In this research, a detailed analysis of the decay heat contributions of both actinides and non-actinides (fission fragments) from spent nuclear fuel (SNF) was made after 50 GWd·tHM−1 burnup of fresh uranium fuel with 4.5% enrichment lasted for 1,350 days. The calculations were made for a long storage period of 300 years divided into four sections 1, 10, 100, and 300 years so that we could study the decay heat and physical disposal ratios of radioactive waste in medium- and long-term storage periods. Fresh fuel burnup calculations were made using the code MCNP, while isotopic content and then decay heat were calculated using the built-in stiff equation solver in the MATLAB code. It is noted that only around 12 isotopes contribute more than 90% of the decay heat at all times. It is also noted that the contribution of actinides persists and is the dominant ether despite decreasing decay heat, while the effect of fission products decreases at a very rapid rate after about 40 years of storage.

Molten Salt Reactor, which employs molten salt mixture as fuel, has many advantages in reactor size and operation compared to conventional nuclear reactor. In developing Molten Salt Reactor, the behavior of fission product in operation should be preliminary evaluated for the correct design of reactor and its associated system including off-gas treatment. In this study, for 100 Mw 46 KCl- 54 UCl3 based Molten Salt Reactor with operating life time of 20 year, the fission product behavior was estimated by thermodynamic modeling employing FactSage 8.2. Total inventory of all fission product were firstly calculated using OpenMC code allowing depletion during neutronic calculation. Then, among all inventory, 46 element species from Uranium to Holmium were chosen and given to the input for equilibrium module of Factsage with its mass. In phase equilibrium calculation, for the correct description of solution phase, KCl-UCl3 solution database based on modified quasichemical model in the quadruplet approximation (ANL/CFCT-21/04) was employed and the coexisting solid phase was assumed to pure state. With the assumption of no oxygen and moisture ingress into reactor system, equilibrium calculation showed that 1% of solid phase and of gas phase were newly formed and, in gas phase, major species were identified : ZrCl4 (47%), Xe (33%), UCl4 (14%), Kr (5%), Ar (1%) and others. This result reveals that off-gas treatment of system should account for the appropriate treatment of ZrCl4 and UCl4 besides treatment of noble gas such as Xe and Kr.

Molybdenum-99 (Mo-99) and, its daughter, technetium-99m (Tc-99m) are the most commonly used medical isotope covering more than 85% of the nuclear diagnostics. Currently, majority of Mo-99 supplied in the market is fission-based Mo-99 produced by the fission of U-235 in research reactors. In spite of substitutive production schemes, fission-based Mo-99 is the major source for its significant advantages of high specific activity and large production capacity. The new research reactor (KJRR) is under construction in Gijang, Busan, Korea. The project is aiming 2,000 Ci/week Mo-99 production. For the objective, KAERI has been developed Mo-99 production process using HANARO. Weekly production of 2,000 Ci (100,000 Ci/yr, 6-day calibration) Mo-99 can cover 100% domestic needs, as well as 20% of international demand. However, overall cost for the fission-based Mo-99 production is continuously increasing. Previously, the most Mo-99 producers used weapon-grade highly enriched uranium (HEU) targets. Recently, the use of HEU in private sector is limited for non-proliferation. As a result, major Mo-99 producers are forced to convert their targets from HEU to low enriched uranium (LEU, 19.75% U-235 enrichment). The conversion of Mo-99 target caused waste issue. It is not only because of the 50% less yield in production, but also increment of the radioactive waste by 200%. Therefore, designing optimal radioactive waste treatment strategy for fission-based Mo-99 production is becoming more important than ever. During the process, irradiated LEU targets are dissolved in alkaline solution in hot cells. Fission products other than Mo-99 removed from the solution via series of separation steps. Then Mo-99 is eluted and purified to meet international standard as an active pharmaceutical ingredients (APIs). Radioisotopes of xenon (Xe) and krypton (Kr) generated from the fission of U-235 during the irradiation of the target in the research reactor. Then, the radioactive gas released during the process. The emission of radioactive noble gas from the medical radioisotope production facility can be controlled via delayed release through large charcoal beds. KAERI developed compact xenon adsorption module with chilled carbon column to meet 5 GBq/ day of CTBTO recommendation. Small volume of chilled charcoal can satisfy the guideline, replacing massive gas tank system. Therefore, development of optimized radioactive gas treatment system for the Mo-99 production is one of the essential piece for the successful construction, licensing and operation of the KJRR project.

In case of damaged spent fuels, it would require additional treatment for their transportation and storage to capture the radioactive fission products in a defined space. The canning container for the damaged spent fuels is one way to seal the radioactive fission products inside the container. In the Post Irradiation Examination Facility (PIEF) of KAERI, the Quiver container has been introduced for canning damaged spent fuels from Westinghouse Sweden. The main container body has been manufactured for particle-tightness of spent fuel. In addition, drying equipment is being prepared for gas-tightness of spent fuel. The drying equipment can remove water and fill the inert gas inside the container. Before drying inside the container, we evaluated the volatile fission products inventory because volatile fission products could be released during the drying process. Despite assuming highly conservative hypotheses for the inventory remaining in damaged fuel rods, the amount that could be released during the drying process was less and dose rate levels around the evacuation piping system were low.

Molten Salt Reactor (MSR) is one of Generation-IV nuclear reactors that uses molten salts as a fuel and coolant in liquid forms at high temperatures. The advantages of MSR, such as safety, economic feasibility, and scalability, are attributed from the fact that the molten salt fuel in a liquid state is chemically stable and has excellent thermo-physical properties. MSR combines the fuel and coolant by dissolving the actinides (U, Th, TRU, etc.) in the molten salt coolant, eliminating the possibility of a core meltdown accident due to loss of coolant (LOCA). Even if the molten salt fuel leaks, the radioactive fission products dissolved in the molten salt will solidify with the fuel salt at room temperature, preventing potential leakage to the outside. MSR was first demonstrated at ORNL starting with the Aircraft Reactor Experiment (ARE) in 1954 and was extended to the 7.4 MWth MSRE developed in 1964 and operated for 5 years. Recently, various start-ups, including TerraPower, Terrestrial Energy, Moltex Energy, and Seaborg, have been conducting research and development on various types of MSR, particularly focusing on its inherent safety and simplicity. While in the past, fluoride-based molten salt fuels were used for thermal neutron reactors, recently, a chlorine-based molten salt fuel with a relatively high solubility for actinides and advantageous for the transmutation of spent nuclear fuel and online reprocessing has been developing for fast neutron spectrum MSRs. This paper describes the development status of the process and equipment for producing highpurity UCl3, a fuel material for the chlorine-based molten salt fuel, and the development status of the gas fission product capturing technologies to remove the gaseous fission products generated during MSR operation. In addition, the results of the corrosion property evaluation of structural materials using a natural circulation molten salt loop will also be included.

Molten salt reactors and pyroprocessing are widely considered for various nuclear applications. The main challenges for monitoring these systems are high temperature and strong radiation. Two harsh environments make the monitoring system needs to measure nuclides at a long distance with sufficient resolution for discriminating many different elements simultaneously. Among available methodologies, laser-induced breakdown spectroscopy (LIBS) has been the most studied. The LIBS method can provide the required stand-off and desired multi-elemental measurable ability. However, the change of the level for molten salts induces uncertainty in measuring the concentration of the nuclides for LIBS analysis. The spectra could change by focusing points due to the different laser fluence and plasma shape. In this study, to prepare for such uncertainties, we evaluated a LIBS monitoring system with machine learning technology. While the machine learning technology cannot use academic knowledge of the atomic spectrum, this technique finds the new variable as a vector from any data including the noise, target spectrum, standard deviation, etc. Herein, the partial least squares (PLS) and artificial neural network (ANN) were studied because these methods represent linear and nonlinear machine learning methods respectively. The Sr (580–7200 ppm) and Mo (480–4700 ppm) as fission products were investigated for constructing the prediction model. For acquiring the data, the experiments were conducted at 550°C in LiCl-KCl using a glassy carbon crucible. The LIBS technique was used for accumulating spectra data. In these works, we successfully obtained a reasonable prediction model and compared each other. The high linearities of the prediction model were recorded. The R2 values are over 0.98. In addition, the root means square of the calibration and cross-validation were used for evaluating the prediction model quantitatively.

In KAERI, the nuclide management technology is currently being developed for the reduction of disposal area required for spent fuel management. Among the all fission products of interest, Cs, I, Kr, Tc are considered to be significantly removed by following mid-temperature and high-temperature treatment, however, a difficulty of spent-fuel thermal treatment experiment limits the development of such thermal treatment. In this study, we applied our previously developed two-stage diffusion release model coupled to UO2 oxidation model to the development of optima thermal treatment scenario. Since the formation of cesium pertechnetate should be avoided and the fission release behavior is considerably affected by the extent of oxygen, we obtained oxygen-content dependent model parameters for two-stage fission release model and applied the model to the evaluation of fission release behavior to different oxygen content and thermal treatment procedure. It was found that the developed fission release model closely describes the experimental behavior of fission product of interest, implying a validity of model prediction and the thermal treatment condition reducing the chemical reaction between cesium and technetium could be developed.

Molten Salt Reactor (MSR) is one of the generation-IV advanced nuclear reactors in which hightemperature molten salt mixture is used as the primary coolant, or even the fuel itself unlike most nuclear reactors that adopt solid fuels. The MSR has received a great attention because of its excellent thermal efficiency, high power density, and structural simplicity. In particular, since the MSR uses molten salts with boiling points higher than the exit temperature of the reactor core, there is no severe accident such as a core melt-down which leads to a hydrogen explosion. In addition, it is possible to remove the residual heat through a completely passive way and when the fuel salt leaks to the outside, it solidifies at room-temperature without releasing radioactive fission products such as cesium, which make the MSR inherently safe. Both fluoride and chloride mixtures can be used as liquid fuel salts by adding actinide halides for MSRs. However, the MSRs using chloride-based salt fuels can be operated for a long time without adding nuclear fuel or online reprocessing because the actinide solubility in chloride salts is about six times higher than that in fluoride salts. Therefore, the chloride-based MSRs are more effective for the transmutation of long-lived radionuclides such as transuranic elements than the fluoride-based MSRs, which is beneficial to resolve the high radioactive spent nuclear fuel generated from light water reactors (LWRs). This paper examines liquid fuel fabrication using an improved U chlorination process for the chloride-based MSRs and presents the strategy for the management of gaseous fission products generated during the operation of MSR.

One of the promising candidates for accident-tolerant fuel (ATF), a ceramic microcell fuel, which can be distinguished by an unusual cell-like microstructure (UO2 grain cell surrounded by a doped oxide cell wall), is being developed. This study deals with the microstructural observation of the constituent phases and the wetting behaviors of the cell wall materials in three kinds of ceramic microcell UO2 pellets: Si-Ti-O (STO), Si-Cr-O (SCO), and Al-Si-Ti-O (ASTO). The chemical and physical states of the cell wall materials are estimated by HSC Chemistry and confirmed by experiment to be mixtures of Si-O and Ti-O for the STO; Si-O and Cr-O for SCO; and Si-O, Ti-O, and Al-Si-O for the ASTO. From their morphology at triple junctions, UO2 grains appear to be wet by the Si-O or Al-Si-O rather than other oxides, providing a benefit on the capture-ability of the ceramic microcell cell wall. The wetting behavior can be explained by the relationships between the interface energy and the contact angle.

몰리브덴-99의 붕괴에 의해 생산되는 테크네튬-99m 은 방사선 진단에 중요한 역할을 담당하고 있다. 몰리브덴-99 는 주로 우라늄-235의 핵분열에 의해 생산되고 있으며, 이를 위해 고농축 우라늄 표적 또는 저농축 우라늄 표적이 연구로에서 조사 된다. 현재는 고농축 우라늄의 사용에 따른 핵확산 문제를 저감하기 위해 저농축 우라늄 표적의 사용이 권장되고 있다. 본 연구는 몰리브덴-99 생산 시설의 계획 단계에서 방사성 폐기물 관리 전략을 정의하기 위하여 저농축 우라늄의 사용이 방사 성 폐기물의 흐름에 미치는 영향을 분석하였다. 저농축 우라늄 표적 사용 시 우라늄 함유 폐기물의 부피가 6배 이상 증가하 기 때문에 우라늄 고밀도 표적의 사용과 고온 정수압 압축법의 활용이 제안되었다.

본 연구는 H2O2가 함유된 (Na2CO3-NaHCO3) 혼합 탄산염 계에서 사용후핵연료를 산화용해할 시 U과 함께 공용해 되는 Cs, Te, Tc, Mo 등의 핵분열생성물로부터 Cs과 Tc의 선택적 침전 제거 거동을 규명하였다. Cs과 Tc은 각각 장수명 핵종으로 지하에서의 빠른 핵종 이동성과 고방열성 등으로 최종 처분 시 처분 환경 을 저해하는 핵종으로 처분 안전성 제고 측면에서 이들의 제거는 중요한 과제 중의 하나이다. Cs과 Re (Tc대용원소)의 선택적 침전제로는 각각 NaTPB, TPPCl를 선정하였으며, NaTPB에 의한 Cs 침전 및 TPPCl에 의한 Re 침전 모두 5분 이내로 매우 빠르게 이루어졌으며, 온도를 50℃, 교반속도를 1000 rpm 까지 증가 시켜도 이들의 침전 속도에는 별 영향이 없었다. NaTPB 침전 및 TPPCl 침전에 있어 가장 중요한 요인은 침전 용액의 pH 이며, 특히 TPPCl에 의한 Re의 선택적 침전의 경우 낮은 pH 에서 Mo가 Re과 공침되므로 pH 9 이상에서 수행하는 것이 효과적이다. 그리고 [NaTPB]/[Cs] 및 [TPPCl]/[Re]의 몰 농도 비 1 이상에서 Cs 및 Re을 각각 99% 이상 선택적으로 침전 제거할 수 있었다.