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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 1,225 records · Page 68
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Retractable Sensor Poster for ANIMMA 2023 Conference

Material Test Reactors (MTRs) such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) are used to irradiate nuclear fuels and materials to evaluate their performance after high levels of exposure to a reactor in-pile environment. The test environment within these reactors are extremely harsh and long-term exposure causes sensor decalibration or failure. Most tests only require data points at daily intervals meaning that the sensor does not necessarily need to be in place continuously for the full experiment. One proposed solution to the issue described above is to have a sensor which can be periodically inserted through a guide tube into the test region and retracted; this would act as a reference check on the sensors installed permanently. If a compact and robust enough design can be made, it has the potential to extend the life of the sensor and keep it within calibration. This LRS entry is for a poster summarizing a presentation that will be given at the ANIMMA 2023 conference. The poster will be on display in the Measurement Sciences Laboratory booth at ANIMMA. Most of the material on this poster came from LRS entry INL/CON-23-72797.

42 ENGINEERING↗
DOE OSTI2020

Data Exploration of UAV Flights Recorded on Smartphones

The increasing accessibility and technological advancements of unmanned aerial vehicles (UAVs) pose legitimate concerns to nuclear non-proliferation efforts. To deter and combat unwanted or nefarious UAV activity near or around nuclear sensitive sites, the ability to quickly and easily detect drones becomes vital. Our team is exploring the use of ubiquitous smartphones as multi-modal sensors (microphone, accelerometer, gyroscope, etc.) to meet this challenge. In conjunction with Idaho National Lab, two UAV flight tests and one stationary in-air test were conducted and recorded with eight Samsung Galaxy S8 smartphones. This work presents initial characterization and annotation of acoustic signals associated with UAV activity.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗
DOE OSTI2021

FY2021 ILAW Glass Ion-Exchange Rate Testing

Approximately 54 to 56 million gallons of radioactive mixed waste is currently stored in underground tanks at The United States Department of Energy’s (DOE’s) Hanford site in the State of Washington. The Hanford Tank Waste Treatment and Immobilization Plant (WTP) will provide DOE’s Office of River Protection (ORP) with a means of treating this waste by vitrification for subsequent disposal. The tank waste will be separated into low- and high-activity waste fractions, which will then be vitrified respectively into Immobilized Low Activity Waste (ILAW) and Immobilized High-Level Waste (IHLW) products. The ILAW product will be disposed of in an engineered facility – the Integrated Disposal Facility (IDF) – on the Hanford site, while the IHLW product will be directed to the national deep geological disposal facility for high-level nuclear waste. The ILAW and IHLW products must meet a variety of requirements with respect to protection of the environment before they can be accepted for disposal. To capitalize on the success of the FY19 and FY20 Atkins/VSL test results on six glasses, Washington River Protection Solutions, LLC (WRPS) has contracted with Atkins/VSL to collect IEX data on four more ILAW glasses using PFT, the results of which are the subject of the present report. The work described herein was performed according to a Test Plan that is responsive to the corresponding WRPS scope of work.

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

Sister Rod Destructive Examinations (FY23) Appendix D: Fission Gas, Fuel Burnup, and Fuel Isotopic Measurements

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. The DEs are being conducted to obtain a baseline of the HBU rod’s condition before dry storage and are focused on understanding overall SNF rod strength and durability. Composite fuel and defueled cladding will be tested to derive material properties. Although the data generated can be used for multiple purposes, one primary goal for obtaining the post-irradiation examination data and the associated measured mechanical properties is to support SNF dry storage licensing and relicensing activities by (1) addressing identified knowledge gaps and (2) enhancing the technical basis for post-storage transportation, handling, and subsequent disposition. This appendix documents the status of the ORNL Phase 1 DE activities related to fission gas sample analysis, fuel burnup analysis, and fuel isotopic analysis of selected sister specimens in Phase 1 of the sister rod test program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Machine learning applied to classifying neutron resonances

The performance of nuclear reactors and other nuclear systems depends on a precise understanding of the neutron interaction cross sections for materials used in these systems. These cross sections exhibit resonance structure whose shape is determined in part by the angular momentum quantum numbers of the resonances. The correct assignment of the quantum numbers of neutron resonances is therefore of paramount importance. In this project, we apply a machine learning technique, namely decision trees, to automate the quantum number assignments. The tree is trained from simulated data generated to mimic the errors found in real data. We explore the use of several physics-motivated features for training our trees, including the nearest neighbor spacing distribution, cumulative level distribution, and channel width distributions. Initial results using random matrix theory motivated fits which demonstrated that we can determine resonance spin groups somewhat reliably. If we use these fits as features in our trees, we can train them to spot outliers corresponding to misassigned resonances. We found that with the large number of features used in this project that the decision tree tended to over t training data resulting in poor performance with respect to the test data. By reducing the number of features, we can achieve nearly perfect assignment of quantum numbers with our training data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2026

SMR safety through HTTF modeling and benchmark efforts for code validation for gas-cooled reactor applications

Accurate modeling and simulation tools for thermal-hydraulics calculations are a key element needed to design and license new advanced reactors including Small Modular Reactors (SMR) and Microreactors. Uncertainties in modeling and simulation can have significant safety and economic implications. The High Temperature Test Facility (HTTF) at Oregon State University (OSU) is a scaled integral effects experiment designed to investigate transient behavior in high-temperature gas-cooled prismatic-block nuclear reactors. High-quality measurement data is available from the HTTF that is suitable for a thermal-hydraulics code validation benchmark for gas-cooled reactor simulations. Here, this paper summarizes individual HTTF modeling efforts to date for tool validation at Idaho National Laboratory (INL), Argonne National Laboratory (ANL), Oregon State University (OSU) and Canadian Nuclear Laboratories (CNL) using system thermal-hydraulics codes, Computational Fluid Dynamics (CFD) codes and system-CFD code couplings. Also, the paper introduces the ongoing OECD Nuclear Energy Agency (NEA) High Temperature Gas Reactor Thermal-Hydraulics (HTGR T/H) benchmark that allows for better comparisons of results between different international modeling teams. The benchmark provides well defined computational problems that include code-to-code comparisons and comparisons to measured data. These problems provide an avenue for quantifying accuracy and identifying sources of uncertainty in thermal-hydraulics calculations, including in measured thermophysical properties, as part of validation for gas-cooled reactor simulation tools.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2014

SESAME 96171, a three-phase equation of state for CeO 2

This report describes an earlier version of an equation of state (EOS) for cerium (IV) oxide, CeO 2 . This work has been superseded by a newer version that is described in LA-UR-14-24689. Everything from this report needed to understand the new version is included in the new report. I have constructed SESAME 96171, an EOS for cerium (IV) oxide that includes two solid phases and the liquid. Unlike its immediate predecessor, this EOS is valid over a larger range in pressure due to including the extra phase. I have described the process of constructing and testing the EOS, emphasizing its continuity with the previous EOS and the modifications needed for the new frame-work. The comparison with data is good, but is still limited by the same issues with nuclear models that were present in the previous work. The remaining significant issue is the comparison of the crystal-density EOS with experimental porous Hugoniot data, which may require using more sophisticated model to represent porosity, as discussed in the previous report.

36 MATERIALS SCIENCE↗
DOE OSTI2013

The Use of Geomorphometrics for Geologic Site Characterization

The Problem: Developing an accurate geologic site characterization is difficult for denied-access sites. A good understanding of the geology (both in terms of the rock types and the fracture networks) is required for event evaluation (e.g., modulation of seismic signals from the source), planning of treaty monitoring sensors (fracture play a key role in the gas migration pathways), and evaluation of containment strategies. The Hypothesis: Quantitative analysis of terrain characteristics provides a way to relate the topography (easily obtained at high resolution) to surface geology. The Method: Integration of Terrain Complexity Index (TCI, derived from surface roughness) with drainage patterns and texture to produce a "predicted" geologic map. The viability of this method was evaluated by applying the algorithm to an area of know geology (Area 15 at the Nevada Test Site - NNSS). A 10-meter digital elevation model (DEM) was used for this evaluation. The Application: We have demonstrated the utility of this method by developing an improved surface geologic map for the North Korean Test Site using a 5-meter DEM. Why Now: Improved geologic characterization of potential nuclear test sites in a number of countries is needed for a number of Nonproliferation objectives. What's Needed Next: 1) A more robust and automated interpretation algorithm for the drainage analysis aspect; 2) improved data fusion of the TCI with drainage patterns; and 3) evaluation of the potential for fracture mapping in granite using a very high resolution DEM at the NNSS site (a 1-meter DEM in on order).

58 GEOSCIENCES↗
DOE OSTI2018

Investigation of Coupled Processes in Argillite Rock: FY18 Progress

The report presents the current status of the TOUGH-FLAC and TOUGH-RBSN simulators for modeling of coupled THM processes in argillite, including fracturing, and verification of the TOUGH-FLAC model against major in situ heater experiments at the Mont Terri Underground Research Laboratory (URL) in Switzerland and at the URL in Bure, France. The heater experiments modeled are the Mont Terri FE (Full-scale Emplacement) Experiment, conducted as part of the Mont Terri Project, and the TED and ALC experiments conducted in Callovo-Oxfordian claystone (COx) at the Meuse/Haute-Marne (MHM) URL in Bure, France. The modeling of the TED and ALC heater experiments is part of a modeling task (Task E) of the international DECOVALEX-2019 project. DECOVALEX, which stands for DEvelopment of COupled Models and their VALidation against EXperiments, is an international collaborative activity, in which DOE and LBNL gain access to unique laboratory and field data defined as modeling test cases. DECOVALEX tasks are studied collectively among several international groups to better understand the processes and to improve numerical models that could eventually be applied in the performance assessment for nuclear waste disposal in clay host rocks and bentonite backfill. The report also includes a description of LBNL’s activities for modeling gas migration in bentonite related to Task A of the international DECOVALEX project, and presents new FY2018 work related to monitoring and characterization of faults using a unique borehole tool.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Product consistency test results for the LAW Phase 4 gases

In this report, the Savannah River National Laboratory provides chemical analysis of Product Consistency Test (PCT) leachates from a series of simulated nuclear waste glasses fabricated at the Pacific Northwest National Laboratory (PNNL). The series included quenched and canister-centerline cooled (CCC) versions of the glasses. The resulting data will be used in the development of enhanced property/composition models for waste vitrification at Hanford. For some of the glass leachates, minor scatter among the triplicate values of some analytes were observed. For other leachates, there were more significant differences among the triplicate values. A review of the PCT data noted that there was little difference between the normalized values based on targeted or measured glass composition. Several of the study glasses have normalized concentration of element “i” (NCi) values that are greater than the Hanford Tank Waste Treatment and Immobilization Plant immobilized low-activity waste constraint of 4 g/L for boron (B), sodium (Na), and silicon (Si). The results of these glasses will help ensure the ability of advanced glass performance models to appropriately predict acceptable compositions. For the study glasses with NCi values exceeding 4 g/L, the CCC heat treatment samples had generally lower NCi values than quenched samples. The samples of the Environmental Assessment (EA) reference glass included with each PCT set had generally consistent NCi values. The release rates for boron (B), potassium (K), lithium (Li), sodium (Na), and silicon (Si) were highly correlated for the study glasses.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Summary of the Initial Post-Irradiation Characterization of HFIR-Irradiated Low-N and High-N HT-9 Steel

Reference cladding systems for sodium fast reactors are based on the historical steel, HT-9. HT-9 is a Fe12Cr ferritic/martensitic steel with additions of Mo, W, V, and other minor elements and demonstrates low irradiation swelling and adequate mechanical properties. Extensive irradiation literature exists on the use of HT-9 as cladding for metal fuel, primarily irradiation on the U-Zr/HT-9 system from the Experimental Breeder Reactor-II (EBR-II) and Fast Flux Test Facility (FFTF) sodium fast reactor, and as a structural material from experiments in the FFTF. The large amount of historical data makes the U-Zr/HT-9 system the reference fuel specification for many nuclear reactor vendors that seek to license modern sodium-cooled fast reactors in the United States. However, it is yet unclear how variations in impurity content within HT-9 fundamentally affect irradiation performance at various irradiation temperatures. Recent work suggests that impurity content may noticeably alter the production of helium through nuclear transmutation. For these reasons, High-Flux Isotope Reactor (HFIR) irradiation of HT-9 steels with known variations in the impurity content is particularly timely to generate data to enable more accurate refinement of the chemical specification for nuclear-grade HT-9 material. This report summarizes the initial transmission electron microscopy characterization of HFIR-irradiated HT-9 steels following mechanical property measurements by the Advanced Fuels Campaign (AFC). This report includes qualitative results of the cavity, dislocation loop and cluster/precipitate microstructures as well as radiation-induced segregation. Quantitative results are being shared with partner institutions and will be included in more detail in a future report in FY2026.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1993

Particle bed reactor modeling

The topics are presented in viewgraph form and include the following: particle bed reactor (PBR) core cross section; PBR bleed cycle; fuel and moderator flow paths; PBR modeling requirements; characteristics of PBR and nuclear thermal propulsion (NTP) modeling; challenges for PBR and NTP modeling; thermal hydraulic computer codes; capabilities for PBR/reactor application; thermal/hydralic codes; limitations; physical correlations; comparison of predicted friction factor and experimental data; frit pressure drop testing; cold frit mask factor; decay heat flow rate; startup transient simulation; and philosophy of systems modeling.

Sapyta, Joe↗
DOE OSTI2019

Novel Low-Cost Approach for Acquiring High Resolution High-Speed Data

The Oak Ridge National Laboratory (ORNL) has developed and tested a novel system architecture for acquiring high fidelity high-speed data. The approach uses a low-cost consumer grade audio recording device coupled with computing technology running open-source software. The main advantage of this approach is perchannel cost; an instrument grade data acquisition system typically costs between $800 to $2,000 per channel compared to less than $50 per channel for these consumer grade components. Three systems, each featuring four channels, have been deployed for acquiring data from geophones and the electrical supply system that supports the High Flux Isotope Reactor (HFIR) and the Radiochemical Engineering Development Center (REDC) at ORNL. Each channel samples at 96 kHz at 24-bit resolution. The deployed systems operate continuously 24/7 and produce about 4 terabytes of data per month per system. This paper provides a technical overview of this approach, its implementation, and some preliminary results from qualification testing. This work was conducted in support of the Multi-Informatics for Nuclear Operations Scenarios (MINOS).

Wetherington Jr, G↗
DOE OSTI2013

Characterization of Canberra's Tomographic Gamma-Ray Can Scanner ('Can-TGS') - 13311

The Tomographic Gamma-ray Scanner (TGS) for large volume drummed waste has been successfully commercialized by Canberra over the last several years. As part of an R and D effort to continually improve this technology, we have developed a scaled down version of the standard commercial product (Can-TGS). The Can-TGS is able to accommodate cans and pails of various sizes, ranging from sub-liter to 20 liter volumes with densities of up to 4 g/cc. The Can-TGS has three diamond-shaped collimators (6.35 mm [0.25''], 12.7 mm [0.5''], and 25.4 mm [1'']) to facilitate a range of container volumes and heights. As with the standard TGS, the Can-TGS has a transmission source sub-system, where the transmission source can be easily swapped between sources of various strengths and type. The acquisition portion of the Can-TGS is powered by the Canberra Lynx{sup R} MCA which accommodates both multi-spectral scaling (MSS) and list-mode. Recently, the Can-TGS has been successfully characterized for an 18.93 L [5-gallon] container for the 25.4 mm diamond-shaped collimator. In principle, a single measurement (with good statistics) is required for each configuration in order to characterize the system. However, for this study, measurements were performed for several different matrices. For each matrix used, 6 different measurements were acquired. For each of these measurements, the drum was rotated 60 deg. with respect to the previous starting position. This procedure was followed in order to average out any radial bias that might be produced from just a single measurement. A description of the Can-TGS system is given. The details of the recent characterization measurements and the associated data analysis and results are presented. TGS results are compared with Segmented Gamma Scanner (SGS) results for the same source configuration. Additionally, the future outlook for Canberra's R and D efforts with this system is discussed. These efforts include TGS algorithm optimization, improving our understanding of the uncertainties associated with TGS results, and the overall benefits and limitations of this technology as compared to other available non-destructive assay techniques. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

Quantitative resilience evaluation on recovery from emergency situations in nuclear power plants

Here, this paper mainly introduces how to develop a resilience evaluation model by quantifying the relationship between resilience and resilience components in regard to recovering from emergency situations in nuclear power plants (NPPs). It is an extension of the author’s previous researches. To develop the model, the first step was to analyze event reports published in Republic of Korea in terms of resilience factors from different safety perspectives. Second, statistical methods such as factor analysis, principal component analysis, and multivariate regression analysis were applied to the dataset to identify relations among resilience factors and develop a quantitative resilience evaluation model for recovery from emergency situations in NPPs. Third, the model was validated using the recent event report data released in Korea. Lastly, we present a discussion on how to apply the quantitative resilience evaluation model to results from stress tests performed in Korea.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Integral Full Core Multi-Physics PWR Benchmark with Measured Data

In recent years, the importance of modeling and simulation has been highlighted extensively in the DOE research portfolio with concrete examples in nuclear engineering with the CASL and NEAMS programs. These research efforts and similar efforts worldwide aim at the development of high-fidelity multi-physics analysis tools for the simulation of current and next-generation nuclear power reactors. Like all analysis tools, verification and validation is essential to guarantee proper functioning of the software and methods employed. The current approach relies mainly on the validation of single physic phenomena (e.g. critical experiment, flow loops, etc.) and there is a lack of relevant multiphysics benchmark measurements that are necessary to validate high-fidelity methods being developed today. This work introduces a new multi-cycle full-core Pressurized Water Reactor (PWR) depletion benchmark based on two operational cycles of a commercial nuclear power plant that provides a detailed description of fuel assemblies, burnable absorbers, in-core fission detectors, core loading and re-loading patterns. This benchmark enables analysts to develop extremely detailed reactor core models that can be used for testing and validation of coupled neutron transport, thermal-hydraulics, and fuel isotopic depletion. The benchmark also provides measured reactor data for Hot Zero Power (HZP) physics tests, boron letdown curves, and three-dimensional in-core flux maps from 58 instrumented assemblies. The benchmark description is now available online and has been used by many groups. However, much work remains to be done on the quantification of uncertainties and modeling sensitivities. This work aims to address these deficiencies and make this benchmark a true non-proprietary international benchmark for the validation of high-fidelity tools. This report details the BEAVRS uncertainty quantification for the first two cycle of operations and serves as the final report of the project.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Spectroscopic online monitoring for process control and safeguarding of radiochemical streams

This paper summarizes application of the absorption and vibrational spectroscopic techniques supplemented by physicochemical measurements for radiochemical process monitoring. In this context, our team experimentally assessed the potential of Raman and spectrophotometric techniques for online real-time monitoring of the U(VI)/nitrate ion/nitric acid and Pu(IV)/Np(V)/Nd(III), respectively, in solutions relevant to spent fuel reprocessing. These techniques demonstrate robust performance in the repetitive batch measurements of each analyte in a wide concentration range using simulant and commercial dissolved spent fuel solutions. Spectroscopic measurements served as training sets for the multivariate data analysis to obtain partial least squares predictive models, which were validated using on-line centrifugal contactor extraction tests. Satisfactory prediction of the analytes concentrations in these preliminary experiments warrants further development of the spectroscopy-based methods for radiochemical process control and safeguarding. Additionally, the ability to identify material intentionally diverted from a liquid-liquid extraction contactor system was successfully tested using on-line process monitoring as a means to detect the amount of material diverted. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Improved Heat Transfer and Volume Scaling through Novel Heater Design

In 2014, the University of California, Berkeley, Nuclear Engineering Thermal Hydraulics Laboratory completed construction of the Compact Integral Effects Test (CIET) facility, a scaled thermal hydraulics scaled integral effects test facility designed to model the Mark-1 pebble-bed fluoride-salt-cooled high-temperature reactor (Mk1 PB-FHR or simply Mk1). CIET uses an organic simulant fluid, Dowtherm A, to match the Prandtl, Reynolds, Froude, and Grashof numbers of important fluoride salt candidates for molten salt reactors, such as flibe. The FHR reactor core is simulated using a resistively heated tube, which heats the oil flowing through the annulus between the heated tube and a non-heated inner tube. However, the volume fraction of the coolant in the CIET heater was dramatically lower than the volume fraction of the primary coolant in the Mk1 PB-FHR design (3% in CIET vs. 17% in Mk1) because the CIET heater was initially scaled to match an earlier FHR design. Also, the maximum core heater power in CIET is limited by the melting point of Teflon components in CIET, approximately 330 deg. C. For conservatism, the CIET heater power is limited to keep the maximum surface temperature of the heater below 250 deg. C. At the nominal mass flow rate of 0.18 kg/s, this power was about 5 kW. A higher heat transfer coefficient from the heater wall to the oil was needed to increase the effective power delivered to the oil. In order to alleviate these two limitations, a new heater insert was designed and constructed for CIET. This new heater insert uses a twisted tape to augment heat transfer. The twisted tape is supported inside a perforated inner tube that has the same diameter as the original inner tube of the heater, in order to make it compatible with CIET's original heater design. The heat transfer and pressure drop properties of the new heater insert designs were tested in CIET and compared to the previous design in terms of pressure drop and convective heat transfer coefficient. The new heater inserts are intended to increase the range of power levels and mass flow rates that can be tested in CIET in order to augment understanding of FHR thermal hydraulic phenomena. Also, obtaining different data points for CIET response dynamics under different heater coolant residence times will be valuable in evaluating the Mk1 design.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗