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

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

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

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

Available from: 2018-11-27 Created: 2018-11-27 Last updated: 2018-11-27Bibliographically approved
Jeltsov, M., 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
Jeltsov, M. (2018). Validation and Application of CFD to Safety-Related Phenomena in Lead-Cooled Fast Reactors. (Doctoral dissertation). KTH Royal Institute of Technology
Open this publication in new window or tab >>Validation and Application of CFD to Safety-Related Phenomena in Lead-Cooled Fast Reactors
2018 (English)Doctoral thesis, comprehensive summary (Other academic)
Abstract [en]

Carbon-free nuclear power production can help to relieve the increasing energy demand in the world. New generation reactors with improved safety, sustainability, reliability, economy and security are being developed. Lead-cooled fast reactor (LFR) is one of them.

Lead-cooled reactors have many advantages related to the use of liquid metal coolants and pool-type primary system designs. Still, several technological challenges are to be overcome before their successful commercial deployment. The decisions on new designs and licensing of LFRs require use of code simulations, due to lack of operational experience and the fact that full scale experimental investigations are impractical. Code development and validation is necessary in order to reduce uncertainty in predictions for risk assessment and decision support. In order to make the decisions robust, i.e. not sensitive to remaining uncertainty, the process of collecting necessary evidences should iteratively converge. Extensive sensitivity and uncertainty analysis is necessary in order to make the process user-independent.

The pool-type design introduces 3D phenomena that require application of advanced modeling tools such as Computational Fluid Dynamics (CFD). The focus of this thesis is on the development and application of CFD modeling and general code validation methodology to the following issues of safety importance for LFR: (i) pool thermal stratification and mixing, (ii) steam generator tube leakage, (iii) seismic sloshing, and (iv) solidification.

Thermal mixing and stratification in a pool-type reactor can affect natural circulation stability and thus decay heat removal capability and pose thermal stresses on the reactor structures. Standalone System Thermal-hydraulics (STH) and CFD codes are either inadequate or too computationally expensive for analysis of such phenomena in prototypic conditions. In this work an approach to coupling of STH and CFD is proposed and implemented. For validation of standalone and coupled codes TALL-3D facility (lead-bismuth eutectic (LBE) loop) and test matrix was designed featuring significant two-way thermal-hydraulic feedbacks between the pool-type test section and the rest of the loop.

The possibility of core voiding due to bubble transport in case of steam generator tube leakage/rupture (SGTL/R) is a serious safety concern that can be a showstopper for LFR licensing. Voiding of the core increases risks for reactivity insertion accident (RIA) and/or local fuel damage due to deteriorated heat transfer. Bubble transport analysis for the ELSY (European Lead-cooled SYstem) reactor design demonstrated the effect of uncertainty in drag correlation, bubble size distributions and leak location on the core and primary system voiding. Possible design solutions for prevention of core voiding in case of SGTL/R are discussed.

The effect of seismic isolation (SI) system on the sloshing of the heavy metal coolant and respective mechanical loads on the vessel structures is studied for design basis and beyond design basis earthquakes. It was found that SI shifts the frequencies closer to resonance conditions for sloshing, thus increasing the loads due to slamming of coolant waves onto the vessel internal structures including reactor lid. Possible mitigation measures, such as partitioning baffles have been proposed and their effectiveness was analyzed. 

Coolant solidification is another serious safety issue in liquid metal cooled reactors that can cause partial or even full blockage of the coolant flow path and thermo-mechanical loads on the primary system structures. Validation of melting/solidification models implemented in CFD is a necessary pre-requisite for its application to risk analysis. In this work, the parametric analysis in support of the design of a solidification test section (STS) in TALL3D facility is carried out. The goals and requirements for the validation experiment and how they are satisfied in the design are discussed in the thesis.

Place, publisher, year, edition, pages
KTH Royal Institute of Technology, 2018. p. 107
Series
TRITA-SCI-FOU ; 2018:11
Keywords
Verification, Validation, Calibration, Sensitivity Analysis, Uncertainty Analysis, CFD, STH, Code Coupling, Liquid Lead Coolant, LFR, SGTL/R, Bubble transport, Core voiding, Seismic Sloshing, Melting/Solidification
National Category
Energy Engineering
Research subject
Physics
Identifiers
urn:nbn:se:kth:diva-228355 (URN)978-91-7729-725-3 (ISBN)
Public defence
2018-06-07, FR4 (Oskar Klein), AlbaNova Universitetcentrum, Roslagstullsbacken 21, Stockholm, 09:30 (English)
Opponent
Supervisors
Note

QC 20180523

Available from: 2018-05-23 Created: 2018-05-22 Last updated: 2018-05-23Bibliographically 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
Bandini, G., Polidori, M., Gerschenfeld, A., Pialla, D., Li, S., Ma, W., . . . Maas, L. (2015). Assessment of systems codes and their coupling with CFD codes in thermal-hydraulic applications to innovative reactors. Nuclear Engineering and Design, 281, 22-38
Open this publication in new window or tab >>Assessment of systems codes and their coupling with CFD codes in thermal-hydraulic applications to innovative reactors
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2015 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 281, p. 22-38Article in journal (Refereed) Published
Abstract [en]

The THINS project of the 7th Framework EU Program on nuclear fission safety is devoted to the investigation of crosscutting thermal hydraulic issues for innovative nuclear systems. A significant effort in the project has been dedicated to the qualification and validation of system codes currently employed in thermal hydraulic transient analysis for nuclear reactors. This assessment is based either on already available experimental data, or on the data provided by test campaigns carried out in the frame of THINS project activities. Data provided by TALL and CIRCE facilities were used in the assessment of system codes for HLM reactors, while the PHENIX ultimate natural circulation test was used as reference for a benchmark exercise among system codes for sodium-cooled reactor applications. In addition, a promising grid-free pool model based on proper orthogonal decomposition is proposed to overcome the limits shown by the thermal hydraulic system codes in the simulation of pool-type systems. Furthermore, multi-scale system-CFD solutions have been developed and validated for innovative nuclear system applications. For this purpose, data from the PHENIX experiments have been used, and data are provided by the tests conducted with new configuration of the TALL-3D facility, which accommodates a 3D test section within the primary circuit. The TALL-3D measurements are currently used for the validation of the coupling between system and CFD codes.

National Category
Energy Engineering
Identifiers
urn:nbn:se:kth:diva-161150 (URN)10.1016/j.nucengdes.2014.11.003 (DOI)000348950400004 ()2-s2.0-84913593586 (Scopus ID)
Funder
EU, FP7, Seventh Framework Programme, FP7-249337
Note

QC 20150319

Available from: 2015-03-19 Created: 2015-03-09 Last updated: 2018-05-23Bibliographically approved
Papukchiev, A., Jeltsov, M., Kööp, K., Kudinov, P. & Lerchl, G. (2015). Comparison of different coupling CFD-STH approaches for pre-test analysis of a TALL-3D experiment. Nuclear Engineering and Design, 290, 135-143
Open this publication in new window or tab >>Comparison of different coupling CFD-STH approaches for pre-test analysis of a TALL-3D experiment
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2015 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 290, p. 135-143Article in journal (Refereed) Published
Abstract [en]

The system thermal-hydraulic (STH) code ATHLET was coupled with the commercial 3D computational fluid dynamics (CFD) software package ANSYS CFX to improve ATHLET simulation capabilities for flows with pronounced 3D phenomena such as flow mixing and thermal stratification. Within the FP7 European project THINS (Thermal Hydraulics of Innovative Nuclear Systems), validation activities for coupled thermal-hydraulic codes are being carried out. The TALL-3D experimental facility, operated by KTH Royal Institute of Technology in Stockholm, is designed for thermal-hydraulic experiments with lead-bismuth eutectic (LBE) coolant at natural and forced circulation conditions. GRS carried out pre-test simulations with ATHLET-ANSYS CFX for the TALL-3D experiment T01, while KTH scientists perform these analyses with the coupled code RELAP5/STAR CCM+. In the experiment T01 the main circulation pump is stopped, which leads to interesting thermal-hydraulic transient with local 3D phenomena. In this paper, the TALL-3D behavior during T01 is analyzed and the results of the coupled pre-test calculations, performed by GRS (ATHLET-ANSYS CFX) and KTH (RELAP5/STAR CCM+) are directly compared.

Keywords
Codes (symbols), Computer software, Experiments, Hydraulics, Two phase flow, Experimental facilities, Forced circulations, Innovative nuclear system, Lead-bismuth eutectics, Royal Institute of Technology, Software package ANSYS, Thermal hydraulics, Thermal-hydraulic codes
National Category
Energy Engineering
Identifiers
urn:nbn:se:kth:diva-165393 (URN)10.1016/j.nucengdes.2014.11.008 (DOI)
Note

QC 20150708

Available from: 2015-04-27 Created: 2015-04-27 Last updated: 2019-09-20Bibliographically approved
Jeltsov, M., Villanueva, W. & Kudinov, P. (2015). Parametric study of sloshing effects in the primary system of an isolated lead-cooled fast reactor. Nuclear Technology, 190(1), 1-10
Open this publication in new window or tab >>Parametric study of sloshing effects in the primary system of an isolated lead-cooled fast reactor
2015 (English)In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 190, no 1, p. 1-10Article in journal (Refereed) Published
Abstract [en]

Risks related to sloshing of liquid metal coolant due to seismic excitation need to be investigated. Sloshing effects on reactor performance include first, fluid-structure interaction and second, gas entrapment in the coolant with subsequent transport of void to the core region. While the first can hypothetically lead to structural damage or coolant spill, the second increases the risk of a reactivity insertion accident and/or local dryout of the fuel. A two-dimensional computational fluid dynamics study is carried out in order to obtain insights into the modes of sloshing depending on the parameters of seismic excitation. The applicability and performance of the numerical mesh and the Eulerian volume of fluid method used to track the free surface are evaluated by modeling a simple dam break experiment. Sloshing in the cold plenum free surface region of the European Lead-cooled SYstem (ELSY) conceptual pool-type lead-cooled fast reactor (LFR) is studied. Various sinusoidal excitations are used to imitate the seismic response at the reactor level. The goal is to identify the domain of frequencies and magnitudes of the seismic response that can lead to loads threatening the structural integrity and possible core voiding due to sloshing. A map of sloshing modes has been developed to characterize the sloshing response as a function of excitation parameters. Pressure forces on vertical walls and the lid have been calculated. Finally, insight into coolant voiding has been provided.

National Category
Energy Engineering
Identifiers
urn:nbn:se:kth:diva-165389 (URN)10.13182/NT14-8 (DOI)000352678000001 ()
Note

QC 20150513

Available from: 2015-04-27 Created: 2015-04-27 Last updated: 2018-05-23Bibliographically approved
Grishchenko, D., Jeltsov, M., Kööp, K., Karbojian, A., Villanueva, W. & Kudinov, P. (2015). The TALL-3D facility design and commissioning tests for validation of coupled STH and CFD codes. Nuclear Engineering and Design, 290, 144-153
Open this publication in new window or tab >>The TALL-3D facility design and commissioning tests for validation of coupled STH and CFD codes
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2015 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 290, p. 144-153Article in journal (Refereed) Published
Abstract [en]

Application of coupled CFD (Computational Fluid Dynamics) and STH (System Thermal Hydraulics) codes is a prerequisite for computationally affordable and sufficiently accurate prediction of thermal-hydraulics of complex systems. Coupled STH and CFD codes require validation for understanding and quantification of the sources of uncertainties in the code prediction. TALL-3D is a liquid Lead Bismuth Eutectic (LBE) loop developed according to the requirements for the experimental data for validation of coupled STH and CFD codes. The goals of the facility design are to provide (i) mutual feedback between natural circulation in the loop and complex 3D mixing and stratification phenomena in the pool-type test section, (ii) a possibility to validate standalone STH and CFD codes for each subsection of the facility, and (iii) sufficient number of experimental data to separate the process of input model calibration and code validation. Description of the facility design and its main components, approach to estimation of experimental uncertainty and calibration of model input parameters that are not directly measured in the experiment are discussed in the paper. First experimental data from the forced to natural circulation transient is also provided in the paper.

National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-164064 (URN)10.1016/j.nucengdes.2014.11.045 (DOI)000357227700015 ()2-s2.0-84937514619 (Scopus ID)
Note

QC 20150623

Available from: 2015-04-13 Created: 2015-04-13 Last updated: 2018-05-23Bibliographically approved
Mickus, I., Kööp, K., Jeltsov, M., Vorobyev, Y., Villanueva, W. & Kudinov, P. (2014). An Approach to Physics Based Surrogate Model Development for Application with IDPSA. In: : . Paper presented at Probabilistic Safety Assessment and Management PSAM 12, June 2014, Honolulu, Hawaii.
Open this publication in new window or tab >>An Approach to Physics Based Surrogate Model Development for Application with IDPSA
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2014 (English)Conference paper, Published paper (Refereed)
Abstract [en]

Integrated Deterministic Probabilistic Safety Assessment (IDPSA) methodology is a powerful tool for identification of failure domains when both stochastic events and physical time dependent processes are important. Computational efficiency of deterministic models is one of the limiting factors for detailed exploration of the event space. Pool type designs of Generation IV heavy liquid metal cooled reactors introduce importance of capturing intricate 3D flow phenomena in safety analysis. Specifically mixing and stratification in 3D elements can affect efficiency of passive safety systems based on natural circulation. Conventional 1D System Thermal Hydraulics (STH) codes are incapable of predicting such complex 3D phenomena. Computational Fluid Dynamics (CFD) codes are too computationally expensive to be used for simulation of the whole reactor primary coolant system. One proposed solution is code coupling where all 1D components are simulated with STH and 3D components with CFD codes. However, modeling with coupled codes is still too time consuming to be used directly in IDPSA methodologies, which require thousands of simulations. The goal of this work is to develop a computationally efficient surrogate model (SM) which captures key physics of complex thermal hydraulic phenomena in the 3D elements and can be coupled with 1D STH codes instead of CFD. TALL-3D is a lead-bismuth eutectic thermal hydraulic loop which incorporates both 1D and 3D elements. Coupled STH-CFD simulations of TALL-3D typical transients (such as transition from forced to natural circulation) are used to calibrate the surrogate model parameters. Details of current implementation and limitations of the surrogate modeling are discussed in the paper in detail.

Keywords
Dynamic PSA, IDPSA, Surrogate model, TALL-3D
National Category
Energy Engineering
Identifiers
urn:nbn:se:kth:diva-165283 (URN)2-s2.0-84925062062 (Scopus ID)
Conference
Probabilistic Safety Assessment and Management PSAM 12, June 2014, Honolulu, Hawaii
Note

QC 20150513

Available from: 2015-04-24 Created: 2015-04-24 Last updated: 2015-06-10Bibliographically approved
Organisations
Identifiers
ORCID iD: ORCID iD iconorcid.org/0000-0001-5653-9206

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