Preliminary analysis of the post-disassembly expansion phase and structural response under unprotected loss of flow accident in prototype sodium cooled fast reactor

For the prototype sodium-cooled fast reactor, MONJU, the mechanical energy and structural response under energetics caused by neutronic power excursion during Unprotected Loss of Flow (ULOF) accident were preliminarily evaluated. In the first licensing of MONJU, pressure-volume relation (P-V relatio...

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Autores principales:	Yuichi ONODA, Ken-ichi MATSUBA, Yoshiharu TOBITA, Tohru SUZUKI
Formato:	article
Lenguaje:	EN
Publicado:	The Japan Society of Mechanical Engineers 2017
Materias:	sodium-cooled fast reactor severe accident unprotected loss of flow energetics simmer-iv autodyn Mechanical engineering and machinery TJ1-1570
Acceso en línea:	https://doaj.org/article/029d69babe974376bf8abaf7ec1bda13
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spelling	oai:doaj.org-article:029d69babe974376bf8abaf7ec1bda132021-11-26T07:03:57ZPreliminary analysis of the post-disassembly expansion phase and structural response under unprotected loss of flow accident in prototype sodium cooled fast reactor2187-974510.1299/mej.16-00597https://doaj.org/article/029d69babe974376bf8abaf7ec1bda132017-03-01T00:00:00Zhttps://www.jstage.jst.go.jp/article/mej/4/3/4_16-00597/_pdf/-char/enhttps://doaj.org/toc/2187-9745For the prototype sodium-cooled fast reactor, MONJU, the mechanical energy and structural response under energetics caused by neutronic power excursion during Unprotected Loss of Flow (ULOF) accident were preliminarily evaluated. In the first licensing of MONJU, pressure-volume relation (P-V relation) was evaluated based on the maximum theoretical work energy possible for an expanding core. It was adopted in the structural response analysis of the reactor vessel as the input. The maximum theoretical work energy is called Fuel Vapor Work Potential (FVWP) in this paper. In the successive studies of the energetics, mechanical energy was evaluated with the code in which mechanistic modelling of core expansion was implemented and this might reduce the Actual Work Potential (AWP) by an order of magnitude below FVWP. In order to evaluate the realistic structural response of the reactor vessel using AWP, method to convert the AWP to the P-V relation is necessary. Therefore, we developed the method to obtain realistic P-V relation based on the AWP by tracing the surface of the expanding core, and then we evaluated the mechanical energy and structural response under energetics during ULOF accident in MONJU using the developed method. The AWP is evaluated to 3 MJ based on the result of the latest ULOF analysis in which FVWP was evaluated to 30MJ, and sodium slug does not impact on the lower surface of the shield plug and no residual strain of the reactor vessel is evaluated. When FVWP is assumed to be 500 MJ as a hypothetical condition covering the conservative energy production, corresponding AWP is evaluated to 33 MJ. In this case, sodium slug impacts on the lower surface of the shield plug and residual strain of the reactor vessel of 0.008% at the maximum is evaluated, however the integrity of the primary boundary is still maintained.Yuichi ONODAKen-ichi MATSUBAYoshiharu TOBITATohru SUZUKIThe Japan Society of Mechanical Engineersarticlesodium-cooled fast reactorsevere accidentunprotected loss of flowenergeticssimmer-ivautodynMechanical engineering and machineryTJ1-1570ENMechanical Engineering Journal, Vol 4, Iss 3, Pp 16-00597-16-00597 (2017)
institution	DOAJ
collection	DOAJ
language	EN
topic	sodium-cooled fast reactor severe accident unprotected loss of flow energetics simmer-iv autodyn Mechanical engineering and machinery TJ1-1570
spellingShingle	sodium-cooled fast reactor severe accident unprotected loss of flow energetics simmer-iv autodyn Mechanical engineering and machinery TJ1-1570 Yuichi ONODA Ken-ichi MATSUBA Yoshiharu TOBITA Tohru SUZUKI Preliminary analysis of the post-disassembly expansion phase and structural response under unprotected loss of flow accident in prototype sodium cooled fast reactor
description	For the prototype sodium-cooled fast reactor, MONJU, the mechanical energy and structural response under energetics caused by neutronic power excursion during Unprotected Loss of Flow (ULOF) accident were preliminarily evaluated. In the first licensing of MONJU, pressure-volume relation (P-V relation) was evaluated based on the maximum theoretical work energy possible for an expanding core. It was adopted in the structural response analysis of the reactor vessel as the input. The maximum theoretical work energy is called Fuel Vapor Work Potential (FVWP) in this paper. In the successive studies of the energetics, mechanical energy was evaluated with the code in which mechanistic modelling of core expansion was implemented and this might reduce the Actual Work Potential (AWP) by an order of magnitude below FVWP. In order to evaluate the realistic structural response of the reactor vessel using AWP, method to convert the AWP to the P-V relation is necessary. Therefore, we developed the method to obtain realistic P-V relation based on the AWP by tracing the surface of the expanding core, and then we evaluated the mechanical energy and structural response under energetics during ULOF accident in MONJU using the developed method. The AWP is evaluated to 3 MJ based on the result of the latest ULOF analysis in which FVWP was evaluated to 30MJ, and sodium slug does not impact on the lower surface of the shield plug and no residual strain of the reactor vessel is evaluated. When FVWP is assumed to be 500 MJ as a hypothetical condition covering the conservative energy production, corresponding AWP is evaluated to 33 MJ. In this case, sodium slug impacts on the lower surface of the shield plug and residual strain of the reactor vessel of 0.008% at the maximum is evaluated, however the integrity of the primary boundary is still maintained.
format	article
author	Yuichi ONODA Ken-ichi MATSUBA Yoshiharu TOBITA Tohru SUZUKI
author_facet	Yuichi ONODA Ken-ichi MATSUBA Yoshiharu TOBITA Tohru SUZUKI
author_sort	Yuichi ONODA
title	Preliminary analysis of the post-disassembly expansion phase and structural response under unprotected loss of flow accident in prototype sodium cooled fast reactor
title_short	Preliminary analysis of the post-disassembly expansion phase and structural response under unprotected loss of flow accident in prototype sodium cooled fast reactor
title_full	Preliminary analysis of the post-disassembly expansion phase and structural response under unprotected loss of flow accident in prototype sodium cooled fast reactor
title_fullStr	Preliminary analysis of the post-disassembly expansion phase and structural response under unprotected loss of flow accident in prototype sodium cooled fast reactor
title_full_unstemmed	Preliminary analysis of the post-disassembly expansion phase and structural response under unprotected loss of flow accident in prototype sodium cooled fast reactor
title_sort	preliminary analysis of the post-disassembly expansion phase and structural response under unprotected loss of flow accident in prototype sodium cooled fast reactor
publisher	The Japan Society of Mechanical Engineers
publishDate	2017
url	https://doaj.org/article/029d69babe974376bf8abaf7ec1bda13
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Preliminary analysis of the post-disassembly expansion phase and structural response under unprotected loss of flow accident in prototype sodium cooled fast reactor

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