1. 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

2. 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

3. 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

4. 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”.

5. 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

6. 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 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

7. 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.

8. 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.

9. 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

10. 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

11. 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.

12. Thank You