Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “CLaDS”

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.

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

At least 1,729 records · Page 96
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Improved Fracture Models for Relocation Modeling

The BISON fuel performance code is being developed to provide a modern tool that has the flexibility to analyze a wide variety of fuel forms and to model conditions and phenomena that could not be represented in legacy tools. There are a number of motivations for this, including providing support for development of advanced fuel with improved accident tolerance for existing light water (LWR) reactors, improved understanding of mechanisms in fuel designs in current use in a wider variety of conditions, and facilitating the development of fuel for advanced reactor designs. To accomplish these goals, it is clear that BISON must rely on models of fuel behavior that are based on fundamental physical behavior, rather than on empirical correlations that represent that behavior in a simplified fashion. BISON still does employ many empirical models that were originally developed for other fuel performance codes, but efforts are underway to replace these with models that are more physically based. Radial relocation in LWR fuel is an example of a phenomenon that is currently represented by an empirical model, but which is ripe for replacement by physically based models. During normal operation, ceramic LWR fuel experiences significant fracturing that is driven by spatially nonuniform volumetric expansion. This occurs due to the significant thermal gradients within the fuel that occur in fresh and irradiated fuel, as well as nonuniform swelling due to fission products that occurs over longer-term irradiation. Fracture and fragmentation of fuel allows the outer radius of the fuel pellet to expand due to the loss of mechanical constraint and outward radial migration of fragments. This radial expansion has a significant effect on the fuel system response because it decreases the size of the gap between the fuel and cladding. This decreases the thermal resistance across that gap, which leads to decreased fuel centerline temperatures. This report documents recent work to improve the ability of BISON to model radial relocation using the extended finite element method (XFEM). These enhancements include the ability to include cohesive zone models, improved code architecture for propagating cracks based on fracture integrals, and improved handling of fuel and cladding interfaces with XFEM. These capabilities are demonstrated on a simulation of radial relocation including a surface interaction model that enforces residual opening.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Low length scale modeling of Cr precipitation in FeCrAl under thermal aging conditions

This report summarizes the progress on lower length scale modeling of Cr precipitation in FeCrAl alloys, which has been identified as a possible cladding material for accident tolerant fuels due to its excellent corrosion resistance via the formation of a protective, slow growing alumina surface layer. The precipitation of Cr-rich α' clusters in FeCrAl can lead to degraded mechanical properties such as the well-known “475oC embrittlement”, and should therefore be avoided to ensure cladding integrity. In FY16, a lattice kinetic Monte Carlo (LKMC) model was developed for FeCr binary alloys. Due to the assumption of constant bond energies, the previous model cannot correctly reproduce the low-temperature part of the Fe-Cr phase diagram. It is also not possible to model the anomalous ordering tendency and repulsive interactions between Cr atoms in α Fe at the dilute concentration regime using constant bond energies. To address these issues, a new local-environment-dependent bond energy formalism has been developed for Fe-Cr in FY17 that can simultaneously reproduce the negative mixing energy at low Cr concentrations and the positive mixing energy at high Cr concentrations, and the consequent large solubility of Cr in bcc Fe at low temperatures. Furthermore, the LKMC model has been extended to include Al to enable the assessment of Cr precipitation tendency in ternary FeCrAl alloys under thermal aging. Results from the present KMC simulations show good agreement with available experiments in the literature. This study is supported by the DOE NEAMS program Accident Tolerant Fuel project.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Preparation of Aluminum Oxide Films Under Water Exposure – Preliminary Report on 1100 Series Alloys

This report describes laboratory experiments performed to produce oxyhydroxide (oxide) films on aluminum 1100 series alloy substrates at various water/temperature conditions. The aluminum alloys selected in this program are types used on aluminum-clad research reactor fuel, and the water/temperature conditions are selected to produce film structures that would be prototypic to those formed on the fuel as a result of reactor exposure and post-discharge water and/or dry storage exposure. The experimental set-up and results to characterize the oxide morphology, thickness, structure, and chemical composition are reported. The results from this work will be compared to oxide characterizations of fuel cladding alloys that have been exposed to actual reactor and post-discharge service conditions. The aim is to develop a laboratory protocol to produce specimens for further studies for response under drying and radiation conditions. The work in this report is interim results and preliminary information prepared under Task 6, Milestone 6.1 in TTQAP SRNL-RP-2018-00610 as part of the action plan described in Ref. [1]. Additional experimental milestones are in progress and will be reported upon completion.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

FY17 end of year MARMOT materials model development update

This document summarizes the lower length scale progress made in FY17 in developing materials models for fuel performance modeling of oxide fuels and their implementation in the MARMOT code. The progress covers the development of a phase field recrystallization model for high-burnup structure formation in oxide fuels, a phase field model for studying stage-3 fission gas release in UO 2 , a phase-field model for irradiation-enhanced densification in UO 2 , a quantitative model for hydride morphology in Zirconium based cladding, and the calculation of hydrogen diffusivity under stress in various Zirconium based alloys as cladding candidates. Further development in the MARMOT code has been carried out as well to improve the multi-phase, multi-component phase field model, the representation of interfacial energy in the Kim-Kim-Suzuki model, and the description of interfacial stress. For most of the above, separated reports have been generated with detailed description. Here a brief summary is provided, with references pointing to the detailed reports to guide interested readers.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Oxyhydroxides on aluminum spent nuclear fuel: formation studies and removal practices to prevent radiolytic gas production

Aluminum-clad nuclear fuel has been at the forefront in research and commercial reactors. After being removed from the reactor, the spent nuclear fuel is placed in large-volume water basins (wet storage) that are constantly monitored for water quality to minimize corrosion/deterioration. The water provides both radiation shielding and cooling for the freshly removed fuel. It is highly desired to have a pathway for fuel to be transitioned to dry storage in sealed casks, so that spent nuclear fuel can be safely transported and dispositioned in a permanent repository, and basin capacity can be made available for temporary wet storage of fresher spent fuel. A challenge for safe, road-ready dry storage is the hydrated aluminum oxides (oxyhydroxides) that are prevalent on aluminum spent nuclear fuel that has been stored under water for years. If water is released from the oxyhydroxides during dry storage, it can create corrosion within the storage canister, and hydrogen can be generated through radiolysis of the oxyhydroxides, causing pressurization and flammability risks. The focus of this project is to develop knowledge of the oxyhydroxide growth and production of gases through radiolysis and to apply empirical models for predicting these phenomena. The expertise developed will be used to formulate conditions and parameters for oxyhydroxide removal/dehydration from fuel cladding through vacuum or forced-air gas drying, to prevent problems during permanent dry storage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Oxide Morphology of C26M at 300 - 600 °C

Iron-chromium-aluminum (FeCrAl) alloys are of interest to the nuclear materials community due to their resistance to high temperature steam oxidation under accident conditions. The present work investigates oxide formation at temperatures relevant to light water reactor cladding operation following extended aging to assess growth kinetics, chemical composition, and microstructure of oxide formation on an Oak Ridge National Laboratory developed Fe-12Cr-6Al- 2Mo-0.2Si-0.03Y alloy (designated ‘C26M’). Aging treatments were performed for 100-2000 hours in stagnant air at 300, 400, 500, and 600 °C, respectively. Oxide growth behavior under the investigated conditions follows a logarithmic time dependence. When the oxidization temperature is 400 °C or below, the oxide is amorphous. At 500 °C, crystalline α-Al 2 O 3 oxide film develops and the correlated logarithmic rate constant decreases significantly, indicating enhanced oxidation resistance of the formed oxide film. In addition, Mo segregation at grain boundaries has been observed when the aging temperature exceeds 500 °C. The results of this study can be viewed as an upper bounding result for potential oxide coarsening on the cladding inner surface during reactor operation.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Design and analysis of oxidation tests to inform FeCrAl ATF severe accident models

Iron-chromium-aluminum (FeCrAl) alloys are being considered as advanced fuel cladding concepts with enhan ced accident tolerance. At high temperatures, FeCrAl alloys have slower oxidation kinetics and higher strength compared with zirconium-based alloys. FeCrAl could be used for fuel cladding and spacer or mixing vane grids in light water reactors and/or as ch annel box material in boiling water reactors (BWRs). There is a need to assess the potential gains afforded by the FeCrAl accident-tolerant-fuel (ATF) concept over the existing zirconium-based materials employed today. To accurately assess the response of FeCrAl aloys under severe accident conditions, a number of FeCrAl properties and characteristics are required. These include thermophysical properties as well as burst characteristics, oxidation kinetics, possible eutectic interactions, and failure temperatures. These properties can vary among different FeCrAl alloys.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Interlaboratory Round Robin Study on Axial Tensile Properties of Tubular SiC/SiC Specimens

Extensive research and development work is being conducted at different governmental, university and industrial laboratories for developing accident tolerant fuel-cladding systems for light water reactors. Continuous SiC fiber-reinforced SiC matrix (SiC/SiC) composites is one of the candidate materials for replacing the traditional zirconium alloy cladding due to its high temperature stability, chemical inertness and stability under neutron irradiation. A comprehensive database on the statistics of the mechanical properties of SiC/SiC composite is much needed to estimate the failure probability in evolving thermo-mechanical conditions as well as rigorous analysis of the material’s behavior under normal and off-normal reactor conditions. In addition to the properties database, test standards, which ensure the unbiased characterization of the material, are needed. The mechanical properties database, particularly for tubular SiC/SiC specimens, is limited and standardized test methods are still under development.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Channel Gap Probe EMPIrE Report

EMPIrE aims to provide data for the "comprehension phase" of the European Low Enriched Uranium (LEU ) U-Mo Fuel Development program for the conversion of European civilian research reactors. The majority of the fuel performance data will be obtained during post-irradiation examinations (PIE) of the fuel. The primary EMPIrE fuel specimen design is based on uranium-molybdenum (U-Mo) alloy particles dispersed in an aluminum matrix clad in the European alloy AG3NE and in the form of mini-plate sized specimens (1" x 4" plates). Secondary EMPIrE specimens include U Mo monolithic fuel clad in European alloy AlFeNi. The experiment is designed in such a way that U-Mo samples achieve three distinct end-of-life burnup groupings. Irradiation has taken place at ATR (Advance Treatment Reactor). The fuel plates are contained in fuel capsules holding four plates each, with 2 plates at 2 levels. Once the capsules with the plates are irradiated, they have to be inspected during the outage in-between cycles. The capsules need to have visual inspection performed and channel gap widths measured prior to reinsertion. These inspections are to ensure that the plates have not deformed significantly. Obstruction of the coolant channels and restriction of the coolant flow are less of a concern in EMPIrE testing as the channels were made intentionally wide. The Channel Gap Probe (CGP) has been developed to measure channel gap widths, which serves as indirect measures of the fuel plate performance and deformation in the course of irradiation which enables the evaluation of the risk from continuing irradiation prior to reinsertion of the plates in reactor.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2018

Initial assessment of clad/base metal systems for MSR

In support of the development of new design and construction rules and analysis methods for cladded structure for use under the anticipated service lifetime and cyclic service conditions in Molten Salt Reactors (MSR), research work has been initiated at ORNL for assessment of physical, metallurgical and mechanical properties of clad/base metal systems. This report summarizes the results for the nickel cladded SS316H system in FY18.

36 MATERIALS SCIENCE↗
DOE OSTI2018

X-ray characterization of atomistic defects causing irradiation creep of SIC

Silicon carbide (SiC) fiber–reinforced SiC matrix composites have been widely studied for use in developing an accident-tolerant fuel (ATF) cladding for light water reactors (LWRs) in recent years because of their superior oxidation and irradiation resistance and low neutron absorption cross section. An important task is the establishment of the thermo-mechanical analysis capability to predict the in-pile performance and failure probability of the cladding. Irradiation creep is known to be a key property in modeling the stress state under irradiation. To better understand the radiation creep mechanism from an atomic point of view, this study conducted synchrotron-based X-ray diffraction (XRD) experiments on SiC neutron-irradiated with and without applied stress at the National Synchrotron Light Source-II (NSLS-II). This report presents progress of the XRD study and future plans. The important finding obtained was that the macroscopic creep strain and strain due to the lattice expansion were similar in case of neutron irradiation at 300°C to 0.1 displacement per atom (dpa), which indicates that the lattice strain is a major source of the transient creep strain.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Test Plan for the Horizontal Dry Cask Simulator

The thermal performance of commercial spent nuclear 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). Canistered dry storage cask systems rely on ventilation between the inner canister and the overpack to convect heat away from the canister to the surrounding environment for both horizontal and vertical configurations. 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 by increasing the internal convection through greater canister helium pressure. Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. While several testing programs have been previously conducted, these earlier validation studies did not integrate all the physics or components important in a modern, horizontal dry cask system. The purpose of the investigation described in this test plan is to produce data sets that can be used to benchmark the codes and best practices presently used to determine cladding temperatures and induced cooling air flows in modern horizontal dry storage systems. The horizontal dry cask simulator(HDCS) has been designed to generate this benchmark data and add to the existing knowledgebase. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure(MAWP)rating of 2,400 kPa at400 °C. An existing electrically heated but otherwise prototypic boiling water reactor(BWR), Incoloy-clad test assembly will be deployed inside of a representative storage basket and canister. An insulated sheet metal enclosure will be used to mimic the thermal properties of the concrete vault enclosure used in a modern horizontal storage system. Radial and axial temperature profiles along with induced cooling air flow will be measured for a wide range of decay powers and representative(and higher)cask pressures using various backfills of helium, argon, or air. The single assembly geometry with well-controlled boundary conditions simplifies computational requirements while preserving relevant physics. The proposed test apparatus integrates all the underlying thermal-hydraulics important to defining the performance of a modern horizontal storage system. These include combined-mode heat transfer from the electrically-heated assembly to the canister walls and the primarily natural-convective heat transfer from the canister to the cooling air flow passing through the horizontal vault enclosure. The objective of the HDCS is not to reproduce the performance of a commercial dry storage system for any given set of operational parameters. Rather ,the objective is to capture the dominant physics in a well-characterized test apparatus. The close coupling between the thermal response of the canister system and the resulting induced cooling air flow rate is of particular importance. While incorporating the best available information based on thermal-hydraulic scaling arguments as well as previous vertical testing, this test plan is subject to changes due to improved understanding or from as built deviations to designs. As-built conditions and actual procedures will be documented in the final test report.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Thermomechanical Design Optimization of Roll-to-Roll Manufactured Hybrid Metal Polymer Heat Exchanger

Metal based heat exchangers for ultra-low temperature (<100C) waste heat recovery (WHR) face high maintenance and material costs that surpass profits from recovery[1]. Low-cost polymer pipes suffer from poor thermal conductivity (~0.2 W/mK), and thus a low overall heat transfer coefficient. Thermal conductivity of polymers can be enhanced through systematic molecular alignment by stretching[2], chemical vapor deposition of extended polymers [3], surface grating [4], etc. However, these approaches suffer from scalability issues for WHR heat exchangers in terms of cost [3], and thermomechanical considerations [5]. In this work, we propose novel hybrid metal-polymer strips that are strategically designed to enhance thermal conductivity by cladding copper around polymer strips. The polymer strips with copper-clad edges are wound helically to form the pipes for a heat exchanger. Through finite element simulations, we optimize the design for enhanced thermal performance of the polymer-metal strips. We predict up to 20 % enhancement in overall heat transfer from a base polymer with a thermal conductivity of 0.2 W/mK. We show that the thermomechanical strains induced in the pipes can be reduced to < 0.2 % by strategically stacking and welding multiple polymer-metal strips. We verified the improved thermal performance of the roll-to-roll manufactured strips through a thermal conductivity measurement setup. The optimized design of hybrid metal-polymer pipes provides a scalable and cost-effective means for harvesting waste heat from ultra-low temperature sources. Thekdi, Arvind, and Sachin U. Nimbalkar. Oak Ridge National Lab, 2015 Shen, Sheng, et al. Nature nanotechnology 5.4 (2010): 251. Xu, Yanfei, et al. Science advances 4.3 (2018): 3031. Smith, M. K., Singh, V., Kalaitzidou, K., & Cola, B. A. (2015). ACS nano, 9(2), 1080-1088. Aymonier, C., Bortzmeyer, D., Thomann, R., & Mülhaupt, R. (2003). Chemistry of Materials, 15(25), 4874-4878.

C Rajagopal, Manjunath↗
DOE OSTI2019

Design and Characterization of High Thermal Conductivity Pipes from Roll-to- Roll Processed Hybrid Copper-Polymer Strips for Waste Heat Recovery

Metal based heat exchangers targeting low temperature (<150°C) waste heat recovery (WHR) face high maintenance and material costs [1] that result in long payback periods (>1 year). Low-cost polymer pipes suffer from poor thermal conductivity (~0.2 W/m.K), and thus a low overall heat transfer coefficient. Previous reports demonstrated thermal conductivity enhancement in polymers through systematic molecular alignment [2, 3, 4], or by metal micro-particle inclusions [5]. However, these approaches suffer from scalability issues for WHR heat exchangers in terms of cost [1, 3], and thermomechanical considerations [6, 7]. In this work, we propose hybrid metal-polymer strips that are strategically designed to enhance thermal conductivity by cladding copper around polymer strips. The polymer strips with copper-clad edges are wound helically to form the pipes for a heat exchanger. Through numerical simulations, we optimize the copper-polymer strip layouts to achieve high wall thermal conductivity (k > 1 W/m.K) and also have low volume fraction of copper (<70%) for the composite pipes. In copper-polymer composite heat exchangers, the bottlenecks for heat transfer are the thermal interfacial resistances of copper-copper and copper-polymer joints. We experimentally measured interfacial resistances to be ~1x10 -5 K.m 2 /W and ~4x10 -5 K.m 2 /W for copper-copper and copper-polymer joints, respectively. Using the measured interfacial resistances, we further optimized the design of the hybrid pipes to find high thermal conductivity (> 1 W/m.K) designs. We predict up to ~20 % enhancement in overall heat transfer relative to an all polymer (k = 0.2 W/m.K) heat exchanger system. Using a cohesive zone modelling (CZM) for predicting delamination at the weld joints, we find that our hybrid pipes can have operating internal pressures > 200 psi. The optimized designs of hybrid metal-polymer pipes provide a scalable and cost- effective means for harvesting waste heat from low temperature sources. Thekdi, A., and Nimbalkar, S. Oak Ridge National Lab (2014). Shen, Sheng, et al. Nature nanotechnology 5.4 (2010). Xu, Yanfei, et al. Science advances 4.3 (2018). Smith et al. ACS nano, 9(2), (2015). Sofian, N.M., et al., Journal of Thermoplastic Composite Materials (2001). Aymonier et al. Chemistry of Materials, 15(25), (2003). Tavman, I.H., Journal of Applied Polymer Science, 62(12), (1996).

C Rajagopal, Manjunath↗
DOE OSTI2019

Program Management Plan EnCore ® Accident Tolerant Fuel (ATF) Program

This Program Management Plan (PMP) is developed to support Westinghouse’s vision of developing and commercializing the Accident Tolerant Fuel. The near term goal of this Program is to deliver 2 fuel assemblies that consists of LTRs for Exelon Byron Unit 2 in spring of 2019. The longer term goal of this program is deployment of LTAs in 2022 of a mixture of coated zirconium cladding and SiC cladding with ADOPT and U3Si2 fuels. This is a high complexity program and is one of the largest research and development programs in Westinghouse and as such, the project scope is expected to change continuously. Integrated change control will be invoked as necessary. The Program consists of multiple funding streams (internal through Westinghouse innovation funding), external (through the DOE). The DOE funding consists of two phases: 1. Phase 2 – funded by DOE award number DE-NE0008222. 2. Phase 2B – funded by DOE – Westinghouse proposal reference # LTR-AMER-MKG-18-826, Rev. 1, August 23, 2018 (10/18 to 2/21), Award No. DE-NE0008824. 3. Cost share contributions from Westinghouse and program partners.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Considerations for Use of Am-241 for Heat Source Material for Radioisotope Power Systems.

There have been several efforts made over the last 60 years to examine the entire periodic table of the elements to determine the best isotope/element for use as a heat source for radioisotope power systems (RPS). Am-241 is typically not in the final group being considered due to its 432-year half-life. The window of isotopes considered has usually been kept to 15-100 year half-life. The criteria in down selecting an isotope for use as a heat source material have been: 1) radiation characteristics (easy to shield), 2) specific power (higher is better), 3) fuel form considerations (chemically compatible with cladding, chemically inert with respect to the environment, robust mechanically, high melting point), and 4) available at a reasonable cost and suitable quantities. There have been a few efforts examining Am-241 for use in RPS during the past 2 decades. None of these has advanced beyond comparison of known chemical or material properties. No heat sources have been made and none have been placed into RPS for testing. No tests have been conducted of the various americium oxides (AmO, Am 2 O 3 , AmO 2 nor nonstoichiometric phases between these oxides) for their behavior with the environment to confirm inert behavior or compatibility with cladding materials. From a crystallographic point of view americium and oxygen can form crystal phases that exhibit face-centered cubic, monoclinic, cubic or tetragonal characteristics. A simple point to make here is that the chemistry of americium and oxygen is complicated and not well understood at this point compared to plutonium and oxygen.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2019

Characterization of 14YWT & ODS Base Metals

Nanostructured ferritic alloys (NFA) with oxide dispersion strengthening are being developed for next generation reactor applications. Applications such as transmutation of long lived isotopes in used fuels require cladding materials that can withstand high irradiation doses, potentially in the hundreds of displacements per atom (dpa). Currently, however, there are no joining processes that retain the strength and fine distribution of oxide precipitates characteristic of ODS alloys. As part of a larger effort to adapt a solid state capacitive discharge welding technique for use with ODS cladding, presented will be characterization of the initial materials including 14YWT and MA956 ODS alloys.

36 MATERIALS SCIENCE↗