VVER-440 Thermal Hydraulics as a Computer Code Validation Challenge

This thesis concentrates on the validation of a generic thermal hydraulic computer code TRACE under the challenges of the VVER-440 reactor type. The code capability to model the VVER-440 geometry and thermal hydraulic phenomena specific to this reactor design has been examined and demonstrated acceptable. The main challenge in VVER-440 thermal hydraulics appeared in the modelling of the horizontal steam generator. The major challenge here is not in the code physics or numerics but in the formulation of a representative nodalization structure. Another VVER-440 specialty, the hot leg loop seals, challenges the system codes functionally in general, but proved readily representable. Computer code models have to be validated against experiments to achieve confidence in code models. When new computer code is to be used for nuclear power plant safety analysis, it must first be validated against a large variety of different experiments. The validation process has to cover both the code itself and the code input. Uncertainties of different nature are identified in the different phases of the validation procedure and can even be quantified. This thesis presents a novel approach to the input model validation and uncertainty evaluation in the different stages of the computer code validation procedure. This thesis also demonstrates that in the safety analysis, there are inevitably significant uncertainties that are not statistically quantifiable; they need to be and can be addressed by other, less simplistic means, ultimately relying on the competence of the analysts and the capability of the community to support the experimental verification of analytical assumptions. This method completes essentially the commonly used uncertainty assessment methods, which are usually conducted using only statistical methods.

Modelling of hydrogen stratification and gas mixing in containment using the APROS code

Hydrogen stratification and atmosphere mixing is a very important phenomenon in nuclear reactor containments when severe accidents are studied and simulated. Hydrogen generation, distribution and accumulation in certain parts of containment may pose a great risk to pressure increase induced by hydrogen combustion, and thus, challenge the integrity of NPP containment. The accurate prediction of hydrogen distribution is important with respect to the safety design of a NPP. Modelling methods typically used for containment analyses include both lumped parameter and field codes. The lumped parameter method is universally used in the containment codes, because its versatility, flexibility and simplicity. The lumped parameter method allows fast, full-scale simulations, where different containment geometries with relevant engineering safety features can be modelled. Lumped parameter gas stratification and mixing modelling methods are presented and discussed in this master’s thesis. Experimental research is widely used in containment analyses. The HM-2 experiment related to hydrogen stratification and mixing conducted at the THAI facility in Germany is calculated with the APROS lump parameter containment package and the APROS 6-equation thermal hydraulic model. The main purpose was to study, whether the convection term included in the momentum conservation equation of the 6-equation modelling gives some remarkable advantages compared to the simplified lumped parameter approach. Finally, a simple containment test case (high steam release to a narrow steam generator room inside a large dry containment) was calculated with both APROS models. In this case, the aim was to determine the extreme containment conditions, where the effect of convection term was supposed to be possibly high. Calculation results showed that both the APROS containment and the 6-equation model could model the hydrogen stratification in the THAI test well, if the vertical nodalisation was dense enough. However, in more complicated cases, the numerical diffusion may distort the results. Calculation of light gas stratification could be probably improved by applying the second order discretisation scheme for the modelling of gas flows. If the gas flows are relatively high, the convection term of the momentum equation is necessary to model the pressure differences between the adjacent nodes reasonably.

Code assessment and modelling for Design Basis Accident Analysis of the European sodium fast reactor design. Part I: System description, modelling and benchmarking

The new reactor concepts proposed in the Generation IV International Forum (GIF) are conceived to improve the use of natural resources, reduce the amount of high-level radioactive waste and excel in their reliability and safe operation. Among these novel designs sodium fast reactors (SFRs) stand out due to their technological feasibility as demonstrated in several countries during the last decades. As part of the contribution of EURATOM to GIF the CP-ESFR is a collaborative project with the objective, among others, to perform extensive analysis on safety issues involving renewed SFR demonstrator designs. The verification of computational tools able to simulate the plant behaviour under postulated accidental conditions by code-to-code comparison was identified as a key point to ensure reactor safety. In this line, several organizations employed coupled neutronic and thermal-hydraulic system codes able to simulate complex and specific phenomena involving multi-physics studies adapted to this particular fast reactor technology. In the “Introduction” of this paper the framework of this study is discussed, the second section describes the envisaged plant design and the commonly agreed upon modelling guidelines. The third section presents a comparative analysis of the calculations performed by each organisation applying their models and codes to a common agreed transient with the objective to harmonize the models as well as validating the implementation of all relevant physical phenomena in the different system codes.

Analysis of the damage domains of sbo sequences with rcp passive thermal shutdown

The integrated Safety Assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council (CSN), has been applied to a thermal-hydraulic analysis of PWR Station Blackout (SBO) sequences in the context of the IDPSA (Integrated Deterministic-Probabilistic Safety Assessment) network objectives. The ISA methodology allows obtaining the damage domain (the region of the uncertain parameters space where the damage limit is exceeded) for each sequence of interest as a function of the operator actuations times. Given a particular safety limit or damage limit, several data of every sequence are necessary in order to obtain the exceedance frequency of that limit. In this application these data are obtained from the results of the simulations performed with MAAP code transients inside each damage domain and the time-density probability distributions of the manual actions. Damage limits that have been taken into account within this analysis are: local cladding damage (PCT>1477 K); local fuel melting (T>2499 K); fuel relocation in lower plenum and vessel failure. Therefore, to every one of these damage variables corresponds a different damage domain. The operation of the new passive thermal shutdown seals developed by several companies since Fukushima accident is considered in the paper. The results show the capability and necessity of the ISA methodology, or similar, in order to obtain accurate results that take into account time uncertainties.

Effect of Noncondensable Gases on Circulation of Primary Coolant in Nuclear Power Plants in Abnormal Situations

The present study focuses on two effects of the presence of a noncondensable gas on the thermal-hydraulic behavior of thecoolant of the primary circuit of a nuclear reactor in the VVER-440 geometry inabnormal situations. First, steam condensation with the presence of air was studied in the horizontal tubes of the steam generator (SG) of the PACTEL test facility. The French thermal-hydraulic CATHARE code was used to study the heat transfer between the primary and secondary side in conditions derived from preliminary experiments performed by VTT using PACTEL. In natural circulation and single-phase vapor conditions, the injection of a volume of air, equivalent to the totalvolume of the primary side of the SG at the entrance of the hot collector, did not stop the heat transfer from the primary to the secondary side. The calculated results indicate that air is located in the second half-length (from the mid-length of the tubes to the cold collector) in all the tubes of the steam generator The hot collector remained full of steam during the transient. Secondly, the potential release of the nitrogen gas dissolved in the water of the accumulators of the emergency core coolant system of the Loviisa nuclear power plant (NPP) was investigated. The author implemented a model of the dissolution and release ofnitrogen gas in the CATHARE code; the model created by the CATHARE developers. In collaboration with VTT, an analytical experiment was performed with some components of PACTEL to determine, in particular, the value of the release time constant of the nitrogen gas in the depressurization conditions representative of the small and intermediate break transients postulated for the Loviisa NPP. Such transients, with simplified operating procedures, were calculated using the modified CATHARE code for various values of the release time constant used in the dissolution and release model. For the small breaks, nitrogen gas is trapped in thecollectors of the SGs in rather large proportions. There, the levels oscillate until the actuation of the low-pressure injection pumps (LPIS) that refill the primary circuit. In the case of the intermediate breaks, most of the nitrogen gas is expelled at the break and almost no nitrogen gas is trapped in the SGs. In comparison with the cases calculated without taking into account the release of nitrogen gas, the start of the LPIS is delayed by between 1 and 1.75 h. Applicability of the obtained results to the real safety conditions must take into accountthe real operating procedures used in the nuclear power plant.

Application and development of numerical methods for the modelling of innovative gas cooled fission reactors

Innovative gas cooled reactors, such as the pebble bed reactor (PBR) and the gas cooled fast reactor (GFR) offer higher efficiency and new application areas for nuclear energy. Numerical methods were applied and developed to analyse the specific features of these reactor types with fully three dimensional calculation models. In the first part of this thesis, discrete element method (DEM) was used for a physically realistic modelling of the packing of fuel pebbles in PBR geometries and methods were developed for utilising the DEM results in subsequent reactor physics and thermal-hydraulics calculations. In the second part, the flow and heat transfer for a single gas cooled fuel rod of a GFR were investigated with computational fluid dynamics (CFD) methods. An in-house DEM implementation was validated and used for packing simulations, in which the effect of several parameters on the resulting average packing density was investigated. The restitution coefficient was found out to have the most significant effect. The results can be utilised in further work to obtain a pebble bed with a specific packing density. The packing structures of selected pebble beds were also analysed in detail and local variations in the packing density were observed, which should be taken into account especially in the reactor core thermal-hydraulic analyses. Two open source DEM codes were used to produce stochastic pebble bed configurations to add realism and improve the accuracy of criticality calculations performed with the Monte Carlo reactor physics code Serpent. Russian ASTRA criticality experiments were calculated. Pebble beds corresponding to the experimental specifications within measurement uncertainties were produced in DEM simulations and successfully exported into the subsequent reactor physics analysis. With the developed approach, two typical issues in Monte Carlo reactor physics calculations of pebble bed geometries were avoided. A novel method was developed and implemented as a MATLAB code to calculate porosities in the cells of a CFD calculation mesh constructed over a pebble bed obtained from DEM simulations. The code was further developed to distribute power and temperature data accurately between discrete based reactor physics and continuum based thermal-hydraulics models to enable coupled reactor core calculations. The developed method was also found useful for analysing sphere packings in general. CFD calculations were performed to investigate the pressure losses and heat transfer in three dimensional air cooled smooth and rib roughened rod geometries, housed inside a hexagonal flow channel representing a sub-channel of a single fuel rod of a GFR. The CFD geometry represented the test section of the L-STAR experimental facility at Karlsruhe Institute of Technology and the calculation results were compared to the corresponding experimental results. Knowledge was gained of the adequacy of various turbulence models and of the modelling requirements and issues related to the specific application. The obtained pressure loss results were in a relatively good agreement with the experimental data. Heat transfer in the smooth rod geometry was somewhat under predicted, which can partly be explained by unaccounted heat losses and uncertainties. In the rib roughened geometry heat transfer was severely under predicted by the used realisable k − epsilon turbulence model. An additional calculation with a v2 − f turbulence model showed significant improvement in the heat transfer results, which is most likely due to the better performance of the model in separated flow problems. Further investigations are suggested before using CFD to make conclusions of the heat transfer performance of rib roughened GFR fuel rod geometries. It is suggested that the viewpoints of numerical modelling are included in the planning of experiments to ease the challenging model construction and simulations and to avoid introducing additional sources of uncertainties. To facilitate the use of advanced calculation approaches, multi-physical aspects in experiments should also be considered and documented in a reasonable detail.

Failure modes of passive decay heat removing safety systems of modern nuclear power plants

The purpose of this master’s thesis is to gain an understanding of passive safety systems’ role in modern nuclear reactors projects and to research the failure modes of passive decay heat removal safety systems which use phenomenon of natural circulation. Another purpose is to identify the main physical principles and phenomena which are used to establish passive safety tools in nuclear power plants. The work describes passive decay heat removal systems used in AES-2006 project and focuses on the behavior of SPOT PG system. The descriptions of the main large-scale research facilities of the passive safety systems of the AES-2006 power plant are also included. The work contains the calculations of the SPOT PG system, which was modeled with thermal-hydraulic system code TRACE. The dimensions of the calculation model are set according to the dimensions of the real SPOT PG system. In these calculations three parameters are investigated as a function of decay heat power: the pressure of the system, the natural circulation mass flow rate around the closed loop, and the level of liquid in the downcomer. The purpose of the calculations is to test the ability of the SPOT PG system to remove the decay heat from the primary side of the nuclear reactor in case of failure of one, two, or three loops out of four. The calculations show that three loops of the SPOT PG system have adequate capacity to provide the necessary level of safety. In conclusion, the work supports the view that passive systems could be widely spread in modern nuclear projects.

Steps Ahead in the Few-Group CRoss-Section Library Generation at the Pin Level

There exists an interest in performing pin-by-pin calculations coupled with thermal hydraulics so as to improve the accuracy of nuclear reactor analysis. In the framework of the EU NURISP project, INRNE and UPM have generated an experimental version of a few group diffusion cross sections library with discontinuity factors intended for VVER analysis at the pin level with the COBAYA3 code. The transport code APOLLO2 was used to perform the branching calculations. As a first proof of principle the library was created for fresh fuel and covers almost the full parameter space of steady state and transient conditions. The main objective is to test the calculation schemes and post-processing procedures, including multi-pin branching calculations. Two library options are being studied: one based on linear table interpolation and another one using a functional fitting of the cross sections. The libraries generated with APOLLO2 have been tested with the pin-by-pin diffusion model in COBAYA3 including discontinuity factors; first comparing 2D results against the APOLLO2 reference solutions and afterwards using the libraries to compute a 3D assembly problem coupled with a simplified thermal-hydraulic model.

Efficient algorithms for dynamic probabilistic safety assessment. An application to convex damage domain.

The design of nuclear power plant has to follow a number of regulations aimed at limiting the risks inherent in this type of installation. The goal is to prevent and to limit the consequences of any possible incident that might threaten the public or the environment. To verify that the safety requirements are met a safety assessment process is followed. Safety analysis is as key component of a safety assessment, which incorporates both probabilistic and deterministic approaches. The deterministic approach attempts to ensure that the various situations, and in particular accidents, that are considered to be plausible, have been taken into account, and that the monitoring systems and engineered safety and safeguard systems will be capable of ensuring the safety goals. On the other hand, probabilistic safety analysis tries to demonstrate that the safety requirements are met for potential accidents both within and beyond the design basis, thus identifying vulnerabilities not necessarily accessible through deterministic safety analysis alone. Probabilistic safety assessment (PSA) methodology is widely used in the nuclear industry and is especially effective in comprehensive assessment of the measures needed to prevent accidents with small probability but severe consequences. Still, the trend towards a risk informed regulation (RIR) demanded a more extended use of risk assessment techniques with a significant need to further extend PSA’s scope and quality. Here is where the theory of stimulated dynamics (TSD) intervenes, as it is the mathematical foundation of the integrated safety assessment (ISA) methodology developed by the CSN(Consejo de Seguridad Nuclear) branch of Modelling and Simulation (MOSI). Such methodology attempts to extend classical PSA including accident dynamic analysis, an assessment of the damage associated to the transients and a computation of the damage frequency. The application of this ISA methodology requires a computational framework called SCAIS (Simulation Code System for Integrated Safety Assessment). SCAIS provides accident dynamic analysis support through simulation of nuclear accident sequences and operating procedures. Furthermore, it includes probabilistic quantification of fault trees and sequences; and integration and statistic treatment of risk metrics. SCAIS comprehensively implies an intensive use of code coupling techniques to join typical thermal hydraulic analysis, severe accident and probability calculation codes. The integration of accident simulation in the risk assessment process and thus requiring the use of complex nuclear plant models is what makes it so powerful, yet at the cost of an enormous increase in complexity. As the complexity of the process is primarily focused on such accident simulation codes, the question of whether it is possible to reduce the number of required simulation arises, which will be the focus of the present work. This document presents the work done on the investigation of more efficient techniques applied to the process of risk assessment inside the mentioned ISA methodology. Therefore such techniques will have the primary goal of decreasing the number of simulation needed for an adequate estimation of the damage probability. As the methodology and tools are relatively recent, there is not much work done inside this line of investigation, making it a quite difficult but necessary task, and because of time limitations the scope of the work had to be reduced. Therefore, some assumptions were made to work in simplified scenarios best suited for an initial approximation to the problem. The following section tries to explain in detail the process followed to design and test the developed techniques. Then, the next section introduces the general concepts and formulae of the TSD theory which are at the core of the risk assessment process. Afterwards a description of the simulation framework requirements and design is given. Followed by an introduction to the developed techniques, giving full detail of its mathematical background and its procedures. Later, the test case used is described and result from the application of the techniques is shown. Finally the conclusions are presented and future lines of work are exposed.

Diseño de un módulo de IFMIF para irradiación de envolturas regeneradoras basadas en metales líquidos para reactores de fusión

La fusión nuclear es, hoy en día, una alternativa energética a la que la comunidad internacional dedica mucho esfuerzo. El objetivo es el de generar entre diez y cincuenta veces más energía que la que consume mediante reacciones de fusión que se producirán en una mezcla de deuterio (D) y tritio (T) en forma de plasma a doscientos millones de grados centígrados. En los futuros reactores nucleares de fusión será necesario producir el tritio utilizado como combustible en el propio reactor termonuclear. Este hecho supone dar un paso más que las actuales máquinas experimentales dedicadas fundamentalmente al estudio de la física del plasma. Así pues, el tritio, en un reactor de fusión, se produce en sus envolturas regeneradoras cuya misión fundamental es la de blindaje neutrónico, producir y recuperar tritio (fuel para la reacción DT del plasma) y por último convertir la energía de los neutrones en calor. Existen diferentes conceptos de envolturas que pueden ser sólidas o líquidas. Las primeras se basan en cerámicas de litio (Li2O, Li4SiO4, Li2TiO3, Li2ZrO3) y multiplicadores neutrónicos de Be, necesarios para conseguir la cantidad adecuada de tritio. Los segundos se basan en el uso de metales líquidos o sales fundidas (Li, LiPb, FLIBE, FLINABE) con multiplicadores neutrónicos de Be o el propio Pb en el caso de LiPb. Los materiales estructurales pasan por aceros ferrítico-martensíticos de baja activación, aleaciones de vanadio o incluso SiCf/SiC. Cada uno de los diferentes conceptos de envoltura tendrá una problemática asociada que se estudiará en el reactor experimental ITER (del inglés, “International Thermonuclear Experimental Reactor”). Sin embargo, ITER no puede responder las cuestiones asociadas al daño de materiales y el efecto de la radiación neutrónica en las diferentes funciones de las envolturas regeneradoras. Como referencia, la primera pared de un reactor de fusión de 4000MW recibiría 30 dpa/año (valores para Fe-56) mientras que en ITER se conseguirían <10 dpa en toda su vida útil. Esta tesis se encuadra en el acuerdo bilateral entre Europa y Japón denominado “Broader Approach Agreement “(BA) (2007-2017) en el cual España juega un papel destacable. Estos proyectos, complementarios con ITER, son el acelerador para pruebas de materiales IFMIF (del inglés, “International Fusion Materials Irradiation Facility”) y el dispositivo de fusión JT-60SA. Así, los efectos de la irradiación de materiales en materiales candidatos para reactores de fusión se estudiarán en IFMIF. El objetivo de esta tesis es el diseño de un módulo de IFMIF para irradiación de envolturas regeneradoras basadas en metales líquidos para reactores de fusión. El módulo se llamará LBVM (del inglés, “Liquid Breeder Validation Module”). La propuesta surge de la necesidad de irradiar materiales funcionales para envolturas regeneradoras líquidas para reactores de fusión debido a que el diseño conceptual de IFMIF no contaba con esta utilidad. Con objeto de analizar la viabilidad de la presente propuesta, se han realizado cálculos neutrónicos para evaluar la idoneidad de llevar a cabo experimentos relacionados con envolturas líquidas en IFMIF. Así, se han considerado diferentes candidatos a materiales funcionales de envolturas regeneradoras: Fe (base de los materiales estructurales), SiC (material candidato para los FCI´s (del inglés, “Flow Channel Inserts”) en una envoltura regeneradora líquida, SiO2 (candidato para recubrimientos antipermeación), CaO (candidato para recubrimientos aislantes), Al2O3 (candidato para recubrimientos antipermeación y aislantes) y AlN (material candidato para recubrimientos aislantes). En cada uno de estos materiales se han calculado los parámetros de irradiación más significativos (dpa, H/dpa y He/dpa) en diferentes posiciones de IFMIF. Estos valores se han comparado con los esperados en la primera pared y en la zona regeneradora de tritio de un reactor de fusión. Para ello se ha elegido un reactor tipo HCLL (del inglés, “Helium Cooled Lithium Lead”) por tratarse de uno de los más prometedores. Además, los valores también se han comparado con los que se obtendrían en un reactor rápido de fisión puesto que la mayoría de las irradiaciones actuales se hacen en reactores de este tipo. Como conclusión al análisis de viabilidad, se puede decir que los materiales funcionales para mantos regeneradores líquidos podrían probarse en la zona de medio flujo de IFMIF donde se obtendrían ratios de H/dpa y He/dpa muy parecidos a los esperados en las zonas más irradiadas de un reactor de fusión. Además, con el objetivo de ajustar todavía más los valores, se propone el uso de un moderador de W (a considerar en algunas campañas de irradiación solamente debido a que su uso hace que los valores de dpa totales disminuyan). Los valores obtenidos para un reactor de fisión refuerzan la idea de la necesidad del LBVM, ya que los valores obtenidos de H/dpa y He/dpa son muy inferiores a los esperados en fusión y, por lo tanto, no representativos. Una vez demostrada la idoneidad de IFMIF para irradiar envolturas regeneradoras líquidas, y del estudio de la problemática asociada a las envolturas líquidas, también incluida en esta tesis, se proponen tres tipos de experimentos diferentes como base de diseño del LBVM. Éstos se orientan en las necesidades de un reactor tipo HCLL aunque a lo largo de la tesis se discute la aplicabilidad para otros reactores e incluso se proponen experimentos adicionales. Así, la capacidad experimental del módulo estaría centrada en el estudio del comportamiento de litio plomo, permeación de tritio, corrosión y compatibilidad de materiales. Para cada uno de los experimentos se propone un esquema experimental, se definen las condiciones necesarias en el módulo y la instrumentación requerida para controlar y diagnosticar las cápsulas experimentales. Para llevar a cabo los experimentos propuestos se propone el LBVM, ubicado en la zona de medio flujo de IFMIF, en su celda caliente, y con capacidad para 16 cápsulas experimentales. Cada cápsula (24-22 mm de diámetro y 80 mm de altura) contendrá la aleación eutéctica LiPb (hasta 50 mm de la altura de la cápsula) en contacto con diferentes muestras de materiales. Ésta irá soportada en el interior de tubos de acero por los que circulará un gas de purga (He), necesario para arrastrar el tritio generado en el eutéctico y permeado a través de las paredes de las cápsulas (continuamente, durante irradiación). Estos tubos, a su vez, se instalarán en una carcasa también de acero que proporcionará soporte y refrigeración tanto a los tubos como a sus cápsulas experimentales interiores. El módulo, en su conjunto, permitirá la extracción de las señales experimentales y el gas de purga. Así, a través de la estación de medida de tritio y el sistema de control, se obtendrán los datos experimentales para su análisis y extracción de conclusiones experimentales. Además del análisis de datos experimentales, algunas de estas señales tendrán una función de seguridad y por tanto jugarán un papel primordial en la operación del módulo. Para el correcto funcionamiento de las cápsulas y poder controlar su temperatura, cada cápsula se equipará con un calentador eléctrico y por tanto el módulo requerirá también ser conectado a la alimentación eléctrica. El diseño del módulo y su lógica de operación se describe en detalle en esta tesis. La justificación técnica de cada una de las partes que componen el módulo se ha realizado con soporte de cálculos de transporte de tritio, termohidráulicos y mecánicos. Una de las principales conclusiones de los cálculos de transporte de tritio es que es perfectamente viable medir el tritio permeado en las cápsulas mediante cámaras de ionización y contadores proporcionales comerciales, con sensibilidades en el orden de 10-9 Bq/m3. Los resultados son aplicables a todos los experimentos, incluso si son cápsulas a bajas temperaturas o si llevan recubrimientos antipermeación. Desde un punto de vista de seguridad, el conocimiento de la cantidad de tritio que está siendo transportada con el gas de purga puede ser usado para detectar de ciertos problemas que puedan estar sucediendo en el módulo como por ejemplo, la rotura de una cápsula. Además, es necesario conocer el balance de tritio de la instalación. Las pérdidas esperadas el refrigerante y la celda caliente de IFMIF se pueden considerar despreciables para condiciones normales de funcionamiento. Los cálculos termohidráulicos se han realizado con el objetivo de optimizar el diseño de las cápsulas experimentales y el LBVM de manera que se pueda cumplir el principal requisito del módulo que es llevar a cabo los experimentos a temperaturas comprendidas entre 300-550ºC. Para ello, se ha dimensionado la refrigeración necesaria del módulo y evaluado la geometría de las cápsulas, tubos experimentales y la zona experimental del contenedor. Como consecuencia de los análisis realizados, se han elegido cápsulas y tubos cilíndricos instalados en compartimentos cilíndricos debido a su buen comportamiento mecánico (las tensiones debidas a la presión de los fluidos se ven reducidas significativamente con una geometría cilíndrica en lugar de prismática) y térmico (uniformidad de temperatura en las paredes de los tubos y cápsulas). Se han obtenido campos de presión, temperatura y velocidad en diferentes zonas críticas del módulo concluyendo que la presente propuesta es factible. Cabe destacar que el uso de códigos fluidodinámicos (e.g. ANSYS-CFX, utilizado en esta tesis) para el diseño de cápsulas experimentales de IFMIF no es directo. La razón de ello es que los modelos de turbulencia tienden a subestimar la temperatura de pared en mini canales de helio sometidos a altos flujos de calor debido al cambio de las propiedades del fluido cerca de la pared. Los diferentes modelos de turbulencia presentes en dicho código han tenido que ser estudiados con detalle y validados con resultados experimentales. El modelo SST (del inglés, “Shear Stress Transport Model”) para turbulencia en transición ha sido identificado como adecuado para simular el comportamiento del helio de refrigeración y la temperatura en las paredes de las cápsulas experimentales. Con la geometría propuesta y los valores principales de refrigeración y purga definidos, se ha analizado el comportamiento mecánico de cada uno de los tubos experimentales que contendrá el módulo. Los resultados de tensiones obtenidos, han sido comparados con los valores máximos recomendados en códigos de diseño estructural como el SDC-IC (del inglés, “Structural Design Criteria for ITER Components”) para así evaluar el grado de protección contra el colapso plástico. La conclusión del estudio muestra que la propuesta es mecánicamente robusta. El LBVM implica el uso de metales líquidos y la generación de tritio además del riesgo asociado a la activación neutrónica. Por ello, se han estudiado los riesgos asociados al uso de metales líquidos y el tritio. Además, se ha incluido una evaluación preliminar de los riesgos radiológicos asociados a la activación de materiales y el calor residual en el módulo después de la irradiación así como un escenario de pérdida de refrigerante. Los riesgos asociados al módulo de naturaleza convencional están asociados al manejo de metales líquidos cuyas reacciones con aire o agua se asocian con emisión de aerosoles y probabilidad de fuego. De entre los riesgos nucleares destacan la generación de gases radiactivos como el tritio u otros radioisótopos volátiles como el Po-210. No se espera que el módulo suponga un impacto medioambiental asociado a posibles escapes. Sin embargo, es necesario un manejo adecuado tanto de las cápsulas experimentales como del módulo contenedor así como de las líneas de purga durante operación. Después de un día de después de la parada, tras un año de irradiación, tendremos una dosis de contacto de 7000 Sv/h en la zona experimental del contenedor, 2300 Sv/h en la cápsula y 25 Sv/h en el LiPb. El uso por lo tanto de manipulación remota está previsto para el manejo del módulo irradiado. Por último, en esta tesis se ha estudiado también las posibilidades existentes para la fabricación del módulo. De entre las técnicas propuestas, destacan la electroerosión, soldaduras por haz de electrones o por soldadura láser. Las bases para el diseño final del LBVM han sido pues establecidas en el marco de este trabajo y han sido incluidas en el diseño intermedio de IFMIF, que será desarrollado en el futuro, como parte del diseño final de la instalación IFMIF. ABSTRACT Nuclear fusion is, today, an alternative energy source to which the international community devotes a great effort. The goal is to generate 10 to 50 times more energy than the input power by means of fusion reactions that occur in deuterium (D) and tritium (T) plasma at two hundred million degrees Celsius. In the future commercial reactors it will be necessary to breed the tritium used as fuel in situ, by the reactor itself. This constitutes a step further from current experimental machines dedicated mainly to the study of the plasma physics. Therefore, tritium, in fusion reactors, will be produced in the so-called breeder blankets whose primary mission is to provide neutron shielding, produce and recover tritium and convert the neutron energy into heat. There are different concepts of breeding blankets that can be separated into two main categories: solids or liquids. The former are based on ceramics containing lithium as Li2O , Li4SiO4 , Li2TiO3 , Li2ZrO3 and Be, used as a neutron multiplier, required to achieve the required amount of tritium. The liquid concepts are based on molten salts or liquid metals as pure Li, LiPb, FLIBE or FLINABE. These blankets use, as neutron multipliers, Be or Pb (in the case of the concepts based on LiPb). Proposed structural materials comprise various options, always with low activation characteristics, as low activation ferritic-martensitic steels, vanadium alloys or even SiCf/SiC. Each concept of breeding blanket has specific challenges that will be studied in the experimental reactor ITER (International Thermonuclear Experimental Reactor). However, ITER cannot answer questions associated to material damage and the effect of neutron radiation in the different breeding blankets functions and performance. As a reference, the first wall of a fusion reactor of 4000 MW will receive about 30 dpa / year (values for Fe-56) , while values expected in ITER would be <10 dpa in its entire lifetime. Consequently, the irradiation effects on candidate materials for fusion reactors will be studied in IFMIF (International Fusion Material Irradiation Facility). This thesis fits in the framework of the bilateral agreement among Europe and Japan which is called “Broader Approach Agreement “(BA) (2007-2017) where Spain plays a key role. These projects, complementary to ITER, are mainly IFMIF and the fusion facility JT-60SA. The purpose of this thesis is the design of an irradiation module to test candidate materials for breeding blankets in IFMIF, the so-called Liquid Breeder Validation Module (LBVM). This proposal is born from the fact that this option was not considered in the conceptual design of the facility. As a first step, in order to study the feasibility of this proposal, neutronic calculations have been performed to estimate irradiation parameters in different materials foreseen for liquid breeding blankets. Various functional materials were considered: Fe (base of structural materials), SiC (candidate material for flow channel inserts, SiO2 (candidate for antipermeation coatings), CaO (candidate for insulating coatings), Al2O3 (candidate for antipermeation and insulating coatings) and AlN (candidate for insulation coating material). For each material, the most significant irradiation parameters have been calculated (dpa, H/dpa and He/dpa) in different positions of IFMIF. These values were compared to those expected in the first wall and breeding zone of a fusion reactor. For this exercise, a HCLL (Helium Cooled Lithium Lead) type was selected as it is one of the most promising options. In addition, estimated values were also compared with those obtained in a fast fission reactor since most of existing irradiations have been made in these installations. The main conclusion of this study is that the medium flux area of IFMIF offers a good irradiation environment to irradiate functional materials for liquid breeding blankets. The obtained ratios of H/dpa and He/dpa are very similar to those expected in the most irradiated areas of a fusion reactor. Moreover, with the aim of bringing the values further close, the use of a W moderator is proposed to be used only in some experimental campaigns (as obviously, the total amount of dpa decreases). The values of ratios obtained for a fission reactor, much lower than in a fusion reactor, reinforce the need of LBVM for IFMIF. Having demonstrated the suitability of IFMIF to irradiate functional materials for liquid breeding blankets, and an analysis of the main problems associated to each type of liquid breeding blanket, also presented in this thesis, three different experiments are proposed as basis for the design of the LBVM. These experiments are dedicated to the needs of a blanket HCLL type although the applicability of the module for other blankets is also discussed. Therefore, the experimental capability of the module is focused on the study of the behavior of the eutectic alloy LiPb, tritium permeation, corrosion and material compatibility. For each of the experiments proposed an experimental scheme is given explaining the different module conditions and defining the required instrumentation to control and monitor the experimental capsules. In order to carry out the proposed experiments, the LBVM is proposed, located in the medium flux area of the IFMIF hot cell, with capability of up to 16 experimental capsules. Each capsule (24-22 mm of diameter, 80 mm high) will contain the eutectic allow LiPb (up to 50 mm of capsule high) in contact with different material specimens. They will be supported inside rigs or steel pipes. Helium will be used as purge gas, to sweep the tritium generated in the eutectic and permeated through the capsule walls (continuously, during irradiation). These tubes, will be installed in a steel container providing support and cooling for the tubes and hence the inner experimental capsules. The experimental data will consist of on line monitoring signals and the analysis of purge gas by the tritium measurement station. In addition to the experimental signals, the module will produce signals having a safety function and therefore playing a major role in the operation of the module. For an adequate operation of the capsules and to control its temperature, each capsule will be equipped with an electrical heater so the module will to be connected to an electrical power supply. The technical justification behind the dimensioning of each of these parts forming the module is presented supported by tritium transport calculations, thermalhydraulic and structural analysis. One of the main conclusions of the tritium transport calculations is that the measure of the permeated tritium is perfectly achievable by commercial ionization chambers and proportional counters with sensitivity of 10-9 Bq/m3. The results are applicable to all experiments, even to low temperature capsules or to the ones using antipermeation coatings. From a safety point of view, the knowledge of the amount of tritium being swept by the purge gas is a clear indicator of certain problems that may be occurring in the module such a capsule rupture. In addition, the tritium balance in the installation should be known. Losses of purge gas permeated into the refrigerant and the hot cell itself through the container have been assessed concluding that they are negligible for normal operation. Thermal hydraulic calculations were performed in order to optimize the design of experimental capsules and LBVM to fulfill one of the main requirements of the module: to perform experiments at uniform temperatures between 300-550ºC. The necessary cooling of the module and the geometry of the capsules, rigs and testing area of the container were dimensioned. As a result of the analyses, cylindrical capsules and rigs in cylindrical compartments were selected because of their good mechanical behavior (stresses due to fluid pressure are reduced significantly with a cylindrical shape rather than prismatic) and thermal (temperature uniformity in the walls of the tubes and capsules). Fields of pressure, temperature and velocity in different critical areas of the module were obtained concluding that the proposal is feasible. It is important to mention that the use of fluid dynamic codes as ANSYS-CFX (used in this thesis) for designing experimental capsules for IFMIF is not direct. The reason for this is that, under strongly heated helium mini channels, turbulence models tend to underestimate the wall temperature because of the change of helium properties near the wall. Therefore, the different code turbulence models had to be studied in detail and validated against experimental results. ANSYS-CFX SST (Shear Stress Transport Model) for transitional turbulence model has been identified among many others as the suitable one for modeling the cooling helium and the temperature on the walls of experimental capsules. Once the geometry and the main purge and cooling parameters have been defined, the mechanical behavior of each experimental tube or rig including capsules is analyzed. Resulting stresses are compared with the maximum values recommended by applicable structural design codes such as the SDC- IC (Structural Design Criteria for ITER Components) in order to assess the degree of protection against plastic collapse. The conclusion shows that the proposal is mechanically robust. The LBVM involves the use of liquid metals, tritium and the risk associated with neutron activation. The risks related with the handling of liquid metals and tritium are studied in this thesis. In addition, the radiological risks associated with the activation of materials in the module and the residual heat after irradiation are evaluated, including a scenario of loss of coolant. Among the identified conventional risks associated with the module highlights the handling of liquid metals which reactions with water or air are accompanied by the emission of aerosols and fire probability. Regarding the nuclear risks, the generation of radioactive gases such as tritium or volatile radioisotopes such as Po-210 is the main hazard to be considered. An environmental impact associated to possible releases is not expected. Nevertheless, an appropriate handling of capsules, experimental tubes, and container including purge lines is required. After one day after shutdown and one year of irradiation, the experimental area of the module will present a contact dose rate of about 7000 Sv/h, 2300 Sv/h in the experimental capsules and 25 Sv/h in the LiPb. Therefore, the use of remote handling is envisaged for the irradiated module. Finally, the different possibilities for the module manufacturing have been studied. Among the proposed techniques highlights the electro discharge machining, brazing, electron beam welding or laser welding. The bases for the final design of the LBVM have been included in the framework of the this work and included in the intermediate design report of IFMIF which will be developed in future, as part of the IFMIF facility final design.

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

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.

The effect of flow arrangement on the pressure drop of plate heat exchangers

For the optimal design of plate heat exchangers (PHEs), an accurate thermal-hydraulic model that takes into account the effect of the flow arrangement on the heat load and pressure drop is necessary. In the present study, the effect of the flow arrangement on the pressure drop of a PHE is investigated. Thirty two different arrangements were experimentally tested using a laboratory scale PHE with flat plates. The experimental data was used for (a) determination of an empirical correlation for the effect of the number of passes and number of flow channels per pass on the pressure drop; (b) validation of a friction factor model through parameter estimation; and (c) comparison with the simulation results obtained with a CFD (computational fluid dynamics) model of the PHE. All three approaches resulted in a good agreement between experimental and predicted values of pressure drop. Moreover, the CFD model is used for evaluating the flow maldistribution in a PHE with two channels Per Pass. (c) 2008 Elsevier Ltd. All rights reserved.

Quantifying rates of landscape evolution and tectonic processes by thermochronology and numerical modeling of crustal heat transport using PECUBE

PECUBE is a three-dimensional thermal-kinematic code capable of solving the heat production-diffusion-advection equation under a temporally varying surface boundary condition. It was initially developed to assess the effects of time-varying surface topography (relief) on low-temperature thermochronological datasets. Thermochronometric ages are predicted by tracking the time-temperature histories of rock-particles ending up at the surface and by combining these with various age-prediction models. In the decade since its inception, the PECUBE code has been under continuous development as its use became wider and addressed different tectonic-geomorphic problems. This paper describes several major recent improvements in the code, including its integration with an inverse-modeling package based on the Neighborhood Algorithm, the incorporation of fault-controlled kinematics, several different ways to address topographic and drainage change through time, the ability to predict subsurface (tunnel or borehole) data, prediction of detrital thermochronology data and a method to compare these with observations, and the coupling with landscape-evolution (or surface-process) models. Each new development is described together with one or several applications, so that the reader and potential user can clearly assess and make use of the capabilities of PECUBE. We end with describing some developments that are currently underway or should take place in the foreseeable future. (C) 2012 Elsevier B.V. All rights reserved.

Heat Transfer Analysis of the EPR Core Catcher Test Facility Volley

Uusi EPR-reaktorikonsepti on suunniteltu selviytymään tapauksista, joissa reaktorinsydän sulaa ja sula puhkaisee paineastian. Suojarakennuksen sisälle on suunniteltu alue, jolle sula passiivisesti kerätään, pidätetään ja jäähdytetään. Alueelle laaditaan valurautaelementeistä ns.sydänsieppari, joka tulvitetaan vedellä. Sydänsulan tuottama jälkilämpö siirtyyveteen, mistä se poistetaan suojarakennuksen jälkilämmönpoistojärjestelmän kautta. Suuri osa lämmöstä poistuu sydänsulasta sen yläpuolella olevaan veteen, mutta lämmönsiirron tehostamiseksi myös sydänsiepparin alapuolelle on sijoitettu vedellä täytettävät jäähdytyskanavat. Jotta sydänsiepparin toiminta voitaisiin todentaa, on Lappeenrannan Teknillisellä Yliopistolla rakennettu Volley-koelaitteisto tätä tarkoitusta varten. Koelaitteisto koostuu kahdesta täysimittaisesta valuraudasta tehdystä jäähdytyskanavasta. Sydänsulan tuottamaa jälkilämpöä simuloidaan koelaitteistossa sähkövastuksilla. Tässä työssä kuvataan simulaatioiden suorittaminen ja vertaillaan saatuja arvoja mittaustuloksiin. Työ keskittyy sydänsiepparista jäähdytyskanaviin tapahtuvan lämmönsiirron teoriaan jamekanismeihin. Työssä esitetään kolme erilaista korrelaatiota lämmönsiirtokertoimille allaskiehumisen tapauksessa. Nämä korrelaatiot soveltuvat erityisesti tapauksiin, joissa vain muutamia mittausparametreja on tiedossa. Työn toinen osa onVolley 04 -kokeiden simulointi. Ensin käytettyä simulointitapaa on kelpoistettuvertaamalla tuloksia Volley 04 ja 05 -kokeisiin, joissa koetta voitiin jatkaa tasapainotilaan ja joissa jäähdytteen käyttäytyminen jäähdytyskanavassa on tallennettu myös videokameralla. Näiden simulaatioiden tulokset ovat hyvin samanlaisiakuin mittaustulokset. Korkeammilla lämmitystehoilla kokeissa esiintyi vesi-iskuja, jotka rikkoivat videoinnin mahdollistavia ikkunoita. Tämän johdosta osassa Volley 04 -kokeita ikkunat peitettiin metallilevyillä. Joitakin kokeita jouduttiin keskeyttämään laitteiston suurten lämpöjännitysten johdosta. Tällaisten testien simulaatiot eivät ole yksinkertaisia suorittaa. Veden pinnan korkeudesta ei ole visuaalista havaintoa. Myöskään jäähdytteen tasapainotilanlämpötiloista ei ole tarkkaa tietoa, mutta joitakin oletuksia voidaan tehdä samoilla parametreilla tehtyjen Volley 05 -kokeiden perusteella. Mittaustulokset Volley 04 ja 05 -kokeista, jotka on videoitu ja voitu ajaa tasapainotilaan saakka, antoivat simulaatioiden kanssa hyvin samankaltaisia lämpötilojen arvoja. Keskeytettyjen kokeiden ekstrapolointi tasapainotilaan ei onnistunut kovin hyvin. Kokeet jouduttiin keskeyttämään niin paljon ennen termohydraulista tasapainoa, ettei tasapainotilan reunaehtoja voitu ennustaa. Videonauhoituksen puuttuessa ei veden pinnan korkeudesta saatu lisätietoa. Tuloksista voidaan lähinnä esittää arvioita siitä, mitä suuruusluokkaa mittapisteiden lämpötilat tulevat olemaan. Nämä lämpötilat ovat kuitenkin selvästi alle sydänsiepparissa käytettävän valuraudan sulamislämpötilan. Joten simulaatioiden perusteella voidaan sanoa, etteivät jäähdytyskanavien rakenteet sula, mikäli niissä on pienikin jäähdytevirtaus, eikä useampia kuin muutama vierekkäinen kanava ole täysin kuivana.

USER'S GUIDE FOR ADVANCED PROCESS SIMULATION SOFTWARE APROS 5

Tässä diplomityössä tehtiin käyttäjän opas kehittyneelle prosessisimulointiohjelmistolle APROS 5. Opas on osa VTT Energialle tehtävää APROS 5 käyttäjän koulutuspakettia, joka julkaistaan myöhemmin CD-ROM -muotoisena. Prosessisimulointiohjelmistoa AAPROS 5 voidaan käyttää termohydraulisten prosessien, automaatiopiirien ja sähköjärjestelmien mallinnuksessa. Ohjelma sisältää myös neutroniikkamallin ydinreaktorin käyttäytymisen mallintamiseksi. APROS:in aikaisemmilla UNIX-ympäristössä toimivilla versioilla on toteutettu useita ydinvoimalaitosten turvallisuustutkimukseen liittyviä analyysejä ja sekä ydinvoimalaitosten että konventionaalisten voimalaitosten koulutussimulaattoreita. APROS 5 toimii Windows NT -ympäristössä ja on oleellisesti erilainen käyttää kuin aikaisemmat versiot. Tämän myötä syntyi tarve uudelle käyttäjän oppaalle. Käyttäjän oppaassa esitetään APROS 5:n tärkeimmät toiminnot, mallinnuksen periaatteet ja termohydraulisten ja neutroniikan ratkaisumallit. Lisäksi oppaassa esitetään esimerkki, jossa mallinnetaan yksinkertaistettu VVER-440 -tyyppisen ydinvoimalaitoksen primääripiiri. Yksityiskohtaisempaa tietoa ohjelmistosta on saatavilla APROS 5 -dokumentaatiosta.