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Apr 19 – 22, 2022
Vienna, Austria
Europe/Vienna timezone
FR22 starts in Vienna 19 - 22 April 2022 Online Stream:

Thermal Hydraulic Simulation of Loss of Flow Without Scram Test in FFTF using DYANA-P code

Apr 20, 2022, 1:40 PM
Vienna, Austria

Vienna, Austria

POSTER Track 2. Fast Reactor Safety Poster Session


Vikram Govindarajan (IGCAR)


System dynamic simulation of loss of flow without SCRAM test carried out in FFTF has been carried out using plant dynamics code DYANA-P. DYANA-P has one-dimensional models for various sub-systems of sodium cooled fast reactor. Thermal models are based on heat balance between various sections exchanging heat. Hydraulic model is based on momentum balance between various flow segments in sodium circuits. Pumps are modeled with characteristics derived from generalized homologous characteristics. Neutronic model for the core is based on point kinetics approximation. A similar modeling methodology is adopted in the formulation of computer codes such as DYNAM and SIFDYN developed for performing dynamic calculations in the Fast Breeder Test Reactor (FBTR). These codes have been validated through various tests carried out in FBTR.
In the hydraulics modeling of FFTF, flow evolutions in different coolant loops as well as core channels are modeled through integral momentum balance approach. However, level variations in coolant plenums are not modeled. The reactor core comprising of 80 fuel sub-assemblies is represented through six core channels and decay power in the core is evaluated using an empirical correlation. Various reactivity feedback effects due to (i) radial expansion of core support structure, (ii) control rod drive line expansion, (iii) volumetric expansion of sodium in the core, (iv) axial expansion of cladding, (v) axial expansion of fuel, (vi) Doppler Effect due to changes in fuel temperature, (vii) sodium level changes inside GEM sub-assemblies, etc. are considered.
As a result of primary pump speed reduction, the pump developed head decreases causing a decrease in pump flow rates. The reduction in primary loop flow rates causes GEM sodium levels to decrease resulting in large negative feedback reactivity addition leading to reactor shutdown. The natural circulation flow sets up in the primary circuit. Initially, because of the reduction in flow rate, the sodium outlet temperatures from core channels start increasing. Later, because of the power reduction due to the action of GEM, the temperature starts decreasing. The rate of power reduction is less than the rate of flow reduction. Hence, a second temperature peak appears. Further, the temperatures decrease continuously as the decay heat decreases continuously, and also the flow rate has become constant. No significant temperature rise is predicted in any reactor component (maximum clad temperature predicted is ~100 °C above its nominal value for a few seconds). Continuous decay heat removal from core to DHX is established through IHX.

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Primary authors

Vikram Govindarajan (IGCAR) Natesan Kumaresan (Scientific Officer) Mr Devan K. (IGCAR) Mr Rajendrakumar M. (Indira Gandhi Centre for Atomic Research) RAGHUPATHY S. (Indira Gandhi Centre for Atomic Research, Kalpakkam)

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