Search NASA⌕ Search

SEARCH · Search NASA

Results for “nuclear cladding”

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 631 records · Page 35
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Enhanced chemical CRUD decontamination combined with electrokinetic process - 15242

This study has tested about the deposition and dissolution of Chalk River Unidentified Deposit (CRUD), and investigated the remediation of CRUD. CRUD is an accumulated corrosion product on external fuel rod cladding surfaces and composed of either dissolved ions or solid particles such as Ni and Co. As nuclear power plants are getting older, decontaminating process has been concerned due to increased contamination. Corrosion product can be activated through nuclear reaction to form radionuclide such as Co-60. Because the CRUD is formed at cladding surface and radionuclides can be contained in CRUD, it can affect to human health and environment. Therefore CRUD needs to be removed safely during NPP's decontaminating process by efficient. In order to decontaminate the CRUD, first thing it has to be dissolved and removed. Deposition behavior of corrosion products has been well studied, but the study about behavior of CRUD dissolution is required in various environments. In this study, the CRUD synthesis, the dissolution method of CRUD, and the CRUD decontamination method was investigated. There are three major approaches the remediation of CRUD. First one is well synthesized the CRUD, second one is finding the optimum condition for CRUD dissolution, and last one is applying the Electro-Kinetic (EK) process to CRUD remediation. Synthesis of simulated CRUD for the decontamination experiments was successful. The combination of EK process with chemical decontamination of CRUD proved to be a more economical method for removing CRUD; it can be directly applied to the NPP decommissioning and decontamination process, and it can decrease the volume of secondary radioactive waste as well. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Critical heat flux impact due to cladding surface condition

This paper demonstrates how the cladding surface condition can impact fuel rod heat transfer and critical heat flux (CHF) performance at Pressurized Water Reactor (PWR) type conditions. The cladding surface that is investigated here is a surface with a crud deposit similar to ones observed in a PWR reactor. Tests are performed in the Westinghouse WALT (Westinghouse Advanced Loop Tester) loop test facility where a single ZIRLO clad electrically heated rod with approximately one-foot length is tested at PWR thermal hydraulic conditions and chemistry. Tests are performed for a normal clean rod surface, an oxide surface and with CRUD deposits to investigate the impact on fuel rod heat transfer and CHF performance. The CRUD deposit is characterized for thickness, chimney size and porosity. Test results are presented for the different cladding surface conditions along with comparisons to evaluations performed in the literature for different surfaces at pool boiling conditions. The results indicate that the WALT loop test facility can produce CHF test results which are repeatable for the same heater rod and for different rods with the same surface condition. CHF test results for a ZIRLO cladding surface with a CRUD deposit near 20 microns typical of PWR deposits show similar CHF performance to clean rod surfaces. A noticeable increase in CHF performance was observed for a ZIRLO cladding surface with a thermal oxide layer based on one CHF measurement.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Modeling of complex wear behavior associated with grid-to-rod fretting in light water nuclear reactors

One significant concern in the operation of light water nuclear reactors is the fretting wear damage to fuel cladding from flow-induced vibrations. For years, research on the grid-to-rod fretting (GTRF) phenomena has been underway in countries where nuclear power production is a significant industry. Under the auspices of the U.S. Department of Energy Consortium for Advanced Simulation of Light Water Reactors, an effort has been underway to develop and test an engineering wear model for zirconium alloy fuel rod cladding against a supporting grid. Furthermore, the multi-stage model accounts for oxide layers and wear rate transitions. Our paper describes the basis for a GTRF engineering wear model, the physical significance of the wear factor it contains, and recent progress toward model validation based on a fretting wear testing apparatus that accounts for coolant temperature, pressure, and the presence of periodic impacts (gaps) in grid/rod contact.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

RESEARCH AND DEVELOPMENT ACTIVITIES AT SAVANNAH RIVER SITE'S H CANYON FACILITY

The Savannah River Site's (SRS) H Canyon Facility is the only large scale, heavily shielded, nuclear chemical separations plant still in operation in the U.S. The facility's operations historically recovered uranium-235 (U-235) and neptunium-237 (Np-237) from aluminum-clad, enriched-uranium fuel tubes from Site nuclear reactors and other domestic and foreign research reactors. Today the facility, in conjunction with HB Line, is working to provide the initial feed material to the Mixed Oxide Facility also located on SRS. Many additional campaigns are also in the planning process. Furthermore, the facility has started to integrate collaborative research and development (R&D) projects into its schedule. H Canyon can serve as the appropriate testing location for many technologies focused on monitoring the back end of the fuel cycle, due to the nature of the facility and continued operation. H Canyon, in collaboration with the Savannah River National Laboratory (SRNL), has been working with several groups in the DOE complex to conduct testing demonstrations of novel technologies at the facility. The purpose of conducting these demonstrations at H Canyon will be to demonstrate the capabilities of the emerging technologies in an operational environment. This paper will summarize R&D testing activities currently taking place in H Canyon and discuss the possibilities for future collaborations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Evaluation of Increased Peak Temperatures for Spent Fuel Cladding Performance during Dry Storage (Rev.1)

NUREG-2214, “Managing Aging Processes in Storage (MAPS) Report,” (2019) documents the technical basis for age-related degradation mechanisms on spent nuclear fuel assemblies during dry storage operations. The technical bases have been reviewed, and hoop cladding stress calculations were conducted using the Fuel Analysis under Steady-state and Transients (FAST) code at elevated normal condition temperatures of 425°C and 450°C, and at an off-normal and accident conditions temperature of 600°C. The NUREG-2214 conclusions are substantiated by the referenced literature. The calculations demonstrate that the aging mechanisms primarily driven by hoop stress such as hydride reorientation, delayed hydride cracking, thermal and athermal (low-temperature) creep, and localized mechanical overload are either not credible or do not compromise the fuel assembly’s performance for increased peak normal condition temperatures of 425°C or 450°C during the up to 60-year dry storage period. Other aging mechanisms considered are also not credible for compromising the fuel assembly’s performance during 60-year dry storage with a peak temperature of 400ºC as specified in ISG – 11 Rev. 3, and the same is indicated for the increased peak normal condition temperatures of 425°C or 450°C.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Impact of LWR Assembly Structural Features on Cladding Burst Behavior under LOCA Conditions

This report builds on the previous work by assessing the impact of assembly structural features on cladding balloon and burst behavior. The BISON fuel performance code was used to generate two dimensional RZ meshes containing structural features commonly used in nuclear fuel assemblies. The results were then compared to the cladding burst temperature and balloon strain results from the previous work. A mesh sensitivity study was performed to ensure spacer grid effects were appropriately considered, resulting in a more refined mesh than the previous study. The balloon deformation and burst times of the cladding were compared. Consideration was also given to the effect of rod initial pressure. Overall, the BISON simulations indicated the spacer grids can locally influence cladding performance, however, the impact on the cladding performance was minor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Comparisons of Peak Cladding Temperature and Effective Thermal Conductivity: CFD Simulation on PWR Spent Fuel Assemblies - 16278

When spent fuel assemblies from the reactor of nuclear power plants (NPPs) are transported or stored, the assemblies are exposed to a variety of environments that can affect the peak cladding temperature of spent fuel assemblies. There are four methods to calculate the peak cladding temperature of spent fuel assemblies in a cask: Manteufel and Todreas's two-region model, Bahney Lotz's effective thermal conductivity model, Wooton-Epstein correlation, and computational fluid dynamics (CFD) simulation. The CFD simulation based on the FLUENT code is used to calculate the peak cladding temperature and effective thermal conductivity of a spent fuel assemblies used in Republic of Korea NPPs: 14x14, 16x16 ACE7, 16x16 PLUS7 and 17x17 PWR spent fuel assemblies. The CFD simulation results show that the effective thermal conductivity of the 16x16 ACE7 was more conservative than those for the other assemblies. Therefore the effective thermal conductivity using 16x16 ACE7 is used for thermal evaluation of the KORAD's developing cask which can load 14x14, 16x16 ACE7, 16x16 PLUS7 and 17x17 spent fuel assembly. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Evolution of pores in the fuel cladding of the electrogenerating channel

The results of reactor tests of carbonitride fuel in a monocrystalline cladding from a molybdenum-based alloy that were carried out earlier on an experimental setup Ya-82 for 8 to 300 h at a temperature of ∼1500°C can be used in order to prove the operational reliability of fuel elements in the design of a megawatt nuclear power plant for a spacecraft. A raster image of the surface of a sample of cladding shows that the interfaces between the layers are decorated by pores. This result is explained in this work by the theory of coalescence. The mechanisms responsible for the evolution of pores taking place at the parameters of a Ya-82 setup are considered. The effect of decoration by pores of the interfaces between the layers of a sample of cladding of the electrogenerating channel by carrying out reactor tests is explained. The dependence of the average radius of pores on the duration of the experiment is obtained. An evaluation of the average sizes of pores arising under the conditions of the experiment gives a value of ∼2 μm, which is in agreement with the experimental data. A computational study of swelling of the cladding material in the process of irradiation is performed. Predictive estimates of the behavior of the porous system and swelling of the cladding material for a megawatt class nuclear power plant are made.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Handbook on the Material Properties of FeCrAl Alloys for Nuclear Power Production Applications (FY18 Version: Revision 1)

FeCrAl alloys are a class of alloys that have seen increased interest for nuclear power applications including accident tolerant fuel cladding, structural components for fast fission reactors, and as first wall and blanket structures for fusion reactors. FeCrAl alloys are under consideration for these applications due to their inherent corrosion resistance, stress corrosion cracking resistance, radiation-induced swelling resistance, and high temperature oxidation resistance. A substantial amount of research effort has been completed to design, develop, and begin commercial scaling of FeCrAl alloys for nuclear power applications over the past half a century. These efforts have led to the development of an extensive database on material properties and process knowledge for FeCrAl alloys but not within a consolidated format. The following report is the first edition of a materials handbook to consolidate the current state of knowledge on FeCrAl alloys for nuclear power applications. This centralized database focuses solely on wrought FeCrAl alloys, oxide dispersion strengthened alloys, although discussed in brief, are not covered. Where appropriate, recommendations for applications of the data is provided and current knowledge gaps are identified.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Overview of the U.S. DOE advanced fuels campaign

Advanced Fuel Campaign (AFC) has responsibility to develop advanced nuclear fuel technologies for the Fuel Cycle Research and Development (FCRD) program. AFC focus areas include 4 topics: advanced LWR fuels, actinide transmutation, research infrastructure development for the study of new nuclear fuels and advanced modeling and simulation. This paper reviews the latest technical advances in the 4 topics. Fuel and cladding materials under investigation include the following general classes of technologies: iron-based cladding, ceramic cladding, coated Zirconium alloy cladding, molybdenum-based cladding, high-density fuel, fully ceramic matrix fuel, and metallic alloy-based nuclear fuel. Concerning actinide transmutation fuel alloy additions to immobilize the lanthanide fission products, sodium-free annular fuel concepts, and cladding coatings and liners for mitigation of fuel-cladding chemical interaction have been investigated. Direct casting of metallic fuel into a zirconium liner was demonstrated. A Talos Transmission Electron Microscope was acquired for the study of advanced nuclear fuels and materials. The conceptual and preliminary designs for a static capsule, a flowing water loop, and a flowing sodium loop to support future transient testing needs have been completed. AFC interfaces with the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program to develop multi-scale, multi-physics fuel performance codes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Characterization of Ion Implantation and Irradiation Effects in Interface-Dominated Metallic Composites

The candidate structural and fuel cladding materials for the next-generation nuclear power reactors must possess excellent irradiation damage tolerance, and high mechanical and thermal stability that present a challenge for materials design. Here we report an approach to engineer interphase boundaries in metallic composites to achieve superior resistance to void swelling and prevent loss of plastic deformability after irradiation. Model composite systems of face-centered-cubic (Cu) and body-centered-cubic (Nb, Mo or V) metals are used to elucidate the effects of boundary structure and spacing on the damage accumulation and mechanical properties after irradiation. Transmission electron microscopy (TEM) and nano-mechanical testing are used to characterize the irradiation response. A recent development in in situ straining in high-resolution TEM to measure lattice strains and thereby estimate local stresses corresponding to dislocation nucleation and migration is discussed as a capability that in future can enable quantitative characterization of defect interactions in irradiated materials. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2017

Nuclear Energy Advanced Modeling and Simulation (NEAMS) Accident Tolerant Fuels High Impact Problem: FeCrAl Modeling Capabilities

Iron-chromium-aluminum alloys are a potential candidate replacement for zirconium-based claddings in light water reactor fuel rods due to their increased strength (lower creep rates and higher yield stress) and slower oxidation kinetics with steam. Their larger neutron absorption cross-section necessitates thinner claddings and larger fuel pellets to maintain the same as-fabricated fuel-to-clad gap. The aggressive development schedule of the United States Department of Energy's (DOE) Advanced Fuels Campaign (AFC) to have potential accident tolerant fuels (ATF) concepts in lead test rods or assemblies by 2022 has necessitated the need for advanced modeling and simulation techniques to investigate the behavior of these materials under normal operating and accident conditions. This report presents the modeling capabilities for FeCrAl claddings at the engineering (continuum) scale developed as part of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) high impact problem (HIP). Sensitivity studies of select material parameters have been completed to determine their significance in predictions of important fuel performance parameters such as fuel centerline and surface temperatures, cladding stresses and strains, fission gas released, and rod internal pressure. Comparative analyses of FeCrAl and Zircaloy-4 claddings for separate effects conditions are presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Improvement of mechanistic fuel-cladding chemical interaction modeling in BISON

This report describes work performed during FY2024 under the auspices of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program to inform and improve mechanistic models of fuel-cladding chemical interaction (FCCI) in metallic fuel. For fuel-side FCCI, atomistic simulations were performed to determine the diffusivity of iron (Fe) in the $\alpha$ and $\gamma$ phases of uranium (U). A model of liquid penetration of cladding due to melting of the fuel-side FCCI region was updated to account for the finite size of the FCCI region, and the model was validated through comparison with tests performed in the Fuel Behavior Test Apparatus (FBTA). For cladding wastage formation, a reduced-order model was improved by comparison with a multi-scale mechanistic model to better quantify the ROM parameters.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Steam Oxidation Testing in the Severe Accident Test Station

Since 2011, Oak Ridge National Laboratory (ORNL) has been conducting high temperature steam oxidation testing of candidate alloys for accident tolerant fuel (ATF) cladding. These concepts are designed to enhance safety margins in light water reactors (LWR) during severe accident scenarios. In the US ATF community, the Severe Accident Test Station (SATS) has been evaluating candidate materials (including coatings) since 2012. Compared to the current UO 2 /Zr-based alloy fuel system, alternative cladding materials need to offer slower oxidation kinetics and a smaller enthalpy of oxidation in order to significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident. The steam oxidation behavior of candidate materials is a key metric in the evaluation of ATF concepts and also an important input into models. However, prior modeling work of FeCrAl cladding has used incomplete information on the physical properties of FeCrAl. Also, the steam oxidation data being collected at 1200°-1700°C is unique as no prior work has considered steam oxidation of alloys at such high temperatures. In some cases, the results have been difficult to interpret and more fundamental information is needed such as the stability of alumina in flowing steam at 1400°-1500°C. This report summarizes recent work to measure the steam oxidation kinetics of candidate alloys, the evaporation rate of alumina in steam and the development of integral data on FeCrAl compared to conventional Zr-based cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Amplicon Sequencing Reveals Microbiological Signatures in Spent Nuclear Fuel Storage Basins

Water quality is an important determinant for the structural integrity of alloy cladded fuels and assemblies during long-term wet storage. Detailed characterization of a water filled storage basin for spent nuclear reactor fuel was performed following the formation and proliferation of an amorphous white flocculent. White precipitant was sampled throughout the storage basin for chemical and spectroscopic characterization, and eDNA was extracted for pyrosequencing of bacterial rRNA gene diversity. Accordingly, spectroscopic analyses indicated the precipitant to be primarily amorphous to crystalline aluminum (oxy) hydroxides with minor associated elemental components including Fe, Si, Ti, and U. High levels of dissolved carbon were co-localized with the precipitant relative to bulk water. Bacterial densities were highly variable between sampling locations and with depth; cell numbers (log scale) ranged from 5.6 to 4.89 cells / mL. Bacterial diversity that was physically associated with the aluminum (oxy) hydroxide complexes exceeded an estimated 4,000 OTUs / amplicon library (3% cutoff) and the greatest percent majority of sequences were aligned to the families Burkholderiales (23%), Nitrospiraceae (23%), Hyphomicrobiaceae (17%), and Comamonadaceae (6%). We surmise that episodic changes in the physical and chemical properties of the basin contribute to the polymerization of aluminum (oxy) hydroxides, which in turn can chemisorb nutrients, carbon ligands and bacterial cells from the surrounding bulk aqueous phase. As such, these precipitants should establish favorable microhabitats for bacterial colonization and growth. Comparative analyses of 16S rRNA gene amplicon libraries across diverse environmental landscapes were performed and microbiological signatures unique to the spent nuclear fuel storage basin environment were revealed. These insights could spur the development of tractable bioindicators that are specific of and diagnostic for water quality at discrete locations and finer scales of resolution, marking an important contribution for improved water quality and management of spent nuclear fuel storage facilities.

aluminum↗
DOE OSTI2018

Mechanism and Conditions for Continuous Hydride Layer Formation in Zirconium Cladding

Local hydrogenation of zirconium fuel-element cladding, resulting in the formation of continuous hydrides, as a rule occurs when coolant or other hydrogen-containing compound finds its way beneath the fuel-element cladding irradiated in the core of a water moderated and cooled reactor. In this article, the set of processes occurring during hydrogenation of zirconium cladding is analyzed, and estimates are given for the parameters of these processes. The conditions and mechanism of formation of a continuous hydride over the thickness of zirconium cladding are established. It is shown that the duration of the nucleation of the hydride is determined primarily by the time to local destruction of the oxide film on the inner surface of the zirconium cladding of the fuel elements. The shortest possible time for continuous hydride to grow through the thickness of fuel-element cladding is determined.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗