Dynamic characteristics assessment of reactor vessel internals with fluid-structure interaction

Je, Sang Yun; Chang, Yoon‐Suk; Kang, Sung-Sik

doi:10.1016/j.net.2017.05.003

Cited by 12 publications

(2 citation statements)

References 16 publications

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“…The design and safety of nuclear power plant components and systems have received much attention. From relatively simple components such as helical tube [8], fuel rods [9], and steam generator drain pipes [10] to the complicated systems such as nuclear coolant pump [11], reactor core support [12,13], shield building [14], and reactor internal [15,16]. Both simulation [9,10,13,14,16] and experiments [11,15,17,18] are adopted to verify the applicability of the design and modification for the components and systems.…”

Section: Introductionmentioning

confidence: 99%

Experimental and Numerical Investigation of Temperature Distribution on Reactor Pressure Vessel Support under Operating Conditions

Lv²,

Liu

2019

Science and Technology of Nuclear Installations

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Reactor pressure vessel (RPV) support is a key safety facility which is categorized as Class 1 in the ASME nuclear safety design. The temperature distribution of RPV support is one of the key considerations for the concrete safety contacting with the bottom of the support. So it is necessary for accurate evaluation on the temperature field characteristics of RPV support, especially the bottom of support. This paper investigates the temperature field characteristics of modified RPV support which will be applied to a large advanced pressurized water reactor. A support entity is manufactured in a ratio of 1:1, and its temperature distribution is measured under simulated reactor operating conditions. Numerical simulation is also used to validate the results by the developed CFD model. The results show that under the operating conditions, of which the inlet cooling air temperature is 35.35°C and the velocity is 6.25 m/s, the temperature distribution of modified RPV support bottom is uneven, and the highest temperature is around 38°C, which is much lower than the demanding design temperature 93.3°C. Therefore, the design of the modified RPV support is reliable. In addition, the results of numerical simulation agree well with the experimental results with the error less than ±4°C, which ensures the reliability of the conclusion. The effects of inlet cooling air temperature and velocity on the RPV support temperature distribution are further studied. Both the temperature decrease and velocity increase can reduce the RPV support temperature. But the effect of inlet cooling air temperature is more obvious than inlet cooling air velocity. So the best way to improve air cooling capacity is to decrease the support inlet cooling air temperature. The results can provide a good guidance to the design of RPV support for the subsequent large advanced pressurized water reactor.

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Section: Introductionmentioning

confidence: 99%

Experimental and Numerical Investigation of Temperature Distribution on Reactor Pressure Vessel Support under Operating Conditions

Lv²,

Liu

2019

Science and Technology of Nuclear Installations

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“…FEA of crankshaft shows smaller NF with fine mesh [11]. The FEA analysis performs respect spectrum analysis of reactor vessel to satisfy the stress criteria [12].…”

Section: Introductionmentioning

confidence: 99%

The Earthquake Safety Assessment of 1000 kVA Cast Resin Transformer by Finite Element Analysis with Mesh Convergence and Seismic Test on the Large Shaking Table Under Korean and IEEE Seismic Standards

Lee

Kim

Mun³

2019

J. Electr. Eng. Technol.

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A 1000 kVA cast resin transformer (CRT) is an electrical device that reduces the amount of voltage flow for end-consumers to use safely. This paper describes seismic response analysis by numerical finite element analysis (FEA) using ANSYS and the seismic test for the Korean peninsula and IEEE-693 seismic standards. The CRT is mounted on the shaking table for seismic test carried out at KIMM (Korea Institute of Machinery and Material) to see the stress level and safety. Several FEA analysis steps are performed to match the fundamental natural frequency from resonance search test. In addition, convergence test of element's mesh size shows that the natural frequency decreases when mesh size becomes smaller. Also, optimization is performed to find the Young's modulus considering special stiffness characteristics of laminated core in CRT. Both the natural frequency and stress results from the numerical computation and seismic test are close and show adequate electrical and structural safety margin judged by the American Society of Civil Engineering design code for the maximum earthquake rate.

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Calculation of coupled added mass of multibody system immersed in fluids

2021

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Dynamic characteristics assessment of reactor vessel internals with fluid-structure interaction

Cited by 12 publications

References 16 publications

Experimental and Numerical Investigation of Temperature Distribution on Reactor Pressure Vessel Support under Operating Conditions

Experimental and Numerical Investigation of Temperature Distribution on Reactor Pressure Vessel Support under Operating Conditions

The Earthquake Safety Assessment of 1000 kVA Cast Resin Transformer by Finite Element Analysis with Mesh Convergence and Seismic Test on the Large Shaking Table Under Korean and IEEE Seismic Standards

Calculation of coupled added mass of multibody system immersed in fluids

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