快中子通量实验堆失流事故三维数值模拟

为验证计算流体动力学（CFD）方法在钠冷快堆失流事故模拟计算中的可靠性和可行性，针对快中子通量实验堆（FFTF），建立了包含冷池、热池、堆芯在内的全三维模型，其中堆芯组件简化为多孔介质模型，堆芯保留了盒间特征，各类隔板简化为无厚度面。失流事故下主要参数计算结果与实验数据的对比表明，CFD方法能有效捕捉冷池、热池以及盒间复杂的流动换热现象，堆芯最热组件的位置在瞬态过程发生了变化，热管段出口温度与实验值符合良好，装有温度测点的组件出口温度模拟值较实验值低。CFD方法仍需针对组件盒间进行相应的模型开发和验证，此外还需进行大量全堆级别的实验验证，以保证计算结果的合理性。

Three-dimensional Numerical Simulation of Loss of Flow Accident for Fast Flux Test Facility

In order to assess transient responses and parameters of sodium cooled fast reactor (SFR) under all type of accidents whether satisfy the acceptance criteria, and validate the reliability and feasibility of computational fluid dynamics model when applying lower Prandtl fluids, a CFD method substituting for traditional system analysis was adopted to simulate loss of flow without scram experiment for fast flux test facility (FFTF), which is a loop type SFR. The three dimensional model of FFTF was constructed, which includes inlet plenum, annular plenum, in vessel storage, peripheral plenum, core basket, hot pool, and core. It should be noted that the fuel assemblies were simplified into porous mediums while the inter wrapper was reserved, and the barriers of pool were simplified into plate without thickness, the standard k ε turbulence model was adopted. In the simulation process, two procedures are adopted, i.e., steady debug and transient analysis. In the steady debug, the porous parameters of fuel assemblies were adjusted by code, for the wire region, the empirical relations were adopted and kept uncharged, for inlet region, because the local form resistance is always hard to get, the porous parameters of those regions were adjusted by comparing the calculated mass flow rate with design mass flow rate until the error was within acceptance range. In the transient analysis step, the loss of flow without scram #13 (LOFWOS#13) was selected to simulate, and the simulation time lasts 900 s. The numerical results were compared with experiments and the local phenomena of FFTF during accidents process were analyzed. For the comparison between numerical results and experiments, the numerical results of outlet temperature have a little difference between experiments results for the assembly which installed thermocouple. For the local phenomena, the CFD method could capture well for the complex thermal hydraulic phenomena of hot pool such as thermal stratification, and the temperature changes only 4 K, which imply that the heat capacity of the huge pool has large thermal-inertia which is a positive influence for safety of SFR. And from the temperature distribution of core, the location of hot channel in the core is changed versus time, and the inter wrapper flow is highly dependent on core temperature distribution, and under the lower power and lower mass flow rate condition, the inter wrapper could flat core temperature distribution. Those phenomena could not be tracked by traditional system code, and show that it is necessity to adopt three dimensional CFD method in the safety analysis. However, due to small difference between experiments results and lack of enough validation of CFD methods, there are still to develop accurate model and be validated by much more experiment especially full reactor experiments to satisfy tools requirements in security review process for nuclear regulatory.