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Concept# Réacteur nucléaire

Résumé

Un réacteur nucléaire est un ensemble de dispositifs comprenant du combustible nucléaire, qui constitue le « cœur » du réacteur, dans lequel une réaction en chaîne peut être initiée et contrôlée par des agents humains ou par des systèmes automatiques, suivant des protocoles et au moyen de dispositifs propres à la fission nucléaire. La chaleur ainsi produite est ensuite évacuée et éventuellement convertie en énergie électrique.
Principe
Dans le cœur d'un réacteur nucléaire, sous l'effet d'une collision avec un neutron, le noyau atomique de certains gros atomes, dits fissiles, peut se casser en deux (il fissionne), en libérant une grande quantité de chaleur et en produisant deux ou trois neutrons, chacun étant capable de produire une nouvelle fission lors d'une collision avec un autre atome (créant potentiellement une réaction en chaîne).
La matière fissile qui constitue le cœur des réacteurs est de l’uranium enrichi ou du plutonium encapsulé dans des crayons regroupés en

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PHYS-443: Physics of nuclear reactors

In this course, one acquires an understanding of the basic neutronics interactions occurring in a nuclear fission reactor as well as the conditions for establishing and controlling a nuclear chain reaction.

ME-464: Introduction to nuclear engineering

This course is intended to understand the engineering design of nuclear power plants using the basic principles of reactor physics, fluid flow and heat transfer. This course includes the following: Reactor designs, Thermal analysis of nuclear fuel, Nuclear safety and Reactor dynamics

ChE-403: Heterogenous reaction engineering

The theoretical background and practical aspects of heterogeneous reactions including the basic knowledge of heterogeneous catalysis are introduced. The fundamentals are given to allow for the use of chemical reactors to study reaction kinetics and test various mechanistic assumptions.

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Uranium

L’uranium est l'élément chimique de numéro atomique 92, de symbole U. Il fait partie de la famille des actinides.
L'uranium est le naturel le plus abondant dans la croûte terrestre, son abondance

Fission nucléaire

redresse=1.67|vignette|Schéma animé (simplifié) d'une fission nucléaire. Un atome (énorme rond rouge) est percuté par un neutron (point bleu). Celui-ci se scinde en deux atomes. La réaction émet d'aut

Plutonium

Le plutonium est l'élément chimique de symbole Pu et de numéro atomique 94. C'est un métal radioactif transuranien de la famille des actinides. Il se présente sous la forme d'un solide cristallisé d

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Nuclear reactors are inherently stochastic systems, in which neutronic and thermal-hydraulic parameters fluctuate continuously even during steady-state conditions. In addition, structural components vibrate due to the coolant hydraulic forces. This stochasticity is the cause of the neutron population fluctuating behavior, a phenomenon referred as neutron noise. The neutron noise is monitored over the reactor lifetime, providing valuable knowledge of the core behavior. More importantly, the neutron noise monitoring is used for the early detection of reactor anomalies. The neutron noise phenomenon is an intensively-studied research topic, leading the development of noise surveillance methods, signal processing techniques and analytical solvers. Nevertheless, an unexpected neutron noise level increase trend, observed in the last decade in several reactors causing undesirable costly operational consequences, triggered an increasing interest. The new observations revealed the need for an improved neutron noise behavior understanding. In this context, the current doctoral research intents to systematically analyze the neutron noise phenomenon by developing innovative noise modelling methodologies. First, main attention is given to the neutron noise characteristics identification of the Swiss Gösgen nuclear reactor (KKG). KKG plant data are analyzed in the time and frequency domains using traditional signal processing techniques, and key aspects of the KKG noise phenomenology are revealed, allowing the better characterization of the noise sources affecting its dynamic behavior. Then, a neutron noise modelling methodology is developed, utilizing advanced neutronic solvers. These codes are used to systematically model key noise sources (i.e. fuel assembly vibration, and inlet coolant temperature and flow fluctuations), and to study their impact on the neutron noise. For the first time, the fuel assembly vibration model in the utilized codes is systematically studied and is qualified at a 3D full core level. This detailed work, demonstrate the capabilities and the robustness of the newly developed PSI neutron noise methodology to successfully model key noise sources and to reproduce neutron noise phenomena. In addition, a comparative study between the simulated results and the KKG measured data revealed that, the stochastic fluctuation of the inlet coolant temperature in combination with the fuel assembly vibration have a primary role on the KKG neutron noise behavior. Most importantly, it is observed that, the neutron noise increase trend observed in KKG could be explained, at some extent, by the introduction in the core of a newer fuel design which is more susceptible in lateral vibrations. Last, a new in-house methodology is established serving as a supportive diagnostic tool for the detailed identification of signals connectivity patterns. To this aim, the PSI connectivity analysis methodology is based on the causality analysis principles to indicate the cause-and-effect interactions between reactor signals. The nuclear core is analyzed utilizing the most prominent causality analysis techniques in the frequency domain (i.e. rPDC and DTF). The current research is the first application of causality analysis techniques on nuclear reactor data. The application of the developed methodology in measured and simulated datasets showed that, it can successfully indicate the causal interconnections and the perturbation root-cause.

Understanding the time-dependent behaviour of a nuclear reactor following an intended or unintended change of the reactor conditions is of crucial importance to the safe operation of nuclear reactors. Safety evaluations of nuclear reactors involve the analysis of normal and abnormal reactor operation states with adequate and validated simulation tools.
There is a growing interest in supplementing the results of best-estimate reactor calculations with uncertainties. A major source of uncertainty in reactor calculations is the nuclear data. The propagation of uncertainties allows to study their impact on response values of reactor calculations and to establish safety margins in nuclear reactor safety evaluations.
For the simulation of reactor transients, the two-step approach consisting of the generation of group constants in infinite lattice geometry and the full-core calculation using a neutron-kinetic/thermal-hydraulic code system has been state-of-the-art.
Nowadays, Monte Carlo codes allow for the determination of group constants based on a three-dimensional full-core model. Thus, effects from neighbouring fuel assemblies can be considered.
In this thesis, the applicability of the neutron-kinetic code DYN3D coupled with the thermal-hydraulic system code ATHLET for the simulation of transients in small reactor cores was assessed. Experimental reactivity accident tests carried out at the SPERT III research reactor were simulated. Appropriate group constants were generated with the Monte Carlo code Serpent. The overall good agreement between the calculated and the experimental results prove the applicability of the applied codes for the analysis of such reactor transients.
Nuclear data uncertainties were propagated through the simulations of two control rod withdrawal transients in a mini-core model of a pressurised water reactor. The random sampling-based method XSUSA in combination with the SCALE code package was used to determine varied two-group constants.
Based on these two-group constants, 1,000 DYN3D-ATHLET calculations were performed and statistically analysed to obtain uncertainty information for the reactor power and the reactivity.
Since the assumption of normality could not be maintained for the reactor power, distribution-free Wilks tolerance limits were applied.
By sensitivity analyses, the number of delayed neutrons per fission of U-235 and the scattering cross sections of U-238 were identified as the major contributors to the reactor power uncertainty.
For a potential improvement of transient simulations, an approach was developed that was comprised of the determination of group constants using a Serpent full-core model at selected times during a transient simulation and the updating of the original infinite lattice group constants with these full-core group constants. Various challenges and possible solutions were identified. The applicability of the group constants obtained by a full-core model was assessed by their application in full-core diffusion calculations.

Noise analysis applied to nuclear reactor physics is a powerful tool to investigate a reactor's kinetic parameters, and more generally underlying physical processes determining core behavior. The kinetic parameters are the coefficients of simplified time-dependent neutron population equations, the so-called point kinetics equations. Experimentally determined kinetic parameters aid to validate codes and to potentially evaluate nuclear data. This thesis focuses on improvements to the kinetic parameter and uncertainty quantification, the experimental techniques, and the direct simulation of noise experiments. In particular, the notion of spatial effects, i.e. effects that render point kinetics assumptions inaccurate for noise measurements, was investigated. This was achieved by drawing on experiments conducted in the zero power reactor CROCUS.
New detection instrumentation for noise experiments was developed for CROCUS, namely current mode amplifiers for neutron detectors to allow for higher detection efficiency and thus precision; and a scintillation-based gamma detection array, called LEAF, to enable the study of gamma noise. A set of reference neutron and gamma noise experiments were conducted to exemplify an improved method of parameter and uncertainty estimation based on bootstrapping. It lead to a full uncertainty budget showing a relative uncertainty of 3.6% and 1.3% on the prompt decay constant with neutron and gamma noise, respectively. Moreover, reporting of full distributions of kinetic parameters was shown to be required in order to provide an accurate representation of the experimental results. Gamma noise was shown to be superior in terms of precision by a factor of at least two compared to neutron noise when determining the prompt decay constant.
To study spatial effects and the spatial extent of the noise field of CROCUS, a set of experiments with varying detector locations, reactivity, and particle type were conducted. The neutron noise field study showed that measurements are limited to the immediate proximity of the core, and that CROCUS is reliably modeled as a point kinetic reactor for static configurations. Nonetheless, a systematic trend was shown when comparing to code predictions, pointing to a weak spatial effect biasing neutron noise measurements at a distance. The observable gamma noise field using high efficiency detectors was shown to extend beyond neutron noise limits, enabling ex-vessel measurements. The prompt decay constant was determined at about 0.9 meters to the core center with comparable accuracy to that of gamma in-core reference experiments using less-efficient detectors. Gamma correlations were be observed outside of the reactor cavity in front of an experimental channel, enabling a prompt decay constant determination at 7 meters to the core center.
To complement the experimental investigation and to study the underlying physics, a simulation methodology to estimate noise responses was developed. Specifically, the explicit simulation of noise experiments using analog Monte Carlo transport coupled to a fission model was studied. An established code with an existing fission library coupling, TRIPOLI-4, was used as reference. In addition, a newly developed coupling of a fission library to Serpent 2 allowed for code-to-code comparison. The full methodology, verification, and validation to the aforementioned CROCUS experiments is presented.