Development of LORL evaluation method and its application to a loop-type sodium-cooled fast reactor

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 failu...

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Autores principales:	Yuya IMAIZUMI, Fumiaki YAMADA, Mitsuhiro ARIKAWA, Hiroki YADA, Yoshitaka FUKANO
Formato:	article
Lenguaje:	EN
Publicado:	The Japan Society of Mechanical Engineers 2018
Materias:	coolant leakage severe accident sodium-cooled fast reactor loss-of-reactor-level (lorl) safety evaluation method Mechanical engineering and machinery TJ1-1570
Acceso en línea:	https://doaj.org/article/a486bebe15b44c138dfe0034be268f2d
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spelling	oai:doaj.org-article:a486bebe15b44c138dfe0034be268f2d2021-11-26T07:21:52ZDevelopment of LORL evaluation method and its application to a loop-type sodium-cooled fast reactor2187-974510.1299/mej.18-00083https://doaj.org/article/a486bebe15b44c138dfe0034be268f2d2018-06-01T00:00:00Zhttps://www.jstage.jst.go.jp/article/mej/5/4/5_18-00083/_pdf/-char/enhttps://doaj.org/toc/2187-9745The 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.Yuya IMAIZUMIFumiaki YAMADAMitsuhiro ARIKAWAHiroki YADAYoshitaka FUKANOThe Japan Society of Mechanical Engineersarticlecoolant leakagesevere accidentsodium-cooled fast reactorloss-of-reactor-level (lorl)safety evaluation methodMechanical engineering and machineryTJ1-1570ENMechanical Engineering Journal, Vol 5, Iss 4, Pp 18-00083-18-00083 (2018)
institution	DOAJ
collection	DOAJ
language	EN
topic	coolant leakage severe accident sodium-cooled fast reactor loss-of-reactor-level (lorl) safety evaluation method Mechanical engineering and machinery TJ1-1570
spellingShingle	coolant leakage severe accident sodium-cooled fast reactor loss-of-reactor-level (lorl) safety evaluation method Mechanical engineering and machinery TJ1-1570 Yuya IMAIZUMI Fumiaki YAMADA Mitsuhiro ARIKAWA Hiroki YADA Yoshitaka FUKANO Development of LORL evaluation method and its application to a loop-type sodium-cooled fast reactor
description	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.
format	article
author	Yuya IMAIZUMI Fumiaki YAMADA Mitsuhiro ARIKAWA Hiroki YADA Yoshitaka FUKANO
author_facet	Yuya IMAIZUMI Fumiaki YAMADA Mitsuhiro ARIKAWA Hiroki YADA Yoshitaka FUKANO
author_sort	Yuya IMAIZUMI
title	Development of LORL evaluation method and its application to a loop-type sodium-cooled fast reactor
title_short	Development of LORL evaluation method and its application to a loop-type sodium-cooled fast reactor
title_full	Development of LORL evaluation method and its application to a loop-type sodium-cooled fast reactor
title_fullStr	Development of LORL evaluation method and its application to a loop-type sodium-cooled fast reactor
title_full_unstemmed	Development of LORL evaluation method and its application to a loop-type sodium-cooled fast reactor
title_sort	development of lorl evaluation method and its application to a loop-type sodium-cooled fast reactor
publisher	The Japan Society of Mechanical Engineers
publishDate	2018
url	https://doaj.org/article/a486bebe15b44c138dfe0034be268f2d
work_keys_str_mv	AT yuyaimaizumi developmentoflorlevaluationmethodanditsapplicationtoalooptypesodiumcooledfastreactor AT fumiakiyamada developmentoflorlevaluationmethodanditsapplicationtoalooptypesodiumcooledfastreactor AT mitsuhiroarikawa developmentoflorlevaluationmethodanditsapplicationtoalooptypesodiumcooledfastreactor AT hirokiyada developmentoflorlevaluationmethodanditsapplicationtoalooptypesodiumcooledfastreactor AT yoshitakafukano developmentoflorlevaluationmethodanditsapplicationtoalooptypesodiumcooledfastreactor
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Development of LORL evaluation method and its application to a loop-type sodium-cooled fast reactor

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