Core coolant flow rates

In a PIUS plant, the core coolant flow rate is determined by the thermal conditions at the reactor core outlet - relative to the reactor pool. The resulting pressure drop across the core and up through the riser must correspond to the static pressure difference between the interface levels in the upper and lower density locks. The main coolant pumps are operated to establish a pressure balance across the lower density lock to keep the reactor system in operation. In case of a severe transient or an acddent, the natural circulation flow loop will be established, providing both reactor shutdown and continued core cooling. 

An example of fuel assembly design of the Super LWR is shown in Fig. 1.13 [35]. An example Super LWR core and fuel characteristics are given in Table 1.4 [24]. The core coolant flow rate of the Super LWR is substantially lower than that of LWRs due to the high enthalpy rise in the core. The gap between fuel... 

Abnormal transients Decrease in core coolant flow rate Partial loss of reactor coolant flow Loss of offsite power Abnormality in reactor pressure Loss of turbine load Isolation of main steam line Pressure control system failure Abnormality in reactivity Loss of feedwater heating Inadvertent startup of AFS Reactor coolant flow control system failure Uncontrolled CR withdrawal at normal operation Uncontrolled CR withdrawal at startup Accidents... 

Decrease in core coolant flow rate Total loss of reactor coolant flow Reactor coolant pump seizure Abnormality in reactivity CR ejection at full power CR ejection at hot standby LOCA... 

The reactor coolant flow abnormality is important for the Super LWR because maintaining the core coolant flow rate is the fundamental safety requirement It should be noted that there are two types of reactor coolant flow abnormalities with and without reactor scram before events the former are abnormal transient types... 

Core coolant flow rate The nominal design value of the core coolant flow rate is 1,420 kg/s, and the typical measurement error of 2% of the nominal value is used. The standard deviation is 1 % for the normal distribution and 1.15% for the uniform distribution. 

The feedwater flow rate decreases stepwise to 95% of the initial value. The control rod position and the turbine control valve opening are kept constant. The results are shown in Figs. 4.11 and 4.12. Due to the once-through coolant cycle, a decrease in the feedwater flow rate directly leads to a decrease in the core coolant flow rate. The main steam temperature increases. The core and main steam pressures... 

Figure 4.15 shows that the main steam pressure is sensitive to the turbine control valve opening. Thus, the main steam pressure is regulated by the turbine control valves like in BWRs. The reactor power is controlled by the control rods. Even though Fig. 4.12 shows that the reactor power is also sensitive to the flow rate, control of the reactor power by regulating the core coolant flow rate as is done in BWRs is not suitable for the Super LWR because ... 

The reactor power of the Super LWR is not very sensitive to the core coolant flow rate as described in Sect. 4.3.2... 

Since the Super LWR does not use saturated steam, the main steam temperature changes with the power to flow rate ratio in the core. It needs to be kept constant in order to avoid too much thermal stress or thermal fatigue on the structures. Since the Super LWR has no superheaters that are utilized to control the main steam temperature as in FPPs, another method is needed. The analysis results described in Sect. 4.3.2 show that the main steam temperature is sensitive to the feedwater flow rate. Thus, the main steam temperature is controlled by regulating the feedwater flow rate. It is also suitable from the viewpoint of the safety principle of the Super LWR, i.e., keeping the core coolant flow rate (described in Sect. 6.2) because the feedwater flow rate indirectly follows the reactor power in this control method. The plant control system employed for the Super LWR is shown in Fig. 4.16. The plant control strategies of the Super LWR, PWRs, BWRs, and FPPs are compared in Table 4.3. 

Loss of feedwater flow is the same as loss of reactor coolant flow for the Super LWR. BWRs have the recirculation system and large coolant inventory in the reactor vessel. PWRs have the secondary system. Therefore, the feedwater of the Super LWR is more important for its safety than that of LWRs. In this chapter, feedwater flow, feedwater system, and feedwater pump of the Super LWR are called main coolant flow, main coolant system, and reactor coolant pump (RCP) , respectively, to distinguish them from those of LWRs. The main coolant flow rate is equal to the core coolant flow rate and the main steam flow rate at the steady-state. 

The abnormal events related to a decrease in core coolant flow rate are the most important for the Super LWR because the core coolant flow rate is the fundamental safety requirement, as described in Sect. 6.2. Since the coolant cycle of the Super LWR is different from those of both PWRs and BWRs, the events need to be carefully selected and classified. 

The once-through coolant cycle of the Super LWR is schematically illustrated in Fig. 6.9 [1]. The feedwater pump is the same as the RCP. The loss of all feedwater flow and the total loss of reactor coolant flow are the same incidents. Classification of this event depends on the frequency. Also, the guidelines for Japanese LWRs are followed. A simultaneous sudden trip of all ptunps that have been directly maintaining the core coolant flow rate must be classified as a total loss of reactor coolant flow accident. These pumps correspond to the primary pumps of PWRs and the recirculation pumps of BWRs. Since the RCPs of the Super LWR also maintain the core coolant flow rate, a simultaneous sudden trip of the RCPs is classified as the total loss of reactor coolant flow accident, assuming that its frequency will be less than 10 per year by system separation and high reliability. 

This is a typical pressure decreasing transient. The maximum turbine control valve opening is assumed and it is 130% of the rated value. The cladding temperature is always below the initial temperature because the main steam flow rate and therefore the core coolant flow rate increase. A scram signal is released when the pressure reaches the low level 1 (24.0 MPa). A depressurization signal is released when the pressure reaches the low level 2 (23.5 MPa). After opening the ADS, the reactor behavior is similar to that shown in Fig. 6.7 [1]. 

This is a typical flow increasing transient. The demand of the main coolant flow rate is assumed to rise stepwise up to 138% of the rated flow as is assumed in the feedwater control system failure of Japanese ABWRs. Since increase in the core coolant flow rate is mild in ABWRs due to the large recirculation flow, the feed-water flow rate is assumed to increase stepwise. This assumption is too conservative for the Super LWR. The main coolant flow rate is gradually increased by the control system in the safety analysis. The calculation results are shown in Fig. 6.31. The reactor power increases with the flow rate due to water density feedback. A scram signal is released when the reactor power reaches 120% of the rated power. The maximum power is 124% while the criterion is 182%. The increase in the pressure is small. The sensitivity analysis is summarized in Table 6.15. 

In contrast to the cold-leg break, a hot-leg break is less important for the Super LWR. This is because the core coolant flow rate naturally increases during blowdown (cf. Fig. 6.7 [1]), and because forced flooding by the LPCIs is expected after the blowdown. 

The calculation results of the total loss of reactor coolant flow at 15 MPa are shown in Fig. 6.56 as an example of the loss of flow events. Due to the coast-down and hence the decrease in the core coolant flow rate, departure-from-nucleate-boiling (DNB) occurs at 3 s and the cladding temperature quickly increases. However, the peak value is much lower than the criterion. Although the water source effect of the water rods is small at subcritical pressure, the core coolant flow can be kept by natural circulation in the recirculation loop before the start of the AFSs. 

The results are shown in Fig. 7.69 [31]. The main steam pressure increases. The core coolant flow rate decreases with the main steam flow rate, which increases the main steam temperature. The increase in the main steam temperature is nearly 20° C while that in the Super LWR is 12°C. 

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