Classes of effect

The effects of radiation upon animals, including man, can conveniently be divided into two groups:

Genetic effects A result of cell transformation in the reproductive cells. Genetic effects are then inherited and, as they are manifested in offspring, they are es­sentially long term.

Somatic effects A result of cell death or cell trans­formation in any cells of the body. Somatic effects are observed in the individual irradiated. It is con­venient to further subdivide somatic effects into early effects (broadly those which appear within a few weeks) and late effects (which may take tens of years to manifest themselves).

Dose/response relationships

For any biologically harmful agent, it is important to correlate the dose administered with the response or damage produced in order to establish acceptable levels of exposure. The response may be the frequency of induction of a particular late effect such as cancer or the severity of an early effect such as skin dam­age. A graph of dose versus response (the dose/ — response curve) is important in ascertaining a dose level ait which’ the associated biological response is acceptable.

Figure 4.4 is the dose/response curve for an effect which increases linearly with dose. The probability of the response appearing rather than its severity is a function of dose. As the dose/response line passes

through the origin there is no dose at which there is no probability of the response occurring. Effects which follow this pattern are known as stochastic effects. Cancer production is a stochastic effect.

Figure 4.5 is the dose/response curve for an effect which is not observed at low doses. The effect exhi­bits a threshold dose. Some mechanism other than dose plays a part in determining the response. Typi­cally the response is the severity of the effect rather than the probability of it occurring. Such effects are termed non-stochastic. Examples are skin reddening (erythema) and cataract formation on the lens of the eye.

It is worth differentiating between two types of radiation exposure. Acute exposure is a single short duration exposure which, by implication, is usually large. Chronic exposure is a continuing or repeated exposure over a long period and may be at low levels.

Monitoring of radioactivity in the marine environment is directed towards the assessment of radiation ex­posure of the public by means of critical pathway analysis. Marine environmental monitoring is therefore carried out as follows:

• Monthly samples of locally caught fish and shellfish are measured for total beta activity and for indi­vidual radionuclides by gamma spectrometry. Where local dietary habits warrant it, individual species are examined specifically, e. g., oysters at Bradwell, trout at Trawsfynydd, which are known to concen­trate certain radionuclides.

• At those stations situated on coastal sites, measure­ments of gamma dose rates are carried out along the foreshore. Special attention is paid to areas of fine sediments, particularly where occupancy by members of the public is likely to be high. Dose rates are measured on a quarterly basis.

• In support of the gamma dose rate measurements, samples of sand and silt from the above areas may also be collected. These are measured for total beta activity, and for individual nuclides by means of gamma spectrometry.

The monitoring programmes described here are those current at the time of writing, although it is likely that in the near future the analysis of milk for fission products will be curtailed.

Consignments of irradiated fuel are conveyed at fre­quent intervals from CEGB nuclear power stations to the British Nuclear Fuels works at Sellafield for reprocessing. A special emergency plan has been pre­pared to achieve a rapid and effective response to any mishap involving an irradiated fuel transport flask whilst in transit by road or rail. For accidents to fuel flasks in transit this emergency plan replaces the National Arrangements for Incidents involving Radioactivity (NAIR), a scheme introduced in 1964 to provide expert advice for the police when con­fronted with an incident which might involve the exposure of the public to radiation. Fuel flasks owned by the South of Scotland Electricity Board, Nuclear Transport Limited, Pacific Nuclear Transport Limited, BNF and the UKAEA are also covered by this emer­gency plan. For simplicity, loaded and empty flasks are treated in the same way.

Essentially the emergency plan aims to provide health physics expertise at the scene of an accident in the shortest practicable time. The health physicist would be supported by an emergency team trained and equipped to limit the spread of any radioactivity that might eventually leak from the flask following a very severe mishap and to assist in the safe transport of the affected flask to a suitable destination.

AGRs in the UK use enriched uranium dioxide (UO2) fuel contained within a stainless steel clad or ‘can’. Each fuel element, w’hich is approximately 1 m in length, consists of a bundle of 36 fuel pins, each 14.5 mm diameter and of the same enrichment, supported within a 190 mm bore double-graphite sleeve assem­bly. This serves both to insulate the moderator bricks within the reactor fuel channels from the hot coolant passing over the fuel, and also to permit the flow of cooler gas around the outer sleeve of the fuel element to further cool the moderator. As shown in Fig 3.53, the fuel pins, arranged in three concentric rings of 6, 12 and 18 are attached to machined grids at their lower ends and traverse two additional support braces at the centre and upper end of the fuel element. A central guide tube accommodates the threading of a tie-bar which takes up the full weight of the fuel stack {usually eight elements) during subsequent handling of the completed stringer by the charge machine.

The cans are made from niobium-stabilised stain­less steel, the overall volume of which needs to be mmimi>ed on grounds of neutron economy, leading a fundamental!} ‘thin can’ design of pin. Heat transter is enhanced via surface-roughening of the can m the term of machined transverse annular ribs, although some of the earlier AGR fuel elements con­tained helically-ribbed cans. At the pin ends, cup-

~E =E’AA NG = G

shaped end caps are welded to the main can wall in two places to provide a double-seal between the fuel and the reactor environment and the pins are locked into the fuel element grid at their lower ends. Each pin contains typically 64 hollow UCb fuel pellets, some of which are specially grooved to receive the can during its pressurisation onto the fuel in a helium gas environment, during the final stages of pin manu­

facture. These ‘anti-stacking’ grooves (ASGs) prevent gaps opening up in the pellet stack during service as a result of differential movements between the fuel and the can. If present, such gaps would allow the can to deform into them under the influence of the external reactor gas pressure, which would eventually lead to localised weaknesses in the can wall and hence provide a possible mechanism for failure. Fuel cur­rently in use at Hinkley Point and Hunterston con­tains either 14 or 22 ASGs per pin. The top two positions within the fuel stack (elements 7 and 8) are assembled only with elements containing 22 ASG pins, since these positions have shown themselves to be more prone to inter-pellet gap formation than those lower down the stack. An increase in the number of ASGs reduces gapping to a minimum. The 14 ASG fuel is assembled into positions 1 to 6. At each end of the pellet stack there is an insulating pellet made from sintered alumina powder, the function of which is to protect the can at the pin ends, a region of poor heat transfer, from excessive temperatures. The UCb fuel pellets contain a hollow central bore (typi­cally 6.35 mm dia. in current Hinkley Point feed fuel) which acts as a void for the accumulation of fission gases produced during reactor operation. Since the rate of release of fission gas depends strongly upon the fuel temperature, the presence of the bore also restricts the quantity of gas actually released from within the fuel by effectively lowering the centre UO2 temperature for a given heat rating. The degree of enrichment of the fuel in U-235 varies across the reactor core and also between stations, but generally lies between l№o and 3% by weight.

In the UK, control over radioactive wastes is exer­cised through the Radioactive Substances Act 1960, (‘the Act’) which came into force in 1963 following a government white paper policy document ‘The Control of Radioactive Wastes’ (Cmnd 884). This policy was reviewed following consideration of the 1976 Sixth Report of the Royal Commission on Environmental ~ Pollution [1]. Further reviews of policy and objectives were given in the 1977 White Paper ‘Nuclear Power and the Environment’ [2] and an Expert Group re­viewed Cmnd 884 in 1979 [3]. Policy objectives were repeated in a 1982 White Paper ‘Radioactive Waste Management’ [4]. The Government has established a Radioactive Waste Management Advisory Committee whose fifth Annual Report was made in 1984 [5]. The specific items of reference of the Committee are: ‘To advise the Secretaries of State for the Environ­ment, Scotland and Wales on major issues relating to the development and implementation of an overall policy for the management of civil radioactive waste, including the waste management implications of nu­clear policy, of the design of nuclear systems and of research and development, and the environmental as­pects of the handling and treatment of wastes.’

Under an independent Chairman, the Committee currently consists of eight independent members, four members from the nuclear and electricity generating industries and three from the trade unions with mem­bers in those industries. Whilst these reviews are re­flected in the interpretation of the Radioactive Sub­stances Act 1960, the Act itself remains unchanged.

Plant faults are those which might arise due to fail­ures or malfunction of the plant itself. They include those which can affect the reactor, the reactor pri­mary and secondary coolant systems, and also those which can affect new and irradiated fuel and radio­active effluent treatment plants. An outline of a few such faults is given in Section 3.5 of this chapter (Fault studies).

Internal hazards are on-site events which have the potential to cause damage to unrelated equipment around the point of origin, or beyond. Those con­sidered include:

• Fire, missiles, turbine disintegration, dropped loads, explosive and toxic gases, and failure of pressurised systems. Failure of pressurised systems could, for example, result in pipe whip, flying fragments, jet impingement and local flooding.

By comparison with natural and other external hazards it is more often possible to place an upper limit on the level of interna! hazard, for example, the magni­tude of a dropped load.

The design approach to demonstrate the safety of the plant against internal hazards depends on the particular hazard under consideration, but the main features of the approach are:

• To reduce the probability of the hazard occurring through a high standard of design, construction and operation.

• To limit the consequences of hazards through se­gregation, protection and layout features.

• To ensure that the reactor can be taken to a hot shutdown condition in the event of the hazard oc­curring. Some repair, temporary connections (for example, cabling), or local manual operation may be required to establish cold shutdown capability. Maintenance requirements and the single failure criterion are taken into account when assessing the adequacy of the plant proposals.

External hazards are events originating outside the site but which could affect plant safety and include both natural and man-made events. Those considered for the PWR for instance include:

• Ground settlement and subsidence, precipitation, lightning, wind, flooding, extreme ambient tempera­tures, industrial activity off-site, gas clouds, earth­quakes, aircraft impact.

Where possible, the criteria for hazards is to ensure that the combination of the frequency of the hazard and the probability of subsequent failure to control the reactor is consistent with the design safety criteria for large uncontrolled releases to the environment. On this basis, the external hazard design criteria have in general been set at a level of frequency of 10“4 per year and the probability of failure to cool and shutdown the reactor in the event of the hazard set at less than 10 ~3 per demand.

Thus, in the case of earthquakes for example, an earthquake with a peak horizontal acceleration of 0.25 g has been adopted for design purposes. This exceeds the 10 ~4 per year level at all currently planned sites; it is above the UK average and would be conservative for most UK locations. This has associated with it a ground motion spectrum which characterises the fre­quency content of the various possible earthquakes which could have a peak acceleration, of 0.25 g. The station and plant design has to be such that the re­actor can be safely shutdown with requisite reliability in the event of such an earthquake, and is known as the safe shutdown earthquake (SSE). It will be noted that earthquakes which exceed this design level are possible. However, in practice the plant will safely withstand more severe hazards than that specified be­cause plant and structures built to normal conservative design codes have a considerable margin of reserve strength between the design limits and the point of collapse or total failure. Sensitivity studies are carried out to show that there is no sudden reduction in safety for events more severe than the SSE.

As in the case of the SSE, the design wind speed is set at a level such that the frequency of its being exceeded is less than 10“4 per year. On this basis, the extreme wind speed arrived at for the design of safety — related buildings and plant is 58.8 m/s {132 mph).

Aircraft crash exemplifies an external hazard for which a modified approach is adopted to establish the acceptability of the design. A different approach is possible because sufficient data for aircraft crashes exist to enable the frequency of both civil and mili­tary aircraft crashes to be predicted for the site. This, together with an evaluation to identify those areas of the site which, if hit, have the potential for signifi­cant radioactive release, leads to a predicted frequency of crash onto potentially vulnerable areas. Because of the inherent strength of some of the buildings, it is possible to discount any hazard arising from impact of light aircraft and some helicopters. Although the arguments are further complicated by consideration of consequential fires due, for example, to spillage of aviation fuel, it is possible from this type of assess­ment to show that major radioactive release from an aircraft crash is of the order of 10“7 per year and therefore considered an acceptably low risk.

For some hazards a probabilistic approach is not appropriate or, because of lack of data, not possible. In the case of precipitation, for example, normal UK practice is used; this is regarded as adequate when taking into consideration the segregation, redundancy and diversity in the plant. For sea flooding, as an­other example, a maximum credible sea level has been evaluated.

4.3.1 Development of the rules

In November 1957 the Nuclear Safety Rules Sub­committee was formed with the terms of reference ‘to formulate rules for the radiological protection of persons employed on CEGB premises’.

The first meeting of the subcommittee was held in 1958 and since that time there have been several changes to the rules, although the underlying princi­ples of the first rules still remain valid today.

In recent years the subcommittee was renamed the Nuclear Safety Rules Advisory Committee. The Chairman is the Director of Health and Safety and the membership comprises the Electricity Council’s Chief Safety Officer, Senior Regional Representatives, the Nuclear Safety Officer of the Health and Safety De­partment, the CEGB’s Medical Adviser, various Power Station Managers, the Senior Research Health Physi­cist of Berkeley Nuclear Laboratories and a representa­tive of the Electrical Power Engineers Association. A technical secretary is provided by the Nuclear Safety Branch of Health and Safety Department^

Originally there were two formal sets of rules, the Safety Rules (Radiological) and the Safety Rules Ion­ising Radiations for Non-Destructive Testing but, with the introduction of the Ionising Radiations Regulations 1985, these were, in effect, combined.

The preparation of the first rules was made all the more difficult by the absence of any significant legislation. The Factories Act legislation (the Sealed Sources and Unsealed Radioactive Substances Regu­lations), only existed in draft form and the Nuclear Installations Act of 1959 and the Radioactive Sub­stances Act of 1960 did not exist. None of the current series of ICRP recommendations existed, although ICRP Publication 1 was introduced later on in 1958.

In spite of all this, some Codes of Practice were in existence and there was also the practical experience of the United Kingdom Atomic Energy Authority.

Responsibility

In accordance with the CEGB QA policy and the adoption of BS5882 by the CEGB as the appropriate basis for the total QA programme for its power stations, it is the responsibility of the CEGB to ensure that a QA programme for each plant is established and implemented. In each phase of a power station’s life, many organisations are involved in providing plant items and services to the CEGB.

Where these organisations lie within the CEGB, their responsibilities are defined within the CEGB Directives, it may be necessary to prepare project — specific QA documentation, for example, in compli­ance with nuclear site licence requirements, in addition to generic procedural documentation.

Where organisations external to the CEGB are sup­pliers, the CEGB will require that each organisation demonstrates that its own QA arrangements are ade­quate and specifies or approves those of its contractors, suppliers and agents. This hierarchical system of iden­tifying detailed responsibility, together with formal CEGB monitoring of the immediate purchaser, is followed throughout the purchaser/supplier arrange­ments whilst not relieving the CEGB of its overall QA responsibility.

Design

Design covers not only how an end item is to be made but also how it is to be used. The designer is responsible not only for the proof of integrity of the design by analysis and testing, and for providing the manufacturer with the instructions of how to make an item, but also for provision to the operator of the item of the instructions for the use of the item.

The objective of QA in design is to provide mea­sures of confidence that the management system re­sponsible for design has adequately considered the requirements for reliability, performance and safety in formulating the design, has provided instructions to the manufacturer for the realisation of the design and has prepared an examination, inspection and test schedule that will confirm the design assumptions.

Procurement

Procurement covers all the activities associated with realising a declared design in an end item, i. e., re­quisition, manufacture and construction, the supplier is responsible for both the design and the procurement of plant systems.

The objective of QA in procurement is to provide measures of confidence in the ability and performance of a supplier in the production of the end item.

The assessment of the quality capabilities of sup­pliers is an essential feature of the procurement pro­cess. The function of the plant item will determine the safety significance and hence the QA actions ap­propriate in the procurement process. This will include the identification of an appropriate quality manage­ment system standard and the particular actions, such as procurement authority approval of QA programmes and quality plans, enforceable through the terms of contract.

In common with most industrial processes nuclear power generation produces a certain amount of waste. The feature peculiar to nuclear generating stations is that a proportion of this waste is radioactie or contaminated by radioactive material.

The use of radioactive substances and the genera­tion of nuclear power have, for a number of years, been subject to comprehensive statutory controls in order to protect the public. All disposals of radio­active wastes require the authorisation of the appro­priate government departments under the Radioactive Substances Act 1960 I he aim of government policy is that the total contribution to radiation exposure from radioactive wastes should remain very small, even though the use of nuclear power increases. The aim is also that the radiation doses received by in­dividuals, which of course vary, must be kept well within acceptable limits.

8.3.1 Waste categories

For convenience, wastes are categorised as follows:

• Low level — These are wastes for which facilities for their disposal are provided for in the design

of the station.

• Intermediate levels — These may be safely stored on site, although facilities for their final disposal may not yet exist. It is worth noting that this temporary site storage can reduce the initial acti­vity to more manageable levels by radioactive decay.

• High level or heat generating wastes — It is the irradiated fuel which falls into this category and the final handling of the waste is dealt with by the BNFL.

Generally, the lower the level of activity, the greater is the volume of waste material involved. The prin­ciple adopted in handling wastes is to reduce the volume for storage and to dilute gases or liquids be­fore discharge. Large volumes of low level wastes in the form of paper, wipes or oil are reduced in volume by incineration before disposal.

The Regulations have the enforcement of law and prosecutions may thus be brought for specific breaches of particular requirements.

In order to provide an interpretation of how speci­fic regulations may be complied with, the Approved Code of Practice (ACOP) is attached to the regula­tions. As this is an approved code, compliance with its requirements automatically ensures compliance with the relevant regulation. However, ACOP does not have the force of law and, where it is not in use, the onus may be placed on an individual to demonstrate com­pliance with the Regulations. In addition to ACOP, there exist several Guidance Notes, which again do not have the force of law, nor the impact of ACOP. Guidance Notes have been written to demonstrate, technically, how certain regulations and associated ACOPs may be complied with.