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)
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
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
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)
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
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.
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)
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)
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
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
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.
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)
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.
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
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)
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....
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
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.
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
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)
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
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
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
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 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)
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
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
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
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)
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)
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
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)
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.
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)
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
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)
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)
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.)
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)
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)
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.
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)
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
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
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
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
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.
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.
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
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.)
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%.
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)
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.
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...
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.)
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.
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.)
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
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)
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
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
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
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)
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.
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}.
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.
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)
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)
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.
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.
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.
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).
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.
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.
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)
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.
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
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.
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
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)
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
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)
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.
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.
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.
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)
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.)
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
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.
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.
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.
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.
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
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)
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.
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
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
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.
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
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
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
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)
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
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.
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.
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.
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
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)
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)
Some aspects of transformation of the nonlinear plasma equations to the space-independent frame
International Nuclear Information System (INIS)
Paul, S.N.; Chakraborty, B.
1982-01-01
Relativistically correct transformation of nonlinear plasma equations are derived in a space-independent frame. This transformation is useful in many ways because in place of partial differential equations one obtains a set of ordinary differential equations in a single independent variable. Equations of Akhiezer and Polovin (1956) for nonlinear plasma oscillations have been generalized and the results of Arons and Max (1974), and others for wave number shift and precessional rotation of electromagnetic wave are recovered in a space-independent frame. (author)
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
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 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
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.
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
Energy Technology Data Exchange (ETDEWEB)
Torres Vida, J
1963-07-01
This programme, written for the UNIVAC-UCT of J.E.N., obtain the time behaviour of neutron density as a function of both positive and negative step change in reactivity. These results are obtained from solutions of the space-independent kinetic equations of a bare thermal reactor based on the Fermi continuous slowing down model and using six groups of delayed neutrons. (Author) 3 refs.
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.
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.
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.
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
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)
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
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
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
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
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
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.
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)
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....
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)
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)
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...
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.
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.
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.
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...
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)
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)
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
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.
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)
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....
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
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)
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)
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
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)
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.
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)
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...
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
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)
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.
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
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
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
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 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)
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.
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
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
International Nuclear Information System (INIS)
Toyama, Masahiro; Kasai, Shigeo.
1978-01-01
Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)
International Nuclear Information System (INIS)
Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.
1979-01-01
Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)
Koontz, Steven L. (Inventor); Davis, Dennis D. (Inventor)
1991-01-01
A flow reactor for simulating the interaction in the troposphere is set forth. A first reactant mixed with a carrier gas is delivered from a pump and flows through a duct having louvers therein. The louvers straighten out the flow, reduce turbulence and provide laminar flow discharge from the duct. A second reactant delivered from a source through a pump is input into the flowing stream, the second reactant being diffused through a plurality of small diffusion tubes to avoid disturbing the laminar flow. The commingled first and second reactants in the carrier gas are then directed along an elongated duct where the walls are spaced away from the flow of reactants to avoid wall interference, disturbance or turbulence arising from the walls. A probe connected with a measuring device can be inserted through various sampling ports in the second duct to complete measurements of the first and second reactants and the product of their reaction at selected XYZ locations relative to the flowing system.
Zhang, Ruobing; Zhang, Chi; Cheng, XingXin; Wang, Liming; Wu, Yan; Guan, Zhicheng
2007-04-02
Removal of amaranth, a commercial synthetic azo dye widely used in the dye and food industry, was examined as a possible remediation technology for treating dye-contaminated water. Effects of various parameters such as gas flow rate, solution conductivity, pulse repetition frequency, etc., on decolorization kinetics were investigated. Experimental results show that an aqueous solution of 24 mg/l dye is 81.24% decolorized following 30 min plasma treatment for a 50 kV voltage and 0.75 m(3)/h gas flow rate. Decolorization reaction of amaranth in the plasma reactor is a pseudo first order reaction. Rate constant (k) of decolorization increases quickly with increasing the applied voltage, pulse repetition frequency and the gas flow rate. However, when the applied voltage is beyond 50 kV and increases further, increase rate of k decreases. In addition, k decreases quickly when the solution conductivity increases from 200 to 1481 microS/cm. The decolorization reaction has a high rate constant (k=0.0269 min(-1)) when the solution pH is beyond 10. Rate constant k decreases with the decrease of pH and reaches minimum at a pH of about 5 (k(min)=0.01603 min(-1)), then increases to 0.02105 min(-1) when pH decreases to 3.07. About 15% of the initial TOC can be degraded only in about 120 min non-thermal plasma treatment.
International Nuclear Information System (INIS)
Zhang Ruobing; Zhang Chi; Cheng Xingxin; Wang Liming; Wu Yan; Guan Zhicheng
2007-01-01
Removal of amaranth, a commercial synthetic azo dye widely used in the dye and food industry, was examined as a possible remediation technology for treating dye-contaminated water. Effects of various parameters such as gas flow rate, solution conductivity, pulse repetition frequency, etc., on decolorization kinetics were investigated. Experimental results show that an aqueous solution of 24 mg/l dye is 81.24% decolorized following 30 min plasma treatment for a 50 kV voltage and 0.75 m 3 /h gas flow rate. Decolorization reaction of amaranth in the plasma reactor is a pseudo first order reaction. Rate constant (k) of decolorization increases quickly with increasing the applied voltage, pulse repetition frequency and the gas flow rate. However, when the applied voltage is beyond 50 kV and increases further, increase rate of k decreases. In addition, k decreases quickly when the solution conductivity increases from 200 to 1481 μS/cm. The decolorization reaction has a high rate constant (k = 0.0269 min -1 ) when the solution pH is beyond 10. Rate constant k decreases with the decrease of pH and reaches minimum at a pH of about 5 (k min = 0.01603 min -1 ), then increases to 0.02105 min -1 when pH decreases to 3.07. About 15% of the initial TOC can be degraded only in about 120 min non-thermal plasma treatment
International Nuclear Information System (INIS)
Dahmani, M.; Baudron, A.M.; Lautard, J.J.; Erradi, L.
2001-01-01
The mixed dual nodal method MINOS is used to solve the reactor kinetics equations with improved quasistatic IQS model and the θ method is used to solve the precursor equations. The speed of calculation which is the main advantage of the MINOS method and the possibility to use the large time step for shape flux calculation permitted by the IQS method, allow us to reduce considerably the computing time. The IQS/MINOS method is implemented in CRONOS 3D reactor code. Numerical tests on different transient benchmarks show that the results obtained with the IQS/MINOS method and the direct numerical method used to solve the kinetics equations, are very close and the total computing time is largely reduced
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
International Nuclear Information System (INIS)
Wang Lizhang; Zhao Yuemin; Fu Jianfeng
2008-01-01
The electrochemical oxidation of phenolic wastewater in a lab-scale reactor, packed into granular activated carbon (GAC) with Ti/SnO 2 anodes and stainless steel cathodes, was interpreted in this study. GAC saturated rapidly if it was only used as sorbent, but application of suitable electric energy for the system simultaneously could recover the adsorption ability of GAC and maintain the continuous running effectively. The titanium dioxide (TiO 2 ) as catalyst and airflow were also applied to the electrochemical reactor to examine the enhancement for phenol oxidation process. Results revealed that the electrochemical degradation of phenol could be reasonably described by first-order kinetics. In addition, it was illustrated that acid region, increased voltage, more dosage of TiO 2 and higher aeration intensity were all beneficial parameters for phenol oxidation rates. By inspecting the relationship between the rate constants (k) and influencing factors, respectively, an overall kinetic model for phenol oxidation was proposed. The kinetics obtained from the experiments under corresponding electrochemical conditions could provide an accurate estimation of phenol concentration effluent and better design of the packed bed reactor
Energy Technology Data Exchange (ETDEWEB)
Wang Lizhang [College of Environment and Spatial Informatics, China University of Mining and Technology, South Jiefang Road, Quanshan District, Xuzhou City, Jiangsu 221008 (China)], E-mail: wlzh0731@126.com; Zhao Yuemin [School of Chemical Engineering and Technology, China University of Mining and Technology, South Jiefang Road, Quanshan District, Xuzhou City, Jiangsu 221008 (China)], E-mail: ymzhao@cumt.edu.cn; Fu Jianfeng [Department of Environmental Engineering, Southeast University, Nanjing City, Jiangsu 210096 (China)
2008-12-30
The electrochemical oxidation of phenolic wastewater in a lab-scale reactor, packed into granular activated carbon (GAC) with Ti/SnO{sub 2} anodes and stainless steel cathodes, was interpreted in this study. GAC saturated rapidly if it was only used as sorbent, but application of suitable electric energy for the system simultaneously could recover the adsorption ability of GAC and maintain the continuous running effectively. The titanium dioxide (TiO{sub 2}) as catalyst and airflow were also applied to the electrochemical reactor to examine the enhancement for phenol oxidation process. Results revealed that the electrochemical degradation of phenol could be reasonably described by first-order kinetics. In addition, it was illustrated that acid region, increased voltage, more dosage of TiO{sub 2} and higher aeration intensity were all beneficial parameters for phenol oxidation rates. By inspecting the relationship between the rate constants (k) and influencing factors, respectively, an overall kinetic model for phenol oxidation was proposed. The kinetics obtained from the experiments under corresponding electrochemical conditions could provide an accurate estimation of phenol concentration effluent and better design of the packed bed reactor.
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.
Directory of Open Access Journals (Sweden)
I. Grčić
2018-04-01
Full Text Available The possibilities of treating industrial effluents and water purification by advanced oxidation processes have been extensively studied; photocatalysis has emerged as a feasible alternative solution. In order to apply the photocatalytic treatment on a larger scale, relevant modeling approaches are necessary. The scope of this work was to investigate the applicability of recently published kinetic models in different reactor systems (batch and CSTR under UVA or UVC irradiation and in combination with two types of TiO2 catalyst, AEROXIDE® P25 and PC-500 for degradation of azo dyes (C.I. Reactive Violet 2, and C.I. Mordant Yellow 10, oxalic acid and their mixtures. The influences of reactor geometry and irradiation intensities on pollutant oxidation efficiency were examined. The effect of photon absorption by dyes in water matrix was thoroughly studied. Relevant kinetic models were introduced to the mass balance for particular reactor system. Resulting models were sufficient for description of pollutant degradation in batch reactors and CSTR. Experimental results showed 1.15 times higher mineralization extents achieved after 7 cycles in CSTR than in batch photoreactor of similar geometry within the equivalent time-span. The application of CSTR in-series could simplify the photocatalytic water treatment on a larger scale.
International Nuclear Information System (INIS)
Kokkoris, George; Panagiotopoulos, Apostolos; Gogolides, Evangelos; Goodyear, Andy; Cooke, Mike
2009-01-01
Gas phase and reactor wall-surface kinetics are coupled in a global model for SF 6 plasmas. A complete set of gas phase and surface reactions is formulated. The rate coefficients of the electron impact reactions are based on pertinent cross section data from the literature, which are integrated over a Druyvesteyn electron energy distribution function. The rate coefficients of the surface reactions are adjustable parameters and are calculated by fitting the model to experimental data from an inductively coupled plasma reactor, i.e. F atom density and pressure change after the ignition of the discharge. The model predicts that SF 6 , F, F 2 and SF 4 are the dominant neutral species while SF 5 + and F - are the dominant ions. The fit sheds light on the interaction between the gas phase and the reactor walls. A loss mechanism for SF x radicals by deposition of a fluoro-sulfur film on the reactor walls is needed to predict the experimental data. It is found that there is a net production of SF 5 , F 2 and SF 6 , and a net consumption of F, SF 3 and SF 4 on the reactor walls. Surface reactions as well as reactions between neutral species in the gas phase are found to be important sources and sinks of the neutral species.
Directory of Open Access Journals (Sweden)
Ignazio Renato Bellobono
2008-01-01
Full Text Available Photomineralization of methane in air (10.0–1000 ppm (mass/volume of C at 100% relative humidity (dioxygen as oxygen donor was systematically studied at 318±3 K in an annular laboratory-scale reactor by photocatalytic membranes immobilizing titanium dioxide as a function of substrate concentration, absorbed power per unit length of membrane, reactor geometry, and concentration of a proprietary vanadium alkoxide as photopromoter. Kinetics of both substrate disappearance, to yield intermediates, and total organic carbon (TOC disappearance, to yield carbon dioxide, were followed. At a fixed value of irradiance (0.30 W⋅cm-1, the mineralization experiments in gaseous phase were repeated as a function of flow rate (4–400 m3⋅h−1. Moreover, at a standard flow rate of 300 m3⋅h−1, the ratio between the overall reaction volume and the length of the membrane was varied, substantially by varying the volume of reservoir, from and to which circulation of gaseous stream took place. Photomineralization of methane in aqueous solutions was also studied, in the same annular reactor and in the same conditions, but in a concentration range of 0.8–2.0 ppm of C, and by using stoichiometric hydrogen peroxide as an oxygen donor. A kinetic model was employed, from which, by a set of differential equations, four final optimised parameters, k1 and K1, k2 and K2, were calculated, which is able to fit the whole kinetic profile adequately. The influence of irradiance on k1 and k2, as well as of flow rate on K1 and K2, is rationalized. The influence of reactor geometry on k values is discussed in view of standardization procedures of photocatalytic experiments. Modeling of quantum yields, as a function of substrate concentration and irradiance, as well as of concentration of photopromoter, was carried out very satisfactorily. Kinetics of hydroxyl radicals reacting between themselves, leading to hydrogen peroxide, other than with substrate or
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.
International Nuclear Information System (INIS)
Nakahara, Yasuaki; Ise, Takeharu; Kobayashi, Kensuke; Itoh, Yasuyuki
1975-12-01
A new method has been developed for numerical solution of a class of nonlinear Volterra integro-differential equations with quadratic nonlinearity. After dividing the domain of the variable into subintervals, piecewise approximations are applied in the subintervals. The equation is first integrated over a subinterval to obtain the piecewise equation, to which six approximate treatments are applied, i.e. fully explicit, fully implicit, Crank-Nicolson, linear interpolation, quadratic and cubic spline. The numerical solution at each time step is obtained directly as a positive root of the resulting algebraic quadratic equation. The point reactor kinetics with a ramp reactivity insertion, linear temperature feedback and delayed neutrons can be described by one of this type of nonlinear Volterra integro-differential equations. The algorithm is applied to the Argonne benchmark problem and a model problem for a fast reactor without delayed neutrons. The fully implicit method has been found to be unconditionally stable in the sense that it always gives the positive real roots. The cubic spline method is divergent, and the other four methods are intermediate in between. From the estimation of the stability, convergency, accuracy and CPU time, it is concluded that the Crank-Nicolson method is best, then the linear interpolation method comes closely next to it. Discussions are also made on the possibility of applying the algorithm to the fusion reactor kinetics in the form of a nonlinear partial differential equation. (auth.)
State-space representation of the reactor dynamics equations
International Nuclear Information System (INIS)
Bernard, J.A.
1995-01-01
This paper describes a novel formulation of the reactor space-independent kinetics equations. The intent is to present these equations in a form that is both compatible with modern control theory and mathematically rigorous. It is desired to write the kinetics equations in the standard state variable representation, x = Ax, where x is the state vector and A is the system matrix and, at the same time, avoid mathematical compromises such as the linearization of an equation about a particular operating point. The advantage to this proposed formulation is that it may allow the lateral transfer of existing control concepts, some that have been developed for other fields, to the operation of nuclear reactors. For example, sliding mode control has been developed to allow robots to function in a robust manner in the presence of changes in the system model. This is necessary because a robot is expected to be capable of picking up an object of unknown mass and moving that object along a specified trajectory. The variability of the object's mass introduces an uncertainty into the system model that is used to deduce the appropriate control action. Thus, the robot controller must be made robust against such variations. Sliding mode control is one means of accomplishing this. A reactor controller might benefit from the same concept if its objective were to cause the reactor power to move along a demanded trajectory despite the presence of some uncertainty in the net amount of reactivity that is present
Energy Technology Data Exchange (ETDEWEB)
Nasir, Rubina; Mirza, Nasir M. [Dept. of, Physics, Air University, Islamabad (Pakistan); Mirza, Sikander M. [Dept. of, Physics and Applied Mathematics, Pakistan Institute of Engineering and Applied Sciences, Post Office Nilore, Islamabad (Pakistan)
2017-06-15
This work looks at the effect of changes in kinetic parameters on simultaneous reactivity insertions and beam tube flooding in a typical material testing reactor-type research reactor with low enriched high density (U{sub 3}Si{sub 2}-Al) fuel. Using a modified PARET code, various ramp reactivity insertions (from $0.1/0.5 s to $1.3/0.5 s) plus beam tube flooding ($0.5/0.25 s) accidents under uncontrolled conditions were analyzed to find their effects on peak power, net reactivity, and temperature. Then, the effects of changes in kinetic parameters including the Doppler coefficient, prompt neutron lifetime, and delayed neutron fractions on simultaneous reactivity insertion and beam tube flooding accidents were analyzed. Results show that the power peak values are significantly sensitive to the Doppler coefficient of the system in coupled accidents. The material testing reactor-type system under such a coupled accident is not very sensitive to changes in the prompt neutron life time; the core under such a coupled transient is not very sensitive to changes in the effective delayed neutron fraction.
DEFF Research Database (Denmark)
Shah, Vivek; Vaz Salles, Marcos António
2018-01-01
The requirements for OLTP database systems are becoming ever more demanding. Domains such as finance and computer games increasingly mandate that developers be able to encode complex application logic and control transaction latencies in in-memory databases. At the same time, infrastructure...... engineers in these domains need to experiment with and deploy OLTP database architectures that ensure application scalability and maximize resource utilization in modern machines. In this paper, we propose a relational actor programming model for in-memory databases as a novel, holistic approach towards......-level function calls. In contrast to classic transactional models, however, reactors allow developers to take advantage of intra-transaction parallelism and state encapsulation in their applications to reduce latency and improve locality. Moreover, reactors enable a new degree of flexibility in database...
Directory of Open Access Journals (Sweden)
Francesco Saverio Marra
2015-09-01
Full Text Available This paper focus on the behavior of a continuous stirred tank reactor (CSTR subject to perturbations of finite amplitude and frequency. Two main objectives are pursued: to determine the extinction line in the equivalence ratio (φ - residence time (τ plane, fixed the thermodynamic state conditions; and to characterize the response of the chemical system to periodic forcing of the residence time. Transient simulations of combustion of methane with air, using both global single-step and detailed chemical kinetic mechanisms, have been conducted and the corresponding asymptotic solutions analyzed. Results indicate very different dynamical behaviors, posing the issue of a proper choice of the kinetic scheme for the numerical study of combustion oscillations.
International Nuclear Information System (INIS)
Hun, N.
2011-01-01
Gas Fast Reactor (GFR) is one of the different Generation IV concepts under investigation for energy production. SiC/SiC composites are candidates of primary interest for a GFR fuel cladding use, thanks to good corrosion resistance among other properties. The mechanisms and kinetics of SiC oxidation under operating conditions have to be identified and quantified as the corrosion can decrease the mechanical properties of the composite. An experimental device has been developed to study the oxidation of silicon carbide under high temperature and low oxygen partial pressure. The results pointed out that not only parabolic oxidation, but also interfacial reactions and volatilization occur under such conditions. After determining the kinetics of each mechanism, as functions of oxygen partial pressure and temperature, the data are used for the modeling of the composites oxidation. The model will be used to predict the lifetime of the composite in operating conditions. (author) [fr
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
International Nuclear Information System (INIS)
Wynveen, R.A.; Smith, W.H.; Sholeen, C.M.; Flynn, K.F.; Justus, A.L.
1981-10-01
Rockwell International's Santa Susana Laboratories in Ventura County, California, have been the site of numerous federally-funded contracted projects involving the use of radioactive materials. Among these was the Kinetics Experiment Water Boiler (KEWB) Reactor which was operated under the auspices of the US Atomic Energy Commission (AEC). The KEWB Reactor was last operated in 1966. The facility was subsequently declared excess and decontamination and decommissioning operations were conducted during the first half of calendar year 1975. The facility was completely dismantled and the site graded to blend with the surrounding terrain. During October 1981, a post-remedial-action (certification) survey of the KEWB site was conducted on the behalf of the US Department of Energy by the Radiological Survey Group (RSG) of the Occupational Health and Safety Division's Health Physics Section (OHS/HP) of Argonne National Laboratory (ANL). The survey confirmed that the site was free from contamination and could be released for unrestricted use
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)
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.
International Nuclear Information System (INIS)
Fujibayashi, Toru.
1976-01-01
Object: To provide a boiling water reactor which can enhance a quake resisting strength and flatten power distribution. Structure: At least more than four fuel bundles, in which a plurality of fuel rods are arranged in lattice fashion which upper and lower portions are supported by tie-plates, are bundled and then covered by a square channel box. The control rod is movably arranged within a space formed by adjoining channel boxes. A spacer of trapezoidal section is disposed in the central portion on the side of the channel box over substantially full length in height direction, and a neutron instrumented tube is disposed in the central portion inside the channel box. Thus, where a horizontal load is exerted due to earthquake or the like, the spacers come into contact with each other to support the channel box and prevent it from abnormal vibrations. (Furukawa, Y.)
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
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.
Rahman, N K; Kamaruddin, A H; Uzir, M H
2011-08-01
The influence of water activity and water content was investigated with farnesyl laurate synthesis catalyzed by Lipozyme RM IM. Lipozyme RM IM activity depended strongly on initial water activity value. The best results were achieved for a reaction medium with an initial water activity of 0.11 since it gives the best conversion value of 96.80%. The rate constants obtained in the kinetics study using Ping-Pong-Bi-Bi and Ordered-Bi-Bi mechanisms with dead-end complex inhibition of lauric acid were compared. The corresponding parameters were found to obey the Ordered-Bi-Bi mechanism with dead-end complex inhibition of lauric acid. Kinetic parameters were calculated based on this model as follows: V (max) = 5.80 mmol l(-1) min(-1) g enzyme(-1), K (m,A) = 0.70 mmol l(-1) g enzyme(-1), K (m,B) = 115.48 mmol l(-1) g enzyme(-1), K (i) = 11.25 mmol l(-1) g enzyme(-1). The optimum conditions for the esterification of farnesol with lauric acid in a continuous packed bed reactor were found as the following: 18.18 cm packed bed height and 0.9 ml/min substrate flow rate. The optimum molar conversion of lauric acid to farnesyl laurate was 98.07 ± 0.82%. The effect of mass transfer in the packed bed reactor has also been studied using two models for cases of reaction limited and mass transfer limited. A very good agreement between the mass transfer limited model and the experimental data obtained indicating that the esterification in a packed bed reactor was mass transfer limited.
Wübker, S M; Laurenzis, A; Werner, U; Friedrich, C
1997-08-20
The kinetics of degradation of toluene from a model waste gas and of biomass formation were examined in a bioscrubber operated under different nutrient limitations with a mixed culture. The applicability of the kinetics of continuous cultivation of the mixed culture was examined for a special trickle-bed reactor with a periodically moved filter bed. The efficiency of toluene elimination of the bioscrubber was 50 to 57% and depended on the toluene mass transfer as evident from a constant productivity of 0.026 g dry cell weight/L . h over the dilution rate. Under potassium limitation the biomass productivity was reduced by 60% to 0.011 g dry cell weight/L . h at a dilution rate of 0.013/h. Conversely, at low dilution rates the specific toluene degradation rates increased. Excess biomass in a trickle-bed reactor causes reduction of interfacial area and mass transfer, and increase in pressure drop. To avoid these disadvantages, the trickle-bed was moved periodically and biomass was removed with outflowing medium. The concentration of steady state biomass fixed on polyamide beads decreased hyperbolically with the dilution rate. Also, the efficiency of toluene degradation decreased from 72 to 56% with increasing dilution rate while the productivity increased. Potassium limitation generally caused a reduction in biomass, productivity, and yield while the specific degradation increased with dilution rate. This allowed the application of the principles of the chemostat to the trickle-bed reactor described here, for toluene degradation from waste gases. (c) 1997 John Wiley & Sons, Inc. Biotechnol Bioeng 55: 686-692, 1997.
Energy Technology Data Exchange (ETDEWEB)
Perez M, C
2004-07-01
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)
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
Kinetic Study of Methyl Acetate Oxidation in a Pt/Al2O3 Fixed-Bed Reactor
Hoy, Michael; Li, K. Y.; Li, Jeffrey S.; Chen, S. M.; Yaws, C. L.; Chu, H. W.; Simon, W. E.
1994-01-01
To support technology development for future long-term missions, a metabolic simulator will be used in a closed chamber to test the functions of a Controlled Ecological Life Support System (CELSS). Methyl acetate (MA) was selected as the fuel because its metabolic respiratory quotient is near that of humans. A kinetic study of the catalytic oxidation of MA over Pt/Al203 was then conducted to support the design and operation of the simulator. Kinetic data were obtained as a conversion percentage of MA versus retention time. The reaction was studied at one atmosphere and temperatures from 220 to 340 deg. C. The inlet MA concentration was varied from 100 to 2000 ppm with retention times from 0.01 to 10 sec. A first-order rate law and a Langmuir-Hinshelwood rate equation were tested by nonlinear regression of the kinetic data to estimate rate constants in the rate law. Regression results of the L-H equation explain the kinetic data better than the results of the first-order rate law. A Taguchi experimental design was used to study the effects of temperature, retention time, and concentrations of MA, CO2, and O2 on the conversion of MA. Results indicate that temperature has greatest effect, followed by retention time, and finally MA concentration. It was further determined that the effects of CO2 and O2 concentrations, and the cross effects, are negligible.
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)
Nichols, K.P.F.; Azoz, Seyla; Gardeniers, Johannes G.E.
2008-01-01
Enzyme kinetics were obtained in a porous silicon microfluidic channel by combining an enzyme and substrate droplet, allowing them to react and deposit a small amount of residue on the channel walls, and then analyzing this residue by directly ionizing the channel walls using a matrix assisted laser
Ignition of DME and DME/CH4 at High Pressure: Flow Reactor Experiments and Kinetic Modeling
DEFF Research Database (Denmark)
Hashemi, Hamid; Christensen, Jakob Munkholt; Glarborg, Peter
The pyrolysis and oxidation of dimethyl ether (DME) and its mixtures with methane were investigated at high pressures (50 and 100 bar) and intermediate temperatures (450―900 K) in a laminar flow reactor. DME pyrolysis started at 825 K (at 50 bar). The onset of DME reaction was detected at 525―550 K...
Kinetics of Ar+*(2G9/2) metastable ions and transport of argon ions in ICP reactor
Sadeghi, N.; Derouard, J.; Grift, van de M.; Kroesen, G.M.W.; Hoog, de F.J.; Tachibana, K.; Watanabe, Y.
1997-01-01
The decay time of the argon Ar~~(2G912) metastable ions was measured in the afterglow of a low pressure pulsed helicon reactor. From the argon pressure and electron density dependence of this decay time, rate coefficients for quenching of these ions by argon atoms and by plasma electrons have been
Energy Technology Data Exchange (ETDEWEB)
Mondal, Prasenjit; Mohanty, Bikash; Majumder, Chandrajit Balo [Department of Chemical Engineering, Indian Institute of Technology Roorkee, Roorkee, Uttrakhand (India)
2012-05-15
This paper deals with kinetics and equilibrium studies on the adsorption of arsenic species from simulated groundwater containing arsenic (As(III)/As(V), 1:1), Fe, and Mn in concentrations of 0.188, 2.8, and 0.6 mg/L, respectively, by Ca{sup 2+} impregnated granular activated charcoal (GAC-Ca). Effects of agitation period and initial arsenic concentration on the removal of arsenic species have also been described. Although, most of the arsenic species are adsorbed within 10 h of agitation, equilibrium reaches after {proportional_to}24 h. Amongst various kinetic models investigated, the pseudo second order model is more adequate to explain the adsorption kinetics and film diffusion is found to be the rate controlling step for the adsorption of arsenic species on GAC-Ca. Freundlich isotherm is adequate to explain the adsorption equilibrium. However, empirical polynomial isotherm gives more accurate prediction on equilibrium specific uptakes of arsenic species. Maximum specific uptake (q{sub max}) for the adsorption of As(T) as obtained from Langmuir isotherm is 135 {mu}g/g. (Copyright copyright 2012 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)
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)
International Nuclear Information System (INIS)
Tanomaru, N.
1979-12-01
The problem of parameter identification in a pontual model for a thermal reactor is dealt with using the quasilinearization technique. The model considers one group of delayed neutrons and a heavily non-linear temperature feedback in the reactivity. The parameter prompt neutron generation time and a parameter of the fuel temperatura reactivity coefficient equation are identified simultaneously, considering discrete measurements of the reactor power, during the transient produced by a change in the external reactivity. The influences of the choice of the external reactivity disturbance, of the two parameters values initial guesses, of the interval between measurements and the measurement noise level in the method accuracy and rate of convergence are analysed. For noiseless or low level noise measurements, the method proved to be very effective. (Author) [pt
Wols, B A; Harmsen, D J H; Wanders-Dijk, J; Beerendonk, E F; Hofman-Caris, C H M
2015-05-15
UV/H2O2 treatment is a well-established technique to degrade organic micropollutants. A CFD model in combination with an advanced kinetic model is presented to predict the degradation of organic micropollutants in UV (LP)/H2O2 reactors, accounting for the hydraulics, fluence rate, complex (photo)chemical reactions in the water matrix and the interactions between these processes. The model incorporates compound degradation by means of direct UV photolysis, OH radical and carbonate radical reactions. Measurements of pharmaceutical degradations in pilot-scale UV/H2O2 reactors are presented under different operating conditions. A comparison between measured and modeled degradation for a group of 35 pharmaceuticals resulted in good model predictions for most of the compounds. The research also shows that the degradation of organic micropollutants can be dependent on temperature, which is relevant for full-scale installations that are operated at different temperatures over the year. Copyright © 2015 Elsevier Ltd. All rights reserved.
Cai, Liming; Sudholt, Alena; Lee, Dongjoon; Egolfopoulos, Fokion N.; Pitsch, Heinz G.; Westbrook, Charles K.; Sarathy, Mani
2014-01-01
The combustion characteristics of promising alternative fuels have been studied extensively in the recent years. Nevertheless, the pyrolysis and oxidation kinetics for many oxygenated fuels are not well characterized compared to those of hydrocarbons. In the present investigation, the first chemical kinetic study of a long-chain linear symmetric ether, di-n-butyl ether (DBE), is presented and a detailed reaction model is developed. DBE has been identified recently as a candidate biofuel produced from lignocellulosic biomass. The model includes both high temperature and low temperature reaction pathways with reaction rates generated using appropriate rate rules. In addition, experimental studies on fundamental combustion characteristics, such as ignition delay times and laminar flame speeds have been performed. A laminar flow reactor was used to determine the ignition delay times of lean and stoichiometric DBE/air mixtures. The laminar flame speeds of DBE/air mixtures were measured in the stagnation flame configuration for a wide rage of equivalence ratios at atmospheric pressure and an unburned reactant temperature of 373. K. All experimental data were modeled using the present kinetic model. The agreement between measured and computed results is satisfactory, and the model was used to elucidate the oxidation pathways of DBE. The dissociation of keto-hydroperoxides, leading to radical chain branching was found to dominate the ignition of DBE in the low temperature regime. The results of the present numerical and experimental study of the oxidation of di-n-butyl ether provide a good basis for further investigation of long chain linear and branched ethers. © 2013 The Combustion Institute.
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)
Cai, Liming
2014-03-01
The combustion characteristics of promising alternative fuels have been studied extensively in the recent years. Nevertheless, the pyrolysis and oxidation kinetics for many oxygenated fuels are not well characterized compared to those of hydrocarbons. In the present investigation, the first chemical kinetic study of a long-chain linear symmetric ether, di-n-butyl ether (DBE), is presented and a detailed reaction model is developed. DBE has been identified recently as a candidate biofuel produced from lignocellulosic biomass. The model includes both high temperature and low temperature reaction pathways with reaction rates generated using appropriate rate rules. In addition, experimental studies on fundamental combustion characteristics, such as ignition delay times and laminar flame speeds have been performed. A laminar flow reactor was used to determine the ignition delay times of lean and stoichiometric DBE/air mixtures. The laminar flame speeds of DBE/air mixtures were measured in the stagnation flame configuration for a wide rage of equivalence ratios at atmospheric pressure and an unburned reactant temperature of 373. K. All experimental data were modeled using the present kinetic model. The agreement between measured and computed results is satisfactory, and the model was used to elucidate the oxidation pathways of DBE. The dissociation of keto-hydroperoxides, leading to radical chain branching was found to dominate the ignition of DBE in the low temperature regime. The results of the present numerical and experimental study of the oxidation of di-n-butyl ether provide a good basis for further investigation of long chain linear and branched ethers. © 2013 The Combustion Institute.
DEFF Research Database (Denmark)
Lackner, Susanne; Smets, Barth F.
2012-01-01
was on the influence of key biokinetic parameters (maximum specific growth rates, oxygen and nitrogen affinity constants of AOB (ammonium oxidizing bacteria) and NOB (nitrite oxidizing bacteria)) and their ratios on nitritation efficiency in these geometries. This exhaustive simulation study revealed that nitritation...... strongly depends on the chosen kinetic parameters of AOB and NOB. The maximum specific growth rates (μmax,AOB and μmax,NOB) had the strongest impact on nitritation efficiency (NE). In comparison, the counter-diffusion geometry yielded more parameter combinations (27.5%) that resulted in high NE than the co...
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.
Energy Technology Data Exchange (ETDEWEB)
Uma, B.; Sandhya, S. [National Environmental Engineering Research Institute, CSIR-Complex, Madras (India)
1998-04-01
Bacillus coagulans strain isolated from contaminated soil was immobilised on activated carbon for degradation of pyridine, toluene and methylene chloride containing synthetic wastewaters. Pyridine was supplied as the only source of nitrogen in the wastewaters. Continuous runs in a packed bed laboratory reactor showed that immobilized B. coagulans can degrade pyridine along with other organics rapidly and the effluent ammonia is also controlled in presence of ``organic carbon``. About 644 mg/l of influent TOC was efficiently degraded (82.85%) at 64.05 mg/l/hr loading. (orig.) With 2 figs., 4 tabs., 15 refs.
Energy Technology Data Exchange (ETDEWEB)
Obradovic, D; Petrovic, M [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)
1966-11-15
This paper describes experimental study of space-time behaviour of nuclear reactors by local complex-periodic perturbation of the absorption cross section and by measuring the local reactor response to this perturbation. Perturbation was done by BOR-1 fast oscillator. Cross correlation between the response and the perturbation was done numerically after completing the measurement by using digital computer. Obtained experimental results are preliminary and are measured with significant errors which were analysed in this paper. Results show qualitative agreement with those obtained by theoretical model. This paper is the first progress report in this field in our country. U radu je opisan eksperimentalni prilaz proucavanju prostorno-vremenskog ponasanja nuklearnih reaktora, vrseci lokalnu slozeno-periodicnu perturbaciju apsorpcionog preseka i mereci lokalni odziv reaktora na tu perturbaciju. Perturbacija je vrsena brzim oscilatorom BOR-1. Kroskorelacija izmedju odziva i perturbacije vrsena je numericki posle obavljenog merenja upotrebom digitalne racunske masine. Dobijeni eksperimentalni rezultati imaju preliminarni karakter i odredjeni su sa znatnim eksperimentalnim greskama koje su, u radu analizirane. Izlozeni rezultati pokazuju kvalitativna slaganja sa teorijski dobijenim modelom. Ovaj rad predstavlja prvi izvestaj o napredovanju na ovoj problematici kod nas (author)
Dev, Subhabrata; Roy, Shantonu; Bhattacharya, Jayanta
2016-07-15
A novel marine waste extract (MWE) as alternative nitrogen source was explored for the growth of sulfate reducing bacteria (SRB). Variation of sulfate and nitrogen (MWE) showed that SRB growth follows an uncompetitive inhibition model. The maximum specific growth rates (μmax) of 0.085 and 0.124 h(-1) and inhibition constants (Ki) of 56 and 4.6 g/L were observed under optimized sulfate and MWE concentrations, respectively. The kinetic data shows that MWE improves the microbial growth by 27%. The packed bed bioreactor (PBR) under optimized sulfate and MWE regime showed sulfate removal efficiency of 62-66% and metals removal efficiency of 66-75% on using mine wastewater. The microbial community analysis using DGGE showed dominance of SRB (87-89%). The study indicated the optimum dosing of sulfate and cheap organic nitrogen to promote the growth of SRB over other bacteria. Copyright © 2016 Elsevier Ltd. All rights reserved.
International Nuclear Information System (INIS)
Grieder, Christoph; Mittweg, Greta; Dhillon, Baldev S.; Montes, Juan M.; Orsini, Elena; Melchinger, Albrecht E.
2012-01-01
Maize (Zea mays L.) is the most competitive crop for methane production in Germany. Methane fermentation yield per unit of dry matter (MFY) is a determinant of methane yield, but little information is available on this trait. Our objectives were to investigate the kinetics of MFY during fermentation of maize, estimate quantitative-genetic parameters for different traits related to MFY and examine the relationship of MFY with chemical composition and silage quality. Whole-plant material of 16 inbreds and their 32 testcrosses was analyzed for MFY over 35 days of fermentation using a discontinuous laboratory assay. Data were also generated on chemical composition and in vitro digestible organic matter (IVDOM). Significant genotypic variances and high heritabilities were observed for MFY at early fermentation stages (up to 5 days) probably due to different concentrations of easily degradable chemical components. However, genotypic variances and heritability of MFY reduced as fermentation progressed, because of complete or partial degradation of all chemical components. Further, there were strong correlations of MFY with chemical components at early fermentation stages but not at later stages. Therefore, MFY at later stages, which is closer to potential MFY, does not seem to be amenable to selection. High heritability of IVDOM and its strong correlation with MFY in testcrosses indicated its possible use for preliminary or indirect selection. Keeping in view the magnitude of genetic variance that was low for MFY and high for dry matter yield (DMY), the other component of methane yield, more emphasis on breeding for DMY seems appropriate. -- Highlights: ► We investigated methane fermentation yield (MFY) of diverse germplasm of maize. ► The kinetics of MFY and its correlations with chemical composition were examined. ► Genetic variance and heritability for MFY decreased with fermentation time. ► Complete fermentation (35 d) reduced correlations of MFY with chemical
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)
Rebo, Hans Petter
1999-07-01
In Norway, the limited offshore oil resources, the abundance of natural gas and the need to recover associated gas from the crude oil production have made the utilisation of natural gas the focus of increased attention. Most products from refineries and chemical industry are formed by gas phase reactions over solid materials like metals, metal oxides and zeolites. Heterogeneous catalysts are in addition frequently used for environmental purposes and energy production. In the work described in this thesis, an experimental set-up was built and used to study some typical processes in heterogeneous catalysis. The set-up included a tapered element oscillating microbalance (TEOM) for measuring mass changes. The following properties of the TEOM were found particularly useful: (1) Frequent frequency counting makes the TEOM suitable for recording transient uptake curves, (2) High sensitivity of the microbalance makes it possible to work with low catalyst loading and still obtain high signal to noise ratio, and (3) Reliable kinetic data are obtained due to the fixed bed characteristics of the TEOM. Adsorption and diffusion of o-xylene and toluene in a commercial HZSM-5 zeolite were studied at 30, 100 and 200 {sup o}C and at partial pressures in the range of 0.002-0.1 bar. The effect of coke on the adsorption and diffusion properties were studied by adsorption experiments at 30 {sup o}C of ethane, toluene and n-hexane before and after coke formation during ethene oligomerisation at 475 {sup o}C and at P(ethene)=0.8 bar. The oligomerisation of ethene over HZSM-5 was used as a model reaction for comparing coke formation in a gravimetric microbalance and in the TEOM. The work also includes a study of coke formation and the effect of coke on the kinetics of propene dehydrogenation over Pt-Sn/Al{sub 2}O{sub 3} catalysts at 500-580 {sup o}C.
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
DEFF Research Database (Denmark)
Zheng, Yuanjing; Jensen, Peter Arendt; Jensen, Anker Degn
2008-01-01
The reactions between gaseous potassium chloride and coal minerals were investigated in a lab-scale high temperature fixed-bed reactor using single sorbent pellets. The applied coal minerals included kaolin, mullite, silica, alumina, bituminous coal ash, and lignite coal ash that were formed...... into long cylindrical pellets. Kaolin and bituminous coal ash that both have significant amounts of Si and Al show superior potassium capture characteristics. Experimental results show that capture of potassium by kaolin is independent of the gas oxygen content. Kaolin releases water and forms metakaolin...... when heated at temperatures above 450°C. The amounts of potassium captured by metakaolin pellet decreases with increasing reaction temperature in the range of 900-1300°C and increases again with further increasing the temperature up to 1500°C. There is no reaction of pre-made mullite with KCl...
International Nuclear Information System (INIS)
Wintergerst, M.
2009-05-01
For the Jules Horowitz new material-testing reactor (JHR), an aluminium base alloy, called AlFeNi, will be used for the cladding of the fuel plates. This alloy (Al - 1% Fe - 1% Ni - 1 % Mg), which is already used as fuel cladding, was developed for its good corrosion resistance in water at high temperatures. However, few studies dealing with the alteration process in water and the relationships with irradiation effects have been performed on this alloy. The conception of the JHR fuel requires a better knowledge of the corrosion mechanisms. Corrosion tests were performed in autoclaves at 70 C, 165 C and 250 C on AlFeNi plates representative of the fuel cladding. Several techniques were used to characterize the corrosion scale: SEM, TEM, EPMA, XRD, Raman spectroscopy. Our observations show that the corrosion scale is made of two main layers: a dense amorphous scale close to the metal and a porous crystalline scale in contact with the water. More than the morphology, the chemical compositions of both layers are different. This duplex structure results from a mixed growth mechanism: an anionic growth to develop the inner oxide and a cationic diffusion followed by a dissolution-precipitation process to form the outer one. Dynamic experiments at 70 C and corrosion kinetics measurements have demonstrated that the oxide growth process is controlled by a diffusion step associated to a dissolution/precipitation process. A corrosion mechanism of the AlFeNi alloy in aqueous media has been proposed. Then post-irradiation exams performed on irradiated fuel plates were used to investigate the effects of the irradiation on the corrosion behaviour in the reactor core. (author)
Energy Technology Data Exchange (ETDEWEB)
Geslot, Benoit; Pepino, Alexandra; Blaise, Patrick; Mellier, Frederic [CEA, DEN, DER/SPEx, Cadarache, F-13108 St Paul Lez Durance (France); Lecouey, Jean-Luc [LPC Caen, ENSICAEN, Universite de Caen, CNRS/IN2P3, 6 Bd. Marechal Juin 14050 Caen cedex (France); Carta, Mario [ENEA, UTFISST-REANUC, C.R. Casaccia, S.P.040 via Anguillarese 301, 00123 S. Maria Di Galeria, Roma (Italy); Kochetkov, Anatoly; Vittiglio, Guido [SCK.CEN, Belgian Nuclear Research Centre, Boeretang 200, BE-2400, Mol (Belgium); Billebaud, Annick [LPSC, CNRS, IN2P3/UJF/INPG, 53 Avenue des Martyrs, 38026 Grenoble cedex (France)
2015-07-01
A pile noise measurement campaign has been conducted by the CEA in the VENUS-F reactor (SCK-CEN, Mol Belgium) in April 2011 in the reference critical configuration of the GUINEVERE experimental program. The experimental setup made it possible to estimate the core kinetic parameters: the prompt neutron decay constant, the delayed neutron fraction and the generation time. A precise assessment of these constants is of prime importance. In particular, the effective delayed neutron fraction is used to normalize and compare calculated reactivities of different subcritical configurations, obtained by modifying either the core layout or the control rods position, with experimental ones deduced from the analysis of measurements. This paper presents results obtained with a CEA-developed time stamping acquisition system. Data were analyzed using Rossi-α and Feynman-α methods. Results were normalized to reactor power using a calibrated fission chamber with a deposit of Np-237. Calculated factors were necessary to the analysis: the Diven factor was computed by the ENEA (Italy) and the power calibration factor by the CNRS/IN2P3/LPC Caen. Results deduced with both methods are consistent with respect to calculated quantities. Recommended values are given by the Rossi-α estimator, that was found to be the most robust. The neutron generation time was found equal to 0.438 ± 0.009 μs and the effective delayed neutron fraction is 765 ± 8 pcm. Discrepancies with the calculated value (722 pcm, calculation from ENEA) are satisfactory: -5.6% for the Rossi-α estimate and -2.7% for the Feynman-α estimate. (authors)
International Nuclear Information System (INIS)
Gollion, H.
1977-01-01
The reasons for the development of fast reactors are briefly reviewed (a propitious neutron balance oriented towards a maximum uranium burnup) and its special requirements (cooling, fissile material density and reprocessing) discussed. The three stages in the French program of fast reactor development are outlined with Rapsodie at Cadarache, Phenix at Marcoule, and Super Phenix at Creys-Malville. The more specific features of the program of research and development are emphasized: kinetics and the core, the fuel and the components [fr
International Nuclear Information System (INIS)
Langlais, F.; Hottier, F.; Cadoret, R.
1982-01-01
Silicon chemical vapour deposition (SiH 2 Cl 2 /H 2 system), under reduced pressure conditions, in a hot-wall reactor, is presented. The vapour phase composition is assessed by evaluating two distinct equilibria. The homogeneous equilibrium , which assumes that the vapour phase is not in equilibrium with solid silicon, is thought to give an adequate description of the vapour phase in the case of low pressure, high gas velocities, good temperature homogeneity conditions. A comparison with heterogeneous equilibrium enables us to calculate the supersaturation so evidencing a highly irreversible growth system. The experimental determination of the growth rates reveals two distinct temperature ranges: below 1000 0 C, polycrystalline films are usually obtained with a thermally activated growth rate (+40 kcal mole -1 ) and a reaction order, with respect to the predominant species SiCl 2 , close to one; above 1000 0 C, the films are always monocrystalline and their growth rate exhibits a much lower or even negative activation energy, the reaction order in SiCl 2 remaining about one. (orig.)
Kinetics study of heterogeneous reactions of ozone with erucic acid using an ATR-IR flow reactor.
Leng, Chunbo; Hiltner, Joseph; Pham, Hai; Kelley, Judas; Mach, Mindy; Zhang, Yunhong; Liu, Yong
2014-03-07
The ozone initiated heterogeneous oxidation of erucic acid (EA) thin film was investigated using a flow system combined with attenuated total reflection infrared spectroscopy (ATR-IR) over wide ranges of ozone concentrations (0.25-60 ppm), thin film thickness (0.1-1.0 μm), temperatures (263-298 K), and relative humidities (0-80% RH) for the first time. Pseudo-first-order rate constants, kapp, and overall reactive uptake coefficients, γ, were obtained through changes in the absorbance of C[double bond, length as m-dash]O stretching bands at 1695 cm(-1), which is assigned to the carbonyl group in carboxylic acid. Results showed that the reaction followed the Langmuir-Hinshelwood mechanism and kapp was largely dominated by surface reaction over bulk phase reaction. In addition, both the kapp and the γ values showed very strong temperature dependences (∼two orders of magnitude) over the temperature range; in contrast, they only slightly increased with increasing RH values from 0-80%. According to the kapp values as a function of temperature, the activation energy for the heterogeneous reaction was estimated to be 80.6 kJ mol(-1). Our results have suggested that heterogeneous reactions between ozone and unsaturated solid surfaces likely have a substantially greater temperature dependence than liquid ones. Moreover, the hygroscopic properties of EA thin films before and after exposure to ozone were also studied by measurement of water uptake. Based on the hygroscopicity data, the insignificant RH effect on reaction kinetics was probably due to the relatively weak water uptake by the unreacted and reacted EA thin films.
International Nuclear Information System (INIS)
Baeten, Peter
2006-01-01
This course gives an introduction to Nuclear Reactor Physics. The first chapter explains the most important parameters and concepts in nuclear reactor physics such as fission, cross sections and the effective multiplication factor. Further on, in the second chapter, the flux distributions in a stationary reactor are derived from the diffusion equation. Reactor kinetics, reactor control and reactor dynamics (feedback effects) are described in the following three chapters. The course concludes with a short description of the different types of existing and future reactors. (author)
Payn, R. A.; Helton, A. M.; Poole, G.; Izurieta, C.; Bernhardt, E. S.; Burgin, A. J.
2012-12-01
Many hypotheses have been proposed to predict patterns of biogeochemical redox reactions based on the availability of electron donors and acceptors and the thermodynamic theory of chemistry. Our objective was to develop a computer model that would allow us to test various alternatives of these hypotheses against data gathered from soil slurry batch reactors, experimental soil perfusion cores, and in situ soil profile observations from the restored Timberlake Wetland in coastal North Carolina, USA. Software requirements to meet this objective included the ability to rapidly develop and compare different hypothetical formulations of kinetic and thermodynamic theory, and the ability to easily change the list of potential biogeochemical reactions used in the optimization scheme. For future work, we also required an object pattern that could easily be coupled with an existing soil hydrologic model. These requirements were met using Network Exchange Objects (NEO), our recently developed object-oriented distributed modeling framework that facilitates simulations of multiple interacting currencies moving through network-based systems. An initial implementation of the object pattern was developed in NEO based on maximizing growth of the microbial community from available dissolved organic carbon. We then used this implementation to build a modeling system for comparing results across multiple simulated batch reactors with varied initial solute concentrations, varied biogeochemical parameters, or varied optimization schemes. Among heterotrophic aerobic and anaerobic reactions, we have found that this model reasonably predicts the use of terminal electron acceptors in simulated batch reactors, where reactions with higher energy yields occur before reactions with lower energy yields. However, among the aerobic reactions, we have also found this model predicts dominance of chemoautotrophs (e.g., nitrifiers) when their electron donor (e.g., ammonium) is abundant, despite the
International Nuclear Information System (INIS)
Kowarski, L.
1955-01-01
It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing
Energy Technology Data Exchange (ETDEWEB)
Santos, Rubens Souza dos [Instituto de Engenharia Nuclear (IEN), Rio de Janeiro, RJ (Brazil); Martinez, Aquilino Senra; Alvim, Antonio Carlos Marques [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear
2002-07-01
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)
Directory of Open Access Journals (Sweden)
Valeria Reginatto
2008-01-01
Full Text Available Neste trabalho, um reator em escala laboratorial de lodo ativado, aplicado ao processo da nitrificação, foi acompanhado por meio de ensaios cinéticos de atividade específica. A atividade de nitrificação da biomassa foi determinada por respirometria nacaracterização do inóculo e na avaliação da biomassa do reator em duas condições: durante a alimentação do reator com meio sintético autotrófico; e após a sua alimentação com efluente de um reator UASB, utilizado para desnitrificação. O reator atingiu uma eficiência em torno de 90% de nitrificação em ambas as condições de operação. O modelo cinético de Andrews, que inclui uma constante da inibição pelo substrato (Ki, ajustou-se melhor aos resultados obtidos nos testes de atividade do que o de Monod. Entretanto, observou-se aumento daconstante de inibição (Ki do lodo após operação do reator em relação ao inóculo, demonstrando a adaptação da biomassa às novas condições (cargas de nitrificação.In this work, an activated sludge lab-scale reactor used fornitrification was monitored by specific activity kinetic assays. The nitrification biomass activity was carried out by respirometric methods in order to characterize the inoculum and the reactor sludge after two different operation conditions: during the feeding of the reactor with synthetic autotrophic medium, and after feeding it with effluent from an UASB reactor used for denitrification. The efficiency of nitrification reached 90% in both operation conditions. Results obtained by the kinetic activity assays were better adjusted by the kinetic model of Andrews, which includes the inhibition constant by the substrate (Ki, than the Monod model. However, an increase was observed in the inhibition constant (Ki of the sludge after the operation of the reactor as compared with the inoculum. This effect demonstrates an adaptation of the biomass to the new nitrification conditions (loading rate.
International Nuclear Information System (INIS)
Shakeri Yekta, Sepehr; Lindmark, Amanda; Skyllberg, Ulf; Danielsson, Åsa; Svensson, Bo H.
2014-01-01
Highlights: • Thermodynamics and kinetics of Fe, Co and Ni added to biogas reactors were studied. • Formation of Fe-sulfide and Fe-thiol aqueous complexes controlled the Fe solubility. • Cobalt solubility was controlled by processes independent of Co-sulfide interaction. • Iron added to the biogas reactors effected the Ni speciation and solubility. - Abstract: The objective of the present study was to assess major chemical reactions and chemical forms contributing to solubility and speciation of Fe(II), Co(II), and Ni(II) during anaerobic digestion of sulfur (S)-rich stillage in semi-continuous stirred tank biogas reactors (SCSTR). These metals are essential supplements for efficient and stable performance of stillage-fed SCSTR. In particular, the influence of reduced inorganic and organic S species on kinetics and thermodynamics of the metals and their partitioning between aqueous and solid phases were investigated. Solid phase S speciation was determined by use of S K-edge X-ray absorption near-edge spectroscopy. Results demonstrated that the solubility and speciation of supplemented Fe were controlled by precipitation of FeS(s) and formation of the aqueous complexes of Fe-sulfide and Fe-thiol. The relatively high solubility of Co (∼20% of total Co content) was attributed to the formation of compounds other than Co-sulfide and Co-thiol, presumably of microbial origin. Nickel had lower solubility than Co and its speciation was regulated by interactions with FeS(s) (e.g. co-precipitation, adsorption, and ion substitution) in addition to precipitation/dissolution of discrete NiS(s) phase and formation of aqueous Ni-sulfide complexes
Energy Technology Data Exchange (ETDEWEB)
Shakeri Yekta, Sepehr, E-mail: sepehr.shakeri.yekta@liu.se [Department of Thematic Studies – Water and Environmental Studies, Linköping University, SE-581 83 Linköping (Sweden); Lindmark, Amanda [Department of Thematic Studies – Water and Environmental Studies, Linköping University, SE-581 83 Linköping (Sweden); Skyllberg, Ulf [Department of Forest Ecology and Management, Swedish University of Agricultural Sciences, SE-901 83 Umeå (Sweden); Danielsson, Åsa; Svensson, Bo H. [Department of Thematic Studies – Water and Environmental Studies, Linköping University, SE-581 83 Linköping (Sweden)
2014-03-01
Highlights: • Thermodynamics and kinetics of Fe, Co and Ni added to biogas reactors were studied. • Formation of Fe-sulfide and Fe-thiol aqueous complexes controlled the Fe solubility. • Cobalt solubility was controlled by processes independent of Co-sulfide interaction. • Iron added to the biogas reactors effected the Ni speciation and solubility. - Abstract: The objective of the present study was to assess major chemical reactions and chemical forms contributing to solubility and speciation of Fe(II), Co(II), and Ni(II) during anaerobic digestion of sulfur (S)-rich stillage in semi-continuous stirred tank biogas reactors (SCSTR). These metals are essential supplements for efficient and stable performance of stillage-fed SCSTR. In particular, the influence of reduced inorganic and organic S species on kinetics and thermodynamics of the metals and their partitioning between aqueous and solid phases were investigated. Solid phase S speciation was determined by use of S K-edge X-ray absorption near-edge spectroscopy. Results demonstrated that the solubility and speciation of supplemented Fe were controlled by precipitation of FeS(s) and formation of the aqueous complexes of Fe-sulfide and Fe-thiol. The relatively high solubility of Co (∼20% of total Co content) was attributed to the formation of compounds other than Co-sulfide and Co-thiol, presumably of microbial origin. Nickel had lower solubility than Co and its speciation was regulated by interactions with FeS(s) (e.g. co-precipitation, adsorption, and ion substitution) in addition to precipitation/dissolution of discrete NiS(s) phase and formation of aqueous Ni-sulfide complexes.
International Nuclear Information System (INIS)
Zhang, S.
2012-01-01
Within the framework of the research in the nuclear reactor safety field, the iodine oxides formation by organic iodides destruction in the containment has been studied with the means of the atmospheric chemistry field. The destruction kinetics and their activation energy of organic iodides by . OH and . O radical has been quantified by a Flash Photolysis system able to monitor the oxidant radicals by resonance fluorescence. Those results have been published and some of them for the first time in the literature. The mechanisms leading to the organic iodides destruction are either by a hydrogen atom abstraction, either by the formation of a complex, depending on the organic iodide involved. Then, certain kinetics reactions have been updated in the IODAIR code. Other reactions have been added based on the recent literature available. A comparison of the kinetics destruction of CH 3 I by . OH and . O with IODAIR and the global kinetics of destruction in ASTEC/IODE showed a difference of about 2 which shows the importance of these two radicals (and mainly . O) in those destruction processes. The other main path of destruction would be by electron radiation. Other radicals like . H and . N would not contribute significantly to organic iodides destruction. A sensitivity analysis highlighted that organic iodides would mostly be destroyed into iodine oxides with a almost complete conversion within a few hours. Finally, an atmospheric chamber has been used to quantify iodine oxides growth, density and composition. Under the conditions studied, their formation is fast. Particles sizes of about 200-400 nm are formed within a few hours. The main parameters influencing their growth are the relative humidity and the presence of ozone (whose function is to create . O and . OH radicals). (author)
Trojanowicz, Karol; Wójcik, Włodzimierz
2011-01-01
The article presents a case-study on the calibration and verification of mathematical models of organic carbon removal kinetics in biofilm. The chosen Harremöes and Wanner & Reichert models were calibrated with a set of model parameters obtained both during dedicated studies conducted at pilot- and lab-scales for petrochemical wastewater conditions and from the literature. Next, the models were successfully verified through studies carried out utilizing a pilot ASFBBR type bioreactor installed in an oil-refinery wastewater treatment plant. During verification the pilot biofilm reactor worked under varying surface organic loading rates (SOL), dissolved oxygen concentrations and temperatures. The verification proved that the models can be applied in practice to petrochemical wastewater treatment engineering for e.g. biofilm bioreactor dimensioning.
International Nuclear Information System (INIS)
Miyano, Hiroshi; Narabayashi, Naoshi.
1990-01-01
The represent invention concerns a reactor system with improved water injection means to a pressure vessel of a BWR type reactor. A steam pump is connected to a heat removing system pipeline, a high pressure water injection system pipeline and a low pressure water injection system pipeline for injecting water into the pressure vessel. A pump actuation pipeline is disposed being branched from a main steam pump or a steam relieaf pipeline system, through which steams are supplied to actuate the steam pump and supply cooling water into the pressure vessel thereby cooling the reactor core. The steam pump converts the heat energy into the kinetic energy and elevates the pressure of water to a level higher than the pressure of the steams supplied by way of a pressure-elevating diffuser. Cooling water can be supplied to the pressure vessel by the pressure elevation. This can surely inject cooling water into the pressure vessel upon loss of coolant accident or in a case if reactor scram is necessary, without using an additional power source. (I.N.)
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)
Energy Technology Data Exchange (ETDEWEB)
Muhammad, Farhan [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, P.O. Nilore, Islamabad 45650 (Pakistan)], E-mail: mfarhan_73@yahoo.co.uk; Majid, Asad [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, P.O. Nilore, Islamabad 45650 (Pakistan)
2008-09-15
The effects of using high density low enriched uranium on the neutronic parameters of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density LEU fuels currently being developed under the RERTR program. Since the alloying elements have different cross-sections affecting the reactor in different ways, therefore fuels U-Mo (9 w/o) which contain the same elements in same ratio were selected for analysis. Simulations were carried out to calculate core excess reactivity, neutron flux spectrum, prompt neutron generation time, effective delayed neutron fraction and feedback coefficients including Doppler feedback coefficient, and reactivity coefficients for change of water density and temperature. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the excess reactivity at the beginning of life does not increase as the uranium density of fuel. Both the prompt neutron generation time and the effective delayed neutron fraction decrease as the uranium density increases. The absolute value of Doppler feedback coefficient increases while the absolute values of reactivity coefficients for change of water density and temperature decrease.
International Nuclear Information System (INIS)
Muhammad, Farhan; Majid, Asad
2008-01-01
The effects of using high density low enriched uranium on the neutronic parameters of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density LEU fuels currently being developed under the RERTR program. Since the alloying elements have different cross-sections affecting the reactor in different ways, therefore fuels U-Mo (9 w/o) which contain the same elements in same ratio were selected for analysis. Simulations were carried out to calculate core excess reactivity, neutron flux spectrum, prompt neutron generation time, effective delayed neutron fraction and feedback coefficients including Doppler feedback coefficient, and reactivity coefficients for change of water density and temperature. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the excess reactivity at the beginning of life does not increase as the uranium density of fuel. Both the prompt neutron generation time and the effective delayed neutron fraction decrease as the uranium density increases. The absolute value of Doppler feedback coefficient increases while the absolute values of reactivity coefficients for change of water density and temperature decrease
International Nuclear Information System (INIS)
Bricmont, R.J.; Hamilton, P.A.; Ming Long Ting, R.
1981-01-01
Reactors, fuel processing plants etc incorporate pipes and conduits for fluids under high pressure. Fractures, particularly adjacent to conduit elbows, produce a jet of liquid which whips the broken conduit at an extremely high velocity. An enormous impact load would be applied to any stationary object in the conduit's path. The design of cellular, corrugated metal impact pads to absorb the kinetic energy of the high velocity conduits is given. (U.K.)
DEFF Research Database (Denmark)
van Leeuwen, Theo; Djonov, Emilia
2014-01-01
After discussing broad cultural drivers behind the development of kinetic typography, the chapter outlines an approach to analysing kinetic typography which is based on Halliday's theory of transitivity, as applied by Kress and Van Leeuwen to visual images.......After discussing broad cultural drivers behind the development of kinetic typography, the chapter outlines an approach to analysing kinetic typography which is based on Halliday's theory of transitivity, as applied by Kress and Van Leeuwen to visual images....
Energy Technology Data Exchange (ETDEWEB)
Seery, D.J.; Freihaut, J.D.; Proscia, W.M. (United Technologies Research Center, East Hartford, CT (USA)); Howard, J.B.; Peters, W.; Hsu, J.; Hajaligol, M.; Sarofim, A. (Massachusetts Inst. of Tech., Cambridge, MA (USA)); Jenkins, R.; Mallin, J.; Espindola-Merin, B. (Pennsylvania State Univ., University Park, PA (USA)); Essenhigh, R.; Misra, M.K. (Ohio State Univ., Columbus, OH (USA))
1989-07-01
This report contains results of a coordinated, multi-laboratory investigation of coal devolatilization. Data is reported pertaining to the devolatilization for bituminous coals over three orders of magnitude in apparent heating rate (100 to 100,000 + {degree}C/sec), over two orders of magnitude in particle size (20 to 700 microns), final particle temperatures from 400 to 1600{degree}C, heat transfer modes ranging from convection to radiative, ambient pressure ranging from near vacuum to one atmosphere pressure. The heat transfer characteristics of the reactors are reported in detail. It is assumed the experimental results are to form the basis of a devolatilization data base. Empirical rate expressions are developed for each phase of devolatilization which, when coupled to an awareness of the heat transfer rate potential of a particular devolatilization reactor, indicate the kinetics emphasized by a particular system reactor plus coal sample. The analysis indicates the particular phase of devolatilization that will be emphasized by a particular reactor type and, thereby, the kinetic expressions appropriate to that devolatilization system. Engineering rate expressions are developed from the empirical rate expressions in the context of a fundamental understanding of coal devolatilization developed in the course of the investigation. 164 refs., 223 figs., 44 tabs.
An internally illuminated monolith reactor: Pros and cons relative to a slurry reactor
Carneiro, Joana T.; Carneiro, J.T.; Berger, Rob; Moulijn, Jacob A.; Mul, Guido
2009-01-01
In the present study, kinetic models for the photo-oxidation of cyclohexane in two different photoreactor systems are discussed: a top illumination reactor (TIR) representative of a slurry reactor, and the so-called internally illuminated monolith reactor (IIMR) representing a reactor containing
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.
Energy Technology Data Exchange (ETDEWEB)
Bruna, J G; Brunet, J P; Clouet D' Orval, Ch; Caizergues, R; Verriere, Ph [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires
1964-07-01
The {alpha} decay constant of prompt neutrons has been studied in the homogeneous plutonium-fueled, light-water-moderated reactor Alecto, by the probability method. In this method, the probability to count one, two,.... neutrons during a given time is measured. The value of {alpha} can be deduced from this measurement, for various subcritical states of the reactor. The experimental results were then compared with values obtained, for the same reactivities, by the pulsed neutron technique. (authors) [French] On a etudie sur Alecto, reacteur homogene au plutonium, modere a l'eau legere, la constante de decroissance {alpha} des neutrons prompts par la methode des probabilites. Celle-ci consiste a mesurer la probabilite de compter un, deux, etc..., neutrons pendant un intervalle de temps donne. On a pu en deduire la valeur de {alpha}, dans divers etats sous-critiques du reacteur. On a compare les resultats experimentaux a d'autres valeurs obtenues, aux memes reactivites, par la methode des neutrons pulses. (auteurs)
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
Reactor physics and thermodynamics of a gaseous core fission reactor
International Nuclear Information System (INIS)
Kuijper, J.C.; Van Dam, H.; Stekelenburg, A.J.C.; Hoogenboom, J.E.; Boersma-Klein, W.; Kistemaker, J.
1990-01-01
Neutron kinetics and thermodynamics of a Gaseous Core Fission Reactor with magnetical pumping are shown to have many unconventional aspects. Attention is focussed on the properties of the fuel gas, the stationary temperature distribution, the non-linear neutron kinetics and the energy balance in thermodynamical cycles
Determination of the kinetics of ethene epoxidation
Schouten, E.P.S.; Schouten, E.P.S.; Borman, P.C.; Borman, P.C.; Westerterp, K.R.
1996-01-01
Several problems and pitfalls in the use of laboratory reactors for the determination of the kinetics of ethene epoxidation over industrial silver on α-alumina catalyst are discussed. Also, commonly used methodologies for kinetic studies are dealt with because of the general nature of some problems.
Kinetic mechanism of the decomposition of dimethyltin dichloride
Mol, van A.M.B.; Croon, de M.H.J.M.; Spee, C.I.M.A.; Schouten, J.C.
1999-01-01
Results are reported of a study of the intrinsic kinetics of gas phase reactions. For this purpose a reactor system is designed in such a way that concentration and temperature variations throughout the reactor can be neglected enabling investigation of intrinsic reaction kinetics. The gas phase
Shi, Wenying; He, Benqiao; Cao, Yuping; Li, Jianxin; Yan, Feng; Cui, Zhenyu; Zou, Zhiqun; Guo, Shiwei; Qian, Xiaomin
2013-02-01
A novel composite catalytic membrane (CCM) was prepared from sulfonated polyethersulfone (SPES) and polyethersulfone (PES) blend supported by non-woven fabrics, as a heterogeneous catalyst to produce biodiesel from continuous esterification of oleic acid with methanol in a flow-through mode. A kinetic model of esterification was established based on a plug-flow assumption. The effects of the CCM structure (thickness, area, porosity, etc.), reaction temperature and the external and internal mass transfer resistances on esterification were investigated. The results showed that the CCM structure had a significant effect on the acid conversion. The external mass transfer resistance could be neglected when the flow rate was over 1.2 ml min(-1). The internal mass transfer resistance impacted on the conversion when membrane thickness was over 1.779 mm. An oleic acid conversion kept over 98.0% for 500 h of continuous running. The conversions obtained from the model are in good agreement with the experimental data. Copyright © 2012 Elsevier Ltd. All rights reserved.
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)
Directory of Open Access Journals (Sweden)
Ferschneider G.
2006-11-01
Full Text Available Fixed bed reactors with a single fluid phase are widely used in the refining or petrochemical industries for reaction processes catalysed by a solid phase. The design criteria for industrial reactors are relatively well known. However, they rely on a one-dimensional writing and on the separate resolution of the equation of conservation of mass and energy, and of momentum. Thus, with complex geometries, the influence of hydrodynamics on the effectiveness of the catalyst bed cannot be taken into account. The calculation method proposed is based on the multi-dimensional writing and the simultaneous resolution of the local conservation equations. The example discussed concerns fixed-bed catalytic reactors. These reactors are distinguished by their annular geometry and the radial circulation of the feedstock. The flow is assumed to be axisymmetric. The reaction process is reflected by a simplified kinetic mechanism involving ten chemical species. Calculation of the hydrodynamic (mean velocities, pressure, thermal and mass fields (concentration of each species serves to identify the influence of internal components in two industrial reactor geometries. The map of the quantity of coke formed and deposited on the catalyst, calculated by the model, reveals potential areas of poor operation. Les réacteurs à lit fixe avec une seule phase fluide sont largement utilisés dans l'industrie du raffinage et de la pétrochimie, pour mettre en oeuvre un processus réactionnel catalysé par une phase solide. Les règles de conception des réacteurs industriels sont relativement bien connues. Cependant, elles reposent sur l'écriture monodimensionnelle et la résolution séparée, d'une part, des équations de conservation de la masse et de l'énergie et d'autre part, de la quantité de mouvement. Ainsi dans le cas de géométries complexes, l'influence de l'hydrodynamique sur l'efficacité du lit catalytique ne peut être prise en compte. La méthode de calcul
International Nuclear Information System (INIS)
Lifschitz, E.M.; Pitajewski, L.P.
1983-01-01
The textbook covers the subject under the following headings: kinetic gas theory, diffusion approximation, collisionless plasma, collisions within the plasma, plasma in the magnetic field, theory of instabilities, dielectrics, quantum fluids, metals, diagram technique for nonequilibrium systems, superconductors, and kinetics of phase transformations
Shakeri Yekta, Sepehr; Lindmark, Amanda; Skyllberg, Ulf; Danielsson, Asa; Svensson, Bo H
2014-03-30
The objective of the present study was to assess major chemical reactions and chemical forms contributing to solubility and speciation of Fe(II), Co(II), and Ni(II) during anaerobic digestion of sulfur (S)-rich stillage in semi-continuous stirred tank biogas reactors (SCSTR). These metals are essential supplements for efficient and stable performance of stillage-fed SCSTR. In particular, the influence of reduced inorganic and organic S species on kinetics and thermodynamics of the metals and their partitioning between aqueous and solid phases were investigated. Solid phase S speciation was determined by use of S K-edge X-ray absorption near-edge spectroscopy. Results demonstrated that the solubility and speciation of supplemented Fe were controlled by precipitation of FeS(s) and formation of the aqueous complexes of Fe-sulfide and Fe-thiol. The relatively high solubility of Co (∼ 20% of total Co content) was attributed to the formation of compounds other than Co-sulfide and Co-thiol, presumably of microbial origin. Nickel had lower solubility than Co and its speciation was regulated by interactions with FeS(s) (e.g. co-precipitation, adsorption, and ion substitution) in addition to precipitation/dissolution of discrete NiS(s) phase and formation of aqueous Ni-sulfide complexes. Copyright © 2014 Elsevier B.V. All rights reserved.
Energy Technology Data Exchange (ETDEWEB)
Cau Dit Coumes, C; Devisme, F [Commissariat a l` Energie Atomique, CE/VRH, Bagnols-sur-Ceze (France); Vargas, S; Chopin-Dumas, J [Laboratoire d` Electrochimie Inorganique, ENSSPICAM, Marseille (France)
1996-12-01
Iodine, which can be released inside the containment building when an accident occurs, can be traced, in normal operating conditions, at the back end of the fuel cycle. Hydroxylamine has been selected as a reagent of potential interest to trap iodine in the dissolution off gas treatment. The kinetics of the reaction between hydroxylamine and iodine has been studied in a narrow range of pH (1-2), with hydroxylamine in excess (ratios of hydroxylamine to iodine initial concentrations varying from 2 to 40), at constant temperature (30{sup o}C) and ionic strength (0.1 mol/L). Spectrophotometry and voltametry have been coupled for analytical investigation. The problem of the rapid mixing of the reactants has been solved using a continuous reactor. Triiodide has been shown non reactive towards hydroxylamine. An initial rate law has been proposed, pointing out the first order of the reaction with respect to hydroxylamine and iodine, and the inhibitory effect of iodide and hydrogen ions. Nitrous acid has been identified as a transitory product. Nitrous oxide and nitrogen monoxide have been detected by gas chromatography, the ratio of the amounts of products formed depending on acidity. The complexity of the overall reaction has been ascribed to the competition of four reactions as previously proposed in the literature. (author) 8 figs., 1 tab., 13 refs.
International Nuclear Information System (INIS)
Cau Dit Coumes, C.; Devisme, F.; Vargas, S.; Chopin-Dumas, J.
1996-01-01
Iodine, which can be released inside the containment building when an accident occurs, can be traced, in normal operating conditions, at the back end of the fuel cycle. Hydroxylamine has been selected as a reagent of potential interest to trap iodine in the dissolution off gas treatment. The kinetics of the reaction between hydroxylamine and iodine has been studied in a narrow range of pH (1-2), with hydroxylamine in excess (ratios of hydroxylamine to iodine initial concentrations varying from 2 to 40), at constant temperature (30 o C) and ionic strength (0.1 mol/L). Spectrophotometry and voltametry have been coupled for analytical investigation. The problem of the rapid mixing of the reactants has been solved using a continuous reactor. Triiodide has been shown non reactive towards hydroxylamine. An initial rate law has been proposed, pointing out the first order of the reaction with respect to hydroxylamine and iodine, and the inhibitory effect of iodide and hydrogen ions. Nitrous acid has been identified as a transitory product. Nitrous oxide and nitrogen monoxide have been detected by gas chromatography, the ratio of the amounts of products formed depending on acidity. The complexity of the overall reaction has been ascribed to the competition of four reactions as previously proposed in the literature. (author) 8 figs., 1 tab., 13 refs
Energy Technology Data Exchange (ETDEWEB)
Caillet, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires
1961-07-01
The author reviews precisely the analogical techniques used for the resolution of the kinetic equations of nuclear reactors. Prior to this, he recalls the reasons which oblige physicians and engineers, even today, to use electronic machines in this domain. The author then considers the technological problems posed by the range of values which the various nuclear parameters adopt. In each case, he shows that a compromise is possible allowing an optimum precision. He compares the results to those obtained by arithmetic calculation and uses the examples chosen in a critical analysis of the present possibilities of the two methods of calculation. (author) [French] L'auteur cherche a faire un point aussi exact que possible des techniques analogiques utilisees pour resoudre les equations cinetiques des reacteurs nucleaires. Il rappelle auparavant les raisons pour lesquelles physiciens et ingenieurs sont obliges, encore aujourd'hui, de faire appel aux machines electroniques dans ce domaine. Puis il etudie les problemes technologiques que souleve le champ des valeurs prises par les differents parametres nucleaires. Dans chacun des cas, il montre l'existence d'un compromis qui permet d'atteindre une precision optimum. Il compare les resultats obtenus a ceux provenant de calculateurs arithmetiques et profite des exemples choisis pour faire une analyse critique des possibilites actuelles offertes par les deux modes de calcul. (auteur)
International Nuclear Information System (INIS)
Hoffman, A. J.; Lee, J. C.
2013-01-01
A new time-dependent neutron transport method based on the method of characteristics (MOC) has been developed. Whereas most spatial kinetics methods treat time dependence through temporal discretization, this new method treats time dependence by defining the characteristics to span space and time. In this implementation regions are defined in space-time where the thickness of the region in time fulfills an analogous role to the time step in discretized methods. The time dependence of the local source is approximated using a truncated Taylor series expansion with high order derivatives approximated using backward differences, permitting the solution of the resulting space-time characteristic equation. To avoid a drastic increase in computational expense and memory requirements due to solving many discrete characteristics in the space-time planes, the temporal variation of the boundary source is similarly approximated. This allows the characteristics in the space-time plane to be represented analytically rather than discretely, resulting in an algorithm comparable in implementation and expense to one that arises from conventional time integration techniques. Furthermore, by defining the boundary flux time derivative in terms of the preceding local source time derivative and boundary flux time derivative, the need to store angularly-dependent data is avoided without approximating the angular dependence of the angular flux time derivative. The accuracy of this method is assessed through implementation in the neutron transport code DeCART. The method is employed with variable-order local source representation to model a TWIGL transient. The results demonstrate that this method is accurate and more efficient than the discretized method. (authors)
International Nuclear Information System (INIS)
Swart, C.A.M. de.
1983-01-01
The author has studied the kinetics of heparin and heparin fractions after intravenous administration in humans and in this thesis the results of this study are reported. Basic knowledge about the physico-chemical properties of heparin and its interactions with proteins resulting in anticoagulant and lipolytic effects are discussed in a review (chapter II), which also comprises some clinical aspects of heparin therapy. In chapter III the kinetics of the anticoagulant effect are described after intravenous administration of five commercial heparin preparations. A mathematical model is presented that fits best to these kinetics. The kinetics of the anticoagulant and lipolytic effects after intravenous injection of various 35 S-radiolabelled heparin fractions and their relationship with the disappearance of the radiolabel are described in chapter IV. Chapter V gives a description of the kinetics of two radiolabels after injection of in vitro formed complexes consisting of purified, 125 I-radiolabelled antithrombin III and various 35 S-radiolabelled heparin fractions. (Auth.)
Energy Technology Data Exchange (ETDEWEB)
Harouiya, N.; Chairat, C.; Kohler, S.J.; Gout, R.; Oelkers, E.H. [Univ Toulouse 3, CNRS, UMR 5563, F-31400 Toulouse (France); Chairat, C. [CEA, LCLT SECM DTCD, Lab Etud Comportement Long Terme, F-30207 Bagnols Sur Ceze, (France)
2007-07-01
The apparent solubility and dissolution rates of natural apatite were measured in closed-system reactors as a function of temperature from 5 to 50 degrees C and pH from 1 to 6. The temporal release rates of Ca, P, and F during the experiments are approximately consistent with stoichiometric dissolution in all experiments. One advantage of closed-system experiments is that they allow determination of reactive fluid evolution and dissolution rates at far-from to near-to equilibrium conditions. Surface area normalized apatite dissolution rates, r, obtained in all experiments are consistent with r = A{sub A}a{sub H{sup +}}{sup n}exp(E{sub A}/RT)(1 -exp(-A/{sigma} RT)) where A{sub A} stands for a rate constant equal to 4 * 10{sup -3} mol/cm{sup 2}/s, a{sub H{sup +}}) denotes the activity of the aqueous H{sup +}, n designates a reaction order equal to 0.6, E{sub A} symbolizes an activation energy equal to 11.0 kcal/mol, A refers to the chemical affinity of the dissolving apatite, {sigma} stands for Temkin's average stoichiometric number equal to 5; R designates the gas constant, and T represents absolute temperature. Logarithms of apparent equilibrium constants obtained from experiments performed at 3 {<=} pH {<=} 5.6 for the apatite dissolution reaction: Ca{sub 5}(PO{sub 4}){sub 3}F + 3H{sup +} = 5Ca{sup 2+} + 3HPO{sub 4}{sup 2-} + F{sup -} are found to be - 29.5 {+-} 0.6, - 29.4 {+-} 0.9 and - 29.9 {+-} 1.3 at 5, 25, and 50 degrees C, respectively. (authors)
International Nuclear Information System (INIS)
1966-01-01
This Volume 1 of the Proceedings of 2. Yugoslav symposium on reactor physics includes nine papers dealing with the following topics: reactor kinetics, reactor noise, neutron detection, methods for calculating neutron flux spatial and time dependence in the reactor cores of both heavy and light water moderated experimental reactors, calculation of reactor lattice parameters, reactor instrumentation, reactor monitoring systems; measuring methods of reactor parameters; reactor experimental facilities
International Nuclear Information System (INIS)
Gicquel, L.
2010-01-01
This PhD study consisted in studying reactive mechanisms and kinetics of sodium-CO 2 interactions, in the frame of the assessment of an energy conversion system with supercritical CO 2 for fast breeder reactors cooled by sodium. The approach was the following. First of all, the interactions between sodium and CO 2 have been brought to light by laboratory experiments associated with products analysis. They have enabled the establishment of a coherent mechanism, in agreement with literature data, and gave preliminary indications on the reaction kinetics. In order to estimate a more detailed reaction kinetics, we tried to approach the phenomenon that appears in the case of a leak in a sodium-CO 2 heat exchanger. Geometry of such heat exchangers is not fixed for the moment, even if the development of compact exchangers is foreseen. Then, free jets of CO 2 in liquid sodium have been modeled in order to obtain, by identification, kinetics parameters of the reaction. Those parameters, estimated with such a geometry, will remain valid with a much complex geometry, that will better represent the real exchanger. An experimental bench has been defined and built to realize those jets. The first laboratory experiments have concluded in the existence of different reactive mechanisms according to the temperature level. A threshold has been brought to light around 500 C. Below this one, reaction appears moderated, or even, slow, with a medium exothermicity, and appears after an induction period that depends on the temperature,and which duration could reach several hours. At contrary, above this threshold, it seems rapid and more exothermic. Below 500 C, sodium oxalate is produced, and then reacts with sodium in an exothermic way, following the reactions: CO 2 + Na →1/4 Na 2 C 2 O 4 + 1/4 CO + 1/4 Na 2 CO 3 (5) 4 Na + Na 2 C 2 O 4 → 3 Na 2 O + CO + C (6) Above 500 C, sodium carbonate is produced, and can then possibly react with sodium in an endothermic way, following the
DEFF Research Database (Denmark)
2009-01-01
A kinetic interface for orientation detection in a video training system is disclosed. The interface includes a balance platform instrumented with inertial motion sensors. The interface engages a participant's sense of balance in training exercises.......A kinetic interface for orientation detection in a video training system is disclosed. The interface includes a balance platform instrumented with inertial motion sensors. The interface engages a participant's sense of balance in training exercises....
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.
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)
International Nuclear Information System (INIS)
Barre, Bertrand
2015-10-01
After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor
Advanced spheromak fusion reactor
International Nuclear Information System (INIS)
Fowler, T.K.
1996-01-01
The spheromak has no toroidal magnetic field coils or other structure along its geometric axis, and is thus more attractive than the leading magnetic fusion reactor concept, the tokamak. As a consequence of this and other attributes, the spheromak reactor may be compact and produce a power density sufficiently high to warrant consideration of a liquid 'blanket' that breeds tritium, converts neutron kinetic energy to heat, and protects the reactor vessel from severe neutron damage. However, the physics is more complex, so that considerable research is required to learn how to achieve the reactor potential. Critical physics problems and possible ways of solving them are described. The opportunities and issues associated with a possible liquid wall are considered to direct future research
Directory of Open Access Journals (Sweden)
Mojtaba Ahmadi
2016-11-01
Full Text Available The aqueous degradation of Reactive Yellow 84 (RY84 by potassium peroxydisulfate (K2S2O8 has been studied in laboratory scale experiments. The effect of the initial concentrations of potassium peroxydisulfate and RY84, pH and temperature on RY84 degradation were also examined. Experimental data were analyzed using first and second-order kinetics. The degradation kinetics of RY84 of the potassium peroxydisulfate process followed the second-order reaction kinetics. These rate constants have an extreme values similar to of 9.493 mM−1min−1 at a peroxydisulfate dose of 4 mmol/L. Thermodynamic parameters such as activation (Ea and Gibbs free energy (ΔG° were also evaluated. The negative value of ΔGo and Ea shows the spontaneous reaction natural conditions and exothermic nature.
International Nuclear Information System (INIS)
Bussac, J.; Reuss, P.
1985-01-01
This book presents the main physical bases of neutron theory and nuclear reactor calculation. 1) Interactions of neutrons with matter and basic principles of neutron transport; 2) Neutron transport in homogeneous medium and the neutron field: kinetic behaviour, slowing-down, resonance absorption, diffusion equation, processing methods; 3) Theory of a reactor constituted with homogeneous zones: critical condition, kinetics, separation of variables, calculation and neutron balance of the fundamental mode, one-group and multigroup theories; 4) Study of heterogeneous cell lattices: fast fission factor, resonance absorption, thermal output factor, diffusion coefficient, computer codes; 5) Operation and control of reactors: perturbation theory, reactivity, fuel properties evolution, poisoning by fission products, calculation of a reactor and fuel management; 6) Study of some types of reactors: PWR and fast breeder reactors, the main reactor types of the present French program [fr
International Nuclear Information System (INIS)
Peters, A.M.; Lavender, J.P.; Saverymuttu, S.H.
1985-01-01
By using density gradient materials enriched with autologous plasma, the authors have been able to isolate granulocutes from other cellular elements and label them with In-111 without separation from a plasma environment. The kinetic behavior of these cells suggests that phenomena attributed to granulocyte activation are greatly reduced by this labeling. Here, they review their study of granulocyte kinetics in health and disease in hope of quantifying sites of margination and identifying principal sites of destruction. The three principle headings of the paper are distribution, life-span, and destruction
Calculation of research reactor RA power at uncontrolled reactivity changes
International Nuclear Information System (INIS)
Cupac, S.
1978-01-01
The safety analysis of research reactor RA involves also the calculation of reactor power at uncontrolled reactivity changes. The corresponding computer code, based on Point Kinetics Model has been made. The short review of method applied for solving kinetic equations is given and several examples illustrating the reactor behaviour at various reactivity changes are presented. The results already obtained are giving rather rough picture of reactor behaviour in considered situations. This is the consequence of using simplified feed back and reactor cooling models, as well as temperature reactivity coefficients, which do not correspond to the actual reactor RA structure (which is now only partly fulfilled with 80% enriched uranium fuel). (author) [sr
Kinetics of cadmium hydroxide precipitation
International Nuclear Information System (INIS)
Patterson, J.W.; Marani, D.; Luo, B.; Swenson, P.
1987-01-01
This paper presents some preliminary results on the kinetics of Cd(OH)/sub 2/ precipitation, both in the absence and the presence of citric acid as an inhibiting agent. Batch and continuous stirred tank reactor (CSTR) precipitation studies are performed by mixing equal volumes of NaOH and Cd(NO/sub 3/)/sub 2/ solutions, in order to avoid localized supersaturation conditions. The rate of metal removal from the soluble phase is calculated from the mass balance for the CSTR precipitation tests. In addition, precipitation kinetics are studied in terms of nucleation and crystal growth rates, by means of a particle counter that allows a population balance analysis for the precipitation reactor at steady state conditions
Production kinetics of zirconium tetrachloride
International Nuclear Information System (INIS)
Sudjoko, D.; Masduki, B.; Sunardjo; Sulistyo, B.
1996-01-01
This research was intended to study the kinetics of zirconium tetrachloride production. The process was carried out in semi continuous reactor, equipped with heater, temperature controller, sublimator and scrubber. The variables investigated were time, temperature and the pellet forming pressure. Within the range of variables studied, the expression of the process in the chemical reaction controller region and diffusion controller region were both presented. (author)
Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...
Simulation development for TRIGA reactor
International Nuclear Information System (INIS)
Handoyo, D.
1997-01-01
A simulator of the dynamic of TRIGA reactor has been made. this simulator is meant to study the reactor kinetic behavior and for operator training to more assure the safety and the reliability of the real operation of TRIGA reactor. the simulator consists of PC (Personal Computer) for processing the calculation of reactivity, neutron flux, period, ect and control panel for regulating the input data such as the change of power range, control rod position as well as cooling flow rate. the result will be displayed on screen monitor of personal computer as given in the real control room of TRIGA reactor. the output of simulator will be verified by comparing with measurement result in the real TRIGA MARK II reactor of Musashi institute of technology. for the change of reactivity of 0.3, 0.5 and 0.7 the reactor power and fuel temperature between the simulator and measurements are comparable
WWER reactor physics code applications
International Nuclear Information System (INIS)
Gado, J.; Kereszturi, A.; Gacs, A.; Telbisz, M.
1994-01-01
The coupled steady-state reactor physics and thermohydraulic code system KARATE has been developed and applied for WWER-1000 and WWER-440 operational calculations. The 3 D coupled kinetic code KIKO3D has been developed and validated for WWER-440 accident analysis applications. The coupled kinetic code SMARTA developed by VTT Helsinki has been applied for WWER-440 accident analysis. The paper gives a summary of the experience in code development and application. (authors). 10 refs., 2 tabs., 5 figs
International Nuclear Information System (INIS)
Moons, F.
1998-01-01
SCK-CEN's programme on fusion reactor materials includes studies (1) to investigate fracture mechanics of neutron-irradiated beryllium; (2) to describe the helium behaviour in irradiated beryllium at atomic scale; (3) to define the kinetics of beryllium reacting with air or steam; (3) to perform a feasibility study for the testing of integrated blanket modules under neutron irradiation. Progress and achievements in 1997 are reported
Thermal decomposition kinetics of ammonium uranyl carbonate
International Nuclear Information System (INIS)
Kim, E.H.; Park, J.J.; Park, J.H.; Chang, I.S.; Choi, C.S.; Kim, S.D.
1994-01-01
The thermal decomposition kinetics of AUC [ammonium uranyl carbonate; (NH 4 ) 4 UO 2 (CO 3 ) 3 [ in an isothermal thermogravimetric (TG) reactor under N 2 atmosphere has been determined. The kinetic data can be represented by the two-dimensional nucleation and growth model. The reaction rate increases and activation energy decreases with increasing particle size and precipitation time which appears in the particle size larger than 30 μm in the mechano-chemical phenomena. (orig.)
Heterogeneous gas core reactor
International Nuclear Information System (INIS)
Han, K.I.
1977-01-01
Preliminary investigations of a heterogeneous gas core reactor (HGCR) concept suggest that this potential power reactor offers distinct advantages over other existing or conceptual reactor power plants. One of the most favorable features of the HGCR is the flexibility of the power producing system which allows it to be efficiently designed to conform to a desired optimum condition without major conceptual changes. The arrangement of bundles of moderator/coolant channels in a fissionable gas or mixture of gases makes a truly heterogeneous nuclear reactor core. It is this full heterogeneity for a gas-fueled reactor core which accounts for the novelty of the heterogeneous gas core reactor concept and leads to noted significant advantages over previous gas core systems with respect to neutron and fuel economy, power density, and heat transfer characteristics. The purpose of this work is to provide an insight into the design, operating characteristics, and safety of a heterogeneous gas core reactor system. The studies consist mainly of neutronic, energetic and kinetic analyses of the power producing and conversion systems as a preliminary assessment of the heterogeneous gas core reactor concept and basic design. The results of the conducted research indicate a high potential for the heterogeneous gas core reactor system as an electrical power generating unit (either large or small), with an overall efficiency as high as 40 to 45%. The HGCR system is found to be stable and safe, under the conditions imposed upon the analyses conducted in this work, due to the inherent safety of ann expanding gaseous fuel and the intrinsic feedback effects of the gas and water coolant
Kreuzer, Hans Jürgen
1986-01-01
This monograph deals with the kinetics of adsorption and desorption of molecules physisorbed on solid surfaces. Although frequent and detailed reference is made to experiment, it is mainly concerned with the theory of the subject. In this, we have attempted to present a unified picture based on the master equation approach. Physisorption kinetics is by no means a closed and mature subject; rather, in writing this monograph we intended to survey a field very much in flux, to assess its achievements so far, and to give a reasonable basis from which further developments can take off. For this reason we have included many papers in the bibliography that are not referred to in the text but are of relevance to physisorption. To keep this monograph to a reasonable size, and also to allow for some unity in the presentation of the material, we had to omit a number of topics related to physisorption kinetics. We have not covered to any extent the equilibrium properties of physisorbed layers such as structures, phase tr...
Nuclear reactor engineering: Reactor design basics. Fourth edition, Volume One
International Nuclear Information System (INIS)
Glasstone, S.; Sesonske, A.
1994-01-01
This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in design and operation of nuclear power plants. Extensively updated, the fourth edition includes new material on reactor safety and risk analysis, regulation, fuel management, waste management, and operational aspects of nuclear power. This volume contains the following: energy from nuclear fission; nuclear reactions and radiations; neutron transport; nuclear design basics; nuclear reactor kinetics and control; radiation protection and shielding; and reactor materials
Reactor simulator development. Workshop material
International Nuclear Information System (INIS)
2001-01-01
The International Atomic Energy Agency (IAEA) has established a programme in nuclear reactor simulation computer programs to assist its Member States in education and training. The objective is to provide, for a variety of advanced reactor types, insight and practice in reactor operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the supply or development of simulation programs and training material, sponsors training courses and workshops, and distributes documentation and computer programs. This publication consists of course material for workshops on development of such reactor simulators. Participants in the workshops are provided with instruction and practice in the development of reactor simulation computer codes using a model development system that assembles integrated codes from a selection of pre-programmed and tested sub-components. This provides insight and understanding into the construction and assumptions of the codes that model the design and operational characteristics of various power reactor systems. The main objective is to demonstrate simple nuclear reactor dynamics with hands-on simulation experience. Using one of the modular development systems, CASSIM tm , a simple point kinetic reactor model is developed, followed by a model that simulates the Xenon/Iodine concentration on changes in reactor power. Lastly, an absorber and adjuster control rod, and a liquid zone model are developed to control reactivity. The built model is used to demonstrate reactor behavior in sub-critical, critical and supercritical states, and to observe the impact of malfunctions of various reactivity control mechanisms on reactor dynamics. Using a PHWR simulator, participants practice typical procedures for a reactor startup and approach to criticality. This workshop material consists of an introduction to systems used for developing reactor simulators, an overview of the dynamic simulation
International Nuclear Information System (INIS)
Colombino, A.; Mosiello, R.; Norelli, F.; Jorio, V.M.; Pacilio, N.
1975-01-01
A nuclear system kinetics is formulated according to a stochastic approach. The detailed probability balance equations are written for the probability of finding the mixed population of neutrons and detected neutrons, i.e. detectrons, at a given level for a given instant of time. Equations are integrated in search of a probability profile: a series of cases is analyzed through a progressive criterium. It tends to take into account an increasing number of physical processes within the chosen model. The most important contribution is that solutions interpret analytically experimental conditions of equilibrium (moise analysis) and non equilibrium (pulsed neutron measurements, source drop technique, start up procedures)
Activity report of Reactor Physics Division - 1993
International Nuclear Information System (INIS)
Indira, R.
1994-01-01
The research and development (R and D) activities of the Reactor Physics Division of Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam during 1993 are reported. The activities are arranged under the headings: Nuclear Data Processing and validation, Core Physics and Operation Studies, Reactor Kinetics and Safety analysis, Reactor Noise Analysis and Radiation Transport and Shielding Studies. List of publication is given at the end. (author). refs., figs., tabs
Activity report of Reactor Physics Division-1995
International Nuclear Information System (INIS)
Gopalakrishnan, V.
1996-01-01
The research and development (R and D) activities of the Reactor Physics Division of Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam during 1995 are reported. The activity are arranged under the headings: Nuclear Data Processing and Validation, Core Physics and Operation Studies, Reactor Kinetics and Safety analysis, Reactor Noise Analysis and Radiation Transport and Shielding Studies. List of publication is given at the end. refs., figs., tabs
Isothermal calorimeter for reactor radiation dosimetry
Energy Technology Data Exchange (ETDEWEB)
Radak, B; Markovic, V [Institute of Nuclear Sciences Boris Kidric, Odeljenje za radijacionu hemiju, Vinca, Beograd (Serbia and Montenegro)
1961-12-15
An isothermal calorimeter with thermistors for measuring absorbed dose rates from 10{sup 4}-5-6.10{sup 5} rad/h in reactor experimental holes has been designed. A kinetics method for determining the equilibrium temperature difference has been developed, and its application in isothermal calorimetry proved. The expected accuracy in measurements within {+-} 2-5% has been proved by measurements carried out in the reactor. Some data obtained by measurements in the reactor RA are presented (author)
Activity report of Reactor Physics Division - 1993
Energy Technology Data Exchange (ETDEWEB)
Indira, R [ed.; Indira Gandhi Centre for Atomic Research, Kalpakkam (India)
1994-12-31
The research and development (R and D) activities of the Reactor Physics Division of Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam during 1993 are reported. The activities are arranged under the headings: Nuclear Data Processing and validation, Core Physics and Operation Studies, Reactor Kinetics and Safety analysis, Reactor Noise Analysis and Radiation Transport and Shielding Studies. List of publication is given at the end. (author). refs., figs., tabs.
International Nuclear Information System (INIS)
Hicks, D.R.; Kraml, M.; Cayen, M.N.; Dubuc, J.; Ryder, S.; Dvornik, D.
1984-01-01
The kinetics of tolrestat, a potent inhibitor of aldose reductase, were examined. Serum concentrations of tolrestat and of total 14 C were measured after dosing normal subjects and subjects with diabetes with 14 C-labeled tolrestat. In normal subjects, tolrestat was rapidly absorbed and disappearance from serum was biphasic. Distribution and elimination t 1/2s were approximately 2 and 10 to 12 hr, respectively, after single and multiple doses. Unchanged tolrestat accounted for the major portion of 14 C in serum. Radioactivity was rapidly and completely excreted in urine and feces in an approximate ratio of 2:1. Findings were much the same in subjects with diabetes. In normal subjects, the kinetics of oral tolrestat were independent of dose in the 10 to 800 mg range. Repetitive dosing did not result in unexpected cumulation. Tolrestat was more than 99% bound to serum protein; it did not compete with warfarin for binding sites but was displaced to some extent by high concentrations of tolbutamide or salicylate
Density variations in a reactor during liquid full dimerization
Golombok, M.; Bruijn, J.
2000-01-01
In a liquid full plug flow reactor during lower olefin dimerization, the assumption of constant density is not valid—the volume of a plug changes as it proceeds along the reactor. The observed kinetics depend on the density variation in the reactor as the conversion proceeds towards a distribution
International Nuclear Information System (INIS)
Ait Abderrahim, A.
2002-01-01
SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised
Energy Technology Data Exchange (ETDEWEB)
Ait Abderrahim, A
2001-04-01
The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.
Energy Technology Data Exchange (ETDEWEB)
Ait Abderrahim, A
2002-04-01
SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised.
International Nuclear Information System (INIS)
Ait Abderrahim, A.
2001-01-01
The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised
Energy Technology Data Exchange (ETDEWEB)
NONE
1966-07-01
This Volume 1 of the Proceedings of 2. Yugoslav symposium on reactor physics includes nine papers dealing with the following topics: reactor kinetics, reactor noise, neutron detection, methods for calculating neutron flux spatial and time dependence in the reactor cores of both heavy and light water moderated experimental reactors, calculation of reactor lattice parameters, reactor instrumentation, reactor monitoring systems; measuring methods of reactor parameters; reactor experimental facilities.
Shaw, J
2013-01-01
Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp
Flash pyrolysis kinetics of pine wood
Wagenaar, B.M.; Prins, W.; van Swaaij, Willibrordus Petrus Maria
1993-01-01
The kinetics of sawdust pyrolysis in the temperature range from 300 to 600 °C has been measured. A thermogravimetric analyser was applied for the temperature range from 300 to 450 °C while for measurements in the temperature range from 450 to 600 °C, an entrained flow reactor has been used. The
Kinetics of Ethyl Acetate Synthesis Catalyzed by Acidic Resins
Antunes, Bruno M.; Cardoso, Simao P.; Silva, Carlos M.; Portugal, Ines
2011-01-01
A low-cost experiment to carry out the second-order reversible reaction of acetic acid esterification with ethanol to produce ethyl acetate is presented to illustrate concepts of kinetics and reactor modeling. The reaction is performed in a batch reactor, and the acetic acid concentration is measured by acid-base titration versus time. The…
International Nuclear Information System (INIS)
Middleton, J.E.
1977-01-01
Reference is made to water cooled reactors and in particular to the cooling system of steam generating heavy water reactors (SGHWR). A two-coolant circuit is described for the latter. Full constructural details are given. (U.K.)
International Nuclear Information System (INIS)
Lawton, H.
1984-01-01
A pioneering project on the decommissioning of the Windscale Advanced Gas-cooled Reactor, by the UKAEA, is described. Reactor data; policy; waste management; remote handling equipment; development; and recording and timescales, are all briefly discussed. (U.K.)
International Nuclear Information System (INIS)
1978-02-01
In addition to basic characteristics of the RA reactor, organizational scheme and financial incentives, this document covers describes the state of the reactor components after 18 years of operation, problems concerned with obtaining the licence for operation with 80% fuel, problems of spent fuel storage in the storage pool of the reactor building and the need for renewal of reactor equipment, first of all instrumentation [sr
International Nuclear Information System (INIS)
Moura Neto, C. de; Nair, R.P.K.
1979-08-01
The study of reflected reactors can be done employing the multigroup diffusion method. The neutron conservation equations, inside the intervals, can be written by fluxes and group constants. A reflected reactor (one and two groups) for a slab geometry is studied, aplying the continuity of flux and current in the interface. At the end, the appropriated solutions for a infinite cylindrical reactor and for a spherical reactor are presented. (Author) [pt
International Nuclear Information System (INIS)
Hattori, Sadao; Sato, Morihiko.
1994-01-01
Liquid metals such as liquid metal sodium are filled in a reactor container as primary coolants. A plurality of reactor core containers are disposed in a row in the circumferential direction along with the inner circumferential wall of the reactor container. One or a plurality of intermediate coolers are disposed at the inside of an annular row of the reactor core containers. A reactor core constituted with fuel rods and control rods (module reactor core) is contained at the inside of each of the reactor core containers. Each of the intermediate coolers comprises a cylindrical intermediate cooling vessels. The intermediate cooling vessel comprises an intermediate heat exchanger for heat exchange of primary coolants and secondary coolants and recycling pumps for compulsorily recycling primary coolants at the inside thereof. Since a plurality of reactor core containers are thus assembled, a great reactor power can be attained. Further, the module reactor core contained in one reactor core vessel may be small sized, to facilitate the control for the reactor core operation. (I.N.)
Mid-space-independent deformable image registration.
Aganj, Iman; Iglesias, Juan Eugenio; Reuter, Martin; Sabuncu, Mert Rory; Fischl, Bruce
2017-05-15
Aligning images in a mid-space is a common approach to ensuring that deformable image registration is symmetric - that it does not depend on the arbitrary ordering of the input images. The results are, however, generally dependent on the mathematical definition of the mid-space. In particular, the set of possible solutions is typically restricted by the constraints that are enforced on the transformations to prevent the mid-space from drifting too far from the native image spaces. The use of an implicit atlas has been proposed as an approach to mid-space image registration. In this work, we show that when the atlas is aligned to each image in the native image space, the data term of implicit-atlas-based deformable registration is inherently independent of the mid-space. In addition, we show that the regularization term can be reformulated independently of the mid-space as well. We derive a new symmetric cost function that only depends on the transformation morphing the images to each other, rather than to the atlas. This eliminates the need for anti-drift constraints, thereby expanding the space of allowable deformations. We provide an implementation scheme for the proposed framework, and validate it through diffeomorphic registration experiments on brain magnetic resonance images. Copyright © 2017 Elsevier Inc. All rights reserved.
International Nuclear Information System (INIS)
1982-11-01
After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)
International Nuclear Information System (INIS)
Merchie, Francois
2015-10-01
This article proposes an overview of research reactors, i.e. nuclear reactors of less than 100 MW. Generally, these reactors are used as neutron generators for basic research in matter sciences and for technological research as a support to power reactors. The author proposes an overview of the general design of research reactors in terms of core size, of number of fissions, of neutron flow, of neutron space distribution. He outlines that this design is a compromise between a compact enough core, a sufficient experiment volume, and high enough power densities without affecting neutron performance or its experimental use. The author evokes the safety framework (same regulations as for power reactors, more constraining measures after Fukushima, international bodies). He presents the main characteristics and operation of the two families which represent almost all research reactors; firstly, heavy water reactors (photos, drawings and figures illustrate different examples); and secondly light water moderated and cooled reactors with a distinction between open core pool reactors like Melusine and Triton, pool reactors with containment, experimental fast breeder reactors (Rapsodie, the Russian BOR 60, the Chinese CEFR). The author describes the main uses of research reactors: basic research, applied and technological research, safety tests, production of radio-isotopes for medicine and industry, analysis of elements present under the form of traces at very low concentrations, non destructive testing, doping of silicon mono-crystalline ingots. The author then discusses the relationship between research reactors and non proliferation, and finally evokes perspectives (decrease of the number of research reactors in the world, the Jules Horowitz project)
The Optimization of power reactor control system
International Nuclear Information System (INIS)
Danupoyo, S.D.
1997-01-01
A power reactor is an important part in nuclear powered electrical plant systems. Success in controlling the power reactor will establish safety of the whole power plant systems. Until now, the power reactor has been controlled by a classical control system that was designed based on output feedback method. To meet the safety requirements that are now more restricted, the recently used power reactor control system should be modified. this paper describes a power reactor control system that is designed based on a state feedback method optimized with LQG (Linear-quadrature-gaussian) method and equipped with a state estimator. A pressurized-water type reactor has been used as the model. by using a point kinetics method with one group delayed neutrons. the result of simulation testing shows that the optimized control system can control the power reactor more effective and efficient than the classical control system
Kinetic theory of radiation effects
International Nuclear Information System (INIS)
Mansur, L.K.
1987-01-01
To help achieve the quantitative and mechanistic understanding of these processes, the kinetic theory of radiation effects has been developed in the DOE basic energy sciences radiation effects and fusion reactor materials programs, as well as in corresponding efforts in other countries. This discipline grapples with a very wide range of phenomena and draws on numerous sub-fields of theory such as defect physics, diffusion, elasticity, chemical reaction rates, phase transformations and thermodynamics. The theory is cast in a mathematical framework of continuum dynamics. Issues particularly relevant to the present inquiry can be viewed from the standpoints of applications of the theory and areas requiring further progress
Design characteristics of zero power fast reactor Lasta
International Nuclear Information System (INIS)
Milosevic, M.; Stefanovic, D.; Pesic, M.; Popovic, D.; Nikolic, D.; Antic, D.; Zavaljevski, N.
1987-01-01
The concept, purpose and preliminary design of a zero power fast reactor LASTA are described. The methods of computing the reactor core parameters and reactor kinetics are presented with the basic calculated results and analysis for one selected LASTA configuration. The nominal parameters are determined according to the selected reactor safety criteria and results of calculations. Important aspects related to the overall safety are examined in detail. (author)
Fundamental aspects of plasma chemical physics kinetics
Capitelli, Mario; Colonna, Gianpiero; Esposito, Fabrizio; Gorse, Claudine; Hassouni, Khaled; Laricchiuta, Annarita; Longo, Savino
2016-01-01
Describing non-equilibrium "cold" plasmas through a chemical physics approach, this book uses the state-to-state plasma kinetics, which considers each internal state as a new species with its own cross sections. Extended atomic and molecular master equations are coupled with Boltzmann and Monte Carlo methods to solve the electron energy distribution function. Selected examples in different applied fields, such as microelectronics, fusion, and aerospace, are presented and discussed including the self-consistent kinetics in RF parallel plate reactors, the optimization of negative ion sources and the expansion of high enthalpy flows through nozzles of different geometries. The book will cover the main aspects of the state-to-state kinetic approach for the description of nonequilibrium cold plasmas, illustrating the more recent achievements in the development of kinetic models including the self-consistent coupling of master equations and Boltzmann equation for electron dynamics. To give a complete portrayal, the...
International Nuclear Information System (INIS)
Young, J.R.
1976-01-01
The fusion reactor currently is being developed as a clean source of electricity with an essentially infinite source of fuel. These reactors are visualized as using a fusion reaction to generate large quantities of high temperature energy which can be used as process heat or for the generation of electricity. The energy would be created primarily as the kinetic energy of neutrons or other reaction products. Neutron energy could be converted to high-temperature heat by moderation and capture of the neutrons. The energy of other reaction products could be converted to high-temperature heat by capture, or directly to electricity by direct conversion electrostatic equipment. An analysis to determine the wastes released as a result of operation of fusion power plants is presented
International Nuclear Information System (INIS)
Ise, Takeharu
1976-12-01
Review studies have been made on algorithms of numerical analysis and benchmark tests on point kinetics and quasistatic approximate kinetics computer codes to perform efficiently benchmark tests on space-dependent neutron kinetics codes. Point kinetics methods have now been improved since they can be directly applied to the factorization procedures. Methods based on Pade rational function give numerically stable solutions and methods on matrix-splitting are interested in the fact that they are applicable to the direct integration methods. An improved quasistatic (IQ) approximation is the best and the most practical method; it is numerically shown that the IQ method has a high stability and precision and the computation time which is about one tenth of that of the direct method. IQ method is applicable to thermal reactors as well as fast reactors and especially fitted for fast reactors to which many time steps are necessary. Two-dimensional diffusion kinetics codes are most practicable though there exist also three-dimensional diffusion kinetics code as well as two-dimensional transport kinetics code. On developing a space-dependent kinetics code, in any case, it is desirable to improve the method so as to have a high computing speed for solving static diffusion and transport equations. (auth.)
International Nuclear Information System (INIS)
Naruse, Yoshihiro.
1990-01-01
The thickness of steel shell plates in a reactor container embedded in sand cussions is monitored to recognize the corrosion of the steel shell plates. That is, the reactor pressure vessel is contained in a reactor container shell and the sand cussions are disposed on the lower outside of the reactor container shell to elastically support the shell. A pit is disposed at a position opposing to the sand cussions for measuring the thickness of the reactor container shell plates. The pit is usually closed by a closing member. In the reactor container thus constituted, the closing member can be removed upon periodical inspection to measure the thickness of the shell plates. Accordingly, the corrosion of the steel shell plates can be recognized by the change of the plate thickness. (I.S.)
International Nuclear Information System (INIS)
Moir, R.W.
1980-01-01
The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of 233 U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m -2 , and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid
International Nuclear Information System (INIS)
Garabedian, G.
1988-01-01
A liquid reactor is described comprising: (a) a reactor vessel having a core; (b) one or more satellite tanks; (c) pump means in the satellite tank; (d) heat exchanger means in the satellite tank; (e) an upper liquid metal conduit extending between the reactor vessel and the satellite tank; (f) a lower liquid metal duct extending between the reactor vessel and satellite tanks the upper liquid metal conduit and the lower liquid metal duct being arranged to permit free circulation of liquid metal between the reactor vessel core and the satellite tank by convective flow of liquid metal; (g) a separate sealed common containment vessel around the reactor vessel, conduits and satellite tanks; (h) the satellite tank having space for a volume of liquid metal that is sufficient to dampen temperature transients resulting from abnormal operating conditions
International Nuclear Information System (INIS)
Ailloud, J.; Millot, J.P.
1964-01-01
circumstances... - experimental investigations on power excursions linked with precise initial conditions: the aim of this work is to define the basis for theoretical research, and the limits beyond which the risks of explosion cease to be negligible. The research work will be done so as to enable checking with outside reactor experiments and to continue them in the explosion field. - studies of the behaviour of the reactor control-instrumentation. - experimental investigations related with transient operation with initial short life (study of boiling, temperature measurements, vacuum pressure and fraction...) with the aim of defining the hypotheses of a theory on swimming-pool reactor kinetics related to heat transfer phenomena, - investigations of the behaviour of fuels in reactors (these experiments are planned to be carried out in loops) Preliminary experimental results. CABRI went critical on the 21 December 1963. The first transient experiments are expected for March 1964. (authors) [fr
International Nuclear Information System (INIS)
Batheja, P.; Huber, R.; Rau, P.
1985-01-01
Particularly for nuclear reactors of small output, the reactor pressure vessel contains at least two heat exchangers, which have coolant flowing through them in a circuit through the reactor core. The circuit of at least one heat exchanger is controlled by a slide valve, so that even for low drive forces, particularly in natural circulation, the required even loading of the heat exchanger is possible. (orig./HP) [de
International Nuclear Information System (INIS)
Moura Neto, C. de; Nair, R.P.K.
1979-08-01
The microscopic study of a cell is meant for the determination of the infinite multiplication factor of the cell, which is given by the four factor formula: K(infinite) = n(epsilon)pf. The analysis of an homogeneous reactor is similar to that of an heterogeneous reactor, but each factor of the four factor formula can not be calculated by the formulas developed in the case of an homogeneous reactor. A great number of methods was developed for the calculation of heterogeneous reactors and some of them are discussed. (Author) [pt
Integral Design Methodology of Photocatalytic Reactors for Air Pollution Remediation
Directory of Open Access Journals (Sweden)
Claudio Passalía
2017-06-01
Full Text Available An integral reactor design methodology was developed to address the optimal design of photocatalytic wall reactors to be used in air pollution control. For a target pollutant to be eliminated from an air stream, the proposed methodology is initiated with a mechanistic derived reaction rate. The determination of intrinsic kinetic parameters is associated with the use of a simple geometry laboratory scale reactor, operation under kinetic control and a uniform incident radiation flux, which allows computing the local superficial rate of photon absorption. Thus, a simple model can describe the mass balance and a solution may be obtained. The kinetic parameters may be estimated by the combination of the mathematical model and the experimental results. The validated intrinsic kinetics obtained may be directly used in the scaling-up of any reactor configuration and size. The bench scale reactor may require the use of complex computational software to obtain the fields of velocity, radiation absorption and species concentration. The complete methodology was successfully applied to the elimination of airborne formaldehyde. The kinetic parameters were determined in a flat plate reactor, whilst a bench scale corrugated wall reactor was used to illustrate the scaling-up methodology. In addition, an optimal folding angle of the corrugated reactor was found using computational fluid dynamics tools.
Programming for a nuclear reactor instrument simulator
International Nuclear Information System (INIS)
Cohn, C.E.
1989-01-01
A new computerized control system for a transient test reactor incorporates a simulator for pre-operational testing of control programs. The part of the simulator pertinent to the discussion here consists of two microprocessors. An 8086/8087 reactor simulator calculates simulated reactor power by solving the reactor kinetics equations. An 8086 instrument simulator takes the most recent power value developed by the reactor simulator and simulates the appropriate reading on each of the eleven reactor instruments. Since the system is required to run on a one millisecond cycle, careful programming was required to take care of all eleven instruments in that short time. This note describes the special programming techniques used to attain the needed performance
A new kinetic biphasic approach applied to biodiesel process intensification
Energy Technology Data Exchange (ETDEWEB)
Russo, V.; Tesser, R.; Di Serio, M.; Santacesaria, E. [Naples Univ. (Italy). Dept. of Chemistry
2012-07-01
Many different papers have been published on the kinetics of the transesterification of vegetable oil with methanol, in the presence of alkaline catalysts to produce biodiesel. All the proposed approaches are based on the assumption of a pseudo-monophasic system. The consequence of these approaches is that some experimental aspects cannot be described. For the reaction performed in batch conditions, for example, the monophasic approach is not able to reproduce the different plateau obtained by using different amount of catalyst or the induction time observed at low stirring rates. Moreover, it has been observed by operating in continuous reactors that micromixing has a dramatic effect on the reaction rate. At this purpose, we have recently observed that is possible to obtain a complete conversion to biodiesel in less than 10 seconds of reaction time. This observation is confirmed also by other authors using different types of reactors like: static mixers, micro-reactors, oscillatory flow reactors, cavitational reactors, microwave reactors or centrifugal contactors. In this work we will show that a recently proposed biphasic kinetic approach is able to describe all the aspects before mentioned that cannot be described by the monophasic kinetic model. In particular, we will show that the biphasic kinetic model can describe both the induction time observed in the batch reactors, at low stirring rate, and the very high conversions obtainable in a micro-channel reactor. The adopted biphasic kinetic model is based on a reliable reaction mechanism that will be validated by the experimental evidences reported in this work. (orig.)
Energy Technology Data Exchange (ETDEWEB)
Kuerten, H; Zehner, P [BASF A.G., Ludwigshafen am Rhein (Germany, F.R.)
1979-08-01
Slurry reactors are designed on the basis of empirical data and model investigations. It is as yet not possible to calculate the flow behavior of such reactors. The swarm of gas bubbles and cluster formations of solid particles and their interaction in industrial reactors are not known. These effects control to a large extent the gas hold-up, the gas-liquid interface and, similarly as in bubble columns, the back-mixing of liquids and solids. These hydrodynamic problems are illustrated in slurry reactors which constructionally may be bubble columns, stirred tanks or jet loop reactors. The expected effects are predicted by means of tests with model systems modified to represent the conditions in industrial hydrogenation reactors. In his book 'Mass Transfer in Heterogeneous Catalysis' (1970) Satterfield complained of the lack of knowledge about the design of slurry reactors and hence of the impossible task of the engineer who has to design a plant according to accepted rules. There have been no fundamental changes since then. This paper presents the problems facing the engineer in designing slurry reactors, and shows new development trends.
International Nuclear Information System (INIS)
Butz, H.P.; Heuser, F.W.; May, H.
1985-01-01
The paper comprises an introduction into nuclear physics bases, the safety concept generally speaking, safety devices of pwr type reactors, accident analysis, external influences, probabilistic safety assessment and risk studies. It further describes operational experience, licensing procedures under the Atomic Energy Law, research in reactor safety and the nuclear fuel cycle. (DG) [de
International Nuclear Information System (INIS)
Mysels, K.J.; Shenoy, A.S.
1976-01-01
A nuclear reactor is described in which the core consists of a number of fuel regions through each of which regulated coolant flows. The coolant from neighbouring fuel regions is combined in a manner which results in an averaging of the coolant temperature at the outlet of the core. By this method the presence of hot streaks in the reactor is reduced. (UK)
International Nuclear Information System (INIS)
Kato, Masami; Nishio, Masahide.
1987-01-01
Purpose: To prevent the rupture of the dry well even when the melted reactor core drops into a reactor pedestal cavity. Constitution: In a reactor container in which a dry well disposed above the reactor pedestal cavity for containing the reactor pressure vessel and a torus type suppression chamber for containing pressure suppression water are connected with each other, the pedestal cavity and the suppression chamber are disposed such that the flow level of the pedestal cavity is lower than the level of the pressure suppression water. Further, a pressure suppression water introduction pipeway for introducing the pressure suppression water into the reactor pedestal cavity is disposed by way of an ON-OFF valve. In case if the melted reactor core should fall into the pedestal cavity, the ON-OFF valve for the pressure suppression water introduction pipeway is opened to introduce the pressure suppression water in the suppression chamber into the pedestal cavity to cool the melted reactor core. (Ikeda, J.)
International Nuclear Information System (INIS)
1989-01-01
This chapter includes the following: General description of the RA reactor, organization of work, responsibilities of leadership and operators team, regulations concerning operation and behaviour in the reactor building, regulations for performing experiments, regulations and instructions for inserting samples into experimental channels [sr
International Nuclear Information System (INIS)
Ait Abderrahim, H.
1998-01-01
Progress in research on reactor physics in 1997 at the Belgian Nuclear Research Centre SCK/CEN is described. Activities in the following four domains are discussed: core physics, ex-core neutron transport, experiments in Materials Testing Reactors, international benchmarks
Aluminum corrosion product release kinetics
Energy Technology Data Exchange (ETDEWEB)
Edwards, Matt, E-mail: Matthew.Edwards@cnl.ca; Semmler, Jaleh; Guzonas, Dave; Chen, Hui Qun; Toor, Arshad; Hoendermis, Seanna
2015-07-15
Highlights: • Release of Al corrosion product was measured in simulated post-LOCA sump solutions. • Increased boron was found to enhance Al release kinetics at similar pH. • Models of Al release as functions of time, temperature, and pH were developed. - Abstract: The kinetics of aluminum corrosion product release was examined in solutions representative of post-LOCA sump water for both pressurized water and pressurized heavy-water reactors. Coupons of AA 6061 T6 were exposed to solutions in the pH 7–11 range at 40, 60, 90 and 130 °C. Solution samples were analyzed by inductively coupled plasma atomic emission spectroscopy, and coupon samples were analyzed by secondary ion mass spectrometry. The results show a distinct “boron effect” on the release kinetics, expected to be caused by an increase in the solubility of the aluminum corrosion products. New models were developed to describe both sets of data as functions of temperature, time, and pH (where applicable)
International Nuclear Information System (INIS)
Azekura, Kazuo; Kurihara, Kunitoshi.
1992-01-01
In a BWR type reactor, a great number of pipes (spectral shift pipes) are disposed in the reactor core. Moderators having a small moderating cross section (heavy water) are circulated in the spectral shift pipes to suppress the excess reactivity while increasing the conversion ratio at an initial stage of the operation cycle. After the intermediate stage of the operation cycle in which the reactor core reactivity is lowered, reactivity is increased by circulating moderators having a great moderating cross section (light water) to extend the taken up burnup degree. Further, neutron absorbers such as boron are mixed to the moderator in the spectral shift pipe to control the concentration thereof. With such a constitution, control rods and driving mechanisms are no more necessary, to simplify the structure of the reactor core. This can increase the fuel conversion ratio and control great excess reactivity. Accordingly, a nuclear reactor core of high conversion and high burnup degree can be attained. (I.N.)
International Nuclear Information System (INIS)
Fukazawa, Masanori.
1991-01-01
A system for controlling combustible gases, it has been constituted at present such that the combustible gases are controlled by exhausting them to the wet well of a reactor container. In this system, however, there has been a problem, in a reactor container having plenums in addition to the wet well and the dry well, that the combustible gases in such plenums can not be controlled. In view of the above, in the present invention, suction ports or exhaust ports of the combustible gas control system are disposed to the wet well, the dry well and the plenums to control the combustible gases in the reactor container. Since this can control the combustible gases in the entire reactor container, the integrity of the reactor container can be ensured. (T.M.)
International Nuclear Information System (INIS)
Kojima, Yoshihiro; Hosomi, Kenji; Otonari, Jun-ichiro.
1997-01-01
In the present invention, a catalyst for oxidizing hydrogen to be disposed in a reactor container upon rupture of pipelines of a reactor primary coolant system is prevented from deposition of water droplets formed from a reactor container spray to suppress elevation of hydrogen concentration in the reactor container. Namely, a catalytic combustion gas concentration control system comprises a catalyst for oxidizing hydrogen and a support thereof. In addition, there is also disposed a water droplet deposition-preventing means for preventing deposition of water droplets in a reactor pressure vessel on the catalyst. Then, the effect of the catalyst upon catalytic oxidation reaction of hydrogen can be kept high. The local elevation of hydrogen concentration can be prevented even upon occurrence of such a phenomenon that various kinds of mobile forces in the container such as dry well cooling system are lost. (I.S.)
International Nuclear Information System (INIS)
Tilliette, Z.
1975-01-01
A description is given of a nuclear reactor and especially a high-temperature reactor in which provision is made within a pressure vessel for a main cavity containing the reactor core and a series of vertical cylindrical pods arranged in spaced relation around the main cavity and each adapted to communicate with the cavity through two collector ducts or headers for the primary fluid which flows downwards through the reactor core. Each pod contains two superposed steam-generator and circulator sets disposed in substantially symmetrical relation on each side of the hot primary-fluid header which conveys the primary fluid from the reactor cavity to the pod, the circulators of both sets being mounted respectively at the bottom and top ends of the pod
Activity report of Reactor Physics Division : 1990
International Nuclear Information System (INIS)
Mohanakrishnan, P.
1991-01-01
The major Research and Development and Project activities carried out during the year 1990 in Reactor Physics Division are presented in the form of summaries in this report. The various activities are organised under the following areas : (1) Nuclear Data Evaluation, Processing and Validation, (2) Core Physics and Analysis, (3) Reactor Kinetics and Safety Analysis, (4) Noise Analysis, and (5) Radiation Transport and Shielding. FBTR was restarted in July 1990 and the power was raised upto 500 kW. A number of low power physics experiments on reactivity coefficients, kinetics and noise, neutron flux and gamma dose in B cells, were performed, which are discussed in this report. (author). figs., tabs
International Nuclear Information System (INIS)
1975-01-01
Full text: The meeting was attended by 15 participants from seven countries and two international organizations. The Eighth Annual Meeting of the International Working Group on Fast Reactors (IWGFR) was attended by representatives from France, Fed. Rep. Germany, Italy, Japan, United Kingdom, Union of Soviet Socialist Republics and the United States of America - countries that have made significant progress in developing the technology and physics of sodium cooled fast reactors and have extensive national programmes in this field - as well as by representatives of the Commission of the European Communities and the IAEA. The design of fast-reactor power plants is a more difficult task than developing facilities with thermal reactors. Different reactor kinetics and dynamics, a hard neutron spectrum, larger integral doses of fuel and structural material irradiation, higher core temperatures, the use of an essentially novel coolant, and, as a result of all these factors, the additional reliability and safety requirements that are imposed on the planning and operation of sodium cooled fast reactors - all these factors pose problems that can be solved comprehensively only by countries with a high level of scientific and technical development. The exchange of experience between these countries and their combined efforts in solving the fundamental problems that arise in planning, constructing and operating fast reactors are promoting technical progress and reducing the relative expenditure required for various studies on developing and introducing commercial fast reactors. For this reason, the meeting concentrated on reviewing and discussing national fast reactor programmes. The situation with regard to planning, constructing and operating fast experimental and demonstration reactors in the countries concerned, the experience accumulated in operating them, the difficulties arising during operation and ways of over-coming them, the search for optimal designs for the power
Cesium removal and kinetics equilibrium: Precipitation kinetics
International Nuclear Information System (INIS)
Barnes, M.J.
1999-01-01
This task consisted of both non-radioactive and radioactive (tracer) tests examining the influence of potentially significant variables on cesium tetraphenylborate precipitation kinetics. The work investigated the time required to reach cesium decontamination and the conditions that affect the cesium precipitation kinetics
Kinetic stability of internal kink mode
International Nuclear Information System (INIS)
Romanelli, F.; Fogaccia, G.
1993-01-01
With reference to studies of the attainment of ignited operations on devices like ITER (International Thermonuclear Experimental Reactor), the stability of the internal kink mode is re-investigated by taking into account the contribution of perpendicular compressibility, obtained by solving the drift kinetic equation. The resulting stability condition yields threshold values typically larger than the conventional Bussac criterion. For the case of ultra-flat safety factor profiles, the mode can be stable also in the absence of line-bending
Performance of neutron kinetics models for ADS transient analyses
International Nuclear Information System (INIS)
Rineiski, A.; Maschek, W.; Rimpault, G.
2002-01-01
Within the framework of the SIMMER code development, neutron kinetics models for simulating transients and hypothetical accidents in advanced reactor systems, in particular in Accelerator Driven Systems (ADSs), have been developed at FZK/IKET in cooperation with CE Cadarache. SIMMER is a fluid-dynamics/thermal-hydraulics code, coupled with a structure model and a space-, time- and energy-dependent neutronics module for analyzing transients and accidents. The advanced kinetics models have also been implemented into KIN3D, a module of the VARIANT/TGV code (stand-alone neutron kinetics) for broadening application and for testing and benchmarking. In the paper, a short review of the SIMMER and KIN3D neutron kinetics models is given. Some typical transients related to ADS perturbations are analyzed. The general models of SIMMER and KIN3D are compared with more simple techniques developed in the context of this work to get a better understanding of the specifics of transients in subcritical systems and to estimate the performance of different kinetics options. These comparisons may also help in elaborating new kinetics models and extending existing computation tools for ADS transient analyses. The traditional point-kinetics model may give rather inaccurate transient reaction rate distributions in an ADS even if the material configuration does not change significantly. This inaccuracy is not related to the problem of choosing a 'right' weighting function: the point-kinetics model with any weighting function cannot take into account pronounced flux shape variations related to possible significant changes in the criticality level or to fast beam trips. To improve the accuracy of the point-kinetics option for slow transients, we have introduced a correction factor technique. The related analyses give a better understanding of 'long-timescale' kinetics phenomena in the subcritical domain and help to evaluate the performance of the quasi-static scheme in a particular case. One
International Nuclear Information System (INIS)
Maruyama, Toru; Murata, Ritsuko.
1996-01-01
In the present invention, a spent fuel storage pool of a BWR type reactor is formed at an upper portion and enlarged in the size to effectively utilize the space of the building. Namely, a reactor chamber enhouses reactor facilities including a reactor pressure vessel and a reactor container, and further, a spent fuel storage pool is formed thereabove. A second spent fuel storage pool is formed above the auxiliary reactor chamber at the periphery of the reactor chamber. The spent fuel storage pool and the second spent fuel storage pool are disposed in adjacent with each other. A wall between both of them is formed vertically movable. With such a constitution, the storage amount for spent fuels is increased thereby enabling to store the entire spent fuels generated during operation period of the plant. Further, since requirement of the storage for the spent fuels is increased stepwisely during periodical exchange operation, it can be used for other usage during the period when the enlarged portion is not used. (I.S.)
International Nuclear Information System (INIS)
Shibata, Satoru; Kawashima, Hiroaki
1984-01-01
Purpose: To optimize the temperature distribution of the reactor container so as to moderate the thermal stress distribution on the reactor wall of LMFBR type reactor. Constitution: A good heat conductor (made of Al or Cu) is appended on the outer side of the reactor container wall from below the liquid level to the lower face of a deck plate. Further, heat insulators are disposed to the outside of the good heat conductor. Furthermore, a gas-cooling duct is circumferentially disposed at the contact portion between the good heat conductor and the deck plate around the reactor container. This enables to flow the cold heat from the liquid metal rapidly through the good heat conductor to the cooling duct and allows to maintain the temperature distribution on the reactor wall substantially linear even with the abrupt temperature change in the liquid metal. Further, by appending the good heat conductor covered with inactive metals not only on the outer side but also on the inside of the reactor wall to introduce the heat near the liquid level to the upper portion and escape the same to the cooling layer below the roof slab, the effect can be improved further. (Ikeda, J.)
Neutron inverse kinetics via Gaussian Processes
International Nuclear Information System (INIS)
Picca, Paolo; Furfaro, Roberto
2012-01-01
Highlights: ► A novel technique for the interpretation of experiments in ADS is presented. ► The technique is based on Bayesian regression, implemented via Gaussian Processes. ► GPs overcome the limits of classical methods, based on PK approximation. ► Results compares GPs and ANN performance, underlining similarities and differences. - Abstract: The paper introduces the application of Gaussian Processes (GPs) to determine the subcriticality level in accelerator-driven systems (ADSs) through the interpretation of pulsed experiment data. ADSs have peculiar kinetic properties due to their special core design. For this reason, classical – inversion techniques based on point kinetic (PK) generally fail to generate an accurate estimate of reactor subcriticality. Similarly to Artificial Neural Networks (ANNs), Gaussian Processes can be successfully trained to learn the underlying inverse neutron kinetic model and, as such, they are not limited to the model choice. Importantly, GPs are strongly rooted into the Bayes’ theorem which makes them a powerful tool for statistical inference. Here, GPs have been designed and trained on a set of kinetics models (e.g. point kinetics and multi-point kinetics) for homogeneous and heterogeneous settings. The results presented in the paper show that GPs are very efficient and accurate in predicting the reactivity for ADS-like systems. The variance computed via GPs may provide an indication on how to generate additional data as function of the desired accuracy.
Variational estimates of point-kinetics parameters
International Nuclear Information System (INIS)
Favorite, J.A.; Stacey, W.M. Jr.
1995-01-01
Variational estimates of the effect of flux shifts on the integral reactivity parameter of the point-kinetics equations and on regional power fractions were calculated for a variety of localized perturbations in two light water reactor (LWR) model problems representing a small, tightly coupled core and a large, loosely coupled core. For the small core, the flux shifts resulting from even relatively large localized reactivity changes (∼600 pcm) were small, and the standard point-kinetics approximation estimates of reactivity were in error by only ∼10% or less, while the variational estimates were accurate to within ∼1%. For the larger core, significant (>50%) flux shifts occurred in response to local perturbations, leading to errors of the same magnitude in the standard point-kinetics approximation of the reactivity worth. For positive reactivity, the error in the variational estimate of reactivity was only a few percent in the larger core, and the resulting transient power prediction was 1 to 2 orders of magnitude more accurate than with the standard point-kinetics approximation. For a large, local negative reactivity insertion resulting in a large flux shift, the accuracy of the variational estimate broke down. The variational estimate of the effect of flux shifts on reactivity in point-kinetics calculations of transients in LWR cores was found to generally result in greatly improved accuracy, relative to the standard point-kinetics approximation, the exception being for large negative reactivity insertions with large flux shifts in large, loosely coupled cores
International Nuclear Information System (INIS)
Rau, P.
1980-01-01
The reactor core of nuclear reactors usually is composed of individual elongated fuel elements that may be vertically arranged and through which coolant flows in axial direction, preferably from bottom to top. With their lower end the fuel elements gear in an opening of a lower support grid forming part of the core structure. According to the invention a locking is provided there, part of which is a control element that is movable along the fuel element axis. The corresponding locking element is engaged behind a lateral projection in the opening of the support grid. The invention is particularly suitable for breeder or converter reactors. (orig.) [de
International Nuclear Information System (INIS)
Elliott, J.A.
1993-01-01
Plasma kinetic theory is discussed and a comparison made with the kinetic theory of gases. The plasma is described by a modified set of fluid equations and it is shown how these fluid equations can be derived. (UK)
Neutron and thermal dynamics of a gaseous core fission reactor
International Nuclear Information System (INIS)
van Dam, H.; Kuijper, J.C.; Stekelenburg, A.J.C.; Hoogenboom, J.E.; Boersma-Klein, W.; Kistemaker, J.
1989-01-01
In this paper neutron kinetics and thermal dynamics of a Gaseous Core Fission Reactor with magnetical pumping are shown to have many unconventional aspects. Attention is focused on the properties of the fuel gas, the non-linear neutron kinetics and the energy balance in thermodynamical cycles
Modeling of heat transfer in wall-cooled tubular reactors
Koning, G.W.; Westerterp, K.R.
1999-01-01
In a pilot scale wall-cooled tubular reactor, temperature profiles have been measured with and without reaction. As a model reaction oxidation of carbon monoxide in air over a copper chromite catalyst has been used. The kinetics of this reaction have been determined separately in two kinetic
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
International Nuclear Information System (INIS)
Prescott, R.F.
1976-01-01
A nuclear reactor containment vessel faced internally with a metal liner is provided with thermal insulation for the liner, comprising one or more layers of compressible material such as ceramic fiber, such as would be conventional in an advanced gas-cooled reactor and also a superposed layer of ceramic bricks or tiles in combination with retention means therefor, the retention means (comprising studs projecting from the liner, and bolts or nuts in threaded engagement with the studs) being themselves insulated from the vessel interior so that the coolant temperatures achieved in a High-Temperature Reactor or a Fast Reactor can be tolerated with the vessel. The layer(s) of compressible material is held under a degree of compression either by the ceramic bricks or tiles themselves or by cover plates held on the studs, in which case the bricks or tiles are preferably bedded on a yielding layer (for example of carbon fibers) rather than directly on the cover plates
International Nuclear Information System (INIS)
Miyashita, Akio.
1981-01-01
Purpose: To facilitate and accelerate a leakage test of valves of a main steam pipe by adding a leakage test partition valve thereto. Constitution: A leakage testing partition valve is provided between a pressure vessel for a nuclear reactor and the most upstream side valve of a plurality of valves to be tested for leakage, a testing branch pipe is communicated with the downstream side of the partition valve, and the testing water for preventing leakage is introduced thereto through the branch pipe. Since main steam pipe can be simply isolated by closing the partition valve in the leakage test, the leakage test can be conducted without raising or lowering the water level in the pressure vessel, and since interference with other work in the reactor can be eliminated, the leakage test can be readily conducted parallel with other work in the reactor in a short time. Clean water can be used without using reactor water as the test water. (Yoshihara, H.)
International Nuclear Information System (INIS)
Abe, Yoshihito; Sano, Tamotsu; Ueda, Sabuo; Tanaka, Kazuhisa.
1987-01-01
Purpose: To improve the liquid surface disturbance in LMFBR type reactors. Constitution: A horizontal flow suppressing mechanism mainly comprising vertical members is suspended near the free liquid surface of coolants in the upper plenum. The horizontal flow of coolants near the free liquid surface is reduced by the suppressing mechanism to effectively reduce the surface disturbance. The reduction in the liquid surface disturbance further prevails to the entire surface region with no particular vertical variations to the free liquid surface to remarkably improve the preventive performance for the liquid surface disturbance. Accordingly, it is also possible to attain the advantageous effects such as prevention for the thermal fatigue in reactor vessel walls, reactor upper mechanisms, etc. and prevention of burning damage to the reactor core due to the reduction of envolved Ar gas. (Kamimura, M.)
Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.
1959-02-17
Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.
Miller, H.I.; Smith, R.C.
1958-01-21
This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.
Advances in Reactor Physics, Mathematics and Computation. Volume 2
Energy Technology Data Exchange (ETDEWEB)
1987-01-01
These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, Volume 2, are divided into 7 sessions bearing on: - session 7: Deterministic transport methods 1 (7 conferences), - session 8: Interpretation and analysis of reactor instrumentation (6 conferences), - session 9: High speed computing applied to reactor operations (5 conferences), - session 10: Diffusion theory and kinetics (7 conferences), - session 11: Fast reactor design, validation and operating experience (8 conferences), - session 12: Deterministic transport methods 2 (7 conferences), - session 13: Application of expert systems to physical aspects of reactor design and operation.
International Nuclear Information System (INIS)
Schulze, I.; Gutscher, E.
1980-01-01
The core contains a critical mass of UN or U 2 N 3 in the form of a noncritical solution with melted Sn being kept below a N atmosphere. The lining of the reactor core consists of graphite. If fission progresses part of the melted metal solution is removed and cleaned from fission products. The reactor temperatures lie in the range of 300 to 2000 0 C. (Examples and tables). (RW) [de
International Nuclear Information System (INIS)
Erdoes, P.
1977-01-01
This is one of a series of articles discussing aspects of nuclear engineering ranging from a survey of various reactor types for static and mobile use to mention of atomic thermo-electric batteries of atomic thermo-electric batteries for cardiac pacemakers. Various statistics are presented on power generation in Europe and U.S.A. and economics are discussed in some detail. Molten salt reactors and research machines are also described. (G.M.E.)
RETRAN-02 one-dimensional kinetics model: a review
International Nuclear Information System (INIS)
Gose, G.C.; McClure, J.A.
1986-01-01
RETRAN-02 is a modular code system that has been designed for one-dimensional, transient thermal-hydraulics analysis. In RETRAN-02, core power behavior may be treated using a one-dimensional reactor kinetics model. This model allows the user to investigate the interaction of time- and space-dependent effects in the reactor core on overall system behavior for specific LWR operational transients. The purpose of this paper is to review the recent analysis and development activities related to the one dimensional kinetics model in RETRAN-02
A three-dimensional nodal neutron kinetics capability for relaps
International Nuclear Information System (INIS)
Judd, J.L.; Weaver, W.L.
1996-01-01
The incorporation of a three-dimensional neutron kinetics capability into the DOE version of the RELAP5/MOD3.2 reactor safety code is discussed. A brief discussion of the kinetics method is given along with a discussion of the cross section parameterization models available in RELAP5/MOD3.2. The RELAP5/MOD3.2 code is then used to perform calculations of the NEACRP rod ejection and rod withdrawal benchmarks, and results are presented
International Nuclear Information System (INIS)
Kawabe, Ryuhei; Yamaki, Rika.
1990-01-01
A water vessel is disposed and the gas phase portion of the water vessel is connected to a reactor container by a pipeline having a valve disposed at the midway thereof. A pipe in communication with external air is extended upwardly from the liquid phase portion to a considerable height so as to resist against the back pressure by a waterhead in the pipeline. Accordingly, when the pressure in the container is reduced to a negative level, air passes through the pipeline and uprises through the liquid phase portion in the water vessel in the form of bubbles and then flows into the reactor container. When the pressure inside of the reactor goes higher, since the liquid surface in the water vessel is forced down, water is pushed up into the pipeline. Since the waterhead pressure of a column of water in the pipeline and the pressure of the reactor container are well-balanced, gases in the reactor container are not leaked to the outside. Further, in a case if a great positive pressure is formed in the reactor container, the inner pressure overcomes the waterhead of the column of water, so that the gases containing radioactive aerosol uprise in the pipeline. Since water and the gases flow being in contact with each other, this can provide the effect of removing aerosol. (T.M.)
International Nuclear Information System (INIS)
Vasile, A.
2001-01-01
Fast reactors have capacities to spare uranium natural resources by their breeding property and to propose solutions to the management of radioactive wastes by limiting the inventory of heavy nuclei. This article highlights the role that fast reactors could play for reducing the radiotoxicity of wastes. The conversion of 238 U into 239 Pu by neutron capture is more efficient in fast reactors than in light water reactors. In fast reactors multi-recycling of U + Pu leads to fissioning up to 95% of the initial fuel ( 238 U + 235 U). 2 strategies have been studied to burn actinides: - the multi-recycling of heavy nuclei is made inside the fuel element (homogeneous option); - the unique recycling is made in special irradiation targets placed inside the core or at its surroundings (heterogeneous option). Simulations have shown that, for the same amount of energy produced (400 TWhe), the mass of transuranium elements (Pu + Np + Am + Cm) sent to waste disposal is 60,9 Kg in the homogeneous option and 204.4 Kg in the heterogeneous option. Experimental programs are carried out in Phenix and BOR60 reactors in order to study the feasibility of such strategies. (A.C.)
Detecting Dark Photons with Reactor Neutrino Experiments
Park, H. K.
2017-08-01
We propose to search for light U (1 ) dark photons, A', produced via kinetically mixing with ordinary photons via the Compton-like process, γ e-→A'e-, in a nuclear reactor and detected by their interactions with the material in the active volumes of reactor neutrino experiments. We derive 95% confidence-level upper limits on ɛ , the A'-γ mixing parameter, ɛ , for dark-photon masses below 1 MeV of ɛ reactors as potential sources of intense fluxes of low-mass dark photons.
Activity report of Reactor Physics Division - 1988
International Nuclear Information System (INIS)
Keshavamurthy, R.S.
1989-01-01
This report highlights the progress of activities carried out during the year 1988 in Reactor Physics Division in the form of brief summaries. The topics are organised under the following subject categories:(1) nuclear data evaluation , processing and validation, (2) core physics and analysis, (3) reactor kinetics and safety analysis, (4) noise analysis and (5) radiation transport and shielding. List of publications by the members of the Division and the Reactor Physics Seminars held during the year 1988, is included at the end of report. (author). refs., figs., tabs
Reactor physics for non-nuclear engineers
International Nuclear Information System (INIS)
Lewis, E.E.
2011-01-01
A one-term undergraduate course in reactor physics is described. The instructional format is strongly influenced by its intended audience of non-nuclear engineering students. In contrast to legacy treatments of the subject, the course focuses on the physics of nuclear power reactors with no attempt to include instruction in numerical methods. The multi-physics of power reactors is emphasized highlighting the close interactions between neutronic and thermal phenomena in design and analysis. Consequently, the material's sequencing also differs from traditional treatments, for example treating kinetics before the neutron diffusion is introduced. (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.
Degradation of aqueous phenol solutions by coaxial DBD reactor
Dojcinovic, B. P.; Manojlovic, D.; Roglic, G. M.; Obradovic, B. M.; Kuraica, M. M.; Puric, J.
2008-07-01
Solutions of 2-chlorophenol, 4-chlorophenol and 2,6-dichlorophenol in bidistilled and water from the river Danube were treated in plasma reactor. In this reactor, based on coaxial dielectric barrier discharge at atmospheric pressure, plasma is formed over a thin layer of treated water. After one pass through the reactor, starting chlorophenols concentration of 20 mg/l was diminished up to 95 %. Kinetics of the chlorophenols degradation was monitored by High Pressure Liquid Chromatography method (HPLC).
Generation IV reactors: reactor concepts
International Nuclear Information System (INIS)
Cardonnier, J.L.; Dumaz, P.; Antoni, O.; Arnoux, P.; Bergeron, A.; Renault, C.; Rimpault, G.; Delpech, M.; Garnier, J.C.; Anzieu, P.; Francois, G.; Lecomte, M.
2003-01-01
Liquid metal reactor concept looks promising because of its hard neutron spectrum. Sodium reactors benefit a large feedback experience in Japan and in France. Lead reactors have serious assets concerning safety but they require a great effort in technological research to overcome the corrosion issue and they lack a leader country to develop this innovative technology. In molten salt reactor concept, salt is both the nuclear fuel and the coolant fluid. The high exit temperature of the primary salt (700 Celsius degrees) allows a high energy efficiency (44%). Furthermore molten salts have interesting specificities concerning the transmutation of actinides: they are almost insensitive to irradiation damage, some salts can dissolve large quantities of actinides and they are compatible with most reprocessing processes based on pyro-chemistry. Supercritical water reactor concept is based on operating temperature and pressure conditions that infers water to be beyond its critical point. In this range water gets some useful characteristics: - boiling crisis is no more possible because liquid and vapour phase can not coexist, - a high heat transfer coefficient due to the low thermal conductivity of supercritical water, and - a high global energy efficiency due to the high temperature of water. Gas-cooled fast reactors combining hard neutron spectrum and closed fuel cycle open the way to a high valorization of natural uranium while minimizing ultimate radioactive wastes and proliferation risks. Very high temperature gas-cooled reactor concept is developed in the prospect of producing hydrogen from no-fossil fuels in large scale. This use implies a reactor producing helium over 1000 Celsius degrees. (A.C.)
The Simulator Development for RDE Reactor
Subekti, Muhammad; Bakhri, Syaiful; Sunaryo, Geni Rina
2018-02-01
BATAN is proposing the construction of experimental power reactor (RDE reactor) for increasing the public acceptance on NPP development plan, proofing the safety level of the most advanced reactor by performing safety demonstration on the accidents such as Chernobyl and Fukushima, and owning the generation fourth (G4) reactor technology. For owning the reactor technology, the one of research activities is RDE’s simulator development that employing standard equation. The development utilizes standard point kinetic and thermal equation. The examination of the simulator carried out comparison in which the simulation’s calculation result has good agreement with assumed parameters and ChemCAD calculation results. The transient simulation describes the characteristic of the simulator to respond the variation of power increase of 1.5%/min, 2.5%/min, and 3.5%/min.
Research reactors - an overview
International Nuclear Information System (INIS)
West, C.D.
1997-01-01
A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs
International Nuclear Information System (INIS)
Al-Chalabi, R.M.; Turinsky, P.J.; Faure, F.-X.; Sarsour, H.N.; Engrand, P.R.
1993-01-01
The NESTLE nodal kinetics code has been developed for utilization as a stand-alone code for steady-state and transient reactor neutronic analysis and for incorporation into system transient codes, such as TRAC and RELAP. The latter is desirable to increase the simulation fidelity over that obtained from currently employed zero- and one-dimensional neutronic models and now feasible due to advances in computer performance and efficiency of nodal methods. As a stand-alone code, requirements are that it operate on a range of computing platforms from memory-limited personal computers (PCs) to supercomputers with vector processors. This paper summarizes the features of NESTLE that reflect the utilization and requirements just noted
Physical measurements in Marcoule reactors (1962)
International Nuclear Information System (INIS)
Teste du Bailler, A.
1962-01-01
A brief description of the physical measurements in Marcoule reactors is given here. During commissioning and subsequent years of operation, various experiments ha been carried out to check design data, and improve the operating conditions and also test theoretical models for kinetic studies. (author) [fr
Reactivity transient calculatios in research reactor
International Nuclear Information System (INIS)
Santos, R.S. dos
1986-01-01
A digital program for reactivity transient analysis in research reactor and cylindrical geometry was showed quite efficient when compared with methods and programs of the literature, as much in the solution of the neutron kinetics equation as in the thermohydraulic. An improvement in the representation of the feedback reactivity adopted on the program reduced markedly the computation time, with some accuracy. (Author) [pt
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)
Energy Technology Data Exchange (ETDEWEB)
Wintergerst, M.
2009-05-15
For the Jules Horowitz new material-testing reactor (JHR), an aluminium base alloy, called AlFeNi, will be used for the cladding of the fuel plates. This alloy (Al - 1% Fe - 1% Ni - 1 % Mg), which is already used as fuel cladding, was developed for its good corrosion resistance in water at high temperatures. However, few studies dealing with the alteration process in water and the relationships with irradiation effects have been performed on this alloy. The conception of the JHR fuel requires a better knowledge of the corrosion mechanisms. Corrosion tests were performed in autoclaves at 70 C, 165 C and 250 C on AlFeNi plates representative of the fuel cladding. Several techniques were used to characterize the corrosion scale: SEM, TEM, EPMA, XRD, Raman spectroscopy. Our observations show that the corrosion scale is made of two main layers: a dense amorphous scale close to the metal and a porous crystalline scale in contact with the water. More than the morphology, the chemical compositions of both layers are different. This duplex structure results from a mixed growth mechanism: an anionic growth to develop the inner oxide and a cationic diffusion followed by a dissolution-precipitation process to form the outer one. Dynamic experiments at 70 C and corrosion kinetics measurements have demonstrated that the oxide growth process is controlled by a diffusion step associated to a dissolution/precipitation process. A corrosion mechanism of the AlFeNi alloy in aqueous media has been proposed. Then post-irradiation exams performed on irradiated fuel plates were used to investigate the effects of the irradiation on the corrosion behaviour in the reactor core. (author)
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 temperature distribution in a gas core fission reactor
Energy Technology Data Exchange (ETDEWEB)
Hoogenboom, J.E.; Dam, H. van; Kuijper, J.C. (Interuniversitair Reactor Inst., Delft (Netherlands)); Kistemaker, J.; Boersma-Klein, W.; Vitalis, F. (FOM-Instituut voor Atoom-en Molecuulfysica, Amsterdam (Netherlands))
1991-01-01
A model is proposed for the heat transport in a nuclear reactor with gaseous fuel at high temperatures taking into account radiative and kinetic heat transfer. A derivation is given of the equation determining the temperature distribution in a gas core reactor and different numerical solution methods are discussed in detail. Results are presented of the temperature distribution. The influence of the kinetic heat transport and of dissociation of the gas molecules is shown. Also discussed is the importance of the temperature gradient at the reactor wall and its dependence on system parameters. (author).
The temperature distribution in a gas core fission reactor
International Nuclear Information System (INIS)
Hoogenboom, J.E.; Dam, H. van; Kuijper, J.C.; Kistemaker, J.; Boersma-Klein, W.; Vitalis, F.
1991-01-01
A model is proposed for the heat transport in a nuclear reactor with gaseous fuel at high temperatures taking into account radiative and kinetic heat transfer. A derivation is given of the equation determining the temperature distribution in a gas core reactor and different numerical solution methods are discussed in detail. Results are presented of the temperature distribution. The influence of the kinetic heat transport and of dissociation of the gas molecules is shown. Also discussed is the importance of the temperature gradient at the reactor wall and its dependence on system parameters. (author)
Study on transient of fluidized bed nuclear reactor
International Nuclear Information System (INIS)
Streck, E.E.
1988-01-01
The point kinetic equations for a Fluidized-Bed Nuclear Reactor are solved by the method of Hansen. Due to the time varying nature of the reactor volume, the equations have a non-conventional formulation (moving boundary problem), but the method of solution preserves its asymptotic convergence and efficiency characteristics under this formulation. A one dimensional and linearized thermal hydraulics feedback model was coupled to the point kinetic equations in order to obtain a more realistic representation of the reactor power. The resulting equations are solved by the Euler explicit method. (author)
International Nuclear Information System (INIS)
Maudlin, P.J.
1979-01-01
A theoretical methodology describing the time dependent growth of large populations of nuclear power reactors of different types is pursued. This methodology is based on the apparent close analogy between the time dependent variations of neutrons and of fuel in nuclear reactors. Methods for the realistic projection of reactor populations, as they develop in a reactor park, are provided using the point park model as kernel in a superposition of reactor deployment elements that form a realistic park scenario. Typical deployment strategy results are presented illustrating the theoretical and computational advantages of the point park model methodology
Research reactors; Les piles de recherche
Energy Technology Data Exchange (ETDEWEB)
Kowarski, L [Commissariat a l' Energie Atomique, Paris (France). Centre d' Etudes Nucleaires; [Organisation europeenne pour la Recherche Nucleaire, Geneve (Switzerland)
1955-07-01
It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing
Research reactors; Les piles de recherche
Energy Technology Data Exchange (ETDEWEB)
Kowarski, L. [Commissariat a l' Energie Atomique, Paris (France). Centre d' Etudes Nucleaires]|[Organisation europeenne pour la Recherche Nucleaire, Geneve (Switzerland)
1955-07-01
It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing
A novel fractional technique for the modified point kinetics equations
Directory of Open Access Journals (Sweden)
Ahmed E. Aboanber
2016-10-01
Full Text Available A fractional model for the modified point kinetics equations is derived and analyzed. An analytical method is used to solve the fractional model for the modified point kinetics equations. This methodical technique is based on the representation of the neutron density as a power series of the relaxation time as a small parameter. The validity of the fractional model is tested for different cases of step, ramp and sinusoidal reactivity. The results show that the fractional model for the modified point kinetics equations is the best representation of neutron density for subcritical and supercritical reactors.
AUS diffusion module POW checkout - 1- and 2-dimensional kinetics calculations
International Nuclear Information System (INIS)
Pollard, J.P.
1977-01-01
POW is the diffusion module 'workhorse' of the AUS reactor neutronics modular code system; its steady state calculations have been checked out against other diffusion codes (particularly CRAM and GOG). Checkout of kinetic aspects, however, is difficult as kinetic codes are not freely available. In this report POW has been checked against three benchmark calculations as well as a calculation on the 100 KW Argonaut reactor Moata. (author)
Energy Technology Data Exchange (ETDEWEB)
Prescott, R F; George, B V; Baglin, C J
1978-05-10
Reference is made to thermal insulation on the inner surfaces of containment vessels of fluid cooled nuclear reactors and particularly in situations where the thermal insulation must also serve a structural function and transmit substantial load forces to the surface which it covers. An arrangement is described that meets this requirement and also provides for core support means that favourably influences the flow of hot coolant from the lower end of the core into a plenum space in the hearth of the reactor. The arrangement comprises a course of thermally insulating bricks arranged as a mosaic covering a wall of the reactor and a course of thermally insulating tiles arranged as a mosaic covering the course of bricks. Full constructional details are given.
International Nuclear Information System (INIS)
Prescott, R.F.; George, B.V.; Baglin, C.J.
1978-01-01
Reference is made to thermal insulation on the inner surfaces of containment vessels of fluid cooled nuclear reactors and particularly in situations where the thermal insulation must also serve a structural function and transmit substantial load forces to the surface which it covers. An arrangement is described that meets this requirement and also provides for core support means that favourably influences the flow of hot coolant from the lower end of the core into a plenum space in the hearth of the reactor. The arrangement comprises a course of thermally insulating bricks arranged as a mosaic covering a wall of the reactor and a course of thermally insulating tiles arranged as a mosaic covering the course of bricks. Full constructional details are given. (UK)
McCarty, Perry L.; Bachmann, Andre
1992-01-01
A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.
Kinetic parameters for source driven systems
International Nuclear Information System (INIS)
Dulla, S.; Ravetto, P.; Carta, M.; D'Angelo, A.
2006-01-01
The definition of the characteristic kinetic parameters of a subcritical source-driven system constitutes an interesting problem in reactor physics with important consequences for practical applications. Consistent and physically meaningful values of the parameters allow to obtain accurate results from kinetic simulation tools and to correctly interpret kinetic experiments. For subcritical systems a preliminary problem arises for the adoption of a suitable weighting function to be used in the projection procedure to derive a point model. The present work illustrates a consistent factorization-projection procedure which leads to the definition of the kinetic parameters in a straightforward manner. The reactivity term is introduced coherently with the generalized perturbation theory applied to the source multiplication factor ks, which is thus given a physical role in the kinetic model. The effective prompt lifetime is introduced on the assumption that a neutron generation can be initiated by both the fission process and the source emission. Results are presented for simplified configurations to fully comprehend the physical features and for a more complicated highly decoupled system treated in transport theory. (authors)
A kinetic model for hydrodesulfurisation
Energy Technology Data Exchange (ETDEWEB)
Sau, M.; Narasimhan, C.S.L.; Verma, R.P. [Indian Oil Corporation Limited, Research and Development Centre, Faridabad (India)
1997-07-01
Due to stringent environmental considerations and related insistence on low sulfur fuels, hydrodesulfurisation has emerged as an important component of any refining scheme globally. The process is used ranging from Naphta/Kerosine hydrotreating to heavy oil hydrotreating. Processes such as Deep gas oil desulfurisation aiming at reduction of sulfur levels to less than 500 ppm have emerged as major players in the scenario. Hydrodesulfurisation (HDS) involves parallel desulfurisation of different organo-sulfur compounds present in the complex petroleum mixtures. In order to design, monitor, optimise and control the HDS reactor, it is necessary to have a detailed, yet simple model which follows the reaction chemistry accurately. In the present paper, a kinetic model is presented for HDS using continuum theory of lumping. The sulfur distribution in the reaction mixture is treated as continuum and parallel reaction networks are devised for kinetic modelling using continuum theory of lumping approach. The model based on the above approach follows the HDS chemistry reasonably well and hence the model parameters are almost feed invariant. Methods are also devised to incorporate heat and pressure effects into the model. The model has been validated based on commercial kero-HDS data. It is found that the model predictions agree with the experimental/commercial data. 17 refs.
International Nuclear Information System (INIS)
Scholz, M.
1976-01-01
An improvement of the accessibility of that part of a nuclear reactor serving for biological shield is proposed. It is intended to provide within the biological shield, distributed around the circumference of the reactor pressure vessel, several shielding chambers filled with shielding material, which are isolated gastight from the outside by means of glass panes with a given bursting strength. It is advantageous that, on the one hand, inspection and maintenance will be possible without great effort and, on the other, a large relief cross section will be at desposal if required. (UWI) [de
Wigner, E.P.; Weinberg, A.W.; Young, G.J.
1958-04-15
A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.
Methane production from cellulosic wastes: kinetic studies
Energy Technology Data Exchange (ETDEWEB)
Goma, G; De La Torre, I; Maugheri, F; Yameogo, T
1979-09-01
The anaerobic fermentation is studied on lignocellulosic materials and sucrose as substrate. With the lignocellulosic material cellulolysis is rate limiting. The reducing sugar concentration in the fermentation broth is less than 15 mgl/sup -1/. With rumen bacteria, 50% of the initial polyoside are used and 12% of the lignin is solubilized. With sucrose studies on the kinetic behaviour of the mixed population responsible of the acidogenesis step permite to find the optimal temperature (40/sup 0/C) feed substrate concentration (50 gl/sup -1/) residence time (60 hours) and pH (5,5 - 6). The better technology is a plug flow fermentor with cell recycling. Modeling of the behaviour of this reactor was performed. The use of this reactor for the selection of acido resistant bacteria is discussed. For methanogenesis, in continuous culture, the ideal technology seems to be an association of two reactors. In the first step, a plug flow reactor must be used for acidogenesis, and in the second step a well mixed reactor permit the conversion of organic acids in methane.
Principles of chemical kinetics
House, James E
2007-01-01
James House's revised Principles of Chemical Kinetics provides a clear and logical description of chemical kinetics in a manner unlike any other book of its kind. Clearly written with detailed derivations, the text allows students to move rapidly from theoretical concepts of rates of reaction to concrete applications. Unlike other texts, House presents a balanced treatment of kinetic reactions in gas, solution, and solid states. The entire text has been revised and includes many new sections and an additional chapter on applications of kinetics. The topics covered include quantitative rela
Introduction to chemical kinetics
Soustelle, Michel
2013-01-01
This book is a progressive presentation of kinetics of the chemical reactions. It provides complete coverage of the domain of chemical kinetics, which is necessary for the various future users in the fields of Chemistry, Physical Chemistry, Materials Science, Chemical Engineering, Macromolecular Chemistry and Combustion. It will help them to understand the most sophisticated knowledge of their future job area. Over 15 chapters, this book present the fundamentals of chemical kinetics, its relations with reaction mechanisms and kinetic properties. Two chapters are then devoted to experimental re
International Nuclear Information System (INIS)
Wende, C.W.J.
1976-01-01
The method of operating a water-cooled neutronic reactor having a graphite moderator is described which comprises flowing a gaseous mixture of carbon dioxide and helium, in which the helium comprises 40--60 volume percent of the mixture, in contact with the graphite moderator. 2 claims, 4 figures
International Nuclear Information System (INIS)
Wende, C.W.J.
1976-01-01
A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield
International Nuclear Information System (INIS)
Suzuki, Hiroaki; Murase, Michio; Yokomizo, Osamu.
1997-01-01
The present invention provides a BWR type reactor facility capable of suppressing the amount of steams generated by the mutual effect of a failed reactor core and coolants upon occurrence of an imaginal accident, and not requiring spacial countermeasures for enhancing the pressure resistance of the container vessel. Namely, a means for supplying cooling water at a temperature not lower by 30degC than the saturated temperature corresponding to the inner pressure of the containing vessel upon occurrence of an accident is disposed to a lower dry well below the pressure vessel. As a result, upon occurrence of such an accident that the reactor core should be melted and flown downward of the pressure vessel, when cooling water at a temperature not lower than the saturated temperature, for example, cooling water at 100degC or higher is supplied to the lower dry well, abrupt generation of steams by the mutual effect of the failed reactor core and cooling water is scarcely caused compared with a case of supplying cooling water at a temperature lower than the saturation temperature by 30degC or more. Accordingly, the amount of steams to be generated can be suppressed, and special countermeasure is no more necessary for enhancing the pressure resistance of the container vessel is no more necessary. (I.S.)
International Nuclear Information System (INIS)
Gilroy, J.E.
1980-01-01
An improved cover structure for liquid metal cooled fast breeder type reactors is described which it is claimed reduces the temperature differential across the intermediate grid plate of the core cover structure and thereby reduces its subjection to thermal stresses. (UK)
International Nuclear Information System (INIS)
Harvie, J.D.
2002-01-01
This presentation discusses reactor licensing and includes the legislative basis for licensing, other relevant legislation , the purpose of the Nuclear Safety and Control Act, important regulations, regulatory document, policies, and standards. It also discusses the role of the CNSC, its mandate and safety philosophy
International Nuclear Information System (INIS)
Hattori, Sadao; Sekine, Katsuhisa.
1987-01-01
Purpose: To decrease the thickness of a reactor container and reduce the height and the height and plate thickness of a roof slab without using mechanical vibration stoppers. Constitution: Earthquake proofness is improved by filling fluids such as liquid metal between a reactor container and a secondary container and connecting the outer surface of the reactor container with the inner surface of the secondary container by means of bellows. That is, for the horizontal seismic vibrations, horizontal loads can be supported by the secondary container without providing mechanical vibration stoppers to the reactor container and the wall thickness can be reduced thereby enabling to simplify thermal insulation structure for the reduction of thermal stresses. Further, for the vertical seismic vibrations, verical loads can be transmitted to the secondary container thereby enabling to reduce the wall thickness in the same manner as for the horizontal load. By the effect of transferring the point of action of the container load applied to the roof slab to the outer circumferential portion, the intended purpose can be attained and, in addition, the radiation dose rate at the upper surface of the roof slab can be decreased. (Kamimura, M.)
International Nuclear Information System (INIS)
Matsuura, Tetsuaki; Nomura, Teiji; Tokunaga, Kensuke; Okuda, Shin-ichi
1990-01-01
Fuel assemblies in the portions where the gradient of fast neutron fluxes between two opposing faces of a channel box is great are kept loaded at the outermost peripheral position of the reactor core also in the second operation cycle in the order to prevent interference between a control rod and the channel box due to bending deformation of the channel box. Further, the fuel assemblies in the second row from the outer most periphery in the first operation cycle are also kept loaded at the second row in the second operation cycle. Since the gradient of the fast neutrons in the reactor core is especially great at the outer circumference of the reactor core, the channel box at the outer circumference is bent such that the surface facing to the center of the reactor core is convexed and the channel box in the second row is also bent to the identical direction, the insertion of the control rod is not interfered. Further, if the positions for the fuels at the outermost periphery and the fuels in the second row are not altered in the second operation cycle, the gaps are not reduced to prevent the interference between the control rod and the channel box. (N.H.)
Cybulskis, Viktor J.; Smeltz, Andrew D.; Zvinevich, Yury; Gounder, Rajamani; Delgass, W. Nicholas; Ribeiro, Fabio H.
2016-01-01
Understanding catalytic chemistry, collecting and interpreting kinetic data, and operating chemical reactors are critical skills for chemical engineers. This laboratory experiment provides students with a hands-on supplement to a course in chemical kinetics and reaction engineering. The oxidation of methane with a palladium catalyst supported on…
Kinetics of the Reaction of CO2 with Aqueous Potassium Salt of Taurine and Glycine
Kumar, P.S.; Hogendoorn, J.A.; Versteeg, G.F.; Feron, P.H.M.
2003-01-01
The kinetics of the reaction between CO2 and aqueous potassium salts of taurine and glycine was measured at 295 K in a stirred-cell reactor with a flat gas–liquid interface. For aqueous potassium taurate solutions, the temperature effect on the reaction kinetics was measured at 285 and 305 K. Unlike
Kinetics of the reaction of CO2 with aqueous potassium salt of taurine and glycine
Kumar Paramasivam Senthil, P.S.; Hogendoorn, Kees; Versteeg, Geert; Feron, P.H.M.
2003-01-01
The kinetics of the reaction between CO2 and aqueous potassium salts of taurine and glycine was measured at 295 K in a stirred-cell reactor with a flat gas-liquid interface. For aqueous potassium taurate solutions, the temperature effect on the reaction kinetics was measured at 285 and 305 K. Unlike
Kinetics of the gas-solid hydroxyethylation of potato starch
Kuipers, N.J M; Beenackers, A.A C M
The kinetics of the reaction between gaseous ethylene oxide and semidry granular potato starch was studied in a pressure-controlled semibatch reactor with and without impregnation of the starch with the catalyst sodium hydroxide. Four parallel reactions are involved: the catalyzed (with reaction
Kinetic and Thermodynamic Studies on Adsorption of Sulphate from ...
African Journals Online (AJOL)
DELL USER
22, No. 1, 2017. 39. Kinetic and Thermodynamic Studies on Adsorption of Sulphate from Aqueous Solution by. Magnetite ... poison catalysts, and affect the .... C for 1 h in a stainless steel reactor placed in a furnace ... N2 gas for 30 min. 50 ml of ...... adsorption for designing and evaluating the ... is the equilibrium liquid-phase.
A model for nuclear research reactor dynamics
Energy Technology Data Exchange (ETDEWEB)
Barati, Ramin, E-mail: Barati.ramin@aut.ac.ir; Setayeshi, Saeed, E-mail: setayesh@aut.ac.ir
2013-09-15
Highlights: • A thirty-fourth order model is used to simulate the dynamics of a research reactor. • We consider delayed neutrons fraction as a function of time. • Variable fuel and temperature reactivity coefficients are used. • WIMS, BORGES and CITVAP codes are used for initial condition calculations. • Results are in agreement with experimental data rather than common codes. -- Abstract: In this paper, a useful thirty-fourth order model is presented to simulate the kinetics and dynamics of a research reactor core. The model considers relevant physical phenomena that govern the core such as reactor kinetics, reactivity feedbacks due to coolant and fuel temperatures (Doppler effects) with variable reactivity coefficients, xenon, samarium, boron concentration, fuel burn up and thermal hydraulics. WIMS and CITVAP codes are used to extract neutron cross sections and calculate the initial neuron flux respectively. The purpose is to present a model with results similar to reality as much as possible with reducing common simplifications in reactor modeling to be used in different analyses such as reactor control, functional reliability and safety. The model predictions are qualified by comparing with experimental data, detailed simulations of reactivity insertion transients, and steady state for Tehran research reactor reported in the literature and satisfactory results have been obtained.
International Nuclear Information System (INIS)
Egorenkov, P.M.
2001-01-01
The multi-purpose research reactor MAPLE (Canada) and concept of new reactor MAPLE-CNF as will substitute the known Canadian research reactor NRU are described. New reactor will be used as contributor for investigations into materials, neutron beams and further developments for the CANDU type reactor. The Budapest research reactor (BRR) and its application after the last reconstruction are considered also [ru
Energy Technology Data Exchange (ETDEWEB)
Plas, R.
1962-07-01
The author reports a study on kinetics equations for a reactor. He uses the conventional form of these equations but by using a dynamic multiplication factor. Thus, constants related to delayed neutrons are not modified by efficiency factors. The author first describes the theoretic kinetic operation of a reactor and develops the associated equations. He reports the development of equations for multiplication factors.
Kinetic equation solution by inverse kinetic method
International Nuclear Information System (INIS)
Salas, G.
1983-01-01
We propose a computer program (CAMU) which permits to solve the inverse kinetic equation. The CAMU code is written in HPL language for a HP 982 A microcomputer with a peripheral interface HP 9876 A ''thermal graphic printer''. The CAMU code solves the inverse kinetic equation by taking as data entry the output of the ionization chambers and integrating the equation with the help of the Simpson method. With this program we calculate the evolution of the reactivity in time for a given disturbance
Energy Technology Data Exchange (ETDEWEB)
Corella, J; Narvaez, I; Orio, A [Madrid Univ. (Spain). Dept. of Chem. Eng.
1997-12-31
Calcined dolomites and commercial steam reforming catalysts are used downstream biomass gasifiers for hot catalytic raw gas cleaning. To further compare these solids under a rigorous basis, a reaction network and a kinetic model are presented. The apparent kinetic constant for the tar reduction is here proposed as a basis of comparison. Tar sampling and analysis, and the units used for the space-time in the catalytic reactor affect the kinetic constants observed. (author) (2 refs.)
Energy Technology Data Exchange (ETDEWEB)
Corella, J.; Narvaez, I.; Orio, A. [Madrid Univ. (Spain). Dept. of Chem. Eng.
1996-12-31
Calcined dolomites and commercial steam reforming catalysts are used downstream biomass gasifiers for hot catalytic raw gas cleaning. To further compare these solids under a rigorous basis, a reaction network and a kinetic model are presented. The apparent kinetic constant for the tar reduction is here proposed as a basis of comparison. Tar sampling and analysis, and the units used for the space-time in the catalytic reactor affect the kinetic constants observed. (author) (2 refs.)
Reactor core of nuclear reactor
International Nuclear Information System (INIS)
Sasagawa, Masaru; Masuda, Hiroyuki; Mogi, Toshihiko; Kanazawa, Nobuhiro.
1994-01-01
In a reactor core, a fuel inventory at an outer peripheral region is made smaller than that at a central region. Fuel assemblies comprising a small number of large-diameter fuel rods are used at the central region and fuel assemblies comprising a great number of smalldiameter fuel rods are used at the outer peripheral region. Since a burning degradation rate of the fuels at the outer peripheral region can be increased, the burning degradation rate at the infinite multiplication factor of fuels at the outer region can substantially be made identical with that of the fuels in the inner region. As a result, the power distribution in the direction of the reactor core can be flattened throughout the entire period of the burning cycle. Further, it is also possible to make the degradation rate of fuels at the outer region substantially identical with that of fuels at the inner side. A power peak formed at the outer circumferential portion of the reactor core of advanced burning can be lowered to improve the fuel integrity, and also improve the reactor safety and operation efficiency. (N.H.)
Proceedings of the 1992 topical meeting on advances in reactor physics. Volume 2
Energy Technology Data Exchange (ETDEWEB)
1992-04-01
This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements & Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)
Energy Technology Data Exchange (ETDEWEB)
Milosevic, M; Stefanovic, D; Pesic, M; Popovic, D; Nikolic, D; Antic, D; Zavaljevski, N [Institut za nuklearne nauke Boris Kidric, Vinca, Beograd (Yugoslavia)
1987-07-01
The concept, purpose and preliminary design of a zero power fast reactor LASTA are described. The methods of computing the reactor core parameters and reactor kinetics are presented with the basic calculated results and analysis for one selected LASTA configuration. The nominal parameters are determined according to the selected reactor safety criteria and results of calculations. Important aspects related to the overall safety are examined in detail. (author)
Proceedings of the 1992 topical meeting on advances in reactor physics
International Nuclear Information System (INIS)
1992-01-01
This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements ampersand Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)
DEFLUORIDATION OF DRINKING WATER BY ELECTROCOAGULATION/ELECTROFLOTATION - KINETIC STUDY
Directory of Open Access Journals (Sweden)
Bennajah Mounir
2010-06-01
Full Text Available A variable order kinetic (VOK model derived from the langmuir-freundlish equation was applied to determine the kinetics of fluoride removal reaction by electrocoagulation (EC. Synthetic solutions were employed to elucidate the effects of the initial fluoride concentration, the applied current and the initial acidity on the simulation results of the model. The proposed model successfully describes the fluoride removal in Airlift reactor in comparison with the experimental results. In this study two EC cells with the same capacity (V = 20 L were used to carry out fluoride removal with aluminum electrodes, the first is a stirred tank reactor (STR the second is an airlift reactor (ALR. The comparison of energy consumption demonstrates that the (ALR is advantageous for carrying out the defluoridation removal process.
Kinetics in radiation chemistry
International Nuclear Information System (INIS)
Hummel, A.
1987-01-01
In this chapter the authors first briefly review the kinetics of first- and second-order processes for continuous and pulsed irradiation, without taking the effects of nonhomogeneous formation of the species into consideration. They also discuss diffusion controlled reactions under conditions where interactions of more than two particles can be neglected, first the kinetics of the diffusion-controlled reaction of randomly generated species (homogeneous reaction) and then that of isolated pairs of reactants. The latter is often called geminate kinetics when dealing with pairs of oppositely charged species; they shall use this term for the kinetics of isolated pairs in general. In the last section they discuss briefly the kinetics of groups of more than two reactants
Method and program for complex calculation of heterogeneous reactor
International Nuclear Information System (INIS)
Kalashnikov, A.G.; Glebov, A.P.; Elovskaya, L.F.; Kuznetsova, L.I.
1988-01-01
An algorithm and the GITA program for complex one-dimensional calculation of a heterogeneous reactor which permits to conduct calculations for the reactor and its cell simultaneously using the same algorithm are described. Multigroup macrocross sections for reactor zones in the thermal energy range are determined according to the technique for calculating a cell with complicate structure and then the continuous multi group calculation of the reactor in the thermal energy range and in the range of neutron thermalization is made. The kinetic equation is solved using the Pi- and DSn- approximations [fr
Non-kinetic capabilities: complementing the kinetic prevalence to targeting
Ducheine, P.
2014-01-01
Targeting is used in military doctrine to describe a military operational way, using (military) means to influence a target (or addressee) in order to achieve designated political and/or military goals. The four factors italicized are used to analyse non-kinetic targeting, complementing our knowledge and understanding of the kinetic prevalence. Paradoxically, non-kinetic targeting is not recognized as a separate concept: kinetic and non-kinetic are intertwined facets of targeting. Kinetic tar...
International Nuclear Information System (INIS)
Gibbons, J.F.; McLaughlin, D.J.
1978-01-01
In the pressure vessel of the water-cooled nuclear reactor there is provided an internal flange on which the one- or two-part core barrel is hanging by means of an external flange. A cylinder is extending from the reactor vessel closure downwards to a seat on the core cupport structure and serves as compression element for the transmission of the clamping load from the closure head to the core barrel (upper guide structure). With the core barrel, subject to tensile stress, between the vessel internal flange and its seat on one hand and the compression of the cylinder resp. hold-down element between the closure head and the seat on the other a very strong, elastic sprung structure is obtained. (DG) [de
International Nuclear Information System (INIS)
Sasaki, Tomozo.
1987-01-01
Purpose: To improve the nuclear reactor availability by enabling to continuously exchange fuels in the natural-slightly enriched uranium region during operation. Constitution: A control rod is withdrawn to the midway of a highly enriched uranium region by means of control rod drives and the highly enriched uranium region is burnt to maintain the nuclear reactor always at a critical state. At the same time, fresh uranium-slightly enriched uranium is continuously supplied gravitationally from a fresh fuel reservoir through fuel reservoir to each of fuel pipes in the natural-slightly enriched uranium region. Then, spent fuels reduced with the reactivity by the burn up are successively taken out from the bottom of each of the fuel pipes through an exit duct and a solenoid valve to the inside of a spent fuel reservoir and the burn up in the natural-slightly enriched uranium region is conducted continuously. (Kawakami, Y.)
International Nuclear Information System (INIS)
Sakurai, Mikio; Yamauchi, Koki.
1983-01-01
Purpose: To improve the channel stability and the reactor core stability in a spontaneous circulation state of coolants. Constitution: A reactor core stabilizing device comprising a differential pressure automatic ON-OFF valve is disposed between each of a plurality of jet pumps arranged on a pump deck. The stabilizing device comprises a piston exerted with a pressure on the lower side of the pump deck by way of a pipeway and a valve for flowing coolants through the bypass opening disposed to the pump deck by the opening and closure of the valve ON-OFF. In a case where the jet pumps are stopped, since the differential pressure between the upper and the lower sides of the pump deck is removed, the valve lowers gravitationally into an opened state, whereby the coolants flow through the bypass opening to increase the spontaneous circulation amount thereby improve the stability. (Yoshino, Y.)
International Nuclear Information System (INIS)
Aleite, W.; Bock, H.W.; Struensee, S.
1976-01-01
The invention concerns the use of burnable poisons in a nuclear reactor, especially in PWRs, in order to improve the controllability of the reactor. An unsymmetrical arrangement in the lattice is provided, if necessary also by insertion of special rods for these additions. It is proposed to arrange the burnable poisons in fuel elements taken over from a previous burn-up cycle and to distribute them, going out from the side facing the control rods, over not more than 20% of the lenth of the fuel elements. It seems sufficient, for the burnable poisons to bind an initial reactivity of only 0.1% and to become ineffective after normal operation of 3 to 4 months. (ORU) [de
International Nuclear Information System (INIS)
Ichiki, Tadaharu; Saba, Kazuhisa.
1979-01-01
Purpose: To improve the earthquake resistance as well as reduce the size of a container for a nuclear reactor with no adverse effects on the decrease of impact shock to the container and shortening of construction step. Constitution: Reinforcing profile steel materials are welded longitudinally and transversely to the inner surface of a container, and inner steel plates are secured to the above profile steel materials while keeping a gap between the materials and the container. Reactor shielding wall planted to the base concrete of the container is mounted to the pressure vessel, and main steam pipeways secured by the transverse beams and led to the outside of container is connected. This can improve the rigidity earthquake strength and the safetiness against the increase in the inside pressure upon failures of the container. (Yoshino, Y.)
International Nuclear Information System (INIS)
Oyamada, Osamu; Furukawa, Hideyasu; Uozumi, Hiroto.
1979-01-01
Purpose: To lower the position of an intermediate slab within a reactor container and fitting a heat insulating material to the inner wall of said intermediate slab, whereby a space for a control rod exchanging device and thermal stresses of the inner peripheral wall are lowered. Constitution: In the pedestal at the lower part of a reactor pressure vessel there is formed an intermediate slab at a position lower than diaphragm floor slab of the outer periphery of the pedestal thereby to secure a space for providing automatic exchanging device of a control rod driving device. Futhermore, a heat insulating material is fitted to the inner peripheral wall at the upper side of the intermediate slab part, and the temperature gradient in the wall thickness direction at the time of a piping rupture trouble is made gentle, and thermal stresses at the inner peripheral wall are lowered. (Sekiya, K.)
International Nuclear Information System (INIS)
Lewis, W.R.
1978-01-01
Disclosed is a graphite-moderated, water-cooled nuclear reactor including a plurality of rectangular graphite blocks stacked in abutting relationship in layers, alternate layers having axes which are normal to one another, alternate rows of blocks in alternate layers being provided with a channel extending through the blocks, said channeled blocks being provided with concave sides and having smaller vertical dimensions than adjacent blocks in the same layer, there being nuclear fuel in the channels
International Nuclear Information System (INIS)
Humphreys, P.; Davidson, D.F.; Thatcher, G.
1980-01-01
The cooling system of a liquid metal cooled fast breeder nuclear reactor of the pool kind is described. It has an intermediate heat exchange module comprising a tube-in-shell heat exchanger and an electromagnetic flow coupler in the base region of the module. Primary coolant is flowed through the heat exchanger being driven by electromagnetic interaction with secondary liquid metal coolant flow effected by a mechanical pump. (author)
International Nuclear Information System (INIS)
Jungmann, A.
1975-01-01
Between a PWR's reactor pressure vessel made of steel and the biological shield made of concrete there is a gap. This gap is filled up with a heat insulation facting the reactor pressure vessel, for example with insulating concrete segments jacketed with sheet steel and with an additional layer. This layer serves for smooth absorption of compressive forces originating in radial direction from the reactor pressure vessel. It consists of cylinder-segment shaped bricks made of on situ concrete, for instance. The bricks have cooling agent ports in one or several rows which run parallel to the wall of the pressure vessel and in alignment with superposed bricks. Between the layer of bricks and the biological shield or rather the heat insulation, there are joints which are filled, however, with injected mortar. That guarantees a smooth series of connected components resistant tom compression. Besides, a slip foil can be set between the heat insulation and the joining joint filled with mortar for the reduction of the friction at thermal expansions. (TK) [de
International Nuclear Information System (INIS)
Ebata, Sakae.
1990-01-01
At least one valve rack is disposed in a reactor building, on which pipeways to a main closure valve, valves and bypasses of turbines are placed and contained. The valve rack is fixed to the main body of the building or to a base mat. Since the reactor building is designed as class A earthquake-proofness and for maintaining the S 1 function, the valve rack can be fixed to the building main body or to the base mat. With such a constitution, the portions for maintaining the S 1 function are concentrated to the reactor building. As a result, the dispersion of structures of earthquake-proof portion corresponding to the reference earthquake vibration S 1 can be prevented. Accordingly, the conditions for the earthquake-proof design of the turbine building and the turbine/electric generator supporting rack are defined as only the class B earthquake-proof design conditions. In view of the above, the amount of building materials can be saved and the time for construction can be shortened. (I.S.)
International Nuclear Information System (INIS)
Yoshioka, Michiko.
1985-01-01
Purpose: To obtain an optimum structural arrangement of IRM having a satisfactory responsibility to the inoperable state of a nuclear reactor and capable of detecting the reactor power in an averaged manner. Constitution: As the structural arrangement of IRM, from 6 to 16 even number of IRM are bisected into equial number so as to belong two trip systems respectively, in which all of the detectors are arranged at an equal pitch along a circumference of a circle with a radius rl having the center at the position of the central control rod in one trip system, while one detector is disposed near the central control rod and other detectors are arranged substantially at an equal pitch along the circumference of a circle with a radius r2 having the center at the position for the central control rod in another trip system. Furthermore, the radius r1 and r2 are set such that r1 = 0.3 R, r2 = 0.5 R in the case where there are 6 IRM and r1 = 0.4 R and R2 = 0.8 R where there are eight IRM where R represents the radius of the reactor core. (Kawakami, Y.)
International Nuclear Information System (INIS)
Ryazantsev, E.P.; Egorenkov, P.M.; Nasonov, V.A.; Smimov, A.M.; Taliev, A.V.; Gromov, B.F.; Kousin, V.V.; Lantsov, M.N.; Radchenko, V.P.; Sharapov, V.N.
1998-01-01
The Material Testing Loop Reactor (MLR) development was commenced in 1991 with the aim of updating and widening Russia's experimental base to validate the selected directions of further progress of the nuclear power industry in Russia and to enhance its reliability and safety. The MLR reactor is the pool-type one. As coolant it applies light water and as side reflector beryllium. The direction of water circulation in the core is upward. The core comprises 30 FA arranged as hexagonal lattice with the 90-95 mm pitch. The central materials channel and six loop channels are sited in the core. The reflector includes up to 11 loop channels. The reactor power is 100 MW. The average power density of the core is 0.4 MW/I (maximal value 1.0 MW/l). The maximum neutron flux density is 7.10 14 n/cm 2 s in the core (E>0.1 MeV), and 5.10 14 n/cm 2 s in the reflector (E<0.625 eV). In 1995 due to the lack of funding the MLR designing was suspended. (author)
International Nuclear Information System (INIS)
Shirakawa, Toshihisa.
1979-01-01
Purpose: To prevent cladding tube injuries due to thermal expansion of each of the pellets by successively extracting each of the control rods loaded in the reactor core from those having less number of notches, as well as facilitate the handling work for the control rods. Constitution: A recycle flow control device is provided to a circulation pump for forcibly circulating coolants in the reactor container and an operational device is provided for receiving each of the signals concerning number of notches for each of the control rods and flow control depending on the xenon poisoning effect obtained from the signals derived from the in-core instrument system connected to the reactor core. The operational device is connected with a control rod drive for moving each of the control rods up and down and a recycle flow control device. The operational device is set with a pattern for the aimed control rod power and the sequence of extraction. Upon extraction of the control rods, they are extracted successively from those having less notch numbers. (Moriyama, K.)
International Nuclear Information System (INIS)
Hidaka, Masataka; Hatamiya, Shigeo; Kawasaki, Terufumi; Fukui, Toru; Suzuki, Hiroaki; Kataoka, Yoshiyuki; Kawabe, Ryuhei; Murase, Michio; Naito, Masanori.
1990-01-01
In order to suppress the pressure elevation in a reactor container due to high temperature and high pressure steams jetted out upon pipeway rupture accidents in the reactor container, the steams are introduced to a pressure suppression chamber for condensating them in stored coolants. However, the ability for suppressing the pressure elevation and steam coagulation are deteriorated due to the presence of inactive incondensible gases. Then, there are disposed a vent channel for introducing the steams in a dry well to a pressure suppression chamber in the reactor pressure vessel, a closed space disposed at the position lower than a usual liquid level, a first channel having an inlet in the pressure suppression chamber and an exit in the closed space and a second means connected by way of a backflow checking means for preventing the flow directing to the closed space. The first paths are present by plurality, a portion of which constitutes a syphon. The incondensible gases and the steams are discharged to the dry well at high pressure by using the difference of the water head for a long cooling time after the pipeway rupture accident. Then, safety can be improved without using dynamic equipments as driving source. (N.H.)
Reactor core in FBR type reactor
International Nuclear Information System (INIS)
Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.
1989-01-01
In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)