Fuel pellets

• Chemical processing of imported ore concentrates by dissolving in HNO3.

• Solvent extraction as uranyl nitrate.

• Conversion through trioxide, dioxide, tetrafluoride to hexafluoride.

• UO: powder prepared from enriched uranium hexafluoride and/or purified directly from enriched uranyl nitrate solution.

• Fuel pellet manufacture from enriched UO2 powder by granulation, pressing, binder removal, sintering and grinding/operations.

• Close control on porosity, grain size and U-235 enrichment level.

Insulating pellet

The insulating pellet is produced from alumina powder by pressing, binder removal, sintering and grinding. Close control on density, thermal neutron absorption cross-section and mechanical strength.

Fuel can

• Reducing vacuum melted billet to hollows by either direct extrusion or boring.

• Tube manufacture from the hollows by combined cold reduction and cold drawing with interstage annealing.

• Cans are machined from seamless drawn tube by die box or grinding technique. [20]

End cap

• Produced by pressing or cupping from annealed strip, manufactured from vacuum melted stock by forging and hot rolling, followed by cold rolling.

* Close control of the product is maintained by checks on chemical analysis, neutron absorption cross — section, materia! cleanliness, hardness and grain size.

Extension piece

This is machined from seamless drawn tube produced by a similar route to that of the fuel can.

Fuel pin assembly

Assembly of the fuel pin takes place under carefully controlled conditions to ensure freedom from damage and contamination, all components being clean and free from grease.

The stack of fuel and insulating pellets is assem­bled and dried. The fuel section, which includes anti­stacking groove pellets, is inserted into the can which is then purged with helium. Insulating pellets and end — caps are inserted at each end of the pin to abut on the fuel stack and each endcap is resistance welded to the can to seal the fuel pin.

The endcap and can at the upper end of the pin, together with the endcap, can and extension piece at the lower end, are then argon arc fusion welded together.

Fuel pins are then pressurised at ambient tempera­ture at a pressure of 120.6 bar to reduce can-to-fuel — pellet clearance.

Straightening is carried out where necessary and the pins are then annealed, leak tested, inspected and finally decontaminated using an electrolytic deplating process.

Bottom support grids

Bottom support grids are machined from 20/25/Nb stabilised stainless steel blanks formed from bar stock. At convenient stages in manufacture, the bar ma­terial is checked for cleanliness and inclusions. Samples are taken for chemical analysis and neutron capture cross-section checks.

Guide tube

This is produced from seamless drawn 20/25/Nb or Ті stabilised stainless steel tube produced by a similar route to that of the fuel can.

Brace

Braces are fabricated by the resistance spot and fu­sion welding of formed annealed strip and rolled and pressed rims.

Locating lugs are fusion welded to the rim. The material is chemically analysed and checked for its neutron absorption cross-section, metallurgical condi­tions, grain size and tensile strength.

Sleeves

Inner and outer sleeves and retaining rings are ma­chined from extruded semi-isotropic impregnated graph­ite bar or tube stock. They are required to meet rigorous standards of acceptance in chemical, phy­sical and mechanical tests and are visually inspected to ensure freedom from cracks, extrusion faults, po­rosity and surface defects. Outer sleeves are also subjected to leakage flow and mechanical proof load testing.

Stringer components

Normal engineering procedures are used in the manu­facture of the upper stabilising brush assembly, top reflector assembly, central inertial collector assembly and bottom support and reflector assembly with spe­cial requirements relating to neutron absorption, chem­ical and metallurgical properties.

Tie bars are machined from air melted and vacuum refined nickel-chromium-molybdenum alloy nimonic PE 16 after conventional extrusion forging, rolling and drawing operations. Final heat treatment and ageing is carried out to provide the required tensile and creep strength.

The two diverse cooling routes are arbitrarily de­signated the X and Y systems.

The X system provides forced gas circulation to cool the fuel and the feed to the decay heat boilers, to remove heat from the primary circuit.

The Y system makes use of the natural gas cir­culation through the core which occurs, in the absence of forced gas circulation, when the main boilers are in service, fed by the emergency boiler feed system.

Both systems comprise four entirely independent trains, designated A to D (X and Y) in line with the associated reactor quadrants; each of these trains comprises:

• An emergency source of electrical power.

• Automatic post-trip sequencing equipment.

• Essential electrical switchboards and associated electrical distribution systems. [21]

• Controls and instrumentation.

The necessary X and Y system operations to achieve satisfactory post-trip heat removal are summarised in Fig 2.113. For pressurised faults, the two diverse X and Y systems will each provide adequate reactor cooling when acting independently. The X system, providing forced gas circulation by the main gas cir­culators operating at IS^o speed and heat rejection to the atmosphere via the decay heat boiler system, is the primary means of shutdown heat removal; re­dundancy is such that only one of the four trains is required to function correctly to provide acceptable cooling. In the unlikely event of total loss of forced gas circulation, the Y system feeding at least two main boilers by means of the emergency boiler feed system and natural gas circulation, provides the di­verse cooling route. In addition to the two diverse cooling routes, the automatic sequencing equipment will establish forced gas circulation (X) in conjunc­tion with the main boilers fed by the emergency feed system (Y) in the event of failure of the decay heat boiler system.

Following reactor depressurisation faults it is ac­ceptable to provide reactor cooling by a single system only due to the low frequency of occurrence of these events. To ensure satisfactory heat removal, the X and Y systems are used together since it is necessary to provide forced gas circulation through the core and heat removal from the primary circuit via two main boilers due to the low gas density; the main boilers are supplied with feedwater by the emergency boiler feed system.

In addition to the X and Y systems operations for reactor safety, the main boilers are fed post-trip by the starting, and standby feed pumps, to prevent the — bottom of the boilers being frequently subjected to hot gas; this feed is terminated automatically when steam, and hence gas, temperatures are sufficiently low (about 20 minutes post-trip for the spurious trip fault). To increase the reliability of this protec­tion the emergency boiler feed system is arranged to function as a back-up whilst steam temperatures are high.

The reactor coolant system (RCS) is the primary re­actor heat transport system, transferring heat from the reactor core to the steam generators. The system is based on the standard Westinghouse four-loop pressurised water reactor which produces 3425 MW of heat. The arrangement of the main components of the system is shown in Fig 2.131.

Four reactor coolant transport loops circulate hot pressurised water at 158 bar through the reactor core. There are no isolating or flow control valves in the reactor coolant loop flow path. The coolant enters the core at 293°C and exits at 325°C. The flow rate through each coolant loop is approximately 4700 kg/s. Each heat transport loop contains a steam generator and a reactor coolant pump, these components being connected to the reactor vessel by stainless steel pipe­work. Reactor coolant is circulated between the steam generators and the reactor core by the pump in each loop, transferring heat from the core to the steam generators where the heat (in turn) is passed to the reactor secondary cooling system.

Inside the_reactor vessel the core is supported by the reactor infernal structures. Control rod drive me­chanisms (CRDM) mounted on the reactor vessel closure head enable the control rods to be raised, low­ered or released into the core to control the reactor power appropriately.

A pressuriser vessel is connected to one of the exit pipes from the reactor vessel by a surge line. The pressuriser acts as a surge volume and enables the reactor coolant pressure to be controlled by electri­cal heaters and water sprays. Protection of the RCS against overpressure is provided by a system of relief valves arranged in a relief header mounted on the top of the pressuriser. Relief fluid from the pressuri­ser is fed by pipework to the pressuriser relief tank.

The principal components of the RCS are located and supported by the reactor building concrete struc­tures within the primary containment by a system of equipment supports and restraints, which allow for ther­mal movement and limited displacements under dyna­mic loadings such as those arising from an earthquake.

A number of systems are connected to the RCS for the following purposes:

Fig, 2.131 Reactor coolant system components

• To control the inventory and chemistry of the water in the circuit and specifically to adjust the concen­tration of soluble absorber (boric acid).

• To enable heat to be removed from the circuit at lower temperature and pressure, particularly during shutdowns.

• To supply emergency make-up of borated water in the unlikely event of a piping failure in the circuit, or other fault causing coolant shrinkage or loss.

The design of the major components of the RCS must be such as to contain the coolant at high pres­sure to a high degree of assurance. The reactor vessel, pressuriser and steam generators (except the tubes and divider plates which are of Inconel) are constructed of carbon steel forgings of controlled composition and manufacture. The surfaces in contact with reactor coolant are lined with stainless steel or Inconel to provide corrosion protection. The coolant pump bowls and interconnecting pipework are constructed of stain­less steel castings or forgings. The pressure containing parts are designed and manufactured in accordance with Section III of the American Society of Mechani­cal Engineers Boiler and Pressure Vessel Code which deals specifically with nuclear power plant. Additional detailed requirements for materials properties, testing, manufacturing controls and inspections have been de­veloped by the CEGB to give added confidence in component integrity, and these are applied specifically to the primary circuit for Sizewell В and subsequent PWR stations.

Each component vessel is subjected to a pressure test before installation. After installation, before commis­sioning and at prescribed intervals during operation­al life, the whole pressure boundary is subjected to further pressure tests and inspections, particularly of welds. All components are provided with externally — mounted thermal insulation constructed of layered stainless steel sheet and spacers. This is demount­able at welds to allow in-service inspections to be performed.

Following a LOCA, hydrogen is generated inside the containment and must be controlled in order to limit the risk of an explosion or fire that could challenge containment integrity or equipment operability. The principal mechanisms of hydrogen production are ra­diolysis of the coolant, reaction between the coolant and zirconium in the fuel cladding, and corrosion of metals and paints on equipment surfaces by spilt or sprayed water. Equipment which contributes to con­trol of hydrogen comprises the combustible gas control system, the containment cooling system and the con­tainment spray system.

The combustible gas control system consists of hy­drogen monitoring, hydrogen mixing and hydrogen recombiner subsystems. The monitors obtain, and mea­sure the hydrogen concentration of, suitable samples of the primary containment atmosphere to confirm that hydrogen concentrations are being kept below the lower flammability limit of about 4%.

Mixing of the atmosphere to ensure that, as far as practicable, a uniform concentration of hydrogen exists throughout the atmosphere is achieved by four hydrogen mixing fans. These fans, and those of the containment fan cooler units perform complementary duties for both post-LOCA hydrogen mixing and nor­mal environmental control purposes. Operation of the containment spray system also assists in mixing of the atmosphere and, by controlled recirculating w’ater chemistry (pH), in limiting the rate of production of hydrogen due to radiolvtic or chemical decomposition of water.

Recombination of hydrogen with oxygen in the at­mosphere is used to limit the overall concentration to an acceptably low level. Tw’o passive electric recom­biner units inside the containment operate by natural convection in heating approximately 170 mVh each of the atmosphere to more than 620°C, at which temperature free hydrogen and oxygen combine with­out catalytic action to form water vapour. The recom­biners are initiated manually about one day after a LOCA and utilise diesel-backed electrical supplies at 41? V AC.

12 References

(I] Nuclear Energy Group Symposium: Refuelling of Gas Cooled Reactors: Proc. IMechE, Vol. 183, Pan 3G: 1968/69

[2j Jervis, M, W.: Control and instrumentation of Large Nuclear Power Stations; A review of future trends: Proc. IEE, Vol. 131, Part A, No. 7: September 1984

Jervis, M. W.: Computers in CEGB Nuclear Power Stations with Special Reference to Heysham 2 AGR station: Paper щ the American Nuclear Society

Mitchie. R. E. and Neal, R.: Heysham 2/Torness power stations; ‘Control and Instrumentation Micros, Minis and

Making them Manage’: BNES ‘PROMAN 88’ Conference: July 1988

The temperature of the reactor may alter because of a change in nuclear power, coolant flow or coolant inlet temperature. Different components of the reactor will respond on different time scales; for instance, the moderator of an AGR responds on a time scale of tens of minutes to external perturbations whereas the fuei responds on a time scale of tens of seconds. When the temperature of a component changes, the balance of the neutron chain reaction will be altered and the overall reactivity of the reactor will change. The relationship between reactivity and temperature of a certain component is described by the temperature coefficient of reactivity of that component and is mea­sured in units of ‘milliNile per degree C (mN/°C) w here 1 Nile corresponds to а 1го change in the mul­tiplication factor k. (In fast reactors and water re­actors, reactivity is sometimes measured in dollars ($) where $ 1 is the reactivity required to take the re­actor from critical to prompt-critical; temperature coefficients are then expressed as $/°C or c/°C (lc = S 0.01); note, however, that miiliNiles are used here throughout.) in view of the different response times of components there are fast — and slow-acting tem­perature feedback effects. In magnox reactors these effects are identified with the fuel and moderator re­spectively. In AGRs, the dividing line between the two is less precise whilst in PWRs the time constants of both fuel and moderator temperature changes are short.

The reactivity worth invested in control rods, parti­cularly bulk rods, must be sufficient to shut down the reactor promptly in an emergency and maintain the reactor sub-critical by an adequate margin when it is shut down. The first requirement, causing the re­actor to shut down, is easy to achieve with relatively few control rods; the prompt neutrons react rapidly to give a prompt drop which reduces the neutron power to a low value, the delayed neutrons then de­cay. Figure 3.19 shows that for a rapid insertion of as little as -200 mN, equivalent to about two bulk rods, the neutron power can be reduced by a quarter in one second. Insertion of a larger number of bulk rods would give a larger prompt drop.

The second requirement however, that of main­taining it sub-critical, requires more thought as indi­cated in the following three paragraphs. The minimum shutdown margin is specified in the Operating Rules, typically 0.5 N.

The bulk rods must be evenly distributed across the reactor core. It may be thought that since the neutron flux is higher in the central region of the core (in a radial sense) than round the periphery when the reactor is at power, bulk rods should be more closely distributed in the central region than round the peri­phery. However, when the reactor is in a shutdown state with all rods inserted, such a rod distribution would give smaller shutdown margins round the peri­phery than in the central region with consequently greater risk of the reactor becoming critical in the peripheral region due to withdrawal of a few control rods in a local area for maintenance, access, etc. (such an event is known as a ‘local criticality’). Therefore the bulk rod distribution is chosen to ensure adequate shutdown margin in all regions of the core.

The core reactivity which the inserted control rods must suppress varies with a number of parameters,

mainly fuel irradiation and temperature (note that any Xe-135 present will assist with the shutdown margin). It would require a whole chapter to do justice to the subject of core reactivity, since reactor operation at power as well as at shutdown must be considered when deciding on the reactivity requirement for the control rods. Suffice to say for the purpose of this section, that these factors are taken into account at the design stage to ensure that under all conceivable conditions the rod worth is such that the reactor can be maintained shut down with an adequate margin of reactivity (with safety rods withdrawn). Allowance is made for a limited number of rods to be removed from the core for maintenance, access, etc.

Reactor internal temperatures are monitored to en­sure that the integrity of structural components is not jeopardised and graphite temperatures are not such as to lead to excessive core corrosion.

In the case of Heysham 2, thermocouples are distributed throughout the reactor internals and pene­trations to measure metal temperatures and are sum­marised below;

These monitoring measurements are routed to the main (level-1) data processing system and used to pro­duce VDU displays, printed logs and to generate alarms, as described in Volume F, Chapter 7.

Standpipe temperatures

Thermocouples are provided to monitor for abnor­mal conditions. At Heysham 2, one thermocouple is attached to the closure unit on each fuel stand­pipe to measure metal temperatures. This acts to give warning of CO2 leaks and any other abnormalities leading to high temperature which could affect pile cap components. An alarm is raised via the data processing system if any one temperature exceeds a preset level.

The pile cap heating and ventilation extraction duct is monitored to give warning of CO2 leakage in the pile cap area.

Pressure vessel temperature and strains Prestressed concrete pressure vessel (PCPV) tempera­tures and strains are monitored to ensure;

• Vessel line temperatures are not excessive, i. e., the thermal shielding and liner cooling systems are functioning satisfactorily.

• Concrete temperatures and strains are not exceeding the design limits.

In the case of Heysham 2, thermocouples are dis­tributed throughout the PCPV to measure concrete and liner temperatures summarised below:

The measurements are routed to the level-2 part of the data processing system described in Volume F, Chapter 7. They are used to:

• Raise data processing system alarms on excessive temperature.

• Produce a printed log of temperatures and alarm

states.

Strain gauges are installed in the PCPV concrete as described in Chapter 2. These are routed through the level-2 data processing system for logging purposes.

Boiler conditions

The majority of boiler instrumentation is for perfor­mance monitoring and has no safety role. Selected sets of thermocouples are for safety and plant pro­tection use within both the quadrant protection system and the post-trip sequencing equipment (PTSE).

In the case of the Heysham 2, thermocouples for monitoring are marshalled in each quadrant and routed to the level-1 data processing system. The measure­ments are used to produce VDU displays, printed logs, and to generate alarms.

Thermocouples for control are marshalled in each quadrant and routed to the control microprocessors forming part of the main computer system, described in Volume F, Chapter 7.

Thermocouples for safety and quadrant protection use are routed directly to the safety system and quad­rant protection system.

The first reactor on each station is specially in­strumented. Extra thermocouples are added for moni­toring. These are located within one quadrant on a special centre unit C2 and a special wing unit Cl.

The distribution of thermocouples is summarised as follows:

An on-load failed beam detection system is provided to give an alarm to the operator in the event of failure of a single main unit support assembly.

When selecting a material for nuclear cladding a large number of factors need to be taken into account. Compatibility with the fuel is obviously important and, amongst the physical properties, one could cite the melting point which should, ideally, be high and the neutron absorption coefficient which should be low. As can be seen from Table 1.9, the metallic elements vary greatly in these respects.

In general, metals with low cross-sections have been selected, magnesium in the magnox reactors, zirconium in water reactors, and aluminium in most low power experimental reactors. During the design stage of the (prototype) Windscale AGR, beryllium was chosen. However, the falling cost of fuel enrichment meant that low neutron cross-section became less of a prior­ity than hitherto and, when problems were encoun­tered (notably with oxidation and ductility), beryllium was abandoned in favour of a niobium stabilized stainless steel containing 20% Cr, 25% Ni. Part of the attraction of this material was the fact that the clad could be quite thin ( = 0.4 mm).

The presence of certain impurities in the clad can have an effect upon neutron absorption which is out of proportion to their concentration. However, whilst this will affect neutron economy it can have other, possibly more serious, consequences, most notably helium production within the clad (which impairs the ductility) or the formation of long-lived radioactive isotopes, which increase the problems of handling, reprocessing and disposal of irradiated fuel.

Chemical properties are other important factors in the choice of cladding material and here the main problems are ones of clad compatibility with fuel, fission products and coolant. Some progress has been made, particularly in the CANDU reactors, with im­permeable coatings to separate fuel and clad but these are not widely used. In general, so far as the clad and the fuel itself are concerned, there are few problems

this respect; this is especially so with oxide fuel

ind is normally the case with uranium metal. More

a nrnr1t though, are clad/eooiant and clad/fission importauii

product reactions. Both ot these aspects will be dis­cussed later.

The mechanical properties are the final group ot important factors to be considered. In most cases the num concerns are creep strength and ductility. Various approaches have been adopted; the magnox reactors, }or instance, have used a clad which is very weak but which is also highly ductile. The zirconium alloys used m water reactors are fairly strong at the rele — ant temperatures with rather lower ductilities than maenox. The problem is one of designing a clad with sufficient strength and ductility to accommodate fuel shape changes so that the integrity of the pin is maintained.

10 2 1 Magnesium alloys and magnox fuel

elements

As can be seen from Table 1.9, magnesium has a low thermal neutron cross-section. Since it is also com­patible with uranium it has good potential as a clad­ding alloy.

Although it has a low melting point ( = 650°C), this is not a serious problem, given that the uranium metal fuel itself cannot be operated above -660°C. Its only major disadvantages, ease of ignition and rapid oxidation above about 320°C are, to a large extent, obviated by alloying with 0.8% aluminium and 0.002-0.050% beryllium. These remain in solid solu­tion in the magnesium to produce the alloy magnox ALSO (magnesium-no-oxidation) after which the re­actors are named.

A magnox fuel can is a finned tube between 0.48 m (Berkeley) and 1.07 m (Sizewell A, Wylfa) long and 1.5- 2.0 mm thick excluding the fins. The can is sealed at both ends by welded caps. The fins take one of two forms: either helical-finned polyzonal or herringbone as shown in Fig 1.37. In all the UK reactors, apart trom the two at Berkeley, between seven and ten fuel elements are arranged in a vertical stack (Table 1.10) each element taking the weight of those about it; this produces longitudinal creep compression of the lower elements.

At Berkeley, the elements are shorter and thirteen ol these are stacked vertically in each channel; a fur­ther difference is that the self-weight of the column of iuel is taken by two vertically aligned graphite struts on each element. Running parallel to the axis of ail the different element types are splitters or lugs (Fig Le ). which act as flow separators but which also bruit the extent ot tue! element bowing (which occurs because ot longitudinal creep). Both lugs and splitters do this by acting as physical supports, contacting the >-hanne[ wall when the element bows. The splitters also serve the additional purpose of reducing the event ol longitudinal creep by being produced from a higher strength magnesium alloy, ZR55 (Table 1. 11), which consists of magnesium with a dispersion of fine zirconium hydride particles produced by hydriding the base material of magnesium/O.55% zirconium. As a result the herringbone elements, which have lugs rather than splitters, tend to bow more.

Finally, several other components (Table 1.12) of the fuel element are made from the magnesium — manganese alloy MN80, which has lower strength than ZR55 but which can be welded rather more easily. Amongst these components is the top end fitting, also known as a spring arm spider, whose purpose it is to brace the element against the channel wall, reducing flow’-induced vibration.

Mechanical properties of magnox fuel element components

Since the clad operates at peak temperatures of -0.7 Tm (Tm is the melting point in K), thermal creep is considerable. The result of this is that the weight of the fuel in the channel is borne almost entirely by the uranium bar which, in each individual element, also supports the clad. The avoidance of fuel failures is thus largely a matter of limiting the deformation of the uranium (w’hich occurs by growth or swelling) whilst also ensuring that the magnox can has suffi­cient ductility to accommodate any fuel shape changes which do occur. Under the temperature and stress conditions existing in a magnox fuel can, the principal deformation mechanism is creep by stress-directed diffusion of vacancies, so-called diffusion or Nabarro — Herring creep. Striking evidence for this mechanism has been presented by Hines et al (1973) [8].

The stress-directed diffusion of vacancies leads to magnesium atoms plating out along grain boundaries which are stressed in tension. In a precipitate hard­ened alloy such as MN80 this leads to precipitate-free zones being formed along these boundaries as shown in Fig 1.38.

The main features of this mechanism are that the creep rate depends linearly on stress and on the inverse square (or cube) of the grain size. The former leads to high ductilities and resistance to necking, whilst the latter leads to the highest ductilities being found in the finest grained material. Thus the fuel cans should contain grains of uniform size so as to ensure that creep deformation brings about stress and strain dis­tributions which are also uniform.

As already noted, the high operating temperatures for magnox clad (in terms of the melting point) result in irradiation having a negligible effect upon the me­chanical properties; the one exception to this is the ductility, which decreases. The reasons for this are not precisely known but could be caused by, amongst other things, helium production by the (n, a) reaction with 6Li, which is present as an impurity.

Since the clad has a low creep strength, shortly after pressurisation of the reactor, it creeps down onto the

Helical-finned polyzonal and herringbone fuel elements showing detail of the heat transfer surfaces

lucl to produce very good metal to metal contact. This :ub the advantage of promoting heat transfer and is nNo useful in the event of can failure, since it limits die area of uranium which is exposed to the oxidising

CO: coolant.

Tritium is mentioned separately because of its radio­logical significance and half-life of 12 years. The principal source of tritium is the reaction of the bo­ron isotope B-10 in primary circuit water:

B-10 (n, 2a) H-3

Within the fuel, ternary fission of U-235 produces tens of curies of tritium per day in a power reactor, a small proportion (0.1%) of which will diffuse through the zircaloy clad in addition to the release from the assumed failed fuel fraction. There are a number of other potential sources such as B-10, В-П, Li-7 and any residual Li-6. However, the overall situation, after taking account of source and loss terms is a
.vnical equilibrium concentration of 10“2 to 10“3 г litre after several fuel cycles and a production rate per fuel cycle of >00 Ci.

10.7 8 Added impurities

The main impurities likely to derive from ion ex — — hanoe operations in the CVCS are sodium and chlo — nde both of which will form active species by the reactions:

Na-23 (n, 7) Na-24 -» (ta 15.03 h)

Cl-37 (n, 7) Cl-38 —[12] (tj_ 37.18 m)

u here the isotopic abundance of the isotopes are Na-23 (100*70) and Cl-37 (24.2%). In addition, the concentration of chloride ion must be minimised in order to prevent stress corrosion cracking of austenitic stainless steels.

Impurity elements such as silicon (Si), aluminium (AD, calcium (Ca) and magnesium (Mg) are likely to derive from added chemicals and make-up water, and be present initially as silica (SiC>2) and the metal sulphates and silicates. It is necessary to control the level of these impurities (cation and anion) because of their tendency to form hard deposits on the high temperature surfaces particularly on the fuel clad. All these added impurities are controlled in the KCS by means of mixed-bed and cation-ion exchange plant in the CVCS.

The use of CO; as the primary coolant with its poor specific heat capacity requires careful optimisation of primary circuit, heat exchanger and turbine design conditions.

The need to circulate gas at the highest possible rate to achieve a high thermal output reduces the contact time between the fuel element surface and the gas. This restricts the attainable reactor gas outlet temperature because of the need to keep fuel element sheath temperatures below 450°C. A low reactor leav­ing temperature and the high gas circulator power associated with the need for the highest gas mass flow reduces the overall cycle efficiency. It therefore becomes necessary to arrange the steam cycle and turbine design to make the most efficient use of the reactor thermal output.

As the primary coolant is clean there is no need to allow for fouling of the heat transfer surface in the steam raising unit. Therefore a smaller temperature difference between primary and secondary coolants than would be allowed in, say, a conventional boiler, is possible. This allows a higher steam outlet tem­perature and steam pressure to be obtained which results in a higher cycle thermal efficiency and larger heat drop in the turbine. If, however, the difference is made too small, the extra cost of the larger boiler nullifies the benefit. The optimum temperature dif­ference is around 15°C.