Jacob Gorton

Followers

Following

Public Views

Jason Hales

Idaho National Laboratory

University of Illinois at Urbana-Champaign

Shripad Revankar

Uploads

Papers by Jacob Gorton

Multiphysics Assessment of Accident Tolerant Fuel, Cladding, and Core Structural Material Concepts

The severe accident at the Fukushima-Daiichi nuclear power plant in 2011 ignited a global researc... more The severe accident at the Fukushima-Daiichi nuclear power plant in 2011 ignited a global research and development effort to replace traditionally-used materials in Light Water Reactors (LWRs) with Accident Tolerant Fuel (ATF) materials. These materials are intended to extend the coping time of nuclear power plants during severe accident scenarios, but must undergo thorough safety and performance evaluations before being implemented. Four ATF concepts are analyzed in this dissertation using state-of-the-art computer modeling tools: (1) iron-chromium-aluminum (FeCrAl) fuel rod cladding, (2) silicon carbide (SiC) fiber-reinforced, SiC matrix composite (SiC/SiC) boiling water reactor (BWR) channel boxes, (3) mixed thorium mononitride (ThN) and uranium mononitride (UN) fuel, (4) and UO2 [uranium dioxide] with embedded high thermal conductivity Mo inserts. The goals and approaches used for each study differed, and portions of this dissertation focused on verifying the accuracy of advanced modeling tools. Although each ATF evaluation is distinct, the underlying theme is the enhancement of safety, efficiency, and economic competitiveness of nuclear power through the use of advanced modeling techniques applied to material characterization studies. Results from the evaluations show the pros and cons of each ATF concept and highlight areas of needed modeling development. Comparisons of simulated and experimental critical heat flux (CHF) data for FeCrAl cladding and subsequent sensitivity analyses emphasized differences between real-world and simulated post-CHF phenomena. The Virtual Environment for Reactor Applications (VERA) multiphysics modeling suite was verified against other widely-used modeling tools for BWR application, and its advanced features were used to generate boundary conditions in SiC/SiC channel boxes used for deformation analyses. Several ThN-UN mixtures were analyzed using reactor physics and thermal hydraulic techniques and were shown to significantly increase the margin to fuel melt compared with UO2 [uranium dioxide] in LWRs. Mo inserts for UO2 [uranium dioxide] were optimized using sensitivity regression techniques and were also shown to significantly increase the margin to fuel melt compared with traditional UO2 [uranium dioxide]. v

Download

A review of neutronics and thermal hydraulics–based screening methods applied to accelerated nuclear fuel qualification

Progress in Nuclear Energy

Sensitivity Analysis of TRISO Fuel Configuration in a MiniFuel Irradiation Experiment

OSTI OAI (U.S. Department of Energy Office of Scientific and Technical Information), Nov 1, 2022

Simulation of a TRISO MiniFuel irradiation experiment with data-informed uncertainty quantification

Nuclear Engineering and Design

Thermal Optimization of UO2-Mo Fuel Using Sensitivity Analysis and Genetic Algorithms

OSTI OAI (U.S. Department of Energy Office of Scientific and Technical Information), Oct 1, 2022

Thermal Hydraulics of Accident Tolerant Fuel Concepts and a Preliminary Demonstration of CASL’s Coupled Tools for BWRs

MiniFuel Experimental Capability at Oak Ridge National Laboratory

U.S. Department of Energy Office of Scientific and Technical Information - OSTI OAI, Jun 1, 2022

Initial Design of High-Temperature MiniFuel Irradiation Experiments in the High Flux Isotope Reactor Removable Beryllium Region

U.S. Department of Energy Office of Scientific and Technical Information - OSTI OAI, Dec 1, 2021

Testing and Simulation of an Updated Cartridge Loop Vehicle

Status Report on Irradiation of MiniFuel Targets bearing TRISO Fuel Compacts

Download

Pre-conceptual high temperature gas cooled microreactor design utilizing two-phase composite moderators. Part II: Design space and safety characteristics

Progress in Nuclear Energy

Simulation of natural circulation cartridge loop experiments and application to molten salt reactors

Nuclear Engineering and Design

A Neutronics and Thermal Hydraulics–Based Approach for Accelerating the Screening and Qualification of Novel Nuclear Fuel Concepts☆

SSRN Electronic Journal

Evaluating accident tolerant fuel concepts in light water reactors using CASL CTF

Transactions of the American Nuclear Society, 2018

Design and Control of a Fueled Molten Salt Cartridge Experiment for the Versatile Test Reactor

Nuclear Science and Engineering, 2022

Comparison of experimental and simulated low-pressure CHF tests using CTF and RELAP5-3D

International Topical Meeting on Advances in Thermal Hydraulics 2018, ATH 2018 - Held in conjunction with the 2018 American Nuclear Society (ANS) Winter Meeting, 2018

A Demonstration of BWR Coupled Analysis and Potential ATF Applications Using CASL’s MPACT/CTF

Sensitivity of Critical Heat Flux for ATF FeCrAl Alloy Using RELAP5-3D and RAVEN

Analysis of Reactivity Accidents in Modular HTGRs

A control rod withdrawal (CRW) is a possible reactivity initiated accident (RIA) in modular high-... more A control rod withdrawal (CRW) is a possible reactivity initiated accident (RIA) in modular high-temperature gas-cooled reactors (mHTGRs). The purpose of this study is to perform a sensitivity analysis of a CRW event in an mHTGR model using the systems code RELAP5-3D with point kinetics feedback to demonstrate the impact of uncertainty in heat transfer and reactor kinetic parameters. The adaptive Sobol decomposition method in the uncertainty quantification code RAVEN is used to perform the sensitivity study and to determine which input parameters have the greatest impact on the figures of merit, which in this case are peak reactor power and maximum fuel temperature. This study addresses a need highlighted by the Nuclear Regulatory Commission (NRC) for transient fuel testing by quantifying the impact of uncertainty in heat transfer and reactor kinetic parameters and by generating potential boundary conditions for transient testing of conventional mHTGR fuel.

Assessment of CASL VERA for BWR analysis and application to accident tolerant SiC/SiC channel box

Nuclear Engineering and Design, 2020

Application of the Virtual Environment for Reactor Applications (VERA) to BWR analysis is assesse... more Application of the Virtual Environment for Reactor Applications (VERA) to BWR analysis is assessed in this paper by comparing results to those calculated using other widely-used modeling tools, namely the U.S. Nuclear Regulatory Commission's PARCS/PATHS and the Serpent Monte Carlo particle transport code. Additionally, VERA is used to calculate 3-D temperature and fast neutron flux distributions in silicon carbide (SiC) fiberreinforced, SiC matrix composite (SiC/SiC) BWR channel boxes, which are being studied as an Accident Tolerant Fuel core structural material concept. The code-to-code comparisons were favorable, and the SiC/SiC channel box evaluation demonstrates the many advanced modeling features VERA provides while also highlighting the non-uniformity in fast neutron flux distributions that can play a role in potential SiC/SiC channel box deformation. Traditional BWR analysis tools do not have the calculation fidelity necessary for coupled assessment of flux and temperature gradients in a SiC/SiC channel box. VERA is a state-of-the-art modeling environment that was developed to increase the safety and economic competitiveness of nuclear power through improved modeling accuracy. While VERA has already been deployed in the nuclear industry for PWR applications, the current study is a vital initial step in the extensive development, validation, and verification that VERA must go through to be useful for BWR applications.

Download

Multiphysics Assessment of Accident Tolerant Fuel, Cladding, and Core Structural Material Concepts

Download

A review of neutronics and thermal hydraulics–based screening methods applied to accelerated nuclear fuel qualification

Progress in Nuclear Energy

Sensitivity Analysis of TRISO Fuel Configuration in a MiniFuel Irradiation Experiment

OSTI OAI (U.S. Department of Energy Office of Scientific and Technical Information), Nov 1, 2022

Simulation of a TRISO MiniFuel irradiation experiment with data-informed uncertainty quantification

Nuclear Engineering and Design

Thermal Optimization of UO2-Mo Fuel Using Sensitivity Analysis and Genetic Algorithms

OSTI OAI (U.S. Department of Energy Office of Scientific and Technical Information), Oct 1, 2022

Thermal Hydraulics of Accident Tolerant Fuel Concepts and a Preliminary Demonstration of CASL’s Coupled Tools for BWRs

MiniFuel Experimental Capability at Oak Ridge National Laboratory

U.S. Department of Energy Office of Scientific and Technical Information - OSTI OAI, Jun 1, 2022

Initial Design of High-Temperature MiniFuel Irradiation Experiments in the High Flux Isotope Reactor Removable Beryllium Region

U.S. Department of Energy Office of Scientific and Technical Information - OSTI OAI, Dec 1, 2021

Testing and Simulation of an Updated Cartridge Loop Vehicle

Status Report on Irradiation of MiniFuel Targets bearing TRISO Fuel Compacts

Download

Pre-conceptual high temperature gas cooled microreactor design utilizing two-phase composite moderators. Part II: Design space and safety characteristics

Progress in Nuclear Energy

Simulation of natural circulation cartridge loop experiments and application to molten salt reactors

Nuclear Engineering and Design

A Neutronics and Thermal Hydraulics–Based Approach for Accelerating the Screening and Qualification of Novel Nuclear Fuel Concepts☆

SSRN Electronic Journal

Evaluating accident tolerant fuel concepts in light water reactors using CASL CTF

Transactions of the American Nuclear Society, 2018

Design and Control of a Fueled Molten Salt Cartridge Experiment for the Versatile Test Reactor

Nuclear Science and Engineering, 2022

Comparison of experimental and simulated low-pressure CHF tests using CTF and RELAP5-3D

International Topical Meeting on Advances in Thermal Hydraulics 2018, ATH 2018 - Held in conjunction with the 2018 American Nuclear Society (ANS) Winter Meeting, 2018

A Demonstration of BWR Coupled Analysis and Potential ATF Applications Using CASL’s MPACT/CTF

Sensitivity of Critical Heat Flux for ATF FeCrAl Alloy Using RELAP5-3D and RAVEN

Analysis of Reactivity Accidents in Modular HTGRs

Assessment of CASL VERA for BWR analysis and application to accident tolerant SiC/SiC channel box

Nuclear Engineering and Design, 2020

Download

Jacob Gorton

Related Authors

Uploads

Papers by Jacob Gorton

Log In