Recently, attention has been drawn to the viability of using S-CO(2) as a working fluid in modern reactor designs. Near the critical point, CO2 has a rapid rise in density allowing a significant reduction in the compressor work of a closed Brayton Cycle. Therefore, 45% efficiency can be achieved at much more moderate temperatures than is optimal for the helium Brayton cycles. An additional benefit of the S-CO2 system is its universal applicability as an indirect secondary Power Conversion System (PCS) coupled to most GEN-IV concept reactors, as well as fusion reactors. The United States DOE's GNEP is now focusing on the liquid Na cooled primary as an alternative to conventional Rankine steam cycles. This primary would also benefit from being coupled to an S-CO2 PCS. Despite current progress on designing the S-CO2 PCS, little work has focused on the principal supporting systems required. Many of the required auxiliary systems are similar to those used in other nuclear or fossil-fired units; others have specialized requirements when CO2 is used as the working fluid, and are therefore given attention in this thesis. Auxiliary systems analyzed within this thesis are restricted to those specific to using CO2 as the working fluid. Particular systems discussed include Coolant Make-up and Storage, Coolant Purification, and Coolant Leak Detection. / Contract number: N62271-97-G-0026. / US Navy (USN) author

Durante um acidente em uma usina nuclear, a integridade do vaso pressurizado deve ser assegurada. Em resposta a um possível derretimento do combustível nuclear, a atual geração de usinas possui um sistema para a injeção de água potável na cavidade do vaso pressurizado com intuito de resfriar sua parede, prevenindo danos a sua estrutura e evitando o vazamento de material radioativo. Esse estudo considerou o uso de água marinha como refrigerante para inundar a cavidade do vaso pressurizado combinado com a fixação de um estrutura porosa em forma de grade em sua parede externa como meio de aprimorar a margem de segurança durante a contenção de acidentes. Experimentos de longa duração para a ebulição em piscina de água marinha artificial foram conduzidos em uma superfície circular de cobre plana com 30 mm de diâmetro. Foi encontrado um fluxo de calor crítico de 1; 6 MW/m2 sob pressão atmosférica. Esse valor é significantemente maior que aquele obtido (1; 0 MW/m2) nas mesmas condições experimentais. Foi verificado que os depósitos de sais marinhos podem aumentar a molhabilidade e a capilaridade da superfície de teste, aprimorando assim o fluxo crítico. Combinando a água marinha e a fixação da estrutura porosa sobre a superfície de teste, verificou-se um melhora no coeficiente de transmissão de calor e no fluxo de calor crítico de até 110 % (2; 1 MW/m2), quando comparado a água destilada na superfície limpa, sem a instalação da estrutura. Após os experimentos, foi identificado que muitos dos poros presentes nas superfícies da estrutura porosa encontravam-se bloqueados devido ao aglutinamento de sais marinhos. Isso levou a conclusão que o aumento no valor do fluxo crítico observado para a água marinha artificial ocorreu devido, principalmente, a separação das fases líquida e gasosa do fluido na região próxima a superfície de teste, efeito proporcionado pela forma de grade da estrutura porosa, e ao aumento da molhabilidade e capilaridade da superfície devido a formação dos depósitos marinhos. / During a severe nuclear power plant accident, the integrity of the reactor pressure vessel must be assured. In response to a possible fuel meltdown, operators of the current generation of nuclear power plants are likely to inject water into the reactor pressure vessel to cool down the reactor vessel wall, preserving its integrity and avoiding leakage of radioactive material. This study considers the use of seawater to flood a reactor pressure vessel combined with the attachment of a honeycomb porous plate (HPP) on the vessel outer wall as a way to improve the safety margins for in-vessel retention of fuel. In long-duration experiments, saturated pool boiling of artificial seawater was performed with an upward-facing plain copper heated surface 30 mm in diameter. The resulting value for critical heat flux (CHF) was 1; 6 MW/m2 at atmospheric pressure, a value significantly higher than the CHF obtained when the working fluid was distilled water (1; 0 MW/m2). It was verified that sea-salt deposits could greatly improve surface wettability and capillarity, enhancing the CHF. The combination of artificial seawater and an HPP attached to the heated surface improved the boiling heat transfer coefficient and increased the CHF up to 110% (2; 1 MW/m2) as compared to distilled water on a bare surface. After the artificial seawater experiments, most of the wall micropores of the HPP were clogged because of sea-salt aggregation on the HPP top and bottom surfaces. Thus, the CHF enhancement observed in this case was attributed mainly to the separation of liquid and vapor phases provided by the HPP channel structure and improvement of surface wettability and capillarity by sea-salt deposition.

Kwong Lop Sam. / Thesis (M.Phil.)--Chinese University of Hong Kong. / Bibliography: leaves 60-63.

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

In this project, we have assessed the structural tolerance of advanced refractory alloys to simulated nuclear fusion reactor environments, by using intense proton beams to mimic fusion neutron damage and analysing the proton damaged structures using in-situ/ex-situ transmission electron microscopy and nano-hardness measurements. Refractory metals such as tungsten or tantalum, and their binary alloy combinations, are considered as promising structural materials to withstand the unprecedented high heat loads and fast neutron/helium fluxes expected in future magnetically-confined fusion reactors. Tungsten is currently the frontrunner for the production of plasma-facing components for fusion reactors. The attractiveness of tungsten as structural material lies in its high resistance to plasma-induced sputtering, erosion and radiation-induced void swelling, together with its thermal conductivity and high-temperature strength. Unfortunately, the brittle nature of tungsten hampers the manufacture of reactor components and can also lead to catastrophic failure during reactor operations. We have focused on two potential routes to enhance the ductility of tungsten-containing materials, namely alloying tungsten with controlled amounts of tantalum, and using alternatively tantalum-based alloys containing specific tungsten additions, either as a full-thickness structural facing material or as a coating of first wall reactor components. The aim was to investigate the formation and evolution of radiation-induced damaged structures in these material solutions and the impact of those structures on the hardness of the material. The main results of this work are: (1) the addition of 5wt%Ta to W leads to saturation in the number density and average dimensions of the radiation-induced a/2 dislocation loops formed at 350C, whereas in W the loop length increases progressively and evolves into dislocation strings, and later into hydrogen bubbles and surface blisters, (2) the recovery behaviour of proton irradiated W5wt.%Ta alloy is characterized by dislocation loop growth at 600-900C, whereas voids form at 1000C by either vacancy absorption or loop collapse, (3) the presence of radiation-induced a loops at 590C in Ta hinders the formation and ordering of voids observed with increasing damage levels at 345C, (4) the addition of 5-10wt.%W to Ta delays the evolution of a/2 dislocation loops with increasing damage levels, and therefore the appearance of random voids. These results expand the composition palette available for the safe selection of refractory alloys for plasma facing components with enhanced, or at least predictable, tolerance to the heat-radiation flux combinations expected in future nuclear fusion plants.

Thesis. 1975. M.S.--Massachusetts Institute of Technology. Dept. of Civil Engineering. / Bibliography: leaf 106. / by Joseph Parker Morray, Jr. / M.S.

Typescript (photocopy). / Digitized by Kansas Correctional Industries

O estudo termo-hidráulico, utilizado para análise de acidentes e transientes em reatores nucleares, é feito com o uso de algumas ferramentas computacionais sofisticadas. Esses programas utilizam uma filosofia realista (best estimate) para análise de acidentes e transientes em reatores refrigerados à água leve do tipo PWR (Pressurized Water Reactor) e sistemas associados. O código RELAP5, objeto de nosso estudo, tem sido usado como uma ferramenta para o licenciamento de instalações nucleares no nosso país. Uma das maiores dificuldades na simulação de acidentes e transientes em uma instalação nuclear com o código RELAP5 é a quantidade de informações necessárias, que na maioria dos casos é muito grande. Além disso, existe a necessidade de uma quantidade razoável de operações matemáticas para os cálculos da geometria dos componentes. Portanto, a fim de facilitar a manipulação destas informações, percebeu-se a necessidade do desenvolvimento de um pré-processador amigável com o usuário, para realização desses cálculos e para elaboração dos dados de entrada do RELAP5. A ferramenta escolhida foi o MS-EXCEL, que apresentou grande potencialidade no desenvolvimento do pré-processador desejado. / Computational program are used for thermal hydraulic analysis of accidents and transients conditions in nuclear power plants. The RELAP5 code has been developed to simulate accidents and transients conditions, performing a best estimate analysis, in Pressurized Water Reactors (PWR) and auxiliary systems. The RELAP5 code, which has been used as a tool for licensing nuclear facilities in Brazil, is the objective of the study performed in this work. The main problem in using the RELAP5 code is the huge amount of information necessary to model the nuclear reactor and thus to simulate thermal-hydraulic accidents. Moreover, the RELAP5 code input data requires a large amount of mathematical operations to calculate the geometry of the plant components. Therefore, in order to make easier the data input for the RELAP5 code a friendly preprocessor has been developed. The preprocessor accepts basic information about the geometry of the plant components and performs all the calculations needed for the RELAP5 input. This preprocessor has been developed based on the MS-EXCEL software.

O presente trabalho visa contribuir com o desenvolvimento sistemático de novas metodologias experimentais da medida da reatividade de arranjos físseis subcríticos, utilizando: estatísticas de alta ordem das contagens de nêutrons com detectores no modo pulso, o recente conceito de reatividade generalizada, e as instalações do reator IPEN/MB-01. Este trabalho reuniu em um só texto diversos aspectos da implementação destes tipos de medidas. Diferentemente das demais técnicas utilizadas nas medidas da reatividade subcrítica, as metodologias apresentadas neste trabalho tem o potencial para permitir a medida experimental da reatividade subcrítica sem a necessidade da estimativa prévia de quaisquer outros parâmetros cinéticos, obtidos de forma teórica ou experimental, calibração de fontes externas ou detectores.A princípio, os métodos estatísticos de alta ordem das contagens de nêutrons permitem obter diretamente o valor da subcriticalidade (ou o fator de multiplicação) de um arranjo físsil, independentemente do modelo de física subcrítica utilizado, sem a utilização de infra-estrutura diferenciada (como uma fonte pulsada de nêutrons), sendo uma extensão natural das metodologias que utilizam estatísticas de ordens inferiores - por exemplo, Feymann-. E este conteúdo estatístico diferenciado dos momentos de altas ordens das contagens de nêutrons, o principal motivador da implementação deste trabalho. Apesar de suas potencialidades, a implementação experimental do método esbarra no tempo e taxa de aquisição de dados; ou seja, na quantidade de conteúdo estatístico necessária para a obtenção de medida útil. Exatamente esta dificuldade impediu a obtenção de uma medida útil/prática nas instalações do reator IPEN/MB-01. Existem, entretanto, outras formas de explorar estatísticas ordem superior. Por exemplo, uma extensão do método de Rossi- sugerida neste trabalho pode utilizar auto bi-correlações (coincidências triplas não acidentais de contagens). A despeito do alto valor das incertezas, os aspectos estatísticos fundamentais de uma medida foram preservados nos métodos empregados neste trabalho. O método das auto bicorrelações é conceitualmente mais robusto contra as influências do tempo morto do sistema de aquisição de dados. Ao longo de sua execução, o presente trabalho visou preencher algumas lacunas de procedimentos experimentais aparentemente pouco abordadas por outros autores, permitindo estabelecer métodos estatisticamente mais rigorosos. Entre as contribuições neste sentido destacam-se, entre outras, as correções por tempo morto ou as geradas pela correlação entre os parâmetros estatísticos em tela. Do ponto de vista teórico, este trabalho sugere duas maneiras originais de abordar o mesmo problema da utilização de estatísticas de altas ordens: (a) auto bicorrelações; e (2) os biespectros de densidade de espectral de potência própria, sendo o primeiro explorado experimentalmente/estatisticamente em detalhes. / The present work aims to contribute to the systematic development of new experimental methods of measuring the reactivity of any subcritical fissile arrangements using: high-order statistics of neutron counts from neutron detectors working in pulse mode, the recent concept general reactivity, and the IPEN/MB-01 facility. This thesis brought together in a single text various aspects concerning the proper implementation of these types of measures. Unlike other techniques used in measurements of subcritical reactivity, the methodologies presented in this thesis has the potential to allow the experimental measurement of subcritical reactivity without the prior estimate of any other kinetic parameters, obtained from experiments or from theoretical considerations, external sources calibrations or detectors e ciency measurements. At first, the high-order statistical methods of neutron counts allow to obtain directly the value of the subcriticality (or multiplication factor) from an fissile arrangement regardless the type of subcritical physical theory, and also without the use of unusual infrastructure (such as a pulsed neutron source). These methods are a natural extension of those that use lower order statistics - for example, Feymann-. The greater information content in high order statistics of neutron counting is the main reason for the implementation of this work. Despite its potential, the experimental implementation of the method found huge problems concerning acquisition time and rate of data acquisition. This difficulty overcome any effort in order to obtain a useful measurement inside the IPEN/MB-01 nuclear reactor (a critical facility). However, there are other ways to exploit higher order statistics. For example, an extension of the Rossi- method suggested in this thesis used self bicorrelations. Though the high variance values of obtained results, the fundamental statistical requirements of a measurement were preserved, once the proposed methodologies are observed. It was proposed a methodology to handle dead time issues, in order to allow one to carry out measurement at higher detection rates. Throughout its execution, this thesis aimed to fulfill some gaps in the experimental procedures apparently not addressed by other authors, allowing the establishment of more rigorous statistical procedures. Regarding those contributions, dead time corrections stands out together with the concerning for correlation treatment between the statistical parameters. From the theoretical point of view, this thesis suggests two new ways to address the same problem of using high order statistics of neutron detections in pulse mode: (1) self-bicorrelations, and (2) self-bispectra (power spectral density in two axis). The first was experimentally tested and exhaustively detailed, the second one was only suggested as a theoretical speculation to be confronted against experimental evidence

Oregon State University has hosted an International Atomic Energy Agency (IAEA) International Collaborative Standard Problem (ICSP) through testing conducted on the Multi-Application Small Light Water (MASLWR) facility. The MASLWR facility features a full-time natural circulation loop in the primary vessel and a unique pressure suppression device for accident scenarios. Automatic depressurization system (ADS) lines connect the primary vessel to a high pressure containment (HPC) which dissipates steam heat through a heat transfer plate thermally connected to another vessel with a large cool water inventory. This feature drew the interest of the IAEA and an ICSP was developed where a loss of feedwater to the steam generators prompted a depressurization of the primary vessel via a blowdown through the ADS lines. The purpose of the ICSP is to evaluate the applicability of thermal-hydraulic computer codes to unique experiments usually outside of the validation matrix of the code itself. RELAP5-3D 2:4:2 was chosen to model the ICSP. RELAP5-3D is a best-estimate code designed to simulate transient fluid and thermal behavior in light water reactors. Modeling was conducted in RELAP5-3D to identify the strengths and weaknesses of the code in predicting the experimental trends of the IAEA ICSP. This extended to nodalization sensitivity studies, an investigation of built-in models and heat transfer boundary conditions. Besides a qualitative analysis, a quantitative analysis method was also performed. / Graduation date: 2013