The primary purpose of the Act is to ensure legis­lative control over radioactive waste. Since radioac­tive wastes arise from radioactive material, the Act extends to radioactive substances as well as waste. Such control ensures that radioactive wastes are not produced unnecessarily. Associated radiological safety such as the health protection of workers exposed to radioactive material and wastes or the transport of radioactive material and wastes is covered by other legislation.

2.5.1 Nature of controls

The Act requires all premises where radioactive ma­terials are used, to be registered unless an Exemption Order made under the Act applies. There is a special registration procedure where mobile radioactive appa­ratus such as industrial radiography equipment is moved from premises to premises.

In a similar manner to registration, no radioactive waste may be accumulated on any premises or dis­posed on or from any premises except in accordance with an authorisation issued under the Act unless an Exemption Order applies.

Radioactive material and radioactive waste are de­fined in the Act which includes exemptions or orders made under it where the amount of radioactive material is too small to justify registration or authorisation.

Nuclear power stations are on sites licensed under the Nuclear Installations Act 1965 (as amended) and exemption is provided under the Act from registration for the keeping and use of radioactive materials at such sites. This exemption is in recognition of the legisla­tion relating to control and surveillance of operations at nuclear licensed sites. However the disposal of ra­dioactive wastes at these sites is subject to the Act,

The quantitative estimation of risk, known as proba­bilistic risk assessment (PRA) or as probabilistic safety assessment (PSA) is still in its infancy, but is increas­ingly becoming accepted as a valuable input to the judgement of whether a reactor design is acceptably safe or not. However, the quantification of risk is not yet a formal requirement of CEGB’s design safety criteria nor is it a licensing requirement.

The use of probabilistic techniques in studying re­actor safety is still the subject of much debate because of the difficulties with the treatment of uncertainties, systems interactions, human factors and the tack of data in specific areas such as hazards and the failure ot the major structural items of plant. In spite of the known shortcomings it is now becoming widely accepted that probabilistic analyses are a valuable and essential element of a reactor safety study, although the specific number which comes out of such analysis should not be regarded as an absolute value. The results of such an analysis can be used as an aid to the judgement of acceptability and provide a guide as to the relative safety ot two different reactor designs, or two safeguard systems, so lone as the analyses have been carried out in a consistent manner.

Underlying the CEGB’s design safety criteria there is the judgement that a level of risk of fatality of 10"6 per year to individual members of the public as a consequence of the operation of a nuclear power station is acceptable. This is in view of the benefits derived by the nation from the production of electri­city by this means, and on the basis that such a risk is negligible when compared with the risks to which we are all subject in everyday living. It is a risk which is also comparable to (or better than) that associated with those non-nuclear industries which the public regard as very safe. From the results of studies, it is also judged that a station which meets this individual risk criterion also leads to an acceptable social risk.

Essentially, risk is the product of frequency and consequences. Thus, for example, the risk (R) of fatali­ty of a given individual for a given potential accident sequence is the product of the estimated frequency of that accident sequence (F) and the probability (C) that the radiological dose received would lead to the fatality of that individual, namely:

R = F x C

To arrive at the total risk, all potential accident se­quences have to be taken into account, each sequence having its own frequency value and consequence val­ue, and these individual contributions to risk then summated.

The concept is simple, but in practice estimation of risk is complex. It proceeds in three distinct steps. First, the frequencies of the sequences are estimated (known as a level 1 PRA). Secondly, the quantity and composition of the radioactive material released to atmosphere is estimated (a level 2 PRA) and thirdly the health effect of this release on the public is es­timated (a level 3 PRA).

A level 1 PRA is restricted to determining the fre­quency of accident sequences. Such an assessment is also necessary to determine, for example, whether the CEGB’s target of 10-6 per year for the frequency of uncontrolled releases of radioactive material (see Section 3.2 of this chapter) is met. The analysis uti­lises the techniques of fault and event trees.

The level 2 PRA extends that of level 1 to deter­mine the quantity and composition of the radioactive material released to atmosphere (the source term). It includes transient analysis to determine how fuel and clad temperatures would vary, determination of the quantities and type of fission products released from failed clad, determination of whether these have a pathway to atmosphere (e. g., through open relief valves or by containment leakage) and estimation of the time, duration and height of release.

Finally, in the level 3 PRA, the ways in which that activity can reach members of the public has to be determined which, inter alia, depends on the weather conditions that might prevail at the time, and on the particular habits of individuals in the locality (what food they eat, etc.).

To carry out such calculations for each and every fault sequence would be prohibitive and, in practice, simplifying and generally pessimistic assumptions are made. Similar fault sequences are grouped together and derived radiological source terms are similarly grouped.

The faults studied are not restricted to those within the design basis, i. e., those for which engineered safe­guards systems are provided, but extends to very low frequency events which would lead, for example, to core meltdown. Study of these very low frequency events (degraded core analysis) necessitates assessment of the progression of the fault as the core melts and the molten material interacts with structural material such as the pressure vessel and subsequently the con­crete containment. Uncertainty exists in these complex phenomena. Also allowance for operator action, bene­ficial or counterproductive, cannot yet be fully taken into account, and some inadvertent omissions in the fault sequences considered cannot be discounted. There are also some deliberate omissions in the assessment, such as the effects of external hazards for which data is sparse.

Despite these uncertainties, the risk assessment for Sizewell В supports the judgement that the design achieves a high level of safety. The estimated risk of fatality for an individual at the site fence is about 10 ~s per year from accidents. It is considered incon­ceivable that the known shortcomings of probabilistic risk assessment could completely erode away the large margin that exists between this calculated risk and the CEGB’s underlying judgement that a risk of 10~6 per year is an acceptable value.

The original Safety Rules (Radiological) applied to nuclear sites, and were also implemented by the then South Western Region at Bedminster Down; the Non — Destructive Testing Rules applied at all the CEGB locations, including the nuclear sites. A difficulty arose at one time with respect to a site licence requirement for the need for safety rules, in that (on nuclear construction sites) the CEGB does not assume respon­sibility for the site until fuel loading. In this case therefore, the safety rules cannot be enforced but the licence required that radiological safety rules be ap­plied. To overcome this difficulty, a set of rules entitled ‘Rules for the control of exposure to ionising radiations at nuclear construction sites prior to fuel loading’ were introduced by the Health and Safety Department. In essence, the function of these rules was to protect the interests of the CEGB’s staff on these sites with respect to the practices of contractors radiographers.

As the combined Radiological Safety Rules reflect the requirements of the Ionising Radiations Regu­lations 1985, which apply equally to all the CEGB’s premises, the rules apply equally to these locations also.

The objective of QA in commissioning is to provide measures of confidence in the correct functioning of the installed plant through the demonstration of con­formance of the plant to the designer’s intent by the results of a test programme.

The experience of the CEGB in the commissioning of plant is integrated in the CEGB Plant Completion and Station Commissioning Procedure.

QA documentation for operational nuclear power stations

The CEGB QA arrangements at operating nuclear stations are based on British Standard BS5882 and the Corporate QA Guides as brought together in the CEGB Quality Memorandum QM(0)1. QM(0)1 gives guidance on the content and preparation of QA docu­mentation at operational locations. It is not restricted to nuclear power stations or the nuclear safety related activities on nuclear stations, but recognises some of the additional safety related requirements.

QM(0)1 has been accepted as the in-house standard for operational QA programmes. It contains:

Part 1 — Introduction and explanatory para­

graphs plus BS5882.

Part 2 — Guidance on the application of the principles of BS5882.

Part 3 — Guidance on the preparation of an

Operation QA Programme.

The programme covers both nuclear and conventional activities.

As power stations built prior to Heysham 2 and Torness were not designed, constructed and commis­sioned under formal QA arrangements, the applica­tion of QA to the present stations involves the review of existing management practices and procedures, the preparation of new top tier documents and the modi­fication of existing documentation into an auditable system.

The documentation of the stations’ QA arrange­ments consists of three tiers of QA documentation:

• Top tier definition of management policy and com­mitments.

• Middle tier description of the means by which the established policies and commitments are to be achieved.

• Lower tier provision of instructions for carrying out specific tasks and recording data or results.

The top tier document will be the station operation QA programme.

The middle tier documents will comprise:

• Departmental manuals which index the procedures and practices within departments to conform to the QA programme.

• Quality plans which identify task-orientated docu­mentation associated with the control and verifi­cation of activities, especially when technically or organisationally complex.

The lower tier documents will be in place on most stations in the form of station procedures/instructions/ standing orders, station operating instructions or their equivalent, plant item operating and maintenance in­structions. The review process inherent in the appli­cation of QA will identify omissions and shortcomings, require revision of existing documents and the pre­paration and issue of new documents.

Decommissioning

The application of QA to the decommissioning of nuclear power stations will provide measures of con­fidence that at each stage of decommissioning the safety of the plant and the environment meets the required standards. For the latest PWR plants, the suppliers are required to consider decommissioning aspects in the design of the plant.

The preparation of the QA programme for decom­missioning will be the responsibility of the owner of the plant. However, the period of application of the programme will exceed that for any other phase of the plant’s life, and it is highly probable that the ownership of the decommissioned plant will change during the decommissioning phase. Special emphasis will be necessary on the clear definition of the re­sponsibilities within the phase to ensure that, in any transfer of ownership, continuity is maintained.

The main sources of radioactive waste are shown in Fig 3.56 and are made up of gases, liquids and solids. The most active of these wastes is the spent irra­diated fuel which is returned to the BNFL plant at Sella field, Cumbria. At Sellafield, the radioactive con­stituents are chemically separated from the bulk of the material (uranium) and are stored in solution in cooled vtainless steel tubes. The remaining waste materials arc dealt with on site.

The gaseous wastes are solely derived from the dis­charge or leakage of the carbon dioxide coolant gas. Generally gas leaks arise from leakage past seals. The Co is collected by ventilation systems and after pass — ,nt through absolute filters to remove particulate mat — tcr 11 о discharged to the atmosphere. The volume afid activity of these discharges are carefully moni­tored to control the activity release and to detect any unusual increase in levels which would demand an immediate investigation as to the source and cause. There is also some controlled discharge of gas from the blowdown of reactor gas, either from a require­ment to depressurise the reactor or for control of gas composition, and from the necessity to depressurise the fuelling machine at the end of its cycle prior to the discharge of fuel. In either case strict controls are enforced to ensure that the discharge conforms to statutory regulations.

The liquid wastes are largely derived from the cool­ing pond water treatment plant, and via drains from change room and laundry wash-water facilities. These are collected in the active effluent plant and after treatment, which is usually pH adjustment, it is passed to a final delay tank for analysis before disposal. If necessary, a time delay may be employed for radio­activity decay or water dilution to control activity release to the environment.

The bulk of the solid wastes is derived from the fuel elements in the form of magnox, steel and graph­ite. This material is stored on site in specially-con­structed above or below ground vaults. Similar vaults are used to contain solid wastes derived from main­tenance procedures on active or contaminated equip­ment. Spent resins and sludges from the active water treatment plant are stored in shielded tanks. A start has been made to contain some of these solid wastes in approved packages and to dispatch them to the BNFL sites at Drigg, Cumbria for trench burial.

The Regulations are arranged into nine parts and ten schedules, each dealing with specific areas, principles and practices.

part 1: Interpretation and General

This part, as implied by its title, deals with the gen­eral aspects of the legislation and definition of terms, etc. The most important here are Regulations 2, Interpretation, defining the meaning of phrases, words and so on; Regulation 4, which says that employers must co-operate with each other to ensure that the Regulations can be complied with; and Regulation 5 requiring notification to the HSE, by an employer, that work with radiation is being undertaken.

Regulation 4 is particularly important when the CEGB is employing contract labour. Such contractors may not have previous experience of work with radia­tion and it is important that both the CEGB and the contractor understand their mutual responsibilities.

Part 2: Dose limitation

This part contains only two regulations, but these are perhaps the most important as they deal with dose limitation. Not only must doses be kept below the limits set out in Schedule 1, but all ‘reasonably prac­ticable steps’ must be taken to restrict radiation ex­posure. This latter aspect is known as the ALARP principle, meaning as low as reasonably practicable, and is taken as being equivalent to the ICRP principle that all doses should be kept as low as reasonably achievable (ALARA), economic and social factors being taken into account

The words ‘reasonably practicable’ have been used throughout the regulations, rather than ‘reasonably achievable’ as the interpretation of it has been es­tablished by case law. Regulation 28 deems that an ALARP investigation must be carried out by the em­ployer if the dose to an employee exceeds three-tenths of any dose limit. This means that for each case where this occurs, the doses received by the workers have to be justified and demonstrated as being ALARP, It should be noted that the same dose limits apply to all persons at work but that some persons may be regarded as not working with radiation. In these cases, for example, administrative staff at a nuclear site, the ACOP states that it is unlikely that their doses would be ALARP if the dose limit exceeds that for members of the public, І. e., one-tenth of the occupa­tional dose limits.

Part 3: Regulation of Work with Ionising Radiation

This part deals with certain administrative arrange­ments to be undertaken by the employer in respect of the nature of the work carried out. Regulation 8 re­quires the designation of controlled and supervised areas. A controlled area is one in which it may be possible for a worker to exceed three-tenths of the dose limit; a supervised area is one in which it may­be possible to exceed one-tenth of the dose limit. On nuclear power stations, the reactor building complex is designated the controlled area, whilst the supervised area is the area within the site fence. Note that it is usual to set the boundaries for such areas where it is physically practicable, rather than the exact dose rate corresponding to the area boundaries, as long as these could not lead to breaches of the requirements. This usually also applies to radiographic work at any location.

Regulation 9 requires the employer to designate as classified persons, those employees who might exceed three-tenths of a relevant dose limit. Here the duty is on the employer to fulfil this requirement. How­ever, for contract labour working on a nuclear li­censed site, Regulation 4 is very relevant as previously mentioned.

Regulations 10 and 11 require the appointment of radiation protection advisers (RPA) and radiation pro­tection supervisors (RPS). The RPA is an independent adviser to the employer and on a nuclear licensed site would be an accredited health physicist. The RPS, as the name suggests, has a more supervisory role and is usually the Senior Authorised Person (NR) on a nuclear site.

Regulation 11 also requires that local rules be draw-n up by the employer, and in the CEGB this require­ment is fulfilled by the Safety Rules (Radiological). However, additional procedures may be enforced at particular locations which would be regarded as local rules, as defined.

Regulation 12 requires the employer to instruct and train the employees so that they may carry out their work in accordance with the Regulations. Also in­cluded, is the requirement to inform those who work with radiation of the health risks associated with the work and the precautions to be taken.

The faults associated with the AGR are very similar to those considered for the magnox reactors. The physical characteristics of the AGR are somewhat dif — terent, and different computer codes are used. The fuel is uranium dioxide in the form of hollow cylin­drical pellets about 1 cm tong and 1 cm diameter.

These are contained within stainless steel tubes about 1 m long. Thirty-six of these steel tubes or pins are mounted in three concentric rings inside a cylindrical graphite sleeve, a little longer than the pins. The complete unit, sleeve and pins is a single fuel element.

Usually, eight elements are coupled together by a tie bar running through the axis of the element and joined to an upper unit composed of a gag unit, shield plugs and a pressure-containing closure to make up a stringer. The stringer is loaded into a single channel of the reactor. There are about 300 similar channels. The uranium is enriched in the isotope U-235 to overcome the increased neutron absorption of the stainless steel of the clad compared with magnox. As a further refinement, burnable poisons contained in toroids incorporated in the graphite sleeve have been added in the latest reactors. In this case higher en­riched uranium is used, permitting longer irradiation times, thus reducing the amount of refuelling neces­sary and improving the economics of the reactor. The moderator is graphite, although of a slightly different type to that used in magnox reactors. As the fuel is irradiated, fission products are produced, some of which escape from the molecular structure in the form of a gas causing a build-up of pressure within the pin. The fault studies for the AGR are largely con­cerned with the risk that these gaseous fission pro­ducts may be released into the coolant gas and from there into the environment, and clearly this will occur if the clad melts. It will also occur if the internal gas pressure exceeds the external pressure to an ex­tent where the pin fails. Fission products may also be released if fuel is damaged during refuelling if, for instance, it is dropped. In this case the fault studies examine the measures which need to be taken in terms of the reactor power at which the refuelling may be done and the reactor tripping arrangement, to limit the consequential release of fission products to an acceptable level.

The fault studies follow a similar pattern to that described for the magnox reactors with some refine­ments. The peak limiting clad temperature is taken to be 1350°C which is about 40°C below the actual melt­ing point. When investigating depressurisation faults it is necessary to calculate the fission product gas pressure within the pin and the strength of the clad, both of which are functions of the irradiation history of the elements of the channel being considered and of the transient temperatures. From all of these studies, the limits for full power operation and the settings of the protection equipment are derived.

One further aspect of fault study work is concerned with the handling and storage of enriched fuel out­side the reactor core. The quantity of fuel required to form a critical assembly depends upon its enrich­ment, the type and quantity of moderator available and the configuration of the fuel pellets. When han­dling fuel outside the reactor, it is conceivable that damage may occur which rearranges the pellet con­figuration or that extraneous moderating material such as oil or water may be introduced. The design of the fuel storage facilities and of the fuel route is arranged to avoid such incidents, but studies are carried out to assess the potential reactivity state for any assem­bly under all credible configurations. This study is required to show a kclr of less than 0.95 and the results are assessed and agreed by a ‘criticality panel’ of experts. Finally, a certificate is drawn up for each area where enriched fuel may be held or handled and for all operations specifying the area, the allowable quantity of fuel and the nature of the operations. These certificates are called Criticality Certificates and, after agreement by the criticality panel, are approved by those Nuclear Safety Committee members from the headquarters departments and divisions.

After removal from the reactor, the spent fuel ele­ments are stored for a period in order to allow some of the short-lived radioactivity to decay prior to their transport off site for reprocessing.

Of the CEGB’s eight magnox power stations, seven employ water filled ponds for fuel storage, the eighth station, Wylfa, uses dry storage facilities. All the AGR stations operating or under construction also have ponds.

The pond water provides shielding against radiation from the fuel and also allows the heat generated by the decay of the radioactive fission products to be safely removed.

On receipt in the pond, the fuel elements are placed into open-topped skips and transferred to the main pond storage area. The skips have holes in the sides and base to allow water circulation. Most of the fuel is stored wet, i. e., the elements are in contact with water. However, a small number of elements for post­irradiation examination (PIE) are placed into dry sealed bottles before pond storage.

At the end of its storage period, each skip of fuel is prepared for transfer to the flask loading area for loading into the fuel transport flask. Magnox fuel is normally desplittered (or delugged) at or before this stage in order to utilise transport skip capacity to the full. At most magnox stations and all AGR stations flask loading is carried out in the pond itself. How­ever, at two magnox stations, the flask is ‘dry loaded’ in a shielded area outside the pond.

The main safety aspects to be considered in respect of cooling ponds are those of radiation and contami­nation control, containment and nuclear criticality.

In the event of an emergency, action would be taken to prevent emergency reference levels (ERL) of ra­diation dose being exceeded. However, experience at Three Mile Island showed that the public would no longer be satisfied with the information that they had not been exposed to a ‘significant level of radiation’. It was therefore decided that, for the recovery of public confidence following an incident, the CEGB should provide the best estimates of individual ef­fective dose equivalent and collective effective dose equivalent at all stages of an emergency and at levels well below the ERL. The following procedures have been adopted to make the necessary population dose assessment:

• The minimum level of individual effective dose equivalent that would be assessed is 0.5 mSv, i. e., one-tenth of the annua! dose limit for members of the public. [41] power station. The readings of these monitors are telemetered to the site emergency control centre and would provide a time profile of any release of airborne radioactivity. This information would assist in the interpolation and extrapolation of the inhalation and deposition measurements made by the off-site survey teams.

• An emergency dose assessment service has been established at Berkeley Nuclear Laboratories. This team of specialists would make an assessment of the effective dose equivalent from data supplied from the OSC by telephone teleprinter and facsimile transmission.

6.3.1 Fuel cycle

Higher coolant temperatures and ratings within the AGR, consistent with the requirement for elevated steam temperatures and pressures in the boilers, has led to the adoption of oxide fuel contained within a stainless steel clad. In order to compensate for the increased neutron absorption characteristics of the steel, use is made of enriched uranium dioxide fuel, a change which is uniquely responsible for the relative complexity of the AGR fuel cycle and the various fuel management considerations which relate to it. The AGR is designed for continuous on-load refuelling and, like magnox, will eventually contain fuel at all stages of its irradiation life (i. e., age). However, any similarities between the two fuel cycles ends here. As we have seen earlier, the power in a nuclear reactor emanates from the fission rate within the fuel, and the capability that the fuel has for producing fissions and therefore free neutrons is known as its reacti­vity. The changes in material composition of enriched AGR fuel as irradiation proceeds, produces a reacti­vity variation over life which differs markedly from that of its magnox predecessor, in which the trans­mutation during irradiation of U-238 to the more effective fissile isotope Pu-239 helps to maintain core reactivity. For the AGR fuel currently in use in the CEGB’s reactors, enriched typically from 1 to 3 w/o of U-235, the depletion in U-235 content which takes place during irradiation swamps any beneficial effect of build-up of Pu-239, with the overall result that reactivity in the fuel, and therefore rating (i. e., power) falls continuously from start of life to eventual dis­charge, the fall-off being further enhanced by the build-up of other ‘poisons’ (neutron absorbers) with­in the fuel. Indeed AGR fuel changes during its life from being a net producer to a net absorber of neutrons. Absolute values of reactivity at any irradia­tion will depend only upon the initial fuel enrichment — the higher the enrichment level (i. e., concentra­tion of U-235 atoms) — the higher the reactivity. It follows, therefore, that since the fuel progressively loses reactivity it eventually poisons the reactor in which it resides, so the timing of its replacement needs to be controlled in practice by working to an upper limit on fuel age {i. e., discharge irradiation). Since the increased proportion of U-235 in the fuel together with the Pu-239 is conducive to longer life, the earlier AGR fuel designs still in use are restricted, on grounds of reactivity only, to a maximum channel average irradiation (CAI) at discharge of 18 GWd/t, to be compared with around 5 GWd/t for magnox fuel (Fig 3.47). No other fuel element endurance pro­blem was thought to be more limiting than the main­tenance of adequate reactivity, and indeed this ori­ginal design duty is now well proven following the considerable manufacturing and operational experience obtained. Furthermore, fuel capable of achieving 21 GWd/t at discharge is already being loaded into the reactors at Hinkley Point В and Hunterston В, and physics calculations have since been performed in order to evaluate AGR feed fuel enrichment levels appropriate to the even higher discharge irradiations of 24, 27 and 30 GWd/t. Although it is hoped that such high irradiations will be achievable through evo­lution of the current design of fuel element, continued research and development will be necessary in other areas.

F’t.. 3.47 Variation of channel power with irradiation
for AGR and magnox fuel

The reactivity "age relationship for enriched AGR fuel
determines that fuel rating also exhibits a gradual
decline with irradiation, falling by some ЗО^о over life.
B> comparison, natural uranium magnox fuel, despite
the beneficial effects of build-up of Pu-239, is
removed from the reactor much earlier in life.
However, in calendar time, the residence of AGR and
magnox fuel is comparable at 4 to 5 years.

Regular refuelling is of course essential in order to both maintain core reactivity and remove fuel which has reached its discharge limit. However, an important consequence of the attainment of very high discharge irradiations, particularly when coupled with the very much smaller AGR core size compared to magnox, is that the replacement of spent fuel dur­ing the refuelling process gives rise to large local ‘swings’ in reactivity, and therefore power. All of these changes, whether age or refuelling induced, must be anticipated and controlled as and when they be­come apparent on the running reactor.