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At least 235 records · Page 13
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Drying and Analysis of Aluminum (Oxy)hydroxide Films for Dry Storage of Aluminum-Clad Spent Nuclear Fuels

In aluminum-clad spent nuclear fuels, an (oxy)hydroxide layer on the surface of the cladding hosts chemisorbed water formed during reactor and post-discharge exposure to water. Any residual water is susceptible to generating hydrogen via radiolysis, which can be a risk associated with dry fuel storage. Engineering-scale forced helium dehydration (FHD) and vacuum drying tests were conducted on mock-up fuel assemblies that included corroded aluminum surrogate plates to assess the removal of bulk and chemisorbed water. Thermogravimetric analysis was performed on samples of the surrogate plates, both undried control samples used to determine onset temperatures associated with a phase change occurring in the oxide layer and samples from drying tests used to determine the effectiveness of each drying method. Both vacuum drying and FHD processes were capable of removing bulk water. However, FHD was determined to provide additional drying capabilities, including partial removal of chemisorbed water from bayerite due to the higher temperatures during drying. In conclusion, the temperature threshold for partial dehydroxylation of the oxide layer was determined to be around 220°C, meaning any drying methods attempting to remove chemisorbed water must exceed 220°C.

chemisorbed water↗
DOE OSTI2022

Developing solid-surface plasma facing components for pilot plants and reactors with replenishable wall claddings and continuous surface conditioning. Part A: concepts and questions

Abstract It is estimated that pilot plants and reactors may experience rates of net erosion and deposition of solid plasma facing component (PFC) material of 10 3 –10 5 kg yr −1 . Even if the net erosion (wear) problem can be solved, the redeposition of so much material has the potential for major interference with operation, including disruptions due to so-called ‘unidentified flying objects (UFOs)’ and unsafe dust levels. The potential implications appear to be no less serious than for plasma contact with the divertor target: a dust explosion or a major UFO-disruption could be as damaging for an actively-cooled deuterium-tritium (DT) tokamak as target failure. It will therefore be necessary to manage material deposits to prevent their fouling operation. This situation appears to require a fundamental paradigm shift with regard to meeting the challenge of taming the plasma–material interface: it appears that any acceptable solid PFC material will in effect be flow-through , like liquid–metal PFCs, although at far lower mass flow rates. Solid PFC material will have to be treated as a consumable , like brake pads in cars. ITER will use high-Z (tungsten) armor on the divertor targets and low-Z (beryllium) on the main walls. The ARIES-AT reactor design calls for a similar arrangement, but with SiC cladding on the main walls. Non-metallic low-Z refractory materials such as ceramics (graphite, SiC, etc) used as in situ replenishable, relatively thin—of order mm—claddings on a substrate which is resistant to neutron damage could provide a potential solution for the main walls, while reducing the risk of degrading the confined plasma. Separately, wall conditioning has proven essential for achieving high performance. For DT devices, however, standard methods appear to be unworkable, but recently powder droppers injecting low-Z material ∼continuously into discharges have been quite effective and may be usable in DT devices as well. The resulting massive generation of low-Z debris, however, has the same potential to seriously disrupt operation as noted above. Powder droppers provide a unique opportunity to carry out controlled studies on the management of low-Z slag in all current tokamaks, independent of whether their protection tiles use low-Z or high-Z material.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Aluminium composites prepared by laser cladding assisted by friction stir processing

Abstract A laser cladding and friction stir processing hybrid method was employed to produce an Al matrix composite layer. The microstructure, phase composition, microhardness and conductivity of the composites were investigated. A laser cladding layer with a thickness of approximately 200 μ m was prepared on a 1060 aluminium plate and it was broken up and distributed on the Al matrix after friction stir processing. The particle/Al interfaces exhibited extremely good interfacial integrity. Microstructural observations revealed that an obvious in situ reaction occurred at the particle/Al interfaces, which effectively improved the bonding between the reinforcement phase and the matrix. TEM analysis and selected area diffraction enabled the identification of the intermetallic compounds and confirmed them to be Al 5 Fe 2 and Al 3 Fe. The average microhardness values of the friction stir processed composites reached approximately 85 HV. The electrical resistivity of the friction stir processed composites is slightly higher than that of the aluminium matrix.

Wei, Yanni (ORCID:0000000244443642)↗
DOE OSTI2020

Copper-Clad Aluminum Windings as an Alternative Conductor for High-Power Electric Vehicle Wireless Charging

In this paper, copper and copper-clad aluminum winding conductors were compared. The effective resistances as functions of frequency for the cored coils were compared using the 3D finite element model frequency sweep analysis for low, medium, and high frequencies. To compare equivalent designs, both coils were designed to have similar direct-current resistance. Benefits of the copper-clad aluminum over the copper coils are presented.

Wojda, Rafal↗
DOE OSTI2025

Mechanical response of HFIR-irradiated M5FRAMATOME cladding under simple and complex loading conditions

This study investigates the mechanical behavior of High Flux Isotope Reactor (HFIR) irradiated M5FRAMATOME cladding under simple and complex loading conditions through axial tensile and reversible cyclic bending. Axial tension specimens were pre-machined prior to HFIR irradiation while cyclic bend specimens were inserted as intact tubes. Tests articles were neutron-irradiated to 4 and 16 dpa, and specimens were tested at ORNL's hotcell facilities. The axial tension tests were conducted under constant displacement control, and the reversible cyclic bend tests were performed using ORNL's Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) apparatus. Results showed that mechanical response of Cr-coated and uncoated M5FRAMATOME cladding were similar and independent of irradiation dose for axial tension tests, while Cr-coated specimens’ reversible cyclic bend behavior differed from uncoated counterparts. For all tests, irradiation temperature showed a significant impact on the mechanical behavior. Below 280°C, all axial tensile specimens whether coated or not behaved similarly. Above 280°C, YS and UTS showed decrease with increasing irradiation temperature. A similar behavior was also observed in cyclic bend tests as well. The mechanical damage during cyclic bend tests was linked to damage accumulation in unirradiated Cr-coated zircaloy-4 specimens, and the effect of irradiation temperature was related to changing characteristics of defect mobility during high temperature irradiation.

Cinbiz, Nedim [ORNL] (ORCID:0000000346268515)↗
DOE OSTI2025

Irradiation Vehicles for Evaluating SiC/SiC Cladding Bowing Under Neutron Flux Gradients

Silicon carbide fiber–reinforced silicon carbide matrix (SiC/SiC) composites are among the most promising candidates for long term accident-tolerant nuclear fuel cladding. A key challenge related to their deployment is lateral bowing caused by differential radiation-induced swelling under dose or temperature gradients, which could obstruct coolant flow or interfere with control rod/blade movements. Although previous modeling efforts have predicted bowing behavior in light-water reactor (LWR) environments, experimental validation remains limited, especially at prototypic LWR temperatures. This study addresses that gap by irradiating six reduced-length SiC/SiC cladding tubes (~600 mm) in the High Flux Isotope Reactor (HFIR) at ~300°C, which is representative of LWR conditions. The tubes will be housed in a sealed vessel with an inert gas gap to maintain target temperatures and prevent direct coolant contact. Arranged in three pairs, each set will receive a different radiation dose (low, medium, high), with the central pair receiving ~0.1 displacements per atom (dpa)—the expected dose for peak bowing. The experiment will determine the dose-dependent bowing behavior and validate predictive finite element models. In this work, the tubes are freely suspended from pins to allow for unconstrained bowing; however, we present a concept for introducing localized constraints to represent grid spacer effects. Post-irradiation examination will include swelling measurements and profilometry to assess bowing and compare the results with model predictions. This work aims to confirm the conditions under which maximum bowing occurs so as to improve the reliability of SiC/SiC performance models in nuclear applications.

Russell, Nick [ORNL] (ORCID:0000000296099820)↗
DOE OSTI2023

Broadband integrated polarization splitter and rotator using subwavelength grating claddings

We present a broadband integrated photonic polarization splitter and rotator (PSR) using adiabatically tapered coupled waveguides with subwavelength grating (SWG) claddings. The PSR adiabatically rotates and splits the fundamental transverse-magnetic (TM 0 ) input to the fundamental transverse-electric (TE 0 ) mode in the coupler waveguide, while passing the TE 0 input through the same waveguide. The SWGs work as an anisotropic metamaterial and facilitate modal conversions, making the PSR efficient and broadband. We rigorously present our design approaches in each section and show the SWG effect by comparing with and without the SWG claddings. The coupling coefficients in each segment explicitly show a stronger coupling effect when the SWGs are included, confirmed by the coupled-mode theory simulations. The full numerical simulation shows that the SWG-PSR operates at 1500–1750 nm (≈250 nm) wavelengths with an extinction ratio larger than 20 dB, confirmed by the experiment for the 1490–1590 nm range. The insertion losses are below 1.3 dB. Since our PSR is designed based on adiabatical mode evolution, the proposed PSR is expected to be tolerant to fabrication variations and should be broadly applicable to polarization management in photonic integrated circuits.

47 OTHER INSTRUMENTATION↗
DOE OSTI2024

Brittle failure analysis and modeling of high-burnup PWR fuel cladding alloys

The aim of this research is the development of methods for predicting mechanical behavior and identification of limiting conditions to prevent brittle failure of high-burnup (HBU) pressure water reactor (PWR) fuel cladding alloys. A finite element (FE) model of the ring compression test (RCT) was created to analyze the failure behavior of zirconium-based alloys with radial hydrides during the RCT. An elastic-plastic material model describes the zirconium alloy. The stress-strain curve needed for the elastic-plastic material model was derived by inverse finite element analyses. Cohesive zone modeling is used to reproduce sudden load drops during RCT loading. Based on the failure mechanism in non-irradiated ZIRLO (R) claddings, a micro-mechanical model was developed that distinguishes between brittle failure along hydrides and ductile failure of the zirconium matrix. Two different cohesive laws representing these types of failure are present in the same cohesive interface. The key differences between these constitutive laws are the cohesive strength, the stress at which damage initiates, and the cohesive energy, which is the damage energy dissipated by the cohesive zone. Statistically generated matrix-hydride distributions were mapped onto the cohesive elements and simulations with focus on the first load drop were performed. Computational results are in good agreement with the RCT results conducted on high-burnup M5 (R) samples. It could be shown that crack initiation and propagation strongly depend on the specific configuration of hydrides and matrix material in the fracture area.

Simbruner, Kai↗
DOE OSTI2021

Visual Examination of Aluminum Containers for Extended Wet Storage of Non-Aluminum-Clad Spent Nuclear Fuel (FY 2021)

The goal of the Augmented Monitoring and Condition Assessment Program (AMCAP) program is to provide a condition assessment of the storage containers for non-aluminum-clad spent nuclear fuel (NASNF) in L Basin and demonstrate continued safe storage of all NASNF pending retrieval for ultimate disposition. The storage configuration of NASNF is vulnerable to inside-out corrosion attack. The approach being used in AMCAP is to develop and deploy examination methods for remote, in-situ inspection of the bundle storage in VTS and for OSC storage in the OSC racks. A method for Visual Examination (VE) using video-recording cameras was readied as part of the full-scale mockup development, and was deployed in L Basin for the initial condition assessment work. This report describes the development and the deployment of the VE for inspection of selected bundles in L Basin. The first deployment of the VE system was for visually examining five spent fuel bundles. Bundles were chosen based on their vulnerability in terms of inner fuel contents, history, storage duration, storage configuration, and possibility of inside out corrosion due to galvanic coupling of fuel cladding with the aluminum bundle wall.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Sister Rod Destructive Examinations (FY2021) Appendix B: Segmentation, Defueling, Metallographic Data and Total Cladding Hydrogen

As a part of the DOE-NE High Burnup Spent Fuel Data Project, Oak Ridge National Laboratory (ORNL) is performing destructive examinations (DEs) of high burnup (HBU) (>45 GWd/MTU) spent nuclear fuel (SNF) rods from the North Anna Nuclear Power Station operated by Dominion Energy. The SNF rods, called sister rods or sibling rods are all HBU and include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin (LT) Zirc-4, ZIRLO®, and M5®. Appendix B provides detailed information regarding defueling activities, metallographic imaging and measurements, and total cladding hydrogen measurements.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Cladding Degradation Model

This report begins with an evaluation of cladding degradation mechanisms deemed important to assessing barrier capability. Unlike similar efforts done in the past, this evaluation accounts for the hypothetical conditions associated with direct dual-purpose canister (DPC) disposal, including conditions resulting from a postulated, in-package, steady-state criticality event. A total of 16 cladding degradation mechanisms are examined assuming direct disposal of DPCs in two different hypothetical repositories: a saturated repository in shale and an unsaturated repository in alluvium.

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

Report summarizing demonstration of thermal conductivity measurement methods for coated zirconium cladding

Thermal diffusivity measurements on Zirconium-based cladding materials have historically been a challenge due to the difficulty to measure on specimens with curved geometries, including nuclear grade Zircaloy cladding materials. In this work, we first used Laser Flash Analysis methods to measure the thermal diffusivity of Zircaloy tubes, which show good agreement with Zircaloy plates in this work, as well as previously published data. The consistent results proved the applicability of the Laser Flash Analysis setup for investigating thermal diffusivity of Zircaloy tubes. We further investigate the hydrogen effect on thermal diffusivity of Zircaloy tubes. Hydrogen plays significant roles in temperature dependent thermal diffusivity trends of Zircaloy tube. For low hydrogen concentration, where hydride dissolved into single phase α-Zr at higher temperature, thermal diffusivity is relatively unchanged at low temperature at two phase regions (α-Zr and δ-hydride), and it increases as a function of temperature at single phase α-Zr region. For high hydrogen concentration, where phase transformation (α-Zr + δ-hydride → α-Zr + β-Zr) occurs at 567 °C, thermal diffusivity decreases as a function of temperature below 567 °C before the transformation (α-Zr and δ-hydride), while increase at higher temperature after the transformation (α-Zr and β-Zr).

36 MATERIALS SCIENCE↗
DOE OSTI2022

BISON Validation to In-Situ Cladding Burst Test and High Burnup LOCA Experiments

The process to develop and qualify nuclear fuels for commercial nuclear application requires fundamental material development, characterization, and design; out-of-pile testing on unirradiated materials; integral fuel rod irradiations, testing, and post irradiation examinations (PIEs); and transient analyses. The historical approach depends on the generation of large empirical datasets and series of integral fuel rod irradiations, and this approach ultimately takes approximately 20 years—or sometimes longer—to acquire data through extensive sequential testing. Thus, the qualification and eventual deployment of new fuel systems constitute a long and drawn-out process. However, recent technological advancements have provided researchers the opportunity to perform out-of-cell, in-situ measurements to assess material performance for the duration of the experiment. One such example of this capability is the use of digital image coordination and thermal imaging to assess Zircaloy cladding performance under a simulated loss-of-coolant accident (LOCA) transient condition. In general, the in-situ measurements provide high-fidelity strain, strain rates, and temperature surface maps. This is critical for the United States nuclear industry, which is actively developing a technical basis to support extending the peak rod average burnup from 62 to ~75 GWd/tU and the deployment of accident-tolerant fuel. However, the Nuclear Regulatory Commission, through their research information letter, outlined a number of technical issues for the industry to address before extending burnup. One topic of specific interest is understanding the cladding balloon and rupture geometry during the LOCA heatup phase. Leveraging these advanced in-situ capabilities, this work intends to use insitu data generated from a simulated LOCA in the Severe Accident Test Station at Oak Ridge National Laboratory to better understand high-temperature creep and its impact on Zircaloy balloon and rupture performance. This work used the BISON fuel performance code to compare BISON’s high-temperature creep model predictions to in-situ data and identify limitations and gaps within the model. Additionally, BISON will subsequently be used to simulate relevant high-burnup LOCA tests. The results will be analyzed and compared to the available post-test data in a manner consistent with the approach outlined in the Nuclear Regulatory Commission research information letter.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Review and Summary of Oxide Thickness Data for Aluminum-Clad Spent Nuclear Fuel

The U.S. Department of Energy (DOE) owns a large inventory of aluminum-clad spent nuclear fuel (ASNF) in interim storage pending ultimate disposition, with more being generated by currently operating research reactors. Dry storage in sealed DOE Standard Canisters is being investigated as an approach for long-term interim storage and/or disposition in a repository for ASNF. A primary challenge for ASNF storage is the presence of aluminum (oxy)hydroxide layers formed on the cladding surfaces during water exposure in the reactor and in wet storage, which forms a reservoir of chemisorbed water not readily removed at low (<100°C) drying temperatures. Free, physisorbed, and chemisorbed water are all susceptible to radiolytic breakdown under irradiation and can release hydrogen gas. Identifying the likely range of (oxy)hydroxide loadings on ASNF that may be placed in dry storage will help to ensure that the impact of the (oxy)hydroxide is adequately accounted for while avoiding over conservatism and enable mitigation strategies to be implemented where needed. The current report summarizes information on (oxy)hydroxide thicknesses and characteristics from the literature and from recent measurements taken in the present research campaign. In general, corrosion studies have indicated that the corrosion and oxide buildup on aluminum are affected by numerous conditions, including pH, temperature and heat flux, coolant flowrates, irradiation (in-reactor vs. unirradiated tests), duration of water exposure, and the amount of oxide already on the surface. Some of these factors are interrelated (e.g., the local temperature, heat flux, and coolant flowrate). Observations of the impact of irradiation (in-reactor versus ex-reactor measurements) on corrosion and oxide thickness are mixed, but multiple studies indicate that there is a significant difference between in-reactor and ex-reactor corrosion kinetics, even when other operating conditions such as pH, flowrate, heat flux, etc., simulate those in a reactor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Sister Rod Destructive Examinations (FY22) Appendix B: Segmentation, Defueling, Metallographic Data and Total Cladding Hydrogen

This report documents work performed under the Spent Fuel and Waste Disposition’s Spent Fuel and Waste Science and Technology program for the US Department of Energy (DOE) Office of Nuclear Energy (NE). This work was performed to fulfill Level 2 Milestone M2SF-23OR010201024, “FY22 Report on ORNL Sibling Rod Testing Results,” within work package SF-23OR01020102 and is an update to the work reported in M2SF-22OR010201047, M2SF-21OR010201032, M2SF-19ORO010201026, and M2SF 19OR010201028. As a part of the DOE-NE High Burnup Spent Fuel Data Project, Oak Ridge National Laboratory (ORNL) is performing destructive examinations (DEs) of high burnup (HBU) (>45 GWd/MTU) spent nuclear fuel (SNF) rods from the North Anna Nuclear Power Station operated by Dominion Energy. The SNF rods, called sister rods or sibling rods are all HBU and include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin (LT) Zirc-4, ZIRLO ® , and M5 ® . Appendix B provides detailed information regarding defueling activities, metallographic imaging and measurements, and total cladding hydrogen measurements.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Independent Technical Review of INL Aluminum-Clad Spent Nuclear Fuel (ASNF) Oxide Layer Radiolytic Gas Generation Resolution: Task 2

PNNL staff performed an independent technical review (ITR) of the aluminum-clad spent nuclear fuel (ASNF) program Task 2. The objective of Task 2 is to determine G-values for the radiolytic production of gaseous molecular hydrogen (H 2 ) from the aluminum oxyhydroxide layers present on ASNF cladding. The G-values need to be defensible, bounding, but not overly conservative. As part of the ITR, the team reviewed program documents, Task 2 work plans, and results to date. In addition, the team held three separate video conferences with the program research staff to ask clarifying questions and obtain additional detail. The PNNL ITR team concludes that if the program follows the methodology and rigor outlined in the test matrices, the program will develop G-values for a defensible technical basis of extended storage of ASNF. Recommendations for improvement are given.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Investigation of Degradation Mechanisms of Cr-coated Zirconium Alloy Cladding in Reactivity-Initiated Accident

Reactivity-initiated accident testing has been performed on two baseline fuel rodlets which were fabricated from Zr-4 cladding and utilized fresh UO 2 fuel pellets. These experiments are part of a larger series of transient irradiations to explore Cr-coated zirconium alloy cladding under Reactivity-initiated accident conditions to evaluate ballooning and burst and sever oxidation characteristic of late-stage failure. The first two irradiations have been completed and post-transient examinations have been performed. In one rod, ballooning and burst occurred, along with oxidation. The second rod leaked and as a result did not balloon as much during the transient. This second rod also experienced increased oxidation compared to the first due to higher temperatures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

High-Temperature SiC Cladding Bowing Irradiation Design and HFIR Readiness

Silicon carbide fiber–reinforced silicon carbide (SiC/SiC) matrix composites are candidate materials for long-term accident-tolerant fuel cladding for use in light water reactors (LWRs). This report documents the design of a dry SiC/SiC composite cladding bowing experiment intended for irradiation in the High Flux Isotope Reactor. The dry experiment allows for testing at representative LWR temperatures for comparison to previous experimental results. This report describes detailed neutronics, thermal, and bowing analyses supporting the irradiation. The design leverages calculations from previous irradiation experiments and production targets for reduced design complexity. The thermal analysis provides a map of SiC thermometer temperatures for comparison during post-irradiation examination to determine the experiment’s irradiation temperatures. Finally, two concepts for introducing localized constraints representative of pressurized water reactor grid spacers are presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗