Loss of Main Feedwater

The AFWS is a non-safety grade system to supply feedwater to the secondary side of the SG to remove heat from the core via the SG under a loss of main feedwater and other transients. Two types of pumps, turbine driven and motor driven, are installed to achieve high reliability and diversity. 

In the Main Feedwater System, three of the pumps are steam turbine driven. The fourth pump is motor driven and is normally only used for startup and shutdown. This pump starts automatically on loss of main feedwater and reactor trip (see CESSAR-DC, Section 10.4.7.2.3.D). 

Guidelines are also established for hot functional tests to be performed by the owner-operator in accordance with BTP ASB 10-2. These tests are to verify that unacceptable feedwater system water hammer does not occur when (a) using normal plant operating procedures for normal and emergency restoration of SG water level following loss of main feedwater, and (b) transferring main feedwater during normal operation from the SG downcomer feedwater inlet nozzles to the economizer inlet nozzles. (See CESSAR-DC, Sections 14.2.12.1.63 and 14.2.12.4.13, respectively.)... 

After the Three Mile Island Unit 2 accident the NRC reviewed the auxiliary feedwater system for availability and reliability of components and decay heat removal capability. In particular, the EFW system was scrutinized with regard to the potential for failure under a variety of loss of main feedwater conditions. The safety concern was that a total loss of feedwater, i.e., loss of both main and emergency feedwater, could result in loss of core cooling. The NRC requested operating plants and plants under construction to review both the reliability and the capability of the EFW system to perform its intended safety function i.e., core decay heat removal. The evaluation by the plants was divided into three parts as discussed below. 

Part one consisted of a limited PRA to determine the potential for EFW system failure under various loss-of-main-feedwater transient conditions, with particular emphasis being placed on determining potential failures from human errors, common causes, single-point vulnerabilities, and test and maintenance outages. This evaluation applies to operating plants and plants under construction and not to advanced or future plants. Part two was composed of a deterministic review of the EFW system using the acceptance criteria of SRP Section 10.4.9 and the associated Branch Technical Position (BTP) ASB 10-1. Part three required a re-evaluation of the decay heat removal capability of the EFW system with respect to EFW system flowrate. Parts two and three apply to advanced or future plants. 

After the Three Mile Island Unit 2 accident, the NRC reviewed auxiliary feedwater system designs with respect to timely initiation, as described in 10 CFR 50, Appendix A, (GDC 20), (Reference 3). Upon completion of the review, the NRC determined that new guidance identified in NUREG-0737, (Reference 4) was necessary in order to assure a timely start of the AFW system after a design basis event (e.g., loss of main feedwater). Among this new guidance was automatic system initiation, environmental and seismic equipment qualification, and single failure criterion. 

Loss of main feedwater to one steam generator lEV-LMFWl 1.92E-01 4.53E-10 2.12E-10... 

The auxiliary feedwater system is designed to compensate for a loss of main feedwater and... 

The AFWS is a 2 division and 4 train system. The AFWS is designed to supply feedwater to the SGs for RCS heat removal in case of loss of main/startup feedwater systems. The reliability of the AFWS has been increased by use of two 100% motor-driven pumps, two 100% turbine-driven pumps and two independent safety-related auxiliary feedwater storage tanks as a water source instead of condensate storage tank. 

Loss of the main heat sink Loss of normal feedwater supply Glosure of the main steam isolation valves Stuck-open safety-relief valve... 

Station Very Small Small LOCA Steam Loss of Main Loss of Others Blaekout LOCA Generator Feedwater Offsite Power... 

During the past few years, most Westinghouse PWRs have developed procedures for using feed and bleed cooling and secondary system blowdown to cope with loss of all feedwater. These procedures have led to substantial reductions in the frequencies of transient core damage sequences involving the loss of main and auxiliary feedwater. Appropriate credit for these actions was given in these analyses. However, there are plant-specific features that will affect the success rate of such actions. For example, the loss of certain... 

A loss of electrical load transient could occur from a generator trip, a turbine trip, or a loss of main condenser vacuum. Generally, the most severe transient would be caused by the loss of condenser vacuum. The main feedwater pumps in many plants are steam turbine-driven and exhaust to the main condenser. Thus, loss of condenser vacuum also could cause a loss of the main feedwater pumps. In this case the sequence of events would be similar to the loss of feedwater transient. The most severe effect of the transient, the peak pressure in the primary system, would be of about the same magnitude as in the loss of feedwater flow transient. 

The output from the turbine might be superheated or partially condensed, as is the case in Fig. 6.32. If the exhaust steam is to be used for process heating, ideally it should be close to saturated conditions. If the exhaust steam is significantly superheated, it can be desuperheated by direct injection of boiler feedwater, which vaporizes and cools the steam. However, if saturated steam is fed to a steam main, with significant potential for heat losses from the main, then it is desirable to retain some superheat rather than desuperheat the steam to saturated conditions. If saturated steam is fed to the main, then heat losses will cause excessive condensation in the main, which is not desirable. On the other hand, if the exhaust steam from the turbine is partially condensed, the condensate is separated and the steam used for heating. 

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). 

Examples of plant parameters monitored that are needed to identify a Loss-of-Feedwater event are steam generator pressure and level (wide range) main and emergency feedwater flow and reactor coolant pressure, temperature and degree of subcooling. 

In contrast, there are disadvantages to automatic isolation of EFW. If both channels of the controlling isolation logic systems were to spontaneously actuate either during normal operation or in the course of a transient, the availability of EFW would be lost and the main steam isolation valves would close. Most newer plants use turbine-driven main feedwater pumps. Thus, main feedwater would also be lost, resulting in complete loss of the secondary heat sink. Capability to lock-out the isolation logic is necessary to preclude such scenarios. 

This group of characteristics specifies components of a standby feedwater system that is started if the main feedwater system is not available because of a failure. These systems should have an independent power source available after loss-of-ofifsite-power events. The auxiliary feedwater pumps are usually connected to the main feedwater tank and may also be used as normal start-up feedwater pumps at low reactor power (see above). The emergency feedwater system is usually designed for use in case of a catastrophic failure of the main feedwater system, e.g. if a common feedwater pipeline ruptures, making both main and auxiliary feedwater pumps unavailable. The emergency feedwater pumps and flowpaths should be different to those of the main and auxiliary feedwater. 

Heat removal from PWR plants following reactor trip and a loss of off-site power is accomplished by the operation of several systems, including the secondary system via the steam relief to the atmosphere. The auxiliary (emergency) feedwater system (AFW) functions as a safety system because it is the only source of makeup water to the steam generators for decay heat removal when the main feedwater systems becomes inoperable. 

International practice considers the analysis of ATWS for a variety of initiating events such as loss of feedwater, loss of load, turbine trip, loss of condenser vacuum, loss of off-site power, closure of main steamline isolation valves, uncontrolled boron dilution, inadvertent control rod withdrawal, etc. ATWS analyses are performed in general by using best-estimate tools to determine the preventive (e.g. a diverse scram system) or mitigative measures (e.g. initiation of turbine trip and emergency feedwater supply) which need to be implemented for strengthening plants defence in depth. 

The decay heat and residual heat could be cooled for about 30 minutes through the natural circulation of primary coolant in the primary system, and through the operation of turbine operation auxiliary water supply pump and the main steam safety valve. Necessary power for the safety protection systems and the turbine-driven auxiliary feedwater systems is supplied from highly reliable batteries to secure the safety of reactor even during the total loss of power. 

For those events where offsite ac power is lost, an appropriate time delay between turbine trip and the postulated loss of offsite ac power is assumed in the analyses. A time delay of 3 seconds is used. This time delay is based on the inherent stability of the offeite power grid (see subsection 8.2 of Reference 5.6). Following the time delay, the effect of the loss of offsite ac power on plant auxiliary equipment (such as reactor coolant pumps, main feedwater pumps, condenser, startup feedwater pumps, and RCCAs) is considered in the analyses. 

Babcock Wilcox claimed that, had there been auxiliary feedwater, the temperature of the reactor coolant might have remained relatively stable until the problem of the condensate pumps was corrected and normal feedwater was reinstated. This view has been contested not only by the NRC but also by the utility-sponsored Nuclear Safety Analysis Center, an investigative arm of the Electric Power Research Institute. Their investigations indicate that, except for adding another dimension to the areas of concern within the main control room, the early unavailability of auxiliary feedwater did not significantly affect the progression of the accident, which was dominated by the uncompensated loss of reactor coolant. 

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. 

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. 

Three units of the AFS are assumed to start. The AFS flow (12% of rated value, 30°C) is added stepwise to the main coolant flow at 0 s. The results are shown in Fig. 6.30. The main coolant flow rate and the fuel channel inlet flow rate increase due to the AFS startup. At the beginning, the fuel channel inlet flow rate is lower than the main coolant flow rate because the feedwater temperature, which is the same as the loss of feedwater heating transient described above, decreases. The... 

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