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Simulation of Reactor Core Accidents


Experimental and Numerical Study of Liquid Metal Boiling as Simulation of Accidents in fast Reactor Core

Tech Area / Field

  • FIR-EXP/Experiments/Fission Reactors

3 Approved without Funding

Registration date

Leading Institute
FEI (IPPE), Russia, Kaluga reg., Obninsk


  • Tokyo Institute of Technology, Japan, Tokyo\nCEA / DRN / DER / CEN Cadarache, France, Cadarache\nENEA, Italy, Bologna

Project summary


Emergency standards of countries involved into the development and operation of nuclear power facilities of different purposes cooled by liquid metal coolants consider an onset of liquid metal boiling within reactor core as accident situation. Previously performed first experimental and numerical investigations into the liquid metal boiling in the pin bundles simulating fast reactor fuel subassembly shown that, under definite conditions, a stable heat removal from the structure to liquid metal can be observed without attainment of an accident, that improved reliability of nuclear power facilities.

The purpose of the project is to study an opportunity for stable heat removal in fast reactor core under low-flow conditions (simulating reduction of coolant flow rate or loss-of-flow accident, emergency heat removal) in case of boiling of liquid metal coolant; to develop reliable advanced prediction approaches, as well as to study detection techniques for liquid metal boiling and ways to delay heat transfer crisis due to using special means.

The final objectives of the suggested project are as follow:

· To define boundaries of steady boiling in view of influence of various factors (patterns of changes in power, flow rate, pressure, state of fuel pin surface, dimension of pin bundle, availability of parallel channels etc.);
· To study the following boiling characteristics: boundaries of boiling patterns; stability of coolant natural circulation in the contour containing one subassembly and system of parallel subassemblies; overheating needed for boiling onset; critical heat flux; characteristics of post dryout heat transfer;
· To give recommendations for setting a steady heat removal in fast reactor core in the event of liquid metal boiling;
· To develop an advanced prediction procedure and fast reactor core boiling thermohydraulic code;
· To have analysed numerically a thermal hydraulic characteristics of fast reactor core in the event of the boiling to occur under conditions of emergency heat removal.


Data on boundaries of steady heat removal in fast reactor core in the event of liquid metal boiling, advanced computer code developed to analyse thermohydraulic characteristics of fast reactor core in the event of liquid metal boiling, recommendations on using of detection means in liquid metal boiling , as well as on ways of delay of heat transfer crisis.


Research Laboratory for Nuclear Reactors at the Tokyo Institute of Technology can contribute to the project by simulating the experiments and by providing the computational results from two-fluid subchannel code for sodium boiling in a fuel subassembly. It can share the technical expertise gained during the code developments and validation. The code developmental efforts at the TIT are directed to the extending the analytical capabilities of sodium boiling two-phase flow to those of fuel pin disintegration including the cladding/fuel melting.


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