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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 559 records · Page 31
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

OPTIMIZATION OPTIONS FOR NARROWING TRANSIENT TESTING PULSE WIDTHS NECESSARY TO ENHANCE LWR RIA SIMULATIONS IN THE TREAT FACILITY

The Transient Reactor Test (TREAT) Facility provided thousands of transient irradiations and plays a crucial role in nuclear-heated safety research. The facility’s flexible design and multi-mission nature saw historic experiments for numerous reactor fuels and transient types, but was never specifically adapted to address very-brief pulse transients akin to postulated Light Water Reactor (LWR) Reactivity Initiated Accidents (RIA). Since the behaviors of fuel under these conditions depends strongly on energy input duration and resulting cladding time/temperature response under pellet cladding mechanical interactions, this three-year project was conceived to investigate new pulse-narrowing capabilities. Kinetic models showed incremental improvements for minor facility enhancements including increased reactivity step insertions (to initiate the power pulse) and slightly-increased transient rod drive speed for pulse termination (“clipping”). Replacing peak fuel assemblies, repositioning non-transient control rods to hold down the “hot side” of the core, and balancing against required excess reactivity needs, the limiting fuel assembly power can be reduced by ~20 %. This is a remarkable discovery of latent capability in TREAT, not only in enabling the subject capabilities for reduced pulse widths, but also significantly “uprating” the core’s transient energy capacity. Incremental improvements can likely enhance TREAT’s capability into BWR-relevant missions; briefer PWR pulse shapes can only be achieved with an advanced clipping system. Helium 3 was found to offer the greatest benefits for clipping design and overall performance. A unique concept showed great promise for enabling a 3He-based system with a reasonable cost; this design concept will likely become the focal point of future work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Narrowing transient testing pulse widths to enhance LWR RIA experiment design in the TREAT facility

The Transient Reactor Test (TREAT) Facility provided thousands of transient irradiations and plays a crucial role in nuclear-heated safety research. The facility’s flexible design and multi-mission nature saw historic experiments for numerous reactor fuels and transient types but was never specifically adapted to address very-brief pulse transients akin to postulated Light Water Reactor (LWR) Reactivity Initiated Accidents (RIA). Since the behaviors of fuel under these conditions depends strongly on energy input duration and resulting cladding time/temperature response under pellet cladding mechanical interactions, this three-year project was conceived to investigate new pulse-narrowing capabilities. Kinetic models showed incremental improvements for minor facility enhancements including increased reactivity step insertions (to initiate the power pulse) and slightly-increased transient rod drive speed for pulse termination (“clipping”). Replacing peak fuel assemblies, repositioning non-transient control rods to hold down the “hot side” of the core, and balancing against required excess reactivity needs, the limiting fuel assembly power can be reduced by ~20 %. Furthermore, this is a remarkable discovery of latent capability in TREAT, not only in enabling the subject capabilities for reduced pulse widths, but also significantly “uprating” the core’s transient energy capacity. Incremental improvements can likely enhance TREAT’s capability into BWR-relevant missions; briefer PWR pulse shapes can only be achieved with an advanced clipping system. Helium-3 ( 3 He) was found to offer the greatest benefits for clipping design and overall performance. A unique concept showed great promise for enabling a 3 He-based system with a reasonable cost; this design concept will likely become the focal point of future work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1974

Advanced-power-reactor design concepts and performance characteristics

Five reactor cooling concepts which allow continued reactor operation following a single rupture of the coolant system are presented for application with the APR. These concepts incorporate convective cooling, double containment, or heat pipes to ensure operation after a coolant line rupture. Based on an evaluation of several control system concepts, a molybdenum clad, beryllium oxide sliding reflector located outside the pressure vessel is recommended.

Davison, H. W.↗
DOE OSTI2016

Determination of water equivalent content in fuel rods

In water-cooled reactors, hydriding failure of zircaloy cladding tubes can be easily caused by high content of total water in fuel rods. WEC, which is the water equivalent content within the cold space of a fuel rod, directly reflects the total water content of the fuel rod. A device measuring WEC has been designed, manufactured and tested. This extraction device is a hermetically sealed system consisting of six parts: the break-off part, the heating unit, the reduction unit, the sample storage part, the collecting unit and the gas flow system. Each part is connected to another one using a special connection, and the device's end part is connected to a gas chromatograph. The determination procedure can be briefly summarized as follows: first, a test fuel rod should be put into the sealed device after lathing two circular grooves on both ends, so that it can be easily broken at these points in the break-off part. Driven by carrier gas, water and hydrogen both contained in helium gas and absorbed on the rod in-wall pass over the hot uranium filament where water is converted into hydrogen. After being concentrated by the collecting unit, hydrogen flows into the gas chromatograph for measurement and data are calculated as total water content A. UO{sub 2} pellets in the test rod should be taken out from the sample storage part and the total hydrogen content of these pellets is determined according to ASTM C1457 standard, followed by its conversion into total water content B. Finally, the sum of A and B being divided by the free volume of the cold space within the test rod, gives the value of WEC.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Interactions and immobilization of lanthanides with dopants in uranium-based metallic fuels

In uranium-based fuels, production and migration of lanthanide fission products and subsequent chemical interactions with cladding constituents lead to fuel–cladding chemical interaction (FCCI), limiting fuel performance. In general, dopant addition to the fuel matrix to arrest lanthanides within the fuel by forming intermetallics is found to be effective in mitigating FCCI. Recently, we proposed ab-initio based alloy-design principles, which can be useful in identifying dopant(s) that can bind a lanthanide (Nd) inside the fuel matrix. Here, we demonstrate the robustness of such principles by performing a systematic study to choose dopants to form compounds with a range of experimentally-observed lanthanide fission products. Results reveal that for a given dopant, lanthanide Ce and La exhibit similar behavior in their compound-forming tendencies compared to Nd and Pr, which is linked to lanthanide intrinsic characteristics such as the electronic configurations. Our predicted lanthanide-dopant intermetallic formation is verified experimentally in selective cases. Finally, we showed alloy design principles that accurately identify previously known dopants like Pd, and also predict new dopants As and Se, that can be effective in binding all lanthanides within the uranium matrix. Altogether, this research helps to develop generic alloy-design principles for complex multi-component systems based on lanthanide and dopant intrinsic characteristics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Characterization of oxide films on aluminum materials following reactor exposure and wet storage in the SRS L-Basin

This report describes the characterization of oxide films on one cropped piece of 6061, one 6063, and one cropped piece that could have been either 6061 or 6063 aluminum-based alloy, stored wet for up to approximately 40 years in L-Basin on the Savannah River Site (SRS). One 6061 cropping was from a Missouri University Research Reactor (MURR) fuel element. The 6063 piece was from a cropping from a Universal Sleeve Housing (USH). The Mark-16B fuel assembly could have been either 6061 or 6063 based on design drawings. Both the Mark-16B and USH were used in SRS Production Reactors. Characterization of the as-received oxides includes the oxide morphology, thickness, structure, and chemical composition. The aim is to characterize the oxide layer on aluminum cladding and fuel components that have reactor service and long-term basin storage service.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1980

Investigation of direct integrated optics modulators

Direct modulation techniques applicable to integrated optics data preprocessors were investigated. Several methods of modulating a coherent optical beam by interaction with an incoherent beam were studied. It was decided to investigate photon induced conductivity changes in thin semiconductor cladding layers on optical waveguides. Preliminary calculations indicate significant changes can be produced in the phase shift in a propagating wave when the conductivity is changed by ten percent or more. Experimental devices to verify these predicted phase changes and experiments designed to prove the concept are described.

Batchman, T. E.↗
NASA NTRS1982

Design solutions for the solar cell interconnect fatigue fracture problem

Mechanical fatigue of solar cell interconnects is a major failure mechanism in photovoltaic arrays. A comprehensive approach to the reliability design of interconnects, together with extensive design data for the fatigue properties of copper interconnects, has been published. This paper extends the previous work, developing failure prediction (fatigue) data for additional interconnect material choices, including aluminum and a variety of copper-Invar and copper-steel claddings. An improved global fatigue function is used to model the probability-of-failure statistics of each material as a function of level and number of cycles of applied strain. Life-cycle economic analyses are used to evaluate the relative merits of each material choce. The copper-Invar clad composites demonstrate superior performance over pure copper. Aluminum results are disappointing.

Mon, G. R.↗
DOE OSTI2022

Thermal Model of Oxide Growth for Full Size Fuel Plate Experiments in the Advanced Test Reactor

The full-size plate number one (FSP-1) irradiation test designed for irradiation testing of plate type fuel is planned to be irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). This FSP-1 experiment is a non-instrumented drop-in test where aluminum-clad fuel plates are cooled directly by the ATR Primary Coolant System (PCS) water. This experiment is one of a series of experiments being irradiated at ATR for the United States High Performance Research Reactor (USHPRR) program. This fueled experiment contains aluminum-clad fuel full-size plates consisting of monolithic U-10Mo. Previous thermal analyses have been documented concerning oxide growth [1] and thermal safety margins [2] for similar U-10Mo mini plate experiments. This analysis is part of the thermal safety analysis of the experiment to be irradiated in ATR. Departure from nucleate boiling ratio (DNBR) and flow instability ratio (FIR) have been calculated for the flow coastdown condition 2 and the reactivity insertion accident condition-2 (RIA2) transient. The purpose of this paper is to investigate the thermal margin of the fuel meat during a RIA2 transient when oxide has accumulated on the plate surfaces acting as an insulator.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Unit mechanisms of fission gas release: Current understanding and future needs

Gaseous fission product transport and release has a large impact on fuel performance, degrading fuel properties and, once the gas is released into the gap between the fuel and cladding, lowering gap thermal conductivity and increasing gap pressure. While gaseous fission product behavior has been investigated with bulk reactor experiments and simplified analytical models, recent improvements in experimental and modeling approaches at the atomistic and mesoscales are being applied to provide unprecedented understanding of the unit mechanisms that define the fission product behavior. In this article, existing research on the basic mechanisms behind the various stages of fission gas release during normal reactor operation are summarized and critical areas where experimental and simulation work is needed are identified. This basic understanding of the fission gas behavior mechanisms has the potential to revolutionize our ability to predict fission product behavior during reactor operation and to design fuels that have improved fission product retention. In addition, this work can serve as a model on how a coupled experimental and modeling approach can be applied to understand the unit mechanisms behind other critical behaviors in reactor materials.

computer simulation↗
DOE OSTI2016

CFD Analysis and Design of Detailed Target Configurations for an Accelerator-Driven Subcritical System

High-fidelity analysis has been utilized in the design of beam target options for an accelerator driven subcritical system. Designs featuring stacks of plates with square cross section have been investigated for both tungsten and uranium target materials. The presented work includes the first thermal-hydraulic simulations of the full, detailed target geometry. The innovative target cooling manifold design features many regions with complex flow features, including 90 bends and merging jets, which necessitate three-dimensional fluid simulations. These were performed using the commercial computational fluid dynamics code STAR-CCM+. Conjugate heat transfer was modeled between the plates, cladding, manifold structure, and fluid. Steady-state simulations were performed but lacked good residual convergence. Unsteady simulations were then performed, which converged well and demonstrated that flow instability existed in the lower portion of the manifold. It was established that the flow instability had little effect on the peak plate temperatures, which were well below the melting point. The estimated plate surface temperatures and target region pressure were shown to provide sufficient margin to subcooled boiling for standard operating conditions. This demonstrated the safety of both potential target configurations during normal operation.

43 PARTICLE ACCELERATORS↗
NASA NTRS1977

A lightweight solar array study

A sample module was assembled to model a portion of a flexible extendable solar array, a type that promises to become the next generation of solar array design. The resulting study of this module is intended to provide technical support to the array designer for lightweight component selection, specifications, and tests. Selected from available lightweight components were 127-micron-thick wrap-around contacted solar cells, 34- micron-thick sputtered glass covers, and as a substrate a 13-micron-thick polyimide film clad with a copper printed circuit. Each component displayed weaknesses. The thin solar cells had excessive breakage losses. Sputtered glass cover adhesion was poor, and the covered cell was weaker than the cell uncovered. Thermal stresses caused some cell delamination from the model solar array substrate.

Josephs, R. H.↗
DOE OSTI2021

Architecture and Component Selection for the Remote Canister-Monitoring System

This report documents the selected system architecture and components surveyed for the development of a remote, canister-monitoring system (RCMS) for determining and monitoring environmental conditions within dry fuel storage canisters containing aluminum-clad spent nuclear fuel (ASNF) at Idaho National Laboratory (INL). These efforts are in support of the Department of Energy Office of Environmental Management (DOE-EM) investigations into the technical issues associated with extended dry storage (50+ years) of ASNF. It is important to resolve questions regarding storage because the ATR is expected to continue producing approximately 100 elements of spent fuel per year for at least another twenty years. The parameters to be monitored by the RCMS include temperature, relative humidity, hydrogen gas concentration, and radiation environment (dose). Components are selected based on the expected environmental conditions within the canister and fuel storage area and the desired performance of the RCMS. Several of these components have been purchased and are currently undergoing component testing to confirm their suitability for use in delivering the RCMS. Additional components may be identified through the course of design and may be selected in the future. Furthermore, changing performance, interfacing, and compatibility requirements and the results of component testing may necessitate revisiting component selection.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1977

Feasibility of modern airships - Design definition and performance of selected concepts

Three promising modern airship system concepts and their associated missions were studied: (1) a heavy-lift airship, employing a nonrigid hull and a significant amount of rotor lift, used for short-range transporting and positioning of heavy military and civil payloads; (2) a VTOL (vertical take-off and landing), metal-clad, partially buoyant airship used as a short-haul commercial transport; (3) a fully-buoyant airship used for long-endurance Navy missions. The heavy-lift airship concept offers a dramatic increase in vertical lift capability over existing systems at significantly lower costs per ton-mile. The VTOL airship transport concept appears to be economically competititve with other VTOL aircraft concepts but can attain significantly lower noise levels. The fully-buoyant airship concept can provide an airborne platform with long endurance that satisfies many Navy mission requirements.

Huston, R. R.↗
DOE OSTI2017

A simulation of I 2 S-LWR selected transients

The Integral Inherently Safe (I 2 S-LWR) reactor was designed to improve the safety performance of PWR types reactors, and several design basis accidents were eliminated using innovative approaches to the I 2 S-LWR plant layout. However, some accidents such as station blackout (SBO), main steam line break (MSLB) type accident – the secondary hot leg break (SHLB) – were deemed statistically relevant, and the work performed here was to simulate these events using the system thermal–hydraulics code RELAP5 coupled to the 3D spatial kinetics code PARCS. The safety response of the reactor to a station blackout is predicted to be acceptable at both beginning of cycle (BOC) and end of cycle EOC core conditions with either the UO 2 or U 3 Si 2 fuel options considered here. In the analysis of the SHLB performed here, an unprotected (no SCRAM) condition was assumed to highlight the I 2 S-LWR capabilities. Results indicate that with the UO 2 core, the fuel temperatures rise notably, but no fuel melting was predicted and departure from nucleate boiling ratio (DNBR) limits were not exceeded. Similarly, the U 3 Si 2 fuel performed well, but with lower fuel temperatures and decreased DNBR. Following the break of the steam line in the I 2 S-LWR, shutdown is achieved after about two minutes via negative reactivity from loss of primary flow. In all cases considered, the I 2 S-LWR appears to accomplish the goals of increased safety performance. Further design improvements and other fuel options may marginally improve on the observed safety performance. Additionally, higher fidelity analysis with the next generation subchannel, Computational Fluid Dynamics (CFD), or full core transport may be useful in refining the prediction of the peak fuel and cladding temperatures, as well limiting DNBR during the accidents.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Multiphysics analysis of fuel fragmentation, relocation, and dispersal susceptibility–Part 1: Overview and code coupling strategies

The US nuclear energy industry is investigating strategies to increase the reactor operating cycle to 24 months, resulting in peak rod average burnups exceeding the current limit of 62 GWd/tU. This increase will in turn increase the probability of fuel fragmentation, relocation, and dispersal (FFRD) in the event of a loss-of-coolant accident (LOCA). This effort couples multiple codes to (1) evaluate full-core power histories for high-burnup fuel operated in a Westinghouse 4-loop pressurized water reactor, (2) model a postulated large-break LOCA, and (3) calculate the mass of fuel susceptible to FFRD. This paper, the first of three describing the work, focuses on code coupling strategies and FFRD susceptibility calculations. The other two companion papers focus on code-specific designs and analyses. Three codes were used in this work. VERA was used to calculate steady-state power histories, TRACE was used to model the transient thermal hydraulics, and BISON was used to model steady-state and transient fuel performance and cladding failure. Herein, several fuel pulverization models were used to calculate FFRD susceptibility in failed rods. Depending on the cladding failure/fuel pulverization model combination, the core-wide FFRD susceptibility during the postulated LOCA range from 0 to over 5,000 kg.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Results for the Aboveground Configuration of the Boiling Water Reactor Dry Cask Simulator

The thermal performance of commercial nuclear spent fuel dry storage casks is evaluated through detailed numerical analysis. These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full-sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask, in part by increasing the efficiency of internal conduction pathways, and also by increasing the internal convection through greater canister helium pressure. These same canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above- and below-ground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of above-ground and below-ground canistered dry cask systems. The purpose of the current investigation was to produce data sets that can be used to test the validity of the assumptions associated with the calculations used to determine steady-state cladding temperatures in modern dry casks that utilize elevated helium pressure in the sealed canister in an above-ground configuration.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Aluminum Clad Spent Nuclear Fuel Task 3: Sealed and Vented Systems Episodic Breathing and Gas Generation Modeling Plan

This document presents the test plan for Task 3 of the Aluminum Clad Spent Nuclear Fuel Long Term Dry Storage Technical Issues Action Plan, which covers modeling and predictions of the generations of corrosive gases within both sealed and vented, subject to episodic breathing, storage systems. This includes aluminum clad spent nuclear fuel (ASNF) that is currently or may in the future be held in dry storage for an extended period. The overall objectives of the Task 3 modeling plan are to: 1. Comprehensively model the combined effects of episodic breathing in vented storage systems/canisters using INL INTEC CPP-603 facility as an example. Both bulk gas phase radiolysis reactions and aluminum cladding surface gas generation reactions will be included into a 3-dimensional computational fluid dynamics (CFD) model to model and predict long-term evolutions of the temperature and concentrations of corrosive gases within unsealed canisters subjected to episodic breathing. 2. Comprehensively model sealed and inert canister storage system using DOE standard canister as an example. 3. Closely integrate with Task 1 and Task 2 for identification of main gas-generation reactions and kinetics and provide simulation results back to Task 1 and 2 to further refine experimental designs. This test plan lays out the technical activities that will be performed to achieve the test objectives described above. Simulations will be performed using the Idaho National Laboratory super computer Falcon.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗