BACKGROUND OF THE INVENTION In a nuclear plant of the pressurized water reactor (PWR) type, coolant fluid, which is borated water, is continuously transferred through a closed circulation loop between a nuclear reactor and one or more steam generators.

During power production, the circulating pressurized coolant absorbs heat released by the thermonuclear reaction occurring in the reactor. The heated coolant flows through a main pipe which is appropriately known as the "hot leg"of the circulation loop. The hot leg delivers the hot cooiant to a steam generator.

In the steam generator, the coolant fluid circulates through a heat exchanger. The heat exchanger cools the primary coolant fluid and uses the heat removed from the coolant to produce steam in the secondary system.

This steam is eventually used to drive turbines and generate electricity.

After the circulating coolant is cooled by a heat exchanger, a circulation pump removes the coolant from the steam generator via a"suction leg"and retums it to the reactor via a"cold leg"and inlet. The coolant is then reheated in the reactor and the cycle repeats.

This circulation of coolant through one or more loops is critical for the operation of the power plant. Not only does it deliver heat energy to the

steam generators where the energy is used to produce steam for driving the turbines, but the circulating coolant also prevents the reactor core in the reactor from over-heating.

Nuclear power plant systems. inciuding the steam generators, require periodic maintenance. During reactor shutdown periods, the fluid circulation system must be inspected for potential degradation of the steam generator tubes and nozzle dams are normaily installe and removed from the steam generator hot and cold legs to allow inspection and maintenance to be performed in a dry environment.

In order to install and remove nozzle dams, the coolant fluid must be drained from the steam generator. This requires iowering the fluid level in the main circulation loop and consequently the hot leg or main pipe. During such a maintenance period. which is termed a"shutdown", the coolant continues to be heated by decay heat frorn the reactor core but it is cooled by an alternate heat exchanger and auxiliary circulatory system know as the"shutdown cooling system".

In order to lower the coolant or water level in the shutdown reactor system to allow maintenance operations on portions of the system above the lowered water ievel. the water levei must be controlled and maintained at a minimum level while the core flow rate must be maintained to provide adequate core cooling. This minimum water level is about midway within the reactor coolant system main loop piping (the hot leg) and is commonly referred to as"midloop".

During midloop operation, primary coolant water is circulated through the system to cool the core. Typically, the shutdown cooling line on each hot leg communicate with the lower region of the hot leg or of the main loop pipe to draw the heated water from the core for cooling by the alternate heat exchanger in the shutdown cooling system and subsequent recircuiation of cooled water to a reactor iniet and thus to the core.

It is possible to experience the formation of a coriolis effect vortex in the shutdown cooling line during midloop operation, if the water level is lowered too far down or if the shutdown cooling flow rate is too high. Such a vortex is an undesirable anomaly because it limits the rate at which coolant flow can be drained from the system and it can lead to air entrainment and cavitation in the shutdown cooling pump (s). These results cause concern for continued cooling of the core.

The current methods to avoid vortex formation attempt to keep the water level as high as possible and/or reduce the flow rate, resulting in a conflit between the need to lower the water level for maintenance service, and the need to keep the water level high and at a sufficient flow rate for safe core cooling.

Midloop liquid level sensor systems in use are related to a detection of the water elevation and inference of the status of any drain vortex therefrom.

A midloop ultrasonic measuring instrument for this purpose is disclosed in co- pending U. S. patent Application Serial No. 08/864,644, assigned to the same assignee as the instant application and filed May 29,1997. Another co- pending U. S. Patent Application Serial No. 08/783,978, assigned to the same assignee as the instant application and filed January 15,1997, is for a system which reduces air entrainment and pump cavitation by inserting a vortex breaker in the shutdown cooling pipe adjacent the main pipe or hot leg.

An instrument which provides direct detection of air being entrained by a vortex in the shutdown cooling system piping which could lead to a total loss of shutdown cooling and to alert the nuclear power plant operator of a dangerously low reactor coolant system water level condition so that corrective action can be taken in a timely manner is provided in co-pending U. S. Application Serial No. 08/927,137, assigned to the same assignee as the instant application and filed September 2.1997. All three of the above- referenced applications are included herein, by reference.

In nuclear power plants, much attention has been given to shutdown cooling system reliability, especially during reactor coolant system midloop water level operation. Midloop operation in a typical pressurized water reactor (PWR) nuclear steam supply system, for example, for the installation and removal of steam generator nozzle dams, can be a very difficult operational process. Typically, the water level allowed tolerance is approximately plus or minus one inch ( 1"). The plant operator must control this water level manually. Current instruments used measure only the average reactor cootant system (RCS) water ievel and the shutdown cooling pump current.

The RCS water level measurement accuracy is limited by the instrument technology used and the process parameter changes such as temperature, pressure and boric acid concentration. The shutdown cooling pump current measurement alarm occurs only after air has already been ingested into the pump, thus it cannot be used to avoid the air vortex.

A shutdown cooling system of a nuclear power plant includes a main pipe or hot leg lower region which is connected to a shutdown cooling pump for circulation of a portion of the hot leg flow through an auxiliary heat exchanger and back to an iniet of the reactor. Provision for by-passing the heat exchanger for draining water to create an appropriate midloop water level is also provided by another system (not shown). In any event, whichever drain method is used, drains water from the reactor coolant system to the desired level for maintenance in the steam generator and shutdown cooling system and recirculates coolant to the reactor core for core cooiing.

Pump cavitation caused by a coriolis effect vortex, air entrainment or other flow anomalies in the hot leg drain pipe robs efficiency by lowering the flow rate of the core recirculation and slowing down the process of achieving the desired level and recirculation rate. Also, air entrainment and pump cavitation threatens the pump's ability to adequately recirculate coolant in the

shutdown coolant system. Accordingly, efficient sustained performance of the pump is mandatory.

SUMMARY OF THE INVENTION The present invention is a safety feed system for a shutdown cooling system of a nuciear power plant to eliminate or minimize entrained air or cavitation in the shutdown cooling drain pump (s). This is accomplished by providing a reservoir with a make-up feed pipe intersecting and in communication with the drain pipe between the hot leg pipe or hot leg and the shutdown cooling pumps. The make-up feed pipe has a modulated control valve therein responsive to signals from a liquid level sensor in the hot leg and an air entrainment or pump cavitation sensor which is within, across or adjacent to the shutdown cooling or drain pump. Signals from the liquid level sensor and the air entrainment or pump cavitation sensor are typically based on pump current, pressure across the pump or flow rate through the pump and are processed in a controller of the electrical circuit which is connected to and controls and modulates the valve in the make-up feed pipe.

The novel safety feed system may utilize for the detection of air entrainment or air vortexing commercially available portable clamp-on ultrasonic flowmeters as disclosed in the above cited Application Serial No. 08/783,978. These instruments are mounted for the detection of air entrainment or air vortexing and provide a cavitation sensor when mounted across or adjacent the shutdown cooling pump either upstream or downstream thereof. In the event of air entrainment, an air vortex, air void or other flow anomalies, a disruption of the flow meter ultrasonic signal triggers an alarm and provides a signal to the controller of the valve in the make-up feed pipe. Pump current sensors within the pump may also be used to trigger a signal and indicate a flow anomaly to the controller which controls and modulates the make-up feed pipe valve.

By the use of the system of the invention which monitors both main pipe coolant level and air entrainment or cavitation of the pump and provides make-up feed in response thereto, potentially dangerous plant situations and any subsequent extensive reguiatory scrutiny and paper work are avoided.

BRIEF DESCRIPTION OF THE DRAWINGS Figure 1 is an isometric schematic drawing of a nuclear power plant having two steam generators with a shutdown cooling system and safety feed system of the invention illustrated in connection with onty one of the steam generators, for clarity.

DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT Figure 1 illustrates a nuciear power plant incorporating the safety feed system of the present invention. The numeral 10 generally designates a pressurized water reactor type of plant in which water is continuously transferred through a closed circulation loop between a reactor 12 and each of two steam generators 14 and 16, respectively.

The water coolant from reactor 12 flows through main pipes or hot legs 18 to the respective steam generators 14 and 16, each of which has similar piping.

In the case of steam generator 16, for illustration, a coolant system circulation pump 20 circulates water, which has been cooled in the steam generator, through a suction leg pipe 22, and back to the reactor 12 via cold leg 24 and reactor vessel inlet 26.

The shutdown cooling system or drain pipe 28, with isolation valves (s) 29 therein, intersects the lower region of the substantially horizontal main pipe or hot leg 18. Within the lower region of the hot leg or main pipe 18 is a shutdown cooling system drain 30.

The opening 30 of pipe 18 is variously termed drain 30 or shutdown cooling system drain 30 and the pipe from it is called drain pipe 28 or shutdown cooling system pipe 28.

Flow from the hot leg pipe 18 into the shutdown cooling system or drain pipe 28 may form a vortex at drain 30 which entrains air in the coolant and may create cavitation in shutdown cooling or drain pump 33 which has direct fluid communication from the hot leg pipe 18 where it intersects with the lower region of hot teg pipe 18. This vortex inhibits flow rate in the drain pipe 28 and drain pump 33 by creating voids and cavitation and can airlock the shutdown cooling pump 33. The shutdown cooling or drain pump 33 discharges through conduit 28'to shutdown cooling auxiliary heat exchanger 34 downstream from it for temperature control of the shutdown cooling system water.

A clamp on portable flowmeter 32 is mounted downstream (but may be upstream or across) of pump 33 to act as a cavitation sensor by sensing entrained air, cavitation or other flow volume anomalies. Transducers for the flowmeter 32 are attached to pipe 28 by chains or straps as taught in Application Serial No. 08/927,137.

The preferred clamp-on portable flowmeter is a Controlotron System 1010 DP which is available from Controlotron, 155 Plant Avenue, Hauppauge, NY 11788. It is specified in their brochure 1010DP-1, a copy of which is attached hereto and which is considered part of the disclosure hereof.

From heat exchanger 34 the water is directed by valve 36 in drain pipe section 28"to a drain 38 or to a section of pipe 40 which is connected to the main pipe cold leg 24 and inlet 26 for recirculation through reactor 12 to cool the core during the shutdown period or in an emergency when the auxiliary heat exchanger's capacity is needed for safety reasons.

The entrained air or cavitation sensor, which is flowmeter 32, is connected in circuit to controller 42 by a cable 46. Controller 42 is also

connected in circuit to a liquid level sensor 44 by means of a cable 48.

Controller 42 is preferably an ABB (IAD Division), Vasteras, Sweden, Model AC110 Programmable Logic Controller. The liquid level sensor is preferably as disclosed in Application Serial No. 08/864,644. It could also be as disclosed in U. S. Patent No. 5,541.969, assigned to the same assignee as the instant application.

Controller 42 acts to control and modulate valve 50 located in make-up feed pipe 52 between reservoir 54 of make-up coolant 56 and the intersection 60 of make-up pipe 52 with loop or drain pipe 28.

Valve 50 may be of the pneumatic type sold by Fisher Controis Company of Marshalltown, Iowa 50158 as a Full Sized Design ED Linear Cage Valve with a Type 667 Diaphragm Actuator. Motor operated and solenoid valves may also be used for valve 50.

In operation, as long as the proper midloop fluid level is sensed by liquid level sensor 44 and no significant air entrainment, cavitation or other flow volume anomaly is sensed by flowmeter 32 or an equivalent entrained air or cavitation sensor within, across or adjacent to the drain pump 33, the valve 50 is kept closed by controller 42 through its circuit connection cable 58 with valve 50. If air entrainment or cavitation is sensed by the cavitation sensor 32 or the liquid level sensor 44 senses a too low fluid level in main pipe 18, coolant 56 is fed through modulating control valve 50 to drain pipe 28 via intersection 60. This will minimize cavitation in pump 33 and prevent a dangerous situation in the nuclear plant.