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At least 937 records · Page 52
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Progress in Fast Modular Reactor Conceptual Design

The Fast Modular Reactor (FMR) is a 100-MW(thermal) gas-cooled fast reactor being developed by General Atomics Electromagnetic System with the goal of developing a FMR for flexible and dispatchable power to the U.S. electricity market in the mid-2030s. The conceptual design aims to develop and verify simplified design features. These include an inert helium gas coolant, pellet-loaded fuel rods, installations with air cooling as ultimate heat sink, and small and passive heat removal systems. The goal is to ensure the development of a safe, maintainable, cost-effective, and distributed nuclear energy-generating station. The baseline technologies selected to achieve this goal are a helium coolant that is an inert gas with no chemical reaction with structural components, not activated, single phase, enabling high-temperature operation and a high thermal efficiency Brayton cycle; conventional uranium dioxide (UO 2 ) fuel, which is the most widely used and well-known fuel material, capable of high burnup (100 MWd/kg) and a long fuel life; and silicon carbide composite (SiGA®) cladding and internal structures that are chemically inert in the helium environment, exceptionally radiation tolerant, and being derisked by accident tolerant fuel technology development. Further, the reactor was specifically designed with passive safety features, including high-temperature in-core materials and a reactor vessel cooling system consisting of cooling panels of naturally circulating water. The passive safety of the core was confirmed for the depressurized loss-of–forced cooling accident, which showed the peak cladding temperature at ~1600°C during the transient, which is below the current design limit of 1800°C. The conceptual design of the FMR has been conducted for the reactor system, vessel system, generator and turbomachine, instrumentation and control, residual heat removal system, plant service system, and containment, as well as pre-application licensing documents.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Fabrication and Preliminary Demonstration of Microwave Resonant Cavity Transducer Performance

We are investigating a microwave cavity-based transducer for in-core high-temperature fluid flow sensing in molten salt cooled reactors (MSCR) and sodium fast reactors (SFR). This sensor is a hollow metallic cylindrical cavity, which can be fabricated from stainless steel, and as such is expected to be resilient to radiation, high temperature and corrosive environment of MSCR and SFR. The principle of sensing consists of making one wall of the cylindrical cavity flexible enough so that dynamic pressure, which is proportional to fluid velocity, will cause membrane deflection. Membrane deflection causes cavity volume change, which leads to a shift in the resonant frequency. We have developed an initial design for proof-of-principle testing of the flow sensor performance in microwave K-band. A cylindrical resonator prototype was fabricated from brass for initial tests in water. The cavity size is matched to the flange of a standard WR-42 waveguide. Microwave field is coupled into the resonant cavity through a subwavelength-size aperture. A test article was developed, consisting of a piping Tee with bulkhead WR-42 microwave waveguide installed in leak-proof design. In the test article, the cylindrical cavity is positioned in the center of the pipe. A microwave waveguide circulator was installed in the setup to suppress microwave reflections at the cavity entrance. Preliminary spectral characterization of cavity spectral response was performed with microwave VNA. Applying mechanical pressure to cavity membrane showed a measurable shift in the microwave resonant frequency. We also investigate mechanical integrity of the flowmeter’s membrane through computer simulations. By calculating the stress on the plate due to deflection and, comparing the stress to the material ultimate tensile strength and yield strength, it can be estimated if the plate will fail. The stress on the plate was calculated with an analytic closed form solution model, and with COMSOL Structural Mechanics Module which does not involve any approximations. Both the analytic model and COMSOL model showed that maximum stresses on the plate, which are at the radial boundary of the plate, are three orders of magnitude smaller than the yield strength and ultimate tensile strength. This indicates that the sensor is at a low risk of mechanical failure.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1999

Demonstration of an Ultra-Short Channel Metal Monolith Catalytic Reactor for Trace Contaminant Control Applications

The International Space Station (ISS) Trace Contaminant Control Subassembly (TCCS) design is based upon proven, highly reliable technology. However, because its core unit operations rely upon expendable activated charcoal and an indirectly heated high temperature catalyst, annual logistics mass, crew time, and power consumption requirements are significant. To address this situation, a unique catalytic reactor design has been developed which is suitable for retrofit into the TCCS's high temperature catalytic oxidizer (HTCO) assembly. The unique design, which employs a metallic, ultra-short channel length monolith (USCM) catalyst substrate, was tested in a flight-like TCCS HTCO assembly to investigate its performance characteristics. Test results indicate that retrofitting the TCCS with a USCM-based catalytic reactor is feasible and that it may provide significant reductions in logistics mass, crew time, and power consumption. Savings indicated by test results are up to 81% for annual logistics, 56% for crew time, and 77% for startup transient duration. In addition, its demonstrated ability to operate in a power saving mode provides up to a 43% savings in average power consumption. A summary of the USCM demonstration test objectives, approach, results, and specific benefits to the TCCS's process economics are presented,

Perry, J. L↗
DOE OSTI2022

Condition assessment of cable insulation materials in advanced reactor environments

While existing cable insulation materials are sufficient for use in the current reactor fleet, environmental conditions inside some advanced reactors, such as small modular reactors, will be substantially harsher. The work herein was conducted to assess the performance and overall survivability of several high-temperature, radiation resistant insulation polymers in a simulated small modular reactor environment. The materials evaluated under this research include polyether ether ketone, polyimide, and silicone rubber. These polymers were subjected to 1000 h of environmental stress exposure under two different conditions: 1) high temperature (i.e., 250 °C) in atmospheric pressure and 2) high temperature in a low-pressure vacuum (i.e., less than 2 psia). During the environmental exposure process, each insulation was periodically tested using several materials characterization techniques including Fourier transform infrared spectroscopy, oxidation induction time, and oxidation induction temperature. The objective of this testing was to assess and monitor changes that occurred in the thermal and chemical properties of the materials during the environmental exposures. The overall survivability of each insulation was assessed based on these property changes. The results of this research revealed that the polyimide and polyether ether ketone insulations had significantly better thermal and chemical stability than the silicone rubber during both the high-temperature atmospheric and vacuum exposures. Both polyimide and polyether ether ketone experienced minimal reductions in their chemical and thermal properties over the course of 1000 h in the simulated environments. Finally, based on this research, polyimide and polyether ether ketone show promise as potential base materials for small modular reactor cable insulation.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Handbook on the Material Properties of Yttrium Hydride for High Temperature Moderator Applications

Yttrium hydride is an optimal choice for a high-temperature moderator material in advanced thermal neutron spectrum reactors that require small core volumes. A complete database of the thermomechanical properties of yttrium hydride is not yet available, although it is much needed to understand and predict the moderator performance during service in reactors. This milestone report presents the properties of unirradiated bulk yttrium hydride as a function of hydrogen concentration—including density, crystal structure, specific heat capacity, thermal diffusivity, thermal conductivity, hardness, elastic/shear moduli, Poisson’s ratio, fracture strength, microstructure, and thermal stability. This information provides a baseline measurement for the subsequent neutron irradiation response study of yttrium hydride. The recommended empirical treatment of the data is suggested. In addition, other properties (i.e., hydrogen retention, thermal hydrogen migration, and irradiation response) requiring investigation are discussed. Also included are the Preliminary post-irradiation examination (PIE) data of yttrium hydride irradiated at 600 and 900°C to 0.1 dpa in the High Flux Isotope Reactor (HFIR). The thermophysical properties have insignificant change following this low dpa irradiation for both irradiation temperatures.

36 MATERIALS SCIENCE↗
DOE OSTI2025

High-temperature stability and thermal expansion behavior of equi-atomic refractory multi-principal element alloys based on MoNbTi system for Gen IV reactor applications

The present study investigates the thermal stability and thermal expansion behavior of seven equi-atomic refractory multi-principal element alloys (MPEAs) based on the MoNbTi ternary system composed of low neutron absorption cross section elements. Through an integrated approach utilizing in-situ high-temperature X-ray diffraction (HT-XRD) in conjunction with differential scanning calorimetry (DSC), dilatometry and ageing heat treatment, the thermal stability of the MPEAs was comprehensively analyzed. In-situ HT-XRD experiment confirmed the stability of the room temperature phases up to 1000 °C with no peaks observed corresponding to additional phases in the HT-XRD patterns at 500, 800 and 1100 °C. DSC thermograms showed the absence of peaks up to 1000 °C, while peaks and valleys corresponding to exothermic and endothermic events were observed above 1000 °C. Coefficient of thermal expansion (CTE) derived from second order polynomial fitting of linear thermal expansion data from the dilatometry experiment showed linear increment up to 1000 °C for all the alloys except those containing Zr. The Cr containing alloys exhibited notably higher CTE values, particularly the Al containing alloy exhibited the highest value. Ageing heat treatment at 800 and 1000 °C for 96 h and subsequent microstructural analysis revealed significant precipitation of secondary phases in MoNbTiZr, MoNbTiZrV and MoNbTiCrAl. In conclusion, a substantial increase in hardness was observed in MoNbTiZr and MoNbTiCrAl due to secondary phase precipitation, while the other alloys maintained hardness values comparable to their as-cast and homogenized states.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRSDate not supplied

Moderator Considerations for Space Nuclear Power and Propulsion Systems

Nuclear reactors have the potential to provide high energy density to enable sustainable surface power and advanced propulsion methods needed for human exploration activities at the moon and mars. Current mission planning is surveying different reactor types for space power and propulsion application. Of these reactor types, the use of a moderator within the reactor can enable reduced enrichment, reduce overall fuel loadings, and minimize the critical size of the reactor compared to unmoderated reference systems. This proceeding summarizes some moderator materials identified for space reactor applications: zirconium hydride, yttrium hydride, beryllium, and beryllium oxide, and the unique design considerations inherent to surface power and nuclear thermal propulsion reactor designs. It was found that there are four key considerations during the moderator selection and design process: nuclear properties, thermophysical & mechanical properties, manufacture & readiness, and environmental compatibility. Surface power reactors can benefit from moderators which minimize overall system mass and are capable of surviving high temperature irradiation environments for years with little degradation. Nuclear thermal propulsion reactors can benefit from moderators which are capable of retaining structural integrity under multiple burns while being exposed to a wide temperature range (40 < T < 500+ K). Moderator materials which exhibit good stability under irradiation and high temperature operation, minimize fuel pitch, and are high readiness are desirable for near term implementation.

nuclear thermal propulsion↗
DOE OSTI2021

Grizzly Development for Light Water Reactor and Advanced Reactor Applications in Fiscal Year 2021

This report summarizes work performed during Fiscal Year (FY) 2021 at Idaho National Laboratory (INL) for the U.S. Department of Energy’s Nuclear Engineering Advanced Modeling and Simulation (NEAMS) program for the Structural Materials and Chemistry Technical Area in the work package entitled “MS-21IN050106 - Structural Materials - INL.” This effort mainly focused on development and application of capabilities for engineering-scale analysis of nuclear reactor structural components in the Grizzly and BlackBear codes. These efforts include performance improvements and application of models for high-temperature component response, cluster dynamics modeling of precipitation in light water reactor pressure vessel steel, engineering-scale capability development in the areas of concrete simulation and reactor pressure vessel analysis, and preparation for and conducting an independent assessment of the adherence of these codes to software quality standards.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Iron surface corrosion in supercritical CO2 at atomic scale investigated by molecular dynamics simulations

Understanding the corrosion behavior of steels in supercritical carbon dioxide (S-CO2) is essential for ensuring the safe application of S-CO2 as a heat-transfer fluid in high-temperature energy systems, including advanced nuclear reactors. In this work, molecular dynamics (MD) simulations using ReaxFF potential are performed to explore the atomic-scale corrosion mechanisms of body-centered cubic iron (BCC-Fe) in S-CO2. The results show that CO2 molecules in S-CO2 decompose at the Fe surface, generating free C and O atoms that form Fe-C and Fe-O bonds and subsequently produce oxides and carbides. Concurrently, Fe atoms dissolve from the surface and diffuse into the S-CO2 region, resulting in interdiffusion of Fe, C and O atoms at the interface. The corrosion-layer thickness calculations show that high pressure and temperature induced by S-CO2 have stronger effects than surface orientation on the corrosion process. In addition, surface Fe atoms undergo substantial displacement under S-CO2 exposure, further accelerating corrosion. When a radiation-induced void is introduced near the Fe surface, the corrosion is enhanced. The void-matrix interface expands the reaction surface area and simultaneously induces corrosion reactions inside the bulk, resulting in a deeper penetration of C and O and thicker corrosion layers. All these results indicate that high-temperature, high-pressure and radiation-induced voids can seriously affect the corrosion of Fe in S-CO2, and must be considered to better use S-CO2 in nuclear facilities.

Li, Wenhua↗
DOE OSTI2022

Passive heat removal in horizontally oriented micro-HTGRs

There are novel, horizontally oriented high-temperature gas-cooled micro-reactor (HTGR) designs under consideration because of their portability, simplicity, and reliable operation. These features of the micro-HTGR design can meet the fission battery attributes, such as economic, standardize, installed, unattended and reliable. Although there have been HTGR related safety studies, micro-HTGRs when horizontally oriented are expected to exhibit significantly different thermal physics. During the unavailability of helium circulation, the internal reactor core is designed to cool by block-to-block conduction and radiation, and the reactor vessel surface is cooled by the ambient air. This scenario is anticipated during the transport of the micro-HTGR in a shipping container. The conduction and radiation between the prismatic micro-HTGR blocks in the core can be influenced by variances in the thermal contacts. This work investigated the conduction within a simulated horizontal HTGR core. An experimental setup was used to validate a numerical model. The experimental setup consisted of a hexagonal assembly with scaled prismatic blocks placed within a high-temperature vacuum environment. The gaps between the blocks were well controlled and monitored. The experimental setup was designed in such a way that the temperature variation in the axial direction was minimal, such that the experiment could be observed as a 2D (r, ) heat transfer problem. The pressurized tube was equipped with Infra-Red (IR) transparent windows, and an IR camera was used to capture the spatio-temporal temperature evolution in the hexagonal block assembly during the cooldown experiments. The experimental scenario was computationally modeled with a finite element analysis (FEA) program. Once validated, the computational model was used to investigate the impact of gap conductance on overall decay heat removal. Using a conservative estimate for gap conductance value (100 W/m 2 -K) between the prismatic blocks, there is a negligible increase in temperature observed during decay heat generation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Results of Initial Alloy 617 High Temperature Crack Growth Testing

Crack propagation data can provide valuable insights when performing a safety evaluation for a component. Alloy 617 is qualified in Section III, Division 5 of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) for elevated-temperature nuclear service up to 950°C. Little is known, however, about subcritical crack-growth phenomena in Alloy 617. Previous crack-growth studies of Alloy 617 did not investigate temperatures across the range which is currently qualified in Section III, Division 5. High-temperature crack-growth testing in air and in reactor-grade helium can provide data for establishing the crack-growth correlations in support of an ASME BPVC Section XI high-temperature flaw evaluation Code Case. Idaho National Laboratory (INL) has previously performed crack-growth testing, but the equipment requires revitalization. This report provides the status of crack-growth testing in Alloy 617 at INL.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Oxidation behavior of Fe-Cr-2Si alloys in high temperature steam

FeCrSi alloys have been proposed as alternatives to FeCrAl alloys for accident-tolerant fuel cladding in light water reactors. Experiments on the oxidation resistance of three Fe-Cr-2Si alloys in high-temperature steam, with 12, 16, and 20 wt. % Cr, show that ≥16 % Cr is sufficient to form a thin, passivating, binary oxide layer. Additionally, at 1200 °C, only Cr 2 O 3 and SiO 2 form on Fe 16 Cr 2 Si and Fe 20 Cr 2 Si, while additional ternary Fe oxides form on Fe 12 Cr 2 Si. Decohesion and spalling of oxide were observed on Fe-Cr-Si alloys exposed to steam at 900 °C and above. Fe 20 Cr 2 Si exhibits the best oxidation resistance of the three in high-temperature steam.

36 MATERIALS SCIENCE↗
DOE OSTI2025

GIF: VHTR Benchmark

Intern research presentation for C-120 Group meeting.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Investigation of Reactivities of Bimetallic Cu-Fe Oxygen Carriers with Coal in High Temperature In-situ Gasification Chemical-Looping Combustion (iG-CLC) and Chemical-Looping with Oxygen Uncoupling (CLOU) Using a Fixed Bed Reactor

This study investigated reduction reactivities and CO2 conversion efficiency of synthesized Fe-Cu-Si OCs for both in-situ gasification CLC (iG-CLC) and Chemical-Looping with Oxygen Uncoupling (CLOU) at high temperatures (950 to 1100°C) with varying ratios of OC to coal char (φ) using a fixed bed reactor-quadrupole mass spectrometer system.

Wang, Ping↗
NASA NTRS1990

H2/O2 three-body rates at high temperatures

The extraction of thrust from air breathing hypersonic propulsion systems is critically dependent on the degree to which chemical equilibrium is reached in the combustion process. In the combustion of H2/Air mixtures, slow three-body chemical reactions involving H-atoms, O-atoms, and the OH radical play an important role in energy extraction. A first-generation high temperature and pressure flash-photolysis/laser-induced fluorescence reactor was designed and constructed to measure these important three-body rates. The system employs a high power excimer laser to produce these radicals via the photolysis of stable precursors. A novel two-photon laser-induced fluorescence technique is employed to detect H-atoms without optical thickness or O2 absorption problems. To demonstrate the feasibility of the technique the apparatus in the program is designed to perform preliminary measurements on the H + O2 + M reaction at temperatures from 300 to 835 K.

Marinelli, William J.↗
NASA NTRS2024

Integrated Simulations of the Sabatier and Carbon Vapor Deposition Reactor to Understand Its Impacts to Operations and Performance

The carbon vapor deposition (CVD) reactor is a technology developed by Honeywell Aerospace to convert methane, at high temperatures, into hydrogen and solid carbon. This element is coupled with a Sabatier reactor to support a closed-loop environmental control and life support system with the aim of achieving nearly complete oxygen recovery (> 95%). Initial open-loop, brassboard CVD reactor tests and simulations have shown the CVD’s ability to achieve moderately high methane conversion and high hydrogen selectivity. However, in an integrated system, additional deficiencies are expected due to recycling of unreacted or extraneous species from the Sabatier reactor (e.g., carbon dioxide, hydrogen, water) and CVD reactor (e.g., hydrocarbons, methane, etc.). Sabatier and CVD reactor models were integrated and simulated to predict potential impacts to individual reactors’ and the overall system’s performance. The simulations showed that increasing the recycle of the CVD effluent hydrogen combined with decreasing the system inlet hydrogen flow rate (i.e., drawing a stoichiometric flow rate from an electrolyzer) can lead to an oxygen recovery of > 95%. However, system integration comes at a detriment to the individual reactors. The simulations show the initial conversion from the integrated system (Sabatier = 87% and CVD = 59%) to be lower than the standalone systems (Sabatier = 91% and CVD = 69%). Furthermore, transient simulations show substrate densification, leading to worsening methane conversion coupled with increasing acetylene production, which is commensurate with soot formation. Simulations predict a shortening of the maintenance interval (i.e., time until CVD methane conversion drops below 50%) in the integrated system, which would increase the consumable substrate mass. These analyses highlight the importance of long-duration, integrated tests to corroborate these findings as well as suggest potential modifications (e.g., intermediate gas separations) to improve performance.

Air Revitalization↗