Change search
Link to record
Permanent link

Direct link
BETA
Alternative names
Publications (10 of 86) Show all publications
Chen, Y., Zhang, H., Villanueva, W., Ma, W. & Bechta, S. (2019). A sensitivity study of MELCOR nodalization for simulation of in-vessel severe accident progression in a boiling water reactor. Nuclear Engineering and Design, 343, 22-37
Open this publication in new window or tab >>A sensitivity study of MELCOR nodalization for simulation of in-vessel severe accident progression in a boiling water reactor
Show others...
2019 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 343, p. 22-37Article in journal (Refereed) Published
Abstract [en]

This paper presents a sensitivity study of MELCOR nodalization for simulation of postulated severe accidents in a Nordic boiling water reactor, with the objective to address the nodal effect on in-vessel accident progression, including thermal-hydraulic response, core degradation and relocation, hydrogen generation, source term release, melt behavior and heat transfer in the lower head, etc. For this purpose, three meshing schemes (coarse, medium and fine) of the COR package of MELCOR are chosen to analyze two severe accident scenarios: station blackout (SBO) accident and large break loss-of-coolant accident (LOCA) combined with station blackout. The comparative results of the MELCOR simulations show that the meshing schemes mainly affect the core degradation and relocation to the lower head of the reactor pressure vessel: the fine mesh leads to a delayed leveling process of a heap-like debris bed in the lower head, and a later breach of the vessel. The simulations with fine mesh also provide more detailed distributions of corium mass and temperature, as well as heat flux which is an important parameter in qualification assessment of the In-Vessel Melt Retention (IVR) strategy.

Place, publisher, year, edition, pages
ELSEVIER SCIENCE SA, 2019
Keywords
Boiling water reactor, Reactor safety, Severe accident, MELCOR simulation, Mesh sensitivity analysis
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-244082 (URN)10.1016/j.nucengdes.2018.12.011 (DOI)000456923500003 ()2-s2.0-85059233155 (Scopus ID)
Note

QC 20190219

Available from: 2019-02-19 Created: 2019-02-19 Last updated: 2019-04-29Bibliographically approved
Manickam, L., Guo, Q., Ma, W. & Bechta, S. (2019). An experimental study on the intense intense heat transfer and phase change during melt and water interactions. Experimental heat transfer, 32(3), 251-266
Open this publication in new window or tab >>An experimental study on the intense intense heat transfer and phase change during melt and water interactions
2019 (English)In: Experimental heat transfer, ISSN 0891-6152, E-ISSN 1521-0480, Vol. 32, no 3, p. 251-266Article in journal (Refereed) Published
Abstract [en]

Accidental contact between hot melt and cold water poses fatal hazard in several industries. Vapor explosion during melt-water contact in nuclear power plant accident can result in catastrophic containment failure. The fast transient phenomena as vapor explosion is not comprehensively understood despite several advances in research. It is not clear why certain parameters of melt and water exhibit differences in fragmentation behavior. To examine the influential parameters, we perform a series of experiments. The interactions between melt and water is visualized by high-speed video and X-ray radiograph.

Place, publisher, year, edition, pages
Taylor & Francis Group, 2019
National Category
Other Physics Topics
Identifiers
urn:nbn:se:kth:diva-239897 (URN)10.1080/08916152.2018.1505786 (DOI)000462344000004 ()2-s2.0-85052292697 (Scopus ID)
Note

QC 20181214

Available from: 2018-12-05 Created: 2018-12-05 Last updated: 2019-05-03Bibliographically approved
Bechta, S., Ma, W., Miassoedov, A., Journeau, C., Okamoto, K., Manara, D., . . . Schyns, M. (2019). On the EU-Japan roadmap for experimental research on corium behavior. Annals of Nuclear Energy, 124, 541-547
Open this publication in new window or tab >>On the EU-Japan roadmap for experimental research on corium behavior
Show others...
2019 (English)In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 124, p. 541-547Article in journal (Refereed) Published
Abstract [en]

A joint research roadmap between Europe and Japan has been developed in severe accident field of light water reactors, focusing particularly on reactor core melt (corium) behavior. The development of this roadmap is one of the main targets of the ongoing EU project SAFEST. This paper presents information about ongoing severe accident studies in the area of corium behavior, rationales and comparison of research priorities identified in different projects and documents, expert ranking of safety issues, and finally the research areas and topics and their priorities suggested for the EU Japan roadmap and future bilateral collaborations. These results provide useful guidelines for (i) assessment of long-term goals and proposals for experimental support needed for proper understanding, interpretation and learning lessons of the Fukushima accident; (ii) analysis of severe accident phenomena; (iii) development of accident prevention and mitigation strategies, and corresponding technical measures; (iv) study of corium samples in European and Japanese laboratories; and (v) preparation of Fukushima site decommissioning.

Place, publisher, year, edition, pages
PERGAMON-ELSEVIER SCIENCE LTD, 2019
Keywords
Light water reactor, Severe accident, Corium, Accident phenomena, Research priority, Experimental facility
National Category
Energy Engineering
Identifiers
urn:nbn:se:kth:diva-240340 (URN)10.1016/j.anucene.2018.10.019 (DOI)000451498100046 ()2-s2.0-85055353709 (Scopus ID)
Note

QC 20181218

Available from: 2018-12-18 Created: 2018-12-18 Last updated: 2018-12-18Bibliographically approved
Manickam, L., Guo, Q., Komlev, A. A., Ma, W. & Bechta, S. (2019). Oxidation of molten zirconium droplets in water. Nuclear Engineering and Design, 354, Article ID UNSP 110225.
Open this publication in new window or tab >>Oxidation of molten zirconium droplets in water
Show others...
2019 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 354, article id UNSP 110225Article in journal (Refereed) Published
Abstract [en]

Zirconium, which is used as the cladding material of nuclear fuel rods in LWRs, can react with steam in the case of a core meltdown accident. This results in the release of hydrogen which poses a significant risk of hydrogen explosion. Oxidation of Zr occurs either during the core degradation when the steam flows over the hot fuel rod surfaces or during an FCI when molten corium falls into a water pool (e.g. in the lower head). An experimental study was performed at the MISTEE-OX facility at KTH to observe and quantify the oxidation of molten zirconium droplets in a water pool. During the experimental runs, single droplets of molten zirconium were discharged into a subcooled water pool and the dynamic events were recorded using a high-speed camera. The bubble dynamics indicate a process of cyclic oxidation and hydrogen release from the rear periphery of a droplet while it is quenched in the water. The debris (solidified state of the droplet) after each run was collected for compositional and microstructural analysis via SEM/EDS. The obtained data were employed to estimate the oxidation fractions of the droplets and the results revealed several interesting insights into the oxidation phenomenon of the Zr melt. The water subcooling was observed to have a significant influence on the oxidation: the degree of oxidation decreased with increase in the water subcooling. Furthermore, the degree of oxidation was also influenced by the depth into the debris, forming compounds whose oxygen content decreases from the outer surface towards the core of the debris. Therefore, the qualitative and quantitative results presented in this paper are important in the context of developing a phenomenological understanding of the oxidation behaviour of zirconium melt during the FCI as well as to improve and validate the currently available models implemented in the state-of-art steam explosion codes.

Place, publisher, year, edition, pages
ELSEVIER SCIENCE SA, 2019
Keywords
Reactor safety, Severe accident, Fuel coolant interaction, Zirconium oxidation, Hydrogen production
National Category
Other Engineering and Technologies
Identifiers
urn:nbn:se:kth:diva-257790 (URN)10.1016/j.nucengdes.2019.110225 (DOI)000481647400019 ()2-s2.0-85069953862 (Scopus ID)
Note

QC 20190913

Available from: 2019-09-13 Created: 2019-09-13 Last updated: 2019-09-13Bibliographically approved
Yu, P., Ma, W., Villanueva, W., Karbojian, A. & Bechta, S. (2019). Validation of a thermo-fluid-structure coupling approach for RPV creep failure analysis against FOREVER-EC2 experiment. Annals of Nuclear Energy, 133, 637-648
Open this publication in new window or tab >>Validation of a thermo-fluid-structure coupling approach for RPV creep failure analysis against FOREVER-EC2 experiment
Show others...
2019 (English)In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 133, p. 637-648Article in journal (Refereed) Published
Abstract [en]

The failure of reactor pressure vessel (RPV) during a severe accident of light water reactors is a thermal fluid-structure interaction (FSI) problem which involves melt pool heat transfer and creep deformation of the RPV. The present study is intended to explore a reliable coupling approach of thermo-fluid-structure analyses which will not only be able to reflect the transient thermal FSI feature, but also apply the advanced models and computational platforms to melt pool convection and structural mechanics, so as to improve simulation fidelity. For this purpose, the multi-physics platform of ANSYS encompassing Fluent and Structural capabilities was employed to simulate the fluid dynamics and structural mechanics in a coupled manner. In particular, the FOREVER-EC2 experiment was chosen to validate the coupling approach. The natural convection in melt pool was modeled with the SST turbulence model with a well-resolved boundary layer, while the creep deformation for the vessel made of 16MND5 steel was analyzed with a new three-stage creep model (modified theta projection model). A utility tool was introduced to transfer the transient thermal loads from Fluent to Structural which minimizes the user effort in performing the coupled analysis. The validation work demonstrated the well-posed capability of the coupling approach for prediction of the key parameters of interest, including temperature profile, total displacement of vessel bottom point and the evolution of wall thickness profile in the experiment. Ltd. All rights reserved.

Place, publisher, year, edition, pages
PERGAMON-ELSEVIER SCIENCE LTD, 2019
Keywords
Reactor pressure vessel, Creep failure, Thermal fluid-structure interaction, Computational fluid dynamics, Computational structural mechanics, Coupled analysis
National Category
Energy Engineering
Research subject
Energy Technology; Energy Technology
Identifiers
urn:nbn:se:kth:diva-260983 (URN)10.1016/j.anucene.2019.06.067 (DOI)000484649800061 ()2-s2.0-85068784394 (Scopus ID)
Note

QC 20191010

Available from: 2019-10-10 Created: 2019-10-10 Last updated: 2019-11-26Bibliographically approved
Fichot, F., Carenini, L., Villanueva, W. & Bechta, S. (2018). A revised methodology to assess in-vessel retention strategy for high-power reactors. In: PROCEEDINGS OF THE 26TH INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING, 18, VOL 7: . Paper presented at 2018 26th International Conference on Nuclear Engineering, ICONE 2018; London; United Kingdom; 22 July 2018 through 26 July 2018. The American Society of Mechanical Engineers, 7
Open this publication in new window or tab >>A revised methodology to assess in-vessel retention strategy for high-power reactors
2018 (English)In: PROCEEDINGS OF THE 26TH INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING, 18, VOL 7, The American Society of Mechanical Engineers , 2018, Vol. 7Conference paper, Published paper (Refereed)
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 design and SAM guidances (SAMGs) of several operating small and medium capacity LWRs (reactors below 500 MWe, e.g. VVER440) and is part of the SAMG strategies for some Gen III+ PWRs of higher power such as the AP1000 or the APR1400. However, the demonstration of IVR feasibility for high power reactors requires using less conservative models as the safety margins are reduced. 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 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 W/m(2)) whereas the first type provides the lowest heat fluxes (around 500 kW/m(2)) but this model is not realistic due to the immiscibility of molten steel with oxide melt. Obviously, there is a need to reach a consensus about best estimate practices for IVR assessment to be used in the major codes used 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 can reach, in many cases, values which are well above 1 MW/m(2). This could reduce the residual thickness of the vessel considerably and threaten its strength and integrity. Therefore, it is clear that the safety demonstration of IVR in 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 the focusing effect or significantly overheated metal, e.g. after oxide and metal layer inversion. Both situations are illustrated in this paper. The demonstration also requires an accurate thermo-mechanical analysis of the ablated vessel. The standard approach based on "yield stress" (plastic behaviour) is compared with more detailed calculations made on realistic profiles of ablated vessels. The validity of the standard approach is discussed. The current approach followed by many 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, e.g. associated with molten pool transient behaviour, 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. Some elements that might help to reach such harmonization are proposed in this paper, with a preliminary revision of the methodology that could be used to address the IVR issue. In the proposed revised methodology, the safety criterion is not based on a comparison of the heat flux and the Critical Heat Flux (CHF) profiles as in the current approaches but on the minimum vessel thickness reached after ablation and the maximum pressure load 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 is more straightforward to be used in a deterministic approach.

Place, publisher, year, edition, pages
The American Society of Mechanical Engineers, 2018
Series
International Conference on Nuclear Engineering, Proceedings, ICONE ; 7
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-248375 (URN)10.1115/ICONE26-82248 (DOI)000461413000039 ()2-s2.0-85062420243 (Scopus ID)978-0-7918-5151-7 (ISBN)
Conference
2018 26th International Conference on Nuclear Engineering, ICONE 2018; London; United Kingdom; 22 July 2018 through 26 July 2018
Note

QC 20190409

Available from: 2019-04-09 Created: 2019-04-09 Last updated: 2019-04-09Bibliographically approved
Yu, P., Villanueva, W., Galushin, S., Ma, W. & Bechta, S. (2018). Coupled Thermo-Mechanical Creep Analysis For A Nordic BWR Lower Head Using Non-Homogeneous Debris Bed Configuration From MELCOR. In: : . Paper presented at 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12).
Open this publication in new window or tab >>Coupled Thermo-Mechanical Creep Analysis For A Nordic BWR Lower Head Using Non-Homogeneous Debris Bed Configuration From MELCOR
Show others...
2018 (English)Conference paper, Published paper (Refereed)
Abstract [en]

We present a coupled thermo-mechanical creep analysis for a Nordic BWR lower head with a non-homogeneous debris bed configuration generated with MELCOR code. A one-way coupling approach was adopted which uses the Phase-Change Effective Convectivity Model implemented in Fluent to simulate the convective heat transfer in the melt pool and the ANSYS Mechanical to simulate the vessel wall deformation induced by the thermal and mechanical load from the debris. An initial non-homogeneity of debris bed was estimated using MELCOR core relocation simulation results specifying the mass of each component (UO2/Zr/ZrO2/SS/SSOX) and temperature in each MELCOR cell of the lower head. A mapping scheme was designed to transfer this non-homogeneities debris bed to Fluent through User Defined Functions. All components were locally treated in Fluent as one ideal phase by averaging the weights of element-specific mass fractions inside each cell. Material properties (density, heat capacity, etc.) and volumetric heat in the debris were both spatial- and temperature-dependent. Meanwhile, additional simulations using homogeneous debris bed configuration but with the same amount of mass compositions were run for comparison. Results including temperature escalation, vessel failure timing and location were analyzed and compared.

Keywords
Boiling Water Reactor, Severe Accident, Vessel Failure, Thermo-Mechanical Analysis
National Category
Energy Engineering
Identifiers
urn:nbn:se:kth:diva-259590 (URN)
Conference
12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12)
Note

QCR 20191015

Available from: 2019-09-18 Created: 2019-09-18 Last updated: 2019-10-15Bibliographically approved
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
Kim, H. Y., Bechta, S., Foit, J. & Hong, S. W. (2018). In-Vessel and Ex-Vessel Corium Stabilization in Light Water Reactor. Science and Technology of Nuclear Installations, Article ID 3918150.
Open this publication in new window or tab >>In-Vessel and Ex-Vessel Corium Stabilization in Light Water Reactor
2018 (English)In: Science and Technology of Nuclear Installations, ISSN 1687-6075, E-ISSN 1687-6083, article id 3918150Article in journal, Editorial material (Other academic) Published
Place, publisher, year, edition, pages
HINDAWI LTD, 2018
National Category
Other Engineering and Technologies
Identifiers
urn:nbn:se:kth:diva-238170 (URN)10.1155/2018/3918150 (DOI)000447449900001 ()2-s2.0-85055513307 (Scopus ID)
Note

QC 20181106

Available from: 2018-11-06 Created: 2018-11-06 Last updated: 2018-11-06Bibliographically approved
Lopukh, D. B., Skrigan, I. N., Vavilov, A. V., Martynov, A. P., Bechta, S. & Komlev, A. A. (2018). Numerical Simulation of Induction Heating for Molten Pool heat Transfer Experiments in Slice Geometry. In: 2018 INTERNATIONAL SCIENTIFIC MULTI-CONFERENCE ON INDUSTRIAL ENGINEERING AND MODERN TECHNOLOGIES (FAREASTCON): . Paper presented at 2018 International Multi-Conference on Industrial Engineering and Modern Technologies (FarEastCon). IEEE
Open this publication in new window or tab >>Numerical Simulation of Induction Heating for Molten Pool heat Transfer Experiments in Slice Geometry
Show others...
2018 (English)In: 2018 INTERNATIONAL SCIENTIFIC MULTI-CONFERENCE ON INDUSTRIAL ENGINEERING AND MODERN TECHNOLOGIES (FAREASTCON), IEEE , 2018Conference paper, Published paper (Refereed)
Abstract [en]

Modeling results of induction heating of experimental facility developed in the frame of European IVMR project "In-Vessel Melt Retention strategy for high power nuclear reactor" have been presented. Facility is intended for heat transfer study in the stratified pool with unmixable liquid layers. Test section consists of a slice-type vessel having a semicircular geometry representing the lower head of the reactor pressure vessel. Electrical parameters of induction heating system and design of inductor have been determined based on the modelling results to ensure the most uniform heating of the bottom layer representing heat-generating oxidic melt. A design of an electromagnetic shield has been developed to minimize electromagnetic heating of the facility vessel and influence of Lorenz forces on the natural convection of the top metal layer.

Place, publisher, year, edition, pages
IEEE, 2018
Keywords
induction heating, numerical modeling, corium, melt pool, in vessel melt retention, Lorenz forces, severe accident
National Category
Energy Engineering
Identifiers
urn:nbn:se:kth:diva-246303 (URN)10.1109/FarEastCon.2018.8602779 (DOI)000459848500196 ()2-s2.0-85061720333 (Scopus ID)
Conference
2018 International Multi-Conference on Industrial Engineering and Modern Technologies (FarEastCon)
Note

QC 20190320

Available from: 2019-03-20 Created: 2019-03-20 Last updated: 2019-05-13Bibliographically approved
Organisations
Identifiers
ORCID iD: ORCID iD iconorcid.org/0000-0001-7816-8442

Search in DiVA

Show all publications