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At least 199 records · Page 11

Nuclear Requalification of the ATR Core

The Advanced Test Reactor (ATR) creates a unique, high neutron flux materials testing environment that subsequently causes neutron embrittlement damage to its beryllium reflector. The reflector is eventually required to be replaced during a reactor Outage. Concurrently, a piece-for-piece core replacement is conducted to maintain reactor functionality called a core internals changeout (CIC). Once changeout is complete, ATR Reactor Engineering utilizes previous CIC data, notably 1977-1994 CIC data, to conduct predictive criticality analysis on the new, unirradiated core. However, CIC data has been inconsistently reported over the years making it difficult, and in some cases impossible, to utilize. These inconsistencies arise in various forms of not using correct units, inconsistent diagram usage, and not portraying the same kind of data similarly across the reports. To work around this inconsistency issue, Reactor Engineering is implementing the use of several new neutronic modeling codes that utilize modern computation speeds and methods to model the new core in different criticality environments. Further validation is then conducted by nuclear testing and bringing the reactor critical and conducting various irradiation tests. Upon completion and certification that the reactor passed all testing requirements, ATR can once again re-start normal operations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Modeling of High-Temperature Corrosion of Zirconium Alloys Using the eXtended Finite Element Method (X-FEM)

Oxidation modeling in modern nuclear fuel performance codes is currently limited by the lack of coupling with mechanics, thus preventing proper description of how high-temperature oxidation impacts mechanical properties. This is mostly due to the fact that the finite difference formalism adopted in corrosion models is incompatible with the direct coupling with mechanics in the finite element modeling employed in modern nuclear fuel performance codes. In this study, a physically based zirconium alloy corrosion model called the Coupled-Current Charge Compensation (C4) model, which was initially developed for operating temperature conditions, has been updated to include high-temperature corrosion in order to provide additional critical information (e.g., oxygen concentration profile) under loss-of-coolant accident (LOCA) conditions—information lacking in existing empirical models. The C4 model was implemented in the MOOSE finite-element framework developed at Idaho National Laboratory, enabling it to be used in the BISON nuclear fuel performance code based on the MOOSE framework. To precisely track the different interfaces at a relatively low computational cost, the eXtended Finite Element Method (X-FEM) was applied in MOOSE. The model’s results were compared to those of existing empirical models as well as metallographic analysis of high-temperature oxidized Zircaloy-4 coupons. Oxygen diffusivities in the a and ß phases resulting from this comparison closely agree with those found in the literature. The C4 model implemented with X-FEM in MOOSE now has the capability to accurately predict oxide, oxygen-stabilized a, and prior ß phase layer growth kinetics under isothermal exposure at high temperature (1000–1500°C). Furthermore, in contrast with the empirical models, the C4 model accounts for the finite thickness of the fuel cladding. It can predict the oxygen concentration profile evolution through the whole cladding, enabling evaluation of the remaining ductile thickness—a crucial variable for modeling the mechanical behavior of the fuel cladding under LOCA. Furthermore, this implementation allows direct coupling with mechanics, at a low computing cost, using finite-element-based nuclear fuel performance codes such as BISON.

36 MATERIALS SCIENCE↗

Verification and Validation of the BISON Fuel Performance Code for PCMI Applications

BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. A brief overview of BISON’s computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described. Validation for application to light water reactor (LWR) PCMI problems is assessed by comparing predicted and measured rod diameter following base irradiation and power ramps. Results indicate a tendency to overpredict clad diameter reduction early in life, when clad creepdown dominates, and more significantly overpredict the diameter increase late in life, when fuel expansion controls the mechanical response. Initial rod diameter comparisons have led to consideration of additional separate effects experiments to better understand and predict clad and fuel mechanical behavior. Results from this study are being used to define priorities for ongoing code development and validation activities.

BISON↗

Creation of fully vectorized FORTRAN code for integrating the movement of dust grains in interplanetary environments

The main objective is to improve the performance of a specific FORTRAN computer code from the Planetary Sciences Division of NASA/Johnson Space Center when used on a modern vectorizing supercomputer. The code is used to calculate orbits of dust grains that separate from comets and asteroids. This code accounts for influences of the sun and 8 planets (neglecting Pluto), solar wind, and solar light pressure including Poynting-Robertson drag. Calculations allow one to study the motion of these particles as they are influenced by the Earth or one of the other planets. Some of these particles become trapped just beyond the Earth for long periods of time. These integer period resonances vary from 3 orbits of the Earth and 2 orbits of the particles to as high as 14 to 13.

Colquitt, Walter↗

Recent MCNP6 ® Code Developments and Improvements for Nuclear Engineering Applications [Slides]

The Los Alamos MCNP Monte Carlo radiation transport code has been the international gold standard for particle transport applications for over three decades. Many developments to the code have taken place with several significant new feature additions, major improvements, and enhancements to existing features. With significant institutional and programmatic investment in the code since the time of the last public release in 2018, important code development and infrastructure modernization has taken place and remains a high priority for all ongoing efforts across the code development team.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Wasatch: An architecture-proof multiphysics development environment using a Domain Specific Language and graph theory

To address the coding and software challenges of modern hybrid architectures, we propose an approach to multiphysics code development for high-performance computing. This approach is based on using a Domain Specific Language (DSL) in tandem with a directed acyclic graph (DAG) representation of the problem to be solved that allows runtime algorithm generation. When coupled with a large-scale parallel framework, the result is a portable development framework capable of executing on hybrid platforms and handling the challenges of multiphysics applications. In addition, we share our experience developing a code in such an environment – an effort that spans an interdisciplinary team of engineers and computer scientists.

97 MATHEMATICS AND COMPUTING↗

fermi-ad/flutter-template

This code is a collection of repositories for code that will help modernize the AD controls system. This is a collection of backend services for applications, libraries with common code, and templates to help initialize applications. The individual repositories are: 1. https://github.com/fermi-ad/flutter-template 2. https://github.com/fermi-ad/grpc-alarms-db 3. https://github.com/fermi-ad/alarms-actions-synchronizer 4. https://github.com/fermi-ad/aeolisten-async 5. https://github.com/fermi-ad/aeolisten 6. https://github.com/fermi-ad/acorn-alarms-service 7. https://github.com/fermi-ad/ap-python-template 8. https://github.com/fermi-ad/rust-template 9. https://github.com/fermi-ad/acsys-python-container-example 10. https://github.com/fermi-ad/dev-containers 11. https://github.com/fermi-ad/grpc-reflection-catalog 12. https://github.com/fermi-ad/rust-db-lib 13. https://github.com/fermi-ad/ap-python-launcher

Neswold, Rich [Fermi National Accelerator Laborato↗

On the Accuracy of CFD-Based Pressure Drop Predictions for Right-Angle Ducts

The predictive capability of computational fluid dynamics (CFD) codes for turbulent flow through curved ducts is of significant importance to the design and performance analysis of modern rocket engine flowpaths. Code calibration and validation studies for this class of flow are desireable to estimate the performance margin and operating range of components designed using Navier-Stokes methods. Parametric experimental studies such as that of Weske (NACA ARR W-39) provided a wealth of performance data for the design of single- and compound elbow configurations with various cross-sections, curvature and aspect ratios at varying Reynolds numbers. In that work, the majority of data is presented in the form of loss coefficients, characterizing pressure losses due to duct curvature, and including losses due to wall friction. Using measured friction coefficients, losses of equivalent straight lengths of duct are subtracted, resulting in performance curves useful for design computations. These data are currently used in a CFD-based parametric study covering a broad range of operating conditions. Of particular interest for the accuracy of CFD predictions are the effects on pressure loss due to inlet boundary layer thickness (dependent on upstream development length), and the wall treatment for the turbulence equations (conventional wall functions vs. wall integration using a two-layer model). The experimental data are reassessed in the form of an error analysis, and are compared with CFD predictions for 18 computational cases. Grid-independence, grid spacing, and convergence requirements of the cases are discussed. Conclusions regarding the relative importance of the parametric variables will be presented.

Brankovic, Andreja↗

Validating the BISON fuel performance code to integral LWR experiments

BISON is a modern finite element-based nuclear fuel performance code that has been under development at the Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. Code validation is underway and is the subject of this study. A brief overview of BISON’s computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described, followed by a summary of the experimental data used to date for validation of Light Water Reactor (LWR) fuel. Validation comparisons focus on fuel centerline temperature, fission gas release, and rod diameter both before and following fuel-clad mechanical contact. Comparisons for 35 LWR rods are consolidated to provide an overall view of how the code is predicting physical behavior, with a few select validation cases discussed in greater detail. Our results demonstrate that 1) fuel centerline temperature comparisons through all phases of fuel life are very reasonable with deviations between predictions and experimental data within ±10% for early life through high burnup fuel and only slightly out of these bounds for power ramp experiments, 2) accuracy in predicting fission gas release appears to be consistent with state-of-the-art modeling and with the involved uncertainties and 3) comparison of rod diameter results indicates a tendency to overpredict clad diameter reduction early in life, when clad creepdown dominates, and more significantly overpredict the diameter increase late in life, when fuel expansion controls the mechanical response. In the initial rod diameter comparisons they were unsatisfactory and have lead to consideration of additional separate effects experiments to better understand and predict clad and fuel mechanical behavior. Results from this study are being used to define priorities for ongoing code development and validation activities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

The elphbolt ab initio solver for the coupled electron-phonon Boltzmann transport equations

elphbolt is a modern Fortran (2018 standard) code for efficiently solving the coupled electron–phonon Boltzmann transport equations from first principles. Using results from density functional and density functional perturbation theory as inputs, it can calculate the effect of the non-equilibrium phonons on the electronic transport (phonon drag) and non-equilibrium electrons on the phononic transport (electron drag) in a fully self-consistent manner and obeying the constraints mandated by thermodynamics. It can calculate the lattice, charge, and thermoelectric transport coefficients for the temperature gradient and electric fields, and the effect of the mutual electron–phonon drag on these transport properties. The code fully exploits the symmetries of the crystal and the transport-active window to allow the sampling of extremely fine electron and phonon wave vector meshes required for accurately capturing the drag phenomena. The corray feature of modern Fortran, which offers native and convenient support for parallelization, is utilized. The code is compact, readable, well-documented, and extensible by design.

electron drag↗

CINDER90 and CINDER2008 Comparison for Second Target Station Analysis

A preliminary target activation analysis for the Second Target Station (STS) at the Spallation Neutron Source (SNS) located at Oak Ridge National Laboratory (ORNL) was preformed using Monte Carlo N-Particle X (MCNPX), CINDER90, and the corresponding multi-group cross section libraries distributed with CINDER90. This preliminary analysis has been used to guide some design decisions of the STS target design. Using the MCNPX output files from this preliminary analysis, a comparison of the activation analysis using CINDER90 and CINDER2008 was performed. While the CINDER90 code has been validated for use in spallation source activation and transmutation calculations, and compares well with measurements, the transition to the modernized version of the code (CINDER2008) and the updated cross section libraries is desirable. The goal of this comparison is to evaluate the differences and implications for the STS activation analysis. Overall, the total activity differs between CINDER2008 and CINDER90 by no more than 20% for the materials evaluated in this report. The largest discrepancies were observed in the decay gamma intensity densities with CINDER90 calculating a factor of 2 higher than CINDER2008.

97 MATHEMATICS AND COMPUTING↗

Light Output and Thermal Blast Analysis of Nuclear Fireballs

During the 1940s, 50s, and 60s, there were 210 atmospheric nuclear tests. Every test was extensively recorded with high quality motion picture films and photographs, providing a unique historic record, containing ~10,000 films, with a total of ~22 million frames of information. Much of what is known about the effects of nuclear weapons has been obtained from these tests. There are approximately 6500 scientific films that are in the process of being scanned by Lawrence Livermore National Laboratory and re-analyzed using modern computer techniques. The four major effects of a nuclear detonation are prompt radiation, shockwave, thermal blast, and fallout. The objective of this interim report is to describe recent efforts to extract benchmark data from the existing streak films using modern Figure-processing techniques. These new and, hopefully, greatly-improved data will provide our code developers with the benchmark data necessary to validate our modern computer codes that simulate nuclear weapon effects.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

The SkyView Data Model: Handling Heterogeneous Data Using Dynamic Software

The SkyView Virtual Telescope allows users to query over 75 distinct datasets of astronomical imagery including more than 190 distinct surveys. These holdings are extraordinarily diverse ranging over more than 15 orders of magnitude in frequency and5 orders of magnitude in resolution. To accommodate supporting these data, SkyView has developed a simple data model for surveys. Nor surprisingly, given the diverse origins of the datasets included, some surveys do not easily fit within this model. The conventional approach to dealing with such outliers is to build a layer on top of the actual data which more faithfully adheres to the data model. With SkyView we have used a somewhat alternative approach. We allow the data model to define new behaviors in the SkyView code itself either by customizing the use of existing capabilities or pointing to new classes that have non-standard behavior. This approach is feasible since in modern computer languages the code that is run can be selected dynamically with new classes completely overriding traditional behavior. This paper describes this approach and considers how it may become more useful as our software structures become increasingly dynamically linked.

Thomas McGlynn↗

The Variable Infiltration Capacity model version 5 (VIC-5): infrastructure improvements for new applications and reproducibility

The Variable Infiltration Capacity (VIC) model is a macroscale semi-distributed hydrologic model. VIC development began in the early 1990s and the model has since been used extensively for basin- to global-scale applications that include hydrologic dataset construction, trend analysis of hydrologic fluxes and states, data evaluation and assimilation, forecasting, coupled climate modeling, and climate change impact assessment. Ongoing operational applications of the VIC model include the University of Washington's drought monitoring and forecasting systems and NASA's Land Data Assimilation System. This paper documents the development of VIC version 5 (VIC-5), which includes a major reconfiguration of the legacy VIC source code to support a wider range of modern hydrologic modeling applications. The VIC source code has been moved to a public GitHub repository to encourage participation by the broader user and developer communities. The reconfiguration has separated the core physics of the model from the driver source code, whereby the latter is responsible for memory allocation, preprocessing and post-processing, and input–output (I–O). VIC-5 includes four drivers that use the same core physics modules, but which allow for different methods for accessing this core to enable different model applications. Finally, VIC-5 is distributed with robust test infrastructure, components of which routinely run during development using cloud-hosted continuous integration. The work described here provides an example to the model development community for extending the life of a legacy model that is being used extensively. The development and release of VIC-5 represents a significant step forward for the VIC user community in terms of support for existing and new model applications, reproducibility, and scientific robustness.

42 ENGINEERING↗

Perturbation of a normal shock - A test case for unsteady Euler codes

Shock tracking capabilities of modern nonoscillatory schemes for solving the Euler equations are studied by comparing their solutions to solutions obtained using the method of characteristics on two test problems. Both test problems involve accelerating and decelerating shock motions. The schemes tested include first order upwind, second order SONIC-A, third order ENO, and third order TONIC-A. Results show that the second and higher order schemes give nearly identical shock trajectories that are closer to the characteristic solution than the first order upwind scheme. All schemes approach the characteristic solution upon mesh refinement.

Suresh, Ambady↗

Calibration and Validation of NRC Fuel Performance Models

Pacific Northwest National Laboratory maintains steady-state and transient fuel performance codes for the U.S Nuclear Regulatory Commission (NRC). The NRC uses these codes in performing confirmatory calculations of licensee fuel performance codes and safety analysis methods during the review and approval of these codes and methods. The NRC also uses these codes to perform calculations to direct regulatory research and to assist in the development of new regulations and regulatory guides. Because of the safety-related uses for these codes, it is critically important that the codes are capable of accurately modeling modern fuel and operations and that the models and code predictions are well validated using data collected from conditions relevant to in-reactor conditions. FRAPCON and FRAPTRAN have an extensive database of separate effects data used to validate individual material property and fuel performance models and an integral effects database used to assess the overall code prediction within the full range of fuel operations. The key code predictions that are of regulatory interest as they relate to various specified acceptable design limits are; fuel and cladding temperature, rod internal pressure, cladding hoop strain, cladding corrosion and hydriding, and fission gas release. To assess the ability of the code to accurately predict these values, the integral assessment database consists of measurements of fuel centerline temperature, fission gas release, void volume, cladding corrosion, and cladding hoop strain. This paper will discuss the activities that are performed to ensure that FRAPCON and FRAPTRAN represent the state of the industry and will describe the separate effect and integral effect databases that are used to validate the models and are continually being expanded to ensure validation over the full range of fuel operations.

Geelhood, Kenneth J.↗

OLTARIS: On-Line Tool for the Assessment of Radiation in Space

The On-Line Tool for the Assessment of Radiation In Space (OLTARIS) is a World Wide Web based tool that assesses the effects of space radiation on humans and electronics in items such as spacecraft, habitats, rovers, and spacesuits. This document explains the basis behind the interface and framework used to input the data, perform the assessment, and output the results to the user as well as the physics, engineering, and computer science used to develop OLTARIS. The transport and physics is based on the HZETRN and NUCFRG research codes. The OLTARIS website is the successor to the SIREST website from the early 2000's. Modifications have been made to the code to enable easy maintenance, additions, and configuration management along with a more modern web interface. Overall, the code has been verified, tested, and modified to enable faster and more accurate assessments.

Singleterry, Robert C., Jr.↗

Validation of CTF Droplet Entrainment and Annular/Mist Closure Models using Riso Steam/Water Experiments

This document summarizes the work to validate the droplet entrainment and de-entrainment models, as well as two-phase closure models in the CTF (Coolant Boiling in Rod Arrays - Two Fluids) code by comparison with experimental data obtained at Risoe National Laboratory. CTF is the modernized version of the subchannel code, COBRA-TF, being jointly developed by Oak Ridge National Laboratory and North Carolina State University for applications in the Consortium for Advanced Simulation of Light Water Reactors (CASL). The Risoe data included a series of more than 250 steam/water experiments that were performed in both tube and annulus geometries over a range of various pressures and outlet qualities. Experimental conditions were set so that the majority of cases was in the annular/mist flow regime. Measurements included liquid film flow rate, droplet flow rate, film thickness, and two-phase pressure drop. CTF was used to model 180 of the tubular geometry cases, matching experimental geometry, outlet pressure, and outlet flow quality to experimental values. CTF results were compared to the experimental data at the outlet of the test section in terms of vapor flow fractions, and entrained liquid flow fractions, and pressure drop per unit length.

42 ENGINEERING↗