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At least 289 records · Page 16

Ultra-strong tungsten refractory high-entropy alloy via stepwise controllable coherent nanoprecipitations

High-performance refractory alloys with ultrahigh strength and ductility are in demand for a wide range of critical applications, such as plasma-facing components. However, it remains challenging to increase the strength of these alloys without seriously compromising their tensile ductility. Here, we put forward a strategy to “defeat” this trade-off in tungsten refractory high-entropy alloys by stepwise controllable coherent nanoprecipitations (SCCPs). The coherent interfaces of SCCPs facilitate the dislocation transmission and relieve the stress concentrations that can lead to premature crack initiation. As a consequence, our alloy displays an ultrahigh strength of 2.15 GPa with a tensile ductility of 15% at ambient temperature, with a high yield strength of 1.05 GPa at 800 °C. The SCCPs design concept may afford a means to develop a wide range of ultrahigh-strength metallic materials by providing a pathway for alloy design.

36 MATERIALS SCIENCE↗

Effect of Solar Exposure on the Atomic Oxygen Erosion of Hubble Space Telescope Aluminized-Teflon Thermal Shields

When exposed to low Earth orbital (LEO) environment, external spacecraft materials degrade due to radiation, thermal cycling, micrometeoroid and debris impacts, and atomic oxygen (AO) interaction. Collisions between AO and spacecraft can result in oxidation of external spacecraft surface materials, which can lead to erosion and severe structural and/or optical property deterioration. It is therefore essential to understand the AO erosion yield (Ey), the volume loss per incident oxygen atom (cu cm/atom), of polymers to assure durability of spacecraft materials. The objective of this study was to determine whether solar radiation exposure can increase the rate of AO erosion of polymers in LEO. The material studied was a section of aluminized-Teflon (DuPont) fluorinated ethylene propylene (Al-FEP) thermal shield exposed to space on the Hubble Space Telescope (HST) for 8.25 years. Retrieved samples were sectioned from the circular thermal shield and exposed to ground laboratory thermal energy AO. The results indicate that the average Ey of the solar facing HST Al-FEP was 1.9 10(exp -24)cu cm/atom, while the average Ey of the anti-solar HST Al-FEP was 1.5 10(exp -24)cu cm/atom. The Ey of the pristine samples was 1.6- 1.7 10(exp -24)cu cm/atom. These results indicate that solar exposure affects the post-flight erosion rate of FEP in a plasma asher. Therefore, it likely affects the erosion rate while in LEO.

Guo, Aobo↗

Insights into elevated temperature tensile deformation mechanisms and kink banding in additively manufactured tungsten

Additive manufacturing (AM) potentially enables fabrication and repair of plasma facing components (PFCs) of tungsten. However, deformation mechanisms in AM-fabricated tungsten under service-relevant thermomechanical conditions remain unknown majorly due to challenges in achieving crack-free W. This study reveals the deformation mechanisms operative under tension at 800 °C and 1200 °C in the electron beam melting powder bed fusion (EBM-PBF) fabricated W. Textured columnar grains with mixed <001 >/<111 >|| build-direction were observed. Strong anisotropy in tensile properties persisted across test temperatures, and all specimens exhibited extensive deformation-induced banding phenomena. Grain boundary character analysis of the deformed specimens indicated that these deformation bands were kink bands with tilt grain boundaries prominently present at the band-matrix interface. Interestingly, the material within the kink bands rotated toward higher resolved shear stress, providing insights into the origins of kink band formation in body-centered cubic refractory metals. Intragranular misorientation axis analysis revealed that plasticity was dominated by {110} <111 > slip at 800 °C, whereas additional (213)[11⁢̄1] and (112)[11⁢̄1] slip systems activated at 1200 °C. A dense low-angle boundary networks observed near fractured surface indicated the onset of dynamic recrystallization. Results reveal plasticity governing mechanisms in AM-fabricated W under power plant-relevant conditions, and provide insights into kink band attributes in refractory metals.

Mayes, Riley [ORNL] (ORCID:000900089307010X)↗

Lithium–Divertor Interactions and Helium/Hydrogen Trapping in Lithiated Metals (Final Technical Report)

The goal of this project was to develop a fundamental understanding of helium and hydrogen behavior in lithium, both in bulk and at interfaces with tungsten, in order to inform the design of lithium based plasma facing components for fusion devices. Over the course of the award, the project produced the first comprehensive, peer reviewed dataset describing helium energetics, migration behaviors, and defect interactions in lithium. The research demonstrated that helium behaves in ways not previously observed in any other body centered cubic (BCC) metal: its interstitial configurations are more stable than substitutional ones, and its migration barriers are extraordinarily low, in some cases more than an order of magnitude below those in tungsten or iron. These discoveries reveal that helium in lithium diffuses so rapidly that its transport may be dominated by translational motion rather than the vibrationally activated mechanisms that underpin conventional solid state diffusion. This work lays a scientific foundation for understanding gas retention, bubble formation, and wall evolution in lithium based fusion environments and provides new computational tools, most notably a newly developed Li–He interatomic potential, for advancing future modeling efforts.

36 MATERIALS SCIENCE↗

Radiation damage in tungsten under high-energy He-ion irradiation

Tungsten (W) is a primary candidate for plasma-facing components (PFCs) in fusion reactors and hybrid fission-fusion reactors. However, W suffers from severe microstructural damage under Helium (He) ion irradiation in operation conditions. Key irradiation parameters, including incident ion energy, fluence, and irradiation temperature, are known to determine the formation of He bubbles and dislocation loops, which ultimately lead to the degradation of mechanical properties, such as radiation hardening and ductile-to-brittle transition temperature (DBTT) shifts. Various strategies, for instance, interface engineering and alloying, have been developed to enhance He irradiation tolerance in W to address these challenges. Studies in the recent two decades have enhanced the understanding of the underlying mechanisms that drive defect evolution and mechanical performance degradation, thereby offering more insights into the design of more resilient W-based materials for extreme environments.

Liu, Yinghang [Purdue University]↗

Hydrogen isotope analysis in W-tiles using fs-LIBS

Abstract Laser-Induced Breakdown Spectroscopy (LIBS) is a promising technology for in-situ analysis of Plasma-Facing Components in magnetic confinement fusion facilities. It is of major interest to monitor the hydrogen isotope retention i.e. tritium and deuterium over many operation hours to guarantee safety and availability of the future reactor. In our studies we use ultraviolet femtosecond laser pulses to analyze tungsten (W) tiles that were exposed to a deuterium plasma in the linear plasma device PSI-2, which mimics conditions at the first wall. A high-resolution spectrometer is used to detect the Balmer- $$\alpha$$ α transition of the surface from implanted hydrogen isotopes (H and D). We use Calibration Free CF-LIBS to quantify the amount of deuterium stored in W. This proof-of-principle study shows the applicability of femtosecond lasers for the detection of low deuterium concentration as present in first wall material of prevailing fusion experiments.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

A constrained-transport embedded boundary method for compressible resistive magnetohydrodynamics

Motivated by the increased interest in pulsed-power magneto-inertial fusion devices in recent years, we present a method for implementing an arbitrarily shaped embedded boundary on a Cartesian mesh while solving the equations of compressible resistive magnetohydrodynamics. The method is built around a finite volume formulation of the equations in which a Riemann solver is used to compute fluxes on the faces between grid cells, and a face-centered constrained transport formulation of the induction equation. The small time step problem associated with the cut cells is avoided by always computing fluxes on the faces and edges of the Cartesian mesh. We extend the method to model a moving interface between two materials with different properties using a ghost-fluid approach, and show some preliminary results including shock-wave-driven and magnetically-driven dynamical compressions of magnetohydrostatic equilibria. In conclusion, we present a thorough verification of the method and show that it converges at second order in the absence of discontinuities, and at first order with a discontinuity in material properties.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Active wall conditioning through boron powder injection compatible $\mathrm{ELM}$ control in $\mathrm{EAST}$

Real-time wall conditioning and edge-localized mode (ELM) suppression through real-time boron powder injection into high-confinement mode discharges of the Experimental Advanced Superconducting Tokamak (EAST) have been investigated. The boron powder injection effectively conditioned the plasma-facing components, as shown by the reductions in low-Z and high-Z impurities and the global recycling level in the whole inner vessel, and suppressed the strong magnetohydrodynamic (MHD) activity in the plasma core. The improved wall conditions also resulted in an earlier transition from L- to H-mode. Boron powder injection also effectively suppressed ELM in the lower single null configuration with BT in the favorable direction while leaving the plasma density and stored energy largely unaffected.

36 MATERIALS SCIENCE↗

Radiation Damage Analysis of FNSF Components Using McCad and MCNP

The Fusion Energy System Studies Fusion Nuclear Science Facility (FESS-FNSF) concept represents a transitional step between ITER and a commercial fusion power plant. The FNSF is a conceptualized D-T fueled tokamak with 518 MW of fusion power that has been extensively used to explore and optimize design features. The energetic 14.1-MeV neutrons can produce significant localized heating and activations, and can cause damage to plasma-facing components, which can determine maintenance/outage scheduling needs and also impact the lifetime of the device as a whole. This study illustrates a neutronics analysis that was conducted on a 22.5-degree symmetric sector of the FNSF with the goal of understanding the neutron heating and radiation damage that can be characterized by quantifying the displacements per atom (dpa). Concurrently, this study also focused on the development of analysis capabilities by converting a three-dimensional computer-aided design model of the FNSF into MCNP6.2 input using the McCad code. Accordingly, some confirmatory results on tritium production and the tritium breeding ratio (TBR) are provided to support model validation. The results produced by MCNP6.2 simulations showed that the highest heating and damage occurred in the outboard region, which concentrated approximately 290 MW of the total nuclear heating, in contrast to 97 MW within the inboard region. These results are consistent with previous studies that employed earlier versions of the FNSF concept and different modeling approaches. Furthermore, this study also provides additional details on neutron wall loading, as well as total heating from neutrons and gammas, results which show the total heating of the device (16 sectors) is approximately 477.83 ± 0.80% MW, indicating a neutron energy multiplication factor of 1.15. Additionally, the capability to calculate hydrogen and helium production, as well as dpa, is illustrated. Finally, the neutronics effects of using alternative materials to tungsten carbide were evaluated for the vacuum vessel, low-temperature shield, and structural ring components, which showed that compounds like YH 2 , Mg(BH 4 ) 2 , and ZrH 2 could reduce the total heating on the magnet and also reduce the TBR.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Wall heating by subcritical energetic electrons generated by the runaway electron avalanche source *

Abstract Subcritical energetic electrons (SEEs) produced by the runaway electron (RE) avalanche source at energies below the runaway threshold are found to be the primary contributor to surface heating of plasma-facing components (PFCs) during final loss events. This finding is supported by theoretical analysis, computational modeling with the Kinetic Orbit Runaway electrons Code (KORC), and qualitative agreement with DIII-D experimental observations. The avalanche source generates significantly more secondary electrons below the runaway threshold, which thermalize rapidly when well-confined. However, during a final loss event, the RE beam impacts the first wall, and SEEs are deconfined before they can thermalize. Additionally, because the energy deposition length decreases faster than energy, the deposited energy density, and thus the maximum PFC surface temperature change, is larger for SEEs than REs. KORC simulations employ an analytic first wall to model particle deconfinement onto a non-axisymmetric wall composed of individual tiles. PFC surface heating is calculated using a 1D model extended to include an energy-dependent deposition length scale. Simulations of DIII-D qualitatively agree with infrared (IR) imaging only when SEEs from the avalanche source are included. These results demonstrate that SEEs are the dominant contributor to PFC surface heating and indicate that the avalanche source plays a critical role in the PFC damage caused during final loss events. The prominence of SEEs also has important implications for interpreting IR imaging, one of the primary diagnostics for RE-wall interaction diagnosis, despite REs dominating the energy and current density. This result improves predictions of wall damage due to post-disruption REs to estimate material lifetime and design RE mitigation systems for ITER and future reactors.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Liquid lithium divertor analysis using coupled plasma material interaction model

A liquid lithium divertor can improve performance of future fusion devices by creating efficient power exhaust and improving the energy confinement via pumping of the hydrogen isotopes. In addition, significantly higher heat fluxes can be handled if controlled vapor shielding is used to redistribute the divertor heat flux over a wider area. Design and optimization of such a system calls for an analysis model which includes a strong two-way coupling between the plasma and divertor material. The incoming plasma heat and particle flux will affect the divertor surface temperature, which is a defining factor of the lithium evaporative and sputtered flux going into the plasma. Results of the coupled model based on the plasma edge code SOLPS-ITER and the computational fluid dynamics (CFD) code ANSYS-CFX will be presented for different configurations. An analytical slab flow model is used as a heat transfer boundary condition for SOLPS, defining particle flux from the wall via calculation of the surface temperature. At the final step, results of the SOLPS analysis are verified using a 3D CFD magnetohydrodynamics (MHD) analysis which uses heat and particle flux from SOLPS as a boundary condition. In addition to plasma heat flux, both analytical and CFD temperature models include several plasma material interaction effects, such as lithium evaporation, condensation and sputtering based on deuterium target flux. New adatom sputtering model based on the available experimental data is presented. Analytical model is expanded to include free surface axisymmetric configurations. Results of parametric studies of the divertor configurations with different lithium inlet temperature and velocity will be presented leading to the optimal design resulting in the lowest possible lithium contamination in the core.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

An Overview of LO-DuSST (Lunar Occupancy Dust Surface Separation Technologies) Objectives

Of the myriad of issues facing lunar exploration and maintaining an extended lunar presence, lunar dust is possibly the most pervasive. These jagged, chemically reactive, electrostatically charged, sometimes magnetic particles can impact every aspect of a lunar surface mission ranging from abrading extra-vehicular activity (EVA) suits to disrupting lunar vehicle thermal management systems to impeding efficacy of excavation equipment to impacting lunar inhabitant health. NASA’s Lunar Occupancy Dust Surface Separation Technologies (LO-DuSST) task, as a part of the broader Lunar Surface Innovation Initiative (LSII) project, seeks to implement synergistic active and passive lunar dust management and mitigation technologies for an array of applications. One such application is protection of power generation capabilities via solar panel arrays. Plasma dust lofting coupled with a piezoelectric-driven vibration technology will be demonstrated on solar panel surfaces contaminated with lunar dust simulant. Intrinsically low adhesion coatings will be applied to these surfaces to enhance dust removal. Electrostatic repulsion in confined geometries and materials to manage high velocity lunar dust wear will also be investigated. Collectively, these active and passive mitigation technologies will facilitate lunar surface operations including occupancy logistics, landing pad operations, power generation, and transportation. Initial results regarding lunar simulant-contaminated solar panel performance experiments, dust lofting using an electron beam, surface interactions of high velocity lunar dust arising from lunar lander plume-surface interactions, and lunar dust simulant-material interaction experiments will be discussed.

Lunar dust↗

Polarized fusion and potential in situ tests of fuel polarization survival in a tokamak plasma

Abstract The use of spin-polarized fusion fuels would provide a significant boost towards the ignition of a burning plasma. The cross section for D + T → α + n, would be increased by 1.5 if the fuels were injected with parallel polarization. Furthermore, our simulations demonstrate additional non-linear power gains in large-scale machines such as ITER, due to increased alpha heating. Such benefits require the survival of spin polarizations for periods comparable to the particle confinement time. During the 1980s, calculations predicted that polarizations could survive a plasma environment, although concerns persisted regarding the cumulative impacts of wall recycling. In that era, technical challenges prevented direct tests and left the large scale fueling of a power reactor beyond reach. Over the last decades, this situation has changed dramatically. Detailed simulations of ITER have predicted negligible wall recycling in a high-power reactor, and recent advances in laser-driven sources project the capability of producing large quantities of ∼100% polarized D and T. The remaining crucial step is an in-situ demonstration of polarization survival in a plasma. For this, we outline a measurement strategy using the isospin-mirror reaction, D + 3 He → α + p. Polarized 3 He avoids the complexities of handling tritium, while encompassing the same spin-physics. We evaluate two methods of delivering deuterium, using dynamically polarized Lithium-Deuteride (with vector polarization P V D of 70%) or frozen-spin Hydrogen-Deuteride (with P V D of 40%), together with a method of injecting optically-pumped 3 He (with 65% polarization). Pellets of these materials all have long polarization decay times (∼6 min for LiD at 2 K, ∼2 months for HD at 2 K, and ∼3 d for 3 He at 77 K), all far greater than a plasma shot in a research tokamak such as DIII-D (∼20 s). Both species can be propelled from a single cryogenic injection gun. We review plasma requirements and strategies for detecting polarization survival. Polarization alters both fusion yields and the angular distribution of fusion products, and each of these provides a potential signal. In this paper we simulate a selection of shots with similar characteristics in a future high-T ion H plasma, and find ratios of yields from shots with fuel spins parallel and antiparallel reaching 1.3 (HD + 3 He) to 1.6 (LiD + 3 He) over a wide range of poloidal angles. (A companion paper finds sensitivity to fusion product angular distributions as reflected in the pitch angles of protons and alphas reaching the plasma facing wall.)

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Experimental and computational studies on high-entropy carbide MoNbTaVWC 5 under high pressures

High-entropy carbide, MoNbTaVWC 5 , was synthesized from oxide precursors of the constituent metals, mixed with graphite powder in a microwave-generated hydrogen plasma at 26.66 kPa and 2100 °C. Ambient x-ray diffraction analysis confirms the full conversion of oxide precursors into a single-phase, face-centered cubic structure with a lattice parameter a = 4.3309 Å. Nanoindentation measured a hardness of 24.5 ± 1.3 GPa and an elastic modulus of 386 ± 22 GPa. The synthesized sample, mixed with a copper pressure marker, was studied by the radial x-ray diffraction technique with beryllium gasketing in a diamond anvil cell up to 70 GPa. The experimentally measured pressure–volume curve and shear strength were compared with theoretical predictions using the special quasi-random structure technique and density functional theory. MoNbTaVWC 5 achieved a 12% volume compression at 70 GPa and exhibited a high shear strength of 6.6 GPa. The present study demonstrates that the high-entropy carbide MoNbTaVWC 5 exhibits exceptional incompressibility and high strength under extreme conditions.

36 MATERIALS SCIENCE↗

Basic Energy Sciences Roundtable: Foundational Science to Accelerate Nuclear Energy Innovation

Energy security, availability, and reliability are among the greatest challenges facing the nation and the planet. An abundant potential source of energy resides in the fundamental atomic building blocks of the universe in the form of nuclear fission and fusion reactions. In fact, energy from nuclear fission currently provides the majority of the world’s zero-carbon electricity, and future fusion energy systems offer great promise; carbon-free nuclear energy technologies can be key to the world’s decarbonized energy future. Although contemporary fission systems use well-established technologies to supply safe and efficient baseload power, they could be more fuel efficient and less costly. Moving beyond massive light-water fission reactors to a variety of advanced nuclear systems—which will vary in size and operate in extremes of temperature, corrosivity, and other parameters—will place stringent conditions on materials and chemical systems. New demands will be placed on the coolants and solvents, the materials, and the monitoring tools used in these reactors. Fusion-based nuclear energy will require superior materials to withstand extremely high temperatures, plasma exposure, radiation damage, and implanted gases. The advantages associated with these new fission and fusion technologies will be realized only through continued advancements in the fundamental science underpinning our knowledge of the physics and chemistry of nuclear systems gained via improved experimental and computational methods. In July 2022, the U.S. Department of Energy’s Office of Basic Energy Sciences—in coordination with the Offices of Nuclear Energy, Fusion Energy Sciences, and Advanced Scientific Computing Research—held a virtual roundtable titled “Foundational Science to Accelerate Nuclear Energy Innovation” to discuss the scientific and technical barriers for advanced nuclear energy systems. Five priority research opportunities were identified to address these scientific and technical challenges and to accelerate progress toward the realization of next-generation fusion and fission energy systems. The foundational science gaps inhibiting the advancement of nuclear energy technologies are identified and tackled in five priority research opportunities. These opportunities pave the way to accelerate the development and ultimately the adoption of new nuclear energy systems. They include the fundamental aspects of ion-electron interactions; novel properties of next-generation coolants and solvents; interfacial dynamics, not only in solids, but in other aspects of nuclear reactors; novel operando and in situ monitoring and sensing; and artificial intelligence to accelerate condensed phases discovery. Building on the foundation established by previous Basic Energy Sciences workshops, these opportunities encompass recent advances in fundamental knowledge and focus on the experimental and computational methods needed to resolve major technical challenges for nuclear energy technologies. Through developing fundamental scientific insight as well as pushing the frontiers of modeling complex systems and probing the operation of materials and chemical systems in extreme environments, research motivated by the priorities identified here will further develop the promise, potential, and utilization of nuclear energy for a clean energy future.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Recent DIII-D progress toward validating models of tungsten erosion, re-deposition, and migration for application to next-step fusion devices

Abstract Fundamental mechanisms governing the erosion and prompt re-deposition of tungsten impurities in tokamak divertors are identified and analyzed to inform the lifetime of tungsten plasma-facing components in ITER and other future devices. Various experiments conducted at DIII-D to benchmark predictive models are presented, leveraging the DiMES removable sample exposure probe capability and the Metal Rings Campaign, in which toroidally symmetric rows of tungsten-coated tiles were installed in the DIII-D divertor. In tokamak divertors, the width of the electric sheath is of the order of the main ion Larmor radius, and a vast majority of sputtered tungsten impurities are typically ionized within the sheath. Therefore, W prompt redeposition is mainly governed by the ratio of the characteristic ionization mean-free path of neutral tungsten to the width of the sheath. In-situ monitoring of the prompt redeposition of tungsten impurities in divertors is demonstrated via the use of WII/WI line ratios and the ionizations/photon (S/XB) method in L-mode discharges. Even with this relatively limited set of emission measurements, net erosion measurements were found to be a consistent upper bound to an analytic scaling based on the ratio of the W ionization length, λ iz , and the width of the magnetic sheath rather than the ratio of λ iz and the W + gyro-radius. In the far-scrape-off layer (SOL) of the ITER divertor, however, it is calculated that the measurement of photon emissions associated with the ionization of tungsten impurities up to W 5 + may be required. Finally, W deposition patterns on DiMES collector probes, interpreted via DIVIMP-WallDYN modelling, reveal the key roles of progressive W erosion/re-deposition staps and E × B drifts in regulating long-range high-Z material migration.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Modeling, analysis, and code/data validation of DIII-D tokamak divertor experiments on ELM and non-ELM plasma tungsten sputtering erosion

Abstract We analyzed recent DIII-D tokamak tungsten divertor probe experiments using advanced, coupled, sputter erosion/redeposition, plasma, and surface response code packages. Modeling is done for ELMing H-mode, and L-mode plasmas, impinging on various size tungsten deposits on Divertor Material Evaluation System (DiMES) carbon probes. The simulations compute 3D, full kinetic, sub-gyromotion, impurity sputtering and transport, including changes in tungsten surface composition and response due to mixed deuterium and carbon ions irradiation. Per our analysis, ELM (edge localized mode) plasma sputtering in DIII-D mostly involves free-streaming high energy (∼500–1000 eV) D + and C +6 ions, with high near-surface plasma density. L-Mode sputtering is due to impurity sputtering (C, W) only, with lower density. All cases show complete redeposition of tungsten on the divertor, with significant redeposition on the tungsten spots themselves, and low self-sputtering. Comparison of ELM plasma gross tungsten erosion simulation results with in-situ spectroscopic data is good, as are code/data comparisons of net erosion using post-exposure Rutherford backscattering (RBS) data for the L-mode probes. The analysis, extrapolated to a full tungsten divertor, implies low net erosion and negligible plasma contamination from sputtering. These results support the use of high-Z plasma facing surfaces in ITER and beyond.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Environmental Exposure Conditions for Teflon FEP on the Hubble Space Telescope Investigated

The Hubble Space Telescope (HST) was launched into low Earth orbit on April 24,1990. During the first servicing mission in December 1993 (3.6 years after launch), multilayer insulation (MLI) blankets were retrieved from the two magnetic sensing systems located on the light shield. Retrieval of one of the solar arrays during this mission also provided MLI blanket material from the solar array drive arm. These MLI materials were analyzed in ground-based facilities, and results indicate that the space-facing outer layer of the MLI, aluminized Teflon FEP (DuPont; fluorinated ethylene propylene), was beginning to degrade. Close inspection of the FEP revealed through-the-thickness cracks in areas with the highest solar exposure and stress concentration. During the second servicing mission in February 1997 (6.8 years after launch), astronauts observed and documented severe cracking in the outer layer of the MLI blankets on both the solar-facing and anti-solar-facing surfaces. During this second mission, some material from the outer layer of the light shield MLI was retrieved and subsequently analyzed in ground-based facilities. After the second servicing mission, a Failure Review Board was convened by NASA Goddard Space Flight Center to address the MLI degradation problem on HST. Members of the Electro-Physics Branch of the NASA Glenn Research Center at Lewis Field participated on this board. To determine possible degradation mechanisms, board researchers needed to consider all environmental constituents to which the FEP MLI surfaces were exposed. On the basis of measurements, models, and predictions, environmental exposure conditions for FEP surfaces on HST were estimated for various time periods from launch in 1990 through 2010, the planned end-of-life for HST. The table summarizes these data including the number and temperature ranges of thermal cycles; equivalent Sun hours; fluence and absorbed radiation dose from solar event x rays; fluence and absorbed dose from solar wind protons and electrons trapped in Earth s magnetic field; fluence of plasma electrons and protons; and atomic oxygen fluence.

Dever, Joyce A.↗