Nuclear power plant coolant pumps

Hazardous Decomp. Prods. Heated to decomp., emits toxic fumes of Hg HMIS Health 3, Flammability 0, Reactivity 1 Uses Amalgam dental fillings catalyst electrical apparatus cathodes for prod, of chlorine and caustic soda instrument indicator fluid thermometers barometers mercury vapor arc lamps extraction metallurgy mirror coating boilers coolant and neutron absorber in nuclear power plants vacuum pumps... 

Fig. 1. Pressurized water reactor (PWR) coolant system having U-tube steam generators typical of the 3—4 loops in nuclear power plants. PWR plants having once-through steam generators contain two reactor coolant pump-steam generator loops. CVCS = chemical and volume-control system.

FIG. 2. "Arctica "-type icebreaker s nuclear power plant configuration 1-reactor, 2-steam generator, 3-reactor coolant pump, 4-auxiliary turbine generator (ATG),... 

SAKHA-92 is a maintenance-free nuclear power plant of increased safety. Plant design was developed on the basis of PWR technology, but implements integrated steam and gas pressurizer systems and relies on natural circulation of the primary coolant (Fig. 1). The use of such designs as leak-tight turbine-generator, canned condensate and feed pumps allows to secure the tightness of both primary and secondary circuits, which in turn make it possible to exclude some auxiliary systems. 

The V-sequence refers to a Loss of Coolant Accident (LOCA) through the Low Pressure Coolant Injection System (LPIS), which in this type of nuclear power plants has a common part with the Residual Heat Removal System (RHRS). The suction pipe of the RHRS connects two of the three hot legs in the primary circuit with the RHRS pumps placed into the Auxiliary Building. It has been assumed that three isolation valves fail in one of these pipelines (two of them are motorized valves) a break near the RHR pump has been postulated as a result of the over-pressurization generated in the pipe. 

Since some of the operational research reactors were constructed more than 30 years ago, it is recommend that surveillance instruments be installed to monitor in primary coolant pump. Inductive accelerometers and displacement sensors are widely used in nuclear power plants, and in the industry in general, and have been proven to be reliable. By monitoring piunp vibration, one can anticipate failure conditions and make the necessary repairs well in advance. 

Two-circuit reactor plant (RP) with a vessel-type pressurized water reactor is used for floating power unit of nuclear head and power station. Basic RP components reactor, steam generators and primary coolant pumps are incorporated by pressure nozzles in a compact steam-generating block. KLT-40C RP characteristics are given in Table 1. 

Several events at nuclear facilities in the United States have resulted in dual-unit transients. The causes were due primarily to electrical disturbances (usually lightning strikes) which resulted in the loss of power to reactor coolant pumps or rod control systems. In one case a plant runback occurred when air compressors were lost due to the loss of electrical control equipment in the control air system. 

The CUSH code has been developed to provide a complete representation of a primary circuit containing.several hundred channels with appropriate pumps, steam drums, headers, tall pipes and other features associated with a practical arrangement. Each channel is represented one-dimensionally and such aspects as the appropriate single- and two-phase pressure drops and coolant density distribution are calculated. A three-dimensional specification of the nuclear power distribution is required as input to CUSH. Specifications of the feed-water flow for a given steam output are also needed to determine the sub-cooling at channel entry and may be derived from the JAVAN complete steam plant model. 

From a very fundamental point of view, a coolant must be capable of removing the heat from a reactor with a reasonable expenditure of pumping power. The heat removal rate per unit frontal area of coolant is simply the product of the coolant density, velocity, specific heat, and the temperature rise across the core, i.e., pvCp AT. Typical values of these parameters for water, helium, and sodium coolants are illustrated in Table 11. The coolant-gas velocities are generally consistent with a pumping power of 5 to 10% of the electrical output of the nuclear plant. 

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