Search NASA⌕ Search

SEARCH · Search NASA

Results for “Nuclear Data Testing”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 613 records · Page 34
All sourcesNASA NTRSDOE OSTI
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↗
DOE OSTI2015

Updated Results of Ultrasonic Transducer Irradiation Test

Ultrasonic technologies offer the potential for high accuracy and resolution in-pile measurement of a range of parameters, including geometry changes, temperature, crack initiation and growth, gas pressure and composition, and microstructural changes. Many Department of Energy-Office of Nuclear Energy (DOE-NE) programs are exploring the use of ultrasonic technologies to provide enhanced sensors for in-pile instrumentation during irradiation testing. For example, the ability of small diameter ultrasonic thermometers (UTs) to provide a temperature profile in candidate metallic and oxide fuel would provide much needed data for validating new fuel performance models. These efforts are limited by the lack of identified ultrasonic transducer materials capable of long term performance under irradiation test conditions. To address this need, the Pennsylvania State University (PSU) was awarded an Advanced Test Reactor National Scientific User Facility (ATR NSUF) project to evaluate the performance of promising magnetostrictive and piezoelectric transducers in the Massachusetts Institute of Technology Research Reactor (MITR) up to a fast fluence of at least 10{sup 21} n/cm{sup 2}. A multi-National Laboratory collaboration funded by the Nuclear Energy Enabling Technologies Advanced Sensors and Instrumentation (NEET-ASI) program also provided initial support for this effort. This irradiation, which started in February 2014, is an instrumented lead test and real-time transducer performance data are collected along with temperature and neutron and gamma flux data. The irradiation is ongoing and will continue to approximately mid-2015. To date, very encouraging results have been attained as several transducers continue to operate under irradiation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRSDate not supplied

Regulatory Approach for Nuclear Thermal Propulsion Reactor Systems

Nuclear Thermal Propulsion (NTP) is being developed to support crewed or cargo transfer missions to Mars. Obtaining regulatory approval for NTP engine testing proves a challenge for not only the technology maturation required but also the fabrication, launch and operational nuclear regulatory environments. Existing regulations currently apply to either (1) high power, long duration commercial power plants, or (2) low power, short duration research reactors. NTP systems find themselves in a separate and unique area due to their higher power but short operating duration. This is supplemented by use of the reactor coolant as propellant. When planning the regulatory approach for NTP systems, operations in space and testing on Earth must be considered – each have their own challenges. This paper identifies a preliminary pathway for regulatory approval to operate a NTP demonstration engine as well as supporting test data that is recommended to be generated during the development program to support major regulatory milestones and deliverables.

Regulatory↗
DOE OSTI2019

Salt Heater Test (FY19), Rev. 2

This report summarizes the 2019 fiscal year (FY19) status of the borehole heater test in salt funded by the US Department of Energy Office of Nuclear Energy (DOE-NE) Spent Fuel and Waste Science & Technology (SFWST) campaign. This report satisfies SFWST level-three milestone report M3SF-19SN010303033. This report is an update of the April 2019 level-two milestone report M2SF-19SNO10303031 to reflect the nearly complete as-built status of the borehole heater test. This report discusses the fiscal year 2019 (FY19) design, implementation, and preliminary data interpretation plan for a set of borehole heater tests call the brine availability tests in salt (BATS), which is funded by the DOE Office of Nuclear Energy (DOE-NE) at the Waste Isolation Pilot Plant (WIPP), a DOE Office of Environmental Management (DOE-EM) site. The organization of BATS is outlined in Project Plan: Salt In-Situ Heater Test (SNL, 2018). An early design of the field test is laid out in Kuhlman et al. (2017), including extensive references to previous field tests, which illustrates aspects of the present test. The previous test plan by Stauffer et al. (2015) places BATS in the context of a multi-year testing strategy, which involves tests of multiple scales and processes, eventually culminating in a drift-scale disposal demonstration. This level-3 milestone report is an update of a level-2 milestone report from April 2019 by the same name. The update adds as-built details of the heater test, which at the time of writing (August 2019) is near complete implementation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Recovery and calibration of legacy analog data from the Leo Brady Seismic Network for the Source Physics Experiment

The Leo Brady Seismic Network (LBSN) was established in 1960 by Sandia National Laboratories for monitoring underground nuclear tests (UGTs) at the Nevada Test Site— renamed in 2010 to the Nevada National Security Site (NNSS). The LBSN has been in various configurations throughout its existence, but it has been generally comprised of four to six stations at regional distances from the NNSS with evenly spaced azimuthal coverage. Between 1962 and the early 1980s, the LBSN—and a sister network operated by Lawrence Livermore National Laboratory—were the most comprehensive U.S. source of regional seismic data of UGTs. During the pre-digital era, LBSN data were transmitted as frequency-modulated (FM) audio over telephone lines to the NTS and recorded in analog on hi-fi 8-track AMPEX tapes. These tapes have been stored in temperature-stable buildings or bunkers on the NNSS and Kirtland Air Force Base in Albuquerque, NM for decades and contain the sole record of this irreplaceable data from the analog era; full waveforms of UGTs during this time were never routinely converted to digital form. We have been developing a process over the past few years to recover and calibrate data from these tapes, converting them from FM audio to digital waveforms in ground motion units. The calibration of legacy data from the LBSN is still ongoing. To date, we have digitized tapes from 592 separate UGTs. As a proof-of-concept, we calibrated data from the BOXCAR event.

58 GEOSCIENCES↗
DOE OSTI2014

Design and Scaling of the Natural Convection Shutdown Heat Removal Test Facility

The Natural convection Shutdown heat removal Test Facility (NSTF) at Argonne National Laboratory (ANL) reflects a 1/2 scale model of one conceptual design for passive safety in advanced reactors. The project was initiated in 2010 primarily to conduct ex-vessel, passive decay heat removal experiments in support of the Advanced Reactor Concepts (ARC), Small Modular Reactor (SMR), and Next Generation Nuclear Plant (NGNP) programs while also generating data for code validation purposes. The facility successfully demonstrated scoping objectives in late 2013, and is expected to begin testing by early 2014. The following paper summarizes some of the key design and scaling considerations used in construction of the experimental facility, along with an overview of the current instrumentation and data acquisition methods. Details of the distributed fiber optic temperature system will be presented, which introduces a level of data density suitable for CFD validation and is a first-of-its-kind for largescale thermal hydraulics facilities.

Lisowski, Darius D.↗
DOE OSTI2020

Neutronic Consideration of TREAT Facility Fuel SiC Recladding

The Transient Reactor Test (TREAT) Facility is an air-cooled reactor composed of graphite-urania fuel encapsulated in zirconium alloy canisters and surrounded by ~2 ft (~61 cm) of nuclear grade graphite. The primary role of TREAT is to provide in-pile nuclear-heated safety research via transient testing [1]. Benchmark efforts have demonstrated sensitivity of TREAT neutronics models to graphite cross section data and key neutron poisons found in the fuel assemblies [2]. During fabrication of TREAT fuel assemblies, a small concentration of boron was inadvertently introduced with variable concentration into the aggregate of fuel blocks prepared for TREAT use. The concentration of boron, and its respective uncertainty, represent the most sensitive component in a neutronics study; unfortunately, nothing can be done to reduce or otherwise mitigate its impact. Hydrogen in the fuel blocks reduces the mean free path of the core and its true quantity and uncertainty is approximated. The hydrogen remains from pitch binder used in the fuel fabrication process and was not completely released due to incomplete fuel graphitization. Destructive fuel testing is necessary to quantify the actual hydrogen content and to improve benchmark model quality. A third neutron poison of some significance is the hafnium impurity in the Zircaloy-3 (Zry-3) assembly cladding. The quantity of hafnium is expected to be within nominal manufacturing limits for nuclear grade zircaloy. Recladding the fuel assemblies with material that is less neutron absorbing is expected to benefit reactor neutron economy. Moreover, TREAT’s total transient energy is restricted based on the rate of Zry-3 oxidation in TREAT’s air coolant environment. Hence, cladding with more benign air reaction kinetics could dramatically uprate the facility’s experimental capability. Silicon carbide (SiC) has been proposed as a candidate material for recladding TREAT assemblies, as SiC/SiC demonstrate favorable properties within nuclear environments [3]. The current operational limits of TREAT are restricted by the oxidation of Zry-3. Change of the cladding to SiC could feasibly double the current temperature limitation. Recladding opportunities allow for previously damaged fuel assemblies to be returned to service, as Zry-3 is not commercially available. The initial effort required to consider proceeding with any recladding effort is to neutronically characterize key reactor parameters while replacing Zry-3 with SiC. A characterization summary, with respective results, is provided herein.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Experimental Investigation of Two-Phase Flow in Consolidating and Intact Salt: Analysis of Capillary Pressure Measurements from Core Labs

Two-phase flow of brine and air through very low permeability porous media is the dominant fluid flow state in and around geologic repositories for nuclear waste disposal. Constitutive models developed for other subsurface media (shale, sandstone, limestone, etc.) have been adopted but never experimentally parameterized for geologic salt. This report summarizes results of scoping capillary pressure and brine saturation measurements on reconsolidated run-of-mine salt and intact geologic salt samples. These measurements were obtained using mercury injection tests and yielded values of threshold pressure, pore-size distribution, and mercury saturation. The brine capillary pressure curve was then inferred from the raw mercury capillary pressure data. This report also summarizes estimates of the constitutive model parameters from the brine saturation and capillary pressure data (Brooks and Corey, 1964; van Genuchten, 1980). The results suggests that the model of van Genuchten (1980) is better suited for brine-air-salt systems. Additional testing is needed for the low brine saturation range of the capillary pressure curve where the present suite of tests did not yield sufficient data.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗