TNAG-1776-FCO40-2530-Daya-Bay-nuclear-power-station-project-safety-concerns-in-Ho-1988 — Page 131

2.5 Heat Extraction

from the

A large commercial power station may be producing up to about 3 000 megawatts (3 thousand million watts) of heat in its core (a typical electric kettle has a 2 kilowatt- 2 thousand watt-element). Some means is needed to extract this heat from the core and use it to generate electricity. In a gas cooled reactor, as its name suggests, a gas usually carbon dioxide—is pumped through the core at high speed. The emerging hot gas is routed through a 'heat exchanger' and then pumped back into the core to close the circuit. In the heat exchanger, heat is transferred

gas in the 'primary circuit' to water in the 'secondary circuit'. This water is turned into steam, which is used to drive a turbine to generate electricity. In a pressurised water reactor (PWR) water is pumped through the core to remove the heat. The water is prevented from boiling by being kept at high pressure. This hot water is then routed to a heat exchanger where heat is transferred to water in the secondary circuit. The primary water is then pumped back into the core to close the primary circuit. As in a gas cooled reactor, the water in the secondary circuit is turned to steam which is used to drive the turbine to generate electricity.

In a PWR all the fuel pins are contained in a single large pressure vessel which forms part of the primary circuit. In the BWR the water in the single large pressure vessel is allowed to boil and the steam is passed directly to the turbine. In the CANDU design each cluster of fuel pins is in a separate pressure tube and the primary water is pumped through a heat exchanger, as in a PWR. In the RBMK design the fuel pins are again in pressure tubes and, as with the BWR, the water is allowed to boil and the steam is passed directly to the turbine.

It is important that any heat produced in the core is extracted, otherwise the fuel would heat up and damage might result. However, safety does not depend on the continuing operation of the main cooling system, whether it be gas or water. Should this system fail, or any other fault develop, safety systems would automatically shut down the reactor, and emergency cooling systems would remove the residual heat, (at Chernobyl some of these safety systems had been blocked, see Chapter 4).

2.6 Temperature and Void Coefficients

In a PWR the water being pumped through the core serves two main functions; to act as a moderator for the neutrons and to remove heat from the fuel. As the primary circuit water passes through the core its temperature increases and as it passes through the heat exchanger its temperature decreases. The use of water both as a moderator and as a coolant has two major safety advantages. Firstly, water is a better moderator at lower temperatures than at higher temperatures. If for some reason the reaction rate increased, that is the

power started to increase, then more heat would be generated in the reactor core. The water temperature would increase but the moderating properties of the water would decrease leading to a reduction in the reaction rate and hence a reduction in power and heat output. This self-regulating mechanism, known as a negative temperature coefficient, is thus an intrinsic safety feature. Secondly, water is a very much better moderator than steam. If a sudden large increase in power were to occur, causing the water in the core to boil, the steam bubbles or ‘voids' so created would have very poor moderating properties and this would tend to decrease the reaction rate in the core very rapidly. This is another intrinsic safety feature of PWRs, known as a 'negative void coefficient'. The Chernobyl RBMK design, although water cooled, did not have this safety feature under all conditions. When operating at low power, the presence of graphite (not present in a PWR) resulted in a positive instead of a negative void coefficient, and the appearance of steam voids led to an increase rather than a decrease in reaction rate. This was essentially the design feature that, combined with serious operator faults, led to the Chernobyl accident. A similar fault sequence could not occur in a PWR.

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