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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
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 DesignArticle in journal (Refereed) Submitted
National Category
Engineering and Technology
Identifiers
urn:nbn:se:kth:diva-242350 (URN)
Note

QC 20190130

Available from: 2019-01-30 Created: 2019-01-30 Last updated: 2019-08-28Bibliographically 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
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
Show others...
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
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. & Kudinov, P. (2018). Analysis of the effect of severe accident scenario on the vessel lower head failure in Nordic BWR using MELCOR code. 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 >>Analysis of the effect of severe accident scenario on the vessel lower head failure in Nordic BWR using MELCOR code
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]

Severe accident management (SAM) in Nordic boiling water reactors (BWR) relies on ex-vessel core debris coolability. In case of core melt and vessel failure, melt is poured into a deep pool of water located under the reactor. The melt is expected to fragment, quench, and form a debris bed, coolable by natural circulation of water. Success of the strategy is contingent upon melt release conditions from the vessel which determine (i) properties and thus coolability of the debris bed, and (ii) potential for energetic steam explosion. Both non-coolable debris bed and steam explosion are credible threats to containment integrity. Melt release conditions are the major source of uncertainty in quantification of the risk of containment failure in Nordic BWRs using ROAAM+ Framework. The melt release conditions, including in-vessel\ex-vessel pressure, lower drywell pool depth and temperature, are affected by aleatory (severe accident scenario) and epistemic (modeling) uncertainties. In this work we use MELCOR code to perform the analysis of the effects of Severe accident scenarios and modelling options in MELCOR on the properties of debris relocated to the lower head, the time and the mode of vessel lower head failure. We identify the most influential uncertain factors and discuss the needs for improvements in the modeling approaches. 

Place, publisher, year, edition, pages
International Association for Probablistic Safety Assessment and Management (IAPSAM), 2018
Keywords
MELCOR, Nordic BWR, ROAAM, Severe accident, Accidents, Boiling water reactors, Debris, Explosions, Uncertainty analysis, Boiling water reactor (BWR), Containment integrity, Natural circulation, Severe accident management, Failure (mechanical)
National Category
Engineering and Technology
Identifiers
urn:nbn:se:kth:diva-252279 (URN)2-s2.0-85063129205 (Scopus ID)
Conference
14th Probabilistic Safety Assessment and Management, PSAM 2018, 16 September 2018 through 21 September 2018
Note

QC20190607

Available from: 2019-06-07 Created: 2019-06-07 Last updated: 2019-09-24Bibliographically approved
Organisations
Identifiers
ORCID iD: ORCID iD iconorcid.org/0000-0002-0683-9136

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