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检索条件"机构=Nuclear Radiological Engineering Program of Georgia Institute of Technology"

共 337 条记录，以下是1-10 订阅

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Global Core Environment Effects on Multigroup Cross Sections for Finer Energy Structures within Microreactors

Global Core Environment Effects on Multigroup Cross Sections...

引用

2024 International Conference on Physics of Reactors, PHYSOR 2024

作者： Sommer, Aaron Rahnema, Farzad Ougouag, Abderrafi M. Zhang, Dingkang DeHart, Mark Georgia Institute of Technology Nuclear & Radiological Engineering and Medical Physics Programs AtlantaGA United States

ISBN: (纸本)9780894487972

This paper uses a combined single-assembly and full-core benchmark problem based on a heat-pipe-cooled microreactor design to study the effect of core environment on homogenized multigroup macroscopic cross sections generated in a fifteen-group energy structure by the continuous-energy Monte Carlo code Serpent 2. These cross sections are evaluated in fuel pins, moderator rods, and heat pipes in selected locations within the single assembly and those corresponding locations in multiple such fuel assemblies at a set of locations within the full core. Effects of core environment on multigroup cross sections are determined using the obtained results. As expected, it is found that the effects of core environment on fifteen-group cross section data are much less marked than the same effects observed in published three-group cross sections. This finding applies across several reaction types in all selected materials. The maximum effect due to core environment found is 7.5% using fifteen-group cross section data versus 26.2% using three-group data. However, the effects remain significant even in the fifteen-group structure, which unequivocally demonstrates the imperative need for applying correction methods. Typical energy bins and reactions where this prominently occurs are intermediate and fast capture and fission in fuel, and intermediate scattering in heat pipes. This study confirms that many-group cross section libraries obtained from single assembly transport calculations produce full-core modeling results that are less sensitive to spectral effects arising from core environment than those generated using fewer group libraries. This study also demonstrates that on-the-fly correction of these libraries is still required for accurate modeling of an entire microreactor core. © 2024 AMERICAN nuclear SOCIETY. All rights reserved.

关键词： Serpentine

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Preliminary Full-Core Analysis of nuclear Thermal Propulsion Reactors using Nodal Diffusion Codes

Preliminary Full-Core Analysis of Nuclear Thermal Propulsion...

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2024 nuclear and Emerging Technologies for Space, NETS 2024

作者： Smith, Jake Kotlyar, Dan Georgia Institute of Technology George W. Woodruff School Nuclear and Radiological Engineering AtlantaGA30332-0405 United States

ISBN: (纸本)9780894487989

This paper presents preliminary full-core neutronics analysis for nuclear Thermal Propulsion (NTP) reactors using nodal diffusion codes. NTP reactor physics analyses generally rely on high-fidelity simulations (e.g., Monte Carlo (MC)), which are computationally expensive and can be prohibitive to use in transient simulations. The primary objective of this study is to develop the framework for an approach to NTP neutronics analysis that is better aligned with conventional methods used in full core modeling of traditional light water reactor systems. This study is only the first step to using a traditional simulation sequence, according to which the full core level simulation is preceded by macroscopic cross-section generation stage. Here, the cross sections (one-group) are generated using a single full-core MC simulation and fed into a nodal diffusion code to obtain the multiplication factor and spatial flux distribution. The homogenous solution obtained by the nodal diffusion code, DYN3D, yields near perfect agreement with the heterogeneous result only when complemented by on-the-fly generated correction parameters, while requiring significantly less computational expense. © 2024 AMERICAN nuclear SOCIETY. All rights reserved.

关键词： Light water reactors

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Preliminary Sensitivity Studies using a Newly Developed System Code for NTP Analysis

Preliminary Sensitivity Studies using a Newly Developed Syst...

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2024 nuclear and Emerging Technologies for Space, NETS 2024

作者： Myers, R. Kotlyar, D. Georgia Institute of Technology George W. Woodruff School Nuclear and Radiological Engineering AtlantaGA30332-0405 United States

ISBN: (纸本)9780894487989

nuclear thermal propulsion is an evolving technology that can be utilized for long-distance space travel. This technology yields the advantage of high thrust efficiencies that provides an edge over the accustomed chemical propulsion processes. Therefore, an examination of potential application arrangements is necessary to enhance its feasibility. The development of nuclear thermal propulsion requires a comprehensive understanding of the property developments throughout operation. This paper investigates the operational flow of the system. To do so involves full implementation of the turbopump and appropriate system. This includes linking the tank, chamber, fuel, moderator, nozzle, reflector, and piping to the turbopump. Analysis of the package capability is done through sensitivity studies and multidimensional thermal-hydraulic analysis. Specifically, the effect of the pressure ratio on the flow and thermal characteristics is studied. © 2024 AMERICAN nuclear SOCIETY. All rights reserved.

关键词： nuclear propulsion

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Computational Modeling of Radiation Effects in Carbon Fiber-Reinforced Polyethylene Composites 38

Computational Modeling of Radiation Effects in Carbon Fiber-...

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38th Technical Conference of the American Society for Composites, ASC 2023

作者： Li, Fuyue Woodrough, Mary Deo, Chaitanya School of Materials Science and Engineering Georgia Institute of Technology North Avenue AtlantaGA30332 United States Nuclear and Radiological Engineering Program GWW School of Mechanical Engineering Georgia Institute of Technology North Avenue AtlantaGA30332 United States

ISBN: (纸本)9781605956916

Radiation effect in carbon fiber-filled polyethylene composites are of importance in space and aircraft applications. A two-fold approach is employed to study radiation effects in carbon nanotube polyethylene as the exemplar. First, a radiation transport model is created with a single-wall carbon nanotube (SWCNT) as thin carbon slab superimposed on a slab of medium-density polyethylene matrix. Different weight fractions of SWCNT are examined and compared with an integrated model for the shielding abilities of galactic cosmic radiation. Second, molecular dynamics simulation with bond-order force field are conducted for 350 eV carbon recoil cascade simulation. Preliminary results discuss the effect of SWCNT on the radiation damage of crystalline polyethylene. Transport simulation demonstrate the efficacy of the slab model in predicting the shielding abilities and concluded that the dose increases with increasing SWCNT filler. Molecular dynamics simulations demonstrate that adding SWCNT enhances the radiation resistance with fewer free molecules produced. © 2023 by DEStech Publications, Inc. and American Society for Composites. All rights reserved.

关键词： Single-walled carbon nanotubes (SWCN)

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Development of a Hardware-in-the-Loop Testbed Using a Full-scope nuclear Power Plant Simulator for Instrumentation & Control and Cybersecurity Education, Training, and Research

Development of a Hardware-in-the-Loop Testbed Using a Full-s...

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2023 Conference on nuclear Training and Education: A Biennial International Forum, CONTE 2023

作者： Chen, Xiangyi Zhang, Fan Nuclear and Radiological Engineering George W. Woodruff School of Mechanical Engineering Georgia Institute of Technology AtlantaGA30332 United States

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Cask Design and Offline Processing of Fuel Salt via Fluorination and Impurity Absorption for MSR nuclear Waste (COFFIN)

Cask Design and Offline Processing of Fuel Salt via Fluorina...

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2023 Transactions of the American nuclear Society Winter Conference and Expo, ANS 2023

作者： Eccleston, Gracie Shade, Jana Welchel, Zach Wilson, Caryanne G. W. Woodruff School Georgia Institute of Technology Nuclear and Radiological Engineering AtlantaGA30332-0745 United States

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Risk-Informed Comparison of Atmospheric Plume Models for Dose-Based Advanced Reactor Licensing Siting 18

Risk-Informed Comparison of Atmospheric Plume Models for Dos...

引用

18th International Probabilistic Safety Assessment and Analysis, PSA 2023

作者： Wang, Jeffrey Clayton, Daniel Dewji, Shaheen Azim Nuclear and Radiological Engineering and Medical Physics Programs Georgia Institute of Technology AtlantaGA United States Sandia National Laboratories AlbuquerqueNM United States

ISBN: (纸本)9780894487927

This study conducts a comparison of atmospheric dispersion models for use in future advanced reactor licensing and Emergency Planning Zone (EPZ) siting. The comparison utilized the existing SOARCA Peach Bottom boiling water reactor (BWR) unmitigated long-term accident scenario source term and site location along with National Oceanic and Atmospheric Administration (NOAA) North American Mesoscale (NAM) 12 km weather data to simulate mean radial peak International Commission on radiological Protection (ICRP) Publication 60 effective dose distributions. Two atmospheric transport and dispersion modeling techniques were compared using the Maccs code: (1) a conventional Gaussian straight-line plume segment model (GPM) built into Maccs, and (2) a newly released integration with HYSPLIT using Lagrangian particle in cell dispersion. The dose distributions were deconstructed into individual exposure pathways to investigate the significant differences caused by the dispersion models. Results demonstrated that in the far-field case, Maccs and HYSPLIT estimate similar mean peak dose quantities and, designating a 0.01 Sv threshold from the NRC approved NuScale risk-informed EPZ technical basis, Maccs and HYSPLIT both place the EPZ limit far beyond current regulatory requirements. However, in the near-field case within a few kilometers of the facility boundary, HYSPLIT predicts a dose distribution several orders of magnitude larger than Maccs does. While HYSPLIT may be based in more realistic assumptions and physical phenomena compared to Gaussian Maccs, the utilization of coarse underlying weather grid sizes and lack of specific near-field dispersal modeling results in overly conservative estimation of consequences near to site boundaries. For advanced reactors such as small and micro reactors, whose source terms generate plumes not expected to reach beyond the near-field, HYSPLIT modeling could result in prohibitively high dose distributions. © 2023 Proceedings of 18th Internati

关键词： Dispersion Gaussian Lagrangian Maccs Near-field

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In-Vivo Detector Arrays for Time-Dependent Inhalation Radiation Contamination Assessment

In-Vivo Detector Arrays for Time-Dependent Inhalation Radiat...

引用

IEEE Symposium on nuclear Science (NSS/MIC)

作者： Y. Wei S. A. Dewji Nuclear and Radiological Engineering and Medical Physics Programs Georgia Institute of Technology Atlanta Georgia United States of America

ISBN: (数字)9798350388152

ISBN: (纸本)9798350388169

In response to radiological or nuclear emergencies, prompt assessment of internal radiation contamination is critical for triage assessment and medical countermeasures administrations. The NAIS-2’2 NaI(Tl) scintillation detectors were selected to construct a scanning system to measure internalized gamma-ray emitting radionuclide contamination in warfighters. A mesh-type adult female anthropomorphic phantom of 165 cm height and 65 kg weight was used, and an eight-point detector grid was implemented covering critical body areas (head, neck, chest, abdomen, and hip) implemented in the Monte-Carlo simulation tool, PHITS. The gamma depositions from critical radionuclides were simulated across all organs of interests. Biokinetic modeling data of each radionuclide were incorporated to determine the time-dependent retention in these organs and the gamma depositions in the detectors. To optimize the number and position of detectors for contamination characterization, a scoring algorithm was developed by analyzing the pseudo planes across the anterior and posterior of the phantom. These planes were subdivided into fine meshes to simulate the gamma flux, which were then integrated with biokinetics to compute the detector counting efficiencies at various times post-inhalation. This integration will enable the identification of the optimal detector array configuration, achieving the best counting statistics for various combinations of radionuclide sources, times post-exposure, and phantom morphologies, and the reconstruction of total contamination patterns.

关键词： Semiconductor device measurement Monte Carlo methods Semiconductor detectors Phantoms Morphology Detectors Biological systems

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Development of Tunnel Oxide Passivating Contact for Next-generation Silicon Detectors

Development of Tunnel Oxide Passivating Contact for Next-gen...

引用

IEEE Symposium on nuclear Science (NSS/MIC)

作者： Y. Tao M. Duce A. Erickson Georgia Institute of Technology Nuclear and Radiological Engineering Atlanta Georgia United States of America

Conventional silicon radiation detectors suffer from severe carrier recombination in the heavily doped p + and n + layers or regions, as well as beneath the metal contact areas, especially on the detector side, which is also known as "dead layer". Here we present the tunnel oxide passivating contact with an outstanding carrier selectivity. Its key element is an ultra-thin (~1.5 nm) interfacial oxide layer that can allow an efficient transportation of majority carriers via tunneling, while effectively block the minority carriers. The extraordinarily low saturation current density of 5 ~ 10 fA/cm 2 even with the metal contact (electrode), which was measured by the quasi-steady-state photoconductance, proves the superiority of tunnel oxide passivating contact. While conventional p-n junction or high-low junction have the saturation current density in the range of 10 ~ 90 fA/cm 2 for the passivated p + and n + layers due to Auger recombination, and 800 ~ 6000 fA/cm 2 for the metal contact areas because of significant metal-induced recombination at the interface. This demonstrates the potential and supremacy of applying the tunnel oxide passivating contact for silicon radiation detectors to thoroughly collect the generated charge carriers, leading to much higher energy resolution.

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Overview of the Development and Verification of the ntpThermo Code for the Design of nuclear Thermal Propulsion Reactors

Overview of the Development and Verification of the ntpTherm...

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2022 Transactions of the American nuclear Society Winter Meeting and technology Expo, ANS 2022

作者： Krecicki, Matt Kotlyar, Dan Georgia Institute of Technology George W. Woodruff School Nuclear and Radiological Engineering AtlantaGA30332-0405 United States

The methodology and results summarized in this paper demonstrate that the ntpThermo code is a suitable reduced-order code for engineering and full-core multiphysicsanalysis of NTP reactors. Additionally, the analysis presented in this paper demonstrates the need to account for T/M feedback on T/H solutions of NTP reactors. © 2022 American nuclear Society. All rights reserved.

关键词： Codes (symbols)

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