Phase III Indo-UK Collaboration New Reactor Design from a Safety Systems Perspective: Passivity and Grace Time A.K. Nayak Bhabha Atomic Research Centre Trombay, Mumbai, India

Indo-UK Scientists Involved in this Project BARC Team Dr A K Nayak Dr D K Chandraker Mr Arnab Dasgupta Mr S V Prasad Mr Alok Visnoi UK Team Prof. Simon Walker Research Scholars

Enhanced safety – a major objective of new reactors In the entire history of commercial nuclear power so far, three major accidents leading to damage of reactor core, have taken place: Three Mile Island (1979), Chernobyl (1986) and recent Fukushima (2011)). These accidents have reinforced the necessity to further improve safety in the nuclear power plant design. The lessons from these accidents have led to a higher attention for improvement in man-machine interfaces to minimize human errors Even though the safety performance of current fleet of nuclear reactors has been excellent, next generation nuclear energy systems need to have even higher safety goals. Criteria for safety should be consistent with the increase in number of nuclear facilities Criteria for safety should address to a possible need to locate nuclear facility close to population centers in accordance with siting rules generally applicable to conventional industrial facilities

Passive safety Systems Play Major Role in New Reactor Designs For meeting enhanced safety goals of “no emergency planning in public domain” – passive safety systems are instrumental which can not only provide large grace period but also are “simpler” and “reliable”.

Issues with Design of Passive Safety Systems Lack of Plant Data and Operational Experience Lack of sufficient experimental data from Integral Facilities or even from Separate Effect Tests in order to understand their performance characteristics not only at normal operation but also during transients and accidents. Difficulty in modeling the physical behaviour of such systems; particularly Multi-dmeinsional natural convection Critical heat flux under oscillatory condition (OICHF) Molten corium coolability – IVR with external vessel cooling Buoyant flows and cold trap formation Heat rejection by passive air condenser

Tests Conducted in BARC for Oscillatory Induced CHF Film flow rate evaluation experiments at higher pressure and higher power allowing flow conditions similar to BWRs, i.e., 70 bar, 285oC and 1000-2000 kg/m2s. OICHF : Critical Heat Flux experiments with flow oscillations. This is relevant to fault conditions in reactors as well as safety systems employing NC. The CHF-LUT is unable to predict OICHF reliably Typical OICHF data point (CHIL measurements) Sudden rise in surface temperature upon sustained flow oscillations A mechanistically based model for OICHF is one of the objectives of the phase-III project

IN-VESSEL RETENTION WITH EXTERNAL COOLING During a postulated severe accident leading to complete or partial core melting, the priority is to prevent release to the environment of the melted corium. Some advanced reactor designs rely on demonstrating in-vessel retention of the corium, via external reactor vessel cooling through natural convection and boiling after flooding the reactor cavity with water. This second strand of our ‘Grace Time’ proposal is a combined experimental and modelling programme addressing this.

MODELLING ISSUES WITH IN-VESSEL RETENTION AND EXTERNAL COOLING In the in-vessel retention approach, the heat transfer from the molten corium to the water outside through the vessel wall is complex, involving multiple heat transfer modes including conduction, convection, boiling and radiation. Molten corium, which still generates heat, loses heat to upper in vessel structures by radiation, and to the vessel wall by natural convection/conduction. A low thermal conductivity solid ‘crust’ is formed between the vessel wall and the melt pool, which can limit the heat transfer. The extent of heat transfer from the molten corium pool depends upon natural convection inside the pool.

BARC CONTRIBUTION – EXPERIMENTS FOR IVR WITH DECAY HEAT Objective: To determine the extent of heat transfer from the melt pool with decay heat to the outside vault water through vessel wall Melt Test section details: Test vessel : 300 mm (1:26 scale) Length : 456 mm (1:13 scale) Thickness : 26 mm (1:1.25 scale) Material : SS 304L (Same as actual) Volume of water : 400 l Melt quantity : 60 kg Melt volume : 24 l Heater :Cartridge type Power : 2.3 kW each Number : 4 Nos. Total Power : 9.2kW Heat generation in melt : 0.67 MW/m3 TV Inner & Outer Water tank

BARC CONTRIBUTION – EXPERIMENTS FOR IVR WITH DECAY HEAT After 3 hrs, steady state of melt temperature was observed and after 4 hrs, the temperature of melt started increasing. With heat generation in melt, it was observed that initially crust thickness grew up to 40 mm and then it stabilised at that thickness. Even though the melt temperature was ~ 1200°C, maximum average inner surface temperature of vessel was 249°C and maximum average outer surface temperature of vessel was 75°C Thus, no steam generation took place When the vessel was opened, no gap between the crust and vessel was observed The crust retained the high temp melt, providing insulation to the vessel wall. Melt temperature Crust thickness CV Inner temperature CV outer temperature Melt Vessel

CLOSING REMARKS Considering the safety goals of advanced reactor designs “passivity” and “grace period” are important aspects Under this project Critical heat flux under oscillatory condition (OICHF) Molten corium coolability – IVR with external vessel cooling Buoyant flows and cold trap formation Experiments are being continued for IVR and OICHF in BARC Data available will be utilized for development of models for the above specific issues.

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