• 한국전산구조공학회논문집
• /
• 제28권5호
• /
• pp.535-542
• /
• 2015

원자로 격납건물은 냉각재상실사고와 같이 내부의 과도한 압력이 유발되는 사고에 있어서도 방사성 물질이 외부로 누출되지 않도록 막는 최종의 방벽이다. 이러한 격납건물의 기능적 중요성에 기인하여, 건설 초기 구조건전성시험(SIT)을 수행한다. 이러한 SIT거동을 가장 실제와 가깝게 예측하기 위한 해석 연구를 수행하였다. 해당 연구의 결과는 2편의 논문으로 정리되었는데, 본 논문은 그 중 II편으로 I편의 해석모델 구성 시의 주요 고려사항의 분석 및 예비해석 결과를 반영한 상세 해석 모델의 구성 과정 및 해석 결과를 제시하고 있다. 특히 비부착식 텐던으로 시공된 구조물에서 덕트관에 의한 강성 저감효과 및 덕트관을 사이에 둔 텐던과 콘크리트간의 밀착 여부에 따른 영향을 해석 시 최대한 고려하고자 하였다. 이러한 과정을 통해 구축된 해석 모델에 따른 변위과 신고리 3호기 SIT 측정변위를 비교한 결과, ASME CC-6000 기준을 충분히 만족시키는 결과가 나타남을 확인하였다.

• 한국가스학회지
• /
• 제14권5호
• /
• pp.13-18
• /
• 2010

본 연구는 9% 니켈강재식 LNG 저장탱크용 통합제어안전관리시스템을 개발하고자 한다. 새로운 통합제어안전관리 시스템은 기존의 측정 및 제어시스템에 비해 압력, 변위, 하중을 측정할 수 있는 장치를 추가하였다. 또한, 측정된 데이터는 새로운 제어장치와 안전관리 시스템에 의해 통합되고 분석하는 프로세스를 동시적으로 진행하는 것이다. 초대형 완전밀폐식 LNG 저장탱크의 안전성과 효율성을 증가시키기 위해 통합제어안전관리시스템은 압력계이지를 추가하고, 내부탱크의 외측벽과 스티프너, 톱거더의 용접지역, 코너 프로텍션 탱크의 외측벽에 새로운 변위센서/압력센서를 설치하였다. 변위센서와 하중센서는 내부탱크와 코너 프로텍션 탱크의 9% 니켈강재 구조물에 대한 파손징후를 제공할 수 있고, 내부탱크로부터 누설되는 LNG를 감시할 수 있다. 기존의 누설센서는 내부탱크와 코너 프로텍션 사이의 단열지역에 설치한 누설감지기에 의해 경고신호가 접수될 때까지도 9% 니켈강재 탱크의 파괴에 대한 적절한 정보를 제공하지 못한다는데 문제가 있다. 따라서, 새로운 통합제어안전관리시스템은 온도, 압력, 변위, 하중, LNG의 밀도 데이터를 수집하고 분석하기 위한 것으로, 탱크시스템의 안전성과 내부탱크의 누설을 제어하기 위한 시스템이다. 또한, 디지털 데이터는 9% 니켈강재로 제작한 탱크의 안전성에 관련된 변위와 하중, LNG의 액위와 밀도, 쿨다운 공정, 누설, 압력 등을 제어하기 위해 측정한다.

• Nuclear Engineering and Technology
• /
• 제52권5호
• /
• pp.1051-1065
• /
• 2020

Safety analysis of nuclear power plant (NPP) especially in accident conditions is a basic and necessary issue for applications and commercialization of reactors. Many previous researches and development works have been conducted. However, most achievements focused on the safety reliability of primary pressure system vessels. Few literatures studied the structural safety of huge concrete structures surrounding primary pressure system, especially for the fourth generation NPP which allows existing of through cracks. In this paper, structural safety reliability of concrete structures of HTR-PM in accidental double-ended break of hot gas ducts was studied by Exceedance Probability Method. It was calculated by Monte Carlo approaches applying numerical simulations by Abaqus. Damage parameters were proposed and used to define the property of concrete, which can perfectly describe the crack state of concrete structures. Calculation results indicated that functional failure determined by deterministic safety analysis was decided by the crack resistance capability of containment buildings, whereas the bearing capacity of concrete structures possess a high safety margin. The failure probability of concrete structures during an accident of double-ended break of hot gas ducts will be 31.18%. Adding the consideration the contingency occurrence probability of the accident, probability of functional failure is sufficiently low.

• Nuclear Engineering and Technology
• /
• 제49권7호
• /
• pp.1431-1441
• /
• 2017

The Fukushima accident reveals the vulnerability of existing active nuclear power plant (NPP) design against prolonged loss of external electricity events. The passive safety system is considered an attractive alternative to cope with this kind of disaster. Also, the passive safety system enhances both the safety and the economics of NPPs. The adoption of a passive safety system reduces the number of active components and can minimize the construction cost of NPPs. In this paper, reflecting on the experience during the development of the APR+ design in Korea, we propose the concept of an innovative Power Reactor (iPower), which is a kind of passive NPP, to enhance safety in a revolutionary manner. The ultimate goal of iPower is to confirm the feasibility of practically eliminating radioactive material release to the environment in all accident conditions. The representative safety grade passive system includes a passive emergency core cooling system, a passive containment cooling system, and a passive auxiliary feedwater system. Preliminary analysis results show that these concepts are feasible with respect to preventing and/or mitigating the consequences of design base accidents and severe accidents.

• Nuclear Engineering and Technology
• /
• 제48권4호
• /
• pp.870-880
• /
• 2016

A steam generator tube rupture (SGTR) event with a stuck-open safety relief valve constitutes one of the most serious accident sequences in pressurized water reactors (PWRs) because it may create an open path for radioactive aerosol release into the environment. The release may be mitigated by the deposition of fission product particles on a steam generator's (SG's) dry tubes and structures or by scrubbing in the secondary coolant. However, the absence of empirical data, the complexity of the geometry, and the controlling processes have, until recently, made any quantification of retention difficult to justify. As a result, past risk assessment studies typically took little or no credit for aerosol retention in SGTR sequences. To provide these missing data, the Paul Scherrer Institute (PSI) initiated the Aerosol Trapping In Steam GeneraTor (ARTIST) Project, which aimed to thoroughly investigate various aspects of aerosol removal in the secondary side of a breached steam generator. Between 2003 and 2011, the PSI has led the ARTIST Project, which involved intense collaboration between nearly 20 international partners. This summary paper presents key findings of experimental and analytical work conducted at the PSI within the ARTIST program.

• Journal of international Conference on Electrical Machines and Systems
• /
• 제2권4호
• /
• pp.460-465
• /
• 2013

Electrical equipments of nuclear power plant are divided into class 1E and non-class 1E. Electrical equipment and systems that are essential to emergency reactor shutdown, containment isolation, reactor core cooling, and containment and reactor heat removal, are classified as class 1E. batteries of nuclear power plant are divided into four channels, which are physically and electrically separate and independent. The battery bank of class 1E DC power system of the nuclear power plant use lead-acid batteries in present. The lead acid battery, which has a high energy density, is the most popular form of energy storage. Kt factor of lead-acid battery is used to determine battery size and it is one of calculatiing coefficient for capacity. this paper analyzes Kt factor of lead-acid battery for the DC power system of nuclear power plant. In addition, correlation between Kt parameter and peukert's exponent of lead-acid battery for nuclear plant are discussed. The analytical results contribute to optimize of determining size Lead-acid battery bank.

• 전기학회논문지
• /
• 제58권7호
• /
• pp.1450-1455
• /
• 2009

Assumed that an IPTV camera system is to be used as an ad-hoc sensor for the surveillance and diagnostics of safety-critical equipments installed in the in-containment building of the nuclear power plant, an major problem is the presence of high dose-rate gamma irradiation fields inside the one. In order to uses an IPTV camera in such intense gamma radiation environment of the in-containment building, the radiation-weakened devices including a CCD imaging sensor, FPGA, ASIC and microprocessors are to be properly shielded from high dose-rate gamma radiation using the high-density material, lead or tungsten. But the passive elements such as mirror, lens and window, which are placed in the optical path of the CCD imaging sensor, are exposed to a high dose-rate gamma ray source directly. So, the gamma-ray irradiation characteristics of the passive elements, is needed to test. A CCD camera lens, made of glass material, have been gamma irradiated at the dose rate of 4.2 kGy/h during an hour up to a total dose of 4 kGy. The radiation induced color-center in the glass lens is observed. The degradation performance of the gamma irradiated lens is explained using an color component analysis.

• Nuclear Engineering and Technology
• /
• 제30권1호
• /
• pp.58-74
• /
• 1998

A PC window-based computer code, CONPAS (CONtainment Performance Analysis System), has been developed to integrate the numerical, graphical, and results-operation aspects of Level 2 probabilistic safety assessments (PSA) for nuclear power plants automatically. As a main logic for accident progression analysis, it employs a concept of the small containment phenomenological event tree (CPET) helpful to trace out visually individual accident progressions and of the detailed supporting event tree (DSET) for its detailed quantification. For the integrated analysis of Level 2 PSA, the code utilizes five distinct, but closely related modules. Its computational feasibility to real PSAs has been assessed through an application to the UCN 3&4 full scope Level 2 PSA. Compared with other existing computer codes for Level 2 PSA, the CONPAS code provides several advanced features: (1) systematic uncertainty analysis / importance analysis / sensitivity analysis, (2) table / graphical display & print, (3) employment of the recent Level 2 PSA technologies, and (4) highly effective user interface. The main purpose of this paper is to introduce the key features of CONPAS code and results of its feasibility study.

• International Journal of Ocean System Engineering
• /
• 제1권4호
• /
• pp.230-235
• /
• 2011

Reinforced polyurethane foam (RPUF) is one of the important materials of Mark III type insulation systems used in liquefied natural gas (LNG) cargo containment systems. However, RPUF is the most difficult material to use with regard to its safety assessment, because there is little public and reliable data on its mechanical properties, and even some public data show relatively large differences. In this study, to investigate the structural response of the system under compressive loads such as sloshing action, time-dependent characteristics of RPUF were examined. A series of compressive load tests of the insulation system including RPUF under various temperature conditions was carried out using specimens with rectangular section. As a result, the relationship between deformation of RPUF and time is linear and dependent on the loading rate, so the concept of strain rate could be applied to the analysis of the insulation system. Also, we found that the spring constant tends to converge to a value as the loading rate increases and that the convergence level is dependent on temperature.

• 한국압력기기공학회 논문집
• /
• 제7권3호
• /
• pp.40-47
• /
• 2011

KSNPs(Korea Standard Nuclear Power Plant) have been applied the break exclusion criteria to the high energy lines passing through containment penetration area to ensure that piping failures would not cause the loss of containment isolation function, and to reduce the resulting dynamic effects. Systems with the criteria are the Main Steam system, Feed Water system, Steam Generator Blowdown system, and Chemical & Volume Control system. In accordance with FSAR(Final Safety Analysis Report), a 100% volumetric examination by augmented in-service inspection of all pipe welds appled the break exclusion criteria is required for the break exclusion application piping. However, it is difficult to fully satisfy the requirements of inspection because 12", 8" and 6" weldolet weldments of Main Steam pipe line have complex structural shapes. To resolve the difficulty on the application of conventional UT(Ultrasonic Testing) technique, realistic mock-ups and UT calibration blocks were made. Simulations of conventional UT were performed utilizing CIVA, a commercial NDE(Nondestructive Examination) simulation software. Phased array UT experiments were performed through mock-up including artificial notch type flaws. A phased array UT technique is finally developed to improve the reliability of ultrasonic test at main steam line pipe to 12", 8" and 6" branch connection weld.