It has been indicated that the radial flux shape is non-uniform across the core and that it may be nat — tened by the loading of absorber. Similarly, the axial flux shape Is non-uniform and although this flux may be shaped it is not usual so to do.

As a consequence ot these flux shapes, the power developed by individual element* (axially in the chan­nels and radial I у across the reactor) о non-uniform. For a typical reactor the relative rating of radial chan­nels and of individually positioned axial elements is shown in Fig 3.39. The axial rating is relative to the mean channel irradiation. Selecting a channel lo­cated in the flattened zone and irradiated to the level of 5500 MWd/t, then from the axial relative rating curve, the relative rating of the No 8 element is 0.46 and of the No 4 element is 1.36. Thus, the No 8 element will have been irradiated to 2530 MWd. t (0.46 x 5500) and the No 4 element to 7480 MWd. t (1.36 x 5500). Although element-to-element irradia­tion differences exist, these can be employed to im­prove the utilisation of the fuel as indicated in the succeeding paragraphs.

Axial shuffling Considering the relative ratings of elements in a channel in the flattened zone, then the irradiation of the No 8 and No 1 elements, when the target channel irradiation of 5500 MWd/t is reached, is 2450 MWd/t and 2780 MWd/t respectively. If on discharge of the channel, the No 8 and No 1 elements are returned to their original positions, having re­placed the other elements in the channel with new fuel, they may be further irradiated to twice the above figures to 4900 MWd/t and 5560 MWd/t. An alter­native, on discharge of the channel at 5500 MWd/t, is to transfer the No 7 and No 2 elements with an initial irradiation of 4780 MWd/t and 5630 MWd/t to the No 8 and No 1 positions for a further life cycle.

The latter method has the advantage that the ex­change of positions is carried out every time the se­lected channel is visited. With the former method, retention of the fuel is on alternate visits. This tech­nique can be applied to the unfiattened zone of the reactor but it must be ensured that the dwell time of the retained elements does not exceed the current dwell time limitations,

Radial shuffling In an ideal reactor in which the channel How gagging is correct and refuelling has been carried out as required, the unnattened zone fuel would be discharged at the same irradiation as that of the Hattened zone. Associated with these conditions, the temperature distribution across the reactor would be a constant. However, variations in the original gag setting and the refuelling procedures frequently give rise to a non-uniform temperature distribution. Be­cause high fuel irradiations give rise to low reacti-

(a.1 AXIAL CHANNELS

Fig. 3.39 Typical relative rating

vity, retention of fuel at these higher levels could result in low power regions producing low channel gas outlet temperatures. As a consequence, it may become necessary to discharge the fuel earlier than at the target irradiation. Fuel utilisation may be improved by discharging the more lowly irradiated fuel and transferring it to the flattened zone thus enabling the fuel to be irradiated to the target discharge level. This process is known as radial shuffling.

Thus, certain areas of the flattened zone would be refuelled by partially irradiated fuel elements transfer­red from a less advantageous location in the unflatten­ed zone. New fuel would always be charged to the unflattened zone. It is necessary to consider each case on its own merits, bearing in mind the temperature distribution and the effect the fuel reactivity has in its original position and in the position to which it may be transferred.

The problem of access applies equally to the proce­dure used for inspection and for maintenance. Above — core access is typically via a restricted diameter of some 200-300 mm with a distance of 10-12 m to gain access to the above core area (the charge pan level is an additional 4-5 m lower). Within the above­core area, there are components which restrict the manoeuvrability of any equipment which has been inserted (Fig 3.57). In the above-core region, the en­vironment is an air or CO; atmosphere with tem­peratures between 60-l20°C. Radiation levels are moderately high, generally in the order of a 100 Rad/h rising to 1000 Rad/h over the top of a fuel channel.

Access to the below-core region is even more re­stricted in that it is often via the core itself through a fuel or control rod channel. In these cases, the ac­cess diameter is of the order of 70 mm and passage has to be made through a radiation field of several thousand Rad/h.

Optical inspection, whether it be photographic or by television techniques, has the additional problem of illumination. The internal surfaces of a reactor tend to be coated with a matt black deposit. Illumina­tion of such surfaces require high power directed — lighting systems. These generate large quantities of heat that have to be removed by some form of cooling and components require protection from radiated and conducted heat.

In the United Kingdom, as in many other countries, the safety standards for transporting radioactive ma­terials are based on the International Atomic Agency’s (IAEA) Regulations for the Safe Transport of Radio­active Materials,

Section 2.8.1 below outlines the scope and provi­sions of these Regulations and indicates how they are incorporated into UK transport legislation.

Against this background, Section 4.3 of this chapter explains CEGB safety standards and practice in the transport of irradiated fuel and other radioactive ma­terials associated with the CEGB’s magnox and AGR

power stations. Л brief assessment of PWR fuel trans­port safety is also given.

The selection of a nuclear site imolves many differing factors, but the CEGB’s prime consideration is the protection of the health and safety of the public in the event of an accident leading to an off-site release of radioactivity. Since a nuclear reactor contains large quantities of fission products, the release of a small fraction of this radioactive inventory could produce serious consequences. Thus the engineered safety fea­tures of a power plant must be examined in relation to the characteristics of the proposed site in order to ensure that adequate protection exists to guard the public and the environment against accidental events. When considering a possible site the characteristics of importance include the population distribution in the vicinity of the site, the meteorology, topography, geology and demography of the area and the ability to provide adequate facilities for dealing with emer­gencies and possible population evacuation associated with any major incident. Other matters which must be considered are the safety of potable water sources, geological stability and the risk to the site of external hazards such as floods, seismic movement and air­craft crash. Where cooling water supplies are a limiting factor, the choice of reactor system is of major im­portance. In order to determine the effect a nuclear power station might have on its environment, early investigative desk studies, rural surveys, and wind tun­nel tests are carried out by the CEGB.

The siting of nuclear power stations has been sub­ject to government policy since the first nuclear power programme was announced in 1955. A cautious ap­proach w-as adopted with limits on nuclear plant being built in heavily populated areas, even though they were of an inherently safe design. Initially little con­sideration was given to detailed site assessment with station selection being on the following basis:

• That only a very few isolated dwellings existed with­in a third of a mile of the reactors.

• That no more than a few hundred people lived within one and a half miles from the reactors.

• That only small communities (less than 10000 people would be found within five miles of the reactors).

In order to ensure that continuing development did not at feet the original bases for site selection, arrange­ments were made for the control of development in the area. At each site it was necessary for the local authority to consult with the Nil on proposed new developments in order to ensure that residential de­velopment could be controlled and the general char­acteristics of the siting area maintained. In particu­lar the system of control included any development likely to lead to (a) any increase of population within 1 mile of the site, (b) an increase of more than 50 people between 1 to 2 miles of the site and fc) an increase of more than 500 people within 2 to 5 miles of the site. Industrial development is also the subject of continuing consideration to ensure that it does not present a hazard to the nuclear plant and that the work force can be readily evacuated if necessary. These arrangements are still in force.

Development of initial siting criteria allowed more flexible treatment of population and took into accouni the results of the assessment of the Windscale inci­dent (1957) which indicated the importance of the radioactive isotopes of iodine in incidents leading to off-site releases.

In February 1968 a revised policy statement on nuclear siting was made in parliament, this indicated that gas-cooled reactors in pre-stressed concrete pres­sure vessels could be constructed and operated much nearer to built-up areas than previously permitted. No rigorous definitions were given for site require­ments but it was considered that the population char­acteristics of the sites at Hartlepool and Heysham were examples of acceptable sites under the new policy. Although these sites could be considered to be near — urban in character, they represent a class of site that in terms of risks to the public are an order of mag­nitude less than that for sites in or near large metro­politan areas.

A description of policy in the UK for siting nuclear power plants is given by Gronow and Gausden (1973). The paper makes mention of the fact that the main protection to the public from the consequences of nuclear accidents is provided by the engineered safe­guards built into the plant and the achievement of high standards in design, construction and operation. It further mentions that although siting is considered of secondary importance, it is nevertheless a most effective means of allowing control of the exposure of the population to radiation in the event of an in­cident leading to an off-site release of radioactivity.

The siting criteria developed to implement the po­licy excludes the use of sites in or near to large cen­tres of population, not least because emergency action becomes less reliable in built up areas on account of the large numbers of people that would be involved. In addition, emergency action would almost always have to cover a larger area than any specified bound­ary based on risk estimations, as it would be desir­able to evacuate whole communities rather than create public apprehension by stopping short at some arbi­trary boundary w’ithin a town or village. The extent to which provision is made for emergency action is an essential feature of present siting policy. Sites are only accepted if it can be show-n that effective emergency evacuation and possible medication can be taken for all persons within 1 km radius of the site and that this is capable of extension, if necessary, to all per­

sons within 3 km of the site. There are no restrictions on purely industrial development provided it does not present a hazard to the safe operation of the nuclear plant or that any incident arising from the nuclear plant presents any unusual situation at the industrial site. It is also necessary that the workforce can be effectively included in the emergency procedures.

In assessing a particular site, consideration is also given to the population characteristics in the 3 to 30 km radial zone. Development in this area is not sub­ject to control but it is unlikely that a licence would be granted for a site where development to full urban density was probable during the life of the station. A method of assessing a proposed site was first pro­posed by Gronow and Charlesworth (1967). This de­rived a set of characteristic curves for the site and the worst 30° sector. These are called the site and sector risk curves, and provide a measure of the po­pulation distribution around the site and give an as­sessment of risk in terms of thyroid dose due to air­borne radioactive iodine in the event of an off-site release, Further information on the derivation and application of this method of site assessment is given by Haire and Shaw (1979).

In addition to ensuring that proposals for a nuclear site meet the Government siting criteria, the CEGB has to consider a wide range of factors during the site investigations. These commence with desk studies of the areas of interest and largely confidential discus­sions are held with planning and other statutory au­thorities to identify sites worthy of detailed attention. Consultation and joint studies are also carried out with statutory authorities on such matters as water supplies, road and rail access, etc., in order to identify the problems of local services and to ensure that both the CEGB’s and local needs are reconciled. Where the formation of the land or meteorological conditions are likely to affect the normal dispersion of ventilated gases, tests are carried out in wind tunnels to deter­mine the optimum height of the stack and so ensure that acceptable ground level concentrations are not exceeded.

Subsoil investigations are used to check foundation conditions and marine surveys of tidal movements, depth of water and littoral drift give information to determine the effect of cooling water extraction and discharge on the coastal regime. In addition, zones of visual influence are prepared and, with topographical models, allow the subjective assessment of alternative station building and transmission arrangements.

It is an essential part of CEGB planning policy that new reactors must be licensable under the more relaxed siting criteria. It is also a continuing require­ment of the Nil that reactor systems new to commer­cial operation in the UK must be built on sites cleared under the siting criteria covering the early magnox stations. Thus, permission has been given to build the PWR at Sizewelh one of the original steel pressure vessel magnox reactor sites.

4 Radiological protection

For the purpose of compliance with the conditions of fuel transport off-site, various limitations are placed on the contents of each fuel storage skip. These in­clude limits on the number of elements and on their radioactivity and decay heat content. A minimum cooling time is also specified and in the case of un­bottled magnox fuel, there is a further limit on the release rate of caesium-137. The latter is determined by placing a hood over the fuel skip to isolate it from the main pond environment, and sampling at intervals to measure the rate of increase in concentration of caesium-137.

A system of administrative controls is applied to ensure that the contents of each skip comply with all the relevant limits.

Summary of experience at magnox and A GR stations Quantitative standards have been established for the physical and chemical conditions in magnox fuel ele­ment storage ponds. Prolonged storage times and other adverse factors have on occasion led to enhanced transfer of radioactivity into the ponds, and plant modifications have been necessary to maintain good conditions in these circumstances.

In respect of radiation from the pond water the most important radionuclide is caesium-137. Concen­trations of this nuclide can generally be maintained at or below 20 to 40 kBq/1, i. e., approximately 0.5 to 1.0 /tCi/І, at which concentrations the correspond­ing gamma radiation dose rates at the pond walk­ways are 5-10 /xSv/h (0.5-1.0 mr/h). The dose rates increase during fuel handling operations and during maintenance w’ork.

Alpha activity associated with transuranic particu­lates is the most important constituent of airborne contamination in pond areas. The sources and transfer mechanisms are now well understood and can be con­trolled such that respiratory protection is generally not required other than during special maintenance work on pond equipment. Because of the particularly stringent controls which are applied to work in con­ditions where airborne contamination is present, the contribution of such contamination to personnel dose is very small.

Operational targets have also been established for AGR pond conditions. Experience to date suggests that spallation of radioactive materials in oxides from the fuel element surfaces may not be as severe a problem as — was anticipated.

The principal contribution to radiation dose rate is from cobalt-60. Levels of this radionuclide in pond water are generally less than 40 Bq/1 (1.0 (iC/) and, correspondingly, radiation dose rates in operating areas are generally less than 20 /iSv/h (2 mr/h).

The mechanisms of airborne contamination transfer are similar to those at magnox stations.

The role of the MAFF in an emergency would be:

• To ensure as far as possible that no one is exposed to the risk of consuming food {including milk) con­taminated by radioactivity to unacceptable levels.

• To ensure that alternative food supplies are available.

• To mitigate as far as possible the effects of an accident on agriculture, fisheries and food.

There is a central MAFF memorandum of proce­dures setting out the Ministry’s policy and each MAFF region has a nuclear contingency plan containing de­tails of the emergency organisation in the region, together with the arrangements for coordinating all actions on behalf of the Ministry.’

In the event of an emergency, the MAFF Chief Regional Officer would be responsible for deciding what counter measures were required and would:

• Proceed to the operational support centre to act as liaison officer on all matters within the purview of the MAFF. The MAFF headquarters atomic energy branch liaison officer would join the chief regional officer at the OSC.

• Establish regional and/or divisional operation centres for the control of field staff.

• Notify National Farmers Union county branch secretaries.

• Decide, on the basis of scientific advice and survey data, whether the distribution of milk or other foodstuffs should be banned.

• Issue instructions to farmers, food manufacturers and fishermen.

• Review monitoring arrangements with health phy­sicists at the OSC and mobilise the regional tech­nical services for the collection of samples.

• Notify the food science division if the decision is taken to ban milk or other foodstuffs.

The simplest initial loading pattern would involve the use of only one fuel enrichment, chosen 50 that the overall ‘built-in’ reactivity of the core would be just sufficient for the attainment of criticality. In fuel cycle terminology this is often referred to as the just critical scheme, but because of its potentially wasteful use of fuel is not an idea adopted in practice. Fuel would need replacing almost immediately after start-up in order to replenish reactivity, and the very low reactivity benefit obtained at each refuelling would also result in the need for rapid refuelling so that core reactivity could be maintained in the longer term. This makes for poor fuel utilisation — too little energy removed per unit mass of initial fuel, and w-ould obviously be very costly. Furthermore, since no provision is made for power flattening, additional output penalties would be incurred. However, the just critical scheme does provide a useful yardstick against which more sophis­ticated ideas can be measured.

Vacancies

All the fuel in the initial charge is new, resulting in no age-induced reactivity differences between chan­nels. Better radial form factors are therefore pro­duced enabling design reactor power to be more easily achieved. This natural benefit at SOL can be com­bined with the choice of slightly higher fuel enrich­ments than that required for the ‘just-critical’ con­dition, in order to build in excess reactivity into the initial charge, which can then be used to provide two major sources of cost savings. In the first instance it can lead to the attainment of design thermal power from the reactor at SOL with fewer fuelled channels than are available in the core as a whole. The un­fuelled channels can be charged with graphite ‘vacancy’ blocks, which are then systematically fuelled at the beginning of the refuelling sequence. The second bene­fit that arises relates to the timing of the start of refuelling (i. e., vacancy replacement), which can be delayed because of the extra built-in reactivity. At Hinkley Point В and Hunterston В, for example, 32 of the 308 fuel channels were loaded with vacancy stringers around the core periphery in the initial charge.

Although the initial enrichments need to be raised in such a scheme, and at some expense, a large net reduction in initial charge costs results from the need to load fewer channels at the outset. The delayed onset of refuelling makes a further contribution to cost savings.

During the studies carried out to evaluate the initial charge enrichments, careful attention also needs to be addressed to the requirements of reactor safety, particularly if the enrichments need to be further increased in schemes such as this. Too large an in­crease could conflict with essential reactor shutdown criteria.

An alternative to the vacancy scheme is one in­volving the use of burnable poisons. With the initial enrichments being increased above the just critical level, the excess reactivity is taken up by absorber or ‘poison’ material present in the fuel elements. This is arranged to subsequently ‘burn-out’ at such a rate that the reactivity of the initial core remains steady for a period of time, during which of course no re­fuelling becomes necessary. The cost of the initial charge in such a scheme is therefore increased but the refuelling cost savings are improved by a substantially greater margin, resulting in an overall reduction in costs.

On balance, the vacancy scheme is usually prefer­red to the use of burnable poisons within the initial charge, particularly when used in conjunction with the ‘mixed enrichment’ idea described in the following

section. Burnable poison schemes require both a higher degree of extra enrichment of the initial fuel as well as the loading of all the channels.

The CEGB is responsible for the operation of the commercial nuclear power stations in England and Wales. To build and operate such stations a site li­cence is granted by the Health and Safety Executive and the licence conditions are administered by HM Nuclear Installations Inspectorate (Nil). At the time of construction of the magnox stations, a period of 20 years was selected for amortisation purposes. As the result of experience it has become apparent that these stations will have a life considerably in excess of 20 years. As a consequence the Nil requested the CEGB to carry out a comprehensive review’ of safety-related aspects of operation and maintenance of each station as it approaches its 20-year period of operation. The review, titled the Long Term Safety Review, is com­monly known as ‘the 20 year review’, albeit the time of 20 years lies in the financial provision for the station rather than any safety implications.

The review of reactor safety with respect to design, operation and maintenance is implemented immediate­ly the design stage is launched. No station can com­mence operation until the N11 is satisfied that an adequate safety study has been presented. Even so, the design criteria are checked stage-by-stage during the commissioning period. Approval is given by the Nil to proceed to each commissioning phase after the results of the previous phases have been analysed and found satisfactory.

The site licence requires the CEGB to establish a Nuclear Safety Committee for each site to consider and advise the site on the safety of modifications to plant and operating rules. The committee comprises experienced senior members of the CEGB power sta­tions, departments and divisions, together with similar personnel from the United Kingdom Atomic Energy Authority and British Nuclear Fuels, Meetings are held monthly at which submissions from the stations are reviewed and recommendations to accept, reject or amend the submissions are made. The proceedings of the committee are notified to the Nil and no changes to operating rules or significant plant modi­fications may be made without endorsment by the committee and with the agreement of the NIL Thus, these procedures constitute an on-going review process.

Each of the CEGB’s reactors are shut down bien­nially for routine inspection and maintenance proce­dures. The procedures are the subject of a maintenance schedule, approved by the Nil, and specifies the nature and frequency of inspections and overhaul. The reactor concerned may not be restarted unless the Nil are satisfied that the requirements of the schedule have been fulfilled in respect of the shutdown pro­cedures and those procedures performed outside and prior to the shutdown.

The N11 has inspectors appointed to every station to enable a continuous review to be made of operating conditions, procedures and current approved modifi­cations in hand. In addition, the CEGB has its own independent Health and Safety Department which also has site-based inspectors functioning in a similar man­ner to the Nil’s personnel.

Thus, nuclear safety of CEGB sites is subject to a continuous and independent review of all safety as­pects throughout their operational life. Nevertheless, the N11 requires the Long Term Safety Reviews to be carried out.

The review presents in one comprehensive set of documents the satisfactory nature of the overall safety case. Operating experience, which is not available at the design stage, is incorporated in the review together with the accumulated review procedures developed over the life of the station. Where possible, plant is compared against present day standards, although ba­sically, comparisons are made against the as-designed criteria. However, modifications are subject to current standards and as such are assessed by staff external to the station and who are aware of present day requirements. The review incorporates mechanisms not fully recognised at the design stage such as the ef­fects of mild steel oxidation in the coolant gas. Areas may be identified where improvements to system de­sign are practicable and worthwhile. Such a case mav be the segregation of essential supplies and protection circuits by physically separating two identical lines of supply, so that damage to one area does not affect the other.

The review covers all aspects of plant maintenance and operation which may have an influence on nu­clear safety, i. e., those aspects which could directly or indirectly result in a radiological incident. The scope of the review is demonstrated by the require­ments of the Nil of areas to be reviewed, as follows:

• Pressure vessels and primary circuit integrity.

• Graphite behaviour.

• Tripping parameters and shutdown systems.

• Cooling systems.

• Electrical supplies.

• Fault studies.

• Consequential damage following depressurisation.

• Operating rules and role of the operator.

• Hazards.

• Fuel handling and fuel route assessment.

• Health physics. —

• Radioactive waste management.

• Decommissioning.

The comprehensive nature of the reviews is amply demonstrated by considering the topic of ‘cooling sys­tems’ listed. The ability to remove heat from a shut­down reactor is important to ensure that overheating of the fuel does not occur. This involves adequate passage of the coolant gas over the fuel in the core, i. e., core (graphite) integrity. Gas circulators drive the coolant gas through the core and boiler, i. e., circula­tor integrity. These in turn derive their power from the ‘essential supplies’ system of which the emergency gas turbine or diesel generators form a vital part. The boilers are required to remove the heat transferred from the fuel by the coolant gas. This in its turn requires the integrity of the feed water supplies to the boiler to be examined. Any steam generated within the boilers has to be passed to the dump condenser and the heat carried in the steam must be removed by the main circulating water system, i. e., CW pump

system integrity. This resume is sufficient to demon­strate the wide scope of the review and the manner in which the various plant systems have to be con­sidered as a whole rather than individual entities.

In contrast to the example given, an examination of the reactor core structure requires a more theo­retical approach, the findings of which are supported bv observation, measurement and sampling techniques. As indicated in Sections 9,1 and 9.3 of this chapter, sampling of materials is earned out to determine material analysis and measurements are taken to con­firm theoretical assessment.

The review’s are co-ordinated centrally for the CEGB by its Nuclear Operations Support Group, whose func­tion ensures a uniform approach to topic investigation and presentation. At the request of the Nil, papers

References

are assessed independently within the CEGB, although those relating to pressure vessel integrity are assessed by the CEGB’s technical insurers. The reviews pro­vide confirmation that the existing review procedures adopted by the CEGB, including monitoring and in­spection, are achieving their objectives and provide a satisfactory basis for the operation of the station beyond the nominal 20-year life.

The purpose of the branch is to provide a focus for non-nuclear safety matters within the CEGB and to formulate policy, provide advice, set targets, conduct examinations, perform audits, collect statistics, con­tribute to the development of new health and safety legislation, study the implications of such legislations on the CEGB’s business and make recommendations as appropriate.

The branch is managed by the Head of Industrial Safety and is supported by five further managers, two at CEGB Headquarters responsible for policy and safety performance, and three field managers each of whom will service all CEGB locations in his area.

The Industrial Safety Branch is the custodian, of the CEGB’s (Electrical and Mechanical) Safety Rules and Codes of Practice and issue other guidance in the form of General Series Codes of Practice grouped under Mechanical, Electrical, Environmental Health, Chemical, and Operational Procedures. The informa­tion contained in the codes provides long-term formal management guidance for controlling industrial safety hazards on site. From time to time, it is found neces­sary to promulgate immediate safety advice to locations because of incidents which have happened within or outside the CEGB. Frequently, the advice is of such urgency that it has to be issued with only a general understanding of the topic and before it has been fully researched. In such cases, preliminary advice is issued to the Area Safety Managers in the form of a Safety Information Co-ordination Note, and they then re­commend a suitable course of action for the Station Manager to follow, to prevent a dangerous situation arising at his station.

2.10.2 Health and Safety Strategy Branch

The purpose of this branch is to carry out studies and assessments of a range of matters relating to health and safety which are strategic in nature and broader than the remit of the other branches in the department.

The branch is led by the Head of Health and Safe­ty Strategy and is divided into two sections, each with its owm manager; General Health and Safety, and Nuclear.

An atom consists of a relatively massive nucleus con­taining a number of positively charged protons and neutral neutrons, collectively known as nucleons.

Surrounding the nucleus are orbiting electrons (each having a mass of about one two-thousandth that of a nucleon) in equal numbers to the number of pro­tons, giving an overall electrically-neutral atom. The nucleus occupies very little space (a diameter of about 10“15 m) but contains virtually all the mass. The whole atom has a diameter of about 10“ n m, thus an atom mainly consists of free space.

The number of protons in an atom determines the number of electrons and thus the chemistry of that atom. Atoms with the same number of protons are of the same element. If the number of neutrons in two atoms is different, but the number of protons is the same, then the atoms are said to be isotopes of the same element.

In heavy elements, if the neutron-to-proton ratio is too low, the atom attempts to rectify this by emit­ting a so called alpha particle. This alpha particle is emitted at high energy from the nucleus and consists of two protons and two neutrons, it is thus elemen­tally the same as the helium nucleus. Alpha particles are not very penetrating because of their relatively massive size. A thin layer of skin or metal foil will easily stop them, and their range in air is only a few centimetres. Consequently they can prove difficult to detect. Because they are easily stopped, alpha particles do not cause a radiation hazard when external to the body. However, when ingested or inhaled, they can create a serious radiological hazard since most of any alpha particle’s energy will be dissipated in a small volume of tissue surrounding the radioactive particle and cause severe local cellular damage.

When the neutron to proton ratio is too high, an atom will attempt to stabilise by getting rid of a ne­gative charge from the neutral neutron, which may be thought of comprising a positive plus a negative charge. This negative charge is emitted in the form of a beta particle which is physically the same as an electron and has a very low mass. In effect therefore, the nucleus of the atom loses a neutron but gains a proton, thereby rectifying the imbalance. Beta parti­cles have the ability, depending upon their energy, to penetrate tissue to varying depths, and can cause an external radiation hazard to skin tissue in particular. Generally, beta particles of average energy (up to 0.5 MeV) termed ‘soft’, are easily stopped by a few mil­limetres of tissue or several centimetres in air. ‘Hard’ beta particles having energy greater than 0.5 MeV travel correspondingly further; up to one metre in air. Like alpha particles, betas may also present a hazard if inhaled or ingested. Beta particles can in­teract with a shielding medium, such as lead, and cause the production of penetrating X-rays (known as bremsstrahlung). Consequently, the shielding against beta particles must be designed carefully.

If the neutron-to-proton ratio is too low, as for alpha decay, but the emission of an alpha particle is energetically impossible, the nucleus may give up a positive charge. If the proton of the nucleus is con­sidered equivalent to a positive charge plus a neutral charge, then the positive charge is emitted by a par­ticle, again of similar mass to the electron and is known as positron. Since positrons are essentially elec­trons, then the radiation hazards associated with them are identical to those of beta particles. Positrons are unstable in the environment outside of the nucleus and very quickly interact with an electron, causing both to annihilate each other and give off gamma radiation.

In some isotopes, if the neutron to proton ratio is too high or too low then, instead of emitting the particles described above, a neutron or proton is di­rectly emitted, although these isotopes are fairly rare. However, sources of neutrons do exist which originate from the collision between an alpha particle and cer­tain elements. As neutrons are electrically neutral they are very penetrating and can cause much biological damage.

Gamma-rays are mono-energetic electromagnetic ra­diations, emitted from the excited nucleus of an atom following radioactive decay. It is the method by which an excited nucleus rids itself of unwanted energy in reaching equilibrium. Since they are in the form of electromagnetic radiations gamma rays penetrate ma­terials, especially tissues, very easily and consequently pose an external radiation hazard. Protection against gamma (and X-rays) is best afforded by a combina­tion of shielding, time of exposure and distance from source (these topics will be dealt with in detail later).

X-rays originate not from the nucleus of an atom, but from the electron shells surrounding this. If an electron is removed somehow, e. g., by a collision with another particle, then a gap is left which must be filled. This vacancy is taken by another electron oribiting at a higher level but lower energy. This energy difference is expressed by the emission of an X-ray, which is electromagnetic radiation.

In a nuclear power station, the reactor is the pri­mary source of all radiation on and off the site. The fission process, which involves the splitting of uranium nuclei after the absorption of a neutron, gives rise to a whole spectrum of radioactive isotopes or fission products. These are mainly beta-gamma emitters, al­though uranium is also radioactive in its own right and emits all types of radiation. The fission products are sealed within the fuel element and do not escape unless there is a failure of the can. However, radiation from the fission products, although being absorbed to a very large extent within the reactor, does penetrate to the outside world.

In addition to the fission products, the fission process produces neutrons, thus enabling the chain reaction to continue. However, as described already, neutrons are very penetrating and may escape from the reactor. The optimum neutron shielding is material consisting mainly of hydrogen, for example, water, paraffin wax and concrete. Hence the main neutron shield around a reactor is a massive wall of concrete.

In addition to interacting with the uranium in the fuel, the neutrons may also interact with virtually any material and make it radioactive. For instance, steel contains a certain proportion of cobalt, mainly as the non-radioactive isotope cobalt-59. The cobalt-59 absorbs a neutron to produce cobalt-60 which is ra­dioactive. The graphite moderator of a reactor is a principal source of tritium gas, produced by neutrons being absorbed by lithium impurities giving rise to an alpha particle and tritium. The carbon dioxide coolant contains some natural argon which, when irradiated by neutrons, gives the radioactive isotope argon-41. The examples are many.