Fast Reactors and Accelerator Driven Systems Knowledge Base

Conference Article: 3-D thermal hydraulic analysis of transient heat removal from fast reactor core using immersion coolers

Chvetsov, I.; Volkov, A. (State Scientific Center of Russian Federation, Institute of Physics and Power Engineering, Obninsk, Kaluga Region (Russian Federation))

Abstract

For advanced fast reactors (EFR, BN-600M, BN-1600, CEFR) the special complementary loop is envisaged in order to ensure the decay heat removal from the core in the case of LOF accidents. This complementary loop includes immersion coolers that are located in the hot reactor plenum. To analyze the transient process in the reactor when immersion coolers come into operation one needs to involve 3-D thermal hydraulics code. Furthermore sometimes the problem becomes more complicated due to necessity of simulation of the thermal hydraulics processes into the core inter-wrapper space. For example on BN-600M and CEFR reactors it is supposed to ensure the effective removal of decay heat from core subassemblies by specially arranged internal circulation circuit: “inter-wrapper space”. For thermal hydraulics analysis of the transients in the core and in the whole reactor including hot plenum with immersion coolers and considering heat and mass exchange between the main sodium flow and sodium that moves in the inter-wrapper space the code GRJFIC (the version of GRIF code family) was developed in IPPE. GRIFIC code was tested on experimental data obtained on RAMONA rig under conditions simulating decay heat removal of a reactor with the use of immersion coolers. Comparison has been made of calculated and experimental result, such as integral characteristics (flow rate through the core and water temperature at the core inlet and outlet) and the local temperatures (at thermocouple location) as well. In order to show the capabilities of the code some results of the transient analysis of heat removal from the core of BN-6OOM - type reactor under loss-of-flow accident are presented.

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key words: fter-heat removal; beloyarsk-3 reactor; bn-1600 reactor; cefr reactor; coolant loops; g codes; heat exchangers; hydrodynamics; joyo reactor; loss of flow; simulation; thermal analysis; three-dimensional calculations; transients
Reference:
Technical committee meeting on methods and codes for calculations of thermohydraulic parameters for fuel, absorber pins and assemblies of LMFR's with traditional and burner cores. Obninsk (Russian Federation) 27-31 Jul 1998
International Atomic Energy Agency, Vienna (Austria)
IAEA-TECDOC--1157, pp:89-102