Asymptotic analysis of a stochastic non-linear nuclear reactor model

The asymptotic behaviour of a stochastic non-linear nuclear reactor modelled by a master equation is analysed in two different limits: the thermodynamic limit and the zero-neutron-source limit. In the first limit a finite steady neutron density is obtained. The second limit predicts the neutron extinction. The interplay between these two limits is studied for different situations.     

Simulation of nuclear reactor dynamics equations using reconfigurable computing

This paper describes the application of a multilayer discrete-time cellular neural network (DT-CNN¹) and its hardware implementation on a field programmable gate array (FPGA²) to model and simulate the nuclear reactor dynamics equations. A new computing architecture model based on FPGA and its detailed hardware implementation are proposed for accelerating the solution of nuclear reactor dynamics equations. The proposed FPGA-based reconfigurable computing platform is implemented on a Xilinx FPGA device and is utilized to simulate step and ramp perturbation transients in typical 2D nuclear reactor cores. Properties of the implemented specialized architecture are examined in terms of speed and accuracy against the numerical solution of the nuclear reactor dynamics equations using MATLAB and C programs. Steady state as well as transient simulations, show a very good comparison between the two methods. Also, the validity of the synthesized architecture as a hardware accelerator is demonstrated.     

Generalized linear model for identifying the kinetic parameters of a nuclear reactor

Methods of real-time identification of the kinetic parameters of a reactor on the basis of an analysis of the linear couplings in transient neutron processes are examined. The algorithms described can be used in various combinations during reactor operation to identify typical transient processed according to their traces.     

An improved porous media model for nuclear reactor analysis

In this study, two modifications are proposed to mitigate drawbacks of the conventional approach of using the ‘‘Porous Media Model'(PMM) for nuclear reactor analysis. In the conventional approach, whole reactor core simplifies to a single porous medium and also, the resistance coefficients that are essential to using this model are constant values. These conditions impose significant errors and restrict the applications of the model for many cases,including accident analysis. In this article, the procedures for calculating the coefficients are modified by introducing a practical algorithm. Using this algorithm will result in obtaining each coefficient as a function of mass flow rate.Furthermore, the method of applying these coefficients to the reactor core is modified by dividing the core into several porous media instead of one. In this method, each porous medium comprises a single fuel assembly. PMM with these two modifications is termed ‘‘multi-region PMM' in this study. Then, the multi-region PMM is introduced to a new CFD-based thermo-hydraulic code that is specifically devised for combining with neutronic codes.The CITVAP code, which solves multi-group diffusion equations, is the selected as the neutronic part for this study. The resulting coupled code is used for simulation of natural circulation in a MTR. A new semi-analytic method,based on steady-state CFD analysis is developed to verify the results of this case. Results demonstrate considerable improvement, compared to the conventional approach.     

一种低温堆核电站的动态数学模型

针对研发的采用一体化布置、全功率自然循环的低温核反应堆电站,建立了一个可用于大功率运行范围控制系统仿真的动态数学模型.模型采用了六组缓发中子动态方程(考虑了慢化剂温度和燃料温度反应性负反馈)、集中参数的堆芯传热模型以及自然循环流动模型,重点考虑了主回路自然循环对堆芯内冷却剂和燃料棒之间的传热系数、主换热器换热系数、主回路时间常数的影响.仿真结果表明,模型能够正确反映低温堆核电站的主要动态特性,可用于电站控制系统仿真.     

Numerical simulation of a Hypothetical Core Disruptive Accident in a small-scale model of a nuclear reactor

In the case of a Hypothetical Core Disruptive Accident (HCDA) in a Liquid Metal Fast Breeder Reactor, it is assumed that the core of the nuclear reactor has melted partially and that the chemical interaction between molten fuel and liquid sodium has created a high-pressure gas bubble in the core. The violent expansion of this bubble loads and deforms the reactor vessel and the internal structures, thus endangering the safety of the nuclear plant.The MARA 10 experimental test simulates a HCDA in a 1/30-scale mock-up schematising a reactor block. In the mock-up, the liquid sodium cooling the reactor core is replaced by water and the argon blanket laying below the reactor roof is simulated by an air blanket. The explosion is triggered by an explosive charge.This paper presents a numerical simulation of the test with the EUROPLEXUS code and an analysis of the computed results. In particular, the evolution of the fluid flows and the deformations of the internal and external structures are analysed in detail. Finally, the current computed results are compared with the experimental ones and with previous numerical results computed with the SIRIUS and CASTEM-PLEXUS codes.     

Three-dimensional harmonic monitoring of a nuclear reactor

Translated from Atomnaya Énergiya, Vol. 67, No. 6, pp. 428–430, December, 1989.     

In-vessel corium catcher of a nuclear reactor

A new concept of an in-vessel corium melt catcher is proposed. The lower part of an elongated reactor vessel, which is filled with a sacrificial material of a proper composition, porosity, and arrangement, is used as such a catcher. The concept accounts of the scientific and design experience with the development of the ex-vessel corium catcher for the Tyan’van NPP with VVER-1000 reactors.     

Simulation of a gas-cooled fluidized bed nuclear reactor. Part I: Mixed oxide fuels

The worldwide use of enriched uranium has resulted over several decades in a stockpile of ²³⁸U. Fertile ²³⁸U can be converted by nuclear reaction into a transuranic mixture with a fissile content. Fuel which includes a limited proportion of this converted material is already used in some reactors. Use is restricted by the smaller delayed neutron yield and lower negative temperature coefficient of reactivity compared with uranium fuels.     

Dynamic model for control system design and simulation of a low temperature nuclear reactor

A lumped parameter dynamic model for the primary-loop and the U-tube steam generator of a low temperature power reactor is developed based on the fundamental conservation laws of fluid mass, energy and momentum. The dynamic model is formulated by coupling the point kinetics with reactivity feedback and the thermal-hydraulics of the reactor. The developed dynamic model is implemented on a personal computer using MATLAB/SIMULINK. Numerical simulation results for steady-state and transient responses are then presented, which show that the steady-state precision of the newly developed dynamic model is acceptable and the trend of the transient responses is correct. In addition, the “swell and shrink” behavior of the U-tube steam generator is also verified by numerical simulation. This newly established model can be utilized to control system design and simulation for the low temperature power reactor.     

Optimal control of a boiling water nuclear reactor

Application of optimal control to a boiling water nuclear reactor is the theme of this paper. The optimal control problem of a linearized model of a reactor is treated as a regulator problem and feedback control laws are derived to drive the system to steady state in the presence of disturbances. The weighting matrices in the performance index of the regulator problem are suitably changed to yield acceptable closed-loop responses for specific disturbances. The disturbances considered are (i) impulse change in temperature of water at inlet to plenum chamber and (ii) step change in throttle valve area. Then the feedback control laws are implemented on the nonlinear model to illustrate their effectiveness both for large and small disturbances.