• Transactions of the Korean Society of Mechanical Engineers B
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• v.34 no.5
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• pp.553-561
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• 2010

A medium-scale helium loop that can simulate a VHTR (very-high-temperature reactor) is now under construction at the Korea Atomic Energy Research Institute. The heaters of the test helium loop electrically heat helium fluid up to $950^{\circ}C$ at pressures of 1 to 9 MPa. To optimize the design specifications of the experimental helium loop, the conjugate heat transfer in the high-temperature helium heater was analyzed by performing a CFD simulation. The analysis results indicate that the maximum temperature does not exceed the allowable limit. It is confirmed that the thermal characteristics of the loop with the given geometry satisfy the design requirements.

• Journal of the Korean Society of Safety
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• v.34 no.3
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• pp.102-110
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• 2019

This paper describes the work and results of the preliminary Probabilistic Safety Assessment (PSA) for a research reactor in the design phase. This preliminary PSA was undertaken to assess the level of safety for the design of a research reactor and to evaluate whether it is probabilistically safe to operate and reliable to use. The scope of the PSA described here is a Level 1 PSA which addresses the risks associated with core damage. After reviewing the documents and its conceptual design, eight typical initiating events are selected regarding internal events during the normal operation of the reactor. Simple fault tree models for the PSA are developed instead of the detailed model at this conceptual design stage. A total of 32 core damage accident sequences for an internal event analysis were identified and quantified using the AIMS-PSA. LOCA-I has a dominant contribution to the total CDF by a single initiating event. The CDF from the internal events of a research reactor is estimated to be 7.38E-07/year. The CDF for the representative initiating events is less than 1.0E-6/year even though conservative assumptions are used in reliability data. The conceptual design of the research reactor is designed to be sufficiently safe from the viewpoint of safety.

• Journal of the Korean Society of Safety
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• v.37 no.4
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• pp.129-138
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• 2022

A new method for injecting cooling water into the Korean research reactor (KRR) in the event of beam tube rupture is proposed in this paper. Moreover, the research evaluates the risk to the reactor core in terms of core damage frequency (CDF). The proposed method maintains the cooling water in the chimney at a certain level in the tank to prevent nuclear fuel damage solely by gravitational coolant feeding from the emergency water supply system (EWSS). This technique does not require sump recirculation operations described in the current procedure for resolving beam tube accidents. The reduction in the risk to the core in the event of beam tube rupture that can be achieved by the proposed change in the cooling water injection design is quantified as follows. 1) The total CDF of the KRR for the proposed design change is approximately 4.17E-06/yr, which is 8.4% lower than the CDF of the current design (4.55E-06/yr). 2) The CDF for beam tube rupture is 7.10E-08/yr, which represents an 84.1% decrease compared with that of the current design (4.49E-07/yr). In addition to this quantitative reduction in risk, the modified cooling water injection design maintains a supply of pure coolant to the EWSS tank. This means that the reactor does not require decontamination after an accident. Thermal hydraulic analysis proves that the water level in the reactor pool does not cause damage to the nuclear fuel cladding after beam tube rupture. This is because the amount of water in the chimney can be regulated by the EWSS function. The EWSS supplies emergency water to the reactor core to compensate for the evaporation of coolant in the core, thus allowing water to cover the fuel assemblies in the reactor core over a sufficient amount of time.

• Transactions of the Korean hydrogen and new energy society
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• v.23 no.1
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• pp.73-82
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• 2012

A 2-dimensional heterogeneous reactor model was developed and simulated for a tube reactor of packed bed where the steam-$CO_2$ combined reforming reaction of natural gas proceeded to produce synthesis gas. Under the reactor feeding rate, 45 $Nm^3$/h, of the reactant gas stream, the 2-dimensional heterogeneous reactor model showed the similar results to those from the ASPEN simulator although there were some discrepancies between the two in the temperature and the $H_2$/CO ratio of the reformed gas at the reactor exit. The calculated enthalpy difference between the reformed gas at the reactor exit and the reactant gas fed to the reactor was closely correspondent to the total amount of heat transferred to the reactor interior from the furnace. This supports that the 2-dimensional heterogeneous reactor model was reasonably established and the numerical solution was properly obtained.

• Nuclear Engineering and Technology
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• v.42 no.3
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• pp.329-338
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• 2010

We developed an achievable control logic for the reactor power level during a power maneuvering event and set up some constraints for the control of the reactor power in a conceptual sodium-cooled fast reactor (KALIMER-600) that was developed at KAERI. For simulating the dynamic behaviors of the plant, we developed a fast-running performance analysis code. Through various simulations of the power maneuvering event, we evaluated some suggested control logic for the reactor power and found an achievable control logic. The objective of the control logic is to search for the position of the control rods that would keep the average temperature of the primary pool constant and, concurrently, minimize the power deviation between the reactor and the BOP cycle during the power maneuvering. In addition, the flow rates of the primary pool and the intermediate loop should be changed according to the power level in order to not violate the constraints set up in this study. Also, we evaluated some movement speeds of the control rods and found that a fast movement of the control rods might cause the power to fluctuate during the power maneuvering event. We suggested a reasonable movement speed of the control rods for the developed control logic.

• 제어로봇시스템학회:학술대회논문집
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• 2003.10a
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• pp.2063-2065
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• 2003

The control rod drive mechanisms in a reactor head are arranged too narrow for a human worker to approach. Moreover, the working environment is in high radiation area. In order to inspect defections in the surfaces of the reactor head and welding parts, a visual inspection device that can approach such a narrow and high radiation area is required. This paper introduces a tele-operated mobile robot for visual inspection of a reactor head, which has pan/tilt camera, fixed rear camera, ultrasonic collision detection system, and so on. Moreover, the host controller and digital video logging system are developed and integrated control software is also developed. The robot is operated by a wireless control, which gives flexibility for the inspection.

• Nuclear Engineering and Technology
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• v.31 no.6
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• pp.1-7
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• 1999

An innovative concept for a passive reactor shutdown system, so called self-actuated shutdown system(SASS), is inevitably required for the inherent safety in liquid metal reactor, which is designed with the totally different concept from the usual reactor shutdown system in LWR. SASS using Curie point electromagnet(CPEM) was selected as the passive reactor shutdown system for KALIMER (Korea Advanced Liquid MEtal Reactor). A mock-up of the SASS was designed, fabricated and tested. From the test it was confirmed that the mockup was self-actuated at the Curie point of the temperature sensing material used in the mockup. An articulated control rod was also fabricated and assembled with the CPEM to confirm that the control rod can be inserted into core even when the control rod guide tube is deformed due to earthquake. The operability of SASS in the actual sodium environment should be confirmed in the future. All the design and test data will be applied to the KALIMER design.

• Nuclear Engineering and Technology
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• v.30 no.2
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• pp.83-90
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• 1998

To predict effects of seismic isolation, seismic isolation bearings were applied to the Wolsong reactor building and the analytical study was performed. For this study, the Wolsong reactor building was modeled using lumped masses and beam elements. Design Basis Earthquake with a ground acceleration of 0.2g was applied. And then, the behavior of the isolated structure was compared with that of the unisolated structure. The horizontal response acceleration at the top of the unisolated reactor building was 0.99g, while that of the isolated one was 0.14g(15% damping) and the acceleration response along the height of the structure was constant. The maximum displacement of the unisolated structure was 8.3mm, while that of the isolated structure was 66mm. The application of isolation bearings on the reactor building reduces seismic loads but increases the displacement of the structure on a large scale. Therefore, when using isolation bearings, the reactor building and BOP should be located on a common mat to cover the large displcement.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.4
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• pp.443-449
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• 2014

To check the characteristics of nuclear fuels during an irradiation test, the nuclear fuel rod needs to be disassembled from the test rig located in the pool of the research reactor. Then, the disassembled fuel rod is delivered to the hot cell for intermediate examination. A fuel rod that passes the intermediate examination is delivered to the reactor pool to be reassembled into the test rig. The irradiation test is resumed with the reassembled test rig. Because nuclear fuel rods irradiated by neutrons are highly radioactive, all the disassembly and reassembly processes should be carried out in the pool of the research reactor to prevent operators being exposed to radiation. In particular, because a test rig is 5.4-m long and the reactor pool of HANARO is 6-m deep, special tools need to be developed for performing the disassembly and reassembly processes. In this study, a new assembly design of nuclear fuel rods for intermediate examination is introduced. Furthermore, tools for treating the irradiated fuel rod assembly are introduced, and their performance is verified by an out pile test.

• Nuclear Engineering and Technology
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• v.52 no.4
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• pp.764-775
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• 2020

Periodic safety reviews (PSRs) are conducted on operating nuclear power plants (NPPs) and have been mandated also for research reactors in Korea, in response to the Fukushima accident. One safety review tool, the probabilistic safety assessment (PSA), aims to identify weaknesses in the design and operation of the research reactor, and to evaluate and compare possible safety improvements. However, the PSA for research reactors is difficult due to scarce data availability. An important element in the analysis of research reactors is the reactor protection system (RPS), with its functionality and importance. In this view, we consider that of the AGN-201K, a zero-power reactor without forced decay heat removal systems, to demonstrate a risk-informed safety improvement study. By incorporating risk- and safety-significance importance measures, and sensitivity and uncertainty analyses, the proposed method identifies critical components in the RPS reliability model, systematically proposes potential safety improvements and ranks them to assist in the decision-making process.