• Proceedings of the Korean Nuclear Society Conference
• /
• 한국원자력학회 2004년도 추계학술발표회 발표논문집
• /
• pp.1307-1308
• /
• 2004

• Proceedings of the Korean Nuclear Society Conference
• /
• 한국원자력학회 2004년도 추계학술발표회 발표논문집
• /
• pp.1301-1302
• /
• 2004

• Nuclear Engineering and Technology
• /
• 제41권1호
• /
• pp.1-10
• /
• 2009

The safe and reliable performance of fusion and fission plants depends on the choice of suitable materials and an assessment of long-term materials degradation. These materials are degraded by their exposure to extreme conditions; it is necessary, therefore, to address the issue of long-term damage evolution of materials under service exposure in advanced plants. The empirical approach to the study of structural materials and fuels is reaching its limit when used to define and extrapolate new materials, new environments, or new operating conditions due to a lack of knowledge of the basic principles and mechanisms present. Materials designed for future Gen IV systems require significant innovation for the new environments that the materials will be exposed to. Thus, it is a challenge to understand the materials more precisely and to go far beyond the current empirical design methodology. Breakthrough technology is being achieved with the incorporation in design codes of a fundamental understanding of the properties of materials. This paper discusses the multi-scale, multi-code computations and multi-dimensional modelling undertaken to understand the mechanical properties of these materials. Such an approach is envisaged to probe beyond currently possible approaches to become a predictive tool in estimating the mechanical properties and lifetimes of materials.

• Nuclear Engineering and Technology
• /
• 제45권3호
• /
• pp.311-322
• /
• 2013

High-Cr martensitic steel HT-9 is one of the candidate materials for advanced nuclear energy systems. Thanks to its excellent thermal conductivity and irradiation resistance, ferritic/martensitic steels such as HT-9 are considered for in-core applications of advanced nuclear reactors. The harsh neutron irradiation environments at the reactor core region pose a unique challenge for structural and cladding materials. Microstructural and microchemical changes resulting from displacement damage are anticipated for structural materials after prolonged neutron exposure. Consequently, various irradiation effects on the service performance of in-core materials need to be understood. In this work, the fundamentals of radiation damage and irradiation effects of the HT-9 martensitic steel are reviewed. The objective of this paper is to provide a background introduction of displacement damage, microstructural evolution, and subsequent effects on mechanical properties of the HT-9 martensitic steel under neutron irradiations. Mechanical test results of the irradiated HT-9 steel obtained from previous fast reactor and fusion programs are summarized along with the information of irradiated microstructure. This review can serve as a starting point for additional investigations on the in-core applications of ferritic/martensitic steels in advanced nuclear reactors.

• Nuclear Engineering and Technology
• /
• 제56권5호
• /
• pp.1887-1894
• /
• 2024

Fatigue strength of the structural materials of lead-cooled fast reactors (LFRs) and accelerator-driven systems (ADS) may be degraded in liquid metal (Lead or lead-bismuth eutectic (LBE)) environments. The fatigue crack growth (FCG) data of structural materials in liquid LBE are necessary for damage tolerance design, safety assessment and life management of key equipment. A novel monitoring system for fatigue crack length was designed on the compliance method and the monitor technology of crack opening displacement (COD) of CT specimens by the linear variable differential transformers (LVDT) system. It can be used to predict the crack length by monitoring the COD of CT specimens in harsh high-temperature liquid LBE using a LVDT system. The prediction accuracy of this system was verified by FCG experiments in room temperature air and liquid LBE at 150, 250 and 350 ℃. The first results obtained in the FCG test for T91 steel in liquid LBE at 350 ℃ are presented.

• Nuclear Engineering and Technology
• /
• 제56권6호
• /
• pp.2282-2291
• /
• 2024

In this work we study the effects of different factors of dislocation loop on its obstacle strength when interacting with an edge dislocation. At first, the interaction model for dislocation and dislocation loop is established and the full and partial absorption mechanism is obtained. Then, the effect of temperature, size and burgers vector of dislocation loop are investigated. The relation between the obstacle strength and irradiation dose has been established, which bridges the irradiation source and microscale properties. Except that, the obstacle strength of C, Cr, Ni, Mn, Mo and P decorated dislocation loop is studied. Results show that the obstacle strength for dislocation loop decorated by alloy element decreases in the sequence of Cr, Ni, Mn, C, P and Mo, which could be used to help parameterize and validate crystal plasticity finite element model and therein integrated constitutive laws to enable accounting for irradiation-induced chemical segregation effects.

• Carbon letters
• /
• 제12권1호
• /
• pp.8-15
• /
• 2011

We study the relationships between the thermal emissivity of nuclear graphites (IG-110, PCEA, IG-430 and NBG-18) and their surface structural change by oxidation using scanning electron microscope and X-ray diffraction (XRD). The nonoxidized (0% weight loss) specimen had the surface covered with glassy materials and the 5% and 10% oxidized specimens, however, showed high roughness of the surface without glassy materials. During oxidation the binder materials were oxidized first and then graphitic filler particles were subsequently oxidized. The 002 interlayer spacings of the non-oxidized and the oxidized specimens were about $3.38{\sim}3.39{\AA}$. There was a slight change in crystallite size after oxidation compared to the nonoxidized specimens. It was difficult to find a relationship between the thermal emissivity and the structural parameters obtained from the XRD analysis.

• Nuclear Engineering and Technology
• /
• 제50권3호
• /
• pp.411-415
• /
• 2018

Nuclear fusion may be among the strongest sustainable ways to replace fossil fuels because it does not contribute to acid rain or global warming. In this context, activated cobalt materials in corrosion products for fusion energy are significant in determination of dose levels during maintenance after a coolant leak in a nuclear fusion reactor. Therefore, cross-section studies on cobalt material are very important for fusion reactor design. In this article, the excitation functions of some nuclear reaction channels induced by proton particles on $^{59}Co$ structural material were predicted using different models. The nuclear level densities were calculated using different choices of available level density models in ALICE/ASH code. Finally, the newly calculated cross sections for the investigated nuclear reactions are compared with the experimental values and TENDL data based on TALYS nuclear code.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• 제13권1호
• /
• pp.75-83
• /
• 2017

This study investigated deformation behavior of major nuclear structural materials under cyclic loading conditions via cyclic stress-strain test. The cyclic stress-strain tests were conducted on SA312 TP316 stainless steel and SA508 Gr.3 Cl.1 low-alloy steel, which are used as materials for primary piping and reactor pressure vessel nozzle respectively, under cyclic load with constant strain amplitude and constant load amplitude at room temperature (RT) and $316^{\circ}C$. From the results of tests, the cyclic hardening and softening behavior, stabilized cyclic stress-strain behavior, and ratcheting behavior of both materials were investigated at both RT and $316^{\circ}C$. In addition, appropriate considerations for cyclic deformation behavior in the structural integrity evaluation of major nuclear components under excessive seismic condition were discussed.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• 제21권2호
• /
• pp.193-204
• /
• 2023

To ensure radiological safety margin in the transport and storage of spent nuclear fuel, it is crucial to perform source term and shielding analyses in advance from the perspective of conservation. When performing source term analysis on UO₂ fuel, which is mostly used in commercial nuclear power plants, uranium and oxygen are basically considered to be the initial materials of the new fuel. However, the presence of impurities in the fuel and structural materials of the fuel assembly may influence the source term and shielding analyses. The impurities could be radioactive materials or the stable materials that are activated by irradiation during reactor power operation. As measuring the impurity concentration levels in the fuel and structural materials can be challenging, publicly available information on impurity concentration levels is used as a reference in this evaluation. To assess the effect of impurities, the results of the source term and shielding analyses were compared depending on whether the assumed impurity concentration is considered. For the shielding analysis, generic cask design data developed by KEPCO-E&C was utilized.