• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.968-978
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• 2017

Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal-hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.1022-1028
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• 2024

A neutronics study of a supercritical CO₂-cooled fast reactor core for nuclear propulsion has been performed in this work. The thermal power of the reactor core is 30 MWth and a ceramic UO₂ fuel can be used to achieve a 20-year lifetime without refueling. In order to make a compact core with inherent safety features, the drum-type reactivity control system and folding-type shutdown system are adopted. In addition, we suggest a cold shutdown system using gadolinium as a spectral shift absorber (SSA) against flooding. Although there is a penalty of U-235 enrichment for the core embedded with the cold shutdown system, it effectively mitigates the increment of reactivity at the flooding of seawater. In this study, the neutronics analyses have been performed by using the continuous energy Monte Carlo Serpent 2 code with the evaluated nuclear data file ENDF/B-VII.1 Library. The supercritical CO₂-cooled fast reactor core is characterized in view of important safety parameters such as the reactivity worth of reactivity control systems, fuel temperature coefficient (FTC), coolant temperature coefficient (CTC), and coolant temperature-density coefficient (CTDC). We can say that the suggested core has inherent safety features and enough flexibility for load-following operation.

• Nuclear Engineering and Technology
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• v.37 no.6
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• pp.537-556
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• 2005

A primary-pipe rupture accident is one of the design-basis accidents of a High-Temperature Gas-cooled Reactor (HTGR). When the primary-pipe rupture accident occurs, air is expected to enter the reactor core from the breach and oxidize in-core graphite structures. This paper describes an experiment and analysis of the air ingress phenomena and the method fur the prevention of air ingress into the reactor during the primary-pipe rupture accident. The numerical results are in good agreement with the experimental ones regarding the density of the gas mixture, the concentration of each gas species produced by the graphite oxidation reaction and the onset time of the natural circulation of air. A hydrogen production system connected to the High-Temperature Engineering Test Reactor (HTTR) Is being designed to be able to produce hydrogen by themo-chemical iodine-Sulfur process, using a nuclear heat of 10 MW supplied by the HTTR. The HTTR hydrogen production system is first connected to a nuclear reactor in the world; hence a permeation test of hydrogen isotopes through heat exchanger is carried out to obtain detailed data for safety review and development of analytical codes. This paper also describes an overview of the hydrogen permeation test and permeability of hydrogen and deuterium of Hastelloy XR.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4560-4570
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• 2022

Dynamic fault tree (DFT) and its related research methods have received extensive attention in safety analysis and reliability engineering. DFT can perform reliability modelling for systems with sequential correlation, resource sharing, and cold and hot spare parts. A technical modelling method of DFT is proposed for modelling ship collision accidents and loss-of-coolant accidents (LOCAs). Qualitative and quantitative analyses of DFT were carried out using the cutting sequence (CS)/extended cutting sequence (ECS) method. The results show nine types of dynamic fault failure modes in ship collision accidents, describing the fault propagation process of a dynamic system and reflect the dynamic changes of the entire accident system. The probability of a ship collision accident is 2.378 × 10^-9 by using CS. This failure mode cannot be expressed by a combination of basic events within the same event frame after an LOCA occurs in a marine nuclear reactor because the system contains warm spare parts. Therefore, the probability of losing reactor control was calculated as 8.125 × 10^-6 using the ECS. Compared with CS, ECS is more efficient considering expression and processing capabilities, and has a significant advantage considering cost.

• Journal of the Korean Society of Systems Engineering
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• v.15 no.2
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• pp.39-46
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• 2019

The purpose of this paper is to present the results of the study, applying Systems Engineering(SE) method, on the feasibility of developing a Melt-down Proof Modular Micro Reactor(MDP-MMR) for its future deployment in Korea. The reactor is being developed by NCSU (North Carolina State University) due to its advantage of melt-down proof nature of the reactor core. For this paper, the characteristics of the MDP-MMR has been studied in terms of fuel characteristics, inherent safety features and passive safety system. The NCSU's development process has been reviewed applying the SE method, and further research is recommended for the feasibility study on deploying such a modular micro reactor in Korea.

• Journal of the Korean Professional Engineers Association
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• v.30 no.5
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• pp.82-95
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• 1997

A Study of Improvement in Reactor Thermal Power Measurement Method using Kalman Filter. The objectives of the safety analysis of nuclear power plants are to maintain the surface temperature of fuel and fuel cladding within limit value in case of Loss of Coolant accident (LOCA) so that it ensures the safety and reliability of nuclear power plants. The new technique evaluating the reactor power and improvement of existing plant system increase the safety margin of nuclear power plant operation, and accordingly, economic effect will be anticipated. Hereby, 1 would like to introduce reactor power measurement method using Kalman filter that enables to calculate the reactor power more precisely combining the parameters, for example, turbine output as the 1 st stage pressure of high pressure turbine, and reactor power using energy equilibrium relation. It is expected that the new technique will enhance the accuracy of measurement of reactor power and maintain the reliability of nuclear power operation by increasing operational safety margin, and gain the economic benefit

• Journal of the Korean Society of Safety
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• v.35 no.5
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• pp.115-120
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• 2020

From 2009 when the CNS facility was installed, the number of reactor failures due to abnormal CNS facility system has increased significantly. Of the total of 19 nuclear reactor shutdowns over the six years from 2009 to 2019, there were 10 nuclear reactor shutdowns associated with the CNS facility, which are very numerous. Therefore, this report intends to analyze the history of nuclear reactor shutdowns due to CNS facility system failure in detail, and to present the root cause and solution to problems. As a result of FMEA implementation of CNS facility system, a total of 76 SPVs were selected. In addition, 10 cases of reactor shutdown history due to CNS facility system abnormalities were analyzed in detailed, and improvement plans for solving the root cause and problem were suggested for each trip history. The results of this study are expected to be able to operate the domestic research reactor and CNS facilities more stably by providing effective measures to prevent recurrence of CNS facilities and reactor trips.

• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.953-967
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• 2017

In this paper, a complete station blackout (SBO) or complete loss of electrical power supplies is simulated and analyzed in a downward cooling 5-MW pool-type Material Testing Reactor (MTR). The scenario is traced in the absence of active cooling systems and operators. The code nodalization is successfully benchmarked against experimental data of the reactor's operating parameters. The passive heat removal system includes downward water cooling after pump breakdown by the force of gravity (where the coolant streams down to the unfilled portion of the holdup tank), safety flapper opening, flow reversal from a downward to an upward cooling direction, and then the upward free convection heat removal throughout the flapper safety valve, lower plenum, and fuel assemblies. Both short-term and long-term natural core cooling conditions are simulated and investigated using the RELAP5 code. Short-term analyses focus on the safety flapper valve operation and flow reversal mode. Long-term analyses include simulation of both complete SBO and long-term operation of the free convection mode. Results are promising for pool-type MTRs because this allows operators to investigate RELAP code abilities for MTR thermal-hydraulic simulations without any oscillation; moreover, the Tehran Research Reactor is conservatively safe against the complete SBO and long-term free convection operation.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.1017-1023
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• 2019

The increase of the alternate current frequency results in increased rotational speed of the electrical motors and connected pumps. The consequence for the reactor coolant pumps is increased flow in primary coolant system. Increase of the current frequency can be initiated by the subsynchronous resonance phenomenon (SSR). This paper analyses the implications of the SSR and consequential increase of the frequency on the nuclear power plant safety. The Simulink $MATLAB^{(R)}$ model of the steam turbine and governor system and RELAP5 computer code of the pressurized water reactor are used in the analysis. The SSR results in fast increase of reactor coolant pumps speed and flow in the primary coolant system. The turbine trip value is reached in short time following SSR. The increase of flow of reactor coolant pumps results in increase of heat removal from reactor core. This results in positive reactivity insertion with reactor power increase of 0.5% before reactor trip is initiated by the turbine trip. The main parameters of the plant did not exceed the values of reactor trip set points. The pressure drop over reactor core is small discarding the possibility of core barrel lift.

• International Journal of Safety
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• v.7 no.2
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• pp.7-11
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• 2008

In this paper, a Magnetic Core Reactor (MCR) which forms a part of the DC reactor type three-phase high-Tc superconducting fault current limiter (SFCL) has been developed. This SFCL is more economical than other types with three coils since it uses only one high-Tc superconducting (HTS) coil. When DC reactor type three-phase high-Tc SFCL is developed using just one coil, fewer power electronic devices and shorter HTS wire are needed. The SFCL proposed in this paper needs a power-linking device to connect the SFCL to the power system. The design concept for this device was sprang from the fact that the magnetic energy could be changed into the electrical energy and vice versa. Ferromagnetic material is used as a path of magnetic flux. When high-Tc superconducting DC reactor is separated from the power system by using SCRs, this device also limits fault current until the circuit breaker is opened. The device mentioned above was named Magnetic Core Reactor (MCR). MCR was designed to minimize the voltage drop and total losses. Majority of the design parameters was tuned through experiments with the design prototype. In the experiment, the current density of winding conductor was found to be $1.3\;A/mm^2$, voltage drop across MCR was 20 V and total losses on normal state was 1.3 kW.