Safety assessment for the MBIR reactor using the RELAP code
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 duringoperation 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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- Idaho National Engineering Laboratory Lockheed Idaho Technologies Company Idaho Falls, Idaho 83415 «RELAP5/MOD3 CODE MANUAL VOLUME IV: MODELS AND CORRELATIONS».
- Kirillov P.L., Bogoslovskaya G.P. Heat transfer in nuclear power plants: Textbook for higher education. Moscow, Energoatomizdat Publ., 2000 (in Russian).