Application of CFD to Safety and Thermal-Hydraulic Analysis of Lead-Cooled Systems

Computational Fluid Dynamics (CFD) is increasingly being used in nuclear reactor safety analysis as a tool that enables safety related physical phenomena occurring in the reactor coolant system to be described in more detail and accuracy. Validation is a necessary step in improving predictive capability of a computationa code or coupled computational codes. Validation refers to the assessment of model accuracy incorporating any uncertainties (aleatory and epistemic) that may be of importance. The uncertainties must be identi ed, quanti ed and if possible, reduced. In the rst part of this thesis, a discussion on the development of an approach and experimental facility for the validation of coupled Computational Fluid Dynamics codes and System Thermal Hydraulics (STH) codes is given. The validation of a coupled code requires experiments which feature signi cant two-way feedbacks between the component (CFD sub-domain) and the system (STH sub-domain). Results of CFD analysis that are used in the development of a exible design of the TALL-3D experimental facility are presented. The facility consists of a lead-bismuth eutectic (LBE) thermal-hydraulic loop operating in forced and natural circulation regimes with a heated pool-type 3D test section. Transient analysis of the mixing and strati cation phenomena in the 3D test section under forced and natural circulation conditions in the loop show that the test section outlet temperature deviates from that predicted by analytical solution (which the 1D STH solution essentially is). Also an experimental validation test matrix according to the key physical phenomena of interest in the new experimental facility is developed. In the second part of the thesis we consider the risk related to steam generator tube leakage or rupture (SGTL/R) in a pool-type design of lead-cooled reactor (LFR). We demonstrate that there is a possibility that small steam bubbles leaking from the SGT will be dragged by the turbulent coolant ow into the core region. Voiding of the core might cause threats of reactivity insertion accident or local damage (burnout) of fuel rod cladding. Trajectories of the bubbles are determined by the bubble size and turbulent ow eld of lead coolant. The main objective of such study is to quantify likelihood of steam bubble transport to the core region in case of SGT leakage in the primary coolant system of the ELSY (European Lead-cooled SYstem) design. Coolant ow eld and bubble motion are simulated by CFD code Star-CCM+. First, we discuss drag correlations for a steam bubble moving in liquid lead. Thereafter the steady state liquid lead ow eld in the primary system is modeled according to the ELSY design parameters of nominal full power operation. Finally, the consequences of SGT leakage are modeled by injecting bubbles in the steam generator region. An assessment of the probability that bubbles can reach the core region and also accumulate in the primary system, is performed. The most dangerous leakage positions in the SG and bubble sizes are identi ed. Possible design solutions for prevention of core voiding in case of SGTL/R are discussed.

Validation and Application of CFD to Safety-Related Phenomena in Lead-Cooled Fast Reactors

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.QC 20180523</p

Parametric Study of Sloshing Effects in the Primary System of an Isolated Lead-Cooled Fast Reactor

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.</p

Validation of Star-CCM plus for liquid metal thermal-hydraulics using TALL-3D experiment

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).</p

Risk of sloshing in the primary system of a lead cooled fast reactor

Pool-type designs of Lead-cooled Fast Reactor (LFR) aim for commercial viability by simplified engineering solutions and passive safety systems. However, such designs carry the risks related to heavy coolant sloshing in case of seismic event. Sloshing can cause (i) structural damage due to fluid-structure interaction (FSI) and (ii) core damage due to void induced reactivity insertion or due to local heat transfer deterioration. The main goal of this study is to identify the domain of seismic excitation characteristics at the reactor vessel level that can lead to exceedance of the safety limits for structural integrity and core damage. Reference pool-type LFR design used in this study is the European Lead-cooled SYstem (ELSY). Liquid lead sloshing is analyzed with Computational Fluid Dynamics (CFD) method. Outcome of the analysis is divided in two parts. First, different modes of sloshing depending on seismic excitation are identified. These modes are characterized by wave shapes, loads on structures and entrapped void. In the second part we capitalize on the framework of Integrated Deterministic-Probabilistic Safety Assessment (IDPSA) to quantify the risk. Specifically, statistical parameters pertaining to mechanical loads and void transport are quantified and combined with the deterministically obtained data about consequences.</p

Thermal bridge effect of vertical diagonal tie connectors in precast concrete sandwich panels: an experimental and computational study

The purpose of this work was to quantify the thermal bridge effect of vertical diagonal tie connectors in precast concrete sandwich panels (PCSPs). Special interest was in cases where the use of rigid insulation (e.g. PIR) would leave air gaps between insulation boards and diagonal ties, thus intensifying the thermal bridge. A climate chamber experiment using 5 different joint types was performed to gather reference data for CFD model validation. In the experiment, natural convection was observed in joints where no additional insulation was used, i.e. in air cavities. Significantly larger heat fluxes were measured in these cavities compared to insulated joints. The thermal bridging effect was evaluated for a typical PCSP (thermal transmittance without thermal bridges U = 0.11 W/(m²·K)) using CFD software taking into account 3D heat conduction and convection. Simulation results indicate that diagonal ties without adjacent air cavities increased the average thermal transmittance (U-value) of the envelope by 8%, diagonal ties with a 6 mm air cavity – 19...33% and diagonal ties with a 10 mm air cavity – 45...56%. In conclusion, it was found that the joints in insulation caused by diagonal ties affect the overall thermal performance of the building envelope significantly when efforts are not made to fill the air cavities around the connectors.</jats:p