Storage

Subsequent to its discharge from the reactor, most magnox fuel is stored on site in a water-filled cooling pond prior to its despatch to the chemical processing plant of BNFL. (The exception to this ‘wet’ storage is the ‘dry’ storage area of Wylfa power station.) The elements are stored under 5-6 m of water which provides the cooling medium for heat removal and also acts as a radiation shield for the highly radio­active fuel elements. The elements are stored for a sufficient length of time to allow the fission products to decay and for the associated heat produced to dissipate and to reduce to an acceptable level for transport requirements.

The elements are accumulated on receipt at the ponds in a skip, which is an open-top box fabricated from mild steel. On discharge, the elements are com­plete with the splitter assembly or with lugs depend­ing on the type of fuel. The presence of these items results in a poor packing factor for the fuel. Later, these items are removed. Since with magnox fuel there are no criticality considerations to be taken into account, the fuel may be moved as found necessary within the pond.

Cooling pond management

The principal objectives of cooling pond management are firstly, to satisfactorily store the fuel pending its despatch off-site and secondly to ensure its safe re­moval from site including its transportation to the processing plant. These objectives are further amplified as follows:

• To preserve the fuel cladding by minimising cor­rosion and handling damage.

• To minimise radiological hazards arising within the

pond area.

• To ensure that adequate cooling is achieved before

despatch.

• To control fuel movements into, within and from the

pond area.

The element canning material is a magnesium alloy and as such is a highly reactive metal which is readily corroded by water, unless a protective oxide/hydroxide film can be maintained over the whole of the can surface. This film, which is initially formed in the reactor environment, continues to grow during storage in the cooling ponds. The film may be damaged by mechanisms which are either chemical or physical in origin.

The chemical preservation of this protective film is effected by stringent chemical control of the cooling water. The film is maintained by employing a high pH level between 11.5 and ll.7.

Demineralised water is used in the ponds and a concentration of 200 mg/kg NaOH achieves the target pH level. To maintain a higher pH, say 12.5, would require a tenfold increase in the NaOH concentration, which would result in loss of pond water treatment plant availability due to increased resin regeneration times. Atmospheric carbon dioxide dissolves in the pond water and this results in a lowering of the pH by it reaction with the sodium hydroxide to produce sodium carbonate. At a pond water pH of 11.5-11.7 the sodium carbonate concentration will be in the range 100-250 mg/kg. The presence of anions, in particular chloride and sulphate, enhances the rate of corrosion of magnox. Thus anions in the pond water are controlled with a target of less then 0.5 mg/kg and an upper limit of 1.0 mg/kg.

A water treatment plant (Fig 3.43) is used to adjust the chemical conditions of the pond. Between З^о and 5% of the pond volume is passed through the plant per day. After filtration to remove particulate matter, the water is passed through a cation resin to remove the sodium ions leaving the hydroxyl ions as water and the carbonate ions as carbonic acid in solution. The latter is then scrubbed out of solution in a ‘decarbonating’ tower. The water is then passed through a mixed bed resin unit to remove all further ions. Finally, the water pH is restored to the required level by sodium hydroxide injection. It will be noted that the mixed bed resin will remove many of the ions associated with radioactive products arising in the pond. These are derived from the activation pro­ducts of the canning material and from fission pro­ducts leached from the uranium bar via leak paths produced by corrosion or mechanical damage to the canning.

It usually provides a special resin bed to cater for the removal of fission products caesium-137 and caesium-134. The resin, Lewatit DN, has the ability to perform efficiently at the pond pH levels. However, since sodium is also removed by this resin, a bed volume capacity of 12 000 to 15 000 is to be ex­pected irrespective of the caesium activity in the pond. Thus, this bed is not employed before the caesium has risen to a predetermined level and then its use is discontinued when the target operating level has been restored.

The fission product decay heat from a large number of elements can raise the pond water temperature by an appreciable amount. A maximum of 30°C is a practical limit which is reasonable to achieve with efficient coolers in the summer. Some stations have installed pond water chiller plants to maintain the water temperature at 10°C. This has the advantage of considerably reducing the rate of corrosion since chemical reactions are temperature dependent. Civil engineering requirements dictate that the rate of change of water temperature and the temperature distribution throughout the ponds shall be at a minimum and con­stant to avoid stresses in the pond structure.

Pond water invariably holds suspended particulate matter which in settling forms a sludge, the presence of which increases the possibility of corrosion. The paniculate matter is removed by filtration.

The skips in which the elements are stored are painted in a durable paint and it is essential that the paint surface is maintained in good condition. This is to avoid the possibility of enhanced corrosion of the fuel brought about by the production of galvanic couples between the bare mild steel and the cladding material.

It has been indicated that mechanical damage to the protective film can precipitate corrosion. To this end. elements are handled as little as possible even though the tools used for this task are purpose de — ‘mned. Similarly, the removal of the splitter cage by J ram and die process and the cropping of lues from lerringbone elements is delayed until the time of despatch ot that fuel off-site is imminent.

Before despatch, the fuel has to be cooled to re — ULj»e the heat burden of a road transport flask to an -^ptable lev el. Contractually, the CEGB is required

Л00′ tuel for a period of 90 days. This is re — 4mred so that the release of iodine-131 in BNFL’s 1 r{XeS4’nS plant is kept to a level suitable to their
operational requirements. Observance of this 90-day limit is obtained in the first instance by administra­tive control. This control is reinforced by the use of a device known as a ‘short cooled element moni­tor’. This device is employed at the displittering/ delugging stage of element handling. In use, the ele­ment is presented to the instrumentation which is designed to identify the lanthanum-146 at 1.6 MeV in the spectrum of energy emissions from the element. This peak is no longer identifiable after 90 days cooling when the fission products associated with the lanthanum will have decayed to a low value of activity.

Whilst a 90-day cooling period is a contractual requirement, the heat burden of a potential skip of fuel for despatch has to be below that required by transportation regulations. Whilst the heat burden varies with the type of fuel and transport flask design, it is of the order of 4.5 kW for magnox fuel.

On discharge from the reactor, the continued decay of fission products results in the production of heat on a decreasing scale. In addition, elements from dif­ferent areas of the core generate varying amounts of heat. Referring to Fig 3.44, it will be noted that at the end of 90 days cooling the heat burden of 200 elements from the flattened zone is still above the

acceptable level for transportation. Indeed, some 130-140 days cooling are required in this case. At the same time, fuel from the unflattened zone and from the edge of the core reaches a satisfactory level within 25 days. Even so this fuel would not have completed the 90-day contractural cooling period. Thus before despatch, and even before desplittering, it is necessary to compute the heat burden of the pro­posed full load. This load may be made up of fuel from various areas of the core because a skip of undesplittered/delugged fuel is 120-140 elements whilst that of fuel ready for despatch is 200-225 elements. This being so, the mixing of elements is unavoidable. To ensure that the selection of fuel for despatch is optimised, it is necessary to have a comprehensive recording and control system.

The graphite core of the reactor consists of machined graphite bricks which are locked together with either graphite kes or zirconium pins. The integrity of this structure must be maintained, since it must withstand
the loads imposed by thermal expansion and by di­mensional changes induced by irradiation. It must also withstand the static weight of the core itself and the associated fuel which together weigh some 1000-3500 t. The carbon dioxide (CCb) coolant of the magnox and

AGR reactors reacts with the core graphite, causing a «eight loss and a consequent loss of strength. It is the loss of strength rather than the loss of mod­erating effectiveness which necessitates the control of tzraphite oxidation.

The process of oxidation may be thermal or radio — Ivtic, the detailed chemistry of the processes being discussed in Chapter I. Since the thermal reaction between the graphite and coolant is insignificant below 600°C, this reaction does not predominate in CEGB reactors. Radiolytic oxidation occurs when CO: is decomposed by ionising radiation to gie rise to re­active oxidising species (positive ions resulting from absorption of energy by the CO:), Some of these ions are able to combine with the graphite and produce carbon monoxide (CO). The oxidising species also combine with the carbon monoxide to reform carbon dioxide. Since carbon monoxide is also formed by radiolysis of the coolant gas (CO:), its concentration is allowed to build up to inhibit the action of the oxi­dising species on the graphite. There is no further increase in inhibition at CO concentrations above 1.5 vol<?o, so that this level is the target maximum for the magnox range of reactors.

Graphite oxidation is a function of radiation in­tensity and gas pressure. The criteria of 1.5 voIVo CO (maximum) was acceptable for the early magnox re­actors. However, the rate of graphite oxidation in­creases through the magnox series mainly because the coolant pressures rise from 9 bar (Berkeley) to 27 bar (Wylfa). A detailed research programme identified hy­drogen as the most suitable inhibitor for the later stations. Ingress of water from boilers and oil from gas circulators give rise to low concentrations of hydrogenous compounds and the total hydrogen equi­valent from (H2O + H2 + CH4) is used for control purposes. It should be noted that at Hinkley Point A,~ the gas circulators have air seals and hydrogen may need to be injected to maintain the required levels. The use of hydrogen brings its own problem since high concentrations accelerate the oxidation of steels. Figure 3.65 illustrates the conflict of requirements with respect to graphite and steel oxidation, i. e., low hydrogen concentrations for steel oxidation control but high levels for control of the graphite reaction. The magnox system chemistry is illustrated in block diagram form by Fig 3.66.

Each reactor presents its own particular problems in coolant chemical control. Much depends on past operating conditions and coolant compositions. Con­trol of CO and hydrogen concentrations is exercised by periodic purges of the CO:, and in some cases a catalytic recombination unit is employed for CO control to combat the uneconomic use of CO: for purging. Gas driers are used for water removal but these are generally employed to remove water from the gas circuit following a prolonged shutdown for maintenance. Those reactors in which air is admitted to the coolant circuit at these times are provided with

‘O’v’.W a’ — ‘-1.

2T

dry air and the circuit is kept at a positive pressure. This is to reduce the quantity of vvater absorbed by the graphite and is primarily intended to reduce the dry-out time at the subsequent start-up of the reactor.

The AGR reactors, operating at higher gas pres­sures, temperatures and flux levels, present the graph­ite core with a more hostile environment than that of the magnox reactors, resulting in the need for a high degree of oxidation inhibition.

— Radiolytic oxidation takes place mainly within the pores of the graphite rather than on exposed surfaces. This is because the oxidising species have extremely short lifetimes and are quickly eliminated within a short distance of their point of production. This fact led to the development of reactor-grade graphites (pro­duced from Gilsonite deposits in Utah, USA), with a reduced pore volume compared with that used in the magnox reactors. The oxidation rate is also dependent on the pore diameter spectrum and geometry. The graphite should preferably contain a small number of large pores rather than a large number of small pores. It should be noted that oxidation increases the pore volume so that the rate of oxidation increases with time.

As in the case of the magnox reactors, carbon monoxide is an inhibitor of the oxidation rate but is of limited value (maximum factor of 2) and has the additional disadvantage of a tendency to form carbonaceous deposits, Further inhibition of the graph­ite corrosion reaction is obtained by providing a sa­crificial carbonaceous film on the surface of the graphite, derived from the radiolytic decomposition of methane. Unfortunately, this process gives rise to carbon deposits on the fuel pins, thus impairing heat transfer.

It has been established that methane is a powerful inhibitor (1000 vpm reduces reaction rate by a factor of 20). However, due to its radiolytic destruction in-core, technical and economic problems limit the maximum concentration that can be used in practice. For example, the use of high concentrations of meth­ane requires a large plant for the production of the gas, a large drier unit to remove the water resulting from the decomposition of the methane, and a re­combination unit to remove the CO so produced. Ob­viously, there is an economic limit to the size of such

equipment and this reflects the level to which the methane concentration can be raised (Fig 3.67).

It has been indicated that the production of the oxidising species takes place within the graphite pores. To gain access to the pores, methane has to diffuse into the graphite and, as a result, marked weight-loss profiles occur within a brick structure despite the provi­sion of methane access holes (Fig 3.68).

The relationship between the coolant composition and. graphite lifetime is shown in Fig 3.69. The term ’effective weight loss’ is a parameter which describes

the graphite weight loss in a brick since this is not uniform throughout. The highest effective weight loss that could be tolerated around the end of design life is presently considered to be about 20%. In Fig 3.69, the corrosion contours link gas compositions of equal effective weight loss and show that high methane concentrations prolong graphite life. Since plant economics preclude the adoption of certain composi­tions, these are bounded by the ‘plant limit’ line. Compositions that are prone to produce carbonaceous deposits are defined as those occurring above the predicted ‘deposition boundary’.

The initial gas composition selected for the CEGB’s lirst AGR commissioned was well below the deposition boundary — (1% СО/130 vpm CHj). Post-irradia­tion examination of the discharged fuel confirmed that carbon deposition was absent (some deposit was pre­sent but it was identified as being derived from lu­bricating oil). It will be noted from Fig 3.69 that the selected composition was far from ideal, in that a brick life of less than 20 years would be expected.

Subsequent to this initial trial, tests were carried out on the reactors at Hinkley Point В and Hunterston В with small changes to the gas compositions such that more inhibiting conditions were achieved. These changes advanced the expected brick life to about 28 years. However, there is a necessary delay between the completion of a trial and the examination of this fuel to assess deposition of carbon.

To overcome the delay in assessment and to pro­vide a more flexible method of detecting deposition, instrumental fuel stringers were developed. These units were provided with a number of thick-walled cans (1.8 mm instead of the normal 0.38 mm) with a hole drilled in the wall to accept a 0.5 mm thermocouple. The fuel pellets used in these cans were specially en­riched to compensate for the additional steel present in these particular cans. Thus the measured can wall temperature matched that of the standard fuel can. In addition, prqansion was made to measure gas inlet and outlet temperatures together with the gas-mass flow. The use of these stringers to detect deposition, requires that the temperature can be calculated taking into account all variables except deposition. The difference between the prediction and the as-measured can tem­perature is a measure of the deposition present.

The first trial was started in October 1982 with a coolant composition of 1.5% CO/300 vpm CHj, It was predicted that there would be a 9% reduction

in the heat transfer from the lowest element within a nine month period. By November 1983, no change had been detected. Subsequent post irradiation ex­amination (PIE) of the fuel confirmed that deposits present were within the range established before the test began. The composition tested was therefore judged to be non-depositing; Fig 3,69 shows that it is strongly inhibiting and provides some 30 full power years of life.

Tests in 1984 ’85 with coolants containing l. S^o CO/415 vpm СНд and 1.2°“o СО/350 vpm CH4 were later shown by PIE of fuels to have produced signi­ficant fuel pin deposition. Pending further evaluation of reactor and research rig evidence, CO and CH4 levels have both been reduced to avoid the rush of further deposition. The present coolant composition is ITo CO/230 vpm CH4, which will provide approxi­mately 25 years of full power life.

9.2 Steel oxidation lifetime The first of the CEGB’s magnox stations was com­missioned in 1962, the design being based on infor­mation available at that time. In 1968, the results of rig work and examination of specimens removed from some of the magnox reactors became available. These showed that steel oxidation rates of in-core compo­nents were unacceptably high at the (then) current operating conditions. The minimum gas outlet tem­perature was limited to 360°C on all the CEGB re­actors (except Berkeley which remained at 355°C) to ensure that the station’s economic lifetime would be achieved. Since that time the limited gas temperatures have been raised, thereby recovering some of the con­sequential loss of power output.

The requirements of the regulations include the de­signation for the purpose of the regulations of a national competent authority.

Competent authority approval certification is re­quired, for example, for package designs of Type В and for packages of fissile material except for some specific exemptions; approval is also required for cer­tain shipments. Special arrangements may be made in respect of consignments which do not fully comply with the regulatory design requirements. These arrange­ments ensure that there is no relaxation in the overall standard of safety. All such special arrangements require competent authority approval. Other require­ments on the competent authority include periodic assessment to ensure that the agency’s radiation safety standards [15] are being complied with.

Consignors are required to establish quality assur­ance programmes covering all aspects of transport activities, including package design, manufacture and use, in order to ensure that the regulations are being complied with. A corresponding responsibility is placed on the competent authority in respect of compliance assurance.

4.3.1 Development of the rules

In November 1957 the Nuclear Safety Rules Sub­committee was formed with the terms of reference ‘to formulate rules for the radiological protection of persons employed on CEGB premises’.

The first meeting of the subcommittee was held in 1958 and since that time there have been several changes to the rules, although the underlying princi­ples of the first rules still remain valid today.

In recent years the subcommittee was renamed the Nuclear Safety Rules Advisory Committee. The Chairman is the Director of Health and Safety and the membership comprises the Electricity Council’s Chief Safety Officer, Senior Regional Representatives, the Nuclear Safety Officer of the Health and Safety De­partment, the CEGB’s Medical Adviser, various Power Station Managers, the Senior Research Health Physi­cist of Berkeley Nuclear Laboratories and a representa­tive of the Electrical Power Engineers Association. A technical secretary is provided by the Nuclear Safety Branch of Health and Safety Department^

Originally there were two formal sets of rules, the Safety Rules (Radiological) and the Safety Rules Ion­ising Radiations for Non-Destructive Testing but, with the introduction of the Ionising Radiations Regulations 1985, these were, in effect, combined.

The preparation of the first rules was made all the more difficult by the absence of any significant legislation. The Factories Act legislation (the Sealed Sources and Unsealed Radioactive Substances Regu­lations), only existed in draft form and the Nuclear Installations Act of 1959 and the Radioactive Sub­stances Act of 1960 did not exist. None of the current series of ICRP recommendations existed, although ICRP Publication 1 was introduced later on in 1958.

In spite of all this, some Codes of Practice were in existence and there was also the practical experience of the United Kingdom Atomic Energy Authority.

Responsibility

In accordance with the CEGB QA policy and the adoption of BS5882 by the CEGB as the appropriate basis for the total QA programme for its power stations, it is the responsibility of the CEGB to ensure that a QA programme for each plant is established and implemented. In each phase of a power station’s life, many organisations are involved in providing plant items and services to the CEGB.

Where these organisations lie within the CEGB, their responsibilities are defined within the CEGB Directives, it may be necessary to prepare project — specific QA documentation, for example, in compli­ance with nuclear site licence requirements, in addition to generic procedural documentation.

Where organisations external to the CEGB are sup­pliers, the CEGB will require that each organisation demonstrates that its own QA arrangements are ade­quate and specifies or approves those of its contractors, suppliers and agents. This hierarchical system of iden­tifying detailed responsibility, together with formal CEGB monitoring of the immediate purchaser, is followed throughout the purchaser/supplier arrange­ments whilst not relieving the CEGB of its overall QA responsibility.

Design

Design covers not only how an end item is to be made but also how it is to be used. The designer is responsible not only for the proof of integrity of the design by analysis and testing, and for providing the manufacturer with the instructions of how to make an item, but also for provision to the operator of the item of the instructions for the use of the item.

The objective of QA in design is to provide mea­sures of confidence that the management system re­sponsible for design has adequately considered the requirements for reliability, performance and safety in formulating the design, has provided instructions to the manufacturer for the realisation of the design and has prepared an examination, inspection and test schedule that will confirm the design assumptions.

Procurement

Procurement covers all the activities associated with realising a declared design in an end item, i. e., re­quisition, manufacture and construction, the supplier is responsible for both the design and the procurement of plant systems.

The objective of QA in procurement is to provide measures of confidence in the ability and performance of a supplier in the production of the end item.

The assessment of the quality capabilities of sup­pliers is an essential feature of the procurement pro­cess. The function of the plant item will determine the safety significance and hence the QA actions ap­propriate in the procurement process. This will include the identification of an appropriate quality manage­ment system standard and the particular actions, such as procurement authority approval of QA programmes and quality plans, enforceable through the terms of contract.

AGRs in the UK use enriched uranium dioxide (UO2) fuel contained within a stainless steel clad or ‘can’. Each fuel element, w’hich is approximately 1 m in length, consists of a bundle of 36 fuel pins, each 14.5 mm diameter and of the same enrichment, supported within a 190 mm bore double-graphite sleeve assem­bly. This serves both to insulate the moderator bricks within the reactor fuel channels from the hot coolant passing over the fuel, and also to permit the flow of cooler gas around the outer sleeve of the fuel element to further cool the moderator. As shown in Fig 3.53, the fuel pins, arranged in three concentric rings of 6, 12 and 18 are attached to machined grids at their lower ends and traverse two additional support braces at the centre and upper end of the fuel element. A central guide tube accommodates the threading of a tie-bar which takes up the full weight of the fuel stack {usually eight elements) during subsequent handling of the completed stringer by the charge machine.

The cans are made from niobium-stabilised stain­less steel, the overall volume of which needs to be mmimi>ed on grounds of neutron economy, leading a fundamental!} ‘thin can’ design of pin. Heat transter is enhanced via surface-roughening of the can m the term of machined transverse annular ribs, although some of the earlier AGR fuel elements con­tained helically-ribbed cans. At the pin ends, cup-

~E =E’AA NG = G

shaped end caps are welded to the main can wall in two places to provide a double-seal between the fuel and the reactor environment and the pins are locked into the fuel element grid at their lower ends. Each pin contains typically 64 hollow UCb fuel pellets, some of which are specially grooved to receive the can during its pressurisation onto the fuel in a helium gas environment, during the final stages of pin manu­

facture. These ‘anti-stacking’ grooves (ASGs) prevent gaps opening up in the pellet stack during service as a result of differential movements between the fuel and the can. If present, such gaps would allow the can to deform into them under the influence of the external reactor gas pressure, which would eventually lead to localised weaknesses in the can wall and hence provide a possible mechanism for failure. Fuel cur­rently in use at Hinkley Point and Hunterston con­tains either 14 or 22 ASGs per pin. The top two positions within the fuel stack (elements 7 and 8) are assembled only with elements containing 22 ASG pins, since these positions have shown themselves to be more prone to inter-pellet gap formation than those lower down the stack. An increase in the number of ASGs reduces gapping to a minimum. The 14 ASG fuel is assembled into positions 1 to 6. At each end of the pellet stack there is an insulating pellet made from sintered alumina powder, the function of which is to protect the can at the pin ends, a region of poor heat transfer, from excessive temperatures. The UCb fuel pellets contain a hollow central bore (typi­cally 6.35 mm dia. in current Hinkley Point feed fuel) which acts as a void for the accumulation of fission gases produced during reactor operation. Since the rate of release of fission gas depends strongly upon the fuel temperature, the presence of the bore also restricts the quantity of gas actually released from within the fuel by effectively lowering the centre UO2 temperature for a given heat rating. The degree of enrichment of the fuel in U-235 varies across the reactor core and also between stations, but generally lies between l№o and 3% by weight.

In the UK, control over radioactive wastes is exer­cised through the Radioactive Substances Act 1960, (‘the Act’) which came into force in 1963 following a government white paper policy document ‘The Control of Radioactive Wastes’ (Cmnd 884). This policy was reviewed following consideration of the 1976 Sixth Report of the Royal Commission on Environmental ~ Pollution [1]. Further reviews of policy and objectives were given in the 1977 White Paper ‘Nuclear Power and the Environment’ [2] and an Expert Group re­viewed Cmnd 884 in 1979 [3]. Policy objectives were repeated in a 1982 White Paper ‘Radioactive Waste Management’ [4]. The Government has established a Radioactive Waste Management Advisory Committee whose fifth Annual Report was made in 1984 [5]. The specific items of reference of the Committee are: ‘To advise the Secretaries of State for the Environ­ment, Scotland and Wales on major issues relating to the development and implementation of an overall policy for the management of civil radioactive waste, including the waste management implications of nu­clear policy, of the design of nuclear systems and of research and development, and the environmental as­pects of the handling and treatment of wastes.’

Under an independent Chairman, the Committee currently consists of eight independent members, four members from the nuclear and electricity generating industries and three from the trade unions with mem­bers in those industries. Whilst these reviews are re­flected in the interpretation of the Radioactive Sub­stances Act 1960, the Act itself remains unchanged.

Plant faults are those which might arise due to fail­ures or malfunction of the plant itself. They include those which can affect the reactor, the reactor pri­mary and secondary coolant systems, and also those which can affect new and irradiated fuel and radio­active effluent treatment plants. An outline of a few such faults is given in Section 3.5 of this chapter (Fault studies).

Internal hazards are on-site events which have the potential to cause damage to unrelated equipment around the point of origin, or beyond. Those con­sidered include:

• Fire, missiles, turbine disintegration, dropped loads, explosive and toxic gases, and failure of pressurised systems. Failure of pressurised systems could, for example, result in pipe whip, flying fragments, jet impingement and local flooding.

By comparison with natural and other external hazards it is more often possible to place an upper limit on the level of interna! hazard, for example, the magni­tude of a dropped load.

The design approach to demonstrate the safety of the plant against internal hazards depends on the particular hazard under consideration, but the main features of the approach are:

• To reduce the probability of the hazard occurring through a high standard of design, construction and operation.

• To limit the consequences of hazards through se­gregation, protection and layout features.

• To ensure that the reactor can be taken to a hot shutdown condition in the event of the hazard oc­curring. Some repair, temporary connections (for example, cabling), or local manual operation may be required to establish cold shutdown capability. Maintenance requirements and the single failure criterion are taken into account when assessing the adequacy of the plant proposals.

External hazards are events originating outside the site but which could affect plant safety and include both natural and man-made events. Those considered for the PWR for instance include:

• Ground settlement and subsidence, precipitation, lightning, wind, flooding, extreme ambient tempera­tures, industrial activity off-site, gas clouds, earth­quakes, aircraft impact.

Where possible, the criteria for hazards is to ensure that the combination of the frequency of the hazard and the probability of subsequent failure to control the reactor is consistent with the design safety criteria for large uncontrolled releases to the environment. On this basis, the external hazard design criteria have in general been set at a level of frequency of 10“4 per year and the probability of failure to cool and shutdown the reactor in the event of the hazard set at less than 10 ~3 per demand.

Thus, in the case of earthquakes for example, an earthquake with a peak horizontal acceleration of 0.25 g has been adopted for design purposes. This exceeds the 10 ~4 per year level at all currently planned sites; it is above the UK average and would be conservative for most UK locations. This has associated with it a ground motion spectrum which characterises the fre­quency content of the various possible earthquakes which could have a peak acceleration, of 0.25 g. The station and plant design has to be such that the re­actor can be safely shutdown with requisite reliability in the event of such an earthquake, and is known as the safe shutdown earthquake (SSE). It will be noted that earthquakes which exceed this design level are possible. However, in practice the plant will safely withstand more severe hazards than that specified be­cause plant and structures built to normal conservative design codes have a considerable margin of reserve strength between the design limits and the point of collapse or total failure. Sensitivity studies are carried out to show that there is no sudden reduction in safety for events more severe than the SSE.

As in the case of the SSE, the design wind speed is set at a level such that the frequency of its being exceeded is less than 10“4 per year. On this basis, the extreme wind speed arrived at for the design of safety — related buildings and plant is 58.8 m/s {132 mph).

Aircraft crash exemplifies an external hazard for which a modified approach is adopted to establish the acceptability of the design. A different approach is possible because sufficient data for aircraft crashes exist to enable the frequency of both civil and mili­tary aircraft crashes to be predicted for the site. This, together with an evaluation to identify those areas of the site which, if hit, have the potential for signifi­cant radioactive release, leads to a predicted frequency of crash onto potentially vulnerable areas. Because of the inherent strength of some of the buildings, it is possible to discount any hazard arising from impact of light aircraft and some helicopters. Although the arguments are further complicated by consideration of consequential fires due, for example, to spillage of aviation fuel, it is possible from this type of assess­ment to show that major radioactive release from an aircraft crash is of the order of 10“7 per year and therefore considered an acceptably low risk.

For some hazards a probabilistic approach is not appropriate or, because of lack of data, not possible. In the case of precipitation, for example, normal UK practice is used; this is regarded as adequate when taking into consideration the segregation, redundancy and diversity in the plant. For sea flooding, as an­other example, a maximum credible sea level has been evaluated.

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.