Risk of sloshing in the primary system of a lead cooled fast reactor
2014 (English)In: PSAM 2014 - Probabilistic Safety Assessment and Management, 2014Conference paper (Refereed)
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.
Place, publisher, year, edition, pages
IdentifiersURN: urn:nbn:se:kth:diva-165399ScopusID: 2-s2.0-84925070484OAI: oai:DiVA.org:kth-165399DiVA: diva2:808237
12th International Probabilistic Safety Assessment and Management Conference, PSAM 2014; Sheraton WaikikiHonolulu; United States; 22 June 2014 through 27 June 2014
QC 201504282015-04-272015-04-272015-04-28Bibliographically approved