Change search
Link to record
Permanent link

Direct link
BETA
Alternative names
Publications (10 of 157) Show all publications
Jeltsov, M., Villanueva, W. & Kudinov, P. (2018). Seismic sloshing effects in lead-cooled fast reactors. Nuclear Engineering and Design, 332, 99-110
Open this publication in new window or tab >>Seismic sloshing effects in lead-cooled fast reactors
2018 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 332, p. 99-110Article in journal (Refereed) Published
Abstract [en]

Pool-type primary system can improve the economy of lead-cooled fast reactors. However, partially filled pool of heavy liquid metal poses safety concerns related to seismic loads. Violent sloshing during earthquake-initiated fluid-structure interaction can lead to structural failures, gas entrapment and potential core voiding. Seismic isolation systems can be used to reduce the structural stresses, but its effect on sloshing is not straightforward. This paper presents a numerical study of seismic sloshing in ELSY reactor. The purpose is to evaluate the effects of seismic isolation system on sloshing at different levels of earthquake. Sloshing is modeled using computational fluid dynamics with a volume of fluid free surface capturing model. Earthquake is simulated using synthetic seismic data produced in SILER project as a boundary condition. Simultaneous verification and validation of the numerical model using a dam break experiment is presented. The adverse resonance effect of seismic isolation system is demonstrated in terms of sloshing-induced hydrodynamic loads and gas entrapment. Effectiveness of seismic isolation system is discussed separately for design and beyond design seismic levels. Partitioning baffles are proposed as a potential mitigation measure in the design and their effect is analyzed.

Place, publisher, year, edition, pages
Elsevier, 2018
Keywords
LFR, CFD, Seismic sloshing, FSI, Gas entrapment, Seismic isolation
National Category
Applied Mechanics
Identifiers
urn:nbn:se:kth:diva-227208 (URN)10.1016/j.nucengdes.2018.03.020 (DOI)000430395700010 ()2-s2.0-85044166706 (Scopus ID)
Note

QC 20180530

Available from: 2018-05-30 Created: 2018-05-30 Last updated: 2018-05-30Bibliographically approved
Galushin, S. & Kudinov, P. (2018). Sensitivity analysis of debris properties in lower plenum of a Nordic BWR. Nuclear Engineering and Design, 332, 374-382
Open this publication in new window or tab >>Sensitivity analysis of debris properties in lower plenum of a Nordic BWR
2018 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 332, p. 374-382Article in journal (Refereed) Published
Abstract [en]

Severe accident management (SAM) in Nordic Boiling Water Reactors (BWR) employs ex-vessel core debris coolability. The melt released from the vessel is poured into a deep pool of water and is expected to fragment, quench, and form a coolable debris bed. Success of the strategy is contingent upon the melt release mode from the vessel, which determine conditions for (i) the debris bed coolability, (ii) steam explosion that present credible threats to containment integrity. Melt release conditions are recognized as the major source of uncertainty in quantification of the risk of containment failure in Nordic BWRs. The characteristics of melt release are determined by the in-vessel accident scenarios and phenomena, subject to aleatory and epistemic uncertainties respectively. Specifically, properties of the debris relocated into the lower head determine conditions for the corium interactions with the vessel structures (such as instrumentation guide tubes IGTs, control rod guide tubes CRGTs), vessel failure and melt release. This work is focused on the evaluation of uncertainty in core degradation progression and its effect on the resultant properties of relocated debris in lower plenum of Nordic BWR. We use MELCOR code for prediction of the accident progression. The main goal of this paper is to characterize the range of possible debris properties in lower plenum and its sensitivity towards different modelling parameters, which is of paramount importance for the analysis of in-vessel debris coolability and vessel failure mode in the risk assessment framework.

Place, publisher, year, edition, pages
Elsevier, 2018
Keywords
Severe accident, Nordic BWR, ROAAM, MELCOR
National Category
Other Chemistry Topics
Identifiers
urn:nbn:se:kth:diva-227209 (URN)10.1016/j.nucengdes.2018.03.029 (DOI)000430395700033 ()
Note

QC 20180529

Available from: 2018-05-29 Created: 2018-05-29 Last updated: 2018-05-29Bibliographically approved
Jeltsov, M., Villanueva, W. & Kudinov, P. (2018). Steam generator leakage in lead cooled fast reactors: Modeling of void transport to the core. Nuclear Engineering and Design, 328, 255-265
Open this publication in new window or tab >>Steam generator leakage in lead cooled fast reactors: Modeling of void transport to the core
2018 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 328, p. 255-265Article in journal (Refereed) Published
Abstract [en]

Steam generator tube leakage and/or rupture (SGTL/R) is one of the safety issues for pool type liquid metal cooled fast reactors. During SGTL/R, water is injected from high-pressure secondary side to low-pressure primary side. The possible consequences of such an event include void transport to the core that has adverse effects on the reactor performance including heat transfer deterioration and reactivity insertion. This paper addresses the potential transport of steam bubbles to the core and subsequent void accumulation in the primary system in ELSY conceptual reactor. A CFD model of the primary coolant system for nominal operation is developed and verified. Bubble motion is simulated using Lagrangian tracking of steam bubbles in Eulerian flow field. The effects of uncertainties in the bubble size distribution and bubble drag are addressed. A probabilistic methodology to estimate the core and primary system voiding rates is proposed and demonstrated. A family of drag correlations by Tomiyama et al. (1998) provide the best agreement with the available experimental data. Primary system and core voiding analysis demonstrate that the smallest (sub-millimeter) bubbles have the highest probability to be entrained and remain in the coolant flow. It is found that leaks at the bottom region of the SG result in larger rates of void accumulation.

Place, publisher, year, edition, pages
Elsevier, 2018
Keywords
Bubble transport, CFD, LFR, Steam generator tube leakage/rupture
National Category
Other Physics Topics
Identifiers
urn:nbn:se:kth:diva-221684 (URN)10.1016/j.nucengdes.2018.01.006 (DOI)000427432300023 ()2-s2.0-85040467440 (Scopus ID)
Funder
EU, FP7, Seventh Framework Programme, 249668
Note

QC 20180122

Available from: 2018-01-22 Created: 2018-01-22 Last updated: 2018-05-23Bibliographically 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, p. 487-498Article 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
Keywords
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, p. 487-498Article 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
Keywords
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
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
Show others...
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
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.

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)
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, p. 182-197Article 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
Keywords
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., Grishchenko, D. & Kudinov, P. (2017). Risk analysis framework for severe accident mitigation strategy in nordic BWR: An approach to communication and decision making. In: International Topical Meeting on Probabilistic Safety Assessment and Analysis, PSA 2017: . Paper presented at 2017 International Topical Meeting on Probabilistic Safety Assessment and Analysis, PSA 2017, 24 September 2017 through 28 September 2017 (pp. 587-594). American Nuclear Society
Open this publication in new window or tab >>Risk analysis framework for severe accident mitigation strategy in nordic BWR: An approach to communication and decision making
2017 (English)In: International Topical Meeting on Probabilistic Safety Assessment and Analysis, PSA 2017, American Nuclear Society , 2017, p. 587-594Conference paper, Published paper (Refereed)
Abstract [en]

Severe accident management (SAM) in Nordic boiling water reactors (BWRs) employ ex-vessel debris cooling in a deep water pool. The success of the strategy requires (i) formation of a coolable porous debris bed; (ii) no energetic steam explosion that can threaten containment integrity. Both scenario (aleatory) and modeling (epistemic) uncertainties are important in the assessment of the failure risks. A consistent approach is necessary for the decision making on whether the strategy is sufficiently effective, or a modification of the SAM is necessary. Risk Oriented Accident Analysis Methodology (ROAAM+) is a tool for assessment of failure probability to enable robust decision making, insensitive to remaining uncertainty. Conditional containment failure probability is considered in this work as an indicator of severe accident management effectiveness for Nordic BWR. The ultimate goal of ROAAM+ application for Nordic BWR is to provide a scrutable background in order to achieve convergence of experts' opinions in decision making. The question is: if containment failure can be demonstrated as physically unreasonable, given severe accident management strategy and state-of-the-art knowledge? If inherent safety margins are large, then the answer to the question is positive and can be demonstrated through risk assessment with consistent conservative treatment of uncertainties and by improving, when necessary, knowledge and data. Otherwise, the risk management should be applied in order to increase margins and achieve the safety goal through modifications of the SAM (e.g. safety design, SAMGs, etc.). The challenge for a decision maker is to distinguish when collecting more knowledge and reduction of uncertainty in risk assessment or application of risk management with SAM modifications would be the most effective and efficient approach. In this work we demonstrate a conceptual approach for communication of ROAAM+ framework analysis results and provide an example of a decision support model. The results of the risk analysis are used in order to provide necessary insights on the conditions when suggested changes in the safety design are justified.

Place, publisher, year, edition, pages
American Nuclear Society, 2017
National Category
Other Civil Engineering
Identifiers
urn:nbn:se:kth:diva-230063 (URN)2-s2.0-85047811802 (Scopus ID)9781510851801 (ISBN)
Conference
2017 International Topical Meeting on Probabilistic Safety Assessment and Analysis, PSA 2017, 24 September 2017 through 28 September 2017
Note

QC 20180612

Available from: 2018-06-12 Created: 2018-06-12 Last updated: 2018-06-12Bibliographically approved
Organisations
Identifiers
ORCID iD: ORCID iD iconorcid.org/0000-0002-0683-9136

Search in DiVA

Show all publications