(PWR)

10 Pressurised water reactor and associated systems

Пн* CEGB pressurised water reactor (PWR) power stations will, as Гаг as practicable, be to a common design. This approach confers advantages in minimis­ing the cost of redesign and development work and in avoiding the need to repeat safety analyses. However, for the purpose of the following discussion, attention will be focussed on the Sizewell В station to be built immediately alongside the existing A (magnox) station.

Sizewell В is rated at 1245 MW(e) gross and incor­porates a 3425 MW(t) 4-loop nuclear steam supply system (NSSS) supplying twin 3000 r/min turbine- generators. The NSSS is based on a well-established

Westinghouse design of considerable maturity and proven performance.

The layout of the NSSS, the arrangement of the plant buildings and the design of most of the aux­iliary plant and systems are based on the standardised nuclear unit power plant system (SNUPPS) design produced by Bechtel Power Corporation in associa­tion with L’S utilities. Two stations (Callaway and Wolf Creek, which first raised power in 1984 and 1985 respectively) have been built in the United States to the SNUPPS design, which is itself based on the Bechtel Generic Design that has been adopted for eleven PWR units in the US and other countries.

The PWR system utilises an indirect heat transfer cycle, in which the primary or reactor coolant is used to heat the secondary fluid which, in turn, drives the turbine-generators. The NSSS comprises the reactor, the reactor coolant sy-vm and a number of auxiliary and safety systems. The reactor coolant system (RCS) consists of the reactor in its pressure vessel and four cooling loops, connecting the reactor vessel to the steam generators (boilers) which form the interface with the secondary circuit. The NSSS is located wholly within the reactor building or primary containment. The turbine-generators, located in the separate turbine house, receive nearly-dry saturated steam from the steam generators through pipes which penetrate the reactor building wall.

The reactor vessel and the cooling loops are filled with demineralised natural water which is pumped through the reactor, over the fuel in the core and through the steam generator tubes. The water takes heat from the core and gives it up, in the steam gen­erators, to the secondary (feed) water surrounding the tubes. The feedwater is boiled to produce saturated steam at about 285°C and 72 bar. The primary water, which is slightly hotter (about 325°C at reactor exit), is held at a pressure of 155 bar to prevent boiling. This pressure is controlled by a control system and a part-full vertical surge tank (the pressuriser) connected to one of the loops.

The reactor itself comprises the core containing the fuel assemblies, together with components which support and locate them and direct the primary cool­ant flow over the fuel, as well as guide the control rods which are used to alter the power output. The reactor pressure vessel provides a high integrity en­velope to confine the reactor and its coolant, facili­tates the entry and exit of coolant, and allows access to the fuel from time to time for its replacement. The vessel is manufactured of carbon steel, clad inter­nally with stainless steel, and is housed in a vertical cylindrical cavity at the bottom of the reactor build­ing. This cavity is surrounded by a massive rein­forced concrete wall which supports the weight of the vessel and provides shielding against the intense radiation emitted from the core.

Outside this primary shield wail are the four com­partments containing the coolant loops. Each loop comprises horizontal hot leg and cold leg pipework, welded to flanged nozzles near the top of the reactor vessel, which take coolant from above and supply it to the bottom of the core respectively. The hot leg is welded to the steam generator inlet and the cold leg to the outlet of the reactor coolant pump. A U-shaped crossover leg completes the loop by connec­ting the steam generator outlet to the reactor coolant pump inlet.

The steam generators comprise vertical shell and inverted U-tube heat exchangers, the bundle of U — tubes connecting the inlet and outlet sides of the pri­mary w-ater channel head. The heat transfer surface of the U-tubes is situated well above the core, be­cause of the arrangement of the primary circuit com­ponents, and this helps to ensure adequate natural circulation of primary coolant to remove decay heat from the fuel w-hen the reactor coolant pumps are not running.

The loop compartments and the pressuriser cell are surrounded by the secondary shield wall, outside which is an outer annular zone containing auxiliary equipment. All high energy pipework and equipment is contained as far as practicable inside the second­ary shield wall. The outer annulus experiences lower temperatures and radiation levels during power op­eration and personnel access is permissible for limited periods.

Principal items of auxiliary equipment inside the reactor building include the following:

• The fuel handling equipment, which transports new and irradiated fuel and core components between the core and the fuel transfer tube and between the reactor and fuel buildings.

• The polar crane for handling the reactor vessel head, its muiti-stud tensioner and other major com­ponents, as necessary for refuelling and maintenance operations.

• Heating, ventilating and air conditioning (HVAC) equipment.

• The accumulators and recirculating sumps of the emergency core cooling system (ECCS).

• The spray headers and supply pipework of the containment spray system.

• The emergency boration system equipment.

• The incoming and outgoing lines, ducts and cables to and from reactor and ancillary equipment that penetrates the containment boundary, with isolating valves or dampers as appropriate.

All these auxiliary systems and equipment, together with other features of the power station, are described in the following sections. After a description of the layout of the site and of the major buildings, together with a discussion of measures adopted to control the spread of radioactive contamination, the major re­actor coolant system components are discussed in some detail, dealing in particular with those aspects which ensure that the necessary very high level of integrity is achieved. As well as the auxiliary items listed above, the various control and protection sys — icrns which contribute to the safe operation of the station and the avoidance or mitigation of faults are

described.

The design philosophy with regard to removal of post-trip heat and assurance of power supply integrity js of particular importance for the Sizewell В design and layout, and this is discussed in some detail.