Thermal-Hydraulic Stability Analyses

The fuel channel thermal-hydraulics model and the chaimel inlet orifice model are used in the thermal-hydraulic stability analyses. The axial power distribution is taken as a cosine distribution. The power generation in the fuel is assumed to be constant and only flow feedback is considered. The block diagram for thermal-hydraulic stability is shown in Fig. 5.27 [10, 11]. The forward transfer function is evaluated from the chaimel inlet orifice model. The feedback transfer function is 

Transfer function from pressure difference to iniet fiow veiocity 

Energy conservation Mass conservation Momentum conservation 

Herein, both the maximum power channel and the average power channel are analyzed to investigate the thermal-hydraulic stability. For the maximum power channel, the maximum linear heat generation rate in the axial direction is 39 kW/m and the core outlet temperature is 575°C. For the average power channel, the maximum linear heat generation rate in the axial direction is 24.6 kW/m and the core outlet temperature is 500°C. The calculation conditions used for the thermal-hydraulic analyses of maximum power channel and average power channel are shown in Table 5.8 [11]. 

The present stability code is checked by the characteristics of the frequency response of the transfer functions at the steady state by using Bode plots. For instance, the frequency response of the inlet velocity to outlet pressure transfer function shows that at low frequency the phase lag of the outlet pressure to that of 

---

Coupled neutronics/thermal hydraulics stability analyses of the STAR reactor at these plant equilibrium states at full and partial load will be required. Such analyses have been conducted already for the STAR-LM which shares the neutronics and thermal hydraulics properties of STAR-H2 reactor, - and stability has been demonstrated. 

Moreover, coupled neutronics/thermal-hydraulics stability analyses would be required for the ending equilibrium states from the passive accommodation of ATWS initiators. Work for the STAR-LM suggests that these states are indeed stable ones. 

Table 5.8 Calculation conditions for thermal-hydraulic stability analyses of the Super LWR (taken from ref. [11] and used with permission from Atomic Energy Society of Japan)...

Like in the thermal-hydraulic stability analyses, the frequency domain analysis method is used here, too. The mathematical model contains six submodels - the neutron kinetics model, the fuel rod heat transfer model, water rod heat transfer model, fuel channel thermal-hydraulic model, water rod thermal-hydraulic model, and the excore circulation system model. The fuel channel thermal-hydraulic model and water rod thermal-hydraulic model are the same as the thermal-hydraulic stability analysis model described in Sect. 5.4.3. 

Coupled Neutronic Thermal-Hydraulic Stability Analyses... 

Chapters 3-5 treat the plant system and behaviors. They include system components and configuration, plant heat balance, the methods of plant control system design, plant dynamics, plant startup schemes, methods of stability analysis, thermal-hydraulic analyses, and coupled neutronic and thermal-hydraulic stability analyses. 

Fig. 1.32 Coupled neutronic thermal-hydraulic stability analysis result at power increase phase...

In the Super LWR analyzed in Chap. 5, the coolant channels of all the fuel assemblies are cooled by upward flow, so that only one hot channel is treated in the thermal analysis and thermal hydraulic stability analysis. On the other hand, the coolant flow scheme in the reactor pressure vessel of the Super FR is the so-called two-pass scheme where part of the seed fuel assemblies and all the blanket fuel assemblies are cooled by downward flow as shown in Fig. 7.35. The fractions of the downward flow rate in the seed assemblies, blanket assemblies, and downcomer at the normal operating condition are determined in the core design as shown in Figs. 7.36 and 7.58 [26]. The flow distribution among those downward flow paths would change during the power raising phase so as to balance the pressure drops. 

Because it is based on well-established LWR technology and implements benchmarked thermal hydraulic safety analysis codes, and because each module is a relatively small reactor, it is possible that all of the information required for design certification may be obtained using a full-scale demonstration of a single module. Of significant interest for design certification would be the impact of neutronic feedback on flow stability, particularly during plant transients. 

T. T. Yi, S. Koshizuka and Y. Oka, A Linear Stability Analysis of Supercritical Water Reactors, (I) Thermal-Hydraulic Stability, Journal of Nuclear Science and Technology, Vol. 41(12), 1166-1175 (2004)... 

Although the coolant flow in the Super LWR is single-phase, the coolant enthalpy and therefore the density change substantially in the core because the coolant flow rate per thermal power in the Super LWR core is less than one eighth of LWR cores. Thus, the Super LWR can be susceptible to flow oscillations as the BWRs are. In Sect. 5.4, thermal hydraulic stability of the Super LWR is analyzed with the frequency domain approach. The analysis includes both supercritical and subcritical pressure conditions. 

The Super LWR employs separate large square water rods as neutron moderators. The time delay of the heat transfer to the water rod is much larger than that of the heat transfer to the coolant. Thus, the reactor system becomes less stable when a water rod model is included than when no water rod model is used. The descending water rods will have a significant effect on the coupled neutronic thermal-hydraulic stability because of the moderator density reactivity feedback from the large square water rods, and it needs to be considered in stability analysis of the Super LWR. 

The modified sliding pressure staitup as proposed by Yi et al. (2005) can be adapted to the proposed operating conditions in the Canadian SCWR concept. To provide a starting point for future analysis of critical performance characteristics (eg, fuel cladding temperatures and thermal-hydraulic and neutron stabilities), reference operating conditions (eg, flow rates, reactor power levels, and mechanical equipment configurations) have been selected. 

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]. 

---