Thesis. 1978. Ph.D.--Massachusetts Institute of Technology. Dept. of Nuclear Engineering. / MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE. / Includes bibliographical references. / by David C. Aldrich. / Ph.D.

Thesis. 1978. M.S.--Massachusetts Institute of Technology. Dept. of Mechanical Engineering. / MICROFICHE COPY AVAILABLE IN ARCHIVES AND ENGINEERING. / Includes bibliographical references. / M.S.

Após os acidentes nucleares ocorridos no mundo, critérios e requisitos extremamente rígidos para a operação das instalações nucleares foram determinados pelos órgãos internacionais que regulam essas instalações. A partir da ocorrência destes eventos, as operadoras de plantas nucleares necessitam simular alguns acidentes e transientes, por meio de programas computacionais específicos, para obter a licença de operação de uma planta nuclear. Com base neste cenário, algumas ferramentas computacionais sofisticadas têm sido utilizadas como o Reactor Excursion and Leak Analysis Program (RELAP5), que é o código mais utilizado para a análise de acidentes e transientes termo-hidráulicos em reatores nucleares no Brasil e no mundo. Uma das maiores dificuldades na simulação usando o código RELAP5 é a quantidade de informações geométricas da planta necessárias para a análise de acidentes e transientes termo-hidráulicos. Para a preparação de seus dados de entrada é necessário um grande número de operações matemáticas para calcular a geometria dos componentes. Assim, a fim de realizar estes cálculos e preparar dados de entrada para o RELAP5, um pré-processador matemático amigável foi desenvolvido, neste trabalho. O Visual Basic for Applications (VBA), combinado com o Microsoft Excel, foi utilizado e demonstrou ser um instrumento eficiente para executar uma série de tarefas no desenvolvimento desse pré-processador. A fim de atender as necessidades dos usuários do RELAP5, foi desenvolvido o Programa de Cálculo do RELAP5 PCRELAP5 onde foram codificados todos os componentes que constituem o código, neste caso, todos os cartões de entrada inclusive os opcionais de cada um deles foram programados. Adicionalmente, uma versão em inglês foi criada para PCRELAP5. Também um design amigável do PCRELAP5 foi desenvolvido com a finalidade de minimizar o tempo de preparação dos dados de entrada e diminuir os erros cometidos pelos usuários do código RELAP5. Nesse trabalho, a versão final desse pré-processador foi aplicada com sucesso para o Sistema de Injeção de Emergência (SIE) da usina Angra 2. / Nuclear accidents in the world led to the establishment of rigorous criteria and requirements for nuclear power plant operations by the international regulatory bodies. By using specific computer programs, simulations of various accidents and transients likely to occur at any nuclear power plant are required for certifying and licensing a nuclear power plant. Based on this scenario, some sophisticated computational tools have been used such as the Reactor Excursion and Leak Analysis Program (RELAP5), which is the most widely used code for the thermo-hydraulic analysis of accidents and transients in nuclear reactors in Brazil and worldwide. A major difficulty in the simulation by using RELAP5 code is the amount of information required for the simulation of thermal-hydraulic accidents or transients. The preparation of the input data requires a great number of mathematical operations to calculate the geometry of the components. Thus, for those calculations performance and preparation of RELAP5 input data, a friendly mathematical preprocessor was designed. The Visual Basic for Application (VBA) for Microsoft Excel demonstrated to be an effective tool to perform a number of tasks in the development of the program. In order to meet the needs of RELAP5 users, the RELAP5 Calculation Program (Programa de Cálculo do RELAP5 PCRELAP5) was designed. The components of the code were codified; all entry cards including the optional cards of each one have been programmed. In addition, an English version for PCRELAP5 was provided. Furthermore, a friendly design was developed in order to minimize the time of preparation of input data and errors committed by users. In this work, the final version of this preprocessor was successfully applied for Safety Injection System (SIS) of Angra 2.

Em instalações nucleares o controle do material nuclear é uma das atividades da maior importância. A Comissão Nacional de Energia Nuclear (CNEN) e a Agencia Internacional de Energia Atomica (AIEA) quando de suas inspeções rotineiras tem os dados fornecidos como um fator de segurança. Ter um sistema de controle de material nuclear que permita a qualquer momento reportar a quantidade e a localização dos diversos itens a serem inspecionados é um fator de primordial importância nos dias de hoje. Neste trabalho objetivou-se aprimorar um sistema já existente utilizando para seu desenvolvimento uma plataforma mais amigável através da linguagem de programação VisualBasic (Microsoft Corporation) para facilitar a equipe de operação do Reator IEA-R1 o fornecimento de dados que possibilitem o melhor controle dos materiais nucleares do Reator IEA-R1. Esses dados tem permitido o desenvolvimento de trabalhos a serem apresentados em congressos nacionais ou internacionais bem como em dissertações de mestrado ou teses de doutorado. O programa foi desenvolvido para atender as exigências das normas de salvaguarda da CNEN e da AIEA, mas suas funções podem ser ampliadas conforme as necessidades futuras. Este sistema poderá ser utilizado em outros reatores que por ventura sejam contruidos no pais, pois é bem pratico e sua utilização permite um um controle efetivo sobre o material nuclear da instalação. / In nuclear facilities, the nuclear material control is one of the most important activities. The National Commission of Nuclear Energy (CNEN) and the International Atomic Energy Agency (IAEA), when inspecting routinely, regards the data provided as a major safety factor. Having a control system of nuclear material that allows the amount and location of the various items to be inspected, at any time, is a key factor today. The objective of this work was to enhance the existing system using a more friendly platform of development, through the VisualBasic programming language (Microsoft Corporation), to facilitate the operation team of the reactor IEA-R1 Reactor tasks, providing data that enable a better and prompter control of the IEAR1 nuclear material. These data have allowed the development of papers presented at national and international conferences and the development of master´s dissertations and doctorate theses. The software object of this study was designed to meet the requirements of the CNEN and the IAEA safeguard rules, but its functions may be expanded in accordance with future needs. The program developed can be used in other reactors to be built in the country, since it is very practical and allows an effective control of the nuclear material in the facilities.

Este trabalho tem como objetivo a avaliação dosimétrica de componentes semicondutores (detectores Barreira de Superfície e fotodiodos PIN) para aplicação em medições de dose equivalente em campos de baixo fluxo de nêutrons (rápidos e térmicos), utilizando uma fonte de AmBe de alto fluxo, a instalação de Neutrongrafia do reator IEA-R1 (fluxos térmicos/epitérmicos) e fluxo de nêutrons rápidos do núcleo do reator IPEN/MB-01 (UCRI Unidade Crítica). Para a detecção de nêutrons (térmicos, epitérmicos e rápidos) foram usados componentes moderadores e conversores (parafina, boro e polietileno). Os fluxos resultantes da moderação e conversão foram utilizados para a irradiação de componentes semicondutores (SSB - Barreira de Superfície e fotodiodos). Foi utilizado também um conversor misto constituído de uma folha de polietileno borado (marca Kodak). O método de simulação por Monte Carlo foi utilizado para avaliar de forma analítica a espessura ótima da parafina. O resultado obtido foi similar ao verificado experimentalmente e serviu para avaliar o fluxo de nêutrons emergentes do moderador (parafina). Da mesma forma, através de simulação, foi avaliado também o fluxo de nêutrons rápidos que atinge o conversor de polietileno que cobre a face sensível dos semicondutores. O nível de radiação gama foi avaliado cobrindo o detector por inteiro com uma folha de cádmio de 1 mm de espessura. O reator IPEN/MB-01 foi usado para avaliar a resposta dos detectores para nêutrons rápidos de alto fluxo. Os resultados, de uma forma geral, mostraram concordância e similaridade com os trabalhos desenvolvidos por outros grupos de pesquisas. Foi também estabelecida uma abordagem para o cálculo de dose equivalente utilizando os espectros obtidos nas experiências. / The main objetive of this research is the dosimetric evaluation of semiconductor componentes (surface barrier detectors and PIN photodiodes) for applications in dose equivalent measurements on low dose fields (fast and thermal fluxes) using an AmBe neutron source, the IEA-R1 reactor neutrongraphy facility (epithermal and thermal fluxes) and the Critical Unit facility IPEN/MB-01 (fast fluxes). As moderator compound to fast neutrons flux from the AmBe source was used paraffin and boron and polyethylene as converter for thermal and fast neutrons measurements. The resulting fluxes were used to the irradiation of semiconductor components (SSB Surface Barrier Detector and PIN photodiodes). A mixed converter made of a borated polyethylene foil (Kodak) was also used. Monte Carlo simulation metodology was employed to evaluate analytically the optimal paraffin thickness. The obtained results were similar to the experimental data and allowed the evaluation of emerging neutron flux from moderator, as well as the fast neutron flux reaching the polyethylene covering the semiconductor sensitive surface. Gamma radiation levels were evaluated covering the whole detector with cadmium foil 1 mm thick, allowing thermal neutrons blockage and gamma radiation measurements. The IPEN/MB-01 facility was employed to evaluate the detector response for high neutron flux. The results were in good agreement with other studies published. Using the obtained spectra an approach to dose equivalent calculation was established.

Typescript. / Digitized by Kansas Correctional Industries

Calculation of the neutron flux in a nuclear reactor core is ideally performed by solving the neutron transport equation for a detailed-geometry model using several tens of energy groups. However, performing such detailed calculations for an entire core is prohibitively expensive from a computational perspective. Full-core neutronic calculations for CANDU reactors are therefore performed customarily using two-energy-group diffusion theory (no angular dependence) for a node-homogenized reactor model. The work presented here is concerned with reducing the loss in accuracy entailed when going from Transport to Diffusion. To this end a new method of calculating the diffusion coefficient was developed, based on equating the neutron balance equation expressed by the transport equation with the neutron balance equation expressed by the diffusion equation. The technique is tested on a simple twelve-node model and is shown to produce transport-like accuracy without the associated computational effort. / UOIT

In this study, the impact of various flow models is assessed under free and forced convection: compressible versus incompressible models for a Pressurized Water Reactor, and Darcy's law vs full momentum equation for High Temperature Gas Reactor. Euler equations with friction forces and a momentum and energy source/sink are used. The geometric model consists of a one-dimensional rectangular loop system. The fluid is heated up and cooled down along the vertical legs. A pressurizer and a pump are included along the horizontal legs. The compressible model is assumed to be the most accurate model in this study. Simulations show that under forced convection compressible and incompressible models yield the same transient and steady-state. As free convection is studied, compressible and incompressible models have different transient but the same final steady-state. As Darcy's law is used, pressure and velocity steady-state profiles yield some differences compared to the compressible model both under free and forced convections. It is also noted some differences in the transient.

Knowledge of the consequences of a nuclear criticality accident in aqueous fissile solutions is necessary to design the processing equipment for such solutions. The data at the disposal of designers before 1967 was provided by actual critically accidents. In 1968, the Service d'Etudes de Criticite of the French Commissariat a L'Energie Atomique initiated a program of systematic experimental aqueous solution nuclear excursions which were initiated intentionally to obtain solution criticality accident data. This program was designated "Consequence Radiologiques d'un Accident de Criticite" (CRAC). Although not intended to study the evolution of a solution nuclear criticality accident, the Kinetic Experiment on Water Boiler (KEWB) demonstrated the dependence of the nuclear excursion on parameters such as solution temperature and radiolytic gas. Similarly, the CRAC program results indicated the excursion was governed by parameters such as the solution addition rate, initial neutron population, solute concentration, and thermal and radiolytic gas feedback. The energy deposited in a fissile solution is the sum of the energies contributed by the radiation sources. The majority of the energy is deposited by the fission fragments. One feature of the energy deposition is a commensurate increase in the system temperatures which affects the solution volume and thereby the neutron leakage probability. A second feature is the decomposition of the water molecule which results in release of H(,2) and O(,2) in the solution. Microbubbles are nucleated in the fissile solution by a localized thermal spike generated by a fission fragment. Initially, the microbubble contains a mixture of radiolytic gas and water vapor. Below the boiling point the vapor condenses quickly, leaving a gas microbubble. Unless the solution is supersaturated, the gas bubble will dissolve in a few microseconds. However, in a supersaturated solution the bubble will grow and produce negative feedback by increasing neutron leakage. The analysis for this study employs two mathematical models for the radiolytic gas feedback. One assumes the radiolytic gas concentration is a linear function of the energy release and the nucleation rate is a linear function of the power (Energy model). The other assumes a correlation between the system pressure and the radiolytic gas feedback (Pressure model). Both models have been incorporated into a space-independent kinetic computer code, MACKIN, while the pressure model was also incorporated into a space-dependent code, AZPAD, (Space-dependent model). The model incorporation provides a numerical tool with which to analyze a nuclear excursion in an aqueous fissile solution. The models have been successful in predicting the peak power, burst energy, and maximum system pressure for the first burst in both KEWB and CRAC experiments.

A steady-state model for the calculation of temperature and pressure distributions, and heat and work balance for the STAR-LM and the STAR-H2 systems was developed. The STAR-LM system is designed for electricity production and consists of the lead cooled reactor on natural circulation and the supercritical carbon dioxide Brayton cycle. The STAR-H2 system uses the same reactor which is coupled to the hydrogen production plant, the Brayton cycle, and the water desalination plant. The Brayton cycle produces electricity for the on-site needs. Realistic modules for each system component were developed. The model also performs design calculations for the turbine and compressors for the CO2 Brayton cycle. The model was used to optimize the performance of the entire system as well as every system component. The size of each component was calculated. For the 400 MWt reactor power the STAR-LM produces 174.4 MWe (44% efficiency) and the STAR-H2 system produces 7450 kg H2/hr. The steady state model was used to conduct quasi-static passive load follow analysis. The control strategy was developed for each system; no control action on the reactor is required. As a main safety criterion, the peak cladding temperature is used. It was demonstrated that this temperature remains below the safety limit during both normal operation and load follow.