Thermal hydraulic parametric investigation of decay heat removal from degraded core of a sodium cooled fast Breeder reactor Lokesh Verma a , Anil Kumar Sharma b,⇑ , K. Velusamy b a Department of Physics and Astrophysics, University of Delhi, Delhi 110007, India b Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam, India highlights  Decay heat removal from degraded core of a typical SFR is highlighted.  Influence of number of DHXs in operation on PAHR is analyzed.  Investigations on structural integrity of the inner vessel and core catcher.  Feasibility study for retention of a part of debris in upper pool of SFR. article info Article history: Received 21 July 2016 Received in revised form 8 December 2016 Accepted 12 December 2016 Keywords: Core disruptive accident Post accident decay heat removal Natural convection Fast reactor Core catcher abstract Ensuring post accident decay heat removal with high degree of reliability following a Core Disruptive Accident (CDA) is very important in the design of sodium cooled fast reactors (SFR). In the recent past, a lot of research has been done towards the design of an in-vessel core catcher below the grid plate to prevent the core debris reaching the main vessel in a pool type SFR. However, during an energetic CDA, the entire core debris is unlikely to reach the core catcher. A significant part of the debris is likely to settle in core periphery between radial shielding subassemblies and the inner vessel. Failure of inner vessel due to the decay heat can lead to core debris reaching the main vessel and threatening its integrity. On the other hand, retention of a part of debris in core periphery can reduce the load on main core catcher. Towards achieving an optimum design of SFR and safety evaluation, it is essential to quantify the amount of heat generating core debris that can be retained safely within the primary vessel. This has been performed by a mathematical simulation comprising solution of 2-D transient form of the gov- erning equations of turbulent sodium flow and heat transfer with Boussinesq approximations. The con- jugate conduction-convection model adopted for this purpose is validated against in-house experimental data. Transient evolutions of natural convection in the pools and structural temperatures in critical com- ponents have been predicted. It is found that 50% of the core debris can be safely accommodated in the gap between radial shielding subassemblies and inner vessel without exceeding structural temperature limit. It is also established that a single plate core catcher can safely accommodate decay heat arising due to 70% of the core debris by establishing natural circulation in the lower sodium pool. The influence of heat removal rate by natural circulation on availability of the number of decay heat exchangers (DHX) dipped in the upper pool is also analyzed. It is seen that the temperatures in the inner vessel, source plate and the maximum debris temperature do not increase significantly even when the DHXs are deployed 5 h after the accident, demonstrating the benefit of large thermal inertia of the pool. Ó 2016 Elsevier B.V. All rights reserved. 1. Introduction A severe accident scenario like the Core Disruptive Accident (CDA) leading to whole core meltdown is a very rare incident (frequency of occurrence <10 6 /ry). A CDA can be initiated by two events in a pool type SFR (Chellapandi et al., 2013). The first event is ULOFA (Unprotected Loss of Flow Accident), which can hamper the structural integrity of the main vessel. The second event is UTOPA (Unprotected Transient Overpower Accident), which occurs due to uncontrolled withdrawal of control rods caus- ing a surge in the reactivity. The comparison of the behavior of two core designs for the ASTRID reactor in case of severe accidents is http://dx.doi.org/10.1016/j.nucengdes.2016.12.017 0029-5493/Ó 2016 Elsevier B.V. All rights reserved. ⇑ Corresponding author. E-mail address: aksharma@igcar.gov.in (A. Kumar Sharma). Nuclear Engineering and Design 313 (2017) 285–295 Contents lists available at ScienceDirect Nuclear Engineering and Design journal homepage: www.elsevier.com/locate/nucengdes