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At least 199 records · Page 11
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Zircaloy Cladding as Starting Precursor for Ceramic Waste-Forms - 15562

Zircaloy cladding of spent fuel is regular waste-stream related to the process of reprocessing. It remains in the apparatus after spent fuel dissolution. The most common technique of zircaloy cladding treatment is cementation and disposal in metallic containers. In view of the high cost of the cladding material (mainly zirconium), there have been a number of proposals for this material to be recovered and re-used. However, the residual contamination of the claddings following the dissolution of the fuel and the chemical stability of the zirconium militate against any proposal for a recycling process that might be economically reasonable. At the same time there is a lot of information available about synthesis of mineral-like materials based on zirconium for immobilization of long-lived radionuclides. The object of this work was to study experimentally electrochemical dissolution of fuel rod cladding for further obtaining starting precursor for the synthesis of ceramic-waste forms. The results of experiments on electrochemical dissolution of zirconium and zirconium alloy (fuel rod cladding) in nitric acid solutions are presented. The peak dissolution rate of specimens is shown to be observed in dilute nitric acid at elevated temperatures. The rate of dissolution rises with temperature, reaching ∼40-70 mg/cm{sup 2}.h at 90 deg. C. The current yield at this point is 0.6-0.7 g/A.h. The results of analysis of the surface oxide film on specimens and residues formed during dissolution are also presented. An electrochemical dissolution of zircaloy claddings is the common technique for preparation of zirconium solutions to be used for ceramic waste form synthesis. (authors)

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

Modeling and simulation of hydrogen behavior in Zircaloy-4 fuel cladding

As a result of corrosion during normal operation in nuclear reactors, hydrogen can enter the zirconium-alloy fuel cladding and precipitate as brittle hydride platelets, which can severely degrade the cladding ductility. Under a heterogeneous temperature distribution, hydrides tend to accumulate in the colder areas, creating local spots of degraded cladding that can favor crack initiation. Therefore, an estimation of the local hydride distribution is necessary to help predict the risk of cladding failure. The hydride distribution is governed by three competing phenomena. Hydrogen in solid solution diffuses under a concentration gradient due to Fick’s law and under a temperature gradient due to the Soret effect. Precipitation of the hydride platelets occurs once the hydrogen solubility limit is reached. A model of these phenomena was implemented in the 3D fuel performance code BISON in order to calculate the hydrogen distribution for arbitrary geometries, such as a nuclear fuel rod, and is now available for BISON users. Simulations have been performed on simple geometries to validate the model and its implementation. The simulations predict that before precipitation occurs, hydrogen tends to accumulate in the colder spots due to the Soret effect. Once the solubility limit is reached, hydrogen precipitates and forms a rim close to the outer edge of the cladding. The simulations also predict that the reactor shut down has little effect on already precipitated hydrides but causes the remaining hydrogen to precipitate homogeneously into hydrides.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Microstructure Evolution in Ion-Irradiated Oxidized Zircaloy-4 Studied with Synchrotron Radiation Microdiffraction and Transmission Electron Microscopy

The corrosion process (oxidation and hydriding) of zirconium alloy fuel cladding is one of the limiting factors on fuel rod lifetime, particularly for Zircaloy-4. The corrosion rate of this alloy shows indeed a great acceleration at high burnup in light water reactors (LWRs). Understanding the corrosion behavior under irradiation for this alloy is an important technological issue for the safety and efficiency of LWRs. In particular, understanding the effect of irradiation on the metal and oxide layers is a key parameter in the study of corrosion behavior of zirconium alloys. In this study, Zircaloy-4 samples underwent helium and proton ion irradiation up to 0.3 dpa, forming a uniform defect distribution up to 1 pm deep. Both as-received and precorroded samples were irradiated to compare the effect of metal irradiation to that of oxide layer irradiation. After irradiation, samples were corroded to study the impact of irradiation defects in the metal and in preexisting oxide layers on the formation of new oxide layers. Synchrotron X-ray microdiffraction and microfluorescence were used to follow the evolution of oxide crystallographic phases, texture, and stoichiometry both in the metal and in the oxide. In particular, the tetragonal oxide phase fraction, which has been known to play an important role in corrosion behavior, was mapped in both unirradiated and irradiated metals at the submicron scale and appeared to be significantly affected by irradiation. These observations, complemented with electron microscopy analyses on samples in carefully chosen areas of interest, were combined to fully characterize changes caused by irradiation in metal and oxide phases of both alloys.

ion-irradiation↗
DOE OSTI2017

Evaluation of the First Generation Dual-purpose Coatings for SiC Cladding

The loss of coolant and station blackout at the Fukushima Daichi nuclear power station highlighted the need to improve the safety margins of current and future Light Water Reactors (LWRs). Accident Tolerant Fuel (ATF) development is designed to find solutions to mitigate or prevent such events through improvement of fuels and fuel cladding. One promising ATF concept is the replacement of zirconium-based alloy cladding with silicon carbide (SiC) fiber-reinforced SiC matrix (SiCf-SiC) composites, due to the inherent accident resistant features of SiC. However, during normal operations, SiC corrodes in reactor coolant potentially at an unacceptable rate. In addition, it is likely for the SiC-based cladding to undergo microcracking under an applied stress, which potentially degrades the hermeticity.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Microbeam X-Ray Absorption Near-Edge Spectroscopy of Alloying Elements in the Oxide Layers of Irradiated Zircaloy-2

Hydrogen pickup of zirconium-based fuel cladding and structural materials during in-reactor corrosion can degrade fuel components because the ingress of hydrogen can lead to the formation of brittle hydrides. In the boiling water reactor (BWR) environment, Zircaloy-2 fuel cladding and structural components such as water rods and channels can experience accelerated hydrogen pickup, whereas Zircaloy-4 components exposed to similar conditions do not. Because the principal difference between the two alloys is that Zircaloy-2 contains nickel, accelerated hydrogen pickup has been hypothesized to result from the presence of nickel. However, an understanding of the mechanism by which this acceleration occurs is still lacking. We investigated the link between hydrogen pickup and the oxidation behavior of alloying elements when incorporated into the oxide layers formed on zirconium alloys when corroded in the reactor. Synchrotron radiation microbeam X-ray absorption near-edge spectroscopy (XANES) at the Advanced Photon Source was performed on carefully selected BWR-corroded Zircaloy-2 water rods at an assembly-averaged burnup ranging from 32.8 to 74.6 GWd/MTU to determine the oxidation states of alloying elements, such as iron and nickel, within the oxide layers as a function of distance from the oxide-metal interface at high burnup. Samples were chosen for comparison based on having similar oxide thicknesses, processing, elevation, reactors, and fluences but different hydrogen pickup fractions. Examinations of the oxide layers formed on these samples showed that (1) the oxidation states of these alloying elements changed with distance from the oxide-metal interface, (2) these elements exhibited delayed oxidation relative to the host zirconium, and (3) nickel in Zircaloy-2 remained metallic in the oxide layer at a longer distance from the oxide-metal interface than iron. An analysis of these results showed an apparent correlation between the delayed oxidation of nickel and higher hydrogen pickup of Zircaloy-2 at high burnup.

alloying elements↗
DOE OSTI2021

Oak Ridge National Laboratory Fission Product Hydrochlorination

This work supports the recovery of high-assay low-enriched uranium (HALEU, uranium enriched in 235 U to >5% and <20%) from used nuclear fuel through a process based on the zirconium extraction (ZIRCEX) process. The ZIRCEX process starts with the dry head-end removal of zirconium or aluminum cladding from the nuclear fuel via chlorination or hydrochlorination, which results in the separation of cladding from the fuel as the metal chloride (e.g., aluminum trichloride or zirconium tetrachloride). The hydrochlorination process is included in the Material Recovery Pilot Plant (MRPP), a ZIRCEX-based demonstration plant at Idaho National Laboratory (INL). The experiments outlined in this report assessed fission product (FP) volatility and transport under the hydrochlorination conditions used to separate cladding from used nuclear fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Oxidation of Advanced Steel Cladding Alloys in Extreme Environments

Accident tolerant claddings must withstand the loss of coolant for longer times than conventional zirconium-based reactor claddings. Advanced steel alloys (such as Alloy 33, APMT) offer enhanced resistance to high-temperature oxidation in steam environments due to the formation of protective oxide layers of chromia and/or alumina oxides on the surface of the alloy, which enhance their stability at high temperatures by acting as an effective diffusion barrier, so resulting in low rate of oxidation. These passivating oxide layers, resistant to accelerated oxidation, offer an improvement over the current designs by reducing the production of hydrogen gas, and by delaying the release of fission products into the environment. Hence, these alloys present a great opportunity to explore alternative cladding materials with better accident tolerance for commercial light water reactors. Alloy 33 and APMT showed enhanced resistance to corrosion under the normal operation and off-nominal conditions of light water reactors.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Study of Thermomechanical Behavior of Zirconium-0.3 Tin Alloy

Zirconium-based alloys are used in the manufacturing of nuclear fuel tubes. These tubes need to be clad with zirconium-tin (Zr-0.3 wt.% Sn) alloy to enhance the lives of tubes. Manufacturing of such clad tubes is a challenging task, due to work hardening and dynamic softening phenomenon. In this study, hot deformation behavior of Zr-0.3Sn alloy is investigated at different temperatures and strain rates. It is found that fully recrystallized microstructure can be obtained by working at around 900 °C and high strain rate of 10 s{sup −1}. Further, constitutive equations are developed for the entire domain of deformation range and fitted into standard FEM-based simulation model for extrusion. The simulation results are validated with experimental data. The difference in flow stresses values obtained from numerical and experimental methods was around 14%.

36 MATERIALS SCIENCE↗
DOE OSTI2016

FY-16 Technology Gap Study Technical Report: Analysis of Undissolved Anode Materials of Mark-IV Electrorefiner

The Experimental Breeder Reactor II (EBR-II) is a sodium cooled fast reactor developed at Argonne National Laboratory (ANL). The used fuels from the EBR-II are currently being treated in the Fuel Conditioning Facility (FCF) at the Idaho National Laboratory (INL). The Mark IV (Mk-IV) electrorefiner (ER) is a unit process in the FCF, which is primarily assigned to treating the used driver fuels. The stainless steel anode baskets hold the chopped spent driver fuel segments. During electrorefining, the anode baskets are immersed into the electrolyte and the used fuel is dissolved electrochemically. Perforated sides and bottoms allow the flow of the electrolyte into and out of the anode baskets. The steel cathode is also immersed into the electrolyte and collects the reduced products. The active metal contents in the used fuel (e.g., Cs, Sr, lanthanides, Pu, etc.) reacts with uranium cations in the electrolyte and progressively reports to the electrolyte. Noble metals are mostly retained in the cladding hulls. Varying quantities of zirconium are retained in the cladding hulls depending on the operational conditions of the Mk-IV ER. The undissolved anode materials are removed from the anode baskets and stored for subsequent metal waste form processing. These undissolved materials typically include undissolved fuels, stainless steel cladding, and adhering electrolyte. A couple of hulls are retrieved for chemical analysis and used for estimating the composition of the entire undissolved anode materials. The mass balance attempt based on this practice of estimating the undissolved anode materials has been a challenge due to inherently high sampling errors associated with heterogeneous undissolved material compositions. Responding to the prescribed challenge, this report investigates chemical analysis data as a whole and finds noticeable trends in the compositions of undissolved anode material samples with respect to the mass of the whole undissolved anode materials. Based upon this discovery, an empirical model is proposed.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Degradation and Failure Phenomena of Advanced Reactor Fuel Concepts: Sodium-Cooled Fast Reactor Metallic Fuel

The U.S. Nuclear Regulatory Commission (NRC) is anticipating licensing applications and commercial use of new fuel types in advanced nuclear power reactors that would be designed and built in the United States. Pacific Northwest National Laboratory (PNNL) is providing technical assistance to the NRC related to the newly proposed nuclear fuel and cladding designs that would be deployed in these reactors. This report focuses specifically on the metallic fuel that is being considered for sodium-cooled fast reactors and specifically on mechanisms that would cause damage or failure to the fuel under reactor operating conditions and design basis accident conditions. There is historic experience with both metallic and oxide fuels in experimental sodium-cooled fast reactors, but this report will focus solely on metallic fuel. Currently two U.S.-based reactor designers are engaged with the NRC in the application or pre-application review and considering a sodium-cooled fast reactor. TerraPower is engaged with NRC in pre-application review of its Natrium reactor. The current design for this fuel uses a sodium-bonded uranium-10wt% zirconium (U10Zr) fuel clad in HT9 stainless steel. ARC Clean Technology is engaged with NRC in pre-application review of its ARC-100 reactor. The current design for this fuel uses a sodium-bonded uranium-10wt% zirconium (U10Zr) fuel with steel cladding. This report will focus on this fuel system specifically, with broader information given regarding other metallic fuel systems with other stainless steel alloy claddings. To support the NRC’s readiness efforts, this report will identify and discuss degradation and failure modes of metallic fuel concepts for sodium-cooled fast reactors, including fuel performance characteristics that may not be addressed within existing regulatory documents.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Recent Advances in Cladding Material Extraction from Fuels in Nuclear Fuel Cycles

An improved recycling and recovery process for the cladding material from spent nuclear fuels is very important toward confirming nuclear energy to support ongoing sustainable development of nuclear management by reducing waste and conserving resources. Nuclear spent fuel cladding materials such as zirconium alloys have economic values and can be recovered, and their recovery eliminates problems in waste disposal and conserves valuable resources. Over 110 published reports and journal articles are reviewed and summarized herein, with a main focus on documenting recovery techniques used to recover cladding materials from spent nuclear fuel and recent developments. Several recovery techniques which are used at present times, such as mechanical separation, chemical dissolution, and hydrometallurgical processes have been covered with examples and discussions. Difficulties within the recovery process are also discussed, and most probable areas for future research in improving efficiency and sustainability of recovering cladding material are identified and discussed at the end. Here, this review could be an important document to the field of spent nuclear fuel reprocessing, recovering valuables and thereby offering guidance on how to effectively manage, safely handle, and reduce nuclear waste. In addition to reducing the volume and radiotoxicity of high-level waste, this review also highlights the potential economic benefit of recovering zirconium from spent fuel cladding by relating typical zirconium metal prices to the mass of cladding per tonne of spent fuel, illustrating that the recoverable material value is non-negligible compared with back-end fuel-cycle costs.

Mondal, Kunal [Oak Ridge National Laboratory (ORNL↗
DOE OSTI2019

Radiation response of Ti 2 AlC MAX phase coated Zircaloy-4 for accident tolerant fuel cladding

A Ti 2 AlC MAX phase material was deposited on Zircaloy-4 substrate using the cold spray process, with the goal of enhancing the accident tolerance of zirconium-alloy fuel cladding in light water reactors. The samples were annealed up to 923 K to study the intermetallic phase formation between the coating and Zircaloy-4 substrate. Three interface compounds, ZrAl 2 , ZrAl, and Zr 3 Al, were identified and the activation energy of the interfacial intermetallic compound growth was determined. The radiation response of the coating material and interfacial intermetallic compounds after 3.5 MeV Zr 2+ ion irradiation was investigated by transmission electron microscopy and nanoindentation. Ion irradiation disrupted the nanolamellar structure of the Ti 2 AlC coating while amorphization was observed some of the intermetallic compounds. Furthermore, nanoindentation revealed overall hardening of the Ti 2 AlC coating and Zr 3 Al phase, with no significant change observed in the ZrAl 2 and ZrAl phases.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Effect of surface oxides on tritium entrance and permeation in FeCrAl alloys for nuclear fuel cladding: a review

Abstract Iron-chromium-aluminum (FeCrAl) alloys are being considered for the cladding of uranium dioxide fuel in light water reactors (LWRs). FeCrAl alloys have good mechanical properties at temperatures of 300 °C and higher, and have superlative resistance to attack by steam at temperatures of up to 1000 °C and higher. A concern has been raised that the use of FeCrAl for cladding would result in a higher content of tritium in the reactor coolant as compared with the current system where the cladding is a zirconium based alloy. This review shows that the flux of tritium from the fuel rod cavities to the coolant across the fuel cladding wall will be greatly reduced by the presence of oxides on the surface of the cladding. The review of current literature and permeation data show that (a) protective oxides are expected to be present on both sides of the FeCrAl cladding, and (b) depending on the characteristics of these oxide layers it is reasonable to expect about two–three orders of magnitude reduction in tritium permeation, relative to the permeation response in clean, unoxidized condition for FeCrAl steels of interest, around 277 °C–377 °C temperatures.

Electrochemistry↗
DOE OSTI2017

The Mechanical Response of Advanced Claddings during Proposed Reactivity Initiated Accident Conditions

This study investigates the failure mechanisms of advanced nuclear fuel cladding of FeCrAl at high-strain rates, similar to design basis reactivity initiated accidents (RIA). During RIA, the nuclear fuel cladding was subjected to the plane-strain to equibiaxial tension strain states. To achieve those accident conditions, the samples were deformed by the expansion of high strength Inconel alloy tube under pre-specified pressure pulses as occurring RIA. The mechanical response of the advanced claddings was compared to that of hydrided zirconium-based nuclear fuel cladding alloy. The hoop strain evolution during pressure pulses were collected in situ; the permanent diametral strains of both accident tolerant fuel (ATF) claddings and the current nuclear fuel alloys were determined after rupture.

Accident tolerant fuel↗
DOE OSTI2021

Investigation of δ zirconium hydride morphology in a single crystal using quantitative phase field simulations supported by experiments

In light water nuclear reactors, waterside corrosion of the cladding material leads to the production of hydrogen, a fraction of which is picked up by the zirconium cladding and precipitates into brittle hydride particles. These nanoscale hydride particles aggregate into mesoscale hydride clusters. The principal stacking direction of the nanoscale hydrides precipitated in the cladding tube changes from circumferential in the absence of applied stress to radial under circumferential applied stress. A quantitative phase field model has been developed to predict the hydride morphology observed experimentally and identify the mechanisms responsible for nanoscale hydride stacking. The model focuses on nanoscale hydride precipitation in a single zirconium grain with a detailed description of the anisotropic elastic contribution. The model predictions concerning the shape, orientation, and stacking behavior of nanoscale hydride are analyzed and compared with experimental observations. The model accurately accounts for the experimentally observed elongated nanoscale hydride shape and the stacking of hydrides along the basal plane of the hexagonal zirconium matrix. When investigating the role of applied stress in hydride morphology, the model challenges some of the mechanisms previously proposed to explain hydride reorientation. Finally, although hydride reorientation has been hypothesized to be caused by a change in nanoscale hydride shape, the current study shows that these mechanisms are unlikely to occur.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Diffusion bonding of tungsten-vanadium-zirconium using vacuum hot pressing for the development of a low decay heat cladding solution for tungsten spallation targets

Tantalum has been used as cladding material for water-cooled solid tungsten targets at many leading spallation neutron production facilities thanks to its high neutron yield, manageable radiation damage behavior, and excellent corrosion/erosion resistance in radiation environments. However, from a safety hazard perspective, thermal neutron capture of tantalum in spallation environments causes a high specific decay heat in the target volume, which often becomes a limiting factor in increasing the beam power on the target. In this paper, we studied vacuum hot pressing (VHP) parameters to diffusion bond zirconium to tungsten to explore the feasibility of using zirconium alloys as an alternative cladding material to tantalum. Zirconium alloys have long been used as cladding material for early generation solid spallation targets, and nuclear fuel rods. In spallation environments zirconium has significantly lower decay heat with shorter decay time compared to tantalum. The hot isostatic pressing (HIP) of zirconium and tungsten is known to produce limited bonding quality due to the formation of the brittle ZrW2 intermetallic layer. To overcome this problem, placing a vanadium interlayer between tungsten and zirconium has been proposed by exploring parameter space in binary alloy phase diagrams. Under the VHP conditions, 860 ◦C at 70 MPa for 4 hours, Zr-V and V-W showed good diffusion bonding, which demonstrates the feasibility of a single step HIP process to make the zirconium alloy clad tungsten spallation volumes.

Lee, Yong Joong↗
DOE OSTI2016

Hydrogen trapping at neutron irradiation produced defects in recrystallized alpha annealed Zircaloy-4 - 2016-0060

Hydrogen is absorbed into zirconium alloy-clad fuel elements and structural components as a result of water-side corrosion in both pressurized water reactor (PWR) and boiling water reactor (BWR) environments. Increasing hydrogen concentrations in the zirconium alloy affect both its mechanical properties and dimensional stability. It is well known that hydrogen can diffuse relatively rapidly through zirconium base alloys (in relation to the lifetime of the components) in response to solid solution concentration, temperature, and stress gradients. Migration of hydrogen within a component through its life must be considered to reliably predict the performance of the component through life and even post-service life. Many of the physical properties needed to describe this process, such as the solubility of hydrogen in zirconium alloys, diffusion coefficients of hydrogen through zirconium alloys, and the Soret effect within zirconium alloys, already have been published. Many of these measurements have been made outside of the reactor core environment on nonirradiated zirconium alloys. This paper presents in-reactor, isothermal hydrogen diffusion data that show an effect of the in-reactor radiation environment on the final measured hydrogen concentrations. Samples used are similar to the hydrogen diffusion couples used by several investigators in the past to measure the hydride dissolution solvus as a function of temperature. The experimental results are interpreted on the basis of a physical model for hydrogen trapping that hypothesizes that hydrogen in solid solution in the zirconium lattice comes into thermodynamic equilibria with the hydride phase as well as into a steady-state equilibrium with trapping at irradiation-induced defects within the lattice. In agreement with the model, the irradiation effect on final measured concentrations is seen to be a decreasing function of the irradiation temperature between ∼271 deg. C and 354 deg. C, with any effect essentially eliminated by an irradiation temperature of 354 deg. C. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

TREAT reactor LEU fuel-clad chemical interaction empirical modeling analysis

DOE is converting TREAT (Transient Reactor Test) facility from its existing highly enriched uranium (HEU) core to a low-enriched uranium (LEU) core. In order to assess the magnitude of chemical interaction between fuel and cladding materials during physical contact under expected TREAT operational limits, planned transients tests, and reactivity accident scenarios, a combination of experimental testing and thermodynamic modeling was performed to predict the expected chemical interactions among fuel and cladding chemical constituents. Thermodynamic calculations were then validated with empirical data from experiments that emulate TREAT's expected upper operational limits. Pellet samples composed of LEU oxide powder dispersed in a graphite matrix had intimate contact with zirconium-based alloy cladding. The samples were subjected to long-term isothermal heating under high vacuum. Specimen characterization consisted of scanning electron microscopy, x-ray diffraction analysis, and x-ray tomography. ThermoCalc software and Ellingham's diagrams were used for thermodynamic calculations. The X-ray tomography and SEM analysis of the pellets showed a homogeneous distribution of UO{sub 2} particles within the graphite matrix with isolated larger UO{sub 2} agglomerates. The SEM analysis showed the formation of a lamellar structure between the large UO{sub 2} agglomerate and the graphite matrix. This is an indication that diffusion is taking place at a rather moderate temperature and reaction time during the pellet fabrication. Secondary phases, with high Al and Mg content, were detected inside a UO{sub 2} agglomerate using SEM in SE/EBSD mode and chemical analysis with EDS. LEU feedstock impurities, as Al and Mg, seem to promote the chemical reactivity between UO{sub 2} particles and graphite matrix. XRD analysis of a pellet sample showed the presence of UO{sub 2} and graphite, but no other crystalline phases were detected with this technique. Thermodynamic analysis through the use of Ellingham diagrams showed that within TREAT's operational to reactivity accident temperature (278-820 Celsius degrees), UO{sub 2} can be reduced to U metal by Zr due to a more favorable ΔG for the oxidation of Zr.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗