Le travail présenté dans ce mémoire est basé sur l’étude et le développement d’une chaine analogique de traitement du signal destinée à la détection de particules issues d’un réacteur nucléaire. Le détecteur de radiations relatif à cette chaine, étant placé au cœur du réacteur (hautes températures et radiations), deux scénarii d’intégration de la chaine analogique en environnement sévère ont été étudiés. Tandis que le premier scénario consiste à placer le préamplificateur et l’amplificateur de mise en forme loin du détecteur, le deuxième est basé sur l’intégration de ces blocs analogiques en environnement contraint. Bien que le premier scénario ait démontré de bonnes performances en termes de linéarité et sensibilité, l’amélioration de la résolution digitale fournie par le système nous a mené à investiguer le placement de la chaine tout près du détecteur. Plusieurs études, basées sur l’évaluation de la fiabilité de la structure vis-à-vis des contraintes de l’environnement sévère ont été donc élaborées. De plus une solution innovante permettant une meilleure caractérisation de la particule incidente a été présentée et détaillée. / The work presented in this thesis is based on the study and development of an analog processing chain dedicated to the detection of particles delivred from a nuclear reactor. The radiation detector related to this chain, being placed at the heart of the reactor (high temperatures and radiation flux), two scenraii related to the integration of the analog chain in harsh environment were studied. While the first scenario consists on placing the preamplifier and the shaping amplifier away from the detector, the second is based on the integration of these analog blocks in harsh environment. Although the first scenario has proven good performances in terms of linearity and sensitivity, the improvement of the digital resolution provided by the system leaded us to investigate the placement of the chain close to the detector. Several studies based on the evaluation of the structure reliability under harsh environment constraints were thus elaborated. In addition, an innovative solution providing better characterization of the incident particle was presented and detailed.

<p>I chose to study how the Swedish Air force carries out their work with setting goals for their work to improve flight safety and compare this with how the Swedish nuclear power plant Forsmark performs their work with improving nuclear safety. The study showed that both organisations have a lot to improve on in their ways of how to work. Especially when it comes to engaging the employees in the process, both in letting the employees be a part of coming up with areas to work on and in letting the employees evaluate which progress that has been made. It also showed that both organisations need to improve their methods for evaluating the work that has been done. This is especially important for the Air Force which evaluates on many levels and locations and did not have a set standard for how to conduct the evaluation. Although, Forsmark also needed to improve but on how to conduct the evaluation on more than one level in the organisation, allowing more people into the evaluation process and maybe letting someone not part of the process be a part of the evaluation to see things from a different perspective. Also, Forsmark had goals set which did not have a set time for when to evaluate them, which is another criteria for success.</p>

I chose to study how the Swedish Air force carries out their work with setting goals for their work to improve flight safety and compare this with how the Swedish nuclear power plant Forsmark performs their work with improving nuclear safety. The study showed that both organisations have a lot to improve on in their ways of how to work. Especially when it comes to engaging the employees in the process, both in letting the employees be a part of coming up with areas to work on and in letting the employees evaluate which progress that has been made. It also showed that both organisations need to improve their methods for evaluating the work that has been done. This is especially important for the Air Force which evaluates on many levels and locations and did not have a set standard for how to conduct the evaluation. Although, Forsmark also needed to improve but on how to conduct the evaluation on more than one level in the organisation, allowing more people into the evaluation process and maybe letting someone not part of the process be a part of the evaluation to see things from a different perspective. Also, Forsmark had goals set which did not have a set time for when to evaluate them, which is another criteria for success.

The system code TRACE is the latest development of the U.S. Nuclear Regulatory Commission (US NRC). TRACE, developed for the analysis of operational conditions, transients and accidents of light water reactors (LWR), is a best-estimate code with two fluid, six equation models for mass, energy, and momentum conservation, and related closure models. Since TRACE is mainly applied to LWR specific issues, the validation process related to innovative nuclear systems (liquid metal cooled systems, systems operated with supercritical water, etc.) is very limited, almost not existing. In this work, essential contribution to the validation of TRACE related to lead and lead alloy cooled systems as well as systems operated with supercritical water is provided in a consistent and corporate way. In a first step, model discrepancies of the TRACE source code were removed. This inconsistencies caused the wrong prediction of the thermo physical properties of supercritical water and lead bismuth eutectic, and hence the incorrect prediction of heat transfer relevant characteristic numbers like Reynolds or Prandtl number. In addition to the correction of the models to predict these quantities, models describing the thermo physical properties of lead and Diphyl THT (synthetic heat transfer medium) were implemented. Several experiments and numerical benchmarks were used to validate the modified TRACE version. These experiments, mainly focused on wall-to-fluid heat transfer, revealed that not only the thermo physical properties are afflicted with inconsistencies but also the heat transfer models. The models for the heat transfer to liquid metals were enhanced in a way that the code can now distinguish between pipe and bundle flow by using the right correlation. The heat transfer to supercritical water was not existing in TRACE up to now. Completely new routines were implemented to overcome that issue. The comparison of the calculations to the experiments showed, on one hand, the necessity of these changes and, on the other hand, the success of the new implemented routines and functions. The predictions using the modified TRACE version were close to the experimental data. After validating the modified TRACE version, two design studies related to the Generation IV International Forum (GIF) were investigated. In the first one, a core of a lead-cooled fast reactor (LFR) was analyzed. To include the interaction between the thermal hydraulic and the neutron kinetic due to temperature and density changes, the TRACE code was coupled to the program system ERANOS2.1. The results gained with that coupled system are in accordance with theory and helped to identify sub-assemblies with the highest loads concerning fuel and cladding temperature. The second design which was investigated was the High Performance Light Water Reactor (HPLWR). Since the design of the HPLWR is not finalized, optimization of vital parameters (power, mass flow rate, etc.) are still ongoing. Since most of the parameters are affecting each other, an uncertainty and sensitivity analysis was performed. The uncertainty analysis showed the upper and lower boundaries of selected parameters, which are of importance from the safety point of view (e.g., fuel and cladding temperature, moderator temperature). The sensitivity study identified the most relevant parameters and their influence on the whole system.

Magistro baigiamajame darbe išanalizuotos ir įvertintos branduolinio tipo ekstremaliųjų situacijų prevencijos ir parengties priemonės. Pirmoje darbo dalyje teoriniu aspektu tiriama branduolinės saugos bei ekstremaliųjų situacijų valdymo ciklo prevencijos ir parengties fazių samprata. Taip pat išanalizuojamas technologinio pobūdžio rizikos suvokimas, jos valdymo ypatumai bei branduolinių tipo ekstremaliųjų situacijų klasifikavimas pagal tarptautinę branduolinių ir radiologinių įvykių skalę. Antroje dalyje nagrinėjos branduolinio tipo ekstremaliųjų situacijų prevencijos ir parengties priemonės bei trumpai apžvelgiama Lietuvos situacija, įvertinta pagal tas priemones, atskleidžiamos spragos teoriniame lygmenyje. Trečioje dalyje aptariama tyrimo, atliekamo trianguliacijos principu metodika bei analizuojami žvalgybis,kokybinis ir kiekybinis tyrimai. Kokybiniu tyrimu atskleidžiama Černobylio avarijos prevencijos ir parengties priemonių situacija, kiekybiniu – visuomenės nuomonė apie branduolinę saugą Lietuvoje, žvalgybiniu – specialistų nuomonė dėl radiacinės saugos. / Master's thesis analyzed and evaluated nuclear emergency prevention and preparedness measures. The first part is an overview of nuclear safety, prevention and preparedness phases conception of emergency management cycle. It is also examined the nature of technological risk perception and its operating features and the type of nuclear emergency classification according to the INES scale. The second part analyzed the nuclear emergency prevention and preparedness measures and it is given a brief overview of the situation in Lithuania, assessed according to these measures. The third section discussed the research carried out by triangulation method: qualitative and quantitative. Qualitative research revealed the Chernobyl accident prevention and preparedness measures and identified gaps in emergency management process. Quantitative research analyzed public opinion on nuclear safety in Lithuania. Also experts in radiological sphere have given their opinion about radiological safety measures.

Control room operating crews are a crucial component in maintaining the safety of nuclear power plants. The primary support to operators during disturbances or emergencies are numerous emergency operating procedures. Further support is provided by reoccurring crew training in which the crews practice on handling anticipated disturbances in full-scale simulators. Due to the complexity of nuclear power plants and the small number of previous accidents to learn from, every possible accident scenario cannot be covered by the procedures and hence not trained on in the simulator. This raises the question of how operators can be prepared and able to cope with unexpected events by other means.   This thesis investigates the possibilities of operating crews to act flexibly in situations where stable responses in the form of prescribed actions sequences from procedures cannot be applied. The study is based on the safety research paradigm of resilience engineering and the four cornerstones of resilience; learning, monitoring, anticipating, and responding (Hollnagel, 2011). The meaning and applicability of the resilience cornerstones were examined by interviewing a domain expert at the time employed by the OECD Halden Reactor Project. Subsequently, eight semi-structured interviews with operator training personnel at a Swedish nuclear power plant provided the main data of this study.   This study shows that the resilience cornerstones were applicable to the work of nuclear power plant crews during emergency operations. In addition, the study provides findings regarding which artefacts (e.g. procedures) or crew characteristics (e.g. ways of communicating) support the cornerstone functions. The base thesis is that procedures always shall be used, but in situations where an operator perceives that no procedure is applicable, the crew have an opportunity to discuss the problem to come up with some other solution, i.e. act flexibly. Some trainers argued that the room for flexibility is there when needed, but it is not as certain how much flexibility and what kind of flexibility the operators are given. However, it does not seem like the flexibility, or lack of flexibility, given to operators is in itself the most problematic issue in the preparation of crews for unexpected events. Instead, this study identified several other problems of training and everyday work that could negatively affect crews’ capability to handle unexpected events. On the other hand, the trainers highlighted communication and teamwork to be important when the unexpected strikes and that much focus have been shifted towards such issues in training. Hence this can be claimed to be an important contribution given by the training today in successfully handling unforeseen events.

Ce travail de thèse porte sur l’étude des pertes de pression pour des écoulements monophasiques et diphasiques inertiels au travers de milieux poreux. Son objectif est d’aider à la compréhension et à la modélisation des transferts de quantité de mouvement à l’intérieur de lits de particules, en lien avec la problématique de la gestion d’un accident grave dans un réacteur nucléaire. En effet, lors d’un tel accident, la dégradation du coeur du réacteur peut amener celui-ci à s’effondrer pour former un lit de débris, que l’on peut assimiler à un milieu poreux à haute température et dégageant de la chaleur. Ce travail de thèse s’inscrit dans un projet de recherche en sûreté nucléaire visant à prédire la refroidissabilité d’un lit de débris par injection d’eau, ou « renoyage ». Une étude expérimentale des pertes de pression pour des écoulements monodimensionnels monophasiques et diphasiques à froid est proposée dans des situations représentatives du cas réacteur, en termes de granulométrie, de formes de particules et de vitesses d’écoulement. Les expériences réalisées apportent un complément important aux données existantes, en permettant notamment d’explorer les domaines d’écoulements diphasiques avec nombres de Reynolds liquides non nuls, tout en mesurant le taux de vide, ce qui est essentiel pour une modélisation. Des modèles prédictifs pour les pertes de pression à l’intérieur d’écoulements monophasiques et diphasiques au travers de lits de particules sont établis à partir des structures d’équations obtenues par une prise de moyenne volumique des équations de conservation locales. L’observation des écoulements monophasiques montrent que des lois de type Darcy-Forchheimer avec une correction quadratique en vitesse de filtration sont à même de prédire les pertes de pression avec une précision supérieure à 10%. Une étude numérique a montré que ce résultat est applicable pour un lit désordonné de particules peu rugueuses. L’étude des écoulements diphasiques montre qu’une structure d’équations de type Darcy-Forchheimer généralisée, incluant des termes supplémentaires pour prendre en compte les effets inertiels et les frottements interfaciaux, permet de reproduire le comportement des pertes de pression dans cette situation. Un nouveau modèle est proposé, et comparé aux données expérimentales et aux modèles utilisés dans les codes de simulation des accidents graves. / This work deals with single and two-phase flow pressure losses in porous media. The aim is to improve understanding and modeling of momentum transfer inside particle beds, in relation with nuclear safety issues concerning the reflooding of debris beds during severe nuclear accidents. Indeed, the degradation of the core during such accidents can lead to the collapse of the fuel assemblies, and to the formation of a debris bed, which can be described as a hot porous medium. This thesis is included in a nuclear safety research project on coolability of debris beds during reflooding sequences. An experimental study of single and two-phase cold-flow pressure losses in particle beds is proposed. The geometrical characteristics of the debris and the hydrodynamic conditions are representative of the real case, in terms of granulometry, particle shapes, and flow velocities. The new data constitute an important contribution. In particular, they contain pressure losses and void fraction measurements in two-phase air-water flows with non-zero liquid Reynolds numbers, which did not exist before. Predictive models for pressure losses in single and two-phase flow through particle beds have been established from experimental data. Their structures are based on macroscopic equations obtained from the volume averaging of local conservation equations. Single-phase flow pressure losses can be described by a Darcy-Forchheimer law with a quadratic correction, in terms of filtration velocity, with a better-than-10 % precision. Numerical study of single-phase flows through porous media shows that this correlation is valid for disordered smooth particle beds. Twophase flow pressure losses are described using a generalized Darcy-Forchheimer structure, involving inertial and cross flow terms. A new model is proposed and compared to the experimental data and to the usual models used in severe accident simulation codes.

This thesis is focused on the design of passive cooling system for a nuclear reactor VVER- 1000.This type of reactor is located in the Czech Republic in the location of Nuclear power plant Temelín. The thesis states an overview of the different cooling systems of nuclear power plants. The thesis is focused on passive safety system especially on passive cooling system, so there was done an overview of currently used passive safety system. In the work is discussed nuclear safety and the maximum projected accident of VVER-1000, which is called LOCA accident. In the design part of the thesis was done thermal calculation of heat exchangers. Exchangers are designed as condensers with a natural flow, where cooling of system is provided by outside airflow in case an accident. The results are evaluated at the end of the thesis.

The Institute of Safety Research (ISR) is one of the six Research Institutes of Forschungszentrum Dresden-Rossendorf e.V. (FZD), which is a member institution of the Wissenschaftsgemeinschaft Gottfried Wilhelm Leibniz (Leibnizgemeinschaft). Together with the Institutes of Radiochemistry and Radiation Physics, ISR implements the research programme „Nuclear Safety Research“, which is one of the three scientific programmes of FZD. The programme includes two main topics, i. e. “Safety Research for Radioactive Waste Disposal” and “Safety Research for Nuclear Reactors”.

The Institute of Radiochemistry (IRC) is one of the seven institutes of the Helmholtz-Zentrum Dresden-Rossendorf (HZDR). The research activities are fully integrated into the “Nuclear Safety Research Program” of the Helmholtz Association and focused on the topic “Safety of Nuclear Waste Disposal”. The research objectives are to generate better process understanding and data for the long-term safety analysis of a nuclear waste disposal in the deep geological underground. A better knowledge about the dominating processes essential for radionuclide (actinide) mobilization and immobilization on the molecular level is needed for the assessment of the macroscopic processes which determine the transport and distribution of radioactivity in the environment. Special emphasis is put on the biological mediated transport of long-lived radionuclides in the geosphere and their interaction with different biosystems like biota and human organism for a better calculation of environmental and health risks. Advanced knowledge is needed for description of the processes dominating at the interfaces between geo- and bio-systems related to the distribution of long-lived radionuclides in various bio-systems along the food chain.