Design modification and shaking table test of the drain pipes of steam generator for vibration reduction

Heo, Yong-Ho; Oh, Seung‐Hwan

doi:10.1016/j.nucengdes.2018.09.025

Cited by 2 publications

(1 citation statement)

References 15 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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