Loss of Feedwater

Total loss of feedwater flow (all loops) 39. Automatic trip - no transient condition... 

Loss of feedwater beater 22. Loss of all reedwater flow... 

Caution should be observed in copying another analysis. No previous analysis would have anticipated the TMI-2 accident. Other near misses were the Davis-Besse loss of feedwater incident of 1985, the Salem breaker problem of 1983, the Brown s Ferry Fire of 1975 and the scram problem at Brown s Ferry. 

NO flow of feedwater. NO level in drum. NO fuel flow. LICV closed. Loss of feedwater pressure. Massive leak. Fuel line closed. Loss of level in drum and explosion of drum by flame impingement on dry shell if flame continues. Same as above. Flame extinguished. No steam generation. 

REVERSE flow in feedwater line. Loss of feedwater pressure. Steam in feedwater system. 

U-12 Loss of feedwater flow to one steam generator loop 3 per loop... 

Such initiating events include loss of feedwater, excessive feedwater flow and reduction of feedwater temperature. 

Total loss of feedwater flow may be caused by de-energization of the feedwater pump motors and by failure of the oil cooling of these pumps, as well as by inadvertent closing of isolating and/or control valves on the feedwater lines. The transient is characterized by a pressure drop in the coolant circuit and by a reduction of the MCP cavitation margin. 

The economizer section is designed in full consideration of operating transients, startup and standby operation, and accident conditions such as loss of feedwater flow and feedwater line break. The structural design of the various parts is adequate to withstand the thermal and pressure loadings from these various conditions, consistent with the appropriate load classifications and design rules in the ASME Code, Section III, see Appendix G. 

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. 

Total loss of feedwater - Active or passiv e scram, passive Decay Heat Removal by emergency condenser (L) ... 

REEXAMINE PRA ESTIMATE OF CORE DAMAGE RISK FROM LOSS OF FEEDWATER... 

MEASURES TO MITIGATE SMALL BREAK IN LOSS-OF-COOLANT ACCIDENTS AND LOSS-OF-FEEDWATER ACCIDENTS—IE BULLETINS... 

Generic Safety Issue (GSI) 122.2 in NUREG-0933 (Reference 1) addresses the Loss Of All Feedwater Event with respect to the provision of enhanced operator training and improved instrumentation to aid the operator in determining that the plant has experienced a total loss of feedwater. 

During routine operation at the Davis-Besse nuclear power generating station, a loss of all feedwater event occurred. Subsequent to the loss of feedwater, the operators delayed initiating f eed-and-bleed to cool the core on the presumption that auxiliary feedwater flow was imminent. 

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. 

The System 80+ Standard Design also incorporates a dedicated safety-related Emergency Feedwater System, as described in CESSAR-DC, Section 10.4.9. This system is not required for normal operation but significantly reduces the probability of a Loss-of-Feedwater event occurring. 

Since (1) the Nuplex 80+ Advanced Control Complex incorporates adequate and reliable instrumentation for the rapid detection of a Loss-of-Feedwater event by the plant operators and for monitoring the subsequent actions to achieve a safe shutdown, and (2), emergency procedure guidelines are provided for this event, this safety issue is resolved for the System 80+ Standard Design. 

However, after installation of the SPDS at operating plants, the Davis-Besse plant Loss-Of-Feedwater event and other operating plant SPDS availability surveys raised concerns regarding SPDS reliability and availability and its impact on plant safety. 

Generic Safety Issue (GSI) A-09 in NUREG-0933 (Reference 1), addresses the issue of assuring that the reactor can attain safe shutdown after incurring an anticipated transient with a failure of the Reactor Trip System (RTS). An ATWS is an expected operational transient (such as a loss of feedwater, loss of condenser vacuum, or loss of offsite power to the reactor) which is accompanied by a failure of the RTS to shut down the reactor. 

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. 

In general, these criteria provide that plant design and operation adequately address both small-break LOCAs and loss-of-feedwater events in accordance with the guidance given in NUREG-0737. 

The plant emergency cooldown system is designed to remove residual heat to the ultimate sink (air) in case of initiating events such as NPP blackout or loss of feedwater supply to the steam generators. The emergency cooldown (aftercooling) system is a safety system and consists of passive elements, which do not require intervention of the operating personnel or actuation of automatic machinery to put them into operation. 

Loss of heat sink (LOHS) accidents- due to an accident in the energy conversion system, such as a turbine trip, loss of feedwater, or a steam line break. 

The damage of the lateral support would affect the vibrational stability and the ability to sustain earthquake and LOCA loadings. The complete fall of the wrapper barrel could lead to the loss of feedwater, damage to the largest radius tube U-bends, loose parts, and tube rupture. 

H2 Procedures used in case of loss of feedwater ARE-ASG (FWFCS (feed water flow control system) -AFWS (Auxiliary feedwater system). Residual heat is extracted using pressurizer relief valves and start up of safety injection and containment spray systems (feed and bleed process). 

On June 9, 1985, Davis-Besse NPP had a partial loss of feedwater while operating at 90% power. Following a reactor trip, the loss of all feedwater occurred. Operating experience and studies performed by NRC staff and the nuclear industry indicated that AFW systems failed at a relatively high rate, for instance ... 

The initiating events and sequences for the total loss of feedwater are plant-specific, and thus, individual plant assessment is needed. 

Total loss of feedwater which already happened in an operating plant implies the loss of heat removal safety function, if no other measures is available. The event significantly contributes to the core-melt frequency. 

Most Westinghouse PWRs adopted the primary side feed and bleed cooling as a backup to the total loss of feedwater event. 

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. 

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