Calibration and Recovery of Nuclear Test Seismic Ground-Motion Data from the Leo Brady Seismic Network.
Abstract not provided.
Young, Brian↗
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MCNP6 Status and plans for criticality safety applications - 14368
The first production version of MCNP6 was released in 2013, with an update in 2014. MCNP6 provides all of the standard methodology for criticality calculations available in previous versions and new capabilities for adjoint-based sensitivity profiles and perturbations. This paper reviews the current status of MCNP6, including new features, performance improvements, verification testing, and nuclear data libraries. For criticality safety support, a new statistical package called Whisper was developed and first applied in 2014 for validation and determining USLs. Whisper includes input files and sensitivity profiles for over 1000 ICSBEP benchmarks and will be distributed with MCNP6. For the long term, an initiative called MCNP 2020 is underway to improve performance, strengthen the code foundation, and prepare for future computer systems. (authors)
42 ENGINEERING↗
Verification and validation testing and tools: comparison between MCNP code versions and nuclear data libraries [Slides]
This presentation discusses the primary goal of software testing which is to test the code for correctness. It also discusses the results for individual suites and the role of validation and verification also referred to in the presentation as V&V. In summation, the V&V framework enables easy comparison between calculations performed with different code versions and/or nuclear data libraries. This entire framework will be distributed with the upcoming MCNP6.3 release. V&V test suites shown and several that were not (Criticality, LAQGSM, Lockwood) will be distributed in the new framework.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
General Guidelines on Criteria for Adoption or Rejection of Evaluated Libraries and Data by the Nuclear Data Team
This memo contains general guidelines on what documentation and tools need to be in place as well as format and data testing requirements such that evaluated nuclear data sets or entire libraries can be adopted by the nuclear data team. Additional requirements beyond this memo might apply for specific nuclear data observables. These guidelines were established based on discussions between J.L. Conlin, M.G. Gray, A.P. McCartney, D. Neudecker, D.K. Parsons and M.C. White.
96 KNOWLEDGE MANAGEMENT AND PRESERVATION↗
User input verification and test driven development in the NJOY21 nuclear data processing code
Before physically-meaningful data can be used in nuclear simulation codes, the data must be interpreted and manipulated by a nuclear data processing code so as to extract the relevant quantities (e.g. cross sections and angular distributions). Perhaps the most popular and widely-trusted of these processing codes is NJOY, which has been developed and improved over the course of 10 major releases since its creation at Los Alamos National Laboratory in the mid-1970’s. The current phase of NJOY development is the creation of NJOY21, which will be a vast improvement from its predecessor, NJOY2016. Designed to be fast, intuitive, accessible, and capable of handling both established and modern formats of nuclear data, NJOY21 will address many issues that many NJOY users face, while remaining functional for those who prefer the existing format. Although early in its development, NJOY21 is quickly providing input validation to check user input. By providing rapid and helpful responses to users while writing input files, NJOY21 will prove to be more intuitive and easy to use than any of its predecessors. Furthermore, during its development, NJOY21 is subject to regular testing, such that its test coverage must strictly increase with the addition of any production code. This thorough testing will allow developers and NJOY users to establish confidence in NJOY21 as it gains functionality. This document serves as a discussion regarding the current state input checking and testing practices of NJOY21.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Discriminating Underground Nuclear Explosions Leading To Late-Time Radionuclide Gas Seeps
Utilizing historical data from the U.S. nuclear test program and freely available barometric pressure data, we performed an analytical barometric-pumping efficiency analysis to determine factors resulting in late-time radionuclide gas seeps from underground nuclear explosions. In this work, we considered 16 underground nuclear explosions with similar geology and test setup, of which five resulted in the measurement of late-time radionuclide gas concentrations at the ground surface. Additionally, the factors we considered include barometric frequency and amplitude, depth of burial, air-filled porosity, intact-rock permeability, fracture aperture, and fracture spacing. The analysis indicates that the best discriminators of late-time radionuclide gas seeps for these explosions are barometric frequency and amplitude and air-filled porosity. While geologic information on fracture aperture and spacing is not available for these explosions, the sensitivity of barometric-pumping efficiency to fracture aperture indicates that it would likely also be a good discriminator.
58 GEOSCIENCES↗
Validation and Testing of ENDF/B-VII Decay Data
The nuclear decay library developed for the ORIGEN code was upgraded from ENDF/B-VI.8 to -VII.0 and released in the SCALE nuclear modeling and simulation code system in 2011. Experience with the ENDF/B-VII.0 library identified serious errors and performance issues in the evaluated decay sublibrary. Initially, errors were identified in the decay scheme for the 238U series, leading to large discrepancies in the gamma emission rate for uranium. Further testing of the fission product decay data revealed significant performance issues with ENDF/B-VII.0 relative to the previous release of ENDF/B-VI. To address the performance issues, the decay data library in SCALE was upgraded to ENDF/B-VII.1 and released as an update for users in 2013. This paper describes the practical impact of the ENDF/B-VII.0 decay data performance issues on common code applications, and the validation and testing performed on the updated ORIGEN libraries developed from the ENDF/B-VII.1 sublibrary.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Nuclear data uncertainty propagation applied to the versatile test reactor conceptual design
We report the Versatile Test Reactor (VTR) currently under development is a 300 MWth sodium-cooled fast reactor (SFR) fueled with ternary metal alloy fuel, which aims to accelerate the testing of advanced nuclear fuels, materials, instrumentation, and sensors in high flux environments that are necessary to license the next generation of advanced reactor concepts. To support the VTR design process, uncertainties associated with the nuclear data has been propagated through the reactor core neutronics calculation to global parameters of interest, such as the core multiplication factor, kinetic parameters, and various reactivity feedback coefficients, following the sensitivity based uncertainty propagation approach. By folding the sensitivity coefficients, separately computed by the generalized perturbation theory code PERSENT and Monte Carlo code Serpent 2, with the variance-covariance matrices from COMMARA-2.0, we obtain the reaction-wise, isotope-wise, and overall uncertainties for each response of interest due to nuclear data uncertainty. With Serpent 2, the statistical error of the uncertainty is obtained by propagating the statistical error of the sensitivity coefficients through the same process using a newly developed uncertainty propagation method. From both codes, the overall top uncertainty contributors are found to be the cross section of Fe-56 elastic scattering, Na-23 elastic scattering, and U 238 inelastic scattering. The large contributions of the Fe-56 elastic scattering cross sections to global parameters are due to its relatively large relative uncertainty of 5–10% in nuclear data and the large volume of Fe-containing reflector assemblies in the fairly compact VTR core design. Both codes agreed well for the overall uncertainty estimates of all responses of interest, except the delayed neutron fraction, prompt neutron generation time, and the coolant density feedback coefficient, where Serpent 2 yielded a much larger value than PERSENT due to the large statistical error of sensitivity coefficients. The calculated uncertainties are also compared to those associated with other SFR cores. Another outcome of this study is a variance-covariance matrix of reactivity coefficients, which can be used in the subsequent uncertainty propagation to the system level to investigate the impact of identified uncertainties on system responses in the safety analysis.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Reaction History Uncertainty Propagation from Flux to α and from Time to α [Slides]
We discuss uncertainty propagation in reaction history data from underground nuclear tests. γ reaction history detectors measured flux as a function of time. After data analysis we report α as a function of time. Flux uncertainty and time uncertainty propagation into α uncertainty are discussed in detail. Discretization of the formulas is discussed.
45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗
Simulation of IST Turbomachinery Power-Neutral Tests with the ANL Plant Dynamics Code
The validation of the Plant Dynamics Code (PDC) developed at Argonne National Laboratory (ANL) for the steady-state and transient analysis of supercritical carbon dioxide (sCO2) systems has been continued with new test data from the Naval Nuclear Laboratory (operated by Bechtel Marine Propulsion Corporation) Integrated System Test (IST). Although data from three runs were provided to ANL, only two of the data sets were analyzed and described in this report. The common feature of these tests is the power-neutral operation of the turbine-compressor shaft, where no external power through the alternator was provided during the tests. Instead, the shaft speed was allowed to change dictated by the power balance between the turbine, the compressor, and the power losses in the shaft. The new test data turned out to be important for code validation for several reasons. First, the power-neutral operation of the shaft allows validation of the shaft dynamics equations in asynchronous mode, when the shaft is disconnected from the grid. Second, the shaft speed control with the compressor recirculation (CR) valve not only allows for testing the code control logic itself, but it also serves as a good test for validation of both the compressor surge control and the turbine bypass control actions, since the effect of the CR action on the loop conditions is similar for both of these controls. Third, the varying compressor-inlet temperature change test allows validation of the transient response of the precooler, a shell-and-tube heat exchanger. The first transient simulation of the compressor-inlet temperature variation Test 64661 showed a much slower calculated response of the precooler in the calculations than the test data. Further investigation revealed an error in calculating the heat exchanger tube mass for the PDC dynamic equations that resulted in a slower change in the tube wall temperature than measured. The transient calculations for both tests were done in two steps. The first step was done in the same fashion as the FY15 IST analysis, where the CR valve position and the turbine-compressor shaft speed were specified through the PDC input based on the test values. On the second step, the turbine-compressor shaft dynamics equations were invoked by specifying that the shaft is disconnected from the grid. In addition, the CR valve control was used to control the shaft speed, based on the turbine bypass control logic already implemented in the PDC. For the shaft power balance, the friction (windage) loss is calculated based on the shaft balance at the steady-state conditions and is assumed to be scaled to the third power of shaft speed in the transient. Both the steady-state and transient simulations of both tests showed good agreement with the test data. The only significant difference was the turbine performance, which was not predicted as well as it was in the previous IST simulation, resulting in the prediction of a somewhat different flow split between the two turbines. This flow split difference is believed to be the result of not addressing the recent turbine modifications in the model. In addition, the full simulation of the turbine-compressor speed variation Test 65261-P with shaft speed control showed greater a difference with the test data later in the transient than the other test. Further analysis of the results revealed that this difference is most likely due to scaling the shaft windage losses only with the shaft speed and ignoring its dependency on the fluid density in the shaft cavity. Based on the results of steady state and transient calculations of the Tests 64661 and 65216-P, several areas of future improvements for the PDC simulation of the IST are identified.
42 ENGINEERING↗
Predictive Data Analytics Framework Using Advanced Test Reactor Acoustic Data
Although a nuclear reactor is a hostile environment for sensing and electrical communications, the reactor core is amenable to acoustic communication. An acoustic measurement infrastructure (AMI) has been installed at the Advanced Test Reactor (ATR) nozzle trench area to record acoustic signals that has the ability to capture different operating regime of the reactor. This AMI includes ATR in-pile structural components, coolant, acoustic receivers, primary coolant pumps (PCP) as signal sources, a data acquisition system, and signal-processing algorithms, enabling real-time. This report will discusses development of recursive Fast Fourier Transform approach to process in real-time acoustic signals, application of short time Fast Fourier Transform to the ATR brush data to understand the vibration level and to develop spectrograms for different primary coolant pump combinations. The combination of primary coolant pumps for normal and power axial locator mechanism of ATR are different and generates different signatures. These acoustic signatures were used to develop machine learning approaches to automatically classify different operating regimes. This lay the foundation for predictive analytic framework that can be leverage by ATR to optimize their operation and maintenance. The path forward involves continued engagement with ATR and expanded implementation of AMI and predictive framework at ATR and other facilities within INL and at other experimental reactors.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
Book Review: Monitoring compliance with the comprehensive Nuclear-Test-Ban Treaty (CTBT). Contributions by the German national data centre edited by Christoph Pilger, Lars Ceranna and Christian Bönnemann
For decades, the nuclear weapons capability of different countries was tracked in part via the interpretation of signals from more than 2000 nuclear test explosions that were conducted in programs of weapons development. The first such explosion took place in July 1945 (in New Mexico, USA, prior to the use of nuclear weapons in World War 2 against Japan). The last nuclear test, prior to the Comprehensive Nuclear-Test-Ban Treaty (CTBT) becoming open for signature in September 1996, was in July 1996 (in China). This treaty, which took nearly forty years to negotiate, has been signed by 184 states as of December 2018, but a few key additional signatures and ratifications are still needed before nuclear tests are banned under international law. The evidence that a nuclear test explosion has occurred can come from electromagnetic, seismic, hydroacoustic, and infrasonic waves; and from unusual radionuclides— which can be present in particulate matter and/or as noble gases. Of course there is a vast infrastructure associated with military programs of nuclear weapons development in a limited number of countries. Less well-known, is the existence of a large infrastructure involving national and international efforts to monitor for the possible occurrence of nuclear test explosions, and thus to support CTBT verification. This book does a very good job of explaining much of that work, as it is conducted in practice by a western nation with sophisticated facilities.
98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
ORNL Testing of Multiple Graphite Benchmarks [Slides]
Integral criticality safety and reactor physics benchmark experiments from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) and the International Handbook of Reactor Physics Experiments (IRPhE Handbook) are essential for nuclear data validation and testing
ENDF↗
Computational Workflows for Uncertainty-Quantified Nuclear Reactions: From Nuclear Theory Inputs to Astrophysical Reaction Rates
Reactions on unstable nuclei, particularly those on the neutron-rich side of stability, are important for both fundamental and applied physics. For fundamental science, the most prevalent use case is astrophysi cal nucleosynthesis by rapid neutron capture—the r-process—by which heavy nuclei are formed in extreme astrophysical environments, such as in supernovae and neutron star mergers; see, e.g., Refs. [1–3]. For ap plications, these processes are relevant for the interpretation of radiochemical data from historic nuclear tests, which contribute to our ability to certify the enduring stockpile in the absence of nuclear testing [4]; see Ref. [5] for a broader discussion of applications. However, reaction cross sections involving unsta ble species are generally poorly understood, for the simple reason that useful data become scarce as one moves away from stability. While there are avenues for improving the amount and quality of data for these species [6], one is fundamentally reliant on nuclear theory to make progress on these fields of study.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
YOLO11 to SAM2 pipeline for feature extraction from nuclear test films
The response to the effects of nuclear detonations is supported by models that describe the evolution of the nuclear fireball and cloud and the associated transport of active debris. Validation of those descriptions relies on data from the nuclear test operations. Video records of those events offer a rich source of information that was exploited to a limited extent in historic analyses. Computer vision and machine learning techniques are powerful tools that can be used to increase the number of measurements that can be obtained from those films. In this work, we apply computer vision techniques to automatically track the temporal evolution of the nuclear fireball. In particular, we apply You Only Look Once 11 (YOLO11) and Segment Anything Model 2 (SAM2) in combination with minimal human intervention to digitized versions of the original nuclear test films. As part of the proposed workflow, the YOLO11 model is applied to films to determine bounding boxes for the fireball within each frame. These are then used as inputs to SAM2, which uses image segmentation to determine the fireball boundaries and their temporal evolution. We assess the accuracy of our approach by using it to determine the energy released during the Trinity nuclear test and comparing the results with previous analyses based on manual measurements.
Van Exel, Kimberly [ORNL] (ORCID:0009000877463894)↗
Updates to the HTR-PROTEUS HALEU Benchmark Using Modern Analysis Methodologies
The HTR-PROTEUS IRPhEP Handbook benchmarks represent some of the highest quality benchmarks available for systems with TRISO-HALEU fuel, graphite pebbles, graphite reflector, and high neutron leakage. Nevertheless, significant variability in computed eigenvalue results was encountered in HTR-PROTEUS configurations that are relevant for transportation configurations. The team will investigate the source of these variabilities in the original benchmark. In addition, the team will investigate the HTR-PROTEUS subcritical measurements, kinetics data, and potential criticality effects from the introduction of hydrogen in the system (water ingress). The project will generate a benchmark evaluation of additional HTR-PROTEUS measurements that can significantly increase the value of these criticality benchmarks in testing nuclear codes and data to support transportation validation needs for industry and the U.S. NRC.
22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
Design and Development of IER:516 - Zirconium Test Assembly (ZTA) - Critical Experiments for Zr Nuclear Data Verification and Validation
Explore the source record for details and available documents.
Siemers, Gregory Jay↗
Design and Development IER:516 - Zirconium Test Assembly (ZTA) - Critical Experiments for Zr Nuclear Data Verification and Validation
Abstract not provided.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗