Boiling-Water Reactor


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boiling-water reactor

[¦bȯil·iŋ ¦wȯd·ər rē′ak·tər]
(nucleonics)
A nuclear reactor in which the coolant is water, maintained at such a pressure as to allow it to boil and form steam. Abbreviated BWR.

Boiling-Water Reactor

 

a nuclear reactor in which the core is cooled by a boiling coolant, usually water. Boiling-water reactors may be used in single-loop schemes of atomic power plants, in which steam produced in the reactor is conducted directly to a turbine. The favorable heat-transfer conditions that may be achieved by boiling water in the reactor core make it possible to attain high specific loads in the core. Factors that limit the increase of the specific power of a boiling-water reactor are the heat flux per unit length of the fuel element at which melting of the nuclear fuel occurs and the heat flux per unit surface at which the critical heat transfer point occurs, causing coating of the surface by a vapor film, a sharp reduction of the heat transfer, and, consequently, burnout of the fuel element jacket.

There are boiling-water reactors of the vessel and tube types. In vessel-type reactors, water is also the moderator, whereas in tube-type reactors the boiling occurs in tubes located within the moderator blocks. Separation of the water-steam mixture takes place either inside the reactor vessel or in external drum separators. The separated water is mixed with the cooler feed water and passes into the evaporator part of the core, where it evaporates partially.

In the USSR, two tube-type reactors are being successfully used at the I. V. Kurchatov Beloiarsk Atomic Power Plant. These reactors have power of 100 and 200 megawatts (MW), respectively, and represent the first industrial application of steam superheating by nuclear means. In the first block reactor, which was started up in 1964, the heat of boiling water in evaporating channels is used to generate secondary steam in steam generators, which is then superheated in the reactor channels of the second loop. The radiation safety of both heat-transfer loops, which was confirmed by operation, made possible the use in the second unit, which was started up in 1967, of single-loop circulation of the boiling water and superheated steam, which is distinguished by simplicity and economy. A power plant with an experimental boiling-water reactor of the VK-50 vessel type, with a power of 50 MW and natural circulation of the coolant, has been in operation in Dimitrovgrad since 1965.

A large number of boiling-water reactors have been designed in various countries—for example, the Oyster Creek vessel-type boiling-water reactor (USA), with a power rating of 515 MW, in which the devices for steam separation and the multiple circulation loop for the coolant are located within the vessel. Positive operational experience with boiling-water reactors, the possibility of providing high power output from a single unit, and the use of superheated steam, as well as the simplicity and economy of atomic power plants containing boiling-water reactors, make this type of reactor very promising in world power production. The Leningrad, Kursk, and Chernobyl’sk unit electric power plants, with carbon-uranium tube boiling-water reactors and with a power of 1,000 MW each, are under construction.

V. P. VASILEVSKII

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