SAC-CFR computer code verification with Experimental Breeder Reactor II loss-of-primary-flow-without-scram tests data

Guo, Chao; Lu, Daogang; Sui, Danting; Zhang, Xun

doi:10.1016/j.anucene.2014.03.038

Cited by 3 publications

(2 citation statements)

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“…SAC-3D uses the parallel-channel approach to conduct the reactor vessel thermal-hydraulics calculation, which does not consider the coolant crossflow between the internal components. Simultaneously, the model does not consider the fuel component's axial heat conduction [9]. In the preliminary work, we used only 9 channels to simulate the core flow path and did not consider the bypass flow [3].…”

Section: Ermal-hydraulics Calculationmentioning

confidence: 99%

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Benchmark Analysis on Loss-of-Flow-without-Scram Test of FFTF Using Refined SAC-3D Models

Lyu

Sui

2021

Science and Technology of Nuclear Installations

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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.

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Section: Ermal-hydraulics Calculationmentioning

confidence: 99%

“…e temperature calculation nodes of the heat tube metal and shell wall are defined in the center of the control volume. e temperature calculation nodes of the coolant are defined at the boundary of the control body [9]. e energy equation of the primary side coolant control volume can be written as…”

Section: Ermal-hydraulics Calculationmentioning

confidence: 99%