The coolant gas temperatures at the AGR stations during normal operation range from about 290°C at reactor inlet to 650°C at reactor gas outlet. As with the magnox station vessels, insulation and a cooling system are attached to the liner to keep the concrete temperatures and thermal gradients at, or below, the design values.

The insulation design is different at each station. There are two basic types which have been used; metallic foil at Dungeness В, Hartlepool and Heysham

1, and ceramic fibre at Hinkley Point В and Heysham

2.

The insulation on the central region of the floor and lower parts of the walls of the Dungeness В vessel is similar in design to that at Oldbury, i. e., corrugated rings, foils, spreader plates and coverplates on the floor, and foil and wire mesh on the walls. The design of the insulation was changed for the upper part of the sidewall and roof of the vessel to enable a multi­stud fixing of the coverplate to be used. This was necessary because of the high turbulence in that part of the vessel and the cyclic loadings which it imposed on the coverplate. The insulation is made up of layers of elements, each element consisting of a stainless steel dimpled foil covered by plain foils. Sealing be­tween adjacent elements is achieved by thin metallic strips. The elements are covered by a hot-face skin to limit the penetration of gas into the insulation, and an acoustic skin to reduce the effects of noise on the elements. The coverplates and insulation are supported by studs, usually four per coverplate, which are welded to the liner and attached to the coverplate with hoops. The hoops are to provide flexibility in the retention system to accommodate the differential thermal expansion between the coverplate and the liner to which it is attached (Fig 2.82).

The insulation for Hartlepool and Heysham 1 is similar to the element-type insulation at Dungeness B. The insulation and coverplates are supported by either studs or strips of metal welded to the liner. Four studs or strips per coverplate are required to give adequate support (Fig 2.83).

A ceramic fibre insulant material, ‘Triton Kaowool’, was selected for Hinkley Point В and the insulation design has been repeated at Heysham 2. (At Heysham 2 the terminology has been changed and the insula­tion is now called a ‘thermal shield’.) The insulation consists of layers of fibre blanket, the number and thickness of the layers varying according to the local gas conditions and the thermal performance require­ments. In between the layers of blanket are stainless steel foils. On the vertical surfaces there are ‘Z sec-

lions’, or shelves, to give support to the blankets. The blankets are contained at the hot face by a sys­tem of overlapping foils and 0.9 mm thick secondary coverplates. The whole of the insulation is retained by primary coverplates which also are used to com­press the blankets. Each coverplate is retained by a stud welded to the vessel liner. Studs positioned at the corners of the coverplates provide additional blanket support and act as anti-rotation devices for the inter — blanket foils, secondary coverplates, etc. Secondary retention of the coverplates is provided by a system of hoops which interconnect adjacent coverplates (Fig 2.84).

Interblanket support in the standpipe region of the roof is by retention plates held in position by washers welded to studs. The coverplates are welded to exten­sions of the fuelling standpipe liners, and secondary retention devices are fitted between coverplates.

In addition to the basic X and Y systems sequences there are some minor variations for particular fault conditions.

Loss of grid supplies

In the event of loss of grid supplies, all plant is cleared from the 3.3 kV and 415 V essential switch — board^-and the diesel generators will be automatically connected to the 3.3 kV switchboards. The plant is then sequenced to start in exactly the same manner as with grid available. Plant sequencing is delayed after all reactor ‘trips to allow sufficient time for the diesel generators to start and run-up to speed should they be needed, and also to provide a satisfactory loading schedule. Should the grid fail sometime after the trip, the switchboards will be cleared and the diesel will be connected at this stage, and all plant restarted in the same sequence as previously.

Since the starting and standby feed pumps are sup­plied at 11 kV, they are not available following loss of grid, and emergency boiler feed will be introduced into the main boilers for plant protection purposes.

The fuel handling and storage system provides for the safe, efficient and economical handling and storage of the fuel from the arrival of new fuel to the despatch of fully irradiated fuel. This includes the receipt, inspection and storage of new fuel and core control
components, the disassembly of the reactor for the re­fuelling operation, the refuelling of the reactor core, the reactor reassembly after refuelling, the storage of irradiated fuel and control components, and the dis­patch of fully irradiated fuel from the power station. The fuel handling facilities. Fig 2.138, are sited in

two areas:

• The refuelling cavity within the reactor building.

• The fuel building situated adjacent to the reactor building.

The fuel building contains the fuel storage pond with­in which are racks for storing both new and irra­diated fuel assemblies. Adjacent to the fuel storage pond and normally separated from it by removable gates, are the fuel transfer canal and two bays for the preparation and filling of the irradiated fuel transport flask, All the bays are of reinforced concrete con­struction with stainless steel line plates. Also within the fuel building are facilities for the receipt and in­spection of new fuel and the loading bay for the road transport vehicles. In addition, the building houses the pumps and heat exchangers required for the re­moval of the decay heat given out by the stored irra­diated fuel, and the heating and ventilation equipment for the building.

The tw’o areas are connected by the fuel transfer tube which penetrates the reactor building contain­ment. This tube is isolated by a blind flange during reactor operation, to maintain the reactor building containment integrity.

Refuelling is carried out at approximately yearly intervals with the reactor off-load. To refuel, the re­actor is shutdown and depressurised and the closure head is lifted off to gain access to the vessel upper internals and, when they have been removed in turn, the fuel assemblies.

After cooldown and clean-up of the reactor coolant system, the electrical and instrumentation cables serv­ing the CRDMs are disconnected, the cooling air duct between the fan plenum and the reactor vessel head cooling shroud is removed, and insulation is removed from the reactor vessel head area. The vessel head assembly is then clear to permit the multistud ten­sioner to be lowered into position over the reactor vessel studs securing nuts. The studs are de-tensioned, the nuts are unscrewed and both they and the studs are removed by the multistud tensioner which is fully automatic and remotely controlled. After the studs have been removed, the reactor vessel head assembly is removed by the reactor building polar crane and stored on the operating floor.

As the head is removed, the refuelling cavity is flooded with water from the refuelling water storage

tank (RWST). The control rod drive shafts are de­coupled from the rod cluster control assemblies and the upper internals are removed, exposing the core to permit refuelling. The internals are stored under water on a stand at the far end of the refuelling cavity.

The normal refuelling operation involves discharg­ing one-third of the core of fully irradiated fuel to the fuel storage pond, shuffling the positions of the remaining two-thirds of the core of partially irradiated fuel and charging the core with one-third of a core of new fuel from the fuel storage pond. During the process, the core control components are transferred between the fuel assemblies by the refuelling machine to maintain their correct position in the core. Alter­natively, when it is required to carry out reactor pres­sure vessel in-service inspection, the complete core of 193 fuel assemblies together with the core control components, is discharged to the fuel storage pond where the new fuel assemblies are already stored. The core control components are shuffled between fuel assemblies in the storage pond, utilising the various components handling tools suspended from the mono- rail hoist of the pond fuel-handling machine. The new core, made up of new and particularly irradiated fuel assemblies, is then charged into the reactor, leaving the fully irradiated fuel in the fuel storage pond.

At all positions along the fuel handling route, until it is placed and sealed into the irradiated fuel trans­port flask, the fuel is handled under water; this pro­vides an effective, economical and optically transparent radiation shield as well as a reliable medium for the removal of decay heat generated by the irradiated fuel. The water used in all the fuel handling facilities contains boron, normally at a concentration of 2000 ppm, to ensure sub-critical conditions during all fuel handling operations.

Upon completion of the refuelling, the upper in­ternals are replaced over the core, the control rod drive shafts are recoupled and the refuelling cavity is drained. The reactor vessel flange shield is lowered into place around the control rod drive shafts to provide radiation protection for the reactor vessel flange and the sealing face cleaning operations.

The flange shield is then removed and stored on the storage floor and the reactor vessel head assembly is lowered, guided by three guide studs, onto the re­actor vessel. The reactor vessel studs and nuts are replaced and tensioned by the multi-stud tensioner and the tensioner removed. The two cable bridges are lowered into position and the cable connections are made, the cooling air duct is refitted and the insulation is installed.

These operations are followed by the reactor start­up tests, plant heat-up and power raising.

The components of the fuel handling system are designed to handle only one fuel assembly at a time. The main components are the refuelling machine, the fuel transfer system, the pond fuel handling ma­chine and the new — fuel elevator. Various tools are also provided for handling fuel assemblies and core control components.

The refuelling machine is essentially a rectilinear bridge and trolley crane system with a vertical mast extending down from the trolley into the refuelling cavity; the bridge spans the refuelling cavity and runs on rails set into the edge of the cavity at the operat­ing floor level. The bridge and trolley motions are used to position the vertical mast over the fuel as­sembly positions in the core and the up-ender position of the fuel transfer system. The bridge and trolley controls automatically position the mast over the op­erator selected position, with co-ordinates programmed into the control console computer. A television mo­nitor display indicates the position selected and the actual position of the mast.

Separate manual controls are provided for non­automatic operation. The vertical mast supports and guides the gripping devices and the main hoist for the handling of core components.

Only one core component, e. g., a rod control clus­ter assembly, a thimble plug or a fuel assembly, which may or may not have inserted a control component, can be handled at any one time within the mast.

The fuel transfer system comprises a fuel assembly container basket mounted on a conveyor carriage that runs on rails from the fuel transfer canal in the fuel building, through the fuel transfer tube, into the re­fuelling cavity in the reactor building. Located at each end of the transfer tube is a hydraulically operated up-ending device which turns the basket between the vertical and the horizontal. The basket is designed to hold one fuel assembly and the control components can only be transferred inserted in the fuel assembly. In the vertical position, the basket is loaded and un­loaded with a fuel assembly by the refuelling machine in the reactor building, or by an irradiated fuel han­dling tool suspended from the pond fuel handling machine in the fuel building. The fuel is passed hori­zontally through the transfer tube.

The pond fuel handling machine is a wheeled bridge structure with a steel deck walkway and two hoists that are carried on an overhead monorail structure. The bridge spans both the fuel storage pond and the fuel transfer canal, and runs on rails set into the edge of the pond and canal at operating floor level. The machine is used primarily to transfer single new and irradiated fuel assemblies between the fuel storage racks and the fuel transfer system, new fuel assemblies between the new fuel elevator and the storage racks, and irradiated fuel assemblies between the storage racks and the irradiated fuel transport flask in the flask fill bay. The machine, together with various tools, also performs any transferring of control com­ponents between the fuel assemblies in the storage racks. All the core components are handled vertically and underwater. Additionally, the machine handles the opening and closing of the fuel transfer gates, utilising the second hoist on the monorail structure.

The new fuel elevator, situated in the flask fill bay, comprises a fuel basket carriage assembly and a guid­ance system of rollers. The rollers are confined by rajls attached to the bay wall in order to maintain the ertical orientation of the carriage as it is raised and lowered by a winch, mounted above the operat­ing floor. The elevator is used exclusively to hold

“d lower single new fuel assemblies to the bottom of the bay; from there they are transferred to the fuel storage racks by the irradiated fuel handling tool suspended from the hoist of the pond fuel handling machine. For new core control components to be introduced into the storage pond, they must first be inserted into new fuel assemblies.

Fuel storage facilities are located in two areas of the fuel building; one is a transit storage area for new fuel assemblies upon arrival at the station, and the second is the storage pond which contains the permanent fuel storage racks. New fuel arrives at the power station in containers on a road transport vehicle. Each container can carry two fuel assemblies that are clamped to a strongback for support during transit. The containers are lifted off the road trans­port vehicles by a monorail hoist suspended from the underside of the fuel building crane, and are placed for temporary storage in the new fuel reception and preparation area on the operating floor. In this area the fuel assemblies are removed from their containers, detached from the strongback and inspected. The fuel is then transferred via the monorail hoist, the new fuel elevator and the pond fuel handling machine to the fuel storage racks.

The fuel storage racks rest on the fuel storage pond floor. The racks are not tied to the stainless steel line plate at the floor or walls. Each rack consists of a number of stainless steel vertical storage cells formed into a square lattice with full height neutron absorber material on all sides. The inclusion of this neutron absorber permits the storage of a fuel assem­bly in any empty position within the storage racks, without regard to the degree of burn-up, decay peri­od, or initial enrichment, thereby minimising the han­dling of irradiated fuel. Allowance is made in the rack design for the passage of cooling water through the stored fuel assemblies. The pitch of the storage cell lattice is 267 mm which results in an ultimate stor­age capacity for 1323 fuel assemblies. This capacity would suffice for sixteen annual discharges of irra­diated fuel, plus one-third of a core of new fuel, plus the contingent capacity for one full core discharge during in-service inspection of the reactor vessel.

The fuel storage pond cooling system consists of two 100% capacity cooling trains for the removal of decay heat generated by the irradiated fuel that is stored in the fuel storage pond. Each cooling train comprises two 50% capacity horizontal centrifugal pumps, one 100°7o shell and U-tube heat exchanger, a strainer, manually operated valves, and the instrumen — tation required for system operation.

The decay heat generated by the irradiated fuel is transferred from the fuel storage pond, through the heat exchangers of the cooling system, to the com­ponent cooling water system.

During normal system operation, both pumps of one train take suction from the fuel storage pond and transfer the water through a heat exchanger back to the storage pond. The pumps’ suction line penetrates the wall of the fuel storage pond near the normal water level; the return line terminates in a distribution header at the bottom of the pond. An ami-siphon hole in each return line is located near the surface of the water to prevent inadvertent draining of the pond. Normal make-up water for the fuel storage pond, to compensate for evaporative losses, is supplied from the reactor make-up water system. For emergency make­up, the borated water in the flask preparation bay water system is available for use. Two further systems are employed to clean-up the water in the fuel storage pond, fuel transfer canal and handling bays, and in the refuelling cavity of the reactor building.

The fuel storage pond clean-up system contains two centrifugal pumps in parallel, two filters in parallel, a mixed bed demineraliser, a strainer and four float — type skimmers. The pumps and filters are designed for 50% of the system capacity and the demineraliser and strainer are designed for 100% system capacity. The filters are provided to remove particulate matter. The demineraliser removes ionic corrosion impurities and fission products, and the strainer downstream of the demineraliser removes resin fines. The float-type skimmers are positioned so that there is one in the fuel transfer canal, one in the flask fill bay and two in the fuel storage pond; each has inlets positioned for the removal of surface debris.

The similar refuelling pool clean-up system provides a capability for purifying the water in the refuelling cavity during refuelling and, at other times, the con­tents of the refuelling water storage tank. It contains one 100% duty centrifugal pump, four filters in par­allel, two 50% mixed-bed demineralisers each with a strainer, and two float type skimmers stationed in the refuelling pool.

The irradiated fuel transport flask is conveyed to and from the station on a road vehicle, fitted with a transport and tilting frame onto which the flask is secured horizontally. Inside the fuel building reception bay, the fuel building crane is used with a special lifting attachment to elevate the flask to the vertical position. The flask is then lifted the minimum distance necessary to disengage it from the transport and tilting frame, and transferred to the flask fill bay where it is filled with irradiated fuel underwater. Finally the flask is returned lidded, decontaminated and in all ways prepared for loading onto the vehicle for trans­port off site.

The flask preparation bay provides the access, tools and equipment for checking and preparation of the flask, including lid bolt removal/replacement, and for flask decontamination. A position is provided at op­erating floor level for the inspection and refurbishing of the flask lid and its seals.

Three water retaining gates are provided between the fuel storage pond and the vehicle loading bay:

• A flask transfer gate between the vehicle loading bay and the flask preparation bay.

• A flask transfer gate between the flask preparation bay and the flask fill bay.

• A fuel transfer gate between the fuel storage pond and the flask fill bay.

During flask handling operations, interlocks prevent more than one of these three water retaining gates from being open at any one time.

The hour-to-hour control and operation of a nuclear power station is under the supervision of the shift charge engineer or shift manager. The title of charge engineer or manager depends upon the structure of staff at the particular location but they have similar responsibilities (shift manager will be used henceforth). The shift manager has a team of staff to help him manage the most likely set of circumstances for w:hich he may be responsible.

The team which he heads will be typically made up as follows:

• A group of engineers who will control through their functional responsibilities all the activities which make up the operating package at the time. [30]

• A group of health physics monitors to carry out monitoring functions, headed by a health physics foreman.

• A group of maintenance craftsmen of mixed disciplines headed by a maintenance foreman. The group’s prime responsibility is to carry out emer­gency work but in the absence of this will do rou­tine maintenance.

• A security group with specific responsibilities to secure the site.

The shift manager will normally have through ex­ternal and internal communications other specialist engineers and managers who can give advice on any topic on which he feels he may require opinions and/ or instructions. In describing the role of the opera­tor the following notes apply to all members of the manager’s staff to a greater or lesser degree depending on the role they play, but with all members contri­buting to the success of the operation.

The time scale for both fuel and moderator tempera­ture effects is similar in PWRs with time constants of a few seconds.

2.2.4 Fuel feedback

Fuel temperature feedback in PWRs is due principally to the Doppler broadening of U-238 and Pu-240 resonance peaks. The Pu-239 thermal resonance is much less important in PWRs because the neutron spectrum is considerably harder than in AGRs and magnox reaciors. The resonance absorption increases somewhat with irradiation as Pu-240 builds up in the core. The point fuel feedback coefficient is typically -3 mN. r’C throughout the fuel cycle. However, be­cause of the relatively short neutron migration length, ■‘tatistical weighting effects are much more important lor P\ R than for AGR and magnox reactors, and the core average coefficient is highly dependent on the neutron flux distribution.

This is the easiest perturbation to understand since there is a simple relationship between cause and ef­fect. It can readily be assumed that gas flow remains constant, as noted in a preceding paragraph. First, consider the disturbing influences which can cause re — activitv дк to deviate from zero and thereby cause a change in neutron power.

A change in the position of one or more control rods in the reactor core will directly change the value of reactivity. Rod movement may be intentional, such as that carried out either automatically or manually for temperature trimming purposes or to remove/ reinstate a rod for on-load control rod actuator main­tenance; or it may be a fault, such as a fault in an electrical control system causing a rod or rods to drive into or out of the core, or a contactor failure or supply failure releasing a rod to run into the core under gravity. Divergent faults such as rod run-out are covered by fault studies and protection is provided to ensure reactor safety, but prompt action by the control engineer may arrest the fault and keep the reactor safely on load. Such faults are rare, rod run — in being more likely, although convergent and safe, and prompt action by the reactor control engineer (assisted where appropriate by the auto control system on reactor gas outlet temperature) might keep the re­actor on load.

6.6.1 Water purity in feedwater and steam circuits

In order to control corrosion and deposition on the surfaces of boiler and turbine plant, stringent con­trol of water chemistry is essential. This control is achieved by a combination of treatment techniques and monitoring. The latter provides evidence that the control has been maintained, monitors the operation of the chemical control processors and initiates alarms if there is a marked departure from acceptable con­ditions, e. g., after a condenser leak.

The later magnox and AGR stations have once — through boilers which introduce special problems be­cause they cannot be blown down to remove the water impurities in a way similar to that used for drum type boilers.

6.6.2 Other water circuits

Other water circuits that require chemical monitoring include generator stator water cooling, and pressure vessel and reactor ancillary water systems. For the latter two, conductivity, pH and CO: in water are used for the detection and location of leaks.

Important aspects of the measurements are the sampling technique which has to provide a represen­tative sample, and the reduction of pressure and tem­perature to make it suitable for the measuring device.

6.6.3 Steel/graphite measurements

The long-term performance of the core is monitored by measurements on steel/graphite samples and is described in detail in Chapter 1.

There would be no problems with interaction be­tween the graphite core and its steel restraint and support structures if:

• The core restraint is temperature compensated to expand at the rate of graphite.

• The graphite is decoupled from the support structure (diagridj.

The degree to which this is designed into magnox reactors varies. An example of where extra tempera­ture instrumentation is necessary to prevent damage to the graphite from uneven expansion rates is Oldbury, where neither of the above is achieved. In some re­actors, where the temperature compensated core re­straint no longer acts as designed due to mild steel oxidation problems, additional displacement instru­mentation has been installed.

6.6.4 Use of computers

Modern digital computers of the type described in Volume F, Chapter 7 provide the most effective means of collecting, analysing and recording data. They are used extensively to provide the basis for the long term monitoring systems required to operate a nuclear power station.

The operating temperatures typical of water reactors are not high in terms of homologous temperatures (<0.4 Tm) so that thermal creep is relatively slow. Furthermore, the creep strength is considerably en­hanced by the presence of dissolved oxygen and hy­drogen present as hydride. In practice, however, the in-reactor deformation seems to be primarily radia­tion, induced. In-pile creep tests show that the creep rate (e) is almost athermal and approximately linearly dependent on the fast neutron flux (Ф) and on the — stress (or) (Wnod and Watkins, 1971 [18]). The me­chanisms werg assessed by Hesketh (1968) [19] who attributed the enhanced in-pile creep to yielding creep [4]. Diameter measurements on PWR fuel rods by Franklin (1982) [20] produced an equation of the form t = к Ф0-6 a1 5.

In ‘fair agreement’ with the results of in-pile creep experiments (k is a constant). In this study, ovalisa­tion of the clad and ridge formation due to the clad taking on the (hour-glassed) shape of the fuel pellets occurred during the second reactor cycle; subsequent (diametral) clad shape changes were controlled by the pellet stack.

In the early days of PWR, problems were encoun­tered with collapse of the clad into large interpellet gaps. These had formed by accumulation of smaller gaps caused by fuel densification (Roberts et al, 1977 [21]). Whilst few failures were produced, such an effect is obviously undesirable and steps were taken to prevent its re-occurrence. These were the use of a higher starting density for the fuel and internal pre — pressurisation of the pin (to about 10 atmospheres, which roughly doubles at operating temperature) to

reduce the driving force for collapse (Frost, 1982 [22]). Experience has shown that these can prevent clad col­lapse into large inter-pellet gaps and, as shown by Franklin’s results, clad/fuel contact is delayed until some way into the second cycle. At the ends of the fuel rod, of course, the plenums always consist of lengths of unsupported tube. In this case clad col­lapse does not occur because both the neutron flux

and the temperature are low, hence there is little scope for either irradiation or thermal creep.

Gro w(h

Just as with uranium (Section 10.1.2 of this chapter), fast neutron irradiation of zirconium alloy cladding results in growth because the vacancies and intersti-

. js which are produced haw different preferences for^sink positions. If (be metal trains were perfectly randomly oriented, this would haw little effect other [han the” generation of high internal stresses. In prac — of course, any metallurgical treatment will result I j” s’ome kind of texture and, although this can be — patrolled to some extent by the fabrication proce­dure in Zircalos tubes г he ‘exture is usuallv sucii dial fast neutron irradiation tends ю make the tubes I oncer and imperceptibly thinner. Other factors which uTect erowth are prior cold work and temperature (Adamson, Tucker and Fileris, 1982 [23]) and, since the process produces rod length changes it is appa­rent that the presence of cross-rod flux and tempera­ture sradients will also be capable of producing dif­ferential length changes which will lead to rod bowing.

Other, less important, contributions to bowing are possible from non-uniform, radiation-induced stress re­laxation (Montgomery, Mayer and French, 1977 [24]). To date rod bowing seems to have caused few failures and this may be due to the very firm built-in support offered by most spacer designs and the fairly short unsupported rod lengths (typically 0.5 m).

The use of high nickel alloys in a PWR (stainless steel, inconel 600) is responsible for the Co-58 pro­duction from Ni-58 during early reactor life. In con­trast, Co-59 present in hard facing alloys and as an impurity in other materials determines the Co-60 pro­duction later in life. The relative contribution of each of these sources to the Co-60 production depends upon operational aspects such as valve seat lapping and upon the established metal corrosion and corro­sion product release rates. There is therefore an ad­vantage in minimising the Co impurity level in inconel and stainless steel, and developing Co-free or very low Co hard facing alloys. Typical Co specification for inconel and stainless steel is 0.1% with inconel available at 0.05%Co.

Table 1.21 shows that since approximately 73% of the surface area is inconel compared to 7% for stain­less steel, and 20% for zircaloy, there is a clear in­centive to pursue lower inconel Co levels, provided the cost benefit analysis is favourable.

Similarly, although only 0.04% of the surface area is hard facing alloy, a typical Co level for stellite is 66% and some material is located in the vicinity of
the core (control rod drive mechanisms) and in valves. Direct release of Co-60 species is consequently pos­sible by wear and corrosion mechanisms. There is therefore a clear long term benefit in developing low or Co-free hard facing alloys, and this is being ad­dressed worldwide. However, apart from the need to resolve questions of availability and fabrication it is possible that a Co-free alloy may wear more exten­sively and require more maintenance. This increased maintenance dose will, to some extent offset the bene­fits of a low system Co inventory.

All the BCD systems installed, from the earliest mag­nox power stations such as Bradwell up to the last at Wyifa, were engineered from the same basic design concept developed for Calder Hall A (commissioned in 1956).

However, with the construction of large magnox reactors, the twin requirements of improving fuel sur­veillance and rationalising the pipework system (par­ticularly within the pressure vessel envelope) resulted in major design development on BCD components, such as standardised precipitators and improved high temperature selector valves. This greatly improved the BCD system overall performance. Examination of the BCD system currently in use has highlighted two dif­ferent design concepts, mainly involving the location of the primary selector valves:

• Positioning of selector valves on or adjacent to standpipe penetrations. These have been taken out horizontally and housed below the pile cap floor, outside the main primary circuit.

• Positioning of selector valves within the primary circuit above the reactor core. These have been taken out horizontally through a special side-mounted penetration.

1.7.5 Typical BCD layout The Wyifa BCD system has been chosen to illustrate generally the design intent of a typical BCD system installation. Like most of the BCD systems Wyifa has three independent operating facilities, each with its own performance regarding sensitivity and scanning frequency. Only the compressor unit is common to all three BCD systems. The complete BCD system pipework arrangement can be seen in the diagrammatic arrange­ment shown in Fig 2.27. The three facilities are com­monly known as bulk, group and single.
hAwMiFE
r SSQN
2 77 -‘OURS З‘0 M:4JTES 32 SECONDS 9.8 SECONDS 3 0 SECONDS?:SECONDS ■ 4 SECONDS 4′ SECONDS ‘tSECONDS ‘ ’ SECONDS ’ SECOND	‘ 7 8 ViNUTES 1 5 4 MINUSES 2’4 M’NJTES ■ 4 MINUTES 8C SECONDS ShOPT ShOpt 32 MiNuTcS 95 UNITES Є6 SECONDS Short Short
Fig, 2.25 Relevant krypton and xenon isotopes