Design and Simulation of Neutron Detectors PDF Download

Author: Mohamed Luis El-Sheikh
Publisher:
ISBN:
Category :
Languages : en
Pages : 186

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Author: Manoj Kumar Parida
Publisher:
ISBN:
Category : Technology & Engineering
Languages : en
Pages : 0

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Author: Christian W. Fabjan
Publisher: Springer Nature
ISBN: 3030353184
Category : Elementary particles (Physics).
Languages : en
Pages : 1083

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Book Description
This second open access volume of the handbook series deals with detectors, large experimental facilities and data handling, both for accelerator and non-accelerator based experiments. It also covers applications in medicine and life sciences. A joint CERN-Springer initiative, the "Particle Physics Reference Library" provides revised and updated contributions based on previously published material in the well-known Landolt-Boernstein series on particle physics, accelerators and detectors (volumes 21A, B1,B2,C), which took stock of the field approximately one decade ago. Central to this new initiative is publication under full open access

Author:
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Category :
Languages : en
Pages :

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Neutron detectors are simulated using Monte Carlo methods in order to gain insight into how they work and optimize their performance. Simulated results for a Micromegas neutron beam monitor using a custom computer code are compared with published experimental data to verify the accuracy of the simulation. Different designs (e.g. neutron converter material, gas chamber width, gas pressure) are tested to assess their impact on detector performance. It is determined that a 10B converter foil and 1 mm drift gap width work best for a neutron beam monitor. The Micromegas neutron beam monitor neutronics are evaluated using the computer code MCNP. An optimized set of design criteria are determined that minimize neutron scattering probability in the device. In a best-case scenario, the thermal neutron scattering probability in the detector is 1.1*10−3. Lastly, composite neutron scintillators consisting of fluorescent dopant particles in a lithiated matrix material are simulated using a custom Monte Carlo code. The effects of design parameters such as dopant particle size, dopant volumetric concentration, and dopant and matrix material densities on scintillator characteristics are quantified. For ZnS:Ag particles in a lithiated glass matrix, it is found that dopant particle radii of 1 micron or less result in approximately Gaussian-shaped pulse height spectra and dopant particle radii of 5 microns or less result in practically all neutron absorption events producing scintillation light emission. Self-absorption of scintillation light is not treated in the simulation. Both the Micromegas and composite neutron scintillator simulations use the TRIM code as a heavy-charged particle transport engine.

Author: Floriana Fasolo
Publisher:
ISBN:
Category :
Languages : en
Pages : 185

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Author: Wenkai Fu
Publisher:
ISBN:
Category :
Languages : en
Pages :

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The Transient REActor Test (TREAT) facility was restarted and will be used to test accident-tolerant fuels to improve nuclear reactor safety. In this work, alternative neutron detectors for use in core and with the hodoscope at the TREAT facility were modeled and simulated using different computational tools to understand the underlying physics. The Hornyak button scintillation detector used in the original TREAT hodoscope to detect fast neutrons and its variants were evaluated using Geant4 to simulate the coupled nuclear and optical physics. The Hornyak-button model predicted an intrinsic efficiency of 0.35% for mono-directional fission neutrons and strong gamma-induced Cherenkov noise, which agree relatively well with the reported experimental observations. The proposed variants use silicon photomultipliers to reduce Cherenkov noise and have optimized layered or homogenized scintillation volumes. The layered and homogenized variants with 5-cm length were predicted to achieve neutron-detection efficiencies of 3.3% and 1.3%, respectively, at a signal-to-noise ratio of 100. Another candidate devices for the hodoscope are the actinide and hydrogenous microstructured semiconductor neutron detectors (MSNDs) evaluated using Geant4 and MCNP. With a sufficient rejection of the gamma noises, the U235 -filled and the hydrogenous MSNDs were predicted to yield neutron-detection efficiencies of 1.2% and 2.5%, respectively, at the length of 2 cm. The micro-pocket fission detectors (MPFDs) were developed to detect in-core neutrons, and the electron collection process in such devices was evaluated using Garfield++-based computational routine. The high-performance Garfield++ application was developed using the built-in, optimized element-search techniques and a hydrid MPI and OpenMP parallelization scheme. The preliminary results indicated that the averaged deposited energy per fission fragment was 7.15 MeV, and the induced current occured within 400 ns.

Author: Martin Rodney Williamson
Publisher:
ISBN:
Category :
Languages : en
Pages : 192

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Due to the eminent shortage of 3He, there exists a significant need to develop a new (or optimize an existing) neutron detection system which would reduce the dependency on the current 3He-based detectors for Domestic Nuclear Detection Office (DNDO) applications. The purpose of this research is to develop a novel methodology for optimizing candidate neutron detector designs using multivariate statistical analysis of Monte Carlo radiation transport code (MCNPX) models. The developed methodology allows the simultaneous optimization of multiple detector parameters with respect to multiple response parameters which measure the overall performance of a candidate neutron detector. This is achieved by applying three statistical strategies in a sequential manner (namely factorial design experiments, response surface methodology, and constrained multivariate optimization) to results generated from MCNPX calculations. Additionally, for organic scintillators, a methodology incorporating the light yield nonproportionality is developed for inclusion into the simulated pulse height spectra (PHS). A Matlab® program was developed to post-process the MCNPX standard and PTRAC output files to automate the process of generating the PHS thus allowing the inclusion of nonlinear light yield equations (Birks equations) into the simulation of the PHS for organic scintillators. The functionality of the developed methodology is demonstrated on the successful multivariate optimization of three neutron detection systems which utilize varied approaches to satisfying the DNDO criteria for an acceptable alternative neutron detector. The first neutron detection system optimized is a 3He-based radiation portal monitor (RPM) based on a generalized version of a currently deployed system. The second system optimized is a 6Li-loaded polymer composite scintillator in the form of a thin film. The final system optimized is a 10B-based plastic scintillator sandwiched between two standard plastic scintillators. Results from the multivariate optimization analysis include not only the identification of which factors significantly affect detector performance, but also the determination of optimum levels for those factors with simultaneous consideration of multiple detector performance responses. Based on the demonstrated functionality of the developed multivariate optimization methodology, application of the methodology in the development process of new candidate neutron detector designs is warranted.

Author: Diego Laramore
Publisher:
ISBN:
Category :
Languages : en
Pages :

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A semiconductor neutron imaging device is proposed (X-MSND) based on high efficiency, Micro-structured Semiconductor Neutron Detector (MSNDs) bump bonded onto a Timepix pixelated readout chip. The device serves as a combined neutron and photon imager with a 256x256 pixel array, using per-pixel time-over-threshold (ToT) energy deposition. The X-MSND design has produced thermal neutron detection efficiency of 14%, significantly greater than the theoretical maximum of less than 5% for planar devices. Simulated pixel clusters showed similar qualitative characteristics as planar neutron sensitive Timepix hybrid detectors. A workflow for simulating the semiconductor physics, ionization by radiation, and charge carrier transport for micro-structured sensors in general has been devised and described in this work. The simulation workflow began by steady state initial conditions of the X-MSND PIN diode sensor at full bias using COMSOL Multiphysics. This simulation step served the combined purpose of providing the necessary electric field solutions used in charge transport, and provided necessary design and operating parameters for device fabrication. Radiation transport code Geant4 was used to simulate the radiation detection characteristics of the sensor: thermal neutron detection efficiency, energy deposition per detection event, and the location of the ionized charge cloud per interaction are all calculated at this step. Dassault SolidWorks was used to generate Computer-Assisted Drafting (CAD) models of the full micro-structured device geometry, which was then converted and imported into a format that can be interpreted by Geant4 for the radiation transport. Geant4 can then interface directly with a modified version of Allpix^2 to perform charge carrier drift over the entire X-MSND geometry. Fabricated X-MSND devices were tested and evaluated at Kansas State University (KSU), and were found capable of producing high quality radiographs with both X-rays and neutrons. The fabrication of functioning X-MSND/Timepix assemblies was a collaborative effort among several research groups at KSU, domestic and international industrial partners, and international research groups. Fabrication of the X-MSND sensors was performed largely by Radiation Detection Technologies, Inc., with parametric design support from the Radiological System Integration Laboratory (RSIL) and Radiological Engineering Analysis Laboratory (REAL). The processes used in the production of X-MSND sensors are conventional micro-electro-mechanical system (MEMS) photolithography techniques; spin-on deposition and ultraviolet development of photoresist, metal lift-off, and wet etching of silicon are all used over the course of fabrication, and are described in detail. The Electronics Design Laboratory (EDL) of Kansas State University assisted in the design of custom printed circuit boards to which the X-MSND/Timepix assemblies are mounted. External to facilities located at KSU, various industrial manufacturing partners who specialize in Very Large Scale Integration (VLSI) and micro-fabrication assembly processes were also contracted to perform the specialized assembly processes required in assembling the full X-MSND/Timepix systems.

Author: Edward K. Kropp
Publisher:
ISBN:
Category : Neutron counters
Languages : en
Pages : 94

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A self-powered neutron detector (SPND) is a device that, coupled with a current meter, provides a readout proportional to neutron population. This thesis discusses the design parameters of an array of such devices, their characteristics, and the use of these devices as a self-powered neutron spectrometer (SPNS) to provide information about the energy distribution in a neutron radiation field. Neutron absorption in an appropriate material produces subsequent beta emissions. In a SPND, some of these beta particles will cross a non-conducting region and stop in a collector material. A net exchange of charge between these regions can be read as a current flowing between the emission region and the collector region. One potential SPNS design was modeled using a Monte Carlo simulation of the device's interaction with a radiation field. The Monte Carlo program used predicts the beta flux which is proportional to the current that would be produced by an actual device. Various beta emitting materials were considered for this device, and a sensitivity study of each was included. The design considered is comprised of a concentric set of these cylindrical SPND detector elements which, in themselves, are currently available technology.

Author: Simon R. Bolding
Publisher:
ISBN:
Category :
Languages : en
Pages :

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Simulations were performed using MCNP5 to optimize the geometry of a neutron spectrometer. The cylindrical device utilizes micro-structured neutron detectors encased in polyethylene moderator to identify sources based on energy spectrum. Sources are identified by comparison of measured detector responses to predetermined detector response templates that are unique to each neutron source. The design of a shadow shield to account for room scattered neutrons was investigated as well. For sufficient source strength in a void, the optimal geometric design was able to detect all sources in 1000 trials, where each trial consists of simulated detector responses from 11 unique sources. When room scatter from a concrete floor was considered, the shadow shield corrected responses were capable of correctly identifying 96.4% of the simulated sources in 1000 trials using the same templates. In addition to spectrometer simulations, a set of neutron multiplicity experiments from a plutonium sphere with various reflector thicknesses were simulated. Perturbations to nuclear data were made to correct a known discrepancy between multiplicity distributions generated from MCNP simulations and experimental data. Energy-dependent perturbations to the total number of mean neutrons per fission [average velocity] of 239Pu ENDF/B-VII. 1 data were analyzed. Perturbations were made using random samples, correlated with corresponding covariance data. Out of 500 unique samples, the best-case [average velocity] data reduced the average deviation in the mean of multiplicity distributions between simulation and experiment to 4.32% from 6.73% for the original data; the average deviation in the second moment was reduced from 13.87% to 8.74%. The best-case [average velocity] data preserved k[subscript]e[subscript]f[subscript]f with a root-mean-square deviation (RMSD) of 0.51% for the 36 Pu cases in the MCNP validation suite, which is comparable to the 0.49% RMSD produced using the original nuclear data. Fractional shifts to microscopic cross sections were performed and multiplicity and criticality results compared. A 1.5% decrease in fission cross section was able to correct the discrepancy in multiplicity distributions greater than the [average velocity] perturbations but without preserving k[subscript]e[subscript]f[subscript]f.