Izvestiya vuzov. Yadernaya Energetika

The peer-reviewed scientific and technology journal. ISSN: 0204-3327

AR-1 experimental model and facility preparation for the purpose of experimental investigation of sodium boiling in fuel subassembly mockup for new generation fast reactor safety justification

5/29/2014 2014 - #01 Global safety, reliability and diagnostics of nuclear power installations

Khafizov R.R. Ashurko Yu.M. Volkov A.V. Ivanov E.F. Privezentsev V.V. Sorokin A.P. Kumskoy V.V.

DOI: https://doi.org/10.26583/npe.2014.1.05

UDC: 621.039.526.034+621.039.546.8:536.26

Sodium coolant boiling in fast reactor core fuel subassemblies is an accidental operating mode of a nuclear power plant (NPP). In case of surge of sodium boiling the reactor core and all NPP technical characteristics must provide stable heat removal from fuel pins surface by means of boiling coolant. Therefore, the design solutions accepted for the reactor core must eliminate any possibilities of cladding melting or core structure damaging and, furthermore, a time reserve should be provided for stabilizing the operational mode.

One of the most vulnerable situations that may lead to coolant boiling in the Liquid metal fast reactor (LMFR) core resulting in a severe accident is an ULOF (Unprotected Loss of Flow) accident implying simultaneous main pumps electric supply failure and emergency shutdown system breakdown.

As part of the program for safety analysis and justification of perspective LMFR, checking some design solutions and obtaining experimental data for computer code verification work is currently underway to prepare sodium boiling experiments at the IPPE. The experiments will be focused on heat exchange analysis inside the fuel subassembly mockup in different sodium boiling regimes. This experimental work is part of the unified calculationexperimental complex developing program which will make possible to accurately determine operation modes for both operating and projected NPP.

The article presents a brief review of the sodium cooled fast reactor accidental operating mode due to coolant boiling in a fuel subassembly channel by the example of ULOF. Also the authors touch upon the issue of experimental investigations required for verification of twophase liquid metal coolant models included in the COREMELT code. In addition, the article includes some information on preparation of the experimental facility for sodium boiling in fuel a subassembly mockup is provided as well as a description of the experimental data acquisition system.


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nuclear fuel cycle non-proliferation of nuclear weapons fissile materials plutonium enriched uranium fast reactors and thermal reactors