Thermal-hydraulic stability

March-Leuba (1990) presented radial nodalization effects on the stability calculations. March-Leuba and Blakeman (1991) reported on out-of-phase power instabilities in BWRs. BWR stability analyses were reported by Anegawa et al. (1990) and by Haga et al. (1990). The experience and safety significance of BWR core-thermal-hydraulic stability was presented by Pfefferlen et al. (1990). 

Pfefferlen, H., R. Raush, and G. Watford, 1990, BWR Core Thermal-Hydraulic Stability Experience and Safety Significance, Proc. Int. Workshop on BWR Instability. Hottsville, NY, CSNI Rep. 178, OECD-NEA, Paris. (6)... 

Intrinsic plant safety characteristics, such as nuclear and thermal-hydraulic stability and thermal inertia of the cooling system. 

Thermal-hydraulic stability The negative void effect is an important contributor to reactor thermal-hydraulic stability. 

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. 

Following the occurrence of an unstable phenomenon in La Salle Nuclear Power Station Unit 2 in 1988, the Ministry of International Trade and Industry of Japan re-examined nuclear thermal hydraulic stability at the Safety Evaluation Sub-Group of the Technical Advisory Group. As a result, the following measures have been identified ... 

The CCR is a direct cycle reactor, with its primary coolant system being also the main heat ransport system (MHT), The function of the MHT is to remove nuclear heat from the reactor core through natural circulation in both operating and shut down conditions. The design objectives are to achieve a Minimum Critical Heat Flux Ratio (MCHFR) of at least 1.5 at 120% full power and to ensure thermal-hydraulic stability during all operating conditions. 

Chatoorgoon, V., 1986. A simple thermal hydraulic stability code. Nuclear Engineering and Design 93, 51—67. 

Thermal-hydraulic stability Decay ratio 0.5 (damping ratio S 0.11) Decay ratio <1.0 (damping ratio > 0)... 

The decay ratios of the thermal-hydraulic stability of the hottest channel and the average channel are obtained as shown in Fig. 1.26. 

The relation between the decay ratios and orifice pressure drop coefficients is shown in Fig. 1.27. The reactor becomes more stable when the orifice pressure drop coefficient increases as is also known for BWRs. It can be seen that the thermal-hydraulic stability criterion is satisfied in the Super LWR at full power normal operation for the average power channel. The maximum power channel can be stabilized by applying a proper orifice pressure drop coefficient. The minimum orifice pressure drop coefficient required for thermal-hydraulic stability at full power operation is found to be 6.18 (a pressure drop of 0.0054 MPa). The total core pressure drop at 100% maximum power operation is 0.133 MPa. The required orifice pressure drop is small compared with the total core pressure drop. 

The block diagram used for coupled neutronic and thermal-hydraulic stability of the Super LWR is shown in Fig. 1.28. The neutronic model is used to find the forward transfer function G(i) and the thermal-hydraulic heat transfer and ex-core models are used to determine the backward transfer function H(s). The fi-equency... 

Fig. 1.27 Orifice pressure drop coefficient versus decay ratio of thermal-hydraulic stability at full power operation. (Taken from ref. [49] and used with permission from Atomic Energy Society of Japan)...

Fig. 1.29 Gain response of closed loop transfer function of coupled neutronic thermal-hydraulic stability...

Figure 1.31 shows the decay ratio contour map for coupled neutronic and thermal-hydraulic stability of the Super LWR. The decay ratio contour line of DR =1.0 indicates the stability boundary on the power versus flow rate map of the Super LWR. At the high power low flow rate region, the reactor becomes unstable. At low power operation and during startup, it is necessary to take care to satisfy the stability criteria. At the low power low flow rate region, the unstable conditions should be avoided by carefully adjusting the flow rate. 

In spite of the low flow rate and large coolant density change, the thermal-hydraulic stability of the Super LWR can be maintained by a sufficient orifice pressure drop coefficient. 

The presence of water rods reduces the density reactivity feedback effect due to the large time delay in the heat transfer to the water rods, and this affects the coupled neutronic and thermal-hydraulic stability. 

In summary, at the subcritical pressure operation during the pressurization phase, thermal criteria are more limiting due to dryout. The startup scheme prior to line switching is mainly determined by thermal criteria. The thermal-hydraulic stability criterion is satisfied by applying a sufficient orifice pressure drop coefficient. The coupled neutronic and thermal-hydraulic stability is also satisfied, since the power to flow rate ratio is low during this phase. 

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

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)... 

For the pressure vessel of the Super LWR, a similar design to that of PWRs is expected to be possible with the power scale similar to that of current LWRs [12, 13]. From the viewpoint of neutron economy, the core height to the equivalent diameter ratio of around 1.0 is desirable. However, from the viewpoint of thermal-hydraulic stability, a greater ratio is favorable. From these arguments, the core active height is determined to be 4.2 m. 

The orifices in BWRs are mainly used for improving the core thermal-hydraulic stabilities. Generally, the BWR channel stability improves when the pressure drops and inertia in the single-phase flow region are increased. This is why inlet orifices are used in BWRs. For LMFBRs, inlet orifices are used to control the coolant flow rate to the fuel assemblies to effectively cool the fuel. 

The same operational goal as used in BWRs is applied here. Thermal-hydraulic stability and coupled neutronic thermal-hydraulic stability of the Super LWR are described in Chap. 5. 

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. 

Thermal-Hydraulic Stability Considerations 5.4.1 Mechanism of Thermal-Hydraulic Instability... 

The axial core power is assumed to follow a cosine distribution for simplicity. The variation of the axial power distribution with fuel bumup is not considered here. It should be noted that the actual distribution of the axial core power may be top-peak, bottom-peak or chopped cosine, depending on the fuel bumup during the cycle. Also, the effect of the axial power distribution on thermal-hydraulic stability is not taken into account here. 

Thermal-Hydraulic Stability Considerations Mass conservation ... 

The following stability criteria of decay ratio for thermal-hydraulic stability are imposed on the Super LWR to guarantee the safety and stability of the reactor. These criteria are taken from those of BWRs. 

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... 

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