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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 253 records · Page 14
All sourcesNASA NTRSDOE OSTI
NASA NTRS1952

Experimental investigation of coolant-flow characteristics of a sintered porous turbine blade

Local cooling-air flow rates through the walls of a sintered porous-metal turbine blade were measured at room temperature for a range of pressure drops. In order to check the validity of the correlation procedure, this procedure was used to correlate air-flow rates thorough two porous disks at temperatures up to 600 degrees F.245:00:Experimental investigation of coolant Data indicate the method permits room-temperature flow data to be used for heat-transfer work at elevated temperatures with reasonable accurach. Cooling-air flow distribution around the periphery of the test blade is presented for two internal cooling air pressures with the blade in a 1000 degrees Fahrenheit gas stream at a cascade-inlet Mach number of 0.45.

Bartoo, Edward R↗
NASA NTRS1997

Nonintrusive Flow Rate Determination Through Space Shuttle Water Coolant Loop Floodlight Coldplate

Using a Nonintrusive Flow Measurement System (NFMS), the flow rates through the Space Shuttle water coolant coldplate were determined. The objective of this in situ flow measurement was to prove or disprove a potential block inside the affected coldplate had contributed to a reduced flow rate and the subsequent ice formation on the Space Shuttle Discovery. Flow through the coldplate was originally calculated to be 35 to 38 pounds per hour. This application of ultrasonic technology advanced the envelope of flow measurements through use of 1/4-inch-diameter tubing, which resulted in extremely low flow velocities (5 to 30 pounds per hour). In situ measurements on the orbiters Discovery and Atlantis indicated both vehicles, on the average, experienced similar flow rates through the coldplate (around 25 pounds per hour), but lower rates than the designed flow. Based on the noninvasive checks, further invasive troubleshooting was eliminated. Permanent monitoring using the NFMS was recommended.

Werlink, Rudolph↗
NASA NTRS2006

Extravehicular Mobility Unit (EMU) / International Space Station (ISS) Coolant Loop Failure and Recovery

Following the Colombia accident, the Extravehicular Mobility Units (EMU) onboard ISS were unused for several months. Upon startup, the units experienced a failure in the coolant system. This failure resulted in the loss of Extravehicular Activity (EVA) capability from the US segment of ISS. With limited on-orbit evidence, a team of chemists, engineers, metallurgists, and microbiologists were able to identify the cause of the failure and develop recovery hardware and procedures. As a result of this work, the ISS crew regained the capability to perform EVAs from the US segment of the ISS.

Lewis, John F.↗
DOE OSTI2022

Coolant delivery via an independent cooling circuit

An embodiment of an independent cooling circuit for selectively delivering cooling fluid to a component of a gas turbine system includes: at least one coolant feed channel fluidly coupled to a supply of cooling fluid; and an interconnected circuit of cooling channels, including: an interconnected circuit of cooling channels embedded within an exterior wall of the component; an impingement plate; and a plurality of feed tubes connecting the impingement plate to the exterior wall of the component and fluidly coupling a supply of cooling fluid to the interconnected circuit of cooling channels; wherein the cooling fluid flows through the plurality of feed tubes into the interconnected circuit of cooling channels only in response to a formation of a breach in the exterior wall of the component that exposes at least one of the cooling channels.

42 ENGINEERING↗
DOE OSTI2023

Development of a Large-Scale Open Geometry for Squealer Tip Coolant Injection Research

In this work, a large-scale symmetric canonical axial turbine squealer tip model is developed, which allows for the investigation of incoming gas path flow interaction with coolant flow injected into the squealer cavity. The model is developed to investigate Reynolds numbers on the order of one million for their applicability to novel supercritical CO2 cycles.

Tuite, Logan M.↗
DOE OSTI2022

Coolant delivery via an independent cooling circuit

A coolant delivery system for a component of a gas turbine system includes: a plurality of independent circuits of cooling channels embedded within an exterior wall of the component, each independent circuit of cooling channels including a plurality of headers and a plurality of feed tubes fluidly coupling the plurality of headers to a supply of cooling fluid; and an impingement plate connected to the exterior wall of the component by the plurality of feed tubes of the independent circuits of cooling channels, wherein, in each of the plurality of independent circuits of cooling channels, the cooling fluid flows through the plurality of feed tubes and the plurality of headers into the circuit of cooling channels only in response to a formation of a breach in the exterior wall of the component that exposes at least one of the cooling channels of the circuit of cooling channels.

Hafner, Matthew Troy↗
DOE OSTI2020

Dismantling of the Bohunice V1 Reactor Coolant System - 20011

About 2.5 kilometers from the village of Jaslovske Bohunice in the Trnava District in western Slovakia at the Bohunice V1 nuclear power plant, Westinghouse and VUJE are working on the dismantling of two VVER-440 units. The consortium was awarded this contract in October 2017 from Jadrova a vyradovacia spolocnost (JAVYS) and has begun onsite work to dismantle the reactor coolant systems of both units that operated at the plant. The Bohunice V1 plant is composed of two V-230 type VVER 440 units (440 MWe each). The first unit ceased operation on December 31, 2006, and the second on December 31, 2008, after 28 years of operation. The project strategy relies on the early release of some plant areas to allow the erection of new facilities to enable a fast and safe dismantling. These up-front activities play a crucial role within the whole project. Since schedule and ALARA are two key elements to be considered, it has been decided to separate the dry cutting activities from the ones occurring underwater. Two Wet Cutting Workshops will allow the segmentation and packaging of the high activity components. To take advantage of the large space available in the turbine building, a Dry Cutting Workshop has been built for segmenting the twelve steam generators of both units. Two more Dry Cutting Workshops (one per unit) have been installed in the reactor building for the fragmentation of the remaining large components and auxiliary systems components. This paper describes all current preparatory works as well as the on-going segmentation and packaging activities. Interesting lessons learned are also provided. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Multi-scale simulation of high burnup UO2 nuclear fuel during loss-of-coolant accident conditions

To improve the economics of commercial nuclear energy generation, U.S. utilities are currently seeking licensing approval to operate UO2 fuel to higher burnups. One significant safety issue that must be addressed to obtain approval is the potential for fine fragmentation/pulverization of the fuel during a loss-of-coolant accident (LOCA). It has been hypothesized that pulverization is caused by the rapid increase of pressure in fission gas bubbles in the high burnup region of the fuel during a LOCA. To better understand this phenomenon, a novel phase-field model of the fission gas bubble microstructure in UO2 has been developed and implemented in Idaho National Laboratory (INL)'s Marmot application for phase-field simulation of nuclear materials. Simulations of the bubble response to steady-state and transient conditions were conducted, and the results were used to inform a mechanistic model of pulverization in BISON, INL’s fuel performance simulation code.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

DEVELOPMENT OF TRANSIENT FISSION GAS RELEASE CAPABILITIES TO STUDY HIGH BURNUP COMMERCIAL FUEL PERFORMANCE UNDER LOSS OF COOLANT ACCIDENT CONDITIONS

The contribution of transient fission gas release to the burst behaviour of high burnup fuel during a loss of coolant accident (LOCA) for commercial light water reactors is a complex phenomenon that is not fully understood. During the temperature ramp associated with a LOCA, there is significant, additional transient fission gas release in excess of steady state fission gas release from normal operation. The timing of this release and the various conditions that impact this release warrant further study. To better characterize this phenomenon, a system was developed to capture and characterize fission gas released from irradiated fuel during simulated LOCAs. This system is an upgrade to the existing Oak Ridge National Laboratory Severe Accident Test Station LOCA test apparatus and detects fission gas by gamma counting cryogenically cooled cold traps and by using an online gas mass spectrometer. The examined fuel can be pressurized or open to the furnace atmosphere. The design and operation of this new capability are discussed. Results from the first tests on high burnup fuel are presented. In the initial test, the LOCA relevant transient fission gas release was 5.3% for an unpressurized segment. Post-test microscopy on the tested samples is presented to demonstrate the likely radial source of transient fission gas release.

Harp, Jason↗
DOE OSTI2024

Computational Investigation of Fuel Dispersal Phenomena during Large-Break Loss of Coolant Accident in Light-Water Reactors

In the event of cladding rupture, which could occur in light water reactor fuel assemblies during a loss-of-coolant accident (LOCA), fuel particles, along with fission gases, can be expelled into the reactor core from the fractured fuel rod. This expulsion of fragmented fuel particles, referred as fuel dispersal, is the subject of investigation to evaluate the safety implications of increasing fuel burnup in light water reactors, with a specific focus on fuel fragmentation, relocation, and dispersal. Particle trajectories and the resulting mass distribution of the settled particles within the reactor pressure vessel can pose a long-term cooling challenge for the reactor core. The fuel dispersal phenomenon is significantly influenced by the ejection characteristics of the fuel fragments, as well as the size and shape of the cladding rupture and fuel rod depressurization history during LOCA transients. In this study, the transport of fuel particles within a scaled 5 × 5 lattice of a pressurized water reactor rod bundle geometry is modeled through a two-fluid Eulerian framework that treats the gas and solid phases as interpenetrating continua. The required boundary conditions are evaluated from the fuel performance code BISON in a postulated large-break LOCA scenario. The modeling framework considers solid fuel particles as granular matter, interacting with the gaseous dry steam phase and fission gases through the governing interfacial momentum and energy exchange between the gas and solid phases. The simulation results provide the volume fraction of the solids settled on the bottom surface of the fuel bundle, quantifying the deposition within the bundle geometry.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Predictive Modeling and Uncertainty Quantification in Condition Monitoring of Active Components: A Reactor Coolant Pump Use Case

This work develops data-driven models for onset of thermal barrier leakage in reactor coolant pumps. It incorporates uncertainty quantification to enhance the reliability and robustness of pre- dictions. Using synthetic data generated by the Generic Pressurized Water Reactor simulator, realistic degradation scenarios were simulated across lifecycle stages—beginning, middle, and end of life. Key variables, including differential pressure, flow rate, vibration, and temperatures, were analyzed using machine learning framework. The fully connected neural network models demonstrated exceptional performance, achieving R2 scores exceeding 0.99 and root mean square errors as low as around 8.23 × 10-2 gallon per minute (gpm) for the three stages of the lifecy- cle. UQ analysis further validated the model’s robustness, with narrow uncertainty bounds during steady-state operations and appropriately wider bounds during transitional phases, reflecting the physical behavior of the system. This work addresses important gaps in real-time condition moni- toring and regulatory compliance by integrating advanced condition monitoring technologies with UQ into IST programs. The ability to detect thermal barrier leakage early and quantify prediction reliability supports optimizing maintenance strategies while ensuring nuclear power plants’ safe and reliable operation.

99 - GENERAL AND MISCELLANEOUS↗
DOE OSTI2023

SPC-70734 Rev 1 MARVEL Primary Coolant System (PCS) Fill and Eutectic NaK Procurement

This specification contains the requirements for the filling the Microreactor Applications Research Validation and Evaluation (MARVEL) reactor Primary Coolant System (PCS) with eutectic sodium-potassium (NaK). The subcontractor will design and fabricate a NaK filling system (referred to herein as the PCS Fill System), the requisite volume of high purity NaK, and perform filling operations in conjunction with the Contractor Operations. The MARVEL reactor system will be operated by the Contractor, and the PCS fill system will be operated by Subcontractor personnel.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2024

ECAR-7813 Rev 0 Secondary Coolant Cover Gas System Stirling Engine Tube Rupture Analysis, Insulation Block Analysis, and Labyrinth Seal Design

The purpose of this ECAR is to document compliance with best engineering practices for the Secondary Coolant Cover Gas System of the MARVEL microreactor which includes a Stirling Engine helium tube rupture analysis, an analysis to verify the strength of the insulation blocks, and the design of the Labyrinth Seal.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗