SEARCH · Engineering Papers

Results for “conditional thermal state”

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.

SourceResearch connectionFrom yearThrough year

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 721 records · Page 40

All sources NASA NTRS DOE OSTI

NASA NTRS1982

Hybrid thermoelectric solar collector design and analysis

A flat-plate solar collector is conceived where energy cascades through thermoelectric power modules generating direct-current electricity. The intent of this work was to choose a collector configuration and to perform a steady-state thermal performance assessment. A set of energy balance equations were written and solved numerically for the purpose of optimizing collector thermal and electrical performance. The collector design involves finned columns of thermoelectric modules imbedded in the absorber plate (hot junction) over a parallel array of vertical tubes. The thermoelectric power output is limited by the small hot-junction/cold-junction temperature difference which can be maintained under steady-state conditions. The electric power per unit tube pass area is found to have a maximum as a function of a geometric parameter, while electric power is maximized with respect to an electric resistance ratio. Although the electric power efficiency is small, results indicate that there is sufficient electric power production to drive a coolant circulator, suggesting the potential for a stand-alone system.

Roberts, A. S., Jr.↗

NASA NTRS1979

TACT 1: A computer program for the transient thermal analysis of a cooled turbine blade or vane equipped with a coolant insert. 2. Programmers manual

A computer program to calculate transient and steady state temperatures, pressures, and coolant flows in a cooled axial flow turbine blade or vane with an impingement insert is described. Coolant-side heat transfer coefficients are calculated internally in the program, with the user specifying either impingement or convection heat transfer at each internal flow station. Spent impingement air flows in a chordwise direction and is discharged through the trailing edge and through film cooling holes. The ability of the program to handle film cooling is limited by the internal flow model. Input to the program includes a description of the blade geometry, coolant-supply conditions, outside thermal boundary conditions, and wheel speed. The blade wall can have two layers of different materials, such as a ceramic thermal barrier coating over a metallic substrate. Program output includes the temperature at each node, the coolant pressures and flow rates, and the coolant-side heat transfer coefficients.

Gaugler, R. E.↗

DOE OSTI2025

Corrosion of 316 L stainless steel under the natural circulation of molten NaCl-MgCl 2 salt

The corrosion behavior of 316 L stainless steel (SS) was studied via the natural circulation of molten eutectic NaCl-MgCl 2 salt through a microloop. The post-corrosion tested 316 L SS microloop sections were characterized with microscopy techniques to determine the microstructural and microchemical changes that occurred at the alloy/salt interface. It was found that 316 L SS showed heterogeneous dissolution at the hot-leg, whereas deposition of corrosion products occurred at the cold-leg. For the first time, experimentally obtained molten salt flow-induced corrosion of 316 L SS results were combined with computational thermodynamic-kinetic models to validate the dissolution and deposition in terms of elemental compositional changes at the alloy/salt interface. The thermodynamic-kinetic modeling predicted that the heterogeneous dissolution of Cr and Fe from the hot-leg section of 316 L SS persisted throughout the salt circulation. The model also estimated that, despite Cr deposition starting earlier than Fe, the total redeposition of Fe is expected to be significantly greater than that of Cr over the circulation of salt. Furthermore, the modeling accurately predicted the subsurface enrichment of Mo which is attributed to the reduced Cr activity and the relatively higher diffusion rate of Mo within the alloy matrix. Here, the agreement between modeling and experimental results confirms that Fe chlorides dissolve at the hot-leg and subsequently deposit at the cold-leg due to activity changes driven by the thermal gradient. By contrast, Cr was not detected in the cold-leg deposits, which is attributed to its weaker temperature dependence on activity, limiting its redeposition under these conditions.

36 - MATERIALS SCIENCE↗

DOE OSTI2020

FAST-1.0: A Computer Code for Thermal-Mechanical Nuclear Fuel Analysis under Steady-state and Transients: Developed under NQA-1-2017

Fuel Analysis under Steady-state and Transients (FAST) is the US Nuclear Regulatory Commission (NRC)’s computer code that calculates the steady-state and transient response of nuclear reactor fuel rods during long-term in-reactor burnup, anticipated operational occurrences (AOOs), design basis accidents (DBAs), and dry storage conditions. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include: heat conduction through the fuel and other materials, heat transfer from the cladding-to-coolant, cladding elastic and plastic deformation, including creep, fuel-cladding mechanical interaction, fission gas release from the fuel, rod internal pressure and void volume, and cladding oxidation. The code contains necessary material and coolant properties, as well as clad-to-coolant heattransfer correlations, for normal operation through postulated accidents for today’s US-based light water reactor (LWR) fuel designs. FAST-1.0 also contains preliminary materials and models for new LWR fuel concepts, such as accident tolerant fuel (ATF), and non-LWR fuel concepts such as metallic fuels for sodium fast reactors (SFRs). FAST has been developed for use on Windows and Linux operating systems. This document describes FAST-1.0, which is the first official version of this code. This document is one of a series of documents on FAST; the other documents detail the material properties used by FAST as well as its integral assessment to experiments and commercial data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

FAST-1.1: A Computer Code for Thermal-Mechanical Nuclear Fuel Analysis under Steady-state and Transients (Developed under NQA-1-2017)

Fuel Analysis under Steady-state and Transients (FAST) is the U.S. Nuclear Regulatory Commission (NRC)’s computer code that calculates the steady-state and transient response of nuclear reactor fuel rods during long-term in-reactor burnup, anticipated operational occurrences (AOOs), design basis accidents (DBAs), and dry storage conditions. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include: * heat conduction through the fuel and other materials * heat transfer from the cladding-to-coolant * cladding elastic and plastic deformation, including creep * fuel-cladding mechanical interaction * fission gas release from the fuel * rod internal pressure and void volume * cladding oxidation The code contains necessary material and coolant properties, as well as clad-to-coolant heat transfer correlations, for normal operation through postulated accidents and AOOs for today’s U.S.-based light water reactor (LWR) fuel designs. FAST-1.1 also contains preliminary materials and models for new LWR fuel concepts, such as accident tolerant fuel (ATF), and non-LWR fuel concepts such as metallic fuels for sodium fast reactors (SFRs). FAST has been developed for use on Windows and Linux operating systems. This document describes FAST-1.1 and is one of a series of documents on the code; the other documents detail the material properties used by FAST as well as its integral assessment to experiments and commercial data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS1974

Prediction and verification of creep behavior in metallic materials and components, for the space shuttle thermal protection system. Volume 1, phase 1: Cyclic materials creep predictions

Cyclic creep response was investigated and design methods applicable to thermal protection system structures were developed. The steady-state (constant temperature and load) and cyclic creep response characteristics of four alloys were studied. Steady-state creep data were gathered through a literature survey to establish reference data bases. These data bases were used to develop empirical equations describing creep as a function of time, temperature, and stress and as a basis of comparison for test data. Steady-state creep tests and tensile cyclic tests were conducted. The following factors were investigated: material thickness and rolling direction; material cyclic creep response under varying loads and temperatures; constant stress and temperature cycles representing flight conditions; changing stresses present in a creeping beam as a result of stress redistribution; and complex stress and temperature profiles representative of space shuttle orbiter trajectories. A computer program was written, applying creep hardening theories and empirical equations for creep, to aid in analysis of test data. Results are considered applicable to a variety of structures which are cyclicly exposed to creep producing thermal environments.

Davis, J. W.↗

DOE OSTI2022

Preliminary Design of Reactor Cavity Cooling System for a Horizontal Compact HTGR

The Horizontal Compact High Temperature Gas Reactor (HC-HTGR) is being designed by a multi-disciplinary team of nuclear, mechanical, and structural engineers under the support of a DOE-NE Advanced Reactor Demonstration Program’s Advanced Reactor Concepts-20 (ARC-20) award. The objective of this ARC-20 project is to deliver a conceptual design for the proposed MIGHTR in 3 years and support its commercialization as a safe and low-cost HTGR. Argonne National Laboratory (Argonne) is responsible for the design and analysis of the reactor cavity cooling system (RCCS) as a safety system for passive decay heat removal of the reactor concept. This report documents the preliminary design study of the RCCS for the HC-HTGR. It includes the establishment of the design requirements, a high-level design study by initial scoping calculations, and preliminary performance calculations of the HC-HTGR RCCS design. Design requirements for the HC-HTGR RCCS have been established to guide preliminary design activities and scoping performance calculations. Initial scoping calculations including estimation of the water inventory, estimation of HVAC thermal capability, and a parametric study on loop dimensions by standalone RCCS analysis. Based on scoping calculation results, a set of baseline dimensions of the HC-HTGR RCCS was derived. A water panel modeling approach was investigated to explore various potential design options for the water panel under consideration for the HC-HTGR RCCS using RELAP5-3D. A test case study was performed to assess the prediction capability of two modeling approaches. The results were compared with CFD simulations conducted in constant RPV temperature and heat flux boundary conditions. It confirms the capability of the RELAP5-3D modeling approach to include all important heat transfer mechanisms expected in the HC-HTGR RCCS operation conditions. Then, a reference RELAP5-3D model for the 1/8 th of a compartment of the preliminary design of the HC-HTGR RCCS was developed. A preliminary performance analysis was conducted to evaluate single-phase natural circulation performance with different top tank temperature values and panel conduction performance in various operation conditions. From single-phase natural circulation performance analysis, the system operation mode was investigated in normal operating and limiting design conditions. It showed operation mode in a subcooled state with a proper top tank water cooling system. Parasitic heat loss by both internal air flow and RCCS was estimated, showing it satisfies maintaining below target maximum heat loss of the HC-HTGR RCCS. From the panel conduction performance analysis, two candidate materials for the riser tube such as carbon steel and stainless steel were compared in the thermal performance of HC-HTGR RCCS. From a single water panel test compared with CFD simulation results, it was confirmed that the current capability of the RELAP5-3D modeling approach for the water panel predicts the thermal conduction of two different materials of the water panel. Then, system-level thermal performance analysis was performed for 1/8 th of the compartment of the preliminary HC-HTGR RCCS design. It was first observed that the current preliminary HC-HTGR RCCS design had minimal impact on the overall thermal performance of the water panel by changing pipe material from carbon steel to stainless steel. From Argonne’s effort on the ongoing water-based NSTF testing program, several considerations other than the thermal performance point of view were addressed to be considered in selecting pipe materials.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

Transport physics dependence of Bohm speed in presheath–sheath transition

We report the ion exit flow speed at the sheath entrance is constrained by the Bohm criterion, which is used as a boundary condition for simulations that do not resolve the sheath region. Traditional Bohm criterion analysis invokes the equation of state and, thus, ignores transport physics in the sheath transition problem. An expression for the Bohm speed away from the asymptotic limit is derived from a set of anisotropic plasma transport equations. The thermal force, collisional temperature isotropization, and heat flux enter into the evaluation of the Bohm speed. By comparison with kinetic simulation results, this expression is shown to be accurate in the presheath–sheath transition region rather than a single point at the sheath entrance over a broad range of collisionality.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2023

Sound Assisted Low Temperature Spalling: Upscaling and Throughput

In evaluating the state-of-the-art spalling techniques, the key factor that is missing is the control of crack propagation, which is responsible for low quality wafers due to high surface roughness and Total Thickness Variation, TTV. This project has shown that acoustic waves can alter the conditions at the crack tip during thermal spalling, modifying the propagation speed and the stress field around it (stress intensity factor, K). Both factors are directly related to the roughness of the spalled surface, so as we control speed and stress, we can fine-tune the surface roughness through the entire wafer as suggested by Arakawa.

42 ENGINEERING↗

DOE OSTI2022

FAST-1.1: Integral Assessment

An integral assessment has been performed to quantify the predictive capabilities of FAST, a thermal-mechanical nuclear fuel performance code designed to analyze fuel behavior from beginning of life to burnup levels allowed by the U.S. Nuclear Regulatory Commission (NRC). FAST code calculations are shown to compare satisfactorily to a preselected set of experimental data with both steady-state and anticipated operating occurrence (AOO) conditions and design basis accident (DBA) transient operating conditions. This document describes the assessment of FAST-1.1, the latest version of FAST, released April 2022. Abstract

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

FAST-1.2: Integral Assessment

An integral assessment has been performed to quantify the predictive capabilities of FAST, a thermal-mechanical nuclear fuel performance code designed to analyze fuel behavior from beginning of life to burnup levels allowed by the U.S. Nuclear Regulatory Commission (NRC). FAST code calculations are shown to compare satisfactorily to a preselected set of experimental data with both steady-state and anticipated operating occurrence (AOO) conditions and design basis accident (DBA) transient operating conditions. This document describes the assessment of FAST-1.2, the latest version of FAST, released March 2023.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2026

FAST-1.2.2: Integral Assessment

This integral assessment quantifies the predictive capabilities of Fuel Analysis under Steady-state and Transients (FAST), a thermal-mechanical nuclear fuel performance code designed to analyze fuel behavior from beginning of life to burnup levels allowed by the U.S. Nuclear Regulatory Commission. FAST code calculations are shown to compare satisfactorily to a pre-selected set of experimental data with both steady-state and anticipated operating occurrence conditions and design basis accident transient operating conditions. This document describes the assessment of FAST-1.2.2, the latest version of FAST, released April 2026.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS1991

Integrated flow and structural modeling for rocket engine component test facility propellant systems

A set of PC-based computational Dynamic Fluid Flow Simulation models is presented for modeling facility gas and cryogenic systems. Data obtained provide important information regarding performance envelope parameters for the facility using different engine components; time-dependent valve setting for controlling steady-state, quasi-steady state, and transient profiles; optimum facility pipe and pipe component sizes and parameters; momentum transfer loads; and fluid conditions at critical points. A set of COSMIC NASTRAN-based finite element models is also presented to evaluate the loads and stresses on test facility piping systems from fluid and gaseous effects, thermal chill down, and occasional wind loads. The models are based on Apple Macintosh software which makes it possible to change numerous parameters.

Dequay, L.↗

NASA NTRS1997

Workshop on Parent-Body and Nebular Modification of Chondritic Materials

The purpose of the workshop was to advance our understanding of solar nebula and asteroidal processes from studies of modification features in chondrites and interplanetary dust particles. As reflected in the program contained in this volume, the workshop included five regular sessions, a summary session, and a poster session. Twenty-three posters and 42 invited and contributed talks were presented. Part 1 of this report contains the abstracts of these presentations. The focus of the workshop included: (1) mineralogical, petrologic, chemical, and isotopic observations of the alteration mineralogy in interplanetary dust particles, ordinary and carbonaceous chondrites, and their components (Ca-Al-rich inclusions, chondrules, and matrix) to constrain the conditions and place of alteration; (2) sources of water in chondrites; (3) the relationship between aqueous alteration and thermal metamorphism; (4) short-lived radionuclides, AI-26, Mn-53, and I-129, as isotopic constraints on timing of alteration; (5) experimental and theoretical modeling of alteration reactions; and (6) the oxidation state of the solar nebula. There were approximately 140 participants at the workshop, probably due in part to the timeliness of the workshop goals and the workshop location. In the end few new agreements were achieved between warring factions, but new research efforts were forged and areas of fruitful future exploration were highlighted. Judged by these results, the workshop was successful.

Krot, A. N.↗

NASA NTRS2012

Wavefront Compensation Segmented Mirror Sensing and Control

The primary mirror of very large submillimeter-wave telescopes will necessarily be segmented into many separate mirror panels. These panels must be continuously co-phased to keep the telescope wavefront error less than a small fraction of a wavelength, to ten microns RMS (root mean square) or less. This performance must be maintained continuously across the full aperture of the telescope, in all pointing conditions, and in a variable thermal environment. A wavefront compensation segmented mirror sensing and control system, consisting of optical edge sensors, Wavefront Compensation Estimator/Controller Soft ware, and segment position actuators is proposed. Optical edge sensors are placed two per each segment-to-segment edge to continuously measure changes in segment state. Segment position actuators (three per segment) are used to move the panels. A computer control system uses the edge sensor measurements to estimate the state of all of the segments and to predict the wavefront error; segment actuator commands are computed that minimize the wavefront error. Translational or rotational motions of one segment relative to the other cause lateral displacement of the light beam, which is measured by the imaging sensor. For high accuracy, the collimator uses a shaped mask, such as one or more slits, so that the light beam forms a pattern on the sensor that permits sensing accuracy of better than 0.1 micron in two axes: in the z or local surface normal direction, and in the y direction parallel to the mirror surface and perpendicular to the beam direction. Using a co-aligned pair of sensors, with the location of the detector and collimated light source interchanged, four degrees of freedom can be sensed: transverse x and y displacements, as well as two bending angles (pitch and yaw). In this approach, each optical edge sensor head has a collimator and an imager, placing one sensor head on each side of a segment gap, with two parallel light beams crossing the gap. Two sets of optical edge sensors are used per segment-to-segment edge, separated by a finite distance along the segment edge, for four optical heads, each with an imager and a collimator. By orienting the beam direction of one edge sensor pair to be +45 away from the segment edge direction, and the other sensor pair to be oriented -45 away from the segment edge direction, all six degrees of freedom of relative motion between the segments can be measured with some redundancy. The software resides in a computer that receives each of the optical edge sensor signals, as well as telescope pointing commands. It feeds back the edge sensor signals to keep the primary mirror figure within specification. It uses a feed-forward control to compensate for global effects such as decollimation of the primary and secondary mirrors due to gravity sag as the telescope pointing changes to track science objects. Three segment position actuators will be provided per segment to enable controlled motions in the piston, tip, and tilt degrees of freedom. These actuators are driven by the software, providing the optical changes needed to keep the telescope phased.

Redding, David C.↗

NASA NTRS2012

Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Upgrade Activities

To support the on-going nuclear thermal propulsion effort, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The facility to perform this testing is referred to as the Nuclear Thermal Rocket Element Environment Simulator (NTREES). This device can simulate the environmental conditions (minus the radiation) to which nuclear rocket fuel components will be subjected during reactor operation. Test articles mounted in the simulator are inductively heated in such a manner so as to accurately reproduce the temperatures and heat fluxes which would normally occur as a result of nuclear fission and would be exposed to flowing hydrogen. Initial testing of a somewhat prototypical fuel element has been successfully performed in NTREES and the facility has now been shutdown to allow for an extensive reconfiguration of the facility which will result in a significant upgrade in its capabilities

Emrich, William J. Jr.↗

NASA NTRS2013

Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Phase II Upgrade Activities

Emrich, William J.↗

NASA NTRS2014

Initial Operation and Shakedown of the Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

To support the on-going nuclear thermal propulsion effort, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The facility to perform this testing is referred to as the Nuclear Thermal Rocket Element Environment Simulator (NTREES). This device can simulate the environmental conditions (minus the radiation) to which nuclear rocket fuel components will be subjected during reactor operation. Prototypical fuel elements mounted in the simulator are inductively heated in such a manner so as to accurately reproduce the temperatures and heat fluxes which would normally occur as a result of nuclear fission in addition to being exposed to flowing hydrogen. Recent upgrades to NTREES now allow power levels 24 times greater than those achievable in the previous facility configuration. This higher power operation will allow near prototypical power densities and flows to finally be achieved in most prototypical fuel elements.

Emrich, William J., Jr.↗