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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 235 records · Page 13
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

A reduced-order modeling of a tubular solar reactor for long duration thermochemical energy storage

The storage of solar energy in a solid form, referred to as a “solar fuel”, can be achieved through a process known as endothermic solar thermochemistry. This process transforms the absorbed solar energy into a stable and retrievable form that can be stored for extended periods of time. This paper presents a low–order heat transfer model of a counter–current tubular falling bed reactor designed to produce thermally reduced magnesium manganese oxide pellets for long duration thermochemical energy storage. The energy required for the endothermic reduction was supplied by concentrated solar energy or renewable electricity via indirect heating of the gas and solid reactants flowing in a ceramic tube. The counter-current gas flow enhances the mixing of the solid particles with the heat recuperation zone, allowing the gas and particles to enter and exit the tubular reactor close to room temperature. Further, the reactor was vertically oriented and was heated circumferentially by an adjustable level heat flux along a finite segment of its length. The temperature distribution of the reactor in response to transient changes along the tube was modeled by considering conduction, convection, and radiation heat transfer. Governing equations for the heat transfer model were solved by discretizing the reactor tube into a finite number of control volumes and using an energy balance for the heat exchange between the reactor wall, gas, and particles within the control volume. The energy absorbed during this endothermic reaction was modeled numerically by fitting the data of the chemical conversion rate with the corresponding temperature of particles in the heating zone. The numerical model has been experimentally validated using a reactor prototype made of a 121.92 cm alumina tube heated by a 7kW electric tube–furnace. The alumina tube receives magnesium manganese oxide pellets of 3.66±0.516 mm in diameter from the top, and a counter–current gas flow from the bottom. The reactor wall temperature was monitored by six thermocouples installed along the reactor tube length. The experimental procedure was numerically simulated, and the temperature variation along the reactor tube was compared with a matrix of experimental runs for a range of particles mass flowrates (0.75–1.25g/s) and corresponding gas flowrates (36–65 SLPM). The reactor system was heated gradually from room temperature to a steady state temperature of 1673K, and then cooled down to room temperature. The heating and cooling processes were simulated, and the numerical and experimental results were compared throughout processes. The numerical model showed similar trends to the experimental results, with an error of 0.69 to 7.9% for the particle inlet and 0.7 to 7.9% for the gas inlet during steady-state operation. The proposed numerical model can be implemented as a simplified physical model to design a feedback control system to regulate reactor temperature.

14 SOLAR ENERGY↗
DOE OSTI2023

A Better Method to Calculate Fuel Burnup in Pebble Bed Reactors Using Machine Learning

Burnup measurement is an important step in material control and accountancy (MC&A) at nuclear reactors, and may be done by examining gamma spectra of fuel samples. Traditional approaches rely on known correlations to specific photopeaks (e.g. 137 Cs) and operate via a standard linear regression method. However, the quality of these regression methods is limited even in the best case, and is significantly poorer at short fuel cool-down times, due to the elevated radiation background by short life-time isotopes, and self-shielding effect of the fuel. For practical operation of pebble bed reactors (PBRs), quick measurements (in minutes) and short cooling times (in hours) are required from a safety and security perspective. We investigated the efficacy and performance of machine learning (ML) methods to predict the burnup of the pebble fuel from full gamma spectra (rather than specific discrete photopeaks) and found a full-spectrum ML approach to far outperform baseline regression predictions in all measurement and cooling conditions - including in operational-like measurement conditions. We also performed model and data ablation experiments to determine the relative performance impact of our ML methods' capacity to model data nonlinearities and the inherent additional information in full spectra. Applying our ML methods, we found a number of surprising results, including improved accuracy at shorter fuel cooling times (the opposite of the norm), remarkable robustness to spectrum compression (via rebinning), and competitive burnup predictions even when using background signal only (i.e. explicitly omitting known isotope photopeaks).

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2022

Controlling a power output of a nuclear reaction without control rods

A nuclear power system includes a reactor vessel that includes a reactor core that includes nuclear fuel assemblies configured to generate a nuclear fission reaction. A representative nuclear power system further includes a riser positioned above the reactor core and a primary coolant flow path that extends from a bottom portion of the reactor vessel, through the reactor core, and through an annulus between the riser and the reactor vessel. A primary coolant circulates through the primary coolant flow path to receive heat from the nuclear fission reaction and release the heat to a power generation system configured to generate electric power. The nuclear power system further includes a control rod assembly system positioned in the reactor vessel and configured to position control rods in only two discrete positions.

Callaway, Allyson↗
DOE OSTI2020

Preliminary design of control rods in the single-fluid double-zone thorium molten salt reactor (SD-TMSR)

Recent studies on Molten Salt Reactors (MSRs) showed that the excess reactivity at the beginning of the operation is large for many fueling strategies and must be compensated by a reactivity control system. The current work introduces a reliable safety system based on control rods in addition to the online feed system reactivity control in the Single-fluid Double-zone Thorium-based Molten Salt Reactor (SD-TMSR). Three different initial fissile loadings are considered: 233 U, reactor-grade Pu, and transuranic (TRU) elements as a startup fuel. We applied six different absorbing materials to investigate the main operational and safety parameters in the SD-TMSR: natural B 4 C, enriched B 4 C with 90% 10 B, HfB 2 , HfH 1.62 , Eu 2 O 3 , and Gd 2 O 3 . The present work focuses on control rod design, integral and differential control rod worth, shutdown margin, and shadowing effects at steady-state. We employed the SERPENT-2 Monte-Carlo code to calculate the reactivity worth and analyze the performance of the reactivity control system. We showed that 233 U and reactor-grade Pu startup cores maintain adequate shutdown margin with all considered absorbers. Finally, this report proposes a design of control rod clusters that compensate the excess reactivity of the SD-TMSR loaded with different initial fissile material.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

HFIR SiC Bowing Test Ready to Insert

This report describes the successful assembly of a High Flux Isotope Reactor (HFIR) irradiation experiment designed to assess radiation-induced lateral bowing of silicon carbide fiber–reinforced, silicon carbide matrix composite (SiC/SiC) components under a radial fast neutron flux gradient. Excessive bowing of a SiC/SiC channel box in a boiling water reactor could potentially interfere with control blade movements. Similar concerns exist for SiC/SiC fuel cladding in light water reactors. The experiment described herein will provide experimental validation of the structural response of a miniature SiC/SiC channel box and tube specimens with pressurized water reactor diameters during irradiation. The significant radial fast neutron flux gradients that exist in the permanent reflector of HFIR were characterized using detailed three-dimensional neutronic calculations. The three-dimensional displacement damage dose rate profile and the resulting volumetric swelling in SiC were used as inputs to structural analyses that predicted the deformation and stresses in the channel box specimen. The specimens were thoroughly characterized prior to irradiation using traditional dimensional inspection and surface profilometry so that these measurements can later be compared with similar measurements that will be made post-irradiation to determine radiation-induced deformations. Furthermore, fine engraving markers were inscribed along all outer surfaces of the specimen and mapped using a digital microscope and a three-dimensional stage. This technique allowed for accurate measurements of the marker spacings, which can be compared with similar measurements that will be made post-irradiation to provide local radiation-induced strain mapping. The experiment was successfully assembled and is scheduled for insertion during HFIR cycle 492, which is currently scheduled to run from May 25, 2021 to June 18, 2021.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Fast Modular Reactor Nuclear Design Parameters of Fuel Cycle and Power Distributions

This paper describes conceptual design of a 112 MW thermal (50 MW electric) Fast Modular Reactor (FMR) system operating at 7 MPa with inlet/outlet temperatures of 509/800 ºC. The reactor system includes the reactor core, fuel assemblies, fuel rods, reactor internals, reflector, neutron control system, flow control, and structural components. The nuclear design and analysis were conducted to search for a baseline core with a cycle length greater than 8-year and the power peaking factor less than 1.5 by adjusting the fuel assembly, reactor core, and reflector configurations. The neutronics calculations of the baseline core showed that a refueling interval of 9-year is achieved with a total peaking factor of 1.47 and a fuel rod average linear power of 3.6 kW/m.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Fast modular reactor nuclear design parameters of fuel cycle and power distributions

This paper describes conceptual design of a 112 MW thermal (50 MW electric) Fast Modular Reactor (FMR) system operating at 7 MPa with inlet/outlet temperatures of 509/800 Celsius degrees. The reactor system includes the reactor core, fuel assemblies, fuel rods, reactor internals, reflector, neutron control system, flow control, and structural components. The nuclear design and analysis were conducted to search for a baseline core with a cycle length greater than 8-year and the power peaking factor less than 1.5 by adjusting the fuel assembly, reactor core, and reflector configurations. The neutron transport calculations of the baseline core showed that a refueling interval of 9-year is achieved with a total peaking factor of 1.47 and a fuel rod average linear power of 3.6 kW/m. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2026

Integrating Phase Change Materials within Catalysts for Effective Heat Management

Here, most fuel conversion technologies, including Fischer-Tropsch synthesis, C-C coupling, methanol synthesis, and aldol condensation reactions, are highly exothermic, making precise temperature control essential to avoid thermodynamic limitations while maintaining selectivity and catalyst stability. Current industrial practice mitigates these effects primarily through reactor-level, ex situ temperature control strategies — such as heat exchangers, multi-tubular reactors, or quench cooling — but these approaches fall short in regulating the heat generated, transported, and dissipated at the catalyst particle scale.

Karakaya, Canan [Oak Ridge National Laboratory (OR↗
DOE OSTI2020

Vendor-Independent Design Requirements for a Boiling Water Reactor Safety System Upgrade

This Vendor-Independent License Amendment Request (LAR) Framework Document is a research product developed for the Idaho National Laboratory Light Water Reactor Sustainability (LWRS) Program. It was developed by MPR Associates, Inc. with technical input from LWRS and Exelon Generation personnel. To accomplish the purpose of this research, an operating plant (Limerick Generating Station [LGS]) was used as a reference to provide a concrete example and to present a "top-down" view of design concepts consistent with functional requirements baseline documents (references X and Y) also developed for this research. While this document is written in a framework to support a complete LAR submittal, its purpose is to communicate research concepts and provide an example of how to present digital information necessary to address expectations with regard to leveraging the Alternate Review Process as provided in Digital Instrumentation and Controls Interim Staff Guidance #06, Revision 2, Licensing Process. To leverage that process, this research document presupposes that a utility would select a vendor platform which has been prequalified for safety-related by the United States Nuclear Regulatory Commission. While this document is written using the Exelon Generation Limerick Generating Station (LGS) Units 1 and 2 as a baseline, this document is a research product. This document contains no commitments and makes no binding design decisions for Exelon Generation. Exelon Generation is leveraging this research to support their in-progress LGS Plant Protection System (PPS) upgrade efforts and plans to leverage it for their Redundant Reactor Control System (RRCS) replacement. The LWRS Program appreciates the research support provided by Exelon Generation in the generation of this document.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

The effect of moisture on hydrocarbon-based solvent liquefaction of pine, cellulose and lignin

In this report we examined the effect of moisture on the solvent liquefaction of loblolly pine, cellulose, and lignin in tetralin at 280 °C and pressures ranging from 15 to 70 bar. Solvent liquefaction experiments were conducted in a quasi-batch reactor capable of continuous pressure control and external vapor condensation. Moisture was varied by the controlled addition of de-ionized water. Control over the system pressure subsequently impacted the removal of water vapor from the reactor. Liquid yield decreased by 25, 21, and 35 wt%, for pine, cellulose, and lignin, respectively, when moisture content was increased from 1 to 50 wt% at 42 bar. Humins were observed in the solid residue from liquefaction of wet cellulose. Wet lignin yielded a substantial amount of solid residue compared to dry lignin with a corresponding decrease in total phenolic monomer production. It was concluded that the ionic dissociation of water was an important factor in loss of liquid yield in the presence of water. Although water was less than 20 wt% of the solvent loading in these experiments, it strongly influenced the liquefaction of biomass and biomass components.

09 BIOMASS FUELS↗
DOE OSTI2025

Steady state radiolysis–polyhalide accumulation products [Poster]

The study of steady-state radiolysis of polyhalides is crucial as it simulates the prolonged radiation exposure that materials (e.g., alloys, molten salts) experience in reactor environments. This research provides valuable insights into the long-term effects on material properties, chemical stability, and structural integrity. It is closely linked with fundamental radiation chemistry, which aids in predicting the chemical behavior of accumulation products. Understanding the initial speciation, yield, and chemical evolution of radiolysis products is essential for managing material degradation, controlling corrosion, and ensuring reactor safety.

36 - MATERIALS SCIENCE↗
DOE OSTI2022

Multiphysics Simulation of the NASA SIRIUS-CAL Fuel Experiment in the Transient Test Reactor Using Griffin

After approximately 50 years, NASA is restarting efforts to develop nuclear thermal propulsion (NTP) for interplanetary missions. Building upon nuclear engine tests performed from the late 1950s to the early 1970s, the present research and testing focuses on advanced materials and fabrication methods. A number of transient tests have been performed to evaluate materials performance under high-temperature, high-flux conditions, with several more experiments in the pipeline for future testing. The measured data obtained from those tests are being used to validate the Griffin reactor multiphysics code for this particular type of application. Griffin was developed at Idaho National Laboratory (INL) using the MOOSE framework. This article describes the simulation results of the SIRIUS-CAL calibration experiment in the Transient Reactor Test Facility (TREAT). SIRIUS-CAL was the first transient test conducted on NASA fuels, and although the test was performed with a relatively low core peak power, the test specimen survived a temperature exceeding 900 K. Griffin simulations of the experiment successfully matched the reactor’s power transient after calibrating the initial control rod position to match the initial reactor period. The thermal-hydraulics model largely matches the time-dependent response of a thermocouple located within the experiment specimen to within the uncertainty estimate. However, the uncertainty range is significant and must be reduced in the future.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2022

Vapor-Phase Aggregation of Cerium Oxide Nanoparticles in a Rapidly Cooling Plasma

Local conditions, such as temperature and oxygen availability, have a pronounced effect on the formation and evolution of fallout following a nuclear explosion. While the behavior of nuclear-relevant materials such as uranium has begun to be explored under a wider range of environments, little is known about the behavior of plutonium. Here, using cerium as a surrogate, we track the vapor-phase aggregation of cerium oxide nanoparticles created in a plasma flow reactor under conditions of controlled temperature at two different oxygen fugacities. In situ optical emission spectroscopy is used to measure the variations in the spectral intensity of atomic and molecular species with temperature and oxygen content. We find that the relative rate of gas-phase oxidation of cerium is highly dependent on both temperature and local redox conditions within the flow reactor, to the extent that doubling the oxygen availability effectively doubles the amount of vapor-phase cerium monoxide at high temperatures (>2000 K). Condensed cerium oxide nanoparticles are also collected and analyzed ex situ via transmission electron microscopy and grazing-incidence small-angle X-ray scattering to determine their elemental composition, crystal structure, and size distribution. The size and morphology of the condensed nanoparticles are independent of local redox conditions, forming the same crystal type with the same size distribution regardless of oxygen availability. Postcondensation particle evolution, however, is found to be predominantly driven by temperature, with the average particle size increasing as particles cool and subsequently aggregate. These results expand our understanding of the chemical and physical behavior of refractory oxides that form during the early stages of fallout formation.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI1988

Nuclear Safety [Vol. 29, No. 3, July-September 1988]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 259 Fifteenth Water Reactor Safety Information Meeting by E. G. Silver; CONTROL AND INSTRUMENTATION: 284 Reliability Technology to Improve and/or Maintain Emergency Diesel Generator Performance by S. Karimian and J. H. Taylor, 293 A Noise Diagnostics System for Operator Advice by G. Hessel, P. Liewers, P. Schumann, W. Schmitt, and F.-P. Weiss; PLANT SAFETY FEATURES: 307 A Passive Containment System for Advanced Light-Water Reactors by O. B. Falls, Jr., and F. W. Kleimola; ENVIRONMENTAL EFFECTS: 318 Data Base Construction for a Computerized Radiological Risk Investigation System by L. M. Hively, J. E. Nyquist, J. L. Bledsoe, and A. L. Sjoreen, 326 Erratum to "Radiation Hormesis and Nuclear Safety," Vol. 29, No. 1; OPERATING EXPERIENCES: 327 Operational Safety Experience and Passive Safety Testing at the Fast Flux Text Facility by Q. L. Baird, J. L. Rathbun, D. D. Stepnewski, R. L. Stover, and A. E. Waltar, 344 Backfilling of Independent Residual Heat Removal Systems in West Germany and Switzerland by G. Eckert and Y. Salomon, 353 Reactor Shutdown Experience Compiled by J. W. Cletcher, 356 Selected Safety-Related Events Compiled by G. A. Murphy, 363 Operating U.S. Power Reactors Compiled by E. G. Silver; RECENT DEVELOPMENTS: 384 General Administrative Activities Compiled by E. G. Silver, 390 Reports, Standards, and Safety Guides by D. S. Queener, 395 Status of Power-Reactor Projects Undergoing Licensing Review Compiled by E. G. Silver, 400 Proposed Rule Changes as of Mar. 31,1988; ANNOUNCEMENTS: 283 International ENS/ANS Conference on Thermal Reactor Safety "NUCSAFE 88", 317 Northwestern University Short Course on Radiation Safety, 407 Second International Seminar on Small- and Medium-Sized Nuclear Reactors, 407 International Workshop on New Developments in Occupational Dose Control and ALARA Implementation at Nuclear Power Plants and Similar Facilities, 408 Fourth International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-4), 404 The Authors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Radioxenon signatures of molten salt reactors

Developments of molten salt reactor (MSR) technologies are making rapid progress across the globe. These reactor designs involve properly controlling radionuclides through off-gas systems. This work examines radioxenon emissions and assesses treaty detection technology to be used for monitoring. This work also includes a sensitivity study with ORIGEN on the nuclear forensic signals possible from a MSR. Multiple Isotope Ratio Comparison (MIRC) plots are used to compare the ratios of radioxenon isotopes in an MSR to a highly enriched uranium (HEU) reactor. Finally, the results show that MSR emissions may significantly overlap with signals produced in a HEU reactor and that changing reactor operations or retaining emissions in an off-gas system can significantly shift the radioxenon signature away from the HEU pulse.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2021

Development of Real-Time, In-pile Creep Test Rigs for Characterizing the Structural Materials of Nuclear Components

New and improved materials are being considered for supporting both existing and next-generation nuclear reactors. Reactor materials can significantly degrade with time, thus limiting or altering their properties in harsh reactor environments. To accurately understand such material degradation, real-time data obtained under prototypic irradiation conditions are required. In particular, understanding the creep behavior of materials exposed to irradiation and elevated temperatures is essential for safety concern evaluations. To provide these capabilities, Idaho National Laboratory (INL)’s High Temperature Test Laboratory (HTTL) developed several instrumented test rigs for obtaining real-time data from specimens in well-controlled pressurized-water reactor (PWR) coolant conditions at the Materials Test Reactor. This technical report focuses on INL’s efforts to evaluate and enhance the former creep test rig prototype that relied on linear variable differential transformers in laboratory settings. Specifically, the test rig can detect changes in the length of creep specimens, which is useful for measuring thermal expansion and creep deformation.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

Scalable Framework of Hybrid Modeling with Anticipatory Control Strategy for Autonomous Operation of Modular and Microreactors

This poster provides a high level overview of the Scalable Framework of Hybrid Modeling with Anticipatory Control Strategy for Autonomous Operation of Modular and Microreactors LDRD program to be presented at a poster session for students of the 2021 MeV School. The purpose of the poster is to provide other participants of the MeV School with a high level overview of the author's research direction.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗