• Nuclear Engineering and Technology
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• 제46권5호
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• pp.595-604
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• 2014

SILER (Seismic-Initiated event risk mitigation in LEad-cooled Reactors) is a Collaborative Project, partially funded by the European Commission in the $7^{th}$ Framework Programme, aimed at studying the risk associated to seismic-initiated events in Generation IV Heavy Liquid Metal reactors, and developing adequate protection measures. The project started in October 2011, and will run for a duration of three years. The attention of SILER is focused on the evaluation of the effects of earthquakes, with particular regards to beyond-design seismic events, and to the identification of mitigation strategies, acting both on structures and components design. Special efforts are devoted to the development of seismic isolation devices and related interface components. Two reference designs, at the state of development available at the beginning of the project and coming from the $6^{th}$ Framework Programme, have been considered: ELSY (European Lead Fast Reactor) for the Lead Fast Reactors (LFR), and MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) for the Accelerator-Driven Systems (ADS). This paper describes the main activities and results obtained so far, paying particular attention to the development of seismic isolators, and the interface components which must be installed between the isolated reactor building and the non-isolated parts of the plant, such as the pipe expansion joints and the joint-cover of the seismic gap.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(1)
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• pp.91-96
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• 1995

In this paper, research efforts to develop computer code system for analysis of the transmutation reactors at KAIST are described Especially the computer code HANCELL for assembly calculation of fast reactors is mainly described. Features and function of the code are identified md current status of the code development is provided

• Nuclear Engineering and Technology
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• 제51권6호
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• pp.1504-1513
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• 2019

Recently, ocean nuclear reactors have received attention due to enhanced safety features. The movable and transportable characteristics distinguish ocean nuclear reactors from land-based nuclear reactors. Therefore, for safety/design analysis of the ocean reactor, the thermos-hydraulics must be investigated in the moving system. However, there are no studies reporting the general two-fluid equations that can be used for multi-dimensional simulations of two-phase flows in moving systems. This study is to systematically formulate the multi-dimensional two-fluid equations in the non-inertial frame of reference. To demonstrate the applicability of the formulated equations, we perform a total of six different simulations in 2D tanks with translational and/or rotational motions.

• 방사성폐기물학회지
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• 제17권1호
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• pp.47-58
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• 2019

The paper presents the results of a multi-criteria comparative evaluation of 12 feasible Russian nuclear energy deployment scenarios with thermal and fast reactors in a closed nuclear fuel cycle. The comparative evaluation was performed based on 6 performance indicators and 5 different MCDA methods (Simple Scoring Model, MAVT / MAUT, AHP, TOPSIS, PROMETHEE) in accordance with the recommendations elaborated by the IAEA/INPRO section. It is shown that the use of different MCDA methods to compare the nuclear energy deployment scenarios, despite some differences in the rankings, leads to well-coordinated and similar results. Taking into account the uncertainties in the weights within a multi-attribute model, it was possible to rank the scenarios in the absence of information regarding the relative importance of performance indicators and determine the preference probability for a certain nuclear energy deployment scenario. Based on the results of the uncertainty/sensitivity analysis and additional analysis of alternatives as well as the whole set of graphical and attribute data, it was possible to identify the most promising nuclear energy deployment scenario under the assumptions made.

• Nuclear Engineering and Technology
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• 제50권7호
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• pp.1051-1059
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• 2018

Fast periodic pulsed reactor is a type of reactor in which the fission bursts are formed entirely with external reactivity modulation with a specified time periodicity. This type of reactors could generate much larger intensity of neutron beams for experimental use, compared with the steady state reactors. In the design of fast periodic pulsed reactors, the time dependent simulation of the power pulse is majorly based on a point kinetic model, which is known to have limitations. A more accurate calculation method is desired for the design analyses of fast periodic pulsed reactors. Monte Carlo computer code MCNP6 is used for this task due to its three dimensional transport capability with a continuous energy library. Some new routines were added to simulate the rotation of the movable reflector parts in the time dependent calculation. Fast periodic pulsed reactor IBR-2M was utilized to validate the new routines. This reactor is periodically in prompt supercritical state, which lasts for ${\sim}400{\mu}s$, during the equilibrium state. This generates long neutron fission chains, which requires tremendously large amount of computation time during Monte Carlo simulations. Russian Roulette was applied for these very long neutron chains in MCNP6 calculation, combined with other approaches to improve the efficiency of the simulations. In the power pulse of the IBR-2M at equilibrium state, there is some discrepancy between the experimental measurements and the calculated results using the point kinetics model. MCNP6 results matches better the experimental measurements, which shows the merit of using MCNP6 calculation relative to the point kinetics model.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.292-297
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• 1997

An alternating conditional expectation (ACE) algorithm, a kind of non-parametric regression method, is proposed to generate empirical correlations automatically. The ACE algorithm yields an optimal relationship between a dependent variable and multiple independent variables without any preprocessing and initial assumption on the functional forms. This algorithm is applied to a collection of 12,879 CHF data points for forced convective boiling hi vertical tubes to develop a new critical heat flux (CHF) correlation. The meat root mean square, and maximum errors of our new correlation are -0.558%, 12.5%, and 122.6%, respectively. Our CHF correlation represents the entire set of CHF data with an overall accuracy equivalent to or better than that of three existing correlations.

• Nuclear Engineering and Technology
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• 제55권12호
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• pp.4477-4490
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• 2023

Fuel locking mechanisms (FLMs) are essential in upward-flow research reactors to prevent accidental fuel separation from the core during reactor operation. This study presents a novel design concept for a remotely controlled plate-type nuclear fuel locking mechanism. By employing electromagnetic field analysis, we optimized the design of the electromagnet for fuel unlocking, allowing the FLM to adapt to various research reactor core designs, minimizing installation space, and reducing maintenance efforts. Computational flow analysis quantified the drag acting on the fuel assembly caused by coolant upflow. Subsequently, we performed finite element analysis and evaluated the structural integrity of the FLM based on the ASME boiler and pressure vessel (B&PV) code, considering design loads such as dead weight and flow drag. Our findings confirm that the new FLM design provides sufficient margins to withstand the specified loads. We fabricated a prototype comprising the driving part, a simplified moving part, and a dummy fuel assembly. Through basic operational tests on the assembled components, we verified that the manufactured products meet the performance requirements. This remote-controlled micro locking mechanism holds promise in enhancing the safety and efficiency of plate-type nuclear fuel operation in upflow research reactors.

• Nuclear Engineering and Technology
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• 제56권2호
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• pp.601-610
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• 2024

This paper gives an extensive analysis on the characterization of xenon isotopic ratios for various nuclear reactors and nuclear explosions through neutronic depletion codes. The results of the characterization can be used for discriminating the sources of the xenon isotopes' release among the nuclear explosions and nuclear reactors. The considered sources of the xenon radionuclides do not only include PWR, CANDU, and nuclear explosions using uranium and plutonium bombs, but also IRT-200 and 5MWe Yongbyon (MAGNOX reactor) research reactors operated in North Korea. A new data base (DB) on xenon isotopic activity ratios was produced using the results of the characterization, which can be used in discrimination of the sources of xenon isotopes. The results of the study show that 5MWe Yongbyon reactor has quite different characteristics in ¹³⁵Xe/¹³³Xe ratio from the PWRs and the nuclear reactors have different characteristics in ¹³⁵Xe/¹³³Xe ratios from the nuclear explosions.

• Environmental Engineering Research
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• 제12권4호
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• pp.148-154
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• 2007

The purpose of this research was to investigate the effects of low concentration of oxygen on reduction of perchlorate, especially low perchlorate influent concentrations in a biofilm reactor, as well as the effect of flow pattern in a biofilm reactor. Dissolved oxygen averaging 1 mg/L did not inhibit reduction of influent perchlorate from 23 to $426\;{\mu}g/L$ in the biofilm reactors when sufficient acetate was added, probably due to limitation of oxygen diffusion into the biofilm. Influent perchlorate ranging from 23 to $426\;{\mu}g/L$ was reduced to below detection level ($4\;{\mu}g/L$) in the presence of 1 mg/L dissolved oxygen (DO). Chloride was produced in a ratio of $0.37gCl^-/g{ClO_4}^-$ and $0.35gCl^-/g{ClO_4}^-$ in plug flow and recirculation biofilm reactor which is similar to stoichiometric amount ($0.36gCl^-/g{ClO_4}^-$) indicating complete perchlorate reduction at $426\;{\mu}g/L$ of ${ClO_4}^-$ feeding. At $23\;{\mu}g/L$L influent perchlorate, total biomass solids were 3.18 g and 2.81 g in the plug flow and recirculation biofilm reactors. The most probable number(MPN) analysis for perchlorate-reducing bacteria showed $10^4$ to $10^5\;cells/cm^2$ in both biofilm reactors throughout the experiments. The effluent perchlorate concentrations were not significantly different in the two different flow regimes, plug flow and recirculation biofilm reactors.

• Nuclear Engineering and Technology
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• 제55권4호
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• pp.1244-1249
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• 2023

Recently, the Kingdom of Saudi Arabia (KSA) announced the development of first-of-a-kind(FOAK) and most advanced futuristic vertical city and named as 'The LINE'. The project will have zero carbon dioxide emissions and will be powered by clean energy sources. Therefore, a study was designed to understand which clean energy sources might be a better choice. Because of its nearly carbon-free footprint, nuclear energy may be a good choice. Nowadays, the development of very small modular reactors (vSMRs) is gaining attention due to many salient features such as cost efficiency and zero carbon emissions. These reactors are one step down to actual small modular reactors (SMRs) in terms of power and size. SMRs typically have a power range of 20 MWe to 300 MWe, while vSMRs have a power range of 1-20 MWe. Therefore, a study was conducted to discuss different vSMRs in terms of design, technology types, safety features, capabilities, potential, and economics. After conducting the comparative test and analysis, the fuel cycle modeling of optimal and suitable reactor was calculated. Furthermore, the levelized unit cost of electricity for each reactor was compared to determine the most suitable vSMR, which is then compared other generation SMRs to evaluate the cost variations per MWe in terms of size and operation. The main objective of the research was to identify the most cost effective and simple vSMR that can be easily installed and deployed.