• 한국소음진동공학회논문집
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• 제21권12호
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• pp.1098-1103
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• 2011

The structural integrity of APR1400 reactor vessel internals has been being assessed referring the US Nuclear Regulatory Commission regulatory guide 1.20, comprehensive vibration assessment program. The program is composed of a vibration and stress analysis, a vibration and stress measurement, and an inspection. This paper covers the vibration and stress analysis on the reactor vessel internals by the pulsation of reactor coolant pump. 3-dimensional models to calculate the hydraulic loads and structural responses were built and the pressure distributions and the structural responses were predicted using ANSYS. This paper presents that APR1400 reactor vessel internals have enough structural integrity against the pulsation of reactor coolant pump as the peak stress of the reactor vessel internals is much lower than the acceptance limit.

• International Journal of Safety
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• 제10권2호
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• pp.1-5
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• 2011

The structural integrity assessment of APR1400 (Advanced Power Reactor 1400) reactor vessel internals has been being performed referring the US Nuclear Regulatory Commission regulatory guide 1.20 comprehensive vibration assessment program prior to commercial operation. The program is composed of a hydraulic and structural analysis, a vibration measurement, and an inspection. This paper describes the hydraulic and structural analysis on the reactor vessel internals due to hydraulic loads caused by the turbulence of reactor coolant. Three-dimensional models were built for the hydraulic and structural analysis and then hydraulic loads and structural responses were predicted for five analysis cases with CFX and ANSYS respectively. The structural responses show that the APR1400 reactor vessel internals have sufficient structural integrity in comparison with the acceptance criteria.

• Nuclear Engineering and Technology
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• 제30권2호
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• pp.164-172
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• 1998

This paper presents the detailed modelling of reactor vessel ledge region for the dynamic analysis of the coupled internals and core model. The dynamic responses due to earthquake and pipe break are calculated using the input motions of reactor vessel taken from Ulchin nuclear power plant units 3 and 4. Two different representations for detailed and simplified models of the RV ledge region are made. The dynamic responses of the reactor internals components are compared between them. Response characteristics are reported and simplified model is suggested for earthquake and pipe break analysis for the future design of the reactor internals.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2011년도 추계학술대회 논문집
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• pp.221-226
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• 2011

The structural integrity of APR1400 reactor vessel internals has been being assessed referring the US Nuclear Regulatory Commission regulatory guide 1.20 comprehensive vibration assessment program. The program is composed of a vibration and stress analysis, a limited vibration measurement, and an inspection. This paper covers the vibration and stress analysis on the reactor vessel internals due to the pulsation of reactor coolant pump. 3-dimensional models to calculate the hydraulic loads and structural responses were built and the pressure distributions and the structural responses were predicted using ANSYS. The peak stress of the reactor vessel internals is much lower than the acceptance limit.

• Nuclear Engineering and Technology
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• 제53권3호
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• pp.988-1003
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• 2021

Existing elastic analysis methods cannot be adhered to in order to assess the structural integrity of a reactor vessel and internals for a beyond design basis earthquake. Elasto-plastic analysis methods are required, and the factors that affect the elasto-plastic behavior of reactor materials should be taken into account. In this study, a material behavior model was developed that considers the irradiation embrittlement effect, which affects the elasto-plastic behavior of the reactor material. This was used to perform the elasto-plastic time history analyses of the reactor vessel and its internals for beyond design basis earthquake. For this investigation, appropriate beyond design basis earthquakes and reliable finite element models were used. Based on the analysis results, consideration was given to the load reduction effect and the margin change. These were transferred to the internals due to the plastic deformation of the reactor vessel.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2004년도 추계학술대회 논문집
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• pp.1327-1330
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• 2004

An underwater robotic system has been developed and applied to visual inspection of reactor vessel internals. The Korea Electric Power Robot for Visual Test (KeproVt) consists of an underwater robot, a vision processor-based measuring unit, a master control station and a servo control station. The robot guided by the control station with the measuring unit can be controlled to have any motion at any position in the reactor vessel with $\pm$1 cm positioning and $\pm$2 degrees heading accuracies with enough precision to inspect reactor internals. A simple and fast installation process is emphasized in the developed system. The developed robotic system was successfully deployed at the Younggwang Nuclear Unit 1 for the visual inspection of reactor internals.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 추계학술대회
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• pp.1358-1363
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• 2003

Due to the application of Integrated Head Assembly (IHA) in KSNP+ reactor design, an investigation of reactor internals integrity is carried out to assure that the adoption of IHA does not affect the safety of reactor operation. One of the postulated accident events is the R.V. closure head fall from 5.5m high directly above the reactor vessel that may occur during the refueling operation. The analysis model consists of lumped mass elements of the entire reactor vessel and internals. Because of extreme load, separate elastic-plastic analyses are done for the members that undergo plastic deformation. The analysis verified that the stresses of the reactor internals and the fuel assemblies are within the bound of allowable stress limits and the integrity of the fuel assemblies is maintained.

• Nuclear Engineering and Technology
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• 제25권1호
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• pp.28-36
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• 1993

최근 국내에서 가압경수로형 원자력발전소를 표준화하기 위한 작업이 이루어지고 있다. 본 논문에서는 설계표준화 작업의 일환으로서 원자력발전소 원자로내부구조물에 대한 내진설계기준을 제시하였다. 영광 3,4호기 최종설계단계에서의 운전기준지진에 대한 원자로용기 플랜지와 스너버의 거동을 입력하중으로 사용하여 지진설계하중을 계산하였고 이로부터 원자로내부구조물의 설계에 허용가능한 원자로용기의 거동을 규정하였다. 해석방법등 해석의 전반적인 개요에 대하여 설명하였고 원자로용기의 거동에 따른 원자로내부구조물 각각의 응답에 대하여 자세히 고찰하였다.

• 한국압력기기공학회 논문집
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• 제9권1호
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• pp.20-24
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• 2013

Reactor Vessel Internals(RVIs), which are installed within the reactor pressure vessel and support the fuel assembly, take responsibility for safety of reactor core. In operating Nuclear Power Plants(NPPs), the RVIs have been exposed to severe conditions such as neutron irradiation, high temperature, high pressure, and high velocity of coolant flow and have degraded by materials aging with long-term operation. Therefore, the effective aging management plan and the appropriate regulatory requirements are necessary to maintain the integrity of RVIs. The purpose of this paper is to provide a review guide for Periodic Safety Review(PSR) of RVIs in presurized water reactor. The review guide is developed based on the revised review guides and reports established from IAEA and USNRC, and the analysis results of design characteristics, aging mechanisms, and operating experiences of RVIs in domestic and international NPPs. Consequently, the developed review guide for PSR of RVIs is expected to contribute an overall strategy and standard for the PSR of RVIs.

• Structural Engineering and Mechanics
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• 제51권6호
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• pp.1005-1016
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• 2014

Roles of reactor internals are to support nuclear fuel, provide insertion and withdrawal channels of nuclear fuel control rods, and carry out core cooling. In case of functional loss of the reactor internals, it may lead to severe accidents caused by damage of nuclear fuel assembly and deterioration of reactor vessel due to attack of fallen out parts. The present study is to examine fluid flows in reactor internals subjected to hydrodynamic loads. In this context, an integrated model was developed and applied to two kinds of numerical analyses; one is to analyze periodic loading effect caused by pump pulsation and the other is to analyze random loading effect employing different turbulent models. Acoustic pressure distributions and flow velocity as well as pressure and temperature fields were calculated and compared to establish appropriate analysis techniques.