The third major group of faults which need to be considered are those where the coolant flow fails or the feedwater flow to the boilers fails. In either case, there is a loss of the capability for removing the heat generated in the reactor core.

Firstly, consider gas circulator failure. If the sup­plies to gas circulators are lost, the motors run down at a rate depending on the inertia of the system. As soon as a mismatch occurs between the heat gen­erated in the core and the heat removal capability, channel gas outlet and fuel temperatures rise to a point where a reactor trip is initiated. However, the reactor is usually tripped on loss of volts to the gas circulator motor, but it is the trip on high tempera­ture which leads to the larger transient. Gas circula­tion is restored post-trip using alternative electrical supplies to subsidiary motors known as pony motors.

In the case of the failure of feedwater flow to the boiler, due either to a fault in boiler feed pumps or a breach of the main feed lines, the first effect is a rise in boiler gas outlet temperature followed by a rise in channel gas inlet temperature and, hence, in fuel can and channel gas outlet temperature leading finally to a reactor trip. It is, however, also of par­ticular importance in this case to study the structural temperatures about the diagrid and the core support at the bottom of the core. Excessive temperatures or temperature gradients in this region could cause dam­age which, although unlikely to cause an immediate hazard in terms of the release of radioactive material, could jeopardise the continued operation of the re­actor as a power-producing unit. Following the reactor trip emergency feed is restored using alternative pumps provided for the purpose.

There are a number of other faults which are con­sidered in the studies but they usually revert to one of those already described. They differ only in the parameter used to trip the reactor and the time from the start of the fault condition when the trip actually occurs. In this category are such items as loss of main grid supplies, tripping of the main turbine, and loss of less than all the main gas circulators.

The CEGB evidence to the Sizewell В Public Enquiry included an assessment of safety aspects of the trans­port of irradiated PWR fuel [31]. The following notes are based on that assessment.

Consideration of the reactor refuelling pattern and of the capacity of the transport flasks, shows that the frequency of fuel movements from a PWR power station would be small, e. g., six to ten per year, compared say to the 50 or so journeys per year from a magnox station. There is considerable on-site fuel storage capacity, which gives operational flexibility, and it is a requirement that fuel for reprocessing should have a decay period of at least five years.

In view of the above, and of the design flexibility in the arrangements for flask handling at a PWR plant, the choice of flask design would not need to be made at an early stage.

There is a considerable amount of experience of transporting PWR and BWR fuel in the UK, mainly by BNFL and associated companies. There are many different flask designs, including ‘dry’ flasks in which inert gas rather than water is used as a heat transfer medium. PWR flasks are cylindrical rather than cu­boid in shape, the lid and base sections being pro­tected by shock absorbers. Like their magnox and AGR counterparts, they are massive and relatively simple structures with valves for venting and filling. The lids are secured by numerous large diameter high tensile bolts, and sealed by elastomeric О-rings. Flasks are subject to a regular programme of maintenance.

Current practice for flask loading and pre-despatch procedures is similar in principle to those for magnox and AGR fuel transport. However, in the case of some wet flasks the PWR fuel is transported in a multi­element bottle (MEB) rather than as individual elements. In the case of dry flasks, the flask is filled with water before fuel loading and after fuel load­ing the water is removed from the flask by vacuum pumping.

From present experience it can be said that fuel transport from a PWR power station would fully meet the requirements of the IAEA Regulations, in par­ticular those in respect of nuclear criticality, heat dispersion, containment and radiation shielding.

In respect of the last mentioned, estimates have been made of the annual radiation doses to transport workers and to members of the public as a result of routine fuel transport from a PWR power station. It is estimated that a dose to an individual member of the public living close to a rail marshalling yard would be about 0.75 Sv/year (0.075 mrem/year). For transport workers the maximum individual dose would be about 20 Sv/year (2.1 mrem/year). These dose rates are negligible compared, for example, to the average natural background radiation level in the UK of 1870 Sv/year (187 mrem/year). The corre­sponding collective radiation doses to members of the public living close to marshalling yards, and to marshalling yard workers are estimated as about 1.4 x 104 man-Sv/year (0.0144 mrem/year) and 1.2 x 10~6 man Sv/year (1.2 x 10 ~4 mrem/year) respec­tively. In addition, the collective dose to members of the public living close to the transport route would be about 1.1 x 10 4 man Sv/year (0.0105 mrem/ year).

6.1.2 The requirement for operational support and press briefing centres

Following the accident to the Unit 2 reactor at Three Mile Island (USA) in March 1979, the opportunity was taken to review the emergency procedures at CEGB nuclear sites. This review indicated that the existing emergency arrangements were adequate for the pro­tection of the public and for incidents of short dura­tion. However, if an incident were to be prolonged beyond a few hours it was evident that the overall responsibility of the site emergency controller would be too great and that additional support would be necessary; for example, for the coordination of off­site activities and for the provision of adequate public information. Certain specific areas were identified as requiring particular attention:

• Managerial responsibilities during an emergency.

• Liaison with external organisations.

• Public relations and news media information.

• The radiation dose received by members of the public.

The arrangements introduced for meeting these re­quirements are as follows.

Operational support centre (OSC)

The first two requirements have been met by the introduction of Operational Support Centres, situated at distances of between 5 and 20 miles from each CEGB nuclear site. In an emergency, a senior CEGB man­agement team would attend the OSC and would be responsible for the coordination of all off-site activi­ties including radiation surveys and liaison with the police, local authorities and government departments, and agencies for the protection of the public and the control and assessment of any effects of the accident. The site emergency controller would remain respon­sible for all on-site actions but could look to the OSC for additional support and advice.

The OSC would be directed by a senior CEGB manager (designated OSC controller) who would be assisted by a team providing operations, engineering, health physics, medical, public relations, administrative and clerical support. The CEGB Director of Health and Safety together with the Principal Health Physicist and Principal Inspector from the Health and Safety Department would attend the OSC to give specialist advice to the OSC controller.

Representatives from outside organisations having responsibilities in the event of an emergency would also be present at the OSC to give and receive advice from the OSC controller, and to act as liaison offi­cers between the OSC and the bodies they represent.

The following criteria have been adopted for the location of OSCs and for the facilities provided. Operational support centres are:

• Located outside the planned evacuation area.

• Close enough to the power station to enable easy transfer of personnel.

• Far enough from the power station to avoid traffic congestion near the site.

• Within reasonable distance of the headquarters of local police and the district and county authorities.

• At or near centres of population having hotel ac­commodation.

Operational support centres have:

• A number of offices for senior CEGB personnel and their supporting staff, and for representatives from outside organisations.

• Reliable telephone links with the power station, CEGB headquarters and the public telephone system.

• Telex and facsimile transmission equipment.

• Secretarial, typing and photocopying facilities.

• Facilities for the control and coordination of off­site radiological surveys.

• Facilities for helicopters.

Press briefing centre (PBC)

Events at Three Mile Island indicated that the ori­ginal arrangements for press briefing in CEGB emer­gency schemes would be inadequate to cope with the high level of public and news media interest during and after a nuclear incident. It was further evident that there must be a single authoritative source of information on the course of events. Press briefing centres with adequate facilities for a large number of reporters have therefore been established for each nuclear power station. Here, CEGB public relations personnel can prepare press statements and the OSC controller together with the government technical ad­viser and other specialists can hold press conferences.

Consideration of the requirements for the location and facilities of both OSCs and PBCs showed that there were a number of common factors, e. g., the presence of senior management, the need for secure telephone links, etc. Press briefing centres have, there­fore, with one exception, been established at or ad­jacent to each operational support centre. The ex­ception is in the South West where the geographical disposition of sites has made it possible for a single operational support centre to serve three nuclear power stations. In this case individual press briefing centres have been established near each site.

Depleted irradiated magnox fuel is processed by British Nuclear Fuels. The chemical process separates the radioactise isotopes produced during irradiation from the depleted natural uranium and in doing so con­centrates them for safe storage. The transport of this fuel is accomplished by its containment in packages known as road transport flasks. The flasks are moved by road and/or rail from the CEGB’s sites to BNFL at Sella field.

The flasks for the transport of magnox fuel ele­ments {Fig 3.45) consist of the following principal items:

• Flask body with integral base heat shield.

• Flask lid with integral heat shield.

• Skip contained within the water-filled flask and to hold the fuel.

The body consists of a forging of box form which is machined internally and externally to the required dimensions. Cooling fins are welded on to the sides, with full penetration welds to give an adequate heat flow path. A laminated heat shield is seal-welded on to the base to reduce the temperature rise of the con­tents during a fire accident condition. A single water level valve is mounted in the side of the flask and is used to determine the required water level in the flask. The valve is protected from accident damage by a padlocked cover plate.

The lid is manufactured from a single forging. It is located on the flask body with a tapered spigot and retained by bolts. Face sealing is incorporated at the joint between the lid and body by double seals re­tained in a single groove provided in the lid. The lid is shaped to provide shock-absorbing character­istics and also incorporates a laminated heat shield

• Code of Practice for the Carriage of Radioac­tive Materials by Road issued by HM Stationery — Office.

The regulations require that conditions are met both for normal transportation and under postulated accident conditions. A safety case is made for each type of flask used and is submitted via the CEGB’s Nuclear Health and Safety Department to the De­partment of Transport for approval.

The safety case provides material specifications to­gether with manufacturing details including quality — assurance procedures. It further defines the flask contents and the handling procedures to be used. Theoretical assessments are made of thermal and stress analysis for normal and accident conditions. The ana­lyses are supported, where possible, by model or full scale tests. With regard to accident conditions, the requirements of the IAEA regulations are met by — considering:

• A free drop from a height of 9 m onto an un­yielding target.

• A thermal test at 800°C for 30 minutes.

• Immersion in water to a depth of 15 m.

Arising from the regulations, certain conditions have to be met to allow the fuel to be transported in a flask of approved design:

• In the case of magnox fuel, this must be stored and cooled on site for a minimum period of 90 days following its discharge from the reactor. This en­sures an adequate reduction of the residual heat and the decay of the short lived isotopes 1-131, thereby limiting the possible release of that isotope under the postulated accident conditions.

• For magnox fuel the water contained in the flask is chemically controlled to provide a pH greater than 11.5. This is achieved by a sodium hydroxide content of 200 ppm.

• To reduce magnox corrosion at elevated tempera­tures following a postulated accident, a sodium tluoride solution is injected into the flask after it has been loaded with fuel such that the resultant fluoride concentration is 1000 ppm. The injection is etfected by a pressurised nitrogen system or by a gravity feed system,

• Before dispatch of the flask, all watertight seals are pressure tested to 20.68 bar and the ullage space above the water is purged with nitrogen, which in the magnox case avoids the possibility of an ex­plosive hydroeen/air mixture. [38]

with the flask design but is typically 5 kW.

• The possibility of exposed uranium enabling fission products to be leached into the water is monitored by caesium release rate measurements. The mea­sured rate is temperature and irradiation dependent and no skip load of fuel may be dispatched if the rate is above clearly defined limits.

Utilisation of the CEGB’s tlasks is determined cen­trally and their movement is planned in collaboration with British Rail. Emergency plans are available to provide for the unlikely event of an accident to a loaded flask in transit. These have been drawn up by the CEGB, British Rail and BNFL. The plans enable a health physics team to be called out from the nearest nuclear power station or BNFL establishment to deal with such an emergency and ensures national coverage and a rapid response.

The oxide is produced by the reduction of CO2 to form the adherent black oxide, magnetite (nominally Fe^Oj). The rate of formation of this scale is tem­perature-dependent and like any chemical reaction increases with temperature. The early experimental evidence suggested that a protective oxide would be built-up and would reach a limiting thickness which was catered for in design. The unforeseen aspect of steel oxidation was the so called ‘breakaway’ behaviour.

The stages of oxidation are illustrated by Fig 3.70. The protective layer is about 50 microns and is strongly adherent to the metal surface. On mild steel, given sufficient time, a porous sub-layer of oxide is formed and this bursts through the protective layer to form excrescences. As this transition stage proceeds, more and more of the protective layer is broken down until the full breakaway condition is reached, when the metal will continue to oxidise and is wasted away.

The oxidation rate in the breakaway condition is temperature, gas coolant composition and material composition dependent. These features are illustrated by Figs 3.71 and 3.72. Except for very low silicon con­tent steels, the oxidation rate is low at temperatures below 360°C and roughly doubles for every 20°C increase in temperature. Hence the reduction of maxi­mum gas temperatures to 360°C in 1968. The carbon monoxide and hydrogen content of the coolant gas also influences the rate of oxidation (the lower the lex els, the less the oxidation). However, low values of these constituents increases the corrosion of graphite, і. e., the core, so that in controlling gas composition a balance has to be made between these two conflict­ing requirements.

There is a general requirement for emergency pro­visions to be made, so that appropriate protective actions can be taken in the event of transport accidents involving radioactive materials.

2.8.2 Regulatory position in the United Kingdom

Under the Radioactive Substances Act 1948 [16], the appropriate Minister may, as regards the transport of radioactive substances, make such regulations as appear necessary to prevent injury being caused. To this end, the IAEA Regulations have been adopted as the basis of current UK legislation.

The IAEA Regulations [12] require each member country to appoint a competent authority which in the case of Great Britain for transport by road, rail, air and sea, is the Secretary of State for Transport. The executive function of the competent authority is invested in the Transport Radiological Adviser of the Department of Transport (DTp).

Road transport was governed in general terms by the Radioactive Substances (Carriage by road) (Great

Britain) Regulations 1974 [17] with additional speci­fications tor the protection of road transport workers [18]. This situation changed in 1986 when new widely embracing Ionising Radiations Regulations were en­acted [19]. These Carriage by Road Regulations are amplified in a Code of Practice [20] which conforms generally to the IAEA Regulations. This code sets out to assist all concerned, and consignors in particular, to discharge their obligations under the law.

Transport by rail is covered by British Rail’s List of Dangerous Goods and Conditions of Acceptance [21] which require, except where otherwise provided in the conditions, compliance with the requirements of the IAEA Regulations.

There are separate International Agreements and Conventions covering various modes of transport when making shipments to and from the UK [22, 23, 24 and 25].

A guide for designers, manufacturers and users of packagings giving advice supplementary to, or in am­plification of, the IAEA recommendations [12,13] have been published by the British Standards Institution [26].

Transport on nuclear sites is governed by the site licences, which require nuclear matter to be carried either in appropriate containers of a type approved by the Health and Safety Executive (HSE), or under rules or arrangements made by the licensee and ap­proved by the HSE. The despatch of nuclear matter from the site is similarly controlled.

Insurance for the carriage of nuclear materials, with some exemptions, is required under the Nuclear In­stallations Act 1965 [27,28]. However, individual co­pies of the Certificates of Financial Security need only be provided by CEGB to the organisations involved, for transport outside the territorial limits of the UK.

These are non-stochastic effects that manifest them­selves in the individual exposed within a few weeks of the exposure. They are all the result of fairly massive acute doses of radiation. Although changes in the blood cell count could be detected by laboratory test, an exposed individual would not feel unwell after doses of up to 0.5 Sv. At higher doses the following effects will become more pronounced until at doses of more than 4 Sv most people would die without specialised medical treatment.

Haematopoietic or bone marrow syndrome

The long lifetime of red blood cells results in the number of circulating red cells not being immediately affected by a large exposure, whereas the shorter-lived white cells show’ a rapid drop in number over the first week. A reduction in the white cell count clearly lowers the body’s defences against infection. The red cell count eventually falls because of parent cell death.

A whole-body dose of 7 Sv will render sufficient numbers of the white parent cell population repro­ductive^ non-viable to cause death from uncontrolled infections in the irradiated individual. Death usually occurs within one or two months of irradiation, but the exact time will depend on the dose and the phy­sical state of the individual exposed.

Gastro-intestinal tract syndrome (GI)

Doses of several sieverts and higher, significantly im­pair the cell renewal system of the gastro-intestinal tract by ‘killing’ the parent cell population. Death may result from a combination of infection by gut organisms, reduced nutrient uptake and loss of blood to the intestine. Death can occur within a few days or weeks for doses over 10 Sv with early symptoms such as vomiting, diarrhoea and dehydration.

The central nervous system syndrome (CNS)

The CNS syndrome only becomes apparent after doses of several tens of sieverts. Tremors, convulsion and general loss of control of movement are the most common symptoms and sometimes occur within min­utes, although death may not occur for a day or two. In this case, unlike the two previous syndromes, the cells involved physically break up.

For that monitoring which is statutory, results are required to be reported to the authorising government departments at quarterly intervals, although generally samples are taken and measurements made more fre­quently than this.

Table 4.15 shows typical levels of radioactivity observed in environmental samples near CEGB sites. In most cases these levels are due only to background radioactivity.

For airborne discharges, the milk route constitutes one of the more important pathways back to man. Except for sulphur-35, there is no reason to believe that radioactivity in milk samples results from station emissions. In the case of sulphur-35 discharges from Hinkley Point, members of the public who consume local milk, do not receive radiation doses in excess of 0.1% of the ICRP recommended limit. Sulphur-35 discharges from other stations result in much lower radiation doses.

For liquid discharges, the critical pathway is usu­ally via consumption of fish or shellfish. Near most stations, the radiation doses to which members of the local population may be exposed as a consequence of liquid effluent discharges is not more than 0.2% of the ICRP limit. The exception is Trawsfynydd where the dose to a small group of trout eaters could in 1983 have been approximately 4% of the ICRP
limit. At Bradwell, the critical exposure pathway was originally considered to be through the consumption of oysters from beds in the Blackwater Estuary, with zinc-65 and silver — 110m as the critical radio­nuclides.

This pathway has now declined to such an extent that the consumption of fish is currently more signi­ficant. These exceptional pathways are summarised in Table 4.16.

In addition, film badges exposed on the site fence enable an estimate to be made of radiation doses to the public due to direct radiation from the nuclear site. Generally, the average dose measured by these film badges around any site does not exceed 5 mSv/ year and the dose to individual members of the public near the site would not be expected to be more than about 0.15 mSv/year. At Berkeley, however, dose rates are usually higher and in a normal year the average dose rate measured by its fence film badge is approximately 10 mSv/year.

This results in an estimated maximum radiation dose to certain members of the public near the site of about 2 mSv in a year.

Further details of environmental monitoring near CEGB sites are reported annually, the most recent report being Heap and Short, 1984 [8]. Additionally, DoE publish an environmental monitoring report deal­ing with each individual station for presentation to that station’s Local Liaison Committee which meets once a year. MAFF publish an annual report dealing with monitoring of the aquatic environment around UK. The most recent was published in 1984 [39].

Transport accidents to fuel flasks are divided into two categories, namely Nuclear Flask Incidents and Nuclear Flask Emergencies.

A nuclear flask emergency will be declared when a railway or road transporter carrying a flask has suf­fered or might suffer a mishap, and any of the fol­lowing conditions exists or is suspected to exist:

• The flask is involved in, or endangered by, fire.

• There are dangerous loads (e. g., petrol, liquified gas) in the vicinity.

• There is visible damage to the flask or its cover, which is suspected to be due to the mishap.

• The flask is tilted out of the upright position, or is displaced from its normal position, on the transport vehicle.

• The flask is endangered in some other way, e. g., by sabotage or a security threat.

A nuclear flask incident will be declared when the transport vehicle has suffered a mishap other than a mechanical failure and none of the above conditions exist.

Every AGR fuel element contains its own unique alphanumeric identification number which defines the design characteristics and manufacturing detail of both the element and its fuel pins. The actual number is marked in 12.7 mm high characters on the outside surface of the outer sleeve of the element, at its upper end. A typical identification code oc­cupies up to 13 characters and would take the form; GADXX200034AB

The first letter defines the power station for which the fuel was made, ‘G’ in this case means Hinkley Point В fuel. A combination of the second and third letters together (‘AD’ above) defines the fuel element ‘TYPE’ number and also the ‘MARK’ number of the fuel pins. The ‘TYPE’ number identifies the fuel element assembly including the fuel pin array, but pays no regard to individual pin detail. For example, Type 6 and Type 6A elements differ only in the supplier of the graphite used during manufacture. The ‘MARK’ number identifies the fuel pin only and refers to vari­ous features of pin design such as ASG array, pellet type, can manufacturing history and so on. Fuel en­richment is denoted by the next two letters (‘XX’ in the example) and the following six digits represent the element’s own unique serial number. The final two letters, which are optional and usually unused, are suffixes to identify special or experimental fuel. An additional letter (not shown), appearing above the first few letters of the fuel element code, and therefore displaced from the 13-character field, relates to special manufacturing tests performed on the element prior to delivery. Each fuel pin also contains identification markings on both its top and bottom end caps. Two identical letters are used on the top end cap to denote isotopic enrichment — these letters will correspond to the enrichment markings contained within the element identification. The lower end cap contains two capitals which are used to define can and pin manufacturing detail as well as the pin ‘MARK’ number.