Severe Accident Simulation of Small Modular Reactors

Since the Three Mile Island Unit 2 (TMI-2), accident in 1979 which led to the meltdown of about one half of the reactor core and to limited releases of radioactive materials to the environment, an important international effort has been made on severe accident research. The present work aims to investigate the behaviour of a Small Modular Reactor during severe accident conditions. In order to perform these analyses, a SMR has been studied for the European reference severe accident analysis code ASTEC, developed by IRSN and GRS. In the thesis will be described in detail the IRIS Small Modular Reactor; the reference reactor chosen to develop the ASTEC input deck. The IRIS model was developed in the framework of a research collaboration with the IRSN development team. In the thesis will be described systematically the creation of the ASTEC IRIS input deck: the nodalization scheme adopted, the solution used to simulate the passive safety systems and the strong interaction between the reactor vessel and the containment. The ASTEC SMR model will be tested against the RELAP-GOTHIC coupled code model, with respect to a Design Basis Accident, to evaluate the capability of the ASTEC code on reproducing correctly the behaviour of the nuclear system. Once the model has been validated, a severe accident scenario will be simulated and the obtained results along with the nuclear system response will be analysed.

Traffic accident simulation. Final report.

Louisiana Transportation Research Center, Baton Rouge

An evaluation of the applicability of existing data to the analytical description of a nuclear-reactor accident : quarterly progress report for ... through ....

Includes bibliographical references.

PCRELAP5 - Programa de cálculo para os dados de entrada do código RELAP5

Dissertação (Mestrado em Tecnologia Nuclear)

The induction plasma chemical reactor: Part I. Equilibrium model

A mathematical model is presented for the numerical simulation of the flow, temperature, and concentration fields in an rf plasma chemical reactor. The simulation is performed assuming chemical equilibrium. The extent of validity of this assumption is discussed. The system considered is the reaction of SiCl4 and NH3 for the production of Si3N4.

Uso de ambientes virtuais para a simulação de acidentes radiológicos

Em um ambiente virtual, construído com o uso de tecnologia computacional, encontram-se presentes entidades virtuais inseridas em um espaço tridimensional, que é utilizado para a simulação de processos críticos, como os acidentes radiológicos. A pronta detecção de um acidente radiológico e a determinação da sua possível extensão são fatores essenciais para o planejamento de respostas imediatas e de ações de emergência. A integração das representações georeferenciadas do espaço tridimensional, com modelos baseados em agentes autônomos, com o objetivo de construir ambientes virtuais que tenham a capacidade de simular acidentes radiológicos é a proposta dessa tese. As representações georeferenciadas do espaço tridimensional candidatas são: i)as representações espaciais usadas nos sistemas de informações geográficas (SIG) e ii) a representação adotada pelo Google MapsTM. Com o uso deste ambiente pode-se: quantificar as doses recebidas pelas pessoas; ter uma distribuição espacial das pessoas contaminadas; estimar o número de indivíduos contaminados; estimar o impacto na rede de saúde; estimar impactos ambientais; gerar zonas de exclusão; construir cenários alternativos; treinar pessoal técnico para lidar com acidentes radiológicos.

Decay heat in fast reactors with transuranic fuels

In this work, we performed an evaluation of decay heat power of advanced, fast spectrum, lead and molten salt-cooled reactors, with flexible conversion ratio. The decay heat power was calculated using the BGCore computer code, which explicitly tracks over 1700 isotopes in the fuel throughout its burnup and subsequent decay. In the first stage, the capability of the BGCore code to accurately predict the decay heat power was verified by performing a benchmark calculation for a typical UO2 fuel in a Pressurized Water Reactor environment against the (ANSI/ANS-5.1-2005, "Decay Heat Power in Light Water Reactors," American National Standard) standard. Very good agreement (within 5%) between the two methods was obtained. Once BGCore calculation capabilities were verified, we calculated decay power for fast reactors with different coolants and conversion ratios, for which no standard procedure is currently available. Notable differences were observed for the decay power of the advanced reactor as compared with the conventional UO2 LWR. The importance of the observed differences was demonstrated by performing a simulation of a Station Blackout transient with the RELAP5 computer code for a lead-cooled fast reactor. The simulation was performed twice: using the code-default ANS-79 decay heat curve and using the curve calculated specifically for the studied core by BGCore code. The differences in the decay heat power resulted in failure to meet maximum cladding temperature limit criteria by ∼100 °C in the latter case, while in the transient simulation with the ANS-79 decay heat curve, all safety limits were satisfied. The results of this study show that the design of new reactor safety systems must be based on decay power curves specific to each individual case in order to assure the desired performance of these systems. © 2009 Elsevier B.V. All rights reserved.