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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 1,099 records · Page 61
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Enhanced Component Performance Study: Emergency Diesel Generators 1998–2018

This report presents an enhanced performance evaluation of emergency diesel generators (EDGs) at U.S. commercial nuclear power plants. This report evaluates component performance over time using (1) Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES) data from 1998 through 2018 and (2) maintenance unavailability (UA) performance data from Mitigating Systems Performance Index (MSPI) Basis Document data from 2002 through 2018. The objective is to show estimates of current failure probabilities and rates related to EDGs, trend these data on an annual basis, determine if the current data are consistent with the probability distributions currently recommended for use in NRC probabilistic risk assessments, show how the reliability data differ for different EDG manufacturers and for EDGs with different ratings; and summarize the subcomponents, causes, detection methods, and recovery associated with each EDG failure mode. Engineering analyses were performed with respect to time period and failure mode without regard to the actual number of EDGs at each plant. The factors analyzed are: sub-component, failure cause, detection method, recovery, manufacturer, and EDG rating. A statistically significant increasing trend was identified in the frequency of FTLR demands for emergency power system (EPS) and high pressure core spray (HPCS) EDGs and a statistically significant decreasing trend was identified in the frequency of run > 1H hours for EPS and HPCS EDGs.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2016

Spent nuclear fuel dynamic performance under normal conditions of transport

The objective of this project is to evaluate the spent nuclear fuel (SNF) vibration integrity under normal condition of transport. The SNF fatigue strength data has been collected from various pressurized water reactors. The CIRFT (Cyclic Integrated Reversible-Bending Fatigue Tester) has been used to simulate transportation environments. CIRFT is an enabling hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL). The CIRFT consists mainly of a U-frame testing setup and a real-time curvature measurement method. The three-component U-frame setup of the CIRFT has two rigid arms and linkages to a universal testing machine. The curvature of rod bending is obtained through a three-point deflection measurement method. Three linear variable differential transformers (LVDTs) are used and clamped to the side connecting plates of the U-frame to capture the deformation of the rod. CIRFT results have provided insight into the fuel/clad system response to transportation related loads. Because of the non-homogeneous composite structure of the SNF system, finite element analyses (FEA) are needed to translate the global moment-curvature measurement into local stress-strain profiles. The detailed mechanisms of the pellet-pellet and pellet-clad interactions and the stress concentration effects at the pellet-pellet interface cannot be readily obtained directly from a CIRFT system measurement. Therefore, detailed FEA is used to understand the global test response. The major findings of CIRFT test on the high burn-up SNF are as follows: -) SNF system interface bonding plays an important role in SNF vibration performance; -) Fuel structure contributes to the SNF system stiffness; -) There are significant variations in stress and curvature of SNF systems during vibration cycles resulting from segment pellets and clad interaction, and -) SNF failure initiates at the pellet-pellet interface region and appears to be spontaneous.

42 ENGINEERING↗
DOE OSTI2015

Cold-cap temperature profile comparison between the laboratory and mathematical model - 15591

The rate of waste vitrification in an electric melter is connected to the feed-to-glass conversion process, which occurs in the cold cap, a layer of reacting feed on top of molten glass. The cold cap consists of two layers: a low-temperature (∼100 deg. C - ∼800 deg. C) region of unconnected, reacting feed and a high-temperature (∼800 deg. C - ∼1100 deg. C) region of molten glass with gas bubbles and cavities. A recently developed mathematical model describes the effect of the cold cap on glass production. For verification of the mathematical model, a cold cap produced in a laboratory-scale melter was subsequently analyzed by correlating structural features of a polished region with those obtained in heat-treated feed samples to determine the temperature profile as a function of cold-cap position. After corrections were made for the diminished volume of bubbles and collapsed cavities that occurred in the cold-cap sample during cooling, the temperature gradients were alike in the model and the laboratory cold cap through the range from ∼500 deg. C to ∼1100 deg. C. The temperature profiles were different in the upper region of the cold cap because the radiant heat from the molten glass in the laboratory-scale melter resulted in dry and hot (∼400 deg. C) surface conditions whereas the model assumed a surface covered by boiling slurry. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Advanced Computational Accident Scenario Modeling for Safety Margin Economic Analysis

The intent of this research is to establish a new tool in the Risk Informed Safety Margin Characterization (RISMC) tool-kit for the fiscal analysis of potential power plant upgrades. This is being done by providing proof-of-concept of the Safety Margin Economic Evaluation (SMEE) methodology of performing cost-benefit analyses of potential Nuclear Power Plant (NPP) upgrades to allow for more data-driven decision making with regards to nuclear safety. To this end, the RAVEN code developed by the Idaho National Laboratory is being used to perform a cost-benefit analysis of several potential power plant safety upgrades to compare the material cost of the upgrades with the probabilistic benefit of avoiding accident scenario consequences to find the most cost effective safety upgrades. Additionally, the MELCOR and MACCS2 codes developed by the Sandia National Laboratory are being used for accident scenario modeling. Evaluating potential NPP upgrades is concluded by doing a detailed cost-benefit analysis of the upgrades. The cost of the upgrade is the direct material cost of implementing, maintaining, and operating the upgrade at an NPP for the projected duration of its lifetime. The benefit of the upgrade is more complex, and is essentially the Risk avoided, where Risk is defined as Consequences multiplied by Probability. Here, Consequences is the economic impact of a severe accident and/or nuclear release and Probability is the likelihood that an NPP upgrade will prevent or mitigate a severe accident, reducing or even eliminating its Consequences. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Benchmark Progress towards TREAT M8CAL Core Loading to Support Validation of TREAT Operations

Benchmark experiment data is necessary to validate modeling and simulation activities to support Transient Reactor Experiment Test (TREAT) Facility restart and operations. Key measurements of interest include criticality, control rod worth, excess reactivity, and shutdown margin for varying core loading sizes. Benchmark evaluations are being developed according the guidelines provided in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook) that can also support advanced modeling and simulation activities, new experiment design for advanced reactors and accident tolerant fuel, and potential low-enriched uranium conversion of the TREAT reactor. This summary discusses development activities supporting a benchmark evaluation of the TREAT M8CAL core loading, which will be submitted for peer-review and publication in the IRPhEP Handbook with the identifier TREAT-FUND-RESR-002. Detailed models of TREAT were developed and evaluated to prepare benchmark models of the minimum critical mass configuration. Calculations were performed using MCNP6.1 with ENDF/B-VII.1 nuclear data libraries. Discussion of results.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Inspection Technologies for Waste Containers at Canadian Nuclear Laboratories - 19429

To provide data for aging management programs, containers for different forms of nuclear waste, ranging from low-level waste to irradiated nuclear fuel, should be inspected. At Canadian Nuclear Laboratories, visual, ultrasonic, electromagnetic, and hybrid forms of non-destructive evaluation (NDE) are used for in situ inspection of carbon steel containers having various configurations, access issues, and material thickness. NDE applications tailored to the requirements of the task provide data for condition monitoring and operational assessments. The examples of fuel baskets stacked in concrete canisters and sheet metal containers arrayed in warehouses are illustrated here. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Nuclear Criticality Safety Margin and Handbook Data: Concepts and Applications [Slides]

At the end of this briefing, personnel should be able to: • Define concepts associated with criticality safety margin • Provide and understand practical applications that illustrate safety margin concepts • Describe how safety margin is addressed in criticality safety evaluations • Improve documentation and communication of safety margin • Connect these concepts to related statements in the ANS-8 Standards • Find numerous applications for handbook data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Benchmark Development for TREAT Minimum Critical Mass Core Loading to Support Validation of TREAT Operations

Benchmark experiment data is necessary to validate modeling and simulation activities to support Transient Reactor Experiment Test (TREAT) Facility restart and operations. Key measurements of interest include criticality, control rod worth, excess reactivity, and shutdown margin for varying core loading sizes. Benchmark evaluations are being developed according the guidelines provided in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook) that can also support advanced modeling and simulation activities, new experiment design for advanced reactors and accident tolerant fuel, and potential low-enriched uranium conversion of the TREAT reactor. This summary discusses development activities supporting a benchmark evaluation of the TREAT minimum critical mass core loading, which will be submitted for peer-review and publication in the IRPhEP Handbook with the identifier TREAT-FUND-RESR-001. Detailed models of TREAT were developed and evaluated to prepare benchmark models of the minimum critical mass configuration. Calculations were performed using MCNP6.1 with ENDF/B-VII.1 nuclear data libraries. Discussion of results.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Benefits of utilizing CellProfiler as a characterization tool for U–10Mo nuclear fuel

Automated image processing techniques have the potential to aid in the performance evaluation of nuclear fuels by eliminating judgment calls that may vary from person-to-person or sample-to-sample. Analysis of in-core fuel performance is required for design and safety evaluations related to almost every aspect of the nuclear fuel cycle. This study presents a methodology for assessing the quality of uranium–molybdenum fuel images and describes image analysis routines designed for the characterization of several important microstructural properties. The analyses are performed in CellProfiler, an open-source program designed to enable biologists without training in computer vision or programming to automatically extract cellular measurements from large image sets. The quality metric scores an image based on three parameters: the illumination gradient across the image, the overall focus of the image, and the fraction of the image that contains scratches. The metric presents the user with the ability to ‘pass’ or ‘fail’ an image based on a reproducible quality score. Passable images may then be characterized through a separate CellProfiler pipeline, which enlists a variety of common image analysis techniques. The results demonstrate the ability to reliably pass or fail images based on the illumination, focus, and scratch fraction of the image, followed by automatic extraction of morphological data with respect to fission gas voids, interaction layers, and grain boundaries. - Graphical abstract: Display Omitted - Highlights: • A technique is developed to score U–10Mo FIB-SEM image quality using CellProfiler. • The pass/fail metric is based on image illumination, focus, and area scratched. • Automated image analysis is performed in pipeline fashion to characterize images. • Fission gas void, interaction layer, and grain boundary coverage data is extracted. • Preliminary characterization results demonstrate consistency of the algorithm.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Rancor-HUNTER: A Virtual Plant and Operator Environment for Predicting Human Performance

Advances in simulation capabilities to model physical systems have outpaced the development of simulations for humans using those physical systems. There is an argument that the infinite span of potential human behaviors inherently render human modeling more challenging than physical systems. Despite this challenge, the need for modeling humans interacting with these complex systems is paramount. As technologies have improved, many of the failure modes originating from the physical systems have been solved. This means the overall proportion of human errors has increased, such that it is not uncommon to be the primary driver of system failure in modern complex systems. Moreover, technologies such as automated systems may introduce emerging contexts that can cause new, unanticipated modes of human error. Therefore, it is now more important than ever to develop models of human behavior to realize overall system error reductions and achieve established safety margins. To support new and novel concepts of operations for the anticipated wave of advanced nuclear reactor deployments, human factors and human reliability analysis researchers need to develop advanced simulation-based approaches. This talk presents a simulation environment suitable to both collect data and then perform Monte Carlo simulations to evaluate human performance and develop better models of human behavior. Specifically, the Rancor Microworld Simulator models a complex energy production system in a simplified manner. Rancor includes computer-based procedures, which serve as a framework to automatically classify human behaviors without manual, subjective experimenter coding during scenarios. This method supports a detailed level of analysis at the task level. It is feasible for collecting large sample sizes required to develop quantitative modelling elements that have historically challenged traditional full-scope simulator study approaches. Additionally, the other portion of this experimental platform, the Human Unimodel for Nuclear Technology to Enhance Reliability (HUNTER), is presented to show how the collected data can be used to evaluate novel scenarios based on the contextual factors, or performance shaping factors, derived from Rancor simulations. Rancor-HUNTER is being used to predict operator performance with new procedures, such as results from control room modernization or new-build situations. Rancor-HUNTER is also proving a useful surrogate platform to model human performance for other complex systems.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2018

Template for estimating uncertainties of measured neutron-induced fission cross-sections

A template for estimating uncertainties (unc.) of measured neutron-induced fission, (n,f), cross-sections (cs) is presented. This preliminary template not only lists all expected unc. sources but also supplies ranges of unc., estimates for correlations between unc. of the same and different experiments which can be used if the information is nonexistent. If this template is applied systematically when estimating experimental covariances for an evaluation, it may help in pinpointing missing unc. for individual datasets, identifying unreasonably low unc., and estimating correlations between different experimental datasets. Thus, a detailed unc. estimate – usually, a time-intensive procedure – can be undertaken more consistently and efficiently. As an example, it is shown that unc. and correlations of 239 Pu(n,f) by Merla et al. [Proceedings of the Conference on Nuclear Data for Science and Technology 1991 Jülich (Springer-Verlag, Berlin, 1992) pp. 510–513], which are questionably low in the GMA database underlying the neutron cs standards evaluations, are distinctly larger at 14.7 MeV and more strongly correlated if this template is used for reestimating the associated covariances.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Neutron Activation Foil and Thermoluminescent Dosimeter Responses to a Polyethylene Reflected Pulse of the CEA Valduc SILENE Critical Assembly

This benchmark experiment was conducted as a joint venture between the US Department of Energy (DOE) and the French Commissariat à l'Energie Atomique (CEA). Staff at the Oak Ridge National Laboratory (ORNL) in the US and the Centre de Valduc in France planned this experiment. The experiment was conducted on October 19, 2010 in the SILENE critical assembly facility at Valduc. Several other organizations contributed to this experiment and the subsequent evaluation, including CEA Saclay, Lawrence Livermore National Laboratory (LLNL), the Y-12 National Security Complex (NSC), Babcock International Group in the United Kingdom, and Los Alamos National Laboratory (LANL). The goal of this experiment was to measure neutron activation and thermoluminescent dosimeter (TLD) doses from a source similar to a fissile solution critical excursion. The resulting benchmark can be used for validation of computer codes and nuclear data libraries as required when performing analysis of criticality accident alarm systems (CAASs). A secondary goal of this experiment was to qualitatively test performance of two CAAS detectors similar to those currently and formerly in use in some US DOE facilities. The detectors tested were the CIDAS MkX and the Rocky Flats NCD-91. The CIDAS detects gammas with a Geiger-Muller tube and the Rocky Flats detects neutrons via charged particles produced in a thin 6 LiF disc depositing energy in a Si solid state detector. These detectors were being evaluated to determine whether they would alarm, so they were not expected to generate benchmark quality data.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

Data Evaluation of Actinide Cross Sections: 230 Th and 231 Th

A new evaluation of the ENDL cross section set for Thorium (Z=90) is developed using the TALYS statistical model cross section code. The primary goal of this effort is to produce an evaluation that attempts to match as closely as possible fission cross sections developed through surrogate reaction techniques on actinide targets 230 Th and 231 Th. This evaluation effort and the processing needed to render its results into data libraries is a necessary step in making the efforts of nuclear experimentalists useful to the broad community of researchers engaged in simulations of nuclear fusion for basic and applied science. Another aspect, verification and validation against various AGEX experiments, is also presented. The end-product is an updated library that includes the latest measurements of fission cross sections for comparison against those measured via traditional techniques. All the steps in the evaluation, processing, validation and verification, and library release are described in the following sections. For completeness, the appendix contains all the parameters used in the TALYS cross section evaluation for neutrons incident on 230 Th and 231 Th.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Nuclear Data Sheets for A = 139

The experimental nuclear spectroscopic data for known nuclides of mass number 139 (Sn, Sb, Te, I, Xe, Cs, Ba, La, Ce, Pr, Nd, Pm, Sm, Eu, Gd, Tb, Dy) have been evaluated and presented together with adopted properties for levels and gamma rays. This evaluation represents a revision of previous one (2001Bu16) by T.W. Burrows. Since then extensive new data have become available for many A=139 nuclides although, no excited state data are yet available for {sup 139}Sn, {sup 139}Sb, {sup 139}Tb, and {sup 139}Dy. The decay schemes of {sup 139}Sn, {sup 139}Sb, {sup 139}I, {sup 139}Tb and {sup 139}Dy nuclei are not known, and those for {sup 139}Gd and {sup 139}Eu are poorly known. Particle-transfer data are available for {sup 139}Ba, {sup 139}La, {sup 139}Ce and {sup 139}Pr. Limited high-spin data are available for {sup 139}Te, {sup 139}I, {sup 139}Cs, {sup 139}Ba, {sup 139}La and {sup 139}Ce, while such data are extensive for {sup 139}Pr, {sup 139}Nd, {sup 139}Pm, {sup 139}Sm, {sup 139}Eu and {sup 139}Gd. Recent {sup 139}Pm high-spin data from two independent studies using the same reaction and similar beam energy are in conflict and could be improved using state-of-the-art large gamma-detector arrays in contrast to the smaller arrays currently used. Cases for which no new experimental information are available since the last update have undergone checking resulting in some changes in the current work, but for the most part these are taken from the previous evaluations. Thus, the present work has greatly benefited from all prior NDS evaluations (2001Bu16, 1989Bu12, 1981Pe04, 1974Gr46), but at the same time data presented herein supersede those in the earlier evaluations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Advancements in Validation of TSLs through Inelastic Neutron Scattering and Transmission Measurements [Abstract]

Historically, the free gas approximation has been used to treat the thermal scattering of neutrons with energies below a few electron-volts (eV) in unevaluated materials. However, this method inadequately reproduces neutron scattering at these energies. Until recently, only a limited number of materials had available thermal scattering law (TSL) files/libraries in the ENDF nuclear data libraries in this energy range. With advancements in atomistic modeling techniques, such as molecular dynamics, ab-initio molecular dynamics, and density functional theory, TSL libraries have become available for many more materials. This is particularly relevant due to the rising interest in several advanced reactor systems that require novel moderator and reflector materials. While quasi-integral and integral benchmarks have been designed to validate historically important moderator materials (such as light water and polyethylene), there is currently a lack of standard validation methods for TSLs, especially when multiple conflicting TSLs exist. To address this issue, the Oak Ridge National Lab Nuclear Data group has been working on utilizing inelastic neutron scattering (INS) measurements combined with transmission (i.e., total cross section) measurements to evaluate and validate TSLs for different materials. We plan to demonstrate how this method has worked on materials such as polyethylene, lucite, and polystyrene. In addition, we will compare the newly created libraries to ENDF libraries for these materials and explain why integral benchmarks should not be used for validation when multiple TSLs exist.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Artificial Intelligence for Multiphysics Nuclear Design Optimization with Additive Manufacturing

The geometric flexibility of additively manufactured metals and ceramics generates a very large and open design space that requires advanced modeling and simulation tools for physics simulations and the rigorous definition of design problems. This effort deploys artificial intelligence (AI) and machine learning (ML) algorithms to understand the design space, evaluate potential designs, and more efficiently generate optimized results. The Transformational Challenge Reactor (TCR) program is leveraging advances in several scientific areas—including materials, manufacturing, sensors and control systems, data analytics, and high-fidelity modeling and simulation—to accelerate the design, manufacturing, qualification, and deployment of advanced nuclear energy systems. Through a manufacturing-informed design approach, the TCR program seeks to integrate digital data for rapid nuclear innovation; accelerate the adoption of advances in manufacturing, materials, and computational sciences for nuclear applications; and dramatically reduce deployment costs and timelines for new nuclear reactor technologies. This report documents efforts under the TCR program to leverage advanced modeling and simulation techniques driven by AI/ML algorithms on high-performance computing (HPC) systems to yield more optimized TCR core designs. A multiphysics ML surrogate model was developed to run on the HPC architectures. The surrogate model is trained on high-fidelity simulation data of coupled neutronics and thermofluidics and is used to quickly evaluate thousands of candidate core designs in parallel, which drives the evolution of the cooling channel shapes to minimize temperature peaking and material stress. Outcomes from these activities provide design information and feedback into the core design efforts.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Study of a Proposed fissionTPC Measurement of 233 U(n,f)/ 235 U(n,f) Cross Section Ratio

A review of the available 233 U( n, f )/ 235 U( n, f ) cross section ratio data compared to the ENDF evaluation in the fast neutron energy range, 0.1–20 MeV, shows that discrepancies exist between some of the various measurements and the evaluation. The discrepancies vary and a function of incident neutron energy and range between 1–5%. In this document we review two potential options for performing new measurements of the cross section ratio to potentially resolve these discrepancies. The two options are compared and contrasted in terms of measurement accuracy and cost. The first option would utilize the fissionTPC, which was designed specifically to achieve measurement uncertainties of better than 1% to resolve a similar data–evaluation discrepancy in the 239 Pu( n, f ) cross section. While this would seem to be an ideal option in this case also, there are challenges. The fissionTPC is a complex instrument and the cost of maintenance and operations are significant. Furthermore, some of the fissionTPC equipment is nearing the end of its useful life which presents a risk to completion of a measurement campaign. The second option is to use a dual-fission chamber, which is the more traditional approach to these types of measurements. The cost is estimated to be about half that of a fissionTPC measurement, while the much simpler design of the detector reduces overall risk. The trade-off is the measurement uncertainty, which would be on order of 1–3% with a careful measurement and analysis. This uncertainty however may not be good enough to fully resolve the discrepancies. The objective of a new measurement would be the improvement in the ENDF evaluation with the end goal of improving the accuracy of models and simulations at the end of the nuclear data pipeline. An understanding of the sensitivity of these models to the data uncertainties would help guide the choice of measurement techniques.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2014

Evaluation of Neutron Radiography Reactor LEU-Core Start-Up Measurements

Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. Experiments include criticality, control-rod worth measurements, shutdown margin, and excess reactivity for four core loadings with 56, 60, 62, and 64 fuel elements. The worth of four graphite reflector block assemblies and an empty dry tube used for experiment irradiations were also measured and evaluated for the 60-fuel-element core configuration. Dominant uncertainties in the experimental k eff come from uncertainties in the manganese content and impurities in the stainless steel fuel cladding as well as the 236 U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 neutron nuclear data are approximately 1.4% (9σ) greater than the benchmark model eigenvalues, which is commonly seen in Monte Carlo simulations of other TRIGA reactors. Simulations of the worth measurements are within the 2σ uncertainty for most of the benchmark experiment worth values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗