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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

Examination of New Theory for Neutron Multiplicity Counting of Non-Point-Like Sources of Special Nuclear Material

The purpose of a nondestructive assay is to accurately verify the declared mass of special nuclear material (SNM) samples in a limited amount of time. One measurement modality is neutron multiplicity counting (NMC), which relates time-correlated neutron detection rates to the mass of SNM present. Traditional theory assumes point-like sources, which can be ill-posed for kilogram-quantity, bulk samples. Recent theory was developed to improve NMC accuracy for non-point-like samples. This work preliminarily examines the theory with measured data. The OSCAR prototype (a 3-by-4 array of 5.08-cm-thick, 5.08-cm-diameter trans-stilbene organic scintillators) measures configurations of 29.41-49.00 kg of highly enriched uranium (93wt% 235 U). Adjusted factorial moments for the emitted neutron multiplicity distributions are used to account for the shape and type of SNM being measured.

252Cf↗
DOE OSTI2023

Examination of a Methane/Diesel RCCI Engine Using Pele: Preprint

Multi-fuel, advanced injection strategies have become increasingly promising as a strategy to mitigate the emissions generated from internal combustion engines. By carefully controlling the combustion phasing in-cylinder, these new multi-pulse, multi-fuel injection strategies are able to burn in the low-temperature combustion regime where both NOx and soot are not readily produced, reducing the need for extensive exhaust gas recirculation systems. In this study, we examine a reactivity-controlled compression ignition (RCCI) strategy that uses an early pre-filled methane-air mixture with low turbulence background as the low-reactivity fuel and a direct injection of four discrete dodecane jets as a surrogate for the high-reactivity diesel fuel. We use the Pele software suite, a highly optimized, exascale-ready, adaptive mesh refinement codebase to perform high-resolution numerical simulations of a scaled down, single cylinder from the RCCI engine. Here, we resolve the ignition kernels down to micrometer scales and present several statistical quantities evaluating the development of the flow and detailing the onset of ignition and subsequent flame development. Particular attention is paid to the conditions surrounding the onset of the first ignition kernels and discussing what led to the development of those conditions.

CFD↗
DOE OSTI2018

Post-Irradiation Examinations of High Burnup PWR Fuel Rods - Initial Results

The High Burnup Spent Fuel Data Project (herein referred to as the “project”), sponsored by the US Department of Energy (DOE) Office of Nuclear Energy (NE), is focused on experimentally evaluating the effects of long-term storage and transportation on high burnup (HBU) (defined as >45 gigawatt days per metric ton uranium) pressurized water reactor (PWR) spent nuclear fuel (SNF) [1]. To support this project, in the fall of 2017, an instrumented bolted-lid cask (project cask) at the North Anna reactor site was loaded with intact, HBU PWR fuel assemblies (project assemblies) having four different kinds of cladding and will be reopened in approximately 10-years for inspection. To support pre- and post-storage SNF rod characterization, 25 “sister rods” that are similar to the rods being stored in the project cask were extracted from seven different fuel assemblies at the North Anna nuclear power plant for post-irradiation examination (PIE).

Bevard, Bruce Balkcom↗
DOE OSTI2012

Examination of temperature dependent subgroup formulations in direct whole core transport calculation for power reactors

The traditional subgroup method which has been applied for lattice transport calculations has an inherent limitation for non-uniform temperature distributions. As a measure to incorporate temperature dependence into the subgroup formulation, the subgroup level and number density adjustment method have been proposed. In this paper, the temperature dependent subgroup formulations employed for reflecting the non-uniform temperature effects on the resonance spatial self-shielding are examined for the whole core transport calculation with the thermal feedback. For 2D pin-cell problem with non-uniform temperature profiles, the inherent limitation of conventional subgroup method is confirmed. And the improvement in terms of reactivity is observed with the proposed adjustment scheme. For the real PWR core calculation with thermal feedback in the hot-full-power condition, the noticeable correction for the fuel temperature coefficient by about 10% more negative is obtained with the correction schemes. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Evaluation of the radiation doses in newborn patients submitted to CT examinations

The number of computed tomography (CT) scans available to the population is increasing, as well as the complexity of such exams. As a result, the radiation doses are increasing as well. Considering the population exposed to CT exams, pediatric patients are considerably more sensitive to radiation than adults. They have a longer life expectancy than adults, and may receive a higher radiation dose than necessary if the CT scan settings are not adjusted for their smaller body size. As a result of these considerations, the risk of developing cancer is of great concern when newborn patients are involved. The objective of this work was to study the radiation doses on radiosensitive organs of newborn patients undergoing a whole body CT examination, utilizing Monte Carlo simulations. The novelty of this work is the use of pediatric virtual anthropomorphic phantoms, developed at the Department of Nuclear Energy at the Federal University of Pernambuco (DEN/UFPE). The CT equipment utilized during the simulations was a Discovery VCT GE PET/CT system, with a tube voltage of 140 kVp. The X-ray spectrum of this CT scanner was generated by the SRS-78 software, which takes into account the X-ray beam energy used in PET/CT procedures. The absorbed organ doses were computed employing the F6 tally (MeV/g). The results were converted to dose coefficients (mGy/100 mA) for all the structures, considering all employed beams. The highest dose coefficients values were obtained for the brain and the thyroid. This work provides useful information regarding the risks involving ionizing radiation in newborn patients, employing a new and reliable technique. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Nuclear fuel behavior at high temperature: new insights from post-test examinations on the VERDON-1 sample

In order to reduce the uncertainties concerning Source Term assessment under Light Water Reactor (LWR) severe accident conditions, 4 VERDON tests were recently performed. They have been defined in the framework of the International Source Term Program and carried out at CEA Cadarache, in the LECA-STAR facility. Among these 4 VERDON tests, VERDON-1 test was devoted to high burn-up UO{sub 2} fuel behavior and FP release under H{sub 2} reducing conditions at very high temperature. This test was performed with the Release Circuit, which is devoted to the precise characterization of the FP release, in terms of FP release kinetics and fission products (FPs) released fraction. The main components of the circuits are a furnace heating the fuel sample, a filter trapping the emitted aerosols and gamma spectrometry detectors for online and post-test measurements. VERDON-1 test was conducted up to about 2600 Celsius degrees without reaching global fuel relocation, allowing thereby post-test micro-analyses. After a brief description of the fuel sample nature and the experimental set-up, we present the progress of the experimental sequence and the associated gamma spectrometry measurements. We then focus on the major results of the VERDON-1 tests regarding the fuel global behavior. These results are obtained by two complementary techniques. First, the fuel sight gamma station can be used either on-line to detect fuel relocation or just after the test, as a longitudinal bench, to get preliminary information on the global shape of the fuel sample inside the furnace. Secondly, these information are completed by post-test examinations performed on the fuel sample after extraction from the furnace. Complete qualitative and quantitative characterizations, including EPMA (Electron Probe Micro-Analysis) and SEM (Scanning Electron Microscopy) analysis, were performed for the first time at the CEA Cadarache on a sample that has been submitted to such severe accident conditions. These results show an advanced state of sample degradation, with locally previously molten areas, meaning that fuel global relocation was imminent by the end of thermal-hydraulic sequence. The relocation temperature is compared to that of previous VERCORS tests and discussed regarding the atmosphere of the test and possible UO{sub 2}-ZrO{sub 2} interaction. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Hanford double-shell tank ultrasonic test examinations - 15287

Washington River Protection Solutions, LLC (WRPS) under contact from the US DOE is responsible for assessing the condition of the double-shell tanks (DSTs) of the Hanford Nuclear Site. WRPS contracted AVERA Federal Services, LLC to perform the ultrasonic testing inspections of the primary tank wall in 28 DSTs to access the condition of the tanks, judge the effects of past corrosion control practices, and satisfy a regulatory requirement to periodically (eight to ten year frequency) assess the integrity of the tanks per Brookhaven National Laboratory (BNL) guidelines for structural integrity programs for tank systems. The major requirements for the Ultrasonic Testing (UT) of each tank are to detect, characterize (identify, size and locate), and record measurements made of any wall thinning, pitting or cracks that might be present in the wall of the primary tank. The UT program examines representative areas of the primary tank and secondary liner by deploying equipment in the annulus of the tank through two 24-inch diameter risers approximately 180 deg. apart. From 1996 to 2006 the first-round of ultrasonic test inspections for all DSTs were completed utilizing a Pulse-Echo (P-Scan) system. The second round of ultrasonic test inspections have continued since 2007 and are complete for 26 of 28 tanks. The two remaining tanks will receive the second inspections during fiscal year 2015. Currently, the ultrasonic testing comparison results between the first and second rounds of inspection have identified no structural integrity concerns at action levels specified in the BNL guidelines. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Department of energy tank closure - an examination of alternative approaches - 15390

Closure of single- and double-shell underground storage tanks at sites across the DOE complex poses a unique technical and regulatory challenge. Some sludge waste residues invariably remain in a tank after waste retrieval - especially in tanks with obstructions or associated piping. In addition, reducing the volume of these waste residues becomes increasingly difficult depending on the constituents and the age of the heel. This paper discusses tank cleaning and closure alternatives to allow a more efficient retrieval and characterization of the tank wastes. The paper also examines the possible use of a tank closure strategy that is informed by risk rather than subjective criteria. Gaps in the technical foundation supporting tank retrieval and closure have traditionally resulted in a number of overly conservative assumptions in performance assessments. An alternate tank closure approach would be to develop a scientific basis for tank closure which may allow greater waste volumes to be left in the tanks post-retrieval, while increasing protectiveness to human health and the environment. The approach is to characterize complex residual tank waste solids, measure contaminant release rates as a function of chemical environment, and build mechanistic release models. (authors)

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

Droning on to Delivery: Examining the Energy Impacts of Using Drones for Moving Goods

The demand for fast, localized delivery has grown significantly in recent years. Whether for cheesy snacks, prepared food, medical solutions, or business deliveries, fast and efficient delivery is increasingly a demand and differentiator. Drone delivery in the freight sector offers to revolutionize last-mile logistics and improve services. This research analyzes the impacts of drone energy for delivery operations, aiming to compare various types of drones (large and small, rotary and VTOL) and various types of business methods. The project executes novel open-air and laboratory-based testing to look at the impacts of weights, operations, temperatures and weather conditions. The study combines this real-world experimental data with fleet optimization mathematical models to assess energy consumption of different scenarios as well as examine the minimum fleet size and additional battery requirements. These fleet optimization models compare energy between types of deployments with existing delivery methods. The model was also extended to look at mixed fleets of aerial drones and ground vehicles to accommodate different restrictions on drones or when weather prohibits their use. The analysis provided insights into factors such as drone design, payload weight, flight distance, weather conditions and operational parameters and impacts of each. It showed how to combine different types of vehicles to reduce energy and improve services. And it showed that drone speed, routing restrictions, and unfavorable weather significantly influence energy consumption. Our open experiment data and optimization models can assist stakeholders and industry in understanding drone package delivery and offers keys to improving deployment.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2016

Computational and Experimental Examination of Simulated Core Damage and Relocation Dynamics of a BWR Fuel Assembly - 16013

The Japan Atomic Energy Agency (JAEA) plans a large-scale test to evaluate damage and relocation behavior of Boiling Water Reactor (BWR) core materials consisting of fuel rods (zirconia as fuel simulant), channel boxes, control blade assemblies as well as lower core-support structures. Its purpose is to contribute to the understanding of the core material relocation behavior in the event of severe accidents with the BWR design conditions for which existing experimental database is quite limited. Prior to large-scale testing anticipated at JAEA in the future, JAEA desires preliminary investigations to examine melting test pieces, being composed of a mixture of zirconia and metals by a non-transferred thermal plasma heating system. The purpose of such tests is to verify the materials and test piece will be heated by plasma as a primary heat source to the target temperature of ca. 2900 K and to collect data about the relocation behavior of the high-temperature materials. This paper describes the first in a series of tests demonstrating the effective use of thermal plasma technology to recreate the dynamics of a typical fuel core bundle channel assembly section. Results from preliminary computational simulations of simplified fuel bundle assemblies are presented illustrating the effectiveness of a 150 kW, non transferred (NTR) configuration of a thermal plasma argon jet to heat the assembly within a highly-insulated containment vessel. An experimental test program using the computational analyses as a basis and a Phoenix Solutions Co (PSC) commercial model PT150, 150 kW plasma torch is described and the test results are presented illustrating the time versus temperature history of the fuel bundle assembly melting dynamics. Details on plasma operation are discussed including: power requirements, current flow, voltage levels achieved, argon gas flow rates, processing times and standoff distances of the torch discharge from the top of the fuel bundle assembly (achieved with remote control of the axial (vertical) translation of the torch within the containment processing chamber). Photos taken from the top of the processing chamber are presented illustrating the various phases of melting characteristics of the zircaloy-clad, zirconium oxide filled fuel tubes, all within a stainless steel tube bundle channel centered within an insulated zirconia crucible. Based on lessons learned from this initial test campaign, a brief discussion of the requirements and features of a follow-on Phase II test program is provided. (authors)

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

Examining Supply Chain Resilience for the Inter-modal Shipment of Spent Nuclear Fuel and High Level Radioactive Materials - 16309

The U.S. Department of Energy (DOE) has a significant programmatic interest in the safe and secure routing and transportation of Spent Nuclear Fuel (SNF) and High Level Waste (HLW) in the United States, including shipments entering the country from locations outside U.S borders. In any shipment of SNF/HLW, there are multiple 'chains'; a jurisdictional chain as the material moves between jurisdictions (state, federal, tribal, administrative), a physical supply chain (which mode), as well as a custody chain (which stakeholder is in charge/possession) of the materials being transported. Given these interconnected networks, there lies vulnerabilities, whether in lack of communication between interested stakeholders or physical vulnerabilities such as interdiction. By identifying key links and nodes as well as administrative weaknesses, decisions can be made to 'harden' the physical network and improve communication between stakeholders. This paper examines the parallel chains of oversight and custody as well as the chain of stakeholder interests for the shipments of SNF/HLW and the potential impacts on systemic resiliency. Using the Crystal River shutdown location as well as a hypothetical international shipment brought into the United States, this paper illustrates the parallel chains and maps them out visually. (authors)

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

Micro/nano-structural examination and fission product identification in neutron irradiated AGR-1 Triso fuel - 18529

Advanced electron microscopic and micro-analysis techniques were developed and applied to study irradiation effects and fission-product behavior in selected low-enriched uranium-oxide/uranium-carbide tri-structural-isotropic (TRISO)-coated particles from fuel compacts in four capsules irradiated to burnups of 11.2 to 19.6% fissions per initial metal atom (FIMA) consisting of Baseline, Variant 1, and Variant 3 fuel types. Trend analysis shows precipitates were mostly random in their distribution along the perimeter of the inner pyrolytic carbon-silicon carbide (IPyC-SiC) interlayer with only weak association with kernel protrusion and buffer fractures. Pd is dominantly found in most precipitates in both intra and intergranular locations. Nano-sized Ag is predominantly found in grain boundaries and triple points with only two findings of Ag inside a SiC grain in two different compacts (Baseline and Variant 3 fueled compacts). Generally, more element combinations exist for precipitates from particles with relatively low Ag retention compared to particles with relatively high Ag-retention irrespective of fuel type. This study shows the presence of Cs in particles from all compacts evaluated. From this work, no single fission product mechanism hypothesis can be reported. The complexity of mechanisms is further highlighted by the multiple variations of elemental combinations found in the more than 700 fission product precipitates examined. It seems that movement of Ag is not assisted by a specific element in all cases. Therefore, it is not necessarily true that a chemical-assisted transport mechanism is dominant. The presence of Ag predominantly on grain boundaries suggests that a grain boundary transport mechanism may be prominent. Studies to determine the effect of neutron damage are recommended for future work. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Examining the Effect of Plume Rise on Radiological Consequences of Nuclear Accidents

A nuclear accident could cause release of radioactive material outside containment. This material will eventually disperse into the surrounding environment, and could pose a serious threat to public safety, and environment. Mitigation of radiological consequences of such an accident is highly desirable, and since Fukushima accident, there has been a growing focus on nuclear utilities to develop capabilities for this purpose. Undoubtedly, suppression of radioactive material release and dispersion is the primary mean for mitigating radiological consequences nevertheless alternative methods should also be explored to broaden the spectrum of choices for actual circumstances. As we know that, dispersion of radioactive material and subsequent radiological consequences are affected by plume rise. Plume rise is an important factor in determining ground level contamination since it can increase effective release height appreciably provided that effluent discharge speed and temperature is high enough. Under most circumstances, one can expect reduction in radiological consequences due to lower levels of radiation dose and ground contamination in the presence of adequate plume rise. Nevertheless, the effect of plume rise is usually ignored in radiological consequence studies for the sake of conservatism, and lack of information. However, this particular attribute of plume rise interested us to investigate the effect of plume rise in the scope of consequence mitigation and how this advantage could be utilized during an accident progression to minimize the impact of radioactivity release to the environment. In this work, we examine the effect of plume rise on radiological consequences of a nuclear accident for plume rise contributions much larger ({approx}500 m) than usually achievable in normal circumstances. The effect of plume rise on radiological consequences is quantified in this study as decontamination credit (DCC) achievable corresponding to maximum annual dose. This analysis is performed for a number of release heights and atmospheric conditions. We discuss available results in the scope of radiological consequence mitigation, devising a strategy for utilizing plume rise phenomenon with the help of an external propelling device like a jet engine for radiological consequence management. This study is continuation of our efforts for devising systems and strategies to mitigate consequences of radioactivity release accidents like Fukushima. We have proposed several ideas for suppressing/preventing the dispersion of released radioactive material to the environment using offsite equipment.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Status Report on Fast Flux Test Facility Mechanistic Fuel Failure Experiment Analysis with BISON for Post Irradiation Examination Support

The renewed interest in metallic U-Zr nuclear fuel alloy has led to a drive for deeper understanding of the mechanisms driving the phenomena observed under irradiation conditions. The Department of Energy Advanced Fuel Campaign has developed infrastructure to support metallic fuel development, including Post Irradiation Examination (PIE) of legacy Fast Flux Test Facility (FFTF) Mechanistic Fuel Failure (MFF) experiments. The PIE performed on legacy FFTF MFF experiments gives insight on metallic fuel performance and can address the lack of knowledge and scarcity of reliable data identified in several studies over recent years. Unfortunately, PIE efforts can cost significant time and resources which can impede the progress of metallic U-Zr fuel development. Metallic U-Zr fuel performance modeling can be used to inform PIE efforts on regions of interest for relevant investigations and can help understand phenomena observed in PIE. This report demonstrates the current progress of FFTF MFF fuel performance simulations using the BISON fuel performance code and discusses the support provided by simulation to PIE efforts. Progress in temperature, profilometry, fission gas release, plenum pressure, and zirconium redistribution simulation results have been demonstrated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Post-Transient Examination Results of RIA Commissioning Teats at the Transient Reactor Test Facility

Six reactivity-initiated-accident (RIA) commissioning tests have been performed at Idaho National Laboratory’s (INL) Transient Reactor Test Facility. Five of these tests were performed using fresh fuel rodlets, and the sixth test was performed using a previously irradiated rodlet from the ATF-2 irradiation experiment. These experiments demonstrate the ability to perform RIA testing including the ability to perform experiments using previously irradiated materials. Post-transient examinations revealed pellet cladding interactions and cladding ballooning depending on the initial boundary conditions of the test. Increased thickness of both zirconium oxide and alpha-zirconium were found in the cladding with tests performed at higher levels of total energy deposition. No oxide or alpha-zirconium was found in the test with the greatest amount of ballooning, which was also the test with the highest initial rod internal pressure. The highest energy deposition test (1110 J/g) resulted in failure of the rodlet. Fragmentation of the fuel pellets occurred in this test with the size of the fragments decreasing inversely to the pellet radius.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Sirius Irradiation Experiments and Post-Irradiation Examinations for Nuclear Thermal Propulsion

Nuclear Thermal Propulsion (NTP) systems hold promise in reducing transit times for exploration of the solar system by both crewed and uncrewed missions. NTP systems currently under investigation include a once-through high temperature gas-cooled fission reactor to provide thermal energy to heat the coolant which also serves as the propellant. The fuel systems of angular UN fuel particles dispersed in a matrix of W/Re, creating a ceramic and metallic composite or cermet, has been irradiated in Idaho National Laboratory’s (INL) Transient Reactor Test Facility (TREAT) enabling evaluation of these materials under representative nuclear heating rates (~95 K/s) and peak temperatures (~2527 K). These tests named Sirius-1 (UN-W/Re), have been irradiated and this paper will present post-irradiation examination results. The Sirius-1 test produced cracks in the fuel specimen and spalling of surface material. Uranium soot was found on the inner wall of the irradiation capsule indicating loss of some fissile material from the fuel specimen. Spalling from the surfaces was also noted upon visual inspection. Uranium diffusion from the fuel particles resulted in the formation of U/Re phases and edge features producing a laminar microstructure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Non-destructive Post-Transient Examination Results of the MOXTOP-1 Experiment

The MOXTOP-1 experiment has been assembled and irradiated in INL’s HFEF and TREAT facilities followed by disassembly of the experiment to enable initial post-transient examination. Results of neutron radiography of the fuel pin still in the THOR capsule were somewhat inconclusive due to the additional scatter and attenuation from the iron heat sink included in the design. Nonetheless, no obvious failure or fuel relocation could be identified in the radiography. Following disassembly of the fuel pin from the capsule, initial visual inspections were performed, and documented via images obtained through the hot cell window. No obvious signs of fuel failure were identified, indicating that the fuel pin remained intact during the transient. Some discoloration was observed near the middle of the fuel pin, likely corresponding to the fueled zone indicating that that region may have reached a relatively high temperature. Pre-transient, and post-transient profilometry of the fuel pin were performed and compared. Some additional strain displacement in the bottom portion of the fuel zone is identified, but remains well below 1%.

Mixed Oxide↗