In the event of an accident, the station personnel will be formed into a number of teams in order to carry out certain essential remedial actions. Each member of an emergency team receives periodic instruction in accordance with an approved training schedule.

The emergency team duties and responsibilities are described in the following paragraphs.

Health physics

Trained survey teams will be available for the imme­diate monitoring of radiological conditions both on and off the site. The on-site team will take measure­ments of airborne activity, CO2 concentration and radiation dose rate. The results will be communicated to the emergency health physicist who will advise the emergency controller of any precautions necessary on the site. The off-site teams, in fully equipped vehicles, will take measurements of airborne activity and radia­tion dose rate in the neighbourhood of the station. Their survey routine will be so arranged that a rapid assessment of conditions can be made in any popu­lated areas close to the site. The survey results will be passed to the emergency control centre by radio and the emergency health physicist will advise the emer­gency controller of the precautions necessary, includ­ing any need for the evacuation of members of the public.

Incident assessment

A rapid initial survey at the scene of an accident will be made by a small team lead by a shift engineer and including a health physics monitor. This team will be responsible for the initial assessment of any damage, the initial location of casualties and for the assess­ment of the radiological conditions at the scene of the accident. During the initial survey, the team leader will make frequent reports of the situation to the emergency control organisation.

Incident control

An incident control point will be established at a safe location near to the scene of the accident to control access to the affected area. The incident control point will be under the command of a shift engineer who will be assisted by personnel trained in breathing apparatus control, protective clothing procedures and contamination monitoring. All teams operating from the incident control point will be accompanied by a health physics monitor.

Fire

A trained fire team will be available on each shift. First aid

Trained personnel will be available to accompany res­cue teams and to render emergency first aid in the medical centre. Arrangements will be made to dispatch casualties to selected hospitals, if necessary.

Rescue

Shift personnel trained in appropriate techniques will report to the incident control point to rescue any casualties located by the incident assessment team.

Damage control

A team of craftsmen will be available to effect any temporary repairs. The team will be led by a shift engineer who will be assisted by mechanical and elec­trical fitters, welders, etc. The team will operate from the incident control point.

Plant operation

In addition to the staff in the main control room, personnel will be available to carry out any emergency plant operations such as the closing of valves.

Since the uranium in the magnox element is natural uranium and contains only 0.7% of the fissile U-235, criticality of the fuel cannot result from its trans­portation. Thus, the sole requirement is the protection of the dements from mechanical damage. This is achieved by packing the wrapped (brown paper and polythene) elements in preformed fibre packing such that the element is constrained from movement and provided with all-round protection. They are trans­ported in this packing in sealed metal boxes designed to hold 20 elements. The lack of criticality problems allows the storage of the boxed fuel to be relatively simple and only precautions against fire damage have to be considered. These are essentially the suitable stacking of the boxes to a maximum of six high with access to at least one side of the box. A fusible salt extinguisher as employed for metal fires is provided for fire control purposes.

New fuel is unpacked in clean conditions, the pur­pose of which is to protect the fuel from damage and contamination. The exclusion of certain materials from this area is essential. In particular, contact between the element canning and metals or alloys which could result in a low melting point magnesium alloy is rigorously avoided, e. g., lead. Personnel han­dling the fuel are provided with ‘clean condition cloth­ing’ (including linen gloves) to avoid contamination of the fuel by salts and grease during handling.

Unpacking usually occurs immediately before in­spection, the purpose of which is to eliminate elements which may give rise to problems whilst they are within the reactor. To date some 2 x 106 magnox elements have been manufactured, the number of elements giv­ing rise to reactor problems and associated with a manufacturing defect being only a few hundred and mostly from the early days of production.

Nevertheless, since the removal of failed elements from reactors at power can be costly and failure of a single element may involve the discharge of a complete channel of fuel, inspection of the fuel before charging is essential. The following typical features are checked: [37]

• Type identification (LTA, HTA, etc.).

• Freedom from grease and dirt.

• Freedom from damage of the can finning.

• Weld damage, cracks, or pin holes in end caps.

• Soundness of fittings such as splitters, braces and lugs.

A particularly vital area examined for damage is the end cap closure weld and the upstand of the can into which the-cap is screwed. If the defect is such that the fission products, which are produced when the re­actor is operating at power, can readily escape from the can, then the BCD equipment will effectively indicate the channel containing the defective fuel ele­ment. If, however, as a consequence of even appa­rently minor damage, the leak is very small or the leak path is long, it may give rise to the phenomenon of a ‘fast burst’. This results from the ingress of the coolant gas through the leak to the uranium and leads to local oxidation of the bar. The accumulation of oxide causes can-swelling and eventual rupture. A long path length allows fission products diffusing outwards to decay before leaving the can and thus evade detection. At a later stage, the fission products are suddenly released in larger quantities due to fail­ure of the can giving rise to the nomenclature of a ‘fast burst’. Normally it is only during unpacking and inspection that the fuel is handled, thereafter the fuel is remotely handled.

Optimisation studies of reactor systems covering core physics, heat transfer characteristics, gas circu­lator power and material constraints are carried out for each design concept.

The following characteristics of magnox elements contribute to the optimisation of the physical size of magnox elements and the steam cycle conditions of the associated reactor plant:

• Uranium experiences a metallurgical phase change at 663°C when the metal changes from an orthor­hombic to a tetragonal crystalline structure. This places an upper limit on the operating temperature of the centre of the bar.

• Magnox has a limiting ignition temperature of the order of 630°C and under operating conditions, including any transients resultant from postulated fault conditions, this temperature must not be exceeded.

These characteristics and other constraints result in an optimum uranium bar diameter of about 25 mm, a finned can of 60 mm diameter and associated with a channel gas outlet temperature of 400°C. The latter will determine the steam cycle conditions and efficien­cy. As a result, there is a limit to achievable steam conditions associated with magnox fuel beyond which progress is marginal.

In the CEGB, as with any organisation, it is impor­tant to know the costs of the operations being carried out, whether it be the production costs of the product or the maintenance costs of the production plant, so that strategies may be reviewed and the economics of the system examined. Details are sought of the com­position of the costs so that an indication is acquired of the areas in which costs can be reduced. Generally in industry, it is the labour and material costs of maintenance which are most easily controlled. How­ever, in the CEGB, the major cost of maintenance is represented by the down time of the reactor unit.

MANIPULATOR DRIVE FUNCTIONS

Fig. 3 59 Manipulator with seven functions of movement

When out of service, lost generation by the unit has to be made up by running less-economical plant elsewhere. This cost varies from hour to hour, day to day and from season to season. Typically, the replacement generation costs of a 250 MW(E) nuclear unit are of the order of £60 000/day; costs for a larger unit would increase pro rata. To this must be added the more familiar costs of manpower and revenue expenditure to arrive at the overall cost of maintenance. The purpose of this section is to illus­trate the magnitude of the costs involved and to indicate the relationship between the replacement costs and the additional costs. Since the replacement costs ‘■ary from station to station and the maintenance programmes are peculiar to individual stations, it is not feasible to give an overall picture. However, an example is taken from an Oldbury shutdown pro­gramme of about 80 days’ duration to fulfil the afore­mentioned object і v es.

The programme, typical of the station, consisted of a short phase of a few days during which the reactor was shut down and cooled. This phase was

followed by the maintenance period which occupied most of the programme time. On completion of main­tenance, a further few days were used to dry-out and re-start the reactor. The maintenance period was made up of three areas of activity which proceeded in parallel, of which the in-core inspection and main­tenance was one. In practice, the critical path for the piogramme could pass through either of these areas. The basic approach to analysing the programme was to calculate the effective proportion of the day used on any given task in terms of days. Due regard was paid to parallel activities so that the total re­placement generation cost of £60k per day was pro­portioned to each task. For this programme, the total replacement generation costs were £5.256 million of which £1.8 million was attributed to the reactor’s inspection and maintenance programme. Added to this was a further cost of £340k for labour, capital and revenue expenditure to give a grand total of £2.1 million.

It is immediately apparent that the replacement of generation costs far outweigh the costs of conducting the inspection and repairs; so there is always an in­centive to reduce the reactor downtime by optimising the overall programme and eliminating unessential work.

The reactor’s replacement generation cost was ana­lysed in the manner previously described to cost the seven main activities of its inspection and mainte-

nance programme. The opportunity was taken to identity w hat is designated the direct, the standpipe and the operational charges associated with each activity. These charges make up part of the costs and are de­fined as follows: [39]

volume of work involved.

• Standpipe charge — this is the cost allocated to gain access to the reactor by the removal, storage and reinstatement of standpipe closure units.

• Operational charge — this is the cost of time spent shutting down and re-starting the reactor. Some of which must be allocated to each specific task.

The distribution of these costs is illustrated by Fig 3.63 which shows that 50ro of the cost was employed on routine and special maintenance procedures at a cost of about £9Q0K. Routine maintenance is those procedures which cannot be carried out whilst the reactor is on load, such as the maintenance of stand­pipe housings, BCD isolation ahes and certain con­trol rod actuators. The special maintenance activities relate to in-reactor repair work, such as the fixing of fish plate bolts referred to earlier. The inspection and oxide investigations each absorbed about 20% (£360k) of the costs. The direct charge can only be reduced by the elimination of work done and this may not be acceptable. The standpipe charge can be reduced by carrying out as many activities as possible at the standpipes which have been opened, and this is the normal practice adopted in preparing a pro­gramme. The operational charge is maintained at a minimum by constantly reviewing the reactor shut­down and start-up procedures so that the minimum of time is involved.

The additional charges of manpower, revenue and capital expenditure (£340k) were calculated from in­formation that was readily available for each of the seven areas of interest. The manpower costs were derived from the salaries and wages of the staff em­ployed on the work. The revenue costs are the value
of material used in maintenance and the capital costs represent the value of capital equipment purchased (including research and deselopment charges) spread over a ten year period. The distribution of these costs is illustrated in Fig 3.64. It will be noted that all the special procedures, inspections and sampling tech­niques require the injection of significant capital. This is because access to the reactor is via the standpipe, which is essentially a 9 m long pipe with an effective 229 mm diameter, into a hostile ensironment.

With total costing assessed, it was possible to es­timate the cost of individual procedures and Table 3.9 illustrates some of these costs. It is worth while to note that the cost of sampling material for further chemical and physical analysis is considerable, e. g.. two bolts were trepanned from the charge pan area at a cost of £15 000 each and approximately 10 g of metal swarf from drill sampling for chemical analysis, cost a total of £26 000.

The above example serves to illustrate that the in­spection of reactors is an expensive procedure and that downtime must be kept to a minimum. Where feasible, access points are utilised for as many procedures as possible and work is carried out simultaneously at sev­eral sites. Every effort is made by the CEGB to reduce the shutdown time to a minimum by careful program­ming and monitoring the progress of work in hand.

RCUT’NE УД NTtNANCE

TV INSPECTION

OXIDE MEASUREMENT

MATERIAL ANALYSIS

OTHER

Packages have to be marked to indicate if they are Type A, Type В and fissile material packages, and the gross weight must be marked if it exceeds 50 kg.

Labelling is prescribed for packages and other con­tainers. Three label designs are specified corresponding to the three radiation categories defined in Table 4.12. In addition, the label indicates the radioactive contents, their activity and the transport index.

Road and rail vehicles, freight containers and tanks carrying radioactive material are required to bear pla­cards of a similar specified design on the sides and also, in some cases, on the ends.

2.8.1 Documentation

In addition to being responsible for marking and la­belling, the consignor is required to provide informa-

cion about the materials transported in the transport document. A declaration is required that the con­tents of the consignment are properly described and are in proper condition for transport according to the relevant regulations.

The transport documents given to the carrier must also include advice on supplementary operational re­quirements, route restrictions and emergency arrange­ments.

The cells of the lining of the intestine are continually lost through the action of stomach acids and digestive juices. Consequently after high radiation doses (5 Sv), insufficient parent cells may be able to divide to maintain the lining. Intake of fluids and nutrients is then reduced and body fluids and salts may be lost. At very high doses the results may be fatal.

The Foetus

Following fertilisation, the growing foetus contains large numbers of rapidly dividing cells. During this period of growth the foetus is very sensitive to radia­tion damage. This leads to the additional restrictions on the radiation dose which may be received by wo­men of child-bearing age.

Bone

Bone is a network of protein fibres with insoluble inorganic salts (mainly calcium phosphates) deposited in them. Certain isotopes chemically similar to cal­cium, can, when ingested, be absorbed onto bone surfaces and heavily irradiate the surrounding tissue. Such isotopes include those of strontium, radium and plutonium.

The Thyroid

The thyroid gland is small and soft and located either side of the Adam’s apple. The hormone produced by the thyroid is a compound of iodine and an amino — acid. As the thyroid is the body’s store for iodine it is particularly ‘sensitive’ to inhaled or ingested isotopes of iodine or chemically similar elements.

The Eye

New cells are continually produced in the lens of the eye. Irradiation produces cell damage or death with a clouding of the lens (cataract). The effect is of greatest significance when low energy X-rays are used and is not observed at doses of less than 5 Gy.

The Testis

The most sensitive cells in the testis are those which continually produce sperm. A single dose of 0.1 Sv to the testis may lower the sperm count for up to a year, 4 Sv may produce permanent sterility. Of greater significance is the possibility that damage to the pa­rent cells may lead to a transformation in the gene­tic information carried by the sperm cell, which may lead to spontaneous abortion of the fertilised ovum or possibly the birth of an abnormal child. Conse­quently, men who have received high doses of ra­diation, e. g., from radiotherapy, are advised against fathering children for a period dependent on the dose received.

The Ovary

Unlike the testis, the ovary does not continually pro­duce new cells. At birth it has a full complement of

undeveloped ova which gradually differentiate and are shed. A dose to the ovary of 3 Sv will result in sterilisation. The same considerations with regard to genetic damage apply as in the case of the testis ex­cept that, as all the ova are present from birth, any damage is permanent.

(a) Milk For the purpose of sampling, the area around the station is again divided into three annuli of radii 1 to 5 km, 5 to 10 km and 10 to 25 km in which are situated inner, outer and control farms respectively. Normally four farms are chosen in each category, disposed at approximately equal intervals of arc in the land sector.

Milk samples are obtained once every two weeks from each farm which is currently in production, and analysed for iodine-131 and sulphur-35. From these samples, a bulked sample (together with for­malin solution) is made up for each category over three months, and this bulked sample is ana­lysed quantitatively for strontium-90 and caesium — 137, and qualitatively for gamma-emitting nu­clides. Some bulking of samples may also take place for iodine-131 and sulphur-35. While every attempt is made to sample continually the same farms, they have sometimes to be changed because of cessation of milk production.

(b) Soil cores Ten of these are obtained from around each farm site; sites are sampled on a rota basis

such that at least one farm is sampled each year and each farm is sampled at least once every five years. Core samples are taken to a depth of 150 mm using a sampling tool about 100 mm in diameter. A bulked sample is analysed quantitative­ly for gross beta, strontium-90 and caesium-137, and qualitatively for other gamma emitters.

(c) Deposition collectors For the purpose of these samples, two concentric circles are drawn around the station, the inner having a radius of about 1 km and the outer about 10 km. Inner sites are disposed at approximately equal intervals of arc around the land sector of the inner circle, and outer sites are similarly distributed around the outer circle.

Deposition collectors consist of rectangular wire frames covered with muslin, the total area of cloth exposed being about 2000 cm2. The collectors are suspended some 3 metres above ground level. During exposure, particulate matter adheres to the muslin. The collectors are exposed in pairs at each site for a period of two months then replaced. The exposed collectors are analysed quantitative­ly for gross beta, and qualitatively for gamma emitters.

The NRPB is responsible for advising government departments and other, bodies on radiological pro­tection matters. In the event of an emergency, the NRPB would undertake the coordination of moni­toring for radioactivity outside the area covered by the site emergency plan. It would also assess the ra­diological hazard within the emergency plan area. The NRPB would liaise with the N11, government depart­ments and the CEGB, and provide advice on the ra­diological hazards.

Representatives from the NRPB would be in at­tendance at the OSC and at the N11 headquarters emergency centre.

In order to understand the achievements made re­garding the general progression towards on-load re­fuelling at maximum reactor power, it is necessary first of all to examine briefly some of the history of AGR operation. At Hinkley Point B, the earliest attempts made to handle fuel stringers with the re­actor at power were restricted to the charging of new fuel only at the beginning of the fuel cycle, when some of the vacancy stringers were exchanged with feed fuel. During subsequent premature discharge of one of these vacancy replacement stringers which was thought to contain failed fuel (i. e., pin failure(s)), it was discovered that the graphite sleeves had failed on two of the fuel elements within the stringer. A period of exhaustive investigation began and it was eventually established that the presence of undetected cracks re­sulted in weaknesses within the graphite sleeves, which were responsible for their failure during loading at high reactor power when they had been subjected to a differential pressure of some 2.5 bar. Furthermore, it had previously been discovered that the normal cooling gas flows within the charge-path could, if the irradiated stringer became stuck at a crucial position during its removal from a reactor at power, result in flow stagnation, leading to consequent overheating of the tie-bar and standpipe seals.

Taking these and many other factors into consi­deration, on-load refuelling operations were suspended until all the various safety issues could be satisfac­torily resolved. In the meantime, refuelling had to proceed in large batches (perhaps 8-12 stringers at a time, or even more) with the reactor shut down. Apart from the economic ramifications, this resulted in unwanted thermal cycling of the fuel and reactor plant, difficulties with regard to reactor optimisation and an uneven loading of the working patterns of the fuel route. There was naturally a large incentive to re-establish an on-load refuelling regime as quickly as possible and its eventual return was only made possible following major modifications to the charge machine, in which a more effective and dependable cooling system and a prompt reactor trip system based on hoist load sensing, were provided. Together, these major system modifications provided guaranteed pro­tection against the likelihood of overheating and con­sequent tie-bar failure during on-load refuelling, as well as ensuring that even if a fuel stringer were accidentally dropped, there would be no significant release of fission products to the environment. In addition, special equipment was provided in the fuel route so that all new fuel elements could be eddy current and pressure tested on site before assembly into fuel stringers, thereby enabling any sleeve de­fects, not visible by eye, to be located prior to load­ing of the fuel to the reactor.

Current practice

On-load refuelling of AGRs now takes place in a totally different climate to that which existed in the early days. New operating techniques have been de­veloped for control of reactor power during the re­fuelling operations and new legislation and procedures (i. e., Operating Rules and related instructions) have therefore followed. Current practice is to refuel the reactor at about 30% power, which produces a small differential pressure of approximately 0.5 bar across the fuel element graphite sleeves. The batch refuel­ling concept has been retained, as shown in Fig 3.52, involving usually 6 to 8 exchanges at a time, which roughly represents the refuelling needs of each reactor following a month’s operation at maximum power. In practice, operational penalties result if the refuel­ling ‘backlog’ is allowed to exceed about half-a-dozen stringers since the initial reduction in power down to the 30% condition (approximately 450 MW(Th) or 150 MW(E)) then needs to be made more slowly so that the reactor does not ‘poison out’ during the en­suing xenon transient. As each exchange is completed, power is raised to an intermediate ‘parking’ condition, currently equivalent to about 500 MW(E) and set from considerations of temperature cycling upon boiler life­time. Under normal circumstances about three ex­changes can be completed in a day.

Much work is already in hand in support of ele­vating the power at which refuelling can safely take place. It is generally considered however that the po­tential of the current design of fuel element is some-

Fig. 3.52 Typical power variation during on-load
batch refuelling

Reactor power is initially reduced from maximum
down to about 450 MW(Th) (150 MW(E)), at which
refuelling can take place. On completion of the batch,
re-establishment of maximum power can take many
hours, depending primarily upon the resulting control
rod positions.

what limited in this respect, and that significant pro­gress can only be made by using a more robust design incorporating a single thick-graphite sleeve and other improved performance features.

The requirement for licensees to obtain a site licence before the start of construction of nuclear installa­tions and the power to attach mandatory licence conditions at any time, form a powerful and effec­tive means of control for the regulatory authority. The licence is specifically granted for a particular site and, until it is issued, the Health and Safety Executive has no formal power over a particular site under the Nuclear Installations Act. It might appear, there­fore, that there is no statutory control in the initial siting, investigative, and informative design stages of a nuclear power plant. In practice, however, a licence would not be granted unless sufficient assurance had previously been given on the suitability of the pro­posed site and the installation as previously described. Therefore consultation between the CEGB and the Nil proceeds in parallel with the pre-site investigations by the CEGB, including consultation with other sta­tutory authorities who also need to obtain informal understanding in relation to a wide range of additional statutory consents at an early stage.

Such informal consultation and discussion between the CEGB and the Nil during the application stage is a necessary preview to the formal licensing procedure. As the civil nuclear power programme has been run­ning since 1955, a large amount of expertise and ex­perience has been accumulated on gas-cooled reactors by the CEGB and the Nil. If, therefore, a licence application is for a plant type which is produced fol­lowing a pattern of steady development and operational experience it is likely that this pre-licence examination would not be as extensive or detailed as it would be if the plant concerned was of a new type, or contained extensively different features from previous design. The safety report system would, of course, still be fol­lowed. If, however, as in the case of the Sizewell В proposals, the design is new it will be subjected to exhaustive, lengthy, and searching examination both by the CEGB and the NIL Once the Nil is satisfied that the proposed plant can be designed and built to the requirements for a specified nuclear site, the ap­plicant is informed that there are no objections to granting a licence. Subject to clearing other statutory obligations, the licence may be granted and the ap­plicant may then commence construction within the terms of the licence and the relevant conditions.

The licence document consists of the signed licence, authorised on behalf of the Health and Safety Exe­cutive by a Senior Officer in the Nil and includes three attached schedules. Part 1 of Schedule 1 pro­vides details of the site and Part 2 of Schedule 1 provides a brief description of the plant. The licence is granted subject to the conditions attached to it and these are contained in Schedule 2. The continuing valid consents, approvals and directions are included in Schedule 3. The Nil, acting on behalf of the Executive, provides itself with direct powers under the licence conditions of which there are three main types. These are consents, approvals and directions. The requirement for a consent prevents the licensee from carrying out a specific operation on the site unless the Nil is satisfied and gives formal agreement to its implementation. The approval procedure is invoked when the licensee is required by a licence condition to furnish its arrangements or procedure for carrying out a specific activity which requires the direct appro­val of the NIL A particular example of a direction in a. site licence is the one which requires the CEGB to provide operational records of its nuclear plant. Cer­tain directions are used very infrequently when the Nil wishes to direct the licensee to carry out a specific action it considers essential in the interests of safety. The initial licence is quite short and has only some 10 or 12 conditions. These cover the project up to the commissioning stage and include requirements on site security, laying of the foundations of the reac­tors, and bringing nuclear fuel onto the site. Require­ments on the furnishing of information on design, and evidence of supporting research and development pro­grammes in advance of specific stages of construction may also be included. Certain conditions may require the licensee to halt construction at specified points, until the Nil is satisfied with the information pro­vided. A further licence condition, which applied to previous nuclear stations, required the licensee to make arrangements for the examination and testing of speci­fied parts of the plant by suitably qualified people, and to ensure that the materials used in construction were suitable for the purpose for which they were intended and that the plant was properly constructed and installed. The arrangements which had to be approved covered the pressure vessel structures and foundations, the pressure vessel liners and penetrations and any other parts of plant designed to contain and control the cooling gas under pressure. On the concrete pressure vessel stations pre-stressing cables, anchorage devices, cooling systems, pressure vessel insulation, and installed instrumentation were also included. The li­cence issued for the Heysham 2 project, included a further requirement for quality assurance arrangements to be provided for all safety related items. This re­quirement will be extended to all future stations.

The licensee is required to set up a station com­missioning committee for the control of the plant, and to furnish the Health and Safety Executive with a commissioning programme and any documents re­lating to the commissioning of the plant. The com­missioning programme must be formally approved, and forms the means by which the nuclear reactors are taken through clearly defined and approved stages of commissioning up to full commercial operation. Each stage is defined in the programme which briefly describes the commissioning tests, associated safety features, and programme of work. At each commis­sioning stage it must be shown that the results ob­tained confirm the design safety intent and that no unexpected features have arisen which could adversely influence the safety of the plant. Following accept­ance of such evidence, approval is given by the Nil for the commissioning programme to continue to the following stage.

At some time before the first reactor becomes op­erational, which is usually when nuclear fuel is first brought to site, the original licence may either be re­voked and replaced by a new licence or the original licence may be varied by the addition of a new set of conditions. Whichever method is used the licence for the operational phase of the station comprises the following schedules:

• Schedule 1, description of the site and plant.

• Schedule 2, conditions attached to the licence.

• Schedule 3, continuing valid consents, approvals and

directions.

At this stage the licence is greatly expanded contain* ing some 33 conditions covering every safety related aspect of nuclear power station operation, including such matters as plant operating rules, operation and maintenance procedures, storage and consignment of nuclear fuel, storage and processing of radioactive waste, plant modification procedures and provision of emergency arrangements. A further most important requirement is the establishment of a safety committee which will consider any matter of relevance to the safety of the plant which may be placed before it, including modifications and additions. The safety committee is referred to in more detail later in this chapter.

Each of the CEGB’s nuclear stations has its own individual licence. In general all licences are basically the same, although there may be minor differences between specific licences. When the Nil is considering licensing changes these are discussed with the CEGB before implementation and will be included in the li­censing process, either by varying the licence through adding additional conditions, changing existing condi­tions or by re-issuing the licence.

The fundamental approach to protection against acci­dents is to identify all the faults which might threaten the safety of the plant and to provide safeguard sys­tems to detect the faults, shut down the reactor and bring it to a stable condition. The CEGB’s aim is to produce a design for which the total predicted fre­quency of all accidents resulting in large uncontrolled releases of radioactivity to the environment should be less than 10“6 per reactor year. In order to ensure that any single accident sequence does not dominate the design, a further criterion states that the predicted frequency of any single accident leading to a large uncontrolled release should be less than 10-7 per re­actor year.

The CEGB criteria for the less severe accidents is that the frequency of accidents from which doses in the range of 0.1 ERL to one ERL (an emergency reference level is taken to be 0.1 Sv (10 rem) to the

whole body or an equivalent dose to particular organs) could be expected, should not exceed 10“4 per reactor year. Smaller doses are permitted at higher frequencies as shown in the Table 4.13.

The criteria are not site specific, and care has been taken to ensure that the accidental release criteria do not generally need reference to site conditions or population densities for interpretation. Such a speci­fication could give rise to differences in protection system standards for reactors at sites with differ­ent population features. This is inconsistent with the CEGB’s aim to achieve a sensible level of standard­isation for new power stations.