Reactor feedwater control system

The reactor feedwater control system automatically controls the flow of reactor feedwater into the reactor vessel to maintain the water in the vessel within predetermined levels during all modes of plant operation. The control system utilizes signals from reactor vessel water level, steam flow, and feedwater flow. 

The reactor feedwater control system provides the signal for the reduction of reactor water recirculation flow to accommodate reduced feedwater flow caused by failure of a single feedwater pump. 

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Since the digital technology is considerably different than analog technology, the criteria appropriate for the safety review of digital computer based system are different. Such systems are being considered for use in the reactor protection, ECCS actuation, feedwater control systems, etc. 

In the other types of abnormalities, the event classification follows those of LWRs because the components such as the valves and the control rod drives are expected to be similar to those of PWRs or BWRs. In the category of the reactivity abnormality, the incidents related to the control rods are taken from those of PWRs. The loss of feedwater heating is taken like BWRs. Most of the incidents of the pressure abnormality are taken from BWRs because the Super LWR also adopts the direct steam cycle. The reactor depressurization is taken from PWRs. The abnormalities categorized into the inadvertent start or malfunction of core cooling system are taken from those of PWRs or BWRs. The inadvertent startup of AFS of the Super LWR corresponds to the inadvertent startup of ECCS of PWRs. The core coolant flow control system failure is the same as the feedwater control system failure for the Super LWR while the two incidents are different in BWRs due to the recirculation system. All the accidents categorized into the loss... 

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. 

The concentrations of the corrosion products in the reactor water are controlled by a number of parameters, including feedwater input, particle deposition and resuspension, precipitation and dissolution, and quality of performance of the reactor water cleanup system. As a consequence, these concentrations vary considerably from plant to plant and also within a plant, frequently without showing a discernible trend. The concentrations of total iron may vary by more than 3 orders of magnitude, primarily as a result of variations in the concentration of insoluble iron species, while that of dissolved iron is relatively constant and typically below 10 ppb. The concentration range for total cobalt in different plants is considerably smaller, and is typically in the range 10 to 200 ppt. In general, even when extreme fluctuations are ignored, there is no consistent trend in the concentrations of the corrosion products in the reactor water. 

A mismatch between reactivity added by the reflectors and the reactivity lost via fuel bum-up is adjusted by the feedwater control of the water/steam system. Therefore, the reactivity control is unnecessary at a reactor side and this is an important factor to simplify the reactor operation. 

BWRs use reactor level instrumentation to perform a number of functions including control functions, such as feedwater control, and protective functions, such as automatic scram and autostart of emergency core cooling system. 

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). 

High-Pressure Systems High-Pressure cooiant injection Reactor core-isoiation cooling Reactor feedwater pumps Control-rod drive Standby liquid 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. 

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