Search NASA⌕ Search

SEARCH · Search NASA

Results for “Nuclear 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 1,603 records · Page 89
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

The Nuclear Fuel Cycle: Safe Management of Spent Nuclear Fuel

The aim for storage of spent nuclear fuel (SNF) either in wet or in dry storage systems is to ensure general safety objective s are met throughout a desired storage period. Staff at the Savannah River National Laboratory (SRNL), in collaborations with partners at other national laboratories, industry research organizations, and the University of South Carolina (UofSC), have performed materials aging testing and analyses, and have established nuclear materials aging management programs to support extended periods of safe storage of research reactor (RR) SNF and of commercial power reactor (PR) SNF pending ultimate disposal. Several example challenges include susceptibility of aluminum SNF from research reactors to corrosion in poor quality water (wet storage). In dry storage, aluminum SNF can release hydrogen via radiolysis of the hydrated oxides on the aluminum cladding. Austenitic stainless steel canisters used for dry storage are susceptible to chloride-induced stress corrosion cracking (outside-in attack) that threaten the confinement boundary provided by the canister. This paper further describes these challenges, among others, and the formulated solutions to support extended safe storage of SNF.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Aluminum Clad Spent Nuclear Fuel Task 3: Sealed and Vented Systems Episodic Breathing and Gas Generation Modeling Plan

This document presents the test plan for Task 3 of the Aluminum Clad Spent Nuclear Fuel Long Term Dry Storage Technical Issues Action Plan, which covers modeling and predictions of the generations of corrosive gases within both sealed and vented, subject to episodic breathing, storage systems. This includes aluminum clad spent nuclear fuel (ASNF) that is currently or may in the future be held in dry storage for an extended period. The overall objectives of the Task 3 modeling plan are to: 1. Comprehensively model the combined effects of episodic breathing in vented storage systems/canisters using INL INTEC CPP-603 facility as an example. Both bulk gas phase radiolysis reactions and aluminum cladding surface gas generation reactions will be included into a 3-dimensional computational fluid dynamics (CFD) model to model and predict long-term evolutions of the temperature and concentrations of corrosive gases within unsealed canisters subjected to episodic breathing. 2. Comprehensively model sealed and inert canister storage system using DOE standard canister as an example. 3. Closely integrate with Task 1 and Task 2 for identification of main gas-generation reactions and kinetics and provide simulation results back to Task 1 and 2 to further refine experimental designs. This test plan lays out the technical activities that will be performed to achieve the test objectives described above. Simulations will be performed using the Idaho National Laboratory super computer Falcon.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Leveraging Optimal Sparse Sensor Placement to Aggregate a Network of Digital Twins for Nuclear Subsystems

Nuclear power plants (NPPs) require continuous monitoring of various systems, structures, and components to ensure safe and efficient operations. The critical safety testing of new fuel compositions and the analysis of the effects of power transients on core temperatures can be achieved through modeling and simulations. They capture the dynamics of the physical phenomenon associated with failure modes and facilitate the creation of digital twins (DTs). Accurate reconstruction of fields of interest (e.g., temperature, pressure, velocity) from sensor measurements is crucial to establish a two-way communication between physical experiments and models. Sensor placement is highly constrained in most nuclear subsystems due to challenging operating conditions and inherent spatial limitations. This study develops optimized data-driven sensor placements for full-field reconstruction within reactor and steam generator subsystems of NPPs. Optimized constrained sensors reconstruct field of interest within a tri-structural isotropic (TRISO) fuel irradiation experiment, a lumped parameter model of a nuclear fuel test rod and a steam generator. The optimization procedure leverages reduced-order models of flow physics to provide a highly accurate full-field reconstruction of responses of interest, noise-induced uncertainty quantification and physically feasible sensor locations. Accurate sensor-based reconstructions establish a foundation for the digital twinning of subsystems, culminating in a comprehensive DT aggregate of an NPP.

42 ENGINEERING↗
DOE OSTI2020

Recent Advances in Nuclear Forensic Chemistry

Nuclear forensics is a rapidly growing field with a widespread impact on global nonproliferation. From the beginning, nuclear forensic capabilities have been inextricably bound to existing limitations within chemistry and materials science. As nuclear forensics has continued to develop as a unique field of research, new advances in chemistry have revolutionized our ability to rapidly and quantitatively analyze radioactive materials for attribution. In particular, microanalytical techniques originally developed for nanoscience, such as micro-XRF, SIMS, and SEM-EDX, have enabled spatially resolved chemical analyses to be performed with only nanograms of material. Such techniques have redefined the paradigm for pre- and postdetonation nuclear forensics, as large quantities of material are no longer necessary for full characterization and attribution. This Review explains how modern nuclear forensic techniques have developed from their origins in chemistry and provides real-world examples of how these techniques have been used to characterize radioactive materials, including fallout from nuclear detonations such as the 1945 Trinity test.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2022

Modeling and Simulation of Austenitic Welds and Coarse-grained Specimens: Part II

The Pacific Northwest National Laboratory (PNNL) is conducting confirmatory research for the U.S. Nuclear Regulatory Commission (NRC) to evaluate commercially available nondestructive examination (NDE) modeling and simulation software used in the nuclear industry. Simulation results from ultrasonic testing (UT) models can inform the design and qualification of inspection techniques and help interpret inspection results. CIVA is a modeling and simulation package developed by the French Alternative Energies and Atomic Energy Commission (CEA). CIVA was selected for this study because it is readily available and has been used for NDE in the US commercial nuclear power industry. This report is focused on completing the efforts initiated in the previous PNNL report to evaluate UT modeling and simulation performance, reliability, and accuracy in relation to common inservice inspection (ISI) scenarios in nuclear power plants (NPP). This work will be used to provide guidance when establishing methods to perform and evaluate simulations for more standardized model implementation, simulation analysis, and interpretation of results.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Spent Nuclear Fuel Mechanical Loads in the General Package Drop Scenario

The US Department of Energy Spent Fuel and Waste Science and Technology program is performing research to determine the mechanical loading conditions applied to spent nuclear fuel (SNF) during normal conditions of transportation to inform mechanical tests of SNF and close an important knowledge gap related to the practical disposition of SNF in the US. A recent multi-national collaborative test campaign measured SNF assembly impact response to the 30 cm horizontal package drop scenario, which is a common regulatory basis test of SNF package design. Researchers at Pacific Northwest National Laboratory (PNNL) are using the test data to validate explicit finite element models to calculate the mechanical loads and structural response of spent nuclear fuel assemblies in the as-tested 30 cm horizontal package drop scenario. Once the as-tested package drop model is validated, the next step is to apply the model to the general 30 cm drop scenario, which includes all impact angles, all fuel assembly types, and all burnup conditions. PNNL is developing a damage model that will incorporate the results of finite element parametric studies to establish trends in SNF mechanical loading to various input parameters, like impact orientation and burnup. The damage model will have the capability to estimate mechanical loads for any single set of input parameters, but it will first be used to describe the upper bounds of potential SNF loading in the 30 cm drop scenario. This paper describes PNNL’s progress toward developing the general solution to the problem of spent nuclear fuel mechanical loads in the 30 cm package drop scenario and it describes the next steps in closing the knowledge gap.

Klymyshyn, Nicholas A.↗
DOE OSTI2022

FAST irradiations and initial post irradiation examinations – Part I

The Advanced Fuels Campaign Fission Accelerated Steady-state Test (FAST) at Idaho National Laboratory (INL) completed its first irradiation cycle within the Advanced Test Reactor (ATR). The test focused on the irradiation of alloy fuel forms for use in sodium fast reactors. Tests included the use of fuel additives (Sn, Sb, and Pd) for reducing the gettering of lanthanide fission products, zirconium liners to mitigate fuel-cladding chemical interaction, and annular fuel geometry for sodium-free fuel designs. The first cycle of FAST testing was completed and a series of rodlets were removed for the initial post irradiation examination (PIE) of FAST rodlets. The rodlet irradiation conditions was evaluated using Monte Carlo N-Particle (MCNP) for as-run power history and COMSOL for temperature analysis. These rodlets include a set of low burnups (~2.5 % heavy metal [%FIMA]) control rodlets and a helium bonded annular rodlet (4.7 %FIMA). Initial non-destructive PIE has been completed and includes, visual inspection, neutron radiography and gamma scanning of the FAST rodlets. Radiography confirmed the integrity of the experiments as well as showing the complete filling of the annulus in the annular fuel at a modest burnup (4.7 %FIMA) and potential slumping of the cooler rodlets at lower burnup. Precision gamma scanning indicated mostly usual fission product behavior with the exception of the ceasium in the He-bonded annular fuel. Future destructive PIE will be necessary in order to fully understand the effects of accelerated irradiation on U-Zr metallic fuel behavior.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Final Seismic Shake Table Test Plan

The Spent Fuel Waste Disposition (SFWD) program is planning to conduct a full-scale seismic shake table test on the dry storage systems of spent nuclear fuel (SNF) to close the gap related to seismic loads on fuel assemblies in dry storage systems. This test will allow for quantifying the strains and accelerations on surrogate fuel assembly hardware and cladding during earthquakes of different magnitudes and frequency content. Full-scale testing is needed because a dry storage system is a complex and highly nonlinear system making it hard to predict (model) the responses to seismic excitations. The non-linearity arises from the multiple spatial gaps in the system – between fuel rods and the basket, between the basket and dry storage canister, between the dry storage canister and the storage cask (overpack), and ventilation gaps. The non-linearities pose significant limitations on the value of tests with scaled systems.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Flexible classification with spatial quantile comparison and novel statistical features applied to spent nuclear fuel analysis

Multivariate classification algorithms are a common tool in contemporary data science that have begun to be applied to nuclear forensic analyses. Using a multivariate generalization of quantile–quantile plots for comparing unknown statistical distributions, a new model-free multivariate classifier called the Quantile–Quantile Comparator has been developed and tested on the analysis of simulated irradiated nuclear fuel. A method to estimate the sample-specific probability of having made an incorrect classification decision by analyzing the distribution of quantile comparisons has been developed, and methods have been developed to analyze the partition of the known data library used to analyze unknown data. By analyzing this partition, the analyst may identify classes that are likely to be confused, identify test data that are inconsistent with the data library and return a “None-of-the-Above” classification decision, and optimize the class boundaries. In this paper, the development of these capabilities is described, and a case study using real experimental assay data collected from nuclear fuel extracted from the newly-released In conclusion, Spent Fuel Isotopic Composition Database to demonstrate the efficacy of these capabilities is presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

An approach for spent nuclear fuel containment integrity verification using gas tagging

Verification of containment integrity is required for spent nuclear fuel (SNF) managed by the commercial nuclear industry and U.S. Department of Energy (DOE), especially after extended storage. Certain SNF storage systems, such as the DOE road-ready dry storage system, hold several packaged containments within a welded over-canister. These packaged containments are called Department of Energy Standard Canisters (DOESCs). DOESC leakage identification is challenging because their containment boundary cannot be accessed for testing and their contents (i.e., SNF and fill gas) are often similar. There are concerns that this could result in costly characterization and repackaging operations of DOE road-ready dry storage systems if compromised DOESCs are suspected. Here, to address these concerns, this paper presents an approach for applying a gas tagging process using xenon to uniquely identify compromised inaccessible containments following extended storage. The containments considered for this application are seven DOESCs, each packaged within a single over-canister. Two different SNF loading configurations from the Advanced Test Reactor and Fort Saint Vrain nuclear power plant are considered. These configurations are used to represent research reactor aluminum-clad spent nuclear fuel (ASNF) and TRi-structural ISOtropic (TRISO) SNF types. Results for this application show that for ASNF and TRISO type fuels for which the selected fuels are representative, the volume of taggant required at loading is determined primarily by the lower detection limit and leak rate of taggant from a compromised DOESC, rather than the amount of fission-generated xenon in the loaded fuel. While the application presented is suited for larger leaks, smaller leaks could be detected by modifying certain design parameters. This gas tagging approach can also be applied to other DOE containments and advanced reactor SNF storage systems.

07 - ISOTOPES AND RADIATION SOURCES↗
DOE OSTI2017

Accelerated Aging Test for Plastic Scintillator Gamma Ray Detectors

Polyvinyl toluene (PVT) and polystyrene (PS), collectively referred to as “plastic scintillator,” are synthetic polymer materials used to detect gamma radiation, and are commonly used in instrumentation. Recent studies have revealed that plastic scintillator undergoes an environmentally related material degradation that adversely affects performance under certain conditions and histories. A significant decrease in gamma ray sensitivity has been seen in some detectors in systems as they age. The degradation of sensitivity of plastic scintillator over time is due to a variety of factors, and the term “aging” is used to encompass all factors. Some plastic scintillator samples show no aging effects (no significant change in sensitivity over more than 10 years), while others show severe aging (significant change in sensitivity in less than 5 years). Aging effects arise from weather (variations in heat and humidity), chemical exposure, mechanical stress, light exposure, and loss of volatile components. The damage produced by these various causes can be cumulative, causing observable damage to increase over time. Damage may be reversible up to some point, but becomes permanent under some conditions. It has been demonstrated that exposure of plastic scintillator in an environmental chamber to 30 days of high temperature and humidity (90% relative humidity and 55°C) followed by a single cycle to cold temperature (-30°C) will produce severe fogging in all PVT samples. This thermal cycle will be referred to as the “Accelerated Aging Test.” This document describes the procedure for performing this Accelerated Aging Test.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2017

Observed Changes in As-Fabricated U-10Mo Monolithic Fuel Microstructures After Irradiation in the Advanced Test Reactor

A low-enriched uranium U-10Mo monolithic nuclear fuel is being developed by the Material Management and Minimization Program, earlier known as the Reduced Enrichment for Research and Test Reactors Program, for utilization in research and test reactors around the world that currently use high-en-riched uranium fuels. As part of this program, reactor experiments are being performed in the Advanced Test Reactor. It must be demonstrated that this fuel type exhibits mechanical integrity, geometric stability, and pre-dictable behavior to high powers and high fission densities in order for it to be a viable fuel for qualification. This paper provides an overview of the microstructures observed at different regions of interest in fuel plates before and after irradiation for fuel samples that have been tested. These fuel plates were fabricated using laboratory-scale fabrication methods. Observations regarding how microstructural changes during irradiation may impact fuel performance are discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2010

Fission Surface Power System Initial Concept Definition

Under the NASA Exploration Technology Development Program (ETDP) and in partnership with the Department of Energy (DOE), NASA has embarked on a project to develop Fission Surface Power (FSP) technology. The primary goals of the project are to 1) develop FSP concepts that meet expected surface power requirements at reasonable cost with added benefits over other options, 2) establish a hardwarebased technical foundation for FSP design concepts and reduce overall development risk, 3) reduce the cost uncertainties for FSP and establish greater credibility for flight system cost estimates, and 4) generate the key products to allow NASA decision-makers to consider FSP as a preferred option for flight development. The FSP project was initiated in 2006 as the Prometheus Program and the Jupiter Icy Moons Orbiter (JIMO) mission were phased-out. As a first step, NASA Headquarters commissioned the Affordable Fission Surface Power System Study to evaluate the potential for an affordable FSP development approach. With a cost-effective FSP strategy identified, the FSP team evaluated design options and selected a Preliminary Reference Concept to guide technology development. Since then, the FSP Preliminary Reference Concept has served as a point-of-departure for several NASA mission architecture studies examining the use of nuclear power and has provided the foundation for a series of "Pathfinder" hardware tests. The long-term technology goal is a Technology Demonstration Unit (TDU) integrated system test using full-scale components and a non-nuclear reactor simulator. The FSP team consists of Glenn Research Center (GRC), Marshall Space Flight Center (MSFC) and the DOE National Laboratories at Los Alamos (LANL), Idaho (INL), Oak Ridge (ORNL), and Sandia (SNL). The project is organized into two main elements: Concept Definition and Risk Reduction. Under Concept Definition, the team performs trade studies, develops analytical tools, and formulates system concepts. Under Risk Reduction the team develops hardware prototypes and conducts laboratory-based testing.

Source record↗
DOE OSTI2018

Steam Oxidation and Burst Testing in the Severe Accident Test Station

Within six months of the March 2011 nuclear accident at Fukushima Daiichi, research began at Oak Ridge National Laboratory (ORNL) to simulate the high temperature steam oxidation accident conditions and identify potential accident tolerant fuel (ATF) cladding [1-14]. The ORNL laboratory equipment for steam oxidation equipment was expanded and made available to the entire ATF research community as the Severe Accident Test Station (SATS) in 2012 [5,1 5], including the capability to conduct the standard integral loss of coolant accident (LOCA) experiment [9]. A second SATS has now been deployed in an ORNL hot cell for handling irradiated specimens including reprocessed UO 2 /Zr fuel [16].

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Electric Form Factor of the Neutron from Asymmetry Measurements

The overwhelming majority of visible mass in the universe is composed of protons and neutrons, collectively known as nucleons, which are arguably the most important strongly interacting systems to study. A nucleon may naively be thought of as a composite object built out of three non-interacting sub-objects, known as quarks, in the large energy limit; however, the structure of the nucleon is much more complicated, and consequently far richer, than a simple valence-quark picture. There are sea quarks in addition to valence quarks that interact strongly via the exchange of gluons resulting in a complex vacuum structure, and the collective system must give rise to the observed properties of the nucleon, e.g. the radius and mass. The nucleon is the most well-studied hadron and yet there are still unresolved complexities in the calculation of properties from first principles of QCD; this represents a central problem in nuclear physics. The theoretical difficulty with the nucleon requires experimentation, which is steadily increasing the knowledge of nucleon structure and the strong interaction. The subject of this thesis pertains to nucleon electromagnetic form factors which are fundamental quantities containing information on the spatial and momentum distributions of charge and current within the nucleon. Nucleon form factors may be accessed through well-understood leading order electromagnetic processes with a leptonic probe, and provide strong constraints on testing non-perturbative QCD and nuclear structure models. The form factor ratio of the neutron has been extracted via a beam-target helicity asymmetry measurement. The Jefferson Lab Hall A experiment E02-013 ran in 2006 utilizing the 6 GeV CEBAF for its high-duty, longitudinally polarized electron beam. The double-arm coincidence experiment detected the quasielastically scattered electrons in a large angular and momentum acceptance spectrometer. The recoiling nucleons were detected and momentum analyzed in a large scintillator-iron based neutron detector. The analysis of a previously unpublished extraction of the electric form factor of the neutron is presented.

Obrecht, Richard↗
DOE OSTI2020

Preliminary Evaluation of Driver Fuel Plates and Reduced Sodium Temperature to Increase Fast Test Reactor Experimental Neutron Irradiation Capabilities

All sodium-cooled fast test reactors operated so far have used essentially the same fuel geometrical ar-rangement, namely cylindrical fuel pins assembled into a triangular array in hexagonal fuel assemblies. Fuel compositions and dimensions may vary but the overall geometrical pattern remains identical. Fur-thermore, most sodium-cooled reactors have operated, and still operate, with core-average inlet and outlet sodium temperatures of, respectively, 350-400°C and 500-550°C. The present preliminary study revisits the standard fuel assembly geometry as well as operating tempera-tures to evaluate potential impact on fast test reactor neutron irradiation capabilities. The reference fuel is U-20Pu-10Zr. Pin (reference) and plate geometries are considered together with two sets of core-average sodium inlet/outlet temperatures: 350/500°C (reference) and 200/400°C. More prototypic environments could be obtained inside ad-hoc irradiation testing vehicles that are independent from the reactor primary coolant. Everything else being the same, fuel temperature is significantly lower in plates than in pins. Therefore, plates could in principle accommodate higher plutonium content, thus, reducing the need for uranium enrichment or allowing the use of lower quality plutonium (U-xPu-10Zr thermal conductivity and solidus temperature decrease as x increases, hence, necessitating additional thermal margins). For the reference pin configuration, lowering the inlet/outlet sodium temperatures from the reference 350/500°C down to 200/400°C provides additional thermal margins that can be used to increase the peak fast flux from about 4.5 × 10 15 n/cm 2 -s to 6 ×10 15 n/cm 2 -s for the same core power of 300 MW. Assuming 300 Equivalent Fuel Power Days (EFPD) of operation per calendar year, a (steel) test article could accu-mulate up to 75 dpa/year. The use of fuel plates provides even more thermal margins which may allow the peak fast flux to reach values as high as 8 × 10 15 n/cm 2 -s for both sets of inlet/outlet temperatures. In this environment, a test article could accumulate up to 95 dpa/year assuming 300 EFPD/year, hence greatly accelerating irradiation testing. Allowing a fast test reactor to operate over a wide range of inlet/outlet temperatures could add significant flexibility to its neutron irradiation capabilities. Steel present in the driver fuel assemblies is not expected to accumulate more than 90 displacements per atom (dpa) over its lifetime. Hence, steel alloys that are not suited for very high dpa but can operate over a wide range of temperatures (from low to high) would be appropriate for this application. Finally, because pin and plate assemblies fit on the same grid plate, a fast test reactor could in principle start-up with standard fuel pin assemblies and, later on, move to a plate-type Mark-II fuel to further in-crease its neutron irradiation capabilities while potentially reducing the level of uranium enrichment needed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Probing chiral interactions up to next-to-next-to-next-to-leading order in medium-mass nuclei

Here, we study ground-state energies and charge radii of closed-shell medium-mass nuclei based on novel chiral nucleon-nucleon (NN) and three-nucleon (3⁢𝑁) interactions, with a focus on exploring the connections between finite nuclei and nuclear matter. To this end, we perform in-medium similarity renormalization group (IM-SRG) calculations based on chiral interactions at next-to-leading order (NLO), N 2 ⁢LO, and N 3 ⁢LO, where the 3⁢𝑁 interactions at N 2⁢ LO and N 3 ⁢LO are fit to the empirical saturation point of nuclear matter and to the triton binding energy. Our results for energies and radii at N 2 ⁢LO and N 3 ⁢LO overlap within uncertainties, and the cutoff variation of the interactions is within the EFT uncertainty band. We find underbound ground-state energies, as expected from the comparison to the empirical saturation point. The radii are systematically too large, but the agreement with experiment is better. We further explore variations of the 3⁢𝑁 couplings to test their sensitivity in nuclei. While nuclear matter at saturation density is quite sensitive to the 3⁢𝑁 couplings, we find a considerably weaker dependence in medium-mass nuclei. In addition, we explore a consistent momentum-space SRG evolution of these NN and 3⁢𝑁 interactions, exhibiting improved many-body convergence. For the SRG-evolved interactions, the sensitivity to the 3⁢𝑁 couplings is found to be stronger in medium-mass nuclei.

ab initio calculations↗