In the first instance there is an obvious loss of ma­terial from exposed surfaces and thin sections may be wasted away or seriously weakened. The effect of oxidation on a low silicon thin-section material is amply illustrated by Fig 3.73. This shows the oxida­tion present on a jubilee clip removed from a reactor and shows the growth of oxide on all surfaces to­gether with the thinning of the material.

When oxide forms, it occupies a larger volume than the material from which it is derived. The ratio of oxide thickness to that of the original metal from which it is formed is about 2.5 so that the volume change is considerable. If the scale is formed between two bolted or welded components then these are forced apart (oxide jacking) once any initial clearance has been taken up. Similarly, the growth of oxide reduces the clearance between moving components and results in seizure. This straining of the material may lead to failure of the component. Interfacial oxide builds up slightly less quickly than that formed on a free sur­face and the rate of formation does not slow down significantly as the jacking forces increase. The effects of interfacial oxide are illustrated in Fig 3.74.

In 1957, a Treaty establishing the European Atomic Energy Community (Euratom) was agreed by the mem­bers of the European Community. When UK became a member of the European Community in 1973, she also became subject to the provisions of the Treaty. The two Articles of the Treaty that have most effect on the operations of the CEGB are Articles 31 and 37.

Article 31 requires that ‘basic standards shall be laid down within the Community for the protection of the health of workers and the genera! public against the dangers arising from ionising radiations’. A Direc­tive on this subject was issued prior to the UK joining the Community, with a date for compliance of June 1978. This reflected the philosophy encompassed in the International Commission on Radiological Protec­tion Publication No 9 (1966). This publication has been followed by ICRP Publication No 26 (1977) w’hich revises and updates this philosophy and contains re­vised recommendations on numerical values for some dose limits. As a result, the group of experts estab­lished to advise the Commission on the Directive re­commended a new text reflecting the recommendations of ICRP 26, and also proposed a two year extension to the period for implementation. Any proposals for change in the current European Directive are subject to consultation between the Health and Safety Exe­cutive and those likely to be affected by such provi­sions. Government legislation and CEGB Safety Rules, with which nuclear power stations must comply, re­flect. ICRP recommendations.

Article 37 of the Treaty requires each Member State to submit to the Commission ‘such general data re­flecting to any plan for the disposal of radioactive waste in whatever form as will make it possible to determine whether the implementation of such plan is liable to result in radioactive contamination of the water, soil or airspace of another Member State’.

The general data is presented as a submission do­cument to the European Economic Commission by DoE on behalf of the UK Government after consulta­tion with the operator — in this case CEGB. The Commission, after consulting its group of expert ad­visers, must deliver its opinion within 6 months. Pro­vided the opinion is favourable, the plant may then be authorised by the Government Authorising Depart­ments, to begin discharging radioactive effluents.

Because the requirements of Article 37 are not re­trospective, submissions have been presented by DoE only for the more recent CEGB power stations, that is to say, Hinkley Point B, Dungeness B, Hartlepool, Heysham / and 2. These were all favourably received by the EEC.

One of the more common effects of ionising radia­tions, which were apparent from the wartime exposures in Japan and also from the effects of radiotherapy, is a reddening of the skin (erythema) often referred to as ‘radiation burns’. This is often associated with hair loss. Erythema degenerates to desquamation (skin loss) which is termed moist or dry according to the degree of severity and can be directly related to cell ‘death’ (and impaired blood flow). Erythema first ap­pears at skin doses of about 4 Sv, as previously mentioned.

Of course, in real cases of massive exposures, a combination of the above syndromes would appear depending on the dose received. Rapid death from CNS damage would only occur after the most massive doses.

Treatment of acute exposures

There are several types of treatment that increase the chances of survival if the acute dose is not too high. Death from CNS damage appears inevitable after very high doses but the other syndromes can be treated. Infection and fluid imbalance can be treated with broad spectrum antibiotics and fluid replacement.

Haemorrhage can be treated with blood transfu­sions and, with much greater difficulty, by bone mar­row transplants. Even in tissues which have been severely damaged, normal recovery is possible. A few viable cells may be left in the damaged tissue and, provided the individual survives, these cells multiply to repopulate the damaged tissue.

Some permanent damage may be produced which is not life threatening. The skin may suffer permanent loss of hair or sweat glands or, after high doses and resultant ulceration, a permanent scar may be formed. Cataracts do not exhibit repair and are a late effect of acute exposures. Other effects such as cancer (sto­chastic effects) may only become apparent after many years.

Acute exposures may be continuous or intermittent. Both cell loss and repopulation take place at the same time. If repair lags behind damage then the cell po­pulation of sensitive tissues gradually declines. If, however, repair keeps up with damage the tissue may appear normal. Continual tissue damage may give rise to late effects such as cancer.

If an individual receives a dose of radiation over only part of the body the biological response will be largely restricted to that part and the prognosis will be better than for a whole-body irradiation. The effect on the individual will depend very much on whether or not the most important tissues were included in the radiation field. Thus І0 Sv to the abdomen may produce death by GI syndrome but the same dose confined to the feet or hands will produce only lo­calised damage.

Radiation which does not penetrate deeply may pro­duce a similarly muted response. If radioactive ma­terial finds its way into the body via ingestion “or inhalation it may also provide partial-body irradiation if it is concentrated in specific organs, e. g., iodine in the thyroid, strontium and other materials in the bone.

5.1 Licensee’s responsibility

Before a nuclear power station can be constructed or operated within the UK, the prospective operator must obtain the agreement of the Health and Safety Executive. The agreement is given in the form of a

Site Licence as required by the Nuclear Installations Act (1965) as amended. This makes the licensee di­rectly responsible for ensuring that plant is operated in such a way that the risk to the public of death or injury from nuclear hazard is kept within acceptable and previously agreed limits. The licensee is also re­quired to ensure that the operators are not unduly exposed to radiation. He must maintain a record of the radiation history of every classified worker and retain that record for the whole of the workforce. The licensee must prepare plans for dealing with ac­cidents which safeguard the population and involve the organisation of police, ambulance and fire service. These requirements are additional to those aspects covered in the Health and Safety Acts.

Rail mishap

In the event of a ‘Nuclear Flask Emergency’ involving a flask in transit by rail, British Rail will notify the ‘Alerting Officer’, who is a CEGB officer on duty at the National Control Centre.

The alerting officer will contact an appropriate nu­clear establishment for specialist advisers to attend the mishap, nominate a nuclear establishment to co­ordinate the operation and initiate the CEGB HQ notification scheme. The nominated station will gen­erally be a CEGB nuclear power station but may also be BNF Sellafield. A senior member of staff at the nominated station will act as ‘Coordinating Officer’ and will arrange for the duty health physicist to at­tend the scene of the mishap immediately. The ‘Flask Emergency Team’, consisting of health physics and fuel handling personnel, will follow as soon as possi­ble. At the scene of the emergency the duty health physicist and the flask emergency team will:

• Liaise with the British Rail Mishap Officer and officers in charge of the emergency services.

• Arrange controlled access to the flask and wagon.

• Carry out radiological surveys including air sampling.

• Monitor personnel involved in the mishap and institute decontamination measures as required.

• Check the internal pressure of the flask and, if necessary, carry out venting procedures.

• Give advice on any radiological precautions neces­sary during procedures to recover the flask and if necessary during its journey to a safe location.

• Give frequent situation reports to the coordinating officer, the British Rail Mishap Officer and the officers in charge of the emergency services.

British Rail will provide breakdown facilities for any righting and recovery of a displaced flask that may be required.

The coordinating officer will, in consultation with the duty health physicist, decide when the emergency is at an end and will notify all parties concerned.

Road mishap

It is assumed that any flask mishap during road transit will occur en-route either from or to a rail head. The mishap will be reported by the vehicle crew to the dispatching or receiving establishment which will as­sume the role of nominated station. The coordinating officer of the nominated station will decide on the category of the mishap, and arrange for the duty health physicist and the flask emergency team to at­tend the scene of the accident.

The coordinating officer will also:

• Inform the local police and advise that NAIR Stage 2 assistance is not required.

• Inform the local fire brigade, if necessary.

• Inform the alerting officer at National Control and request initiation of the CEGB HQ notification scheme.

• After consultation with the duty health physicist advise the alerting officer when the mishap has ended.

If the flask sealing is impaired, the following organisa­tions should be informed by the coordinating officer:

• Radioactive Materials Transport Division, Depart­ment of Transport.

• Department of the Environment,

• The relevant local authority.

• The relevant local water authority.

• The Chief Constable for the police area in which the mishap has occurred.

The duty health physicist and the flask emergency team will, at the scene of the mishap: [42]

• Carry out radiological surveys including air sam­pling.

• Monitor personnel involved in the mishap and institute decontamination measures as required.

• Check the internal pressure of the flask and, if necessary, carry out venting procedures.

• Give frequent situation reports to the coordinating officer and officers in charge of the emergency services.

• Give advice on any radiological precautions ne­cessary during the recovery of the flask and its delivery to the nominated station.

6.2.1 External health physics support

In certain circumstances a health physicist from an agency, i. e., a location other than a CEGB nuclear power station, e. g., AEE Winfrith, BNF Springfields, Rolls Royce Ltd., may be called to attend by the alerting officer if there is a possibility that this health physicist will reach the scene of the accident well before the duty health physicist from the nominated station. Additional health physics support is also avail­able in the Greater London Area and may be called upon to supplement the initial response at the dis­cretion of the coordinating officer.

Since the AGR fuel element contains both enriched uranium as well as an amount of integral moderator, in theory it is just possible that an uncontrolled mix of fuel and moderator could result in accidental criticality. Because of this, special fuel reactivity cal­culations are performed in which the potential for criticality in all the various circumstances which could prevail during transport and storage of AGR fuel is assessed. These studies, known as ‘Criticality Safety Assessments’, are deliberately pessimised to remove any doubts about the adequacy of the assumptions made, by assuming, for example, artificially high fuel enrichments together with the worst possible configu­rations of elements in storage, and further allowances are made for unusal conditions such as flooding of the fuel stores. Each part of the fuel route is sub­jected to a Criticality Safety Assessment so that suit­able administrative guidelines and controls can be provided for the safe Spelling and storage of the fuel. Safety of the fuel in transport is also appraised in a similar manner and the end-product of all such assessments is the setting of safe working standards. Good housekeeping is of the utmost importance since extraneous moderating materials could be placed ad­jacent to the fuel store. Graphite, in the form of broken sleeves, as well as protective polythene bags and new fuel box packing materials, which are known to contain hydrogen (an excellent moderator), are the most obvious examples. It has been demonstrated that the small quantities of polythene bags and packing materials present are insufficient to cause a criticality hazard, but it is nevertheless considered essential that such materials should not be allowed to accumulate unnecessarily.

The basic objectives of radioactive waste management are to ensure that the system of radiation dose lim­itation recommended in ICRP 26 [6] and expanded in NRPB ASP2 [7] are applied. These are that:

• All practices giving rise to radioactive wastes must be justified, i. e. the need for the practice must be proven in respect of its overall benefit.

• Radiation exposure of individuals and the collective dose to the population arising from radioactive wastes shall be reduced to levels which are as low and reasonably achievable, economic and social fac­tors being taken into account.

• The average effective dose equivalent from all sources excluding natural background radiation and medical procedures, to representative members of a critical group of the general public shall not ex­ceed 5 mSv (0.5 rem) in any one year. The use of a limit of 5 mSv in a year is expected in most cases to result in an average dose rate equivalent to a critical group of less than 1 mSv (0.1 rem) per year of life-long whole body exposure from all sources of radiation. Hence the lifetime whole body ex­posure of an individual is unlikely to exceed 70 mSv (7 rem).

At nuclear power stations in England, applications for the discharge of radioactive wastes is made jointly to the Secretary of State for the Environment and the Minister of Agriculture Fisheries and Food. In Wales and Scotland the control is exercised by the respective Secretaries of State.

The Certificates of Authorisation to discharge ra­dioactive effluent in general are subject to limitations which take the form of attached conditions. These relate to the quantities and character of the radio­active wastes to be discharged, the methods of ex­aminations of the wastes and the returns and records of disposal. A typical authorisation for discharge of liquid radioactive waste from a nuclear power station is given in Fig 4.1.

A nuclear power station may only be constructed and operated in the UK if the licensing authorities are satisfied that the agreed criteria for nuclear safety have been met. In order to convince the authorities that the necessary approvals may be issued, the li­censee instructs the contractor to write a comprehen­sive description of the plant and the way in which it is to be operated. This document, which also contains detailed arguments supported by any necessary calcu­lations demonstrating the safety of the plant under both normal operating and fault conditions, is known as the Station Safety Report, There has been a general improvement over the years in the degree of detail and the quality of Station Safety Reports for succes­sive stations. This reflects not only the increasing at­tention paid by the Licensing Authorities to nuclear safety, but also the development in the philosophy of nuclear safety.

The Station Safety Report has always been pro­duced in three quite distinct stages although for the most recent stations at Heysham 2 and Torness, this approach has been somewhat modified.

A preliminary safety report is produced by the contractors when the contract is first awarded. This version of the report is produced primarily for the purchaser of the station and gives an outline of the intended safety case. The first stage of the Station Safety Report proper, known variously as the Stage 1 Safety Report or the Preconstruction Safety Report (PCSR), is produced and issued to the licensing au­thority in support of a request for their consent to proceed with construction beyond the initial site pre­paration, i. e., beyond completion of the ‘blinding’ stage of the foundation. Because this stage of the report is produced so early, much of the design detail is unavailable and the report lacks the rigor which would be required before agreement to load fuel or raise power could be obtained. The report does, how­ever, contain an indication of the information which will be provided and a general description of the plant to be constructed. Clearly, if a station to be con­structed is of a proven design with little intention to change that design the detail of the Stage l report can be considerably enhanced.

After the issue of the consent to continue construc­tion beyond the foundations and as the design detail is completed, the second stage of the Safety Report, the Stage 2 Safety Report is written. This issue of the report is intended to provide the licensing authorities with sufficient information and argument to enable them to agree to fuel being loaded into the reactor. It still lacks some information, in particular that which will come from commissioning tests or as a result of late modifications to plant, but it nevertheless forms the basis of the final version of the Safety Report.

The Stage 3 Safety Report or Final Station Safety Report is produced as a series of amendments and additions to the Stage 2 report. For the earlier stations the Final Safety Report appeared as two sets of vol­umes, the Stage 2 Safety Report, and the volumes containing the amendment sheets and addenda. Lat­terly, these amendments and addenda have been in­corporated with the Stage 2 Safety Report into a single set of volumes. Whilst the Final Safety Report contains substantial detail, this is distilled from a vast number of detailed project documents which appear in the list of references. The Final Safety Report gives the base line for the design of the station, and the de­finitive safety arguments, when commercial operation is achieved on the second reactor of a two-reactor station. The Final Safety Report for all stations prior to Heysham 2/Torness was therefore frozen at some point in time, agreed with the Licensing Authorities as being about the time when the second reactor nominally achieved full power. It is from this base line that modifications to plant or to the safety arguments are considered by the Nuclear Safety Committee.

For Heysham 2 power station and Torness a slightly different approach has been adopted. In the first place these stations were very similar to those built at Hinkley Point В and Hunterston В and hence the Preconstruction Safety Report (PCSR) was more de­tailed than had been the ca^e for the earlier stations. In the second place, the concept of the Reference Safety Statement (RSS) was implemented. The RSS is a continuously updated compendium of documents which describe at any time the state of the Safety Case. This compendium of documents is initially simply the PCSR. Further submissions, information supplied in letters and minutes of meetings, are pro­gressively added to the documentation and a record kept in a Schedule of the Reference Safety Statement. The report produced for the fuel loading consent at Heysham 2 is the Station Report and is equivalent to the Stage 2 Safety Report for the earlier stations. At the time of its issue, 6 months before the fuel loading consent is required, it largely represents the total content of the RSS. In other words it replaces the PCSR and most of the information in the other sub­missions, letters and minutes of meetings. The RSS is continually added to in the time following the issue of the Station Safety Report and will include results from the commissioning tests, experience during early operation and any modifications to plant. It is in­tended from time to time to incorporate the infor­mation in the RSS into new versions of the Station Safety Report hence maintaining throughout the life of the station a coherent up to date Safety Case. One further point about the Heysham 2/Torness Safety Reports is that they should be suitable for general publication if required. The Safety Reports for the earlier stations included an extensive list of references containing the detailed calculations supporting the case made. It was decided that it would be quite impractical to have this arrangement if the report were to be made generally available, and hence the infor­mation in the detailed documents was distilled into a number of identified references to be included as extra volumes of the Safety Report itself. Thus, the Station Safety Report, but not necessarily the full com­pendium of documents making up the RSS, is in­tended to be essentially self-explanatory.

For the Sizewell В PWR, a PCSR was also pro­duced and made pubiicaily available for the Sizewell В enquiry. This document went into still greater detail than any previous PCSR. On the basis of this docu­ment, together with additional information requested subsequently by the licensing authority (Nil), the Nil issued a Site Licence. This document contained many references to letters sent to the N11 and is to be up­dated that the substance of the safety arguments made in those letters are included in the text. The updated PCSR is also to be published.

During manufacture, construction and commission­ing the PCSR will be modified, amplified and updated into the Final Safety Report which records the state of the plant in the ‘as-built’ condition. In the past, each of the CEGB nuclear power stations has com­prised two reactors and the Final Safety Report has been produced and issued prior to power raising on the second of the two reactors, enabling experience gained in power raising on the first to be included in the report. For a PWR station with only one reactor, it is probable that the Final Safety Report would be required before the Licensing Authority give consent to raise power.

It is on the Final Safety Report that the Operating Rules, Operating Instructions and general requirements for safe operation of the plant will be based, and it is this report which will be provided to the Station Manager as the basis for the safe operation of his station.

It is noteworthy that Sir Frank Lay field in his report following the Sizewell В Public Enquiry stated that І see no reason why the revised PCSR and FSR should not be made pubiicaily available, and as many supporting documents to the PCSR as practi­cable placed on deposit where they can be inspected.’ He also added that The PCSR, its supporting docu­ments and the FSR are technical in nature. It is im­portant that reports are published at a later stage in the licensing process which are suitable for lay public and parliamentary scrutiny. These documents should not be over-simplified, but should explain the principal safety aspects thoroughly and clearly.’ The CEGB has already taken the decision to publish the PCSR and is likely to follow the remainder of Layfield’s advice.

The second fundamental principle which has remained in the rules since outset is the Safety Document sys­tem. The rules state that the forms used for Safety Documents in controlled areas of nuclear licensed sites, shall contain a Radiological Control Document which specifies the radiological precautions which are required.

Under normal operations the dose rates and con­tamination levels of an area will be relatively constant and thus will be zoned as appropriate. There is no need to specify additional precautions to those already in existence in the area. However, if maintenance is carried out on a plant item in the area, then the work itself may significantly increase the hazards and thus it is necessary to specify the radiological precautions necessary on the radiological control document.

Any maintenance work in radiation areas R3 or R4, or contamination areas C2, C3 or C4 must not be undertaken unless it is covered by a Safety Document with Radiological Control Document. Further, access only to R4 or C4 areas must also be covered by a Safety Document with Radiological Control Document.

A Safety Document with Radiological Control Do­cument is issued by a Senior Authorised Person (Nu­clear Radiations), SAP (NR) who is responsible for specifying the radiological precautions. The SAP (NR) may call on the Health Physicist for advice in pre­paring the content of the radiological section at any time if it is thought necessary. However, the rules specify that this advice must be called for when work is to be done in a region which is normally sealed and which may contain loose radioactivity. This advice is issued in the form of a Health Physies Certificate by an Accredited Health Physicist.

QA represents a management control system that the nuclear industry can use to establish and attain safety objectives for nuclear installations, as it produces measures of confidence in their achievement. It must be noted that assurance of quality is a corporate management goal and not the responsibility of a par­ticular QA function or QA group.

QA is a technique of good management that has been implicitly recognised for many years, especially in nuclear power stations, in responding to the sta­tutory requirements of the nuclear site licence. How­ever, the particular emphasis that quality assurance has brought is to make good management explicit in documentation and records that can be examined and assessed. The application of quality assurance to the whole life-cycle of nuclear power stations will involve analysis of arrangements, existing or proposed, to ensure that they provide adequate confidence and that their effective implementation can be verified objectively.