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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 1,495 records · Page 83
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

ML-PSA

The computer code uses a parallel simulated annealing framework with embedded machine learning components to solve multi-constrained optimization problems. The software automatically balances the execution of low and high fidelity physics models within the optimization procedure. The low fidelity model is used to rapidly explore the design space while the high fidelity physics model is executed sparingly to account for complex design constraints that are not resolved by the quickly executing low fidelity model.

Gurecky, William↗
DOE OSTI2022

Web application for machining parameter recommendation using visual analytics

The computer codes create a web application including multiple user interfaces, visualizations, interaction modules between the user, and visualizations. The visualizations will be interactive, allowing users to query the data directly, and they will be arranged in a dashboard. When selections are made in one visualization, filters will be applied to all other visualizations providing a coordinated system for asking questions about the data. The visualization can be utilized for manufacturing data management, analysis, and recommendation.

Chae, Junghoon↗
DOE OSTI2022

PUFFIn

The PUFFIn software package is a graphical user interface that allows for photon and electron beam calculations using the PENELOPE computer code. It prepares the required input files and displays the resulting output data. It can be used by anyone interested in the interaction of photon or electron beams with matter

Schwarz, Randy↗
DOE OSTI2023

CTH v.13.0

SAND2023-06647O The latest version of the CTH suite of computer codes is a flexible software system designed to treat a wide range of shock wave propagation and material motion phenomena in 1-D, 2-D, or 3-D. CTH has models suitable for most conditions encountered in shock physics including material strength, fracture, distended materials, high explosives, and a variety of boundary conditions. Finite-volume analogs of the Lagrangian equations of momentum and energy conservation are employed with continuous rezoning to construct Eulerian differencing. Adaptive mesh refinement is available for maximizing resolution in regions of interest. Full Lagrangian numerical methods are available for mechanical or structural response.

Schumacher, Shane↗
DOE OSTI2025

PyPTkinetics

This basic research code computes the time-dependent phase transition between two solid crystalline phases under dynamic conditions.

Blaschke, Daniel N.↗
DOE OSTI2025

HALLUFIELD: DETECTING LLM HALLUCINATIONS VIA FIELD-THEORETIC MODELING

A research-focused Python package that implements our hallucination-detection method for large language models(LLMs). The code computes stability signals from LLM predictions across various hyperparameters sweeps and combines free-energy/entropy–style metrics to flag likely hallucinations, with tunable thresholds for batch scoring. The repo includes evaluation scripts, config files, and example notebooks to reproduce benchmark results and ablations; it depends on standard open-source libraries (PyTorch, Hugging Face) and runs on CPU/GPU. The repository examples only use public models/datasets only.

Bhattarai, Manish [Los Alamos National Labs]↗
DOE OSTI2022

Initial study on cross-section generation requirements for a PBR equilibrium core

A Serpent model of the equilibrium core HTR-PM small modular nuclear reactor in China, was developed for use in cross-section preparation studies in order to guide methods development for the Griffin reactor multiphysics application. The model includes detailed isotopics for 10 distinct pebble burnup groups in 126 core zones with unique fuel and moderator temperatures obtained from a coupled neutronics-thermal-fluids equilibrium core calculation using Griffin-Pronghorn. A sensitivity study of the fuel and moderator temperatures for various core regions was performed with the MOOSE stochastic tools. The results show that the uncertainties are, not unexpectedly, dominated by the value of the fluid temperature and that the power level, heat transfer coefficient and effective conduction to neighboring pebbles and fluid constitute, at best, second order effects. The temperature uncertainty range varies from 28 K to 57 K between the core entry and exit planes, respectively, but these values are probably higher. We still have to quantify the significance of these uncertainties in the preparation of cross-sections in future work. In addition, we verified that the effective pebble approximation used in the PEBBED and V.S.O.P. computer codes works well for the preparation of region averaged cross-sections. Nevertheless, there are some discrepancies in the cross-sections when compared to the multi-pebble model, which could affect the prediction of peak values and the depletion calculation. We conclude that is highly desirable for future studies with Griffin to be able to handle both the 'effective' pebble approximation and the multi-pebble approach for various pebble burnup groups. This enables Griffin users with the flexibility to perform higher-fidelity studies. Finally, we initiated the preparation of cross-sections for various core regions from the full core Serpent reference model. We quantified the differences in 26 group cross-sections from infinite domain models versus the full core approach. These reference cross-sections will serve to verify the double heterogeneity, self-shielding, and spectrum-correction methods in Griffin. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Investigation of Application Specific k eff Sensitivities to Existing and New Subcritical and Critical Experiments

Integral subcritical and critical experiments are essential in providing validation of computational codes which improve confidence in the predicted k eff results of specific applications. This work discusses the relevance of existing integral subcritical and critical experiments to applications far reaching across the nuclear complex. These applications include Travelling Wave Reactors (TWR), Kilopower Reactor Using Stirling Technology (KRUSTY), and plutonium reflected by tantalum. The primary goal of this work is to improve knowledge of energy-dependent cross-section sensitivities for specific applications, identify nuclear data gaps in existing experiments, enable an optimized design of new experiments to address these gaps, and design practical new experiments that could be executed at the National Criticality Experiments Research Center (NCERC) that have a high-value to these applications.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Optimizing hybrid rotational femtosecond/picosecond coherent anti-Stokes Raman spectroscopy in nitrogen at high pressures and temperatures

We demonstrate the use of hybrid rotational femtosecond/picosecond (fs/ps) coherent anti-Stokes Raman spectroscopy (HR-CARS) as a technique for temperature measurements in nitrogen gas at high pressures and temperatures. A broadband pulse shaper-adjusted 42 fs pulse interacts with a narrow-bandwidth, frequency-upconverted 5.5 ps pulse in a cell containing N 2 at pressures of 1–70 atm and temperatures of 300–1000 K. A computational code is used to model spectra and fit experimental results to obtain best-fit temperatures. We demonstrate good qualitative fits as well as good accuracy and precision between thermocouple measured and best-fit temperatures over the explored pressure and temperature regimes. The overall average percentage temperature difference between thermocouple measurements and best-fit temperatures is − <#comment/> 0 . 3 % <#comment/> with a standard deviation of 7.1%, showing the suitability of HR-CARS for characterizing high-pressure and -temperature environments.

Torge Mecker, Nils (ORCID:0000000238977170)↗
DOE OSTI1991

Nuclear Safety [Vol. 32, No. 4, October-December 1991]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 477 Report on the International Symposium on the Use of Probabilistic Safety Assessment for Operational Safety—PSA '91, S. Chakraborty and M. Khatib-Rahbar; 488 Good Relationships Are Pivotal in Nuclear Data Bases, A. S. Heger and B. V. Koen; 494 Technical Note: The Interagency Nuclear Safety Review Panel's Evaluation of the Ulysses Space Mission, J. A. Sholtis, Jr., D. A. Huff, L. B. Gray, N. P. Klug, and R. O. Winchester; ACCIDENT ANALYSIS: 502 The Severe Accident Analysis Program for the Savannah River Nuclear Production Reactors, M. L. Hyder; CONTROL AND INSTRUMENTATION: 511 A Framework for Selecting Suitable Control Technologies for Nuclear Power Plant Systems, R. A. Kisner; DESIGN FEATURES: 521 Containments for Gas-Cooled Power Reactors History and Status, P.W. Williams; ENVIRONMENTAL EFFECTS: 537 Indoor Radon: A Natural Risk, N. H. Harley and J.H. Harley; On the Importance of the Atmospheric Parameters in the Fission Products Distribution of a Severe Reactor Accident, M. C. Barla and A. R. Bayulken; Book Review: Health Effects of Exposure to Low Levels of Ionizing Radiation BEIR V, C. R. Richmond; WASTE AND SPENT FUEL MANAGEMENT: 555 Activities Related to Waste and Spent Fuel Management, M. D. Muhlheim and E. G. Silver OPERATING EXPERIENCES: 567 Effects of Component Aging on the Westhinghouse Control Rod Drive System, K. Sullivan and W. Gunther; 577 Reactor Shutdown Experience, Compiled by J. W. Cletcher, 580 Selected Safety-Related Events, Compiled by G. A. Murphy; 582 Operating U 8 Power Reactors, Compiled by M. D. Muhlheim and E. G. Silver, RECENT DEVELOPMENTS: 596 General Administrative Activities, Compiled by M. D. Muhlheim and E. G. Silver; 610 Reports, Standards, and Safety Guides, D. S. Queener; 614 Proposed Rule Changes as of June 30, 1991; ANNOUNCEMENTS: 520 Workshop on PC PRAISE: A Probabilistic Fracture Mechanics Personal Computer Code for Nuclear Power Plant Piping Reliability Assessment; 536 Fifth Workshop on Nuclear Power Plant Containment Integrity; 624 New OECD "International Information System on Occupational Exposure (ISOE)"; 625 Harvard School of Public Health Announces Short Courses; 625 Eighth Power Plant Dynamics, Control and Testing Symposium; 626 Second Training Course on Off-Site Emergency Planning and Response for Nuclear Accidents; 618 The Authors; 622 Reviewers of Nuclear Safety, Vol 32.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Criticality Safety Validation of SCALE 6.2.2

The computational bias of criticality safety computer codes must be established through the validation of the codes to critical experiments. A large collection of suitable experiments has been vetted by the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and made available in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook). More than 600 cases from this handbook ha ve been prepared and reviewed within the Verified, Archived Library of Inputs and Data (VALID) , which is maintained by the Reactor and Nuclear Systems Division at Oak Ridge National Laboratory. The performance of the KENO V.a and KENO-VI Monte Carlo codes within the SCALE 6.2.2 code system is assessed using the VALID models of benchmark experiments. A range of nuclear cross section libraries based on both ENDF/B-VII.1 and ENDF/B-VII.0 in both multigroup (MG) and continuous energy (CE) formats is considered. The TSUNAMI tools for sensitivity and uncertainty analysis are used to examine some systems further to try to identify potential causes of unexpected results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Analysis of Radiological Impacts to Terrestrial Biota in Support of Environmental Assessment for Use of DOE-Owned High-Assay Low-Enriched Uranium Stored at Idaho National Laboratory

There are many species of plants and animals (biota) on the INL Site and facilities such as INTEC and MFC have surrounding habitat supporting resident plant and animal populations. Consequently, radionuclide releases from the HALEAU processing facilities could pose a potential radiological impact to nearby flora and fauna populations. Therefore, the RESRAD-BIOTA (Version 1.8) computer code (http://resrad.evs.anl.gov/codes/resrad-biota/) was used to assess the impact of radiation exposures to terrestrial biota resulting from soil contaminated by airborne releases from the HALEAU processing facilities proposed for the INTEC and/or MFC complexes. The contaminated soil subsequently results in contamination in air and in different food sources used by biota. Both external radiation and internal radiation are considered in the dose evaluation.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2019

Failures and Implications of Heat Pipe Systems

Under Department of Energy (DOE), Office of Nuclear Energy (NE), Gateway for Accelerated Innovation in Nuclear (GAIN), Sandia National Laboratories (SNL) was awarded DOE-NE GAIN voucher GA-19SN020107, "Risk-informed mechanistic source term calculations for a sodium fast reactor." Under this GAIN voucher, SNL supported the industry partners development in preparation for licensing and commercialization by providing subject matter expertise on heat pipe technologies, providing computer code training and support, and perform first-of-a-kind experiments demonstrating the safety/risk impacts of heat pipe breach failures. The experiments that were performed had two primary goals: measure the peak heat fluxes that lead to heat pipe dry out and subsequent wall breach; and observe the consequences that result from catastrophic failure of a heat pipe wall. Intentional breaching of the heat pipe walls took advantage of heat pipe physics and operating limits. Large and nearly instantaneous heat fluxes were applied to the heat pipe to first cause localized dry out at the evaporator section which then leads to melting of the heat pipe wall. The hour glass heat pipe (Test 1) experienced dry out at 112 W/cm 2 and after 45 seconds, wall temperatures measure about 1,280°C and intentional failure of the heat pipe wall was achieved. The cylindrical heat pipe (Test 2) experienced dry out at 125 W/cm 2 and after 65 seconds, wall temperatures exceeded 1,400°C and intentional failure of the heat pipe wall was achieved. Both experiments characterize the parameters needed to lead to heat pipe wall failure. Furthermore, the failure of the heat pipes characterizes the safety/risk impacts from sodium-oxygen reactions that occur following the intentional failure. There were two major conclusions of these intentional failure tests: the heat pipes were able to continue operating beyond expected performance limits, and the failure behavior validated decades of operational experience.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

NEUP Final Report: Analysis of Thermal Transients for a Sodium-CO 2 Heat Exchanger

The design of heat exchangers for use in a supercritical carbon dioxide (sCO 2 ) Brayton cycle power converter must provide for acceptable performance for duty cycle events encompassing anticipated transients and postulated accidents. Argonne National Laboratory participated in the NEUP project, “Thermal Hydraulic and Structural Testing and Modeling of Compact Diffusion-Bonded Heat Exchangers for Supercritical CO 2 Brayton Cycle” with the task of conducting a comprehensive analysis of thermal transients for sCO 2 cycle heat exchangers, with emphasis on the sodium-to-CO 2 heat addition heat exchanger. A range of transients, from normal operation to severe accidents, were simulated with the coupled PDC and SAS4A/SASSYS-1 system level dynamic analysis computer codes. For each transient, the calculated change in the heat exchanger wall temperature is determined as a measure of the thermal loading. In addition, an analysis to find an equivalent helium-CO 2 heat exchanger was carried out to facilitate testing with helium rather than sodium on the hot side.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Dose Modeling for Harborview Research and Training Facility

The Harborview Research and Training (R&T) Facility at the University of Washington (UW) was accidently contaminated from a breached sealed source of Cesium-137. The release occurred during removal of the sealed source from its holder within a portable hot cell in the loading dock area near room 220. The loading dock and northeast corner of the building above the ground floor were impacted by the release, whereas the remaining portions of the R&T Facility were minimally affected. Detailed descriptions of the R&T Facility and the affected area/rooms are included in Section 2. The RESRAD Team in the Environmental Science Division of Argonne National Laboratory visited the site on August 12, 2019, and interviewed the R&T Facility building manager and staff working in the building at that time. Site- and building-specific data were collected including building floor plan CAD drawings, building occupants’ working hours, contaminated rooms photos and their dimensions, building air exchange rates, building materials and “curtain wall” details, and so forth. These data are being used in the dose modeling for potential occupants of the building after it is remediated. The dose modeling was performed using the RESRAD-BUILD computer code, which is designed for the evaluation of radiologically contaminated buildings (Yu et al. 2003). Two hypothetical building occupants of the Harborview R&T Facility were evaluated: a researcher occupying room 726 (a research laboratory on the seventh floor) and a custodian occupying room 220 (loading dock and trash compactor room with a rollup door). A maintenance worker that will be spending less time in impacted areas was not modeled because it will be bounded by other two hypothetical building occupants modeled. A summary description of the RESRAD-BUILD is provided in Appendix A, and the modeling of potential doses to a researcher and a custodian is described in detail in Section 3. A summary and a discussion of the dose modeling results are included in Section 4. The parameters used in the RESRAD-BUILD modeling are included in Appendix B. This report documents the dose modeling results; it does not describe measurements of the actual radioactivity levels at the Harborview R&T Facility, nor does it set any expectation for the actual final radioactivity levels. The radiation doses presented are modeling results based on the generic screening values of Cs-137 and several conservative assumptions and are meant only to assist in decision making regarding decontamination of the Harborview R&T Facility.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2019

Utah State University's contribution to Enhancing Understanding of High Energy Density Plasmas from Wire Array and Solid Liner Implosions Using Fluid Modeling with Kinetic Closures (Final Report)

This final report pertains to work performed at Utah State University (USU) on the subject of kinetic closures for fluid simulations of high-energy-density (HED) plasmas. This report was written following the end date of the three year grant that started August 1, 2016, and ended July 31, 2019. The key aspect of the work at USU was to develop and implement kinetic closures for fluid simulations of more strongly coupled plasmas. Funds for this project provided some summer salary for the PI at USU, Dr. Eric Held, but mainly provided a robust stipend for his graduate student, David Hansen, who has worked on the project since the start date. Mr. Hansen is nearing completion of his PhD thesis which he will defend either in December of 2019 or the spring of 2020. With this award Dr. Eric Held and his PhD student David Hansen developed a computer code called {\bf X-section} that can be used to generate collisional scattering cross sections which summarize particle interactions in high-energy-density plasmas.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Verification of the DIF3D Software to Support Fast Reactor Analysis

Ongoing design activities at Argonne National Laboratory are requiring a thorough verification of the Argonne Reactor Computation codes be performed. DIF3D is central to this system. The driver for this effort requires the 3D Cartesian, triangular-Z, and hexagonal-Z core geometry options of DIF3D be verified. Previous work identified the DIF3D features required to be verified to support current design activities, features of which are generally applicable to hexagonal-Z fast reactor designs. The scope of this verification effort includes verifying DIF3D’s ability to correctly translate the user’s model in to DIF3D’s preferred format, verifying that options planned for use have the desired effect, and verifying the correctness of the eigenvalue, fixed-source, forward, and adjoint solvers in DIF3D-FD and DIF3D-VARIANT. This manuscript provides the verification tasks and their results with respect to the features needed for current design activities. Since analytic solutions of the neutron diffusion and transport equations are either limited in scope or not possible, multiple tiers of problems unique to each solver and geometry type were implemented. Each of these tiers tests features independent and complementary arguments for why the separate testing of functionalities is acceptable. Finally, this separate testing was also supplemented with a high-level integral check of each the diffusion and transport capabilities and applicable geometries. To accommodate cases which an analytic solution is not feasible, MCNP6.2 was relied upon to provide a higher-order reference solution. This therefore required that the capabilities within MCNP6.2 which were relied upon for this work are also verified in this work. No MCNP discrepancies were noted in this effort. Note that the MCNP6.2 verification included in this work does not stand as a full verification of MCNP6.2, but merely verifies the features used in verifying DIF3D. The verification effort identified no issues that are debilitating or otherwise impactful to design usage of DIF3D, and thus DIF3D version 11.0, release 3012 is considered verified. The types of issues that were identified were predominantly in the areas of: unclear documentation, software bugs which were inconsequential to final results, editing options which were ignored in favor of printing more information than requested, bugs in the outputs of intermediate results, or secondary output binary file information which was not present. While not a bug, this verification report also identified that the algorithm used to evaluate the peak fast flux in a nodal transport solution can be quite unreliable due to the polynomial order used and the location of the peak within the mesh. The authors of the report therefore recommend the usage of the EvaluateFlux software (distributed with ARC) as a more robust alternative.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Air Dispersion Modeling of Radioactive Releases During Proposed 236Z Rubble Pile Removal Activities

This report is part of the planning process for the removal of the 236Z rubble remaining as of January 1, 2018, at the Plutonium Finishing Plant (PFP) on the Hanford Site. This report documents an analysis considering standard techniques for the removal of demolition rubble. It does not document any decisions about the removal approaches. This document is intended to guide the establishment of a boundary for the application of an air monitoring regime. Atmospheric dispersion modeling using estimated release rates has been conducted to provide information on the location and levels of radioactive contamination that may be expected as the result of rubble removal activities. Hourly meteorological data collected over a 6-year period (2004–2009) were used to examine the effects of wind speed, direction, and stability on projected concentrations of contaminants in air and deposited on nearby surfaces. The source terms modeled in this report were based on the residual transuranic contamination levels that are anticipated to remain in the 236Z rubble following building demolition. The potential airborne concentrations were established using the five-factor formula from DOE-HDBK-3010 (DOE 1994) considering material-at-risk, damage ratio, airborne release fraction, respirable fraction, and leak path factor. Radioactive contamination emissions were calculated by release mechanism and rubble pile area for on-shift and off-shift activities. The emissions from the applicable sources were combined to provide emissions estimates for each day from each cleanup area. The U.S. Environmental Protection Agency’s (EPA’s) AERMOD computer code was used to estimate atmospheric dispersion and deposition of the released radioactive materials in the immediate vicinity of the planned rubble removal activities Both airborne and surface concentrations were modeled with AERMOD. Hourly derived air concentrations (DAC) were modeled for an array of receptors covering the rubble pile site and surrounding area. Peak (95th and 99th percentile) values of time-integrated air concentrations at these receptor points are derived from these hourly values, with modeling results reported as total incremental air concentrations in DAC-hours occurring over the selected time period. Total accumulated deposition amounts were also evaluated with AERMOD using the same array of receptors, with results reported as disintegrations per minute (dpm) per 100 cm2. Each rubble pile was considered in terms of its size, location, and estimated contamination levels. The modeling was based on the assumed sequence of the rubble removal activities. The results in this report are based on cleanup schedules and methods proposed by CHPRC. As of January 1, 2018, there are broadly three large rubble piles (west, footprint, and east) remaining of the 236Z Building process cell. It is assumed that the overall cleanup will require either 1) 30 days over a 6-week period, or 2) 60 days over a 12-week period. Work was assumed to be performed during 10-hour day shifts. The analysis shows that some releases of radioactive material are to be anticipated during the cleanup of the 236Z rubble piles. The modeling results presented here are closely tied to the details of how the rubble removal is to be conducted. The results indicate that for the rubble removal, the radiological exposures from the planned activities are predicted to be well below the designated limits for air and soil exposures. The cleanup of the 236Z main process cell rubble is expected to release some alpha-emitting radionuclides. However, results based on 6 years of hourly climatological data, two representative cleanup periods, realistic mitigating actions, and a bounding source term indicate that concentrations at the fence line of the PFP facilities should remain well below 12 DAC-hours/week. The results presented in this report are a reevaluation of a subset of the results presented in PNNL-20173 Rev. 4.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗