Thermal-hydraulic Transient Analysis of Benchmark for FFTF LOFWOS Test #13 Based on FR-Sdaso Code

FFTF LOFWOS Test #13 is an unprotected loss of flow test conducted on the FFTF sodium-cooled loop fast reactor in the United States, aimed at demonstrating the inherent safety characteristics of the reactor. To verify advanced neutron science, thermal-hydraulic, and safety analysis programs for sodium-cooled fast reactors and improve simulation and analysis capabilities in the field of sodium-cooled fast reactors, IAEA initiated the FFTF LOFWOS Test #13 benchmark in 2017. To further verify the system program FR-Sdaso independently developed by China Institute of Atomic Energy, the code development team participated in the benchmark project. A FFTF benchmark simulation model was established using the FR-Sdaso code with the primary pump speed, secondary circuit flow rate, and sodium temperature at the outlet of the air heat exchanger as boundary conditions. The calculation of core fission power adopted a point reactor model, while the decay power adopted a lumped parameter model, considering Doppler, sodium density, axial expansion, radial expansion, control rod drive mechanism expansion, and GEM feedback reactivity. The core thermal calculation adopted a single channel model, dividing the core into 9 thermal channels. The inlet and outlet areas of the primary reactor core adopted a lumped parameter model, while the heat exchanger and pipelines adopt a one-dimensional control volume model. The key parameters that could be directly compared with experimental measurement results were analyzed and compared, such as flow rate of each loop in the first circuit, core power, total reactivity of the core, sodium temperature at the outlet of PIOTA subassemblies in the second and sixth rows of the core, temperature in the cold and hot legs of the first and the secondary circuits. The comparison analysis between the calculation results and the FFTF LOFWOS Test #13 test data shows that the FR-Sdaso program can effectively predict the transient changes in reactor power, as well as the temperature and flow rate of the first and second circuits after an unprotected loss of flow accident. The maximum relative deviation between the calculated and experimental values of the reactor decay power during the natural circulation stage is −7.1%, and the maximum deviation of the natural circulation flow rate of the three circuits in the first circuit from the initial value is 0.65%. For the second and sixth row PIOTA components, the calculated outlet temperature is approximately 25.5 ℃ higher than the experimental measurement due to the lack of consideration of transient power distribution changes and heat transfer between components in the simulation. The calculated results are more conservative. The analysis results of the FFTF LOFWOS Test #13 benchmark example using the FR-Sdaso program preliminarily verify the correctness of the program and the thermal-hydraulic models of the first and second circuits.