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Diffusion welding of heat-resistant alloys has attracted interest in the manufacturing of components with complex configurations. Controlling secondary precipitates along the interface is necessary to enhance the quality of the diffusion welding. Surface treatment to increase the solubility product (Ksp) of Ti-rich carbide is proposed to accomplish such an enhancement. The reduction of secondary precipitates along the interface induced grain boundary migration across the interface. The chemical compositions at/near the interface satisfied the material specifications. The mechanical properties of the diffusion weldment were similar to those of Alloy 800H that underwent the same thermo-mechanical processes in the range of 25–700 °C. At 25 °C, the tensile strength was 553 MPa, which satisfied the minimum specified tensile strength described in ASTM B: 409-22. The location of failure was random in the gauge section, and dimples, the evidence of the macroscopic plastic deformation, were observed on the fracture surface. [ABSTRACT FROM AUTHOR]

The effect of strain range and temperature on the low-cycle fatigue behaviour and microstructure change during cyclic deformation of Alloy 617 for use in very high temperature gas-cooled reactor components were studied at elevated temperature starting from ambient condition. Increasing the strain range and the temperature was noticed to reduce the fatigue resistance of nickel-based Alloy 617 due to facilitating the transformation behavior of the carbides in the grain interior, precipitates along the grain boundary, and oxidation behavior inducing surface connected precipitates cracking. Initial hardening behavior was observed at room temperature condition during cyclic due to the pile-up dislocation of micro-precipitates. The grain size was also taking a role due to the formation of an obstacle in the matrix. In the high temperature regime, the alloy 617 was found to soften for its entire life due to the fast recovery deformation, proved by its higher plasticity compared with lower temperature. The deformation behavior also showing high environmentally assisted damage. Oxidation behavior was found to become the primary crack initiation, resulting in early intergranular surface cracking.

The Korea Atomic Energy Research Institute (KAERI) has developed the Core Reliable Optimization and thermofluid Network Analysis (CORONA) code for core thermofluid analysis of a prismatic high-temperature gas-cooled reactor (HTGR). KAERI performed scaled-down standard fuel block (SFB) heated tests at a helium experimental loop to validate the CORONA code. The scaled-down SFB was designed based on the core thermofluid design for a 350-MW(thermal) HTGR. The reference test condition was selected to maintain the Reynolds number of the coolant channels and the bypass gaps. The test section had seven coolant holes and 12 fuel holes considering KAERI's helium loop circulator design. The material of the fuel block was Al2O3, selected to simulate the low thermal conductivity of the irradiated graphite at the high-temperature condition. The bypass gap structure was made of stainless steel 304 to minimize gap size deformation at the heated condition. This paper presents a comparison between the test results and the CORONA analysis results. The test parameter was the nitrogen flow velocity (3.6 to 6.0 kg/min) and constant heated condition. [ABSTRACT FROM AUTHOR]

The potential of reducing the oxidation of the supporting graphite components during normal and/or accident conditions in the Very High Temperature Reactor (VHTR) design has been studied. In this work efforts have been made to slow the oxidation process of the graphite with a thin SiC coating (∼ 10 μm). Upon heating at ≥ 1173 K in air, the spallations and cracks were formed in the dense columnar structured SiC coating layer grown on the graphite with a functionally gradient electron beam physical vapor deposition (EB-PVD. In accordance with the formations of these defects, the sample was vigorously oxidized, leaving only the SiC coating layer. Then, efforts were made to heal the surface defects using additional EB-PVD with ion beam bombardment and chemical vapor deposition (CVD). The EB-PVD did not effectively heal the cracks. But, the CVD was more appropriate for crack healing, likely due to its excellent crack line filling capability with a high density and high aspect ratio. It took ∼ 34 min for the 20% weight loss of the CVD crack healed sample in the oxidation test with annealing at 1173 K, while it took ∼ 8 min for the EB-PVD coated sample, which means it took ∼4 times longer at 1173 K for the same weight reduction in this experimental set-up. [ABSTRACT FROM AUTHOR]

Abstract: Atomistic structure change in a nuclear grade graphite irradiated at 353 K to 3.4×1017 ion/cm2 with 3 MeV H+ was characterized by measuring positron lifetime and Raman spectrum at room temperature. It is evident from the positron lifetime results that the pre-existing structural defect is disoriented crystalline boundaries, and vacancy clusters ranging from di- to quadruple-vacancies were newly formed after ion irradiation. The relative intensity ratio of the Raman D and G peaks increased from 0.25 to 0.67 after ion irradiation. The concentration of radiation-induced vacancies was reasonably estimated by the Raman intensity ratio. [ABSTRACT FROM AUTHOR]

Plate-type heat exchangers are anticipated to be used in the next-generation nuclear industry, and solid-state diffusion welding is a critical technology for building plate-type heat exchangers with high integrity. In this study, we manufactured a diffusion weldment and evaluated its creep behavior. Microscopic analysis revealed that Al-rich oxides were developed along the interface, significantly impeding grain-boundary movement across the interface. Oxide-containing planar grain boundaries resulted in premature brittle fracture at the interface with less than 9% creep strain under all test conditions. The time to rupture and time to 1% creep strain of the diffusion weldment were less than those of the as-received alloy, while the slopes in double-logarithmic plots were almost identical for both alloys. In a Larson–Miller parameter study, the stress to rupture of the diffusion weldment reached 95.59% of that of the as-received alloy, whereas the stress to 1% creep strain steeply decreased in the low-stress range. [ABSTRACT FROM AUTHOR]

Alloy 800H is currently under consideration to be used in a very high temperature reactor system. This paper describes the experimental low-cycle fatigue study on Alloy 800H base metal and weldments in the air with fully-reversed strain control at 700 °C. The recommended modern Alloy 800H weldments were fabricated by gas tungsten arc welding. The observed fatigue life is found to be comparable between base metal and weldments at this temperature, indicating the welding technique can compete with base material. The influence of hold time was further investigated at the maximum tensile strain of 0.6% total strain range condition for 60 s, and it is found to reduce the fatigue life. For most of the tests, the base metal revealed an initial cyclic hardening stress response, otherwise, the weldment was entirely softened with cycles. The addition of hold time is also found to induce an unequal cyclic stress response in both materials. Accordingly, for continuous low-cycle fatigue tests, fatigue cracks were initiated and propagated in a classical transgranular fracture mode. However, the hold time affected the mixed-mode fracture mode by means of a transition from transgranular to intergranular fracture. [ABSTRACT FROM AUTHOR]

• We obtained various optimal thicknesses of the coating layers of an 800-μm UO 2 TRISO. • Failure fraction and fission product release are evaluated for the optimal TRISO. • The optimal thicknesses of buffer, IPyC, SiC, OPyC are 110/40/57/35, 120/35/60/27 μm. • They are suitable for a 10-MWth block-type micro modular HTGR within 7600 days. The study evaluates the optimal thicknesses of the coating layers of an 800-μm UO 2 tri-structural isotropic coated fuel particle (TRISO) for a 10-MW th block-type micro modular high temperature gas-cooled reactor (HTGR) using a response surface methodology (RSM) and an HTGR fuel performance analysis code, COPA. The objective function to be minimized in the optimal design of a TRISO consists of the product of the packing fraction of TRISO particles in a compact, the fractional releases of major metallic fission products from a TRISO and the failure probability of the silicon carbide (SiC) layers. A constraint on the sum of the coating layer thicknesses was made using a maximum packing fraction. Two fuel performance analyses such as a failure fraction analysis and a fission product release analysis are applied to the optimal designs of a TRISO that the RSM produced. An 800-μm UO 2 TRISO having buffer, inner high-density pyrocarbon (IPyC), SiC, outer high-density pyrocarbon (OPyC), each with a thickness of 110, 40, 57, 35 μm, or 120, 35, 60, 27 μm is suitable for the operation of a 10-MW th block-type micro modular HTGR within 7600 days when both the SiC and through-coatings failure fractions do not occur and the fractional release of silver, cesium, strontium and krypton are below 1 × 10−9. The packing fraction of the optimal TRISOs in a compact is 40%. [ABSTRACT FROM AUTHOR]

To obtain high tritium breeding performance with limited blanket thickness, most of solid breeder blanket concepts employ a combination of lithium ceramic as a breeder and beryllium as a multiplier. In this case, considering that huge amount of beryllium are needed in fusion power plants, its handling difficulty and cost can be a major factor to be accounted for commercial use. Korea has proposed a Helium-Cooled Ceramic Reflector (HCCR) breeding blanket concept relevant to fusion power plants. Here, graphite is used as a reflector material by reducing the amount of beryllium multiplier. Its feasibility has been investigated in view point of the nuclear performance as well as material-related issues. In this paper, a nuclear analysis is performed under the fusion reactor condition to address the feasibility of graphite reflector in breeding blanket, considering tritium breeding capability and neutron shielding and activation aspects. Also, the chemical stability of the graphite is investigated considering the chemical stability under accident conditions, resulting in that the adaptation of graphite reflector in breeding blanket is intrinsically safe and plausible under fusion reactor condition. [ABSTRACT FROM AUTHOR]