1.1 Layout and radiological protection

In common with fossil fuel fired plant one of the
prime objectives in deciding the layout of magnox stations was to minimise capital cost without preju­dicing reliability, safety, efficiency and ease of opera­tion. For the early magnox stations perceived safety needs and, to some extent, construction needs dictated that each reactor building in a two-unit station and the turbine-generator building should be separate. This led to separate housing of many of the common re-

r services with long interconnecting routes and ‘■onsiderable ground area requirements. The aim of ^his section is to indicate the effect on station layout of radiological protection needs.

Radioloeical hazards may be categorised as follows:

• Direct radiation lrom radioactive materials.

• Inhalation of active gases and dusts.

• Inaestion of active material primarily through the food chain or from contaminated hands.

• Skin contamination.

Л number of UK statutory regulations exist to safe­guard both operators and the public. In particular, design and operation must comply with the nuclear site licence and the authorisations to discharge radio­active substances which are granted by UK govern­ment departments for each station. The CEGB Safety Rules (Radiological) were produced to ensure that stations are operated in accordance with statutory requirements and stations must be so designed that they can be operated in accordance with the rules and requirements (see Chapter 4).

Insofar as station layout is affected radiological protection needs may be considered in five categories:

(a) Plant and structures required specifically for radio­logical protection.

(b) Plant and equipment required for reactor protection.

(c) Special maintenance requirements.

(d) Active waste storage (see Chapter 4).

(d Control of personnel and protection against con­tamination.

In category (a) the most significant items are:

• Shielding to give protection against direct radiation lrom the core, irradiated fuel, active coolant and components and active waste.

• Filters tor removing active particulate matter from reactor coolant and coolant blowdown.

• Filter for removing active iodine from CO: blow­down tollowing an accident.

• ^ emulation and filtration plant to prevent the spread of airborne particulate activity arising outside the reactor and to prevent the accumulation of toxic voneerurations of CO: in accessible areas. [14]

Category (b) includes such items as emergency gen­erators and batteries, reactor protection systems, burst cartridge (or can) detection (BCD) equipment, control rod motor generators, etc.

Considering category (c), there are a number of plant and equipment items which become activated or contaminated in normal operation and which require routine maintenance. For magnox stations it was not considered necessary to provide facilities specifically for maintaining equipment which becomes highly ac­tivated in the core. However, it was considered ne­cessary to be able to maintain contaminated plant such as gas circulators, pond equipment, active liquid pumps and valves, refuelling machines, etc. For this purpose a decontamination centre and special work­shops are needed. A specially equipped radio-chemistry laboratory is needed for handling active gases and liquids and a special laundry for cleaning contaminated protective clothing.

With regard to category (d) shielded storage faci­lities are required for any parts removed from irra­diated fuel before it is despatched from the station, for discarded grabs and other fuel handling equipment and for discarded in-core instruments. Storage is re­quired also for spent desiccant from the CO: drying plant, spent resins and sludges from pond cooling and active effluent treatment plant, discarded CO: filters, shield cooling air filters and ventilation filters, con­taminated gas circulator lubricating oil, clothing, tem­porary coverings and swabs. Of a somewhat different nature is the three month store needed for the decay of spent fuel before despatch from the site.

Considering category (e), the plant must be arranged and shielding provided so that normal operations can be carried out without exceeding the maximum per­missible dose to operators or the public. Arrange­ments must be such that only authorised persons can gain access to radiation zones. Where infrequent access is required to high radiation zones these must be normally locked-off. For protection against surface and airborne contamination, potential contamination zones must be environmentally isolated from neigh­bouring areas by suitable enclosures and a ventilation system which ensures that no air can escape from the enclosure except through a filtered extract. Facilities for changing into and out of protective clothing with washing, showering and personal monitoring facilities must be provided. As in the case of radiation zones, access to potential contamination zones must be con­trolled so that only authorised persons can enter.

The requirements for control of access, the protec­tion of personnel against contamination and the pre­vention of spread of surface contamination are met by segregating the radiation and contamination zones from the remainder of the plant and by providing only one controlled point for entry and exit. At this point a main change room, with shower and monitoring equipment is located with the appropriate number of dean and dirty clothes lockers and an office for a

‘guardian’. A ‘sub-change’ room with washing facilities connected to the active drains is placed adjacent to each potential contamination area to which access is required. Where only infrequent access is needed it is sufficient to provide a space with water supply and active drains in which temporary change facilities can be erected.

Where the station layout demands the movement of activated or contaminated equipment outside a building to, for example, workshops, decontamination centre or waste stores it is necessary for the intercon­necting roads to be within the controlled area and therefore fenced off from the remainder of the site.

Hinkley Point A is one of the earlier magnox sta­tions and has separate reactor and turbine room build­ings with some of the common reactor services also in separate buildings. Figure 2.1 shows the Hinkley Point site plan. The disposition of the A station reactor and turbine room buildings can be seen together with the separate buildings for new-fuel store, emergency generators (diesels), decontamination centre, active workshop, irradiated fuel cooling pond, active waste store, cooling pond water and active liquid effluent treatment plant and other services. The station control room and the main change room are both situated at ground level in an annexe on the reactor side of the turbine hall on the station centre line. Figure 2.1 also shows the A station controlled area fence and the road system within.

Special ventilation plant is provided to ensure that any leakage of CO2 is safely dispersed from areas such as the circulator houses, boiler rooms, bypass filter and drier rooms, burst can detector (BCD) pre­cipitator and compressor rooms, CO2 safety valve, vacuum pumps and filter rooms. Filters are provided in the shield cooling air discharge and the ventilation discharge from areas such as the reactor equipment building, fuelling machine maintenance bay and maga­zine filling pit, the irradiated fuel discharge shaft and decontamination building. Figure 2.2 shows the loca­tion of the BCD precipitators, CO2 filter room, shield cooling air fans and filters, fuelling machine main­tenance bay and irradiated fuel discharge shaft.

There are four solid active waste stores:

• Pile cap mortuary holes into which discarded con­trol rods and charge chutes are lowered by the fuel machine.

• Voids between the primary and secondary shields near the fuel discharge points into which the fuel machine discharges control rod cables and connec­tors, flux flattening elements, control rod stools, grabheads, fuel element bottom support struts and thermocouple leads and ion chambers. [15]

• A store with airlock type access to prevent release of dust for fuel element splitter vanes.

1.2 Fuel

The first magnox reactors at Calder Hall started op­eration in 1956; in 1962 the CEGB commissioned the Berkeley and Bradwejl plants which represented the first of the eight Mk 1 Nuclear Power Stations to be operated.

When fuel for these reactors was first designed, little information was available on endurance, and as a result the initial fuel cycles were based on reject irradiation of 1750 MWd/t (channel average) and three years’ irradiation time.

The plants were designed for on-load refuelling by means of a system whereby individual fuel elements could be removed through standpipes which had ac­cess to groups of channels. Low burn-up limits in­volved on-load refuelling rates which were at the limit of charge machine capacity.

There was therefore an urgent need to gather in­formation on fuel endurance in order to raise these limits to provide improved plant margins and operating economics.

Although oxidation of certain steel reactor circuit components has subsequently necessitated some re­duction in output, the fuel endurance has been pro­gressively improved allowing fuel discharge limits to be raised to levels in excess of 5500 MWd/t (chan­nel average) and nine years’ dwell, with experimental quantities now exceeding 6500 MWd/t (channel aver­age) and eleven years (1986 targets).

The success of the magnox fuel can be judged against the irradiation of several million fuel elements to date with only limited failures which have caused negligible lossiof generation.

The following loads must be supplied under all cre­dible fault conditions:

• Plant identified as essential to maintain adequate reactor cooling post-trip. Whilst the full power of gas circulators is not required, some circulation is needed and pony motors are usually provided for this duty. Emergency boiler feedwater and some CW will also be required.

• Information systems required by the operator in­cluding instrumentation, computer systems, etc.

• Essential controls of the above plant and equipment.

• Emergency lighting.

4.1 Essential electrical supplies systems

Normal auxiliary supplies from the grid and station busbars may be lost in the event of a fault, and sup­plies from the turbine-generator will be lost whenever a trip occurs. Electrical equipment which must remain available post-fault is therefore supplied from an es­sential supplies system which normally forms part of
the auxiliary system, but following a loss of normal supplies is separated from it. Battery systems are used to supply loads which must be maintained in continuous operation (no break), but are very expen­sive to supply large loads. Loads such as gas circu­lation and feedwater flow which can be lost for a few minutes without endangering safety are supplied from diesel generators. In the event of loss of normal

3.3 kV supplies, the busbars are cleared and diesel generators automatically started: after a ‘short break’ of no more than two or three minutes, supplies are restored and the short-break loads are reconnected as required by the operator. These short-break sup­plies are also used to feed the battery chargers and prevent the batteries being completely discharged. Small gas turbines are used at Oldbury and Wylfa instead of diesels, as they provide a more economical way of meeting the essential loads.

The AGR fuel assembly designs for Dungeness B, Hinkley Point Bt Hunterston В, Hartlepool/Heysham l and Heysham 2/Torness reactors are developments of the basic type proven in the Windscale advanced gas cooled reactor (WAGR) prototype operation since 1963.

They comprise a plug unit and fuel stringer joined together by a central tie bar or coupling to form a

long flexible assembly.

The upper plug unit incorporates a closure unit, gag actuator mechanism, gag and shields and is described in detail in Section 6,3 of this chapter.

As originally proposed, the lower section of the first civil fuel assembly design comprised a number of 36-pin fuel elements of the floating pin type, develop­ments of prototype WAGR 9 and 18 pin configura­tions. However, early developments of this design led to the adoption of a fixed pin element arrangement in which the fuel was attached to the support grid and located radially at central and upper positions by braces mounted from the graphite sleeves which form the coolant channel (see Fig 2.72).

Both improved cluster performance and vibrational stability were achieved by fixing the fuel to the grid and reducing the inter-element gap. Increases in fuel length and reductions in inter-sleeve leakage and ab­sorber content were also possible.

Improvements in fuel pin endurance were also found to be possible when hollow fuel pellets were used in place of the earlier solid design, by assisting in the problem of accommodating fission product gases in providing increased pin voidage and eliminating the hot central core region associated with solid fuel. The use of fuel cans manufactured from 25cold-worked tubing followed by assembly annealing was also found to be beneficial.

Requirements of on load refuelling have necessi­tated the incorporation of charge/discharge and in­pile stabilising features and the use of element anti­gapping devices.

More recently a Stage 2 fuel element has been developed adopting a single thicker graphite sleeve and integrally spot-welded grid and braces, which with additional strength is more impact resistant providing prospects of full power on-load refuelling (see Fig 2.73).

The need for operational control on the radioactive contamination of boilers and ancillary reactor parts has also involved the development and installation of inertial collectors in stringers in order to filter gas — borne particular matter from the coolant.

Each station has active maintenance facilities where plus units and other reactor assemblies (such as con­trol rod assemblies) handled by the fuelling machine can be serviced. Glove boxes with appropriate shield­ing are used at the higher levels, with remote handling tools at the lowrer levels such as at the neutron scatter plug where activity is high. Some sub-assemblies such as closure units or control rod actuators may be re­moved for detail maintenance in special workshops for contaminated components.

The other facilities include a test and training faci­lity which embodies a full scale replication of the reactor charge path for use with the fuelling machine. Fuelling machine maintenance facilities include pro­vision for nose unit, grab and hoist servicing. Pro­vision is made for operations such as fuel element bottle inspection and test before re-use. Recovery equipment for visual inspection of the reactor charge path and removal of damaged components is pro­vided either as part of the fuelling machine or as a separate unit.

8 AGR post-trip heat removal systems and essential electrical supplies

Comprehensive protection systems are provided to shutdown the reactor under fault conditions (see Sec­tion 9 of this chapter), but these actions alone are insufficient to ensure that the reactor remains in a safe long term shutdown state, since heat continues to be generated in the core from the radioactive decay of fission products. This shutdown decay heat is an inevitable consequence of the nuclear fission process und it is entirely outside the control of the operators, wen after a normal controlled shutdown. In order to prevent overheating of the fuel, and the consequential damage to the reactor structure and primary pressure circuit components which could result in a release of radioactivity, it is necessary to maintain a reactor ‘-ooling function following shutdown of the reactor.

The post-trip cooling function is achieved in prin­ciple in a similar manner to that employed to cool the reactor core vvhen at power, by:

• Circulating sufficient coolant gas to transfer the fission decay heat from the fuel to the boilers.

• Providing sufficient feedwater to the boiler systems to remove this heat from the primary circuit and reject it to the environment.

Whilst cooling the reactor core, the plant must be operated in such a manner that the maximum compo­nent temperatures and stresses are maintained within design limits to avoid unacceptable plant damage. This cooling function is effected through the combined operation of a number of plant systems. Some of these are normally operational during power operation of the reactor, whilst others are provided specifically for post-trip heat removal duties, and are shutdown during normal power operation. Automatic sequencing equipment is provided to carry out the large number of operations required to terminate the normal power operational cooling system and to establish the post­trip heat removal systems in service.

The same principles of post-trip heat removal are applicable to all AGR power stations. However, the plant systems differ considerably in detail between stations and Heysham 2 is described in depth, being the latest station and thus most representative of current CEGB practice.

The reactor core is situated in the pressure vessel between the upper and lower core plates. It comprises 193 vertical fuel assemblies each containing 264 fuel rods, 24 guide thimbles and one instrumentation tube arranged in a 17 x 17 square array. The fuel assem­blies are constrained in the vertical direction by hold — down springs (which are attached to their top nozzles) impinging on the core upper plate and, in the hori­zontal direction, by the core baffle plates at the core periphery. A general view of the pressure vessel in­ternal components, including the core, appears in Fig 2.129.

During power operation, the natural water cool­ant flowing upwards through the core is kept under a pressure of about 155 bar which is sufficiently high to prevent fuel dryout with a core inlet tem­perature of 292.4°C and a coolant flow rate of 17.9 kg/s.

The fuel assembly design is shown in Fig 2.130. The 264 fuel rods are each 3.85 m long and 9.5 mm in diameter and consist of a stack of pellets of low enrichment uranium dioxide, approximately cylindri­cal in shape, clad in 0.57 mm thick zirconium alloy (Zircaioy). The fuel rods are backfilled with helium during manufacture to a pressure of about 25 bar to improve the pellet-to-clad conductance and keep fuel temperatures down. The pellet stack is held in place by a spring located in the upper plenum of the fuel rod. The fuel rods are retained by springs in the eight stainless steel grids. Six of the grids have vanes which serve to mix the coolant as it flows through the core, thereby reducing the potential for hotspots.

Each assembly contains 24 guide thimbles which are also constructed of Zircaioy and are joined to the grids and the top and bottom nozzles. They can accommodate control rods, burnable poison rods, neu­tron sources, or simply thimble plugs when they are not required for anything else. In addition, there is a central thimble which can permit the periodic in­sertion of in-core instrumentation which is used to monitor neutron flux distributions. The bulk of the structural material in the active core is constructed of Zircaioy to reduce the parasitic absorption of neutrons. This results in a reduction in the level of U-235 enrichment that would otherwise be needed to keep the reactor critical.

The reactor is designed for batch reloading at ap­proximately one yearly intervals when about one-third of the most highly irradiated fuel is discharged and replaced with fresh fuel. The neutron spectrum is such that the reduction in reactivity due to fission of the U-235 atoms and the build up of absorbing fission products are not compensated for by a fissile plutonium build-up due to U-238 neutron capture. This leads to a continuous reduction in reactivity throughout the cycle.

To enable the core to remain critical throughout the cycle, sufficient reactivity must be invested in the fuel by enrichment to, typically, 3.1% U-235 from a natural uranium enrichment of 0.71% U-235. This is mainly compensated for at the beginning of the cycle by boric acid, which is a strong neutron ab­sorber, dissolved in the coolant. It is only possible to change this boric acid concentration slowly during operation, and so rapid reactivity control is provided by control rods made up of a mixture of silver, indium and cadmium, and clad in stainless steel.

Control rods are designed to travel vertically in the guide thimbles of fuel assemblies and a cluster of 24 such rods is called a rod cluster control as­sembly (RCCA). There are 53 such RCCAs grouped together in either control or shutdown banks. Each control bank is constrained to move as a unit with a fixed overlap with respect to other control banks.

The RCCAs are designed to fall freely under gravity to insert negative reactivity very rapidly and render the core sub-critical in the event of a fault or incor­rect operation.

For reasons of safety, the core and fuel design are such as to have negative temperature feedback characteristics. In the fuel, this occurs naturally be­cause the dominant U-238 isotope of uranium has
very strong resonance absorption bands at epithermal neutron energies. These widen as the fuel is heated, due to the Doppler effect, thus increasing the neutron absorption and bringing about a reduction in reac­tivity. In the moderator, a negative coefficient is brought about by designing the core such that it is always slightly undermoderated. An increase in mod­erator temperature reduces the moderator density and

lakes the uranium-to-moderator ratio even further away from the optimum. At very high boric acid concentrations, however, it is possible for the reduc­tion in the moderator boric acid density to more than offset the reduction in moderation, resulting in a positive moderator temperature coefficient. Thus there is an upper limit to the negative reactivity which is permitted to be held by the boric acid.

Excessively high boric acid concentrations are avoid­ed by introducing burnable poison rods, which can be situated in any of the 24 thimble locations in assemblies not associated with control rods. These burnable poison rods consist of hollow borosilicate glass rods containing 4% by weight of boron (as B1O3) sheathed in stainless steel. These are designed in such a way that at the end of the fuel cycle the
boron is almost completely burnt out, thus minimising the U-235 enrichment penalty. Similarly the fuel U-235 enrichment selected is such that, at the end of the cvcle, the reactor coolant boric acid concentration is virtually zero.

The distribution of fuel assemblies in the core is arranged such that the resulting power distribution peakina factors are minimised. This necessitates the most reactive fuel being loaded in the outer annulus of the core where the neutron leakage is greatest, leading to a flat power distribution.

The first core is arranged to have 3.1% enriched fuel assemblies loaded in the outer annulus, and 2,6% and 2.1% enriched fuel assemblies arranged in a chequerboard fashion in the core interior.

After the first cycle, the fuel with the lowest re­activity (the 2.1% enriched fuel) is discharged. The 3.1% enriched fuel is moved inboard and shuffled with the 2.6% enriched fuel and fresh 3.1% enriched fuel is again loaded in the periphery. After cycle 2, and each subsequent cycle, the 2.6% enriched fuel is discharged and, again, the fuel in the periphery is shuffled inboard and replaced with fresh 3.1% en­riched fuel.

Equipment is provided to contribute directly to re­ducing the amount of radioactivity which reaches the external environment, once it has been released into the atmosphere inside the reactor building or primary containment. This equipment comprises the primary containment itself; its isolation system; the contain­ment spray svstem; and the secondary containment.

The primary containment is a massive prestressed reinforced concrete structure designed to withstand the high pressures reached following an energetic re­lease of mass and energy into its internal atmosphere, for example due to a LOCA, The basic shape is a right circular cylinder, about 46 m diameter and 42 m high, topped by a hemispherical dome. The general wall thickness is 1.3 m and the design pressure is 4.46 bar. The whole of the containment is lined inter­nally with a 6 mm thick welded carbon steel liner, which is contiguous with pre-engineered steel pene­tration assemblies that accommodate pipes and cables passing through the containment wall.

The containment isolation system ensures that all those penetrations not needed to remain open to pro­vide emergency services are closed automatically, should a rise in pressure or activity inside the containment be detected. This arrangement ensures that leakage through the containment boundary іь reduced to a very low level (less than 0.1го of the containment atmosphere during a 24 hour period) even at full de­sign pressure. Suitable leak rates tests are performed periodically to check that building integrity is being maintained and leakage is within limits.

The containment spray system is provided to assist in removing radioactivity from the primary contain­ment atmosphere so that it is no longer available for release via leakage from the containment building. This is achieved by spraying the maximum practicable proportion of the containment atmosphere with water droplets from sets of spray nozzles attached to a pair of spray headers affixed to the underside of the containment dome. Each header is supplied with re­latively cool borated water at about 800 m3/h from an external storage tank via a centrifugal pump located in the; auxiliary building basement; these pumps are started automatically when the atmospheric pressure rises above 2.4 bar.

The droplets tend to capture both particulate and volatile radioactive fission products as they fall through the atmosphere. Eventually the spray water reaches the containment floor and mixes with any water spilt from the primary or secondary circuits. It is then re­circulated as required by both the ECCS and the spray pumps drawing from the recirculation sumps. The pH of the recirculated water is adjusted by allowing it to dissolve quantities of tri-sodium phosphate ОІазРОд), normally stored as dry chemical in racks at the bot­tom of the containment; this offsets the boric acid dissolved in the water for reactivity control purposes and promotes slightly alkaline conditions, which are conducive to the continuing removal of fission pro­ducts such as elemental iodine.

The secondary containment, as its name implies, deals with any radioactive material that escapes from the primary containment. It consists of a steel framed and metalclad enclosure building, which completely covers the dome and other exposed faces of the pri­mary containment; the auxiliary building and the steam and feed cell, which are structural steel and reinforced concrete structures that adjoin the primary contain­ment; and part of the fuel building, comprising a buffer zone adjacent to the primary containment in the area of the fuel transfer tube, and connected to the enclosure building itself.

The enclosure building is partially supported by, but structurally separate from, the auxiliary building, to which it is sealed. It encloses all parts of the primary containment not surrounded by other buildings, and incorporates a housing which covers the platform outside the equipment hatch.

The auxiliary building, containing the fluid au- jliarv systems, encloses areas of the containment wall containing the majority of mechanical and electrical penetrations. Ducts and pipes penetrating the aux­iliary building wall are provided with redundant val­ues or dampers arranged to do>e automatical!;, on detection of high activity in the auxiliary building, or a safety injection signal indicative of a LOCA or SLB/FLB.

Activity leaking into the secondary containment is collected and processed by filtration equipment of at least 99% efficiency before it reaches the external environment. An emergency exhaust system is pro­vided to draw potentially-contaminated air from the auxiliary building, the steam and feed cell and the enclosure building. The action of the exhaust system creates a slight atmospheric depression inside the sec­ondary containment boundary to prevent untreated out-leakage. Redundant trains of high and moderate efficiency particulate air filters, charcoal absorber units and dehumidifiers are incorporated in the outlet duct­work, and the multiple redundant exhaust fans are powered from diesel-backed supplies. The secondary containment provisions described above are claimed to reduce the radioactive dose to members of the public from LOCA faults by at least a factor of 10.

Irradiation produces important effects on the graph­ite moderator of an AGR which could lead to the moderator being the life-limiting feature of AGRs, The integrity of the structure of an AGR core, which is essential for the maintenance of coolant How, operation of control rods and refuelling, depends on the interlocked features of the moderator bricks. The strength of these graphite keys and keyways must therefore remain adequate to accommodate the im­posed stresses throughout reactor life in normal op­eration and under seismic loads. Neutron irradiation not only affects the strength of the moderation but also produces dimensional changes which result in internal stresses. These effects are compounded with changes due to oxidation of the graphite, the oxida­tion itself being radiolytically induced.

Considering first the changes in graphite strength, the effect of neutron irradiation is in fact to strengthen the graphite by up to 50*Го over reactor life. The de­tailed changes occurring are complex, but essentially the radiation induced defects obstruct the disloca­tions in the material, thus increasing the strength. This gain in strength is however counteracted by the effects of graphite weight loss due to corrosion. Graphite does not oxidise thermally in carbon dioxide at the temperature existing in AGR moderator, about 450°C, but irradiation of the carbon dioxide produces highly oxidising species which, although short-lived, do cause graphite corrosion from within the pores of the mod­erator. The rate of oxidation is suppressed by addi­tion of corrosion inhibitors carbon monoxide and methane to the coolant, but the amount of inhibitor has to be restricted in order to avoid carbon deposi­tion on fuel pins and in the boilers. Corrosion of graphite equivalent to a weight loss of 20To reduces graphite strength by about a half.

The internal stresses in the graphite bricks arise Irom differential dimensional changes. Neutron irra­diation initially causes graphite to shrink. In the iso­tropic graphite used in AGRs this shrinkage is due to displaced atoms taking up places in the graphite pores. While the graphite is shrinking, the shrinkage is greatest near the channel wall where fast neutron dose is highest. This results in a tensile stress near the channel wall and a compressive stress at the brick edges where the kevways are located. This is a non-damaging situation. However, when the shrinkage eventual!) stops and reverses, then the stresses reverse and the critical keywav area comes under tensile stress. It is this localised stress relative to the local strength which determines whether keyways will fail. The reversal of graphite shrinkage is delayed by oxida­tion of the graphite because oxidation takes place preferentially on the walls of the pores, thus keeping pores open.

The ‘turn-around’ time is therefore dependent on coolant composition, i. e.. on the corrosion inhibitors introduced to the coolant.

The net changes which occur in stress and strength at keyways are shown in Fig 3.14 for typical AGR conditions with a modestly inhibiting coolant. The residual strength, which is the strength minus the in­ternal stress, is seen to increase initially, reduce to its initial value at around 15 full-power years and then continue to reduce quite rapidly as strength decreases and stress increases monotonically.

With the reactor at full gas flow, final trimming to full power commences. A comprehensive survey of operational parameters is carried out to ensure no abnormalities. Some temperature trimming on regu­lating rods is carried out to achieve a level tempera­ture distribution at the reactor outlet for optimum boiler performance. A detailed assessment of reactor power and temperatures is carried out to determine the margins to operating limits, temperatures can then be raised as permitted by the margins. It will be two days before the xenon concentration reaches its equi­librium level, and throughout this period the bulk rods will be inched out to maintain criticality. Tem­perature is limited to a few degrees below the normal operating limit by the deep bulk rod insertion asso­ciated with low xenon concentration, also it takes some time to achieve optimum operating conditions because of the several inter-related factors to be op­timised, so full power will not be achieved for up to two days.

Control of reactor metal temperatures is important to avoid over-stressing the pressure vessel and core support systems on a ‘shutdown’ reactor. The Operat­ing Rules state certain minimum temperatures for the vessel when pressurised. When the reactor is operat­ing there are preset temperature differentials between upper gas duct and vessel which must be maintained. Recorders in the CCR display these metal tempera­tures to aid the operator in complying with the rules. Loss of individual temperature measurements would not cause any problem.

Reactor shield concrete temperatures

The monitoring of reactor concrete temperatures is necessary to maintain the integrity of the biological shield. These temperatures tend not to change very rapidly and under normal operation remain virtually static. There is only one 12-point recorder situated in the annexe of the CCR from which the operator can detect trends and make arrangements for the neces­sary adjustments to cooling air flow via dampers. Loss of any of these measurements would not cause the operator any short term problems as adjustments rare­ly have to be made.

The neutron flux is not constant throughout the re­actor but decreases from the reactor centre outwards. As a result, the rate of irradiation of the fuel is not the same in all parts of the reactor.

To obtain the reactivity of the whole reactor, the average channel reactivities must be determined after first weighting the contribution from each channel by the square of the mean flux in that channel. A typical curve for the build up of reactivity in a virgin core is shown in Fig 3.36. It is assumed that refuelling is beina carried out continuously and at the rate re­quired by the fuel cde. Initially, the reactivity of the reactor follows a oune similar to that for an individual channel, i. e., after a small drop in the reactivitv (Samarium dip), it then rises to a maximum which is numerically lower in value than for a chan­nel. This is because some of the fuel in the core will have been replaced by new fuel which depresses the overall reactivity level. Following the peak, the re­activity falls and levels off at a value known as the equilibrium reactivity. The numerical value of this reactivity is approximately half that equivalent to the tareet irradiation at which the fuel is discharged. As the target irradiation is increased the numerical value of the equilibrium reactivity is reduced. There is therefore a limit to the target irradiation beyond which there will be insufficient reactivity in the core to sustain a chain reaction. The limiting channel aver­age target irradiation is of the order of 6000-6500 MVd/t, so that for reactivity reasons the potential of 14 000 MWd/t is not achievable.