Power reactor feedwater control system

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. 

Excessive feedwater flow may result from improper operation of the feedwater controlling device. The fast controlled power reduction and FPR systems protect the reactor against overfilling of the coolant circuit, acting on the corresponding set points for water levels in the drum separators. 

LEADIR-PS 200 has a graceful and safe response to all anticipated transients. For example, an overcooling event (as could be caused by loss of feedwater control or spurious opening of steam relief valves in combination with control system failure) causes the core inlet temperature (normally 350°C) to fall as the freezing point of 327°C is approached the coolant viscosity increases, coolant flow decreases, and in the absence of any control system action, the negative temperature coefficients of the fuel and moderator reduce reactor power. Heat removal is maintained by natural convection. 

Reactor pressure increase Several events may cause this e.g., inadvertent closure of one turbine control valve, pressure regulator downscale failure, generator load rejection, turbine trip MSIV closure, loss of condenser vacuum, loss of nonemergency AC power to station auxiliaries, loss of feedwater etc. All these have been analysed. Features are included in the instrumentation and control systems or redundancies to maintain reactor pressure through a combination of component automatic responses or operator actions, depending on the identified cause. 

Compared with current commercial LWR designs a number of safety-grade systems have been eliminated the control rods and the safety injection boron system are replaced by the density locks, the automatic depressurization system is not required, the auxiliary feedwater supply system for RHR is replaced by the reactor pool, the containment heat removal and containment spray systems are replaced by the passive cooling of the reactor pool. The safety-grade closed cooling water stem, HVAC sterns, and a.c. power supply systems have been replaced by non-safety-grade systems, allowing major simplification of the plant. 

Feedwater system malfunctions causing an increase in feedwater flow (two cases were modelled the accidental opening of one feedwater control valve with the reactor just critical at zero load conditions and the accidental opening of one feedwater control valve with the reactor in automatic control at full power). This fault models the failure of one protection division as the limiting single failure. This is fault 4.2.2 in the fault schedule. 

The capability to accept a turbine trip from full-power operation without reactor trip. This capability is provided with the normally available systems (sueh as steam dump and feedwater control). 

The plant s integrated control system attempted automatically to reduce reactor/turbine power in accordance with the reduced feedwater flow. The control rods were being inserted into the core and reactor power had been reduced to about 80%. At the same time the primary-side reactor operator held open the pressurizer spray valve in an attempt to keep the reactor coolant pressure below the high pressure reactor trip set point of 2300 psig (normal pressure is 2150 psig). However, the reduction of feedwater and subsequent degradation of heat removal from the... 

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... 

The covering of all major abnormal transients by these proposed models are confirmed by comparing the results obtained by them with results obtained from detailed fuel rod analyses modeling each abnormal transient event. The following eight abnormal transient events are analyzed for confirmation inadvertent startup of the auxiliary feedwater system (AFS) loss of feedwater heating loss of load without turbine bypass withdrawal of control rods at normal operation main coolant flow control system failure pressure control system failure partial loss of reactor coolant flow and loss of offsite power. 

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. 

The improved control system with the feedwater controller described as (7.28) is designated Control system (A). The gain A p2a is tuned by analyzing the plant dynamics against the 10% stepwise decrease in the setpoint of the reactor power. The results are shown in Fig. 7.73 [31]. The changes in the power to flow rate ratio and the main steam temperature can be decreased from the case with the reference control system. 0.4 is chosen as Kp2 so that these changes are minimized. 

The same event is analyzed in Sects. 7.9.3 and 7.9.4. The results are shown in Fig. 7.76 [31]. The reactor power settles to the new setpoint at around 100 s with all the control systems. Since the feedwater flow rate follows the reactor power more closely with the improved control systems than the reference case, the changes in the main steam temperature are kept smaller. 

Since the reactor power temporarily decreases due to the coolant density feedback in this event the change in the feedwater flow rate is slowed down by the second term of (7.29), so that the main steam temperature reaches the new setpoint without overshoot with Control system (B). On the other hand, the change in the feedwater flow rate is accelerated by the second term of (7.30), so that the settling time is shorter with Craitrol system (C) than that with the reference control system. 

With Control system (C), the feedwater flow rate decreases further at the beginning and hence the change in the main steam temperature is larger than the reference case. This is because the derivative control term of (7.30) decreases the feedwater flow rate by detecting the initial decrease in the reactor power. 

Control system (B), where the reactor power is fed back to the feedwater controller, keeps the main steam temperature more stable than the reference control system when the power level is changed by the CRs as in Fig. 7.76 [31]. However, this control system gives similar plant dynamics to those with the reference control system unless the reactor power is significantly changed by the CRs. This is because the reactor power is less sensitive to the flow rate and mainly influenced by CRs. 

Increase in reactor heat removal inadvertent opening of steam relief valves secondary pressure control malfunctions leading to an increase in steam flow rate feedwater system malfunctions leading to an increase in the heat removal rate. —Decrease in reactor heat removal feedwater pump trips reduction in the steam flow rate for various reasons (control malfunctions, main steam valve closure, turbine trip, loss of external load, loss of power, loss of condenser vacuum). 

The fundamental concept of the 4S is that of continuous monitoring rather than active operation . The reactor operates using a system of pre-programmed movable reflectors and the power control is executed from the outside, through feedwater flow rate changes in the power circuit. The plant and component conditions and/or unauthorized access could be continuously monitored from outside the site, e.g. by satellite systems. 

The control rods are carefully withdrawn to start nuclear heating and raise the power. The power rise is kept slow in order to satisfy the limit of temperature rise (55°C/h) as is also applied in LWRs. At the same time, the motor-driven feedwater pumps are started, and the water circulates in the feedwater loop, which is provided to warm the feedwater system during startup. This procedure is necessary to avoid cold shock on the reactor vessel inlet nozzles when the recirculation mode is switched to the once-through mode. 

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