Neutronic-coupled Thermal Hydraulic Analysis of Two-phase Cooled Flow Blockage Condition in LFR


Student thesis: Doctoral Thesis

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Awarding Institution
  • Minghuang WANG (External person) (External Supervisor)
  • Taosheng LI (External person) (External Supervisor)
  • Jiyun ZHAO (Supervisor)
Award date20 Feb 2024


Flow blockage in lead-based fast reactors is considered an important issue in safety analysis. The existing researches mainly focus on characteristics of flow blockage in LBE-cooled single-phase flow under uniform heat flux conditions. However, the LBE-argon two phase flow should be considered in gas-lift enhanced natural circulations. Also, power distributions along the radial and axial directions in the reactor core should be non-uniform in actual operating conditions. Thus, this study proposes a neutronic-coupled thermal-hydraulic model to analyze the coupling effect on flow blockage in LBE-argon two-phase flow conditions in the scale of a single fuel assembly and the CLEAR-S core simulator. The thesis is composed of three parts, i.e., the model validation part, the self-developed coding and performance analysis, and the code-application and condition-analysis of flow blockage.

Firstly, for the model validation part, the calculating results of flow blockage module have been validated through comparison with reported experimental data and empirical correlations. The optimal parameters of mesh grids as well as the turbulent models are selected. The relative error of temperature is within 5%, and the error for the friction factor is within 10%. The outputs of the neutronic-coupled module are consistent with reported benchmarks, and the difference schemes and approximation methods of neutron density are compared to make sure the relative error is within 10-5 order of magnitude. For validation of LBE-argon two-phase module, on the one hand, the coefficient for the fitting curve of two-phase flow velocities is acceptable. On the other hand, comparisons of shapes for steady rising bubbles with theoretical diagrams are conducted, which shows good consistency. By the way, the dominant forces for interface of two-phase flow are determined by calculating the dimensionless numbers. It is concluded that the VOF method is suitable for interface calculation of LBE-argon two-phase flow, and the inertial effect and surface tension cannot be neglected.

Secondly, from the perspective of coding performance, comparisons have been conducted for difference schemes, approximation methods and reactivity insertion manners. For the magnitude function of neutron flux, the 6-groups neutron density approximation under implicit backward Euler scheme using the Crack-Nicolson (C-N) finite difference method is considered to be better than the full- implicit scheme. For the shape function of neutron flux, using the exported Bessel function is precise enough to realize the radial power distribution. By comparison of different reactivity insertion manners, the dimensionless temperature is more sensitive when there is a large step insertion of reactivity, and the coolant temperature and rod temperature rise close to linearly when there is a ramp insertion of reactivity.

Then, the neutronic-coupled thermal hydraulic model has been applied to analyze the porous blockage condition in LBE-cooled fuel assemblies, LBE-argon two-phase cooled fuel assemblies, and the heated zone of LBE-cooled CLEAR-S core simulator. At the scale of fuel assemblies, flow resistance and backward flow upstream or inside the blockage regions are affected by porosities. Flow recovery phenomenon in the wake region downstream of the blockage could lead to severer overheating than the solid blockage itself. For example, the maximum temperature for central six-subchannel blockage in steady state condition is 589 K, which is 50 K higher than the solid blockage region. In the two-phase cooled blockage scenarios, the behavior of argon bubbles can be divided into three categories, i.e., gas escape, dispersion and being sealed. High volume fraction of gas phase could cause local overheating and an "increase" of static pressure. Two boundary conditions are simulated for comparison, the overheating to the average temperature of VOFAr 0.6 when Re equals to 5000 at 0.15 s is 5.4 K for internal boundary, which is 17.2 K lower than wall boundary condition of the same VOFAr and inlet Re. At the scale of the CLEAR-S core simulator, the flow blockage phenomenon may lead to stratifications to temperature and velocity distributions at the outlet of the core region. For the transient state, the effect of reactivity change to the reactor power are calculated and compared. The insertion manners include the constant insertion, ramp insertion, and different magnitudes of step insertions. Results show that the reactor power increase more obviously for the larger decreasing magnitude of absorption cross-section, especially in the radial directions.

Coupling schemes based on the Point Kinetic model and the 3D quasi-static model have been proposed in this research, respectively. By means of the function-export of Bessel function, it can realize spatial power distributions considering reactivity feedback and real-time coupling analysis of the blockage phenomenon at the scales of both the assemblies and the heated core simulator. The porous approximation of flow blockage is proposed from obtaining the heterogenous velocities, and the blockage characteristics cooled by LBE-argon two-phase flow are given in this numerical analysis.

This research can provide a reference for the flow blockage safety analysis in lead-based fast reactors or loop designs.

    Research areas

  • Lead-based fast reactor, Fuel assembly, Flow blockage, Two-phase flow, Neutronic-thermal hydraulic coupling, Computational fluid dynamics