In a lead-cooled fast reactor (LFR), small bubbles (in the order of one mm or less) may enter the coolant from a leaking steam generator. If such a leakage is undetected the small bubbles may eventually coalesce into a larger bubble in local stagnation zones under the active core. If such a bubble or void releases and passes through the core, it could drive the reactor into prompt criticality. It is therefore desirable to be able to detect the initial stages of such void formation. In this thesis, a methodology to detect such leaks is presented together with a study on void-induced reactivity effects in various LFR's. The methodology developed is based on information from two fission chambers positioned radially outside the core. The fissile content of the fission chambers consist either of 235U or 242Pu making them sensitive to different parts of the neutron spectrum. It is shown that the information from the fission chambers can be used to obtain an early indication of the presence of a small leak within typically a month. Furthermore, it is shown that for all but the smallest LFR’s, prompt criticality due to voids passing the core cannot be excluded. One conclusion is that the methodology may form an attractive complement to the general monitoring system of future LFR’s but, as is noted, it has potential for further developments.

One of the future reactor technologies defined by the Generation-IV International Forum (GIF) is the Lead-Cooled Fast Reactor (LFR). An advantage with this reactor technology is that steam production is accomplished by means of heat exchangers located within the primary reactor vessel, which decreases costs and increases operational safety. However, a crack in a heat exchanger tube may create steam (void) into the coolant and this process has the potential to introduce reactivity changes, which may cause criticality issues. This fact motivates the development of a methodology to detect such voids. This thesis comprises theoretical investigations on a possible route to detect voids by studying changes of the neutron spectrum in a small LFR as a function of various types of in-core voids .The methodology includes a combination of fission chambers loaded with U-235 and Pu-242 operating in various positions. It is shown that such a combination results in information that can be made independent on reactor power, a feasible property in order to detect the relatively small spectral changes due to void. A sensitivity analysis of various combinations of detectors, fuel burnup and void has also been included in the investigation. The results show that the proposed methodology yields a reasonably large sensitivity to voids down to (1-2) % of the coolant volume. The results obtained so far point in the direction that the proposed methodology is an interesting subject for further studies.

Doctor of Philosophy / Department of Mechanical and Nuclear Engineering / Douglas S. McGregor / Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Measurement of the neutron-flux distribution within the reactor core provides a more complete understanding of the operating conditions in the reactor than typical ex-core sensors. Micro-Pocket Fission Detectors (MPFDs) have been developed and tested previously but have been limited to single-node operation and have utilized highly specialized designs. The development of a widely deployable, multi-node MPFD assembly will enhance nuclear research capabilities. In-core neutron flux measurements include many challenges because of the harsh environment within the reactor core. Common methods of in-core neutron measurement are also limited by geometry and other physical constraints. MPFDs are designed to be small and robust while offering a real-time, spatial measurement of neutron flux. Improvements to the MPFD design were developed based on shortcomings of prior research in which many of the theoretical considerations for MPFDs were examined. Fabrication techniques were developed for the preparation of MPFD components and electrodeposition of fissile material. Numerous arrays of MPFDs were constructed for test deployments at the Kansas State University TRIGA Mk. II research nuclear reactor, University of Wisconsin Nuclear Reactor, Transient REActor Test facility at the Idaho National Laboratory (INL), and Advanced Test Reactor at INL. Preliminary testing of a single MPFD sensor at KSU yielded a linear response to reactor power between 10 kWth and 750 kWth and followed both positive and negative reactivity insertions in real-time. A $1.50 reactor pulse was monitored from the Intra-Reflector Irradiation System, located in reflector region of the KSU TRIGA Mk. II core with 1-ms time resolution. Improved multi-node MPFD arrays were then designed, fabricated, and deployed in flux ports between fuel rods and within an iron-wire flux port which was inserted into the central thimble of the KSU TRIGA Mk. II research nuclear reactor. Work continues to develop MPFDs for deployment at research reactors at INL and elsewhere. Results from the MPFD measurements will be useful for future validation of computational modeling and as part of advanced nuclear fuel development efforts.

Doctor of Philosophy / Department of Mechanical and Nuclear Engineering / Douglas S. McGregor / A Micro-Pocket Fission Detector (MPFD) is a miniaturized type of fission chamber developed for use inside a nuclear reactor. Their unique design allows them to be located between or even inside fuel pins while being built from materials which give them an operational lifetime comparable to or exceeding the life of the fuel. While other types of neutron detectors have been made for use inside a nuclear reactor, the MPFD is the first neutron detector which can survive sustained use inside a nuclear reactor while providing a real-time measurement of the neutron flux. This dissertation covers the deployment of MPFDs as a large three-dimensional array inside the Kansas State University TRIGA Mark-II Nuclear Reactor for real-time neutron flux measurements. This entails advancements in the design, construction, and packaging of the Micro-Pocket Fission Detector Triads with incorporated Thermocouple, or MPFD[superscript]3-T. Specialized electronics and software also had to be designed and built in order to make a functional system capable of collecting real-time data from up to 60 MPFD[superscript]3-Ts, or 180 individual MPFDs and 60 thermocouples. Design of the electronics required the development of detailed simulations and analysis for determining the theoretical response of the detectors and determination of their size. The results of this research shows that MPFDs can operate for extended times inside a nuclear reactor and can be utilized toward the use as distributed neutron detector arrays for advanced reactor control systems and power mapping. These functions are critical for continued gains in efficiency of nuclear power reactors while also improving safety through relatively inexpensive redundancy.

Master of Science / Department of Mechanical and Nuclear Engineering / Douglas S. McGregor / Fission chambers are ideal neutron flux surveillance instruments to ensure nuclear reactor control and safety. They can provide online, in-core, real-time measurements covering the dynamic range of neutron flux including pulse, Campbell, and current mode over decades of reactor operation cycles. The first patented fission chamber was developed by Baer et al. in 1957. It was a cylindrical assembly thermal fission counter having sensitivity of 0.7 count/neutron cm⁻² for a background measurement of 5 counts/second with ability to operate at a temperature range of 20-80 ºC [3]. Since then, fission chamber technology was developed to come up with miniature and sub-miniature dimensions withstanding high irradiation and high temperature environment making them suitable for in-core online diagnosis. Since the introduction of high temperature fission chamber technology starting in the 1970’s, the need of the advancement in modeling of the fission chambers to improve their performance has become important. The development of modeling depends upon the understanding and consideration of underlying physics of these detectors. The validation of modeling of fission chambers will need the quantification of uncertainty introduced at every stage from neutron-deposit interaction to signal shaping. Based on this objective, a detailed review was performed on fission chamber modeling and simulation covering neutron flux self-shielding, fissile deposit evolution, fission product emission, auto-absorption, electron-ion pair creation, charge recombination and avalanche, space charge effect, charge transport, propagation of electronic pulse and pulse shaping. The analytical methods, algorithmic treatments, simulation, and computation codes used so far in case of modeling different aspects of fission chambers were reviewed. Along with the numerical methods and computer codes for simulating electron drift and charge transport for the usual gas chamber detectors, the use of several fissile material evolution techniques and computation codes were observed in case of fission chamber modeling. The use higher order statistics to handle fluctuation mode and to treat noisy data were observed. In recent years, fission chamber modeling made reasonable improvement in detail physics modeling. Several analytical methods like advanced statistics for Campbellng mode and electric field distortion due to space charge effect need to be incorporated in computation codes. More progress in the areas of evolution of gas behavior, consideration of Penning, recombination, and avalanche effect still needed.

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No description available.

Safety and reliability are one of the key objectives for future Generation IV nuclear energy systems. The neutron flux monitoring system forms an integral part of the safety design of a nuclear reactor and must be able to detect any irregularities during all states of reactor operation. The work in this thesis mainly concerns the detection of in-core perturbations arising from unwanted movements of control rods with in-vessel neutron detectors in a sodium-cooled fast reactor. Feasibility study of self-powered neutron detectors (SPNDs) with platinum emitters as in-core power profile monitors for SFRs at full power is performed. The study shows that an SPND with a platinum emitter generates a prompt current signal induced by neutrons and gammas of the order of 600 nA/m, which is large enough to be measurable. Therefore, it is possible for the SPND to follow local power fluctuations at full power operation. Ex-core and in-core detector locations are investigated with two types of detectors, fission chambers and self-powered neutron detectors (SPNDs) respectively, to study the possibility of detection of the spatial changes in the power profile during two different transient conditions, i.e. inadvertent withdrawal of control rods (IRW) and one stuck rod during reactor shutdown (OSR). It is shown that it is possible to detect the two simulated transients with this set of ex-core and in-core detectors before any melting of the fuel takes place. The detector signal can tolerate a noise level up to 5% during an IRW and up to 1% during an OSR.

La fission nucléaire est un phénomène complexe dont tous les mécanismes ne sont pas entièrement compris. Dans le cadre d'une coopération internationale, le CEA/DAM/DIF étudie les spectres en énergie des neutrons prompts émis lors de la fission induite par des neutrons rapides, et plus particulièrement la zone à basse énergie de ces spectres (<1 MeV). Ce travail de thèse a consisté à optimiser un dispositif expérimental de mesure de neutrons prompts de fission. Dans un premier temps, de nouveaux détecteurs de fission ont été développés. Nous en rapportons ici la conception et étudions leurs performances en termes de discrimination alpha-fission, de résolution en temps et de distorsion sur le spectre mesuré. Le second axe de développement abordé au cours de cette thèse est celui de la détection des neutrons. Plusieurs types de détecteurs ont été comparés (discrimination neutron-gamma, efficacité de détection), en vue d'optimiser la détection des neutrons de basse énergie (<1 MeV). Ce mémoire présente les résultats de ces études. Enfin, le dispositif expérimental ainsi optimisé est utilisé pour mesurer le spectre en énergie des neutrons prompts émis lors de la fission induite par neutrons de l' ²³⁸UU. Après avoir présenté la méthode utilisée pour l'analyse des données, les résultats obtenus sont interprétés en termes de modèles et d'évaluations. / The nuclear fission is a complex phenomenon whose mechanisms are not fully understood. Within the framework of an international cooperation, the CEA/DAM/DIF is taking part in the study of prompt fission neutron energy spectra from fast neutron induced fission, focusing on the low energy domain of these spectra (<1 MeV). This PhD was dedicated to the optimization of the experimental setup. New fission detectors were developed. We report on their conception and their performances in terms of alpha-fission discrimination, timing resolution and distortion on the measured spectrum. In a second step, several neutron detectors were studied (neutron-gamma discrimination, detection efficiency), so as to optimize the detection of low energy neutrons (<1 MeV). In the present document, we report on the results of this comparative study. Finally, the optimized experimental setup was used to measure prompt fission neutron energy spectra for the fast-neutron induced fission of ²³⁸U. After detailing the data analysis method, the results are interpreted in terms of models and evaluations.

In dieser Forschungsarbeit wurde die ortsabhängige Neutronenflussdichteverteilung des Ausbildungskernreaktors AKR-2 untersucht. Ziel der Arbeit war es, die radiale und tangentiale Verteilung der Neutronenflussdichte experimentell zu messen und diese Ergebnisse mit Simulationen zu vergleichen. Für die Experimente wurden Gold- und Manganelemente im Reaktor aktiviert und ihre Neutronenflussdichten mit verschiedenen Methoden gemessen, darunter die Verwendung einer Spaltkammer und die Anwendung des Monte-Carlo-Programms SERPENT zur Simulation. Es zeigte sich, dass die experimentellen Ergebnisse in allen Messmethoden eine ähnliche Verteilung der Neutronenflussdichte aufwiesen. Allerdings gab es Abweichungen zwischen den Simulationen und den experimentellen Daten, was auf die vereinfachte, eindimensionale Modellierung und die Vernachlässigung von Neutronen unterschiedlicher Energien zurückgeführt wurde. Die Arbeit kommt zu dem Schluss, dass mehrdimensionale Simulationen notwendig sind, um die Neutronenverteilung in einem Reaktor wie dem AKR-2 präziser abzubilden.:1 EINLEITUNG 1 1.1 Allgemein ................................................................................................................... 1 1.2 Versuchsstandort: AKR-2 .......................................................................................... 1 1.3 Grundlagen ................................................................................................................ 2 1.3.1 Neutroneneinfang ........................................................................................2 1.3.2 Prinzip der Kernreaktion .............................................................................. 2 2 Experimente 3 2.1 Allgemein ................................................................................................................... 3 2.2 Reaktorstart ............................................................................................................... 4 2.3 Experimente mit Mangan ......................................................................................... 4 2.3.1 Versuchsausrüstung ..................................................................................... 4 2.3.2 Datenmessungen radial durch den Reaktor .............................................. 5 2.3.3 Messung entlang des Tangentialkanals des Reaktors...............................9 2.4 Experimente mit Gold.............................................................................................10 2.4.1 Versuchsausrüstung ...................................................................................10 2.4.2 Ziel................................................................................................................12 2.4.3 Versuchsdurchführung...............................................................................12 2.4.4 Datenauswertung .......................................................................................13 2.5 Experimente mit einer Spaltkammer (F4CA) ........................................................16 2.5.1 Versuchsausrüstung ...................................................................................16 2.5.2 Messung direkt mit F4CA ...........................................................................17 2.5.3 Messung mit F4CA und Referenzspaltkammer........................................19 2.6 Vergleich der experimentellen Ergebnisse mit Simulationsdaten ......................22 Inhaltsverzeichnis 3 Simulation 24 3.1 Allgemein .................................................................................................................24 3.2 Diffusionsgleichung und numerische Lösung ......................................................24 3.3 Vergleich der drei Randbedingungen....................................................................27 3.3.1 Randbedingung 1........................................................................................27 3.3.2 Randbedingung 2........................................................................................28 3.3.3 Randbedingung 3........................................................................................29 3.3.4 Vergleich der drei Randbedingungen und Bestimmung der Randbedingung .......................................................................................................31 3.4 Simulationsmethode...............................................................................................32 3.5 Konstruktion des Reaktormodells und Abbildung der Neutronenverteilung....36 3.6 Vergleich der Simulationsergebnisse 4 Zusammenfassung Literaturverzeichnis Abkürzungsverzeichnis Symbolverzeichnis Abbildungsverzeichnis Tabellenverzeichnis A radiale Mssungsdaten von Mangan B Tangentiale Mssungsdaten von Mangan