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All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

First-Principles Studies of Tritium Species Diffusivity Across the Interfaces of Ni-NiZr Alloy-Zircaloy-4

A tritium producing assembly, also known as tritium producing burnable absorber rods (TPBARs), consists of a metal getter tube located between the cladding and γ-LiAlO2 pellets. The metal getter tube is composed of a nickel (Ni) layer coated on Zircaloy-4. The getter assembly is used to capture tritium (3H) species (mainly 3H2 and 3H2O) generated from γ-LiAlO2 pellets during irradiation. Exploring 3H species (3H2, 3H2O) dissociation on the Ni surface and diffusion in the Ni layer and across the interface of Ni-plated Zircaloy-4 getters can provide insights on tritium transport and retention in the pellets and the getter materials. . In FY25, based on the ideal interface model of Ni-Zircaloy-4 generated in FY24, we will further explore the diffusion pathways of 3H across such Ni-ZrNi-alloy layer-Zircalory-4 interfaces under different conditions, including with oxide/hydroxide clusters on the Ni layer, and impurities located in ZrNi-alloy and Zircalory-4 layers. Due to the size limitation of DFT simulation, we will separate the 3-layer system into 3 sub-systems: Ni-ZrNi interface, ZrNi alloy layer, ZrNi-alloy-Zircaloy-4 interface, and simulate the diffusion barriers for tritium.

diffusion barrier↗
DOE OSTI2018

Small-angle neutron scattering measurements of δ-phase deuteride (hydride) precipitates in Zircaloy 4

Small-angle neutron scattering (SANS) measurements have been performed under ambient conditions to characterize deuteride (hydride) particles in Zircaloy 4, a fuel cladding material used in pressurized light-water nuclear reactors. Hydrogen pickup by the cladding leads to a rim structure in which large circumferential hydride plate-like particles preferentially form on the cooler water-side region of the cladding. Deuterium substitution has been used to increase the coherent response and decrease the incoherent background of the SANS measurements. Four bulk deuterium concentrations were investigated, approximately 100, 400, 500 and 1000 parts per million by weight (w.p.p.m.) deuterium, as well as a zero-deuterium-concentration reference sample. The net SANS response from the deuteride phase was determined at all concentration values after subtraction of the reference SANS response, which effectively subtracted the strong scattering from second-phase particles in as-received Zircaloy. The net SANS response consisted of strong Porod scattering from deuteride particles over the entire measured Q range (0.005–0.4 Å -1 ). The net SANS response was anisotropic at concentrations greater than 100 w.p.p.m. and required elliptical averaging analysis. A significant sample orientation effect on the intensity of the SANS response was observed, due to preferential alignment of deuteride particles. The effect of ex situ applied stress at elevated temperature on deuteride phase dissolution and reprecipitation was investigated; a weak effect was observed with SANS that could not be confirmed by optical microscopy.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Small-Angle Neutron Scattering Measurements of the δ Deuteride Phase in Zircaloy 4

Hydrogen pickup by Zr-based LWR fuel cladding during service leads to the formation of circumferential hydride platelet particles embedded within the Zr matrix. The precipitation of the hydride phase occurs preferentially on the water-side due to the temperature gradient within the clad wall. The non-uniform hydride morphology is known as a rim structure. Significant tensile hoop stresses within the clad wall develop during the fuel lifetime due to fission gas release and pellet-clad mechanical interaction. One potential post-service cladding failure mechanism is delayed hydride cracking (DHC) initiated by hydride reorientation under this tensile hoop stress. As a consequence, hydride reorientation and DHC have been extensively studied over the last few decades to understand the interdependencies of tensile stress, temperature, hydride morphology, and hydrogen concentration. Much of the most recent work in this area has taken advantage of 3. generation synchrotron X-ray sources to perform in situ measurements of hydride phase stability. Neutron scattering based techniques offer unique advantages in the study of hydrogen in metals (typically with deuterium substituted for hydrogen) but are under utilized in the study of hydrides in Zr. We present small-angle neutron scattering (SANS) measurements of the deuteride phase in Zircaloy 4 (Zy4) here. The SANS measurements presented here demonstrate the sensitivity of the technique to deuteride phase particles embedded in a Zy4 matrix. Quantifiable net scattering at 50 wppm effective hydrogen concentration is derived from a subtraction of a reference sample. This demonstrates that the strong χ phase Porod scattering can be effectively eliminated, revealing the response from the deuteride phase. A similar experimental protocol using small-angle X-ray scattering (SAXS) would likely fail due to poor contrast between the hydride phase and the Zr matrix. Stated in different terms, the SAXS response would be dominated by SPP Porod scattering with unresolvable net scattering from hydride particles. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2019

Material Properties of Non-Irradiated Zircaloy 4 in Support of ASME Code Acceptance for Pressure Vessel Design

A program underway at Oak Ridge National Laboratory (ORNL) to support private industry design of a system for production of 99Mo includes review of Zircaloy-4 (Zry-4) properties as a potential material for a primary vessel in the system. Because Zry-4 is not a material specified in the ASME Boiler and Pressure Vessel Code, a review has been performed to obtain the necessary data in the non-irradiated condition from the literature and from additional testing of Zry-4 to enable creation of tables of material properties required for ASME Code Section III Article ND-3000 design. A strategy for an in-vessel surveillance program is also recommended for early detection of potential corrosion and irradiation degradation issues during operational life of the vessel. Additionally, the ASME Code Section II, Part D was reviewed to ascertain the requirements for proposing a new material for Code approval. A plate of Zry-4 procured by ORNL was used to obtain tensile properties that were used as the basis for determining the maximum allowable stresses applicable to the vessel design. The materials properties of primary interest for ASME Code application are tensile properties, fracture toughness, fatigue, and five specific physical properties required by the Code for approval of a new material. The literature review revealed that all reported experimental data showed tensile strengths to substantially exceed the minimum specifications in ASTM B352. Sufficient fracture toughness and fatigue data are available to conclude that Zry-4 has sufficient fracture and fatigue resistance in the unirradiated condition for vessel design and operation. Of the physical properties reviewed, only thermal conductivity revealed a discrepancy that is recommended for further review. A surveillance program plan has been developed that is probably the maximum that would need to be implemented for the specific vessel design considered. Once adequate non-irradiated and irradiated data are available from preproduction testing of the Zry-4 and weld metal and HAZ materials, the program could be reduced to reflect that either initial properties are very good and/or the effects of embrittlement appear to be minimal. However, the overall design of a surveillance program is, of course, also dependent on the ASME Code design class for the pressure vessel.

Nanstad, Randy↗
DOE OSTI2026

First-principles Study of the Tritium Species Diffusion Across Ni-Zircaloy-4 Getter Interface

Tritium is a critical fuel component for experimental nuclear fusion reactors (like tokamaks), typically used in combination with deuterium. In tritium-producing burnable absorber rods (TPBARs), due to its high lithium density, LiAlO2 is used in the form of an annular ceramic pellet enriched with the Li_6 isotope located between the Zircaloy-4 liner and nickel-plated Zircaloy-4 tritium getter. Employing DFT, we explored the tritium diffusion across the pellet and getter.

36 MATERIALS SCIENCE↗
DOE OSTI2013

SiC-CMC-Zircaloy-4 Nuclear Fuel Cladding Performance during 4-Point Tubular Bend Testing

The U.S. Department of Energy Office of Nuclear Energy (DOE NE) established the Light Water Reactor Sustainability (LWRS) program to develop technologies and other solutions to improve the reliability, sustain the safety, and extend the life of current reactors. The Advanced LWR Nuclear Fuel Development Pathway in the LWRS program encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. Recent investigations of potential options for “accident tolerant” nuclear fuel systems point to the potential benefits of silicon carbide (SiC) cladding. One of the proposed SiC-based fuel cladding designs being investigated incorporates a SiC ceramic matrix composite (CMC) as a structural material supplementing an internal Zircaloy-4 (Zr-4) liner tube, referred to as the hybrid clad design. Characterization of the advanced cladding designs will include a number of out-of-pile (nonnuclear) tests, followed by in-pile irradiation testing of the most promising designs. One of the out-of-pile characterization tests provides measurement of the mechanical properties of the cladding tube using four point bend testing. Although the material properties of the different subsystems (materials) will be determined separately, in this paper we present results of 4-point bending tests performed on fully assembled hybrid cladding tube mock-ups, an assembled Zr-4 cladding tube mock-up as a standard and initial testing results on bare SiC-CMC sleeves to assist in defining design parameters. The hybrid mock-up samples incorporated SiC-CMC sleeves fabricated with 7 polymer impregnation and pyrolysis (PIP) cycles. To provide comparative information; both 1- and 2-ply braided SiC-CMC sleeves were used in this development study. Preliminary stress simulations were performed using the BISON nuclear fuel performance code to show the stress distribution differences for varying lengths between loading points and clad configurations. The 2-ply sleeve samples show a higher bend momentum compared to those of the 1-ply sleeve samples. This is applicable to both the hybrid mock-up and bare SiC-CMC sleeve samples. Comparatively both the 1- and 2-ply hybrid mock-up samples showed a higher bend stiffness and strength compared with the standard Zr-4 mock-up sample. The characterization of the hybrid mock-up samples showed signs of distress and preliminary signs of fraying at the protective Zr-4 sleeve areas for the 1-ply SiC-CMC sleeve. In addition, the microstructure of the SiC matrix near the cracks at the region of highest compressive bending strain shows significant cracking and flaking. The 2-ply SiC-CMC sleeve samples showed a more bonded, cohesive SiC matrix structure. This cracking and fraying causes concern for increased fretting during the actual use of the design. Tomography was proven as a successful tool to identify open porosity during pre-test characterization. Although there is currently insufficient data to make conclusive statements regarding the overall merit of the hybrid cladding design, preliminary characterization of this novel design has been demonstrated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Creating Hydride Rim Structures in Zircaloy 4 Cladding Tubes to Mimic High Burnup

During their long service in commercial light water reactors, zirconium alloy cladding tubes are subject to oxidation and subsequent hydrogen pickup which can degrade the materials mechanical properties [1]. These degradations can lead to premature cladding ruptures during reactor power transients [2] as well as negatively impact the claddings post quench ductility following high temperature steam oxidation in thermal hydraulic transients [3]. Hydrogen has low solid solubility in zirconium and usually precipitates as brittle zirconium hydride platelets in the cladding metal matrix. Due to temperature driven diffusion (Soret effect) and the grain texture of stress relieved zirconium metal tubes, these hydrides precipitate with a circumferential orientation near the outer surface of the cladding tubes forming a hydride rim which can serve as a site for crack initiation during circumferential ‘hoop’ loading of the cladding [4rim formation][5]. In order to better understand the effect of hydride rims on the fracture behavior of stress relieved zirconium alloy cladding tubes a procedure to produce these rim-like structures is being developed at Idaho National Laboratory (INL).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Pre-Irradiation Testing and Analysis to Support the LWRS Hybrid SiC-CMC-Zircaloy-4 Unfueled Rodlet Irradiation

Nuclear fuel performance is a significant driver of nuclear power plant operational performance, safety, economics and waste disposal requirements. The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Pathway focuses on improving the scientific knowledge basis to enable the development of high-performance, high burn-up fuels with improved safety and cladding integrity and improved nuclear fuel cycle economics. To achieve significant improvements, fundamental changes are required in the areas of nuclear fuel composition, cladding integrity, and fuel/cladding interaction.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Development of Plasticity Model Using Non-Associated Flow Rule for HCP Materials Including Zirconium for Nuclear Applications

In this report (prepared in collaboration with Prof. Jeong Whan Yoon, Deakin University, Melbourne, Australia) a research effort was made to develop a non associated flow rule for zirconium. Since Zr is a hexagonally close packed (hcp) material, it is impossible to describe its plastic response under arbitrary loading conditions with any associated flow rule (e.g. von Mises). As a result of strong tension compression asymmetry of the yield stress and anisotropy, zirconium displays plastic behavior that requires a more sophisticated approach. Consequently, a new general asymmetric yield function has been developed which accommodates mathematically the four directional anisotropies along 0 degrees, 45 degrees, 90 degrees, and biaxial, under tension and compression. Stress anisotropy has been completely decoupled from the r value by using non associated flow plasticity, where yield function and plastic potential have been treated separately to take care of stress and r value directionalities, respectively. This theoretical development has been verified using Zr alloys at room temperature as an example as these materials have very strong SD (Strength Differential) effect. The proposed yield function reasonably well models the evolution of yield surfaces for a zirconium clock rolled plate during in plane and through thickness compression. It has been found that this function can predict both tension and compression asymmetry mathematically without any numerical tolerance and shows the significant improvement compared to any reported functions. Finally, in the end of the report, a program of further research is outlined aimed at constructing tensorial relationships for the temperature and fluence dependent creep surfaces for Zr, Zircaloy 2, and Zircaloy 4.

42 ENGINEERING↗
DOE OSTI2014

Physical and mechanical metallurgy of zirconium alloys for nuclear applications: a multi-scale computational study

In the post-Fukushima world, the stability of materials under extreme conditions is an important issue for the safety of nuclear reactors. Because the nuclear industry is going to continue using advanced zirconium cladding materials in the foreseeable future, it become critical to gain fundamental understanding of the several interconnected problems. First, what are the thermodynamic and kinetic factors affecting the oxidation and hydrogen pick-up by these materials at normal, off-normal conditions, and in long-term storage? Secondly, what protective coatings (if any) could be used in order to gain extremely valuable time at off-normal conditions, e.g., when temperature exceeds the critical value of 2200°F? Thirdly, the kinetics of oxidation of such protective coating or braiding needs to be quantified. Lastly, even if some degree of success is achieved along this path, it is absolutely critical to have automated inspection algorithms allowing identifying defects of cladding as soon as possible. This work strives to explore these interconnected factors from the most advanced computational perspective, utilizing such modern techniques as first-principles atomistic simulations, computational thermodynamics of materials, diffusion modeling, and the morphological algorithms of image processing for defect identification. Consequently, it consists of the four parts dealing with these four problem areas preceded by the introduction and formulation of the studied problems. In the 1st part an effort was made to employ computational thermodynamics and ab initio calculations to shed light upon the different stages of oxidation of ziraloys (2 and 4), the role of microstructure optimization in increasing their thermal stability, and the process of hydrogen pick-up, both in normal working conditions and in long-term storage. The 2nd part deals with the need to understand the influence and respective roles of the two different plasticity mechanisms in Zr nuclear alloys: twinning (at low T) and crystallographic slip (higher T’s). For that goal, a description of the advanced plasticity model is outlined featuring the non-associated flow rule in hcp materials including Zr. The 3rd part describes the kinetic theory of oxidation of the several materials considered to be perspective coating materials for Zr alloys: SiC and ZrSiO 4 . In the 4th part novel and advanced projectional algorithms for defect identification in zircaloy coatings are described. In so doing, the author capitalized on some 12 years of his applied industrial research in this area. Our conclusions and recommendations are presented in the 5th part of this work, along with the list of used literature and the scripts for atomistic, thermodynamic, kinetic, and morphological computations.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Characterization of the Chlorination Products of Zirconium Alloys used for Nuclear Fuel Cladding

The United States produces over 2000 metric tons of used nuclear fuel material per year, of which 25% is irradiated fuel cladding, and it is expected to increase for the next 50 years. Previous simulation studies have shown that in the irradiated cladding, the major activation products found are {sup 125}Sb, {sup 60}Co, and {sup 99m}Nb, and the major fission product found is {sup 137}Cs. These radioisotopes make it impossible to treat the cladding as low level waste. A chlorination process has been proposed for zirconium reprocessing as it has been proven more efficient than previous mechanical processes, redox processes, and other chemical methods (e.g. iodination). In the chlorination process, zirconium alloy cladding reacts with Cl{sub 2} at 350- 380 deg. C to produce mainly ZrCl{sub 4}. The zirconium tetrachloride has a relatively low sublimation point (331 deg. C) that allows its separation from other products, and so it can be treated as low level waste for disposal or recycling. However, previous work has observed element impurities (iron, tin, chromium, cesium, antimony, and niobium) in the recovered ZrCl{sub 4}, meaning that additional steps are necessary in a chlorination process previous to the reduction of ZrCl, to metallic Zr. The emphasis of this research summary is on the progress that has been made to isolate and characterize the impurities produced after chlorination of three previously characterized non-irradiated zirconium alloys (Zircaloy-2, Zircaloy-4, and a commercial Zr-Nb alloy) widely used as cladding materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Effect of pH on hydrogen pick-up and corrosion in zircaloy-4

Thermal Desorption Spectroscopy (TDS) has been used to investigate the hydrogen pick-up behaviour in different samples of Zircaloy-4. This includes as-received, cold rolled, hydrogen charged, and autoclave-oxidised in pure water and at an elevated pH (with 50% deuterated water) compared to commercial reactors. A characteristic desorption peak for hydrogen has been found at about 650 Celsius degrees, which is theorised to come from hydrides. Cold rolled samples show the same general shape spectrum as the as-received samples, with more hydrogen detected per unit volume of sample. This suggests more hydrogen is in trapping sites or outward diffusion paths are facilitated by dislocations, since no more hydrogen has been added. The peak broadens towards higher temperature, which indicates that additional dislocations could provide trapping sites with a slightly higher energy than the average in the as-received samples. The hydrogen charged samples form a hydride rim, which contains much more hydrogen than the as-received samples. The peak broadens but on average does not shift. The surface oxide, when present, acts as a barrier to the desorption of hydrogen, until the temperature rises enough for the oxygen to migrate into the bulk. The oxide itself only contains a small amount of hydrogen, as is evidenced when the oxide layer is removed, whereupon the desorption peak returns to the characteristic shape seen in as-received samples. Samples oxidised in pure water exhibit a slightly higher oxidation rate, and pass through kinetic transition much earlier than samples exposed to an elevated pH. Before transition pH has little effect on the hydrogen pick-up. After transition the hydrogen pick-up is much higher in pure water, which implies that the H{sup +} concentration in the environment affects hydrogen pickup at this stage. The elevated pH samples showed peaks for D-H and D{sub 2}, which follow the H{sub 2} peaks. The hydrogen content in these samples increases at around 70% of the time to the first transition. This may correlate with development of through-thickness percolation paths, which could lead to the pH inside cracks being modified to a lower value, increasing the availability of H{sup +} ions. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2018

High- and Low-Temperature Deformation Behavior of Different Orientation Hot-Rolled Annealed Zircaloy-4

In this paper, the hot-rolled annealed Zircaloy-4 samples with different orientation were subjected to uniaxial compression with a strain rate of 0.001 s{sup −1} to obtain the stress-strain curves of different initial orientation samples at different temperatures. Electron backscatter diffraction (EBSD) technique and transmission electron microscope (TEM) technique were used to analyze the microstructures and textures of compressed samples. The mechanical properties and microstructural evolution of rolling directions (RD), transverse directions (TD) and normal directions (ND) were investigated under the conditions of – 150 °C low temperature, room temperature and 200 °C high temperature (simulated lunar temperature environment). The results show that the strength of Zircaloy-4 decreases with the increase in deformation temperature, and the strength in three orientations is ND > TD > RD. The deformation mechanism of hot-rolled annealed Zircaloy-4 with different orientation is different. In RD, {101̄0}〈a〉 prismatic slip has the highest Schmid factor (SF), so it is most easy to activate the slip, followed by TD orientation, and ND orientation is the most difficult to activate. The deformed grains abide slip→twinning→slip rule, and the different orientation Zircaloy-4 deformation mechanisms mainly are the twinning coordinated with the slip.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Synergetic Strengthening of Grain Refinement and Texture in Gradient Zircaloy-4 by Surface Mechanical Rolling Treatment

A Zircaloy-4 rod was subjected to surface mechanical rolling treatment (SMRT) to form a gradient structure, and the evolution of the resulting sub-grain-boundary property and texture across the gradient structure was characterized using electron backscattered diffraction. Dual-gradient structures in grain size and orientation were formed; the grain size was refined from several microns at the center to approximately 400 nm at the topmost surface. Texture analysis revealed that the c-axis gradually tilted from a random orientation toward the parallel-to-radial direction on the radial–tangential plane. The SMRT-induced formation of the dual-gradient microstructure is attributed to the formation of gradient distributions of stress and strain, which resulted in various deformation mechanisms (twinning and dislocation) being active at different depths. During the SMRT process, twinning and dislocations were activated to refine the grains. When the dual-gradient microstructure formed, twinning was mainly activated at the subsurface near the matrix, whereas dislocations were activated across the entire gradient. The geometrically necessary dislocation density increased with decreasing depth and then slightly decreased near the surface. The synergetic strengthening of the dual-gradient microstructure resulted in a gradient distribution of the microhardness near the surface. Thus, the Zircaloy-4 rod exhibited a good combination of strength and ductility.

36 MATERIALS SCIENCE↗
DOE OSTI2013

Design study of the geometry of the blanking tool to predict the burr formation of Zircaloy-4 sheet

In this work, we investigated factors that influence burr formation for zircaloy-4 sheet used for spacer grids of nuclear fuel roads. Factors we considered are geometric factors of punch. We changed clearance and velocity in order to consider the failure parameters, and we changed shearing angle and corner radius of L-shaped punch in order to consider geometric factors of punch. First, we carried out blanking test with failure parameter of GTN model using L-shaped punch. The tendency of failure parameters and geometric factors that affect burr formation by analyzing sheared edges is investigated. Consequently, geometric factor's influencing on the burr formation is also high as failure parameters. Then, the sheared edges and burr formation with failure parameters and geometric factors is investigated using FE analysis model. As a result of analyzing sheared edges with the variables, we checked geometric factors more affect burr formation than failure parameters. To check the reliability of the FE model, the blanking force and the sheared edges obtained from experiments are compared with the computations considering heat transfer.

97 MATHEMATICS AND COMPUTING↗