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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.

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50 records · Page 3
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Fundamental Studies of Tritium Diffusivity in Pure and Defective Zircaloy-4 Getters (Tritium Science Program FY2020 Report)

Zirconium (Zr) alloys are extensively used as tritium (T) getters in nuclear reactors due to their low absorption cross section to thermal neutron, good mechanical properties at high temperature and resistant to corrosion in different environmental conditions. Zr alloys have better thermal properties than many other refractory alloys including stainless steel. The nuclear characteristic is simulated, and T is produced in tritium-producing burnable absorber rods (TPBAR) by the irradiation of neutron flux in pressurized water reactor (PWR). T thus produced diffuses through the pellets and is captured by the getter where it chemically reacts with Zr to form metal hydride (ZrTx). These hydrides are brittle and affect adversely the mechanical properties of alloys. In addition to that, mismatch in the lattice structures after hydride formation creates a stress which needs to be considered in component design and their life evaluation. Therefore, understanding the behavior of T and its species becomes significant as fuel burnup is increased, which leads to increase in hydrogen pickup and oxide formation. The transport of T and its species in pure Zr alloys, and in alloys with certain alloying elements and defects are important to understand in order to enhance the performance of the materials.

36 MATERIALS SCIENCE↗
DOE OSTI2021

First-principles Studies of Tritium Species Formation on and Desorption from Surfaces of $γ- LiAlO_2$ Pellets with Related Secondary Phase

The $γ- LiAlO_2$ ceramics enriched with the $^6Li$ isotope are used in the form of annular pellets in tritium-producing burnable absorber rods (TPBARs). When irradiated in a pressurized water reactor (PWR), the $^6Li$ absorbs neutrons, simulating the nuclear characteristics of a burnable absorber rod, and produces tritium $(_{1}^{3}H, T)$. Once the $_{1}^{3}H$ is produced in the bulk of $γ- LiAlO_2$, it diffuses to its surface, forming different $_{1}^{3}H$ species (possibly $T, OT, T_2, or T_2O$). The $T_2O$ is expected to be generated from experiment, to avoid $_{1}^{3}H$ leakage. However, up to now, it is still not clear which $_{1}^{3}H$ species are formed on the $γ- LiAlO_2$ surface and travel to the getter. Besides, with the dramatic Li loss under irradiation, the secondary phase (possibly $LiAl_5O_8$) precipitates and participates into the $_{1}^{3}H$ diffusivity and $_{1}^{3}H$ species formation. Therefore, exploring $_{1}^{3}H$ species formation on and desorption from surfaces of $γ- LiAlO_2$ with related secondary phases becomes a very important task to provide a better understanding of $_{1}^{3}H$H species escaping from pellets.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Fundamental Studies of Tritium Formation and Diffusivity in Pure and Defective Zircaloy-4 Getters

Zirconium (Zr) and its alloys are used as fuel cladding in nuclear reactors. These materials have their low absorption cross-section for thermal neutrons and have good mechanical and thermal properties. In addition, they are used as tritium getters in tritium-producing burnable absorber rods (TPBARs), in the form of nickel-plated Zircaloy-4 tubes. To produce 3 H, the breading blankets such as lithium aluminate are irradiated in pressurized water. 3 H produced in this process is captured by the getter and 3 H reacts chemically with Zr to form metal hydride (ZrTx). The precipitation of metal hydride results in not only the volume increase and hydrogen embrittlement of zircaloy but also changes in getter’s chemical properties. This degrades the getter’s performance. Therefore, understanding the behavior of 3 H (such as diffusion and solubility) in zircaloy is important to understand getter performance.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

First-Principles Studies of Tritium Species Dissociability & Diffusivity Across the Interface of Nickel-Plated Zircaloy-4

Zirconium (Zr) and its alloys (Zircaloy-4) are widely used in nuclear reactors due to their low neutron adsorption cross-section and excellent corrosion resistance. In tritium-producing burnable absorber rods (TPBARs), the metal getter tube located between the cladding and the γ-LiAlO 2 pellets is composed of nickel (Ni)-plated Zircaloy-4, which is used to capture tritium ( 3 H) species (mainly 3 H 2 and 3 H 2 O) generated from γ-LiAlO 2 pellets during irradiation. The 3 H-related products transfer to the surface of metal Ni upon adsorption and dissociation to form new 3 H species and diffuse into the Zircaloy-4 getters to form metal hydrides (Zr 3 H x ). Therefore, exploring 3 H species ( 3 H 2 , 3 H 2 O) dissociation on the surface of Ni and diffusion across the interface of Ni-plated Zircaloy-4 getters can provide a better understanding of 3H species formation and transport from pellets into the getters.

36 MATERIALS SCIENCE↗
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 OSTI2025

Investigating the Impact of Absorbed Surface Impurities (CO, CO 2 , CH 4 , and H 2 O) on Getter Rate Performance

This report provides an overview of engineering and experimental efforts to establish a new capability for measuring H2 uptake rates for Tritium Producing Abosrber Rod (TPBAR) getter materials. This new system was used to gain a better understanding of the impact of contaminant species (CO 2 , CO, CH 4 , and H 2 O) on getter performance under relevant hydrogen uptake conditions (350 °C, 20 torr). Herein, we successfully established a high-resolution gravimetric sorption (HRGS) system in a flow through configuration that can measure sample mass changes with high resolution (1 µg or 10 µg) over long experiment times (days), high temperature, and varying pressures. We coupled this capability with an in-line universal gas analyzer (UGA) system to characterize potential surface chemistry side reactions occurring after contaminant gas exposure. The major findings of this study suggest that methane and CO have the most detrimental effects on getter H 2 uptake performance showing large changes in rate after exposure compared to CO 2 and water. The effect of these contaminant species did not influence getter sample capacity (~ 1.8 wt% H 2 ). After CO and methane exposure, changes in getter rate were accompanied by observable Ni-surface defects and pinholes via SEM imaging suggesting possible Ni- layer corrosion. Additionally, the increase in C and O signal in the zircaloy layer, observed via EDX, point to C and O species transport through the getter sample resulting in the formation of carbides and oxide species. Thus, co-exposure studies where mixed gases or water vapor are introduced together will be important to further understand what mechanisms are contributing to Ni-layer corrosion and changes in getter performance.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Atomistic Modeling of the Interphase Structure between Alumina and Al12Fe2Cr Quasicrystal Approximant

Molecular dynamics simulations are performed to study the stability of an fcc metallic interlayer phase between an iron aluminide quasicrystal approximant (QCA) phase (84.2 Al, 5.7 Fe, 5.4 Ni, and 4.7 at.% Cr) and aluminum oxides (α-Al 2 O 3 and γ-Al 2 O 3 ). The presence of the interlayer phase was experimentally observed in some regions of the inner diameter surface of a TPBAR cladding tube. The simulations employed existing many-body (MEAM and EAM) interatomic potentials to describe atomic interactions in the interlayer and QCA phases, and Buckingham+Coulomb pair interactions for the alumina and between the alumina and the QCA (or fcc). Interface systems of QCA/alumina, fec/alumina, and QCA/fcc are modeled and the fracture energy of the interfaces is calculated. Based on the fracture energies, for the γ-Al 2 O 3 , both MEAM and EAM suggest that the fcc interlayer phase would not form. For the α-Al 2 O 3 , the MEAM predicts that the fcc phase would form, however, the EAM predicts that it forms only if the α-Al 2 O 3 is Al-terminated at the interface. For the O-terminated α-Al 2 O 3 , the EAM predicts the fcc phase would not form, consistent with the results for γ-Al 2 O 3 .

36 MATERIALS SCIENCE↗
DOE OSTI2021

Fundamental Studies of Tritium Formation and Diffusivity in Pure and Defective Zircaloy-4 Getters

Zirconium (Zr) alloys have excellent mechanical properties at high temperature and are resistant to corrosion in different environmental conditions. These alloys are used as tritium (3H or T) getters in nuclear reactors due to their low absorption cross section to thermal neutrons and excellent thermo-mechanical properties. Zr alloys have better mechanical and thermal properties than many other refractory alloys including stainless steel. After the irradiation of TPBAR by neutron flux, 3H produced during irradiation diffuses though the pellets and is captured by the getter. 3H chemically reacts with Zr metal to form metal hydride (ZrTx). Hydrides thus formed are brittle and adversely affect the mechanical and thermal properties of alloys. The lattice mismatch at the interface of alloy and metal hydrides also creates a stress that significantly reduces the alloy’s performance and operation life. Therefore, understanding the behavior of 3H and its species becomes significant as fuel burnup is increased, which leads to increase in hydrogen pickup and oxide formation.

Duan, Yuhua↗
DOE OSTI2021

Molecular dynamics studies of lattice defect effects on tritium diffusion in zirconium

Tritium diffusion in α-Zr containing point defects such as vacancies or self-interstitial atoms (SIAs) is simulated using molecular dynamics. Point defects rapidly aggregate to form extended defects, such as 3D nanoclusters and Frank loops. The geometry of extended defects is affected by the presence of tritium. At low temperature and in the absence of tritium, vacancies aggregate to form stacking fault pyramids. Addition of tritium at these temperatures promotes aggregation of vacancies to form 3D nanoclusters, within which the tritium concentration can be sufficiently high to suggest that these defects may serve as nucleation sites for hydride precipitation. Trapping of tritium in vacancy nanocluster reduces the calculated bulk diffusivity by an amount proportional to the vacancy concentration. At high temperature, vacancy clusters change shape to form planar basal dislocation loops, which bind tritium less strongly, leading to a sharp reduction in the fraction of trapped tritium and a corresponding increase in tritium diffusivity at high temperature. In contrast, SIAs increase tritium diffusion through α-Zr. Analysis of atomic trajectories shows that tritium does not interact directly with SIAs. In conclusion, diffusion enhancement is instead related to expansion of the lattice.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Molecular dynamics study of interstitial He clusters in nickel

This study presents a molecular dynamics analysis focusing on the behavior of interstitial helium (He) clusters in nickel (Ni), examining their formation, stability, and migration energetics. Consistent with previous research, we found that the binding energies of interstitial helium within a helium cluster are positive and increase with the cluster size, indicating a preference for helium atoms to cluster together. However, our findings also reveal that while the formation energy increases monotonically with cluster size, the increase in binding energy is non-monotonic. Importantly, small He clusters were observed to be thermally unstable at reactor operational temperatures (approximately 600 K), with the He 2 cluster exhibiting instability even at room temperature. With a binding energy of 0.49 eV for a He 4 cluster, we hypothesize that for helium bubbles to form via homogeneous nucleation (i.e., through trap mutation) at reactor operating temperatures, the helium concentration must be high enough to facilitate the formation of helium clusters of at least size 4 or larger. As expected, interstitial helium and small helium clusters are highly mobile. This mobility was observed not only at room temperature but also at temperatures as low as approximately 200 K. Furthermore, the mean squared displacement method has been utilized to determine the migration barriers and the corresponding prefactors for clusters ranging from He 1 to He 6

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

PORFLOW Modeling of Vadose Zone Flow and Transport for E-Area Intermediate Level Vault

In support of the E-Area Performance Assessment, a two-dimensional model of water flow and radionuclide transport through the E-Area Intermediate Level Vault (ILV) and local vadose zone has been developed using the PORFLOW TM software. The purpose of the model is to calculate flux to the water table for radionuclides eluted from the ILV during its operational life, the period of institutional control, and times following site closure. Results of model calculations will be used by a three-dimensional PORFLOW model of transport through the aquifer to determine radionuclide concentrations at a hypothetical 100 meter well and at the site boundary where contaminated groundwater is accessible to members of the public following site closure.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗