Master of Science / Department of Mechanical and Nuclear Engineering / Douglas S. McGregor / Fission chambers are a common type of detector used to determine the neutron flux and power of a nuclear reactor. Due to the limited space and high neutron flux in a reactor core, it is difficult to perform real-time flux measurements with present-day in-core instrumentation. Micro-pocket fission detectors, or MPFDs, are relatively small in size and have low neutron sensitivity while retaining a large neutron to gamma ray discrimination ratio, thereby, allowing them to be used as active neutron flux monitors inside a nuclear reactor core. The micro-pocket fission chamber allows for multiple detectors to be inserted into a flux port or other available openings within the nuclear reactor core. Any material used to construct the MPFD must be rugged and capable of sustaining radiation damage for long periods of time. Each calibrated MPFD provides measurements of the flux for a discrete location. The size of these detectors allows for a spatial map of the flux to be developed, enabling real-time analysis of core burnup, power peaking, and rod shadowing. Small diameter thermocouples can be included with the array to also measure the temperature at each location. The following document details the research and development of MPFDs for long term use in nuclear power reactors. Previous MPFD designs were improved, miniaturized, and optimized for long term operations in reactor test ports designed for passive measurements of fluence using iron wires. Detector chambers with dimensions of 0.08 in x 0.06 in x 0.04 in were attached to a common cathode and individual anodes to construct an array of the MPFDs. Each array was tested at the Kansas State University TRIGA Mark II nuclear reactor to demonstrate functionality. The linear response in reactor power was measured. These arrays have also demonstrated reactor power tracking by following reactivity changes in steady state operations and reactor pulsing events. Stability testing showed consistent operation at 100 kW for several hours. The MPFDs have been demonstrated to be a viable technology for in-core measurements.

The smuggling of special nuclear material (SNM) has long been a concern. In April 2009, President Obama declared that a terrorist acquiring a nuclear weapon was the most immediate threat to global security. The Second Line of Defense (SLD) initiative was stood up by the National Nuclear Security Administration to deter, detect, and interdict illicit trafficking of nuclear and radioactive materials across international borders and maritime shipping. The SLD initiative does not provide for the detection of SNM being carried on small, personal watercraft. Previous work examined the possibility of using active neutron detectors to induce fission in SNM and detect the response. This thesis examines the possibility of detecting SNM using passive 3He neutron detectors. Monte Carlo N-Particle (MCNP) simulations were run to determine the best detector configuration. Detecting sources at increasing depths, detecting moving sources and the effects of waves were also simulated in MCNP. Comparisons with experimental measurements showed that detectors parallel to the surface of water were best at detecting neutron sources below the surface. Additionally, stacking detectors and placing a cadmium sheet between the polyethylene blocks resulted in a greater ability to determine the height of a source by taking the ratio of count rates in the lower and upper detectors. Using this configuration, a source of strength 3.39 x 10^5 n/s could be detected to a depth of 12.00 in below the water surface. Count rates in the presence of waves did not average out to count rates taken above a flat plane of water. Detectors closer to the water performed worse than above a flat plane while detectors placed higher recorded more counts than above a flat plane. Moving sources were also simulated; sources under water, 3.00 ft from the detectors, and moving at 5.8 kts could be detected above background.

Boron carbide-aromatic composites, formed by plasma-enhanced co-deposition of carboranes and aromatic precursors, present enhanced electron-hole separation as neutron detector. This is achieved by aromatic coordination to the carborane icosahedra and results in improved neutron detection efficiency. Photoemission (XPS) and FTIR suggest that chemical bonding between B atoms in icosahedra and aromatic contents with preservation of π system during plasma process. XPS, UPS, density functional theory (DFT) calculations, and variable angle spectroscopic ellipsometery (VASE) demonstrate that for orthocarborane/pyridine and orthocarborane/aniline films, states near the valence band maximum are aromatic in character, while states near the conduction band minimum include those of either carborane or aromatic character. Thus, excitation across the band gap results in electrons and holes on carboranes and aromatics, respectively. Further such aromatic-carborane interaction dramatically shrinks the indirect band gap from 3 eV (PECVD orthocarborane) to ~ 1.6 eV (PECVD orthocarborane/pyridine) to ~1.0 eV (PECVD orthocarborane/aniline), with little variation in such properties with aromatic/orthocarborane stoichiometry. The narrowed band gap indicate the potential for greatly enhanced charge generation relative to PECVD orthocarborane films, as confirmed by zero-bias neutron voltaic studies. The results indicate that the enhanced electron-hole separation and band gap narrowing observed for aromatic/orthocarborane films relative to PECVD orthocarborane, has significant potential for a range of applications, including neutron detection, photovoltaics, and photocatalysis. Acknowledgements: This work was supported by the Defense Threat Reduction Agency (Grant No.HDTRA1-14-1-0041). James Hilfiker is also gratefully acknowledged for stimulating discussions.

We present a preliminary design for a neutron detection system that is compact, lightweight, and low power consuming, utilizing the CubeSat platform making it suitable for space-based applications. This is made possible using the scintillating crystal lithium indium diselenide ((LiInSe2)-Li-6), the first crystal to include Li-6 in the crystalline structure, and a silicon avalanche photodiode. The schematics of this instrument are presented as well as the response of the instrument to initial testing under alpha radiation. A principal aim of this work is to demonstrate the feasibility of such a neutron detection system within a CubeSat platform. The entire end-to-end system presented here is 10 x 10 x 15 cm(3), weighs 670 g, and requires 5 V direct current at 3 W. (C) 2016 Society of Photo-Optical Instrumentation Engineers (SPIE)

Doctor of Philosophy / Department of Mechanical and Nuclear Engineering / Douglas S. McGregor / Interest in high-efficiency replacements for thin-film-coated thermal neutron detectors led to the development of single-sided microstructured semiconductor neutron detectors (MSNDs). MSNDs are designed with micro-sized trench structures that are etched into a vertically-oriented pvn-junction diode, and backfilled with a neutron converting material, such as ⁶LiF. Neutrons absorbed by the converting material produce a pair of charged-particle reaction products that can be measured by the diode substrate. MSNDs have higher neutron-absorption and reaction-product counting efficiencies than their thin-film-coated counterparts, resulting in up to a 10x increase in intrinsic thermal neutron detection efficiency. The detection efficiency for a single-sided MSND is reduced by neutron streaming paths between the conversion-material filled regions that consequently allow neutrons to pass undetected through the detector. Previously, the highest reported intrinsic thermal neutron detection efficiency for a single MSND was approximately 30%. Methods for double-stacking and aligning MSNDs to reduce neutron streaming produced devices with an intrinsic thermal neutron detection efficiency of 42%. Presented here is a new type of MSND that features a complementary second set of trenches that are etched into the back-side of the detector substrate. These dual-sided microstructured semiconductor neutron detectors (DS-MSNDs) have the ability to absorb and detect neutrons that stream through the front-side, effectively doubling the detection efficiency of a single-sided device. DS-MSND sensors are theoretically capable of achieving greater than 80% intrinsic thermal neutron detection efficiency for a 1-mm thick device. Prototype DS-MSNDs with diffused pvp-junction operated at 0-V applied bias have achieved 53.54±0.61%, exceeding that of the single-sided MSNDs and double-stacked MSNDs to represent a new record for detection efficiency for such solid-state devices.

<div> <div> <div> <p>This dissertation describes work conducted in pursuit of interests in adapting Tension Metastable Fluid Detectors (TMFDs) for dosimetry-related applications with the specific intent of engineering a neutron ambient dose spectrometer. TMFDs possess several charac- teristics desirable for neutron spectrometry, including high efficiencies, complete blindness to gamma and beta radiation, and tailorable-threshold response functions. Prior spectro- scopic work with TMFDs, aptly named Single Atom Spectroscopy (SAS), was constrained to a specific subset of detection fluids who’s composition includes hydrogen and only one other higher Z element (e.g. hydrocarbons), where only one element is assumed capable of initiating a cavitation detection event (CDE). The present work alleviates these restrictions, enabling spectroscopy in detection fluids with multiple constituent elements. </p> <p>Simulating the detector’s response predicates knowledge of the energy necessary for ra- diation induced nucleation, which has been theoretically derived with nucleation theory for superheated fluids, but remains unbeknownst for tensioned metastable states. This limi- tation was overcome using MCNPX-PoliMI to model the spatial recoil nuclei spectra from isotope sources and coupled with SRIM to generate the ion energy deposition probabil- ity density within a critical length scale of each interaction event. Thereafter, the energy deposition threshold necessary to generate a detection event, and corresponding response matrix, was derived empirically by solving for the solution curve that minimizes the residual difference between the measured and simulated count rates. </p> <p>The accuracy of the derived response matrix was evaluated through comparisons with a 6LiI Bonner Sphere Spectrometer in which, for 252Cf and 239PuBe/241AmBe isotope source neutron spectra, the two systems offered results within ±10% of each other for ambient equivalent fluences on the order of 100 μRem/hr fields. Notably, when under ultra-low (10 μRem/hr) fields the Bonner spectrometer and other traditional detectors proved impractical. In contrast, the TMFD system was capable of resolving underlying spectral features and corresponding ambient dose rates within ±5% of MCNP predictions. </p> </div> </div> </div>

Master of Science / Department of Mechanical and Nuclear Engineering / Jeffrey Geuther / William L. Dunn / The properties of hexagonal boron nitride (h-BN) as a semiconductor neutron detection medium were investigated. Single h-BN crystal domains were synthesized by the Chemical Engineering department at Kansas State University (KSU) using crystallization from molten metal solutions. At Texas Tech University (TTU), a detector was fabricated using epitaxial h-BN growth on a sapphire substrate where metallic micro-strip contacts 5 [mu]m apart and 5 nm thick where deposited onto the un-doped h-BN. In this research both the crystal domains synthesized at KSU and the detector fabricated at TTU were tested for neutron response. Neutron irradiation damage/effects were studied in pyrolytic h-BN by placing samples in the central thimble of the TRIGA MARK II reactor at KSU and irradiating at increasing neutron fluences. The domains synthesized at KSU as well as the detector fabricated at TTU showed no response to neutron activity on a MCA pulse height spectrum. Conductivity analysis showed abrupt increases in the conductivity of the pyrolytic h-BN at around a fluence of 10[superscript]1[superscript]4 neutrons per cm[superscript]2. Bandgap analysis by photoluminescence on the irradiated pyrolytic h-BN samples showed shifts in energy due to towards plane stacking disorders upon neutron irradiation. Future efforts may include the introduction of dopants in h-BN growth techniques for charge carrier transport improvement, and mitigation of plane stacking disorders.

A Terapia por Captura de Nêutrons pelo Boro (BNCT) é uma técnica radioterapêutica, que visa o tratamento de alguns tipos de câncer, em que sua energia útil é proveniente da reação nuclear promovida pela incidência de nêutrons térmicos no isótopo de 10B. No Brasil existe uma instalação, localizada junto ao canal de irradiação número 3 do Reator de Pesquisas IEA-R1 do IPEN, que foi projetada para o desenvolvimento de pesquisas em BNCT. Para uma aplicação adequada da técnica é necessário que o feixe de irradiação na posição de amostra, seja composto predominantemente por nêutrons térmicos com reduzida contaminação dos componentes do feixe, correspondente aos nêutrons epitérmicos e rápidos e à radiação gama. Este trabalho tem como objetivo monitorar e avaliar o feixe de irradiação na posição de irradiação de amostras, através do uso de detectores de ativação (folhas de ativação), e a partir de simulações utilizando o código de transporte de radiação, MCNP, avaliar mudanças na instalação, mais especificamente no conjunto de filtros e moderadores, para que se aprimore as condições de irradiação na instalação. O trabalho propos uma nova metodologia de cálculo para estudos de otimização do feixe a partir do recurso de redução de variância presente no MCNP, o wwg (weight window generation). Com os resultados obtidos através da adoção de um conjunto maior de folhas de ativação, foi possível a discriminação experimental do feixe de nêutrons em 5 faixas de energia e concluir que a instalação para estudos em BNCT do IPEN possui fluxo de nêutrons térmicos de 108 n/cm2.s, intensidade suficiente para que os estudos na área possam ser realizados com grande potencial de alteração de suas componentes conforme demanda experimental. / Boron Neutron Capture Therapy (BNCT) is a radiotherapeutic technique for the treatement of some types of cancer whose useful energy comes from a nuclear reaction that occurs when thermal neutron impinges upon a Boron-10 atom. In Brazil there is a research facility built along the beam hole number 3 of the IEA-R1 research reactor at IPEN, which was designed to perform BNCT research experiments. For a good performance of the technique, the irradiation beam should be mostly composed of thermal neutrons with a minimum as possible gamma and above thermal neutron components. This work aims to monitor and evaluate the irradiation beam on the sample irradiation position through the use of activation detectors (activation foils) and also to propose, through simulation using the radiation transport code, MCNP, new sets of moderators and filters which shall deliver better irradiation fields at the irradiation sample position In this work, a simulation methodology, based on a MCNP card, known as wwg (weight window generation) was studied, and the neutron energy spectrum has been experimentally discrimanted at 5 energy ranges by using a new set o activation foils. It also has been concluded that the BNCT research facility has the required thermal neutron flux to performe studies in the area and it has a great potencial for improvement for tailoring the irradiation field.

A Terapia por Captura de Nêutrons pelo Boro (BNCT) é uma técnica radioterapêutica, que visa o tratamento de alguns tipos de câncer, em que sua energia útil é proveniente da reação nuclear promovida pela incidência de nêutrons térmicos no isótopo de 10B. No Brasil existe uma instalação, localizada junto ao canal de irradiação número 3 do Reator de Pesquisas IEA-R1 do IPEN, que foi projetada para o desenvolvimento de pesquisas em BNCT. Para uma aplicação adequada da técnica é necessário que o feixe de irradiação na posição de amostra, seja composto predominantemente por nêutrons térmicos com reduzida contaminação dos componentes do feixe, correspondente aos nêutrons epitérmicos e rápidos e à radiação gama. Este trabalho tem como objetivo monitorar e avaliar o feixe de irradiação na posição de irradiação de amostras, através do uso de detectores de ativação (folhas de ativação), e a partir de simulações utilizando o código de transporte de radiação, MCNP, avaliar mudanças na instalação, mais especificamente no conjunto de filtros e moderadores, para que se aprimore as condições de irradiação na instalação. O trabalho propos uma nova metodologia de cálculo para estudos de otimização do feixe a partir do recurso de redução de variância presente no MCNP, o wwg (weight window generation). Com os resultados obtidos através da adoção de um conjunto maior de folhas de ativação, foi possível a discriminação experimental do feixe de nêutrons em 5 faixas de energia e concluir que a instalação para estudos em BNCT do IPEN possui fluxo de nêutrons térmicos de 108 n/cm2.s, intensidade suficiente para que os estudos na área possam ser realizados com grande potencial de alteração de suas componentes conforme demanda experimental. / Boron Neutron Capture Therapy (BNCT) is a radiotherapeutic technique for the treatement of some types of cancer whose useful energy comes from a nuclear reaction that occurs when thermal neutron impinges upon a Boron-10 atom. In Brazil there is a research facility built along the beam hole number 3 of the IEA-R1 research reactor at IPEN, which was designed to perform BNCT research experiments. For a good performance of the technique, the irradiation beam should be mostly composed of thermal neutrons with a minimum as possible gamma and above thermal neutron components. This work aims to monitor and evaluate the irradiation beam on the sample irradiation position through the use of activation detectors (activation foils) and also to propose, through simulation using the radiation transport code, MCNP, new sets of moderators and filters which shall deliver better irradiation fields at the irradiation sample position In this work, a simulation methodology, based on a MCNP card, known as wwg (weight window generation) was studied, and the neutron energy spectrum has been experimentally discrimanted at 5 energy ranges by using a new set o activation foils. It also has been concluded that the BNCT research facility has the required thermal neutron flux to performe studies in the area and it has a great potencial for improvement for tailoring the irradiation field.

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