• Nuclear Engineering and Technology
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• 제44권7호
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• pp.807-816
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• 2012

The Three Mile Island Unit 2 (TMI-2) accident has been studied extensively, as part of both post-accident technical assessment and follow-up computer code calculations. The models used in computer codes for severe accidents have improved significantly over the years due to better understanding. It was decided to reanalyze the severe accident scenario using current state of the art codes and methodologies. This reanalysis was adopted as a part of the joint standard problem exercise for the Atomic Energy Regulatory Board (AERB) - United States Regulatory Commission (USNRC) bilateral safety meet. The accident scenario was divided into four phases for analysis viz., Phase 1 covers from the accident initiation to the shutdown of the last Reactor Coolant Pumps (RCPs) (0 to 100 min), Phase 2 covers initial fuel heat up and core degradation (100 to 174 min), Phase 3 is the period of recovery of the core water level by operating the reactor coolant pump, and the core reheat that followed (174 to 200 min) and Phase 4 covers refilling of the core by high pressure injection (200 to 300 min). The base case analysis was carried out for all four phases. The majority of the predicted parameters are in good agreement with the observed data. However, some parameters have significant deviations compared to the observed data. These discrepancies have arisen from uncertainties in boundary conditions, such as makeup flow, flow during the RCP 2B transient (Phase 3), models used in the code, the adopted nodalisation schemes, etc. In view of this, uncertainty and sensitivity analyses are carried out using simulation based techniques. The paper deals with uncertainty and sensitivity analyses carried out for the first three phases of the accident scenario.

• 대한핵의학회:학술대회논문집
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• 대한핵의학회 2001년도 제40차 춘계학술대회 및 연수교육
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• pp.66-75
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• 2001

Although diverse mechanisms are involved in multidrug resistance for chemotherapeutic drugs, the development of cellular P-glycoprotein(Pgp) and multidrug-resistance associated protein (MRP) are important factors in the chemotherapy failure to cancer. Various detection assays provide information about the presence of drug efflux pumps at the mRNA and protein levels. However these methods do not yield information about dynamic function of Pgp and MRP un vivo. Single photon emission tomography (SPECT) and positron emission tomography (PET) are available for the detection of Pgp and MRP-mediated transport. $^{99m}Tc$-sestaMIBl and other $^{99m}Tc$-radiopharmaceuticals are substrates for Pgp and MRP, and have been used in clinical studies for tumor imaging, and to visualize blockade of Pgp-mediated transport after modulation of Pgp pump. Colchicine, verapamil and daunorubicin labeled with $^{11}C$ have been evaluated for the quantification of Pgp-mediated transport with PET in vivo and reported to be feasible substrates with which to image Pgp function in tumors. Leukotrienes are specific substrates for MRP and N-$[^{11}C]$acetyl-leukotriene E4 provides an opportunity to study MRP function non-invasively in vivo. Results obtained from recent publications are reviewed to confirm the feasibility of using SPECT and PET to study the functionality of MDR transporters in vivo.

• 에너지공학
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• 제11권2호
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• pp.106-113
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• 2002

급격히 증가하는 에너지 수요에 비해, 현재 주에너지원인 화석에너지나 원자력은 지구 온난화 및 핵폐기물 등의 환경오염 문제로 인해 그 공급을 계속적으로 증가시키는 데에는 많은 제약을 받고 있다. 이러한 에너지 공급의 한계를 극복하기 위해서는 풍부한 자연에너지를 활용할 필요가 있는데, 최근의 열펌프 기술의 발전과 더불어 하천수·해수 하수처리수 등의 온도차에너지에 대한 미활용에너지의 활용지술의 기대가 크게 증가되고 있다. 따라서 미활용에너지의 활용을 위한 기초자료를 확보하기 위하여 열원별·지역별 미활용에너지의 부존량 및 이용가능량을 조사·산정하고, 미활용에너지의 이용가능성에 대하여 평가하였다.

• Nuclear Engineering and Technology
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• 제33권3호
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• pp.339-348
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• 2001

NSSS Control Systems automatically mitigate transient conditions and leads to a stable plant condition without operator actions when a transient occurs during normal power operation. In this paper, the function and performance of NSSS control systems were examined and evaluated by comparing the predicted results with the measured data for the selected events. Loss of a Main Feedwater Pump and Load Rejection to House Load Operation events were selected for the evaluation among the transient tests peformed during the Power Ascension Test (PAT) of UCN unit 4. The overall schematic control actions of NSSS control systems can be evaluated easily through the observation of these two typical events. The selected events were analyzed by the KISPAC computer code[l] which had been used in developing the control logic and determining the control setpoints during the plant design. Additionally, the performance of FWCS during low power operation was evaluated. The result of evaluation showed that the NSSS control systems were designed properly and the performance of the NSSS control systems was excellent and also the computer code had a good prediction capability.

• Nuclear Engineering and Technology
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• 제52권6호
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• pp.1131-1136
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• 2020

Xenon behaves differently in molten salt reactors (MSRs) compared to solid fuel reactors. This behavior needs exploring due to the large reactivity effect of the ¹³⁵Xe isotope, given the current interest in MSR power plant development for commercial deployment. This paper focuses on select topics in xenon transport, reviews relevant past works, and proposes specific research questions to advance the state of the art in each of the focus areas. Specifically, the paper discusses the issue of xenon solubility in MSRs, the behavior of particulates circulating in MSR fuel salt and its influence on the xenon transport, the possibility of ionization of xenon atoms which changes its effective size and thus affects its mass transport, and finally the issue of circulating void fraction and how it is measured. This work presents specific recommendations for MSR designers to research the limits of Henry's law validity, circulating particulate scrubbers, validity of mass transport coefficients in high radiation fields, and the effects of pump speed on circulating void fraction.

• 한국유체기계학회 논문집
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• 제17권6호
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• pp.84-88
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• 2014

A long term passive cooling system is considered as the most important safety feature for the nuclear design after the Fukushima Daiichi nuclear power plant accident in 2011. The conventional active pump driven safety systems are not available during a station Black Out (SBO) accident. The current design requirement on cooling time of the Passive Auxiliarly Feedwater System (PAFS) is about 8 hours only. To meet the 72 hours cooling time, the pool capacity of cooling water tank should be increased as much as 3~4 times larger than that of current water cooling tank. In order to extend the cooling time for 72 hours, a new passive air-water combined cooling system is proposed. This paper provides the feasibility of the combined passive air-water cooling system. The current pool capacity of water cooling system is preserved, and the cooling capability is extended by an additional air cooler.

• 대한기계학회논문집
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• 제16권11호
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• pp.2126-2135
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• 1992

본 연구에서는 국내 원자력 발전소중 영광 3,4호기의 설계자료를 토대로 1/6 크기로 축소한 모델실험을 통해서 공기흡입이 발생하는 임계수위를 결정하는 상관식을 개발하였으며 또한 공기흡입구를 reducer type으로 개선함으로써 공기흡입을 방지할 수 있음을 밝혔다.

• 한국안전학회지
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• 제31권6호
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• pp.129-134
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• 2016

Standard design of APR+(advanced power reactor plus) was certified at 2014 by Korea regulatory body. Based on the experience gained from OPR1000 and APR1400, the APR1400 was being developed as a 1,500MWe class reactor using Korean technologies for design code, reactor coolant pump, and man-machine interface system. APR+ has been basically designed to have the seismic design basis of safe shutdown earthquake (SSE) 0.3g, a 4-train safety concept based on N+2 design philosophy, and a passive auxiliary feedwater system (PAFS). Also, safety issues on the Fukushima-type accidents have been extensively reviewed and applied to enhance APR+ safety. APR+ provides higher reliability and safety against tsunami and earthquake. The purpose of this paper is to implement probabilistic safety assessment considering these design features and to analyze sensitivity of core damage frequency for large loss of coolant accident of APR+.

• Nuclear Engineering and Technology
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• 제27권5호
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• pp.687-700
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• 1995

제안된 CANDU-9 원자로의 열수력 과도변화상태가 해석되었으며 주요한 몇개의 과도변화가 열수송 계통의 설계요건을 만족시키는지에 대해 평가되었다. 열수송계통의 과도변화시 핵연료의 건전성과 계통압력상승의 제한 측면에서 분석된 본 해석결과에 따라서 제안된 열수송계통형상과 열수송계통기기의 예비 크기가 확정 및 검증되었다. AECB R-77 요구조건에 대한 CANDU-9 원자로의 만족여부를 평가하였다. 해석결과, 각 과도변화시 원자로 모관의 고압첨두치가 ASME코드의 요구조건에 따른 허용범주내에 있었으며 핵연료의 건전성이 확인되었다. 원자로 가동운전시 제안된 CANDU-9 원자로의 고유적인 핵연료채널을 통한 역류현상을 규명하기 위하여 한개의 펌프가 시동될때의 과도변화현상을 해석하였다.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2005년도 추계학술대회논문집
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• pp.394-399
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• 2005

A case history is presented pertaining to high frequency piping vibration and noise caused by globe valve in the spent fuel pool cooling system of nuclear power plant. Frequency analyses were performed on the system to diagnose the problem and develop a solution to reduce the piping vibration and noise. The source of the high frequency and noise energy was traced to the globe valve located immediately downstream of the centrifugal pump by performing valve throttling test. Measurements of vibration and noise are presented to show that the high frequency vibration and noise amplitude was dependent upon the valve disc position and flow rate. Strouhal vortex shedding frequencies were generated at the exit of the globe valve which exited structural resonance of valve disc and amplified the high frequency vibration and noise. The problem was identified as an interaction between the flow inside globe valve and the valve disc structure. Attempts to reduce the vibration and noise amplitudes of the piping system were successfully achieved by the modification of guide-disc diameter and disc-edge figure The valve disc was replaced by an alternative to eliminate the source of the harmful high frequency vibration and noise.