This study was carried out to suggest a improvement plan for nuclear administration system in korea by comparing our country with other leading countries for nuclear power policy, current state and plans. It is suggested that NSSC, directly responsible to the prime minster should be upgraded to the president, the independence of NSSC will be maintained. A region safety committee directly responsible to the local government should be established, forge the dualistic regulation framework for mutual help and check, and consolidate the supervision of management for local nuclear power plant. NSSC is in charge of the engineered safety and protection of facilities, disaster protection etc., regional safety committee plays a special role in the safety measures for inhabitants and environmental monitoring

According to the statistics of IEA (International Energy Agency), China has the strong dependency on the coal energy. However, the coal energy creates the highest level of CO2 emission among the energy resource. Since 2006, China became the most important CO2 emission country. In order to control the air polution and the CO2 emission, Chinese government is diversifying the energy resource for its economy. After Fukushima incident, Chinese government checked the safety of nuclear facility and suspended the new nuclear power plant construction. This governmental measures were similar to those of Germany. However, Chinese government resumed the new nuclear power plant construction while Germen government turn its nuclear policy into the shutdown. The important role of Chinese governmental nuclear companies is similar to the French situation of nuclear energy development by the Government-owned company. Like the French nuclear company which acquired its nuclear technology from Westinghouse, China introduced and learned the nuclear technology from USA (AP1000) and from France (EPR). Especially, French government and French nuclear company are willing to cooperate with China in the field of nuclear energy. French nuclear company, Areva expect that the over than 50% of potential nuclear energy development in the World will be created in China.

This study presents a structural safety analysis method for a plant annunciator panel under the seismic effect. Seismic qualification analysis for the nuclear plant annunciator panel is carried out to confirm the structural integrity and the results are represented by required response spectra. For the numerical analysis, finite element method is adopted. Mode combinations are also used to obtain the reliability of the spectrum analysis. The analysis results shows that the nuclear plant annunciator panel is designed as a dynamically rigid assembly, without any resonance frequency blow 33Hz. The calculated stress of the nuclear plant annunciator panel is much less than yield stress of used steel.

In this study, dynamic characteristics and seismic capacity of the nuclear power plant piping system are evaluated by model test results using multi-platform shake table. The model is 21.2 m long and consists of straight pipes, elbows, and reducers. The stainless steel pipe diameters are 60.3 mm (2 in.) and 88.9 mm (3 in.) and the system was assembled in accordance with ASME code criteria. The dynamic characteristics such as natural frequency, damping and acceleration responses of the piping system were estimated using the measured acceleration, displacement and strain data. The natural frequencies of the specimen were not changed significantly before and after the testing and the failure and leakage of the piping system was not observed until the final excitation. The damping ratio was estimated in the range of 3.13 ~ 4.98 % and it is found that the allowable stress(345 MPa) according to ASME criteria is 2.5 times larger than the measured maximum stress(138 MPa) of the piping system even under the maximum excitation level of this test.

Involved in a research for the application of seismic isolation to the nuclear industry, this study evaluates firstly the responses of seismic isolation system considering general ranges of structural period and damping ratio by using preliminary design formula. Secondly, coupling effects of input motions were evaluated to find out appropriate conditions of excitations and effect of the iteration for calculating yield displacement of lead core was also assessed in terms of response of a seismically isolated structure. Finally, the results of preliminary design calculation were compared with those of dynamic analysis and the propriety of the formula was evaluated and appropriate ranges of reduction factor were also suggested from the results.

Since commercial operation of Kori-1 nuclear power plant in 1978, twenty units are under operation and six units are under construction at 2011 present. Nuclear power become a main energy source in our country. However, the plant was constructed by a USA

Traditionally Nuclear Research and Development (R&D) result has been big influence on other industries and societies and it requires large scale investments and study period. So it is essential to apply Quality Assurance (QA) for systematic R&D management

An Emergency Diesel Generator (EDG) installed in a nuclear power plant is the primary power source, supplying AC power to Class 1E power systems when the main turbine generator and offsite power source are not available. Thus, reliability of the EGD is essential for plant safety and availability. In this paper, the EDG is selected for a Long Term Asset Management (LTAM) strategy and the results are summarized briefly. The LTAM strategy is intended to provide an effective long-term planning tool for minimizing unplanned capability loss and then optimizing maintenance programs and capital investments consistent with plant safety and an identified plant operating strategy. Such an operating strategy might include license renewal or retaining the option for license renewal.

원자력발전소에 설치되는 안전관련 캐비닛형 전기기기는 설치 전에 내진검증이 요구된다. 전기기기의 동특성분석은 내진 검증에 포함된 중요한 과정이며, 기기의 정확한 해석모델을 작성하기 위해서도 필수적으로 요구되는 업무이다. 이 연구에서는 입력진동수준에 따른 기기의 동특성 변화를 분석하기 위하여 원전 지진감시시스템 캐비닛을 대상으로 진동대시험을 수행하고, 입력진동운동의 수준별로 계측된 진동응답신호를 진동수영역분해법으로 분석하였다. 분석결과, 대상기기는 입력진동수준의 크기에 따라 동특성이 비선형적으로 변화하고, 국내 원전의 안전정지지진 수준 이하의 진동에서도 동특성이 비선형적 거동을 보이고 있음을 확인하였다. 이러한 입력진동 수준에 따라 전기기기의 동특성이 비선형적으로 변하는 원인은 대상기기의 특성과 입력진동수준을 고려할 때 일반적인 재료 비선형보다는 각 부품들의 마찰력과 기하학적인 비선형성에 기인하는 것으로 판단된다. 따라서 전기 캐비닛의 입력진동수준에 따른 동특성의 비선형적 변화는 향후 안전관련 기기의 내진검증 업무에서 중요하게 검토되어야 할 것으로 판단된다.

월성원자력발전소 주변해역에서 출현한 동물플랑크톤은 종 수준까지 동정이 가능한 32종을 포함하여, 총 63종과 85~28,087개체 m-3 범위였다. 연구해역에 분포하는 동물플랑크톤 군집의 구조를 분석한 결과, 전반적으로 배수구를 중심으로 북쪽과 동쪽에 위치한 정점군(그룹 A)과 취수구 아래쪽 정점군(그룹 B)의 두 개의 그룹으로 구분되었다. 동물플랑크톤 군집의 분리는 10월에 출현 종수가, 그 밖의 시기인 1월, 5월 및 8월에는 개체수가 영향

본 연구는 월성 원자력발전소 주변해역에 서식하는 대형저서동물의 출현 종수, 생물량 및 군집구조를 파악하기 위하여 2007년 10월부터 2008년 7월까지 계절별로 현장조사를 실시하였다. 총 163종의 대형저서동물이 출현하였고, 단위면적당 (m2) 개체수와 생체량은 각각 1,005개체와 21.81 gWWt이었다. 출현 개체수의 자료를 기초로 LeBris index (1988)를 이용하여 상위 10위까지의 우점종을 선정하였다. 개체수에 근거한 상위 1

A institute developed Quality Assurance(QA) program for nuclear R&D projects to meet the demands of its customers' requirements for recognized quality standards and nuclear industry accepted practices. It was implemented by project quality assurance plan as a new process. This paper is designed to introduce the process of establishment and execution of nuclear quality assurance programs for R&D as a case study. This QA program can be used as a reference to other organization on implementation of QA for R&D projects.

본 연구는 원자력발전소에 설치되는 캐비닛형 전기기기의 동적 진동시험 자료를 이용하여 캐비닛의 지진응답을 예측할 수 있는 기법을 제안하였다. 제안된 기법은 1) 절점질량 이상화 모델에 기반한 등가 지진하중 산정, 2) 진동시험자료에 기반한 캐비닛 구조의 입출력 상태방정식 규명, 3) 산정된 등가지진하중과 규명된 입출력 상태방정식을 사용한 지진응답산정의 과정으로 구성된다. 제안된 기법은 유한요소기법(FEM) 모델 개선(Model Updating)에 기반한 지진응답예측기법에 비하여 모델링 오차가 개입 되지 않는 장점을 가진다. 캐비넷 구조를 이상화한 2차원 프레임 모델과 3차원 상세 모델에 대한 수치검증을 통하여 제안된 기법이 지진응답을 매우 정확하게 예측을 함을 관찰하였고, 측정 노이즈에 대해서도 강인함을 관찰하였다. 추후연구로 실험검증이 요구된다.

An emergency diesel generator(EDG) manufactured by a French company Wartsila SACM is a tandem type engine and consisted of two 10 cylindered diesel engines on each side. The maintenance manual provided by the manufacturer recommends that engine bearing be inspected every 15 years. However, it is difficult to inspect them because the manhole located in the lower compartment of the engine is too small for maintenance worker to access engine internals. Furthermore, the EDG should be disassembled and then overturned to inspect bearings unlike other EDG type. Such process will take longer period time than ordinary maintenance period. So it is not possible to inspect the main engine bearing and crank shaft during a routine or scheduled maintenance. In this paper, five methods are proposed and estimated to resolve the problem and the optimal maintenance method is chosen among them. The proposed optimal maintenance plan makes it possible to perform proper maintenance during regular maintenance period and to lower maintenance cost considerably.

The main objective of this paper is to search whether containment vessel's best pressure may increase until how long when loss of coolant accident (LOCA) happened in containment vessel of Ulchin nuclear power plant 1 and 2. Another goal of this research is to find the influential factors that increase containment vessel pressure. Model for this research is Ulchin nuclear power plant 1 with 10 cycles. Data were collected by simulator of Ulchin nuclear power plant 1 and design of experiment was used for data analysis. For the experiment, seven factors that are going to influence in containment vessel pressure were chosen. It was found that fatter which influences in early rise of containment vessel pressure after LOCA is only explosion size. Also, containment vessel's best pressure (3.74 bar.a) was much lower than limit (4.86 bar.a) of FSAR (Final Safety Analysis Report).

The Main Control Room(MCR) of Korean Next Generation Reactor(KNGR) has faced entirely new operation environment. The Computerized Procedure System(CPS) is completely different from existed Paper-based Procedure(PBP). The CPS which is improved by evaluation in this study is ready to be applied to the ergonomic foundations. In this study, we redesigned displays of the CPS as a Man-Machine Interface System(MMIS) to maximize human performance and to minimize human error and proposed an alternative display designs of the CPS applying principles related with MMIS.

This study presents a method of quality category classification by safety, maturity, complexity, and what types and extent of controls and verifications are applied to specific products and services during the various stages of a nuclear facility life cycle. All products, services and processes have various controls and verifications built in to ensure they perform their functions satisfactorily. The highest grade should require the most stringent application of the quality assurance requirements ; while, the lowest grade should require the least stringent. When products or services are modified, the assigned grade of quality assurance requirements could become more stringent or less stringent depending on the significance in nuclear safety. Applying QA program always costs money, and they should be applied and focused to the extent where necessary and not applied or applied to a lesser degree for less important activities. An efficient QA program should be developed to satisfy the necessary requirements and to ensure the required confidence in quality, but without unnecessary stipulations. Not all the requirements of QA standard must be applied identically to all products and services which are to be provided.

원자력 발전의 고온 가스로(high temperature gas-cooled reactor, HTGR)의 냉각제로 사용되는 He가스의 열에너지를 이용하여 물을 분해해서 수소를 생산하는 "열화학적 수소제조 IS프로세스"에 대해 설명하였다. 특히, 분리막 기술의 이용에 관한 연구를 중점으로 정리하였다. 고온 원자력 열에너지를 이용한 열화학적 수소 제조법은 실현 가능한 단계까지 왔다고 생각되며, 아직 연구 개발 과제가 많이 남아 있지만, 미래의 청정에너지 중의 하나인 수소를 대량 생산할 수 있는 가능성을 갖고 있다.