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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 667 records · Page 37
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Characterization of high thermal conductivity fuel surrogates before and after ion irradiation

High thermal conductivity nuclear fuels offer important potential advantages over traditional oxide-based fuels such as higher burnup, reduction in fission gas release, and better overall safety of the system. One proposed approach to high thermal conductivity fuels utilizes high thermal conductivity nonfissile additives with UO 2 fuel to lower the fuel operating temperature and thereby take advantage of the highly favorable radiation resistance of UO 2 at lower operating temperatures. However, differential swelling in the matrix and high conductivity additive phases during high dose irradiation could lead to internal cracking and poor performance. In the current study, ceria (CeO 2 ) and zirconia (ZrO 2 ) surrogate matrices were used to model UO2 behavior. Additives of 10 vol. % Al 2 O 3 or SiC in the form of short fibers or platelets were used for the high conductivity second phase. The nuclear fuel surrogates were sintered to achieve densities greater than 93% of the ideal values. Scanning electron microscopy (SEM) imaging and X-ray diffraction confirmed the uniform distribution of the second phase and that no intermetallic second phase was formed during sintering. The thermal conductivity of the sintered samples was measured from 50 °C to 900 °C and confirmed the desirable increase compared to pure CeO 2 /ZrO 2 pellets. Samples were irradiated with 20 MeV Ni 6+ ions at midrange doses ranging from 1 to 15 displacements per atom (dpa) and temperatures from 300 °C to 700 °C. Post irradiation characterization revealed a good stability of the samples at low to medium doses with matrix lattice parameter swelling of < 0.14 % but showed a significant microstructural deterioration and decrease of the mechanical properties at 15 dpa.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Hierarchically porous and single Zn atom-embedded carbon molecular sieves for H2 separations

Abstract Hierarchically porous materials containing sub-nm ultramicropores with molecular sieving abilities and microcavities with high gas diffusivity may realize energy-efficient membranes for gas separations. However, rationally designing and constructing such pores into large-area membranes enabling efficient H 2 separations remains challenging. Here, we report the synthesis and utilization of hybrid carbon molecular sieve membranes with well-controlled nano- and micro-pores and single zinc atoms and clusters well-dispersed inside the nanopores via the carbonization of supramolecular mixed matrix materials containing amorphous and crystalline zeolitic imidazolate frameworks. Carbonization temperature is used to fine-tune pore sizes, achieving ultrahigh selectivity for H 2 /CO 2 (130), H 2 /CH 4 (2900), H 2 /N 2 (880), and H 2 /C 2 H 6 (7900) with stability against water vapor and physical aging during a continuous 120-h test.

42 ENGINEERING↗
DOE OSTI2018

AGR-3/4 Experiment Preliminary Mass Balance

AGR-3/4 was an experiment primarily aimed at studying fission product transport in graphite and graphitic materials. To accomplish this, 80 designed-to-fail (DTF) particles coated only with a thin pyrocarbon layer were incorporated among the roughly 7500 tristructural isotropic-coated driver fuel particles in each capsule. It was anticipated that intact DTF particles would behave like TRISO particles with SiC layer failures (releasing cesium and other metallic fission products to some extent, but retaining fission gases), and failed DTF particles would behave like TRISO particles with failed TRISO coatings (releasing both cesium and other metallic fission products and fission gases). The DTF particles provided a known source of fission products to migrate out into the fuel compact matrix and out into the surrounding concentric rings of graphite and graphitic matrix material for study. Post-irradiation examinations have focused on measuring the total releases of fission products from the fuel compacts (mass balance) and the spatial distribution of fission products in the rings. The total mass balance is an important parameter for comparison to fission product transport simulations of the AGR-3/4 experiment, and the spatial distribution within carbon rings is being used to derive fission product diffusion coefficients. To determine the fission product mass balance, each of the 12 irradiation capsules was disassembled, and their component parts analyzed via gamma counting and destructive leach or burn-leach methods. In “standard” capsules (Capsules 1, 3-5, 7-8, 10, and 12), inner rings, outer rings, sink rings, spacers, foils, felts, and through tubes were all analyzed for gamma-emitting fission products. Beta-emitting Sr-90 was measured as well, but because of the analysis program on the inner and outer rings, the Sr-90 inventory of the inner and outer rings is not yet available. “Fuel body” Capsules 2, 6, 9, and 11 were retained intact for future heating tests; thus, the inner and outer rings from these capsules were not measured. The most commonly detected radionuclide fission products were Ag-110m, Cs-134, Cs-137, Eu-154, Eu-155, and Sr-90. Sb-125 was also frequently detected in the sink rings and spacers; however, the zircaloy spacers used in Capsules 1 through 6 contained natural antimony that transmuted to Sb-125 and made it impossible to separate Sb-125 released from the fuel from that generated in the spacers. Summing the fission product inventory measured on each capsule component made it clear that the presence of 80 DTF particles in each capsule resulted in noticeably higher releases of cesium, europium, and strontium from AGR-3/4 fuel when compared to AGR-1 fuel which did not have DTF particles. Greater than 30 particles worth of Cs-134 was measured outside of the fuel compacts in Capsules 3-5, 7, 8, and 10. In Capsule 11 (an intact fuel body for which the inner and outer rings were not measured), 26 particle equivalents of Cs-134 was measured outside of the outer ring. Based on experience from AGR 1, it is very unlikely that any driver particles failed during the irradiation; therefore, this cesium is assumed to be overwhelmingly dominated by release from DTF particles.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Inert gas enhanced laser-assisted purification of platinum electron-beam-induced deposits

Electron-beam-induced deposition patterns, with composition of PtC 5 , were purified using a pulsed laser-induced purification reaction to erode the amorphous carbon matrix and form pure platinum deposits. Enhanced mobility of residual H 2 O molecules via a localized injection of inert Ar–H 2 (4%) is attributed to be the reactive gas species for purification of the deposits. Surface purification of deposits was realized at laser exposure times as low as 0.1 s. The ex situ purification reaction in the deposit interior was shown to be rate-limited by reactive gas diffusion into the deposit, and deposit contraction associated with the purification process caused some loss of shape retention. To circumvent the intrinsic flaws of the ex situ anneal process, in situ deposition and purification techniques were explored that resemble a direct write atomic layer deposition (ALD) process. First, we explored a laser-assisted electron-beam-induced deposition (LAEBID) process augmented with reactive gas that resulted in a 75% carbon reduction compared to standard EBID. Lastly, a sequential deposition plus purification process was also developed and resulted in deposition of pure platinum deposits with high fidelity and shape retention.

43 PARTICLE ACCELERATORS↗
DOE OSTI2023

Robust Millisecond Coherence Times of Erbium Electron Spins

Erbium doped solids are prime candidates for quantum memories in optical quantum networks given their telecom-compatible photon emission. An electron spin of erbium with millisecond coherence time is desirable for generating remote entanglement between adjacent quantum network nodes. Here, in this study, we report GHz-range electron spin transitions of 167 Er 3+ in yttrium oxide (Y 2 O 3 ) matrix with coherence times that are consistently longer than a millisecond. By polarizing para magnetic impurity spins we achieve a spin T 2 up to 1.46 ms, and up to 7.1 ms after dynamical decoupling. These coherence lifetimes are among the longest found for erbium in crystalline matrices despite the presence of host nuclear spins. We further enhance the coherence time beyond conventional dynamical decoupling, using customized sequences to simultaneously mitigate spectral diffusion and dipolar interactions. Our study not only establishes 167 Er 3+ : Y 2 O 3 as a significant quantum memory platform but also provides a guideline for engineering long-lived erbium spins in a variety of host materials for quantum technologies.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2020

Non-Destructive Imaging of a Liquid Moving Through Porous Media Using a Computer Tomography Scanner

The most common approach used by modelers to describe movement of liquids through a porous solid, such as cement, sediment, or glass, is to assume a uniform flow rate, such as a Darcy Flux or a diffusion constant. However, this convenient simplification is problematic because in many cases it ignors the presence of fractures and macropores, which commonly dominate water flow and contaminant transport. For this reason, it is common that such modeling results do not reflect the multi-modal flow detected in laboratory and field studies. The objective of this seedling study was to develop a new capability for SRNL to track liquid moving through micro- (matrix-) and macro-flow using an X-ray Computed Topography (CT) Scanner. 4-dimensional anamated renditions were created that permited quantifying traditional matrix flow and macropore flow. These animations were modelled using a public domain software, HYDRUS 1-D, describing one dimension, dual-porosity and dual-permeability processes. This new capability provides a proof of concept for reducing model uncertainty applicable to waste disposal risk calculations, environmental remediation, and waste form development when describing the movement of liquids as they pass through glass, cement, fractured rock, or soil.

47 OTHER INSTRUMENTATION↗
DOE OSTI2019

VeloxChem: A Python-driven density-functional theory program for spectroscopy simulations in high-performance computing environments

An open-source program named VeloxChem has been developed for the calculation of electronic real and complex linear response functions at the levels of Hartree–Fock and Kohn–Sham density functional theories. With an object-oriented program structure written in a Python/C++ layered fashion, VeloxChem enables time-efficient prototyping of novel scientific approaches without sacrificing computational efficiency, so that molecular systems involving up to and beyond 500 second-row atoms (or some 10,000 contracted and in part diffuse Gaussian basis functions) can be routinely addressed. In addition, VeloxChem is equipped with a polarizable embedding scheme for the treatment of the classical electrostatic interactions with an environment that in turn is modeled by atomic site charges and polarizabilities. The underlying hybrid message passing interface (MPI)/open multiprocessing (OpenMP) parallelization scheme makes VeloxChem suitable for execution in high-performance computing cluster environments, showing even slightly beyond linear scaling for the Fock matrix construction with use of up to 16,384 central processing unit (CPU) cores. An efficient—with respect to convergence rate and overall computational cost—multifrequency/gradient complex linear response equation solver enables calculations not only of conventional spectra, such as visible/ultraviolet/X-ray electronic absorption and circular dichroism spectra, but also time-resolved linear response signals as due to ultra-short weak laser pulses.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2022

How to Get the Best Gas Separation Membranes from State-of-the-Art Glassy Polymers

Herein we focus on fundamental polymer science factors in membrane-based gas separation to open another chapter on this important topic. Realistically thinking about nano and angstrom scale matrix attributes that are translatable into higher performance thin glassy polymer asymmetric membrane layers in practical large-scale devices is crucial. Such thinking should be guided by expertise in both polymer and membrane communities to advance the state of the art. Polymer structures, membrane morphologies and effects of operating conditions are discussed here, using specific examples to illustrate key issues. Rubbery polymers are not the focus of this discussion, since they lack diffusive discrimination for size-similar penetrants, which glassy polymers provide and make them the focus here. Four scalable subtopics guide necessary thinking: (i) plasticization; (ii) anti-plasticization†2; (iii) dual-mode transport involving saturation of unrelaxed free volume, and (iv) non-uniform free volume and stress profiles in thin skin glassy polymer asymmetric or composite membranes. Furthermore, using these subtopics in the context of macromolecular science we discuss issues needed to expand the state-of-the-art in gas separation membranes.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2015

Ballistic-diffusive approximation for the thermal dynamics of metallic nanoparticles in nanocomposite materials

Based on ballistic-diffusive approximation, a method is presented to model heat transfer in nanocomposites containing metal nanoparticles. This method provides analytical expression for the temperature dynamics of metallic nanoparticles embedded in a dielectric medium. In this study, nanoparticles are considered as spherical shells, so that Boltzmann equation is solved using ballistic-diffusive approximation to calculate the electron and lattice thermal dynamics in gold nanoparticles, while thermal exchange between the particles is taken into account. The model was used to investigate the influence of particle size and metal concentration of the medium on the electron and lattice thermal dynamics. It is shown that these two parameters are crucial in determining the nanocomposite thermal behavior. Our results showed that the heat transfer rate from nanoparticles to the matrix decreases as the nanoparticle size increases. On the other hand, increasing the metal concentration of the medium can also decrease the heat transfer rate.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2020

Mechanical and Thermophysical Properties of 3D-Printed SiC-FY20

In the Transformational Challenge Reactor (TCR), the fuel blocks consist of an additively-manufactured silicon carbide (SiC) matrix and uranium nitride tristructural isotropic (UN TRISO) fuel particles, which are stacked to form fuel columns. The SiC matrix is manufactured using binderjet 3D printing followed by loading the TRISO fuel particles and the chemical vapor infiltration (CVI) process. Because the fuel matrix is a primary component of the TCR core and its response to mechanical and thermal loads during operation is one of the most influential factors on the integrity of TCR core, testing and evaluation have focused on producing mechanical and thermophysical properties data for the binderjet/CVI SiC. Mechanical and thermophysical properties were measured from various types of specimens printed for two or three orientations, which included equibiaxial flexural failure strength, elastic constants, thermal diffusivity and conductivity, density, and the coefficient of thermal expansion. Flexural failure strength datasets showed similar Weibull distributions regardless of sample variants including different orientations. The mean failure strengths of the 3D-printed SiC variants were in the range of 286–306 MPa, which are 22–27% lower than that of the CVD SiC. Thermophysical test results showed that specific heat and thermal expansion are not sensitive to the build directions of SiC samples, while thermal diffusivity is highly dependent on the build direction and can be correlated to the anisotropic character of the 3D-printed SiC. This report also includes discussions on the uniaxial tensile properties of the as-printed SiC before CVI and on ongoing efforts for irradiation effects studies.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Mechanical and Thermophysical Properties of 3D-Printed SiC before and after Neutron Irradiation – FY21

This report presents the mechanical and thermophysical properties of 3D-printed SiC before and after neutron irradiation that have been evaluated to assess the fuel matrix material for the Transformational Challenge Reactor (TCR). The TCR fuel form consists of an additively-manufactured silicon carbide (SiC) matrix and uranium nitride tristructural isotropic (UN TRISO) fuel particles, which is manufactured through a newly developed processing route combining binderjet 3D printing, TRISO fuel particle loading, and chemical vapor infiltration (CVI). Because the fuel matrix is a primary component of the TCR core and its response to mechanical and thermal loads during operation is one of the most influential factors on the integrity of TCR core, testing and evaluation have focused on producing mechanical and thermophysical properties data for the binderjet/CVI SiC. Baseline mechanical and thermophysical properties were measured from the disk specimens printed for different and sizes orientations, which included equibiaxial flexural failure strength, elastic constants, thermal diffusivity and conductivity, density, and the coefficient of thermal expansion. Flexural failure strength datasets showed similar Weibull distributions regardless of sample variants including different orientations. The mean failure strengths of the 3D-printed SiC variants were in the range of 280–310 MPa, which are slightly lower than that of the chemical vapor deposition (CVD) SiC. Thermophysical test results showed that specific heat and thermal expansion are not sensitive to the build direction of SiC samples, while thermal conductivity is highly dependent on the build direction and can be correlated to the anisotropic character of the 3D-printed SiC. Neutron irradiation tests were carried out on the 3D-printed 6-mm diameter SiC disk specimens. Irradiation was carried to 2.3 dpa over a temperature range of 360–880°C. No significant degradation in strength was observed in SiC after irradiations in various conditions and with different orientations. Anisotropy that had been observed in the thermal conductivity of 3D-printed SiC prior to irradiation vanished after irradiation as the irradiation defect thermal resistivity accumulated in the material. Electron microscopy of the microstructure after neutron irradiation showed distinct defect morphologies in the heterogenous material, but no evidence for irradiation-induced cracking or degradation in the microstructure was observed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Effect of Interaction Layer and U-Mo Size on Pore Growth in U-Mo/Al Fuel

U-Mo/Al dispersion fuel, composed of U-Mo alloy fuel particles embedded in the aluminum (Al) matrix, has been developed to convert high-enriched uranium fuel to low-enriched uranium for high-performance research reactors. The fueled zone, referred to as 'fuel meat,' is clad with aluminum alloy. Fuel meat and cladding are metallurgically bonded during fabrication. Irradiated U-Mo/Al dispersion fuel has shown stable performance in low fission rate and burnup ranges. However, excessive fuel meat swelling or breakaway swelling, which is believed an incipient failure, was observed at higher fission rate and burnup conditions. This breakaway swelling is believed to be a result from the combination of accelerated fuel swelling and interconnection of large pores formed outside of the UMo particles. These large pores were frequently found, particularly, at the interface between the Al and the fuel-matrix interaction layer (IL) that formed by irradiation-enhanced interdiffusion of U-Mo and Al. Kim et al. developed a model for pore growth using a mechanical equilibrium between gas pressure in the pore, interfacial energies, and the external stress at the pore surface. In this model, IL provides a path way for fission gas to reach the pores and also generates tensile stress fields. Hence, suppressing IL growth by applying a diffusion barrier, like a coating of the U-Mo particles, appears to be an effective method to prevent pore formation. Another remedy to prevent pore formation is to use large U-Mo fuel particles. Using larger U-Mo particles was thought advantageous because it reduces fission gas release. In this study, the effect of IL growth and U-Mo particle size on pore formation is examined using the model. The advantage was quantified in terms of porosity.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Microstructural Characteristics of HIP-bonded Monolithic Nuclear Fuels with a Diffusion Barrier

Due to the limitation of maximum uranium load achievable by dispersion fuel type, the Global Threat Reduction Initiative (GTRI) is developing an advanced monolithic fuel to convert US high performance research reactors to low-enriched uranium. Hot-isostatic-press bonding was the single process down-selected to bond monolithic U-Mo fuel meat to aluminum alloy cladding. A diffusion barrier was applied to the U–Mo fuel meat by roll-bonding process to prevent extensive interaction between fuel meat and aluminum-alloy cladding. Microstructural characterization was performed on fresh fuel plates fabricated at Idaho National Laboratory. Interfaces between fuel meat, cladding, and diffusion barrier, as well as U–10Mo fuel meat and Al–6061 cladding were characterized by scanning electron microscopy. Preliminary results indicate that the interfaces contain many different phases while decomposition, second phases, and chemical banding were also observed in the fuel meat. The important attributes of the HIP-bonded monolithic fuel are • A typical Zr diffusion barrier of thickness 25 µm • Transverse cross section that exhibits relatively equiaxed grains with an average grain diameter of 10 µm • Chemical banding, in some areas more than 100 µm in length, that is very pronounced in longitudinal (i.e., rolling) direction with Mo concentration varying from 7–13 wt% • Decomposed areas containing plate-shaped low-Mo phase • A typical Zr/cladding interaction layer of thickness 1-2 µm • A visible UZr2 bearing layer of thickness 1-2 µm • Mo-rich precipitates (mainly Mo2Zr, forming a layer in some areas) followed by a Mo-depleted sub-layer between the visible UZr2-bearing layer and the U–Mo matrix • No excessive interaction between cladding and the uncoated fuel edge • Cladding-to-cladding bonding that exhibits no cracks or porosity with second phases high in Mg, Si, and O decorating the bond line. • Some of these attributes might be critical to the irradiation performance of monolithic U-10Mo nuclear fuel. There are several issues or concerns that warrant more detailed study, such as precipitation along cladding-to-cladding bond line, chemical banding, uncovered fuel-zone edge, and interaction layer between U–Mo fuel meat and zirconium. Future post-irradiation examination results will focus, among other things, on identifying in-reactor failure mechanisms and, eventually, directing further fresh fuel characterization efforts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Evaluation of AGR-3/4 In-pile Silver Release Predictions Against Post-Irradiation Examination Measurements

Fuel performance modeling codes that accurately predict the transport of radionuclides in high-temperature gas-cooled reactors that utilize tristructural isotopic (TRISO) fuel particles are an important aspect of reactor safety analyses. One objective of the Advanced Gas Reactor (AGR)-3/4 experiment was to assess the transport of fission products through fuel particles and their subsequent release into the compact matrix and structural graphite materials. This was accomplished by irradiating uranium oxycarbide (UCO) driver fuel particles and designed-to-fail (DTF) particles to serve as known sources of fission products. The fission product of particular interest when it comes to such transport is silver (Ag-110 m), as it has a 250-day half-life and has relatively high mobility in the TRISO coating layers. Furthermore, to assess the current modeling capabilities and diffusion parameters employed in the fuel performance codes PARFUME and BISON, the fractional release of silver release predicted by the two codes were compared against post-irradiation examination measurements from the AGR-3/4 experiment.

AGR-3/4 Experiment↗
DOE OSTI2025

Time-resolved tracking of cellulose biosynthesis and assembly during cell wall regeneration in live Arabidopsis protoplasts

Cellulose, the most abundant polysaccharide on earth composing plant cell walls, is synthesized by coordinated action of multiple enzymes in cellulose synthase complexes embedded within the plasma membrane. Multiple chains of cellulose fibrils form intertwined extracellular matrix networks. It remains largely unknown how newly synthesized cellulose is assembled into an intricate fibril network on cell surfaces. Here, we have established an in vivo time-resolved imaging platform to continuously visualize cellulose biosynthesis and fibril network assembly onArabidopsis thalianaprotoplast surfaces as the primary cell wall regenerates. Our observations provide the basis for a model of cellulose fibril network development in protoplasts driven by an interplay of multiscale dynamics that includes rapid diffusion and coalescence of nascent cellulose fibrils, processive elongation of single fibrils, and cellulose fibrillar network rearrangement during maturation. This study provides fresh insights into the dynamic and mechanistic aspects of cell wall synthesis at the single-cell level.

Science & Technology - Other Topics↗
DOE OSTI2023

Low-Order Preconditioning for the High-Order Finite Element de Rham Complex

Here, we present a unified framework for constructing spectrally equivalent low-order-refined discretizations for the high-order finite element de Rham complex. This theory covers diffusion problems in H 1 , H(curl), and H(div) and is based on combining a low-order discretization posed on a refined mesh with a high-order basis for Nédélec and Raviart–Thomas elements that makes use of the concept of polynomial histopolation (polynomial fitting using prescribed mean values over certain regions). This spectral equivalence, coupled with algebraic multigrid methods constructed using the low-order discretization, results in highly scalable matrix-free preconditioners for high-order finite element problems in the full de Rham complex. Additionally, a new lowest-order (piecewise constant) preconditioner is developed for high-order interior penalty discontinuous Galerkin (DG) discretizations, for which spectral equivalence results and convergence proofs for algebraic multigrid methods are provided. In all cases, the spectral equivalence results are independent of polynomial degree and mesh size; for DG methods, they are also independent of the penalty parameter. These new solvers are flexible and easy to use; any “black-box” preconditioner for low-order problems can be used to create an effective and efficient preconditioner for the corresponding high-order problem. A number of numerical experiments are presented, based on an implementation in the finite element library MFEM. A range of challenging three-dimensional problems are used to corroborate the theoretical properties and demonstrate the flexibility and scalability of the method.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2018

Development of a mechanistic thermal aging model for Grade 91

This report provides a summary on understanding and predicting the effects of long-term thermal aging on microstructure and tensile properties of G91 to corroborate the ASME Code rules in strength reduction due to elevated temperature service. The research is to support the design and long-term operation of G91 structural components in sodium-cooled fast reactors (SFRs). The report is a Level 2 deliverable in FY18 (M2NT-18AN050502061), under the Work Package NT-18AN05050206, “G91 Code Extension Testing” performed by the Argonne National Laboratory (ANL), as part of the Advanced Materials Development Program for Fast Reactors. The thermal aging study of G91 involves three types of thermally-aged specimens: (1) specimens machined from two heats of G91, G91-H1 and H30176 that have been tested for the Advanced Materials Development Programs in the past years. These specimens were aged at 550, 600, and 650°C for times up to ~64,000 h; (2) specimens fabricated from the heads of the archived crept specimens (13 in total) of six heats of G91, H5349, H30394, H30383, H10148, H30176. These specimens were aged at 427 - 538°C and for times up to 132,647 h; (3) specimens made from a tube removed from the Kingston coal-fired power plant after exposure for 155,000 h at 550°C. These three sets of specimens provide a comprehensive data set for understanding and predicting the effect of long-term thermal aging on microstructure and tensile properties of G91. Microstructural analysis of the thermally-aged G91 revealed that thermal aging causes significant microstructural changes: (1) the tempered martensite formed during tempering is unstable, and suffers significant recovery during aging, manifested by an increase in subgrain width, a decrease in subgrain length, and a decrease in subgrain aspect ratio (i.e. subgrains become more equiaxed) as the aging temperature and time increase. The subgrain recovery is facilitated by the reduction of dislocation density within subgrains; (2) M23C6 carbides in the as-received G91 are distributed along grain and subgrain boundaries and play an important role as pinning obstacles against subgrain coarsening. M23C6 carbides coarsen during aging, and the coarsening of M23C6 carbides reduces the pinning force on the boundaries; (3) MX carbonitrides in the as-received G91 are distributed uniformly within subgrains, providing the precipitation hardening effect. They coarsen with increasing aging temperature and time, but the coarsening rate is low; (3) a new phase, the Laves phase intermetallic forms during aging. The formation of the Laves phase removes the Mo solutes from the matrix, reducing the solid solution strengthening effect. The Laves phase particles grow rapidly once they are nucleated, and they precipitate primarily along boundaries. They may have a pinning effect on boundaries, but should provide little to the alloy strength. The evolution of each microstructural constituent during thermal aging can be well described by individual microstructural models. The estimated activation energy for MX coarsening was 175 kJ/mol, equivalent to the activation energy for boundary and core diffusion in a-Fe, while the estimated activation energy for subgrain coarsening was significantly lower, ~96 kJ/mol. Thermal aging of G91 results in the reduction in the yield stress and the ultimate tensile strength but has an insignificant effect on the uniform and total elongations. The reduction in tensile strength is more pronounced as the aging temperature and time increases. The in situ tensile tests with high-energy X-rays revealed that sub-boundary strengthening plays a dominant role, and the strengthening role of M23C6 particles may be accounted for by subgrian boundary hardening, and three strengthening mechanisms can be superimposed to describe the tensile strength of G91: subgrain boundary hardening, MX precipitation hardening, and Mo solid solution strengthening. A microstructure – strength model was establ

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗