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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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At least 865 records · Page 48
All sourcesNASA NTRSDOE OSTI
NASA NTRS1984

Formation of ion damage tracks

A new model is proposed to explain both localized damage regions and preferential etching of damage tracks caused by the passage of energetic ions in insulators. The formation of each region of extended defects is initiated by the Auger decay of a vacancy produced in an inner electronic shell of an atom of the insulator by an incident ion. This decay produces an intense source of ionization within a small volume around the decaying atom, which causes decomposition of the material in a manner similar to that observed in pulsed laser irradiation. The resulting chemical or crystalline modification of the material is the latent track, which can be preferentially etched due to its changed structure.

Tombrello, T. A.↗
DOE OSTI2025

Development of Surveillance Test Articles for Materials Degradation Management in MSR Environments

Materials in molten salt reactors (MSR) undergo accelerated degradation from corrosion, irradiation, and cyclic loads at elevated temperatures. Establishing a materials surveillance program to enable the assessment of material deterioration is critical to assure structural integrity of MSRs components. This presentation summarizes recent efforts towards the development of surveillance test articles for collecting various damages for monitoring materials degradation. Surveillance test articles with reduced dimensions were designed to capture creep-fatigue damage from cyclic loading at elevated temperatures. The strain evolution of test articles during thermal cycling was analyzed both numerically and experimentally. Out-of-reactor thermal cycling demonstrated successful capture of strain range for materials assessment. Moreover, test articles were subject to both mechanical loads and molten salt exposure, and the damage due to stress and corrosion was investigated. Additionally, damage inference models were developed to predict the remaining life of materials based on the accumulated damage in surveillance test articles.

36 - MATERIALS SCIENCE↗
DOE OSTI2021

High Entropy Alloys: Irradiation

High entropy alloys (HEAs) have been considered as structural materials for nuclear applications due to their promising mechanical properties and radiation resistance. For this purpose, irradiation-induced defect evolution and microstructure change have been characterized to evaluate the irradiation performance of different HEAs. This article reviews recent advances in understanding the irradiation response of HEAs, including the effect of disordered states on the defect energy landscape and defect evolution, irradiation-induced microstructure changes, void swelling, phase stability, mechanical properties under irradiation, as well as the He irradiation effects.

Zhao, Shijun↗
DOE OSTI2023

A quantitative method to determine the region not influenced by injected interstitial and surface effects during void swelling in ion-irradiated metals

We propose and demonstrate a microstructurally-based experimental method to quantitively determine the depth regions in self-ion-irradiated metals that are affected by the injected interstitial effect and various surface effects, focusing on the choice of safe analysis zones to minimize the impact of these phenomena. The goal is to define the depth ranges where extracted data can be confidently applied to ion-neutron correlations for reactor application. Since ion energies in the range of 1–5 MeV are most frequently employed by the radiation effects community, irradiations were conducted at four energies in this range, all proceeding at 475 °C. The experiment was conducted on relatively pure single crystal iron to focus only on physical phenomena, avoiding the influence of possibly confounding chemical or segregation processes. Care was also taken to minimize the influence of other physical factors such as crystalline orientation. It was shown that, at 475 °C, ion energies of ≤1 MeV were too shallow in penetration and did not yield a safe depth range, but irradiations at 2.5 MeV and above yielded useful safe zones with predicted swelling behavior becoming independent of ion energy. The surface-affected zone width was found to be roughly twice that of the void-denuded zone width and to be independent of accumulated displacement dose. The largest injected-interstitial effect arises from the injected interstitial depression of void nucleation and growth. The interstitial-affected region starts at about one half of the projected range and does not show any “spreading” of its influence in depth as the peak damage level increases from 50 to 100 dpa. This study provides some confidence that enhances the credibility of ion simulation when applied to prediction of void swelling in neutron environments.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Evaluation of two extraction chromatography resins for scandium and titanium separation for medical isotope production

Scandium-47 ( 47 Sc) can be used in nuclear medicine as a therapeutic-diagnostic, or "theragnostic," radioactive medical isotope for cancer detection and treatment. The 47 Sc isotope can be produced through the photonuclear reaction 48 Ti(γ,p) 47 Sc by irradiating enriched 48 Ti target material. The enriched target material necessary for production is costly; 48 TiO 2 costs ~ $\$$1550/g, and targets can be > 50 g ($\$$77,500) to produce medically relevant amounts of 47 Sc. In order to keep costs low, a highly efficient separation of scandium from bulk titanium is desired, along with efficient methods for recycling the target material. Here this research is focused on evaluating efficient methods for the separation of scandium from bulk quantities of titanium using commercially available diglycolamide-based and hydroxamate-based extraction chromatography resins (DGA resin and ZR resin, respectively). The sorption of 47 Sc and Ti on these resins were investigated at varying concentrations of HNO 3 , HCl, H 2 SO 4 , and HF to explore how they might be used in a large-scale production/processing setting.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2020

Temperature effects on damage evolution in ion-irradiated NiCoCr concentrated solid-solution alloy

In this work, the damage evolution in equiatomic NiCoCr single crystals irradiated at 138 and 500 K with 1.5 MeV Ni+ ions has been investigated using ion channeling analysis and compared with previous results obtained at 300 K under identical irradiation conditions. As expected, irradiation temperature has a major influence on damage evolution and long-range effects in ion-irradiated NiCoCr. A change in relative irradiation resistance as a function of temperature is revealed when the ion channeling data are compared with previous ion channeling results from Ni-irradiated NiFe at 150, 300 and 500 K. Additionally, NiCoCr has superior irradiation resistance than NiFe under ion irradiation at ≤ 300 K; surprisingly, the relative irradiation resistance is reversed under ion irradiation performed at > 300 K (i.e., 500 K), which may be attributed to local chemical ordering. These observations indicate that the NiFe alloy exhibits a more favorable chemical environment for suppressing deleterious microstructural degradation under high temperature irradiation than the NiCoCr alloy.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Experimental results of conductive inserts to reduce nuclear fuel temperature during nuclear volumetric heating

Advanced fuel designs that incorporate thinner fuel UO2 pellets interspaced by high thermal conductivity inserts have been proposed, with the primary goals of reducing peak centerline temperatures and temperature gradients across fuel pellets and enhancing heat transfer from the fuel to the coolant. An initial series of experiments has been performed on this design, including laboratory experiments and a series of experiments using the Idaho National Laboratory (INL) Transient Reactor Test (TREAT) Facility, the latter of which compared thermal gradient driven fracture of standard pellet designs with that in the proposed advanced fuel design. Although reducing fracture is not the primary objective of the new fuel design, the lower thermal gradients are expected to reduce fracture, so it can serve as an indicator of the thermal behavior of this fuel in the reactor. The in-reactor tests were conducted at multiple linear heat generation rates and confirm the expected result that fracture in both the standard and advanced fuel pellets occurs during the first ramp to power in standard light-water reactor conditions. Post-irradiation examination of the experiment material was performed and included quantification of the extent of fracture in the fuel pellets. It was found that the advanced-design pellets reduce the extent of fracture in a statistically significant way. This confirms the expected behavior predicted by two-dimensional axisymmetric models of this experiment. Here in this study is an important first experimental confirmation of the efficacy of the proposed inserts for achieving their desired effect on the thermal behavior of the fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Integrated simulation of U-10Mo monolithic fuel swelling behavior

Here, a separate computational branch has been implemented within the DART (Dispersion Analysis Research Tool) computational code to simulate the swelling behavior of U-10Mo monolithic fuel under the operating conditions of high-power research and test reactors (RTRs). The monolithic branch of the DART code implements a mechanistic rate-theory-based fission-gas-behavior model for the calculation of fission gas swelling, as well as a suite of thermal, physical, and mechanical models to take into account various processes occurring in RTR fuels during irradiation. In order to accurately simulate and eventually predict U-10Mo monolithic fuel irradiation behavior, the code uses materials properties calculated with lower length-scale computational methods, such as gas atom diffusivity and U-Mo surface energy from atomic simulations and grain-morphology-specific recrystallization kinetics (recrystallized fuel volume fractions vs. fission density) predicted using the phase-field method. The remainder of fission gas behavior parameters used in the model were calibrated with measured intergranular bubble size distributions. With this integrated simulation approach, the swelling behavior of U-10Mo monolithic fuel was simulated for various initial grain sizes at different operating conditions and compared with measured data. Furthermore, because limited experimental data exist for parameter calibration detailed sensitivity studies for the important parameters used in the fission gas behavior model were performed in order to examine their impact on both intergranular gas bubble morphology at low fission density, and on total porosity at high fission density.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Review of SiC material development for nuclear fusion applications: Cross-cutting research and emerging opportunities

The SiC-based materials, particularly SiC-fiber-reinforced SiC matrix (SiC/SiC) composites, show strong potential for structural and functional applications in future fusion power plants because they can operate at high temperatures with a range of coolants and breeders, thereby enabling higher energy conversion efficiency. Here, this paper presents recent advancements in the development of SiC-based materials, focusing on processing techniques and material performance and resistance under fusion-relevant environments. The processing activities have emphasized near-net-shape fabrication and the joining of SiC subcomponents, with processing methods and material compositions informed by previous irradiation experiments on various grades of SiC. Research on irradiation effects has remained focused on degradation mechanisms and the microstructural optimization of SiC/SiC composites irradiated to high neutron damage levels. Analysis of irradiation defects in SiC has advanced via the application of cutting-edge characterization methods, among which Raman spectroscopy is becoming a common tool to assess atomic-scale chemical disorder. Fusion–fission crosscutting irradiation research has explored combined effects in SiC/SiC composites with application-relevant geometries, including bowing of SiC/SiC composite channels under neutron flux gradients, stress evolution in SiC/SiC composite tubes under through-thickness temperature gradients, and irradiation-enhanced corrosion in SiC. Finally, research opportunities for component testing and assessment under fusion-relevant conditions, in support of emerging concepts from the private fusion sector, are discussed.

Advanced manufacturing↗
DOE OSTI2026

3D probabilistic fracture mechanics / computational fluid dynamics simulation of a reactor pressure vessel under transient conditions

Reactor pressure vessels (RPVs) are safety-critical light-water-reactor components that, under irradiation, experience long-term material degradation in the form of embrittlement. This can increase their susceptibility to fracture under thermal-shock conditions, which could occur during off-normal transients such as loss-of-coolant accidents (LOCAs). During a LOCA, the most severe conditions for the RPV occur when emergency core cooling water is injected through the cold legs into the water-and-steam-filled RPV. The rapid cooling of the downcomer and internal RPV surface causes decreased temperature and elevated thermally driven tensile stresses in the RPV wall. This, combined with long-term material embrittlement, may cause fracture initiation at pre-existing flaws, challenging the integrity of the RPV. Assessing RPV integrity during transients with a large spatial variation in the coolant temperature requires a modeling approach that considers the effects of spatially varying coolant temperature on the fracture probability of a population of flaws distributed throughout the RPV, accounting for spatially varying embrittlement. Here, the present study addresses this need by demonstrating first-of-its-kind coupling of a high-fidelity 3D computational fluid dynamics code with 3D probabilistic fracture mechanics This was accomplished using representative models of a pressurized-water reactor subjected to small- and medium-break LOCA conditions, both of which can result in large spatial temperature variations. While the observed impact of accounting for 3D effects was minimal under the small-break LOCA this study indicates a significant increase in the probability of fracture initiation under the medium-break LOCA when 3D effects are considered, relative to a spatially uniform cooling scenario.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Monte Carlo simulations demonstrating physics of equivalency of gamma, electron-beam, and X-ray for radiation sterilization

We report the sterilization of medical devices using the gamma rays from the decay of cobalt-60 has accumulated decades of experience of the performance of the materials and devices that are irradiated. The use of radiation using electron beams and x-rays has much less experience and this leads to questions of equivalency between these three technologies. Computer simulations were conducted to model the relevant physical processes of the interactions of each of the three forms of radiation in order to compare the spectra of electron energies at energies below 500 keV. It is predominantly the electrons below this threshold that produce the sterilization dose. No difference in energy spectra was seen between the three types of initial radiation. It is concluded that there is no energy dependent difference between gamma, e-beam, and x-ray for radiation sterilization.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Unconventional Pathways to Carbide Phase Synthesis via Thermal Decomposition of UI 4 (1,4-dioxane) 2

UI4(1,4-dioxane) 2 was subjected to laser-based heating-a method that enables localized, fast heating (T > 2000 °C) and rapid cooling under controlled conditions (scan rate, power, atmosphere, etc.)-to understand its thermal decomposition. A predictive computational thermodynamic technique estimated the decomposition temperature of UI 4 (1,4-dioxane) 2 to uranium (U) metal to be 2236 °C, a temperature achievable under laser irradiation. Dictated by the presence of reactive, gaseous byproducts, the thermal decomposition of UI 4 (1,4-dioxane) 2 under furnace conditions up to 600 °C revealed the formation of UO 2 , UI x , and U(C 1–x O x ) y , while under laser irradiation, UI 4 (1,4-dioxane) 2 decomposed to UO 2 , U(C 1–x O x ) y , UC 2–z O z , and UC. Despite the fast dynamics associated with laser irradiation, the central uranium atom reacted with the thermal decomposition products of the ligand (1,4-dioxane = C 4 H 8 O 2 ) instead of producing pure U metal. In conclusion, the results highlight the potential to co-develop uranium precursors with specific irradiation procedures to advance nuclear materials research by finding new pathways to produce uranium carbide.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Design of Stand-Alone Cryomodule based on Superconducting Quarter-Wave Resonator

Present superconducting accelerators require a central refrigeration system and piping to distribute liquid helium and nitrogen. Stand-alone cryocooler-based systems are of significant interest both to scientific and industrial customers, as they allow applications at facilities where no cryogenic equipment and expertise exists, where it is undesirable to develop additional cryogenic systems, or where a mobile system that can be transported between beamlines or facilities is required. The technical challenge of this approach is to minimize the heat losses to the liquid helium so that commercially available, self-contained cryocoolers are able to handle the thermal load. This paper presents an improved design for a stand-alone quarter-wave resonator (QWR) cryomodule a based on commercially available cryocoolers. This cryomodule has a particular application in the ATLAS intensity upgrade project where it will house a 72.75 MHz QWR for beam energy adjustments to experimental areas, where no helium cryogenic systems exist. The described cryomodule may also be the critical part of a turnkey system with application to material processing, semiconductor manufacturing, food irradiation and homeland security. The development of low heat loss peripheral components reduces cost and widens applicability of the system. We will discuss the cryomodule and RF coupler designs, including numerical thermal and structural analyses.

43 PARTICLE ACCELERATORS↗
DOE OSTI2021

OpenMC

OpenMC is an open source, community-developed framework for performing Monte Carlo particle transport calculations.It is capable of simulating neutron and photon transport below ~50 MeV and also includes a model for bremsstrahlung production of photons by electrons/positrons. Fixed source, k-eigenvalue, and subcritical multiplication problems can be solved. Problem geometries can be modeled either using constructive solid geometry or a CAD representation. A flexible and efficient tally system enables a wide variety of physical quantities to be tallied and analyzed. OpenMC can run in parallel using a hybrid MPI and OpenMP programming model and has been extensively tested on leadership class supercomputers. A built-in Bateman equation solver enables the modeling of material composition changes due to irradiation. One of the unique features of OpenMC is its rich, extensible Python and C/C++ programming interfaces that enable programming pre- and post-processing, multigroup cross section generation, workflow automation, multiphysics coupling, and the visualization of geometry and tally results. In addition to the core Monte Carlo transport solver and associated APIs, OpenMC includes a Python-based nuclear data interface that enables power users to inspect, modify,and perform various types of nuclear data processing on ENDF, ACE, and OpenMC¿s native HDF5 files

ROMANO, PaulK.↗
DOE OSTI2024

Proton beam power limits for stationary water-cooled tungsten target with different cladding materials

The proton beam power limit for a solid-tungsten spallation target is largely determined by beam induced thermomechanical structural loads and decay heat power deposition, while its lifetime is limited by radiation damage and fatigue life of the target materials. In this paper, we studied the power limits of a stationary water-cooled solid tungsten target concept. Tantalum clad tungsten was considered as a reference case. Being a low activation material, zircaloy 2 cladding option was studied and its decay heat driven power limit was compared with the reference case. Zirconium alloys have proven operations records in spallation target and nuclear fission environments, supported by materials data obtained from post irradiation examinations. Recent study also demonstrated feasibility of diffusion bonding zirconium to tungsten using vanadium foil inter layer. Particle transport simulations code FLUKA was used to calculate energy deposition and decay heat power deposition in the target, based on the beam parameters technically feasible at the Second Target Station of the Spallation Neutron Source at Oak Ridge National Laboratory. The energy deposition data were used for flow, thermal, and structural analyses to determine the beam intensity limit on the target concept studied. The decay heat deposition data were used to calculate the transient temperature evolution in the tungsten volumes in a loss of coolant accident (LOCA) scenario to determine its beam power limit. For a 1.3 GeV proton beam, the power limit on a stationary target was 400 kW for a tantalum clad target model and 800 kW for a zircaloy 2 clad target model.

Lee, Yong Joong↗
DOE OSTI2021

Planarization, densification, and exfoliation of porous materials by high-energy ion beams

A method and system for providing at least one of planarization, densification, and exfoliation of a porous material using ion beams. The method may use an ion beam generator to generate an ion beam, the ion beam having energy above 0.1 MeV. The ion beam generator may irradiate the surface of a porous material with the ion beam to produce at least one of planarization, densification, and exfoliation of the porous material.

Kucheyev, Sergei O.↗
DOE OSTI2022

Cracking in first ramp to power of standard versus high thermal conductivity UO2 pellets via internal nuclear heating

Advanced fuel designs that incorporate thinner fuel UO2 pellets interspaced by high thermal conductivity inserts have been proposed, with the primary goals of reducing peak centerline temperatures and temperature gradients across fuel pellets and enhancing heat transfer from the fuel to the coolant. An initial series of experiments has been performed on this design, including laboratory experiments and a series of experiments using the Idaho National Laboratory (INL) Transient Reactor Test (TREAT) Facility, the latter of which compared thermal gradient driven fracture of standard pellet designs with that in the proposed advanced fuel design. Although reducing fracture is not the primary objective of the new fuel design, the lower thermal gradients are expected to reduce fracture, so it can serve as an indicator of the thermal behavior of this fuel in the reactor. The in-reactor tests were conducted at multiple linear heat generation rates and confirm the expected result that fracture in both the standard and advanced fuel pellets occurs during the first ramp to power in standard light-water reactor conditions. Post-irradiation examination of the experiment material was performed and included quantification of the extent of fracture in the fuel pellets. It was found that the advanced-design pellets reduce the extent of fracture in a statistically significant way. This confirms the expected behavior predicted by two-dimensional axisymmetric models of this experiment. This study is an important first experimental confirmation of the efficacy of the proposed inserts for achieving their desired effect on the thermal behavior of the fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗