Mechanical Engineering Journal
Online ISSN : 2187-9745
Nuclear Power- Safe, Reliable and Clean Energy
Development of LORL evaluation method and its application to a loop-type sodium-cooled fast reactor
Yuya IMAIZUMIFumiaki YAMADAMitsuhiro ARIKAWAHiroki YADAYoshitaka FUKANO
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2018 Volume 5 Issue 4 Pages 18-00083

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Abstract

The loss-of-reactor-level (LORL) where the coolant circulation path is lost is one of the important accident types of loss-of-heat-removal-system (LOHRS) in loop-type sodium-cooled fast reactors (SFRs). Due to the non-negligible possibility obtained by probabilistic risk assessment (PRA), pipe failures and succeeding sodium leakage in two points both occurred in primary heat transport system (PHTS) was assumed in this study, unlike the conventional studies assuming single pipe failures. The sodium level in reactor vessel (RV) is affected by many elements such as leakage position, guard vessel volume, operational state of pumps, and countermeasures to maintain sodium level which are sodium pump-up into RV and siphon-breaking of the pipe between RV and pump. Thus, a calculation program was developed in this study to evaluate and discuss the effectiveness of the countermeasures and safety margins for the loss of coolant circulation path. In addition, the crack size was discussed and evaluated realistically, and t 2 (t : pipe thickness) was obtained for a sufficiently conservative value, instead of Dt/4 (D : pipe diameter) that was assumed in the conventional studies. Time interval between two leakages was also given by PRA, considering failure rates of the pipes and components. Representative sequences and leakage positions where the sodium level can decline below emergency sodium level (EsL) were chosen, and the sodium level transient in RV was calculated where the crack size of the second leakage was set t 2. The calculations were also conducted where the larger crack size, Dt/4, was set for both the first and second leakage, in order to investigate additional requirements to maintain the RV sodium level. The evaluation results clarified that the coolant circulation loop can be maintained even after the second leakage in PHTS, taking into account the effects by the countermeasures.

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© 2018 The Japan Society of Mechanical Engineers
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