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Li, H., Villanueva, W., Puustinen, M., Laine, J. & Kudinov, P. (2018). Thermal stratification and mixing in a suppression pool induced by direct steam injection. Annals of Nuclear Energy, 111, 487-498.
Open this publication in new window or tab >>Thermal stratification and mixing in a suppression pool induced by direct steam injection
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2018 (English)In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 111, 487-498 p.Article in journal (Refereed) Published
Abstract [en]

An experimental and numerical investigation of thermal stratification and mixing in a suppression pool is presented. Steam injected into a drywell flows through a blowdown pipe and then down to the pressure suppression pool where direct contact condensation occurs. The steam venting and condensation is a source of heat and momentum. A complex interplay between the two leads either to thermal stratification or mixing of the pool. The experiments are conducted in a scaled down PPOOLEX facility at Lappeenranta University of Technology (LUT). The corresponding numerical simulations are performed using GOTHIC with the Effective Heat Source (EHS) and Effective Momentum Source (EMS) models. The EHS/EMS models, that have been previously proposed, predict the development of thermal stratification and mixing during a steam injection into a large pool of water. The experiments exhibit the development of thermal stratification in the pool at relatively low mass flow rates and then pool mixing when the mass flow rates are increased but later thermal stratification can re-develop even at the same relatively high mass flow rates, which is due to the increasing pool temperature that shifts the condensation to a different regime. The numerical simulations quantitatively capture this complex transient pool behavior and are in excellent agreement with the transient averaged pool temperature and water level in the pool.

Place, publisher, year, edition, pages
PERGAMON-ELSEVIER SCIENCE LTD, 2018
Keyword
Pressure suppression pool, Direct steam injection, Thermal stratification and mixing
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-217397 (URN)10.1016/j.anucene.2017.09.014 (DOI)000413877800044 ()2-s2.0-85029704434 (Scopus ID)
Note

QC 20171121

Available from: 2017-11-21 Created: 2017-11-21 Last updated: 2017-11-21Bibliographically approved
Li, H., Villanueva, W., Puustinen, M., Laine, J. & Kudinov, P. (2018). Thermal stratification and mixing in a suppression pool induced by direct steam injection. Annals of Nuclear Energy, 111, 487-498.
Open this publication in new window or tab >>Thermal stratification and mixing in a suppression pool induced by direct steam injection
Show others...
2018 (English)In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 111, 487-498 p.Article in journal (Refereed) Published
Abstract [en]

An experimental and numerical investigation of thermal stratification and mixing in a suppression pool is presented. Steam injected into a drywell flows through a blowdown pipe and then down to the pressure suppression pool where direct contact condensation occurs. The steam venting and condensation is a source of heat and momentum. A complex interplay between the two leads either to thermal stratification or mixing of the pool. The experiments are conducted in a scaled down PPOOLEX facility at Lappeenranta University of Technology (LUT). The corresponding numerical simulations are performed using GOTHIC with the Effective Heat Source (EHS) and Effective Momentum Source (EMS) models. The EHS/EMS models, that have been previously proposed, predict the development of thermal stratification and mixing during a steam injection into a large pool of water. The experiments exhibit the development of thermal stratification in the pool at relatively low mass flow rates and then pool mixing when the mass flow rates are increased but later thermal stratification can re-develop even at the same relatively high mass flow rates, which is due to the increasing pool temperature that shifts the condensation to a different regime. The numerical simulations quantitatively capture this complex transient pool behavior and are in excellent agreement with the transient averaged pool temperature and water level in the pool. 

Place, publisher, year, edition, pages
Elsevier Ltd, 2018
Keyword
Direct steam injection, Pressure suppression pool, Thermal stratification and mixing, Condensation, Mass transfer, Mixing, Numerical models, Steam, Steam condensers, Thermal stratification, Water injection, Water levels, Direct contact condensation, Heat sources, Low mass flow rates, Mass flow rate, Momentum sources, Numerical investigations, Pool mixing, Steam injection, Lakes
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-216805 (URN)10.1016/j.anucene.2017.09.014 (DOI)000413877800044 ()2-s2.0-85029704434 (Scopus ID)
Note

Export Date: 24 October 2017; Article; CODEN: ANEND; Correspondence Address: Villanueva, W.; Division of Nuclear Power Safety, Royal Institute of Technology (KTH), Roslagstullsbacken 21, Sweden; email: walterv@kth.se. QC 20171205

Available from: 2017-12-05 Created: 2017-12-05 Last updated: 2017-12-05Bibliographically 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, 1064-1072 p.Article 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
Phung, V.-A., Grishchenko, D., Galushin, S. & Kudinov, P. (2017). Prediction of In-Vessel Debris Bed Properties in BWR Severe Accident Scenarios using MELCOR and Neural Networks. Annals of Nuclear Energy.
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
2017 (English)In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100Article in journal (Refereed) Submitted
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.

Keyword
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)
Note

QC 20170309

Available from: 2017-03-08 Created: 2017-03-08 Last updated: 2017-11-29Bibliographically approved
Kudinov, P., Grishchenko, D., Konovalenko, A. & Karbojian, A. (2017). Premixing and steam explosion phenomena in the tests with stratified melt-coolant configuration and binary oxidic melt simulant materials. Nuclear Engineering and Design, 314, 182-197.
Open this publication in new window or tab >>Premixing and steam explosion phenomena in the tests with stratified melt-coolant configuration and binary oxidic melt simulant materials
2017 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 314, 182-197 p.Article in journal (Refereed) Published
Abstract [en]

Steam explosion phenomena in stratified melt-coolant configuration are considered in this paper. Liquid corium layer covered by water on top can be formed in severe accident scenarios with (i) vessel failure and release of corium melt into a relatively shallow water pool; (ii) with top flooding of corium melt layer. In previous assessments of potential energetics in stratified melt-coolant configuration, it was assumed that melt and coolant are separated by a stable vapor film and there is no premixing prior to the shock wave propagation. This assumption was instrumental for concluding that the amount of energy that can be released in such configuration is not of safety importance. However, several recent experiments carried out in Pouring and Under-water Liquid Melt Spreading (PULiMS) facility with up to 78 kg of binary oxidic corium simulants mixtures have resulted in spontaneous explosions with relatively high conversion ratios (order of one percent). The instability of the melt-coolant interface, melt splashes and formation of premixing layer were observed in the tests. In this work, we present results of experiments carried out more recently in steam explosion in stratified melt-coolant configuration (SES) facility in order to shed some light on the premixing phenomena and assess the influence of the test conditions on the steam explosion energetics.

Place, publisher, year, edition, pages
ELSEVIER SCIENCE SA, 2017
Keyword
Severe accident, Steam explosion, Stratified melt-coolant configuration, Premixing
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-206284 (URN)10.1016/j.nucengdes.2017.01.029 (DOI)000396947800016 ()2-s2.0-85012164081 (Scopus ID)
Note

QC 20170509

Available from: 2017-05-09 Created: 2017-05-09 Last updated: 2017-05-09Bibliographically approved
Galushin, S. & Kudinov, P. (2016). Analysis of core degradation and relocation phenomena and scenarios in a Nordic-type BWR. NUCLEAR ENGINEERING AND DESIGN, 310, 125-141.
Open this publication in new window or tab >>Analysis of core degradation and relocation phenomena and scenarios in a Nordic-type BWR
2016 (English)In: NUCLEAR ENGINEERING AND DESIGN, ISSN 0029-5493, Vol. 310, 125-141 p.Article in journal (Refereed) Published
Abstract [en]

Severe Accident Management (SAM) in Nordic Boiling Water Reactors (BWR) employs ex-vessel cooling of core melt debris. The melt is released from the failed vessel and 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 a natural circulation and evaporation of water. Success of the strategy is contingent upon melt release conditions from the vessel and melt-coolant interaction that determine (i) properties of the debris bed and its coolability (ii) potential for energetic melt-coolant interactions (steam explosions). Risk Oriented Accident Analysis Methodology (ROAAM+) framework is currently under development for quantification of the risks associated with formation of non-coolable debris bed and occurrence of steam explosions, both presenting a credible threats to containment integrity. The ROAAM+ framework consist of loosely coupled models that describe each stage of the accident progression. Core relocation analysis framework provides initial conditions for melt vessel interaction, vessel failure and melt release frameworks. The properties of relocated debris and melt release conditions, including in-vessel and ex-vessel pressure, lower drywell pool depth and temperature, are sensitive to the accident scenarios and timing of safety systems recovery and operator actions. This paper illustrates a methodological approach and relevant data for establishing a connection between core relocation and vessel failure analysis in ROAAM+ approach. MELCOR code is used for analysis of core degradation and relocation phenomena. Properties of relocated debris are obtained as functions of the accident scenario parameters. Pattern analysis is employed in order to characterize typical behavior of core relocation transients. Clustering analysis is employed for grouping of different accident scenarios, which result in similar core relocation behavior and properties of the debris.

Place, publisher, year, edition, pages
Elsevier, 2016
National Category
Other Physics Topics
Identifiers
urn:nbn:se:kth:diva-200215 (URN)10.1016/j.nucengdes.2016.09.029 (DOI)000390736400011 ()2-s2.0-84993993448 (Scopus ID)
Note

QC 20170202

Available from: 2017-02-02 Created: 2017-01-23 Last updated: 2017-02-02Bibliographically approved
Phung, V.-A., Koop, K., Grishchenko, D., Vorobyev, Y. & Kudinov, P. (2016). Automation of RELAP5 input calibration and code validation using genetic algorithm. Nuclear Engineering and Design, 300, 210-221.
Open this publication in new window or tab >>Automation of RELAP5 input calibration and code validation using genetic algorithm
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2016 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 300, 210-221 p.Article in journal (Refereed) Published
Abstract [en]

Validation of system thermal-hydraulic codes is an important step in application of the codes to reactor safety analysis. The goal of the validation process is to determine how well a code can represent physical reality. This is achieved by comparing predicted and experimental system response quantities (SRQs) taking into account experimental and modelling uncertainties. Parameters which are required for the code input but not measured directly in the experiment can become an important source of uncertainty in the code validation process. Quantification of such parameters is often called input calibration. Calibration and uncertainty quantification may become challenging tasks when the number of calibrated input parameters and SRQs is large and dependencies between them are complex. If only engineering judgment is employed in the process, the outcome can be prone to so called "user effects". The goal of this work is to develop an automated approach to input calibration and RELAP5 code validation against data on two-phase natural circulation flow instability. Multiple SRQs are used in both calibration and validation. In the input calibration, we used genetic algorithm (GA), a heuristic global optimization method, in order to minimize the discrepancy between experimental and simulation data by identifying optimal combinations of uncertain input parameters in the calibration process. We demonstrate the importance of the proper selection of SRQs and respective normalization and weighting factors in the fitness function. In the code validation, we used maximum flow rate as the SRQ of primary interest. The ranges of the input parameter were defined based on the experimental data and results of the calibration process. Then GA was used in order to identify combinations of the uncertain input parameters that provide maximum deviation of code prediction results from the experimental data. Such approach provides a conservative estimate of the possible discrepancy between the code result and the experimental data.

Place, publisher, year, edition, pages
Elsevier, 2016
Keyword
Thermal hydraulics
National Category
Other Physics Topics
Identifiers
urn:nbn:se:kth:diva-185600 (URN)10.1016/j.nucengdes.2016.01.003 (DOI)000372840400019 ()2-s2.0-84958025786 (Scopus ID)
Note

QC 20160428

Available from: 2016-04-28 Created: 2016-04-25 Last updated: 2017-11-30Bibliographically approved
Grishchenko, D., Basso, S. & Kudinov, P. (2016). Development of a surrogate model for analysis of ex-vessel steam explosion in Nordic type BWRs. NUCLEAR ENGINEERING AND DESIGN, 310, 311-327.
Open this publication in new window or tab >>Development of a surrogate model for analysis of ex-vessel steam explosion in Nordic type BWRs
2016 (English)In: NUCLEAR ENGINEERING AND DESIGN, ISSN 0029-5493, Vol. 310, 311-327 p.Article in journal (Refereed) Published
Abstract [en]

Severe accident mitigation strategy adopted in Nordic type Boiling Water Reactors (BWRs) employs ex vessel core melt cooling in a deep pool of water below reactor vessel. Energetic fuel coolant interaction (steam explosion) can occur during molten core release into water. Dynamic loads can threaten containment integrity increasing the risk of fission products release to the environment. Comprehensive uncertainty analysis is necessary in order to assess the risks. Computational costs of the existing fuel coolant interaction (FCI) codes is often prohibitive for addressing the uncertainties, including the effect of stochastic triggering time. This paper discusses development of a computationally efficient surrogate model (SM) for prediction of statistical characteristics of steam explosion impulses in Nordic BWRs. The TEXAS-V code was used as the Full Model (FM) for the calculation of explosion impulses. The surrogate model was developed using artificial neural networks' (ANNs) and the database of FM solutions. Statistical analysis was employed in order to treat chaotic response of steam explosion impulse to variations in the triggering time. Details of the FM and SM implementation and their verification are discussed in the paper.

Place, publisher, year, edition, pages
Elsevier, 2016
National Category
Other Physics Topics
Identifiers
urn:nbn:se:kth:diva-200216 (URN)10.1016/j.nucengdes.2016.10.014 (DOI)000390736400026 ()2-s2.0-85002512598 (Scopus ID)
Note

QC 20170202

Available from: 2017-02-02 Created: 2017-01-23 Last updated: 2017-02-02Bibliographically approved
Basso, S., Konovalenko, A. & Kudinov, P. (2016). Empirical closures for particulate debris bed spreading induced by gas-liquid flow. Nuclear Engineering and Design, 297, 19-25.
Open this publication in new window or tab >>Empirical closures for particulate debris bed spreading induced by gas-liquid flow
2016 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 297, 19-25 p.Article in journal (Refereed) Published
Abstract [en]

Efficient removal of decay heat from the nuclear reactor core debris is paramount for termination of severe accident progression. One of the strategies is based on melt fragmentation, quenching and cooling in a deep pool of water under the reactor vessel. Geometrical configuration of the debris bed is among the important factors which determine possibility of removing the decay heat from the debris bed by natural circulation of the coolant. For instance, a tall mound-shape debris bed can be non-coolable, while the same debris can be coolable if spread uniformly. Decay heat generates a significant amount of thermal energy which goes to production of steam inside the debris bed. Two-phase flow escaping through the top layer of the bed becomes a source of mechanical energy which can move the particulate debris along the slope of the bed. The motion of the debris will lead to flattening of the bed. Such process is often called "self-leveling" phenomenon. Spreading of the debris bed by the self-leveling process can take significant time, depending on the initial debris bed configuration and other parameters. There is a competition between the time scales for reaching (i) a coolable configuration of the bed, and (ii) onset of dryout and re-melting of the debris. In the previous work we have demonstrated that the rate of particulate debris spreading is determined by local gas velocity and local slope angle of the bed. In this work we develop a scaling approach and a closure for prediction of debris spreading rate based on generalization of available experimental data. We demonstrate that introduced scaling criteria are universal for particles of different shapes and size distributions.

Place, publisher, year, edition, pages
Elsevier, 2016
National Category
Energy Engineering
Identifiers
urn:nbn:se:kth:diva-180926 (URN)10.1016/j.nucengdes.2015.10.016 (DOI)000369167700003 ()2-s2.0-84950119479 (Scopus ID)
Funder
Swedish Radiation Safety Authority
Note

QC 20160126. QC 20160304

Available from: 2016-01-26 Created: 2016-01-25 Last updated: 2017-11-30Bibliographically approved
Konovalenko, A., Basso, S., Kudinov, P. & Yakush, S. E. (2016). Experimental investigation of particulate debris spreading in a pool. Nuclear Engineering and Design, 297, 208-219.
Open this publication in new window or tab >>Experimental investigation of particulate debris spreading in a pool
2016 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 297, 208-219 p.Article in journal (Refereed) Published
Abstract [en]

Termination of severe accident progression by core debris cooling in a deep pool of water under reactor vessel is considered in several designs of light water reactors. However, success of this accident mitigation strategy is contingent upon the effectiveness of heat removal by natural circulation from the debris bed. It is assumed that a porous bed will be formed in the pool in the process of core melt fragmentation and quenching. Debris bed coolability depends on its properties and system conditions. The properties of the bed, including its geometry are the outcomes of the debris bed formation process. Spreading of the debris particles in the pool by two-phase turbulent flows induced by the heat generated in the bed can affect the shape of the bed and thus influence its coolability. The goal of this work is to provide experimental data on spreading of solid particles in the pool by large-scale two-phase flow. The aim is to provide data necessary for understanding of separate effects and for development and validation of models and codes. Validated codes can be then used for prediction of debris bed formation under prototypic severe accident conditions. In PDS-P (Particulate Debris Spreading in the Pool) experiments, air injection at the bottom of the test section is employed as a means to create large-scale flow in the pool in isothermal conditions. The test section is a rectangular tank with a 2D slice geometry, it has fixed width (72 mm), adjustable length (up to 1.5 m) and allows water filling to the depth of up to 1 m. Variable pool length and depth allows studying two-phase circulating flows of different characteristic sizes and patterns. The average void fraction in the pool is determined by video recording and subsequent image processing. Particles are supplied from the top of the facility above the water surface. Results of several series of PDS-P experiments are reported in this paper. The influence of the gas flow rate, pool dimensions, particle density and size on spreading of the particles is addressed. A preliminary scaling approach is proposed and shown to provide good agreement with the experimental findings.

Place, publisher, year, edition, pages
Elsevier, 2016
National Category
Energy Engineering
Identifiers
urn:nbn:se:kth:diva-180933 (URN)10.1016/j.nucengdes.2015.11.039 (DOI)000369167700022 ()2-s2.0-84951121625 (Scopus ID)
Funder
Swedish Radiation Safety Authority
Note

QC 20160126. QC 20160304

Available from: 2016-01-26 Created: 2016-01-25 Last updated: 2017-11-30Bibliographically approved
Organisations
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ORCID iD: ORCID iD iconorcid.org/0000-0002-0683-9136

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