• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2373-2379
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• 2023

In sodium systems, leakage is one of the safety concerns; it can cause chemical reactions, which may result in fires. There are contact and non-contact types of leak detectors, and the conventional method of non-contact type detection is by gas sampling. Because of the complexity of this method, there has always been a need for a simple gas sensor, and the resistive-type nanostructure ZnO sensor is a promising option with various advantages. In this study, a ZnO sensor was fabricated, and the concept was tested as a leak detector using a dedicated experiment facility. The experiment results showed distinctive changes in resistance with the presence of sodium aerosol under various conditions. Replacing the conventional gas sampling with the ZnO sensors is expected to enable identification of the leakage location if used as a point-wise instrumentation and to greatly reduce the total cost, making the system simple, light, and effective. For further study, more tests will be performed to evaluate the sensitivity of key parameters under various conditions.

• Nuclear Engineering and Technology
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• 제56권1호
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• pp.283-291
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• 2024

Sodium cooled Fast Reactors (SFR) are built with several engineered safety features and hence a severe accident such as a core melt accident is hypothetical with a probability of <10^-6/ry. However, in case of such accidents, the mixture of the molten fuel and structural materials interacts with sodium. This phenomenon is known as Molten Fuel Coolant Interaction (MFCI) and results in fragmentation of the melt due to various instabilities. The fragmented particles settle as a debris bed on the core catcher at the bottom of the reactor vessel, and continue to generate decay heat. Characteristics of the debris particles play a vital role in heat transfer from the bed and need thorough investigation. The size, shape, and physical state of the debris depend on the associated fragmentation mechanism, superheating of the melt, and sodium temperature. Experiments have been conducted by releasing simulated corium, a molten mixture of alumina and iron generated by the aluminothermy process at ~2400 ℃ into liquid sodium, to study the fragmentation phenomena. After the experiment, the fragmented debris was retrieved and the particle size distribution was determined by sieve analysis. The debris was subjected to microscopic investigation for obtaining morphological characteristics. Based on the characteristics of debris, an attempt has been made to assess of fragmentation mechanism of simulated corium in sodium.

• Nuclear Engineering and Technology
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• 제52권5호
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• pp.925-936
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• 2020

To ascertain the characteristics of pool sloshing behavior that might be encountered during a core disruptive accident of sodium-cooled fast reactors, in our earlier work several series of experiments were conducted under various scenarios including the condition with mono-sized solid particles. It is found that under the particle-bed condition, three typical flow regimes (namely the bubble-impulsion dominant regime, the transitional regime and the bed-inertia dominant regime) could be identified and a flow-regime model (base model) has been even successfully established to estimate the regime transition. In this study, aimed to further understand this behavior at more realistic particle-bed conditions, a series of simulated experiments is newly carried out using mixed-size particles. Through analyses, it is verified that for present scenario, by applying the area mean diameter, our previously-developed base model can provide the most appropriate predictive results among the various effective diameters. To predict the regime transition with a form of extension scheme, a correction factor which is based on the volume-mean diameter and the degree of convergence in particle-size distribution is suggested and validated. The conducted analyses in this work also indicate that under certain conditions, the potential separation between different particle components might exist during the sloshing process.

• Nuclear Engineering and Technology
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• 제41권4호
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• pp.427-446
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• 2009

In order to meet the increasing demand for electricity, Korea has to rely on nuclear energy due to its poor natural resources. In order for nuclear energy to be expanded in its utilization, issues with uranium supply and waste management issues have to be addressed. Fast reactor system is one of the most promising options for electricity generation with its efficient utilization of uranium resources and reduction of radioactive waste, thus contributing to sustainable development. The Korea Atomic Energy Research Institute (KAERI) has been performing R&Ds on Sodium-cooled Fast Reactors (SFRs) under the national nuclear R&D program. Based on the experiences gained from the development of KALIMER conceptual designs of a pool-type U-TRU-10%Zr metal fuel loaded reactor, KAERI is currently developing Advanced SFR design concepts that can better meet the Generation IV technology goals. This also includes developing, Advanced SFR technologies necessary for its commercialization and basic key technologies, aiming at the conceptual design of an Advanced SFR by 2011. KAERI is making R&D efforts to develop advanced design concepts including a passive decay heat removal system and a supercritical $CO_2$ Brayton cycle energy conversion system, as well as developing design methodologies, computational tools, and sodium technology. The long-term Advanced SFR development plan will be carried out toward the construction of an Advanced SFR demonstration plant by 2028.

• Nuclear Engineering and Technology
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• 제54권11호
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• pp.4125-4133
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• 2022

KAERI (Korea Atomic Energy Research Institute) developed the conceptual design of PGSFR (Prototype Gen-IV Sodium Cooled Fast Reactor) and Burner Reactor. Since the reactor characteristics of the PGSFR and Burner Reactor are different, the shape, size and the arrangement of the main components in the reactors must be different. Therefore, the conceptual design for the fuel handling and transfer systems needs to be performed coinciding with the structure of the reactor. Especially, because a redan structure dividing hot and cold pool is installed in the reactor vessel, the conceptual design of the fuel handling and transfer system largely changes depending on the location of the redan structure. Various elements of the conceptual design and an integral arrangement for the fuel handling and transfer system were arranged according to the characteristics, sizes and shapes of the reactors. In this paper, the conceptual designs of the fuel handling and transfer system for PGSFR and Burner Reactor are described. Especially, an A-frame method is selected as the fuel handling and transfer system for the Burner Reactor, considering the layout of the internal structure. The tilt angle, diameter and length of A-frame is determined and the strength evaluation of the A-frame is performed.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2077-2083
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• 2022

In the unlikely event of core disruptive accident in sodium cooled fast reactors, the reactor containment building would be bottled up with sodium and fission product aerosols. The behavior of these aerosols is crucial to estimate the in-containment source term as a part of nuclear reactor safety analysis. In this work, the evolution of sodium aerosol characteristics (mass concentration and size) is simulated using HAARM-S code. The code is based on the method of moments to solve the integro-differential equation. The code is updated to FORTRAN-77 and run in Microsoft FORTRAN PowerStation 4.0 (on Desktop). The sodium aerosol characteristics simulated by HAARM-S code are compared with the measured values at Aerosol Test Facility. The maximum deviation between measured and simulated mass concentrations is 30% at initial period (up to 60 min) and around 50% in the later period. In addition, the influence of humidity on aerosol size growth for two different aerosol mass concentrations is studied. The measured and simulated growth factors of aerosol size (ratio of saturated size to initial size) are found to be matched at reasonable extent. Since sodium is highly reactive with atmospheric constituents, the aerosol growth factor depends on the hygroscopic growth, chemical transformation and density variations besides coagulation. Further, there is a scope for the improvement of the code to estimate the aerosol dynamics in confined environment.

• Nuclear Engineering and Technology
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• 제56권1호
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• pp.132-140
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• 2024

In the event of a severe accident in Sodium Cooled Fast Reactors (SFR), the sodium combustion aerosols along with fission product aerosols would migrate to the environment through leak paths of the Reactor Containment Building (RCB) concrete wall under positive pressure. Understanding the characteristics of sodium aerosol transport through concrete leak paths is important as it governs the environmental source term. In this context, experiments are conducted to study the influence of various parameters like pressure, initial mass concentration, leak path diameter, humidity etc., on the transport and deposition of sodium aerosols in straight leak paths of concrete. The leak paths in concrete specimens are prepared by casting and the diameter of the leak path is measured using thermography technique. Aerosol transport experiments are conducted to measure the transported and plugged aerosol mass in the leak paths and corresponding plugging times. The values of differential pressure, aerosol concentration and relative humidity taken for the study are in the ranges 10-15 kPa, 0.65-3.04 g/m³ and 30-90% respectively. These observations are numerically simulated using 1-Dimensional transport equation. The simulated values are compared with the experimental results and reasonable agreement among them is observed. From the safety assessment view of reactor, the approach presented here is conservative as it is with straight leak paths.

• Nuclear Engineering and Technology
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• 제45권6호
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• pp.731-744
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• 2013

Energy security is a topic of high importance to many countries throughout the world. Countries with access to vast energy supplies enjoy all of the economic and political benefits that come with controlling a highly sought after commodity. Given the desire to diversify away from fossil fuels due to rising environmental and economic concerns, there are limited technology options available for baseload electricity generation. Further complicating this issue is the desire for energy sources to be sustainable and globally scalable in addition to being economic and environmentally benign. Nuclear energy in its current form meets many but not all of these attributes. In order to address these limitations, TerraPower, LLC has developed the Traveling Wave Reactor (TWR) which is a near-term deployable and truly sustainable energy solution that is globally scalable for the indefinite future. The fast neutron spectrum allows up to a ~30-fold gain in fuel utilization efficiency when compared to conventional light water reactors utilizing enriched fuel. When compared to other fast reactors, TWRs represent the lowest cost alternative to enjoy the energy security benefits of an advanced nuclear fuel cycle without the associated proliferation concerns of chemical reprocessing. On a country level, this represents a significant savings in the energy generation infrastructure for several reasons 1) no reprocessing plants need to be built, 2) a reduced number of enrichment plants need to be built, 3) reduced waste production results in a lower repository capacity requirement and reduced waste transportation costs and 4) less uranium ore needs to be mined or purchased since natural or depleted uranium can be used directly as fuel. With advanced technological development and added cost, TWRs are also capable of reusing both their own used fuel and used fuel from LWRs, thereby eliminating the need for enrichment in the longer term and reducing the overall societal waste burden. This paper describes the origins and current status of the TWR development program at TerraPower, LLC. Some of the areas covered include the key TWR design challenges and brief descriptions of TWR-Prototype (TWR-P) reactor. Selected information on the TWR-P core designs are also provided in the areas of neutronic, thermal hydraulic and fuel performance. The TWR-P plant design is also described in such areas as; system design descriptions, mechanical design, and safety performance.

• Nuclear Engineering and Technology
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• 제50권6호
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• pp.915-922
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• 2018

Metallic fuel has been considered for sodium-cooled fast reactors because it can maximize the uranium resources. It generates rare earth elements as fission products, where it is reported by aggravating the fuel-cladding chemical interaction at the operating temperature. Rare earth elements form a multicomponent alloy (Ce-Nd-Pr-La-Sm-etc.) during reactor operation, where it shows a higher reaction thickness than a single element. Experiments have been carried out by simplifying multicomponent alloys for mono or binary systems because complex alloys have difficulty in the analysis. In previous experiments, xCe-yNd was fabricated with two elements, Ce and Nd, which have a major effect on the fuel-cladding chemical interaction, and the thickness of the reaction layer reached maximum when the rare earth elements ratio was 1:1. The objective of this study is to evaluate the effect and relationship of rare earth elements on such synergistic behavior. Single and binary rare earth model alloys were prepared by selecting five rare earth elements (Ce, Nd, Pr, La, and Sm). In the single system, Nd and Pr behaviors were close to diffusion, and Ce showed a eutectic reaction. In the binary system, Ce and Sm further increased the reaction layer, and La showed a non-synergy effect.

• 한국기계기술학회지
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• 제13권2호
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• pp.145-153
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• 2011

The irradiation tests of materials in HANARO have been performed usually at temperatures below $300^{\circ}C$ at which the RPV(Reactor Pressure Vessel) materials of the commercial reactors such as the light water reactor and CANDU are operated. As VHTR(Very High Temperature Reactor) and SFR(Sodium-cooled Fast Reactor) projects are being carried as a part of the present Gen-IV program in Korea, the requirements for irradiation of materials at temperatures higher than $500^{\circ}C$ are recently being gradually increased. To overcome the restriction in the use at high temperature of the existing Al thermal media, a new capsule with double thermal media composed of two kinds of materials such as Al-Ti and Al-graphite was designed and fabricated more advanced than the single thermal media capsule. At the irradiation test of the capsule, the temperature of the specimens successfully reached $700^{\circ}C$ and the integrity of Al, Ti and graphite material was maintained.