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Publications (10 of 180) Show all publications
Estévez-Albuja, S., Gallego-Marcos, I., Kudinov, P. & Jiménez, G. (2020). Modelling of a Nordic BWR containment and suppression pool behavior during a LOCA with GOTHIC 8.1. Annals of Nuclear Energy, 136, Article ID 107027.
Open this publication in new window or tab >>Modelling of a Nordic BWR containment and suppression pool behavior during a LOCA with GOTHIC 8.1
2020 (English)In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 136, article id 107027Article in journal (Refereed) Published
Abstract [en]

Boiling water reactors use the Pressure Suppression Pool (PSP) to relieve the containment pressure in case of an accident. During the event of a Loss of Coolant Accident (LOCA), drywell air and steam are injected into the PSP through blowdown pipes. This may lead to thermal stratification, which is a relevant safety issue as it leads to higher water surface temperatures than in mixed conditions and thus, to higher containment pressures. The Effective Heat (EHS) and Momentum (EMS) Source models were previously introduced to predict the effect of small-scale direct contact condensation phenomena on the large-scale pool water circulation. In this paper, the EHS/EMS models are extended by adding the effect of non-condensable gases on the chugging regime. The EHS/EMS models are implemented in the GOTHIC code to model a full-scale Nordic BWR containment under different LOCA scenarios. The results show that thermal stratification can be developed in the PSP.

Place, publisher, year, edition, pages
Elsevier, 2020
Keywords
BWR, Chugging, Effective momentum source, GOTHIC, LOCA, Pressure suppression pool, Architecture, Lakes, Loss of coolant accidents, Thermal stratification, Direct contact condensation, Momentum sources, Non-condensable gas, Safety issues, Source models, Water surface temperature, Boiling water reactors
National Category
Energy Engineering Other Physics Topics
Research subject
Physics, Nuclear Engineering
Identifiers
urn:nbn:se:kth:diva-263428 (URN)10.1016/j.anucene.2019.107027 (DOI)000498274900033 ()2-s2.0-85072249872 (Scopus ID)
Funder
EU, Horizon 2020
Note

QC 20191205

Available from: 2019-12-05 Created: 2019-12-05 Last updated: 2019-12-12Bibliographically approved
Galushin, S. & Kudinov, P. (2020). Sensitivity and Uncertainty Analysis of the Vessel Lower Head Failure Mode and Melt Release Conditions in Nordic BWR using MELCOR Code. Annals of Nuclear Energy, 135, Article ID 106976.
Open this publication in new window or tab >>Sensitivity and Uncertainty Analysis of the Vessel Lower Head Failure Mode and Melt Release Conditions in Nordic BWR using MELCOR Code
2020 (English)In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 135, article id 106976Article in journal (Refereed) Published
Abstract [en]

Nordic boiling water reactors (BWR) employ ex-vessel debris coolability as a severe accident management (SAM) strategy. Melt release mode into a deep water pool located in the lower drywell is a major source of uncertainty for success of such strategy. The main goal of this work is to identify the major contributors to the uncertainty in prediction of vessel failure mode, timing and melt release conditions in Nordic BWR using MELCOR severe accident analysis code. There is a forest of control rod guide tubes (CRGTs) and instrumentation guide tubes (IGTs) in the lower head of a BWR. Failure of the penetrations is considered in this work along with the gross creep rupture of the vessel lower head. Modelling of penetrations failure in MELCOR code is based on parametric models, allowing a user to select different assumptions, such as temperature threshold for penetrations failure, modelling of penetrations assembly and respective failure modes, mode of solid and liquid debris ejection from the vessel. In this work we perform sensitivity analysis to the MELCOR modelling options and sensitivity parameters in unmitigated station blackout scenario (SBO). Results of analysis suggest that (i) vessel breach due to penetration failure is predicted to occur before creep-rupture failure of the vessel lower head, (ii) the mode of debris ejection from the vessel has the dominant effect on the likelihood of creep-rupture failure of the vessel lower head and debris ejection rate from the vessel.

National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-242352 (URN)10.1016/j.anucene.2019.106976 (DOI)
Note

QC 20191204

Available from: 2019-01-30 Created: 2019-01-30 Last updated: 2019-12-04Bibliographically approved
Galushin, S. & Kudinov, P. (2020). Sensitivity and uncertainty analysis of the vessel lower head failure mode and melt release conditions in Nordic BWR using MELCOR code. Annals of Nuclear Energy, 135, Article ID 106976.
Open this publication in new window or tab >>Sensitivity and uncertainty analysis of the vessel lower head failure mode and melt release conditions in Nordic BWR using MELCOR code
2020 (English)In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 135, article id 106976Article in journal (Refereed) Published
Abstract [en]

Nordic boiling water reactors (BWR) employ ex-vessel debris coolability as a severe accident management (SAM) strategy. Melt release mode into a deep water pool located in the lower drywell is a major source of uncertainty for success of such strategy. The main goal of this work is to identify the major contributors to the uncertainty in prediction of vessel failure mode, timing and melt release conditions in Nordic BWR using MELCOR severe accident analysis code. There is a forest of control rod guide tubes (CRGTs) and instrumentation guide tubes (IGTs) in the lower head of a BWR. Failure of the penetrations is considered in this work along with the gross creep rupture of the vessel lower head. Modelling of penetrations failure in MELCOR code is based on parametric models, allowing a user to select different assumptions, such as temperature threshold for penetrations failure, modelling of penetrations assembly and respective failure modes, mode of solid and liquid debris ejection from the vessel. In this work we perform sensitivity analysis to the MELCOR modelling options and sensitivity parameters in unmitigated station blackout scenario (SBO). Results of analysis suggest that (i) vessel breach due to penetration failure is predicted to occur before creep-rupture failure of the vessel lower head, (ii) the mode of debris ejection from the vessel has the dominant effect on the likelihood of creep-rupture failure of the vessel lower head and debris ejection rate from the vessel.

Place, publisher, year, edition, pages
Elsevier Ltd, 2020
Keywords
MELCOR, Melt release, Nordic BWR, Sensitivity, Uncertainty, Vessel failure, Accidents, Boiling water reactors, Codes (symbols), Creep, Debris, Failure modes, Sensitivity analysis, Uncertainty analysis, Failure (mechanical)
National Category
Energy Engineering Other Physics Topics
Research subject
Physics, Nuclear Engineering
Identifiers
urn:nbn:se:kth:diva-263441 (URN)10.1016/j.anucene.2019.106976 (DOI)000496898500044 ()2-s2.0-85070926736 (Scopus ID)
Funder
Swedish Radiation Safety Authority
Note

QC 20191205

Available from: 2019-12-05 Created: 2019-12-05 Last updated: 2019-12-17Bibliographically approved
Galushin, S. & Kudinov, P. (2019). Analysis of the Effect of MELCOR Modelling Parameters on In-Vessel Accident Progression in Nordic BWR. Nuclear Engineering and Design, 350, 243-258
Open this publication in new window or tab >>Analysis of the Effect of MELCOR Modelling Parameters on In-Vessel Accident Progression in Nordic BWR
2019 (English)In: Nuclear Engineering and Design, Vol. 350, p. 243-258Article in journal (Refereed) Published
Abstract [en]

Nordic Boiling Water Reactors (BWRs) rely on the flooding of the lower drywell (LDW) as a severe accident management (SAM) strategy. The termination of a SA is achieved by fragmenting and quenching of the melt released from the vessel. Success of SAM strategy depends on melt release and water pool conditions. The characteristics of the melt release are the major source of uncertainty in quantification of the risk of SAM failure. Vessel failure and melt release modes are subject to aleatory and epistemic uncertainties at the in-vessel accident progression stage. In this work we focus on predicting the properties of debris relocated to the lower plenum using MELCOR code. We address the effect of epistemic uncertainty in modeling parameters and models in the MELCOR code in different severe accident scenarios on main characteristics of the in-vessel accident progression in Nordic BWRs. Sensitivity analysis is performed to rank the importance of MELCOR modelling parameters and the effect of different MELCOR models is addressed by using different versions of the code. The results provide valuable insights regarding the effect of MELCOR models, modelling parameters and sensitivity coefficients on code predictions.

Place, publisher, year, edition, pages
Elsevier, 2019
National Category
Engineering and Technology
Identifiers
urn:nbn:se:kth:diva-242350 (URN)10.1016/j.nucengdes.2019.04.040 (DOI)000470690900024 ()2-s2.0-85066260379 (Scopus ID)
Note

QC 20191204

Available from: 2019-01-30 Created: 2019-01-30 Last updated: 2019-12-04Bibliographically approved
Galushin, S. & Kudinov, P. (2019). Analysis of the Effect of Severe Accident Scenario on Debris Properties in Lower Plenum of Nordic BWR Using Different Versions of MELCOR Code. Science and Technology of Nuclear Installations, Article ID 5310808.
Open this publication in new window or tab >>Analysis of the Effect of Severe Accident Scenario on Debris Properties in Lower Plenum of Nordic BWR Using Different Versions of MELCOR Code
2019 (English)In: Science and Technology of Nuclear Installations, ISSN 1687-6075, E-ISSN 1687-6083, article id 5310808Article in journal (Refereed) Published
Abstract [en]

Nordic Boiling Water Reactors (BWRs) employ ex-vessel debris coolability as a severe accident management strategy (SAM). Core melt is released into a deep pool of water where formation of noncoolable debris bed and ex-vessel steam explosion can pose credible threats to containment integrity. Success of the strategy depends on the scenario of melt release from the vessel that determines the melt-coolant interaction phenomena. The melt release conditions are determined by the in-vessel phase of severe accident progression. Specifically, properties of debris relocated into the lower plenum have influence on the vessel failure and melt release mode. In this work we use MELCOR code for prediction of the relocated debris. Over the years, many code modifications have been made to improve prediction of severe accident progression in light-water reactors. The main objective of this work is to evaluate the effect of models and best practices in different versions of MELCOR code on the in-vessel phase of different accident progression scenarios in Nordic BWR. The results of the analysis show that the MELCOR code versions 1.86 and 2.1 generate qualitatively similar results. Significant discrepancy in the timing of the core support failure and relocated debris mass in the MELCOR 2.2 compared to the MELCOR 1.86 and 2.1 has been found for a domain of scenarios with delayed time of depressurization. The discrepancies in the results can be explained by the changes in the modeling of degradation of the core components and changes in the Lipinski dryout model in MELCOR 2.2.

Place, publisher, year, edition, pages
Hindawi Publishing Corporation, 2019
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-251723 (URN)10.1155/2019/5310808 (DOI)000466590400001 ()2-s2.0-85065230309 (Scopus ID)
Note

QC 20190520

Available from: 2019-05-20 Created: 2019-05-20 Last updated: 2019-05-29Bibliographically approved
Galushin, S. & Kudinov, P. (2019). Analysis of the Effect of Severe Accident Scenario on Debris Properties in Lower Plenum of Nordic BWR Using Different Versions of MELCOR Code. Science and Technology of Nuclear Installations
Open this publication in new window or tab >>Analysis of the Effect of Severe Accident Scenario on Debris Properties in Lower Plenum of Nordic BWR Using Different Versions of MELCOR Code
2019 (English)In: Science and Technology of Nuclear InstallationsArticle in journal (Refereed) Submitted
National Category
Other Engineering and Technologies
Identifiers
urn:nbn:se:kth:diva-242349 (URN)
Note

QC 20190130

Available from: 2019-01-30 Created: 2019-01-30 Last updated: 2019-08-28Bibliographically approved
Yakush, S. E. & Kudinov, P. (2019). On the evaluation of dryout conditions for a heat-releasing porous bed in a water pool. International Journal of Heat and Mass Transfer, 895-905
Open this publication in new window or tab >>On the evaluation of dryout conditions for a heat-releasing porous bed in a water pool
2019 (English)In: International Journal of Heat and Mass Transfer, ISSN 0017-9310, E-ISSN 1879-2189, p. 895-905Article in journal (Refereed) Published
Abstract [en]

This work is motivated by the problem of restraining temperature escalation inside a porous heat-releasing media submerged in a pool of liquid coolant. When coolant temperature reaches saturation, boiling begins in the bulk of the porous bed, with void generation rate determined by the heating power. Amount of void determines hydrostatic pressure difference that drives natural circulation of two-phase flow through the porous material. At a certain critical value of the heat release rate, the driving head cannot overcome drag of the two phase porous media flow, which results in complete evaporation of coolant in some zone. Temperature of material in the dry zone increases significantly due to deterioration of heat exchange with single phase vapor flow in comparison with boiling flow heat transfer. The paper considers the problem of determining the critical conditions for onset of dryout in a heat-releasing porous bed of an arbitrary shape. The well-known one-dimensional problem for a flat top-flooded bed is revisited, and the functional form of the dryout boundary (expressed as the dryout heat flux, DHF) is derived using non-dimensional parameters. Asymptotic behavior of the solution is analyzed, and, by the method of asymptotic interpolation, a surrogate model is proposed consisting of three single-argument, non-dimensional functions. It is shown that such a model provides acceptable accuracy even in the cases where complete similarity of solutions is not achieved. The results obtained provide important insights into the physics of the problem, reduce the number of free parameters, and enable fast evaluation of dryout conditions without the need of numerical solution of algebraic equations involved in the exact formulation. The ultimate goal of the surrogate model development, i.e. its application to multidimensional configurations, is discussed.

Place, publisher, year, edition, pages
Elsevier, 2019
Keywords
Coolability, Dryout, Porous bed, Severe accident, Two-phase flow
National Category
Energy Engineering
Identifiers
urn:nbn:se:kth:diva-246442 (URN)10.1016/j.ijheatmasstransfer.2019.01.083 (DOI)000462418300076 ()2-s2.0-85060648383 (Scopus ID)
Note

QC 20190326

Available from: 2019-03-26 Created: 2019-03-26 Last updated: 2019-04-24Bibliographically approved
Moreau, V., Profir, M., Alemberti, A., Frignani, M., Merli, F., Belka, M., . . . Martelli, D. (2019). Pool CFD modelling: lessons from the SESAME project. Nuclear Engineering and Design, 355, Article ID UNSP 110343.
Open this publication in new window or tab >>Pool CFD modelling: lessons from the SESAME project
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2019 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 355, article id UNSP 110343Article in journal (Refereed) Published
Abstract [en]

The current paper describes the Computational Fluid-Dynamics (CFD) modelling of Heavy Liquid Metal (HLM) flows in a pool configuration and in particular how this is approached within the Horizon 2020 SESAME project. SESAME's work package structure, based on a systematic approach of redundancy and diversification, is explained along with its motivation. The main achievements obtained and the main lessons learned during the project are illustrated. The paper focuses on the strong coupling between the experimental activities and CFD simulations performed within the SESAME project. Two different HLM fluids are investigated: pure lead and Lead-Bismuth Eutectic. The objective is to make CFD a valid instrument used during the design of safe and innovative Gen-IV nuclear plants. Some effort has also been devoted to Proper Orthogonal Decomposition with Galerkin projection modelling (POD-Galerkin), a reduced order model suited for Uncertainty Quantification that operates by post-processing CFD results. Assessment of Uncertainty highly improves the reliability of CFD simulations. Dedicated experimental campaigns on heavily instrumented facilities have been conceived with the specific objective to build a series of datasets suited for the calibration and validation of the CFD modelling. In pool configuration, the attention is focused on the balance between conductive and convective heat transfer phenomena, on transient test-cases representative of incidental scenarios and on the possible occurrence of solidification phenomena. Four test sections have been selected to generate the datasets: (i) the CIRCE facility from ENEA, (ii) the TALL-3D pool test section from KTH, (iii) the TALL-3D Solidification Test Section (STS) from KTH and (iv) the SESAME Stand facility from CVR. While CIRCE and TALL-3D were existing facilities, the STS and SESAME Stand facility have been conceived, built and operated within the project, heavily relying on the use of CFD support. Care has been taken to ensure that almost all tasks were performed by at least two partners. Specific examples are given on how this strategy has allowed to uncover flaws and overcome pitfalls. Furthermore, an overview of the performed work and the achieved results is presented, as well as remaining or new uncovered issues. Finally, the paper is concluded with a description of one of the main goals of the SESAME project: the construction of the Gen-IV ALFRED CFD model and an investigation of its general circulation.

Place, publisher, year, edition, pages
ELSEVIER SCIENCE SA, 2019
Keywords
CFD, Numerical simulation, Pool thermal-hydraulics, Lead solidification, Gen-IV reactors
National Category
Mechanical Engineering
Identifiers
urn:nbn:se:kth:diva-264342 (URN)10.1016/j.nucengdes.2019.110343 (DOI)000493898800029 ()2-s2.0-85072246527 (Scopus ID)
Note

QC 20191126

Available from: 2019-11-26 Created: 2019-11-26 Last updated: 2019-11-26Bibliographically approved
Gallego-Marcos, I., Kudinov, P., Villanueva, W., Kapulla, R., Paranjape, S., Paladino, D., . . . Kotro, E. (2019). Pool stratification and mixing induced by steam injection through spargers: CFD modelling of the PPOOLEX and PANDA experiments. Nuclear Engineering and Design, 347, 67-85
Open this publication in new window or tab >>Pool stratification and mixing induced by steam injection through spargers: CFD modelling of the PPOOLEX and PANDA experiments
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2019 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 347, p. 67-85Article in journal (Refereed) Published
Abstract [en]

Spargers are multi-hole injection pipes used in Boiling Water Reactors (BWR) and Generation III/III+ Pressurized Water Reactors (PWR) to condense steam in large water pools. During the steam injection, high pool surface temperatures induced by thermal stratification can lead to higher containment pressures compared with completely mixed pool conditions, the former posing a threat for plant safety. The Effective Heat Source (EHS) and Effective Momentum Source (EMS) models were previously developed and validated for the modelling of a steam injection through blowdown pipes. The goal of this paper is to extend the EHS/EMS model capabilities towards steam injection through multi-hole spargers. The models are implemented in ANSYS Fluent 17.0 Computational Fluid Dynamics (CFD) code and calibrated against the spargers experiments performed in the PPOOLEX and PANDA facilities, analysed by the authors in Gallego-Marcos et al. (2018b). CFD modelling guidelines are established for the adequate simulation of the pool behaviour. A new correlation is proposed to model the turbulent production and dissipation caused by buoyancy. Sensitivity studies addressing the effect of different assumptions on the effective momentum magnitude, profile, angle and turbulence are presented. Calibration of the effective momentum showed an inverse proportionality to the sub-cooling. Differences between the effective momentum calibrated for PPOOLEX and PANDA are discussed. Analysis of the calculated flow above the cold stratified layer showed that the erosion of the layer is induced by the action of turbulence rather than mean shear flow.

Place, publisher, year, edition, pages
ELSEVIER SCIENCE SA, 2019
Keywords
Thermocline, Turbulence production buoyancy, Richardson, C-3e coefficient, Oscillatory bubble regime
National Category
Energy Engineering
Identifiers
urn:nbn:se:kth:diva-251271 (URN)10.1016/j.nucengdes.2019.03.011 (DOI)000465217900008 ()2-s2.0-85063478019 (Scopus ID)
Note

QC 20190514

Available from: 2019-05-14 Created: 2019-05-14 Last updated: 2019-05-29Bibliographically approved
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
Organisations
Identifiers
ORCID iD: ORCID iD iconorcid.org/0000-0002-0683-9136

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