Nuclear graphite is a critical material in high-temperature nuclear reactors due to its superior thermal and mechanical properties. The manufacturing process leaves multi-scale ‘pre-existing’ defects that can affect thermal transport characteristics. Because these defects are remnant of graphitization temperature, they cannot be thermally annealed. This study employs a non-thermal electron wind force (EWF) annealing technique to avoid this obstacle. 2 min of EWF treatment of the as-received graphite IG-110 at temperatures below 100 °C led up to 67

Achieving precise control over crack propagation in nuclear graphite and conducting quantitative analysis remain challenging. In this study, the through-thickness fracture behavior of pristine and neutron-irradiated (700 degrees C, similar to 7 dpa) NBG-17 nuclear graphite was investigated using split-disc testing coupled with micro-computed tomography (micro-CT). A notably lower number of micropores was observed in the neutron-irradiated specimen. The fracture toughness of neutron-irradiated NBG-17 was measured to be 1.45 MPa root m, compared to 1.17 +/- 0.05 MPa root m for the pristine specimen. In both materials, cracks were found to initiate at the filler-binder interface, and often correlated with microstructural features such as pores and thermal cracks. Crack bridging and deflections emerged as the primary toughening mechanisms in both unirradiated and irradiated NBG-17. However, compared with the pristine specimen, the cracks in the neutron-irradiated specimen were more likely to grow trans granularly, resulting in less deflected crack paths. The reduced micro-porosity and strengthened filler-binder boundaries were considered to be the cause of the observed differences in crack morphologies. This study provides a qualitative analysis of the fracture behavior of neutron-irradiated nuclear graphite in the absence of radiolytic oxidation.

Fiber-reinforced ceramic matrix composites are attractive for high-temperature nuclear applications due to excellent thermal and mechanical properties as well as reasonable-to-outstanding radiation resistance. Over the past 20 years, the use of ceramic matrix composite applications expanded to many commercial non-nuclear industries as fabrication and application of the technologies mature.The ASME Boiler Pressure Vessel Code, under Section III Division 5, provides the design and construction rules for High Temperature Reactor components. It published the first rules for ceramic matrix composites to be used for reactor core components. The rules lay out the quality requirements together with the design and materials criteria for the use and application of silicon carbide- and carbon-based matrix material technologies. As with the established graphite rules, the ceramic composite material rules are structured in Subsection HH (from Section III), that addresses the criteria for class SN nonmetallic core components. The code rules rely heavily on the development and publication of standards for composite specification, classification, and testing of mechanical, thermal, and other properties. These test methods are developed in ASTM Committee C28 on Advanced Ceramics, with a current focus on ceramic composite tubes. This article describes the detail of the composites code, the design methodology and similarities to the graphite code, the guidance for the development of specifications for ceramic composites (for nuclear applications) including recent standard developments, and it mentions the next steps to support licensing aspects by validating the code with benchmarking data

For oxidation data from diverse studies to be compared and applied to nuclear graphite degradation behavior, all testing must quantify the oxidation rate as a material-specific property under tightly controlled test conditions. Graphite oxidation behavior is affected by a wide range of parameter variations beyond graphite grade and oxidation temperature. ASTM D7542-15 attempts to control the graphite oxidation with specific furnace dimensions and careful test parameters designed to establish surplus oxygen conditions over a limited temperature range. Practical constraints for relating the observed oxidation of the standardized test sample size to smaller sample geometries must be considered along with the testing conditions to enforce an intrinsic, or material-specific, response. Similarly, these same test parameters must be examined when altering the oxidizing furnace (i.e., oxidizing within a thermogravimetric analyzer [TGA]) to ensure consistent intrinsic response rather than responding to the extrinsic test conditions. Here, data are presented from both vertical furnace and TGA experiments. Substantial differences in sample dimensions and geometric proportions require substantial verification of the test conditions. Because these variations can influence oxygen availability, the presence of excess oxygen must be confirmed to validate the range of conditions over which the data from disparate oxidizing systems can be compared. Statistical reproduction of testing over a suitable range of parameters appears to be more important than accommodating larger specimens, competing priorities that are particularly challenging for irradiated materials. The major considerations addressed are oxygen supply (partial pressure, gas flow, availability at the interface), specimen dimensions and purity, grade-specific microstructural effects (mass loss progression, oxidation penetration, and density gradient evolution), concurrent with temperature effects for tested conditions.

The acute oxidation behaviors for a number of nuclear graphite grades pertinent to gas-cooled high-temperature reactor (HTR) designs are compared using established American Society for Testing and Materials oxidation testing standards. In the unlikely event of an air-ingress accident, graphite components within the HTR core region are anticipated to oxidize if oxygen is allowed to enter the hot core region. Specimens from six disparate nuclear graphite grades were exposed to air/helium mixtures (100% air and 10% air-in-helium) within and beyond the kinetic oxidation temperature regime to ascertain the oxidation performance for each grade. All specimens were oxidized to a prescribed mass loss within a constant total gas flow of 10 l/min at multiple isothermal temperatures for calculation of oxidation rate and Arrhenius parameters for each graphite grade. A comparison of the oxidation performance among the different grades, as well as the effects resulting from high impurity levels, are discussed. (C) 2020 Elsevier B.V. All rights reserved.

A vein structure, which becomes more pronounced with increasing ion dose, was found on the surface of polycrystalline HOPG (highly oriented pyrolytic graphite) implanted by ex situ C+ (up to 1.8 x 10(17) ions/cm(2)), and in situ Ar+ in a transmission electron microscope (TEM). These veins are found to be independent of the crystallographic orientations and are associated with the formation of pores. Underneath the veins, a triangular-shaped core was formed with the graphite platelet inside the core displaced up towards the surface. A macro-scale 'ruck&tuck' geometry was thus generated at these triangle structure boundaries. Progressive movement of dislocations along basal planes during irradiation was observed, and a mechanistic model was proposed on this basis to explain the vein formation. A small increase of cspacing was observed with irradiation but it is believed that macro-scale vein formation plays a more vital role in the dimensional and property changes in polycrystalline graphite, especially when a stress gradient is present. The model proposed also explains the change of thermal expansion in HOPG with irradiation. Together with Heggie's 'ruck&tuck' and Barsoum's 'ripplocations' models, the present model is considered to have provided an additional experimentally proven mechanism responsible for irradiation behaviour in graphite materials. (C) 2020 Elsevier Ltd. All rights reserved.

The modular high temperature reactor is a leading candidate for near-term deployment of an advanced reactor concept. The US Department of Energy supports the development of high temperature reactors through qualification of fuels, materials, and analysis methods that exploit the capabilities of the national laboratories (mainly Idaho, Oak Ridge, and Argonne) to generate the fundamental fuel, material, and core behavior data needed to support design and licensing efforts by vendors. Industrial vendors of high temperature reactor concepts provide input to the R&D program and engage in the design of the systems, structures, and components that are unique to their designs. Although there is considerable variation in the plant designs being developed, they all exploit the properties and behavior of the TRISO fuel, material, and analysis methods being tested in the DOE qualification programs. This paper summarizes the goals and recent progress of the Department of Energy's Gas-Cooled Reactor research and development campaign and how it supports a wide variety of industrial efforts to deploy high temperature reactors to serve modern energy markets.

•The DOE Advanced Reactor Technologies (ART) Graphite R&D program was initiated in 2005 to provide unirradiated and irradiated material data of new nuclear graphite grades.•Three primary areas of research (irradiation testing, unirradiated testing, and material science analysis) combine to provide a comprehensive understanding of the available commercial graphite grades.•The Advanced Graphite Creep (AGC) Experiment will irradiate over 2000 graphite specimens to dose levels of 15 dpa and over irradiation temperatures of 600 °C and 800 °C.•The unirradiated as-manufactured material properties test program (Baseline program) utilizes full-sized ASTM test specimens, a large sample population, and established test standards.•Material analysis program focuses on analyzing the data from the AGC and Baseline programs, effects of degradation (oxidation, fracture, and irradiation damage), and the development of predictive behavior models.

Atomic level processes involved in the swelling and crack-closing in nuclear grade graphite under electron irradiation have been observed in real-time using transmission electron microscopy. Noise-filtered lattice images show the formation of vacancy loops, interstitial loops and resulting dislocations with unprecedented clarity. The dislocation dipoles formed via vacancy loops were found to undergo climb resulting in extra basal planes. Concurrent EELS studies showed a reduction in the atomic density because of the breakage of hexagonal carbon rings. The formation of new basal planes via dislocation climb in addition to the bending/breaking of basal planes leads to swelling and closing of micro-cracks.

Graphite will be used as a structural and moderator material in next-generation nuclear reactors. While the overall nature of the production of nuclear graphite is well understood, the historic nuclear grades of graphite are no longer available. This paper reports the virgin microstructural characteristics of filler particles and macro-scale porosity in virgin nuclear graphite grades of interest to the Next Generation Nuclear Plant program. Optical microscopy was used to characterize filler particle size and shape as well as the arrangement of shrinkage cracks. Computer aided image analysis was applied to optical images to quantitatively determine the variation of pore structure, area, eccentricity, and orientation within and between grades. The overall porosity ranged between ∼14% and 21%. A few large pores constitute the majority of the overall porosity. The distribution of pore area in all grades was roughly logarithmic in nature. The average pore was best fit by an ellipse with aspect ratio of ∼2. An estimated 0.6–0.9% of observed porosity was attributed to shrinkage cracks in the filler particles. Finally, a preferred orientation of the porosity was observed in all grades.

The U.S. Department of Energy is sponsoring the Generation IV Initiative in the United States for the purposes of developing future-generation nuclear energy systems. Six systems have been selected for Generation IV consideration: gas-cooled fast reactor, lead-cooled fast reactor, molten salt-cooled reactor, sodium-cooled fast reactor, supercritical water-cooled reactor, and very high temperature reactor. Critical to the development of Generation IV concepts is successful development and deployment of materials that operate successfully in the aggressive operating environments envisioned in the Generation IV concepts. This paper summarizes the Generation IV operating environments and describes materials challenges and potential solutions, including crosscutting solutions applicable to multiple Generation IV concepts.

A discussion of fabrication techniques and performance testing of solid oxide components for use in hydrogen steam electrolysis is presented. Novel plasma spray techniques are utilized to deposit the thin ceramic oxide electrode, electrolyte, and interconnect layers on a planar intermetallic bipolar plate. Optimal porosity is achieved within the electrode microstructure through mixed feed techniques that are a combination of dry powder feed and liquid solution injection. The perovskite anode coatings formed from liquid precursor feedstock require post-deposition annealing in an oxygen-rich atmosphere to form the desired perovskite structures. Electrical conductivity measurements were measured for all electrodes and interconnect materials as a function of temperature to 1000 °C.

The U.S Department of Energy (DOE) has selected the Very High Temperature Reactor (VHTR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production without greenhouse gas emissions. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed, thermal neutron spectrum reactor that will produce electricity and hydrogen in a state-of-the-art thermodynamically efficient manner. The NGNP will use very high burn-up, low-enriched uranium, TRISO-coated fuel and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Project is envisioned to demonstrate the following: (1) A full-scale prototype VHTR by about 2021; (2) High-temperature Brayton Cycle electric power production at full scale with a focus on economic performance; (3) Nuclear-assisted production of hydrogen (with about 10% of the heat) with a focus on economic performance; and (4) By test, the exceptional safety capabilities of the advanced gas-cooled reactors. Further, the NGNP program will: (1) Obtain a Nuclear Regulatory Commission (NRC) License to construct and operate the NGNP, this process will provide a basis for future performance based, risk-informed licensing; and (2) Support the development, testing, and prototyping of hydrogen infrastructures. The NGNP Materials Research and Development (R&D) Program is responsible for performing R&D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. The NGNP Materials R&D Program includes the following elements: (1) Developing a specific approach, program plan and other project management tools for managing the R&D program elements; (2) Developing a specific work package for the R&D activities to be performed during each government fiscal year; (3) Reporting the status and progress of the work based on committed deliverables and milestones; (4) Developing collaboration in areas of materials R&D of benefit to the NGNP with countries that are a part of the Generation IV International Forum; and (5) Ensuring that the R&D work performed in support of the materials program is in conformance with established Quality Assurance and procurement requirements. The objective of the NGNP Materials R&D Program is to provide the essential materials R&D needed to support the design and licensing of the reactor and balance of plant, excluding the hydrogen plant. The materials R&D program is being initiated prior to the design effort to ensure that materials R&D activities are initiated early enough to support the design process and support the Project Integrator. The thermal, environmental, and service life conditions of the NGNP will make selection and qualification of some high-temperature materials a significant challenge; thus, new materials and approaches may be required.