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The first 30 years of nuclear power have seen considerable developments in reactor design. The early stations in Britain, the Magnox type were improved versions of Calder Hall. The Magnox design has been superseded by the Advanced Gas-cooled Reactor (AGR) which, with higher steam temperature and pressure, operates at much higher efficiency.

Parallel development of water-cooled reactors, mainly in the United States, produced the Pressurized Water Reactor (PWR). Much research and development has gone into the fast reactor, which can use uranium far more effectively. A commercial demonstration fast reactor, featuring a common European design, is expected to be built by the turn of the century.

In electrical power production, there is no fundamental difference between fossil fuel and nuclear systems. Heat generated from splitting atoms or burning coal, oil or gas is used to boil water, make steam, drive turbines and generate electricity. This book is basically concentrated on the nuclear power reactor

The reactors considered here are the pressurized-water reactor, the boiling water reactor, the heavy water reactor and the gas-cooled reactor. All the types considered are thermal reactors, that is, the neutrons are moderated to make use of the higher fission cross-sections at low neutron energy.

The predominant commercial reactor is the light-water reactor (LWR). The light-water reactor is found in two types, the pressurized water reactor (PWR) and the boiling water reactor (BWR). These reactors use ordinary water as the moderator and coolant. Natural uranium cannot be used as the fuel in a LWR. All the fuel is, indeed, uranium. But the concentration of the fissile U235 has to be increased from its natural 0.7 % to almost 3 % in order for such a reactor to operate. The 24 life-time uranium requirement of a LWR is 4000 to 6000 tons of natural U3O8.

Pressurized-Water Reactor

The core of a pressurized-water reactor (PWR) consists of a large number of square fuel assemblies or bundles. Each of these contains about 200 fuel pins closely held together in a matrix with no outer sheath. A full-sized (i.e. 1000 MWe) PWR may contain nearly 40,000 fuel pins, containing about 110 tons of UO2. The power generation density in the core is about 98 kW/liter. All assemblies have provision for the passage of control rods through the bundle. These rods are manipulated by drives at the top of the reactor. Additional control is available by addition of neutron absorbers (such as boric acid) to the coolant. The reactor vessel itself may be 40 ft high by 14 ft in diameter, made of steel 8 in, or more thick. The top head is removable for refueling. The coolant enters the reactor vessel near the top of the core and constrained by a "core barrel" between the vessel and the core, flows to the bottom of the core, then up through the core itself and out to the steam generators from which it is recirculated by large pumps. Maintaining the pressure at about 2250 psi prevents the formation of steam in this "primary” system.

Instead, steam is raised in a "secondary" system by allowing heat to flow from the high pressure primary coolant to the lower pressure secondary fluid. This heat transfer occurs through the walls of large numbers of tubes through which the primary coolant circulates in the team generators. Afterpassing through separators to remove water droplets, the steam leaves the steam generator for the production of electricity. After condensation, it returns as liquid to the steam generators. The overall thermal efficiency of a PWR is about 32 %.

A large PWR has four external circuits, each with its own steam generatorand pump. Since maintenance of the pressure near the designvalue la crucial (to avoid the formation of steam, on the one hand, andrupture of the primary circuit, on the other), a PWR system also includesa "pressurizer", connected to the "hot" leg of one of the steam generatorcircuits. The pressurizer is filled partly by water and partly by steam, itinitiates condensation or vaporization as needed to keep the pressure

within specified operating limits.

All of these components of the "primary" system - the reactor vessel,

steam generators, pumps, and pressurizer - are in the containment building.This structure is steel-lined, reinforced concrete, designed to with25stand the overpressures expected if all the primary coolant were releasedin an accident. Sprays or other means are available for condensing steam(thereby reducing pressure) and for removing any radioactive materialreleased into the containment atmosphere.

Various features are available for limitation of abnormalities. Theseinclude introduction of control to shut down rapidly (i.e. to "trip” or“scram") the chain reaction, and systems for continuation of core coolingthereafter. In case much of the primary coolant is lost through apipe-break or other leak, one of a number of redundant emergency corecooling systems (ECCS) becomes operative. The first is an accumulatorwhich automatically injects water into the system to provide continuedcooling of the core. In addition, there are independently-powered systemsfor actively pumping stored water into the primary system. Thesesystems rely on redundancy of components to assure their availability inemergencies. This is also true of other devices. For example, each of thefour primary coolant pumps has a capacity large enough to provide byitself for removal of decay heat after shutdown of the reactor.

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