Basic CFD investigation of decay heat removal in a pool type research reactor

Safety is one of the most important and desirable characteristic in a nuclear plant. Natural circulation cooling systems are noted for providing passive safety. These systems can be used as mechanism for removing the residual heat from the reactor, or even as the main cooling system for heated sections, such as the core. In this work, a computational fluid-dynamics (CFD) code is used to simulate the process of natural circulation in an open pool research reactor after its shutdown. The physical model studied is similar to the Open Pool Australian Light water reactor (OPAL), and contains the core, cooling pool, reflecting tank, circulation pipes and chimney. For best computing performance, the core region was modeled as a porous media, where the parameters were obtained from a separately detailed CFD analysis

Computational simulation of fuel burnup estimation for research reactors plate type

The aim of this study is to estimate the spatial fuel burnup, through computational simulation, in two research reactors plate type, loaded with dispersion fuel: the benchmark Material Test Research – International Atomic Energy Agency (MTR–IAEA) and a typical multipurpose reactor (MR). The first composed of plates with uranium oxide dispersed in aluminum (UAlx–Al) and a second composed with uranium silicide (U3Si2) dispersed in aluminum. To develop this work we used the deterministic code, WIMSD-5B, which performs the cell calculation solving the neutron transport equation, and the DF3DQ code, written in FORTRAN, which solves the three-dimensional neutron diffusion equation using the finite difference method. The methodology used was adequate to estimate the spatial fuel burnup , as the results was in accordance with chosen benchmark, given satisfactorily to the proposal presented in this work, even showing the possibility to be applied to other research reactors. For future work are suggested simulations with other WIMS libraries, other settings core and fuel types. Comparisons the WIMSD-5B results with programs often employed in fuel burnup calculations and also others commercial programs, are suggested too. Another proposal is to estimate the fuel burnup, taking into account the thermohydraulics parameters and the Xenon production

Simulação computacional de escoamentos viscosos compressíveis/quase incompressíveis

A new finite element formulation designed for both compressible and nearly incompressible viscous flows is presented. The formulation combines conservative and non-conservative dependent variables, namely, the mass-velocity (density*velocity), internal energy and pressure. The central feature of the method is the derivation of a discretised equation for pressure, where pressure contributions arising from the mass, momentum and energy balances are taken implicitly in the time discretisation. Numerical examples, covering a wide range of Mach number, demonstrate the robustness and versatility of the new method.É apresentada uma nova formulação, usando a técnica de elementos finitos, para escoamentos viscosos compressíveis e/ou quase-incompressíveis. A formulação combina as seguintes variáveis dependentes conservativas e não-conservativas: velocidade mássica (densidade*velocidade), energia interna e pressão. A principal característica do método é a derivação de uma equação discretizada para a pressão, onde as contribuições para a pressão provenientes dos balanços de massa, quantidade de movimento e energia são tomadas implicitamente na discretização temporal. Exemplos numéricos abrangendo uma larga faixa de número de Mach demonstram a robustez e versatilidade do novo método

Analise de um acidente hipotético de perda de vazão forçada em um reator tipo LMFBR

This report uses a model to analyse a postulated accident scenario involving loss of forced flow in the reactor vessel of a LMFBR. Five phases of the accident are analysed: Natural Circulation, Subcooled Boiling, Nucleate Boiling, Core Dryout and Cladding melt. The heat conduction in the fuel, cladding, coolant and lower and upper plenum are calculated by a lump-parameter model. Physical data of a prototype LMFBR reactor were used for the calculation.Este trabalho utiliza um modelo de calculo para se analisar um acidente hipotético de perda total de vazão forçada no vaso de um reator. São analisadas cinco fases do acidente: Circulação Natural, Ebulição Subresfriada, Ebulição Nucleada, Secagem do Núcleo e Derretimento do Revestimento. A difusão radial de calor no combustível, revestimento, canal médio de refrigeração do núcleo e câmaras superior e inferior e representada por um modelo de parâmetros concentrados. Os cálculos foram efetuados utilizando-se dados de um protótipo de reator LMFBR

One-dimensional termohidraulic model of the nucleus and lower and upper pool regions of the reactor RMB

The computer codes used in construction of nuclear reactors projects, specifically with regard to the thermo-hydraulics concepts part of your core, whose main goal reproduce actual operating conditions in order to predict quantitatively the limiting conditions of operation so that the safety limit is not exceeded. Computational methods for studies of fluid flow are developed around the world, including Brazil. With the evolution of computers application of numerical methods greatly reduced response time results, and tends to further decrease the extent that the computers and processors develop, making it feasible to use programming accident simulations of heat transfer reactors. The software developed in this paper presents a method for analyzing the thermo-hydraulic behavior of the Brazilian Multipurpose Reactor (RMB) after its shutdown. The software solves the conservations equations applied to the core and also the lower and upper regions of the RMB. The thermo-hydraulics characteristics studied are: the temperatures of the core, cladding and refrigerant, the mass flow and the heat transfer. The numerical resolution was performed using the Matlab language and the outputs are presented in graphs and tables forms

An experimental low-pressure facility to study boron transients in the pressurizer of an a integral modular nuclear reactor

Small and medium size modular reactors offer many advantages when compared with typical nuclear plants in various circumstances, such as offering greater simplicity of design, economy of mass production, and reducing siting costs. The integral configuration is characterized by having most of its components inside the pressure vessel, eliminating the probability of accidents. However, for this configuration there is no spray system for boron homogenization, which may cause power transients. Thus, it is necessary to investigate boron mixing. The Federal University of Pernambuco (UFPE), in a partnership with the Regional Center of Nuclear Sciences of Northeast (CRCN-NE) and the Nuclear Engineering Institute (IEN/CNEN-RJ), is developing a project that aims to analyze transients in a compact modular integral reactor. This analysis will be made by using the data obtained from one experimental bench that is mounted at CRCN-NE. A study accomplished in 2012 using a simplified bench (built in reduced scale with a test section manufactured with transparent acrylic) showed that it was possible to obtain preliminary experimental results for the boron homogenizing process

Pressure drop calculation in a fuel element of a pool type reactor

Even with the advances of hardware in computer sciences, sometimes it is necessary to simplify the simulation in order to optimize the results given the same calculation runtime. The object of this study is a thermodynamic analysis of the core of a pool type research reactor, focusing on natural circulation. Due to the high geometrical complexity of the core, the scale transfer process becomes an essential step to the thermodynamic study of the reactor. This process takes place by determining the effective equivalent properties obtained from a detailed simulation of the core and transferring them to a porous medium having a coarse mesh while preserving the overall characteristics. In this way, it will be able to obtain the quadratic resistance coefficient KQ by calculating the pressure drop inside the fuel element. To observe in detail the behavior of this flow, longitudinal and transversal cross sections will be made in different points, thereby observing the velocity and pressure distributions. The analysis will provide detailed data on the fluid flow between the fuel plates enabling the observation of possible critical points or undesired behavior. The whole analysis was made by using the commercial code ANSYS CFX ver. 12.1. This is study will provide data, as a first step to enable future simulations which will consider the entire reactor

Methodology for studies of natural circulation in closed circuits

This work presents an analysis of stability of the phenomenon of natural circulation for onedimension single-phase flow in a closed loop. The computer program uses a stabilized finite element formulation for the solution of the Navier-Stokes and energy equations in cartesian coordinates. The formulation has been developed and tested in a computer code developed at the Nuclear Engineering Institute (IEN-CNEN) and is now available either for future analysis or design of nuclear system