• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.17-22
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• 2016

Alloy 600 widely used in nuclear power plant is susceptible to primary water stress corrosion cracking (PWSCC). It is important to prioritize the inspection of Alloy 600 components using PWSCC susceptibility index. Plant-specific model for the susceptibility index was reviewed. The normalized PWSCC susceptibility index to a reference value is suggested and applied. The result was found to be reasonable.

• Nuclear Engineering and Technology
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• v.25 no.2
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• pp.343-352
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• 1993

다목적연구로(KMRR)는 일반 발전용 원자로와는 매우 다른 특성을 가지고 있으며, 설계 개념 또한 특이하다. 위와 같은 특이한 설계 특성을 파악하기 위하여 열수력 실험을 수행하였으며 시운전 시험도 설계 개념의 입증에 중점을 두고 수행될 예정이다. 실증실험은 크게 설계 자료 생산을 위한 실험, 기기 설계 검증 시험, 시운전 성능 시험으로 나눌 수 있다. 설계 자료 생산을 위한 실험으로 핵연료의 열수력학적 특성을 규명하는 실험, 우회 유동에 의한 노심 출구 냉각수 상승 억제를 입증 또는 해석하기 위한 자료 생산용 실험 등이 이루어졌다. 기기 설계 검증 시험으로는 Pump 특성 시험, Flap valve 특성 시험 등을 들 수 있다. 또한, 시운전 성능 시험으로는 설계 개념을 입증하기 위한 여러 시험들이 행해질 예정이다. 이러한 실험들을 통하여 설계에 필요한 많은 자료들이 생산되었고, 시운전 시험을 통하여 설계를 검증하고 실제 운전에 필요한 많은 자료를 얻을 수 있으리라 기대된다. 본 기고를 통하여 이러한 실험의 중요성 및 내용에 대해 간략하게 기술하고자 한다.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.28 no.2
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• pp.167-173
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• 2004

Fouling of heat exchangers is generated by water-borne deposits, commonly known as foulants including particulate matter from the air, migrated corrosion produces; silt, clays, and sand suspended in water; organic contaminants; and boron based deposits in plants. The fouling is known to interfere with normal flow characteristics and reduce thermal efficiencies of heat exchangers. This paper describes the fouling analysis technique developed in this study which can analyze the thermal performance for heat exchangers and estimate the future fouling variations. To develop the fouling analysis technique fur heat exchangers, fouling factor was introduced based on the ASME O&M codes and TEMA standards. For the purpose or verifying the fouling analysis technique, the routing analyses were performed for four heat exchangers in several nuclear power plants; two residual heat removal heat exchangers of the residual heat removal system and two component cooling water heat exchangers of the component cooling water system.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.28 no.11
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• pp.1384-1389
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• 2004

As operating time of heat exchangers progresses, fouling generated by water-borne deposits and the number of plugged tubes increase and thermal performance decreases. Both fouling and tube plugging are known to interfere with normal flow characteristics and to reduce thermal efficiencies of heat exchangers. The heat exchangers of domestic nuclear power plants have been analyzed in terms of the heat flux and heat transfer coefficient at test conditions as a means of heat exchanger management. Except for the fouling level generated in operation of heat exchangers, also, all of the tubes of heat exchangers have been replaced when the number of plugged tubes exceeds the plugging criteria based on design performance sheet. This paper describes the plugging margin evaluation mettled reflected the fouling of shell-and-tube heat exchangers, which can evaluate the thermal performance for heat exchangers, estimate the future fouling variations, and reflect the current fouling level. To identify the effectiveness of the developed method, the fouling and plugging margin evaluations were performed for a component cooling heat exchanger in a nuclear power plant.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.17 no.2
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• pp.90-100
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• 2021

Molten corium-concrete interaction (MCCI) is one of the most important phenomena that can lead to the potential hazard of late containment failure due to basemat penetration during a severe accident. In this study, MCCI analytical models of the CORQUENCH code were prepared through verification calculations of several experiments, which had been performed using concrete types similar to those of the calandria vault floor in CANDU-6 plants. The behaviors of thermal-hydraulic variables related to MCCI phenomena were analyzed under the conditions of dry floor and water flooding during the severe accident stemming from a hypothetic station blackout. Uncertainty analyses on the ablation depth were also carried out. It was estimated that the concrete ablation was not interrupted due to the continuous MCCI process under the dry condition but was terminated within 24 hours under the water flooding condition. It was confirmed that the water flooding as a mitigating action was effective to achieve the quenching and thermal stabilization of the melt discharged from the calandria vessel, showing that the present models are capable of reasonably simulating MCCI phenomena in CANDU-6 plants. This study is expected to provide the technical bases to the accident management strategy during the late-phase severe accidents.

• Journal of Energy Engineering
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• v.8 no.4
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• pp.525-532
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• 1999

The standardized Korea Next Generation Reactor (KNGR) NSSS has developed in the basis of the ABB-CE System 80+ design concept. In this study, several regulatory requirements for the KNGR shutdown cooling system (SCS) operation are investigated. The purpose of this study is to establish the technical self-reliance for SCS design by supporting fundamental data such as SDCHX effective area and reactor CCW flow rate. Thermal power of KNGR would be increased to about 4,000 $MW_{th}$ in comparison with thermal power 2.825 $MW_{th}$ of UCN 3&4, therefore, SCS design data shall b recalculated by using the KDESCENT Code, which could evaluate cooling capability of SCS. It is found that SCS minimum flow rate is able to remove the primary sensible heat. Reviewing the major components such as heat exchanger, pump, value, and operating procedure, it is concluded as follows.

• Proceedings of the KSME Conference
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• 2001.11b
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• pp.505-510
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• 2001

Advanced Pressurized Reactor 1400(APR1400), which is a standard evolutionary advanced light water reactor(ALWR), has been developed from 1992 as one of long-term Government Project(G-7). The APR-1400 is designed to operate at the rated output of 4000MWt to produce an electric power output of around 1450MWe. Due to the increased electric power, In Nuclear Power plant huge quantities of heat are generated in the thermo-dynamic process used for producing electrical energy. So, There is considerationly additional cooling, Heat transfer area and increased cooling water of Heat Exchanger which take care of the different smaller cooling duties within the nuclear power plant. We review applying to PRE instead of Shell-and-Tube Heat exchanger. In this paper, we describe the major design features of PRE, Comparison between a PHE and a Shell-and-Tube Heat Exchanger, and then Applicability of Plate Heat Exchanger in Nuclear Power Plant Component Cooling water systems.

• Journal of Energy Engineering
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• v.23 no.3
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• pp.186-190
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• 2014

Nuclear Power Plants require the control ability of chemical condition (pH) because pH control during transient accident such as LOCA makes an able the fission product removal capability to be maintained, stress corrosion cracking of stainless steel equipment to be prevented and the production of hydrogen by aluminum and zinc to be minimized. An NPP is designed to control the pH of containment spray and sump coolant using the spray additives 30% NaOH in the event of loss of coolant accident. In this paper, the pH of sump coolant of an NPP during LOCA was analyzed and the fission products removal constant and decontamination factor were calculated according to Standard Review Plan 6.5.2 related to spray chemical conditions of pH. The calculated pH value of recirculation mode using the computer code corresponds to 8.09~9.67, which meets the chemical environment regulation requirements. The fission product removal capability caused by containment spray system is performed to provide input to radiation analysis.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.2
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• pp.20-27
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• 2010

In order to prevent the corrosion of component contacted primary water designed alloy 600 material in the nuclear power plant. But the primary water stress corrosion cracking(PWSCC) of alloy 600 and weld area occurs continuously due to the residual stress. The leakage accident resulted from PWSCC in the drain nozzle of the steam generator of domestic power plants. Heater sleeves of the pressurizer are welded with alloy 600 weld material and therefore exposed to the primary water environment. PWSCC occurred in heater sleeve material and weld area of many foreign power plants. The current issue of domestic nuclear power plants are consequently concentrated to PWSCC of similar material. In order to improve the detection and the sizing of the PWSCC in the welding sleeve of the pressurizer, the automatic UT system and multi-directions probe sets have been developed. The experimental studies have been performed using the mock-up block containing artificial reflectors(ID connected EDM notch) and semi-artificial cracks made from thermal fatigue. The automatic UT System is applied in the detection and the length sizing of the ID/OD on the tube and the J-groove weld area of the artificial reflectors and results of the detection and the sizing are compared respectively. Also, the developed automatic UT system is successfully accomplished to inspect the heater sleeve and the J-groove weld area on the pressurizer for the detection of PWSCC.

• Proceedings of the Korean Vacuum Society Conference
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• 2012.02a
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• pp.539-539
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• 2012

전북대학교 고온플라즈마 응용연구센터는 교육과학기술부 기초연구사업 중 고가연구장비 구축사업의 일환으로, 고 엔탈피, 초음속 유동 환경을 모사하여, 항공우주, 군사기기, 핵융합 분야 등의 고온 재료 개발을 위한 기초 연구 장치로써, 0.4MW급 플라즈마 풍동 장치를 구축하고 있다. 0.4MW 플라즈마 풍동 장치의 플라즈마 발생부는 DC 전원 공급장치와 디스크 형태의 양극과 음극 사이에 동일 형태의 간극을 삽입한 0.4MW급 분절형 아크 플라즈마 토치로 구성되었으며, 토치에서 발생된 아크 플라즈마는 노즐을 통과하며 마하 2~4의 초음속을 나타내도록 설계 제작되었다. 시험 챔버는 노즐에서 나온 초음속 플라즈마의 특성 및 재료 시험을 위한 3차원 이송식 기판이 장착되어 있으며, 고 엔탈피 유동을 관측하기 위한 광학창을 구비하였다. 시험 챔버 하류에는 유동 안정을 위한 디퓨저(diffuser)가 설치되어 있으며, 디퓨저(diffuser)로부터 배출되는 고온가스는 열교환기를 통해 냉각된 후 진공펌프를 통해 대기로 배출되게 된다. 장치의 압력조절을 위하여 $1,000m^3/min$의 용량의 진공펌프 시스템이 설치될 예정이며 가스공급장치, 냉각수 공급장치, 디퓨져, 열교환기는 1MW급 용량으로 설계 제작되었다. 본 장치는 400kW의 전원 공급, 15 g/s의 공기유량 주입 시 약 13 MJ/kg의 고엔탈피를 가진, mach 2~4의 초음속 유동을 나타내는 것을 특징으로 한다.