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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 559 records · Page 31
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Remote In-bore Laser Cutting and Welding Tools for use in Future Nuclear Fusion Reactors - 18348

The maintenance, replacement and decommissioning of future nuclear fusion reactors will require quick and reliable cutting and joining of in-vessel pipework. Initial design studies for nuclear fusion reactors for power generation have estimated cutting and welding could account for up to 60% of the maintenance duration using conventional in-situ processing techniques (dry-mechanical cutting and arc welding) and that new methods are required to expedite the process. Additionally, the expected radioactivity and limited access at the cutting and welding sites mean that these processes cannot be done manually and robotic tools are required. To this end, remote in-bore laser cutting and welding tools have been developed for use in 90 mm internal diameter steel pipes. Here, we will present the designs of the remote in-bore laser cutting and welding tools, and how the tools fit within the overall remote maintenance strategy of a nuclear fusion reactor. Initial high-power laser trials were performed with prototype tools which successfully demonstrated their laser processing, thermal management and dust protection functionality. Although the tools described herein have been developed specifically for use within the fusion reactor environment, however the tool design and technology demonstrated here is readily transferable to many remote applications in challenging environments such as fission reactor maintenance, nuclear decommissioning and other in-accessible pipework. (authors)

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

Status of GTA Gear Inspection after Testing in Sodium

The Mechanisms Engineering Test Loop (METL) is an intermediate-scale liquid metal experimental facility that provides purified reactor-grade (R-grade) sodium to various experimental test vessels. In these test vessels, components that are required to operate in an advanced fast reactor can be tested in a prototypical sodium environment. Experiments conducted in METL significantly assist in the development and maturation of systems and components for advanced reactors. The METL facility consists of multiple test segments including: a purification and diagnostic loop (cold trap was designed after EBR-II), two 18” diameter test vessels with 150 L capacities capable of operating in static or dynamic flow at 538°C, two 28” diameter test vessels with 644 L capacities capable of operating in static or dynamic flow at 650°C, a 3,180 L dump tank with 21 instrumentation ports, and the infrastructure to expand capacity in the future.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

LX-17 Thermal Decomposition-Characterization of Solid Residues from Cook-Off in a Small-Scale Vessel Under Confinement

Concerns surround whether insensitive (or any) energetic materials are more dangerous to handle when exposed to abnormal thermal environments. This study characterizes the residual material remaining after LX-17 (92.5 % 1,3,5-triamino 2,4,6-trinitro benzene (TATB) and 7.5 % Kel-F) is exposed to various thermal environments in a sealed small-scale vessel cook-off test reactor (heated at 0.1 to 100 °C/min until the reactor opened at 3000 psi (20.7 MPa)). Previous work has shown no additional sensitivity of these residues as evaluated by small-scale safety analysis, but characterization on the molecular scale indicates the TATB is transformed to more reactive compounds as well as the residue could be precursors to toxic gases. The solids and chars were characterized by various analytical methods. Heat-flow measurements indicated exothermic release is due to a mixture of residual TATB and related decomposition products (which may be more energetic). The N/C and O/N ratios indicated a material much more degraded than TATB. Primarily, the solids were a network of amorphous C inter-dispersed with N and O. Types of bonding include C–C, C–N, N–H, N–C, N=C, N≡C, C–O=, and –OH. Solvent extracts of the solids showed TATB decomposition intermediates benzo-furazans and benzo-furoxans, substituted TATB (mono-nitroso, hydroxyl, and chlorinated) along with several unidentified smaller molecules. These results indicate thermal treatment produces an amorphous carbon residue with heteroatoms incorporated through differing functionality, varying depending upon the thermal severity of exposure. These structures also could further decompose producing toxic light gases (such as cyanide).

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

The effect of helium on cavity swelling in dual-ion irradiated Fe and Fe-10Cr ferritic alloys

We report Ferritic-martensitic (FM) steels for in-vessel components in proposed fusion reactors are expected to suffer from high levels of displacement damage and helium (He) generation by neutron transmutation. However, a thorough understanding of the He synergistic effects on the cavity swelling in FM steels is still not well established. To gain fundamental insights into the He effect on cavity swelling, high purity Fe and Fe-10 wt.% Cr ferritic model alloys were irradiated with 8 MeV Ni ions and co-implanted He ions at 400 to 550 °C up to 30 displacements per atom (dpa) with He implantation rates of 0.1, 10 and 50 appm He/dpa. The current study focuses on the 50 appm He/dpa, 500 °C behavior vs. 0.1 and 10 appm He/dpa. Irradiation-induced defects, including cavities, dislocation loops, and dislocation networks were characterized using transmission electron microscopy (TEM). In the grain interior, a bimodal cavity size distribution was observed in the 10 and 50 appm He/dpa samples, but not for 0.1 appm He/dpa. Cavity swelling was maximized at intermediate He implantation rates of ~10 appm He/dpa for both ion-irradiated Fe and Fe-10Cr alloys. The cavity swelling behavior as a function of He implantation rate appears to be controlled by the He/dpa-dependent variation of cavity sink strengths. Treating small bubbles as biased sinks for interstitial absorption can significantly increase the ratio of biased to unbiased sink strengths (Q) and results in maximized cavity swelling for a Q ratio close to one.

36 MATERIALS SCIENCE↗
DOE OSTI2021

A laser inverse Compton scattering diagnostic to study runaway electron dynamics during tokamak disruptions

Both technical and simulation progress has been made toward implementing a laser inverse Compton scattering diagnostic on the DIII-D tokamak. This is a novel diagnostic which uses a short (sub-ns) pulse laser together with a fast (sub-ns) gated x-ray detector to measure relativistic electrons inside a tokamak plasma. The study of these relativistic electrons is important to make sure they can be removed from any tokamak-based power reactor, as they could damage vessel walls and cooling lines. This diagnostic is completely new for magnetic fusion research and has never been implemented before on a tokamak. It is found that good signals and radial profiles could be obtained with an optimized high-energy (7 Joule) laser. Using a standard off-the-shelf Nd:YAG (0.7 Joule) laser would give some signals above the noise but no ability to measure the relativistic electron radial profile.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
NASA NTRS1973

Reactivity change in a fast-spectrum space power reactor due to a 328-meter-per-second (1075-ft/sec) impact

Application of nuclear reactors in space will present operational problems. One such problem is the possibility of an earth impact at velocities in excess of 305 m/sec (1000 ft/sec). This report shows the results of an impact against concrete at 328 m/sec (1075 ft/sec) and examines the deformed core to estimate the range of activity inserted as a result of the impact. The results of this examination are that the deformation of the reactor core within the containment vessel left only an estimated 2.7 percent void in the core and that the reactivity inserted due to this impact deformation could be from 4.0 to 10.25 dollars.

Peoples, J. A., Jr.↗
NASA NTRS2007

Gravity Scaling of a Power Reactor Water Shield

A similarity analysis on a water-based reactor shield examined the effect of gravity on free convection between a reactor shield inner and outer vessel boundaries. Two approaches established similarity between operation on the Earth and the Moon: 1) direct scaling of Rayleigh number equating gravity-surface heat flux products, 2) temperature difference between the wall and thermal boundary layer held constant. Nusselt number for natural convection (laminar and turbulent) is assumed of form Nu = CRa(sup n).

Reid, Robert S.↗
DOE OSTI2020

Thermal reactor systems and methods

An exemplary embodiment of the present invention provides a reactor system comprising: a reaction vessel comprising a reactant, a heat transfer fluid and a first reaction product, wherein the heat transfer fluid has a greater density than the first reaction product such that at least a portion of the first reaction product floats on a surface of the heat transfer fluid; a first outlet positioned at a surface level of the first reaction product, the first outlet configured to output a first outlet flow comprising at least a portion of the first reaction product and at least a portion of the heat transfer fluid; wherein the heat transfer fluid is configured to provide thermal energy to the reactant in the reaction vessel to form the first reaction product.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

Electrochemically modulated molten salt reactor

An electrochemically modulated molten salt reactor (EMMSR) that contains a vessel and a power source. The vessel houses a fuel salt, at least a portion of a neutron moderator, and at least a portion of an insulator. The fuel salt includes enough dissolved fissile isotopes to cause continued self-sustaining fission reactions during the operation of the EMMSR. The neutron moderator is configured to slow down fast neutrons produced by the dissolved fissile isotopes. The insulator is configured to electrically isolate the neutron moderator from the vessel. The power source has a positive potential and a negative potential. The positive potential is received by the neutron moderator and the negative potential is received by the vessel.

Jiang, Junhua↗
DOE OSTI2020

Simulation of DWPF Chemical Process Cell Using a Reaction Calorimeter

The Defense Waste Processing Facility (DWPF) converts liquid radioactive waste into glass using a Sludge Receipt Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) in addition to other processes. The SRAT and SME processes can be simulated on a small-scale using the Mettler-Toledo Reaction Calorimeter (RC1). This equipment is able to automate the execution and monitoring of experiments involving nonradioactive chemical simulants. During the chemical processing of sludge in the SRAT process, foam begins to form and can foam over the reaction vessel. The current antifoam used at DWPF creates a risk of increased flammability and has a limited pH range. Usually, mercury is circulated back into the reaction vessel during the SRAT process, making it difficult to remove. To combat foam over, novel antifoams will be tested during the boil-up process using the RC1. To improve mercury extraction, 3D-printed coalescer columns will be tested within the Mercury Water Wash Tank (MWWT) of the RC1. Testing of Simulant Sludge Batch 6 (SB6) was performed to ensure: the sludge would form a stable foam at boiling during full SRAT/SME runs; the antifoam would reduce the amount of foam within the vessel The sludge and a Precipitate Reactor Feed Tank (PRFT) simulant with monosodium titanate (MST) were mixed separately, combined, and then heated to boiling. As shown below, the antifoam used was able to eliminate the foam within the vessel, as expected. The FormLabs Form2 3D Printer can be used to create roughly 3-inch tall and 1-inch diameter columns from clear and ceramic resins. The columns will utilize chemical and physical forces to draw the mercury out of a condensed offgas stream leaving the RC1 reaction vessel. These columns will be tested during full SRAT/SME runs on the basis of how well they are able to collect elemental mercury. In preparation for full-scale runs, the RC1 equipment was assembled onto a custom table and tested with water runs. The software used to control automation, iControl, has a user-friendly interface that allows the experiment to be altered as needed. An example of the data recorded during an experimental run is represented in Figure 1 below. Water runs were utilized as a way to prove if the RC1 system was functioning in the following areas: Boil-up rates, Chemical additions using pumps, Temperature readings, Balance readings, Probe readings, Offgas system, Agitation. Future Plans: Five novel antifoams will be tested during full SRAT/SME runs. The most effective antifoam will be utilized in DWPF's Chemical Process Cell. The design of the Mercury Extraction Coalescer can be further manipulated to increase the extraction of not only elemental mercury, but other heavy metals in environmental remediation processes as well.

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

In-situ Condition Monitoring of Components in Small Modular Reactors Using Process and Electrical Signature Analysis. Final report, volume 1. Development of experimental flow control loop, data analysis and plant monitoring

The research and development under this project was focused on the following three major objectives: Objective 1: Identification of critical in-vessel SMR components for remote monitoring and development of their low-order dynamic models, along with a simulation model of an integral pressurized water reactor (iPWR). Objective 2: Development of an experimental flow control loop with motor-driven valves and pumps, incorporating data acquisition and on-line monitoring interface. Objective 3: Development of stationary and transient signal processing methods for electrical signatures, machinery vibration, and for characterizing process variables for equipment monitoring. This objective includes the development of a data analysis toolbox. The following is a summary of the technical accomplishments under this project: - A detailed literature review of various SMR types and electrical signature analysis of motor-driven systems was completed. A bibliography of literature is provided at the end of this report. Assistance was provided by ORNL in identifying some key references. - A review of literature on pump-motor modeling and digital signal processing methods was performed. - An existing flow control loop was upgraded with new instrumentation, data acquisition hardware and software. The upgrading of the experimental loop included the installation of a new submersible pump driven by a three-phase induction motor. All the sensors were calibrated before full-scale experimental runs were performed. - MATLAB-Simulink model of a three-phase induction motor and pump system was completed. The model was used to simulate normal operation and fault conditions in the motor-pump system, and to identify changes in the electrical signatures. - A simulation model of an integral PWR (iPWR) was updated and the MATLAB-Simulink model was validated for known transients. The pump-motor model was interfaced with the iPWR model for testing the impact of primary flow perturbations (upsets) on plant parameters and the pump electrical signatures. Additionally, the reactor simulation is being used to generate normal operation data and data with instrumentation faults and process anomalies. A frequency controller was interfaced with the motor power supply in order to vary the electrical supply frequency. The experimental flow control loop was used to generate operational data under varying motor performance characteristics. Coolant leakage events were simulated by varying the bypass loop flow rate. The accuracy of motor power calculation was improved by incorporating the power factor, computed from motor current and voltage in each phase of the induction motor.- A variety of experimental runs were made for steady-state and transient pump operating conditions. Process, vibration, and electrical signatures were measured using a submersible pump with variable supply frequency. High correlation was seen between motor current and pump discharge pressure signal; similar high correlation was exhibited between pump motor power and flow rate. Wide-band analysis indicated high coherence (in the frequency domain) between motor current and vibration signals. - Wide-band operational data from a PWR were acquired from AMS Corporation and used to develop time-series models, and to estimate signal spectrum and sensor time constant. All the data were from different pressure transmitters in the system, including primary and secondary loops. These signals were pre-processed using the wavelet transform for filtering both low-frequency and high-frequency bands. This technique of signal pre-processing provides minimum distortion of the data, and results in a more optimal estimation of time constants of plant sensors using time-series modeling techniques.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

SPC-3185 Rev 5 Guard Vessel Build to Print Fabrication Specification

Idaho National Laboratory (INL) is developing a microreactor to produce electrical power utilizing a small nuclear core and Stirling engines under the Microreactor Application Research Validation and Evaluation (MARVEL) program. This procurement specification defines the requirements for fabrication of the Guard Vessel.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2017

Light Water Reactor Sustainability Program, U.S. Efforts in Support of Examinations at Fukushima Daiichi-2017 Evaluations

Although the accident signatures from each unit at the Fukushima Daiichi Nuclear Power Station (NPS) [Daiichi] differ, much is not known about the end-state of core materials within these units. Some of this uncertainty can be attributed to a lack of information related to cooling system operation and cooling water injection. There is also uncertainty in our understanding of phenomena affecting: a) in-vessel core damage progression during severe accidents in boiling water reactors (BWRs), and b) accident progression after vessel failure (ex-vessel progression) for BWRs and Pressurized Water Reactors (PWRs). These uncertainties arise due to limited full scale prototypic data. Similar to what occurred after the accident at Three Mile Island Unit 2, these Daiichi units offer the international community a means to reduce such uncertainties by obtaining prototypic data from multiple full-scale BWR severe accidents. Information obtained from Daiichi is required to inform Decontamination and Decommissioning activities, improving the ability of the Tokyo Electric Power Company Holdings, Incorporated (TEPCO Holdings) to characterize potential hazards and to ensure the safety of workers involved with cleanup activities. This document, which has been updated to include FY2017 information, summarizes results from U.S. efforts to use information obtained by TEPCO Holdings to enhance the safety of existing and future nuclear power plant designs. This effort, which was initiated in 2014 by the Reactor Safety Technologies Pathway of the Department of Energy Office of Nuclear Energy Light Water Reactor (LWR) Sustainability Program, consists of a group of U.S. experts in LWR safety and plant operations that have identified examination needs and are evaluating TEPCO Holdings information from Daiichi that address these needs. Each year, annual reports include examples demonstrating that significant safety insights are being obtained in the areas of component performance, fission product release and transport, debris end-state location, and combustible gas generation and transport. In addition to reducing uncertainties related to severe accident modeling progression, these insights are being used to update guidance for severe accident prevention, mitigation, and emergency planning. Furthermore, reduced uncertainties in modeling the events at Daiichi will improve the realism of reactor safety evaluations and inform future D&D activities by improving the capability for characterizing potential hazards to workers involved with cleanup activities. Highlights in this FY2017 report include new insights with respect to the forces required to produce the observed Daiichi Unit 1 (1F1) shield plug endstate, the observed leakage from 1F1 components, and the amount of combustible gas generation required to produce the observed explosions in Daiichi Units 3 and 4 (1F3 and 1F4). This report contains an appendix with a list of examination needs that was updated after U.S. experts reviewed recently obtained information from examinations at Daiichi. Additional details for higher priority, near-term, examination activities are also provided. This report also includes an appendix with a description of an updated website that has been reformatted to better assist U.S. experts by providing information in an archived retrievable location, as well as an appendix summarizing U.S. Forensics activities to host the TMI-2 Knowledge Transfer and Relevance to Fukushima Meeting that was held in Idaho Falls, ID, on October 10-14, 2016.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Robotics Technologies for Decommissioning of Fukushima Daiichi - 17077

The Fukushima Daiichi Nuclear Power Station (NPS) of Tokyo Electric Power Company (TEPCO) was the site of a severe nuclear accident after the 2011 Off the Pacific coast of Tohoku Earthquake and subsequent tsunami. Extensive structural damage occurred in the NPS facilities, including such environments as the reactor buildings and the primary containment vessels (PCVs). These are high radiation areas, having underwater areas, narrow spaces and so on. It is necessary to determine the extent of contamination and the location of debris to plan decommissioning operations. Survey targets in the past few years have been focused on inspections of the basement of the reactor buildings and the inside of the PCVs. Human operators cannot access these environments and work in them directly. Therefore, remotely operated robots and measurement technologies are needed to carry out various operations. For underwater surveys, we have developed a submersible crawling swimming robot that can avoid obstacles and inspect wide areas as it moves around. A shape-changing robot has been developed for passing through narrow spaces and crawling on rough floors. An ultrasonic imaging method combined with convex scanning and a divergent beam has been proposed to visualize environments in turbid water. In this paper, we describe details of the two survey robots and the underwater imaging method. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Endurance test of helium circulator for HTR-PM - 18531

High-temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM) will be built as the world's first pebble-bed type modular HTGR commercial demonstration plant according to Chinese National science and technology program. HTR-PM will be expected to begin generating electricity in the end of 2017 in Shidao bay of China. Helium circulator is one of critical component of HTR-PM, which circulates reactor coolant through the reactor core and steam generators. The helium circulator was designed to be a vertical layout and its rotor was supported by active magnet bearings (AMBs). The whole circulator including motor and cooler was sealed in a pressure vessel. Since it is a large mechanical and electrical equipment and the design is totally advanced integrated with many new technologies, experiment study of helium circulator would be significant for safe and reliable operation of HTR-PM plant in the future. For the lack of engineering experience of this equipment, it was necessary to manufacture a prototype for tests. Meanwhile, Engineering Test Facility- Helium Circulator. ETF-HC was built in Institute of Nuclear and New Energy Technology (INET) for the tests. The prototype is completely same as the real machine for the plant. After prototype fabrication, nitrogen test was firstly conducted in the factory by the manufacturer, designer and users. After successful factory test, the prototype was delivered to INET to be tested in ETF-HC. Although nitrogen simulation test had been completed successfully, helium endurance test has been continually carried out in ETF-HC for one and half years because of several sealing problems caused by helium. After solving all kinds of problem in the field, helium circulator run continuously 50 hours under rated condition. Furthermore, 50 service cycles test was conducted in ETF-HC. More than 500 operation hours was accumulated during endurance test. Throughout the final performance test, a conclusion can be made that the helium circulator performance met the design requirements and its feasibility, availability and reliability was verified. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Application of Combined Sustained and Cyclic Loading Test Results to Alloy 617 Elevated Temperature Design Criteria

Alloy 617 is a reference structural material for very high temperature components of advanced-gas cooled reactors with outlet temperatures in the range of 900-950°C . In order for designers to be able to use Alloy 617 for these high temperature components, Alloy 617 has to be approved for use in Section III (the nuclear section) of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. A plan has been developed to submit a draft code for Alloy 617 to ASME Section III by 2015. However, the current rules in Subsection NH for the evaluation of strain limits and creep-fatigue damage using simplified methods based on elastic analysis have been deemed inappropriate for Alloy 617 at temperatures above 1200°F (650°C). The rationale for this exclusion is that at higher temperatures it is not feasible to decouple plasticity and creep deformation, which is the basis for the current simplified rules. This temperature, 1200 °F, is well below the temperature range of interest for this material in High Temperature Gas Cooled Reactor (HTGR) applications. The only current alternative is, thus, a full inelastic analysis which requires sophisticated material models which have been formulated but not yet verified. To address this issue, proposed code rules have been developed which are based on the use of elastic-perfectly plastic (EPP) analysis methods and which are expected to be applicable to very high temperatures.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Water Sealing With NOH{sub 2}O on the Fukushima Site - 17477

The Fukushima Daiichi nuclear site is still recovering from the natural disaster of March 11, 2011. Much of the recent recovery work has been focused on mitigating ground water intrusion into the radioactively contaminated basements of the reactor and turbine buildings of units 1 - 4. Other recovery efforts involve potentially sealing the vent lines in the primary containment vessels of units 1 - 3 to maintain a flooded vessel for core cooling and shielding. This paper details preliminary test results of a proprietary grout - NOH{sub 2}O{sup TM} - in sealing/plugging those vent lines to allow flooding of the primary containment vessel. Several partial and full-scale tests were conducted on primary containment vessel vent line mock-ups between March and October 2015 to ascertain the effectiveness of NOH{sub 2}O in sealing vent lines that have flowing water, which is problematic for conventional cement-based grouts that require relatively protracted set times. In contrast, NOH{sub 2}O can be used with a proprietary accelerant to provide instant set. The other key advantage of NOH{sub 2}O is its ability to penetrate even the smallest of cracks and porous soils, following and moving with the water flow and creating a continuous, impermeable barrier. It is also extremely flexible once set, conforming to its enclosure as upstream water head increases. Finally, the material's extremely high radiation resistance and longevity make it an ideal candidate for a permanent solution in high rad environments. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗