Ydinreaktorin painesäiliön ulkoisen jäähdytyksen lämmönsiirtokokeita

The purpose of this work was to design and carry out thermal-hydraulic experiments dealing with overcooling transients of a VVER-440-type nuclear reactor pressure vessel. Sudden overcooling accident could have negative effect on the mechanical strength of the pressure vessel. If part of the pressure vessel is compromised, the intense pressure inside a pressurized water reactor could cause the wall to fracture. Information on the heat transfer along the outside of the pressure vessel wall is necessary for stress analysis. Basic knowledge of the overcooling accident and heat transfer types on the outside of the pressure vessel is presented as background information. Test facility was designed and built based to study and measure heat transfer during specific overcooling scenarios. Two test series were conducted with the first one concentrating on the very beginning of the transient and the second one concentrating on steady state heat transfer. Heat transfer coefficients are calculated from the test data using an inverse method, which yields better results in fast transients than direct calculation from the measurement results. The results show that heat transfer rate varies considerably during the transient, being very high in the beginning and dropping to steady state in a few minutes. The test results show that appropriate correlations can be used in future analysis.

Vertailuanalyysin käyttö termohydraulisten koelaitteistojen tulosten laitosmittakaavaan skaalautumisen tutkimisessa

Työn tavoitteena oli tarkastella termohydraulisten koelaitteistojen skaalauksessa käytettäviä periaatteita ja menettelyjä sekä vertailla Apros-simulaattoriohjelmalla laskettuja kanden koelaitteistomallin ja EPR-mallin tuloksia. Tarkoituksena oli saada käsitys siitä, miten hyvin tarkastellut koelaitteistot kuvaavat EPR-Iaitostyypin käyttäytymistä onnettomuustilanteessa. Malleilla tutkittiin jäähdytteen määrän vaikutusta primääripiirin käyttäytymiseen. Koelaitteistomallien tuloksissa toistuvat samat ilmiöt kuin EPR-mallin tuloksissa. Laskettuja PKL-koelaitteistomallin tuloksia vertailtiin myös koelaitteistolla suoritettuun kokeeseen. PKL-mallin todettiin toistavan hyvin kokeen tulokset. Koelaitteistojen tuloksien perusteella kelpoistetaan laskentaohjelmia, joita käytetään ydinvoimalaitosten turvallisuustutkimuksessa. Erityistä harkintaa tulee käyttää koelaitteistojen tulosten hyödyntämisessä, sillä mittakaava vaikuttaa ilmiöiden esiintymiseen.

Heat Transfer in Plate Heat Exchangers

This study illustrates the different types of plate heat exchangers that are commonly used in various domestic and industrial applications. The main purpose of this paper was to devise a methodology that is capable of calculating optimum number of plates in the design of a plate heat exchanger. To obtain the appropriate number of plates, typically several iterations must be made before a final acceptable design is completed, since plate amount depends on many factors such as, flow velocities, physical properties of the streams, flow channel geometry, allowable pressure drop, plate dimensions, and the gap between the plates. The methodology presented here can be used as a general guide for designing a plate heat exchanger. To investigate the effects of relevant parameters on the thermal-hydraulic design of a plate heat exchanger, several experiments were carried out for single-phase and counter flow arrangement with two brazed plate heat exchangers by varying the flow rates and the inlet temperatures of the fluid streams. The actual heat transfer coefficients obtained based on the experiment were nearly close to the calculated values and to improve the design, a correction factor was introduced. Besides, the effect of flow channel velocity on the pressure drop inside the unit is presented.

Improving the calculation of rate of stratification in APROS

APROS (Advanced Process Simulation Environment) is a computer simulation program developed to simulate thermal hydraulic processes in nuclear and conventional power plants. Earlier research at VTT Technological Research Centre of Finland had found the current version of APROS to produce inaccurate simulation results for a certain case of loop seal clearing. The objective of this Master’s thesis is to find and implement an alternative method for calculating the rate of stratification in APROS, which was found to be the reason for the inaccuracies. Brief literature study was performed and a promising candidate for the new method was found. The new method was implemented into APROS and tested against experiments and simulations from two test facilities and the current version of APROS. Simulation results with the new version were partially conflicting; in some cases the new method was more accurate than the current version, in some the current method was better. Overall, the new method can be assessed as an improvement.

Monte Carlo -reaktorifysiikkalaskennan ja laskennallisen virtausmekaniikan kytkentä kuulakekoreaktorissa

Monte Carlo -reaktorifysiikkakoodit nykyisin käytettävissä olevilla laskentatehoilla tarjoavat mielenkiintoisen tavan reaktorifysiikan ongelmien ratkaisuun. Neljännen sukupolven ydinreaktoreissa käytettävät uudet rakenteet ja materiaalit ovat haasteellisia nykyisiin reaktoreihin suunnitelluille laskentaohjelmille. Tässä työssä Monte Carlo -reaktorifysiikkakoodi ja CFD-koodi yhdistetään kytkettyyn laskentaan kuulakekoreaktorissa, joka on yksi korkealämpötilareaktorityyppi. Työssä käytetty lähestymistapa on uutta maailmankin mittapuussa ajateltuna.

Onset of Nucleate Boiling and Onset of Fully Developed Subcooled Boiling Using Pressure Transducers Signals Spectral Analysis

The experimental technique used for detection of subcooled boiling through analysis of the fluctuation contained in pressure transducer signals is presented. This work was partly conducted at the Institut für Kerntechnik und zertörungsfreie Prüfverfahren von Hannover (IKPH, Germany) in a thermal-hydraulic circuit with one electrically heated rod with annular geometry test section. Piezoresistive pressure sensors are used for onset of nucleate boiling (ONB) and onset of fully developed boiling (OFDB) detection using spectral analysis/ signal correlation techniques. Experimental results are interpreted by phenomenological analysis of these two points and compared with existing correlation. The results allow us to conclude that this technique is adequate for the detection and monitoring of the ONB and OFDB.

Coolability of porous core debris beds : Effects of bed geometry and multi-dimensional flooding

This thesis addresses the coolability of porous debris beds in the context of severe accident management of nuclear power reactors. In a hypothetical severe accident at a Nordic-type boiling water reactor, the lower drywell of the containment is flooded, for the purpose of cooling the core melt discharged from the reactor pressure vessel in a water pool. The melt is fragmented and solidified in the pool, ultimately forming a porous debris bed that generates decay heat. The properties of the bed determine the limiting value for the heat flux that can be removed from the debris to the surrounding water without the risk of re-melting. The coolability of porous debris beds has been investigated experimentally by measuring the dryout power in electrically heated test beds that have different geometries. The geometries represent the debris bed shapes that may form in an accident scenario. The focus is especially on heap-like, realistic geometries which facilitate the multi-dimensional infiltration (flooding) of coolant into the bed. Spherical and irregular particles have been used to simulate the debris. The experiments have been modeled using 2D and 3D simulation codes applicable to fluid flow and heat transfer in porous media. Based on the experimental and simulation results, an interpretation of the dryout behavior in complex debris bed geometries is presented, and the validity of the codes and models for dryout predictions is evaluated. According to the experimental and simulation results, the coolability of the debris bed depends on both the flooding mode and the height of the bed. In the experiments, it was found that multi-dimensional flooding increases the dryout heat flux and coolability in a heap-shaped debris bed by 47–58% compared to the dryout heat flux of a classical, top-flooded bed of the same height. However, heap-like beds are higher than flat, top-flooded beds, which results in the formation of larger steam flux at the top of the bed. This counteracts the effect of the multi-dimensional flooding. Based on the measured dryout heat fluxes, the maximum height of a heap-like bed can only be about 1.5 times the height of a top-flooded, cylindrical bed in order to preserve the direct benefit from the multi-dimensional flooding. In addition, studies were conducted to evaluate the hydrodynamically representative effective particle diameter, which is applied in simulation models to describe debris beds that consist of irregular particles with considerable size variation. The results suggest that the effective diameter is small, closest to the mean diameter based on the number or length of particles.

Analysis of PWR-PACTEL Small Break LOCA experiment using TRACE

Currently, the power generation is one of the most significant life aspects for the whole man-kind. Barely one can imagine our life without electricity and thermal energy. Thus, different technologies for producing those types of energy need to be used. Each of those technologies will always have their own advantages and disadvantages. Nevertheless, every technology must satisfy such requirements as efficiency, ecology safety and reliability. In the matter of the power generation with nuclear energy utilization these requirements needs to be highly main-tained, especially since accidents on nuclear power plants may cause very long term deadly consequences. In order to prevent possible disasters related to the accident on a nuclear power plant strong and powerful algorithms were invented in last decades. Such algorithms are able to manage calculations of different physical processes and phenomena of real facilities. How-ever, the results acquired by the computing must be verified with experimental data.

Simulação numérica de trocadores de calor multitubulares equipados com fitas helicoidais externamente aos tubos

Classical shell-and-tube heat exchangers are usually equipped with segmental baffles. These baffles serve two basic functions: (a) they provide tube supports, thereby preventing or reducing mechanical problems, such as sagging or vibration; (b) they direct the fluid flow over the tubes so as to introduce a cross-flow component, thereby increasing the heat transfer. Segmented baffles have several sources of performance loss, some due to various leakage flows and others caused by stagnation zones. A new concept of longitudinal flow heat exchanger - based on placing twisted tapes along the tube bundle subchannels - was developed to mitigate drawbacks of other types of tubular heat exchangers. In this paper, a numerical model has been implemented in order to simulate the thermal-hydraulic feature of tubular heat exchangers equipped either with segmental baffles or with subchannel twisted tapes. The tube bundle has been described by means of an equivalent porous medium type model, allowing a macroscopic description of the shell-side flow. The basic equations - continuity, momentum and energy - have been solved by using the finite volume method. Typical numerical results have been compared with experimental data, reaching a very good agreement. A comparative analysis of different types of heat exchangers has been carried out, revealing the satisfactory thermal-hydraulic efficiency level of the twisted tapes heat exchangers.

Selected papers from OECD-NEA PSBT benchmark

Historically, the prediction of safety margins has been based on system level thermal-hydraulic calculations employing suitable empirical formulations for assembly specific geometries and fuel-element grid spacers. These works have assessed response, margins, and consequences for the system based on one-dimensional two-fluid or drift-flux type thermalhydraulics formulations with fuel-vendor specific hydraulic losses and heat transfer characteristics for various fuel assemblies, including the so-called hot channel. Analysis of the hot channel gives important information on flow rates, fuel element centerline temperature, fuel sheath temperature, and margin to the departure from nucleate boiling. Given the reliance of the above approaches on empirical formulations obtained from complex and often difficult experiments, there is significant interest in obtaining reliable and accurate results from computation tools which employ more fundamental empirical relationships which can be obtained from subsets of the domain or from other scaled experiments.

Tritium permeation experiment at IFMIF Medium Flux Test Module

Tritium release experiments using different breeding material candidates are planned for the medium flux region of the IFMIF Test Cell. Nowadays, only ceramic breeder materials have been suggested to be tested in the Tritium Release Module located in the Medium Flux Test Module of IFMIF. Liquid breeder blankets are very promising and for that reason, several concepts will be tested in ITER. One of the main problems concerning the liquid blankets is the permeation of the generated tritium in the breeder throughout the walls. Since tritium permeation is highly influenced by irradiation conditions, IFMIF is a suitable scenario to perform tritium permeation related experiments. In this paper, a preliminary design of a tritium permeation experiment for the Medium Flux Test Module of IFMIF is proposed, in order to contribute to the progress of the liquid breeder blanket concept validation. The conceptual design of the capsule in which the experiment will be performed is carried out, taking into consideration the experiment necessities and its implementation in the Tritium Release Module. In addition to this, some thermal hydraulic calculations have been performed to evaluate the thermal behaviour of the irradiation capsule

Intensificação da transferência de calor e otimização de trocadores de calor compactos tipo venezianas com geradores de vórtices tipo delta-winglets.

Esta pesquisa visa a análise da contribuição de cinco variáveis de entrada e a otimização do desempenho termo-hidráulico de trocadores de calor com venezianas combinados com geradores de vórtices delta-winglets. O desempenho termohidráulico de duas geometrias distintas, aqui nomeadas por GEO1 e GEO2, foram avaliadas. Smoothing Spline ANOVA foi usado para avaliar a contribuição dos parâmetros de entrada na transferência de calor e perda de carga. Considerando aplicação automotiva, foram investigados números de Reynolds iguais a 120 e 240, baseados no diâmetro hidráulico. Os resultados indicaram que o ângulo de venezianas é o maior contribuidor para o aumento do fator de atrito para GEO1 e GEO2, para ambos os números de Reynolds. Para o número de Reynolds menor, o parâmetro mais importante em termos de transferência de calor foi o ângulo das venezianas para ambas as geometrias. Para o número de Reynolds maior, o ângulo de ataque dos geradores de vórtices posicionados na primeira fileira é o maior contribuidor para a tranfesferência de calor, no caso da geometria GEO1, enquanto que o ângulo de ataque dos geradores de vórtices na primeira fileira foi tão importante quanto os ângulos das venezianas para a geometria GEO2. Embora as geometrias analisadas possam ser consideradas como técnicas compostas de intensificação da transferência de calor, não foram observadas interações relevantes entre ângulo de venezianas e parâmetros dos geradores de vórtices. O processo de otimização usa NSGA-II (Non-Dominated Sorting Genetic Algorithm) combinado com redes neurais artificiais. Os resultados mostraram que a adição dos geradores de vórtices em GEO1 aumentaram a transferência de calor em 21% e 23% com aumentos na perda de carga iguais a 24,66% e 36,67% para o menor e maior números de Reynolds, respectivamente. Para GEO2, a transferência de calor aumentou 13% e 15% com aumento na perda de carga de 20,33% e 23,70%, para o menor e maior número de Reynolds, respectivamente. As soluções otimizadas para o fator de Colburn mostraram que a transferência de calor atrás da primeira e da segunda fileiras de geradores de vórtices tem a mesma ordem de magnitude para ambos os números de Reynolds. Os padrões de escoamento e as características de transferência de calor das soluções otimizadas apresentaram comportamentos vi particulares, diferentemente daqueles encontrados quando as duas técnicas de intensificação de transferência de calor são aplicadas separadamente.

Aplicação de mapas auto-organizáveis na classificação de padrões de escoamento bifásico

O escoamento bifásico de gás-líquido é encontrado em muitos circuitos fechados que utilizam circulação natural para fins de resfriamento. O fenômeno da circulação natural é importante nos recentes projetos de centrais nucleares para a remoção de calor. O circuito de circulação natural (Circuito de Circulação Natural - CCN), instalado no Instituto de Pesquisas Energéticas e Nucleares, IPEN / CNEN, é um circuito experimento concebido para fornecer dados termo-hidráulicos relacionados com escoamento monofásico ou bifásico em condições de circulação natural. A estimativa de transferência de calor tem sido melhorada com base em modelos que requerem uma previsão precisa de transições de padrão de escoamento. Este trabalho apresenta testes experimentais desenvolvidos no CCN para a visualização dos fenômenos de instabilidade em ciclos de circulação natural básica e classificar os padrões de escoamento bifásico associados aos transientes e instabilidades estáticas de escoamento. As imagens são comparadas e agrupadas utilizando mapas auto-organizáveis de Kohonen (SOM), aplicados em diferentes características da imagem digital. Coeficientes da Transformada Discreta de Cossenos de Quadro Completo (FFDCT) foram utilizados como entrada para a tarefa de classificação, levando a bons resultados. Os protótipos de FFDCT obtidos podem ser associados a cada padrão de escoamento possibilitando uma melhor compreensão da instabilidade observada. Uma metodologia sistemática foi utilizada para verificar a robustez do método.

Desenvolvimento de um elemento combustível instrumentado para o reator de pesquisa IEA-R1

Após o aumento de potência do reator IEA-R1 de 2 MW para 5 MW observou-se um aumento da taxa de corrosão nas placas laterais de alguns elementos combustíveis e algumas dúvidas surgiram com relação ao valor de vazão utilizada nas análises termo-hidráulicas. A fim de esclarecer e medir a distribuição de vazão real pelos elementos combustíveis que compõe o núcleo do reator IEA-R1, um elemento combustível protótipo, sem material nuclear, chamado DMPV-01 (Dispositivo para Medida de Pressão e Vazão), em escala real, foi projetado e construído em alumínio. A vazão no canal entre dois elementos combustíveis é muito difícil de estimar ou ser medida. Esta vazão é muito importante no processo de resfriamento das placas laterais. Este trabalho apresenta a concepção e construção de um elemento combustível instrumentado para medir a temperatura real nestas placas laterais para melhor avaliar as condições de resfriamento do combustível. Quatorze termopares foram instalados neste elemento combustível instrumentado. Quatro termopares em cada canal lateral e quatro no canal central, além de um termopar no bocal de entrada e outro no bocal de saída do elemento. Existem três termopares para medida de temperatura do revestimento e um para a temperatura do fluido em cada canal. Três séries de experimentos, para três configurações distintas, foram realizadas com o elemento combustível instrumentado. Em dois experimentos uma caixa de alumínio foi instalada ao redor do núcleo para reduzir o escoamento transverso entre os elementos combustíveis e medir o impacto na temperatura das placas externas. Dada a tamanha quantidade de informações obtidas e sua utilidade no projeto, melhoria e capacitação na construção, montagem e fabricação de elementos combustíveis instrumentados, este projeto constitui um importante marco no estudo de núcleos de reatores de pesquisa. As soluções propostas podem ser amplamente utilizadas para outros reatores de pesquisa.

Proceedings of the U.S. Nuclear Regulatory Commission Eleventh Water Reactor Safety Research Information Meeting, held at National Bureau of Standards, Gaithersburg, Maryland, October 24-28, 1983 /

Includes bibliographies.