Application of CFD to Safety and Thermal-Hydraulic Analysis of Lead-Cooled Systems
Independent thesis Advanced level (degree of Master (Two Years)), 20 credits / 30 HE creditsStudent thesis
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.
Place, publisher, year, edition, pages
2011. , 55 p.
Trita-FYS, ISSN 0280-316X
Coupled Codes, Verification&Validation, CFD, System Thermal-Hydraulics, Lead Cooled sys-
IdentifiersURN: urn:nbn:se:kth:diva-37806OAI: oai:DiVA.org:kth-37806DiVA: diva2:435252
Master of Science - Nuclear Energy Engineering
UppsokPhysics, Chemistry, Mathematics
Arzhanov, Vasily, Researcher