Subchannel analysis code

A subchannel analysis code at supercritical pressure was developed at the University of Tokyo [23, 24]. It has been applied to thermal-hydraulic fuel assembly design and has been used to evaluate PCST at supercritical pressure. The subchannnel analysis model and some results obtained by it for the Super LWR fuel assembly design are described in this section. 

Subchannel analysis is based on a control volume approach. A coolant flow channel is treated as a control volume which interacts with an adjacent control volume through gaps between the fuel pins. The subchannel analysis method consists of following four governing equations at the steady state  

The first term in (2.23) represents axial mass flow change and the second term denotes mass transfer from adjacent subcharuiels, j. 

The left-hand side of (2.24) represents the change of axial force. The first term on the right-hand side denotes the axial change of pressure force and the second term is a pressure loss term by frictional and form loss. The third term represents gravitational force and the last gives axial momentum exchange between adjacent channels. 

The transverse momentum equation represents the momentum exchange in the transverse direction by cross-flow. The first term of the left-hand side represents the transverse momentum change coming from the axial direction and the second term is the transverse momentum coming from adjacent channels. The first term of right-hand side is the pressure force between adjacent charuiels, the second is the frictional loss by cross-flow and the last is the gravitational force. 

---

The worst operating condition in a common design practice consists of overly conservative assumptions on the hot-channel input. These assumptions must be realistically evaluated in a subchannel analysis by the help of in-core instrumentation measurements. In the early subchannel analysis codes, the core inlet flow conditions and the axial power distribution were preselected off-line, and the most conservative values were used as inputs to the code calculations. In more recent, improved codes, the operating margin is calculated on-line, and the hot-channel power distributions are calculated by using ex-core neutron detector signals for core control. Thus the state parameters (e.g., core power, core inlet temper-... 

Total calculated subassembly flow rate, as well as design data on core neutronics are input to subchannel analysis codes that predict coolant flow and temperature distribution in the subchannels of the core subassemblies. The peripheral subchannel temperatures and flow rates used for duct temperature prediction, as well as peak subchannel coolant temperature used for prediction of hot spot temperature of the hottest fuel elements are of particular interest. 

Subchannel analysis code for the calculation of coolant local thermo hydraulic conditions based on the rod bundles data obtained from the CHF data bank (subchannel CHF data bank)... 

Computer code CALPER — a thermal hydraulic subchannel analysis code for the assessment of coolant local conditions in the fuel assemblies and in the core of PWRAVWER-type nuclear reactors... 

Subchannel analysis codes, ASFRE for single-phase flow and SABENA for two-phase flow, have been developed for the purpose of predicting fuel element temperature and thermalhydraulic characteristics in the FBR fuel assemblies. ASFRE has the detailed wire-spacer model called distributed flow resistance model, which calculates the effect of wire-spacer on thermalhydraulics. Also planer and porous blockage models are implemented for fuel assembly accident analysis. In this reporting period, three dimensional thermal conduction model was used for the evaluation of local blockage in a fuel assembly. In addition, the comparison of pressure losses in the assembly with the water experimental data has been performed. Regarding SABENA, based on the two-fluid model, no activity is reported. 

The cladding temperature that was obtained by the three-dimensional coupled core calculation is the average temperature over the assembly. The peak cladding temperature of a fuel rod is necessary for the evaluation of the fuel cladding integrity. The subchannel analysis code of the Super LWR is coupled with the fuel assembly bum-up calculation code for this purpose [25]. Fuel pin-wise power distributions are produced for various bum-ups, coolant densities, and control rod positions. The pin-wise power distributions are combined with the homogenized fuel assembly power distribution to reconstmct the pin-wise power distribution of the core fuel assembly. The power distribution over the fuel assembly is taken into account as shown in Fig. 1.11. The reconstracted pin-wise power distribution is used in the evaluation of peak cladding temperature with the subchaimel analysis. 

The change of cross flow within a subassembly may occur during transients. The MCST may change from the result of the single-channel calculation. A transient subchannel analysis code was developed and the safety analysis of a Super LWR was carried out [70]. The temperature rises from the steady state value are about 20°C at the abnormal transients and about 130°C at accidents. The maximum values still stay below the MCST criteria for transients and accidents. The development and application of the transient subchannel analysis code are sununarized in Sect. 6.8. 

The plant dynamics code for the analysis of plant control and startup thermal considerations are described in ref. [115]. The subchannel analysis code and the analysis are found in refs. [116, 117]. Thermal-hydraulic and coupled stability calculations at supercritical and at subcritical pressure as well as startup considerations are described in ref. [118]. 

An improved core design procedure of the Super LWR that coupled the subchannel analysis with three-dimensional coupled core calculations is described in ref. [28]. The time-dependent subchannel analysis code for safety analysis of the Super LWR is described in ref. [123]. 

K. Yoshimura, Y. Ishiwatari, et al., Development of Transient Subchannel Analysis Code of Super LWR and Application to Flow Decreasing Events, Proc. NURETH-13, Kanazawa, Japan, Septemberl7-October 2, 2009, N13P1434 (2009)... 

K. Yoshimura, Development of a Transient Subchannel analysis code and safety analysis for the Super LWR, Graduate thesis, the University of Tokyo (2009) (in Japanese)... 

Fig. 2.78 Flow diagram of subchannel analysis code. (Taken from [24])...

Statistical thermal design uncertainty was evaluated by a Monte Carlo sampling technique combined with the subchannel analysis code. The engineering imcer-tainty of the Super LWR is evaluated as 31.88°C based oti the Monte Carlo Statistical Thermal Design Procedure. 

T. Tanabe, S. Koshizuka and Y. Oka, A Subchannel Analysis Code for Supercritical-Pressure LWR with Downward-Flowing Water Rods, Proc. ICAPP 04, Pittsbiugh, PA, Jime 13-17, 2004, Paper 4333 (2004)... 

Development of a Transient Subchannel Analysis Code and Application... 

The geometry and mesh arrangement in the fluid region are exactly the same as those of the steady-state subchannel analysis code. Figure 6.60 shows the entire algorithm. The momentum conservation equations for three directions and a mass conservation equation are solved with the Simplified Marker And Cell (SMAC) method [32]. In the SMAC method, a temporary velocity field is calculated, the Poison equation is solved, and then the velocity and pressure fields are calculated as shown in Fig. 6.61. The Successive Over-Relaxation (SOR) method is used to solve a matrix. 

For the verification of this code, three typical steady-states are calculated and compared to the results by the steady-state subchannel analysis code. Table 6.24 summarizes the three steady-state cases. Figure 6.62 shows the pin power distributions in those cases. The steady-state cmiditions are obtained using the transient... 

Open circle-. Used in the subchannel analysis code... 

Since the transient subchannel analysis code does not have the functions prepared in typical system analysis codes, several parameters are taken from the calculation results by the single channel safety analyses performed in Sect. 6.7. These parameters are the flow rate, temperature and pressure at the inlet of the hot fuel assembly, and the relative power. The radial and axial power distributions are assumed not to change with time. This is reasonable because the reactivity is not locally changed at the flow decreasing events. 

In order to investigate the influence of cross flow in the fuel assemblies on the safety margin, a transient subchannel analysis code for the Super LWR is prepared. It is found that the safety margin decreases at the flow decreasing events by considering the cross flow but the Super LWR still has a considerable safety margin. 

---