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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.

Place, publisher, year, edition, pages
2015. Vol. 281, p. 22-38
National Category
Energy Engineering
Identifiers
URN: urn:nbn:se:kth:diva-161150DOI: 10.1016/j.nucengdes.2014.11.003ISI: 000348950400004Scopus ID: 2-s2.0-84913593586OAI: oai:DiVA.org:kth-161150DiVA, id: diva2:796498
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
In thesis
1. Validation and Application of CFD to Safety-Related Phenomena in Lead-Cooled Fast Reactors
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
Keyword
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

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Kudinov, PavelJeltsov, Marti

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