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

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

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

International Nuclear Information System (INIS)

Ise, Takeharu

1976-12-01

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

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

Empiric model for mean generation time adjustment factor for classic point kinetics equations

Energy Technology Data Exchange (ETDEWEB)

Goes, David A.B.V. de; Martinez, Aquilino S.; Goncalves, Alessandro da C., E-mail: david.goes@poli.ufrj.br, E-mail: aquilino@lmp.ufrj.br, E-mail: alessandro@con.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Departamento de Engenharia Nuclear

2017-11-01

Point reactor kinetics equations are the easiest way to observe the neutron production time behavior in a nuclear reactor. These equations are derived from the neutron transport equation using an approximation called Fick's law leading to a set of first order differential equations. The main objective of this study is to review classic point kinetics equation in order to approximate its results to the case when it is considered the time variation of the neutron currents. The computational modeling used for the calculations is based on the finite difference method. The results obtained with this model are compared with the reference model and then it is determined an empirical adjustment factor that modifies the point reactor kinetics equation to the real scenario. (author)

Empiric model for mean generation time adjustment factor for classic point kinetics equations

International Nuclear Information System (INIS)

Goes, David A.B.V. de; Martinez, Aquilino S.; Goncalves, Alessandro da C.

2017-01-01

Point reactor kinetics equations are the easiest way to observe the neutron production time behavior in a nuclear reactor. These equations are derived from the neutron transport equation using an approximation called Fick's law leading to a set of first order differential equations. The main objective of this study is to review classic point kinetics equation in order to approximate its results to the case when it is considered the time variation of the neutron currents. The computational modeling used for the calculations is based on the finite difference method. The results obtained with this model are compared with the reference model and then it is determined an empirical adjustment factor that modifies the point reactor kinetics equation to the real scenario. (author)

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

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)

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.

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)

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.

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

International Nuclear Information System (INIS)

Jain, V.K.

1989-01-01

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

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

An Explicit Finite Difference scheme for numerical solution of fractional neutron point kinetic equation

International Nuclear Information System (INIS)

Saha Ray, S.; Patra, A.

2012-01-01

Highlights: ► In this paper fractional neutron point kinetic equation has been analyzed. ► The numerical solution for fractional neutron point kinetic equation is obtained. ► Explicit Finite Difference Method has been applied. ► Supercritical reactivity, critical reactivity and subcritical reactivity analyzed. ► Comparison between fractional and classical neutron density is presented. - Abstract: In the present article, a numerical procedure to efficiently calculate the solution for fractional point kinetics equation in nuclear reactor dynamics is investigated. The Explicit Finite Difference Method is applied to solve the fractional neutron point kinetic equation with the Grunwald–Letnikov (GL) definition (). Fractional Neutron Point Kinetic Model has been analyzed for the dynamic behavior of the neutron motion in which the relaxation time associated with a variation in the neutron flux involves a fractional order acting as exponent of the relaxation time, to obtain the best operation of a nuclear reactor dynamics. Results for neutron dynamic behavior for subcritical reactivity, supercritical reactivity and critical reactivity and also for different values of fractional order have been presented and compared with the classical neutron point kinetic (NPK) equation as well as the results obtained by the learned researchers .

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-04-15

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

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

Introduction to the neutron kinetics of nuclear power reactors

CERN Document Server

Tyror, J G; Grant, P J

2013-01-01

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

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

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

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)

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)

Time discretization of the point kinetic equations using matrix exponential method and First-Order Hold

International Nuclear Information System (INIS)

Park, Yujin; Kazantzis, Nikolaos; Parlos, Alexander G.; Chong, Kil To

2013-01-01

Highlights: • Numerical solution for stiff differential equations using matrix exponential method. • The approximation is based on First Order Hold assumption. • Various input examples applied to the point kinetics equations. • The method shows superior useful and effective activity. - Abstract: A system of nonlinear differential equations is derived to model the dynamics of neutron density and the delayed neutron precursors within a point kinetics equation modeling framework for a nuclear reactor. The point kinetic equations are mathematically characterized as stiff, occasionally nonlinear, ordinary differential equations, posing significant challenges when numerical solutions are sought and traditionally resulting in the need for smaller time step intervals within various computational schemes. In light of the above realization, the present paper proposes a new discretization method inspired by system-theoretic notions and technically based on a combination of the matrix exponential method (MEM) and the First-Order Hold (FOH) assumption. Under the proposed time discretization structure, the sampled-data representation of the nonlinear point kinetic system of equations is derived. The performance of the proposed time discretization procedure is evaluated using several case studies with sinusoidal reactivity profiles and multiple input examples (reactivity and neutron source function). It is shown, that by applying the proposed method under a First-Order Hold for the neutron density and the precursor concentrations at each time step interval, the stiffness problem associated with the point kinetic equations can be adequately addressed and resolved. Finally, as evidenced by the aforementioned detailed simulation studies, the proposed method retains its validity and accuracy for a wide range of reactor operating conditions, including large sampling periods dictated by physical and/or technical limitations associated with the current state of sensor and

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.

Application of the Wiener-Hermite functional method to point reactor kinetics driven by random reactivity fluctuations

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

Taylor's series method for solving the nonlinear point kinetics equations

International Nuclear Information System (INIS)

Nahla, Abdallah A.

2011-01-01

Highlights: → Taylor's series method for nonlinear point kinetics equations is applied. → The general order of derivatives are derived for this system. → Stability of Taylor's series method is studied. → Taylor's series method is A-stable for negative reactivity. → Taylor's series method is an accurate computational technique. - Abstract: Taylor's series method for solving the point reactor kinetics equations with multi-group of delayed neutrons in the presence of Newtonian temperature feedback reactivity is applied and programmed by FORTRAN. This system is the couples of the stiff nonlinear ordinary differential equations. This numerical method is based on the different order derivatives of the neutron density, the precursor concentrations of i-group of delayed neutrons and the reactivity. The r th order of derivatives are derived. The stability of Taylor's series method is discussed. Three sets of applications: step, ramp and temperature feedback reactivities are computed. Taylor's series method is an accurate computational technique and stable for negative step, negative ramp and temperature feedback reactivities. This method is useful than the traditional methods for solving the nonlinear point kinetics equations.

A highly accurate algorithm for the solution of the point kinetics equations

International Nuclear Information System (INIS)

Ganapol, B.D.

2013-01-01

Highlights: • Point kinetics equations for nuclear reactor transient analysis are numerically solved to extreme accuracy. • Results for classic benchmarks found in the literature are given to 9-digit accuracy. • Recent results of claimed accuracy are shown to be less accurate than claimed. • Arguably brings a chapter of numerical evaluation of the PKEs to a close. - Abstract: Attempts to resolve the point kinetics equations (PKEs) describing nuclear reactor transients have been the subject of numerous articles and texts over the past 50 years. Some very innovative methods, such as the RTS (Reactor Transient Simulation) and CAC (Continuous Analytical Continuation) methods of G.R. Keepin and J. Vigil respectively, have been shown to be exceptionally useful. Recently however, several authors have developed methods they consider accurate without a clear basis for their assertion. In response, this presentation will establish a definitive set of benchmarks to enable those developing PKE methods to truthfully assess the degree of accuracy of their methods. Then, with these benchmarks, two recently published methods, found in this journal will be shown to be less accurate than claimed and a legacy method from 1984 will be confirmed

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}.

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

International Nuclear Information System (INIS)

Orso, J A

2012-01-01

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

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

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)

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

International Nuclear Information System (INIS)

Kozomara-Maic, S.

1987-06-01

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

Verification of a three-dimensional neutronics model based on multi-point kinetics equations for transient problems

Energy Technology Data Exchange (ETDEWEB)

Park, Kyung Seok; Kim, Hyun Dae; Yeom, Choong Sub [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-07-01

A computer code for solving the three-dimensional reactor neutronic transient problems utilizing multi-point reactor kinetics equations recently developed has been developed. For evaluating its applicability, the code has been tested with typical 3-D LWR and CANDU reactor transient problems. The performance of the method and code has been compared with the results by fine and coarse meshes computer codes employing the direct methods.

Analytical solution of point kinetic equations for sub-critical systems

International Nuclear Information System (INIS)

Henrice Junior, Edson; Goncalves, Alessandro C.

2013-01-01

This article presents an analytical solution for the set of point kinetic equations for sub-critical reactors. This solution stems from the ordinary, non-homogeneous differential equation that rules the neutron density and that presents the incomplete Gamma function in its functional form. The method used proved advantageous and allowed practical applications such as the linear insertion of reactivity, considering an external constant source or with both varying linearly. (author)

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)

Theory of fluctuations and parametric noise in a point nuclear reactor model

International Nuclear Information System (INIS)

Rodriguez, M.A.; San Miguel, M.; Sancho, J.M.

1984-01-01

We present a joint description of internal fluctuations and parametric noise in a point nuclear reactor model in which delayed neutrons and a detector are considered. We obtain kinetic equations for the first moments and define effective kinetic parameters which take into account the effect of parametric Gaussian white noise. We comment on the validity of Langevin approximations for this problem. We propose a general method to deal with weak but otherwise arbitrary non-white parametric noise. Exact kinetic equations are derived for Gaussian non-white noise. (author)

Methods for solving the stochastic point reactor kinetic equations

International Nuclear Information System (INIS)

Quabili, E.R.; Karasulu, M.

1979-01-01

Two new methods are presented for analysis of the statistical properties of nonlinear outputs of a point reactor to stochastic non-white reactivity inputs. They are Bourret's approximation and logarithmic linearization. The results have been compared with the exact results, previously obtained in the case of Gaussian white reactivity input. It was found that when the reactivity noise has short correlation time, Bourret's approximation should be recommended because it yields results superior to those yielded by logarithmic linearization. When the correlation time is long, Bourret's approximation is not valid, but in that case, if one can assume the reactivity noise to be Gaussian, one may use the logarithmic linearization. (author)

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)

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

Generalized treatment of point reactor kinetics driven by random reactivity fluctuations via the Wiener-Hermite functional method

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

The solution of the neutron point kinetics equation with stochastic extension: an analysis of two moments

Energy Technology Data Exchange (ETDEWEB)

Silva, Milena Wollmann da; Vilhena, Marco Tullio M.B.; Bodmann, Bardo Ernst J.; Vasques, Richard, E-mail: milena.wollmann@ufrgs.br, E-mail: vilhena@mat.ufrgs.br, E-mail: bardobodmann@ufrgs.br, E-mail: richard.vasques@fulbrightmail.org [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Programa de Pos-Graduacao em Engenharia Mecanica

2015-07-01

The neutron point kinetics equation, which models the time-dependent behavior of nuclear reactors, is often used to understand the dynamics of nuclear reactor operations. It consists of a system of coupled differential equations that models the interaction between (i) the neutron population; and (II) the concentration of the delayed neutron precursors, which are radioactive isotopes formed in the fission process that decay through neutron emission. These equations are deterministic in nature, and therefore can provide only average values of the modeled populations. However, the actual dynamical process is stochastic: the neutron density and the delayed neutron precursor concentrations vary randomly with time. To address this stochastic behavior, Hayes and Allen have generalized the standard deterministic point kinetics equation. They derived a system of stochastic differential equations that can accurately model the random behavior of the neutron density and the precursor concentrations in a point reactor. Due to the stiffness of these equations, this system was numerically implemented using a stochastic piecewise constant approximation method (Stochastic PCA). Here, we present a study of the influence of stochastic fluctuations on the results of the neutron point kinetics equation. We reproduce the stochastic formulation introduced by Hayes and Allen and compute Monte Carlo numerical results for examples with constant and time-dependent reactivity, comparing these results with stochastic and deterministic methods found in the literature. Moreover, we introduce a modified version of the stochastic method to obtain a non-stiff solution, analogue to a previously derived deterministic approach. (author)

The solution of the neutron point kinetics equation with stochastic extension: an analysis of two moments

International Nuclear Information System (INIS)

Silva, Milena Wollmann da; Vilhena, Marco Tullio M.B.; Bodmann, Bardo Ernst J.; Vasques, Richard

2015-01-01

The neutron point kinetics equation, which models the time-dependent behavior of nuclear reactors, is often used to understand the dynamics of nuclear reactor operations. It consists of a system of coupled differential equations that models the interaction between (i) the neutron population; and (II) the concentration of the delayed neutron precursors, which are radioactive isotopes formed in the fission process that decay through neutron emission. These equations are deterministic in nature, and therefore can provide only average values of the modeled populations. However, the actual dynamical process is stochastic: the neutron density and the delayed neutron precursor concentrations vary randomly with time. To address this stochastic behavior, Hayes and Allen have generalized the standard deterministic point kinetics equation. They derived a system of stochastic differential equations that can accurately model the random behavior of the neutron density and the precursor concentrations in a point reactor. Due to the stiffness of these equations, this system was numerically implemented using a stochastic piecewise constant approximation method (Stochastic PCA). Here, we present a study of the influence of stochastic fluctuations on the results of the neutron point kinetics equation. We reproduce the stochastic formulation introduced by Hayes and Allen and compute Monte Carlo numerical results for examples with constant and time-dependent reactivity, comparing these results with stochastic and deterministic methods found in the literature. Moreover, we introduce a modified version of the stochastic method to obtain a non-stiff solution, analogue to a previously derived deterministic approach. (author)

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.

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.)

On a closed form approach to the fractional neutron point kinetics equation with temperature feedback

International Nuclear Information System (INIS)

Schramm, Marcelo; Bodmann, Bardo E.J.; Vilhena, Marco T.M.B.; Petersen, Claudio Z.; Alvim, Antonio C.M.

2013-01-01

Following the quest to find analytical solutions, we extend the methodology applied successfully to timely fractional neutron point kinetics (FNPK) equations by adding the effects of temperature. The FNPK equations with temperature feedback correspond to a nonlinear system and “stiff” type for the neutron density and the concentration of delayed neutron precursors. These variables determine the behavior of a nuclear reactor power with time and are influenced by the position of control rods, for example. The solutions of kinetics equations provide time information about the dynamics in a nuclear reactor in operation and are useful, for example, to understand the power fluctuations with time that occur during startup or shutdown of the reactor, due to adjustments of the control rods. The inclusion of temperature feedback in the model introduces an estimate of the transient behavior of the power and other variables, which are strongly coupled. Normally, a single value of reactivity is used across the energy spectrum. Especially in case of power change, the neutron energy spectrum changes as well as physical parameters such as the average cross sections. However, even knowing the importance of temperature effects on the control of the reactor power, the character of the set of nonlinear equations governing this system makes it difficult to obtain a purely analytical solution. Studies have been published in this sense, using numerical approaches. Here the idea is to consider temperature effects to make the model more realistic and thus solve it in a semi-analytical way. Therefore, the main objective of this paper is to obtain an analytical representation of fractional neutron point kinetics equations with temperature feedback, without having to resort to approximations inherent in numerical methods. To this end, we will use the decomposition method, which has been successfully used by the authors to solve neutron point kinetics problems. The results obtained will

On a closed form approach to the fractional neutron point kinetics equation with temperature feedback

Energy Technology Data Exchange (ETDEWEB)

Schramm, Marcelo; Bodmann, Bardo E.J.; Vilhena, Marco T.M.B., E-mail: marceloschramm@hotmail.com, E-mail: bardo.bodmann@ufrgs.br, E-mail: mtmbvilhena@gmail.com [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Departamento de Engenharia Mecanica; Petersen, Claudio Z., E-mail: claudiopetersen@yahoo.com.br [Universidade Federal de Pelotas (UFPel), RS (Brazil). Departamento de Matematica; Alvim, Antonio C.M., E-mail: alvim@nuclear.ufrj.br [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Instituto Alberto Luiz Coimbra de Pos-Graduacao e Pesquisa em Engenharia

2013-07-01

Following the quest to find analytical solutions, we extend the methodology applied successfully to timely fractional neutron point kinetics (FNPK) equations by adding the effects of temperature. The FNPK equations with temperature feedback correspond to a nonlinear system and “stiff” type for the neutron density and the concentration of delayed neutron precursors. These variables determine the behavior of a nuclear reactor power with time and are influenced by the position of control rods, for example. The solutions of kinetics equations provide time information about the dynamics in a nuclear reactor in operation and are useful, for example, to understand the power fluctuations with time that occur during startup or shutdown of the reactor, due to adjustments of the control rods. The inclusion of temperature feedback in the model introduces an estimate of the transient behavior of the power and other variables, which are strongly coupled. Normally, a single value of reactivity is used across the energy spectrum. Especially in case of power change, the neutron energy spectrum changes as well as physical parameters such as the average cross sections. However, even knowing the importance of temperature effects on the control of the reactor power, the character of the set of nonlinear equations governing this system makes it difficult to obtain a purely analytical solution. Studies have been published in this sense, using numerical approaches. Here the idea is to consider temperature effects to make the model more realistic and thus solve it in a semi-analytical way. Therefore, the main objective of this paper is to obtain an analytical representation of fractional neutron point kinetics equations with temperature feedback, without having to resort to approximations inherent in numerical methods. To this end, we will use the decomposition method, which has been successfully used by the authors to solve neutron point kinetics problems. The results obtained will

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)

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)

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 $

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

International Nuclear Information System (INIS)

Loureiro, Cesar Augusto Domingues; Santos, Adimir dos

2009-01-01

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

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.

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)

Power Trip Set-points of Reactor Protection System for New Research Reactor

International Nuclear Information System (INIS)

Lee, Byeonghee; Yang, Soohyung

2013-01-01

This paper deals with the trip set-point related to the reactor power considering the reactivity induced accident (RIA) of new research reactor. The possible scenarios of reactivity induced accidents were simulated and the effects of trip set-point on the critical heat flux ratio (CHFR) were calculated. The proper trip set-points which meet the acceptance criterion and guarantee sufficient margins from normal operation were then determined. The three different trip set-points related to the reactor power are determined based on the RIA of new research reactor during FP condition, over 0.1%FP and under 0.1%FP. Under various reactivity insertion rates, the CHFR are calculated and checked whether they meet the acceptance criterion. For RIA at FP condition, the acceptance criterion can be satisfied even if high power set-point is only used for reactor trip. Since the design of the reactor is still progressing and need a safety margin for possible design changes, 18 MW is recommended as a high power set-point. For RIA at 0.1%FP, high power setpoint of 18 MW and high log rate of 10%pp/s works well and acceptance criterion is satisfied. For under 0.1% FP operations, the application of high log rate is necessary for satisfying the acceptance criterion. Considering possible decrease of CHFR margin due to design changes, the high log rate is suggested to be 8%pp/s. Suggested trip set-points have been identified based on preliminary design data for new research reactor; therefore, these trip set-points will be re-established by considering design progress of the reactor. The reactor protection system (RPS) of new research reactor is designed for safe shutdown of the reactor and preventing the release of radioactive material to environment. The trip set point of RPS is essential for reactor safety, therefore should be determined to mitigate the consequences from accidents. At the same time, the trip set-point should secure margins from normal operational condition to avoid

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-05-15

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

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

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.)

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)

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....

Point kinetics modeling

International Nuclear Information System (INIS)

Kimpland, R.H.

1996-01-01

A normalized form of the point kinetics equations, a prompt jump approximation, and the Nordheim-Fuchs model are used to model nuclear systems. Reactivity feedback mechanisms considered include volumetric expansion, thermal neutron temperature effect, Doppler effect and void formation. A sample problem of an excursion occurring in a plutonium solution accidentally formed in a glovebox is 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)

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)

Reactor kinetics methods development. Final report

International Nuclear Information System (INIS)

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

1978-01-01

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

A 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)

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

International Nuclear Information System (INIS)

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

2014-01-01

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

Turning points in reactor design

International Nuclear Information System (INIS)

Beckjord, E.S.

1995-01-01

This article provides some historical aspects on nuclear reactor design, beginning with PWR development for Naval Propulsion and the first commercial application at Yankee Rowe. Five turning points in reactor design and some safety problems associated with them are reviewed: (1) stability of Dresden-1, (2) ECCS, (3) PRA, (4) TMI-2, and (5) advanced passive LWR designs. While the emphasis is on the thermal-hydraulic aspects, the discussion is also about reactor systems

Turning points in reactor design

Energy Technology Data Exchange (ETDEWEB)

Beckjord, E.S.

1995-09-01

This article provides some historical aspects on nuclear reactor design, beginning with PWR development for Naval Propulsion and the first commercial application at Yankee Rowe. Five turning points in reactor design and some safety problems associated with them are reviewed: (1) stability of Dresden-1, (2) ECCS, (3) PRA, (4) TMI-2, and (5) advanced passive LWR designs. While the emphasis is on the thermal-hydraulic aspects, the discussion is also about reactor systems.

Novel swirl-flow reactor for kinetic studies of semiconductor photocatalysis

NARCIS (Netherlands)

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

1997-01-01

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

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.

Point Genetics: A New Concept to Assess Neutron Kinetics

International Nuclear Information System (INIS)

Klein Meulekamp, R.; Kuijper, J.C.; Schikorr, M.

2005-01-01

Point genetic equations are introduced. These equations are similar to the well-known point kinetic equations but characterize and couple individual fission generations in subcritical systems. Point genetic equations are able to describe dynamic behavior of source-driven subcritical systems on shorter timescales than is possible with point kinetic equations. Point genetic parameters can be used as a first-order characterization of the system and can be calculated using standard Monte Carlo techniques; the implementation in other calculational schemes seems straightforward. A Godiva sphere is considered to show the applicability of the point genetic equations in describing a detector response on short timescales. For this system the point genetic parameters are calculated and compared with reference calculations. Typical dynamic source behavior is considered by studying a transient in which the neutron source energy decreases from 20 to 1 MeV. For all cases studied, the point genetic equations are compared to full space-time kinetic solutions, and it is shown that point genetics performs well

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)

Improved point-kinetics model for the BWR control rod drop accident

International Nuclear Information System (INIS)

Neogy, P.; Wakabayashi, T.; Carew, J.F.

1985-01-01

A simple prescription to account for spatial feedback weighting effects in RDA (rod drop accident) point-kinetics analyses has been derived and tested. The point-kinetics feedback model is linear in the core peaking factor, F/sub Q/, and in the core average void fraction and fuel temperature. Comparison with detailed spatial kinetics analyses indicates that the improved point-kinetics model provides an accurate description of the BWR RDA

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

International Nuclear Information System (INIS)

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

1992-01-01

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

Reaction kinetic analysis of reactor surveillance data

Energy Technology Data Exchange (ETDEWEB)

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

2017-02-15

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

Kinetic 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

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

Energy Technology Data Exchange (ETDEWEB)

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

1965-11-15

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

Modified mean generation time parameter in the neutron point kinetics equations

Energy Technology Data Exchange (ETDEWEB)

Diniz, Rodrigo C.; Gonçalves, Alessandro C.; Rosa, Felipe S.S., E-mail: alessandro@nuclear.ufrj.br, E-mail: frosa@if.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil)

2017-07-01

This paper proposes an approximation for the modified point kinetics equations proposed by NUNES et. al, 2015, through the adjustment of a kinetic parameter. This approximation consists of analyzing the terms of the modified point kinetics equations in order to identify the least important ones for the solution, resulting in a modification of the mean generation time parameter that incorporates all influences of the additional terms of the modified kinetics. This approximation is applied on the inverse kinetics, to compare the results with the inverse kinetics from the modified kinetics in order to validate the proposed model. (author)

Modified mean generation time parameter in the neutron point kinetics equations

International Nuclear Information System (INIS)

Diniz, Rodrigo C.; Gonçalves, Alessandro C.; Rosa, Felipe S.S.

2017-01-01

This paper proposes an approximation for the modified point kinetics equations proposed by NUNES et. al, 2015, through the adjustment of a kinetic parameter. This approximation consists of analyzing the terms of the modified point kinetics equations in order to identify the least important ones for the solution, resulting in a modification of the mean generation time parameter that incorporates all influences of the additional terms of the modified kinetics. This approximation is applied on the inverse kinetics, to compare the results with the inverse kinetics from the modified kinetics in order to validate the proposed model. (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

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)

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

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.

3D simulation of CANDU reactor regulating system

International Nuclear Information System (INIS)

Venescu, B.; Zevedei, D.; Jurian, M.

2013-01-01

Present paper shows the evaluation of the performance of the 3-D modal synthesis based reactor kinetic model in a closed-loop environment in a MATLAB/SIMULINK based Reactor Regulating System (RRS) simulation platform. A notable advantage of the 3-D model is the level of details that it can reveal as compared to the coupled point kinetic model. Using the developed RRS simulation platform, the reactor internal behaviours can be revealed during load-following tests. The test results are also benchmarked against measurements from an existing (CANDU) power plant. It can be concluded that the 3-D reactor model produces more realistic view of the core neutron flux distribution, which is closer to the real plant measurements than that from a coupled point kinetic model. It is also shown that, through a vectorization process, the computational load of the 3-D model is comparable with that of the 14-zone coupled point kinetic model. Furthermore, the developed Graphical User Interface (GUI) software package for RRS implementation represents a user friendly and independent application environment for education training and industrial utilizations. (authors)

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

International Nuclear Information System (INIS)

Lu Guiping; Peng Feng; Yi Jieyi

1988-01-01

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

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

Science.gov (United States)

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

2016-01-01

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

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)

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)

High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

Energy Technology Data Exchange (ETDEWEB)

Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-03-01

A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

Energy Technology Data Exchange (ETDEWEB)

Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-01-01

A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

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)

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

High-Temperature Gas-Cooled Test Reactor Point Design

Energy Technology Data Exchange (ETDEWEB)

Sterbentz, James William [Idaho National Laboratory; Bayless, Paul David [Idaho National Laboratory; Nelson, Lee Orville [Idaho National Laboratory; Gougar, Hans David [Idaho National Laboratory; Kinsey, James Carl [Idaho National Laboratory; Strydom, Gerhard [Idaho National Laboratory; Kumar, Akansha [Idaho National Laboratory

2016-04-01

A point design has been developed for a 200 MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technological readiness level, licensing approach and costs.

Reactor coolant flow measurements at Point Lepreau

International Nuclear Information System (INIS)

Brenciaglia, G.; Gurevich, Y.; Liu, G.

1996-01-01

The CROSSFLOW ultrasonic flow measurement system manufactured by AMAG is fully proven as reliable and accurate when applied to large piping in defined geometries for such applications as feedwater flows measurement. Its application to direct reactor coolant flow (RCF) measurements - both individual channel flows and bulk flows such as pump suction flow - has been well established through recent work by AMAG at Point Lepreau, with application to other reactor types (eg. PWR) imminent. At Point Lepreau, Measurements have been demonstrated at full power; improvements to consistently meet ±1% accuracy are in progress. The development and recent customization of CROSSFLOW to RCF measurement at Point Lepreau are described in this paper; typical measurement results are included. (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)

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

Science.gov (United States)

Kumar, B Shiva; Venkateswarlu, Ch

2014-08-01

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

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

NARCIS (Netherlands)

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

2005-01-01

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

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

NARCIS (Netherlands)

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

2005-01-01

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

Detection of kinetic change points in piece-wise linear single molecule motion

Science.gov (United States)

Hill, Flynn R.; van Oijen, Antoine M.; Duderstadt, Karl E.

2018-03-01

Single-molecule approaches present a powerful way to obtain detailed kinetic information at the molecular level. However, the identification of small rate changes is often hindered by the considerable noise present in such single-molecule kinetic data. We present a general method to detect such kinetic change points in trajectories of motion of processive single molecules having Gaussian noise, with a minimum number of parameters and without the need of an assumed kinetic model beyond piece-wise linearity of motion. Kinetic change points are detected using a likelihood ratio test in which the probability of no change is compared to the probability of a change occurring, given the experimental noise. A predetermined confidence interval minimizes the occurrence of false detections. Applying the method recursively to all sub-regions of a single molecule trajectory ensures that all kinetic change points are located. The algorithm presented allows rigorous and quantitative determination of kinetic change points in noisy single molecule observations without the need for filtering or binning, which reduce temporal resolution and obscure dynamics. The statistical framework for the approach and implementation details are discussed. The detection power of the algorithm is assessed using simulations with both single kinetic changes and multiple kinetic changes that typically arise in observations of single-molecule DNA-replication reactions. Implementations of the algorithm are provided in ImageJ plugin format written in Java and in the Julia language for numeric computing, with accompanying Jupyter Notebooks to allow reproduction of the analysis presented here.

The asymptotic behaviour of a critical point reactor in the absence of a controller

International Nuclear Information System (INIS)

Bansal, N.K.; Borgwaldt, H.

1976-11-01

A method is presented to calculate the first and second moments of neutron and precursor populations for a critical reactor system described by point kinetic equations and possessing inherent reactivity fluctuations. The equations have been linearised on the assumption that the system has a large average neutron population and that the amplitude of reactivity fluctuations is sufficiently small. The reactivity noise is assumed to be band limited white with a corner frequency higher than all the time constants of the system. Explicit expressions for the exact time development of the moments have been obtained for the case of a reactor without reactivity feedback and with one group of delayed neutrons. It is found that the expected values of the neutron and delayed neutron precursor numbers tend asymptotically to stationary values, whereas the mean square deviations increase linearly with time at an extremely low rate. (orig.) [de

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)

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

International Nuclear Information System (INIS)

Ravoire, Jean

1979-11-01

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

A coupled 3-D kinetics/system thermal-hydraulic analysis of main steam line break accident for Optimized Power Reactor 1000

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

A coupled 3-D kinetics/system thermal-hydraulic analysis of main steam line break accident for Optimized Power Reactor 1000

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

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

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

International Nuclear Information System (INIS)

Bae, Moo Hoon; Joo, Han Gyu

2009-01-01

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

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

International Nuclear Information System (INIS)

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

1992-01-01

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

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

Science.gov (United States)

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

2018-01-01

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

Optimal configuration of spatial points in the reactor cell

International Nuclear Information System (INIS)

Bosevski, T.

1968-01-01

Optimal configuration of spatial points was chosen in respect to the total number needed for integration of reactions in the reactor cell. Previously developed code VESTERN was used for numerical verification of the method on a standard reactor cell. The code applies the collision probability method for calculating the neutron flux distribution. It is shown that the total number of spatial points is twice smaller than the respective number of spatial zones needed for determination of number of reactions in the cell, with the preset precision. This result shows the direction for further condensing of the procedure for calculating the space-energy distribution of the neutron flux in a reactors cell [sr

Method of nuclear reactor control using a variable temperature load dependent set point

International Nuclear Information System (INIS)

Kelly, J.J.; Rambo, G.E.

1982-01-01

A method and apparatus for controlling a nuclear reactor in response to a variable average reactor coolant temperature set point is disclosed. The set point is dependent upon percent of full power load demand. A manually-actuated ''droop mode'' of control is provided whereby the reactor coolant temperature is allowed to drop below the set point temperature a predetermined amount wherein the control is switched from reactor control rods exclusively to feedwater flow

INDIAN POINT REACTOR STARTUP AND PERFORMANCE

Energy Technology Data Exchange (ETDEWEB)

Deddens, J. C.; Batch, M. L.

1963-09-15

The testing program for the Indian Point Reactor is discussed. The thermal and hydraulic evaluation of the primary coolant system is discussed. Analyses of fuel loading and initial criticality, measurement of operating coefficients of reactivity, control rod group reactivity worths, and xenon evaluation are presented. (R.E.U.)

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

Science.gov (United States)

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

2018-02-01

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

AIREKMOD-RR, Reactivity Transients in Nuclear Research Reactors

International Nuclear Information System (INIS)

Baggoura, B.; Mazrou, H.

2001-01-01

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

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.

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

Science.gov (United States)

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

1988-01-01

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

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.

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

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)

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

Science.gov (United States)

Deepanraj, B; Sivasubramanian, V; Jayaraj, S

2015-11-01

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

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

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

KAUST Repository

Wang, Jui-Yang

2017-06-01

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

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

International Nuclear Information System (INIS)

Barros, R.C. de.

1992-05-01

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

Initial value problem for the equations of reactor kinetics

International Nuclear Information System (INIS)

Kyncl, J.

1987-08-01

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

SPQR: a Monte Carlo reactor kinetics code

International Nuclear Information System (INIS)

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

1980-02-01

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

Development, validation and application of multi-point kinetics model in RELAP5 for analysis of asymmetric nuclear transients

Energy Technology Data Exchange (ETDEWEB)

Pradhan, Santosh K., E-mail: santosh@aerb.gov.in [Nuclear Safety Analysis Division, Atomic Energy Regulatory Board, Mumbai 400094 (India); Obaidurrahman, K. [Nuclear Safety Analysis Division, Atomic Energy Regulatory Board, Mumbai 400094 (India); Iyer, Kannan N. [Department of Mechanical Engineering, IIT Bombay, Mumbai 400076 (India); Gaikwad, Avinash J. [Nuclear Safety Analysis Division, Atomic Energy Regulatory Board, Mumbai 400094 (India)

2016-04-15

Highlights: • A multi-point kinetics model is developed for RELAP5 system thermal hydraulics code. • Model is validated against extensive 3D kinetics code. • RELAP5 multi-point kinetics formulation is used to investigate critical break for LOCA in PHWR. - Abstract: Point kinetics approach in system code RELAP5 limits its use for many of the reactivity induced transients, which involve asymmetric core behaviour. Development of fully coupled 3D core kinetics code with system thermal-hydraulics is the ultimate requirement in this regard; however coupling and validation of 3D kinetics module with system code is cumbersome and it also requires access to source code. An intermediate approach with multi-point kinetics is appropriate and relatively easy to implement for analysis of several asymmetric transients for large cores. Multi-point kinetics formulation is based on dividing the entire core into several regions and solving ODEs describing kinetics in each region. These regions are interconnected by spatial coupling coefficients which are estimated from diffusion theory approximation. This model offers an advantage that associated ordinary differential equations (ODEs) governing multi-point kinetics formulation can be solved using numerical methods to the desired level of accuracy and thus allows formulation based on user defined control variables, i.e., without disturbing the source code and hence also avoiding associated coupling issues. Euler's method has been used in the present formulation to solve several coupled ODEs internally at each time step. The results have been verified against inbuilt point-kinetics models of RELAP5 and validated against 3D kinetics code TRIKIN. The model was used to identify the critical break in RIH of a typical large PHWR core. The neutronic asymmetry produced in the core due to the system induced transient was effectively handled by the multi-point kinetics model overcoming the limitation of in-built point kinetics model

Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype

Science.gov (United States)

Bragg-Sitton, Shannon M.

2005-01-01

The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .tests, the point kinetics model was based on core thermal expansion via deflection measurements. It was found that core deflection was a strung function of how the SAFE-100 modules were fabricated and assembled (in terms of straightness, gaps, and other tolerances). To remove the added variable of how this particular core expands as compared to a different concept, it was decided to use a temperature based feedback model (based on several thermocouples placed throughout the core).

On a closed form solution of the point kinetics equations with reactivity feedback of temperature

International Nuclear Information System (INIS)

Silva, Jeronimo J.A.; Vilhena, Marco T.M.B.; Petersen, Claudio Z.; Bodmann, Bardo E.J.; Alvim, Antonio C.M.

2011-01-01

An analytical solution of the point kinetics equations to calculate reactivity as a function of time by the Decomposition method has recently appeared in the literature. In this paper, we go one step forward, by considering the neutron point kinetics equations together with temperature feedback effects. To accomplish that, we extended the point kinetics by a temperature perturbation, obtaining a second order nonlinear ordinary differential equation. This equation is then solved by the Decomposition Method, that is, by expanding the neutron density in a series and the nonlinear terms into Adomian Polynomials. Substituting these expansions into the nonlinear ordinary equation, we construct a recursive set of linear problems that can be solved by the methodology previously mentioned for the point kinetics equation. We also report on numerical simulations and comparisons against literature results. (author)

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)

Fast neutron reactors: the safety point of view

International Nuclear Information System (INIS)

Laverie, M.; Avenas, M.

1984-01-01

All versions of nuclear reactors present favourable and unfavourable characteristics from the point of view of safety. The safety of the installations is obtained by making efforts to utilize in the best possible way those which are favourable and by taking proper steps in the face of those which are unfavourable. The present article shows how this general principle has been applied as regards the fast neutron reactors of integrated design which have been developped in France, taking into account the specific features of this version. A qualitative method to compare the safety of this version with that of pressurized water reactors which has been widely put to the test commercially all over the world is presented. These analyses make, generally speaking, several positive characteristics stand out for these fast neutron reactors from the safety aspects [fr

Study on neutron diffusion and time dependence heat ina fluidized bed nuclear reactor

International Nuclear Information System (INIS)

Vilhena, M.T. de.

1988-01-01

The purpose of this work is to model the neutron diffusion and heat transfer for a Fluidized Bed Nuclear Reactor and its solution by Laplace Transform Technique with numerical inversion using Fourier Series. Also Gaussian quadrature and residues techniques were applied for numerical inversion. The neutron transport, diffusion, and point Kinetic equation for this nuclear reactor concept are developed. A matricial and Taylor Series methods are proposed for the solution of the point Kinetic equation which is a time scale problem of Stiff type

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)

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Kinetic study of treatment of wastewater contains food preservative agent by anaerobic baffled reactor : An overview

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.

Kinetic study of treatment of wastewater contains food preservative agent by anaerobic baffled reactor : An overview

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

Kinetic study of treatment of wastewater contains food preservative agent by anaerobic baffled reactor : An overview

Science.gov (United States)

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.

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

Application of Extreme Learning Machines to inverse neutron kinetics

International Nuclear Information System (INIS)

Picca, Paolo; Furfaro, Roberto

2017-01-01

Highlights: • The paper applies the Extreme Learning Machines (ELMs) to inverse reactor problems. • Multi-group transport model is used for the inversion as opposed to point kinetics. • ELMs are compared against Artificial Neural Networks (ANNs). • Various options are tested to improve the reliability of the estimation. • Results highlight the potential of the ELM approach. - Abstract: The paper presents the application of Extreme Leaning Machines (ELMs) for inverse reactor kinetic applications. ELMs were proposed by Huang and co-workers (2004, 2006a,b, 2015), which showed their enhances capabilities in terms of training speed and generalization with respect to classical Artificial Neural Networks (ANNs). ELMs are here implemented for reactivity determination as an alternative to ANNs (e.g. Picca et al. (2008)) and Gaussian Processes (Picca and Furfaro, 2012). After a review of the main features of ELMs, their application to inverse kinetic problems is proposed. The ELMs performance is tested on a typical accelerator drive system configuration (Yalina reactor) and the inversion is carried out on an accurate kinetic model (multi-group transport).

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

Directory of Open Access Journals (Sweden)

Mohammadreza Khani

2016-11-01

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

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

Science.gov (United States)

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

2015-01-01

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

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

KAUST Repository

Wang, Jui-Yang

2017-01-01

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

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)

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-03-15

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

Science.gov (United States)

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

2017-09-08

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

Reactivity changes in hybrid thermal-fast reactor systems during fast core flooding

International Nuclear Information System (INIS)

Pesic, M.

1994-09-01

A new space-dependent kinetic model in adiabatic approximation with local feedback reactivity parameters for reactivity determination in the coupled systems is proposed in this thesis. It is applied in the accident calculation of the 'HERBE' fast-thermal reactor system and compared to usual point kinetics model with core-averaged parameters. Advantages of the new model - more realistic picture of the reactor kinetics and dynamics during local large reactivity perturbation, under the same heat transfer conditions, are underlined. Calculated reactivity parameters of the new model are verified in the experiments performed at the 'HERBE' coupled core. The model has shown that the 'HERBE' safety system can shutdown reactor safely and fast even in the case of highly set power trip and even under conditions of big partial failure of the reactor safety system (author)

Generalized saddle point condition for ignition in a tokamak reactor with temperature and density profiles

International Nuclear Information System (INIS)

Mitari, O.; Hirose, A.; Skarsgard, H.M.

1989-01-01

In this paper, the concept of a generalized ignition contour map, is extended to the realistic case of a plasma with temperature and density profiles in order to study access to ignition in a tokamak reactor. The generalized saddle point is found to lie between the Lawson and ignition conditions. If the height of the operation path with Goldston L-mode scaling is higher than the generalized saddle point, a reactor can reach ignition with this scaling for the case with no confinement degradation effect due to alpha-particle heating. In this sense, the saddle point given in a general form is a new criterion for reaching ignition. Peaking the profiles for the plasma temperature and density can lower the height of the generalized saddle point and help a reactor to reach ignition. With this in mind, the authors can judge whether next-generation tokamaks, such as Compact Ignition Tokamak, Tokamak Ignition/Burn Experimental Reactor, Next European Torus, Fusion Experimental Reactor, International Tokamak Reactor, and AC Tokamak Reactor, can reach ignition with realistic profile parameters and an L-mode scaling law

Reaction kinetics in open reactors and serial transfers between closed reactors

Science.gov (United States)

Blokhuis, Alex; Lacoste, David; Gaspard, Pierre

2018-04-01

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

Effect of ultrasound on flotation kinetics in the reactor-separator

International Nuclear Information System (INIS)

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

2013-01-01

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

Effect of ultrasound on flotation kinetics in the reactor-separator

Science.gov (United States)

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

2013-03-01

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

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

Directory of Open Access Journals (Sweden)

V. K. Bityukov

2015-01-01

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

Operating point considerations for the Reference Theta-Pinch Reactor (RTPR)

International Nuclear Information System (INIS)

Krakowski, R.A.; Miller, R.L.; Hagenson, R.L.

1976-01-01

Aspects of the continuing engineering design-point reassessment and optimization of the Reference Theta-Pinch Reactor (RTPR) are discussed. An updated interim design point which achieves a favorable energy balance and involves relaxed technological requirements, which nonetheless satisfy more rigorous physics and engineering constraints, is presented

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

Subcriticality calculation in nuclear reactors with external neutron sources

Energy Technology Data Exchange (ETDEWEB)

Silva, Adilson Costa da; Martinez, Aquilino Senra; Silva, Fernando Carvalho da [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia (COPPE). Programa de Engenharia Nuclear]. E-mails: asilva@con.ufrj.br; aquilino@lmp.ufrj.br; fernando@con.ufrj.br

2007-07-01

The main objective of this paper consists on the development of a methodology to monitor subcriticality. We used the inverse point kinetic equation with 6 precursor groups and external neutron sources for the calculation of reactivity. The input data for the inverse point kinetic equation was adjusted, in order to use the neutron counting rates obtained from the subcritical multiplication (1/M) in a nuclear reactor. In this paper, we assumed that the external neutron sources strength is constant and we define it in terms of a known initial condition. The results obtained from inverse point kinetic equation with external neutron sources were compared with the results obtained with a benchmark calculation, and showed good accuracy (author)

Subcriticality calculation in nuclear reactors with external neutron sources

International Nuclear Information System (INIS)

Silva, Adilson Costa da; Martinez, Aquilino Senra; Silva, Fernando Carvalho da

2007-01-01

The main objective of this paper consists on the development of a methodology to monitor subcriticality. We used the inverse point kinetic equation with 6 precursor groups and external neutron sources for the calculation of reactivity. The input data for the inverse point kinetic equation was adjusted, in order to use the neutron counting rates obtained from the subcritical multiplication (1/M) in a nuclear reactor. In this paper, we assumed that the external neutron sources strength is constant and we define it in terms of a known initial condition. The results obtained from inverse point kinetic equation with external neutron sources were compared with the results obtained with a benchmark calculation, and showed good accuracy (author)

Different seeds to solve the equations of stochastic point kinetics using the Euler-Maruyama method

International Nuclear Information System (INIS)

Suescun D, D.; Oviedo T, M.

2017-09-01

In this paper, a numerical study of stochastic differential equations that describe the kinetics in a nuclear reactor is presented. These equations, known as the stochastic equations of punctual kinetics they model temporal variations in neutron population density and concentrations of deferred neutron precursors. Because these equations are probabilistic in nature (since random oscillations in the neutrons and population of precursors were considered to be approximately normally distributed, and these equations also possess strong coupling and stiffness properties) the proposed method for the numerical simulations is the Euler-Maruyama scheme that provides very good approximations for calculating the neutron population and concentrations of deferred neutron precursors. The method proposed for this work was computationally tested for different seeds, initial conditions, experimental data and forms of reactivity for a group of precursors and then for six groups of deferred neutron precursors at each time step with 5000 Brownian movements per seed. In a paper reported in the literature, the Euler-Maruyama method was proposed, but there are many doubts about the reported values, in addition to not reporting the seed used, so in this work is expected to rectify the reported values. After taking the average of the different seeds used to generate the pseudo-random numbers the results provided by the Euler-Maruyama scheme will be compared in mean and standard deviation with other methods reported in the literature and results of the deterministic model of the equations of the punctual kinetics. This comparison confirms in particular that the Euler-Maruyama scheme is an efficient method to solve the equations of stochastic point kinetics but different from the values found and reported by another author. The Euler-Maruyama method is simple and easy to implement, provides acceptable results for neutron population density and concentration of deferred neutron precursors and

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.

Photocatalytic mineralization of commercial herbicides in a pilot-scale solar CPC reactor: photoreactor modeling and reaction kinetics constants independent of radiation field.

Science.gov (United States)

Colina-Márquez, Jose; Machuca-Martínez, Fiderman; Li Puma, Gianluca

2009-12-01

The six-flux absorption-scattering model (SFM) of the radiation field in the photoreactor, combined with reaction kinetics and fluid-dynamic models, has proved to be suitable to describe the degradation of water pollutants in heterogeneous photocatalytic reactors, combining simplicity and accuracy. In this study, the above approach was extended to model the photocatalytic mineralization of a commercial herbicides mixture (2,4-D, diuron, and ametryne used in Colombian sugar cane crops) in a solar, pilot-scale, compound parabolic collector (CPC) photoreactor using a slurry suspension of TiO(2). The ray-tracing technique was used jointly with the SFM to determine the direction of both the direct and diffuse solar photon fluxes and the spatial profile of the local volumetric rate of photon absorption (LVRPA) in the CPC reactor. Herbicides mineralization kinetics with explicit photon absorption effects were utilized to remove the dependence of the observed rate constants from the reactor geometry and radiation field in the photoreactor. The results showed that the overall model fitted the experimental data of herbicides mineralization in the solar CPC reactor satisfactorily for both cloudy and sunny days. Using the above approach kinetic parameters independent of the radiation field in the reactor can be estimated directly from the results of experiments carried out in a solar CPC reactor. The SFM combined with reaction kinetics and fluid-dynamic models proved to be a simple, but reliable model, for solar photocatalytic applications.

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Antibiotic Fermentation Broth Treatment by a pilot upflow anaerobic sludge bed reactor and kinetic modeling.

Science.gov (United States)

Coskun, T; Kabuk, H A; Varinca, K B; Debik, E; Durak, I; Kavurt, C

2012-10-01

In this study, an upflow anaerobic sludge blanket (UASB) mesophilic reactor was used to remove antibiotic fermentation broth wastewater. The hydraulic retention time was held constant at 13.3 days. The volumetric organic loading value increased from 0.33 to 7.43 kg(COD)m(-3)d(-1) using antibiotic fermentation broth wastewater gradually diluted with various ratios of domestic wastewater. A COD removal efficiency of 95.7% was obtained with a maximum yield of 3,700 L d(-1) methane gas production. The results of the study were interpreted using the modified Stover-Kincannon, first-order, substrate mass balance and Van der Meer and Heertjes kinetic models. The obtained kinetic coefficients showed that antibiotic fermentation broth wastewater can be successfully treated using a UASB reactor while taking COD removal and methane production into account. Copyright © 2012 Elsevier Ltd. All rights reserved.

The research reactor as a tool in the master in nuclear reactors in Argentina

International Nuclear Information System (INIS)

Notari, Carla

2003-01-01

complete the Master with a seminar: Nuclear Power Plants, and a Thesis. In the frame of the academic plan, multiple activities are organized related to research reactors and also to nuclear power plants. Since the very beginning the performance of selected experiments in a nuclear reactor was recognized as an extraordinary tool to give the students an insight in the principal phenomena associated with the chain reaction and the related engineering problems. This experiments have an intrinsic elevated cost, associated with the relevance of the installation and with the specialized personnel involved. CNEA provides the career with this educational instrument through the Ra-1 and RA-3 reactors located at Constituyentes and Ezeiza Atomic Centers respectively. Various activities are under way but the most established, in the Reactor Physics Course, is the estimation of kinetic parameters in RA-1 reactor. The practice includes three different experiments: Approach to critical and calibration of control rods by the compensation method: Starting in a subcritical state with source the calibration of control rod B1 vs B2 is done by introduction of the first and withdrawal of the second. The methods used are based on the Point Kinetic Model; Measurement of control rods effectivity by the rod-drop method: Separate Rod Drop of rods B1 B2 B3 of the overall ensemble B1 B2 B3 B4 and total scram starting with three withdrawn and one partially inserted, is the procedure followed to estimate the reactivity worth of B1 B2 B3 and scram. The Point Kinetic Model and the Modal Kinetic Model are used; Reactor noise technique for the estimation of reactor parameters: α and Λ. The kinetic parameters are estimated assuring that the Point Kinetic Model is valid (detection chambers near to the core), that the fluctuation of the fission density is the dominant source of the correlated part of neutron noise (measurement at low power, <10kw), the dominance of the fundamental armonic (simultaneous use of

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.

Sensitivity analysis of power excursion in RSG-GAS reactor due to reactivity insertion

International Nuclear Information System (INIS)

Pinem, Surian; Sembiring, Tagor Malem

2002-01-01

Reactor kinetics has a very important role in reactor operation safety and nuclear reactor control. One of the important aspects in reactor kinetics is power behavior as function of time due to chain reaction in the core. The calculation was performed using kinetic equation and feedback reactivity and evaluated using static power coefficient. Analysis was performed for oxide 250 g, silicide 250 g and silicide 300 g fuel elements with insertion of positive reactivity, negative reactivity and reactivity close to delay neutron fraction. The calculation of power excursion sensitivity showed that the insertion of 0,5 % Δk/k, in the fuel element of silicide 300 g is bigger 5 % than the one of oxide 250 g or silicide 250 g. If inserted by - 1,2 % Δk/k, there is no change among three fuel elements. Therefore, in kinetic point of view, it is showed there is no significant influence in the RSG-GAS reactor operation safety is the current core of oxide 250 g is converted to silicide 250 g or to silicide 300 g

Compact reversed-field pinch reactors (CRFPR): sensitivity study and design-point determination

International Nuclear Information System (INIS)

Hagenson, R.L.; Krakowski, R.A.

1982-07-01

If the costing assumptions upon which the positive assessment of conventional large superconducting fusion reactors are based proves overly optimistic, approaches that promise considerably increased system power density and reduced mass utilization will be required. These more compact reactor embodiments generally must operate with reduced shield thickness and resistive magnets. Because of the unique, magnetic topology associated with the Reversed-Field Pinch (RFP), the compact reactor embodiment for this approach is particularly attractive from the viewpoint of low-field resistive coils operating with Ohmic losses that can be made small relative to the fusion power. A comprehensive system model is developed and described for a steady-state, compact RFP reactor (CRFPR). This model is used to select a unique cost-optimized design point that will be used for a conceptual engineering design. The cost-optimized CRFPR design presented herein would operate with system power densities and mass utilizations that are comparable to fission power plants and are an order of magnitude more favorable than the conventional approaches to magnetic fusion power. The sensitivity of the base-case design point to changes in plasma transport, profiles, beta, blanket thickness, normal vs superconducting coils, and fuel cycle (DT vs DD) is examined. The RFP approach is found to yield a point design for a high-power-density reactor that is surprisingly resilient to changes in key, but relatively unknown, physics and systems parameters

Comparative analysis among several methods used to solve the point kinetic equations

International Nuclear Information System (INIS)

Nunes, Anderson L.; Goncalves, Alessandro da C.; Martinez, Aquilino S.; Silva, Fernando Carvalho da

2007-01-01

The main objective of this work consists on the methodology development for comparison of several methods for the kinetics equations points solution. The evaluated methods are: the finite differences method, the stiffness confinement method, improved stiffness confinement method and the piecewise constant approximations method. These methods were implemented and compared through a systematic analysis that consists basically of confronting which one of the methods consume smaller computational time with higher precision. It was calculated the relative which function is to combine both criteria in order to reach the goal. Through the analyses of the performance factor it is possible to choose the best method for the solution of point kinetics equations. (author)

Comparative analysis among several methods used to solve the point kinetic equations

Energy Technology Data Exchange (ETDEWEB)

Nunes, Anderson L.; Goncalves, Alessandro da C.; Martinez, Aquilino S.; Silva, Fernando Carvalho da [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear; E-mails: alupo@if.ufrj.br; agoncalves@con.ufrj.br; aquilino@lmp.ufrj.br; fernando@con.ufrj.br

2007-07-01

The main objective of this work consists on the methodology development for comparison of several methods for the kinetics equations points solution. The evaluated methods are: the finite differences method, the stiffness confinement method, improved stiffness confinement method and the piecewise constant approximations method. These methods were implemented and compared through a systematic analysis that consists basically of confronting which one of the methods consume smaller computational time with higher precision. It was calculated the relative which function is to combine both criteria in order to reach the goal. Through the analyses of the performance factor it is possible to choose the best method for the solution of point kinetics equations. (author)

An accurate technique for the solution of the nonlinear point kinetics equations

International Nuclear Information System (INIS)

Picca, Paolo; Ganapol, Barry D.; Furfaro, Roberto

2011-01-01

A novel methodology for the solution of non-linear point kinetic (PK) equations is proposed. The technique is based on a piecewise constant approximation of PK system of ODEs and explicitly accounts for reactivity feedback effects, through an iterative cycle. High accuracy is reached by introducing a sub-mesh for the numerical evaluation of integrals involved and by correcting the source term to include the non-linear effect on a finer time scale. The use of extrapolation techniques for convergence acceleration is also explored. Results for adiabatic feedback model are reported and compared with other benchmarks in literature. The convergence trend makes the algorithm particularly attractive for applications, including in multi-point kinetics and quasi-static frameworks. (author)

Optimal power and distribution control for weakly-coupled-core reactor

International Nuclear Information System (INIS)

Oohori, Takahumi; Kaji, Ikuo

1977-01-01

A numerical procedure has been devised for obtaining the optimal power and distribution control for a weakly-coupled-core reactor. Several difficulties were encountered in solving this optimization problem: (1) nonlinearity of the reactor kinetics equations; (2) neutron-leakage interaction between the cores; (3) localized power changes occurring in addition to the total power changes; (4) constraints imposed on the states - e.g. reactivity, reactor period. To obviate these difficulties, use is made of the generalized Newton method to convert the problem into an iterative sequence of linear programming problems, after approximating the differential equations and the integral performance criterion by a set of discrete algebraic equations. In this procedure, a heuristic but effective method is used for deriving an initial approximation, which is then made to converge toward the optimal solution. Delayed-neutron one-group point reactor models embodying transient temperature feed-back to the reactivity are used in obtaining the kinetics equations for the weakly-coupled-core reactor. The criterion adopted for determining the optimality is a norm relevant to the deviations of neutron density from the desired trajectories or else to the time derivatives of the neutron density; uniform control intervals are prescribed. Examples are given of two coupled-core reactors with typical parameters to illustrate the results obtained with this procedure. A comparison is also made between the coupled-core reactor and the one-point reactor. (auth.)

End point control of an actinide precipitation reactor

International Nuclear Information System (INIS)

Muske, K.R.

1997-01-01

The actinide precipitation reactors in the nuclear materials processing facility at Los Alamos National Laboratory are used to remove actinides and other heavy metals from the effluent streams generated during the purification of plutonium. These effluent streams consist of hydrochloric acid solutions, ranging from one to five molar in concentration, in which actinides and other metals are dissolved. The actinides present are plutonium and americium. Typical actinide loadings range from one to five grams per liter. The most prevalent heavy metals are iron, chromium, and nickel that are due to stainless steel. Removal of these metals from solution is accomplished by hydroxide precipitation during the neutralization of the effluent. An end point control algorithm for the semi-batch actinide precipitation reactors at Los Alamos National Laboratory is described. The algorithm is based on an equilibrium solubility model of the chemical species in solution. This model is used to predict the amount of base hydroxide necessary to reach the end point of the actinide precipitation reaction. The model parameters are updated by on-line pH measurements

A benchmark on the calculation of kinetic parameters based on reactivity effect experiments in the CROCUS reactor

International Nuclear Information System (INIS)

Paratte, J.M.; Frueh, R.; Kasemeyer, U.; Kalugin, M.A.; Timm, W.; Chawla, R.

2006-01-01

Measurements in the CROCUS reactor at EPFL, Lausanne, are reported for the critical water level and the inverse reactor period for several different sets of delayed supercritical conditions. The experimental configurations were also calculated by four different calculation methods. For each of the supercritical configurations, the absolute reactivity value has been determined in two different ways, viz.: (i) through direct comparison of the multiplication factor obtained employing a given calculation method with the corresponding value for the critical case (calculated reactivity: ρ calc ); (ii) by application of the inhour equation using the kinetic parameters obtained for the critical configuration and the measured inverse reactor period (measured reactivity: ρ meas ). The calculated multiplication factors for the reference critical configuration, as well as ρ calc for the supercritical cases, are found to be in good agreement. However, the values of ρ meas produced by two of the applied calculation methods differ appreciably from the corresponding ρ calc values, clearly indicating deficiencies in the kinetic parameters obtained from these methods

Coalescence kinetics of dispersed crude oil in a laboratory reactor

International Nuclear Information System (INIS)

Sterling, M.C. Jr.; Ojo, T.; Autenrieth, R.L.; Bonner, J.S.; Page, C.A.; Ernst, A.N.S.

2002-01-01

A study was conducted to examine the effects of salinity and mixing energy on the resurfacing and coalescence rates of chemically dispersed crude oil droplets. This kinetic study involved the use of mean shear rates to characterize the mixing energy in a laboratory reactor. Coagulation kinetics of dispersed crude oil were determined within a range of mean shear rates of 5, 10, 15, and 20 per second, and with salinity values of 10 and 30 per cent. Observed droplet distributions were fit to a transport-reaction model to estimate collision efficiency values and their dependence on salinity and mixing energy. Dispersant efficiencies were compared with those derived from other laboratory testing methods. Experimentally determined dispersant efficiencies were found to be 10 to 50 per cent lower than predicted using a non-interacting droplet model, but dispersant efficiencies were higher than those predicted using other testing methods. 24 refs., 1 tab., 3 figs

CFD analysis of the dynamic behaviour of a fuel rod subchannel in a supercritical water reactor with point kinetics

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

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.)

Decontamination of the Douglas Point reactor, May 1983

International Nuclear Information System (INIS)

Lesurf, J.E.; Stepaniak, R.; Broad, L.G.; Barber, W.G.

1983-01-01

The Douglas Point reactor primary heat transport system including the fuel, was successfully decontaminated by the CAN-DECON process in 1975. A second decontamination, also using the CAN-DECON process, was successfully performed in May 1983. This paper outlines the need for the decontamination, the process used, the results obtained, and the benefits to the station maintenance and operation

Stability analysis of the Ghana Research Reactor-1 (GHARR-1)

International Nuclear Information System (INIS)

Della, R.; Alhassan, E.; Adoo, N.A.; Bansah, C.Y.; Nyarko, B.J.B.; Akaho, E.H.K.

2013-01-01

Highlights: • We developed a theoretical model to study the stability of the Ghana Research Reactor-1. • The neutronics transfer function was described by the point kinetics model for a single group of delayed neutrons. • The thermal hydraulics transfer function was based on the modified lumped parameter concept. • A computer code, RESA (REactor Stability Analysis) was developed. • Results show that the closed-loop transfer function was stable and well damped for variable operating power levels. - Abstract: A theoretical model has been developed to study the stability of the Ghana Research Reactor one (GHARR-1). The closed-loop transfer function of GHARR-1 was established based on the model, which involved the neutronics and the thermal hydraulics transfer functions. The reactor kinetics was described by the point kinetics model for a single group of delayed neutrons, whilst the thermal hydraulics transfer function was based on the modified lumped parameter concept. The inherent internal feedback effect due to the fuel and the coolant was represented by the fuel temperature coefficient and the moderator temperature coefficient respectively. A computer code, RESA (REactor Stability Analysis), entirely in Java was developed based on the model for systems analysis. Stability analysis of the open-loop transfer function of GHARR-1 based on the Nyquist criterion and Bode diagrams using RESA, has shown that the closed-loop transfer function was marginally stable for variable operating power levels. The relative stability margins of GHARR-1 were also identified

Biodegradation of phenol with chromium(VI) reduction in an anaerobic fixed-biofilm process-Kinetic model and reactor performance

International Nuclear Information System (INIS)

Lin, Yen-Hui; Wu, Chih-Lung; Hsu, Chih-Hao; Li, Hsin-Lung

2009-01-01

A mathematical model system was derived to describe the simultaneous removal of phenol biodegradation with chromium(VI) reduction in an anaerobic fixed-biofilm reactor. The model system incorporates diffusive mass transport and double Monod kinetics. The model was solved using a combination of the orthogonal collocation method and Gear's method. A laboratory-scale column reactor was employed to validate the kinetic model system. Batch kinetic tests were conducted independently to evaluate the biokinetic parameters used in the model simulation. The removal efficiencies of phenol and chromium(VI) in an anaerobic fixed-biofilm process were approximately 980 mg/g and 910 mg/g, respectively, under a steady-state condition. In the steady state, model-predicted biofilm thickness reached up to 350 μm and suspended cells in the effluent were 85 mg cell/l. The experimental results agree closely with the results of the model simulations.

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

Coupling of kinetic Monte Carlo simulations of surface reactions to transport in a fluid for heterogeneous catalytic reactor modeling

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.

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.

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

International Nuclear Information System (INIS)

Castellotti, U.

1987-01-01

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

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

International Nuclear Information System (INIS)

Santos, R.S. dos

1993-01-01

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

Kinetics of concentration decay of specific organic matter in UASB reactors operating with and without return of aerobic sludge.

Science.gov (United States)

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.

The influence of TiO2 and aeration on the kinetics of electrochemical oxidation of phenol in packed bed reactor

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

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

Comparison of Wims-Aecl / Dragon / RFSP and MCNP results with Zed-2 measurements for control device worth and reactor kinetics - 037

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)

Point design for deuterium-deuterium compact reversed-field pinch reactors

International Nuclear Information System (INIS)

Dabiri, A.E.; Dobrott, D.R.; Gurol, H.; Schnack, D.D.

1984-01-01

A deuterium-deuterium (D-D) reversed-field pinch (RFP) reactor may be made comparable in size and cost to a deuterium-tritium (D-T) reactor at the expense of high-thermal heat load to the first wall. This heat load is the result of the larger percentage of fusion power in charged particles in the D-D reaction as compared to the D-T reaction. The heat load may be reduced by increasing the reactor size and hence the cost. In addition to this ''degraded'' design, the size may be kept small by means of a higher heat load wall, or by means of a toroidal divertor, in which case most of the heat load seen by the wall is in the form of radiation. Point designs are developed for these approaches and cost studies are performed and compared with a D-T reactor. The results indicate that the cost of electricity of a D-D RFP reactor is about20% higher than a D-T RFP reactor. This increased cost could be offset by the inherent safety features of the D-D fuel cycle

Modeling Taylor series approximations for prompt neutron kinetics with lab view simulations

International Nuclear Information System (INIS)

Adzri, E. P.

2012-09-01

The reactor point kinetics equations have been subjected to intense research in an effort to find simple yet accurate numerical solutions methods. The equations are very stiff numerically, meaning that there is a wide variation in the decay constants, so that using a particular time step in the numerical solution may provide sufficient accuracy for the group, but not for another. Several solutions techniques have been presented on the point kinetics equations with varying degrees of complexity. These include Power Series Solutions, CORE, PCA, Genapol and Taylor series methods. In this research, algorithms were developed based on the first and second order Taylor series expansion and simulated in LabVIEW to solve the Reactor Point Kinetics equations using block diagram nodes implemented within stacked sequences. The algorithms developed were fast,accurate and simple to code. Several reactivity insertions were used to simulate the change in neutron population with time. The LabVIEW- Taylor series solutions were compared with other solution techniques such as Power Series Solutions, CORE, PCA, Genapol and McMahon and Pierson's Taylor series approximation. The results of LabVIEW-Taylor series technique used by McMahon and Pearson The LabVIEW-implemented techniques were found to agree very well with these other methods. At 1x10 -8 s the neutron population was 1.000220 neutrons / cm 3 , at 1 x 10 -2 s it was 2.007681 neutrons / cm 3 and at 1x10 -1 s it was 2.075317 neutrons / cm 3 ; same results reported by Genapol for a fast reactor, it produced good and accurate results and compared very favorably with other methods found in the literature. Using much smaller time steps to the order or 10 -8 s commensurate with fast reactor parameters also produced very satisfactory results, indicating that the LabVIEW-based Taylor series technique is suitable for simulating the kinetics of fast reactors as well as thermal reactors. Algorithms developed that included second order terms

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

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

International Nuclear Information System (INIS)

Vargas, L.

1988-01-01

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

[Kinetics of catalytic wet air oxidation of phenol in trickle bed reactor].

Science.gov (United States)

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.

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)

Kinetic modelling and characterization of microbial community present in a full-scale UASB reactor treating brewery effluent.

Science.gov (United States)

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.

Three-dimensional coupled kinetics/thermal- hydraulic benchmark TRIGA experiments

International Nuclear Information System (INIS)

Feltus, Madeline Anne; Miller, William Scott

2000-01-01

This research project provides separate effects tests in order to benchmark neutron kinetics models coupled with thermal-hydraulic (T/H) models used in best-estimate codes such as the Nuclear Regulatory Commission's (NRC) RELAP and TRAC code series and industrial codes such as RETRAN. Before this research project was initiated, no adequate experimental data existed for reactivity initiated transients that could be used to assess coupled three-dimensional (3D) kinetics and 3D T/H codes which have been, or are being developed around the world. Using various Test Reactor Isotope General Atomic (TRIGA) reactor core configurations at the Penn State Breazeale Reactor (PSBR), it is possible to determine the level of neutronics modeling required to describe kinetics and T/H feedback interactions. This research demonstrates that the small compact PSBR TRIGA core does not necessarily behave as a point kinetics reactor, but that this TRIGA can provide actual test results for 3D kinetics code benchmark efforts. This research focused on developing in-reactor tests that exhibited 3D neutronics effects coupled with 3D T/H feedback. A variety of pulses were used to evaluate the level of kinetics modeling needed for prompt temperature feedback in the fuel. Ramps and square waves were used to evaluate the detail of modeling needed for the delayed T/H feedback of the coolant. A stepped ramp was performed to evaluate and verify the derived thermal constants for the specific PSBR TRIGA core loading pattern. As part of the analytical benchmark research, the STAR 3D kinetics code (, STAR: Space and time analysis of reactors, Version 5, Level 3, Users Guide, Yankee Atomic Electric Company, YEAC 1758, Bolton, MA) was used to model the transient experiments. The STAR models were coupled with the one-dimensional (1D) WIGL and LRA and 3D COBRA (, COBRA IIIC: A digital computer program for steady-state and transient thermal-hydraulic analysis of rod bundle nuclear fuel elements, Battelle

Kinetics of vinyl acetate emulsion polymerization in a pulsed tubular reactor: comparison between experimental and simulation results

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.

Wastewater treatment using photo-impinging streams cyclone reactor: Computational fluid dynamics and kinetics modeling

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.

Measurement and analysis of pressure tube elongation in the Douglas Point reactor

International Nuclear Information System (INIS)

Causey, A.R.; MacEwan, S.R.; Jamieson, H.C.; Mitchell, A.B.

1980-02-01

Elongations of zirconium alloy pressure tubes in CANDU reactors, which occur as a result of neutron-irradiation-induced creep and growth, have been measured over the past 6 years, and the consequences of thses elongations have recently been analysed. Elongation rates, previously deduced from extensive measurements of elongations of cold-worked Zircaloy-2 pressure tubes in the Pickering reactors, have been modified to apply to the pressure tubes in the Douglas Point (DP) reactor by taking into account measured diffences in texture and dislocation density. Using these elongation rates, and structural data unique to the DP reactor, the analysis predicts elongation behaviour which is in good agreement with pressure tube elongations measured during the ten years of reactor operation. (Auth)

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.)

The Simulator Development for RDE Reactor

Science.gov (United States)

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.

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

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

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

International Nuclear Information System (INIS)

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

1978-01-01

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

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

Science.gov (United States)

Abedi-Varaki, Mehdi

2017-08-01

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

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.

The reflector effect on the neutron lifetimes in the IPEN/MB-01 reactor

International Nuclear Information System (INIS)

Gonnelli, Eduardo

2013-01-01

The aim of this study is to present the reflector effect on the neutron lifetimes in the IPEN/MB-01 reactor. The proposed method requires an approach which takes into account both the reflector and the core, so that the point kinetics equations, which constitute the theoretical basis of all mathematical development, contemplate both regions of the reactor. From these equations, as known as two regions kinetics point equations, theoretical expressions are obtained for the Auto Power Spectral Densities (APSD), which are used for least squares fit of the experimental data of APSD obtained in several subcritical states. The prompt neutron generation time, the neutron lifetimes in the reflector and the neutron return fraction from the reflector to the core are derived from the fitting. (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

Programme of hot points eradication (Co-60) led on French PWR type reactors

International Nuclear Information System (INIS)

Rocher, A.; Ridoux, P.; Anthoni, S.; Brun, C.

1998-01-01

The question of hot points (pellets rich in cobalt 59 or in cobalt 60 in a PWR type reactor), is studied from the radiation protection point of view. The purpose is to see how to optimize the radiation protection, the elimination of these hot points can bring an improvement. (N.C.)

Simulation and calculation of three-reactor system of catalytic reforming

International Nuclear Information System (INIS)

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

1999-01-01

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

Thermo-kinetic instabilities in model reactors. Examples in experimental tests

Science.gov (United States)

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

Lipase kinetics: hydrolysis of triacetin by lipase from Candida cylindracea in a hollow-fiber membrane reactor

NARCIS (Netherlands)

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

Determination of melting point of mixed-oxide fuel irradiated in a fast breeder reactor

International Nuclear Information System (INIS)

Tachibana, Toshimichi

1985-01-01

The melting point of fuel is important to set its in-reactor maximum temperature in fuel design. The fuel melting point measuring methods are broadly the filament method and the capsule sealing method. The only instance of measuring the melting point of irradiated mixed oxide (U, Pu)O 2 fuel by the filament method is by GE in the United States. The capsule sealing method, while the excellent means, is difficult in weld sealing the irradiated fuel in a capsule within the cell. In the fast reactor development program, the remotely operated melting point measuring apparatus in capsule sealing the mixed (U, Pu)O 2 fuel irradiated in the experimental FBR Joyo was set in the cell and the melting point was measured, for the first time in the world. (Mori, K.)

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.

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

Point kinetics equations for subcritical systems based on the importance function associated to an external neutron source

International Nuclear Information System (INIS)

Carvalho Gonçalves, Wemerson de; Martinez, Aquilino Senra; Carvalho da Silva, Fernando

2015-01-01

Highlights: • We define the new function importance. • We calculate the kinetic parameters Λ, β, Γ and Q to: 0.95, 0.96, 0.97, 0.98 and 0.99. • We compared the results with those obtained by the main important functions. • We found that the calculated kinetic parameters are physically consistent. - Abstract: This paper aims to determine the parameters for a new set of equations of point kinetic subcritical systems, based on the concept of importance of Heuristic Generalized Perturbation Theory (HGPT). The importance function defined here is related to both the subcriticality and the external neutron source worth (which keeps the system at steady state). The kinetic parameters defined in this work are compared with the corresponding parameters when adopting the importance functions proposed by Gandini and Salvatores (2002), Dulla et al. (2006) and Nishihara et al. (2003). Furthermore, the point kinetics equations developed here are solved for two different transients, considering the parameters obtained with different importance functions. The results collected show that there is a similar behavior of the solution of the point kinetics equations, when used with the parameters obtained by the importance functions proposed by Gandini and Salvatores (2002) and Dulla et al. (2006), specially near the criticality. However, this is not verified as the system gets farther from criticality

Nonlinear dynamics of ITU TRIGA reactor

International Nuclear Information System (INIS)

Hizal, N.A.; Gencay, S.; Gungordu, E.; Geckinli, M.; Ciftcioglu, O.; Can, B.

1988-01-01

Complete dynamics of a reactor could be developed starting from the very basic principles. However such a detailed approach is often not worth the effort for a rather simple pool type reactor which may be subjected to various power excursion maneuvers without challenging its safety system. Therefore a coupled point kinetics-lumped thermal hydraulics model is taken up as the basis of the system model. Response of the reactor to ramp insertion of reactivity is observed by sampling the power channel, water, and fuel temperatures with the help of a PC. One of the important model parameters, fuel temperature feedback effect is studied during power excursions and the results are compared with those of static tests. (author)

New applications of neutron noise theory in power reactor physics

Energy Technology Data Exchange (ETDEWEB)

Arzhanov, Vasiliy

2000-04-01

The present thesis deals with neutron noise theory as applied to three comparatively different topics (or problems) in power reactor physics. Namely they are: theoretical investigation of the possibility to use a newly proposed current-flux (C/F) detector in Pressurized Water Reactors (PWRs) for the localisation of anomalies; both definition and studies on the point kinetic and adiabatic approximations for the relatively recently proposed Accelerator Driven Systems (ADS); development of the general theory of linear reactor kinetics and neutron noise in systems with varying size. One important practical problem is to detect and localise a vibrating control rod pin. The significance comes from the operational experience which indicates that individual pins can execute excessive mechanical vibrations that may lead to damage. Such mechanical vibrations induce neutron noise that can be detected. While the detection is relatively easy, the localisation of a vibrating control rod is much more complicated because only one measuring position is available and one needs to have at least three measured quantities. Therefore it has currently been proposed that the fluctuations of the neutron current vector, called the current noise, can be used in addition to the scalar noise in reactor diagnostic problems. The thesis investigates the possibility of the localization of a vibrating control rod pin in a PWR control assembly by using the scalar neutron noise and the 2-D radial current noise as measured at one central point in the control assembly. An explicit localisation technique is elaborated in which the searched position is determined as the absolute minimum of a minimisation function. The technique is investigated in numerical simulations. The results of the simulation tests show the potential applicability of the method. By design accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and

New applications of neutron noise theory in power reactor physics

International Nuclear Information System (INIS)

Arzhanov, Vasiliy

2000-04-01

The present thesis deals with neutron noise theory as applied to three comparatively different topics (or problems) in power reactor physics. Namely they are: theoretical investigation of the possibility to use a newly proposed current-flux (C/F) detector in Pressurized Water Reactors (PWRs) for the localisation of anomalies; both definition and studies on the point kinetic and adiabatic approximations for the relatively recently proposed Accelerator Driven Systems (ADS); development of the general theory of linear reactor kinetics and neutron noise in systems with varying size. One important practical problem is to detect and localise a vibrating control rod pin. The significance comes from the operational experience which indicates that individual pins can execute excessive mechanical vibrations that may lead to damage. Such mechanical vibrations induce neutron noise that can be detected. While the detection is relatively easy, the localisation of a vibrating control rod is much more complicated because only one measuring position is available and one needs to have at least three measured quantities. Therefore it has currently been proposed that the fluctuations of the neutron current vector, called the current noise, can be used in addition to the scalar noise in reactor diagnostic problems. The thesis investigates the possibility of the localization of a vibrating control rod pin in a PWR control assembly by using the scalar neutron noise and the 2-D radial current noise as measured at one central point in the control assembly. An explicit localisation technique is elaborated in which the searched position is determined as the absolute minimum of a minimisation function. The technique is investigated in numerical simulations. The results of the simulation tests show the potential applicability of the method. By design accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and

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

International Nuclear Information System (INIS)

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

2012-01-01

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

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

Power reactor noise studies and applications

Energy Technology Data Exchange (ETDEWEB)

Arzhanov, V

2002-03-01

The present thesis deals with the neutron noise arising in power reactor systems. Generally, it can be divided into two major parts: first, neutron noise diagnostics, or more specifically, novel methods and algorithms to monitor nuclear industrial reactors; and second, contributions to neutron noise theory as applied to power reactor systems. Neutron noise diagnostics is presented by two topics. The first one is a theoretical study on the possibility to use a newly proposed current-flux (C/F) detector in Pressurised Water Reactors (PWR) for the localisation of anomalies. The second topic concerns various methods to detect guide tube impacting in Boiling Water Reactors (BWR). The significance of these problems comes from the operational experience. The thesis describes a novel method to localise vibrating control rods in a PWR by using only one C/F detector. Another novel method, based on wavelet analysis, is put forward to detect impacting guide tubes in a BWR. Neutron noise theory is developed for both Accelerator Driven Systems (ADS) and traditional reactors. By design the accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and methods that have been developed for traditional reactors and also it poses a number of new problems. As for the latter, the thesis investigates the space-dependent neutron noise caused by a fluctuating source. It is shown that the frequency-dependent spatial behaviour exhibits some new properties that are different from those known in traditional critical systems. On the other hand, various reactor physics approximations (point kinetic, adiabatic etc.) have not been defined yet for the subcritical systems. In this respect the thesis presents a systematic formulation of the above mentioned approximations as well as investigations of their properties. Another important problem in neutron noise theory is the treatment of moving boundaries. In this case one

Power reactor noise studies and applications

International Nuclear Information System (INIS)

Arzhanov, V.

2002-03-01

The present thesis deals with the neutron noise arising in power reactor systems. Generally, it can be divided into two major parts: first, neutron noise diagnostics, or more specifically, novel methods and algorithms to monitor nuclear industrial reactors; and second, contributions to neutron noise theory as applied to power reactor systems. Neutron noise diagnostics is presented by two topics. The first one is a theoretical study on the possibility to use a newly proposed current-flux (C/F) detector in Pressurised Water Reactors (PWR) for the localisation of anomalies. The second topic concerns various methods to detect guide tube impacting in Boiling Water Reactors (BWR). The significance of these problems comes from the operational experience. The thesis describes a novel method to localise vibrating control rods in a PWR by using only one C/F detector. Another novel method, based on wavelet analysis, is put forward to detect impacting guide tubes in a BWR. Neutron noise theory is developed for both Accelerator Driven Systems (ADS) and traditional reactors. By design the accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and methods that have been developed for traditional reactors and also it poses a number of new problems. As for the latter, the thesis investigates the space-dependent neutron noise caused by a fluctuating source. It is shown that the frequency-dependent spatial behaviour exhibits some new properties that are different from those known in traditional critical systems. On the other hand, various reactor physics approximations (point kinetic, adiabatic etc.) have not been defined yet for the subcritical systems. In this respect the thesis presents a systematic formulation of the above mentioned approximations as well as investigations of their properties. Another important problem in neutron noise theory is the treatment of moving boundaries. In this case one

MAGNETIC NULL POINTS IN KINETIC SIMULATIONS OF SPACE PLASMAS

International Nuclear Information System (INIS)

Olshevsky, Vyacheslav; Innocenti, Maria Elena; Cazzola, Emanuele; Lapenta, Giovanni; Deca, Jan; Divin, Andrey; Peng, Ivy Bo; Markidis, Stefano

2016-01-01

We present a systematic attempt to study magnetic null points and the associated magnetic energy conversion in kinetic particle-in-cell simulations of various plasma configurations. We address three-dimensional simulations performed with the semi-implicit kinetic electromagnetic code iPic3D in different setups: variations of a Harris current sheet, dipolar and quadrupolar magnetospheres interacting with the solar wind, and a relaxing turbulent configuration with multiple null points. Spiral nulls are more likely created in space plasmas: in all our simulations except lunar magnetic anomaly (LMA) and quadrupolar mini-magnetosphere the number of spiral nulls prevails over the number of radial nulls by a factor of 3–9. We show that often magnetic nulls do not indicate the regions of intensive energy dissipation. Energy dissipation events caused by topological bifurcations at radial nulls are rather rare and short-lived. The so-called X-lines formed by the radial nulls in the Harris current sheet and LMA simulations are rather stable and do not exhibit any energy dissipation. Energy dissipation is more powerful in the vicinity of spiral nulls enclosed by magnetic flux ropes with strong currents at their axes (their cross sections resemble 2D magnetic islands). These null lines reminiscent of Z-pinches efficiently dissipate magnetic energy due to secondary instabilities such as the two-stream or kinking instability, accompanied by changes in magnetic topology. Current enhancements accompanied by spiral nulls may signal magnetic energy conversion sites in the observational data

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

International Nuclear Information System (INIS)

Setiyanto; Sembiring, Tagor M.; Pinem, Surian

2007-01-01

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

The influence of TiO{sub 2} and aeration on the kinetics of electrochemical oxidation of phenol in 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.

Calculation of kinetic parameters of Caliban metallic core experimental reactor from stochastic neutron measurements

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)

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)

Numerical instability of time-discretized one-point kinetic equations

International Nuclear Information System (INIS)

Hashimoto, Kengo; Ikeda, Hideaki; Takeda, Toshikazu

2000-01-01

The one-point kinetic equations with numerical errors induced by the explicit, implicit and Crank-Nicolson integration methods are derived. The zero-power transfer functions based on the present equations are demonstrated to investigate the numerical stability of the discretized systems. These demonstrations indicate unconditional stability for the implicit and Crank-Nicolson methods but present the possibility of numerical instability for the explicit method. An upper limit of time mesh spacing for the stability is formulated and several numerical calculations are made to confirm the validity of this formula

Space nuclear reactor concepts for avoidance of a single point failure

International Nuclear Information System (INIS)

El-Genk, M. S.

2007-01-01

This paper presents three space nuclear reactor concepts for future exploration missions requiring electrical power of 10's to 100's kW, for 7-10 years. These concepts avoid a single point failure in reactor cooling; and they could be used with a host of energy conversion technologies. The first is lithium or sodium heat pipes cooled reactor. The heat pipes operate at a fraction of their prevailing capillary or sonic limit. Thus, when a number of heat pipes fail, those in the adjacent modules remove their heat load, maintaining reactor core adequately cooled. The second is a reactor with a circulating liquid metal coolant. The reactor core is divided into six identical sectors, each with a separate energy conversion loop. The sectors in the reactor core are neurotically coupled, but hydraulically decoupled. Thus, when a sector experiences a loss of coolant, the fission power generated in it will be removed by the circulating coolant in the adjacent sectors. In this case, however, the reactor fission power would have to decrease to avoid exceeding the design temperature limits in the sector with a failed loop. These two reactor concepts are used with energy conversion technologies, such as advanced Thermoelectric (TE), Free Piston Stirling Engines (FPSE), and Alkali Metal Thermal-to- Electric Conversion (AMTEC). Gas cooled reactors are a better choice to use with Closed Brayton Cycle engines, such as the third reactor concept to be presented in the paper. It has a sectored core that is cooled with a binary mixture of He-Xe (40 gm/mole). Each of the three sectors in the reactor has its own CBC and neutronically, but not hydraulically, coupled to the other sectors

A direct method for numerical solution of a class of nonlinear Volterra integro-differential equations and its application to the nonlinear fission and fusion reactor kinetics

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.)

The kinetics of the partial dehydration of gibbsite to activated alumina in a reactor for pneumatic transport

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.

Analytical modelling and study of the stability characteristics of the Advanced Heavy Water Reactor

International Nuclear Information System (INIS)

Nayak, A.K.; Vijayan, P.K.; Saha, D.

2000-04-01

An analytical model has been developed to study the thermohydraulic and neutronic-coupled density-wave instability in the Indian Advanced Heavy Water Reactor (AHWR) which is a natural circulation pressure tube type boiling water reactor. The model considers a point kinetics model for the neutron dynamics and a lumped parameter model for the fuel thermal dynamics along with the conservation equations of mass, momentum and energy and equation of state for the coolant. In addition, to study the effect of neutron interactions between different parts of the core, the model considers a coupled multipoint kinetics equation in place of simple point kinetics equation. Linear stability theory was applied to reveal the instability of in-phase and out-of-phase modes in the boiling channels of the AHWR. The results indicate that the design configuration considered may experience both Ledinegg and Type I and Type II density-wave instabilities depending on the operating condition. Some methods of suppressing these instabilities were found out. In addition, it was found that the stability behavior of the reactor is greatly influenced by the void reactivity coefficient, fuel time constant, radial power distribution and channel inlet orificing. The delayed neutrons were found to have strong influence on the Type I and Type II instabilities. Decay ratio maps were predicted considering various operating parameters of the reactor, which are useful for its design. (author)

Disclosure of the oscillations in kinetics of the reactor pressure vessel steel damage at fast neutron intensity decreasing

Science.gov (United States)

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.

PC-Reactor-core transient simulation code

International Nuclear Information System (INIS)

Nakata, H.

1989-10-01

PC-REATOR, a reactor core transient simulation code has been developed for the real-time operator training on a IBM-PC microcomputer. The program presents capabilities for on-line exchange of the operating parameters during the transient simulation, by friendly keyboard instructions. The model is based on the point-kinetics approximation, with 2 delayed neutron percursors and up to 11 decay power generating groups. (author) [pt

Safety reviews of the Brazilian multipurpose reactor

International Nuclear Information System (INIS)

Soares, Humberto Vitor

2014-01-01

This work presents a model developed for thermal hydraulic (TH) simulation of the Multipurpose Brazilian Reactor (RMB), whose Brazilian proposal for design, construction and operation was established in 2007. This reactor has as main proposed the production of radioisotopes for use in exams of nuclear medicine, material tests and utilization of neutrons beams. Besides of the TH modeling and safety analysis of the reactor, the application of a methodology to perform coupled calculation thermal-hydraulic/neutron kinetic (TH/NK) is also presented. Initially, the RMB was modeled in the safety analysis RELAP5 code. This code performs the thermal hydraulic calculation using point kinetics. Subsequently, the model was adapted and verified to the RELAP5-3D© code. This code performs the process of internal coupling through the option of nodal neutron kinetics calculation using the NESTLE code which solves the neutron diffusion equation. To generate the neutronic group constants, which are macroscopic cross sections that serve as input data for the neutronic codes, it was used the WIMSD-5B cell calculation code. The neutron analysis code PARCS was also used to model the 3D RMB core in order to compare the results of radial and axial average power distribution with the results generated by RELAP5-3D© code and with the available results of the CITATION neutron kinetic code. The safety analyses demonstrated safe behavior of the reactor through situations of possible transients. The 3D coupled calculations to the steady state operation also showed expected behavior, as well as the RMB neutronic analyzes performed with the codes NESTLE and PARCS.(author)

Kinetic Study of COS with Tertiary Alkanolamine Solutions. 2. Modeling and Experiments in a Stirred Cell Reactor

NARCIS (Netherlands)

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

Omega-mode perturbation theory and reactor kinetics for analyzing accelerator-driven subcritical systems

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)

High temperature mechanisms and kinetics of SiC oxidation under low partial pressures of oxygen: application to the fuel cladding of gas fast reactors

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

Beam transient analyses of Accelerator Driven Subcritical Reactors based on neutron transport method

Energy Technology Data Exchange (ETDEWEB)

He, Mingtao; Wu, Hongchun [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, Shaanxi (China); Zheng, Youqi, E-mail: yqzheng@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, Shaanxi (China); Wang, Kunpeng [Nuclear and Radiation Safety Center, PO Box 8088, Beijing 100082 (China); Li, Xunzhao; Zhou, Shengcheng [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, Shaanxi (China)

2015-12-15

Highlights: • A transport-based kinetics code for Accelerator Driven Subcritical Reactors is developed. • The performance of different kinetics methods adapted to the ADSR is investigated. • The impacts of neutronic parameters deteriorating with fuel depletion are investigated. - Abstract: The Accelerator Driven Subcritical Reactor (ADSR) is almost external source dominated since there is no additional reactivity control mechanism in most designs. This paper focuses on beam-induced transients with an in-house developed dynamic analysis code. The performance of different kinetics methods adapted to the ADSR is investigated, including the point kinetics approximation and space–time kinetics methods. Then, the transient responds of beam trip and beam overpower are calculated and analyzed for an ADSR design dedicated for minor actinides transmutation. The impacts of some safety-related neutronics parameters deteriorating with fuel depletion are also investigated. The results show that the power distribution varying with burnup leads to large differences in temperature responds during transients, while the impacts of kinetic parameters and feedback coefficients are not very obvious. Classification: Core physic.

The ARIES-I high-field-tokamak reactor: Design-point determination and parametric studies

International Nuclear Information System (INIS)

Miller, R.L.

1989-01-01

The multi-institutional ARIES study has examined the physics, technology, safety, and economic issues associated with the conceptual design of a tokamak magnetic-fusion reactor. The ARIES-I variant envisions a DT-fueled device based on advanced superconducting coil, blanket, and power-conversion technologies and a modest extrapolation of existing tokamak physics. A comprehensive systems and trade study has been conducted as an integral and ongoing part of the reactor assessment in order to identify an acceptable design point to be subjected to detailed analysis and integration as well as to characterize the ARIES-I operating space. Results of parametric studies leading to the identification of such a design point are presented. 15 refs., 6 figs., 2 tabs

Qualitative performance comparison of reactivity estimation between the extended Kalman filter technique and the inverse point kinetic method

International Nuclear Information System (INIS)

Shimazu, Y.; Rooijen, W.F.G. van

2014-01-01

Highlights: • Estimation of the reactivity of nuclear reactor based on neutron flux measurements. • Comparison of the traditional method, and the new approach based on Extended Kalman Filtering (EKF). • Estimation accuracy depends on filter parameters, the selection of which is described in this paper. • The EKF algorithm is preferred if the signal to noise ratio is low (low flux situation). • The accuracy of the EKF depends on the ratio of the filter coefficients. - Abstract: The Extended Kalman Filtering (EKF) technique has been applied for estimation of subcriticality with a good noise filtering and accuracy. The Inverse Point Kinetic (IPK) method has also been widely used for reactivity estimation. The important parameters for the EKF estimation are the process noise covariance, and the measurement noise covariance. However the optimal selection is quite difficult. On the other hand, there is only one parameter in the IPK method, namely the time constant for the first order delay filter. Thus, the selection of this parameter is quite easy. Thus, it is required to give certain idea for the selection of which method should be selected and how to select the required parameters. From this point of view, a qualitative performance comparison is carried out

Is automated kinetic measurement superior to end-point for advanced oxidation protein product?

Science.gov (United States)

Oguz, Osman; Inal, Berrin Bercik; Emre, Turker; Ozcan, Oguzhan; Altunoglu, Esma; Oguz, Gokce; Topkaya, Cigdem; Guvenen, Guvenc

2014-01-01

Advanced oxidation protein product (AOPP) was first described as an oxidative protein marker in chronic uremic patients and measured with a semi-automatic end-point method. Subsequently, the kinetic method was introduced for AOPP assay. We aimed to compare these two methods by adapting them to a chemistry analyzer and to investigate the correlation between AOPP and fibrinogen, the key molecule responsible for human plasma AOPP reactivity, microalbumin, and HbA1c in patients with type II diabetes mellitus (DM II). The effects of EDTA and citrate-anticogulated tubes on these two methods were incorporated into the study. This study included 93 DM II patients (36 women, 57 men) with HbA1c levels > or = 7%, who were admitted to the diabetes and nephrology clinics. The samples were collected in EDTA and in citrate-anticoagulated tubes. Both methods were adapted to a chemistry analyzer and the samples were studied in parallel. In both types of samples, we found a moderate correlation between the kinetic and the endpoint methods (r = 0.611 for citrate-anticoagulated, r = 0.636 for EDTA-anticoagulated, p = 0.0001 for both). We found a moderate correlation between fibrinogen-AOPP and microalbumin-AOPP levels only in the kinetic method (r = 0.644 and 0.520 for citrate-anticoagulated; r = 0.581 and 0.490 for EDTA-anticoagulated, p = 0.0001). We conclude that adaptation of the end-point method to automation is more difficult and it has higher between-run CV% while application of the kinetic method is easier and it may be used in oxidative stress studies.

Homotopy analysis solutions of point kinetics equations with one delayed precursor group

International Nuclear Information System (INIS)

Zhu Qian; Luo Lei; Chen Zhiyun; Li Haofeng

2010-01-01

Homotopy analysis method is proposed to obtain series solutions of nonlinear differential equations. Homotopy analysis method was applied for the point kinetics equations with one delayed precursor group. Analytic solutions were obtained using homotopy analysis method, and the algorithm was analysed. The results show that the algorithm computation time and precision agree with the engineering requirements. (authors)

PKI, Gamma Radiation Reactor Shielding Calculation by Point-Kernel Method

International Nuclear Information System (INIS)

Li Chunhuai; Zhang Liwu; Zhang Yuqin; Zhang Chuanxu; Niu Xihua

1990-01-01

1 - Description of program or function: This code calculates radiation shielding problem of gamma-ray in geometric space. 2 - Method of solution: PKI uses a point kernel integration technique, describes radiation shielding geometric space by using geometric space configuration method and coordinate conversion, and makes use of calculation result of reactor primary shielding and flow regularity in loop system for coolant

Dynamic Response Testing in an Electrically Heated Reactor Test Facility

Science.gov (United States)

Bragg-Sitton, Shannon M.; Morton, T. J.

2006-01-01

Non-nuclear testing can be a valuable tool in development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system (Bragg-Sitton, 2005). The current paper applies the same testing methodology to a direct drive gas cooled reactor system, demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. In each testing application, core power transients were controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. Although both system designs utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility.

Growth kinetics of dislocation loops in irradiated ceramic materials

International Nuclear Information System (INIS)

Ryazanov, A.I.; Kinoshita, C.

2002-01-01

Ceramic materials are expected to be applied in the future fusion reactor as radio frequency (RF) windows, toroidal insulating breaks and diagnostic probes. The radiation resistance of ceramic materials, degradation of the electrical properties and radiation induced conductivity of these materials under neutron irradiation are determined by the kinetics of the accumulation of point defects in the matrix and point defect cluster formation (dislocation loops, voids, etc.). Under irradiation, due to the ionization process, excitation of electronic subsystem and covalent type of interaction between atoms the point defects in ceramic materials are characterized by the charge state (e.g. an F + center, an oxygen vacancy with a single trapped electron) and the effective charge. For the investigation of radiation resistance of ceramic materials for future fusion applications it is very important to understand the physical mechanisms of formation and growth of dislocation loops and voids under irradiation taking into account in this system the effective charge of point defects. In the present paper the physical mechanisms of dislocation loop growth in ceramic material are investigated. For this aim a theoretical model is suggested for the description of the kinetics of point defect accumulation in the matrix taking into account the charge state of the point defects and the effect of an electric field on diffusion migration process of charged point defects. A self-consistent system of kinetic equations describing the generation of electrical fields near dislocation loops and diffusion migration of charged point defects in elastic and electrical fields is formulated. The solution of the kinetic equations allows to find the growth rate of dislocation loops in ceramic materials under irradiation taking into account the charge state of the point defects and the effect of electric and elastic stress fields near dislocation loop on the diffusion processes

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

Energy Technology Data Exchange (ETDEWEB)

Alca Ruiz, F

1982-07-01

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

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

International Nuclear Information System (INIS)

Risher, D.H. Jr.

1975-01-01

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

ENERGY DISSIPATION IN MAGNETIC NULL POINTS AT KINETIC SCALES

International Nuclear Information System (INIS)

Olshevsky, Vyacheslav; Lapenta, Giovanni; Divin, Andrey; Eriksson, Elin; Markidis, Stefano

2015-01-01

We use kinetic particle-in-cell and MHD simulations supported by an observational data set to investigate magnetic reconnection in clusters of null points in space plasma. The magnetic configuration under investigation is driven by fast adiabatic flux rope compression that dissipates almost half of the initial magnetic field energy. In this phase powerful currents are excited producing secondary instabilities, and the system is brought into a state of “intermittent turbulence” within a few ion gyro-periods. Reconnection events are distributed all over the simulation domain and energy dissipation is rather volume-filling. Numerous spiral null points interconnected via their spines form null lines embedded into magnetic flux ropes; null point pairs demonstrate the signatures of torsional spine reconnection. However, energy dissipation mainly happens in the shear layers formed by adjacent flux ropes with oppositely directed currents. In these regions radial null pairs are spontaneously emerging and vanishing, associated with electron streams and small-scale current sheets. The number of spiral nulls in the simulation outweighs the number of radial nulls by a factor of 5–10, in accordance with Cluster observations in the Earth's magnetosheath. Twisted magnetic fields with embedded spiral null points might indicate the regions of major energy dissipation for future space missions such as the Magnetospheric Multiscale Mission

Neutron transport. Physics and calculation of nuclear reactors with applications to pressurized water reactors and fast neutron reactors. 2 ed.

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

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.

Solution of the two-dimensional space-time reactor kinetics equation by a locally one-dimensional method

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 and simulation of CANDU reactor and its regulating system

Science.gov (United States)

Javidnia, Hooman

Analytical computer codes are indispensable tools in design, optimization, and control of nuclear power plants. Numerous codes have been developed to perform different types of analyses related to the nuclear power plants. A large number of these codes are designed to perform safety analyses. In the context of safety analyses, the control system is often neglected. Although there are good reasons for such a decision, that does not mean that the study of control systems in the nuclear power plants should be neglected altogether. In this thesis, a proof of concept code is developed as a tool that can be used in the design. optimization. and operation stages of the control system. The main objective in the design of this computer code is providing a tool that is easy to use by its target audience and is capable of producing high fidelity results that can be trusted to design the control system and optimize its performance. Since the overall plant control system covers a very wide range of processes, in this thesis the focus has been on one particular module of the the overall plant control system, namely, the reactor regulating system. The center of the reactor regulating system is the CANDU reactor. A nodal model for the reactor is used to represent the spatial neutronic kinetics of the core. The nodal model produces better results compared to the point kinetics model which is often used in the design and analysis of control system for nuclear reactors. The model can capture the spatial effects to some extent. although it is not as detailed as the finite difference methods. The criteria for choosing a nodal model of the core are: (1) the model should provide more detail than point kinetics and capture spatial effects, (2) it should not be too complex or overly detailed to slow down the simulation and provide details that are extraneous or unnecessary for a control engineer. Other than the reactor itself, there are auxiliary models that describe dynamics of different

Blankets for fusion reactors : materials and neutronics

International Nuclear Information System (INIS)

Carvalho, S.H. de.

1980-03-01

The studies about Fusion Reactors have lead to several problems for which there is no general agreement about the best solution. Nevertheless, several points seem to be well defined, at least for the first generation of reactors. The fuel, for example, should be a mixture of deuterium and tritium. Therefore, the reactor should be able to generate the tritium to be burned and also to transform kinetic energy of the fusion neutrons into heat in a process similar to the fission reactors. The best materials for the composition of the blanket were first selected and then the neutronics for the proposed system was developed. The neutron flux in the blanket was calculated using the discrete ordinates transport code, ANISN. All the nuclides cross sections came from the DLC-28/CTR library, that processed the ENDF/B data, using the SUPERTOG Program. (Author) [pt

Aging management program of the reactor building concrete at Point Lepreau Generating Station

Science.gov (United States)

Aldea, C.-M.; Shenton, B.; Demerchant, M. M.; Gendron, T.

2011-04-01

In order for New Brunswick Power Nuclear (NBPN) to control the risks of degradation of the concrete reactor building at the Point Lepreau Generating Station (PLGS) the development of an aging management plan (AMP) was initiated. The intention of this plan was to determine the requirements for specific structural components of concrete of the reactor building that require regular inspection and maintenance to ensure the safe and reliable operation of the plant. The document is currently in draft form and presents an integrated methodology for the application of an AMP for the concrete of the reactor building. The current AMP addresses the reactor building structure and various components, such as joint sealant and liners that are integral to the structure. It does not include internal components housed within the structure. This paper provides background information regarding the document developed and the strategy developed to manage potential degradation of the concrete of the reactor building, as well as specific programs and preventive and corrective maintenance activities initiated.

Models and Stability Analysis of Boiling Water Reactors

Energy Technology Data Exchange (ETDEWEB)

John Dorning

2002-04-15

We have studied the nuclear-coupled thermal-hydraulic stability of boiling water reactors (BWRs) using a model that includes: space-time modal neutron kinetics based on spatial w-modes; single- and two-phase flow in parallel boiling channels; fuel rod heat conduction dynamics; and a simple model of the recirculation loop. The BR model is represented by a set of time-dependent nonlinear ordinary differential equations, and is studied as a dynamical system using the modern bifurcation theory and nonlinear dynamical systems analysis. We first determine the stability boundary (SB) - or Hopf bifurcation set- in the most relevant parameter plane, the inlet-subcooling-number/external-pressure-drop plane, for a fixed control rod induced external reactivity equal to the 100% rod line value; then we transform the SB to the practical power-flow map used by BWR operating engineers and regulatory agencies. Using this SB, we show that the normal operating point at 100% power is very stable, that stability of points on the 100% rod line decreases as the flow rate is reduced, and that operating points in the low-flow/high-power region are least stable. We also determine the SB that results when the modal kinetics is replaced by simple point reactor kinetics, and we thereby show that the first harmonic mode does not have a significant effect on the SB. However, we later show that it nevertheless has a significant effect on stability because it affects the basin of attraction of stable operating points. Using numerical simulations we show that, in the important low-flow/high-power region, the Hopf bifurcation that occurs as the SB is crossed is subcritical; hence, growing oscillations can result following small finite perturbations of stable steady-states on the 100% rod line at points in the low-flow/high-power region. Numerical simulations are also performed to calculate the decay ratios (DRs) and frequencies of oscillations for various points on the 100% rod line. It is

Models and Stability Analysis of Boiling Water Reactors

International Nuclear Information System (INIS)

Dorning, John

2002-01-01

We have studied the nuclear-coupled thermal-hydraulic stability of boiling water reactors (BWRs) using a model that includes: space-time modal neutron kinetics based on spatial w-modes; single- and two-phase flow in parallel boiling channels; fuel rod heat conduction dynamics; and a simple model of the recirculation loop. The BR model is represented by a set of time-dependent nonlinear ordinary differential equations, and is studied as a dynamical system using the modern bifurcation theory and nonlinear dynamical systems analysis. We first determine the stability boundary (SB) - or Hopf bifurcation set- in the most relevant parameter plane, the inlet-subcooling-number/external-pressure-drop plane, for a fixed control rod induced external reactivity equal to the 100% rod line value; then we transform the SB to the practical power-flow map used by BWR operating engineers and regulatory agencies. Using this SB, we show that the normal operating point at 100% power is very stable, that stability of points on the 100% rod line decreases as the flow rate is reduced, and that operating points in the low-flow/high-power region are least stable. We also determine the SB that results when the modal kinetics is replaced by simple point reactor kinetics, and we thereby show that the first harmonic mode does not have a significant effect on the SB. However, we later show that it nevertheless has a significant effect on stability because it affects the basin of attraction of stable operating points. Using numerical simulations we show that, in the important low-flow/high-power region, the Hopf bifurcation that occurs as the SB is crossed is subcritical; hence, growing oscillations can result following small finite perturbations of stable steady-states on the 100% rod line at points in the low-flow/high-power region. Numerical simulations are also performed to calculate the decay ratios (DRs) and frequencies of oscillations for various points on the 100% rod line. It is

SALLY, Dynamic Behaviour of Reactor Cooling Channel by Point Model

International Nuclear Information System (INIS)

Reiche, Chr.; Ziegenbein, D.

1981-01-01

1 - Nature of the physical problem solved: The dynamical behaviour of a cooling channel is calculated. Starting from an equilibrium state a perturbation is introduced into the system. That may be an outer reactivity perturbation or a change in the coolant velocity or in the coolant temperature. The neutron kinetics is treated in the framework of the one-point model. The cooling channel consists of a cladded and cooled fuel rod. The temperature distribution is taken into account as an array above a mesh of radial zones and axial layers. Heat transfer is considered in radial direction only, the thermodynamical coupling of the different layers is obtained by the coolant flow. The thermal material parameters are considered to be temperature independent. Reactivity feedback is introduced by means of reactivity coefficients for fuel, canning, and coolant. Doppler broadening is included. The first cooling cycle can be taken into account by a simple model. 2 - Method of solution: The integration of the point kinetics equations is done numerically by the P11 scheme. The system of temperature equations with constant heat resistance coefficients is solved by the method of factorization. 3 - Restrictions on the complexity of the problem: Given limits are: 10 radial fuel zones, 25 axial layers, 6 groups of delayed neutrons

Experimental and kinetic modeling study of 3-methylheptane in a jet-stirred reactor

KAUST Repository

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.

Development of a computer code for Dalat research reactor transient analysis

International Nuclear Information System (INIS)

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

2003-01-01

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

A KINETIC MODEL FOR H2O2/UV PROCESS IN A COMPLETELY MIXED BATCH REACTOR. (R825370C076)

Science.gov (United States)

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...

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-03-15

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

An internally illuminated monolith reactor: Pros and cons relative to a slurry reactor

NARCIS (Netherlands)

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

Robust observer based control for axial offset in pressurized-water nuclear reactors based on the multipoint reactor model using Lyapunov approach

Energy Technology Data Exchange (ETDEWEB)

Zaidabadinejad, Majid; Ansarifar, Gholam Reza [Isfahan Univ. (Iran, Islamic Republic of). Dept. of Nuclear Engineering

2017-11-15

In nuclear reactor imbalance of axial power distribution induces xenon oscillations. These fluctuations must be maintained bounded within allowable limits. Otherwise, the nuclear power plant could become unstable. Therefore, bounded these oscillations is considered to be a restriction for the load following operation. Also, in order to design the nuclear reactor control systems, poisons concentrations, especially xenon must be accessible. But, physical measurement of these parameters is impossible. In this paper, for the first time, in order to estimate the axial xenon oscillations and ensures these oscillations are kept bounded within allowable limits during load-following operation, a robust observer based nonlinear control based on multipoint kinetics reactor model for pressurized-water nuclear reactors is presented. The reactor core is simulated based on the multi-point nuclear reactor model (neutronic and thermal-hydraulic). Simulation results are presented to demonstrate the effectiveness of the proposed observer based controller for the load-following operation.

Robust observer based control for axial offset in pressurized-water nuclear reactors based on the multipoint reactor model using Lyapunov approach

International Nuclear Information System (INIS)

Zaidabadinejad, Majid; Ansarifar, Gholam Reza

2017-01-01

In nuclear reactor imbalance of axial power distribution induces xenon oscillations. These fluctuations must be maintained bounded within allowable limits. Otherwise, the nuclear power plant could become unstable. Therefore, bounded these oscillations is considered to be a restriction for the load following operation. Also, in order to design the nuclear reactor control systems, poisons concentrations, especially xenon must be accessible. But, physical measurement of these parameters is impossible. In this paper, for the first time, in order to estimate the axial xenon oscillations and ensures these oscillations are kept bounded within allowable limits during load-following operation, a robust observer based nonlinear control based on multipoint kinetics reactor model for pressurized-water nuclear reactors is presented. The reactor core is simulated based on the multi-point nuclear reactor model (neutronic and thermal-hydraulic). Simulation results are presented to demonstrate the effectiveness of the proposed observer based controller for the load-following operation.

Analysis of neutronics and dynamic characteristics with reactivity injection in LBE cooled sub-critical reactor

International Nuclear Information System (INIS)

Chen Sen; Wu Yican; Jin Ming; Chen Zhibin; Bai Yunqing; Zhao Zhumin

2014-01-01

Accelerator Driven Sub-critical System (ADS) has particular neutronics behaviors compared with the critical system. Prompt jump approximation point reactor kinetics equations taken external source into account have been deduced using an approach of prompt jump approximation. And the relationship between injection reactivity and power ampliation has been achieved. In addition, based on the RELAP5 code the prolong development of point reactor kinetics code used into assessing sub-critical system have been promoted. Different sub-criticality (k eff = 0.90, 0.95, 0.97, 0.98 and 0.99) have been assessed in preliminary design of a type of natural circulation cooling sub-critical reactor under conditions of reactivity injection +1 β in one second. It shows that the external source prompt transient approximation method has an accurate solution after injecting reactivity around short time and has a capacity to solve the dynamic equation, and the sub-critical system has an inner stability while the deeper sub-criticality the less impact on the sub-critical system. (authors)

Reactor safeguards

CERN Document Server

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

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Investigation of boiling water reactor stability and limit-cycle amplitude

International Nuclear Information System (INIS)

Damiano, B.; March-Leuba, J.A.; Euler, J.A.

1991-01-01

Galerkin's method has been applied to a boiling water reactor (BWR) dynamics model consisting of the point kinetics equations, which describe the neutronics, and a feedback transfer function, which describes the thermal hydraulics. The result is a low-order approximate solution describing BWR behavior during small-amplitude limit-cycle oscillations. The approximate solution has been used to obtain a stability condition, show that the average reactor power must increase during limit-cycle oscillations, and qualitatively determine how changes in transfer function values affect the limit-cycle amplitude. 6 refs., 2 figs., 2 tabs

Experimental study of radiation dose rate at different strategic points of the BAEC TRIGA Research Reactor.

Science.gov (United States)

Ajijul Hoq, M; Malek Soner, M A; Salam, M A; Haque, M M; Khanom, Salma; Fahad, S M

2017-12-01

The 3MW TRIGA Mark-II Research Reactor of Bangladesh Atomic Energy Commission (BAEC) has been under operation for about thirty years since its commissioning at 1986. In accordance with the demand of fundamental nuclear research works, the reactor has to operate at different power levels by utilizing a number of experimental facilities. Regarding the enquiry for safety of reactor operating personnel and radiation workers, it is necessary to know the radiation level at different strategic points of the reactor where they are often worked. In the present study, neutron, beta and gamma radiation dose rate at different strategic points of the reactor facility with reactor power level of 2.4MW was measured to estimate the rising level of radiation due to its operational activities. From the obtained results high radiation dose is observed at the measurement position of the piercing beam port which is caused by neutron leakage and accordingly, dose rate at the stated position with different reactor power levels was measured. This study also deals with the gamma dose rate measurements at a fixed position of the reactor pool top surface for different reactor power levels under both Natural Convection Cooling Mode (NCCM) and Forced Convection Cooling Mode (FCCM). Results show that, radiation dose rate is higher for NCCM in compared with FCCM and increasing with the increase of reactor power. Thus, concerning the radiological safety issues for working personnel and the general public, the radiation dose level monitoring and the experimental analysis performed within this paper is so much effective and the result of this work can be utilized for base line data and code verification of the nuclear reactor. Copyright © 2017 Elsevier Ltd. All rights reserved.

Aging management program of the reactor building concrete at Point Lepreau Generating Station

Directory of Open Access Journals (Sweden)

Gendron T.

2011-04-01

Full Text Available In order for New Brunswick Power Nuclear (NBPN to control the risks of degradation of the concrete reactor building at the Point Lepreau Generating Station (PLGS the development of an aging management plan (AMP was initiated. The intention of this plan was to determine the requirements for specific structural components of concrete of the reactor building that require regular inspection and maintenance to ensure the safe and reliable operation of the plant. The document is currently in draft form and presents an integrated methodology for the application of an AMP for the concrete of the reactor building. The current AMP addresses the reactor building structure and various components, such as joint sealant and liners that are integral to the structure. It does not include internal components housed within the structure. This paper provides background information regarding the document developed and the strategy developed to manage potential degradation of the concrete of the reactor building, as well as specific programs and preventive and corrective maintenance activities initiated.

Kinetic mechanism of the decomposition of dimethyltin dichloride

NARCIS (Netherlands)

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

The theory of dissipative structures of the kinetic system for defects of nonlinear physical system 'metal+loading+irradiation'. Part 3

International Nuclear Information System (INIS)

Tarasov, V.A.; Borikov, T.L.; Kryzhanovskaya, T.V.; Chernezhenko, S.A.; Rusov, V.D.

2007-01-01

The kinetic system for defects of physical nonlinear system 'metal + load + irradiation' is specified [1, 2, 3]. Developing the approaches offered in [4], where distinctions of mechanisms of radiating creep and areas of their applicability are formalized (depending on external parameters) for fuel and constructional metals, division of kinetic systems for defects of constructional and fuel metals is carrying out. Thus the accent on the autocatalytic features of kinetic system for defects of reactor fuel metals, resulting from the exoenergic autocatalytic character of nuclear fission reactions being the main point defect source is done. In this part of the article the basic attention is given to the kinetic of sink drains for point defects. For kinetic systems of sinks-sources new approaches for the task of boundary conditions are offered. The possible structure of the computer program modelling kinetic system for defects of nonlinear physical system 'metal + load + irradiation' is considered

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)

Development of a generalized stochastic model for the analysis of monoenergetic space-time nuclear factor Kinetics

International Nuclear Information System (INIS)

Pham, Nhu Viet Ha

2011-02-01

(SSKM), which can be considered as a generalization of the stochastic point-kinetics equations, was then validated by testing against analog Monte Carlo calculations for cases of slab reactors. The comparative results showed that the SSKM agrees well within about 5% with the Monte Carlo method. Also, the SSKM was confirmed to provide a fast calculation method in comparison with the Monte Carlo computations. Finally, an additional numerical investigation was conducted using the SSKM in order to observe the random fluctuations in population dynamics under various reactor transients. The numerical results, which appear plausible on a physical basis, demonstrate that the behavior of the stochastic neutron and precursor distributions within a reactor can be explicitly induced by using the SSKM. Thus, it is realized that the SSKM can be applied to reactor transient and safety analysis with possibility of predicting the trends in variances of the interacting populations in advance

Solution of Point Reactor Neutron Kinetics Equations with Temperature Feedback by Singularly Perturbed Method

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.

Determination of the parameters of the point kinetics equation of a nuclear reactor by the quasilinearization technique

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

Coupled hydro-neutronic calculations for fast burst reactor accidents

International Nuclear Information System (INIS)

Paternoster, R.; Kimpland, R.; Jaegers, P.; McGhee, J.

1994-01-01

Methods are described for determining the fully coupled neutronic/hydrodynamic response of fast burst reactors (FBR) under disruptive accident conditions. Two code systems, PAD (1 -D Lagrangian) and NIKE-PAGOSA (3-D Eulerian) were used to accomplish this. This is in contrast to the typical methodology that computes these responses by either single point kinetics or in a decoupled manner. This methodology is enabled by the use of modem supercomputers (CM-200). Two examples of this capability are presented: an unreflected metal fast burst assembly, and a reflected fast burst assembly typical of the Skua or SPR-III class of fast burst reactor

Photocatalytic Oxidation of Azo Dyes and Oxalic Acid in Batch Reactors and CSTR: Introduction of Photon Absorption by Dyes to Kinetic Models

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.

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)

Experimental measurement of zero power reactor transfer function

International Nuclear Information System (INIS)

Liang Shuhong

2011-01-01

In order to study the zero power reactor (ZPR) transfer function, the ZPR transfer function expression was deduced with the point reactor kinetics equation, which was disturbed by reactivity input response. Based on the Fourier analysis for the input of triangular wave, the relation between the transfer function and reactivity was got. Validating research experiment was made on the DF-VI fast ZPR. After the disturbed reactivity was measured, the experimental value of the transfer function was got. According to the experimental value and the calculated value, the expression of the ZPR transfer function is proved, whereas the disturbed reactivity is got from the transfer function. (authors)

Detailed Reaction Kinetics for CFD Modeling of Nuclear Fuel Pellet Coating for High Temperature Gas-Cooled Reactors

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

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

International Nuclear Information System (INIS)

Jeong, Jae Jun; Chung, Bub Dong

2005-09-01

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

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

Study of decomposition kinetics of volatile β-diketonates of yttrium, barium and copper in flow reactor

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

The solution of the point kinetics equations via converged accelerated Taylor series (CATS)

Energy Technology Data Exchange (ETDEWEB)

Ganapol, B.; Picca, P. [Dept. of Aerospace and Mechanical Engineering, Univ. of Arizona (United States); Previti, A.; Mostacci, D. [Laboratorio di Montecuccolino, Alma Mater Studiorum - Universita di Bologna (Italy)

2012-07-01

This paper deals with finding accurate solutions of the point kinetics equations including non-linear feedback, in a fast, efficient and straightforward way. A truncated Taylor series is coupled to continuous analytical continuation to provide the recurrence relations to solve the ordinary differential equations of point kinetics. Non-linear (Wynn-epsilon) and linear (Romberg) convergence accelerations are employed to provide highly accurate results for the evaluation of Taylor series expansions and extrapolated values of neutron and precursor densities at desired edits. The proposed Converged Accelerated Taylor Series, or CATS, algorithm automatically performs successive mesh refinements until the desired accuracy is obtained, making use of the intermediate results for converged initial values at each interval. Numerical performance is evaluated using case studies available from the literature. Nearly perfect agreement is found with the literature results generally considered most accurate. Benchmark quality results are reported for several cases of interest including step, ramp, zigzag and sinusoidal prescribed insertions and insertions with adiabatic Doppler feedback. A larger than usual (9) number of digits is included to encourage honest benchmarking. The benchmark is then applied to the enhanced piecewise constant algorithm (EPCA) currently being developed by the second author. (authors)

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

Energy Technology Data Exchange (ETDEWEB)

Isaac K. Gamwo; Dimitri Gidaspow

1999-09-01

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

Effect of kinetic parameters on simultaneous ramp reactivity insertion plus beam tube flooding accident in a typical low enriched U{sub 3}Si{sub 2}-Al fuel-based material testing reactor-type research reactor

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.

"Batch" kinetics in flow: online IR analysis and continuous control.

Science.gov (United States)

Moore, Jason S; Jensen, Klavs F

2014-01-07

Currently, kinetic data is either collected under steady-state conditions in flow or by generating time-series data in batch. Batch experiments are generally considered to be more suitable for the generation of kinetic data because of the ability to collect data from many time points in a single experiment. Now, a method that rapidly generates time-series reaction data from flow reactors by continuously manipulating the flow rate and reaction temperature has been developed. This approach makes use of inline IR analysis and an automated microreactor system, which allowed for rapid and tight control of the operating conditions. The conversion/residence time profiles at several temperatures were used to fit parameters to a kinetic model. This method requires significantly less time and a smaller amount of starting material compared to one-at-a-time flow experiments, and thus allows for the rapid generation of kinetic data. Copyright © 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Noise analysis of the Dodewaard boiling water reactor: characteristics and time history

International Nuclear Information System (INIS)

Veer, J.H.C. v.d.; Kema, N.V.

1982-01-01

Reactor noise measurements have been performed in the Dodewaard BWR since the eighth fuel cycle (1978). Analysis of the noise characteristics of the ex-core neutron detectors are reported. As a result characteristics of the global component of the boiling noise and the influence of oscillatory effects in reactor pressure control and steam flow rate are described. The influence of power feedback effects on the detection of global noise at low frequencies is given using point kinetic reactor theory. Results are reported on the behaviour of the neutron noise characteristics during one fuel cycle and on the behaviour from fuel cycle 8 to 11. (author)

PREMOR: a point reactor exposure model computer code for survey analysis of power plant performance

International Nuclear Information System (INIS)

Vondy, D.R.

1979-10-01

The PREMOR computer code was written to exploit a simple, two-group point nuclear reactor power plant model for survey analysis. Up to thirteen actinides, fourteen fission products, and one lumped absorber nuclide density are followed over a reactor history. Successive feed batches are accounted for with provision for from one to twenty batches resident. The effect of exposure of each of the batches to the same neutron flux is determined

PREMOR: a point reactor exposure model computer code for survey analysis of power plant performance

Energy Technology Data Exchange (ETDEWEB)

Vondy, D.R.

1979-10-01

The PREMOR computer code was written to exploit a simple, two-group point nuclear reactor power plant model for survey analysis. Up to thirteen actinides, fourteen fission products, and one lumped absorber nuclide density are followed over a reactor history. Successive feed batches are accounted for with provision for from one to twenty batches resident. The effect of exposure of each of the batches to the same neutron flux is determined.

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-01

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

Intrinsic reaction kinetics of coal char combustion by direct measurement of ignition temperature

International Nuclear Information System (INIS)

Kim, Ryang-Gyoon; Jeon, Chung-Hwan

2014-01-01

A wire heating reactor that can use a synchronized experimental method was developed to obtain the intrinsic kinetics of large coal char particles ranging in size from 0.4 to 1 mm. This synchronization system consists of three parts: a thermocouple wire for both heating and direct measurement of the particle temperature, a photodetector sensor for determining ignition/burnout points by measuring the intensity of luminous emission from burning particles, and a high-speed camera–long-distance microscope for observing and recording the movement of luminous zone directly. Coal char ignition was found to begin at a spot on the particle's external surface and then moved across the entire particle. Moreover, the ignition point determined according to the minimum of dT/dt is a spot point and not a full growth point. The ignition temperature of the spot point rises as the particle diameter increases. A spot ignition model, which describes the ignition in terms of the internal conduction and external/internal oxygen diffusion, was then developed to evaluate the intrinsic kinetics and predict the ignition temperature of the coal char. Internal conduction was found to be important in large coal char particles because its effect becomes greater than that of oxygen diffusion as the particle diameter increases. In addition, the intrinsic kinetics of coal char obtained from the spot ignition model for two types of coal does not differ significantly from the results of previous investigators. -- Highlights: • A novel technique was used to measure the coal char particle temperature. • The ignition point determined from a dT/dt minimum is a spot ignition point. • A spot ignition model was suggested to analyze the intrinsic reaction kinetics of coal char. • Internal conduction has to be considered in order to evaluate the intrinsic kinetics for larger particle (above 1 mm)

Ozonation kinetics of winery wastewater in a pilot-scale bubble column reactor.

Science.gov (United States)

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.

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.)

Kinetic modeling of particle acceleration in a solar null point reconnection region

DEFF Research Database (Denmark)

Baumann, Gisela; Haugbølle, Troels; Nordlund, Åke

2013-01-01

The primary focus of this paper is on the particle acceleration mechanism in solar coronal 3D reconnection null-point regions. Starting from a potential field extrapolation of a SOHO magnetogram taken on 2002 November 16, we first performed MHD simulations with horizontal motions observed by SOHO...... particles and 3.5 billion grid cells of size 17.5\\,km --- these simulations offer a new opportunity to study particle acceleration in solar-like settings....... applied to the photospheric boundary of the computational box. After a build-up of electric current in the fan-plane of the null-point, a sub-section of the evolved MHD data was used as initial and boundary conditions for a kinetic particle-in-cell model of the plasma. We find that sub...

Dynamical Analysis of a Continuous Stirred-Tank Reactor with the Formation of Biofilms for Wastewater Treatment

Directory of Open Access Journals (Sweden)

Karen López Buriticá

2015-01-01

Full Text Available This paper analyzes the dynamics of a system that models the formation of biofilms in a continuous stirred-tank reactor (CSTR when it is utilized for wastewater treatment. The growth rate of the microorganisms is modeled using two different kinetics, Monod and Haldane kinetics, with the goal of studying the influence of each in the system. The equilibrium points are identified through a stability analysis, and the bifurcations found are characterized.

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)

A 3D nodal mixed dual method for nuclear reactor kinetics with improved quasistatic model and a semi-implicit scheme to solve the precursor equations

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

Reactor for in situ measurements of spatially resolved kinetic data in heterogeneous catalysis

Science.gov (United States)

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.

Further experience in simulation of rod drop experiments in the Loviisa and Mochovce reactors

International Nuclear Information System (INIS)

Siltanen, P.; Kaloinen, E.; Tanskanen, A.; Mattila, R.

2001-01-01

Simulations of reactor scram experiments using the 3-dimensional kinetics code HEXTRAN have been updated for the initial cores of Loviisa-1 and 2 Mochovce-1 and have been extended to burned cores of Loviisa-1. In these simulations, the entire experiment is simulated dynamically, including the behaviour of the core, the signal of the ionization chamber, and the inverse point kinetics of the reactivity meter. The predicted output of the reactivity meter is compared with the output observed during the experiment (Authors)

Dynamic power behavior of a PWR type nuclear reactor

International Nuclear Information System (INIS)

Moreira, F.J.

1984-01-01

A methodology for the power level evaluation (dynamic behavior) in a Pressurized Water Reactor, during a transient is developed, by solving the point kinetic equation related to the control rod insertion effects and fuel or moderator temperature 'feed-back'. A new version of the thermal-hydraulic code COBRA III P/MIT, is used. In this new version was included, as an option, the methodology developed. (E.G.) [pt

Studies of tokamak fusion reactor dynamics. Progress report, June 1, 1975--February 15, 1976

International Nuclear Information System (INIS)

Mills, R.G.; Gralnick, S.L.

1976-01-01

An investigation of the effect of plasma shape and position on the inductive coupling between the plasma and the external poloidal field coils is briefly described. Research on a multi-node time-dependent point kinetics code with which to study the operating dynamics of a tokamak reactor is also mentioned

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.

Calculation of limit cycle amplitudes in commercial boiling water reactors

International Nuclear Information System (INIS)

March-Leuba, J.; Perez, R.B.; Cacuci, D.G.

1984-01-01

This paper describes an investigation of the dynamic behavior of a boiling water reactor (BWR) in the nonlinear region corresponding to linearly unstable conditions. A nonlinear model of a typical BWR was developed. The equations underlying this model represent a one-dimensional void reactivity feedback, point kinetics with a single delayed neutron group, fuel behavior, and recirculation loop dynamics (described by a single-node integral momentum equation)

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

Characters of neutron noise in full-size molten salt reactor

International Nuclear Information System (INIS)

Wang, Jiangmeng; Cao, Xinrong

2015-01-01

Highlights: • The larger system size makes full-size MSR deviate from point kinetic behavior. • The increasing velocity has non-monotonic effect on the effective delayed neutron fraction. • The amplitude of Green’s function at low frequencies is inversely proportional to the effective delayed neutron fraction. • The range of plateau region is smaller due to the more prominent point kinetic effect. - Abstract: In the present paper, the frequency-dependent and space-dependent behavior of the neutron noise in a full-size Molten Salt Reactor (MSR) is investigated. The theoretical models considering the fuel circulation are established based on one-group neutron diffusion theory. Green’s function of the neutron noise induced by a propagating perturbation is introduced with linear noise theory, due to the small perturbation. The equations are numerically calculated by developing a code, in which the eigenfunction expansion method is adopted. The static results show that the effective delayed neutron fraction changes non-monotonically with the increasing fuel velocity. In the dynamic case, the results are compared to those obtained in 1D MSR and a traditional reactor, in order to figure out the effects of both the fuel circulation and the system size. It is found that there is no difference in 1D and 3D MSR systems from the view of fuel circulation, i.e., the fuel circulation enhances the spatial neutronic coupling and leads to the stronger point kinetic effect. The more prominent space-dependent effect founded in 3D traditional reactors is also observed in the MSR, due to the looser neutronic coupling and the unique singularity of Green’s function in the location of the perturbation. Another interesting finding is that Green’s function for low frequencies changes non-monotonically with increasing velocity. The largest magnitude of Green’s function is observed at the velocity where the effective delayed neutron fraction reaches its minimum. Finally, the

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.

De-mercurization of wastewater by Bacillus cereus (JUBT1): Growth kinetics, biofilm reactor study and field emission scanning electron microscopic analysis

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.

An automatic regulating control system for a graphite moderated reactor using digital techniques

International Nuclear Information System (INIS)

Carvalho Goncalves Junior, J. de.

1989-01-01

The work propose an automatic regulating control system for a graphite moderated reactor using digital techniques. The system uses a microcomputer to monitor the power and the period, to run the control algorithm, and to generate electronic signals to excite the motor, which moves vertically the control rod banks. A nuclear reactor simulator was developed to test the control system. The simulator consists of a software based on the point kinetic equations and implanted in an analogical computer. The results show that this control system has a good performance and versatility. In addition, the simulator is capable of reproducing with accuracy the behavior of a nuclear reactor. (author)

Determination of the kinetic parameters of the CALIBAN metallic core reactor from stochastic neutron measurements

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)

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

International Nuclear Information System (INIS)

Gui Xuewen; Cai Qi; Luo Bangqi

2007-01-01

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

MIDAS/PK code development using point kinetics model

International Nuclear Information System (INIS)

Song, Y. M.; Park, S. H.

1999-01-01

In this study, a MIDAS/PK code has been developed for analyzing the ATWS (Anticipated Transients Without Scram) which can be one of severe accident initiating events. The MIDAS is an integrated computer code based on the MELCOR code to develop a severe accident risk reduction strategy by Korea Atomic Energy Research Institute. In the mean time, the Chexal-Layman correlation in the current MELCOR, which was developed under a BWR condition, is appeared to be inappropriate for a PWR. So as to provide ATWS analysis capability to the MIDAS code, a point kinetics module, PKINETIC, has first been developed as a stand-alone code whose reference model was selected from the current accident analysis codes. In the next step, the MIDAS/PK code has been developed via coupling PKINETIC with the MIDAS code by inter-connecting several thermal hydraulic parameters between the two codes. Since the major concern in the ATWS analysis is the primary peak pressure during the early few minutes into the accident, the peak pressure from the PKINETIC module and the MIDAS/PK are compared with the RETRAN calculations showing a good agreement between them. The MIDAS/PK code is considered to be valuable for analyzing the plant response during ATWS deterministically, especially for the early domestic Westinghouse plants which rely on the operator procedure instead of an AMSAC (ATWS Mitigating System Actuation Circuitry) against ATWS. This capability of ATWS analysis is also important from the view point of accident management and mitigation

Neutron lifetimes behavior analysis considering the two-region kinetic model in the IPEN/MB-01 reactor

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.

Neutron lifetimes behavior analysis considering the two-region kinetic model in the IPEN/MB-01 reactor

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

Physics of nuclear reactors

International Nuclear Information System (INIS)

Baeten, Peter

2006-01-01

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

Estimation of Adjoint-Weighted Kinetics Parameters in Monte Carlo Wieland Calculations

International Nuclear Information System (INIS)

Choi, Sung Hoon; Shim, Hyung Jin

2013-01-01

The effective delayed neutron fraction, β eff , and the prompt neutron generation time, Λ, in the point kinetics equation are weighted by the adjoint flux to improve the accuracy of the reactivity estimate. Recently the Monte Carlo (MC) kinetics parameter estimation methods by using the self-consistent adjoint flux calculated in the MC forward simulations have been developed and successfully applied for the research reactor analyses. However these adjoint estimation methods based on the cycle-by-cycle genealogical table require a huge memory size to store the pedigree hierarchy. In this paper, we present a new adjoint estimation in which the pedigree of a single history is utilized by applying the MC Wielandt method. The effectiveness of the new method is demonstrated in the kinetics parameter estimations for infinite homogeneous two-group problems and the Godiva critical facility

Kinetic parameters of biomass growth in a UASB reactor treating wastewater from coffee wet processing (WCWP

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.

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

Energy Technology Data Exchange (ETDEWEB)

Jeong, Jae Jun; Chung, Bub Dong

2005-09-15

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

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)

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)

CENTAR code for extended nonlinear transient analysis of extraterrestrial reactor systems

International Nuclear Information System (INIS)

Nassersharif, B.; Peer, J.S.; DeHart, M.D.

1987-01-01

Current interest in the application of nuclear reactor-driven power systems to space missions has generated a need for a systems simulation code to model and analyze space reactor systems; such a code has been initiated at Texas A and M, and the first version is nearing completion; release was anticipated in the fall of 1987. This code, named CENTAR (Code for Extended Nonlinear Transient Analysis of Extraterrestrial Reactor Systems), is designed specifically for space systems and is highly vectorizable. CENTAR is composed of several specialized modules. A fluids module is used to model fluid behavior throughout the system. A wall heat transfer module models the heat transfer characteristics of all walls, insulation, and structure around the system. A fuel element thermal analysis module is used to predict the temperature behavior and heat transfer characteristics of the reactor fuel rods. A kinetics module uses a six-group point kinetics formulation to model reactivity feedback and control and the ANS 5.1 decay-heat curve to model shutdown decay-heat production. A pump module models the behavior of thermoelectric-electromagnetic pumps, and a heat exchanger module models not only thermal effects in thermoelectric heat exchangers, but also predicts electrical power production for a given configuration. Finally, an accumulator module models coolant expansion/contraction accumulators

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)

Protection set-points lines for the reactor core and considerations about power distribution and peak factors

International Nuclear Information System (INIS)

Furieri, E.B.

1981-01-01

In order to assure the reactor core integrity during the slow operational transients (power excursion above the nominal value and the high coolant temperature), the formation of a steam film (DNB-Departure from Nucleate Boiling) in the control rods must be avoided. The protection set points lines presents the points where DNBR (relation between critical heat flux-q sub(DNB) and the local heat flux-q' sub(local) is equal to 1.30, corrected by peak factors and uncertainty in function of ΔTr and T sub(R), respectively coolant elevation and medium coolant temperature in reactor pressure vessel. The curve set-points were determined using a new version of COBRA-IIIF (CUPRO) computer code, implemented with new subroutines and linearized convergence scheme. Pratical results for Angra-1 core were obtained and its were compared with the results from the fabricator. (E.G.) [pt

Assessments of the kinetic and dynamic transient behavior of sub-critical systems (ADS) in comparison to critical reactor systems

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

Ab-initio modelling of thermodynamics and kinetics of point defects in indium oxide

International Nuclear Information System (INIS)

Agoston, Peter; Klein, Andreas; Albe, Karsten; Erhart, Paul

2008-01-01

The electrical and optical properties of indium oxide films strongly vary with the processing parameters. Especially the oxygen partial pressure and temperature determine properties like electrical conductivity, composition and transparency. Since this material owes its remarkable properties like the intrinsic n-type conductivity to its defect chemistry, it is important to understand both, the equilibrium defect thermodynamics and kinetics of the intrinsic point defects. In this contribution we present a defect model based on DFT total energy calculations using the GGA+U method. Further, the nudged elastic band method is employed in order to obtain a set of migration barriers for each defect species. Due to the complicated crystal structure of indium oxide a Kinetic Monte-Carlo algorithm was implemented, which allows to determine diffusion coefficients. The bulk tracer diffusion constant is predicted as a function of oxygen partial pressure, Fermi level and temperature for the pure material

Research Reactor Power Control System Design by MATLAB/SIMULINK

International Nuclear Information System (INIS)

Baang, Dane; Suh, Yong Suk; Kim, Young Ki; Im, Ki Hong

2013-01-01

In this study it is presented that MATLAB/SIMULINK can be efficiently used for modeling and power control system design for research reactors. The presented power control system deals with various functions including reactivity control, signals processing, reactivity calculation, alarm request generation, etc., thus it is required to test all the software logic using proper model for reactor, control rods, and field instruments. In MATLAB/SIMULINK tool, point kinetics, thermal model, control absorber rod model, and other instrument models were developed based on reactor parameters and known properties of each component or system. The software for power control system was invented and linked to the model to test each function. From the simulation result it is shown that the power control performance and other functions of the system can be easily tested and analyzed in the proposed simulation structure

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Two Proposals for determination of large reactivity of reactor

International Nuclear Information System (INIS)

Kaneko, Yoshihiko; Nagao, Yoshiharu; Yamane, Tsuyoshi; Takeuchi, Mituo

1999-01-01

Two Proposals for determination of large reactivity of reactors are presented. One is for large positive reactivity. The other is for large negative reactivity. Existing experimental methods for determination of large positive reactivity, the fuel addition method and the neutron adsorption substitution method were analyzed. It is found that both the experimental methods are possibly affected to the substantially large systematic error up to ∼ 20%, when the value of the excess multiplication factor comes into the range close to ∼20%Δk. To cope with this difficulty, a revised method is validly proposed. The revised method evaluates the value of the potential excess multiplication factor as the consecutive increments of the effective multiplication factor in a virtual core, which are converted from those in an actual core by multiplying a conversion factor f to it. The conversion factor f is to be obtained in principle by calculation. Numerical experiments were done on a slab reactor using one group diffusion model. The rod drop experimental method is widely used for determination of large negative negative reactivity values. The decay of the neutron density followed by initiating the insertion of the rod is obliged to be slowed down according to its speed. It is proved by analysis based on the one point reactor kinetics that in such a case the integral counting method hitherto used tend to significantly underestimate the absolute values of negative reactivity, even if the insertion time is in the range of 1-2 s. As for the High Temperature Engineering Test Reactor (HTTR), the insertion time will be lengthened up to 4-6 s. In order to overcome the difficulty , the delayed integral counting method is proposed, in which the integration of neutron counting starts after the rod drop has been completed and the counts before is evaluated by calculation using one point reactor kinetics. This is because the influence of the insertion time on the decay of the neutron

Simulation of the accumulation kinetics for radiation point defects in a metals with impurity

International Nuclear Information System (INIS)

Iskakov, B.M.; Nurova, A.B.

2001-01-01

In the work a kinetics of vacancies (V) and interstitial atoms (IA) accumulation for cases when the V and IA are recombining with each other, absorbing by drain and capturing by impurity atoms has been simulated. The differential equations system numerical solution was carried out by the Runge-Kutta method. The dynamical equilibrium time achievement for the point radiation defects accumulation process in the metal with impurity is considered

RIA Analysis of Unprotected TRIGA Reactor

Directory of Open Access Journals (Sweden)

M.H. Altaf

2017-07-01

Full Text Available An RIA (reactivity initiated accident analysis has been carried out for the TRIGA Mark II research reactor considering both step and ramp reactivity ranges within 0.5 % dk/k (< $1 to 2.0 % dk/k (>$2. The insertion time was set at 10 s. Based on the fact that a reactor becomes unprotected if scram does not work at the event of danger, to define unprotected conditions, the time to actuate scram (trip was taken as close to total simulation time. In this long duration of scram inactivity, it is obtained from the present analysis that the reactor remained safe to up to 1.8 % dk/k ($2.57 for step reactivity and 1.99 % dk/k ($2.84 for ramp reactivity. In addition to negative temperature coefficient of reativity, probably the longer time of reactivity insertion keeps TRIGA safe even at larger magnitudes of reactivity during unprotected reactor transients. Coupled point kinetics, neutronics, and thermal hydraulics code EUREKA-2/R has been utilized for this work. It appears that EUREKA-2/RR predicts the sequence of unprotected transient scenario of TRIGA core with good approximation and the results will definitely be helpful for the reactor operators.

A generalized electron energy probability function for inductively coupled plasmas under conditions of nonlocal electron kinetics

Science.gov (United States)

Mouchtouris, S.; Kokkoris, G.

2018-01-01

A generalized equation for the electron energy probability function (EEPF) of inductively coupled Ar plasmas is proposed under conditions of nonlocal electron kinetics and diffusive cooling. The proposed equation describes the local EEPF in a discharge and the independent variable is the kinetic energy of electrons. The EEPF consists of a bulk and a depleted tail part and incorporates the effect of the plasma potential, Vp, and pressure. Due to diffusive cooling, the break point of the EEPF is eVp. The pressure alters the shape of the bulk and the slope of the tail part. The parameters of the proposed EEPF are extracted by fitting to measure EEPFs (at one point in the reactor) at different pressures. By coupling the proposed EEPF with a hybrid plasma model, measurements in the gaseous electronics conference reference reactor concerning (a) the electron density and temperature and the plasma potential, either spatially resolved or at different pressure (10-50 mTorr) and power, and (b) the ion current density of the electrode, are well reproduced. The effect of the choice of the EEPF on the results is investigated by a comparison to an EEPF coming from the Boltzmann equation (local electron kinetics approach) and to a Maxwellian EEPF. The accuracy of the results and the fact that the proposed EEPF is predefined renders its use a reliable alternative with a low computational cost compared to stochastic electron kinetic models at low pressure conditions, which can be extended to other gases and/or different electron heating mechanisms.

Fuel cycle kinetics

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

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%

The extension of the SWS period or CANDU reactors with particular reference to Douglas Point

International Nuclear Information System (INIS)

Bennett, C.R.

1985-01-01

The foregoing approach to the determination of the fate of a concrete containment building is worth much consideration. The expenditure of $10 8 or its escalated equivalent is too much to pay for the probable saving of fraction of a statistical life. The unquestioning adoption of the dogma of reactor dismantlement displays a complete misunderstanding of the numerics of ''risk'', even the place of reactor dismantling in the spectrum of nuclear risk. The position of the risk of reactor dismantling is more than an order of magnitude lower than the former of these. The most altruistic criterion for any engineering activity is the achievement of the greatest expected net benefit (or the least expected net detriment) when all the consequences of the activity are taken into account. As has been shown this criterion leads to the conclusion that, at least in CANDU reactors and particularly Douglas Point, there is apparently no reason why the S.W.S. period should not be extended indefinitely

Simplified methodology for control cell constant calculations of the reactor cores for the space kinetics

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)

Initial verification and validation of RAZORBACK - A research reactor transient analysis code

Energy Technology Data Exchange (ETDEWEB)

Talley, Darren G. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

2015-09-01

This report describes the work and results of the initial verification and validation (V&V) of the beta release of the Razorback code. Razorback is a computer code designed to simulate the operation of a research reactor (such as the Annular Core Research Reactor (ACRR)) by a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This initial V&V effort was intended to confirm that the code work to-date shows good agreement between simulation and actual ACRR operations, indicating that the subsequent V&V effort for the official release of the code will be successful.

The reflector effect on the neutron lifetimes in the IPEN/MB-01 reactor; O efeito do refletor sobre o tempo de vida neutronico no reator IPEN/MB-01

Energy Technology Data Exchange (ETDEWEB)

Gonnelli, Eduardo

2013-07-01

The aim of this study is to present the reflector effect on the neutron lifetimes in the IPEN/MB-01 reactor. The proposed method requires an approach which takes into account both the reflector and the core, so that the point kinetics equations, which constitute the theoretical basis of all mathematical development, contemplate both regions of the reactor. From these equations, as known as two regions kinetics point equations, theoretical expressions are obtained for the Auto Power Spectral Densities (APSD), which are used for least squares fit of the experimental data of APSD obtained in several subcritical states. The prompt neutron generation time, the neutron lifetimes in the reflector and the neutron return fraction from the reflector to the core are derived from the fitting. (author)

Sensitivity analysis of the kinetic behaviour of a Gas Cooled Fast Reactor to variations of the delayed neutron parameters

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)

Qualification of SiC materials for fusion and fission reactors

International Nuclear Information System (INIS)

Ryazanov, Alexander

2009-01-01

Ceramic materials such as silicon carbide (SiC) and SiC/SiC composites are both considered, due to their high-temperature strength, pseudo-ductile fracture behavior and low-induced radioactivity, as candidate materials for fusion reactor (test blanket module for ITER) and high temperature gas-cooled reactors (HTGR). The radiation swelling and creep of SiC are very important physical phenomena that determine the radiation resistance of them in these reactors. Other important problem which exists especially in fusion reactor is an effect of accumulation of high concentrations of helium atoms in SiC (up to 15000-20000 at.ppm) due to (n,α) nuclear reaction on physical mechanical properties. An understanding of the physical mechanism of this phenomenon is very important for the investigations of helium atom effect on radiation swelling in SiC. In this report a compilation of non-irradiated and irradiated properties of SiC are provided and analyzed in terms of their application to fusion and high temperature gas cooled reactors. Special topic of this report is oriented on the micro structural changes in chemically vapor-deposited (CVD) high-purity beta-SiC during neutron and ion irradiations at elevated temperatures. The evolutions of various radiation induced defects including dislocation loops, network dislocations and cavities are presented here as a function of irradiation temperature and fluencies. These observations are discussed in relation with such irradiation phenomena in SiC as low temperature swelling and cavity swelling. One of the main difficulties in the radiation damage studies of SiC materials lies in the absence of theoretical models and interpretation of many physical mechanisms of radiation phenomena including the radiation swelling and creep. The point defects in ceramic materials are characterized by the charge states and they can have an effective charge. The internal effective electrical field is formed due to the accumulation of charged point

Safety reviews of the Brazilian multipurpose reactor; Avaliacoes de seguranca do reator multiproposito brasileiro

Energy Technology Data Exchange (ETDEWEB)

Soares, Humberto Vitor

2014-08-01

This work presents a model developed for thermal hydraulic (TH) simulation of the Multipurpose Brazilian Reactor (RMB), whose Brazilian proposal for design, construction and operation was established in 2007. This reactor has as main proposed the production of radioisotopes for use in exams of nuclear medicine, material tests and utilization of neutrons beams. Besides of the TH modeling and safety analysis of the reactor, the application of a methodology to perform coupled calculation thermal-hydraulic/neutron kinetic (TH/NK) is also presented. Initially, the RMB was modeled in the safety analysis RELAP5 code. This code performs the thermal hydraulic calculation using point kinetics. Subsequently, the model was adapted and verified to the RELAP5-3D© code. This code performs the process of internal coupling through the option of nodal neutron kinetics calculation using the NESTLE code which solves the neutron diffusion equation. To generate the neutronic group constants, which are macroscopic cross sections that serve as input data for the neutronic codes, it was used the WIMSD-5B cell calculation code. The neutron analysis code PARCS was also used to model the 3D RMB core in order to compare the results of radial and axial average power distribution with the results generated by RELAP5-3D© code and with the available results of the CITATION neutron kinetic code. The safety analyses demonstrated safe behavior of the reactor through situations of possible transients. The 3D coupled calculations to the steady state operation also showed expected behavior, as well as the RMB neutronic analyzes performed with the codes NESTLE and PARCS.(author)

Experimental determination of neutron lifetimes through macroscopic neutron noise in the IPEN/MB-01 reactor

Energy Technology Data Exchange (ETDEWEB)

Gonnelli, Eduardo; Diniz, Ricardo [Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP Travessa R-400, 05508-900, Cidade Universitaria, Sao Paulo (Brazil)

2013-05-06

The neutron lifetimes of the core, reflector, and global were experimentally obtained through macroscopic neutron noise in the IPEN/MB-01 reactor for five levels of subcriticality. The theoretical Auto Power Spectral Densities were derived by point kinetic equations taking the reflector effect into account, and one of the approaches consider an additional group of delayed neutrons.

Dynamic Modeling for the Design and Cyclic Operation of an Atomic Layer Deposition (ALD Reactor

Directory of Open Access Journals (Sweden)

Curtisha D. Travis

2013-08-01

Full Text Available A laboratory-scale atomic layer deposition (ALD reactor system model is derived for alumina deposition using trimethylaluminum and water as precursors. Model components describing the precursor thermophysical properties, reactor-scale gas-phase dynamics and surface reaction kinetics derived from absolute reaction rate theory are integrated to simulate the complete reactor system. Limit-cycle solutions defining continuous cyclic ALD reactor operation are computed with a fixed point algorithm based on collocation discretization in time, resulting in an unambiguous definition of film growth-per-cycle (gpc. A key finding of this study is that unintended chemical vapor deposition conditions can mask regions of operation that would otherwise correspond to ideal saturating ALD operation. The use of the simulator for assisting in process design decisions is presented.

Kinetic modeling of cellulosic biomass to ethanol via simultaneous saccharification and fermentation: Part I. Accommodation of intermittent feeding and analysis of staged reactors.

Science.gov (United States)

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.

Microbial ureolysis in the seawater-catalysed urine phosphorus recovery system: Kinetic study and reactor verification.

Science.gov (United States)

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.

RAZORBACK - A Research Reactor Transient Analysis Code Version 1.0 - Volume 3: Verification and Validation Report.

Energy Technology Data Exchange (ETDEWEB)

Talley, Darren G.

2017-04-01

This report describes the work and results of the verification and validation (V&V) of the version 1.0 release of the Razorback code. Razorback is a computer code designed to simulate the operation of a research reactor (such as the Annular Core Research Reactor (ACRR)) by a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, the equation of motion for fuel element thermal expansion, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This V&V effort was intended to confirm that the code shows good agreement between simulation and actual ACRR operations.

Influence of Irradiance, Flow Rate, Reactor Geometry, and Photopromoter Concentration in Mineralization Kinetics of Methane in Air and in Aqueous Solutions by Photocatalytic Membranes Immobilizing Titanium Dioxide

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

New signal acquisition and processing system for the execution of initial criticality after refueling and physical tests at low power in Angra-2, with the incorporation of the real time resolution of the inverse point kinetic equation - IPK

Energy Technology Data Exchange (ETDEWEB)

Júnior, Décio Brandes M.F.; Oliveira, Mônica Georgia N.; Silva, Cristiano da, E-mail: deciobr@eletronuclear.gov.br, E-mail: mongeor@eletronuclear.gov.br, E-mail: cdsilva@eletronuclear.gov.br [Eletrobrás Termonuclear S.A. (ELETRONUCLEAR), Angra dos Reis, RJ (Brazil). Departamento DDD.O - Física de Reatores

2017-07-01

The goal of this work is present the new System of Acquisition and Signal Processing for the execution of the initial criticality after refueling and physical tests at low power with the incorporation of the real time resolution of Inverse Point Kinetic Equations (IPK). The system was developed using cRIO 9082 hardware (compactRIO), which is a programmable logic controller (PLC) and, the National Lab's LabVIEW programming language. The developed system enabled a better visualization and monitoring interface of the neutron flux evolution during the first criticality of cycle and following the low power physical tests, which allows the Reactor Physics Group and Reactor Operators of Angra 2 guide faster and accurately the reactivity variations at physical tests. The digital reactivity meter developed reinforces in Angra-2 the set of operational practices of reactivity management. (author)

New signal acquisition and processing system for the execution of initial criticality after refueling and physical tests at low power in Angra-2, with the incorporation of the real time resolution of the inverse point kinetic equation - IPK

International Nuclear Information System (INIS)

Júnior, Décio Brandes M.F.; Oliveira, Mônica Georgia N.; Silva, Cristiano da

2017-01-01

The goal of this work is present the new System of Acquisition and Signal Processing for the execution of the initial criticality after refueling and physical tests at low power with the incorporation of the real time resolution of Inverse Point Kinetic Equations (IPK). The system was developed using cRIO 9082 hardware (compactRIO), which is a programmable logic controller (PLC) and, the National Lab's LabVIEW programming language. The developed system enabled a better visualization and monitoring interface of the neutron flux evolution during the first criticality of cycle and following the low power physical tests, which allows the Reactor Physics Group and Reactor Operators of Angra 2 guide faster and accurately the reactivity variations at physical tests. The digital reactivity meter developed reinforces in Angra-2 the set of operational practices of reactivity management. (author)

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

Development of the coupled 'system thermal-hydraulics, 3D reactor kinetics, and hot channel' analysis capability of the MARS code

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.

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

A Point Kinetics Model for Estimating Neutron Multiplication of Bare Uranium Metal in Tagged Neutron Measurements

Science.gov (United States)

Tweardy, Matthew C.; McConchie, Seth; Hayward, Jason P.

2017-07-01

An extension of the point kinetics model is developed to describe the neutron multiplicity response of a bare uranium object under interrogation by an associated particle imaging deuterium-tritium (D-T) measurement system. This extended model is used to estimate the total neutron multiplication of the uranium. Both MCNPX-PoliMi simulations and data from active interrogation measurements of highly enriched and depleted uranium geometries are used to evaluate the potential of this method and to identify the sources of systematic error. The detection efficiency correction for measured coincidence response is identified as a large source of systematic error. If the detection process is not considered, results suggest that the method can estimate total multiplication to within 13% of the simulated value. Values for multiplicity constants in the point kinetics equations are sensitive to enrichment due to (n, xn) interactions by D-T neutrons and can introduce another significant source of systematic bias. This can theoretically be corrected if isotopic composition is known a priori. The spatial dependence of multiplication is also suspected of introducing further systematic bias for high multiplication uranium objects.

Numerical solution of multi groups point kinetic equations by simulink toolbox of Matlab software

International Nuclear Information System (INIS)

Hadad, K.; Mohamadi, A.; Sabet, H.; Ayobian, N.; Khani, M.

2004-01-01

The simulink toolbox of Matlab Software was employed to solve the point kinetics equation with six group delayed neutrons. The method of Adams-Bash ford showed a good convergence in solving the system of simultaneous equations and the obtained results showed good agreements with other numerical schemes. The flexibility of the package in changing the system parameters and the user friendly interface makes this approach a reliable educational package in revealing the affects of reactivity changes on power incursions

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

Modeling and Control of a Large Nuclear Reactor A Three-Time-Scale Approach

CERN Document Server

Shimjith, S R; Bandyopadhyay, B

2013-01-01

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

Assessment of fuel damage of pool type research reactor in the case of fuel plates blockage

Energy Technology Data Exchange (ETDEWEB)

Jalil, Jafari; Samad, Khakshournia [AEOI, Karegar Ave. School of R and D of Nuclear Reactors and Accelerators, Teheran (Iran, Islamic Republic of); D' Auria, F. [Pisa Univ., DIMNP (Italy)

2007-07-01

Tehran Research Reactor (TRR) is a pool type 5 MW research reactor. It is assumed that external objects or debris that may fall down to reactor core cause obstruction of coolant flow through one of the fuel assemblies. Thermal hydraulic analysis of this event, using the RELAP5 system code has been studied. The reported transient is related to the partial and total obstruction of a single Fuel Element (FE) cooling channel of 27 FE equilibrium core of TRR. Such event constitutes a severe accident for this type of reactor since it may lead to local dryout and eventually to loss of the FE integrity. Two scenarios are analysed to emphasize the severity of the accident. The first one is a partial blockage of an average FE considering four different obstruction levels: 25%, 50%, 75% and 97% of nominal flow area. The second one is an extreme scenario consisting of total blockage of the same FE. This study constitutes the first step of a larger work which consists of performing a 3-dimensional simulation using the Best Estimate coupled code technique. However, as a first approach the instantaneous reactor power is derived through the point kinetic calculation included in the RELAP5 code. Main results obtained from the RELAP5 calculations are as following. First, in the case of flow blockage under 97% of the nominal flow area of an average FE, only an increase of the coolant and clad temperatures is observed without any consequences for the integrity of the FE. The mass flow rate remains sufficient to cool the clad safely. Secondly, in the case of total obstruction of the nominal flow area, it is seen that transient turns out to be a severe accident due to the dryout conditions are reached shortly and melting of the cladding occurs. Thirdly, the use of the point kinetic approach leads to conservative results. A best estimate simulation of such kind of transients requires the use of 3-dimensional kinetic calculations, which could be done using the current Coupled Codes

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

Kinetics of coal pyrolysis

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.

Nanostructure evolution of neutron-irradiated reactor pressure vessel steels: Revised Object kinetic Monte Carlo model

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.

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.

Determination of the kinetics of ethene epoxidation

NARCIS (Netherlands)

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.

Experimental, kinetic and numerical modeling of hydrogen production by catalytic reforming of crude ethanol over a commercial catalyst in packed bed tubular reactor and packed bed membrane reactor

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)

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

Effect of DUPIC cycle on CANDU reactor safety parameters

Energy Technology Data Exchange (ETDEWEB)

Mohamed, Nader M. A. [Atomic Energy Authority, ETRR-2, Cairo (Egypt); Badawi, Alya [Dept. of Nuclear and Radiation Engineering, Alexandria University, Alexandria (Egypt)

2016-10-15

Although, the direct use of spent pressurized water reactor (PWR) fuel in CANda Deuterium Uranium (CANDU) reactors (DUPIC) cycle is still under investigation, DUPIC cycle is a promising method for uranium utilization improvement, for reduction of high level nuclear waste, and for high degree of proliferation resistance. This paper focuses on the effect of DUPIC cycle on CANDU reactor safety parameters. MCNP6 was used for lattice cell simulation of a typical 3,411 MWth PWR fueled by UO{sub 2} enriched to 4.5w/o U-235 to calculate the spent fuel inventories after a burnup of 51.7 MWd/kgU. The code was also used to simulate the lattice cell of CANDU-6 reactor fueled with spent fuel after its fabrication into the standard 37-element fuel bundle. It is assumed a 5-year cooling time between the spent fuel discharges from the PWR to the loading into the CANDU-6. The simulation was carried out to calculate the burnup and the effect of DUPIC fuel on: (1) the power distribution amongst the fuel elements of the bundle; (2) the coolant void reactivity; and (3) the reactor point-kinetics parameters.

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

Prompt Neutron Lifetime for the NBSR Reactor

Energy Technology Data Exchange (ETDEWEB)

Hanson, A.L.; Diamond, D.

2012-06-24

In preparation for the proposed conversion of the National Institute of Standards and Technology (NIST) research reactor (NBSR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel, certain point kinetics parameters must be calculated. We report here values of the prompt neutron lifetime that have been calculated using three independent methods. All three sets of calculations demonstrate that the prompt neutron lifetime is shorter for the LEU fuel when compared to the HEU fuel and longer for the equilibrium end-of-cycle (EOC) condition when compared to the equilibrium startup (SU) condition for both the HEU and LEU fuels.

Development and application of a system analysis code for liquid fueled molten salt reactors based on RELAP5 code

Energy Technology Data Exchange (ETDEWEB)

Shi, Chengbin [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Cheng, Maosong, E-mail: mscheng@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Liu, Guimin [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China)

2016-08-15

Highlights: • New point kinetics and thermo-hydraulics models as well as a numerical method are added into RELAP5 code to be suitable for liquid fueled molten salt reactor. • The extended REALP5 code is verified by the experimental benchmarks of MSRE. • The different transient scenarios of the MSBR are simulated to evaluate performance during the transients. - Abstract: The molten salt reactor (MSR) is one of the six advanced reactor concepts declared by the Generation IV International Forum (GIF), which can be characterized by attractive attributes as inherent safety, economical efficiency, natural resource protection, sustainable development and nuclear non-proliferation. It is important to make system safety analysis for nuclear power plant of MSR. In this paper, in order to developing a system analysis code suitable for liquid fueled molten salt reactors, the point kinetics and thermo-hydraulic models as well as the numerical method in thermal–hydraulic transient code Reactor Excursion and Leak Analysis Program (RELAP5) developed at the Idaho National Engineering Laboratory (INEL) for the U.S. Nuclear Regulatory Commission (NRC) are extended and verified by Molten Salt Reactor Experiment (MSRE) experimental benchmarks. And then, four transient scenarios including the load demand change, the primary flow transient, the secondary flow transient and the reactivity transient of the Molten Salt Breeder Reactor (MSBR) are modeled and simulated so as to evaluate the performance of the reactor during the anticipated transient events using the extended RELAP5 code. The results indicate the extended RELAP5 code is effective and well suited to the liquid fueled molten salt reactor, and the MSBR has strong inherent safety characteristics because of its large negative reactivity coefficient. In the future, the extended RELAP5 code will be used to perform transient safety analysis for a liquid fueled thorium molten salt reactor named TMSR-LF developed by the Center

Development and application of a system analysis code for liquid fueled molten salt reactors based on RELAP5 code

International Nuclear Information System (INIS)

Shi, Chengbin; Cheng, Maosong; Liu, Guimin

2016-01-01

Highlights: • New point kinetics and thermo-hydraulics models as well as a numerical method are added into RELAP5 code to be suitable for liquid fueled molten salt reactor. • The extended REALP5 code is verified by the experimental benchmarks of MSRE. • The different transient scenarios of the MSBR are simulated to evaluate performance during the transients. - Abstract: The molten salt reactor (MSR) is one of the six advanced reactor concepts declared by the Generation IV International Forum (GIF), which can be characterized by attractive attributes as inherent safety, economical efficiency, natural resource protection, sustainable development and nuclear non-proliferation. It is important to make system safety analysis for nuclear power plant of MSR. In this paper, in order to developing a system analysis code suitable for liquid fueled molten salt reactors, the point kinetics and thermo-hydraulic models as well as the numerical method in thermal–hydraulic transient code Reactor Excursion and Leak Analysis Program (RELAP5) developed at the Idaho National Engineering Laboratory (INEL) for the U.S. Nuclear Regulatory Commission (NRC) are extended and verified by Molten Salt Reactor Experiment (MSRE) experimental benchmarks. And then, four transient scenarios including the load demand change, the primary flow transient, the secondary flow transient and the reactivity transient of the Molten Salt Breeder Reactor (MSBR) are modeled and simulated so as to evaluate the performance of the reactor during the anticipated transient events using the extended RELAP5 code. The results indicate the extended RELAP5 code is effective and well suited to the liquid fueled molten salt reactor, and the MSBR has strong inherent safety characteristics because of its large negative reactivity coefficient. In the future, the extended RELAP5 code will be used to perform transient safety analysis for a liquid fueled thorium molten salt reactor named TMSR-LF developed by the Center

UABUC - Single energy point model burnup computer code for water reactors

International Nuclear Information System (INIS)

El-Meshad, Y.; Morsy, S.; El-Osery, I.A.

1981-01-01

UABUC is a single energy point reactor burnup computer program in FORTRAN language. The program calculates the change in the isotopic composition of the uranium fuel as a function of irradiation time with all its associated quantities such as the average point flux, the conversion ratio, macroscopic fuel cross sections, and the point reactivity profile. A step-wise time analytical solution was developed for the nonlinear first order burnup differential equations. The ''Westcott'' convention of the effective cross sections was used except for plutonium-240 and uranium-238. For plutonium-240, an effective microscopic cross section was derived from the direct physical arguments taking into account the selfshielding effect of plutonium-240 as well as the 1 ev. resonance absorption. For uranium-238, an effective cross section, reflecting the effect of fast fission and resonance absorption was used. The fission products were treated in the three groups with 50, 300, and 800 barns. The yields in the groups were treated as functions of the type of fissionable nuclides, the effective neutron temperature, and the epithermal index. Xenon-135 and Samarium-149 were treated separately as functions of irradiation time. (author)

Perturbation measurements in reactor LR-0 and their evaluation

International Nuclear Information System (INIS)

Rypar, W.; Faehrmann, K.H.

1988-07-01

To investigate space-dependent kinetic effects in reactors of the WWER-1000 type, two central and one eccentric perturbation measurements were performed in the zero power reactor LR-0 of the UJV Rez (CSSR) by trapeze-form movements of an absorber cluster. The measurements were based ona computer aided CAMAC system for the simultaneous data acquisition of 20 spatially distributed neutron detectors and for cluster movement control. The measurements were followed by a detailed evaluation in the ZfK Rossendorf (GDR) with respect to the calculation results of flux response obtained by nodal code HEXDYN3D, the aim of which was to demostrate the limits of the point reactor model and to account for space-dependent effects by approximative methods. A sensitive check of the calculation methods was made possible especially by the eccentric perturbation where the space dependent effects,due to a larger distance of cluster movement, were most significant. (author). 17 figs., 9 refs

Advanced methodology to simulate boiling water reactor transient using coupled thermal-hydraulic/neutron-kinetic codes

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

Controlled biomass formation and kinetics of toluene degradation in a bioscrubber and in a reactor with a periodically moved trickle-bed.

Science.gov (United States)

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.

A Point Kinetics Model for Estimating Neutron Multiplication of Bare Uranium Metal in Tagged Neutron Measurements

International Nuclear Information System (INIS)

Tweardy, Matthew C.; McConchie, Seth; Hayward, Jason P.

2017-01-01

An extension of the point kinetics model is developed in this paper to describe the neutron multiplicity response of a bare uranium object under interrogation by an associated particle imaging deuterium-tritium (D-T) measurement system. This extended model is used to estimate the total neutron multiplication of the uranium. Both MCNPX-PoliMi simulations and data from active interrogation measurements of highly enriched and depleted uranium geometries are used to evaluate the potential of this method and to identify the sources of systematic error. The detection efficiency correction for measured coincidence response is identified as a large source of systematic error. If the detection process is not considered, results suggest that the method can estimate total multiplication to within 13% of the simulated value. Values for multiplicity constants in the point kinetics equations are sensitive to enrichment due to (n, xn) interactions by D-T neutrons and can introduce another significant source of systematic bias. This can theoretically be corrected if isotopic composition is known a priori. Finally, the spatial dependence of multiplication is also suspected of introducing further systematic bias for high multiplication uranium objects.

Removal of oxytetracycline (OTC) in a synthetic pharmaceutical wastewater by a sequential anaerobic multichamber bed reactor (AMCBR)/completely stirred tank reactor (CSTR) system: biodegradation and inhibition kinetics.

Science.gov (United States)

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.

Fluoride Salt-Cooled High-Temperature Demonstration Reactor Point Design

Energy Technology Data Exchange (ETDEWEB)

Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carbajo, Juan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrell, Jerry W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wysocki, Aaron J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-02-01

The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would use tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 LiF-BeF₂) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include TRISO particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Several preconceptual and conceptual design efforts that have been conducted on FHR concepts bear a significant influence on the FHR DR design. Specific designs include the Oak Ridge National Laboratory (ORNL) advanced high-temperature reactor (AHTR) with 3400/1500 MWt/megawatts of electric output (MWe), as well as a 125 MWt small modular AHTR (SmAHTR) from ORNL. Other important examples are the Mk1 pebble bed FHR (PB-FHR) concept from the University of California, Berkeley (UCB), and an FHR test reactor design developed at the Massachusetts Institute of Technology (MIT). The MIT FHR test reactor is based on a prismatic fuel platform and is directly relevant to the present FHR DR design effort. These FHR concepts are based on reasonable assumptions for credible commercial prototypes. The FHR DR concept also directly benefits from the operating experience of the Molten Salt Reactor Experiment (MSRE), as well as the detailed design efforts for a large molten salt reactor concept and its breeder variant, the Molten Salt Breeder Reactor. The FHR DR technology is most representative of the 3400 MWt AHTR

Non-Linear Response to Periodic Forcing of Methane-Air Global and Detailed Kinetics in Continuous Stirred Tank Reactors Close to Extinction Conditions

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.

Study of Real-Time Programming for Simulation of Nuclear Reactor Dynamics

International Nuclear Information System (INIS)

Aliq; Widi Setiawan; Hendro Tjahjono

2003-01-01

Many aspects of real-time system are reviewed including the method, programming techniques, and its possibility to be applied in research reactor. The main point of real-time system is that it must designed to have a characteristics not only fast response but the most important is on-time response. In order to cover this requirements, real-time system need also a simple operating system consist of a kernel and application software. At the level of programming, real-time system require a modular approach, hard and soft time division and interprocess communications. The implementation can include some real-time (RT) operation system such as: RT-Linux, RT-OS9 and RT-Mat lab. Because of fast and on-time response requirements, if this system is going to be applied to research reactor, the transfer function model maybe more appropriate model compared to point kinetics model for the reason of computation time. (author)

Spectral shift reactor control method

International Nuclear Information System (INIS)

Impink, A.J. Jr.

1981-01-01

A method of operating a nuclear reactor having a core and coolant displacer elements arranged in the core wherein is established a reator coolant temperature set point at which it is desired to operate said reactor and first reactor coolant temperature band limits are provided within which said set point is located and it is desired to operate said reactor charactrized in that said reactor coolant displacer elements are moved relative to the reactor core for adjusting the volume of reactor coolant in said core as said reactor coolant temperature approaches said first band limits thereby to maintain said reactor coolant temperature near said set point and within said first band limits

Detection of gaseous heavy water leakage points in CANDU 6 pressurized heavy water reactors

International Nuclear Information System (INIS)

Park, T-K.; Jung, S-H.

1996-01-01

During reactor operation, the heavy water filled primary coolant system in a CANDU 6 Pressurized Heavy Water (PHWR) may leak through routine operations of the plant via components, mechanical joints, and during inadvertent operations etc. Early detection of leak points is therefore important to maintain plant safety and economy. There are many independent systems to monitor and recover heavy water leakage in a CANDU 6 PHWR. Methodology for early detection based on operating experience from these systems, is investigated in this paper. In addition, the four symptoms of D 2 O leakage, the associated process for clarifying and verifying the leakage, and the probable points of leakage are discussed. (author)

Modelling of alcohol fermentation in a tubular reactor with high biomass recycle

Energy Technology Data Exchange (ETDEWEB)

Narodoslawsky, M; Mittmannsgruber, H; Nagl, W; Moser, A

1988-05-30

Fermentation in tubular recycle reactors with high biomass concentrations is a way to boost productivity in alcohol production. A computer model has been developed to investigate the potential as well as to establish the limits of this process from a chemical engineering point of view. The model takes into account the kinetics of the reaction, the nonideality of flow and the segregation in the bioreactor. In accordance with literature, it is shown that tubular reactors with biomass recycle can improve productivity of alcohol fermentation substantially. With the help of the computer based reactor model it was also possible to estimate the detrimental effects of cell damage due to pumping. These effects are shown to play a major role, if the biomass separation is performed by filtration units which need high flow rates, e.g. tangential flow filters.

A small modular fast reactor as starting point for industrial deployment of fast reactors

International Nuclear Information System (INIS)

Chang, Yoon I.; Lo Pinto, Pierre; Konomura, Mamoru

2006-01-01

The current commercial reactors based on light water technology provide 17% of the electricity worldwide owing to their reliability, safety and competitive economics. In the near term, next generation reactors are expected to be evolutionary type, taking benefits of extensive LWR experience feedbacks and further improved economics and safety provisions. For the long term, however, sustainable energy production will be required due to continuous increase of the human activities, environmental concerns such as greenhouse effect and the need of alternatives to fossil fuels as long term energy resources. Therefore, future generation commercial reactors should meet some criteria of sustainability that the current generation cannot fully satisfy. In addition to the current objectives of economics and safety, waste management, resource extension and public acceptance become other major objectives among the sustainability criteria. From this perspective, two questions can be raised: what reactor type can meet the sustainability criteria, and how to proceed to an effective deployment in harmony with the high reliability and availability of the current nuclear reactor fleet. There seems to be an international consensus that the fast spectrum reactor, notably the sodium-cooled system is most promising to meet all of the long term sustainability criteria. As for the latter, we propose a small modular fast reactor project could become a base to prepare the industrial infrastructure. The paper has the following contents: - Introduction; - SMFR project; - Core design; - Supercritical CO 2 Brayton cycle; - Near-term reference plant; - Advanced designs; - Conclusions. To summarize, the sodium-cooled fast reactor is currently recognized as the technology of choice for the long term nuclear energy expansion, but some research and development are required to optimize and validate advanced design solutions. A small modular fast reactor can satisfy some existing near-term market niche

Photocatalytic reactors for treating water pollution with solar illumination. I: a simplified analysis for batch reactors

Energy Technology Data Exchange (ETDEWEB)

Sagawe, G.; Bahnemann, D. [Inst. fuer Technische Chemie, Univ. Hannover, Hannover (Germany); Brandi, R.J.; Cassano, A.E. [INTEC (Univ. Nacional del Litoral and CONICET), Santa Fe (Argentina)

2003-07-01

Usual applications of photocatalytic reactors for treating wastewater exhibit the difficulty of handling fluids having varying composition and/or concentrations; thus, a detailed kinetic representation may not be possible. When the catalyst activation is obtained employing solar illumination an additional complexity always coexists: solar fluxes are permanently changing with time. For comparing different reacting systems under similar operating conditions and to provide approximate estimations for scaling up purposes, simplified models may be useful. For these approximations the model parameters should be restricted as much as possible to initial physical and boundary conditions such as: initial concentrations (expressed as such or as TOC measurements), flow rate or reactor volume, irradiated reactor area, incident radiation fluxes and a fairly simple experimental observation such as the photonic efficiency. A combination of a new concept: the ''actual observed photonic efficiency'' with ideal reactor models and empirical kinetic rate expressions can be used to provide rather simple working equations that can be efficiently used to describe the performance of practical reactors. In this paper, the method has been developed for the case of a photocatalytic batch reactor (PBR). (orig.)

Neutron flux shape effects in large fast reactor safety calculations

International Nuclear Information System (INIS)

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

1978-01-01

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

Determination of the protection set-points lines for the Angra-1 reactor core

International Nuclear Information System (INIS)

Furieri, E.B.

1980-03-01

In this work several thermo-hidraulic calculation were performed to obtain Protection set-points lines for the Angra-1 reactor core in order to compare with the values presented by the vendor in the FSAR. These lines are the locus of points where DNBR min = 1,3 and power = 1,18 x P nominal as a function of ΔT m and T m , the temperature difference and the average coolant temperature between hot and cold legs. A computation scheme was developed using COBRA-IIIF as a subroutine of a new main program and adding new subroutines in order to obtain the desired DNBR. The solution is obtained through a convergentce procedure using parameters estimated in a sensivity study. (author) [pt

Breeder reactors

International Nuclear Information System (INIS)

Gollion, H.

1977-01-01

The reasons for the development of fast reactors are briefly reviewed (a propitious neutron balance oriented towards a maximum uranium burnup) and its special requirements (cooling, fissile material density and reprocessing) discussed. The three stages in the French program of fast reactor development are outlined with Rapsodie at Cadarache, Phenix at Marcoule, and Super Phenix at Creys-Malville. The more specific features of the program of research and development are emphasized: kinetics and the core, the fuel and the components [fr

Kinetics of ethylcyclohexane pyrolysis and oxidation: An experimental and detailed kinetic modeling study

KAUST Repository

Wang, Zhandong

2015-07-01

Ethylcyclohexane (ECH) is a model compound for cycloalkanes with long alkyl side-chains. A preliminary investigation on ECH (Wang et al., Proc. Combust. Inst., 35, 2015, 367-375) revealed that an accurate ECH kinetic model with detailed fuel consumption mechanism and aromatic growth pathways, as well as additional ECH pyrolysis and oxidation data with detailed species concentration covering a wide pressure and temperature range are required to understand the ECH combustion kinetics. In this work, the flow reactor pyrolysis of ECH at various pressures (30, 150 and 760Torr) was studied using synchrotron vacuum ultraviolet (VUV) photoionization mass spectrometry (PIMS) and gas chromatography (GC). The mole fraction profiles of numerous major and minor species were evaluated, and good agreement was observed between the PIMS and GC data sets. Furthermore, a fuel-rich burner-stabilized laminar premixed ECH/O2/Ar flame at 30Torr was studied using synchrotron VUV PIMS. A detailed kinetic model for ECH high temperature pyrolysis and oxidation was developed and validated against the pyrolysis and flame data performed in this work. Further validation of the kinetic model is presented against literature data including species concentrations in jet-stirred reactor oxidation, ignition delay times in a shock tube, and laminar flame speeds at various pressures and equivalence ratios. The model well predicts the consumption of ECH, the growth of aromatics, and the global combustion properties. Reaction flux and sensitivity analysis were utilized to elucidate chemical kinetic features of ECH combustion under various reaction conditions. © 2015 The Combustion Institute.

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)

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)

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.

Chemical kinetics in H2O and D2O under hydrothermal conditions

International Nuclear Information System (INIS)

Ghandi, K.; Alcorn, C.D.; Legate, G.; Percival, P.W.; Brodovitch, J.-C.

2010-01-01

Muonium (Mu = μ + e - ) is a light analogue of the H-atom. Studies of Mu chemical kinetics have been extended to supercritical water, a medium in some designs of future generation nuclear reactors. The Supercritical-Water-Cooled Reactor (SCWR) would operate at higher temperatures than current pressurized water-cooled reactors, and the lack of knowledge of water radiolysis under supercritical conditions constitutes a technology gap for SCWR development. Accurate modeling of chemistry in a SCWR requires data on kinetics of reactions involved in the radiolysis of water. In this paper, we first review our measurements of kinetics in H 2 O and then describe new data for D 2 O under sub- and supercritical conditions. (author)

Source term determination from subcritical multiplication measurements at Koral-1 reactor

International Nuclear Information System (INIS)

Blazquez, J.B.; Barrado, J.M.

1978-01-01

By using an AmBe neutron source two independent procedures have been settled for the zero-power experimental fast-reactor Coral-1 in order to measure the source term which appears in the point kinetical equations. In the first one, the source term is measured when the reactor is just critical with source by taking advantage of the wide range of the linear approach to critical for Coral-1. In the second one, the measurement is made in subcritical state by making use of the previous calibrated control rods. Several applications are also included such as the measurement of the detector dead time, the determinations of the reactivity of small samples and the shape of the neutron importance of the source. (author)

TRIGASIM: A computer program to simulate a TRIGA Mark I Reactor

International Nuclear Information System (INIS)

Ruby, Lawrence

1992-01-01

A Fortran-77 computer program has been written which simulates the operation of a TRIGA Mark I Reactor. The 'operator' has options at 1-second intervals, of raising rods, lowering rods, maintaining rods steady, dropping a rod, or scramming the reactor. Results are printed to the screen, and to 2 output files - a tabular record and a logarithmic plot of the power. The Point Kinetic Equations are programmed with 6 delayed groups, quasi-static power feedback, and forward differencing. A pulsing option is available, with simulation which employs the Fuchs Model. A pulse-tail model has been devised to simulate behavior for a few minutes following a pulse. Both graphic and tabular output are also available for the pulses. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-12-15

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

Determination of Model Kinetics for Forced Unsteady State Operation of Catalytic CH4 Oxidation

Directory of Open Access Journals (Sweden)

Effendy Mohammad

2016-01-01

Full Text Available The catalytic oxidation of methane for abating the emission vented from coal mine or natural gas transportation has been known as most reliable method. A reverse flow reactor operation has been widely used to oxidize this methane emission due to its capability for autothermal operation and heat production. The design of the reverse flow reactor requires a proper kinetic rate expression, which should be developed based on the operating condition. The kinetic rate obtained in the steady state condition cannot be applied for designing the reactor operated under unsteady state condition. Therefore, new approach to develop the dynamic kinetic rate expression becomes indispensable, particularly for periodic operation such as reverse flow reactor. This paper presents a novel method to develop the kinetic rate expression applied for unsteady state operation. The model reaction of the catalytic methane oxidation over Pt/-Al2O3 catalyst was used with kinetic parameter determined from laboratory experiments. The reactor used was a fixed bed, once-through operation, with a composition modulation in the feed gas. The switching time was set at 3 min by varying the feed concentration, feed flow rate, and reaction temperature. The concentrations of methane in the feed and product were measured and analysed using gas chromatography. The steady state condition for obtaining the kinetic rate expression was taken as a base case and as a way to judge its appropriateness to be applied for dynamic system. A Langmuir-Hinshelwood reaction rate model was developed. The time period during one cycle was divided into some segments, depending on the ratio of CH4/O2. The experimental result shows that there were kinetic regimes occur during one cycle: kinetic regime controlled by intrinsic surface reaction and kinetic regime controlled by external diffusion. The kinetic rate obtained in the steady state operation was not appropriate when applied for unsteady state operation

Electro-Fenton treatment of mature landfill leachate in a continuous flow reactor

Energy Technology Data Exchange (ETDEWEB)

Zhang, Hui, E-mail: eeng@whu.edu.cn [Department of Environmental Engineering, Hubei Biomass-Resource Chemistry and Environmental Biotechnology Key Laboratory, Wuhan University, P.O. Box C319, Luoyu Road 129, Wuhan 430079 (China); Ran, Xiaoni; Wu, Xiaogang [Department of Environmental Engineering, Hubei Biomass-Resource Chemistry and Environmental Biotechnology Key Laboratory, Wuhan University, P.O. Box C319, Luoyu Road 129, Wuhan 430079 (China)

2012-11-30

Highlights: Black-Right-Pointing-Pointer CSTR mode was used for COD removal of landfill leachate by Fered-Fenton process. Black-Right-Pointing-Pointer The complete mixing condition in the CSTR was verified. Black-Right-Pointing-Pointer A modified pseudo-first order kinetic model was developed. Black-Right-Pointing-Pointer The effects of important parameters on COD removal were investigated. Black-Right-Pointing-Pointer The organic components before and after treatment were determined by GC-MS. - Abstract: The treatment of mature landfill leachate by EF-Fere (also called Fered-Fenton) method was carried out in a continuous stirred tank reactor (CSTR) using Ti/RuO{sub 2}-IrO{sub 2}-SnO{sub 2}-TiO{sub 2} mesh anodes and Ti mesh cathodes. The effects of important parameters, including initial pH, inter-electrode gap, H{sub 2}O{sub 2} to Fe{sup 2+} molar ratio, H{sub 2}O{sub 2} dosage and hydraulic retention time, on COD removal were investigated. The results showed that the complete mixing condition was fulfilled in the electrochemical reactor employed in this study and COD removal followed a modified pseudo-first order kinetic model. The COD removal efficiency increased with the decrease of H{sub 2}O{sub 2} to Fe{sup 2+} molar ratio and hydraulic retention time. There existed an optimal inter-electrode gap or H{sub 2}O{sub 2} dosage so that the highest COD removal was achieved. Nearly the same COD removal was obtained at initial pH 3 and 5, but the steady state was quickly achieved at initial pH 3. The organic pollutants in the leachate were analyzed through a gas chromatography coupled with mass spectrometry (GC-MS) system. About 73 organics were detected in the leachate, and 52 of which were completely removed after EF-Fere process.

Physics of the pebble-bed high temperature reactor in massive water ingress accidents

International Nuclear Information System (INIS)

Scherer, W.

1989-10-01

A point-kinetics model was developed to describe qualitatively hypothetical water ingress transients in the primary loop of High Temperature Reactors. Neutron kinetics, heat-flow balance and the chemical reaction of graphite corrosion together with their mutual influence are included. The qualitative behaviour of the transients is calculated and discussed for two fictitious examples, namely the long-term water ingress into a medium sized HTR (HTR-500) and the 'startup' of a small HTR after an intensive water flooding of the core. The model developed and the computer code KINKOR are thought to be tools for the general understanding of the water ingress phenomena and should be looked at as basis for more elaborated systems. (orig./HP) [de

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

International Nuclear Information System (INIS)

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

1982-09-01

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

Spectral shift reactor control method

International Nuclear Information System (INIS)

Impink, A.J.

1982-01-01

A method of operating a nuclear reactor having a core and coolant displacer elements arranged in the core where there is established a reactor coolant temperature set point at which it is desired to operate the reactor and first reactor coolant temperature band limits within which the set point is characterized. The reactor coolant displacer elements are moved relative to the reactor core for adjusting the volume of reactor coolant in the core as the reactor coolant temperature approaches the first band limits to maintain the reactor coolant temperature near the set point and within the first band limits. The reactivity charges associated with movement of respective coolant displacer element clusters is calculated and compared with a calculated derived reactivity charge in order to select the cluster to be moved. (author)

Inverse kinetics for subcritical systems with external neutron source

International Nuclear Information System (INIS)

Carvalho Gonçalves, Wemerson de; Martinez, Aquilino Senra; Carvalho da Silva, Fernando

2017-01-01

Highlights: • It was developed formalism for reactivity calculation. • The importance function is related to the system subcriticality. • The importance function is also related with the value of the external source. • The equations were analyzed for seven different levels of sub criticality. • The results are physically consistent with others formalism discussed in the paper. - Abstract: Nuclear reactor reactivity is one of the most important properties since it is directly related to the reactor control during the power operation. This reactivity is influenced by the neutron behavior in the reactor core. The time-dependent neutrons behavior in response to any change in material composition is important for the reactor operation safety. Transient changes may occur during the reactor startup or shutdown and due to accidental disturbances of the reactor operation. Therefore, it is very important to predict the time-dependent neutron behavior population induced by changes in neutron multiplication. Reactivity determination in subcritical systems driven by an external neutron source can be obtained through the solution of the inverse kinetics equation for subcritical nuclear reactors. The main purpose of this paper is to find the solution of the inverse kinetics equation the main purpose of this paper is to device the inverse kinetics equations for subcritical systems based in a previous paper published by the authors (Gonçalves et al., 2015) and by (Gandini and Salvatores, 2002; Dulla et al., 2006). The solutions of those equations were also obtained. Formulations presented in this paper were tested for seven different values of k eff with external neutrons source constant in time and for a powers ratio varying exponentially over time.

Development of a digital reactivity meter for criticality prediction and control rod worth evaluation in pressurized water reactors

Energy Technology Data Exchange (ETDEWEB)

Kuramoto, Renato Y.R.; Miranda, Anselmo F.; Valladares, Gastao Lommez; Prado, Adelk C. [Eletrobras Termonuclear S.A. - ELETRONUCLEAR, Angra dos Reis, RJ (Brazil). Central Nuclear Almirante Alvaro Alberto], e-mail: kuramot@eletronuclear.gov.br

2009-07-01

In this work, we have proposed the development of a digital reactivity meter in order to monitor subcriticality continuously during criticality approach in a PWR. A subcritical reactivity meter can provide an easy prediction of the estimated critical point prior to reactor criticality, without complicated hand calculation. Moreover, in order to reduce the interval of the Physics Tests from the economical point of view, a subcritical reactivity meter can evaluate the control rod worth from direct subcriticality measurement. In other words, count rate of Source Range (SR) detector recorded during the criticality approach could be used for subcriticality evaluation or control rod worth evaluation. Basically, a digital reactivity meter is based on the inverse solution of the kinetic equations of a reactor with the external neutron source in one-point reactor model. There are some difficulties in the direct application of a digital reactivity meter to the subcriticality measurement. When the Inverse Kinetic method is applied to a sufficiently high power level or to a core without an external neutron source, the neutron source term may be neglected. When applied to a lower power level or in the sub-critical domain, however, the source effects must be taken in account. Furthermore, some treatments are needed in using the count rate of Source Range (SR) detector as input signal to the digital reactivity meter. To overcome these difficulties, we have proposed a digital reactivity meter combined with a methodology of the modified Neutron Source Multiplication (NSM) method with correction factors for subcriticality measurements in PWR. (author)