The reactor coolant system has four reactor coolant pumps (RCPs), one in each of the four reactor cool­ant loops, in the return pipework from the steam gen­erators to the reactor essel. The design Пои through each pump is approximately 6.3 m "

The RCP assembly comprises a vertical sinele-staffp

m. xed now type pump with an overhung &

and a three-stage controlled leakage shaft §$Єа1 .The pump is driven by an above-mounted air cooled three Phase induction motor, solidly coupled to the pump shaft by a removable intermediate spool piece A cut away perspective view of the RCP assemblv is shown in Fig 2.132.

The more important elements of the RCP art — Hp scribed below; e’

• The pump casing is fabricated from a single-piece austenitic stainless steel casting complete with in­tegral suction and discharge nozzles and support lugs. This arrangement eliminates fabrication welds within the casing and hence reduces the in-service inspection of this component to surface examina­tion only, with a consequent minimisation of person­nel radiation exposure.

• The pump is a high specific speed unit and hence the overhung impeller is of mixed flow design with double curvature Francis vanes, which are shrouded along upper and lower sides. The diffuser converts a portion of the velocity head into static pressure and guides the flow out of the radial discharge nozzle. A thermal barrier, attached to the diffuser flange, restricts the transfer of heat from the hot reactor coolant to the pump bearing and seal areas.

• The seal system comprises three seals arranged in series such that reactor coolant leakage to the re­actor building is negligible. The number 1 seal is a controlled-leakage film-riding face seal and the num­ber 2 and 3 seals are rubbing face seals, mounted in a single cartridge type of assembly for ease of maintenance.

• The complete pump and motor shaft assembly runs on three radial bearings, of which two are located in the motor and the third in the pump. The bear­ing within the pump is of the hydrodynamically water-lubricated sleeve type with a self-aligning ca­pability. The bearing material is a carbon matrix impregnated with graphite whilst the journal is stellite on stainless steel. The two radial bearings within the motor are conventional oil-lubricated babbit-on-steel pivoted pad guide bearings.

• A removable spool, located between the pump shaft and the motor shaft, is provided to facilitate in-service inspection and maintenance of the shaft seal system without removing the motor. This de­sign feature reduces personnel radiation exposure and pump maintenance downtime by minimising the number of operations involved in seal inspec­tion and maintenance. [24] reverse rotation of the pump in an idle reactor coolant loop due to back-flow generated by the operating loops. This situation is encountered dur­ing the sequential start-up of the four RCPs.

10.2.5 Steam generator

The four steam generators provide the physical inter­facing link between the primary pressure boundary of the reactor coolant system and the secondary circuit system. Their function is to transfer heat from the reactor coolant water to the feedwater supplied from the turbine condensate system, thus converting the latter to essentially dry and saturated steam. The steam is then used to drive the turbine-generators.

Each loop of the reactor coolant system contains a recirculation type steam generator, as shown in Fig 2.133. The unit is basically a vertical shell and U-tube heat exchanger, in which heat is transferred from a single-phase fluid at a higher temperature and pres­sure on the tube side, to generate a two-phase steam/ water mixture at lower temperature and pressure on the secondary (or ‘shelf) side.

The reactor coolant water at 155 bar and 323.6°C, is pumped at a constant flow rate through the U — tubes, where its heat is transmitted through the tube walls to the water/steam mixture on the secondary side of the unit. During steady state operation, the steam generated in the secondary side is balanced by the addition of feedwater, thus maintaining a constant fluid inventory and heat content on both secondary and primary sides of the unit.

The primary side consists of the inlet and outlet plena jocated in the hemispherical bottom channel head, and the inside of the U-tubes. Reactor cool­ant flows into one half of the channel head, thence through the U-tubes and back into the other side of the channel head, from where it rejoins the reactor loop.

The secondary side consists of a lower cylindrical shell surrounding the tube bundle and an upper steam drum which contains the moisture separation equip­ment, joined by a transition cone. Feedwater enters through the feedwater nozzle and flows into the feedwater ring. It then spills out from inverted U — tubes located on the feedwater ring and flows down an annular gap formed by the inside of the lower shell and a wrapper barrel which surrounds the U-tube bundle.

The location of the wrapper barrel ensures that the mixture of recirculated water and feedwater enters the U-tube bundle at the level of the tubesheet upper face.

The feedwater is supplied from the plant feedwater system at approximately 56°C, below its saturation temperature. Inside the steam generator, the feed — water is joined by the water recirculating from the moisture separators, producing a feed mixture for the tube bundle that is close to the saturation tern-

perature. Thus, only a small portion of the tube bundle, located just above the tubesheet, functions as a ‘preheater’ to raise the fluid temperature to sat­uration point. The majority of the tube bundle thus operates in the heat transfer efficient nucleate boiling regime.

As the water flows upward through the tube bun­dle, the heat transferred from the reactor coolant raises its temperature to saturation value. Thereafter steam is generated until, at exit from the tube bun­dle, approximately 25% of the water has been con­verted to steam. The steam/water mixture rises into
the upper section where a set of centrifugal moisture separators removes most of the entrained water from the steam. The steam continues to the secondary separators located just below the outlet nozzle, where most of the remaining moisture is removed. The steam quality at exit from the unit is at least 99.75(ro. The entrained water removed by the primary and secondary separators is returned to mix with the in­coming feedwater, the mixture then being recirculated through the tube bundle.

An inherently useful characteristic of the design is that it is relatively insensitive to reductions of tube surface area. The reason for this is the relatively small temperature difference between the primary and sec­ondary side fluids. At the hot end of the steam gen­erator, the temperature difference is approximately 40°C, whilst at the cold end the temperature differ­ence is approximately 8.5°C. This results in an overall logarithmic mean temperature difference of approx­imately 20°C. The effect of this is that if for any reason tubes have to be blanked off, a significant change in surface area can be compensated by a small change in operating temperatures.

Some further details of the major steam generator components are as follows.

The hemispherical channel head is made from low alloy steel and is of forged construction. It is welded to the underside of the tubesheet and is divided inter­nally by a vertical divider plate, made from Inconel, which is welded to the channel head itself and to the underside of the tubesheet. Each chamber has a nozzle which is welded onto the primary loop piping, and an access manway. The manways are provided with bolted covers which are isolated from the re­actor coolant by an Inconel insert diaphragm. The internal surfaces of the channel head are clad with austenitic stainless steel, which provides a corrosion — resistant barrier between the channel head material and the reactor coolant.

The tubesheet is a single forging of low alloy steel, approximately 4190 mm in diameter and 610 mm thick. Together with the U-tubes, it forms the bound­ary between the primary and secondary sides of the steam generator. The outer circumferential portion of the forging is a flat ring into which the vertical supports for the unit connect. Tube holes are drilled through the tube plate into which the U-tubes are inserted. The underside face of the tubesheet is clad with Inconel to provide a corrosion-resistant barrier between the forging and reactor coolant. The Inconel cladding is the same alloy material as used for the tubing, thus enabling the tube-to-tubesheet weld to be made between similar materials.

The tube bundle consists of 5626 U-tubes, fabricated from Inconel 600 material. The tubes are 17.48 mm outside diameter and 1.016 mm wall thickness. The total length of tubing within one steam generator is approximately 100 km. The tubes are given a special thermal treatment after being drawn to size, which improves the corrosion resistance of the material. Additionally, tubes with the smallest radius bends are given a further heat treatment to reduce the resid­ual stresses resulting from the bending operation. The tubes are inserted through, and supported by, the tubesheet and tube support plates. The tube ends are welded to the tubesheet cladding, after which each tube is hydraulically expanded over virtually the full depth of the tubesheet.

The tube bundle is supported by seven tube sup­port plates which are themselves supported and held in position by stay rods and spacer tubes.

The U-tubes pass through quatrefoil-shaped holes in the support plates formed by drilling and broaching. The hole shape consists of four lobes and four flat support lands around each tube. This design of hole shape reduces the tendency to dry-out and, hence, to increase the concentration of corrosive chemicals in the region of the tube and support plate intersection. The support plate material is a ferritic stainless steel.

A flow distribution baffle, made of stainless steel, is located between the bottom tube support plate and tubesheet. Its purpose is to promote cross-flow in the tubesheet region, thus reducing the potential for tube dry-out in this region. The tubes pass through the baffle in either individual round clearance holes or through the circular central cut-out. In the U-bend region, anti-vibration bars are provided to stiffen the bundle and restrain any vibration tendency of the tubes.

The secondary side pressure boundary consists of the lower and upper shells, the transition cone and the top head. The lower and upper shells are constructed from low alloy steel forged into cylinders of requi­site diameter. The transition cone between lower and upper shells is also of forged construction. The ellip­soidal head is either a single forging or made from two shaped plates. The shell contains several access openings consisting of manways (upper shell) and handholes (upper and lower shells). These allow ac­cess for the maintenance of moisture separation equip­ment, visual inspection and sludge lancing operations at the top surface of the tubesheet. In addition to the main feed and steam nozzles, various other small nozzles are provided for water level instrumentation.

The secondary lower shell contains the tube bundle wrapper and the upper shell contains the moisture separation equipment. The carbon steel wrapper as­sembly encloses the tube bundle and forms the inner surface of the downcomer annulus. A conical transi­tion piece welded to its upper end provides a transi­tion to the assembly barrels of the primary moisture

separators.

f!,e main feedwater nozzle located at one side of the upper shell connects to an internal distribution rinc Feedwater is discharged from the ring through j. nozzles located on its top side. The design of the j nozzle, inlet ring and J-nozzIes inhibits drain — — a 0f the ring if the water level should drop below it’thereby reducing the possibility of water hammer effects.

The steam outlet nozzle at the top of the unit is pro­vided with a flow restrictor which is designed to limit steam flow in the unlikely event of a break in the main steam line. It consists of seven Inconel venturi inserts, installed in holes in a low alloy steel forging.

The moisture separation equipment located within the upper shell region comprises two stages. The sixteen first stage separators are located directly above the tube bundle and consist of 508 mm diameter by 3048 mm high assemblies containing static swirl vanes. These vanes impart rotary motion to the steam-water mix­ture causing the heavier water to be thrown outward by centrifugal force and diverted to a concentric — downcomer barrel. The water returns to the recir­culating water plenum.

The second stage separators consist of banks of contoured vanes contained in a housing. The contours in the vanes produce multiple changes in the flow direction, thus causing removal of the remaining en-> trained water from the steam.

A steam generator blowdown system is provided to draw a small water flow continuously from each steam generator, just above the tubesheet, to purify it and return it to the condensate system, or reject it as etfluent. The purpose of this system is to prevent excessive concentration of non-volatile impurities due to continuous evaporation in the steam generator.

The steam generator rests on four vertical steel col­umns with spherical plain bearings at top and bottom. This arrangement allows for the thermal expansion of the coolant loop to be accommodated by lateral movement of the unit. The support columns are at­tached at four locations 90 degrees apart on the outer circumferential ring of the tubesheet forging, in addition to vertical support, lateral support is provided at tubesheet level and just below the shell transition cone. These lateral supports consist of steel bumpers and dampers. The bumpers are provided with clearances to allow for normal thermal expan­sion. In the event of earthquake or pipe rupture.

however, their function is to prevent excessive move­ment of the steam generator. The dampers provided at the upper support location are designed to allow slow thermal movements but also prevent any exces­sive sudden movement by earthquake or pipe rupture.

10.2.6 Reactor coolant pressure control system

During normal station operation and followina fault conditions, the reactor coolant system experiences tem­perature changes which cause the volume of the cool­ant to increase or decrease, which in turn affects the pressure. If the pressure were to rise unchecked, the design limits of the pressure parts might be ex­ceeded; an uncontrolled pressure reduction would cause boiling in the system. The purpose of the pressure control system, of which the pressuriser is the major component, is to maintain the pressure within pre­scribed limits during normal operations and in fault conditions, to prevent design limits being exceeded.

The pressuriser is a vertical cylindrical vessel fa­bricated from carbon steel and clad internally with austenitic stainless steel. It has an overall height of approximately 16 m, an internal volume of 51 m which is filled partly with water and partly with steam in thermal equilibrium with each other. The steam space in the pressuriser acts as a buffer to limit the magnitude of pressure changes which result from changes in the volume of the reactor coolant. Such volume changes cause a flow of water into or out of the pressuriser through the surge line, which connects the bottom of the pressuriser to the hot leg of one of the reactor coolant loops. Relatively small changes in pressure are counteracted by the use of either elec­trical immersion heaters or a water spray in the pres — suriser, both of which are under the control of the station control system. The 78 heaters, with a total power of 1800 kW, are located in vertical sheaths which pass through the bottom head of the pres­suriser in three concentric rings around the surge line nozzle. A general view of the pressuriser is shown in Fig 2.134.

During normal operation a small quantity of water is sprayed continuously through the spray nozzle at the top of the pressuriser, being supplied via two control valves from the cold legs of two of the loops. This small flow avoids temperature fluctuations and ensures that the boron concentration of the water in the pressuriser does not differ significantly from that in the remainder of the reactor coolant. Should the pressure rise, the spray control valves will open to increase the flow of spray water at 294°C into the steam at 345°C, thus condensing some of the steam and reducing the pressure. Conversely, should the pressure fall, power to the electrical heaters will be increased from the normal steady state value required to offset the effect of the small spray flow and heat losses from the pressuriser. This additional heating will generate steam which will increase the pressure.

There are two sets of safety valves connected to the top of the pressuriser which prevent overpres — surisation of the reactor coolant system under all normal fault conditions. The first set of valves con­sists of three tandem pairs of pilot-operated safety relief valves (POSRV) produced by the French com­pany SEBIM. In each pair, the valves are connected in series; the upstream valve being the normally closed relief valve, and the downstream valve being the nor­

mally open isolation valve. The second set of safe­ty valves consists of two spring-loaded safety valves (SRV), which both act as a diverse (back-up) relief system for frequent faults and also contribute to the total system relief capacity needed for extremely un­likely faults, which have a requirement for a large relief capacity.

The POSRVs are primarily self-actuating valves op­erating under the action of system pressure alone, but

they can also be actuated by an electrical signal to a solenoid control valve in the pilot unit. Each POSRV consists of two basic parts; the valve itself and the

ilot unit which is mounted remote from the valve. The pressuriser pressure is sensed by a small bore pipework connection between the pressuriser and the

pilot unit.

When the pressuriser rises to the relief valve open­ing set pressure, a small valve in the pilot unit opens and depressurises the space above the valve’s actuating piston, thereby allowing the system pressure acting on the underside of the valve disc to open the valve. Following sufficient relief from the system, the pres­sure in the pressuriser will fall to below the relief valve closing set point and another small valve in the pilot unit will open and apply pressuriser pressure to the top of the relief valve’s actuating piston, thereby closing the valve.

Should the relief valve fail to close, the pressuriser pressure will continue to fall until the closure set point of the downstream isolation valve is reached, whereupon it will close and terminate the discharge.

This automatic isolation feature reduces the risk of coolant loss due to the POSRV sticking open. The three pairs of POSRVs have staggered set pressures, the lowest being 162 bar, ensuring that the number of valves which open under any particular fault tran­sient is minimised. The two SRVs have a higher set point (172 bar) and are only required to open under extreme fault conditions.

The electrical opening facility of the POSRVs is used to provide cold overpressure protection to the reactor pressure vessel at or near cold shutdown con­ditions. The tendency for reduced toughness of the reactor vessel material at low temperature is exacer­bated by neutron irradiation during station life. Hence, pressure relief is required at pressures below the nor­mal operating set pressures when the vessel material is at a low temperature. To provide protection against faults involving a pressure increase at low tempera­ture, the reactor protection system calculates the maxi­mum permissible pressure at the measured reactor coolant temperature and opens the POSRVs if this pressure is exceeded.

All the safety valves discharge into a pipeline which is connected to the pressuriser relief tank located at a low level in the reactor building. This tank is kept partly filled with coid water and the discharge pipe­line is connected to a submerged sparger pipe so that the discharged steam is condensed in the water and contained within the tank, thereby preventing any discharge to the reactor building. However, bursting discs are fitted to this tank to protect it against over — pressurisation should the steam discharge be excessive.

The system described is generally similar to that of ^estinghouse PWR units in the USA, but significant improvements hae been made for Sizewell B. The, irst improvement is in the use of SEBUM POSRVs v hіch have been developed by Eiectricite de France

especially for use in French PWR stations. After a detailed assessment of all potential suitable valves, the SEBIM type was selected as being the best quali­fied for this particularly arduous duty of both steam and water relief. The second major improvement con­cerns the fabrication details and materials properties of the pressuriser vessel, which have been enhanced by the use of SA 508 class 3 forgings for the three cylindrical shells and for the top and bottom heads. The nozzles on these heads are integrally forged, there­by eliminating the head-to-nozzle welds required using normal fabrication techniques. Longitudinal shell welds are also eliminated.

In order to give added assurance that the pressuriser vessel will not suffer disruptive failure in service, both the forgings and the welds are subject to redundant and diverse non-destructive inspection during fabrica­tion as well as further inspections during the plant lifetime.

10.2.7 Chemical and volume control system (CVCS)

The chemical and volume control system is one of the principal auxiliary systems connected to the reactor coolant circuit and is essential to normal operation of the reactor. It comprises mechanical plant and equip­ment located mainly on the lower floors of the aux­iliary building.

Careful control of the primary coolant chemistry in the PWR is necessary because of the potentially severe corrosive effect of the high temperature water environment on the materials of construction of the reactor circuit and on the fuel. Corrosion rates must be minimised for two main reasons:

„ • To prevent undue degradation in the structural in-
tegrity of the primary circuit pressure boundary.

• To limit the rate of formation of radioactive ‘crud’ which, when deposited around the circuit, contributes to the radioactive dose received by plant operators.

Despite careful chemistry control there will inevitably be some corrosion. Moreover, small unavoidable leak­ages of fission products through the fuel cladding and radiolysis of the coolant will also give rise to gaseous, liquid and solid impurities which must not be allowed to build up in the primary coolant. To control these impurities, the chemical and volume control system uses demineralisation and filtration processes to clean up primary coolant circulating in a bypass loop con­nected to the primary circuit.

Initial filling of the circuit with very high purity demineralised water, to which only boric acid and the necessary pH correction agent (lithium hydroxide) are added, is followed by rigorous degassing procedures to achieve a very low level of dissolved impurities,

particularly oxygen, prior to reaching operating tem­perature. Thereafter, chemistry control is exercised via the CVCS, a simplified arrangement of which is shown in Fig 2.135.

The main flow path of the CVCS consists of a bypass loop through which is circulated a small, more or less constant, fraction of the primary coolant flow extracted or letdown from one of the cross-over legs. After processing as required, the coolant is returned to one of the cold legs as a charging flow. The CVCS also provides the principal method of controlling the amount of coolant in the primary circuit by regulating the charging flow, as required by the reactor make-up control system, so as to maintain the correct level of liquid in the pressuriser.

Since the ion-exchange processes used by the CVCS for purification take place at low temperature (60°C or less) it is necessary to cool the letdown flow prior to processing. To reduce the amount of useful heat energy lost, the first part of the required temperature reduc­tion (to about 150°C) is achieved regeneratively. Further cooling and two stages of pressure reduction
then bring the letdown flow to suitable conditions for purification.

Filtration and demineralisation are used to remove suspended (particulate) and dissolved impurities re­spectively from the letdown flow. A 2-stage filtering process down to 5 fim is followed by passage through mixed-bed and, if required, cation bed ion-exchange resin beds. The filters and demineralisers, which ac­cumulate high concentrations of radioactive corrosion products and other contaminants, are heavily shielded and provided with the means of remotely changing spent filter elements and resins, which are disposed of via the radioactive waste management systems.

After filtration and demineralisation, the letdown flow reaches the volume control tank (VCT). This tank of 7 m3 capacity is maintained partly full of coolant under a continuously fed blanket of hydrogen, at about 5 bar overpressure. At this low pressure point in the system, gaseous impurities in the coolant come out of solution and mix with the hydrogen which transports them to the radioactive waste man­agement plant. The partial pressure of hydrogen in

the blanket causes hydrogen to dissolve in the reactor coolant and this maintains the dissolved oxygen content at an extremely low level because of recombination.

primary coolant is taken from the VCT by one of the two centrifugal charging pumps and supplied con­tinuously to the reactor circuit at a rate controlled either automatically or manually by the operator. The wo pumps are 10-siage horizontal machines driven pv ‘ ‘ kV electric motors supplied with power from redundant switchboards of the station essential elec­trical system. One pump only is normally in use, supplying 7.5 kg s of charging flow to the reactor t 155 bar. Most of this coolant flow reaches the circuit via the regenerative heater exchanger, in which it cools (and is heated by) the letdown flow. However, the charging pumps also supply flow via a suitable filter to the shaft seals of the reactor coolant pumps, and some of this flow also reaches the reactor circuit, the balance that leaks off being returned to the suc­tions of the charging pumps.

During normal plant operation at steady load, the net charging and letdown flows and their boron con­centrations are identical. As burn-up of the fuel pro­ceeds and its reactivity is steadily reduced, a reduction in primary coolant boron level is needed (from about 700 ppm boron at the beginning of a core cycle to about 50 ppm at the end) to maintain the optimum control rod configuration. Thus, from time to time the CVCS is used to dilute the primary coolant with fresh coolant having a lower dissolved boric acid "Content.

This is achieved by mixing demineralised water with the flow leaving the volume control tank; the excess of incoming borated coolant is diverted away to the boron recycle system (BRS) for possible re-use. Con­versely, when reactor shutdown for refuelling is re­quired, the primary coolant must be borated to 2000 ppm and a supply of concentrated boric acid solu — — tion is fed to the charging pumps, the relatively dilute letdown How being diverted away to the BRS. Other chemical additives, such as lithium hydroxide (for pH adjustment) and hydrazine (for oxygen scavenging during reactor start-up), are also introduced from a separate mixing facility. These operations involve an element of manual control.

In the event of small leakages from the primary ‘.ircuit, the reduction in inventory that would result is olivet by an increase in the charging flowrate. The charging pumps can each provide sufficient flow to make up the loss of coolant from any anticipated operational leakage, or from a single instrument line connected to the primary circuit in the unlikely event oi its rupture. The charging flow can also be increased! o compensate for shrinkage of the coolant due to temperature changes. This make-up role of the CVCS! S important in assuring reactor safety following faults involving reduction in primary coolant inventory; a larger external tank of borated water known as the ■ etueliing water storage tank (RWST) provides a se — ‘-nre supply of coolant to the charging pumps if

needed, In the very unlikely event of total failure of this route, a completely separate pair of positive dis­placement emergency charging pumps, driven by steam turbines and drawing supplies from a dedicated stor­age tank, provide a diverse and redundant alternative means of RCP seal injection and borated water make­up.

10.2.8 Control rod drive mechanism

Control of the nuclear chain reaction in the core is achieved by the movement of neutron absorbing rods arranged in clusters, called the rod cluster control as­semblies (RCCA), and by varying the concentration of boric acid in the primary coolant using the CVCS as just described. Each of the 53 RCCAs consists of an assembly of 24 individual rods which are attached to a central hub. These individual rods run in the thimble tubes of selected fuel assemblies. When they are withdrawn from the core, the RCCAs are sup­ported by the guide tubes which are part of the reac­tor internal structures. Each RCCA is held, withdrawn or inserted by a control rod drive mechanism (CRDM), which utilises latches to engage in grooves in a drive rod which is connected to the central hub of the RCCA. A CRDM is illustrated in Fig 2.136.

The actuating mechanism and drive rod of each CRDM are completely enclosed within a pressure hous­ing which is attached to a penetration in the reactor pressure vessel closure head. Each head penetration consists of a plain Inconel tube welded to the vessel head with an externally-threaded stainless steel section welded to the top of the tube.

JEach CRDM pressure housing comprises two forg- inp manufactured from type 304 austenitic stainless steel. The lower forging is the latch housing which encloses the drive rod when the RCCA is withdrawn from the care. The top of the rod travel housing is closed with a threaded plug containing a small vent valve. All joints in the pressure housings are threaded, with a canopy seal weld over each threaded joint to provide a hermetically sealed pressure bound­ary. The latches are magnetically actuated and there are thus no penetrations through the CRDM pressure boundaries.

The tops of the CRDMs are restrained by sleeves which fit over the tops of the pressure housings and locate in the holes in a targe steel plate, termed the missile shield, which is supported by four legs attached to the reactor vessel head. During normal pow-er operation of the reactor, the missile shield is attached to the sides of the refuelling cavity by tie rods which restrain the whole assembly in the event of a seismic event. It also serves to prevent any mis­siles generated in the unlikely event of a CRDM hous­ing failure from impacting vulnerable items inside the reactor building.

Each drive rod and its RCCA is held, raised and lowered by the latch mechanism which is located inside

the latch housing. Actuation is effected by a stack of three electrical coils which surround the outside of the latch housing. A latch mechanism consists of two sets of three gripper latches, operated by armatures which move when their corresponding electrical coils are energised with a direct current. The structural parts of the latch mechanism are made from type 304 austenitic stainless steel, whilst the armatures are made from type 410 martensitic stainless steel.

The lower fixed set of latches normally holds the drive rod; to raise the drive rod a sequence of DC pulses is applied to the three actuating coils. This causes the upper movable set of latches to grip the drive rod; the lower latches are then withdrawn and the upper latches raised sufficiently to allow the lower latches to grip the groove below that previously held. The upper latches are then withdrawn and pushed downwards by a spring, and at the end of this se­
quence the drive rod is raised by approximately 16 mm. The control system has a maximum operating speed of 72 such cycles per minute. Rod insertion is essentially the reverse of the sequence described above.

The 53 control rods are arranged in 6 shutdown banks and 3 control banks. The shutdown banks are withdrawn completely prior to criticality, whilst the three shutdown banks are withdrawn sequentially with a predetermined overlap to control the reactor power. When the reactor is tripped, the protection systems open circuit-breakers in the electrical supply to the rod control system, causing the coils to be de-energised on all 53 of the CRDMs. The weight of the drive rods forces the latches to withdraw as the rods fall into the reactor under gravity. A dashpot action between the RCCAs and the bottom part of the fuel assembly thimbles causes the RCCAs to decelerate towards the bottom of their fall, which takes up to two seconds.

The position of each control rod is measured by a column of detector coils which surround its rod travel housing. These coils detect the presence of the drive rod by its effect upon their electrical impedence; the top of the drive rod is between these two coils which have a significant difference in their AC im­pedance. The measurements of rod position are used by the reactor protection system to provide the core limits protection function, and also to actuate the emergency boration system in the unlikely event of failure of sufficient rods to insert fully following a reactor trip. It is also used to check that all of the rods within a bank are inside their alignment tolerance.

When the reactor vessel head is removed, the drive rods and RCCAs are left behind. Using a special tool, the drive rods are decoupled from the RCCAs and removed with the reactor upper internals, leaving the RCCAs behind in their respective fuel assemblies.

The CRDMs used for Sizewell В are identical to those used in many Westinghouse plants, this design having evolved from the original design produced by Westinghouse. Considerable operational experience has been gained with this design of CRDM which has proved to be extremely reliable. Experimental life test­ing of these mechanisms has demonstrated a design lite in excess of 2.5 million steps before wear effects would potentially start to prevent the correct stepping action. The CRDMs in some of the control banks may exceed this limit if Sizewell В is operated in a load-following or frequency regulating mode, but re­placement of latch mechanisms would not present any significant problems.

The equipment mounted on the vessel head forms an integrated head-package which is lifted off with Tie head to facilitate refuelling. This equipment in­cludes the CRDMs, the missile shield and lifting rig, together with the CRDM cabling and connectors and ‘he cooling system shroud, through which air is drawn to cool the CRDM coils. The head package for Size — w. l В has been specifically designed to minimise the amount of equipment which must be removed during refuelling operations, as this reduces both time and personnel radiation exposures. One major constraint imposed upon the design of the package was the maximum diameter which will still permit the use of the complete ring type of multistud tensioner to re­move and replace the reactor vessel flange bolting, It is considered that the optimum design has been de­veloped for minimising both the radiation exposure and the time taken to remove and replace the reactor vessel head.