A committee was appointed by the government in 1970 to consider the whole range of law relating to occupational health and safety. It reported in 1972 and fundamentally criticised the existing legal frame­work. Following this the government introduced le­gislative proposals based on a report in May 1973 and these were passed into law on 31 July 1974.

The Act provides for the establishment of the Health and Safety Commission and the Health and Safety Executive and makes further provision for securing the health and safety and welfare for persons at work, for protecting others against risks to health or safety in connection with the activities of persons at work, and other activities related to health and safety at work.

Provisions are contained for the mass of detailed and technical legislation which was previously adminis-

tered and enforced by a variety of statutory agencies to be replaced by a simpler coherent and co-ordinated body of regulations supported by practical guidance and coming under the scrutiny of one central policy making and enforcing body. That body is the Health and Safety Commission with its enforcing arm, the Health and Safety Executive. The Nuclear Installations Act 1965 (as amended) provides the necessary relevant statutory provisions of the main Act and thus contains the main safety legislation regulating the health and safety standards of nuclear installations in the United Kingdom.

The Health and Safety Executive may issue nuclear site licences under the powers granted by the Nuclear installations Act and attach such conditions to these licences as considered necessary. Within the Executive, the Nil has the responsibility for determining and maintaining the relevant safety requirements and for ensuring that the necessary precautions are effected by the licensee.

In addition to the flexible and effective control over the nuclear licensee given by the conditions attached to the site licence, the Health and Safety at Work Act contains additional means of enforcement by giving the Nil the power to serve improvement notices or prohibition notices.

The Ionising Radiations Regulations 1985 were is­sued under the requirements of the Health and Safety at Work Etc. Act and replace the radiological control requirements previously included as conditions in the nuclear site licence.

The CEGB believes that to assist in the achievement of a satisfactory tec el of safety for an operating re­actor, it is desirable to set down safety criteria at the beginning of a project against which the developing design can be assessed. The criteria should take the form of specific basic targets for designers, but should not be such as to inhibit them from making legiti­mate balances between system configuration and plant standards to produce optimum and economic design solutions.

The general safety requirements for the early mag — nox and AGR nuclear power stations have been subject to a continuous process of improvement and refine­ment in the light of technical and scientific develop­ments as successive designs of reactors have been through the design process. In 1974 the CEGB de­veloped the present detailed design safety requirements in order to consolidate experience and worldwide de­velopments in the safety field, and to encourage the use of probabilistic techniques in safety assessments. The fundamental criteria are now set down in the CEGB’s Health and Safety Department’s (HSD) docu­ment ‘Design Safety Criteria’ (HS/R167/81 Revised, March 1982) which lays down the safety requirements for all types of nuclear power stations. A more de­tailed guide to designers and the implementation of the criteria is given in the CEGB’s Generation Develop­ment and Construction Division’s (GDCD) document ‘Design Safety Guidelines* for each reactor type (DSG2 for the PWR).

The Design Safety Criteria and Guidelines provide guidance on the important safety-related factors which need to be taken into account during design. Several of these are based on the concept of acceptable risk and are expressed in probability terms as design tar­gets for each reactor on a site, while others describe targets in qualitative or engineering terms.

The criteria fall into three broad groups. The first, setting out the fundamental or basic criteria, specifies the targets and methods of assessment for doses to operating staff and the public under normal operating conditions. The second deals with the assessment of faults and hazards; that is, faults originating within the plant itself and hazards (e. g., earthquakes) arising from outside it. The third group details the engineer­ing criteria specifying system reliability criteria and the requirements for segregation of plant, separation of functions, inspection, testing and monitoring, emer­gency control and operator actions. It also contains requirements for items such as the control of radio­active discharges and specifies a requirement for a comprehensive quality assurance programme with re­sults fully documented and retained.

The criteria are not station operational limits but targets for designers, so that if a particular reactor design does not meet the criteria in all respects, it does not automatically follow that the design is un­acceptable to the CEGB from a safety point of view.

Nonetheless, while there is some latitude in the criteria quoted for operator exposures, releases and dose, it is generally the case that these cannot be significantly exceeded if the general safety objectives are to be met and an acceptable design produced.

Radiation surveys, performed by the Health Physics Department, are carried out both on and off the site. These are an important method of controlling radia­tion as they decide the zone classifications on the site and establish that no undue discharges are taking or have taken place. They are also a method of detect­ing unexpected increases in radiation intensity before radiation exposures have been incurred.

The introduction of maximum permissible dose limits implies a requirement to measure the dose received by the individual and there are two possible methods whereby this requirement can be fulfilled. If detailed surveys are undertaken, as just described, of the radiation fields present in all work areas then, with a knowledge of the time workers have spent in each location, an assessment can be made of their accumulated doses. Such a monitoring regime, al­though providing useful information for dose control is extremely difficult to implement in a nuclear power station environment. Therefore, each individual is issued with a dosemeter to be worn whenever access is required to radiation areas. The dosemeter is gen­erally just a sampling device and does not measure the doses to the individual body organs of interest. It does, however, give an assessment of the. radia­tion exposure to the body in the vicinity of the dose­meter. The doses to the various body parts have to be assessed or separately measured by additional dose — meters if these are exposed to a significantly different extent.

The principal control system jusl. rdescribed is based on the time spent in radiation areas of specific dose rates. However, an additional factor which is also important in the total dose equation is the distance parameter.

If one considers a point source of radiation, e. g., a piece of irradiated swarf or a commercial radioactive source, then the dose rate varies as the inverse square of the distance from the source. That is, if the dis­tance from a point source is doubled then the dose rate reduces by a factor of four; conversely, if the distance is halved, the dose rate increases fourfold. With linear sources, e. g., pipework, the dose rate varies directly with distance; for other shapes of source the exact relationship can be very complex but in all cases the principle is the same, i. e., to reduce the accu­mulated dose, the distance from the source must be increased.

It may not always be practicable to have large distances between personnel and the radiation source they may be working with so, as well as restricting the time, the effective distance is increased with the introduction of shielding. The type of shielding de­pending on the type and strength of the radiation, as previously discussed.

The nuclear site licence places a specific responsibility on the licensee to appoint duly authorised persons for covering identified areas of responsibility. The station manager has the delegated authority to make the appointments and will ensure that the appointee has the necessary qualifications and experience to carry out the function for which he has been ap­pointed. Details of all duly authorised persons’ train­ing, qualifications and experience are entered on a duly authorised persons’ (DAP) list which is sent to the Nil by HSD. The list is formally updated every three months and renewed biennially. The list of duly au­thorised persons is displayed on the nuclear site.

Duly authorised persons are those who control and supervise the plant and who authorise the maintenance of safety mechanisms. Under the CEGB Safety Rules (Radiological) authorisations are required for senior authorised persons (Nuclear Radiations), competent persons (Nuclear Radiations) and the accredited health physicists.

The senior authorised person (NR) is appointed, in writing, after a formal interview by a panel com­prised of senior station management and an inde­pendent manager from HSD. The SAP (NR) has full responsibility for radiological safety and the sole au­thority to prepare, issue, revalidate and cancel radio­logical control documents. SAP (NR)s are responsible (among other things) for controlling and co-ordinating the radiological safety of work on the plant and are appointed from the operations staff; being a shift manager, assistant charge engineer or, in some cases, assistant engineers (operations). Other grades of opera­tions staff are appointed as considered necessary to maintain adequate control of the safety of personnel and plant.

The accredited health physicist, is also appointed after formal interview by senior station management and an independent manager from HSD, and is a person suitably qualified by knowledge and experience to assess and advise on the health physics precautions necessary for radiological work on the nuclear licence site. After being appointed in writing, the person can sign and issue health physics certificates which provide detailed professional advice on radiological protection. Health physics certificates are issued at the request of a SAP (NR) and provide powerful advice on the necessary precautions to be taken prior to the work in radiation and contamination zones. It is, however, the ultimate responsibility of the SAP (NR) issuing the safety documentation to decide to implement such advice.

The competent person (Nuclear Radiations) is a person who can implement the requirements of a radiological control document and has been appointed in writing by the location manager to receive and clear such documents.

Graphite corrosion can be expressed either as G(-C) (the number of carbon atoms removed per 100 eV of
energy adsorbed) or as oxidation rate (the fractional weight of graphite oxidised per mWh/g of energy adsorbed). Graphite corrosion is measured by either a weight loss or a C-14 technique. Weight loss mea­surements involve weighing a sample, irradiating to a known dose in a known coolant composition and then reweighing the sample. During irradiation in-pore carbon deposits are formed, increasing with increasing G(-CH. t), and thus allowance must be made for these. The quantity of deposit is measured using a differ­ential thermal oxidation measurement in air. Using the graphite weight loss the G(-C) value is calculated using:

D x £ x P

where T is the absolute temperature, К

W is the graphite weight change, ppm D is the cumulative dose, mWh/g £ is the graphite open pore volume, cm3/g P is the gas pressure, bar

C-14 measurements are made by initially labelling a small graphite sample (typically 10-50 g) with C-14 by either manufacturing the graphite from C-14 la­belled coke and pitch or by heating a sample of C-12 graphite to 3000°C in the presence of C-14 labelled carbon monoxide when exchange between the gas and solid phases occur. The sample is then enclosed in a steel capsule and irradiated in a known radiation field with a gas of known composition passing through the capsule. The graphite oxidises and releases CuO into the gas stream, this being oxidised to CI4C>2 which

oitected and measured using either a gas or solid phase scintillation counter. G(-C) is calculated:

6.69 x 1010 x C Q T

here c is the effluent activity, Bq/cm3 q is the gas flow rate, cm3TP/s S is the graphite specific activity, Bq/g D is the dose rate, mW/g £ is the graphite open pore volume, cm3 p is the gas pressure, bar M is the molecular weight of CO2

The relationship between G(-C) and oxidation rate (OR) is given by: OR = G(-C) x £ x 8.1 x ІСГ10 at 41 bar and 673°C, where £ is the graphite open pore volume cm3/(100 cm3).

The mechanism of radiolytic graphite corrosion is a complex process proceeding via the formation of an ionised carbon dioxide molecule, to a carbon monoxide molecule and an oxidising species. In the absence of other materials these will recombine to carbon dioxide giving it its apparent radiation sta­bility. The lifetime of the oxidising species is typi­cally 10“7 s, equivalent to a range of a few microns at reactor pressures. Radiolytic corrosion therefore occurs throughout the graphite brick and not solely at the geometric surface. Early experiments demonstrated that the reaction rate was controlled by the rate of energy deposition within the coolant in the pores, i. e., the rate increases with increasing pore volume and increasing gas density (increasing pressure and decreasing temperature).

In magnox reactors the primary inhibitor of the reaction is carbon monoxide, which is permitted to build up to a maximum of 1.5v/o. The mechanism of inhibition is that as the carbon monoxide in­creases, the probability that the oxidising species will recombine before they react with the graphite surface also increases. In addition the small quantities of hydrogen and water (total of 25-100 vpm) form a slightly protective complex on the surface.

In AGR reactors, where methane concentrations up to 400 vpm are used together with up to 1.5 v/o carbon monoxide, the formation of a protective sur­face complex due to the radiolytic destruction of methane is the main mechanism of graphite inhibi­tion. In large diameter pores only a fraction of the oxidising species will reach the graphite surface be­fore being deactivated resulting in a net formation of protective species and a build up of carbon deposits. In the smaller diameter pores all the oxidising species feach the graphite surface leading to a net removal of carbon. Hence in a AGR a most important para­meter is the volume of small diameter pores (0.05 f*m -* 5 цт).

The difference of the pore distribution in which corrosion occurs between magnox and AGR coolant compositions has a very significant effect on the re­lationship between weight loss and cumulative dose. The basic equation relating the two parameters at constant reactor power is

(А2/100тге) log (1 + Ct A) — ACt 100 — g0t

where go is the initial oxidation rate given by go = 2.10 x 103 £G(-C) D ~ /<r0у

D is the dose rate, W/g

£ is the graphite open pore volume, cm-Vg

P is the gas pressure, bar

T is the temperature, °A

A = 100xe/(l — 7Te)

xe is the effective open pore volume, cm3/cm3 Ct is the % weight loss after time, t years.

For magnox coolant compositions the effective open pore volume to be used is the total initial open pore volume, but for AGR coolant compositions the vol­ume is only a fraction (20-30%) of the initial open pore volume. The equation indicates that, for any given initial oxidation rate, the cumulative weight loss increases with decreasing initial pore volume. This has been experimentally verified in a realistic AGR coolant composition and the results are shown in Fig 1.42 as a comparison of xe = open pore volume and irt ш 0.2 open pore volume. This clearly demon­strates the effect; but at the highest cumulative doses the weight loss becomes increasingly lower than pre­dicted by the above simple model, being due to the increasing diameter of the small diameter pores. More complicated corrosion models have been developed which allow for this effect as well as the influence of closed pore volume opening due to oxidation of the graphite.

Effect of coolant composition on graphite oxidation rates

In magnox reactors the two primary impurities in the coolant are carbon monoxide and hydrogen/water. For the former, increases up to 1.5 v/o decrease the corrosion rate by a factor of up to 3 compared to pure carbon dioxide, but further increases in carbon monoxide concentration do not significantly reduce the rate further. Increases in hydrogen/water concen­tration also reduce the oxidation rate up to 150 vpm but at higher concentrations little further decrease occurs. The operating levels of carbon monoxide and hydrogen/water which should be maximised to reduce

graphite corrosion have to be optimised with respect to carbon deposition on fuel pins and corrosion of circuit materials.

In AGR reactors the water concentration is in the range 200-500 vpm, due to its formation from methane destruction, and over this range has virtual­ly no effect on graphite oxidation rate. Therefore the main impurities that exert the major effect on graphite oxidation rate are the carbon monoxide and methane concentrations. A wide range of experiments using moderator graphite were carried out in the DIDO Materials Testing Reactor at AERE Harwell, the Siloe Reactor at CEA Grenoble and in the gamma irra­diation facility at Berkeley Nuclear Laboratory using both weight loss and C-14 techniques. Analysis of the results, ignoring thin specimens (< 2—4 mm) when geometric surface effects dominate, has shown good agreement between all the facilities, and the ‘mean’ oxidation rate against coolant composition is shown in Fig 1.43 and is given by the equation:

OR = 2.7 [0.26 — 1.53 log jo [(CO) + 8(CH4)]j

Г 1 1

————————- 0,5

0.55 + 1.29(CO)

I

1 + 2is(™±)

CO /

Where OR is the oxidation rate at 41 bar and 673 K, and (CO) and (CH4) are the carbon monoxide and methane concentrations, Vo.

Variations of ±30% about the mean data are found between different samples but no difference has been observed between the materials of two manufacturers. Anglo Great Lakes Corporation and British Acheson Electrode Ltd.

The effect of increasing methane concentration is always to decrease the oxidation rate although the effect tends to level out above 400-500 vpm. The effect of carbon monoxide concentration is more com­plex and demonstrates the relative effects of the two mechanisms of graphite inhibition. At low carbon monoxide concentrations the gas phase recombina­tion of active species is low and hence at low meth­ane concentrations the oxidation rate is high. As the methane concentration is increased the rate of meth­ane destruction increases and the increased surface inhibition leads to very low oxidation rates. At high carbon monoxide concentrations the gas phase recom­
bination of active species is high and hence at low methane concentration the oxidation rate is low. As the methane concentration is increased the rate of destruction of methane increases but, because this reaction is inversely proportional to carbon monoxide concentration, the increase in methane destruction and consequent decrease in oxidation rate is not as great as for low carbon monoxide concentrations. In the technological range of interest of methane concen­tration (200-400 vpm) the carbon monoxide concen­tration has little effect on oxidation rate and the exact choice has to involve other constraints.

A similar but more limited range of irradiation tests has been completed for AGR sleeve graphite and generally the mean oxidation rate has been shown to be twice the mean oxidation rate of AGR moderator graphite namely:

OR = 5.4 [0.26 — 1.53 logio [(CO) + 8(CH4>]]

[———- !———

[ 0.55 + 1.29(CO)

of predicted low methane concentration. A greater benefit can be achieved if the coolant is forced through the graphite by applying a pressure drop across the brick. In this situation the major graphite parameter controlling the methane depletion is the ‘permeability’, particularly at reactor pressure, the ‘viscous flow co­efficient’, Bo. Again this parameter should be maxi­mised within other constraints. However, ‘permeability’ is a difficult parameter to control and can vary by up to two orders of magnitude from one brick to another even within one graphite type. This is overcome in reactor design by testing every brick and preferentially laying the core so that bricks of high permeability are positioned in areas of high load or high cumulative dose. This has minimised the depletion of methane within the brick and maximised core integrity.

A series of computer programs generically known as ‘DIFFUSE’ have been written to solve the com­plex interactions. An example of the predicted corro­sion profile resulting from the programs is shown in Fig 1.44.

10.4.1 Graphite corrosion (thermal)

Section Ю.7 has described the operation of a PWR primary coolant system (reactor coolant system), and has emphasised the importance of an optimised chemical control regime.

Most of the technical issues discussed have been and will continue to be subjected to continuous re­search and re-assessment, particularly in response to changes in design and operation of PWR systems, improved monitoring techniques, the desirability of further reducing dose rates, and the emergence of new materials.

An attempt has been made to cover all the features that are included in current PWR primary circuit op­eration and to indicate either specific or general fea­tures that may be incorporated in future reactor de­signs following successful research and development.

10.8 Behaviour of magnox fuel in storage After discharge from a magnox reactor, fuel elements are stored to allow decay of radioactivity, particularly iodine, and of their heat content before they are transported to Sellafield for reprocessing. At all mag­nox stations (except Wylfa), the storage medium is water, the elements generally being stored horizontally in skips in ponds deep enough for adequate shield­ing purposes. At Wylfa, the fuel elements from the reactor are transferred to a dry store in which they are held in carbon dioxide and then, after a period to allow thermal decay, moved either to an air-filled dry store on site, or transported directly to Sellafield.

2.1 Design objectives

Following a trip or planned shutdown, a reactor still produces heat due to fission product decay. This is about 7% initially falling rapidly during the first hour to around 1.4%. It then falls more gradually to about 0.6% after 1 day and 0.3% after 1 week. Even after several weeks it is still significant. Normally this heat is readily removed.

However, if due to a fault in the normal cooling system the heat is not removed, the temperature of the reactor and coolant gas could slowly increase even­tually causing fuel failure and release of the active fission products to the gas coolant. The gas tem­perature increase would also cause the pressure to rise with the eventual opening of the safety relief valves. This would allow a path to atmosphere for the fission products, although there are filters on the gas side safety relief valves to remove particulate activity and limit any release.

There is therefore a safety requirement to cool the reactor fuel in the short and long term post-trip. The

approach adopted in design is to provide adequate cooling systems so that whatever the circumstances, normal or fault, the fuel and plant are maintained below prescribed safety limits.

In addition to safety limits, there will be economic constraints. If these were to be exceeded, the plant or fuel might sustain damage which would delay return to service or require some fuel to be discharged. The post-trip cooling systems must therefore also be de­signed for compliance with economic limits.

The same basic constraints must, of course, also be met when the reactor trips in the event of a fault. The reactor post-trip cooling systems protect against a wide range of faults including loss of primary or secondary circulation where the primary circuit inte­grity is maintained and for other faults where it may be breached.

The effect of such faults on the design of post-trip cooling systems is three fold. First, the nature of the fault may enhance the demand on the system; for example, where reactor pressure is lost. Second, the fault may arise from a plant fault which prevents part of the post-trip cooling system from acting. Third, plant provided for post-trip cooling may be damaged as a consequence of the fault. Each of these aspects must be taken into account in developing the design.

The overriding design objective is to remove post trip decay heat from the reactor in a reliable manner under all credible operational and fault conditions so that safety limits are not exceeded. The need to comply with economic constraints is less crucial but still im­portant in safeguarding the capital investment in the plant.

5.6.1 Condition monitoring of coolant

It is important to monitor the level of radioactivity in the COi cooiant to provide warning of the fol­lowing potential hazards:

• Release of long-lived fission products that could contaminate and limit access to reactor internals.

• Release of gaseous and volatile fission products from the vessel through leakage in normal opera­tion and in a depressurisation accident.

A recent development is the use of an online gamma spectrometer using a gas chamber near a Ge/Li-drifted semi-conductor gamma detector cooled with liquid nitrogen. This enables gaseous radio isotopes to be identified and their quantitative concentrations dis­played for each isotope. This technique enables very small leaks in the fuel cans to be detected and sup­plements the main failed fuel detection system.

The Hinkley Point В fuelling machine weighs 600 tonnes. It comprises a vessel some 26 m high with a design pressure of 65.5 bar. The hoist is at the top and the grab is suspended from two roller chains which pass over a double hoist sprocket, the chain slack passing into an external chain locker tube (Fig 2.107).

The multicore electrical cables from the grab run between and with the support chains. The hoist system embodies a soft suspension system of weights and springs so that the grab chain tension does not sud­denly change if a fuel assembly snags during movement.

Below the hoist head shaft is a turret comprising three storage tubes. When a fuel assembly has been lifted into a storage tube, latches are engaged under the closure to support it. The grab can then be dis­engaged and the turret rotated to bring another stor­age tube below the grab. The bottom of the pressure vessel is closed when required by a slide valve. Below the valve is a telescopic snout, which can be lowered to connect with either an access hole to one of the fuel handling facilities in the central block or the make-up shield pressure vessel.

The lower half of the pressure vessel is surrounded by external iron shot/concrete/steel shielding. The vessel and its shielding is carried on a crab running on a gantry bridge which traverses the length of the charge hall.

At Hinkley Point В, the reactor shield floor is 1.5 m below the charge hall floor level. A make-up shield is therefore coupled to the fuelling machine when it is at the reactor. The elastomer seals in the machine, the make-up shield telescopic snouts and the grab are cooled by a cold CO2 supply, the flow totalling 0.5 kg/s under normal circumstances.

The grab uses spring-assisted gravity to engage and lock the grab hooks onto the closure-pin cage lifting ring. The grab locking system is electrically energised to unlock and disengage (Fig 2.108). The grab has duplicated load cells so that the load on the grab is recorded and monitored by the machine interlock/ control system, which also monitors the outputs from the microswitches which register the positions of es­sential components in the grab.

At the reactor, the fuelling machine is aligned to the standpipe, prior to engaging the snouts, by op­erator control using TV cameras in the base of the make-up shield. TV cameras are also fitted in the

successfully completed, all relevant components are jn their correct starting positions and conditions, and other movements are locked out.

When refuelling at the reactor, the standpipe is pre-

ared by removing a number of slabs, disconnecting the closure cables, removing the gag motor/gearbox and placing an extension sleeve over the standpipe. The machine, containing a new fuel assembly and fitted with the make-up shield, is pressurised to re­actor pressure and connected to the reactor by lower­ing the make-up shield snout to engage with the sleeve. The machine valve is opened after the inter­space between the valve and the closure has been pressurised and all pressures have been equalised.

The fuelling machine grab then engages with the dosure/Iifting head of the spent fuel assembly and lifts it into the storage tube. After latching and grab disengagement, the turret is rotated and the new fuel assembly is lowered and locked into the standpipe. The machine valve is closed and, after interspace de — pressurisation, the machine is disconnected from the standpipe which is then reinstated. After disconnection of the make-up shield, the machine moves to the fuel handling facilities in the central block to dis­charge the spent fuel assembly into either a buffer storage tube, where it remains for a period of time to allow the fuel decay heat to reduce, or directly to the irradiated fuel disposal facility (IFDF).

The Dungeness В fuelling machine, which weighs some 1000 tonnes, is similar to that at Hinkley Point В in that it has a З-tube turret, but each storage tube has its own grab, chain system and hoist sprocket driven by an external hoist motor when a tube is in the operating position. The charge hall floor is all at one level so there is no make-up shield. The lowest part of the machine shielding has interlocked doors which are opened to give access for manual coupling of the telescopic adaptor (attached to the grab), to the closure and operation of its interlock key system. The Dungeness refuelling system permits the passage of hot reactor gas up into the machine which has a recirculating gas cooling system operating from start of hoisting until the fuel assembly is in the irradiated fuel disposal facility (IFDF). (At Hinkley Point B, the cold CO; injection prevents the reactor hot gas from rising into the bulk volume of the fuelling ma­chine pressure vessel, and when the assembly is stored within the machine its decay heat is dissipated by natural convection cooling.) The refuelling sequence at the Dungeness В reactor is similar to that at Hinkley Point B.

The Hartlepool/Heysham I fuelling machine is very different {Fig 2.109). It is a single-chamber ma­chine weighing some 300 tonnes. The pressure vessel is lowered on jacks through the shielding to seal the snout on a standpipe. There are two similar grabs, the fuelling grab and the plug grab. The fuel­ling grab (Fig 2.110) is suspended on hoist chains, the chain passing over hoist sprockets at the top of the vessel and the chain slack passing into a sepa­rate chain storage tube. When at the bottom of its travel the grab seals the bottom of the pressure vessel, The portion of the grab below the seal is telescopic and is extended and rotated by external drives engaging with the grab at this level to en­gage with the lifting features on the fuel assembly closure. The plug grab (Fig 2.111) has similar features and is stored in the lower portion of the machine pressure vessel. This grab carries a standpipe clo-.ure seal plug which is swung over and inserted into the standpipe.

When refuelling at the reactor, the empty machine is jacked dowm onto the standpipe and the hoist grab seals the bottom of the pressurised machine vessel. The interspace is pressurised and the fuelling grab is extended and engaged with the spent fuel assembly closure which is rotated by the grab to unlock it. The grab and fuel assembly are hoisted up into the machine. The plug grab is now swung over, lowered and operated to seal both the machine vessel and standpipe, the fuel channel being left empty. This grab is disengaged from the plug, the interspace is depressurised and the machine vessel is jacked off the standpipe.

The fuelling machine travels to a buffer storage tube, where the plug grab picks up the plug which seals the pressurised tube. The spent fuel assembly is lowered and locked into the buffer tube, the hoist grab sealing the machine. The fuelling machine now travels to another buffer storage tube, or to the new fuel facility, and picks up a new’ fuel assembly which is subsequently inserted into the empty fuel channel in the reactor.

With the exception of the mechanical annexe and turbine house, the buildings of the main power block are all designed to remain structurally intact, leak — tight {in the case of pressure-retaining parts) and functionally operable, to the extent required by their safety role, despite the occurrence of defined hazards including a safe shutdown earthquake. Because of its position close to the safety classified buildings, the mechanical annexe also has some capability for resisting an earthquake such that it will not fail and impair the functioning of the other buildings.

Other buildings designed to withstand seismic load­ings are the reserve ultimate heat sink auxiliary shut­down and diesel building, essential diesel buildings and the radioactive waste process and storage build­ing. The last-mentioned building is designed to retain liquid and solid wastes that may be spilled under the specified earthquake conditions.

The main building foundations are constructed of reinforced concrete. In the main power block, the reac­tor building, auxiliary and control building, fuel build­ing and the turbine house have separate foundations.

The buildings of the main power block are con­structed of reinforced concrete with some internal steel framing. The reactor building comprises the pre­stressed concrete primary containment vessel and its internal structures supporting the primary coolant sys­tem components. The containment base is of rein­forced concrete and contains a keyhole-shaped slot to accommodate the reactor pressure vessel and its instrument guide tubes. For leak tightness the con­crete containment structure has a 6 mm thick steel liner attached to its internal surfaces. Penetrations are provided for pipes and cables to enter the con­tainment and for man access. The two man access penetrations each incorporate two interlocked doors uuh an airlock between. There is an equipment hatch tor bringing large plant components into and out of the containment; this is provided for plant construc­tion, repair and replacement purposes. All these pene­trations are firmly anchored and sealed into the con­tainment wall. Except where it butts the auxiliary and fuel buildings, the containment is enclosed by a
steel-framed secondary containment enclosure build­ing, which provides for collection and filtration of leakages from the primary containment.

The containment is designed to contain the radio­active products and withstand the effects arising from postulated major reactor fault conditions such as a loss of coolant accident (LOCA). It also serves to protect the reactor primary coolant plant including the reactor vessel, steam generators, pressuriser and reactor coolant pumps. Massive reinforced concrete internal structures support the primary coolant plant within the containment. Structures embedded in these are designed to restrain the plant and prevent dam­age in the event of an earthquake and to limit dam­age due to postulated failure of major high pressure pipes. The concrete internal structures also function as radiation shields to shield operating and mainte­nance staff from excessive radiation exposure from the reactor coolant system components. The contain­ment internal height is 64 m, its internal diameter is 45.7 m, and its wall thickness is 1.3 m. It supports

within it a high level polar crane of 260 tonnes lifting capacity.