• Nuclear Engineering and Technology
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• 제20권4호
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• pp.253-264
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• 1988

본 논문에서는 불란서에서 건설한 900 MWe급 가압경수형 원자로의 중성자 잡음해석 결과를 제시하였다. 중성자 잡음해석이란 노심내의 반응도 변화 및 노심의 수평운동으로 인한 노외검출기 신호의 변화를 해석하는 기법을 의미한다 이러한 방법은 Deterministic Dynamic Testing 기법중에서도 발전소의 정상운전 조건을 유지시키며 기존의 발전소 계측설비를 이용할 수 있다는 장점을 지니고 있다. 본 논문에 사용된 잡음신호는 울진 1호기 원자로의 시운전 시험기간에 구하였으며 이를 통계적 기술함수인 에너지 밀도함수(PSD), 검출기간의 상관함수 (CF)및 위상차(Phase Difference)로 나타내었다. 실험결과, 원자로 용기내의 냉각수 흐름 및 압력맥동 등에 의해 유도되는 Core Support Barrel(CSB)의 진동 주파수가 8Hz 근처임을 규명하였다.

• 한국유체기계학회 논문집
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• 제17권1호
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• pp.5-11
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• 2014

The purpose of this paper is to investigate an analytical approach for opening performance evaluation of the nuclear pressure safety valve based on standard codes such as ASME or KEPIC. It is well-known that safety valve is considered as one of pressure relief valves for protecting a boiler or pressure vessel from exceeding the maximum allowable working pressure. When pressure in a container reaches its set pressure, the safety valve commences discharging the internal fluid by a sudden opening called as popping. Safety valve is usually evaluated by set pressure, full open, blow-down, leakage and flow capacity. The test procedure and technical requirement for performance evaluation is described in international code of ASME code such as BPVC. The opening characteristics of steam safety valve can be analyzed by computational fluid dynamics (CFD) and steam shaft dynamics. First, the flow analysis along opening process is simulated by running the CFD models of the ten types of opening steps from 0 to 100%. As a analysis result, the various CFD outputs of flow pattern, pressure, forces on the disc and mass flow at each simulation step is demonstrated. The lift force is calculated by using the forces applied on disc from static pressure and secondary flow. And, the effect of huddle chamber or control chamber is studied by dynamic analysis based on CFD simulation results such as lift force. As a result, dynamics analysis shows opening features according to the sizes of control chamber.

• Nuclear Engineering and Technology
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• 제24권3호
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• pp.274-284
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• 1992

가압경수로 최적 열수력 분석용 전산코드인 CATHRE의 모델 평가를 위하여 가압경수로의 가상 냉각재 상실사고시 원자로 용기내의 유동현상을 모의한 1/15축소의 CREARE 실험을 모의 계산하였다. 이 실험에서 주요변수들은 비상노심 탱각재 주입량과 아냉정도 그리고 계통압력 및 노심에서 발생되는 증기유량이지만. 본 연구에서는 우선 Downcomer에서 역방향유동의 정성적 분석에 촛점을 맞추었다. 모의 계산 결과와 실험 결과를 비교할 때 정량적인 값 뿐 아니라 변화의 경향에서도 차이가 나타난 것은 주로 적절하지 못한 일부의 수치해석 모델과 상간의 계면마찰 때문으로 판단된다. 따라서 매개변수적 민감도 분석을 통하여 CATHARE 전산코드의‘VOLUME’에 접한 접합점에서 운동량 보존방정식의 상세연구 혹은 다차원 분석을 통해서 이 경우의 물리적 현상을 보다 현실적으로 나타낼 수 있다는 결론을 얻었다.

• Nuclear Engineering and Technology
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• 제26권2호
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• pp.168-178
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• 1994

사고관리란 사고발생시에 이용가능한 모든 자원, 즉 인원과 설비를 효율적으로 활용함으로써 발전소를 안전상태로 회복시키거나 사고의 피해를 완화시키기 위한 제반 활동을 말한다. 사고관리의 접근방식은 첫째, 후보사고관리방안의 사전 평가, 둘째, 효과적으로 적절한 조치를 수행하게 하는 세부 절차서의 개발, 그리고 셋째, 그러한 조치수행에 필요한 도구와 자원의 준비, 실현 가능한 원전 시스템의 변경등을 포함한다. 사고관리 전략을 평가할 때에는 그 전략의 효율성분만 아니라 부작용, 타당성, 필요한 정보, 기존 절차서와의 양립성 등을 종합적으로 고려하여야 한다. 이 논문의 목적은 여러가지 사고관리 전략을 모델링하고 평가하기위한 체제를 개발하기 위한 의사결정 수목과 영향도의 해석도구를 소개하는 것이다. 이 해석도구와 관련한 여러가지 특징들이 제시되었으며 이 해석도구에 근거하여 세워진 사고관리전략의 평가체제가 간단한 예제문제에 적용되었다.

• 유체기계공업학회:학술대회논문집
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• 유체기계공업학회 2003년도 유체기계 연구개발 발표회 논문집
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• pp.161-166
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• 2003

It is the common features of the integral reactors that the main components of the RCS are installed within the reactor vessel, and so there are no any flow pipes connecting the coolant pumps or steam generators. Due to no any flow pipes, it is impossible to measure the differential pressure at the RCS of the integral reactors, and it also makes impossible measure the flow-rate of the reactor coolant. As a alternative method, the method by the measurement of coolant pump power has been introduced in this study. Up to now, we did not found out a precedent which the coolant pump power is used for the real-time flow-rate calculation at normal operation of the commercial nuclear power plants. The objective of the study is to embody the real-time flow-rate calculation method by the measurement of coolant pump power in an integral reactor. As a result of the study, we could theoretically reason that the capacity-head curve and capacity-shaft power curve around the rated capacity with the high specific-speeded axial flow pumps have each diagonally steep incline but show the similar shape. Also, we could confirm the above theoretical reasoning from the measured result of the pump motor inputs, So, it has been concluded that it is possible to calculate the real-time flow-rate by the measurement of pump motor inputs. In addition, the compensation for a above new method can be made by HBM being now used in the commercial nuclear power plants.

• Nuclear Engineering and Technology
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• 제50권5호
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• pp.627-638
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• 2018

This article describes an integrated Level 1-Level 2 probabilistic risk assessment (PRA) methodology to evaluate the radiological risk during postulated accident scenarios initiated during the decommissioning phase of a typical Mark I containment boiling water reactor. The fuel damage scenarios include those initiated while the reactor is permanently shut down, defueled, and the spent fuel is located into the spent fuel storage pool. This article focuses on the integrated Level 1-Level 2 PRA aspects of the analysis, from the beginning of the accident to the radiological release into the environment. The integrated Level 1-Level 2 decommissioning PRA uses event trees and fault trees that assess the accident progression until and after fuel damage. Detailed deterministic severe accident analyses are performed to support the fault tree/event tree development and to provide source term information for the various pieces of the Level 1-Level 2 model. Source terms information is collected from accidents occurring in both the reactor pressure vessel and the spent fuel pool, including simultaneous accidents. The Level 1-Level 2 PRA model evaluates the temporal and physical changes in plant conditions including consideration of major uncertainties. The goal of this article is to provide a methodology framework to perform a decommissioning Probabilistic Risk Assessment (PRA), and an application to a real case study is provided to show the use of the methodology. Results will be derived from the integrated Level 1-Level 2 decommissioning PSA event tree in terms of fuel damage frequency, large release frequency, and large early release frequency, including uncertainties.

• Nuclear Engineering and Technology
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• 제26권4호
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• pp.493-501
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• 1994

하향 가열 평판에서의 풀비등 임계열유속 실험이 수행되었다. 이는 원자로에서의 노심용융사고 발생시 그 결과를 완화시키는 한 방법으로 고려되고 있는 원자로용기 외부 냉각 (Ex-Vessel Flooding) 개념과 연관된다. 대기압하, 포화상태 물에서 너비가 다른 두 개의 평판 (20mm$\times$200mm및 25mm$\times$200mm)을 이용, -90$^{\circ}$(평형 하향), -88$^{\circ}$, -86$^{\circ}$, -84$^{\circ}$, -60$^{\circ}$와 -40$^{\circ}$의 경사 가도에 대한 임계열유속이 측정되었다. 실험 결과 너비가 큰 평판에서, 그리고 수직 위치로부터의 각도가 클수록 임계열유속이 낮게 나타났다. 이는 가열면에서 발생된 기포들의 이탈이 어려워지기 때문인 것으로 판단된다. 경사가도에 따른 전체적 인 임계열유속 경향은 기존 연구들과 대체로 일치하나, 임계열유속 감소율이 변화하는 천이 각도가 -80$^{\circ}$ 근방에서 발견되었다.

• 대한핵의학회지
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• 제28권2호
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• pp.177-185
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• 1994

Pharmacologic coronary vasodilation in conjunction with myocardial perfusion scintigraphy has become an alternative to dynamic exercise test for the diagnosis and risk stratification of coronary artery disease, especially in patients who are unable to perform adequate exercise. Dipyridamole and adenosine have been used for pharmacologic stress testing with myocardial perfusion imaging. Adenosine is a potent coronary vasodilator with rapid onset of action, short half-life, near maximal coronary vasodilation and less serious side effects. ST segment depression has been reported in about 7-15% of patients with coronary artery disease receiving dipyridamole in conjunction with myocardial perfusion imaging. The exact cause and clinical significance are not known. In order to evaluate the relationship between adenosine-induced ST segment depression during $^{99m}Tc$-MIBI myocardial perfusion scintigraphy and the severity of coronary artery disease, we performed $^{99m}Tc$-MIBI imaging after intravenous Infusion of adenosine In 120 patients with suspected coronary artery disease. Of the 120 patients, 28 also performed coronary angiography. There were 24 patients with ST segment depression during $^{99m}Tc$-MIBI scintigraphy and 96 patients without ST segment depression. Adenosine was infused Intravenously at a dose of 0.14mg/kg per minute lot 6minutes and $^{99m}Tc$-MIBI was injected at 3 minute. We then com-pared the hemodynamic changes, side effects, scintigraphic and angiographic findings. Heart rate increased $90{\pm}19$ beats/minute in the group with ST depression compared with $80{\pm}16$ beats/minute in the group without ST depression(p<0.05). Baseline systolic blood pressure was significantly higher in the group with ST depression($152{\pm}27$ mmHg) than in the group without 57 depression($140{\pm}21$mmHg, p<0.05). Double product at baseline($10.90{\pm}2.77$ versus $9.55{\pm}2.34\;beats/minute{\times}mmHg$) and during adenosine infusion($12.72{\pm}3.89$ versus $10.83{\pm}2.98\;beats/minute{\times}mmHg$) were significantly higher in the group with ST depression(p<0.05). The incidence of anginal chest pain was also significantly higher in the group with ST depression(ST versus 29%, p<0.0001). The $^{99m}Tc$-MIBI images were abnormal in 23(96%) patients with ST segment depression and 66(69%) patients without ST segment depression(p<0.05). In patients with ST segment depression, there were more reversible perfusion defects than in patients without ST segment depression(83 versus 55%, p<0.05). The number of abnormal segments were significantly higher in the group with ST depression($3.05{\pm}2.01$ versus $1.51{\pm}1.45$, p<0.005). In patients with ST segment depression, there were more segments of reversible perfusion defects than in patients without segment depression($2.15{\pm}2.11$ versus $0.89{\pm}1.24$, p<0.05). There were no differences in the angiographic severity by vessel(p ; NS). We concluded that ST segment depression during $^{99m}Tc$-MIBI myocardial perfusion scintigraphy with Intravenous adenosine is related to the severity of coronary artery disease.

• 대한금속재료학회지
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• 제49권8호
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• pp.628-634
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• 2011

SA508 Gr.4N Ni-Cr-Mo low alloy steel, in which Ni and Cr contents are higher than in commercial SA508 Gr.3 Mn-Mo-Ni low alloy steels, may be a candidate reactor pressure vessel (RPV) material with higher strength and toughness from its tempered martensitic microstructure. The inner surface of the RPV is weld-cladded with stainless steels to prevent corrosion. The goal of this study is to evaluate the microstructural properties of the clad interface between Ni-Cr-Mo low alloy steel and stainless weldment, and the effects of post weld heat treatment (PWHT) on the properties. The properties of the clad interface were compared with those of commercial Mn-Mo-Ni low alloy steel. Multi-layer welding of model alloys with ER308L and ER309L stainless steel by the SAW method was performed, and then PWHT was conducted at $610^{\circ}C$ for 30 h. The microstructural changes of the clad interface were analyzed using OM, SEM and TEM, and micro-Vickers hardness tests were performed. Before PWHT, the heat affected zone (HAZ) showed higher hardness than base and weld metals due to formation of martensite after welding in both steels. In addition, the hardness of the HAZ in Ni-Cr-Mo low alloy steel was higher than that in Mn-Mo-Ni low alloy steel due to a comparatively high martensite fraction. The hardness of the HAZ decreased after PWHT in both steels, but the dark region was formed near the fusion line in which the hardness was locally high. In the case of Mn-Mo-Ni low alloy steel, formation of fine Cr-carbides in the weld region near the fusion line by diffusion of C from the base metal resulted in locally high hardness in the dark region. However, the precipitates of the region in the Ni-Cr-Mo low alloy steel were similar to that in the base metal, and the hardness in the region was not greatly different from that in the base metal.

• 에너지공학
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• 제20권1호
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• pp.21-25
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• 2011

본 연구에서는 방사성 폐기물을 처리하는 원통형 회전 분리기를 감시하기 위해 전기 저항법을 제안하고 수학적 모델을 연구하였다. 회전형 방사성 폐기물 분리기에서 관 벽에 설치된 한 쌍의 전극의 전기 저항은 표면을 따라 형성된 불용해성 입자의 환상 영역의 두께와 그 영역의 불용해성 입자의 농도와 관련이 있다. 본 연구는 전위 방정식에 대한 2차원 해와 실험적인 전도도-농도 관계를 기반으로 전술한 인자들 사이의 해석적 관계를 기술하였다. 또한, 원통형 회전 분리기를 감시하기 위한 전기 저항법의 적용 가능성을 논의 하였다.