic model and the three-phase reactor model. Thus, using the kinetics determined from limited data in the batch mode, it is possible to predict continuous flow multiphase reactor performance.

Three Dimensional Dirac Semimetals

Science.gov (United States)

Zaheer, Saad

2014-03-01

Dirac points on the Fermi surface of two dimensional graphene are responsible for its unique electronic behavior. One can ask whether any three dimensional materials support similar pseudorelativistic physics in their bulk electronic spectra. This possibility has been investigated theoretically and is now supported by two successful experimental demonstrations reported during the last year. In this talk, I will summarize the various ways in which Dirac semimetals can be realized in three dimensions with primary focus on a specific theory developed on the basis of representations of crystal spacegroups. A three dimensional Dirac (Weyl) semimetal can appear in the presence (absence) of inversion symmetry by tuning parameters to the phase boundary separating a bulk insulating and a topological insulating phase. More generally, we find that specific rules governing crystal symmetry representations of electrons with spin lead to robust Dirac points at high symmetry points in the Brillouin zone. Combining these rules with microscopic considerations identifies six candidate Dirac semimetals. Another method towards engineering Dirac semimetals involves combining crystal symmetry and band inversion. Several candidate materials have been proposed utilizing this mechanism and one of the candidates has been successfully demonstrated as a Dirac semimetal in two independent experiments. Work carried out in collaboration with: Julia A. Steinberg, Steve M. Young, J.C.Y. Teo, C.L. Kane, E.J. Mele and Andrew M. Rappe.

Antibiotic Fermentation Broth Treatment by a pilot upflow anaerobic sludge bed reactor and kinetic modeling.

Science.gov (United States)

Coskun, T; Kabuk, H A; Varinca, K B; Debik, E; Durak, I; Kavurt, C

2012-10-01

In this study, an upflow anaerobic sludge blanket (UASB) mesophilic reactor was used to remove antibiotic fermentation broth wastewater. The hydraulic retention time was held constant at 13.3 days. The volumetric organic loading value increased from 0.33 to 7.43 kg(COD)m(-3)d(-1) using antibiotic fermentation broth wastewater gradually diluted with various ratios of domestic wastewater. A COD removal efficiency of 95.7% was obtained with a maximum yield of 3,700 L d(-1) methane gas production. The results of the study were interpreted using the modified Stover-Kincannon, first-order, substrate mass balance and Van der Meer and Heertjes kinetic models. The obtained kinetic coefficients showed that antibiotic fermentation broth wastewater can be successfully treated using a UASB reactor while taking COD removal and methane production into account. Copyright © 2012 Elsevier Ltd. All rights reserved.

Three-dimensional instability of standing waves

Science.gov (United States)

Zhu, Qiang; Liu, Yuming; Yue, Dick K. P.

2003-12-01

We investigate the three-dimensional instability of finite-amplitude standing surface waves under the influence of gravity. The analysis employs the transition matrix (TM) approach and uses a new high-order spectral element (HOSE) method for computation of the nonlinear wave dynamics. HOSE is an extension of the original high-order spectral method (HOS) wherein nonlinear wave wave and wave body interactions are retained up to high order in wave steepness. Instead of global basis functions in HOS, however, HOSE employs spectral elements to allow for complex free-surface geometries and surface-piercing bodies. Exponential convergence of HOS with respect to the total number of spectral modes (for a fixed number of elements) and interaction order is retained in HOSE. In this study, we use TM-HOSE to obtain the stability of general three-dimensional perturbations (on a two-dimensional surface) on two classes of standing waves: plane standing waves in a rectangular tank; and radial/azimuthal standing waves in a circular basin. For plane standing waves, we confirm the known result of two-dimensional side-bandlike instability. In addition, we find a novel three-dimensional instability for base flow of any amplitude. The dominant component of the unstable disturbance is an oblique (standing) wave oriented at an arbitrary angle whose frequency is close to the (nonlinear) frequency of the original standing wave. This finding is confirmed by direct long-time simulations using HOSE which show that the nonlinear evolution leads to classical Fermi Pasta Ulam recurrence. For the circular basin, we find that, beyond a threshold wave steepness, a standing wave (of nonlinear frequency Omega) is unstable to three-dimensional perturbations. The unstable perturbation contains two dominant (standing-wave) components, the sum of whose frequencies is close to 2Omega. From the cases we consider, the critical wave steepness is found to generally decrease/increase with increasing radial

SUPERCRITICAL FLUID TREATMENT OF THREE-DIMENSIONAL HYDROGEL MATRICES, COMPOSED OF CHITOSAN DERIVATIVES

Directory of Open Access Journals (Sweden)

P. S. Timashev

2016-01-01

Full Text Available Aim. Controlled treatment of the physico-chemical and mechanical properties of a three-dimensional crosslinked matrix based on reactive chitosan. Materials and methods. The three-dimensional matrices were obtained using photosensitive composition based on allyl chitosan (5 wt%, poly(ethylene glycol diacrylate (8 wt% and the photoinitiator Irgacure 2959 (1 wt% by laser stereolithography setting. The kinetic swelling curves were constructed for structures in the base and salt forms of chitosan using gravimetric method and the contact angles were measured using droplet spreading. The supercritical fl uid setting (40 °C, 12 MPa was used to process matrices during 1.5 hours. Using nanohardness Piuma Nanoindenter we calculated values of Young’s modulus. The study of cytotoxicity was performed by direct contact with the culture of the NIH 3T3 mouse fi broblast cell line. Results. Architectonics of matrices fully repeats the program model. Matrices are uniform throughout and retain their shape after being transferred to the base form. Matrices compressed by 5% after treatment in supercritical carbon dioxide (scCO2 . The elastic modulus of matrices after scCO2 treatment is 4 times higher than the original matrix. The kinetic swelling curves have similar form. In this case the maximum degree of swelling for matrices in base form is 2–2.5 times greater than that of matrices in salt form. There was a surface hydrophobization after the material was transferred to the base form: the contact angle is 94°, and for the salt form it is 66°. The basic form absorbs liquid approximately 1.6 times faster. The fi lm thickness was increased in the area of contact with the liquid droplets after absorption by 133 and 87% for the base and the salt forms, respectively. Treatment of samples in scCO2 reduces their cytotoxicity from 2 degree of reaction (initial samples down to 1 degree of reaction. Conclusion. The use of supercritical carbon dioxide for scaffolds

Drying kinetics characteristic of Indonesia lignite coal (IBC) using lab scale fixed bed reactor

Energy Technology Data Exchange (ETDEWEB)

Kang, TaeJin; Jeon, DoMan; Namkung, Hueon; Jang, DongHa; Jeon, Youngshin; Kim, Hyungtaek [Ajou Univ., Suwon (Korea, Republic of). Div. of Energy Systems Research

2013-07-01

Recent instability of energy market arouse a lot of interest about coal which has a tremendous amount of proven coal reserves worldwide. South Korea hold the second rank by importing 80 million tons of coal in 2007 following by Japan. Among various coals, there is disused coal. It's called Low Rank Coal (LRC). Drying process has to be preceded before being utilized as power plant. In this study, drying kinetics of LRC is induced by using a fixed bed reactor. The drying kinetics was deduced from particle size, the inlet gas temperature, the drying time, the gas velocity, and the L/D ratio. The consideration on Reynold's number was taken for correction of gas velocity, particle size, and the L/D ratio was taken for correction packing height of coal. It can be found that active drying of free water and phase boundary reaction is suitable mechanism through the fixed bed reactor experiments.

Cylindrical Three-Dimensional Porous Anodic Alumina Networks

Directory of Open Access Journals (Sweden)

Pedro M. Resende

2016-11-01

Full Text Available The synthesis of a conformal three-dimensional nanostructure based on porous anodic alumina with transversal nanopores on wires is herein presented. The resulting three-dimensional network exhibits the same nanostructure as that obtained on planar geometries, but with a macroscopic cylindrical geometry. The morphological analysis of the nanostructure revealed the effects of the initial defects on the aluminum surface and the mechanical strains on the integrity of the three-dimensional network. The results evidence the feasibility of obtaining 3D porous anodic alumina on non-planar aluminum substrates.

Multiparallel Three-Dimensional Optical Microscopy

Science.gov (United States)

Nguyen, Lam K.; Price, Jeffrey H.; Kellner, Albert L.; Bravo-Zanoquera, Miguel

2010-01-01

Multiparallel three-dimensional optical microscopy is a method of forming an approximate three-dimensional image of a microscope sample as a collection of images from different depths through the sample. The imaging apparatus includes a single microscope plus an assembly of beam splitters and mirrors that divide the output of the microscope into multiple channels. An imaging array of photodetectors in each channel is located at a different distance along the optical path from the microscope, corresponding to a focal plane at a different depth within the sample. The optical path leading to each photodetector array also includes lenses to compensate for the variation of magnification with distance so that the images ultimately formed on all the photodetector arrays are of the same magnification. The use of optical components common to multiple channels in a simple geometry makes it possible to obtain high light-transmission efficiency with an optically and mechanically simple assembly. In addition, because images can be read out simultaneously from all the photodetector arrays, the apparatus can support three-dimensional imaging at a high scanning rate.

Backlund transformations and three-dimensional lattice equations

NARCIS (Netherlands)

Nijhoff, F.W.; Capel, H.W.; Wiersma, G.L.; Quispel, G.R.W.

1984-01-01

A (nonlocal) linear integral equation is studied, which allows for Bäcklund transformations in the measure. The compatibility of three of these transformations leads to an integrable nonlinear three-dimensional lattice equation. In appropriate continuum limits the two-dimensional Toda-lattice

Arching in three-dimensional clogging

Science.gov (United States)

Török, János; Lévay, Sára; Szabó, Balázs; Somfai, Ellák; Wegner, Sandra; Stannarius, Ralf; Börzsönyi, Tamás

2017-06-01

Arching in dry granular material is a long established concept, however it remains still an open question how three-dimensional orifices clog. We investigate by means of numerical simulations and experimental data how the outflow creates a blocked configuration of particles. We define the concave surface of the clogged dome by two independent methods (geometric and density based). The average shape of the cupola for spheres is almost a hemisphere but individual samples have large holes in the structure indicating a blocked state composed of two-dimensional force chains rather than three-dimensional objects. The force chain structure justifies this assumption. For long particles the clogged configurations display large variations, and in certain cases the empty region reaches a height of 5 hole diameters. These structures involve vertical walls consisting of horizontally placed stable stacking of particles.

Three-dimensional reconstruction of the biliary tract using spiral computed tomography. Three-dimensional cholangiography

International Nuclear Information System (INIS)

Gon, Masanori; Ogura, Norihiro; Uetsuji, Shouji; Ueyama, Yasuo

1995-01-01

In this study, 310 patients with benign biliary diseases, 20 with gallbladder cancer, and 8 with biliary tract carcinoma underwent spiral CT (SCT) scanning at cholangiography. Depiction rate of the shape of the conjunction site of the gallbladder and biliary tract was 27.5% by conventional intravenous cholangiography (DIC), 92.5% by ERC, and 90.0% by DIC-SCT. Abnormal cystic duct course was admitted in 14.1%. Multiplanar reconstruction by DIC-SCT enabled identification of the common bile duct and intrahepatic bile duct stone. Three-dimensional reconstruction of DIC-SCT was effective in evaluating obstruction of the anastomosis or passing condition of after hepatico-jejunostomy. Two-dimensional SCT images through PTCD tube enabled degree of hepatic invasion in bile duct cancer, and three-dimensional images were useful in grasping the morphology of the bile duct branches near the obstruction site. DIC-SCT is therefore considered a useful procedure as non-invasive examination of bile duct lesions. (S.Y.)

A simple three-dimensional-focusing, continuous-flow mixer for the study of fast protein dynamics.

Science.gov (United States)

Burke, Kelly S; Parul, Dzmitry; Reddish, Michael J; Dyer, R Brian

2013-08-07

We present a simple, yet flexible microfluidic mixer with a demonstrated mixing time as short as 80 μs that is widely accessible because it is made of commercially available parts. To simplify the study of fast protein dynamics, we have developed an inexpensive continuous-flow microfluidic mixer, requiring no specialized equipment or techniques. The mixer uses three-dimensional, hydrodynamic focusing of a protein sample stream by a surrounding sheath solution to achieve rapid diffusional mixing between the sample and sheath. Mixing initiates the reaction of interest. Reactions can be spatially observed by fluorescence or absorbance spectroscopy. We characterized the pixel-to-time calibration and diffusional mixing experimentally. We achieved a mixing time as short as 80 μs. We studied the kinetics of horse apomyoglobin (apoMb) unfolding from the intermediate (I) state to its completely unfolded (U) state, induced by a pH jump from the initial pH of 4.5 in the sample stream to a final pH of 2.0 in the sheath solution. The reaction time was probed using the fluorescence of 1-anilinonaphthalene-8-sulfonate (1,8-ANS) bound to the folded protein. We observed unfolding of apoMb within 760 μs, without populating additional intermediate states under these conditions. We also studied the reaction kinetics of the conversion of pyruvate to lactate catalyzed by lactate dehydrogenase using the intrinsic tryptophan emission of the enzyme. We observe sub-millisecond kinetics that we attribute to Michaelis complex formation and loop domain closure. These results demonstrate the utility of the three-dimensional focusing mixer for biophysical studies of protein dynamics.

Comparison of finite element and influence function methods for three-dimensional elastic analysis of boiling water reactor feedwater nozzle cracks

International Nuclear Information System (INIS)

Besuner, P.M.; Caughey, W.R.

1976-11-01

The paper compares the finite element (FE) and influence function (IF) methods for a three-dimensional elastic analysis of postulated circular-shaped surface cracks in the feedwater nozzle of a typical boiling water reactor (BWR). The FE method is incorporated in a direct manner. The nozzle and crack geometry and the complex loading are all included in the model which simulates the structural crack problem. The IF method is used to compute stress intensity factors only when the uncracked stress field (that is, the stress in the uncracked solid at the locus of the crack to be eventually considered) has been computed previously. The IF method evaluates correctly the disturbance of this uncracked stress field caused by the crack by utilizing a method of elastic superposition. Both the IF and FE methods are described in detail in the paper and are applied to several test cases chosen for their similarity to the nozzle crack problem and for the availability of an accurate published result obtained from some recognized third method of solution. Results are given which summarize both the accuracy and the direct computer costs of the two methods

Three-dimensional printed knotted reactors enabling highly sensitive differentiation of silver nanoparticles and ions in aqueous environmental samples

International Nuclear Information System (INIS)

Su, Cheng-Kuan; Hsieh, Meng-Hsuan; Sun, Yuh-Chang

2016-01-01

Whether silver nanoparticles (AgNPs) persist or release silver ions (Ag + ) when discharged into a natural environment has remained an unresolved issue. In this study, we employed a low-cost stereolithographic three-dimensional printing (3DP) technology to fabricate the angle-defined knotted reactors (KRs) to construct a simple differentiation scheme for quantitative assessment of Ag + ions and AgNPs in municipal wastewater samples. We chose xanthan/phosphate-buffered saline as a dispersion medium for in situ stabilization of the two silver species, while also facilitating their extraction from complicated wastewater matrices. After method optimization, we measured extraction efficiencies of 54.5 and 32.3% for retaining Ag + ions and AgNPs, respectively, in the printed KR (768-turn), with detection limits (DLs) of 0.86 and 0.52 ng L −1 when determining Ag + ions and AgNPs, respectively (sample run at pH 11 without a rinse solution), and 0.86 ng L −1 when determining Ag + ions alone (sample run at pH 12 with a 1.5-mL rinse solution). The proposed scheme is tolerant of the wastewater matrix and provides more reliable differentiation between Ag + /AgNPs than does a conventional filtration method. The concept and applicability of adopting 3DP technology to renew traditional KR devices were evidently proven by means of these significantly improved analytical performance. Our analytical data suggested that the concentrations of Ag + ions and AgNPs in the tested industrial wastewater sample were both higher than those in domestic wastewater, implying that industrial activity might be a main source of environmental silver species, rather than domestic discharge from AgNP-containing products. - Highlights: • 3D printed knotted reactors are utilized to differentiate AgNPs and Ag + ions. • Xanthan/phosphate-buffered saline is used for stabilizing the two silver species. • Extraction efficiency up to 54.5% is available for retaining Ag + ion species. • The

Three-dimensional analyses of fluid flow and heat transfer for moderator integrity assessment in PHWR

International Nuclear Information System (INIS)

Yu, S.-O.; Kim, M.; Kim, H.-J.

2002-01-01

A CANDU reactor has the unique features and the intrinsic safety related characteristics that distinguish it from other water-cooled thermal reactors. If there is the loss of coolant accident (LOCA) and a coincident failure of the emergency coolant injection (ECI) system, the heavy water moderator is continuously cooled, providing a heat sink for decay heat produced in the fuel. Therefore, it is one of major concerns to estimate the local subcooling of moderator inside the calandria vessel under postulated accident in CANDU safety analyses. The Canadian Nuclear Safety Commission (CNSC), a regulatory body in Canada, categorized the integrity of moderator as a generic safety issue and recommended that a series of experimental works be performed to verify the safety evaluation codes for individual simulated condition of nuclear power plant, comparing with the results of three-dimensional experimental data. In this study, three-dimensional analyses of fluid flow and heat transfer have been performed to assess thermal-hydraulic characteristics for moderator simulation conducted by SPEL (Sheridan Park Experimental Laboratory) experimental facility. The parametric study has also carried out to investigate the effect of major parameters such as flowrate, temperature, and heat load generated from the heaters on the temperature and flow distribution inside the moderator. Three flow patterns have been identified in the moderator with flowrate, heat generation, or both. As the transition of fluid flow is progressed, it is found that the dimensionless numbers (Ar) and the ratio of buoyancy to inertia forces are constant. (author)

A Nodal and Finite Difference Hybrid Method for Pin-by-Pin Heterogeneous Three-Dimensional Light Water Reactor Diffusion Calculations

International Nuclear Information System (INIS)

Lee, Deokjung; Downar, Thomas J.; Kim, Yonghee

2004-01-01

An innovative hybrid spatial discretization method is proposed to improve the computational efficiency of pin-wise heterogeneous three-dimensional light water reactor (LWR) core neutronics analysis. The newly developed method employs the standard finite difference method in the x and y directions and the well-known nodal methods [nodal expansion method (NEM) and analytic nodal method (ANM) as needed] in the z direction. Four variants of the hybrid method are investigated depending on the axial nodal methodologies: HYBRID A, NEM with the conventional quadratic transverse leakage; HYBRID B, the conventional NEM method except that the transverse-leakage shapes are obtained from a fine-mesh local problem (FMLP) around the control rod tip; HYBRID C, the same as HYBRID B except that ANM with a high-order transverse leakage obtained from the FMLP is used in the vicinity of the control rod tip; and HYBRID D, the same as HYBRID C except that the transverse leakage is determined using the buckling approximation instead of the FMLP around the control rod tip. Benchmark calculations demonstrate that all the hybrid algorithms are consistent and stable and that the HYBRID C method provides the best numerical performance in the case of rodded LWR problems with pin-wise homogenized cross sections

Kinetic Theory of a Confined Quasi-Two-Dimensional Gas of Hard Spheres

Directory of Open Access Journals (Sweden)

J. Javier Brey

2017-02-01

Full Text Available The dynamics of a system of hard spheres enclosed between two parallel plates separated a distance smaller than two particle diameters is described at the level of kinetic theory. The interest focuses on the behavior of the quasi-two-dimensional fluid seen when looking at the system from above or below. In the first part, a collisional model for the effective two-dimensional dynamics is analyzed. Although it is able to describe quite well the homogeneous evolution observed in the experiments, it is shown that it fails to predict the existence of non-equilibrium phase transitions, and in particular, the bimodal regime exhibited by the real system. A critical revision analysis of the model is presented , and as a starting point to get a more accurate description, the Boltzmann equation for the quasi-two-dimensional gas has been derived. In the elastic case, the solutions of the equation verify an H-theorem implying a monotonic tendency to a non-uniform steady state. As an example of application of the kinetic equation, here the evolution equations for the vertical and horizontal temperatures of the system are derived in the homogeneous approximation, and the results compared with molecular dynamics simulation results.

Three-dimensional appearance of the lips muscles with three-dimensional isotropic MRI: in vivo study

Energy Technology Data Exchange (ETDEWEB)

Olszewski, Raphael; Reychler, H. [Universite Catholique de Louvain, Department of Oral and Maxillofacial Surgery, Cliniques Universitaires Saint Luc, Brussels (Belgium); Liu, Y.; Xu, T.M. [Peking University School and Hospital of Stomatology, Department of Orthodontics, Beijing (China); Duprez, T. [Universite Catholique de Louvain, Department of Radiology, Cliniques Universitaires Saint Luc, Brussels (Belgium)

2009-06-15

Our knowledge of facial muscles is based primarily on atlases and cadaveric studies. This study describes a non-invasive in vivo method (3D MRI) for segmenting and reconstructing facial muscles in a three-dimensional fashion. Three-dimensional (3D), T1-weighted, 3 Tesla, isotropic MRI was applied to a subject. One observer performed semi-automatic segmentation using the Editor module from the 3D Slicer software (Harvard Medical School, Boston, MA, USA), version 3.2. We were able to successfully outline and three-dimensionally reconstruct the following facial muscles: pars labialis orbicularis oris, m. levatro labii superioris alaeque nasi, m. levator labii superioris, m. zygomaticus major and minor, m. depressor anguli oris, m. depressor labii inferioris, m. mentalis, m. buccinator, and m. orbicularis oculi. 3D reconstruction of the lip muscles should be taken into consideration in order to improve the accuracy and individualization of existing 3D facial soft tissue models. More studies are needed to further develop efficient methods for segmentation in this field. (orig.)

Three-dimensional appearance of the lips muscles with three-dimensional isotropic MRI: in vivo study.

Science.gov (United States)

Olszewski, Raphael; Liu, Y; Duprez, T; Xu, T M; Reychler, H

2009-06-01

Our knowledge of facial muscles is based primarily on atlases and cadaveric studies. This study describes a non-invasive in vivo method (3D MRI) for segmenting and reconstructing facial muscles in a three-dimensional fashion. Three-dimensional (3D), T1-weighted, 3 Tesla, isotropic MRI was applied to a subject. One observer performed semi-automatic segmentation using the Editor module from the 3D Slicer software (Harvard Medical School, Boston, MA, USA), version 3.2. We were able to successfully outline and three-dimensionally reconstruct the following facial muscles: pars labialis orbicularis oris, m. levatro labii superioris alaeque nasi, m. levator labii superioris, m. zygomaticus major and minor, m. depressor anguli oris, m. depressor labii inferioris, m. mentalis, m. buccinator, and m. orbicularis oculi. 3D reconstruction of the lip muscles should be taken into consideration in order to improve the accuracy and individualization of existing 3D facial soft tissue models. More studies are needed to further develop efficient methods for segmentation in this field.

Reactor kinetics calculated in the summation method and key delayed-neutron data

International Nuclear Information System (INIS)

Oyamatsu, Kazuhiro

2001-01-01

The point-reactor kinetics after a step reactivity insertion to a critical condition is solved directly form fission-product (FP) data (fission yields and decay data) for the first time. Numerical calculations are performed with the FP data in ENDF/B-VI. The inhour equation obtained directly from the FP data shows a different behavior at long periods from the one obtained from Tuttle's six-group parameter sets. The behavior is quite similar to the one obtained from the six-group parameter sets in ENDF/B-VI, that were obtained from FP data in a preliminary version of ENDF/B-VI. To identify the erroneous FP data, we examine the asymptotic form of the inhour equation at an infinitely long period. It is found that the most important precursors for long reactor periods are found 137 I, 88 Br and 87 Br. They cover more than 60% of the reactivity. It is remarkable that 137 I alone covers 30-50% depending on the fissioning system. In addition to the three precursors, 136 Te is found a candidate precursor for the peculiarity from the time dependence of the delayed neutron activity. It is recommended that the precision of their Pn values should be improved experimentally. For 137 I, 88 Br, and 87 Br, the relative uncertainty, dPn/Pn, should be decreased down to 2% and for 136 Te to 5%. (author)

Coupling of the 3D neutron kinetic core model DYN3D with the CFD software ANSYS-CFX

International Nuclear Information System (INIS)

Grahn, Alexander; Kliem, Sören; Rohde, Ulrich

2015-01-01

Highlights: • Improved thermal hydraulic description of nuclear reactor cores. • Possibility of three-dimensional flow phenomena in the core, such as cross flow, flow reversal, flow around obstacles. • Simulation at higher spatial resolution as compared to system codes. - Abstract: This article presents the implementation of a coupling between the 3D neutron kinetic core model DYN3D and the commercial, general purpose computational fluid dynamics (CFD) software ANSYS-CFX. In the coupling approach, parts of the thermal hydraulic calculation are transferred to CFX for its better ability to simulate the three-dimensional coolant redistribution in the reactor core region. The calculation of the heat transfer from the fuel into the coolant remains with DYN3D, which incorporates well tested and validated heat transfer models for rod-type fuel elements. On the CFX side, the core region is modeled based on the porous body approach. The implementation of the code coupling is verified by comparing test case results with reference solutions of the DYN3D standalone version. Test cases cover mini and full core geometries, control rod movement and partial overcooling transients

Three-Dimensional Messages for Interstellar Communication

Science.gov (United States)

Vakoch, Douglas A.

One of the challenges facing independently evolved civilizations separated by interstellar distances is to communicate information unique to one civilization. One commonly proposed solution is to begin with two-dimensional pictorial representations of mathematical concepts and physical objects, in the hope that this will provide a foundation for overcoming linguistic barriers. However, significant aspects of such representations are highly conventional, and may not be readily intelligible to a civilization with different conventions. The process of teaching conventions of representation may be facilitated by the use of three-dimensional representations redundantly encoded in multiple formats (e.g., as both vectors and as rasters). After having illustrated specific conventions for representing mathematical objects in a three-dimensional space, this method can be used to describe a physical environment shared by transmitter and receiver: a three-dimensional space defined by the transmitter--receiver axis, and containing stars within that space. This method can be extended to show three-dimensional representations varying over time. Having clarified conventions for representing objects potentially familiar to both sender and receiver, novel objects can subsequently be depicted. This is illustrated through sequences showing interactions between human beings, which provide information about human behavior and personality. Extensions of this method may allow the communication of such culture-specific features as aesthetic judgments and religious beliefs. Limitations of this approach will be noted, with specific reference to ETI who are not primarily visual.

A benchmark on the calculation of kinetic parameters based on reactivity effect experiments in the CROCUS reactor

International Nuclear Information System (INIS)

Paratte, J.M.; Frueh, R.; Kasemeyer, U.; Kalugin, M.A.; Timm, W.; Chawla, R.

2006-01-01

Measurements in the CROCUS reactor at EPFL, Lausanne, are reported for the critical water level and the inverse reactor period for several different sets of delayed supercritical conditions. The experimental configurations were also calculated by four different calculation methods. For each of the supercritical configurations, the absolute reactivity value has been determined in two different ways, viz.: (i) through direct comparison of the multiplication factor obtained employing a given calculation method with the corresponding value for the critical case (calculated reactivity: ρ calc ); (ii) by application of the inhour equation using the kinetic parameters obtained for the critical configuration and the measured inverse reactor period (measured reactivity: ρ meas ). The calculated multiplication factors for the reference critical configuration, as well as ρ calc for the supercritical cases, are found to be in good agreement. However, the values of ρ meas produced by two of the applied calculation methods differ appreciably from the corresponding ρ calc values, clearly indicating deficiencies in the kinetic parameters obtained from these methods

Coalescence kinetics of dispersed crude oil in a laboratory reactor

International Nuclear Information System (INIS)

Sterling, M.C. Jr.; Ojo, T.; Autenrieth, R.L.; Bonner, J.S.; Page, C.A.; Ernst, A.N.S.

2002-01-01

A study was conducted to examine the effects of salinity and mixing energy on the resurfacing and coalescence rates of chemically dispersed crude oil droplets. This kinetic study involved the use of mean shear rates to characterize the mixing energy in a laboratory reactor. Coagulation kinetics of dispersed crude oil were determined within a range of mean shear rates of 5, 10, 15, and 20 per second, and with salinity values of 10 and 30 per cent. Observed droplet distributions were fit to a transport-reaction model to estimate collision efficiency values and their dependence on salinity and mixing energy. Dispersant efficiencies were compared with those derived from other laboratory testing methods. Experimentally determined dispersant efficiencies were found to be 10 to 50 per cent lower than predicted using a non-interacting droplet model, but dispersant efficiencies were higher than those predicted using other testing methods. 24 refs., 1 tab., 3 figs

Three-dimensional topological insulators and bosonization

Energy Technology Data Exchange (ETDEWEB)

Cappelli, Andrea [INFN, Sezione di Firenze,Via G. Sansone 1, 50019 Sesto Fiorentino - Firenze (Italy); Randellini, Enrico [INFN, Sezione di Firenze,Via G. Sansone 1, 50019 Sesto Fiorentino - Firenze (Italy); Dipartimento di Fisica e Astronomia, Università di Firenze,Via G. Sansone 1, 50019 Sesto Fiorentino - Firenze (Italy); Sisti, Jacopo [Scuola Internazionale Superiore di Studi Avanzati (SISSA),Via Bonomea 265, 34136 Trieste (Italy)

2017-05-25

Massless excitations at the surface of three-dimensional time-reversal invariant topological insulators possess both fermionic and bosonic descriptions, originating from band theory and hydrodynamic BF theory, respectively. We analyze the corresponding field theories of the Dirac fermion and compactified boson and compute their partition functions on the three-dimensional torus geometry. We then find some non-dynamic exact properties of bosonization in (2+1) dimensions, regarding fermion parity and spin sectors. Using these results, we extend the Fu-Kane-Mele stability argument to fractional topological insulators in three dimensions.

Two-dimensional analytical solution for nodal calculation of nuclear reactors

International Nuclear Information System (INIS)

Silva, Adilson C.; Pessoa, Paulo O.; Silva, Fernando C.; Martinez, Aquilino S.

2017-01-01

Highlights: • A proposal for a coarse mesh nodal method is presented. • The proposal uses the analytical solution of the two-dimensional neutrons diffusion equation. • The solution is performed homogeneous nodes with dimensions of the fuel assembly. • The solution uses four average fluxes on the node surfaces as boundary conditions. • The results show good accuracy and efficiency. - Abstract: In this paper, the two-dimensional (2D) neutron diffusion equation is analytically solved for two energy groups (2G). The spatial domain of reactor core is divided into a set of nodes with uniform nuclear parameters. To determine iteratively the multiplication factor and the neutron flux in the reactor we combine the analytical solution of the neutron diffusion equation with an iterative method known as power method. The analytical solution for different types of regions that compose the reactor is obtained, such as fuel and reflector regions. Four average fluxes in the node surfaces are used as boundary conditions for analytical solution. Discontinuity factors on the node surfaces derived from the homogenization process are applied to maintain averages reaction rates and the net current in the fuel assembly (FA). To validate the results obtained by the analytical solution a relative power density distribution in the FAs is determined from the neutron flux distribution and compared with the reference values. The results show good accuracy and efficiency.

ENERGY RELEASE AND TRANSFER IN SOLAR FLARES: SIMULATIONS OF THREE-DIMENSIONAL RECONNECTION

International Nuclear Information System (INIS)

Birn, J.; Fletcher, L.; Hesse, M.; Neukirch, T.

2009-01-01

Using three-dimensional magnetohydrodynamic simulations we investigate energy release and transfer in a three-dimensional extension of the standard two-ribbon flare picture. In this scenario, reconnection is initiated in a thin current sheet (suggested to form below a departing coronal mass ejection) above a bipolar magnetic field. Two cases are contrasted: an initially force-free current sheet (low beta) and a finite-pressure current sheet (high beta), where beta represents the ratio between gas (plasma) and magnetic pressure. The energy conversion process from reconnection consists of incoming Poynting flux turned into up- and downgoing Poynting flux, enthalpy flux, and bulk kinetic energy flux. In the low-beta case, the outgoing Poynting flux is the dominant contribution, whereas the outgoing enthalpy flux dominates in the high-beta case. The bulk kinetic energy flux is only a minor contribution in the downward direction. The dominance of the downgoing Poynting flux in the low-beta case is consistent with an alternative to the thick target electron beam model for solar flare energy transport, suggested recently by Fletcher and Hudson, whereas the enthalpy flux may act as an alternative transport mechanism. For plausible characteristic parameters of the reconnecting field configuration, we obtain energy release timescales and energy output rates that compare favorably with those inferred from observations for the impulsive phase of flares. Significant enthalpy flux and heating are found even in the initially force-free case with very small background beta, resulting mostly from adiabatic compression rather than Ohmic dissipation. The energy conversion mechanism is most easily understood as a two-step process (although the two steps may occur essentially simultaneously): the first step is the acceleration of the plasma by Lorentz forces in layers akin to the slow shocks in the Petschek reconnection model, involving the conversion of magnetic energy to bulk kinetic

Effect of local automatic control rods on three-dimensional calculations of the power distribution in an RBMK

International Nuclear Information System (INIS)

Pogosbekyan, L.R.; Lysov, D.A.; Bronitskii, L.L.

1993-01-01

Numerical simulators and information systems that support nuclear reactor operators must have fast models to estimate how fuel reloads and control rod displacement affect neutron and power distributions in the core. The consequences of reloads and control rod displacement cannot be evaluated correctly without considering local automatic control-rod operations in maintaining the radial power distribution. Fast three-dimensional models to estimate the effects of reloads and displacement of the control and safety rods have already been examined. I.V. Zonov et al. used the following assumptions in their calculational model: (1) the full-scale problem could be reduced a three-dimensional fragment of a locally perturbed core, and (2) the boundary conditions of the fragment and its total power were constant. The last assumption considers approximately how local automatic control rods stabilize the radial power distribution, but three dimensional calculations with these rods are not considered. These assumptions were introduced to obtain high computational speed. I.L. Bronitskii et al. considered in more detail how moving the local automatic control rods affect the power dimensional in the three-dimensional fragment, because, with on-line monitoring of the reload process, information on control rod positions is periodically renewed, and the calculations are done in real time. This model to predict the three-dimensional power distribution to (1) do a preliminary reload analysis, and (2) prepare the core for reloading did not consider the effect of perturbations from the local automatic control rods. Here we examine a model of a stationary neutron distribution. On one hand it gives results in an acceptable computation time; on the other it is a full-scale three-dimensional model and considers how local automatic control rods affect both the radial and axial power distribution

High-resolution three-dimensional mapping of semiconductor dopant potentials

DEFF Research Database (Denmark)

Twitchett, AC; Yates, TJV; Newcomb, SB

2007-01-01

Semiconductor device structures are becoming increasingly three-dimensional at the nanometer scale. A key issue that must be addressed to enable future device development is the three-dimensional mapping of dopant distributions, ideally under "working conditions". Here we demonstrate how a combin......Semiconductor device structures are becoming increasingly three-dimensional at the nanometer scale. A key issue that must be addressed to enable future device development is the three-dimensional mapping of dopant distributions, ideally under "working conditions". Here we demonstrate how...... a combination of electron holography and electron tomography can be used to determine quantitatively the three-dimensional electrostatic potential in an electrically biased semiconductor device with nanometer spatial resolution....

A new nodal kinetics method for analyzing fast control rod motions in nuclear reactor cores

International Nuclear Information System (INIS)

Kaya, S.; Yavuz, H.

2001-01-01

A new nodal kinetics approach is developed for analyzing large reactivity accidents in nuclear reactor cores. This method shows promising that it has capability of inspecting promt criticality transients and it gives comparable results with respect to those of other techniques. (orig.)

Vectorization of three-dimensional neutron diffusion code CITATION

International Nuclear Information System (INIS)

Harada, Hiroo; Ishiguro, Misako

1985-01-01

Three-dimensional multi-group neutron diffusion code CITATION has been widely used for reactor criticality calculations. The code is expected to be run at a high speed by using recent vector supercomputers, when it is appropriately vectorized. In this paper, vectorization methods and their effects are described for the CITATION code. Especially, calculation algorithms suited for vectorization of the inner-outer iterative calculations which spend most of the computing time are discussed. The SLOR method, which is used in the original CITATION code, and the SOR method, which is adopted in the revised code, are vectorized by odd-even mesh ordering. The vectorized CITATION code is executed on the FACOM VP-100 and VP-200 computers, and is found to run over six times faster than the original code for a practical-scale problem. The initial value of the relaxation factor and the number of inner-iterations given as input data are also investigated since the computing time depends on these values. (author)

Equilibrium: three-dimensional configurations

International Nuclear Information System (INIS)

Anon.

1987-01-01

This chapter considers toroidal MHD configurations that are inherently three-dimensional. The motivation for investigation such complicated equilibria is that they possess the potential for providing toroidal confinement without the need of a net toroidal current. This leads to a number of advantages with respect to fusion power generation. First, the attractive feature of steady-state operation becomes more feasible since such configurations no longer require a toroidal current transformer. Second, with zero net current, one potentially dangerous class of MHD instabilities, the current-driven kink modes, is eliminated. Finally, three-dimensional configurations possess nondegenerate flux surfaces even in the absence of plasma pressure and plasma current. Although there is an enormous range of possible three-dimensional equilibria, the configurations of interest are accurately described as axisymmetric tori with superimposed helical fields; furthermore, they possess no net toroidal current. Instead, two different and less obvious restoring forces are developed: the helical sideband force and the toroidal dipole current force. Each is discussed in detail in Chapter 7. A detailed discussion of the parallel current constraint, including its physical significance, is given in section 7.2. A general analysis of helical sideband equilibria, along with a detailed description of the Elmo bumpy torus, is presented in sections 7.3 and 7.4. A general description of toroidal dipole-current equilibria, including a detailed discussion of stellarators, heliotrons, and torsatrons, is given in sections 7.5 and 7.6

Dimensional measurement of fresh fuel bundle for CANDU reactor

International Nuclear Information System (INIS)

Jo, Chang Keun; Cho, Moon Sung; Suk, Ho Chun; Koo, Dae Seo; Jun, Ji Su; Jung, Jong Yeob

2005-01-01

This report describes the results of the dimensional measurement of fresh fuel bundles for the CANDU reactor in order to estimate the integrity of fuel bundle in two-phase flow in the CANDU-6 fuel channel. The dimensional measurements of fuel bundles are performed by using the 'CANDU Fuel In-Bay Inspection and Dimensional Measurement System', which was developed by this project. The dimensional measurements are done from February 2004 to March 2004 in the CANDU fuel storage of KNFC for the 36 fresh fuel bundles, which are produced by KNFC and are waiting for the delivery to the Wolsong-3 plant. The detail items of dimensional measurements are included fuel rod and bearing pad profiles of the outer ring in fuel bundle, diameter of fuel bundle, bowing of fuel bundle, fuel rod length, and surface profile of end plate profile. The measurement data will be compared with those of the post-irradiated bundles cooled in Wolsong-3 NPP spent fuel pool by using the same bundles and In-Bay Measurement System. So, this analysis of data will be applied for the evaluation of fuel bundle integrity in two-phase flow of the CANDU-6 fuel channel

Volume scanning three-dimensional display with an inclined two-dimensional display and a mirror scanner

Science.gov (United States)

Miyazaki, Daisuke; Kawanishi, Tsuyoshi; Nishimura, Yasuhiro; Matsushita, Kenji

2001-11-01

A new three-dimensional display system based on a volume-scanning method is demonstrated. To form a three-dimensional real image, an inclined two-dimensional image is rapidly moved with a mirror scanner while the cross-section patterns of a three-dimensional object are displayed sequentially. A vector-scan CRT display unit is used to obtain a high-resolution image. An optical scanning system is constructed with concave mirrors and a galvanometer mirror. It is confirmed that three-dimensional images, formed by the experimental system, satisfy all the criteria for human stereoscopic vision.

Levy-Lieb-Based Monte Carlo Study of the Dimensionality Behaviour of the Electronic Kinetic Functional

Directory of Open Access Journals (Sweden)

Seshaditya A.

2017-06-01

Full Text Available We consider a gas of interacting electrons in the limit of nearly uniform density and treat the one dimensional (1D, two dimensional (2D and three dimensional (3D cases. We focus on the determination of the correlation part of the kinetic functional by employing a Monte Carlo sampling technique of electrons in space based on an analytic derivation via the Levy-Lieb constrained search principle. Of particular interest is the question of the behaviour of the functional as one passes from 1D to 3D; according to the basic principles of Density Functional Theory (DFT the form of the universal functional should be independent of the dimensionality. However, in practice the straightforward use of current approximate functionals in different dimensions is problematic. Here, we show that going from the 3D to the 2D case the functional form is consistent (concave function but in 1D becomes convex; such a drastic difference is peculiar of 1D electron systems as it is for other quantities. Given the interesting behaviour of the functional, this study represents a basic first-principle approach to the problem and suggests further investigations using highly accurate (though expensive many-electron computational techniques, such as Quantum Monte Carlo.

Three-dimensional bio-printing.

Science.gov (United States)

Gu, Qi; Hao, Jie; Lu, YangJie; Wang, Liu; Wallace, Gordon G; Zhou, Qi

2015-05-01

Three-dimensional (3D) printing technology has been widely used in various manufacturing operations including automotive, defence and space industries. 3D printing has the advantages of personalization, flexibility and high resolution, and is therefore becoming increasingly visible in the high-tech fields. Three-dimensional bio-printing technology also holds promise for future use in medical applications. At present 3D bio-printing is mainly used for simulating and reconstructing some hard tissues or for preparing drug-delivery systems in the medical area. The fabrication of 3D structures with living cells and bioactive moieties spatially distributed throughout will be realisable. Fabrication of complex tissues and organs is still at the exploratory stage. This review summarize the development of 3D bio-printing and its potential in medical applications, as well as discussing the current challenges faced by 3D bio-printing.

Acid-base properties of complexes with three-dimensional polyligands. Complexes with three-dimensional polyphosphoric acids

International Nuclear Information System (INIS)

Kopylova, V.D.; Bojko, Eh.T.; Saldadze, K.M.

1985-01-01

By the method of potentiometric titration acid-base properties of uranyl (2) complexes with three-dimensional polyphosphoric acids, KRF-8p, KF-1, KF-7 prepared by phosphorylation of copolymer of styrene and divinylbenzene or saponification of the copolymers of di-2,2'-chloroethyl ester of vinylphosphonic acid with divinyl benzene are studied. It is shown that in case of formation in the phase of three-dimensional polyphosphoric acids of UO 2 2+ complexes with the growth of bond covalence of metal ion-phosphonic group the acidjty of the second hydroxyl of the phosphonic group increases

Three dimensional periodic foundations for base seismic isolation

International Nuclear Information System (INIS)

Yan, Y; Mo, Y L; Cheng, Z; Shi, Z; Menq, F; Tang, Y

2015-01-01

Based on the concept of phononic crystals, periodic foundations made of periodic materials are investigated in this paper. The periodic foundations can provide low frequency band gaps, which cover the main frequency ranges of seismic waves. Therefore, the periodic foundations are able to protect the upper structures during earthquake events. In this paper, the basic theory of three dimensional periodic foundations is studied and the finite element method was used to conduct the sensitivity study. A simplified three-dimensional periodic foundation with a superstructure was tested in the field and the feasibility of three dimensional periodic foundations was proved. The test results showed that the response of the upper structure with the three dimensional periodic foundation was reduced under excitation waves with the main frequency falling in the attenuation zones. The finite element analysis results are consistent with the experimental data, indicating that three dimensional periodic foundations are a feasible way of reducing seismic vibrations. (paper)

Simulation on three dimensional bubble formation using MARS

International Nuclear Information System (INIS)

Kunugi, Tomoaki

1997-01-01

This paper describes a numerical simulation on three-dimensional bubble formation by means of the MARS (Multi-interfaces Advection and Reconstruction Solver) developed by the author. The comparison between two-dimensional and three-dimensional simulation on an agglomeration of two bubbles is discussed. Moreover, some simulation results regarding a phase change phenomena such as a boiling and condensation in a two dimensional enclosure with heated and cooled walls are presented. (author)

Development of LMR basic design technology - Development of 3-D multi-group nodal kinetics code for liquid metal reactors

Energy Technology Data Exchange (ETDEWEB)

Kim, Myung Hyun [Kyunghee University, Seoul (Korea, Republic of)

1996-07-01

A development project of 3-dimensional kinetics code for ALMR has three level of works. In the first level, a multi-group, nodal kinetics code for the HEX-Z geometry has been developed. A code showed very good results for the static analysis as well as for the kinetics problems. At the second level, a core thermal-hydraulic analysis code was developed for the temperature feedback calculation in ALMR transients analysis. This code is coupled with kinetics code. A sodium property table was programmed and tested to the KAERI data and thermal feedback model was developed and coupled in code. Benchmarking of T/H calculation has been performed and showed fairly good results. At the third level of research work, reactivity feedback model for structure thermal expansion is developed and added to the code. At present, basic model was studied. However, code development in now on going. Benchmarking of this model developed can not be done because of lack of data. 31 refs., 17 tabs., 38 figs. (author)

Improvement of neutron kinetics module in TRAC-BF1code: one-dimensional nodal collocation method

Energy Technology Data Exchange (ETDEWEB)

Jambrina, Ana; Barrachina, Teresa; Miro, Rafael; Verdu, Gumersindo, E-mail: ajambrina@iqn.upv.es, E-mail: tbarrachina@iqn.upv.es, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Universidade Politecnica de Valencia (UPV), Valencia (Spain); Soler, Amparo, E-mail: asoler@iberdrola.es [SEA Propulsion S.L., Madrid (Spain); Concejal, Alberto, E-mail: acbe@iberdrola.es [Iberdrola Ingenieria y Construcion S.A.U., Madrid (Spain)

2013-07-01

The TRAC-BF1 one-dimensional kinetic model is a formulation of the neutron diffusion equation in the two energy groups' approximation, based on the analytical nodal method (ANM). The advantage compared with a zero-dimensional kinetic model is that the axial power profile may vary with time due to thermal-hydraulic parameter changes and/or actions of the control systems but at has the disadvantages that in unusual situations it fails to converge. The nodal collocation method developed for the neutron diffusion equation and applied to the kinetics resolution of TRAC-BF1 thermal-hydraulics, is an adaptation of the traditional collocation methods for the discretization of partial differential equations, based on the development of the solution as a linear combination of analytical functions. It has chosen to use a nodal collocation method based on a development of Legendre polynomials of neutron fluxes in each cell. The qualification is carried out by the analysis of the turbine trip transient from the NEA benchmark in Peach Bottom NPP using both the original 1D kinetics implemented in TRAC-BF1 and the 1D nodal collocation method. (author)

Biodegradation of phenol with chromium(VI) reduction in an anaerobic fixed-biofilm process-Kinetic model and reactor performance

International Nuclear Information System (INIS)

Lin, Yen-Hui; Wu, Chih-Lung; Hsu, Chih-Hao; Li, Hsin-Lung

2009-01-01

A mathematical model system was derived to describe the simultaneous removal of phenol biodegradation with chromium(VI) reduction in an anaerobic fixed-biofilm reactor. The model system incorporates diffusive mass transport and double Monod kinetics. The model was solved using a combination of the orthogonal collocation method and Gear's method. A laboratory-scale column reactor was employed to validate the kinetic model system. Batch kinetic tests were conducted independently to evaluate the biokinetic parameters used in the model simulation. The removal efficiencies of phenol and chromium(VI) in an anaerobic fixed-biofilm process were approximately 980 mg/g and 910 mg/g, respectively, under a steady-state condition. In the steady state, model-predicted biofilm thickness reached up to 350 μm and suspended cells in the effluent were 85 mg cell/l. The experimental results agree closely with the results of the model simulations.

Performance of the coupled thermalhydraulics/neutron kinetics code R/P/C on workstation clusters and multiprocessor systems

International Nuclear Information System (INIS)

Hammer, C.; Paffrath, M.; Boeer, R.; Finnemann, H.; Jackson, C.J.

1996-01-01

The light water reactor core simulation code PANBOX has been coupled with the transient analysis code RELAP5 for the purpose of performing plant safety analyses with a three-dimensional (3-D) neutron kinetics model. The system has been parallelized to improve the computational efficiency. The paper describes the features of this system with emphasis on performance aspects. Performance results are given for different types of parallelization, i. e. for using an automatic parallelizing compiler, using the portable PVM platform on a workstation cluster, using PVM on a shared memory multiprocessor, and for using machine dependent interfaces. (author)

Advanced numerical methods for three dimensional two-phase flow calculations in PWR

International Nuclear Information System (INIS)

Toumi, I.; Gallo, D.; Royer, E.

1997-01-01

This paper is devoted to new numerical methods developed for three dimensional two-phase flow calculations. These methods are finite volume numerical methods. They are based on an extension of Roe's approximate Riemann solver to define convective fluxes versus mean cell quantities. To go forward in time, a linearized conservative implicit integrating step is used, together with a Newton iterative method. We also present here some improvements performed to obtain a fully implicit solution method that provides fast running steady state calculations. This kind of numerical method, which is widely used for fluid dynamic calculations, is proved to be very efficient for the numerical solution to two-phase flow problems. This numerical method has been implemented for the three dimensional thermal-hydraulic code FLICA-4 which is mainly dedicated to core thermal-hydraulic transient and steady-state analysis. Hereafter, we will also find some results obtained for the EPR reactor running in a steady-state at 60% of nominal power with 3 pumps out of 4, and a thermal-hydraulic core analysis for a 1300 MW PWR at low flow steam-line-break conditions. (author)

HEXBU-3D, a three-dimensional PWR-simulator program for hexagonal fuel assemblies

International Nuclear Information System (INIS)

Karvinen, E.

1981-06-01

HEXBU-3D is a three-dimensional nodal simulator program for PWR reactors. It is designed for a reactor core that consists of hexagonal fuel assemblies and of big follower-type control assemblies. The program solves two-group diffusion equations in homogenized fuel assembly geometry by a sophisticated nodal method. The treatment of feedback effects from xenon-poisoning, fuel temperature, moderator temperature and density and soluble boron concentration are included in the program. The nodal equations are solved by a fast two-level iteration technique and the eigenvalue can be either the effective multiplication factor or the boron concentration of the moderator. Burnup calculations are performed by tabulated sets of burnup-dependent cross sections evaluated by a cell burnup program. HEXBY-3D has been originally programmed in FORTRAN V for the UNIVAC 1108 computer, but there is also another version which is operable on the CDC CYBER 170 computer. (author)

Advanced numerical methods for three dimensional two-phase flow calculations

Energy Technology Data Exchange (ETDEWEB)

Toumi, I. [Laboratoire d`Etudes Thermiques des Reacteurs, Gif sur Yvette (France); Caruge, D. [Institut de Protection et de Surete Nucleaire, Fontenay aux Roses (France)

1997-07-01

This paper is devoted to new numerical methods developed for both one and three dimensional two-phase flow calculations. These methods are finite volume numerical methods and are based on the use of Approximate Riemann Solvers concepts to define convective fluxes versus mean cell quantities. The first part of the paper presents the numerical method for a one dimensional hyperbolic two-fluid model including differential terms as added mass and interface pressure. This numerical solution scheme makes use of the Riemann problem solution to define backward and forward differencing to approximate spatial derivatives. The construction of this approximate Riemann solver uses an extension of Roe`s method that has been successfully used to solve gas dynamic equations. As far as the two-fluid model is hyperbolic, this numerical method seems very efficient for the numerical solution of two-phase flow problems. The scheme was applied both to shock tube problems and to standard tests for two-fluid computer codes. The second part describes the numerical method in the three dimensional case. The authors discuss also some improvements performed to obtain a fully implicit solution method that provides fast running steady state calculations. Such a scheme is not implemented in a thermal-hydraulic computer code devoted to 3-D steady-state and transient computations. Some results obtained for Pressurised Water Reactors concerning upper plenum calculations and a steady state flow in the core with rod bow effect evaluation are presented. In practice these new numerical methods have proved to be stable on non staggered grids and capable of generating accurate non oscillating solutions for two-phase flow calculations.

Advanced numerical methods for three dimensional two-phase flow calculations

International Nuclear Information System (INIS)

Toumi, I.; Caruge, D.

1997-01-01

This paper is devoted to new numerical methods developed for both one and three dimensional two-phase flow calculations. These methods are finite volume numerical methods and are based on the use of Approximate Riemann Solvers concepts to define convective fluxes versus mean cell quantities. The first part of the paper presents the numerical method for a one dimensional hyperbolic two-fluid model including differential terms as added mass and interface pressure. This numerical solution scheme makes use of the Riemann problem solution to define backward and forward differencing to approximate spatial derivatives. The construction of this approximate Riemann solver uses an extension of Roe's method that has been successfully used to solve gas dynamic equations. As far as the two-fluid model is hyperbolic, this numerical method seems very efficient for the numerical solution of two-phase flow problems. The scheme was applied both to shock tube problems and to standard tests for two-fluid computer codes. The second part describes the numerical method in the three dimensional case. The authors discuss also some improvements performed to obtain a fully implicit solution method that provides fast running steady state calculations. Such a scheme is not implemented in a thermal-hydraulic computer code devoted to 3-D steady-state and transient computations. Some results obtained for Pressurised Water Reactors concerning upper plenum calculations and a steady state flow in the core with rod bow effect evaluation are presented. In practice these new numerical methods have proved to be stable on non staggered grids and capable of generating accurate non oscillating solutions for two-phase flow calculations

Three-dimensional tori and Arnold tongues

Energy Technology Data Exchange (ETDEWEB)

Sekikawa, Munehisa, E-mail: sekikawa@cc.utsunomiya-u.ac.jp [Department of Mechanical and Intelligent Engineering, Utsunomiya University, Utsunomiya-shi 321-8585 (Japan); Inaba, Naohiko [Organization for the Strategic Coordination of Research and Intellectual Property, Meiji University, Kawasaki-shi 214-8571 (Japan); Kamiyama, Kyohei [Department of Electronics and Bioinformatics, Meiji University, Kawasaki-shi 214-8571 (Japan); Aihara, Kazuyuki [Institute of Industrial Science, the University of Tokyo, Meguro-ku 153-8505 (Japan)

2014-03-15

This study analyzes an Arnold resonance web, which includes complicated quasi-periodic bifurcations, by conducting a Lyapunov analysis for a coupled delayed logistic map. The map can exhibit a two-dimensional invariant torus (IT), which corresponds to a three-dimensional torus in vector fields. Numerous one-dimensional invariant closed curves (ICCs), which correspond to two-dimensional tori in vector fields, exist in a very complicated but reasonable manner inside an IT-generating region. Periodic solutions emerge at the intersections of two different thin ICC-generating regions, which we call ICC-Arnold tongues, because all three independent-frequency components of the IT become rational at the intersections. Additionally, we observe a significant bifurcation structure where conventional Arnold tongues transit to ICC-Arnold tongues through a Neimark-Sacker bifurcation in the neighborhood of a quasi-periodic Hopf bifurcation (or a quasi-periodic Neimark-Sacker bifurcation) boundary.

The reactor kinetics code tank: a validation against selected SPERT-1b experiments