2025.4. ~ 2025.12., National Research Foundation of Korea, Hogyu Yi
2021.5. ~ 2023.12., Nuclear Safety and Security Commission (NSSC), Woong Ha
2019.4. ~ 2021.12., National Research Foundation of Korea, Geon Kim & Yunsong Jung
To successfully manage the safety of used nuclear fuel (UNF) pools for unexpectedly long-term due to unsettled public acceptance on the establishment of permanent disposal site, material integrity of neutron absorber has to be monitored and ensured perhaps over 40 years, which may not so easy without some possible helps from the multi-simulation code to be developed in this project. Many neutron absorbers used in UNF pools are based on boron-10 which emits 1~2 MeV alpha particle and lithium ion when absorbing a neutron. Until today, there is no known simulation code to calculate the radiation damage, and the distribution of helium and lithium in the absorber, resulted from the energetic ions emitted from B-10(n, alpha)Li-7 reactions. Without these information, we cannot accurately predict the corrosion behavior of the type of absorbers since the induced radiations change practically almost all parameters that determine the speed of material corrosion: (1) surface porosity; (2) effective absorber temperatures; (3) athermal diffusivity via ballistic mixing; and even (4) minor electric potential difference, a maybe.
Schematic illustration of 10B(n,𝛼)7Li reaction
2018.10. ~ 2020.12., National Research Foundation of Korea, Changhyeon Nam & Myeongkyu Lee
Diesel-powered ships around the ocean are emitting truly creepy amount of CO2 and SOx, like 16 Republic of Korea floating around the world in emission standpoint. Those ships have to be replaced in the incoming several decades (if human beings are not insane enough) with some viable alternative energy-powered ones like nuclear. To achieve this end, one may need to invent a concept of very high density (or Zr-lean) Zr-U alloy fuel to enable the use of LEU (instead of 90% enriched HEU) as the inherently safe fuel for marine propulsion systems that would not be a proliferation issue. In this regard, we would like to demonstrate acceptable corrosion and radiation resistance of U-rich Zr-U fuel and its chemical interaction with Zircaloy and a few advanced claddings.
Fabrication of Zr-U alloy with arc melting furnace
2018.2. ~ 2020.2., Korea Hydro & Nuclear Power Co., Ltd. (KHNP), Yunsong Jung & Myeongkyu Lee
Used nuclear fuel assemblies stored in Korea has to endure longer times these days, although damaged surface of neutron absorber materials and their encapsulations already has been spotted. For the safe longer term underwater storage of UNF, one may need to monitor the status of neutron absorber and be able to predict their long-term performance. To this end, the degradation mechanisms of neutron absorber, corrosion and radiation damage, have to be investigated fully. It is also well known scientific fact that the underwater corrosion of nuclear materials can be remarkably expedited by radiation damage that could create porous layer at the surface of materials. Major kinds of neutron absorber materials used in Korea will be tested in this project for the next two years, which is a bit short but directly asked by the client.
Helium bubble formation in aluminum matrix
2017.9. ~ 2022.9., National Research Foundation of Korea, Gyeonghun Kim & Jungsu Ahn
This particular design of MMR is hiring super-critical CO2 (S-CO2) as coolant to achieve its smallest dimension with high power density, hence the transportability and economics at the same time. By the adoption of S-CO2 as its coolant, however, the fuel-coolant compatibility becomes a big question mark that has not been experimentally explored. Very high temperature coolant is also unavoidable to achieve the highest thermal efficiency. All together, the design of fuel system for this MMR has to be conducted in a delicate manner considering required high fissile density and high melting point. While all the options are kept on the table, a high smear density refractory ceramic fuel encapsulated with Oxide-dispersed-strengthened (ODS) steel cladding could be a viable candidate that may deserve the first trial to assess its feasibility as the fuel system of this first-of-its-kind reactor.
Schematic illustration of solid monolithic fuel
2017.4. ~ 2021.12., National Research Foundation of Korea, Gyunghoon Kim & Jungsu Ahn
There are few viable options for the LFR fuel: perhaps, MOX and UN are the only two considerable. We are going to compare these options to suggest an optimized fuel design for Korean version of LFR. Experimental focus will be on interactions between candidate fuels and advanced cladding materials to deliver the better accident-tolerance. In the same spirit, fuel-coolant interaction experiments are also planned.
2017.4. ~ 2019.9., Korea Hydro & Nuclear Power Co., Ltd. (KHNP), Geon Kim
Numerical simulations of nuclear fuel are hardly ever believed, especially by skeptical experimentalists, mainly due the fact that many interplay between a bunch of phenomena strongly coupled to each other have been separately modeled with too much simplification. This situation has to be put to a change. As a start, a fuel performance code that can simulate entire geometry of a fuel pin without any axisymmetric overlooking is now under development phase in the URANUM laboratory. This project will be a meaningful first step for the development of a multi-physics fuel performance code for the Republic of Korea.
Fuel performance code calling sequence & axisymmetric heat transfer result
2017.4. ~ 2018.11., Korea Institute of Nuclear Nonproliferation and Control (KINAC), Gyunghoon Kim
The IAEA has been leaning on Thermo-Gravimetric Analysis (TGA) for UO2 stoichiometry measurement for the practice of nuclear material accountancy (NMA), which is not applicable for on-site inspection. In the spirit of making global effort to address this issue and to strengthen our State System of Accounting for and Control of nuclear material (SSAC), the Korea Institute of Nuclear Nonproliferation and Control (KINAC) recently launched this project in cooperation with the URANUM laboratory to develop a better methodology that could potentially provide on-site inspection capability with higher precision than the TGA method. Since thermal conductivity of UO2 is strongly tied to its stoichiometry, on-going investigation is focused on (but certainly not limited to) the utilization of Laser Flash Analysis (LFA). The ultimate goal of this project is to deliver a neat conceptual design for a portable (preferably hand-carry) high precision device that can measure UO2 stoichiometry of any samples collected from the IAEA and KINAC inspection for nuclear safeguards.
2016.6. ~ 2022.12., National Research Foundation of Korea, Jungsu Ahn & Woong Ha
The ‘ATOM’ is an innovative concept of Small Modular Reactor (SMR) that enables automatic load-following operation without boron chemical shim in water coolant, which, in exchange, mandates lower fuel temperature. Hence, the invention of a high thermal conductivity ceramic nuclear fuel has to precede the development of the ATOM. To address this issue, UO2–based hybrid ceramic fuel options are being explored during the first phase of this project. Other ceramic fuel, perhaps U3Si2 or UN variants would be the focus of the second phase (2020-2022) of this project.
Fabrication of U3Si2 pellet with Dr.sinter SPS 210 equipment
2016.6. ~ 2019.5., National Research Foundation of Korea, Myeongkyu Lee
Materials inside advanced nuclear reactors have to withstand extreme radiation damage at high temperature. Particularly, the fuel cladding of Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) has to endure ~200 displacement per atom (dpa) at ~600 °C. To this end, advanced cladding materials for the PGSFR were recently developed by the Korea Atomic Energy Research Institute (KAERI), namely, FC.92-N and FC.92-B, which are 9Cr Ferritic/Martensitic Steels (FMS) designed in particular order to provide superior mechanical performance under high temperature (~550 °C) coolant. In-core performance and radiation resistance of these F/M steels are currently under test in a modified fast-neutron spectrum reactor in Russia, BOR-60, which requires at least 5 to 10 years of irradiation and over tens of millions of dollars to be paid outside our country. On the other hand, utilizing ion-beam accelerators, we can complete numerous irradiation tests on various F/M steels and Fe-Cr alloy specimens within a couple of years at much lower cost. The test could provide very useful supplementary information for fuel performance modeling.
