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DOE OSTI2020

LANL Critical Benchmark Comparison Study and Subsequent Revision

As part of an international collaboration within the DOE Nuclear Criticality Safety Program (NCSP), LANL is involved in a comparison study to quantify differences in k-effective results from neutron transport simulations of critical benchmark experiments. The DOE NCSP Mission and Vision details the activity in which the French Institut De Radioprotection et De Sûreté Nucléaire (IRSN) leads the study with LANL and in conjunction with ORNL and LLNL to compare results of various neutron transport codes and nuclear data libraries to compute k-effective for ICSBEP benchmarks held in common by the entities. This report documents results obtained through partial completion of the overall effort with a focus on the changes made to LANL benchmarks modeled with MCNP6 using ENDF/B-VII.1 nuclear data that appeared to have discrepant results when compared with results of other codes. The feedback received through participation in the comparison collaboration has prompted an effort to review particular input files for benchmarks and revise when necessary. This report documents the results of review and revision of specific benchmarks highlighted as possibly discrepant in the comparison study. In addition, this effort prompted a new collaboration between LANL XCP and NCS Divisions in the development of a shared review/revision procedure and use of a new benchmark repository.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Nuclear Data and Benchmarking Program: Nuclear Data Performance Assessment for Advanced Reactors

Over the last several Evaluated Nuclear Data File (ENDF)/B releases, many updates have been introduced to the nuclear data libraries that can significantly change computational results. These changes can be particularly important for advanced reactor concepts, which are not as thoroughly investigated as those used in light water reactor (LWR) systems. Therefore, a performance assessment of the ENDF/B libraries using the SCALE 6.2.3 code package was conducted under the auspices of the U.S., Department of Energy Office of Nuclear Energy (DOE-NE). The following systems relevant to the advanced reactor community were chosen as subject for investigation in this report: sodium cooled fast reactor systems, graphite moderated high temperature gas-cooled reactors, and several molten salt reactor models. At first, the similarity of the integral performance of the ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data libraries caused by compensating errors in important differential data was demonstrated. In calculations for a shipping container of high-assay low enriched uranium, the data of individual isotopes were systematically swapped between the two ENDF/B libraries. An eigenvalue difference of up to 450 pcm was found due to the use of 235 U and 238 U from one library and 1 H and 16 O from the other library. This clearly demonstrates a cross-correlation between reaction data sets of different isotopes within a library that should be reported in the evaluations. For sodium-cooled fast reactor (SFR) fuel assemblies, significant eigenvalue (k inf ) differences (200– 450 pcm) were found between calculations using the 2011 ENDF/B-VII.1 data and 2018 ENDF/B-VIII.0 data. These differences were mainly caused by updates of 238 U and 239 Pu neutron cross sections. Results from the multigroup (MG) calculations further revealed that the group structure of the applied MG library strongly influences the MG bias due to the importance of an appropriate energy resolution of the resonances in higher energy ranges. The application of previously used MG libraries requires new verification for advanced reactor simulation in comparisons with reference continuous-energy calculations. Nuclear data uncertainty analyses of these SFR systems resulted in eigenvalue uncertainties between 1,400 and 1,800 pcm, which are three to four times higher than corresponding uncertainties in light water reactor (LWR) systems. The main contributor to this uncertainty was found to be inelastic scattering on 238 U, which shows an uncertainty of up to 50% in the fast energy range. Other important contributors are the scattering reactions of 56 Fe and 23 Na which have so far not appeared in LWR analysis. Since the largest contribution to the eigenvalue uncertainty of SFR systems is coming from scattering reactions, it is expected that uncertainties in the angular scattering distributions also have a significant impact. Those uncertainties are, however, not available for the majority of nuclides, and the capabilities to determine sensitivities to this data are currently underdeveloped. For two graphite moderated high temperature gas-cooled reactor (HTGR) benchmarks, only the ENDF/BVII. 1 calculations resulted in consistent eigenvalues (k eff ) with the corresponding experimental measurements. Eigenvalue differences of about 1,000 pcm were found between the 2006 ENDF/B-VII.0 data and the 2011 ENDF/B-VII.1 data, primarily due to an updated carbon capture cross section in the thermal energy range in the later ENDF/B release. The ENDF/B-VIII.0 eigenvalue was larger than the ENDF/B-VII.1 eigenvalue by about 300 pcm mainly due to updates in multiple 235 U neutron cross sections, with offsetting updates in the 238 U cross sections. With ENDF/B-VIII.0, graphite can be modeled as perfect crystal or with two different porosities. The choice of the graphite evaluation can have a significant influence of the eigenvalue. For a HTGR benchmark, maximum eigenvalue differences as high as 650 pcm due to different porosities were observed. More detailed studies of the impact of the porosity are necessary while considering that the graphite porosity can vary between the used materials and changes as a function of neutron fluence. xii The uncertainty of the HTGR eigenvalues due to nuclear data uncertainties was found between 500 and 600 pcm, with the top contributor being the neutron multiplicity of 235 U. A gap in the form of missing uncertainties in graphite thermal scattering data, that might have significant impact on HTGR reactors, was identified. This gap is being investigated by the DOE-NE Nuclear Data and Benchmarking Program, as documented in a separate report The fast spectrum molten salt reactor calculations revealed similar comparisons as for the SFR assemblies. The eigenvalue uncertainties were significantly influenced by inelastic scattering on 238 U, while the impact of angular scattering distributions is unknown. For the graphite and zirconium hydride moderated system, eigenvalue uncertainties of up to 700 pcm were observed, requiring in depth analyses of the cross sections and corresponding uncertainties of the included materials. In general, it was observed that the flux spectra of the various molten salt reactor systems show significant differences, even between a fresh state and a depleted state. The choice of the energy group structure of MG calculations is therefore highly relevant; the applicability of previously used MG libraries needs to be verified. Due the unavailability of pin power measurements of advanced reactor systems, pin power calculations with both ENDF/B-VII.1 and VIII.0 data were performed for a light water reactor and compared to corresponding measurements. Both calculations show good agreement with the measurements. A comparison between the ENDF/B-VII.1 and VIII.0 results revealed small differences, mostly in the range of about 0.2%.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

2009.1 Revision of the Evaluated Nuclear Data Library (ENDL2009.1)

LLNL’s Computational Nuclear Data and Theory Group have created a 2009.1 revised release of the Evaluated Nuclear Data Library (ENDL2009.1). This library is designed to support LLNL’s current and future nuclear data needs and will be employed in nuclear reactor, nuclear security and stockpile stewardship simulations with ASC codes. The ENDL2009 database was the most complete nuclear database for Monte Carlo and deterministic transport of neutrons and charged particles. It was assembled with strong support from the ASC PEM and Attribution programs, leveraged with support from Campaign 4 and the DOE/Office of Science’s US Nuclear Data Program. This document lists the revisions and fixes made in a new release called ENDL2009.1, by comparing with the existing data in the original release which is now called ENDL2009.0. These changes are made in conjunction with the revisions for ENDL2011.1, so that both the .1 releases are as free as possible of known defects.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

LLNL Response to the DOE ASCR RFI, "Stewardship of Software for Scientific and High-Performance Computing"

For decades, Lawrence Livermore National Laboratory (LLNL) has been engaged in significant research, development, and support for software to enable scientific computing and, particularly, the use of high performance computing (HPC) in the NNSA mission space. In particular, the move in the mid-1990’s to simulation as a leading component of stockpile stewardship through the ASCI and the successor ASC programs, as well as the need for reliable data acquisition and control software for the National Ignition Facility, have been important drivers in building expertise in production-quality software development at LLNL. LLNL has also been a leader in the DOE SciDAC FASTMath Institute and the DOE Exascale Computing Project (ECP), both of which have striven to make scientific computing software – in particular, the enabling technologies underpinning simulation capabilities – more widely adopted and sustainable. As such, we believe that our experience can inform the broader goal of software stewardship for scientific and high-performance computing. LLNL strongly supports the formation of a new DOE ASCR program element in software stewardship and sustainment. Historically, DOE ASCR has funded applied mathematics and computer science research that has led to the development of important new capabilities and algorithms that are expressed as artifacts in research software. Such frameworks, libraries, and tools have seldom been directly funded to address the important issues of code maintenance, documentation, robustness, and community building. Software engineering and support have typically been done on the side in support of the ASCR-driven research products. DOE funding priorities have been slow to recognize that good software engineering, the kind that ensures research investments have more adoption and longevity, requires significant resources. Based upon our experiences, we have prepared this response to highlight the concerns and issues we believe to be important as DOE ASCR considers its role in scientific software stewardship. We believe that role is important and will require a significant investment of new funding to legitimately support the technologies past and future DOE ASCR investments have and will produce to facilitate their uptake and adoption in the broader scientific computing community. Following a summary of our involvement in scientific software development, the remainder our response is organized around the nine topics specifically identified in the RFI.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2015

The linear and nonlinear response of infinite periodic systems to static and/or dynamic electric fields. Implementation in CRYSTAL code

An implementation of the vector potential approach (VPA) for treating the response of infinite periodic systems to static and dynamic electric fields has been initiated within the CRYSTAL code. The VPA method is based on the solution of a time-dependent Hartree-Fock or Kohn-Sham equation for the crystal orbitals wherein the usual scalar potential, that describes interaction with the field, is replaced by the vector potential. This equation may be solved either by perturbation theory or by finite field methods. With some modification all the computational procedures of molecular ab initio quantum chemistry can be adapted for periodic systems. Accessible properties include the linear and nonlinear responses of both the nuclei and the electrons. The programming of static field pure electronic (hyper)polarizabilities has been successfully tested. Dynamic electronic (hyper)polarizabilities, as well as infrared and Raman intensities, are in progress while the addition of finite fields for calculation of vibrational (hyper)polarizabilities, through nuclear relaxation procedures, will begin shortly.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2016

MO-FG-BRA-07: Intrafractional Motion Effect Can Be Minimized in Tomotherapy Stereotactic Body Radiotherapy (SBRT)

Purpose: Tomotherapy has unique challenges in handling intrafractional motion compared to conventional LINAC. In this study, we analyzed the impact of intrafractional motion on cumulative dosimetry using actual patient motion data and investigated real time jaw/MLC compensation approaches to minimize the motion-induced dose discrepancy in Tomotherapy SBRT treatment. Methods: Intrafractional motion data recorded in two CyberKnife lung treatment cases through fiducial tracking and two LINAC prostate cases through Calypso tracking were used in this study. For each treatment site, one representative case has an average motion (6mm) and one has a large motion (10mm for lung and 15mm for prostate). The cases were re-planned on Tomotherapy for SBRT. Each case was planned with 3 different jaw settings: 1cm static, 2.5cm dynamic, and 5cm dynamic. 4D dose accumulation software was developed to compute dose with the recorded motions and theoretically compensate motions by modifying original jaw and MLC to track the trajectory of the tumor. Results: PTV coverage in Tomotherapy SBRT for patients with intrafractional motion depends on motion type, amplitude and plan settings. For the prostate patient with large motion, PTV coverage changed from 97.2% (motion-free) to 47.1% (target motion-included), 96.6% to 58.5% and 96.3% to 97.8% for the 1cm static jaw, 2.5cm dynamic jaw and 5cm dynamic jaw setting, respectively. For the lung patient with large motion, PTV coverage discrepancies showed a similar trend of change. When the jaw and MLC compensation program was engaged, the motion compromised PTV coverage was recovered back to >95% for all cases and plans. All organs at risk (OAR) were spared with < 5% increase from original motion-free plans. Conclusion: Tomotherapy SBRT is less motion-impacted when 5cm dynamic jaw is used. Once the motion pattern is known, the jaw and MLC compensation program can largely minimize the compromised target coverage and OAR sparing.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2013

SHARP Assembly-Scale Multiphysics Demonstration Simulations

The NEAMS Reactor Product Line effort aims to develop an integrated multiphysics simulation capability for the design and analysis of future generations of nuclear power plants. The Reactor Product Line code suite’s multi-resolution hierarchy is being designed to ultimately span the full range of length and time scales present in relevant reactor design and safety analyses, as well as scale from desktop to petaflop computing platforms. In this report, we describe our efforts to integrate thermal/hydraulics, neutronics, and structural mechanics modeling codes to perform coupled analysis of a representative fast sodium-cooled reactor fuel assembly. Over the past five years, the Reactor Product Line effort has developed high-fidelity single-physics codes for neutron transport modeling, in the PROTEUS code, and Large Eddy Simulation-based thermal/fluid modeling in the Nek5000 code. Both these codes have been exercised on over 100,000 processors of the IBM Blue Gene/P. The Diablo code has been used to perform structural mechanics and thermo-mechanical modeling. MOAB, the Reactor Geometry Generator (RGG), and MeshKit have been developed to generate and manipulate mesh and mesh-based data, in both serial and parallel environments. These tools together form a strong basis on which to build a multi-physics modeling capability. The goal of developing such a tool is to perform multi-physics neutronics, thermal/fluid, and structural mechanics modeling of the components inside a reactor core. While the focus of this report is on modeling a fast sodium-cooled reactor, it is also the goal that this simulation tool be useful for most reactor types. Here we report on the integration of PROTEUS and Nek5000 into the NEAMS framework. Both codes have been modified to read and write data from and to the MOAB library, in both serial and parallel environments. We focused on broad support for as many original physics code options as possible, rather than focusing only on near-term needs. Both codes were demonstrated through this interface on a relatively simple, seven-pin hexagonal fuel assembly. To support coupled analysis, the new multi-physics driver library COUPE was developed; this library provides flexible support for both loose and tight coupling, and can interface flexibly with codes in multiple programming languages. To couple results between the separate meshes used in each physics, the driver, PROTEUS, and Nek5000 all interface with the MBCoupler solution transfer tool provided with MOAB. Together, these tools form the basis of a new, coupled neutronics and thermal/fluid modeling code, that works in both serial and parallel. In a related effort, a driver in the Diablo code was used to access Nek5000-computed results written to a MOAB database, as well as the temperature data used to drive structural mechanics computations.

Tautges, Timothy J.↗
DOE OSTI2013

Final Report A Multi-Language Environment For Programmable Code Optimization and Empirical Tuning

This report summarizes our effort and results of building an integrated optimization environment to effectively combine the programmable control and the empirical tuning of source-to-source compiler optimizations within the framework of multiple existing languages, specifically C, C++, and Fortran. The environment contains two main components: the ROSE analysis engine, which is based on the ROSE C/C++/Fortran2003 source-to-source compiler developed by Co-PI Dr.Quinlan et. al at DOE/LLNL, and the POET transformation engine, which is based on an interpreted program transformation language developed by Dr. Yi at University of Texas at San Antonio (UTSA). The ROSE analysis engine performs advanced compiler analysis, identifies profitable code transformations, and then produces output in POET, a language designed to provide programmable control of compiler optimizations to application developers and to support the parameterization of architecture-sensitive optimizations so that their configurations can be empirically tuned later. This POET output can then be ported to different machines together with the user application, where a POET-based search engine empirically reconfigures the parameterized optimizations until satisfactory performance is found. Computational specialists can write POET scripts to directly control the optimization of their code. Application developers can interact with ROSE to obtain optimization feedback as well as provide domain-specific knowledge and high-level optimization strategies. The optimization environment is expected to support different levels of automation and programmer intervention, from fully-automated tuning to semi-automated development and to manual programmable control.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2019

Improving the Field Portability of Neutron and Gamma Counting Techniques

Savannah River National Laboratory (SRNL) puts science to work by developing techniques that better meet customers needs for in-situ measurements. There is high demand for laboratory quality measurements in field situations, improving the cost and speed for first responders and environmental monitoring. This work focused on two improvements (1) reading a signal from a neutron detector using a Raspberry Pi (2) testing the Peak-to-Valley method for determining the depth of a source in soil contamination measurements. A Raspberry Pi is a cheap, compact computer. This affordable, compact computer is able to do simple tasks that a desktop computer is able to do, like spreadsheets, browse the internet, play games and even program a simple robot. Additionally, a Raspberry Pi can run off of a battery, so it could be deployed for long term measurements in adverse conditions like neutron monitors deployed in a shipping port. Using the Raspberry Pi and libraries easily accessible on the internet, a code was written in Python that allows the Pi to keep track of how many times a button is pushed. This work will be used with the future goal of creating low power-budget neutron detectors. 1) Make sure Python was correctly installed on the Raspberry Pi. 2) Write a code that will keep track of how many times a button connected to the Pi is pressed. A code was written to use the Raspberry Pi to count the number of times the button was pushed. Moving forward the Raspberry Pi will be tested with a neutron counter. Ideally, each signal from the neutron pulse will be equivalent of 'push the button' and be counted. In the characterization of a soil contamination area both the location and the depth of the source is important. Attenuation of the source by the soil can make determining its depth challenging, particularly for monoenergetic radionuclides like Cs-137. To address this issue, a method of calculating the 'peak-to-valley' ratio of the gamma count spectra is implemented to help calculate the known depth of the source. 1) Determine the 'peak-to-valley' ratio of a Cs-137 source at different depths. 2) Find an appropriate function to serve as a model for the peak-to-valley ratio as a function of depth. Several Petri dishes, filled with sand, were stacked with one containing a Cs-137 source. The individual Petri dishes allowed for the source to be placed at different depths. A bucket was placed over the sand and source, to keep the detector at a consistent height. A Petri dish containing a Eu-155 source, which acted as an internal calibration standard, was placed on top of the bucket. A NaI detector was place on top of the Eu-155 source and the assembly was surrounded by a tungsten shield. Count times were 20 minutes and the data was analyzed using PeakEasy. Data was recorded for Cs-137 at a series of depths in sand. Both exponential and logarithmic functions provided a reasonable fit to the data. The experimental data showed a decreasing 'peak-to-valley' ratio as the depth was increased. The next steps for this project involves determining the optimum fit function and combining the results from sand to different soil types to get a better representation of the soil found at SRS. Eventually, the peak-to-valley ratio will be used to estimate the depth of contamination in field measurements.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2017

VERA Core Physics Benchmark Progression Problem Results for VERA 3.6

This report contains a comparison of results obtained with VERA 3.6 Release Candidate 0 for the Core Physics Benchmark Progression Problems 1 through 9 to both computational tools and measured data from Watts Bar Unit 1 (WBN1). (Ref. 1 and Ref. 2) These problems progress from single pin cell analyses to full core depletion calculations. VERA results obtained for Problems 1 through 5 are compared against results obtained using the SCALE 6.2 Beta code KENO-VI. Results for Problem 6 and 7 are compared against those obtained using a coupled neutronic and thermal-hydraulic combination of the KAPL software MC21 and the CASL tool CTF. (Ref. 3, Ref. 4 and Ref. 5) VERA results from problems 5 and 9 are compared to data from WBN1. For all of the analyses contained in this report, VERA provided similar results to other state-of-the-art software programs and real reactor data. This indicates that the program is capable of modeling a full-scale PWR in a more discretized and computationally efficient manner than any other reactor physics program. Since the descriptions of the Core Physics Benchmark Progression Problems and the logic behind their selection is discussed in-depth elsewhere it is not covered in the scope of this document. Additionally, in-depth discussions of the software programs used in this analysis are not provided for the same rationale.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Implementing Ordinary Differential Equation Solvers in Rust Programming Language for Modeling Vehicle Powertrain Systems: Preprint

Efficient and accurate ordinary differential equation (ODE) solvers are necessary for powertrain and vehicle dynamics modeling. However, current commercial ODE solvers can be financially prohibitive, leading to a need for accessible, effective, open-source ODE solvers designed for powertrain modeling. Rust is a compiled programming language that has the potential to be used for fast and easy-to-use powertrain models, given its exceptional computational performance, robust package ecosystem, and short time required for modelers to become proficient. However, of the three commonly used (>3,000 downloads) packages in Rust with ODE solver capabilities, only one has more than four numerical methods implemented, and none are designed specifically for modeling physical systems. Therefore, the goal of the Differential Equation System Solver (DESS) was to implement accurate ODE solvers in Rust designed for the component-based problems often seen in powertrain modeling. DESS is a text-based software package that provides a flexible framework for building and solving systems of ODEs. This allows DESS to be included as a dependency for automotive powertrain models that require a variety of solvers and solver configurations. Seven explicit ODE solver methods have been implemented in DESS: Euler’s, Heun’s, midpoint, Ralston’s, classic Runge-Kutta, Bogacki-Shampine, and Cash-Karp. These represent five fixed-step methods and two adaptive-step methods. This paper shows that the solver implementations increase accuracy and computational efficiency compared to Euler's method when modeling a system of three thermal masses in Rust. DESS also includes features designed for modeling component-based physical systems. Users can define relationships between nodes in their system, which the package then translates into a system of equations, leading to simpler and more intuitive code. In the case of a three-thermal-mass system, the user can specify node thermal properties (e.g., thermal capacitance), how nodes are interconnected, and thermal conductance between nodes rather than providing a system of equations. The core contribution from this work is an open-source, text-based Rust package with ODE solvers for automotive powertrain modeling to support cost-free, fast, and accurate simulation.

ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2021

EKAT v.1.0

E3SM Kokkos Application Toolkit (EKAT) is a collection of C++, Fortran, and CMake utilities for providing a single implementation of common kernels based on the Kokkos programming model. The library contains utilities for vectorization, tridiagonal linear system solvers, and linear interpolation as well as some general-purpose utilities such as testing utilities, parameter lists, representation of physical units, and additional interfaces. The goal is to provide a centralized implementation for high-performance computing structures and common utilities that reduce code duplication and streamline maintenance efforts. Sandia National Laboratories is a multimission laboratory managed and operated by National Technology & Engineering Solutions of Sandia, LLC, a wholly owned subsidiary of Honeywell International Inc., for the U.S. Department of Energy's National Nuclear Security Administration under contract DE-NA0003525. SAND2022-1327 O

Bertagna, Luca↗
DOE OSTI2025

LHC Event Generation in the Exascale Era

MCFM is a dedicated Monte-Carlo simulation program for collider phenomenology at highest energies. Designed during the Tevatron era, it has successfully incorporated the latest developments needed for LHC precision calculations and remained on the forefront of collider phenomenology. The Fortran code includes interfaces to modern PDF and loop reduction libraries but has been unchanged structurally compared to the earlier versions. Parallel computing has been enabled using OpenMP and MPI. MCFM provides numerically highly stable one-loop amplitudes and superior phase-space efficiency, leading to excellent performance in NXLO calculations using jettiness or qT subtraction techniques for IR regularization.

Campbell, John [Fermilab]↗
DOE OSTI2022

POSTER: Automatic Differentiation of Parallel Loops with Formal Methods

The accompanying poster to this short paper presents a combination of reverse mode AD and formal methods to enable efficient differentiation of (or backpropagation through) shared-memory parallel code. Compared to the state of the art, our approach can more often avoid the need for atomic updates or private data copies during the parallel derivative computation, even in the presence of unstructured or data-dependent data access patterns. This is achieved by gathering information about the memory access patterns from the input program, which is assumed to be correctly parallelized. This information is then used to build a model of assertions in a theorem prover, which can be used to check the safety of shared memory accesses during the parallel derivative computation

Automatic Differentiation↗
DOE OSTI2024

GPU-enabled extreme-scale turbulence simulations: Fourier pseudo-spectral algorithms at the exascale using OpenMP offloading

Fourier pseudo-spectral methods for nonlinear partial differential equations are of wide interest in many areas of advanced computational science, including direct numerical simulation of three-dimensional (3-D) turbulence governed by the Navier-Stokes equations in fluid dynamics. This paper presents a new capability for simulating turbulence at a new record resolution up to 35 trillion grid points, on the world's first exascale computer, Frontier, comprising AMD MI250x GPUs with HPE's Slingshot interconnect and operated by the US Department of Energy's Oak Ridge Leadership Computing Facility (OLCF). Key programming strategies designed to take maximum advantage of the machine architecture involve performing almost all computations on the GPU which has the same memory capacity as the CPU, performing all-to-all communication among sets of parallel processes directly on the GPU, and targeting GPUs efficiently using OpenMP offloading for intensive number-crunching including 1-D Fast Fourier Transforms (FFT) performed using AMD ROCm library calls. With 99% of computing power on Frontier being on the GPU, leaving the CPU idle leads to a net performance gain via avoiding the overhead of data movement between host and device except when needed for some I/O purposes. Memory footprint including the size of communication buffers for MPI_ALLTOALL is managed carefully to maximize the largest problem size possible for a given node count. Detailed performance data including separate contributions from different categories of operations to the elapsed wall time per step are reported for five grid resolutions, from 2048 3 on a single node to 32768 3 on 4096 or 8192 nodes out of 9408 on the system. Both 1D and 2D domain decompositions which divide a 3D periodic domain into slabs and pencils respectively are implemented. The present code suite (labeled by the acronym GESTS, GPUs for Extreme Scale Turbulence Simulations) achieves a figure of merit (in grid points per second) exceeding goals set in the Center for Accelerated Application Readiness (CAAR) program for Frontier. The performance attained is highly favorable in both weak scaling and strong scaling, with notable departures only for 2048 3 where communication is entirely intra-node, and for 32768 3 , where a challenge due to small message sizes does arise. Communication performance is addressed further using a lightweight test code that performs all-to-all communication in a manner matching the full turbulence simulation code. Performance at large problem sizes is affected by both small message size due to high node counts as well as dragonfly network topology features on the machine, but is consistent with official expectations of sustained performance on Frontier. Overall, although not perfect, the scalability achieved at the extreme problem size of 32768 3 (and up to 8192 nodes — which corresponds to hardware rated at just under 1 exaflop/sec of theoretical peak computational performance) is arguably better than the scalability observed using prior state-of-the-art algorithms on Frontier's predecessor machine (Summit) at OLCF. New science results for the study of intermittency in turbulence enabled by this code and its extensions are to be reported separately in the near future.

3D fast Fourier transform↗
DOE OSTI2018

Generation of Reversible C++ Code for Optimistic Parallel Discrete Event Simulation

The reversible execution of C/C++ code has been a target of research and engineering for more than a decade as reversible computation has become a central notion in large-scale parallel discrete event simulation (PDES). The simulation models that are implemented for PDES are of increasing complexity and size and require various language features to support abstraction, encapsulation, and composition when building a simulation model. In this paper, we focus on parallel simulation models that are written with user-defined C++ abstractions and abstractions of the C++ Standard Library. We present an approach based on incremental state saving for establishing reversibility of C++ and an evaluation for a kinetic Monte-Carlo simulation implemented in C++. Although a significant runtime overhead is introduced with our technique, it is an enormous win that it allows using the entire C++ language, and has that code automatically transformed into reversible code to enable parallel execution with the Rensselaer’s optimistic simulation system (ROSS).

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2015

Warthog: A MOOSE-Based Application for the Direct Code Coupling of BISON and PROTEUS

The Nuclear Energy Advanced Modeling and Simulation (NEAMS) program from the Department of Energy's Office of Nuclear Energy provides a robust toolkit for the modeling and simulation of current and future advanced nuclear reactor designs. This toolkit provides these technologies organized across product lines: two divisions targeted at fuels and end-to-end reactor modeling, and a third for integration, coupling, and high-level workflow management. The Fuels Product Line and the Reactor Product line provide advanced computational technologies that serve each respective field well, however, their current lack of integration presents a major impediment to future improvements of simulation solution fidelity. There is a desire for the capability to mix and match tools across Product Lines in an effort to utilize the best from both to improve NEAMS modeling and simulation technologies. This report details a new effort to provide this Product Line interoperability through the development of a new application called Warthog. This application couples the BISON Fuel Performance application from the Fuels Product Line and the PROTEUS Core Neutronics application from the Reactors Product Line in an effort to utilize the best from all parts of the NEAMS toolkit and improve overall solution fidelity of nuclear fuel simulations. To achieve this, Warthog leverages as much prior work from the NEAMS program as possible, and in doing so, enables interoperability between the disparate MOOSE and SHARP frameworks, and the libMesh and MOAB mesh data formats. This report describes this work in full. We begin with a detailed look at the individual NEAMS framework technologies used and developed in the various Product Lines, and the current status of their interoperability. We then introduce the Warthog application: its overall architecture and the ways it leverages the best existing tools from across the NEAMS toolkit to enable BISON-PROTEUS integration. Furthermore, we show how Warthog leverages a tool known as DataTransferKit to seamlessly enable the transfer for solution data between disparate frameworks and mesh formats. To end, we demonstrate tests for the direct software coupling of BISON and PROTEUS using Warthog, and discuss current impediments and solutions to the construction of physically realistic input models for this coupled BISON-PROTEUS system.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗