Analysis of unprotected loss of flow accident in a sodium cooled fast reactor using coupled thermal hydraulics neutronics code ASTRA

• Published in: Progress in nuclear energy (New series) 2023-09, Vol.163, p.104801, Article 104801
• Main Authors: Narendran, Yadu, Natesan, K., Devan, K., John Arul, A.
• Format: Article
• Language: English
• Subjects: Coupled thermal hydraulics neutronics model; Sodium boiling; Sodium cooled fast reactor; Unprotected loss of flow accident
• ISSN: 0149-1970
• DOI: 10.1016/j.pnucene.2023.104801

The Unprotected loss of flow accident (ULOFA) is a severe accident that may lead to core disruption in a sodium cooled fast reactor (SFR). It is investigated as a part of the defense in depth concept in nuclear safety. A coupled thermal hydraulics neutronics model is required to model the severe accident scenario where reactivity is affected by the changes in the temperature of the core components, coolant boiling, and material distribution during the transient. The coupled code previously developed for the analysis of total instantaneous blockage (ASTRA) is improved for the study of ULOFA by adding various models, viz., (i) a one-dimensional two-fluid sodium boiling model, (ii) a simple primary hydraulics model, (iii) a simple clad motion model, and (iv) an improved point kinetics solver. The boiling model is validated with different loss of flow experiments. The improved code is used to analyze ULOFA in a 500 MWe medium size sodium cooled fast reactor up to the onset of fuel melting. At 21.5s, coolant boiling is initiated in the central channel. The temperature evolution of the core components, reactivity feedbacks, power evolutions, and coolant voiding propagation are calculated. The code predicts early fuel melting at 24.9s due to the power excursion by the early voiding in the fuel channels compared to the study without including the boiling model. Boiling is started in five out of ten representative fuel subassembly (SA) channels at the time of fuel melting. •Unprotected loss of flow accident (ULOFA) in a sodium cooled fast reactor is analyzed using the coupled thermal hydraulics neutronics code ASTRA.•One dimensional two fluid sodium boiling model is developed and validated with loss of flow experiments.•Voiding front, reactivity feedbacks, and power evolution during ULOFA are predicted.•Fast voiding and early fuel melting in the core within less than 25s has been observed based on the analysis.