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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 595 records · Page 33
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Fission barriers at the end of the chart of the nuclides

We present calculated fission-barrier heights for 5239 nuclides for all nuclei between the proton and neutron drip lines with 171 ≤ A ≤ 330. The barriers are calculated in the macroscopic-microscopic finite-range liquid-drop (FRLDM) with a 2002 set of macroscopic-model parameters. The saddle-point energies are determined from potential-energy surfaces based on more than five million different shapes, defined by five deformation parameters in the three-quadratic-surface shape parametrization: elongation, neck diameter, left-fragment spheroidal deformation, right-fragment spheroidal deformation, and nascent-fragment mass asymmetry. The energy of the ground state is determined by calculating the lowest-energy configuration in both the Nilsson perturbed-spheroid (ϵ) and the spherical-harmonic (β) parametrizations, including axially asymmetric deformations. The lower of the two results (correcting for zero-point motion) is defined as the ground-state energy. The effect of axial asymmetry on the inner barrier peak is calculated in the (ϵ,γ) parametrization. We have earlier benchmarked our calculated barrier heights to experimentally extracted barrier parameters and found average agreement to about one MeV for known data across the nuclear chart. Here we do additional benchmarks and investigate the qualitative and, when possible, quantitative agreement and/or consistency with data on β-delayed fission, isotope generation along prompt-neutron-capture chains in nuclear-weapons tests, and superheavy-element stability. In addition these studies all indicate that the model is realistic at considerable distances in Z and N from the region of nuclei where its parameters were determined.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2018

Radionuclide NESHAP Annual Report for CY 2017 Sandia National Laboratories, Nevada Tonopah Test Range

This report provides a summary of the radionuclide releases from the United States (U.S.) Department of Energy (DOE) National Nuclear Security Administration facilities at Sandia National Laboratories/Tonopah Test Range (SNL/TTR) during Calendar Year (CY) 2017, including the data, calculations, and supporting documentation for demonstrating compliance with 40 Code of Federal Regulation (CFR) 61, Subpart H--NATIONAL EMISSION STANDARDS FOR EMISSIONS OF RADIONUCLIDE OTHER THAN RADON FROM DEPARTMENT OF ENERGY FACILITIES (Radionuclide NESHAP).

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2021

Listen: How to make a Fat Man Step-by-step book recently added to Lab collections

The National Security Research Center has tens of millions of materials. This is the Lab’s classified library, which also houses artifacts from LANL history. Think of still-secret nuclear weapons data, old photos of the town in the 1940s, an old pair of goggles from the testing days, and so much more.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2025

Multiphysics Modeling of Microreactors with NEAMS codes, and Validation Based on KRUSTY Reactivity Insertion

The NEAMS Multiphysics Applications team continues to assess code usability and functionality for microreactor design and safety analyses, while demonstrating that NEAMS tools capture both steady-state and transient behavior across distinct microreactor concepts. In FY2025, the team advanced full-core, high-fidelity, multiphysics models that solve more complex problems and strengthen verification/validation for several microreactor systems: heat-pipe microreactor (HPMR), gas-cooled microreactor (GCMR), and the KRUSTY experiment. These models employ the MOOSE MultiApp/Transfers architecture with Griffin for neutronics, BISON for heat conduction/thermomechanics, Sockeye for heat pipes, SAM/THM for coolant channels and loops, and SWIFT for hydride behavior, with meshes generated via the MOOSE Reactor Module. The graphite models available in the Grizzly code were also investigated for future analyses. For the HPMR, a Na-HPMR variant was constructed to align with recently validated heat-pipe experiments and Sockeye’s LCVF capability, enabling mechanistic heat-pipe transients and startup modeling. The Na-HPMR will serve as the primary model for HPMR investigations in upcoming tasks. The load-following and single heat-pipe failure scenarios (Griffin/BISON/Sockeye), which were previously modeled for the K-HPMR, were replicated for the Na-HPMR, showing strong negative temperature feedback and highly localized thermal effects, respectively, while the startup case captured vapor-front progression and heat-removal activation. Solid mechanics was added to the previously built K-HPMR full-core model in BISON, showing minimal impact on steady-state reactivity yet enabling stress-field predictions that prepare the path for full-core TRISO performance analyses. For the GCMR, automated steady-state and four transient scenarios were executed using Griffin/BISON/SAM/SWIFT. Results confirm robust inherent safety: power collapses promptly in loss-of-cooling events, the inlet-temperature drop settles to a new equilibrium, and a single-channel blockage yields only a ~30 K local fuel-temperature rise with <0.4% power decrease. SWIFT-predicted hydrogen redistribution affects reactivity during both steady-state and transient conditions, underscoring its importance. A Brayton-cycle balance of plant (BOP) model in SAM/THM demonstrated stable startup behavior, and xenon-driven reactivity during load following was analyzed. To improve TRISO-compact temperature fidelity, a fast multiscale Heat Source Decomposition (HSD) treatment was implemented. Against heterogeneous benchmarks, HSD reduces underprediction of kernel temperatures and lowers predicted peak powers in reactivity-insertion transients compared to previous homogenized models. KRUSTY warm-critical validation progressed from FY2024 baselines: the 15Ȼ insertion shows excellent agreement in peak power (~2% high) and temperature trends, and the 30Ȼ case was automated via a feedback controller that maintained power near 3 kW for ~150 s with close agreement to data. The successful modeling of the warm critical tests has laid a strong foundation for simulating more complex nuclear system tests in the years ahead. Throughout FY2025, developer feedback was provided (e.g., MOOSE batch mesh generation, distributed pre-split meshes, Griffin sweeper on displaced meshes), several new models were contributed to the Virtual Test Bed, and an OECD-NEA WPRS multiphysics benchmark based on the HPMR was initiated to enable broader cross-comparison and best-practice development with the nuclear community at large.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Review of the Current State of Knowledge on the Effects of Radiation on Concrete

A review of the current state of knowledge on the effects of radiation on concrete in nuclear applications is presented. Emphasis is placed on the effects of radiation damage as reflected by changes in engineering properties of concrete in the evaluation of the long-term operation (LTO) and for Plant Life or Aging Management of nuclear power plants (NPPs) in Japan, Spain, and the United States. National issues and concerns are described for Japan and the US followed by a discussion of the fundamental understanding of the effects radiation on concrete. Specifically, the effects of temperature, moisture content, and irradiation on ordinary Portland cement paste and the role of temperature and neutron energy spectra on radiation induced volumetric expansion (RIVE) of aggregate-forming minerals are described. This is followed by a discussion of the bounding conditions for extended operation, the significance of accelerated irradiation conditions, the role of temperature, creep, and how these issues are being incorporated into numerical and meso-scale models. From these insights on radiation damage, analyses of these effects on concrete structures are reviewed and the current status of work in Japan and the US are described. Also discussed is the recent formation of a new international scientific and technical organization, the International Committee on Irradiated Concrete (ICIC), to provide a forum for timely information exchanges among organizations pursuing the identification, quantification, and modeling of the effects of radiation on concrete in commercial nuclear applications. Lastly, the paper concludes with a discussion of research gaps including: 1) interpreting test-reactor data, 2) evaluating service-irradiated concrete for aging management and to inform radiation damage models with the Zorita NPP (Spain) serving as the first comprehensive test case, 3) irradiated-assisted alkali-silica reactions, and 4) RIVE under constrained conditions.

36 MATERIALS SCIENCE↗
NASA NTRS1974

Performance of the electrical controls for the Mini-Brayton system

The design theory and performance of a breadboard of the proposed Mini-Brayton electrical control system is presented. The Mini-Brayton is a nuclear isotope powered dynamic power conversion system. Testing was performed with an electronic simulation of a turbine alternator. Data on the voltage regulation, speed control, power consumption, reliability and transient response are presented for the breadboard.

Birchenough, A. G.↗
DOE OSTI2016

Interfacial Area Transport through Vertical-Downward Elbows in Air-water Two-phase Flow

In the study of two-phase flow, models are often developed by the data acquired in a test section with fixed orientation and with no flow restrictions. In view of practical nuclear reactor systems, however, two-phase flow through flow restrictions is important. Flow restrictions and orientations can alter both global and local two-phase flow parameters such as: cause additional pressure loss; alter the flow regime transition; change bubble distributions; affect the bubble velocity and the bubble interaction mechanisms. However, there is insufficient database and predictive models that can account for such effects. In view of above, the objectives of the current study are: (1) to establish database for local and global two-phase flow parameters downstream of a horizontal-to-vertical glass elbow with a non-dimensional radius of curvature, R{sub c}/D, of three, (2) to develop predictive models in the elbow region, and (3) to apply the developed model to the interfacial area transport equation (IATE) and evaluate it in elbow regions. (authors)

42 ENGINEERING↗
DOE OSTI2014

Eikonal Corrections for Nucleus-Nucleus Reaction Cross Sections - Paper 90

The goal of this research is to modify a Fortran computer code, which uses an optical potential approximation to quantum mechanical multiple-scattering theory to compute total and reaction cross sections for nucleon-nucleus and nucleus-nucleus collisions for use in space radiation protection and shielding applications. These nuclear collision cross sections are utilized in space radiation transport codes that describe the passage of solar and galactic cosmic radiation fields through matter. In the code being modified by this work, an Eikonal approximation, in analogy with the complex index of refraction used to describe absorption in optics, is used to formulate the nuclear scattering amplitudes that describe the total and reaction cross sections in the formalism. Previous work found that the cross sections calculated with this model sometimes underestimated the experimental data. It has been noted that the Eikonal phase functions, which are used to calculate these cross sections and which contain all of the scattering information in the model, can be improved by incorporating higher order corrections In this work, a previously-developed first-order correction to the Eikonal phase function is coded in Fortran and incorporated into the optical model computer code that calculates reaction cross sections for nucleon-nucleus and nucleus-nucleus collisions at energies relevant to space radiation protection and shielding. The main thrust of this work is to test the use of this higher order correction by calculating reaction cross sections for a variety of nuclear collisions and comparing the results to experimental data. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2021

Quasielastic electromagnetic scattering cross sections and world data comparisons in the GENIE Monte Carlo event generator

The usage of Monte Carlo neutrino event generators (MC$\textit{ν}$EGs) is a norm within the high-energy $\textit{ν}$ scattering community. The relevance of quasielastic (QE) energy regimes to $\textit{ν}$ oscillation experiments implies that accurate calculations of $\textit{ν}$A cross sections in this regime will be a key contributor to reducing the systematic uncertainties affecting the extraction of oscillation parameters. In spite of this, many MC$\textit{ν}$EGs utilize highly phenomenological, parametrized models of QE scattering cross sections. Moreover, a culture of validation of MC$\textit{ν}$EGs against prolific electron ($\textit{e}$) scattering data has been historically lacking. In this work, we implement new eA cross sections obtained from nuclear ab initio approaches in GENIE, the primary MC$\textit{ν}$EG utilized by the FNAL community. In particular, we utilize results from quantum MC methods which solve the many-body nuclear problem in the short-time approximation (STA), allowing consistent retention of two-nucleon dynamics which are crucial to explain available nuclear electromagnetic (electroweak) data over a wide range of energy and momentum transfers. This new implementation in GENIE is fully tested against the world QE electromagnetic data, finding agreement with available data below ~2 GeV of beam energy with the aid of a scaling function formalism. The STA is currently limited to study $\textit{A}$ ≤ 12 nuclei, however, its semi-inclusive multibody identity components are exportable to other many-body computational techniques such as auxiliary field diffusion MC which can reach $\textit{A}$ ≤ 40 systems while continuing to realize the factorization contained within the STA’s multinucleon dynamics. Together, these developments promise to make future experiments such as DUNE more accurate in their assessment of MC$\textit{ν}$EG systematics, $\textit{ν}$ properties, and potentially empower the discovery of physics beyond the Standard Model.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2025

Stockpile Stewardship and Nuclear Testing: A Technical Assessment

Since 1992, the United States has retained confidence in its nuclear weapon stockpile without performing any tests that produce nuclear yield. Instead, it has invested in a stockpile stewardship program (SSP) based on the same approach that was validated through fifty years of experience during the nuclear testing era. As the test moratorium continues and new information becomes available, it is both necessary and appropriate to periodically revisit the question of whether or not nuclear testing should resume. In a recent article by Dr. Mark Schneider, he asserts that the United States should return to nuclear testing to address the issues that negatively affect our country’s nuclear deterrent. A serious, technically informed evaluation of the most significant of the issues Dr. Schneider has raised is provided in this paper. The conclusion of this evaluation is that the strategy of a well-funded SSP coupled with rigorous assessments to identify if there is a specific need for a nuclear test is sound and cost-effective approach. Any decision otherwise needs to take into careful consideration all the classified SSP successes and data pertaining to the particular issues in question.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗
DOE OSTI2020

Small-Scale Drying: FY2020 Interim Report

This report documents the interim experimental status of small-scale drying tests currently being performed at Pacific Northwest National Laboratory. These small-scale drying tests are to develop the technical bases for sensors, techniques, and approaches that will be used to address the challenges faced when determining moisture content from gas samples taken at the North Anna Nuclear Generating Station for the High Burnup Demonstration Project. Specifically, the purpose of these tests is to support Sandia National Laboratories gas sample methods and analyses through a series of small-scale experiments. Key focus areas include performing tests to: • Correlate sample bottle measurements to in-cask conditions • Estimate the amount of trapped and absorbed water on cask features and surfaces that can contribute to water vapor • Assess the accuracy of the hygrometer used in the measurements under similar conditions • Identify additional methods for measuring humidity in casks easily and accurately using lessons-learned from Sandia National Laboratories, including direct gravimetric and isotopic tracer techniques. To address these topics this report outlines a series of five tests and any data collected to date that were planned for FY 2020 that include 1) humidity measurement method development, 2) gas bottle sampling tests, 3) surface drying tests on cladding, 4) small-scale drying of cladding, and 5) small-scale drying of guide-tube/dashpots. This interim report documents the experimental background, design and progress of these small-scale drying tests. Further data gathered will be summarized in the FY 2021 report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Nuclear Facility Accident (NFAC) Unit Test Report For HPAC Version 6.3

This is a unit test report for the Nuclear Facility Accident (NFAC) model for the Hazard Prediction and Assessment Capability (HPAC) version 6.3. NFAC’s responsibility as an HPAC component is three-fold. First, it must present an interactive graphical user interface (GUI) by which users can view and edit the definition of an NFAC incident. Second, for each incident defined, NFAC must interact with RTH to create activity table inputs and associate them with pseudo materials to be transported via SCIPUFF. Third, NFAC must create SCIPUFF releases with the associated pseudo materials for transport and dispersion. The goal of NFAC unit testing is to verify that the inputs it produces are correct for the source term or model definition as specified by the user via the GUI.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Experimental test plan for PWR sister rods in the high burnup spent fuel data project

The High Burnup Spent Fuel Data Project (herein referred to as the 'project'), sponsored by the US Department of Energy (DOE) Office of Nuclear Energy (NE), is focused on experimentally defining the effects of long-term storage and transportation on high burnup (HBU) pressurized water reactor (PWR) fuel (>45 gigawatt days per metric ton uranium). The goals of the project are to provide confirmatory data for model validation and potential improvement; to provide input to future dry storage cask design; to support license renewals and new licenses for Independent Spent Fuel Storage Installation facilities; and to support transportation licensing for high burnup used fuel. In support of these project goals, an instrumented TN-32B bolted-lid cask at the North Anna reactor site will be loaded with intact, HBU PWR fuel assemblies (project assemblies) having four different kinds of cladding: standard Zircaloy-4, low-tin Zircaloy-4, ZIRLO, and M5. In approximately 10 years, the instrumented cask will be opened for inspection and several fuel rods from the project assemblies will be characterized. To support the project goals, the project pulled 25 'sister rods', 9 from the project assemblies and 16 from similar HBU assemblies, for characterization. The 25 sister rods are all high burnup and cover the range of cladding alloys. The 25 sister rods were shipped to Oak Ridge National Laboratory (ORNL) in early 2016 for detailed nondestructive and destructive examination. Examinations are intended to provide baseline data on the initial physical state of the cladding and fuel prior to the loading, drying, and long-term dry storage process. Further examinations are focused on determining the effects of temperatures encountered during and following drying. Similar tests will be performed on rods taken from the project assemblies at the end of the storage period (the 'cask rods') to identify any significant changes in the fuel rods that may have occurred during the dry storage period. Additionally, some of the sister rods will be used for separate effects testing to expand the applicability of the project data to the fleet, and to address some of the data-related gaps associated with extended storage and subsequent transportation of high burnup fuel. A draft test plan is being developed that describes the experimental work to be conducted on the sister rods. This paper summarizes the draft test plan and necessary coordination activities for the multi-year experimental program to supply data relevant to the assessment of the safety of long-term storage followed by transportation of high burnup used fuel. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Adjoint sensitivity analysis and data assimilation for verification of dry storage cask contents

Dry cask storage is a method for interim storage of spent fuel assemblies which contain fissile isotopes of uranium and plutonium. These can present a proliferation concern and consequently there is a need for non-destructive testing methods to verify a dry cask's contents for proliferation protection. We present an application of adjoint sensitivity analysis and data assimilation to a multigroup diffusion model of dry cask storage. Adjoint sensitivity analysis allows the efficient calculation of sensitivities for use in data assimilation to calibrate imprecisely known parameter values and data consistency tests to detect diversion scenarios. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

IER 480: TEX-Pu Benchmark (PU-MET-THERM-004) to Test Polyethylene and Lucite Thermal Scattering Laws [Slides]

This presentation finds that PMT-004 has four new benchmark cases highly sensitive to PE TSL (2 cases) and PMMA TSL (2 Cases). PE cases were well predicted using MCNP6.2 and ENDF/B-VIII.0. PMMA cases overpredicted by approximately 0.6-0.7% in k eff at 20°C. Accepted into 2023 version of the ICSBEP Handbook. Temperature had a large impact on reactivity of the critical configuration. Implications for validation work for thermal cases- need to adjust TSL data to correct temperature as it can have hundreds of pcm effects for a few °C. Future thermal experiments should try and measure reactivity at multiple temperatures to aid in data testing and benchmark adjustment.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

DOE Advanced Gas Reactor Fuel Development and Qualification Program Overview

AGR-3/4 post-irradiation examination and data analysis AGR-5/6/7 PIE and safety testing Supplemental fuel microanalysis and method development Fuel oxidation testing Air/moisture Ingress Experiment (AMIX) system development (deployed in FY23) Single particle testing in FITT Data management and analysis Fuel performance modeling

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Sister Rod Destructive Examinations (FY2021)

As a part of DOE NE High Burnup Spent Fuel Data Project, Oak Ridge National Laboratory (ORNL) is performing destructive examinations (DEs) of high burnup (HBU) (>45 GWd/MTU) spent nuclear fuel (SNF) rods from the North Anna Nuclear Power Station operated by Dominion Energy. The SNF rods, called sister rods or sibling rods, are all HBU and include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin (LT) Zirc-4, ZIRLO, and M5. This report documents the status of the ORNL Phase 1 DEs of 7 sister rods and outlines the DE tasks performed and the data collected to date, as guided by the sister rod test plans.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗