• 한국수소및신에너지학회논문집
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• 제23권3호
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• pp.219-226
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• 2012

A nuclear fusion fuel cycle plant is composed of various subsystems such as a hydrogen isotope storage and delivery system, a tokamak exhaust processing system, and a hydrogen isotope separation system. Korea shares in the construction of the International Thermonuclear Experimental Reactor fuel cycle plant with the EU, Japan and US, and is responsible for the development and supply of the storage and delivery system. We thus present details on the hydrogen isotope process safety. The main safety analysis procedure is to use a hazard and operability study. Nine segments were studied how the plant might deviate from its design purpose. We present a detailed description of the process, examine every part of it to determine how deviations from the design intent can occur and decide whether these deviations can give rise to hazards. We determine possible causes and note protective systems, evaluate the consequences of the deviation, and recommend actions to achieve our safety goal.

• Nuclear Engineering and Technology
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• 제54권3호
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• pp.860-866
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• 2022

An accurate prediction of the pressure drop along the flow paths is crucial in the design of advanced passive systems cooled by heavy liquid metal coolants. To date, a generic pressure drop correlation over spacer grids by Rehme has been applied extensively, which was obtained from substantial experimental data with multiple types of components. However, a few experimental studies have reported that the correlation may give large discrepancies. To provide a more reliable correlation for ring-type spacer grids, the current numerical study aims at figuring out the most critical factor among four hypothetical parameters, namely the flow area blockage ratio, number of fuel rods, type of fluid, and thickness of the spacer grid in the flow direction. Through a set of computational fluid dynamics simulations, we observed that the flow area blockage ratio dominantly influences the pressure loss characteristics, and thus its dependence should be more emphasized, whereas the other parameters have little impact. Hence, we suggest a new correlation for the drag coefficient as C_B = C_ν,m/ε^2.7, where C_ν,m is formulated by a nonlinear fit of simulation data such that C_ν,m = -11.33 ln(0.02 ln(Re_b)).

• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 춘계학술대회
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• pp.1833-1838
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• 2003

In the nuclear power plant, emergency core coolant system(ECCS) is furnished at reactor coolant system(RCS) in order to cool down high temperature water in case of emergency. However, in this coolant system, thermal stratification phenomenon can be occurred due to coolant leaking in the check valve. The thermal stratification produces excessive thermal stresses at the pipe wall so as to yield thermal fatigue crack(TFC) accident. In the present study, when the turbulence penetration occurs in the branch piping, the maximum temperature differences of fluid at the pipe cross-sections of the T-branch with thermal stratification are examine

• Nuclear Engineering and Technology
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• 제33권4호
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• pp.423-440
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• 2001

For sodium cooling liquid metal reactors during the last 25 years, it was most important to verify the safety of the steam generator, which absolutely requires a water leak detection system with fine sensitivity and response. This study describes the structure and leak classification of the HAMMER (Korea Advanced Liquid Metal Reactor) steam generator, compared with other classifications, and explains the effects of leak development. The requirements and experimental situations for the development of the KALIMER acoustic leak detection system (KADS) which detects micro leaks, not intermediate leaks, are introduced. We proposed four frequency bands, 1∼8kHz, 8∼20kHz, 20∼40kHz and 40∼200kHz, split effectively for analyzing the detected acoustic leak signals obtained from the sodium-water reaction model or water model in the mock-up system.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2003년도 학술회의 논문집 정보 및 제어부문 B
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• pp.567-570
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• 2003

This paper deals with the design, implementation, and test of a CRCS for nuclear power plants. Although CRCS is still classified into non-safety class, much attention on its reliability issue has been given so far because of its importance for the stable operation of the reactor in the plant. In terms of technical aspects, our system adopts a full-duplex configuration to enhance reliability in contrast to the existing systems that are all simplex.

• 한국IT서비스학회지
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• 제8권2호
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• pp.147-156
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• 2009

As the result of the rapid development of IT technology, an on-line diagnostic system using the field bus communication network coupled with a smart sensor module will be widely used at the nuclear power plant in the near future. The smart sensor system is very useful for the prompt understanding of abnormal state of the key equipments installed in the nuclear power plant. In this paper, it is assumed that a smart sensor system based on the fieldbus communication network for the surveillance and diagnostics of safety-critical equipments will be installed in the harsh-environment of the nuclear power plant. It means that the key components of fieldbus communication system including microprocessor, FPGA, and ASIC devices, are to be installed in the RPV (reactor pressure vessel) and the RCS (reactor coolant system) area, which is the area of a high dose-rate gamma irradiation fields. Gamma radiation constraints for the DBA (design basis accident) qualification of the RTD sensor installed in the harsh environment of nuclear power plant, are typically on the order of 4 kGy/h. In order to use a field bus communication network as an ad-hoc diagnostics sensor network in the vicinity of the RCS pump area of the nuclear power plant, the robust survivability of IT-based micro-electronic components in such intense gamma-radiation fields therefore should be verified. An intelligent CCD camera system, which are composed of advanced micro-electronics devices based on IT technology, have been gamma irradiated at the dose rate of about 4.2kGy/h during an hour UP to a total dose of 4kGy. The degradation performance of the gamma irradiated CCD camera system is explained.

• 한국화재소방학회논문지
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• 제33권6호
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• pp.105-113
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• 2019

본 연구에서는 원자력발전소용 비안전등급케이블 2종(보안상 A사, B사로 지칭키로 함)을 대상으로 가속열화 기간에 따른 독성 특성을 분석하였다. NES 713 시험장비 및 규격에 의거하여 비노화, 20년, 40년으로 가속열화한 케이블에 대해 피복재 및 절연재로 구분하여 시험하였다. 시험결과 20년, 40년 가속열화 케이블의 독성지수가 비노화 케이블의 독성지수보다 높았으며 A, B사 케이블 공통적으로 20년 가속열화 케이블에서의 독성지수가 높게 산출되었다. 이는 급증한 일산화탄소의 방출량과 더불어 염화수소와 브롬화수소의 할로겐계 가스 방출량이 높은 것이 주된 원인으로 파악되었다. 또한, 피복재와 절연재를 구분하여 분석 시 A, B사 케이블 일부 피복재의 독성지수가 높게 나타났다. 또한, 피복재와 절연재의 독성지수를 구체적으로 분석하고자 미국 국방성 규격인 MIL-DTL을 적용하여 독성지수 허용치 초과 여부를 판단하였으며, 이 중 절연재의 경우 상당량 초과하여 방출되는 결과를 보였다.

• 조명전기설비학회논문지
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• 제25권4호
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• pp.71-78
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• 2011

This paper contains the selection of operation parameters for protection and control of steam turbine in a Korea Standard Type Nuclear Power Plant. The safety of nuclear reactor must be ensured which generates nuclear energy and produces steam. Also, the safety of turbine, which consume the nuclear energy as a core machine, must be ensured. For the purpose of this, we describe how the operating parameters were selected, reviewed, implemented into the operator console and finally put into actual operation of the system.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.833-837
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• 1998

This paper presents preliminary findings regarding a modeling framework under development for use in a multi-attribute decision model for advanced emergency operating procedures(EOPs). This model provides a means for optimal decision making strategy for advanced emergency operating procedures conceptualizing the dynamic coordination of responsibilities and information in the human system interactions with advanced reactor systems. For the purpose of evaluation of the applicability of this modeling framework, an empirical case study for a post-cooldown strategy during an steam generator tube rupture (SGTR) accident was carried out. As a result, it was found empirically that the multi-attribute decision model is a useful tool for establishing advanced EOPs that reduce the operator's cognitive and decision making burden during the accident mitigation process.

• 해양환경안전학회지
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• 제4권1호
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• pp.49-56
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• 1998

The Microfiltration and Ultrafiltration were used to treat effluent of secondary municipal wastewater treatment system(Sequencing Batch Reactor). The cross-flow hollow fiber, UF 500,000(NMWC) and MF 0.65$\mu$ membrane were selected as suitable membrane. Short term and long term fouling effect were measured as a factor of flux decrease and the fouling removal effect of mixing air bubble in the penetrant was studied. The removal of anionic sulfactants before and after formation of micelle with several kinds of oil were checked. The test results show that removal of TOC was 70~80%, TN 28% and TP 16%. The decrease of flux due to fouling were 85%(UF) and 90%(MF) after running of 100hrs. The removal of anionic sulfactants were 60~70% notwithstanding micelle or not.