Doctor of Philosophy / Department of Mechanical and Nuclear Engineering / Douglas S. McGregor / The microstructured semiconductor neutron detector (MSND) was investigated and previous designs were improved and optimized. In the present work, fabrication techniques have been refined and improved to produce three-dimensional microstructured semiconductor neutron detectors with reduced leakage current, reduced capacitance, highly anisotropic deep etched trenches, and increased signal-to-noise ratios. As a result of these improvements, new MSND detection systems function with better gamma-ray discrimination and are easier to fabricate than previous designs. In addition to the microstructured diode fabrication improvement, a superior batch processing backfill-method for 6LiF neutron reactive material, resulting in a nearly-solid backfill, was developed. This method incorporates a LiF nano-sizing process and a centrifugal batch process for backfilling the nanoparticle LiF material. To better transition the MSND detector to commercialization, the fabrication process was studied and enhanced to better facilitate low cost and batch process MSND production. The research and development of the MSND technology described in this work includes fabrication of variant microstructured diode designs, which have been simulated through MSND physics models to predict performance and neutron detection efficiency, and testing the operational performance of these designs in regards to neutron detection efficiency, gamma-ray rejection, and silicon fabrication methodology. The highest thermal-neutron detection efficiency reported to date for a solid-state semiconductor detector is presented in this work. MSNDs show excellent neutron to gamma-ray (n/γ) rejection ratios, which are on the order of 106, without significant loss in thermal-neutron detection efficiency. Individually, the MSND is intrinsically highly sensitive to thermal neutrons, but not extrinsically sensitive because of their small size. To improve upon this, individual MSNDs were tiled together into a 6x6-element array on a single silicon chip. Individual elements of the array were tested for thermal-neutron detection efficiency and for the n/γ reject ratio. Overall, because of the inadequacies and costs of other neutron detection systems, the MSND is the premier technology for many neutron detection applications.

With an increase in the capabilities and sophistication of terrorist networks worldwide comes a corresponding increase in the probability of a radiological or nuclear device being detonated within the borders of the United States. One method to decrease the risk associated with this threat is to interdict the material during transport into the US. Current RPMS have limitations in their ability to detect shielded nuclear materials. It was proposed that directionally sensitive neutron detectors might be able to overcome many of these limitations. This thesis presents a method to create a directionally sensitive neutron detector using a unique characteristic of 10B. This characteristic is the Doppler broadening of the de-excitation gamma-ray from the 10B(n, alpha) reaction. Using conservation principles and the method of cone superposition, the mathematics for determining the incoming neutron direction vector from counts in a boron loaded cloud chamber and boron loaded semiconductor were derived. An external routine for MCNPX was developed to calculate the Doppler broaden de-excitation gamma-rays. The calculated spectrum of Doppler broadened de-excitation gamma-rays was then compared to measured and analytical spectrums and matched with a high degree of accuracy. MCNPX simulations were performed for both a prototype 10B loaded cloud chamber and prototype 10B loaded semiconductor detector. These simulations assessed the detectors' abilities to determine incoming neutron direction vectors using simulated particle reactant data. A sensitivity analysis was also performed by modifying the energy and direction vector of the simulated output data for 7Li* particles. Deviation coefficients showed a respective angular uncertainty of 1.86 degrees and 6.07 degrees for the boron loaded cloud chamber and a boron loaded semiconductor detectors. These capabilities were used to propose a possible RPM design that could be implemented.

Master of Science / Department of Mechanical and Nuclear Engineering / J. Kenneth Shultis / Perforated neutron detectors suffer the unfortunate effect that their efficiency is a strong function of the direction of neutron incidence. It is found, by Monte Carlo simulation of many perforation shapes, that sinusoidal-type perforations greatly reduce the variation of detector efficiency. Detectors with rod-type perforations are modeled using a hybrid transport method linking the MCNP transport code and a specialized ion-transport code to calculate the probability that a neutron is detected. Channel, chevron, and sinusoidal perforations are modeled using other customized transport codes. Detector efficiency calculations are performed for neutrons incident at various polar and azimuthal angles. It is discovered that the efficiency losses of the detectors result from the decreasing solid angle subtended by the detector from the source and streaming through the detector at specific azimuthal angles. Detectors achieving an efficiency in excess of 10% and having a relatively flat ± 1% angular dependence in all azimuthal angles and polar angles between 0 and 60 degrees are predicted. Efficiencies up to 25% are achievable at the loss of directional independence. In addition to minimizing the directional dependence of the perforated detectors, the feasibility of developing a neutron detector for deployment in cargo containers to locate nuclear weapon pits is investigated using the MCNP transport code. The detector considered is a 7-mm diameter, 6LiF, rod-perforated detector surrounded in a cylinder of polyethylene. The optimum thicknesses of surrounding polyethylene, to maximize the response of the detector, is determined to be 10 cm of radial, 5 cm of front, and 5 cm of back polyethylene for end-on neutron incidence. Such a detector is predicted to produce a count rate between 12 and 15 cpm from a nuclear-weapon pit composed of 90% 239Pu and 10% 240Pu at a distance of 3 m. Side incidence is also considered, and the optimum moderator dimensions are 8 cm of radial, 10 cm of front, and 10 cm of back polyethylene that produce approximately the same count rate.

Doctor of Philosophy / Department of Mechanical and Nuclear Engineering / Douglas S. McGregor / A novel design of neutron sensor was investigated and developed. The perforated, or micro-structured, diode neutron sensor is a concept that has the potential to enhance neutron sensitivity of a common solid-state sensor configuration. The common thin-film coated diode neutron sensor is the only semiconductor-based neutron sensor that has proven feasible for commercial use. However, the thin-film coating restricts neutron counting efficiency and severely limits the usefulness of the sensor. This research has shown that the perforated design, when properly implemented, can increase the neutron counting efficiency by greater than a factor of 4. Methods developed in this work enable detectors to be fabricated to meet needs such as miniaturization, portability, ruggedness, and adaptability. The new detectors may be used for unique applications such as neutron imaging or the search for special nuclear materials. The research and developments described in the work include the successful fabrication of variant perforated diode neutron detector designs, general explanations of fundamental radiation detector design (with added focus on neutron detection and compactness), as well as descriptive theory and sensor design modeling useful in predicting performance of these unique solid-state radiation sensors. Several aspects in design, fabrication, and operational performance have been considered and tested including neutron counting efficiency, gamma-ray response, perforation shapes and depths, and silicon processing variations. Finally, the successfully proven technology was applied to a 1-dimensional neutron sensor array system.

The purpose of this research is to determine the feasibility of creating an affordable and durable neutron detector for usage in the field surveys, site inspections, and transportation hub monitoring. Currently, organic scintillating detectors are an established method of detecting neutrons but are either costly, fragile solids like stilbene, or flammable liquids like benzene. In this work, several scintillation mixtures were tested with a PuBe source, which emits both neutrons and gamma rays. The pulse shape discrimination method was utilized to separate the signal pulses created from the mixed radiation field of the PuBe source. Two candidate mixtures were selected for solidification with elastomers for their verified neutron detection capabilities. The solid detectors measured high energy neutrons and gamma rays from the PuBe source. The solidified detectors have a Figure of Merit for separating neutrons of 0.859 ±0.419 and cost $0.13 per gram, while commercially available stilbene separates neutrons from gammas with a Figure of Merit of 4.70 and costs $64.36 per gram. This research shows that it is feasible to create affordable solid organic scintillators sensitive to high energy neutrons.

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.

Doctor of Philosophy / Department of Mechanical and Nuclear Engineering / Douglas S. McGregor / The 3He gas shortage for neutron detection has caused an increase in research efforts to develop viable alternative technologies. 3He neutron detectors cover areas ranging from 10–1000 cm2 in cylindrical form factors and are ideal for many nuclear applications due to their high intrinsic thermal neutron detection efficiency (> 80%) and gamma-ray discrimination (GRR ≤ 1 x 10-6) capabilities. Neutron monitoring systems for nuclear security applications include Radiation Portal Monitors (RPM’s), backpack, briefcase, and hand-held sensors. A viable replacement technology is presented here and compares three neutron detectors, each with different neutron absorber materials, to current 3He standards. These materials include Li and/or B silica aerogels, LiF impregnated foams, and metallic Li foils. Additionally, other neutron absorbing materials were investigated in this work and include LiF coated Mylar, B foils, BN coated carbon foam, and BN coated plastic honeycomb. From theoretical calculations, the Li foil material showed the greatest promise as a viable 3He alternative, thus a majority of the research efforts were focused on this material. The new neutron detector was a multi-wire proportional counter (MWPC) constructed using alternating banks of anode wires and 95% enriched 6Li foils sheets spaced 1.63 cm apart. In total, six anode banks and five layers of foil were used, thus an anode wire bank was positioned on each side of a suspended foils. Reaction products from the 6Li(n,α)3H reaction were able to escape both side of a foil sheet simultaneously and be measured in the surrounding gas volume concurrently. This new concept of measuring both reaction products from a single neutron absorption in a solid-form absorber material increased the intrinsic thermal neutron detection efficiency and gamma-ray discrimination compared to coated gas-filled detectors. Three different sizes of Li foil MWPC neutron detectors were constructed ranging from 25–1250 cm2 and included detectors for RPM’s, backpacks, and hand-held systems. The measured intrinsic thermal neutron detection efficiency of these devices was approximately 54%, but it is possible to exceed 80% efficiency with additional foils. The gamma-ray discrimination abilities of the detector exceeded 3He tubes by almost three orders of magnitude (GRR = 7.6 x 10-9).

No description available.

A new set of thermal neutron detectors has been developed as a near term 3He tube replacement. The zinc oxide scintillator is an ultrafast scintillator which can be doped to have performance equal to or superior to 3He tubes. Originally investigated in the early 1950s, this room temperature semiconductor has been evaluated as a thermal neutron scintillator. Zinc oxide can be doped with different nuclei to tune the band gap, improve optical clarity, and improve the thermal neutron detection efficiency. The effects of various dopant effects on the scintillation properties, materials properties, and crystal growth parameters have been analyzed. Two different growth modalities were investigated: bulk melt grown materials as well as thin film scintillators grown by metalorganic chemical vapor deposition (MOCVD). MOCVD has shown significant advantages including precise thickness control, high dopant incorporation, and epitaxial coatings of neutron target nuclei. Detector designs were modeled and simulated to design an improved thermal neutron detector using doped ZnO layers, conformal coatings and light collection improvements including Bragg reflectors and photonic crystal structures. The detectors have been tested for crystalline quality by XRD and FTIR spectroscopy, for scintillation efficiency by photo-luminescence spectroscopy, and for neutron detection efficiency by alpha and neutron radiation tests. Lastly, a novel method for improving light collection efficiency has been investigated, the creation of a photonic crystal scintillator. Here, the flow of optical light photons is controlled through an engineered structure created with the scintillator materials. This work has resulted in a novel radiation detection material for the near term replacement of 3He tubes with performance characteristics equal to or superior to that of 3He.

Despite nuclear fission prominence in nuclear physics, there are still several fundamental open questions about this process. One uncertainty is the energy distribution of neutrons emitted immediately after fission. In particular the relative energy distribution of neutrons above 8~MeV has been difficult to measure. This experiment measured the prompt neutron energy spectrum of n+235U from 3-10~MeV. The measurement took place at Los Alamos National Laboratory (LANL) and used a double time-of-flight technique to measure both the beam and fission neutron kinetic energies. Fission event timing was measured with a parallel plate avalanche counter. The fission neutron time-of-flight was measured with 2~m long plastic scintillation detectors. By combining the time-of-flight information with a known flight path the kinetic energy spectrum of neutrons was measured. To eliminate backgrounds various time-of-flight and energy cuts were imposed and an accidental coincidence background was subtracted. An MCNP simulation, including the 2~m neutron detector geometry, was done using the Madland and Nix model as the input energy distribution for the simulated neutrons. Finally, the measured energy spectrum was compared with the MCNP simulated n+235U fission neutron energy spectrum.