• Title/Summary/Keyword: Main steam line break

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Development of Ceramic Humidity Sensor for the Korean Next Generation Reactor

  • Lee, Na-Young;Hwang, Il-Soon;Yoo, Han-Ill;Song, Chang-Rock
    • Proceedings of the Korean Nuclear Society Conference
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    • 1996.11a
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    • pp.183-190
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    • 1996

  • Leak-before-break(LBB) approach has been shown to be both cost and risk effective by reducing maintenance cost and occupational exposure when applied to high energy piping in nuclear power plants. For Korean Next Generation Reactor(KNGR) development, LBB is considered for the Main Steam Line(MSL) piping inside containment. Unlike the reactor coolant piping leakages which can be detected by particulate and gaseous radiation monitoring, main steam line leak detection systems must be based on principles that do not involve radioactivity. Ceramics are widely used as humidity sensor materials which can be further developed for nuclear applications. In this paper, we describe the progress in the development of ceramic humidity sensors for use with the main steam lines of KNGR.

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SAFETY ANALYSIS OF INCREASE IN HEAT REMOVAL FROM REACTOR COOLANT SYSTEM WITH INADVERTENT OPERATION OF PASSIVE RESIDUAL HEAT REMOVAL AT NO-LOAD CONDITIONS

  • SHAO, GE;CAO, XUEWU
    • Nuclear Engineering and Technology
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    • v.47 no.4
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    • pp.434-442
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    • 2015

  • The advanced passive pressurized water reactor (PWR) is being constructed in China and the passive residual heat removal (PRHR) system was designed to remove the decay heat. During accident scenarios with increase of heat removal from the primary coolant system, the actuation of the PRHR will enhance the cooldown of the primary coolant system. There is a risk of power excursion during the cooldown of the primary coolant system. Therefore, it is necessary to analyze the thermal hydraulic behavior of the reactor coolant system (RCS) at this condition. The advanced passive PWR model, including major components in the RCS, is built by SCDAP/RELAP5 code. The thermal hydraulic behavior of the core is studied for two typical accident sequences with PRHR actuation to investigate the core cooling capability with conservative assumptions, a main steam line break (MSLB) event and inadvertent opening of a steam generator (SG) safety valve event. The results show that the core is ultimately shut down by the boric acid solution delivered by Core Makeup Tank (CMT) injections. The effects of CMT boric acid concentration and the activation delay time on accident consequences are analyzed for MSLB, which shows that there is no consequential damage to the fuel or reactor coolant system in the selected conditions.

  • https://doi.org/10.1016/j.net.2015.01.007 인용 KSCI