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Konovalenko, A., Sköld, P., Kudinov, P., Bechta, S. & Grishchenko, D. (2017). Controllable Generation of a Submillimeter Single Bubble in Molten Metal Using a Low-Pressure Macrosized Cavity. Metallurgical and materials transactions. B, process metallurgy and materials processing science, 48(2), 1064-1072.
Open this publication in new window or tab >>Controllable Generation of a Submillimeter Single Bubble in Molten Metal Using a Low-Pressure Macrosized Cavity
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2017 (English)In: Metallurgical and materials transactions. B, process metallurgy and materials processing science, ISSN 1073-5615, E-ISSN 1543-1916, Vol. 48, no 2, 1064-1072 p.Article in journal (Refereed) Published
Abstract [en]

We develop a method for generation of a single gas bubble in a pool of molten metal. The method can be useful for applications and research studies where a controllable generation of a single submillimeter bubble in opaque hot liquid is required. The method resolves difficulties with bubble detachment from the orifice, wettability issues, capillary channel and orifice surfaces degradation due to contact with corrosive hot liquid, etc. The macrosized, 5- to 50-mm(3) cavity is drilled in the solid part of the pool. Flushing the cavity with gas, vacuuming it to low pressure, as well as sealing and consequent remelting cause cavity implosion due to a few orders in magnitude pressure difference between the cavity and the molten pool. We experimentally demonstrate a controllable production of single bubbles ranging from a few milliliters down to submillimeter size. The uncertainties in size and bubble release timing are estimated and compared with experimental observations for bubbles ranging within 0.16 to 4 mm in equivalent-volume sphere diameter. Our results are obtained in heavy liquid metals such as Wood's and Lead-Bismuth eutectics at 353 K to 423 K (80 A degrees C to 150 A degrees C).

Place, publisher, year, edition, pages
SPRINGER, 2017
National Category
Metallurgy and Metallic Materials
Identifiers
urn:nbn:se:kth:diva-206279 (URN)10.1007/s11663-017-0914-z (DOI)000396028600030 ()2-s2.0-85011313828 (Scopus ID)
Note

QC 20170509

Available from: 2017-05-09 Created: 2017-05-09 Last updated: 2017-05-09Bibliographically approved
Manickam, L., Bechta, S. & Ma, W. (2017). On the fragmentation characteristics of melt jets quenched in water. International Journal of Multiphase Flow, 91, 262-275.
Open this publication in new window or tab >>On the fragmentation characteristics of melt jets quenched in water
2017 (English)In: International Journal of Multiphase Flow, ISSN 0301-9322, E-ISSN 1879-3533, Vol. 91, 262-275 p.Article in journal (Refereed) Published
Abstract [en]

Experiments were carried out to investigate the characteristics of jet breakup and debris formation after melt jets fall into a subcooled water pool, which may occur in industrial processes such as the interactions of molten corium with coolant during a severe accident of light water reactors. A high-speed visualization system developed previously at KTH was used to capture the jet fragmentation phenomenon. Molten metal (Woods metal or tin) and mixture of binary oxides (WO3-Bi2O3 or WO3-ZrO2) were employed separately as melt materials to address different breakup mechanisms (e.g., hydrodynamic vs. thermodynamic fragmentation) and material effect. Moreover, the parameters related to melt and water conditions, including superheat, jet diameter and velocity of melt as well as subcooling of water, were scrutinized for their influences on jet fragmentation characteristics. The experimental data were acquired on the melt jet fragmentation patterns, breakup length, droplet size spectrum, droplet breakup and solidification as well as debris morphology, which can be useful for validation of the codes used for the prediction of debris formation phenomena.

Place, publisher, year, edition, pages
Elsevier, 2017
Keyword
Multiphase flow; Jet breakup; Fragmentation; Debris formation
National Category
Other Physics Topics
Research subject
Physics
Identifiers
urn:nbn:se:kth:diva-204647 (URN)10.1016/j.ijmultiphaseflow.2017.02.005 (DOI)000398752500019 ()2-s2.0-85013499337 (Scopus ID)
Note

QC 20170412

Available from: 2017-03-30 Created: 2017-03-30 Last updated: 2017-05-10Bibliographically approved
Fischer, M., Bechta, S., Bezlepkin, V. V., Hamazaki, R. & Miassoedov, A. (2016). Core Melt Stabilization Concepts for Existing and Future LWRs and Associated Research and Development Needs. Paper presented at 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH), AUG 30-SEP 04, 2015, Chicago, IL. NUCLEAR TECHNOLOGY, 196(3), 524-537.
Open this publication in new window or tab >>Core Melt Stabilization Concepts for Existing and Future LWRs and Associated Research and Development Needs
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2016 (English)In: NUCLEAR TECHNOLOGY, ISSN 0029-5450, Vol. 196, no 3, 524-537 p.Article in journal (Refereed) Published
Abstract [en]

In the event of a severe accident in a nuclear power plant with the core melting, the stabilization of the molten corium is an important mitigation issue, as it can avoid late containment failure caused by basemat penetration, overpressure, or severe damage to internal structures. The related failure modes may result in significant long-term radiological consequences and related high costs. Because of this, the licensing frameworks of several countries now include a requirement to implement mitigative core melt stabilization measures. This applies not only to new builds but also to existing light water reactors. The paper gives an overview of the ex-vessel core melt stabilization strategies developed during the last decades. These strategies are based on a variety of physical principles, like melt fragmentation in a deep water pool or during the molten core-concrete interaction with top flooding, water injection from the bottom (COMET), and retention in an outside-cooled crucible structure. This overview covers the physical background and functional principles of these concepts, as well as their validation status and, if applicable, the remaining open issues and research and development needs. For the concepts based on melt retention inside a cooled crucible that have reached sufficient maturity to be implemented in current Generation III+ designs, like the VVER-1000/1200 and the European Pressurized Water Reactor, more detailed descriptions are provided, which include key aspects of the related technical realization. The paper is compiled using contributions from the main developers of the individual concepts.

Place, publisher, year, edition, pages
American Nuclear Society, 2016
Keyword
Severe accident mitigation, LWR, core catcher
National Category
Water Engineering
Identifiers
urn:nbn:se:kth:diva-200230 (URN)10.13182/NT16-19 (DOI)000390524000010 ()2-s2.0-85002876985 (Scopus ID)
Conference
16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH), AUG 30-SEP 04, 2015, Chicago, IL
Note

QC 20170213

Available from: 2017-02-13 Created: 2017-02-13 Last updated: 2017-02-16Bibliographically approved
Manickam, L., Kudinov, P., Ma, W., Bechta, S. & Grishchenko, D. (2016). On the influence of water subcooling and melt jet parameters on debris formation. Nuclear Engineering and Design, 309, 265-276.
Open this publication in new window or tab >>On the influence of water subcooling and melt jet parameters on debris formation
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2016 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 309, 265-276 p.Article in journal (Refereed) Published
Abstract [en]

Breakup of melt jet and formation of a porous debris bed at the base-mat of a flooded reactor cavity is expected during the late stages of a severe accident in light water reactors. Debris bed coolability is determined by the bed properties including particle size, morphology, bed height and shape as well as decay heat. Therefore understanding of the debris formation phenomena is important for assessment of debris bed coolability. A series of experiments was conducted in MISTEE-jet facility by discharging binary-oxide mixtures of WO3-Bi2O3 and WO3-ZrO2 into water in order to investigate properties of resulting debris. The effect of water subcooling, nozzle diameter and melt superheat was addressed in the tests. Experimental results reveal significant influence of water subcooling and melt superheat on debris size and morphology. Significant differences in size and morphology of the debris at different melt release conditions is attributed to the competition between hydrodynamic fragmentation of liquid melt and thermal fracture of the solidifying melt droplets. The particle fracture rate increases with increased sub cooling. Further the results are compared with the data from larger scale experiments to discern the effects of spatial scales. The present work provides data that can be useful for validation of the codes used for the prediction of debris formation phenomena.

Place, publisher, year, edition, pages
Elsevier, 2016
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-198956 (URN)10.1016/j.nucengdes.2016.09.017 (DOI)000387629900022 ()2-s2.0-84991702020 (Scopus ID)
Note

QC 20170113

Available from: 2017-01-13 Created: 2016-12-22 Last updated: 2017-11-29Bibliographically approved
Journeau, C., Bouyer, V., Cassiaut-Louis, N., Fouquart, P., Piluso, P., Ducros, G., . . . Zdarek, J. (2016). SAFEST ROADMAP FOR CORIUM EXPERIMENTAL RESEARCH IN EUROPE. In: PROCEEDINGS OF THE 24TH INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING, 2016, VOL 4: . Paper presented at 2016 24th International Conference on Nuclear Engineering, ICONE 2016; Charlotte; United States; 26 June 2016 through 30 June 2016. ASME Press, 4, Article ID V004T14A019.
Open this publication in new window or tab >>SAFEST ROADMAP FOR CORIUM EXPERIMENTAL RESEARCH IN EUROPE
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2016 (English)In: PROCEEDINGS OF THE 24TH INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING, 2016, VOL 4, ASME Press, 2016, Vol. 4, V004T14A019Conference paper, Published paper (Refereed)
Abstract [en]

SAFEST (Severe Accident Facilities for European Safety Targets) is a European project networking the European corium experimental laboratories with the objective to establish coordination activities, enabling the development of a common vision and research roadmaps for the next years, and of the management structure to achieve these goals. In this frame, a European roadmap on corium experimental research has been written to define research challenges to contribute to further reinforcement of Gen II and III NPP safety. It is based on the research priorities deteimined by SARNET SARP group as well as those from the recently formulated in the NUGENIA Roadmap for severe accidents and the recently published NUGENIA Global Vision report. It also takes into account issues identified in the analysis of the European stress tests and from the interpretation of the Fukushima accident. 19 relevant issues related to corium have been selected during these prioritization efforts. These issues have been compared to a survey of the European corium experimental facilities and corium analysis laboratories. Finally, the coherence between European infrastructures and R&D needs has been assessed and a table linking issues and infrastructures has been derived. It shows a few lacks in EU corium infrastructures, especially in the domains of core late reflooding impact on source term, Reactor Pressure Vessel failure and corium release, Spent Fuel Pool accidents, as well as the need for a large mass (100500 kg) prototypic corium facility.

Place, publisher, year, edition, pages
ASME Press, 2016
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-199003 (URN)10.1115/ICONE24-60916 (DOI)000387190900093 ()2-s2.0-84995618157 (Scopus ID)978-079185004-6 (ISBN)
Conference
2016 24th International Conference on Nuclear Engineering, ICONE 2016; Charlotte; United States; 26 June 2016 through 30 June 2016
Note

QC 20170118

Available from: 2017-01-18 Created: 2016-12-22 Last updated: 2017-01-18Bibliographically approved
Sehgal, B. R. & Bechta, S. (2016). Severe accident progression in the BWR lower plenum and the modes of vessel failure. Annals of Nuclear Energy.
Open this publication in new window or tab >>Severe accident progression in the BWR lower plenum and the modes of vessel failure
2016 (English)In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100Article in journal (Refereed) Published
Abstract [en]

Most of our knowledge base on the severe accident progression in the lower plenum of LWRs is based on the data obtained from the TMI-2 accident. It should be recognized that the lower plenum of a BWR is very different from that of a PWR. Unlike the PWR, the BWR plenum is full of control rod guide tubes (CRGTs) with their axial structural variations. These CRGTs are arranged in a cellular fashion with each CRGT supporting 4 rod bundles. There are also a large number of instrument guide tubes (IGTs), each generally placed in the middle of 4CRGTs. Both the CRGTs and IGTs traverse the thick vessel bottom wall and are welded to their extensions which come to bottom of the core. The core-melt progression in the lower plenum is controlled by the structures present and they, in turn, influence the timings and the modes of vessel failure for a BWR.The uranium oxide-zirconium oxide core melt formed in the 4 fuel bundles is directed by the structure below toward the water regions in-between the 4 CRGTs. The FCI will take place in those water regions and some particulate debris will be created, although there is insufficient water for quenching the melt. A FCI may occur inside a CRGT if and when the melt enters the CRGT at its top opening or the melt in the water region between the four CRGTs breaches the wall of the CRGT.The important issue is whether the welding holding the IGT inside the vessel will fail and the bottom part of the IGT falls out creating a hole in the vessel with release of water and melt/particulate debris from the vessel to the dry well of the BWR containment. Similarly, the failure of CRGT could have water and melt/particulate debris coming out of the vessel. These modes of vessel failure appear to be credible and they could occur before any large-scale melting and melt pool convection takes place. These modes of vessel failure and the melt release to the containment will have very different consequences than those generated by the other modes of vessel failure.Such BWR plenum melt progression scenarios have been considered in this paper. Some results of analyses performed at KTH have been described. We believe that the issues raised are important enough to consider a set of experiments for verification and validation of the melt progression in a BWR plenum. Such experiments are proposed.

Place, publisher, year, edition, pages
Elsevier, 2016
Keyword
BWR lower plenum, CRGT weld failure, Heat transfer analysis, IGT weld failure, Mode of BWR vessel failure, Severe accident, Accidents, Boiling water reactors, Debris, Heat transfer, Hydrophilicity, Knowledge based systems, Welding, Welding rods, Welds, Vessel failure, Weld failure, Failure analysis
National Category
Mechanical Engineering
Identifiers
urn:nbn:se:kth:diva-184191 (URN)10.1016/j.anucene.2015.12.030 (DOI)2-s2.0-84957068899 (Scopus ID)
Note

QC 20160405

Available from: 2016-03-30 Created: 2016-03-30 Last updated: 2017-11-30Bibliographically approved
Komlev, A., Gusarov, V., Bechta, S. V., Almjashev, V. & Khabensky, V. (2016). Single phase sacrificial material for core catcher. ru 102883.
Open this publication in new window or tab >>Single phase sacrificial material for core catcher
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2016 (Russian)Patent (Other (popular science, discussion, etc.))
National Category
Engineering and Technology
Identifiers
urn:nbn:se:kth:diva-184202 (URN)
Patent
RU 102883
Note

QS 2016

Available from: 2016-03-30 Created: 2016-03-30 Last updated: 2016-04-14Bibliographically approved
Kumar, R., Bechta, S., Kudinov, P., Curnier, F., Marquès, M. & Bertrand, F. (2015). A PSA Level-1 method with repairable components: An application to ASTRID Decay Heat Removal systems. In: Safety and Reliability: Methodology and Applications - Proceedings of the European Safety and Reliability Conference, ESREL 2014. Paper presented at European Safety and Reliability Conference, ESREL 2014, 14 September 2014 through 18 September 2014, Wroclaw (pp. 1611-1617). .
Open this publication in new window or tab >>A PSA Level-1 method with repairable components: An application to ASTRID Decay Heat Removal systems
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2015 (English)In: Safety and Reliability: Methodology and Applications - Proceedings of the European Safety and Reliability Conference, ESREL 2014, 2015, 1611-1617 p.Conference paper, Published paper (Refereed)
Abstract [en]

Technological advancements in area of sensor-based online maintenance systems have made the possibility of repairing some failed safety support systems of Nuclear Power Plants (NPP) such as electrical supply, I&C systems, ventilation systems. However, the possibility of repair during accident situation is yet to be included into PSA level-1. Therefore, this paper presents a scheme of PSA level-1 by implementing an integrated method of Repairable Event Tree (RET) and Repairable Fault Tree (RFT) analysis. The Core Damage Frequency (CDF) is calculated from consequence probabilities of the RET. An initiating event of Decay Heat Removal (DHR) systems of ASTRID reactor is analyzed. The proportionate CDFs estimated with repair and without repair have been compared and found that the recoveries can reduce CDF. In sum, this paper attempts to deal with the possibility of repair of some safety systems in PSA and its impacts on CDF of the NPP.

Keyword
Nuclear power plants, Online systems, Reliability, Ventilation, Accident situation, Core damage frequency, Decay heat removal, Decay heat removal systems, Online maintenance, Safety support systems, Technological advancement, Ventilation systems, Repair
National Category
Other Engineering and Technologies
Identifiers
urn:nbn:se:kth:diva-167393 (URN)10.1201/b17399-219 (DOI)000380543400207 ()2-s2.0-84906679503 (Scopus ID)9781138026810 (ISBN)
Conference
European Safety and Reliability Conference, ESREL 2014, 14 September 2014 through 18 September 2014, Wroclaw
Note

QC 20150601

Available from: 2015-06-01 Created: 2015-05-22 Last updated: 2016-12-06Bibliographically approved
Fischer, M., Bechta, S., Bezlepkin, V. V., Hamazaki, R. & Miassoedov, A. (2015). Core melt stabilization concepts for existing and future LWRs and associated R&D needs. In: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015: . Paper presented at 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2015; Chicago; United States (pp. 7578-7592). , 9.
Open this publication in new window or tab >>Core melt stabilization concepts for existing and future LWRs and associated R&D needs
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2015 (English)In: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015, 2015, Vol. 9, 7578-7592 p.Conference paper, Published paper (Refereed)
Abstract [en]

In the event of a severe accident with core melting in a NPP the stabilization of the molten corium is an important mitigation issue, as it can avoid late containment failure caused by basemat penetration, overpressure, or severe damage of internal structures. The related failure modes may result in significant long-term radiological consequences and high related costs. Because of this, the licensing framework of several countries now includes the request to implement mitigative core melt stabilization measures. This does not only apply to new builds but also to existing LWR plants. The paper gives an overview of the ex-vessel core melt stabilization strategies developed during the last decades. These strategies are based on a variety of physical principles like: melt fragmentation in a deep water pool or during molten core concrete interaction with top-flooding, water injection from the bottom (COMET concept), and retention in an outside-cooled crucible structure. The provided overview covers the physical background and functional principles of these concepts, as well as their status of validation and, if applicable, the remaining open issues and R&D needs. For concepts based on melt retention inside a cooled crucible that reached sufficient maturity to be implemented in current Gen-III+ designs, like the VVER-1000/1200 and the EPR™, more detailed descriptions are provided, which include key aspects of the related technical realization. The paper is compiled using contributions from the main developers of the individual concepts.

Keyword
Core catcher, LWR, Severe accident mitigation
National Category
Information Systems
Identifiers
urn:nbn:se:kth:diva-187404 (URN)2-s2.0-84964068542 (Scopus ID)978-151081184-3 (ISBN)
Conference
16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2015; Chicago; United States
Note

QC 20160525

Available from: 2016-05-25 Created: 2016-05-23 Last updated: 2018-01-10Bibliographically approved
Dietrich, P., Kretzschmar, F., Miassoedov, A., Class, A., Villanueva, W. & Bechta, S. (2015). Coupling of melcor with the pecm for improved modelling of a core melt in the lower plenum. In: International Conference on Nuclear Engineering, Proceedings, ICONE: . Paper presented at 23rd International Conference on Nuclear Engineering: Nuclear Power - Reliable Global Energy, ICONE 2015, 17 May 2015 through 21 May 2015. JSME.
Open this publication in new window or tab >>Coupling of melcor with the pecm for improved modelling of a core melt in the lower plenum
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2015 (English)In: International Conference on Nuclear Engineering, Proceedings, ICONE, JSME , 2015Conference paper, Published paper (Refereed)
Abstract [en]

MELCOR contains a coupling interface based on the MPI-Standard, which enables the communication to other codes such as RELAP5 or GASFLOW. However a detailed knowledge of this coupling interface in MELCOR is necessary to use this possibility. Therefore, at the KIT the software tool DINAMO (Direct Interface for Adding Models) has been developed. This program contains the coupling routines as well as an interface to communicate with other programs. Using DINAMO it is also possible to utilize new or enhanced models for phenomena, which occur during a severe accident in a nuclear power plant, in MELCOR without modification of the MELCOR source code. In the present work MELCOR calculations of experiments in the LIVE-Facility are presented. The LIVE-Facility is used to simulate the behavior of a melt in the lower plenum of a reactor pressure vessel (RPV). For these calculations we coupled MELCOR via DINAMO with the Phase-Change Effective Convectivity Model (PECM), which has been developed at the KTH in Stockholm. Using the PECM it is possible to improve the prediction of a core melt in the lower plenum of a RPV in case of a core melt accident.

Place, publisher, year, edition, pages
JSME, 2015
Keyword
Coupled codes, LIVE, Lower head, MELCOR, PECM
National Category
Engineering and Technology
Identifiers
urn:nbn:se:kth:diva-184194 (URN)2-s2.0-84959049239 (Scopus ID)
Conference
23rd International Conference on Nuclear Engineering: Nuclear Power - Reliable Global Energy, ICONE 2015, 17 May 2015 through 21 May 2015
Note

QC  20160330

Available from: 2016-03-30 Created: 2016-03-30 Last updated: 2016-03-30Bibliographically approved
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ORCID iD: ORCID iD iconorcid.org/0000-0001-7816-8442

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