• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.149-154
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• 1998

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 from 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 of 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.

• Nuclear Engineering and Technology
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• v.43 no.6
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• pp.523-530
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• 2011

In a previous study, the stochastic space-dependent kinetics model (SSKM) based on the forward stochastic model in stochastic kinetics theory and the Ito stochastic differential equations was proposed for treating monoenergetic space-time nuclear reactor kinetics in one dimension. The SSKM was tested against analog Monte Carlo calculations, however, for exemplary cases of homogeneous slab reactors with only one delayed-neutron precursor group. In this paper, the SSKM is improved and evaluated with more realistic and complicated cases regarding several delayed-neutron precursor groups and heterogeneous slab reactors in which the extraneous source or reactivity can be introduced locally. Furthermore, the source level and the initial conditions will also be adjusted to investigate the trends in the variances of the neutron population and fission product levels across the reactor. The results indicate that the improved SSKM is in good agreement with the Monte Carlo method and show how the variances in population dynamics can be controlled.

• Journal of Korean Society of Water and Wastewater
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• v.25 no.6
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• pp.971-979
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• 2011

Palm Oil Mill Effluent (POME) is the mixed organic wastewater generated from palm oil industry. In this study, kinetic analysis with treating POME in an anaerobic hybrid reactor (AHR) was performed. Therefore, the AHR was monitored for its performances with respect to the changes of COD concentrations and hydraulic retention time (HRT). Batch tests were performed to find out the substrate removal kinetics by granular sludge from POME. Modified Stover Kincannon, First-order, Monod, Grau second-order kinetic models were used to analyze the performance of reactor. The results from the batch test indicate that the substrate removal kinetics of granular sludge is corresponds to follow Monod's theory. However, Grau second-order model were the most appropriate models for the continuous test in the AHR. The second order kinetic constant, saturation value constant, maximum substrate removal rate, and first-order kinetic constant were 2.60/day, 41.905 g/L-day, 39.683 g/L-day, and 1.25/day respectively. And the most appropriate model was Grau second-order kinetic model comparing the model prediction values and measured COD concentrations of effluent, whereas modified Stover-Kincannon model showed the lowest correlation.

• Nuclear Engineering and Technology
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• v.28 no.5
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• pp.452-457
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• 1996

A system, such as a reactor point kinetics equation, can be solved with Adomian Decomposition Method (ADM) which uses the notion that all solutions and operators can be expressed as an infinite sum of those basis states, like Adomian polynomials. In this work, ADM is applied to point reactor kinetics equations for step reactivity insertion, ramp input of reactivity, and nonlinear feedback cases without linearization approximation. The results of ADM are more accurate and faster than those of other existing methods, even though we use comparatively large time step sizes.

• Nuclear Engineering and Technology
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• v.11 no.3
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• pp.219-229
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• 1979

The current status of the computational methods and computer codes for the analysis of reactor kinetics is reviewed. Computational methods which have been developed for space-dependent transient analyses are presented and recent progress in the development of methods is discussed.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 1994.11a
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• pp.141-145
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• 1994

A computer code for solving the 3-D time-dependent multigroup neutron diffusion equation by a coupled reactor kinetics method recently developed has been developed and for evaluating its applicability in CANDU transient analysis applied to a 3-D kinetics benchmark problem which reveals non-uniform loss of coolant accident followed by an asymmetric insertion of shutdown devices. The performance of the method and code has been compared with the CANDU design code, CERBERUS, employing a finite difference improved quasistatic method.

• Applied Chemistry for Engineering
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• v.23 no.2
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• pp.204-209
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• 2012

In this study, kinetics data was obtained for steam reforming reaction of ethane over the commercial ruthenium catalyst. The variables of ethane steam reforming were the reaction temperature, partial pressure of ethane, and steam/ethane mole ratio. Parameters for the power rate law kinetic model and the Langmuir-Hinshelwood model were obtained from the kinetic data. Also, sizing of steam reforming reactor was performed by using PRO/II simulator. The reactor size calculated by the power rate law kinetic model was bigger than that of using the Langmuir-Hinshelwood model for the same conversion of ethane. Reactor size calculated by the Langmuir-Hinshelwood model seems to be more suitable for the reactor design because the Langmuir-Hinshelwood model was more consistent with the experimental results.

• Nuclear Engineering and Technology
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• v.27 no.4
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• pp.463-469
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• 1995

The point-kinetics reactor system which is subject to feedback effects may have multiple steady-state solutions for some operating conditions. A necessary and sufficient condition for multiple steady-state solutions of the point-kinetics reactor feedback system for an external input reactivity is obtained through their theoretical approach. If and only if the steady-state feedback reactivity of the reactor system is not strictly monotonic on some values of the feedback variables, then the reactor system has multiple steady-state solutions for the equilibrium operating conditions corresponding to the values of the feedback variables. Also, if and only if the steady--state feedback reactivity is strictly monotonic on all the feedback variables, then the reactor system has only one steady-state solution for all the operating conditions.

• Journal of Energy Engineering
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• v.17 no.1
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• pp.23-30
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• 2008

The neutron kinetic analysis methods for the molten-salt reactors are quite different from those for conventional solid-fuel reactors, which do not take into account the flowing-fuel-induced neutronics effects. Therefore, for dynamics and safety analyses of the molten-salt reactor systems, the conventional kinetics codes would not be appropriate to accurately predict its transient behaviors. A point-kinetics with flowing- fuel model has been used to assess the fluid-fuel reactor system safety, but recognized as not to be sufficient to simulate spatial distributions of delayed-neutron precursors and neutron populations during transients for given detail reactor models. In order to meet this requirement, AMBIKIND, a 2-group, 2-dimensional neutron kinetics code suitable for the molten-salt reactor systems was developed. This paper explains the code's theoretical and numerical descriptions and, as a part of its verification, includes some simulation results of MSRE stability experiments. Even though the present reactor model does not include the recirculation effect of the fuel-salt through the reactor system, the AMBIKIN2D code should be able to predict the power and phase shift at various power levels and reactivity insertions with better accuracy.

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1731-1735
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• 2021

The operation of the nuclear reactor requires accurate and fast methods and techniques for analysing its kinetics. These techniques become even more important when the MOX-fuel is used due to the lower value of delayed neutron fraction 𝛽 for ²³⁹Pu. Based on a Birth-and-Death process review, the mathematical model of thermal reactor core has been proposed different from existing ones. The analytical method for thermal point-reactor parameters evaluation is described within this work. The proposed method is applied for analysis of the unsteady transient processes taking place in a thermal reactor at its start-up or shutdown power change, as well as during small accidental power variation from the rated value. Theoretical determination of MASURCA reactor core reactivity through the analysis of experimental data on neutron time spectra was made.