When the short term transient is terminated with flow restoration after a short period of stagnation, it is
necessary to ensure that the reactor remains in a sale state. Heat is generated post-trip by fission product decay at a rate which reduces with time. When the clad temperatures are at their peak, a considerable amount of heat is lost to the graphite by radiation. The introduction of a small amount of coolant How is sufficient to bring these clad temperatures down somewhat. Graphite temperatures however continue to rise. Since the reactor pressure has reduced from 10-15 bar and since, as a result of the incident, a number of gas circulators and boilers may not be available immediately, less heat may be removed from the core as a whole than is being generated. When the clad temperatures approach the rising graphite temperatures, they too may begin to rise again. Fur­ther, if the breach cannot be sealed, air mav enter the vessel. The heat of oxidation of the graphite is trivial compared with fission product heat at 400- 500°C, but above this it becomes increasingly sig­nificant. In addition, air is a less satisfactory heat

MAGNOX IGNITION TEMP. 64D°C

Fie. 4,5 A topical short term transient

transfer medium than CO: at any given pressure because it has a much lower density.

A computer code is used capable of modelling all these characteristics as they affect a single channel in the reactor. The study is usually carried out to deter­mine the highest initial graphite temperature con­sistent with the graphite transient ultimately being controlled. Each of the probable initial steady states in terms of numbers of gas circuits available, essential plant available and reactor power is examined and a graphite temperature limit derived. In addition, the effect of possible operator action is investigated and where this action is acceptably reliable an alleviation of the graphite temperature limit may be allowed. The actions which an operator may take to control the long term transient are; to inject CO2 reducing the oxygen/graphite reaction, reduce gas inlet tem­perature by over feeding the boilers thus increasing heat removal capability, seal the breach in the vessel, bring into service extra gas circulators or increase the speed of those already running.

A wide variety of radioactive materials are transported

from nuclear power stations. The main types are samples sent for analysis, radioactive wastes sent for disposal, items of contaminated/irradiated plant re­quiring off-site testing or maintenance. In addition, radioactive sources that are used at both nuclear and conventional stations for measurement and test pur­poses such as radiography and tracer experiments.

Wastes for sea disposal were usually sent by rail to the port of shipment. Otherwise most consignments are sent by road.

Effluent samples are invariably low in activity con­tent and are sent in excepted packages, which might consist of a glass or plastic bottle surrounded by ab­sorbent material within a cardboard box. On the other hand, some samples of reactor structural materials such as steel and graphite are highly radioactive and are transported in specially designed Type A or Type В packages.

Radioactive wastes arise as a result of normal op­erations on nuclear stations and those routinely dis­posed of include used protective clothing, redundant items of equipment, incinerator ash, contaminated oil and general trash. Wastes such as pond sludges and spent ion exchange resins arise from the treatment of effluent or of cooling pond water.

The low level wastes are sent for shallow land burial. Their specific activity or surface contamination levels are such that most meet the specifications for low specific activity material (LSA) or surface contami­nated objects (SCO). They can therefore be sent in ordinary industrial drums, sometimes having a plastic bag with a liner. Some treatment of contaminated items may be required to reduce the readily dispersi­ble component of the contamination. Drums are then loaded into a reinforced freight container to give added integrity. Contaminated items too large to be drummed are also transported in a reinforced freight container. An ordinary road tanker is used to take contaminated oil to oil-fired power stations where it is burnt after being mixed with normal fuel. Oily residues which have a higher activity are absorbed on solids and sent in drums to regional incinerators.

The higher activity spent ion exchange resins are currently stored on stations but if they are to be trans­ported to disposal sites they will have to be processed into solid compact blocks, as were the pond sludges which were sent for sea disposal. These wastes require specially designed packaging that provides shielding as well as containment. As the solid compact block provides some containment, the containment provided by the packaging can be reduced. In order to make the disposal operations more cost effective, the future trends may be to have larger packages incorporating more waste. This will involve using Type В packages and having re-usable rather than disposable shielding.

The packaging required for the miscellany of other contaminated plant cannot be specified in advance as it will depend upon the activity levels and the facilities at the off-site testing establishment. After decontami­nation, the item may be suitable for transport un­packaged or in excepted packaging, i. e., a wooden box. However, packaging having a higher integrity may have to be employed.

Radiography sources mainly require Type A pack­aging, although the higher activity sources may require Type В packaging. In general the source container itself provides the packaging.

Industrial packaging, used for LSA and SCO, and Type A packaging do not require competent authority approval provided they are fissile exempt (see Sec­tion 2.8.3 of this chapter), but approvals are issued by the health and safety department to ensure a uniform standard throughout the CEGB. The approval and supporting documentation, including any test re­ports, are kept centrally and are available for inspec­tion by the competent authority.

The emergency arrangements for each nuclear licensed site are contained in two documents: an Emergency Plan, approved by the Health and Safety Executive, which specifies the essential principles and an Emergen­cy Handbook which contains comprehensive details of emergency procedures, manpower and equipment. The contents of the latter document are not subject to

formal approval but may be varied at the request of the HSE.

The emergency plan contains the definitions of var­ious accident categories ranging from a site incident standby to a full emergency alert, together with the conditions under which they would be declared and the staff who are empowered to make or cancel such declarations. The definitions of the accident categories are as follows:

• Site Incident This is a hazardous condition which is confined in its effect within the site security fence.

• Site Incident Standby This is the stage when a warning is given on the station that a potential site incident exists or is believed to be imminent.

• Site Incident Alert This is the stage when a warning is given that a site incident exists.

• Emergency This is a hazardous condition, the ef­fect of which is to cause, or is likely to cause, a radiological hazard to the public in the vicinity of the station.

• Emergency Standby This is the stage when a warn­ing is given that a potential state of emergency exists or is believed to be imminent.

• Emergency Alert This is the stage when a warning is given that an emergency exists.

Typical examples of plant conditions that would re­quire the declaration of the various categories of accident at a magnox station with a concrete pressure vessel are:

Condition

A rapid rise of all burst car — tride detection gear readings with abnormal BCDG filter activity.

A fire exists in the reactor with the pressure vessel intact.

A loss of coolant gas several times greater than the normal daily make-up occurs, together with a high level of radio­activity in the coolant gas.

Measurements indicate that a discharge of radioactive material which is likely to cause a hazard to the public has occurred.

The emergency plan also contains a summary of the basic emergency control organisation, the arrange­ments for communication and liaison with outside
organisations, and the level of airborne radioactivity at which action would be taken to evacuate members of the public from the vicinity of the station.

The Station Emergency Handbook is a detailed re­ference document for all personnel having emergency duties on or near to the site. Such details include the arrangements for mustering and roll-call, the consti­tution and operational procedures of all the various emergency teams, lists of emergency equipment and the operating instructions for that equipment, com­munications procedure and the location and staffing of the emergency control centres.

It will be apparent from our consideration of the magnox fuel cycle that it is essential to carry out the exchange of fuel on a continuous basis whilst the reactor is at power. Refuelling machinery has been designed to meet this requirement and whilst details vary from site to site the basic principles remain the same. A typical cycle of activities for a magnox station is as follows:

• New fuel will have been inspected and loaded to a transfer device. This usually takes the form of a carousel into which the fuel is loaded by hand.

• The refuelling machine is positioned over this trans­fer device and the fuel picked up element by ele­ment using the element grab and stored in suitable magazines within the machine.

• After pressurisation with carbon dioxide to reactor pressure the machine is attached to the reactor via an access point known as a standpipe. Often, a separate coupling device is required between the machine and standpipe.

• The interspace between the machine and the stand­pipe closure unit (shield plug) is pressurised with carbon dioxide to equalise the machine/reactor pressures. This procedure allows the closure device to unlock from the standpipe enabling the plug to Pe withdrawn tor >torage. In some cases, a grab ^uidinti device (charge chute) is loaded to the stand­pipe.

• Irradiated fuel is withdrawn from the selected chan­nel and stored within the machine. This is followed by loading of new fuel to the discharge channel.

• The discharee/charge of fuel is repeated as required until all selected channels have been dealt with,

• On completion of refuelling, the closure device is reloaded to the reactor and the standpipe/machine interspace depressurised allowing the shield plug to be relocked into the standpipe. The machine moves off the standpipe, is depressurised and the irra­diated fuel discharged from the machine to the storage ponds.

The more recent machines have been designed such that, in addition to carrying out the exchange of fuel, they are able to handle standpipe shield plugs, charge chutes and control rod shield plugs. These items are stored in the service turrets of the machine, A typical machine is shown in Fig 3.40.

The management of this cycle of activities is directed to achieving the following objectives:

• Ensuring that the selected channels are visited and discharged/charged as required.

• Ensuring that the irradiated fuel is safely discharged to the storage ponds.

• Ensuring that all operations are carried out with no effects on personnel or the environment.

The control of these objectives is embodied in ad­ministrative and operational procedures and in equip­ment design. In the first instance, documentation plays a major role in establishing the selection of the cor­rect type of fuel. (Reference to Section 8.4 on docu­mentation will indicate the extent to which this form oi control is employed.) Usually, during the discharge/ charge sequence, the operator has sight of the new tuel ма TV cameras and the distinguishing features ot polyzonal herringbone fuel and LTA/HTA type dі! Ierences can be identified.

The selection ol the correct channel is dependent to some degree on the design of the machine control circuits. However, each channel is provided with a bur>i cartridge detection (BCD) sample point and the behaviour of the channel signals can be used to con­orm the correct channel selection. A decrease in signal occurs as tuel is discharged or, alternatively, a signal change is induced by the insertion of a contaminated probe to the selected channel. Either method enables correct charge visitation to be confirmed. Usually the refuelling machines are provided with grab weight and height instrumentation. Observation of the erab weight-height behaviour is used to account for ele­ment movements.

All refuelling operations are conducted with close health physics monitoring of the activities for radio­logical and carbon dioxide hazards. Particular atten­tion is directed to the operation^ involving the cou­pling or uncoupling of the refuelling machine to the standpipe and during the discharge of the irradiated fuel to the ponds.

On-load refuelling has associated with it special pro­blems which influence the operation of the reactor as opposed to the handling of fuel and these must be taken into account during the refuelling process. In this context, recognition must be given to the fact that the fuelling machine when it is coupled to the reactor becomes an integral part of the pressure vessel. Also, the handling of new or irradiated fuel within the reactor is effecting a local change in the core which may result in variations of nearby temperature instrumentation.

In the first instance, the coupling of the refuelling equipment to the reactor via a standpipe is accom­plished by the equalising of a gas pressure above some form of closure device (Fig 3.41) with that of the reactor gas below that device. This process releases pins or ball catches to allow the closure device and shield plug to be withdrawn from the standpipe. Too great a pressure differential across the closure unit may result in seals being damaged or rolled, with the consequence that it may be impossible to remove the fuelling machine without a reactor shutdown and de- pressurisation. Although it is usual to carry a spare closure unit within the refuelling machine, tilting of the closure unit resulting from adverse pressure dif­ferentials or machine misalignment can result in the same situation due to the failure to withdraw the unit or re-seal it. In many cases, the reactor gas pressure is employed to maintain the react or/machine seal faces against the event of loss of operating gas pressure. With such designs, there may be operating conditions in which a lower than normal gas pressure is insuf­ficient to maintain this seal and a proper machine connection cannot be sustained, resulting in leakage of gas at pile cap level.

Fuel is handled by a grab constructed from steel components and on its insertion in a fuel channel re­presents the loading of some 13.6 kg of absorber into the core. The local temperature transients caused by the insertion and removal of the grab during re­fuelling may require special measures to be adopted to avoid excessive variations in the signals from safety line thermocouples.

The refuelling machine control circuits are designed so that any one feature is monitored by two inde­pendent means. Using the equalising of gas pressure

RETRACTABLE MAKE-UP SHIELDING

Fig. 3.40 Typical magnox fuelling machine

across the standpipe closure unit as an example, in one case the direct measurement of the pressure dif­ferential across it is to be zero and in the second instance the relative movement of the seal’s mechani­cal components is monitored by position switches which indicate that the seal is unlocked. When these two conditions are met the service hoist circuit is energised allowing the closure unit to be withdrawn.

The need to carry out maintenance on in-core items resulted from the oxidation problems in 1968. At that

time, fractured bolts were discovered within the core area of the Bradueil reactors. The bolts were identi­fied as being associated with the steel sample baskets fixed to the roof of the reactor. Equipment was pro­duced and used to torque-off remaining bolts and to remove the brackets from the reactor. It was also found necessary to reinforce the top restraint beams of the core. This was effected by the placement of new beams onto the existing beams. The beam end> were coupled by a tongue and tee-slot coupling and the whole ring structure tensioned into place. An awareness ot the oxidation problems prompted theo­retical assessments of in-core components at all the CEGB reactors. The assessments identify areas re­quiring attention, and subsequent monitoring by in­spection is used to ensure their continued satisfactory condition. Where areas are identified as being prone to failure, programmes of repair or reinforcements are initiated. The programmes have individually led to the development of an extensive range of purpose — built equipment. Most engineering bench operations have been carried out within reactors and include drill­ing, grinding, cutting and welding. There has been only a limited exchange of equipment between stations because each reactor is of unique dimensional design, necessitating the development of purpose-built equip­ment. Repair packages in themselves frequently require the provision and development of built-in viewing/ illumination devices.

It is not possible to detail all the repairs that have been carried out on the CEGB reactors, so comment is confined to a number of examples. Further details may be obtained by reference to the proceedings and symposia of the institutional bodies, in particular the Institutions of Mechanical and Nuclear Engineers and the British Nuclear Energy Society.

A repair at Bradwell necessitated the removal of- . a component from the lower end of the standpipe. The component (latch assembly) is larger than the standpipe internal diameter and this required it to be reduced in size by plasma arc cutting to facilitate its removal. Retaining bolt nuts were torqued off, the latch ring lowered to a cutting table and then reduced in size by plasma arc cutting. The debris was held within the device which was then withdrawn from the reactor for disposal of the pieces. Thirty operational — equences were required for each latch and a com­puter prompt program was used as an operational guide.

An early ‘routine’ repair operation at Oldbury re­quired the removal of the jubilee clips used to clamp thermocouple runs and their replacement by suitably designed clips. The runs were located at eight sites tn each reactor and extended over some 3 m of the mield wall. A manipulator (Fig 3.59) was designed with >een functions of movement. A variety of tools be attached to the arm of the unit so that up to 1- different operations can be carried out. There being ■wwera] hundred Jip$ in each reactor necessitated the

work being spread over a Ю-year period. Work was carried out on a 24-hour shift cycle for about 10 weeks at each shut down.

Using the same manipulator another ‘routine’ task was that of fixing a flanged assembly to the boiler shield wall (Fig 3.60). In this case a repair package was designed to drill holes through the flanges and into the wall, insert and butt weld a retaining stud and to finally nut-torque the retaining plate into position. The manipulators are powered by air motors and capable of a 36,3 kg pay load at 2 m radius.

Retaining plates in the reactor roof were predicted to fail by weld cracking (via oxide jacking). A more powerful hydraulic manipulator was build for this re­pair — 90.7 kg at 2.75 m radius (Fig 3.61). The unit handled a drilling package and a bolt loading package and this was used to drill holes and to insert and torque-up the bolts. Again, a large number of bolts were involved and located around the periphery of the reactor roof thus requiring the work to be extended over a number of years.

Inevitably, welding repairs will be required within the core space of reactors, and work on a remotely controlled welding process has developed to the stage where extensive repairs have already been completed. The system known as ‘Warrior’ (Welding And Repair Robot In Oldbury Reactor) consists of a heavy duty manipulator known as the ‘serving manipulator’. This handles a six movement of freedom ‘work performing manipulator’. The latter carries a MIG welding head incorporating a laser ranging system (Fig 3.62).

The general aim of the Regulations is to provide ade­quate standards of safety to protect people, property and the environment from the hazards of radioactive material. It is also desirable that packages should be moved with the minimum of delay and that require­ments for special actions by the carrier are minimised.

The Regulations are written in such a form that they can be incorporated into national legislation. They prescribe what has to be done, rather than how or why it is done, the latter aspects being subjects for sup­porting documents [13,19].

As far as possible, the required protection is af­forded by package design and contents control, so that in general the carriers’ responsibilities are limited to segregating and limiting the numbers of packages on conveyances and during in-transit storage.

Adequate protection has to be provided against external radiation from packages, and human intake of any radioactive material on or escaping from the package under both normal and accident conditions. In addition to prescribed limits on dose and dose rate [14], the general principle that radiation exposures shall be kept as low as reasonably achievable (ALARA) applies.

l or packages containing fissile materials, precau­tions are required against criticality, with exemption in a few cases under specified conditions.

Arrangements must be made for adequate dissipa­tion ot heat from packages.

The human body can be considered as a collection of interrelated parts or organs all made up of spe­cialised cells. It is possible to identify those organs which will be of most significance when considering radiation damage, usually because of the higher radio­sensitivity of rapidly dividing cells. It is worth con­sidering the general effect of radiation on a number of body components.

The Blood

The two main cellular components of blood are red cells and white cells. Red cells carry oxygen to and carbon dioxide from other cells in the body. They are produced from specialised parent cells in bone mar­row and, as they have no nucleus, are unable to re­plicate by division. There is a continuous renewal of red blood cells as they only have a lifetime of about four months.

White blood cells are of several types. They are produced in bone marrow, the spleen and lymph nodes and are concerned with fighting infection in the body.

A third type of cell, platelets, is concerned with the clotting of the blood.

The effect of a high radiation dose to the blood — forming organs is to induce cell death in the spe­cialised parent cells which prevents the formation of new blood cells. This has serious consequences, both for the transport of oxygen and carbon dioxide and for the body’s resistance to infection and its ability to heal wounds.

There must be means of confirming that disposals of radioactive waste do not result in members of the public receiving radiation doses in excess of the ICRP recommended limit. It is rarely practicable to measure doses to individuals by direct means and in most cases they can only be estimated from the in­formation gained through environmental monitoring programmes. Furthermore, information must be avail­able from which collective doses to populations can be estimated; again environmental monitoring pro­grammes are a source of such data.

Before an authorisation to dispose of radioactive waste to the environment is granted by the Authorising Government Departments (MAFF and DoE in England or Welsh Office in Wales), an assessment is made of the environmental pathways by which the radionuclides may be returned to man. Environmental monitoring is needed to confirm the assessment made in arriving at the authorised disposal limits and to confirm that the control objectives are being met.

In order to demonstrate compliance with the terms of authorisations granted to dispose of radioactive waste, the responsible Minister is empowered to direct the operator to take and analyse samples of the waste and environmental materials. Results of this monitor­ing are reported to the authorising departments, whose representatives check periodically on the accuracy and effectiveness of such monitoring. It is thus seen that monitoring is carried out in two separate stages, the sampling of disposable waste (discharge monitoring) and the sampling of environmental materials (environ­mental monitoring).

Environmental monitoring must begin before any new major installation comes into operation. This makes it possible to determine the normal level of background radiation in the area, which includes nat­ural radiation, and to establish and rehearse proce­dures for when the establishment becomes operational. Measurements taken when the installation is opera­tional may be compared with pre-operational mea­surements and, where necessary, an assessment may be made of the contribution from radioactive waste discharges.

The purposes of environmental monitoring may be defined as follows:

(a) To assess the radiation exposure of the public This is the fundamental objective of environmen­tal monitoring for radioactivity. In most cases the direct measurement of radiation exposure of the public is not practicable. Instead exposure must be assessed from environmental measurements used in association with data on the habits of the po­pulation concerned.

(b) To confirm control measures The sampling and analysis of radioactive waste discharges serve as a check on the discharge control system. Environ­mental monitoring provides a further check and may result in the detection of releases that might otherwise have gone unnoticed. Monitoring for this purpose may include sampling indicator materials such as fresh water mosses and inedible seaweeds which have no radiological significance. Such monitoring is particularly valuable in the early stages of development of a nuclear facility when discharges are too low to be measurable in those pathways contributing to public radiation expo­sure. It can also give an early warning of changes which may subsequently have a bearing on public radiation exposure.

(c) To contribute to research Monitoring carried out for purposes of the control system — items (a) and (b) — may add to the overall state of knowledge of the behaviour of radionuclides in the environment. It is however, sometimes ne­cessary to improve that state of knowledge by carrying out more extensive studies on a research basis. In particular, at such time as an authorised discharge has resulted in measurable radioactivity in the environment, it is usual to engage in a comprehensive monitoring programme in order to re-assess the relationship between discharges and levels in the environment. Measurements are made embracing the pathways which lead to public exposure. Some of the pathways will be found to be critical.

(d) To provide public information and assurance It is important to provide public information and assurance on the safety discharges from nuclear establishments. Monitoring, based strictly on those discharges and pathways which give rise to envi­ronmental radioactivity leading to a significant level of public radiation exposure, may not on its own be sufficient to satisfy the public of the safety of discharges. It is quite often necessary to engage in monitoring which equivocally demon­strates the safety of discharges, even when such monitoring shows that measured levels are in all senses trivial and in some cases are not discern­ible from natural background levels. Monitoring carried out for this purpose is often based on potentially critical pathways, but there are also occasions when monitoring with little scientific justification is required.

(e) To establish pre-operational ambient levels of radiation Monitoring programmes are needed to determine background levels of radiation and radioactivity in order that long term variations may be observed. Information derived from such programmes serves to describe the state of the environment, that state being made up of radio­activity levels produced by nature and by man.

(f) To help decide on the action required during an emergency In the event of an emergency there must exist the capability to initiate monitoring immediately. The existence of gamma dose rate monitoring programmes which serve to provide an assessment of environmental levels of radioac­tivity will in many cases also enable an emergency programme to be launched without delay.

Monitoring near CEGB sites may be divided into three parts; Land gamma radiation measurements, Land samples and Marine environmental monitor­ing. The three parts may be described briefly as follows, noting that there are minor variations among stations.

The Department of Energy would be the lead depart­ment for all civil nuclear emergencies in England and Wales and would be responsible for the nuclear emer­gency briefing room in London. The briefing room would not be an operations room but would be the focal point in Whitehall for the information and brief­ing of government ministers and departments on the emergency. The department would be responsible for the formal appointment of a government technical adviser if an emergency was considered to be serious or likely to be prolonged beyond 24 hours. The gov­ernment technical adviser would be one of the Deputy Chief Inspectors of Nuclear Installations. The GTA would go to the operational support centre as a govern­ment representative and would:

• Provide authoritative advice to the police, local authorities and the health authorities on any action necessary to protect the public.

• Provide an assessment of the cause of the accident to government departments.

• Act as chief government spokesman on all off-site operations authorities and other bodies before ad­vising people that they might return home after an evacuation.

The government technical adviser would not be re­sponsible for taking any executive action.

The department would provide a central press brief­ing centre at their offices in London where govern­ment statements on the emergency would be issued.

A representative of the department would attend the operational support centre as liaison officer.

In common with all manufacturing processes, a pri­mary aim of the fuel cycle is to minimise production costs. In the case of the AGR this objective can be realised by reducing the refuelling rate, thereby re­laxing the demands on station fuel handling machinery and reducing the risk of expensive breakdowns. This is particularly important on the AGR since, in the absence of refuelling, reactor power will begin to decline rapidly after only two or three months.

The incorporation of a neutron ‘poison’ within a fuel element and arranging for it to ‘burn-out’ as ir­radiation proceeds can be used to advantage in AGRs in tw’o principal areas. Loading of fuel containing burnable poison within the initial charge can be used to delay the onset of regular refuelling, but the main area of interest lies in the use of burnable poison within feed fuel in order to assist in the attainment of higher discharge irradiations, and therefore reduced refuelling needs. Fuel which is required to last longer must be more highly enriched in U-235 in the interests of core reactivity, but high enrichment fuel would, if unchecked, produce very high channel powers at SOL, therefore poor form factors and related power penalties. The presence of poison in the fuel, however, suppresses high SOL powers and with suitable care the rate of destruction of the poison can be arranged such that the reactivity so released matches the reac­tivity lost through enrichment depletion. The overall result is that fuel reactivity, and therefore power, remains steady until the poison is completely used up (Fig 3.51).

This represents current AGR practice. For example, shortly after the beginning of loading the second replacement (feed) charge at Hinkley Point В (i. e., first feed discharged, second feed loaded), loadings began of fuel with higher feed enrichments, designed to reach 21 GWd/t at discharge and containing burn­able poison to assist in the maintenance of manage­able form factors. The poison used is gadolinium (which has a very high neutron absorption cross — section) in the form of gadolinium oxide, and is located in stainless steel cables or ‘toroids’ within the fuel element grid and brace support structure. The number of toroids within each element determines the reactivity taken up at SOL. In practice this means that the reactivity swing at refuelling is reduced by the presence of the poison, therefore assisting in the con­trol of channel powers. The rate of reactivity release from the poison burn-up is controlled by the mass of gadolinium within each toroid.

The choice of the most appropriate poison loading to achieve a given CAI at discharge is very compli­cated. In addition to the original enrichment increase, the presence of poisons in the fuel results in a lower average reactivity level over life when compared with the same enrichment unpoisoned, and accordingly the fuel needs further enriching in order to compensate. The use of burnable poisons therefore produces a reactivity ‘penalty’, resulting ultimately in more ex­pensive manufacturing costs, offsetting the original financial incentive to increase the discharge irradia­tion. However, there is usually a net gain because the reduced total fuel manufacturing costs more than compensates for the increased enrichment cost per stringer. At Hinkley Point В and Hunterston B, the 21 GWd/t fuel currently being loaded contains four toroids per fuel element, each with 25% gadolinium density so that the poison is completely burnt up after approximately 4.5 GWd/t. Eventually the re­fuelling rate will reduce in proportion to the dis­charge irradiations, i. e., in the ratio of 18 to 21. This represents a 14% reduction, equivalent to a saving of some 20 stringers per year for the station as a whole.