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Grishchenko, Dmitry
Publications (10 of 19) Show all publications
Galushin, S., Grishchenko, D. & Kudinov, P. (2019). Surrogate Model Development for Prediction of Vessel Failure Mode and Melt Release Conditions in Nordic BWR based on MELCOR code. In: : . Paper presented at ICONE-27, 27th International Conference on Nuclear Engineering, Tsukuba, Ibaraki, Japan, May 19-24, 2019.
Open this publication in new window or tab >>Surrogate Model Development for Prediction of Vessel Failure Mode and Melt Release Conditions in Nordic BWR based on MELCOR code
2019 (English)Conference paper, Published paper (Refereed)
National Category
Other Engineering and Technologies
Identifiers
urn:nbn:se:kth:diva-242348 (URN)
Conference
ICONE-27, 27th International Conference on Nuclear Engineering, Tsukuba, Ibaraki, Japan, May 19-24, 2019
Note

QC 20190130

Available from: 2019-01-30 Created: 2019-01-30 Last updated: 2019-01-30Bibliographically approved
Galushin, S., Grishchenko, D. & Kudinov, P. (2019). The Effect of the Uncertainty in Prediction of Vessel Failure Mode and Melt Release Conditions on Risk of Containment Failure due to Ex-Vessel Steam Explosion in Nordic BWR. In: : . Paper presented at ICONE-27, 27th International Conference on Nuclear Engineering, Tsukuba, Ibaraki, Japan, May 19-24, 2019.
Open this publication in new window or tab >>The Effect of the Uncertainty in Prediction of Vessel Failure Mode and Melt Release Conditions on Risk of Containment Failure due to Ex-Vessel Steam Explosion in Nordic BWR
2019 (English)Conference paper, Published paper (Refereed)
National Category
Engineering and Technology
Identifiers
urn:nbn:se:kth:diva-242346 (URN)
Conference
ICONE-27, 27th International Conference on Nuclear Engineering, Tsukuba, Ibaraki, Japan, May 19-24, 2019
Note

QC 20190130

Available from: 2019-01-30 Created: 2019-01-30 Last updated: 2019-01-30Bibliographically approved
Gallego-Marcos, I., Grishchenko, D. & Kudinov, P. (2019). Thermal stratification and mixing in a Nordic BWR pressure suppression pool. Annals of Nuclear Energy, 132, 442-450
Open this publication in new window or tab >>Thermal stratification and mixing in a Nordic BWR pressure suppression pool
2019 (English)In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 132, p. 442-450Article in journal (Refereed) Published
Abstract [en]

The pressure suppression pool of a Nordic Boiling Water Reactor (BWR) serves as a heat sink to condense steam from the primary coolant system in normal operation and accident conditions. Thermal stratification can develop in the pool when buoyancy forces overcome the momentum created by the steam injection. In this case, hot condensate forms a hot layer at the top of the pool, reducing the pool cooling and condensation capacity compared to mixed conditions. The Effective Heat Source and Effective Momentum Source (EHS/EMS) models were previously proposed to model the large-scale pool behavior during a steam injection. In this work, we use CFD code of ANSYS Fluent with the EHS/EMS models to simulate the transient behavior of a Nordic BWR pool during a steam injection through spargers. First, a validation against a Nordic BWR pool test with complete mixing is presented. Prediction of the pool behavior for other possible injection scenarios show that stratification can occur at prototypic steam injection conditions, and that the hot layer temperature above the injection point can be non-uniform. In cases with significant steam condensation inside the sparger pipes, the 95 degrees C pool temperature limit for the Emergency Core Cooling System (ECCS) pumps was reached similar to 7 h after the beginning of the blowdown.

Place, publisher, year, edition, pages
PERGAMON-ELSEVIER SCIENCE LTD, 2019
Keywords
Sparger, Relief vales, Steam injection, Condensation, CFD, Effective momentum
National Category
Atom and Molecular Physics and Optics
Identifiers
urn:nbn:se:kth:diva-259408 (URN)10.1016/j.anucene.2019.04.054 (DOI)000482247600042 ()2-s2.0-85065229097 (Scopus ID)
Note

QC 20190925

Available from: 2019-09-25 Created: 2019-09-25 Last updated: 2019-09-25Bibliographically approved
Jeltsov, M., Grishchenko, D. & Kudinov, P. (2019). Validation of Star-CCM plus for liquid metal thermal-hydraulics using TALL-3D experiment. Nuclear Engineering and Design, 341, 306-325
Open this publication in new window or tab >>Validation of Star-CCM plus for liquid metal thermal-hydraulics using TALL-3D experiment
2019 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 341, p. 306-325Article in journal (Refereed) Published
Abstract [en]

Computational Fluid Dynamics (CFD) provides means for high-fidelity 3D thermal-hydraulics analysis of Generation IV pool-type nuclear reactors. However, to be used in the decision making process a proof of code adequacy for intended application is required. This paper describes the Verification, Validation and Uncertainty Quantification (VVUQ) of a commercial CFD code Star-CCM + for forced, natural and mixed convection regimes in lead-bismuth eutectic (LBE) coolant pool flows. Code qualification is carried out according to an iterative VVUQ process aiming to reduce user effects. Validation data is produced in TALL-3D experimental facility - a 7 m high LBE loop featuring a 3D pool-type test section. Accurate prediction of mutual interaction between thermal stratification and mixing in the pool and the loop dynamics requires 3D analysis, especially during natural circulation. Solution verification is used to reduce the numerical uncertainty during code validation activities. Sensitivity Analysis (SA) is used to identify the effect of the most influential uncertain input parameters (UIPs) on numerical results. Two new visualization methods are proposed to enhance interpretation of the SA results. Dedicated experiments are performed according to the SA results to reduce the uncertainties in the most important UIPs. Automated calibration method for large CFD models is tested and demonstrated in combination with manual calibration using detailed temperature profile measurements in the pool. Calibration reveals the deficiencies in the modeling of heat losses owing to the presence of thermal bridges and other local effects in thermal insulation that are not explicitly modeled. It is demonstrated that Star-CCM + is able to predict thermal stratification and mixing phenomena in the pool type geometries. The results are supported by an Uncertainty Analysis (UA).

Place, publisher, year, edition, pages
Elsevier, 2019
Keywords
LFR, CFD, VVUQ, Pool thermal-hydraulics, Star-CCM
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-240701 (URN)10.1016/j.nucengdes.2018.11.015 (DOI)000453016700028 ()
Funder
EU, FP7, Seventh Framework Programme, FP7-249337
Note

QC 20190111

Available from: 2019-01-11 Created: 2019-01-11 Last updated: 2019-06-11Bibliographically approved
Galushin, S., Ranlöf, L., Bäckström, O., Adolfsson, Y., Grishchenko, D., Kudinov, P. & Marklund, A. R. (2018). Joint application of risk oriented accident analysis methodology and PSA level 2 to severe accident issues in Nordic BWR. In: PSAM 2018 - Probabilistic Safety Assessment and Management: . Paper presented at 14th Probabilistic Safety Assessment and Management, PSAM 2018, 16 September 2018 through 21 September 2018. International Association for Probablistic Safety Assessment and Management (IAPSAM)
Open this publication in new window or tab >>Joint application of risk oriented accident analysis methodology and PSA level 2 to severe accident issues in Nordic BWR
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2018 (English)In: PSAM 2018 - Probabilistic Safety Assessment and Management, International Association for Probablistic Safety Assessment and Management (IAPSAM) , 2018Conference paper, Published paper (Refereed)
Abstract [en]

A comprehensive and robust assessment of severe accident management effectiveness in preventing unacceptable releases is a challenge for a today’s real life PSA. This is mainly due to the fact that major uncertainty is determined by the physical phenomena and timing of the events. The static PSA is built on choosing scenario parameters to describe the accident progression sequence and typically uses a limited number of simulations in the underlying deterministic analysis. Risk Oriented Accident Analysis Methodology framework (ROAAM+) is being developed in order to enable consistent and comprehensive treatment of both epistemic and aleatory uncertainties. The framework is based on a set of deterministic models that describe different stages of the accident progression. The results are presented in terms of distributions of conditional containment failure probabilities for given combinations of the scenario parameters. This information is used for enhanced modeling in the PSA-L2. Specifically, it includes improved definitions of the sequences determined by the physical phenomena rather than stochastic failures of the equipment, improved knowledge of timing in sequences and estimation of probabilities determined by the uncertainties in the phenomena. In this work we present an example of application of the dynamic approach in a large scale PSA model and show that the integration of the ROAAM+ results and the PSA model can potentially lead to a considerable change in PSA Level 2 analysis results. 

Place, publisher, year, edition, pages
International Association for Probablistic Safety Assessment and Management (IAPSAM), 2018
Keywords
Nordic BWR, PSA L2, ROAAM, Severe accident, Accident prevention, Accidents, Probability distributions, Risk analysis, Risk assessment, Stochastic systems, Uncertainty analysis, Aleatory uncertainty, Deterministic analysis, Deterministic models, Severe accident management, Boiling water reactors
National Category
Probability Theory and Statistics
Identifiers
urn:nbn:se:kth:diva-252277 (URN)2-s2.0-85063145364 (Scopus ID)
Conference
14th Probabilistic Safety Assessment and Management, PSAM 2018, 16 September 2018 through 21 September 2018
Note

QC20190610

Available from: 2019-06-10 Created: 2019-06-10 Last updated: 2019-06-10Bibliographically approved
Phung, V.-A., Grishchenko, D., Galushin, S. & Kudinov, P. (2018). Prediction of In-Vessel Debris Bed Properties in BWR Severe Accident Scenarios using MELCOR and Neural Networks. Annals of Nuclear Energy, 461-476
Open this publication in new window or tab >>Prediction of In-Vessel Debris Bed Properties in BWR Severe Accident Scenarios using MELCOR and Neural Networks
2018 (English)In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, p. 461-476Article in journal (Refereed) Published
Abstract [en]

Severe accident management strategy in Nordic boiling water reactors (BWRs) employs ex-vessel corium debris coolability. In-vessel core degradation and relocation provide initial conditions for further accident progression. Characteristics of debris bed in the reactor vessel lower plenum affect corium interactions with the vessel structures, vessel failure and melt ejection mode. Melt release from the vessel determines conditions for ex-vessel accident progression and potential threats to containment integrity, i.e. steam explosion, debris bed formation and coolability. Outcomes of core relocation depend on the interplay between (i) accident scenarios, e.g. timing and characteristics of failure and recovery of safety systems and (ii) accident phenomena. Uncertainty analysis is necessary for comprehensive risk assessment. However, computational efficiency of system analysis codes such as MELCOR is one of the big obstacles. Another problem is that code predictions can be quite sensitive to small variations of the input parameters and numerical discretization, due to the highly non-linear feedbacks and discrete thresholds employed in the code models.

 

The goal of this work is to develop a computationally efficient surrogate model (SM) for prediction of main characteristics of corium debris in the vessel lower plenum of a Nordic BWR. Station black out scenario with a delayed power recovery is considered. The SM has been developed using artificial neural networks (ANNs). The networks were trained with a database of MELCOR solutions. The effect of the noisy data in the full model (FM) database was addressed by introducing scenario classification (grouping) according to the ranges of the output parameters. SMs using different number of scenario groups with/without weighting between predictions of different ANNs were compared. The obtained SM can be used for failure domain and failure probability analysis in the risk assessment framework for Nordic BWRs.

Keywords
Core relocation; boiling water reactor; MELCOR; surrogate model; artificial neural network.
National Category
Other Engineering and Technologies not elsewhere specified
Identifiers
urn:nbn:se:kth:diva-202955 (URN)10.1016/j.anucene.2018.06.007 (DOI)000441485700040 ()2-s2.0-85048449676 (Scopus ID)
Note

QC 20170309

Available from: 2017-03-08 Created: 2017-03-08 Last updated: 2019-04-26Bibliographically approved
Jeltsov, M., Kööp, K., Grishchenko, D. & Kudinov, P. (2018). Pre-test analysis of an LBE solidification experiment in TALL-3D. Nuclear Engineering and Design, 339, 21-38
Open this publication in new window or tab >>Pre-test analysis of an LBE solidification experiment in TALL-3D
2018 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 339, p. 21-38Article in journal (Refereed) Published
Abstract [en]

Coolant solidification is a phenomenon of potential safety importance for Liquid Metal Cooled Fast Reactors (LMFRs). Coolant solidification can affect local flow, heat transfer and lead to partial or complete blockage of the coolant flow paths jeopardizing decay heat removal function. It is also possible that reduced flow circulation may increase coolant temperature, counteract solidification and prevent complete blockage of the flow. Complex interactions between local physical phenomena of solidification and system scale natural circulation make modelling of solidification uncertain. Development and validation of adequate models requires validation grade experimental data. In this work we discuss results of analysis carried out in support of experiment development, specifically, design of a solidification test section and a test matrix for TALL-3D experimental facility (lead-bismuth eutectic (LBE) thermal-hydraulic loop). The aim of the analysis is experimental design that satisfies requirements stemming from the process of model qualification. We focus on two aspects: (i) design of solidification test section (STS) for investigation of solidification phenomena in lead-bismuth eutectic (LBE), and (ii) effect of the STS pool on the system scale behavior of the TALL-3D facility. Selection of the STS characteristics and experimental test matrix is supported using computational fluid dynamic (CFD) and system thermal-hydraulic (STH) codes. 

Place, publisher, year, edition, pages
Elsevier Ltd, 2018
Keywords
CFD, Coolant solidification, Experiment design, LMFR, STH, Bismuth, Computational fluid dynamics, Coolants, Design of experiments, Eutectics, Fast reactors, Heat transfer, Liquid metal cooled reactors, Testing, Coolant temperature, Experimental facilities, Lead-bismuth eutectics, Liquid-metal-cooled fast reactors, Natural circulation, Thermal hydraulics, Solidification
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-236566 (URN)10.1016/j.nucengdes.2018.08.014 (DOI)000446333700003 ()2-s2.0-85052758449 (Scopus ID)
Note

 Funding details: STS, Society of Thoracic Surgeons; Funding text: This work has received funding the Euratom research and training programme 2014–2018 under the grant agreement No. 654935 (SESAME). The authors are also thankful to Vincent Moreau and Manuela Profir for their contribution in the discussions and support during the STS design process. QC 20181127

Available from: 2018-11-27 Created: 2018-11-27 Last updated: 2018-11-27Bibliographically approved
Jeltsov, M., Grishchenko, D. & Kudinov, P. (2018). Validation of Star-CCM+ for liquid metal thermal-hydraulics using TALL-3D experiment. Nuclear Engineering and Design
Open this publication in new window or tab >>Validation of Star-CCM+ for liquid metal thermal-hydraulics using TALL-3D experiment
2018 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759XArticle in journal (Refereed) Submitted
National Category
Other Engineering and Technologies not elsewhere specified
Identifiers
urn:nbn:se:kth:diva-228354 (URN)
Note

QC 20180607

Available from: 2018-05-22 Created: 2018-05-22 Last updated: 2018-06-07Bibliographically 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
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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, 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
Grishchenko, D., Galushin, S., Basso, S. & Kudinov, P. (2017). Failure domain analysis and uncertainty quantification using surrogate models for steam explosion in a nordic type BWR. In: 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017: . Paper presented at 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017, 3 September 2017 through 8 September 2017. Association for Computing Machinery, Inc
Open this publication in new window or tab >>Failure domain analysis and uncertainty quantification using surrogate models for steam explosion in a nordic type BWR
2017 (English)In: 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017, Association for Computing Machinery, Inc , 2017Conference paper, Published paper (Refereed)
Abstract [en]

Sever accident mitigation strategy adopted in Nordic Boiling Water Reactors (BWRs) employs a deep water pool below the reactor vessel in order to fragment and quench core melt and provide long term cooling of the debris. One of the risk factors associated with this accident management strategy is early failure of the containment due to steam explosion. Assessment of the risk is subject to significant epistemic and aleatory uncertainties in (i) modelling of steam explosion and (ii) scenarios of melt release from the vessel and water pool conditions. High computational efficiency of the models is required for such assessment. A surrogate model (SM) approach has been previously developed using artificial neural network and the database of Texas-V code solutions for steam explosion loads in the Nordic type BWRs. In this paper we extend our surrogate model to allow analysis of steam explosion in relatively shallow water pools (>2 m), address effects of melt emissivity and resolve more accurately variation of pressure in the drywell. We provide detailed comparison of metallic vs oxidic melt release scenarios, incorporate uncertainty of the SM into modelling and analyze the sensitivity of our results to SM uncertainty. We estimate risks of containment failure with non-reinforced and reinforced hatch door and demonstrate the effect of the surrogate model uncertainty on the results. We analyze the results and develop a simplified approach for decision making considering predicted failure probabilities, expected costs and scenario frequencies. 

Place, publisher, year, edition, pages
Association for Computing Machinery, Inc, 2017
Keywords
Aleatory and epistemic uncertainties, Artificial Neural Networks, Severe accident, Surrogate model uncertainty, Boiling water reactors, Computational efficiency, Explosions, Failure (mechanical), Fuel additives, Hydraulics, Lakes, Neural networks, Nuclear reactor accidents, Reinforcement, Risk assessment, Risk perception, Steam, Aleatory uncertainty, Failure Probability, Mitigation strategy, Shallow water pools, Surrogate model, Uncertainty quantifications, Uncertainty analysis
National Category
Other Civil Engineering
Identifiers
urn:nbn:se:kth:diva-236832 (URN)2-s2.0-85052599683 (Scopus ID)
Conference
17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017, 3 September 2017 through 8 September 2017
Funder
Swedish Radiation Safety Authority
Note

QC 20181221

Available from: 2018-12-21 Created: 2018-12-21 Last updated: 2018-12-21Bibliographically approved
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