Safety assessment for the MBIR reactor using the RELAP code

УДК 621.039.526

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The possibility of performing analysis of an emergency situation at sodium%cooled
reactors using the RELAP code is discussed. The difficulty is that RELAP does not take into
account the liquid metal coolant.
The problem was considered in relation to the objective analysis of the emergency
situation at the MBIR reactor associated with the introduction of positive reactivity of the
reactor control and safety system. Scenarios of ejection of one of the control rods in the
reactor control and safety system and unauthorized extraction with regular speed during
operation of the reactor at nominal power level were considered.
Imitation of sodium coolant was performed by superheated steam with preservation of
the exhaust coolant capacity. To achieve it, the equivalent steam flow was calculated and
heat%transfer coefficients were replaced by those for sodium.
As a result, a calculation model of the MBIR nuclear reactor was developed using the
syntax of the RELAP code and the model was used to calculate the transients. Analysis of
the results obtained and their comparison with those obtained by other software codes
showed good agreement.

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2. SC «ROSATOM». Conceptual design of a multipurpose fast research reactor. Explanatory note No. 4486/3032 from 15.12.2008 (in Russian).
3. Idaho National Engineering Laboratory Lockheed Idaho Technologies Company Idaho Falls, Idaho 83415 «RELAP5/MOD3 CODE MANUAL VOLUME IV: MODELS AND CORRELATIONS».
4. Kirillov P.L., Bogoslovskaya G.P. Heat transfer in nuclear power plants: Textbook for higher education. Moscow, Energoatomizdat Publ., 2000 (in Russian).

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