Dynamic transient analysis of the reactor core barrel due to sudden rupture of the recirculation line piping

Dubyk, Y. R.; Orynyak, I. V.

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DC pole	Hodnota	Jazyk
dc.contributor.author	Dubyk, Y. R.
dc.contributor.author	Orynyak, I. V.
dc.coverage.temporal	19-22 вересня 2017 року
dc.coverage.temporal	19-22 September 2017
dc.date.accessioned	2017-12-07T15:56:21Z	-
dc.date.available	2017-12-07T15:56:21Z	-
dc.date.created	2017-09-19
dc.date.issued	2017-09-19
dc.identifier.citation	Dubyk Y. R. Dynamic transient analysis of the reactor core barrel due to sudden rupture of the recirculation line piping / Y. R. Dubyk, I. V. Orynyak // Праці конференції „Пошкодження матеріалів під час експлуатації, методи його діагностування і прогнозування“, 19-22 вересня 2017 року. — Т. : ТНТУ, 2017. — С. 187–190. — (Оцінювання залишкового ресурсу елементів конструкцій).
dc.identifier.isbn	978-966-305-083-6
dc.identifier.uri	http://elartu.tntu.edu.ua/handle/lib/22585	-
dc.description.abstract	We have analyzed sudden rupture of the primary cooling loop which causes a water hammer event for the reactor core barrel. Assuming that core barrel is a thin shell, we have performed dynamic stress and strain calculations in the frequency domain. The Duhamel integral was used to calculate the transient response of a shell to an impulse load caused by the water hammer event. The results obtained were used to estimate structural stability of the core barrel
dc.format.extent	187-190
dc.language.iso	en
dc.publisher	ТНТУ
dc.publisher	TNTU
dc.relation.ispartof	Праці Ⅴ Міжнародної науково-технічної конференції „Пошкодження матеріалів під час експлуатації, методи його діагностування і прогнозування“
dc.relation.ispartof	Proceeding of the International Conference “In-Service Damage of Materials, its Diagnostics and Prediction”
dc.title	Dynamic transient analysis of the reactor core barrel due to sudden rupture of the recirculation line piping
dc.type	Conference Abstract
dc.rights.holder	© Тернопільський національний технічний університет імені Івана Пулюя, 2017; © Ternopil Ivan Pulu’uj National Technical University, 2017
dc.coverage.placename	Тернопіль
dc.coverage.placename	Ternopil
dc.format.pages	4
dc.relation.references	1. NUREG-0609, “Asymmetric Blowdown Loads on PWR (Pressurized-Water-Reactor) Primary Systems: Resolution of Generic Task Action Plan A-2,” Nuclear Regulatory Commission, January 1981.
dc.relation.references	2. BARC/1998/E/032, Fluid Structure Interaction Studies on Acoustic Load Response of Light Water NuclearReactor Core Internals Under Blowdown Condition, Bhabha Atomic Research Centre, 1998, Mumbai, India.
dc.relation.references	3. “Coolant Blowdown Studies of a Reactor Simulator Vessel Containing a Perforated Sieve Plate Separator,” AEC Research and Development Report, Battelle Memorial Institute Pacific Northwest Laboratories, BNWL-1463.
dc.relation.references	4. Sommerville, D., Karpanan, K., “Boiling Water Reactor Core Shroud Acoustic Loads Resulting from a Recirculation Outlet Line Break Loss of Coolant Accident – A Case Study,” 2011 ASME PVP Conference, PVP2011-57743.
dc.relation.references	5. Antti Timperi et al “Validation of fluid-structure interaction calculations in a large break loss of coolant accident” ICONE1648206 May 1115, 2008, Orlando, Florida, USA.
dc.relation.references	6. Y. Murakami, (Editor-in-chief) //Stress Intensity Factors Handbook Volume 2, Pergamon Press (1987).
dc.relation.references	7. Novozhilov, V.V., 1970. Thin Shell Theory. Wolters-Noordhoff, Groningen.
dc.relation.references	8. Dubyk I. R. Analysis of water hammer due to sudden rupture of reactor coolant system/I. R. Dubyk, I. V. Orynyak // Proceedings of the ASME 2016 Pressure Vessels and Piping Conference PVP2016-63589. — July 17-21, 2016, Vancouver. — 9p
dc.relation.references	9. M. Bergman“Stress intensity factors for circumferential surface cracks in pipes”, Fatigue Fract. Eng. Mater. Struct Vol 18. No10 pp.1155-1172, 1995
dc.relation.references	10. Little, E. A., “Dynamic J-Integral Toughness and Fractographic Studies of Fast Reactor IrradiatedType 321 Stainless Steel,” Effects of Radiation on Material, Properties: 12th Intl. Symp., ASTM STP 870, American Society of Testing and Materials, Philadelphia, PA, pp. 563-579, 1985.
dc.relation.referencesen	1. NUREG-0609, "Asymmetric Blowdown Loads on PWR (Pressurized-Water-Reactor) Primary Systems: Resolution of Generic Task Action Plan A-2," Nuclear Regulatory Commission, January 1981.
dc.relation.referencesen	2. BARC/1998/E/032, Fluid Structure Interaction Studies on Acoustic Load Response of Light Water NuclearReactor Core Internals Under Blowdown Condition, Bhabha Atomic Research Centre, 1998, Mumbai, India.
dc.relation.referencesen	3. "Coolant Blowdown Studies of a Reactor Simulator Vessel Containing a Perforated Sieve Plate Separator," AEC Research and Development Report, Battelle Memorial Institute Pacific Northwest Laboratories, BNWL-1463.
dc.relation.referencesen	4. Sommerville, D., Karpanan, K., "Boiling Water Reactor Core Shroud Acoustic Loads Resulting from a Recirculation Outlet Line Break Loss of Coolant Accident – A Case Study," 2011 ASME PVP Conference, PVP2011-57743.
dc.relation.referencesen	5. Antti Timperi et al "Validation of fluid-structure interaction calculations in a large break loss of coolant accident" ICONE1648206 May 1115, 2008, Orlando, Florida, USA.
dc.relation.referencesen	6. Y. Murakami, (Editor-in-chief) //Stress Intensity Factors Handbook Volume 2, Pergamon Press (1987).
dc.relation.referencesen	7. Novozhilov, V.V., 1970. Thin Shell Theory. Wolters-Noordhoff, Groningen.
dc.relation.referencesen	8. Dubyk I. R. Analysis of water hammer due to sudden rupture of reactor coolant system/I. R. Dubyk, I. V. Orynyak, Proceedings of the ASME 2016 Pressure Vessels and Piping Conference PVP2016-63589, July 17-21, 2016, Vancouver, 9p
dc.relation.referencesen	9. M. Bergman"Stress intensity factors for circumferential surface cracks in pipes", Fatigue Fract. Eng. Mater. Struct Vol 18. No10 pp.1155-1172, 1995
dc.relation.referencesen	10. Little, E. A., "Dynamic J-Integral Toughness and Fractographic Studies of Fast Reactor IrradiatedType 321 Stainless Steel," Effects of Radiation on Material, Properties: 12th Intl. Symp., ASTM STP 870, American Society of Testing and Materials, Philadelphia, PA, pp. 563-579, 1985.
dc.identifier.citationen	Dubyk Y. R., Orynyak I. V. (2017) Dynamic transient analysis of the reactor core barrel due to sudden rupture of the recirculation line piping. Proceedings of the Conference „In-service damage of materials, its diagnostics and prediction“ (Tern., 19-22 September 2017), pp. 187-190 [in English].
dc.contributor.affiliation	G.S. Pisarenko Institute for Problems of Strength, National Academy of Sciences of Ukraine
dc.citation.journalTitle	Праці Ⅴ Міжнародної науково-технічної конференції „Пошкодження матеріалів під час експлуатації, методи його діагностування і прогнозування“
dc.citation.spage	187
dc.citation.epage	190
dc.citation.conference	Ⅴ Міжнародна науково-технічна конференція „Пошкодження матеріалів під час експлуатації, методи його діагностування і прогнозування“
Vyskytuje se v kolekcích:	Ⅴ Міжнародна науково-технічна конференція „Пошкодження матеріалів під час експлуатації, методи його діагностування і прогнозування“ (2017)

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