Abstract
The Fast Flux Test Facility (FFTF) is a liquid sodium-cooled nuclear reactor designed by the Westinghouse Electric Corporation for the U.S. Department of Energy. In July 1986, a series of unprotected transients were performed to demonstrate the passive safety of FFTF. Among these, a total of 13 loss-of-flow-without scram (LOFWOS) tests were conducted to confirm the liquid metal reactor safety margins, provide data for computer code validation, and demonstrate the inherent and passive safety benefits of specific design features. In our preliminary work, we have performed relatively coarse modeling of the FFTF. To better predict the transient behavior of FFTF LOFWOS test #13, we modeled it using a more refined thermal-hydraulics model. In this paper, we simulate FFTF LOFWOS test #13 with the system safety analysis code SAC-3D according to the benchmark specifications provided by Argonne National Laboratory (ANL). The simulation range includes the primary and secondary circuits. The reactor core was modeled by the built-in 3D neutronics calculation module and the parallel-channel thermal-hydraulics calculation module. To better predict the reactivity feedback introduced by coolant level variations within the GEMs, a real-time macro cross-section homogenization processing module was developed. The steady-state power distribution was calculated as the transient simulation initial boundary conditions. In general, both the steady-state calculation results and the whole-plant transient behavior predictions are in good agreement with the measured data. The relatively large deviations in transient simulation occur in the outlet temperature predictions of the PIOTA in row 6. It can be preliminarily explained by the reason for neglecting the heat transfer between channels in this model.
Highlights
In 2017, the International Atomic Energy Agency (IAEA) initialed a Coordinated Research Project (CRP) on benchmark analysis of Fast Flux Test Facility (FFTF) loss-of-flow-without scram (LOFWOS) test
A total of 25 organizations from 13 countries participated in this CRP. e overall objective of the CRP is to improve the Member States’ analytical capabilities in the field of fast reactor simulation and design. e test defined in this benchmark is LOFWOS test #13, which was initiated at 50% power and 100% flow with the pump pony motors turned off. e conditions of the LOFWOS test, along with the feedback from FFTF’s limited free bow core restraint system and the novel passive safety gas expansion modules (GEMs), pose a very challenging and uniquely valuable benchmark exercise
There are a total of eight types of assemblies within the core. e in core shim assemblies (ICSAs), fracture mechanics assembly (FMA), and materials open test assembly (MOTA) have all different geometries and material specifications, as the ICSA and MOTA are very similar to driver fuel assemblies (DFA) while, for instance, the FMA is mostly comparable to reflector assemblies (REFL)
Summary
In 2017, the International Atomic Energy Agency (IAEA) initialed a Coordinated Research Project (CRP) on benchmark analysis of FFTF loss-of-flow-without scram (LOFWOS) test. North China Electric Power University (NCEPU) participated in both neutronics and thermal-hydraulics benchmark analysis tasks of the CRP. E fast reactor system analysis code SAC-3D was selected to simulate the steadystate and transient-state for this benchmark. We have performed relatively coarse modeling of the FFTF and calculated neutronics and thermal-hydraulics transients [3]. E neutron calculation results are in good agreement, but the thermal-hydraulic transient results deviate significantly. E blind calculation phase is over; most of the measured data has been made available to the benchmark participants Some data, such as power distribution in a steady state, are still “blind”. For the “blind” part, we compared the simulation results with the results of other participants who applied similar system analysis software for modeling calculations. For the “blind” part, we compared the simulation results with the results of other participants who applied similar system analysis software for modeling calculations. e neutronics calculation results in both steady-state and transient-state obtained good agreement in general. e trends of the thermal-hydraulics transient calculations are in good agreement, but the values somewhat deviate
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