Category Archives: Modern Power Station Practice

Radiation detection and measurement

In order to evaluate the radiation dose to an indi­vidual, or the radiation and contamination levels of an area, there must be an effective means of mea­suring and detecting radiation.

Radiation detection

The detection of radiation, and ultimately its mea­surement, utilises the principles of the interaction of radiation with matter, and this usually means ion­isation. Ionisation is simply the removal of electrons from the outer shells of atoms due to the interaction between radiation and the atoms. If an electric field is present, then the electrons will flow and a current is produced which can be amplified and displayed. This principle gives rise to the most common radiation devices known as gas-filled detectors which include ionisation chambers, proportional counters and Geiger- Muller counters.

A solid state equivalent of the gas ionisation de­tectors are the semiconductor detectors, usually ger­manium-lithium, silicon-lithium or cadmium sulphide. These types of detector are usually used when it is important to know the energy of the radiation.

With scintillation detectors the radiation interacts with a scintillant, e. g., zinc sulphide, and produces flashes of light which may then be used to produce an electrical output. Scintillation detectors are very often used for measuring alpha and beta radiation.

A device which combines the properties of solid state detectors and those using the scintillation principle is the thermoluminescence detector (TLD). When ir­radiated, the TLD ‘stores’ the energy produced by the radiation until it is heated. When heated this energy is released in the form of light, the amount being proportional to the initial amount of radiation. Com­mon substances used for TLD material are lithium fluoride, calcium sulphate and lithium borate. TLD detectors are mainly used for personal monitoring and are particularly useful for extremity monitoring, e. g., fingertips, when it is used in the powder form sealed into a plastic sachet.

In addition to physical changes, radiation may also produce chemical or physio-chemical changes in a material. The most important device using this prin­ciple is the film badge dosemeter, which works on the same principle as the photographic process. The radiation interacts with the film emulsion, which when developed shows a darkening proportional to the amount of radiation exposure.

Radiation and contamination monitoring The choice of technique for the measurement of an individual’s dose depends on the nature of the work to be performed, the type and level of radiation en­countered and whether contamination is present. In most operational circumstances the film badge or TLD device is used. These can discriminate, depending on the design, between thJivarious types of radiation. If the job to be performed entails handling radioactive material, or if any other body part is irradiated to a different extent, then additional devices are worn on these body parts. It is important to recognise that the film badge and TLD cannot be directly read, they must be processed in a laboratory to determine the dose received. If high radiation doses are involved then electronic dosemeters are issued, based on the — ionisation chamber principle. These have an electronic display of the accumulated dose and may also be coupled to audible alarm circuits, which set off at preselected doses or dose rates.

If there is likely to be an airborne contribution to the dose, or if it is required to demonstrate that there is no airborne hazard, then a personal air sam­pler may be issued. This is simply a filter paper con­nected to a pump which draws through air at the worker’s breathing zone. Subsequent measurement determines whether or not there has been airborne contamination present.

In addition to personnel being monitored to deter­mine their dose, areas of plant are also monitored. This, as already described, may be in the form of radiation and contamination surveys to determine the classification of the zone. In addition certain areas may have fixed instrumentation to detect the radia­tion. These may be gamma detectors, for the mea­surement of direct radiation, or fixed air samplers, where airborne contamination is present. These may be connected to alarm circuits to warn of potential or actual hazards and may even be interlocked to

such things as cranes, to prevent raising or lowering high dose rate objects, or door locks to prevent entry to (but not exit from) high dose rate areas.

CEGB policy

The CEGB Executive Minute 1198, September 1982: NOTING that the CEGB had for many years displayed a firm commitment to application and development of quality assurance, the Executive REAFFIRMED the Statement of Policy as follows:

‘It is the policy of the CEGB that for all items of power generating and transmission plant and asso­ciated systems, that there shall be in force appropriate arrangements for providing assurance of quality at all stages from design to decommissioning.’

CEGB corporate quality assurance guidance

For the current list of corporate quality assurance guidance documentation reference should be made to the Quality Assurance Policy Development Unit. Two corporate guides have been issued, which are broadly based, identify fundamental principles and provide general guidance to those responsible for QA in each phase of a power station’s life. The titles of these guides are:

CQA1 — A Guide to the QA Practices for Con­ventional Generation and Transmission Plant and Equipment.

CQA2 — A guide to the QA Practices for Nuclear Safety Related Plant and Equipment.

Additional guidance is provided by a series of CEGB Quality Memoranda (QM), some of which are issued in direct support of CQA1 and CQA2. For example, CQA1 and CQA2 provide general comments on the application of QA to the operational phase of an installation, but QM(0)1 *A Guide to the Development and Documentation of a Quality Assurance System for an Operational Location’ provides detailed guid­ance as discussed in Section 5.8.2 below.

Fuel handling — normal practice

Before describing the standard route followed by AGR fuel once it has served its useful life in the reactor, it is first necessary to give a brief account of two other fuel handling and storage facilities in­volved so that the sequence of events which takes place can be more clearly understood.

Buffer store Located beneath the floor of the central services block in the charge hall are several vertical storage tubes. Some of these are unpressurised and are used for storing new fuel stringers, plug units, control rod assemblies, etc., but a number are water — cooled ‘decay’ tubes, pressurised to about 21 bar of carbon dioxide. These are made specifically for the storage of irradiated fuel stringers, the closures of which are used to form the pressure seal. The buffer store at Hinkley Point В comprises 22 decay tubes and 20 unpressurised storage tubes.

Irradiated fuel dismantling cell In order that irra­diated fuel stringers can be dismantled and individual fuel elements despatched to the cooling ponds, the fuel route on AGR power stations is provided with an irradiated fuel dismantling cell (IFD). Situated below the floor of the central block facilities, it comprises a cell with a hoist room above it and servicing rooms below. Adjacent to the cell is a control room in which viewing of the dismantling operations is pro­vided with the aid of a zinc-bromide shielded window. An IFD hoist is located on a circular rotatable slab in the floor of the hoist room and a single vertical hole with fitted guide tube in the slab allows the passage of items in and out of the cell below. In the floor of the cell are 13 vertical tube lined holes ar­ranged in circular formation so that the guide tube in the rotatable slab in the hoist room above can, by slab rotation, be positioned above any one of them. The tubes consist of a disposal tube for despatch of waste items to debris vaults, a dismantling tube containing a platform on which the irradiated fuel stack is lowered for the actual dismantling process, two pond discharge tubes, a bottling station tube used for placing fuel elements into bottles and eight storage tubes. When irradiated fuel is installed in the cell, the bottling station, storage tubes and pond tubes are supplied with cooling air which is blown from below into the cell by fans and then drawn down the dis­mantling tube by suction pumps.

When an empty fuel channel has been reloaded with a new fuel stringer, the charge machine is moved off the reactor and usually blown down to about 21 bar pressure so that it can be connected to a buffer store decay tube (at the same pressure) enabling the irradiated stringer to be loaded into it and stored for further cooling. If conditions permit, the stringer is sometimes taken directly to the IFD for dismantling. However, it is normal practice for the machine to revisit the buffer store at a later date, picking up the stringer en route to the IFD, whereupon the machine valve is opened and the machine depressurised, but with the IFD isolated, When the pressure approaches atmospheric, suction pumps are automatically started which draw air down through the fuel stack from an air inlet in the machine, thereby maintaining cooling flow. Eventually the IFD is de-isolated, the air How is re-directed down through the cell and its dismantl­ing tube, and the stringer lowered by the charge ma­chine into the cell and onto the dismantling platform (in its fully raised position), enabling the tie-bar end fittings to be removed and disposed of to the debris vaults. The machine grab and IFD dismantling plat­form are then lowered in ‘dual’, i. e., simultaneously and at the same speed, so that the fuel stack descends into the dismantling tube. When this operation is complete the machine lifts the stringer plug unit and tie-bar away from the fuel stack, the tie-bar is cut into pieces for disposal and the plug unit removed to an active maintenance facility for essential mainte­nance prior to re-use. The IFD dismantling platform is then raised and, starting with the number eight, each element is picked up and lowered down one of the pond discharge tubes. Cooling at this time is provided by air which is fan-blown up through the discharge tube, then joining the main cooling flow being drawn down into the dismantling tube by the suction pumps; However, if the fuel is required for PIE, each element is bottled at the IFD cell bottling station, and following pressurisation with nitrogen to about 0.8 bar, each bottle is despatched to the ponds via the discharge tube in the normal way. On arrival in the pond reception tube, each bottled or unbottled element receives a 20-minute wash with recycled pond — water, following which the reception tube is stood upright and the dement or bottle lifted out by the element manipulator and placed in a skip. Following a delay to accommodate cooling and a reduction in radiation levels, a full skip is loaded to a flask for transportation to BNFL in the case of standard fuel, or the Atomic Energy Establishment at Winfrith when PIE is required.

Ionising Radiations Regulations 1985

When the United Kingdom became a member of the Council of European Communities (CEC), in addi­tion to being bound by the more commonly known Economic Community, it also became subject to the provisions of the treaty establishing the European Atomic Energy Community (Euratom). The Euratom Treaty required that basic standards be laid down by the Community for the protection of workers and the public against the dangers of ionising radiation. In June 1976 a Directive was issued by the CEC to this effect based upon the then current recommenda­tions of the International Commission on Radiological Protection (ICRP). However, in 1977 ICRP revised their recommendations and consequently the original Directive was never instigated. An amending Directive was issued in July 1980 which is implemented in the UK through the Ionising Radiations Regulations 1985, and includes the 1977 ICRP recommendations.

The enabling legislation within the UK is the Re­gulations in the Health and Safety at Work Etc. Act 1974, enforced by the Health and Safety Commission (HSC). In 1978 the HSC issued a consultative docu­ment proposing a legislative package in compliance with the original Directive. In 1979 a supplementary consultative document was issued jointly by the HSC and National Radiological Protection Board (NRPB) taking account of the then draft 1980 Directive. In order to advise on the many comments received on these consultative documents, and in order to pre­pare the legislation itself, the HSC set up a technical working party. This consisted of four members no­minated by the Trades Union Congress, four by the Confederation of British Industry and six independent experts from various areas of work with ionising radia­tion. An observer was present from the NRPB and the Chairman and Secretariat were provided by the Health and Safety Executive. The working party’s deliberations went on until June 1982 and a further consultative document was published. Comment was again received on this and final draft versions of the Regulations and Approved Code of Practice (see later) were issued in 1984. During 1985 the Regulations were given the approval of the HSC and were also subjected to ratification by the CEC for their agree­ment that the legislation adequately reflected the re­quirements of the Directive. In order to become law, the Regulations had then to be sanctioned by the Secretary of State before being enforceable from 1st January 1986.

2.7.1 Previous legislation

Prior to the Ionising Radiations Regulations, legisla­tion covering the use of, and work with, radiation and radioactive materials was fragmented and often in­consistent. The only legislation which existed was that made under the Factories Act 1961, namely, the Ionis­ing Radiations (Sealed Sources) Regulations 1969 and the Ionising Radiations (Unsealed Substances) Regu­lations 1968, and the Radioactive Substances (Road Transport Workers) (GB) Regulations 19~0, made under the Radioactive Substances Act. As the Factories Act only applies to factories, as defined, no legislation specifically dealing with ionising radiations applied to hospitals, teaching establishments or laboratories. However, there did exist codes of practice for these establishments w’hich w-еге usually implemented as if having the force of law.

At CEGB nuclear licensed sites, which are factories in the legal sense, radiological conditions of working were laid down in the site licence, issued under the terms of the Nuclear Installations Act 1965, in addi­tion to the Sealed Sources Regulations. The Unsealed Substances Regulations were not enforced on the sites.

One of the major discrepancies between the nuclear site licence conditions and Factories Act legislation was the radiation dose limit. The Factories Act re­gulations had a cumulative dose limit of 5(N-I8) rems, subject to a maximum of three rems per calendar quarter (N being the persons age in years) whereas the site licence conditions imposed the current limit of 5 rems per year, subject to a maximum of three rems per quarter. This meant that a person normally working only under the Factories Act could receive up to 12 rems per year in certain circumstances. Thus there were several instances of radiographers turning up to work at CEGB nuclear sites, who, quite legally, had more than the site licence limit of 5 rems, and thus were not able to work on the site.

With the introduction of the Ionising Radiations Regulations, legislation is now harmonised and as they were made under the Health and Safety at Work Etc. Act 1974, they apply equally at all locations where the principal Act applies, i. e., virtually everywhere where work is carried on.

Asymmetric faults

These faults give a large power increase in very few channels in a small area of the core. Temperatures do rise to some extent over the whole core, but the rise of temperature away from the fault rods is usu­ally insufficient to activate trip thermocouples. Hence only those few thermocouples in the fault region are effective. Furthermore, flux measurements are made at three discrete points on the periphery of the core set at 120° from each other. It is necessary for two of these instruments to detect the excursion before a trip will occur and, for most asymmetric faults, at best only one will be effective. The study of these faults is very extensive since large numbers of possi­ble combinations of rods being withdrawn need to be considered. In addition, the behaviour of the regulat­ing rods is most important in limiting the numbers — of thermocouples which provide effective protection. This occurs because the regulatory rods act to main­tain the temperature in their area to a predetermined value. Hence, if a single rod runs out, the regulatory rods in compensation for this may trip thermocouples and a very sharp peak in the temperature distribution across the core results.

A special case of the asymmetric faults is the problem of local criticality. If with a reactor in the shutdown condition, a number of control rods are withdrawn from a local area of the core, criticality can result. Although the reactivity of the whole core in­creases, and ultimately the core kerf becomes greater than unity, the flux distribution is very peaked in the area from where the rods have been withdrawn. Further, since the reactor is shut down, the mass flow is very low or even zero and channel gas outlet thermocouples are ineffective. Flux protection is also largely ineffective because of the distribution of the sensors. In this case reliance is placed on adminis­trative control to prevent sufficient rods being removed to allow criticality to occur. A fault study is carried out to determine the minimum number of rods needed to be withdrawn to produce local criticality; the number of adjacent rods allowed to be removed for maintenance and repair is then limited to a figure considerably below this. Typically seven or eight would be required in the most sensitive area of the core, and the allowable number would be two or three.

Main transport operations for CEGB

Irradiated magnox fuel

Approximately 500 consignments or irradiated fuel per year are made from the CEGB magnox power stations to BNFL for reprocessing. In general, the fuel loads and the flask and transport conditions are con­trolled to within limits which are specified either di­rectly or by reference in the approval certificates. They are based on a Package Design Safety Report in which an analysis is made to show that under normal transport and accident conditions, the regula­tory requirements will be satisfied.

The bulk of the fuel is consigned in Mk 2 magnox flasks (Fig 4.6), although AGR flasks are also used at WFylfa power station.

There are two design variants of the Mk 2 magnox flask, one having a body made from a single-piece forging and the other being made using full penetra­tion welds. Basically the flask is a massive steel box

Л lid-to-body seat is achieved using two compressed viion rubber rings recessed into grooves in the lid. The! id and seal area are protected by a bolted-on aluminium impact absorber. The base incorporates a heat shield comprising 25 mm of Marinite 36 and a 13 mm steel cover plate.

There are two valves, one in the flask wall to allow the water level to be controlled and one in the lid. These valves also allow the ullage space to be purged or vented in a similar fashion to the magnox flask. The water not only acts as a heat transfer medium but also as a radiation shield. The water used is pond water, i. e., demineralised water dosed with hydroxide and boric acid to maintain a specified boron concentration and pH. Boron is a thermal neutron absorber and enhances the criticality safety margins, although its presence is not essential. The standard type of fuel skip is sub-divided by boron-loaded steel plate to produce 20 compartments which allows unbottled ele­ments to be held in vertical positions during storage and transport. Twelve compartment skips are used
for bottled fuel. Holes are provided in the skip lead liners to facilitate cooling water circulation and give enhanced heat dissipation through the finning on the outer flask surface.

The minimum allowable cooling periods before des­patch are 35 days for bottled fuel and 60 days for un­bottled. However, to meet restrictions on the overall thermal load, 12 and 14 kW respectively, and to meet BNFL requirements, it is usually necessary to extend the period beyond 90 days to optimise the number of elements despatched. The use of covers on flatrols has a very small effect on the maximum allowed heat toad.

The despatch procedures are similar to those em­ployed for magnox flasks.

Liaison with external organisations

A number of central and local government depart­ments and agencies would have duties and responsibi­lities in the event of an accident at a nuclear licensed site. In particular the police and local authorities would play an essential part in the management of actions for the protection and welfare of the public. All these organisations would be notified according to a planned communications procedure and would, where neces­sary, send liaison officers to the appropriate control centre. The following organisations in particular would be involved in emergency arrangements at or near the affected site:

• Police Advice to the public, evacuation of the public and the issue of potassium iodate tablets, control of access to effected areas.

• Fire Service Fire fighting and rescue assistance to the emergency controller.

• Ambulance Service Transport of casualties from the site to hospital.

• County Welfare Service Temporary accommoda­tion for evacuees.

• Nuclear Installations Inspectorate Investigation of the circumstances of an emergency and the assess­ment of its consequences. The provision of informa­tion and advice to the Health and Safety Executive, government departments and the operator.

• Ministry of Agriculture Fisheries and Food Con­trol of the distribution of milk and agricultural products. Advice to farmers.

• Department of the Environment Ensuring supplies of potable water and for arranging the disposal of any radioactive waste arising from the incident.

• Meteorological Office The provision of frequent local weather forecasts.

Irradiated fuel


Subsequent to its discharge from the reactor, most magnox fuel is stored on site in a water-filled cooling pond prior to its despatch to the chemical processing plant of BNFL. (The exception to this ‘wet’ storage is the ‘dry’ storage area of Wylfa power station.) The elements are stored under 5-6 m of water which provides the cooling medium for heat removal and also acts as a radiation shield for the highly radio­active fuel elements. The elements are stored for a sufficient length of time to allow the fission products to decay and for the associated heat produced to dissipate and to reduce to an acceptable level for transport requirements.

The elements are accumulated on receipt at the ponds in a skip, which is an open-top box fabricated from mild steel. On discharge, the elements are com­plete with the splitter assembly or with lugs depend­ing on the type of fuel. The presence of these items results in a poor packing factor for the fuel. Later, these items are removed. Since with magnox fuel there are no criticality considerations to be taken into account, the fuel may be moved as found necessary within the pond.

Cooling pond management

The principal objectives of cooling pond management are firstly, to satisfactorily store the fuel pending its despatch off-site and secondly to ensure its safe re­moval from site including its transportation to the processing plant. These objectives are further amplified as follows:

• To preserve the fuel cladding by minimising cor­rosion and handling damage.

• To minimise radiological hazards arising within the

pond area.

• To ensure that adequate cooling is achieved before


• To control fuel movements into, within and from the

pond area.

The element canning material is a magnesium alloy and as such is a highly reactive metal which is readily corroded by water, unless a protective oxide/hydroxide film can be maintained over the whole of the can surface. This film, which is initially formed in the reactor environment, continues to grow during storage in the cooling ponds. The film may be damaged by mechanisms which are either chemical or physical in origin.

The chemical preservation of this protective film is effected by stringent chemical control of the cooling water. The film is maintained by employing a high pH level between 11.5 and ll.7.

Demineralised water is used in the ponds and a concentration of 200 mg/kg NaOH achieves the target pH level. To maintain a higher pH, say 12.5, would require a tenfold increase in the NaOH concentration, which would result in loss of pond water treatment plant availability due to increased resin regeneration times. Atmospheric carbon dioxide dissolves in the pond water and this results in a lowering of the pH by it reaction with the sodium hydroxide to produce sodium carbonate. At a pond water pH of 11.5-11.7 the sodium carbonate concentration will be in the range 100-250 mg/kg. The presence of anions, in particular chloride and sulphate, enhances the rate of corrosion of magnox. Thus anions in the pond water are controlled with a target of less then 0.5 mg/kg and an upper limit of 1.0 mg/kg.

A water treatment plant (Fig 3.43) is used to adjust the chemical conditions of the pond. Between З^о and 5% of the pond volume is passed through the plant per day. After filtration to remove particulate matter, the water is passed through a cation resin to remove the sodium ions leaving the hydroxyl ions as water and the carbonate ions as carbonic acid in solution. The latter is then scrubbed out of solution in a ‘decarbonating’ tower. The water is then passed through a mixed bed resin unit to remove all further ions. Finally, the water pH is restored to the required level by sodium hydroxide injection. It will be noted that the mixed bed resin will remove many of the ions associated with radioactive products arising in the pond. These are derived from the activation pro­ducts of the canning material and from fission pro­ducts leached from the uranium bar via leak paths produced by corrosion or mechanical damage to the canning.

It usually provides a special resin bed to cater for the removal of fission products caesium-137 and caesium-134. The resin, Lewatit DN, has the ability to perform efficiently at the pond pH levels. However, since sodium is also removed by this resin, a bed volume capacity of 12 000 to 15 000 is to be ex­pected irrespective of the caesium activity in the pond. Thus, this bed is not employed before the caesium has risen to a predetermined level and then its use is discontinued when the target operating level has been restored.

The fission product decay heat from a large number of elements can raise the pond water temperature by an appreciable amount. A maximum of 30°C is a practical limit which is reasonable to achieve with efficient coolers in the summer. Some stations have installed pond water chiller plants to maintain the water temperature at 10°C. This has the advantage of considerably reducing the rate of corrosion since chemical reactions are temperature dependent. Civil engineering requirements dictate that the rate of change of water temperature and the temperature distribution throughout the ponds shall be at a minimum and con­stant to avoid stresses in the pond structure.

Pond water invariably holds suspended particulate matter which in settling forms a sludge, the presence of which increases the possibility of corrosion. The paniculate matter is removed by filtration.

The skips in which the elements are stored are painted in a durable paint and it is essential that the paint surface is maintained in good condition. This is to avoid the possibility of enhanced corrosion of the fuel brought about by the production of galvanic couples between the bare mild steel and the cladding material.

It has been indicated that mechanical damage to the protective film can precipitate corrosion. To this end. elements are handled as little as possible even though the tools used for this task are purpose de — ‘mned. Similarly, the removal of the splitter cage by J ram and die process and the cropping of lues from lerringbone elements is delayed until the time of despatch ot that fuel off-site is imminent.

Before despatch, the fuel has to be cooled to re — ULj»e the heat burden of a road transport flask to an -^ptable lev el. Contractually, the CEGB is required

Л00′ tuel for a period of 90 days. This is re — 4mred so that the release of iodine-131 in BNFL’s 1 r{XeS4’nS plant is kept to a level suitable to their
operational requirements. Observance of this 90-day limit is obtained in the first instance by administra­tive control. This control is reinforced by the use of a device known as a ‘short cooled element moni­tor’. This device is employed at the displittering/ delugging stage of element handling. In use, the ele­ment is presented to the instrumentation which is designed to identify the lanthanum-146 at 1.6 MeV in the spectrum of energy emissions from the element. This peak is no longer identifiable after 90 days cooling when the fission products associated with the lanthanum will have decayed to a low value of activity.

Whilst a 90-day cooling period is a contractual requirement, the heat burden of a potential skip of fuel for despatch has to be below that required by transportation regulations. Whilst the heat burden varies with the type of fuel and transport flask design, it is of the order of 4.5 kW for magnox fuel.

On discharge from the reactor, the continued decay of fission products results in the production of heat on a decreasing scale. In addition, elements from dif­ferent areas of the core generate varying amounts of heat. Referring to Fig 3.44, it will be noted that at the end of 90 days cooling the heat burden of 200 elements from the flattened zone is still above the

acceptable level for transportation. Indeed, some 130-140 days cooling are required in this case. At the same time, fuel from the unflattened zone and from the edge of the core reaches a satisfactory level within 25 days. Even so this fuel would not have completed the 90-day contractural cooling period. Thus before despatch, and even before desplittering, it is necessary to compute the heat burden of the pro­posed full load. This load may be made up of fuel from various areas of the core because a skip of undesplittered/delugged fuel is 120-140 elements whilst that of fuel ready for despatch is 200-225 elements. This being so, the mixing of elements is unavoidable. To ensure that the selection of fuel for despatch is optimised, it is necessary to have a comprehensive recording and control system.

Graphite oxidation lifetime

The graphite core of the reactor consists of machined graphite bricks which are locked together with either graphite kes or zirconium pins. The integrity of this structure must be maintained, since it must withstand
the loads imposed by thermal expansion and by di­mensional changes induced by irradiation. It must also withstand the static weight of the core itself and the associated fuel which together weigh some 1000-3500 t. The carbon dioxide (CCb) coolant of the magnox and

AGR reactors reacts with the core graphite, causing a «eight loss and a consequent loss of strength. It is the loss of strength rather than the loss of mod­erating effectiveness which necessitates the control of tzraphite oxidation.

The process of oxidation may be thermal or radio — Ivtic, the detailed chemistry of the processes being discussed in Chapter I. Since the thermal reaction between the graphite and coolant is insignificant below 600°C, this reaction does not predominate in CEGB reactors. Radiolytic oxidation occurs when CO: is decomposed by ionising radiation to gie rise to re­active oxidising species (positive ions resulting from absorption of energy by the CO:), Some of these ions are able to combine with the graphite and produce carbon monoxide (CO). The oxidising species also combine with the carbon monoxide to reform carbon dioxide. Since carbon monoxide is also formed by radiolysis of the coolant gas (CO:), its concentration is allowed to build up to inhibit the action of the oxi­dising species on the graphite. There is no further increase in inhibition at CO concentrations above 1.5 vol<?o, so that this level is the target maximum for the magnox range of reactors.

Graphite oxidation is a function of radiation in­tensity and gas pressure. The criteria of 1.5 voIVo CO (maximum) was acceptable for the early magnox re­actors. However, the rate of graphite oxidation in­creases through the magnox series mainly because the coolant pressures rise from 9 bar (Berkeley) to 27 bar (Wylfa). A detailed research programme identified hy­drogen as the most suitable inhibitor for the later stations. Ingress of water from boilers and oil from gas circulators give rise to low concentrations of hydrogenous compounds and the total hydrogen equi­valent from (H2O + H2 + CH4) is used for control purposes. It should be noted that at Hinkley Point A,~ the gas circulators have air seals and hydrogen may need to be injected to maintain the required levels. The use of hydrogen brings its own problem since high concentrations accelerate the oxidation of steels. Figure 3.65 illustrates the conflict of requirements with respect to graphite and steel oxidation, i. e., low hydrogen concentrations for steel oxidation control but high levels for control of the graphite reaction. The magnox system chemistry is illustrated in block diagram form by Fig 3.66.

Each reactor presents its own particular problems in coolant chemical control. Much depends on past operating conditions and coolant compositions. Con­trol of CO and hydrogen concentrations is exercised by periodic purges of the CO:, and in some cases a catalytic recombination unit is employed for CO control to combat the uneconomic use of CO: for purging. Gas driers are used for water removal but these are generally employed to remove water from the gas circuit following a prolonged shutdown for maintenance. Those reactors in which air is admitted to the coolant circuit at these times are provided with

‘O’v’.W a’ — ‘-1.


0 20 AO 60 ac


Fig. 3.65 Radiolytic graphite oxidation rate and post — breakaway oxidation rate of mild steel as a tunc:ion of the hydrogen content of СО; Ко CO coolant gas

dry air and the circuit is kept at a positive pressure. This is to reduce the quantity of vvater absorbed by the graphite and is primarily intended to reduce the dry-out time at the subsequent start-up of the reactor.

The AGR reactors, operating at higher gas pres­sures, temperatures and flux levels, present the graph­ite core with a more hostile environment than that of the magnox reactors, resulting in the need for a high degree of oxidation inhibition.

— Radiolytic oxidation takes place mainly within the pores of the graphite rather than on exposed surfaces. This is because the oxidising species have extremely short lifetimes and are quickly eliminated within a short distance of their point of production. This fact led to the development of reactor-grade graphites (pro­duced from Gilsonite deposits in Utah, USA), with a reduced pore volume compared with that used in the magnox reactors. The oxidation rate is also dependent on the pore diameter spectrum and geometry. The graphite should preferably contain a small number of large pores rather than a large number of small pores. It should be noted that oxidation increases the pore volume so that the rate of oxidation increases with time.

As in the case of the magnox reactors, carbon monoxide is an inhibitor of the oxidation rate but is of limited value (maximum factor of 2) and has the additional disadvantage of a tendency to form carbonaceous deposits, Further inhibition of the graph­ite corrosion reaction is obtained by providing a sa­crificial carbonaceous film on the surface of the graphite, derived from the radiolytic decomposition of methane. Unfortunately, this process gives rise to carbon deposits on the fuel pins, thus impairing heat transfer.

It has been established that methane is a powerful inhibitor (1000 vpm reduces reaction rate by a factor of 20). However, due to its radiolytic destruction in-core, technical and economic problems limit the maximum concentration that can be used in practice. For example, the use of high concentrations of meth­ane requires a large plant for the production of the gas, a large drier unit to remove the water resulting from the decomposition of the methane, and a re­combination unit to remove the CO so produced. Ob­viously, there is an economic limit to the size of such

equipment and this reflects the level to which the methane concentration can be raised (Fig 3.67).

It has been indicated that the production of the oxidising species takes place within the graphite pores. To gain access to the pores, methane has to diffuse into the graphite and, as a result, marked weight-loss profiles occur within a brick structure despite the provi­sion of methane access holes (Fig 3.68).

The relationship between the coolant composition and. graphite lifetime is shown in Fig 3.69. The term ’effective weight loss’ is a parameter which describes

Fig. 3.68 Typical weight loss profile for a CAGR moderator brick

the graphite weight loss in a brick since this is not uniform throughout. The highest effective weight loss that could be tolerated around the end of design life is presently considered to be about 20%. In Fig 3.69, the corrosion contours link gas compositions of equal effective weight loss and show that high methane concentrations prolong graphite life. Since plant economics preclude the adoption of certain composi­tions, these are bounded by the ‘plant limit’ line. Compositions that are prone to produce carbonaceous deposits are defined as those occurring above the predicted ‘deposition boundary’.

The initial gas composition selected for the CEGB’s lirst AGR commissioned was well below the deposition boundary — (1% СО/130 vpm CHj). Post-irradia­tion examination of the discharged fuel confirmed that carbon deposition was absent (some deposit was pre­sent but it was identified as being derived from lu­bricating oil). It will be noted from Fig 3.69 that the selected composition was far from ideal, in that a brick life of less than 20 years would be expected.


Fig. 3.69 Relationship between the coolant composi­tion and graphite lifetime

Subsequent to this initial trial, tests were carried out on the reactors at Hinkley Point В and Hunterston В with small changes to the gas compositions such that more inhibiting conditions were achieved. These changes advanced the expected brick life to about 28 years. However, there is a necessary delay between the completion of a trial and the examination of this fuel to assess deposition of carbon.

To overcome the delay in assessment and to pro­vide a more flexible method of detecting deposition, instrumental fuel stringers were developed. These units were provided with a number of thick-walled cans (1.8 mm instead of the normal 0.38 mm) with a hole drilled in the wall to accept a 0.5 mm thermocouple. The fuel pellets used in these cans were specially en­riched to compensate for the additional steel present in these particular cans. Thus the measured can wall temperature matched that of the standard fuel can. In addition, prqansion was made to measure gas inlet and outlet temperatures together with the gas-mass flow. The use of these stringers to detect deposition, requires that the temperature can be calculated taking into account all variables except deposition. The difference between the prediction and the as-measured can tem­perature is a measure of the deposition present.

The first trial was started in October 1982 with a coolant composition of 1.5% CO/300 vpm CHj, It was predicted that there would be a 9% reduction

in the heat transfer from the lowest element within a nine month period. By November 1983, no change had been detected. Subsequent post irradiation ex­amination (PIE) of the fuel confirmed that deposits present were within the range established before the test began. The composition tested was therefore judged to be non-depositing; Fig 3,69 shows that it is strongly inhibiting and provides some 30 full power years of life.

Tests in 1984 ’85 with coolants containing l. S^o CO/415 vpm СНд and 1.2°“o СО/350 vpm CH4 were later shown by PIE of fuels to have produced signi­ficant fuel pin deposition. Pending further evaluation of reactor and research rig evidence, CO and CH4 levels have both been reduced to avoid the rush of further deposition. The present coolant composition is ITo CO/230 vpm CH4, which will provide approxi­mately 25 years of full power life.

9.2 Steel oxidation lifetime The first of the CEGB’s magnox stations was com­missioned in 1962, the design being based on infor­mation available at that time. In 1968, the results of rig work and examination of specimens removed from some of the magnox reactors became available. These showed that steel oxidation rates of in-core compo­nents were unacceptably high at the (then) current operating conditions. The minimum gas outlet tem­perature was limited to 360°C on all the CEGB re­actors (except Berkeley which remained at 355°C) to ensure that the station’s economic lifetime would be achieved. Since that time the limited gas temperatures have been raised, thereby recovering some of the con­sequential loss of power output.

Competent authority

The requirements of the regulations include the de­signation for the purpose of the regulations of a national competent authority.

Competent authority approval certification is re­quired, for example, for package designs of Type В and for packages of fissile material except for some specific exemptions; approval is also required for cer­tain shipments. Special arrangements may be made in respect of consignments which do not fully comply with the regulatory design requirements. These arrange­ments ensure that there is no relaxation in the overall standard of safety. All such special arrangements require competent authority approval. Other require­ments on the competent authority include periodic assessment to ensure that the agency’s radiation safety standards [15] are being complied with.

Consignors are required to establish quality assur­ance programmes covering all aspects of transport activities, including package design, manufacture and use, in order to ensure that the regulations are being complied with. A corresponding responsibility is placed on the competent authority in respect of compliance assurance.