Change search
Link to record
Permanent link

Direct link
BETA
Alternative names
Publications (10 of 75) Show all publications
Khabensky, V. B., Granovsky, V. S., Almjashev, V. I., Vitol, S. A., Krushinov, E. V., Kotova, S. J., . . . Tromm, W. (2018). Effect of temperature gradient on chemical element partitioning in corium pool during in-vessel retention. Nuclear Engineering and Design, 327, 82-91
Open this publication in new window or tab >>Effect of temperature gradient on chemical element partitioning in corium pool during in-vessel retention
Show others...
2018 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 327, p. 82-91Article in journal (Refereed) Published
Abstract [en]

The paper presents some results of the ISTC (International Science and Technology Center)-financed project ‘Investigation of Corium Melt Interaction with NPP Reactor Vessel Steel’ (METCOR). In the METCOR experiments the metallic phase of a two-liquid system was produced by the interaction between hot suboxidized corium and cooled VVER vessel steel, with the steel being corroded. Models of corrosion mechanisms in the considered conditions are used to systematize data on the limiting temperature of corrosion/(dissolution) of the vessel steel. A considerable influence of thermal gradient conditions is shown, which has to be taken into account in the analysis of molten pool behaviour. 

Place, publisher, year, edition, pages
Elsevier Ltd, 2018
Keywords
Chemical elements, Corrosion, Nuclear reactors, Thermal gradients, Corrosion mechanisms, Effect of temperature, Element partitioning, In-vessel retention, International science, Limiting temperature, Reactor vessel steel, Two-liquid systems, Temperature
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-223155 (URN)10.1016/j.nucengdes.2017.11.030 (DOI)000425302400008 ()2-s2.0-85038017789 (Scopus ID)
Note

Export Date: 13 February 2018; Article; CODEN: NEDEA; Correspondence Address: Bottomley, D.; Invited Researcher at JAEA/CLADSJapan; email: dboksb3@gmail.com; Funding details: ISTC, International Science and Technology Center. QC 20180314

Available from: 2018-03-14 Created: 2018-03-14 Last updated: 2018-03-14Bibliographically approved
Fichot, F., Carénini, L., Sangiorgi, M., Hermsmeyer, S., Miassoedov, A., Bechta, S., . . . Guenadou, D. (2018). Some considerations to improve the methodology to assess In-Vessel Retention strategy for high-power reactors. Annals of Nuclear Energy, 119, 36-45
Open this publication in new window or tab >>Some considerations to improve the methodology to assess In-Vessel Retention strategy for high-power reactors
Show others...
2018 (English)In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 119, p. 36-45Article in journal (Refereed) Published
Abstract [en]

The In-Vessel Retention (IVR) strategy for Light Water Reactors (LWR) intends to stabilize and isolate corium and fission products in the reactor pressure vessel and in the primary circuit. This type of Severe Accident Management (SAM) strategy has already been incorporated in the SAM guidance (SAMG) of several operating small size LWR (reactor below 500 MWe (like VVER440)) and is part of the SAMG strategies for some Gen III + PWRs of higher power like the AP1000 or the APR1400. However, for high power reactors, estimations using current level of conservatism show that RPV failure caused by thermo-mechanical rupture takes place in some cases. A better estimation of the residual risk (probability of cases with vessel rupture) requires the use of models with a lower level of conservatism. In Europe, the IVMR project aims at providing new experimental data and a harmonized methodology for IVR. A synthesis of the methodology applied to demonstrate the efficiency of IVR strategy for VVER-440 in Europe (Finland, Slovakia, Hungary and Czech Republic) was made. It showed very consistent results, following quite comparable methodologies. The main weakness of the demonstration was identified in the evaluation of the heat flux that could be reached in transient situations, e.g. under the “3-layers” configuration, where the “focusing effect” may cause higher heat fluxes than in steady-state (due to transient “thin” metal layer on top). Analyses of various designs of reactors with a power between 900 and 1300 MWe were also made. Different models for the description of the molten pool were used: homogeneous, stratified with fixed configuration, stratified with evolving configuration. The last type of model provides the highest heat fluxes (above 3 MW/m2) whereas the first type provides the lowest heat fluxes (around 500 MW/m2) but is not realistic due to the non-miscibility of steel with UO2. Obviously, there is a need to reach a consensus about best estimate practices for IVR assessment to be used in the major codes for safety analysis, such as ASTEC, MELCOR, SOCRAT, MAAP, ATHLET-CD, SCDAP/RELAP, etc. Despite the model discrepancies, and leaving aside the unrealistic case of homogeneous pool, the average calculated heat fluxes in many cases are well above 1 MW/m2 which could reduce the residual thickness of the vessel considerably and threaten its integrity. Therefore, it is clear that the safety demonstration of IVR for high power reactors requires a more careful evaluation of the situations which can lead to formation of either a very thin top metal layer provoking focusing effect or significantly overheated metal, e.g. after oxide and metal layer inversion. It also requires an accurate mechanical analysis of the ablated vessel. The current approach followed by most experts for IVR is a compromise between a deterministic analysis using the significant knowledge gained during the last two decades and a probabilistic analysis to take into account large uncertainties due to the lack of data for some physical phenomena (such as transient effects) and due to excessive simplifications of models. A harmonization of the positions of safety authorities on the IVR strategy is necessary to allow decision making based on shared scientific knowledge. Currently, the acceptance criteria of a safety demonstration for IVR may be differently defined from one country to the other and the differences should be further discussed to reach harmonization on this important topic. This includes the accident scenarios to be considered in the demonstration and the modelling of the phenomena in the vessel. Such harmonization is one of the goals of IVMR project. A revised methodology is proposed, where the safety criterion is based not only on a comparison of the heat flux and the Critical Heat Flux (CHF) profiles as in current approaches but also on the minimum vessel thickness reached after ablation and the maximum integral loads that is applied to the vessel during the transient. The main advantage of this revised criterion is in consideration of both steady-state and transient loads on the RPV. Another advantage is that this criterion may be used in both probabilistic and deterministic approaches, whereas the current approaches are mostly deterministic (with deterministic calculations used only for estimates of uncertainty ranges of input parameters).

Place, publisher, year, edition, pages
Elsevier, 2018
Keywords
Corium molten pool, In-Vessel Retention, IVMR, Light water reactor, Safety, Severe accident, Vessel integrity
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-229257 (URN)10.1016/j.anucene.2018.03.040 (DOI)2-s2.0-85046369762 (Scopus ID)
Funder
EU, Horizon 2020, 662157
Note

QC 20180601

Available from: 2018-06-01 Created: 2018-06-01 Last updated: 2018-06-01Bibliographically approved
Zambaux, J. A., Manickam, L., Meignen, R., Ma, W. M., Bechta, S. & Picchi, S. (2018). Study on thermal fragmentation characteristics of a superheated alumina droplet. Annals of Nuclear Energy, 119, 352-361
Open this publication in new window or tab >>Study on thermal fragmentation characteristics of a superheated alumina droplet
Show others...
2018 (English)In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 119, p. 352-361Article in journal (Refereed) Published
Abstract [en]

In the frame of the European Commission FP7 SAFEST project, IRSN proposed to experimentally investigate the steam explosion triggering mechanisms of a superheated alumina droplet falling into water, through a set of experiments in the Micro Interactions in Steam Explosion Energetics facility (MISTEE) at KTH. Since thermal fragmentation is considered to be a likely process for the triggering of Steam Explosions in the KROTOS tests (performed at CEA) with alumina, the ability of a single droplet of such material to undergo thermally induced fine fragmentation is studied on the MISTEE facility with a close-up visualization. A series of experiments were conducted, where droplets of molten alumina were discharged into a water pool and potentially exposed to a small pressure wave. The intense interactions were recorded with a high-speed camera along with the pressure in the droplet vicinity. The ability of alumina to undergo thermal fragmentation is expected to be firstly contingent on the stability of the vapour film enshrouding the melt droplet. The water and melt temperatures may then play a crucial role on the vapour film stability, and therefore on the observation of a steam explosion. Indeed, under high to moderate water sub-cooling conditions, experimental observations indicate that fine fragmentation of the melt can occur when the droplet is exposed to even a weak pressure wave, in the range of 0.15 MPa. In contrast, melt fine fragmentation is suppressed at low water sub-cooling conditions (less than 30 °C), where the formation of a thick vapour film (and large wake) is observed, and which is probably too stable to be destabilized by the weak pressure wave. The effect of the melt temperature on thermal fragmentation is also assessed. This parameter influences the solidification of the droplet and the strength of the explosion as it determines the available heat energy. In the present conditions, fine fragmentation of melt occurred even at quite low melt superheat (≈60 °C). For a high melt superheat (above 200 °C) a very energetic spontaneous steam explosion was observed. A physical analysis on the debris particles acquired indicates a mass median diameter of ≈100 µm, comparable to the one observed in the KROTOS alumina experiments. The MISTEE experimental results are finally used to assess the heat and mass transfer modelling of the coolant during the fragmentation process in the FCI code MC3D.

Place, publisher, year, edition, pages
Elsevier, 2018
Keywords
Alumina, MC3D, Steam explosion, Thermal fragmentation
National Category
Energy Engineering
Identifiers
urn:nbn:se:kth:diva-228702 (URN)10.1016/j.anucene.2018.05.029 (DOI)2-s2.0-85047160015 (Scopus ID)
Funder
EU, FP7, Seventh Framework Programme
Note

QC 20180530

Available from: 2018-05-30 Created: 2018-05-30 Last updated: 2018-05-30Bibliographically approved
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
Show others...
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, p. 1064-1072Article 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, p. 262-275Article 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
Keywords
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
Show others...
2016 (English)In: NUCLEAR TECHNOLOGY, ISSN 0029-5450, Vol. 196, no 3, p. 524-537Article 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
Keywords
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
Show others...
2016 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 309, p. 265-276Article 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
Show others...
2016 (English)In: PROCEEDINGS OF THE 24TH INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING, 2016, VOL 4, ASME Press, 2016, Vol. 4, article id 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
Keywords
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
Show others...
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
Organisations
Identifiers
ORCID iD: ORCID iD iconorcid.org/0000-0001-7816-8442

Search in DiVA

Show all publications