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 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
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
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)
The analysis of one-dimensional reactor kinetics benchmark computations
International Nuclear Information System (INIS)
Sidell, J.
1975-11-01
During March 1973 the European American Committee on Reactor Physics proposed a series of simple one-dimensional reactor kinetics problems, with the intention of comparing the relative efficiencies of the numerical methods employed in various codes, which are currently in use in many national laboratories. This report reviews the contributions submitted to this benchmark exercise and attempts to assess the relative merits and drawbacks of the various theoretical and computer methods. (author)
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)
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)
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)
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)
A stochastic physical-mathematical method for reactor kinetics analysis
International Nuclear Information System (INIS)
Velickovic, Lj.
1966-01-01
The developed theoretical model is concerned with BF 3 counter placed in the core of a low power reactor (a few MW) where statistical neutron effects are most evident. Our experiments were somewhat different. The detector used was and ionization chamber with double sampling, in ADC and in the time analyzer. The objective of this model was not to obtain precise numerical calculations, but to explain the method and the essentials of the correlation. Introducing all the six groups of delayed neutrons and possibly photoneutrons the model could be improved to obtained more realistic results
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)
Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12
Energy Technology Data Exchange (ETDEWEB)
Aji, Indarta Kuncoro, E-mail: indartaaji@s.itb.ac.id [Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesa 10 Bandung 40132 (Indonesia); Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Permana, Sidik [Nuclear Physics & Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesa 10 Bandung 40132 (Indonesia)
2015-09-30
Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF{sub 2}-ThF{sub 4}-{sup 233}UF{sub 4} respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 data library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155.
Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12
International Nuclear Information System (INIS)
Aji, Indarta Kuncoro; Waris, Abdul; Permana, Sidik
2015-01-01
Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF 2 -ThF 4 - 233 UF 4 respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 data library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155
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)
Kinetics Analysis of Synthesis Reaction of Struvite With Air-Flow Continous Vertical Reactors
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.
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.
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
Energy Technology Data Exchange (ETDEWEB)
Gonnelli, Eduardo; Diniz, Ricardo [Instituto de Pesquisas Energéticas e Nucleares - IPEN/CNEN-SP Travessa R-400, 05508-900, Cidade Universitária, São Paulo (Brazil)
2014-11-11
This is a complementary work about the behavior analysis of the neutron lifetimes that was developed in the IPEN/MB-01 nuclear reactor facility. The macroscopic neutron noise technique was experimentally employed using pulse mode detectors for two stages of control rods insertion, where a total of twenty levels of subcriticality have been carried out. It was also considered that the neutron reflector density was treated as an additional group of delayed neutrons, being a sophisticated approach in the two-region kinetic theoretical model.
International Nuclear Information System (INIS)
Gonnelli, Eduardo; Diniz, Ricardo
2014-01-01
This is a complementary work about the behavior analysis of the neutron lifetimes that was developed in the IPEN/MB-01 nuclear reactor facility. The macroscopic neutron noise technique was experimentally employed using pulse mode detectors for two stages of control rods insertion, where a total of twenty levels of subcriticality have been carried out. It was also considered that the neutron reflector density was treated as an additional group of delayed neutrons, being a sophisticated approach in the two-region kinetic theoretical model
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
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.
Model description and kinetic parameter analysis of MTBE biodegradation in a packed bed reactor
DEFF Research Database (Denmark)
Waul, Christopher Kevin; Arvin, Erik; Schmidt, Jens Ejbye
2008-01-01
A dynamic modeling approach was used to estimate in-situ model parameters, which describe the degradation of methyl tert-butyl ether (MTBE) in a laboratory packed bed reactor. The measured dynamic response of MTBE pulses injected at the reactor's inlet was analyzed by least squares and parameter...
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)
Nuclear reactor kinetics and plant control
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
Introduction to the neutron kinetics of nuclear power reactors
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
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)
International Nuclear Information System (INIS)
Paixao, S.B.; Marzo, M.A.S.; Alvim, A.C.M.
1986-01-01
The calculation method used in WIGLE code is studied. Because of the non availability of such a praiseworthy solution, expounding the method minutely has been tried. This developed method has been applied for the solution of the one-dimensional, two-group, diffusion equations in slab, axial analysis, including non-boiling heat transfer, accountig for feedback. A steady-state program (CITER-1D), written in FORTRAN 4, has been implemented, providing excellent results, ratifying the developed work quality. (Author) [pt
International Nuclear Information System (INIS)
Van Rooijen, W. F. G.; Lathouwers, D.
2007-01-01
In advanced Generation IV (fast) reactors an integral fuel cycle is envisaged, where all Heavy Metal is recycled in the reactor. This leads to a nuclear fuel with a considerable content of Minor Actinides. For many of these isotopes the nuclear data is not very well known. In this paper the sensitivity of the kinetic behaviour of the reactor to the dynamic parameters λ k , β k and the delayed spectrum χ d,k is studied using first order perturbation theory. In the current study, feedback due to Doppler and/or thermohydraulic effects are not treated. The theoretical framework is applied to a Generation IV Gas Cooled Fast Reactor. The results indicate that the first-order approach is satisfactory for small variations of the data. Sensitivities to delayed neutron data are similar for increasing and decreasing transients. Sensitivities generally increase with reactivity for increasing transients. For decreasing transients, there are less clearly defined trends, although the sensitivity to the delayed neutron spectrum decreases with larger sub-criticality, as expected. For this research, an adjoint capable version of the time-dependent diffusion code DALTON is under development. (authors)
RELAP5 kinetics model development for the Advanced Test Reactor
International Nuclear Information System (INIS)
Judd, J.L.; Terry, W.K.
1990-01-01
A point-kinetics model of the Advanced Test Reactor has been developed for the RELAP5 code. Reactivity feedback parameters were calculated by a three-dimensional analysis with the PDQ neutron diffusion code. Analyses of several hypothetical reactivity insertion events by the new model and two earlier models are discussed. 3 refs., 10 figs., 6 tabs
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)
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
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
International Nuclear Information System (INIS)
Ghoshal, Sanjukta; Bhattacharya, Pinaki; Chowdhury, Ranjana
2011-01-01
Graphical abstract: The assembly of biofilm reactor, based on attached growth of Bacillus cereus (JUBT1) on rice husk packing, and an activated carbon filter has been able to ensure the removal of mercury up to near-zero level. Highlights: → A new mercury resistant bacterial strain, Bacillus cereus (JUBT1), has been isolated. → Growth kinetics has been determined. → Biofilm reactor using attached growth of bacteria ensures near-zero level of mercury. → Confinement of mercury is confirmed through energy dispersive spectrometric analysis. - Abstract: Removal of mercuric ions by a mercury resistant bacteria, called Bacillus cereus (JUBT1), isolated from the sludge of a local chlor-alkali industry, has been investigated. Growth kinetics of the bacteria have been determined. A multiplicative, non-competitive relationship between sucrose and mercury ions has been observed with respect to bacterial growth. A combination of biofilm reactor, using attached growth of Bacillus cereus (JUBT1) on rice husk packing, and an activated carbon filter has been able to ensure the removal of mercury up to near-zero level. Energy dispersive spectrometry analysis of biofilm and the activated carbon has proved the transformation of Hg 2+ to Hg 0 and its confinement in the system.
Energy Technology Data Exchange (ETDEWEB)
Jeong, J. J.; Chung, B. D.; Lee, W.J
2005-02-01
The subchannel analysis capability of the MARS 3D module has been improved. Especially, the turbulent mixing and void drift models for flow mixing phenomena in rod bundles have been assessed using some well-known rod bundle test data. Then, the subchannel analysis feature was combined to the existing coupled 'system Thermal-Hydraulics (T/H) and 3D reactor kinetics' calculation capability of MARS. These features allow the coupled 'system T/H, 3D reactor kinetics, and hot channel' analysis capability and, thus, realistic simulations of hot channel behavior as well as global system T/H behavior. In this report, the MARS code features for the coupled analysis capability are described first. The code modifications relevant to the features are also given. Then, a coupled analysis of the Main Steam Line Break (MSLB) is carried out for demonstration. The results of the coupled calculations are very reasonable and realistic, and show these methods can be used to reduce the over-conservatism in the conventional safety analysis.
Calculation of Kinetic Parameters of TRIGA Reactor
International Nuclear Information System (INIS)
Snoj, Luka; Kavcic, Andrej; Zerovnik, Gasper; Ravnik, Matjaz
2008-01-01
Modern Monte Carlo transport codes in combination of fast computer clusters enable very accurate calculations of the most important reactor kinetic parameters. Such are the effective delayed neutron fraction, β eff , and mean neutron generation time, Λ. We calculated the β eff and Λ for various realistic and hypothetical annular TRIGA Mark II cores with different types and amount of fuel. It can be observed that the effective delayed neutron fraction strongly depends on the number of fuel elements in the core or on the core size. E.g., for 12 wt. % uranium standard fuel with 20 % enrichment, β eff varies from 0.0080 for a small core (43 fuel rods) to 0.0075 for a full core (90 fuel rods). It is interesting to note that calculated value of β eff strongly depends also on the delayed neutron nuclear data set used in calculations. The prompt neutron life-time mainly depends on the amount (due to either content or enrichment) of 235 U in the fuel as it is approximately inversely proportional to the average absorption cross-section of the fuel. E.g., it varies from 28 μs for 30 wt. % uranium content fuelled core to 48 μs for 8.5 wt. % uranium content LEU fuelled core. The results are especially important for pulse mode operation and analysis of the pulses. (authors)
Sandia reactor kinetics codes: SAK and PK1D
International Nuclear Information System (INIS)
Pickard, P.S.; Odom, J.P.
1978-01-01
The Sandia Kinetics code (SAK) is a one-dimensional coupled thermal-neutronics transient analysis code for use in simulation of reactor transients. The time-dependent cross section routines allow arbitrary time-dependent changes in material properties. The one-dimensional heat transfer routines are for cylindrical geometry and allow arbitrary mesh structure, temperature-dependent thermal properties, radiation treatment, and coolant flow and heat-transfer properties at the surface of a fuel element. The Point Kinetics 1 Dimensional Heat Transfer Code (PK1D) solves the point kinetics equations and has essentially the same heat-transfer treatment as SAK. PK1D can address extended reactor transients with minimal computer execution time
International Nuclear Information System (INIS)
Ampomah-Amoako, Emmanuel; Akaho, Edward H.K.; Nyarko, Benjamin J.B.; Ambrosini, Walter
2013-01-01
Highlights: • The analysis of flow stability of nuclear fuel subchannels with supercritical water is presented. • The results obtained by a CFD code are compared with those of a system code. • The model includes also heat conduction in the fuel rod and point neutron kinetics. - Abstract: The paper presents the analysis by a CFD code of coupled neutronic–thermal hydraulic instabilities in a subchannel slice belonging to a square lattice assembly. The work represents a further phase in the assessment of the suitability of CFD codes for studies of flow stability of supercritical fluids; the research started in previous work with the analysis of bare 2D circular pipes and already addressed 3D subchannel slices with no allowance for heat conduction or neutronic effects. In the present phase, a more realistic system is considered, dealing with a slice of a fuel assembly subchannel containing the regions of the pellet, the gap and the cladding and including also the effect of inlet and outlet throttling. The adopted neutronic model is a point kinetics one, including six delayed neutron groups with global Doppler and fluid density feedbacks. The response of the model to perturbations applied starting from a steady-state condition at the rated power is compared with that of a similar model developed for a 1D system code. Upward, horizontal and downward flow orientations are addressed making use of a uniform power profile and changing relevant parameters as the gap equivalent conductance and the density reactivity coefficient. A bottom-peaked power profile is also considered in the case of vertical upward flow. Though the adopted model can still be considered simple in comparison with actual details of fuel assemblies, the obtained results demonstrate the potential of the adopted methodology for more accurate analyses to be made with larger computational resources
Energy Technology Data Exchange (ETDEWEB)
Jin, Yung Kwon; Choi, Chul Jin; Kim, Eun Kee; Lee, Sang Yong [Korea Power Engineering Company, Inc, 150 Deokjin-dong, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)
2006-07-01
This paper presents the results of the coupled 3-D neutronics/thermal-hydraulic analysis of hypothetical main steam line break (MSLB) accident for Optimized Power Reactor 1000. One of the major concerns of this accident is a return-to-power occurrence accompanied with extremely large radial peaking near the stuck Control Element Assembly (CEA). The conventional point kinetics application does not properly account for this kind of asymmetric and local core behavior. Therefore, the current licensing method of point kinetics application introduces some uncertainties and conservatisms in the physics parameters generation, e.g., the static net scram rod worth, moderator cooldown reactivity, Doppler reactivity, and a 3-D peaking factor. The recently developed UNICORN-TM code system is applied for the 3-D coupled calculation, where neutronics code MASTER is coupled with the best-estimate system transient code RETRAN. The 3-D coupled results were assessed in comparison with those by point kinetics application using stand-alone RETRAN application. To quantify the 3-D reactivity benefits over point kinetics, both calculations assumed the accidents to be initiated from the same core state, e.g., end of cycle burnup, fuel and CEA configuration with the same initial moderator and Doppler temperature coefficient, and with initial system thermal-hydraulic condition. The core physics parameters required for point kinetics application were produced using MASTER with the method and procedure consistent with the current licensing application. The occurrence of return-to-power was simulated by intentionally reducing the net CEA worth in order to assess the spatial power distribution and local T-H effect on the dynamic reactivity feedback. The results have demonstrated that the 3-D analysis removes some of the conservatisms inherent in point kinetics analysis mainly caused by the inability to properly account for local reactivity feedback effects during return-to-power transient
International Nuclear Information System (INIS)
Jin, Yung Kwon; Choi, Chul Jin; Kim, Eun Kee; Lee, Sang Yong
2006-01-01
This paper presents the results of the coupled 3-D neutronics/thermal-hydraulic analysis of hypothetical main steam line break (MSLB) accident for Optimized Power Reactor 1000. One of the major concerns of this accident is a return-to-power occurrence accompanied with extremely large radial peaking near the stuck Control Element Assembly (CEA). The conventional point kinetics application does not properly account for this kind of asymmetric and local core behavior. Therefore, the current licensing method of point kinetics application introduces some uncertainties and conservatisms in the physics parameters generation, e.g., the static net scram rod worth, moderator cooldown reactivity, Doppler reactivity, and a 3-D peaking factor. The recently developed UNICORN-TM code system is applied for the 3-D coupled calculation, where neutronics code MASTER is coupled with the best-estimate system transient code RETRAN. The 3-D coupled results were assessed in comparison with those by point kinetics application using stand-alone RETRAN application. To quantify the 3-D reactivity benefits over point kinetics, both calculations assumed the accidents to be initiated from the same core state, e.g., end of cycle burnup, fuel and CEA configuration with the same initial moderator and Doppler temperature coefficient, and with initial system thermal-hydraulic condition. The core physics parameters required for point kinetics application were produced using MASTER with the method and procedure consistent with the current licensing application. The occurrence of return-to-power was simulated by intentionally reducing the net CEA worth in order to assess the spatial power distribution and local T-H effect on the dynamic reactivity feedback. The results have demonstrated that the 3-D analysis removes some of the conservatisms inherent in point kinetics analysis mainly caused by the inability to properly account for local reactivity feedback effects during return-to-power transient
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
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.
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)
Identification of nuclear reactor characteristics by the reactor noise analysis
International Nuclear Information System (INIS)
Yashima, Hideyuki
1980-01-01
Reactor noise analysis method was applied to TRIGA II Research Reactor (Atomic Research Laboratory, Musashi Institute of Technology) and computed power spectral density (PSD) from the CIC current record. PSD has provided many valuable informations regarding to the reactor kinetics, including the effect of control rods vibration. Another information of neutron physics parameters were obtained and this result was compared with the parameter which was formerly measured by the Feynman-α experiment. Through these experiments we could find overall frequency characteristics of TRIGA II Reactor. (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.)
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.
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
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....
Energy Technology Data Exchange (ETDEWEB)
Velickovic, Lj [Institut za nuklearne nauke ' Boris Kidric' , Vinca, Belgrade (Yugoslavia)
1966-07-01
The developed theoretical model is concerned with BF{sub 3} counter placed in the core of a low power reactor (a few MW) where statistical neutron effects are most evident. Our experiments were somewhat different. The detector used was and ionization chamber with double sampling, in ADC and in the time analyzer. The objective of this model was not to obtain precise numerical calculations, but to explain the method and the essentials of the correlation. Introducing all the six groups of delayed neutrons and possibly photoneutrons the model could be improved to obtained more realistic results.
International Nuclear Information System (INIS)
Arien, B.
1998-01-01
The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of four main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents, the development of an expert system for the aid to diagnosis; the development and application of a probabilistic reactor dynamics method. Main achievements in 1999 are reported
Shao, Xiongjun; Lynd, Lee; Wyman, Charles; Bakker, André
2009-01-01
The model of South et al. [South et al. (1995) Enzyme Microb Technol 17(9): 797-803] for simultaneous saccharification of fermentation of cellulosic biomass is extended and modified to accommodate intermittent feeding of substrate and enzyme, cascade reactor configurations, and to be more computationally efficient. A dynamic enzyme adsorption model is found to be much more computationally efficient than the equilibrium model used previously, thus increasing the feasibility of incorporating the kinetic model in a computational fluid dynamic framework in the future. For continuous or discretely fed reactors, it is necessary to use particle conversion in conversion-dependent hydrolysis rate laws rather than reactor conversion. Whereas reactor conversion decreases due to both reaction and exit of particles from the reactor, particle conversion decreases due to reaction only. Using the modified models, it is predicted that cellulose conversion increases with decreasing feeding frequency (feedings per residence time, f). A computationally efficient strategy for modeling cascade reactors involving a modified rate constant is shown to give equivalent results relative to an exhaustive approach considering the distribution of particles in each successive fermenter.
RB reactor noise analysis; Analiza sumova reaktora RB
Energy Technology Data Exchange (ETDEWEB)
Petrovic, M; Velickovic, Lj; Markovic, V; Jovanovic, S [Institut za nuklearne nauke Boris Kidric, Vinca, Beograd (Yugoslavia)
1964-07-01
Statistical fluctuations of reactivity represent reactor noise. Analysis of reactor noise enables determining a series of reactor kinetic parameters. Fluctuations of power was measured by ionization chamber placed next to the tank of the RB reactor. The signal was digitized by an analog-digital converter. After calculation of the mean power, 3000 data obtained by sampling were analysed.
International Nuclear Information System (INIS)
Arien, B.
2000-01-01
The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of two main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents. Main achievements in 1999 are reported
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.
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
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)
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.)
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.
One-dimensional reactor kinetics model for RETRAN
International Nuclear Information System (INIS)
Gose, G.C.; Peterson, C.E.; Ellis, N.L.; McClure, J.A.
1981-01-01
This paper describes a one-dimensional spatial neutron kinetics model that was developed for the RETRAN code. The RETRAN -01 code has a point kinetics model to describe the reactor core behavior during thermal-hydraulic transients. A one-dimensional neutronics model has been developed for RETRAN-02. The ability to account for flux shape changes will permit an improved representation of the thermal and hydraulic feedback effects for many operational transients. 19 refs
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
Adsorption analysis equilibria and kinetics
Do, Duong D
1998-01-01
This book covers topics of equilibria and kinetics of adsorption in porous media. Fundamental equilibria and kinetics are dealt with for homogeneous as well as heterogeneous particles. Five chapters of the book deal with equilibria and eight chapters deal with kinetics. Single component as well as multicomponent systems are discussed. In kinetics analysis, we deal with the various mass transport processes and their interactions inside a porous particle. Conventional approaches as well as the new approach using Maxwell-Stefan equations are presented. Various methods to measure diffusivity, such
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
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
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)
Kinetics of two phase fuel reflected reactors
International Nuclear Information System (INIS)
Buzano, M.L.; Corno, S.E.; Mattioda, F.
2000-01-01
In the present work a self-consistent mathematical model for the local dynamics of a quite particular class of fission reactors has been developed and solved. These devices consist of an innermost multiplying region, in which a significant fraction of the fissile fuel is diluted into a liquid phase, while the complementary fuel fraction operates as a standing solid matrix. This unconventional active region is surrounded by a standard peripheral reflector. For cooling purposes, the fluid fraction of the fuel needs to be circulated through external heat exchangers. The pump driven circulation causes the delayed neutron precursors, dissolved inside the fluid phase, to be spatially homogenized in the core volume well before decaying, while a continuous removal of precursor nuclei from the core takes place as a consequence of the outside circulation. Furthermore, the fraction of the extracted precursors still surviving after the solenoidal trip through the heat exchangers is continuously reinserted into the core. A new type of dynamical model is required to account for these unusual technological features. The mathematical structure of the evolution model presented in this paper consists of a system of integro-differential-difference equations, whose solution is derived in closed-form, by means of fully analytical techniques. Many dynamics and safety features of reactors of this type can be clarified a priori, upon inspection of the mathematical properties of the solution of the model. The rigorous time-eigenvalue generating equation can be explicitly established in the present theoretical context, together with the evaluation of any kind of transients. A short survey on the possible fields of application of these reactors is also presented
International Nuclear Information System (INIS)
Arien, B.
1998-01-01
Risk assessments of nuclear installations require accurate safety and reliability analyses to estimate the consequences of accidental events and their probability of occurrence. The objective of the work performed in this field at the Belgian Nuclear Research Centre SCK-CEN is to develop expertise in probabilistic and deterministic reactor safety analysis. The four main activities of the research project on reactor safety analysis are: (1) the development of software for the reliable analysis of large systems; (2) the development of an expert system for the aid to diagnosis; (3) the development and the application of a probabilistic reactor-dynamics method, and (4) to participate in the international PHEBUS-FP programme for severe accidents. Progress in research during 1997 is described
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)
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.
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)
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)
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)
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
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.
Reactor operational transient analysis
International Nuclear Information System (INIS)
Shin, W.K.; Chae, S.K.; Han, K.I.; Yang, K.S.; Chung, H. D.; Kim, H.G.; Moon, H.J.; Ryu, Y.H.
1983-01-01
To build up efficient capability of safety review and inspection for the nuclear power plants, four area of studies have performed as follows: 1) In order to search the most optimized operating method during load follow operating schemes, automatic control and normal control, are compared each other under the CAOC condition. The analysis performed by DDID code has shown that the reactor has to be controlled by the operator manually during load follow operation. 2) Through the sensitivity analysis by COBRA code, the operating parameters, such as coolant pressure, flow rate, inlet temperature, and power distribution are shown to be important to the determination of DNBR. Expecially, inlet temperature of primary coolant system is appeared as the most senstive parameter on DNBR. 3) FRAPCON code is adapted to study the sensitivity of several operational parameters on the mechanical properties of reactor fuel rod. 4) The calculations procedure which is required to be obtained the neutron fluence at the reactor vessel and the spectrum at the surveillance capsule is established. The results of computation are conpared with those of FSAR and SWRI report and proved its applicability to reactor surveillance program. (Author)
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)
Kinetic parameters from thermogravimetric analysis
Kiefer, Richard L.
1993-01-01
High performance polymeric materials are finding increased use in aerospace applications. Proposed high speed aircraft will require materials to withstand high temperatures in an oxidative atmosphere for long periods of time. It is essential that accurate estimates be made of the performance of these materials at the given conditions of temperature and time. Temperatures of 350 F (177 C) and times of 60,000 to 100,000 hours are anticipated. In order to survey a large number of high performance polymeric materials on a reasonable time scale, some form of accelerated testing must be performed. A knowledge of the rate of a process can be used to predict the lifetime of that process. Thermogravimetric analysis (TGA) has frequently been used to determine kinetic information for degradation reactions in polymeric materials. Flynn and Wall studied a number of methods for using TGA experiments to determine kinetic information in polymer reactions. Kinetic parameters, such as the apparent activation energy and the frequency factor, can be determined in such experiments. Recently, researchers at the McDonnell Douglas Research Laboratory suggested that a graph of the logarithm of the frequency factor against the apparent activation energy can be used to predict long-term thermo-oxidative stability for polymeric materials. Such a graph has been called a kinetic map. In this study, thermogravimetric analyses were performed in air to study the thermo-oxidative degradation of several high performance polymers and to plot their kinetic parameters on a kinetic map.
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
Mu, Lin; Chen, Jianbiao; Yao, Pikai; Zhou, Dapeng; Zhao, Liang; Yin, Hongchao
2016-12-01
Co-pyrolysis characteristics of petrochemical wastewater sludge and Huolinhe lignite were investigated using thermogravimetric analyzer and packed-bed reactor coupled with Fourier transform infrared spectrometer and gas chromatography. The pyrolysis characteristics of the blends at various sludge blending ratios were compared with those of the individual materials. Thermogravimetric experiments showed that the interactions between the blends were beneficial to generate more residues. In packed-bed reactor, synergetic effects promoted the release of gas products and left less liquid and solid products than those calculated by additive manner. Fourier transform infrared spectrometer analysis showed that main functional groups in chars gradually disappeared with pyrolysis temperatures increasing, and H 2 O, CH 4 , CO, and CO 2 appeared in volatiles during pyrolysis. Gas compositions analysis indicated that, the yields of H 2 and CO clearly increased as the pyrolysis temperature and sludge blending ratio increasing, while the changes of CH 4 and CO 2 yields were relatively complex. Copyright © 2016 Elsevier Ltd. All rights reserved.
One-dimensional reactor kinetics model for RETRAN
International Nuclear Information System (INIS)
Gose, G.C.; Peterson, C.E.; Ellis, N.L.; McClure, J.A.
1981-01-01
Previous versions of RETRAN have had only a point kinetics model to describe the reactor core behavior during thermal-hydraulic transients. The principal assumption in deriving the point kinetics model is that the neutron flux may be separated into a time-dependent amplitude funtion and a time-independent shape function. Certain types of transients cannot be correctly analyzed under this assumption, since proper definitions for core average quantities such as reactivity or lifetime include the inner product of the adjoint flux with the perturbed flux. A one-dimensional neutronics model has been included in a preliminary version of RETRAN-02. The ability to account for flux shape changes will permit an improved representation of the thermal and hydraulic feedback effects. This paper describes the neutronics model and discusses some of the analyses
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
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
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)
Linear and nonlinear stability analysis, associated to experimental fast reactors
International Nuclear Information System (INIS)
Amorim, E.S. do; Moura Neto, C. de; Rosa, M.A.P.
1980-07-01
Phenomena associated to the physics of fast neutrons were analysed by linear and nonlinear Kinetics with arbitrary feedback. The theoretical foundations of linear kinetics and transfer functions aiming at the analysis of fast reactors stability, are established. These stability conditions were analitically proposed and investigated by digital and analogic programs. (E.G.) [pt
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}.
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)
Supercritical kinetic analysis in simplified system of fuel debris using integral kinetic model
International Nuclear Information System (INIS)
Tuya, Delgersaikhan; Obara, Toru
2016-01-01
Highlights: • Kinetic analysis in simplified weakly coupled fuel debris system was performed. • The integral kinetic model was used to simulate criticality accidents. • The fission power and released energy during simulated accident were obtained. • Coupling between debris regions and its effect on the fission power was obtained. - Abstract: Preliminary prompt supercritical kinetic analyses in a simplified coupled system of fuel debris designed to roughly resemble a melted core of a nuclear reactor were performed using an integral kinetic model. The integral kinetic model, which can describe region- and time-dependent fission rate in a coupled system of arbitrary geometry, was used because the fuel debris system is weakly coupled in terms of neutronics. The results revealed some important characteristics of coupled systems, such as the coupling between debris regions and the effect of the coupling on the fission rate and released energy in each debris region during the simulated criticality accident. In brief, this study showed that the integral kinetic model can be applied to supercritical kinetic analysis in fuel debris systems and also that it can be a useful tool for investigating the effect of the coupling on consequences of a supercritical accident.
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.
Simulation of styrene polymerization reactors: kinetic and thermodynamic modeling
Directory of Open Access Journals (Sweden)
A. S. Almeida
2008-06-01
Full Text Available A mathematical model for the free radical polymerization of styrene is developed to predict the steady-state and dynamic behavior of a continuous process. Special emphasis is given for the kinetic and thermodynamic models, where the most sensitive parameters were estimated using data from an industrial plant. The thermodynamic model is based on a cubic equation of state and a mixing rule applied to the low-pressure vapor-liquid equilibrium of polymeric solutions, suitable for modeling the auto-refrigerated polymerization reactors, which use the vaporization rate to remove the reaction heat from the exothermic reactions. The simulation results show the high predictive capability of the proposed model when compared with plant data for conversion, average molecular weights, polydispersity, melt flow index, and thermal properties for different polymer grades.
Experimental and Kinetic Modeling Study of Ethyl Levulinate Oxidation in a Jet-Stirred Reactor
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).
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)
Safety analysis for research reactors
International Nuclear Information System (INIS)
2008-01-01
The aim of safety analysis for research reactors is to establish and confirm the design basis for items important to safety using appropriate analytical tools. The design, manufacture, construction and commissioning should be integrated with the safety analysis to ensure that the design intent has been incorporated into the as-built reactor. Safety analysis assesses the performance of the reactor against a broad range of operating conditions, postulated initiating events and other circumstances, in order to obtain a complete understanding of how the reactor is expected to perform in these situations. Safety analysis demonstrates that the reactor can be kept within the safety operating regimes established by the designer and approved by the regulatory body. This analysis can also be used as appropriate in the development of operating procedures, periodic testing and inspection programmes, proposals for modifications and experiments and emergency planning. The IAEA Safety Requirements publication on the Safety of Research Reactors states that the scope of safety analysis is required to include analysis of event sequences and evaluation of the consequences of the postulated initiating events and comparison of the results of the analysis with radiological acceptance criteria and design limits. This Safety Report elaborates on the requirements established in IAEA Safety Standards Series No. NS-R-4 on the Safety of Research Reactors, and the guidance given in IAEA Safety Series No. 35-G1, Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, providing detailed discussion and examples of related topics. Guidance is given in this report for carrying out safety analyses of research reactors, based on current international good practices. The report covers all the various steps required for a safety analysis; that is, selection of initiating events and acceptance criteria, rules and conventions, types of safety analysis, selection of
RIA Analysis of Unprotected TRIGA Reactor
Directory of Open Access Journals (Sweden)
M.H. Altaf
2017-07-01
Full Text Available An RIA (reactivity initiated accident analysis has been carried out for the TRIGA Mark II research reactor considering both step and ramp reactivity ranges within 0.5 % dk/k (< $1 to 2.0 % dk/k (>$2. The insertion time was set at 10 s. Based on the fact that a reactor becomes unprotected if scram does not work at the event of danger, to define unprotected conditions, the time to actuate scram (trip was taken as close to total simulation time. In this long duration of scram inactivity, it is obtained from the present analysis that the reactor remained safe to up to 1.8 % dk/k ($2.57 for step reactivity and 1.99 % dk/k ($2.84 for ramp reactivity. In addition to negative temperature coefficient of reativity, probably the longer time of reactivity insertion keeps TRIGA safe even at larger magnitudes of reactivity during unprotected reactor transients. Coupled point kinetics, neutronics, and thermal hydraulics code EUREKA-2/R has been utilized for this work. It appears that EUREKA-2/RR predicts the sequence of unprotected transient scenario of TRIGA core with good approximation and the results will definitely be helpful for the reactor operators.
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)
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)
Analysis of kinetic reaction mechanisms
Turányi, Tamás
2014-01-01
Chemical processes in many fields of science and technology, including combustion, atmospheric chemistry, environmental modelling, process engineering, and systems biology, can be described by detailed reaction mechanisms consisting of numerous reaction steps. This book describes methods for the analysis of reaction mechanisms that are applicable in all these fields. Topics addressed include: how sensitivity and uncertainty analyses allow the calculation of the overall uncertainty of simulation results and the identification of the most important input parameters, the ways in which mechanisms can be reduced without losing important kinetic and dynamic detail, and the application of reduced models for more accurate engineering optimizations. This monograph is invaluable for researchers and engineers dealing with detailed reaction mechanisms, but is also useful for graduate students of related courses in chemistry, mechanical engineering, energy and environmental science and biology.
Analysis of water cooled reactors stability; Analiza stabilnosti reaktorskih sistema hladjenih vodom
Energy Technology Data Exchange (ETDEWEB)
Marinkovic, P; Pesic, M [Boris Kidric Institute of Nuclear Sciences, Vinca, Belgrade (Yugoslavia)
1980-07-01
A model for stability analysis of non-boiling water cooled nuclear system is developed. The model is based on linear reactor kinetics and space averaged heat transfer in reactor and heat-exchanger. The transfer functions are defined and the analysis was applied to nuclear reactor RA at 'Boris Kidric' Institute - Vinca. (author)
Parallelised Krylov subspace method for reactor kinetics by IQS approach
International Nuclear Information System (INIS)
Gupta, Anurag; Modak, R.S.; Gupta, H.P.; Kumar, Vinod; Bhatt, K.
2005-01-01
Nuclear reactor kinetics involves numerical solution of space-time-dependent multi-group neutron diffusion equation. Two distinct approaches exist for this purpose: the direct (implicit time differencing) approach and the improved quasi-static (IQS) approach. Both the approaches need solution of static space-energy-dependent diffusion equations at successive time-steps; the step being relatively smaller for the direct approach. These solutions are usually obtained by Gauss-Seidel type iterative methods. For a faster solution, the Krylov sub-space methods have been tried and also parallelised by many investigators. However, these studies seem to have been done only for the direct approach. In the present paper, parallelised Krylov methods are applied to the IQS approach in addition to the direct approach. It is shown that the speed-up obtained for IQS is higher than that for the direct approach. The reasons for this are also discussed. Thus, the use of IQS approach along with parallelised Krylov solvers seems to be a promising scheme
Reactor surveillance by noise analysis
International Nuclear Information System (INIS)
Ciftcioglu, Ozer
1988-01-01
A real-time noise analysis system is designed for the TRIGA reactor at Istanbul Technical University. By means of the noise techniques, reactor surveillance is performed together with failure diagnosis. The fast data processing is carried out by FFT in real-time so that malfunction or non-stationary operation of the reactor in long term can be identified by comparing the noise power spectra with the corresponding reference patterns while the decision making procedure is accomplished by the method of hypothesis testing. The system being computer based safety instrumentation involves CAMAC in conjunction with the RT-11 (PDP-11) single user dedicated environment. (author)
Kinetic parameters of the RB and RA reactors
Energy Technology Data Exchange (ETDEWEB)
Petrovic, M; Obradovic, D [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)
1965-12-15
In the paper the expressions for transfer functions of the zero power reactors, as well as power reactors of the RA reactor type are given, based on the space independent model. The modulation method for reactor transfer function measurements is explained. The results of the measurement and interpretation are given. The measurement were done on the RB and RA reactors in 'Boris Kidrich' Institute for Nuclear Sciences in Vincha (author)
Methods for solving the stochastic point reactor kinetic equations
International Nuclear Information System (INIS)
Quabili, E.R.; Karasulu, M.
1979-01-01
Two new methods are presented for analysis of the statistical properties of nonlinear outputs of a point reactor to stochastic non-white reactivity inputs. They are Bourret's approximation and logarithmic linearization. The results have been compared with the exact results, previously obtained in the case of Gaussian white reactivity input. It was found that when the reactivity noise has short correlation time, Bourret's approximation should be recommended because it yields results superior to those yielded by logarithmic linearization. When the correlation time is long, Bourret's approximation is not valid, but in that case, if one can assume the reactivity noise to be Gaussian, one may use the logarithmic linearization. (author)
Experimental and Kinetic Modeling Study of Ethyl Levulinate Oxidation in a Jet-Stirred Reactor
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
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
The shock tube as wave reactor for kinetic studies and material systems
Energy Technology Data Exchange (ETDEWEB)
Bhaskaran, K.A. [Indian Institute of Technology, Chennai (India). Department of Mechanical Engineering; Roth, P. [Gerhard Mercator Universitat, Duisberg (Germany). Institut fur Verbrennung und Gasdynamik
2002-07-01
Several important reviews of shock tube kinetics have appeared earlier, prominent among them being 'Shock Tube Technique in Chemical Kinetics' by Belford and Strehlow (Ann Rev Phys Chem 20 (1969) 247), 'Chemical Reaction of Shock Waves' by Wagner (Proceedings of the Eighth International Shock Tube Symposium (1971) 4/1), 'Shock Tube and Shock Wave Research' by Bauer and Lewis (Proceedings of the 11th International Symposium on Shock Tubes and Waves (1977) 269), 'Shock Waves in Chemistry' edited by Assa Lifshitz (Shock Waves in Chemistry, 1981) and 'Shock Tube Techniques in Chemical Kinetics' by Wing Tsang and Assa Lifshitz (Annu Rev Phys Chem 41 (1990) 559). A critical analysis of the different shock tube techniques, their limitations and suggestions to improve the accuracy of the data produced are contained in these reviews. The purpose of this article is to present the current status of kinetic research with emphasis on the diagnostic techniques. Selected studies on homogeneous and dispersed systems are presented to bring out the versatility of the shock tube technique. The use of the shock tube as high temperature wave reactor for gas phase material synthesis is also highlighted. (author)
Energy Technology Data Exchange (ETDEWEB)
Casoli, P.; Authier, N.; Chapelle, A. [Commissariat a l' Energie Atomique et Aux Energies Alternatives, CEA, DAM, F-21120 Is sur Tille (France)
2012-07-01
Several experimental devices are operated by the Criticality and Neutron Science Research Dept. of the CEA Valduc Laboratory. One of these is the Caliban metallic core reactor. The purpose of this study is to develop and perform experiments allowing to determinate some of fundamental kinetic parameters of the reactor. The prompt neutron decay constant and particularly its value at criticality can be measured with reactor noise techniques such as Rossi-{alpha} and Feynman variance-to-mean methods. Subcritical, critical, and even supercritical experiments were performed. Fission chambers detectors were put nearby the core and measurements were analyzed with the Rossi-{alpha} technique. A new value of the prompt neutron decay constant at criticality was determined, which allows, using the Nelson number method, new evaluations of the effective delayed neutron fraction and the in core neutron lifetime. As an introduction of this paper, some motivations of this work are given in part 1. In part 2, principles of the noise measurements experiments performed at the CEA Valduc Laboratory are reminded. The Caliban reactor is described in part 3. Stochastic neutron measurements analysis techniques used in this study are then presented in part 4. Results of fission chamber experiments are summarized in part 5. Part 6 is devoted to the current work, improvement of the experimental device using He 3 neutron detectors and first results obtained with it. Finally, conclusions and perspectives are given in part 7. (authors)
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
Reaction kinetics in open reactors and serial transfers between closed reactors
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.
Novel swirl-flow reactor for kinetic studies of semiconductor photocatalysis
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
New method for the determination of precipitation kinetics using a laminar jet reactor
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
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
Nuclear Reactor Engineering Analysis Laboratory
International Nuclear Information System (INIS)
Carlos Chavez-Mercado; Jaime B. Morales-Sandoval; Benjamin E. Zayas-Perez
1998-01-01
The Nuclear Reactor Engineering Analysis Laboratory (NREAL) is a sophisticated computer system with state-of-the-art analytical tools and technology for analysis of light water reactors. Multiple application software tools can be activated to carry out different analyses and studies such as nuclear fuel reload evaluation, safety operation margin measurement, transient and severe accident analysis, nuclear reactor instability, operator training, normal and emergency procedures optimization, and human factors engineering studies. An advanced graphic interface, driven through touch-sensitive screens, provides the means to interact with specialized software and nuclear codes. The interface allows the visualization and control of all observable variables in a nuclear power plant (NPP), as well as a selected set of nonobservable or not directly controllable variables from conventional control panels
Thermogravimetric analysis of reactor-neutrons-irradiated LEXAN polycarbonate film
International Nuclear Information System (INIS)
Kalsi, P.C.
2000-01-01
The effects of reactor-neutrons irradiation on the thermogravimetric (TG) analysis of LEXAN polycarbonate film in air were studied. Irradiation enhances the degradation rate and the effect increases further with increasing neutron fluence. The kinetics of the different steps of degradation were also evaluated from the TG curves. The activation energy values calculated for all the degradation stages decrease on irradiation. (author)
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.
Stability analysis of the Ghana Research Reactor-1 (GHARR-1)
International Nuclear Information System (INIS)
Della, R.; Alhassan, E.; Adoo, N.A.; Bansah, C.Y.; Nyarko, B.J.B.; Akaho, E.H.K.
2013-01-01
Highlights: • We developed a theoretical model to study the stability of the Ghana Research Reactor-1. • The neutronics transfer function was described by the point kinetics model for a single group of delayed neutrons. • The thermal hydraulics transfer function was based on the modified lumped parameter concept. • A computer code, RESA (REactor Stability Analysis) was developed. • Results show that the closed-loop transfer function was stable and well damped for variable operating power levels. - Abstract: A theoretical model has been developed to study the stability of the Ghana Research Reactor one (GHARR-1). The closed-loop transfer function of GHARR-1 was established based on the model, which involved the neutronics and the thermal hydraulics transfer functions. The reactor kinetics was described by the point kinetics model for a single group of delayed neutrons, whilst the thermal hydraulics transfer function was based on the modified lumped parameter concept. The inherent internal feedback effect due to the fuel and the coolant was represented by the fuel temperature coefficient and the moderator temperature coefficient respectively. A computer code, RESA (REactor Stability Analysis), entirely in Java was developed based on the model for systems analysis. Stability analysis of the open-loop transfer function of GHARR-1 based on the Nyquist criterion and Bode diagrams using RESA, has shown that the closed-loop transfer function was marginally stable for variable operating power levels. The relative stability margins of GHARR-1 were also identified
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
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.
Evaluation of energy collapsing effect on reactor kinetics parameters by diffusion theory
International Nuclear Information System (INIS)
Unesaki, Hironobu
1989-01-01
Reactor kinetics parameters play an important role as scaling factors between observed and calculated reactivities in the analysis of reactor physics experiments. In this report, energy collapsing errors in two kinetic parameters, the effective delayed neutron fraction and the neutron life time, are investigated by means of the diffusion theory. Coarse group calculations are made for various energy group structures. Cores of various moderator-to-fuel volume ratios are selected to investigate the influence of neutron spectrum changes on the energy collapsing error. The energy collapsing errors in the effective delayed neutron fraction and neutron life time are much larger than those in k eff . This might be because the former two parameters are functions of both the foward and adjoint flux, whereas the latter is a function of the forward flux alone. The use of coarse constants will cause errors in both fluxes, and the resulting errors in the former will be much more emphasized. As the effective delayed neutron fraction is sensitive to the treatment of an energy region in the vicinity of the fission spectrum peak, the coarse group error in it might differ between cores with different enrichment and composition. Inaccurate weighting of group constants leads to neutron spectra which do not conserve the fine group spectra, and those errors will be emphasized in calculated integral parameters. (N.K.)
Robustness Analysis of Kinetic Structures
DEFF Research Database (Denmark)
Kirkegaard, Poul Henning; Sørensen, John Dalsgaard
2009-01-01
Kinetic structures in architecture follows a new trend which is emerging in responsive architecture coined by Nicholas Negroponte when he proposed that architecture may benefit from the integration of computing power into built spaces and structures, and that better performing, more rational...
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)
Reactor for in situ measurements of spatially resolved kinetic data in heterogeneous catalysis
Horn, R.; Korup, O.; Geske, M.; Zavyalova, U.; Oprea, I.; Schlögl, R.
2010-06-01
The present work describes a reactor that allows in situ measurements of spatially resolved kinetic data in heterogeneous catalysis. The reactor design allows measurements up to temperatures of 1300 °C and 45 bar pressure, i.e., conditions of industrial relevance. The reactor involves reactants flowing through a solid catalyst bed containing a sampling capillary with a side sampling orifice through which a small fraction of the reacting fluid (gas or liquid) is transferred into an analytical device (e.g., mass spectrometer, gas chromatograph, high pressure liquid chromatograph) for quantitative analysis. The sampling capillary can be moved with μm resolution in or against flow direction to measure species profiles through the catalyst bed. Rotation of the sampling capillary allows averaging over several scan lines. The position of the sampling orifice is such that the capillary channel through the catalyst bed remains always occupied by the capillary preventing flow disturbance and fluid bypassing. The second function of the sampling capillary is to provide a well which can accommodate temperature probes such as a thermocouple or a pyrometer fiber. If a thermocouple is inserted in the sampling capillary and aligned with the sampling orifice fluid temperature profiles can be measured. A pyrometer fiber can be used to measure the temperature profile of the solid catalyst bed. Spatial profile measurements are demonstrated for methane oxidation on Pt and methane oxidative coupling on Li/MgO, both catalysts supported on reticulated α -Al2O3 foam supports.
To the analysis of reactor noise in boiling water reactors
International Nuclear Information System (INIS)
Seifritz, W.
1972-01-01
The paper contains some basic thoughts on the problem of neutron flux oscillations in power reactors. The advantages of self-powered detectors and their function are explained. In addition, noise measurements of the boiling water reactors at Lingen and Holden are described, and the possibilities of an employment of vanadium detectors for the analysis of reactor noise are discussed. The final pages of the paper contain a complete list of the author's publications in the field of reactor noise analysis. (RW/AK) [de
Tang, Wen-Tao; Dai, Ji; Liu, Rulong; Chen, Guang-Hao
2015-12-15
Our previous study has confirmed the feasibility of using seawater as an economical precipitant for urine phosphorus (P) precipitation. However, we still understand very little about the ureolysis in the Seawater-based Urine Phosphorus Recovery (SUPR) system despite its being a crucial step for urine P recovery. In this study, batch experiments were conducted to investigate the kinetics of microbial ureolysis in the seawater-urine system. Indigenous bacteria from urine and seawater exhibited relatively low ureolytic activity, but they adapted quickly to the urine-seawater mixture during batch cultivation. During cultivation, both the abundance and specific ureolysis rate of the indigenous bacteria were greatly enhanced as confirmed by a biomass-dependent Michaelis-Menten model. The period for fully ureolysis was decreased from 180 h to 2.5 h after four cycles of cultivation. Based on the successful cultivation, a lab-scale SUPR reactor was set up to verify the fast ureolysis and efficient P recovery in the SUPR system. Nearly complete urine P removal was achieved in the reactor in 6 h without adding any chemicals. Terminal Restriction Fragment Length Polymorphism (TRFLP) analysis revealed that the predominant groups of bacteria in the SUPR reactor likely originated from seawater rather than urine. Moreover, batch tests confirmed the high ureolysis rates and high phosphorus removal efficiency induced by cultivated bacteria in the SUPR reactor under seawater-to-urine mixing ratios ranging from 1:1 to 9:1. This study has proved that the enrichment of indigenous bacteria in the SUPR system can lead to sufficient ureolytic activity for phosphate precipitation, thus providing an efficient and economical method for urine P recovery. Copyright © 2015 Elsevier Ltd. All rights reserved.
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.)
MBS Analysis Of Kinetic Structures Using ADAMS
DEFF Research Database (Denmark)
Kirkegaard, Poul Henning; Nielsen, Søren R.K.
2009-01-01
The present paper considers multibody system (MBS) analysis of kinetic structures using the software package ADAMS. Deployable, foldable, expandable and reconfigurable kinetic structures can provide a change in the geometric morphology of the envelope by contributing to making it adaptable to e.......g. changing external climate factors, in order to improve the indoor climate performance of the building. The derivation of equations of motion for such spatial mechanical systems is a challenging issue in scientific community. However, with new symbolic tools one can automatically derive equations in so......-called multibody system (MBS) formalism. The present paper considers MBS modeling of kinetic architectural structures using the software packages ADAMS. As a result, it is found that symbolic MBS simulation tools facilitate a useful evaluation environment for MBS users during a design phase of responsive kinetic...
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%.
Enitan, Abimbola M; Kumari, Sheena; Swalaha, Feroz M; Adeyemo, J; Ramdhani, Nishani; Bux, Faizal
2014-02-01
The performance of a full-scale upflow anaerobic sludge blanket (UASB) reactor treating brewery wastewater was investigated by microbial analysis and kinetic modelling. The microbial community present in the granular sludge was detected using fluorescent in situ hybridization (FISH) and further confirmed using polymerase chain reaction. A group of 16S rRNA based fluorescent probes and primers targeting Archaea and Eubacteria were selected for microbial analysis. FISH results indicated the presence and dominance of a significant amount of Eubacteria and diverse group of methanogenic Archaea belonging to the order Methanococcales, Methanobacteriales, and Methanomicrobiales within in the UASB reactor. The influent brewery wastewater had a relatively high amount of volatile fatty acids chemical oxygen demand (COD), 2005 mg/l and the final COD concentration of the reactor was 457 mg/l. The biogas analysis showed 60-69% of methane, confirming the presence and activities of methanogens within the reactor. Biokinetics of the degradable organic substrate present in the brewery wastewater was further explored using Stover and Kincannon kinetic model, with the aim of predicting the final effluent quality. The maximum utilization rate constant U max and the saturation constant (K(B)) in the model were estimated as 18.51 and 13.64 g/l/day, respectively. The model showed an excellent fit between the predicted and the observed effluent COD concentrations. Applicability of this model to predict the effluent quality of the UASB reactor treating brewery wastewater was evident from the regression analysis (R(2) = 0.957) which could be used for optimizing the reactor performance.
Reactor noise analysis of experimental fast reactor 'JOYO'
International Nuclear Information System (INIS)
Ohtani, Hideji; Yamamoto, Hisashi
1980-01-01
As a part of dynamics tests in experimental fast reactor ''JOYO'', reactor noise tests were carried out. The reactor noise analysis techniques are effective for study of plant characteristics by determining fluctuations of process signals (neutron signal, reactor inlet temperature signals, etc.), which are able to be measured without disturbances for reactor operations. The aims of reactor noise tests were to confirm that no unstable phenomenon exists in ''JOYO'' and to gain initial data of the plant for reference of the future data. Data for the reactor noise tests treated in this paper were obtained at 50 MW power level. Fluctuations of process signals were amplified and recorded on analogue tapes. The analysis was performed using noise code (NOISA) of digital computer, with which statistical values of ASPD (auto power spectral density), CPSD (cross power spectral density), and CF (coherence function) were calculated. The primary points of the results are as follows. 1. RMS value of neutron signal at 50 MW power level is about 0.03 MW. This neutron fluctuation is not disturbing reactor operations. 2. The fluctuations of A loop reactor inlet temperatures (T sub(AI)) are larger than the fluctuations of B loop reactor inlet temperature (T sub(BI)). For this reason, the major driving force of neutron fluctuations seems to be the fluctuations of T sub(AI). 3. Core and blanket subassemblies can be divided into two halves (A and B region), with respect to the spacial motion of temperature in the reactor core. A or B region means the region in which sodium temperature fluctuations in subassembly are significantly affected by T sub(AI) or T sub(BI), respectively. This phenomenon seems to be due to the lack of mixing of A and B loop sodium in lower plenum of reactor vessel. (author)
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)
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.
Accident analysis in research reactors
International Nuclear Information System (INIS)
Adorni, M.; Bousbia-salah, A.; D'Auria, F.; Hamidouche, T.
2007-01-01
With the sustained development in computer technology, the possibilities of code capabilities have been enlarged substantially. Consequently, advanced safety evaluations and design optimizations that were not possible few years ago can now be performed. The challenge today is to revisit the safety features of the existing nuclear plants and particularly research reactors in order to verify that the safety requirements are still met and - when necessary - to introduce some amendments not only to meet the new requirements but also to introduce new equipment from recent development of new technologies. The purpose of the present paper is to provide an overview of the accident analysis technology applied to the research reactor, with emphasis given to the capabilities of computational tools. (author)
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
Catalyst dynamics: consequences for classical kinetic descriptions of reactors
DEFF Research Database (Denmark)
Johannessen, Tue; Larsen, Jane Hvolbæk; Chorkendorff, Ib
2001-01-01
in situ studies and surface science investigations has brought added attention to the fact that catalysts may behave in a dynamic manner and reconstruct depending on the reaction conditions. This feature severely limits traditional kinetic descriptions. In the present paper, we present examples...
Thermo-kinetic instabilities in model reactors. Examples in experimental tests
Lavadera, Marco Lubrano; Sorrentino, Giancarlo; Sabia, Pino; de Joannon, Mara; Cavaliere, Antonio; Ragucci, Raffaele
2017-11-01
The use of advanced combustion technologies (such as MILD, LTC, etc.) is among the most promising methods to reduce emission of pollutants. For such technologies, working temperatures are enough low to boost the formation of several classes of pollutants, such as NOx and soot. To access this temperature range, a significant dilution as well as preheating of reactants is required. Such conditions are usually achieved by a strong recirculation of exhaust gases that simultaneously dilute and pre-heat the fresh reactants. These peculiar operative conditions also imply strong fuel flexibility, thus allowing the use of low calorific value (LCV) energy carriers with high efficiency. However, the intersection of low combustion temperatures and highly diluted mixtures with intense pre-heating alters the evolution of the combustion process with respect to traditional flames, leading to features such as the susceptibility to oscillations, which are undesirable during combustion. Therefore, an effective use of advanced combustion technologies requires a thorough analysis of the combustion kinetic characteristics in order to identify optimal operating conditions and control strategies with high efficiency and low pollutant emissions. The present work experimentally and numerically characterized the ignition and oxidation processes of methane and propane, highly diluted in nitrogen, at atmospheric pressure, in a Plug Flow Reactor and a Perfectly Stirred Reactor under a wide range of operating conditions involving temperatures, mixture compositions and dilution levels. The attention was focused particularly on the chemistry of oscillatory phenomena and multistage ignitions. The global behavior of these systems can be qualitatively and partially quantitatively modeled using the detailed kinetic models available in the literature. Results suggested that, for diluted conditions and lower adiabatic flame temperatures, the competition among several pathways, i.e. intermediate- and
Spatial kinetics in nuclear reactor systems. Chapter 4
International Nuclear Information System (INIS)
Owens, D.H.
1980-01-01
The problem of constructing a low-order linear lumped-parameter model of xenon-induced spatial power oscillations in a large, cylindrical nuclear power reactor to replace an (assumed known) nonlinear distributed parameter model is examined. Model expansion and finite difference methods are used together to provide a successful solution to the problem. (U.K.)
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.
Automated Determination of Oxygen-Dependent Enzyme Kinetics in a Tube-in-Tube Flow Reactor.
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.
"Batch" kinetics in flow: online IR analysis and continuous control.
Moore, Jason S; Jensen, Klavs F
2014-01-07
Currently, kinetic data is either collected under steady-state conditions in flow or by generating time-series data in batch. Batch experiments are generally considered to be more suitable for the generation of kinetic data because of the ability to collect data from many time points in a single experiment. Now, a method that rapidly generates time-series reaction data from flow reactors by continuously manipulating the flow rate and reaction temperature has been developed. This approach makes use of inline IR analysis and an automated microreactor system, which allowed for rapid and tight control of the operating conditions. The conversion/residence time profiles at several temperatures were used to fit parameters to a kinetic model. This method requires significantly less time and a smaller amount of starting material compared to one-at-a-time flow experiments, and thus allows for the rapid generation of kinetic data. Copyright © 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.
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.
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
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.)
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)
Analysis of reactor noise; Analiza reaktorskih sumova
Energy Technology Data Exchange (ETDEWEB)
Velickovic, Lj [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)
1967-11-15
This paper describes the theoretical model for interpretation of experimental results, experimental method for study of reactor noise at the RB reactor, numerical treatment of experimental results by correlation technique for analysis of reactor noise. A computer code was written to obtain autocorrelation function and spectral density function. Experimental results obtained by oscillator technique, pulse technique, and autocorrelation method are presented and discussed.
Count-to-count time interval distribution analysis in a fast reactor
International Nuclear Information System (INIS)
Perez-Navarro Gomez, A.
1973-01-01
The most important kinetic parameters have been measured at the zero power fast reactor CORAL-I by means of the reactor noise analysis in the time domain, using measurements of the count-to-count time intervals. (Author) 69 refs
Kinetic parameters of hydroprocessing reactions in a flow reactor
Energy Technology Data Exchange (ETDEWEB)
Raychaudhuri, U.; Banerjee, T.S.; Ghar, R.N. (Indian Institute of Technology, Kharagpur (India))
1994-01-01
The change in distillation properties of a blend of light and heavy distillates over a commercial hydrotreating catalyst was studied using a small packed bed reactor. The results were interpreted assuming a pseudo-component model that took into account the physical and chemical complexity of the system. A first order series-parallel reaction mechanism was found to be valid for the operating conditions involved. Pore diffusion effects were also taken into consideration. 8 refs., 7 figs., 1 tab.
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
International Nuclear Information System (INIS)
Muhammad, Farhan; Majid, Asad
2009-01-01
The effects of using different low enriched uranium fuels, having same uranium density, on the kinetic parameters of a material test research reactor were studied. For this purpose, the original aluminide fuel (UAl x -Al) containing 4.40 gU/cm 3 of an MTR was replaced with silicide (U 3 Si-Al and U 3 Si 2 -Al) and oxide (U 3 O 8 -Al) dispersion fuels having the same uranium density as of the original fuel. Simulations were carried out to calculate prompt neutron generation time, effective delayed-neutron fraction, core excess reactivity and neutron flux spectrum. Nuclear reactor analysis codes including WIMS-D4 and CITATION were used to carry out these calculations. It was observed that both the silicide fuels had the same prompt neutron generation time 0.02% more than that of the original aluminide fuel, while the oxide fuel had a prompt neutron generation time 0.05% less than that of the original aluminide fuel. The effective delayed-neutron fraction decreased for all the fuels; the decrease was maximum at 0.06% for U 3 Si 2 -Al followed by 0.03% for U 3 Si-Al, and 0.01% for U 3 O 8 -Al fuel. The U 3 O 8 -Al fueled reactor gave the maximum ρ excess at BOL which was 21.67% more than the original fuel followed by U 3 Si-Al which was 2.55% more, while that of U 3 Si 2 -Al was 2.50% more than the original UAl x -Al fuel. The neutron flux of all the fuels was more thermalized, than in the original fuel, in the active fuel region of the core. The thermalization was maximum for U 3 O 8 -Al followed by U 3 Si-Al and then U 3 Si 2 -Al fuel.
Subchannel analysis in nuclear reactors
International Nuclear Information System (INIS)
Ninokata, H.; Aritomi, M.
1992-01-01
This book contains 10 informative papers, presented at the International Seminar on Subchannel Analysis 1992 (ISSCA '92), organized by the Institute of Applied Energy, in collaboration with Atomic Energy Society of Japan, Tokyo Electric Power Company, Kansai Electric Power Company, Nuclear Power Engineering Corporation and the Japan Atomic Energy Research Institute, and held at the TIS-Green Forum, Tokyo, Japan, 30 October 1992. The seminar ISSCA '92 was intended to review the current state-of-the-arts of the method being applied to advanced nuclear reactors including Advanced BWRs, Advanced PWRs and LMRs, and to identify the problems to be solved, improvements to be made, and the needs of R and Ds that were required from the new fuel bundles design. The critical review was to focus on the performances of currently available subchannel analysis codes with regard to heat transfer and fluid flows in various types of nuclear reactor bundles under both steady-state and transient operating conditions, CHF, boiling transition (BT) or dryout behaviors and post BT. The behaviors of physical modeling and numerical methods in these extreme conditions were discussed and the methods critically evaluated in comparison with experiments. (author) (J.P.N.)
Basic dye decomposition kinetics in a photocatalytic slurry reactor
International Nuclear Information System (INIS)
Wu, C.-H.; Chang, H.-W.; Chern, J.-M.
2006-01-01
Wastewater effluent from textile plants using various dyes is one of the major water pollutants to the environment. Traditional chemical, physical and biological processes for treating textile dye wastewaters have disadvantages such as high cost, energy waste and generating secondary pollution during the treatment process. The photocatalytic process using TiO 2 semiconductor particles under UV light illumination has been shown to be potentially advantageous and applicable in the treatment of wastewater pollutants. In this study, the dye decomposition kinetics by nano-size TiO 2 suspension at natural solution pH was experimentally studied by varying the agitation speed (50-200 rpm), TiO 2 suspension concentration (0.25-1.71 g/L), initial dye concentration (10-50 ppm), temperature (10-50 deg. C), and UV power intensity (0-96 W). The experimental results show the agitation speed, varying from 50 to 200 rpm, has a slight influence on the dye decomposition rate and the pH history; the dye decomposition rate increases with the TiO 2 suspension concentration up to 0.98 g/L, then decrease with increasing TiO 2 suspension concentration; the initial dye decomposition rate increases with the initial dye concentration up to a certain value depending upon the temperature, then decreases with increasing initial dye concentration; the dye decomposition rate increases with the UV power intensity up to 64 W to reach a plateau. Kinetic models have been developed to fit the experimental kinetic data well
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
Data acquisition and processing system for reactor noise analysis
International Nuclear Information System (INIS)
Costa Oliveira, J.; Morais Da Veiga, C.; Forjaz Trigueiros, D.; Pombo Duarte, J.
1975-01-01
A data acquisition and processing system for reactor noise analysis by time correlation methods is described, consisting in one to four data feeding channels (transducer, associated electronics and V/f converter), a sampling unit, a landline transmission system and a PDP 15 computer. This system is being applied to study the kinetic parameters of the 'Reactor Portugues de Investigacao', a swimming-pool 1MW reactor. The main features that make such a data acquisition and processing system a useful tool to perform noise analysis are: the improved characteristics of analog-to-digital converters employed to quantize the signals; the use of an on-line computer which allows a great accumulation and a rapid treatment of data together with an easy check of the correctness of the experiments; and the adoption of the time cross-correlation technique using two-detectors which by-pass the limitation of low efficiency detectors. (author)
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.
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.
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.
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
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
Energy Technology Data Exchange (ETDEWEB)
Hartmann, Christoph Oliver
2016-06-13
Coupled Thermal-hydraulic/Neutron-kinetic (TH/NK) simulations of Boiling Water Reactor transients require well validated and accurate simulation tools. The generation of cross-section (XS) libraries, depending on the individual thermal-hydraulic state parameters, is of paramount importance for coupled simulations. Problem-dependent XS-sets for 3D core simulations are being generated mainly by well validated, fast running commercial and user-friendly lattice codes such as CASMO and HELIOS. In this dissertation a computational route, based on the lattice code SCALE6/TRITON, the cross-section interface GenPMAXS, the best-estimate thermal-hydraulic system code TRACE and the core simulator PARCS, for best-estimate simulations of Boiling Water (BWR) transients has been developed and validated. The computational route has been supplemented by a subsequent uncertainty and sensitivity study based on Monte Carlo sampling and propagation of the uncertainties of input parameters to the output (SUSA code). The analysis of a single BWR fuel assembly depletion problem with PARCS using SCALE/TRITON cross-sections has been shown a good agreement with the results obtained with CASMO cross-section sets. However, to compensate the deficiencies of the interface program GenPMAXS, PYTHON scripts had to be developed to incorporate missing data, as the yields of Iodine, Xenon and Promethium, into the cross-section-data sets (PMAXS-format) generated by GenPMAXS from the SCALE/TRITON output. The results of the depletion analysis of a full BWR core with PARCS have indicated the importance of considering history effects, adequate modeling of the reflector region and the control rods, as the PARCS simulations for depleted fuel and all control rods inserted (ARI) differs significantly at the fuel assembly top and bottom. Systematic investigations with the coupled codes TRACE/PARCS have been performed to analyse the core behaviour at different thermal conditions using nuclear data (XS
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
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.
Safety analysis of reactor's cooling system
International Nuclear Information System (INIS)
1999-01-01
Results of the analysis of reactor's RBMK-1500 coolant system during normal operation mode, hydrodynamic testing and in the case of earthquake are presented. Analysis was performed using RELAP5 code. Calculations showed the most vulnerable place in the reactor's coolant system. It was found that in the case of earthquake the horizontal support system of drum separator could be damaged
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%
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.
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.
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.
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.
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)
HYDROGEN KINETICS LIMITATION OF AN AUTOTROPHIC SULPHATE REDUCTION REACTOR
Directory of Open Access Journals (Sweden)
CÉSAR SÁEZ-NAVARRETE
2012-01-01
Full Text Available El uso de sustratos inorgánicos podría reducir los costos y simplificar la operación de sistemas de tratamiento de aguas que utilizan bacterias reductoras de sulfato. Sin embargo, el uso de H2 como sustrato energético y la bioproducción de H2S podrían provocar limitaciones cinéticas. El objetivo de este estudio fue evaluar las condiciones en las que la capacidad de transferencia de masa de un bioreactor de reducción de sulfato, limita su cinética de reducción. La cinética del reactor fue obtenida monitoreando la presión del sistema en condiciones de no limitación por sulfato. Se concluyó que el diseño del bioreactor debería basarse en sus propiedades de transferencia. La tasa de consumo de H2 alcanzó un máximo de 10-4 M/min, para una tasa de reducción de sulfato de 3.4 g·L-1·d-1. Para evitar limitación por H2 se requirió un kLa de 1.48 min-1 a 1.2·109 cells/L (1.23·10-9 L·min-1·cell-1, valor relevante para propósitos de escalamiento.
Chemical kinetic functional sensitivity analysis: Elementary sensitivities
International Nuclear Information System (INIS)
Demiralp, M.; Rabitz, H.
1981-01-01
Sensitivity analysis is considered for kinetics problems defined in the space--time domain. This extends an earlier temporal Green's function method to handle calculations of elementary functional sensitivities deltau/sub i//deltaα/sub j/ where u/sub i/ is the ith species concentration and α/sub j/ is the jth system parameter. The system parameters include rate constants, diffusion coefficients, initial conditions, boundary conditions, or any other well-defined variables in the kinetic equations. These parameters are generally considered to be functions of position and/or time. Derivation of the governing equations for the sensitivities and the Green's funciton are presented. The physical interpretation of the Green's function and sensitivities is given along with a discussion of the relation of this work to earlier research
Flow analysis in a supercritical water oxidation reactor
International Nuclear Information System (INIS)
Oh, C.H.; Kochan, R.J.; Beller, J.M.
1996-01-01
Supercritical water oxidation (SCWO), also known as hydrothermal oxidation (HTO), involves the oxidation of hazardous waste at conditions of elevated temperature and pressure (e.g., 500 C--600 C and 234.4 bar) in the presence of approximately 90% of water and a 10% to 20% excess amount of oxidant over the stoichiometric requirement. Under these conditions, organic compounds are completely miscible with supercritical water, oxygen and nitrogen, and are rapidly oxidized to carbon dioxide and water. The essential part of the process is the reactor. Many reactor designs such as tubular, vertical vessel, and transpiring wall type have been proposed, patented, and tested at both bench and pilot scales. These designs and performances need to be scaled up to a waste throughput 10--100 times that currently being tested. Scaling of this magnitude will be done by creating a numerical thermal-hydraulic model of the smaller reactor for which test data is available, validating the model against the available data, and then using the validated model to investigate the larger reactor performance. This paper presents a flow analysis of the MODAR bench scale reactor (vertical vessel type). These results will help in the design of the reactor in an efficient manner because the flow mixing coupled with chemical kinetics eventually affects the process destruction efficiency
Kinetic study on the effect of temperature on biogas production using a lab scale batch reactor.
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.
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.
Research reactor job analysis - A project description
International Nuclear Information System (INIS)
Yoder, John; Bessler, Nancy J.
1988-01-01
Addressing the need of the improved training in nuclear industry, nuclear utilities established training program guidelines based on Performance-Based Training (PBT) concepts. The comparison of commercial nuclear power facilities with research and test reactors owned by the U.S. Department of Energy (DOE), made in an independent review of personnel selection, training, and qualification requirements for DOE-owned reactors pointed out that the complexity of the most critical tasks in research reactors is less than that in power reactors. The U.S. Department of Energy (DOE) started a project by commissioning Oak Ridge Associated Universities (ORAU) to conduct a job analysis survey of representative research reactor facilities. The output of the project consists of two publications: Volume 1 - Research Reactor Job Analysis: Overview, which contains an Introduction, Project Description, Project Methodology,, and. An Overview of Performance-Based Training (PBT); and Volume 2 - Research Reactor Job Analysis: Implementation, which contains Guidelines for Application of Preliminary Task Lists and Preliminary Task Lists for Reactor Operators and Supervisory Reactor Operators
A KINETIC MODEL FOR H2O2/UV PROCESS IN A COMPLETELY MIXED BATCH REACTOR. (R825370C076)
A dynamic kinetic model for the advanced oxidation process (AOP) using hydrogen peroxide and ultraviolet irradiation (H2O2/UV) in a completely mixed batch reactor (CMBR) is developed. The model includes the known elementary chemical and photochemical reac...
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.
Kinetic modeling of the photocatalytic degradation of clofibric acid in a slurry reactor.
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.
Experimental estimations of the kinetics parameters of the IBR-2M reactor by stochastic noises
International Nuclear Information System (INIS)
Pepelyshev, Yu.N.; Tajybov, L.A.; Garibov, A.A.; Mekhtieva, R.N.
2012-01-01
Experimental investigations of stochastic fluctuations of pulse energy of the IBR-2M reactor have been carried out which allowed us to obtain some of the parameters of the reactor kinetics. At different levels of average power a sequence of values of pulse energy was recorded with the calculation of the distribution parameters. An ionization chamber with boron installed near the active zone was used as a neutron detector. The research results allowed us to estimate the average lifetime of prompt neutrons τ = (6.53±0.2)·10 -8 s, absolute power of the reactor and intensity of the source of spontaneous neutrons S sp ≤(6.72±0.12)·10 6 s -1 . It was shown that the experimental results are close to the calculated ones
Energy Technology Data Exchange (ETDEWEB)
Royaee, Sayed Javid; Shafeghat, Amin [Research Institute of Petroleum Industry, Tehran (Iran, Islamic Republic of); Sohrabi, Morteza [Amirkabir University of Technology, Tehran (Iran, Islamic Republic of)
2014-02-15
A photo impinging streams cyclone reactor has been used as a novel apparatus in photocatalytic degradation of organic compounds using titanium dioxide nanoparticles in wastewater. The operating parameters, including catalyst loading, pH, initial phenol concentration and light intensity have been optimized to increase the efficiency of the photocatalytic degradation process within this photoreactor. The results have demonstrated a higher efficiency and an increased performance capability of the present reactor in comparison with the conventional processes. In the next step, residence time distribution (RTD) of the slurry phase within the reactor was measured using the impulse tracer method. A CFD-based model for predicting the RTD was also developed which compared well with the experimental results. The RTD data was finally applied in conjunction with the phenol degradation kinetic model to predict the apparent rate coefficient for such a reaction.
Accident analysis for PRC-II reactor
International Nuclear Information System (INIS)
Wei Yongren; Tang Gang; Wu Qing; Lu Yili; Liu Zhifeng
1997-12-01
The computer codes, calculation models, transient results, sensitivity research, design improvement, and safety evaluation used in accident analysis for PRC-II Reactor (The Second Pulsed Reactor in China) are introduced. PRC-II Reactor is built in big populous city, so the public pay close attention to reactor safety. Consequently, Some hypothetical accidents are analyzed. They include an uncontrolled control rod withdrawal at rated power, a pulse rod ejection at rated power, and loss of coolant accident. Calculation model which completely depict the principle and process for each accident is established and the relevant analysis code is developed. This work also includes comprehensive computing and analyzing transients for each accident of PRC-II Reactor; the influences in the reactor safety of all kind of sensitive parameters; evaluating the function of engineered safety feature. The measures to alleviate the consequence of accident are suggested and taken in the construction design of PRC-II Reactor. The properties of reactor safety are comprehensively evaluated. A new advanced calculation model (True Core Uncovered Model) of LOCA of PRC-II Reactor and the relevant code (MCRLOCA) are first put forward
Reactor accident analysis and evaluation
International Nuclear Information System (INIS)
Chang, J.W.
1983-01-01
Reactor Management Division of Korea Advanced Energy Research Institute has, so far, adopted, modified and developed quite a number of large programs for nuclear core analysis. During the course of this work, it was found necessary to employ some standard subroutines for handling data, input procedures, core memory management and search files. Many programs share lots of common subroutines and/or functions with other programs. Above all, some of them are in lack of transmittal. During the installation of big codes for CYBER computer, it has drawn our keen attention that many elementary subroutines are heavily machine-dependent and that their conversion is extremely difficult. After having collected and modified the subroutines to fit in different codes, it was finally named KINEP (KAERI Improved Nuclear Environmental Package). KINEP has been proved to be convenient even for smaller programs for general purpose. The KINEP includes about one hundred subroutines to facilitate data handling, operator communications, storage allocation, decimal input, file maintence and scratch I/O. (Author)
Economic analysis of EBT reactor
International Nuclear Information System (INIS)
Woo, J.T.; Uckan, N.A.; Lidsky, L.M.
1977-01-01
In order to establish the economic potential of the Elmo Bumpy Torus (EBT) reactor, two independent system-costing models have been developed. Both models predict capital costs of approximately $400/kW(th). These relatively low costs reflect the simplicity of the EBTR design. In particular, the modular nature of the individual blanket-shield segments, the low costs ''accelerator style'' containment building, high beta, and steady-state operation lead to relatively low reactor costs. A detailed cost breakdown for subsystems is analyzed. High cost and high uncertainty subsystems are identified to direct further emphasis into those areas. The calculated capital costs for the EBT reactor are compared with those costs quoted for tokamak reactors
Kinetic Analysis of Horizontal Plyometric Exercise Intensity.
Kossow, Andrew J; Ebben, William P
2018-05-01
Kossow, AJ, DeChiara, TG, Neahous, SM, and Ebben, WP. Kinetic analysis of horizontal plyometric exercise intensity. J Strength Cond Res 32(5): 1222-1229, 2018-Plyometric exercises are frequently performed as part of a strength and conditioning program. Most studies assessed the kinetics of plyometric exercises primarily performed in the vertical plane. The purpose of this study was to evaluate the multiplanar kinetic characteristics of a variety of plyometric exercises, which have a significant horizontal component. This study also sought to assess sex differences in the intensity progression of these exercises. Ten men and 10 women served as subjects. The subjects performed a variety of plyometric exercises including the double-leg hop, standing long jump, single-leg standing long jump, bounding, skipping, power skipping, cone hops, and 45.72-cm hurdle hops. Subjects also performed the countermovement jump for comparison. All plyometric exercises were evaluated using a force platform. Dependent variables included the landing rate of force development and landing ground reaction forces for each exercise in the vertical, frontal, and sagittal planes. A 2-way mixed analysis of variance with repeated-measures for plyometric exercise type demonstrated main effects for exercise type for all dependent variables (p ≤ 0.001). There was no significant interaction between plyometric exercise type and sex for any of the variable assessed. Bonferroni-adjusted pairwise comparisons identified a number of differences between the plyometric exercises for the dependent variables assessed (p ≤ 0.05). These findings should be used to guide practitioners in the progression of plyometric exercise intensity, and thus program design, for those who require significant horizontal power in their sport.
Immune adherence: a quantitative and kinetic analysis
Energy Technology Data Exchange (ETDEWEB)
Sekine, T [National Cancer Center, Tokyo (Japan). Research Inst.
1978-09-01
Quantitative and kinetic analysis of the immune-adherence reaction (IA) between C3b fragments and IA receptors as an agglutination reaction is difficult. Analysis is possible, however, by use of radio-iodinated bovine serum albumin as antigen at low concentrations (less than 200 ng/ml) and optimal concentration of antibody to avoid precipitation of antigen-antibody complexes with human erythrocytes without participation of complement. Antigen and antibody are reacted at 37/sup 0/C, complement is added, the mixture incubated and human erythrocytes added; after further incubation, ice-cold EDTA containing buffer is added and the erythrocytes centrifuged and assayed for radioactivity. Control cells reacted with heated guinea pig serum retained less than 5% of the added radioactivity. The method facilitates measurement of IA reactivity and permits more detailed analysis of the mechanism underlying the reaction.
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
Analysis of dynamic stability and safety of reactor system by reactor simulator
International Nuclear Information System (INIS)
Raisic, N.
1963-11-01
In order to enable qualitative analysis of dynamic properties of reactors RA and RB, mathematical models of these reactors were formulated and adapted for solution on analog computer. This report contains basic assessments for creating the model and complete equations for each reactor. Model was used to analyse three possible accidents at the RA reactor and possible hypothetical accidents at the RB reactor
Study of carbon dioxide gas treatment based on equations of kinetics in plasma discharge reactor
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.
Economic analysis of nuclear reactors
International Nuclear Information System (INIS)
Owen, P.S.; Parker, M.B.; Omberg, R.P.
1979-05-01
The report presents several methods for estimating the power costs of nuclear reactors. When based on a consistent set of economic assumptions, total power costs may be useful in comparing reactor alternatives. The principal items contributing to the total power costs of a nuclear power plant are: (1) capital costs, (2) fuel cycle costs, (3) operation and maintenance costs, and (4) income taxes and fixed charges. There is a large variation in capital costs and fuel expenses among different reactor types. For example, the standard once-through LWR has relatively low capital costs; however, the fuel costs may be very high if U 3 O 8 is expensive. In contrast, the FBR has relatively high capital costs but low fuel expenses. Thus, the distribution of expenses varies significantly between these two reactors. In order to compare power costs, expenses and revenues associated with each reactor may be spread over the lifetime of the plant. A single annual cost, often called a levelized cost, may be obtained by the methods described. Levelized power costs may then be used as a basis for economic comparisons. The paper discusses each of the power cost components. An exact expression for total levelized power costs is derived. Approximate techniques of estimating power costs will be presented
Thermohydraulic analysis of pressurized water reactors
International Nuclear Information System (INIS)
Veloso, M.A.
1980-01-01
The computer program PANTERA is applied in the thermo-hydraulic analysis of Pressurized Water Reactor Cores (PWR). It is a version of COBRA-IIIC in which a new thermal conduction model for fuel rods was introduced. The results calculated by this program are compared with experimental data obtained from bundles of fuel rods, simulating reactor conditions. The validity of the new thermal model is checked too. The PANTERA code, through a simplified procedure of calculation, is used in the thermo-hydraulic analysis of Indian Point, Unit 2, reactor core, in stationary conditions. The results are discussed and compared with design data. (Autor) [pt
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 $
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)
Russell, Charles R
1962-01-01
Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor
Directory of Open Access Journals (Sweden)
NADEZDA JOVANOVIC
2001-04-01
Full Text Available The dehidration kinetics of gibbsite to activated alumina was investigated at four different temperatures between 883 K and 943 K in a reactor for pneumatic transport in the dilute two phase flow regime. The first order kinetic behavior of this reactionwith respect to the water content of the solid material was proved and an activation energy of 66.5 kJ/mol was calculated. The effect of residence time on the water content is given and compared with theoretical calculations. The water content and other characteristics of the products depend on two main parameters, one is the short residence time and the other is the temperature of the dehydration of gibbsite. The short residence time of the gibbsite particles in a reactor for pneumatic transport prevents crystallization into new phases, as established from XRD analysis data. Reactive amorphous alumina powder, with a specific surface area of 250 m2/g, suitable as a precursor for catalyst supports is obtained.
Analysis of lipid peroxidation kinetics. I
DEFF Research Database (Denmark)
Doktorov, Alexander B.; Lukzen, Nikita N.; Pedersen, Jørgen Boiden
2008-01-01
concentrations of reactants or diﬀerent ways of initiating the re- action. Nor has it been possible to predict the time dependence of the products. The reason for these problems is the complicated structure of the kinetic scheme, which includes a chain reaction. In this work we perform an in depth analysis...... The kinetics of the lipid peroxidation reaction is only partly under- stood. Although the set of reactions constituting the overall reaction is believed to be known, it has not been possible to predict how the reaction will respond to a change of one or more of the parameters, e.g. initial...... of the importance of the individual reaction steps and we introduce a new quasi-stationary concentration method based on the assumption that one or more concentrations vary much slower than the others. We show that it is justiﬁed to use a quasi-stationary concentration approximation for the alkyl radical L...
Activation analysis with small mobile reactors
International Nuclear Information System (INIS)
Chung, C.
1990-01-01
A small nuclear reactor (a low-power reactor without heat removal devices) usually has thermal power output under 100 W and an average in-core thermal neutron flux below 10 9 n/cm 2 s. Conventional activation analysis is restricted to determination of specific elements with large neutron capture cross sections in sizable samples. In-vivo prompt gamma activation analysis (IVPGAA) can be used for diagnosis of elemental composition of the human body, particularly the essential elements Ca, Cl, N, and P in the whole body, and toxic Cd and Hg in contaminated organs. In this chapter, activation analysis using an external neutron beam from the Tsing Hua Mobile Educational Reactor (THMER) for in vivo activation is described. Characteristics of the mobile reactor, in-vivo medical diagnosis, and radiation safety are emphasized. 17 refs, 12 figs, 3 tabs
Kinetics study of the fluorination of uranium tetrafluoride in a fluidized bed reactor
International Nuclear Information System (INIS)
Khani, M.H.; Pahlavanzadeh, H.; Ghannadi, M.
2008-01-01
The kinetics of reaction of the uranium tetrafluoride conversion to the uranium hexafluoride with fluorine gas taking place in a fluidized bed reactor operating in industrial conditions have been studied. The external and internal diffusion effects are investigated by Mears and Weisz-Prater criterions. The kinetic equation for the fluorination of uranium tetrafluoride is developed in the absence of diffusional limitation using an integral method by assuming that the gas flow is of plug or perfectly mixed type. A good agreement is observed between the experimental data and a first-order model with respect to fluorine in the CSTR system. The activation energy of the reaction and the pre-exponential factor are obtained using analytical results from a better model
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.
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)
[Kinetics of catalytic wet air oxidation of phenol in trickle bed reactor].
Li, Guang-ming; Zhao, Jian-fu; Wang, Hua; Zhao, Xiu-hua; Zhou, Yang-yuan
2004-05-01
By using a trickle bed reactor which was designed by the authors, the catalytic wet air oxidation reaction of phenol on CuO/gamma-Al2O3 catalyst was studied. The results showed that in mild operation conditions (at temperature of 180 degrees C, pressure of 3 MPa, liquid feed rate of 1.668 L x h(-1) and oxygen feed rate of 160 L x h(-1)), the removal of phenol can be over 90%. The curve of phenol conversion is similar to "S" like autocatalytic reaction, and is accordance with chain reaction of free radical. The kinetic model of pseudo homogenous reactor fits the catalytic wet air oxidation reaction of phenol. The effects of initial concentration of phenol, liquid feed rate and temperature for reaction also were investigated.
Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element
Energy Technology Data Exchange (ETDEWEB)
Mohammed, Abdul Aziz, E-mail: azizM@uniten.edu.my; Rahman, Shaik Mohmmed Haikhal Abdul [Universiti Tenaga Nasional. Jalan Ikram-UNITEN, 43000 Kajang, Selangor (Malaysia); Pauzi, Anas Muhamad, E-mail: anas@uniten.edu.my; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad [Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia)
2016-01-22
In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 ({sup 233}U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.
Spatial kinetics studies in liquid-metal fast breeder reactor critical assemblies
International Nuclear Information System (INIS)
Brumback, S.B.; Goin, R.W.; Carpenter, S.G.
1988-01-01
Recent measurements in the zero-power physics reactor have been used to study the effect of spatial decoupling in fast reactor critical assemblies of various sizes and compositions. Flux distributions in these assemblies had varying degrees of sensitivity to perturbations. Decoupling was investigated using rod-drop, boron-oscillator, and noise-coherence techniques, which emphasized different times following perturbations. Equilibrium flux distributions were also measured for subcritical configurations with inserted control rods. For most assemblies, accurate reactivity measurements were obtained by analyzing the power history from a single detector using inverse kinetics methods, assuming an instantaneous efficiency change for the detector. The instantaneous efficiency change assumption broke down, however, in assemblies with zones in which normal plutonium fuel was replaced by /sup 235/U fuel or fuel with a high /sup 240/Pu content. Flux redistributions caused by perturbations in these cores took several minutes to evolve
Effect of ultrasound on flotation kinetics in the reactor-separator
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.
Generic analysis of kinetically driven inflation
Saitou, Rio
2018-04-01
We perform a model-independent analysis of kinetically driven inflation (KDI) which (partially) includes generalized G-inflation and ghost inflation. We evaluate the background evolution splitting into the inflationary attractor and the perturbation around it. We also consider the quantum fluctuation of the scalar mode with a usual scaling and derive the spectral index, ignoring the contribution from the second-order products of slow-roll parameters. Using these formalisms, we find that within our generic framework the models of KDI which possess the shift symmetry of scalar field cannot create the quantum fluctuation consistent with the observation. Breaking the shift symmetry, we obtain a few essential conditions for viable models of KDI associated with the graceful exit.
International Nuclear Information System (INIS)
Santos, Rubens Souza dos; Martinez, Aquilino Senra; Alvim, Antonio Carlos Marques
2002-01-01
In this work is presented a methodology which focuses the distribution of neutron absorber rods in nuclear reactor power plants, for utilizing in space kinetic calculations, principally in the cluster ejection transients of control rods. A numerical model for macroscopic constant calculations based on the knowledge of the neutron flux without the control rods is proposed, as alternative to the analytical models, based on the hypothesis of the null current on the cell super boundaries. The proposed model in this work has itself showed adequate to deal with problems with strong space dependence, once that the model showed consistence in the global average built in the analytical model. (author)
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.
Analysis of reactivity insertion accidents in PWR reactors
International Nuclear Information System (INIS)
Camargo, C.T.M.
1978-06-01
A calculation model to analyze reactivity insertion accidents in a PWR reactor was developed. To analyze the nuclear power transient, the AIREK-III code was used, which simulates the conventional point-kinetic equations with six groups of delayed neutron precursors. Some modifications were made to generalize and to adapt the program to solve the proposed problems. A transient thermal analysis model was developed which simulates the heat transfer process in a cross section of a UO 2 fuel rod with Zircalloy clad, a gap fullfilled with Helium gas and the correspondent coolant channel, using as input the nulcear power transient calculated by AIREK-III. The behavior of ANGRA-i reactor was analized during two types of accidents: - uncontrolled rod withdrawal from subcritical condition; - uncontrolled rod withdrawal at power. The results and conclusions obtained will be used in the license process of the Unit 1 of the Central Nuclear Almirante Alvaro Alberto. (Author) [pt
Modeling of reaction kinetics in bubbling fluidized bed biomass gasification reactor
Energy Technology Data Exchange (ETDEWEB)
Thapa, R.K.; Halvorsen, B.M. [Telemark University College, Kjolnes ring 56, P.O. Box 203, 3901 Porsgrunn (Norway); Pfeifer, C. [University of Natural Resources and Life Sciences, Vienna (Austria)
2013-07-01
Bubbling fluidized beds are widely used as biomass gasification reactors as at the biomass gasification plant in Gussing, Austria. The reactor in the plant is a dual circulating bubbling fluidized bed gasification reactor. The plant produces 2MW electricity and 4.5MW heat from the gasification of biomass. Wood chips as biomass and olivine particles as hot bed materials are fluidized with high temperature steam in the reactor. As a result, biomass undergoes endothermic chemical reaction to produce a mixture of combustible gases in addition to some carbon-dioxide (CO2). The combustible gases are mainly hydrogen (H2), carbon monoxide (CO) and methane (CH4). The gas is used to produce electricity and heat via utilization in a gas engine. Alternatively, the gas is further processed for gaseous or liquid fuels, but still on the process of development level. Composition and quality of the gas determine the efficiency of the reactor. A computational model has been developed for the study of reaction kinetics in the gasification rector. The simulation is performed using commercial software Barracuda virtual reactor, VR15. Eulerian-Lagrangian approach in coupling of gas-solid flow has been implemented. Fluid phase is treated with an Eulerian formulation. Discrete phase is treated with a Lagrangian formulation. Particle-particle and particle-wall interactions and inter-phase heat and mass transfer have been taken into account. Series of simulations have been performed to study model prediction of the gas composition. The composition is compared with data from the gasifier at the CHP plant in Güssing, Austria. The model prediction of the composition of gases has good agreements with the result of the operating plant.
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
International Nuclear Information System (INIS)
Kataoka, H.
1975-01-01
The 'not-linear' or '2nd-class-nonlinear' characteristics in nuclear reactor kinetics with the feedback effect in the high-power operation and induce the increase in the amplitude of the neutron flux noise, specially in the very low frequency region. The fundamental behaviour of 'not-linear' characteristics and its effect for the reactor noise was investigated. Application of the reactor noise analysis technique to power reactors has not been successful because of unknown large disagreement between the result of the conventional theoretical analysis and the experimental facts. When the cause of this discrepancy is clear, reactor noise analysis techniques can be effectively applied to instrumentation, control, monitoring and diagnosis of power reactors. (author)
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)
Linear and nonlinear stability analysis, associated to experimental fast reactors. Part 2
International Nuclear Information System (INIS)
Amorim, E.S. do; Moura Neto, C. de; Rosa, M.A.P.
1980-07-01
The nonlinear effects in fast reactors kinetics and their stability are studied. The Lyapunov criteria and the Lurie-Letov functions for nonlinear systems were established and simulated. Small oscillations were studied by a Fourier analysis to clarify particular aspects of feedback and load functions in fast reactor at zero power, or/and in normal power level. The results were in agreement with the experimental data existing in the literature. (E.G.) [pt
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)
Finite element application to global reactor analysis
International Nuclear Information System (INIS)
Schmidt, F.A.R.
1981-01-01
The Finite Element Method is described as a Coarse Mesh Method with general basis and trial functions. Various consequences concerning programming and application of Finite Element Methods in reactor physics are drawn. One of the conclusions is that the Finite Element Method is a valuable tool in solving global reactor analysis problems. However, problems which can be described by rectangular boxes still can be solved with special coarse mesh programs more efficiently. (orig.) [de
Accident analysis of heavy water cooled thorium breeder reactor
International Nuclear Information System (INIS)
Yulianti, Yanti; Su’ud, Zaki; Takaki, Naoyuki
2015-01-01
Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The
Accident analysis of heavy water cooled thorium breeder reactor
Energy Technology Data Exchange (ETDEWEB)
Yulianti, Yanti [Department of Physics, University of Lampung Jl. Sumantri Brojonegoro No.1 Bandar Lampung, Indonesia Email: y-yanti@unila.ac.id (Indonesia); Su’ud, Zaki [Department of Physics, Bandung Institute of Technology Jl. Ganesha 10 Bandung, Indonesia Email: szaki@fi.itb.ac.id (Indonesia); Takaki, Naoyuki [Department of Nuclear Safety Engineering Cooperative Major in Nuclear Energy (Graduate School) 1-28-1 Tamazutsumi,Setagayaku, Tokyo158-8557, Japan Email: ntakaki@tcu.ac.jp (Japan)
2015-04-16
Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The
Treating experimental data of inverse kinetic method by unitary linear regression analysis
International Nuclear Information System (INIS)
Zhao Yusen; Chen Xiaoliang
2009-01-01
The theory of treating experimental data of inverse kinetic method by unitary linear regression analysis was described. Not only the reactivity, but also the effective neutron source intensity could be calculated by this method. Computer code was compiled base on the inverse kinetic method and unitary linear regression analysis. The data of zero power facility BFS-1 in Russia were processed and the results were compared. The results show that the reactivity and the effective neutron source intensity can be obtained correctly by treating experimental data of inverse kinetic method using unitary linear regression analysis and the precision of reactivity measurement is improved. The central element efficiency can be calculated by using the reactivity. The result also shows that the effect to reactivity measurement caused by external neutron source should be considered when the reactor power is low and the intensity of external neutron source is strong. (authors)
Neutronics methods for transient and safety analysis of fast reactors
Energy Technology Data Exchange (ETDEWEB)
Marchetti, Marco
2017-07-01
Modeling the evolution of possible or postulated accidents in nuclear reactors is fundamental in designing safe systems. For the next generation of reactors, in particular fast reactors, fuel movement during an accident can, in principle, drive an energetic event. Such is the issue of recriticality. The thermal energy produced during these events will, possibly, be converted into mechanical energy by some mechanisms. For example, the nuclear heat deposited in the fuel could cause fuel vaporization and its subsequent expansion. This movement would accelerate the surrounding sodium: part of the initial energy in the fuel is thus converted into sodium kinetic energy. This mechanical energy will finally be absorbed, in some way or another, by the reactor vessel. Providing an accurate estimate for the maximum mechanical work that any accidental sequence can do onto the reactor vessel is an essential step in designing a reactor containment that would withstand any load generated by any accident. That would assure accident containment, without consequences for the general public. Fast reactor accident modeling is a complicated task. The outcome of an accident is determined by different physical phenomena, all acting at almost the same time. Safety analysts must track all these different phenomena. Multi-physics codes have been developed for this task. They must contain accurate models for fluid-dynamics, neutronics, and structures. This work has to do with neutronics modeling of such accidents. Past and recent analyses have been limited to the approximate description of the neutronic field, for example by using a rough description of the energy and/or of the angular dependence of the neutron flux. In this work, different neutronic solvers are selected and coupled into a general multi-physics code for fast reactor accident analysis. Performances of each of them is then assessed. Some emphasis has been put also in assessing the speed of these solvers for determining the
Progress of electromagnetic analysis for fusion reactors
International Nuclear Information System (INIS)
Takagi, T.; Ruatto, P.; Boccaccini, L.V.
1998-01-01
This paper describes the recent progress of electromagnetic analysis research for fusion reactors including methods, codes, verification tests and some applications. Due to the necessity of the research effort for the structural design of large tokamak devices since the 1970's with the help of the introduction of new numerical methods and the advancement of computer technologies, three-dimensional analysis methods have become as practical as shell approximation methods. The electromagnetic analysis is now applied to the structural design of new fusion reactors. Some more modeling and verification tests are necessary when the codes are applied to new materials with nonlinear material properties. (orig.)
Operational reactor physics analysis codes (ORPAC)
International Nuclear Information System (INIS)
Kumar, Jainendra; Singh, K.P.; Singh, Kanchhi
2007-07-01
For efficient, smooth and safe operation of a nuclear research reactor, many reactor physics evaluations are regularly required. As part of reactor core management the important activities are maintaining core reactivity status, core power distribution, xenon estimations, safety evaluation of in-pile irradiation samples and experimental assemblies and assessment of nuclear safety in fuel handling/storage. In-pile irradiation of samples requires a prior estimation of the reactivity load due to the sample, the heating rate and the activity developed in it during irradiation. For the safety of personnel handling irradiated samples the dose rate at the surface of shielded flask housing the irradiated sample should be less than 200 mR/Hr.Therefore, a proper shielding and radioactive cooling of the irradiated sample are required to meet the said requirement. Knowledge of xenon load variation with time (Startup-curve) helps in estimating Xenon override time. Monitoring of power in individual fuel channels during reactor operation is essential to know any abnormal power distribution to avoid unsafe situations. Complexities in the estimation of above mentioned reactor parameters and their frequent requirement compel one to use computer codes to avoid possible human errors. For efficient and quick evaluation of parameters related to reactor operations such as xenon load, critical moderator height and nuclear heating and reactivity load of isotope samples/experimental assembly, a computer code ORPAC (Operational Reactor Physics Analysis Codes) has been developed. This code is being used for regular assessment of reactor physics parameters in Dhruva and Cirus. The code ORPAC written in Visual Basic 6.0 environment incorporates several important operational reactor physics aspects on a single platform with graphical user interfaces (GUI) to make it more user-friendly and presentable. (author)
Lead reactor strategy economical analysis
International Nuclear Information System (INIS)
Ciotti, Marco
2013-01-01
Conclusions: • A first attempt to evaluate LFR power plant electricity production cost has been performed; • Electricity price is similar to Gen III + plants; • The estimation accuracy is probably low; • Possible costs reduction could arise from coolant characteristics that may improve safety and simplicity by design; • Accident perception, not acceptable by public opinion, may be changed with low potential energy system (non exploding coolant); • Sustainability improvement could open to a better Public acceptance, depending on us. • Problems may arise in coupling a high capital cost low fuel cost plant in a grid with large amount of intermittent sources with priority dispatch. • Lead fast reactors can compete
Reactor applications of quantitative diffraction analysis
International Nuclear Information System (INIS)
Feguson, I.F.
1976-09-01
Current work in quantitative diffraction analysis was presented under the main headings of: thermal systems, fast reactor systems, SGHWR applications and irradiation damage. Preliminary results are included on a comparison of various new instrumental methods of boron analysis as well as preliminary new results on Zircaloy corrosion, and materials transfer in liquid sodium. (author)
Seismic analysis of fast breeder reactor block
International Nuclear Information System (INIS)
Gantenbein, F.
1990-01-01
Seismic analysis of LMFBR reactor block is complex due mainly to the fluid structure interaction and the 3D geometry of the structure. Analytical methods which have been developed for this analysis will be briefly described in the paper and applications to a geometry similar to SPX1 will be shown
International Nuclear Information System (INIS)
Sumantri, Indro; Purwanto,; Budiyono
2015-01-01
The characteristic of wastewater of food industries with preservative substances is high content of organic substances, degradable and high total suspended solid. High organic content in this waste forced the treatment is biologically and pointed out to anaerobic treatment. Anaerobic showed the better performance of degradation than aerobic for high content organic and also for toxic materials. During that day the treatment of food wastewater is aerobically which is high consume of energy required and high volume of sludge produced. The advantage of anaerobic is save high energy, less product of sludge, less requirement of nutrients of microorganism and high efficiency reduction of organic load. The high efficiency of reduction will reduce the load of further treatment, so that, the threshold limit based on the regulation would be easy to achieve. Research of treatment of wastewater of food industries would be utilized by both big scale industries and small industries using addition of preservative substances. The type reactor of anaerobic process is anaerobic baffled reactor that will give better contact between wastewater and microorganism in the sludge. The variables conducted in this research are the baffled configuration, sludge height, preservative agent contents, hydralic retention time and influence of micro nutrients. The respons of this research are the COD effluent, remaining preservative agent, pH, formation of volatile fatty acid and total suspended solid. The result of this research is kinetic model of the anaerobic baffled reactor, reaction kinetic of preservative agent degradation and technology of treatment wastewater contains preservative agent. The benefit of this research is to solve the treatment of wastewater of food industries with preservative substance in order to achieve wastewater limit regulation and also to prevent the environmental deterioration
Sumantri, Indro; Purwanto, Budiyono
2015-12-01
The characteristic of wastewater of food industries with preservative substances is high content of organic substances, degradable and high total suspended solid. High organic content in this waste forced the treatment is biologically and pointed out to anaerobic treatment. Anaerobic showed the better performance of degradation than aerobic for high content organic and also for toxic materials. During that day the treatment of food wastewater is aerobically which is high consume of energy required and high volume of sludge produced. The advantage of anaerobic is save high energy, less product of sludge, less requirement of nutrients of microorganism and high efficiency reduction of organic load. The high efficiency of reduction will reduce the load of further treatment, so that, the threshold limit based on the regulation would be easy to achieve. Research of treatment of wastewater of food industries would be utilized by both big scale industries and small industries using addition of preservative substances. The type reactor of anaerobic process is anaerobic baffled reactor that will give better contact between wastewater and microorganism in the sludge. The variables conducted in this research are the baffled configuration, sludge height, preservative agent contents, hydralic retention time and influence of micro nutrients. The respons of this research are the COD effluent, remaining preservative agent, pH, formation of volatile fatty acid and total suspended solid. The result of this research is kinetic model of the anaerobic baffled reactor, reaction kinetic of preservative agent degradation and technology of treatment wastewater contains preservative agent. The benefit of this research is to solve the treatment of wastewater of food industries with preservative substance in order to achieve wastewater limit regulation and also to prevent the environmental deterioration.
Energy Technology Data Exchange (ETDEWEB)
Sumantri, Indro; Purwanto,; Budiyono [Chemical Engineering Department, Faculty of Engineering, Diponegoro University Jl. Prof. H. Soedarto, SH, Kampus Baru Tembalang, Semarang (Indonesia)
2015-12-29
The characteristic of wastewater of food industries with preservative substances is high content of organic substances, degradable and high total suspended solid. High organic content in this waste forced the treatment is biologically and pointed out to anaerobic treatment. Anaerobic showed the better performance of degradation than aerobic for high content organic and also for toxic materials. During that day the treatment of food wastewater is aerobically which is high consume of energy required and high volume of sludge produced. The advantage of anaerobic is save high energy, less product of sludge, less requirement of nutrients of microorganism and high efficiency reduction of organic load. The high efficiency of reduction will reduce the load of further treatment, so that, the threshold limit based on the regulation would be easy to achieve. Research of treatment of wastewater of food industries would be utilized by both big scale industries and small industries using addition of preservative substances. The type reactor of anaerobic process is anaerobic baffled reactor that will give better contact between wastewater and microorganism in the sludge. The variables conducted in this research are the baffled configuration, sludge height, preservative agent contents, hydralic retention time and influence of micro nutrients. The respons of this research are the COD effluent, remaining preservative agent, pH, formation of volatile fatty acid and total suspended solid. The result of this research is kinetic model of the anaerobic baffled reactor, reaction kinetic of preservative agent degradation and technology of treatment wastewater contains preservative agent. The benefit of this research is to solve the treatment of wastewater of food industries with preservative substance in order to achieve wastewater limit regulation and also to prevent the environmental deterioration.
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
Development of a point-kinetic verification scheme for nuclear reactor applications
Energy Technology Data Exchange (ETDEWEB)
Demazière, C., E-mail: demaz@chalmers.se; Dykin, V.; Jareteg, K.
2017-06-15
In this paper, a new method that can be used for checking the proper implementation of time- or frequency-dependent neutron transport models and for verifying their ability to recover some basic reactor physics properties is proposed. This method makes use of the application of a stationary perturbation to the system at a given frequency and extraction of the point-kinetic component of the system response. Even for strongly heterogeneous systems for which an analytical solution does not exist, the point-kinetic component follows, as a function of frequency, a simple analytical form. The comparison between the extracted point-kinetic component and its expected analytical form provides an opportunity to verify and validate neutron transport solvers. The proposed method is tested on two diffusion-based codes, one working in the time domain and the other working in the frequency domain. As long as the applied perturbation has a non-zero reactivity effect, it is demonstrated that the method can be successfully applied to verify and validate time- or frequency-dependent neutron transport solvers. Although the method is demonstrated in the present paper in a diffusion theory framework, higher order neutron transport methods could be verified based on the same principles.
Directory of Open Access Journals (Sweden)
Claudio Milton Montenegro Campos
2014-10-01
Full Text Available This study evaluated the treatment of wastewater from coffee wet processing (WCWP in an anaerobic treatment system at a laboratory scale. The system included an acidification/equalization tank (AET, a heat exchanger, an Upflow Anaerobic Sludge Blanket Reactor (UASB, a gas equalization device and a gas meter. The minimum and maximum flow rates and volumetric organic loadings rate (VOLR were 0.004 to 0.037 m 3 d -1 and 0.14 to 20.29 kgCOD m -3 d -1 , respectively. The kinetic parameters measured during the anaerobic biodegradation of the WCWP, with a minimal concentration of phenolic compounds of 50 mg L - ¹, were: Y = 0.37 mgTVS (mgCODremoved -1 , Kd = 0.0075 d-1 , Ks = 1.504mg L -1 , μmax = 0.2 d -1 . The profile of sludge in the reactor showed total solids (TS values from 22,296 to 55,895 mg L -1 and TVS 11,853 to 41,509 mg L -1 , demonstrating a gradual increase of biomass in the reactor during the treatment, even in the presence of phenolic compounds in the concentration already mentioned.
Experimental and kinetic modeling study of 3-methylheptane in a jet-stirred reactor
Karsenty, Florent
2012-08-16
Improving the combustion of conventional and alternative fuels in practical applications requires the fundamental understanding of large hydrocarbon combustion chemistry. The focus of the present study is on a high-molecular-weight branched alkane, namely, 3-methylheptane, oxidized in a jet-stirred reactor. This fuel, along with 2-methylheptane, 2,5-dimethylhexane, and n-octane, are candidate surrogate components for conventional diesel fuels derived from petroleum, synthetic Fischer-Tropsch diesel and jet fuels derived from coal, natural gas, and/or biomass, and renewable diesel and jet fuels derived from the thermochemical treatment of bioderived fats and oils. This study presents new experimental results along with a low- and high-temperature chemical kinetic model for the oxidation of 3-methylheptane. The proposed model is validated against these new experimental data from a jet-stirred reactor operated at 10 atm, over the temperature range of 530-1220 K, and for equivalence ratios of 0.5, 1, and 2. Significant effort is placed on the understanding of the effects of methyl substitution on important combustion properties, such as fuel reactivity and species formation. It was found that 3-methylheptane reacts more slowly than 2-methylheptane at both low and high temperatures in the jet-stirred reactor. © 2012 American Chemical Society.
Energy Technology Data Exchange (ETDEWEB)
Casoli, P.; Authier, N.; Baud, J. [Commissariat a l' energie Atomique, Centre de Valduc, 21120 Is-sur-Tille (France)
2009-07-01
Several experimental devices are operated by the Criticality and Neutron Science Research Department of the CEA Valduc Laboratory. One of these is the metallic core reactor Caliban. The knowledge of the fundamental kinetic parameters of the reactor is very useful, indeed necessary, to the operator. The purpose of this study was to develop and perform experiments allowing to determinate some of these parameters. The prompt neutron decay constant and particularly its value at criticality can be measured with reactor noise techniques such as the interval-distribution, the Feynman variance-to-mean, and the Rossi-{alpha} methods. By introducing the Nelson number, the effective delayed neutron fraction and the average neutron lifetime can also be calculated with the Rossi-{alpha} method. Subcritical, critical, and even supercritical experiments were performed. With the Rossi-{alpha} technique, it was found that the prompt neutron decay constant at criticality was (6.02*10{sup 5} {+-} 9%). Experiments also brought out the limitations of the used experimental parameters. (authors)
Energy Technology Data Exchange (ETDEWEB)
Chiapetto, M., E-mail: mchiapet@sckcen.be [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium); Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France); Messina, L. [DEN-Service de Recherches de Métallurgie Physique, CEA, Université Paris-Saclay, F-91191 Gif-sur-Yvette (France); KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-114 21 Stockholm (Sweden); Becquart, C.S. [Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France); Olsson, P. [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-114 21 Stockholm (Sweden); Malerba, L. [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium)
2017-02-15
This work presents a revised set of parameters to be used in an Object kinetic Monte Carlo model to simulate the microstructure evolution under neutron irradiation of reactor pressure vessel steels at the operational temperature of light water reactors (∼300 °C). Within a “grey-alloy” approach, a more physical description than in a previous work is used to translate the effect of Mn and Ni solute atoms on the defect cluster diffusivity reduction. The slowing down of self-interstitial clusters, due to the interaction between solutes and crowdions in Fe is now parameterized using binding energies from the latest DFT calculations and the solute concentration in the matrix from atom-probe experiments. The mobility of vacancy clusters in the presence of Mn and Ni solute atoms was also modified on the basis of recent DFT results, thereby removing some previous approximations. The same set of parameters was seen to predict the correct microstructure evolution for two different types of alloys, under very different irradiation conditions: an Fe-C-MnNi model alloy, neutron irradiated at a relatively high flux, and a high-Mn, high-Ni RPV steel from the Swedish Ringhals reactor surveillance program. In both cases, the predicted self-interstitial loop density matches the experimental solute cluster density, further corroborating the surmise that the MnNi-rich nanofeatures form by solute enrichment of immobilized small interstitial loops, which are invisible to the electron microscope.
Distributed computing and nuclear reactor analysis
International Nuclear Information System (INIS)
Brown, F.B.; Derstine, K.L.; Blomquist, R.N.
1994-01-01
Large-scale scientific and engineering calculations for nuclear reactor analysis can now be carried out effectively in a distributed computing environment, at costs far lower than for traditional mainframes. The distributed computing environment must include support for traditional system services, such as a queuing system for batch work, reliable filesystem backups, and parallel processing capabilities for large jobs. All ANL computer codes for reactor analysis have been adapted successfully to a distributed system based on workstations and X-terminals. Distributed parallel processing has been demonstrated to be effective for long-running Monte Carlo calculations
Reactor sensor surveillance using noise analysis
International Nuclear Information System (INIS)
Hashemian, H.M.; Thie, J.A.; Upadhyaya, B.R.
1986-01-01
Reactor noise signals, as measured by neutron detectors and process sensors, contain information about the dynamics of the process and sensor characteristics. The extent of sensor characteristics that can be determined from such measurements depends on the sensor type, the property of the process noise exciting the sensor and its location. This paper addresses degradation monitoring of temperature and pressure sensors, analysis methods and results of application to operating pressurized water reactors. In addition, the use of noise analysis for monitoring of pressure sensing lines in nuclear power plants is discussed
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
LOCA analysis of the IRIS reactor
International Nuclear Information System (INIS)
Bajs, T.; Grgic, D.; Cavlina, N.
2003-01-01
The IRIS reactor (International Reactor Innovative and Secure) is an integral, light water cooled, medium power reactor. IRIS has been selected as an International Near Term Deployable (INTD) reactor, within the Generation IV International Forum activities. The IRIS concept addresses the key-requirements defined by the US DOE for next generation reactors, i.e. enhanced reliability and safety, and improved economics. It features innovative, advanced engineering, but it is firmly based on the proven technology of pressurized water reactors (PWR). An innovative safety approach has been developed to mitigate the IRIS response to small-to-medium Loss of Coolant Accident (LOCA). This strategy is based on the interaction of IRIS compact containment with the reactor vessel to limit initial blowdown, and on depressurization through the use of a passive Emergency Heat Removal System (EHRS). A small Automatic Depressurization System (ADS) provides supplementary depressurization capability. A pressure suppression system is provided to limit the pressure peak following the initial blowdown to well below the containment design limit. The ultimate result is that during a small-to-medium LOCA, the core remains covered for an extended period of time, without credit for emergency water injection or external core makeup. The IRIS LOCA response is based on 'maintaining water inventory' rather than on the principle of safety injection. This novel safety approach poses significant issues for computational and analysis methods since the IRIS vessel and containment are strongly coupled, and the system response is based on the interaction between the two. The small break LOCA was calculated using RELAP5/mod3.3 and GOTHIC codes. Break of the largest line connected to the IRIS Reactor Pressure Vessel (RPV) was analyzed. The results of the calculations confirmed good performance of the IRIS system during LOCA. (author)
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)
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.
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)
Sensitivity analysis of power excursion in RSG-GAS reactor due to reactivity insertion
International Nuclear Information System (INIS)
Pinem, Surian; Sembiring, Tagor Malem
2002-01-01
Reactor kinetics has a very important role in reactor operation safety and nuclear reactor control. One of the important aspects in reactor kinetics is power behavior as function of time due to chain reaction in the core. The calculation was performed using kinetic equation and feedback reactivity and evaluated using static power coefficient. Analysis was performed for oxide 250 g, silicide 250 g and silicide 300 g fuel elements with insertion of positive reactivity, negative reactivity and reactivity close to delay neutron fraction. The calculation of power excursion sensitivity showed that the insertion of 0,5 % Δk/k, in the fuel element of silicide 300 g is bigger 5 % than the one of oxide 250 g or silicide 250 g. If inserted by - 1,2 % Δk/k, there is no change among three fuel elements. Therefore, in kinetic point of view, it is showed there is no significant influence in the RSG-GAS reactor operation safety is the current core of oxide 250 g is converted to silicide 250 g or to silicide 300 g
Noise analysis of the Dodewaard boiling water reactor: characteristics and time history
International Nuclear Information System (INIS)
Veer, J.H.C. v.d.; Kema, N.V.
1982-01-01
Reactor noise measurements have been performed in the Dodewaard BWR since the eighth fuel cycle (1978). Analysis of the noise characteristics of the ex-core neutron detectors are reported. As a result characteristics of the global component of the boiling noise and the influence of oscillatory effects in reactor pressure control and steam flow rate are described. The influence of power feedback effects on the detection of global noise at low frequencies is given using point kinetic reactor theory. Results are reported on the behaviour of the neutron noise characteristics during one fuel cycle and on the behaviour from fuel cycle 8 to 11. (author)
The PEC reactor. Safety analysis: Detailed reports
Energy Technology Data Exchange (ETDEWEB)
1988-01-01
In the safety-analysis of the PEC Brasimone reactor (Italy), attention was focused on the role of plant-incident analysis during the design stage and the conclusions reached. The analysis regarded the following: thermohydraulic incidents at full power; incidents with the reactor shut down; reactivity incidents; core local faults; analysis of fuel-handling incidents; engineered safeguards and passive safety features; coolant leakage and sodium fires; research and development studies on the seismic behaviour of the PEC fast reactor; generalized sodium fire; severe accidents, accident sequences with shudown; reference accident. Both the theoretical and experimental analyses demonstrated the adequacy of the design of the PEC fast reactor, aimed at minimizing the consequences of a hypothetical disruptive core accident with mechanical energy release. It was shown that the containment barriers were sized correctly and that the residual heat from a disassembled core would be removed. The re-evaluation of the source term emphasized the conservative nature of the hypotheses assumed in the preliminary safety analysis for calculating the risk to the public.
Effect of lattice-level adjoint-weighting on the kinetics parameters of CANDU reactors
International Nuclear Information System (INIS)
Nichita, Eleodor
2009-01-01
Space-time kinetics calculations for CANDU reactors are routinely performed using the Improved Quasistatic (IQS) method. The IQS method calculates kinetics parameters such as the effective delayed-neutron fraction and generation time using adjoint weighting. In the current implementation of IQS, the direct flux, as well as the adjoint, is calculated using a two-group cell-homogenized reactor model which is inadequate for capturing the effect of the softer energy spectrum of the delayed neutrons. Additionally, there may also be fine spatial effects that are lost because the intra-cell adjoint shape is ignored. The purpose of this work is to compare the kinetics parameters calculated using the two-group cell-homogenized model with those calculated using lattice-level fine-group heterogeneous adjoint weighting and to assess whether the differences are large enough to justify further work on incorporating lattice-level adjoint weighting into the IQS method. A second goal is to evaluate whether the use of a fine-group cell-homogenized lattice-level adjoint, such as is the current practice for Light Water Reactors (LWRs), is sufficient to capture the lattice effects in question. It is found that, for CANDU lattices, the generation time is almost unaffected by the type of adjoint used to calculate it, but that the effective delayed-neutron fraction is affected by the type of adjoint used. The effective delayed-neutron fraction calculated using the two-group cell-homogenized adjoint is 5.2% higher than the 'best' effective delayed-neutron fraction value obtained using the detailed lattice-level fine-group heterogeneous adjoint. The effective delayed-neutron fraction calculated using the fine-group cell-homogenized adjoint is only 1.7% higher than the 'best' effective delayed-neutron fraction value but is still not equal to it. This situation is different from that encountered in LWRs where weighting by a fine-group cell-homogenized adjoint is sufficient to calculate the
Chemical analysis of reactor and commercial columbium
International Nuclear Information System (INIS)
Anon.
1981-01-01
The methods cover the chemical analysis of reactor and commercial columbium having chemical compositions within specified limits. The following analytical procedures are discussed along with apparatus, reagents, photometric practice, safety precautions, sampling, and rounding calculated values: nitrogen, by distillation (photometric) method; molybdenum and tungsten by the dithiol (photometric) method; iron by the 1,10-phenanthroline (photometric) method
A century of enzyme kinetic analysis, 1913 to 2013.
Johnson, Kenneth A
2013-09-02
This review traces the history and logical progression of methods for quantitative analysis of enzyme kinetics from the 1913 Michaelis and Menten paper to the application of modern computational methods today. Following a brief review of methods for fitting steady state kinetic data, modern methods are highlighted for fitting full progress curve kinetics based upon numerical integration of rate equations, including a re-analysis of the original Michaelis-Menten full time course kinetic data. Finally, several illustrations of modern transient state kinetic methods of analysis are shown which enable the elucidation of reactions occurring at the active sites of enzymes in order to relate structure and function. Copyright © 2013 Federation of European Biochemical Societies. Published by Elsevier B.V. All rights reserved.
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)
Taipower's transient analysis methodology for pressurized water reactors
International Nuclear Information System (INIS)
Huang, Pinghue
1998-01-01
The methodology presented in this paper is a part of the 'Taipower's Reload Design and Transient Analysis Methodologies for Light Water Reactors' developed by the Taiwan Power Company (TPC) and the Institute of Nuclear Energy Research. This methodology utilizes four computer codes developed or sponsored by Electric Power Research institute: system transient analysis code RETRAN-02, core thermal-hydraulic analysis code COBRAIIIC, three-dimensional spatial kinetics code ARROTTA, and fuel rod evaluation code FREY. Each of the computer codes was extensively validated. Analysis methods and modeling techniques were conservatively established for each application using a systematic evaluation with the assistance of sensitivity studies. The qualification results and analysis methods were documented in detail in TPC topical reports. The topical reports for COBRAIIIC, ARROTTA. and FREY have been reviewed and approved by the Atomic Energy Council (ABC). TPC 's in-house transient methodology have been successfully applied to provide valuable support for many operational issues and plant improvements for TPC's Maanshan Units I and 2. Major applications include the removal of the resistance temperature detector bypass system, the relaxation of the hot-full-power moderator temperature coefficient design criteria imposed by the ROCAEC due to a concern on Anticipated Transient Without Scram, the reduction of boron injection tank concentration and the elimination of the heat tracing, and the reduction of' reactor coolant system flow. (author)
Directory of Open Access Journals (Sweden)
Sayer C.
2002-01-01
Full Text Available A new reactor, the pulsed sieve plate column (PSPC, was developed to perform continuous emulsion polymerization reactions. This reactor combines the enhanced flexibility of tubular reactors with the mixing behavior provided by sieved plates and by the introduction of pulses that is important to prevent emulsion destabilization. The main objective of this work is to study the kinetics of vinyl acetate (VA emulsion polymerization reactions performed in this PSPC. Therefore, both experimental studies and reaction simulations were performed. Results showed that it is possible to obtain high conversions with rather low residence times in the PSPC.
Energy Technology Data Exchange (ETDEWEB)
Jimenez, Santiago [LITEC-CSIC (Spanish Council for Scientific Research), Maria de Luna 10, 50018 Zaragoza (Spain); Remacha, Pilar; Ballester, Javier [LITEC-CSIC (Spanish Council for Scientific Research), Maria de Luna 10, 50018 Zaragoza (Spain); Fluid Mechanics Group, University of Zaragoza, Maria de Luna 3, 50018 Zaragoza (Spain); Ballesteros, Juan C.; Gimenez, Antonio [ENDESA GENERACION, S.A., Ribera del Loira 60, 28042 Madrid (Spain)
2008-03-15
In this paper the results of a complete set of devolatilization and combustion experiments performed with pulverized ({proportional_to}500 {mu}m) biomass in an entrained flow reactor under realistic combustion conditions are presented. The data obtained are used to derive the kinetic parameters that best fit the observed behaviors, according to a simple model of particle combustion (one-step devolatilization, apparent oxidation kinetics, thermally thin particles). The model is found to adequately reproduce the experimental trends regarding both volatile release and char oxidation rates for the range of particle sizes and combustion conditions explored. The experimental and numerical procedures, similar to those recently proposed for the combustion of pulverized coal [J. Ballester, S. Jimenez, Combust. Flame 142 (2005) 210-222], have been designed to derive the parameters required for the analysis of biomass combustion in practical pulverized fuel configurations and allow a reliable characterization of any finely pulverized biomass. Additionally, the results of a limited study on the release rate of nitrogen from the biomass particle along combustion are shown. (author)
Kinetic analysis of the human knee joint
Directory of Open Access Journals (Sweden)
E Wiczkowski
2008-03-01
Full Text Available The pathology of the calcaneal (Achilles tendon constitutes a serious therapeutical and social problem. Indeed, this tendon is the strongest plantar flexor of the foot that plays an important role in the humangait. Although well known for a long time, no explicit description of the spontaneous subcutaneous rupture of the Achilles tendon can be found in medical or biomechanical literature. So far, neither pathomechanism nor the underlying causes of the tendon’s disruption have been fully elucidated. Many authors concentrate mostly on medical and biological aspects of the condition. The commonly held view is that it is the vascular supply to the tendon that plays a crucial role in pathogenesis of the tendon’s injuries. In fact, the vasculature a change with time and after the age of 30 is significantly reduced leading to the development of regressive alterations within as well as the decrement of the mechanical strength of the tendon. Obviously, interdisciplinary approach encompassing not only medical and biological but also the broadly taken mechanical viewpoint is needed to more comprehensively describe and explain this phenomenon. In the present paper, kinetic analysis of the knee was employed to define the trajectory of the point of initial insertion of the medial head of gastrocnemius, which was then used to determine the point’s route within the motor area extending from the flexion to the full extension of the knee. The obtained data on the trajectory are further utilized to present and define the pathomechanism of the spontaneous rupture of the calcaneal tendon.
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
Safety Analysis for Power Reactor Protection System
International Nuclear Information System (INIS)
Eisawy, E.A.; Sallam, H.
2012-01-01
The main function of a Reactor Protection System (RPS) is to safely shutdown the reactor and prevents the release of radioactive materials. The purpose of this paper is to present a technique and its application for used in the analysis of safety system of the Nuclear Power Plant (NPP). A more advanced technique has been presented to accurately study such problems as the plant availability assessments and Technical Specifications evaluations that are becoming increasingly important. The paper provides the Markov model for the Reactor Protection System of the NPP and presents results of model evaluations for two testing policies in technical specifications. The quantification of the Markov model provides the probability values that the system will occupy each of the possible states as a function of time.
Energy Technology Data Exchange (ETDEWEB)
Vazquez, M.; Martin-Fuertes, F.
2013-07-01
As a complement to the attached code 3D neutron-CIEMAT thermohydraulic added a module to simulate transient. Temporary kinetics is resolved by factoring flow in a spatial part and another storm. MCNP provides the reactivity and updated spatial function and COBRA-IV calculates the temperature distribution. Temporary dependence of amplitude is calculated using time delayed neutron Kinetic equations. As an example of application, examines a transient loss of flow in MYRRHA, a lead-cooled experimental reactor.
The reactor kinetics code tank: a validation against selected SPERT-1b experiments
International Nuclear Information System (INIS)
Ellis, R.J.
1990-01-01
The two-dimensional space-time analysis code TANK is being developed for the simulation of transient behaviour in the MAPLE class of research reactors. MAPLE research reactor cores are compact, light-water-cooled and -moderated, with a high degree of forced subcooling. The SPERT-1B(24/32) reactor core had many similarities to MAPLE-X10, and the results of the SPERT transient experiments are well documented. As a validation of TANK, a series of simulations of certain SPERT reactor transients was undertaken. Special features were added to the TANK code to model reactors with plate-type fuel and to allow for the simulation of rapid void production. The results of a series of super-prompt-critical reactivity step-insertion transient simulations are presented. The selected SPERT transients were all initiated from low power, at ambient temperatures, and with negligible coolant flow. Th results of the TANK simulations are in good agreement with the trends in the experimental SPERT data
Consequence analysis for nuclear reactors, Yongbyon
International Nuclear Information System (INIS)
Kang, Taewook; Jae, Moosung
2017-01-01
Since the Fukushima nuclear power plant accidents in 2011, there have been an increased public anxiety about the safety of nuclear power plants in Korea. The lack of safeguards and facility aging issues at the Yongbyon nuclear facilities have increased doubts. In this study, the consequence analysis for the 5-MWe graphite-moderated reactor in North Korea was performed. Various accident scenarios including accidents at the interim spent fuel pool in the 5-MWe reactor have been developed and evaluated quantitatively. Since data on the design and safety system of nuclear facilities are currently insufficient, the release fractions were set by applying the alternative source terms made for utilization in the analysis of a severe accident by integrating the results of studies of severe accidents occurred before. The calculation results show the early fatality zero deaths and latent cancer fatality about only 13 deaths in Seoul. Thus, actual impacts of a radiological release will be psychological in terms of downwind perceptions and anxiety on the part of potentially exposed populations. Even considering the simultaneous accident occurrence in both 5-MWe graphite-moderated reactor and 100-MWt light water reactor, the consequence analysis using the MACCS2 code shows no significant damage to people in South Korea. (author)
Analysis of dynamic stability and safety of the reactor system by reactor simulator
International Nuclear Information System (INIS)
Raisic, N.
1963-11-01
This document defines the approximations done for establishing a mathematical model of a reactor. Since the model should be used for safety analysis, it was important to choose a mathematical model less stable than the reactor itself. The analysis was performed on the analog computer RAS. Results obtained and conclusions concerned with three possible reactor accidents are presented [sr
International Nuclear Information System (INIS)
Ellis, R.J.
1990-10-01
This report discusses the analysis of a postulated loss-of-regulation (LOR) accident in a metal-fuelled MAPLE Research Reactor. The selected transient scenario involves a slow LOR from low reactor power; the control rods are assumed to withdraw slowly until a trip at 12 MW halts the withdrawal. The simulation was performed using the space-time reactor kinetics computer code TANK, and modelling the reactor in detail in two dimensions and in two neutron-energy groups. Emphasis in this report is placed on the modelling techniques used in TANK and the physics considerations of the analysis
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
Development of a safety analysis code for molten salt reactors
International Nuclear Information System (INIS)
Zhang Dalin; Qiu Suizheng; Su Guanghui
2009-01-01
The molten salt reactor (MSR) well suited to fulfill the criteria defined by the Generation IV International Forum (GIF) is presently revisited all around the world because of different attractive features of current renewed relevance. The MSRs are characterized by using the fluid-fuel, so that their technologies are fundamentally different from those used in the conventional solid-fuel reactors. In this work, in particular, the attention is focused on the safety characteristic analysis of the MSRs, in which a point kinetic model considering the flow effects of the fuel salt is established for the MSRs and calculated by developing a microcomputer code coupling with a simplified heat transfer model in the core. The founded models and developed code are applied to analyze the safety characteristics of the molten salt actinide recycler and transmuter system (MOSART) by simulating three types of basic transient conditions including the unprotected loss of flow, unprotected overcooling accident and unprotected transient overpower. Some reasonable results are obtained for the MOSART, which show that the MOSART conceptual design is an inherently stable reactor design. The present study provides some valuable information for the research and design of the new generation MSRs.
Littel, Rob J.; Versteeg, Geert F.; Swaaij, Wim P.M. van
1992-01-01
Absorption experiments of COS into aqueous solutions of MDEA and DEMEA at 303 K have been carried out in a stirred cell reactor. An absorption model, based on Higbie’s penetration theory, has been developed and applied to interpret the absorption experiments, using the kinetic data obtained in part
Guit, R.P.M.; Kloosterman, M.; Meindersma, G.W.; Mayer, M.; Meijer, E.M.
1991-01-01
The aptitude of a hollow-fiber membrane reactor to det. lipase kinetics was investigated using the hydrolysis of triacetin catalyzed by lipase from C. cylindracea as a model system. The binding of the lipase to the membrane appears not to be very specific (surface adsorption), and probably its
Analysis of nuclear reactor pressure vessel flanges
International Nuclear Information System (INIS)
Oliveira, C.A.N. de; Augusto, O.B.
1985-01-01
This work proposes a methodology for the structural analysis of high diameter nuclear reactor pressure vessel flanges. In the analysis the vessel is divided into shell-of-revolution elements, the flanges are represented by rigid rings, and the bolts are treated as beams. The flexibility method is used for solving the problem, and the results are compared with results obtained by the finite element method. (Author) [pt
Ozonation kinetics of winery wastewater in a pilot-scale bubble column reactor.
Lucas, Marco S; Peres, José A; Lan, Bing Yan; Li Puma, Gianluca
2009-04-01
The degradation of organic substances present in winery wastewater was studied in a pilot-scale, bubble column ozonation reactor. A steady reduction of chemical oxygen demand (COD) was observed under the action of ozone at the natural pH of the wastewater (pH 4). At alkaline and neutral pH the degradation rate was accelerated by the formation of radical species from the decomposition of ozone. Furthermore, the reaction of hydrogen peroxide (formed from natural organic matter in the wastewater) and ozone enhances the oxidation capacity of the ozonation process. The monitoring of pH, redox potential (ORP), UV absorbance (254 nm), polyphenol content and ozone consumption was correlated with the oxidation of the organic species in the water. The ozonation of winery wastewater in the bubble column was analysed in terms of a mole balance coupled with ozonation kinetics modeled by the two-film theory of mass transfer and chemical reaction. It was determined that the ozonation reaction can develop both in and across different kinetic regimes: fast, moderate and slow, depending on the experimental conditions. The dynamic change of the rate coefficient estimated by the model was correlated with changes in the water composition and oxidant species.
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 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)
ANCON, Space-Independent Reactor Kinetics with Linear or Nonlinear Thermal Feedback
International Nuclear Information System (INIS)
Vigil, John C.; Dugan, E.T.
1988-01-01
1 - Description of problem or function: ANCON solves the point-reactor kinetic equations including thermal feedback. Lump-type heat balance equations are used to represent the thermodynamics, and the heat capacity of each lump can vary with temperature. Thermal feedback can be either a linear or a non-linear function of lump temperature, and the impressed reactivity can be either a polynomial or sinusoidal function. 2 - Method of solution: In ANCON the system of coupled first-order differential equations is solved by a method based on continuous analytic continuation (references 2 and 3). The basic procedure consists of expanding all the dependent variables except reactivity in Taylor series, with a truncation error criterion, over successive intervals on the time axis. Variations of the basic procedure are used to increase the efficiency of the method in special situations. Automatic switching from the basic procedure to one of its variations (and vice-versa) may occur during the course of a transient. The method yields an analytic criterion for the magnitude of the time-step at any point in the transient. 3 - Restrictions on the complexity of the problem: The program is currently restricted to a maximum of six delayed neutron groups and a maximum of 56 lumps. Larger problems can be accommodated on a 65 K computer by increasing the dimensions of a few subscripted variables. Also, the code is currently restricted to a constant external transport delays, only the open-loop response of a reactor can be computed with ANCON
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.
Analytic method study of point-reactor kinetic equation when cold start-up
International Nuclear Information System (INIS)
Zhang Fan; Chen Wenzhen; Gui Xuewen
2008-01-01
The reactor cold start-up is a process of inserting reactivity by lifting control rod discontinuously. Inserting too much reactivity will cause short-period and may cause an overpressure accident in the primary loop. It is therefore very important to understand the rule of neutron density variation and to find out the relationships among the speed of lifting control rod, and the duration and speed of neutron density response. It is also helpful for the operators to grasp the rule in order to avoid a start-up accident. This paper starts with one-group delayed neutron point-reactor kinetics equations and provides their analytic solution when reactivity is introduced by lifting control rods discontinuously. The analytic expression is validated by comparison with practical data. It is shown that the analytic solution agrees well with numerical solution. Using this analytical solution, the relationships among neutron density response with the speed of lifting control rod and its duration are also studied. By comparing the results with those under the condition of step inserted reactivity, useful conclusions are drawn
Preliminary safety analysis of molten salt breeder reactor
International Nuclear Information System (INIS)
Cheng Maosong; Dai Zhimin
2013-01-01
Background: The molten salt reactor is one of the six advanced reactor concepts identified by the Generation IV International Forum as a candidate for cooperative development, which is characterized by remarkable advantages in inherent safety, fuel cycle, miniaturization, effective utilization of nuclear resources and proliferation resistance. ORNL finished the conceptual design of Molten Salt Breeder Reactor (MSBR) based on the design, building and operation of Molten Salt Reactor Experiment (MSRE). Purpose: We attempt to implement the preliminary safety analysis of MSBR in order to provide a reference for the design and optimization of MSBR in the future. Methods: According to the conceptual design of MSBR, a model of safety analysis using point kinetics coupled with the simplified heat transfer mechanism is presented. The model is applied to simulate the transient phenomena of MSBR initiated by an abnormal step reactivity addition and an abnormal ramp reactivity addition at full-power equilibrium condition. Results: The thermal power in the core increases rapidly at the beginning and is accompanied by a rise of the fuel and graphite temperatures after 100, 300, 500 and 600 pcm reactivity addition. The maximum outlet temperature of the fuel in the core is at 1250℃ in 500 pcm reactivity addition, but up to 1350℃ in 600 pcm reactivity addition. The maximum of the power and the temperature are delayed and lower in the ramp reactivity addition rather than in the step reactivity addition. Conclusions: Based on the results, when the reactivity inserted is less than 500 pcm in maximum at full power equilibrium condition, the structural material in Hastelloy-N is not melted and can keep integrity without external control action. And it is necessary to try to avoid inserting a reactivity at short time. (authors)
Models and Stability Analysis of Boiling Water Reactors
International Nuclear Information System (INIS)
Dorning, John
2002-01-01
We have studied the nuclear-coupled thermal-hydraulic stability of boiling water reactors (BWRs) using a model that includes: space-time modal neutron kinetics based on spatial w-modes; single- and two-phase flow in parallel boiling channels; fuel rod heat conduction dynamics; and a simple model of the recirculation loop. The BR model is represented by a set of time-dependent nonlinear ordinary differential equations, and is studied as a dynamical system using the modern bifurcation theory and nonlinear dynamical systems analysis. We first determine the stability boundary (SB) - or Hopf bifurcation set- in the most relevant parameter plane, the inlet-subcooling-number/external-pressure-drop plane, for a fixed control rod induced external reactivity equal to the 100% rod line value; then we transform the SB to the practical power-flow map used by BWR operating engineers and regulatory agencies. Using this SB, we show that the normal operating point at 100% power is very stable, that stability of points on the 100% rod line decreases as the flow rate is reduced, and that operating points in the low-flow/high-power region are least stable. We also determine the SB that results when the modal kinetics is replaced by simple point reactor kinetics, and we thereby show that the first harmonic mode does not have a significant effect on the SB. However, we later show that it nevertheless has a significant effect on stability because it affects the basin of attraction of stable operating points. Using numerical simulations we show that, in the important low-flow/high-power region, the Hopf bifurcation that occurs as the SB is crossed is subcritical; hence, growing oscillations can result following small finite perturbations of stable steady-states on the 100% rod line at points in the low-flow/high-power region. Numerical simulations are also performed to calculate the decay ratios (DRs) and frequencies of oscillations for various points on the 100% rod line. It is
Models and Stability Analysis of Boiling Water Reactors
Energy Technology Data Exchange (ETDEWEB)
John Dorning
2002-04-15
We have studied the nuclear-coupled thermal-hydraulic stability of boiling water reactors (BWRs) using a model that includes: space-time modal neutron kinetics based on spatial w-modes; single- and two-phase flow in parallel boiling channels; fuel rod heat conduction dynamics; and a simple model of the recirculation loop. The BR model is represented by a set of time-dependent nonlinear ordinary differential equations, and is studied as a dynamical system using the modern bifurcation theory and nonlinear dynamical systems analysis. We first determine the stability boundary (SB) - or Hopf bifurcation set- in the most relevant parameter plane, the inlet-subcooling-number/external-pressure-drop plane, for a fixed control rod induced external reactivity equal to the 100% rod line value; then we transform the SB to the practical power-flow map used by BWR operating engineers and regulatory agencies. Using this SB, we show that the normal operating point at 100% power is very stable, that stability of points on the 100% rod line decreases as the flow rate is reduced, and that operating points in the low-flow/high-power region are least stable. We also determine the SB that results when the modal kinetics is replaced by simple point reactor kinetics, and we thereby show that the first harmonic mode does not have a significant effect on the SB. However, we later show that it nevertheless has a significant effect on stability because it affects the basin of attraction of stable operating points. Using numerical simulations we show that, in the important low-flow/high-power region, the Hopf bifurcation that occurs as the SB is crossed is subcritical; hence, growing oscillations can result following small finite perturbations of stable steady-states on the 100% rod line at points in the low-flow/high-power region. Numerical simulations are also performed to calculate the decay ratios (DRs) and frequencies of oscillations for various points on the 100% rod line. It is
COMPARATIVE ANALYSIS OF SOME EXISTING KINETIC MODELS ...
African Journals Online (AJOL)
The biosorption of three heavy metal ions namely; Zn2+, Cu2+ and Mn2+ using five microorganisms namely; Bacillus circulans, Pseudomonas aeruginosa, Staphylococcus xylosus, Streptomyces rimosus and Yeast (Saccharomyces sp.) were studied. In this paper, the effectiveness of six existing and two proposed kinetic ...
Fault tree analysis of a research reactor
International Nuclear Information System (INIS)
Hall, J.A.; O'Dacre, D.F.; Chenier, R.J.; Arbique, G.M.
1986-08-01
Fault Tree Analysis Techniques have been used to assess the safety system of the ZED-2 Research Reactor at the Chalk River Nuclear Laboratories. This turned out to be a strong test of the techniques involved. The resulting fault tree was large and because of inter-links in the system structure the tree was not modularized. In addition, comprehensive documentation was required. After a brief overview of the reactor and the analysis, this paper concentrates on the computer tools that made the job work. Two types of tools were needed; text editing and forms management capability for large volumes of component and system data, and the fault tree codes themselves. The solutions (and failures) are discussed along with the tools we are already developing for the next analysis
Design and analysis of prestressed reactor vessels
International Nuclear Information System (INIS)
Burrow, R.E.D.
1978-01-01
This review is intended to draw attention to subjects of interest from papers given at two sessions of the SMiRT 4 conference. The first of these is the structural engineering of prestressed reactor vessels. The topics include developments in the general design of prestressed vessels, structural analysis of PCVRs, model tests and design of penetration, closures and liners for PCVRs. The question of gas cracks was amongst other issues raised. The second of the sessions was concerned with loading conditions and structural analysis of reactor containment. Reference is made to a variety of topics discussed in this session. Particular attention is given to the effects caused by missiles. In concluding, the reviewer suggests the need for a critical assessment of the existing mass of information to sort out the essentials and to bring back some simplicity into design analysis. (UK)
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.
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.
Energy Technology Data Exchange (ETDEWEB)
Perez-Navarro Gomez, A
1973-07-01
The most important kinetic parameters have been measured at the zero power fast reactor CORAL-I by means of the reactor noise analysis in the time domain, using measurements of the count-to-count time intervals. (Author) 69 refs.
comparative analysis of some existing kinetic models with proposed
African Journals Online (AJOL)
IGNATIUS NWIDI
two statistical parameters namely; linear regression coefficient of correlation (R2) and ... Keynotes: Heavy metals, Biosorption, Kinetics Models, Comparative analysis, Average Relative Error. 1. ... If the flow rate is low, a simple manual batch.
Structural analysis of reactor fuel elements
International Nuclear Information System (INIS)
Weeks, R.W.
1977-01-01
An overview of fuel-element modeling is presented that traces the development of codes for the prediction of light-water-reactor and fast-breeder-reactor fuel-element performance. It is concluded that although the mathematical analysis is now far advanced, the development and incorporation of mechanistic constitutive equations has not kept pace. The resultant reliance on empirical correlations severely limits the physical insight that can be gained from code extrapolations. Current efforts include modeling of alternate fuel systems, analysis of local fuel-cladding interactions, and development of a predictive capability for off-normal behavior. Future work should help remedy the current constitutive deficiencies and should include the development of deterministic failure criteria for use in design
Swimming pool reactor reliability and safety analysis
International Nuclear Information System (INIS)
Li Zhaohuan
1997-01-01
A reliability and safety analysis of Swimming Pool Reactor in China Institute of Atomic Energy is done by use of event/fault tree technique. The paper briefly describes the analysis model, analysis code and main results. Meanwhile it also describes the impact of unassigned operation status on safety, the estimation of effectiveness of defense tactics in maintenance against common cause failure, the effectiveness of recovering actions on the system reliability, the comparison of occurrence frequencies of the core damage by use of generic and specific data
Computer graphics in reactor safety analysis
International Nuclear Information System (INIS)
Fiala, C.; Kulak, R.F.
1989-01-01
This paper describes a family of three computer graphics codes designed to assist the analyst in three areas: the modelling of complex three-dimensional finite element models of reactor structures; the interpretation of computational results; and the reporting of the results of numerical simulations. The purpose and key features of each code are presented. The graphics output used in actual safety analysis are used to illustrate the capabilities of each code. 5 refs., 10 figs
Light-water reactor safety analysis codes
International Nuclear Information System (INIS)
Jackson, J.F.; Ransom, V.H.; Ybarrondo, L.J.; Liles, D.R.
1980-01-01
A brief review of the evolution of light-water reactor safety analysis codes is presented. Included is a summary comparison of the technical capabilities of major system codes. Three recent codes are described in more detail to serve as examples of currently used techniques. Example comparisons between calculated results using these codes and experimental data are given. Finally, a brief evaluation of current code capability and future development trends is presented
Research reactor operations for neutron activation analysis
International Nuclear Information System (INIS)
Tv'ehlov, Yu.
2002-01-01
The IAEA Special Manual devoted to quality control during neutron activation analysis (NAA) on research and test reactors is discussed. Three parts of the publication involve presentation of common rules for performance of NAA, quantitative and qualitative analyses, statistic and systematic errors, safety regulations and radioactive waste management. Besides, the publication contains practical manual for the performance of NAA, and examples of different NAA regulating registration forms are presented [ru
N Reactor updated safety analysis report, NUSAR
International Nuclear Information System (INIS)
1978-01-01
An update of the N Reactor safety analysis is presented to reconfirm that the continued operation does not pose undue risk to DOE personnel and property, the public, or the environment. A reanalysis of LOCA and reactivity transients utilizing current codes and methods is made. The principal aspects of the overall submission, a general description, and site characteristics including geography and demography, nearby industrial, transportation and military facilities, meteorology, hydraulic engineering, and geology and seismology are described
Nodal method for fast reactor analysis
International Nuclear Information System (INIS)
Shober, R.A.
1979-01-01
In this paper, a nodal method applicable to fast reactor diffusion theory analysis has been developed. This method has been shown to be accurate and efficient in comparison to highly optimized finite difference techniques. The use of an analytic solution to the diffusion equation as a means of determining accurate coupling relationships between nodes has been shown to be highly accurate and efficient in specific two-group applications, as well as in the current multigroup method
International Nuclear Information System (INIS)
Schaefer, C.; Jansen, A. P. J.
2013-01-01
We have developed a method to couple kinetic Monte Carlo simulations of surface reactions at a molecular scale to transport equations at a macroscopic scale. This method is applicable to steady state reactors. We use a finite difference upwinding scheme and a gap-tooth scheme to efficiently use a limited amount of kinetic Monte Carlo simulations. In general the stochastic kinetic Monte Carlo results do not obey mass conservation so that unphysical accumulation of mass could occur in the reactor. We have developed a method to perform mass balance corrections that is based on a stoichiometry matrix and a least-squares problem that is reduced to a non-singular set of linear equations that is applicable to any surface catalyzed reaction. The implementation of these methods is validated by comparing numerical results of a reactor simulation with a unimolecular reaction to an analytical solution. Furthermore, the method is applied to two reaction mechanisms. The first is the ZGB model for CO oxidation in which inevitable poisoning of the catalyst limits the performance of the reactor. The second is a model for the oxidation of NO on a Pt(111) surface, which becomes active due to lateral interaction at high coverages of oxygen. This reaction model is based on ab initio density functional theory calculations from literature.
A study on naphtha catalytic reforming reactor simulation and analysis.
Liang, Ke-min; Guo, Hai-yan; Pan, Shi-wei
2005-06-01
A naphtha catalytic reforming unit with four reactors in series is analyzed. A physical model is proposed to describe the catalytic reforming radial flow reactor. Kinetics and thermodynamics equations are selected to describe the naphtha catalytic reforming reactions characteristics based on idealizing the complex naphtha mixture by representing the paraffin, naphthene, and aromatic groups by single compounds. The simulation results based above models agree very well with actual operation unit data.
A study on naphtha catalytic reforming reactor simulation and analysis
Liang, Ke-min; Guo, Hai-yan; Pan, Shi-wei
2005-01-01
A naphtha catalytic reforming unit with four reactors in series is analyzed. A physical model is proposed to describe the catalytic reforming radial flow reactor. Kinetics and thermodynamics equations are selected to describe the naphtha catalytic reforming reactions characteristics based on idealizing the complex naphtha mixture by representing the paraffin, naphthene, and aromatic groups by single compounds. The simulation results based above models agree very well with actual operation uni...
Kinetic analysis of dynamic PET data
Energy Technology Data Exchange (ETDEWEB)
Knittel, B.
1983-12-01
Our goal is to quantify regional physiological processes such as blood flow and metabolism by means of tracer kinetic modeling and positron emission tomography (PET). Compartmental models are one way of characterizing the behavior of tracers in physiological systems. This paper describes a general method of estimating compartmental model rate constants from measurements of the concentration of tracers in blood and tissue, taken at multiple time intervals. A computer program which applies the method is described, and examples are shown for simulated and actual data acquired from the Donner 280-Crystal Positron Tomograph.
Kinetic analysis of dynamic PET data
International Nuclear Information System (INIS)
Knittel, B.
1983-12-01
Our goal is to quantify regional physiological processes such as blood flow and metabolism by means of tracer kinetic modeling and positron emission tomography (PET). Compartmental models are one way of characterizing the behavior of tracers in physiological systems. This paper describes a general method of estimating compartmental model rate constants from measurements of the concentration of tracers in blood and tissue, taken at multiple time intervals. A computer program which applies the method is described, and examples are shown for simulated and actual data acquired from the Donner 280-Crystal Positron Tomograph
WRAP: a water reactor analysis package
International Nuclear Information System (INIS)
Anderson, M.M.
1977-06-01
The modular computational system known as the Water Reactor Analysis Package (WRAP) has been developed at the Savannah River Laboratory. WRAP is essentially a reprogrammed version of the RELAP4 computer code with an extensively restructured input format, a dynamic dimensioning capability and additional computational capabilities such as an automatic steady-state option for pressurized water reactors and an automatic restart capability with provision for renodalization. The report describes the capabilities of WRAP at its current stage of development. The addition of new capabilities (e.g., a BWR steady-state capability), the inclusion of improved models (e.g., models in RELAP4/M0D8) and the development of improved numerical techniques to reduce execution time are being planned at this time
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)
Systems analysis of the CANDU 3 Reactor
Energy Technology Data Exchange (ETDEWEB)
Wolfgong, J.R.; Linn, M.A.; Wright, A.L.; Olszewski, M.; Fontana, M.H. [Oak Ridge National Lab., TN (United States)
1993-07-01
This report presents the results of a systems failure analysis study of the CANDU 3 reactor design; the study was performed for the US Nuclear Regulatory Commission. As part of the study a review of the CANDU 3 design documentation was performed, a plant assessment methodology was developed, representative plant initiating events were identified for detailed analysis, and a plant assessment was performed. The results of the plant assessment included classification of the CANDU 3 event sequences that were analyzed, determination of CANDU 3 systems that are ``significant to safety,`` and identification of key operator actions for the analyzed events.
Reliability analysis of reactor pressure vessel intensity
International Nuclear Information System (INIS)
Zheng Liangang; Lu Yongbo
2012-01-01
This paper performs the reliability analysis of reactor pressure vessel (RPV) with ANSYS. The analysis method include direct Monte Carlo Simulation method, Latin Hypercube Sampling, central composite design and Box-Behnken Matrix design. The RPV integrity reliability under given input condition is proposed. The result shows that the effects on the RPV base material reliability are internal press, allowable basic stress and elasticity modulus of base material in descending order, and the effects on the bolt reliability are allowable basic stress of bolt material, preload of bolt and internal press in descending order. (authors)
Probabilistic risk analysis of Angra-1 reactor
International Nuclear Information System (INIS)
Spivak, R.C.; Collussi, I.; Silva, M.C. da; Onusic Junior, J.
1986-01-01
The first phase of probabilistic study for safety analysis and operational analysis of Angra-1 reactor is presented. The study objectives and uses are: to support decisions about safety problems; to identify operational and/or project failures; to amplify operator qualification tests to include accidents in addition to project base; to provide informations to be used in development and/or review of operation procedures in emergency, test and maintenance procedures; to obtain experience for data collection about abnormal accurences; utilization of study results for training operators; and training of evaluation and reliability techniques for the personnel of CNEN and FURNAS. (M.C.K.) [pt
Multidimensional space-time kinetics of a heavy water moderated nuclear reactor
International Nuclear Information System (INIS)
Winn, W.G.; Baumann, N.P.; Jewell, C.E.
1980-01-01
Diffusion theory analysis of a series of multidimensional space-time experiments is appraised in terms of the final experiment of the series. In particular, TRIMHX diffusion calculations were examined for an experiment involving free-fall insertion of a 235 U-bearing rod into a heavy water moderated reactor with a large reflector. The experimental transient flux-tilts were accurately reproduced after cross section adjustments forced agreement between static diffusion calculations and static reactor measurements. The time-dependent features were particularly well modeled, and the bulk of the small discrepancies in space-dependent features should be removable by more refined cross-section adjustments. This experiment concludes a series of space-time experiments that span a wide range of delayed neutron holdback effects. TRIMHX calculations of these experiments demonstrate the accuracy of the modeling employed in the code
Energy Technology Data Exchange (ETDEWEB)
Sagawe, G.; Bahnemann, D. [Inst. fuer Technische Chemie, Univ. Hannover, Hannover (Germany); Brandi, R.J.; Cassano, A.E. [INTEC (Univ. Nacional del Litoral and CONICET), Santa Fe (Argentina)
2003-07-01
Usual applications of photocatalytic reactors for treating wastewater exhibit the difficulty of handling fluids having varying composition and/or concentrations; thus, a detailed kinetic representation may not be possible. When the catalyst activation is obtained employing solar illumination an additional complexity always coexists: solar fluxes are permanently changing with time. For comparing different reacting systems under similar operating conditions and to provide approximate estimations for scaling up purposes, simplified models may be useful. For these approximations the model parameters should be restricted as much as possible to initial physical and boundary conditions such as: initial concentrations (expressed as such or as TOC measurements), flow rate or reactor volume, irradiated reactor area, incident radiation fluxes and a fairly simple experimental observation such as the photonic efficiency. A combination of a new concept: the ''actual observed photonic efficiency'' with ideal reactor models and empirical kinetic rate expressions can be used to provide rather simple working equations that can be efficiently used to describe the performance of practical reactors. In this paper, the method has been developed for the case of a photocatalytic batch reactor (PBR). (orig.)
Seal analysis technology for reactor pressure vessel
International Nuclear Information System (INIS)
Zheng Liangang; Zhang Liping; Yang Yu; Zang Fenggang
2009-01-01
There is the coolant with radiation, high temperature and high pressure in the reactor pressure vessel (RPV). It is closely correlated to RPV sealing capability whether the whole nuclear system work well or not. The aim of this paper is to study the seal analysis method and technology, such as the pre-tensioning of the bolt, elastoplastic contact and coupled technology of thermal and structure. The 3 D elastoplastic seal analysis method really and generally consider the loads and model the contact problem with friction between the contact plates. This method is easier than the specialized seal program and used widely. And it is more really than the 2 D seal analysis method. This 3 D elastoplastic seal analysis method has been successfully used in the design and analysis of RPV. (authors)
Evaluation of nuclear reactor based activation analysis techniques
International Nuclear Information System (INIS)
Obrusnik, I.; Kucera, J.
1977-09-01
A survey is presented of the basic types of activation analysis applied in environmental control. Reactor neutron activation analysis is described (including the reactor as a neutron source, sample activation in the reactor, methodology of neutron activation analysis, sample transport into the reactor and sample packaging after irradiation, instrumental activation analysis with radiochemical separation, data measurement and evaluation, sampling and sample preparation). Sources of environmental contamination with trace elements, sampling and sample analysis by neutron activation are described. The analysis is described of soils, waters and biological materials. Methods are shown of evaluating neutron activation analysis results and of their interpretation for purposes of environmental control. (J.B.)
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
2013-01-01
The Handbook of Nuclear Engineering is an authoritative compilation of information regarding methods and data used in all phases of nuclear engineering. Addressing nuclear engineers and scientists at all academic levels, this five volume set provides the latest findings in nuclear data and experimental techniques, reactor physics, kinetics, dynamics and control. Readers will also find a detailed description of data assimilation, model validation and calibration, sensitivity and uncertainty analysis, fuel management and cycles, nuclear reactor types and radiation shielding. A discussion of radioactive waste disposal, safeguards and non-proliferation, and fuel processing with partitioning and transmutation is also included. As nuclear technology becomes an important resource of non-polluting sustainable energy in the future, The Handbook of Nuclear Engineering is an excellent reference for practicing engineers, researchers and professionals.
The analysis for inventory of experimental reactor high temperature gas reactor type
International Nuclear Information System (INIS)
Sri Kuntjoro; Pande Made Udiyani
2016-01-01
Relating to the plan of the National Nuclear Energy Agency (BATAN) to operate an experimental reactor of High Temperature Gas Reactors type (RGTT), it is necessary to reactor safety analysis, especially with regard to environmental issues. Analysis of the distribution of radionuclides from the reactor into the environment in normal or abnormal operating conditions starting with the estimated reactor inventory based on the type, power, and operation of the reactor. The purpose of research is to analyze inventory terrace for Experimental Power Reactor design (RDE) high temperature gas reactor type power 10 MWt, 20 MWt and 30 MWt. Analyses were performed using ORIGEN2 computer code with high temperatures cross-section library. Calculation begins with making modifications to some parameter of cross-section library based on the core average temperature of 570 °C and continued with calculations of reactor inventory due to RDE 10 MWt reactor power. The main parameters of the reactor 10 MWt RDE used in the calculation of the main parameters of the reactor similar to the HTR-10 reactor. After the reactor inventory 10 MWt RDE obtained, a comparison with the results of previous researchers. Based upon the suitability of the results, it make the design for the reactor RDE 20MWEt and 30 MWt to obtain the main parameters of the reactor in the form of the amount of fuel in the pebble bed reactor core, height and diameter of the terrace. Based on the main parameter or reactor obtained perform of calculation to get reactor inventory for RDE 20 MWT and 30 MWT with the same methods as the method of the RDE 10 MWt calculation. The results obtained are the largest inventory of reactor RDE 10 MWt, 20 MWt and 30 MWt sequentially are to Kr group are about 1,00E+15 Bq, 1,20E+16 Bq, 1,70E+16 Bq, for group I are 6,50E+16 Bq, 1,20E+17 Bq, 1,60E+17 Bq and for groups Cs are 2,20E+16 Bq, 2,40E+16 Bq, 2,60E+16 Bq. Reactor inventory will then be used to calculate the reactor source term and it
A fast reactor transient analysis methodology for personal computers
International Nuclear Information System (INIS)
Ott, K.O.
1993-01-01
A simplified model for a liquid-metal-cooled reactor (LMR) transient analysis, in which point kinetics as well as lumped descriptions of the heat transfer equations in all components are applied, is converted from a differential into an integral formulation. All 30 differential balance equations are implicitly solved in terms of convolution integrals. The prompt jump approximation is applied as the strong negative feedback effectively keeps the net reactivity well below prompt critical. After implicit finite differencing of the convolution integrals, the kinetics equation assumes a new form, i.e., the quadratic dynamics equation. In this integral formulation, the initial value problem of typical LMR transients can be solved with large item steps (initially 1 s, later up to 256 s). This then makes transient problems amenable to a treatment on personal computer. The resulting mathematical model forms the basis for the GW-BASIC program LMR transient calculation (LTC) program. The LTC program has also been converted to QuickBASIC. The running time for a 10-h transient overpower transient is then ∼40 to 10 s, depending on the hardware version (286, 386, or 486 with math coprocessors)
Xing, Zhi L; Zhao, Tian T; Gao, Yan H; Yang, Xu; Liu, Shuai; Peng, Xu Y
2017-02-23
Changing of CH 4 oxidation potential and biological characteristics with CH 4 concentration was studied in a landfill cover soil reactor (LCSR). The maximum rate of CH 4 oxidation reached 32.40 mol d -1 m -2 by providing sufficient O 2 in the LCSR. The kinetic parameters of methane oxidation in landfill cover soil were obtained by fitting substrate diffusion and consumption model based on the concentration profile of CH 4 and O 2 . The values of [Formula: see text] (0.93-2.29%) and [Formula: see text] (140-524 nmol kg soil-DW -1 ·s -1 ) increased with CH 4 concentration (9.25-20.30%), while the values of [Formula: see text] (312.9-2.6%) and [Formula: see text] (1.3 × 10 -5 to 9.0 × 10 -3 nmol mL -1 h -1 ) were just the opposite. MiSeq pyrosequencing data revealed that Methylobacter (the relative abundance was decreased with height of LCSR) and Methylococcales_unclassified (the relative abundance was increased expect in H 80) became the key players after incubation with increasing CH 4 concentration. These findings provide information for assessing CH 4 oxidation potential and changing of biological characteristics in landfill cover soil.
AIREK-MOD, Time Dependent Reactor Kinetics with Feedback Differential Equation
International Nuclear Information System (INIS)
Tamagnini, C.
1984-01-01
1 - Nature of physical problem solved: Solves the reactor kinetic equations with respect to time. A standard form for the reactivity behaviour has been introduced in which the reactivity is given by the sum of a polynomial, sine, cosine and exponential expansion. Tabular form is also included. The presence of feedback differential equations in which the dependence on variables different from the considered one is considered enables many heat-exchange problems to be dealt with. 2 - Method of solution: The method employed for the solution of the differential equations is the one developed 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. Within this limitation there may be n delayed neutron groups (n less than or equal to 25), on m other linear feedback equations (n+m less than or equal to 49). CDC 1604 version was offered by EIR (Institut Federal de Recherches en matiere de reacteurs, Switzerland)
TASK, 1-D Multigroup Diffusion or Transport Theory Reactor Kinetics with Delayed Neutron
International Nuclear Information System (INIS)
Buhl, A.R.; Hermann, O.W.; Hinton, R.J.; Dodds, H.L. Jr.; Robinson, J.C.; Lillie, R.A.
1975-01-01
1 - Description of problem or function: TASK solves the one-dimensional multigroup form of the reactor kinetics equations, using either transport or diffusion theory and allowing an arbitrary number of delayed neutron groups. The program can also be used to solve standard static problems efficiently such as eigenvalue problems, distributed source problems, and boundary source problems. Convergence problems associated with sources in highly multiplicative media are circumvented, and such problems are readily calculable. 2 - Method of solution: TASK employs a combination scattering and transfer matrix method to eliminate certain difficulties that arise in classical finite difference approximations. As such, within-group (inner) iterations are eliminated and solution convergence is independent of spatial mesh size. The time variable is removed by Laplace transformation. (A later version will permit direct time solutions.) The code can be run either in an outer iteration mode or in closed (non-iterative) form. The running mode is dictated by the number of groups times the number of angles, consistent with available storage. 3 - Restrictions on the complexity of the problem: The principal restrictions are available storage and computation time. Since the code is flexibly-dimensioned and has an outer iteration option there are no internal restrictions on group structure, quadrature, and number of ordinates. The flexible-dimensioning scheme allows optional use of core storage. The generalized cylindrical geometry option is not complete in Version I of the code. The feedback options and omega-mode search options are not included in Version I
Kinetic compartmental analysis of carnitine metabolism in the dog
International Nuclear Information System (INIS)
Rebouche, C.J.; Engel, A.G.
1983-01-01
This study was undertaken to quantitate the dynamic parameters of carnitine metabolism in the dog. Six mongrel dogs were given intravenous injections of L-[methyl-3H]carnitine and the specific radioactivity of carnitine was followed in plasma and urine for 19-28 days. The data were analyzed by kinetic compartmental analysis. A three-compartment, open-system model [(a) extracellular fluid, (b) cardiac and skeletal muscle, (c) other tissues, particularly liver and kidney] was adopted and kinetic parameters (carnitine flux, pool sizes, kinetic constants) were derived. In four of six dogs the size of the muscle carnitine pool obtained by kinetic compartmental analysis agreed (+/- 5%) with estimates based on measurement of carnitine concentrations in different muscles. In three of six dogs carnitine excretion rates derived from kinetic compartmental analysis agreed (+/- 9%) with experimentally measured values, but in three dogs the rates by kinetic compartmental analysis were significantly higher than the corresponding rates measured directly. Appropriate chromatographic analyses revealed no radioactive metabolites in muscle or urine of any of the dogs. Turnover times for carnitine were (mean +/- SEM): 0.44 +/- 0.05 h for extracellular fluid, 232 +/- 22 h for muscle, and 7.9 +/- 1.1 h for other tissues. The estimated flux of carnitine in muscle was 210 pmol/min/g of tissue. Whole-body turnover time for carnitine was 62.9 +/- 5.6 days (mean +/- SEM). Estimated carnitine biosynthesis ranged from 2.9 to 28 mumol/kg body wt/day. Results of this study indicate that kinetic compartmental analysis may be applicable to study of human carnitine metabolism
Guidelines for nuclear reactor equipments safety-analysis
International Nuclear Information System (INIS)
1978-01-01
The safety analysis in approving the applications for nuclear reactor constructions (or alterations) is performed by the Committee on Examination of Reactor Safety in accordance with various guidelines prescribed by the Atomic Energy Commission. In addition, the above Committee set forth its own regulations for the safety analysis on common problems among various types of nuclear reactors. This book has collected and edited those guidelines and regulations. It has two parts: Part I includes the guidelines issued to date by the Atomic Energy Commission: and Part II - regulations of the Committee. Part I has collected 8 categories of guidelines which relate to following matters: nuclear reactor sites analysis guidelines and standards for their applications; standard exposure dose of plutonium; nuclear ship operation guidelines; safety design analysis guidelines for light-water type, electricity generating nuclear reactor equipments; safety evaluation guidelines for emergency reactor core cooling system of light-water type power reactors; guidelines for exposure dose target values around light-water type electricity generating nuclear reactor equipments, and guidelines for evaluation of above target values; and meteorological guidelines for the safety analysis of electricity generating nuclear reactor equipments. Part II includes regulations of the Committee concerning - the fuel assembly used in boiling-water type and in pressurized-water type reactors; techniques of reactor core heat designs, etc. in boiling-water reactors; and others
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
Analysis of a kinetic multi-segment foot model part II: kinetics and clinical implications.
Bruening, Dustin A; Cooney, Kevin M; Buczek, Frank L
2012-04-01
Kinematic multi-segment foot models have seen increased use in clinical and research settings, but the addition of kinetics has been limited and hampered by measurement limitations and modeling assumptions. In this second of two companion papers, we complete the presentation and analysis of a three segment kinetic foot model by incorporating kinetic parameters and calculating joint moments and powers. The model was tested on 17 pediatric subjects (ages 7-18 years) during normal gait. Ground reaction forces were measured using two adjacent force platforms, requiring targeted walking and the creation of two sub-models to analyze ankle, midtarsal, and 1st metatarsophalangeal joints. Targeted walking resulted in only minimal kinematic and kinetic differences compared with walking at self selected speeds. Joint moments and powers were calculated and ensemble averages are presented as a normative database for comparison purposes. Ankle joint powers are shown to be overestimated when using a traditional single-segment foot model, as substantial angular velocities are attributed to the mid-tarsal joint. Power transfer is apparent between the 1st metatarsophalangeal and mid-tarsal joints in terminal stance/pre-swing. While the measurement approach presented here is limited to clinical populations with only minimal impairments, some elements of the model can also be incorporated into routine clinical gait analysis. Copyright © 2011 Elsevier B.V. All rights reserved.
Statistical hot spot analysis of reactor cores
International Nuclear Information System (INIS)
Schaefer, H.
1974-05-01
This report is an introduction into statistical hot spot analysis. After the definition of the term 'hot spot' a statistical analysis is outlined. The mathematical method is presented, especially the formula concerning the probability of no hot spots in a reactor core is evaluated. A discussion with the boundary conditions of a statistical hot spot analysis is given (technological limits, nominal situation, uncertainties). The application of the hot spot analysis to the linear power of pellets and the temperature rise in cooling channels is demonstrated with respect to the test zone of KNK II. Basic values, such as probability of no hot spots, hot spot potential, expected hot spot diagram and cumulative distribution function of hot spots, are discussed. It is shown, that the risk of hot channels can be dispersed equally over all subassemblies by an adequate choice of the nominal temperature distribution in the core
Directory of Open Access Journals (Sweden)
Karen López Buriticá
2015-01-01
Full Text Available This paper analyzes the dynamics of a system that models the formation of biofilms in a continuous stirred-tank reactor (CSTR when it is utilized for wastewater treatment. The growth rate of the microorganisms is modeled using two different kinetics, Monod and Haldane kinetics, with the goal of studying the influence of each in the system. The equilibrium points are identified through a stability analysis, and the bifurcations found are characterized.
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)
International Nuclear Information System (INIS)
Devyatykh, G.G.; Gavrishchuk, E.M.; Gibin, A.M.; Dadanov, A.Yu.; Dzyubenko, N.G.; Kaul', A.R.; Nichiporuk, R.V.; Snezhko, N.T.; Ul'yanov, A.A.
1990-01-01
Heterogeneous oxidative decomposition of adduct of yttrium acetylacetonate with o-phenanthroline, copper acetylacetonate and barium dipivaloylmethanate in a flow-type reactor was carried out. The basic kinetic characteristics of chemical precipitation processes of films of yttrium, copper and barium oxides, which are components of high-temperature superconductors, were obtained. The values of activation energy of precipitation process of yttrium, copper and barium oxides constituted 76±10, 108±15, 81±12 (t 600 deg C) respectively
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)
Friction analysis of kinetic schemes : the friction coefficient
Lolkema, Juke S.
1995-01-01
Friction analysis is proposed as the application of general control analysis to single enzymes to describe the control of elementary kinetic steps on the overall catalytic rate. For each transition, a friction coefficient is defined that measures the sensitivity of the turnover rate to the free
Impact of spatial kinetics in severe accident analysis for a large HWR
International Nuclear Information System (INIS)
Morris, E.E.
1994-01-01
The impact on spatial kinetics on the analysis of severe accidents initiated by the unprotected withdrawal of one or more control rods is investigated for a large heavy water reactor. Large inter- and intra-assembly power shifts are observed, and the importance of detailed geometrical modeling of fuel assemblies is demonstrated. Neglect of space-time effects is shown to lead to erroneous estimates of safety margins, and of accident consequences in the event safety margins are exceeded. The results and conclusions are typical of what would be expected for any large, loosely coupled core
A kinetic and equilibrium analysis of silicon carbide chemical vapor deposition on monofilaments
Gokoglu, S. A.; Kuczmarski, M. A.
1993-01-01
Chemical kinetics of atmospheric pressure silicon carbide (SiC) chemical vapor deposition (CVD) from dilute silane and propane source gases in hydrogen is numerically analyzed in a cylindrical upflow reactor designed for CVD on monofilaments. The chemical composition of the SiC deposit is assessed both from the calculated total fluxes of carbon and silicon and from chemical equilibrium considerations for the prevailing temperatures and species concentrations at and along the filament surface. The effects of gas and surface chemistry on the evolution of major gas phase species are considered in the analysis.
Safety Analysis Of Actinide Recycled Fast Power Reactor
International Nuclear Information System (INIS)
Taufik, Mohammad
2001-01-01
Simulation for safety analysis of actinide recycled fast power reactor has been performed. The objective is to know reactor response about ULOF and ULOF and UTOP simultaneous accident. From parameter result such reactivity feedback, power, temperature, and cooled flow rate can conclusion that reactor have inherent safety system, which can back to new Equilibrium State
A numerical technique for reactor subchannel analysis
International Nuclear Information System (INIS)
Fath, Hassan E.S.
1983-01-01
A numerical technique is developed for the solution of the transient boundary layer equations with a moving liquid-vapour interface boundary. The technique uses the finite difference method with the velocity components defined over an Eulerian mesh. A system of interface massless markers is defined where the markers move with the flow field according to a simple kinematic relation between the interface geometry and the fluid velocity. Different applications of nuclear engineering interest are reported with some available results. The present technique is capable of predicting the interface profile near the wall which is important in the reactor subchannel analysis
Computation system for nuclear reactor core analysis
International Nuclear Information System (INIS)
Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.; Petrie, L.M.
1977-04-01
This report documents a system which contains computer codes as modules developed to evaluate nuclear reactor core performance. The diffusion theory approximation to neutron transport may be applied with the VENTURE code treating up to three dimensions. The effect of exposure may be determined with the BURNER code, allowing depletion calculations to be made. The features and requirements of the system are discussed and aspects common to the computational modules, but the latter are documented elsewhere. User input data requirements, data file management, control, and the modules which perform general functions are described. Continuing development and implementation effort is enhancing the analysis capability available locally and to other installations from remote terminals
Physics: A New Reactor Physics Analysis Toolkit
International Nuclear Information System (INIS)
Rabiti, C.; Wang, Y.; Palmiotti, G.; Hiruta, H.; Cogliati, J.; Alfonsi, A.
2011-01-01
In the last year INL has internally pursued the development of a new reactor analysis tool: PHISICS. The software is built in a modular approach to simplify the independent development of modules by different teams and future maintenance. Most of the modules at the time of this summary are still under development (time dependent transport driver, depletion, cross section I/O and interpolation, generalized perturbation theory), while the transport solver INSTANT (Intelligent Nodal and Semi-structured Treatment for Advanced Neutron Transport) has already been widely used1, 2, 3, 4. For this reason we will focus mainly on the presentation of the transport solver INSTANT
Accident analysis for US fast burst reactors
International Nuclear Information System (INIS)
Paternoster, R.; Flanders, M.; Kazi, H.
1994-01-01
In the US fast burst reactor (FBR) community there has been increasing emphasis and scrutiny on safety analysis and understanding of possible accident scenarios. This paper summarizes recent work in these areas that is going on at the different US FBR sites. At this time, all of the FBR facilities have or in the process of updating and refining their accident analyses. This effort is driven by two objectives: to obtain a more realistic scenario for emergency response procedures and contingency plans, and to determine compliance with changing regulatory standards
Pontes, P P; Chernicharo, C A L; Von Sperling, M
2014-08-01
This study aimed at assessing the influence of the return of excess aerobic sludge from a trickling filter (TF) upon the anaerobic digestion process in an upflow anaerobic sludge blanket (UASB) reactor, by evaluating its effect on the kinetics of the decay of specific organic matter (carbohydrates, proteins and lipids), as well as on the concentrations of volatile fatty acids in the UASB reactor. A pilot-scale UASB/TF system was used to perform the experiments, operating with (phase 2) and without (phase 1) excess sludge return from the TF to the UASB reactor. Sampling was carried out at different heights of the UASB reactor (0, 25, 125 and 225-cm height), and profile concentrations were determined for the following parameters: carbohydrates, proteins, lipids and volatile fatty acids. First-order kinetics showed the best fit to the decay of concentrations of carbohydrates, proteins, lipids and chemical oxygen demand (COD) in the UASB reactor. The parameters showing the best fit to the first-order kinetics were proteins and COD, during the sludge return phase. The occurrence of higher apparent reaction constants was further observed during the sludge return phase. For an influent COD concentration of 600 mg L-1 and hydraulic retention times of 2.1, 2.6 and 3.0 h in phase 1, the effluent COD concentrations were 125.3, 88.4 and 62.4 mg L-1, respectively, whereas in phase 2, the effluent COD concentrations were 75.5, 47.6 and 30.1 mg L-1, respectively.
Human Reliability Analysis for Small Modular Reactors
Energy Technology Data Exchange (ETDEWEB)
Ronald L. Boring; David I. Gertman
2012-06-01
Because no human reliability analysis (HRA) method was specifically developed for small modular reactors (SMRs), the application of any current HRA method to SMRs represents tradeoffs. A first- generation HRA method like THERP provides clearly defined activity types, but these activity types do not map to the human-system interface or concept of operations confronting SMR operators. A second- generation HRA method like ATHEANA is flexible enough to be used for SMR applications, but there is currently insufficient guidance for the analyst, requiring considerably more first-of-a-kind analyses and extensive SMR expertise in order to complete a quality HRA. Although no current HRA method is optimized to SMRs, it is possible to use existing HRA methods to identify errors, incorporate them as human failure events in the probabilistic risk assessment (PRA), and quantify them. In this paper, we provided preliminary guidance to assist the human reliability analyst and reviewer in understanding how to apply current HRA methods to the domain of SMRs. While it is possible to perform a satisfactory HRA using existing HRA methods, ultimately it is desirable to formally incorporate SMR considerations into the methods. This may require the development of new HRA methods. More practicably, existing methods need to be adapted to incorporate SMRs. Such adaptations may take the form of guidance on the complex mapping between conventional light water reactors and small modular reactors. While many behaviors and activities are shared between current plants and SMRs, the methods must adapt if they are to perform a valid and accurate analysis of plant personnel performance in SMRs.
Introduction to Safety Analysis Approach for Research Reactors
International Nuclear Information System (INIS)
Park, Suki
2016-01-01
The research reactors have a wide variety in terms of thermal powers, coolants, moderators, reflectors, fuels, reactor tanks and pools, flow direction in the core, and the operating pressure and temperature of the cooling system. Around 110 research reactors have a thermal power greater than 1 MW. This paper introduces a general approach to safety analysis for research reactors and deals with the experience of safety analysis on a 10 MW research reactor with an open-pool and open-tank reactor and a downward flow in the reactor core during normal operation. The general approach to safety analysis for research reactors is described and the design features of a typical open-pool and open-tank type reactor are discussed. The representative events expected in research reactors are investigated. The reactor responses and the thermal hydraulic behavior to the events are presented and discussed. From the minimum CHFR and the maximum fuel temperature calculated, it is ensured that the fuel is not damaged in the step insertion of reactivity by 1.8 mk and the failure of all primary pumps for the reactor with a 10 MW thermal power and downward core flow
Cost benefit analysis of reactor safety systems
International Nuclear Information System (INIS)
Maurer, H.A.
1984-01-01
Cost/benefit analysis of reactor safety systems is a possibility appropriate to deal with reactor safety. The Commission of the European Communities supported a study on the cost-benefit or cost effectiveness of safety systems installed in modern PWR nuclear power plants. The following systems and their cooperation in emergency cases were in particular investigated in this study: the containment system (double containment), the leakage exhaust and control system, the annulus release exhaust system and the containment spray system. The benefit of a safety system is defined according to its contribution to the reduction of the radiological consequences for the environment after a LOCA. The analysis is so far performed in two different steps: the emergency core cooling system is considered to function properly, failure of the emergency core cooling system is assumed (with the possible consequence of core melt-down) and the results may demonstrate the evidence that striving for cost-effectiveness can produce a safer end result than the philosophy of safety at any cost. (orig.)
International Nuclear Information System (INIS)
Cleveland, J.C.
1977-01-01
CORTAP (Core Transient Analysis Program) was developed to predict the dynamic behavior of the High Temperature Gas Cooled Reactor (HTGR) core under normal operational transients and postulated accident conditions. CORTAP is used both as a stand-alone component simulation and as part of the HTGR nuclear steam supply (NSS) system simulation code ORTAP. The core thermal neutronic response is determined by solving the heat transfer equations for the fuel, moderator and coolant in an average powered region of the reactor core. The space independent neutron kinetics equations are coupled to the heat transfer equations through a rapidly converging iterative technique. The code has the capability to determine conservative fuel, moderator, and coolant temperatures in the ''hot'' fuel region. For transients involving a reactor trip, the core heat generation rate is determined from an expression for decay heat following a scram. Nonlinear effects introduced by temperature dependent fuel, moderator, and coolant properties are included in the model. CORTAP predictions will be compared with dynamic test results obtained from the Fort St. Vrain reactor owned by Public Service of Colorado, and, based on these comparisons, appropriate improvements will be made in CORTAP
Energy Technology Data Exchange (ETDEWEB)
Geslot, Benoit; Gruel, Adrien; Pepino, Alexandra; Di Salvo, Jacques; Izarra, Gregoire de; Jammes, Christian; Destouches, Christophe; Blaise, Patrick [CEA, DEN, DER/SPEx, Cadarache, F-13108 St Paul Lez Durance (France)
2015-07-01
MINERVE is a two-zone pool type zero power reactor operated by CEA (Cadarache, France). Kinetic parameters of the core (prompt neutron decay constant, delayed neutron fraction, generation time) have been recently measured using various pile noise experimental techniques, namely Feynman-α, Rossi-α and Cohn-α. Results are discussed and compared to each other's. The measurement campaign has been conducted in the framework of a tri-partite collaboration between CEA, SCK.CEN and PSI. Results presented in this paper were obtained thanks to a time-stamping acquisition system developed by CEA. PSI performed simultaneous measurements which are presented in a companion paper. Signals come from two high efficiency fission chambers located in the graphite reflector next to the core driver zone. Experiments were conducted at critical state with a reactor power of 0.2 W. The core integral fission rate is obtained from a calibrated miniature fission chamber located at the center of the core. Other results obtained in two sub-critical configurations will be presented elsewhere. Best estimate delayed neutron fraction comes from the Cohn-α method: 747 ± 15 pcm (1σ). In this case, the prompt decay constant is 79 ± 0.5 s{sup -1} and the generation time is 94.5 ± 0.7 μs. Other methods give consistent results within the confidence intervals. Experimental results are compared to calculated values obtained from a full 3D core modeling with the CEA-developed Monte Carlo code TRIPOLI4.9 associated with its continuous energy JEFF3.1.1-based library. A very good agreement is observed for the calculated delayed neutron fraction (748.7 ± 0.4 pcm at 1σ), that is a difference of -0.3% with the experiment. On the contrary, a 10% discrepancy is observed for the calculated generation time (104.4 ± 0.1 μs at 1σ). (authors)
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
International Nuclear Information System (INIS)
Shi, Chengbin; Cheng, Maosong; Liu, Guimin
2016-01-01
Highlights: • New point kinetics and thermo-hydraulics models as well as a numerical method are added into RELAP5 code to be suitable for liquid fueled molten salt reactor. • The extended REALP5 code is verified by the experimental benchmarks of MSRE. • The different transient scenarios of the MSBR are simulated to evaluate performance during the transients. - Abstract: The molten salt reactor (MSR) is one of the six advanced reactor concepts declared by the Generation IV International Forum (GIF), which can be characterized by attractive attributes as inherent safety, economical efficiency, natural resource protection, sustainable development and nuclear non-proliferation. It is important to make system safety analysis for nuclear power plant of MSR. In this paper, in order to developing a system analysis code suitable for liquid fueled molten salt reactors, the point kinetics and thermo-hydraulic models as well as the numerical method in thermal–hydraulic transient code Reactor Excursion and Leak Analysis Program (RELAP5) developed at the Idaho National Engineering Laboratory (INEL) for the U.S. Nuclear Regulatory Commission (NRC) are extended and verified by Molten Salt Reactor Experiment (MSRE) experimental benchmarks. And then, four transient scenarios including the load demand change, the primary flow transient, the secondary flow transient and the reactivity transient of the Molten Salt Breeder Reactor (MSBR) are modeled and simulated so as to evaluate the performance of the reactor during the anticipated transient events using the extended RELAP5 code. The results indicate the extended RELAP5 code is effective and well suited to the liquid fueled molten salt reactor, and the MSBR has strong inherent safety characteristics because of its large negative reactivity coefficient. In the future, the extended RELAP5 code will be used to perform transient safety analysis for a liquid fueled thorium molten salt reactor named TMSR-LF developed by the Center
Energy Technology Data Exchange (ETDEWEB)
Shi, Chengbin [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Cheng, Maosong, E-mail: mscheng@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Liu, Guimin [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China)
2016-08-15
Highlights: • New point kinetics and thermo-hydraulics models as well as a numerical method are added into RELAP5 code to be suitable for liquid fueled molten salt reactor. • The extended REALP5 code is verified by the experimental benchmarks of MSRE. • The different transient scenarios of the MSBR are simulated to evaluate performance during the transients. - Abstract: The molten salt reactor (MSR) is one of the six advanced reactor concepts declared by the Generation IV International Forum (GIF), which can be characterized by attractive attributes as inherent safety, economical efficiency, natural resource protection, sustainable development and nuclear non-proliferation. It is important to make system safety analysis for nuclear power plant of MSR. In this paper, in order to developing a system analysis code suitable for liquid fueled molten salt reactors, the point kinetics and thermo-hydraulic models as well as the numerical method in thermal–hydraulic transient code Reactor Excursion and Leak Analysis Program (RELAP5) developed at the Idaho National Engineering Laboratory (INEL) for the U.S. Nuclear Regulatory Commission (NRC) are extended and verified by Molten Salt Reactor Experiment (MSRE) experimental benchmarks. And then, four transient scenarios including the load demand change, the primary flow transient, the secondary flow transient and the reactivity transient of the Molten Salt Breeder Reactor (MSBR) are modeled and simulated so as to evaluate the performance of the reactor during the anticipated transient events using the extended RELAP5 code. The results indicate the extended RELAP5 code is effective and well suited to the liquid fueled molten salt reactor, and the MSBR has strong inherent safety characteristics because of its large negative reactivity coefficient. In the future, the extended RELAP5 code will be used to perform transient safety analysis for a liquid fueled thorium molten salt reactor named TMSR-LF developed by the Center
Theoretical analysis of nuclear reactors (Phase I), I-V, Part III, Reactor poisoning
International Nuclear Information System (INIS)
Pop-Jordanov, J.
1962-07-01
Method was developed for calculation of Xe 135 static effect and kinetic effects of Xe 135 and Sm 149 with separate treatment of iodine effect and influence of reactor poisoning during power increase. Mentioned effects are treated first for uranium fuel and then the basic formulae were generalized for a mixture of fissile material. The annex contains a table with data needed for calculations and the Xe 13 5 microscopic capture cross section dependent on temperature [sr
Parametric systems analysis for ICF hybrid reactors
International Nuclear Information System (INIS)
Berwald, D.H.; Maniscalco, J.A.; Chapin, D.L.
1981-01-01
Parametric design and systems analysis for inertial confinement fusion-fission hybrids are presented. These results were generated as part of the Electric Power Research Institute (EPRI) sponsored Feasibility Assessment of Fusion-Fission Hybrids, using an Inertial Confinement Fusion (ICF) hybrid power plant design code developed in conjunction with the feasibility assessment. The SYMECON systems analysis code, developed by Westinghouse, was used to generate economic results for symbiotic electricity generation systems consisting of the hybrid and its client Light Water Reactors (LWRs). These results explore the entire fusion parameter space for uranium fast fission blanket hybrids, thorium fast fission blanket hybrids, and thorium suppressed fission blanket types are discussed, and system sensitivities to design uncertainties are explored
Integrated systems analysis of the PIUS reactor
Energy Technology Data Exchange (ETDEWEB)
Fullwood, F.; Kroeger, P.; Higgins, J. [Brookhaven National Lab., Upton, NY (United States)] [and others
1993-11-01
Results are presented of a systems failure analysis of the PIUS plant systems that are used during normal reactor operation and postulated accidents. This study was performed to provide the NRC with an understanding of the behavior of the plant. The study applied two diverse failure identification methods, Failure Modes Effects & Criticality Analysis (FMECA) and Hazards & Operability (HAZOP) to the plant systems, supported by several deterministic analyses. Conventional PRA methods were also used along with a scheme for classifying events by initiator frequency and combinations of failures. Principal results of this study are: (a) an extensive listing of potential event sequences, grouped in categories that can be used by the NRC, (b) identification of support systems that are important to safety, and (c) identification of key operator actions.
Integrated systems analysis of the PIUS reactor
International Nuclear Information System (INIS)
Fullwood, F.; Kroeger, P.; Higgins, J.
1993-11-01
Results are presented of a systems failure analysis of the PIUS plant systems that are used during normal reactor operation and postulated accidents. This study was performed to provide the NRC with an understanding of the behavior of the plant. The study applied two diverse failure identification methods, Failure Modes Effects ampersand Criticality Analysis (FMECA) and Hazards ampersand Operability (HAZOP) to the plant systems, supported by several deterministic analyses. Conventional PRA methods were also used along with a scheme for classifying events by initiator frequency and combinations of failures. Principal results of this study are: (a) an extensive listing of potential event sequences, grouped in categories that can be used by the NRC, (b) identification of support systems that are important to safety, and (c) identification of key operator actions
Rachidi, Mariam El; Thion, Sé bastien; Togbé , Casimir; Dayma, Guillaume; Mehl, Marco; Dagaut, Philippe; Pitz, William J.; Zá dor, Judit; Sarathy, Mani
2016-01-01
This study is concerned with the identification and quantification of species generated during the combustion of cyclopentane in a jet stirred reactor (JSR). Experiments were carried out for temperatures between 740 and 1250K, equivalence ratios from 0.5 to 3.0, and at an operating pressure of 10atm. The fuel concentration was kept at 0.1% and the residence time of the fuel/O/N mixture was maintained at 0.7s. The reactant, product, and intermediate species concentration profiles were measured using gas chromatography and Fourier transform infrared spectroscopy. The concentration profiles of cyclopentane indicate inhibition of reactivity between 850-1000K for ϕ = 2.0 and ϕ = 3.0. This behavior is interesting, as it has not been observed previously for other fuel molecules, cyclic or non-cyclic. A kinetic model including both low- and high-temperature reaction pathways was developed and used to simulate the JSR experiments. The pressure-dependent rate coefficients of all relevant reactions lying on the PES of cyclopentyl+O, as well as the C-C and C-H scission reactions of the cyclopentyl radical were calculated at the UCCSD(T)-F12b/cc-pVTZ-F12//M06-2X/6-311++G(d,p) level of theory. The simulations reproduced the unique reactivity trend of cyclopentane and the measured concentration profiles of intermediate and product species. Sensitivity and reaction path analyses indicate that this reactivity trend may be attributed to differences in the reactivity of allyl radical at different conditions, and it is highly sensitive to the C-C/C-H scission branching ratio of the cyclopentyl radical decomposition.
Rachidi, Mariam El
2016-06-23
This study is concerned with the identification and quantification of species generated during the combustion of cyclopentane in a jet stirred reactor (JSR). Experiments were carried out for temperatures between 740 and 1250K, equivalence ratios from 0.5 to 3.0, and at an operating pressure of 10atm. The fuel concentration was kept at 0.1% and the residence time of the fuel/O/N mixture was maintained at 0.7s. The reactant, product, and intermediate species concentration profiles were measured using gas chromatography and Fourier transform infrared spectroscopy. The concentration profiles of cyclopentane indicate inhibition of reactivity between 850-1000K for ϕ = 2.0 and ϕ = 3.0. This behavior is interesting, as it has not been observed previously for other fuel molecules, cyclic or non-cyclic. A kinetic model including both low- and high-temperature reaction pathways was developed and used to simulate the JSR experiments. The pressure-dependent rate coefficients of all relevant reactions lying on the PES of cyclopentyl+O, as well as the C-C and C-H scission reactions of the cyclopentyl radical were calculated at the UCCSD(T)-F12b/cc-pVTZ-F12//M06-2X/6-311++G(d,p) level of theory. The simulations reproduced the unique reactivity trend of cyclopentane and the measured concentration profiles of intermediate and product species. Sensitivity and reaction path analyses indicate that this reactivity trend may be attributed to differences in the reactivity of allyl radical at different conditions, and it is highly sensitive to the C-C/C-H scission branching ratio of the cyclopentyl radical decomposition.
International Nuclear Information System (INIS)
Buckner, M.R.; Hostetler, D.E.; Anderson, M.M.; Dodds, H.L.
1977-01-01
GRASS is a three-dimensional, coupled neutronic and engineering code for analysis of the radioisotope production reactors at the Savannah River Plant. The capabilities of GRASS are reviewed with emphasis on recent additions to model accident conditions involving the transport of molten fuel material and to accurately characterize neutronic and engineering feedback. The general application of GRASS to the Savannah River reactors is discussed, and results are presented for the analyses of severla reactor transient calculations
An overview-probabilistic safety analysis for research reactors
International Nuclear Information System (INIS)
Liu Jinlin; Peng Changhong
2015-01-01
For long-term application, Probabilistic Safety Analysis (PSA) has proved to be a valuable tool for improving the safety and reliability of power reactors. In China, 'Nuclear safety and radioactive pollution prevention 'Twelfth Five Year Plan' and the 2020 vision' raises clearly that: to develop probabilistic safety analysis and aging evaluation for research reactors. Comparing with the power reactors, it reveals some specific features in research reactors: lower operating power, lower coolant temperature and pressure, etc. However, the core configurations may be changed very often and human actions play an important safety role in research reactors due to its specific experimental requirement. As a result, there is a necessary to conduct the PSA analysis of research reactors. This paper discusses the special characteristics related to the structure and operation and the methods to develop the PSA of research reactors, including initiating event analysis, event tree analysis, fault tree analysis, dependent failure analysis, human reliability analysis and quantification as well as the experimental and external event evaluation through the investigation of various research reactors and their PSAs home and abroad, to provide the current situation and features of research reactors PSAs. (author)
ANALYSIS OF KINETICS OF CAST IRON ALLOYING THROUGH SLAG PHASE
Directory of Open Access Journals (Sweden)
O. S. Komarov
2012-01-01
Full Text Available The mechanism of cast iron alloying through slag phase due to use of nickel and copper oxides is considered and the analysis of kinetics regularity of alloying in case of absence of fuse in the form of milled cast-iron chips in slag and at their presence in it is carried out.
Systemization of Design and Analysis Technology for Advanced Reactor
International Nuclear Information System (INIS)
Kim, Keung Koo; Lee, J.; Zee, S. K.
2009-01-01
The present study is performed to establish the base for the license application of the original technology by systemization and enhancement of the technology that is indispensable for the design and analysis of the advanced reactors including integral reactors. Technical reports and topical reports are prepared for this purpose on some important design/analysis methodology; design and analysis computer programs, structural integrity evaluation of main components and structures, digital I and C systems and man-machine interface design. PPS design concept is complemented reflecting typical safety analysis results. And test plans and requirements are developed for the verification of the advanced reactor technology. Moreover, studies are performed to draw up plans to apply to current or advanced power reactors the original technologies or base technologies such as patents, computer programs, test results, design concepts of the systems and components of the advanced reactors. Finally, pending issues are studied of the advanced reactors to improve the economics and technology realization
Argentinean integrated small reactor design and scale economy analysis of integrated reactor
International Nuclear Information System (INIS)
Florido, P. C.; Bergallo, J. E.; Ishida, M. V.
2000-01-01
This paper describes the design of CAREM, which is Argentinean integrated small reactor project and the scale economy analysis results of integrated reactor. CAREM project consists on the development, design and construction of a small nuclear power plant. CAREM is an advanced reactor conceived with new generation design solutions and standing on the large experience accumulated in the safe operation of Light Water Reactors. The CAREM is an indirect cycle reactor with some distinctive and characteristic features that greatly simplify the reactor and also contribute to a highly level of safety: integrated primary cooling system, self pressurized, primary cooling by natural circulation and safety system relying on passive features. For a fully doupled economic evaluation of integrated reactors done by IREP (Integrated Reactor Evaluation Program) code transferred to IAEA, CAREM have been used as a reference point. The results shows that integrated reactors become competitive with power larger than 200MWe with Argentinean cheapest electricity option. Due to reactor pressure vessel construction limit, low pressure drop steam generator are used to reach power output of 200MWe for natural circulation. For forced circulation, 300MWe can be achieved. (author)
Reactor noise analysis applications in NPP I and C systems
Energy Technology Data Exchange (ETDEWEB)
Gloeckler, O. [International Atomic Energy Agency, Wagramer Strosse 5, A-1400 Vienna, Austria Ontario Power Generation, 230 Westney Road South, Ajax, Ont. L1S 7R3 (Canada)
2006-07-01
Reactor noise analysis techniques are used in many NPPs on a routine basis as 'inspection tools' to get information on the dynamics of reactor processes and their instrumentation in a passive, non-intrusive way. The paper discusses some of the tasks and requirements an NPP has to take to implement and to use the full advantages of reactor noise analysis techniques. Typical signal noise analysis applications developed for the monitoring of the reactor shutdown system and control system instrumentation of the Candu units of Ontario Power Generation and Bruce Power are also presented. (authors)
A tool model for predicting atmospheric kinetics with sensitivity analysis
Institute of Scientific and Technical Information of China (English)
无
2001-01-01
A package( a tool model) for program of predicting atmospheric chemical kinetics with sensitivity analysis is presented. The new direct method of calculating the first order sensitivity coefficients using sparse matrix technology to chemical kinetics is included in the tool model, it is only necessary to triangularize the matrix related to the Jacobian matrix of the model equation. The Gear type procedure is used to integrate amodel equation and its coupled auxiliary sensitivity coefficient equations. The FORTRAN subroutines of the model equation, the sensitivity coefficient equations, and their Jacobian analytical expressions are generated automatically from a chemical mechanism. The kinetic representation for the model equation and its sensitivity coefficient equations, and their Jacobian matrix is presented. Various FORTRAN subroutines in packages, such as SLODE, modified MA28, Gear package, with which the program runs in conjunction are recommended.The photo-oxidation of dimethyl disulfide is used for illustration.
JAERI thermal reactor standard code system for reactor design and analysis SRAC
International Nuclear Information System (INIS)
Tsuchihashi, Keichiro
1985-01-01
SRAC, JAERI thermal reactor standard code system for reactor design and analysis, developed in Japan Atomic Energy Research Institute, is for all types of thermal neutron nuclear design and analysis. The code system has undergone extensive verifications to confirm its functions, and has been used in core modification of the research reactor, detailed design of the multi-purpose high temperature gas reactor and analysis of the experiment with a critical assembly. In nuclear calculation with the code system, multi-group lattice calculation is first made with the libraries. Then, with the resultant homogeneous equivalent group constants, reactor core calculation is made. Described are the following: purpose and development of the code system, functions of the SRAC system, bench mark tests and usage state and future development. (Mori, K.)
Report of the Panel on Kinetics and Applications of Pulsed Research Reactors
International Nuclear Information System (INIS)
1966-03-01
The question of the dynamic behaviour of a reactor subjected to a highly supercritical condition has had special interest for reactor physicists because of the reactor safety implications involved. The large amount of experimental and theoretical work done during the past dozen years or sc to understand fast transient behaviour and the inherent safety characteristics of reactors has not only helped to ease the concern of reactor designers about the consequences of a prompt critical excursion, but, by demonstrating the feasibility of operating certain types of reactors in a pulsed fashion has led to the development of an extremely useful research tool. Pulsed research reactors of a number of different kinds are in operation, while newer, higher performance systems are presently being designed and constructed. Such devices are being used more and more for research in physics, chemistry and reactor engineering, and with the advent of the newer machines, new research areas will become accessible. Because of the rapidly growing interest in the utilization of pulsed reactors for research, the IAEA convened a panel of experts in this field to review recent progress in the design and application of pulsed reactors to consider the problems of converting an existing pool type research reactor to a pulsing types and to consider future potentialities. The panel met in Vienna from 17 to 21 May 1965. This report of the panel summarizes the discussions
A Preliminary Analysis of Reactor Performance Test (LOEP) for a Research Reactor
Energy Technology Data Exchange (ETDEWEB)
Kim, Hyeonil; Park, Su-Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2015-10-15
The final phase of commissioning is reactor performance test, which is to prove the integrated performance and safety of the research reactor at full power with fuel loaded such as neutron power calibration, Control Absorber Rod/Second Shutdown Rod drop time, InC function test, Criticality, Rod worth, Core heat removal with natural mechanism, and so forth. The last test will be safety-related one to assure the result of the safety analysis of the research reactor is marginal enough to be sure about the nuclear safety by showing the reactor satisfies the acceptance criteria of the safety functions such as for reactivity control, maintenance of auxiliaries, reactor pool water inventory control, core heat removal, and confinement isolation. After all, the fuel integrity will be ensured by verifying there is no meaningful change in the radiation levels. To confirm the performance of safety equipment, loss of normal electric power (LOEP), possibly categorized as Anticipated Operational Occurrence (AOO), is selected as a key experiment to figure out how safe the research reactor is before turning over the research reactor to the owner. This paper presents a preliminary analysis of the reactor performance test (LOEP) for a research reactor. The results showed how different the transient between conservative estimate and best estimate will look. Preliminary analyses have shown all probable thermal-hydraulic transient behavior of importance as to opening of flap valve, minimum critical heat flux ratio, the change of flow direction, and important values of thermal-hydraulic parameters.
Light water reactor lower head failure analysis
International Nuclear Information System (INIS)
Rempe, J.L.; Chavez, S.A.; Thinnes, G.L.
1993-10-01
This document presents the results from a US Nuclear Regulatory Commission-sponsored research program to investigate the mode and timing of vessel lower head failure. Major objectives of the analysis were to identify plausible failure mechanisms and to develop a method for determining which failure mode would occur first in different light water reactor designs and accident conditions. Failure mechanisms, such as tube ejection, tube rupture, global vessel failure, and localized vessel creep rupture, were studied. Newly developed models and existing models were applied to predict which failure mechanism would occur first in various severe accident scenarios. So that a broader range of conditions could be considered simultaneously, calculations relied heavily on models with closed-form or simplified numerical solution techniques. Finite element techniques-were employed for analytical model verification and examining more detailed phenomena. High-temperature creep and tensile data were obtained for predicting vessel and penetration structural response
Light water reactor lower head failure analysis
Energy Technology Data Exchange (ETDEWEB)
Rempe, J.L.; Chavez, S.A.; Thinnes, G.L. [EG and G Idaho, Inc., Idaho Falls, ID (United States)] [and others
1993-10-01
This document presents the results from a US Nuclear Regulatory Commission-sponsored research program to investigate the mode and timing of vessel lower head failure. Major objectives of the analysis were to identify plausible failure mechanisms and to develop a method for determining which failure mode would occur first in different light water reactor designs and accident conditions. Failure mechanisms, such as tube ejection, tube rupture, global vessel failure, and localized vessel creep rupture, were studied. Newly developed models and existing models were applied to predict which failure mechanism would occur first in various severe accident scenarios. So that a broader range of conditions could be considered simultaneously, calculations relied heavily on models with closed-form or simplified numerical solution techniques. Finite element techniques-were employed for analytical model verification and examining more detailed phenomena. High-temperature creep and tensile data were obtained for predicting vessel and penetration structural response.
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
CENTAR code for extended nonlinear transient analysis of extraterrestrial reactor systems
International Nuclear Information System (INIS)
Nassersharif, B.; Peer, J.S.; DeHart, M.D.
1987-01-01
Current interest in the application of nuclear reactor-driven power systems to space missions has generated a need for a systems simulation code to model and analyze space reactor systems; such a code has been initiated at Texas A and M, and the first version is nearing completion; release was anticipated in the fall of 1987. This code, named CENTAR (Code for Extended Nonlinear Transient Analysis of Extraterrestrial Reactor Systems), is designed specifically for space systems and is highly vectorizable. CENTAR is composed of several specialized modules. A fluids module is used to model fluid behavior throughout the system. A wall heat transfer module models the heat transfer characteristics of all walls, insulation, and structure around the system. A fuel element thermal analysis module is used to predict the temperature behavior and heat transfer characteristics of the reactor fuel rods. A kinetics module uses a six-group point kinetics formulation to model reactivity feedback and control and the ANS 5.1 decay-heat curve to model shutdown decay-heat production. A pump module models the behavior of thermoelectric-electromagnetic pumps, and a heat exchanger module models not only thermal effects in thermoelectric heat exchangers, but also predicts electrical power production for a given configuration. Finally, an accumulator module models coolant expansion/contraction accumulators
Krasikov, E.; Nikolaenko, V.
2017-01-01
Fast neutron intensity influence on reactor materials radiation damage is a critically important question in the problem of the correct use of the accelerated irradiation tests data for substantiation of the materials workability in real irradiation conditions that is low neutron intensity. Investigations of the fast neutron intensity (flux) influence on radiation damage and experimental data scattering reveal the existence of non-monotonous sections in kinetics of the reactor pressure vessels (RPV) steel damage. Discovery of the oscillations as indicator of the self-organization processes presence give reasons for new ways searching on reactor pressure vessel (RPV) steel radiation stability increasing and attempt of the self-restoring metal elaboration. Revealing of the wavelike process in the form of non monotonous parts of the kinetics of radiation embrittlement testifies that periodic transformation of the structure take place. This fact actualizes the problem of more precise definition of the RPV materials radiation embrittlement mechanisms and gives reasons for search of the ways to manage the radiation stability (nanostructuring and so on to stimulate the radiation defects annihilation), development of the means for creating of more stableness self recovering smart materials.
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
International Nuclear Information System (INIS)
Pham, Nhu Viet Ha
2011-02-01
To predict the space-time dependent behavior of a nuclear reactor, the conventional space-dependent kinetics equations are widely used for treating the spatial variables. However, the solutions of such deterministic space-dependent kinetics equations, which give only the mean values of the neutron population and the delayed neutron precursor concentrations, do not offer sufficient insight into the actual dynamic processes within a reactor, where the interacting populations vary randomly with space and time. It is also noted that at high power levels, the random behavior of a reactor is negligible but at low power levels, such as at start-up, random fluctuations in population dynamics can be significant. To mathematically describe the evolution of the state of a nuclear reactor using a set of stochastic kinetics equations, the forward stochastic model (FSM) in stochastic kinetics theory is devised through the concept of reactor transition probability and its probability generating function as the spatial domain of a reactor is partitioned into a number of space cells. Nevertheless, the FSM equations for the mean value of neutron and precursor distribution are deterministic-like. Furthermore, the numerical treatment of the FSM equations for the means, variances, and covariances is quite complicated and time-consuming. In the present study, a generalized stochastic model (called the stochastic space-dependent kinetics model or SSKM) based on the FSM and the Its stochastic differential equations was newly developed for the analysis of monoenergetic spacetime nuclear reactor kinetics in one dimension. First, the FSM equations for determining the mean values of neutron and delayed-neutron precursor populations were considered as the deterministic ones without taking into account their variances and covariances. Second, the system of interest was randomized again in the light of the Its stochastic differential equations in order to derive the SSKM. The proposed model
Probabilistic safety analysis applied to RBMK reactors
International Nuclear Information System (INIS)
Gerez Martin, L.; Fernandez Ramos, P.
1995-01-01
The project financed by the European Union ''Revision of RBMK Reactor Safety was divided into nine Topic Groups dealing with different aspects of safety. The area covered by Topic Group 9 was Probabilistic Safety Analysis. TG9 will have touched on some of the problems discussed by other groups, although in terms of the systematic quantification of the impact of design characteristics and RBMK reactor operating practices on the risk of core damage. On account of the reduced time scale and the resources available for the project, the analysis was made using a simplified method based on the results of PSAs conducted in Western countries and on the judgement of the group members. The simplifies method is based on the concepts of Qualification, Redundancy and Automatic Actuation of the systems considered. PSA experience shows that systems complying with the above-mentioned concepts have a failure probability of 1.0E-3 when redundancy is simple, ie two similar equipment items capable of carrying out the same function. In general terms, this value can be considered to be dominated by potential common cause failures. The value considered above changes according to factors that have a positive effect upon it, such as an additional redundancy with a different equipment item (eg a turbo pumps and a motor pump), individual trains with good separations, etc, or a negative effect, such as the absence of suitable periodical tests, the need for operators to perform manual operations, etc. Similarly, possible actions required by the operator during accident sequences are assigned failure probability values between 1 and 1.0E-4, according to the complexity of the action (including local actions to be performed outside the control room) and the time available
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.
Laser excitation kinetic phosphorimetry for uranium analysis
International Nuclear Information System (INIS)
Bushaw, B.A.
1983-02-01
Laser induced phosphorescence with time resolved photon counting detection has been used to measure uranium, as uranyl phosphate, in aqueous solution at room temperature. Demonstrated detection limits are below 10 parts per trillion. Multichannel scaler (MCS) photon counting is used for the rejection of laser and raman scattering, discrimination against prompt fluorescing species such as organics, and to provide a simultaneous determination of the uranyl emission lifetime. The lifetime is then used as a direct (photophysically correct) instrumental correction for matrix quenching and temperature effects. The MCS counting mode also allows rapid measurements (a few seconds at the ppb level, 5 minutes for pptr). Sample volumes of 2 to 3 ml have been used although 1/2 ml would be sufficient. The method has been used for: (1) evaluating uranium contamination in ultrapure solvents and trace analytical procedures, in the low part per trillion range, (2) directly measuring (without chemical pretreatment or separation) urine samples at the 10 ppb level, (3) direct measurements on drinking and ground waters at the sub ppb level, and (4) using oxidative digestion as the only preparation, both urine and brine waters have been measured at the sub ppb level. For the brine water analysis the determined value (0.36 ppb) agreed within 10% of the number determined by U-233 tracer - iron hydroxide coprecipitation - filament source mass spectrometric analysis
Comparison of one-dimensional and point kinetics for various light water reactor transients
International Nuclear Information System (INIS)
Naser, J.A.; Lin, C.; Gose, G.C.; McClure, J.A.; Matsui, Y.
1985-01-01
The object of this paper is to compare the results from the three kinetics options: 1) point kinetics; 2) point kinetics by not changing the shape function; and 3) one-dimensional kinetics for various transients on both BWRs and PWRs. A systematic evaluation of the one-dimensional kinetics calculation and its alternatives is performed to determine the status of these models and to identify additional development work. In addition, for PWRs, the NODEP-2 - NODETRAN and SIMULATE - SIMTRAN paths for calculating kinetics parameters are compared. This type of comparison has not been performed before and is needed to properly evaluate the RASP methodology of which these codes are a part. It should be noted that RASP is in its early pre-release stage and this is the first serious attempt to examine the consistency between these two similar but different methods of generating physics parameters for the RETRAN computer code
THYDE-NEU: Nuclear reactor system analysis code
International Nuclear Information System (INIS)
Asahi, Yoshiro
2002-03-01
THYDE-NEU is applicable not only to transient analyses, but also to steady state analyses of nuclear reactor systems (NRSs). In a steady state analysis, the code generates a solution satisfying the transient equations without external disturbances. In a transient analysis, the code calculates temporal NRS behaviors in response to various external disturbances in such a way that mass and energy of the coolant as well as the number of neutrons conserve. The first half of the report is the description of the methods and models for use in the THYDE-NEU code, i.e., (1) the thermal-hydraulic network model, (2) the spatial kinetics model, (3) the heat sources in fuel, (4) the heat transfer correlations, (5) the mechanical behavior of clad and fuel, and (6) the steady state adjustment. The second half of the report is the users' mannual containing the items; (1) the program control, (2) the input requirements, (3) the execution of THYDE-NEU jobs, (4) the output specifications and (5) the sample calculation. (author)
Dynamic analysis and application of fuel elements pneumatic transportation in a pebble bed reactor
International Nuclear Information System (INIS)
Liu, Hongbing; Du, Dong; Han, Zandong; Zou, Yirong; Pan, Jiluan
2015-01-01
Almost 10,000 spherical fuel elements are transported pneumatically one by one in the pipeline outside the core of a pebble bed reactor every day. Any failure in the transportation will lead to the shutdown of the reactor, even safety accidents. In order to ensure a stable and reliable transportation, it's of great importance to analyze the motion and force condition of the fuel element. In this paper, we focus on the dynamic analysis of the pneumatic transportation of the fuel element and derive kinetic equations. Then we introduce the design of the transportation pipeline. On this basis we calculate some important data such as the velocity of the fuel element, the force between the fuel element and the pipeline and the efficiency of the pneumatic transportation. Then we analyze these results and provide some suggestions for the design of the pipeline. The experiment was carried out on an experimental platform. The velocities of the fuel elements were measured. The experimental results were consistent with and validated the theoretical analysis. The research may offer the basis for the design of the transportation pipeline and the optimization of the fuel elements transportation in a pebble bed reactor. - Highlights: • The kinetic equations of the fuel element in pneumatic transportation are derived. • The dynamic characteristics of the fuel element are analyzed. • Some important parameters are calculated based on the kinetic equations. • The experimental results were consistent with the analysis and verified the analysis. • This paper may offer an important guide to the research of a pebble bed reactor
Development of comprehensive and versatile framework for reactor analysis, MARBLE
International Nuclear Information System (INIS)
Yokoyama, Kenji; Hazama, Taira; Numata, Kazuyuki; Jin, Tomoyuki
2014-01-01
Highlights: • We have developed a neutronics code system for reactor analysis. • The new code system covers all five phases of the core design procedures. • All the functionalities are integrated and validated in the same framework. • The framework supports continuous improvement and extension. • We report results of validation and practical applications. - Abstract: A comprehensive and versatile reactor analysis code system, MARBLE, has been developed. MARBLE is designed as a software development framework for reactor analysis, which offers reusable and extendible functions and data models based on physical concepts, rather than a reactor analysis code system. From a viewpoint of the code system, it provides a set of functionalities utilized in a detailed reactor analysis scheme for fast criticality assemblies and power reactors, and nuclear data related uncertainty quantification such as cross-section adjustment. MARBLE includes five sub-systems named ECRIPSE, BIBLO, SCHEME, UNCERTAINTY and ORPHEUS, which are constructed of the shared functions and data models in the framework. By using these sub-systems, MARBLE covers all phases required in fast reactor core design prediction and improvement procedures, i.e. integral experiment database management, nuclear data processing, fast criticality assembly analysis, uncertainty quantification, and power reactor analysis. In the present paper, these functionalities are summarized and system validation results are described
International Nuclear Information System (INIS)
Chen, G.S.; Christenson, J.M.
1985-01-01
In this paper, the authors present some initial results from an investigation of the application of a locally one-dimensional (LOD) finite difference method to the solution of the two-dimensional, two-group reactor kinetics equations. Although the LOD method is relatively well known, it apparently has not been previously applied to the space-time kinetics equations. In this investigation, the LOD results were benchmarked against similar computational results (using the same computing environment, the same programming structure, and the same sample problems) obtained by the TWIGL program. For all of the problems considered, the LOD method provided accurate results in one-half to one-eight of the time required by the TWIGL program
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.
Energy Technology Data Exchange (ETDEWEB)
Zbinden, M; Durbec, V
1996-12-01
A new concept of industrial wear model adapted to components of nuclear plants is proposed. Its originality is to be supported, on one hand, by experimental results obtained via wear machines of relatively short operational times, and, on the other hand, by the information obtained from the operating feedback over real wear kinetics of the reactors components. The proposed model is illustrated by an example which corresponds to a specific real situation. The determination of the coefficients permitting to cover all assembly of configurations and the validation of the model in these configurations have been the object of the most recent work. (author). 34 refs.
International Nuclear Information System (INIS)
Zbinden, M.; Durbec, V.
1996-12-01
A new concept of industrial wear model adapted to components of nuclear plants is proposed. Its originality is to be supported, on one hand, by experimental results obtained via wear machines of relatively short operational times, and, on the other hand, by the information obtained from the operating feedback over real wear kinetics of the reactors components. The proposed model is illustrated by an example which corresponds to a specific real situation. The determination of the coefficients permitting to cover all assembly of configurations and the validation of the model in these configurations have been the object of the most recent work. (author)
Automated Determination of Oxygen-Dependent Enzyme Kinetics in a Tube-in-Tube Flow Reactor
DEFF Research Database (Denmark)
Ringborg, Rolf Hoffmeyer; Pedersen, Asbjørn Toftgaard; Woodley, John
2017-01-01
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 KMO enzyme systems. For the first time, this paves the way to integrate kinetic data into the protein engineering cycle....
Seismic analysis of the Aguirre Nuclear Reactor
International Nuclear Information System (INIS)
Sepulveda Soza, Cristian
1999-01-01
This thesis aims to verify the seismic design of the Aguirre Nuclear Reactor using the finite elements method and comparing the results with the original analysis. The study focused on the dynamic interaction of soil and structures, using the ANSYS program for the analysis, which was implemented for a work station under a UNIX platform belonging to the Chilean Nuclear Energy Commission. The modeling of the structures was carried out following International Atomic Energy recommendations, those of the makers of the Swanson Analysis Systems program and the prior study by S y S Ingenieros Consultores. Two-dimensional models were developed with axial and symmetry and three-dimensional models with symmetric and asymmetric plans, where the retaining building, the pond block and the soil down to the basal rock were included. The seismic stresses were defined according to the Chilean Standard NCh433.of96, using the spectrum of design accelerations for type II soils for the structural models and type IV for the soil-structure interaction models.The results of interest for this study are: the compression and cutting tensions, the unitary cut distortions and the displacements, which are shown graphically and are compared between the different models and with the original analysis. A sensitivity analysis was prepared for the models with axial symmetry considering soil reaction coefficient values of 20, 10, 5, 2, 1 and 0.5 kp/cm 3 ; and four screens with maximum sizes of 100, 50, 25 and 12.5 cm. The behavior of the stressed materials was studied as well as the result of the seismic stress (CS)
Reactor neutron activation for multielemental analysis
International Nuclear Information System (INIS)
Reddy, A.V.R.
1999-01-01
Neutron Activation Analysis using single comparator (K 0 NAA method) has been used for obtaining multielemental profiles in a variety of matrices related to environment. Gold was used as the comparator. Neutron flux was characterised by determining f, the epithermal to thermal neutron flux ratio and cc, the deviation from ideal shape of the neutron spectrum. The f and a were determined in different irradiation positions in APSARA reactor, PCF position in CIRUS reactor and tray rod position in Dhruva reactor using both cadmium cut off and multi isotope detector methods. High resolution gamma ray spectrometry was used for radioactive assay of the activation products. This technique is being used for multielement analysis in a variety of matrices like lake sediments, sea nodules and crusts, minerals, leaves, cereals, pulses, leaves, water and soil. Elemental profiles of the sediments corresponding to different depths from Nainital lake were determined and used to understand the history of natural absorption/desorption pattern of the previous 160 years. Ferromanganese crusts from different locations of Indian Ocean were analysed with a view to studying the distribution of some trace elements along with Fe and Mn. Variation of Mn/Fe ratio was used to identify the nature of the crusts as hydrogenous or hydrothermal. Fe-rich and Fe-depleted nodules from Indian Ocean were analysed to understand the REE patterns and it is proposed that REE-Th associated minerals could be the potential Th contributors to the sea water and thus reached ferromanganese nodules. Dolomites (unaltered and altered), two types of serpentines and intrusive rock dolerite from the asbestos mines of Cuddapah basin were analysed for major, minor and trace elements. The elemental concentrations are used for distinguishing and characterising these minerals. From our investigations, it was concluded that both dolomite and dolerite contribute elements in the serpentinisation process. Chemical neutron
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)
International Nuclear Information System (INIS)
Aboudheir, Ahmed; Akande, Abayomi; Idem, Raphael
2006-01-01
reactor. The model was based on the coupling of mass, momentum and energy balance equations as well as our new kinetic model developed for this process.The simulation results for crude ethanol conversion were found to be in accordance with the experimental data obtained at various operating conditions. In addition, the predicted variations of the concentration and temperature profiles for our process. In the radial direction indicate that the assumption of plug flow and isothermal behaviour is justified within certain kinetics operating conditions. However, even within these operating conditions, our results have proven that the axial dispersion terms in the balance equations (mass, momentum and energy) cannot be neglected, especially in the hypothetical industrial case presented in this work for the reforming of crude ethanol. In addition, in the experimental study of the application of a porous membrane reactor for the crude ethanol reforming process conducted to compare with that for the packed bed tubular reactor, it was found that the membrane reactor outperformed the packed bed tubular reactor in terms of crude ethanol conversion and hydrogen production. This is due to the function of the membrane reactor to shift the thermodynamic equilibrium in favour of the conversion of crude ethanol to hydrogen according to Le Catelier-Braun's law.(Author)
Study on statistical analysis of nonlinear and nonstationary reactor noises
International Nuclear Information System (INIS)
Hayashi, Koji
1993-03-01
For the purpose of identification of nonlinear mechanism and diagnosis of nuclear reactor systems, analysis methods for nonlinear reactor noise have been studied. By adding newly developed approximate response function to GMDH, a conventional nonlinear identification method, a useful method for nonlinear spectral analysis and identification of nonlinear mechanism has been established. Measurement experiment and analysis were performed on the reactor power oscillation observed in the NSRR installed at the JAERI and the cause of the instability was clarified. Furthermore, the analysis and data recording methods for nonstationary noise have been studied. By improving the time resolution of instantaneous autoregressive spectrum, a method for monitoring and diagnosis of operational status of nuclear reactor has been established. A preprocessing system for recording of nonstationary reactor noise was developed and its usability was demonstrated through a measurement experiment. (author) 139 refs
Accident analysis of heat pipe cooled and AMTEC conversion space reactor system
International Nuclear Information System (INIS)
Yuan, Yuan; Shan, Jianqiang; Zhang, Bin; Gou, Junli; Bo, Zhang; Lu, Tianyu; Ge, Li; Yang, Zijiang
2016-01-01
Highlights: • A transient analysis code TAPIRS for HPS has been developed. • Three typical accidents are analyzed using TAPIRS. • The reactor system has the self-stabilization ability under accident conditions. - Abstract: A space power with high power density, light weight, low cost and high reliability is of crucial importance to future exploration of deep space. Space reactor is an excellent candidate because of its unique characteristics of high specific power, low cost, strong environment adaptability and so on. Among all types of space reactors, heat pipe cooled space reactor, which adopts the passive heat pipe (HP) as core cooling component, is considered as one of the most promising choices and is widely studied all over the world. This paper develops a transient analysis code (TAPIRS) for heat pipe cooled space reactor power system (HPS) based on point reactor kinetics model, lumped parameter core heat transfer model, combined HP model (self-diffusion model, flat-front startup model and network model), energy conversion model of Alkali Metal Thermal-to-Electric Conversion units (AMTEC), and HP radiator model. Three typical accidents, i.e., control drum failure, AMTEC failure and partial loss of the heat transfer area of radiator are then analyzed using TAPIRS. By comparing the simulation results of the models and steady state with those in the references, the rationality of the models and the solution method is validated. The results show the following. (1) After the failure of one set of control drums, the reactor power finally reaches a stable value after two local peaks under the temperature feedback. The fuel temperature rises rapidly, however it is still under safe limit. (2) The fuel temperature is below a safe limit under the AMTEC failure and partial loss of the heat transfer area of radiator. This demonstrates the rationality of the system design and the potential applicability of the TAPIRS code for the future engineering application of
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.
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.
Prism reactor system design and analysis of postulated unscrammed events
International Nuclear Information System (INIS)
Van Tuyle, G.J.; Slovik, G.C.
1991-08-01
Key safety characteristics of the PRISM reactor system include the passive reactor shutdown characteristic and the passive shutdown heat removal system, RVACS. While these characteristics are simple in principle, the physical processes are fairly complex, particularly for the passive reactor shutdown. It has been possible to adapt independent safety analysis codes originally developed for the Clinch River Breeder Reactor review, although some limitations remain. In this paper, the analyses of postulated unscrammed events are discussed, along with limitations in the predictive capabilities and plans to correct the limitations in the near future. 6 refs., 4 figs
PRISM reactor system design and analysis of postulated unscrammed events
International Nuclear Information System (INIS)
Van Tuyle, G.J.; Slovik, G.C.
1991-01-01
Key safety characteristics of the PRISM reactor system include the passive reactor shutdown characteristic and the passive shutdown heat removal system, RVACS. While these characteristics are simple in principle, the physical processes are fairly complex, particularly for the passive reactor shutdown. It has been possible to adapt independent safety analysis codes originally developed for the Clinch River Breeder Reactor review, although some limitations remain. In this paper, the analyses of postulated unscrammed events are discussed, along with limitations in the predictive capabilities and plans to correct the limitations in the near future. (author)
PRISM reactor system design and analysis of postulated unscrammed events
International Nuclear Information System (INIS)
Van Tuyle, G.J.; Slovik, G.C.; Rosztoczy, Z.; Lane, J.
1991-01-01
Key safety characteristics of the PRISM reactor system include the passive reactor shutdown characteristics and the passive shutdown heat removal system, RVACS. While these characteristics are simple in principle, the physical processes are fairly complex, particularly for the passive reactor shutdown. It has been possible to adapt independent safety analysis codes originally developed for the Clinch River Breeder Reactor review, although some limitations remain. In this paper, the analyses of postulated unscrammed events are discussed, along with limitations in the predictive capabilities and plans to correct the limitations in the near future. 6 refs., 4 figs
Computer System Analysis for Decommissioning Management of Nuclear Reactor
International Nuclear Information System (INIS)
Nurokhim; Sumarbagiono
2008-01-01
Nuclear reactor decommissioning is a complex activity that should be planed and implemented carefully. A system based on computer need to be developed to support nuclear reactor decommissioning. Some computer systems have been studied for management of nuclear power reactor. Software system COSMARD and DEXUS that have been developed in Japan and IDMT in Italy used as models for analysis and discussion. Its can be concluded that a computer system for nuclear reactor decommissioning management is quite complex that involved some computer code for radioactive inventory database calculation, calculation module on the stages of decommissioning phase, and spatial data system development for virtual reality. (author)
Stochastic processes analysis in nuclear reactor using ARMA models
International Nuclear Information System (INIS)
Zavaljevski, N.
1990-01-01
The analysis of ARMA model derived from general stochastic state equations of nuclear reactor is given. The dependence of ARMA model parameters on the main physical characteristics of RB nuclear reactor in Vinca is presented. Preliminary identification results are presented, observed discrepancies between theory and experiment are explained and the possibilities of identification improvement are anticipated. (author)
A systems analysis of the ARIES tokamak reactors
International Nuclear Information System (INIS)
Bathke, C.G.
1992-01-01
The multi-institutional ARIES study has completed a series of cost-of-electricity optimized conceptual designs of commercial tokamak fusion reactors that vary the assumed advances in technology and physics. A comparison of these designs indicates the cost benefit of various design options. A parametric systems analysis suggests a possible means to obtain a marginally competitive fusion reactor
Development of fault diagnostic technique using reactor noise analysis
International Nuclear Information System (INIS)
Park, Jin Ho; Kim, J. S.; Oh, I. S.; Ryu, J. S.; Joo, Y. S.; Choi, S.; Yoon, D. B.
1999-04-01
The ultimate goal of this project is to establish the analysis technique to diagnose the integrity of reactor internals using reactor noise. The reactor noise analyses techniques for the PWR and CANDU NPP(Nuclear Power Plants) were established by which the dynamic characteristics of reactor internals and SPND instrumentations could be identified, and the noise database corresponding to each plant(both Korean and foreign one) was constructed and compared. Also the change of dynamic characteristics of the Ulchin 1 and 2 reactor internals were simulated under presumed fault conditions. Additionally portable reactor noise analysis system was developed so that real time noise analysis could directly be able to be performed at plant site. The reactor noise analyses techniques developed and the database obtained from the fault simulation, can be used to establish a knowledge based expert system to diagnose the NPP's abnormal conditions. And the portable reactor noise analysis system may be utilized as a substitute for plant IVMS(Internal Vibration Monitoring System). (author)
An approach to neutronics analysis of candu reactors
International Nuclear Information System (INIS)
Gul, S.; Arshad, M.
1982-12-01
An attempt is made to tackle the problem of neutronics analysis of CANDU reactors. Until now CANDU reactors have been analysed by the methods developed at AECL and CGE using mainly receipe methods. Relying on multigroup transport codes GAM-GATHER in combination with diffusion code CITATION a package of codes is established to use it for survey as well as production purposes. (authors)
Development of fault diagnostic technique using reactor noise analysis
Energy Technology Data Exchange (ETDEWEB)
Park, Jin Ho; Kim, J. S.; Oh, I. S.; Ryu, J. S.; Joo, Y. S.; Choi, S.; Yoon, D. B
1999-04-01
The ultimate goal of this project is to establish the analysis technique to diagnose the integrity of reactor internals using reactor noise. The reactor noise analyses techniques for the PWR and CANDU NPP(Nuclear Power Plants) were established by which the dynamic characteristics of reactor internals and SPND instrumentations could be identified, and the noise database corresponding to each plant(both Korean and foreign one) was constructed and compared. Also the change of dynamic characteristics of the Ulchin 1 and 2 reactor internals were simulated under presumed fault conditions. Additionally portable reactor noise analysis system was developed so that real time noise analysis could directly be able to be performed at plant site. The reactor noise analyses techniques developed and the database obtained from the fault simulation, can be used to establish a knowledge based expert system to diagnose the NPP's abnormal conditions. And the portable reactor noise analysis system may be utilized as a substitute for plant IVMS(Internal Vibration Monitoring System). (author)
Neutronic analysis of the ford nuclear reactor leu core
International Nuclear Information System (INIS)
Raza, S.S.; Hayat, T.
1989-08-01
Neutronic analysis of the ford nuclear reactor low enriched uranium core has been carried out to gain confidence in the com puting methodology being used for Pakistan Research Reactor-1 core conversion calculations. The computed value of the effective multiplication factor (Keff) is found to be in good agreement with that quoted by others. (author). 6 figs
Code system for fast reactor neutronics analysis
International Nuclear Information System (INIS)
Nakagawa, Masayuki; Abe, Junji; Sato, Wakaei.
1983-04-01
A code system for analysis of fast reactor neutronics has been developed for the purpose of handy use and error reduction. The JOINT code produces the input data file to be used in the neutronics calculation code and also prepares the cross section library file with an assigned format. The effective cross sections are saved in the PDS file with an unified format. At the present stage, this code system includes the following codes; SLAROM, ESELEM5, EXPANDA-G for the production of effective cross sections and CITATION-FBR, ANISN-JR, TWOTRAN2, PHENIX, 3DB, MORSE, CIPER and SNPERT. In the course of the development, some utility programs and service programs have been additionaly developed. These are used for access of PDS file, edit of the cross sections and graphic display. Included in this report are a description of input data format of the JOINT and other programs, and of the function of each subroutine and utility programs. The usage of PDS file is also explained. In Appendix A, the input formats are described for the revised version of the CIPER code. (author)
Neutronics analysis of Nigerian Research Reactor-1
International Nuclear Information System (INIS)
Azande, T.S.; Balogun, G.I.
2010-01-01
Feasibility studies for the conversion of the Nigerian Research Reactor-1 (NIRR-1) have been performed using WIMS and CITATION codes (Azande et al, 2009 and Balogun, 2003) at the Centre for Energy Research and Training (CERT), Ahmadu Bello University, Zaria Kaduna State. In this work, the neutronics analysis of NIRR-1 core concerning mass loading of U-235 in the core, shut down margin (SDM), safety reactivity factor (SRF), control rod worth, and control rod critical depth of insertion were investigated at low enrichment. Two fuel types (UAl 4 and UO 2 ) were considered and the uranium densities required for the conversion of NIRR-1 core to low enrichment were computed to be 1201g/cc with 20% enrichment, 1144 g/cc with 19.75% enrichment, 1274 g/cc with 15% enrichment, 1448 g/cc with 10% enrichment for UAl 4 fuel type and 1141g/cc with 20% enrichment, 1144 g/cc with 19.75% enrichment, 1216 g/cc with 15% enrichment, and 1389 g/cc with 10% enrichment for UO 2 fuel type. Signi ficantly, higher uranium densities are required to convert NIRR-1 from HEU to LEU - indicating a drastic review of the NIRR-1 core.
Kinetic model for torrefaction of wood chips in a pilot-scale continuous reactor
DEFF Research Database (Denmark)
Shang, Lei; Ahrenfeldt, Jesper; Holm, Jens Kai
2014-01-01
accordance with the model data. In an additional step a continuous, pilot scale reactor was built to produce torrefied wood chips in large quantities. The "two-step reaction in series" model was applied to predict the mass yield of the torrefaction reaction. Parameters used for the calculation were...... at different torrefaction temperatures, it was possible to predict the HHV of torrefied wood chips from the pilot reactor. The results from this study and the presented modeling approach can be used to predict the product quality from pilot scale torrefaction reactors based on small scale experiments and could...
Occupational analysis for the Angra-1 reactor
International Nuclear Information System (INIS)
Moraes, A.
1991-01-01
Due to several modifications which were imposed to its time schedule during construction, the Angra-1 reactor did not enter to the grid in 1982 as it was initially foreseen. These modifications occurred due to an unforeseen scenario that was verified in steam generators (serie D-3, Westinghouse) of power stations with similar configurations which had been installed in other countries such as Ringhals-3 (Sweden), Almaraz-1 (Spain) and McGuine-1 (USA). So, among the main events that occurred in the Angra-1 reactor, which were of interest from the point of view of radiation protection, it could be pointed out the personnel monitoring, and the occupational exposure measurements at different reactor power, during the reactor fueling and during modification and tests performed at the steam generators and at ducts of the primary coolant circuit. (author)
Linear regression and sensitivity analysis in nuclear reactor design
International Nuclear Information System (INIS)
Kumar, Akansha; Tsvetkov, Pavel V.; McClarren, Ryan G.
2015-01-01
Highlights: • Presented a benchmark for the applicability of linear regression to complex systems. • Applied linear regression to a nuclear reactor power system. • Performed neutronics, thermal–hydraulics, and energy conversion using Brayton’s cycle for the design of a GCFBR. • Performed detailed sensitivity analysis to a set of parameters in a nuclear reactor power system. • Modeled and developed reactor design using MCNP, regression using R, and thermal–hydraulics in Java. - Abstract: The paper presents a general strategy applicable for sensitivity analysis (SA), and uncertainity quantification analysis (UA) of parameters related to a nuclear reactor design. This work also validates the use of linear regression (LR) for predictive analysis in a nuclear reactor design. The analysis helps to determine the parameters on which a LR model can be fit for predictive analysis. For those parameters, a regression surface is created based on trial data and predictions are made using this surface. A general strategy of SA to determine and identify the influential parameters those affect the operation of the reactor is mentioned. Identification of design parameters and validation of linearity assumption for the application of LR of reactor design based on a set of tests is performed. The testing methods used to determine the behavior of the parameters can be used as a general strategy for UA, and SA of nuclear reactor models, and thermal hydraulics calculations. A design of a gas cooled fast breeder reactor (GCFBR), with thermal–hydraulics, and energy transfer has been used for the demonstration of this method. MCNP6 is used to simulate the GCFBR design, and perform the necessary criticality calculations. Java is used to build and run input samples, and to extract data from the output files of MCNP6, and R is used to perform regression analysis and other multivariate variance, and analysis of the collinearity of data
Radioimmunoassay for phencyclidine: application to kinetic analysis in the rat
International Nuclear Information System (INIS)
Ward, D.P.; Trevor, A.J.
1980-01-01
We report the development of a radioimmunoassay for phencyclidine (PCP) that is simple, rapid and sensitive to 0.5 ng/ml. Antibodies were raised in rabbits against the hapten, N-succinyl-3-aminophencyclidine. These antibodies proved to be very specific for PCP and exhibited less than 4% cross reactivity with the drug's two major metabolites. The assay was used for kinetic analysis of PCP in the rat following subcutaneous injection of 5 mg/kg of the drug. Serum and brain tissues were analyzed for PCP and the respective half lives were calculated to be 36 and 29 min for the α phase and 130 and 121 min for the β phase. The accuracy of the method was verified by concomitant assay of a number of kinetic samples by gas chromatography employing a nitrogen-phosphorus detector
Energy Technology Data Exchange (ETDEWEB)
Carbajo, Juan (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin, Madison, WI); Schmidt, Rodney Cannon; Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Ludewig, Hans (Brookhaven National Laboratory, Upton, NY); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache %3CU%2B2013%3E CEA, France)
2011-06-01
This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the
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)
Preliminary conceptual design and analysis on KALIMER reactor structures
International Nuclear Information System (INIS)
Kim, Jong Bum
1996-10-01
The objectives of this study are to perform preliminary conceptual design and structural analyses for KALIMER (Korea Advanced Liquid Metal Reactor) reactor structures to assess the design feasibility and to identify detailed analysis requirements. KALIMER thermal hydraulic system analysis results and neutronic analysis results are not available at present, only-limited preliminary structural analyses have been performed with the assumptions on the thermal loads. The responses of reactor vessel and reactor internal structures were based on the temperature difference of core inlet and outlet and on engineering judgments. Thermal stresses from the assumed temperatures were calculated using ANSYS code through parametric finite element heat transfer and elastic stress analyses. While, based on the results of preliminary conceptual design and structural analyses, the ASME Code limits for the reactor structures were satisfied for the pressure boundary, the needs for inelastic analyses were indicated for evaluation of design adequacy of the support barrel and the thermal liner. To reduce thermal striping effects in the bottom are of UIS due to up-flowing sodium form reactor core, installation of Inconel-718 liner to the bottom area was proposed, and to mitigate thermal shock loads, additional stainless steel liner was also suggested. The design feasibilities of these were validated through simplified preliminary analyses. In conceptual design phase, the implementation of these results will be made for the design of the reactor structures and the reactor internal structures in conjunction with the thermal hydraulic, neutronic, and seismic analyses results. 4 tabs., 24 figs., 4 refs. (Author)
A sustainability analysis of the Brazilian multipurpose reactor project
International Nuclear Information System (INIS)
Obadia, I.J.; Perrotta, J.A.
2010-01-01
The project of a new research reactor in Brazil for radioisotope production, support of the nuclear energy program and scientific research has received a positive sign of the government and is starting to be developed by the Brazilian Commission of Nuclear Energy. International Atomic Energy Agency points out that the implementation of a new research reactor is a major undertaking for a country, requiring an analysis to identify to which extent the conditions of the national nuclear program are proper and adequate to lead to a sustainable research reactor life cycle. This paper introduces the Brazilian Multipurpose Reactor Project (RMB) and describes the sustainability analysis performed, which has shown that the national nuclear infrastructure presents a very favourable condition to the implementation of the RMB project as well as to provide a sustainable life cycle for this new research reactor. (author)
Radioactivity analysis of KAMINI reactor coolant from regulatory perspectives
International Nuclear Information System (INIS)
Srinivasan, T.K.; Sulthan, Bajeer; Sarangapani, R.; Jose, M.T.; Venkatraman, B.; Thilagam, L.
2016-01-01
KAMINI (a 30kWt) research reactor is operated for neutron radiography of fuel subassemblies and pyro devices and activation analysis of various samples. The reactor is fueled by 233 U and DM water is used as the coolant. During reactor operation, fission product noble gasses (FPNGs) such as 85m Kr, 87 Kr, 88 Kr, 135 Xe, 135m Xe and 138 Xe are detected in the coolant water. In order to detect clad failure, the water is sampled during reactor operation at regular intervals as per the technical specifications. In the present work, analysis of measured activities in coolant samples collected during reactor operation at 25 kWt are presented and compared with computed values obtained using ORIGEN (Isotope Generation) code
International Nuclear Information System (INIS)
Perez M, C.
2004-01-01
The development of a software is presented that simulates the punctual kinetics of a nuclear reactor of investigation model TRIGA Mark III, generating the answers of the reactor low different algorithms of control of power. The user requires a graphic interface that allows him easily interacting with the simulator. To achieve the proposed objective, first the system was modeled in open loop, not using a mathematical model of the consistent reactor in a system of linear ordinary differential equations. For their solution in real time the numeric method of Runge-Kutta-Fehlberg was used. As second phase, it was modeled to the system in closed loop, using for it an algorithm of control of the power based on fuzzy logic. This software has as purpose to help the investigator in the control area who will be able to prove different algorithms for the control of the power of the reactor. This is achieved using the code source in language C, C++, Visual Basic, with which a file is generated. DLL and it is inserted in the simulator. Then they will be able to visualize the results as if their controller had installed in the reactor, analyzing the behavior of all his variables that will be stored in files, for his later study. The easiness of proving these control algorithms in the reactor without necessity to make it physically has important consequences as the saving in the expense of fuel, the not generation of radioactive waste and the most important thing, one doesn't run any risk. The simulator can be used how many times it is necessary until the total purification of the algorithm. This program is the base for following investigation processes, enlarging the capacities and options of the same one. The program fulfills the time of execution satisfactorily, assisting to the necessity of visualizing the behavior in real time of the reactor, and it responds from an effective way to the petitions of changes of power on the part of the user. (Author)
Progress of the DUPIC fuel compatibility analysis (I) - reactor physics
Energy Technology Data Exchange (ETDEWEB)
Choi, Hang Bok; Jeong, Chang Joon; Roh, Gyu Hong; Rhee, Bo Wook; Park, Jee Won
2003-12-01
Since 1992, the direct use of spent pressurized water reactor fuel in CANada Deuterium Uranium (CANDU) reactors (DUPIC) has been studied as an alternative to the once-through fuel cycle. The DUPIC fuel cycle study is focused on the technical feasibility analysis, the fabrication of DUPIC fuels for irradiation tests and the demonstration of the DUPIC fuel performance. The feasibility analysis was conducted for the compatibility of the DUPIC fuel with existing CANDU-6 reactors from the viewpoints of reactor physics, reactor safety, fuel cycle economics, etc. This study has summarized the intermediate results of the DUPIC fuel compatibility analysis, which includes the CANDU reactor physics design requirements, DUPIC fuel core physics design method, performance of the DUPIC fuel core, regional overpower trip setpoint, and the CANDU primary shielding. The physics analysis showed that the CANDU-6 reactor can accommodate the DUPIC fuel without deteriorating the physics design requirements by adjusting the fuel management scheme if the fissile content of the DUPIC fuel is tightly controlled.
International Nuclear Information System (INIS)
Rastogi, B.P.
1989-01-01
This report discusses various reactor physics codes developed for neutronic design, fuel-management, reactor operation and safety analysis of PHWRs. These code packages have been utilized for nuclear design of 500 MWe and new 235 MWe PHWRs. (author)
Application of autoregressive moving average model in reactor noise analysis
International Nuclear Information System (INIS)
Tran Dinh Tri
1993-01-01
The application of an autoregressive (AR) model to estimating noise measurements has achieved many successes in reactor noise analysis in the last ten years. The physical processes that take place in the nuclear reactor, however, are described by an autoregressive moving average (ARMA) model rather than by an AR model. Consequently more correct results could be obtained by applying the ARMA model instead of the AR model to reactor noise analysis. In this paper the system of the generalised Yule-Walker equations is derived from the equation of an ARMA model, then a method for its solution is given. Numerical results show the applications of the method proposed. (author)
Establishment of computer code system for nuclear reactor design - analysis
International Nuclear Information System (INIS)
Subki, I.R.; Santoso, B.; Syaukat, A.; Lee, S.M.
1996-01-01
Establishment of computer code system for nuclear reactor design analysis is given in this paper. This establishment is an effort to provide the capability in running various codes from nuclear data to reactor design and promote the capability for nuclear reactor design analysis particularly from neutronics and safety points. This establishment is also an effort to enhance the coordination of nuclear codes application and development existing in various research centre in Indonesia. Very prospective results have been obtained with the help of IAEA technical assistance. (author). 6 refs, 1 fig., 1 tab
Applications in nuclear data and reactor physics
International Nuclear Information System (INIS)
Cullen, D.E.; Muranaka, R.; Schmidt, J.
1986-01-01
This book presents the papers given at a conference on reactor kinetics and nuclear data collections. Topics considered at the conference included nuclear data processing, PWR core design calculations, reactor neutron dosimetry, in-core fuel management, reactor safety analysis, transients, two-phase flow, fuel cycles of research reactors, slightly enriched uranium, highly enriched uranium, reactor start-up, computer codes, and the transport of spent fuel elements
International Nuclear Information System (INIS)
Battaglia, Francine
2008-01-01
The research project was related to the Advanced Fuel Cycle Initiative and was in direct alignment with advancing knowledge in the area of Nuclear Fuel Development related to the use of TRISO fuels for high-temperature reactors. The importance of properly coating nuclear fuel pellets received a renewed interest for the safe production of nuclear power to help meet the energy requirements of the United States. High-temperature gas-cooled nuclear reactors use fuel in the form of coated uranium particles, and it is the coating process that was of importance to this project. The coating process requires four coating layers to retain radioactive fission products from escaping into the environment. The first layer consists of porous carbon and serves as a buffer layer to attenuate the fission and accommodate the fuel kernel swelling. The second (inner) layer is of pyrocarbon and provides protection from fission products and supports the third layer, which is silicon carbide. The final (outer) layer is also pyrocarbon and provides a bonding surface and protective barrier for the entire pellet. The coating procedures for the silicon carbide and the outer pyrocarbon layers require knowledge of the detailed kinetics of the reaction processes in the gas phase and at the surfaces where the particles interact with the reactor walls. The intent of this project was to acquire detailed information on the reaction kinetics for the chemical vapor deposition (CVD) of carbon and silicon carbine on uranium fuel pellets, including the location of transition state structures, evaluation of the associated activation energies, and the use of these activation energies in the prediction of reaction rate constants. After the detailed reaction kinetics were determined, the reactions were implemented and tested in a computational fluid dynamics model, MFIX. The intention was to find a reduced mechanism set to reduce the computational time for a simulation, while still providing accurate results
Sponza, Delia Teresa; Çelebi, Hakan
2012-01-01
An anaerobic multichamber bed reactor (AMCBR) was effective in removing both molasses-chemical oxygen demand (COD), and the antibiotic oxytetracycline (OTC). The maximum COD and OTC removals were 99% in sequential AMCBR/completely stirred tank reactor (CSTR) at an OTC concentration of 300 mg L(-1). 51%, 29% and 9% of the total volatile fatty acid (TVFA) was composed of acetic, propionic acid and butyric acids, respectively. The OTC loading rates at between 22.22 and 133.33 g OTC m(-3) d(-1) improved the hydrolysis of molasses-COD (k), the maximum specific utilization of molasses-COD (k(mh)) and the maximum specific utilization rate of TVFA (k(TVFA)). The direct effect of high OTC loadings (155.56 and -177.78 g OTC m(-3) d(-1)) on acidogens and methanogens were evaluated with Haldane inhibition kinetic. A significant decrease of the Haldane inhibition constant was indicative of increases in toxicity at increasing loading rates. Copyright © 2011 Elsevier Ltd. All rights reserved.
Seismological analysis of group pile foundation for reactor
International Nuclear Information System (INIS)
Wang Demin.
1984-01-01
In the seismic analysis for reactor foundation of nuclear power plant, the local raise of base mat is of great significance. Base on the study of static and dynamic stability as well as soil-structure interaction of group piles on stratified soil, this paper presents a method of seismic analysis for group piles of reactor foundation at abroad, and a case history is enclosed. (Author)
Kinetics parameter measurements on RSG-GAS, a low-enriched fuel reactor
International Nuclear Information System (INIS)
Jujuratisbela, U; Arbie, B; Pinem, S.; Tukiran; Suparlina, L.; Singh, O.P.
1995-01-01
Kinetics parameter measurements, such as reactivity worths of control rods and fuel elements, beam tube void reactivity, power reactivity coefficient and xenon poisoning reactivity have been performed on different cores of Reaktor Serba Guna G.A. Siwabessy (RSG-GAS). In parallel, a programme was also initiated to measure the other kinetics parameters like effective delayed neutron life time, prompt neutron decay constant, validation of period reactivity relationship and zero power frequency response function. The paper provides the results of these measurements. (author)
Thermal-hydraulic analysis of nuclear reactors
Zohuri, Bahman
2015-01-01
This text covers the fundamentals of thermodynamics required to understand electrical power generation systems and the application of these principles to nuclear reactor power plant systems. It is not a traditional general thermodynamics text, per se, but a practical thermodynamics volume intended to explain the fundamentals and apply them to the challenges facing actual nuclear power plants systems, where thermal hydraulics comes to play. Written in a lucid, straight-forward style while retaining scientific rigor, the content is accessible to upper division undergraduate students and aimed at practicing engineers in nuclear power facilities and engineering scientists and technicians in industry, academic research groups, and national laboratories. The book is also a valuable resource for students and faculty in various engineering programs concerned with nuclear reactors. This book also: Provides extensive coverage of thermal hydraulics with thermodynamics in nuclear reactors, beginning with fundamental ...
Study on effects of development of reactor constant in fast reactor analysis
International Nuclear Information System (INIS)
Chiba, Gou
2002-12-01
Evaluation was carried out about an effect of development of the new generation reactor constant system that substitutes for the JFS library in fast reactor analysis. Analyzed cores were ZPPR in JUPITER critical experiment and several power reactor cores that were designed in the feasibility study. In the JUPITER analysis, large effects, over 10%, were observed in sodium void reactivity and sample Doppler reactivity. The former resulted from several factors, while the latter was due to an accurate of a resonance interaction effect between Doppler sample and core fuel. In the previous study, the effect had been evaluated in power reactor cores. The effect included an effect of corrosion of weighting spectrum because JFS-3-J3.2, which had been made with the incorrect weighting spectrum, was used in the evaluation. In the present study, JFS-3-J3.2R, which had been made with the correct weighting spectrum, was used. It was confirmed that the effect of development of reactor constant in power reactor was not as large as that in critical assembly. (author)
Kinetic evaluation of an anaerobic fluidised-bed reactor treating slaughterhouse wastewater
Energy Technology Data Exchange (ETDEWEB)
Borja, R. [Consejo Superior de Investigaciones Cientificas, Seville (Spain). Inst. de la Grasa; Banks, C.J.; Zhengjian Wang [Manchester Univ. (United Kingdom). Inst. of Science and Technology
1995-09-01
An anaerobic fluidised-bed reactor for purification of slaughterhouse wastewater was modelled as a continuous-flow, completely-mixed homogeneous microbial system, with the feed COD as the limiting-substrate concentration. The average microbial residence time in the reactor was defined in terms of conventional sludge-retention-time. The experimental data obtained indicated that the Michaelis-Menten expression was applicable to a description of substrate utilisation (i.e. COD removal) in the anaerobic fluidised-bed system. The maximum substrate utilisation rate, k, and the Michaelis constant, K{sub s}, were determined to be 1.2/day and 0.039 g/l. The observed biomass yield in the reactor decreased with increasing sludge-retention-time. The specific methane production rate observed was a linear function of the specific substrate-utilisation rate. (Author)
DEFF Research Database (Denmark)
Hashemi, Hamid; Christensen, Jakob Munkholt; Glarborg, Peter
A series of experimental and numerical investigations into hydrogen oxidation at high pressures and intermediate temperatures has been conducted. The experiments were carried out in a high pressure laminar flow reactor at 50 bar pressure and a temperature range of 600–900 K. The equivalence ratio......, the mechanism is used to simulate published data on ignition delay time and laminar burning velocity of hydrogen. The flow reactor results show that at reducing, stoichiometric, and oxidizing conditions, conversion starts at temperatures of 750–775 K, 800–825 K, and 800–825 K, respectively. In oxygen atmosphere......, ignition occurs at the temperature of 775–800 K. In general, the present model provides a good agreement with the measurements in the flow reactor and with recent data on laminar burning velocity and ignition delay time....
Analysis of an accelerator-driven subcritical light water reactor
International Nuclear Information System (INIS)
Kruijf, W.J.M. de; Wakker, P.H.; Wetering, T.F.H. van de; Verkooijen, A.H.M.
1997-01-01
An analysis of the basic characteristics of an accelerator-driven light water reactor has been made. The waste in the nuclear fuel cycle is considerably less than in the light water reactor open fuel cycle. This is mainly caused by the use of equilibrium nuclear fuel in the reactor. The accelerator enables the use of a fuel composition with infinite multiplication factor k ∞ < 1. The main problem of the use of this type of fuel is the strongly peaked flux distribution in the reactor core. A simple analytical model shows that a large core is needed with a high peak power factor in order to generate net electric energy. The fuel in the outer regions of the reactor core is used very poorly. 7 refs., 4 figs., 1 tab
Review of the treat upgrade reactor scram system reliability analysis
International Nuclear Information System (INIS)
Montague, D.F.; Fussell, J.B.; Krois, P.A.; Morelock, T.C.; Knee, H.E.; Manning, J.J.; Haas, P.M.; West, K.W.
1984-10-01
In order to resolve some key LMFBR safety issues, ANL personnel are modifying the TREAT reactor to handle much larger experiments. As a result of these modifications, the upgraded Treat reactor will not always operate in a self-limited mode. During certain experiments in the upgraded TREAT reactor, it is possible that the fuel could be damaged by overheating if, once the computer systems fail, the reactor scram system (RSS) fails on demand. To help ensure that the upgraded TREAT reactor is shut down when required, ANL personnel have designed a triply redundant RSS for the facility. The RSS is designed to meet three reliability goals: (1) a loss of capability failure probability of 10 -9 /demand (independent failures only); (2) an inadvertent shutdown probability of 10 -3 /experiment; and (3) protection agaist any known potential common cause failures. According to ANL's reliability analysis of the RSS, this system substantially meets these goals
International Nuclear Information System (INIS)
Schikorr, W.M.
2001-01-01
The neutron kinetic and the reactor dynamic behavior of Accelerator Driven Systems (ADS) is significantly different from those of conventional power reactor systems currently in use for the production of power. It is the objective of this study to examine and to demonstrate the intrinsic differences of the kinetic and dynamic behavior of accelerator driven systems to typical plant transient initiators in comparison to the known, kinetic and dynamic behavior of critical thermal and fast reactor systems. It will be shown that in sub-critical assemblies, changes in reactivity or in the external neutron source strength lead to an asymptotic power level essentially described by the instantaneous power change (i.e. prompt jump). Shutdown of ADS operating at high levels of sub-criticality, (i.e. k eff ∼0.99), without the support of reactivity control systems (such as control or safety rods), may be problematic in case the ability of cooling of the core should be impaired (i.e. loss of coolant flow). In addition, the dynamic behavior of sub-critical systems to typical plant transients such as protected or unprotected loss of flow (LOF) or heat sink (LOH) transients are not necessarily substantially different from the plant dynamic behavior of critical systems if the reactivity feedback coefficients of the ADS design are unfavorable. As expected, the state of sub-criticality and the temperature feedback coefficients, such as Doppler and coolant temperature coefficient, play dominant roles in determining the course and direction of plant transients. Should the combination of these safety coefficients be very unfavorable, not much additional margin in safety may be gained by making a critical system only sub-critical (i.e. k eff ∼0.95). A careful optimization procedure between the selected operating level of sub-criticality, the safety reactivity coefficients and the possible need for additional reactivity control systems seems, therefore, advisable during the early
International Nuclear Information System (INIS)
Pencer, J.; Choy Wong, F.; Bromley, B.P.; Atfield, J.; Zeller, M.
2010-01-01
This paper summarizes comparisons between MCNP5 and WIMS-AECL / DRAGON / RFSP calculations and experimental results obtained from the Zero Energy Deuterium (ZED-2) critical facility at AECL Chalk River Laboratories. MCNP5 and WIMS-AECL / DRAGON / RFSP were used to calculate reactivity worths for two reactivity devices, a mechanical zone controller (MZC) and shut-off rod (SOR) in a lattice similar to that of the ACR-1000 R . WIMS-AECL / DRAGON / RFSP was also used to obtain kinetics parameters for a transient based on a rod drop of a ZED-2 standby absorber rod (SAR). ZED-2 experiments were performed using 43-element ACR Low Enriched Uranium (ACR-LEU) fuel bundles with H 2 O- or air-cooled fuel bundles arranged in a 24-cm pitch square lattice. Calculations with MCNP5 gave biases in device worths that were within 0.2 mk of measured values, while WIMS-AECL / DRAGON / RFSP gave values that were within 0.3 mk of measured values. Transient analyses using the CERBERUS module within RFSP yielded a total delayed neutron fraction (β) that was within 4% of the value derived by point kinetics analysis of experimental data. The corresponding delayed photo-neutron fraction (β photo-neutron ) from CERBERUS was within 5% of that derived by point kinetics. This study has helped quantify the agreement between calculation and measurement for codes that are used in the safety analysis of the ACR-1000 reactor. Results demonstrate good agreement in code predictions. (authors)
Issues affecting advanced passive light-water reactor safety analysis
International Nuclear Information System (INIS)
Beelman, R.J.; Fletcher, C.D.; Modro, S.M.
1992-01-01
Next generation commercial reactor designs emphasize enhanced safety through improved safety system reliability and performance by means of system simplification and reliance on immutable natural forces for system operation. Simulating the performance of these safety systems will be central to analytical safety evaluation of advanced passive reactor designs. Yet the characteristically small driving forces of these safety systems pose challenging computational problems to current thermal-hydraulic systems analysis codes. Additionally, the safety systems generally interact closely with one another, requiring accurate, integrated simulation of the nuclear steam supply system, engineered safeguards and containment. Furthermore, numerical safety analysis of these advanced passive reactor designs wig necessitate simulation of long-duration, slowly-developing transients compared with current reactor designs. The composite effects of small computational inaccuracies on induced system interactions and perturbations over long periods may well lead to predicted results which are significantly different than would otherwise be expected or might actually occur. Comparisons between the engineered safety features of competing US advanced light water reactor designs and analogous present day reactor designs are examined relative to the adequacy of existing thermal-hydraulic safety codes in predicting the mechanisms of passive safety. Areas where existing codes might require modification, extension or assessment relative to passive safety designs are identified. Conclusions concerning the applicability of these codes to advanced passive light water reactor safety analysis are presented
Discussion on safety analysis approach for sodium fast reactors
International Nuclear Information System (INIS)
Hong, Soon Joon; Choo, Yeon Joon; Suh, Nam Duk; Shin, Ahn Dong; Bae, Moo Hoon
2012-01-01
Utilization of nuclear energy is increasingly necessary not only because of the increasing energy consumption but also because of the controls on greenhouse emissions against global warming. To keep step with such demands, advanced reactors are now world widely under development with the aims of highly economical advances, and enhanced safety. Recently, further elaborating is encouraged on the research and development program for Generation IV (GEN IV) reactors, and in collaboration with other interested countries through the Generation IV International Forum (GIF). Sodium cooled Fast Reactor (SFR) is a strong contender amongst the GEN IV reactor concepts. Korea also takes part in that program and plans to construct demonstration reactor of SFR. SFR is under the development for a candidate of small modular reactors, for example, PRISM (Power Reactor Innovative Small Module). Understanding of safety analysis approach has also advanced by the demand of increasing comprehensive safety requirement. Reviewing the past development of the licensing and safety basis in the advanced reactors, such approaches seemed primarily not so satisfactory because the reference framework of licensing and safety analysis approach in the advanced reactors was always the one in water reactors. And, the framework is very plant specific one and thereby the advanced reactors and their frameworks don't look like a well assorted couple. Recently as a result of considerable advances in probabilistic safety assessment (PSA), risk informed approaches are increasingly applied together with some of the deterministic approaches like as the ones in water reactors. Technology neutral framework (TNF) can be said to be the utmost works of such risk informed approaches, even though an intensive assessment of the applicability has not been sufficiently accomplished. This study discusses the viable safety analysis approaches for the urgent application to the construction of pool type SFR. As discussed in
Nonlinear dynamic analysis of nuclear reactor primary coolant systems
International Nuclear Information System (INIS)
Saffell, B.F. Jr.; Macek, R.W.; Thompson, T.R.; Lippert, R.F.
1979-01-01
The ADINA computer code is utilized to perform mechanical response analysis of pressurized reactor primary coolant systems subjected to postulated loss-of-coolant accident (LOCA) loadings. Specifically, three plant analyses are performed utilizing the geometric and material nonlinear analysis capabilities of ADINA. Each reactor system finite element model represents the reactor vessel and internals, piping, major components, and component supports in a single coupled model. Material and geometric nonlinear capabilities of the beam and truss elements are employed in the formulation of each finite element model. Loadings applied to each plant for LOCA dynamic analysis include steady-state pressure, dead weight, strain energy release, transient piping hydraulic forces, and reactor vessel cavity pressurization. Representative results are presented with some suggestions for consideration in future ADINA code development
A calculational study on neutron kinetics and thermodynamics of a gaseous core fission reactor
International Nuclear Information System (INIS)
Kuijper, J.C.
1992-06-01
A numerical and analytical study of the static and dynamic properties of a GCFR with oscillating fuel gas (uranium and carbon fluorides) is presented. Neutron kinetics parts of combined GCFR models are introduced. Thermodynamic properties of the GCFR and of the fuel gas are treated. (HP)
Solid KHT tumor dispersal for flow cytometric cell kinetic analysis
International Nuclear Information System (INIS)
Pallavicini, M.G.; Folstad, L.J.; Dunbar, C.
1981-01-01
A bacterial neutral protease was used to disperse KHT solid tumors into single cell suspensions suitable for routine cell kinetic analysis by flow cytometry and for clonogenic cell survival. Neutral protease disaggregation under conditions which would be suitable for routine tumor dispersal was compared with a trypsin/DNase procedure. Cell yield, clonogenic cell survival, DNA distributions of untreated and drug-perturbed tumors, rates of radioactive precursor incorporation during the cell cycle, and preferential cell cycle phase-specific cell loss were investigated. Tumors dispersed with neutral protease yielded approximately four times more cells than those dispersed with trypsin/DNase and approximately a 1.5-fold higher plating efficiency in a semisolid agar system. Quantitative analysis of DNA distributions obtained from untreated and cytosine-arabinoside-perturbed tumors produced similar results with both dispersal procedures. The rates of incorporation of tritiated thymidine during the cell cycle were also similar with neutral protease and trypsin/DNase dispersal. Preferential phase-specific cell loss was not obseved with either technique. We find that neutral protease provides good single cell suspensions of the KHT tumor for cell survival measurements and for cell kinetic analysis of drug-induced perturbations by flow cytometry. In addition, the high cell yields facilitate electronic cell sorting where large numbers of cells are often required
Heat transfer capability analysis of heat pipe for space reactor
International Nuclear Information System (INIS)
Li Huaqi; Jiang Xinbiao; Chen Lixin; Yang Ning; Hu Pan; Ma Tengyue; Zhang Liang
2015-01-01
To insure the safety of space reactor power system with no single point failures, the reactor heat pipes must work below its heat transfer limits, thus when some pipes fail, the reactor could still be adequately cooled by neighbor heat pipes. Methods to analyze the reactor heat pipe's heat transfer limits were presented, and that for the prevailing capillary limit analysis was improved. The calculation was made on the lithium heat pipe in core of heat pipes segmented thermoelectric module converter (HP-STMC) space reactor power system (SRPS), potassium heat pipe as radiator of HP-STMC SRPS, and sodium heat pipe in core of scalable AMTEC integrated reactor space power system (SAIRS). It is shown that the prevailing capillary limits of the reactor lithium heat pipe and sodium heat pipe is 25.21 kW and 14.69 kW, providing a design margin >19.4% and >23.6%, respectively. The sonic limit of the reactor radiator potassium heat pipe is 7.88 kW, providing a design margin >43.2%. As the result of calculation, it is concluded that the main heat transfer limit of HP-STMC SRPS lithium heat pipe and SARIS sodium heat pipe is prevailing capillary limit, but the sonic limit for HP-STMC SRPS radiator potassium heat pipe. (authors)
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)
Ethanol steam reforming kinetics of a Pd-Ag membrane reactor
Tosti, S.; Basile, A.; Borelli, R.; Borgognoni, F.; Castelli, S.; Fabbricino, M.; Gallucci, F.; Licusati, C.
2009-01-01
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
International Nuclear Information System (INIS)
Moriyama, Kiyofumi; Maruyama, Yu; Nakamura, Hideo
2011-03-01
An iodine chemistry simulation tool, Kiche, was developed for analyses of chemical kinetics relevant to iodine volatilization in the containment vessel of light water reactors (LWRs) during a severe accident. It consists of a Fortran code to solve chemical kinetics models, reaction databases written in plain text format, and peripheral tools to convert the reaction databases into Fortran codes to solve corresponding ordinary differential equation sets. Potential advantages of Kiche are the text format reaction database separated from the code that provides flexibility of the chemistry model, and, being a Fortran code which is relatively easily coupled with other Fortran codes such as severe accident analysis codes. This document describes the model, solution method, code structure, and examples of application of Kiche for simulation of experiments. The calculation results by the present model agreed well with the experimental data and it indicates the model properly includes the most important processes in the volatilization of iodine from irradiated iodide solutions with or without organic impurities. The appendixes give practical information for the usage of Kiche. (author)
RAP-2A Computer code for transients analysis in fast reactors
International Nuclear Information System (INIS)
Iftode, I.; Popescu, C.; Turcu, I.; Biro, L.
1975-10-01
The RAP-2A computer code is designed for analyzing thermohydraulic transients and/or steady state problems for large LMFBR cores. Physical and mathematical models, main input-output data, the flow chart of the code and a sample problem are given. RAP-2A calculates the power and the thermoydraulic transients initiated by a flow or reactivity changes, from a normal operating state of the reactor up to core disassembly. In this analysis a representative fuel pin is considered: a one-group space-independent (point) kinetics model to describe the neutron kinetics and a one-dimensional model describing the heat transfer (radial in the fuel and axial in the coolant) are used. Mechanical deformations due to temperature gradient, pressure losses, fuel melting, etc., are also calculated. The code is written in FORTRAN-4 language and is running on a IBM-370/135 computer
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
International Nuclear Information System (INIS)
Behringer, K.
1991-02-01
In a recent paper by Behringer et al. (1990), the Wiener-Hermite Functional (WHF) method has been applied to point reactor kinetics excited by Gaussian random reactivity noise under stationary conditions, in order to calculate the neutron steady-state value and the neutron power spectral density (PSD) in a second-order (WHF-2) approximation. For simplicity, delayed neutrons and any feedback effects have been disregarded. The present study is a straightforward continuation of the previous one, treating the problem more generally by including any number of delayed neutron groups. For the case of white reactivity noise, the accuracy of the approach is determined by comparison with the exact solution available from the Fokker-Planck method. In the numerical comparisons, the first-oder (WHF-1) approximation of the PSD is also considered. (author) 4 figs., 10 refs
International Nuclear Information System (INIS)
Behringer, K.; Pineyro, J.; Mennig, J.
1990-06-01
The Wiener-Hermite functional (WHF) method has been applied to the point reactor kinetic equation excited by Gaussian random reactivity noise under stationary conditions. Delayed neutrons and any feedback effects are disregarded. The neutron steady-state value and the power spectral density (PSD) of the neutron flux have been calculated in a second order (WHF-2) approximation. Two cases are considered: in the first case, the noise source is low-pass white noise. In both cases the WHF-2 approximation of the neutron PSDs leads to relatively simple analytical expressions. The accuracy of the approach is determined by comparison with exact solutions of the problem. The investigations show that the WHF method is a powerful approximative tool for studying the nonlinear effects in the stochastic differential equation. (author) 5 figs., 29 refs
Preliminary analysis of basic reactor physics of the Dual Fluid Reactor - 15270
International Nuclear Information System (INIS)
Wang, X.; Macian-Juan, R.; Seidl, M.
2015-01-01
The Dual Fluid Reactor (DFR) is a novel fast nuclear reactor concept invented by the IFK based on the Generation IV Molten Salt Reactor and the Liquid Metal Cooled Reactor. The DFR uses a chloride based molten fuel salt in order to harden the neutron spectrum. The molten fuel salt is cooled with a separated liquid lead loop, which in principle allows for higher power densities and better breeding performance. The DFR does not combine heat removal and breeding into a single circuit but separates the two functions into two independent circuits. Since there are attractive features mentioned in this design, the main task of this paper is to verify the model of the whole reactor based on this concept. For this purpose several calculations are presented, including steady state calculations, sensitivity calculations with regard to the nuclide cross sections, the temperature and geometry coefficient of k eff as well as the burnup calculation. The Monte Carlo calculation codes MCNP, SERPENT and SCALE are used for the analysis. As expected the study shows a significant negative reactivity feedback with temperature in the overall fission zone. For the coupled coolant and reflector design the temperature feedback is rather small for practical purposes such as reactor control during normal operation. In the view of these results the DFR in principle can be self-regulated totally by the temperature change of its own fuel salt and consequently can rely on fully passive safety systems for accident management
SACI - O: A code for the analysis of transients in a pressurized water reactor core
International Nuclear Information System (INIS)
Resende Lobo, A.A. de; Soares, P.A.
1979-03-01
The SACI-O digital computer code consists basically of a pressurized water reactor core model. It is useful in the analysis of fast reactivity transients shorter than the loop transit time. The program can also be used for evaluating the core behaviour, during other transients, when the inlet coolant conditions are known. SACI-O uses point model neutron kinetics taking into account moderator and fuel reactivity effects, and fission products decay. The neutronic and thermal-hydraulic equations are solved for an average fuel pin described by a single axial node. To perform a more detailed calculation, the modeling of another cooling channel, which can be divided into axial segments, is included in the program. The reactor trip system is also partially simulated. (Author) [pt
Numerical analysis of magnetoelastic coupled buckling of fusion reactor components
International Nuclear Information System (INIS)
Demachi, K.; Yoshida, Y.; Miya, K.
1994-01-01
For a tokamak fusion reactor, it is one of the most important subjects to establish the structural design in which its components can stand for strong magnetic force induced by plasma disruption. A number of magnetostructural analysis of the fusion reactor components were done recently. However, in these researches the structural behavior was calculated based on the small deformation theory where the nonlinearity was neglected. But it is known that some kinds of structures easily exceed the geometrical nonlinearity. In this paper, the deflection and the magnetoelastic buckling load of fusion reactor components during plasma disruption were calculated
Electromagnetic analysis for fusion reactors: status and needs
International Nuclear Information System (INIS)
Turner, L.R.
1983-01-01
Electromagnetic effects have far-reaching implications for the design, operation, and maintenance of future fusion reactors. Two-dimensional (2-D) eddy current computer codes are available, but are of limited value in analyzing reactors. Three-dimensional (3-D) codes are needed, but are only beginning to be developed. Both 2-D and 3-D codes need verification against experimental data, such as that provided by the upcoming FELIX experiments. Coupling between eddy currents and deflections has application in fusion reactor design and is being studied both by analysis and experiment
Analysis of short-term reactor cavity transient
International Nuclear Information System (INIS)
Cheng, T.C.; Fischer, S.R.
1981-01-01
Following the transient of a hypothetical loss-of-coolant accident (LOCA) in a nuclear reactor, peak pressures are reached within the first 0.03 s at different locations inside the reactor cavity. Due to the complicated multidimensional nature of the reactor cavity, the short-term analysis of the LOCA transient cannot be performed by using traditional containment codes, such as CONTEMPT. The advanced containment code, BEACON/MOD3, developed at the Idaho National Engineering Laboratory (INEL), can be adapted for such analysis. This code provides Eulerian, one and two-dimensional, nonhomogeneous, nonequilibrium flow modeling as well as lumped parameter, homogeneous, equilibrium flow modeling for the solution of two-component, two-phase flow problems. The purpose of this paper is to demonstrate the capability of the BEACON code to analyze complex containment geometry such as a reactor cavity
International Nuclear Information System (INIS)
Tokuhara, K.; Nakata, T.; Murata, I.; Yamashita, K.; Shindo, R.
1991-01-01
This report describes the calculational methods which were employed to determine the temperature coefficients and the kinetic parameters for the safety analysis in the HTTR (High Temperature Engineering Test Reactor). The temperature coefficients (doppler, moderator temperature) and the kinetic parameters (prompt neutron life time; l, effective delayed neutron fraction; β eff) are important for the point model core dynamic analysis and should be evaluated properly. The temperature coefficients were calculated by the whole core model. Doppler coefficient was evaluated under the conditions of all control rods withdrawn and the uniform change of fuel temperature. The minimum and the maximum value of the evaluated doppler coefficients in a burnup cycle are -4.6x10 -5 and -1.5x10 -5 ΔK/K/deg. C respectively. The moderator temperature coefficient was evaluated under the conditions of all control rods withdrawn and the uniform change of moderator temperature. The minimum and the maximum value of the evaluated moderator temperature coefficients in a burnup cycle are -17.1x10 -5 and 0.99x10 -5 ΔK/K/deg. C respectively. In spite of positive moderator temperature coefficient, the power coefficient is always negative. Therefore the HTTR possesses inherent power-suppressing feed back characteristic in all operating condition. We surveyed the effects of the Xe existence, the control rods existence, the fuel temperature and the region in which the temperature was changed on the moderator temperature coefficients. The kinetic parameters were calculated by the perturbation method with the whole core model. The minimum and the maximum value of the evaluated effective delayed neutron fraction (β eff) are 0.0047 and 0.0065 respectively. These of the evaluated prompt neutron life time (l) are 0.67 and 0.78 ms respectively. We have surveyed the effects of the fuel depletion and the core power level on these parameters, and considered these effects on the kinetic parameters. From
International Nuclear Information System (INIS)
Ren-Tai, Chiang
2003-01-01
An ω-mode first-order perturbation theory is developed for analyzing the time- and space-dependent neutron behavior in Accelerator-Driven Subcritical Systems (ADSS). The generalized point-kinetics equations are systematically derived using the ω-mode first-order perturbation theory and Fredholm Alternative Theorem. Seven sets of the ω-mode eigenvalues exist with using six groups of delayed neutrons and all ω eigenvalues are negative in ADSS. Seven ω-mode adjoint and forward eigenfunctions are employed to form the point-kinetic parameters. The neutron flux is expressed as a linear combination of the products of seven ω-eigenvalue-mode shape functions and their corresponding time functions up to the first order terms, and the lowest negative ω-eigenvalue mode is the dominant mode. (author)
Core conversion effects on the safety analysis of research reactors
International Nuclear Information System (INIS)
Anoussis, J.N.; Chrysochoides, N.G.; Papastergiou, C.N.
1982-07-01
The safety related parameters of the 5 MW Democritus research reactor that will be affected by the scheduled core conversion to use LEU instead of HEU are considered. The analysis of the safety related items involved in such a core conversion, mainly the consequences due to MCA, DBA, etc., is of a general nature and can, therefore, be applied to other similar pool type reactors as well. (T.A.)
Automation of reactor neutron activation analysis
International Nuclear Information System (INIS)
Pavlov, S.S.; Dmitriev, A.Yu.; Frontasyeva, M.V.
2013-01-01
The present status of the development of a software package designed for automation of NAA at the IBR-2 reactor of FLNP, JINR, Dubna, is reported. Following decisions adopted at the CRP Meeting in Delft, August 27-31, 2012, the missing tool - a sample changer - will be installed for NAA in compliance with the peculiar features of the radioanalytical laboratory REGATA at the IBR-2 reactor. The details of the design are presented. The software for operation with the sample changer consists of two parts. The first part is a user interface and the second one is a program to control the sample changer. The second part will be developed after installing the tool.
Depleted Reactor Analysis With MCNP-4B
International Nuclear Information System (INIS)
Caner, M.; Silverman, L.; Bettan, M.
2004-01-01
Monte Carlo neutronics calculations are mostly done for fresh reactor cores. There is today an ongoing activity in the development of Monte Carlo plus burnup code systems made possible by the fast gains in computer processor speeds. In this work we investigate the use of MCNP-4B for the calculation of a depleted core of the Soreq reactor (IRR-1). The number densities as function of burnup were taken from the WIMS-D/4 cell code calculations. This particular code coupling has been implemented before. The Monte Carlo code MCNP-4B calculates the coupled transport of neutrons and photons for complicated geometries. We have done neutronics calculations of the IRR-1 core with the WIMS and CITATION codes in the past Also, we have developed an MCNP model of the IRR-1 standard fuel for a criticality safety calculation of a spent fuel storage pool
Parameter analysis calculation on characteristics of portable FAST reactor
International Nuclear Information System (INIS)
Otsubo, Akira; Kowata, Yasuki
1998-06-01
In this report, we performed a parameter survey analysis by using the analysis program code STEDFAST (Space, TErrestrial and Deep sea FAST reactor-gas turbine system). Concerning the deep sea fast reactor-gas turbine system, calculations with many variable parameters were performed on the base case of a NaK cooled reactor of 40 kWe. We aimed at total equipment weight and surface area necessary to remove heat from the system as important values of the characteristics of the system. Electric generation power and the material of a pressure hull were specially influential for the weight. The electric generation power, reactor outlet/inlet temperatures, a natural convection heat transfer coefficient of sea water were specially influential for the area. Concerning the space reactor-gas turbine system, the calculations with the variable parameters of compressor inlet temperature, reactor outlet/inlet temperatures and turbine inlet pressure were performed on the base case of a Na cooled reactor of 40 kWe. The first and the second variable parameters were influential for the total equipment weight of the important characteristic of the system. Concerning the terrestrial fast reactor-gas turbine system, the calculations with the variable parameters of heat transferred pipe number in a heat exchanger to produce hot water of 100degC for cogeneration, compressor stage number and the kind of primary coolant material were performed on the base case of a Pb cooled reactor of 100 MWt. In the comparison of calculational results for Pb and Na of primary coolant material, the primary coolant weight flow rate was naturally large for the former case compared with for the latter case because density is very different between them. (J.P.N.)
Directory of Open Access Journals (Sweden)
Aparna Sarkar
2015-01-01
Full Text Available Newspaper waste was pyrolysed in a 50 mm diameter and 640 mm long reactor placed in a packed bed pyrolyser from 573 K to 1173 K in nitrogen atmosphere to obtain char and pyro-oil. The newspaper sample was also pyrolysed in a thermogravimetric analyser (TGA under the same experimental conditions. The pyrolysis rate of newspaper was observed to decelerate above 673 K. A deactivation model has been attempted to explain this behaviour. The parameters of kinetic model of the reactions have been determined in the temperature range under study. The kinetic rate constants of volatile and char have been determined in the temperature range under study. The activation energies 25.69 KJ/mol, 27.73 KJ/mol, 20.73 KJ/mol and preexponential factors 7.69 min−1, 8.09 min−1, 0.853 min−1 of all products (solid reactant, volatile, and char have been determined, respectively. A deactivation model for pyrolysis of newspaper has been developed under the present study. The char and pyro-oil obtained at different pyrolysis temperatures have been characterized. The FT-IR analyses of pyro-oil have been done. The higher heating values of both pyro-products have been determined.
Analysis of a molten salt reactor benchmark
International Nuclear Information System (INIS)
Ghosh, Biplab; Bajpai, Anil; Degweker, S.B.
2013-01-01
This paper discusses results of our studies of an IAEA molten salt reactor (MSR) benchmark. The benchmark, proposed by Japan, involves burnup calculations of a single lattice cell of a MSR for burning plutonium and other minor actinides. We have analyzed this cell with in-house developed burnup codes BURNTRAN and McBURN. This paper also presents a comparison of the results of our codes and those obtained by the proposers of the benchmark. (author)
SMART - Structure mechanical analysis in reactor technology
International Nuclear Information System (INIS)
Argyris, J.H.; Faust, G.; Szimmat, J.; Warnke, E.P.; Willam, K.J.
1975-01-01
The programme system SMART was developed in the years 1970-75 to calculate prestressed-concrete reactor pressure vessels with finite elements. The present report outlines the course and present state of research and development work. Following the specification of SMART, a brief presentation of the analytical possibilities and of the expansions for investigating creep, ultimate load behaviour and thermodiffusion is given. In conclusion, the fields of application of SMART are illustrated by means of examples. (orig./LH) [de
Analysis of stirred-tank carbonation reactors
International Nuclear Information System (INIS)
Sheppard, N.F.; Rizo-Patron, R.C.; Sun, W.H.
1978-01-01
The removal of CO 2 from air in a calcium hydroxide slurry-agitated reactor was investigated to aid the design of such vessels. Gas-liquid interfacial areas were calculated using theoretical rate expression and experimental data at specific operating conditions. A correlation for interfacial areas was then determined as a function of impeller speed, impeller diameter, gas flow rate, and concentration of the slurry. Decontamination factors were also determined
Norepinephrine metabolism in humans. Kinetic analysis and model
International Nuclear Information System (INIS)
Linares, O.A.; Jacquez, J.A.; Zech, L.A.; Smith, M.J.; Sanfield, J.A.; Morrow, L.A.; Rosen, S.G.; Halter, J.B.
1987-01-01
The present study was undertaken to quantify more precisely and to begin to address the problem of heterogeneity of the kinetics of distribution and metabolism of norepinephrine (NE) in humans, by using compartmental analysis. Steady-state NE specific activity in arterialized plasma during [ 3 H]NE infusion and postinfusion plasma disappearance of [ 3 H]NE were measured in eight healthy subjects in the supine and upright positions. Two exponentials were clearly identified in the plasma [ 3 H]NE disappearance curves of each subject studied in the supine (r = 0.94-1.00, all P less than 0.01) and upright (r = 0.90-0.98, all P less than 0.01) positions. A two-compartment model was the minimal model necessary to simultaneously describe the kinetics of NE in the supine and upright positions. The NE input rate into the extravascular compartment 2, estimated with the minimal model, increased with upright posture (1.87 +/- 0.08 vs. 3.25 +/- 0.2 micrograms/min per m2, P less than 0.001). Upright posture was associated with a fall in the volume of distribution of NE in compartment 1 (7.5 +/- 0.6 vs. 4.7 +/- 0.3 liters, P less than 0.001), and as a result of that, there was a fall in the metabolic clearance rate of NE from compartment 1 (1.80 +/- 0.11 vs. 1.21 +/- 0.08 liters/min per m2, P less than 0.001). We conclude that a two-compartment model is the minimal model that can accurately describe the kinetics of distribution and metabolism of NE in humans
CFD Modeling of Flow and Ion Exchange Kinetics in a Rotating Bed Reactor System
DEFF Research Database (Denmark)
Larsson, Hilde Kristina; Schjøtt Andersen, Patrick Alexander; Byström, Emil
2017-01-01
A rotating bed reactor (RBR) has been modeled using computational fluid dynamics (CFD). The flow pattern in the RBR was investigated and the flow through the porous material in it was quantified. A simplified geometry representing the more complex RBR geometry was introduced and the simplified...... model was able to reproduce the main characteristics of the flow. Alternating reactor shapes were investigated, and it was concluded that the use of baffles has a very large impact on the flows through the porous material. The simulations suggested, therefore, that even faster reaction rates could...... be achieved by making the baffles deeper. Two-phase simulations were performed, which managed to reproduce the deflection of the gas–liquid interface in an unbaffled system. A chemical reaction was implemented in the model, describing the ion-exchange phenomena in the porous material using four different...
Kinetic calculations for miniature neutron source reactor using analytical and numerical techniques
International Nuclear Information System (INIS)
Ampomah-Amoako, E.
2008-06-01
The analytical methods, step change in reactivity and ramp change in reactivity as well as numerical methods, fixed point iteration and Runge Kutta-gill were used to simulate the initial build up of neutrons in a miniature neutron source reactor with and without temperature feedback effect. The methods were modified to include photo neutron concentration. PARET 7.3 was used to simulate the transients behaviour of Ghana Research Reactor-1. The PARET code was capable of simulating the transients for 2.1 mk and 4 mk insertions of reactivity with peak powers of 49.87 kW and 92.34 kW, respectively. PARET code however failed to simulate 6.71 mk of reactivity which was predicted by Akaho et al through TEMPFED. (au)
Reliability analysis of reactor protection systems
International Nuclear Information System (INIS)
Alsan, S.
1976-07-01
A theoretical mathematical study of reliability is presented and the concepts subsequently defined applied to the study of nuclear reactor safety systems. The theory is applied to investigations of the operational reliability of the Siloe reactor from the point of view of rod drop. A statistical study conducted between 1964 and 1971 demonstrated that most rod drop incidents arose from circumstances associated with experimental equipment (new set-ups). The reliability of the most suitable safety system for some recently developed experimental equipment is discussed. Calculations indicate that if all experimental equipment were equipped with these new systems, only 1.75 rod drop accidents would be expected to occur per year on average. It is suggested that all experimental equipment should be equipped with these new safety systems and tested every 21 days. The reliability of the new safety system currently being studied for the Siloe reactor was also investigated. The following results were obtained: definite failures must be detected immediately as a result of the disturbances produced; the repair time must not exceed a few hours; the equipment must be tested every week. Under such conditions, the rate of accidental rod drops is about 0.013 on average per year. The level of nondefinite failures is less than 10 -6 per hour and the level of nonprotection 1 hour per year. (author)
Neutronics analysis of Dalat Research Reactor
International Nuclear Information System (INIS)
Pham Van Lam; Luong Ba Vien; Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Kien Cuong; Nguyen Manh Hung; Pham Hong Son; Tran Quoc Duong
2006-01-01
Many neutronics codes have been used to calculate for Dalat Research Reactor (DRR) from 1983 (the first critical of DRR in December, 1983). The purposes of all calculations are to know exactly many important parameters related to Reactor Physics and Neutron Physics in reactor core. The results from calculation play important role in core and fuel management for DRR. Especially basing on the results we can predict about fuel cycle, fuel burn up distribution and plan for using optimize remain fresh fuel assemblies of DRR. By using system neutronics code including transport codes, diffusion codes and Mote Carlo code, many characteristics of fuel assemblies and other parameters of whole core were received such as main features of VVR-M2 fuel assembly type, multiplication factor, neutron flux distribution, power distribution, burn up distribution, excess reactivity, control rods worth, neutron spectrum, temperature reactivity coefficient ect. In the paper, brief description all computer codes to being used in DRR and the calculation results from the codes above are presented. (author)
Transient analysis of the IRIS reactor
International Nuclear Information System (INIS)
Bajs, T.; Oriani, L.; Ricotti, M.E.; Barroso, A.C.
2002-01-01
An international consortium of industry, laboratory, university and utility establishments, led by Westinghouse, is developing a modular, integral, light water cooled, small to medium power reactor, the International Reactor Innovative and Secure (IRIS). IRIS features innovative, advanced engineering, but it is firmly based on the proven technology of pressurized water reactors (PWR). Given the large number of organizations involved in the IRIS design, the RELAP5/MOD 3.3 code has been selected as the main system code. A nodalization of the reference IRIS design has been developed with a basic set of protective functions and controls. Engineered Safety Features of the concept are being also implemented, and in particular the Emergency Heat Removal System that is used for safety grade decay heat removal and in the small break LOCA response of IRIS (Large break LOCAs are eliminated in IRIS by the adoption of the Integral layout) This paper discusses developed model and transient behavior of the system for representative transient sequences.(author)
Vibration analysis of reactor assembly internals for Prototype Fast Breeder Reactor
International Nuclear Information System (INIS)
Chellapandi, P.; Jalaldeen, S.; Srinivasan, R.; Chetal, S.C.; Bhoje, S.B.
2003-01-01
Vibration analysis of the reactor assembly components of 500 MWe Prototype Fast Breeder Reactor (PFBR) is presented. The vibration response of primary pump as well as dynamic forces developed at its supports are predicted numerically. The stiffness properties of hydrostatic bearing are determined by formulating and solving governing fluid and structural mechanics equations. The dynamic forces exerted by pump are used as input data for the dynamic response of reactor assembly components, mainly inner vessel, thermal baffle and control plug. Dynamic response of reactor assembly components is also predicted for the pressure fluctuations caused by sodium free level oscillations. Thermal baffle (weir shell) which is subjected to fluid forces developed at the associated sodium free levels is analysed by formulating and solving a set of non-linear equations for fluids, structures and fluid structure interaction (FSI). The control rod drive mechanism is analysed for response under flow induced forces on the parts subjected to cross flow in the zone just above the core top, taking into account FSI between sheaths of control and safety rod and absorber pin bundle. Based on the analysis results, it is concluded that the reactor assembly internals are free from any risk of mechanical as well as flow induced vibrations. (author)
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)
International Nuclear Information System (INIS)
Zbinden, M.
1996-01-01
Certain internal components of Pressurized Water Reactors are damaged by wear when subjected to vibration induced by flow. In order to enable predictive calculation of such wear, one must have a model which takes account reliably of real damages. The modelling of wear represents a final link in a succession of numerical calculations which begins by the determination of hydraulic excitations induced by the flow. One proceeds, then, in the dynamic response calculation of the structure to finish up with an estimation of volumetric wear and of the depth of wear scars. A new concept of industrial wear model adapted to components of nuclear plants is proposed. Its originality is to be supported, on one hand, by experimental results obtained via wear machines of relatively short operational times, and, on the other hand, by the information obtained from the operating feedback over real wear kinetics of the reactors components. The proposed model is illustrated by an example which correspond to a specific real situation. The determination of the coefficients permitting to cover all assembly of configurations and the validation of the model in these configurations have been the object of the most recent work
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.
Neutronic analysis of two-fluid thorium molten salt reactor
International Nuclear Information System (INIS)
Frybort, Jan; Vocka, Radim
2009-01-01
The aim of this paper is to evaluate features of the two-fluid MSBR through a parametric study and compare its properties to one-fluid MSBR concepts. The starting point of the analysis is the original ORNL 1000 MWe reactor design, although simplified to some extent. We studied the influence of dimensions of distinct reactor parts - fuel and fertile channels radius, plenum height, design etc. - on fundamental reactor properties: breeding ratio and doubling time, reactor inventory, graphite lifetime, and temperature feedback coefficients. The calculations were carried out using MCNP5 code. Based on obtained results we proposed an improved reactor design. Our results show clear advantages of the concept with two separate fluoride salts if compared to the one fluid concept in breading, doubling time, and temperature feedback coefficients. Limitations of the two-fluid concept - particularly the graphite lifetime - are also pointed out. The reactor design can be a subject of further optimizations, namely from the viewpoint of reactor safety. (author)
Station Blackout Analysis of HTGR-Type Experimental Power Reactor
Syarip; Zuhdi, Aliq; Falah, Sabilul
2018-01-01
The National Nuclear Energy Agency of Indonesia has decided to build an experimental power reactor of high-temperature gas-cooled reactor (HTGR) type located at Puspiptek Complex. The purpose of this project is to demonstrate a small modular nuclear power plant that can be operated safely. One of the reactor safety characteristics is the reliability of the reactor to the station blackout (SBO) event. The event was observed due to relatively high disturbance frequency of electricity network in Indonesia. The PCTRAN-HTR functional simulator code was used to observe fuel and coolant temperature, and coolant pressure during the SBO event. The reactor simulated at 10 MW for 7200 s then the SBO occurred for 1-3 minutes. The analysis result shows that the reactor power decreases automatically as the temperature increase during SBO accident without operator’s active action. The fuel temperature increased by 36.57 °C every minute during SBO and the power decreased by 0.069 MW every °C fuel temperature rise at the condition of anticipated transient without reactor scram. Whilst, the maximum coolant (helium) temperature and pressure are 1004 °C and 9.2 MPa respectively. The maximum fuel temperature is 1282 °C, this value still far below the fuel temperature limiting condition i.e. 1600 °C, its mean that the HTGR has a very good inherent safety system.
Thermal hydraulics analysis of the Advanced High Temperature Reactor
Energy Technology Data Exchange (ETDEWEB)
Wang, Dean, E-mail: Dean_Wang@uml.edu [University of Massachusetts Lowell, One University Avenue, Lowell, MA 01854 (United States); Yoder, Graydon L.; Pointer, David W.; Holcomb, David E. [Oak Ridge National Laboratory, 1 Bethel Valley RD #6167, Oak Ridge, TN 37831 (United States)
2015-12-01
Highlights: • The TRACE AHTR model was developed and used to define and size the DRACS and the PHX. • A LOFF transient was simulated to evaluate the reactor performance during the transient. • Some recommendations for modifying FHR reactor system component designs are discussed. - Abstract: The Advanced High Temperature Reactor (AHTR) is a liquid salt-cooled nuclear reactor design concept, featuring low-pressure molten fluoride salt coolant, a carbon composite fuel form with embedded coated particle fuel, passively triggered negative reactivity insertion mechanisms, and fully passive decay heat rejection. This paper describes an AHTR system model developed using the Nuclear Regulatory Commission (NRC) thermal hydraulic transient code TRAC/RELAP Advanced Computational Engine (TRACE). The TRACE model includes all of the primary components: the core, downcomer, hot legs, cold legs, pumps, direct reactor auxiliary cooling system (DRACS), the primary heat exchangers (PHXs), etc. The TRACE model was used to help define and size systems such as the DRACS and the PHX. A loss of flow transient was also simulated to evaluate the performance of the reactor during an anticipated transient event. Some initial recommendations for modifying system component designs are also discussed. The TRACE model will be used as the basis for developing more detailed designs and ultimately will be used to perform transient safety analysis for the reactor.
Comparative analysis of solution methods of the punctual kinetic equations
International Nuclear Information System (INIS)
Hernandez S, A.
2003-01-01
The following one written it presents a comparative analysis among different analytical solutions for the punctual kinetics equation, which present two variables of interest: a) the temporary behavior of the neutronic population, and b) The temporary behavior of the different groups of precursors of delayed neutrons. The first solution is based on a method that solves the transfer function of the differential equation for the neutronic population, in which intends to obtain the different poles that give the stability of this transfer function. In this section it is demonstrated that the temporary variation of the reactivity of the system can be managed as it is required, since the integration time for this method doesn't affect the result. However, the second solution is based on an iterative method like that of Runge-Kutta or the Euler method where the algorithm was only used to solve first order differential equations giving this way solution to each differential equation that conforms the equations of punctual kinetics. In this section it is demonstrated that only it can obtain a correct temporary behavior of the neutronic population when it is integrated on an interval of very short time, forcing to the temporary variation of the reactivity to change very quick way without one has some control about the time. In both methods the same change is used so much in the reactivity of the system like in the integration times, giving validity to the results graph the one the temporary behavior of the neutronic population vs. time. (Author)
Kinetics of devolatilisation of forestry wastes from thermogravimetric analysis
Energy Technology Data Exchange (ETDEWEB)
Lapuerta, M.; Hernandez, J.J. [Universidad de Castilla-La Mancha, Ciudad real (Spain). Escuela Tecnica Superior de Ingenieros Industriales; Rodriguez, J.J. [Repsol-YPF, Puertollano (Spain). Chemical Department
2004-10-01
The great potential of Maritime pine (Pinus pinaster) wastes in the middle regions of Spain has motivated an increasing interest about the energy use of this material, either through combustion or gasification processes. Samples of these biomass wastes have been analysed by thermogravimetry under both inert and oxidant atmospheres, from room temperature up to 1100 K, at different heating rates: 10, 30, 40, 50 and 60 K min{sup -1}. An estimation of the proximate analysis of the samples was made from combination of both resulting weight loss curves. The devolatilisation process of the samples was divided into three non-interacting mass-loss events described as parallel first-order reactions, being the first event identified as the moisture loss process, the second one as the hemicellulose and cellulose decomposition process and the third one as that of lignin decomposition. A fitting algorithm to obtain the kinetic parameters permitted a good agreement with experimental results, as well as a good discrimination of the effect of the heating rate. Due to the non-homogeneous nature of the tested samples, the use of other conventional methods for obtaining the kinetic parameters has been proved to be inadequate. (author)
Consequence analysis of core meltdown accidents in liquid metal fast reactor
International Nuclear Information System (INIS)
Suk, S.D.; Hahn, D.
2001-01-01
Core disruptive accidents have been investigated at Korea Atomic Energy Research Institute(KAERI) as part of work to demonstrate the inherent and ultimate safety of the conceptual design of the Korea Advanced Liquid Metal Reactor(KALIMER), a 150 Mw pool-type sodium cooled prototype fast reactor that uses U-Pu-Zr metallic fuel. In this study, a simple method was developed using a modified Bethe-Tait method to simulate the kinetics and hydraulic behavior of a homogeneous spherical core over the period of the super-prompt critical power excursion induced by the ramp reactivity insertion. Calculations of energy release during excursions in the sodium-voided core of the KALIMER were subsequently performed using the method for various reactivity insertion rates up to 100 $/s, which has been widely considered to be the upper limit of ramp rates due to fuel compaction. Benchmark calculations were made to compare with the results of more detailed analysis for core meltdown energetics of the oxide fuelled fast reactor. A set of parametric studies was also performed to investigate the sensitivity of the results on the various thermodynamics and reactor parameters. (author)
International Nuclear Information System (INIS)
Tian Wenxi; Qiu Suizheng; Su Guanghui; Jia Dounan; Liu Xingmin; Zhang Jianwei
2007-01-01
A thermohydraulic and safety analysis code-TSACC has been developed using Fortran90 language to evaluate the transient thermohydraulic behavior of the China advanced research reactor (CARR) under station blackout accident (SBA). For the development of TSACC, a series of corresponding mathematical and physical models were applied. Point reactor neutron kinetics model was adopted for solving the reactor power. All possible flow and heat transfer conditions under station blackout accident were considered and the optional correlations were supplied. The usual finite difference method was abandoned and the integral technique was adopted to evaluate the temperature field of the plate type fuel elements. A new simple and convenient equation was proposed for the resolution of the transient behaviors of the main pump instead of the complicated four-quadrant model. Gear method and Adams method were adopted alternately for a better solution to the stiff differential equations describing the dynamic behavior of the CARR. The computational result of TSACC showed the adequacy of the safety margin of CARR under SBA. For the purpose of Verification and Validation (V and V), the simulated results of TSACC were compared with those of RELAP5/MOD3 and a good agreement was obtained. The adoption of modular programming techniques enables TASCC to be applied to other reactors by easily modifying the corresponding function modules
Nuclear reactor descriptions for space power systems analysis
Mccauley, E. W.; Brown, N. J.
1972-01-01
For the small, high performance reactors required for space electric applications, adequate neutronic analysis is of crucial importance, but in terms of computational time consumed, nuclear calculations probably yield the least amount of detail for mission analysis study. It has been found possible, after generation of only a few designs of a reactor family in elaborate thermomechanical and nuclear detail to use simple curve fitting techniques to assure desired neutronic performance while still performing the thermomechanical analysis in explicit detail. The resulting speed-up in computation time permits a broad detailed examination of constraints by the mission analyst.
Contour analysis of steady state tokamak reactor performance
International Nuclear Information System (INIS)
Devoto, R.S.; Fenstermacher, M.E.
1990-01-01
A new method of analysis for presenting the possible operating space for steady state, non-ignited tokamak reactors is proposed. The method uses contours of reactor performance and plasma characteristics, fusion power gain, wall neutron flux, current drive power, etc., plotted on a two-dimensional grid, the axes of which are the plasma current I p and the normalized beta, β n = β/(I p /aB 0 ), to show possible operating points. These steady state operating contour plots are called SOPCONS. This technique is illustrated in an application to a design for the International Thermonuclear Experimental Reactor (ITER) with neutral beam, lower hybrid and bootstrap current drive. The utility of the SOPCON plots for pointing out some of the non-intuitive considerations in steady state reactor design is shown. (author). Letter-to-the-editor. 16 refs, 3 figs, 1 tab
Transient Analysis Needs for Generation IV Reactor Concepts
International Nuclear Information System (INIS)
Siefken, L.J.; Harvego, E.A.; Coryell, E.W.; Davis, C.B.
2002-01-01
The importance of nuclear energy as a vital and strategic resource in the U. S. and world's energy supply mix has led to an initiative, termed Generation IV by the U.S. Department of Energy (DOE), to develop and demonstrate new and improved reactor technologies. These new Generation IV reactor concepts are expected to be substantially improved over the current generation of reactors with respect to economics, safety, proliferation resistance and waste characteristics. Although a number of light water reactor concepts have been proposed as Generation IV candidates, the majority of proposed designs have fundamentally different characteristics than the current generation of commercial LWRs operating in the U.S. and other countries. This paper presents the results of a review of these new reactor technologies and defines the transient analyses required to support the evaluation and future development of the Generation IV concepts. The ultimate objective of this work is to identify and develop new capabilities needed by INEEL to support DOE's Generation IV initiative. In particular, the focus of this study is on needed extensions or enhancements to SCDAP/RELAP5/3D code. This code and the RELAP5-3D code from which it evolved are the primary analysis tools used by the INEEL and others for the analysis of design-basis and beyond-design-basis accidents in current generation light water reactors. (authors)
A powerful methodology for reactor vessel pressurized thermal shock analysis
International Nuclear Information System (INIS)
Boucau, J.; Mager, T.
1994-01-01
The recent operating experience of the Pressurized Water Reactor (PWR) Industry has focused increasing attention on the issue of reactor vessel pressurized thermal shock (PTS). More specifically, the review of the old WWER-type of reactors (WWER 440/230) has indicated a sensitive behaviour to neutron embrittlement. This led already to some remedial actions including safety injection water preheating or vessel annealing. Such measures are usually taken based on the analysis of a selected number of conservative PTS events. Consideration of all postulated cooldown events would draw attention to the impact of operator action and control system effects on reactor vessel PTS. Westinghouse has developed a methodology which couples event sequence analysis with probabilistic fracture mechanics analyses, to identify those events that are of primary concern for reactor vessel integrity. Operating experience is utilized to aid in defining the appropriate event sequences and event frequencies of occurrence for the evaluation. Once the event sequences of concern are identified, detailed deterministic thermal-hydraulic and structural evaluations can be performed to determine the conditions required to minimize the extension of postulated flaws or enhance flaw arrest in the reactor vessel. The results of these analyses can then be used to better define further modifications in vessel and plant system design and to operating procedures. The purpose of the present paper will be to describe this methodology and to show its benefits for decision making. (author). 1 ref., 3 figs
Structural dynamics in fast reactor accident analysis
International Nuclear Information System (INIS)
Fistedis, S.H.
1975-01-01
Analyses and codes are under development combining the hydrodynamics and solid mechanics (and more recently the bubble dynamics) phenomena to gage the stresses, strains, and deformations of important primary components, as well as the overall adequacy of primary and secondary containments. An arbitrary partition of the structural components treated evolves into (1) a core mechanics effort; and (2) a primary system and containment program. The primary system and containment program treats the structural response of components beyond the core, starting with the core barrel. Combined hydrodynamics-solid mechanics codes provide transient stresses and strains and final deformations for components such as the reactor vessel, reactor cover, cover holddown bolts, as well as the pulses for which the primary piping system is to be analyzed. Both, Lagrangian and Eulerian two-dimensional codes are under development, which provide greater accuracy and longer durations for the treatment of HCDA. The codes are being augmented with bubble migration capability pertaining to the latter stages of the HCDA, after slug impact. Recent developments involve the adaptation of the 2-D Eulerian primary system code to the 2-D elastic-plastic treatment of primary piping. Pulses are provided at the vessel-primary piping interfaces of the inlet and outlet nozzles, calculation includes the elbows and pressure drops along the components of the primary piping system. Recent improvements to the primary containment codes include introduction of bending strength in materials, Langrangian mesh regularization techniques, and treatment of energy absorbing materials for the slug impact. Another development involves the combination of a 2-D finite element code for the reactor cover with the hydrodynamic containment code
Thermodynamic analysis of a supercritical water reactor
International Nuclear Information System (INIS)
Edwards, M.
2007-01-01
A thermodynamic model has been developed for a hypothetical design of a Supercritical Water Reactor, with emphasis on Canadian design criteria. The model solves for cycle efficiency, mass flows and physical conditions throughout the plant based on input parameters of operating pressures and efficiencies of components. The model includes eight feedwater heaters, three feedwater pumps, a deaerator, a condenser, the core, three turbines and two reheaters. To perform the calculations, Microsoft Excel was used in conjunction with FLUIDCAL-IAPWS95 and VBA code. The calculations show that a thermal efficiency of 47.5% can be achieved with a core outlet temperature of 625 o C. (author)
WATER-GAS SHIFT KINETICS OVER IRON OXIDE CATALYSTS AT MEMBRANE REACTOR CONDITIONS; A
International Nuclear Information System (INIS)
Carl R.F. Lund
2001-01-01
This report covers the second year of a project investigating water-gas shift catalysts for use in membrane reactors. It has been established that a simple iron high temperature shift catalyst becomes ineffective in a membrane reactor because the reaction rate is severely inhibited by the build-up of the product CO(sub 2). During the past year, an improved microkinetic model for water-gas shift over iron oxide was developed. Its principal advantage over prior models is that it displays the correct asymptotic behavior at all temperatures and pressures as the composition approaches equilibrium. This model has been used to explore whether it might be possible to improve the performance of iron high temperature shift catalysts under conditions of high CO(sub 2) partial pressure. The model predicts that weakening the surface oxygen bond strength by less than 5% should lead to higher catalytic activity as well as resistance to rate inhibition at higher CO(sub 2) partial pressures. Two promoted iron high temperature shift catalysts were studied. Ceria and copper were each studied as promoters since there were indications in the literature that they might weaken the surface oxygen bond strength. Ceria was found to be ineffective as a promoter, but preliminary results with copper promoted FeCr high temperature shift catalyst show it to be much more resistant to rate inhibition by high levels of CO(sub 2). Finally, the performance of sulfided CoMo/Al(sub 2)O(sub 3) catalysts under conditions of high CO(sub 2) partial pressure was simulated using an available microkinetic model for water-gas shift over this catalyst. The model suggests that this catalyst might be quite effective in a medium temperature water-gas shift membrane reactor, provided that the membrane was resistant to the H(sub 2)S that is required in the feed
Dynamic operator actions analysis for inherently safe fast reactors and light water reactors
International Nuclear Information System (INIS)
Ho, V.; Apostolakis, G.
1988-01-01
A comparative dynamic human actions analysis of inherently safe fast reactors (ISFRs) and light water reactors (LWRs) in terms of systems response and estimated human error rates is presented. Brief overviews of the ISFR and LWR systems are given to illustrate the design differences. Key operator actions required by the ISFR reactor shutdown and decay heat removal systems are identified and are compared with those of the LWR. It is observed that, because of the passive nature of the ISFR safety-related systems, a large time window is available for operator actions during transient events. Furthermore, these actions are fewer in number, are less complex, and have lower error rates and less severe consequences than those of the LWRs. We expect the ISFR operator errors' contribution to risk is smaller (at least in the context of the existing human reliability models) than that of the LWRs. (author)
Automatic control of scale range applied for analog study of reactor kinetics
International Nuclear Information System (INIS)
Sergent, O.; Tellier, N.
1967-01-01
We study the response of a reactor, initially in a sub-critical state, for linear release of reactivity obeying to the following criteria, a rod drop comes in 10 seconds after the moment when the neutron power becomess equal to 10 -3 times the nominal power. We are interested in the maximum reactivity reached and in the energy released during the power excursion. For the power varying in a range from 1 to 10 10 we have used the method of automatic change scale which was installed and described in a previous report [fr
Analysis of reactor cavity radiation streaming: some practical considerations
International Nuclear Information System (INIS)
Simmons, G.L.
1979-01-01
A description is presented of a cost effective analysis procedure for use in the prediction of radiation environments in the cavity and containment building of a nuclear power reactor. Comments are offered on potential problems in certification of analysis procedures and the availability of benchmarkable data sets, both measurements and calculations
Lifetime analysis for fusion reactor first walls and divertor plates
International Nuclear Information System (INIS)
Horie, T.; Tsujimura, S.; Minato, A.; Tone, T.
1987-01-01
Lifetime analysis of fusion reactor first walls and divertor plates is performed by (1) a one-dimensional analytical plate model, and (2) a two-dimensional elastic-plastic finite element method. Life-limiting mechanisms and the limits of applicability for these analysis methods are examined. Structural design criteria are also discussed. (orig.)
Reactor physics and reactor computations
International Nuclear Information System (INIS)
Ronen, Y.; Elias, E.
1994-01-01
Mathematical methods and computer calculations for nuclear and thermonuclear reactor kinetics, reactor physics, neutron transport theory, core lattice parameters, waste treatment by transmutation, breeding, nuclear and thermonuclear fuels are the main interests of the conference
Gas-cooled reactor safety and accident analysis
International Nuclear Information System (INIS)
1985-12-01
The Specialists' Meeting on Gas-Cooled Reactor Safety and Accident Analysis was convened by the International Atomic Energy Agency in Oak Ridge on the invitation of the Department of Energy in Washington, USA. The meeting was hosted by the Oak Ridge National Laboratory. The purpose of the meeting was to provide an opportunity to compare and discuss results of safety and accident analysis of gas-cooled reactors under development, construction or in operation, to review their lay-out, design, and their operational performance, and to identify areas in which additional research and development are needed. The meeting emphasized the high safety margins of gas-cooled reactors and gave particular attention to the inherent safety features of small reactor units. The meeting was subdivided into four technical sessions: Safety and Related Experience with Operating Gas-Cooled Reactors (4 papers); Risk and Safety Analysis (11 papers); Accident Analysis (9 papers); Miscellaneous Related Topics (5 papers). A separate abstract was prepared for each of these papers
Kinetic analysis of pulsed laser induced phosphorescence for uranium determination
International Nuclear Information System (INIS)
Serdeiro, Nelida H.
2003-01-01
The laser induced kinetic phosphorescence allows the uranium determination in different kind of matrices, with a lower detection limit than those reached by other spectroscopic methods. It involves the uranyl ions excitation by a pulsed dye-laser source, followed by temporal resolution of the phosphorescence. This method is used for the determination of trace quantities of uranium in aqueous solution, with a suitable complexant agent, without chemical separation before the analysis. The objective of this paper is to present the results of uranium determinations in different standard samples, water, soil, filter and urine, and a comparison with other methods such as fluorimetry, alpha spectrometry and mass spectrometry. Moreover, the measurement conditions, the advantages and disadvantages, the sample preparation, the interferences and the detection limit are described. (author)
Molten Salt Breeder Reactor Analysis Methods
Energy Technology Data Exchange (ETDEWEB)
Park, Jinsu; Jeong, Yongjin; Lee, Deokjung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)
2015-05-15
Utilizing the uranium-thorium fuel cycle shows considerable potential for the possibility of MSR. The concept of MSBR should be revised because of molten salt reactor's advantage such as outstanding neutron economy, possibility of continuous online reprocessing and refueling, a high level of inherent safety, and economic benefit by keeping off the fuel fabrication process. For the development of MSR research, this paper provides the MSBR single-cell, two-cell and whole core model for computer code input, and several calculation results including depletion calculation of each models. The calculations are carried out by using MCNP6, a Monte Carlo computer code, which has CINDER90 for depletion calculation using ENDF-VII nuclear data. From the calculation results of various reactor design parameters, the temperature coefficients are all negative at the initial state and MTC becomes positive at the equilibrium state. From the results of core rod worth, the graphite control rod alone cannot makes the core subcritical at initial state. But the equilibrium state, the core can be made subcritical state only by graphite control rods. Through the comparison of the results of each models, the two-cell method can represent the MSBR core model more accurately with a little more computational resources than the single-cell method. Many of the thermal spectrum MSR have adopted a multi-region single-fluid strategy.
Preliminary shielding analysis of VHTR reactors
International Nuclear Information System (INIS)
Flaspoehler, Timothy M.; Petrovic, Bojan
2011-01-01
Over the last 20 years a number of methods have been established for automated variance reduction in Monte Carlo shielding simulations. Hybrid methods rely on deterministic adjoint and/or forward calculations to generate these parameters. In the present study, we use the FWCADIS method implemented in MAVRIC sequence of the SCALE6 package to perform preliminary shielding analyses of a VHTR reactor. MAVRIC has been successfully used by a number of researchers for a range of shielding applications, including modeling of LWRs, spent fuel storage, radiation field throughout a nuclear power plant, study of irradiation facilities, and others. However, experience in using MAVRIC for shielding studies of VHTRs is more limited. Thus, the objective of this work is to contribute toward validating MAVRIC for such applications, and identify areas for potential improvement. A simplified model of a prismatic VHTR has been devised, based on general features of the 600 MWt reactor considered as one of the NGNP options. Fuel elements have been homogenized, and the core region is represented as an annulus. However, the overall mix of materials and the relatively large dimensions of the spatial domain challenging the shielding simulations have been preserved. Simulations are performed to evaluate fast neutron fluence, dpa, and other parameters of interest at relevant positions. The paper will investigate and discuss both the effectiveness of the automated variance reduction, as well as applicability of physics model from the standpoint of specific VHTR features. (author)
Analysis of reactor trips involving balance-of-plant failures
International Nuclear Information System (INIS)
Seth, S.; Skinner, L.; Ettlinger, L.; Lay, R.
1986-01-01
The relatively high frequency of plant transients leading to reactor trips at nuclear power plants in the US is of economic and safety concern to the industry. A majority of such transients is due to failures in the balance-of-plant (BOP) systems. As a part of a study conducted for the US Nuclear Regulatory Commission, Mitre has carried out a further analysis of the BOP failures associated with reactor trips. The major objectives of the analysis were to examine plant-to-plant variations in BOP-related trips, to understand the causes of failures, and to determine the extent of any associated safety system challenges. The analysis was based on the Licensee Event Reports submitted on all commercial light water reactors during the 2-yr period, 1984-1985
Gas cooled fast reactor materials: compatibility and reaction kinetics of fuel/matrices couples
International Nuclear Information System (INIS)
Lechelle, J.; Aufore, L.; Basini, V.; Belin, R.; Vaudez, S.
2004-01-01
Fourth Generation Gas cooled Fast Reactor concept implies a fast neutron spectrum and aims to lead to an iso-generation of minor actinides. Criteria have been defined for these fuels such as: high core filling factor, efficient fuel cooling, low operation temperature, i.e. 400-850 deg C, good fission product retention, burn-ups in the range of 5-8 atom%, Pu content in the range of 15-25%. Materials matching this demand are considered: mixed uranium - plutonium nitrides and carbides as fuels, whereas TiN, TiC, ZrN, ZrC, SiC are investigated as inert matrices. Thermo-chemical compatibility studies have been carried out, mostly for (U,Pu)N/SiC and (U,Pu)N/TiN couples. They have been associated to matching diffusional studies. For the first studies, accidental reactor conditions have been chosen (1600 deg C) so as to select a couple. Results are presented in terms of nature and quantity of resulting phases identified by XRD and SEM for thermodynamical equilibrium experiments. (authors)
Piezoresistive microcantilever aptasensor for ricin detection and kinetic analysis
Directory of Open Access Journals (Sweden)
Zhi-Wei Liu
2015-04-01
Full Text Available Up to now, there has been no report on target molecules detection by a piezoresistive microcantilever aptasensor. In order to evaluate the test performance and investigate the response dynamic characteristics of a piezoresistive microcantilever aptasensor, a novel method for ricin detection and kinetic analysis based on a piezoresistive microcantilever aptasensor was proposed, where ricin aptamer was immobilised on the microcantilever surface by biotin-avidin binding system. Results showed that the detection limit of ricin was 0.04μg L−1 (S/N ≥ 3. A linear relationship between the response voltage and the concentration of ricin in the range of 0.2μg L−1-40μg L−1 was obtained, with the linear regression equation of ΔUe = 0.904C + 5.852 (n = 5, R = 0.991, p < 0.001. The sensor showed no response for abrin, BSA, and could overcome the influence of complex environmental disruptors, indicating high specificity and good selectivity. Recovery and reproducibility in the result of simulated samples (simulated water, soil, and flour sample determination met the analysis requirements, which was 90.5∼95.5% and 7.85%∼9.39%, respectively. On this basis, a reaction kinetic model based on ligand-receptor binding and the relationship with response voltage was established. The model could well reflect the dynamic response of the sensor. The correlation coefficient (R was greater than or equal to 0.9456 (p < 0.001. Response voltage (ΔUe and response time (t0 obtained from the fitting equation on different concentrations of ricin fitted well with the measured values.
Analysis of Network Topologies Underlying Ethylene Growth Response Kinetics.
Prescott, Aaron M; McCollough, Forest W; Eldreth, Bryan L; Binder, Brad M; Abel, Steven M
2016-01-01
Most models for ethylene signaling involve a linear pathway. However, measurements of seedling growth kinetics when ethylene is applied and removed have resulted in more complex network models that include coherent feedforward, negative feedback, and positive feedback motifs. The dynamical responses of the proposed networks have not been explored in a quantitative manner. Here, we explore (i) whether any of the proposed models are capable of producing growth-response behaviors consistent with experimental observations and (ii) what mechanistic roles various parts of the network topologies play in ethylene signaling. To address this, we used computational methods to explore two general network topologies: The first contains a coherent feedforward loop that inhibits growth and a negative feedback from growth onto itself (CFF/NFB). In the second, ethylene promotes the cleavage of EIN2, with the product of the cleavage inhibiting growth and promoting the production of EIN2 through a positive feedback loop (PFB). Since few network parameters for ethylene signaling are known in detail, we used an evolutionary algorithm to explore sets of parameters that produce behaviors similar to experimental growth response kinetics of both wildtype and mutant seedlings. We generated a library of parameter sets by independently running the evolutionary algorithm many times. Both network topologies produce behavior consistent with experimental observations, and analysis of the parameter sets allows us to identify important network interactions and parameter constraints. We additionally screened these parameter sets for growth recovery in the presence of sub-saturating ethylene doses, which is an experimentally-observed property that emerges in some of the evolved parameter sets. Finally, we probed simplified networks maintaining key features of the CFF/NFB and PFB topologies. From this, we verified observations drawn from the larger networks about mechanisms underlying ethylene
Analysis of Network Topologies Underlying Ethylene Growth Response Kinetics
Directory of Open Access Journals (Sweden)
Aaron M. Prescott
2016-08-01
Full Text Available Most models for ethylene signaling involve a linear pathway. However, measurements of seedling growth kinetics when ethylene is applied and removed have resulted in more complex network models that include coherent feedforward, negative feedback, and positive feedback motifs. However, the dynamical responses of the proposed networks have not been explored in a quantitative manner. Here, we explore (i whether any of the proposed models are capable of producing growth-response behaviors consistent with experimental observations and (ii what mechanistic roles various parts of the network topologies play in ethylene signaling. To address this, we used computational methods to explore two general network topologies: The first contains a coherent feedforward loop that inhibits growth and a negative feedback from growth onto itself (CFF/NFB. In the second, ethylene promotes the cleavage of EIN2, with the product of the cleavage inhibiting growth and promoting the production of EIN2 through a positive feedback loop (PFB. Since few network parameters for ethylene signaling are known in detail, we used an evolutionary algorithm to explore sets of parameters that produce behaviors similar to experimental growth response kinetics of both wildtype and mutant seedlings. We generated a library of parameter sets by independently running the evolutionary algorithm many times. Both network topologies produce behavior consistent with experimental observations and analysis of the parameter sets allows us to identify important network interactions and parameter constraints. We additionally screened these parameter sets for growth recovery in the presence of sub-saturating ethylene doses, which is an experimentally-observed property that emerges in some of the evolved parameter sets. Finally, we probed simplified networks maintaining key features of the CFF/NFB and PFB topologies. From this, we verified observations drawn from the larger networks about mechanisms
Energy Technology Data Exchange (ETDEWEB)
D. E. Shropshire
2009-01-01
The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.
International Nuclear Information System (INIS)
Shropshire, D.E.
2009-01-01
The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program's understanding of the cost drivers that will determine nuclear power's cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-irradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.
Development of the next generation reactor analysis code system, MARBLE
International Nuclear Information System (INIS)
Yokoyama, Kenji; Hazama, Taira; Nagaya, Yasunobu; Chiba, Go; Kugo, Teruhiko; Ishikawa, Makoto; Tatsumi, Masahiro; Hirai, Yasushi; Hyoudou, Hideaki; Numata, Kazuyuki; Iwai, Takehiko; Jin, Tomoyuki
2011-03-01
A next generation reactor analysis code system, MARBLE, has been developed. MARBLE is a successor of the fast reactor neutronics analysis code systems, JOINT-FR and SAGEP-FR (conventional systems), which were developed for so-called JUPITER standard analysis methods. MARBLE has the equivalent analysis capability to the conventional system because MARBLE can utilize sub-codes included in the conventional system without any change. On the other hand, burnup analysis functionality for power reactors is improved compared with the conventional system by introducing models on fuel exchange treatment and control rod operation and so on. In addition, MARBLE has newly developed solvers and some new features of burnup calculation by the Krylov sub-space method and nuclear design accuracy evaluation by the extended bias factor method. In the development of MARBLE, the object oriented technology was adopted from the view-point of improvement of the software quality such as flexibility, expansibility, facilitation of the verification by the modularization and assistance of co-development. And, software structure called the two-layer system consisting of scripting language and system development language was applied. As a result, MARBLE is not an independent analysis code system which simply receives input and returns output, but an assembly of components for building an analysis code system (i.e. framework). Furthermore, MARBLE provides some pre-built analysis code systems such as the fast reactor neutronics analysis code system. SCHEME, which corresponds to the conventional code and the fast reactor burnup analysis code system, ORPHEUS. (author)
Pressurizer pump reliability analysis high flux isotope reactor
International Nuclear Information System (INIS)
Merryman, L.; Christie, B.
1993-01-01
During a prolonged outage from November 1986 to May 1990, numerous changes were made at the High Flux Isotope Reactor (HFIR). Some of these changes involved the pressurizer pumps. An analysis was performed to calculate the impact of these changes on the pressurizer system availability. The analysis showed that the availability of the pressurizer system dropped from essentially 100% to approximately 96%. The primary reason for the decrease in availability comes because off-site power grid disturbances sometimes result in a reactor trip with the present pressurizer pump configuration. Changes are being made to the present pressurizer pump configuration to regain some of the lost availability
Neutron analysis of the fuel of high temperature nuclear reactors
International Nuclear Information System (INIS)
Bastida O, G. E.; Francois L, J. L.
2014-10-01
In this work a neutron analysis of the fuel of some high temperature nuclear reactors is presented, studying its main features, besides some alternatives of compound fuel by uranium and plutonium, and of coolant: sodium and helium. For this study was necessary the use of a code able to carry out a reliable calculation of the main parameters of the fuel. The use of the Monte Carlo method was convenient to simulate the neutrons transport in the reactor core, which is the base of the Serpent code, with which the calculations will be made for the analysis. (Author)
Transient analysis for PWR reactor core using neural networks predictors
International Nuclear Information System (INIS)
Gueray, B.S.
2001-01-01
In this study, transient analysis for a Pressurized Water Reactor core has been performed. A lumped parameter approximation is preferred for that purpose, to describe the reactor core together with mechanism which play an important role in dynamic analysis. The dynamic behavior of the reactor core during transients is analyzed considering the transient initiating events, wich are an essential part of Safety Analysis Reports. several transients are simulated based on the employed core model. Simulation results are in accord the physical expectations. A neural network is developed to predict the future response of the reactor core, in advance. The neural network is trained using the simulation results of a number of representative transients. Structure of the neural network is optimized by proper selection of transfer functions for the neurons. Trained neural network is used to predict the future responses following an early observation of the changes in system variables. Estimated behaviour using the neural network is in good agreement with the simulation results for various for types of transients. Results of this study indicate that the designed neural network can be used as an estimator of the time dependent behavior of the reactor core under transient conditions
20% inlet header break analysis of Advanced Heavy Water Reactor
International Nuclear Information System (INIS)
Srivastava, A.; Gupta, S.K.; Venkat Raj, V.; Singh, R.; Iyer, K.
2001-01-01
The proposed Advanced Heavy Water Reactor (AHWR) is a 750 MWt vertical pressure tube type boiling light water cooled and heavy water moderated reactor. A passive design feature of this reactor is that the heat removal is achieved through natural circulation of primary coolant at all power levels, with no primary coolant pumps. Loss of coolant due to failure of inlet header results in depressurization of primary heat transport (PHT) system and containment pressure rise. Depressurization activates various protective and engineered safety systems like reactor trip, isolation condenser and advanced accumulator, limiting the consequences of the event. This paper discusses the thermal hydraulic transient analysis for evaluating the safety of the reactor, following 20% inlet header break using RELAP5/MOD3.2. For the analysis, the system is discretized appropriately to simulate possible flow reversal in one of the core paths during the transient. Various modeling aspects are discussed in this paper and predictions are made for different parameters like pressure, temperature, steam quality and flow in different parts of the Primary Heat Transport (PHT) system. Flow and energy discharges into the containment are also estimated for use in containment analysis. (author)
One-dimensional hydrodynamical kinetics model of a cylindrical DBD reactor with N2
International Nuclear Information System (INIS)
Flores-Moreno, M; De la Piedad-Beneitez, A; Barocio-Delgado, S; Mercado-Cabrera, A; López-Callejas, R; Peña-Eguiluz, R; Rodríguez-Méndez, B; Muñoz-Castro, A
2012-01-01
A numerical 1-D model of the chemical kinetics related hydrodynamics of room pressure N 2 plasma at 25 degrees C is reported. This generic discharge is assumed to take place between two cylindrical concentric electrodes, coated in a dielectric material, biased between 1 kV and 10 kV at 60Hz - 3kHz. The model includes the integration of particles conservation and the momentum equations as well as the local field approximation and the Poisson equations for the sake of completeness. The outcome shows that an accumulation of electrons takes place in the close vicinity of the higher voltage electrode, due to the electric field convergence to the internal electrode. Thus, this is a region of intense ionization whereas the generation of free radicals would occur away from the internal electrode. The model predicts no significant influence of the electric field on the heavier particles whose density remains practically constant.
Noise analysis role in reactor safety, Spectral analysis (PSD)
International Nuclear Information System (INIS)
Jovanovic, S.; Velickovic, Lj.
1967-11-01
Spectral power density of a zero power reactor is frequency dependent and related to transfer function of the reactor and to spectral density of the input disturbance. Measurement of spectral power density of a critical system is used to obtain the ratio (β/l), β is the effective yield of delayed neutrons, and l is the effective mean neutron lifetime. When reactor is subcritical, if the effective yie ald of delayed neutrons, the effective mean neutron lifetime are known, the shutdown margin can be determined by relation α = (1 - k (1- β0)/l, k is the effective multiplication factor. Output neutron spectrum at the RB reactor in Vinca was measured for a few reactor core configurations and for a few levels of heavy water at subcritical state. Measured values were satisfactory when the reactor was critical, but the reactor noise of subcritical system was covered by the white noise of the detector and electronic equipment. The Ra-Be source was under the reactor vessel when measurements of subcritical system were done. More efficient detector or external random stimulus for increasing the intensity of neutron fluctuations would be needed to obtain results for subcritical system
Probabilistic safety analysis for the Triga reactor Vienna
International Nuclear Information System (INIS)
Boeck, H.; Kirchsteiger, C.
1988-07-01
Triga-type reactors are the most widely used low power research reactors with power levels up to 3 MW. Although Triga reactors are considered inherently safe, due to their unique features such as prompt negative temperature coefficient and low power density, the reactor core still contains a respectable amount of activity which could lead under very adverse circumstances to radiation exposure both of staff members and of public. Such circumstances could be external events, accidents during fuel element manipulation or a loss of coolant water with exposure of the core. Therefore, it was decided to look more closely to various accident pathways and to calculate their probability, if possible. A major drawback is the lack of statistical material because no centralized registration of failures is carried out. Therefore, in many cases values from other research reactor types or even from power reactor statistics had to be used, thus increasing the uncertainty of the results. As most undesired event or TOP-event in this analysis a radiation exposure of staff members, the public or both together was selected and the probabilities of different pathways leading to this exposure was calculated. In the present case 'radiation exposure' are dose rates or activity concentration above the international accepted limits for occupational staff or public. 20 refs., 10 figs. (Author)
International Nuclear Information System (INIS)
Nopharatana, Annop; Pullammanappallil, Pratap C.; Clarke, William P.
2007-01-01
A series of batch, slurry anaerobic digestion experiments were performed where the soluble and insoluble fractions, and unwashed MSW were separately digested in a 200 l stirred stainless steel vessel at a pH of 7.2 and a temperature of 38 deg. C. It was found that 7% of the total MSW COD was readily soluble, of which 80% was converted to biogas; 50% of the insoluble fraction was solubilised, of this only 80% was converted to biogas. The rate of digesting the insoluble fraction was about four times slower than the rate of digesting the soluble fraction; 48% of the total COD was converted to biogas and 40% of the total nitrogen was converted to ammonia. Soluble and insoluble fractions were broken down simultaneously. The minimum time to convert 95% of the degradable fraction to biogas was 20 days. The lag phase for the degradation of insoluble fraction of MSW can be overcome by acclimatising the culture with the soluble fraction. The rate of digestion and the methane yield was not affected by particle size (within the range of 2-50 mm). A dynamic model was developed to describe batch digestion of MSW. The parameters of the model were estimated using data from the separate digestion of soluble and insoluble fractions and validated against data from the digestion of unwashed MSW. Trends in the specific aceticlastic and formate-utilising methanogenic activity were used to estimate initial methanogenic biomass concentration and bacterial death rate coefficient. The kinetics of hydrolysis of insoluble fraction could be adequately described by a Contois equation and the kinetics of acidogenesis, and aceticlastic and hydrogen utilising methanogenesis by Monod equations
Trip setpoint analysis for the reactor protection system of an advanced integral reactor
International Nuclear Information System (INIS)
Yang, Soo Hyung; Kim, Soo Hyung; Chung, Young Jong; Zee, Sung Quun
2007-01-01
The trip setpoints for the reactor protection system of a 65-MWt advanced integral reactor have been analyzed through sensitivity evaluations by using the Transients and Setpoint Simulation/System-integrated Modular Reactor code. In the analysis, an inadvertent control rod withdrawal event has been considered as an initiating event because this event results in the worst consequences from the viewpoint of the minimum critical heat flux ratio and its consequences are considerably affected by the trip setpoints. Sensitivity evaluations have been performed by changing the trip setpoints for the ceiling of a variable overpower trip (VOPT) function and the pressure of a high pressurizer pressure trip function. Analysis results show that a VOPT function is an effective means to satisfy the acceptance criteria as the control rod rapidly withdraws: on the other hand, a high pressurizer pressure trip function is an essential measure to preserve the safety margin in the case of a slow withdrawal of the control rod because a reactor trip by a VOPT function does not occur in this case. It is also shown that the adoptions of 122.2% of the rated core power and 16.25 MPa as the trip setpoint for the ceiling of a VOPT function and the pressure of a high pressurizer pressure trip function are good selections to satisfy the acceptance criteria
SRAC: JAERI thermal reactor standard code system for reactor design and analysis
International Nuclear Information System (INIS)
Tsuchihashi, Keichiro; Takano, Hideki; Horikami, Kunihiko; Ishiguro, Yukio; Kaneko, Kunio; Hara, Toshiharu.
1983-01-01
The SRAC (Standard Reactor Analysis Code) is a code system for nuclear reactor analysis and design. It is composed of neutron cross section libraries and auxiliary processing codes, neutron spectrum routines, a variety of transport, 1-, 2- and 3-D diffusion routines, dynamic parameters and cell burn-up routines. By making the best use of the individual code function in the SRAC system, the user can select either the exact method for an accurate estimate of reactor characteristics or the economical method aiming at a shorter computer time, depending on the purpose of study. The user can select cell or core calculation; fixed source or eigenvalue problem; transport (collision probability or Sn) theory or diffusion theory. Moreover, smearing and collapsing of macroscopic cross sections are separately done by the user's selection. And a special attention is paid for double heterogeneity. Various techniques are employed to access the data storage and to optimize the internal data transfer. Benchmark calculations using the SRAC system have been made extensively for the Keff values of various types of critical assemblies (light water, heavy water and graphite moderated systems, and fast reactor systems). The calculated results show good prediction for the experimental Keff values. (author)
Core test reactor shield cooling system analysis
International Nuclear Information System (INIS)
Larson, E.M.; Elliott, R.D.
1971-01-01
System requirements for cooling the shield within the vacuum vessel for the core test reactor are analyzed. The total heat to be removed by the coolant system is less than 22,700 Btu/hr, with an additional 4600 Btu/hr to be removed by the 2-inch thick steel plate below the shield. The maximum temperature of the concrete in the shield can be kept below 200 0 F if the shield plug walls are kept below 160 0 F. The walls of the two ''donut'' shaped shield segments, which are cooled by the water from the shield and vessel cooling system, should operate below 95 0 F. The walls of the center plug, which are cooled with nitrogen, should operate below 100 0 F. (U.S.)
Analysis of fast reactor steam generator performance
International Nuclear Information System (INIS)
Hulme, G.; Curzon, A.F.
1992-01-01
A computer model for the prediction of flow and temperature fields within a fast reactor steam generator unit is described. The model combines a commercially available computational fluid dynamics (CFD) solver (PHOENICS) with a steam-tube calculation and provides solutions for the fully coupled flow and temperature fields on both the shell side and the tube side. The model includes the inlet and outlet headers and the bottom end stagnant zone. It also accounts for the effects of support grids and edge-gaps. Two and three dimensional and transient calculations have been performed for both straight tube and J-tube units. Examples of the application of the model are presented. (7 figures) (Author)
Analysis of the radiometric survey during the Argonauta reactor operation
International Nuclear Information System (INIS)
Oliveira, Eara de S.L.; Cardozo, Katia K.M.; Silva, Joao Carlos P.; Santos, Joao Regis dos
2013-01-01
The Argonaut reactor at the Institute of Nuclear Engineering-IEN/CNEN, operates normally, the powers between 1.7 and 340 W on neutrongraphy procedures, production of radionuclides and experimental reactor physics lessons to postgraduate courses. The doses from neutrons and gamma radiation are measured when the reactor is critical, inside the reactor hall and surrounding regions. A study of the data obtained was performed to evaluate the daily need of this survey in the reactor hall. Taking into account the principle ALARA, which aims to optimize and minimize the dose received by the individual, we propose, in this work, through an analysis of the acquired data in occupational radiometric surveys, a reformulation of the area monitoring routine practiced by the team of radiological protection of the Institute of Nuclear Engineering - IEN/CNEN-RJ, whereas other monitoring routines regarding the radiological protection are also applied in the routine of the reactor. The operations under review occurred with the reactor operating 340 W power at intervals of 60, 120 and 180 minutes, in monitoring points in controlled areas, supervised and free. The results showed significant dose values in the output of the J-Channel 9 when the operation occurs with this open. With 180 minutes of operation, the measured values of dose rate were lower than the values at 60 min and 120 operations min. At the point in the supervised area, offsite to the reactor hall, situated in the direction of the J-Channel 9, the value reduces more than 14% in any operating time in relation to the dose rate measured at the point opposite the canal. There is a 50% reduction in the dose rates for operations with and J-9 closed. The results suggest a new frequency of radiometric survey whose mode of operation is maintained in similar conditions, since combined with other relevant practices of radiation protection
Design windows and cost analysis on helical reactors
International Nuclear Information System (INIS)
Kozaki, Y.; Imagawa, S.; Sagara, A.
2007-01-01
The LHD type helical reactors are characterized by a large major radius but slender helical coil, which give us different approaches for power plants from tokamak reactors. For searching design windows of helical reactors and discussing their potential as power plants, we have developed a mass-cost estimating model linked with system design code (HeliCos), thorough studying the relationships between major plasma parameters and reactor parameters, and weight of major components. In regard to cost data we have much experience through preparing ITER construction. To compare the weight and cost of magnet systems between tokamak and helical reactors, we broke down magnet systems and cost factors, such as weights of super conducting strands, conduits, support structures, and winding unit costs, through estimating ITER cost data basis. Based on FFHR2m1 deign we considered a typical 3 GWth helical plant (LHD type) with the same magnet size, coil major radius Rc 14 m, magnetic energy 120 GJ, but increasing plasma densities. We evaluated the weight and cost of magnet systems of 3 GWth helical plant, the total magnet weights of 16,000ton and costs of 210 BYen, which are similar values of tokamak reactors (10,200 ton, 110 BYen in ITER 2002 report, and 21,900 ton, 275 BYen in ITER FDR1999). The costs of strands and winding occupy 70% of total magnet costs, and influence entire power plants economics. The design windows analysis and comparative economics studies to optimize the main reactor parameters have been carried out. Economics studies show that it is misunderstanding to consider helical coils are too large and too expensive to achieve power plants. But we should notice that the helical reactor design windows and economics are very sensitive to allowable blanket space (depend on ergodic layer conditions) and diverter configuration for decreasing heat loads. (orig.)
Muravyev, Nikita V; Koga, Nobuyoshi; Meerov, Dmitry B; Pivkina, Alla N
2017-01-25
This study focused on kinetic modeling of a specific type of multistep heterogeneous reaction comprising exothermic and endothermic reaction steps, as exemplified by the practical kinetic analysis of the experimental kinetic curves for the thermal decomposition of molten ammonium dinitramide (ADN). It is known that the thermal decomposition of ADN occurs as a consecutive two step mass-loss process comprising the decomposition of ADN and subsequent evaporation/decomposition of in situ generated ammonium nitrate. These reaction steps provide exothermic and endothermic contributions, respectively, to the overall thermal effect. The overall reaction process was deconvoluted into two reaction steps using simultaneously recorded thermogravimetry and differential scanning calorimetry (TG-DSC) curves by considering the different physical meanings of the kinetic data derived from TG and DSC by P value analysis. The kinetic data thus separated into exothermic and endothermic reaction steps were kinetically characterized using kinetic computation methods including isoconversional method, combined kinetic analysis, and master plot method. The overall kinetic behavior was reproduced as the sum of the kinetic equations for each reaction step considering the contributions to the rate data derived from TG and DSC. During reproduction of the kinetic behavior, the kinetic parameters and contributions of each reaction step were optimized using kinetic deconvolution analysis. As a result, the thermal decomposition of ADN was successfully modeled as partially overlapping exothermic and endothermic reaction steps. The logic of the kinetic modeling was critically examined, and the practical usefulness of phenomenological modeling for the thermal decomposition of ADN was illustrated to demonstrate the validity of the methodology and its applicability to similar complex reaction processes.
Scram and nonlinear reactor system seismic analysis for a liquid metal fast reactor
International Nuclear Information System (INIS)
Morrone, A.; Brussalis, W.G.
1975-01-01
The paper presents the analysis and results for a LMFBR system which was analyzed for both scram times and seismic responses such as bending moments, accelerations and forces. The reactor system was represented with a one-dimensional nonlinear mathematical model with two degrees of freedom per node (translational and rotational). The model was developed to incorporate as many reactor components as possible without exceeding computer limitations. It consists of 12 reactor components with a total of 71 nodes, 69 beam and pin-jointed elements and 27 gap elements. The gap elements were defined by their clearances, impact spring constants and impact damping constants based on a 50% coefficient of restitution. The horizontal excitation input to the model was the response of the containment building at the location of the reactor vessel supports. It consists of a ten seconds Safe Shutdown Earthquake acceleration-time history at 0.005 seconds intervals and with a maximum acceleration of 0.408 g. The analysis was performed with two Westinghouse special purpose computer programs. The first program calculated the reactor system seismic responses and stored the impact forces on tape. The impact forces on the control rod driveline were converted into vertical frictional forces by multiplying them by a coefficient of friction, and then used by the second program for the scram time determination. The results give time history plots of various seismic responses, and plots of scram times as a function of control rod travel distance for the most critical scram initiation times. The total scram time considering the effects of the earthquake was still acceptable but about 4 times longer than that calculated without the earthquake. The bending moment and shear force responses were used as input for the structural analysis (stresses, deflections, fatigue) of the various components, in combination with the other applicable loading conditions. (orig./HP) [de
Directory of Open Access Journals (Sweden)
Wenzhen Chen
2013-01-01
Full Text Available The singularly perturbed method (SPM is proposed to obtain the analytical solution for the delayed supercritical process of nuclear reactor with temperature feedback and small step reactivity inserted. The relation between the reactivity and time is derived. Also, the neutron density (or power and the average density of delayed neutron precursors as the function of reactivity are presented. The variations of neutron density (or power and temperature with time are calculated and plotted and compared with those by accurate solution and other analytical methods. It is shown that the results by the SPM are valid and accurate in the large range and the SPM is simpler than those in the previous literature.
Abstract of programs for nuclear reactor calculation and kinetic equations solution
International Nuclear Information System (INIS)
Marakazov, A.A.
1977-01-01
The collection includes about 50 annotations of programmes,developed in the Kurchatov Atomic Energy Institute in 1971-1976. The programmes are intended for calculating the neutron flux, for solving systems of multigroup equations in P 3 approximation, for calculating the reactor cell, for analysing the system stability, breeding ratio etc. The programme annotations are compiled according to the following diagram: 1.Programme title. 2.Computer type. 3.Physical problem. 4.Solution method. 5.Calculation limitations. 6.Characteristic computer time. 7.Programme characteristic features. 8.Bound programmes. 9.Programme state. 10.Literature allusions in the programme. 11.Required memory resourses. 12.Programming language. 13.Operation system. 14.Names of authors and place of programme adjusting
Energy Technology Data Exchange (ETDEWEB)
Deng, De-Ming; Chang, Cheng-Hung [Institute of Physics, National Chiao Tung University, Hsinchu 300, Taiwan (China)
2015-05-14
Conventional studies of biomolecular behaviors rely largely on the construction of kinetic schemes. Since the selection of these networks is not unique, a concern is raised whether and under which conditions hierarchical schemes can reveal the same experimentally measured fluctuating behaviors and unique fluctuation related physical properties. To clarify these questions, we introduce stochasticity into the traditional lumping analysis, generalize it from rate equations to chemical master equations and stochastic differential equations, and extract the fluctuation relations between kinetically and thermodynamically equivalent networks under intrinsic and extrinsic noises. The results provide a theoretical basis for the legitimate use of low-dimensional models in the studies of macromolecular fluctuations and, more generally, for exploring stochastic features in different levels of contracted networks in chemical and biological kinetic systems.