On the declaration of an emergency, the Nil would send one or more inspectors to the affected site and to the operational support centre to assess the situa­tion and the actions taken to restore control.

The Nil would set up its own emergency centre at its headquarters. From information provided by the CEGB and the inspectors on the site and at the OSC, the Nil would make independent assessments on the likely cause of the accident and its conse­quences. Advice based on these assessments would be given to the Health and Safety Executive, government departments and the CEGB, as appropriate.

Having outlined the main concepts underlying the AGR fuel cycle, brief reference will now be made to some of the practicalities of trying to ensure that the principles are adhered to on the operating reactor. It is all very well to conceive the ideal initial loading pattern and subsequent optimum refuelling strategy, but for one reason or another it sometimes becomes necessary to depart from what could be termed ‘nor­mal fuel cycle practice’. For example, the design fuel cycle at Hinkley Point В and Hunterston В assumed that refuelling would take place continuously at a rate of one exchange every four or five days whilst the reactor remained at full power, This has not been possible in practice and a very significant proportion of the earlier re fuellings were carried out in large batches, perhaps 8, 12 or even 20 stringers at a time, with the reactor shutdown. Furthermore, when low power refuelling became possible the batch-refuelling regime was still retained. Such operational practice led to a re-examination of the earlier studies, using the sophisticated techniques of 2- and 3-dimensional com­puter modelling of core physics (i. e., enrichments, burn-up, reactivity, form factor, etc.), and revised schemes of refuelling were evolved based on the batch — refuelling idea. These took into account the different fuel irradiation and therefore reactivity distributions which were produced within the reactor both during the approach to equilibrium and in the later charges.

As we have seen, the Hinkley Point and Hunterston AGRs originally worked to an 18 GWd/t fuel cycle, but currently the reactors are moving away from an 18 GWd/t equilibrium condition to one at 21 GWd/t using fuel of new feed enrichments and with burnable poison. Such a transition also required a completely new study of the fuel cycle with yet another revised optimum refuelling strategy resulting.

The determination of the detailed sequence of re­fuelling is one of the most important end-products of a fuel cycle study and compliance with it is vital to optimum reactor performance both in the short and long term. It predetermines the sequence of re­fuelling to be followed in all the replacement charges at equilibrium and, therefore, whatever the fuel irra­diation discharge limit and irrespective of whether continuous or hatch refuelling is assumed, it is ex­tremely important that the sequence is rigidly followed in practice. Nevertheless there are a few occasions when this cannot be honoured, such as the need to prematurely discharge for a stringer containing ex­perimental fuel (post irradiation examination) in the generic interests of fuel design and development. It is also possible that fuel may fail in service, therefore requiring urgent removal in order to restrict the spread of contamination within the reactor. In both these instances the exchange of relatively ‘young’ fuel for new would not be conducive to either optimum fuel utilisation or reactor performance, and the discon­tinuity produced in the refuelling sequence would re­main in all later charges.

The regulation of safety of civil nuclear power plants in the United Kingdom has been enacted in the main — by successive Nuclear Installations Acts, and amend­ments and regulations since 1959. Responsibility for administering the legislation rested with different Gov­ernment Departments until 1974, when responsibility for nuclear regulation was transferred in January 1975 to the Health and Safety Executive established by the Health and Safety at Work Etc. Act 1974. The Nuclear Installations Act provides very wide regulatory powers and no body or person other than the United Kingdom Atomic Energy Authority or a Government Department may construct or operate a nuclear re­actor without the site being licensed by the Health and Safety Executive. Nuclear licences may only be issued to corporate bodies and, in practice, conditions attached to the licence during the construction and operational stages provide the major system of regu­latory control.

In addition to the national regulatory controls the United Kingdom is, as a member of the European Community, subject to the provisions of the treaty dated 25 March 1957 which established the European Atomic Energv Community (Euratom). The specific

provisions of this treaty are dealt with later in this chapter.

In the United Kingdom the application for a licence to build and operate a nuclear power station under the Nuclear Installations Act 1965 (as amended) is paralleled by an application to the Secretary of State tor Energy to build a power station under the Electric Uighting Act of 1909. This requirement applies to all power stations both conventional and nuclear. The site licence relates to a total area of land which may be occupied by one or more installations. Prior to making a formal application the utility submits a Safety Report to the Nil setting out the safety case under normal and fault operating conditions along with information on the containment system, normal and emergency cooling arrangements, site layout, ra­diation contours, arrangements for dealing with radio­active effluents, and storage and handling of irradiated fuel and high active waste. The objective at this stage is to enable the Nil to make a preliminary assessment of the safety of the proposed plant and the suitability of the site.

If the Nil’s assessment is favourable, formal appli­cation for consent is made to the responsible Minister who will then notify Local Authorities, River Boards, Water Authorities and other such interested parties. There then follows a period of consultation after which it is for the Secretary of State to decide whether or not there should be a Public Inquiry before issuing a site licence. If, however, the Local Authority lodges an objection or there is a sufficient weight of public objection, the Secretary of State must call a Public Inquiry, following which he will make the necessary decision. Once the licence is issued the CEGB will start to build the reactor(s) in compliance with the requirements of the licence, whilst maintaining a full dialogue with, the Regulatory Authorities. Development of the Safety Report into the more detailed aspects of the plant continues throughout the ensuing stages and is dealt with more fully in Section 3.4 of this chapter.

The fundamental requirement in controlling the radia­tion exposure of operating staff during normal opera­tion of the power station is to ensure that doses re­ceived by operators are kept below the limits recom­mended by ICRP and, further, to keep those doses as low as reasonably practicable (ALARP). An annual target figure for the effective dose-equivalent for in­dividual members of the power station staff resulting from normal operation of the power station was chosen by the CEGB to be equal to 10 mSV (1 rem), in the belief that this represents the lower limit of what is reasonably achievable. In addition to the individual dose target a station collective dose-equivalent of 2 man-mSv (0.2 man-rem) per MW(E) installed capacity was also specified.

When the CEGB safety criteria were first developed, the target for the maximum dose to any member of the public from routine liquid and gaseous effluent discharges and direct radiation, was arbitrarily set at 0.25 mSv (0.025 rem). However, this was later changed to 0.17 mSv (0.017 rem) which is one-thirtieth of the ICRP limit for the general public, with a target of 0.05 mSv (0ЮО5 rem) dose from direct radiation.

Radioactive contamination refers simply to the pre­sence of radioactive material (as opposed to radiation) in places we would rather it was not there.

Ideally, if we are able to contain or seal all the radioactive material on a power station so that there is no possibility of escape, then there would be no contamination to worry about. There would still be the possibility of radiation emitted by the contents penetrating the walls of the container causing a pos­sible external radiation hazard if the radiation is pene­trating enough. It is only when radioactive material is not contained that contamination is present. For example in fuel elements, fission product radioiso­topes are constantly being formed (e. g., Sr-90, 1-131, Cs-137 and the radioactive isotopes of the gases Xe and Kr). Also, radioactive activation products may be produced by neutron bombardment of a number of materials in the vessel, as described earlier. Some of these radioisotopes may eventually escape into the carbon dioxide pressure circuit, e. g., a leaking fuel element can, or corrosion of metal containing the ac­tivated isotopes. Thus the gas circuit becomes contami­nated with radioactive gases and particulate activity.

The contaminated gas is contained within the pres­sure circuit, however, if there is a leak somewhere along the line, the gas with its radioactive contents will contaminate the surrounding area, causing air contamination if it remains^airbofne or surface con­tamination if it settles on surfaces, e. g., walls, floors pipes, etc.

^The maximum permissible doses referred to earlier aVe ‘the total doses from external and internal sources. The CBGB has adopted the policy of measuring doses from external radiation, and where the risk of internal intake of radioactive material occurs, personnel are required to follow laid-down procedures for working in contaminated areas and to use respiratory or other protection where necessary.

The chief danger with contamination is the possi­bility of the radioactive material entering the body and emitting radiation inside the body. Radioactive material can enter the body in the following ways:

• Inhalation — contaminated air breathed in through the mouth or nose and into the lungs, where the contamination may settle out and irradiate the lungs or be absorbed into the bloodstream and taken to various parts of the body. Hence the requirement for breathing protection above certain concentra­tions of contamination.

• Ingestion — into the stomach by eating contami­nated food or by transfer of contamination via the fingers to the mouth and subsequent swallowing. Once in the stomach the radioactive material may be absorbed through the intestinal walls into the bloodstream and carried to various parts of the body. This leads to the ban on eating and drinking inside the controlled area.

• Injection — directly into the bloodstream by contamination settling on a skin break, or by cut­ting oneself with a contaminated object. There is a requirement that no one is allowed to enter a contamination zone without medical department permission.

• Absorption — directly through the skin without a break being present. An example is tritiated water which is known to be absorbed through the skin

walls and into the bloodstream. This leads to the

special requirements for tritium protection.

The method of entry and the contamination hazard depends on the physical form of the contaminant. Dry and dusty contamination presents not only a di­rect contact hazard but also may become airborne. Wet contamination is more easily transferred by con­tact, but wetting dry contamination reduces the air­borne hazard and is an effective method of control. Solid contaminated objects may present little hazard if the contamination is fixed, but there is the danger that if the material is worked, e. g., polished, ground, milled or abraded, the contaminant may become air­borne.

Problems of both radiation and contamination con­trol can be largely dealt with by good design. The ultimate principle in design is to consider every source of radiation and contamination and reduce the levels to minimise the risk to personnel and the public. Un­fortunately this costs money and so a cost benefit analysis is usually performed. In this, a balance is struck between any radiation dose saved by increasing the precautions, and the cost of introducing these.

5.8.1 Statutory requirements, standards, CEGB policy and guidance

Externa( requirements

All nuclear site licences include a quality assurance condition, the text of which is given in Appendix B.

Other requirements to establish quality assurance arrangements come from the Department of Transport in respect of the transport of radioactive materials (which are required to comply with the IAEA Trans­port Regulations) and the BNFL Conditions of Ac­ceptance of Radioactive Waste.

Standards

The IAEA have published a code of practice for QA for safety in nuclear power plants (50-C-QA). This is supported by safety guides covering a range of topics including operational QA (50-SG-QA5), records (50-SG-QA2) and auditing (50-SG-QA10).

British Standard BS5882: ‘Specification for a Total

Quality Assurance Programme for Nuclear Installa­tions’ has been adopted by the CEGB as the standard specification for its QA arrangements as the ‘owner’ of nuclear power stations.

British Standard BS5750: ‘Quality Systems’ is ap­

propriate to companies providing materials or services to the CEGB. BS5750 has three parts which are se­lected by consideration of the nature of the service provided by the company rather than the level of QA required. It is important to realise that each contract must adequately specify what is required, and not rely on a company system to produce the required quality automatically.

Nil guidance

When the Nil first imposed a site licence condition requirement for the CEGB to submit formal QA do­cumentation for approval, a guide was produced for the Nil Assessors to use in their consideration of the documentation and for the information of the licensee. This has been reissued as the Nil Guide to Quality Assurance Programmes for Nuclear Instal­lations (NII/R/8/85) Issue 3 — March 1985.

These are constructed in reinforced concrete and with the aid of two pond gates can be separated into three bays — a flask bay for the storage of road transport flasks and two skip bays. The latter contain rectangular box-shaped stainless steel storage skips divided into compartments by fixed boron steel in­serts, so that each is capable of holding 20 unbottled or 12 bottled fuel elements.

Each. skip bay is provided with a reception tube which is placed at the opening of the appropriate IFD discharge tube to receive the fuel elements as they arrive from, the cell. An element washing system is provided in conjunction with the reception tubes, the objective of which is to restrict the spread of par­ticulate matter to the main ponds. Deionised water is used in the ponds and is dosed with boron in the form of boric acid. It is continuously cooled and cleaned by circulation through specially provided cool­ing and filtration systems whilst the element washing system uses its own independent cleaning loop. Since the fuel still generates heat during pond storage, the water acts as a coolant as well as a radiological shield and unacceptable radiation levels are further prevented by handling equipment design features which prevent fuel being raised too high during lifting operations.

Members of the public are subject to radiation expo­sure as a consequence of the radioactive discharges. Exposure from gaseous effluents arises from:

• Direct radiation from the gaseous plume passing overhead.

• Ingested activity via the agricultural food chain from deposited material.

• Inhaled materials.

Public exposure from liquid effluents is through:

• Ingestion of radioactivity via the sea food and fish pathway,

• External exposure from occupation of beaches and shorelines as a result of radioactive material de­position.

The radiation exposure of the public from liquid and gaseous effluents discharged from nuclear power sta­tions is very low. Typically, the annual exposure of a member of the critical group (the most highly ex­posed members of the public at any particular lo­cation) from liquid and gaseous effluents discharged by a magnox power station of the early type is given by Pepper [11] to be 0.11 mSv (0.011 rem). This dose may be compared with the average annual UK expo­sure from natural sources of 1.86 mSv (0.186 rem) and the International Commission on Radiological

Protection (ICRP) recommended dose limit for members of the public of 5 mSv (0.5 rem) per year.

Solid wastes which are not readily dispersed to the environment when buried on land and/or discharged as packages to sea, give rise to negligible radiation exposure of the public.

For a symmetric fault, the reactivity increase is uni­form over the core, becoming modified as tempera­tures increase as a result of the effect of the fuel and moderator temperature coefficients of reactivity. The temperatures increase most rapidly in the chan­nels with the highest powers. The protection may involve fuel element thermocouples, channel gas outlet thermocouples or neutron flux measuring instrumen­tation. Which of these lines of protection is effec­tive depends on the initial starting point of the fault in terms of absolute reactor power, the channel-by­channel power distribution and the rate of the reac­tivity release.

Within the CEGB, irradiated fuel flask transport op­erations are carried out by the nuclear power stations and by Berkeley Nuclear Laboratories. The CEGB Nuclear Operations Support Group (NOSG) co-ordi­nates ‘user’ requirements, controlling flask movements, advising on operational procedures and identifying requirements for new or improved plant. The design and procurement of fuel flasks to the requirements specified by NOSG, and the preparation of safety cases on which applications for competent authority approval are based, is the responsibility of the Gen­eration Development and Construction Division.

The CEGB’s Health and Safety Department is re­sponsible for advice and assessment at all stages of the transport operation, in order to ensure that the required safety standards are being achieved. They also liaise with the competent authority and apply for and negotiate approval certificates.

Research in support of transport is carried out by the Technology Planning and Research Division.

Transmission Division provides independent inspec­tion as part of the CEGB’s overall quality assurance arrangements and in particular through its regional production, inspection and testing field services (PITFS) offices, during the manufacture and maintenance of fuel flasks. The Structures Testing Centre at Cheddar carries out regulatory compliance drop tests on flasks and flask scale models.

Several external organisations are involved in the CEGB’s irradiated fuel transport operations. These include British Rail, which in addition to being the main carrier, also supplies, to the requirements spe­cified by NOSG, the special rail wagons called flatrols used for transporting flasks. British Nuclear Fuels Ltd (BNFL) carry out the routine sequence of re­ceipt, discharge, inspection and return of fuel flasks, and they also carry out routine and non-routine flask maintenance.