• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 2005년도 춘계학술발표회
• /
• pp.1243-1244
• /
• 2005

• 전력전자학회논문지
• /
• 제22권1호
• /
• pp.9-17
• /
• 2017

A parallel control algorithm of thyristor dual-converter power system for the DC power supply of railway is proposed. The circulating current and current imbalance generated during parallel operation can be limited to control the output voltage of each power system by using the proposed parallel control algorithm. The proposed control algorithm can also eliminate output current sensor to achieve the same output response without additional costs. The validity of the proposed algorithm is verified through simulation and experiment.

• Nuclear Engineering and Technology
• /
• 제42권1호
• /
• pp.47-54
• /
• 2010

The SuperCritical Water-cooled Reactor (SCWR) pursues high power density to reduce its capital cost. The fast spectrum SCWR, called a super fast reactor, can be designed with a higher power density than thermal spectrum SCWR. The mechanism of increasing the average power density of the super fast reactor is studied theoretically and numerically. Some key parameters affecting the average power density, including fuel pin outer diameter, fuel pitch, power peaking factor, and the fraction of seed assemblies, are analyzed and optimized to achieve a more compact core. Based on those sensitivity analyses, a compact super fast reactor is successfully designed with an average power density of 294.8 W/$cm^3$. The core characteristics are analyzed by using three-dimensional neutronics/thermal-hydraulics coupling method. Numerical results show that all of the design criteria and goals are satisfied.

• 한국초전도ㆍ저온공학회논문지
• /
• 제25권2호
• /
• pp.19-24
• /
• 2023

Recently, interest in renewable energy such as solar and wind power has increased as an alternative to fossil fuels. Renewable energy sources such as large wind farms require long-distance power transmission because they are located inland or offshore, far from the city where power is required. High-Temperature Superconducting (HTS) power cables have more than 5 times the transmission capacity and less than one-tenth the transmission loss compared to the existing cables of the same size, enabling large-capacity transmission at low voltage. For commercialization of HTS power cables, unmanned operation and long-distance cooling technology of several kilometers is essential, and pressure drop characteristic is important. The cryostat's spiral corrugation tube is easier to bend, but unlike the round tube, the pressure drop cannot be calculated using the Moody chart. In addition, it is more difficult to predict the pressure drop characteristics due to the irregular surface roughness of the binder wound around the cable core. In this paper, a CFD model of a spiral corrugation tube with a core was designed by referring to the water experiments from previous studies. In the four cases geometry, when the surface roughness of the core was 10mm, most errors were 15% and the maximum errors were 23%. These results will be used as a reference for the design of long-distance HTS power cables.

• 한국안전학회지
• /
• 제29권6호
• /
• pp.166-171
• /
• 2014

Nuclear power plants are equipped with the reactor trip system (RTS) and the engineered safety features actuation system (ESFAS) to improve safety on the normal operation. In the event of the design basis accident (DBA), a various of post accident monitor(PAM)systems support to provide important details (e.g. Containment pressure, temperature and pressure of reactor cooling system and core exit temperature) to determine action of main control room (MCR). Operator should be immediately activated for the accident mitigation with the information. Especially, core exit temperature is a critical parameter because the operating mode converts from normal mode to emergency mode when the temperature of core exit reaches $649^{\circ}C$. In this study, uncertainty which was caused by exterior environment, characteristic of thermocouple/connector and accuracy of calibrator/indicator was evaluated in accordance with ANSI-ISA 67.04. The square root of the sum of square (SRSS) methodology for combining uncertainty terms that are random and independent was used in the synthesis. Every uncertainty that may exist in the hardware which is used to measure the core exit temperature was conservatively applied and the associative relation between the elements of uncertainty was considered simultaneously. As a result of uncertainty evaluation, the channel statistical allowance (CSA) of single channel of core exit temperature was +1.042%Span. The range of uncertainty, -0.35%Span ($-4.05^{\circ}C$) ~ +2.08%Span($24.25^{\circ}C$), was obtained as the operating criteria of core exit temperature.

• 대한기계학회논문집B
• /
• 제36권10호
• /
• pp.1043-1047
• /
• 2012

유럽지역으로의 원전 수출을 위해서는 유럽의 원전 인허가요건을 충족시켜야 하며, 이에 따르면 원전의 중대사고 대처설비로 통상 Core Catcher로 불리는 노외 노심 용융물 냉각설비를 갖출 것을 권장하고 있다. 이에 따라 본 논문에서는 노심 용융물 직접냉각방식과 간접냉각방식에 대해 각각의 개념 안의 장/단점을 비교, 검토하였으며, 그 결과 직접냉각방식은 냉각효율 측면에서, 간접냉각방식은 중대사고 사고관리 측면에서 각각 우위를 보였다.

• Journal of Magnetics
• /
• 제16권2호
• /
• pp.124-128
• /
• 2011

This study estimated experimentally the loss distribution caused by magnetic friction in magnetic parts of a superconductor flywheel energy storage system (SFES) to obtain information for the design of high efficiency SFES. Through the spin down experiment using the manufactured vertical shaft type SFES with a journal type superconductor magnetic bearing (SMB), the coefficients of friction by the SMB, the stator core of permanent magnet synchronous motor/generator (PMSM/G), and the leakage flux of the metal parts were calculated. The coefficients of friction by the stator core of PMSM/G in case of using Si-steel and an amorphous core were calculated. The energy loss by magnetic friction in the stator core of PMSM/G was much larger than that in the other parts. The level of friction loss could be reduced dramatically using an amorphous core. Energy loss by the leakage magnetic field was small. On the other hand, the energy loss could be increased under other conditions according to the type of metal nearby the leakage magnetic fields. In manufactured SFES, the rotational loss by the amorphous core was approximately 2 times the loss of the superconductor and leakage. Moreover, the rotational loss by the Si-steel core is approximately 3~3.5 times the loss of superconductor and leakage.

• Nuclear Engineering and Technology
• /
• 제50권5호
• /
• pp.673-682
• /
• 2018

One of the most widely used methods to estimate core damage during a nuclear power plant accident is containment radiation measurement. The evolution of severe accidents is extremely complex, leading to uncertainty in the containment dose rate (CDR). Therefore, it is difficult to accurately determine core damage. This study proposes to conduct uncertainty analysis of CDR for core damage assessment. First, based on source term estimation, the Monte Carlo (MC) and point-kernel integration methods were used to estimate the probability density function of the CDR under different extents of core damage in accident scenarios with late containment failure. Second, the results were verified by comparing the results of both methods. The point-kernel integration method results were more dispersed than the MC results, and the MC method was used for both quantitative and qualitative analyses. Quantitative analysis indicated a linear relationship, rather than the expected proportional relationship, between the CDR and core damage fraction. The CDR distribution obeyed a logarithmic normal distribution in accidents with a small break in containment, but not in accidents with a large break in containment. A possible application of our analysis is a real-time core damage estimation program based on the CDR.

• Nuclear Engineering and Technology
• /
• 제46권4호
• /
• pp.513-520
• /
• 2014

In this paper, three types of operational and industrial absorbers used at research reactors, including Ag-In-Cd alloy, $B_4C$, and Hf are selected for sensitivity analyses. Their integral effects on the main neutronic core parameters important to safety issues are investigated. These parameters are core excess reactivity, shutdown margin, total reactivity worth of control rods, thermal neutron flux, power density distribution, and Power Peaking Factor (PPF). The IAEA 10 MW benchmark core is selected as the case study to verify calculations. A two-dimensional, three-group diffusion model is selected for core calculations. The well-known WIMS-D4 and CITATION reactor codes are used to carry out these calculations. It is found that the largest shutdown margin is gained using the $B_4C$; also the lowest PPF is gained using the Ag-In-Cd alloy. The maximum point power densities belong to the inside fuel regions surrounding the central flux trap (irradiation position), surrounded by control fuel elements, and the peripheral fuel elements beside the graphite reflectors. The greatest and least fluctuation of the point power densities are gained by using $B_4C$ and Ag-In-Cd alloy, respectively.

• 대한전기학회:학술대회논문집
• /
• 대한전기학회 2003년도 하계학술대회 논문집 B
• /
• pp.825-827
• /
• 2003

There are many problems when air core reactor is used. Especially, measure instruments are influenced from magnetic field. This paper shows the research results for the magnetic field shielding on air core reactor.