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All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

SAM Developments to Support Transient Safety Analysis of Advanced non-LWRs

System Analysis Module (SAM) is under development at Argonne National Laboratory as a modern system-level modeling and simulation tool for advanced non-light water reactor safety analyses. It utilizes the object-oriented application framework MOOSE to leverage the modern software environment and advanced numerical methods. The capabilities of SAM are being extended to enable the transient modeling, analysis, and design of various advanced nuclear reactor systems. This report summarizes major fiscal year 2019 (FY19) progress in SAM code development, capability enhancements, demonstration, and validation to support transient safety analysis of advanced non-LWRs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Transient and Steady-State Inverse Problems in Sierra/Aria

Inverse problems arise in a wide range of applications, whenever unknown model parameters cannot be measured directly. Instead, the unknown parameters are estimated using experimental data and forward simulations. Thermal inverse problems, such as material characterization problems, are often large-scale and transient. Therefore, they require intrusive adjoint-based gradient implementations in order to be solved efficiently. The capability to solve large-scale transient thermal inverse problems using an adjoint-based approach was recently implemented in SNL Sierra Mechanics, a massively parallel capable multiphysics code suite. This report outlines the theory, optimization formulation, and path taken to implement thermal inverse capabilities in Sierra within a unit test framework. The capability utilizes Sierra/Aria and Sierra/Fuego data structures, the Rapid Optimization Library, and an interface to the Sierra/InverseOpt library. The existing Sierra/Aria time integrator is leveraged to implement a time-dependent adjoint solver.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2019

Experimental and Modeling Investigation of Overcooling Transients that Include Freezing, in Fluoride-Salt Cooled High-Temperature Reactors (FHRs). Final Report

The goal of this project was an experimental and modeling investigation of overcooling transients that include freezing in Fluoride-Salt Cooled High-Temperature Reactors (FHRs). The most impactful product of this research collaboration between Professor Scarlat’s group at UW Madison and subsequently at UC Berkeley, and Dr. Rui Hu’s group at Argonne National Laboratory was the initial demonstration of SAM application to FHR transients, and the generation of a freezing module for the SAM code. The national laboratory and industry graduate student internships enabled by this project were valuable for workforce development, and also for knowledge transfer between university, national lab, and industry.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Transient Studies of Nonequilibrium Transport in Two-Dimensional Semiconductors

The overarching objective of research under DOE support has been to investigate emergent carrier phenomena in different two-dimensional (2D) systems, under conditions where they are driven far from equilibrium. The description of nonequilibrium transport has long represented one of the most enduring challenges in condensed-matter physics, in large part due to the need to invoke the role of complex many-body phenomena (electron-phonon and electron-electron coupling) in this regime. In this project, we have explored aspects of this problem in a variety of different two-dimensional systems. These include the modulation-doped two-dimensional electrons gas, or 2DEG, in GaAs/AlGaAs heterostructures; the electron and hole systems in graphene and its related hetero-structures, and; the strongly-confined electron layers in transition-metal dichalcogenide (MoS 2 and WS 2 ) and trichalcogenide (TiS 3 ) materials. Significant understanding has been achieved in a number of different areas, as we elaborate upon in subsequent sections of this proposal: Nanoscale spectroscopy of defect-related barriers in MoS2 Negative differential conductance due to inter-valley scattering in WS 2 under high fields Incipient localization due to scattering from frozen-phonon disorder in quantum point contacts High-field transient transport in graphene/hexagonal BN heterostructures Universal voltage scaling of quantum corrections in graphene under nonequilibrium A defining feature of much of this research has been an approach to problem solving in which novel experimental techniques (such as transient electrical pulsing and scanning-probe microscopy) are combined with the development (in collaboration with colleagues from condensed-matter theory) of sophisticated models, a strategy that we propose to continue in our future research. In recent years, we have developed an increasing capacity for handling atomically-thin materials, and their van der Waals systems, a capacity that is reflected in the balance of topics listed above.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2020

FY20 SAM Code Developments and Validations for Transient Safety Analysis of Advanced non-LWRs

The System Analysis Module (SAM) is under development at Argonne National Laboratory as a modern system-level modeling and simulation tool for advanced non-light water reactor safety analyses. It utilizes the object-oriented application framework MOOSE to leverage the modern software environment and advanced numerical methods. The capabilities of SAM are being extended to enable the transient modeling, analysis, and design of various advanced nuclear reactor systems. This report summarizes major progress in SAM code development, capability enhancements, demonstration, and validation to support transient safety analysis of advanced non-LWRs. Rapid developments continued in fiscal year 2020 (FY20) to support various needs of the advanced reactor community, especially the NRC and industry on the licensing safety analysis of advanced reactor designs. Significant code changes were made to provide various capability enhancements, bug fixes, and user friendliness improvements. Major code updates are summarized in Section 1, while four important enhancements are detailed in Sections 2-5, including a multi-dimension flow model; reactivity feedback and decay heat models; control and trip system modeling, and additional fluid and solid thermophysical property models. Code validation activities in FY20 include using test data from the Fast Flux Test Facility (FFTF), the High Temperature Test Facility (HTTF), and several separate effects test facilities for pebble-bed modeling.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Transient Thermal Analysis of Calorimeters Used in Characterization of the ACRR Radiation Environments

Silicon calorimeters have been used for active radiation dosimetry in the central cavity of the Annular Core Research Reactor (ACRR) for over a decade. Recently, there has been interest in using other materials for calorimetry to accurately measure the prompt gamma-ray energy deposition in the mixed neutron and gamma-ray environment. The calorimeters used in the ACRR use a thermocouple (TC) to measure the change in temperature of specific materials in the radiation environment. The temperature change is related to the instantaneous dose received by the material in a pulse-transient operation. SOLIDWORKS Simulation and ANSYS Mechanical were used to model the calorimeter and analyze the thermal behavior under pulse-transient conditions. This report compares the results from modeling to experimental results for selected calorimeter materials and radiation environments. These materials include bismuth, tin, zirconium, and silicon. Calorimeters assembled with each material were irradiated in the ACRR central cavity in the free- field, LB44, CdPoly, and PLG radiation environments. The neutronics code Monte-Carlo N- Particle (MCNP) was used to calculate the neutron and gamma-ray response of the calorimeter materials at the experimental locations in the central cavity. Different response tallies were used and found to give different results for the gamma-ray energy deposition. It was determined that performing the neutron/gamma-ray/electron transport in MCNP using the *F8 electron tally gave the overall best agreement with the experimental results. The *F8 tally, however, is much more computationally intensive than the neutron/gamma-ray transport calculations. Also, this report contains parametric analyses that examine the ways to improve the current design of the calorimeters. One finding from the parametric analysis was that the TC should be placed closer to the outer radius of the disks to obtain a measurement closer to the maximum temperature of the disk. Also, the parametric analysis showed that the most dominant mechanism of heat loss in the calorimeters is conduction through the alumina posts. In future designs, the conduction should be minimized to reduce the effect of heat loss on the measurements.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Early-Time (E1) High-Altitude Electromagnetic Pulse Effects on Transient Voltage Surge Suppressors

Determining the effectiveness of surge and pulse protection devices in the United States power grid against effects of a High-Altitude Electromagnetic Pulse (HEMP) is crucial in determining the present state of grid resilience. Transient Voltage Surge Suppressors (TVSS) are used to protect loads in substations from transient overvoltages. Designed to mitigate the effects of lightning, their response to a HEMP event is unknown and was determined. TVSSs were tested in two unique configurations using a pulser that generates pulses in the tens of nanoseconds scale to determine their protective capability as well as to determine their self-resilience against HEMP pulses. Testing concluded that TVSS devices adequately protect against microsecond scale pulses like lightning but do not protect against pulses resembling HEMP events. It suggests that TVSS devices should not be relied upon to mitigate the effects of HEMP pulses.

24 POWER TRANSMISSION AND DISTRIBUTION↗
DOE OSTI2021

Understanding Transients: Needs from the Laboratory Astrophysics Community

A growing number of astrophysical transients are pushing astronomers to develop increasingly complex computational tools to model both radiation hydrodynamics and electron transport. Here we review the physics needs and simulation uncertainties associated with modeling these transients. Although this review will focus on theory and simulation aspects of this problem, we also review some experiments designed to study this physics.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2021

Coupling of CTF and TRACE for Modeling of Transients

This report documents the improvements that have been made to the capabilities for coupling CTF to systems codes-specifically, the US Nuclear Regulatory Commission (NRC) TRACE code. An initial systems coupling capability had been set up previously using a nonoverlapping domain approach with the codes exchanging data at the core boundaries. The present work adds a new approach using overlapping domains, in which the system code models the core as well. A new input format has been added to allow the user to specify the physical quantities to be exchanged and their location in the system model, which gives the flexibility of applying one-way or two-way coupling between the codes using the desired data exchanges. In addition to applying thermal-hydraulics (T/H) boundary condition (BC) values obtained from TRACE, a capability was added to allow CTF to apply flow resistance feedback to TRACE to match the CTF core pressure drop. Support was added for executing parallel CTF models within the CTF-systems coupling. The system coupling capability was successfully applied to a parallel MSLB transient, demonstrating that both the one-way and two-way coupling behaved as expected and provided substantial improvements to numerical stability and routine compared to the previous nonoverlapping domain coupling. An initial capability was also developed for performing restart calculations in CTF which will be used in the future for restarting CTF-TRACE simulations at specific points in the transient simulation.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Online Optimization of NSLS-II Dynamic Aperture and Injection Transient (Facility Improvement Project)

The Facility Improvement Project (FIP) “Methods of online optimization of NSLS-II storage ring concurrent with user operations” proposed and approved in 2018 is now complete. The first goal is the online optimization of nonlinear beam dynamics to increase the beam lifetime and injection efficiency. We have developed a model-independent optimization technique using advanced algorithms. Using this technique, we increased the NSLS-II dynamic aperture by 20% and reduced the amplitude-dependent tune shift by a factor of two. We applied sextupole optimization to the new high-chromaticity lattice, which has been developed to improve the beam stability and to increase the single-bunch beam intensity. We were able to double the injection efficiency and exceed 90%. The second goal of this project is to provide the top-off injection with minimized perturbations of the beamline user operations. To achieve the transparent injection, we optimize the matching of four storage ring injection kickers reducing the perturbation of the stored beam orbit. The first application of the online optimization resulted in a reduction of the amplitude of residual beam oscillation by a factor of six, from 1300 μm down to 200 μm. Further improvement was limited by the timing jitter of the trigger boards. Then, we replaced all trigger boards with a new design and reduced the timing jitter by a factor of five, from 5 ns down to 1 ns. This upgrade resulted in the injection transient reduction from 200 μm to 120 μm. To optimize the full set of kicker parameters, including the trigger timing, amplitude, and pulse width, we upgraded all kicker power supplies with the capability of tunable waveform width. As a result, we have reduced the injection transient to the limit of 60 μm.

43 PARTICLE ACCELERATORS↗
DOE OSTI2023

Physics-Based Real-time Analysis and Control to Achieve Transient-Free Operations for the ITER Era (Final Report)

This is the final report for the early career award for Prof. Egemen Kolemen that aimed to develop the physics and technical basis of real-time transient – i.e. disruptions and Edge Localized Modes (ELMs) – avoidance and feedback control at the DIII-D nuclear fusion facility for the ITER era. During this project, we reconstructed high quality real-time DIII-D equilibria using kinetic information from the motional Stark effect (MSE), Thomson Scattering (TS), and Charge Exchange Recombination (CER) spectroscopy. These equilibria were analyzed in real-time to determine the low-n ideal MHD stability and the pedestal stability thresholds, and to obtain plasma response models. These calculations were used to understand which physical process evolves to lead to the onset of disruption in ITER-relevant scenarios at DIII-D, and how disruption can be predicted using stability analysis as opposed to the current methods of correlation of various signals, which are descriptive but lack predictive power. In the disruption studies, we focused on tearing modes, since these constitute the vast majority of disruptions in this regime. With this knowledge, real-time disruption avoidance algorithms were developed and tested. The enhanced situation awareness and physics insight were used in advanced real-time feedback control of pedestal and core profiles, which allowed the pursuit and sustainment of transient-free high-confinement regimes that cannot be achieved via feed-forward design based on in-between shot analysis.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Enhanced LWR High Burnup Transient Simulation Capabilities to Support AOO Margin Identification

As part of ongoing efforts to support the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program’s development of a fuel fragmentation, relocation, and dispersal (FFRD) screening methodology, a number of improvements are required for the NEAMS core simulation capabilities, namely the Virtual Environment for Reactor Applications (VERA). Three areas of improvement were identified in VERA which are important for continued development and application of the FFRD screening methodology. First, the FFRD screening methodology will soon be extended to boiling water reactors (BWRs), requiring development and validation of the VERA BWR capabilities. Second, the screening methodology occasionally requires that VERA be used to simulate a transient in addition to nominal operations. Thus, improvements to both accuracy and performance of the VERA transient capabilities are necessary. Third, the VERAOneWay component of VERA is used to develop BISON fuel performance inputs using the rod-by-rod histories calculated by VERA. Prior use of VERAOneWay exposed significant accuracy, robustness, and performance issues with VERAOneWay; these must be addressed for it to be an effective tool in the NEAMS FFRD methodology. This report documents the efforts in FY23 in each of these three areas to enable successful use of VERA and VERAOneWay for FFRD calculations in FY24 and following years.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Tech-X contributions to the Center for Tokamak Transients (CTTS) Scientific Discovery through Advanced Computing (SciDAC) project (Final Report)

The Center for Tokamak Transients Simulations (CTTS) consists of computational plasma physicists, computer scientists, and applied mathematicians who are together developing and applying simulation software to improve our understanding of and ability to predict and control transient events in tokamak discharges, particularly those that lead to violent disruptions. We have major simulation activities concerning disruptions that are caused by violating ideal MHD stability criteria, vertical displacement events, resistive wall modes, and tearing modes. We are performing comprehensive modeling of disruption mitigation techniques, particularly by shattered pellet injection (SPI). The Center is built around two mature extended MHD codes: NIMROD and M3D-C1. These codes are being extended in a number of ways as required by the demands of disruption modeling, and are being optimized to work well for larger problem size and modern computer architectures as required to model tokamak plasmas with realistic parameters over the required timescales. The codes will be coupled to the FronTier code for state-of-the-art modeling of SPI. Validation with experimental data on DIII-D, NSTX, and JET is being facilitated by rigorous uncertainty quantification (UQ). The high-level objectives of the center are to: (1) Develop better understanding and improved predictive capability of when crossing a linear stability boundary will likely lead to a disruption, (2) Provide a tool for calculating vessel forces, both axisymmetric and non-axisymmetric, for worst case VDE and other disruptions in tokamaks, (3) Develop increased understanding and improved predictive capability for locked mode disruptions and how best to avoid them, (4) Develop, verify, and validate 3D models for disruption mitigation by shattered pellets that can be used to design and optimize a system for future tokamaks, and (5) Increase the efficiency and scalability of the M3D-C1 and NIMROD codes so that they can model disruptions in ITER scale devices using the most powerful DOE HPC hardware, and (6) Demonstrate code-coupling via FronTier coupling with NIMROD and M3D-C1 and by interfacing these codes within a WDM framework.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2020

FAST-1.0: A Computer Code for Thermal-Mechanical Nuclear Fuel Analysis under Steady-state and Transients: Developed under NQA-1-2017

Fuel Analysis under Steady-state and Transients (FAST) is the US Nuclear Regulatory Commission (NRC)’s computer code that calculates the steady-state and transient response of nuclear reactor fuel rods during long-term in-reactor burnup, anticipated operational occurrences (AOOs), design basis accidents (DBAs), and dry storage conditions. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include: heat conduction through the fuel and other materials, heat transfer from the cladding-to-coolant, cladding elastic and plastic deformation, including creep, fuel-cladding mechanical interaction, fission gas release from the fuel, rod internal pressure and void volume, and cladding oxidation. The code contains necessary material and coolant properties, as well as clad-to-coolant heattransfer correlations, for normal operation through postulated accidents for today’s US-based light water reactor (LWR) fuel designs. FAST-1.0 also contains preliminary materials and models for new LWR fuel concepts, such as accident tolerant fuel (ATF), and non-LWR fuel concepts such as metallic fuels for sodium fast reactors (SFRs). FAST has been developed for use on Windows and Linux operating systems. This document describes FAST-1.0, which is the first official version of this code. This document is one of a series of documents on FAST; the other documents detail the material properties used by FAST as well as its integral assessment to experiments and commercial data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

FAST-1.1: A Computer Code for Thermal-Mechanical Nuclear Fuel Analysis under Steady-state and Transients (Developed under NQA-1-2017)

Fuel Analysis under Steady-state and Transients (FAST) is the U.S. Nuclear Regulatory Commission (NRC)’s computer code that calculates the steady-state and transient response of nuclear reactor fuel rods during long-term in-reactor burnup, anticipated operational occurrences (AOOs), design basis accidents (DBAs), and dry storage conditions. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include: * heat conduction through the fuel and other materials * heat transfer from the cladding-to-coolant * cladding elastic and plastic deformation, including creep * fuel-cladding mechanical interaction * fission gas release from the fuel * rod internal pressure and void volume * cladding oxidation The code contains necessary material and coolant properties, as well as clad-to-coolant heat transfer correlations, for normal operation through postulated accidents and AOOs for today’s U.S.-based light water reactor (LWR) fuel designs. FAST-1.1 also contains preliminary materials and models for new LWR fuel concepts, such as accident tolerant fuel (ATF), and non-LWR fuel concepts such as metallic fuels for sodium fast reactors (SFRs). FAST has been developed for use on Windows and Linux operating systems. This document describes FAST-1.1 and is one of a series of documents on the code; the other documents detail the material properties used by FAST as well as its integral assessment to experiments and commercial data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Center for Tokamak Transients Simulations

The major goals for the Center for Tokamak Transients Simulations (CTTS) were to develop detailed numerical simulations of disruptive transients in tokamak experiments and to use them in theoretical studies of the thermal-quench (TQ) and current-quench (CQ) phases of disruption. The University of Wisconsin-Madison component of CTTS contributed relevant development for the NIMROD code (https://nimrodteam.org) and applied it to disruptions that involve global vertical displacement. Code-development contributions include boundary conditions that represent sheath effects which occur when tokamak plasma is in contact with the surrounding vessel wall during vertical displacement events (VDEs). We also implemented a reduced model of energetic runaway electrons (REs) to simulate the effect of these energetic electrons on the CQ phase of tokamak disruptions. Our computations of VDEs include 2D and 3D verification through benchmarks with two other widely used macroscale plasma simulation codes, M3D-C1 and JOREK. Over nonlinear evolution to the final termination of plasma current, the 2D simulation results on magnetic axis location, toroidal current, and halo current track each other well. With the 3D benchmark computation, all three codes found the same qualitative behavior of the asymmetric instabilities having increasing growth rate after contact with the wall, followed by nonlinear excitation of other toroidal harmonics. Our NIMROD computations for an idealized configuration show agreement with an analytical assessment by V. D. Pustovitov [Nucl. Fusion 55, 113032 (2015)] that the plasma and non-ideal wall remain in force-balance such that net forces only result from magnetic stresses over the outer surface of the wall. Our study of the influence of boundary conditions on VDEs shows that extended-MHD simulations of VDEs are sensitive to electron energy transport to the wall and that boundary conditions on flow are only significant to the extent that they influence the electron energy transport. Resistive-MHD NIMROD simulations of the tokamak discharges in the Madison Symmetric Torus show that magnetic perturbations of poloidal wavenumber m=3 lead to chaotic magnetic topology over the edge region, which helps deconfine energetic electrons. A second area of RE study shows that the coupled resistive MHD/reduced RE model reproduces the resistive-hose particle-beam instability. With respect to educational opportunities, this effort involved and supported four graduate students at the University of Wisconsin-Madison and one postdoctoral associate. The results of this study have been disseminated through journal publications, conference presentations, technical reports, and PhD dissertations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Transient Testing of Nuclear Fuels Performed in the Original Operation of TREAT, Third Edition

TREAT core was upgraded at the end of the 1980s. The tests performed since about 1970 are summarized in considerable detail. Earlier tests are noted by their designated series and by some parameters that help to identify and distinguish each series from the others. The detailed summary descriptions are intended to help inform future researchers regarding the large, historical, empirical data base generated from that 35-year span during which hundreds of TREAT experiments were performed to investigate and demonstrate the transient behavior of a wide variety of nuclear fuels during severe off-normal and accident conditions -- using direct nuclear heating of the test samples. That empirical basis is foundational and valuable for guiding future qualification and safety evaluation of improved fuel designs and for future transient test planning and performance in TREAT. The second edition of this report added descriptions of the following test series: D, E, EOS, F, H, J, R, RX, and S. This third edition includes descriptions of an additional four test series: the HUT, HOP (& HUC), ORNL-TR, and PNL.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

FAST-1.2: A Computer Code for Thermal-Mechanical Nuclear Fuel Analysis under Steady-state and Transients

Fuel Analysis under Steady-state and Transients (FAST) is the U.S. Nuclear Regulatory Commission (NRC)’s computer code that calculates the steady-state and transient response of nuclear reactor fuel rods during long-term in-reactor burnup, anticipated operational occurrences (AOOs), design basis accidents (DBAs), and dry storage conditions. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include: • heat conduction through the fuel and other materials • heat transfer from the cladding-to-coolant • cladding elastic and plastic deformation, including creep • fuel-cladding mechanical interaction • fission gas release from the fuel • rod internal pressure and void volume • cladding oxidation The code contains necessary material and coolant properties, as well as clad-to-coolant heat-transfer correlations, for normal operation through postulated accidents and AOOs for today’s U.S.-based light water reactor (LWR) fuel designs. FAST-1.2 also contains preliminary materials and models for new LWR fuel concepts, such as accident tolerant fuel (ATF), and non-LWR fuel concepts such as metallic fuels for sodium fast reactors (SFRs). FAST has been developed for use on Windows and Linux operating systems. This document describes FAST-1.2 and is one of a series of documents on the code; the other documents detail the material properties used by FAST as well as its integral assessment to experiments and commercial data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗