The fact that ionising radiations could cause biological damage became apparent very soon after Roentgen’s discovery of X-rays in 1895 when a researcher using early X-ray tubes developed radiation burns to the skin of his hand. The first case of radiation-induced cancer was reported seven years later. Evidence of the harmful effects of large exposures to ionising ra­diations grew as the use of radiation and radioactive materials spread throughout industry and medicine. Early examples were damage to the hands and cancer among X-ray workers, lung cancer among uranium miners and bone cancer among workers in the radium industry, particularly luminous dial painters. How­ever, the long-term biological significance of smaller, repeated doses of radiation was not widely appreciated until relatively recently and most of our knowledge of these effects has been accumulated since World War 2.

All radiation is absorbed by matter with a transfer of energy from the radiation to the absorber. Radiant heat and light, when absorbed, produce a temperature rise in the absorber. Nuclear radiation differs only in that each particle or photon possesses sufficient energy to cause ionisations in the absorber; thus it is known as ionising radiation. Body cells are about 80<Го water, so radiation absorbed in body tissue will usually cause ionisation of the water in the cells. Ionisation in the cell leads to the formation of chemical species that may damage the cell. If sufficient cells are damaged then the body itself may be affected. The process of radiation damage to cells after interaction with cell water can be characterised in four stages:

(a) Ionisation, typically

H2O radiation H20 + + e“

This process has a reaction time of the order of 10“16 seconds.

(b) Formation of free radicals H20 + H+ + OH*

or H20 + + e" H20′ H* + OH-

H+ and OH’ are ions which are always present in water and so have no significance in this case, but H* and OH* are free radicals and are chemically highly reactive.

(c) The chemical stage The free radicals are able to react with organic molecules in the functional parts of the cell and alter them. This process takes a few seconds.

(d) The biological stage Over a period varying from a few minutes to a few years the biological con­sequences of the chemical changes in the cell be­come apparent.

Alternatively there is the possibility, although with lower probability, that radiation may interact directly with functional organic molecules in the cell, such as DNA, and produce direct damage to them by ion­isation. Biological damage will then become apparent as in (d). The effect of this damage varies with the nature of the damage. It may be so slight as to be insignificant or so subtle that it goes unnoticed. If it is severe enough to affect the function of the cell a reaction may be triggered. Cells have the ability to recover from limited radiation damage by repair of the damaged structure. The repair may be suc­cessful or may result in a functioning cell which is different from the original.

From this we have two effects of radiation damage, the first, where the cell is altered but continues to re­plicate, is called cell transformation. The second, where the cell is unable to divide, is cell death, which re­quires in general a higher radiation dose than the first.

Knowing the ability of cells to repair damage we can make observations about the incidence of cell death. Firstly, cells which are dividing rapidly (e. g., skin cells) will be more susceptible than those which are not (e. g., nerve cells) because there is less time between cell division for repair. Secondly, dose rate as well as total dose will be important because, put simply, at low dose rates repair is better able to keep pace with damage.

The safety considerations in respect of ponds at PWR stations are similar to those at magnox and AGR stations.

In the Sizewell В design, the main fuel storage pond is linked to three bays which are respectively for the purpose of fuel transfer into the pond, loading fuel into the transport flask and preparation of the flask before and after fuel loading. Each bay can be sealed off from the pond and/or adjacent bay by means of stop gates. The flask preparation bay is routinely drained and refilled for flask transfer purposes, and the fuel canal can be drained for maintenance, if necessary.

The fuel, which consists of assemblies of rods of uranium dioxide pellets in zircalloy cladding, is stored in racks which stand on the pond floor.

A range of protective measures is incorporated into the design to control radiation and contamina­tion exposure in normal operation, and to prevent accidents including criticality and damage to the pond containment.

There are interlocks on fuel handling machinery to prevent dropping or collisions during fuel transfer operations. The height to which skips may be raised during transfer is also limited.

Measures to prevent inadvertent pond drainage include the extraction of water from near the pond surface and fitting of anti-syphon devices on the pond water return line. Installed instrumentation in­cludes radiation monitors and high and low water level alarms.

The fuel storage racks incorporate neutron absorb­ing plates, and they are so arranged as to be safe even in the absence of boron from the pond water.

A pond water treatment plant is provided, incor­porating water/coolers, filters and ion exchange resins, together with a boric acid dosing facility.

Skimmers are included for the removal of debris and surface microlayers and an active ventilation sys­tem with filtered extract is provided.

The pond water is dosed to 2000 ppm with boron for criticality control purposes. The water would be maintained at pH near neutral by dosing with lithium hydroxide.

The pond construction is of reinforced concrete with an internal stainless steel liner which is keyed to the concrete. A leakage monitoring system associated with the liner welds is incorporated.

There is now considerable experience in the op­eration of cooling ponds at PWR and other water reactor stations, both in the United States and in other countries. This experience [38] supports the view that good radiological protection standards can be achi­eved and radiation exposure maintained at acceptably low levels.

In the event of an emergency at a nuclear licensed site, the DoE would be responsible for ensuring sup­plies of potable water and for arranging for the dis­posal of any radioactive waste arising from the ac­cident. The DoE has a general interest in any effects such an incident may have on the environment and the Department’s radiochemical inspectors would be available to provide assistance or to give advice.

On the declaration of an emergency standby or alert, the DoE and the local water authorities would be notified by the emergency control organisation of the affected site. A radiochemical inspector would go to the operational support centre and from there would give advice to the water authorities and keep the senior radiochemical inspector in London advised of the situation. The water authority would arrange for duplicate samples of all suspect water sources to be taken, one for analysis by the CEGB and one for analysis by an independent laboratory.

When the initial enrichments have been chosen, fur­ther cost benetits can be secured it the loading pattern itself is arranged as a chequerboard array of ‘high’ and ‘low’ enrichment channels in a large inner-region, together with a third higher enrichment loaded into some of the outer-region channels (the others con­taining vacancy stringers) in order to provide some enrichment flattening of the radial power shape (Fig 3.49). Following replacement of the majority of the vacancies, refuelling is then arranged of the inner — region low’ enrichment channels, punctuated only by the occasional visit to the outer zone to keep the power shape under control. The intent will be to load the centre again with one enrichment, the outer re­gion with another, with possibly an intermediate en­richment region between the two in the interests of radial power shaping (Fig 3.50). Remembering that the inner region originally consisted of a chequerboard array of high and low enrichments, it is clear that the greatest reactivity benefit will be obtained by re-

fuelling the low enrichment (low reactivity) channels firsts leaving the higher enrichment channels to do more work until they too become refuelled later in the cycle. In this way, core reactivity is sustained and the desired irradiation of the initial charge is achieved with- the last stringer of initial fuel being discharged at (or near) the target discharge limit. The overall effect of adopting this strategy is that the CAI at discharge is maximised for all the initial fuel with the added bonus of a significantly reduced re­fuelling rate.

The feed fuel enrichments need to be higher than those of the initial charge in order to compensate for the higher average irradiation level of the fuel at equilibrium and the correspondingly lower average reactivity level. However, the increased average age of the fuel at equilibrium means that the higher rated fuel elements, situated in more central channel posi­tions ‘burn up’ more rapidly than the other elements within the channel, and consequently reactivity and therefore rating declines more rapidly, producing a degree of ‘automatic’ axial power flattening. During the approach to equilibrium, however, the fuel on average is considerably younger and axial power flat­tening is accommodated by introducing higher enrich­ment elements into the outer positions of the fuel stack (positions 1, 7 and 8 usually) during the initial
loading. Such fuel assemblies represent the inner — region low’ channels within the mixed enrichment scheme, whilst the adjacent channels containing fuel uniformly enriched to the same level as the outer elements of the low enrichment channels will therefore constitute the inner ‘high’ positions.

1 Introduction

The British civil nuclear power programme was de­veloped directly from the nuclear weapons programme and the associated radiological risks were always clear­ly recognised. As a result, strict government controls have been placed upon the nuclear industry to ensure that the dangers from ionising radiation to both the general public and to workers in the nuclear installa­tions are reduced to a minimum. Such controls are embodied in various Acts of Parliament and have resulted in safety standards which are far more de­manding than for any other modern industry.

The CEGB as a major nuclear licensee has been governed by such regulatory control since the begin­ning of its nuclear power programme, but has also acted to ensure that its safety standards are main­tained as high as reasonably achievable. Taking into account experience gained from the design, construc­tion and operation of large power reactors, an im­portant part of the CEGB’s nuclear safety policy (and that of its nuclear successor, Nuclear Electric pic) is the need for constant vigilance and understanding of changing circumstances which can affect the safety _ of nuclear generation. In this respect, the safety as­sessment by functional teams at all stages in the de­sign, manufacture, construction and operational phases plays a most important part, with the station man­ager having the delegated authority for the operational safety of his plant and compliance with regulatory requirements. The safety of decommissioning is also becoming increasingly important as the earlier nuclear stations begin to reach the end of their working lives.

In order to provide independent consideration and specialist advice to all levels of management on the nuclear health and safety aspects of the CEGB’s bu­siness, the Health and Safety Department (originally the Nuclear Health and Safety Department) was set up at the start of the civil nuclear programme to provide independent, dispassionate, and objective as­sessment of nuclear safety, and to act as the focal point for discussion and agreement with the regulatory authorities.

From the beginning of its nuclear power programme, the paramount objective of the CEGB has been the attainment of high standards of nuclear safety and

the protection of the public and personnel from harm­ful exposure to radiation due to its nuclear operations. This chapter describes the regulatory controls which govern nuclear licensed sites in the UK. the CEGB’s fundamental safety philosophy, its methods of safety assessment and management, and the way in which it complies with the nuclear regulatory requirements.

1.2 Fundamental safety philosophy

The highest priority is given by the CEGB to the maintenance of nuclear safety standards in order to ensure the radiological protection of both people and the environment and, in particular, to safeguard the public. From an early stage of the development of the first nuclear power programme, the CEGB adopted a policy for nuclear safety which embodied certain fundamental principles, the main features of which are:

• As a result of normal operation of a nuclear power station, no person shall receive doses of radiation in_exce$s of the appropriate limits.

• The exposure of individuals to radiation shall be kept as low as reasonably practicable.

• The collective dose equivalent to operators and the general public as a result of the operation of a nuclear installation shall be kept as low as reason­ably practicable.

• All reasonably practicable steps shall be taken to prevent accidents.

• AH reasonably practicable steps shall be taken to minimise the radiological consequences of any ac­cident.

From these fundamental principles the CEGB has developed a safety philosophy which is based on es­tablishing safety criteria at the beginning of a project against which the developing designs can be assessed. The criteria are comprehensive and cover the stand­ards of engineering to be adopted, the levels of re­dundancy and diversity in the protective system, the treatment of man-made and natural hazards and the means by which quality in design, manufacture, con­struction and operation is assured. The design safety criteria are described in detail later. The overall safety aim is to show that the risks to the station operators and to members of the general public are acceptably small.

As discussed elsewhere in this chapter, the law im­poses radiation dose limits on several categories of person, but principally persons at work are ‘allowed’ to receive 50 mSv per year. However, if they are non-classified workers this drops to 15 mSv per year. Members of the public may receive 5 mSv per year.

The dose to the general public from radiation ori­ginating from the power station is controlled by the positioning of the site boundary fence. The radiation dose to non-classified persons on a licensed site (e. g., administration staff, canteen staff, etc.) are limited by the positioning of the controlled area. The controlled area boundary encompasses the reactor building, al­though with the provision of additional facilities, for instance a laundry building, it may be necessary to have other parts of a site contained within a controlled area fence. Access to the controlled area on a day — to-day basis is restricted to classified personnel and is made via a guardians post. All classified staff must wear a whole-body film badge whilst in the controlled area. All other personnel (e. g., visitors and contractors) are only allowed into the controlled area under specified procedures.

Section 4,3 of this chapter on the Radiological Safety Rules, describes how a nuclear site is zoned in terms of both radiation and contamination levels and how access to the various zones is controlled. This ensures that under normal working conditions that annual dose limits, or any other imposed level, cannot be exceeded.

The site licence for each station requires the CEGB to set up a senior committee to consider any proposal of major safety significance affecting the operating of the station. This committee is known as the Nu­clear Safety Committee (NSC). Even without the li­cence requirement, there would be a need for such a committee to provide authoritative advice to the station manager on major nuclear safety issues and to endorse, if appropriate, any actions he may take. Further, the committee assists in maintaining common standards between stations.

The NSC is chaired by the Director of the Nuclear Operations Support Group and has representatives from the health and safety department, the design department, the research department and two repre­sentatives from outside organisations, British Nuclear Fuels and UKAEA, who provide independent opin­ions. The station manager from the station in ques­tion or his representative is also a member of the committee. It is his attendance which identifies a particular meeting with a particular station, although at a single sitting several managers may be present and proposals from all those stations may be con­sidered. The technical secretary is provided by the Nuclear Operation Support Group and an agreed set of minutes is sent to the HSE within 14 days. The HSE are also supplied with the qualifications training and experience of the members of the NSC and the terms of reference.

The NSC is responsible for providing on behalf of the CEGB, the formal endorsement of any major nuclear safety submission associated with the operat­ing stations and the nuclear laboratories. The consent of the HSE is very rarely sought to any such proposal before the NSC have indicated their agreement.

The committee is free to consider any nuclear safety issue brought before it by the station manager, or any other member. There are, however, a few par­ticular proposals which under the conditions of the site licence must be endorsed by the NSC before the Nil will give their consent. These proposals include the content and any changes to the Operating Rules, and major modifications to plant.

Within the definitions of modifications to plant are changes to any of the safety arguments, and any major repair. These modifications to plant are covered by a procedure, approved by the Nil, which defines the degree of safety significance of any proposal re­quiring the attention of the NSC. Category 1 modi­fications which must be endorsed by the NSC and the HSE before implementation, are those involving a change in the principles upon which the safety case was based, or where an error in conception or implementation could materially affect the risk of a release of radioactive material to the environment. Category 2 modifications, although having safety sig­nificance, do not involve a change in principle and are not likely to have any major impact on the risk of a release of radioactive material. These need the endorsement of the NSC but the HSE only needs to be informed of the proposals; its explicit endorsement is not required. There is machinery for clearing urgent proposals out of committee.

The initial version of the Operating Rules required for fuel loading on the first reactor of a two-reactor station, is approved by the NSC prior to receiving HSE approval. Henceforth, any alteration or suspen­sion of the rules must be considered and endorsed by the committee before being put to the HSE, ex­cept in exceptional and urgent situations. In these exceptional circumstances, the agreement of the Di­rector of Health and Safety is sufficient to allow a change to the rules to be put to the HSE. In all cases, however, the agreement of the HSE to a proposal covering the operating rules is required before it can be implemented.

The nuclear safety committee considers and gives agreement for any experiments or non-routine tests on the reactor except when a station commissioning committee is in existence. The NSC may also be asked to consider any exceptional or unusual occurrence involving nuclear safety and to endorse, or otherwise, recommendations on future actions.

There are certain proposals which require the agree­ment only of those members of the NSC represent­ing the headquarters departments and divisions. These proposals, put forward by the station manager, in­clude changes to identified operating instructions, or to criticality certificates. Identified operating instruc­tions are detailed instructions which augment the in­formation in the Operating Rules. Criticality certifi­cates identify the manner and location of operations involving the handling and storage of enriched fuel.

There are many reasons why improvements to AGR fuel design are constantly being sought. Despite the success so far achieved, the optimum design to meet the longer term requirements of AGR operation has yet to be achieved. We have already seen how large savings in fuel cycle costs can materialise through the attainment of longer fuel life, i. e., higher burn-up. Consequently much effort is being devoted towards this fundamental objective in the hope that AGR fuel will be capable of reaching 21, 24 or even 30 GWd/t at discharge, by the incorporation of burnable poi­son cables or ‘toroids’ within the grid and brace sup­port grooves of more highly-enriched elements. The achievement of greater burn-up, however, must not be realised at the expense of fuel integrity. Therefore, in parallel with any related fuel element modifications, future fuel design and manufacturing objectives will need to make allowances for the implications of in­creased burn-up on the safety and performance aspects of the fuel pins.

Currently, the AGRs w’ithin the UK have not achi­eved their design objective of continuous on-load re­fuelling at full reactor power. The need to regularly reduce reactor power in order to refuel is obviously costly. It is considered that the potential for refuelling at higher reactor powers can only be realised by the development and use of a stronger single-sleeve design of fuel element, which would be capable of with­standing the forces exerted on it by the reactor coolant under conditions of high mass flow. This type of fuel element is known as the Stage 2 (single sleeve) design, as different from the Stage 1 (double sleeve) current design, and its development will undoubtedly be ac­companied by the inclusion of many other age, safety and performance improvements currently under in­vestigation (see Chapter 2).

Spent nuclear fuel from magnox reactors when re­moved is stored in cooling ponds for about 100 days, to allow for fission product heat decay and radio-

wastes is by filtration before discharge, as with liquid wastes from AGR and PWR reactors. Typical annual discharge of liquid effluent from a PWR power station is given by Passant [9j and is shown in Table 4.7.