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Steam generator leakage in lead cooled fast reactors: Modeling of void transport to the core
KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.ORCID iD: 0000-0001-5653-9206
KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.ORCID iD: 0000-0003-3132-7252
KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.ORCID iD: 0000-0002-0683-9136
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. Vol. 328, p. 255-265
Keywords [en]
Bubble transport, CFD, LFR, Steam generator tube leakage/rupture
National Category
Other Physics Topics
Identifiers
URN: urn:nbn:se:kth:diva-221684DOI: 10.1016/j.nucengdes.2018.01.006ISI: 000427432300023Scopus ID: 2-s2.0-85040467440OAI: oai:DiVA.org:kth-221684DiVA, id: diva2:1176299
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
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
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)
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Supervisors
Note

QC 20180523

Available from: 2018-05-23 Created: 2018-05-22 Last updated: 2018-05-23Bibliographically approved

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Jeltsov, MartiVillanueva, WalterKudinov, Pavel

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