• 한국안전학회지
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• 제36권2호
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• pp.111-119
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• 2021

Multi-dimensional analysis of reactor safety-KINS standard (MARS-KS) is a thermal-hydraulic code to simulate multiple design basis accidents in reactors. The code has been essential to assess nuclear safety, but has mainly focused on light water reactors, which are in the majority in South Korea. Few previous studies considered pressurized heavy water reactor (PHWR) applications. To verify the code applicability for PHWRs, it is necessary to develop MARS-KS input decks under various transient conditions. This study proposes an input model to simulate small-break loss of coolant accidents for PHWRs. The input model includes major equipment and experimental conditions for test B9802. Calculation results for selected variables during steady-state closely follow test data within ±4%. We adopted the Henry-Fauske model to simulate break flow, with coefficients having similar trends to integrated break mass and trip time for the power supply. Transient calculation results for major thermal-hydraulic factors showed good agreement with experimental data, but further study is required to analyze heat transfer and void condensation inside steam generator u-tubes.

• Nuclear Engineering and Technology
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• 제41권4호
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• pp.493-520
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• 2009

During the last four decades, 16 Pressurized Water Reactors (PWR) and 4 Pressurized Heavy Water Reactors (PHWR) have been constructed and operated in Korea, and nuclear fuel technology has been developed to a self-reliant state. At first, the PWR fuel design and manufacturing technology was acquired through international cooperation with a foreign partner. Then, the PWR fuel R&D by Korea Atomic Energy Research Institute (KAERI) has improved fuel technology to a self-reliant state in terms of fuel elements, which includes a new cladding material, a large-grained $UO_2$ pellet, a high performance spacer grid, a fuel rod performance code, and fuel assembly test facility. The MOX fuel performance analysis code was developed and validated using the in-reactor test data. MOX fuel test rods were fabricated and their irradiation test was completed by an international program. At the same time, the PWR fuel development by Korea Nuclear Fuel (KNF) has produced new fuel assemblies such as PLUS7 and ACE7. During this process, the design and test technology of fuel assemblies was developed to a self-reliant state. The PHWR fuel manufacturing technology was developed and manufacturing facility was set up by KAERI, independently from the foreign technology. Then, the advanced PHWR fuel, CANFLEX(CANDU Flexible Fuelling), was developed, and an irradiation test was completed in a PHWR. The development of the CANFLEX fuel included a new design of fuel rods and bundles.. The nuclear fuel technology in Korea has been steadily developed in many national R&D programs, and this advanced fuel technology is expected to contribute to a worldwide nuclear renaissance that can create solutions to global warming.

• 한국전산유체공학회지
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• 제17권3호
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• pp.75-86
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• 2012

We developed a preliminary CFD analysis methodology to predict a pressure build up due to hydrogen flame acceleration in the APR1400 IRWST on the basis of CFD analysis results for test data of hydrogen flame acceleration in a scaled-down test facility performed by Korea Atomic Energy Research Institute. We found out that ANSYS CFX-13 with a combustion model of the so-called turbulent flame closure and a model constant of A = 5.0, a grid model with a hexahedral cell length of 5.0 mm, and a time step size of $1.0{\times}10^{-5}$ s can be a useful tool to predict the pressure build up due to the hydrogen flame acceleration in the test results. Through the comparison of the simulated results with the test results, we found out that the proposed CFD analysis methodology enables us to predict the peak pressure within an error range of about ${\pm}29%$ for the hydrogen concentration of 19.5%. However, the error ranges of the peak pressure for the hydrogen concentration of 15.4% and 18.6% were about 66% and 51%, respectively. To reduce the error ranges in case of the hydrogen concentration of 15.4% and 18.6%, some uncertainties of the test conditions should be clarified. In addition, an investigation for a possibility of flame extinction in the test results should be performed.

• Nuclear Engineering and Technology
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• 제48권6호
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• pp.1330-1337
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• 2016

A visualization test of the fuel-coolant interaction in the Test for Real cOrium Interaction with water (TROI) test facility was carried out. To experimentally simulate the In-Vessel corium Retention (IVR)- External Reactor Vessel Cooling (ERVC) conditions, prototypic corium was released directly into the coolant water without a free fall in a gas phase before making contact with the coolant. Corium (34.39 kg) consisting of uranium oxide and zirconium oxide with a weight ratio of 8:2 was superheated, and 22.54 kg of the 34.39 kg corium was passed through water contained in a transparent interaction vessel. An image of the corium jet behavior in the coolant was taken by a high-speed camera every millisecond. Thermocouple junctions installed in the vertical direction of the coolant were cut sequentially by the falling corium jet. It was clearly observed that the visualization image of the corium jet taken during the fuel-coolant interaction corresponded with the temperature variations in the direction of the falling melt. The corium penetrated through the coolant, and the jet leading edge velocity was 2.0 m/s. Debris smaller than 1 mm was 15% of the total weight of the debris collected after a fuel-coolant interaction test, and the mass median diameter was 2.9 mm.

• 한국가시화정보학회지
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• 제10권3호
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• pp.30-36
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• 2012

We are planning to construct a scaled-down moderator facility to simulate the CANDU-6 moderator circulation phenomena during steady state operating and accident conditions. In the present work a preliminary experiment using a 1/40 scaled-down moderator tank has been performed to investigate the anticipated problems of the flow visualization and measurement in the planning scaled-down moderator facility. We shortly describe CFD analysis result for the 1/40 scaled-down test model and the flow measurement techniques used for this test facility under isothermal flow conditions. The Particle Image Velocimetry (PIV) method is used to visualize and measure the velocity field of water in a transparent Plexiglas tank. Planar Laser Induced Fluorescence (PLIF) technique is used to evaluate the feasibility of temperature field measurement in the range of $20-40^{\circ}C$ of water temperature using an one-color method.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2003년도 가을 학술논문집
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• pp.509-513
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• 2003

중수로 원전에서 환경으로 방출되는 방사성탄소는 비록 소량이지만 반감기(5730년)가 길고 에너지(0.156MeV)가 높은 방사선을 내기 때문에 각별한 관리가 요구되는 핵종으로 다른 방사성 화합물보다 각별한 관리와 감시가 요구된다. C-14은 원자로 구조의 특성상 경수로에 비해 6배정도 많이 발생하며 방출되는 C-14의 약 90%는 감속재 계통이 차지하고 있고 주로 감속재 상층기체의 퍼지 및 배기 방출을 통해 환경으로 빠져나가게 된다. 본 연구는 발전소 계통 운전 및 중수 누설 등으로 인해 방출되는 C-14을 흡착, 제거할 수 있는 장치 개발에 초점을 맞추었으며 시험 운전결과, C-14 제거 성능이 매우 우수한 것으로 평가되었다.

• 한국가시화정보학회지
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• 제12권1호
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• pp.18-24
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• 2014

The Korea Atomic Energy Research Institute (KAERI) has a scaled-down moderator test program to simulate the CANDU6 moderator circulation phenomena during steady state operation and accident conditions. In the present work a preliminary experiment using a 1/8 scaled-down moderator tank has been performed to identify the potential problems of the flow visualization and measurement in the scaled-down moderator test facility. With a transparent moderator tank model, a velocity field is measured with a Particle Image Velocimetry (PIV) technique under an isothermal state. The flow patterns from the inlet nozzles to the top region of the tank are investigated using PIV for a 1/8 scale moderator tank.