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Pre-test analysis of an LBE solidification experiment in TALL-3D
KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnenergiteknik.ORCID-id: 0000-0001-5653-9206
KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnenergiteknik.ORCID-id: 0000-0003-1213-0032
KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnenergiteknik.
KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnenergiteknik.ORCID-id: 0000-0002-0683-9136
(engelsk)Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759XArtikkel i tidsskrift (Fagfellevurdert) Submitted
Abstract [en]

This paper presents the process of development of a solidification test section design for TALL-3D experimental facility (lead-bismuth eutectic (LBE) thermal-hydraulic loop of prototypic height). Coolant solidification is a phenomenon of potential safety importance for liquid metal cooled fast reactors (LMFRs). Solidification, e.g. in case of excessive performance of the passive decay heat removal systems, can affect local heat transfer and even lead to partial or complete blockage of the coolant flow paths. This might lead to failure of decay heat removal function. In case of reduced flow circulation, the temperature of the coolant will increase, which might prevent complete blockage of the flow. Prediction of possible outcomes of such scenarios with complex interactions between local physical phenomena of solidification and system scale natural circulation behavior is subject to modelling (epistemic) uncertainty. Development and validation of adequate models requires validation grade experimental data. In this work we discuss results of analysis carried out in support experiment development. The aim of the experimental design is to satisfy requirements stemming from the process of qualification of the model that can be used for addressing the safety-related concerns. In this work we focus on two aspects: (i) design of solidification test section (STS) for investigation of local solidification phenomena of lead-bismuth eutectic (LBE), and (ii) effect of the STS on the system scale behavior of the experimental facility. We discuss selection of the STS characteristics and experimental test matrix using computational fluid dynamic (CFD) and system thermal-hydraulic (STH) codes.

HSV kategori
Identifikatorer
URN: urn:nbn:se:kth:diva-228331OAI: oai:DiVA.org:kth-228331DiVA, id: diva2:1209157
Merknad

QC 20180522

Tilgjengelig fra: 2018-05-22 Laget: 2018-05-22 Sist oppdatert: 2018-05-23bibliografisk kontrollert
Inngår i avhandling
1. Application of a system thermal-hydraulics code to development of validation process for coupled STH-CFD codes
Åpne denne publikasjonen i ny fane eller vindu >>Application of a system thermal-hydraulics code to development of validation process for coupled STH-CFD codes
2018 (engelsk)Doktoravhandling, med artikler (Annet vitenskapelig)
Abstract [en]

Generation IV reactors are designed to provide sustainable energy generation, minimize waste production and excel in safety. Due to lack of operational experience, ever evolving design and stringent safety requirements, these novel reactors have to rely heavily on simulations.

Best estimate one-dimensional (1D) system thermal-hydraulics (STH) codes, originally intended for simulating water-cooled reactor systems with high coolant mass flow rates, are unable to capture complex three-dimensional (3D) phenomena in liquid metal cooled pool-type reactors. Computational fluid dynamics (CFD) codes are capable of resolving the 3D effects, however applying these methods with high resolution for the whole primary system results in prohibiting computational cost.

At the same time, there are system components where flow can, with reasonable accuracy, be approximated with 1D models (e.g. core channels, some heat exchangers, etc.). One of the proposed solutions in order to achieve adequate accuracy and affordable computational efficiency in modelling of a Generation IV reactor is to divide the primary system into 1D and 3D regions and apply coupled STH and CFD codes on the respective sub-domains.

Successful validation is a prerequisite for application of both, standalone and coupled STH and CFD codes in design and safety analysis of Generation IV systems. In this work we develop and apply different aspects of code validation methodology with an emphasis on (i) STH code analysis in support of validation experiment design (facility and test conditions), (ii) calibration of uncertain code input parameters and validation of standalone STH code, (iii) development of an approach to couple STH and CFD codes.

A considerable part of the thesis work is related to the development of a loop-type, 3 leg, liquid metal experimental facility TALL-3D for code validation. Particular focus was on identification of test conditions featuring complex feedbacks between 1D-3D phenomena, which can be challenging for the codes. Standalone STH code (RELAP5) was validated against experimental data. The domain of natural circulation instabilities in TALL-3D operation parameters was discovered using a validated STH code and global optimum search algorithms. Then existence of growing natural circulation oscillations was experimentally confirmed. An international benchmark was initiated in the framework of EU SESAME project based on the obtained experimental data.

Simulations were performed to define dimensions and location of a new test section for coolant solidification experiments that would also enhance possibilities for studying natural circulation instabilities in the future tests.

An approach to automated input calibration and code validation is developed in order to minimize possible “user effect” in case of multiple uncertain input parameters (UIPs) and system response quantities (SRQs). These methods were applied extensively in the development of RELAP5 input models and identification of the natural circulation instability regions.

Domain overlapping approach to coupling of RELAP5 and Star-CCM+ codes was proposed and resulted in considerable improvement of the predictive capabilities in comparison to standalone RELAP5.

sted, utgiver, år, opplag, sider
KTH Royal Institute of Technology, 2018. s. 68
Serie
TRITA-SCI-FOU ; 2018:10
HSV kategori
Forskningsprogram
Fysik
Identifikatorer
urn:nbn:se:kth:diva-228332 (URN)978-91-7729-727-7 (ISBN)
Disputas
2018-06-07, FB52, AlbaNova University Center, Stockholm, 14:00 (engelsk)
Opponent
Veileder
Merknad

QC 20180522

Tilgjengelig fra: 2018-05-22 Laget: 2018-05-22 Sist oppdatert: 2018-05-22bibliografisk kontrollert
2. Validation and Application of CFD to Safety-Related Phenomena in Lead-Cooled Fast Reactors
Åpne denne publikasjonen i ny fane eller vindu >>Validation and Application of CFD to Safety-Related Phenomena in Lead-Cooled Fast Reactors
2018 (engelsk)Doktoravhandling, med artikler (Annet vitenskapelig)
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.

sted, utgiver, år, opplag, sider
KTH Royal Institute of Technology, 2018. s. 107
Serie
TRITA-SCI-FOU ; 2018:11
Emneord
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
HSV kategori
Forskningsprogram
Fysik
Identifikatorer
urn:nbn:se:kth:diva-228355 (URN)978-91-7729-725-3 (ISBN)
Disputas
2018-06-07, FR4 (Oskar Klein), AlbaNova Universitetcentrum, Roslagstullsbacken 21, Stockholm, 09:30 (engelsk)
Opponent
Veileder
Merknad

QC 20180523

Tilgjengelig fra: 2018-05-23 Laget: 2018-05-22 Sist oppdatert: 2018-05-23bibliografisk kontrollert

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