Feedwater control system failure

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. 

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. 

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. 

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. 

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. 

The System 80+ Standard Design has a Main Feedwater Isolation System (see CESSAR-DC Section 10.4.7.2.2) to protect the SGs from overfill. The system includes redundant remotely operated isolation valves in each main feedwater line to each SG. The valve actuation system (see CESSAR-DC Section 7.3.1.1.10.3) is composed of redundant trains A and B, and each train s instrumentation and controls are physically and electrically separate from and independent of those of the other train. A failure of one train will not impair the action of the other. The main feedwater isolation valves are automatically actuated by a Main Steam Isolation Signal (MSIS) from the Engineered Safety Features Actuation System (ESFAS, see CESSAR-DC Section 7.3.1). High SG water level, in a 2-out-of-4 logic, is one of the initiators for the MSIS. The main feedwater isolation valves can be in-service tested in accordance with ASME Code Section XI, Subsection IWV. A Technical Specification (CESSAR-DC Chapter 16) will establish testing requirements for the valve actuation system. These requirements will also be incorporated into the plant maintenance procedures. 

Generic Safety Issue (GSI) II.E.1.2 in NUREG-0933 (Reference 1), addresses the TMI requirement for plants to install a control-grade system for automatic initiation of the auxiliary feedwater (AFW) system. This requirement can be achieved by meeting the criteria identified in IEEE Standard 279-1971 (Reference 2), (e.g., timely system initiation, single failure... 

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. 

Since the HTS boundary is preserved, the safety aspect is release of a portion of any radioactivity contained in the secondary side. Generally, the behaviour is bounded by steam and feedwater line failures. The secondary side controls are modelled in some detail to ensure that either their proper functioning, or lack of response, does not impair any safety system actions. Both normal and alternative modes of plant control are assessed. 

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