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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 631 records · Page 35
All sourcesNASA NTRSDOE OSTI
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 OSTI2011

Pre-test CFD Calculations for a Bypass Flow Standard Problem

The bypass flow in a prismatic high temperature gas-cooled reactor (HTGR) is the flow that occurs between adjacent graphite blocks. Gaps exist between blocks due to variances in their manufacture and installation and because of the expansion and shrinkage of the blocks from heating and irradiation. Although the temperature of fuel compacts and graphite is sensitive to the presence of bypass flow, there is great uncertainty in the level and effects of the bypass flow. The Next Generation Nuclear Plant (NGNP) program at the Idaho National Laboratory has undertaken to produce experimental data of isothermal bypass flow between three adjacent graphite blocks. These data are intended to provide validation for computational fluid dynamic (CFD) analyses of the bypass flow. Such validation data sets are called Standard Problems in the nuclear safety analysis field. Details of the experimental apparatus as well as several pre-test calculations of the bypass flow are provided. Pre-test calculations are useful in examining the nature of the flow and to see if there are any problems associated with the flow and its measurement. The apparatus is designed to be able to provide three different gap widths in the vertical direction (the direction of the normal coolant flow) and two gap widths in the horizontal direction. It is expected that the vertical bypass flow will range from laminar to transitional to turbulent flow for the different gap widths that will be available.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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↗
DOE OSTI2018

Nevada National Security Site Integrated Groundwater Sampling Plan, Revision 1, with ROTC-1

The purpose of the Nevada National Security Site (NNSS) Integrated Groundwater Sampling Plan (referred to herein as the Plan) is to provide a comprehensive, integrated approach for collecting and analyzing groundwater samples to meet the needs and objectives of the U.S. Department of Energy (DOE), Environmental Management (EM) Nevada Program’s Underground Test Area (UGTA) Activity. Implementation of this Plan will provide high-quality data required by the UGTA Activity for ensuring public protection in an efficient and cost-effective manner. The Plan is designed to ensure compliance with the UGTA Quality Assurance Plan (QAP) (NNSA/NFO, 2015). The Plan’s scope comprises sample collection and analysis requirements relevant to assessing both the extent of groundwater contamination from underground nuclear testing and impact of testing on water quality in downgradient communities. This Plan identifies locations to be sampled by corrective action unit (CAU) and location type, sampling frequencies, sample collection methodologies, and the constituents to be analyzed. In addition, the Plan defines data collection criteria such as well purging, detection levels, and accuracy requirements/recommendations; identifies reporting and data management requirements; and provides a process to ensure coordination between NNSS groundwater sampling programs for sampling analytes of interest to UGTA. Information used in the Plan development—including the rationale for selection of wells, sampling frequency, and the analytical suite—is discussed under separate cover (N-I, 2014) and is not reproduced herein.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2015

Benchmarking nuclear fission theory

We suggest a small set of fission observables to be used as test cases for validation of theoretical calculations. Thus, the purpose is to provide common data to facilitate the comparison of different fission theories and models. The proposed observables are chosen from fission barriers, spontaneous fission lifetimes, fission yield characteristics, and fission isomer excitation energies.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Graphite Baseline Status

Subtask Relevancy Inherent variation exists in nuclear-grade graphite due to processing techniques and raw material sources Comprehensive graphite properties database does not exist that fully establishes physical and mechanical property relationships Predictability of long-term behavior in a nuclear environment requires full analysis of variability Technical Approach, Accomplishments/Results Perform comprehensive mechanical and physical properties testing to populate a fully-validated database Grade selection is based upon candidate graphites from four international suppliers 15,000+ property values measured to date Physical properties testing that is based upon AGC geometries is a key component to establishing baseline/AGC data link

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

A correlation-based approach for evaluating mechanical properties of nuclear fuel cladding tubes

The present work offers and evaluates a correlation-based approach to determine the tensile properties of nuclear fuel cladding tubes. A set of model materials with known properties (i.e., austenitic and ferritic steels, fcc- and hcp-alloys) were tested to build a correlation between conventional tensile tests and ring (hoop) and tube (axial) tests. It was shown that close-to-linear relationships exist for the basic mechanical properties (i.e., yield and ultimate stress, uniform and total elongation) between the uniaxial tensile test data and the ring and tube test data. Using the ring and tube specimen geometries, the feasibility of the approach was demonstrated via manufacturing and testing specimens from irradiated Zr and FeCrAl nuclear fuel cladding tubes. Limitations caused by specific deformation mechanisms (e.g., high ductility of 304L steel) were discussed in detail. The approach can be extended to other tubular products with different geometry and dimensions.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Thermonuclear reactions probed at stellar-core conditions with laser-based inertial-confinement fusion

Stars are giant thermonuclear plasma furnaces that slowly fuse the lighter elements in the universe into heavier elements, releasing energy, and generating the pressure required to prevent collapse. To understand stars, we must rely on nuclear reaction rate data obtained, up to now, under conditions very different from those of stellar cores. Here we show thermonuclear measurements of the 2 H(d, n) 3 He and 3 H(t,2n) 4 He S-factors at a range of densities (1.2–16 g cm –3 ) and temperatures (2.1–5.4 keV) that allow us to test the conditions of the hydrogen-burning phase of main-sequence stars. The relevant conditions are created using inertial-confinement fusion implosions at the National Ignition Facility. Here, our data agree within uncertainty with previous accelerator-based measurements and establish this approach for future experiments to measure other reactions and to test plasma-nuclear effects present in stellar interiors, such as plasma electron screening, directly in the environments where they occur.

42 ENGINEERING↗
DOE OSTI2016

The Nuclear Energy Knowledge and Validation Center – Summary of Activities Conducted in FY15

The Nuclear Energy Knowledge and Validation Center (NEKVaC) is a new initiative by the Department of Energy and the Idaho National Laboratory to coordinate and focus the resources and expertise that exist with the DOE Complex toward solving issues in modern nuclear code validation. In time, code owners, users, and developers will view the Center as a partner and essential resource for acquiring the best practices and latest techniques for validating codes, for guidance in planning and executing experiments, for facilitating access to, and maximizing the usefulness of, existing data, and for preserving knowledge for continual use by nuclear professionals and organizations for their own validation needs. The scope of the center covers many inter-related activities which will need to be cultivated carefully in the near-term and managed properly once the Center is fully functional. Three areas comprise the principal mission: 1) identification and prioritization of projects that extend the field of validation science and its application to modern codes, 2) adapt or develop best practices and guidelines for high fidelity multiphysics/multiscale analysis code development and associated experiment design, and 3) define protocols for data acquisition and knowledge preservation and provide a portal for access to databases currently scattered among numerous organizations. These mission areas, while each having a unique focus, are inter-dependent and complementary. Likewise, all activities supported by the NEKVaC, both near-term and long-term), must possess elements supporting all three. This cross-cutting nature is essential to ensuring that activities and supporting personnel do not become ‘stove-piped’, i.e. focused so much on a specific function that the activity itself becomes the objective rather than the achieving the larger vision. Achieving the broader vision will require a healthy and accountable level of activity in each of the areas. This will take time and significant DOE support. Growing too fast (budget-wise) will not allow ideas to mature, lessons to be learned, and taxpayer money to be spent responsibly. The process should be initiated with a small set of tasks, executed over a short but reasonable term, that will exercise most if not all aspects of the Center’s potential operation. The initial activities described in this report have a high potential for near-term success in demonstrating Center objectives but also to work out some of the issues in task execution, communication between functional elements, and the ability to raise awareness of the Center and cement stakeholder buy-in. This report begins with a description of the Mission areas; specifically the role played by each and the types of activities for which they are responsible. It then lists and describes the proposed near-term tasks upon which future efforts can build.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗