The fuel for magnox reactors is natural uranium clad in a magnesium/aluminium alloy (magnox) supported in a graphite core which is also the moderator and cooled by pressurised ССЬ. The early magnox reactors are contained w’ithin a mild steel pressure vessel and the later ones in a prestressed concrete pressure vessel.

Magnox melts at about 640°C and its structural strength becomes significantly reduced some 10-20°C below this. At temperatures above 640°C in dry air or CO: there is an increasing risk that magnox will ignite with a large energy release. In wet air or CO;, ignition may occur at temperatures some 10-20°C below 640°C.

Natural uranium oxidises in CO: at a very signifi­cant rate at temperatures aboe 400°C. with a large increase in volume releasing particulate uranium oxide and fission products into the circuit. In air, this re­action occurs at temperatures as low as 250°C.

Graphite oxidises exothermallv in air, producing significant heat at temperatures above 500°C. The oxidation rate, and hence heat release rate, increases rapidly with temperature. It is against these proper­ties that the fault studies for magnox reactors must be carried out.

1.4.1 Role of Health Physics Department

The Health Physics Department of a nuclear power station exists to provide a comprehensive radiological service to the station. In doing so the department fulfils the role of adviser on radiological aspects to the Station Manager and other departments and is independent of the production process. Nevertheless, the department is involved with work planning and execution and also carries out an emergency scheme function. The primary objective of the department is to ensure that the work of the station is carried out in a radiologically safe manner with respect to both on-site personnel and off-site members of the public. In enabling this primary objective to be achieved, the department is responsible for effecting a number of specialist health physics work functions:

• Routine contamination and radiation surveys.

• The measurement of radiation doses to employees.

• The assessment of radioactive discharges from site.

• Determining the amount of radioactivity in the environment before and during commissioning and through the operational life of the station.

• Ensuring that radioactive materials which are discharged from site meet the relevant packaging and transport requirements.

• The maintenance and upkeep of records in respect of each of the above.

In addition to these principal work areas, a number of other services may be provided which include asbes­tos monitoring, radiological and conventional first aid duties, noise surveying, carbon dioxide monitoring, specialist training, breathing apparatus and respiratory protection maintenance and repair, job monitoring and a laundry service.

In general the base load operating pattern of the nu­clear stations limits the opportunity to gain experience of start-up and shutdown procedures so that the op­erators may. remain familiar with possible operational malfunctions, low probability accident conditions, and abnormal operating situations. This problem is em­phasised by the increasing degree of automatic control being included, particularly on the AGR stations. The Three Mile Island incident emphasised the necessity of operating staff being continually updated with the problems of safe plant operation during recovery from various accident and malfunction situations. In order to ensure that operating staff are regularly updated, they are required to satisfactorily complete a period of revision training every two years following their initial appointment. The revision training is a one week course arranged with the purpose of ensuring that the operating procedures for safe and efficient op­eration during start-up, shutdown, fuel changing, ab­normal and accident conditions are regularly reviewed and fully understood.

Plant familiarisation

This training follows the introduction course and takes place at the station over a period of at least four to six weeks. The pattern of training will vary according to individual needs and will be determined by the station manager, taking into account the post involved and the individual’s previous experience and perfor­mance. It is regarded as an extension of the formal training programme, and to ensure the necessary con­tinuity the engineer is required to complete a written report. The report is agreed between the Nuclear Power Training Centre (NPTC) and the station. Ty­pically it may cover some part of the station plant, such as control rod systems, reactor safety systems, reactor control, or cooling pond management.

Delays in refuelling reflect on the performance of the reactor and the effect is dependent on the mag­nitude of the delay. Referring again to the channel reactivity curve, any fuel not discharged at its due irradiation will become even less reactive and will de­press the overall reactor reactivity. If the delay is continued for any length of time, it may become necessary to discharge absorber to support the falling reactivity, which will be reflected in lower than nor­mal operating temperatures. Whilst some overrun of channel irradiation is acceptable in the flattened zone, there will be a limit which cannot be exceeded. Simi­larly, inner regions of the unflattened zone may be irradiation-limited whilst the outer regions will be dwell-limited.

In either case, the imposed loss of reactivity tends to reduce power output. The resumption of refuelling at too high a rate, in an attempt to redress the imbal­ance of refuelling, results in the insertion of a block of low reactivity fuel. This may require the removal of absorber even if this has not already taken place due to the loss of reactivity by the more highly irra­diated fuel removed. At a later stage, this block of new fuel will become more reactive as the fuel rises to іь peak of reactivity and may require the re-insertion of absorber. This enforced changing pattern of reactivi­ty will be reflected in changes of control rod positions of both coarse rods and regulating rods and will be re Пес ted in the temperature distribution across the core, i. e., low reactivity zones will tend to have lower than normal temperatures and high reactivity zones to have higher than normal temperatures.

Anv significant asymmetry of the delayed fuel load­ing will serve to enhance the resultant effects and in the extreme could limit the reactor output due to an asymmetric temperature distribution depressing the overall bulk gas outlet temperature.

9.1 Reactor inspection and repairs

A condition attached to the Site Licence of a nuclear power station requires the biennial inspection of each reactor and its associated conventional plant. As well as an internal assessment of the reactor, the inspec­tion includes the maintenance and assessment of the various reactor systems. Included in these systems are such items as the burst cartridge detection equipment, control rod system, cooling and ventilation plant. In general, these items do not present unique problems of maintenance or inspection nor do they require specialised techniques or equipment to assess their condition. The outcome of the inspections, and any maintenance deemed necessary, is to provide the Nu­clear Installation Inspectorate (Nil) with an assess­ment of the continued safe and satisfactory service of the whole plant. Such assessment enables the In­spectorate to issue a ‘start-up Certificate’ allowing the continued operation of the reactor for a further two years.

The inspection and maintenance of the conventional plant follows well established electromechanical tech­niques and in many cases the examinations are more frequent than on a biennial basis. For example, the refuelling machinery, considered to be part of the reactor system, is usually subject to frequent checks of its control system to ensure that no untoward incidents may occur whilst attached to the reactor. Similarly, the reactor’s guard line trip circuitry is checked several times per year. All such external features of the reactor are readily available for routine check procedures. However, the internal items of the reactor are not so accessible and are only available during the extended period of a shutdown. It is the special requirements of the reactor internal inspection and repair that are dealt with in this section.

9.1.1 Inspection requirements

Internal reactor inspection includes all those proce­dures devised to establish the component quality and condition. In this sense metallurgical samples, quan­titative material analysis and metrology techniques are considered to be part of the inspection procedure. The procedure is required to:

• Confirm component behaviour predictions.

• Provide back-up data of material quality.

• Establish a base line of component condition.

• Monitor the condition of any installed items

resulting from maintenance procedures.

This part deals with technical administrative aspects

of the Regulations.

Schedules

The Schedules, of which there are ten, specify a num­ber of requirements in more detail than those given in the corresponding Regulations. To all intents and purposes they may be regarded as appendices to the Regulations, but still carrying the same force.

Schedule I gives the numeric values of the dose limits for employees over 18. trainees under IS and other persons, i. e., members of the public. Separate dose limits apply to each of these as described in the following table (values are in mSv);

In the case where the employee is a female of repro­ductive capacity, the dose limit to the abdomen is restricted to 13 mSv in any consecutive three month interval. In the case of a woman who is pregnant, the dose limit is 10 mSv during the declared term of the pregnancy.

Schedule 2 contains listings of all the radionuclides and gives against each, certain numeric values corre­sponding to the requirements of a number of regula­tions. Here for instance will be found the values for air and surface concentrations for designation of areas, notification, levels, etc.

Schedules 3, 4 and 5 contain the formal notification of work procedures and itemises where this need not be done.

Schedule 6 refers to the designation of controlled areas in a more detailed form than that contained in the relevant regulation. For instance, it quotes dose rate values and internal radiation levels, which are of more practical use in defining such areas.

Schedules 7 and 8 contain the detailed requirements of the assessment report, referred to in Regulation 26.

Schedule 9 exempts certain types of sealed radioac­tive sources from the requirements of Regulation 26.

Schedule 10 lists those regulations which have been revoked due to the implementation of the Regulations. As mentioned in the introduction, the Factories Act Regulations and the Road Transport Workers Regu­lations have been revoked. Certain other Regulations and Orders have been amended.

Faults which do not involve, as an initiating event, a breach in the coolant boundary are referred to as pressurised faults. However, the consequent pressure transient may be sufficient to operate the primary cir­cuit pressure relief system or, if steam generator tubes leak, may operate the secondary circuit pressure re­lief system, There is no significant radiological release unless the fault causes substantial fuel clad damage.

Fault progression differs widely with different ini­tiating events and with which additional faults are as­sumed to occur. Reactivity faults have the potential to produce increased fuel temperatures and to cause clad failure. Some of these can be sufficiently severe to cause not only the release of fission products from the fuel/clad gap, but also release some of those nor­mally trapped in the fuel matrix itself. The faults also cause heat-up of the coolant and can increase pressure sufficiently to open the pressure relief valves.

Faults in the secondary side can give rise to radio­logical release. A major steam line break causes de — pressurisation of the secondary coolant and can lead to direct release of steam to atmosphere. The rapid boiling of the secondary coolant leads to its rapid cooldown and, in turn, to rapid transfer of heat in the heat exchangers from the primary to the secondary coolant. The cooldown of the primary coolant in­creases core reactivity (coolant heat-up, the concern in most accidents, has the desirable effect of reducing reactivity — but here the situation is reversed) and, although control rods are automatically inserted, it can actually lead to temporary re-criticality of the reactor.

The extent to which the steam released to atmos­phere carries with it radioactivity depends on the amount and activity of the primary coolant which en­ters the secondary circuit due to steam generator leak­age. Isolating valves are provided in the secondary side to prevent continued release to atmosphere, al­though fault studies include assessment of the con­sequences of one isolating valve failing to close.

Anticipated transients without trip (ATWT) are also studied. ‘Anticipated transients’ are those caused by initiating faults more frequent than about 10_1 per year.

Lower frequency initiating events, coupled with fail­ure to trip control rods into the reactor, need not be studied since failure to trip is assessed to occur with a probability of 10 "6 per demand and such ATWTs would have a frequency of less than 10“7 per year, (i. e., less than 10~1 x 10-6 per year) and hence are beyond the design basis. There are seven initiating faults, the most frequent being inadvertent reactor trip (i. e., receipt of a signal demanding a trip, assumed to occur ten times each year). The most limiting in terms of the required performance of the emergency boration system is the total loss of feedwater to all steam generators. The emergency boration system, a shutdown system diverse to that of the control rods, is initiated if the control rod position monitoring equipment detects that any rod Las failed to enter the core.

The above brief description of PWR faults is in­dicative only of some of the fault studies carried out. The studies are in fact very extensive and the results have been shown to be in conformity with the design

safety criteria. This has not been achieved by chance but by redesign where shortcomings were identified or anticipated.

Magnox ponds It can be shown by calculation that a critical configuration of natural uranium fuel could not be attained in the cooling pond and therefore no criticality controls are required at the magnox stations.

AGR ponds Because it has been postulated that in certain fault/accident situations a critical mass of fuel

Fig. 4.9 Simplified flow diagram of the pond water treatment system at Hinkley Point В (AGR) power station

could be formed, criticality controls are required in respect of storage and handling of AGR fuel, the uranium of which is enriched in U-235 to 3.5 w/o.

For this reason the fuel elements are stored ver­tically in specially compartmented stainless steel skips, the walls of which contain boron as a neutron ab­sorber. In addition, all fuel is stored in pond water containing a normal concentration of 1250 ppm boron ions derived from the 0,7% boric acid dissolved in the pond water.

Precautions against criticality can therefore be summarised as:

• Boronation of pond water to 1250 ppm boron.

• Control of the location and distribution of fuel in the cooling pond.

In normal conditions both these factors are operative. The risk of breakdown of the controls through error
is minimised. All water added to the pond must be boronated and the boron concentration is continuously monitored by installed instruments which alarm if the boron level falls below a prescribed limit. In ad­dition there is manual sampling and analysis of the pond water.

Plant interlocks and standard procedures are de­signed to minimise the risk of occurrences such as the spurious introduction of elements into the ponds, the dropping of elements during handling, or collisions during skip transfer operations. Nevertheless, assess­ments have been made of the full range of accident situations which could conceivably arise in order to ensure that none of these would lead to unacceptable levels of reactivity. These include:

• Loss of pond boron.

• Presence of fuel elements or pins outside their skips, e. g., as a result of a skip being dropped or knocked over during transfer.

• Non-standard water distributions.

• Dropping of individual fuel elements during transfer

from pond receipt to in-skip storage.

• Liaison between the OSC and the Department of Energy nuclear emergency briefing room in London.

Nuclear Installations Inspectorate representative

• Assessment of the likely cause of the accident and its consequences.

• Advice to the Health and Safety Executive and government departments.

• Liaison with the Nil emergency centre.

• Advice to the OSC controller.

National Radiological Protection Board representative

• Assessment of the radiological hazards of the emer­gency.

• Advice to the Nil representatives and to the N11 emergency centre.

• Liaison with the NRPB emergency centre for the coordination of monitoring outside the area covered by the site emergency plan,

• Liaison with health physicists from the CEGB, MAFF and the Department of the Environment.

Local authority representative

• Arrangements for the welfare of members of the public including temporary accommodation, feeding and transport.

• Liaison with local authority headquarters in accor­dance with the county disaster plan.

Health Authority representative

• Arrangements for the treatment of casualties.

• Medical advice to people who have been (or think they have been) exposed to radiation.

Police liaison officer

Liaison with force headquarters for:

• Actions to protect the public including evacuation.

• Advising and reassuring the public, e. g., informa­tion on those evacuated.

• Control of access to the station site.

• Statements on behalf of the police at media briefings.

6.2 Government emergency planning

Nuclear site emergency plans together with the rele­vant local authority disaster scheme are considered adequate to deal with all foreseeable accidents. The role of central government departments in the event of an emergency at a nuclear power station would be concerned primarily with reassuring parliament, the public and the media that the emergency was being competently handled. Some central government de­partments and agencies would, however, be actively involved in a nuclear emergency and have their own emergency plans for this purpose.

A fuel loading scheme employing different enrich­ments is usfd in the AGR as a means of ‘flattening’ the radial power shape. This arranges for fuel situated near the edge of the core to produce power levels closer to that achieved by fuel in central positions, thereby making the overall shape of the power distri­bution more even. Higher enrichments are used nearer the core periphery in order to offset the effects of the greater neutron leakage. Usually two or three different radial enrichment zones in both initial and replacement charges are therefore used. The number of zones adopted varies between stations.

A useful criterion for measurement of the degree of power flattening in existence at any time is the ‘radial form factor’, defined quite simply as the ratio of peak to mean channel power over the whole re­actor. The total reactor power will always equal the sum of the individual stringer powers, but since there will be a limit on peak channel power in normal op­eration, total reactor power will depend upon the form factor and hence the constant incentive to keep this as close to unity as possible.

The concept of ‘axial form factor’, the ratio of peak to mean pow-er within a single channel, is also

commonly used. It exhibits wide variations from chan­nel to channel, since individual values are influenced by local regulating rod penetrations, which in turn influence the axial distributions of power. During the initial loading of AGR cores, two different enrich­ments are often used within the same channel, with the higher enrichments in certain top and bottom elements to provide some axial power flattening.

The in ilia l core and approach to equilibrium Initially all the fuel is fresh and unirradiated, but as soon as operation begins reactivity is reduced. Although this can be counteracted to a very small degree by withdrawing absorber (i. e., control rods), refuelling eventually becomes essential. Each of the initial charge fuel stringers is then progressively re­placed by feed fuel until, after about five years, all the initial fuel will have been replaced. The core will then consist of a mix of fuel with irradiations rang­ing from zero (i. e., fresh fuel just loaded) to the discharge limit itself (i. e., fuel about to be replaced). This signifies the attainment of ‘equilibrium’ — a phase in which a uniform spread of fuel irradiations is maintained by refuelling stringers as they reach their irradiation limit. This condition will remain for the rest of the working life of the reactor and can only be disturbed by further changes to the irradia­tion discharge limit.

Since the AGR will spend most of its life at fuel cycle equilibrium it follows that its core size, chosen by design and the refuelling strategy eventual­ly adopted, will be largely influenced by fuel cycle economics at this condition rather than at ‘start of life’ (SOL). Adequate consideration must nevertheless be given to the cost of assembling the initial charge, and accordingly certain schemes are designed to also improve fuel cycle economics at SOL. The financial outlay associated with the initial charge is usually included within the overall capital cost of the in­stallation,