A neutron flux on the order of 2·10²° neutrons/cm² at 0.18 MeV induces formation of point defects (vacancies and interstitials) in single crystal 3C-SiC causing a volume lattice expansion (swelling) of over 3% that can be measured by X-Ray diffraction. The crystal lattice can be completely restored with an annealing temperature equal to or higher than the irradiation temperature. This phenomenon serves as a basis for temperature measurements and allows the determination of the maximum temperature, if the exposure time is known. The single crystal 3C-SiC sensor is applicable to small, rotating and hard to access parts due to its size of 300-500 microns, wide temperature range of 100-1450 °C, "no-lead" installation, inert chemical properties and high accuracy of temperature measurements. These features make it possible to use the sensor in gas turbine blades, automotive engines, valves, pistons, space shuttle ceramic tiles, thermal protection system design, etc. This work describes the mechanism of neutron irradiation of single crystal 3C-SiC, the formation of point defects and their concentration, the different temperature measurement techniques, and the application of Maximum Temperature Crystal Sensors (MTCS) for maximum temperature measurements in both stationary and non-stationary regimes.

A new nanodosimetry-based cell survival model for mixed high- and low-LET radiations has been developed. The new model employs three dosimetry quantities and three biological quantities. The three dosimetry quantities are related to energy depositions at two nanometer scales, 5nm and 25nm. The three biological quantities are related to lesion production and interaction probabilities, and lesion repair rate. The model assumes that the lesions created at the two nanometer scales are directly or indirectly responsible for cell death depending on the lesions interaction and repair rate. The cell survival fraction derived from the new model can be expressed by the familiar dose-dependent linear quadratic formula. The coefficients alpha and beta are based on the three nanodosimetry quantities and the three biological quantities. Validation of the new model has been performed both by using published data and by the experimental data obtained. Published cell survival curves for V-79 Chinese hamster cells irradiated with various LET of radiations were used for validation. The new model was applied to radiation therapy by irradiating V-79 cells with mixed fission neutron and gamma-rays. The results show that the new model has been successfully used in a mixed n+g field to predict the synergistic effect between neutron and gamma-ray lesions and the RBE for fission neutrons.

There are a variety of uses for reference neutron fields including detector response and dosimeter studies. The Georgia Institute of Technology has a 252Cf spontaneous fission source and an AmBe (α, n) source available for use in its research programs. In addition, it has iron, lead, beryllium, tantalum, heavy water, and polyethylene spheres to modify the neutron energy distributions from these neutron sources. This research characterized the neutron leakage spectra from the source inside spherical shells using a Bonner sphere spectrometer. All the neutron fields measured were also computed with a Monte Carlo code to determine the neutron fluence rate and ambient dose equivalent rate. The comparison of experimental data and calculations are used to provide further insight into the neutron spectra as modified by the spheres. The characterization of these modified sources will provide data to assist in using the resulting neutron fields in other research activities. To measure each neutron field combination, one of the two sources was placed in the center of an attenuating sphere. The neutron field was first measured at a variety of source-to-detector distances with a Bonner Sphere System. The spectrometer measurements, specifically the count rates of the different Bonner spheres, as a function of distance from the source is fitted to obtain corrections for room-scatter and air-scatter of neutrons using the Eisenhauer, Schwartz, and Johnson method. Using these corrections, the count rates free of room return is obtained at 1 m from the source and unfolded using the BUMS software to obtain the reported fluence and dose equivalent rates. These results are compared to those generated by the Monte Carlo Neutral Particle (MCNP) code. Models were made in MCNP for each of the source and moderating sphere combinations. The neutron fluence and dose rates were tallied during the MCNP simulation. The unfolded experimental data and the MCNP calculations showed good agreement for most of source-attenuating sphere combinations, thereby reinforcing the experimental results.

In dieser Arbeit wurde die durch Bestrahlung mit schnellen Neutronen bedingte Materialalterung von Reaktordruckbehälterstählen untersucht. Das Probenmaterial umfasste unbestrahlte, bestrahlte und ausgeheilte RDB-Stähle russischer und westlicher Reaktoren sowie Eisenbasis-Modelllegierungen. Mittels Neutronen-Kleinwinkelstreuung ließen sich bestrahlungsinduzierte Leerstellen/Fremdatom-Cluster unterschiedlicher Zusammensetzung mit mittlerem Radius um 1.0 nm nachweisen. Ihr Volumenanteil steigt mit der Strahlenbelastung monoton, aber im allgemeinen nicht linear an. Der Einfluss der Elemente Cu, Ni und P auf den Prozess der Clusterbildung konnte herausgearbeitet werden. Eine Wärmebehandlung oberhalb der Bestrahlungstemperatur reduziert den Anteil der Strahlendefekte bis hin zu deren vollständiger Auflösung. Die Änderungen der mechanischen Eigenschaften der Werkstoffe lassen sich eindeutig auf die beobachteten Gefügemodifikationen zurückführen. Die abgeleiteten Korrelationen können als Hilfsmittel zur Vorhersage des Materialverhaltens bei fortgeschrittener Betriebsdauer von Leistungsreaktoren mit herangezogen werden.

Advanced nano-structured ferritic alloys (NFAs) containing a high density of ultra-fine (2-5 nm) nanoclusters (NCs) enriched in Y, Ti, and O are considered promising candidates for structural components in future nuclear systems. The superior tensile strengths of NFAs relative to conventional oxide dispersion strengthened (ODS) ferritic alloys are attributed to the high number density of NCs, which may provide effective trapping centers for point defects and transmutation products generated during neutron irradiation. This study consists of production, irradiation, and characterization of an advanced NFA, designated 14YWT, currently being developed at Oak Ridge National Laboratory (ORNL), in Oak Ridge, Tennessee. The purpose of this study was to characterize the tensile and fracture toughness properties of 14YWT produced during this project at ORNL before and after irradiation to evaluate it's resistance to radiation-induced changes in mechanical properties. Another alloy, designated 14WT, was produced during this project using identical production parameters used for 14YWT but without the Y2O3 addition during ball milling required for NC formation. Tensile and fracture toughness specimens were produced from both alloys and irradiated in small "rabbit" capsules in the High Flux Isotope Reactor (HFIR) at ORNL. Five other structural alloys that are currently being evaluated for applications in nuclear environments were irradiated and tested during this project to serve as comparison materials. Microstructural characterization was performed using optical microscopy, scanning electron microscopy, transmission electron microscopy, and atom probe tomography. Tensile strengths for 14YWT were found to be far superior to the other alloys for both irradiated and unirradiated conditions, with yield strength for 14YWT decreasing from ~1,450 MPa at 26°C to ~700 MPa at 600°C. Moderate radiationinduced hardening (50-200 MPa) and reduction in ductility was observed for 14YWT for all irradiation conditions and test temperatures. Fracture toughness results showed 14YWT in the unirradiated condition had a fracture toughness transition temperature (FTTT) around -150°C and upper-shelf K[subscript JIc] values around 175 MPa m. Results from irradiated 14YWT fracture toughness tests were found to closely mirror the unirradiated data and no shift in FTTT or decrease in K[subscript JIc] values were observed following neutron irradiation to 1.5 dpa at 300°C. Master curve analysis of the fracture toughness data show 14YWT to have a T[subscript o] reference temperature of -188 and -176°C in the unirradiated and irradiated condition, respectively, which is unprecedented for a high-strength dispersion strengthened ferritic alloy. The results from this study show 14YWT to be resistant to radiation-induced changes in mechanical properties and a promising candidate for structural applications in future nuclear systems. / text

Anomalous Hall behavior of HgCdTe refers to a "double cross-over" feature of the Hall coefficient in p-type material, or a peak in the Hall mobility or Hall coefficient in n-type material. A magnetoconductivity tensor approach was utilized to identify presence of two electrons contributing to the conduction as well as transport properties of each electron in the material. The two electron model for the mobility shows that the anomalous Hall behavior results from the competition of two electrons, one in the energy gap graded region near the CdZnTe/HgCdTe interface with large band gap and the other in the bulk of the LPE film with narrow band gap. Hg0.78Cd0.22Te samples grown by LPE on CdZnTe(111B)-oriented substrates were exposed to various doses of thermal neutrons (~1.7 x 1016 - 1.25 x 1017 /cm2) and subsequently annealed at ~220oC for ~24h in Hg saturated vapor to recover damage and reduce the presence of Hg vacancies. Extensive Magnetotransport measurements were performed on these samples. SIMS profile for impurities produced by neutron irradiation was also obtained. The purpose for this study is to investigate the influence of neutron irradiation on this material as a basis for further study on HgCdTe74Se. The result shows that total mobility is observed to decrease with increased neutron dose and can be fitted by including a mobility inverse proportional to neutron dose. Electron introduction rate of thermal neutron is much smaller than that of fission neutrons. Total recovering of the material is suggested to have longer time annealing. Using Kane's model, we also fitted carrier concentration change at low temperature by introducing a donor level with activation energy changing with temperature. Results on Se diffusion in liquid phase epitaxy (LPE) grown HgCdTe epilayers is reported. The LPE Hg0.78Cd0.22Te samples were implanted with Se of 2.0×1014/cm2 at 100keV and annealed at 350-450oC in mercury saturated vapor. Secondary ions mass spectrometry (SIMS) profiles were obtained for each sample. From a Gaussian fit we find that the Se diffusion coefficient DSe is about one to two orders of magnitude smaller than that of arsenic. The as-implanted Se distribution is taken into account in case of small diffusion length in Gaussian fitting. Assuming a Te vacancy based mechanism, the Arrhenius relationship yields an activation energy 1.84eV. Dislocations introduced in HgCdTe materials result in two energy levels, where one is a donor and one is an acceptor. Hydrogenation treatment can effectively neutralize these dislocation defect levels. Both experimental results and theoretical calculation show that the mobility due to dislocation scattering remains constant in the low temperature range (<77K), and increases with temperature between 77K and 150K. Dislocation scattering has little effect on electrical transport properties of HgCdTe with an EPD lower than 107/cm2. Dislocations may have little effect on carrier concentration for semiconductor material with zinc blende structure due to self compensation.

Les transistors HEMTs (High Electron Mobility Transistors) de la filière GaN sont destinés à des applications dans les domaines militaire et spatial. C’est pourquoi nous avons étudié l’influence de trois types de stress électriques : à canal ouvert, à canal pincé et NGB (Negative Gate Bias), ainsi que l’influence de neutrons thermalisés avec une fluence pouvant aller jusqu’à 1,7.1012 neutrons.cm-2, sur leurs performances électriques dc.Dans un premier temps, nous avons étudié des HEMTs AlInN/GaN de laboratoire. Pour les trois stress, nous avons observé une dégradation due à la création de pièges accepteurs et donneurs au cours des différents stress et à la présence de pièges préexistants. Nous avons ensuite irradié ces composants par des neutrons thermalisés et avons observé une légère dégradation des performances électriques des transistors non stressés et stressés à canal ouvert ou pincé. En revanche, nous avons mis en lumière une légère amélioration pour les transistors ayant subi un stress NGB. Nous avons également irradié des MOS-HEMTs AlInN/GaN et conclu que ceux-ci étaient plus sensibles vis à vis des irradiations.Dans un deuxième temps, nous avons stressé de manière analogue des HEMTs AlGaN/GaN du commerce. Dans le cas du stress à canal ouvert, nous avons observé une diminution importante du courant de drain tandis que pour les stress à canal pincé et NGB le courant de drain augmente légèrement à cause d’une libération de pièges préexistants sous l’action du champ électrique vertical. Lors des irradiations avec des neutrons thermalisés, ces transistors, stressés ou non, subissent là encore des dégradations. / The GaN based HEMTs (High Electron Mobility Transistors) are excellent candidates for military and spatial applications. That’s why we have analysed the influence of three different types of bias stress: on-state stress, off-state stress and NGB (Negative Gate Bias), and the influence of thermalized neutrons with a fluence up to 1.7x1012 neutrons.cm-2, on their dc electrical performances.First, we have studied laboratory AlInN/GaN HEMTs. For the three conditions of stress, we have observed a degradation due to pre-existing traps and to the creation of acceptor and donor traps during the stress. Then, we have irradiated these components with thermalized neutrons and we have found a small degradation of the electrical performances of unstressed and on-state stressed and off-state stressed transistors. On the other hand, we have highlighted a slight improvement for NGB stressed components. We have also irradiated AlInN/GaN MOS-HEMTs and we have concluded that they are more sensible to irradiation.In a second time we have stressed in the same way commercial AlGaN/GaN HEMTs. For the on-state stress, we have observed an important increase in the drain current. However, the drain current increases for the on-state and NGB stressed components due to a release of electrons from pre-existing traps under vertical electrical field. During the irradiation with thermalized neutrons, the unstressed and stressed transistors are degraded and a small decrease in the drain current is visible.

In dieser Arbeit wurde die durch Bestrahlung mit schnellen Neutronen bedingte Materialalterung von Reaktordruckbehälterstählen untersucht. Das Probenmaterial umfasste unbestrahlte, bestrahlte und ausgeheilte RDB-Stähle russischer und westlicher Reaktoren sowie Eisenbasis-Modelllegierungen. Mittels Neutronen-Kleinwinkelstreuung ließen sich bestrahlungsinduzierte Leerstellen/Fremdatom-Cluster unterschiedlicher Zusammensetzung mit mittlerem Radius um 1.0 nm nachweisen. Ihr Volumenanteil steigt mit der Strahlenbelastung monoton, aber im allgemeinen nicht linear an. Der Einfluss der Elemente Cu, Ni und P auf den Prozess der Clusterbildung konnte herausgearbeitet werden. Eine Wärmebehandlung oberhalb der Bestrahlungstemperatur reduziert den Anteil der Strahlendefekte bis hin zu deren vollständiger Auflösung. Die Änderungen der mechanischen Eigenschaften der Werkstoffe lassen sich eindeutig auf die beobachteten Gefügemodifikationen zurückführen. Die abgeleiteten Korrelationen können als Hilfsmittel zur Vorhersage des Materialverhaltens bei fortgeschrittener Betriebsdauer von Leistungsreaktoren mit herangezogen werden.

<p>Low enriched uranium (LEU, < 20 % 235U)-molybdenum (U-Mo) alloy is the primary nuclear fuel candidate for research and test reactors, and it is also considered one of the fuel candidates for fast reactors. Furthermore, U-Mo monolithic fuel is currently undergoing a qualification process to replace highly-enriched uranium (≥ 20 % 235U) fuel for high-performance research and test reactors. As part of the fuel qualification process, it is critical to examine the microstructural evolution in the final form of U-Mo fuel under irradiation. However, there is a lack of knowledge on the microstructural evolution of the rolled U-Mo alloy foil, which is a proposed geometry for research and test reactors/high-performance research and test reactors, as well as the U-Mo alloy fuel that is cast into a slug form without rolling, which is a more suitable geometry for other advanced reactor fuel types. The effects of the fabrication methods, specifically arc-casting and cold-rolling, on the phase decomposition in U-10Mo alloy subjected to low neutron fluence (0.01 displacements per atom (dpa) and 0.1 dpa) in the temperature range of 150–350oC are evaluated using synchrotron X-ray techniques and scanning electron microscopy (SEM). The as-cast U-10Mo alloys demonstrated better irradiation performance than the U-10Mo alloy foils in research reactor conditions for the investigated regimes. This study will help optimize fuel fabrication techniques to tune phase decomposition under irradiation. </p> <p>Additionally, irradiation behavior in Mo, a critical element in U-Mo fuel, as well as a candidate for cladding material for the next generation of nuclear power plants, is investigated at the atomistic scale following low neutron fluence regimes (0.01 dpa and 0.1 dpa) in the temperature range of 150–800oC using synchrotron X-ray techniques. Lattice contraction was observed in irradiated Mo by synchrotron XRD, indicating that interstitial diffusion is faster than vacancy diffusion in Mo. More interstitials diffuse into the sinks, such as grain boundaries, while fewer vacancies diffuse into the sinks due to slow diffusion, resulting in a higher steady-state concentration of vacancies than that of interstitials under irradiation. The synchrotron PDF also supported the synchrotron XRD results by demonstrating a decrease in the atomic distances under irradiation. </p>

Ferritisch/martensitische Cr-Stähle und deren oxiddispersionsverfestigte Varianten gehören zu den potenziellen Konstruktionswerkstoffen für Komponenten zukünftiger kerntechnischer Einrichtungen, wie z. B. Fusionsreaktoren und Spaltreaktoren der IV. Generation, die Strahlungsfeldern mit hohem Neutronenfluss ausgesetzt sind. Ein Hauptproblem dieser Materialgruppen ist das Auftreten des Spröd-duktil-Übergangs und dessen maßgeblich durch die Strahlenhärtung verursachte Verschiebung zu höheren Temperaturen. In der vorliegenden Arbeit wird das Bestrahlungsverhalten von binären Fe-Cr-Modelllegierungen untersucht, die ein vereinfachtes Modell für ferritisch/martensitische Cr-Stähle darstellen. Dabei werden Bestrahlungen mit Eisenionen zur Simulation der durch Neutronen hervorgerufenen Schädigung verwendet. Die auf wenige Mikrometer begrenzte Eindringtiefe der Ionen macht es erforderlich, dass für dünne Schichten geeignete Charakterisierungsmethoden eingesetzt werden. Im Rahmen dieser Arbeit sind das Nanohärtemessungen und Transmissionselektronenmikroskopie (TEM). Im Ergebnis liegen die bestrahlungsinduzierte Härteänderung der Schicht in Abhängigkeit von Chromgehalt, Bestrahlungsfluenz und –temperatur sowie, für ausgewählte Zustände, quantitative TEM-Analysen vor. Zusammen mit begleitenden Ergebnissen von Neutronenkleinwinkelstreuexperimenten an neutronenbestrahlten Proben der gleichen Werkstoffe ermöglichen sie die Identifizierung von bestrahlungsinduzierten Versetzungsringen und nm-großen α’-Ausscheidungen als Quellen der Strahlenhärtung. Im Rahmen eines vereinfachten Modells, das auf Orowan zurückgeht, werden die Hindernisstärken dieser Gitterbaufehler für das Gleiten von Versetzungen abgeschätzt. Darauf aufbauend erfolgt ausblickartig eine Erweiterung des Untersuchungsgegenstands auf komplexere Situationen hinsichtlich der Bestrahlungsbedingungen und des Werkstoffs. Durch das Einbeziehen simultaner und sequentieller Bestrahlungen mit Eisen- und Heliumionen kann gezeigt werden, dass der Effekt von Helium auf die Strahlenhärtung von der Bestrahlungsreihenfolge abhängt und dass der simultane Eintrag fusionsrelevanter Mengen von Helium zu einer Verstärkung der Strahlenhärtung führt, die auf einem synergistischen Effekt beruht. Für Cr-Stähle mit 9 % Cr und deren oxiddispersionsverfestigte Varianten wird kein grundlegend anderes Bestrahlungsverhalten beobachtet als für binäres Fe-9at%Cr. Es gibt jedoch Hinweise, dass Oxid-dispersionsverfestigung die Strahlenhärtung unter bestimmten Bedingungen reduzieren kann. Im Ergebnis der Arbeit zeigt sich, dass Ionenbestrahlungen in Kombination mit Nanohärtemessungen zu einem vertiefenden Verständnis der Strahlenhärtung in Werkstoffen auf Fe-Cr-Basis sowie zu einer effektiven Materialvorauswahl beitragen können. Voraussetzung ist, dass der Eindruckgrößeneffekt und der Substrateffekt auf geeignete Weise in Rechnung gestellt werden.