Robert Montgomery

Magnetic hardening from the suppression of domain walls by nonmagnetic particles

Magnetic domain switching and hysteresis loops in a single crystal α-iron with and without nonmagnetic particles were simulated based on the Landau-Lifshitz-Gilbert equation of magnetization dynamics. Both the nonmagnetic particle and the 360° domain wall are nucleation sites of an antidirection domain during domain switching; however, the 360° Bloch domain wall is the easiest nucleation site. The nucleation occurs by splitting the 360° Bloch domain wall into two 180° Bloch domain walls. The existence of nonmagnetic particles could prevent the formation of 360° Bloch domain walls and cause magnetic hardening. Simulations demonstrate the impact of nonmagnetic particle sizes on magnetic domain switching and coercive field.

Use of multiscale zirconium alloy deformation models in nuclear fuel behavior analysis

Accurate prediction of cladding mechanical behavior is a key aspect of modeling nuclear fuel behavior, especially for conditions of pellet-cladding interaction (PCI), reactivity-initiated accidents (RIA), and loss of coolant accidents (LOCA). Current approaches to fuel performance modeling rely on empirical constitutive models for cladding creep, growth and plastic deformation, which are limited to the materials and conditions for which the models were developed. To improve upon this approach, a microstructurally-based zirconium alloy mechanical deformation analysis capability is being developed within the United States Department of Energy Consortium for Advanced Simulation of Light Water Reactors (CASL). Specifically, the viscoplastic self-consistent (VPSC) polycrystal plasticity modeling approach, developed by Lebensohn and Tome 1, has been coupled with the BISON engineering scale fuel performance code to represent the mechanistic material processes controlling the deformation behavior of light water reactor (LWR) cladding. A critical component of VPSC is the representation of the crystallographic nature (defect and dislocation movement) and orientation of the grains within the matrix material and the ability to account for the role of texture on deformation. A future goal is for VPSC to obtain information on reaction rate kinetics from atomistic calculations to inform the defect and dislocation behavior models described in VPSC. The multiscale modeling of cladding deformation mechanisms allowed by VPSC far exceed the functionality of typical semi-empirical constitutive models employed in nuclear fuel behavior codes to model irradiation growth and creep, thermal creep, or plasticity. This paper describes the implementation of an interface between VPSC and BISON and provides initial results utilizing the coupled functionality.

Materials Degradation and Detection (MD2): Deep Dive Final Report

An effort is underway at Pacific Northwest National Laboratory (PNNL) to develop a fundamental and general framework to foster the science and technology needed to support real-time monitoring of early degradation in materials used in the production of nuclear power. The development of such a capability would represent a timely solution to the mounting issues operators face with materials degradation in nuclear power plants. The envisioned framework consists of three primary and interconnected “thrust” areas including 1) microstructural science, 2) behavior assessment, and 3) monitoring and predictive capabilities. A brief state-of-the-art assessment for each of these core technology areas is discussed in the paper.

Progress towards developing neutron tolerant magnetostrictive and piezoelectric transducers

Current generation light water reactors (LWRs), sodium cooled fast reactors (SFRs), small modular reactors (SMRs), and next generation nuclear plants (NGNPs) produce harsh environments in and near the reactor core that can severely tax material performance and limit component operational life. To address this issue, several Department of Energy Office of Nuclear Energy (DOE-NE) research programs are evaluating the long duration irradiation performance of fuel and structural materials used in existing and new reactors. In order to maximize the amount of information obtained from Material Testing Reactor (MTR) irradiations, DOE is also funding development of enhanced instrumentation that will be able to obtain in-situ, real-time data on key material characteristics and properties, with unprecedented accuracy and resolution. Such data are required to validate new multi-scale, multi-physics modeling tools under development as part of a science-based, engineering driven approach to reactor development. It is n...

FY16 Status Report for the Uranium-Molybdenum Fuel Concept

The Fuel Cycle Research and Development program of the Office of Nuclear Energy has implemented a program to develop a Uranium-Molybdenum metal fuel for light water reactors. Uranium-Molybdenum fuel has the potential to provide superior performance based on its thermo-physical properties. With sufficient development, it may be able to provide the Light Water Reactor industry with a melt-resistant, accident-tolerant fuel with improved safety response. The Pacific Northwest National Laboratory has been tasked with extrusion development and performing ex-reactor corrosion testing to characterize the performance of Uranium-Molybdenum fuel in both these areas. This report documents the results of the fiscal year 2016 effort to develop the Uranium-Molybdenum metal fuel concept for light water reactors. FY16 Status Report for the Uranium-Molybdenum Fuel Concept September 20, 2016 v Acronyms and Abbreviations 14YWT and 9YWT steel cladding material that has been used in liquid metal reactors AOO Anticipated Operational Occurrence ATF Advanced Test Facility DOE-NE U.S. Department of Energy, Office of Nuclear Energy FCRD Fuel Cycle Research and Development Fe iron FY fiscal year LOCA Loss of Coolant Accident LWR light water reactors MA mechanical alloyed Nb niobium ODS Oxide Dispersion Strengthened OM Optical Microscopy PNNL Pacific Northwest National Laboratory psi pounds per square inch TE triple extrusion TWT thick-wall(ed) tube U-Mo Uranium-Molybdenum FY16 Status Report for the Uranium-Molybdenum Fuel Concept vi September 20, 2016