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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 847 records · Page 47
All sourcesNASA NTRSDOE OSTI
NASA NTRS1971

Mini gas-core propulsion concept.

A propulsion concept is presented that can achieve specific impulses of 1000 to 2000 seconds for thrust levels on the order of 100 pounds (445 N) and maintain a small (1.22 m) reactor diameter for a total power-plant weight of approximately 10,000 kg. The concept combines a small (0.61 m) diameter spherical gas core centrally located, followed by a moderator (BeO-15 cm), surrounded by a driver fuel region and contained in a pressure vessel (ID of 1.22 m). Best results (i.e., highest ratio of cavity power to total power) occur with U-233 isotope as the fuel in both the cavity and the driver, although U-235 can also be used. Concept may also have application in the testing of a gas-core reactor and MHD devices and in high temperature-pressure materials research.

Hyland, R. E.↗
DOE OSTI2020

Example Evaluation of a Representative Heat Pipe Test Article Design for Structural Acceptability using ASME Design Rules

This report describes a design analysis of a test article mimicking a core block of a heat-pipe mi-croreactor using the ASME Boiler & Pressure Vessel Code Section III, Division 5, Subsection HB, Subpart B rules. These rules cover the design and construction of elevated temperature nu-clear reactor structural components, like the core block of a heat-pipe microreactor emulated by the test article. The purposes of the report are to: (1) provide a step-by-step example of applying the ASME rules to a reasonable core block geometry and corresponding loading conditions, (2) to provide design feedback on the test article, and (3) to evaluate the ASME rules for use in the design of complex heat-pipe microreactor components. The test article design has a relatively short ASME design life, limited by high thermal stresses. The current ASME rules are adequate for evaluating the core block component designs, though they could be optimized to increase the design efficiency and reduce the amount of designer/analyst time required to execute the design process. The report covers both the current, 2019 edition ASME rules and modified, currently unapproved, rules designed to improve the design process for components with complex geometries and high secondary stresses, like the core block test article.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

ECAR-7547 Rev 0 MARVEL Intermediate Heat Exchanger Liner ASME Analysis

This ECAR analyzes the Secondary Coolant System liner, otherwise known as the Intermediate Heat Exchanger (IHX) Liner, as an ASME Section VIII, Division 1 pressure vessel. This analysis is only necessary in the event of a failure of the primary coolant boundary, in which case the IHX liner acts as a secondary confinement boundary to prevent primary coolant leakage. The liner main body or vessel is analyzed, as well as the liner flange ring.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2021

The high-power helicon program at DIII-D: gearing up for first experiments

Helicon current drive, also called fast wave current drive in the lower hybrid range of frequencies, has long been regarded as a promising current drive tool for reactor grade plasmas. A newly installed MW-level system at DIII-D will be the first test of this technology in reactor-relevant plasmas, in the sense that full single-pass absorption is expected. A 30-module traveling wave antenna has been installed and optimized in-vessel in early 2020. The linear electromagnetic characteristics of the unloaded module array have been extensively tested both on the bench and in the vessel at instrumentation power levels. Excellent performance has been achieved, ~2% reflected power and ~1.5% dissipated power per module in air, in a 10 MHz band around 476 MHz. Stripline feeds on both ends of the antenna allow either co or counter current drive. The installation of a 1.2 MW klystron and associated high-power electronics was completed in Fall 2020. Commissioning of the antenna is ongoing. An important goal of this experiment is to validate the helicon current drive physics basis using an extensive set of new and upgraded diagnostics.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

The MELTSPREAD Code for Modeling of Ex-Vessel Core Debris Spreading Behavior, Code Manual – Version3-beta

MELTSPREAD3 is a transient one-dimensional computer code that has been developed to predict the gravity-driven flow and freezing behavior of molten reactor core materials (corium) in containment geometries. Predictions can be made for corium flowing across surfaces under either dry or wet cavity conditions. The spreading surfaces that can be selected are steel, concrete, a user-specified material (e.g., a ceramic), or an arbitrary combination thereof. The corium can have a wide range of compositions of reactor core materials that includes distinct oxide phases (predominantly Zr, and steel oxides) plus metallic phases (predominantly Zr and steel). The code requires input that describes the containment geometry, melt “pour” conditions, and cavity atmospheric conditions (i.e., pressure, temperature, and cavity flooding information). For cases in which the cavity contains a preexisting water layer at the time of RPV failure, melt jet breakup and particle bed formation can be calculated mechanistically given the time-dependent melt pour conditions (input data) as well as the heatup and boiloff of water in the melt impingement zone (calculated). For core debris impacting either the containment floor or previously spread material, the code calculates the transient hydrodynamics and heat transfer which determine the spreading and freezing behavior of the melt. The code predicts conditions at the end of the spreading stage, including melt relocation distance, depth and material composition profiles, substrate ablation profile, and wall heatup. Code output can be used as input to other models such as CORQUENCH that evaluate long term core-concrete interaction behavior following the transient spreading stage. MELTSPREAD3 was originally developed to investigate BWR Mark I liner vulnerability, but has been substantially upgraded and applied to other reactor designs (e.g., the EPR), and more recently to the plant accidents at Fukushima Daiichi. The most recent round of improvements that are documented in this report have been specifically implemented to support industry in developing Severe Accident Water Management (SAWM) strategies for Boiling Water Reactors.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

VERA User's Guide for Ex-core Applications

The Virtual Environment for Reactor Applications, or VERA, allows users to set up models to calculate time-dependent and fully coupled solutions for ex-core quantities of interest such as vessel and coupon fluence, and detector responses for multiple statepoints and cycles. MPACT and COBRA-TF together perform in-core transport calculations with temperature feedback while Shift performs the fluence and detector response calculations in the ex-core region. The in-core region is modeled using VERA’s native input format and the ex-core region is defined using Shift’s general geometry package, also known as Omnibus General Geometry. Fixed source ex-core calculations with Shift can be run in forward mode without advanced variance reduction (VR) methods, or with Consistent Adjoint Driven Importance Sampling (CADIS), which is an automated VR method. This document serves as a guide for setting up inputs, running ex-core calculations and post-processing the results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

SU-E-T-112: Dose Distribution of Praseodymium-142 Microspheres in Microcapillary Using Radiochromic Film Dosimetry and Applications in Hepatocellular Carcinoma Microsphere Brachytherapy

Purpose: This work verified simulations of beta-minus emitter Praseodymium-142 (Pr-142) for microsphere brachytherapy by performing absolute dose measurements for Pr 142 microspheres in a microcapillary as a simplified model for a single blood vessel for the treatment of Hepatocellular Carcinoma (HCC). Methods: Pr-142 microspheres (mass: 0.169g, average diameter: 29.7±3.9μm) were activated by thermal neutron activation at the University of Missouri Research Reactor. Experimental setup consisted of a microsphere solution (initial activity 36.6mCi in 0.1ml of sterile water) within a glass microcapillary (internal and external diameter: 305μm and 453μm, respectively) placed for 51h in a custom made Gammex Solid Water™ phantom. GAFCHROMIC™ EBT2 film calibrated with a 6MeV electron beam was used to access the dose fall-off of microspheres. The microcapillary was modeled in MCNPX2.6 in order to compare with experiments. Results: The radial dose fall-off on the transverse plane due to scatter and attenuation in the solid water phantom was analyzed using ImageJ for both film and MCNPX2.6 simulations. Isodose analysis showed close agreement among the methods used, i.e. measurements and simulations agree within 3.9% for doses below 1600cGy. Experimental and simulated doses obtained at 0.5 cm radially from the source were 1547cGy and 1610cGy respectively. Discrepancies for points close to the microcapillary surface were observed between MCNPX2.6 and measurements due to film saturation for high doses. Dose due to Pr-142 3.7% gamma emission was below the threshold of detection for the film. Conclusion: A detailed dosimetric study was performed for Pr-142 glass microspheres within a single microcapillary. MCNPX2.6 simulations were verified by means of direct measurement. Based on these results, Pr-142 appears to be a viable choice of radionuclide for treating HCC.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

Mechanisms Engineering Test Loop (METL) Experimenter's Guide

The Mechanisms Engineering Test Loop (METL) was built to streamline and accelerate the in-sodium testing of systems and components under conditions that simulate a sodium-cooled fast reactor pool environment. The METL team at Argonne National Laboratory (ANL) can assist experimenters in achieving their technical goals by providing liquid-metal expertise and access to infrastructure required for most alkali metal related research. This document offers a brief overview of METL and provides a basic design guide for researchers interested in conducting research at the facility. Additional information regarding the history and operations of METL can be found in §6.1. Furthermore, high resolution images found in this document as well as CAD files of aforementioned vessels can be provided upon request.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Finite Element Based Full-Life Cyclic Stress Analysis of 316 Grade Nuclear Reactor Stainless Steel Under Constant, Variable, and Random Fatigue Loading

Although S~N curve-based approaches are widely followed for fatigue evaluation of nuclear reactor components and other safety critical structural systems, there is a chance of large uncertainty in estimated fatigue lives. This uncertainty may be reduced by using a more mechanistic approach such as physics based three-dimensional (3D) finite element (FE) methods. In a recent paper (Barua et al., 2018, ASME J. Pressure Vessel Technol., 140(1), p. 011403), a fully mechanistic fatigue modeling approach which is based on time-dependent stress–strain evolution of material over the entire fatigue life was presented. Based on this approach, FE-based cyclic stress analysis was performed on 316 nuclear grade reactor stainless steel (SS) fatigue specimens, subjected to constant, variable, and random amplitude loading, for their entire fatigue lives. The simulated results are found to be in good agreement with experimental observation. An elastic-plastic analysis of a pressurized water reactor (PWR) surge line (SL) pipe under idealistic fatigue loading condition was performed and compared with experimental results.

36 MATERIALS SCIENCE↗
DOE OSTI2015

High internal inductance for steady-state operation in ITER and a reactor

Increased confinement and ideal stability limits at relatively high values of the internal inductance (ℓ i ) have enabled an attractive scenario for steady-state tokamak operation to be demonstrated in DIII-D. We achieved normalized plasma pressure in the range appropriate for a reactor in high elongation and triangularity double-null divertor discharges with beta(N) approximate to 5 at ℓ i approximate to 1.3, near the ideal n = 1 kink stability limit calculated without the effect of a stabilizing vacuum vessel wall, with the ideal-wall limit still higher at β N > 5.5. Moreover, confinement is above the H-mode level with H 98(y,H- 2) approximate to 1.8. At q 95 approximate to 7.5, the current is overdriven, with bootstrap current fraction f BS approximate to 0.8, noninductive current fraction f NI > 1 and negative surface voltage. For ITER (which has a single-null divertor shape), operation at ℓ i approximate to 1 is a promising option with f BS approximate to 0.5 and the remaining current driven externally near the axis where the electron cyclotron current drive efficiency is high. This scenario has been tested in the ITER shape in DIII-D at q 95 = 4.8, so far reaching f NI = 0.7 and f BS = 0.4 at beta N approximate to 3.5 with performance appropriate for the ITER Q=5 mission, H 89 beta N /q 2 95 approximate to 0.3. Finally, modeling studies explored how increased current drive power for DIII-D could be applied to maintain a stationary, fully noninductive high ℓ i discharge. Stable solutions in the double-null shape are found without the vacuum vessel wall at beta N = 4, ℓ i = 1.07 and f BS = 0.5, and at beta N = 5 with the vacuum vessel wall.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Magnetic Bearing Pumps for Molten Salt Fusion Energy Devices

The Commonwealth Fusion Systems (CFS) ARC reactor will employ a novel liquid immersion blanket design that uses a molten lithium salt as both the tritium-breeding material and the vacuum vessel coolant. fluoride lithium beryllium (FLiBe) (a lithium fluoride–beryllium difluoride mixture) is currently the leading salt candidate. Its low-Z components provide good moderation, it has adequate heat transfer properties and low electrical conductivity, and it has good neutron multiplication properties, which further enhance tritium production.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
NASA NTRS1972

The external gamma radiation environment from the Kiwi Phoebus, and Pewee reactors

During the past few years, ground tests of high-powered propulsion-prototype reactors have provided several opportunities to observe the external radiation environment. Reactor tests have been conducted in free air and inside of open well shields. Measurements were taken over distances ranging from contact with the pressure vessel out to greater than 5000' both during operation and after shutdown. Some measurements characteristic of each of the systems are presented and compared with results of calculations.

Malenfant, R. E.↗
DOE OSTI2019

Gear Test Assembly - Status Report For FY2019

The final stages of pre-sodium commissioning were completed for the Gear Test Assembly (GTA) in late 2018 and all the components were moved to building 308 for final preparations for insertion into the Mechanisms Engineering Test Loop (METL). The GTA was fully assembled in the experimental test assembly workstation and all supporting electrical and hardware systems were installed on the METL Mezzanine. The GTA was installed in Test Vessel 1 after performing a final systems check, and the vessel and article were raised to operating temperature and held for two weeks to bake-out any moisture present in the components. Test Vessel 1 was filled with sodium and a purification process was completed using the cold trap system. When the sodium was sufficiently clean, Test Vessel 1 was isolated from the rest of METL, and experimental operation of the GTA commenced. The experimental operations consisted of rotating two shafts using Parker servomotors to apply a prototypic torque of a fuel handling mechanism removing and inserting a sodium fast reactor fuel assembly. This removing and inserting motion was repeated until the system seized, indicating component degradation of either the gears or the bearings in the test gearbox. A total of 19,600 cycles were completed, corresponding to the removal and insertion of 9,800 fuel assemblies. The system was shut down and then prepared for removal. The Flexicask System was successfully used to remove the GTA from Test Vessel 1 while keeping both vessel and article inert during the process. The GTA was installed in the Carbonation System, where the process of slowly reacting the sodium away using moist CO 2 has been started. Once the GTA is fully cleaned, it will be disassembled and the gears will undergo the same NDE process performed before testing to quantify the wear that occurred. A report on this, along with the data analysis from the experimental operations will be submitted in late 2019.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Mechanisms Engineering Test Loop - Flexicask Design and Testing Report

This report documents the design development and testing of the Flexicask system for the Mechanisms Engineering Test Loop (METL). Currently, METL has completed Phase I construction and has been operating for over a year. The Flexicask system (Figure 1) was designed to allow for the removal and insertion of test articles into a METL test vessel while maintaining an acceptable inert atmosphere to minimize contamination of the METL sodium. This report discusses the design development of the Flexicask system, the commissioning and testing of the system, and the initial use of the Flexicask system to remove the Gear Test Assembly (GTA) from the METL Test Vessel 1.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Status Update for the Bearing Test Article Design Development – FY2025

Argonne National Laboratory’s (ANL) Mechanism Engineering Test Loop (METL) team has completed the design of a new Bearing Test Article (BTA) to test mechanical bearings for sodium fast reactor (SFR) mechanism applications. Recognition of the need for further bearing testing arose due to failure of rolling element bearings during testing of components for in-vessel fuel handling machines (FHMs) conducted at ANL in the METL facility. While the initial concept of the BTA arose from a need for bearing testing for FHM applications, the BTA design is configurable to test bearings of various sizes, types, and materials submerged in liquid sodium. Requirements for loading, shaft speed, temperature, and bearing size range were chosen based on requirements for SFR fuel handling. Modular components in the test bearing housings and shaft sleeves allow the design to accommodate a wide range of bearing sizes without modification to the larger assembly. Loading will be applied with screw jacks on the air side of the test article that will be adjusted to apply radial, thrust, or combined loading conditions. Applied force will be measured by load cells in-line with the loading mechanisms. In FY25 we have completed the design of the test article that will be fabricated and undergo testing at the METL facility. Drawing packages for the fabrication of the test article have been drafted and are being reviewed to add part tolerances and we have started ordering off-the-shelf components. The report will summarize the development of the Bearing Test Article design starting with an overview of the entire test article, then a more detailed examination of the bearing test section, loading mechanisms, test article ports and seals, and support structures, including analysis that went into significant design choices.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Thermal-Hydraulic Response of the TREAT Multi-Vessel Static Environment Test Vehicle

As part of the resumption of transient testing in the United States, multiple test vehicles are currently under design, which are ideally suited for achieving various test environments. Of these, the multi- static environment rodlet transient test apparatus (SERTTA) is the most mature design. The test vehicle is designed for testing up to four isolated specimens under a variety of conditions ranging from gas, steam, water, and liquid sodium. During proposed transient events such as reactivity initiated accidents (RIA) and loss of coolant accidents (LOCA) in reactors, prototypic environments dictate a wide range of thermal-hydraulic boundary conditions. The purpose of the this work is to provide the thermal hydraulic response of the Multi-SERTTA test vehicle subjected to fuel specimen energy depositions representative of pressurized water reactor (PWR) conditions. Calculations are performed for a RIA-type power pulse in a 10 cm long UO2-zircaloy rodlet immersed in a water environment in Multi-SERTTA. The relatively small volume available in the Multi-SERTTA causes the vessel pressure to be very sensitive to total energy deposition. The effects of total energy deposition and free gas volume in the pressure vessel are presented. The design provides additional flexibility of environment conditions through use of the expansion tank volume. The results show the influence of the vehicle geometric and safety constraints and provide indication of the expected thermal-hydraulic conditions relative to typical RIA transient events.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2015

Mars Propellant Production with Ionic Liquids Project

This project seeks to develop a single vessel for carbon dioxide (CO2) capture and electrolysis for in situ Mars propellant production by eliminating several steps of CO2 processing, two cryocoolers, a high temperature reactor, a recycle pump, and a water condenser; thus greatly reducing mass, volume, and power.

Falker, John↗
DOE OSTI2025

Experimental demonstration of transient CHI start-up using a floating biased electrode configuration

Transient coaxial helicity injection (CHI) current start-up in a spherical tokamak (ST) has been demonstrated for the first time using a reactor-relevant floating single biased (FSB) electrode configuration on QUEST. Unlike previous applications of CHI on HIT-II and on NSTX in which the entire ST vessel was split in two using two toroidal ceramic rings as vacuum breaks, in the FSB electrode configuration on QUEST, only the lower outer divertor plate is electrically insulated from the rest of the vessel. This is the CHI injector region. A CHI discharge is generated by driving current, using an external power supply, along magnetic field lines that connect the inner and outer divertor plates in the CHI injector region. The CHI plasma evolution and closed flux formation on QUEST are very similar to those on NSTX and have generated toroidal currents of 150 kA with 100 kA of current flowing on closed flux surfaces. An alternate, force-free model approach has been used to derive the CHI scaling relations. The resulting CHI scaling relations from the model are consistent with the conventional CHI scaling relations, which state that the CHI generated toroidal current is proportional to the CHI injector flux and the CHI injector current is inversely proportional to the toroidal field. These results from QUEST bode well for the application of CHI in a new generation of compact high field STs and low aspect ratio tokamaks in which the space for the central solenoid is very restricted.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗