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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 811 records · Page 45
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Fire-Induced Pressure Response and Failure Characterization of PCV/SCV/3013 Containers - Phase 3

Several Department of Energy (DOE) facilities have materials stored in robust, welded, stainless - steel containers with presumed fire - induced pressure response behaviors. Lack of test data related to fire exposure requires conservative safety analysis assumptions for container response at these facilities. This conservatism can in turn result in the implementation of challenging operational restrictions with costly nuclear safety controls. To help address this issue for sites that store DOE 3013 stainless - steel containers, a series of ten tests were undertaken at Sandia National Laboratories. The goal of this test series was to obtain the response behavior for various configurations of DOE 3013 containers with various payload compositions when exposed to one of two ASTM fire conditions. Key parameters measured in the test series included identification of failure - specific characteristics such as pressure, temperature, and whether or not a vessel was breached during a test . Numerous failure - specific characteristics were identified from the ten tests. This report describes the implementation and execution of the test series performed to identify these failure - specific characteristics. Discussions on the test configurations, payload compositions, thermal insults, and experimental setups are presented. Test results in terms of pressurization and vessel breach (or no - breach) are presented along with corresponding discussions for each test.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Modeling Astrophysical Explosions and the Nucleosynthesis of the Heavy Elements

The heavier elements in the Universe were created and dispersed in the supernova explosions of stars. These events involve a fascinating interplay of nuclear physics, gravity, hydrodynamics, neutrino and photon transport, and radioactive decay, all occurring under extreme conditions. Astronomical experiments provide a rich and constraining data set, but our theoretical understanding of supernovae remains incomplete. The goal of this project is to advance numerical models of astrophysical explosions and use them to connect the nuclear theory of supernovae to current observational and experimental programs. Our contributions focus on one important aspect of the problem --modeling the transport of radiation (both neutrino and electromagnetic) which is essential to understanding the supernova explosion mechanism, the resulting nucleosynthesis, and the observable spectra and light curves. We will exploit recent advances in high performance computing to improve 3-dimensional simulations of core collapse supernovae and neutron star mergers. The benefits include: 1) Further defining the explosive astrophysical environments that govern the nucleosynthesis of r-process nuclei. 2) Preparing the way for more realistic treatments of neutrino interactions, with the ultimate goal of making supernovae a testing ground for probing fundamental neutrino physics, e.g., unknown mixing angles and new aspects of the MSW mechanism. 3) Providing a means for testing explosion simulations and nucleosynthesis calculations against data from optical and x-ray telescopes, thereby permitting the experimental validation or falsification of supernova models. 4) Helping position nuclear physics, through the development of new algorithms of the type incorporated in our simulation codes, to play a leading role in DOE large scale computing.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

SNM Radiation Signature Classification Using Different Semi-Supervised Machine Learning Models

The timely detection of special nuclear material (SNM) transfers between nuclear facilities is an important monitoring objective in nuclear nonproliferation. Persistent monitoring enabled by successful detection and characterization of radiological material movements could greatly enhance the nuclear nonproliferation mission in a range of applications. Supervised machine learning can be used to signal detections when material is present if a model is trained on sufficient volumes of labeled measurements. However, the nuclear monitoring data needed to train robust machine learning models can be costly to label since radiation spectra may require strict scrutiny for characterization. Therefore, this work investigates the application of semi-supervised learning to utilize both labeled and unlabeled data. As a demonstration experiment, radiation measurements from sodium iodide (NaI) detectors are provided by the Multi-Informatics for Nuclear Operating Scenarios (MINOS) venture at Oak Ridge National Laboratory (ORNL) as sample data. Anomalous measurements are identified using a method of statistical hypothesis testing. After background estimation, an energy-dependent spectroscopic analysis is used to characterize an anomaly based on its radiation signatures. In the absence of ground-truth information, a labeling heuristic provides data necessary for training and testing machine learning models. Supervised logistic regression serves as a baseline to compare three semi-supervised machine learning models: co-training, label propagation, and a convolutional neural network (CNN). In each case, the semi-supervised models outperform logistic regression, suggesting that unlabeled data can be valuable when training and demonstrating value in semi-supervised nonproliferation implementations.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2026

Advanced Multiphysics Code Coupling for Cladding Surface Thermocouples During Two-Phase Heat Transfer from Nuclear Fuel

Transient testing of nuclear fuel involves the evaluation of fuel performance under off-normal and accident conditions and is essential for proving design performance. Instrumentation included in such experiments commonly includes thermocouples attached to the outer surface of the cladding to provide temperature measurements throughout the transient. However, the presence of thermocouples on the cladding surface can alter the local heat transfer characteristics with the surrounding coolant. These localized effects can influence the temperature of the nearby cladding surface and introduce uncertainties in interpreting the thermocouple data. Understanding the impact of thermocouples attached to the outer surface of the cladding is crucial for accurate data interpretation as well as its effect on the thermomechanical behavior of the cladding. This paper presents a novel methodology for simulating the impact of outer cladding thermocouples during transient testing of nuclear fuels. The simulation framework leverages the thermal-hydraulic capabilities of RELAP5-3D coupled to the BISON fuel performance code through the RELAPCouplingApp interface. The methodology is compared against Accident Tolerant Fuel Reactivity Initiated Accident-1-E experiment performed at Idaho National Laboratory. The results reveal approximately 100°C difference between thermocouple-altered temperature and virgin cladding surface. The model overpredicts surface rewet time due to conservative correlations.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Data Evaluation of Actinide Cross Sections: 240 Am, 241 Am, and 242 Am

Evaluations for neutron-induced reactions on the isotopes 240 Am, 241 Am, and 242 Am were per formed by optimizing reaction-model calculations based on existing experimental data from both direct measurements and surrogate reactions. The evaluated data were based on results obtained from the LLNL reaction codes YAHFC and FRESCO using up-to-date nuclear models; including an improvement upon some previous treatments for inelastic processes. Nuclear reaction codes generated data for all relevant reaction channels, which were then translated into Generalized Nu clear Data Structure and LLNL’s Evaluated Neutron Data Libraries (ENDL). These data libraries were tested for potential errors and applied to verification and validation (V&V) tests suites where applicable for comparisons.

42 ENGINEERING↗
DOE OSTI2017

PNNL Measurement Results for the 2016 Criticality Accident Dosimetry Exercise at the Nevada National Security Stite (IER-148)

The Pacific Northwest National Laboratory (PNNL) participated in a criticality accident dosimetry intercomparison exercise held at the Nevada National Security Site (NNSS) May 24-27, 2016. The exercise was administered by Lawrence Livermore National Laboratory (LLNL) and consisted of three exposures performed using the Godiva-IV critical assembly housed in the Device Assembly Facility (DAF) located on the NNSS site. The exercise allowed participants to test the ability of their nuclear accident dosimeters to meet the performance criteria in ANSI/HPS N13.3-2013, Dosimetry for Criticality Accidents and to obtain new measurement data for use in revising dose calculation methods and quick sort screening methods where appropriate. PNNL participated with new prototype Personal Nuclear Accident Dosimeter (PNAD) and Fixed Nuclear Accident Dosimeter (FNAD) designs as well as the existing historical PNAD design. The new prototype designs incorporate optically stimulated luminescence (OSL) dosimeters in place of thermoluminescence dosimeters (TLDs), among other design changes, while retaining the same set of activation foils historically used. The default dose calculation methodology established decades ago for use with activation foils in PNNL PNADs and FNADs was used to calculate neutron dose results for both the existing and prototype dosimeters tested in the exercise. The results indicate that the effective cross sections and/or dose conversion factors used historically need to be updated to accurately measure the operational quantities recommended for nuclear accident dosimetry in ANSI/HPS N13.3-2013 and to ensure PNAD and FNAD performance meets the ANSI/HPS N13.3-2013 performance criteria. The operational quantities recommended for nuclear accident dosimetry are personal absorbed dose, Dp(10), and ambient absorbed dose, D*(10).

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2011

Nuclear corrections in {nu}A DIS and their compatibility with global NPDF analyses

We perform a global {chi}{sup 2}-analysis of nuclear parton distribution functions using data from charged current neutrino--nucleus({nu}A) deep inelastic scattering (DIS), charged-lepton--nucleus (l{sup {+-}A}) DIS, and the Drell-Yan (DY) process. We show that the nuclear corrections in {nu}A DIS are not compatible with the predictions derived from l{sup {+-}A} DIS and DY data. We quantify this result using a hypothesis-testing criterion based on the {chi}{sup 2} distribution which we apply to the total {chi}{sup 2} as well as to the {chi}{sup 2} of the individual data sets. We find that it is not possible to accommodate the data from {nu}A and l{sup {+-}A} DIS by an acceptable combined fit. Our result has strong implications for the extraction of both nuclear and proton PDFs using combined neutrino and charged lepton data sets.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2023

Updated evaluation of potential ultralow Q -value β -decay candidates

“Ultralow” Q-value β decays are referred to as such due to their low decay energies of less than ≈1 keV. Such a low energy decay is possible when the parent nucleus decays to an excited state in the daughter, with an energy close to that of the Q value. These decays are of interest as potential new candidates for neutrino mass determination experiments and as a testing ground for studies of atomic interference effects in the nuclear decay process. In this paper, we provide an updated evaluation of atomic mass data and nuclear energy-level data to identify potential ultralow Q-value β decay candidates. For many of these candidates, more precise and accurate atomic mass data is needed to determine if the Q value of the potential ultralow decay branch is energetically allowed and in fact ultralow. Furthermore, the relevant precise atomic mass measurements can be achieved via Penning trap mass spectrometry.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

Concepts for Small-Scale Testing of Used Nuclear Fuel

This report documents a concept for a small-scale test involving between one and three Boiling Water Rector (BWR) high burnup (HBU) fuel assemblies. This test would be similar to the DOE funded High Burn-Up (HBU) Confirmatory Data Project to confirm the behavior of used high burn-up fuel under prototypic conditions, only on a smaller scale. The test concept proposed would collect data from fuel stored under prototypic dry storage conditions to mimic, as closely as possible, the conditions HBU UNF experiences during all stages of dry storage: loading, cask drying, inert gas backfilling, and transfer to an Independent Spent Fuel Storage Installation (ISFSI) for multi-year storage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Code Benchmark of the HTTF Pressurized Conduction Cooldown Test Using SAM

The High Temperature Test Facility (HTTF) at Oregon State University is an integral system test facility to simulate postulated reactor transients of prismatic high-temperature gas-cooled reactors(HTGRs). A series of test campaigns was launched, providing abundant test data that could be used to benchmark reactor system analysis codes like the System Analysis Module (SAM). In this study, a SAM model of the facility is developed based on the two-dimensional (2D) ring model approach. All components including the ceramic matrix, graphite heaters, helium coolant channels, core barrel, upcomer, pressure vessel, and reactor cavity cooling system are modeled as concentric cylindrical rings. The model is used to simulate one of the benchmark problems-Pressurized Conduction Cooldown (PCC)-within the scope of the Organisation for Economic Co-operation and Development Nuclear Energy Agency International HTTF Benchmark. The simulations consist of two parts. In the first part, operating and boundary conditions as well as thermophysical properties of materials are specified for the benchmark problem. In this work, results from the first part will be used in code-to-code comparison. In the second part, the SAM model is used to simulate Test PG-27, which is the first PCC test carried out in the HTTF, with only two of the ten heater banks activated. The results in the second part are used for code-to-data comparison. Because the helium coolant flow rate is not measured in this facility, it is estimated using the input power and inlet/outlet coolant temperatures. Additionally, radial heat flow in the ceramic blocks is complicated by hundreds of cylindrical coolant channels and heater rods embedded in them. As such, it is necessary to deduce an effective thermal conductivity for the ceramic to analyze the core thermal behavior. SAM predictions of the helium coolant and ceramic temperatures are compared with test data measured in three equivalent sectors. Overall, the SAM results agree reasonably well with test data within the variation of data among the three sectors, which demonstrates SAM's capability in capturing transient effects in HTGR using the simplified 2D ring model.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Transient Testing of Nuclear Fuels Performed in the Original Operation of TREAT (2nd ed.)

This report provides summary information on experiments performed in the Transient Reactor Test Facility (TREAT) during its original operation from 1959-1994, and more specifically, before the TREAT core was upgraded at the end of the 1980s. The tests performed since about 1970 are summarized in considerable detail. Earlier tests are noted by their designated series and by some parameters that help to identify and distinguish each series from the others. The detailed summary descriptions are intended to help inform future researchers regarding the large, historical, empirical data base generated from that 35-year span during which hundreds of TREAT experiments were performed to investigate and demonstrate the transient behavior of a wide variety of nuclear fuels during severe off-normal and accident conditions -- using direct nuclear heating of the test samples. That empirical basis is foundational and valuable for guiding future qualification and safety evaluation of improved fuel designs and for future transient test planning and performance in TREAT. Since Edition 1 of this report, descriptions of the following test series have been added: D, E, EOS, F, H, J, R, RX, and S.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

A Methodology for the Development of a Reliability Database for an Advanced Reactor Probabilistic Risk Assessment

GE Hitachi Nuclear Energy (GEH) and Argonne National Laboratory are currently engaged in a joint effort to modernize and develop probabilistic risk assessment (PRA) techniques for advanced non-light water reactors. At a high level the primary outcome of this project will be the development of next-generation PRA methodologies that will enable risk-informed prioritization of safety- and reliability-focused research and development, while also identifying gaps that may be resolved through additional research. A subset of this effort is the development of a reliability database (RDB) methodology to determine applicable reliability data for inclusion in the quantification of the PRA. The RDB method developed during this project seeks to satisfy the requirements of the Data Analysis element of the ASME/ANS Non-LWR PRA standard. The RDB methodology utilizes a relevancy test to examine reliability data and determine whether it is appropriate to include as part of the reliability database for the PRA. The relevancy test compares three component properties to establish the level of similarity to components examined as part of the PRA. These properties include the component function, the component failure modes, and the environment/boundary conditions of the component. The relevancy test is used to gauge the quality of data found in a variety of sources, such as advanced reactor-specific databases, non-advanced reactor nuclear databases, and non-nuclear databases. The RDB also establishes the integration of expert judgment or separate reliability analysis with past reliability data. This paper provides details on the RDB methodology, and includes an example application of the RDB methodology for determining the reliability of the intermediate heat exchanger of a sodium fast reactor. The example explores a variety of reliability data sources, and assesses their applicability for the PRA of interest through the use of the relevancy test.

Grabaskas, Dave↗
DOE OSTI2025

Corrosion analysis of Al-Ce-Ni and Al-Cu-Ce cast alloys in dilute boric acid at room and elevated temperatures

Two Al-Ce-Ni cast alloys, and an Al-Cu-Ce cast alloy, as candidate wet storage materials for spent nuclear fuels, were tested in 0.23 wt% H 3 BO 3 solution to assess the alloy corrosion resistance. Electrochemical and gravimetric corrosion data suggest that the three Al alloys are unlikely to undergo any severe corrosion in dilute H 3 BO 3 at or below 50 °C. Post-exposure characterization of the three Al alloys, including scanning and transmission electron microscopy and electron dispersive spectroscopy, revealed a corrosion product layer, mostly Al 2 O 3 , on the exposed surface and local penetration of oxygen into the alloy matrix. The degree of oxide layer growth and oxygen penetration is greater at 80 °C than the temperature at/below 50 °C. The Al-Cu-Ce alloy is considered less corrosion resistant than the other two alloys studied.

Al-Ce alloy↗
DOE OSTI2021

The effects of microstructures and radiation damage on the deformation behavior of a HT-9 alloy using microtensile testing

Understanding the influence of radiation damage on the mechanical properties of HT-9 and other tempered martensitic alloys is part of the mission in developing radiation-tolerant materials for the next generation of nuclear reactors. Although there has been extensive data on the macroscopic mechanical properties of the irradiated HT-9 alloys and microstructural changes, it is not well-understood how radiation damage and the resulting microstructural changes influence the local mechanical properties. In this study, we utilized in situ SEM microtensile testing to investigate the deformation behavior of specific martensitic boundaries in 1 dpa proton-irradiated HT-9 and answered the question of what is the weakest link. Additionally, we provided the direct observation of the failure modes of microtensiles containing high angle and low angle martensitic boundaries. In the unirradiated condition, the deformation is ductile and no martensitic boundary failure is observed. In the irradiated condition, the high angle martensitic boundaries are more susceptible to radiation-induced boundary failure as compared to the low angle martensitic boundaries.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

Demonstration of INCC6 for advanced list-mode data acquisition and analysis using ALMM

The International Neutron Coincidence Counting (INCC) software plays an important role in nuclear safeguards and nuclear material control and accounting (NMC&A) measurements. While the current version, INCC5, represents the traditional standard utilized by inspectorates as well as facilities and practitioners alongside shift register hardware, LANL has recently developed an upgraded version of the software, INCC6. INCC6 offers the same analysis tools as INCC5 while adding new capabilities including acquisition and analysis of list-mode data, recording and storage of list-mode data files, and an expanded set of advanced analysis tools that make use of the additional information available from list-mode data. Here, this paper presents the first demonstration of data acquisition and analysis using INCC6 and the Advanced List Mode Multiplicity Module (ALMM). The capability of INCC6 to perform live list-mode data acquisition and analysis is demonstrated for a variety of neutron sources with a range of neutron emission rates relevant for practical applications and validated against the results obtained with INCC5 and traditional shift register hardware. Advanced list-mode analysis tools introduced by INCC6, such as time-interval and coincidence matrix analysis, are also demonstrated. Finally, the capability of INCC6 to analyze list-mode data files is tested and validated against traditional shift register results using INCC5.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2011

Evaluation of HEU-Beryllium Benchmark Experiments to Improve Computational Analysis of Space Reactors

An assessment was previously performed to evaluate modeling capabilities and quantify preliminary biases and uncertainties associated with the modeling methods and data utilized in designing a nuclear reactor such as a beryllium-reflected, highly-enriched-uranium (HEU)-O2 fission surface power (FSP) system for space nuclear power. The conclusion of the previous study was that current capabilities could preclude the necessity of a cold critical test of the FSP; however, additional testing would reduce uncertainties in the beryllium and uranium cross-section data and the overall uncertainty in the computational models. A series of critical experiments using HEU metal were performed in the 1960s and 1970s in support of criticality safety operations at the Y-12 Plant. Of the hundreds of experiments, three were identified as fast-fission configurations reflected by beryllium metal. These experiments have been evaluated as benchmarks for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (IHECSBE). Further evaluation of the benchmark experiments was performed using the sensitivity and uncertainty analysis capabilities of SCALE 6. The data adjustment methods of SCALE 6 have been employed in the validation of an example FSP design model to reduce the uncertainty due to the beryllium cross section data.

Bess, John↗
DOE OSTI2023

High Energy Arcing Fault (HEAF) Photometrics 2022 Test Report

High Energy Arcing Faults (HEAFs) are hazardous events in which an electrical arc leads to the rapid release of energy in the form of heat, vaporized metal, and mechanical force. In Nuclear Power Plants, these events are often accompanied by loss of essential power and complicated shutdowns. To confirm the probabilistic risk analysis (PRA) methodology in NUREG/CR-6850, which was formulated based on limited observational data, the NRC led an international experimental campaign from 2014 to 2016. The results of these experiments uncovered an unexpected hazard posed by aluminum components in or near electrical equipment and the potential for unanalyzed equipment failures. Sandia National Laboratories (SNL), in support of the NRC work, collaborated with NIST, BSI, KEMA, and NRC to support the full-scale HEAF test campaign in 2022. SNL provided high speed visible and infrared video/data of ten tests that collected data from HEAFs originated on copper and aluminum buses inside switchgears and bus ducts. Part of the SNL scope was to place cameras with high-speed data collection at different vantage points within the test facility to provide NRC a more complete and granular view of the test events.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2011

REMORA 3: The first instrumented fuel experiment with on-line gas composition measurement by acoustic sensor

With the aim to improve the knowledge of nuclear fuel behaviour, the development of advanced instrumentation used during in-pile experiments in Material Testing Reactor (MTR) is necessary. To obtain data on high Burn-Up MOX fuel performance under transient operating conditions, especially in order to differentiate between the kinetics of fission gas and helium releases and to acquire data on the degradation of the fuel conductivity, a highly instrumented in-pile experiment called REMORA 3 has been conducted by CEA and IES (Southern Electronic Inst. - CNRS - Montpellier 2 Univ.). A rodlet extracted from a fuel rod base irradiated for five cycles in a French EDF commercial PWR has been re-instrumented with a fuel centerline thermocouple, a pressure transducer and an advanced acoustic sensor. This latter, patented by CEA and IES, is 1 used in addition to pressure measurement to determine the composition of the gases located in the free volume and the molar fractions of fission gas and helium. This instrumented fuel rodlet has been re-irradiated in a specific rig, GRIFFONOS, located in the periphery of the OSIRIS experimental reactor core at CEA Saclay. First of all, an important design stage and test phases have been performed before the irradiation in order to optimize the response and the accuracy of the sensors: - To control the influence of the temperature on the acoustic sensor behaviour, a thermal mock-up has been built. - To determine the temperature of the gas located in the acoustic cavity as a function of the coolant temperature, and the average temperature of the gases located in the rodlet free volume as a function of the linear heat rate, thermal calculations have been achieved. The former temperature is necessary to calculate the molar fractions of the gases and the latter is used to calculate the total amount of released gas from the internal rod pressure measurements. - At the end of the instrumented rod manufacturing, specific internal free volume and pressure measurements have been carried out. Preliminary calculations of the REMORA 3 experiments have been performed from these measurements, with the aim to determine free volume evolution as a function of linear heat rate history. - A tracer gas has been added to the filling gas in order to optimize the accuracy of the helium balance at the time of the post irradiation examination. The two phases of the REMORA 3 irradiation have been achieved at the end of 2010 in the OSIRIS reactor. Slight acoustic signal degradation, observed during the test under high neutron and gamma flux, has led to an efficiency optimization of the signal processing. The instrumentation ran smoothly and allowed to reach all the experimental objectives. After non destructive examination performed in the Osiris reactor pool, typically gamma spectrometry and neutron radiography, the instrumented rod and the device have been disassembled. Then the instrumented rod has been transported to the LECA facility in Cadarache Centre for post irradiation examination. The internal pressure and volume of the rodlet as well as precise gas composition measurements will be known after puncturing step performed in a hot cell of this facility. That will allow us to qualify the in-pile measurements and to finalize the data which will be used for the validation of the fuel behaviour computer codes. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗